DSNF and Other Waste Form Degradation Abstraction Rev 01, ICN 00

ANL-WIS-MD-000004

October 2000

1. PURPOSE 
The purpose of this analysis/model report (AMR) is to select and/or abstract conservative 
degradation models for DOE- (U.S. Department of Energy) owned spent nuclear fuel (DSNF) 
and the immobilized ceramic plutonium (Pu) disposition waste forms for application in the 
proposed monitored geologic repository (MGR) postclosure Total System Performance 
Assessment (TSPA). Application of the degradation models abstracted herein for purposes other 
than TSPA should take into consideration the fact that they are, in general, very conservative. 
This activity is covered by item 6 of the Development Plan for Waste Package Materials 
Department Analysis and Modeling Reports Supporting the Waste Form PMR (CRWMS M&O 
2000e). Using these models, the forward reaction rate for the mobilization of radionuclides, as 
solutes or colloids, away from the waste form/water interface by contact with repository 
groundwater can then be calculated. This forward reaction rate generally consists of the 
dissolution reaction at the surface of spent nuclear fuel (SNF) in contact with water, but the 
degradation models, in some cases, may also include and account for the physical disintegration 
of the SNF matrix. The models do not, however, account for retardation, precipitation, or 
inhibition of the migration of the mobilized radionuclides in the engineered barrier system 
(EBS). These models are based on the assumption that all components of the DSNF waste form 
are released congruently with the degradation of the matrix. The rate of release would be related 
to the rate of waste-form degradation as follows: 
Ri = (degradation rate in mg/m2-d) · (total surface area of SNF exposed to water in m2 ) · (mass 
fraction of species i) 
Several hundred distinct types of DSNF may potentially be stored in the MGR (DOE 1999, 
Appendix D). Therefore, each type cannot be examined viably for its effect on either repository 
preclosure design basis event (DBE) safety analyses or for postclosure TSPA. To enable 
analyses of a limited number of potential DSNF types to represent, or bound, the behavior of all 
DSNF disposed in the repository, the U.S. Department of Energy Office of Civilian Radioactive 
Waste Management (DOE OCRWM) and the National Spent Nuclear Fuel Program (NSNFP) 
collaborated to identify the DSNF groups. These groups are individual categories of spent fuel 
into one of which all DSNF types would fall for DBE and/or TSPA analysis purposes. The 
description of, and justification for, these spent fuel groupings for repository criticality, DBE, 
and TSPA analysis purposes are in the NSNFP report, DOE Spent Nuclear Fuel Grouping in 
Support of Criticality, DBE, and TSPA-LA (DOE 2000b). Also, the NSNFP has compiled a 
report that contains characteristics of the DSNF groups related to their postclosure performance 
and suggested models for waste-form dissolution: DOE Spent Nuclear Fuel Information in 
Support of TSPA-SR (DOE 1999). 
The DOE SNF groups for total system performance analysis for the Site Recommendation 
(TSPA-SR) contained in the above documents and described in Sections 6.3.1 through 6.3.11 of 
this document (and a typical type of SNF in the group) follow: 
· Group 1–Naval SNF 
· Group 2–Pu/U alloy (Fermi SNF) 
· Group 3–Pu/U carbide (FFTF-TFA SNF) 
· Group 4–MOX and Pu oxide (FFTF-DFA/TDFA SNF) 
· Group 5–Th/U carbide (Fort St. Vrain SNF) 
· Group 6–Th/U oxide (Shippingport light water breeder reactor (LWBR) reflector SNF) 
· Group 7–U-metal (N-Reactor SNF) 
· Group 8–U oxide (Three Mile Island [TMI]-2 core debris) 
· Group 9–Al-based SNF (Foreign Research Reactor [FRR] SNF) 
· Group 10–unknown (miscellaneous SNF) 
· Group 11–U-Zr-Hx (TRIGA SNF). 
It should be noted that this grouping of DSNF types is different than the groupings recommended 
for DBE and criticality analyses in DOE (2000b). Thus, the degradation model for a single fuel 
type recommended herein for the TSPA analyses may not be applicable for that fuel type for 
DBE and/or criticality analyses. 
Parameters and/or characteristics of these waste forms, used as input to the TSPA release rate 
models, may be reasonable and/or bounding assumptions concerning the materials; therefore, 
they need not necessarily be fully qualified to support repository licensing. Dissolution models 
for DSNF waste forms to be used in TSPA-LA (License Application) may also be assumptions 
or reasonable models and may not need to be fully validated. The TSPA analyses that use these 
assumptions as inputs will be appropriately validated and will document the consequences of 
using unqualified data as required by AP-3.10Q, Analyses and Models. The license granted to 
operate the repository would contain technical specifications based on these parameters/models. 
Also, the final recommended models apply to the direct degradation of the waste form only. The 
models do not explicitly cover radionuclide migration retardation or enhancement mechanisms 
such as colloid formation, other than to recommend that the formation/migration dynamics for 
colloid formation from the DSNF be taken as both qualitatively and quantitatively similar to that 
of CSNF and HLW glass. 
If post-licensing analyses are required to demonstrate conformance of individual waste forms to 
waste acceptance criteria (WAC), they may require validated waste form dissolution models. If 
this approach is not taken, the post-licensing analysis will need to demonstrate through such 
activities as bounding analyses that using unqualified data and/or unvalidated models is 
acceptable (i.e., in compliance with AP-3.10Q). 
Surplus plutonium (Pu) will be handled in the DOE complex in two ways: some of it will be 
converted into an immobilized ceramic waste form, and the rest will be converted to mixed oxide 
(MOX) fuel for use in commercial light water reactors. Therefore, a Group 12, representing the 
immobilized ceramic Pu disposition waste form, will be added to the other DSNF groups. This 
waste form will consist of disks of a Pu-containing titania-based ceramic enclosed in stainless 
steel cans that are, in turn, encased in a borosilicate high level waste (HLW) glass matrix 
(CRWMS M&O 1998a). This AMR will not select and/or abstract a degradation model for 
mixed oxide (MOX) SNF as a Pu disposition waste form because this form, if used, would be 
treated as commercial light water reactor (LWR) SNF. Therefore, throughout the remainder of 
this document, MOX will refer to the MOX SNF in Group 4. 
The objective of this document is to use the two DOE SNF group-related studies, published 
analyses, and results of experimental degradation tests to abstract models for the degradation rate

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of the SNF groups. Further, this AMR will recommend using a subset of these models to 
provide bounding TSPA analyses of the DOE SNF and immobilized Pu ceramic disposal waste 
forms. 
The intended usage of the abstracted DSNF and immobilized Pu ceramic waste form models is to 
provide forward-reaction-rates (unmitigated by back-reaction) as input into TSPA analyses. 
Although directly used in analyses that estimate the rate of mobilization of radionuclides, the 
expected insensitivity of the TSPA to the specific degradation rates calculated by the model 
limits the model’s importance to the TSPA (CRWMS M&O 2000f, Section 7, bullets 11-16). 
The models provide a source term for other radionuclide transport/inhibition processes. The 
models derived in this report do this by combining the material degradation models, which 
predict degradation in units of mass of DSNF dissolved per unit of exposed surface area per unit 
time, with conservative estimates of the exposed surface area of the SNF available for water 
contact. By using models for degradation that bound the actual rates for the DSNF and 
conservative estimates of the exposed surface area, assurance is provided that the repository 
performance is bounded by the rates. Should the TSPA indicate in future analyses (due to design 
changes or other unforeseen conditions) that the performance of the MGR is, in fact, sensitive to 
the degradation, the model may require reanalysis to remove excessive conservatism. 
2. QUALITY ASSURANCE 
The Quality Assurance (QA) program (DOE 2000c) applies to this analysis. All types of waste 
packages were classified as Quality Level-1 in Classification of the MGR Uncanistered Spent 
Nuclear Fuel Disposal Container System (CRWMS M&O 1999e, p. 7). This analysis applies to 
all of the waste package designs included in the Monitored Geologic Repository (MGR) 
Classification Analyses. Reference CRWMS M&O (1999e) is cited as an example. The 
development of this analysis is conducted under activity evaluation 1101213FM3 Waste Form 
Analyses & Models - PMR (CRWMS M&O 1999g). The models abstracted in this document 
involved no direct use of experimental data, and, therefore, no electronic data management was 
required. 
3. COMPUTER SOFTWARE AND MODEL USAGE 
No computer software was used in the analysis or abstraction of the degradation models 
discussed in this report. No experimental data or data obtained from calculational models were 
used to produce the models abstracted in this report. The degradation models analyzed herein are 
intended to support radionuclide release source terms in the TSPA analyses performed by the 
Performance Assessment Office (PA) of the CRWMS M&O. 
4. INPUTS 
4.1 DATA AND PARAMETERS 
Data, information, and models for the degradation of DSNF and Pu disposition waste forms were 
obtained from laboratory experiments, DOE reports, NSNFP reports, and OCRWM AMRs.

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· Documents by Thornton (1998a, 1998b) and CRWMS M&O (2000f) provide analyses 
demonstrating that even under extremely conservative assumptions for the degradation 
and/or pyrophoric behavior of the DSNF, the contribution of the DSNF to the postclosure site 
dose is negligible. Input values for the degradation rate used in the sensitivity analyses did 
not require qualification because it is a parametric study that used the maximum conceivable 
degradation rates of the waste form to perform the analysis. This analysis does not require 
qualification because it does not provide direct input to the abstraction of any of the 
degradation models. 
· CRWMS M&O (2000f, Section 6.7.1 and 7, bullets 11-16) provides analyses that show that 
the dose rates at the site boundary resulting from the failure of DSNF-containing waste 
packages are insensitive to the degradation rates of the DSNF waste forms. The Repository 
Integration Program (RIP) code used by the PA for this analysis is qualified software. 
However, the code input values for the degradation rate used in this AMR do not require 
qualification because this is a parametric study using the maximum conceivable degradation 
rates of the waste form to perform the analysis. 
· DOE (2000b) establishes and justifies the DSNF groups to be used for TSPA-SR analysis 
purposes. This report proposes that the degradation behavior of each DSNF group can be 
adequately represented by the behavior of a surrogate SNF type within the group. The 
conclusions of this report are supported by TSPA analyses performed using the RIP TSPA 
code (Thornton 1998a, 1998b; and CRWMS M&O 2000f, Section 6.7.1 and Section 7, 
bullets 11-16) which show that the postcontainment site-boundary dose is insensitive to the 
degradation rate of the DSNF or individual components of the DSNF. This information does 
not require qualification because no parameter from this report is directly used in selecting or 
abstracting degradation models. Appendix D of DOE (2000b) contains the equivalent metric 
tons heavy metal (MTHM) inventory for each of the TSPA-SR DSNF groups. The data used 
in this AMR compares the relative importance of each group in formulating a composite 
degradation model for all DSNF. While the data are required to be qualified for other 
documents supporting the TSPA, they are used here to indicate the relative quantities of the 
DSNF groups only. Therefore, they do not require qualification for use in abstracting group 
dissolution models in this AMR. 
· DOE (1999) contains proposed dissolution models and MTHM inventory for each of the 
DSNF TSPA groups identified in DOE (2000b). The suggested models were evaluated 
and/or abstracted in, or selected for, this AMR. The information in the document related to 
the models may need to be validated to demonstrate conformance to waste acceptance 
criteria (WAC) of specific waste forms for emplacement. Appendix A of the referenced 
DOE report contains information on the surface area for each of the TSPA-SR DSNF groups, 
and Appendix D contains the equivalent MTHM inventory. The data from this report is used 
for comparative purposes to indicate the conservatism inherent in the dissolution models 
analyzed in this AMR and is not directly used in the generation of the N-reactor SNF-based 
dissolution model that is recommended for use for all DSNF (except the naval SNF). The 
data, therefore, does not need to be qualified for its usage herein.

