Radionuclide Transport Models Under Ambient Conditions Rev 01, ICN 00 MDL-NBS-HS-000008 November 2003 1. PURPOSE The purpose of this Model Report is to document the Unsaturated Zone Radionuclide Transport Model (UZ RTM), which evaluates, by means of three-dimensional (3-D) numerical models, the transport of radioactive solutes and colloids in the UZ under ambient conditions from the repository horizon to the water table at Yucca Mountain, Nevada. This Model Report also addresses comments made in the Model Validation Status Review (BSC 2001 [156257]), validates the Active Fracture Model with Matrix Diffusion, and justifies the values of sorption coefficients (Kds). This is in accordance with the Technical Work Plan (TWP) for: Performance Assessment Unsaturated Zone (BSC 2002 [160819], Section 1.11). This Model Report used flow fields provided by Model Report UZ Flow Models and Submodels (BSC 2003 [163045]). The output of this Model Report is a set of breakthrough curves for use as benchmarks to validate another model, FEHM, which is to be used for UZ transport calculations for Total System Performance Assessment for License Application (TSPA-LA). Validation of the FEHM based model is documented in Model Report Particle Tracking Model and Abstraction of Transport Processes (BSC 2003 [163933]). This validation is also specified in Section I-2-2-2 of the TWP (BSC 2002 [160819]) under the validation of the FEHM particle tracking model, where it states “Agreement with other codes will be demonstrated by running the same problem that has previously been run with DCPT, T2R3D, or TOUGH2.” Transport calculations presented in this report using DCPT and T2R3D are intended to be used as part of the validation of the FEHM particle tracking model. In the UZ RTM simulations presented in this Model Report, radionuclides are considered to be uniformly distributed at time zero in the fractures throughout the repository elements, and their subsequent transport to the water table is simulated. The UZ RTM does not consider the waste form, waste package, drip shield, or any other component of the Engineered Barrier System, and therefore the results are not to be construed as predictions of potential radionuclide transport after the repository closure, but only as examples of radionuclide transport calculations for this initial condition (i.e., corresponding to a time frame starting at the onset of radionuclide release). This Model Report documents the UZ RTM. This model considers 1. The transport of radionuclides through fractured tuffs 2. The effects of changes in the intensity and characteristics of fracturing between different hydrogeologic units 3. Colloid transport 4. Physical and retardation processes 5. The effects of different conceptual representations of the matrix-fracture system 6. The effects of different modeling approaches in the description of the transport problem. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 20 November 2003 In the present Model Report, we document the capabilities of the UZ RTM, which can describe flow (saturated and/or unsaturated) and transport, and accounts for (a) advection, (b) molecular diffusion, (c) hydrodynamic dispersion (with full 3-D tensorial representation), (d) kinetic or equilibrium physical and/or chemical sorption (linear, Langmuir, Freundlich, or combined), (e) first-order linear chemical reaction, (e) radioactive decay and tracking of daughters, (f) colloid filtration (equilibrium, kinetic, or combined), and (g) colloid-assisted solute transport. The present Model Report also includes determination and discussion of the distribution coefficients Kd of the various sorbing radioactive solutes in the UZ rocks, as well as a corresponding uncertainty analysis (see Attachment I, this Model Report). Transport of radioactive solutes and colloids (incorporating the processes described above) from the repository horizon to the water table are simulated to support model development and studies for Performance Assessment (PA). The results of the simulations are used to evaluate the transport of radioactive solutes and colloids, and to determine the processes, mechanisms, and geologic features that have a significant effect on transport. We evaluated the contribution of radioactive-decay daughter products to the total radionuclide inventory transport and from the bottom of the repository to the water table, as well as the effect on transport of different conceptual models for fracture-matrix systems. We also consider the effect of different modeling approaches on transport predictions. The primary uncertainty for using the modeling results documented in this report is that the input transport parameters were based on limited site data. For some input parameters, best estimates were used because no specific data were available. An additional uncertainty is that the RTM is based on the conceptual models and numerical approaches used for developing flow fields and infiltration maps, and thus they share the same limitations (see BSC 2003 [163045]). Compared to the TWP (BSC 2002 [160819]), there are some variations from the criteria used in the validation component of this study. For the validation of the Radionuclide Transport Model (RTM) using the numerical codes, their simulation results are compared to relevant analytical solutions. The validation criterion we used in this case is one or more of the following: (a) Mass fractions agree within 5% when CR = C/C0 = 10-4, or (b) The location of the tracer fronts agree within 5% when CR = C/C0 = 10-4, or (c) The tracer mass balances agree within 5%. The results presented in Section 7.2.1 show that this criterion was met in all instances. Note that the criterion presented above is somewhat different from that discussed in BSC (2002 [160819], Section I-2-1-1) and listed in Section 7.1.1 of this Model Report because the latter (a) is inadequate in that it captures only one of the three possible aspects for quantitative agreement, and (b) can be unrealistic near the tail end of the CR curve (where CR values are very low, and both analytical and numerical predictions may be unreliable because of roundoff error). For calibration and additional validation of the RTM using numerical codes (through comparison of their predictions to field measurements—Method 1), the criterion used is the following: Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 21 November 2003 (1) Overall agreement between predictions and field measurements are within 50% of each other, and, (2) If this agreement is not possible, document the analysis and discussion of the reasons for the observed deviations. Note that inability of the RTM model to match/predict field observations is not necessarily a sign of inability to validate the model if: (a) the quality of the field measurements are suspect (i.e., the test design may be responsible for uncertainties in the measurement), (b) there is insufficient information to fully describe the field test, or (c) field data involve very steep gradients (involving very significant parameter changes over a short distance or time) that lead to measurements prone to uncertainties and inaccuracies. The approach and criteria for the Method 1 validation in this Model Report is somewhat different from the one discussed in BSC (2002 [160819], Section I-2-1-1) and listed in Section 7.1.1 of this Model Report. This is because the TWP criterion is inadequate because it only addresses the option of time-series data (i.e., breakthrough) without accounting for the possibility of data involving concentration distribution in space at a particular time. The validation criterion used addresses this shortcoming, is more inclusive, and is also stricter. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 22 November 2003 INTENTIONALLY LEFT BLANK Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 23 November 2003 2. QUALITY ASSURANCE Development of this Model Report and the supporting modeling activities have been determined to be subject to the Yucca Mountain Project’s quality assurance (QA) program as indicated in Technical Work Plan for: Performance Assessment Unsaturated Zone (BSC 2002 [160819], Section 8.2, Work Package (WP) AUZM07). Approved QA procedures identified in the TWP (BSC 2002 [160819], Section 4) have been used to conduct and document the activities described in this Model Report. The TWP also identifies the methods used to control the electronic management of data (BSC 2002 [160819], Section 8.4, WP AUZM07) during the modeling and documentation activities. This Model Report discusses ambient radionuclide transport through hydrogeologic units (HGUs) which are a natural barrier and are classified in the Q-List (BSC 2003 [165179]) as “Safety Category” because it is important to waste isolation, as defined in AP-2.22Q, Classification Analyses and Maintenance of the Q-List. The results of this report are important to the demonstration of compliance with the postclosure performance objectives prescribed in 10 CFR 63.113 [156605]. The report contributes to the analysis data used to support performance assessment; the conclusions do not directly impact engineered features important to safety, as defined in AP-2.22Q. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 24 November 2003 INTENTIONALLY LEFT BLANK Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 25 November 2003 3. USE OF SOFTWARE The software codes and routines used in this study are listed in Table 3-1. All the software items used in this study, i.e., TOUGH2 V1.11MEOS9NTV1.0 (LBNL 1999 [113943]), TOUGH2 V1.4 Module EOS9 V1.4 (LBNL 2000 [146496]), TOUGH2 V1.6 Module EOS9 (LBNL 2003 [160242]; LBNL 2003 [161491]), T2R3D V1.4 (LBNL 1999 [146654]), DCPT V1.0 (LBNL 2000 [132448]), DCPT V2.0 (LBNL 2001 [154342]), iTOUGH2 V4.0 (LBNL 1999 [139918]), PHREEQC V2.3 (BSC 2001 [155323]), Bkread.f V1.0 (LBNL 2002 [162143]), XtractG.f90 V1.0 (LBNL 2003 [162786]), and Smesh.f V1.0 (LBNL 2002 [162142]), were appropriate for the intended application used only within the range of their software validation, and obtained from software configuration management per AP-SI.1Q, Software Management. Table 3-1. Qualified Software Used in this Report Software Name Version Software Tracking Number (STN) Document Input Reference System (DIRS) Number iTOUGH2 4.0 10003-4.0-00 139918 TOUGH2 1.11MEOS9NTV1.0 10065-1.11MEOS9NTV1.0-00 113943 TOUGH2 1.4 10007-1.4-01 146496 TOUGH2 1.6 10007-1.6-01 161491 TOUGH2 1.6 10007-1.6-00 160242 T2R3D 1.4 10006-1.4-00 146654 DCPT 1.0 10078-1.0-00 132448 DCPT 2.0 10078-2.0-00 154342 PHREEQC 2.3 10068-2.3-00 155323 Bkread.f 1.0 10894-1.0-00 162143 XtractG.f90 1.0 10930-1.0-00. 162786 Smesh.f 1.0 10896-1.0-00 162142 The software code TOUGH2 V1.11MEOS9NTV1.0 (Module EOS9nTV1.0) (LBNL 1999 [113943]) simulates flow and the decoupled transport of multiple radioactive solutes and/or colloids (parents and daughters) in complex subsurface systems involving porous and/or fractured media. The transport equations account for advection, molecular/colloidal diffusion, hydrodynamic dispersion, kinetic or equilibrium physical and chemical sorption (linear, Langmuir, Freundlich or combined), first-order linear chemical reaction, colloid filtration, and colloid-assisted solute transport. The code T2R3D V1.4 (LBNL 1999 [146654]) simulates flow and the coupled transport of a single radioactive solute tracer in complex subsurface systems involving porous and/or fractured media. The transport equations account for advection, molecular diffusion, hydrodynamic dispersion, and linear equilibrium sorption. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 26 November 2003 The software codes TOUGH2 V1.4 (Module EOS9 V1.4) (LBNL 2000 [146496]) and TOUGH2 V1.6 (Module EOS9 V1.6) (LBNL 2002 [160242]; LBNL 2003 [161491]) solve the Richards equation to simulate flow in complex subsurface systems (involving porous and/or fractured media). The software codes DCPT V1.0 (LBNL 2000 [132448]) and DCPT V2.0 (LBNL 2001 [154342]) involve the particle-tracking method to simulate transport of a single radioactive solute tracer in complex subsurface systems involving porous and/or fractured media. The transport equations account for advection, molecular diffusion, hydrodynamic dispersion, and linear equilibrium sorption. The software iTOUGH2 V4.0 (LBNL 1999 [139918]) and Bkread.f V1.0 (LBNL 2002 [162143]) are used for model calibration and prediction of Alcove 8/Niche 3 tests documented in Section 7.4. A numerical model, generated with a software routine Smesh.f V1.0 (LBNL 2002 [162142]), was developed for the test site to compare the simulation results with the relevant field observations. For elements that sorb primarily through surface complexation reactions, the experimental data are augmented with the results of modeling calculations using PHREEQC V2.3 (BSC 2001 [155323]). The inputs for the modeling calculations include groundwater compositions, surface areas, binding constants for the elements of interest, and thermodynamic data for solution species. These modeling calculations provide a basis for interpolation and extrapolation of the experimentally derived sorption distribution coefficient dataset (see Attachment I). The software XtractG.f90 (LBNL 2003 [162786]) was used to prepare the mesh files for the EOS9nT simulations, for the preparation of the corresponding initial conditions files, and the extraction of desired data subsets for plotting. Standard visual display graphics programs (i.e., IGOR Pro. 4, Tecplot V8.0, Adobe Illustrator V8.0) were also used to illustrate information but are not subject to software quality assurance requirements under Quality Assurance Requirements and Description (QARD) per Section 2.0 of AP-SI.1Q. This Model Report documents the UZ RTM. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 27 November 2003 4. INPUTS The input used in this Model Report consist of the following: • Transport properties • Calibrated fracture and matrix properties • Base case flow fields • Geochemical data • Numerical grids. 4.1 DIRECT INPUT Specific input data sets and their associated Data Tracking Numbers (DTNs) and sources are listed in Table 4.1-1. The Q-status of these data is provided in the DIRS. Discussions of uncertainties in the input data and parameters, and detailed explanations about the selection of corresponding choice of values, are addressed in Section 6. The molecular diffusion coefficients and half lives of radionuclides were obtained from accepted sources and have been qualified in the TDMS. Sorption coefficients were measured on rock samples from YM and using approved procedures, rock properties, including matrix permeability and porosity, fracture permeability and spacing were calibrated with field tests on YM. The grid for numerical simulation was based upon the Geological Framework Model of YM. Therefore, these data are all appropriate for simulations of radionuclide transports. Table 4.1-1. Input Data Description Parameters Data Tracking Number / Technical Information Source Molecular diffusion coefficient Lide 1992 [166224], pp. 5-111–5-112, LA000000000034.002 [148603] Half-life for the first-order radioactive decay T1/2 Lide 1992 [166224], pp. 11-13–11-133 Sorption distribution coefficient Kd LA0305AM831341.001 [163789] LA0306AM831343.001 [164949] LA0309AM831341.002 [165523] LA0309AM831341.003 [165524] LA0309AM831341.004 [165525] LA0309AM831341.005 [165526] LA0309AM831341.006 [165527] LA0309AM831341.007 [165528] LA0310AM831341.001 [165865] Properties of the main radionuclides in the transport simulations (Tc, Np, Pu, U, Pa, U, Th, Am, Cs, Ra, Sr) (Section 6.5 to Section 6.18) Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 28 November 2003 Table 4.1-1. Input Data (continued) Description Parameters Data Tracking Number / Technical Information Source Data for 3-D site-scale radionuclide (solute) transport modeling (Section 6.5 to Section 6.18) 3-D site-scale grid, hydraulic properties/parameters of the various hydrogeologic units, pressures, water saturations, 3- D flow fields for three scenarios (mean, high and low) for each of three climatic conditions (present-day, monsoon and glacial) LB03013DSSCP3I.001 [162379], LB03023DSSCP9I.001 [163044] Rock physical properties of the various hydrogeologic units LB0210THRMLPRP.001 [160799] EOS9nT 3-D site-scale simulations of colloid transport (perched-water model #1, mean present-day infiltration) (Section 6.7) 3-D site-scale grid, hydraulic properties/parameters of the various hydrogeologic units, pressures, water saturations, 3- D flow fields LB03013DSSCP3I.001 [162379], LB03023DSSCP9I.001 [163044] Pore size data for pore size exclusion estimation GS950608312231.008 [144662] GS980908312242.039 [145272] EOS9nT 3-D site-scale simulations (radioactive solutes and colloids, Section 6.15 to Section 6.18) 3-D grid used in the simulations LB03023DSSCP9I.001 [163044] 4.1.1 Data Used Corroboratively Specific data sets used corroboratively are listed in Table 4.1-2. These data were either used to simulate UZ tracer test at Busted Butte, or are field data from that test. Additional data previously cited in Table 4.1-1 were also used for Busted Butte simulations. Discussions of uncertainties in the input data and parameters, and detailed explanations about the selection of corresponding choice of values, are addressed in Section 7. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 29 November 2003 Table 4.1-2. Data Used for Busted Butte Simulations Description Parameters Data Tracking Number / Technical Information Source Modeling the Phase1A test of Busted Butte (Section 7.3.3) Matrix porosity f, permeability k, initial saturation, and rock grain density .s of the field samples GS990308312242.007 [107185], GS990708312242.008 [109822] Diffusion coefficient and sorption parameters Cussler (1984 [146653]), Benson and Bowman (1994 [122788]) Model validation—Phase1A test of Busted Butte (Section 7.3.3) Field measurements/data LA0302WS831372.001 [162765], LA9910WS831372.008 [147156] Modeling the Phase1B test of Busted Butte (Section 7.3.4) Matrix porosity, permeability, and initial saturation of the field samples GS990308312242.007 [107185], GS990708312242.008 [109822] Diffusion coefficient for Li+ and 2,6-DFBA Cussler (1984 [146653]), Benson and Bowman (1994 [122788]) Sorption parameters LA9912WS831372.001 [156586] Field measurements/data LA0201WS831372.008 [162766] Model Validation–Phase 2C test of Busted Butte (Section 7.3.5) Field measurements/data LA0008WS831372.001 [156582] LA0112WS831372.001 [157100] LA0112WS831372.002 [157115] LA0112WS831372.003 [157106] LA0211WS831372.001 [162763] NOTE: The data in this table are being used as reference only to support model validation. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 30 November 2003 4.2 CRITERIA The licensing criteria for postclosure performance assessment are stated in 10 CFR 63.114 [156605]. The requirements to be satisfied by TSPA are identified in the Yucca Mountain Project Requirements Document (Canori and Leitner 2003 [161770]). Acceptance criteria that will be used by the Nuclear Regulatory Commission (NRC) to determine whether the technical requirements have been met are identified in Yucca Mountain Review Plan, Final Report (YMRP; NRC 2003 [163274]). The pertinent requirements and criteria for this Model Report are summarized in Table 4.2-1. Table 4.2-1. Project Requirements and YMRP Acceptance Criteria Applicable to This Model Report Requirement Numbera Requirement Titlea 10 CFR 63 Link YMRP Acceptance Criteria PRD-002/T-016 PRD-002/T-015 Requirements for Performance Assessment Requirements for Multiple Barriers 10 CFR 63.114(a-c) [156605] 10 CFR 63.115(a-c) [156605] Criteria 1 to 4 for Radionuclide Transport in the Unsaturated Zone b Criteria 1 to 3 for Demonstration of Multiple Barriers c NOTES: a from Canori and Leitner (2003 [161770]) b from NRC (2003 [163274], Section 2.2.1.3.7.3) c from NRC (2003 [163274], Section 2.2.1.1.3) The acceptance criteria identified in Section 2.2.1.3.7.3 of the YMRP (NRC 2003 [163274]) are given below, followed by a short description of their applicability to this Model Report: • Acceptance Criterion 1, System Description and Model Integration Are Adequate: Transport relationships at the interface between the waste emplacement drift and the rock are accounted for in a manner that is consistent with the various flow and transport model abstractions used in the TSPA. The methodology accounts for important design features and physical phenomena at a level consistent with the available data and uncertainty such that the radionuclide transport in the unsaturated zone is not underestimated. In particular, this model provides suitable estimates for the boundary conditions for the UZ mountain-scale transport abstraction in terms of the partitioning of radionuclide releases from waste emplacement drifts to the fractures and rock matrix supporting the inclusion of related FEPs. Information supporting Acceptance Criterion 1 is presented in the following Sections: 6.1 to 6.2, 6.5 to 6.23, and 6.25. • Acceptance Criterion 2, Data Are Sufficient for Model Justification: Geological and hydrological characteristics used in the safety case are adequately justified. Adequate descriptions of how the data were used, interpreted, and appropriately synthesized into the parameters are provided. Data on the geology and hydrology of the UZ, including fracture distributions, fracture properties and stratigraphy, used in the TSPA, are based on appropriate techniques. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 31 November 2003 Data that are input to the model are identified in Section 4.1. Comparison of data not used as input to simulation for model validation is presented in Section 7. Information supporting Acceptance Criterion 2 is presented in the following Sections: 6.2, 6.5 to 6.11, 6.23, and 6.25. • Acceptance Criterion 3, Data Uncertainty Is Characterized and Propagated Through the Model Abstraction The parameters used in and derived by the radionuclide transport process model are technically defensible; they are based on and consistent with available data from Yucca Mountain; uncertainties and variabilities are evaluated and reasonably accounted for and adequately represented. Uncertainty is adequately represented in parameter development for conceptual models, process-level models, and alternative conceptual models, considered in developing the abstraction of radionuclide transport in the unsaturated zone. Information supporting Acceptance Criterion 3 is presented in the following Sections: 6.5, 6.6, 6.12 to 6.14, 6.23, and 6.25. • Acceptance Criterion 4, Model Uncertainty is Characterized and Propagated Through the Model Abstraction: Alternative modeling approaches of features, events, and processes are considered and are consistent with available data and current scientific understanding, and the results and limitations are appropriately considered in the abstraction. Conceptual model uncertainties are adequately defined and documented, and effects on conclusions regarding performance are properly assessed. Appropriate alternative modeling approaches are consistent with available data and current scientific knowledge, and appropriately consider their results and limitations. Information supporting Acceptance Criterion 4 is presented in the following Sections: 6.5, 6.6, 6.12 to 6.23, and 6.25. The acceptance criteria identified in Sections 2.2.1.1.3 of the YMRP (NRC 2003 [163274]) are given below, followed by a short description of their applicability to this Model Report: • Acceptance Criterion 1, Identification of Barriers Is Adequate: The unsaturated rock layers below the repository (and above the water table) form a natural barrier important to waste isolation. This barrier functions by delaying radionuclide movement. The barrier capability is determined by the hydrological and transport properties as implemented in this model for UZ transport. Information supporting Acceptance Criterion 1 is presented in the following Sections: 6.6 to 6.10, 6.12 to 6.23, and 6.25. • Acceptance Criterion 2, Description of Barrier Capability to Isolate Waste Is Acceptable: Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 32 November 2003 The capability of the identified barrier to substantially delay the movement of radionuclides is adequately identified and described. The uncertainty associated with barrier capabilities is adequately described. Information supporting Acceptance Criterion 2 is presented in the following Sections: 6.6 to 6.10, 6.12 to 6.23, and 6.25. • Acceptance Criterion 3, Technical Basis for Barrier Capability Is Adequately Presented: The technical bases are consistent with the technical basis for the performance assessment. The technical basis for assertions of barrier capability is commensurate with the importance of each barrier’s capability and the associated uncertainties. Information supporting Acceptance Criterion 3 is presented in Section 6.25. Two of these criteria (criteria 3 and 4 for Radionuclide Transport in the UZ, Section 2.2.1.3.7.3 of NRC 2003 [163274]) were not identified in the TWP (BSC 2002 [160819], Table 3.1), but are appropriate for this Model Report. One criterion that was identified for this Model Report (criterion 2 for Flow Paths in the UZ, Section 2.2.1.3.6.3 of NRC 2003 [163274], Flow Paths in the Unsaturated Zone: Data are Sufficient for Model Justification) is not appropriate for this Model Report. 4.3 CODES AND STANDARDS No specific formally established codes or standards have been identified as applying to this Model Report. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 33 November 2003 5. ASSUMPTIONS In this section we discuss the basic data assumptions involved in the transport model, and we provide a short discussion of the supporting rationale. The main data assumptions in this study are listed in Table 5-1. A physical basis for an assumption for the tortuosity factor is provided by Farrell and Reinhard (1994 [122803], p. 64) and Grathwohl (1998 [141512], pp. 28–35), while the work of Hu and Brusseau (1995 [122846]) experimentally validated this approach. A more thorough analysis on this subject can be found in Section 6.1.2.4. No further confirmation is required for the assumptions listed in Table 5-1. The transfer coefficient Ki indicates the ratio of the concentration of dissolved species in immobile and mobile water. By setting Ki = 1, it is assumed that immobile water is equilibrated instantaneously with mobile water. This assumption was also made by de Marsily (1986 [100439], pp. 238–239) and is consistent with the assumption of linear adsorptive equilibrium. This topic is discussed further in Section 6.2.3.1; no further confirmation is required. James and Chrysikopoulos (1999 [109517], p. 707) found that larger colloidal particles travel in fractures faster than smaller particles. With a parabolic velocity profile (James and Chrysikopoulos (1999 [109517], Figure 1) between parallel plates, the maximum liquid velocity is 1.5 times the average velocity. Therefore the largest particles (450 nm) were assigned that velocity adjustment factor as shown in Table 5-1 and Table 6.18-1. No further confirmation is required for this assumption. Dispersivity values ranging from 0.06 m to 2.63 m were measured in gas-tracer experiments conducted in Yucca Mountain (LeCain et al. 2000 [144612], Tables 18 and 19). Based on those measurements, the values shown in Table 5-1 were assumed. Further discussion of longitudinal dispersivity is presented in section 6.1.2.2. For both solutes and colloids, advection is much more important than dispersion; the assumption is therefore adequate for the purpose of predicting radionuclide transport, and no further confirmation is required. The kinetics of declogging (release of colloidal particles after being immobilized) is not well understood. Therefore, a wide range of values have been assumed (Section 6.18.3). Simulation results shown in Sections 6.18.4 through 6.18.7 show that transport is not sensitive to the declogging kinetic coefficient; the assumption is therefore adequate for the purpose of predicting radionuclide transport, and no further confirmation is required. Because a fracture is by definition devoid of matrix material, assigning a porosity of 1 to fractures is reasonable and requires no confirmation. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 34 November 2003 Table 5-1. Main data assumptions used in this study Parameter Sections Value Justification Tortuosity t 6.6, 6.8 to 6.20 f See Table 6.8-1 Transfer coefficient Ki for solutes 6.8 to 6.20 1 Conventionally used value Velocity adjustment factor fv 6.18 1.5, 1.2, 1.1, 1.0 for the 450, 200, 100, 6 nm Reasonable estimates (covers entire range of fv values, see Section 6.2.4.1) Longitudinal dispersivity aL (solutes) 6.7 to 6.17, 6.20 1 m in fracture, 0.1 m in matrix Scientific judgment (based on LeCain et al. (2000 [144612]). Longitudinal dispersivity aL (colloids) 6.18 0.5 m in fracture, 0.05 m in matrix Scientific judgment (based on LeCain et al. (2000 [144612]). Backward (reverse) kinetic filtration (declogging) coefficient .- 6.18 100 .+, 0.1 .+, 0 Reasonable estimates bracketing the range of .- (ranges are sufficiently wide to cover likely values) Fracture porosity f 6.6 to 6.17, 6.20 1 Reasonable estimate (reasonable for open fractures) The immediately preceding upstream document to this Model Report is entitled UZ Flow Models and Submodels (BSC 2003 [163045]). No assumptions involving numerical values of input parameters were used in that report. However, several approximations and idealization were used for the model development of the UZ Flow Models and Submodels Model Report. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 35 November 2003 6. MODEL DISCUSSION The objective of this Model Report is to provide a defensible and credible model of radionuclide transport in individual vertical quasi 2-D slices of the UZ and in large 3-D systems using the flow fields submitted for use in TSPA calculations. This is in accordance with the Technical Work Plan (TWP) for: Performance Assessment Unsaturated Zone (BSC 2002 [160819], Section 1.11). The output/product of this Model Report is a set of breakthrough curves for use as benchmarks for the particle-tracking calculations performed for TSPA (BSC 2003 [163933]). In this section, we describe the development of the RTM. The primary objectives of the RTM are: • Using the comprehensive calibrated 3-D model of the unsaturated flow developed in BSC (2003 [163045], Section 6.2), to integrate the available data for the development of a comprehensive model of radionuclide transport through the UZ of Yucca Mountain under a range of current and future climate conditions. • To identify the controlling transport processes and phenomena, and to evaluate the effectiveness of matrix diffusion and sorption as retardation processes. • To identify the geologic features that are important to radionuclide transport. • To estimate the migration of important radionuclide solutes and their daughter products from the repository toward the water table. • To evaluate the effects of various climatic conditions on radionuclide transport. • To estimate the migration of radioactive colloids from the repository toward the water table, and to determine the sensitivity of colloid transport to the kinetic coefficients of colloid filtration. • To evaluate the effect of fracture spacing, intensity and configuration on radionuclide transport and retardation through important hydrogeologic units. The RTM is a dual permeability model that uses steady state flow fields, and simulates transports of solutes and colloids advectively and by diffusion between fracture and matrix. The solutes also are subject to sorption and filtrations for the case of colloids. This section consists of the following subsections: Section 6.1. Geological Model and Physical Processes. In this section we focus on (a) the geology and stratigraphy in the UZ of Yucca Mountain, the site for the repository, (b) the processes and phenomena involved in and affecting transport, and (c) important issues that may have an impact on predictions about and understanding of transport behavior of radioactive solutes and colloids in the UZ. We also list the basic modeling approaches. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 36 November 2003 Section 6.2. Mathematical Model of Transport. In this section we present the mathematical basis and discuss the implications of the various processes and phenomena involved in the transport of solutes and colloids in the UZ. Section 6.3. The Numerical Models. In this section, we discuss the codes that implement the numerical models developed from the principles discussed in Section 6.2. These codes are T2R3D V1.4 (hereafter referred to as T2R3D; LBNL 1999 [146654]) and TOUGH2 V1.11MEOS9NTV1.0 (hereafter referred to as EOS9nT; LBNL 1999 [113943]), and were used for the simulations discussed in Sections 6.6 to 6.20, and in Section 7. Section 6.4. Features, Events and Processes (FEPs). This section focuses on the discussion of selected Features, Events and Processes (FEPs) taken from the Licensing Application FEP List, and which are associated with the subject matter of this Model Report. Section 6.5. Sorption Model and Parameters. This section focuses on the sorption behavior of the radionuclides of interest in the various UZ rocks and discusses the determination of the sorption coefficients used in this Model Report. Section 6.6. A Preliminary Conceptual Model of Transport Based on Geology and Stratigraphy. In this section, we develop a conceptual model of radionuclide transport in vertical columns covering the geologic spectrum from the repository to the water table based on the geology and stratigraphy at three representative UZ locations. Section 6.7. Three-Dimensional Transport Simulations. In this section, we discuss the grids, climatic conditions, perched water model, radionuclides, flow fields, and general approach used in the ensuing 3-D site-scale simulations of transport through the UZ (Sections 6.8 to 6.18, and 6.20), as well as the conditions and general options used in these EOS9nT simulations. Section 6.8. Three-Dimensional Simulations of 99Tc Transport—Instantaneous Release. In this section, we study the migration of 99Tc through the UZ following a single release event (instantaneous release). We investigate the effects of various climatic scenarios on transport, as well as the impact of uncertainty in the contribution of diffusion (as quantified by the choice of the molecular diffusion coefficient) on transport. We also identify transport-controlling geologic features, transport patterns, and important retardation mechanisms. Section 6.9. Three-Dimensional Transport of 237Np—Instantaneous Release. In this section, we investigate the transport of 237Np through the UZ under various climatic scenarios following a single radionuclide release event. We also study the sensitivity of transport to uncertainty in matrix diffusion. We identify transport-controlling geologic features and important retardation mechanisms, and we compare the concentration distributions and the transport patterns of 237Np to those of 99Tc. Section 6.10. Three-Dimensional Transport of 239Pu—Instantaneous Release. In this section, we study the transport of 239Pu through the UZ under various climatic scenarios following a single radionuclide release event, and for various levels of matrix diffusion (as controlled by the choice of the diffusion coefficient). We identify transport-controlling geologic features and Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 37 November 2003 important retardation mechanisms, and we compare the concentration distributions and the transport patterns of 239Pu to those of 99Tc and 237Np. Section 6.11. Three-Dimensional Transport of 233U and 235U—Instantaneous Release. In this section, we study the transport of 233U and 235U through the UZ under various climatic scenarios following a single radionuclide release event. Section 6.12. Three-Dimensional Transport of 241Am and 90Sr —Instantaneous Release. This section focuses on the transport of 241Am and 90Sr through the UZ under various climatic scenarios following a single radionuclide release event. Section 6.13. Three-Dimensional Transport of 135Cs—Instantaneous Release. This section addresses the issue of transport of 135Cs through the UZ under various climatic scenarios following a single radionuclide release event. Section 6.14. Three-Dimensional Transport of 226Ra, 229Th, and 231Pa—Instantaneous Release. In this section, we study the transport of 226Ra, 229Th, and 221Pa, all of which sorb very strongly onto the Yucca Mountain rocks. We also study the effect of various climatic scenarios on transport following a single radionuclide release event. Section 6.15. Three-Dimensional Simulations of 99Tc, 237Np, and 239Pu Transport— Continuous Release. In this section, we investigate the migration of 99Tc, 237Np, and 239Pu through the UZ when the radioactive species are released continuously from a decaying source. We identify transport-controlling geologic features, transport patterns, and important retardation mechanisms, and discuss the differences from the instantaneous release scenario. Section 6.16. Three-Dimensional Transport of the 239Pu 235U 231Pa Chain— Continuous Release. In this section, we study the transport of the three-member 239Pu chain through the UZ when the radioactive species are continuously released from a decaying source. The importance of accounting for the transport of the whole chain rather than individual species is discussed. Section 6.17. Three-Dimensional Transport of the 241Am 237Np 233U 229Th Chain—Continuous Release. In this section, we study the transport of the four-member 241Am chain through the UZ when the radioactive species are continuously released from a decaying source. The importance of accounting for the transport of the whole chain rather than individual species is discussed. Section 6.18. Three-Dimensional Site-Scale Transport of PuO2 Colloids—Continuous Release. In this section, we study the transport of four radioactive colloids through the UZ for a mean present-day infiltration. We identify transport-controlling geological features and important retardation mechanisms, and discuss the differences between the transport patterns of the four colloids. Section 6.19. Alternative Models. In this section, we investigate Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 38 November 2003 1. Different representations of the matrix-fracture system (multiple interactive continua vs. dual-permeability systems), and 2. Different conceptual methods of describing the transport problem (particle tracking vs. conventional representation). We investigate the effect of these alternative approaches on predictions of transport through a 2-D vertical cross section of the UZ at Yucca Mountain. Section 6.20. Barrier Evaluation, Uncertainties, and a Note of Caution. In this section, we review the results of the analysis and discuss the overall barrier performance of the geological setting underneath the repository. Additionally, we review the issue of uncertainty in the predictions, and possible implications stemming from this analysis. Finally, we discuss the framework within which the results of this Model Report should be interpreted. The scientific notebooks (SN) used for the activities in this Model Report are listed in Table 6.1. Table 6.1. Scientific Notebooks Used for Model Report LBNL Scientific Notebook YMP M&O Scientific Notebook ID Page Numbers Citation YMP-LBNL-GJM-5 SN-LBNL-SCI-099-V3 1-109 Wang 2003 [164021] YMP-LBNL-GJM-YS-1 SN-LBNL-SCI-230-V1 8–90 Wang 2003 [164021] YMP-LBNL-GJM-GZ-1 SN-LBNL-SCI-235-V1 9–37 Wang 2003 [164021] YMP-LBNL-GSB-LHH-3 SN-LBNL-SCI-215-V1 107–109, 111–112, 114–115 Wang 2003 [164021] YMP-LBNL-GSB-QH-1 SN-LBNL-SCI-168-V1 1–38 Hu 2001 [163148] Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 39 November 2003 6.1 GEOLOGICAL MODEL AND PHYSICAL PROCESSES 6.1.1 Geological Layering The geological model used in this Model Report is based on the Geological Framework Model (GFM2000) and Integrated Site Model (ISM). The subsurface formations at Yucca Mountain consist of heterogeneous layers of anisotropic, fractured volcanic rocks. There are alternating layers of welded and nonwelded ash flow and air fall tuffs. The cooling history of these volcanic rock units determines their mechanical and hydrological properties. Beginning from the land surface, the Yucca Mountain geologic units are the Tiva Canyon, Yucca Mountain, Pah Canyon, the Topopah Spring Tuffs, and interbedded tuffs of the of the Paintbrush Group. Underlying these are the Calico Hills Formation, and the Prow Pass, Bullfrog, and Tram Tuffs of the Crater Flat Group. These formations have been divided into major hydrogeologic units based roughly on the degree of welding. These are the Tiva Canyon welded units (TCw); the Paintbrush nonwelded units (PTn), consisting primarily of the Yucca Mountain and Pah Canyon members and the interbedded tuffs; the Topopah Spring welded units (TSw); the Calico Hills nonwelded units (CHn); and the Crater Flat undifferentiated (CFu) hydrogeologic units (Scott and Bonk 1984 [104181]; BSC 2003 [160109], Section 6). Conceptual models of flow and transport at Yucca Mountain are described in BSC (2003 [163045], Section 6). In the present Model Report, we focus on the subject of radionuclide transport in the hydrogeologic units beneath the repository horizon. The repository will be sited in the TSw unit (CRWMS M&O 1999 [103773]), and more specifically the tsw34, tsw35, and tsw36 layers of the unsaturated zone (UZ), depending on the location. More information on these layers can be found in BSC (2003 [160109], Section 6) and BSC (2003 [161773], Section 6). Unsaturated flow in the TSw is primarily through the fractures, because the matrix permeability in many of the TSw layers can support flows of only a few millimeters per year, and the average fracture spacing in the TSw layers is on the order of 0.5 m (BSC 2003 [161773], Section 6). The CHn unit and the Prow Pass (PP) unit (formally a part of CHn, but studied separately in the present Model Report) below the repository horizon are complex geological systems with strongly heterogeneous distributions of fracture and matrix hydrological properties, which are expected to have pronounced effects on flow and transport of radionuclides in the UZ. There is limited information on the CHn unit, and even less on PP. The permeability of nonwelded tuffs is strongly dependent on the degree of alteration of the rock minerals into zeolites. Zeolitic alteration in the CHn (a common occurrence in its lower layers) can reduce the matrix permeability by orders of magnitude in relation to that of the welded tuffs (BSC 2003 [161773], Section 6). In nonwelded vitric tuffs, the matrix and fracture permeabilities are on the same order of magnitude (BSC 2003 [161773], Section 6). Thus, the non-zeolite layers behave as porous (rather than fractured) media, and flow is matrix dominated. This has important implications for transport, because the longer contact times in these nonwelded tuff units allow increased radionuclide sorption. The CHn major hydrogeologic unit is composed of Calico Hills vitric (CHv) and Calico Hills zeolitic (CHz) units (BSC 2003 [160109], Section 6). Typically, radionuclides are more strongly adsorbed onto zeolitic units than onto vitric units (see Attachment I and Table 6.5-1). For example, the Kd (see Equation 6-9) of 237Np in the vitrified and the zeolitic tuffs is 0.5 mL/g and Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 40 November 2003 1 mL/g, respectively (Output-Data Tracking Number (DTN): LA0302AM831341.002, see Section 6.5). Consequently, migration of 237Np is expected to be more retarded in the zeolitic than in the vitrified units, if the contact time is the same. Flow in the CHz units is expected to be concentrated in the fractures because of the large permeability contrast between matrix and fractures (BSC 2003 [161773], Section 6); the permeability in the fractures is about four orders of magnitude larger than in the matrix (DTNs: LB0205REVUZPRP.001 [159525] and LB0207REVUZPRP.002 [159672]). Fracture-dominated flow is associated with short contact times, limited radionuclide removal through diffusion and sorption, and thus transport over longer distances. The fracture and matrix permeability values of the zeolitized Calico Hills unit are based on product outputs from the report Calibrated Properties Model (BSC 2003 [160240] and the report Analysis of Hydrological Properties Data (BSC 2003 [161773]). These hydrological properties, derived by inverse modeling to match pneumatic responses through fracture network and watercontent and potential values in the matrix, are regarded as reasonable representation of the actual fracture and matrix permeability values of the zeolitized Calico Hills unit. 6.1.2 Transport Radioactive contaminants can escape from the wastes stored in the repository. These contaminants can migrate through the UZ as a dissolved molecular species or in colloidal form. Transport of these radioactive solutes or colloids involves advection, hydrodynamic dispersion, sorption (solutes) or filtration (colloids), matrix diffusion, and radioactive decay. The transport of radionuclides is also affected by factors such as solubility limits, the presence of perched water, and heating effects from the repository. In this section, we briefly discuss the phenomena, processes, and factors affecting transport. 6.1.2.1 Advection Advection is the transport of dissolved or colloidal species by flowing water. In the Yucca Mountain UZ, flow is predominately downward (in response to gravitational differentials), and so is advective transport (DOE 1998 [100550], p. 3-112). Some lateral advection is also expected in response to lateral flow diversion at the boundaries of hydrogeologic units with sharp contrast in their hydraulic properties. Such diversion occurs in the perched-water bodies of the UZ (BSC 2003 [163045], Section 6.2.2; DOE 1998 [100550], p. 3–112). Laterally diverted flow ultimately finds a pathway to the water table through other, more permeable zones (e.g., faults). Advection is probably the most important transport process in this study, because it controls the speed at which radionuclides move through the UZ to reach the water table. Advection in the fractures is expected to be the dominant transport mechanism in many layers of the various hydrogeologic units. This is because the expected flow rates in the matrix exceed the matrix permeability (which gives the measure of flow capacity under the gravitational gradient). This leads inevitably to flow focusing in the more permeable fractures. Advection is the dominant transport mechanism in the fractures because of high permeability, limited fracture pore volumes, limited contact area, and short contact times between the radionuclide-carrying liquid phase and the matrix (only at the fracture walls). In a few hydrogeologic units, such as the CHv, Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 41 November 2003 matrix flow is dominant, resulting in much slower transport velocities (compared to those in the fractures of other units) and longer radionuclide-matrix contact times (DOE 1998 [100550], p. 3-112). 6.1.2.2 Hydrodynamic Dispersion Hydrodynamic dispersion combines mechanical dispersion, caused by localized velocity variations, with molecular diffusion, caused by Brownian motion and is proportional to the concentration gradient. The dispersion of the radionuclides can occur both along and transverse to the average flow direction. Hydrodynamic dispersion leads to the smoothing of sharp concentration fronts and reduces the breakthrough time (which can be defined as the arrival time of the edge of the contaminant front) at the water table. Little information exists on the values of the longitudinal (aL) and transverse (aT) dispersivity ([L]) in the various hydrogeologic units of the UZ. In past simulations (DOE 1998 [100550], p. 3-122), the longitudinal dispersivity (aL) values for both the fractures and the matrix of all units had a mean of 20 m and a standard deviation of 5 m. Analysis of the Northern Ghost Dance Fault test (LeCain et al. 2000 [144612], Tables 18–19) indicates an aL in the 0.06 m to 2.63 m range, but the validity of these estimates is uncertain because they are based on gas (rather than liquid) phase transport, short transport times (<200 min), and a small scale experiment (<10 m). Analysis of the C-well transport test indicate that aL ranged between 1.9 and 6.2 m (BSC 2003 [162415], Section 6.3) however, these results corresponded to the saturated conditions in the deeper formations of YM. Note that dispersion is not expected to play a significant part in the transport of radionuclides in the fractures because of the predominant role of advection as the main transport mechanism and the limited water flow into the matrix. Dispersion is affected by the scale of observation. The value of dispersivity appears to increase with the scale of observation (Gelhar et al. 1992 [122808], p. 1955; Fetter 1993 [102009], pp. 65–66). Thus, aL increased from 10–2 to 104 m when the observation scale increased from 10-1 to 105 m, and its value did not appear to be significantly affected by the texture (porous versus fractured) of the aquifer medium (Gelhar et al. 1992 [122808], p. 1955). The ratio aL / aT controls the shape of a contaminant plume in multidimensional transport. There is a paucity of data on the relationship between aL and aT. Based on that limited field data, Fetter (1993 [102009], pp. 65–66) reported that aL /aT ranged between 6 and 20. Gelhar et al. (1992 [122808], p. 1955) indicated that vertical aT is typically an order of magnitude smaller than the horizontal aL. 6.1.2.3 Sorption Sorption is a general term that describes a combination of chemical interactions between the dissolved radionuclides and the solid phases, that is, either the immobile rock matrix or colloids (mobile or immobile). In transport studies, the concept of sorption does not identify the specific underlying interactions, such as surface adsorption, precipitation, and ion exchange. By removing a portion of the dissolved species from the mobile liquid phase and transferring it to the immobile solid phase, sorption reduces the rate of advance of (i.e., retards) the concentration front of a dissolved or suspended species. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 42 November 2003 In Yucca Mountain studies, the effective Kd approach (see Equation 6-9) is employed to quantify the extent of radionuclide-rock interactions by measuring the overall partitioning between the aqueous and the solid phase. Three basic rock types (devitrified tuffs, vitric tuffs, and zeolitic tuffs) have been studied as having distinctively different (i.e., from each other) sorption interactions with the radionuclides (DOE 1998 [100550], p. 3-118). The estimation of Kd values for several radionuclides in each of these rock types is the subject of the detailed analysis of Section 6.4. Note that sorption is not only a function of the sorptive strength (as quantified by the value of Kd), but also of the contact time of the radionuclides with the rock matrix during transport through the UZ. The Kd values used in the Yucca Mountain studies have been estimated from batch experiments using crushed tuffs with a particle size of 75–500 µm under saturated conditions (Section 6.4, this Model Report; BSC 2001 [160828], Section 6.4). Under true linear equilibrium sorption conditions, the Kd value is independent of concentration or of the time of contact. The effective Kd value of a particular radionuclide sorbing onto a rock can vary depending on the estimation approach. Dependence of the effective Kd on concentration indicates nonlinear sorption, and Kd dependence on contact time indicates kinetic behavior. The importance of kinetic, nonlinear, and irreversible sorption may need to be evaluated against the linear equilibrium isotherm assumed in UZ transport studies. The Kd values obtained from experiments involving low concentration solutions tend to be higher than those from higher concentration ones because the most active sorption sites are immediately occupied by the dilute solutes (while denser solutes will involve less active ones, i.e., indicating nonlinear sorption isotherms such as Langmuir or Freundlich). The obvious implication is that Kd values from experiments involving concentrations similar to those expected in the field should be used to avoid overestimation or underestimation. Lower Kd values can be estimated if (a) the duration of the sorption experiments is short, and (b) a linear equilibrium sorption model is assumed but sorption is actually kinetically controlled. Note that the determination approach in the YM studies (Attachments I and II of present Model Report) tends to underestimate the Kd values because (a) it involves an averaging of Kd values obtained from experiments using both low and high solution concentrations, while (b) the release concentrations are expected to be low at the repository. If kinetic behavior is not an issue, batch experiments with dilute concentrations may overestimate the Kd values because of the aforementioned reasons, and because the prevailing conditions (involving saturated crushed rock samples in complete contact with a large amount of solute-bearing liquid relative to the sample mass) may not be representative of the UZ, which is characterized by unsaturated conditions, a limited contact area, and a limited amount of solute in the finite invading stream (see Section 6.1.3.1 for a more thorough discussion). Sorption kinetics could be important, especially for radionuclides sorbing onto zeolitic tuffs. Column experiments of 237Np transport in tuffs (Viswanathan et al. 1998 [134579], p. 267) indicate the existence of kinetic sorption under flowing conditions, possibly as a result of the slow diffusion of 237Np into the tuff pores. It is not possible to analyze these experimental data without considering a kinetic sorption model. Kinetic sorption can have a substantial effect on transport in fast fracture flow because it reduces sorption in the matrix, thus allowing larger Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 43 November 2003 radionuclide concentrations and migration over longer distances in the fractures. Nonlinear and irreversible sorption are evident from the diffusion and transport studies discussed in BSC (2001 [160828], Section 6.4). Note that, although there is an appreciation of the potential kinetic behavior, sorption is treated as a linear equilibrium process in this Model Report because (a) the kinetic effects are unlikely to persist over the long time frames (up to a million years) of this study, the duration of which strongly suggests an equilibrium process after a relatively short initial kinetic period, and (b) the generally low Kd values used in this study account for the early kinetic effects. 6.1.2.4 Matrix Diffusion Diffusion can play an important role in radionuclide exchange between the fractures and the rock matrix. This process transfers radionuclides into the matrix (where water flow is slow and sorption occurs), thus removing them from (and slowing the advance of their front in) the fast fracture flow. Diffusive flux across a given interface is a function of the concentration gradient, the temperature, the size of the dissolved species and its electric charge, the matrix pore structure, and the water saturation (DOE 1998 [100550], pp. 3-116–3-117). Matrix diffusion is a linear function of the tortuosity coefficient t, which describes the tortuous nature of the pore networks (Grathwohl 1998 [141512], pp. 28–35), including dead-end pores and steric hindrance (in extremely narrow pores). The term t is defined as the distance between any two given points in a porous medium over the actual travel path through the pores between these two points, i.e., t < 1. There are very limited experimental data on the t distribution in the various Yucca Mountain hydrogeologic units. In this study, t is approximated by the value of porosity f (Farrell and Reinhard 1994 [122803], p. 64; Grathwohl 1998 [141512], pp. 28–35). This approach to define t was confirmed experimentally from rock diffusion experiments using devitrified tuffs where the effective diffusion coefficient De (= f t Sw D0, where D0 is the molecular diffusion coefficient [L2 T-1] and Sw is the water saturation) for tritiated water was obtained from six tuff samples. Based on the literature value of D0 for tritiated water (Hu and Brusseau 1995 [122846]), a t estimate of 0.0806±0.035 was obtained for media in which the reported average porosity was 0.0767±0.019 (Triay et al. 1993 [145123], p. 1530 and Scientific Notebook SN-LBNL-SCI-168-V1 (Hu 2001 [163148], p. 23)). The good match between these two numbers validates the approach of using the porosity value as the tortuosity coefficient. Even less information exists on the effect of Sw on De. Porter et al. (1960 [123115]) reported that the t f Sw product decreased consistently in medium- and fine-textured soils as the capillary pressure increased from 0.33 to 15 atm. The effect of saturation on De may be more complex than the linear relationship currently assumed. Supporting evidence was provided from experiments by Conca and Wright (1990 [101582]), who determined De values for K+ ions for a variety of water contents and grain sizes on four types of angular crushed gravel. For volumetric water contents ranging from 0.5% to 6%, the De ranged from 10–14 m2/s to 10–11 m2/s, i.e., the dependence was much stronger than what could be expected from the linear relationship between De and Sw. The linear relationship assumed in this Model Report results in more conservative transport estimates because diffusion from the fractures into the matrix is reduced. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 44 November 2003 6.1.2.5 Solubility The concentration of any radionuclide (released from the stored radioactive wastes at the repository) in the water cannot exceed the radionuclide solubility limit, unless suspended colloids are involved. Limitations to radionuclide solubility in the water infiltrating the repository constitute the first barrier to transport and can reduce the extent of radionuclide migration by limiting the available source. The solubilities of the radionuclide of interest are reported in BSC (2001 [160828], Section 6). The lower solubility of many radionuclides can lead to slower and continuous release over time at the outer boundaries of engineered barrier systems surrounding the radioactive wastes because the source is not exhausted. For example, the solubility of 237Np is sufficiently low to extend the period of its release to tens of thousands of years (Viswanathan et al. 1998 [134579], p. 273). Note that, by using relative concentrations, radionuclide solubility is not explicitly accounted for in this Model Report. 6.1.2.6 Colloids Colloids are fine particles, between 0.001 and 1 µm in diameter, that become suspended and are transportable in a moving liquid. The generation and mobilization of colloids are considered important issues in contaminant transport, particularly the transport of radioactive true (intrinsic) colloids (e.g., colloidal Pu(IV) and Pu(V)) and the colloid-assisted transport of radioactive species (e.g., 239Pu, 237Np, 243Am, and 247Cm from high-level radionuclide wastes, or 137Cs, 90Sr, and 60Co from low-level radioactive wastes, see BSC (2001 [160828], Section 6)) sorbed on pseudocolloids (e.g., naturally occurring clay colloids). See Sections 6.1.3.3 to 6.1.3.6 for a more detailed discussion. 6.1.2.7 Perched Water Perched water is defined as a saturated zone located at a higher elevation than the static water table, to which it is not directly connected. Perched water usually occurs where low permeability horizons do not permit the rapid downward flow of water. Such bodies have been reported at several locations in UZ boreholes (Wu et al. 1999 [117167]). The presence of perched water has implications for the travel times and flow paths of water and radionuclides through the UZ. The majority of the perched-water bodies detected in the UZ boreholes were observed in formations overlaying relatively impermeable matrix material, such as the TSw basal vitrophyre (a glassy cooling unit). Although the vitrophyre is extensively fractured, many of the fractures have been filled with zeolitic material, thus limiting flow. A portion of the Calico Hills formation has been extensively altered to zeolites, creating perched-water bodies (Bodvarsson et al. 1999 [120055], pp. 14–15). The blockage of fracture flow which occurs below these perchedwater bodies can lead to lateral diversion of radionuclide migration if the percolation flux is sufficiently large. Three perched-water models, namely no perched water, perched-water model #1 (flow through the perched water), and perched-water model #2 (flow bypassing the perched water) have been proposed in BSC (2003 [163045], Sections 6.2.2 and 6.2.5). The effects of each of these Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 45 November 2003 perched-water models on radionuclide transport will be investigated in the 3-D site-scale simulations discussed in Section 6.15. 6.1.2.8 Daughter Products Chain-decay adds another layer of complexity because of the need to account for the transport of the daughter products, i.e., the new radionuclides created from the decay of a parent radionuclide. The daughter products may have significantly different transport behavior than the parent radionuclide. Thus, the migration and fate of all the important members of the decay chain must be considered, rather than just the parent radionuclide. 6.1.2.9 Effect of Heat on Transport Radionuclide transport may also be influenced by the heat generated by the decaying radioactive waste, which affects the ambient hydrological and chemical conditions and can thermally alter the rock near the repository. For example, if the zeolitic layers below the repository horizon are thermally altered, the sorption of radionuclides is reduced. In addition to the thermal effect on the flow field and radionuclide sorption coefficient, temperature also enhances diffusion by increasing Brownian motion. This effect is quantified by an increase in the D0 of radionuclides. The temperature-dependent D0 is described by the following equation (Robin et al. 1987 [123119], pp. 1105–1106): 2 1 0 0 T T T D T D .. . .. . µ = .. . .. . µ where T is the absolute temperature and µ is the viscosity of water. The effect of temperature on sorption has been discussed in BSC (2001 [160828], Section 6.4.6). In general, an increase in temperature leads to increased sorption of cationic species and decreased sorption of anionic species. In addition to theoretical predictions, BSC (2001 [160828], Section 6.4.6) includes a discussion of experimental data that show an increase of sorption by an order of magnitude when temperature increases from 25ºC to 75ºC. Temperatures in the UZ increase naturally with depth because of the geothermal gradient. Additionally, heat from the stored waste is expected to lead to higher temperatures over a large volume of the UZ for as long as 100,000 years (BSC 2003 [162050], Section 6.5). The combined effect of higher temperatures is increased retardation of the radionuclides because of (a) increased diffusion from the fractures into the matrix and (b) increased sorption. The assumption of isothermal 25ºC conditions in the studies of the present Model Report reflects a conservative approach and yields results that describe the worst-case scenario. 6.1.3 Important Transport Issues 6.1.3.1 Measurements of Kd Values The sorption behavior of radionuclides is usually described by the distribution coefficients (Kd), which quantifies the partitioning of radionuclides between the solid and aqueous phases under a linear equilibrium isotherm. Most of the Kd values are obtained from batch experiments using Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 46 November 2003 crushed rock. Effective Kd values obtained from batch experiments involving high concentration solutions will tend to underestimate the field Kd values if (a) the expected field concentrations are low and (b) nonlinear and/or kinetic sorption are involved (see discussion in Section 6.1.2.3, this Model Report). Conversely, batch experiments using crushed rock samples may overestimate the Kd values (compared to intact rock samples and for the same solute concentration) if kinetic effects are involved and the contact time is short, but the difference in the Kd estimates from the two sample types vanish for sufficiently long contact times. Crushing of the material creates new contact surfaces and also increases the radionuclide accessibility to pores that may not contribute to sorption to intact rocks. Comparison of the Kd values obtained from batch sorption experiments and from through-diffusion experiments (which involve relatively large samples) show significant differences, which are attributed to differences in the surface area of the samples and the water/solid ratio. A detailed discussion of sorption of important radionuclides onto various UZ rocks, and a listing of the corresponding Kd values, can be found in Triay et al. (1997 [100422], p. 181). Note that the Triay et al. (1997 [100422], p. 181) study involved analysis of laboratory data, a large portion of which are included in the Kd determination discussed in this Model Report (see Attachments I and II) is based. Bradbury and Stephen (1986 [122792]) investigated the sorption of 85Sr, 85Tc, 125I, and 137Cs onto Darley Dale sandstone samples. Comparing Kd values obtained from batch (using <0.1 mm and 1–2 mm) and diffusion-sorption experiments (using disks of 25 mm in diameter and about 5 mm in thickness), they determined that crushed rock tests can overestimate sorption by as much as one or two orders of magnitude. They concluded that the magnitude of the difference in the value of Kd depends on the radionuclide, its concentration, the rock, the water/solid ratio, and the particle size distribution in the batch tests. Holtta et al. (1997 [122832]) studied the effect of specific surface area on the sorption of 22Na, 45Ca, and 85Sr on crushed crystalline rocks of six size fractions (0.071–0.15, 0.15–0.20, 0.2–0.3, 0.3–0.85, 0.85–1.25, >1.25 mm). Sorption of 22Na, 45Ca, and 85Sr on unaltered tonalite and mica gneiss was slight and exhibited virtually no dependence on the fraction size. Considerably higher sorption of 22Na and 85Sr occurred on altered tonalites of smaller fractions because of their larger specific surface areas. Kd values from thin sections (0.030 mm in thickness) were in good agreement with those obtained from batch experiments, possibly because the thin-section thickness is within the range of the crushed-rock size fractions. Johansson et al. (1997 [123001]; 1998 [123004]) conducted batch sorption tests of alkali- and alkaline earth metals (22Na, 137Cs, 45Ca, 85Sr, and 133Ba) using six crushed size fractions (0.045-0.090, 0.090-0.25, 0.25–0.5, 0.5–1, 1–2, 2–4 mm) of medium-grained Aspo diorite and fine-grained granite. The Kd increased with decreasing particle size and differed by approximately one order of magnitude between the largest and smallest particle (because of different specific surface areas). They also determined that the best agreement between the Kd values from diffusion experiments (using 20 mm rock disks) and from batch experiments was observed for the largest size fractions (2-4 mm) in the batch studies. They explained this observation by suggesting that, unlike the smaller fractions, the largest size fraction involved a large number of whole mineral grains. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 47 November 2003 Tachi et al. (1998 [134571]) studied the sorption and diffusion behavior of Se in Tono tuffs in batch experiments (using crushed rock with sample sizes ranging between 0.075 and 0.355 mm) and in through-diffusion experiments (using samples 30 mm in diameter and 5 mm in thickness). The Kd values from the diffusion experiments were one order of magnitude lower than those from the batch experiments. Correcting for the difference in the specific surface areas could not fully account for the Kd discrepancy. Mercury porosimetry suggested that the differences were caused by sorption in microscopic pores (less than 20 nm in diameter) in the crushed rock samples. When the contribution of these pores to sorption was not considered, the Kd values from the batch sorption experiments and from the diffusion experiments were consistent. The effect of the specific area surface of ground rock samples on the sorption of 137Cs, 85Sr, and 237Np was studied in nine size fractions (< 0.038, 0.038–0.063, 0.063–0.075, 0.075–0.106, 0.106–0.25, 0.25–0.50, 0.5–1, 1–2, 2–4 mm) of devitrified Topopah Spring tuff and zeolitized Calico Hills tuff (Rogers and Meijer 1993 [123127]). They showed that the grinding process does not influence the sorption behavior for particle sizes larger than about 63 µm. They also indicated that ground samples must be washed carefully to remove very fine particles generated during the grinding process, which could lead to irreproducible or anomalously high Kd values. 6.1.3.2 Solute Sorption Issues 6.1.3.2.1 Potential Sorption of Anions Adsorption of negatively charged ionic species from the aqueous phase onto mineral substrates is likely to occur whenever the mineral surface exhibits a net positive surface charge. Under typical ambient conditions in shallow systems (saturated or unsaturated), most rock-forming alumino-silicate minerals have negatively charged surfaces. It is possible, however, for these surfaces to become positively charged in the presence of acids or if the surfaces dry out and only residual oligolayers of water remain. The transition from negatively to positively charged surface occurs at the point of zero charge (PZC), where the surface charge is zero and is usually expressed in terms of pH. The PZC differs from one mineral to another, but for most rock-forming minerals it is less than 4. However, in some materials (such as iron oxides, sulfides, and some organic matter), the PZC can be as high as 8 or 9. These minerals display a tendency to adsorb anionic species such as I– and 4 TcO- under ambient water table conditions. For example, the extent of 3 4 AsO - sorption on hydrous ferric oxide is a function of pH (Dzombak and Morel 1990 [105483], pp. 200–204). The edge sites on clays are also reported to contain positive sites. Under low pH conditions (e.g., pH = 5), significant sorption of anionic tracers occurs in systems with high concentrations of iron oxides and kaolinite (Boggs and Adams 1992 [122790]; Seaman 1998 [134563]). Despite the availability of positively charged adsorption sites, 129I– and 99 4 TcO- are in competition with other anionic species in solution, such as 24 SO - , 3 NO- , OH–, etc., which are usually present in concentrations an order of magnitude larger than either 129I– or 99 4 TcO- . Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 48 November 2003 Because of competition, their effective sorption is minimal. There are, however, exceptions, and these should be taken into account when evaluating their retardation. Kaplan and Serne (1995 [123010]) determined that iodine sorbed on all Hanford sediments, and its Kd ranged from 0.7 to 15 mL/g, with a median value of 7 mL/g. This finding could have significant implications for radionuclide transport. Positively charged adsorption sites may exist on the edges of 2:1 clays such as smectite and illite, on Al- and Fe-oxide surfaces, and on 1:1 clays such as kaolinite. Anions may sorb onto these locally positive-charged sites, even though their numbers are limited under typical pH conditions. Note that the study of Kaplan and Serne (1995 [123010]) involved extremely low iodine concentrations (approximately 12 parts per trillion (ppt)). Therefore, only a small number of positively charged sites would be needed for the sorption of a measurable portion of the dissolved iodine. Had an initially larger concentration been used in these experiments (e.g., 1 ppm), it is reasonable to expect that a small fraction of iodine would have sorbed onto the solid phase, resulting in practically undetectable iodine sorption. It must be pointed out that in a low-level waste plume 129I concentrations are expected to be in the ppt range (Kaplan and Serne 1995 [123010]). Similarly, significant sorption of technetium could occur in soils containing considerable amounts of natural organic matter, which tends to sorb anionic species and could reduce technetium to its +4 oxidation state, causing precipitation and/or sorption (Kaplan and Serne 1995 [123010]). In the Yucca Mountain UZ, however, conditions are both oxidizing and deficient in organic matter. Therefore, significant sorption is unlikely to occur until more reducing conditions are encountered within the saturated zone. Currently, there is no information on the occurrence and extent of localized positively charged sorption sites in Yucca Mountain tuffs. But the existence of such sites cannot be ruled out. Clays and oxides can potentially contribute to positively charged sites. Smectite is a significant constituent of some unwelded tuff horizons at Yucca Mountain. Although hematite and illite are present in minor or trace amounts in the unwelded tuffs, hematite is widely distributed in the matrix of the devitrified units. Smectite and hematite also coat fracture walls in Yucca Mountain tuffs (BSC 2001 [160828], Section 6). No estimate exists of the potential effects of such sites on the site-scale transport of radioactive anionic species. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 49 November 2003 6.1.3.2.2 Solute Size Effects Another aspect of technetium and iodine transport is the exclusion effect resulting from their relatively large ionic radius and low charge density. Technetium travels faster than tritium in both diffusion and column studies (BSC 2001 [160828], Section 6). This is attributable to the large size of the pertechnetate anion, which is excluded from tuff pores. In contrast, in crushed-rock column experiments, anion-exclusion effects for pertechnetate are essentially negligible, except in the case of technetium transport through zeolitic tuffs in the J-13 well water. In this case, the anion-exclusion effect was small but measurable (BSC 2001 [160828], Section 6). It should be emphasized that 3H+ can sorb weakly onto tuffs, as can 4 TcO- The faster 4 TcO- breakthrough curves (compared to those for 3H+) reported in BSC (2001 [160828], Section 6.5.3.3) should not be necessarily interpreted to indicate 4 TcO- size-exclusion, but could also denote larger 3H+ sorption. This is supported by the slight retardation of 3H2O reported by Hu and Brusseau (1996 [122849]). The apparent sorption is in essence ion-exchange and has been postulated to occur via OH exchange on the clay lattices. 6.1.3.3 Colloidal Behavior Radioactive true colloids or radionuclides adsorbed onto pseudocolloids can be transported over significant distances (McCarthy and Zachara 1989 [100778], pp. 496–497). The significant migration of strongly sorbing Pu and Am (more than 30 m) from a low-level nuclear waste site at Los Alamos National Laboratory through unsaturated tuff over a period of approximately 30 years is attributed to colloid and/or colloid assisted transport, a hypothesis confirmed by laboratory experiments (Buddemeier and Hunt 1988 [100712], p. 536). Using the 240Pu/239Pu isotope ratio to fingerprint the source of Pu in the water table, Kersting et al. (1999 [103282], pp. 56–59) recently demonstrated that the soluble (ionic) Pu is practically immobile in the subsurface of the Nevada Test Site (NTS) because of its strong sorption, but can be transported over significant distances (1.3 km over a 30-year period) in a colloidal form. Colloids are very fine particles (such as clay minerals, metal oxides, viruses, bacteria, and organic macromolecules) that range in size between 1 and 10,000 nm (McCarthy and Zachara 1989 [100778], pp. 497–498) and have high specific surface areas (~300 m2/g). Their chemical behavior is dominated by surface processes (EPRI 1999 [113923], p. 3-1), and can have a high sorptive capacity for contaminants. Colloids are deposited on porous and fractured media by surface filtration, straining filtration, and physical-chemical filtration. Colloid transport differs from solute transport because the colloidal particle interactions (e.g., flocculation), mechanical clogging effects, and surface reactions (e.g., deposition or attachment) are substantially different from solute processes and phenomena. A complete description of colloid-facilitated radionuclide transport requires consideration of a large number of processes (EPRI 1999 [113923], pp. 4-5 to 4-6), including advection, diffusion, colloid generation, colloid stability, colloid-solute-matrix interactions, affinity of colloids for the gas-water interface, colloid filtration (surface and straining), and kinetically controlled physical-chemical filtration. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 50 November 2003 6.1.3.3.1 Colloid Types and Classes The analysis in BSC (2003 [161620], Section 6.3.1) identifies the following types of colloids based on their origin and characteristics: (a) True (or intrinsic) colloids are generated from a solute when its concentration exceeds its solubility (Saltelli et al. 1984 [117338]). Such true Pu(IV) colloids have been produced by the agglomeration of hydrolyzed Pu(IV) ions under acidic conditions (EPRI 1999 [113923], p. 3-2). When immature, actinide true colloids display hydrophilic properties, but become hydrophobic with increasing age. (b) Waste form colloids are formed from the nucleation of colloids from waste form dissolution and from spallation of colloid-sized waste from alteration products. Waste form colloids are believed to be one of the most significant contributors to radionuclide transport in the UZ. (c) Pseudocolloids are all other colloidal particles, i.e., natural colloids that can be inorganic (e.g., clay, iron oxyhydroxides, silica) or organic (microbes and humic acids (Ibaraki and Sudicky 1995 [109297], p. 2945). Pseudocolloids become radioactive when solute actinides sorb onto them. (Note that in BSC (2003 [161620], Section 6.3.1) only radioactive pseudocolloids are referred to as such, while nonradioactive ones are referred to as seepage/groundwater colloids). In terms of mathematical description of their radioactive and transport behavior, these colloids are classified as follows: Class I: In this class the entire non-aqueous component of the colloidal particle is radioactive. True colloids and waste form colloids are Class I colloids. It is not generally known if decayinduced recoil in such colloids is sufficiently strong to cause ejection of daughters and shrinkage of the colloidal particle, if the colloidal size is maintained but its density is reduced after daughter ejection, or if the daughters remain trapped within the colloidal structure. If there is ejection, then either the colloidal size or the density decreases at a rate one-third that of decay (leading to larger diffusion). Class II: In this class only a portion of the colloidal particle (usually a very small one) is radioactive. This class includes radioactive pseudocolloids, in which the actinide is irreversibly sorbed onto the colloid or incorporated into the colloidal structure (e.g., through ion exchange). In this case, the actinide remains confined onto the colloid and does not exchange mass with its surroundings (i.e., the liquid phase or adjacent colloids). If the time of actinide production is the same (a valid approximation in YM studies, especially in radionuclides with long half lives), then the colloidal particle concentration in the liquid and solid phases, and the radionuclide concentration on each colloid are independent of each other and can be computed separately. Radioactive decay does not change the dimensions of Class II colloids because, even in the event of daughter ejection due to recoil, the actinide mass represents a very small portion of the total colloid mass and the colloidal structure is maintained. Compared to the transport of Class I colloids of the same size with decaying sources, Class II colloids result in lower radioactivity Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 51 November 2003 concentrations (because actinides are a small portion of the total mass), but the relative concentrations (with respect to that at the release points) are the same. Class III: Class III includes radioactive pseudocolloids in which actinides are reversibly sorbed onto the underlying natural colloid. As in Class II colloids, only a small portion of the colloidal particle is radioactive. In this case, the actinide remains is not confined onto the colloid, but can exchange mass with its surroundings (i.e., the liquid phase or adjacent colloids). For the same reasons discussed in Class II colloids, radioactive decay does not change the dimensions of Class III colloids. However, determination of the radioactivity concentration is much more complex in Class III colloids than in Class I and II colloids (the transport equations of which can be linearized) because the corresponding equations are nonlinear (see Equation 6-17, Section 6- 2). A fourth class of colloids, Class IV colloids, includes all non-radioactive colloids, and can belong to any colloid type (true colloids, waste form colloids, or pseudocolloids). Such colloids are not considered in this study of UZ transport. The study of transport of these colloids is identical to that of the colloidal particle of Class II hydrates. More details on the mathematical treatment of the transport of colloids can be found in Section 6.2 (present Model Report). 6.1.3.3.2 Colloid Generation and Stability The formation of mobile colloidal suspensions in the subsurface is attributed to a number of mechanisms: (1) matrix dissolution caused by changes in pH or redox conditions; (2) supersaturation with respect to the inorganic species; (3) disruption of the mineral matrix by large changes in the flow regimes due to injection, pumping, or large episodic rainfall infiltrations; (4) release and movement of viruses and bacteria; and (5) formation of micelles from the agglomeration of humic acids (Abdel-Salam and Chrysikopoulos 1995 [146647], pp. 199–200). Buddemeier and Hunt (1988 [100712], p. 536) indicate that submicrometer colloids can easily be released from mineral and glass surfaces that are chemically, hydrodynamically, or mechanically stressed. Colloid stabilization/destabilization include steric stabilization by mechanisms such as organic coating of inorganic colloids, and the effects of pH and ionic strength on coagulation and precipitation. The stability of colloid suspension is very sensitive to changes in ionic strength (EPRI 1999 [113923], p. 4-6). 6.1.3.3.3 Colloid Deposition Colloid deposition (physical-chemical filtration) during saturated flow through a porous medium is commonly assumed to occur in two steps: (1) transport of colloids to matrix surfaces by Brownian diffusion, interception, or gravitational sedimentation (i.e., colloid-matrix collision) and (2) attachment of colloids to matrix surfaces. The attachment efficiency (i.e., the fraction of collisions resulting in attachment) is strongly influenced by interparticle forces between colloids and matrix surfaces, such as van der Waals and electric double-layer interactions, steric stabilization, and hydrodynamic forces (Kretzschmar et al. 1995 [123019], p. 435). Kretzschmar et al. (1997 [123029], p. 1,129) demonstrated that colloid deposition generally follows a Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 52 November 2003 first-order kinetic rate law and experimentally determined the corresponding collision efficiencies. 6.1.3.3.4 Colloid–Contaminant–Matrix Interactions Colloid attachment to the host rock is strongly dependent on electrostatic interactions. Once attached, colloid detachment (declogging) is generally slow to irreversible. Sorption of radionuclides on colloids is controlled by a range of chemical processes such as ion exchange, surface complexation, and organic complexation (EPRI 1999 [113923], p. 4–9). If sorption of metal ions onto colloids is assumed to follow an equilibrium isotherm, metals are stripped very fast from the colloids when these enter a clean part of the porous medium. This approach was incapable of explaining the long transport distances observed in field experiments (van de Weerd and Leijnse 1997 [109249], p. 246). The problem was addressed by assuming (a) kinetic sorption of radionuclides onto the humic colloids and (b) kinetic colloid deposition (van de Weerd and Leijnse 1997 [109249], pp. 245, 255). The sorption of dissolved ionic Pu(IV) onto hematite, goethite, montmorillonite, and silica colloids in both natural and carbonate-rich synthetic water table was reported to be fast (BSC 2001 [160828], Section 6). Under equilibrium conditions, the Kd values of Pu(V) was about 100 mL/g for hematite and montmorillonite colloids. These very large values indicate that iron oxide and clay colloids in the water table can significantly enhance the transport of 239Pu. Very large Kd values of actinide sorption are also confirmed in BSC (2003 [161620], Section 6.3.3.1). The values are so large that they can support an approximation of treating such radioactive Class III colloids as Class II colloids. 6.1.3.3.5 Colloid Migration in Macroporous and Fractured Systems In conducting infiltration experiments with intact structured sandy loam cores using two types of colloidal suspensions, Jacobsen et al. (1997 [122995], pp. 185–186) observed significant transport of clay and silt colloid particles through the macropores. Macropores can enhance the transport of colloids because all types of filtration are less pronounced in large pores, and the large water velocities can lead to increased detachment (as the large hydrodynamic forces can overcome the colloid-grain bonding forces). In a study of colloid-facilitated transport of radionuclides through fractured media, Smith and Degueldre (1993 [144658], pp. 143, 162–163) determined that the assumption of fast, linear, and reversible radionuclide sorption onto colloids is non-conservative. They observed that the time for radionuclide desorption (days or weeks) significantly exceeded the time for sorption (seconds or minutes). Such radionuclide-laden colloids can migrate over long distances in macropores and fractures, and the larger colloids exhibit little retardation because their size prevents them from entering the wall-rock pores. The Smith and Degueldre (1993 [144658], pp. 143, 162–163) study concluded that the transport of radionuclides sorbed irreversibly on to colloids depends strongly on the extent of colloid interaction with the fractures. Vilks and Bachinski (1996 [109252], pp. 269, 272–278) studied particle migration and conservative tracer transport in a natural fracture within a large granite block, with overall dimensions of 83×90×60 cm. Flushing experiments showed that suspended particles as large as 40,000 nm could be mobilized from the fracture surface. The mobility of suspended particles Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 53 November 2003 with diameters of 1,000–40,000 nm was significantly less than that of colloids (<90 nm). They observed that the conservative tracer lagged slightly behind the colloid front, but colloid mobility was significantly reduced when the average water table velocity decreased. Compared to dissolved tracers, the migration of colloids was more affected by the flow path and flow direction. This tendency can have a significant effect in the fracture-dominated flow of the Yucca Mountain UZ. In field-scale colloid migration experiment, Vilks et al. (1997 [109254], pp. 203, 212-213) showed that silica colloids as small as 20 nm can migrate through open fractures over distances of 17 m, when a colloidal suspension was injected at average flow velocities of 1.6 and 2.9 m/h (orders of magnitude higher than the natural flow rates at the site). Based on the analysis of the results of the study, they also suggested that although colloid migration appeared to behave conservatively, colloids may have followed different pathways than dissolved conservative tracers. 6.1.3.4 Colloidal Behavior in Unsaturated Media Most of the available literature focuses on the behavior of colloids in saturated media. The behavior of colloids in unsaturated media (BSC 2003 [162729]) is a relatively new area of study and has significant additional complexity. 6.1.3.4.1 Filtration as a Function of Water Saturation Colloid filtration is a physical retardation process. Wan and Tokunaga (1997 [108285], p. 2413) distinguish two types of straining: conventional staining (if the colloid is larger than the pore throat diameter or the fracture aperture) and film straining (if the colloid is larger than the thickness of the adsorbed water film coating the grains of the rock). Wan and Tokunaga (1997 [108285], pp. 2413, 2419) developed a conceptual model to describe colloid transport in unsaturated media as a function of water saturation Sw. If the rock Sw exceeds a critical saturation value Sc, colloids move through the system entirely within the aqueous phase. For Sw < Sc, colloids can only move in the thin film of water that lines the grain boundaries, and colloid transport through the water film depends on two parameters: the ratio of the colloid size to the film thickness, and the flow velocity. Temporal variations in the Sw in the subsurface profile and in the infiltration rate can lead to strongly nonlinear colloid mobility in the vicinity of Sc. McGraw and Kaplan (1997 [123043], p. 5.2) investigated the effect of colloid size (from 52 to 1900 nm) and Sw (from 6 to 100%) on colloid transport through unsaturated media in Hanford sediments. They showed a very strong dependence of filtration on the colloid size under unsaturated conditions. At a volumetric water content of 6%, (the expected water content in the Hanford vadose zone), colloid removal increased exponentially with colloid size. The decrease in colloid mobility at low volumetric contents was attributed to resistance due to friction (as the colloids were dragged along the sand grains). Colloid retardation increased as the ratio between the water film thickness and colloid diameter decreased. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 54 November 2003 6.1.3.4.2 Gas–Water Interface Under unsaturated conditions, colloid transport may be either inhibited or enhanced (compared to saturated conditions) because of the presence of the air-water interface. Wan and Wilson (1994 [114430], pp. 857, 863) determined that retention of both hydrophilic and hydrophobic colloids increased with the gas content of the porous medium. They showed that colloids preferentially concentrate in the gas-water interface rather than on the matrix surface. This tendency increases with the colloid surface hydrophobicity or contact angle, with hydrophobic colloids having the strongest affinity for the gas-water interface. The implications of this colloid behavior are important. Colloid affinity for the gas-water interface may retard their transport through the unsaturated zone. Conversely, if the subsurface conditions permit the stability and migration of bubbles, colloid transport may be enhanced. 6.1.3.5 Colloids at Yucca Mountain The available data on colloid occurrence and concentrations at the Yucca Mountain site are limited, and pertain to saturated zone studies. There is no published information on colloids in the unsaturated zone. The current state of understanding is that (a) the natural colloid concentrations in native waters at the Yucca Mountain site are in the 106–1010 particles/mL range, and (b) waste-form colloids will be the dominant colloidal species of concern to transport studies (Kung 1999 [146992], pp. L-I-5, L-8-3 to L-8-5). This conclusion is supported by BSC (2003 [161620], Section 6), which includes the most recent information on colloid occurrence, generation, stability, properties and transport behavior. Additional information from nearby sites, however, cannot be ignored. Preliminary results from a survey of the water table from the Nevada Test Site (NTS) have shown that, in the Pahute Mesa, drainage (both on and off the NTS) has colloidal particle (>3 nm in diameter) loadings of 0.8–6.9 mg/L. Such relatively high concentrations can explain the observed transport of strongly sorbing radionuclides over significant distances, particularly because the ionic composition of the NTS water table is not expected to promote the coagulation of clay colloids (Buddemeier and Hunt 1988 [100712], p. 537). This is because NTS water table is oxygenated and low in organic matter, and a substantial fraction of the colloidal material is composed of stable natural minerals (Buddemeier and Hunt 1988 [100712], pp. 543–544). In addition to natural colloids, anthropogenic colloids may be created from the waste itself or from repository construction and sealing materials. Waste- and repository-derived colloids at Yucca Mountain are likely to include organic colloids, iron oxyhydroxides, and aluminosilicate colloids (EPRI 1999 [113923], p. 2-2, 6-22). In a 50-month experiment involving simulated weathering of a high-level nuclear waste glass, spallation and nucleation were identified as the main mechanisms of colloid genesis (Bates et al. 1992 [100704]). The created colloids were identified as inorganic. The same study determined that Pu and Am released from waste were predominantly in the colloidal, rather than in the dissolved, form. 6.1.3.6 Impact of Colloids on Radionuclide Transport in Yucca Mountain EPRI (1999 [113923], pp. xv–xvii and 2-2) indicated that it is important for the Yucca Mountain Project (YMP) to consider the migration of radionuclides, and to assess the potential role of Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 55 November 2003 colloids on radionuclide transport, within both the saturated and the unsaturated systems. Furthermore, EPRI (1999 [113923], pp. 4-6 to 4-7) identified the following key areas of uncertainty for further study: 1. Colloid-matrix interactions. 2. Thermodynamic data for certain radionuclides and kinetic data to describe virtually all rate-dependent reactions. 3. Representation of the hydrological system: colloid transport differs from the transport of conservative tracers and is influenced by the details of the permeability structure, such as fracture geometry and the hydraulic conditions within the fracture zones. 4. The effects of partially saturated conditions on colloid transport. 6.1.4 Important Points for Consideration in Colloid Transport The assessment of the potential role of colloids in the transport of radionuclides at Yucca Mountain is particularly challenging for the reasons discussed in the previous sections, and is further complicated by the following factors: 1. Different types of colloids (minerals, organics, microbes, and polymeric actinide colloids) are involved, with substantially different characteristics and properties (e.g., hydrophobicity and surface charges) that strongly affect their behavior in the subsurface. 2. Colloid generation, stability, size distribution, and concentration are dynamic. Knowledge about the temporal and spatial distributions of colloid populations in the different hydrogeologic units of the UZ is difficult to obtain. Changes in the ionic strength of the aqueous solution (e.g., when percolating water in an episodic infiltration event encounters resident pore water) can affect colloid stability. There are also uncertainties about the colloid stability in the aqueous phase of the UZ and about the colloid generation from mineral-coated fracture walls (given the predominance of fracture flow in the UZ). 3. Measurements of colloid concentration in the UZ are difficult. Because of the low concentrations and low saturations in the UZ, the determination of colloid concentration in pore water is challenging, and serious questions arise about sample representativity and integrity. 4. The concentration of waste-form colloids is a key uncertainty. There is evidence to suggest that the low concentration of natural colloids in the Yucca Mountain UZ will not lead to significant colloid-assisted transport, and that waste-form colloids will be the dominant colloidal transport problem (Kung 1999 [146992], pp. L-1-5, L-8-3 to L- 8-5). Thus, there is a need for reliable estimates of the types and generation rates of waste-form colloids. Currently, there is considerable uncertainty on this subject (EPRI 1999 [113923], p. 6-7). Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 56 November 2003 Important processes to consider are: 6.1.4.1 Colloid Diffusion Colloids diffuse more slowly than dissolved species because of their larger size (see Section 6.2.5, Equation 6-23). For the largest colloids, diffusion is approximately three orders of magnitude slower than that for molecular species (Nuttall et al. 1991 [123103], p. 189). For example, the diffusion coefficient D0 of a 0.1 µm colloid at 20ºC is 4.29×10-12 m2 /sec (see Equation 6-23 in the present Model Report), while the D0 of - r B is 2.08×10-9 m2 /sec (Cussler 1984 [146653], p.147). 6.1.4.2 Pore Exclusion Using mercury porosimetry, Roberts and Lin (1997 [101710], pp. 577–578) determined that the average pore diameters of welded and densely welded TSw tuff samples were 53.1 nm and 19.7– 21.4 nm, respectively. These extremely small pores are certain to exclude a significant proportion of colloids. It is possible for colloids to accumulate on the fracture walls and thus clog the matrix pores open to the fracture. This can lead to reduction in the matrix permeability and in the colloid diffusion into the matrix. Pore exclusion is not expected to be significant in the fractures, and colloids can travel significant distances in the fractures (especially given the limited diffusion into the matrix). 6.1.4.3 Role of Air-Water Interface The affinity of colloids for the air-water interface depends on their hydrophobicity and electrostatic charge. Hydrophilic colloids, such as mineral fragments, have a low affinity for the interface, in contrast to hydrophobic colloids (such as organic colloids and microbes). This affinity increases with the positive charge on the colloids (EPRI 1999 [113923], p. 6-11). The flow and saturation conditions in the UZ will determine whether this will enhance or retard transport. Note that the potential impact of the air-water interfaces on colloid transport has not yet been quantified. 