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· Pasupathi (2000) is information included in a report generated by the NSNFP, which contains 
an analysis of the oxidation rates of metallic uranium in dry air, humid air, saturated water 
vapor, and liquid water. The report also provides correlation for the oxidation rates of 
metallic uranium based on literature data for unirradiated/uncorroded uranium metal. The 
correlation provided in the reference forms a partial basis for the dissolution rate expression 
for N-reactor SNF recommended/derived in Section 6.3.7 of this AMR. The data analyzed in 
the NSNFP report is a comprehensive representation of the available literature on uranium 
metal oxidation. The NSNFP report shows that the reference data have been compiled in 
numerous publications and used extensively in the past for the analysis of the behavior of 
metallic uranium. The data and analyses in the report, thereby, provide a comprehensive 
database adequate for the usage both in the NSNFP report and for the purpose of providing a 
basis for the models abstracted in this AMR. For these reasons the correlation derived in the 
report for the oxidation rate of uranium metal in liquid water as a function of temperature 
may be regarded as validated and may be used in Section 6.3.7 as a partial basis for the 
N-reactor SNF dissolution model. 
· Gray and Einziger’s (1998) report contains the results of flow-through dissolution tests on 
samples of N-reactor SNF. The model described in their report is used here to abstract a 
model for the degradation of U-metal DSNF. This model, in turn, will be used to abstract 
degradation models for other DSNF groups; therefore, validation to support the 
demonstration of conformance to waste acceptance criteria of the specific waste forms for 
repository emplacement may be necessary. The data/information is currently unqualified 
pending AP-3.15Q, Managing Technical Product Inputs. and/or AP-SIII.2Q, Qualification of 
Unqualified Data and the Documentation of Rationale for Accepted Data, evaluation 
analysis. 
· Wiersma and Mickalonis’ (1998) report contains the results of dissolution testing conducted 
on samples of aluminum-based SNF from the Savannah River Site (SRS). This AMR used 
their data to select a model for the degradation of aluminum-based SNF. The new model, in 
turn, will be used to show only the low-dissolution kinetics comparison to uranium metalbased 
SNF. Because of this restricted use, qualification is not necessary for the purposes of 
this AMR. It should be noted, however, that the data could require qualification to support 
the demonstration of conformance to WAC of the aluminum-based DSNF waste forms for 
emplacement. 
· CRWMS M&O (2000b) is an AMR generated under AP-3.10Q that abstracts models for the 
formation and transport of radionuclides in colloidal form from CSNF and HLW glass waste 
packages. Although no direct information is provided in this AMR concerning the formation 
or transport of colloids from DSNF, the case is made that they may be treated similarly as the 
colloids from the CSNF. This information is used in Section 5 to support the assumption that 
colloid formation from DSNF are not significantly different than for CSNF and HLW glass, 
and, thus, do not need to be specifically addressed in this AMR and in Section 7.2.1 to justify 
using the CSNF and HLW glass colloid models in CRWMS M&O (2000b) as surrogates for 
DSNF colloid behavior.

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· Batt (1999) and Hurt (2000) are National Spent Nuclear Fuel Program letter reports that 
provide information concerning the formation of colloids from the degradation of metallic 
uranium-based (N-reactor) and MOX DSNF waste forms. This information is used in 
Section 5 to support the assumption that colloid formation from DSNF is not significantly 
different than for CSNF and HLW glass and, thus, does not need to be specifically addressed 
in this AMR and in Section 7.2.1 to justify using the CSNF and HLW glass colloid models as 
surrogates for the DSNF colloid behavior. As DSNF colloid testing continues at ANL 
(Mertz 2000), the results will be used as confirmatory data for the assumptions and usage in 
this AMR, or, if appropriate, revisions to these assumptions and conclusions will be made. 
· CRWMS M&O (2000c) is an AMR generated under AP-3.10Q that abstracts models for the 
dissolution rate of HLW glass under repository groundwater exposure conditions. The 
models in CRWMS M&O (2000c) are not yet validated pending completion of the AP-3.10Q 
process. However, the abstracted model is used here to represent the degradation of the 
immobilized ceramic Pu disposition waste form. 
· CRWMS M&O (1998b) contains abstracted models for waste-form materials. In some cases, 
these models were a source of the models used for comparison with the models abstracted 
here. The individual models do not require validation because it is not expected that they 
will be used for licensing. 
· CRWMS M&O (2000a) is an AMR generated under AP-3.10Q that abstracts models for the 
dissolution of uranium dioxide-based commercial LWR SNF (CSNF) under repository 
groundwater exposure conditions. The model described in CRWMS M&O (2000a) is used to 
abstract a degradation model for uranium-oxide based DSNF. The model has completed the 
AP-3.10Q process and is considered validated. 
· CRWMS M&O (1998a) contains a description of the immobilized ceramic Pu disposal waste 
form and describes how it is incorporated into a waste canister. The information is used in 
this AMR to justify the types of dissolution models, a model for glass dissolution and a 
model for ceramic dissolution, used respectively for the conservative and best-estimate 
models for the Pu waste form addressed in Section 6.3.12. The information is not used in 
this AMR to directly formulate the degradation model for the immobilized ceramic Pu 
disposition waste form; therefore, the model does not need to be qualified for this purpose. 
· CRWMS M&O (2000g, Section 2.2.2) indicates that the Pu-ceramic corrosion rate will be 
considerably less than that of the HLW glass. This is further supported by analyses in 
CRWMS M&O (2000c), which demonstrate that the HLW glass rates invoked in CRWMS 
M&O (2000g) are consistent with the HLW glass model. This is also supported by the 
indication in the TSPA-VA models and the initial corrosion data for immobilized Pu ceramic 
in Shaw (1999) that the dissolution rate of borosilicate glass is significantly greater than that 
of titanate-based crystalline material. In the immobilized ceramic waste form, the Pu is to be 
encapsulated in disks of ceramic matrix similar to the chemical composition of synroc, a 
titanate-based pyrochlore. The ceramic disks would then be enclosed in stainless steel cans 
that would then be encased in a borosilicate glass monolith inside the HLW waste canister 
(CRWMS M&O 1998a, Section 2.2). Water, therefore, would have to first penetrate the

 13  
stainless steel HLW canister, the glass matrix, and the stainless steel can before it could 
contact the Pu-containing crystalline material. Accounting for the dissolution of the HLW 
glass in the overall TSPA for the immobilized Pu ceramic, therefore, may not be necessary 
considering the relative durability of the ceramic in which the Pu is actually encased. 
However, the similarity of the glass composition of the Pu-ceramic disposition waste form to 
the HLW glass compositions may need to be verified if the degradation of the glass in the 
canister is part of the TSPA analysis. 
· Shaw (1999, Section 6.1) is a preliminary report from the plutonium immobilization project 
at Lawrence Livermore National Laboratory (LLNL). Shaw’s report contains a 
recommended correlation for the rate of dissolution of the currently envisioned Pu ceramic 
waste form that is recommended as the best estimate model for this type of waste form. 
Data/models for the immobilized Pu ceramic waste form may need to be qualified to support 
the demonstration of conformance to WAC of this waste form for emplacement. 
· CRWMS M&O (2000g, Section 2.2, Table 2-10) contains a comparison of HLW glass and 
Pu-ceramic corrosion rates, which shows that the HLW corrosion rates bound the Pu-ceramic 
rates. 
· LMITCO (Lockheed Martin Idaho Technologies Company) (1997, Section 2.1.8) contains 
information on the dissolution behavior of unirradiated U-Zr-Hx fuel at high-temperature 
conditions. The data are not required to be qualified because they are not used as direct 
inputs into the abstraction model. Information from LMITCO (1997) is used here to 
qualitatively indicate the very low dissolution rate of U-Zr-Hx compared to the other DSNF 
groups. Neither is data qualification required for use in this AMR as the data are used for 
comparison and corroboration purposes only. However, data/models for the degradation of 
the U-Zr-Hx waste form may need to be qualified to support the demonstration of 
conformance to WAC of this waste form for emplacement. 
4.2 CRITERIA 
The models selected and/or abstracted in this AMR are not based on project-level criteria. The 
criteria for the recommended models are that they provide appropriate bounding values for the 
DSNF for use in the TSPA. 
4.3 CODES AND STANDARDS 
ASTM Standard C 1174-97, Standard Practice for Prediction of the Long-Term Behavior of 
Materials, Including Waste Forms, Used in Engineered Barrier Systems (EBS) for Geologic 
Disposal of High-Level Radioactive Waste, is used to support the degradation model 
development methodology, categorize the models developed with respect to their usage for longterm 
TSPA, and relate the information/data used to develop the model to the requirements of the 
standard.

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5. ASSUMPTIONS 
The NSNFP is continuing to support testing related to the dissolution/degradation behavior of 
DSNF in the MGR per the NSNFP Release-Rate Test Program Plan (DOE 1998). The testing 
performed under this program supports the assumptions used in this analysis. The naval SNF 
disposition program will provide appropriate confirmation of Assumptions 5.3 and 5.4. Testing 
in progress and to be performed will be evaluated with respect to continuing verification of the 
assumptions listed below. Assumptions used for the analyses contained herein and those that 
may require verification through the continuing dissolution testing supported by the NSNFP and 
PIP follow: 
5.1 Degradation of the DSNF waste form is congruent; all components of the waste form 
matrix are released at the same rate as the matrix material. This assumption is based on the 
observations of the condition of the uranium dioxide-based commercial light water reactor 
(LWR) SNF and the N-reactor metallic uranium-based SNF degradation test samples in the 
Pacific Northwest National Laboratory (PNNL) tests (Gray and Einziger 1998). 
This assumption is used to justify the use of the degradation models examined in Section 6 
for the congruent release of radionuclides from the SNF. The uranium dioxide-based LWR 
SNF showed congruent dissolution in flow-through testing (CRWMS M&O 2000a) (Gray 
and Einziger 1998, Section 3.2). Some limited testing of MOX and aluminum-based SNF 
supports the congruent dissolution assumption but is not extensive enough to be conclusive. 
The N-reactor SNF testing has not yet involved enough of the somewhat heterogeneous Nreactor 
samples to firmly establish congruent radionuclide release for the U-metal SNF. 
Additionally, the immobilized ceramic Pu waste form has shown indications of some 
incongruent release behavior in as yet incomplete testing (Bourcier 1999). Therefore, this 
assumption needs to be verified as conservative to support licensing of the DSNF waste 
forms for emplacement, unless such verification can be demonstrated to be unnecessary. 
5.2 N-reactor SNF is an adequate conservative surrogate for the degradation rate of the DSNF. 
N-reactor SNF comprises ~84% by MTHM of the total quantity of DSNF other than the 
naval SNF. This assumption is based on the observation in preliminary studies (Gray and 
Einziger 1998) that the rate of dissolution of the metallic uranium-based N-reactor SNF has 
significantly exceeded that of the uranium oxide-based commercial SNF that will comprise 
most of the total SNF to be emplaced in the MGR. Also, examination of the literature 
available for other waste forms (DOE 1999) indicates that the degradation rate of N-reactor 
fuel generally exceeds that of the other forms. This assumption is used in Section 6.4 to 
support the judgement that even if an individual type of DSNF for which the inventory is 
low has a dissolution rate greater than N-reactor SNF, the contribution to the total DSNF 
release is negligible. 
5.3 Commercial LWR SNF will be used as the surrogate for naval SNF in repository 
performance analyses. The basis for this assumption is the robust design of the naval SNF. 
Expected releases from naval SNF waste packages were provided in Mowbray (2000). 
Because of its robust design, the radionuclide releases from naval SNF waste packages are 
considerably less than releases from commercial LWR SNF waste packages; accordingly, 
this assumption is conservative. Naval SNF represents approximately one-tenth of one

 15  
percent of the spent fuel MTHM inventory. The information contained in Mowbray (2000) 
is qualified data. This assumption is used in Section 6 to support the use of LWR SNF as a 
conservative surrogate for the naval SNF. 
5.4 The formation and transport of colloids (resulting from the degradation of the DSNF and 
immobilized ceramic Pu waste forms) is similar, both qualitatively and quantitatively, to 
colloids formed from the degradation of CSNF and HLW glass forms. Thus, the colloid 
models developed in CRWMS M&O (2000b, Section 6.2) may be applied to the DSNF and 
immobilized ceramic Pu waste forms for the purposes of the TSPA-SR. The bases for this 
assumption are (1) the DSNF represents a small fraction of the total inventory of CSNF and 
HLW glass in the repository, and so the colloid contribution to the boundary dose will be 
dominated by these two waste forms (CRWMS M&O 2000b); (2) the DSNF will be 
codisposed in waste packages with HLW glass containers (CRWMS M&O 2000d, 
Attachment 1, Section 3); and (3) the borosilicate glass matrix for the Pu waste form is 
compositionally very similar to the HLW glass. It is recognized in making this assumption 
that some limited preliminary information currently available concerning colloid formation 
from the degradation of N-reactor SNF (the largest component by weight of the DSNF 
inventory) indicates there may be significantly more colloids formed from this waste form 
than the CSNF and/or HLW glass (Mertz 2000, Section 1.0). Testing is being conducted at 
Argonne National Laboratory (Mertz 2000) to characterize colloid formation from the 
degradation of N-reactor and MOX SNF, and this information will be used as it becomes 
available to support either the validation or the modification of this assumption. This 
assumption is used in Sections 6.4 and 7.2.1 to justify recommendations for addressing the 
immobilized ceramic Pu and DSNF colloids by methods similar to those given in CRWMS 
M&O (2000b, Section 6.2).

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6. ANALYSIS/MODEL 
6.1 MODEL REQUIREMENTS 
The purpose of the analyses contained in this document is to provide upper-limit, conservative, 
and best-estimate models for the degradation of DSNF and the immobilized ceramic Pu 
disposition waste form under environmental conditions appropriate to the MGR. Since 
degradation of the DSNF is not identified as a principal factor but is related to the other factors 
Defense Spent Nuclear Fuel, Navy Fuel, and Pu Disposition Waste Form Performance, and 
Colloid Associated Radionuclide Performance in the TSPA analyses, the relative importance of 
the model is medium. It also has been indicated in current TSPA analyses (CRWMS M&O 
2000f) that repository performance is insensitive to the DSNF degradation rates. To develop 
these models, a surrogate SNF is selected to represent the degradation behavior of each group for 
each of these three kinds of models. As discussed below and shown in Table 1b, the degradation 
model surrogate selected for any given group/model combination will not necessarily be a SNF 
type within the group. Neither will the same surrogate SNF necessarily be chosen for each kind 
of degradation model within a given group. 
An examination of the available experimental or analytical degradation rate behavior data for the 
DSNF groups in this report shows that, in general, insufficient data exists to formulate or abstract 
separate derived models for each group. Because of this, the limited degradation behavior data 
for each group are generally compared with the degradation behavior of ceramic, uranium 
dioxide SNF, or uranium metal SNF for which more extensive data are available. 
The reason for providing the three types of models (upper-limit, conservative, and best-estimate) 
is to provide the user of the models in the M&O PA office appropriate flexibility in the 
application of the degradation model(s) to any particular postclosure performance scenario. In 
some cases, a bounding, or very conservative, model would be appropriate, and such models 
would require less rigorous experimental data to support them. To the extent DOE can use 
margin and conservatism to support licensing of a given DSNF waste form, the data applicable to 
that waste form may not need to be qualified. In cases where best-estimate models are 
appropriate, more rigorous data support and qualification would be required. 
The model selection/abstraction in this analysis incorporates the model logic shown in Figure 3 
of ASTM Standard C 1174-97, Standard Practice for Prediction of the Long-Term Behavior of 
Materials, Including Waste Forms, Used in Engineered Barrier Systems (EBS) for Geologic 
Disposal of High-Level Radioactive Waste. Validation of the degradation models used or 
abstracted in this AMR may not be necessary if the TSPA or other performance analyses are 
shown to be very insensitive to the degradation rate of the waste form or if the use of a highly 
conservative model, such as instantaneous release model, still results in acceptable performance 
of the MGR. If validation of the release model selected were required, it would be performed 
using confirmatory testing per the requirements of ASTM C 1174-97. In this standard the types 
of materials degradation models are defined as follows: 
Mechanistic Model–A model derived from fundamental laws governing the behavior of matter 
and energy. It corresponds to one end of a spectrum of models with varying degrees of 
empiricism.

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Semi-empirical Model–A model based partially on one or more mechanisms and partially on 
data from experiments. 
Empirical Model–A model based only on observations or data from experiments, without 
regard to mechanism or theory. 
Bounding Model–A model that yields values for dependent variables or effects that are expected 
to be either always greater than or always less than those expected for the variables or effects to 
be bounded. 
The types of models described in the ASTM standard (mechanistic, semi-empirical, empirical, 
bounding) are classified as such by their degree of mechanistic representation of the mode of 
degradation; whereas, the models described in this AMR analysis (best-estimate, conservative, 
upper-limit) are classified by their degree of conservatism in predicting degradation rates. There 
is, however, a general correlation between the ASTM model logic and that of this analysis. In 
this analysis the upper-limit degradation models would generally correspond to bounding-toempirical 
models per the ASTM concept; the conservative degradation models would generally 
correspond to empirical-to-semi-empirical models per the ASTM concept; and the best-estimate 
degradation models would correspond to semi-empirical-to-mechanistic models per the ASTM 
concept. The degradation models examined in this analysis will also be identified by their 
correspondence to the ASTM concept (see Table 1B). 
6.2 TYPES OF DEGRADATION MODELS FOR TSPA 
These analyses will provide three types of models for TSPA application purposes: best-estimate, 
conservative, and upper-limit. 
Upper-limit Degradation Models-The upper-limit degradation model provides the most 
conservative estimate of dissolution rate to be used in any postclosure waste package or EBS 
performance case. The upper-limit model generally gives unrealistically high estimates of the 
degradation rate of the waste forms. An upper-limit model may be appropriate in cases where 
the results of the TSPA or other performance analyses are either very insensitive to the 
degradation rate of the waste form or where the use of such a model still results in acceptable 
performance of the MGR. An upper-limit model uses dissolution data, or models abstracted 
from experimental data, only in that such data clearly shows that the bounding model predicts 
release rates always well in excess of actual dissolution rates. 
An example of a situation in which an upper-limit degradation model would be appropriate is 
that of the postclosure TSPA analysis of the N-reactor SNF disposal. Preliminary TSPA analysis 
of the DSNF in the MGR (Thornton 1998a, 1998b) had indicated that for large incremental time 
steps (> 1,000 years) in the TSPA analyses, the boundary dose resulting from the failure of waste 
packages containing DSNF is very insensitive to the release rates from the DSNF waste form. 
Subsequent more detailed analyses (CRWMS M&O 2000f) have supported the conclusion that 
even complete dissolution of the waste form still results in calculated site boundary doses below 
the required limits. The upper-limit release rate for the DSNF in this case would be the 
assumption of complete release within a TSPA (i.e., RIP code) analysis time step or 
instantaneous release.