6.1.5 Modeling Stipulations In this section we discuss the basic stipulations in the mathematical basis of the transport model, and we provide a short discussion of the supporting rationale. This section is structured as follows: Section 6.1.5.1 addresses the general stipulations underlying the flow component of the transport model of the Yucca Mountain unsaturated zone (UZ). Section 6.1.5.2 lists the assumptions of the transport processes. The stipulations involved in the treatment and mathematical representation of the fractured rocks using the dual-continuum approach are discussed in Section 6.1.5.3. Stipulations related to the initial and boundary conditions of the model domain are presented in Section 6.1.5.4. Sections 6.1.5.1 to 6.1.5.4 address issues related to large-scale 3-D numerical simulations. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 57 November 2003 6.1.5.1 Stipulations Involved in the Flow Component of the Transport Processes The transport phenomena and processes in the Yucca Mountain UZ have a flow component and a transport component, each of which are described by a set of governing equations. These conserve mass, energy and momentum in the system under study, while quantifying the system response to external mass and energy inputs and interrelationships between the various processes involved. The basic flow stipulations are consistent with those discussed in UZ Flow Models and Submodels (BSC 2003 [163045], Sections 5 and 6). They share identical conceptual models, which are stated below. 1. The macroscopic-continuum approach is a valid concept for the description of the flow and transport processes in the fractured UZ rocks. Justification: The rationale for this approach is provided in BSC (2003 [163045], Section 6.1). Applicability: This approach is used in all numerical (i.e., 3-D site-scale) simulations of flow and transport in this Model Report (Sections 6.8 to 6.20, and Section 7). No further confirmation is required for the purposes of this study. 2. Darcy’s law is a valid model to describe the flow of gas and water in the matrix and fractures of the UZ. Justification: Given the applicability of the macroscopic continuum approach, the gaseous and aqueous flows under ambient conditions in the UZ are sufficiently slow to correspond to a Reynolds number = 10, i.e., the upper limit of applicability of Darcy’s law (Bear 1972 [156269], p. 127). Applicability: This approach is used in all 3-D site-scale simulations of flow and transport in this Model Report (Sections 6.8 to 6.20, and Section 7). No further confirmation is required for the purposes of this study. 3. Richards’ equation (Richards 1931 [104252]) is a valid model of unsaturated water flow in both the matrix and the fractures of the UZ. Justification: Under ambient conditions, the gas-phase pressure in the UZ is atmospheric, corrected for elevation. The absence of gas pressurization makes possible the adoption of Richards’ equation, because the gas phase can be neglected and the aqueous phase flow occurs in response to gravitational and capillary pressure differentials. Applicability: This approach is used in all 3-D site-scale simulations of flow and transport in this Model Report (Sections 6.8 to 6.20, and Section 7). No further confirmation is required for the purposes of this study. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 58 November 2003 4. Relative permeabilities and capillary pressures follow the van Genuchten (1980 [100610]) and Mualem (1976 [100599]) model and are continuous functions of the effective liquid and gas saturations. Justification: This model is consistent with the macroscopic continuum approach, has been successfully used to determine relative permeability and capillary pressure parameters from UZ rocks (matrix blocks), and is reasonable for fractures (BSC 2003 [160240], Section 6.1). Applicability: This approach is used in all 3-D site-scale simulations of flow and transport in this Model Report (Sections 6.8 to 6.20, and Section 7). No further confirmation is required for the purposes of this study. 5. The water flow is isothermal. Justification: This is a reasonable approximation. The flow parameters affected by temperature are water density and water viscosity. Ambient temperature extremes at the domain boundaries are about 20°C at the top and about 30°C at the water table at the bottom (BSC 2003 [163045], Section 6.3). Between 20°C and 30°C, the water density decreases from 998.21 kg/m3 to 995.65 kg/m3 (Lide 1993 [123032], p. 6–10), i.e., the change is very small. The effect on viscosity is more pronounced. Between 20°C and 30°C, the water viscosity decreases from 1.002x10–3 Pa · s to 0.977x10–4 Pa · s (Lide 2002 [160832], p. 6–10), i.e., a reduction of about 20%. Although this variation is not excessive (given the uncertainty in the values of the UZ system hydraulic properties), its effects are minimized by conducting the isothermal flow simulations at 25°C. Isothermal flow may be a less valid approximation in the immediate vicinity of the waste package because of the heat generated by the radioactive decay process. The nonisothermal flow under these conditions is the subject of another report (BSC 2003 [162050]). Applicability: This approximation is used in all 3-D site-scale simulations of flow and transport in this Model Report (Sections 6.8 to 6.20, and Section 7). No further confirmation is required for the purposes of this study. 6. Water flow through the UZ in the numerical simulations of radionuclide transport is considered time-invariant (steady-state). Justification: This is a reasonable approximation, given the very long simulation periods (= 100,000 years). This approach allows the determination of the transportbehavior envelope by considering nine different infiltration scenarios (and, consequently, flow fields), and is consistent with the flow regimes discussed in BSC (2003 [163045], Section 6). Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 59 November 2003 Applicability: This approximation is used in all 3-D site-scale simulations of flow and transport in this Model Report (Sections 6.8 to 6.20, and Section 7). No further confirmation is required for the purposes of this study. 6.1.5.2 Stipulations/Approaches/Simplifications Involved in Transport Processes 7. The individual and combined effects of diffusion (molecular and/or colloidal), surface diffusion, and hydrodynamic dispersion follow a Fickian model. Justification: Given the macroscopic continuum approach, this is a valid model (de Marsily 1986 [100439], pp. 228–277). Applicability: This model is used in all studies in this Model Report (Sections 6.8– 6.20, and Section 7). No further confirmation is required for the purposes of this study. 8. Transport occurs isothermally at 25°C. Justification: Isothermal transport at the average ambient temperature of the UZ is consistent with the assumption of isothermal flow at the same temperature. The transport parameters affected by temperature are (a) the diffusion coefficient D0 of the dissolved or colloidal species and (b) the sorption parameters of the dissolved species or the filtration parameters of the suspended colloid. Natural temperature differentials in the undisturbed UZ profile occur because of the geothermal gradient. Substantial temperature increases over the ambient are expected after radioactive waste emplacement in the repository (BSC 2003 [162050], Section 6). An increasing temperature leads to a higher D0 value according to the relationship discussed in Section 6.1.2.9 of this Model Report. Based on this relationship, an increase in temperature from 20°C (at the top of the UZ domain) to 30°C (at the water table, i.e., the bottom of the UZ domain) leads to an increase of D0 by about 30%. The effect of temperature on sorption and/or filtration is less well-defined. The general effect of an increasing temperature is a decrease in the sorption of anionic species and an increase in the sorption of cationic species. Theoretical and experimental studies indicate an increase of the distribution coefficient Kd with temperature when sorption follows a linear equilibrium isotherm (BSC 2003 [162050], Section 6). Colloid filtration (deposition) generally follows a kinetic process (see Sections 6.2.3 and 6.18.2 in this Model Report). Equation 6-31 in this Model Report indicates that an increase in temperature increases the forward filtration coefficient .+, indicating an increase in the filtration (deposition, clogging) rate. There is no information on the effect of temperature on the reverse filtration coefficient .–. Thus, an increasing temperature in the UZ enhances diffusion (a particularly important mechanism in species mass transfer from the flow-dominating fractures to the matrix) and increases sorption and/or filtration. The cumulative effect is slower transport. The assumption of isothermal transport should not be viewed as an approximation of the Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 60 November 2003 prevailing conditions in the UZ, but rather as a condition that reflects a worst-case transport scenario and leads to conservative estimates of radionuclide travel times to the water table. Investigation of the effect of water-phase changes on transport may be included in future revisions of the present Model Report. Applicability: This simplification is used in all studies in this Model Report (Sections 6.8 to 6.20, Section 7). No further confirmation is required for the purposes of this study. 9. The concentration of the radioactive solutes or colloids is at a tracer level, i.e., too low to have any measurable effect on the flow regime. Justification: Ambient tracers and radionuclides escaping from the repository are expected to occur at concentrations that are too low to affect the aqueous solution density (BSC 2003 [162050], Section 6). Applicability: This approximation is used in all studies in this Model Report (Sections 6.8 to 6.20 and Section 7). No further confirmation is required for the purposes of this study. 10. There is no phase change, i.e., no water evaporation and condensation. Justification: The rationale for this simplification is covered by the discussion in Stipulation (5). Water evaporation and condensation resulting from the heat generated by the radioactive decay of the wastes is covered by BSC (2003 [162050], Section 6), in which it is shown that phase changes are not expected to last longer than the first 10,000 years after waste emplacement (i.e., a rather short time compared to the 100,000 to 1,000,000 years covered by the studies in this Model Report) and are limited to a rather small volume in the immediate vicinity of the repository. However, investigation of the effect of water-phase changes on transport may be included in future revisions of the present Model Report. Applicability: This simplification applies to all studies in this Model Report (Sections 6.8 to 6.20, and Section 7). No further confirmation is required for the purposes of this study. 11. Filtration of colloids is limited to deep filtration, i.e., it does not affect the medium porosity and permeability. Justification: Pseudocolloids (i.e., colloidal waste forms) and natural pseudocolloids (such as clays) under natural conditions are present in sufficiently small concentrations (Kung 1999 [146992], pp. L-1-5, L-8-3 to L-8-5) to justify this simplification. This reference also indicates that, as a result of adverse chemical conditions (e.g., pH, ionic strength) in the immediate vicinity of the repository, true colloids (i.e., radionuclides in colloidal form) will be released in low concentrations for a long time. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 61 November 2003 Applicability: This simplification applies to the colloid studies in this Model Report (Section 6.18). No further confirmation is required for the purposes of this study. These stipulations allow decoupling of the flow and transport equations. Richards’ equation is first solved in the flow component of transport, followed by the sequential solution of the n independent tracer transport equations. 6.1.5.3 Stipulations Involved in the Dual-Continuum Approach The treatment of fracture-matrix interactions is a critical issue in the simulation of flow and transport under the two-phase flow conditions of the fractured UZ rocks. This Model Report closely follows the approach of UZ Flow Models and Submodels (BSC 2003 [163045], Section 6): 12. The dual-permeability model is a valid approximation for flow and transport simulations of the fractured UZ rocks. Justification: In addition to its computational efficiency, this model has a strong conceptual basis because it describes the matrix and fractures as separate but interconnected gridblocks and permits flow and transport between matrix gridblocks, fracture gridblocks, and fractures and matrix. A more detailed discussion on the rationale for this assumption can be found in (BSC 2003 [162267], Section 5) Applicability: This approach applies to all 3-D site-scale studies in this Model Report (Sections 6.8 to 6.20). No further confirmation is required for the purposes of this study. 6.1.5.4 Stipulations Involving Initial and Boundary Conditions 13. In the numerical simulations the top boundary of the UZ model domain is maintained at spatially variable conditions of (a) Temporally constant gas pressure and saturation; (b) Temporally constant temperature; (c) Temporally constant infiltration rates (steady state). Justification: The spatial variations of the parameters reflect differences in elevation, climate and topography, and represent realistic approximations of the prevailing conditions. For sufficiently long simulation periods, temporal variations in these parameters tend to diminish. Moreover, at a relatively short distance below the land surface, such temporal variations diminish rapidly. Applicability: These boundary conditions are used in all 3-D site-scale studies in this Model Report (Sections 6.8 to 6.20). No further confirmation is required for the purposes of this study. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 62 November 2003 14. The bottom boundary of the UZ model domain in the 3-D simulations coincides with the water table. It is maintained at spatially variable conditions of (a) Temporally constant water pressure and saturation; (b) Temporally constant temperature. Justification: This is a good representation of conditions in the saturated zone, for the reasons discussed in the rationale for Assumption 13. Applicability: These boundaries conditions apply to all 3-D site-scale studies in this Model Report (Sections 6.8 to 6.20). No further confirmation is required for the purposes of this study. 15. In the numerical studies of radionuclide transport, the boundaries of the UZ domain through which flow and transport occur are the top and bottom boundaries (i.e., the ground surface and the groundwater, respectively). No lateral flow and/or transport occur across any other boundaries. Justification: The flow and transport through the top and bottom boundaries is consistent with the patterns of rainfall-fed infiltration and gravity-driven flow and drainage. The distance between the repository and these boundaries is sufficiently large to justify the assumption of no lateral flow and/or transport. Applicability: These boundaries conditions apply to all 3-D site-scale studies in this Model Report (Sections 6.8 to 6.20). No further confirmation is required for the purposes of this study. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 63 November 2003 6.2 MATHEMATICAL MODEL OF TRANSPORT The mathematical model of the flow component in this Model Report is thoroughly discussed in BSC (2003 [163045], Section 6). Here we focus on the mathematical model of transport and discuss the implications of the various processes for radionuclide transport. There are eight subsections in Section 6.2. The basic mass-balance equation of solutes and colloids are discussed in Section 6.2.1. Section 6.2.2 focuses on the accumulation terms of the mass balance equation. The equations of sorption (for solutes), filtration (for colloids) and colloid-assisted transport of solutes are described in Sections 6.2.3, 6.2.4, and 6.2.5, respectively. The flux terms of the mass-balance equation (Section 6.2.1) are described mathematically in Section 6.2.6. Radioactive decay is discussed in Section 6.2.7, and Section 6.2.8 describes the equations of transport for the daughter products of radioactive decay. 6.2.1 General Mass Balance Equations Following Pruess (1987 [100684]; 1991 [100413]), mass balance considerations in a control volume dictates that d dt () Vn M.dV = () Gn F.•ndG + () Vn q.dV (Eq. 6-1) where V, Vn = volume, volume of subdomain n[ L3] M. = mass accumulation term of component (tracer) . [ML - 3] G, Gn = surface area, surface area of subdomain n [ L2] F. = Darcy flux vector of component . [ML - 2T - 1] n = inward unit normal vector [ L0] q. = source/sink term of tracer . [ML - 3T - 1] t = time [ T] . The conservation of mass for any subdomain n is given by Equation 6-1, which in space-discretized form assumes the form of the following ordinary differential equation: . . .dM. dt . . . n = 1 Vn .m Anm( F.) nm + ( q.) n (Eq. 6-2) where Anm is the surface segment between elements n and m [L2] , (F.)nm is the mass flux of tracer . between elements n and m [M L-2 T -1], and (qi)n is the mass rate of the source/sink of tracer . in element n [ML - 3T - 1] . Equation 6-2 is general, and applies to solutes and colloids. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 64 November 2003 6.2.2 Accumulation Terms 6.2.2.1 Equations of the Accumulation Terms The accumulation term M of a tracer . (solute or colloid) in a porous or fractured medium (PFM) is given by M M M M M M . , . , . , . , . , . + d + + ... = F L c A c g A L s d i o l l o c r o f s e t u l o s r o f (Eq. 6-3) where ML,. = the mass of tracer . in the aqueous phase [ML - 3] MAg,. = the mass of solute tracer . adsorbed onto the PFM grains [ML - 3] MAc,. = the mass of solute tracer . adsorbed onto colloidal particles [ML-3] MF,. = the mass of filtered colloidal tracer . [ML-3] and the parameter . . . . . . = d solutes for 0 colloid III Class for 1 C (Eq. 6-4) A detailed discussion on the different colloid classes can be found in Section 6.1.3.3.1 of this Model Report. Omitting for simplicity the . subscript, ML is obtained from ML = f ( Sw - Sr) . X + fSr . X (Eq. 6-5) where X = the mass fraction of the tracer in the mobile fraction of the aqueous phase [M/M] X = the mass fraction of the tracer in the immobile fraction of the aqueous phase [M/M] Sw = the water saturation [L3/L3] Sr = the immobile water saturation (can be set equal to the irreducible) [L3/L3] f = the porosity (matrix or fracture) [L3/L3] . = the water density [M L-3]. Equation 6-5 reflects the fact that solute concentrations are different in the mobile and immobile water fractions. Because water is very strongly bound (in electric double layers) to the PFM grain surface, Brownian motion is limited and solubility in the immobile water is lower than in the mobile water fraction. The importance of this boundary layer has been recognized by de Marsily (1986 [100439], p. 234), who differentiates X and X, and Moridis (1999 [117241]), who used the mobile fraction of water in the analysis of diffusion experiments. Using the linear equilibrium relationship (Moridis 1999 [117241]), Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 65 November 2003 X = X Ki (Eq. 6-6) where Ki is a dimensionless mass transfer coefficient, for solutes 1 =Ki >0. Because of their double layers and their relatively large size (compared to solutes), colloids are expected to concentrate in the mobile water fraction and to be less abundant in the immobile water fraction. Thus, a good approximation for colloids is Ki < 1, but there is no supporting information on the subject. Substitution into Equation 6-5 then leads to ML = fh . X, where h = Sw - Sr + Ki Sr (Eq. 6-7) 6.2.2.2 Implications for Transport in the UZ For rocks with high irreducible water saturations at the Yucca Mountain UZ, Ki<1 can lead to smaller h values, indicating lower accumulation in the liquid phase. Under conditions of steady-state release, this can result in faster breakthroughs and, consequently, shorter travel times to the water table. 6.2.3 Sorption Terms The discussion in this section is limited to sorption onto the matrix and fractures. Sorption onto pseudocolloids will be discussed in Section 6.2.4. 6.2.3.1 Equilibrium Physical Sorption Omitting again for simplicity the . subscript, the mass of a solute sorbed onto the PFM grains and following a linear equilibrium isotherm is given by MAg = ( 1 - f) .s F (Eq. 6-8) where .s = the rock density [ML - 3] ; F = Fp+Fc, the total sorbed mass of solute per unit mass of the PFM, [M/M] ; Fp = the physically sorbed mass of solute per unit mass of the PFM [M/M] ; Fc = the chemically sorbed mass of solute per unit mass of the PFM [M/M] ; Following the concepts of de Marsily (1986 [100439], p. 234) and considering that sorption onto the soil grains occurs as the dissolved species diffuses through the immobile water fraction (Moridis 1999 [117241]), the equilibrium physical sorption is described by the equation Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 66 November 2003 X K K X K K X K K X K K F i i i F i d p ) ( . + . , ) ( ) . ( , ) ( . ........... = n o i t p r o s E A L m u i r b i l i u q e r i u m g n a L r o f 1 n o i t p r o s E F m u i r b i l i u q e h c i l d n u e r F r o f n o i t p r o s E L m u i r b i l i u q e r a e n i l r o f ß 2 1 (Eq. 6-9) where Kd [M -1L3], KF [M -ßL3ß], ß, K1[M -1L3], and K2 [M -1L3] are sorption parameters specific to each solute and rock type. Of particular interest is the parameter Kd, called the distribution coefficient, which is the constant slope of the linear equilibrium adsorption isotherm of a solute in relation to the medium. The sorption of the various radionuclides in this Model Report follows a linear equilibrium isotherm (See Section 6.1.5.2). 6.2.3.2 Kinetic Physical Sorption If a kinetic isotherm is followed, then sorption is described by equations given by F X K K X K K k F X K K k F X K K k t d F d p i i L p i F F p p i d p , ) ( ... - . + . ... , ) ( ] - ) . ( [ , ) ( ) d - . ( ........... = l n o i t p r o s P K A L c i t e n i k r i u m g n a L r o f 1 n o i t p r o s P K F c i t e n i k h c i l d n u e r F r o f n o i t p r o s P K L c i t e n i k r a e n i l r o f ß 2 1 (Eq. 6-10) where , ) ( ; ) ( ... = dp n o i t p r o s P I L l a c i s y h p e l b i s r e v e r r i r a e n i l r o f 0 n o i t p r o s P K L l a c i s y h p c i t e n i k r a e n i l r o f 1 (Eq. 6-11) and kl, kF, and kL are the kinetic constants for linear, Freundlich and Langmuir sorption, respectively [T -1] (de Marsily (1986 [100439]). For dp=0, the linear kinetic expression in Equation 6-9 can also be used to describe the chemical process of salt precipitation. 6.2.3.3 Kinetic Chemical Sorption The first-order reversible chemical sorption is represented by the linear kinetic chemical (LKC) model c c i c c F k X K k dt dF - + - = . (Eq. 6-12) Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 67 November 2003 where kc + [M - 1L3T - 1] and kc - [ T - 1] are the forward and backward kinetic constants, respectively. Note that Equation 6-11 can be used in conjunction with the physical sorption equations to describe combined sorption (Cameron and Klute 1977 [117172]), e.g., physical and chemical sorption. Combined sorption accounts for the different rates at which a species is sorbed onto different PFM constituents. Thus, sorption onto organic components may be instantaneous (LE), while sorption onto mineral surfaces may be much slower and kinetically controlled (Cameron and Klute 1977 [117172]). 6.2.3.4 Implications for Transport in the UZ Sorption is the main mechanism of radionuclide mass removal from the transporting liquid phase. Nonsorbing radionuclides (such as 3H or 99Tc) will not be retarded. Strongly sorbing radionuclides (such as Np, Pu, U, Th, Am) will exhibit much longer arrival times to the water table. The transfer coefficient Ki is obviously quite important, because a Ki<1 reduces the radionuclide sorbed onto the particle surfaces. Thus, the larger Ki is, the more the radionuclide is retarded. Although this may not be significant in the case of strong sorbers, it may be important in less-strongly sorbing tracers. Also of interest is sorption onto the fracture surfaces, as opposed to the volume-base sorption in the matrix. Currently, no information exists on the subject. The conventional approach in fractures with f=1 results in zero sorption. It is possible to obtain an estimate of surface sorption by (a) assuming a reasonable grain size for a matrix, (b) computing the internal surface area (available for sorption) per unit volume of the matrix, and (c) computing the surface area of a corresponding fracture for the same grain size. This approach usually produces very small surface Kd values. 6.2.4 Colloid Filtration Terms 6.2.4.1 Equations of Colloid Filtration Colloidal particles moving through porous media are subject to filtration, the mechanisms of which have been the subject of several investigations (e.g., Herzig et al. 1970 [117519]). The mass of filtered colloids is then given by . . . s . = , , c F M (Eq. 6-13) where .c,. is the density of the colloidal particles of colloid . [ML - 3] and s. is the filtered concentration of the colloid expressed as volume of colloids per volume of the porous medium. When colloid deposition is a relatively fast process compared to the water velocity, it is possible to describe colloid filtration as a linear equilibrium process (James and Chrysikopoulos 1999 [109517]). Omitting the . subscript, linear equilibrium filtration is then described by s = KsKi . X (Eq. 6-14) where Ks is a distribution coefficient [M - 1L3] . Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 68 November 2003 Colloid filtration is more accurately described by a linear kinetic model (Çorapçioglu et al. 1987 [117300], pp. 269–342), which can take the following form: s . - . = s d - . . = s - + s X X K K dt d p i ) ( (Eq. 6-15) where . [ T - 1] is a kinetic coefficient, and .+ and .- [ T - 1] are the kinetic forward and reverse colloid deposition rates (clogging and declogging coefficients), respectively, which are specific to each colloid and rock type. The term .- is commonly assumed to be zero (Bowen and Epstein 1979 [117219]), but there is insufficient evidence to support this. As will be seen in Section 6.18 of the present Model Report, colloid transport in the UZ is very sensitive to this parameter. The parameter dp is analogous to that for sorption in Equations 6-10 and 6-11, and describes the reversibility of filtration. From de Marsily (1986 [100439], p. 273) and Ibaraki and Sudicky (1995 [109297], p. 2,948), the following expression for the .+ coefficient can be derived: G u f e = .+ (Eq. 6-16) where e is the filter coefficient of the porous medium [ L - 1] , f is a velocity modification factor, u is the Darcy velocity [L T - 1] , and G is a dynamic blocking function that describes the variation of the PFM porosity and specific surface with s (James and Chrysikopoulos 1999 [109517]). The factor f (1=f =1.5) accounts for the velocity of the colloidal particle flow being larger than that of water (Ibaraki and Sudicky 1995 [109297], p. 2,948). This results from the relatively large size of the colloids, which tends to concentrate them in the middle of the pores where the water velocity is larger than the bulk average velocity. The factor f tends to increase with decreasing ionic strength, but cannot exceed 1.5 because colloids cannot move faster than the maximum water velocity, which occurs at the middle of the pores and is equal to 1.5 the average pore velocity (Ibaraki and Sudicky 1995 [109297], p. 2,948). For deep filtration (i.e., in the case of very dilute colloidal suspensions), there is no interaction among the colloidal particles and no effects on the medium porosity and permeability, i.e., f is constant, and G=1. Note that it is possible for EOS9nT to have combined filtration, in which two different types of filtration (e.g., equilibrium and kinetic, or two kinetic filtrations with different .+ and .-) occur simultaneously. 6.2.4.2 Implications for Transport in the UZ Filtration is one of the main mechanisms of radioactive colloid removal from the transporting liquid phase, the others being chemical destabilization (e.g., because of pH changes) and flocculation. Three types of filtration mechanisms may affect the transport of colloids through the PFM: surface filtration, straining filtration and physical-chemical filtration. Surface filtration occurs when particles are larger than the pores, in which case a filter cake is formed. Straining filtration is determined by the ratio Rd = dg/dp, where dg is the diameter of the grains of the porous medium and dp is the suspended particle diameter. Based on the experimental data of Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 69 November 2003 Sakthivadivel (1969 [134556]), Rd =10 leads to cake filtration, 10 < Rd =20 corresponds to substantial straining filtration (permeability reductions by a factor of 7–15 and particles occupying 0.3 f), and Rd > 20 results in limited straining (only 2–5 % of f occupied by particles and permeability reductions by 10–50%). Herzig et al. (1970 [117519], p. 15) indicated that little straining was expected when Rd > 12, and calculated that when Rd = 50, only 0.053 % of f would be occupied by particles. A detailed discussion of the physical-chemical filtration of colloids can be found in Çorapçioglu et al. (1987 [117300], pp. 269–342). The physical-chemical colloidal filtration by the porous/fractured medium incorporates three mechanisms: (a) contact with the pore walls, (b) colloid fixation onto the walls, and (c) release of previously fixed colloids (Herzig et al. 1970 [117519]; Çorapçioglu et al. 1987 [117300]). Contact with the pore walls and colloidal capture can be the result of sedimentation (caused by a density differential between the colloidal particle and the carrier liquid), inertia (deviation of colloidal trajectories from the liquid streamlines because of their mass), hydrodynamic effects (caused by a variation in the velocity field of the liquid), direct interception (caused by collisions with the pore walls at convergent areas) and diffusion (Brownian motion causing colloids to move toward pore walls or dead-end pores). Fixation on the pore walls occurs at retention sites that include edges between two convex surfaces, pore throats smaller than the colloidal size, and dead-end pores or regions of near-zero liquid velocity. Fixation is caused by retentive forces, which include axial pressure of the fluid at constriction sites, friction forces, Van der Waals forces, electrical forces and chemical forces (Herzig et al. 1970 [117519], pp. 4, 11–17). Finally, remobilization of colloidal particles may be caused by a number of factors, including collision between a loosely held colloid with a moving particle, an increase in pressure as colloids constrict flow, and a change in external conditions. All three mechanisms are expected to be present in the Yucca Mountain UZ. The EOS9nT simulations can account for surface filtration by using a particle size versus pore size criterion, and by not allowing colloidal entry into media that do not meet this criterion. Straining filtration (pore-size exclusion) is described by using appropriate colloid accessibility factors (see Section 6.18), and physical-chemical filtration is represented by using appropriate parameters in Equations 6-14 or 6-15. Note that it is not possible to account for cake filtration or for the effects of filtration on permeability and porosity, as this would violate the linearity in the models. However, these scenarios would be unlikely in the Yucca Mountain UZ because (a) natural pseudocolloids (such as clays) under natural conditions occur in small concentrations, (b) it is expected that, owing to adverse chemical conditions (e.g., pH, ionic strength) in the immediate vicinity of the repository, true colloids will be released at low concentrations for a long time, and (c) the expected concentrations of the waste form colloids (BSC 2003 [161620], Section 6.3.2) are insufficient to support cake filtration. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 70 November 2003 6.2.5 Colloid-Assisted Transport Terms 6.2.5.1 Equations of Colloid-Assisted Transport The mass of a tracer . sorbed onto pseudocolloids is described by X M j j j j N j c A , . = . , ) . + s . ( = . 1 c F (Eq. 6-17) where F.,j denotes the sorbed mass of solute . per unit mass of the pseudocolloid j [M/M] and Nc is the total number of pseudocolloid species (Moridis et al. 1999 [123093]). The first term in the sum inside the parenthesis of Equation 6-17 describes the filtered (deposited) colloid concentration, and the second describes the concentration of the suspended colloids in the liquid phase. F.,j is computed from any combination of Equations 6-9—6-12, with the appropriate sorption parameters corresponding to each colloid. Note that Equation 6-17 applies to Class III colloids only. 6.2.5.2 Implications for Transport in the UZ The formulation of Equation 6-17 allows consideration of the whole size spectrum of a particular colloid, as each size would have different transport behavior (see Equation 6-16) and different sorption properties (resulting from different surface area). Of particular interest in UZ radioactive transport models is the potential of colloid-assisted transport to significantly enhance the migration of tracers whose normally strong sorbing behavior would confine them to the vicinity of the release point. Sorption of such a radionuclide (e.g., Pu) onto pseudocolloids renders the whole colloid radioactive. The transport of such a colloid is no longer dictated by the strong Pu sorption behavior, but by the kinetics of colloid filtration. Coupled with the fact that colloids can move at velocities that can be up to 50% higher than the Darcy velocity (and especially so in fractures, where most of the UZ flow occurs; see Equation 6-16), this can result in travel times to the water table orders of magnitude shorter than the ones predicted for the corresponding sorbing solute. Note that the higher colloid velocities are not a function of the water flow, but depend entirely on the colloid size (see discussion in Section 6.2.4.1). 6.2.6 Flux Terms 6.2.6.1 Equations of the Flux Terms The flux term has contributions from advective, diffusive, and dispersive transport processes and is given by F. = F.X. - . D. . X. - Fs, . (Eq. 6-18) where Fs, . is the flux due to surface diffusion and D. is the dispersion tensor of tracer ., a second order symmetric tensor with a principal axis aligned with the Darcy flow vector. Omitting the . subscript, D. is described by the equations Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 71 November 2003 u u I D , + =D u D D T T L - 2 (Eq. 6-19) DL = f ( Sw - Sr) tD0 + f Sr tD0 + aL u , (Eq. 6-20) DT = f ( Sw - Sr) tD0 + f Sr tD0 + aT u , (Eq. 6-21) where I = the unit vector t = the tortuosity coefficient of the pore paths [L/L] D0 = the molecular diffusion coefficient of tracer i in water [ L2T - 1] aL, aT = longitudinal and transverse dispersivities, respectively [L] u = the Darcy velocity vector [ LT - 1] Equation 6-18 accounts for surface diffusion, which can be responsible for significant transport in strongly sorbing media (Moridis 1999 [117241]; Cook 1989 [117314]). The surface diffusion flux is given by Jahnke and Radke (1987 [117398]) as Fs = ( 1 - f) .stsDs.Fp, (Eq. 6-22) where ts is the tortuosity coefficient of the surface path [L0], Ds is the surface diffusion coefficient [L2T -1], and Fp is the physical diffusion computed from equations (19) or (10). There is theoretical justification for the relationship t = t 3 2 s (Cook 1989 [117314], p. 10). 6.2.6.2 Application to Colloid Fluxes Equations 6-17 through 6-20 apply to solutes, but need the following modifications to render them suitable to colloidal transport. More specifically: 1. Fs = 0 because surface diffusion does not occur in colloids. 2. The flux Fs and the Darcy velocities u are multiplied by the factor f (see Section 6.2.4.1). 3. The dispersivities aL and aT are generally different from those for solutes Ibaraki and Sudicky (1995 [109297]) and may be a function of the colloidal particle size. 4. The term D0 is the colloidal diffusion coefficient in water [L2T-1] and is described by the Stokes-Einstein equation, according to Bird et al. (1960 [103524], p. 514), as D0 = k T 3p µwdp , (Eq. 6-23) Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 72 November 2003 where k is the Boltzmann constant (1.38× 10 - 23JK - 1 in SI units), T is the absolute water temperature [K] , µw is the dynamic viscosity of water [ML - 1T - 1] , and dp is the colloid diameter [L]. 5. The fluxes in Equation 6-18 are multiplied by the colloid accessibility factors fc (0 =fc =1) at the interface of different media. The fc factor describes the portion of the colloidal concentration in a medium that is allowed to enter an adjacent medium of different characteristics, and quantifies pore size exclusion (straining). In the treatment of the general 3-D dispersion tensor, velocities are averaged by using the projected area weighting method (Wu et al. 1996 [100649], p. 23), in which a velocity component uj (j =x,y,z) of the vector u is determined by vectorial summation of the components of all local connection vectors in the same direction, weighted by the projected area in that direction. This approach allows the solution of the transport problem in irregularly shaped grids, in which the velocities normal to the interface areas are not aligned with the principal axes. 6.2.6.3 Implications for Transport in the UZ Because it is directly proportional to the water fluxes, advection in the fractures is by far the dominant mechanism of transport in the UZ of Yucca Mountain. This is further enhanced by longitudinal dispersion, molecular diffusion, and (possibly) surface diffusion (in decreasing order). Coupled with the fracture orientation and gravitational differentials, the result of the fracture and flow characteristics is a mainly downward migration of the radionuclides. Note that advection also occurs in the matrix (and is accounted for in the simulations), albeit at significantly lower rates. Lateral spreading of the contaminants can be achieved through transverse dispersion, molecular diffusion, and (possibly) surface diffusion. As sorption occurs from the liquid phase onto the solids, and fractures are by far the main conduits of water, retardation of sorbing radionuclides will be controlled by the rate of tracer movement from the fractures to the sorbing matrix. This is represented by the sum of diffusive, dispersive, and surface diffusion fluxes, and results in the lateral migration of the contaminants. Currently, field study estimates of aL and aT in the UZ are very limited. When not ignored in the numerical simulations (justified by the relative magnitude of advection), reasonable aL estimates are used, and aT is usually set to zero (especially in the fractures). In past simulations (DOE 1998 [100550], p. 3-122), an estimate of aL = 20 m was used. From the Northern Ghost Dance Fault test (LeCain et al. 2000 [144612]) aL estimates in the 0.4 m to 2.6 m range were obtained but their validity is uncertain because they are based on gas (rather than liquid) phase transport, short transport times (<200 min), and a small scale experiment (<10 m). Accurate values of D0, coupled with reasonable tortuosity coefficients (shown from laboratory experiments to be approximated by the porosity, see Section 5, this Model Report), provide a rather representative estimate of diffusive fluxes in the UZ simulations. No information exists on whether the UZ media support surface diffusion (a possibility in zeolites). Surface diffusion can be important in tracers that exhibit strong sorption (e.g., Pu). A Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 73 November 2003 larger Kd clearly indicates stronger sorption, but this does not mean immobilization of the dissolved species when the porous/fractured medium supports surface diffusion. On the contrary, the stronger the sorption (i.e., the larger the Kd), the larger the diffusion rate, with practically all of it attributable to the surface process (Moridis 1999 [117241], pp. 1,735–1,736). 6.2.7 Radioactive Decay 6.2.7.1 Equations of Radioactive Decay When a tracer . undergoes radioactive decay, the rate of mass change is described by the first-order decay law dM. dt = - .. M. , where .. = ln2 ( T1/2) . (Eq. 6-24) and (T1/2). is the half life of tracer .. Substitution of Equations 6-18 and 6-24 into Equation 6-2 yields (Moridis et al. 1999 [123093] Eq. 32) [ ] . . . . . . . . . + - . . - = . + .. . .. . m n s nm nm n n n q X X F A V M dt dM ) ( F D 1 ) ( , (Eq. 6-25) 6.2.7.2 Implications for Transport in the UZ The decay of radioactive substance is completely predictable and well documented. Thus, the decay of radioactive substances in UZ simulations can be computed very reliably. 6.2.8 Daughter Products of Radioactive Decay 6.2.8.1 Transport Equations of Daughters If a radioactive tracer . is a daughter product of the decay of tracer j, then the mass accumulation terms are adjusted (de Marsily 1986 [100439], pp. 265–266) to yield the equation of transport of the daughter products as ) ( + - . . - [ = ) ( . - ) ( . + ... ... q X X F A V M m M t d M d . , . . . . . n k s i m n m n m n n j r j n n 1 . F D (Eq. 6-26) where mr =W./Wj, and W. and Wj are the molecular weights of the daughter and parent species. Equation 6-26 applies to radioactive solutes or radioactive true colloids. For daughters following an isotherm other than LE, Equations 6-9 and 6-10 need to account for the generation of daughter mass from the decay of the sorbed parent, and become Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 74 November 2003 .F. .t + ..F. - .. - 1mr ..F. - 1 = kA . X. - kB ( F. + mr ..F. - 1) (Eq. 6-27) where F.-1 is the sorbed mass of the parent, . . . = + sorption LKC for sorption LKP/LIP for i c i d A K k K K k k l kB = . . .kpdp for LKP/LIP sorption, k - c for LKC sorption, (Eq. 6-28) and .. is the fraction of the mass of the decayed sorbed parent that remains sorbed as a daughter (0 = .. =1). The term .. is introduced to account for the different sorption behavior of parents and daughters, and the fact that daughters can be ejected from grain surfaces due to recoil (e.g., the ejection of 234Th from grain surfaces during the alpha decay of 238U) (Faure 1977 [122805], pp. 288–289). 6.2.8.2 Implications for Transport in the UZ If decay results in radioactive daughters, UZ simulations must compute the total radioactivity distribution, i.e., the sum of the concentrations of all the members of the radioactive chain. This is especially true if the daughters have long half lives. Intermediate chain products with short half lives relative to the simulation periods can be ignored with impunity. Although the EOS9nT can theoretically obtain the transport scenario of any number of daughters in the radioactive chain, this is especially true for daughters with short half lives. With longer half-lives (such as for 237Np, 239Pu, etc.), machine accuracy considerations and roundoff errors (with the daughter concentrations becoming increasingly smaller moving down the radioactive chain) limit their number to a maximum of four or five with comparable half lives. The . factor is a function of the type of decay, as well as of the chemical form of the sorbed cation. In alpha decay (e.g., 237Np, 239Pu), .=0. There is no information on the behavior of . for other types of decay. The . factor can have significant implications for the transport behavior of daughters if (a) the large sorbed masses of strongly sorbing parents are ejected back into the aqueous phase after decay, (b) the daughter is a much weaker sorber, and (c) its sorption is kinetically controlled. Such a possibility could arise in the kinetic sorption of radionuclides onto pseudocolloids (BSC 2001 [160828], Section 6). For equilibrium isotherms, this is not an issue. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 75 November 2003 6.3 THE NUMERICAL MODELS In mathematical simulations, the conventional meaning of the term model is the mathematical description of the physics governing the behavior of the simulated system, whereas the term code is the implementation of the mathematical model in a computer language. Because model and code are intertwined, these terms have historically been used interchangeably. In the context of T2R3D and EOS9nT applications in the present Model Report, when the word model is used, it is preceded by the qualifying words numerical or semianalytical. 6.3.1 The T2R3D Code The code T2R3D V1.4 (LBNL 1999 [146654]; Wu et al. 1996 [100649]) simulates the flow (saturated and/or unsaturated) and coupled transport of a single radioactive solute tracer in complex subsurface systems involving porous and/or fractured media. The transport equations account for advection, molecular diffusion, hydrodynamic dispersion, and linear equilibrium sorption. T2R3D can simulate problems involving the coupled (a) flow of subsurface liquids, (b) transport of a solute, and (c) the transport of heat in complex heterogeneous subsurface systems. When a flow system has reached a steady state, T2R3D allows the option of solving only the solute transport equation. 6.3.2 The EOS9nT Code The code used for the 3-D site-scale transport simulations in the Model Report is TOUGH2 V1.11MEOS9NTV1.0 (LBNL 1999 [113943]; Moridis et al. 1999 [123093]; 2003 [161902]), which is a member of the TOUGH2 family of codes (Pruess 1991 [100413]). It can simulate flow and transport of an arbitrary number n of nonvolatile tracers (solutes and/or colloids) in the subsurface. EOS9nT first solves the Richards equation, which describes saturated or unsaturated water flow in subsurface formations, and obtains the flow regime. The set of n linearly independent transport equations (corresponding to the n solutes/colloids) are then solved sequentially. The n tracer transport equations account for (a) advection, (b) molecular diffusion, (c) hydrodynamic dispersion (with full 3-D tensorial representation), (d) kinetic or equilibrium physical and chemical sorption (linear, Langmuir, Freundlich, or combined), (e) first-order linear chemical reaction, (e) radioactive decay, (f) colloid filtration (equilibrium, kinetic or combined), and (g) colloid assisted solute transport. A total of n -1 daughter products of radioactive decay (or of a linear, first-order reaction chain) can be tracked. EOS9nT includes two types of Laplace transform formulations of the tracer equations, in addition to conventional timestepping. The Laplace transform is applicable to steady-state flow fields and allows a practically unlimited time-step size and more accurate solution (as numerical diffusion is significantly reduced). Additional information on the EOS9nT numerical model can be found in BSC (2001 [161340], Attachment I). For the simulation of isothermal transport of a tracer that follows a linear equilibrium isotherm, T2R3D and EOS9nT have been shown to produce identical results (Section 7.1, this Model Report; DTN: LB03093RADTRNS.001 [166225]). Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 76 November 2003 6.3.3 Other Computational Tools In addition to T2R3D and EOS9nT, the following computational tools were used: 1. The software codes TOUGH2 V1.4 Module EOS9 (LBNL 2000 [146496]) and TOUGH2 V1.6 Module EOS9 (LBNL 2002 [160242]; LBNL 2003 [161491]), which solve the Richards equation to simulate flow (saturated and/or unsaturated) in complex subsurface systems (involving porous and/or fractured media). These codes were used to obtain the flow fields for the various climatic scenarios discussed in BSC (2003 [163045], Section 6). 2. The software code PHREEQC V2.3 (BSC 2001 [155323]) to perform the computations for the estimations of the Kd (Section 6.4). 3. The XtractG.f90 V1.0 routine (LBNL 2003 [162786]), which was used for the EOS9nT runs for the following applications: (a) To modify the grid system for use in the EOS9nT runs. The modifications involved (1) determining the grid elements corresponding to the repository, (2) moving them to the end of the element file, where (3) the upper and lower boundary elements are also moved. This was necessitated by the fact that the scenario of continuous contaminant release was investigated in this Model Report. Thus the repository elements represent internal boundary points. After moving these elements to the bottom of the element file, the volume of the first element is set to zero or a negative number. Then, these elements are considered in the determination of fluxes but are not included in mass balance computations. Additionally, XtractG.f90 can be used to render inactive all the elements above a given height (e.g., the highest repository element) which do not contribute to transport, thus significantly reducing the size of the problem and making possible the solution within the array sizes of EOS9nT. (b) To obtain initial condition files (INCON) using the SAVE files from the EOS9 runs. (c) To create steady state velocity profiles using the flow field output files from the EOS9 runs. (d) To extract concentrations at desired grid elements within the domain. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 77 November 2003 6.4 FEATURES, EVENTS, AND PROCESSES (FEPS) 6.4.1 FEPs Addressed in this Model Report The following table of Features, Events, and Processes (FEPs) were taken from the LA FEP List (DTN: MO0307SEPFEPS4.000 [164527]). The LA FEP List is a revision to the previous project FEP list (Freeze et al. 2001 [154365]) used to develop the list of included FEPs in the Technical Work Plan for: Performance Assessment Unsaturated Zone (BSC 2002 [160819], Table 2-6). The selected FEPs are those taken from the LA FEP List that are associated with the subject matter of this report, regardless of the anticipated status for exclusion or inclusion in TSPA-LA as represented in BSC (2002 [160819]). The results of this model (or analysis) are part of the basis for the treatment of FEPs as discussed in the Total System Performance Assessment-License Application Methods and Approach (BSC 2002 [160146], Section 3.2.2). The cross-reference for each FEP to the relevant section (or sections) of this report is also given below. Table 6.4-1. FEPs Addressed in This Model Report LA FEP Number FEP Name Section Where FEP is Addressed Summary Description 1.2.02.01.0A Fractures The effects of fractures on radionuclide transport are addressed through the use of fracturecontinuum flow fields (output from the UZ Flow Model) that are part of the input data used for transport simulations (Section 4.1). Simulations showing the effects of fractures are presented in Section 6.8.1.2, and in Attachments IV and V. Groundwater flow in the Yucca Mountain region and transport of any released radionuclides may take place along fractures. The rate of flow and the extent of transport in fractures are influenced by characteristics such as orientation, aperture, asperity, fracture length, connectivity, and the nature of any linings or infills. 1.2.02.02.0A Faults The effects of faults are addressed through the use of dual-continuum flow fields that are part of the input data used for transport simulations (Section 4.1). Simulations showing the effects of faults are presented in Sections 6.8.1.2 and 6.20.2. Numerous faults of various sizes have been noted in the Yucca Mountain Region and in the repository area in specific. Faults may represent an alteration of the rock permeability and continuity of the rock mass, alteration or shortcircuiting of the flow paths and flow distributions close to the repository, and represent unexpected pathways through the repository. 1.3.01.00.0A Climate change Climate change is addressed through the use of different dual-continuum flow fields (output from the UZ Flow Model), for different future climates, that are part of the input data used for transport simulations (Section 4.1). Simulations for different climates are presented in Section 6.8.1 and 6.8.2. The effect of varying infiltration resulting from climate change is shown in Figures 6.8-1, 6.8-26, and 6.8-27. Additional discussion is presented in Sections 6.7.1, and 6.8 through 6.14. Simulations are presented in this Model Report to evaluate the effects of climate change. Climate change may affect the longterm performance of the repository. This includes the effects of long-term change in global climate (e.g., glacial/interglacial cycles) and shorter-term change in regional and local climate. Climate is typically characterized by temporal variations in precipitation and temperature. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 78 November 2003 Table 6.4-1. FEPs Addressed in this Model Report (Continued) LA FEP Number FEP Name Section Where FEP is Addressed Summary Description 1.4.01.01.0A Climate modification increases recharge Steady state dual-continuum flow fields (output from the UZ Flow Model) that are part of the input data used for transport simulations (Section 4.1) account for changes in recharge resulting from climate changes. Simulations for different climates are presented in Section 6.8.1.2. Additional discussion is presented in Sections 6.7.1, and 6.8 through 6.14. Climate modification causes an increase in recharge in the Yucca Mountain region. Increased recharge might lead to increased flux through the repository, perched water, or water table rise. 2.1.08.01.0A Water influx at the repository Water influx at the repository under present and projected future climates is addressed through the use of dual-continuum flow fields (output from the UZ Flow Model) that are part of the input data used for transport simulations (Section 4.1). Water influx at the repository may result in seepage into drifts (The process model for seepage is presented in another Model Report, BSC 2003 [163226]). Seepage may result in radionuclide transport. Sections 6.8 through 6.14 present the transport simulations for radionuclides instantly released from drifts to fractures. Additional discussion is presented in Sections 6.7.1, 6.8.2, 6.9.2, 6.11.1 and 6.11.2. An increase in the unsaturated water flux at the repository affects thermal, hydrologic, chemical, and mechanical behavior of the system. Increases in flux could result from climate change, but the cause of the increase is not an essential part of the FEP. 2.2.03.01.0A Stratigraphy Stratigraphy is addressed in the conceptual model of transport presented in Section 6.6. In 3-D simulations, it is addressed in the material properties included in input data for simulations by using the grid developed in BSC 2003 [160109] and used as direct input in Section 4.1. Stratigraphy affects radionuclide transport through its effects on flow (predominantly matrix flow or fracture flow depending upon the hydrogeologic unit) and through the differing sorptive properties of different hydrogeologic units. The effect of sorption on transport is presented in Sections 6.9.1.6 and 6.10.1.4. Additional discussion is presented in Sections 6.1 and 6.1.1 . Stratigraphic information is necessary information for the performance assessment. This information should include identification of the relevant rock units, soils and alluvium, and their thickness, lateral extents, and relationships to each other. Major discontinuities should be identified. 2.2.03.02.0A Rock properties of host rock and other units Rock Properties are addressed in the conceptual model of transport presented in Section 6.6. In 3-D simulations, rock properties are addressed in the material properties included in input data for simulations by using the grid developed in BSC 2003 [160109] and used as direct input in Section 4.1. Rock physical properties affect radionuclide transport through their effects on flow as shown in the transport simulations shown in Section 6.8 and Attachments IV through VI. The effect of sorption on transport is presented in Sections 6.9.1.6 and 6.10.1.4. Physical properties such as porosity and permeability of the relevant rock units, soils, and alluvium are necessary for the performance assessment. Possible heterogeneities in these properties should be considered. Questions concerning events and processes that may cause these physical properties to change over time are considered in other FEPs. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 79 November 2003 Table 6.4-1. FEPs Addressed in this Model Report (Continued) LA FEP Number FEP Name Section Where FEP is Addressed Summary Description 2.2.07.02.0A Unsaturated groundwater flow in the geosphere Unsaturated groundwater flow is addressed through the use of dual-continuum flow fields (output from the UZ Flow Model) that are part of the input data used for transport simulations (Section 4.1). The effect of unsaturated flow on advective transport is evident in the simulations presented in Section 6.8 and Attachments IV through VI. Groundwater flow occurs in unsaturated rocks in most locations above the water table at Yucca Mountain, including at the location of the repository. See related FEPs for discussions of specific issues related to unsaturated flow. 2.2.07.04.0A Focusing of unsaturated flow (fingers, weeps) Steady state dual-continuum flow fields (output from the UZ Flow Model) that are part of the input data used for transport simulations (Section 4.1) account for focusing of unsaturated flow at the gridblock scale. Subgridblock scale flow focusing in the fracture continuum is accounted for through the active fracture model (Section 7.7.1). This is incorporated into the RTM as input data through the flow fields used for transport simulations (Section 4.1). Unsaturated flow can differentiate into zones of greater and lower saturation (fingers) that may persist as preferential flow paths. Heterogeneities in rock properties, including fractures and faults, may contribute to focusing. Focused flow may become locally saturated. 2.2.07.06.0B Long-term release of radionuclides from the repository The effects of long-term release of radionuclides are addressed in Sections 6.15 to 6.18. Simulation results presented in Attachments V through VII show the effects of long-term release. The release of radionuclides from the repository may occur over a long period of time, as a result of the timing and magnitude of the waste packages and drip shield failures, waste form degradation, and radionuclide transport through the invert. 2.2.07.07.0A Perched water develops The effects of perched water are addressed through the use of dual-continuum flow fields that are part of the input data used for transport simulations (Section 4.1). The transport simulations presented in this Model Report include the effect of perched water. Additional discussion is presented in Sections 6.1.2, 6.1.2.7, 6.7.2, 6.18.5.2, and 8.2.1 (conclusion 7). Zones of perched water may develop above the water table. If these zones occur above the repository, they may affect UZ flow between the surface and the waste packages. If they develop below the repository, for example at the base of the Topopah Spring welded unit, they may affect flow pathways and radionuclide transport between the waste packages and the saturated zone. 2.2.07.08.0A Fracture flow in the UZ The effects of fractures on radionuclide transport are addressed through the use of fracture-continuum flow fields (output from the UZ Flow Model) that are part of the input data used for transport simulations (Section 4.1). Simulations showing the effects of fractures are presented in Section 6.8.1.2, and in Attachments IV and V.A discussion is presented in Sections 6.1.2.1 and 6.1.5.3. Fractures or other analogous channels act as conduits for fluids to move into the subsurface to interact with the repository and as conduits for fluids to leave the vicinity of the repository and be conducted to the SZ. Water may flow through only a portion of the fracture network, including flow through a restricted portion of a given fracture plane. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 80 November 2003 Table 6.4-1. FEPs Addressed in This Model Report (Continued) LA FEP Number FEP Name Section Where FEP is Addressed Summary Description 2.2.07.09.0A Matrix imbibition in the UZ Steady state dual-continuum flow fields (output from the UZ Flow Model) that are part of the input data used for transport simulations (Section 4.1) account for matrix imbibition by changes between one climate stage and the next. Figures showing radionuclide concentrations in the matrix in Section 6.8.1.2 and Attachments IV through VII include the effects of both matrix flow and imbibition from fractures into the matrix. An alternative treatment of matrix imbibition is discussed in Section 6.19. Water flowing in fractures or other channels in the unsaturated zone is imbibed into the surrounding rock matrix. This may occur during steady flow, episodic flow, or into matrix pores that have been dried out during the thermal period. 2.2.07.15.0B Advection and dispersion in the UZ Advection is more important than dispersion for radionuclide transport. The effects of advection are discussed in the transport model in Sections 6.1.2.1, 6.2.6 and 6.8.1.3. Advective transport is calculated using the fracture-continuum flow fields (output from the UZ Flow Model) that are part of the input data used for transport simulations (Section 4.1). All transport simulations presented in this Model Report show the effect of advection. The effects of dispersion are discussed in the transport model in Section 6.1.2.2, 6.2.6. and 6.18.4. Advection and dispersion processes may affect contaminant transport in the UZ. 2.2.08.01.0B Chemical characteristics of groundwater in the UZ This FEP is included; the TSPA Disposition is provided in Section 6.4.2. Attachment I, Section I.4 also discusses the chemical characteristics of UZ groundwater. Chemistry and other characteristics of groundwater in the unsaturated zone may affect groundwater flow and radionuclide transport of dissolved and colloidal species. Groundwater chemistry and other characteristics, including temperature, pH, Eh, ionic strength, and major ionic concentrations, may vary spatially throughout the system as a result of different rock mineralogy. 2.2.08.05.0A Diffusion in the UZ Diffusion in the UZ is addressed in the simulation codes described in Sections 6.3.1, 6.3.2, and 6.3.3. Molecular diffusion processes may affect radionuclide transport in the UZ. This includes osmotic processes in response to chemical gradients. 2.2.08.06.0B Complexation in the UZ This FEP is included; the TSPA Disposition is provided in Section 6.4.2. Complexation on mobile complexing agents such as humic and fulvic acids is treated as part of colloid transport in Section 6.1.3 and 6.18. Complexation on mineral surfaces is treated as part of sorption in Attachment I, Section I.7. Complexing agents such as humic and fulvic acids present in natural groundwaters could affect radionuclide transport in the UZ. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 81 November 2003 Table 6.4-1. FEPs Addressed in This Model Report (Continued) LA FEP Number FEP Name Section Where FEP is Addressed Summary Description 2.2.08.08.0B Matrix diffusion in the UZ Matrix diffusion is addressed in Section 6.1.2.4. Figures showing radionuclide concentrations in the matrix in Section 6.8.1.2 and Attachments IV and V include the effects of both diffusion from fractures into the matrix. Uncertainty of radionuclide transport is addressed in Sections 6.8.3.2, 6.9.3.1 and 6.10.2. Matrix diffusion is the process by which radionuclides and other species transported in the UZ by advective flow in fractures or other pathways move into the matrix of the porous rock by diffusion. Matrix diffusion can be a very efficient retarding mechanism, especially for strongly sorbed radionuclides due to the increase in rock surface accessible to sorption. 2.2.08.09.0B Sorption in the UZ This FEP is included; the TSPA Disposition is provided in Section 6.4.2. The sorption model and the distributions of Kd values are presented in 6.1 and Table 6.5-1. Data supporting the Kd distributions are presented in Attachment I. Colloid transport is addressed in Section 6.1.3 and 6.2.3. The model for colloid filtration in the UZ and simulation results are presented in Section 6.18. Sorption of dissolved and colloidal radionuclides in the UZ can occur on the surfaces of both fractures and matrix in rock or soil along the transport path. Sorption may be reversible or irreversible, and it may occur as a linear or nonlinear process. Sorption kinetics and the availability of sites for sorption should be considered. Sorption is a function of the radioelement type, mineral type, and groundwater composition. 2.2.08.10.0B Colloidal transport in the UZ Colloidal transport in the UZ is addressed in Section 6.18. Mathematical model of colloid transport is addressed in Sections 6.2.4 and 6.2.5. The effect of transport of PuO2 colloidal particles is shown in Attachments VI and VII. Radionuclides may be transported in groundwater in the UZ as colloidal species. Types of colloids include true colloids, pseudo colloids, and microbial colloids. 2.2.09.01.0B Microbial activity in the UZ This FEP is included; the TSPA Disposition is provided in Section 6.4.2. The effects of Microbial activity on colloid transport and on complexing agents are addressed in Sections 6.1.3 and 6.18. The presence of representative biota in sorption experiments is addressed in Section I.5. Microbial activity in the UZ may affect radionuclide mobility in rock and soil through colloidal processes, by influencing the availability of complexing agents, or by influencing groundwater chemistry. Changes in microbial activity could be caused by the response of the soil zone to changes in climate. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 82 November 2003 Table 6.4-1. FEPs Addressed in This Model Report (Continued) LA FEP Number FEP Name Section Where FEP is Addressed Summary Description 3.1.01.01.0A Radioactive decay and ingrowth Radioactive decay and ingrowth are addressed in the simulations of transport of radionuclide chains in Sections 6.2.7, 6.2.8, 6.7.4, 6.7.6, 6.8.1.1, 6.16, and 6.17. Radioactivity is the spontaneous disintegration of an unstable atomic nucleus that results in the emission of subatomic particles. Radioactive isotopes are known as radionuclides. Radioactive decay of the fuel in the repository changes the radionuclide content in the fuel with time and generates heat. Radionuclide quantities in the system at any time are the result of the radioactive decay and the growth of daughter products as a consequence of that decay (i.e., ingrowth). Over a 10,000-year performance period, these processes will produce daughter products that need to be considered in order to adequately evaluate the release and transport of radionuclides to the accessible environment. 6.4.2 FEPs Included for TSPA 2.2.08.01.0B, Chemical Characteristics of Groundwater in the UZ FEP Description: Chemistry and other characteristics of groundwater in the unsaturated zone may affect groundwater flow and radionuclide transport of dissolved and colloidal species. Groundwater chemistry and other characteristics, including temperature, pH, Eh, ionic strength, and major ionic concentrations, may vary spatially throughout the system as a result of different rock mineralogy. Screening Decision and Regulatory Basis: Included Screening Argument: None. TSPA Disposition: The effects of groundwater chemical characteristics are included in the radionuclide sorption coefficients under ambient conditions. The sorption coefficient data on which the distributions are based are obtained in laboratory experiments in which crushed rock samples from the Yucca Mountain site are contacted with groundwaters (or simulated groundwaters) representative of the site, spiked with one or more of the elements of interest (this Model Report, Section I.5). The chemistry of pore waters and perched waters in the UZ along potential flowpaths to the accessible environment is discussed in BSC (2002 [160247]). In the UZ, two distinct water types exist in the ambient system. One is perched water and the other is pore water. Perched water is generally more dilute than pore water. The J-13 and UE p#1 waters were used in sorption experiments as end-member compositions intended to bracket the impact of water composition on sorption coefficients (see Section I.4, this Model Report). Some spatial trends in water composition through the TSw and CHn geologic units have been noted (BSC Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 83 November 2003 2002 [160247], Section 6.5.3.1). However, the uncertainty in these spatial variations (BSC 2003 [162050], Section 6.2.2.1) and the uncertainty with respect to the effects of the bounding water compositions on sorption (see Sections I.8.C, I.8.D, and I.8.I of this Model Report) have led to the treatment of natural variability in water composition as uncertainty. Sorption experiments have been carried out as a function of time, element concentration, atmospheric composition, particle size, and temperature. In some cases, the solids remaining from sorption experiments were contacted with unspiked groundwater in desorption experiments. The experimental data used to determine the sorption Kds are provided in the following DTNs: LA0305AM831341.001 [163789], LA0309AM831341.002 [165523], LA0309AM831341.003 [165524], LA0309AM831341.004 [165525], LA0309AM831341.005 [165526], LA0309AM831341.006 [165527], LA0309AM831341.007 [165528], and LA0310AM831341.001 [165865]. The sorption and desorption experiments together provide information on the equilibration rates of the forward and backward sorption reactions. For elements that sorb primarily through surface complexation reactions, the experimental data are augmented with the results of modeling calculations using PHREEQC V2.3 (BSC 2001 [155323]). The inputs for the modeling calculations include groundwater compositions, surface areas, binding constants for the elements of interest, and thermodynamic data for solution species. These modeling calculations provide a basis for interpolation and extrapolation of the experimentally derived sorption coefficient dataset. The effects of nonlinear sorption are approximated by capturing the effective Kd range (this Model Report, Section I.8). The effects of groundwater composition with respect to sorption coefficients are provided in terms of probability distributions for the sorption coefficient of each element of interest among the three major rock types (devitrified, zeolitic, and vitric) found in the UZ. The influence of expected variations in water chemistry, radionuclide concentrations, and variations in rock surface properties within one of the major rock types are incorporated into these probability distributions. These distributions are specified for each radionuclide/rock type combination (this Model Report, Section I.2 and I.4) and are sampled in the TSPA to account for the effects of natural variations in pore water chemistry and mineral surfaces on sorption. Correlations for sampling sorption coefficient probability distributions have been derived for the elements investigated (this Model Report, Section II). To derive the correlations, a rating system was first developed to rate the impact of six different variables on the sorption coefficient for a given element in each of the three major rock types. The six variables are pH, Eh, water chemistry, rock composition, rock surface area, and radionuclide concentration. Water chemistry refers to the major ion concentrations and silica. Rock composition refers to both the mineralogic composition of the rocks and the chemical composition of the minerals (e.g., zeolite compositions). The output DTN for the sorption Kds and correlations is LA0302AM831341.002. 2.2.08.06.0B, Complexation in the UZ FEP Description: Complexing agents such as humic and fulvic acids present in natural groundwaters could affect radionuclide transport in the UZ. Screening Decision and Regulatory Basis: Included Screening Argument: None. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 84 November 2003 TSPA Disposition: The effects of complexation are implicitly included in the radionuclide sorption coefficients under ambient conditions. The sorption coefficient data on which the distributions are based are obtained in laboratory experiments in which crushed rock samples from the Yucca Mountain site are contacted with groundwaters (or simulated groundwaters) representative of the site, spiked with one or more of the elements of interest (this Model Report, Section I.5). As such, the sorption experiments contain representative ligands responsible for complex formation, such as carbonates (Triay et al. 1997 [100422], p. 85). Sorption experiments have been carried out as a function of time, element concentration, atmospheric composition, particle size, and temperature. In some cases, the solids remaining from sorption experiments were contacted with unspiked groundwater in desorption experiments. The experimental data used to determine the sorption Kds are provided in the following DTNs: LA0305AM831341.001 [163789], LA0309AM831341.002 [165523], LA0309AM831341.003 [165524], LA0309AM831341.004 [165525], LA0309AM831341.005 [165526], LA0309AM831341.006 [165527], LA0309AM831341.007 [165528], and LA0310AM831341.001 [165865]. The sorption and desorption experiments together provide information on the equilibration rates of the forward and backward sorption reactions. For elements that sorb primarily through surface complexation reactions, the experimental data are augmented with the results of modeling calculations using PHREEQC V2.3 (BSC 2001 [155323]). The inputs for the modeling calculations include groundwater compositions, surface areas, binding constants for the elements of interest, and thermodynamic data for solution species. These modeling calculations provide a basis for interpolation and extrapolation of the experimentally derived sorption coefficient dataset. The effects of nonlinear sorption are approximated by capturing the effective Kd range (this Model Report, Section I.8). The effects of organics on sorption were also investigated by Triay et al. (1997 [100422], Section IV.B). Their experiments tested the effects of organic materials (DOPA (dihydroxyphenylalanine) and NAFA (Nordic Aquatic Fulvic Acid)) on the sorption of Pu and Np on tuff materials. The results of these tests showed very little effect of the organic materials for sorption of these radionuclides in tuffs. The effects of complexation with respect to sorption coefficients are provided in terms of probability distributions for the sorption coefficient of each element of interest among the three major rock types (devitrified, zeolitic, and vitric) found in the UZ. The influence of expected variations in water chemistry, radionuclide concentrations, and variations in rock surface properties within one of the major rock types are incorporated into these probability distributions. These distributions are specified for each radionuclide/rock type combination (this Model Report, Section I.2 and I.4) and are sampled in the TSPA to account for the effects of natural variations in pore water chemistry and mineral surfaces on sorption. Correlations for sampling sorption coefficient probability distributions have been derived for the elements investigated (this Model Report, Section II). To derive the correlations, a rating system was first developed to rate the impact of six different variables on the sorption coefficient for a given element in each of the three major rock types. The six variables are pH, Eh, water chemistry, rock composition, rock surface area, and radionuclide concentration. Water chemistry refers to the major ion concentrations and silica. Rock composition refers to both the mineralogic composition of the rocks and the chemical composition of the minerals (e.g., zeolite Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 85 November 2003 compositions). The output DTNs for the sorption Kds and correlations are LA0302AM831341.002 and LA0311AM831341.001. 2.2.08.09.0B, Sorption in the UZ FEP Description: Sorption of dissolved and colloidal radionuclides in the UZ can occur on the surfaces of both fractures and matrix in rock or soil along the transport path. Sorption may be reversible or irreversible, and it may occur as a linear or nonlinear process. Sorption kinetics and the availability of sites for sorption should be considered. Sorption is a function of the radioelement type, mineral type, and groundwater composition. Screening Decision and Regulatory Basis: Included Screening Argument: None. TSPA Disposition: Sorption is included in the TSPA model for mountain-scale unsaturated zone radionuclide transport as a linear equilibrium sorption (Kd) model (this Model Report, Attachment I). Sorption is only accounted for in the matrix continuum; there is no sorption modeled in the fracture continuum. Sorption characteristics of the rock minerals are assumed to be static in time. Sorption Kds have been derived for the elements Am, Cs, Np, Pa, Pu, Ra, Sr, Th, and U. Other radionuclide elements treated by TSPA (e.g. Tc) are modeled as non-sorbing. The sorption coefficient data on which the distributions are based are obtained in laboratory experiments in which crushed rock samples from the Yucca Mountain site are contacted with groundwaters (or simulated groundwaters) representative of the site, spiked with one or more of the elements of interest (this Model Report, Section I.5). Sorption experiments have been carried out as a function of time, element concentration, atmospheric composition, particle size, and temperature. In some cases, the solids remaining from sorption experiments were contacted with unspiked groundwater in desorption experiments. The experimental data used to determine the sorption Kds are provided in the following DTNs: LA0305AM831341.001 [163789], LA0309AM83341.002 [165523], LA0309AM83341.003 [165524], LA0309AM83341.004 [165525], LA0309AM83341.005 [165526], LA0309AM83341.006 [165527], LA0309AM83341.007 [165528], and LA0310AM831341.001 [165865]. The sorption and desorption experiments together provide information on the equilibration rates of the forward and backward sorption reactions. For elements that sorb primarily through surface complexation reactions, the experimental data are augmented with the results of modeling calculations using PHREEQC V2.3 (BSC 2001 [155323]). The inputs for the modeling calculations include groundwater compositions, surface areas, binding constants for the elements of interest, and thermodynamic data for solution species. These modeling calculations provide a basis for interpolation and extrapolation of the experimentally derived sorption coefficient dataset. The effects of nonlinear sorption are approximated by capturing the effective Kd range (this Model Report, Section I.8). Sorption coefficients are provided in terms of probability distributions for the sorption coefficient of each element of interest among the three major rock types (devitrified, zeolitic, and vitric) found in the UZ. The influence of expected variations in water chemistry, radionuclide concentrations, and variations in rock surface properties within one of the major rock types are Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 86 November 2003 incorporated into these probability distributions. These distributions are specified for each radionuclide/rock type combination (this Model Report, Section I.2 and I.4) and are sampled in the TSPA to account for the effects of natural variations in pore water chemistry and mineral surfaces on sorption. Correlations for sampling sorption coefficient probability distributions have been derived for the elements investigated (this Model Report, Section II). To derive the correlations, a rating system was first developed to rate the impact of six different variables on the sorption coefficient for a given element in each of the three major rock types. The six variables are pH, Eh, water chemistry, rock composition, rock surface area, and radionuclide concentration. Water chemistry refers to the major ion concentrations and silica. Rock composition refers to both the mineralogic composition of the rocks and the chemical composition of the minerals (e.g., zeolite compositions). The output DTNs for the sorption Kds and correlations are LA0302AM831341.002 and LA0311AM831341.001. 2.2.09.01.0B, Microbial Activity in the UZ FEP Description: Microbial activity in the UZ may affect radionuclide mobility in rock and soil through colloidal processes, by influencing the availability of complexing agents, or by influencing groundwater chemistry. Changes in microbial activity could be caused by the response of the soil zone to changes in climate. Screening Decision and Regulatory Basis: Included Screening Argument: None. TSPA Disposition: The effects of microbes on sorption are included in the distributions for sorption coefficients used in TSPA. The sorption coefficient data on which the distributions are based are obtained in laboratory experiments in which crushed rock samples from the Yucca Mountain site are contacted with groundwaters (or simulated groundwaters) representative of the site, spiked with one or more of the elements of interest (this Model Report, Section I.5). The basic technique for the laboratory determination of sorption coefficients involved the contact of a groundwater sample, spiked with the radionuclide of interest, with a crushed sample of tuff or alluvium. The rock sample was generally obtained as a core sample. The rock and water samples were not sterilized and therefore contain representative microbial biota from the UZ. Sorption experiments have been carried out as a function of time, element concentration, atmospheric composition, particle size, and temperature. In some cases, the solids remaining from sorption experiments were contacted with unspiked groundwater in desorption experiments. The experimental data used to determine the sorption Kds are provided in the following DTNs: LA0305AM831341.001 [163789], LA0309AM831341.002 [165523], LA0309AM831341.003 [165524], LA0309AM831341.004 [165525], LA0309AM831341.005 [165526], LA0309AM831341.006 [165527], LA0309AM831341.007 [165528], and LA0310AM831341.001 [165865]. The sorption and desorption experiments together provide information on the equilibration rates of the forward and backward sorption reactions. For elements that sorb primarily through surface complexation reactions, the experimental data are augmented with the results of modeling calculations using PHREEQC V2.3 (BSC 2001 [155323]). The inputs for the modeling calculations include groundwater compositions, surface areas, binding constants for the elements of interest, and thermodynamic data for solution species. These modeling calculations provide a basis for interpolation and extrapolation of the Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 87 November 2003 experimentally derived sorption coefficient dataset. The effects of nonlinear sorption are approximated by capturing the effective Kd range (this Model Report, Section I.8). The effects of microbial activity with respect to sorption coefficients are provided in terms of probability distributions for the sorption coefficient of each element of interest among the three major rock types (devitrified, zeolitic, and vitric) found in the UZ. The influence of expected variations in water chemistry, radionuclide concentrations, and variations in rock surface properties within one of the major rock types are incorporated into these probability distributions. These distributions are specified for each radionuclide/rock type combination (this Model Report, Section I.2 and I.4) and are sampled in the TSPA to account for the effects of natural variations in pore water chemistry and mineral surfaces on sorption. Correlations for sampling sorption coefficient probability distributions have been derived for the elements investigated (this Model Report, Section II). To derive the correlations, a rating system was first developed to rate the impact of six different variables on the sorption coefficient for a given element in each of the three major rock types. The six variables are pH, Eh, water chemistry, rock composition, rock surface area, and radionuclide concentration. Water chemistry refers to the major ion concentrations and silica. Rock composition refers to both the mineralogic composition of the rocks and the chemical composition of the minerals (e.g., zeolite compositions). The output DTNs for the sorption Kds and correlations are LA0302AM831341.002 and LA0311AM831341.001. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 88 November 2003 6.5 SORPTION MODEL, DIFFUSION, RADIOACTIVE DECAY, AND PARAMETERS In this section we discuss the radionuclide sorption model. The diffusion coefficients and the decay parameters of the radionuclides considered in this Model Report are also discussed. Based on the analysis in Attachments I and II, we substantiate the arguments for the validity of the linear equilibrium sorption model assumed to govern sorption in the UZ. The corresponding distribution coefficients Kd of the various radionuclides (describing sorption onto the UZ rocks) are listed in Table 6.5-1. A detailed discussion of the derivation process of these estimates and of the supporting laboratory data can be found in Attachment I. In this Model Report, all computations involving sorption are based on the linear equilibrium model and use Kd estimates of Table 6.5-1. Table 6.5-1. Kd in the Rocks of the Unsaturated Zone Species Unit/Analysis Distributiona Coefficients describing distributionb (ml/g) U Zeolitic Cumulative (Kd value, probability) (0, 0) (0.5, 0.5) (30, 1.0) Devitrified Cumulative (Kd value, probability) (0, 0) (0.2, 0.5) (4, 1.0) Vitric Cumulative (Kd value, probability) (0, 0) (0.2, 0.5) (3, 1.0) Np Zeolitic Cumulative (Kd value, probability) (0, 0) (0.5, 0.5) (6, 1.0) Devitrified Cumulative (Kd value, probability) (0, 0) (0.5, 0.5) (6, 1.0) Vitric Cumulative (Kd value, probability) (0, 0) (1.0, 0.5) (3, 1.0) Pu Zeolitic Cumulative (Kd value, probability) (10, 0) (100, 0.5) (200, 1.0) Devitrified Cumulative (Kd value, probability) (10, 0) (70, 0.5) (200, 1.0) Vitric Cumulative (Kd value, probability) (10, 0) (100, 0.5) (200, 1.0) Am Zeolitic Uniform Range = 100 – 1000 (500) Devitrified Uniform Range = 100 – 2000 (1,000) Vitric Cumulative (Kd value, probability) (100, 0) (400, 0.5) (1,000, 1.0) Pa Zeolitic Uniform Range = 1000 – 20,000 (10,000) Devitrified Uniform Range = 1000 – 20,000 (10,000) Vitric Uniform Range = 1000 – 20,000 (10,000) Cs Zeolitic Cumulative (Kd value, probability) (425, 0) (5,000, 0.5) (20,000, 1.0) Devitrified Uniform Range = 1 – 15 (7.5) Vitric Cumulative (Kd value, probability) (0, 0) (2, 0.5) (100, 1.0) Sr Zeolitic Uniform Range = 50 – 2000 (1000) Devitrified Uniform Range = 10 – 70 (40) Vitric Uniform Range = 0 – 50 (25) Ra Zeolitic Uniform Range = 1000 – 5,000 (2,500) Devitrified Uniform Range = 100 – 1,000 (500) Vitric Uniform Range = 50 – 600 (300) Th Zeolitic Uniform Range = 1,000 - 30,000 (15,000) Devitrified Uniform Range = 1,000 - 10,000 (5,000) Vitric Uniform Range = 1,000 - 10,000 (5,000) Output-DTN: LA0302AM831341.002 NOTE: a The term “cumulative” indicates a piecewise linear distribution b The numbers in boldface were used in the simulations Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 89 November 2003 The diffusion coefficients and the decay properties are listed in Table 6.5-2. Of the listed D0, only the first ones were used for the base 3-D simulations, while the rest were used for uncertainty bounding and sensitivity analysis (Sections 6.7). Table 6.5-2. Properties of Radionuclides in the Transport Simulations Radionuclide D0 (m2/s) T1/2 (years) † ) ( 2 ln 1 2 / 1 - = . s T Decay Mode † 99Tc 4.55× 10-10 (a) 4.55× 10-11 (f) 10-9 (f) 2.13× 105 1.031× 10-13 ß- 237Np 1.65× 10-10 (b) 7.12× 10-10 (f) 1.65× 10-11 (f) 10-9 (f) 2.14× 106 1.026× 10-14 a 239Pu 4.81× 10-10 (c) 4.81× 10-11 (f) 6.08× 10-10 (f) 10-9 (f) 2.41× 104 9.114× 10-13 a 241Am 3.69×10-10(d) 4.322×102 5.082× 10-11 ß- 233U 4.94×10-10(c) 1.59×105 1.3814×10-13 a 235U 4.89×10-10(c) 7.08×108‡ 3.1023×10-17 a 231Pa 4.98×10-10(c) 3.25×104‡ 6.7583×10-13 a 229Th 5.02×10-10(c) 7.90×103 2.7803×10-12 a 226Ra 8.89×10-10(e) 1.599×103 1.374×10-11 a 135Cs 2.06×10-9(e) 2.30×106 9.55×10-15 ß- 90Sr 7.91×1010(e) 2.90×101 7.574×10-10 ß- NOTE: (a): DTN: LA000000000034.002 [148603] (b): Using Mn+2 as analog, corrected for mass, see Lide (1992 [166224], pp. 5-111, 5-112) (c): Using UO+2 as analog, corrected for mass, see Lide (1992 [166224], pp. 5-111, 5-112) (d): Using Sm+3 and Ce+3 as analogs, corrected for mass, see Lide (1992 [166224], pp. 5- 111, 5-112 (e):directly used, see Lide (1992 [166224], 5-111, 5-112) (f): Values used for sensitivity analysis, see Section 6.8.3.2 † : From Lide (1992 [166224], pp. 11-28 – 11-133) ‡ : Values adjusted from Lide (1992 [166224], 11-13 – 11-133) Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 90 November 2003 6.6 A PRELIMINARY CONCEPTUAL MODEL OF TRANSPORT BASED ON GEOLOGY AND STRATIGRAPHY 6.6.1 General Issues This section focuses on the development of a conceptual model of radionuclide transport in composite vertical cross sections at select locations of the UZ. The conceptual model is based on the geology, stratigraphy and transport properties of the geohydrologic units in the profiles from the repository horizon to the water table at these locations. The focus is on the expected difference of transport patterns at different UZ locations (because of different geologies and conditions), and on the relative importance of particular geohydrologic units on overall transport. 6.6.2 Hydrogeologic Profiles Transport is considered in three hydrogeologic profiles (columns) called Cross Sections 1, 2, and 3. The three cross sections are located in the vicinity of the USW SD-6, SD-12, and UZ-14 boreholes, respectively. The locations of the three boreholes in the context of the UZ site-scale model are shown in Figure 6.6-1. Thus, Cross Section 3 is located in the northern part of the repository site, while Cross Sections 1 and 2 are located in the southern part. The geologic profiles in Cross Sections 1, 2, and 3, including the layers of all the hydrogeologic units involved in radionuclide transport, their elevations and thicknesses, are shown in Figures 6.6-2 through 6.6-4. Additional information on the three profiles can be found in Attachment III. 6.6.3 Conceptual Model of Transport in Cross Sections 1 and 2 Figure 6.6-5 illustrates a conceptual model of transport in Cross Sections 1 and 2. Water-borne radionuclides enter the TSw formation through the bottom of the repository. Radionuclide transport in the TSw hydrogeologic unit occurs mostly in the fractures. At this location the vitric layers in the underlying CHn unit (i.e., layers ch1v through ch5v) behave as porous (rather than fractured) media because of the parity of permeability in the matrix and in the fractures (BSC 2001 [161340], Section 6.7.2). Thus, transport occurs in both the matrix and the fractures, and the contact times between the radionuclides and the media is longer. The vitric layers are effective transport barriers because of the increased sorption (for sorbing radionuclides) and retardation (see Section 6.5, this Model Report). Once the radionuclide reaches the zeolitic ch6z unit, flow becomes again fracture-dominated, and, consequently, so does transport. Transport in the pp4 layer, the top layer in the PP hydrogeologic unit, is controlled by its zeolitic nature, which leads to fractured-dominated flow and transport. The next two layers (i.e., pp3 and pp2) are devitrified and behave similarly to the vitric layers in the CHn unit. Transport in the thick zeolitic pp1 layer (the last layer in the PP unit) is again fractured-dominated, as is transport in the underlying layers of the BF unit. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 91 November 2003 LEGEND UZ2002 Model Boundary 2002 Repository Boundary 2002 Repository Lower Block ESF and ECRB GFM2000 Faults Sever Wash Fault Exile Hill Fault Solitario Canyon Fault Ghost Dance (West) Fault Ghost Dance (West) Fault Solitario Canyon Fault (West) Dune Wash Fault Dune (West 1) Ghost Dance Fault Imbricate Fault "Toe" Fault Splay "S" Dune "X" Bow Ridge Fault Pagany Wash Fault "SolJFat" Fault "SolJFat" Fault Drill Hole Wash Fault Sundance Fault Sundance Fault Splay "g" Splay "g" Splay "N" Splay "N" NSP Northing (m) NSP Easting (m) UZ-14 SD-6 SD-12 Source: Modified from BSC 2003 [160109], Section 6, Figure 1a Figure 6.6-1. Two-Dimensional (Plan View) of the Yucca Mountain Site Identifying the Locations of Boreholes USW SD-6, SD-12, and UZ-14, Representing Cross Sections 1, 2, and 3 Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 92 November 2003 750 850 950 1050 1150 0 100 Elevation (m) tsw35 tsw36 tsw37 tsw39 ch1v ch4v ch3v ch2v ch5v ch6v pp4 pp2 pp3 pp1 bf3 groundwater table Repository tsw38 DTN: LB03013DSSCP31.001 [162379] Figure 6.6-2. Hydrogeologic Units (and their Layers) of the UZ Model Domain from the Repository to the Groundwater in Cross Section 1 Near the USW SD-6 Borehole Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 93 November 2003 700 800 900 1000 1100 0 100 Elevation (m) tsw36 tsw37 tsw38 tsw39 ch1v ch4z ch3v ch2v ch5v ch6z pp4 pp2 pp3 tsw35 pp1 groundwater table tsw34 Repository ch1z DTN: LB03013DSSCP31.001 [162379] Figure 6.6-3. Hydrogeologic Units (and their Layers) of the UZ Model Domain from the Repository to the Groundwater Table in Cross Section 2 Near the USW SD-12 Borehole Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 94 November 2003 700 800 900 1000 1100 0 100 Elevation (m) tsw35 tsw36 tsw37 tsw38 tsw39 ch1z ch4z ch3z ch2z ch5z ch6z pp4 pp2 pp3 tsw35 groundwater table Repository DTN: LB03013DSSCP31.001 [162379] Figure 6.6-4. Hydrogeologic Units (and their Layers) of the UZ Model Domain from the Repository to the Groundwater in Cross Section 3 Near the USW UZ-14 Borehole Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 95 November 2003 Repository ch (1-5)v tsw 36-39 ch6z pp4 pp3 pp2 pp1 bf3 bf2 SOURCE: Modified from CRWMS M&O 2000 [151940], Figure 3.11-3. Figure 6.6-5. A Conceptual Model of Transport in Cross Section 1. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 96 November 2003 6.6.4 Conceptual Model of Transport in Cross Section 3 Figure 6.6-6 illustrates a conceptual model of transport in Cross Section 3. Water-borne radionuclides enter the TSw formation through the bottom of the repository. Radionuclide transport in the TSw hydrogeologic unit occurs mostly in the fractures. In contrast to Cross Section 1, Cross Section 3 is characterized by the absence of vitric layers in the underlying CHn. The zeolitic layers here have fracture spacing and fracture permeability similar to those in Cross Section 1, but the matrix permeability is about five orders of magnitude lower than that in the fractures (BSC 2003 [163045], Section 6; BSC 2001 [161340], Section 6). The disparity in permeability directs practically all flow into the fractures, leading to fast transport. The pp4 layer, the top layer in the PP hydrogeologic unit, behaves like a zeolitic layer, leading to fractured-dominated flow and transport. The next two layers (i.e., pp3 and pp2) are devitrified and behave like porous media, and the pp1 layer (a thick zeolitic layer, and last one in the PP unit) is again fracture-dominated. 6.6.5 Synopsis and Discussion A review of the conceptual model in Section 6.6 indicates that while the geological profiles in Cross Sections 1 and 2 appear likely to afford good retardation of radionuclide transport (and especially of strongly sorbing radionuclides), this is unlikely to be the case in the domain of Cross Section 3, because of the preponderance of highly permeable (compared to the matrix) fractures of the zeolitic layers at this location. Based on general transport principles, a decreasing Kd is expected to lead to faster transport to the water table, although strongly sorbing radionuclides (such as 239Pu, 231Pa, 229Th) are unlikely to be significantly affected. Increases in the infiltration rate, brought about by possible changes in climatic patterns, can have a substantial impact on breakthrough times, especially at the Cross Section 3 location. The effect of changes in infiltration is likely to become more pronounced with weaker-sorbing radionuclides. The conceptual models supported by these results are consistent with the observations that were based on a simplified 2-D analysis included in the previous revision of this Model Report (CRWMS M&O 2000 [122799], Section 6) and in CRWMS M&O 2000 [151940], from which Figures 6.6-4 and 6.6-5 originated. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 97 November 2003 UZPMR3.11-3REV1 tsw35 tsw36 tsw37 tsw38 tsw39 ch6z ch5z ch4z ch3z ch2z ch1z pp4 pp3 pp2 pp1 Repository SOURCE: Modified from CRWMS M&O 2000 [151940], Figure 3.11-3. Figure 6.6-6. A Conceptual Model of Transport in Cross Section 3. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 98 November 2003 6.7 THREE-DIMENSIONAL TRANSPORT SIMULATIONS In this section, we discuss the general issues involved in the 3-D radionuclide transport through the UZ of Yucca Mountain, i.e., climatic conditions, hydrologic conceptual model, grids, flow systems, the radioactive species considered and the mode of their release, data inputs and outputs, in addition to the general approach and basic stipulations. All the 3-D numerical studies of radionuclide transport use a grid identical to that employed in the flow analysis documented in the Model Report UZ Flow Models and Submodels (BSC 2003 [163045]). The development and features of the 3-D grids for this numerical modeling effort are documented in the Analysis Report Development of Numerical Grids for UZ Flow and Transport Modeling (BSC 2003 [160109]) and further discussed in BSC (2003 [163045], Section 6). 6.7.1 Climatic Conditions The three climatic scenarios investigated here are identical to those discussed in BSC (2003 [163045], Section 6): present-day, monsoon and glacial. In this Model Report, reference to a “glacial” regime is intended to indicate glacial transition climate, as described in USGS 2001 [158378], Table 2, p. 66, rather than a full glacial climate. For each climatic scenario, a high, mean and low infiltration case was studied. The annual infiltration rates corresponding to each of the nine climatic cases are listed in Table 6.7-1. 