 18  
Conservative Degradation Models-The conservative degradation models provide an estimate 
of dissolution rate that reflects the higher rate end of dissolution data available. A conservative 
model for waste-form degradation would be appropriate in cases where the dissolution database, 
from which the model was developed, showed wide data spreads or sensitive dependency on 
waste-form characteristics that could not be definitively controlled, described, or determined for 
the emplacement condition. A conservative model would be expected to encompass the 
dissolution kinetics of all SNF types within a DSNF TSPA group. 
An example of a situation in which a conservative degradation model would be appropriate 
might be that of the postclosure TSPA of HLW glass-containing waste packages. Preliminary 
TSPA analyses showed that the release expected from the HLW glass-containing waste packages 
were not as significant a contributor to boundary dose as the release from waste packages 
containing commercial LWR spent fuel. By using a release model that represents the release rate 
from HLW glass compositions that dissolve at the highest rates and applying these rates to the 
entire HLW glass inventory, the highest plausible release rates could be used while boundary 
doses could still be shown to be within regulatory limits. 
Best-estimate Degradation Models-Best-estimate models would be appropriate when the use 
of overly conservative formulations in the TSPA produce results that indicate marginal MGR 
performance. Best-estimate models might also be used in analyses not directly related to TSPA, 
such as parametric studies, waste package design support, or other such analyses where full 
validation of the model might not be required. Best-estimate models would generally require the 
most extensive experimental data to support validation. Moreover, in many cases the best 
estimate model itself is the result of a conservative analysis of the experimental data (e.g., the Pu 
ceramic model derived in Shaw (1999, Section 6.1). A best-estimate model would be used when 
sufficient dissolution data exists to abstract one, and the characteristics of the waste form can be 
shown to correspond to the characteristics of the materials that provided the dissolution database. 
The modes of degradation of interest are oxidation, corrosion, and/or dissolution of the waste 
forms in air, water vapor, liquid water, and associated radionuclide release rates. These modes 
are to be encompassed under the general heading of degradation model. The basis for the 
analysis will be data/analyses for those parameters pertaining to the degradation of N-reactor 
spent fuel, LWR SNF, and the immobilized ceramic Pu disposal waste form. Section 6.3 will 
describe the models suggested for each DSNF group. 
6.3 MODEL SELECTION/ABSTRACTION 
This section describes the upper-limit, conservative, and best-estimate dissolution models 
suggested for use in licensing each DSNF group, and a summary of these models is in Table 1. 
This table shows the three types of models recommended for each DSNF group: (1) the 
immobilized ceramic Pu disposition form, the surrogate material for which the model was 
developed, (2) the literature/document source reference for the model, and (3) the ASTM C 
1174-97 (1997) classification of the model. A discussion of the development of the individual 
models for each fuel group follows.

 19  
The Program will determine for each model whether that model needs to be validated to support 
the demonstration of conformance to WAC of individual waste forms or whether sufficient 
margin/conservatism exists to make model validation unnecessary. 
In some cases, adjustments of one or more orders of magnitude to degradation rates are 
recommended. These adjustments, which were extracted from DOE (1999), are based on only a 
limited amount of data available for the behavior of the waste forms or similar materials but are 
considered conservative. Moreover, in some cases (such as the uranium metal and immobilized 
Pu ceramic waste form) even the best-estimate models abstracted are the result of conservative 
analyses of the experimental dissolution data. 
Models for degradation of all the surrogate waste forms discussed in this section, Section 6.4, 
and Table 1B do not need to be validated to support license application submittal because the 
degradation response of the surrogate is a characteristic to be used to develop technical 
specifications for acceptance and emplacement of the DSNF. As documented in this section, 
DOE will use scientific judgement supported with available data and models to make the 
assumed performance of DSNF a reasonable estimate of the actual performance. Degradation 
models that will show specific waste forms perform as well as or better than the assumed 
performance of the surrogate may need to be qualified to support the demonstration of 
conformance to WAC of those waste forms for emplacement. 
6.3.1 DSNF Group 1 (Naval SNF) Models 
Commercial LWR SNF will be used as the surrogate for naval SNF in repository performance 
analyses. Expected releases from naval SNF waste packages was provided in Mowbray (2000). 
Because of its robust design, the radionuclide releases from naval SNF waste packages are 
considerably less than releases from commercial LWR SNF waste packages; accordingly, this 
assumption is conservative. Naval SNF represents approximately one-tenth of one percent of the 
spent fuel MTHM inventory. The information contained in Mowbray (2000) is qualified data. 
6.3.2 DSNF Group 2 (Pu/U Alloy) Models 
There are several individual types of Pu/U alloy-based DSNF primarily comprised of U-Mo and 
U-Zr alloys although smaller quantities of U-Th and U-Fe alloy are part of this group (DOE 
1999, Section 6.2). These alloy fuels are generally clad in zirconium alloy, but some small 
quantities have aluminum, stainless steel, or tantalum alloy cladding. The largest single fuel 
types in this group, comprising slightly over 90% by weight of uranium of the total Group 2 
inventory, are the zirconium-clad U-Mo Fermi reactor SNF and stainless steel-clad U-Zr alloy 
Annular Core Research Reactor SNF. Studies of the dissolution behavior of U-Mo and U-Zr 
alloys, reported in DOE (1999), give dissolution rates for the alloy, which depend sensitively on 
the amount of alloying molybdenum and zirconium. They generally show U-Mo dissolution 
rates to be slightly higher than U-metal and U-Zr alloy slightly lower than U-metal. Also, the 
total inventory of this SNF waste form is very small compared to the other DSNF types (see 
Table 3b). From this behavior it is recommended that the best-estimate degradation rate for the 
Group 2 fuels under wet oxic and humid oxic conditions be taken as the model for U-metal from 
DOE (1999, Section 6.7), as follows:

 20  
R [kg/m2-s] = 1.88 x 103 exp (-7970/TK) (wet oxic conditions) (Eq. 1a) 
which converts to mg/m2-d units as follows: 
R [mg/m2-d] = 1.88 x 103 exp (-7970/TK) [kg/m2-s][106 mg/kg][3600 s/h][24 h/d] 
= 1.62 x 1014 exp (-7970/TK) 
and 
R [kg/m2-s] = 0.27 x 102 exp (-7240/TK) (humid oxic conditions) (Eq. 1b) 
which converts to mg/m2-d units as follows: 
R [mg/m2-d] = 0.27 x 102 exp (-7240/TK) [kg/m2-s][106 mg/kg][3600 s/h][24 h/d] 
= 2.33 x 1012 exp (-7970/TK) 
and the conservative model should be taken as 10X this value. 
These equations were derived from data in the open literature for the corrosion of unirradiated 
metallic uranium and uranium alloy, and the suggested degradation model for Group 2 SNF was 
obtained from DOE (1999, Section 6.7) for unirradiated uranium metal. Because the 
data/information upon which the suggested model in this reference is based is from open 
literature sources, it is from uncontrolled sources and, thus, may need to be validated to support 
the demonstration of conformance to WAC of specific waste forms. 
6.3.3 DSNF Group 3 (Pu/U Carbide SNF) Models 
Group 3 SNF consists primarily of fuel from the Fast Flux Test Facility (FFTF) with most of the 
balance from the Sodium Reactor Experiment (SRE). Both consist of mixed-carbide-fissile fuel 
particles in a nongraphite matrix. Only a very limited amount of information concerning the 
chemical reactivity is available. Because of the lack of specific information concerning 
degradation behavior and the indication that the fissile particles could be very reactive, the 
abstracted best-estimate and conservative-degradation rate models abstracted should be taken as 
100X the dissolution rate of the uranium metal SNF (DOE 1999, Section 6.3). That is, no credit 
are taken for lower degradation rates in the best-estimate model formulation for this waste form. 
The suggested degradation model for Group 3 SNF was obtained from DOE (1999, Section 6.3). 
The data/information upon which the suggested model in this reference is based is scarce and 
from old open literature sources. 
6.3.4 DSNF Group 4 (MOX and Pu Oxide SNF) Models 
Group 4 SNF is composed of a mixture of uranium and Pu oxides with various cladding 
materials. Although several dozen SNF types are in this category, the largest single type is the

 21  
FFTF DFA (demonstration fuel assembly) and TFA (test fuel assembly) fuel, contributing over 
50% of the uranium by weight. Since the fuel material is either uranium oxide or Pu oxide, the 
dissolution kinetics of the fuel form is not expected to be materially different from that for 
commercial LWR SNF. The best-estimate and conservative models abstracted are those for the 
LWR SNF, i.e., no credit will be taken for lower degradation rates in the best-estimate model 
formulation for this waste form. 
The suggested degradation model for Group 4 SNF is that abstracted in an AMR titled CSNF 
Waste Form Degradation: Summary Abstraction (CRWMS M&O 2000a) for commercial UO2 
fuel. The data/information upon which the suggested model in this reference is based is qualified 
data generated in support of the CRWMS M&O by PNNL and the Lawrence Livermore National 
Laboratory (LLNL) for uranium-dioxide based commercial LWR SNF, a material similar, but 
not identical, to the SNF in this group. 
6.3.5 DSNF Group 5 (Th/U Carbide SNF) Models 
This SNF group consists primarily of thorium or uranium-carbide particles coated with pyrolytic 
carbon or silicon carbide embedded in a carbonaceous matrix. Over 90% by weight of MTHM 
of this group is Fort St. Vrain (FSV) fuel, with the remainder being Peach Bottom (PB) fuel. 
The PB fuel may be more damaged than the FSV fuel although there is little qualified 
information concerning the condition of either. Fuel in this group, whose protective coatings and 
matrix are intact, would be expected to have dissolution kinetics similar to those of pure silicon 
carbide. SNF with a damaged coating or matrix would be expected to have dissolution kinetics 
similar to uranium carbide (DOE 1999, Section 6.5; CRWMS M&O 1998a, Section 6.3.2.1). 
DOE (1999, Sections 6.5 and 6.7) contains comparative values for the corrosion rates of silicon 
carbide, uranium carbide, and uranium metal. The suggested uranium carbide corrosion rate was 
about 10X that of uranium metal. For these reasons, the recommended best-estimate degradation 
model is that for silicon carbide given in DOE (1999) and the conservative rate model is taken as 
10X that for uranium metal. 
The suggested degradation model for Group 5 SNF was obtained from DOE (1999, Section 6.5). 
The data/information upon which the suggested model in this reference is based is from open 
literature sources. 
6.3.6 DSNF Group 6 (Th/U Oxide SNF) Models 
Thorium uranium oxide spent fuels primarily consist of the Shippingport light-water breeder 
reactor reflector SNF with the remainder from the Dresden and experimental research reactor 
(ERR) thorium-uranium-oxide SNF. The Shippingport fuel was clad in zirconium alloy, and the 
Dresden and ERR fuels were clad in stainless steel. The thorium-uranium-oxide fuel consisted 
of sintered pellets similar to commercial LWR fuel pellets. 
Several reports discussed in DOE (1999, Section 6.6) attest that the mixed thorium uranium was 
more corrosion resistant than pure uranium dioxide, as much as five orders of magnitude more 
corrosion resistant. Since the dissolution of the thorium-uranium oxide was not specifically 
measured, the approach taken is to use a ceramic release model that conservatively bounded data 
for unirradiated LWBR fuel (DOE 1999, Section 6.6). Thus, the use of the ceramic synroc