6.7.2 Conceptual Hydrologic Model, Grids and Flow Simulations The grid, properties, and calibrated hydraulic parameters used in the transport simulations (DTN: LB03023DSSCP9I.001 [163044]) were identical to those used for the analysis of flow in BSC (2003 [163045], Section 6), in the course in which they were derived. These correspond to the #1 conceptual model of perched water (BSC 2001 [158726], Section 6.2). This is the permeability barrier model, which uses the calibrated perched-water parameters for fractures and matrix in the northern part of the model domain, and modified property layers (including the tsw38, tsw39, CH1z and CH2z layers) where the lower basal vitrophyre of the TSw is above the zeolites of the CHn. A detailed discussion of this perched-water model can be found in BSC (2003 [163045], Section 6.2). A 2-D (plan view) of the grid at the repository level is shown in Figure 6.7-1. All 3-D transport simulations that involved the use of the T2R3D numerical model were conducted using this grid and the steady-state flow fields from BSC (2003 [163045], Section 6) for the 9 climatic scenarios (DTN: LB03023DSSCP9I.001 [163044]). Table 6.7-1 provides averaged percolation values of all 9 climatic scenarios in order to show how there is more percolation flux under some climates than others. For the 3-D EOS9nT simulations, the same grid was used, but its numbering was altered. This modification involved a very minor rearrangement of the element order number using the XtractG.f90 routine, which moved (a) the elements above the repository, (b) those corresponding to the location of the repository, and (c) the bottom boundary elements (corresponding to the water table) to the bottom of the element file. Moving these elements to the bottom of the element file made use of the ability of the TOUGH2 family of codes to treat them as inactive elements (i.e., elements that have time-invariant properties and conditions, contribute to flow and transport as Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 99 November 2003 boundaries, and are not accounted for in mass and energy balances). To accomplish that, the only modification was making the volume of the first repository element zero or negative. In the TOUGH2 convention, this rendered all the following elements inactive, thus maintaining their properties and conditions constant throughout the simulations. Making these elements inactive was possible because (a) the steady state of the flow fields is unaffected by making these elements inactive, (b) in the EOS9nT simulations the release rate was considered constant over time, a condition which necessitated treating the repository elements as internal boundary elements, and (c) the elements above the repository do not contribute to transport and can also be treated as boundary elements. This modification reduced the number of active elements by about 50% in the EOS9nT calculations, made the simulation possible within the array size of the code, and significantly reduced the execution times, with absolutely no effect on or compromise of the transport computations . Note that all the properties and the attributes of the original grid were maintained, and the only change was the numbering sequence of the elements. Because the element names remained unaltered, the initial pressure and saturation conditions from the flow computations BSC (2003 [163045], Section 6) were applied unchanged, but the flow field could not be used as was provided by the simulations in BSC (2003 [163045], Section 6) because those did not reflect the changed element numbers. Thus, an updated steady state flow field (involving the same information as the old one, but in a different sequence) was obtained from a single TOUGH 2 V1.4 (Module EOS9 V1.4) (LBNL 2000 [146496]) simulation that used the modified grid and the initial pressure and saturation conditions from the flow computations (BSC 2003 [163045], Section 6). Only a single run was needed because all the EOS9nT simulations were made assuming a mean present-day infiltration scenario (see Table 6.7-1). After the EOS9 simulation, the output file SAVE was compared to those from the simulations using the unmodified grids to confirm that they were identical. Note that EOS9nT (LBNL 1999 [113943]) shares identical flow solution routines with EOS9, and confirmatory tests confirmed the identity of the flow solutions predicted by both EOS9 and EOS9nT. It was decided, however, to use the EOS9 module in the TOUGH2 flow simulations to ensure complete compatibility with the results in BSC (2003 [163045], Section 6). The velocity distribution (stored in the input file VELOC for the EOS9nT simulations) was then extracted from the EOS9 output by using the XtractG.f90 routine. The DTN of the modified grid, the XtractG.f90 input and output, and the EOS9 input (modified) and output files are Output- DTN: LB0307MR0060R1.003. Table 6.7-1. Percolation Fluxes (mm/year) for Different Climatic Regimes Value Present-Day Infiltration Monsoon Infiltration Glacial Infiltration Lower 1.25 4.43 2.35 Mean 4.43 11.83 17.02 Upper 10.74 19.23 31.69 Values averaged from DTN: GS000308311221.005 [147613] Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 100 November 2003 6.7.3 Modes of Radionuclide Release In all the ensuing 3-D studies of site-scale radionuclide transport, the radioactive substances were released in the liquid phase in the fractures of the elements corresponding to the repository. The initial radionuclide concentration was spatially uniform in the area of release, which covered the entire footprint of the repository. We investigated two modes of radionuclide release: instantaneous release and continuous release. 6.7.3.1 Instantaneous release An instantaneous release scenario corresponds to a single catastrophic event during which a finite mass of radionuclides is suddenly released in the liquid phase of the fractures of the repository. Study of the system performance following an instantaneous release provides the measure of the ability of the UZ to store and retard the migration of a limited mass of radioactive substances. 6.7.3.2 Continuous release Under a continuous release scenario, radionuclides are continuously released into the fractures of the elements corresponding to the repository. This is a realistic scenario, given the mass of the radioactive substances to be stored at the site, the expected deterioration of the canisters, and the planned closure of the repository after the storage of the radioactive materials (thus preventing access and repairs that might control continuous release). The concentration of the released radionuclides changes over time because the released radionuclides undergo radioactive decay at the input points (boundaries). Such an approach involves the assumption that all the radionuclides were produced as a batch at the same time, and is valid for all but the shortest-lived isotopes. This is far more consistent with reality than an assumption of a constant-boundary concentration, which implies continuous radionuclide production at the release point. Study of an instantaneous release scenario provides a criterion of absolute system performance and evaluates the time for a given radionuclide mass to migrate past the UZ boundaries. Such a measure of performance is inadequate and inapplicable under a continuous-release regime, the study of which necessitates a criterion of a relative system performance based on the relationship of inflow and outflows across the system boundaries. This provides a measure of the ability or capacity of the domain to accumulate radionuclide mass that enters its boundary at a given rate, and is just as important (if not more so) than the conventional breakthrough study that can only describe response to a finite mass release. Thus, under instantaneous releases, masses are compared, while in continuous release the relationship of fluxes is established. For a complete evaluation of the repository, performance under both release scenarios must be evaluated. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 101 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 Solitario Canyon Fault D rillho le Wash Fault PaganyWash Fault S ever Wash Fault Ghost Dance Fault Bow Ridge Fault Imbricate Fault Repository block DTN: LB990701233129.001 [106785] Figure 6.7-1. Two-Dimensional (Plan View) of the UZ Model Grid Design at the Repository Level Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 102 November 2003 6.7.4 Radionuclides Considered The following radioactive solutes were considered: (a) 99Tc (nonsorbing), (b) 237Np, 235U, 233U (moderately sorbing), (c) 241Am, 239Pu, 231Pa, 229Th, 226Ra, 90Sr (strongly sorbing). Their properties are listed in Tables 6.5-1 and 6.6-1. The radionuclides are released at the gridblocks corresponding to the location of the repository. Additionally, for the 3-D EOS9nT simulations of continuous release, all the important members in the decay chains of 241Am and 239Pu were considered, according to the decay equations (Pigford et al. 1980 [123113]) 241Am 237 Np 233 U 229 Th (Eq. 6-29) 239Pu 235 U 231 Pa (Eq. 6-30) Note that only the most important members of the radioactive chain are included in these decay equations, which omit daughters with short half-lives. Given that alpha decay is the decay mode of all the members in the 237Np and 239Pu chains, .=0 (Equation 6-27) in all the simulations reported in this Model Report because the daughters are ejected from grain surfaces due to recoil (Faure 1977 [122805], pp. 288–289). The transport of radioactive colloids was also investigated. Spherical PuO2 colloids with diameters of 6 nm, 100 nm, 200 nm and 450 nm were used in this study. 6.7.5 Important Geologic Features As will be clearly shown in the ensuing Sections 6.7 through 6.17, the role of faults in the 3-D site-scale transport model is very important because faults provide fast pathways for the radionuclide migration to the water table. In the EOS9, T2R3D and EOS9nT simulations, the faults are represented as thin vertical domains characterized by large permeabilities. More detailed description of the faults, their properties and their role in water flow, can be found in BSC (2003 [161773], Section 6.4), BSC (2002 [159124], Sections 6.2, 6.4, and 6.5), and BSC (2003 [163045], Section 6). 6.7.6 Transport Simulation Options The 3-D simulations involving the instantaneous release of radionuclides were conducted using the T2R3D code with the steady-state flow field option. T2R3D employs conventional timestepping. All the 3-D simulations describing continuous release of solutes and colloids were conducted by using the EOS9nT code with the De Hoog et al. (1982 [117308]) implementation of the Laplace transform formulation. This formulation was selected because of its speed and accuracy, its Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 103 November 2003 ability to provide information over the whole spectrum of the simulation period, and its ability to drastically reduce numerical diffusion. The resulting matrices were very well behaved, requiring no more than 10–15 conjugate gradient iterations to reduce residuals to below the 10-11 level in a matrix of order of about 220,000. Thus, simulations were very fast and efficient, requiring 1,800–2,200 seconds of execution time to cover a simulation period of 1,000,000 years. EOS9nT allows the simultaneous solution of the concentration profiles of any number of radionuclides. Thus, all three parents (99Tc, 237Np, 239Pu) were considered simultaneously in the runs. Similarly, exploiting the fact that EOS9nT allows the tracking of any daughter products, the solutions for each of the parent radionuclides and its daughters in the decay Equations 6-29 and 6-30 were obtained simultaneously. 6.7.7 Basic Data Inputs and Outputs All the 3-D T2R3D simulations for all radionuclides used the same grid, calibrated hydraulic parameters, and steady-state flow field. The initial condition files were different, as they reflected the nine different climatic conditions. All the corresponding data files were provided (and used without modification) from the flow analysis studies of BSC (2003 [163045]). Input files were developed for the T2R3D simulations for the radionuclides studied in this Model Report. The 3-D EOS9nT radionuclide simulations and the T2R3D simulations shared the same calibrated hydraulic parameters and initial flow conditions. The EOS9nT simulations used a slightly modified grid from that in T2R3D (Section 6.7.2) (composed of the same elements as the T2R3D grid, but with different ordering), and the corresponding flowfield for mean present-day infiltration. Using the latter, input files conforming to the EOS9nT requirements (including both flow and transport information for the various radionuclides under study) were created. The DTNs of the input and output files used and created in the course of this study of radionuclide transport through the UZ are listed in Table 6.7-2. 6.7.8 Review of the Approach and Basic Approximations in the 3-D Site-Scale Radionuclide Transport Studies For a better understanding and a more realistic interpretation of the results of the 3-D simulations, it is important to understand the basic approach and approximations employed in the study of transport in the UZ. Analysis of the simulation results without considering the extremely conservative approach and the succession of worst-case scenarios involved in the 3-D studies can lead to erroneous conclusions and a distorted view of the transport regimes through the UZ. To that effect, the note of caution discussed in Section 6.20.3 should be carefully reviewed. The numerical simulations are based on the following approximations and approach (not all of which will be incorporated in the TSPA model): (a) No drip shields are considered, and flow through the canisters is assumed. Thus, water from the drift roof is allowed to fall onto, and then flow uninhibited through the Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 104 November 2003 waste package, carrying radionuclides as it emerges from the ruptured container. This is a worst-case scenario, and rather unrealistic. (b) All the radioactive packages in the entire repository rupture simultaneously, with all of them releasing their contents simultaneously at the beginning of the simulation time. This is an improbable (if not outright impossible) release scenario. (c) The radionuclides are released directly into the fractures. The retardation effects of the invert (with porous media properties) underneath the waste packages are ignored. The presence of the invert can significantly delay the onset of radionuclide release into the underlying fractures. (d) The effects of the shadow zone are ignored. However, the presence of the shadow zone can lead to significant retardation of radionuclide transport (BSC 2003 [164889]). The presence of the shadow zone can lead to significant retardation of radioactive solute transport (BSC 2003 [164889]) because of the generally lower saturations. These lead to drastically lower water velocities in the shadow zone (into which water moves mainly in response to capillary pressure differentials), and to significantly lower diffusive fluxes across the matrix-fracture interfaces. The effect of the shadow zone on the transport of colloids is expected to be far more severe because advective fluxes are extremely low, and diffusive fluxes (the only mechanism capable of transporting radionuclides past the boundaries of the shadow zone) are dramatically depressed (as the colloid diffusion coefficient is generally orders of magnitude lower than the molecular diffusion coefficient of solutes). (e) The vertical fractures are modeled as continuous throughout the UZ from the repository to the water table. Additionally, all the vertical fractures are open (i.e., they are not even partially filled with a porous medium), and do not retard transport through solute sorption and/or colloid attachment onto the fracture walls. The implication of this approach/assumption of uninterrupted vertical fractures is fast and unimpeded advective transport of radionuclides from the repository to the water table. This worst-case scenario is highly unlikely. (f) The horizontal fractures are modeled as interconnected, and are also connected (directly or indirectly) to the vertical fractures. This scenario leads to accelerated migration of radionuclides to the water table. (g) The radioactive tracers (solutes or colloids) are stable, unaffected by the near field conditions (thermal, geochemical, physical), and are not subject to chemical immobilization (e.g., through irreversible sorption or precipitation) anywhere in the UZ. (h) The distribution coefficients were estimated over longer concentration intervals using an approach that results in milder slopes and lower Kd values. This results in lower overall sorption, and leads to more conservative estimates of transport that correspond to faster radionuclide arrivals at the water table. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 105 November 2003 (i) For simulation of the instantaneous radionuclide release scenario, the relative concentration (CR = C/C0) in the water saturation of all the elements corresponding to the fractures of the repository subdomain is set at an initial value of CR = 1, i.e. the initial release concentration is uniformly distributed in the repository fractures. Note that the water saturation in the fractures is that corresponding to the steady state flow field for the given infiltration regime. Given these simplifications (each one of which can accelerate release by thousands of years), the simulation results reflect a rather unrealistically conservative radionuclide release scenario, and are relevant (though by no means representative) after the radioactive species escape the immediate vicinity of the repository in a stable form. The onset of such release can occur a long time (thousands to hundreds of thousands) of years after the initial placement of the radioactive wastes in the repository. Thus, the radionuclide arrival times discussed in the following sections are meaningful only after (a) the drip shield and the canisters fail completely, (b) the radionuclides escape the effects of near-field hydrology, chemistry, and hydrology, and (c) enter a fully interconnected fracture system. Note that the transport model discussed here is intertwined with the UZ flow model discussed in BSC (2003 [163045]). As such, it shares the same conceptual strengths and weaknesses. Changes in the conceptual model of flow can and will have a significant effect on transport. In particular, if an alternative flow model (in which faults act as barriers rather than conduits to flow) is validated, dramatic changes in the transport performance of the UZ are not just possible but likely. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 106 November 2003 Table 6.7-2. DTNs of Input and Output Files of the 3-D Site-Scale Transport Simulations Item DTN 3-D grid for T2R3D runs LB03023DSSCP9I.001 [163044] Steady state 3-D flow fields for 9 climatic scenarios (Table 6.7-1) LB03023DSSCP9I.001 [163044] Calibrated rock properties, hydraulic parameters, flow parameters LB03023DSSCP9I.001 [163044], LB0210THRMLPRP.001 [160799] Input/Output files for the UZ radionuclide transport studies under instantaneous release (11 radionuclides, 9 climatic scenarios, sensitivity analyses, with direct radionuclide release into the fault fracture) LB0307MR0060R1.001 Data summaries and plot files corresponding to DTN: LB0307MR0060R1.001 LB0307MR0060R1.002 Input/Output files for the UZ radionuclide transport studies under mean present-day infiltration-continuous release (3 radioactive parent species, a 3- and a 4- member radioactive chain, 4 radioactive colloids, sensitivity analyses, with direct radionuclide release into the fault fracture) LB0307MR0060R1.003 Data summaries and plot files corresponding to DTN: LB0307MR0060R1.003 LB0307MR0060R1.004 Input/Output files for UZ radionuclide transport studies without direct releases in the faults-continuous and instantaneous release (4 radionuclides, mean present-day infiltration) LB0307MR0060R1.005 Data summaries and plot files corresponding to DTN: LB0307MR0060R1.005 LB0307MR0060R1.006 Data summaries and corresponding breakthrough curves for the radionuclide transport scenarios in DTNs: LB0307MR0060R1.001; LB0307MR0060R1.003; and LB0307MR0060R1.005 LB0307MR0060R1.007 Supplemental Radionuclide Transport Simulations: Input/Output files (Influence of three main geologic units (TSw, CHv and CHz), irreversibly sorbed radionuclides onto colloids, transport of Tc, Np, and Pu with no diffusion (instantaneous, continuous releases)) LB0310MR0060R1.010 Data summaries and plot files corresponding to DTN: LB0310MR0060R1.010 LB0310MR0060R1.011 Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 107 November 2003 6.8 3-D SIMULATIONS OF 99Tc TRANSPORT-INSTANTANEOUS RELEASE In this section we investigate the transport of 99Tc by means of 3-D site-scale simulations representing the entire UZ system of Yucca Mountain. The study analyzes the migration of this finite radionuclide mass through the UZ, and investigates the processes and phenomena that affect its arrival at the water table. Section 6.8 includes three subsections. In Section 6.8.1, we discuss the transport of the nonsorbing 99Tc for three levels of present-day infiltration. The results of this study are shown in Figures 6.8-11 to 6.8-24. In Section 6.8.2, we study the transport of 99Tc for three levels of (a) monsoon infiltration (Figure 6.8-25) and (b) glacial infiltration (Figure 6.8-26). Finally, in Section 6.8.3, we discuss uncertainties involved in the predictions, and we conduct sensitivity analyses to develop the envelop of possible solutions (Figure 6.8-27). 6.8.1 99Tc Transport Under Present-Day Infiltration The DTNs of the input and output files for the three present-day infiltration scenarios are listed in Table 6.7-2. The input parameters used in the T2R3D simulations of 3-D transport of the nonsorbing 99Tc are listed in Table 6.8-1. 6.8.1.1 Breakthrough Curves The release of a finite amount of radionuclide mass allows the use of the conventional concept of breakthrough. This is quantified by the evolution over time of the cumulative breakthrough of the 99Tc mass fraction arriving at the Yucca Mountain water table RM, defined as RM = M0 -M M0 , where M0 is the instantaneously released initial mass of the radionuclide (at t=0) and M is the radionuclide mass remaining in the domain of interest (i.e., in the UZ) at the time of observation. Note that the radionuclide mass M undergoes radioactive decay and decreases over time. In the instantaneous release studies, only the mass of the parent radionuclide is considered, i.e. the daughter products are not tracked. The RM of 99Tc at the repository in Figure 6.8-1 shows a very strong dependence on the infiltration regime. As the infiltration rate increases from low to mean, the t10 time, defined as the time at which RM = 0.1, decreases from about 13,900 years to about 83 years. Thus, t10 is an indicator of the effects of the most conductive geologic features (i.e., certain fractures) on transport. The t50, defined as the time at which RM =0.5, is not reached within 1,000,000 years for the low infiltration regime, but is at 6,640 years for mean infiltration regime. Further increase of infiltration to high present-day levels leads to a decrease of t10 to about 6 years, and of t50 to 230 years. The t10 and t50 of all the radionuclides discussed in this Model Report are summarized in tabular form in Tables 6.20-1, 6.20-2 and 6.20-3. It should be kept in mind that the term RM is relative, and these findings are only important if the magnitude of the released mass at the repository becomes significant. Figure 6.8-1 also shows that the maximum attainable RM increases with the infiltration rate. This is expected because lower infiltration results in lower velocities and longer travel times through UZ, and more radioactive decay occurs. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 108 November 2003 1.0 0.8 0.6 0.4 0.2 0.0 Normalized Mass Fraction 10-1 100 101 102 103 104 105 106 Time (years) 99 Tc - Present Infiltration Lower Mean Upper Output-DTN: LB0307MR0060R1.007 data submitted with this Model Report Figure 6.8-1. Cumulative Breakthrough of the 99Tc Mass Fraction RM at the Water Table for Varying Present-Day Climatic Scenarios Table 6.8-1. Input Parameters for the 3-D Site-Scale Simulations of Solute Transport (#1 Perched-Water Model) Parameters Source Tortuosity t˜ f Grathwohl (1998 [141512], pp. 28-35), Farrell and Reinhard (1994 [122803], p. 64) Properties and characteristics of the geologic units, steady-state pressures, water saturations and flow fields Table 6.7-2 (DTN: LB03023DSSCP9I.001 [163044], LB0210THRMLPRP.001 [160799]) Sorption distribution coefficients Kd Table 6.5-1 (see Attachment I) Molecular diffusion coefficient D0 Table 6.5-2 NOTE: See Section 5, Table 5-1 for assumptions Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 109 November 2003 6.8.1.2 Transport-Controlling Features and Flow Patterns In the study of radionuclide transport, we use the relative mass fraction XR, defined as XR = X X0 = C C0 = CR where X is the tracer mass fraction in the liquid phase (M/M), C is the tracer concentration in the liquid phase (M/L3), and the subscript 0 denotes the value of the subscripted parameter in the water released from the repository. For mean present-day infiltration, the distribution of the 99Tc XR in the aqueous phase within fractures and the matrix in the gridblocks directly above the TSw-CHn interface (i.e., at the bottom of the TSw, corresponding to the tsw39 layer) is given in Figures 6.8-2 through 6.8-11 for t = 10, 100, 1,000, 10,000, and 100,000 years. Note that the XR distributions in Figures 6.8-2 through 6.8-11 do not correspond to a plan view of a horizontal cross section, but follow the uneven topography of the bottom of the tsw39 layer. The XR distributions of 99Tc in the fractures and in the matrix immediately above the water table at the same times are given in Figures 6.8-12 through 6.8-21. From the comparison of these figures with Figure 6.7-1, it becomes apparent that transport in the RTM is both dominated and controlled by the faults. From the distribution of the fracture XR at the bottom of the TSw, it is evident that, immediately above the TSw-CHn interface (i.e., in the tsw39 layer), the Pagany Wash fault and the Drillhole Wash fault in the northern part of are the main transport-facilitating feature. As expected, radionuclide transport to that layer (immediately below the repository) is rapid. The impact of the faults on transport is evident as early as t=10 yrs, at which time the Pagany Wash fault appears to provide the fastest pathway (Figure 6.8-2), as indicated by the early appearance of radionuclides. Although 99Tc arrives later at the Drillhole Wash fault, it appears to have a larger contribution than the Pagany Wash fault for t =100 years. Both faults register a strong signature, and their presence is easily identified from the 99Tc distributions (Figure 6.8-4). At the same time, it is noteworthy that the Sundance fault is beginning to cast a faint shadow on the fracture XR map. The predominance of the Drillhole Wash fault and the Pagany Wash fault as the main transport-facilitating geologic features at the tsw39 level is pervasive during the entire simulation period, as evidenced by Figures 6.8-6, 6.8-8, and 6.8-10. At later times (t =10,000 years), the Solitario Canyon fault becomes an important transport conduit. The northern reach of the Ghost Dance fault becomes visible at t =1,000 years, but appears to be of secondary importance to transport to the tsw39 layer. Review of the fracture XR distributions at the water table (Figures 6.8-12, 6.8-14, 6.8-16, 6.8-18, and 6.8-20) indicates that the Drillhole Wash fault and the Pagany Wash fault continue to be the main transport conduits throughout the geologic profile, from the level of the repository to the water table. It is remarkable that these two faults show measurable radionuclide presence at the water table as early as t = 10 years, and at concentrations that are higher and more widely distributed than those observed at the bottom of the TSw (Figures 6.8-4 and 6.8-6). In other words, transport through these two faults is so fast that the water table concentrations of 99Tc exceed those immediately below the release point at t = 10 years because the radionuclides released into the repository fractures have already reached the water table. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 110 November 2003 In general, the appearance of 99Tc at the water table appears to be earlier and more pronounced at the water table than at the bottom of the TSw. Additionally, more faults active in radionuclide transport are evident at the water table. Note the early and strong signature of the Sundance fault and of the Sever Wash fault (which appears to intersect the radionuclide-carrying fractures emanating from the Pagany Wash fault) at t=10 years (Figure 6.8-12). The significance of the Drillhole Wash fault as the dominant conduit of 99Tc to the water table is further confirmed in Figure 6.8-14 (t = 100 years), where it is shown to have resulted in a large zone of high 99Tc concentrations that extends outward from the fault axis. At this time, in addition to the Sundance fault, the northern portion of the Ghost Dance fault and the Solitario Canyon fault appear to be making significant contributions to transport. Analysis of the matrix XR distributions at the bottom of the TSw and at the water table offers additional insights into the transport patterns. It is remarkable that, in tsw39, the matrix XR distributions at early times (Figures 6.8-3 and 6.8-5) show the highest concentrations in the southern part of the repository, in sharp contrast to the fractures that indicate transport exclusively in the northern part at the same times (Figures 6.8-2 and 6.8-4) and practically no sign of significant 99Tc presence in the matrix next to the rapidly transporting fractures. The reason for these vastly diverse transport patterns is the significantly different geology at the base of the TSw, which is dominated by the vitric CHv layers below the southern part of the repository, while the zeolitic CHz layers are predominant in the north. The permeabilities of the fractures and of the matrix of the vitric layers are similar in magnitude (BSC 2003 [160240], Section 6). This permeability parity, coupled with a large fracture spacing, result in a behavior similar to that of a nonfractured porous medium (BSC 2003 [160240], Section 6). Thus there is no early evidence of fracture flow in the southern part of the repository. In contrast, although the zeolitic CHz layers in the northern part of the repository have fracture spacing and fracture permeability similar to the vitric CHv layers in the south, their matrix permeability is about five orders of magnitude lower than their fracture permeability (BSC 2003 [160240], Section 6). Consequently, matrix flow in the zeolitic layers in the northern part of the repository is extremely slow and practically all the flow occurs in the fractures, leading to the fast transport observed in the fractures. The eastward movement of 99Tc in the matrix (Figures 6.8-5, 6.8-7 and 6.8-9) results from site geology. Once contamination reaches the interface, it moves primarily eastward, moving with the draining water that hugs the downward sloping (in this direction) low-permeability TSw-CHn interface. At later times, the matrix XR footprint shifts northward, as 99Tc diffusing from the fractures advances in matrix. This is evidenced by the transition apparent in Figures 6.8-5, 6.8-7 and 6.8-9). The presence of 99Tc in the matrix at the water table is significantly delayed, and we do not observe a pattern similar to that at the bottom of the TSw at early times. Despite significant radionuclide concentrations in the fractures associated with the faults discussed earlier, the matrix shows scant evidence of 99Tc presence (Figures 6.8-13 and 6.8-15). When it occurs, it follows the fault geometry and orientation, and is localized and limited to the immediate vicinity Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 111 November 2003 of the faults, indicating an origin consistent with diffusion from the fault fractures with no sign of downward migration through the matrix. This matrix concentration pattern further attests to the preponderance of fractures as the main transport-facilitating features. A pervasive feature of the transport pattern at the water table is that the area affected by significant radionuclide concentration is significantly larger in the fracture system than in the matrix. This is the opposite of what we observe at the bottom of the TSw, where the areal distribution of the matrix XR exceeds the footprint of the fracture XR. The obvious conclusion is that the fractures are practically the exclusive source of radionuclides appearing in the matrix, and that the large water velocities in the fractures lead to the limited areal extent and penetration of 99Tc into the matrix. 6.8.1.3 Evidence Supporting the Transport Patterns Of particular interest is the emerging transport pattern, which indicates that radionuclide transport to the water table is faster in the northern part of the repository, where it is also areally concentrated. Based on the properties of the layers beneath the repository, this appears to be qualitatively consistent with expectations of transport in the vertical cross sections discussed in Section 6.6. Thus, the general area of fastest, largest, and most extensive transport is in the northern part of the repository site, where the very low matrix permeability of the CHz directs practically all flow through the fractures. Figures 6.8-22a and 6.8-22b show the areally increasing extent of the fractured zeolitic tuffs with depth (Output-DTN: LB0307MR0060R1.002). As the distance from the repository increases, an increasing portion of the water flow occurs through the zeolites, in which the matrix permeability is very small and the fractured dominated flow (and, consequently, the advective transport) is fast. In addition to geology, support for the observed transport pattern is provided by the infiltration and percolation distributions. A review of the infiltration pattern (Figure 6.8-23) at the surface, the percolation flux at the repository level (Figure 6.8-24) and the percolation flux at the groundwater level (Figure 6.8-25) indicates that they closely reflect the transport patterns in Figures 6.8-2 through 6.8-21. These figures indicate that the water flow pattern dictates the advective transport pattern. The dominance of fracture flow (and, consequently, of advection-dominated transport) is supported by the relative magnitudes of hydraulic properties in the fractures and in the matrix. The permeability of the fractures in the faults can be as high as hundreds of darcies (BSC 2003 [160240], Section 6; Bodvarsson et al. 1999 [120055], p. 15), i.e., orders of magnitude larger. The resulting fast advective transport is the reason for the transport pattern observed in Figures 6.8-2 through 6.8-21. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 112 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-01 9.50E-02 9.00E-02 8.50E-02 8.00E-02 7.50E-02 7.00E-02 6.50E-02 6.00E-02 5.50E-02 5.00E-02 4.50E-02 4.00E-02 3.50E-02 3.00E-02 2.50E-02 2.00E-02 1.50E-02 1.00E-02 5.00E-03 0.00E+00 Tc-TSw-FL-10 Relative Mass Fraction Output-DTN: LB0307MR0060R1.002, data submitted with this Model Report Figure 6.8-2. Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures of the tsw39 Layer at t = 10 Years for Mean Present-Day Infiltration Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 113 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-02 9.50E-03 9.00E-03 8.50E-03 8.00E-03 7.50E-03 7.00E-03 6.50E-03 6.00E-03 5.50E-03 5.00E-03 4.50E-03 4.00E-03 3.50E-03 3.00E-03 2.50E-03 2.00E-03 1.50E-03 1.00E-03 5.00E-04 0.00E+00 Tc-TSw-ML-10 Relative Mass Fraction Output-DTN: LB0307MR0060R1.002, data submitted with this Model Report Figure 6.8-3. Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix of the tsw39 Layer at t = 10 Years for Mean Present-Day Infiltration Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 114 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-03 9.50E-04 9.00E-04 8.50E-04 8.00E-04 7.50E-04 7.00E-04 6.50E-04 6.00E-04 5.50E-04 5.00E-04 4.50E-04 4.00E-04 3.50E-04 3.00E-04 2.50E-04 2.00E-04 1.50E-04 1.00E-04 5.00E-05 0.00E+00 Tc-TSw-FL-100 Relative Mass Fraction Output-DTN: LB0307MR0060R1.002, data submitted with this Model Report Figure 6.8-4. Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures of the tsw39 Layer at t = 100 Years for Mean Present-Day Infiltration Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 115 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-03 9.50E-04 9.00E-04 8.50E-04 8.00E-04 7.50E-04 7.00E-04 6.50E-04 6.00E-04 5.50E-04 5.00E-04 4.50E-04 4.00E-04 3.50E-04 3.00E-04 2.50E-04 2.00E-04 1.50E-04 1.00E-04 5.00E-05 0.00E+00 Relative Mass Fraction Tc-TSw-ML-100 Output-DTN: LB0307MR0060R1.002, data submitted with this Model Report Figure 6.8-5. Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix of the tsw39 layer at t = 100 Years for Mean Present-Day Infiltration Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 116 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-04 9.50E-05 9.00E-05 8.50E-05 8.00E-05 7.50E-05 7.00E-05 6.50E-05 6.00E-05 5.50E-05 5.00E-05 4.50E-05 4.00E-05 3.50E-05 3.00E-05 2.50E-05 2.00E-05 1.50E-05 1.00E-05 5.00E-06 0.00E+00 Tc-TSw-FL-1000 Relative Mass Fraction Output-DTN: LB0307MR0060R1.002, data submitted with this Model Report Figure 6.8-6. Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures of the tsw39 Layer at t = 1,000 Years for Mean Present-Day Infiltration Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 117 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-04 9.50E-05 9.00E-05 8.50E-05 8.00E-05 7.50E-05 7.00E-05 6.50E-05 6.00E-05 5.50E-05 5.00E-05 4.50E-05 4.00E-05 3.50E-05 3.00E-05 2.50E-05 2.00E-05 1.50E-05 1.00E-05 5.00E-06 0.00E+00 Tc-PM-TSw-M-1000 Relative Mass Fraction Output-DTN: LB0307MR0060R1.002, data submitted with this Model Report Figure 6.8-7. Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix of the tsw39 Layer at t = 1,000 Years for Mean Present-Day Infiltration Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 118 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-05 9.50E-06 9.00E-06 8.50E-06 8.00E-06 7.50E-06 7.00E-06 6.50E-06 6.00E-06 5.50E-06 5.00E-06 4.50E-06 4.00E-06 3.50E-06 3.00E-06 2.50E-06 2.00E-06 1.50E-06 1.00E-06 5.00E-07 0.00E+00 Tc-TSw-FL-10,000 Relative Mass Fraction Output-DTN: LB0307MR0060R1.002, data submitted with this Model Report Figure 6.8-8. Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures of the tsw39 Layer at t = 10,000 Years for Mean Present-Day Infiltration Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 119 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-05 9.50E-06 9.00E-06 8.50E-06 8.00E-06 7.50E-06 7.00E-06 6.50E-06 6.00E-06 5.50E-06 5.00E-06 4.50E-06 4.00E-06 3.50E-06 3.00E-06 2.50E-06 2.00E-06 1.50E-06 1.00E-06 5.00E-07 0.00E+00 Tc-TSw-ML-10,000 Relative Mass Fraction Output-DTN: LB0307MR0060R1.002, data submitted with this Model Report Figure 6.8-9. Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix of the tsw39 Layer at t = 10,000 Years for Mean Present-Day Infiltration Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 120 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-07 9.50E-08 9.00E-08 8.50E-08 8.00E-08 7.50E-08 7.00E-08 6.50E-08 6.00E-08 5.50E-08 5.00E-08 4.50E-08 4.00E-08 3.50E-08 3.00E-08 2.50E-08 2.00E-08 1.50E-08 1.00E-08 5.00E-09 0.00E+00 Tc-TSw-FL-100,000 Relative Mass Fraction Output-DTN: LB0307MR0060R1.002, data submitted with this Model Report Figure 6.8-10. Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures of the tsw39 Layer at t = 100,000 Years for Mean Present-Day Infiltration Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 121 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-06 9.50E-07 9.00E-07 8.50E-07 8.00E-07 7.50E-07 7.00E-07 6.50E-07 6.00E-07 5.50E-07 5.00E-07 4.50E-07 4.00E-07 3.50E-07 3.00E-07 2.50E-07 2.00E-07 1.50E-07 1.00E-07 5.00E-08 0.00E+00 Tc-TSw-ML-100,000 Relative Mass Fraction Output-DTN: LB0307MR0060R1.002, data submitted with this Model Report Figure 6.8-11. Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix of the tsw39 Layer at t = 100,000 Years for Mean Present-Day Infiltration Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 122 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-03 9.50E-04 9.00E-04 8.50E-04 8.00E-04 7.50E-04 7.00E-04 6.50E-04 6.00E-04 5.50E-04 5.00E-04 4.50E-04 4.00E-04 3.50E-04 3.00E-04 2.50E-04 2.00E-04 1.50E-04 1.00E-04 5.00E-05 0.00E+00 Tc-WT-FL-10 Relative Mass Fraction Output-DTN: LB0307MR0060R1.002, data submitted with this Model Report Figure 6.8-12. Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures Immediately Above the Water Table at t = 10 Years for Mean Present-Day Infiltration Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 123 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-05 9.50E-06 9.00E-06 8.50E-06 8.00E-06 7.50E-06 7.00E-06 6.50E-06 6.00E-06 5.50E-06 5.00E-06 4.50E-06 4.00E-06 3.50E-06 3.00E-06 2.50E-06 2.00E-06 1.50E-06 1.00E-06 5.00E-07 0.00E+00 Tc-PM-WT-M-10 Relative Mass Fraction Output-DTN: LB0307MR0060R1.002, data submitted with this Model Report Figure 6.8-13 Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix Immediately Above the Water Table at t = 10 Years for a Mean Present-Day Infiltration Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 124 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-04 9.50E-05 9.00E-05 8.50E-05 8.00E-05 7.50E-05 7.00E-05 6.50E-05 6.00E-05 5.50E-05 5.00E-05 4.50E-05 4.00E-05 3.50E-05 3.00E-05 2.50E-05 2.00E-05 1.50E-05 1.00E-05 5.00E-06 0.00E+00 Relative Mass Fraction Tc-WT-FL-100 Output-DTN: LB0307MR0060R1.002, data submitted with this Model Report Figure 6.8-14 Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures Immediately Above the Water Table at t = 100 Years for Mean Present-Day Infiltration Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 125 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-05 9.50E-06 9.00E-06 8.50E-06 8.00E-06 7.50E-06 7.00E-06 6.50E-06 6.00E-06 5.50E-06 5.00E-06 4.50E-06 4.00E-06 3.50E-06 3.00E-06 2.50E-06 2.00E-06 1.50E-06 1.00E-06 5.00E-07 0.00E+00 Tc-PM-WT-M-100 Relative Mass Fraction Output-DTN: LB0307MR0060R1.002, data submitted with this Model Report Figure 6.8-15 Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix Immediately Above the Water Table at t = 100 Years for a Mean Present-Day Infiltration Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 126 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-05 9.50E-06 9.00E-06 8.50E-06 8.00E-06 7.50E-06 7.00E-06 6.50E-06 6.00E-06 5.50E-06 5.00E-06 4.50E-06 4.00E-06 3.50E-06 3.00E-06 2.50E-06 2.00E-06 1.50E-06 1.00E-06 5.00E-07 0.00E+00 Tc-WT-FL-1000 Relative Mass Fraction Output-DTN: LB0307MR0060R1.002, data submitted with this Model Report Figure 6.8-16 Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures Immediately Above the Water Table at t = 1,000 Years for Mean Present-Day Infiltration Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 127 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-05 9.50E-06 9.00E-06 8.50E-06 8.00E-06 7.50E-06 7.00E-06 6.50E-06 6.00E-06 5.50E-06 5.00E-06 4.50E-06 4.00E-06 3.50E-06 3.00E-06 2.50E-06 2.00E-06 1.50E-06 1.00E-06 5.00E-07 0.00E+00 Tc-WT-ML-1000 Relative Mass Fraction Output-DTN: LB0307MR0060R1.002, data submitted with this Model Report Figure 6.8-17 Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix Immediately Above the Water Table at t = 1000 years for a Mean Present-Day Infiltration Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 128 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-05 9.50E-06 9.00E-06 8.50E-06 8.00E-06 7.50E-06 7.00E-06 6.50E-06 6.00E-06 5.50E-06 5.00E-06 4.50E-06 4.00E-06 3.50E-06 3.00E-06 2.50E-06 2.00E-06 1.50E-06 1.00E-06 5.00E-07 0.00E+00 Tc-WT-FL-10,000 Relative Mass Fraction Output-DTN: LB0307MR0060R1.002, data submitted with this Model Report Figure 6.8-18 Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures Immediately Above the Water Table at t = 10,000 Years for Mean Present-Day Infiltration Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 129 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-05 9.50E-06 9.00E-06 8.50E-06 8.00E-06 7.50E-06 7.00E-06 6.50E-06 6.00E-06 5.50E-06 5.00E-06 4.50E-06 4.00E-06 3.50E-06 3.00E-06 2.50E-06 2.00E-06 1.50E-06 1.00E-06 5.00E-07 0.00E+00 Tc-WT-ML-10,000 Relative Mass Fraction Output-DTN: LB0307MR0060R1.002, data submitted with this Model Report Figure 6.8-19 Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix Immediately Above the Water Table at t = 10,000 Years for a Mean Present-Day Infiltration Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 130 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-06 9.50E-07 9.00E-07 8.50E-07 8.00E-07 7.50E-07 7.00E-07 6.50E-07 6.00E-07 5.50E-07 5.00E-07 4.50E-07 4.00E-07 3.50E-07 3.00E-07 2.50E-07 2.00E-07 1.50E-07 1.00E-07 5.00E-08 0.00E+00 Tc-WT-FL-100,000 Relative Mass Fraction Output-DTN: LB0307MR0060R1.002, data submitted with this Model Report Figure 6.8-20 Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures Immediately Above the Water Table at t = 100,000 Years for Mean Present-Day Infiltration Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 131 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-05 9.50E-06 9.00E-06 8.50E-06 8.00E-06 7.50E-06 7.00E-06 6.50E-06 6.00E-06 5.50E-06 5.00E-06 4.50E-06 4.00E-06 3.50E-06 3.00E-06 2.50E-06 2.00E-06 1.50E-06 1.00E-06 5.00E-07 0.00E+00 Tc-WT-ML-100,000 Relative Mass Fraction Output-DTN: LB0307MR0060R1.002, data submitted with this Model Report Figure 6.8-21 Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix Immediately Above the Water Table at t = 100,000 Years for a Mean Present-Day Infiltration Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 132 November 2003 ch1 ch2 NSP Northing (m) NSP Northing (m) NSP Easting (m) NSP Easting (m) tsw39 NSP Easting (m) NSP Northing (m) LEGEND Boreholes Vitric Region UZ Model Boundary Fault UZ#16 NOTE: Non-hachured areas within UZ Model boundary indicate "zeolitic" material Source: BSC 2003 [160109], Figure 6a Figure 6.8-22a Mineralogy Model Plots Below the Repository Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 133 November 2003 ch5 ch6 ch4 ch3 NSP Easting (m) NSP Northing (m) NSP Easting (m) NSP Northing (m) NSP Easting (m) NSP Northing (m) NSP Easting (m) NSP Northing (m) Source: BSC 2003 [160109], Figure 6b Figure 6.8-22b Mineralogy Model Plots Below the Repository Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 134 November 2003 15 14.25 13.5 12.75 12 11.25 10.5 9.75 9 8.25 7.5 6.75 6 5.25 4.5 3.75 3 2.25 1.5 0.75 0 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 Solitario Canyon Fault Drillhole Wash Fault PaganyWash Fault Sever Wash Fault Ghost Dance Fault Bow Ridge Fault Imbricate Fault (mm/year) Present Day Infiltration (Mean) Source: BSC 2003 [163045], Figure 6.1-2 Figure 6.8-23 Mean Present-Day Infiltration Rates at the Surface Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 135 November 2003 15 14.25 13.5 12.75 12 11.25 10.5 9.75 9 8.25 7.5 6.75 6 5.25 4.5 3.75 3 2.25 1.5 0.75 0 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 S olitario Canyon Fault Drillhole Wash Fault PaganyWash Fault Sever Wash Fault Ghost Dance Fault Bow Ridge Fault Imbricate Fault Vertical flux for preq_mA at repository layer mm/year Source: BSC 2003 [163045], Figure 6.6-1 Figure 6.8-24 Percolation Fluxes at the Repository Level Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 136 November 2003 15 14.25 13.5 12.75 12 11.25 10.5 9.75 9 8.25 7.5 6.75 6 5.25 4.5 3.75 3 2.25 1.5 0.75 0 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 S olitario Canyon Fault Drillhole Wash Fault PaganyWash Fault Sever Wash Fault Ghost Dance Fault Bow Ridge Fault Imbricate Fault Vertical flux for preq_mA at bottom boundary mm/year Source: BSC 2003 [163045], Figure 6.6-6 Figure 6.8-25. Percolation Fluxes at the Water Table Level Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 137 November 2003 6.8.2 99Tc Transport Under Monsoon and Glacial Infiltrations The DTNs of the input and output files for the three monsoon and the three glacial infiltration scenarios are listed in Table 6.7-2. The monsoon and glacial infiltration rate is higher than that for present-day infiltration of the same level (Table 6.7-1). The simulation results are consistent with the observations made in Section 6.8.1. The higher infiltration rates result in lower t10 and t50 values than those for present-day infiltration. Thus, for low, mean, and high monsoon infiltration, the t10 is 22, 9.4 and 2.4 years, respectively (Figure 6.8-26). For the same infiltration regimes, the corresponding t50 is 1,310, 417, and 92.4 years. These times are significantly smaller than the ones for the present-day infiltration scenario (Figure 6.8-1), indicating faster transport through the system and earlier appearance of 99Tc at the water table. Because transport is faster than under present-day infiltration, travel times to the water table are shorter and thus the amount of 99Tc lost to radioactive decay is smaller. This is evident in Figure 6.8-26, in which the maximum attainable RM values are shown (a) to increase with the infiltration rate and (b) to be higher and closer to each other than those for present-day infiltration (Figure 6.8-1). 1.0 0.8 0.6 0.4 0.2 0.0 Normalized Mass Fraction 10-1 100 101 102 103 104 105 106 Time (years) 99 Tc - Monsoon Infiltration Lower Mean Upper Output-DTN: LB0307MR0060R1.007, data submitted with this Model Report Figure 6.8-26 Cumulative Breakthrough of the 99Tc Mass Fraction RM at the Water Table for Varying Monsoon Climatic Scenarios Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 138 November 2003 1.0 0.8 0.6 0.4 0.2 0.0 Normalized Mass Fraction 10-1 100 101 102 103 104 105 106 Time (years) 99 Tc - Glacial Infiltration Lower Mean Upper Output-DTN: LB0307MR0060R1.007, data submitted with this Model Report Figure 6.8-27 Cumulative Breakthrough of the 99Tc Mass Fraction RM at the Water Table for Varying Glacial Climatic Scenarios For transport under glacial infiltration, Figure 6.8-27 indicates very early arrivals at the water table. Thus, the t10 for high, mean and low glacial infiltration are only 1.25, 5.68, and 102 years, respectively, while the corresponding t50 values are 42.3, 164, and 8,140 years, respectively. As in the case of monsoon infiltration (Figure 6.8-26), the maximum attainable RM for high and mean glacial infiltration is almost 1, indicating fast transport through the 3-D system. 6.8.3 Uncertainties in 99Tc Transport Predictions The predictions of 99Tc transport are subject to three general types of uncertainties. 6.8.3.1 Uncertainties in the Estimates of Flow and Hydraulic Properties The impact of uncertainties in the estimate of flow and hydraulic properties is in the estimation of the magnitude and relative sizes of flows and saturations in the fractures and in the matrices of the UZ. These directly affect both advective transport (and, consequently, travel times to the water table) and diffusive fluxes (see Equations 6-18 - 6-21). Although the importance of these uncertainties is readily recognized, addressing them is beyond the scope of this study, but rather Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 139 November 2003 the subject of the Model Report on the UZ Flow Model (BSC 2003 [163045], Section 6). Because the radionuclide transport study is intertwined with UZ flow model (from which it derives its flow fields), it has the same flow-related limitations and uncertainties. 6.8.3.2 Uncertainties in the Retardation Processes The only non-decay mechanisms of 99Tc retardation in the process of transport through the UZ is matrix diffusion (including dispersion) because it is nonsorbing and chemical immobilization is not considered. Uncertainties in matrix diffusion are reflected in the values of the diffusion coefficient D0. Given the uncertainties in the estimation of the parameters in Equations 6-20 and 6-21, the choice of D0 in general influences the effective diffusion coefficient. To address this issue, we investigated the sensitivity of the 99Tc transport through the UZ for the D0 values shown in Table 6.5-2, which cover the possible D0 range. The results are shown in Figure 6.8-28. As expected, D0 has a significant impact on breakthrough predictions, resulting in faster arrival times for lower D0 values. 1.0 0.8 0.6 0.4 0.2 0.0 Normalized Mass Fraction 10-1 100 101 102 103 104 105 106 Time (years) 99 Tc - Diffusion Coefficients Base case (Do=4.55x10-10 m2/s) Low Do (Do=4.55x10-11 m2/s) High Do (Do=10-9 m2/s) Output-DTN: LB0307MR0060R1.007, data submitted with this Model Report Figure 6.8-28 Effect of Uncertainty in D0 on the Cumulative Breakthrough of the 99Tc Mass Fraction RM Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 140 November 2003 6.8.3.3 Climatic Uncertainties As the results in the previous sections on 99Tc transport attest to, climatic changes affect the infiltration regime and can have a profound effect on transport. This uncertainty was addressed by estimating transport under all possible nine climatic scenarios of Table 6.7-1 (Figures 6.8-1, 6.8-26, and 6.8-27), thus bounding the possible solution. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 141 November 2003 6.9 3-D TRANSPORT OF 237Np – INSTANTANEOUS RELEASE In this section we investigate the transport of 237Np through the UZ of Yucca Mountain following an instantaneous release (Section 6.7). The 3-D site-scale simulations are conducted by means of T2R3D. Section 6.9 includes three subsections. In Section 6.9.1 we discuss the transport of the moderately sorbing 237Np for three levels of present-day infiltration. The results of this study are shown in Figure 6.9-1, and in Figures IV.1 to IV.24 of Attachment IV. In Section 6.9.2, we study the transport of 237Np for three levels of (a) monsoon infiltration (Figure 6.9-3) and (b) glacial infiltration (Figure 6.9-4). Section 6.9.3 addresses uncertainties involved in the predictions of 237Np transport, and involves sensitivity analyses to develop the envelope of possible solutions. 6.9.1 Transport of 237Np Under Present-Day Infiltration The grids, flow fields, and conditions are as discussed in Section 6.7. The parameters used in the T2R3D simulations of 3-D transport of the moderately sorbing 237Np are listed in Tables 6.5-1, 6.6-1, 6.7-1 and 6.7-2. The DTN of the input and output files for the three present-day infiltration scenarios is Input/Output-DTN: LB0307MR0060R1.001 (Table 6.7-2), and Output DTN: LB0307MR0060R1.002. 6.9.1.1 Breakthrough Curves The moderate sorption of 237Np results in retardation of its transport through the UZ system, as the cumulative breakthrough RM of the 237Np mass fraction at the water table indicates (Figure 6.9-1). Comparison to that of the nonsorbing 99Tc (Figure 6.8-1) reveals a very different pattern. Thus, the moderate sorption of 237Np is sufficient to result in a large increase in t10 and t50 compared to those for the nonsorbing 99Tc. Increased infiltration rate leads to faster transport and shorter travel times to the repository. When infiltration increases from mean to high, t10 and t50 are reduced from 410 and 25,400 to 4.5 and 1,600 years, respectively. Reduction of infiltration to the low present-day level results in analogous increase of t10 to 33,800, while t50 is not reached until after 1,000,000 years. Because of significant retardation, the maximum attainable RM varies over a larger value range and does not necessarily reach a plateau within the simulation period. At t = 1,000,000 years, RM for mean and low infiltration is 0.96 and 0.48, respectively, and the maximum possible values for either infiltration are not yet reached. At the same time, the RM for high infiltration is approaching its maximum at RM=0.99. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 142 November 2003 1.0 0.8 0.6 0.4 0.2 0.0 Normalized Mass Fraction 10-1 100 101 102 103 104 105 106 Time (years) 237Np - Present-Day Infiltration Lower Mean Upper Output-DTN: LB0307MR0060R1.007, data submitted with this Model Report Figure 6.9-1 Cumulative Breakthrough of the 237Np Mass Fraction RM at the Water Table for Varying Present-Day Climatic Scenarios 6.9.1.3 Transport Mechanisms and Patterns The importance of fractures on the 237Np transport becomes evident in Figure IV.1 (DTN: LB0307MR0060R1.002), which shows the areal distribution of XR (see Section 6.9.2.2) in the aqueous phase in the fractures immediately above the TSw-CHn interface (i.e., in the tsw39 layer) at t = 10 years (under mean present-day infiltration). As in the case of 237Np, the Drillhole Wash fault and Pagany Wash fault are clearly identified, as they define the extent of radionuclide presence and distribution at this level. The contrast with the distribution of XR in the aqueous phase of the matrix (Figure IV.2), which shows no discernible difference from the background XR = 0 in the vicinity of these faults, is dramatic. Figure IV.2 shows significant concentrations in the matrix in the southern part of the repository, where the permeability in the fractures is about the same as that in the matrix (see discussion in Sections 6.8.1.2 and 6.8.1.3). Figures IV.3 and IV.4 show the fracture XR and the matrix XR distribution of 237Np in the tsw39 layer at t = 100 years. Note the significant similarities to the corresponding distribution patterns in Figures IV.1 and IV.2. It is noteworthy that the fracture XR distribution shows significantly lower peak values, and that the contributions of the Sever Wash fault and the Sundance fault to transport to the tsw39 unit are becoming apparent at this time. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 143 November 2003 The fracture XR and matrix XR at t = 1,000 years are shown in Figures IV.5 and IV.6, respectively. The magnitude of the fracture XR follows the downward trend identified in Figure IV.3, although the areal distribution of 237Np occurrence in the fracture continuum expands, and the signatures of the main faults (Drillhole Wash, Pagany Wash, Sever Wash, Sundance) become more prominent. The matrix XR in Figure IV.6 shows the same footprint as those at t = 10 years and t = 100 years (Figures IV.2 and IV.4), but the maximum XR level is substantially lower. This indicates that the bulk of the mass of the instantaneously released 237Np has moved past the tsw39 level at t=100 years in the southern part of the repository. Figures IV.7 and IV.8 show the fracture XR and the matrix XR distribution of 237Np in the tsw39 layer at t = 10,000 years, respectively, and reveal a change in the transport regimes. The areal extent of the fracture XR increases compared to that at earlier times. This is attributed to transport through fractures over a larger area, as 237Np arrives from the repository at the bottom of the TSw through less conductive fractures, and/or after the radionuclide reaches such fractures (following diffusion into the matrix from the main fast fractures). The matrix XR distribution in Figure IV.8 confirms the earlier indication that the bulk of the 237Np mass has moved past the tsw39 layer because the extent and magnitude of the radionuclide signature in the southern part keeps decreasing. Conversely, the matrix in the vicinity of the main faults in the northern part of the repository is showing increasing evidence of 237Np, the source of which is diffusion into the matrix through the fracture walls (since sufficient time has elapsed to allow it) and/or direct matrix flow. At t = 100,000 years, the Solitario Canyon fault begins to contribute to the fracture XR (Figure IV.9), although the matrix XR values continue to decrease as the radionuclide mass moves toward the water table and some of it sorbs onto the UZ rocks. This is further supported by the matrix XR distribution of 237Np at the same time (Figure IV.10), which shows a faint signal practically everywhere as the bulk of the 237Np mass has already receded past the tsw39 layer. At t = 1,000,000 years (Figures IV.11 and IV.12), there is only very faint evidence of 237Np in the vicinity of the Pagany Wash fault. The XR distribution immediately above the water table provides another indication of the importance of fractures. For mean present-day infiltration, the fracture XR of 237Np immediately above the water table at t = 10 years (Figures IV.13) clearly defines the Drillhole Wash and Pagany Wash faults. The corresponding matrix XR (Figure IV.14) shows a much weaker signature along the Drillhole Wash fault. At t = 100 years, the fracture XR (Figure IV.15) indicates concentration of the radionuclide along the main faults (i.e., the Drillhole Wash and Pagany Wash faults, and, to a lesser extent, the Sever Wash and Sundance faults). It is remarkable that the matrix XR at the same time (Figure IV.16) does not show the same pattern, but begins to show the effect of transport through the Sundance fault. Thus, the fast-conducting faults limit the extent of the 237Np lateral penetration into the matrix through the fracture walls because of the limited radionuclide mass, the high water velocity (which does not permit substantial diffusion), and the 237Np sorption onto the matrix. At 1,000 and 10,000 years, the patterns described in Figures IV.15 and IV.16 persist. The Drillhole Wash and Pagany Wash faults continue to dominate transport (Figure IV.17 and IV.19), while the vicinity of the Sundance fault is the location of significant matrix XR values Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 144 November 2003 (Figures IV.18 and IV.20 at t=1,000 years and t=10,000 years, respectively). The larger areal extent of fracture XR at t = 10,000 years results in the very low matrix XR signature of the Drillhole Wash fault. The faults lead to relatively fast transport from the repository to the water table, resulting in short residence times and limited penetration into the matrix (and, consequently, limited retardation through sorption). At longer times, the effects of limited radionuclide mass (a consequence of the instantaneous release scenario), sorption, and decay are demonstrated by the decreasing magnitude and areal extent of the fracture and matrix XR (Figures IV.21 to IV.24). 6.9.1.4 Transport-Controlling Features Review of Figures IV.1 through IV.24 (see Attachment IV), and comparison to the corresponding 99Tc figures (Figures 6.8-2 through 6.8-17) indicate that transport is both dominated and controlled by the same faults identified and discussed in Sections 6.8.1.2 and 6.8.1.3. Thus, the Drillhole Wash fault and the Pagany Wash fault are the main transport-facilitating geologic features, and leads to the early appearance of 237Np in the tsw39 layer and at the water table. What is remarkable in the case of 237Np is that that the Drillhole Wash fault and the Pagany Wash fault play an even more important role in transport, as they appear to convey a larger portion of the released mass and in a shorter time than in the case of 99Tc. These faults are very clearly identified as conduits of very fast transport to the water table. The Sever Wash fault and the Sundance fault are important at later times, but their contributions to transport are of lesser importance. The faults appear to act as barriers to the lateral migration of radionuclides. As in the case of 99Tc transport, radionuclide transport to the groundwater is much faster in the northern part of the repository, where it is also areally concentrated. This is caused by (a) the infiltration and percolation patterns, (b) the fast conduit to transport provided by the faults; (c) the permeability regimes (and their effect on flow and to transport) in the northern part; and (d) the distribution, fracture characteristics, and flow behavior of the CHz and CHv layers. A detailed discussion of transport patterns can be found in Section 6.9.2.3. 6.9.1.5 Comparison to the 99Tc Transport The obvious reason for the difference between the 99Tc and the 237Np transport behavior is sorption. This results in (a) the universally lower (than 99Tc) concentrations and the more limited areal extent (footprint) of 237Np in the fractures and in the matrix at the same times, (b) the persistence over a longer time of the area underneath the southern part of the repository in tsw39 as the location where the matrix XR is significant (Figures IV.6 and IV.8, vs. Figures 6.8-7 and 6.8-9), (c) the slower breakthroughs indicated by Figures IV.12 through IV.24 and Table 6.20-1, and (d) the more limited extent of the radionuclide penetration into the UZ. 6.9.1.6 Effect of Sorption in the Main Geologic Units on 237Np Transport Of particular interest is the effect of the main hydrogeologic units (i.e., TSw, CHv and CHz) on the transport and retardation of 237Np. This was investigated by setting the corresponding distribution coefficients Kd in the all the layers of each of these units to zero, and then comparing the resulting cumulative breakthrough curve to that for the base case. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 145 November 2003 The results (for a mean present-day infiltration scenario) are shown in Figure 6.9-2, and indicate that TSw is by far the most important unit in the transport and retardation of 237Np. By eliminating sorption in the all the TSw layers, water table arrivals of 237Np occur significantly earlier. The effect of no sorption in the CHz and CHv units are secondary (if not marginal), and become apparent only at later times. Given that no sorption occurs in the fractures, the lack of any appreciable retardation during advection through the CHz and CHv tends to further indicate that fracture flow is the overwhelmingly dominant mechanism of transport through these units, while advection through matrix flow does not play a significant role. 1.0 0.8 0.6 0.4 0.2 0.0 Normalized Mass Fraction 10-1 100 101 102 103 104 105 106 Time (years) 237 Np Base case Kd = 0 in TSw Kd = 0 in CHz Kd = 0 in CHv Output-DTN: LB0310MR0060R1.010, data submitted with this Model Report Figure 6.9-2. The effect of sorption in TSw, CHz and CHv on the cumulative breakthrough of 237Np at the water table (mean present-day simulation). 6.9.2 237Np Transport Under Monsoon and Glacial Infiltration The DTN of the input and output files for the three monsoon infiltration scenarios is Input/Output-DTN: LB0307MR0060R1.001 (Table 6.7-2), Output DTN: Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 146 November 2003 LB0307MR0060R1.002. The higher monsoon and glacial infiltration rates (compared to those in present-day regimes, see Table 6.7-1) lead to the faster transport indicated in Figures 6.9-3 and 6.9-4 (when compared to that shown in Figure 6.9-1). The t10 for low, mean, and high monsoon infiltration is 14.8, 8.5, and 1.6 years, respectively. The corresponding t50 are 6,140, 2,120, and 714 years, respectively. These times are significantly lower than the ones for the present-day infiltration scenario, as the faster transport through the system would warrant. Because transport is faster than under present-day infiltration, radioactive decay of the released 237Np has a less pronounced effect in the reduction of the RM at a given time. The t10 for low, mean, and high glacial infiltration is 1830, 4 and 0.8 years, respectively. The corresponding t50 are 34,400, 1,070, and 336 years, respectively. For high and mean glacial infiltration, the t10 and t50 times are the shortest in the three infiltration regimes. 1.0 0.8 0.6 0.4 0.2 0.0 Normalized Mass Fraction 10-1 100 101 102 103 104 105 106 Time (years) 237Np - Monsoon Infiltration Lower Mean Upper Output-DTN: LB0307MR0060R1.007, data submitted with this Model Report). Figure 6.9-3. Cumulative Breakthrough of the 237Np Mass Fraction RM at the Yucca Mountain Water Table for Varying Monsoon Climatic Scenarios Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 147 November 2003 1.0 0.8 0.6 0.4 0.2 0.0 Normalized Mass Fraction 10-1 100 101 102 103 104 105 106 Time (years) 237 Np - Glacial Infiltration Lower Mean Upper Output-DTN: LB0307MR0060R1.007, data submitted with this Model Report Figure 6.9-4. Cumulative Breakthrough of the 237Np Mass Fraction RM at the Yucca Mountain Water Table for Varying Glacial Climatic Scenarios 6.9.3 Uncertainties in 237Np Transport Predictions The predictions of 237Np transport are subject to three general types of uncertainties discussed in Section 6.8.3. For the reasons discussed in that section, here we will limit our discussion to the issue of uncertainties in the transport retardation processes. Because 237Np is a sorbing radionuclide, uncertainties involve both diffusion and sorption issues. 6.9.3.1 Uncertainties in Matrix Diffusion As previously indicated, uncertainties in matrix diffusion are reflected in the values of the diffusion coefficient D0. To address this issue, we investigated the sensitivity of the 237Np transport through the UZ for the D0 values shown in Table 6.5-2, which cover the possible D0 range. The results are shown in Figure 6.9-5. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 148 November 2003 As expected, D0 has a significant impact on breakthrough predictions, resulting in faster arrival times for lower D0 values. Note that the relative effect of the D0 uncertainty in predicting transport of the sorbing 237Np is comparatively less pronounced than that of the nonsorbing 99Tc, i.e., the impact of D0 uncertainty on radionuclide breakthrough predictions appears to be mitigated by sorption. 6.9.3.2 Uncertainties in Sorption Such uncertainties in sorption are reflected in the values of the distribution coefficient Kd. To address this issue, we investigated the sensitivity of the 237Np transport through the UZ for the Kd values shown on Figure 6.9-6. These cover the range between zero (no sorption, in which case behavior similar to that of 99Tc was expected) to the maximum values reported in Table 6.5- 1. The results are shown in Figure 6.9-6. As expected, Kd has a significant impact on breakthrough predictions, resulting in faster arrival times for lower Kd values. Note that the relative effect of Kd uncertainty predicting transport of the sorbing 237Np appears to be much more pronounced than that for the D0 uncertainty for the range tested here. 1.0 0.8 0.6 0.4 0.2 0.0 Normalized Mass Fraction 10-1 100 101 102 103 104 105 106 Time (year) 237 Np - Diffusion Coefficients Base case (Do =1.65x10-10 m2/s) Low Do (Do=1.65x10-11 m2/s) High Do (Do=7.12x10-10 m2/s) Output-DTN: LB0307MR0060R1.007, data submitted with this Model Report Figure 6.9-5 Effect of Uncertainty in D0 on 237Np Breakthrough at the UZ Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 149 November 2003 1.0 0.8 0.6 0.4 0.2 0.0 Normalized Mass Fraction 10-1 100 101 102 103 104 105 106 Time (year) 237 Np - Sorption Coefficients Base Case (K d=0.3-0.5 mL/g) Low K (Kd d =0 mL/g) High Kd (Kd=3-6 mL/g) Output-DTN: LB0307MR0060R1.007, data submitted with this Model Report Figure 6.9-6 Effect of Uncertainty in Kd on 237Np Breakthrough at the UZ Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 150 November 2003 6.10 THREE DIMENSIONAL TRANSPORT OF 239Pu – INSTANTANEOUS RELEASE In this section we investigate the transport of 239Pu through the UZ of Yucca Mountain following an instantaneous release (Section 6.7). The 3-D site-scale simulations are conducted by using the T2R3D code. In Section 6.10.1, we discuss the transport of the strongly sorbing 239Pu for three levels of (a) present-day, (b) monsoon, and (c) glacial infiltration. The results of this study are shown in Figures 6.10-1 and 6.10-2. In Section 6.10.2 we address the issue of uncertainty involved in the predictions of 239Pu transport and conduct sensitivity analyses to develop the envelope of possible solutions (Figure 6.10-4). In Section 6.10.3, we discuss the applicability of the uncertainty studies conducted thus far on other radionuclides. 6.10.1 Transport of 239 Pu Under Present-Day Infiltration The grids, flow fields, and conditions discussed in Section 6.7 are used in the simulation of all infiltration regimes. The parameters used in the EOS9nT simulations of 3-D transport of the strongly sorbing 239Pu are listed in Tables 6.5-1, 6.6-1, 6.7-1, and 6.7-2. The DTN of the input and output files for the three present-day infiltration scenarios is Input/Output-DTN: LB0307MR0060R1.001 and Output DTN: LB0307MR0060R1.002. The t10 and t50 of 239Pu for the various infiltration regimes are listed in Table 6.20-1. 6.10.1.1 Breakthrough Curves The cumulative breakthrough RM of the 239Pu mass fraction at the water table for present-day infiltration is shown in Figure 6.10-1. If the contributions of the daughter products (i.e., 235U and 231Pa) are not accounted for, RM first reaches a plateau at the RM= 0.1 level (at t = 20 years, a clear indication of fracture flow), and then a second plateau at about 100,000 years (denoting matrix flow). This behavior results from the strong sorption of 239Pu. The faster transport depicted in Figures 6.10-2 is the result of the higher monsoon and glacial infiltration rates, respectively. The changes in climatic conditions lead to breakthrough curves that follow the same pattern established in the 239Pu present-day transport. A higher infiltration results in two plateaus that are higher than those for present-day infiltration, but occurs at about the same times (denoting fracture and matrix transport). As in the cases of 99Tc and 237Np, a higher infiltration leads to earlier arrival of 239Pu at the water table. However, the mass arriving at the water table represents a very small portion of the initial release because of strong sorption. 6.10.1.2 Transport Mechanisms and Patterns, and Transport-Controlling Features The importance of fractures on the 239Pu transport is very significant, as can be attested to by the contour plots available in the output submitted with this Model Report (Output-DTN: LB0307MR0060R1.002). Thus, the fracture XR at times as early as t = 10 years shows a sizable 239Pu presence in tsw39 and a measurable presence at the water table. Review of the transport patterns of the fracture XR in tsw39 and at the water table confirms the earlier observations (from the 99Tc and 237Np studies) that the Drillhole Wash and the Pagany Wash faults are the main conduits of fast radionuclide transport through the UZ to the water table. The Sever Wash and the Sundance faults also contribute to transport at later times, but the Solitario Canyon fault Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 151 November 2003 appears to have no effect because of the strong sorption of 239Pu onto the UZ rocks. The matrix XR distribution closely follows that of the fracture XR, but has a more limited areal extent and substantially lower magnitude. This is consistent with a 239Pu transport pattern dominated by the fault fractures, with practically no matrix-to-matrix advection, and with 239Pu occurrence in the matrix due to diffusion from the fractures. 6.10.1.3 Comparison to the 99Tc and 237Np Transport A comparison of the 239Pu transport to the transport of 99Tc (Section 6.8) and 237Np (Section 6.9 and Attachment IV) clearly indicates the same general patterns. They all exhibit the same behavior, are affected by the same mechanisms and are controlled by the same geologic features, but the magnitude of the observed fracture and matrix XR varies considerably because of the very different sorption affinity of these radionuclides onto the various UZ rocks. Increasing sorption affinity leads to lower concentrations (especially in the fractures), and consequently more limited areal distributions of mobile radionuclides. 1.0 0.8 0.6 0.4 0.2 0.0 Normalized Mass Fraction 10-1 100 101 102 103 104 105 106 Time (years) 239Pu - Present-Day Infiltration Lower Mean Upper Output-DTN: LB0307MR0060R1.007, data submitted with this Model Report Figure 6.10-1. Cumulative Breakthrough of the 239Pu Mass Fraction RM at the Water Table for Varying Present-Day Climatic Scenarios Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 152 November 2003 (a) 1.0 0.8 0.6 0.4 0.2 0.0 Normalized Mass Fraction 10-1 100 101 102 103 104 105 106 Time (years) 239Pu - Monsoon Infiltration Lower Mean Upper (b) 1.0 0.8 0.6 0.4 0.2 0.0 Normalized Mass Fraction 10-1 100 101 102 103 104 105 106 Time (years) 239Pu - Glacial Infiltration Lower Mean Upper Output-DTN: LB0307MR0060R1.007, data submitted with this Model Report Figure 6.10-2. Cumulative Breakthrough of the 239Pu Mass Fraction RM at the Water Table for (a) Monsoon and (b) Glacial Infiltration Thus, 239Pu transport exhibits (a) a XR that is universally lower than in the mildly sorbing 237Np, and even more so than in the nonsorbing 99Tc, and a more limited areal extent (footprint) of 239Pu in the fractures and in the matrix at the same times, (b) persistence over a longer time of the area underneath the southern part of the repository in tsw39 as the location where the matrix XR is significant, (c) slower breakthroughs, and (d) a more limited extent of the radionuclide penetration into the UZ. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 153 November 2003 6.10.1.4 Effect of Sorption in the Main Geologic Units on 239Pu Transport The effect of the main hydrogeologic units (i.e., TSw, CHv and CHz) on the transport and retardation of 239Pu was also investigated. The cumulative breakthrough curves corresponding to a zero Kd for the TSw, CHv and CHz units are shown in Figure 6.10-3. The results (for a mean present-day infiltration scenario) are analogous to, but far more pronounced than, those discussed in the case of 237Np transport (see Section 6.9.1.6, this Model Report). By setting the Kd to zero in the all the TSw layers, larger amounts of 239Pu arrive earlier at the water table, and a much higher RM value is attained, indicating that much larger amounts of 239Pu can cross the UZ. The overall effect of no sorption in the CHz and CHv units is practically insignificant compared to the base case. As in the case of 237Np transport, these results further underscore the importance of the TSw unit on the transport and retardation of 239Pu, and illustrate the overwhelming dominance of fractures in UZ transport. 0.4 0.3 0.2 0.1 0.0 Normalized Mass Fraction 10-1 100 101 102 103 104 105 106 Time (years) 239 Pu Base case Kd = 0 in TSw Kd = 0 in CHz Kd = 0 in CHv Output-DTN: LB0310MR0060R1.010, data submitted with this Model Report Figure 6.10-3. The Effect of Sorption in TSw, CHz and CHv on the Cumulative Breakthrough of 239Pu at the Water Table (Mean Present-Day Simulation). Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 154 November 2003 6.10.2 Uncertainties in 239Pu Transport Predictions The predictions of 239Pu transport are subject to three general types of uncertainties discussed in Sections 6.8.3 and 6.9.3. For the reasons discussed in these sections, here we will limit our discussion to the issue of uncertainties in the transport retardation processes. Although 239Pu is a sorbing radionuclide subject to the impact of uncertainties in diffusion and sorption, only uncertainty in diffusion is addressed here. The reason for this approach is that 239Pu exhibits sorbing behavior that is so exceptionally strong that it would be physically unrealistic to reduce Kd to levels that may have an impact on transport. Even order-of-magnitude reduction in Kd is insufficient to significantly affect 239Pu transport behavior. The study of uncertainties in diffusion follows the pattern discussed in Sections 6.8.3 and 6.9.3. Uncertainties in matrix diffusion are reflected in the values of the diffusion coefficient D0. We investigated the sensitivity of the 239Pu transport through the UZ for the D0 values shown in Table 6.5-2, which cover the possible D0 range. The results are shown in Figure 6.10-3. As expected, D0 has an impact on breakthrough predictions, resulting in faster arrival times for lower D0 values. However, the relative effect of D0 uncertainty in predicting transport of the strongly sorbing 239Pu is comparatively even less pronounced than that of the mildly sorbing 237Np. That is, the impact of D0 uncertainty on radionuclide breakthrough predictions appears to be mitigated by sorption. 1.0 0.8 0.6 0.4 0.2 0.0 Normalized Mass Fraction 10-1 100 101 102 103 104 105 106 Time (years) 239 Pu - Diffusion Coefficients Base case (Do=4.81x10-10 m2/s) Low Do (Do=4.81x10-11 m2/s) High Do (Do=10-9 m2/s) Output-DTN: LB0307MR0060R1.007, data submitted with this Model Report Figure 6.10-4. Effect of Uncertainty in D0 on the 239Pu Breakthrough at the UZ Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 155 November 2003 6.10.3 Applicability of Uncertainty Studies to Other Radionuclides. The insights obtained from the analysis of uncertainty in the cases of 99Tc, 237Np, and 239Pu are directly applicable to all the other radionuclides studied in this Model Report. By covering the spectrum of sorption behavior and diffusion in the uncertainty study, it is possible to develop general rules to broadly predict the effect of uncertainty. Thus, the effect of uncertainties in the breakthrough of nonsorbing radionuclides will be akin to that of 99Tc. Mildly sorbing radionuclides can be described by using the 237Np analog, while strong sorbers will be described by the 239Pu behavior. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 156 November 2003 6.11 THREE DIMENSIONAL TRANSPORT OF 233U AND 235U – INSTANTANEOUS RELEASE Here we study the transport of 233U and 235U through the Yucca Mountain UZ system after an instantaneous release scenario (Section 6.7). The two isotopes share the same physical properties (e.g., diffusion and sorption behavior), but differ in their half-lives (Table 6.5-2). The grids, flow fields, and conditions are as discussed in Section 6.7. The parameters used in these T2R3D simulations of 3-D transport of the moderately sorbing 233U and 235U are listed in Tables 6.5-1, 6.6-1, 6.7-1, and 6.7-2. The DTN of the input and output files for the nine present-day infiltration scenarios is Input/ Output-DTN: LB0307MR0060R1.001, and Output DTN LB0307MR0060R1.002 (Table 6.7-2). 6.11.1 Transport of 233 U Under Different Climatic Conditions The breakthrough curves of 233U (a mildly sorbing species; see Table 6.5-1) transport through the UZ are shown in Figure 6.11-1(a), (b), and (c), for mean present-day, monsoon and glacial infiltration regimes, respectively. A comparison of these figures to those of 237Np (Figure 6.9-1 through 6.9-3) indicates that their behavior is analogous at early times, but deviates at later times because of the larger T1/2 of 233U (leading to different maximum attainable RM values). The t10 and t50 of 233U for the various infiltration regimes are listed in Table 6.20-1. 6.11.2 Transport of 235U Under Different Climatic Conditions The breakthrough curves of 235U transport through the UZ are shown in Figure 6.11-2(a), (b), and (c), for mean present-day, monsoon and glacial infiltration regimes, respectively. Compared to the equivalent figures for 233U, the breakthrough curves show differences at later times (e.g., in the maximum attainable RM level), reflecting the differences in T1/2 (Table 6.5-1). Because of the isotopic relationship to 233U, the behavior 235U (a mildly sorbing species, see Table 6.5-2) is analogous to that 237Np. The t10 and t50 of 235U for the various infiltration regimes are listed in Table 6.20-1. A review of the figures in this section confirms the well-established pattern observed thus far: increasing infiltration enhances transport by resulting in a shorter travel time through the UZ and leads to a higher maximum attainable RM. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 157 November 2003 (a) 1.0 0.8 0.6 0.4 0.2 0.0 Normalized Mass Fraction 10-1 100 101 102 103 104 105 106 Time (years) 233 U - Present-Day Infiltration Lower Mean Upper (b) 1.0 0.8 0.6 0.4 0.2 0.0 Normalized Mass Fraction 10-1 100 101 102 103 104 105 106 Time (years) 233 U - Monsoon Infiltration Lower Mean Upper (c) 1.0 0.8 0.6 0.4 0.2 0.0 Normalized Mass Fraction 10-1 100 101 102 103 104 105 106 Time (years) 233 U - Glacial Infiltration Lower Mean Upper Output-DTN: LB0307MR0060R1.007, data submitted with this Model Report Figure 6.11-1. Cumulative Breakthrough of the 233U Mass Fraction RM at the Water Table for (a) Present-Day, (b) Monsoon and (c) Glacial Infiltration Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 158 November 2003 (a) 1.0 0.8 0.6 0.4 0.2 0.0 Normalized Mass Fraction 10-1 100 101 102 103 104 105 106 Time (years) 235 U - Present Infiltration Lower Mean Upper (b) 1.0 0.8 0.6 0.4 0.2 0.0 Normalized Mass Fraction 10-1 100 101 102 103 104 105 106 Time (years) 235 U - Monsoon Infiltration Lower Mean Upper (c) 1.0 0.8 0.6 0.4 0.2 0.0 Normalized Mass Fraction 10-1 100 101 102 103 104 105 106 Time (years) 235 U - Glacial Infiltration Lower Mean Upper Output-DTN: LB0307MR0060R1.007, data submitted with this Model Report Figure 6.11-2. Cumulative Breakthrough of the 235U Mass Fraction RM at the Water Table for (a) Present-Day, (b) Monsoon and (c) Glacial Infiltration Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 159 November 2003 6.12 THREE DIMENSIONAL TRANSPORT OF 241Am AND 90Sr – INSTANTANEOUS RELEASE In this section, we analyze the transport of 241Am and 90Sr through the Yucca Mountain UZ after an instantaneous release scenario (Section 6.7). Both 241Am and 90Sr are very strong sorbers (even stronger than Pu; see Table 6.5-1), but have the two shortest T1/2 among the radionuclides studied in this Model Report (Table 6.5-2). The grids, flow fields, and conditions are as discussed in Section 6.7.2. The parameters used in these T2R3D simulations are listed in Tables 6.5-1, 6.6-1, 6.7-1, and 6.7-2. The DTN of the input and output files for the nine infiltration scenarios is Input/Output-DTN: LB0307MR0060R1.001 (Table 6.7-2). 6.12.1 241Am Transport The breakthrough curves of 241Am transport through the UZ are shown in Figure 6.12-1(a), (b) and (c), for mean present day, monsoon, and glacial infiltration regimes, respectively. These figures indicate that the maximum attainable RM is quite low, and that, although an increasing infiltration results invariably in a larger RM, the increase is not significant. As in all radionuclides in this Model Report, the breakthrough curves are characterized by early arrival at the water table and an early single plateau at a low RM level (below 0.2), denoting fracture transport. A second plateau (denoting matrix transport, as observed in the case of the 233Pu transport) does not materialize because of (a) the much stronger sorption of 241Am onto the matrix (thus reducing the radionuclide available for matrix transport) and (b) its rather short halflife (4.322×102 years, Table 6.5-2). The t10 and t50 of 241Am for the various infiltration regimes are listed in Table 6.20-1. 6.12.2 90Sr Transport The breakthrough curves of 90Sr transport through the UZ are shown in Figures 6.12-2(a), (b) and (c), for mean present day, monsoon, and glacial infiltration regimes, respectively. As expected because of the similarity to 241Am, the maximum attainable RM is quite low and does not increase significantly with an increasing infiltration. The breakthrough curves are characterized by early arrival at the water table and an early single plateau at a low RM level (below 0.15), denoting fracture transport. The second plateau (denoting matrix transport) observed in the case of 233Pu transport (Section 6.10) does not materialize because of (a) the much stronger sorption of 90Sr onto the matrix (thus reducing the radionuclide available for matrix transport) and (b) its very short half-life (2.9×101 years, Table 6.5-2). The t10 and t50 of 90Sr for the various infiltration regimes are listed in Table 6.20-1. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 160 November 2003 (a) 0.20 0.15 0.10 0.05 0.00 Normalized Mass Fraction 10-1 100 101 102 103 104 Time (years) 241Am - Present-Day Infiltration Lower Mean Upper (b) 0.20 0.15 0.10 0.05 0.00 Normalized Mass Fraction 10-1 100 101 102 103 104 Time (years) 241Am - Monsoon Infiltration Lower Mean Upper (c) 0.20 0.15 0.10 0.05 0.00 Normalized Mass Fraction 10-1 100 101 102 103 104 Time (years) 241Am - Glacial Infiltration Lower Mean Upper Output-DTN: LB0307MR0060R1.007, data submitted with this Model Report Figure 6.12-1. Cumulative Breakthrough of the 241Am Mass Fraction RM at the Water Table for (a) Present-day, (b) Monsoon and (c) Glacial Infiltration Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 161 November 2003 (a) 0.20 0.15 0.10 0.05 0.00 Normalized Mass Fraction 10-1 100 101 102 103 104 105 Time (years) 90 Sr - Present-Day Infiltration Lower Mean Upper (b) 0.20 0.15 0.10 0.05 0.00 Normalized Mass Fraction 10-1 100 101 102 103 104 105 Time (years) 90 Sr - Monsoon Infiltration Lower Mean Upper (c) 0.20 0.15 0.10 0.05 0.00 Normalized Mass Fraction 10-1 100 101 102 103 104 105 Time (years) 90 Sr - Glacial Infiltration Lower Mean Upper Output-DTN: LB0307MR0060R1.007, data submitted with this Model Report Figure 6.12-2. Cumulative Breakthrough of the 90Sr Mass Fraction RM at the Water Table for (a) Present-Day, (b) Monsoon and (c) Glacial Infiltration Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 162 November 2003 6.13 THREE DIMENSIONAL TRANSPORT OF 135Cs – INSTANTANEOUS RELEASE In this section, we analyze the transport of 135Cs through the Yucca Mountain UZ following an instantaneous release event. 135Cs is a very strong sorber (even stronger than Pu on zeolitic rocks, but less on vitric and devitrified; see Table 6.5-1), but is also among the most diffusive species (Table 6.5-2). The grids, flow fields, and conditions are as discussed in Section 6.7. The parameters used in the T2R3D simulations of 3-D transport of 135Cs are listed in Tables 6.5-1, 6.6-1, 6.7-1, and 6.7-2. The DTN of the input and output files for all nine scenarios is Input/Output-DTN: LB0307MR0060R1.001 (Table 6.7-2). The breakthrough curves of 135Cs transport through the UZ are shown in Figures 6.13-1(a), (b) and (c), for mean present-day, monsoon, and glacial infiltration regimes, respectively. In all climatic scenarios, the very strong sorption of 135Cs demonstrates itself by the very low RM level, which persists for a very long time. The very strong diffusion of 135Cs into the matrix prevents its early arrival at the water table through advective transport in the fractures, taking about 1,000 years for concentrations at the water table to register an upward trend. After this initial delay (during which the effect of the infiltration rates do not substantially affect transport), significant RM values are attainable, and breakthrough begins to be affected by the climatic regime. The t10 and t50 of 135Cs for the various infiltration regimes are listed in Table 6.20-1. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 163 November 2003 (a) 1.0 0.8 0.6 0.4 0.2 0.0 Normalized Mass Fraction 10-1 100 101 102 103 104 105 Time (years) 135 Cs - Present-day Infiltration Lower Mean Upper (b) 1.0 0.8 0.6 0.4 0.2 0.0 Normalized Mass Fraction 10-1 100 101 102 103 104 105 Time (years) 135 Cs - Monsoon Infiltration Lower Mean Upper (c) 1.0 0.8 0.6 0.4 0.2 0.0 Normalized Mass Fraction 10-1 100 101 102 103 104 105 Time (years) 135 Cs - Glacial Infiltration Lower Mean Upper Output-DTN: LB0307MR0060R1.007, data submitted with this Model Report Figure 6.13-1. Cumulative Breakthrough of the 135Cs Mass Fraction RM at the Water Table for (a) Present-Day, (b) Monsoon and (c) Glacial Infiltration Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 164 November 2003 6.14 THREE DIMENSIONAL TRANSPORT OF 226Ra, 229Th AND 231Pa– INSTANTANEOUS RELEASE In this section, we analyze the transport 226Ra, 229Th and 231Pa, three radionuclides that are among the most insoluble species and the strongest known sorbers. These are instantaneously released at the repository level, and their migration through the UZ zone and breakthrough at the water table is studied. These three radionuclides have medium half-lives, ranging from a few thousand to a few tens of thousands of years (Table 6.5-2). The grids, flow fields, and conditions are as discussed in Section 6.7. The parameters used in these T2R3D 3-D simulations are listed in Tables 6.5-1, 6.6-1, 6.7-1, and 6.7-2. The DTN of the input and output files for the all nine scenarios is Input/Output-DTN: LB0307MR0060R1.001. Breakthrough curves of 226Ra transport through the UZ are shown in Figures 6.14-1(a), (b), and (c), for mean present day, monsoon, and glacial infiltration regimes, respectively. The corresponding figures for the 229Th breakthrough are 6.14-2(a), (b), and (c), and those for 231Pa are 6.14-3(a), (b), and (c), respectively. As expected, these three radionuclides exhibit very similar behavior, and similar to those of 241Am and 90Sr. Their maximum attainable RM is quite low, and although larger infiltration results invariably in a larger RM, the increase is not significant. Breakthrough curves are characterized by early arrival at the water table and an early plateau at a low RM level (below 0.2), denoting fracture transport. The second plateau (denoting matrix transport) that was observed in the case of the 239Pu transport does not materialize because of (a) their much stronger sorption onto the matrix that does not permit transport through it, and (b) their medium half-life (dwarfed by the 1,000,000-year-long period of observation depicted in the figures; see Table 6.5-2). The t10 and t50 of 226Ra, 229Th, and 231Pa for the various infiltration regimes are listed in Table 6.20-1. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 165 November 2003 (a) 0.20 0.15 0.10 0.05 0.00 Normalized Mass Fraction 10-1 100 101 102 103 104 105 Time (years) 226 Ra - Present-Day Infiltration Lower Mean Upper (b) 0.20 0.15 0.10 0.05 0.00 Normalized Mass Fraction 10-1 100 101 102 103 104 105 Time (years) 226 Ra - Monsoon Infiltration Lower Mean Upper (c) 0.20 0.15 0.10 0.05 0.00 Normalized Mass Fraction 10-1 100 101 102 103 104 105 Time (years) 226 Ra - Glacial Infiltration Lower Mean Upper Output-DTN: LB0307MR0060R1.007, data submitted with this Model Report Figure 6.14-1. Cumulative Breakthrough of the 226Ra Mass Fraction RM at the Water Table for (a) Present-Day, (b) Monsoon and (c) Glacial Infiltration Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 166 November 2003 (a) 0.20 0.15 0.10 0.05 0.00 Normalized Mass Fraction 10-1 100 101 102 103 104 105 106 Time (years) 229 Th - Present-Day Infiltration Lower Mean Upper (b) 0.20 0.15 0.10 0.05 0.00 Normalized Mass Fraction 10-1 100 101 102 103 104 105 106 Time (years) 229 Th - Monsoon Infiltration Lower Mean Upper (c) 0.20 0.15 0.10 0.05 0.00 Normalized Mass Fraction 10-1 100 101 102 103 104 105 106 Time (years) 229 Th - Glacial Infiltration Lower Mean Upper Output-DTN: LB0307MR0060R1.007, data submitted with this Model Report Figure 6.14-2. Cumulative Breakthrough of the 229Th Mass Fraction RM at the Water Table for (a) Present-Day, (b) Monsoon and (c) Glacial Infiltration. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 167 November 2003 (a) 0.20 0.15 0.10 0.05 0.00 Normalized Mass Fraction 10-1 100 101 102 103 104 105 106 Time (years) 231 Pa - Present-Day Infiltration Lower Mean Upper (b) 0.20 0.15 0.10 0.05 0.00 Normalized Mass Fraction 10-1 100 101 102 103 104 105 106 Time (years) 231 Pa - Monsoon Infiltration Lower Mean Upper (c) 0.20 0.15 0.10 0.05 0.00 Normalized Mass Fraction 10-1 100 101 102 103 104 105 106 Time (years) 231 Pa - Glacial Infiltration Lower Mean Upper Output-DTN: LB0307MR0060R1.007, data submitted with this Model Report Figure 6.14-3. Cumulative Breakthrough of the 231Pa Mass Fraction RM at the Water Table for (a) Present-Day, (b) Monsoon and (c) Glacial Infiltration Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 168 November 2003 6.15 THREE DIMENSIONAL SIMULATIONS OF 99Tc, 237Np AND 239Pu TRANSPORT - CONTINUOUS RELEASE In this section, we investigate the three radionuclides discussed in Sections 6.8 to 6.10 under a scenario of continuous release. 6.15.1 The Continuous-Release Scenario The concept of the continuous-release scenario investigated in these simulations was discussed in Section 6.7.3.2. Only mean present-day infiltration was considered in these simulations. The grids, flow fields, and conditions are as discussed in Section 6.7. The parameters used in these EOS9nT simulations of 3-D transport of 99Tc, 237Np, and 239Pu are listed in Tables 6.5-1, 6.6-1, 6.7-1, and 6.7-2. The DTN of the input and output files for the three present-day infiltration scenarios is Input/Output-DTN: LB0307MR0060R1.003 (Table 6.7-2). 6.15.2 The Definition of the Breakthrough Curves for Continuous Release As the radionuclide concentration in the repository and in the bottom boundary elements (corresponding to the water table) remain constant over time, the breakthrough concept is based on the normalized or relative release rate RF, which is defined as RF = ] [ repository proposed at the rate release Mass ] [ boundary r groundwate at the rate release Mass 1 - -1 MT MT Thus, RF involves a ratio of rates, while the conventional RM is a ratio of masses. As in RM, t10 and t50 are defined as the times at which RF = 0.1 and RF = 0.5, respectively. For radioactive chains, the quantity in the denominator reflects the sum of the fluxes of all members in the release stream. For decaying release concentration, the denominator reflects the original radionuclide release rate from repository elements. Of particular interest is the distribution of the relative sorbed or filtered concentration FR [ML-3] defined as FR = F X0 where F is as defined in Equation 6-8, and X0 is the species mass fraction in the water released from the repository. This is important in the evaluation of sorbing radionuclides such as 237Np and 239Pu. 6.15.3 Transport of Continuously Released 99Tc, 237Np, and 239Pu Under Present-Day Infiltration The breakthrough curves for continuously released 99Tc, 237Np and 239Pu are shown in Figure 6.15-1, which also shows the decaying fluxes (reflecting decaying concentrations, as the water fluxes remain time-invariant) for reference. The t10 and t50 of 99Tc, 237Np and 239Pu under a continuous release scenario for a mean present-day infiltration are listed in Table 6.20-2. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 169 November 2003 Comparison of the t10 and t50 from Figure 6.15-1 to those from the instantaneous-release scenarios (Sections 6.8, 6.9, and 6.10) shows that, in this case, they are generally consistent. Note, however, that in general, the two can be very different. Of interest is the shape of the RF curves, which begin to show a decline after reaching a maximum before the point where of the influx and outflow curves coincide. The existence of such a maximum is a direct consequence of considering a decaying source. RF, the ratio of outflow to initial inflow, initially increases as radionuclides reach the water table in increasing quantities. However, the increase can only proceed up to a point, beyond which there is a balance between influx and outflow (indicating by the coincidence of the two curves in Figure 6.15-1), which involves ever-declining radionuclide concentrations. 6.15.4 Transport Patterns of Continuously Released 99Tc, 237Np, and 239Pu Under Present-Day Infiltration The distribution of 99Tc fracture and matrix XR in tsw39 and at the watertable at various times after the radionuclide release are shown in Attachment V. Under continuous release, it is expected that the radionuclide distributions at early times would be similar to those from instantaneous release, but the two would diverge at later times because of the different masses involved in each scenario. 1.0 0.8 0.6 0.4 0.2 0.0 Relative release rate (current/initial at the repository) 101 102 103 104 105 106 Time (years) WATER TABLE 99Tc 237Np 239Pu REPOSITORY 99Tc 237Np 239Pu 0.1 fraction line 0.5 fraction line Output-DTN: LB0307MR0060R1.007, data submitted with this Model Report Figure 6.15-1. Normalized Relative Release RF of 99Tc, 237Np and 239Pu at the Water Table for Continuous Release and Mean Present-Day Climatic Scenarios Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 170 November 2003 One of the objectives of this study was to determine whether transport under continuous release is dominated by different geologic features than those controlling instantaneous release. Comparison of Figures V.1 through V.12 to Figures 6.8.2 to 6.8-11 indicate that at early times (t = 100 years) the concentration distributions in tsw39 show the same patterns and are in line with expectations. The magnitude of concentrations is different, but this was entirely expected because the total radionuclide mass in the UZ is substantially larger in a continuous release scenario. Differences are observed for t = 1,000 years, which show increasing fracture and matrix concentrations over time. The Drillhole Wash and Pagany Wash faults are the main geologic features responsible for fast transport to the bottom of the TSw and the water table. Note that the shift of the matrix concentrations from underneath the southern part toward the northern part of the repository (Figures 6.8-7, 6.8-9, 6.8-11) does not occur under continuous release, and the southern part continues to exhibit the largest concentrations. The areal extent of radionuclide distribution in both the fractures and the matrix appears to reach a maximum at t = 1,000 years (Figures V.7 and V.8). Although these footprints appear constant after this time, the concentrations change over time and reach a maximum at t = 10,000 years (Figures V.9 and V.10), after which time they decline because of radioactive decay. An important observation is the large footprint and the corresponding high values of the matrix XR under continuous release. Thus, the amount of 99Tc in the mobile liquid in the matrix is larger and over a larger area under continuous release. Comparison of Figures V.11 through V.24 to Figures 6.8-12 to 6.8-21 indicates that the concentration distributions above the water table show the same general patterns. For obvious reasons, the magnitude of concentrations is different, but the areal distributions are remarkably similar. Significant differences are observed at t = 100,000 years, when both the areal extent and the magnitude of the 99Tc concentrations are significantly larger under continuous release. This is entirely consistent with the much larger total radionuclide mass released in the UZ. The corresponding distributions of 237Np and 239Pu, respectively, in tsw39 and at the water table under continuous release for a mean present-day infiltration are included in Output-DTN: LB0307MR0060R1.004, which also include figures with the FR distributions. Comparison to the figures for 237Np and 239Pu to those for 99Tc in Attachment V indicate that the same general patterns persist. The differences in the magnitude and areal extent of concentrations are mainly due to the different sorbing behavior of the radionuclides. Thus, with an increasing sorption affinity to the UZ rocks from 99Tc to 237Np to 239Pu, the areal extent of the radionuclide occurrence decreases and the corresponding maximum XR in the fractures decrease, the maximum XR in the matrix decreases even more severely, and the matrix XR increases. Note that at t = 100,000 years, significant amounts of 237Np and 239Pu are sorbed over areas that exceed the footprint of the repository at both the tsw39 level and at the water table. The implication is that these sorbed radionuclides can become sources for long-term release to the water table even if releases from the repository cease. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 171 November 2003 6.16 THREE DIMENSIONAL TRANSPORT OF THE 239Pu 235U 231Pa CHAIN – CONTINUOUS RELEASE In this section we investigate the transport of the three-member chain 239Pu 235U 231Pa chain continuously released from the repository. We study the transport of the members of this chain by means of EOS9nT 3-D site-scale simulations of the entire UZ system of Yucca Mountain. Only mean present-day infiltration was considered in these simulations. The grids, flow fields, and conditions are as discussed in Section 6.7, and the parameters used in these EOS9nT simulations are listed in Tables 6.5-1, 6.6-1, 6.7-1, and 6.7-2. The DTN of the input and output files for the three present-day infiltration scenarios is Input/Output-DTN: LB0307MR0060R1.003 (Table 6.7-2). The evolution of the mass fractions RF in the release stream at the repository and at the water table are shown in Figures 6.16-1(a) and (b). Figure 6.16-1(a) reflects decay and generation of each of the members of the 239Pu chain, and, being a boundary condition, is unaffected by flow or transport conditions in the UZ. Note that the released stream becomes depleted in 239Pu due to decay after about 100,000 years, at which time 235U is practically the only species entering the UZ from the repository. The mass fractions at the water table in Figure 6.16-1(b) incorporate the cumulative effects of migration through the UZ, and reflect the effects of matrix diffusion and sorption on the effluent concentration. Comparison of Figures 6.16-1(a) and (b) reveals a very important difference. The radionuclide stream at the water table shows a much earlier enrichment in 235U, and 235U is the increasingly dominant species after t=2000 years. The reason for the difference is the much stronger sorption of 239Pu in the UZ, coupled with the relatively weak sorption of 235U. The contribution of 231Pa appears practically insignificant in the first 1,000,000 years because of the very long half-life of 235U and its very strong affinity for sorption (Section 6.11). Note that Figures 6.16-1(a) and (b) reflect relative concentrations of radionuclides in the inflow and outflow streams, but provide no information on the actual inflow and outflow relationship. This is addressed by the relative release rates at the water table in Figure 6.16-2, computed as the ratio of the total outflow rates (i.e., sum of individual member fluxes) versus the total inflow rates at the repository. A comparison of Figure 6.16-2 to that for the 239Pu in Figure 6.16-1(a) (obtained under the same conditions and in the same domain) indicates the importance of considering all the members of the chain in the computations, and the significant error that will be introduced if only the parents are considered. The t10 and t50 of this chain under a continuous release scenario for a mean present-day infiltration are listed in Table 6.20-2. Given the long half-life of 239Pu and the much longer half-life of 235U (see Table 6.5-2), the daughter contributions are very important and cannot be neglected. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 172 November 2003 (a) 10-8 10-7 10-6 10-5 10-4 10-3 10-2 10-1 100 Mass fraction in release stream at repository 101 102 103 104 105 106 Time (years) 239Pu 235U 231Pa (b) 10-8 10-7 10-6 10-5 10-4 10-3 10-2 10-1 100 Mass fraction in release stream at the water table 101 102 103 104 105 106 Time (years) 239Pu 235U 231Pa Output-DTN: LB0307MR0060R1.007, data submitted with this Model Report Figure 6.16-1. Mass Fractions of Each Member of the 239Pu Chain in the Release Stream at (a) the Repository and (b) the Water Table – Continuous Release Mean Present-Day Climatic Scenarios Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 173 November 2003 1.0 0.8 0.6 0.4 0.2 0.0 Normalized release rate at the water table 101 102 103 104 105 106 Time (years) 0.1 fraction line 0.5 fraction line Output-DTN: LB0307MR0060R1.007, data submitted with this Model Report Figure 6.16-2. Normalized Relative Release RF of the Sum of All Members of the 239 Pu Chain at the Water Table – Continuous Release Mean Present-Day Climatic Scenarios Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 174 November 2003 6.17 THREE DIMENSIONAL TRANSPORT OF THE 241Am 237Np 233U 229Th CHAIN – CONTINUOUS RELEASE In this section, we investigate the transport of the four-member 241Am 237Np 233U 229Th chain continuously released from the repository. Each member of the chain has its own distinct transport behavior. Only mean present-day infiltration was considered in these simulations. The grids, flow fields, and conditions are as discussed in Section 6.7. The parameters used in these EOS9nT simulations are listed in Tables 6.5-1, 6.6-1, 6.7-1, and 6.7-2. The DTN of the input and output files for the three present-day infiltration scenarios is Input/Output-DTN: LB0307MR0060R1.003 (Table 6.7-2). The evolution of the mass fractions RF in the release stream at the repository and at the water table is shown in Figures 6.17-1(a) and (b). Figure 6.17-1(a) is a straightforward representation of decay and generation for each member of the 241Am chain, and is unaffected by flow or transport conditions in the UZ because it reflects boundary conditions. Note that, because of the short half-life of 241Am, its concentration in the release stream falls to practically zero after about 3,000 years, at which time 237Np is overwhelmingly the dominant species leaving the repository. The mass fractions at the water table in Figure 6.17-1(b) reflect the cumulative effects of passage through the UZ, i.e., the effects of matrix diffusion and sorption on the radionuclide concentration. Comparison of Figures 6.17-1(a) and (b) reveals that, as in the 239Pu chain case, the effluent at the water table is enriched in 237Np much earlier (about t=500 years). This behavior is caused by the much stronger sorption of 241Am in the UZ, coupled with the relatively weak sorption of 237Np (Table 6.5-1 and Section 6.9). The contribution of 233U exceeds only 1% after about 1,000,000 years, while the long cumulative half-lives of its parents make the contribution of 229Th negligible. The relative release rate at the water table is computed as the ratio of the total outflow rates (i.e., sum of individual member fluxes) versus the total inflow rates at the repository, and is shown in Figure 6.17-2. For the reasons discussed in Section 6.15, it is important to consider all the members of the chain in the computations of transport. Given the long half-life of 237Np (much longer than that of its 241Am parent; see Table 6.5-2), the daughter contributions cannot be neglected with impunity. The daughter contributions are even more important under wetter infiltration scenarios. The t10 and t50 of this chain under a continuous release scenario for a mean present-day infiltration are listed in Table 6.20-2. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 175 November 2003 (a) 10-6 10-5 10-4 10-3 10-2 10-1 100 Mass fraction in the release stream at the repository 101 102 103 104 105 Time (years) 241Am 237Np 233U 229Th (b) 10-6 10-5 10-4 10-3 10-2 10-1 100 Mass fraction in the release stream at the water table 101 102 103 104 105 Time (years) 241Am 237Np 233U 229Th Output-DTN: LB0307MR0060R1.007, data submitted with this Model Report Figure 6.17-1. Mass Fractions of Each Member of the 241Am Chain in the Release Stream at (a) the Repository and (b) the Water Table – Continuous Release Mean Present-Day Climatic Scenarios Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 176 November 2003 1.0 0.8 0.6 0.4 0.2 0.0 Normalized release rate at the water table 101 102 103 104 105 Time (years) 0.1 fraction line 0.5 fraction line Output-DTN: LB0307MR0060R1.007, data submitted with this Model Report Figure 6.17-2. Normalized Relative Release RF of the Sum of All Members of the 241Am Chain at the Water Table – Continuous Release Mean Present-Day Climatic Scenario Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 177 November 2003 6.18 THREE DIMENSIONAL SITE-SCALE TRANSPORT OF PuO2 COLLOIDS – CONTINUOUS RELEASE In this section, we study the large-scale 3-D site-scale transport of true colloids in the Yucca Mountain UZ for mean present-day infiltration. The flow field and the 3-D grid are identical to the ones discussed in Sections 6.14 to 6.17. Radioactive colloids undergoing decay at the source were continuously released into the fractures of the elements corresponding to the repository. Section 6.18 includes eight subsections. In Section 6.18.1, we discuss the colloidal forms and properties. In Section 6.18.2 we present the mathematical model of filtration used in EOS9nT and discuss its parameters for the four colloids considered in the simulations. In Section 6.18.3, we describe the four cases of colloid transport investigated in this Model Report, and in Sections 6.18.4 to 6.18.8 we discuss the simulation results for each case. 6.18.1 Colloidal Forms and Properties Two classes of colloids, Class I and Class II, were considered in this study (see Section 6.1.3.3.1). Class I includes true colloids and waste form colloids, in which the entire nonaqueous component of the colloid is composed of the radioactive substance. Those were taken to have the properties of PuO2 and are subject to radioactive decay. Additionally, the Class I colloid size and density were considered invariable during the simulation. This results in more conservative predictions because (a) matrix diffusion does not increase in response to a smaller colloid diameter, and (b) colloid filtration (the parameters of which are based on the actinide mass per unit volume) does not decrease due to lower concentrations. Class II colloids (i.e., pseudocolloids onto which radionuclides have sorbed irreversibly) rather than Class III colloids were considered because of the extraordinarily strong sorption of radionuclides on natural colloids (BSC 2003 [161620], Section 6.3.3.1). Computation of the transport of Class II colloids involves solution of the transport equation of the underlying nonradioactive pseudocolloid (i.e., that of the corresponding Class IV colloid -- see Section 6.1.3.3.1, this Model Report). Then the radioactivity concentration is computed from Cr = CcXr0 ft (Eq. 6-31) where Cr is the actinide concentration [ML-3], Cc is the underlying non-radioactive colloid particle (Class IV colloid) concentration [ML-3], Xr0 is the initial mass fraction (i.e., at the time of release) of the irreversibly sorbed radionuclide in the colloidal particle, and ft is the remaining radionuclide mass fraction at time t. For a parent radionuclide, ft = exp(- .t). In this study, the irreversibly sorbed radionuclide in the Class II colloids was assumed to be 239Pu. In both colloidal classes considered in this study, four different colloid sizes were considered. Their sizes and their accessibility factors were calculated based on DTNs: GS950608312231.008 [144662], and GS980908312242.039 [145272], as reported in CRWMS M&O (2000 [122799], Section 6.2.1) and are shown in Table 6.18-1. The parameters used in these EOS9nT simulations of 3-D transport of the colloids are listed in Table 6.7-1, 6.7-2, 6.18- 1, and 6.18-2. The DTN of the input and output files for the three present-day infiltration scenarios is Input/Output-DTN: LB0307MR0060R1.003 (Table 6.7-2). Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 178 November 2003 6.18.2 Colloid Filtration Model and Coefficients Pore-size exclusion (straining filtration) was described using the accessibility factors shown in Table 6.18-1. The linear kinetic model of colloid filtration described by Equation 6-15 was used in the simulations. The forward kinetic coefficient .+ was computed from Equation 6-16, in which the velocity coefficients are as shown in Table 6.18-1, and the filter coefficient e is computed using Equations 2 and 6 of Harvey and Garabedian (1991 [109256]). In our nomenclature, this is e = 1.5 1 - f dm ac .c (Eq. 6-32) where dm is the particle size of the medium grains or the fracture aperture [L], ac is the collision efficiency factor and, and . is the single collector efficiency, and u gd d d u d d T k c c m c m c w B c µ . - . + . .. . . .. . + . .. . . .. . µ = . 18 ) ( 5 . 1 9 . 0 2 2 3 / 2 (Eq. 6-33) in which kB is the Boltzmann constant, dc is the colloid diameter [L], T is the absolute temperature, and all other terms remain as previously defined. The clogging (forward) kinetic coefficients .+ are computed internally in EOS9nT. No information exists on the kinetic declogging (reverse) coefficient .-. To alleviate the problem, .- in the EOS9nT coefficients was entered as a fraction of .+ (see Table 5-1). We believe that this is a sounder approach than the constant .- employed in CRWMS M&O (2000 [122799], Section 6) because it maintains dependence on the flow velocity. Because the kinetic coefficient .+ is a linear function of the flow velocity, dependence of .- on velocity is conceptually sounder than the constant value approach. Table 6.18-1. Properties of the Four Colloids in the EOS9nT Simulations Parameter 450 nm 200 nm 100 nm 6 nm Diffusion coefficient D0 (m2/s) * 9.53×10-13 2.15×10-12 4.29×10-12 7.15×10-11 Velocity adjustment factor f † 1.5 1.2 1.1 1.0 Accessibility factor f c in the TSw ‡ 0.05 0.10 0.20 0.65 Accessibility factor f c in the CHv ‡ 0.45 0.50 0.55 0.80a Accessibility factor f c in the CHz ‡ 0.20 0.25 0.25 0.65 NOTES: *: From Grathwohl 2000 [141512] †: Reasonable estimates, see Table 5-1 ‡: DTN: GS950608312231.008 [144662], GS980908312242.039 [145272], CRWMS M&O (2000 [122799], Section 6.2.1) a: Although a value of 0.85 was reported in CRWMS M&O [122799], the 0.8 value fits better the lognormal distribution of pore size exclusion. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 179 November 2003 Table 6.18-2. Input Parameters for the EOS9nT 3-D Site-scale Simulations of Colloid Transport (# 1 perched-water model, mean present-day infiltration) Parameters Source aTortuosity t ~ f Farrell and Reinhard (1994 [122803], p. 64) Properties and characteristics of the geologic units, steady-state pressures, water saturations and flow fields DTN: LB0210THRMLPRP.001 [160799], LB03023DSSCP9I.001 [163044] (Table 6.8-1) Colloid density .c = 11,640 kg/m3 (PuO2) Lide (1993 [123032], p. 4-83) Forward kinetic filtration (clogging) coefficient .+: from Equations 6-6, 6-32, and 6-33, parameters computed at 25 o C Harvey and Garabedian 1991 [109256], p. 180, Equation 6 NOTE: a see Section 5, Table 5-1 for the assumption 6.18.3 The Class I Colloid Transport Simulation Cases The following four cases were investigated: Case 1. .- = 0. This corresponds to a case of no declogging, in which colloids, once filtered, do not detach from the pore/fracture walls. Case 2. .-/.+ = 100. This corresponds to strong kinetic declogging and will provide an estimate of maximum colloidal transport. Case 3. .-/.+ = 0.1. This corresponds to weak kinetic declogging and approaches equilibrium filtration behavior. Case 4. Same as in Case 2, but the fractures are assumed to have the same colloidal transport properties as the corresponding matrix. This study provides an estimate of the importance of fractures in the transport of colloids. Note that the change in the filtration properties affects only 1% of the fracture pore volume (because f = 0.99 in the fractures; see Equation 6-30). There is no filtration (attachment) of colloids onto the fracture walls in Cases 1 to 3. The t10 and t50 of the 239PuO2 colloid transport under a continuous release scenario and for a mean presentday infiltration are listed in Table 6.20-3. 6.18.4 Colloid Transport Case 1 (Class I) This simulation assumes that no declogging occurs once colloids are attached to the matrix. Figure 6.18-1(a) shows the relative release rate RF of the four colloid sizes at the water table. For reference, the decaying release rate of the colloids at the repository (the same for all 239Pu colloids) is included in the figure. Two observations appear particularly important. The first is the very fast breakthrough of the larger colloids (characterized by a rapid rise of the breakthrough curves). The fast breakthrough is about the same for all the larger colloids. The Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 180 November 2003 t10 for the three larger colloids is only about 5 years. After the fast rise, t50 is not reached for any colloid, and the RF begins declining as decay affects concentrations. The second observation is that, contrary to expectations, the smallest colloid (6 nm) exhibits the slowest breakthrough. It is really noteworthy that this colloid never reaches the 0.1 fraction level, indicating that it is sufficiently small to enter the matrix and get attached to the fracture walls. This behavior results from a combination of the following factors: 1. The faster transport velocity of larger colloids, which, by virtue of their size, can only move in the center of pores/fractures where velocities are larger than the average water velocity. In this case, the 450 nm colloid moves at 1.5 times the water velocity (i.e., f=1.5; see Equation 6-16 in Section 6.2.4.1). 2. Larger colloids are unable to penetrate the matrix from the fractures because of size exclusion. Thus, the colloid mass in the fractures is not reduced through colloidal diffusion and/or hydrodynamic dispersion, and practically all of it moves exclusively in the fractures. The 6 nm colloid can diffuse into the matrix, a process which in Figure 6.18-1(a) manifests itself by the substantially slower breakthrough, the very slow and mild rise of the curve, and the low maximum RF value. After the initial rapid rise, the slope of the RF curve in the larger colloids becomes much milder, and is then followed by a short upward trend again, before it begins to decline at a point that coincides with the beginning of the rapid decline in concentration due to decay. This second upward trend is associated with the beginning of matrix flow of colloids. A similar trend, but much delayed, is observed in the 6 nm colloid. 6.18.5 Colloid Transport Case 2 (Class I) 6.18.5.1 Breakthrough Curves Figure 6.18-1(b) shows the relative release rate RF of the four colloid sizes at the water table. The differences between Case 1 and Case 2 are very small, and are most prominent in the case of the 6 nm colloid. The FR is higher in Case 2 because the few colloids that manage to enter the matrix are not irreversibly attached as in Case 1, but declog rather rapidly. The relative insensitivity to the clogging model in the matrix indicates the dominant role of the fractures in the 3-D site scale system, with the matrix appearing to have a minuscule contribution. Because of the similarity between Figures 6.18-1(a) and (b), the comments made in Case 1 apply here too. 6.18.5.2 Transport Mechanisms and Patterns of the 6 nm Colloid The importance of fractures on the transport of the 6 nm colloid (Co0060) is evident in Figure VI.1 (see Attachment VI), which shows the areal distribution of XR in the aqueous phase in the fractures in the tsw39 layer at t = 10 years. Despite the very early time, the colloid concentration in the fracture is substantial. The reasons for the significant presence (in terms of areal extent and level of concentration) of the 6 nm colloid are that (a) diffusion into the matrix is limited because of its relatively large size (compared to solutes) and pore exclusion, and (b) the tsw39 layer is above the TSw-CHn interface, where significant filtration occurs. Note that the distribution pattern of the fracture XR in Figure VI.1 indicates that the colloids accumulate at the Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 181 November 2003 TSw-CHn interface and, unable to cross it, move along the sloping interface in an easterly direction. This transport is more pronounced in the 6 nm colloid than in the solutes. In contrast to Figure VI.1, the matrix XR in Figure VI.2 shows no difference from the background at the same time. In Figure VI.3, the distribution of matrix FR at t = 10 years shows a sizable presence of the 6 nm colloid filtered onto the matrix of the tsw39 layer. As in the case of sorbing solutes, the area of nonzero FR (Figure VI.3) is larger than that of the fracture XR footprint (Figure VI.3) at the same time, and shows the presence of filtered 6 nm colloids at locations where there appear to be no colloids suspended in the aqueous phase in the fractures. At t = 100 years, the fracture XR in Figure VI.4 is significant over a large area (larger than that for solutes at the same time) and is at its maximum value over a large portion of this area. Although the fracture XR (Figure VI.5) does not register a measurable difference from the UZ background, the matrix FR (Figure VI.6) shows a significant presence of the 6 nm colloid over an area larger than that covered by the fracture XR (Figure VI.4). Note that the matrix FR distribution indicates more filtration in the northern part of the repository. This is consistent with the permeability barriers that result in the perched-water bodies at this location. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 182 November 2003 (a) 1.0 0.8 0.6 0.4 0.2 0.0 Relative release rate (current/initial at the repository) 100 101 102 103 104 105 106 Time (years) REPOSITORY dc = 450 nm dc = 200 nm dc = 100 nm dc = 6 nm WATER TABLE dc = 450 nm dc = 200 nm dc = 100 nm dc = 6 nm Colloid transport Case 1 (b) 1.0 0.8 0.6 0.4 0.2 0.0 Relative release rate (current/initial at the repository) 100 101 102 103 104 105 106 Time (year) REPOSITORY 450 nm 200 nm 100 nm 6 nm WATER TABLE 450 nm 200 nm 100 nm 6 nm Colloid transport Case 2 Output-DTN: LB0307MR0060R1.007, data submitted with this Model Report Figure 6.18-1. Normalized Release at the Water Table in (a) Case 1 and (b) Case 2 of Colloidal Transport for Mean Present-Day Infiltration (All the Release Curves at the Repository are Superimposed). Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 183 November 2003 A comparison of the matrix FR of the 6 nm colloid (Figure VI.6) to that of the dissolved 239 Pu (Figure VIII.3) shows a drastically different pattern. While the matrix FR distribution of 239 Pu shows concentration of sorption in the southern part of the repository, the FR of the 6 nm colloid shows filtration in both the northern and the southern part, i.e., colloids and sorbing radioactive solutes (Sections 6.12 to 6.15) are not similarly affected under the #1 perched-water model. The reason for the different behavior is attributed to (a) the lower diffusion of the colloid into (and the lack of filtration onto) the matrix (thus the faster transport of colloids in larger amounts than solutes to the TSw-CHn interface), (b) the lack of sorption and (c) the combined effects of pore-size exclusion and filtration at the interface that are quantitatively larger than the effects of sorption of the limited dissolved 239Pu amounts. The fracture XR, matrix XR, and matrix FR distributions at t = 1,000 years and t = 10,000 years are shown in Figures IX.7 to IX.12, and they follow the same pattern. The matrix XR registers measurable deviations from background at t = 1,000 years (Figure IX.8), and it shows extensive areas of relatively high concentrations at t = 10,000 years (Figure IX.11). At t = 10,000 years, the concentration in the fractures exceeds that in the water released from the repository (i.e., XR = 1.07 in tsw39). This is caused by pore-exclusion (straining) at the TSw-CHn interface, which leads to the accumulation of the 6 nm colloid in the tsw39. Transport at the water table is shown in Figures VI.13 to VI.24, and follows patterns analogous to those in the tsw39 layer. The high concentrations at the water table at early times (t = 100 years) are noteworthy. Matrix XR begins to show signs of colloid presence at t = 1,000 years, and shows highly localized areas of large concentration at t = 10,000 years. The nonzero area of the matrix FR is consistently larger than that for fracture XR. Note that the fracture XR above the water table exceeds that in the water released from the repository, and the differences between the two increases with time. At t = 10,000 years, the maximum fracture XR > 2, i.e., the concentration of the 6 nm colloid in the fractures above the water table is more than double that in the water released at the repository. This occurs because of the fast transport of the colloids in the fractures (where limited attachment occurs because of the large fracture porosity, i.e., f = 1.0) and their accumulation. This accumulation is caused by the very significant pore-size exclusion and kinetic (physical-chemical) filtration at the water table, which prevent colloids from entering the saturated zone while allowing water to flow into it. This occurs because the saturated zone behaves as a porous (rather than a fractured) medium with the properties of the TSw matrix. Review of Figures VI.1 through VI.24 indicates that transport of the 6 nm colloid is dominated by the faults identified and discussed in Sections 6.10 to 6.12. Diffusion from the fractures into the matrix is delayed for the reasons discussed earlier, but the areal extent and the magnitude of the filtered concentration (indicated by the FR distribution) indicates that its effect is sizable. 6.18.5.3 Transport Mechanisms and Patterns of the 450 nm Colloid The fracture XR, matrix XR and matrix FR distributions of the 450 nm colloid (Co450) in the tsw39 and above the water table are shown in Figures VII.1–VII.24 (see Attachment VII). These figures correspond to the same times discussed in the 6 nm colloid. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 184 November 2003 Comparison of the figures in Attachment VII to those for the 6 nm colloid (Attachment VI) indicates that fractures are far more important to the transport of the 450 nm colloid. Supporting evidence is provided by (a) the larger areal extent and values of the fracture XR, (b) the smaller areal extent and lower values of the matrix FR, and (c) the substantial delay in the appearance of 450 nm colloids in the matrix aqueous phase, when compared to the 6 nm case. Thus, the fracture XR in Figure VII.4 (t = 100 years) indicates contributions from the whole area (at least) of the repository, and shows a much larger footprint than the one corresponding to the 6 nm colloid. At the water table and at early times (<10,000 years), the maximum fracture XR exceeds 1 (=1.50) because of straining (pore-size exclusion). Because of its larger size, the straining is more pronounced for the 450 nm colloid than in the 6 nm colloid, and leads to higher concentrations at earlier times behind the interface. An XR>1 does not necessarily mean clogging and decrease in permeability (at least for a long time) if the concentrations of the colloid waste form are sufficiently low. Compared to the 6 nm colloid, the 450 nm colloid exhibits consistently lower matrix FR values over smaller areas (see Figures VII.3, VII.6, VII.9, VII.12, VII.15, VII.18, VII.21, VII.24). This is consistent with the expectation of lower diffusion from the fractures into the matrix and increased straining because of its larger size. 6.18.6 Colloid Transport Case 3 (Class I) The change in the magnitude of the reverse kinetic filtration coefficient .- (as a fraction of .+ ) has a very small effect on colloid transport (Figure 6.18-2(a)), attesting to the fact that the role of advection through fracture flow (the same in Cases 1 through 3) is by far the dominant mechanism of colloid transport. Although this case is closer to equilibrium filtration, the effect on transport is apparent only in the case of the 6 nm colloid (which can enter the pores, and thus be subject to filtration), which shows an RF even lower than that of Case 2. The insensitivity of transport of the larger colloids to the sorption model in the matrix indicates that the matrix has practically no participation in the retardation, which is attributed to straining at matrix/fracture interfaces. The pattern that emerges is the same as the one discussed in Case 1, and all the comments made and conclusions reached therein apply. 6.18.7 Colloid Transport Case 4 (Class I) By assigning a porosity of 1% to the fractures, and setting the fracture filtration properties equal to those of the matrix in Case 2, we create a system in which the fractures are partially filled with a porous medium. The simulation results are shown in Figure 6.18-2(b). The effect of limited diffusion on transport (because of pore-size exclusion and filtration) becomes more obvious in this case. While the more-freely-diffusing 6 nm colloids exhibit a behavior similar to the one in Cases 1 through 4, the effect on the larger colloids is more dramatic. The occurrence of even a minor fillup retards the colloid transport, increasing t10 to about 30 years, and registering a t50 of about 100 (for the 450 nm colloid) to 150 years (for the 200 nm colloid). The maximum RF exceeds that in Case 2 because of temporary colloid storage in (and subsequent release from) the porous medium that partially fills the fractures in Case 4. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 185 November 2003 Additionally, the breakthrough curve assumes a decidedly different shape in the larger colloids, marked by a sudden and dramatic reversal that is consistent in all the three larger colloids. As the occurrence of this phenomenon coincides with the onset of rapid decline in concentrations due to radioactive decay, it is assumed that the kinetic filtration model is very sensitive to minute changes in concentration at this point—hence the sudden slope reversal. However, this issue is not well understood and will require further investigation. Note that the significant differences among Figures 6.18-2(b) and 6.18-1(a), 6.18-1(b), and 6.18- 2(a) result from assigning matrix filtration properties to only 1% of the pore volume of the fractures (by no means an unphysical or unlikely scenario). The significant retardation of the colloids gives a measure of the importance of fractures and their relative contribution to the total colloid transport through the UZ system. 6.18.8 Transport of Class II Colloids (Case 2 and 3) The transport simulations were repeated for Class II colloids with the filtration behavior of Cases 2 and 3 (see Section 6.18.3). As discussed earlier, radioactive concentrations (and fluxes) were obtained by first computing the transport of the non-radioactive corresponding Class IV colloid, and then multiplying it by the remaining radionuclide mass fraction ft (see Eq. 6-31, Section 6.18.1). Given the approximation of constant colloid size in the Class I simulations, the RF results of the Class II simulations were not expected to be very different. Although the radioactive concentrations are much lower in the transport of Class II colloids than in Class I (of the same size and colloid population per unit volume) because of the radioactive component constitutes a small fraction of the total mass, the relative release rate RF (with respect to the initial release rate at the repository) in both cases are practically identical. This is explained by the fact that, mathematically speaking, there is very little difference in the transport equations between these two classes of colloids of the same sizes when the same colloid filtration parameters are assumed. Realistic differences between these two colloid classes are certain to exist in the properties of these two very different (physically and chemically) colloid classes, which will affect their filtration behavior and the corresponding filtration parameters. However, information on the subject is scant. Figure 6.18-3 confirms the expectation discussed above. A comparison of the RF curves over time shows practically no difference from the corresponding results for the Class I colloids in Figures 6.18-1 and 6.18-2. 6.18.9 Uncertainties and Limitations While the results in this section provide some elucidation of colloid transport, caution should be exercised in the interpretation of the simulation results. The reason for this caution is the realization of the substantial knowledge gaps that hamper the study of colloid behavior. Thus, these results should be viewed as indicative and qualitative, rather than quantitative and predictive. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 186 November 2003 Thus, it must be pointed out that basic knowledge about the kinetic clogging and declogging coefficients .+ and .- is embryonic at best, thus severely limiting our predictive capabilities. Despite the insensitivity of transport to the filtration parameters in Cases 1 through 3, the significant change in transport behavior when 1% of the fracture volume is given the attributes of the matrix should provide grounds for caution. There are significant uncertainties in colloid modeling. It is not known whether the applicability limits of the currently available kinetic models, developed from theoretical principles (Herzig et al. 1970 [117519]) and tested in uniform sandy laboratory experiments (van de Weerd and Leijnse 1997 [109249]) or small-scale field tests (Harvey and Garabedian 1991 [109256]), are breached under the UZ conditions at Yucca Mountain. The C-Well experiments (BSC 2003 [162729]) addressed some of the colloid-related uncertainties under realistic field conditions. However, the applicability of the information gleaned from those tests (conducted in the saturated zone) to the UZ has not been established. The affinity of colloids for air-water interfaces can have a significant effect on their transport. The limitations in the equations for the prediction of the forward kinetic filtration (clogging coefficient) have not been tested under the UZ conditions, and the subject of the kinetic declogging coefficient has barely been raised (let alone studied). Additionally, it is unclear how representative the current size-exclusion (straining filtration) models are. Additional challenges and uncertainties are discussed in detail in Section 6.1.3. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 187 November 2003 (a) 1.0 0.8 0.6 0.4 0.2 0.0 Relative release rate (current/initial at the repository) 100 101 102 103 104 105 106 Time (year) REPOSITORY 450 nm 200 nm 100 nm 6 nm WATER TABLE 450 nm 200 nm 100 nm 6 nm Colloid transport Case 3 (b) 1.0 0.8 0.6 0.4 0.2 0.0 Relative release rate (current/initial at the repository) 100 101 102 103 104 105 106 Time (years) REPOSITORY dc= 450 nm dc= 200 nm dc= 100 nm dc= 6 nm WATER TABLE dc = 450 nm dc = 200 nm dc = 100 nm dc = 6 nm Colloid transport Case 4 Output-DTN: LB0307MR0060R1.007, data submitted with this Model Report Figure 6.18-2. Normalized Release at the Water Table in (a) Case 3 and (b) Case 4 of Colloidal Transport for Mean Present-Day Infiltration (all the release curves at the repository are superimposed). Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 188 November 2003 (a) 1.0 0.8 0.6 0.4 0.2 0.0 Relative release rate (current at water table / initial at repository) 100 101 102 103 104 105 106 Time (years) WATER TABLE dc = 450 nm dc = 200 nm dc = 100 nm dc = 6 nm Case 2 - Transport of Colloids With Irreversibly Sorbed 239Pu (b) 1.0 0.8 0.6 0.4 0.2 0.0 Relative release rate (current at water table / initial at repository) 100 101 102 103 104 105 106 Time (years) WATER TABLE dc = 450 nm dc = 200 nm dc = 100 nm dc = 6 nm Case 3 - Transport of Colloids With Irreversibly Sorbed 239Pu Output-DTN: LB0310MR0060R1.010, data submitted with this Model Report Figure 6.18-3. Normalized Release at the Water Table in (a) Case 2 and (b) Case 3 of Colloidal Transport with irreversibly sorbed radionuclide (Mean Present-Day Infiltration) Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 189 November 2003 6.19 ALTERNATIVE MODELS 6.19.1 Alternative Representations and Methods In this section, we investigate alternative conceptual models involving: a) Different representations of the matrix-fracture system (multiple interactive continua versus dual-permeability systems) b) Different conceptual methods of describing the transport problem (particle tracking versus conventional representation). Thus, we consider the effects of employing the concept of the Multiple Interactive Continua (MINC, as implemented by the development of a different grid system in TOUGH2 simulations) (Pruess and Narasimhan 1985 [101707]) on transport predictions related to the RTM. In the MINC method, the steep gradients at the matrix fracture surface are resolved by subgridding of the matrix blocks in an appropriate number of shells. The MINC concept is based on the notion that changes at the fracture-matrix interface will propagate rapidly through the fracture system, while invading the tight matrix comparatively slowly. Thus, changes in the matrix conditions are controlled locally by the distance from the fractures, leading to the creation of a system of nested subdomains. The MINC behavior is expected to result in slower breakthrough curves (as the enhanced fracture-matrix interaction allows for increased diffusion), longer contact times, and more effective sorption (in sorbing media/solute systems). As members of the TOUGH2 family of codes, both T2R3D and EOS9nT are capable of handling MINC grids. Additionally, we investigate the potential of analyzing the transport problem using particle-tracking-based numerical models (DCPT V1.0; LBNL 2000 [132448]; DCPT V2.0; LBNL 2002 [154342]) that appear attractive in terms of execution and storage requirements compared to the conventional models (e.g., T2R3D and EOS9nT). The grid that described transport of a conservative tracer through a 2-D vertical cross section (slice) of the UZ was provided by DTN: LB0308AMRU0185.001 [165172]. The input and output files and data for this study are in DTNs: LB03093RADTRNS.001 [166225] and LB03093RADTRNS.002 [166071]. The results of the simulation are shown in Figure 6.19-1. Predictions from T2R3D and EOS9nT with a conventional dual-permeability grid and a MINC grid are in very good agreement. The result with the MINC grid conforms to expectations, resulting in slower breakthrough curves. Note that despite its scientific and conceptual appeal, the application of the MINC concept to the 3-D UZ site-scale model is challenging because it necessitates replacement of the single matrix block in the current dual permeability system with several MINC subdomains. This increases the already large size of the problem, the memory requirements, and the execution times. The DCPT V1.0 and DCPT V2.0 (LBNL 2000 [132448]; 2002 [154342]) particle-tracking code could provide an alternative, because it enjoys the advantages of low memory requirements and fast execution. Figure 6.19-1 shows that the DCPT V1.0 and V2.0 (LBNL 2000 [132448]; 2002 [154342]) solutions are close to the predictions from the MINC grid (although somewhat more diffusive). Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 190 November 2003 6.19.2 Alternative Diffusion Model As shown in Sections 6.8 to 6.18, the main mechanism of radionuclide retardation in the UZ is diffusion from the fractures into the matrix. This process transfers radionuclides from the fractures (the pathways of fast flow and the main conduits of transport) to the matrix, where their transport is retarded because of the much slower water velocities and matrix sorption. 1.0 0.8 0.6 0.4 0.2 0.0 100 101 102 103 104 105 106 Time (years) EOS9nT Dual k, D = 0 Dual k, D › 0 MINC, D = 0 MINC, D › 0 T2R3D Dual k, D = 0 Dual k, D › 0 MINC, D = 0 MINC, D › 0 YM 2-D DCPT, Dual k v2.0, D › 0 v1.0, D › 0 DTNs: LB03093RADTRNS.001 [166225] and LB03093RADTRNS.002 [166071] Figure 6.19-1. Effect of (a) the MINC Concept and (b) Application of Particle-Tracking Approaches on Breakthrough Predictions at a 2-D Vertical Cross Section of the UZ In this section, we investigate the effect on transport of an alternative conceptual model that does not consider matrix diffusion. In other words, transport in this alternative model is limited to advection in the matrix and in the fractures. Thus, there is no mechanism for radionuclide removal from the fractures other than (a) the relatively small sorption on the fracture walls where the majority in fracture flow (the dominant portion of flow) and (b) sorption in the matrix, where only a small fraction of flow occurs. For non- and moderately sorbing radionuclides, this means that there is no means of retardation, and the contaminants in the fractures are transported virtually unhindered to the groundwater. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 191 November 2003 Note that this alternative model is far more relevant to solutes than to colloids. Because of the very small diffusion coefficient of colloids, transport is advection dominated and diffusive fluxes for the larger colloids (100 nm to 450 nm colloids) are minuscule, as attested to by the practical coincidence of the early portions of the RF curves (see Section 6.18, this Model Report). 6.19.2.1 Instantaneous release The grids, flow fields, and conditions are identical to those in Section 6.7. The parameters used in these T2R3D simulations are listed in Tables 6.5-1, 6.7-1, and 6.7-2. The input and output files for these simulations are included in DTNs: LB0310MR0060R1.010 and LB0310MR0060R1.011, respectively (Table 6.7-2). For instantaneous release and a mean present-day infiltration, the cumulative breakthrough curves of the normalized mass fraction RM in Figure 6.19-2 confirm the expectation of virtually unimpeded transport to the groundwater. The contrast to the RM evolution when matrix diffusion is considered (also included in Figure 6.19-2) is dramatic. Despite their very different sorption affinity to the UZ rocks, all three radionuclides (the nonsorbing 99Tc, the moderately sorbing 237Np and the strongly-sorbing 239Pu) move unhindered in the fractures in the absence of matrix diffusion, and their breakthrough curves to t = 100 years coincide. Transport in this case is extremely fast. The t10 and t50 of all three radionuclides are about 5 years and 20 years, respectively. The very fast early transport and the coincidence of the RM curves are the result of advection (linked to the to rapid fracture flow) and the lack of matrix diffusion. This result is consistent with expectation because there is no other mechanism that can affect transport during this early stage. Note that, by setting f = 1 in the fractures, sorption onto the fracture walls is not considered in these simulations. After the very fast rise of RM (indicated by the steep slope for t > 100 years in figure 6.19-2), the RM slope becomes much flatter and the three breakthrough curves begin to differentiate because of contributions of matrix flow (and advection) to transport. The change of the slope indicates exhaustion of the radionuclide supply in the fractures (from the initial from the instantaneous release). The breakthrough curve of 239Pu daughters shows a plateau after the initial steep rise portion. This is caused by the very strong 239Pu sorption onto the matrix of the UZ rocks, which does not allow 239Pu in the matrix flow to reach the water table. Thus, all the 239Pu in the fractures reaches the water table in a very short time, and all 239Pu in the matrix is trapped by sorption. Because of its moderate sorption affinity to the UZ rocks, the 237Np breakthrough shows a very mild increase from t = 100 years to t = 20000 years, after which time it registers a fast rise. This is attributed to arrival at the water table (through matrix flow and advection) of 237Np that had sorbed onto and traveled through the matrix, into which it had been transported by fracture-tomatrix flow. Because of lack of sorption, the 99Tc breakthrough curve is marked by a shorter period of the mild slope (up to about t = 3,000 years) and an earlier arrival of 99Tc through matrix advection. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 192 November 2003 The results of this study emphasize the importance of diffusion from the fractures into the matrix as the main mechanism for retardation of radionuclide transport. When this process is not considered in the simulations, radionuclides reach the water table in significant quantities in less than 5 years. 1.0 0.8 0.6 0.4 0.2 0.0 Normalized mass fraction 100 101 102 103 104 105 106 Time (years) D = 0 99Tc 237Np 239Pu D > 0 99Tc 237Np 239Pu 0.1 fraction line 0.5 fraction line Figure 6.19-2. Cumulative breakthrough of radioactive solutes at the water table for the no-diffusion alternative model (mean present-day infiltration, instantaneous release). 6.19.2.2 Continuous release The grids, flow fields, and conditions are identical to those in Section 6.15. The parameters used in these EOS9nT simulations are listed in Tables 6.5-1, 6.7-1, and 6.7-2. The input and output files for these simulations are included in DTN: LB0310MR0060R1.010 and LB0310MR0060R1.011, respectively (Table 6.7-2). Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 193 November 2003 For continuous release and a mean present-day infiltration, the relative release rate RF curves in Figure 6.19-3 confirm the expectation of virtually unimpeded transport to the groundwater. For reference, the relative release rates of the decaying parents at the repository are also included. The contrast to the RF evolution when matrix diffusion is considered (see Figure 6.15-1, this Model Report) is remarkable. Despite their very different sorption affinity to the UZ rocks, all three radionuclides (the nonsorbing 99Tc, the moderately sorbing 237Np, and the strongly-sorbing 239Pu) move unhindered in the fractures in the absence of matrix diffusion, and their breakthrough curves to t = 70 years practically coincide. Transport in this case is extremely fast. The t10 of all three radionuclides is about 2.5 years, while t50 ranges from 70 (for 99Tc) to 80 (for 237Np) to 85 years (for 239Pu). The early behavior reflects fracture transport exclusively, and the identity of the RF curves of the three radionuclides results from the absence of diffusion into the matrix. Note that, by setting f = 1 in the fractures, sorption onto the fracture walls is not considered in these simulations. The RF curves show a very mildly sloping region (almost a plateau) at t = 30 years. The near-flat portion of the RF curves indicates a roughly constant radionuclide release rate at the water table, all of which is attributable to fracture flow. The matrix effects begin to become evident for t > 70 years, when matrix flow (and the contaminants it transports) begins to arrive at the water table, at which time the RF curves of the three radionuclides begin to diverge. The arrival of matrix flow is indicated by the rising (steeper) portion of the RF curve after the first plateau (associated with fracture-based advection). After t = 100 years, the nonsorbing 99Tc is not retarded during matrix flow and arrives at the water table earlier (indicated by higher RF values) than the moderately sorbing 237Np, while the 239Pu arrival at the water table is the slowest of the three and consistent with its strong sorption affinity to the UZ rocks. After attaining maxima at different times (consistent with the different half-lives of the actinides), RF curves begin to decline because of radioactive decay Accounting for the 239Pu daughters leads to a somewhat different pattern. A plateau in RF is observed from t = 1,000 to t = 10,000 years (denoting stability of the ratio of repository and water table fluxes), followed by a sharp increase. This is attributed (almost exclusively) to the arrival at the water table of larger amounts of the 235U daughter, which sorbs less strongly and has a longer half-life than 239Pu. The 235U arriving at the water table originates from direct 235U source releases at the repository (transported mostly by fracture advection), and from the decay of both free and sorbed 239Pu in the matrix of the UZ rocks. The difference in the RF patterns and the much higher levels of RF attained in this case underscore the importance of accounting for daughters in radionuclide transport studies (see Sections 6.16 and 6.17). The results of this study confirm the observations made in the case of instantaneous release on the importance of diffusion from the fractures into the matrix as the main mechanism for retardation of radionuclide transport. When this process is not considered in the simulations, very fast travel times are observed, and radionuclides reach the water table in less than 5 years. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 194 November 2003 1.0 0.8 0.6 0.4 0.2 0.0 Relative release rate 101 102 103 104 105 106 Time (years) REPOSITORY 99Tc 237Np 239Pu 239Pu+235U+231Pa WATER TABLE 99Tc 237Np 239Pu 239Pu+235U+231Pa 0.1 fraction line 0.5 fraction line Figure 6.19-3. Relative release rate of radioactive solutes at the water table for the no-diffusion alternative model (mean present-day infiltration, continuous release). Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 195 November 2003 6.20 BARRIER EVALUATION, UNCERTAINTIES, AND A NOTE OF CAUTION In this section, we distill the results presented in Sections 6.6 to 6.20 into an overall evaluation of the performance of the Yucca Mountain UZ as a barrier. We discuss the issue of uncertainty and the effect it may have on predictions of barrier performance. Finally, we address the issue of the framework within which the results and predictions in the Model Report should be understood and interpreted. 6.20.1 Barrier Performance The t10 and t50 values for all the radionuclides, all the climatic scenarios, and the two release scenarios (instantaneous and continuous) are listed in Tables 6.20-1 to 6.20-3. A long t10 and a long t50 are desirable because they indicate desirable barrier performance. Of the two, t10 provides a measure of influence for the most conductive fractures on the UZ barrier behavior, while t50 accounts for the effects of both fractures and matrix. A short t10 and a short t50 could be an indication of inadequate barrier performance. A short t10 but an adequately long t50 is an indication of a system with strong early arrivals at the water table because of fracture-dominated flow (and, consequently, transport), but with an adequate overall barrier performance in the long run. A review of the results in Tables 6.20-1 to 6.20-3 indicates that, in conformance with expectations, the barrier performance of the UZ is a function of the radionuclide under investigation. Additionally, there is an almost universal early arrival of a small portion of the total released mass (usually = 10%) at the water table. The early arrivals are caused by the very permeable fractures at the Drillhole Wash fault and the Pagany Wash fault, and, to a much lesser extent, the Sundance fault, which lead to fast advective transport. Thus, by having the repository footprint straddling important faults, and by following an approach that releases radionuclides throughout the repository, the barrier of the UZ is breached, once the WP and EBS systems are breached. Based on the flow model described in BSC (2003 [163045], Section 6) and incorporated into this transport study (and despite the adverse release patterns and approaches discussed above), the UZ appears to be an effective barrier in the transport of the following radionuclides: a) 90Sr and 226Ra: In the 1,000,000-year study period, their migration through the UZ registers arrivals of less than 10% of the original mass (under instantaneous release) at the water table, even under the most humid of the present-day and monsoon infiltration regimes, and for low and mean glacial infiltration. This is caused by an early arrival (related to fracture flow and the related advection) of a small portion of the total mass (>10%) only for an upper glacial infiltration. Early arrivals are caused by the very permeable fractures at the faults discussed earlier. The low fraction of the total mass reaching the water table is caused by diffusion into the matrix, strong sorption in the matrix, and the relatively short half lives of these radionuclides, combining to prevent any matrix-associated radionuclide advection to the water table. b) 229Th: With the exception of the upper monsoon and glacial infiltrations, less than 10% of the initially released mass reaches the water table. Although a short t10 is observed for the Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 196 November 2003 most humid of the glacial and monsoon infiltration regimes, only a small portion of the total mass (<20%) arrives at the water table in the 1,000,000-year study period. This is attributed to the strong sorption (after diffusion into the matrix) and the relatively short half-life of 229Th, which eliminate any matrix-related 229Th advection to the water table. c) 241Am and 221Pa: With the exception of the most humid (upper) conditions of all three climatic scenarios (present-day, monsoon, and glacial), less than 10% of the initially released mass reaches the water table. As appears to be the case with all radionuclides, there is an early arrival, but only a small portion of the total mass (<20%) arrives at the water table in the 1,000,000-year study period. This is caused by diffusion into the matrix, strong sorption in the matrix, and the relatively short half lives of these radionuclides. Thus, radionuclide arrivals are limited to early advection due to fracture flow, with no apparent contributions of matrix transport to water table arrivals. d) 239Pu: The transport of 239Pu is characterized by the early arrival (common to all radionuclides at the UZ) of a small portion of the total released mass. As in the case of 241Am and 221Pa, with the exception of the most humid (upper) conditions of all three climatic scenarios (present-day, monsoon, and glacial), less than 10% of the initially released mass reaches the water table in this early arrival. Although 239Pu sorbs strongly onto the matrix after diffusion from the fractures (the overwhelmingly dominant transport features), the breakthrough curves in Figures 6.10.1 to 6.10.3 indicate contributions of matrix-related advection to transport (indicated by the second plateau in the figures). This is possible because of the long half-life of 239Pu, which exceeds the travel time of matrix flow to the water table. However, the fraction of the total mass that arrives at the water table over the 1,000,000-year study period does not exceed the 50% level under any climatic conditions. The effects of having (a) the repository straddle important faults and releasing radionuclides directly into the fault fractures (the approach followed in this study) and (b) the improbably conservative (if not practically impossible) approach described in Section 6.7.8, are apparent in the UZ barrier performance in the transport of the remaining radionuclides considered in this study. By allowing these conditions, the barrier appears far less effective, as indicated by the breakthrough curves of the variably sorbing 135Cs (strongly on zeolites, much less on other rocks), the mildly sorbing 233U, 235U, 237Np, and the nonsorbing 99Tc. All these radionuclides have long half-lives (indicating matrix-based advective contributions to transport) and show substantial arrivals at the water table at times that are shorter than their half-lives. Ensuring that the repository is not extended across the important faults discussed earlier is deemed critical in maintaining the integrity of the UZ barrier. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 197 November 2003 Table 6.20-1. Radionuclide Travel Times to the Water Table (Instantaneous Release) Present-Day Monsoon Glacial-Transition Radionuclide Infiltration Scenario t10 (years) t50 (years) t10 (years) t50 (years) t10 (years) t50 (years) 241Am Lower - - - - - - Mean - - - - - - Upper 12 - 3 - 1 - 237Np Lower 33,800 >1,000,000 15 6,160 185 34,400 Mean 410 25,400 8 2,120 4 1,070 Upper 4 1,600 2 714 1 336 231Pa Lower - - - - - - Mean - - - - - - Upper 13 - 4 - 2 - 239Pu Lower - - 86,000 - - - Mean - - 10,400 - 3,710 - Upper 1,530 - 4 - 2 - 226Ra Lower - - - - - - Mean - - - - - - Upper - - - - 3 - 90Sr Lower - - - - - - Mean - - - - - - Upper - - - - 3 - 99Tc Lower 13,900 >1,000,000 22 1,310 102 8,140 Mean 83 6,640 9 417 6 164 Upper 6 230 2 92 1 42 229Th Lower - - - - - - Mean - - - - - - Upper - - 4 - 2 - 233U Lower 65,200 >1,000,000 103 6,730 549 36900 Mean 433 29,100 34 2,130 16 893 Upper 12 1,120 3 458 2 208 235U Lower 55,300 784,800 101 6480 540 32,600 Mean 430 26,500 34 2080 15 882 Upper 12 1,100 3 450 2 206 135Cs Lower >1,000,000 >1,000,000 22,400 >1,000,000 150,000 >1,000,000 Mean 52,500 >1,000,000 4,690 309,000 2,460 120,000 Upper 2,170 71,200 753 24,500 305 990 Output-DTN: LB0307MR0060R1.007 NOTE: (-): t10 or t50 has never been reached because of radioactive decay. (>1,000,000): Simulations only up to 1,000,000 years. The t10 or t50 can be reached after 1,000,000 years. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 198 November 2003 Table 6.20-2. Radionuclide Travel Times to the Water Table under Continuous Release Case Species t10 (years) t50 (years) Three-Parents 99Tc 74 3,901 237Np 781 22,940 239Pu - - 239Pu-Chain 239Pu+235U+231Pa 6,419(a) 33,660(a) 241Am-Chain 241Am+237Np+233U+229Th 1,027(a) 23,450(a) Output-DTN: LB0307MR0060R1.007 NOTE: (a) Corresponds to the sum of the chain members (-):t10 or t50 has never been reached because of radioactive decay. Table 6.20-3. Colloid Travel Times to the Water Table (Continuous Release) Casea Colloid Size (nm) t10 (years) t50 (years) 1 450 4.35 - 200 4.39 - 100 4.53 - 6 - - 2 450 4.35 - 200 4.39 - 100 4.53 - 6 - - 3 450 4.35 - 200 4.39 - 100 4.52 - 6 - - 4 450 32.4 243 200 27.8 251 100 27.6 - 6 - - Output-DTN: LB0307MR0060R1.007 NOTE: (-):t10 or t50 has never been reached because of radioactive decay. a=Cases are defined in Section 6.18.3 Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 199 November 2003 6.20.2 The Effect of Radionuclide Release Directly Into The Fault Fractures Based on the breakthrough results and the transport patterns determined in Sections 6.8 to 6.19, the early arrival of radionuclides at the Yucca Mountain water table (indicated by the low t10 values) was attributed to advective transport through the fractures associated with the Drillhole Wash fault and the Pagany Wash fault. The concentration distributions in the fractures at both the bottom of the TSw and at the water table even at very early times (t = 10 years) and for both types of release (instantaneous and continuous) coincide with the fault outline. Thus, it was important to evaluate the impact of releasing radionuclides directly into the fractures of these faults. This was accomplished by repeating select radionuclide transport simulations in which the initial radionuclide concentration in repository gridblocks (a) representing the fault or (b) being adjacent to it had been set to zero. In essence, radionuclide sources were removed from the faults and from gridblocks on either side of them. These simulations were limited to (a) meanpresent day infiltration and (b) the radioactive solutes 99Tc, 237Np, 239Pu and 235U, which cover the spectrum of sorption behavior. Both release scenarios (instantaneous and continuous) were investigated. The grids, flow fields, and conditions are as discussed in Section 6.7. The parameters used in 3-D simulations are listed in Tables 6.5-1, 6.6-1, 6.7-1, and 6.7-2. The initial condition files INCON were the only difference between the input files and those of the corresponding simulations discussed in Sections 6.8 to 6.11 and 6.15. The DTN of the input and output files for the three present-day infiltration scenarios is Input/Output-DTN: LB0307MR0060R1.005 (Table 6.7-2). Contrary to expectations, eliminating potential sources from the vicinity of the fault fractures appears to have only a small effect on transport and arrivals at the water table. For instantaneous release (when a finite radionuclide mass is involved), the RM breakthrough curves in Figures 6.20-1 and 6.20-2 show a small increase in t10 (in line with expectations because there is no release into the fast-conducting fault fractures), but t50 is practically unchanged. The RF breakthrough curves in Figures 6.20-3 and 6.20-4 for continuous release show practically no discernible difference from those that account for release into the faults (Section 6.15). The reason for this is demonstrated in the contour plots of the 99Tc distribution of XR in Attachment VIII. The effects of eliminating releases from the immediate vicinity of the faults are obvious at the bottom of the TSw, at which level the radionuclide distribution shows significant retardation compared to that shown in Figures 6.8-2 to 6.8-11. This is particularly evident in the northern part of the repository, where the outlines of the Drillhole Wash fault and of the Pagany Wash fault (which dominated transport when radionuclides were released throughout the area of the repository) are absent from the contour plots in Figures VIII.1 to VIII.10. Remarkably, the XR distribution at the water table (Figures VIII.11 to VIII.20) is very similar to that with direct releases into the faults (Figures 6.8-12 to 6.8-21), and is dominated by transport through the Drillhole Wash fault and the Pagany Wash fault. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 200 November 2003 (a) 1.0 0.8 0.6 0.4 0.2 0.0 Normalized Mass Fraction 100 101 102 103 104 105 106 Time (years) 99Tc - Instantaneous Release WATER TABLE No release in vicinity of faults Release directly into faults (b) 1.0 0.8 0.6 0.4 0.2 0.0 Normalized Mass Fraction 100 101 102 103 104 105 106 Time (years) 237Np - Instantaneous Release WATER TABLE No release in vicinity of faults Release directly into faults Output-DTN: LB0307MR0060R1.007, data submitted with this Model Report Figure 6.20-1. Cumulative Breakthrough of the 99Tc and 237Np Mass Fractions RM at the Water Table (Instantaneous Release, Mean present-day infiltration, no Fault Releases) Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 201 November 2003 (a) 0.10 0.08 0.06 0.04 0.02 0.00 Normalized Mass Fraction 100 101 102 103 104 105 106 Time (years) 239Pu - Instantaneous Release WATER TABLE No release in vicinity of faults Release directly into faults (b) 1.0 0.8 0.6 0.4 0.2 0.0 Normalized Mass Fraction 100 101 102 103 104 105 106 Time (years) 235U - Instantaneous Release WATER TABLE No release in vicinity of faults Release directly into faults Output-DTN: LB0307MR0060R1.007, data submitted with this Model Report Figure 6.20-2. Cumulative Breakthrough of the 239Pu and 235U Mass Fractions RM at the Water Table (Instantaneous Release, Mean present-day infiltration, No Fault Releases) Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 202 November 2003 (a) 1.0 0.8 0.6 0.4 0.2 0.0 Relative release rate (current/initial at the repository) 100 101 102 103 104 105 106 Time (years) Repository WATER TABLE No release in vicinity of faults Release directly into faults 99Tc - Continuous Release (b) 1.0 0.8 0.6 0.4 0.2 0.0 Relative release rate (current/initial at the repository) 100 101 102 103 104 105 106 Time (years) Repository WATER TABLE No release in vicinity of faults Release directly into faults 237Np - Continuous Release Output-DTN: LB0307MR0060R1.007, data submitted with this Model Report Figure 6.20-3. Cumulative Breakthrough of the 99Tc and 237Np Normalized Release Rates RF at the Water Table (Continuous Release of Decaying Source, Mean present-day infiltration, No Fault Releases) Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 203 November 2003 (a) 10-5 10-4 10-3 10-2 10-1 100 Relative release rate (current/initial at the repository) 100 101 102 103 104 105 106 Time (years) Repository WATER TABLE No release in vicinity of faults Release directly into faults 235Pu - Continuous Release (b) 1.0 0.8 0.6 0.4 0.2 0.0 Relative release rate (current/initial at the repository) 100 101 102 103 104 105 106 Time (years) Repository WATER TABLE No release in vicinity of faults Release directly into faults 235U - Continuous Release Output-DTN: LB0307MR0060R1.007, data submitted with this Model Report Figure 6.20-4. Cumulative Breakthrough of the 239Pu and 235U Normalized Release Rates RF at the Water Table (Continuous Release of Decaying Source, Mean present-day infiltration, No Fault Releases) Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 204 November 2003 This pattern indicates that (a) avoiding radionuclide releases directly into and in the vicinity of fractures does retard transport to shallow planes of reference such as the bottom of the TSw, but, (b) given a sufficient deep UZ, lateral transport through an interconnected (both vertically and horizontally) fracture network compensates for the lack of direct releases into the faults. Thus, the interconnected fracture system leads to radionuclide transport to (and emergence at) the water table through fault fractures into which no direct releases at the repository ever occurred. The same pattern is evident in the XR distribution of all other radionuclides and release scenarios investigated here (see Output-DTN: LB0307MR0060R1.006). 6.20.3 Interpreting the Model Report Results While the results presented in this Model Report provide some elucidation of the issue of tracer (solute and colloid) transport in the UZ, caution should be exercised in their interpretation. This is because of the conceptual approach and assumptions of this study, and some knowledge gaps (especially in the area of colloid behavior). It is important to keep in mind that this study uses relative quantities (with respect to the tracer concentration in the water released from the repository), and consequently, all the concentration predictions are relative in nature and presuppose the ability (by no means guaranteed) of the radioactive solutes and colloids to reach the underlying fractures. Thus, predictions of large relative concentrations may mean little unless and until the tracer release becomes possible, significant, and known. Moreover, the presented analysis assumes that conditions for the creation and stability of solute and colloidal species exist, and that the effects of the near-field chemical, physical, mineralogical, and thermal conditions on their creation and stability over time can be ignored. While these assumptions may provide a worst-case scenario, they may be unrealistically conservative. Note that the 3-D site-scale simulations discussed here do not describe a realistic (expected-case) scenario: radioactive tracer release is modeled as occurring continuously, uniformly (even into faults), over the whole area of the repository. This presupposes the near-simultaneous rupture of all the waste-containing vessels, the ability of water dripping into the repository to focus exclusively on the ruptured vessels and to flow through them, the spatially and temporally constant contaminant release from each waste package, and the lack of any immobilization process (e.g., precipitation, colloid flocculation, etc.) before they enter the fractures. The geologic model used in the analysis involves continuous fracture-to-fracture flow in certain portions of the UZ, greatly facilitating transport from the repository to the groundwater. While these assumptions provide the upper limit of a worst-case scenario, this is an implausible (if not a virtually impossible) situation. In that respect, the results discussed here should be viewed as an attempt to identify and evaluate the mechanisms, processes, and geological features that control UZ transport using the largest possible input signal (i.e., the worst-case scenario), rather than as an effort to quantitatively predict the effects of a realistic radionuclide release regime. Finally, there are significant uncertainties in colloid modeling. The currently available kinetics models were developed from theoretical principles (Herzig et al. 1970 [117519]; Yao et al. 1971 [101169]) and were tested under saturated conditions, in uniform sandy laboratory experiments (van de Weerd and Leijnse 1997 [109249]), and small-scale field tests (Harvey and Garabedian 1991 [109256]) that involved microspheres or bacteria. It is not known if these models are Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 205 November 2003 applicable to the unsaturated complex fracture-matrix system and the expected colloidal waste forms in the UZ of Yucca Mountain. The limitations of the equations predicting .+ have not been tested under the UZ conditions, and the subject of .- has barely been raised. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 206 November 2003 INTENTIONALLY LEFT BLANK Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 207 November 2003 7. VALIDATION Section 7 is composed of seven subsections. In Section 7.1 we discuss the main approach and validation criteria for the UZ Radionuclide Transport Model (RTM). In Section 7.2 we validate the RTM by means of 3-D and 2-D simulations, respectively. The validation of the matrix diffusion model is discussed in Section 7.4, while the sorption model is discussed in Section 7.5. The colloid model is discussed in Section 7.6. Finally, the Active Fracture Model (AFM) with matrix diffusion is addressed in Section 7.7. 7.1 VALIDATION OF THE RADIONUCLIDE TRANSPORT MODEL (RTM) 7.1.1 Stipulated Validation Requirements, Methods, and Criteria In this section we discuss the validation requirements for the RTM and the corresponding criteria. The validation approaches and criteria are based on the Technical Work Plan (TWP) stipulations (BSC 2002 [160819], Section I-2-1-1), which indicate the following: Quantitative, qualitative, and review criteria will be used for the model validation. One or more of following quantitative criteria will be used: • Method 1—Corroboration with Experimental Data: The Busted Butte Unsaturated Zone Transport Test (UZTT) will be simulated, and the predicted and experimental results (normalized concentration vs. time) will be compared. The validation criteria is: the time at which UZ RTM predicts a value of normalized concentration between 0.03 and 0.97 will be less than twice as long, (or any amount shorter than), the time indicated by a smoothed breakthrough curve fit through the data. The following section (Section 7.1.2) explains that Method 1 is not applicable to the analysis of Busted Butte test data. However, for purposes of post-development model validation for RTM field measurements of porewater chloride concentrations from the ESF and of gas-phase 14C ages from boreholes were used. The latter model validation activities are described in detail in BSC (2003 [163045], Section 7), and summarized in Section 7.3.6. • Method 3—Corroboration with Data from Literature: Comparison with the results of laboratory studies that have been described in peerreviewed journals and to which the model can be applied. This validation step will be considered successful if the difference between the model and the laboratory data (value of normalized concentration at a given location and time) does not exceed 50%. • Method 2—Corroboration with Alternative Mathematical Models: Comparison to analytical solutions (when available). This step is necessary, but not sufficient. This validation step will be considered successful if the difference between the Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 208 November 2003 model and the analytical solutions does not exceed 1% (value of concentration at a given location and time). • Method 5— Independent Technical Review • Publication in a Refereed Professional Journal (for corroboration) 7.1.2 Validation Approach and Criteria The UZ RTM was validated using the numerical codes (T2R3D and EOS9nT, see Sections 7.2 and 7.3). The validation process is based on a combination of (a) comparisons to known analytical solutions and models (Method 2), (b) scientific acceptance, as demonstrated by publications in refereed journals, (c) model acceptance by other organizations involved in radionuclide transport studies (Method 5). Comparisons and model performance evaluation in the study of the Busted Butte field tests (or other field data) of tracer transport are also presented in Section 7.3 as a calibration and validation exercise conducted during model development to enhance confidence in the RTM. In addition, field measurements of porewater chloride concentrations from the ESF and of gasphase 14C ages from boreholes were used in post-development model validation of the RTM. The latter model validation activities are described in detail in BSC (2003 [163045], Section 7), and summarized in Section 7.3.6. For the RTM validation through predictions of the numerical (TOUGH2 V1.11 MEOS9nTV1.0 (LBNL 1999 [113943]) and T2R3D V1.4 (LBNL 1999 [146654])) codes, three conditions must be met: (1) The relevant physical processes must be accounted for and accurately represented by the mathematics. (2) The mathematical equations must be accurately solved by the numerical methods. (3) The physical properties of the rocks must be correct. For the validation of the RTM using the numerical code, their simulation results are compared to relevant analytical solutions (Method 2). The validation criterion we used in this case is one or more of the following (a) Mass fractions agree within 5% when CR = C/C0 = 10-4, or (b) The location of the tracer fronts agree within 5% when CR = C/C0 = 10-4, or (c) The tracer mass balances agree within 5%. The results presented in Table 7.2-6 show that this criterion was met in all instances. Note that the criterion here is somewhat different from the one discussed in BSC (2002 [160819], Section I-2-1-1) and listed in Section 7.1.1 of this Model Report because the latter (a) is inadequate in that it captures only one of the three possible aspects for quantitative agreement, and (b) can be Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 209 November 2003 unrealistic near the tail end of the CR curve (where CR values are very low, and both analytical and numerical predictions may be unreliable because of roundoff error). For validation through scientific acceptance, the criterion we used is publication (or acceptance for publication) in peer-reviewed journals in the fields of hydrology and contaminant transport. Similarly, the criterion through acceptance by other organizations is involvement of the organization in radionuclide transport studies, and reporting of the results of these studies in topical reports or peer-reviewed journals. For additional validation of the RTM using numerical codes through comparison of their predictions to field measurements, the criterion we used is (1) Overall agreement within 50% of each other, and, (2) If this is not possible, capturing the correct trend, accompanied by a documented analysis and discussion of the reasons for the observed deviations. Note that inability of the model to match/predict field observations is not necessarily a sign of inability to validate the model if (a) the quality of the measurements is suspect, (b) the test design may be responsible for uncertainties in the measurements, (c) there is insufficient information to fully describe the field test, or (d) the problem is demonstrated to involve very steep gradients (involving very significant parameter changes over a short distance or time) that lead to measurements prone to uncertainties and inaccuracies. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 210 November 2003 7.2 RTM VALIDATION THROUGH COMPARISON TO ANALYTICAL SOLUTIONS AND SCIENTIFIC ACCEPTANCE The two numerical codes used for the transport studies in this Model Report are TOUGH2 V1.11 Module EOS9nT V1.0 (LBNL 1999 [113943]) and T2R3D V1.4 (LBNL 1999 [146654]). The software code TOUGH2 V1.11 Module EOS9nT V1.0 (LBNL 1999 [113943]) simulates flow (saturated and/or unsaturated) and the decoupled transport of multiple radioactive solutes and/or colloids (parents and daughters) in complex subsurface systems involving porous and/or fractured media. The transport equations account for radioactive decay, advection, molecular/colloidal diffusion, hydrodynamic dispersion, kinetic or equilibrium physical and chemical sorption (linear, Langmuir, Freundlich, or combined), first-order linear chemical reaction, colloid filtration, and colloid-assisted solute transport. EOS9nT can employ conventional time stepping or two Laplace transform formulations. The code T2R3D V1.4 (LBNL 1999 [146654]) simulates flow (saturated and/or unsaturated) and the coupled transport of a single radioactive solute tracer in complex subsurface systems involving porous and/or fractured media. The transport equations account for radioactive decay, advection, molecular diffusion, hydrodynamic dispersion, and linear equilibrium sorption. These two codes have been shown to produce identical results in simulations of transport involving (a) conventional time stepping and (b) nonsorbing tracers or tracers whose sorption follows an equilibrium linear sorption model. A comparison of the performance of the two codes is shown in Figure 7.2-1, which depicts predictions of transport of a conservative (nondecaying, nonsorbing) tracer along a vertical plane (slice) of Yucca Mountain. The data corresponding to this simulation can be found in DTN: LB0308AMRU0185.001 [165172]. Figure 7.2-1 (Section 7.2, this Model Report) shows that the EOS9nT and T2R3D results practically coincide. Further evidence of the identity of the EOS9nT and T2R3D predictions is provided in BSC (2001 [161340], Figure 6.4.5) and in DTNs: LB03093RADTRNS.001 [166225], LB03093RADTRNS.002 [166071]. 7.2.1 Validation Through Comparison to Analytical Solutions In this validation phase, several problems (relevant to conditions and possible transport scenarios at Yucca Mountain) with known analytical solutions were studied. The specifics of these problems and the code performance are discussed in detail below. 7.2.1.1 Test 1: Transport of a Nonsorbing, Nondecaying Tracer This problem involved transport of a nonsorbing, nondecaying tracer through a porous medium. A family of analytical solutions to this problem were developed by Bear (1979 [105038]). The properties and parameters for this problem are listed in Table 7.2-1, and the comparison of the analytical to the numerical solutions is listed in Table 7.2-6. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 211 November 2003 1.0 0.8 0.6 0.4 0.2 0.0 10-1 100 101 102 103 104 105 106 Time (years) Dual k system EOS9nT T2R3D YM 1-D Column @SD9 DTN: LB03093RADTRNS.001 [166225] AND LB03093RADTRNS.002 [166071] Figure 7.2-1. Comparison between the Numerical Predictions of Breakthrough from TOUGH2 V1.11 Module EOS9nT V1.0 (LBNL 1999 [113943]) and from T2R3D V1.4 (LBNL 1999 [146654]). Both T2R3D V1.4 (LBNL 1999 [146654]) and TOUGH2 V1.11 Module EOS9nT V1.0 (LBNL 1999 [113943]) were used in this test. Figure 7.2-2 shows a comparison between the analytical solution and the numerical model predictions, and includes solutions from three different time treatments (designated by T, H, S to indicate regular time stepping, De Hoog Laplace space solution and Stehfest Laplace space solution, respectively). Note that the results for regular time stepping (i.e., the T-solutions) reflect both the EOS9nT and T2R3D estimates, which are not shown separately because they are practically indistinguishable from each other, and from the Laplace space formulation solutions. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 212 November 2003 1.0 0.8 0.6 0.4 0.2 0.0 60 50 40 30 20 10 0 Distance (m) t = 400 days R = 1 D = 0.1 m2 day-1 V = 0.1 m day-1 t = 50 days .x = 0.25 m H,S,T solutions Analytical DTN: LB0308MR0060R1.009 [Output] NOTE: Note that the results from three different time treatments (designated by T, H, S to indicate regular time stepping, De Hoog Laplace space solution, and Stehfest Laplace space solution, respectively) are indistinguishable. Figure 7.2-2. Comparison between the Analytical Solution and the Numerical Predictions in Test 1 7.2.1.2 Test 2: Transport of Sorbing Radioactive Tracers The transport of three sorbing radioactive tracers is studied in this test. The analytical solutions to this problem were developed by Bear (1979 [105038]). Properties and parameters for this problem are listed in Table 7.2-2, and the comparison of the analytical to the numerical solutions is listed in Table 7.2-6. The input and output files for this test are in DTN: LB0308MR0060R1.008. Both T2R3D V1.4 (LBNL 1999 [146654]) and TOUGH2 V1.11 Module EOS9nT V1.0 (LBNL 1999 [113943]) were used in this test. Figure 7.2-3 shows a comparison between the analytical solution and the numerical solutions, and includes solutions from three different time treatments Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 213 November 2003 (i.e., the T, H, S solutions). As in Test 1, the results for regular time stepping reflect both the EOS9nT and T2R3D estimates. They are not shown separately because they are practically indistinguishable from each other, and from the Laplace space formulation solutions. 7.2.1.3 Test 3: Transport of the 234U -> 230Th -> 226Ra Chain This problem involved transport of the 3-member radioactive chain 234U -> 230Th -> 226Ra through a porous medium. All the members of this chain exist in species that sorb onto the porous medium. The analytical solution to this problem was developed by Pigford et al. (1980 [123113]). The properties and parameters for this problem are listed in Table 7.2-3, and the comparison of the analytical to the numerical solutions is listed in Table 7.2-6. The input and output files for this test are in DTN: LB0308MR0060R1.008 [Output]. 1.0 0.8 0.6 0.4 0.2 0.0 60 50 40 30 20 10 0 Distance (m) T1/2 = 69.32 days R = 1 .x = 0.25 m t = 200 days D = 0.05 m2 day-1 V = 0.2 m day-1 T1/2 = 138.63 days R = 2 T1/2 = 1039.72 days R = 1.5 Analytical Numerical DTN: LB0308MR0060R1.009 [Output] Figure 7.2-3. Comparison between the Numerical Predictions (TOUGH2 V1.11 Module EOS9nT V1.0 (LBNL 1999 [113943]) and T2R3D V1.4 (LBNL 1999 [146654])) and the Analytical Solutions (Bear 1979 [105038]) in Test 2 Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 214 November 2003 10-6 10-5 10-4 10-3 10-2 10-1 100 1 10 100 1000 Distance (m) T solutions H solutions S solutions Analytical U-234 Th-230 Ra-226 DTN: LB0308MR0060R1.009 [Output] NOTE: While the results from regular timestepping and the De Hoog Laplace space solution (designated by T and H, respectively) are in excellent agreement with the analytical solutions, the results from the Stehfest Laplace space solution for the daughters become increasingly inaccurate. Figure 7.2-4. Comparison between the Analytical Solution and the TOUGH2 V1.11 Module EOS9nT V1.0 (LBNL 1999 [113943]) Predictions in Test 3 Because it involves transport of parent and daughter species, T2R3D cannot be used for this study. The results in Figure 7.2-4 were obtained with EOS9nT, and show a practical coincidence of the T- and H-solutions with the analytical solution. When the Stehfest algorithm is implemented (S-solution), accurate estimates of the parent species are obtained, but significant deviations occur in the transport prediction of daughter species. Thus, the Stehfest formulation should not be used for transport studies of radioactive chains. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 215 November 2003 Table 7.2-1. Parameters in Test 1 Parameter Value V (m/day) 0.1 D (m2/day) - Equation 20 0.1 Kd (m3/kg) 0 Water saturation Sw 1 t 1 NOTE: DTN: LB0308MR0060R1.008 [Output] Table 7.2-2. Parameters in Test 2 Parameter Value V (m/day) 0.2 D (m2/day) - Equation 20 0.05 Water saturation Sw 1 Tortuosity factor t 1 . (kg/m3) 2600 Kd (m3/kg) 0, 2.13675x10-5, 4. 2735x10-5 T1/2 (days) 69.32, 1039.72,138.63 NOTE: DTN: LB0308MR0060R1.008 [Output] Table 7.2-3. Parameters in Test 3 Parameter Value V (m/day) 0.273785 D (m2/day) - Equation 20 2.73785 Water saturation Sw 1 Tortuosity factor t 1 . (kg/m3) 2600 Kd (m3/kg) 0, 2.13675x10-5, 4. 2735x10-5 T1/2 (days) 69.32, 1039.72,138.63 Kd of 234U (m3/kg) 1.64819 Kd of 230Th (m3/kg) 8.24159 Kd of 226Ra (m3/kg) 0.0822528 T1/2 of 234U (years)a 2.45x105 T1/2 of 230Th (years)a 7.54x104 T1/2 of 226Ra (years)a 1.60x103 NOTE: a From Pigford et al. (1980 [123113]) NOTE: DTN: LB0308MR0060R1.008 [Output] Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 216 November 2003 7.2.1.4 Test 4: 3H Transport in a Fracture-Matrix System This problem describes the transport of a radioactive substance in a system of parallel fractures. Advection, dispersion, and diffusion occur in the fractures, while the species diffuses through the fracture walls into the matrix. The analytical solution to this problem was developed by Sudicky and Frind (1982 [105043]). Properties and parameters for this problem are as in Sudicky and Frind (1982 [105043]), and are listed in Table 7.2-4. The comparison of the analytical to the numerical solutions is listed in Table 7.2-6. The input and output files for this test are in DTN: LB0308MR0060R1.008 [Output]. Both T2R3D V1.4 (LBNL 1999 [146654]) and TOUGH2 V1.11 Module EOS9nT V1.0 (LBNL 1999 [113943]) were used in this test. Figure 7.2-5 shows a comparison between the analytical solution and the numerical solutions, and includes solutions from three different time treatments (i.e., the T, H, S solutions). As in Tests 1 and 2, the results for regular time stepping reflect both the EOS9nT and T2R3D estimates. The numerical results are not reported separately because, at a minimum, they coincide in the first three decimal digits. Table 7.2-4. Parameters in Test 4 Parameter a Value S 1 . (kg/m3) 2600 D0 (m2/day) 1.6x10-9 Fracture aperture (m) 10-4 Fracture V (m/day) 0.1 Fracture S 1 Fracture f 1 Fracture t 1 aL (m) 0.1 Matrix V (m/day) 0 Matrix block width (m) 0.5 Matrix S 1 Matrix f 1 Matrix t 1 Kd (m3/kg) 0 (in fractures and matrix) T1/2 (years) 12.35 NOTE: a From Sudicky and Frind (1982 [105043]) DTN: LB0308MR0060R1.008 Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 217 November 2003 10-5 10-4 10-3 10-2 10-1 100 25 20 15 10 5 0 Distance z (m) Analytical Numerical 3H Transport in Fractures DTN: LB0308MR0060R1.009 [Output] NOTE: The numerical results from the two codes and the three different time treatments are indistinguishable, and are indicated by the single curve. Figure 7.2-5. Comparison between the analytical solution of Sudicky and Frind (1982 [105043]) and the Numerical (TOUGH2 V1.11 Module EOS9nT V1.0 (LBNL 1999 [113943]) and T2R3D V1.4 (LBNL 1999 [146654])) Predictions in Test 4 7.2.1.5 Test 5: Colloid and Colloid-Assisted Radionuclide Transport This transport of a colloid and a solute is described in this problem. The colloid filtration (attachment) is controlled by a kinetic process. The analytical solution to the problem of transport of colloids undergoing kinetic filtration is presented by de Marsily (1986 [100439], pp. 273-274). The solute is radioactive and is irreversibly sorbed onto the colloid; thus, it cannot enter the liquid phase and is transported only by the mobile colloids. All the radionuclides (produced in a single batch at the same time) that enter the system (through the boundaries or through sources) are sorbed onto colloids. This being the case, the radionuclide decays uniformly, and its concentration is the same on all colloids. The radionuclide concentration on each colloid is controlled by radioactive decay, and by choosing an observation time equal to T1/2, its relative concentration is 0.5. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 218 November 2003 Properties and parameters for this problem are listed in Table 7.2-5, and the comparison of the analytical to the numerical solutions is listed in Table 7.2-6. The input and output files are in Output-DTN: LB0308MR0060R1.008. Because it involves transport of a kinetically filtering colloid, T2R3D cannot be used for this study. The numerical results in Figure 7.2-6 were obtained with EOS9nT and show a practical coincidence with the analytical solution, while the radionuclide relative concentration on the colloids is confirmed to be a uniform 0.5. This is validation by Method 2. Table 7.2-5. Parameters in Test 5 Parameter Value S 1 . (kg/m3) 2600 D0 (m2/day) From Eq. 23 (Section 6.2) .c (kg/m3) 1000 U (m/day) 2 S 1 f 0.3 t 1 aL (m) 0.15 e (1/m) 30 DTN: LB0308MR0060R1.008 Table 7.2-6. Model-validation – Comparison of Numerical to Analytical Solutions in Tests 1 to 5 Test # % Difference Pass/Fail Criterion 1 0.48 (max of T,S,H solutions) Pass 2 0.93 (max of T,S,H solutions, 3 cases) Pass 3 234U: 0.93 (T solution), 0.09 (H-solution), 0.97 (H-solution) 230Th: 0.87 (T solution), 0.16 (H-solution) 226Ra: 0.98 (T solution), 0.21 (H-solution) Pass 4 0.83 (max of T,S,H solutions) Pass 5 0.33 (max of T,S,H solutions) Pass DTN: LB0308MR0060R1.009 Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 219 November 2003 0.001 2 3 4 5 6 7 8 9 0.01 2 3 4 5 6 7 8 9 0.1 2 3 4 5 6 7 8 9 1 0.001 2 3 4 5 6 7 8 0.01 2 3 4 5 6 7 8 0.1 2 3 4 5 6 7 8 1 Distance (m) Analytical EOS9nT t = 7200 s Colloid transport DTN: LB0308MR0060R1.009 [Output] Figure 7.2-6. Comparison between the Numerical (TOUGH2 V1.11 Module EOS9nT V1.0 (LBNL 1999 [113943])) and the Analytical (de Marsily 1986 [100439]) Solution of Colloid Transport in Test 5. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 220 November 2003 7.2.2 Validation through Scientific Acceptance Using T2R3D, Wu and Pruess (2000 [153972]) studied the site-scale 3-D transport of radioactive solutes in the journal Advances in Water Resources. A scientific study by Moridis et al. (2003 [161902]) of site-scale 3-D transport of radioactive colloids in the UZ of Yucca Mountain using EOS9nT was recently published in the Journal of Contaminant Hydrology. Both studies are based on essentially the same RTM as the one discussed here, but it involved a 3-D grid from the previous revision of the relevant Model Reports (BSC 2001 [161340]; BSC 2001 [158726]). Additionally, EOS9nT is currently in use by Gesellschaft für Anlagen und Reaktorsicherheit (GRS) mbH, a research and engineering organization owned by the Federal Government of Germany, the State of Bavaria, the state of North Rhine-Westphalia, the Technical Inspection Organization (TÜV) and the Germanischer Lloyd. GRS has a substantial involvement in radioactivity issues, focusing on studies of the effects of potential radioactivity releases. The work of GRS using the RTM embodied in EOS9nT is documented in two GRS reports (Javeri 2001 [161966]; 2002 [161967]). This is validation by Method 3. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 221 November 2003 7.3 RTM VALIDATION THROUGH CORROBORATION WITH THE BUSTED BUTTE FIELD TESTS, CHLORIDE DATA FROM THE ESF, AND THE 14C DATA The Busted Butte studies analyzed in the Model Report are presented in Section 7.3.1 through 7.3.5 as a calibration and validation exercise conducted during model development to enhance confidence in the RTM. Section 7.3.6 summarizes the post-development validation of the RTM using field measurements of pore water chloride concentrations and gas-phase 14C ages. 7.3.1 The Busted Butte Tests The Busted Butte Unsaturated Zone Transport Test (UZTT) was a long-term experiment conducted in Busted Butte near Yucca Mountain (BSC 2001 [160828]) to investigate flow and transport issues in the UZ site-process models for Yucca Mountain, including: 1) The effect of heterogeneities on flow and transport under unsaturated conditions in the CHn hydrogeologic unit near the TSw-CHn interface, particularly fracture-matrix interactions and permeability contrast boundaries; 2) The migration behavior of colloids in the CHn layers; 3) The validation of laboratory sorption results; 4) Scaling effects (from laboratory to field to site) on 3-D site-scale flow and transport simulations. The site was selected because of the presence of a readily accessible interface of the TSw and CHv units, and the similarity of these layers to those beneath the potential repository horizon. The study of the TSw/CHv interface is important because of the significant role that the vitric layers of the CHv unit play in radionuclide retardation (see Section 6.10, this Model Report). The test proceeded in two phases that differ in design, purpose, and experimental scales, among other factors. A detailed description of the tests can be found in BSC (2001 [160828], Section 6.8). The Busted Butte studies analyzed in this Model Report are presented as a calibration and validation exercise conducted during model development to enhance confidence in the RTM. This analysis and modeling includes portions of the data sets obtained during the main phases of the test, namely Phases 1A, 1B, and 2C. The reasons for the choice of tracers and of the data set segments used in this analysis are discussed in detail in the following sections. 7.3.2 Calibration vs. Validation It is important to articulate the approach employed in the RTM calibration and validation using the Busted Butte test data in this Model Report, and emphasize the difference between validation and calibration. It would be extremely desirable and convenient to be able to describe the flow and transport properties of particular hydrogeologic units (and especially ones that extend over large areas of Yucca mountain) using a set of uniform, homogeneous and isotropic values. Unfortunately, this is not possible because of (a) heterogeneity and anisotropy in hydraulic and transport parameters are unavoidable even in small-scale systems (let alone the mountain scale domain under consideration in the study of the UZ), and (b) scale effects (an issue entirely separate from heterogeneity), which further compound the problem. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 222 November 2003 The expected variability in hydraulic and transport parameters makes problematic (if not unwise) the use of single parameters derived from large-scale inversion processes (such as the process involved in the determination of the hydraulic properties of the various hydrogeologic units in this Model Report) to describe the small-scale experiments at the Busted Butte site. Thus, it was necessary to calibrate the RTM model at each one of the three sites involved in the three phases of the field test (i.e., phases 1A, 1B and 2C). The process of calibration was intended to determine the main hydraulic parameters of the medium under consideration. The calibrated hydraulic parameters were then used in an attempt to validate and enhance confidence in the transport model using transport parameters available from other independent sources. To accomplish the task, two separate data sets of tracer concentrations were used from each of the three test phases, of which one was used for calibration and the other for validation. This is a valid approach because the transport behavior of the two tracers used for calibration and validation was completely independent, i.e., there is no sorption-related correlation. If a particular tracer was known (or expected) to have a sorbing affinity for the tested rock, the corresponding data set was used for the calibration effort, which included an estimation of the tracer’s transport parameters (i.e., the Kd distribution coefficient and tortuosity) in addition to the hydraulic parameters. The non-sorbing tracer (Br in all three of the test phases) was used for validation because (a) it is more demanding as a conservative species, and (b) it has a wellknown (and independently determined) Kd (= 0) and a molecular diffusion coefficient D0 (=2.080x10-9 m2/s, see Table 7.3-1). In Test Phases 1A and 1B, the tracer used for calibration was either non-sorbing (fluorescein in Test Phase 1A) or very mildly sorbing (2,6-DFBA in Test Phase 1B). Thus, the retardation of the calibration tracer was none or very small, and the time frame used for calibration and validation was the same. In the case of the Li tracer in Test Phase 2C, its relatively strong sorption resulted inevitably in a significant retardation, leading to later appearance (than the Br) at the collection point. When Li began registering significant concentrations, the Br concentration was invariant at the collection point because it had already reached a maximum (equal to the injection concentration). In this case, the Li data set corresponding to the later times was used for calibration, while the earlier Br data set was used for validation. This further strengthens validation because of the further degree of separation between the Li and Br data sets. 7.3.3 Phase 1A Test The purpose of this modeling study was to calibrate and validate the model through the reconciliation of field measurements and predictions of the concentration and water saturation, following the injection of nonreactive tracers into the ch1v, and the subsequent flow and transport in the ch1v and ch2v layers. Of the four tracers injected during the field experiment, only the transport of two (the nonsorbing Br and fluorescein) is analyzed here because of the poor quality and unreliability of the data from the other two tracers. In this analysis, (a) the fluorescein data are used for calibration, while (b) the Br data are used for validation. Additional discussion on the calibration and validation approach can be found in Section 7.3.2 of this Model Report. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 223 November 2003 7.3.3.1 Procedure and Layout A schematic of the borehole layout in the Phase 1A test can be found in BSC (2001 [161340], Attachment VI), while more detailed information is provided in BSC (2001 [160828], Section 6.8). Parameters describing the site conditions and test operation (and the corresponding sources) are listed in Table 7.3-1. Only injection into borehole 3 (located in the ch1v layer of the CHn unit) was considered. Borehole 3 is located in the ch1v layer and is about 20 cm above the ch1v-ch2v interface. The injection tests into boreholes 1 (located in the ch1v layer, but is further removed from the interface), 2 and 4 (located in the ch2v layer of the CHn unit) were not analyzed because of the poor quality of the data. An extensive discussion of the test can be found in BSC (2001 [160828], Section 6.8). Mineback of the Phase1A test block provided the field data for this analysis. During mineback, as successive vertical slices were being removed, digital photographs under visible and ultraviolet light were taken to record the distribution of moisture and fluorescein. This distribution in the vicinity of Borehole 3 is used for the calibration of the simulation results discussed in the present section because it is the largest fluorescein plume (in addition to being the most uniform). Additionally, rock samples were collected by augering, and the exposed plane was accurately surveyed. These samples were analyzed for tracer concentration and are the basis for the validation of the Br transport. 7.3.3.2 Conceptual and Numerical Model For this 3-D numerical study, the underlying geologic model considered a homogeneous and anisotropic unfractured rock matrix with the properties of the ch1v and ch2v layers. Because of the injection configuration (as described in BSC 2001 [160828], Section 6.8), only half the domain (i.e., the portion of the domain to the right of the injection point; see Figure 7.3-1) was simulated using a grid consisting of 22,463 elements. The TOUGH2 V1.11MEOS9nTV1.0 (LBNL 1999 [113943]) was used for the simulation. The determination of the hydraulic parameters was a significant component of the calibration effort. Initial estimates for these parameters were provided by two different sources: the UZ99 calibrated flow parameter set (DTN: LB997141233129.001 [104055]) and a data set based on hydraulic property measurements from field-collected samples (DTNs: GS990308312242.007 [107185], GS990708312242.008 [109822]). Initial (uncalibrated) data, as well as all relevant flow and transport parameters, are listed in Table 7.3-1 (which also includes the DTN of the field data and measurements used in the validation) and Table 7.3-2. The DTN of the input and output files of this simulation is DTN: LB0308MR0060R1.008 [Output]. 7.3.3.3 Calibration Using the Fluorescein Data Figures 7.3-1 and 7.3-2 show the fluorescein distribution (as recorded digitally in the field using a UV light) and the corresponding numerically predicted distribution (calibrated) at the y = 0.90 m mineback face. The final (calibrated) transport properties that resulted in the distribution of Figure 7.3-2 are listed in Table 7.3-3. Note that the scale included in Figure 7.3-1 is incorrect, the correct one being the faint feature in the middle of the photograph below the plume. This Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 224 November 2003 was confirmed by reviewing the original scientific notebooks describing the mineback operation (Bussod 1999 [146978], p. 130). A comparison of the two figures indicates good agreement between predictions and observations. The absence of “stringers” and other features indicative of fracture presence, coupled with the ellipsoidal size of the plume, indicating that the assumption of an unfractured (porous) medium is valid (visual evidence of local heterogeneity in Figure 7.3-1 notwithstanding, the plume is quite uniform at the scale of observation). Although the size of the numerically predicted horizontal axis is somewhat larger than the observed one, the compressed shape of the fluorescein plume is accurately captured, the compression of the lower half of the plume due to its vicinity with the less permeable ch2v is accurately rendered, and the size of the vertical axis of the ellipsoid fluorescein plume is in very good agreement with that measured in Figure 7.3-1. However, it is not possible to further refine the calibration because critical information on the minimum concentration at which the tracer will fluoresce is unavailable, thus depriving the study of more reliable information on the extent of the outer boundaries of the plume. Additionally, the lack of scaling information in any of the other digital photographs from the mineback did not allow the use of a larger data set to more accurately determine of the appropriate parameters (and the site heterogeneity distribution) and effectively constrain the solution. A quantitative measure of the model calibration stemming from the comparison of field observations to model predictions is provided in Table 7.3-9. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 225 November 2003 Table 7.3-1. Operational, Transport, and Field Parameters in the Analysis of the Busted Butte Phase 1 Field Test Parameter Value Source Phase 1A: Injection rate (mL/hr) 7 LA0008WS831372.001 [156582] Phase 1A: Injection duration (days) 286 LA0008WS831372.001 [156582] Phase 1B: Injection rate (mL/hr) 8 LA0008WS831372.001 [156582] Phase 1B: Injection duration (days) 191 LA0008WS831372.001 [156582] aL (m) - Ch1v and Ch2v 0.1 Reasonable estimate aT (m) - Ch1v and Ch2v 0 Reasonable estimate D0 of Br (m2/s) 2.08x10-9 Lide (1992 [166224]), p. 5-111 D0 of Fluorescein (m2/s) 8x10-10 Initial estimate D0 of 2,6-DFBA (m2/s) 7.6x10-10 Benson and Bowman (1994 [122788]) Kd of Br (m3/kg) 0 LA0203WS831372.002 [161526] Kd of Fluorescein (m3/kg) 0 LA0203WS831372.002 [161526] Kd of 2,6-DFBA (m3/kg) 0 LA0203WS831372.002 [161526] Phase 1A: Br concentration measurements LA9910WS831372.008 [147156] Phase 1A: Fluorescein distribution (plume images) in the rock LA0302WS831372.001 [162765] Phase 1B: Br concentration measurements LA0201WS831372.008 [162766] Phase 1A: 2,6-DFBA concentration measurements LA0201WS831372.008 [162766] Table 7.3-2. Hydrological Parameters in the Analysis of the Busted Butte Phase 1A Field Test Parametera Ch1va Ch2va Ch1vb Ch2vb f 0.273 0.345 0320 0.360 kx = ky= kz (m2) 9.9x10-13 9.27x10-14 3.14x10-13 1.82x10-14 t 0.7 0.7 0.7 0.7 a (1/m) c 0.140 0.503 0.471 0.741 n c 1.538 1.427 1.332 1.200 Sr 0.03 0.07 0.07 0.07 Initial Sw 0.3 0.3 0.3 0.3 NOTE: aUZ99 calibrated properties, DTN: LB997141233129.001 [104055] bbased on USGS measured properties, DTNs: GS990308312242.007 [107185], GS990708312242.008 [109822], as reported in DTN: LB991220140160.010 [164858] cvan Genuchten (1980 [100610]) model parameter Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 226 November 2003 7.3.3.4 Validation of the Bromide Transport Model Figure 7.3-3 shows the numerically predicted distribution at the y = 0.90 m mineback face (at validation), as well as the field measurements of the Br relative concentration (obtained from samples acquired during the mineback operations). The properties that resulted in the distribution of Figure 7.3-3 are those listed in Table 7.3-3, and were provided by the calibration of the fluorescein data (Section 7.3.3.3). For the prediction of Br transport, the simulations were conducted using the fluorescein-derived calibrated parameters (without any adjustment) and the known D0 of Br. A comparison of the two concentration curves in Figure 7.3-3 indicates good agreement between predictions and observations. The model accurately reproduces the rather uniform concentrations along the x-axis that passes by the injection point, as well as the magnitude of the observed concentrations. Toward the outer extents of the horizontal axis of the plume, there appear to exist discrepancies between observations and predictions. Some of the discrepancies are consistent with the rather elongated shape of the predicted tracer plume. Part of this may be attributed to the fact that these occur in areas of steep gradients (i.e., in areas of condensation of the contour lines, when large differences occur within a short distance). An additional issue is uncertainty in the measurements, as reflected by a measured relative concentration of 2.77 (with respect to the injection concentration) in the center of the plume. While this high concentration is possible (if, for example, the sample was inadvertently allowed to dry), it may be indicative of some error in the measurements. Based on these results, the Br transport model is validated because it was based on calibrated parameters from a different data set (that of fluorescein). 7.3.3.5 Remarks and Observations For all calibrations, we adjust the following parameters: f, k, t, Kd (for known sorbing species), D0 (only if not available from Lide (1992 [166224])). For validation, the relevant calibrated parameters are used unchanged, and only the fixed parameters that uniquely describe individual transport behavior (i.e., the known D0 of Br) are different. A review of the parameters obtained in the process of calibration in Table 7.3-3 indicates that they are within a reasonable range and they are not in obvious conflict with the initial data listed in Table 7.3-2. While the possibility of a non-unique solution cannot be summarily dismissed, the calibrated solution (fluorescein) and the corresponding validated solution (of Br transport) are both mathematically correct and physically meaningful. A quantitative measure of the model validation, stemming from the comparison of field observations to model predictions, is provided in Table 7.3-9. Although there are some discrepancies, the agreement between observations and numerical predictions is sufficiently good to enhance confidence in the RTM. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 227 November 2003 DTN: LA0302WS831372.001 [162765] Figure 7.3-1. Fluorescein Plume at the y = 0.9 m Mineback Face at Borehole 3. Table 7.3-3. Calibrated Parameters of Flow and Transport from the Analysis of the Busted Butte Phase 1A Field Test Parametera Ch1v Ch2v f 0.320 0.360 kx (m2) 2.14x10-13 8.20x10-13 ky (m2) 4.14x10-13 2.82x10-13 kz(m2) 6.28x10-14 3.64x10-14 t 0.22 0.12 a (1/m) c 0.471 0.741 n c 1.332 1.200 Sr 0.07 0.07 D0 of Fluorescein (m2/s) 4x10-10 DTN: LB0308MR0060R1.008 [Output] NOTE: aAll other parameters remaining as in Tables 7.3-1 and 7.3-2. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 228 November 2003 -2.0 -1.5 -1.0 -0.5 0.0 Z (m) 1.4 1.2 1.0 0.8 0.6 0.4 0.2 0.0 X (m) 0.9 0.9 0.8 0.8 0.7 0.6 0.5 0.4 0.3 0.2 0.1 TpTpv1 Tac Test 1A Fluorescein distribution DTN: LB0308MR0060R1.009 Figure 7.3-2. Numerical Prediction of the Fluorescein Plume Using Calibrated Parameters (Busted Butte Test Phase 1A) Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 229 November 2003 -2.0 -1.5 -1.0 -0.5 0.0 Z (m) 1.4 1.2 1.0 0.8 0.6 0.4 0.2 0.0 X (m) 0.8 0.8 0.7 0.6 0.5 0.4 0.3 0.2 0.1 TpTpv1 Tac Test 1A 0.74 0.49 0.00 0.00 2.77 0.87 0.75 0.81 0.79 0.32 0.01 0.75 Bromide distribution DTN: LB0308MR0060R1.009 [Output] Figure 7.3-3. Field Measurements and Numerical Prediction of the Bromide Distribution in Busted Butte Test Phase 1A. The Solid Circles Indicate the Location of Measurements (DTN: LA9910WS831372.008 [147156]), Which Appear in the Corresponding Boxes. 7.3.4 Phase 1B Test The purpose of this modeling study was to calibrate and validate the model through the reconciliation of field measurements and predictions for breakthrough curves of tracers injected into Borehole 5 (located in the tsw39 layer, see BSC 2001 [160828], Section 6.8). Of the five tracers injected during Phase 1B of the Busted Butte test, only Br (nonsorbing) and 2,6- difluorobenzoic acid (2,6-DFBA, assumed to be nonsorbing) were used due to the poor quality of other tracer data. Of these, the 2,6-DFBA data were used for calibration, and the Br data set Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 230 November 2003 was used for validation. Additional discussion on the calibration and validation approach can be found in Section 7.3.2 of this Model Report. 7.3.4.1 Procedure and Layout An extensive discussion of the Phase 1B test can be found in BSC (2001 [160828], Section 6.8). Only a short description is presented here. In the Phase 1B field test, the tracers were injected into the lower portion of the Topopah Spring basal vitrophyre (Tptpv2 in the lithostratigraphic units; tsw39 in the UZ99 layers of the hydrogeologic units), which is a relatively low-permeability fractured rock. The design, configuration, and important dimensions of the Phase 1B test are shown in BSC (2001 [161340], Attachment VI, while more detailed information is provided in BSC (2001 [160828], Section 6.8), which depicts the injection/collection pair of boreholes 5 and 6. The tracer solution was injected at y = 1.30 m from the rock face into boreholes 5. Water samples from boreholes were collected and analyzed regularly during the injection period. The parameters describing the site conditions and test operation (and the corresponding sources) are listed in Table 7.3-1, which also includes the DTN of the data used for this validation effort. Although the Phase 1B test involved injection and collection into another set of boreholes (boreholes 7 and 8, respectively), that portion of the test is not needed for this discussion. The determination of the hydraulic parameters was a significant component of the calibration effort. Initial estimates for these parameters (listed in Table 7.3-3) were provided by two different sources: the UZ99 calibrated flow parameter set (DTN: LB997141233129.001 [104055]) and a data set based on hydraulic property measurements from field-collected core samples (DTNs: GS990308312242.007 [107185], GS990708312242.008 [109822]). 7.3.4.2 Conceptual and Numerical Model The initial (uncalibrated) input parameters of flow and transport for the simulation of the Phase 1B test are listed in Tables 7.3-1 and 7.3-4. For this 3-D numerical study, the geologic model treated the domain as a homogeneous and anisotropic unfractured rock matrix. Although this geologic layer is known to be fractured, the assumption of an unfractured rock matrix as the domain model in the simulation appears to be a valid one because the system behavior during the injections did not exhibit evidence of fracture flow. The same grid used in the simulation of Phase 1A was used here, but with the appropriate domain properties, i.e., those of the Tptpv2 unit. The DTN of the input and output files of this simulation is DTN: LB0308MR0060R1.008 [Output]. The TOUGH2 V1.11MEOS9nTV1.0 (LBNL 1999 [113943]) was used for the simulation. For this calibration and validation process, only the data from the first 100 days of the Phase 1B test were used. The reason for this selective truncation of the data set was the existence of obvious inconsistencies and errors in the data. More specifically, measured concentrations of both Br and 2,6-DFBA exhibited pronounced fluctuations and appeared to drop to zero (undetectable) levels after 100 days. This occurred at a time when injection not only proceeded uninterrupted, but also continued to do so at the same injection levels for another 90 days. If Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 231 November 2003 sampling and measurement errors are not considered a possibility, then the only reasonable explanation for such a response would be a sudden and practically infinite dilution (e.g., if very large amounts of pure water reached the sample collection point), a scenario not supported by the facts. This being the case, the data beyond 100 days were deemed unreliable, and thus they were not included in the validation. 7.3.4.3 Calibration Using the 2,6-DFBA Data The calibrated parameters obtained from the 2,6-DFBA transport model in Phase 1B of the Busted Butte test are listed in Table 7.3-5. A comparison of the measured and the numerically predicted breakthrough curves of 2,6-DFBA (based on the calibrated parameters) in Figure 7.3-4 shows a very good agreement. Note that the pattern and response of the breakthrough curves confirm the visual observation that the system did not exhibit fracture flow behavior during the Phase 1B test, thus supporting the validity of the unfractured medium approach made in the simulations. 7.3.4.4 Validation of the Br transport model The calibrated parameters (based on the 2,6-DFBA transport analysis, Table 7.3-5) were used for the prediction of the Br transport in Phase 1B of the Busted Butte test. The known D0 of Br and its Kd = 0 value were used in the validation process. A comparison of the measured and the numerically predicted breakthrough curves of Br in Figure 7.3-5 shows a very good agreement. A quantitative measure of the model validation stemming from the comparison of field observations to model predictions is provided in Table 7.3-9. The level of agreement between observations and numerical predictions enhances confidence in the RTM. 7.3.4.5 Remarks and Observations A review of the calibrated parameters in Table 7.3-5 indicates that they are within a reasonable range, and that they are not in obvious conflict with the initial data listed in Tables 7.3-1 and 7.3- 3. The only possible deviation between initial estimates and the final parameters is the need for a nonzero (although small) Kd for 2,6-DFBA. This appears to be in conflict with the conventional (though not fully supported) assumption that 2,6-DFBA is nonsorbing. Careful consideration of the chemistry of the system indicates that it is entirely possible for an organic substance to sorb onto a rock. In this case, the very low permeability of the rock allows a long contact time, making such a sorption scenario plausible, especially given the fact that organic substances are generally considered to be slow sorbers (Cameron and Klute 1977 [117172]). Additionally, the long residence time in the rock may allow a slow reaction of 2,6- DFBA (a weak acid) with minerals in the rock, a process that would result in a response consistent with an apparent sorption. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 232 November 2003 1.0 0.8 0.6 0.4 0.2 0.0 100 80 60 40 20 0 Time (days) 2,6-Difluorobenzoic Acid Test 1B Measurements Model predictions Sensitivity study Figure 7.3-4. Observed and Numerically Predicted (Calibrated) Breakthrough Curves of 2,6-DFBA in the Busted Butte Phase 1B Test. Measurements from DTN: LA0201WS831372.008 [162766]). Prediction from DTN: LB0308MR0060R1.009 [Output] 1.0 0.8 0.6 0.4 0.2 0.0 100 80 60 40 20 0 Time (days) Test 1B Bromide Measurements Model predictions Sensitivity study Figure 7.3-5. Observed and Numerically Predicted (at Validation) Breakthrough Curves of Br in the Busted Butte Phase 1B Test. Measurements from DTN: LA0201WS831372.008 [162766]), Prediction from DTN: LB0308MR0060R1.009 Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 233 November 2003 To investigate the system sensitivity to spatial resolution, Figures 7.3-4 and 7.3-5 include breakthrough curves at points 0.05 m below the collection point. It is obvious that concentrations in the vicinity of the collection points exhibit very steep gradients. This indicates that minor inaccuracies in the recording of the location of the collection point can register as a strong deviation. In light of this spatial sensitivity, the agreement between predictions and observations is deemed very good, and the observed small deviations can be easily explained. Table 7.3-4. Initial Parameters in the Analysis of the Busted Butte Phase 1B Field Test Parametera UZ99a USGSb f 0.173 0.360 kx = ky= kz (m2) 5.46x10-19 1.82x10-14 t 0.7 0.7 a (1/m) c 0.225 0.225 n c 1.612 1.612 Sr 0.29 0.29 Initial Sw 0.3 0.3 D0 of 2,6-DFBA (m2/s) 7.6x10-10 NOTE: aBased on UZ99 calibrated properties, DTN: LB997141233129.001 [104055], as reported in DTN LB991220140160.010 [164858] bBased on USGS measured properties, DTNs: GS990308312242.007 [107185], GS990708312242.008 [109822], as reported in DTN LB991220140160.010 [164858] cvan Genuchten (1980 [100610]) model parameter Table 7.3-5 Calibrated Flow and Transport Parameters from the Analysis of the Busted Butte Phase 1B Field Test Parametera Value f 0.270 kx (m2) 3.06x10-17 ky (m2) 4.06x10-17 kz(m2) 1.53x10-17 t 0.07 Kd of 2,6-DFBA (m3/kg) 1.47x10-5 DTN: LB0308MR0060R1.008 [Output] NOTE: a All other parameters remaining as in Table 7.3-1 and 7.3-3. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 234 November 2003 7.3.5 Phase 2C Test Phase 2 differs from Phase 1 in that the injection systems were designed to activate large volumes of the rock. A detailed description of this test phase can be found in BSC (2001 [160828], Section 6.8, Figure 38). The injection points in this phase were distributed in horizontal parallel planes at locations designed to test the properties of the lower Topopah Spring Tuff (Tptpv2) and the underlying Calico Hills (Tptpv1) unit. 7.3.5.1 Procedure and Layout In this study, we limit our analysis to Phase 2C, which appears to have yielded the best quality data. In Phase 2C, injection occurred in three horizontal boreholes (#18, #20 and #21), each instrumented with nine injectors located at 0.61 m intervals along the borehole. All three boreholes were located in the Tptpv2 unit. Because of the relatively short injection spacing and the long injection period (695 days), the injection can be considered uniform along the three boreholes. While this assumption introduces an inaccuracy, it is rather small, is limited to the very early stages of injection (when flow is spherical rather than the quasi-cylindrical at the later stages), and keeps decreasing over time. The implication of this uniformity is that it is possible to model the system using a 2-D grid, thus allowing higher spatial resolution. After reviewing the tracer concentrations recorded at the various collection boreholes, this study concentrated in borehole #16 because it is the closest to the horizontal plane of the three injection wells and, thus, has registered the strongest signals (in terms of tracer concentrations). Borehole #16 run was located on a horizontal plane (roughly coinciding with the Tptpv2- Tptpv1 interface) about 0.6 m below that of the injection boreholes and perpendicular to their main axis. A rather large number of tracers were injected in the injection boreholes. After reviewing the concentration data, the analysis was limited to the transport of two tracers: Br and Li. This decision was reached because the data quality of the other tracers was rather poor and often devoid of any logical pattern (possibly attributed to the sample collection method that entailed removal and reintroduction of the same collection pads in the collection boreholes, replacement of pads with new ones, and difficulty in consistently placing the pads at their exact previous location following removal for sampling and testing). The nonsorbing Br and the sorbing Li had registered rather strong and generally (but not always) consistent signals in borehole #16. In addition, having being injected as a solution of LiBr, Li and Br were intertwined and offered an additional level of checking through their very different transport properties and the need to maintain mass balance. The parameters describing the site conditions and test operation of Phase 2C (and the corresponding sources) are listed in Table 7.3-6, which also includes the DTN of the data used for the exercise. In this analysis, the Li data were used for calibration, and the Br data for validation. Additional discussion on the calibration and validation approach can be found in Section 7.3.2 of this Model Report. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 235 November 2003 7.3.5.2 Conceptual and Numerical Model Early exploratory studies indicated that, if the porous medium (unfractured rock) assumption were made, it was very difficult to reproduce the individual responses of Br or Li, and impossible to obtain a consistent joint Br and Li behavior. Using just a consistent fracture domain (i.e., limiting the study to fracture flow and transport) produced a very fast breakthrough for Br and an impossibly fast one for the sorbing Li, because of the inability to account for matrix diffusion and the lack of Li sorption. Thus, modeling the domain as a fracture-matrix system was the only alternative. A dual-permeability model was employed. The 2-D grid had a rather fine resolution ( .x = .y = 0.05 m). The initial (uncalibrated) input parameters of flow and transport for the simulation of the Phase 2C test are listed in Table 7.3-7. The DTN of the input and output files of this simulation is DTN: LB0308MR0060R1.008 [Output]. The TOUGH2 V1.11MEOS9nTV1.0 (LBNL 1999 [113943]) was used for the simulation. The initial steady-state flow field was obtained by setting an infiltration rate corresponding to 5 mm/yr at the top of the domain and running the flow simulation to steady state. Note that these runs were needed at every phase of the validation because the changes in the hydraulic parameters in the process of validation affected the flow field. For the Li calibration process, only the data for t = 337 days and t = 440 days were used. Because of the sorbing behavior of Li, the data for t<337 days were marked by very low concentrations, significant variability, and the corresponding uncertainty. Concentration data for t>440 days showed rather inconsistent behavior (attributed to issues discussed in Section 7.3.5.3). For the Br validation process, only the data for t = 125 days and t = 183 days were used. The data for t<125 days were marked by very low concentrations and significant variability (inevitable because of the relatively large component of error, inaccuracy, and heterogeneity in the early data). Concentration data for t>183 days indicated a rather uniform distribution along the well bore (as the concentration peak had already been reached), differed only in the extent of the zone that had reached the maximum concentration (not a very reliable source of information for reasons explained later), and were not very useful. A quantitative measure of the model calibration and validation stemming from the comparison of field observations to model predictions is provided in Table 7.3-9. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 236 November 2003 Table 7.3-6. Operational, Transport, and Field Parameters in the Analysis of the Busted Butte Phase 2 Field Test Parameter Value Source Injection rate (mL/hr) 120 LA0008WS831372.001 [156582] Injection duration (days) 695 LA0008WS831372.001 [156582] aL (m) - Fractures 0.1 Reasonable estimate aL (m) - Matrix 0.0 Reasonable estimate aT (m) - Everywhere 0 Reasonable estimate D0 of Br (m2/s) 2.08x10-9 Lide (1992 [166224], p. 5-111) D0 of Li (m2/s) 1.03x10-9 Lide (1992 [166224], p. 5-111) Concentration measurements LA0112WS831372.001 [157100] LA0112WS831372.002 [157115] LA0112WS831372.003 [157106] LA0211WS831372.001 [162763] Table 7.3-7. Hydrological and Transport Parameters in the Analysis of the Busted Butte Phase 2 Field Test Parameter F-TpTpv1b M-TpTpv1b F-TpTpv2 M-TpTpv2 fa 1 0.354 1 0.457 kx = ky= kz (m2) a 2.2x10-13 6.649x10-13 8.1x10-13 1.488x10-13 ta 1 0.7 1 0.7 a (1/m) a,c 21.582 0.856 14.715 0.477 n a,c 2.725 1.316 2.725 1.414 Sr a 0.01 0.06 0.01 0.13 Kd of Br (m3/kg)d - 0 - 0 Kd of Br (m)e 0 - 0 - Kd of Li (m3/kg)f - 3.5x10-5 - 8.8x10-5 Kd of Li (m)g 0 - 0 - NOTE: aDTNs: LB03013DSSCP3I.001 [162379], LB0210THRMLPRP.001 [160799] bF: fracture properties, M: matrix properties cvan Genuchten (1980 [100610]) model parameter dDTN: LA0203WS831372.002 [161526] eKd denotes surface distribution coefficient - reasonable estimate fDTN: LA9912WS831372.001 [156586], LA0203WS831372.002 [161526] gInitial estimate Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 237 November 2003 7.3.5.3 Calibration Using the Li Data The parameters from the calibration of the Li transport model in Phase 2C of the Busted Butte test are listed in Table 7.3-8. Figure 7.3-6 shows the measured and the numerically predicted Li distributions along the collection borehole. Note that the Kd values in Table 7.3-8 are consistent with previously reported data (DTN: LA9912WS831372.001 [156586], LA0203WS831372.002 [161526]). Note that the data set used for the Li-based calibration corresponds to the later part of the field test because of the sorption-induced retardation (see discussion in Section 7.3.2 of this Model Report). The Li distribution in Figure 7.3-6 shows a rather good match between predictions and measurements. The peaks, their rough locations, and the distributions are matched. For t>440 days, Li concentrations start decreasing, and then they resume increasing. This is consistent with concentrations in the pad being diluted by Li-poor water until the main Li front arrives at the collection pad (at which time concentrations begin increasing again). Note that the analysis is complicated by lack of knowledge about the sorption behavior of Li in the pad (see discussion in Section 7.3.5.4). 7.3.5.4 Validation of the Br Transport Model Figure 7.3-7 shows the measured and the numerically predicted Br distribution along the collection borehole. This distribution was obtained by using the calibrated properties (Table 7.3- 8) from the analysis of the Li data, and the known D0 and Kd (=0 m3/kg) of Br. Compared to Li, the Br distribution indicates a less diffusive behavior, attributed to the lack of sorption. Although different magnitudes are involved, the concentration distributions patterns (both observed and predicted) are consistent between the two tracers. Note that the data set used for the Br-based validation corresponds to the earlier part of the field test because of the reasons discussed in Section 7.3.2 of this Model Report. A review of the Br predicted distribution reveals that the numerical solution accurately predicts the location and magnitude of the concentration peaks, but exhibits a narrower, more focused pattern and registers deviations away from these peaks. More specifically, the measured concentration of Br is more uniform along the collection borehole axis than what the numerical simulation predicts. In other words, the measurements indicate a system that is more diffusive than the advective one indicated by the simulation. However, adjustment of permeabilities and diffusive fluxes through further calibration was not the correct approach (even disregarding the obvious danger that, to increase the diffusive component, one would have to reach rather unreasonable relations between the permeabilities of the fractures and the matrix). Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 238 November 2003 1.0 0.8 0.6 0.4 0.2 0.0 CR 9 8 7 6 5 4 3 2 X (m) Model t = 337 d t = 440 d Measurements t = 337 d t = 440 d Well #16 Lithium Observed Data from DTN: LA0201WS831372.007 [164721], Prediction from DTN: LB0308MR0060R1.009 [Output]. Figure 7.3-6. Observed and Numerically Predicted (Calibrated) Breakthrough Curves of Li in the Busted Butte Phase 2C Test. 1.0 0.8 0.6 0.4 0.2 0.0 CR 9 8 7 6 5 4 3 2 X (m) Measurements t = 125 d t = 183 d Model t = 125 d t = 183 d Well #16 Bromide Observed Data from DTN: LA0112WS831372.003 [157106], Prediction from DTN: LB0308MR0060R1.009 [Output]. Figure 7.3-7. Observed and Numerically Predicted Breakthrough Curves of Br in the Busted Butte Phase 2C Test (at Validation). Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 239 November 2003 The discrepancy problem had to be recast under the light of what the measurements meant. Use of the collection pad introduced a system with drastically different properties than the host rock, characterized by high permeability, porosity, irreducible water saturation, and capillary pressure. Tracer carrying fracture flow first reaching the collection pad arrives at distinct points (the points where the fractures intercept the borehole), and is quickly redistributed on the initially dry pad, thus leading to an apparently more uniform distribution than the one in the overlying rock. This is more evident on the left lobe of the distribution, corresponding to borehole #18. Being relatively far from the other boreholes, it receives no cross contribution of fracture flow from injection into the other boreholes, and the distribution on the pad shows a more diffusive pattern than what the simulation predicts. In other words, the collection pads skew the data in a direction that indicates a more diffusive system. This being the case, it is expected that the most accurate data will be at early times and roughly under the injection boreholes (not necessarily directly underneath, as the fracture density, distribution, and orientation are anything but known a priori). Thus, it is expected that matching the peaks is a sufficient indication of validation. Heterogeneity of fracture property and distribution is also a substantial issue at this small scale, and it is possible to create the appearance of enhanced diffusion if the fracture frequency is locally higher (e.g., on the left lobe of the curve). If the effect of the pad is as described above, it is expected that it will lead to decreasing peak concentrations and more uniform spatial distribution as time advances. This is because after the arrival of tracers in the faster flowing fractures, the contributions of slower-flowing fractures (with more interactions with the matrix) and of the matrix itself release water of lower concentration that is mixed and redistributed on the pad, and leads to temporarily lower concentrations resulting from dilution. Of course, once the solute front crosses the matrix, concentrations are expected to rise again. A review of the Br measurement data indicates this exact pattern. An attempt was made to simulate the effect of the pad. However, these simulations indicated that its behavior is sensitive to the initial conditions (whether it is an old one or a new replacement), its location (whether the old pad is placed at the exact same location from where it was removed), its wettability properties vis-à-vis the host rock, etc. This being the case, and given the uncertainties and the lack of an adequate description of the fracture-matrix system at the gridblock scale of this problem (0.05 m)—after all, the mathematical description of the matrix-fracture system is averaging even at this scale, and can be quite different from the actual system –no further attempt was made to refine the solution. The solution currently reflects concentration predictions in the overlying rock and not in the pad. A quantitative measure of the model validation stemming from the comparison of field observations to model predictions is provided in Table 7.3-9. The level of agreement between observations and numerical predictions enhances confidence in the RTM. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 240 November 2003 Table 7.3-8. Calibrated Flow and Transport Parameters from the Analysis of the Busted Butte Phase 2C Field Test Parametera F-TpTpv1 M-TpTpv1 F-TpTpv2 M-TpTpv2 f 1 0.354 1 0.060 kx (m2) 3x10-13 1.3x10-13 1.96x10-12 1.2x10-13 kz(m2) 10-13 8x10-14 7.1x10-13 8x10-14 t 2 0.9 2 0.654 Kd of Li (m3/kg) - 5.5x10-4 - 9.3x10-4 Kd of Li (m)b 2.5x10-6 - 4.3x10-6 - DTN: LB0308MR0060R1.009 [Output] NOTE: aAll other parameters remaining as in Tables 7.3-6 and 7.3-7 bKd denotes surface distribution coefficient in fractures Table 7.3-9. Comparison of Numerical Solutions to Field Tests Field Test # Difference* 1A Bromide <50% (concentration, distribution) 1B Bromide: <50% (breakthrough) 2C Bromide: <50% (breakthrough) NOTE: * The 50% agreement refers to either concentration comparison or arrival time comparison. 7.3.5.5 Remarks and Observations A review of the calibration parameters in Table 7.3-8 indicates that they are within a reasonable range, and that they are not in obvious conflict with the initial data listed in Tables 7.3-6 and 7.3-7. Note that the distribution parameter Kd of Li is within the experimentally determined range (DTN: LA9912WS831372.001 [156586], LA0203WS831372.002 [161526]). A surface sorption distribution coefficient had to be considered for the fractures—which is consistent with the rather strong sorbing behavior of Li. Given the degree of match between predictions and observations and the uncertainties involved in this study, the level of agreement between observations and numerical predictions is deemed acceptable and supports the claim of model validation. 7.3.6 Post-Development Validation Using Chloride and 14C Field Data Post-development validation of the RTM on a larger scale is provided by the analysis of field measurements of 14C ages in the gas samples, and of pore water Cl concentrations in the ESF. These studies and the corresponding validation-supported analysis were conducted using the T2R3D numerical model, and are discussed in detail in BSC (2003 [163045]), Sections 7.5 and 7.8, respectively. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 241 November 2003 7.4 VALIDATION OF THE MATRIX DIFFUSION MODEL BY CORROBORATION WITH DATA FROM ALCOVE 8/NICHE 3 FAULT TEST Matrix diffusion can play an important role in radionuclide exchange between the fractures and the rock matrix (Moridis et al. 2003 [161902], p. 255). This process transfers a radionuclide into the matrix (where water flow is slow), thus removing them from fast fracture flow. The diffusive flux is a function of the concentration gradient at the corresponding fracture-matrix interface, fracture interface area, and molecular diffusion coefficient (Section 6.2 of this report). The matrix diffusion model is an important component of the UZ Transport Model. This section validates the Matrix Diffusion Model by corroboration with data from the Alcove 8/Niche 3 fault test. The criteria for validation is that the predicted breakthrough curves collected at Niche 3 qualitatively agree with data and predicted tracer transport time is not longer than observed (BSC 2002 [160819], Section I-2-1-2, Attachment I, p. 26). As demonstrated in the subsections below discussing modeling results, the criteria are met. This modeling activity is also documented in a Scientific Notebook by Wang (2003 [164021], SN-LBNL-SCI-215-V1, pp. 107-109). The Alcove 8/Niche 3 fault test is described in Section 7.4.1. Data from this test include both hydrologic data and tracer data. The numerical model of the test is described in Section 7.4.2. Rock properties in the model were calibrated to the hydrologic data in two calibration runs as described in Section 7.4.3.1. Tracer transport with matrix diffusion was then simulated as described in Section 7.4.3.2, where it is demonstrated that the validation criteria are met. 7.4.1 Field Observations This section highlights experimental observations used in this model analyses. The test was carried out in the upper lithophysal and middle nonlithophysal subunits in the Yucca Mountain UZ. These geological subunits correspond to model layers tsw33 and tsw34, respectively, in the site-scale model for the Yucca Mountain UZ. The tsw33 has some lithophysal cavities that may intersect fractures. Liquid water with and without tracers was released at the floor of an alcove along the fault (about 5 m long (DTN: GS020508312242.001 [162129])) within tsw33. Seepage from the fault into a niche and tracer concentrations of seeping liquid as a function of time were monitored. The niche is located within tsw34, about 20 m below the floor of the alcove; the interface between tsw33 and tsw34 is about 15 m below the floor of the alcove (DTN: LB0301N3SURDAT.001 [162130]). A water-pressure head of 2 cm was applied at an infiltration plot along the fault at the alcove. The plot consists of four trenches that have different infiltration rates as a result of subsurface heterogeneity along the fault. Figure 7.4-1 shows the total infiltration rate as a function of time (DTNs: GS020508312242.001 [162129] and GS020908312242.002 [162141]; Wang 2003 [164021], SN-LBNL-SCI-215-V1, p. 107). For simplicity, our model considers the uniformly distributed infiltration rate along the infiltration plot to be consistent with the use of the uniform property distribution in the site-scale model of the Yucca Mountain UZ. One consideration in our modeling study is to evaluate approaches used in the site-scale model. Considerable temporal variability of infiltration rate occurred during the test, as a result of infill materials within the fault just below the infiltration plot (Figure 7.4-1). In other words, the effective Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 242 November 2003 permeability of the fault just below the plot changed with time. It is also expected that most portions of the fault and the surrounding fractures away from the plot would still be unsaturated, although capillary pressure at the plot was positive during the test. Based on these observations, total infiltration rate (instead of a pressure head of 2 cm) was used as the boundary condition in our model. Seepage from the fault into the niche was measured during the test, with a number of trays used to cover the areas where seepage might occur. Seepage was found to be highly spatially variable. The total seepage rate as a function of time is given in Figure 7.4-2 (DTN: LB0303A8N3LIQR.001 [162570]). Several boreholes were installed around the niche. Water arrival times at these boreholes were monitored through electrical resistivity probes. Figure 7.4-3 shows average water travel velocities determined from the arrival times from two boreholes just above the ceiling of the niche (DTN: LB0303A8N3LIQR.001 [162570]). The fault is about 2 m from the borehole collars in Figure 7.4-3 (DTN: LB0303A8N3LIQR.001 [162570]). Note that relatively uniform water-travel-velocity distribution within and near the fault was observed from these two boreholes. After 209 days, two tracers with different molecular diffusion coefficients (Br and PFBA) were introduced into infiltrating water at the infiltration plot. Tracer concentrations in three of the trays (at the niche) capturing seeping water from the fault were measured (DTN: LB0303A8N3LIQR.001 [162570]). For technical reasons, seepage rates corresponding to these three trays were not measured during the period of tracer concentration measurement. In this study, a flux-averaged breakthrough curve (concentration as a function of time) from these trays was used to represent average breakthrough curve for all trays at the niche where seepage was captured (Wang 2003 [164021], SN-LBNL-SCI-215-V1, pp. 108–109). A constant flux value for each of the three trays was used for calculating the flux-averaged breakthrough curve shown in Figure 7.4-4. The constant flux values for the three trays were determined as the averaged value over 56 days before tracers were introduced. This flux-averaged breakthrough curve is comparable to simulation results. 7.4.2 Numerical Model A numerical model, generated with a software routine Smesh.f V1.0 (LBNL 2002 [162142]), was developed for the fault test site to compare the simulation results with the relevant field observations. While comparison results will be presented below in Section 7.4.3, here we will focus on schemes used for developing the numerical model. A three-dimensional numerical grid was constructed for simulating the fault test (Figure 7.4-5). The fault was represented as a vertical fracture, and surrounding fractured rock is approximated as a dual-continuum system consisting of overlapped, interacting fracture and matrix continua. Global water flow and solute transport are allowed to occur in both continua. Figure 7.4-5 shows a cross section of the grid within the fault. The thickness of the grid in the direction perpendicular to fault walls is 3 m along each side of the fault. The fracture frequency used for generating the dual-continuum grid is 1.03 m-1 for tsw33 (determined from the fracture map at the alcove floor) and 1.72 m-1 for tsw34 (determined from the fracture map at the ceiling of the niche) (DTN: GS030108314224.001 [162131]; Wang 2003 [164021], SN-LBNL-SCI-215-V1, pp. 111-112). As shown in Figure 7.4-5, within a cross section of the grid along the fault, the Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 243 November 2003 grid spacing is 10 cm just above the ceiling of the niche, enabling the seepage process to be accurately simulated. Grid spacings in the direction perpendicular to the fault are 0.024 m, 0.168 m, 0.456 m, 0.756 m and 1.44 m, respectively. The smallest spacing is adjacent to the fault, so that water imbibition and tracer diffusion into the fractured rock from the fault can be accurately captured. Cross sections parallel to the fault walls have identical grid meshes (Figure 7.4-5) for different distances from the fault. The niche is represented by an opening at the bottom of the grid (Figure 7.4-5), with the geometry of the opening determined from the survey data of the niche near the fault. Note that this is only an approximation of the geometry of the test site; a three-dimensional geometry of the niche with an underground tunnel connected to the niche is not incorporated into the model. However, since our main concern is flow and transport processes within the fault, this geometry representation should be adequate. Temporally variable inflow rates are imposed on the top boundary, corresponding to the infiltration plot at the alcove floor. The side boundary corresponds to zero-flow (in the direction perpendicular to the simulation domain) conditions. The niche wall boundary is modeled by a zero capillary-pressure condition, representing capillary barrier effects (Birkholzer et al. 1999 [105170]). The bottom boundary was assigned a constant matrix saturation of 0.85, which is consistent with field observations under ambient conditions (Flint 1998 [100033]). Also based on field observations of Flint (1998 [100033]), matrix saturations are initially assigned to be 0.72 for tsw33 and 0.85 for tsw34. Other initial conditions for the rock mass within the model domain are absence of solutes and low water saturation (1.05E-2) in both the fractures and the fault. Rock properties used in model simulations are presented in the next section. Model calibration was performed using an inverse modeling code iTOUGH2 V4.0 (LBNL 1999 [139918]). The model calibration is defined herein as the adjustment of rock hydraulic parameters to make simulation results match the corresponding data. The goodness of match is measured using the standard least-squares approach, which minimizes the sum of the squared residuals weighted by the inverse of variance of the data. T2R3D V1.4 (LBNL 1999 [146654]), was used for modeling tracer transport. 7.4.3 Model Simulations and Discussions The numerical model was first calibrated against the seepage and water-travel-velocity data to obtain the calibrated rock properties and the corresponding water flow field. Then, tracer transport simulations with different transport parameters were carried out to evaluate the effects of matrix diffusion and other related processes on solute transport in the fault. 7.4.3.1 Calibration of Rock Properties using Seepage-Rate Data and the Average Water-Travel-Velocity Data Both fracture and matrix properties were assumed to be homogeneous within each geological subunit (tsw33 and tsw34). Fault properties were assumed to be homogeneous within both units. This is based mainly on the following three considerations: (1) Consideration of the heterogeneity within each subunit would introduce a large number of rock properties that need to be determined by more data than was available from the test site; Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 244 November 2003 (2) As previously indicated, these treatments have been used by the site-scale model of the Yucca Mountain UZ. It is of interest to examine how well this simple representation of subsurface heterogeneity can be used to model the fault test. (3) A recent study by Zhou et al. (2003 [162133]) indicates that flow and transport in the Yucca Mountain UZ are mainly determined by large-scale heterogeneity, characterized by property differences between different geological units, rather than by property variability within a geological unit. Rock hydraulic properties needed as inputs into the model include fracture and matrix permeabilities, fracture and matrix porosities, fault aperture and permeabilities, van Genuchten (1980 [100610]) parameters (for matrix, fractures, and the fault), and the parameter of the active fracture model, ., for fractures (DTNs: LB997141233129.001 [104055], LB980901233124.101 [136593], LB990861233129.001 [110226] and LB990501233129.001 [106787]; Wang 2003 [164021], SN-LBNL-SCI-215-V1, pp. 114–115). Because fracture van Genuchten parameters for tsw33 and tsw34 are similar, a simple average of these parameters was used as the corresponding parameters for the fault. The averaged k/f (where k is fracture permeability and f is the corresponding fracture porosity) was calculated as fault permeability. Note that in this section 7.4.3, fracture porosity refers to the volume fraction of fractures in the entire rock volume, rather than the void space within the fracture. Also, note that because there is no matrix in the fault in our model (or f =1), the weighted k/f (rather than weighted k) is employed for estimating fault permeability. The aperture of the fault was estimated as the average of fracture apertures of the two subunits. Note that the active fracture model was developed for fracture networks rather than for a single fracture. Consequently, the active fracture model does not apply to the fault here. In fact, most of the parameter values mentioned above and given in Table 7.4-1 are not site specific for the fault test site. These values were used as initial guesses for model calibration against the seepage rate and water-travel-velocity data observed from the fault test. To reduce the number of variables in model calibration (or inverse modeling), parameters expected to significantly affect simulated water travel time and seepage rate were varied in the calibration, while other parameters were kept unchanged. The varied parameters were fracture and fault permeabilities, fracture porosity, fault aperture, and fracture and fault van Genuchten values. Infiltration-seepage processes in the fault and the surrounding fractured rock were determined by several mechanisms. Liquid water applied at the alcove floor (Figure 7.4-5) flowed first into the fault and then into fractured networks connected to the fault. Matrix imbibition occurred at interfaces between fractures and the matrix and between the fault and the matrix. When water arrived at the intersection between the fault and the niche, it could not immediately seep into the niche until the capillary pressure became zero because of capillary barrier effects (Philip et al. 1989 [105743]; Birkholzer et al. 1999 [105170]). The capillary barrier can divert flow away from the opening, resulting in only a portion of the water arriving at the niche ceiling actually seeping into the niche. Water travel time was determined by fracture porosity, fault aperture, and the matrix imbibition process. Figure 7.4-6 shows a comparison between seepage-rate data and the simulation result from a model calibration (Run #1) without considering the water-travel-velocity data. In this calibration Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 245 November 2003 run, fracture porosity and fault aperture were not varied. A fairly good match was obtained (Figure 7.4-6); however, water travel velocity is significantly overestimated (Figure 7.4-7). Water travel velocities were calculated from water arrival times at locations about 1 m above the middle of the opening in Figure 7.4-5. The travel time was defined as the time when fault or fracture saturation was increased from the initial value of 1.05E-2 to 1.06E-2. This comparison implies that seepage rate as a function of time may be mainly controlled by rock properties near seepage locations (influence zone of capillary barrier (Liu et al. 2002 [160230])). On the other hand, water travel velocities are determined by rock properties from the infiltration plot to the locations where water travel velocities are monitored. Table 7.4-2 gives the calibrated properties obtained from Run #1. The overestimation of the water travel velocities may result from the following: (1) some cavities in tsw33 are connected to fractures and could contribute to increasing the storage in the fracture continuum; (2) in reality, the fault is a zone rather than a single fracture. The effective aperture from this zone may be much larger than the assumed aperture value for the fault (Table 7.4-1). Neither of these factors was considered in Run #1 (first calibration). Taking these factors into consideration, the new calibration (Run #2) allowed both fault aperture and fracture porosity in tsw33 to be varied. The resultant values are 3 cm for fault aperture and 0.066 for fracture porosity of tsw33 (Table 7.4-3). While the actual width of the fault zone is unknown, the estimated equivalent fault aperture (3 cm) is considered to be acceptable. The estimated fracture porosity is consistent with those estimated from water release tests performed in the same geological unit (BSC 2001 [158463], Section 6.11.3.1, p. 246). Figure 7.4-7 shows a comparison among calculated water travel velocities from two calibration runs and the velocity data observed from the fault test. The simulated water travel velocities from Run #2 are much closer to the observed data than those from Run #1 (especially near the fault). However, the water travel velocities away from the fault are still overestimated. One possible explanation is that matrix imbibition from fractures above the niche were underestimated because the dual-continuum approach considerably underestimates the pressure gradient near a fracture matrix interface during transient flow conditions (Pruess and Narasimhan 1985 [101707]). While this problem could be resolved with the multiple interacting continua model of Pruess and Narasimhan (1985 [101707]), the computational intensity of the inverse model problem under consideration would be significantly increased. Note that a model calibration involves a great number of forward simulation runs. Considering that (1) the transient flow effects would be considerably reduced at later time of the test and that (2) our focus here is on flow and transport within and near fault, simulated flow field and calibrated rock properties from Run #2 were used for simulating tracer transport at the test site. Figure 7.4-8 also shows a comparison between simulated seepage rates as a function of time (Run #2) and field observations. The match is reasonable. 7.4.3.2 Prediction of Tracer Transport with Matrix Diffusion Tracer transport within the fault is controlled by several processes, including advection, diffusion into the matrix blocks (matrix diffusion), mass exchange between the fault and the surrounding fracture networks, and dispersion. Our special attention in this study is given to evaluating the relative importance of matrix diffusion. To do so, we used the flow field obtained from Run #2 to simulate tracer transport processes and compare simulation results with field Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 246 November 2003 observations (Figure 7.4-4). The breakthrough curve is obtained from the output of T2R3D V1.4 (LBNL 1999 [146654]) using a software routine Bkread.f V1.0 (LBNL 2002 [162143]). Two conservative tracers with different molecular diffusion coefficients (2.08 E-9 m2/s for Br (Lide 2002 [160832], p. 5-96) and 7.60 E-10 m2/s for PFBA (Benson and Bowman 1996 [122788])) were used in the fault test. Based on analyses of the relevant diffusion experiment results for tuff matrix samples in the Yucca Mountain UZ, Moridis et al. (2003 [161902], Table 1) reported that the tortuosity factor for the tuff matrix could be approximated by the corresponding matrix porosity. As a consequence, the average matrix porosity for tsw33 and tsw34 (0.13) was used as the tortuosity factor. The effective diffusion coefficient for the matrix diffusion process is the product of the molecular diffusion coefficient and tortuosity factor. Figure 7.4-9 compares simulated breakthrough curves at the niche with the observed data. In this simulation, the dispersivity is assumed set to zero. (The relative importance of the dispersion will be discussed below.) Since the diffusive flux from the fault to the matrix is proportional to the product of the tortuosity factor and the fault-matrix interface area, changes in the interface area for a given tortuosity factor are equivalent to changes in tortuosity factor for a given interface area. For simplicity, the tortuosity factor value was changed in actual simulations, but the numerical grid (defining the interface area) was kept unchanged. Note that changes in the interface area should not significantly alter the flow field during the period of the tracer test. Tracers were introduced into infiltrating water at about 200 days after infiltration started, resulting in the matrix near the fault being almost saturated during the tracer test and the matrix imbibition being insignificant. As shown in Figure 7.4-9, the simulated breakthrough curve exhibits a much larger concentration peak values but only slightly earlier arrival times for the tracer peaks. This meets the specified validation criteria. 7.4.3.3 A Further Discussion of Matrix Diffusion Effects Although the Matrix Diffusion Model has been validated sufficiently for the purposes of License Application, it is of interest to study its sensitivity to the interface area between fault and matrix. The need to increase interface areas between fractures (or faults) and the matrix in matching the field observations of tracer transport in fractured rock has been recently reported by several researchers. Shapiro (2001 [162132]) reported an interpretation of concentration measurements for tritium and dichlorodifluoromethane collected from a glacial drift and fractured crystalline rock over 4 km2 in central New Hampshire. He found that the effective diffusion coefficient at the kilometer scale is at least three orders of magnitude greater than laboratory estimates of diffusion in crystalline rock. Neretnieks (2002 [162140]) presented comparisons between several analytical solutions with tracer test results at the Äspö site and reported a need for a factor 30–50 times larger for the fracture-matrix interface area than expected. He also indicated that nine other research groups reached a similar conclusion in their interpretation of the same test data set. Our results in this study are consistent with these previous findings. Several mechanisms regarding the increase in the interface area have been reported in the literature. They include (1) advective mass exchanges from high-permeability fractures to lowpermeability fractures (Shapiro 2001 [162132]), (2) diffusion into stagnant water zones (Neretnieks 2002 [162140]), and (3) enhanced fracture-matrix interface areas for fractures with small-trace lengths that do not contribute to global flow and are not considered in the survey data Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 247 November 2003 (and therefore in the numerical grid). In addition to these potential mechanisms, two other factors also contribute to the increase in the interface area. First, in the relevant analytical and numerical solutions to tracer transport, fracture walls are generally assumed to be flat. However, it is now well known that fracture walls are rough and characterized by fractal geometry (National Research Council 1996 [139151]). Consequently, the actual interface areas between fractures (and faults) and the matrix are larger than what are calculated using flat fracture walls. Second, a fault zone may include a great number of crushed matrix blocks that have smaller sizes than the fracture spacing in a nonfault zone. These crushed matrix blocks can make a significant contribution to the matrix diffusion within the fault, but are not considered in our numerical grid, where the fault is simply treated as a vertical fracture. To compensate for the effects of these mechanisms mentioned above, the model’s fault-matrix interface areas obviously need to be increased. Also shown in Figure 7.4-9 is a simulation with the interface area increased by a factor of 45, which is within the range reported by Neretnieks (2002 [162140]). With this adjustment, the simulations match the data more closely. Although simulation results with the increased interface area reasonably match the observed data (Figure 7.4-9), the concentration difference at a given time for the two tracers is generally overestimated by the model. One plausible explanation is that the crushed matrix blocks within the fault zone have much smaller sizes than the fracture spacing. This, however, is not considered in our model, in which the matrix block size is characterized by fracture spacing. The smaller sizes correspond to shorter times needed by the equilibrating tracer concentration at the center and outer surface of a matrix block, reducing the difference between the effects of matrix diffusion on overall solute transport behavior for different molecular-diffusion coefficients. This can be further illustrated by an extreme case: an infinitely small block size within the fault and without mass exchange between the fault and nonfault fractured rock. In this case, the concentrations of the matrix block within the fault can be equilibrated simultaneously with those at the outer surface of the block for two tracers with different molecular diffusion coefficients. Consequently, although the existence of this kind of matrix block can still significantly retard tracer transport within the fault, identical breakthrough curves should be observed at Niche 3 for the two tracers. This issue was not further explored in the current modeling study because the matrix block-size distribution within the fault cannot be independently estimated or observed. Compared with matrix diffusion, the macrodispersion process is not considered to be significant within the fault for this particular test. Field measurements indicate that water travel-velocity distribution is quite uniform within and near the fault (Figure 7.4-3), whereas macrodispersion results from variability in water velocity. This is further confirmed by observed breakthrough curves from three flow paths within the fault. They have arrival times similar to peak concentrations (DTN: LB0303A8N3L1QR.001 [162570]). These experimental observations are consistent with the findings from our model analyses: the observed data were very difficult to match when a considerable degree of dispersion was included in the model. For example, Figure 7.4-10 shows simulated breakthrough curves with a longitudinal dispersivity value of 1 m and a transverse dispersivity value of 0.1 m (and with the increased fault-matrix interface area), compared to results in Figure 7.4-9 (without considering dispersion). Larger dispersivity values generally correspond to earlier arrival times of peak concentrations and to a larger difference between these peak concentrations for the two different tracers. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 248 November 2003 The Yucca Mountain site (where the fault test was carried out) is the proposed geological disposal site for high-level nuclear waste in the United States. Radionuclide transport from the repository horizon to the water table is a key factor affecting the performance of the proposed repository. The study results herein have some important implications for the performance assessment of the proposed site. Matrix diffusion has been identified as a key mechanism for retarding the radionuclide transport in both unsaturated and saturated fractured rock (e.g., Bodvarsson et al. 2001 [160133]; Neretnieks 2002 [162140]). The enhancement of the fracture (fault)-matrix interface area (or effective matrix diffusion coefficient) seems to be common for matching field-scale solute transport observations, as suggested by this study and previous studies (Shapiro 2001 [162132]; Neretnieks 2002 [162140]). Some physical explanations for this enhancement are available, as discussed in Section 7.4.3.2. The current site-scale model for the Yucca Mountain UZ does not consider the effects of this enhancement. Consequently, the performance of the proposed repository, estimated based on the site-scale model, may be underestimated. The other related issue is the effects of cavities (existing in several geological layers at the Yucca Mountain site) on water flow and radionuclide transport processes. One may intuitively expect the cavities connected to fractures to act as capillary barriers under unsaturated conditions, because the cavity openings are much larger than fracture apertures. However, both this study and analyses of water-release tests performed in the related geological units at the Yucca Mountain site suggest that cavities are accessible by water within fracture networks, and therefore are retarding the downward water flow and radionuclide transport processes. This is also supported by field observation that mineral coatings exist in many cavities (BSC 2002 [160247], Section 6.10). The coating is a signature for liquid-water flow paths. Although the cavity openings are larger than fracture apertures, the roughness of cavity walls could result in film flow (along cavity walls) from fractures to the cavities (Tokunaga and Wan 1997 [139195]). The effects of cavities are also not considered in the site-scale model for the Yucca Mountain UZ. This omission would result in further underestimating the performance of the proposed repository. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 249 November 2003 Table 7.4-1. Uncalibrated Rock Properties tsw33 tsw34 Rock property Fault Fracture Matrix Fracture Matrix Permeability (m2) 4.34E-11 5.5E-13 3.08E-17 0.35E-13 4.07E-18 Porosity 1.00 6.6E-3 0.154 1.E-2 0.11 Fracture frequency (1/m) 1.03 1.5 Fracture aperture (m) 1.12E-3 1.49E-3 1.14E-3 Active fracture model parameter . 0.0 0.41 0.41 van Genuchten a (Pa-1) 1.0E-3 1.46E-3 2.13E-5 5.16E-4 3.86E-6 van Genuchten m 0.608 0.608 0.298 0.608 0.291 DTNs: LB997141233129.001 [104055], LB980901233124.101 [136593], LB990861233129.001 [110226] and LB990501233129.001 [106787] (Also see Wang 2003 [164021], SN-LBNL-SCI-215-V1, pp. 114-115) Table 7.4-2. Rock Properties Calibrated from Seepage Rate Data (Run #1) Rock property Fault tsw33 tsw34 Fracture Permeability (m2) 6.67E-11 8.93E-13 3.16E-14 Fracture van Genuchten a (Pa-1) 1.15E-3 1.67E-3 4.59E-4 Source: DTN: LB0303A8N3MDLG.001 [162773] NOTE: All other rock properties are the same as those in Table 7.4-1. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 250 November 2003 Table 7.4-3. Rock Properties Calibrated from both Seepage Rate and Water Travel Velocity Data (Run #2) Rock property Fault tsw33 tsw34 Fracture Permeability (m2) 1.12E-10 1.23E-12 5.01E-13 Fracture Porosity 0.066 Fracture aperture (m) 0.03 Fracture van Genuchten a (Pa-1) 1.24E-3 2.19E-3 1.09E-3 Source: DTN: LB0303A8N3MDLG.001 [162773] NOTE: All other rock properties are the same as those in Table 7.4-1. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 251 November 2003 0 100 200 300 400 Time (d) 0 50 100 150 200 250 300 350 400 450 Infiltration Rate (L/d) DTNs: GS020508312242.001 [162129] and GS020908312242.001 [162141] Figure 7.4-1. Infiltration Rate as a Function of Time 0 100 200 300 Time (day) 0 10 20 30 40 50 60 70 80 90 100 Seepage rate (L/day) DTN: LB0303A8N3LIQR.001 [162570] Figure 7.4-2. Total Seepage Rate as a Function of Time. Data for 3/6/2001 through 1/28/2002. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 252 November 2003 0 1 2 3 4 5 6 Distance from Collar (m) 0 0.25 0.5 0.75 1 1.25 1.5 1.75 2 Velocity of Wetting Front (m/d) DTN: LB0303A8N3LIQR.001 [162570] Figure 7.4-3. Water Travel Velocity Data at two boreholes (9 and 10) 0 50 100 150 Time (day) 0 0.1 0.2 0.3 0.4 0.5 Relative Concentration Br PFBA DTN: LB0303A8N3LIQR.001 [162570] (Also see Wang 2003 [164021], SN-LBNL-SCI-215-V1, p.108-109) Figure 7.4-4. Observed Flux-Average Breakthrough Curve Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 253 November 2003 0 10 20 30 x (m) -20 -15 -10 -5 0 z (m) small plot Niche 3 Alcove 8 TSw33 TSw34 For illustration purposes only. Figure 7.4-5. A Cross Section View of 3-D Numerical Grid 0 100 200 300 Time (day) 0 10 20 30 40 50 60 70 80 90 100 Seepage rate (L/d) data simulation Source: DTN: LB0303A8N3MDLG.001 [162773] Figure 7.4-6. A Comparison between Simulated Seepage Rates as a Function of Time (Run #1) and Field Observations Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 254 November 2003 0 1 2 3 4 5 6 Distance from Collar (m) 0 0.25 0.5 0.75 1 1.25 1.5 1.75 2 Velocity of Wetting Front (m/d) Data Run #1 Run #2 Source: DTN: LB0303A8N3MDLG.001 [162773] Figure 7.4-7. A Comparison among Calculated Water Travel Velocities from Two Calibration Runs and the Velocity Data Observed from the Fault Test 0 100 200 300 Time (day) 0 10 20 30 40 50 60 70 80 90 100 Seepage rate (L/d) data simulation Source: DTN: LB0303A8N3MDLG.001 [162773] Figure 7.4-8. A Comparison between Simulated Seepage Rates as a Function of Time (Run #2) and Field Observations Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 255 November 2003 0 25 50 75 100 Time (day) 0 0.05 0.1 0.15 0.2 0.25 0.3 0.35 0.4 0.45 0.5 0.55 0.6 0.65 0.7 0.75 Relative Concentration data (Br) Data (PFBA) Original interface area (Br) Original interface area (PFBA) Increased interface area (Br) Increased interface area (PFBA) Source: DTN: LB0303A8N3MDLG.001 [162773] Figure 7.4-9. Comparisons between Simulated Breakthrough Curves at the Niche for Two Different Fault-Matrix Interface Areas and the Observed Data. 0 25 50 75 100 Time (day) 0 0.05 0.1 0.15 0.2 0.25 0.3 0.35 0.4 0.45 0.5 0.55 0.6 0.65 0.7 0.75 Relative Concentration data (Br) Data (PFBA) Simulation (Br) Simulation (PFBA) Source: DTN: LB0303A8N3MDLG.001 [162773] Figure 7.4-10. Comparisons between Simulated Breakthrough Curves (Considering Dispersion) at the Niche for the Increased Fault-Matrix Interface Areas and the Observed Data. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 256 November 2003 7.5 VALIDATION OF THE SORPTION MODEL The validation approach and criteria are discussed in the TWP (BSC 2002 [160819], Section I-2- 1-3), which stipulates that: The sorption model is linear equilibrium, defined by Kd values, between sorbed and aqueous species in the UZ. … Kd values will be provided by comparison with laboratory data from the TDMS. This comparison constitutes validation of the sorption model by Method 1, corroboration with experimental data. … Kd values will be justified by comparison with laboratory data from the TDMS. This comparison constitutes validation of the sorption model by Method 1, corroboration with experimental data. The criterion for validation is that Kd values calculated from experimental data be within a factor of 3 of the Kd values used in the model. The model will also be corroborated by comparison with published data. The acceptance criterion, as stated in the TWP (BSC 2002 [160819]), is inadequate in this case because it presupposes estimation of a single Kd from the analysis of the experimental data. A far more rigorous approach is used in the Kd analysis of this Model Report (discussed in Attachments I and II), which aims to (a) identify the type and develop the parameters that describe the statistical distribution of Kd based on data from a number of samples belonging to the same general hydrogeologic unit, and (b) provide representative Kd estimates from these statistical distributions. Validation is provided by (a) confirmation that the analysis of the laboratory data (described in Attachments I and II) supports the linear equilibrium sorption model (onto which the Kd concept is based) for a given radionuclide and rock, and (b) using Kd values that are within the range determined from these laboratory experiments and correspond to the expected values (as estimated from the statistical distributions). Confirmation of the validity of the linear equilibrium sorption model is provided in Attachments I and II. Table 6.5-1 presents the Kd range and the types of the Kd probability distributions of the various radionuclides in the UZ rocks. The Kd values used in the model (indicated by boldface numbers) are also listed in this table, and are shown to be within the measured range and to coincide with the median Kd value from the statistical distributions. Thus, the sorption model is validated. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 257 November 2003 7.6 VALIDATION OF THE COLLOID TRANSPORT MODEL The validation approach and criteria are discussed in the TWP (BSC 2002 [160819], Section I-2- 1-4), which indicates that: The Colloid Transport Model was judged in the Model Validation Status Report (BSC 2001 [156257], p. 28) to have been validated satisfactorily for its purposes in CRWMS M&O (2000 [122799]). Therefore, no further validation of the Colloid Transport Model is planned. However, some additional validation was performed in the Model Report. This additional validation of the colloid transport model is included in the validation of the RTM, and is discussed in Sections 7.2.1.5 (Method 2) and 7.2.2 (Method 3) of this Model Report. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 258 November 2003 7.7 VALIDATION OF THE ACTIVE FRACTURE MODEL WITH MATRIX DIFFUSION Validation of the Active Fracture Model (AFM) with matrix diffusion for transport was planned in Section I-2-1-5 of the TWP (BSC 2002 [160819]). The TWP outlines the required level of confidence, validation approaches, and validation criteria. The validation for the AFM with Matrix Diffusion combines two parts: .. The AFM, already validated for its intended use to represent site-scale flow, in a separate Model Report (BSC 2001 [158726]). .. The Matrix Diffusion model, already validated for saturated or nearly-saturated conditions in Section 7.4 of this report. The resulting model for AFM with Matrix Diffusion is intended to represent transport of radionuclides in the UZ for in situ saturation conditions (high saturation in the rock matrix, and low saturation in the fractures). The processes represented by the AFM and Matrix Diffusion are intertwined and it is impossible to discriminate their separate effects (and the efficacy of the matrix diffusion part) from presently available UZ testing data. Instead the validation strategy presented in this report draws from (1) separate validation of the AFM description of flow behavior, and (2) limited available evidence of matrix diffusion behavior from field testing in the UZ at Yucca Mountain. Section 7.7 includes two subsections. Section 7.7.1. addresses the issue of confidence building during the development of the AFM with matrix diffusion (based on sensitivity analysis of the AFM for transport, aiming to ensure that the incorporation of UZ transport into the TSPA is conservative for fracture-release cases), while Section 7.7.2 focuses on post-development validation of the model. 7.7.1 Confidence Building during Model Development In the AFM conceptualization (Liu et al. 1998 [105729], pp. 2638–2641) only a portion of fracture networks are active (hydraulically conductive) under unsaturated conditions. The active portion is defined as a function of water saturation S to the power of the active fracture parameter . (0=.=1) (Liu et al. 1998 [105729], pp. 2638–2641). When .=0 or S=1 (corresponding to a saturated condition) all fractures are active, while .=1 indicates the smallest active fracture portion for a given saturation. Using carbon-14 data (and incorporating the uncertainty involved in these data), the active fracture parameter . for tsw32–tsw38 were estimated to be in the 0.2–0.4 range (BSC 2003 [161773], pp. 78–82). There is a discrepancy between these estimates and the 0.57 – 0.6 values obtained through inverse modeling in the calibrated flow model (DTN: LB0208UZDSCPMI.002 [161243]). Note that the calibrated values of . (approximately 0.6) were used in all 3-D site-scale simulations in this Model Report. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 259 November 2003 The impact of uncertainties in . on 3-D site-scale flow and transport through the UZ was investigated in BSC 2003 [163045], Section 6.8. The results of this sensitivity analysis indicated that (a) the percolation flux through the repository layers, and (b) the saturation and water potential distributions were not sensitive to large variations in the value ., i.e., reductions by 50% (Figure 7.7-1). However, the sensitivity studies indicated that transport is more sensitive to uncertainty in the value of . than flow. Additionally, the effect is more pronounced at earlier times because diffusive fluxes become progressively smaller as the concentrations in the matrix and fractures approach equilibrium. The sensitivity of transport to . is demonstrated in Figure 7.7-2, which shows a substantial retardation in the arrival of a conservative tracer (with the diffusion coefficient of 99Tc) at the water table when . is reduced by 50% from its original calibrated value of approximately 0.6. In this case (which involved tracer release directly into the repository fractures), t10 increased by over an order of magnitude (from 40 to 650 years), while t50 doubled (from 3500 years to 7000 years). This study demonstrates that the calibrated . value of approximately 0.6 for the hydrostratigraphic units underneath the repository horizon provided conservative estimates of transport times through the UZ (BSC 2003 [163045] Section 6.8.2) when the tracers were released directly into the repository fractures. Note, however, that this value of . leads to an optimistic breakthrough estimate if the tracers are released into the matrix of the repository gridblocks (BSC 2003 [164889] Section 6.3.3). 7.7.2 Post-Development Validation The post-development validation for the AFM with Matrix Diffusion combines the two parts described in the beginning of Section 7.7, i.e., the AFM flow model and the Matrix Diffusion model. Evidence of matrix diffusion from the Alcove 8/Niche 3 Phase 1 Line-Release (Fault) test is somewhat limited because the fracture saturation conditions along the transport pathway (i.e., at or near the fault) are believed to have been highly saturated (inferred from the 2 cm positive pressure head applied at the infiltration plot, see Section 7.4.1). Further testing in Yucca Mountain tuffs, under unsaturated conditions with multiple tracers, could significantly strengthen the validation arguments presented in this report. Two such tests are planned: the Alcove 8/Niche 3 Phase 2 large-plot test, and the cubic-meter-block laboratory test. Relevant unsaturated transport data are currently unavailable from either of these tests, a possibility that was foreseen in the TWP. The TWP covering this Model Report selected independent technical review as a second method for post-development model validation. However, the independent technical review was not utilized in this case. The independent technical review was not considered necessary for adequate validation given the independent validations conducted for the AFM (BSC 2003 [161773], Section 7.2) and matrix diffusion models (see Section 7.4). Future confirmation testing of the UZ transport model is addressed in the Performance Confirmation Plan (Snell et al. 2003 [166219], Section 4.1.1.2). Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 260 November 2003 The validation approach for the AFM with Matrix Diffusion is based on the validation of the AFM flow model and the Matrix Diffusion model separately. It combines the AFM, which is validated in a separate Model Report (BSC 2001 [158726], with the Matrix Diffusion model, validated in Section 7.4 of this report. The validity of the AFM with Matrix Diffusion for Transport Model, for its intended use, is inferred from that of the two separate component models. Additional validation is not necessary because both the active fracture model validation and the matrix diffusion model validation have both active fracture (unsaturated flow through fractures) and matrix diffusion (diffusive exchange between fracture and matrix) processes represented in the validation data. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 261 November 2003 (a) Saturation Elevation (m) 0 0.25 0.5 0.75 1 700 800 900 1000 1100 1200 1300 1400 Calibrated AFP Smaller AFP for under repository Model unit interface TCw PTn TSw CHn (b) Saturation Elevation (m 0 0.25 0.5 0.75 1 700 800 900 1000 1100 Smaller AFP for under repository Model unit interface TSw CHn Water potential (bar) Elevation (m) 0 2 4 6 8 10 700 800 900 1000 1100 1200 1300 1400 Calibrated AFP Smaller AFP for under repository Model unit interface TSw CHn PTn TCw Model Results–DTN LB0304RDTRNSNS.001 [165992] Figure 7.7-1. Comparison of (a) Simulated Matrix Liquid Saturation and (b) Water Potentials Using Calibrated Hydraulic Properties (Solid Line) with That Obtained Using a Smaller (Half) Value of . in the Units of Under Repository (Dashed Line) for Borehole USW SD-6 – From BSC 2003 [163045], Section 6.8, Figure 6.8-2. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 262 November 2003 1.00E+0 1.00E+1 1.00E+2 1.00E+3 1.00E+4 1.00E+5 1.00E+6 Time (year) 0.00 0.20 0.40 0.60 0.80 1.00 Fractional Mass Breakthrough Calibrated gamma (base case) Changed gamma of TSw units Changed gamma of units below the repository Model Results–DTN: LB0304RDTRNSNS.001 [165992] Figure 7.7-2. Comparison of a Simulated Breakthrough Curve of Relative Tracer Mass at the Groundwater Table Obtained for the Base Case (Using Calibrated Rock Hydraulic Properties, Red Solid Line), a Case Using a Smaller (Half) Value of . of the TSw Units (Blue Dash Line), and Another Case Using a Smaller (Half) Value of . of All Units below the Repository (Green Solid Line) – From BSC 2003 [163045], Section 6.8, Figure 6.8-3. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 263 November 2003 8. CONCLUSIONS Section 8 is composed of three subsections. In Section 8.1 we present a summary of the modeling activity in the present Model Report. The main conclusions from the individual studies discussed in Section 6 are presented in Section 8.2, which includes conclusions from the 2-D studies, the 3-D site-scale studies for instantaneous and continuous release of solutes and colloids, and corresponding sensitivity analyses. Section 8.3 focuses on the model output, uncertainties, and limitations. 8.1 SUMMARY OF MODELING ACTIVITY Modeling in Sections 6.8-6.14 involved the application of T2R3D V1.4 (LBNL 1999 [146654]; Wu et al. 1996 [100649]) to study the transport of important radionuclides (99Tc, 237Np, 239Pu, 233U, 235U, 241Am, 135Cs, 90Sr, 226Ra, 229Th, and 221Pa) in the entire Yucca Mountain system, using large-scale 3-D grids and assuming an instantaneous (single event) release scenario. Modeling in Section 6.15 involved the application of TOUGH2 V1.4 Module EOS9 V1.4 (LBNL 2000 [146496]) study the transport of 99Tc, 237Np, and 239Pu in the entire Yucca Mountain system, using large-scale 3-D grids and assuming a continuous release scenario. In Section 6.16 and 6.17, TOUGH2 V1.4 Module EOS9 V1.4 (LBNL 2000 [146496]) was used to study the 3-D site scale transport of the 239Pu 235U 231Pa chain and of the 241Am 237Np 233U 229Th chain, respectively, under a continuous release scenario. The modeling activity in Section 6.18 involved the application of EOS9nT to study the transport of four true colloids of different sizes and properties in the entire Yucca Mountain system, using large-scale 3-D grids. Modeling using the T2R3D (LBNL 1999 [146654]) and EOS9nT codes in Section 6.19 aimed at evaluating the Yucca Mountain UZ system performance under (a) an alternative conceptual model of representing the fracture-matrix system, and (b) an alternative modeling approach involving particle tracking. T2R3D and EOS9nT simulations in Section 6.20 evaluated the impact of the standard radionuclide release model (in which radionuclides are released simultaneously throughout the entire repository) by limiting releases to locations that are not in the immediate vicinity of permeable faults. 8.2. MODEL REPORT CONCLUSIONS The following conclusions can be drawn from this study: 8.2.1 General Conclusions From the 3-D Studies 1. Three dimensional site-scale simulations using EOS9nT indicate that, under the #1 perchedwater model, radionuclide transport from the bottom of the proposed repository to the water table is both dominated and controlled by the faults. These provide fast pathways for downward migration to the water table, but also limit lateral transport across the fault walls into the formation. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 264 November 2003 2. The 3-D site-scale EOS9nT simulations for both an instantaneous and a continuous release regime indicate that radioactive solutes from the repository move faster and reach the water table earlier and over a larger area in the northern part of the repository. This appears to be consistent with the geologic model of the UZ, which is characterized by the presence of highly fractured zeolitic CHz layers in that area. The fractures in these layers are several orders of magnitude more conductive than the corresponding matrices. Thus, flow is fracture-dominated, and results in the transport pattern observed in this study. 3. The transport patterns follow the infiltration and percolation distributions. A review of the infiltration pattern at the surface, the percolation flux at the repository level and the percolation flux at the groundwater level indicates that they closely reflect the transport patterns discussed in Conclusion (2) above. Thus, the water flow pattern dictates the advective transport pattern. 4. For both an instantaneous and a continuous release regime, the highly conductive Drillhole Wash fault and Pagany Wash fault are the main pathways of transport (and related to the percolation patterns) for the dominance of the northern part of the repository. Radionuclides released directly into these faults or reaching them from adjacent areas move rapidly downward to reach the water table at an earlier time. Additionally, the Sundance fault, the Solitario fault, and the Sever Wash fault are transport-facilitating geological features, but their effect becomes significant at later times. 5. Fractures are the main pathways of radionuclide transport. Diffusion from the fractures into the matrix is one of the main retardation processes in radionuclide transport. When radionuclides sorb onto the matrix into which they diffuse, their migration is further retarded. Sorption also leads to retardation in the fractures by resulting in larger concentration gradients (thus enhanced diffusion) between the liquid phases in the fractures and in the matrix. 6. Caution should be exercised in the analysis and interpretation of the 3-D simulation results because the underlying approach is extremely conservative, involves successive (and multiplicative) levels of worst-case scenarios, and is improbable (if not outright impossible). Thus, in this Model Report (a) all the waste packages are assumed to rupture at the same time throughout the repository, (b) water is allowed to seep into the repository by overcoming the capillary pressure barrier, to enter uninhibited the breached canisters, and to carry radionuclides directly into the fractures, (c) drip shields are not considered, (d) the retardation effects of the invert and of the shadow zone are ignored, (e) the vertical fractures are assumed to be open and continuous throughout the UZ from the repository to the water table, (f) the horizontal fractures are assumed to be interconnected, and are also connected (directly or indirectly) to the vertical fractures, and (g) the radioactive tracers (solutes or colloids) are stable, unaffected by the near field conditions (thermal, geochemical, physical), and are not subject to chemical immobilization (e.g., through irreversible sorption or precipitation) anywhere in the UZ. 7. Note that the importance of faults and perched-water bodies in transport is directly dependent on the underlying geologic and perched-water conceptual models. It is reasonable to expect that changing geologic and perched-water models may lead to substantially different results, given the sensitivity of transport to these geologic features. This realization underlines the importance of representative conceptual models. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 265 November 2003 8.2.2 Conclusions From the 3-D Studies of Instantaneously-Released Radioactive Solutes 8. For present-day climatic conditions (Table 6.7-1) and instantaneously-released radionuclides, the t10 (see definition in Section 6.8) of the nonsorbing 99Tc for low, mean, and high infiltration is 13,900, 83, and 6 years, and the t50 (see definition in Section 6.8) is over 1,000,000, 6,640, and 230 years, respectively (Table 6.20-1). Under the same conditions, the t10 of the moderately sorbing 237Np for low, mean, and high infiltration is 33,800, 410, and 4 years, and the corresponding t50 is over 1,000,000, 25,400, and 1,600 years, respectively. The strongly sorbing 239Pu did not exhibit a sufficiently high breakthrough level at the water table for t10 or t50 to become meaningful in most cases (the only exception was a t10 = 1,530 year for a high present-day infiltration. 9. For present-day climatic conditions (Table 6.7-1) and instantaneously-released radionuclides, the t10 (see definition in Section 6.8) of the moderately sorbing 233U for low, mean, and high infiltration is 65,200, 433, and 12 years, and the t50 (see definition in Section 6.8) is over 1,000,000, 29,100, and 1,120 years, respectively (Table 6.20-1). The t10 and t50 for 235U are very close. These results indicate that the 233U and 235U behavior is similar to that of 237Np. Under the same conditions, the t10 of the variably sorbing 135Cs (very different sorption affinity for different rocks, see Table 6.5-1) for low, mean, and high infiltration is 1,000,000, 52,500, and 2,170 years, and the corresponding t50 is over 1,000,000, over 1,000,000, and 71,200 years, respectively. 10. For present-day climatic conditions (Table 6.7-1) and instantaneously-released radionuclides, the cumulative breakthrough curves of 241Am, 231Pa, 229Th, 226Ra, 90Sr indicate very limited (if any) radionuclide arrivals at the water table (Table 6.20-1) because of very strong sorption (all) and short half lives (241Am and 90Sr). 11. The results discussed in (8)-(10) lead to the conclusion that transport to the water table is strongly dependent on the sorption affinity of the radionuclide to the geohydrologic units it encounters in the UZ. Generally speaking, lower Kd values (quantifying weaker sorption) lead to faster radionuclide transport to the water table. 12. The 3-D site-scale simulations indicate that a change in the future climatic conditions can have a significant effect on transport. Increasing infiltration (as the climatic regime changes from present-day to monsoon and glacial conditions and its level changes from low to mean to high) leads to faster radionuclide transport to the water table. The acceleration in transport increases with decreasing sorption. 13. Matrix diffusion (as quantified by the molecular diffusion coefficient D0) has a significant impact on breakthrough predictions, resulting in faster arrival times for lower D0 values. The relative effect of D0 on breakthrough predictions appears to be mitigated by sorption, i.e., the importance of D0 in transport decreases with an increasing Kd. 8.2.3 Conclusions From the 3-D Studies of Continuously-Released Radioactive Solutes 14. For mean present-day climatic conditions (Table 6.7-1) and continuously-released radionuclides from a decaying source, the t10 (see definition in Section 6.15) of 99Tc and 237Np is Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 266 November 2003 74 and 781 years, and the t50 (see definition in Section 6.15) is 3,901 and 22,940 years, respectively (Table 6.20-2). In the case of the strongly sorbing 239Pu, RF did not reach the 0.1 level in the 1,000,000 years of simulation. 15. In the simulation of 3-D transport of 239Pu, the members of the decay chain may have to be considered (i.e., 239Pu, 235U, and 231Pa) if the 239Pu release rate at the repository is significant (Table 6.20-2). Although the t10 for 239Pu transport is always larger than 1,000,000 years, 235U becomes the main contributor in the radionuclide release at the water table after 10,000 years (at most) because of its weaker sorption onto the matrix and its longer half-life. The contributions of 231Pa are not important. Under mean present-day climatic conditions, the t10 and the t50 of the sum of the members of the 239Pu chain are computed from the 3-D simulations as 6,419 and 33,660 years, respectively. 16. The analysis of 3-D transport of the members of the 241Am chain indicates that only the 237Np daughter is significant in the 1,000,000 time frame of this study. The strong sorption affinity and the short T1/2 (432.2 years) of 241Am precludes a significant role in the radionuclide stream arriving at the water table, while the long half life of 237Np prevents its daughters from dominating the releases at the water table. Note that the t10 and t50 of the sum of the 241Am chain members (1,027 and 23,450, respectively) are roughly the sum of t10 and t50 of the transport of pure 237Np (see Conclusion 17 and Table 6.20-2) and the T1/2 (432.2 years) of 241Am. 8.2.4 Conclusions From the 3-D Studies of Continuously-Released Radioactive Colloids 17. For continuous colloid release under a mean present-day infiltration regime and for a given kinetic clogging (forward filtration) coefficient .+, the transport of radioactive true colloids is not appreciably influenced by the kinetic declogging (reverse filtration) coefficient .- in larger colloids. This is caused by (a) the inability of the colloids to reach the matrix because of straining and a low D0 (a result of their large size) and (b) the corresponding very limited filtration in the matrix. 18. The colloid size has a significant effect on transport. Large colloids reach the water table faster. Because fractures are the main transport conduit at Yucca Mountain, the inability of larger colloids to diffuse into the matrix because of smaller D0 values and size exclusion (straining) result in faster transport to the groundwater. Smaller colloidal particles can diffuse more easily into the matrix, and their transport is thus retarded. 19. When there is no filtration in the fractures (Cases 1 to 3), the results of the simulations of transport of PuO2 colloids of four different sizes (450 nm, 200 nm, 100 nm, and 6 nm) indicate that there is very limited retardation of the larger colloids, all of which have a very similar t10 (Table 6.20-3). Note that t10 increases with a decreasing size, but the change is very small (from 4.35 to 4.53 years when the colloid diameter decreases from 450 nm to 100 nm). Although RF does not reach the 0.5 level (thus t50 cannot be defined), it rises rapidly and reaches 0.4 in about 20 years. Conversely, the retardation of the smallest colloid (6 nm) is very significant, and RF is well below the 0.1 level. 20. When limited filtration in the fractures is considered (Case 4), the results of the simulations of transport of PuO2 colloids of four different sizes (450 nm, 200 nm, 100 nm, and 6 nm) are Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 267 November 2003 similar to those for no filtration in the fractures. However, although t10 is still low, it is significantly larger than that for no fracture filtration, and decreases from 32.4 to 27.6 years when the colloid diameter decreases from 450 nm to 100 nm (Table 6.20-3). In this case RF rises rapidly and reaches 0.4 in about 100-150 years. As in the case for no fracture filtration, the retardation of the smallest colloid (6 nm) is very significant, and RF remains well below 0.1 during the simulation period. 21. Size exclusion (straining) at the interfaces of different hydrogeologic layers leads to colloid concentrations immediately above the interface that can be higher than that in the water released from the proposed repository. This phenomenon is more pronounced in larger colloids. 8.2.5 Conclusions From the Study of Alternative Conceptual Models 22. When the alternative conceptual model of multiple interactive continua (to describe the fracture-matrix system) is employed, transport is delayed due to enhanced fracture-matrix interaction. 8.2.6 Conclusions From the Barrier Evaluation 23. Even under the improbably conservative approach involved in the 3-D site-scale studies (Section 6.7.8 and Conclusion 9), the UZ of Yucca Mountain appears to be an effective barrier to the transport of the strongly sorbing radionuclides (90Sr, 226Ra, 229Th, 241Am, 221Pa and 239Pu) to the water table. 