 22  
model was abstracted as the best-estimate model and 1000X the ceramic model as the 
conservative model. 
6.3.7 DSNF Group 7 (U-metal SNF) Models 
The zirconium-clad N-reactor SNF constitutes over 95% of this group with small quantities of 
aluminum-clad-single-pass reactor (SPR) and EBR-II-metallic uranium SNF. A significant 
fraction of the N-reactor fuel is visibly damaged, and much of the rest could have small 
pinholes/cracks in the cladding. The exposed uranium metal surfaces of the N-reactor fuel 
elements show extensive corrosion resulting from the many years of direct exposure to the 
K-basin water (Abrefah et al. 1995; Abrefah et al. 1999, Figure 3.1; Welsh et al. 1997, Section 
1.0, Figures 3.3-3.7). 
Experimental studies have been conducted in the past on unirradiated and uncorroded uranium 
metal and uranium-metal alloy, and some of this work is summarized in DOE (1999, Section 
6.7). Additionally, the NSNFP has conducted an analysis of the kinetics of uranium metal 
oxidation (Pasupathi 2000) in which a correlation for the dissolution kinetics of 
unirradiated/uncorroded uranium metal in water was derived from several literature sources as 
follows: 
k (mg/cm2-hr ) = 5.03 x 109 exp(-A/RTK), or 
Eq. 2A 
k (mg/m2-day) = 1.21 x 1015 exp(-A/RTK) 
where k is the corrosion rate, A is the activation energy for dissolution (66.4 ± 2.0 kJ/mol), R is 
the gas constant (8.314 J/mol-K), and TK is the temperature in kelvins. 
PNNL has conducted dissolution tests on samples of the N-reactor SNF which generally indicate 
that the rates of dissolution somewhat exceed those of the unirradiated/uncorroded uranium 
metal or alloy. The PNNL experimental work was performed on N-reactor SNF samples taken 
from an undamaged/uncorroded area of an N-reactor fuel element (Gray and Einziger 1998, 
Section 2.2 and Figure 1A). The results indicated that there were two stages in the dissolution 
behavior of the N-reactor SNF samples tested: an initial Stage 1 rate and, after an incubation 
period, a faster Stage 2 dissolution rate. The Stage 1 dissolution resembled the congruent 
dissolution noted in the similar PNNL dissolution experiments on uranium-dioxide based LWR 
SNF samples; that is, the uranium went directly into solution as the soluble uranyl species 
(UO2
++) via the following reactions: 
U + O2 à UO2 
UO2 + H2O + ½ O2 à UO2
++ + 2OHThis 
Stage 1 congruent dissolution of the matrix was correlated with the following expression: 
log R (mg/m2-d) = 8.52 + 0.347 log[CO3
--] + 0.088 pH – 1929/TK l (Eq. 2B)

 23  
where [CO3
--] is the molar concentration of carbonate in the contacting solution. The flowthrough 
test data that was used to generate this expression qualified as service condition test 
results under the testing/modeling logic of ASTM C 1174-97 (1997). Unsaturated drip testing of 
metallic uranium, which serves as characterization testing under the logic of ASTM (1997), 
showed that alteration phases could form, which would potentially retard the release of solutes 
from the N-reactor SNF, and this could be responsible for the lower Stage 1 rates (Hurt 2000). 
During Stage 2, the dissolution ceased to be congruent and disintegration of some of the fuel 
matrix, which formed a sludge-like material primarily consisting of U4O9 and/or the mineral 
form Schoepite (UO3 · 2H2O), was observed. It was also postulated that the Stage 2 dissolution 
may have been faster because it coincided with the depletion of dissolved oxygen and 
consequent formation of an anoxic condition in the contacting water, represented by the 
following equation: 
U + 2H2O à UO2 + 2H2 
The rates of dissolution that were observed for Stage 2 ranged from 10,000 mg/m2-d at 25oC to 
290,000 mg/m2-d at 75oC for some samples. The temperature dependence of the dissolution was 
similar to that for unirradiated/uncorroded uranium metal in the Pasupathi (2000) study, although 
the absolute rates were higher for the N-reactor SNF (see Table 1a below). Stage 2 dissolution 
generally began around sixty days into the testing. Table 2 gives values for the Stage 1 
dissolution rate as calculated from the above rate expression and for the experimental Stage 2 
dissolution rates of the PNNL study. It should be noted that the N-reactor SNF dissolution 
testing is still underway; and, therefore, the results upon which this model was derived are 
preliminary. 
In most of the cases where Stage 2 dissolution was initiated, it began around 60 days into the 
test, a very short period in terms of TSPA analyses. A significant fraction of the N-reactor fuel 
had been stored in a damaged condition under water at the K-basins at Hanford, and the metallic 
uranium has been exposed to the water environment. Therefore, the degradation model for 
unirradiated and unexposed U-metal derived in DOE (1999) and recommended for the Group 2 
(U/Pu alloy), Group 3 (Pu/U carbide), and Group 10 (unknown) SNF is not recommended for the 
N-reactor SNF, which forms the basis for Group 7. Although the PNNL experimental data is 
limited, it indicates that Stage 2 dissolution kinetics should be followed for the N-reactor SNF. 
Based on these analyses, the abstracted best-estimate degradation rate model for the U-metal 
SNF is selected so as to encompass the highest directly observed congruent dissolution rates in 
the PNNL studies or 1.8 x 105 mg/m2-d (0.75 mg/cm2-hr) at 75oC and 13,000 mg/m2-d (0.05 
mg/cm2-hr) at 25oC (Gray and Einziger 1998, Section 4.3). 
It was noted above that the samples of N-reactor SNF tested in the flow-through testing at PNNL 
were from an undamaged and uncorroded portion of an N-reactor fuel element. No dissolution 
testing has been conducted on samples of damaged/corroded N-reactor SNF, although this 
damaged/corroded condition would be expected to be the case in repository environment 
exposure after breach of the waste package in the repository. In other studies at PNNL 
concerning the ignition properties (Abrefah et al. 1999, Table S.1) and oxidation kinetics of Nreactor 
SNF (Abrefah et al. 1998), samples of N-reactor SNF taken from the damaged/corroded

 24  
areas of the N-reactor fuel elements showed higher oxidation rates than undamaged/uncorroded 
samples. 
Table 1a compares values for Stage 2 dissolution rates of N-reactor SNF obtained experimentally 
by Gray and Einziger (1998) with rates calculated from the correlation recommended for water 
dissolution of uranium metal in Pasupathi (2000, Table 2-5), and 5 times and 25 times this value. 
Note that the reaction rates given for the N-reactor SNF are approximately a factor of two to five 
higher than those given using Equation 2A. Note also that use of the lower (64.4 kJ/mol) one 
standard deviation (1s) uncertainty given in the equation 2A expression for the activation energy 
(66.4 ± 2.0 kJ/mol) would result in another factor of 2 to 3 in the reaction rate in this temperature 
range. Thus, the combination of these two factors would result in roughly a factor of 5 increase 
on the nominal equation 2A reaction rate, which would represent the most defensible bestestimate 
model kinetics. 
In the PNNL N-reactor SNF experimental work (Gray and Einziger 1998, Section 4.3), the 
highest dissolution rate inferred by cesium release results (but not directly observed as congruent 
uranium dissolution) was 290,000 mg/m2-day (1.21 mg/cm2-hr). This is approximately a factor 
of 1.6 higher than the maximum value of 180,000 mg/m2-day observed for congruent dissolution 
of the uranium. Combining this inferred rate for the N-reactor SNF and a 2s uncertainty (95%) 
lower activation energy of 62.2 kJ/mol would result in another approximate factor of 5 over the 
best-estimate rate for use as the conservative model. 
Table 1a. Comparison of Unirradiated/Uncorroded U Metal and N-Reactor SNF Corrosion Rates 
Temp 
(oC) 
N-Reactor 
SNF(1) 
U-metal (2) U-metal (3) 5X U-metal(2) 25X U-metal (2) 
25 0.05 mg/cm2-hr 
(1.30 x 104 
mg/m2-day) 
0.01 mg/cm2-hr 
(2.77 x 103 
mg/m2-day) 
0.03 mg/cm2-hr 
(6.21x103 
mg/m2-day) 
0.06 mg/cm2-hr 
(1.39 x 104 
mg/m2-day) 
0.29 mg/cm2-hr 
(6.93 x 104 
mg/m2-day) 
75 0.75 mg/cm2-hr 
(1.80 x 105 
mg/m2-day) 
0.54 mg/cm2-hr 
(1.31 x 105 
mg/m2-day) 
1.08 mg/cm2-hr 
(2.60 x 105 
mg/m2-day) 
2.71 mg/cm2-hr 
(6.51 x 105 
mg/m2-day) 
13.57 mg/cm2-hr 
(3.26 x 106 
mg/m2-day) 
(1) from Gray and Einziger 1998, Section 4.3 
(2) from Equation 2A using the nominal activation energy of 66.4 kJ/mol 
(3) from Equation 2A using the 1s minimum activation energy of 64.4 kJ/mol 
For these reasons it is recommended that the best-estimate model for the dissolution of the Nreactor 
SNF be taken as 5 times, and the conservative model be taken as 25 times, the Pasupathi 
(2000) model represented by Equation 2A. 
6.3.8 DSNF Group 8 (U Oxide SNF) Models 
TSPA Group 8 consists of uranium dioxide-based SNF removed from commercial LWRs or 
similar SNF from test reactors. About half of the total inventory of approximately 178 MTHM