24. The variably sorbing 135Cs (strongly on zeolites, much less on other rocks), the mildly sorbing 233U, 235U, 237Np, and the nonsorbing 99Tc arrive at the water table at times that are fractions of their respective half-lives. However, this is not necessarily an indication of a breached or ineffective UZ barrier, but can be a direct consequence of the conceptual model of UZ flow and of the improbably conservative (if not practically impossible) approach described in Section 6.7.8. A more definitive evaluation of the barrier can be provided by using a more realistic scenario accounting for the successive retardation effects of the processes discussed in Section 6.7.8. 25. Eliminating potential sources from the vicisnity of the fault fractures appears to have only a small effect on transport and arrivals at the water table. For instantaneous release (when a finite radionuclide mass is involved), the RM breakthrough curves in Figures 6.20-1 and 6.20-2 show a small increase in t10 (in line with expectations because there is no release into the fast-conducting fault fractures), but t50 is practically unchanged. The RF breakthrough curves in Figures 6.20-3 and 6.20-4 for continuous release show practically no discernible difference from those that account for release into the faults (Section 6.15). 8.3. MODEL OUTPUT, UNCERTAINTIES, AND LIMITATIONS 8.3.1 Model Output As indicated in TWP (BSC 2002 [160819], p. 54), the output/product of this Model Report is a set of breakthrough curves for use as benchmarks for the particle-tracking calculations Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 268 November 2003 performed for TSPA (BSC 2003 [163933]). All the breakthrough curves presented and discussed in this Model Report are included in DTN: LB0307MR0060R1.007, which incorporates both the graphical representations and the model output used for their development. All output DTNs from this Model Report are summarized in Table 8.1 8.3.2 Model Uncertainties As indicated in Section 6, the predictions of transport are subject to three general types of uncertainties. 8.3.2.1 Uncertainties in the Flow Conceptual Model and the Corresponding Parameters The impact of uncertainties in the flow conceptual model and the corresponding hydraulic parameters is in the estimation of the magnitude and relative sizes of flows and saturations in the fractures and in the matrices of the UZ, which affect directly both advective transport (and, consequently, travel times to the water table following a radionuclide release) and diffusive fluxes (see Equations 18–21). Although the importance of these uncertainties is readily recognized, addressing them is beyond the scope of this study, but rather the subject of the Model Report on the UZ Flow Model (BSC 2003 [163045], Section 6). Because the radionuclide transport study is intertwined with the UZ flow uncertainties (from which it derives its flow fields), it suffers from the same flow-related limitations and uncertainties. 8.3.2.2 Climatic Uncertainties The impact of these uncertainties is assessed by estimating transport under all possible nine climatic scenarios of Table 6.7-1, thus bounding the possible solution. From these simulations, the universal conclusion is that the travel time to the water table decreases, and overall transport increases, with an increasing infiltration. 8.3.2.3 Uncertainties in Matrix Diffusion Uncertainties in matrix diffusion are reflected in the values of the diffusion coefficient D0. To address this issue, we investigated the sensitivity of transport through the UZ for the D0 values shown in Table 6.5-2, which cover the possible D0 range for all three possible types of radionuclides (strongly sorbing, mildly sorbing, and nonsorbing). The results of these studies indicated that the D0 value has a significant impact on breakthrough predictions, resulting in faster arrival times for lower D0 values. The impact increases as the sorption affinity of the radionuclides decreases. 8.3.2.4 Uncertainties in Sorption Such uncertainties in sorption are reflected in the values of the distribution coefficient Kd and are important only in the case of mildly sorbing radionuclides. (We did not investigate the effects for strongly sorbing radionuclides, because the studies presented in the Model Report covered the whole range of possibilities and showed minimal effects). To address this issue, we investigated the sensitivity of 237Np transport through the UZ for the Kd values shown on Figure 6.9-5. These cover the range between zero (no sorption, in which case behavior similar to that of 99Tc was expected) to the maximum values reported in Table 6.5-2. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 269 November 2003 As expected, Kd has a significant impact on breakthrough predictions, resulting in faster arrival times for lower Kd values. Note that the relative effect of Kd uncertainty in the prediction of transport of the sorbing 237Np appears to be much more pronounced than that for the D0 uncertainty for the range tested here. 8.3.2.5 Uncertainties in Filtration The impact of these uncertainties is assessed by estimating (a) transport for a very wide range of the kinetic filtration parameters (clogging and declogging coefficients), and (b) the impact of filtration in the fractures (by considering minor fracture fillup), thus bounding the possible solutions. The simulations indicate that the transport is not significantly affected by significantly varying the filtration parameters, travel time to the water table decreases with the colloid size, larger colloids show little retardation while very small ones are retarded significantly, and fracture filtration can have a substantial impact on transport. 8.3.3 Limitations of Applicability The results reported in this Model Report should be viewed as representative of the upper bound of possible solutions when the approach discussed in Section 6.1.5 is employed. When viewed under this light, these results are applicable without limitations. Generally speaking, the results and observations in the Model Report are not applicable to conditions that are not adequately described by the Model Report approach and underlying assumptions. This scenario could include a different conceptual model of UZ flow and/or considerations of the mitigating processes discussed in Sections 6.7.8 and 6.20.3 which are likely to lead to greater delay of radionuclide transport. Therefore, the model is appropriate for comparison with other transport models supporting TSPA to ensure that radionuclide transport is not underestimated. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 270 November 2003 Table 8-1. Output Data and Data Tracking Numbers Location in this report DTN Text Figure Table Remarks LA0302AM831341.002 6.1.1 6.4.2 6.5-1 6.5-2 6.5-3 Unsaturated Zone Distribution Coefficients (KD) For U, Np, Pu, Am, Pa, Cs, Sr, Ra, and Th. LA0311AM831341.001 Attachment II II-2 Correlation Matrix for Sampling of Sorption Coefficient Probability Distributions. Submittal date: 11/06/2003. LB0307MR0060R1.001 6.9.1 6.9.2 6.10.1 6.11 6.12 6.13 6.14 6.7-2 Ambient Radionuclide Transport - Uniform Instantaneous Release: Simulations (Input/output files). LB0307MR0060R1.002 6.8.1.3 6.9.1 6.9.1.3 6.9.2 6.10.1 6.10.1.2 6.11 6.8-2 through 6.8-21 6.7-2 Ambient Radionuclide Transport – Uniform Instantaneous Release: Summaries LB0307MR0060R1.003 6.7.2 6.15.1 6.16 6.17 6.18.1 6.7-2 Ambient Radionuclide Transport – Uniform Continuous Release: Simulations (Input/output files) LB0307MR0060R1.004 6.15.4 6.7-2 Ambient Radionuclide Transport – Uniform Continuous Release: Summaries. LB0307MR0060R1.005 6.20.2 6.7-2 Ambient Radionuclide Transport – No radionuclide releases in the vicinity of major faults, Instantaneous and Continuous Release: Simulations (Input/output files). LB0307MR0060R1.006 6.20.2 6.7-2 Ambient Radionuclide Transport - No radionuclide releases in the vicinity of major faults, Instantaneous and Continuous Release: Summaries. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 271 November 2003 Table 8-1. Output Data and Data Tracking Numbers (Continued) Location in this report DTN Text Figure Table Remarks LB0307MR0060R1.007 8.3.1 6.8-1 6.8-26 6.8-27 6.8-28 6.9-1 6.9-2 6.9-3 6.9-4 6.9-5 6.10-1 6.10-2 6.10-4 6.11-1 6.11-2 6.12-1 6.12-2 6.13-1 6.14-1 6.14-2 6.14-3 6.15-1 6.16-1 6.16-2 6.17-1 6.17-2 6.18-1 6.18-2 6.20-1 6.20-2 6.20-3 6.20-4 6.7-2 6.20-1 6.20-2 6.20-3 Ambient Radionuclide Transport – TSPA Deliverable Extractions. LB0308MR0060R1.008 7.2.1.2 7.2.1.3 7.2.1.4 7.2.1.5 7.3.3.2 7.3.4.2 7.3.5.2 7.2-1 7.2-2 7.2-3 7.2-4 7.2-5 7.3-3 7.3-5 Ambient Radionuclide Transport – Model Validation: Simulations (Input/output files). LB0308MR0060R1.009 7.2-2 7.2-3 7.2-4 7.2-5 7.2-6 7.3-2 7.3-3 7.3-4 7.3-5 7.3-6 7.3-7 7.2-6 7.3-8 Ambient Radionuclide Transport – Model Validation: Data Summaries. LB0310MR0060R1.010 6.19.2.1 6.19.2.2 6.9-2 6.10-3 6.18-3 6.7-2 Supplemental Radionuclide Transport Simulations: Simulations (Input/output files). LB0310MR0060R1.011 6.19.2.1 6.19.2.2 6.7-2 Supplemental Radionuclide Transport Simulations: Data Summaries. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 272 November 2003 INTENTIONALLY LEFT BLANK Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 273 November 2003 9. INPUTS AND REFERENCES The following is a list of the references cited in this document. Column 1 represents the unique six digit numerical identifier (the Document Input Reference System [DIRS] number), which is placed in the text following the reference callout (e.g., BSC 2002 [160819]). The purpose of these numbers is to assist the reader in locating a specific reference. Within the reference list, multiple sources by the same author (e.g., BSC 2002) are sorted alphabetically by title. 9.1 CITED DOCUMENTS 146647 Abdel-Salam, A. and Chrysikopoulos, C.V. 1995. “Modeling of Colloid and Colloid- Facilitated Contaminant Transport in a Two-Dimensional Fracture with Spatially Variable Aperture.” Transport in Porous Media, 20, (3), 197-221. Dordrecht, The Netherlands: Kluwer Academic Publishers. TIC: 247547. 104512 Allard, B. 1982. Sorption of Actinides in Granitic Rock. SKB TR-82-21. Stockholm, Sweden: Svensk Kärnbränsleförsörjning A.B. TIC: 205892. 104410 Allard, B.; Beall, G.W.; and Krajewski, T. 1980. “The Sorption of Actinides in Igneous Rocks.” Nuclear Technology, 49, (3), 474-480. La Grange Park, Illinois: American Nuclear Society. TIC: 245772. 162982 Allard, B.; Olofsson, U.; Torstenfelt, B.; and Kipatsi, H. 1983. Sorption Behaviour of Well-Defined Oxidation States. SKB TR-83-61. Stockholm, Sweden: Svensk Kärnbränsleförsörjning A.B. TIC: 206122. 100704 Bates, J.K.; Bradley, J.P.; Teetsov, A.; Bradley, C.R.; and Buchholtz ten Brink, M. 1992. “Colloid Formation During Waste Form Reaction: Implications for Nuclear Waste Disposal.” Science, 256, 649-651. Washington, D.C.: American Association for the Advancement of Science. TIC: 239138. 162983 Beall, G.W.; Lee, W.W.-L.; and Van Luik, A.E. 1986. “Americium Speciation and Distribution Coefficients in a Granitic Ground Water.” Scientific Basis for Nuclear Waste Management IX, Symposium held September 9-11, 1985, Stockholm, Sweden. Werme, L.O., ed. 50, 501-508. Pittsburgh, Pennsylvania: Materials Research Society. TIC: 203664. 156269 Bear, J. 1972. Dynamics of Fluids in Porous Media. Environmental Science Series. Biswas, A.K., ed. New York, New York: Elsevier. TIC: 217356. 105038 Bear, J. 1979. Hydraulics of Groundwater. New York, New York: McGraw-Hill. TIC: 217574. 122788 Benson, C.F. and Bowman, R.S. 1994. “Tri- and Tetrafluorobenzoates as Nonreactive Tracers in Soil and Groundwater.” Soil Science Society of America Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 274 November 2003 Journal, 58, (4), 1123-1129. Madison, Wisconsin: Soil Science Society of America. TIC: 246741. 144728 Berry, J.A.; Hobley, J.; Lane, S.A.; Littleboy, A.K.; Nash, M.J.; Oliver, P.; Smith- Briggs, J.L.; and Williams, S.J. 1989. “Solubility and Sorption of Protactinium in the Near-Field and Far-Field Environments of a Radioactive Waste Repository.” Analyst, 114, 339-347. Cambridge, England: Royal Society of Chemistry. TIC: 247004. 103524 Bird, R.B.; Stewart, W.E.; and Lightfoot, E.N. 1960. Transport Phenomena. New York, New York: John Wiley & Sons. TIC: 208957. 105170 Birkholzer, J.; Li, G.; Tsang, C-F.; and Tsang, Y. 1999. “Modeling Studies and Analysis of Seepage into Drifts at Yucca Mountain.” Journal of Contaminant Hydrology, 38, (1-3), 349-384. New York, New York: Elsevier. TIC: 244160. 120055 Bodvarsson, G.S.; Boyle, W.; Patterson, R.; and Williams, D. 1999. “Overview of Scientific Investigations at Yucca Mountain—The Potential Repository for High- Level Nuclear Waste.” Journal of Contaminant Hydrology, 38, (1-3), 3-24. New York, New York: Elsevier. TIC: 244160. 160133 Bodvarsson, G.S.; Liu, H.H.; Ahlers, C.F.; Wu, Y-S.; and Sonnenthal, S. 2001. “Parameterization and Upscaling in Modeling Flow and Transport in the Unsaturated Zone of Yucca Mountain.” Chapter 11 of Conceptual Models of Flow and Transport in the Fractured Vadose Zone. Washington, D.C.: National Academy Press. TIC: 252777. 122790 Boggs, J.M. and Adams, E.E. 1992. “Field Study of Dispersion in a Heterogeneous Aquifer, 4. 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TIC: 208542. 101355 Rundberg, R.S.; Ogard, A.E.; and Vaniman, D.T., eds. 1985. Research and Development Related to the Nevada Nuclear Waste Storage Investigations, April 1– June 30, 1984. LA-10297-PR. Los Alamos, New Mexico: Los Alamos National Laboratory. ACC: NNA.19920922.0018. 134556 Sakthivadivel, R. 1969. Clogging of a Granular Porous Medium by Sediment, Final Report. HEL 15-7. Berkeley, California: University of California, College of Engineering, Hydraulic Engineering Laboratory. TIC: 250299. 117338 Saltelli, A.; Avogadro, A.; and Bidoglio, G. 1984. “Americium Filtration in Glauconitic Sand Columns.” Nuclear Technology, 67, (2), 245-254. La Grange Park, Illinois: American Nuclear Society. TIC: 223230. 104181 Scott, R.B. and Bonk, J. 1984. Preliminary Geologic Map of Yucca Mountain, Nye County, Nevada, with Geologic Sections. Open-File Report 84-494. Denver, Colorado: U.S. Geological Survey. ACC: HQS.19880517.1443. 134563 Seaman, J.C. 1998. “Retardation of Fluorobenzoate Tracers in Highly Weathered Soil and Groundwater Systems.” Soil Science Society of America Journal, 62, (2), 354-361. Madison, Wisconsin: Soil Science Society of America. TIC: 246908. 162132 Shapiro, A.M. 2001. “Effective Matrix Diffusion in Kilometer-Scale Transport in Fractured Crystalline Rock.” Water Resources Research, 37, (3), 507-522. [Washington, D.C.]: American Geophysical Union. TIC: 253979. 144658 Smith, P.A. and Degueldre, C. 1993. “Colloid-Facilitated Transport of Radionuclides Through Fractured Media.” Journal of Contaminant Hydrology, 13, 143-166. Amsterdam, The Netherlands: Elsevier. TIC: 224863. 166219 Snell, R.D.; Beesley, J.F.; Blink, J.A.; Duguid, J.O.; Jenni, K.E.; Lin, W.; Monib, A.M.; Nieman, T.L.; and Hommel, S. 2003. Performance Confirmation Plan. TDRPCS- SE-000001 REV 02 ICN 02. Las Vegas, Nevada: Bechtel SAIC Company. ACC: DOC.20031124.0003. 105043 Sudicky, E.A. and Frind, E.O. 1982. “Contaminant Transport in Fractured Porous Media: Analytical Solutions for a System of Parallel Fractures.” Water Resources Research, 18, (6), 1634-1642. Washington, D.C.: American Geophysical Union. TIC: 217475. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 286 November 2003 134571 Tachi, Y.; Shibutani, T.; Sato, H.; and Yui, M. 1998. “Sorption and Diffusion Behavior of Selenium in Tuff.” Journal of Contaminant Hydrology, 35, (1-3), 77-89. Amsterdam, The Netherlands: Elsevier. TIC: 246891. 139195 Tokunaga, T.K. and Wan, J. 1997. “Water Film Flow Along Fracture Surfaces of Porous Rock.” Water Resources Research, 33, (6), 1287-1295. Washington, D.C.: American Geophysical Union. TIC: 242739. 145123 Triay, I.R.; Birdsell, K.H.; Mitchell, A.J.; and Ott, M.A. 1993. “Diffusion of Sorbing and Non-Sorbing Radionuclides.” High Level Radioactive Waste Management, Proceedings of the Fourth Annual International Conference, Las Vegas, Nevada, April 26-30, 1993. 2, 1527-1532. La Grange Park, Illinois: American Nuclear Society. TIC: 208542. 101023 Triay, I.R.; Cotter, C.R.; Huddleston, M.H.; Leonard, D.E.; Weaver, S.C.; Chipera, S.J.; Bish, D.L.; Meijer, A.; and Canepa, J.A. 1996. Batch Sorption Results for Neptunium Transport Through Yucca Mountain Tuffs. LA-12961-MS. Los Alamos, New Mexico: Los Alamos National Laboratory. ACC: MOL.19980924.0050. 104129 Triay, I.R.; Meijer, A.; Cisneros, M.R.; Miller, G.G.; Mitchell, A.J.; Ott, M.A.; Hobart, D.E.; Palmer, P.D.; Perrin, R.E.; and Aguilar, R.D. 1991. “Sorption of Americium in Tuff and Pure Minerals Using Synthetic and Natural Groundwaters.” Radiochimica Acta, 52/53, 141-145. München, Germany: R. Oldenbourg Verlag. TIC: 222704. 100422 Triay, I.R.; Meijer, A.; Conca, J.L.; Kung, K.S.; Rundberg, R.S.; Strietelmeier, B.A.; and Tait, C.D. 1997. Summary and Synthesis Report on Radionuclide Retardation for the Yucca Mountain Site Characterization Project. Eckhardt, R.C., ed. LA-13262- MS. Los Alamos, New Mexico: Los Alamos National Laboratory. ACC: MOL.19971210.0177. 162989 Turner, D.R.; Pabalan, R.T.; and Bertetti, F.P. 1998. “Neptunium(V) Sorption on Montmorillonite: An Experimental and Surface Complexation Modeling Study.” Clays and Clay Minerals, 46, (3), 256-269. Boulder, Colorado: Clay Minerals Society. TIC: 254532. 158378 USGS (U.S. Geological Survey) 2001. Future Climate Analysis. ANL-NBS-GS- 000008 REV 00 ICN 01. Denver, Colorado: U.S. Geological Survey. ACC: MOL.20011107.0004. 109249 van de Weerd, H. and Leijnse, A. 1997. “Assessment of the Effect of Kinetics on Colloid Facilitated Radionuclide Transport in Porous Media.” Journal of Contaminant Hydrology, 26, 245-256. Amsterdam, The Netherlands: Elsevier. TIC: 245731. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 287 November 2003 100610 van Genuchten, M.T. 1980. “A Closed-Form Equation for Predicting the Hydraulic Conductivity of Unsaturated Soils.” Soil Science Society of America Journal, 44, (5), 892-898. Madison, Wisconsin: Soil Science Society of America. TIC: 217327. 109252 Vilks, P. and Bachinski, D.B. 1996. “Colloid and Suspended Particle Migration Experiments in a Granite Fracture.” Journal of Contaminant Hydrology, 21, 269- 279. Amsterdam, The Netherlands: Elsevier. TIC: 245730. 109254 Vilks, P.; Frost, L.H.; and Bachinski, D.B. 1997. “Field-Scale Colloid Migration Experiments in a Granite Fracture.” Journal of Contaminant Hydrology, 26, 203- 214. Amsterdam, The Netherlands: Elsevier. TIC: 245732. 134579 Viswanathan, H.S.; Robinson, B.A.; Valocchi, A.J.; and Triay, I.R. 1998. “A Reactive Transport Model of Neptunium Migration from the Potential Repository at Yucca Mountain.” Journal of Hydrology, 209, 251-280. Amsterdam, The Netherlands: Elsevier. TIC: 243441. 108285 Wan, J. and Tokunaga, T.K. 1997. “Film Straining on Colloids in Unsaturated Porous Media: Conceptual Model and Experimental Testing.” Environmental Science & Technology, 31, (8), 2413-2420. [Washington, D.C.: American Chemical Society]. TIC: 234804. 114430 Wan, J. and Wilson, J.L. 1994. “Colloid Transport in Unsaturated Porous Media.” Water Resources Research, 30, (4), 857-864. Washington, D.C.: American Geophysical Union. TIC: 222359. 164021 Wang, J.S. 2003. “Scientific Notebooks Referenced in Model Report U0060, Radionuclide Transport Models Under Ambient Conditions MDL-NBS-HS-000008 REV 01.” Interoffice correspondence from J.S. Wang (BSC) to File, November 18, 2003, with attachments. ACC: MOL.20031118.0184. 153972 Wu, Y-S. and Pruess, K. 2000. “Numerical Simulation of Non-Isothermal Multiphase Tracer Transport in Heterogeneous Fractured Porous Media.” Advances in Water Resources, 23, (7), 699-723. New York, New York: Elsevier. TIC: 249626. 100649 Wu, Y.S.; Ahlers, C.F.; Fraser, P.; Simmons, A.; and Pruess, K. 1996. Software Qualification of Selected TOUGH2 Modules. LBL-39490. Berkeley, California: Lawrence Berkeley National Laboratory. ACC: MOL.19970219.0104. 117167 Wu, Y.S.; Ritcey, A.C.; and Bodvarsson, G.S. 1999. “A Modeling Study of Perched Water Phenomena in the Unsaturated Zone at Yucca Mountain.” Journal of Contaminant Hydrology, 38, (1-3), 157-184. New York, New York: Elsevier. TIC: 244160. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 288 November 2003 101169 Yao, K.M.; Habibian, M.T.; and O’Melia, C.R. 1971. “Water and Waste Water Filtration: Concepts and Applications.” Environmental Science & Technology, 5, (11), 1105-1112. Washington, D.C.: American Chemical Society. TIC: 239214. 162133 Zhou, Q.; Liu, H-H.; Bodvarsson, G.S.; and Oldenburg, C.M. 2003. “Flow and Transport in Unsaturated Fractured Rock: Effects of Multiscale Heterogeneity of Hydrogeologic Properties.” Journal of Contaminant Hydrology, 60, ([1-2]), 1-30. New York, New York: Elsevier. TIC: 253978. Software Cited 155323 BSC (Bechtel SAIC Company) 2001. Software Code: PHREEQC. V2.3. PC. 10068-2.3-00. 139918 LBNL (Lawrence Berkeley National Laboratory) 1999. Software Code: iTOUGH2. V4.0. SUN, DEC. 10003-4.0-00. 146654 LBNL (Lawrence Berkeley National Laboratory) 1999. Software Code: T2R3D. V1.4. FORTRAN 77, SUN, DEC / ALPHA. 10006-1.4-00. 113943 LBNL (Lawrence Berkeley National Laboratory) 1999. Software Code: TOUGH2. V1.11 MEOS9nTV1.0. MAC, SUN, DEC/Alpha, PC. 10065-1.11MEOS9NTV1.0- 00. 132448 LBNL (Lawrence Berkeley National Laboratory) 2000. Software Code: DCPT. V1.0. PC. 10078-1.0-00. 146496 LBNL (Lawrence Berkeley National Laboratory) 2000. Software Code: TOUGH2. V1.4. Sun Workstation and DEC/ALPHA. 10007-1.4-01. 162143 LBNL (Lawrence Berkeley National Laboratory) 2002. Software Code: Bkread.f. V1.0. SunOS 5.5.1. 10894-1.0-00. 154342 LBNL (Lawrence Berkeley National Laboratory) 2002. Software Code: DCPT. V2.0. PC, Windows. 10078-2.0-00. 162142 LBNL (Lawrence Berkeley National Laboratory) 2002. Software Code: Smesh.f. V1.0. SunOS 5.5.1. 10896-1.0-00. 160242 LBNL (Lawrence Berkeley National Laboratory) 2002. Software Code: TOUGH2. V1.6. PC/MS-DOS under Windows 98, Sun UltraSparc OS 5.5.1, DEC-Alpha OSF1 V4.0. 10007-1.6-00. 161491 LBNL (Lawrence Berkeley National Laboratory) 2003. Software Code: TOUGH2. V1.6. PC/MS-DOS under Windows 98, Sun UltraSparc OS 5.5.1, DEC-Alpha OSF1 V4.0. 10007-1.6-01. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 289 November 2003 162786 LBNL (Lawrence Berkeley National Laboratory) 2003. Software Code: XtractG.f90. V1.0. MAC, OS 9.2; DEC Alpha, OSF1 V4.0 and OSF1 V5.1. 10930-1.0-00. 9.2 CODES, STANDARDS, REGULATIONS, AND PROCEDURES 156605 10 CFR 63. Energy: Disposal of High-Level Radioactive Wastes in a Geologic Repository at Yucca Mountain, Nevada. Readily available. AP-2.22Q, Rev. 1, ICN 0. Classification Criteria and Maintenance of the Monitored Geologic Repository Q-List. Washington, D.C.: U.S. Department of Energy, Office of Civilian Radioactive Waste Management. ACC: DOC.20030807.0002. AP-SI.1Q, Rev. 05, ICN 2. Software Management. Washington, D.C.: U.S. Department of Energy, Office of Civilian Radioactive Waste Management. ACC: DOC.20030902.0003. 9.3 SOURCE DATA, LISTED BY DATA TRACKING NUMBER 147613 GS000308311221.005. Net Infiltration Modeling Results for 3 Climate Scenarios for FY99. Submittal date: 03/01/2000. 153407 GS000608312271.001. Pore-Water Hydrochemistry and Isotopic Data for Boreholes USW NRG-6, USW NRG-7A, USW SD-7, USW SD-9, USW SD-12, USW UZ-14 and UE-25 UZ#16 from 10/1/96 to 1/31/97. Submittal date: 06/23/2000. 165860 GS010608312272.001. Chemical Analysis of Pore Water from Boreholes USW UZ- 7A, USW WT-24, USW SD-6, USW SD-7 and USW SD-12 During FY 1997 and FY 1998. Submittal date: 07/09/2001. 156375 GS010708312272.002. Chemical Data for Pore Water from Tuff Cores of USW NRG-6, USW NRG-7/7A, USW UZ-14, USW UZ-N55 and UE-25 UZ#16. Submittal date: 09/05/2001. 165859 GS011008312272.004. Analysis for Chemical Composition of Pore Water from Boreholes USW WT-24 and USW SD-6 During FY98 and FY99. Submittal date: 12/20/2001. 160899 GS020408312272.003. Collection and Analysis of Pore Water Samples for the Period from April 2001 to February 2002. Submittal date: 04/24/2002. 162129 GS020508312242.001. Trench Fault Infiltration in Alcove 8 Using Permeameters from March 5, 2001 to June 1, 2001. Submittal date: 05/22/2002. 162141 GS020908312242.002. Trenched Fault Infiltration in Alcove 8 Using Permeameters from June 1, 2001 to March 26, 2002. Submittal date: 09/17/2002. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 290 November 2003 162131 GS030108314224.001. Geotechnical Data for Alcove 8 (ECRB) and Niche 3 (ESF): Full Periphery Geologic Map (Drawing OA-46-356). Submittal date: 02/05/2003. 144662 GS950608312231.008. Moisture Retention Data from Boreholes USW UZ-N27 and UE-25 UZ#16. Submittal date: 06/06/1995. 165858 GS951208312272.004. Analysis for Chemical Composition of Perched-Water from Boreholes USW UZ-14, USW NRG-7A, USW SD-9, USW SD-7 and Groundwater from Boreholes UE-25 ONC#1 and USW G-2 from 8/18/89 to 3/21/95. Submittal date: 09/12/2001. 145272 GS980908312242.039. Unsaturated Water Retention Data for Lexan-Sealed Samples from USW SD-6 Measured Using a Centrifuge. Submittal date: 09/22/1998. 146134 GS990208312272.001. Analysis for Chemical Composition of Pore Water from Borehole USW UZ-14 and UE-25 UZ#16 and Groundwater from UE-25 UZ#16. Submittal date: 02/23/1999. 107185 GS990308312242.007. Laboratory and Centrifuge Measurements of Physical and Hydraulic Properties of Core Samples from Busted Butte Boreholes UZTT-BB-INJ- 1, UZTT-BB-INJ-3, UZTT-BB-INJ-4, UZTT-BB-INJ-6, UZTT-BB-COL-5 and UZTT-BB-COL-8. Submittal date: 03/22/1999. 109822 GS990708312242.008. Physical and Hydraulic Properties of Core Samples from Busted Butte Boreholes. Submittal date: 07/01/1999. 148603 LA000000000034.002. Diffusion of Sorbing and Non-Sorbing Radionuclides. Submittal date: 06/22/1993. 156582 LA0008WS831372.001. Calculated Daily Injection Rates for the Busted Butte Unsaturated Zone Transport Tests. Submittal date: 08/23/2000. 157100 LA0112WS831372.001. Busted Butte UZ Transport Test: Phase II Collection Pad Tracer Loading. Submittal date: 12/06/2001. 157115 LA0112WS831372.002. Busted Butte UZ Transport Test: Phase II Collection Pad Tracer Concentrations. Submittal date: 12/06/2001. 157106 LA0112WS831372.003. Busted Butte UZ Transport Test: Phase II Normalized Collection Pad Tracer Concentrations. Submittal date: 12/06/2001. 164721 LA0201WS831372.007. Busted Butte UZ Transport Test: Phase II Normalized Collection Pad Metal Tracer Concentrations. Submittal date: 01/22/2002. 162766 LA0201WS831372.008. Busted Butte UZ Transport Test: Phase I Collection Pad Normalized Tracer Concentrations. Submittal date: 01/22/2002. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 291 November 2003 161526 LA0203WS831372.002. Freundlich and Linear Isotherm Parameters for Radionuclides and Sorbing Tracers Used in the Busted Butte UZ Transport Test. Submittal date: 03/20/2002. 160051 LA0206AM831234.001. Eh-pH Field Measurements on Nye County EWDP Wells. Submittal date: 06/21/2002. 163852 LA0206AM831234.002. Geochemical Field Measurements on Nye County EWDP Wells. Submittal date: 06/21/2002. 162763 LA0211WS831372.001. Busted Butte Rock Extractions: Measured and Normalized Anion and FBA Tracer Concentrations. Submittal date: 11/19/2002. 162765 LA0302WS831372.001. Fluorescein Plume Images from the Phase 1A Mineback at Busted Butte. Submittal date: 02/26/2003. 163789 LA0305AM831341.001. 1977 to 1987 Sorption Measurements of AM, BA, CS, NP, PU, PA, SR, TH, and U with Yucca Mountain Tuff Samples. Submittal date: 05/21/2003. 164949 LA0306AM831343.001. Modeling Calculations of Radionuclide Sorption via Surface-Complexation Reactions. Submittal date: 06/09/2003. 165523 LA0309AM831341.002. Batch Sorption Coefficient Data for Barium on Yucca Mountain Tuffs in Representative Water Compositions. Submittal date: 09/25/2003. 165524 LA0309AM831341.003. Batch Sorption Coefficient Data for Cesium on Yucca Mountain Tuffs in Representative Water Compositions. Submittal date: 09/25/2003. 165525 LA0309AM831341.004. Batch Sorption Coefficient Data for Neptunium on Yucca Mountain Tuffs in Representative Water Compositions. Submittal date: 09/25/2003. 165526 LA0309AM831341.005. Batch Sorption Coefficient Data for Plutonium on Yucca Mountain Tuffs in Representative Water Compositions. Submittal date: 09/25/2003. 165527 LA0309AM831341.006. Batch Sorption Coefficient Data for Strontium on Yucca Mountain Tuffs in Representative Water Compositions. Submittal date: 09/25/2003. 165528 LA0309AM831341.007. Batch Sorption Coefficient Data for Uranium on Yucca Mountain Tuffs in Representative Water Compositions. Submittal date: 09/25/2003. 165865 LA0310AM831341.001. Sorption/Desorption Measurements of Cesium on Yucca Mountain Tuff. Submittal date: 10/21/2003. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 292 November 2003 147156 LA9910WS831372.008. Busted Butte UZ Transport Test: Gravimetric Moisture Content and Bromide Concentration in Selected Phase 1A Rock Samples. Submittal date: 11/03/1999. 156586 LA9912WS831372.001. Sorption of Fluorinated Benzoic Acids and Lithium on Rock Samples Form Busted Butte, NV. Submittal date: 02/22/2000. 159525 LB0205REVUZPRP.001. Fracture Properties for UZ Model Layers Developed from Field Data. Submittal date: 05/14/2002. 159672 LB0207REVUZPRP.002. Matrix Properties for UZ Model Layers Developed from Field and Laboratory Data. Submittal date: 07/15/2002. 161243 LB0208UZDSCPMI.002. Drift-Scale Calibrated Property Sets: Mean Infiltration Data Summary. Submittal date: 08/26/2002. 160799 LB0210THRMLPRP.001. Thermal Properties of UZ Model Layers: Data Summary. Submittal date: 10/25/2002. 162379 LB03013DSSCP3I.001. 3-D Site Scale Calibrated Properties: Data Summaries. Submittal date: 01/27/2003. 162130 LB0301N3SURDAT.001. Niche 3107 Measurements and Elevations Used for Grid Generation. Submittal date: 01/29/2003. 163044 LB03023DSSCP9I.001. 3-D Site Scale UZ Flow Field Simulations for 9 Infiltration Scenarios. Submittal date: 02/28/2003. 162570 LB0303A8N3LIQR.001. Alcove 8 - Niche 3 Seepage Data Compilation. Submittal date: 03/19/2003. 162773 LB0303A8N3MDLG.001. Alcove 8 - Niche 3 Seepage Modeling: Simulations. Submittal date: 03/31/2003. 165992 LB0304RDTRNSNS.001. Supporting Files of 3D Flow and Transport Sensitivity Analyses. Submittal date: 04/29/2003. 165172 LB0308AMRU0185.001. Section 6.3.2 Matrix Block Discretization and its Effects on UZ Flow and Transport Simulations. Submittal date: 08/29/2003. 166225 LB03093RADTRNS.001. Three Way Transport Model Comparison: Input/Output Files. Submittal date: 09/24/2003. 166071 LB03093RADTRNS.002. Three Way Transport Model Comparison: Data Summaries. Submittal date: 09/24/2003. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 293 November 2003 136593 LB980901233124.101. Pneumatic Pressure and Air Permeability Data from Niche 3107 and Niche 4788 in the ESF from Chapter 2 of Report SP33PBM4: Fracture Flow and Seepage Testing in the ESF, FY98. Submittal date: 11/23/1999. 106787 LB990501233129.001. Fracture Properties for the UZ Model Grids and Uncalibrated Fracture and Matrix Properties for the UZ Model Layers for AMR U0090, “Analysis of Hydrologic Properties Data.” Submittal date: 08/25/1999. 106785 LB990701233129.001. 3-D UZ Model Grids for Calculation of Flow Fields for PA for AMR U0000, “Development of Numerical Grids for UZ Flow and Transport Modeling.” Submittal date: 09/24/1999. 110226 LB990861233129.001. Drift Scale Calibrated 1-D Property Set, FY99. Submittal date: 08/06/1999. 164858 LB991220140160.010. Model Prediction of Busted Butte Using T2R3D Input and Output Files. AMR U0060, “Radionuclide Transport Models Under Ambient Conditions.”. Submittal date: 03/14/2000. 104055 LB997141233129.001. Calibrated Basecase Infiltration 1-D Parameter Set for the UZ Flow and Transport Model, FY99. Submittal date: 07/21/1999. 151523 MO0007MAJIONPH.010. Major Ion Content of Groundwater from Borehole UE-25 P #1 Extracted from ANL-NBS-HS-000021, Geochemical and Isotopic Constraints on Groundwater Flow Directions, Mixing and Recharge at Yucca Mountain, Nevada. Submittal date: 07/27/2000. 151530 MO0007MAJIONPH.013. Major Ion Content of Groundwater from Selected YMP and Other Boreholes Extracted from ANL-NBS-HS-000021, Geochemical and Isotopic Constraints on Groundwater Flow Directions, Mixing and Recharge at Yucca Mountain, Nevada. Submittal date: 07/27/2000. 164527 MO0307SEPFEPS4.000. LA FEP List. Submittal date: 07/31/2003. 119199 MO9910MWDISMMM.003. ISM3.1 Mineralogic Models. Submittal date: 10/01/1999. 9.4 OUTPUT DATA, LISTED BY DATA TRACKING NUMBER LA0302AM831341.002. Unsaturated Zone Distribution Coefficients (KDS) For U, NP, PU, AM, PA, CS, SR, RA, AND TH. Submittal date: 02/04/2003. LA0311AM831341.001. Correlation Matrix for Sampling of Sorption Coefficient Probability Distributions. Submittal date: 11/06/2003. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 294 November 2003 LB0307MR0060R1.001. Ambient Radionuclide Transport – Uniform Instantaneous Release. Submittal date: 07/19/2003. LB0307MR0060R1.002. Ambient Radionuclide Transport – Uniform Instantaneous Release. Submittal date: 07/19/2003. LB0307MR0060R1.003. Ambient Radionuclide Transport – Uniform Continuous Release. Submittal date: 07/19/2003. LB0307MR0060R1.004. Ambient Radionuclide Transport – Uniform Continuous Release. Submittal date: 07/19/2003. LB0307MR0060R1.005. Ambient Radionuclide Transport - Unfaulted Domain, Instantaneous and Continuous Release. Submittal date: 07/19/2003. LB0307MR0060R1.006. Ambient Radionuclide Transport - Unfaulted Domain, Instantaneous and Continuous Release. Submittal date: 07/19/2003. LB0307MR0060R1.007. Ambient Radionuclide Transport – TSPA Deliverable Extractions. Submittal date: 07/19/2003. LB0308MR0060R1.008. Ambient Radionuclide Transport – Model Validation: Simulations. Submittal date: 08/29/2003. LB0308MR0060R1.009. Ambient Radionuclide Transport – Model Validation: Data Summaries. Submittal date: 08/29/2003. LB0310MR0060R1.010. Supplemental Radionuclide Transport Simulations: Input/Output files. Submittal date: 10/23/2003. LB0310MR0060R1.011. Supplemental Radionuclide Transport Simulations: Data Summaries. Submittal date: 10/23/2003. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-1 November 2003 ATTACHMENT I TECHNICAL BASIS FOR Kd PROBABILITY DISTRIBUTION FUNCTIONS Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-2 November 2003 INTENTIONALLY LEFT BLANK Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-3 November 2003 FIGURES Page I-1. Americium Sorption Coefficients on Devitrified Tuff versus Calculated Final Cesium Concentration in Solution ............................................................................................I-18 I-2. Americium Sorption Coefficients for Devitrified Tuff as a Function of Experiment Duration .......................................................................................................................I-19 I-3. Americium Sorption Coefficients on Devitrified Tuff Versus pH of J-13 Water...............I-20 I-4a. Americium Sorption Coefficients on Quartz from Beall et al. (1986 [162983]) and Model Fit......................................................................................................................I-21 I-4b. Americium Sorption Coefficients on Zeolitic Tuff versus Calculated Final Americium Concentration in Solution .........................................................................I-22 I-5. Americium Sorption Coefficients for Zeolitic Tuff as Function of Duration of Sorption and Desorption Experiments .........................................................................I-23 I-6. Americium Sorption Coefficient Data and Modeling Results for Zeolitic Tuff as a Function of pH .............................................................................................................I-24 I-7. Americium Sorption Coefficients on Vitric Tuff versus Calculated Final Americium Concentration in Solution ............................................................................................I-25 I-8. Americium Sorption Coefficient Data and Modeling Results for Vitric Tuff as a Function of pH .............................................................................................................I-25 I-9. Cesium Sorption Coefficients on Devitrified Tuff versus Calculated Final Cesium Concentration in Solution ............................................................................................I-27 I-10. Freundlich Isotherm Fit to Sorption Coefficient Data for Cesium on Devitrified Tuff Sample G1-2840 in J-13 Water....................................................................................I-28 I-11. Cesium Sorption Coefficients on Devitrified Tuff versus Experiment Duration for Sorption (Forward) and Desorption (Backward) Experiments....................................I-28 I-12. Cesium Sorption Coefficients on Zeolitic Tuff versus Calculated Final Cesium Concentration in Solution ............................................................................................I-29 I-13. Freundlich Isotherm Fit to Sorption Coefficient Data for Sample YM-38 in J-13 Water...........................................................................................................................I-29 I-14. Cesium Sorption Coefficients on Zeolitic Tuff versus Experiment Duration for Sorption (Forward) and Desorption (Backward) Experiments....................................I-30 I-15. Cesium Sorption Coefficients on Vitric Tuff versus Calculated Final Cesium Concentration in Solution ............................................................................................I-31 I-16. Cesium Sorption Coefficients on Vitric Tuff versus Experiment Duration for Sorption (Forward) and Desorption (Backward) Experiments....................................I-31 I-17. Neptunium Sorption Coefficients on Devitrified Tuff versus Calculated Final Neptunium Concentration in Solution .........................................................................I-33 I-18. Neptunium Sorption Coefficients on Devitrified Tuff versus Calculated Final Neptunium Concentration in Solution. Experiments oversaturated with Np2O5 have been omitted. .......................................................................................................I-34 I-19. Neptunium Sorption Coefficient on Devitrified Tuff versus Experiment Duration for Sorption (Forward) and Desorption (Backward) Experiments. Experiments oversaturated with Np2O5 have been omitted. .............................................................I-34 Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-4 November 2003 FIGURES Page I-20. Neptunium Sorption Coefficient on Devitrified Tuff in J-13 and Synthetic p#1 versus Solution pH in Sorption (Forward) and Desorption (Backward) Experiments. PHREEQC model results for J-13 and p#1 waters also plotted ............I-35 I-21. Neptunium Sorption Coefficients on Zeolitic Tuff versus Calculated Final Neptunium Concentration in Solution .........................................................................I-36 I-22. Neptunium Sorption Coefficients on Zeolitic Tuff versus Calculated Final Neptunium Concentration in Solution. Oversaturated experiments have been removed........................................................................................................................I-36 I-23. Neptunium Sorption Coefficient on Zeolitic Tuff versus Experiment Duration for Sorption (Forward) and Desorption (Backward) Experiments. Oversaturated experiments have been omitted....................................................................................I-37 I-24. Neptunium Sorption Coefficient on Devitrified Tuff in J-13 and Synthetic p#1 versus Solution pH in Sorption (Forward) and Desorption (Backward) Experiments. Oversaturated experiments have been omitted. ....................................I-38 I-25. Neptunium Sorption Coefficients on Vitric Tuff versus Calculated Final Neptunium Concentration in Solution. Oversaturated experiments have been omitted................I-39 I-26. Neptunium Sorption Coefficient on Vitric Tuff versus Experiment Duration for Sorption (Forward) and Desorption (Backward) Experiments. Oversaturated experiments have been omitted....................................................................................I-39 I-27. Neptunium Sorption Coefficient on Devitrified Tuff in J-13 and Synthetic p#1 versus Solution pH in Sorption (Forward) and Desorption (Backward) Experiments. Oversaturated experiments have been omitted. ....................................I-40 I-28. Plutonium Sorption Coefficients on Devitrified Tuff versus Calculated Final Plutonium Concentration in Solution...........................................................................I-41 I-29. Plutonium Sorption Coefficient versus Calculated Final Plutonium Solution Concentration in Moles/Liter for Experiments with Samples YM-22 and G4-272.....I-42 I-30. Plutonium Sorption Coefficient on Devitrified Tuff versus Experiment Duration for Sorption (Forward) and Desorption (Backward) Experiments....................................I-43 I-31. Plutonium Sorption Coefficient on Devitrified Tuff in J-13 and Synthetic p#1 versus Solution pH in Sorption (Forward) and Desorption (Backward) Experiments. Experiments lasting 40 days or more are plotted separately from experiments lasting less than 40 days...............................................................................................I-43 I-32. Plutonium Sorption Coefficient on Devitrified Tuff in J-13 and Synthetic p#1 versus Solution pH in Sorption (Forward) and Desorption (Backward) Experiments with Durations Greater than 40 Days...................................................................................I-44 I-33. Plutonium Sorption Coefficients versus Eh as Predicted by PHREEQC V2.3 Model. Separate curves are shown for different pH values. ....................................................I-45 I-34. Plutonium Sorption Coefficients on Zeolitic Tuff versus Calculated Final Plutonium Concentration in Solution ............................................................................................I-47 I-35. Plutonium Sorption Coefficient on Zeolitic Tuff versus Experiment Duration for Sorption (Forward) and Desorption (Backward) Experiments....................................I-48 Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-5 November 2003 FIGURES Page I-36. Plutonium Sorption Coefficient on Zeolitic Tuff in J-13 and Synthetic p#1 versus Solution pH in Sorption (Forward) and Desorption (Backward) Experiments with Durations Greater than 40 Days...................................................................................I-48 I-37. Plutonium Sorption Coefficients on Vitric Tuff versus Calculated Final Plutonium Concentration in Solution ............................................................................................I-49 I-38. Plutonium Sorption Coefficient on Vitric Tuff versus Experiment Duration for Sorption (Forward) and Desorption (Backward) Experiments....................................I-50 I-39. Plutonium Sorption Coefficient on Vitric Tuff in J-13 and Synthetic p#1 versus Solution pH in Sorption (Forward) and Desorption (Backward) Experiments with Durations Greater than 40 Days...................................................................................I-51 I-40. Protactinium Sorption Coefficients vs. pH .......................................................................I-52 I-41. Barium and Radium Sorption Coefficients on Devitrified Tuff versus Calculated Final Barium or Radium Concentrations in Solution ..................................................I-54 I-42. Barium and Radium Sorption Coefficients on Devitrified Tuff versus Experiment Duration for Sorption (Forward) and Desorption (Backward) Experiments ...............I-55 I-43. Barium and Radium Sorption Coefficients on Devitrified Tuff versus Experiment Duration for Sorption (Forward) and Desorption (Backward) Experiments ...............I-56 I-44. Barium and Radium Sorption Coefficients on Zeolitic Tuff versus Calculated Final Barium or Radium Concentrations in Solution............................................................I-57 I-45. Isotherm Diagram for Ba Sorption on Zeolitic Tuff YM-38 in J-13 Water......................I-57 I-46. Barium and Radium Sorption Coefficient on Zeolitic Tuff in J-13 versus Experiment Duration for Sorption (Forward) and Desorption (Backward) Experiments ...............I-58 I-47. Barium and Radium Sorption Coefficients on Vitric Tuff versus Calculated Final Barium or Radium Concentrations in Solution............................................................I-59 I-48. Barium and Radium Sorption Coefficient on Vitric Tuff versus Experiment Duration for Sorption (Forward) and Desorption (Backward) Experiments ..............................I-59 I-49. Strontium Sorption Coefficients on Devitrified Tuff versus Calculated Final Strontium Concentration in Solution ...........................................................................I-61 I-50. Strontium Sorption Coefficients on Devitrified Tuff versus Experiment Duration for Sorption (Forward) and Desorption (Backward) Experiments. Data points include experiments in which fines were not removed from sample...........................I-62 I-51. Strontium Sorption Coefficients on Devitrified Tuff versus Experiment Duration for Sorption (Forward) and Desorption (Backward) Experiments. Fines fraction removed from all experiments. ....................................................................................I-63 I-52. Strontium Sorption Coefficients on Zeolitic Tuff versus Calculated Final Strontium Concentration in Solution ............................................................................................I-64 I-53. Strontium Sorption Coefficients versus Calculated Final Solution Concentration in M/L for Sample YM-38 in J-13 Water ........................................................................I-65 I-54. Strontium Sorption Coefficients on Zeolitic Tuff versus Experiment Duration for Sorption (Forward) and Desorption (Backward) Experiments....................................I-65 I-55. Strontium Sorption Coefficients on Zeolitic Tuff with Fine Fraction Removed versus Experiment Duration for Sorption (Forward) and Desorption (Backward) Experiments .................................................................................................................I-66 Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-6 November 2003 FIGURES Page I-56. Strontium Sorption Coefficients on Vitric Tuff versus Calculated Final Strontium Concentration in Solution ............................................................................................I-67 I-57. Strontium Sorption Coefficients on Vitric Tuff versus Experiment Duration for Sorption (Forward) and Desorption (Backward) Experiments....................................I-67 I-58. Thorium Sorption Coefficients on Tuff versus Calculated Final Thorium Concentration in Solution ............................................................................................I-69 I-59. Thorium Sorption Coefficients on Tuff Versus pH ..........................................................I-70 I-60. Uranium Sorption Coefficients on Devitrified Tuff Versus Calculated Final Uranium Concentration in Solution .............................................................................I-71 I-61. Uranium Sorption Coefficients on Devitrified Tuff Versus Experiment Duration for Sorption (Forward) and Desorption (Backward) Experiments....................................I-72 I-62. Uranium Sorption Coefficients on Devitrified Tuff versus pH. Model curves are from the PHREEQC Surface Complexation Model. ...................................................I-73 I-63. Uranium Sorption Coefficients on Devitrified Tuff in p#1 Water versus pH. Model curve is from the PHREEQC Surface Complexation Model.......................................I-73 I-64. Uranium Sorption Coefficients on Zeolitic Tuff versus Calculated Final Uranium Concentration in Solution ............................................................................................I-74 I-65. Uranium Sorption Coefficients on Zeolitic Tuff as a Function of Experiment Duration .......................................................................................................................I-75 I-66. Uranium Sorption Coefficients for Zeolitic Tuff in J-13 Plotted as a Function of pH. Model curves derived with PHREEQC surface complexation modeling are also shown. ..........................................................................................................................I-76 I-67. Uranium Sorption Coefficients for Zeolitic Tuff in Synthetic p#1 Plotted as a Function of pH. Model curves derived with PHREEQC surface complexation modeling are also shown..............................................................................................I-76 I-68. Uranium Sorption Coefficients on Vitric Tuff versus Calculated Final Uranium Concentration in Solution ............................................................................................I-77 I-69. Uranium Sorption Coefficients on Vitric Tuff as a Function of Experiment Duration ....I-78 I-70. Uranium Sorption Coefficients for Vitric Tuff in J-13 Plotted as a Function of pH. Model curves derived with PHREEQC surface complexation modeling are also shown. ..........................................................................................................................I-79 I-71. Uranium Sorption Coefficients for Vitric Tuff in Synthetic p#1 Plotted as a Function of pH. Model curves derived with PHREEQC surface complexation modeling are also shown..............................................................................................................I-79 Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-7 November 2003 TABLES Page I-1. Surface Areas1 in m2/g for Yucca Mountain Tuffs .................................................................11 I-2. Composition of Yucca Mountain UZ Waters ........................................................................12 Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-8 November 2003 INTENTIONALLY LEFT BLANK Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-9 November 2003 I.1. INTRODUCTION This attachment provides the bases for the derivation of sorption coefficient probability distributions used in the unsaturated zone (UZ) transport calculations. The sorption coefficient data on which the distributions are based were obtained in laboratory experiments in which crushed rock samples from the Yucca Mountain site were contacted with groundwaters (or simulated groundwaters) representative of the site, spiked with one or more of the elements of interest. Sorption experiments have been carried out as a function of time, element concentration, atmospheric composition, particle size, and temperature. In some cases, the solids remaining from sorption experiments were contacted with unspiked groundwater in desorption experiments. The sorption and desorption experiments together provide information on the equilibration rates of the forward and backward sorption reactions. For elements that sorb primarily through surface complexation reactions, the experimental data are augmented with the results of modeling calculations using PHREEQC V2.3 (BSC 2001 [155323]). The inputs for the modeling calculations include groundwater compositions, surface areas, binding constants for the elements of interest, and thermodynamic data for solution species. These modeling calculations provide a basis for interpolation and extrapolation of the experimentally derived sorption coefficient dataset. The primary controls on sorption behavior of the elements of interest in the UZ flow system include the detailed characteristics of mineral surfaces in the rock units through which water flows from the repository to the saturated zone. They also include the detailed chemistry of pore waters and perched waters in the UZ along this flow path, the sorption behavior of each element, and the concentrations of the various radionuclides in the groundwaters. These parameters will be discussed in the following sections. This Attachment provides probability distributions for the sorption coefficient of each element of interest, on the three major rock types (devitrified, zeolitic, and vitric) found in the UZ. The influence of expected variations in water chemistry, radionuclide concentrations, and variations in rock surface properties within one of the major rock types are incorporated into these probability distributions. I.2. AQUIFER MATRIX COMPOSITIONS ALONG MOST PROBABLE TRANSPORT PATHWAYS Data on rock mineralogic compositions in the UZ are provided in DTN: MO9910MWDISMMM.003 [119199]. These data have been incorporated into a site mineralogic model. There are three dominant rock types in the part of the UZ along potential flow paths from the repository to the saturated zone: devitrified tuff, zeolitic tuff, and vitric tuff. Devitrified tuff is composed primarily of silica (quartz and cristobalite) and alkali feldspar. It may also contain trace amounts of mica, hematite, calcite, tridymite, kaolinite, and hornblende, as well as minor amounts (<25%) of smectite and/or zeolite. For the purposes of this analysis, sorption coefficient distributions for devitrified tuff are based on data obtained on samples that are composed primarily of silica phases and feldspar with only trace amounts of other phases. Although devitrified tuff samples that contain significant amounts (> 5%) of clays or zeolites generally have higher sorption coefficients Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-10 November 2003 than samples that do not, the distribution of these types of altered rocks along the flowpath is not well enough established to reliably incorporate these rock types into the sorption coefficient probability distributions. Sorption coefficient distributions for zeolitic tuff exclusively use samples that contain more than 50% zeolite, with the balance made up of clay, silica phases, alkali feldspar, and/or glass. For vitric tuffs, sorption coefficient distributions are based on samples that contain more than 50% glass, with the remainder made up predominantly of feldspar, silica phases, zeolites, and clay. The total number of samples for each rock type used in the analysis in this section included 38 devitrified tuffs, 34 zeolitic tuffs, and 7 vitric tuffs. Not all of these samples were obtained from the UZ. Nonetheless, these samples as a whole are taken to be representative of devitrified, zeolitic, and vitric tuffs in the Yucca Mountain UZ. As explained below, surface areas for the tuffs are inputs to PHREEQC V2.3 (BSC 2001 [155323]) calculations. Triay et al. (1996 [101023], p. 62) presented surface-area analyses for 23 tuff samples. A table of the analyses with averages (omitting these outlying samples) is presented in Table I-1. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-11 November 2003 Table I-1. Surface Areas1 in m2/g for Yucca Mountain Tuffs Not Ground Dry-sieved2 Wet sieved2 With J-13 water trial #1 Wet sieved2 With J-13 water trial #2 Wet sieved2 With UE-25 p#1 water USW G1-732 Devitrified 2.1 2.7 2.6 2.5 3.3 USW G1-1936 Devitrified 4.5 4.9 3.6 3.9 3.7 USW G4-270 Devitrified 2.0 6.4 5.1 5.0 USW G4-275 Devitrified 2.9 4.5 USW G4-2570 Devitrified 2.8 3.6 2.9 2.9 2.8 USW GU3-747 Devitrified 2.2 2.9 2.8 2.4 2.8 USW GU3-2325 Devitrified 1.8 2.5 2.2 1.8 2.5 Average 2.6 3.9 3.2 2.7 3.4 USW G1-1405 Zeolitic 32 28 26 31 USW G2-1813 Zeolitic 34 USW G2-1951 Zeolitic 66 USW GU3-1992 Zeolitic 32 USW G4-1506 Zeolitic 22 30 27 25 USW G4-1529 Zeolitic 37 21 22 31 USW G4-1530 Zeolitic 40 41 USW G4-1625 Zeolitic 28 27 28 33 USW G4-1772 Zeolitic 23 22 23 23 23 USW G4-2077 Zeolitic 19 18 Average 28.7 26.7 32.3 23 28.6 USW G2-767 Vitric 0.89 1.1 0.62 0.71 0.87 USW GU3-1249 Vitric 0.52 0.92 0.99 0.87 USW GU3-1407 Vitric 1.7 3.3 3.0 3.2 Average 1.0 1.8 1.5 0.71 1.6 Source: Triay et al. (1996 [101023], pp. 5, 6, 8, 62, and Appendix C) NOTE: 1 BET method 2 75-500 µm fraction I.3. RADIONUCLIDES OF INTEREST AND THEIR POTENTIAL CONCENTRATION RANGES IN TRANSPORT SYSTEM The list of radionuclides for which sorption coefficient data are required was derived in Radionuclide Screening (BSC 2002 [160059], Table 10). The list includes isotopes of Am, Cs, Np, Pa, Pu, Ra, Sr, Th, and U. Because different isotopes of a given (heavy) element behave the same in chemical reactions, this discussion will focus on this list of elements. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-12 November 2003 Sorption coefficients for the radionuclides of interest can be a function of the concentrations of the radionuclides present in solution. Thus, experiments have been carried out as a function of radionuclide concentrations for most of the elements of interest. In most cases, experiments were carried out over a range of concentrations up to a solubility limit. Above the solubility limit, a solid phase incorporating the element of interest is precipitated out of solution. Therefore, the concentration of an element in solution cannot rise much higher than the solubility limit. Experiments in which the final solution concentrations for a given element of interest exceeded the solubility limit are not useful in this analysis and will be rejected. Because experiments have been carried out at concentrations up to the solubility limit for most elements of interest, the experimental results and the probability distributions derived from them include this dependency. The only element for which the experimental concentrations did not approach a solubility limit was Cs. The solubility of cesium compounds is very high in Yucca Mountain waters (BSC 2001 [163152], Section 6.17). Thus, the sorption coefficient probability distributions for cesium must be calibrated to the Cs concentrations expected in the UZ. I.4. WATER COMPOSITIONAL RANGES ALONG TRANSPORT PATHWAYS The chemistry of pore waters and perched waters in the UZ along potenial flowpaths to the accessible environment is discussed in the Analysis of Geochemical Data for the Unsaturated Zone (BSC 2002 [160247]). In the UZ, two rather distinct water types exist in the ambient system. One is perched water and the other is pore water. Perched water is generally more dilute than pore water. Table I-2. Composition of Yucca Mountain UZ Waters Element Units J-13b Range of concentrations in perched watersc Range of concentrations in pore waters within and below repository horizond UE 25 p#1 Carbonate aquifera Synthetic p#1 Ca mg/L 12 2.9–45 0.3–91.8 100 Mg mg/L 2.1 0–4.1 0–24.6 39 Na mg/L 42 34–98 3–207 150 261 K mg/L 5 3.6–5.8 1.4–148.7 12 SiO2 mg/L 57 7.7–64 5–352 41 Cl mg/L 7.1 4.1–16 6–130 28 SO4 mg/L 17 4–223 6–101.1 160 HCO3 mg/L 124 112–197 8–384 694 691 pH 7.2 7.6–8.7 6.7–9.7 6.6 SOURCE: aDTN: MO0007MAJIONPH.010 [151523] bDTN: MO0007MAJIONPH.013 [151530] cDTN: GS951208312272.004 [165858] dDTN: GS010708312272.002 [156375]; GS011008312272.004 [165859]; GS990208312272.001 [146134]; GS010608312272.001 [165860]; GS000608312271.001 [153407] Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-13 November 2003 The J-13 and UE p#1 waters were used in sorption experiments as end-member compositions intended to bracket the impact of water composition on sorption coefficients. Because water from the UE p#1 well was not available to the experimental program at all times, a synthetic p#1 water was developed. This water (Table I-2) was primarily intended to have a bicarbonate concentration similar to that found in UE p#1. It was used primarily in experiments with uranium, neptunium, and plutonium because the solution and sorption behavior of these elements is sensitive to the bicarbonate and carbonate concentrations in solution. I.5. EXPERIMENTAL TECHNIQUES Most of the experimental data upon which the conclusions of this section are based were obtained at Los Alamos National Laboratories over a period of 25 years. Over this time period, there were minor changes in the experimental techniques used to obtain sorption coefficient data. However, the main technique was developed early on and was maintained through most of the 25-year time frame. It is the intent of this section to provide an overview of the experimental techniques used to obtain sorption coefficients. The details of the various techniques used (e.g., radioactivity counting techniques) are provided in the quality assurance procedures developed for this purpose. These procedures can be obtained from the YMP. The basic technique for the laboratory determination of sorption coefficients involved the contact of a groundwater sample, spiked with the radionuclide of interest, with a crushed sample of tuff or alluvium. The rock sample was generally obtained as a core sample. The rock and water samples were not sterilized and therefore contain representative microbial biota from the UZ. The rock was crushed in a jaw crusher and subsequently sieved to a selected grain-size fraction. The sieving process was usually carried out under water. Initially, several different grain-size fractions were used in the experiments. With experience, it was concluded that the 75–500 µm fraction was the most appropriate for use in these types of experiments. Results for samples which included all sizes below a certain grain size (e.g., <35 µm) tended to produce higher sorption coefficient values than the 75– 500 µm fraction. This was thought to result from increased surface area, the creation of active surface sites, or mineral fractionation. Because the tuffs are very fine grained (i.e., crystal sizes on the order of 10—20 µm), crushing would not produce significant increases in mineral surface area. The idea that active sites were created by crushing the rock samples was negated by analysis of sorption coefficients over a range of grain sizes from a single rock sample. This analysis (Rogers and Meijer 1993 [123127], pp. 1511–1512) showed that different size fractions do not necessarily produce different sorption coefficient values. This was thought to reflect the very fine-grained nature of the tuffs. Mineral fractionation is the process by which certain minerals such as clays become enriched in a very fine-grain-size fraction. This is the most likely explanation for the higher sorption coefficients obtained in experiments that included fines. One gram of crushed rock material was added to a test tube with 20 mL groundwater spiked with the radionuclide(s) of interest. The test tube was put on a shaker table for a pre-determined period to allow reaction to occur. After the predetermined time had passed, Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-14 November 2003 the solution was separated from the solid phase by either centrifugation or filtration. Centrifugation was preferred for those elements thought to have an affinity for the filter medium. The separations were not always perfect, due to various experimental constraints. In some cases, the solid fraction was separately counted. A sorption coefficient was usually calculated from the difference between the initial and final solution concentrations. Corrections were generally made for sorption onto the surface of the test tube during the equilibration (shaking) period. Some potential sources of errors and experimental artifacts that may pertain to the sorption coefficients obtained at Los Alamos include weighing errors, counting errors, errors resulting from solutions being oversaturated with the element of interest, errors from imperfect solid/liquid separations, errors from inaccurate correction for sorption onto container walls, recording errors, transcription errors, inadvertent laboratory errors, and calculation errors. These errors cannot be quantitatively assessed. However, their existence will become apparent in the scatter of the data on diagrams presented in a later section. I.6. APPROACH TO THE DERIVATION OF Kd RANGES FOR MAJOR ROCK TYPES IN THE YUCCA MOUNTAIN FLOW SYSTEM. The derivation of sorption coefficient probability distributions for the elements of interest on the major rock types in Yucca Mountain involves both an evaluation of available experimental data and sorption modeling. Experimental data will be used to evaluate the impact of variations in rock sorption properties within each of the rock types, radionuclide concentrations, sorption kinetics, and water chemistry on sorption coefficients for the elements of interest. The radionuclides of interest are divided into three groups of radioelements. For the first group, including Am, Np, Pa, Pu, Th and U, experimental data is used to evaluate the impact of radionuclide concentrations, sorption kinetics, and variations in water chemistry on sorption coefficients. Surface complexation modeling is used to further evaluate the impact of variations in water chemistry and surface area on sorption coefficients. The surface complexation models used in this analysis are based on the code PHREEQC V2.3 (BSC 2001 [155323]). The binding constants required for surface complexation modeling are either obtained from the literature or derived from experimental data involving sorption of the radioelement on quartz (see discussion in Section I.7). In the second group of elements, including Cs, Pa, Ra, and Sr, the ranges of sorption coefficient values for the major rock types are derived directly from the available experimental data and the ranges for environmental variables expected in the transport system. Although it would be preferable to have a theoretical model to evaluate /the impact of variations in water chemistry and rock chemistry on sorption coefficients for these radionuclides, sufficient data are not available to properly constrain such a model. For the third group, including carbon, iodine, and technetium, the sorption coefficient is set to zero in volcanic rocks and in alluvium. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-15 November 2003 I.7. SURFACE COMPLEXATION MODELING The PHREEQC surface complexation model used in this analysis is a nonelectrostatic model (Parkhurst 1995 [142177], p. 10–11). Inputs required for modeling include deprotonation constants, binding constants for elements of interest, total site concentration, water chemistry, and a thermodynamic database for solution species and solids. Because the tuffs contain up to 80 weight percent silica (Broxton et al. 1986 [100023], p. 39), a silica surface was used to represent the mineral surfaces in the tuffs. The surface complexation models included the effects of competition from common constituents in the rock such as Ca, Mg, Na, K, and Al. Deprotonation constants for silica were obtained from Dixit and Van Cappellen (2002 [162985], p. 2565). Binding constants were obtained from Dixit and Van Cappellen (2002 [162985], p. 2565) for aluminum on silica, from Triay et al. (1996 [100422], Table 22) for potassium and calcium on silica, and from Marmier et al. (1999 [162986], p. 228) for sodium on silica (the same value was used for potassium). The total site concentrations were obtained from surface areas reported in Triay et al. (1996 [101023], p. 62) and a site density of 2.3 sites/nm2 as recommended by Pabalan et al. (1998 [162987], p. 124). The thermodynamic database used for the modeling was the LLNL.DAT database supplied with the baselined YMP version (V2.3; STN: 10068-2.3-00 [155323]) of the PHREEQC code. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-16 November 2003 I.8. DISCUSSION AND ANALYSIS OF EXPERIMENTAL SORPTION COEFFICIENT DATA AND SURFACE COMPLEXATION MODELING The data and modeling results for each element are discussed in separate sections. The sections are arranged in alphabetical order as follows: A. AMERICIUM B. CESIUM C. NEPTUNIUM D. PLUTONIUM E. PROTACTINIUM F. RADIUM G. STRONTIUM H. THORIUM I. URANIUM Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-17 November 2003 I.8.A. AMERICIUM According to Nitsche et al. (1993 [155218], p. 78, 86), the solubility of americium in J-13 water is controlled by Am(OH)CO3. At pH = 8.5, the solubility is 2.4 (+/-1.9) × 10-9 moles/Liter at 25°C. At pH = 7.0, the solubility is 1.2 (+/-0.3) × 10-9 moles/Liter at 25°C. Devitrified Tuff Sorption coefficients measured on devitrified tuffs are plotted versus calculated final solution concentrations in Figure I-1. The points identified by the term “sorption” refer to experiments in which J-13 water spiked with americium was contacted with devitrified tuff. The points identified by the term “desorption” refer to experiments in which the solid remaining after the “sorption” step was contacted with unspiked J-13. The steep negative slopes evident at the higher concentrations for individual samples reflect a mass-balance constraint. That is, the final solution concentration is calculated using the starting solution concentration and the measured sorption coefficient. This forces a linear dependence of the sorption coefficient on the final solution concentration. The offset between the points for the various samples primarily reflects different starting concentrations. The position of a sample along a slope reflects the degree of equilibration among various experiments on the same sample with the same starting concentration. When the sorption experiments are initiated, these trends start on the abscissa at the initial concentration. As experiment duration is increased, americium is sorbed to the solid phase, the solution concentration decreases correspondingly, and the Kd increases. At equilibrium, the solution concentration remains constant with time, and the Kd is at a maximum value. For desorption experiments, the trend is reversed. As the duration of a desorption experiment increases, the solution concentrations increase as americium is released from the solid, and the Kd decreases. At equilibrium, the Kd obtained from a desorption experiment should approach the Kd value obtained from the associated sorption experiment, unless the isotherm is quite nonlinear (in which case the trends could be offset). The offset would be caused by a different total americium concentration in the sorption and desorption experiments. The Kd value on which the two trends converge is considered the equilibrium value. As shown in Figure I-1, the calculated final solution concentrations were higher than the solubility of Am(OH)CO3 in numerous experiments with samples JA-32 and YM-54. Thus, the results for these experiments must be discounted. For the remaining experiments, the sorption coefficients range from 1,000 to more than 10,000 mL/g. In one sample (GU3-688), americium concentrations were analyzed by isotope dilution mass spectrometry (Triay et al. 1991 [104129], p. 144). This technique has higher sensitivity than the (radioactivity) counting techniques normally used, thereby allowing the use of lower americium concentrations. The result obtained for this sample is in the range of the results obtained for the other samples. The sorption coefficients derived from desorption experiments are generally larger than those derived from the sorption experiments on a given sample (Figure I-1). This could result from slow desorption kinetics or a nonlinear isotherm. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-18 November 2003 Americium sorption coefficients are plotted versus duration of the sorption and desorption experiments in Figure I-2. The trend in the data (ignoring data from samples JA-32 and YM-54) suggests americium sorption coefficient values are not very sensitive to the duration of the experiments except in those experiments with very short durations. Thus, the kinetics of the americium sorption reactions appear to be relatively fast. Americium on Devit. Tuff in J-13 10 100 1000 10000 100000 1.00E-15 1.00E-12 1.00E-09 1.00E-06 Final Am Concentration (mol/L) Am Kd (mL/g) Sorption G1-1883 Desorption G1-1883 Sorption GU3-0433 Desorption GU3-0433 Sorption JA-32 Desorption JA-32 Sorption YM-22 Desorption YM-22 Sorption YM-54 Desorption YM-54 GU3-688 DTN: LA0305AM831341.001 [163789] Figure I-1. Americium Sorption Coefficients on Devitrified Tuff versus Calculated Final Cesium Concentration in Solution Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-19 November 2003 Americium on Devit. Tuff in J-13 10 100 1000 10000 100000 0 50 100 Experiment Duration (Days) Am Kd (mL/g) Sorption G1-1883 Desorption G1-1883 Sorption GU3-0433 Desorption GU3-0433 Sorption JA-32 Desorption JA-32 Sorption YM-22 Desorption YM-22 Sorption YM-54 Desorption YM-54 GU3-688 DTN: LA0305AM831341.001 [163789] Figure I-2. Americium Sorption Coefficients for Devitrified Tuff as a Function of Experiment Duration The effects of variations in water chemistry on americium sorption coefficients have been not been tested experimentally except for variations in final solution pH. All the americium sorption and desorption experiments carried out with Yucca Mountain samples used J-13 water. Americium sorption and desorption coefficients in J-13 water are plotted versus pH in Figure I-3. It is evident that there is no clear trend among the data points plotted. The variations in sorption coefficients observed in multiple experiments with the same rock sample could reflect (radioactivity) counting statistics, long-term stability of the counting equipment, sorption kinetics, imperfect separation of the solid and liquid phases, and/or the consistency of adsorption of americium to the walls of the experimental containers. Counting statistics are generally a small percentage of the error in the measured sorption coefficient. Sorption kinetics may explain some of the scatter but not all of it. The other factors are likely significant, but they cannot be quantified with the available data. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-20 November 2003 Americium on Devit. Tuff in J-13 10 100 1000 10000 100000 6 7 8 9 10 pH Am Kd (mL/g) Sorption G1-1883 Desorption G1-1883 Sorption GU3-0433 Desorption GU3-0433 Sorption JA-32 Desorption JA-32 Sorption YM-22 Desorption YM-22 Sorption YM-54 Desorption YM-54 GU3-688 Model J-13 (2.8m2/g) Model p#1-v (2.8 m2/g) DTNs: LA0306AM831343.001 [164949]; LA0305AM831341.001 [163789] Figure I-3. Americium Sorption Coefficients on Devitrified Tuff Versus pH of J-13 Water To gauge the potential impact of variations in water chemistry on americium sorption coefficients, surface complexation modeling was carried out with PHREEQC. In the modeling, it was assumed all surface sites on devitrified tuff were equivalent to surface sites on silica. Binding constants for americium species on silica were derived by fitting data presented by Beall et al. (1986 [162983], p. 502) for the sorption of americium onto quartz (2.8 m2/g; Allard et al. 1980 [104410], p. 478) in Oak Ridge National Laboratory (ORNL) Standard Water. This water is similar in composition to J-13. As shown in Figure I-4a, the sorption coefficients for americium on quartz (Beall et al. 1986 [162983], p. 502) as a function of pH could be fit very well using two surface reactions involving an americium sulfate complex and an americium carbonate complex. These reactions are listed in Table I-3. Table I-3. Surface Complexation Reactions for Americium SiOH + Am3+ + SO4 2- . SiOAmSO4 + H+ Log K = 5.5 SiOH + Am3+ + CO3 2- . SiOAmCO3 + H+ Log K = 6.5 Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-21 November 2003 Americium on Qtz in ORNL Water (Beall et al., 1986) 1 10 100 1000 10000 3 5 7 9 pH Kd (mL/g) Data Model DTN: LA0306AM831343.001 [164949]; LA0305AM831341.001 [163789] Figure I-4a. Americium Sorption Coefficients on Quartz from Beall et al. (1986 [162983]) and Model Fit Using a surface area of 2.8 m2/g, americium sorption coefficients were calculated for “devitrified tuff” in J-13 as a function of pH. The model curves are shown in Figure I-3. The fact that the model sorption coefficient for J-13 at pH = 8.5 lies in the middle of the range of experimental values provides confidence that the model is reasonable. The model curves move up or down linearly with surface area. Thus, at least part of the range of experimental values could reflect variations in devitrified tuff surface areas. The model curve shown in Figure I-3 indicates that sorption coefficients for americium will increase with decreasing pH in the pH range from 7.0 to 9.0. Thus, in terms of pH, the experimentally derived sorption coefficients are at the low end of the range of coefficients to be expected in the saturated zone. The effect of variation in major ion chemistry of groundwater is shown by the curve calculated for p#1-v groundwater. Americium sorption coefficients calculated using p#1-v water are similar to those calculated using J-13 water. Thus, variations in water chemistry are not expected to have a major impact on americium sorption coefficients. On the basis of the experimental data and model curves plotted in Figure I-3, the range of americium sorption coefficients expected for devitrified tuffs in the saturated volcanic section at Yucca Mountain is selected as 1,000–10,000 mL/g. The probability distribution type selected is a truncated normal distribution. The data for JA- 32 and YM-54 were discounted in derivation of the distribution because the final solutions in the experiments with these samples were oversaturated with Am(OH)CO3. Zeolitic Tuff The measured sorption coefficients for zeolitic tuff are plotted versus calculated final solution concentrations in Figure I-4b. The calculated final solution concentrations were higher than the solubility of Am(OH)CO3 in essentially all the experiments with sample JA-18. Therefore, the sorption coefficients obtained for this sample are discounted in the derivation of the sorption coefficient probability distribution for americium on zeolitic tuff. Sorption experiments with Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-22 November 2003 sample G4-1952 were close to saturation with Am(OH)CO3. The sorption coefficients obtained for this sample will be used in the derivation of the distribution. Am on Zeolitic Tuff 10 100 1000 10000 100000 1.00E-11 1.00E-10 1.00E-09 1.00E-08 1.00E-07 1.00E-06 Final Am Concentration (mol/L) Am Kd (mL/g) Sorption G2-1952 Desorption G4-1952 Sorption G4-1502 Sorption YM-38 Desorption YM-38 Sorption YM-49 Desorption YM-49 Sorption JA-18 Desorption JA-18 DTN: LA0305AM831341.001 [163789] Figure I-4b. Americium Sorption Coefficients on Zeolitic Tuff versus Calculated Final Americium Concentration in Solution Americium sorption coefficients for zeolitic tuff are plotted versus duration of the sorption and desorption experiments in Figure I-5. As with the devitrified tuffs, the trend in the data (excluding sample JA-18) suggests americium sorption coefficient values obtained in experiments with solutions undersaturated with Am(OH)CO3 are not very sensitive (<10X) to the duration of the experiments, except in those experiments with short durations (e.g., sample G4-1502). Thus, the kinetics of the americium sorption reactions on zeolitic tuff appear to be relatively fast. The surface complexation model developed for zeolitic tuff is similar to the model developed for devitrified tuff, except that a surface area of 28 m2/g was used. The model sorption coefficient at pH = 8.5 lies at the high end of the range of experimental values (Figure I-6). It might be expected that the sorption coefficients for zeolitic tuffs would be on the order of 10 times larger than those for devitrified tuffs to reflect the factor of 10 increase in surface area. The fact that the americium sorption coefficients obtained for zeolitic tuffs are not, on average, a factor of 10 larger than those measured in devitrified tuffs suggests not all the surface area in the zeolitic tuffs may be accessible to americium. Alternatively, the binding sites in the interior of zeolite crystals may not have a high affinity for americium species. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-23 November 2003 Americium on Zeolitic Tuff in J-13 10 100 1000 10000 100000 0 50 100 Experiment Duration (Days) Am Kd (mL/g) Sorption G4-1502 Sorption G2-1952 Desorption G2-1952 Sorption JA=18 Desorption JA-18 Sorption YM-38 Desorption YM-38 Sorption YM-49 Desorption YM-49 DTN: LA0305AM831341.001 [163789] Figure I-5. Americium Sorption Coefficients for Zeolitic Tuff as Function of Duration of Sorption and Desorption Experiments As with devitrified tuff, the model curves for zeolitic tuffs shown in Figure I-6 indicate that sorption coefficients for americium will increase with decreasing pH in the pH range from 7.0 to 9.0. Thus, in terms of pH, the experimentally derived sorption coefficients are at the low end of the range of coefficients to be expected in the unsaturated zone. The effect of variation in the major ion chemistry of groundwater is shown by the curve calculated for p#1 groundwater. Americium sorption coefficients calculated using p#1 water are similar to the coefficients calculated using J-13 water. On the basis of the experimental data and model curves plotted in Figure I-6, the range of americium sorption coefficients selected for zeolitic tuffs in the saturated volcanic section at Yucca Mountain is 1,000–10,000 mL/g. The probability distribution type selected is a truncated normal distribution. The upper end of this distribution could have been set at 100,000 mL/g instead of 10,000 mL/g based on the available data and modeling results. However, the 10,000 mL/g value is large enough to effectively keep americium from being transported over the regulatory time frame. In addition, by using an upper limit of 10,000 mL/g, the same distribution can be used for devitrified and zeolitic tuffs. The data for JA-18 were discounted in derivation of the distribution because the final solutions in the experiments with this sample were oversaturated with Am(OH)CO3. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-24 November 2003 Americium on Zeolitic Tuff in J-13 10 100 1000 10000 100000 1000000 6.00 7.00 8.00 9.00 10.00 pH Am Kd (mL/g) Sorption G4-1502 Sorption G2-1952 Desorption G2-1952 Sorption JA=18 Desorption JA-18 Sorption YM-38 Desorption YM-38 Sorption YM-49 Desorption YM-49 Model J-13 (28 m2/g) Model p#1-v (28 m2/g) DTNs: LA0305AM831341.001 [163789]; LA0306AM831343.001 [164949] Figure I-6. Americium Sorption Coefficient Data and Modeling Results for Zeolitic Tuff as a Function of pH Vitric Tuff The measured sorption coefficients for vitric tuff are plotted versus calculated final solution concentrations in Figure I-7. The calculated final solution concentrations were close to Am(OH)CO3 solubility in essentially all the experiments with samples GU3-1203 and GU3- 1301. Therefore, the sorption coefficients obtained for these samples are minimum values. Americium sorption experiments on vitric tuffs were all carried out for a duration of 42 days. Thus, there is no information available on the kinetics of the americium sorption reactions on vitric tuff. The dependence of Am sorption coefficients on water chemistry in vitric tuff will be very similar to the dependence calculated in devitrified tuff (Figure I-3). Thus, americium sorption coefficients will decrease with increasing pH. Since the sorption experiments with vitric tuffs were carried out at the high end of the pH range expected in the natural system, sorption coefficients in the natural system are expected to be larger than those experimentally observed. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-25 November 2003 Am on Vitric Tuff in J-13 100 1000 10000 100000 1000000 1.0E-13 1.0E-12 1.0E-11 1.0E-10 1.0E-09 1.0E-08 Calculated Final Am Concentration (mol/L) Am Kd (mL/g) Sorption G2-547 Desorption G2-547 Sorption G2-723 Desorption G2-723 Sorption GU3-1203 Desorption GU3-1203 Sorption GU3-1301 Desorption GU3-1301 DTN: LA0305AM831341.001 [163789] Figure I-7. Americium Sorption Coefficients on Vitric Tuff versus Calculated Final Americium Concentration in Solution Am Kd on Vitric Tuff in J-13 100 1000 10000 100000 1000000 6.0 7.0 8.0 9.0 pH Am Kd (mL/g) Sorption G2-547 Desorption G2-547 Sorption G2-723 Sorption GU3-1203 Desorption GU3-1203 Sorption GU3-1301 Desorption GU3-1301 DTN: LA0305AM831341.001 [163789] Figure I-8. Americium Sorption Coefficient Data and Modeling Results for Vitric Tuff as a Function of pH On the basis of the experimental data plotted in Figure I-8 and the inferred pH dependence from Figure I-3, the americium sorption-coefficient probability distribution selected for vitric tuff is a cumulative distribution with a value of 100 mL/g at 0.0, 400 mL/g at 0.5, and 1,000 mL/g at 1.0. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-26 November 2003 The range of values in this distribution could have been set at between 1,000 mL/g and 1,000,000 mL/g based on the available data and modeling results. However, the lower values were selected to acknowledge the potential for vitric units to have smaller surface areas than those reported for vitric tuff in Table I-1. Surface areas of true vitrophyres are likely to be substantially lower than the surface areas listed in Table I-1 for vitric tuffs. The data for GU3- 1203 and GU3-1301 were rejected in derivation of the distribution because the final solutions in the experiments with this sample were oversaturated with Am(OH)CO3. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-27 November 2003 I.8.B CESIUM The solubility of cesium in J-13 water at 25°C is very high. In fact, cesium concentrations in Yucca Mountain groundwaters will not have a solubility limitation (BSC 2001 [155455], p. 43). Devitrified Tuff Experimentally derived sorption coefficients for cesium on devitrified tuff are plotted against the calculated final cesium concentrations of the experiments in Figure I-9. The data points are separated into groups on the basis of when the experiments were carried out (pre-1990 = “old” and post-1990 = “new”), water type, and whether the sorption coefficient was determined from a sorption or a desorption experiment. There are also data for sorption isotherms on samples YM- 22 and G1-2840. Cs on Devitrified Tuff 1 10 100 1000 10000 1.0E-12 1.0E-10 1.0E-08 1.0E-06 1.0E-04 1.0E-02 Calculated Final Cs Conc. (mol/L) Cs Kd (mL/g) Sorption New J-13 Sorption Old J-13 Desorption Old J-13 Sorption Old p#1 YM-22 Old J-13 G1-2840 Old J-13 DTNs: LA0309AM831341.003 [165524]; LA0305AM831341.001 [163789] Figure I-9. Cesium Sorption Coefficients on Devitrified Tuff versus Calculated Final Cesium Concentration in Solution Sorption coefficients obtained in “new” experiments lie within the range defined by the “old” experiments at similar cesium concentrations. Although most of the sorption experiments resulted in Kd values greater than 100 mL/g, these experiments were carried out at cesium concentrations below 1.0 × 10-6 mol/L. At higher cesium concentrations, the Kd values obtained were between 10 and 100 mL/g. Nonlinear sorption isotherms were obtained for samples YM- 22 and G1-2840 in J-13 water. The data for sample G1-2840 were fit using a Freundlich equation as shown in Figure I-10. Sorption coefficients obtained in experiments with p#1 water fall in the middle of the cluster of points in Figure I-9 near 1 × 10-10 mol/L. Thus, there is little or no impact caused by water-chemistry variations on cesium sorption coefficients in devitrified tuff. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-28 November 2003 Cs on Devitrified Tuff G1-2840 in J-13 y = 0.0024x0.7212 1.0E-12 1.0E-10 1.0E-08 1.0E-06 1.0E-04 1.0E-12 1.0E-10 1.0E-08 1.0E-06 1.0E-04 1.0E-02 Calculated Final Cs Conc (mol/L) Cs on Solid (mol/g) DTN: LA0305AM831341.001 [163789] Figure I-10. Freundlich Isotherm Fit to Sorption Coefficient Data for Cesium on Devitrified Tuff Sample G1-2840 in J-13 Water The effects of experiment duration on the cesium sorption coefficients for devitrified tuff are shown in Figure I-11. The large range in sorption coefficients obtained at a given duration (e.g., 21 days) mainly reflects variations in solution cesium concentrations, although variations in ion exchange capacities may also contribute to the range. The range of sorption coefficient values is fairly consistent with duration when the results for samples YM-22 and G1-2840 are excluded. This range extends from just above 100 mL/g to above 1,000 mL/g. The consistency indicates cesium sorption reactions are fast. Cs on Devitrified Tuff 1 10 100 1000 10000 0 20 40 60 80 100 Experiment Duration (days) Cs Kd (mL/g) Sorption New J-13 Sorption Old J-13 Desorption Old J-13 Sorption Old p#1 YM-22 Old J-13 G1-2840 Old J-13 DTNs: LA0309AM831341.003 [165524]; LA0305AM831341.001 [163789] Figure I-11. Cesium Sorption Coefficients on Devitrified Tuff versus Experiment Duration for Sorption (Forward) and Desorption (Backward) Experiments The cesium sorption-coefficient probability distribution derived for devitrified tuff in the UZ is a uniform distribution with a range of 1–15 mL/g. This distribution was chosen to acknowledge the potential for high cesium concentrations (e.g., >10-3 mol/L) to be transported in the UZ. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-29 November 2003 Zeolitic Tuff Experimentally derived sorption coefficients for cesium on zeolitic tuff are plotted against the calculated final cesium concentrations of the experiments in Figure I-12. The data points are separated into groups on the basis of when the experiments were carried out (pre-1990 = “old” and post-1990 = “new”), water type, and on whether the sorption coefficient was determined from a sorption or a desorption experiment. There are also data for a sorption isotherm on sample YM-38. Cs on Zeolitic Tuff in J-13 1000 10000 100000 1.00E-13 1.00E-10 1.00E-07 1.00E-04 Calculated Final Cs Conc. (mol/L) Cs Kd (mL/g) Sorption New J-13 Sorption Old J-13 Desorption Old J-13 Sorption Old p#1 Desorption Old p#1 DTNs: LA0310AM831341.001[165865] Figure I-12. Cesium Sorption Coefficients on Zeolitic Tuff versus Calculated Final Cesium Concentration in Solution The range of measured cesium sorption coefficients for zeolitic tuffs is 3,500-72,000 mL/g. A sorption isotherm was obtained for sample YM-38 in J-13 water. As shown in Figure I-13, the isotherm is nearly linear. Sorption coefficients obtained in experiments with p#1 water fall at the lower end of the range of values obtained for experiments with J-13 water (Figure I-12). Thus, there is some impact from variations in water chemistry on cesium sorption coefficients in zeolitic tuff, although this impact is minor. Cs on Zeolitic Tuff YM-38 in J-13 y = 1.8197x0.9397 1.0E-12 1.0E-10 1.0E-08 1.0E-06 1.0E-04 1.0E-12 1.0E-10 1.0E-08 1.0E-06 1.0E-04 Calculated Final Cs Conc (mol/L) Cs on Solid (mol/g) DTN: LA0310AM831341.001 [165865] Figure I-13. Freundlich Isotherm Fit to Sorption Coefficient Data for Sample YM-38 in J-13 Water Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-30 November 2003 The effects of experiment duration on cesium Kd for zeolitic tuff are shown in Figure I-14. The large range in sorption coefficients obtained at a given duration mainly reflects variations in cesium solution concentrations, although there must also be some contribution from variations in ion exchange capacities of the zeolitic tuff samples used in the experiments. The range of sorption coefficient values is fairly consistent with duration. For example, the range of sorption coefficient values for 3.6-day experiments is similar to the range for the 42-day experiments. This indicates the sorption reaction kinetics are fast. Cs on Zeolitic Tuff in J-13 1000 10000 100000 0 20 40 60 80 100 Calculated Final Cs Conc. (mol/L) Cs Kd (mL/g) Sorption New J-13 Sorption Old J-13 Desorption Old J-13 Sorption Old p#1 Desorption Old p#1 YM-38 Old J-13 DTNs: LA0309AM831341.003 [165524]; LA0305AM831341.001 [163789] Figure I-14. Cesium Sorption Coefficients on Zeolitic Tuff versus Experiment Duration for Sorption (Forward) and Desorption (Backward) Experiments The cesium sorption-coefficient probability distribution selected for zeolitic tuff in the UZ is a cumulative distribution, starting with a value of 425 mL/g at 0.0, 5,000 mL/g at 0.5, and 20,000 mL/g at 1.0. The low end of this distribution was selected to acknowledge the potential for high cesium concentrations and lower-than-average ion exchange capacities during transport in the UZ. The middle value of the distribution was selected as a more representative value for cesium concentrations below 10-4 mol/L. The upper end of the range was selected as a minimum upper limit, given the potential impact of lower cesium solution concentrations and higher-thanaverage ion exchange capacities. Vitric Tuff Experimentally derived sorption coefficients for cesium on vitric tuff are plotted against the calculated final cesium concentrations of the experiments in Figure I-15. The data points are separated into groups on the basis of when the experiments were carried out (pre-1990 = “old” and post-1990 = “new”), water type, and on whether the sorption coefficient was determined from a sorption or a desorption experiment. Experiment Duration (Days) Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-31 November 2003 Cs on Vitric Tuff 10 100 1000 10000 1.0E-11 1.0E-10 1.0E-09 1.0E-08 Calculated Final Cs Conc. (mol/L) Cs Kd (mL/g) Sorption J-13 (New ) Sorption J-13 (Old) Desorption J-13 (Old) Sorption p#1 (Old) Desorption p#1 (Old) DTNs: LA0309AM831341.003 [165524]; LA0310AM831341.001[165865] Figure I-15. Cesium Sorption Coefficients on Vitric Tuff versus Calculated Final Cesium Concentration in Solution The range of measured cesium sorption coefficients for zeolitic tuffs is 50–1,000 mL/g. Sorption coefficients obtained in experiments with p#1 water fall at the lower end of the range of values obtained for experiments with J-13 water (Figure I-15). However, the p#1 experiments were performed with higher initial cesium solution concentrations than all the experiments performed with J-13 water. Thus, it is unclear to what extent water chemistry caused the sorption coefficients to be lower in p#1 water compared to J-13. The effects of experiment duration on the cesium Kd for vitric tuff are shown in Figure I-16. That the sorption coefficient values obtained in short-term experiments exceed those obtained in longer-term experiments indicates that cesium sorption reactions on vitric tuff are fast. Cs on Vitric Tuff in J-13 10 100 1000 10000 0 10 20 30 40 50 Calculated Final Cs Conc. (mol/L) Cs Kd (mL/g) Sorption New J-13 Sorption Old J-13 Desorption Old J-13 Sorption Old p#1 Desorption Old p#1 DTNs: LA0309AM831341.003 [165524]; LA0305AM831341.001 [163789] Figure I-16. Cesium Sorption Coefficients on Vitric Tuff versus Experiment Duration for Sorption (Forward) and Desorption (Backward) Experiments Experiment Duration Days Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-32 November 2003 The cesium-sorption-coefficient probability distribution selected for vitric tuff in the UZ is a cumulative distribution, starting with a value of 0.0 mL/g at 0.0, rising to a value of 2.0 mL/g at 0.5, and ending at a value of 100 mL/g at 1.0. The low-end value was selected based on the possibility that cesium concentrations could be high in the UZ (i.e., higher than those plotted in Figure I-15) and the possibility that surface areas in vitrophyric zones could be lower than those listed in Table I-1. The middle value was selected to represent more representative surface areas, high cesium concentrations, and water chemistries like p#1. The upper end of the distribution was chosen as a minimum upper limit, given the potential impact of lower cesium solution concentrations (Figure I-15), higher surface areas in ash layers, and more dilute water chemistry (e.g., like J-13). Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-33 November 2003 I.8.C. NEPTUNIUM According to CRWMS M&O (2001 [154629], p. 19), the solubility of neptunium in J-13 water at 25°C, pH = 8.4, is 7.0 ± 0.8 × 10-6 mol/L and the solubility controlling solid is Np2O5 under oxidizing conditions (Eh >180 mv). At pH = 7.0, the solubility is higher at 6.9 ± 0.8 × 10-5 mol/L under oxidizing conditions (Eh >250 mv). The solubility of neptunium in p#1 water at 25°C, pH = 8.5, is 2.5 ± 0.1 × 10-5 mol/L, and 7.3 ± 0.4 × 10-5 mol/L at 25°C, pH = 7.0 according to CRWMS M&O (2001 [154629], p.19). Thus, the solubility of neptunium in J-13 is somewhat lower than it is in p#1. Devitrified Tuff The results of sorption experiments with devitrified tuff are shown in Figure I-17. Some of the experiments with J-13 water had final neptunium concentrations above 7.0 × 10-6 mol/L. Thus, the results for these experiments should be discounted because the experiments could have been oversaturated with Np2O5. All but four of the experiments with synthetic p#1 water had final neptunium solution concentrations of less than 2.5 × 10-5 mol/L. The results for the four experiments with over-saturated final solutions will be discounted. The remaining data points, plotted in Figure I-17, suggest a dependence of the sorption coefficient on the final neptunium solution concentration. Np on Devitrified Tuff 0.01 0.1 1 10 100 1E-12 1E-10 1E-08 1E-06 1E-04 Calculated Final Np Conc. (M/l) Np Kd (mL/g) New Sorption J-13 <40 Days Old Sorption J-13 >40 Days Old Desorption J-13 >40 days New Sorption p#1 <40 Days Calculated Final Np Conc. (mol/L) DTNs: LA0309AM831341.004 [165525]; LA0305AM831341.001 [163789] Figure I-17. Neptunium Sorption Coefficients on Devitrified Tuff versus Calculated Final Neptunium Concentration in Solution Neptunium sorption experiments carried out as a function of experiment duration are shown in Figure I-18. A significant difference appears between the results for “old” and “new” experiments, with the “old” results generally having higher values than the “new” results. The most straightforward explanation is that the difference is caused by the “old” results representing Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-34 November 2003 experiments with longer durations than the “new” results. Within the “old” data points, the sorption coefficient values appear to reach a steady-state level after 42 days. Np on Devitrified Tuff 0.01 0.1 1 10 100 1E-12 1E-10 1E-08 1E-06 1E-04 Np Kd (mL/g) New Sorption J-13 <40 Days Old Sorption J-13 >40 Days Old Desorption J-13 >40 days New Sorption p#1 <40 Days Calculated Final Np Conc. (mol/L) DTNs: LA0309AM831341.004 [165525]; LA0305AM831341.001 [163789] Figure I-18. Neptunium Sorption Coefficients on Devitrified Tuff versus Calculated Final Neptunium Concentration in Solution. Experiments oversaturated with Np2O5 have been omitted. Np on Devitrified Tuff 0.01 0.1 1 10 100 0.1 10 1000 Experiment Duration (days) Np Kd (mL/g) Sorption New J-13 Sorption Old J-13 Desorption Old J-13 Sorption New p#1 DTNs: LA0309AM831341.004 [165525]; LA0305AM831341.001 [163789] Figure I-19. Neptunium Sorption Coefficient on Devitrified Tuff versus Experiment Duration for Sorption (Forward) and Desorption (Backward) Experiments. Experiments oversaturated with Np2O5 have been omitted. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-35 November 2003 The impact of variations in pH on neptunium coefficients on devitrified tuffs is shown in Figure I-20. There is a lot of scatter in the “new” data, and there do not appear to be clear positive or negative trends among these data points. Nor does there appear to be much difference between the J-13 and synthetic p#1 results. The “old” data points are more consistent and show that the neptunium sorption coefficient depends very little on pH, except at pH values less than 7.0. PHREEQC surface complexation model curves for “devitrified tuff” in J-13 and in p#1 are also plotted. Neptunium binding constants on silica were obtained from Turner et al. (1998 [162989], p. 264). The J-13 curve lies between the “old” sorption and desorption points. This placement suggests that the curve may reflect the equilibrium values of neptunium sorption coefficients on devitrified tuff better than the experimental data. In this interpretation, the sorption data points reflect experiments that have not reached an equilibrium state. Np on Devitrified Tuff 0.01 0.1 1 10 100 6 7 8 9 10 pH Np Kd (mL/g) Sorption New J-13 Sorption Old J-13 Desorption Old J-13 Sorption New p#1 Model 2.8 m2/g J-13 Model 2.8 m2/g p#1 DTNs: LA0309AM831341.004 [165525]; LA0305AM831341.001 [163789]; LA0306AM831343.001 [164949] Figure I-20. Neptunium Sorption Coefficient on Devitrified Tuff in J-13 and Synthetic p#1 versus Solution pH in Sorption (Forward) and Desorption (Backward) Experiments. PHREEQC model results for J-13 and p#1 waters also plotted The neptunium sorption-coefficient probability distribution selected for devitrified tuff in the UZ is a cumulative distribution, starting with a value of 0.0 mL/g at 0.0, increasing to a value of 0.5 mL/g at 0.5, and to a value of 6.0 mL/g at 1.0. The low end of the chosen range is selected based on the minimum value obtained in short-term experiments (up to 21 days). The upper end of the distribution was chosen as a minimum upper limit for a neptunium concentration near the solubility limit (Figure I-18), with emphasis on results from experiments with p#1 water at pH near 7.0. It is acknowledged that a higher limit could be selected for the upper end of the distribution, based on the available data and modeling results. Zeolitic Tuff Sorption coefficients on zeolitic tuff are shown as a function of calculated final solution concentration in Figure I-21. Many of the experiments with J-13 water had final neptunium concentrations above 7.0 × 10-6 mol/L. Thus, the results for these experiments should be Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-36 November 2003 discounted because the experiments could have been oversaturated with Np2O5. Many of the experiments with synthetic p#1 water had final neptunium solution concentrations close to or greater than 2.5 × 10-5 M/L. Results for these experiments must also be discounted. Np on Zeolitic Tuff 0.1 1 10 100 1.E-12 1.E-10 1.E-08 1.E-06 1.E-04 1.E-02 Calculated Final Np Conc. (M/l) Np Kd (ml/g) New Sorption J-13 <40 Days Old Sorption J-13 >40 Days Old Desorption J-13 >40 days New Sorption p#1 <40 Days Old Sorption p#1 >40 Days Old Desorption p#1 >40 Days DTNs: LA0309AM831341.004 [165525]; LA0305AM831341.001 [163789] Figure I-21. Neptunium Sorption Coefficients on Zeolitic Tuff versus Calculated Final Neptunium Concentration in Solution Removal of the oversaturated experiments from the dataset results in Figure I-22. With the oversaturated data points removed, the sorption coefficient has virtually no dependence on the calculated final solution concentration for J-13 experiments. However, there is clearly a dependence of sorption coefficient on water chemistry in the short-term experiments. If the solubility of neptunium in synthetic p#1 is less than the solubility in p#1, this dependence on water chemistry may not be real. Np on Zeolitic Tuff 0.1 1 10 100 1.E-12 1.E-10 1.E-08 1.E-06 1.E-04 Calculated Final Np Conc. (mol/L) Np Kd (mL/g) New Sorption J-13 <40 Days Old Sorption J-13 >40 Days Old Desorption J-13 >40 days New Sorption p#1 <40 Days Old Sorption p#1 >40 Days Old Desorption p#1 >40 Days DTNs: LA0309AM831341.004 [165525]; LA0305AM831341.001 [163789] Figure I-22. Neptunium Sorption Coefficients on Zeolitic Tuff versus Calculated Final Neptunium Concentration in Solution. Oversaturated experiments have been removed. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-37 November 2003 Neptunium sorption experiments carried out as a function of experiment duration are shown in Figure I-23. There is a significant difference between the results for “old” and “new” experiments, with the “old” results generally having higher values than the “new” results. The most straightforward explanation is that the difference is caused by the “old” results representing experiments with longer durations than the “new” results. Within the “old” data points, the sorption coefficient values appear to reach a steady-state level after 42 days. Np on Zeolitic Tuff 0.1 1 10 100 0.1 10 1000 Experiment Duration (days) Np Kd (mL/g) Sorption New J-13 Sorption Old J-13 Desorption Old J-13 Sorption New p#1 Sorption Old p#1 Desorption Old p#1 DTNs: LA0309AM831341.004 [165525]; LA0305AM831341.001 [163789] Figure I-23. Neptunium Sorption Coefficient on Zeolitic Tuff versus Experiment Duration for Sorption (Forward) and Desorption (Backward) Experiments. Oversaturated experiments have been omitted. The impact of variations in pH on neptunium coefficients for devitrified tuffs is shown in Figure I-24. There is a lot of scatter in the “new” data, and there do not appear to be clear positive or negative trends among these data points. Nor does there appear to be much difference between the J-13 and synthetic p#1 results. The “old” data points are more consistent and show that the neptunium sorption coefficient has very little dependence on pH, except at pH values less than 7.0. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-38 November 2003 Np on Zeolitic Tuff 0.01 0.1 1 10 100 6 7 8 9 10 pH Np Kd (mL/g) Sorption New J-13 Sorption Old J-13 Desorption Old J-13 Sorption New p#1 Sorption Old p#1 Desorption Old p#1 DTNs: LA0309AM831341.004 [165525]; LA0305AM831341.001 [163789] Figure I-24. Neptunium Sorption Coefficient on Devitrified Tuff in J-13 and Synthetic p#1 versus Solution pH in Sorption (Forward) and Desorption (Backward) Experiments. Oversaturated experiments have been omitted. The neptunium sorption-coefficient probability distribution selected for zeolitic tuff in the UZ is a cumulative distribution, starting at 0.0 mL/g at 0.0, increasing to a value of 0.5 mL/g at 0.5, and ending at a value of 6 mL/g at 1.0. The low end of the chosen range is selected based on the minimum value obtained in short-term experiments (up to 21 days). The upper end of the distribution was chosen as a minimum upper limit for a neptunium concentration near the solubility limit (Figure I-21), with emphasis on results from experiments with p#1 water at pH near 7.0. It is acknowledged that a higher limit could be selected for the upper end of the distribution, based on the available data. Vitric Tuff Sorption coefficients on vitric tuff are shown as a function of calculated final solution concentration in Figure I-25. Many of the experiments with J-13 water had final neptunium concentrations above 7.0 × 10-6 mol/L. Thus, the results for these experiments were omitted from the diagram because the experiments could have been oversaturated with Np2O5. Several of the experiments with synthetic p#1 water had final neptunium solution concentrations close to or greater than 2.5 × 10-5 mol/L. Results for these experiments were also omitted. The remaining data points show a clear dependence on final neptunium solution concentration. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-39 November 2003 Np on Vitric Tuff 0.01 0.1 1 10 100 1.E-12 1.E-10 1.E-08 1.E-06 1.E-04 Calculated Final Np Conc. (mol/L) Np Kd (mL/g) New Sorption J-13 <40 Days Old Sorption J-13 >40 Days Old Desorption J-13 >40 days New Sorption p#1 <40 Days Old Sorption p#1 >40 Days Old Desorption p#1 >40 Days DTNs: LA0309AM831341.004 [165525]; LA0305AM831341.001 [163789] Figure I-25. Neptunium Sorption Coefficients on Vitric Tuff versus Calculated Final Neptunium Concentration in Solution. Oversaturated experiments have been omitted. Neptunium sorption experiments carried out as a function of experiment duration are shown in Figure I-26. There is a trend of increasing sorption coefficient with increasing duration. Np on Vitric Tuff 0.01 0.1 1 10 100 1.0 10.0 100.0 1000.0 Experiment Duration (days) Np Kd (mL/g) Sorption New J-13 Sorption Old J-13 Desorption Old J-13 Sorption New p#1 Sorption Old p#1 Desorption Old p#1 DTNs: LA0309AM831341.004 [165525]; LA0305AM831341.001 [163789] Figure I-26. Neptunium Sorption Coefficient on Vitric Tuff versus Experiment Duration for Sorption (Forward) and Desorption (Backward) Experiments. Oversaturated experiments have been omitted. The impact of variations in pH on neptunium coefficients on vitric tuffs is shown in Figure I-27. There is a lot of scatter in the “new” data and there do not appear to be clear positive or negative trends among these data points. Nor does there appear to be much difference between the J-13 and synthetic p#1 results. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-40 November 2003 Np on Vitric Tuff 0.01 0.1 1 10 100 6.0 7.0 8.0 9.0 10.0 pH Np Kd (mL/g) Sorption New J-13 Sorption Old J-13 Desorption Old J-13 Sorption New p#1 Sorption Old p#1 Desorption Old p#1 DTNs: LA0309AM831341.004 [165525]; LA0305AM831341.001 [163789] Figure I-27. Neptunium Sorption Coefficient on Devitrified Tuff in J-13 and Synthetic p#1 versus Solution pH in Sorption (Forward) and Desorption (Backward) Experiments. Oversaturated experiments have been omitted. The neptunium sorption-coefficient probability distribution selected for vitric tuff in the UZ is a cumulative distribution starting at 0.0 mL/g at 0.0, increasing to a value of 1.0 mL/g at 0.5, and ending with a value of 3.0 mL/g at 1.0. The low end of the chosen range is selected based on the minimum value obtained in short-term experiments (up to 21 days). The upper end of the distribution was chosen as a minimum upper limit for a neptunium concentration near the solubility limit (Figure I-25), with emphasis on results from experiments with p#1 water at pH near 7.0. It is acknowledged that a higher limit could be selected for the upper end of the distribution, based on the available data. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-41 November 2003 I.8.D PLUTONIUM According to Nitsche et al. (1993 [155218], p. 54), the solubility of plutonium in J-13 water at 25°C, pH = 8.4, is 2.9 ± 0.8 × 10-7 mol/L and is not very sensitive to pH over the range from 7.0 to 8.4. The solubility of plutonium in p#1 water at 25°C, pH = 8.5, is 1.0 ± 0.1 × 10-6 mol/L, and 4.5 ± 0.4 × 10-7 mol/L at 25°C, pH = 7.0 according to Nitsche et al. (1994 [155218], p. 39). Thus, the solubility of plutonium in J-13 is somewhat lower than it is in p#1. Devitrified Tuff The experimentally derived sorption coefficients for plutonium on devitrified tuff are plotted against the calculated final plutonium concentrations of the experiments in Figure I-28. The data points are separated into groups on the basis of when the experiments were carried out (pre-1990 = “old” and post-1990 = “new”), water chemistry, experiment duration, and whether the sorption coefficient was determined from a sorption or a desorption experiment. The plotted data indicate that the “new” and “old” data show similar ranges of Kd values, that the longer-term experiments generally yield higher sorption coefficient values than the shorter term experiments, and that desorption experiments yield higher sorption coefficient values than sorption experiments. These points are discussed in greater detail below. Pu on Devitrified Tuff in J-13 1 10 100 1000 10000 1.0E-15 1.0E-13 1.0E-11 1.0E-09 1.0E-07 Calculated Final Pu Conc. (mol/L) Pu Kd (mL/g) New Sorption J-13 <40 Days New Sorption J-13 >40 Days New Desorption J-13 <40 Days Old Sorption J-13 <40 Days Old Sorption J-13 >40 Days Old Desorption J-13 <40 days Old Desorption J-13 >40 days New Sorption p#1 <40 Days New Desorption p#1 <40 Days DTNs: LA0309AM831341.005 [165526]; LA0305AM831341.001 [163789] Figure I-28. Plutonium Sorption Coefficients on Devitrified Tuff versus Calculated Final Plutonium Concentration in Solution The maximum calculated final plutonium concentration plotted in Figure I-28 is slightly less than the solubility determined by Nitsche et al. (1993 [155218], p. 54) for plutonium in J-13. Thus, the plutonium sorption coefficients plotted in Figure I-28 reflect solutions that were undersaturated with respect to the solid plutonium phase precipitated in the experiments reported by Nitsche et al. (1993 [155218], p. 54). Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-42 November 2003 The effect of the solution plutonium concentration on the sorption coefficient value obtained is shown more clearly in Figure I-29. Data are plotted for two devitrified tuff samples that contain only trace amounts of secondary phases (e.g., clays, zeolites). In addition, plotted data are restricted to 21-day experiment durations and pH between 7.2 and 9.2. As is evident in the figure, within the errors of the analyses, the measured plutonium sorption coefficients are nearly independent of the final plutonium solution concentration. Pu on Devitrified Tuff in J-13 (pH = 8.4; 21 Days) 1 10 100 1000 10000 1.0E-14 1.0E-12 1.0E-10 1.0E-08 Calculated Final Solution Pu Conc. (M/l) Pu Kd (ml/g) Sorption YM-22 Desorption YM-22 Sorption G4-272 Desorption G4-272 (pH = 7.2-9.2; 21 Days) mol/L) mL/g) DTNs: LA0309AM831341.005 [165526]; LA0305AM831341.001 [163789] Figure I-29. Plutonium Sorption Coefficient versus Calculated Final Plutonium Solution Concentration in Moles/Liter for Experiments with Samples YM-22 and G4-272. The effect of experiment duration on the plutonium Kd is shown in Figure I-31. As expected, the Kd values for sorption experiments increase with increasing duration and the Kd values for desorption experiments decrease with increasing duration. However, the increase in the sorption values is much greater than the decrease in the desorption values over the time-frame of the experiments. This may reflect the reduction of plutonium +5 and/or +6 to plutonium +4 on the mineral surfaces present in devitrified tuff as discussed in greater detail below. The trends in the sorption and desorption data points suggest they would converge to values somewhere between 100 and 1,000 mL/g. Based on the data plotted, such convergence would require more than 100 days. Note that the “old” sorption data points exceeding 1,000 mL/g are for samples that contain significant amounts of clay or zeolite. For this reason, these data are discounted in the derivation of the sorption coefficient probability distribution. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-43 November 2003 Pu on Devitrified Tuff in J-13 1 10 100 1000 10000 0 20 40 60 80 100 Experiment Duration (days) Pu Kd (mL/g) Sorption New J-13 Desorption New J-13 Sorption Old J-13 Desorption Old J-13 Sorption New p#1 Desorption p#1 New DTNs: LA0309AM831341.005 [165526]; LA0305AM831341.001 [163789] Figure I-30. Plutonium Sorption Coefficient on Devitrified Tuff versus Experiment Duration for Sorption (Forward) and Desorption (Backward) Experiments The impact of variations in pH for plutonium sorption coefficients on devitrified tuffs is shown in Figure I-31. There is a lot of scatter in the data, and there do not appear to be clear positive or negative trends in any of the data groupings. Pu on Devitrified Tuff 1 10 100 1000 10000 6 7 8 9 10 pH Pu Kd (mL/g) Sorption-New J-13 <40 Days Sorption-New J-13 >40 Days Desorption-New J-13 Sorption-Old J-13 <40 Days Sorption-Old J-13 >40 Days Desorption-Old J-13 Sorption-New p#1 <40 days Sorption-New p#1 >40 days Desorption-New p#1 <40 days DTNs: LA0309AM831341.005 [165526]; LA0305AM831341.001 [163789] Figure I-31. Plutonium Sorption Coefficient on Devitrified Tuff in J-13 and Synthetic p#1 versus Solution pH in Sorption (Forward) and Desorption (Backward) Experiments. Experiments lasting 40 days or more are plotted separately from experiments lasting less than 40 days. The lack of clear trends is also evident when the results of short-term experiments (<40 days) are removed from the dataset. This is shown in Figure I-32, in which only the longer-term data are plotted. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-44 November 2003 Pu on Devitrified Tuff 10 100 1000 10000 6 7 8 9 10 pH Pu Kd (mL/g) Sorption-New J-13 >40 Days Sorption-Old J-13 >40 Days Desorption-Old J-13 Sorption-New p#1 >40 days DTNs: LA0309AM831341.005 [165526]; LA0305AM831341.001 [163789] Figure I-32. Plutonium Sorption Coefficient on Devitrified Tuff in J-13 and Synthetic p#1 versus Solution pH in Sorption (Forward) and Desorption (Backward) Experiments with Durations Greater than 40 Days The impact of variations in the major ion composition of groundwaters are also shown in Figures I-30 through I-32. Although there is a limited set of experiments with synthetic p#1 water, the results are within the range of the results obtained in experiments with J-13 water. Thus, there is no clear evidence of the impact of water chemistry variations on plutonium sorption coefficients in devitrified tuff. A major factor not explicitly accounted for in the experimental program is the impact of variations in the Eh (oxidation/reduction potential normalized to the Standard Hydrogen Electrode) of Yucca Mountain groundwaters on plutonium sorption coefficients. The laboratory experiments upon which the data discussed in this section are based were invariably conducted under oxidizing conditions, because the waters used in the experiments contained dissolved oxygen and were in contact with the atmosphere. Nitsche et al. (1993 [155218], pp. 60–61) found that plutonium dissolved in J-13 water is present predominantly in the +5 and +6 oxidation states. If plutonium in the +5 and +6 oxidation states behaves similarly to neptunium +5 and uranium +6, respectively, as has been suggested by many investigators (e.g., Keeney-Kennicutt and Morse 1985 [106313], pp. 2577–2578), then small values (<10 mL/g) would be expected for plutonium sorption coefficients under oxidizing conditions (see Neptunium and Uranium attachment sections). The fact that plutonium sorption coefficients measured under oxidizing conditions are up to 2–3 orders of magnitude larger than expected (Figure I-28) suggests that either plutonium +5 and +6 do not behave like neptunium +5 and uranium +6 in sorption reactions, or plutonium is reduced to the +4 oxidation state on rock/mineral surfaces. Data presented by Keeney-Kennicutt and Morse (1985 [106313], p. 2577) support the latter alternative. To further pursue the latter alternative, a PHREEQC model was developed to calculate plutonium sorption coefficients as a function of Eh. To develop this model, binding constants for neptunium +5 and uranium +6 species were used for plutonium +5 and plutonium +6 species, Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-45 November 2003 respectively. For plutonium +4, binding constants were derived by fitting experimental data for plutonium sorption on quartz in Oak Ridge Reference Water published by Allard (1982 [104512], p. 61). This water composition has somewhat reducing characteristics, so that plutonium +4 would be expected to be sorbed onto the solid phase. The sorption coefficients calculated with this model using J-13 water are plotted as a function of Eh in Figure I-33. The curves plotted in Figure I-33 suggest that at Eh values less than 500 mv, plutonium is predominantly in the +4 oxidation state in the solution and on the solid phase. This Eh value is higher than the highest value measured in Yucca Mountain groundwater (Ogard and Kerrisk 1984 [100783], pp. 10, 15; DTNs: LA0206AM831234.001 [160051]; LA0206AM831234.002 [163852]). This implies that plutonium +4 will be the dominant oxidation state sorbed in most Yucca Mountain groundwaters. At higher Eh values, plutonium in solution will become progressively oxidized to plutonium +5 and eventually to plutonium +6. The end points of the curves (high Eh) plotted in Figure I-33 represent the Eh values expected when the water is in contact with the atmosphere. In the p#1 water composition, the lower ends of the curves are shifted downward somewhat to lower Kd values at a given Eh value (not shown). Pu on Devitrified Tuff (2.8 m2/g) in J-13 0.1 1 10 100 1000 10000 0 200 400 600 800 1000 Eh (mv) Pu Kd (mL/g) pH 7.0 pH 7.5 pH 8.5 DTN: LA0306AM831343.001 [164949] Figure I-33. Plutonium Sorption Coefficients versus Eh as Predicted by PHREEQC V2.3 Model. Separate curves are shown for different pH values. The large range in plutonium sorption coefficients measured in devitrified tuffs (e.g., Figure I- 28) could be explained if plutonium is present in more than one oxidation state on the rock/mineral surfaces in the devitrified tuffs. However, in the absence of definitive data for the Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-46 November 2003 oxidation state of plutonium on rock/mineral surfaces, the experimentally determined sorption coefficient values will be used to derive the probability distributions. The plutonium sorption-coefficient probability distribution derived for devitrified tuff in the UZ is a cumulative distribution starting at 10 mL/g at 0.0, increasing to a value of 70 mL/g at 0.5, and ending at a value of 200 at 1.0. A cumulative distribution was chosen for the sake of simplicity given the limited data set for p#1 water. The low end of the chosen range is selected to allow for the possibility of relatively fast flow in the UZ. It also captures the potential impact of variations in surface areas among samples used in the experiments, variations in water chemistry, and variations in plutonium concentrations up to the solubility limit. The upper end of the distribution was chosen as a minimum upper limit given the potential impacts of sorption kinetics during potentially fast flow in the UZ. Zeolitic Tuff The experimentally derived sorption coefficients for plutonium on zeolitic tuff are plotted against the calculated final plutonium concentrations in Figure I-34. As before, the data points are separated into groups on the basis of when the experiments were carried out (pre-1990 = “old” and post-1990 = “new”), water chemistry, experiment duration, and whether the sorption coefficient was determined from a sorption or a desorption experiment. The plotted data indicate that there is no clear trend of plutonium Kd with solution concentration. Most of the data from sorption experiments plot between sorption coefficient values of 100 and 1,000 mL/g. A series of “old” experiments at concentrations between 10-7 and 10-9 M/L yielded sorption coefficient values less than 100 mL/g. However, the same rock type when used in “new”experiments yielded a series of sorption coefficient values >100 mL/g. Why there is a difference of almost a factor of ten between these two sets of data is not known with certainty. Part of the answer lies in the fact that, on average, the “Old” data points represent shorter duration experiments than the “new” data points. However, other factors are likely involved. For example, the oxidation state of plutonium in the starting solution may play a part. As with devitrified tuffs, the plutonium sorption coefficients for zeolitic tuff plotted in Figure I- 34 reflect solutions that were undersaturated with the plutonium phase precipitated in the Nitsche et al. (1993 [155218], p. 54; 1994 [144515], p. 39) experiments. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-47 November 2003 Pu on Zeolitic Tuff 1 10 100 1000 10000 1.0E-15 1.0E-13 1.0E-11 1.0E-09 1.0E-07 Calculated Final Pu Conc. (mol/L) Pu Kd (mL/g) New Sorption J-13 <40 Days New Sorption J-13 >40 Days New Desorption J-13 <40 Days Old Sorption J-13 <40 Days Old Sorption J-13 >40 Days Old Desorption J-13 <40 days Old Desorption J-13 >40 days New Sorption p#1 <40 Days New Desorption p#1 <40 Days DTNs: LA0309AM831341.005 [165526]; LA0305AM831341.001 [163789] Figure I-34. Plutonium Sorption Coefficients on Zeolitic Tuff versus Calculated Final Plutonium Concentration in Solution The effects of experiment duration on the plutonium Kd for zeolitic tuff are shown in Figure I-35. As expected, the Kd values for sorption experiments increase with increasing duration and the Kd values for desorption experiments decrease with increasing duration, at least in short-term experiments. Interestingly, the decrease in longer-term desorption experiments is not very pronounced. The trends in the sorption and desorption data points suggest they would converge to values somewhere between the values of 100 and 1,000 mL/g. This is similar to the range of values predicted for devitrified tuffs. Thus, the higher surface areas of zeolitic tuffs compared to devitrified tuffs (approximately 10×) do not appear to result in higher sorption values. This effect was also observed by Pabalan et al. (1998 [162987], p. 113) in experiments with uranium sorption on zeolite. The cause for the large range of values obtained at a given value for experiment duration is not known, but may largely result from variations in the oxidation state of plutonium in the starting solutions. However, variations in surface areas and surface chemistry among the samples used in the experiments must also contribute to the range observed. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-48 November 2003 Pu on Zeolitic Tuff 1 10 100 1000 10000 0 20 40 60 80 100 Experiment Duration (days) Pu Kd (mL/g) Sorption J-13 New Desorption J-13 New Sorption J-13 Old Desorption J-13 Old Sorption p#1 New Desorption p#1 New DTNs: LA0309AM831341.005 [165526]; LA0305AM831341.001 [163789] Figure I-35. Plutonium Sorption Coefficient on Zeolitic Tuff versus Experiment Duration for Sorption (Forward) and Desorption (Backward) Experiments The impact of variations in pH on plutonium sorption coefficients on devitrified tuffs is shown in Figure I-36. Although there is a lot of scatter in the data, there do not appear to be clear positive or negative trends among any of the data groupings. Pu on Zeolitic Tuff 1 10 100 1000 10000 6 7 8 9 10 pH Pu Kd (mL/g) New Sorption <40 Days New Sorption >40 Days New Desorption <40 Days Old Sorption <40 Days Old Sorption >40 Days Old Desorption <40 days Old Desorption >40 days New Sorption P#1 <40 Days New Desorption p#1 <40 Days DTNs: LA0309AM831341.005 [165526]; LA0305AM831341.001 [163789] Figure I-36. Plutonium Sorption Coefficient on Zeolitic Tuff in J-13 and Synthetic p#1 versus Solution pH in Sorption (Forward) and Desorption (Backward) Experiments with Durations Greater than 40 Days The impact of variations in the major ion composition of groundwaters is also shown in Figure I- 36. Although there are a limited set of experiments with synthetic p#1 water, the results are largely within the range of the results obtained in experiments with J-13 water. Thus, there is no Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-49 November 2003 clear evidence that water chemistry variations have an impact on plutonium sorption coefficients in devitrified tuff. The plutonium sorption-coefficient probability distribution derived for zeolitic tuff in the UZ is a cumulative distribution, starting at 10 mL/g with a value of 100 mL/g at 0.5 and 200 mL/g at 1.0. The value at 0.5 was increased from 70 mL/g for devitrified tuff to 100 mL/g for zeolitic tuff to reflect the higher sorption coefficient values obtained with zeolitic tuff at lower pH values (7– 7.5). Vitric Tuff The experimentally derived sorption coefficients for plutonium on vitric tuff are plotted against the calculated final plutonium concentrations in Figure I-37. As before, the data points are separated into groups on the basis of when the experiments were carried out (pre-1990 = “old” and post-1990 = “new”), water chemistry, experiment duration, and on whether the sorption coefficient was determined from a sorption or a desorption experiment. The plotted data indicate that there is no clear trend of plutonium Kd with final solution concentration. As with devitrified tuffs, the plutonium sorption coefficients for vitric tuff plotted in Figure I-37 reflect solutions that were undersaturated with respect to the plutonium phase precipitated in the Nitsche et al. (1993 [155218], p.54; 1994 [144515], p. 39) experiments. Pu on Vitric Tuff 10 100 1000 10000 1.0E-15 1.0E-13 1.0E-11 1.0E-09 1.0E-07 Calculated Final Pu Conc. (mol/L) Pu Kd (mL/g) New Sorption J-13 <40 Days New Sorption J-13 >40 Days New Desorption J-13 <40 Days Old Sorption J-13 <40 Days Old Sorption J-13 >40 Days Old Desorption J-13 <40 days Old Desorption J-13 >40 days New Sorption p#1 <40 Days New Sorption p#1 >40 Days New Desorption p#1 <40 Days DTNs: LA0309AM831341.005 [165526]; LA0305AM831341.001 [163789] Figure I-37. Plutonium Sorption Coefficients on Vitric Tuff versus Calculated Final Plutonium Concentration in Solution The data in Figure I-37 for experiments with synthetic p#1 water plot in the middle of the data for experiments with J-13 water. Thus, there does not appear to be a significant impact of variations in water composition. The effects of experiment duration on the plutonium Kd for vitric tuff are shown in Figure I-38. As expected, the Kd values for sorption experiments generally increase with increasing duration. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-50 November 2003 The Kd values for the “new” and “old” desorption experiments also increase with increasing duration, unlike the results for devitrified and zeolitic tuffs. However, as a group, the desorption results decrease with increasing duration. The trends in the sorption and desorption data points suggest they would converge to values somewhere between 100 and 1,000 mL/g. This is similar to the range of values predicted for devitrified and zeolitic tuffs. The cause for the large range of values obtained at a given value for experiment duration is not known, but may largely result from variations in the oxidation state of plutonium in the starting solutions in addition to analytical errors. Variations in surface areas and surface chemistry among the samples used in the experiments must also contribute to the range observed. Pu on Vitric Tuff 10 100 1000 10000 0 20 40 60 80 100 Experiment Duration (days) Pu Kd (mL/g) Sorption New J-13 Desorption New J-13 Sorption Old J-13 Desorption Old J-13 Sorption New p#1 Desorption New p#1 DTNs: LA0309AM831341.005 [165526]; LA0305AM831341.001 [163789] Figure I-38. Plutonium Sorption Coefficient on Vitric Tuff versus Experiment Duration for Sorption (Forward) and Desorption (Backward) Experiments The impact that variations in pH have on plutonium sorption coefficients in vitric tuffs is shown in Figure I-39. Although there is a lot of scatter in the data, there are no obvious trends among any of the data groupings. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-51 November 2003 Pu on Vitric Tuff 10 100 1000 10000 6 7 8 9 10 pH Pu Kd (mL/g) Sorption J-13 New <40 Days Sorption J-13 New >40 Days Desorption J-13 New <40 Days Sorption J-13 Old <40 Days Sorption J-13 Old >40 Days Desorption J-13 Old <40 days Desorption J-13 Old >42 Days Sorption p#1 New <40 Days Desorption p#1 New <40 Days DTNs: LA0309AM831341.005 [165526]; LA0305AM831341.001 [163789] Figure I-39. Plutonium Sorption Coefficient on Vitric Tuff in J-13 and Synthetic p#1 versus Solution pH in Sorption (Forward) and Desorption (Backward) Experiments with Durations Greater than 40 Days The impact of variations in the major ion composition of groundwaters is also represented in Figures I-38 and I-39. The limited set of experiments with synthetic p#1 water are within the range of the results obtained in experiments with J-13 water, although they tend to be at the lower end of this range. The plutonium sorption-coefficient probability distribution selected for vitric tuff in the UZ is a cumulative distribution starting with a value of 10 mL/g at 0.0, increasing to a value of 100 mL/g at 0.5 and ending with a value of 200 mL/g at 1.0. The 10 mL/g value is selected to compensate for the potential for fast transport through the UZ. The value of 100 mL/g at 0.5 is intended to emphasize the median value among the sorption coefficients obtained in experiments with synthetic p#1 water. The upper limit of 200 mL/g was selected also to emphasize the p#1 water chemistry. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-52 November 2003 I.8.E PROTACTINIUM Protactinium is very insoluble in waters of the type found at Yucca Mountain. The best estimates of protactinium solubility in these waters range from 10-15 to 10-13 mol/L (Berry et al. 1989 [144728], p. 346). Devitrified Tuff The oxidation state of protactinium is +5 in groundwaters of the type found at Yucca Mountain (Cotton and Wilkinson 1988 [105732], p. 1002). No sorption coefficient data have been obtained for protactinium on devitrified tuffs from Yucca Mountain. Allard et al. (1983 [162982], p. 12) have reported protactinium sorption coefficient data for experiments with a silica sample having a surface area similar to that measured for devitrified tuffs. The solution composition used in the experiments was 0.01 M NaClO4. The initial protactinium concentration used in all experiments was 4.0 × 10-12 mol/L. Allard's data indicate that protactinium sorption coefficients vary substantially (approximately 2 orders of magnitude) as a function of pH (as shown in Figure I-40). The cause for this variation in sorption coefficients with pH is unknown. Over the pH range expected in saturated zone waters at Yucca Mountain (7–8.5), the sorption coefficients reported by Allard et al. (1983 [162982], p. 12) range from approximately 7,500 to 20,000 mL/g. The results for 6-hour experiments were similar to the results for experiments lasting up to 6 weeks. Thus, sorption kinetics for protactinium sorption reactions appear to be fast. Allard et al. (1983 [162982], p. 12) reported results for alumina that were very similar to the results they reported for silica. Protactinium 1 10 100 1000 10000 100000 0 5 10 pH Pa Kd (mL/g) Allard-SiO2 YMP-Zeol DTN: LA0305AM831341.001 [163789] Figure I-40. Protactinium Sorption Coefficients vs. pH Hydrolysis reactions appear to dominate the solution chemistry of protactinium. The hydrolysis reactions of pentavalent protactinium in water are very complex even in relatively acidic solutions (Cotton and Wilkinson 1988 [105732], p. 1002). In near-neutral solutions, the Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-53 November 2003 hydrolysis behavior of protactinium is essentially unknown. However, because hydrolysis reactions appear to dominate the solution behavior of protactinium, changes in the major ion chemistry of groundwaters are not expected to impact the sorption behavior of protactinium. Unfortunately, insufficient thermodynamic data are available to model protactinium sorption behavior using a surface complexation model. The sorption coefficient range for americium (1,000 to 10,000 mL/g) was used as a default for protactinium. This range is well within the range reported by Allard et al. (1983 [162982], p. 12) for silica and alumina. As with americium sorption coefficients, a truncated normal distribution was selected. Corroboration of this range of sorption coefficient values is provided by protactinium sorption experiments performed by Berry et al. (1989 [144728], p. 347). These authors report a range of 1,000 to 1,000,000 mL/g in protactinium sorption coefficients for rock samples including sandstone, shale, granite, and clay in contact with natural groundwaters. Zeolitic Tuff Sorption coefficients for protactinium on zeolitic tuffs from Yucca Mountain in J-13 water were reported by Rundberg et al. (1985 [101355], p. 63). The reported sorption coefficients ranged from 3.3 to 10.1 mL/g (Figure I-40). The initial solution concentrations in the experiments ranged from 1 × 10-11 to 5 × 10-14 molar. The initial solution concentration used by Allard et al. (1983 [162982], p. 6) was in the low end of this range. Thus, initial solution concentration does not appear to explain the difference between the Allard et al. (1983 [162982], p. 12) results and the results on Yucca Mountain samples. It is possible that the results on Yucca Mountain samples reflect oversaturation due to co-precipitation of protactinium with some other easily hydrolyzed species. The pH values of the final solutions in the experiments on Yucca Mountain samples were in the range 6.3–6.7. These pH values are below the range expected in saturated zone waters at Yucca Mountain. Thus, the reported sorption coefficients do not directly apply to conditions in the Yucca Mountain UZ. As shown in Figure I-40, there appears to be an adsorption edge at a pH close to 6.8, and sorption coefficients increase by approximately 2 orders of magnitude at pH>6.8. The cause for this increase is not known, but is likely related to hydrolysis reactions. Because zeolitic tuffs have greater surface area than devitrified tuffs, the sorption coefficient distribution for devitrified tuff is used as a default for zeolitic tuff. This will lead to conservative predictions of protactinium transport rates. Vitric Tuff No sorption coefficient data is available for vitric tuffs. Because vitric tuffs have surface areas similar to the surface areas of devitrified tuffs, the sorption coefficient probability distribution for devitrified tuffs is used for vitric tuffs. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-54 November 2003 I.8.F RADIUM The chemistry of radium is very similar to that of barium (Cotton and Wilkinson 1988 [105732], p. 144). Because barium has a radioactive isotope more readily measured by gamma counting than the radium isotopes, barium was used to measure sorption coefficients for radium. A limited number of experiments were performed with radium to confirm its sorption behavior relative to barium. Barite is the solubility-controlling solid for barium in Yucca Mountain groundwaters. The solubility of barite (barium sulfate) in J-13 water at 25°C is 9.0 × 10-7 mol/L (DTN: LA0306AM831343.001 [164949]). The solubility of barite in p#1 water at 25°C is 2.0 × 10-7 mol/L. At 25°C, the solubility of radium sulfate is 3.2 × 10-7 mol/L in J-13 water. Devitrified Tuff The experimentally derived sorption coefficients for barium on devitrified tuff are plotted against the calculated final barium concentrations of the experiments in Figure I-41. The data points are separated into groups on the basis of when the experiments were carried out (pre-1990 = “old” and post-1990 = “new”), water chemistry, experiment duration, and whether the sorption coefficient was determined from a sorption or a desorption experiment. Ba and Ra on Devitrified Tuff 100 1000 10000 100000 1.0E-13 1.0E-11 1.0E-09 1.0E-07 1.0E-05 Calculated Final Ba Conc. (mol/L) Ba Kd (mL/g) New Sorption J-13 <40 Days Old Sorption J-13 <40 Days Old Sorption J-13 >40 Days Old Desorption J-13 <40 Days Old Desorption J-13 >40 Days Radium J-13 >40 Days Old Sorption p#1 >40 Days DTNs: LA0309AM831341.002 [165523]; LA0305AM831341.001 [163789] Figure I-41. Barium and Radium Sorption Coefficients on Devitrified Tuff versus Calculated Final Barium or Radium Concentrations in Solution The calculated final concentration of barium in the experiments with J-13 was below saturation with barite in all but one experiment (Figure I-41). The results for this one experiment will be omitted from further consideration. The calculated final barium concentrations in experiments with p#1 are all lower than the barite-saturation value of 2.0 × 10-7 mol/L. Thus, oversaturation was not an issue in the barium sorption experiments with p#1 water. Similarly, the radium solution concentrations were all lower than the radium sulfate saturation value in J-13. Thus, these experiments were undersaturated with radium sulfate. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-55 November 2003 The data plotted in Figure I-41 show a general increase in sorption coefficient with decreasing concentration. This suggests that the isotherms for barium sorption onto devitrified tuff are nonlinear. The impact of this nonlinearity must be included in the sorption coefficient probability distribution. The effects of experiment duration on the barium Kd for devitrified tuff are shown in Figure I-42. The large range in sorption coefficients obtained at a given duration reflects variations in grain size of the crushed tuff samples used in the experiments, variations in solution concentrations, variations in surface chemistry, and analytical errors and artifacts. Experiments with crushed tuff samples that include the fines (e.g., <30 µm) often have sorption coefficients larger than samples from which the fines have been removed (e.g., 75–500 µm). This is partly due to the higher surface area of samples with fines and partly due to mineral fractionation. Mineral fractionation can occur during the sieving process and cause the preferential concentration of very fine grained minerals (e.g., clays) in the fine fraction. Ba and Ra on Devitrified Tuff 100 1000 10000 100000 0 20 40 60 80 100 Experiment Duration (Days) Ba Kd (mL/g) Sorption J-13 New Sorption J-13 Old Sorption J-13 Old Desorption J-13 Old Desorption J-13 Old Sorption p#1 Old Desorption p#1 Old Radium J-13 Old DTNs: LA0309AM831341.002 [165523]; LA0305AM831341.001 [163789] Figure I-42. Barium and Radium Sorption Coefficients on Devitrified Tuff versus Experiment Duration for Sorption (Forward) and Desorption (Backward) Experiments Figure I-43 shows the effects of experiment duration on sorption coefficients for samples with the fines removed. The few sorption coefficients with values near 10,000 mL/g represent experiments with very low solution concentrations (Figure I-41). For the remaining experiments, the total range of sorption coefficient values is substantially reduced. Further, it appears likely the range would converge to a range between 100 and 1,000 mL/g with increasing duration. The horizontal trend among data points at durations other than 21 and 42 days suggests that barium and radium sorption reactions are relatively fast. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-56 November 2003 Ba and Ra on Devitrified Tuff 100 1000 10000 100000 0 20 40 60 80 100 Experiment Duration (Days) Ba Kd (mL/g) Sorption J-13 New Sorption J-13 Old Desorption J-13 Old Sorption p#1 Old Desorption p#1 Old Radium J-13 Old DTNs: LA0309AM831341.002 [165523]; LA0305AM831341.001 [163789] Figure I-43. Barium and Radium Sorption Coefficients on Devitrified Tuff versus Experiment Duration for Sorption (Forward) and Desorption (Backward) Experiments The impact of variations in the major ion composition of groundwaters is also shown in Figures I-42 and I-43. Although there is a limited set of experiments with p#1 water, the sorption coefficients obtained (at 42 days) fall at the low end of the range of results for J-13 water. Thus, there is some water-chemistry impact, although it is substantially less than an order of magnitude. The radium sorption-coefficient probability distribution derived for devitrified tuff in the UZ is a uniform distribution with a range of 100–1,000 mL/g. The low end of the chosen range was selected based on the minimum value observed in long-term experiments (>40 days) and the potential impact of variations in water chemistry and surface areas among devitrified tuffs at Yucca Mountain. Because there are experiments within the dataset that have solution concentrations close to saturation with a barium/radium sulfate, the effect of nonlinear isotherms is incorporated into the distribution. The upper end of the distribution was chosen as a minimum upper limit, given the potential impact of sorption kinetics, radium solution concentrations, and variation in surface areas. Zeolitic Tuff The experimentally derived sorption coefficients for barium and radium on zeolitic tuff are plotted against the calculated final barium or radium concentrations in Figure I-44. As before, the data points are separated into groups on the basis of when the experiments were carried out (pre-1990 = “old” and post-1990 = “new”), experiment duration, and whether the sorption coefficient was determined from a sorption or a desorption experiment. The plotted data indicate that there is no clear trend of barium Kd with solution concentration. Essentially all the sorption coefficients exceed a value of 10,000 mL/g and some desorption experiments approach values of 1,000,000 mL/g. The radium results are in the range of the results for barium, confirming the similar sorption behavior of these two elements. As with devitrified tuffs, the barium and radium sorption coefficients for zeolitic tuff plotted in Figure I-44 reflect solutions that were undersaturated with barium and radium sulfate. The sorption coefficients obtained with p#1 Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-57 November 2003 water are well within the range defined by the experiments with J-13 water. Thus, varying groundwater chemistry does not appear to have a clear impact. Ba and Ra on Zeolitic Tuff 1000 10000 100000 1000000 1.0E-14 1.0E-12 1.0E-10 1.0E-08 Calculated Final Conc. (mol/L) Ba Kd (mL/g) New Sorption J-13 <40 Days Old Sorption J-13 <40 Days Old Sorption J-13 >40 Days Old Desorption J-13 <40 Days Old Desorption J-13 >40 Days Radium J-13 >40 Days Old Sorption p#1 >40 Days DTNs: LA0309AM831341.002 [165523]; LA0305AM831341.001 [163789] Figure I-44. Barium and Radium Sorption Coefficients on Zeolitic Tuff versus Calculated Final Barium or Radium Concentrations in Solution Data plotted in Figure I-44 also indicate that there is no clear trend of barium Kd with solution concentration. Data were obtained for a sorption isotherm on sample YM-38 (Figure I-45). As indicated by the exponent in the equation on the diagram, the calculated isotherm is essentially linear. Ba on Zeolitic Tuff YM-38 in J-13 y = 23.23x0.9499 1.0E-09 1.0E-08 1.0E-07 1.0E-06 1.0E-05 1.0E-11 1.0E-10 1.0E-09 1.0E-08 1.0E-07 Calculated Final Ba Conc (mol/L) Ba on Solid (mol/g) DTNs: LA0309AM831341.002 [165523]; LA0305AM831341.001 [163789] Figure I-45. Isotherm Diagram for Ba Sorption on Zeolitic Tuff YM-38 in J-13 Water Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-58 November 2003 The effects of experiment duration on the barium Kd for zeolitic tuff are shown in Figure I-46. The barium and radium sorption reactions are quite fast, as indicated by the range of values obtained in 6-day experiments (Figure I-46) being similar to the range of values obtained in 84 day experiments. Ba and Ra on Zeolitic Tuff 1000 10000 100000 1000000 0 20 40 60 80 100 Experiment Duration (Days) Ba Kd (mL/g) Sorption J-13 New Sorption J-13 Old Desorption J-13 Old Sorption p#1 Old Desorption p#1 Old Radium J-13 Old DTNs: LA0309AM831341.002 [165523]; LA0305AM831341.001 [163789] Figure I-46. Barium and Radium Sorption Coefficient on Zeolitic Tuff in J-13 versus Experiment Duration for Sorption (Forward) and Desorption (Backward) Experiments The impact of variations in the major ion composition of ground waters is also shown in Figure I-46. Although there is a limited set of experiments with synthetic p#1 water, the results are largely within the range of the results obtained in experiments with J-13 water. Thus, there is no clear evidence of water-chemistry variations having an impact on barium/radium sorption coefficients in zeolitic tuff. The radium sorption-coefficient probability distribution derived for zeolitic tuff in the unsaturated zone is a uniform distribution with a range of 1,000–5,000 mL/g. The low end of the chosen range was selected based on the possibility that at concentrations near the solubility limit, sorption coefficients may be lower than the minimum value observed in long-term experiments (i.e., isotherms may be nonlinear) and that rock chemistry (Ca+Mg/Na+K in zeolites) may show greater variation than that represented among the samples used in the experiments. The upper end of the distribution was chosen as a minimum upper limit given the potential impacts of radium solution concentrations and variation in rock chemistry. Vitric Tuff The experimentally derived sorption coefficients for barium and radium on vitric tuff are plotted against the calculated final barium or radium concentrations in Figure I-47. As before, the data points are separated into groups on the basis of when the experiments were carried out (pre-1990 = “old” and post-1990 = “new”), experiment duration, and on whether the sorption coefficient was determined from a sorption or a desorption experiment. Although there are a limited number of datapoints, the data plotted in Figure I-47 show a general increase in sorption coefficient with decreasing concentration. This suggests that the isotherms for barium sorption onto vitric tuff are nonlinear. The impact of this nonlinearity must be included in the sorption coefficient probability distribution. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-59 November 2003 The radium sorption coefficient values obtained are larger than the barium coefficients obtained at similar solution concentrations. This indicates that barium sorption coefficients provide a conservative estimate of radium sorption coefficients on vitric tuff. As with devitrified tuffs, the barium and radium sorption coefficients for vitric tuff plotted in Figure I-47 reflect solutions that were undersaturated with barium and radium sulfate. The sorption coefficients obtained with p#1 water have values that are lower than those obtained in experiments with J-13 water. This indicates there is a groundwater composition effect for barium/radium sorption onto vitric tuff. Ba and Ra on Vitric Tuff 10 100 1000 10000 1.0E-12 1.0E-10 1.0E-08 1.0E-06 Calculated Final Ba Conc. (mol/L) Ba Kd (mL/g) Sorption New J-13 Sorption Old J-13 Desorption Old J-13 Radium Old J-13 Sorption Old p#1 Desorption Old p#1 DTNs: LA0309AM831341.002 [165523]; LA0305AM831341.001 [163789] Figure I-47. Barium and Radium Sorption Coefficients on Vitric Tuff versus Calculated Final Barium or Radium Concentrations in Solution The effects of experiment duration on the barium Kd for vitric tuff are shown in Figure I-48. The fact that the short-term experiments show higher sorption coefficient values than the longer-term experiments indicates that barium and radium sorption reactions on vitric tuff are fast. Ba and Ra on Vitric Tuff 10 100 1000 10000 100000 0 10 20 30 40 50 Experiment Duration (days) Ba Kd (mL/g) Sorption J-13 New Sorption J-13 Old Desorption J-13 Old Radium J-13 Old Sorption p#1 Old Desorption p#1 Old DTNs: LA0309AM831341.002 [165523]; LA0305AM831341.001 [163789] Figure I-48. Barium and Radium Sorption Coefficient on Vitric Tuff versus Experiment Duration for Sorption (Forward) and Desorption (Backward) Experiments Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-60 November 2003 The impact of variations in the major ion composition of groundwaters are also shown in Figure I-48. Although there is a limited set of experiments with synthetic p#1 water, the sorption coefficient values obtained are clearly below the range of the values obtained in experiments with J-13 water. Thus, there is evidence that water chemistry variations have an impact on barium sorption coefficients in devitrified tuff. The radium sorption-coefficient probability distribution derived for vitric tuff in the UZ is a uniform distribution with a range from 50–600 mL/g. The value of 50 mL/g was derived based on experiments with p#1 water at final solution concentrations near saturation with barite. The value of 600 mL/g was derived based on the potential impact of more dilute water chemistry (J- 13-like) combined with the potential for radium concentrations near the solubility limit. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-61 November 2003 I.8.G STRONTIUM The solubility of strontium in J-13 water at 25°C ranges from 2.7 × 10-5 mol/L at pH = 7.1 to 1.4 × 10-6 mol/L at pH = 8.5 and the solubility-controlling phase is strontianite (strontium carbonate). The solubility of strontianite in p#1 water at 25°C ranges from 2.4 × 10-5 mol/L at pH = 6.9 to 9.9 × 10-7 mol/L at pH = 8.6 (see DTN: LA0306AM831343.001 [164949]). Devitrified Tuff The experimentally derived sorption coefficients for strontium on devitrified tuff are plotted against the calculated final strontium concentrations of the experiments in Figure I-49. The data points are separated into groups on the basis of when the experiments were carried out (pre-1990 = “old” and post-1990 = “new”), water type, and whether the sorption coefficient was determined from a sorption or a desorption experiment. There is also data for a sorption isotherm on sample G1-2840. Sr on Devitrified Tuff 1 10 100 1000 10000 1.00E-12 1.00E-09 1.00E-06 1.00E-03 Calculated Final Sr Conc. (mol/L) Sr Kd (mL/g) Sorption J-13 New Sorption J-13 Old Desorption J-13 Old G1-2840 J-13 Old p#1 Old DTNs: LA0309AM831341.006 [165527]; LA0305AM831341.001 [163789] Figure I-49. Strontium Sorption Coefficients on Devitrified Tuff versus Calculated Final Strontium Concentration in Solution Most of the calculated final solution concentrations for the sorption experiments are nearly all below saturation with strontianite in J-13. The data points with concentrations greater than 1.0 × 10-6 mol/L reflect experiments that were oversaturated with strontianite. Thus, the sorption coefficients obtained in these five experiments will not be used in the derivation of the strontium sorption coefficient probability distribution. The calculated final strontium concentrations used in experiments with p#1 are all lower than the saturation value. Thus, oversaturation was not an issue in the strontium sorption experiments with p#1 water. A sorption isotherm was obtained for sample G1-2840 in J-13 water. The isotherm is linear at concentrations below approximately 5 × 10-5 mol/L as shown in Figure I-49. Further, the Kd value in the linear portion of the isotherm is at the low end of the range of strontium sorption coefficients obtained with J-13 and p#1 waters. Thus, the isotherms for the bulk of the samples would lie at higher Kd values compared to the isotherm for G1-2840. The sorption coefficients Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-62 November 2003 obtained in experiments with p#1 water fall in the middle of the range of values obtained for experiments with J-13 water. Thus, there is little or no impact of variations in water chemistry on strontium sorption coefficients on devitrified tuff. The effects of experiment duration on the strontium Kd for devitrified tuff are shown in Figure I- 50. The large range in sorption coefficients obtained at a given duration reflects variations in grain size of the crushed tuff samples used in the experiments, variations in solution strontium concentrations, variations in surface chemistry, and analytical error and artifacts. Experiments with crushed tuff samples that include the fines (e.g., particle size <30 µm) usually have sorption coefficients that are larger than samples from which the fines have been removed (e.g., particle sizes 75–500 µm). This is partly caused by the higher surface area of samples with fines and partly by mineral fractionation. Mineral fractionation can occur during the sieving process and cause the preferential concentration of very fine grained minerals (e.g., clays) in the fine fraction. Sr on Devitrified Tuff 1 10 100 1000 10000 0 20 40 60 80 100 Experiment Duration (days) Sr Kd (mL/g) Sorption J-13 New Sorption J-13 Old Desorption J-13 Old G1-2840 J-13 Old p#1 Old DTNs: LA0309AM831341.006 [165527]; LA0305AM831341.001 [163789] Figure I-50. Strontium Sorption Coefficients on Devitrified Tuff versus Experiment Duration for Sorption (Forward) and Desorption (Backward) Experiments. Data points include experiments in which fines were not removed from sample. Figure I-51 shows the effects of duration on sorption coefficients for crushed tuff samples with the fines removed. The total range of sorption coefficient values is reduced compared to Figure I-50. Further, it appears likely this range would converge to a range between 50 and 500 mL/g with increasing duration. The horizontal trend among data points at durations other than 21 and 42 days suggests strontium sorption reactions are relatively fast. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-63 November 2003 Sr on Devitrified Tuff 1 10 100 1000 0 20 40 60 80 100 Experiment Duration (days) Sr Kd (mL/g) Sorption J-13 New Sorption J-13 Old Desorption J-13 Old G1-2840 J-13 Old p#1 Old DTNs: LA0309AM831341.006 [165527]; LA0305AM831341.001 [163789] Figure I-51. Strontium Sorption Coefficients on Devitrified Tuff versus Experiment Duration for Sorption (Forward) and Desorption (Backward) Experiments. Fines fraction removed from all experiments. The strontium sorption coefficient probability distribution derived for devitrified tuff in the unsaturated zone is a uniform distribution with a range of 10–70 mL/g. The low end of the chosen range was selected based on the minimum value observed in isotherm experiments. Because the minimum value observed in isotherm experiments corresponds to solution concentrations close to saturation with a strontium carbonate, the effect of nonlinear isotherms is incorporated into the distribution. The upper end of the distribution was chosen as a minimum upper limit, given the potential impact of natural strontium in pore waters. Natural strontium concentrations measured in pore waters can be as high as 4.7 × 10-5 mol/L but the average is closer to 1.1 × 10-5 mol/L (DTN: GS020408312272.003 [160899]). The isotherm shown in Figure I-51 would indicate a sorption coefficient of 50–70 mL/g at this concentration. The high end of this range was chosen for the upper limit of the distribution. A uniform distribution was chosen to equally weight the sorption coefficient values in the selected range. Zeolitic Tuff The experimentally derived sorption coefficients for strontium on zeolitic tuff are plotted against the calculated final strontium concentrations of the experiments in Figure I-52. The data points are separated into groups on the basis of water type, whether or not the solid phase included fines, and whether the sorption coefficient was determined from a sorption or a desorption experiment. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-64 November 2003 Sr on Zeolitic Tuff 1000 10000 100000 1000000 1.0E-13 1.0E-10 1.0E-07 1.0E-04 Calculated Final Sr Conc. (M/l) Sr Kd (ml/g) Sorption-Old J-13 w / fines Sorption-Old J-13 no fines Desorption-Old J-13 w / fines Desorption-Old J-13 no fines Sorption-Old p#1 no fines Desorption-Old p#1 no fines DTNs: LA0309AM831341.006 [165527]; LA0305AM831341.001 [163789] Figure I-52. Strontium Sorption Coefficients on Zeolitic Tuff versus Calculated Final Strontium Concentration in Solution All except two of the calculated final solution concentrations were below the saturation level with strontianite. The calculated final strontium concentrations used in experiments with p#1 are all lower then the saturation value. Thus, oversaturation was not an issue in the strontium sorption experiments. The few sorption coefficients obtained with p#1 water tend to lie at the low end of the range of values shown in Figure I-52. This suggests variations in water composition will have an impact on the strontium sorption coefficient for zeolitic tuff. Sorption experiments were carried out at a number of different starting concentrations to obtain an isotherm for sample YM-38. As shown in Figure I-53, the sorption coefficients obtained in these experiments are not consistent with a simple relationship between strontium concentration and sorption coefficient. Nonetheless, the sorption coefficients plotted in Figure I-52 were obtained over a range of strontium concentrations. Thus, the concentration dependence of the sorption coefficient is included in the dataset. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-65 November 2003 Sr on Zeolitic Tuff YM-38 1000 10000 100000 1.0E-11 1.0E-09 1.0E-07 1.0E-05 1.0E-03 Calculated Final Sr Conc. (mol/L) Sr Kd (mL/g) Sorption-Old J-13 no fines DTNs: LA0309AM831341.006 [165527]; LA0305AM831341.001 [163789] Figure I-53. Strontium Sorption Coefficients versus Calculated Final Solution Concentration in M/L for Sample YM-38 in J-13 Water The effects of experiment duration on the strontium Kd for zeolitic tuff are shown in Figure I-54. The large range in sorption coefficients obtained at a given duration reflects variations in the grain size of the crushed tuff samples used in the experiments, variations in solution strontium concentrations, variations in surface chemistry, and analytical errors and artifacts. Experiments with crushed tuff samples that include the fines (e.g., particle size <30 µm) usually have sorption coefficients larger than samples from which the fines have been removed (e.g., particle sizes 75– 500 µm). This is partly a result of the higher surface area of samples with fines and partly due to mineral fractionation. Mineral fractionation can occur during the sieving process and cause the preferential concentration of very fine grained minerals (e.g., clays) in the fine fraction. Sr on Zeolitic Tuff 1000 10000 100000 1000000 0 20 40 60 80 100 Experiment Duration (days) Sr Kd (mL/g) Sorption-New J-13 no fines Sorption-Old J-13 w / fines Sorption-Old J-13 no fines Desorption-Old J-13 w / fines Desorption-Old J-13 no fines Sorption-Old p#1 no fines Desorption-Old p#1 no fines DTNs: LA0309AM831341.006 [165527]; LA0305AM831341.001 [163789] Figure I-54. Strontium Sorption Coefficients on Zeolitic Tuff versus Experiment Duration for Sorption (Forward) and Desorption (Backward) Experiments. Figure I-55 shows the effects of duration on sorption coefficients for crushed tuff samples with the fines removed. The total range of sorption coefficient values is reduced slightly compared to Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-66 November 2003 Figure I-54. The rather limited range of values at 84 days is likely caused by the limited number of experiments carried at this duration. There is no clear trend of sorption coefficient value with duration. This likely reflects fast sorption kinetics. Sr on Zeolitic Tuff 1000 10000 100000 1000000 0 20 40 60 80 100 Experiment Duration (days) Sr Kd (mL/g) Sorption-New J-13 no fines Sorption-Old J-13 no fines Desorption-Old J-13 no fines Sorption-Old p#1 no fines Desorption-Old p#1 no fines DTNs: LA0309AM831341.006 [165527]; LA0305AM831341.001 [163789] Figure I-55. Strontium Sorption Coefficients on Zeolitic Tuff with Fine Fraction Removed versus Experiment Duration for Sorption (Forward) and Desorption (Backward) Experiments The strontium sorption coefficient probability distribution derived for devitrified tuff in the UZ is a uniform distribution with a range of 50–2000 mL/g. The low end of the chosen range was selected based on the minimum value observed in isotherm experiments and the trend of lower sorption coefficients with increasing strontium concentration. When the trend is extended to the solubility limit for p#1 water, the value obtained is well below 1,000 mL/g. A value of 50 mL/g was chosen as a conservative value. The upper end of the distribution was chosen as a minimum upper limit, given the potential impact of natural strontium in pore waters. Natural strontium concentrations measured in pore waters can be as high as 4.7 × 10-5 mol/L but the average is closer to 1.1 × 10-5 mol/L (DTN: GS020408312272.003 [160899]). The data shown in Figure I- 52 would indicate a sorption coefficient of approximately 2,000 mL/g at this concentration. A uniform distribution was chosen to equally weight the sorption coefficient values in the selected range. Vitric Tuff The experimentally derived sorption coefficients for strontium on vitric tuff are plotted against the calculated final strontium concentrations of the experiments in Figure I-56. The data points are separated into groups on the basis of when the experiments were carried out (pre-1990 = “old” and post-1990 = “new”), water type, and whether the sorption coefficient was determined from a sorption or a desorption experiment. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-67 November 2003 Sr on Vitric Tuff 1 10 100 1000 1.E-10 1.E-09 1.E-08 1.E-07 1.E-06 1.E-05 Calculated Final Sr Conc. (mol/L) Sr Kd (mL/g) Sorption J-13 (New ) Sorption J-13 (Old) Desorption J-13 (Old) Sorption p#1 (Old) Desorption p#1 (Old) DTNs: LA0309AM831341.006 [165527]; LA0305AM831341.001 [163789] Figure I-56. Strontium Sorption Coefficients on Vitric Tuff versus Calculated Final Strontium Concentration in Solution The calculated final solution concentrations for the sorption experiments are all below saturation with strontianite. Thus, oversaturation was not an issue in the strontium sorption experiments. Sorption coefficients obtained in experiments with p#1 water fall in the middle of the range of values obtained for experiments with J-13 water. Thus, variations in water chemistry have little or no impact on strontium sorption coefficients in devitrified tuff. The effects of experiment duration on the strontium Kd for devitrified tuff are shown in Figure I- 57. Because the short-term experiments have higher sorption coefficient values than the longerterm experiments, strontium sorption kinetics must be relatively fast. Sr on Vitric Tuff 1 10 100 1000 0 20 40 60 80 100 Experiment Duration (days) Sr Kd (mL/g) Sorption-New J-13 no fines Sorption-Old J-13 no fines Desorption-Old J-13 no fines Sorption-Old p#1 no fines Desorption-Old p#1 no fines DTNs: LA0309AM831341.006 [165527]; LA0305AM831341.001 [163789] Figure I-57. Strontium Sorption Coefficients on Vitric Tuff versus Experiment Duration for Sorption (Forward) and Desorption (Backward) Experiments Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-68 November 2003 The strontium sorption coefficient probability distribution derived for vitric tuff in the unsaturated zone is a uniform distribution with a range of 0–50 mL/g. The low end of the chosen range was selected based on the minimum value observed in experiments with p#1 water and acknowledgment that the solution concentrations could get close to the saturation concentration (8 × 10-4 mol/L) with the addition of natural strontium to the radioactive fraction released from the repository (1.1–4.7 × 10-5 mol/L; DTN: GS020408312272.003 [160899]). The upper end of the distribution was chosen as a minimum upper limit in J-13 water with natural strontium. A uniform distribution was chosen to equally weight the sorption coefficient values in the selected range. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-69 November 2003 I.8.H THORIUM The solubility of thorium dioxide in waters such as the UZ waters is estimated at 3.2 × 10-9 mol/L at pH > 6.0 (Hummel et al. 2002 [161904], p. 377). Devitrified Tuff Experiments with Yucca Mountain tuffs were carried out with initial concentrations in the 1.0 × 10-7 to 6 × 10-8 mol/L range. Thus, the experiments were initially oversaturated with thorium dioxide. The calculated final thorium concentrations shown in Figure I-58 indicate that Th sorption onto the rock sample brought the final solution concentrations below saturation with thorium dioxide—in some cases but not all. The results of experiments oversaturated with thorium dioxide are of questionable value. For the remaining experiments, the sorption coefficients range from 1,213 to 23,800 mL/g. There are no data available for the effect of experimental duration on sorption coefficient values for Yucca Mountain samples. However, Allard et al. (1983 [162982], p. 10) reported that over experiment durations of 6 hours to 6 weeks, time had little influence on the measured sorption coefficients for Th on silica in 0.01 M NaClO4. Note that the starting concentrations reported by Allard et al. (1983 [162982], p. 6) were below the saturation level for thorium dioxide. Th Sorption Coefficients in J-13 100 1000 10000 100000 1.E-11 1.E-10 1.E-09 1.E-08 1.E-07 Calculated Final Th Conc. Th Kd (mL/g) Devitrified Tuff Zeolitic Tuff Vitric Tuff DTN: LA0305AM831341.001 [163789] Figure I-58. Thorium Sorption Coefficients on Tuff versus Calculated Final Thorium Concentration in Solution There are no data available to evaluate the impact of variations in water chemistry on Th sorption coefficients. However, thorium forms primarily hydroxide complexes at near neutral pH (Langmuir and Herman 1980 [147527], p. 1753). Therefore, water chemistry is expected to have very little influence on Th sorption coefficient values in Yucca Mountain groundwaters. However, water chemistry (i.e., pH) does impact the solubility of thorium dioxide. This is the reason some of the experiments in Figure I-58 were oversaturated with thorium dioxide. As shown in Figure I-59, the lowest sorption coefficients were obtained at near-neutral pH values where J-13 water was oversaturated with thorium dioxide. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-70 November 2003 Th Sorption Coefficients in J-13 100 1000 10000 100000 4 5 6 7 8 9 pH Th Kd (mL/g) Devitrified Tuff Zeolitic Tuff Vitric Tuff DTN: LA0305AM831341.001 [163789] Figure I-59. Thorium Sorption Coefficients on Tuff Versus pH On the basis of the experimental data in Figures I-58 and I-59, the range of thorium sorption coefficients expected for devitrified tuffs in the saturated volcanic section at Yucca Mountain is 1,000-10,000 mL/g. This range is intended to reflect the range in surface areas found in devitrified tuffs in the saturated zone and the range in thorium concentrations expected during saturated zone transport. The lower end of the range reflects sorption coefficients at thorium concentrations near the solubility limit. Given the sparseness of the available data, the probability distribution type selected is a uniform distribution. Zeolitic Tuff Sorption coefficient data for zeolitic tuff are plotted in Figures I-58 and I-59. Based on the available data, zeolitic tuffs have sorption coefficients for thorium that are similar to those obtained for devitrified tuffs. On the basis of the data plotted in Figures I-58 and I-59, the range of thorium sorption coefficients selected for zeolitic tuffs in the saturated volcanic section at Yucca Mountain is 1,000–30,000 mL/g. The upper end of this range was selected to reflect the higher surface areas of zeolitic tuffs relative to devitrified tuffs. A uniform distribution is selected for zeolitic tuffs. Vitric Tuff Sorption coefficient data for vitric tuff are plotted in Figures I-58 and I-59. Based on the available data, vitric tuffs have sorption coefficients for thorium in the range of those obtained for the other tuffs in those experiments that were undersaturated in thorium dioxide. On the basis of the data plotted in Figures I-58 and I-59, the range of thorium sorption coefficients selected for vitric tuffs in the saturated volcanic section at Yucca Mountain is 1,000-10,000 mL/g. A uniform distribution is selected for vitric tuffs. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-71 November 2003 I.8.I URANIUM The solubility of uranium in J-13 water under oxidizing conditions ranges from 1.8 × 10-4 at pH = 7.1 to 2.0 × 10-4 mol/L at pH = 8.5 (DTN: LA0306AM831343.001 [164949]). The solubility of uranium in synthetic p#1 water under oxidizing conditions ranges from 6.7 × 10-4 at pH = 6.9 to 9.0 × 10-4 mol/L at pH = 8.6 (DTN: LA0306AM831343.001 [164949]). The solubility-controlling solid in both waters is schoepite. Devitrified Tuff As shown in Figure I-60, the calculated final uranium concentrations in the sorption experiments were generally below saturation with schoepite. The sorption coefficients obtained in sorption experiments with devitrified tuffs do not show a correlation with the calculated final uranium solution concentrations. U on Devitrified Tuff -10 -5 0 5 10 15 20 1.E-08 1.E-06 1.E-04 Calculated Final U Conc. U Kd (mL/g) Sorption Old J-13 Sorption New J-13 Sorption New p#1 Desorption Old J-13 DTNs: LA0305AM831341.001 [163789]; LA0309AM831341.007 [165528] Figure I-60. Uranium Sorption Coefficients on Devitrified Tuff Versus Calculated Final Uranium Concentration in Solution Sorption experiments carried out as a function of time are shown in Figure I-61. Beyond approximately 3 days, there is no clear correlation between the sorption coefficients obtained and the duration of the experiments. The data imply that uranium sorption reactions on devitrified tuffs must be relatively fast (i.e., they reach steady state in a few days). Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-72 November 2003 U on Devitrified Tuff -10 -5 0 5 10 15 20 0 20 40 60 Experiment Duration (days) U Kd (mL/g) Sorption Old J-13 Sorption New J-13 Sorption New p#1 Desorption Old J-13 DTNs: LA0305AM831341.001 [163789]; LA0309AM831341.007 [165528] Figure I-61. Uranium Sorption Coefficients on Devitrified Tuff Versus Experiment Duration for Sorption (Forward) and Desorption (Backward) Experiments The dependence of the uranium sorption coefficient on water chemistry was tested with experiments using two water compositions (J-13 and synthetic p#1). The J-13 experimental data are shown as a function of pH in Figure I-62. The “old” data were obtained in the 1980s and the “new” data were obtained in the 1990s. The difference between them is not statistically significant. The range of values obtained at a given pH (e.g., 8.4) reflects experimental errors and natural variations in rock properties (e.g., surface area and mineral chemistry). It is not possible to discriminate between these possible causes with the available data. On the basis of the experimental data points, there does not appear to be a correlation between Kd and pH. However, surface complexation modeling with PHREEQC, using binding constants derived by Pabalan et al. (1998 [162987], p. 124) for uranium on silica, points to a clear pH dependence as shown in Figure I-62. The two model curves reflect two different surface areas (2.8 and 5.6 m2/g). The 2.8 m2/g surface area is approximately an average value for devitrified tuffs at Yucca Mountain. The sorption coefficients obtained in experiments with “synthetic p#1” water are shown in Figure I-63. The data plotted have substantial experimental errors associated with them, as indicated by the magnitude of some of the negative Kd values. These experimental errors result from counting statistics, the stability of counters over time, corrections made for adsorption to container walls, and the pH of the tracer solution added to the experiment. Taken at face value, the experimental data suggest a trend of increasing Kd with increasing pH. However, the surface complexation modeling predicts a decrease in Kd with increasing pH, although the absolute Kd values are rather small. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-73 November 2003 Uranium on Devit Tuff in J-13 -4 -2 0 2 4 6 8 10 6 7 8 9 pH Uranium Kd (mL/g) Model 5.6 m2/g Model 2.8 m2/g Old Data New Data DTNs: LA0305AM831341.001 [163789]; LA0309AM831341.007 [165528]; LA0306AM831343.001 [164949] Figure I-62. Uranium Sorption Coefficients on Devitrified Tuff versus pH. Model curves are from the PHREEQC Surface Complexation Model. Uranium on Devit Tuff in p#1 -6 -4 -2 0 2 4 6 6 7 8 9 10 pH Uranium Kd (mL/g) YMP Data Model 2.8 m2/g DTNs: LA0305AM831341.001 [163789]; LA0309AM831341.007 [165528]; LA0306AM831343.001 [164949] Figure I-63. Uranium Sorption Coefficients on Devitrified Tuff in p#1 Water versus pH. Model curve is from the PHREEQC Surface Complexation Model. On the basis of the experimental data and model curves plotted in Figures I-62 and I-63, a cumulative probability distribution type was selected for the sorption coefficient probability distribution for devitrified tuff, with a value of 0.0 mL/g at 0.0, 0.2 mL/g at 0.5, and 4.0 mL/g at 1.0. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-74 November 2003 Zeolitic Tuff As shown in Figure I-64, the sorption coefficients obtained in sorption experiments with zeolitic tuffs do not show a clear correlation with the calculated final uranium solution concentrations. The high end of the concentrations plotted is below saturation with a solid uranium phase. U Kd on Zeolitic Tuff -10 0 10 20 30 40 1.0E-07 1.0E-06 1.0E-05 1.0E-04 1.0E-03 Calculated Final U Concentration (mol/L) U Kd (mL/g) J-13 "New" J-13 "Old" Synthetic p#1 DTNs: LA0305AM831341.001 [163789]; LA0309AM831341.007 [165528] Figure I-64. Uranium Sorption Coefficients on Zeolitic Tuff versus Calculated Final Uranium Concentration in Solution Uranium sorption experiments on zeolitic tuffs carried out as a function of time are shown in Figure I-65. Beyond a period of approximately 3 days, there is no clear correlation between the sorption coefficients obtained and the duration of the experiments. These data imply that uranium sorption reactions on zeolitic tuffs are relatively fast (i.e., they reach steady state in a few days). Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-75 November 2003 Uranium on Zeolitic Tuff -10 0 10 20 30 40 0 20 40 Experiment Duration (days) U Kd (mL/g) J-13 "Old" J-13 "New" Synthetic p#1 DTNs: LA0305AM831341.001 [163789]; LA0309AM831341.007 [165528] Figure I-65. Uranium Sorption Coefficients on Zeolitic Tuff as a Function of Experiment Duration The dependence of the uranium sorption coefficient on water chemistry was tested with experiments using two water compositions (J-13 and synthetic p#1). The J-13 data are shown as a function of pH in Figure I-66. The “old” data were obtained in the 1980s and the “new” data were obtained in the 1990s. The difference between them is not statistically significant. The range of values observed at a given pH (e.g., 8.4) reflects variations in rock properties and experimental errors. The experimental errors result from such things as counting statistics, the stability of counters over time, the accuracy of corrections for adsorption to container walls, and other experimental artifacts. In some cases, the pH of the tracer solution added to the experiment seems to have an effect. Some of these errors are random (e.g., counting errors) and others (e.g., adsorption to container walls) may have a nonrandom bias. It is not possible to separately evaluate these errors with the information available. Note that the distribution of data points in Figure I-66 does not indicate a strong correlation between Kd and pH. Surface complexation modeling was carried out with PHREEQC to provide a framework in which to interpret the experimental data. Binding constants for uranium on silica derived by Pabalan et al. (1998 [162987], p. 124) were used in the modeling. The modeling results show a clear pH dependence (Figure I-66). The two model curves reflect two different surface areas. A surface area of 28 m2/g was used because it approximates an average value for zeolitic tuffs and because it is an order of magnitude larger than the average value used for modeling devitrified tuffs. A surface area of 14 m2/g was also used to show the impact of a factor of 2 change in surface area. The sorption coefficients obtained in experiments with “synthetic p#1” water are shown in Figure I-67. The magnitudes of the negative Kd values plotted are similar to the magnitudes of the positive values plotted. Thus, the net values may be very close to zero. Taken at face value, the experimental data suggest a trend of increasing Kd with increasing pH. However, the surface Experiment Duration Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-76 November 2003 complexation modeling predicts an increase in Kd with decreasing pH, although the absolute Kd values are rather small, in agreement with the net values obtained from the experimental data. Uranium on Zeolitic Tuff in J-13 -25 0 25 50 6 7 8 9 pH Uranium Kd (mL/g) Model (28 m2/g) Model (14 m2/g) YMP "Old Data" YMP "New Data" DTNs: LA0305AM831341.001 [163789]; LA0309AM831341.007 [165528]; LA0306AM831343.001 [164949] Figure I-66. Uranium Sorption Coefficients for Zeolitic Tuff in J-13 Plotted as a Function of pH. Model curves derived with PHREEQC surface complexation modeling are also shown. Uranium on Zeolitic Tuff in Synthetic p#1 -8 -6 -4 -2 0 2 4 6 8 6.000 7.000 8.000 9.000 10.000 pH Uranium Kd (mL/g) Model (14 m2/g) YMP Data DTNs: LA0305AM831341.001 [163789]; LA0309AM831341.007 [165528]; LA0306AM831343.001 [164949] Figure I-67. Uranium Sorption Coefficients for Zeolitic Tuff in Synthetic p#1 Plotted as a Function of pH. Model curves derived with PHREEQC surface complexation modeling are also shown Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-77 November 2003 On the basis of the experimental data and model curves plotted in Figures I-66 and I-67, a cumulative probability distribution type was selected for the sorption coefficient probability distribution for zeolitic tuff with a value of 0.0 mL/g at 0.0, 0.2 mL/g at 0.5 and 30.0 mL/g at 1.0. Uranium–Vitric Tuff As shown in Figure I-68, the high end of the concentration range plotted is below saturation with schoepite. Thus, oversaturation was not a problem in sorption experiments with vitric tuffs. Uranium on Vitric Tuffs -5 0 5 10 15 1.E-09 1.E-07 1.E-05 1.E-03 Final Uranium Conc. (mol/L) Uranium Kd (mL/g) Sorption New J-13 Sorption New s-p#1 Sorption Old J-13 Desorption Old J-13 DTNs: LA0305AM831341.001 [163789]; LA0309AM831341.007 [165528] Figure I-68. Uranium Sorption Coefficients on Vitric Tuff versus Calculated Final Uranium Concentration in Solution Uranium sorption experiments on vitric tuffs carried out as a function of time are shown in Figure I-69. Beyond a period of approximately 3 days, there is no clear correlation between the sorption coefficients obtained and the duration of the experiments. These data imply that uranium sorption reactions on vitric tuffs are relatively fast (i.e., they reach steady state in a few days). Sorption coefficients obtained from desorption experiments are larger, on average, than sorption coefficients obtained from sorption experiments. The desorption experiments were performed on a sample that contained 15% zeolite. Perhaps the presence of zeolite in this sample caused it to have a higher affinity for uranium. The point with the largest sorption coefficient value from sorption experiments (12 mL/g) was also obtained on this particular sample. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-78 November 2003 Uranium on Vitric Tuff -5 0 5 10 15 0 10 20 30 40 50 Experiment Duration (days) Uranium Kd (mL/g) Sorption New J-13 Sorption Old J-13 Desorption Old J-13 Sorption New s-p#1 DTNs: LA0305AM831341.001 [163789]; LA0309AM831341.007 [165528] Figure I-69. Uranium Sorption Coefficients on Vitric Tuff as a Function of Experiment Duration The dependence of the uranium sorption coefficient on water chemistry was tested with experiments using two water compositions (J-13 and synthetic p#1). The J-13 data are shown as a function of pH in Figure I-70. The coefficients obtained in sorption experiments are smaller than the coefficients obtained in desorption experiments, as noted above. As before, the range of values observed at a given pH (e.g., 8.4) reflects variations in rock properties and experimental errors. The experimental errors result from such things as counting statistics, the stability of counters over time, the accuracy of corrections for adsorption to container walls, and other experimental artifacts. In some cases, the pH of the tracer solution added to the experiment seems to have an effect. Some of these errors are random (e.g., counting errors), and others (e.g., adsorption to container walls) may have a nonrandom bias. It is not possible to separately evaluate these errors with the information available. However, the magnitude of the negative values suggest error values of up to 100%. A model using 1.5 m2/g surface area to represent the average vitric tuff is shown on the Figure I-70. This model predicts sorption coefficient values in the pH range 8.5 to 7.0 between 0.2 and 1.5 mL/g, respectively. Uranium sorption coefficients obtained with synthetic p#1 water are plotted in Figure I-71. The average of the points plotted is close to zero. The results of a model calculation using a 1.5 m2/g surface area and the p#1 water composition as inputs are shown in Figure I-71. This model predicts sorption coefficient values of 0.004 and 0.12 mL/g for pH values of 8.5 and 7.0, respectively. Thus, the predicted uranium sorption coefficient for vitric tuff in p#1-v water is close to zero, consistent with the experimental data. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-79 November 2003 Uranium on Vitric Tuff in J-13 -5 0 5 10 15 6 7 8 9 10 pH Uranium Kd (mL/g) Sorption Desorption Model J-13 DTNs: LA0305AM831341.001 [163789]; LA0309AM831341.007 [165528] Figure I-70. Uranium Sorption Coefficients for Vitric Tuff in J-13 Plotted as a Function of pH. Model curves derived with PHREEQC surface complexation modeling are also shown. Uranium on Vitric Tuff in Synthetic p#1 -5 0 5 10 6 7 8 9 10 pH Uranium Kd (mL/g) Sorption New s-p#1 Model p#1-v DTNs: LA0305AM831341.001 [163789]; LA0309AM831341.007 [165528] Figure I-71. Uranium Sorption Coefficients for Vitric Tuff in Synthetic p#1 Plotted as a Function of pH. Model curves derived with PHREEQC surface complexation modeling are also shown. On the basis of the experimental data and model curves plotted in Figures I-70 and I-71, a cumulative probability distribution type was selected for the sorption coefficient probability distribution for vitric tuff, with a value of 0.0 mL/g at 0.0, 0.2 mL/g at 0.5 and 3.0 mL/g at 1.0. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment I-80 November 2003 I.9. SUMMARY Sorption-coefficient probability distribution functions were derived for radionuclides of interest to be used in transport calculations. Experimental and modeling results were used to constrain the distributions. In general, the approach used in derivation of the distributions tended to underestimate the range and median or expected value. This was done to provide some conservatism in the derivation, given potential scaling uncertainties in the application of these distributions to transport calculations at the Yucca Mountain site. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment II-1 November 2003 ATTACHMENT II CORRELATIONS FOR SAMPLING OF SORPTION COEFFICIENT PROBABILITY DISTRIBUTIONS Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment II-2 November 2003 INTENTIONALLY LEFT BLANK Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment II-3 November 2003 TABLES Page II-1. Ratings of Controls on Sorption Behavior..............................................................................6 II-2. Correlations For Sampling Sorption Coefficient Probability Distributions ...........................7 Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment II-4 November 2003 INTENTIONALLY LEFT BLANK Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment II-5 November 2003 In the TSPA, UZ transport calculations are carried out separately for each radionuclide. The sorption-coefficient probability distribution for each radionuclide could be sampled independently in each transport calculation. However, such independent sampling could potentially lead to dose dilution. That is, independent sampling of the distributions could cause radionuclides to travel at independent rates, such that the calculated dose at the accessible environment would not be representative of maximum possible doses. Similarities in the chemical dependencies of sorption coefficients for the various radionuclides suggest transport rates in the UZ are likely to be correlated for some radionuclides. Correlations for sampling sorption coefficient probability distributions have been derived for the elements Am, Cs, Np, Pa, Pu, Ra, Sr, Th, and U. To derive the correlations, a rating system was first developed to rate the impact of six different variables on the sorption coefficient for a given element in each of the three major rocks types. The matrix containing the ratings is shown in Table II-2. The six variables are pH, Eh, water chemistry, rock composition, rock surface area, and radionuclide concentration. Most of these parameters are self explanatory. Water chemistry refers to the major ion concentrations and silica. Rock composition refers to both the mineralogic composition of the rocks and the chemical composition of the minerals (e.g., zeolite compositions). The ratings are based on the sorption data and modeling presented in Section I.5 of Attachment I, combined with professional judgement. The rating system presented in Table II-1 was used to develop correlations between the sorption coefficient probability distributions for the elements of interest. Identical parameter ratings resulted in a correlation of 100%. If the two highest rating were in the same parameters in the same order, a correlation of 75% was assigned. If the two highest ratings were in the same parameters but not in the same order, a correlation of 50% was assigned. If the three highest ratings were in the same parameters but not in the same order, a correlation of 25% was assigned. If the four highest ratings were in the same parameters but not in the same order, a correlation of 10% was assigned. If the three highest ratings were not in the same parameters, a correlation of 0% was assigned. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment II-6 November 2003 Table II-1. Ratings of Controls on Sorption Behavior Element pH Eh Water Chemistry Rock Comp. Surface Area RN Conc. Am (Devit.) 2 No 3 4 1 N/D Am (Zeol.) 2 No 3 4 1 N/D Am (Vitric) 2 No 3 4 1 N/D Cs (Devit.) No No 3 4 2 1 Cs (Zeol.) No No 2 1 3 4 Cs (Vitric) No No 3 4 2 1 Np (Devit.) 3 No 1 4 2 5 Np (Zeol.) 3 No 1 4 2 5 Np (Vitric) 3 No 1 4 2 5 Pa (Devit.) 2 No N/D N/D 1 N/D Pa (Zeol.) 2 No N/D N/D 1 N/D Pa (Vitric) 2 No N/D N/D 1 N/D Pu (Devit.) 4 (No) 2 1 3 5 Pu (Zeol.) 4 (No) 2 1 3 5 Pu (Vitric) 4 (No) 2 1 3 5 Ra (Devit.) No No 3 4 2 1 Ra (Zeol.) No No 1 2 3 4 Ra (Vitric) No No 3 4 2 1 Sr (Devit.) No No 2 4 1 3 Sr (Zeol.) No No 1 2 3 4 Sr (Vitric) No No 2 4 1 3 Th (Devit.) No No 2 N/D 1 N/D Th (Zeol.) No No 2 N/D 1 N/D Th (Devit.) No No 2 N/D 1 N/D U (Devit.) 2 (No) 1 No 4 3 U (Zeol.) 2 (No) 1 No 4 3 U (Vitric) 2 (No) 1 No 4 3 DTN: LA0311AM831341.001 [Output] NOTE: 1 = High impact, 5 = Low impact. (No) = No impact under oxidizing conditions. Note, however, that rock composition may provide locally reducing environments for plutonium to sorb as plutonium IV. For example, ferrous ironbearing minerals may provide such local microenvironments. N/D = No data available. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment II-7 November 2003 Table II-2. Correlations For Sampling Sorption Coefficient Probability Distributions Element Am (Devit.) Am (Zeol.) Am (Vitric) Cs (Devit.) Cs (Zeol.) Cs (Vitric) Np (Devit.) Np (Zeol.) Np (Vitric) Pa (Devit.) Pa (Zeol.) Pa (Vitric) Pu (Devit.) Pu (Zeol.) Pu (Vitric) Ra (Devit.) Ra (Zeol.) Ra (Vitric) Sr (Devit.) Sr (Zeol.) Sr (Vitric) Th (Devit.) Th (Zeol.) Th (Vitric) U (Devit.) U (Zeol.) U (Vitric) Am (Devit.) 100 Am (Zeol.) 100 100 Am (Vitric) 100 100 100 Cs (Devit.) 0 0 0 100 Cs (Zeol.) 0 0 0 10 100 Cs (Vitric) 0 0 0 100 10 100 Np (Devit.) 25 25 25 0 0 0 100 Np (Zeol.) 25 25 25 0 0 0 100 100 Np (Vitric) 25 25 25 0 0 0 100 100 100 Pa (Devit.) 75 75 75 0 0 0 0 0 0 100 Pa (Zeol.) 75 75 75 0 0 0 0 0 0 100 100 Pa (Vitric) 75 75 75 0 0 0 0 0 0 100 100 100 Pu (Devit.) 10 10 10 0 75 0 10 10 10 0 0 0 100 Pu (Zeol.) 10 10 10 0 75 0 10 10 10 0 0 0 100 100 Pu (Vitric) 10 10 10 0 75 0 10 10 10 0 0 0 100 100 100 Ra (Devit.) 0 0 0 100 10 100 0 0 0 0 0 0 0 0 0 100 Ra (Zeol.) 0 0 0 10 50 10 0 0 0 0 0 0 50 50 50 10 100 Ra (Vitric) 0 0 0 100 10 100 0 0 0 0 0 0 0 0 0 100 10 100 Sr (Devit.) 0 0 0 25 10 25 50 50 50 0 0 0 0 0 0 25 10 25 100 Sr (Zeol.) 0 0 0 10 50 10 0 0 0 0 0 0 50 50 50 10 100 10 10 100 Sr (Vitric) 0 0 0 25 10 25 50 50 50 0 0 0 0 0 0 25 10 25 100 10 100 Th (Devit.) 0 0 0 0 0 0 50 50 50 0 0 0 0 0 0 0 0 0 75 0 75 100 Th (Zeol.) 0 0 0 0 0 0 50 50 50 0 0 0 0 0 0 0 0 0 75 0 75 100 100 Th (Vitric) 0 0 0 0 0 0 50 50 50 0 0 0 0 0 0 0 0 0 75 0 75 100 100 100 U (Devit.) 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 100 U (Zeol.) 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 100 100 U (Vitric) 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 100 100 100 DTN: LA0311AM831341.001 [Output] Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01J Attachment II-8 November 2003 INTENTIONALLY LEFT BLANK Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment II-9 November 2003 II.1 CORRELATION OF SAMPLING OF SORPTION-COEFFICIENT PROBABILITY DISTRIBUTIONS BETWEEN THE UNSATURATED ZONE AND SATURATED ZONE The sampling of sorption coefficient probability distributions for the UZ should not be correlated with sampling of sorption coefficient probability distributions for the saturated zone (SZ) except possibly for the elements radium and cesium in devitrified tuffs. This conclusion is based on an analysis of variations in the parameters that have the greatest impact on the sorption behavior of the radionuclides of interest (i.e., pH, Eh, water chemistry, rock composition, radionuclide concentration). This analysis is discussed in the following. • Parameter: pH Beneath the repository footprint, variation in pH in SZ groundwater is limited (7.0–7.2). This is likely to be the case in any climate scenario, because the primary control on pH is elevated 2 CO P (log = -2.0) maintained from below (i.e., Paleozoic aquifer). Outside the repository footprint, pH shows somewhat greater variation (7.0–7.5). In the UZ, the pH of pore water is more variable, particularly in zeolitic units. There is no evidence that the pH of SZ groundwaters are correlated with the pH of UZ pore waters or perched waters. Therefore, there is no reason to correlate the sorption-coefficient probability distributions on the basis of pH. • Parameter: Eh In the UZ, Eh will always be oxidizing. In the SZ, Eh can be reducing or oxidizing. Therefore, there is no reason to correlate the sorption-coefficient probability distributions on the basis of Eh. • Parameter: Water Chemistry (i.e., ionic strength, alkalinity) SZ appears to have groundwater chemistry of glacial times. UZ pore water has chemistry of an arid climate. Further, UZ water chemistry appears to be layered, with variations mainly in the vertical direction. There is no evidence that UZ pore-water chemistry is areally correlated with SZ groundwater chemistry. Therefore, there is no reason to correlate the sorption-coefficient probability distributions on the basis of water chemistry. • Parameter: Rock Composition Rock composition (e.g., zeolite compositions) is known to be variable horizontally and vertically in both the UZ and SZ. There is no evidence that rock compositions in the UZ are areally correlated with rock compositions in the SZ within the repository footprint. In fact, rock compositions tend to be layered, with variations primarily in the vertical direction. Therefore, there is no reason to correlate the sorption coefficient probability distributions on the basis of rock composition. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment II-10 November 2003 • Parameter: Surface Area Rock surface area (see Table I-1 of Attachment I, this Model Report) will not change significantly over the regulatory time frame. There is no evidence that rock surface area in the UZ is correlated areally with rock surface area in the SZ. This also reflects the horizontal layering of rock compositions. Therefore, there is no reason to correlate the sorption coefficient probability distributions on the basis of rock surface area. • Parameter: Radionuclide Concentration If the concentration of a radionuclide in the UZ is on the high end of the range, the concentration of the radionuclide in the SZ will also be on the high end of the range. Thus, there is a reason to correlate the sorption-coefficient probability distributions. However, radionuclide concentrations are the dominant control on radionuclide sorption behavior only for radium and cesium on devitrified and vitric tuffs (Table II-1). Thus, it would seem appropriate to correlate the sampling of sorption-coefficient distributions for these two elements in these two rock types. However, vitric tuffs are not a component of the saturated zone. Thus, only sorption-coefficient probability distributions for cesium and radium on devitrified tuff in the UZ and the saturated zone might need to be correlated. Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment III-1 November 2003 ATTACHMENT III SUPPORTING GEOLOGICAL DATA FOR THE CONCEPTUAL MODEL Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment III-2 November 2003 INTENTIONALLY LEFT BLANK Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment III-3 November 2003 TABLES Page III-1. Thickness and Elevation of Hydrogeologic Units in the Vertical Column on the Location of USW SD-6 Borehole............................................................................... III-4 III-2. Thickness and Elevation of Hydrogeologic Units in the Vertical Column on the Location of USW SD-12 Borehole............................................................................. III-5 III-3. Thickness and Elevation of Hydrogeologic Units in the Vertical Column on the Location of USW UZ-14 Borehole............................................................................. III-6 Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment III-4 November 2003 Table III-1. Thickness and Elevation of Hydrogeologic Units in the Vertical Column on the Location of USW SD-6 Borehole. Hydrogeologic unit Thickness (m) Elevation (m) tswF6 30.70 1090.00 tswF7 15.34 1059.30 tswF8 14.33 1043.96 tswFv 5.19 1029.63 ch1Fv 12.50 1024.44 ch2Fv 7.85 1011.94 ch3Fv 7.85 1004.09 ch4Fv 7.85 996.23 ch5Fv 7.85 988.38 ch6Fv 15.54 980.53 pp4Fz 7.59 964.98 pp3Fd 40.25 957.39 pp2Fd 11.00 917.14 pp1Fz 65.34 906.14 bf3Fd 96.57 840.80 Source: DTN LB03013DSSCP3I.001 [162379] Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment III-5 November 2003 Table III-2. Thickness and Elevation of Hydrogeologic Units in the Vertical Column on the Location of USW SD-12 Borehole. Hydrogeologic unit Thickness (m) Elevation (m) tswF5 83.70 1075.30 tswF6 43.15 991.60 tswF7 21.57 948.45 pcF38 9.10 926.88 pcF39 9.01 917.78 pcF1z 4.50 908.77 ch1Fv 18.02 904.26 ch2Fv 14.33 886.25 ch3Fv 14.33 871.91 ch4Fz 14.33 857.58 ch5Fv 14.33 843.24 ch6Fz 14.96 828.91 pp4Fz 8.65 813.95 pp3Fd 33.54 805.30 pp2Fd 23.76 771.75 bf3Fd 96.57 840.80 Source: DTN LB03013DSSCP3I.001 [162379] Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment III-6 November 2003 Table III-3. Thickness and Elevation of Hydrogeologic Units in the Vertical Column on the Location of USW UZ-14 Borehole. Hydrogeologic unit Thickness (m) Elevation (m) tswF6 29.10 1088.40 tswF7 15.34 1059.30 tswF8 14.33 1043.96 tswFv 5.19 1029.63 ch1Fv 12.50 1024.44 ch2Fv 7.85 1011.94 ch3Fv 7.85 1004.09 ch4Fv 7.85 996.23 ch5Fv 7.85 988.38 ch6Fv 15.54 980.53 pp4Fz 7.59 964.98 pp3Fd 40.25 957.39 pp2Fd 11.00 917.14 pp1Fz 65.34 906.14 bf3Fd 96.57 840.80 Source: DTN LB03013DSSCP3I.001 [162379] Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment IV-1 November 2003 ATTACHMENT IV FIGURES FROM THE 237Np 3-D TRANSPORT STUDIES (INSTANTANEOUS RELEASE, MEAN PRESENT-DAY INFILTRATION) OUTPUT-DTN: LB0307MR0060R1.002 Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment IV-2 November 2003 INTENTIONALLY LEFT BLANK Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment IV-3 November 2003 FIGURES Page IV.1. Distribution of the Relative Mass Fraction XR of 237Np in the Fractures of the tsw39 Layer at t = 10 Years for Mean Present-Day Infiltration and Instantaneous Release ..............................................................................................................................5 IV.2. Distribution of the Relative Mass Fraction XR of 237Np in the Matrix of the tsw39 layer at t = 10 Years for Mean Present-Day Infiltration and Instantaneous Release ..............................................................................................................................5 IV.3. Distribution of the Relative Mass Fraction XR of 237Np in the Fractures of the tsw39 Layer at t = 100 years for Mean Present-Day Infiltration and Instantaneous Release ..............................................................................................................................6 IV.4. Distribution of the Relative Mass Fraction XR of 237Np in the Matrix of the tsw39 Layer at t = 100 years for Mean Present-Day Infiltration and Instantaneous Release ..............................................................................................................................6 IV.5. Distribution of the Relative Mass Fraction XR of 237Np in the Fractures of the tsw39 layer at t = 1,000 years for Mean Present-Day Infiltration and Instantaneous Release .......................................................................................................7 IV.6. Distribution of the Relative Mass Fraction XR of 237Np in the Matrix of the tsw39 Layer at t = 1,000 Years for Mean Present-Day Infiltration and Instantaneous Release .......................................................................................................7 IV.7. Distribution of the Relative Mass Fraction XR of 237Np in the Fractures of the tsw39 Layer at t = 10,000 Years for Mean Present-Day Infiltration and Instantaneous Release .......................................................................................................8 IV.8. Distribution of the Relative Mass Fraction XR of 237Np in the Matrix of the tsw39 Layer at t = 10,000 Years for Mean Present-Day Infiltration and Instantaneous Release .......................................................................................................8 IV.9. Distribution of the Relative Mass Fraction XR of 237Np in the Fractures of the tsw39 Layer at t = 100,000 Years for Mean Present-Day Infiltration and Instantaneous Release .......................................................................................................9 IV.10. Distribution of the Relative Mass Fraction XR of 237Np in the Matrix of the tsw39 Layer at t = 100,000 Years for Mean Present-Day Infiltration and Instantaneous Release .......................................................................................................9 IV.11. Distribution of the Relative Mass Fraction XR of 237Np in the Fractures of the tsw39 Layer at t = 1,000,000 Years for Mean Present-Day Infiltration and Instantaneous Release .....................................................................................................10 IV.12. Distribution of the Relative Mass Fraction XR of 237Np in the Matrix of the tsw39 Layer at t = 1,000,000 Years for Mean Present-Day Infiltration and Instantaneous Release .....................................................................................................10 IV.13. Distribution of the Relative Mass Fraction XR of 237Np in the Fractures Immediately Above the Groundwater Table at t = 10 Years for Mean Present- Day Infiltration and Instantaneous Release ....................................................................11 IV.14. Distribution of the Relative Mass Fraction XR of 237Np in the Matrix Immediately Above the Groundwater at t = 10 Years for Mean Present-Day Infiltration and Instantaneous Release ............................................................................11 Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment IV-4 November 2003 FIGURES (Continued) Page IV.15. Distribution of the Relative Mass Fraction XR of 237Np in the Fractures Immediately Above the Groundwater at t = 100 Years for Mean Present-Day Infiltration and Instantaneous Release ............................................................................12 IV.16. Distribution of the Relative Mass Fraction XR of 237Np in the Matrix Immediately Above the Groundwater at t = 100 Years for Mean Present-Day Infiltration and Instantaneous Release ............................................................................12 IV.17. Distribution of the Relative Mass Fraction XR of 237Np in the Fractures Immediately Above the Groundwater at t = 1,000 Years for Mean Present-Day Infiltration and Instantaneous Release ............................................................................13 IV.18. Distribution of the Relative Mass Fraction XR of 237Np in the Matrix Immediately Above the Groundwater at t = 1,000 Years for Mean Present-Day Infiltration and Instantaneous Release ............................................................................13 IV.19. Distribution of the Relative Mass Fraction XR of 237Np in the Fractures Immediately Above the Groundwater at t = 10,000 Years for Mean Present-Day Infiltration and Instantaneous Release ............................................................................14 IV.20. Distribution of the Relative Mass Fraction XR of 237Np in the Matrix Immediately Above the Groundwater at t = 10,000 Years for Mean Present-Day Infiltration and Instantaneous Release ............................................................................14 IV.21. Distribution of the Relative Mass Fraction XR of 237Np in the Fractures Immediately Above the Groundwater at t = 100,000 Years for Mean Present-Day Infiltration and Instantaneous Release ............................................................................15 IV.22. Distribution of the Relative Mass Fraction XR of 237Np in the Matrix Immediately Above the Groundwater at t = 100,000 Years for Mean Present-Day Infiltration and Instantaneous Release ............................................................................15 IV.23. Distribution of the Relative Mass Fraction XR of 237Np in the Fractures Immediately Above the Groundwater at t = 1,000,000 Years for Mean Present- Day Infiltration and Instantaneous Release ....................................................................16 IV.24. Distribution of the Relative Mass Fraction XR of 237Np in the Matrix Immediately Above the Groundwater at t = 1,000,000 Years for Mean Present- Day Infiltration and Instantaneous Release ....................................................................16 Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment IV-5 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-02 9.50E-03 9.00E-03 8.50E-03 8.00E-03 7.50E-03 7.00E-03 6.50E-03 6.00E-03 5.50E-03 5.00E-03 4.50E-03 4.00E-03 3.50E-03 3.00E-03 2.50E-03 2.00E-03 1.50E-03 1.00E-03 5.00E-04 0.00E+00 Np-PM-TSw-FL-10 Relative Mass Fraction Output-DTN: LB0307MR0060R1.002 Figure IV.1. Distribution of the Relative Mass Fraction XR of 237Np in the Fractures of the tsw39 Layer at t = 10 Years for Mean Present-Day Infiltration and Instantaneous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-04 9.50E-05 9.00E-05 8.50E-05 8.00E-05 7.50E-05 7.00E-05 6.50E-05 6.00E-05 5.50E-05 5.00E-05 4.50E-05 4.00E-05 3.50E-05 3.00E-05 2.50E-05 2.00E-05 1.50E-05 1.00E-05 5.00E-06 0.00E+00 Np-TSw-ML-10 Relative Mass Fraction Output-DTN: LB0307MR0060R1.002 Figure IV.2. Distribution of the Relative Mass Fraction XR of 237Np in the Matrix of the tsw39 layer at t = 10 Years for Mean Present-Day Infiltration and Instantaneous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment IV-6 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-04 9.50E-05 9.00E-05 8.50E-05 8.00E-05 7.50E-05 7.00E-05 6.50E-05 6.00E-05 5.50E-05 5.00E-05 4.50E-05 4.00E-05 3.50E-05 3.00E-05 2.50E-05 2.00E-05 1.50E-05 1.00E-05 5.00E-06 0.00E+00 Np-TSw-FL-100 Relative Mass Fraction Output-DTN: LB0307MR0060R1.002 Figure IV.3. Distribution of the Relative Mass Fraction XR of 237Np in the Fractures of the tsw39 Layer at t = 100 years for Mean Present-Day Infiltration and Instantaneous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-04 9.50E-05 9.00E-05 8.50E-05 8.00E-05 7.50E-05 7.00E-05 6.50E-05 6.00E-05 5.50E-05 5.00E-05 4.50E-05 4.00E-05 3.50E-05 3.00E-05 2.50E-05 2.00E-05 1.50E-05 1.00E-05 5.00E-06 0.00E+00 Relative Mass Fraction Np-TSw-ML-100 Output-DTN: LB0307MR0060R1.002 Figure IV.4. Distribution of the Relative Mass Fraction XR of 237Np in the Matrix of the tsw39 Layer at t = 100 years for Mean Present-Day Infiltration and Instantaneous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment IV-7 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-05 9.50E-06 9.00E-06 8.50E-06 8.00E-06 7.50E-06 7.00E-06 6.50E-06 6.00E-06 5.50E-06 5.00E-06 4.50E-06 4.00E-06 3.50E-06 3.00E-06 2.50E-06 2.00E-06 1.50E-06 1.00E-06 5.00E-07 0.00E+00 Np-PM-TSw-FL-1000 Relative Mass Fraction Output-DTN: LB0307MR0060R1.002 Figure IV.5. Distribution of the Relative Mass Fraction XR of 237Np in the Fractures of the tsw39 layer at t = 1,000 years for Mean Present-Day Infiltration and Instantaneous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-04 9.50E-05 9.00E-05 8.50E-05 8.00E-05 7.50E-05 7.00E-05 6.50E-05 6.00E-05 5.50E-05 5.00E-05 4.50E-05 4.00E-05 3.50E-05 3.00E-05 2.50E-05 2.00E-05 1.50E-05 1.00E-05 5.00E-06 0.00E+00 Np-TSw-ML-1000 Relative Mass Fraction Output-DTN: LB0307MR0060R1.002 Figure IV.6. Distribution of the Relative Mass Fraction XR of 237Np in the Matrix of the tsw39 Layer at t = 1,000 Years for Mean Present-Day Infiltration and Instantaneous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment IV-8 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-05 9.50E-06 9.00E-06 8.50E-06 8.00E-06 7.50E-06 7.00E-06 6.50E-06 6.00E-06 5.50E-06 5.00E-06 4.50E-06 4.00E-06 3.50E-06 3.00E-06 2.50E-06 2.00E-06 1.50E-06 1.00E-06 5.00E-07 0.00E+00 Np-TSw-FL-10,000 Relative Mass Fraction Output-DTN: LB0307MR0060R1.002 Figure IV.7. Distribution of the Relative Mass Fraction XR of 237Np in the Fractures of the tsw39 Layer at t = 10,000 Years for Mean Present-Day Infiltration and Instantaneous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-05 9.50E-06 9.00E-06 8.50E-06 8.00E-06 7.50E-06 7.00E-06 6.50E-06 6.00E-06 5.50E-06 5.00E-06 4.50E-06 4.00E-06 3.50E-06 3.00E-06 2.50E-06 2.00E-06 1.50E-06 1.00E-06 5.00E-07 0.00E+00 Np-TSw-ML-10,000 Relative Mass Fraction Output-DTN: LB0307MR0060R1.002 Figure IV.8. Distribution of the Relative Mass Fraction XR of 237Np in the Matrix of the tsw39 Layer at t = 10,000 Years for Mean Present-Day Infiltration and Instantaneous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment IV-9 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-06 9.50E-07 9.00E-07 8.50E-07 8.00E-07 7.50E-07 7.00E-07 6.50E-07 6.00E-07 5.50E-07 5.00E-07 4.50E-07 4.00E-07 3.50E-07 3.00E-07 2.50E-07 2.00E-07 1.50E-07 1.00E-07 5.00E-08 0.00E+00 Np-TSw-FL-100,000 Relative Mass Fraction Output-DTN: LB0307MR0060R1.002 Figure IV.9. Distribution of the Relative Mass Fraction XR of 237Np in the Fractures of the tsw39 Layer at t = 100,000 Years for Mean Present-Day Infiltration and Instantaneous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-05 9.50E-06 9.00E-06 8.50E-06 8.00E-06 7.50E-06 7.00E-06 6.50E-06 6.00E-06 5.50E-06 5.00E-06 4.50E-06 4.00E-06 3.50E-06 3.00E-06 2.50E-06 2.00E-06 1.50E-06 1.00E-06 5.00E-07 0.00E+00 Np-TSw-ML-100,000 Relative Mass Fraction Output-DTN: LB0307MR0060R1.002 Figure IV.10. Distribution of the Relative Mass Fraction XR of 237Np in the Matrix of the tsw39 Layer at t = 100,000 Years for Mean Present-Day Infiltration and Instantaneous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment IV-10 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-07 9.50E-08 9.00E-08 8.50E-08 8.00E-08 7.50E-08 7.00E-08 6.50E-08 6.00E-08 5.50E-08 5.00E-08 4.50E-08 4.00E-08 3.50E-08 3.00E-08 2.50E-08 2.00E-08 1.50E-08 1.00E-08 5.00E-09 0.00E+00 Np-TSw-FL-1,000,000 Relative Mass Fraction Output-DTN: LB0307MR0060R1.002 Figure IV.11. Distribution of the Relative Mass Fraction XR of 237Np in the Fractures of the tsw39 Layer at t = 1,000,000 Years for Mean Present-Day Infiltration and Instantaneous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-06 9.50E-07 9.00E-07 8.50E-07 8.00E-07 7.50E-07 7.00E-07 6.50E-07 6.00E-07 5.50E-07 5.00E-07 4.50E-07 4.00E-07 3.50E-07 3.00E-07 2.50E-07 2.00E-07 1.50E-07 1.00E-07 5.00E-08 0.00E+00 Np-TSw-ML-1,000,000 Relative Mass Fraction Output-DTN: LB0307MR0060R1.002 Figure IV.12. Distribution of the Relative Mass Fraction XR of 237Np in the Matrix of the tsw39 Layer at t = 1,000,000 Years for Mean Present-Day Infiltration and Instantaneous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment IV-11 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-02 9.50E-03 9.00E-03 8.50E-03 8.00E-03 7.50E-03 7.00E-03 6.50E-03 6.00E-03 5.50E-03 5.00E-03 4.50E-03 4.00E-03 3.50E-03 3.00E-03 2.50E-03 2.00E-03 1.50E-03 1.00E-03 5.00E-04 0.00E+00 Np-WT-FL-10 Relative Mass Fraction Output-DTN: LB0307MR0060R1.002 Figure IV.13. Distribution of the Relative Mass Fraction XR of 237Np in the Fractures Immediately Above the Groundwater Table at t = 10 Years for Mean Present-Day Infiltration and Instantaneous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-06 9.50E-07 9.00E-07 8.50E-07 8.00E-07 7.50E-07 7.00E-07 6.50E-07 6.00E-07 5.50E-07 5.00E-07 4.50E-07 4.00E-07 3.50E-07 3.00E-07 2.50E-07 2.00E-07 1.50E-07 1.00E-07 5.00E-08 0.00E+00 Np-WT-ML-10 Relative Mass Fraction Output-DTN: LB0307MR0060R1.002 Figure IV.14. Distribution of the Relative Mass Fraction XR of 237Np in the Matrix Immediately Above the Groundwater at t = 10 Years for Mean Present-Day Infiltration and Instantaneous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment IV-12 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-04 9.50E-05 9.00E-05 8.50E-05 8.00E-05 7.50E-05 7.00E-05 6.50E-05 6.00E-05 5.50E-05 5.00E-05 4.50E-05 4.00E-05 3.50E-05 3.00E-05 2.50E-05 2.00E-05 1.50E-05 1.00E-05 5.00E-06 0.00E+00 Np-WT-FL-100 Relative Mass Fraction Output-DTN: LB0307MR0060R1.002 Figure IV.15. Distribution of the Relative Mass Fraction XR of 237Np in the Fractures Immediately Above the Groundwater at t = 100 Years for Mean Present-Day Infiltration and Instantaneous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-05 9.50E-06 9.00E-06 8.50E-06 8.00E-06 7.50E-06 7.00E-06 6.50E-06 6.00E-06 5.50E-06 5.00E-06 4.50E-06 4.00E-06 3.50E-06 3.00E-06 2.50E-06 2.00E-06 1.50E-06 1.00E-06 5.00E-07 0.00E+00 Np-WT-ML-100 Relative Mass Fraction Output-DTN: LB0307MR0060R1.002 Figure IV.16. Distribution of the Relative Mass Fraction XR of 237Np in the Matrix Immediately Above the Groundwater at t = 100 Years for Mean Present-Day Infiltration and Instantaneous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment IV-13 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-04 9.50E-05 9.00E-05 8.50E-05 8.00E-05 7.50E-05 7.00E-05 6.50E-05 6.00E-05 5.50E-05 5.00E-05 4.50E-05 4.00E-05 3.50E-05 3.00E-05 2.50E-05 2.00E-05 1.50E-05 1.00E-05 5.00E-06 0.00E+00 Np-WT-FL-1000 Relative Mass Fraction Output-DTN: LB0307MR0060R1.002 Figure IV.17. Distribution of the Relative Mass Fraction XR of 237Np in the Fractures Immediately Above the Groundwater at t = 1,000 Years for Mean Present-Day Infiltration and Instantaneous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-05 9.50E-06 9.00E-06 8.50E-06 8.00E-06 7.50E-06 7.00E-06 6.50E-06 6.00E-06 5.50E-06 5.00E-06 4.50E-06 4.00E-06 3.50E-06 3.00E-06 2.50E-06 2.00E-06 1.50E-06 1.00E-06 5.00E-07 0.00E+00 Np-WT-ML-1000 Relative Mass Fraction Output-DTN: LB0307MR0060R1.002 Figure IV.18. Distribution of the Relative Mass Fraction XR of 237Np in the Matrix Immediately Above the Groundwater at t = 1,000 Years for Mean Present-Day Infiltration and Instantaneous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment IV-14 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-05 9.50E-06 9.00E-06 8.50E-06 8.00E-06 7.50E-06 7.00E-06 6.50E-06 6.00E-06 5.50E-06 5.00E-06 4.50E-06 4.00E-06 3.50E-06 3.00E-06 2.50E-06 2.00E-06 1.50E-06 1.00E-06 5.00E-07 0.00E+00 Np-WT-FL-10,000 Relative Mass Fraction Output-DTN: LB0307MR0060R1.002 Figure IV.19. Distribution of the Relative Mass Fraction XR of 237Np in the Fractures Immediately Above the Groundwater at t = 10,000 Years for Mean Present-Day Infiltration and Instantaneous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-05 9.50E-06 9.00E-06 8.50E-06 8.00E-06 7.50E-06 7.00E-06 6.50E-06 6.00E-06 5.50E-06 5.00E-06 4.50E-06 4.00E-06 3.50E-06 3.00E-06 2.50E-06 2.00E-06 1.50E-06 1.00E-06 5.00E-07 0.00E+00 Np-WT-ML-10,000 Relative Mass Fraction Output-DTN: LB0307MR0060R1.002 Figure IV.20. Distribution of the Relative Mass Fraction XR of 237Np in the Matrix Immediately Above the Groundwater at t = 10,000 Years for Mean Present-Day Infiltration and Instantaneous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment IV-15 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-05 9.50E-06 9.00E-06 8.50E-06 8.00E-06 7.50E-06 7.00E-06 6.50E-06 6.00E-06 5.50E-06 5.00E-06 4.50E-06 4.00E-06 3.50E-06 3.00E-06 2.50E-06 2.00E-06 1.50E-06 1.00E-06 5.00E-07 0.00E+00 Np-WT-FL-100,000 Relative Mass Fraction Output-DTN: LB0307MR0060R1.002 Figure IV.21. Distribution of the Relative Mass Fraction XR of 237Np in the Fractures Immediately Above the Groundwater at t = 100,000 Years for Mean Present-Day Infiltration and Instantaneous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-05 9.50E-06 9.00E-06 8.50E-06 8.00E-06 7.50E-06 7.00E-06 6.50E-06 6.00E-06 5.50E-06 5.00E-06 4.50E-06 4.00E-06 3.50E-06 3.00E-06 2.50E-06 2.00E-06 1.50E-06 1.00E-06 5.00E-07 0.00E+00 Np-WT-ML-100,000 Relative Mass Fraction Output-DTN: LB0307MR0060R1.002 Figure IV.22. Distribution of the Relative Mass Fraction XR of 237Np in the Matrix Immediately Above the Groundwater at t = 100,000 Years for Mean Present-Day Infiltration and Instantaneous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment IV-16 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-07 9.50E-08 9.00E-08 8.50E-08 8.00E-08 7.50E-08 7.00E-08 6.50E-08 6.00E-08 5.50E-08 5.00E-08 4.50E-08 4.00E-08 3.50E-08 3.00E-08 2.50E-08 2.00E-08 1.50E-08 1.00E-08 5.00E-09 0.00E+00 Np-WT-FL-1,000,000 Relative Mass Fraction Output-DTN: LB0307MR0060R1.002 Figure IV.23. Distribution of the Relative Mass Fraction XR of 237Np in the Fractures Immediately Above the Groundwater at t = 1,000,000 Years for Mean Present-Day Infiltration and Instantaneous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-06 9.50E-07 9.00E-07 8.50E-07 8.00E-07 7.50E-07 7.00E-07 6.50E-07 6.00E-07 5.50E-07 5.00E-07 4.50E-07 4.00E-07 3.50E-07 3.00E-07 2.50E-07 2.00E-07 1.50E-07 1.00E-07 5.00E-08 0.00E+00 Np-WT-ML-1,000,000 Relative Mass Fraction Output-DTN: LB0307MR0060R1.002 Figure IV.24. Distribution of the Relative Mass Fraction XR of 237Np in the Matrix Immediately Above the Groundwater at t = 1,000,000 Years for Mean Present-Day Infiltration and Instantaneous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment V-1 November 2003 ATTACHMENT V FIGURES FROM THE 99Tc 3-D TRANSPORT STUDIES (CONTINUOUS RELEASE, MEAN PRESENT-DAY INFILTRATION) Output-DTN: LB0307MR0060R1.004 Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment V-2 November 2003 INTENTIONALLY LEFT BLANK Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment V-3 November 2003 FIGURES Page V.1. Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures of the tsw39 Layer at t = 10 Years for Mean Present-Day Infiltration and Continuous Release...........................................................................................................5 V.2. Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix of the tsw39 Layer at t = 10 Years for Mean Present-Day Infiltration and Continuous Release ..........5 V.3. Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures of the tsw39 Layer at t = 100 Years for Mean Present-Day Infiltration and Continuous Release...........................................................................................................6 V.4. Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix of the tsw39 Layer at t = 100 Years for Mean Present-Day Infiltration and Continuous Release ........6 V.5. Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures of the tsw39 Layer at t = 1,000 Years for Mean Present-Day Infiltration and Continuous Release...........................................................................................................7 V.6. Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix of the tsw39 Layer at t = 1,000 Years for Mean Present-Day Infiltration and Continuous Release...........................................................................................................7 V.7. Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures of the tsw39 Layer at t = 10,000 Years for Mean Present-Day Infiltration and Continuous Release...........................................................................................................8 V.8. Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix of the tsw39 Layer at t = 10,000 Years for Mean Present-Day Infiltration and Continuous Release...........................................................................................................8 V.9. Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures of the tsw39 Layer at t = 100,000 Years for Mean Present-Day Infiltration and Continuous Release...........................................................................................................9 V.10. Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix of the tsw39 Layer at t = 100,000 Years for Mean Present-Day Infiltration and Continuous Release ................9 V.11. Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures Immediately Above the Groundwater Table at t = 10 Years for Mean Present- Day Infiltration and Continuous Release........................................................................10 V.12. Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix Immediately Above the Groundwater at t = 10 Years for Mean Present-Day Infiltration and Continuous Release................................................................................10 V.13. Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures Immediately Above the Groundwater at t = 100 Years for Mean Present-Day Infiltration and Continuous Release................................................................................11 V.14. Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix Immediately Above the groundwater at t = 100 Years for Mean Present-Day Infiltration and Continuous Release.........................................................................................................11 V.15. Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures Immediately Above the Groundwater at T = 1,000 Years for Mean Present-Day Infiltration and Continuous Release................................................................................12 V.16. Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix Immediately Above the Groundwater at t = 1,000 Years for Mean Present-Day Infiltration and Continuous Release.........................................................................................................12 Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment V-4 November 2003 FIGURES Page V.17. Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures Immediately Above the Groundwater at t = 10,000 Years for Mean Present-Day Infiltration and Continuous Release................................................................................13 V.18. Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix Immediately Above the Groundwater at t = 10,000 Years for Mean Present-Day Infiltration and Continuous Release..................................................................................................13 V.19. Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures Immediately Above the Groundwater at t = 100,000 Years for Mean Present-Day Infiltration and Continuous Release................................................................................14 V.20. Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix Immediately Above the Groundwater at t = 100,000 Years for Mean Present-Day Infiltration and Continuous Release..................................................................................................14 Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment V-5 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00 0.95 0.90 0.85 0.80 0.75 0.70 0.65 0.60 0.55 0.50 0.45 0.40 0.35 0.30 0.25 0.20 0.15 0.10 0.05 0.00 Fracture Mass Fraction at Bottom of TSw (for Tc at 10 years) Output-DTN: LB0307MR0060R1.004 Figure V.1. Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures of the tsw39 Layer at t = 10 Years for Mean Present-Day Infiltration and Continuous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00 0.95 0.90 0.85 0.80 0.75 0.70 0.65 0.60 0.55 0.50 0.45 0.40 0.35 0.30 0.25 0.20 0.15 0.10 0.05 0.00 Matrix Mass Fraction at Bottom of TSw (for Tc at 10 years) Output-DTN: LB0307MR0060R1.004 Figure V.2. Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix of the tsw39 Layer at t = 10 Years for Mean Present-Day Infiltration and Continuous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment V-6 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00 0.95 0.90 0.85 0.80 0.75 0.70 0.65 0.60 0.55 0.50 0.45 0.40 0.35 0.30 0.25 0.20 0.15 0.10 0.05 0.00 Fracture Mass Fraction at Bottom of TSw (for Tc at 100 years) Output-DTN: LB0307MR0060R1.004 Figure V.3. Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures of the tsw39 Layer at t = 100 Years for Mean Present-Day Infiltration and Continuous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00 0.95 0.90 0.85 0.80 0.75 0.70 0.65 0.60 0.55 0.50 0.45 0.40 0.35 0.30 0.25 0.20 0.15 0.10 0.05 0.00 Matrix Mass Fraction at Bottom of TSw (for Tc at 100 years) Output-DTN: LB0307MR0060R1.004 Figure V.4. Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix of the tsw39 Layer at t = 100 Years for Mean Present-Day Infiltration and Continuous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment V-7 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00 0.95 0.90 0.85 0.80 0.75 0.70 0.65 0.60 0.55 0.50 0.45 0.40 0.35 0.30 0.25 0.20 0.15 0.10 0.05 0.00 Fracture Mass Fraction at Bottom of TSw (for Tc at 1000 years) Output-DTN: LB0307MR0060R1.004 Figure V.5. Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures of the tsw39 Layer at t = 1,000 Years for Mean Present-Day Infiltration and Continuous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00 0.95 0.90 0.85 0.80 0.75 0.70 0.65 0.60 0.55 0.50 0.45 0.40 0.35 0.30 0.25 0.20 0.15 0.10 0.05 0.00 Matrix Mass Fraction at Bottom of TSw (for Tc at 1000 years) Output-DTN: LB0307MR0060R1.004 Figure V.6. Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix of the tsw39 Layer at t = 1,000 Years for Mean Present-Day Infiltration and Continuous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment V-8 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00 0.95 0.90 0.85 0.80 0.75 0.70 0.65 0.60 0.55 0.50 0.45 0.40 0.35 0.30 0.25 0.20 0.15 0.10 0.05 0.00 Fracture Mass Fraction at Bottom of TSw (for Tc at 10000 years) Output-DTN: LB0307MR0060R1.004 Figure V.7. Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures of the tsw39 Layer at t = 10,000 Years for Mean Present-Day Infiltration and Continuous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00 0.95 0.90 0.85 0.80 0.75 0.70 0.65 0.60 0.55 0.50 0.45 0.40 0.35 0.30 0.25 0.20 0.15 0.10 0.05 0.00 Matrix Mass Fraction at Bottom of TSw (for Tc at 10000 years) Output-DTN: LB0307MR0060R1.004 Figure V.8. Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix of the tsw39 Layer at t = 10,000 Years for Mean Present-Day Infiltration and Continuous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment V-9 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00 0.95 0.90 0.85 0.80 0.75 0.70 0.65 0.60 0.55 0.50 0.45 0.40 0.35 0.30 0.25 0.20 0.15 0.10 0.05 0.00 Fracture Mass Fraction at Bottom of TSw (for Tc at 100000 years) Output-DTN: LB0307MR0060R1.004 Figure V.9. Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures of the tsw39 Layer at t = 100,000 Years for Mean Present-Day Infiltration and Continuous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00 0.95 0.90 0.85 0.80 0.75 0.70 0.65 0.60 0.55 0.50 0.45 0.40 0.35 0.30 0.25 0.20 0.15 0.10 0.05 0.00 Matrix Mass Fraction at Bottom of TSw (for Tc at 100000 years) Output-DTN: LB0307MR0060R1.004 Figure V.10. Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix of the tsw39 Layer at t = 100,000 Years for Mean Present-Day Infiltration and Continuous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment V-10 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 0.50 0.48 0.45 0.43 0.40 0.38 0.35 0.33 0.30 0.28 0.25 0.23 0.20 0.18 0.15 0.13 0.10 0.08 0.05 0.03 0.00 Fracture Mass Fraction atWater Table (for Tc at 10 years) Output-DTN: LB0307MR0060R1.004 Figure V.11. Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures Immediately Above the Groundwater Table at t = 10 Years for Mean Present-Day Infiltration and Continuous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-03 9.50E-04 9.00E-04 8.50E-04 8.00E-04 7.50E-04 7.00E-04 6.50E-04 6.00E-04 5.50E-04 5.00E-04 4.50E-04 4.00E-04 3.50E-04 3.00E-04 2.50E-04 2.00E-04 1.50E-04 1.00E-04 5.00E-05 0.00E+0 Matrix Mass Fraction at Water Table (for Tc at 10 years) Output-DTN: LB0307MR0060R1.004 Figure V.12. Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix Immediately Above the Groundwater at t = 10 Years for Mean Present-Day Infiltration and Continuous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment V-11 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 0.50 0.48 0.45 0.43 0.40 0.38 0.35 0.33 0.30 0.28 0.25 0.23 0.20 0.18 0.15 0.13 0.10 0.08 0.05 0.03 0.00 Fracture Mass Fraction atWater Table (for Tc at 100 years) Output-DTN: LB0307MR0060R1.004 Figure V.13. Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures Immediately Above the Groundwater at t = 100 Years for Mean Present-Day Infiltration and Continuous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 0.100 0.095 0.090 0.085 0.080 0.075 0.070 0.065 0.060 0.055 0.050 0.045 0.040 0.035 0.030 0.025 0.020 0.015 0.010 0.005 0.000 Matrix Mass Fraction at Water Table (for Tc at 100 years) Output-DTN: LB0307MR0060R1.004 Figure V.14. Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix Immediately Above the groundwater at t = 100 Years for Mean Present-Day Infiltration and Continuous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment V-12 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00 0.95 0.90 0.85 0.80 0.75 0.70 0.65 0.60 0.55 0.50 0.45 0.40 0.35 0.30 0.25 0.20 0.15 0.10 0.05 0.00 Fracture Mass Fraction atWater Table (for Tc at 1000 years) Output-DTN: LB0307MR0060R1.004 Figure V.15. Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures Immediately Above the Groundwater at T = 1,000 Years for Mean Present-Day Infiltration and Continuous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00 0.95 0.90 0.85 0.80 0.75 0.70 0.65 0.60 0.55 0.50 0.45 0.40 0.35 0.30 0.25 0.20 0.15 0.10 0.05 0.00 Matrix Mass Fraction at Water Table (for Tc at 1000 years) Output-DTN: LB0307MR0060R1.004 Figure V.16. Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix Immediately Above the Groundwater at t = 1,000 Years for Mean Present-Day Infiltration and Continuous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment V-13 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00 0.95 0.90 0.85 0.80 0.75 0.70 0.65 0.60 0.55 0.50 0.45 0.40 0.35 0.30 0.25 0.20 0.15 0.10 0.05 0.00 Fracture Mass Fraction atWater Table (for Tc at 10000 years) Output-DTN: LB0307MR0060R1.004 Figure V.17. Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures Immediately Above the Groundwater at t = 10,000 Years for Mean Present-Day Infiltration and Continuous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00 0.95 0.90 0.85 0.80 0.75 0.70 0.65 0.60 0.55 0.50 0.45 0.40 0.35 0.30 0.25 0.20 0.15 0.10 0.05 0.00 Matrix Mass Fraction at Water Table (for Tc at 10000 years) Output-DTN: LB0307MR0060R1.004 Figure V.18. Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix Immediately Above the Groundwater at t = 10,000 Years for Mean Present-Day Infiltration and Continuous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment V-14 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00 0.95 0.90 0.85 0.80 0.75 0.70 0.65 0.60 0.55 0.50 0.45 0.40 0.35 0.30 0.25 0.20 0.15 0.10 0.05 0.00 Fracture Mass Fraction atWater Table (for Tc at 100000 years) Output-DTN: LB0307MR0060R1.004 Figure V.19. Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures Immediately Above the Groundwater at t = 100,000 Years for Mean Present-Day Infiltration and Continuous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00 0.95 0.90 0.85 0.80 0.75 0.70 0.65 0.60 0.55 0.50 0.45 0.40 0.35 0.30 0.25 0.20 0.15 0.10 0.05 0.00 Matrix Mass Fraction at Water Table (for Tc at 100000 years) Output-DTN: LB0307MR0060R1.004 Figure V.20. Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix Immediately Above the Groundwater at t = 100,000 Years for Mean Present-Day Infiltration and Continuous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VI-1 November 2003 ATTACHMENT VI FIGURES FROM THE 3-D TRANSPORT STUDIES OF THE 6 nm 239PuO2 COLLOID IN CASE 2 (CONTINOUS RELEASE, MEAN PRESENT-DAY INFILTRATION) Output-DTN: LB0307MR0060R1.004 Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VI-2 November 2003 INTENTIONALLY LEFT BLANK Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VI-3 November 2003 FIGURES Page VI.1. Distribution of the Relative Mass Fraction XR of the 6 nm 239PuO2 Colloid in the Fractures of the tsw39 Layer at t = 10 Years for Mean Present-Day Infiltration and Continuous Release....................................................................................................7 VI.2. Distribution of the Relative Mass Fraction XR of the 6 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 10 Years for Mean Present-Day Infiltration and Continuous Release...........................................................................................................7 VI.3. Distribution of the Relative filtered concentration FR of the 6 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 10 Years for Mean Present-Day Infiltration and Continuous Release..................................................................................8 VI.4. Distribution of the Relative Mass Fraction XR of the 6 nm 239PuO2 Colloid in the Fractures of the tsw39 Layer at t = 100 Years for Mean Present-Day Infiltration and Continuous Release....................................................................................................8 VI.5. Distribution of the Relative Mass Fraction XR of the 6 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 100 Years for Mean Present-Day Infiltration and Continuous Release...........................................................................................................9 VI.6. Distribution of the Relative filtered concentration FR of the 6 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 100 Years for Mean Present-Day Infiltration and Continuous Release..................................................................................9 VI.7. Distribution of the Relative Mass Fraction XR of the 6 nm 239PuO2 Colloid in the Fractures of the tsw39 Layer at t = 1,000 Years for Mean Present-Day Infiltration and Continuous Release..................................................................................................10 VI.8. Distribution of the Relative Mass Fraction XR of the 6 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 1,000 Years for Mean Present-Day Infiltration and Continuous Release..................................................................................................10 VI.9. Distribution of the Relative filtered concentration FR of the 6 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 1,000 Years for Mean Present-Day Infiltration and Continuous Release................................................................................11 VI.10. Distribution of the Relative Mass Fraction XR of the 6 nm 239PuO2 Colloid in the Fractures of the tsw39 Layer at t = 10,000 Years for Mean Present-Day Infiltration and Continuous Release................................................................................11 VI.11. Distribution of the Relative Mass Fraction XR of the 6 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 10,000 Years for Mean Present-Day Infiltration and Continuous Release..................................................................................................12 Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VI-4 November 2003 FIGURES Page VI.12. Distribution of the Relative filtered concentration FR of the 6 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 10,000 Years for Mean Present-Day Infiltration and Continuous Release................................................................................12 VI.13. Distribution of the Relative Mass Fraction XR of the 6 nm 239PuO2 Colloid in the Fractures of the tsw39 Layer at t = 100,000 Years for Mean Present-Day Infiltration and Continuous Release................................................................................13 VI.14. Distribution of the Relative Mass Fraction XR of the 6 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 100,000 Years for Mean Present-Day Infiltration and Continuous Release..................................................................................................13 VI.15. Distribution of the Relative filtered concentration FR of the 6 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 100,000 Years for Mean Present-Day Infiltration and Continuous Release................................................................................14 VI.16. Distribution of the Relative Mass Fraction XR of the 6 nm239PuO2 Colloid in the Fractures immediately above the groundwater table at t = 10 Years for Mean Present-Day Infiltration and Continuous Release...........................................................14 VI.17. Distribution of the Relative Mass Fraction XR of the 6 nm 239PuO2 Colloid in the Matrix immediately above the groundwater at t = 10 Years for Mean Present- Day Infiltration and Continuous Release........................................................................15 VI.18. Distribution of the Relative filtered concentration FR of the 6 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 10 Years for Mean Present-Day Infiltration and Continuous Release................................................................................15 VI.19. Distribution of the Relative Mass Fraction XR of the 6 nm239PuO2 Colloid in the Fractures immediately above the groundwater at t = 100 Years for Mean Present- Day Infiltration and Continuous Release........................................................................16 VI.20. Distribution of the Relative Mass Fraction XR of the 6 nm 239PuO2 Colloid in the Matrix immediately above the groundwater at t = 100 Years for Mean Present- Day Infiltration and Continuous Release........................................................................16 VI.21. Distribution of the Relative filtered concentration FR of the 6 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 100 Years for Mean Present-Day Infiltration and Continuous Release................................................................................17 Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VI-5 November 2003 FIGURES Page VI.22. Distribution of the Relative Mass Fraction XR of the 6 nm239PuO2 Colloid in the Fractures immediately above the groundwater at t = 1,000 Years for Mean Present-Day Infiltration and Continuous Release...........................................................17 VI.23. Distribution of the Relative Mass Fraction XR of the 6 nm 239PuO2 Colloid in the Matrix immediately above the groundwater at t = 1,000 Years for Mean Present- Day Infiltration and Continuous Release........................................................................18 VI.24. Distribution of the Relative filtered concentration FR of the 6 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 1,000 Years for Mean Present-Day Infiltration and Continuous Release................................................................................18 VI.25. Distribution of the Relative Mass Fraction XR of the 6 nm239PuO2 Colloid in the Fractures immediately above the groundwater at t = 10,000 Years for Mean Present-Day Infiltration and Continuous Release...........................................................19 VI.26. Distribution of the Relative Mass Fraction XR of the 6 nm 239PuO2 Colloid in the Matrix immediately above the groundwater at t = 10,000 Years for Mean Present-Day Infiltration and Continuous Release...........................................................19 VI.27. Distribution of the Relative filtered concentration FR of the 6 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 10,000 Years for Mean Present-Day Infiltration and Continuous Release................................................................................20 VI.28. Distribution of the Relative Mass Fraction XR of the 6 nm 239PuO2 Colloid in the Fractures immediately above the groundwater at t = 100,000 Years for Mean Present-Day Infiltration and Continuous Release...........................................................20 VI.29. Distribution of the Relative Mass Fraction XR of the 6 nm 239PuO2 Colloid in the Matrix immediately above the groundwater at t = 100,000 Years for Mean Present-Day Infiltration and Continuous Release...........................................................21 VI.30. Distribution of the Relative filtered concentration FR of the 6 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 100,000 Years for Mean Present-Day Infiltration and Continuous Release................................................................................21 Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VI-6 November 2003 INTENTIONALLY LEFT BLANK Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VI-7 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00 0.95 0.90 0.85 0.80 0.75 0.70 0.65 0.60 0.55 0.50 0.45 0.40 0.35 0.30 0.25 0.20 0.15 0.10 0.05 0.00 Fracture Mass Fraction at Bottom of TSw (for Co006 at 10 years) Output-DTN: LB0307MR0060R1.004 Figure VI.1. Distribution of the Relative Mass Fraction XR of the 6 nm 239PuO2 Colloid in the Fractures of the tsw39 Layer at t = 10 Years for Mean Present-Day Infiltration and Continuous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 2.00E-0 1.90E-0 1.80E-0 1.70E-0 1.60E-0 1.50E-0 1.40E-0 1.30E-0 1.20E-0 1.10E-0 1.00E-0 9.00E-0 8.00E-0 7.00E-0 6.00E-0 5.00E-0 4.00E-0 3.00E-0 2.00E-0 1.00E-0 0.00E+ Matrix Mass Fraction at Bottom of TSw (for Co006 at 10 years) Output-DTN: LB0307MR0060R1.004 Figure VI.2. Distribution of the Relative Mass Fraction XR of the 6 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 10 Years for Mean Present-Day Infiltration and Continuous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VI-8 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E+04 3.98E+03 1.58E+03 6.31E+02 2.51E+02 1.00E+02 3.98E+01 1.58E+01 6.31E+00 2.51E+00 1.00E+00 3.98E-01 1.58E-01 6.31E-02 2.51E-02 1.00E-02 3.98E-03 1.58E-03 6.31E-04 2.51E-04 1.00E-04 Matrix Filtered Concentration at Bottom of TSw (for Co006 at 10 years) kg/m3 Output-DTN: LB0307MR0060R1.004 Figure VI.3. Distribution of the Relative filtered concentration XR of the 6 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 10 Years for Mean Present-Day Infiltration and Continuous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00 0.95 0.90 0.85 0.80 0.75 0.70 0.65 0.60 0.55 0.50 0.45 0.40 0.35 0.30 0.25 0.20 0.15 0.10 0.05 0.00 Fracture Mass Fraction at Bottom of TSw (for Co006 at 100 years) Output-DTN: LB0307MR0060R1.004 Figure VI.4. Distribution of the Relative Mass Fraction XR of the 6 nm 239PuO2 Colloid in the Fractures of the tsw39 Layer at t = 100 Years for Mean Present-Day Infiltration and Continuous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VI-9 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 2.00E-02 1.90E-02 1.80E-02 1.70E-02 1.60E-02 1.50E-02 1.40E-02 1.30E-02 1.20E-02 1.10E-02 1.00E-02 9.00E-03 8.00E-03 7.00E-03 6.00E-03 5.00E-03 4.00E-03 3.00E-03 2.00E-03 1.00E-03 0.00E+00 Matrix Mass Fraction at Bottom of TSw (for Co006 at 100 years) Output-DTN: LB0307MR0060R1.004 Figure VI.5. Distribution of the Relative Mass Fraction XR of the 6 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 100 Years for Mean Present-Day Infiltration and Continuous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E+04 3.98E+03 1.58E+03 6.31E+02 2.51E+02 1.00E+02 3.98E+01 1.58E+01 6.31E+00 2.51E+00 1.00E+00 3.98E-01 1.58E-01 6.31E-02 2.51E-02 1.00E-02 3.98E-03 1.58E-03 6.31E-04 2.51E-04 1.00E-04 Matrix Filtered Concentration at Bottom of TSw (for Co006 at 100 years) kg/m3 Output-DTN: LB0307MR0060R1.004 Figure VI.6. Distribution of the Relative filtered concentration XR of the 6 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 100 Years for Mean Present-Day Infiltration and Continuous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VI-10 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00 0.95 0.90 0.85 0.80 0.75 0.70 0.65 0.60 0.55 0.50 0.45 0.40 0.35 0.30 0.25 0.20 0.15 0.10 0.05 0.00 Fracture Mass Fraction at Bottom of TSw (for Co006 at 1000 years) Output-DTN: LB0307MR0060R1.004 Figure VI.7. Distribution of the Relative Mass Fraction XR of the 6 nm 239PuO2 Colloid in the Fractures of the tsw39 Layer at t = 1,000 Years for Mean Present-Day Infiltration and Continuous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 2.00E-01 1.90E-01 1.80E-01 1.70E-01 1.60E-01 1.50E-01 1.40E-01 1.30E-01 1.20E-01 1.10E-01 1.00E-01 9.00E-02 8.00E-02 7.00E-02 6.00E-02 5.00E-02 4.00E-02 3.00E-02 2.00E-02 1.00E-02 0.00E+00 Matrix Mass Fraction at Bottom of TSw (for Co006 at 1000 years) Output-DTN: LB0307MR0060R1.004 Figure VI.8. Distribution of the Relative Mass Fraction XR of the 6 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 1,000 Years for Mean Present-Day Infiltration and Continuous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VI-11 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E+04 3.98E+03 1.58E+03 6.31E+02 2.51E+02 1.00E+02 3.98E+01 1.58E+01 6.31E+00 2.51E+00 1.00E+00 3.98E-01 1.58E-01 6.31E-02 2.51E-02 1.00E-02 3.98E-03 1.58E-03 6.31E-04 2.51E-04 1.00E-04 Matrix Filtered Concentration at Bottom of TSw (for Co006 at 1000 years) kg/m3 Output-DTN: LB0307MR0060R1.004 Figure VI.9. Distribution of the Relative filtered concentration XR of the 6 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 1,000 Years for Mean Present-Day Infiltration and Continuous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00 0.95 0.90 0.85 0.80 0.75 0.70 0.65 0.60 0.55 0.50 0.45 0.40 0.35 0.30 0.25 0.20 0.15 0.10 0.05 0.00 Fracture Mass Fraction at Bottom of TSw (for Co006 at 10000 years) Output-DTN: LB0307MR0060R1.004 Figure VI.10. Distribution of the Relative Mass Fraction XR of the 6 nm 239PuO2 Colloid in the Fractures of the tsw39 Layer at t = 10,000 Years for Mean Present-Day Infiltration and Continuous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VI-12 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00 0.95 0.90 0.85 0.80 0.75 0.70 0.65 0.60 0.55 0.50 0.45 0.40 0.35 0.30 0.25 0.20 0.15 0.10 0.05 0.00 Matrix Mass Fraction at Bottom of TSw (for Co006 at 10000 years) Output-DTN: LB0307MR0060R1.004 Figure VI.11. Distribution of the Relative Mass Fraction XR of the 6 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 10,000 Years for Mean Present-Day Infiltration and Continuous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E+04 3.98E+03 1.58E+03 6.31E+02 2.51E+02 1.00E+02 3.98E+01 1.58E+01 6.31E+00 2.51E+00 1.00E+00 3.98E-01 1.58E-01 6.31E-02 2.51E-02 1.00E-02 3.98E-03 1.58E-03 6.31E-04 2.51E-04 1.00E-04 Matrix Filtered Concentration at Bottom of TSw (for Co006 at 10000 years) kg/m3 Output-DTN: LB0307MR0060R1.004 Figure VI.12. Distribution of the Relative filtered concentration FR of the 6 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 10,000 Years for Mean Present-Day Infiltration and Continuous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VI-13 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 0.100 0.095 0.090 0.085 0.080 0.075 0.070 0.065 0.060 0.055 0.050 0.045 0.040 0.035 0.030 0.025 0.020 0.015 0.010 0.005 0.000 Fracture Mass Fraction at Bottom of TSw (for Co006 at 100000 years) Output-DTN: LB0307MR0060R1.004 Figure VI.13. Distribution of the Relative Mass Fraction XR of the 6 nm 239PuO2 Colloid in the Fractures of the tsw39 Layer at t = 100,000 Years for Mean Present-Day Infiltration and Continuous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 0.100 0.095 0.090 0.085 0.080 0.075 0.070 0.065 0.060 0.055 0.050 0.045 0.040 0.035 0.030 0.025 0.020 0.015 0.010 0.005 0.000 Matrix Mass Fraction at Bottom of TSw (for Co006 at 100000 years) Output-DTN: LB0307MR0060R1.004 Figure VI.14. Distribution of the Relative Mass Fraction XR of the 6 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 100,000 Years for Mean Present-Day Infiltration and Continuous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VI-14 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E+04 3.98E+03 1.58E+03 6.31E+02 2.51E+02 1.00E+02 3.98E+01 1.58E+01 6.31E+00 2.51E+00 1.00E+00 3.98E-01 1.58E-01 6.31E-02 2.51E-02 1.00E-02 3.98E-03 1.58E-03 6.31E-04 2.51E-04 1.00E-04 Matrix Filtered Concentration at Bottom of TSw (for Co006 at 100000 years) kg/m3 Output-DTN: LB0307MR0060R1.004 Figure VI.15. Distribution of the Relative filtered concentration FR of the 6 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 100,000 Years for Mean Present-Day Infiltration and Continuous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 0.50 0.48 0.45 0.43 0.40 0.38 0.35 0.33 0.30 0.28 0.25 0.23 0.20 0.18 0.15 0.13 0.10 0.08 0.05 0.03 0.00 Fracture Mass Fraction atWater Table (forCo006 at 10 years) Output-DTN: LB0307MR0060R1.004 Figure VI.16. Distribution of the Relative Mass Fraction XR of the 6 nm239PuO2 Colloid in the Fractures immediately above the groundwater table at t = 10 Years for Mean Present-Day Infiltration and Continuous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VI-15 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-07 9.50E-08 9.00E-08 8.50E-08 8.00E-08 7.50E-08 7.00E-08 6.50E-08 6.00E-08 5.50E-08 5.00E-08 4.50E-08 4.00E-08 3.50E-08 3.00E-08 2.50E-08 2.00E-08 1.50E-08 1.00E-08 5.00E-09 0.00E+00 Matrix Mass Fraction at Water Table (for Co006 at 10 years) Output-DTN: LB0307MR0060R1.004 Figure VI.17. Distribution of the Relative Mass Fraction XR of the 6 nm 239PuO2 Colloid in the Matrix immediately above the groundwater at t = 10 Years for Mean Present-Day Infiltration and Continuous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E+04 3.98E+03 1.58E+03 6.31E+02 2.51E+02 1.00E+02 3.98E+01 1.58E+01 6.31E+00 2.51E+00 1.00E+00 3.98E-01 1.58E-01 6.31E-02 2.51E-02 1.00E-02 3.98E-03 1.58E-03 6.31E-04 2.51E-04 1.00E-04 Matrix Filtered Concentration at Water Table ( for Co006 at 10 years ) kg/m3 Output-DTN: LB0307MR0060R1.004 Figure VI.18. Distribution of the Relative filtered concentration FR of the 6 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 10 Years for Mean Present-Day Infiltration and Continuous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VI-16 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 0.50 0.48 0.45 0.43 0.40 0.38 0.35 0.33 0.30 0.28 0.25 0.23 0.20 0.18 0.15 0.13 0.10 0.08 0.05 0.03 0.00 Fracture Mass Fraction atWater Table (for Co006 at 100 years) Output-DTN: LB0307MR0060R1.004 Figure VI.19. Distribution of the Relative Mass Fraction XR of the 6 nm239PuO2 Colloid in the Fractures immediately above the groundwater at t = 100 Years for Mean Present-Day Infiltration and Continuous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-05 9.50E-06 9.00E-06 8.50E-06 8.00E-06 7.50E-06 7.00E-06 6.50E-06 6.00E-06 5.50E-06 5.00E-06 4.50E-06 4.00E-06 3.50E-06 3.00E-06 2.50E-06 2.00E-06 1.50E-06 1.00E-06 5.00E-07 0.00E+00 Matrix Mass Fraction at Water Table (for Co006 at 100 years) Output-DTN: LB0307MR0060R1.004 Figure VI.20. Distribution of the Relative Mass Fraction XR of the 6 nm 239PuO2 Colloid in the Matrix immediately above the groundwater at t = 100 Years for Mean Present-Day Infiltration and Continuous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VI-17 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E+04 3.98E+03 1.58E+03 6.31E+02 2.51E+02 1.00E+02 3.98E+01 1.58E+01 6.31E+00 2.51E+00 1.00E+00 3.98E-01 1.58E-01 6.31E-02 2.51E-02 1.00E-02 3.98E-03 1.58E-03 6.31E-04 2.51E-04 1.00E-04 Matrix Filtered Concentration at Water Table ( for Co006 at 100 years ) kg/m3 Output-DTN: LB0307MR0060R1.004 Figure VI.21. Distribution of the Relative filtered concentration FR of the 6 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 100 Years for Mean Present-Day Infiltration and Continuous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 0.50 0.48 0.45 0.43 0.40 0.38 0.35 0.33 0.30 0.28 0.25 0.23 0.20 0.18 0.15 0.13 0.10 0.08 0.05 0.03 0.00 Fracture Mass Fraction atWater Table (forCo006 at 1000 years) Output-DTN: LB0307MR0060R1.004 Figure VI.22. Distribution of the Relative Mass Fraction XR of the 6 nm239PuO2 Colloid in the Fractures immediately above the groundwater at t = 1,000 Years for Mean Present-Day Infiltration and Continuous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VI-18 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-04 9.50E-05 9.00E-05 8.50E-05 8.00E-05 7.50E-05 7.00E-05 6.50E-05 6.00E-05 5.50E-05 5.00E-05 4.50E-05 4.00E-05 3.50E-05 3.00E-05 2.50E-05 2.00E-05 1.50E-05 1.00E-05 5.00E-06 0.00E+00 Matrix Mass Fraction at Water Table (for Co006 at 1000 years) Output-DTN: LB0307MR0060R1.004 Figure VI.23. Distribution of the Relative Mass Fraction XR of the 6 nm 239PuO2 Colloid in the Matrix immediately above the groundwater at t = 1,000 Years for Mean Present- Day Infiltration and Continuous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E+04 3.98E+03 1.58E+03 6.31E+02 2.51E+02 1.00E+02 3.98E+01 1.58E+01 6.31E+00 2.51E+00 1.00E+00 3.98E-01 1.58E-01 6.31E-02 2.51E-02 1.00E-02 3.98E-03 1.58E-03 6.31E-04 2.51E-04 1.00E-04 Matrix Filtered Concentration at Water Table ( for Co006 at 1000 years ) kg/m3 Output-DTN: LB0307MR0060R1.004 Figure VI.24. Distribution of the Relative filtered concentration FR of the 6 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 1,000 Years for Mean Present-Day Infiltration and Continuous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VI-19 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 0.50 0.48 0.45 0.43 0.40 0.38 0.35 0.33 0.30 0.28 0.25 0.23 0.20 0.18 0.15 0.13 0.10 0.08 0.05 0.03 0.00 Fracture Mass Fraction atWater Table (for Co006 at 10000 years) Output-DTN: LB0307MR0060R1.004 Figure VI.25. Distribution of the Relative Mass Fraction XR of the 6 nm239PuO2 Colloid in the Fractures immediately above the groundwater at t = 10,000 Years for Mean Present-Day Infiltration and Continuous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-02 9.50E-03 9.00E-03 8.50E-03 8.00E-03 7.50E-03 7.00E-03 6.50E-03 6.00E-03 5.50E-03 5.00E-03 4.50E-03 4.00E-03 3.50E-03 3.00E-03 2.50E-03 2.00E-03 1.50E-03 1.00E-03 5.00E-04 0.00E+00 Matrix Mass Fraction at Water Table (for Co006 at 10000 years) Output-DTN: LB0307MR0060R1.004 Figure VI.26. Distribution of the Relative Mass Fraction XR of the 6 nm 239PuO2 Colloid in the Matrix immediately above the groundwater at t = 10,000 Years for Mean Present- Day Infiltration and Continuous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VI-20 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E+04 3.98E+03 1.58E+03 6.31E+02 2.51E+02 1.00E+02 3.98E+01 1.58E+01 6.31E+00 2.51E+00 1.00E+00 3.98E-01 1.58E-01 6.31E-02 2.51E-02 1.00E-02 3.98E-03 1.58E-03 6.31E-04 2.51E-04 1.00E-04 Matrix Filtered Concentration at Water Table ( for Co006 at 10000 years ) kg/m3 Output-DTN: LB0307MR0060R1.004 Figure VI.27. Distribution of the Relative filtered concentration FR of the 6 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 10,000 Years for Mean Present-Day Infiltration and Continuous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 0.100 0.095 0.090 0.085 0.080 0.075 0.070 0.065 0.060 0.055 0.050 0.045 0.040 0.035 0.030 0.025 0.020 0.015 0.010 0.005 0.000 Fracture Mass Fraction atWater Table (for Co006 at 100000 years) Output-DTN: LB0307MR0060R1.004 Figure VI.28. Distribution of the Relative Mass Fraction XR of the 6 nm 239PuO2 Colloid in the Fractures immediately above the groundwater at t = 100,000 Years for Mean Present-Day Infiltration and Continuous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VI-21 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 3.21E-02 3.06E-02 2.91E-02 2.75E-02 2.60E-02 2.45E-02 2.29E-02 2.14E-02 1.99E-02 1.84E-02 1.68E-02 1.53E-02 1.38E-02 1.22E-02 1.07E-02 9.18E-03 7.65E-03 6.12E-03 4.59E-03 3.06E-03 1.53E-03 Matrix Mass Fraction at Water Table (for Co006 at 100000 years) Output-DTN: LB0307MR0060R1.004 Figure VI.29. Distribution of the Relative Mass Fraction XR of the 6 nm 239PuO2 Colloid in the Matrix immediately above the groundwater at t = 100,000 Years for Mean Present-Day Infiltration and Continuous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E+04 3.98E+03 1.58E+03 6.31E+02 2.51E+02 1.00E+02 3.98E+01 1.58E+01 6.31E+00 2.51E+00 1.00E+00 3.98E-01 1.58E-01 6.31E-02 2.51E-02 1.00E-02 3.98E-03 1.58E-03 6.31E-04 2.51E-04 1.00E-04 Matrix Filtered Concentration at Water Table ( for Co006 at 100000 years ) kg/m3 Output-DTN: LB0307MR0060R1.004 Figure VI.30. Distribution of the Relative filtered concentration FR of the 6 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 100,000 Years for Mean Present-Day Infiltration and Continuous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VI-22 November 2003 INTENTIONALLY LEFT BLANK Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VII-1 November 2003 ATTACHMENT VII FIGURES FROM THE 3-D TRANSPORT STUDIES OF THE 450 nm 239PuO2 COLLOID IN CASE 2 (CONTINOUS RELEASE, MEAN PRESENT-DAY INFILTRATION) Output-DTN: LB0307MR0060R1.004 Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VII-2 November 2003 INTENTIONALLY LEFT BLANK Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VII-3 November 2003 FIGURES Page VII.1. Distribution of the Relative Mass Fraction XR of the 450 nm 239PuO2 Colloid in the Fractures of the tsw39 Layer at t = 10 Years for Mean Present-Day Infiltration and Continuous Release..................................................................................7 VII.2. Distribution of the Relative Mass Fraction XR of the 450 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 10 Years for Mean Present-Day Infiltration and Continuous Release....................................................................................................7 VII.3. Distribution of the Relative Filtered Concentration XR of the 450 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 10 Years for Mean Present-Day Infiltration and Continuous Release..................................................................................8 VII.4. Distribution of the Relative Mass Fraction XR of the 450 nm 239PuO2 Colloid in the Fractures of the tsw39 Layer at t = 100 Years for Mean Present-Day Infiltration and Continuous Release..................................................................................8 VII.5. Distribution of the Relative Mass Fraction XR of the 450 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 100 Years for Mean Present-Day Infiltration and Continuous Release....................................................................................................9 VII.6. Distribution of the Relative Filtered Concentration XR of the 450 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 100 Years for Mean Present-Day Infiltration and Continuous Release..................................................................................9 VII.7. Distribution of the Relative Mass Fraction XR of the 450 nm 239PuO2 Colloid in the Fractures of the tsw39 Layer at t = 1,000 Years for Mean Present-Day Infiltration and Continuous Release................................................................................10 VII.8. Distribution of the Relative Mass Fraction XR of the 450 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 1,000 Years for Mean Present-Day Infiltration and Continuous Release................................................................................10 VII.9. Distribution of the Relative Filtered Concentration XR of the 450 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 1,000 Years for Mean Present- Day Infiltration and Continuous Release........................................................................11 VII.10. Distribution of the Relative Mass Fraction XR of the 450 nm 239PuO2 Colloid in the Fractures of the tsw39 Layer at t = 10,000 Years for Mean Present-Day Infiltration and Continuous Release................................................................................11 Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VII-4 November 2003 FIGURES Page VII.11. Distribution of the Relative Mass Fraction XR of the 450 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 10,000 Years for Mean Present-Day Infiltration and Continuous Release................................................................................12 VII.12. Distribution of the Relative Filtered Concentration XR of the 450 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 10,000 Years for Mean Present- Day Infiltration and Continuous Release........................................................................12 VII.13. Distribution of the Relative Mass Fraction XR of the 450 nm 239PuO2 Colloid in the Fractures of the tsw39 Layer at t = 100,000 Years for Mean Present-Day Infiltration and Continuous Release................................................................................13 VII.14. Distribution of the Relative Mass Fraction XR of the 450 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 100,000 Years for Mean Present-Day Infiltration and Continuous Release................................................................................13 VII.15. Distribution of the Relative Filtered Concentration XR of the 450 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 100,000 Years for Mean Present- Day Infiltration and Continuous Release........................................................................14 VII.16. Distribution of the Relative Mass Fraction XR of the 450 nm 239PuO2 Colloid in the Fractures Immediately Above the Groundwater Table at t = 10 Years for Mean Present-Day Infiltration and Continuous Release.................................................14 VII.17. Distribution of the Relative Mass Fraction XR of the 450 nm 239PuO2 Colloid in the Matrix Immediately Above the Groundwater at t = 10 Years for Mean Present-Day Infiltration and Continuous Release...........................................................15 VII.18. Distribution of the Relative Filtered Concentration XR of the 450 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 10 Years for Mean Present-Day Infiltration and Continuous Release................................................................................15 VII.19. Distribution of the Relative Mass Fraction XR of the 450 nm 239PuO2 Colloid in the Fractures Immediately Above the Groundwater at t = 100 Years for Mean Present-Day Infiltration and Continuous Release...........................................................16 VII.20. Distribution of the Relative Mass Fraction XR of the 450 nm 239PuO2 Colloid in the Matrix Immediately Above the Groundwater at t = 100 Years for Mean Present-Day Infiltration and Continuous Release...........................................................16 Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VII-5 November 2003 FIGURES Page VII.21. Distribution of the Relative Filtered Concentration XR of the 450 nm 239PuO2 Colloid in the Matrix of the tsw39 layer at t = 100 Years for Mean Present-Day Infiltration and Continuous Release................................................................................17 VII.22. Distribution of the Relative Mass Fraction XR of the 450 nm 239PuO2 Colloid in the Fractures Immediately Above the Groundwater at t = 1,000 Years for Mean Present-Day Infiltration and Continuous Release...........................................................17 VII.23. Distribution of the Relative Mass Fraction XR of the 450 nm 239PuO2 Colloid in the Matrix Immediately Above the Groundwater at t = 1,000 Years for Mean Present-Day Infiltration and Continuous Release...........................................................18 VII.24. Distribution of the Relative Filtered Concentration XR of the 450 nm 239PuO2 Colloid in the Matrix of the tsw39 layer at t = 1,000 Years for Mean Present-Day Infiltration and Continuous Release................................................................................18 VII.25. Distribution of the Relative Mass Fraction XR of the 450 nm 239PuO2 Colloid in the Fractures Immediately Above the Groundwater at t = 10,000 Years for Mean Present-Day Infiltration and Continuous Release...........................................................19 VII.26. Distribution of the Relative Mass Fraction XR of the 450 nm 239PuO2 Colloid in the Matrix Immediately Above the Groundwater at t = 10,000 Years for Mean Present-Day Infiltration and Continuous Release...........................................................19 VII.27. Distribution of the Relative Filtered Concentration XR of the 450 nm 239PuO2 Colloid in the Matrix of the tsw39 layer at t = 10,000 Years for Mean Present- Day Infiltration and Continuous Release........................................................................20 VII.28. Distribution of the Relative Mass Fraction XR of the 450 nm 239PuO2 Colloid in the Fractures Immediately Above the Groundwater at t = 100,000 Years for Mean Present-Day Infiltration and Continuous Release.................................................20 VII.29. Distribution of the Relative Mass Fraction XR of the 450 nm 239PuO2 Colloid in the Matrix Immediately Above the Groundwater at t = 100,000 Years for Mean Present-Day Infiltration and Continuous Release...........................................................21 VII.30. Distribution of the Relative Filtered Concentration XR of the 450 nm 239PuO2 Colloid in the Matrix of the tsw39 layer at t = 100,000 Years for Mean Present- Day Infiltration and Continuous Release........................................................................21 Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VII-6 November 2003 INTENTIONALLY LEFT BLANK Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VII-7 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00 0.95 0.90 0.85 0.80 0.75 0.70 0.65 0.60 0.55 0.50 0.45 0.40 0.35 0.30 0.25 0.20 0.15 0.10 0.05 0.00 Fracture Mass Fraction at Bottom of TSw (for Co450 at 10 years) Output-DTN: LB0307MR0060R1.004 Figure VII.1. Distribution of the Relative Mass Fraction XR of the 450 nm 239PuO2 Colloid in the Fractures of the tsw39 Layer at t = 10 Years for Mean Present-Day Infiltration and Continuous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-03 9.50E-04 9.00E-04 8.50E-04 8.00E-04 7.50E-04 7.00E-04 6.50E-04 6.00E-04 5.50E-04 5.00E-04 4.50E-04 4.00E-04 3.50E-04 3.00E-04 2.50E-04 2.00E-04 1.50E-04 1.00E-04 5.00E-05 0.00E+00 Matrix Mass Fraction at Bottom of TSw (for Co450 at 10 years) Output-DTN: LB0307MR0060R1.004 Figure VII.2. Distribution of the Relative Mass Fraction XR of the 450 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 10 Years for Mean Present-Day Infiltration and Continuous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VII-8 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E+02 7.08E+01 5.01E+01 3.55E+01 2.51E+01 1.78E+01 1.26E+01 8.91E+00 6.31E+00 4.47E+00 3.16E+00 2.24E+00 1.58E+00 1.12E+00 7.94E-01 5.62E-01 3.98E-01 2.82E-01 2.00E-01 1.41E-01 1.00E-01 Matrix Filtered Concentration at Bottom of TSw (for Co450 at 10 years) kg/m3 Output-DTN: LB0307MR0060R1.004 Figure VII.3. Distribution of the Relative Filtered Concentration XR of the 450 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 10 Years for Mean Present-Day Infiltration and Continuous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00 0.95 0.90 0.85 0.80 0.75 0.70 0.65 0.60 0.55 0.50 0.45 0.40 0.35 0.30 0.25 0.20 0.15 0.10 0.05 0.00 Fracture Mass Fraction at Bottom of TSw (for Co450 at 100 years) Output-DTN: LB0307MR0060R1.004 Figure VII.4. Distribution of the Relative Mass Fraction XR of the 450 nm 239PuO2 Colloid in the Fractures of the tsw39 Layer at t = 100 Years for Mean Present-Day Infiltration and Continuous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VII-9 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 0.020 0.019 0.018 0.017 0.016 0.015 0.014 0.013 0.012 0.011 0.010 0.009 0.008 0.007 0.006 0.005 0.004 0.003 0.002 0.001 0.000 Matrix Mass Fraction at Bottom of TSw (for Co450 at 100 years) Output-DTN: LB0307MR0060R1.004 Figure VII.5. Distribution of the Relative Mass Fraction XR of the 450 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 100 Years for Mean Present-Day Infiltration and Continuous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E+04 3.98E+03 1.58E+03 6.31E+02 2.51E+02 1.00E+02 3.98E+01 1.58E+01 6.31E+00 2.51E+00 1.00E+00 3.98E-01 1.58E-01 6.31E-02 2.51E-02 1.00E-02 3.98E-03 1.58E-03 6.31E-04 2.51E-04 1.00E-04 Matrix Filtered Concentration at Bottom of TSw (for Co450 at 100 years) kg/m3 Output-DTN: LB0307MR0060R1.004 Figure VII.6. Distribution of the Relative Filtered Concentration XR of the 450 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 100 Years for Mean Present-Day Infiltration and Continuous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VII-10 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00 0.95 0.90 0.85 0.80 0.75 0.70 0.65 0.60 0.55 0.50 0.45 0.40 0.35 0.30 0.25 0.20 0.15 0.10 0.05 0.00 Fracture Mass Fraction at Bottom of TSw (for Co450 at 1000 years) Output-DTN: LB0307MR0060R1.004 Figure VII.7. Distribution of the Relative Mass Fraction XR of the 450 nm 239PuO2 Colloid in the Fractures of the tsw39 Layer at t = 1,000 Years for Mean Present-Day Infiltration and Continuous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 0.20 0.19 0.18 0.17 0.16 0.15 0.14 0.13 0.12 0.11 0.10 0.09 0.08 0.07 0.06 0.05 0.04 0.03 0.02 0.01 0.00 Matrix Mass Fraction at Bottom of TSw (for Co450 at 1000 years) Output-DTN: LB0307MR0060R1.004 Figure VII.8. Distribution of the Relative Mass Fraction XR of the 450 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 1,000 Years for Mean Present-Day Infiltration and Continuous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VII-11 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E+04 3.98E+03 1.58E+03 6.31E+02 2.51E+02 1.00E+02 3.98E+01 1.58E+01 6.31E+00 2.51E+00 1.00E+00 3.98E-01 1.58E-01 6.31E-02 2.51E-02 1.00E-02 3.98E-03 1.58E-03 6.31E-04 2.51E-04 1.00E-04 Matrix Filtered Concentration at Bottom of TSw (for Co450 at 1000 years) kg/m3 Output-DTN: LB0307MR0060R1.004 Figure VII.9. Distribution of the Relative Filtered Concentration XR of the 450 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 1,000 Years for Mean Present-Day Infiltration and Continuous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00 0.95 0.90 0.85 0.80 0.75 0.70 0.65 0.60 0.55 0.50 0.45 0.40 0.35 0.30 0.25 0.20 0.15 0.10 0.05 0.00 Fracture Mass Fraction at Bottom of TSw (for Co450 at 10000 years) Output-DTN: LB0307MR0060R1.004 Figure VII.10. Distribution of the Relative Mass Fraction XR of the 450 nm 239PuO2 Colloid in the Fractures of the tsw39 Layer at t = 10,000 Years for Mean Present-Day Infiltration and Continuous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VII-12 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00 0.95 0.90 0.85 0.80 0.75 0.70 0.65 0.60 0.55 0.50 0.45 0.40 0.35 0.30 0.25 0.20 0.15 0.10 0.05 0.00 Matrix Mass Fraction at Bottom of TSw (for Co450 at 10000 years) Output-DTN: LB0307MR0060R1.004 Figure VII.11. Distribution of the Relative Mass Fraction XR of the 450 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 10,000 Years for Mean Present-Day Infiltration and Continuous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E+04 3.98E+03 1.58E+03 6.31E+02 2.51E+02 1.00E+02 3.98E+01 1.58E+01 6.31E+00 2.51E+00 1.00E+00 3.98E-01 1.58E-01 6.31E-02 2.51E-02 1.00E-02 3.98E-03 1.58E-03 6.31E-04 2.51E-04 1.00E-04 Matrix Filtered Concentration at Bottom of TSw (for Co450 at 10000 years) kg/m3 Output-DTN: LB0307MR0060R1.004 Figure VII.12. Distribution of the Relative Filtered Concentration XR of the 450 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 10,000 Years for Mean Present-Day Infiltration and Continuous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VII-13 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 0.100 0.095 0.090 0.085 0.080 0.075 0.070 0.065 0.060 0.055 0.050 0.045 0.040 0.035 0.030 0.025 0.020 0.015 0.010 0.005 0.000 Fracture Mass Fraction at Bottom of TSw (for Co450 at 100000 years) Output-DTN: LB0307MR0060R1.004 Figure VII.13. Distribution of the Relative Mass Fraction XR of the 450 nm 239PuO2 Colloid in the Fractures of the tsw39 Layer at t = 100,000 Years for Mean Present-Day Infiltration and Continuous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 0.20 0.19 0.18 0.17 0.16 0.15 0.14 0.13 0.12 0.11 0.10 0.09 0.08 0.07 0.06 0.05 0.04 0.03 0.02 0.01 0.00 Matrix Mass Fraction at Bottom of TSw (for Co450 at 100000 years) Output-DTN: LB0307MR0060R1.004 Figure VII.14. Distribution of the Relative Mass Fraction XR of the 450 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 100,000 Years for Mean Present-Day Infiltration and Continuous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VII-14 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E+04 3.98E+03 1.58E+03 6.31E+02 2.51E+02 1.00E+02 3.98E+01 1.58E+01 6.31E+00 2.51E+00 1.00E+00 3.98E-01 1.58E-01 6.31E-02 2.51E-02 1.00E-02 3.98E-03 1.58E-03 6.31E-04 2.51E-04 1.00E-04 Matrix Filtered Concentration at Bottom of TSw (for Co450 at 100000 years) kg/m3 Output-DTN: LB0307MR0060R1.004 Figure VII.15. Distribution of the Relative Filtered Concentration XR of the 450 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 100,000 Years for Mean Present- Day Infiltration and Continuous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.50 1.43 1.35 1.27 1.20 1.13 1.05 0.97 0.90 0.82 0.75 0.67 0.60 0.53 0.45 0.38 0.30 0.22 0.15 0.07 0.00 Fracture Mass Fraction atWater Table (forCo450 at 10 years) Output-DTN: LB0307MR0060R1.004 Figure VII.16. Distribution of the Relative Mass Fraction XR of the 450 nm 239PuO2 Colloid in the Fractures Immediately Above the Groundwater Table at t = 10 Years for Mean Present-Day Infiltration and Continuous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VII-15 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-05 9.50E-06 9.00E-06 8.50E-06 8.00E-06 7.50E-06 7.00E-06 6.50E-06 6.00E-06 5.50E-06 5.00E-06 4.50E-06 4.00E-06 3.50E-06 3.00E-06 2.50E-06 2.00E-06 1.50E-06 1.00E-06 5.00E-07 0.00E+00 Matrix Mass Fraction at Water Table (for Co450 at 10 years) Output-DTN: LB0307MR0060R1.004 Figure VII.17. Distribution of the Relative Mass Fraction XR of the 450 nm 239PuO2 Colloid in the Matrix Immediately Above the Groundwater at t = 10 Years for Mean Present-Day Infiltration and Continuous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E+04 3.98E+03 1.58E+03 6.31E+02 2.51E+02 1.00E+02 3.98E+01 1.58E+01 6.31E+00 2.51E+00 1.00E+00 3.98E-01 1.58E-01 6.31E-02 2.51E-02 1.00E-02 3.98E-03 1.58E-03 6.31E-04 2.51E-04 1.00E-04 Matrix Filtered Concentration at Water Table ( for Co450 at 10 years ) kg/m3 Output-DTN: LB0307MR0060R1.004 Figure VII.18. Distribution of the Relative Filtered Concentration XR of the 450 nm 239PuO2 Colloid in the Matrix of the tsw39 Layer at t = 10 Years for Mean Present-Day Infiltration and Continuous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VII-16 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.50 1.43 1.35 1.27 1.20 1.13 1.05 0.97 0.90 0.82 0.75 0.67 0.60 0.53 0.45 0.38 0.30 0.22 0.15 0.07 0.00 Fracture Mass Fraction atWater Table (for Co450 at 100 years) Output-DTN: LB0307MR0060R1.004 Figure VII.19. Distribution of the Relative Mass Fraction XR of the 450 nm 239PuO2 Colloid in the Fractures Immediately Above the Groundwater at t = 100 Years for Mean Present-Day Infiltration and Continuous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-04 9.50E-05 9.00E-05 8.50E-05 8.00E-05 7.50E-05 7.00E-05 6.50E-05 6.00E-05 5.50E-05 5.00E-05 4.50E-05 4.00E-05 3.50E-05 3.00E-05 2.50E-05 2.00E-05 1.50E-05 1.00E-05 5.00E-06 0.00E+00 Matrix Mass Fraction at Water Table (for Co450 at 100 years) Output-DTN: LB0307MR0060R1.004 Figure VII.20. Distribution of the Relative Mass Fraction XR of the 450 nm 239PuO2 Colloid in the Matrix Immediately Above the Groundwater at t = 100 Years for Mean Present-Day Infiltration and Continuous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VII-17 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E+04 3.98E+03 1.58E+03 6.31E+02 2.51E+02 1.00E+02 3.98E+01 1.58E+01 6.31E+00 2.51E+00 1.00E+00 3.98E-01 1.58E-01 6.31E-02 2.51E-02 1.00E-02 3.98E-03 1.58E-03 6.31E-04 2.51E-04 1.00E-04 Matrix Filtered Concentration at Water Table ( for Co450 at 100 years ) kg/m3 Output-DTN: LB0307MR0060R1.004 Figure VII.21. Distribution of the Relative Filtered Concentration XR of the 450 nm 239PuO2 Colloid in the Matrix of the tsw39 layer at t = 100 Years for Mean Present-Day Infiltration and Continuous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.50 1.43 1.35 1.27 1.20 1.13 1.05 0.97 0.90 0.82 0.75 0.67 0.60 0.53 0.45 0.38 0.30 0.22 0.15 0.07 0.00 Fracture Mass Fraction atWater Table (forCo450 at 1000 years) Output-DTN: LB0307MR0060R1.004 Figure VII.22. Distribution of the Relative Mass Fraction XR of the 450 nm 239PuO2 Colloid in the Fractures Immediately Above the Groundwater at t = 1,000 Years for Mean Present-Day Infiltration and Continuous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VII-18 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-02 9.50E-03 9.00E-03 8.50E-03 8.00E-03 7.50E-03 7.00E-03 6.50E-03 6.00E-03 5.50E-03 5.00E-03 4.50E-03 4.00E-03 3.50E-03 3.00E-03 2.50E-03 2.00E-03 1.50E-03 1.00E-03 5.00E-04 0.00E+00 Matrix Mass Fraction at Water Table (for Co450 at 1000 years) Output-DTN: LB0307MR0060R1.004 Figure VII.23. Distribution of the Relative Mass Fraction XR of the 450 nm 239PuO2 Colloid in the Matrix Immediately Above the Groundwater at t = 1,000 Years for Mean Present-Day Infiltration and Continuous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E+04 3.98E+03 1.58E+03 6.31E+02 2.51E+02 1.00E+02 3.98E+01 1.58E+01 6.31E+00 2.51E+00 1.00E+00 3.98E-01 1.58E-01 6.31E-02 2.51E-02 1.00E-02 3.98E-03 1.58E-03 6.31E-04 2.51E-04 1.00E-04 Matrix Filtered Concentration at Water Table ( for Co450 at 1000 years ) kg/m3 Output-DTN: LB0307MR0060R1.004 Figure VII.24. Distribution of the Relative Filtered Concentration XR of the 450 nm 239PuO2 Colloid in the Matrix of the tsw39 layer at t = 1,000 Years for Mean Present-Day Infiltration and Continuous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VII-19 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.50 1.43 1.35 1.27 1.20 1.13 1.05 0.97 0.90 0.82 0.75 0.67 0.60 0.53 0.45 0.38 0.30 0.22 0.15 0.07 0.00 Fracture Mass Fraction atWater Table (for Co450 at 10000 years) Output-DTN: LB0307MR0060R1.004 Figure VII.25. Distribution of the Relative Mass Fraction XR of the 450 nm 239PuO2 Colloid in the Fractures Immediately Above the Groundwater at t = 10,000 Years for Mean Present-Day Infiltration and Continuous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 0.20 0.19 0.18 0.17 0.16 0.15 0.14 0.13 0.12 0.11 0.10 0.09 0.08 0.07 0.06 0.05 0.04 0.03 0.02 0.01 0.00 Matrix Mass Fraction at Water Table (for Co450 at 10000 years) Output-DTN: LB0307MR0060R1.004 Figure VII.26. Distribution of the Relative Mass Fraction XR of the 450 nm 239PuO2 Colloid in the Matrix Immediately Above the Groundwater at t = 10,000 Years for Mean Present-Day Infiltration and Continuous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VII-20 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E+04 3.98E+03 1.58E+03 6.31E+02 2.51E+02 1.00E+02 3.98E+01 1.58E+01 6.31E+00 2.51E+00 1.00E+00 3.98E-01 1.58E-01 6.31E-02 2.51E-02 1.00E-02 3.98E-03 1.58E-03 6.31E-04 2.51E-04 1.00E-04 Matrix Filtered Concentration at Water Table ( for Co450 at 10000 years ) kg/m3 Output-DTN: LB0307MR0060R1.004 Figure VII.27. Distribution of the Relative Filtered Concentration XR of the 450 nm 239PuO2 Colloid in the Matrix of the tsw39 layer at t = 10,000 Years for Mean Present-Day Infiltration and Continuous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 0.30 0.28 0.27 0.26 0.24 0.22 0.21 0.20 0.18 0.16 0.15 0.14 0.12 0.11 0.09 0.07 0.06 0.04 0.03 0.01 0.00 Fracture Mass Fraction atWater Table (for Co450 at 100000 years) Output-DTN: LB0307MR0060R1.004 Figure VII.28. Distribution of the Relative Mass Fraction XR of the 450 nm 239PuO2 Colloid in the Fractures Immediately Above the Groundwater at t = 100,000 Years for Mean Present-Day Infiltration and Continuous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VII-21 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 0.030 0.029 0.027 0.026 0.024 0.022 0.021 0.020 0.018 0.017 0.015 0.014 0.012 0.011 0.009 0.007 0.006 0.005 0.003 0.002 0.000 Matrix Mass Fraction at Water Table (for Co450 at 100000 years) Output-DTN: LB0307MR0060R1.004 Figure VII.29. Distribution of the Relative Mass Fraction XR of the 450 nm 239PuO2 Colloid in the Matrix Immediately Above the Groundwater at t = 100,000 Years for Mean Present-Day Infiltration and Continuous Release Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E+04 3.98E+03 1.58E+03 6.31E+02 2.51E+02 1.00E+02 3.98E+01 1.58E+01 6.31E+00 2.51E+00 1.00E+00 3.98E-01 1.58E-01 6.31E-02 2.51E-02 1.00E-02 3.98E-03 1.58E-03 6.31E-04 2.51E-04 1.00E-04 Matrix Filtered Concentration at Water Table ( for Co450 at 100000 years ) kg/m3 Output-DTN: LB0307MR0060R1.004 Figure VII.30. Distribution of the Relative Filtered Concentration XR of the 450 nm 239PuO2 Colloid in the Matrix of the tsw39 layer at t = 100,000 Years for Mean Present- Day Infiltration and Continuous Release Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VII-22 November 2003 INTENTIONALLY LEFT BLANK Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VIII-1 November 2003 ATTACHMENT VIII FIGURES FROM THE 99Tc 3-D TRANSPORT STUDIES (INSTANTANEOUS RELEASE ONLY FROM NON-FAULTED DOMAIN, MEAN PRESENT-DAY INFILTRATION) Output-DTN: LB0307MR0060R1.002 Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VIII-2 November 2003 INTENTIONALLY LEFT BLANK Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VIII-3 November 2003 FIGURES Page VIII.1. Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures of the TSw39 Layer at t = 10 Years for Mean Present-Day Infiltration and Instantaneous Release only from Non-Faulted Domain. ..................................................5 VIII.2. Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix of the TSw39 Layer at t = 10 Years for Mean Present-Day Infiltration and Instantaneous Release only from Non-Faulted Domain. .........................................................................5 VIII.3. Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures of the TSw39 Layer at t = 100 Years for Mean Present-Day Infiltration and Instantaneous Release only from Non-Faulted Domain. ..................................................6 VIII.4. Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix of the TSw39 Layer at t = 100 Years for Mean Present-Day Infiltration and Instantaneous Release only from Non-Faulted Domain. ..................................................6 VIII.5. Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures of the TSw39 Layer at t = 1,000 Years for Mean Present-Day Infiltration and Instantaneous Release only from Non-Faulted Domain. ..................................................7 VIII.6. Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix of the TSw39 Layer at t = 1,000 Years for Mean Present-Day Infiltration and Instantaneous Release only from Non-Faulted Domain. ..................................................7 VIII.7. Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures of the TSw39 Layer at t = 10,000 Years for Mean Present-Day Infiltration and Instantaneous Release only from Non-Faulted Domain. ..................................................8 VIII.8. Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix of the TSw39 Layer at t = 10,000 Years for Mean Present-Day Infiltration and Instantaneous Release only from Non-Faulted Domain. ..................................................8 VIII.9. Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures of the TSw39 Layer at t = 100,000 Years for Mean Present-Day Infiltration and Instantaneous Release only from Non-Faulted Domain. ...........................................9 VIII.10. Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix of the TSw39 Layer at t = 100,000 Years for Mean Present-Day Infiltration and Instantaneous Release only from Non-Faulted Domain. ..................................................9 VIII.11. Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures immediately above the groundwater table at t = 10 Years for Mean Present-Day Infiltration and Instantaneous Release only from Non-Faulted Domain. .......................10 VIII.12. Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix immediately above the groundwater at t = 10 Years for Mean Present-Day Infiltration and Instantaneous Release only from Non-Faulted Domain. ................................................10 VIII.13. Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures immediately above the groundwater at t = 100 Years for Mean Present-Day Infiltration and Instantaneous Release only from Non-Faulted Domain. .......................11 Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VIII-4 November 2003 FIGURES Page VIII.14. Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix immediately above the groundwater at t = 100 Years for Mean Present-Day Infiltration and Instantaneous Release only from Non-Faulted Domain. ................................................11 VIII.15. Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures immediately above the groundwater at t = 1,000 Years for Mean Present-Day Infiltration and Instantaneous Release only from Non-Faulted Domain. .......................12 VIII.16. Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix immediately above the groundwater at t = 1,000 Years for Mean Present-Day Infiltration and Instantaneous Release only from Non-Faulted Domain. ................................................12 VIII.17. Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures immediately above the groundwater at t = 10,000 Years for Mean Present-Day Infiltration and Instantaneous Release only from Non-Faulted Domain. .......................13 VIII.18. Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix immediately above the groundwater at t = 10,000 Years for Mean Present-Day Infiltration and Instantaneous Release only from Non-Faulted Domain. ................................................13 VIII.19. Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures immediately above the groundwater at t = 100,000 Years for Mean Present-Day Infiltration and Instantaneous Release only from Non-Faulted Domain. .......................14 VIII.20. Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix immediately above the groundwater at t = 100,000 Years for Mean Present-Day Infiltration and Instantaneous Release only from Non-Faulted Domain. .........................................14 Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VIII-5 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-02 9.50E-03 9.00E-03 8.50E-03 8.00E-03 7.50E-03 7.00E-03 6.50E-03 6.00E-03 5.50E-03 5.00E-03 4.50E-03 4.00E-03 3.50E-03 3.00E-03 2.50E-03 2.00E-03 1.50E-03 1.00E-03 5.00E-04 0.00E+00 Relative Mass Fraction Tc-TSw-NF-FL-10 Output-DTN: LB0307MR0060R1.002 Figure VIII.1. Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures of the TSw39 Layer at t = 10 Years for Mean Present-Day Infiltration and Instantaneous Release Only from Non-Faulted Domain Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-02 9.50E-03 9.00E-03 8.50E-03 8.00E-03 7.50E-03 7.00E-03 6.50E-03 6.00E-03 5.50E-03 5.00E-03 4.50E-03 4.00E-03 3.50E-03 3.00E-03 2.50E-03 2.00E-03 1.50E-03 1.00E-03 5.00E-04 0.00E+00 Relative Mass Fraction Tc-TSw-NF-ML-10 Output-DTN: LB0307MR0060R1.002 Figure VIII.2. Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix of the TSw39 Layer at t = 10 Years for Mean Present-Day Infiltration and Instantaneous Release Only from Non-Faulted Domain Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VIII-6 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-03 9.50E-04 9.00E-04 8.50E-04 8.00E-04 7.50E-04 7.00E-04 6.50E-04 6.00E-04 5.50E-04 5.00E-04 4.50E-04 4.00E-04 3.50E-04 3.00E-04 2.50E-04 2.00E-04 1.50E-04 1.00E-04 5.00E-05 0.00E+00 Fraction Tc-TSw-NF-FL-100 Relative Mass Output-DTN: LB0307MR0060R1.002 Figure VIII.3. Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures of the TSw39 Layer at t = 100 Years for Mean Present-Day Infiltration and Instantaneous Release Only from Non-Faulted Domain Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-03 9.50E-04 9.00E-04 8.50E-04 8.00E-04 7.50E-04 7.00E-04 6.50E-04 6.00E-04 5.50E-04 5.00E-04 4.50E-04 4.00E-04 3.50E-04 3.00E-04 2.50E-04 2.00E-04 1.50E-04 1.00E-04 5.00E-05 0.00E+00 Fraction Tc-TSw-NF-ML-100 Relative Mass Output-DTN: LB0307MR0060R1.002 Figure VIII.4. Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix of the TSw39 Layer at t = 100 Years for Mean Present-Day Infiltration and Instantaneous Release Only from Non-Faulted Domain Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VIII-7 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-04 9.50E-05 9.00E-05 8.50E-05 8.00E-05 7.50E-05 7.00E-05 6.50E-05 6.00E-05 5.50E-05 5.00E-05 4.50E-05 4.00E-05 3.50E-05 3.00E-05 2.50E-05 2.00E-05 1.50E-05 1.00E-05 5.00E-06 0.00E+00 Fraction Tc-TSw-NF-FL-1000 Relative Mass Output-DTN: LB0307MR0060R1.002 Figure VIII.5. Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures of the TSw39 Layer at t = 1,000 Years for Mean Present-Day Infiltration and Instantaneous Release Only from Non-Faulted Domain Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-04 9.50E-05 9.00E-05 8.50E-05 8.00E-05 7.50E-05 7.00E-05 6.50E-05 6.00E-05 5.50E-05 5.00E-05 4.50E-05 4.00E-05 3.50E-05 3.00E-05 2.50E-05 2.00E-05 1.50E-05 1.00E-05 5.00E-06 0.00E+00 Relative Mass Fraction Tc-TSw-NF-ML-1000 Output-DTN: LB0307MR0060R1.002 Figure VIII.6. Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix of the TSw39 Layer at t = 1,000 Years for Mean Present-Day Infiltration and Instantaneous Release Only from Non-Faulted Domain Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VIII-8 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-05 9.50E-06 9.00E-06 8.50E-06 8.00E-06 7.50E-06 7.00E-06 6.50E-06 6.00E-06 5.50E-06 5.00E-06 4.50E-06 4.00E-06 3.50E-06 3.00E-06 2.50E-06 2.00E-06 1.50E-06 1.00E-06 5.00E-07 0.00E+00 Relative Mass Fraction Tc-TSw-NF-FL-10,000 Output-DTN: LB0307MR0060R1.002 Figure VIII.7. Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures of the TSw39 Layer at t = 10,000 Years for Mean Present-Day Infiltration and Instantaneous Release Only from Non-Faulted Domain Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-05 9.50E-06 9.00E-06 8.50E-06 8.00E-06 7.50E-06 7.00E-06 6.50E-06 6.00E-06 5.50E-06 5.00E-06 4.50E-06 4.00E-06 3.50E-06 3.00E-06 2.50E-06 2.00E-06 1.50E-06 1.00E-06 5.00E-07 0.00E+00 Fraction Tc-TSw-NF-ML-10,000 Relative Mass Output-DTN: LB0307MR0060R1.002 Figure VIII.8. Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix of the TSw39 Layer at t = 10,000 Years for Mean Present-Day Infiltration and Instantaneous Release Only from Non-Faulted Domain Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VIII-9 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-07 9.50E-08 9.00E-08 8.50E-08 8.00E-08 7.50E-08 7.00E-08 6.50E-08 6.00E-08 5.50E-08 5.00E-08 4.50E-08 4.00E-08 3.50E-08 3.00E-08 2.50E-08 2.00E-08 1.50E-08 1.00E-08 5.00E-09 0.00E+00 Fraction Tc-TSw-NF-FL-100,000 Relative Mass Output-DTN: LB0307MR0060R1.002 Figure VIII.9. Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures of the TSw39 Layer at t = 100,000 Years for Mean Present-Day Infiltration and Instantaneous Release Only from Non-Faulted Domain Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-06 9.50E-07 9.00E-07 8.50E-07 8.00E-07 7.50E-07 7.00E-07 6.50E-07 6.00E-07 5.50E-07 5.00E-07 4.50E-07 4.00E-07 3.50E-07 3.00E-07 2.50E-07 2.00E-07 1.50E-07 1.00E-07 5.00E-08 0.00E+00 Relative Mass Fraction Tc-TSw-NF-ML-100,000 Output-DTN: LB0307MR0060R1.002 Figure VIII.10. Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix of the TSw39 Layer at t = 100,000 Years for Mean Present-Day Infiltration and Instantaneous Release Only from Non-Faulted Domain Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VIII-10 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-03 9.50E-04 9.00E-04 8.50E-04 8.00E-04 7.50E-04 7.00E-04 6.50E-04 6.00E-04 5.50E-04 5.00E-04 4.50E-04 4.00E-04 3.50E-04 3.00E-04 2.50E-04 2.00E-04 1.50E-04 1.00E-04 5.00E-05 0.00E+00 Relative Mass Fraction Tc-WT-NF-FL-10 Output-DTN: LB0307MR0060R1.002 Figure VIII.11. Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures immediately above the groundwater table at t = 10 Years for Mean Present-Day Infiltration and Instantaneous Release Only from Non-Faulted Domain Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-05 9.50E-06 9.00E-06 8.50E-06 8.00E-06 7.50E-06 7.00E-06 6.50E-06 6.00E-06 5.50E-06 5.00E-06 4.50E-06 4.00E-06 3.50E-06 3.00E-06 2.50E-06 2.00E-06 1.50E-06 1.00E-06 5.00E-07 0.00E+00 Relative Mass Fraction Tc-WT-NF-ML-10 Output-DTN: LB0307MR0060R1.002 Figure VIII.12. Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix immediately above the groundwater at t = 10 Years for Mean Present-Day Infiltration and Instantaneous Release Only from Non-Faulted Domain Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VIII-11 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-04 9.50E-05 9.00E-05 8.50E-05 8.00E-05 7.50E-05 7.00E-05 6.50E-05 6.00E-05 5.50E-05 5.00E-05 4.50E-05 4.00E-05 3.50E-05 3.00E-05 2.50E-05 2.00E-05 1.50E-05 1.00E-05 5.00E-06 0.00E+00 Relative Mass Fraction Tc-WT-NF-FL-100 Output-DTN: LB0307MR0060R1.002 Figure VIII.13. Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures immediately above the groundwater at t = 100 Years for Mean Present-Day Infiltration and Instantaneous Release Only from Non-Faulted Domain Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-05 9.50E-06 9.00E-06 8.50E-06 8.00E-06 7.50E-06 7.00E-06 6.50E-06 6.00E-06 5.50E-06 5.00E-06 4.50E-06 4.00E-06 3.50E-06 3.00E-06 2.50E-06 2.00E-06 1.50E-06 1.00E-06 5.00E-07 0.00E+00 Relative Mass Fraction Tc-WT-NF-ML-100 Output-DTN: LB0307MR0060R1.002 Figure VIII.14. Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix immediately above the groundwater at t = 100 Years for Mean Present-Day Infiltration and Instantaneous Release Only from Non-Faulted Domain Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VIII-12 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-05 9.50E-06 9.00E-06 8.50E-06 8.00E-06 7.50E-06 7.00E-06 6.50E-06 6.00E-06 5.50E-06 5.00E-06 4.50E-06 4.00E-06 3.50E-06 3.00E-06 2.50E-06 2.00E-06 1.50E-06 1.00E-06 5.00E-07 0.00E+00 Fraction Tc-WT-NF-FL-1,000 Relative Mass Output-DTN: LB0307MR0060R1.002 Figure VIII.15. Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures immediately above the groundwater at t = 1,000 Years for Mean Present-Day Infiltration and Instantaneous Release Only from Non-Faulted Domain Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-05 9.50E-06 9.00E-06 8.50E-06 8.00E-06 7.50E-06 7.00E-06 6.50E-06 6.00E-06 5.50E-06 5.00E-06 4.50E-06 4.00E-06 3.50E-06 3.00E-06 2.50E-06 2.00E-06 1.50E-06 1.00E-06 5.00E-07 0.00E+00 Relative Mass Fraction Tc-WT-NF-ML-1,000 Output-DTN: LB0307MR0060R1.002 Figure VIII.16. Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix immediately above the groundwater at t = 1,000 Years for Mean Present-Day Infiltration and Instantaneous Release Only from Non-Faulted Domain Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VIII-13 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-05 9.50E-06 9.00E-06 8.50E-06 8.00E-06 7.50E-06 7.00E-06 6.50E-06 6.00E-06 5.50E-06 5.00E-06 4.50E-06 4.00E-06 3.50E-06 3.00E-06 2.50E-06 2.00E-06 1.50E-06 1.00E-06 5.00E-07 0.00E+00 Relative Mass Fraction Tc-WT-NF-FL-10,000 Output-DTN: LB0307MR0060R1.002 Figure VIII.17. Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures immediately above the groundwater at t = 10,000 Years for Mean Present-Day Infiltration and Instantaneous Release Only from Non-Faulted Domain Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-05 9.50E-06 9.00E-06 8.50E-06 8.00E-06 7.50E-06 7.00E-06 6.50E-06 6.00E-06 5.50E-06 5.00E-06 4.50E-06 4.00E-06 3.50E-06 3.00E-06 2.50E-06 2.00E-06 1.50E-06 1.00E-06 5.00E-07 0.00E+00 Relative Mass Fraction Tc-WT-NF-ML-10,000 Output-DTN: LB0307MR0060R1.002 Figure VIII.18. Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix immediately above the groundwater at t = 10,000 Years for Mean Present-Day Infiltration and Instantaneous Release Only from Non-Faulted Domain Radionuclide Transport Models Under Ambient Conditions U0060 MDL-NBS-HS-000008 REV 01 Attachment VIII-14 November 2003 Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-06 9.50E-07 9.00E-07 8.50E-07 8.00E-07 7.50E-07 7.00E-07 6.50E-07 6.00E-07 5.50E-07 5.00E-07 4.50E-07 4.00E-07 3.50E-07 3.00E-07 2.50E-07 2.00E-07 1.50E-07 1.00E-07 5.00E-08 0.00E+00 Fraction Tc-WT-NF-FL-100,000 Relative Mass Output-DTN: LB0307MR0060R1.002 Figure VIII.19. Distribution of the Relative Mass Fraction XR of 99Tc in the Fractures immediately above the groundwater at t = 100,000 Years for Mean Present-Day Infiltration and Instantaneous Release Only from Non-Faulted Domain Nevada Coordinate E-W (m) Nevada Coordinate N-S (m) 168000 170000 172000 174000 230000 232000 234000 236000 238000 1.00E-05 9.50E-06 9.00E-06 8.50E-06 8.00E-06 7.50E-06 7.00E-06 6.50E-06 6.00E-06 5.50E-06 5.00E-06 4.50E-06 4.00E-06 3.50E-06 3.00E-06 2.50E-06 2.00E-06 1.50E-06 1.00E-06 5.00E-07 0.00E+00 Relative Mass Fraction Tc-WT-NF-ML-100,000 Output-DTN: LB0307MR0060R1.002 Figure VIII.20. Distribution of the Relative Mass Fraction XR of 99Tc in the Matrix immediately above the groundwater at t = 100,000 Years for Mean Present-Day Infiltration and Instantaneous Release Only from Non-Faulted Domain