 25  
comes from the TMI-2 core and, therefore, is not intact fuel. The other half is substantially 
undamaged SNF from other commercial reactors. The dissolution kinetics per unit area of the 
damaged fuel, such as the TMI rubble, would be expected to be similar to the kinetics of the 
uranium-dioxide based LWR spent fuel reported in CRWMS M&O (2000a, Eq. 11, Table 14) 
but with a substantially enhanced effective surface area for release. 
log10 R (mg/m2-d) = 5.479057 + [-2457.050662 (1/TK)] + [1.510878 (-log10(CO3
--))] + [- 
1.729906 (-log10(O2))] + [0.234718 pH] + [-0.799526 log10 BU] + 
[400.755947 (-log10 O2) (1/TK)] + [780.806133 (log10 BU) (1/TK)] + 
[0.172305 (log10 BU) (-log10(CO3
--))] + [0.174428 (log10BU) (-log10 O2)] + 
[-0.271203 (log10BU) (pH)] + [-0.339535 (-log10(CO3
--))2] (Eq. 3) 
where TK is the temperature in kelvins, CO3
-- is the molar concentration of carbonate ion in the 
liquid phase, BU is the spent fuel burnup in MWd/kgU, and O2 is the oxygen partial pressure in 
atmospheres in the gas phase. 
This model for the degradation rate of uranium-dioxide spent fuel was derived from dissolution 
tests that qualified as service condition and/or characterization tests under the testing/modeling 
logic of ASTM C 1174-97 (1997). These tests represented the maximum forward-dissolutionreaction 
rate for the material with no back reactions that would inhibit dissolution. The fact that 
the dissolution reaction represented the maximum rate was verified by the observations of 
congruent dissolution and the lack of precipitated alteration phases on the test specimens during 
the tests. Vapor-phase tests were also performed on samples of the commercial LWR SNF that 
served as characterization tests per the logic of ASTM C 1174-97. These tests indicated that 
alteration phases could form on the SNF, and these phases would inhibit the overall dissolution 
rate (CRWMS M&O 2000a, Sections 6.3.1 and 6.3.2). Therefore, the testing overall 
demonstrated the conservatism inherent in using Equation 3 for the dissolution rate for uraniumdioxide 
based SNF. 
DOE (1999, Section 6.8) suggests a surface-area enhancement factor of 100 for a release model 
representing the Group 8 SNF. Thus, the abstracted best-estimate-degradation model for intact 
uranium dioxide based Group 8 SNF should be the same as the degradation model for 
commercial LWR SNF (Equation 3), and the conservative degradation model for intact Group 8 
fuel should be 100X the best-estimate model. The abstracted best-estimate and conservativedegradation 
models for the non-intact Group 8 SNF is 100X the intact fuel best-estimate model, 
i.e., no credit is taken for lower degradation rates in the best-estimate model formulation for this 
waste form. 
6.3.9 DSNF Group 9 (Al-based SNF) Models 
This group consists of fuels based on a uranium aluminide, uranium silicide, or uranium oxide 
particle phase dispersed in a continuous aluminum alloy matrix. The fission product 
radionuclides remain in the dispersed phase; therefore, the dissolution of the dispersed phase 
material is the parameter most germane to the release of the radionuclides upon contact with 
groundwater. The dissolution rate of interest is expressed in terms of mgU/m2-day.

 26  
Much of this spent fuel (~36%) is from foreign research reactor sources, with the balance from 
domestic research reactors such as the high flux irradiation reactor (HFIR), the advanced test 
reactor (ATR), and university research reactors (DOE 1999, Appendix D). SRS has conducted 
dissolution studies on SRS reactor spent fuel samples at 25oC in both J-13 well water and 
bicarbonate solutions (Wiersma and Mickalonis 1998, Table 3) and on unirradiated samples at 
25oC and 90oC (Wiersma and Mickalonis 1998, Table 4). Flow-through dissolution tests at 250C 
subsequently performed at PNNL on similar samples gave results similar to the SRS study (Gray 
2000, Table 7). The SRS data showed dissolution rates for the irradiated SNF of 0.19-0.22 
mgU/m2-day in J-13 well water and 22-36 mgU/m2-day in bicarbonate solution. The corrosion 
data for the unirradiated fuel samples showed that the dissolution rate at 90oC was approximately 
10X the rate at 25oC. The J-13 well-water data is selected for the best-estimate degradation 
model because the groundwater chemistry at the time of waste-package failure is expected to be 
approximately that of the J-13 well water. Since bicarbonate is a potentially more aggressive 
water condition (CRWMS M&O 2000a, Table 23), the bicarbonate data is used as the basis for 
the conservative model. 
6.3.10 DSNF Group 10 (Unknown SNF) Models 
Group 10 SNF consists of a small amount of uranium nitride SNF and fuel with unknown 
matrices. This group consists of only about 0.2% of the total inventory of DSNF in MTHM. 
Because of the unknown fuel matrix material and the small volume, the degradation models for 
this group are based on the dissolution kinetics of unirradiated uranium metal, similar to the 
Group 2 SNF. Therefore, the abstracted best-estimate and conservative models are the same as 
for the Group 2 SNF. 
6.3.11 DSNF Group 11 (U-Zr-Hx) Models 
TRIGA test reactor fuel comprises ~97% of the total Group 11 SNF inventory of ~1.6 MTHM 
(DOE 1999, Appendix D). The TRIGA fuel consists of a dispersion of fine particles of metallic 
uranium dispersed in a zirconium hydride matrix (U-Zr-Hx). Uranium loadings varied from 
approximately 8 to 45 wt% (LMITCO 1997, Section 1). 
Unirradiated U-Zr-Hz fuel has been shown to have good elevated temperature corrosion 
resistance (LMITCO 1997, page 2-5). However, there is no known qualified data for the 
dissolution rate of this material in repository-relevant water and temperature conditions. For this 
reason the NSNFP DSNF information report (DOE 1999, Section 6.11) proposes that the 
dissolution rate for this DSNF form be taken as 0.1X the uranium oxide SNF dissolution rate. 
The low total inventory in MTHM of the U-Zr-Hx SNF makes the repository TSPA performance 
of this material insensitive to the degradation model for this waste form. 
6.3.12 Immobilized Ceramic Pu Disposition Waste Form Models 
The waste form for the immobilized Pu will be cold-pressed, titania-based pyrochlore ceramic 
disks containing approximately 10.5 wt% embedded Pu as PuO2. These disks will be stacked in 
stainless steel cans, which are, in turn, embedded inside a canister filled with a vitrified 
borosilicate filler glass similar to the HLW glass waste form (CRWMS M&O 1998a, pp. 2-3,

 27  
Tables 2.2.3-1 and 2.2.4-1). Although the ceramic waste form is similar to synroc (CRWMS 
M&O 1998a, Section 6.3.2.2) in that it is titania-based, there is limited dissolution rate data 
(Shaw 1999, Section 6.1) specifically for this ceramic. Shaw (1999, Section 6.1) and Bourcier 
(2000, Section IV) give the best-estimate model proposed in this report for the dissolution rate, 
K, at 25oC of a combination zirconolite brannerite pyrochlore ceramic: 
For pH < 7 
log10 K = -0.167 (pH) – 4.66 
For pH ³ 7 
log10 K = 0.25 (pH) – 8 (Eq. 4) 
With a temperature dependence given by, 
ln(K2/K1) = (Ea/R) (T2-T1)/(T2T1) 
and K is the dissolution rate in g/m2-day, T is the temperature in kelvins, Ea is 16 kcal/mol, and 
R is 8.314 J/mol-K (1.987 cal/mol-K). 
The dissolution rates represented by these expressions are inherently somewhat conservative 
since they are based on the results of Single Pass Flow-Through Tests (Shaw 1999, Section 6.1), 
which are designed to measure a forward reaction rate under conditions far from thermodynamic 
equilibrium and do not take into account the buildup of dissolved components in the contacting 
water or the development of protective layers. 
Bourcier (2000, Figure 14) shows that these expressions for the best-estimate model provide 
dissolution rates, which generally fit the spread of experimental data in the low pH region, while 
slightly overpredicting the experimentally determined rates in the high pH region. A factor of 
10X the best-estimate model would bound all the dissolution rates and is, therefore, 
recommended for use of the conservative model for the immobilized Pu ceramic. 
This model represents the dissoluton of the titanate ceramic matrix of the waste form. The 
studies, which supported the development of this model, showed that the release of plutonium 
was generally congruent with the dissolution of the ceramic matrix, but that this needed to be 
verified by further testing (Bourcier 1999). A rate model similar to this best-estimate model has 
been used in external criticality analyses for degraded waste packages (CRWMS M&O 1998a), 
and dissolution rate experiments at ANL/LLNL sponsored by the PIP continue to provide data in 
support of model development and validation. It is anticipated that validation of this model 
would enable its usage in the TSPA safety or reasonable representation cases for the immobilized 
ceramic Pu waste form. 
Since the stainless steel cans into which the immobilized Pu ceramic is emplaced are in turn to 
be embedded in a HLW glass matrix, a degradation model for the glass may be required if 
dissolution of the glass is a necessary precondition for exposure of the ceramic to water. The 
HLW glass dissolution models that should be used in this case should be that developed in 
(CRWMS M&O 2000c, Section 7) as follows:

 28  
R/Sim (g/m2-d) = keff · 10h·pH · exp(-Ea/RTK) (Eq. 5) 
where 
Sim = the effective reacting surface area of the glass in which the ceramic disks are embedded 
when immersed in the groundwater and is taken as 20X the geometric surface area, 
R = 0.00831 kJ/mol-K 
for pH < pHm; log10 keff = 9 ± 1, h = -0.6 ± 0.2, and Ea = 58 ± 15 kJ/mol 
for pH ³ pHm; log10 keff = 6.9 ± 0.5, h = 0.4 ± 0.1, and Ea = 80 ± 10 kJ/mol 
and pHm = 2.1 + 1149/TK 
6.4 ABSTRACTION SUMMARY-DSNF AND PU DISPOSITION FORM 
DEGRADATION MODELS 
6.4.1 Model Basis 
A summary of the selected/abstracted group-specific-upper-limit, conservative, and best-estimate 
degradation models to be used to license the DSNF and immobilized ceramic Pu waste forms is 
given in Table 1b. 
For all groups, other than the naval DSNF group, the recommended upper-limit model to be used 
in safety case TSPA analyses is complete dissolution of the waste form during a single TSPA 
code (e.g., RIP) time step upon exposure of the waste form to groundwater. This is chosen as the 
upper-limit model in part because TSPA analyses performed for the DSNF (Thornton 1998a, 
1998b) have shown that the overall effect of the failure of DSNF-containing waste packages does 
not significantly contribute to the repository site boundary dose. This document also 
recommends that the U-metal (Group 7) conservative and best-estimate models be used, along 
with the U-metal fuel surface area (7 x 10-5 cm2/g) given in DOE (1999, Appendix B), as 
surrogate for all but the naval DSNF in cases where less conservative models are required or 
deemed appropriate by PA personnel. 
The group-specific conservative and best-estimate degradation models described in Section 6.3 
may be used in the demonstration of conformance to WAC analyses for specific waste forms. It 
is recognized, however, that they are currently based on limited and generally unqualified 
corrosion, dissolution, or oxidation data for most of the DSNF groups. 
Qualified data/information are used in other AMRs to abstract the uranium dioxide-based SNF 
degradation model (CRWMS M&O 2000a) and borosilicate glass (as a surrogate for the 
immobilized ceramic Pu disposition waste form) degradation model (CRWMS M&O 2000c). 
The data used to generate the metallic uranium-based SNF degradation models were taken under 
a PNNL QA program. 
Commercial LWR SNF will be used as the surrogate for naval SNF in repository performance 
analyses. Expected releases from naval SNF waste packages was provided in Mowbray (2000). 
Because of its robust design, the radionuclide releases from naval SNF waste packages are 
considerably less than releases from commercial LWR SNF waste packages; accordingly, this

 29  
assumption is conservative. Naval SNF represents approximately one-tenth of one percent of the 
spent fuel MTHM inventory. The information contained in Mowbray (2000) is qualified data. 
The other data and/or information on which the DSNF group models were based are unqualified; 
therefore, in this analysis they are used for comparison purposes only because it is recommended 
herein that the N-reactor SNF (Group 7) best-estimate and conservative models be applied to the 
total DSNF radionuclide inventory for TSPA. The reasons for proposing that these models be 
used to represent the degradation behavior of the entire DSNF inventory follow: 
· The N-reactor SNF model predicts dissolution rates greater than most other groups. 
· The total inventory of the Group 3 SNF (the only group showing reaction rates greater than 
the uranium metal SNF) is only 0.10 MTHM compared to the total inventory of ~2,400 
MTHM (DOE 1999, Appendix D). 
· There is insufficient qualified information/data to support detailed degradation models for 
each DSNF type or group. 
· TSPA calculations for DSNF (CRWMS M&O 2000f, Section 6.7.1 and Section 7, bullets 11- 
16) showed that the postclosure site boundary doses are insensitive to the DSNF release 
model. 
Table 3a gives a comparison of the dissolution rates for each fuel group at 50oC (328 K) and 
100oC (373 K) for pH 8, 0.002 molar CO3
-- in the contacting water and 20% oxygen in the gas 
phase. Also included in this table is the approximate inventory in MTHM of each of the DSNF 
groups. Table 3b further gives an example of the mass fractional degradation rate at 50oC using 
Equation 2 above, which demonstrates the rapid corrosion rates of these waste forms compared 
to the long time frames of the TSPA RIP code representation. When the dissolution rates 
calculated from the models and the relative MTHM are compared for the DSNF groups, it can be 
seen that use of the Group 7 degradation model to conservatively bound the entire DSNF 
inventory is justified. Aside from the Group 7 U-metal SNF, the groups that have the highest 
release rates also have the lowest MTHM. The only DSNF type that the available data indicates 
dissolves faster than metallic uranium SNF is the Group 2 Pu/U carbide, and the inventory of this 
fuel is four orders of magnitude lower than that of the N-reactor SNF. 
The abstracted best estimate and conservative models for the degradation of the immobilized Pu 
ceramic disposition waste form are based on the titanate ceramic dissolution model described in 
Shaw (1999, Section 6.1) and Bourcier (2000, Section IV). The best-estimate model uses the 
parameters given in Boucier (2000, Table 5) and the conservative model is taken as 10 times the 
rate given by the best-estimate model. Instantaneous release in a single RIP time step is 
recommended as the bounding model for postclosure TSPA purposes (CRWMS M&O 2000f, 
Section 7). If required for the TSPA, the HLW glass model to be used in conjunction with the 
immobilized Pu ceramic model is that given in CRWMS M&O (2000c). 
The fraction of the released radionuclides that may be in colloids versus solutes has not been 
determined experimentally for the DSNF and immobilized ceramic Pu waste forms, although 
there is NSNFP-sponsored work in progress concerning colloid formation from the degradation 
of N-reactor (U-metal) and MOX SNF. Until the results from these experiment are available, it

 30  
is proposed herein that the colloid formation and radionuclide attachment/detachment kinetics 
for the DSNF be taken as that for the CSNF given in CRWMS M&O (2000b, Section 6.2.1.2). 
Due to the compositional similarity of the borosilicate glass matrix to the HLW glass, the 
formation and radionuclide attachment/detachment characteristics of colloids released as a result 
of the degradation of the immobilized ceramic Pu waste form should be taken as that of the 
HLW glass in CRWMS M&O (2000b, Section 6.3). 
6.4.2 Model Uncertainties 
The application of the degradation models abstracted in this document involves the extrapolation 
of data over periods of time that are orders of magnitude greater than the experimental test 
periods used to generate the data. ASTM C1174-97 (1997, Section 24) recommends that 
uncertainties in the extrapolation of such models be minimized through the use of models whose 
mathematical forms are as mechanistic as possible. However, it can be seen from the 
abstractions above that the lack of any directly relevant experimental dissolution/degradation 
data for many of the DSNF waste forms, and the relatively small amount of data for those that 
have been tested, makes the generation of mechanistic models problematic. Additionally, 
uncertainties in the data used to generate the models-such as in the surface area measurements 
used to calculate normalized dissolution rates (Gray and Einziger 1998, Section 2.5)-produce 
significant uncertainties even in the short term application of the models. For this reason, and 
because preliminary TSPA analyses (Thornton 1998a, 1998b) have shown that the overall 
performance of the MGR is very insensitive to the degradation rate of the DSNF, the emphasis in 
this document, whenever possible, is on the application of upper-limit or bounding degradation 
models. 
6.4.3 Surface Area of DSNF Groups and Immobilized Pu Ceramic Waste Form 
The abstracted models for the degradation rates of the DSNF groups are generally expressed in 
terms of the weight of material released per unit surface area of SNF exposed to water per unit 
time, typically in units of mg/m2-day. In order to use the conservative and best-estimate models 
abstracted in this report to calculate the actual release rate from a waste form exposed to water 
after failure of a waste package, it is, therefore, necessary to use a conservative estimate for the 
exposed surface area of the fuel. It is not necessary to use a value for the surface area for the 
upper-limit model since this model, which is total release of the inventory, is not dependant on 
exposed surface area. 
The N-reactor SNF, represented in group 7, constitutes approximately 85% by weight of the total 
DSNF inventory, and is known to have extensively damaged cladding, in terms of both gross 
damage and cracking. Although some of the other fuel groups contain fuel types described as 
intact, such as MOX and the aluminum-based fuels, there is no known documentation of 
analyses that demonstrate the cladding will maintain fully intact until the time of delivery to the 
repository, and likewise no known studies showing the cladding will remain intact after waste 
package failure. DOE (1999, Appendix B) calculated the geometric surface area of potentially 
exposed fuel based on the geometric characteristics of representative fuels within the groups. 
Where the fuel group contained damaged fuel, or fuel suspected to be damaged, a roughness 
factor for the surface area of 5 was applied. The result is the estimate of exposed surface area 
per gram of fuel group given in the third column of Table 3b. If the best-estimate and/or

 31  
conservative degradation models for each DSNF group are required for licensing of specific 
waste forms for repository emplacement, then those models should use the estimated surface area 
for the individual groups in calculating overall release rates for waste packages containing those 
groups. If a repository TSPA analysis, such as, for example, the reasonable representation case, 
requires a calculation of the overall release from DSNF waste forms, then the weighted average 
specific surface of 0.03 m2/g calculated using column 4 and footnote (c) of Table 3b could be 
used. 
For the immobilized Pu ceramic waste form, the multiplication factor (to account for the 
combined effects of enhanced surface area and radiation damage effects) to be used in 
conjunction with the best estimate and conservative models should be 30 (Shaw 1999, Section 
6.2). The effective surface area to be used in conjunction with the these models is the geometric 
surface area of the Pu-containing ceramic disks embedded in the HLW glass, 0.0123 m2 
per 
ceramic disk (CRWMS M&O 1998a, footnote (b) of Table 3b), multiplied by this factor of 30. 
This translates to an effective surface area of 0.369 m2 per ceramic disk for the disk design 
described in CRWMS M&O (1998a, Section 2.2.1). 
7. CONCLUSIONS 
This document and its conclusions may be affected by technical product input information and 
assumptions that require confirmation. Any changes to the document or its conclusions that may 
occur as a result of completing the confirmation activities will be reflected in subsequent 
revisions. The status of the input information quality may be confirmed by review of the 
Document Input Reference System database. 
In particular, the conclusions may be affected by the confirmation of currently unvalidated inputs 
concerning the dissolution of U-metal SNF (Gray and Einziger 1998), the dissolution of 
unirradiated/uncorroded U-metal (Pasupathi 2000), the dissolution rates of the immobilized PU 
ceramic (Shaw 1999), and unvalidated assumptions concerning the use of LWR SNF as a 
surrogate for naval SNF (Sections 5.3 and 5.4) and the similarity of DSNF colloid behavior to 
that of CSNF and HLW glass (Section 5.5). There are analytical activities and tests sponsored by 
OCRWM, NNPP, and NSNFP currently underway that should enable confirmation of these 
inputs and assumptions. Should these inputs and assumptions not be confirmed/validated, then 
the only DSNF degradation model, which could be used for the TSPA-SR or TSPA-LA, would 
be the Upper Limit model (i.e., complete dissolution of the waste form during one TSPA time 
step). This model does not require confirmation/validation because it represents the fastest 
possible release kinetics. 
7.1 DSNF DEGRADATION MODELS VALIDITY 
7.1.1 Dissolution Models 
The only DSNF waste forms for which there are directly applicable dissolution data suitable for 
generating a degradation model are the Group 7 (uranium metal), and Group 9 (Al-based) DSNF. 
It is anticipated that new dissolution data for Group 4 (MOX) DSNF (DOE 1998) will become 
available in the near future. This DSNF data is currently unqualified. In contrast to this, 
sufficient qualified experimental data for the dissolution of commercial uranium dioxide LWR

 32  
spent fuel similar to the Group 8 DSNF (CRWMS M&O 2000a) and HLW glass (CRWMS 
M&O 2000c) have been generated under the OCRWM QARD program to generate adequate 
semi-empirical models for the dissolution rate of these waste forms. 
The NSNFP has sponsored dissolution testing at PNNL of N-reactor SNF, but this testing is not 
yet complete. Likewise, the dissolution testing for the Al-based DSNF sponsored by NSNFP at 
SRS is not yet complete; therefore, the results should be regarded as preliminary. NSNFP has 
also initiated dissolution testing of MOX DSNF, but the results are, at this time, not yet sufficient 
to formulate a degradation model. There is also testing of candidate Pu immobilized ceramic 
compositions underway at ANL, PNNL, and LLNL that may enable improved release models in 
future revisions of this AMR. 
There are no known current plans at NSNFP or elsewhere to perform dissolution tests on 
surrogates of the other DSNF groups. Thus, the models referred to in Table 1 for those groups 
are not expected to be validated. However, these models do not require validation if the models 
recommended for TSPA-SR (i.e., the U-metal SNF models) conservatively bound the 
degradation rates calculated from these models, and the analyses in this report demonstrate that 
they do. 
The results of TSPA sensitivity analyses for DSNF (CRWMS M&O 2000f) indicate that the 
performance of the repository is very insensitive to the DSNF degradation kinetics. If, because 
of this insensitivity, the upper-limit model is the only one used for TSPA analyses, then 
validation of the other models would be unnecessary since they would not be used. Since the 
upper-limit release model is that of instantaneous release of all radionuclides, it does not require 
validation since it represents the maximum possible rate for TSPA analyses. If validation of the 
best-estimate and/or conservative degradation models for the individual DSNF groups is required 
for licensing of specific waste forms for repository emplacement, then those models would need 
to be validated using confirmatory test data per the requirements in ASTM C 1174-97. The 
criteria for validation would be the similarity or closeness of the confirmatory test data to the 
release rates predicted by the models. There is currently some limited information concerning the 
degradation of N-reactor and aluminum-based DSNF from the NSNFP-sponsored testing 
programs. Further NSNFP-sponsored confirmatory testing is either underway or planned for 
N-reactor, MOX, and aluminum-based DSNF. Testing for the degradation rates of samples from 
the other fuel groups is not currently anticipated. 
Therefore, the conservative and best-estimate models selected/abstracted in this document for the 
various DSNF waste-form groups have various degrees of technical data support in their 
generation. Only the Group 8 (U oxide) dissolution model will become a validated model 
(CRWMS M&O 2000a), but usage of this model for the rest of the DSNF in TSPA would be 
non-conservative. The other DSNF models would require qualification and/or confirmatory 
testing per the ASTM C 1174-97 logic if they were to be directly invoked to license specific 
waste forms for emplacement. Additionally, the available information shows that the only 
DSNF group that has dissolution rates greater than those of the Group 7 uranium metal SNF is 
the Group 3 (U/Pu carbide) SNF, and the inventory of this material is extremely small. For these 
reasons it is concluded that the Group 7 conservative and best-estimate dissolution models would 
likely bound the degradation rate of the total DSNF inventory. The supportive data based on the

 33  
dissolution testing sponsored by the NSNFP at PNNL would need to be qualified per 
AP-SIII.2Q, Section 5.5 of AP-3.10Q, or the model derived accepted under AP-3.15Q. 
The upper-limit model for the DSNF inventory would be that of complete release during one 
time step of the TSPA analysis. This model does not require qualification, validation, or 
confirmatory testing since it represents essentially the instantaneous mobilization of all the 
contents of a waste package containing DSNF. Only the total inventory of radionuclides in the 
inventory of DSNF would need to be qualified. Thus, the upper-limit model proposed for the 
DSNF and Pu disposal waste forms are impacted primarily by the total inventory of 
radionuclides that are present in the SNF. 
The conservative and best-estimate models abstracted herein for use in TSPA for the DSNF 
waste forms are primarily impacted by the validity of the uranium metal-based SNF dissolution 
models. The conservative and best-estimate models recommended herein for the degradation of 
the immobilized ceramic Pu waste form are primarily impacted by the validity of the titanateceramic 
degradation models (CRWMS M&O 2000c). 
If used as direct input to the TSPA, both the conservative and best estimate models (based on the 
degradation kinetics of the N-reactor SNF) need to be further validated. Validation would be 
done primarily by comparison of the degradation analysis results (i.e., the degradation rates 
predicted by the models) with accepted or qualified data from performance confirmation studies. 
These studies would probably primarily consist of NSNFP-sponsored flow-through testing at 
PNNL and unsaturated test data generated at ANL on N-reactor SNF specimens. The criterion 
for validation would be the closeness of the new test data to the best-estimate model predictions 
and the consequent verification that the conservative model(s) overpredict the experimental 
degradation rates. 
The other models, which may require validation, are the conservative and best-estimate models 
for the degradation of the immobilized Pu ceramic waste form and the naval fuel models. Tests 
underway concerning the dissolution of the immobilized Pu ceramic waste form are expected to 
produce data in support of the validation of the titanate ceramic model (see Section 7.2.2 below). 
Information in support of the demonstration that the degradation of the naval SNF is bounded by 
the CSNF model is to be provided by NNPP. 
Other performance confirmation studies that may produce information useful for helping validate 
these models could include the NSNFP-sponsored testing of MOX and aluminum-based DSNF 
and testing of N-reactor SNF at Hanford in support of the Hanford K-basin N-reactor SNF dry 
storage program. The testing of the MOX and Al-based DSNF could support validation by 
demonstrating that the degradation rate of these fuel forms is always less than that predicted 
from the N-reactor SNF-based models. 
7.1.2 Colloid Formation 
Compared to colloids formed from the corrosion of HLW glass, very little radionuclide-bearing 
colloidal material was detected in tests on the corrosion product of the uranium dioxide-based 
commercial LWR SNF (CRWMS M&O 2000b, Section 6.1.1.1). In particular, no colloids with 
embedded radionuclide phases similar to those observed in HLW glass tests were observed in the

 34  
much fewer number of CSNF tests. Even less information is currently available concerning the 
quantity or types of colloids formed from DSNF corrosion, but preliminary tests at ANL (Batt 
1999; Hurt 2000) indicate that colloid formation from metallic uranium-based DSNF may be 
significant. DSNF colloid formation and characterization testing has been initiated by NSNFP at 
ANL for N-reactor and MOX DSNF samples, but the results are still preliminary. For modeling 
purposes it will be assumed that only colloids with reversibly attached radionuclides are formed 
in the corrosion of the DSNF (CRWMS M&O 2000b, Section 6.3). This essentially means that 
DSNF colloids speciation and attachment/detachment kinetics will be similar to that of the CSNF 
colloids. This assumption is potentially non-conservative and will require validation as pertinent 
results of ongoing confirmatory tests become available. 
7.2 RECOMMENDED MODEL USAGE FOR TSPA 
7.2.1 DSNF 
The upper-limit model for each DSNF group and the Pu disposal waste form should be taken as 
complete release of radionuclides upon exposure to groundwater within one time step of the 
TSPA analysis. The upper-limit model should be used in the TSPA safety case since the 
conservative and best-estimate models are currently unqualified and require further test data for 
qualification. Use of a model other that the upper-limit model for TSPA could result in 
difficulties during subsequent efforts to demonstrate conformance to WAC requirements for 
waste form emplacement. Upper-limit models should also be used in cases where their usage in 
TSPA analyses results in acceptable boundary doses, and when other less conservative models 
are not needed. It should be noted that, as shown by the fractional release values in Table 3b, the 
release rates of the various waste forms on current TSPA time scales are very rapid for all the 
models, essentially releasing all material within a single time step. Future applications of the RIP 
TSPA codes may or may not require more refined models for the various DSNF waste forms. 
The conservative release models represent practical bounding cases for the degradation of a 
specific DSNF group or waste form. It is recommended that the N-reactor SNF conservative 
degradation model, with the weighted average DSNF effective exposed specific surface area of 
0.10 m2/g given in Section 6.4.3 and Table 3b and as adjusted for the relative density of the 
particular SNF matrix with respect to that of metallic uranium, be used to represent the entire 
DSNF inventory (with the exception of the naval SNF, which uses the CSNF model) for TSPA 
analyses in cases where PA personnel determine that a less conservative model than the upperlimit 
model is justified as follows: 
Rate in mg/cm2-hr = 1.26 x 1011 exp(-66,400/RTK)(rmatrix/rU-metal) 
where R is the gas constant (8.314 J/mol-K). (Eq. 5) 
where Dmatrix is the density of the spent fuel matrix, and DU-metal is the density of uranium metal. 
This model could be used in the TSPA analyses for the reasonable representation case. This 
model has not as yet been validated and may require confirmation through confirmatory testing 
per the requirements of ASTM C 1174-97 (1997).

 35  
The best-estimate degradation model represents a best estimate of the actual dissolution rate for 
the DSNF group. The best-estimate N-reactor SNF release-rate model (as adjusted for the 
relative density of the particular SNF with respect to that of metallic uranium) could be used to 
represent the release rate of the entire DSNF inventory (with the exception of the naval SNF) in 
TSPA cases that PA personnel determine require a more realistic representation of the actual 
degradation kinetics: 
Rate in mg/cm2-hr = 2.52 x 1010 exp(-66,400/RTK)(rmatrix/rU-metal) 
where R is the gas constant (8.314 J/mol-K). (Eq. 6) 
where Dmatrix is the density of the spent fuel matrix, and DU-metal is the density of uranium metal. 
The conservative and best estimate release models should not be used in safety case applications 
of the performance assessment codes since they are currently largely based on unvalidated 
data/analyses. Validation of the conservative and best-estimate U-metal dissolution models 
would require analysis per AP-SIII.2Q and may require further service condition and 
confirmatory testing per ASTM C 1174-97. Saturated flow-through testing at PNNL and 
unsaturated drip testing at ANL continues on N-reactor SNF that could be used to support the 
validation this model. 
It is expected that the application of the DSNF degradation models to the safety case for the 
TSPA-SR will involve usage of the upper-limit models or perhaps conservative models. 
Decisions as to exactly which models will be used will depend upon the sensitivity of the TSPA 
results to the models and the generation of dissolution test data sufficient to abstract and support 
the model. It is anticipated that the overall postclosure TSPA will be insensitive to the 
dissolution rate of the DSNF or Pu disposition waste forms (CRWMS M&O 2000f, Section 7.). 
The upper-limit group 7 (U metal) degradation model should also be used in the case of TSPA 
analyses wherein the metallic uranium-based SNF (primarily the N-reactor SNF) is assumed or 
analyzed to be pyrophoric in the emplacement condition. For bounding calculations, it should be 
assumed that the entire radionuclide inventory in the DOE SNF is released during the TSPA time 
step analyzed after waste package failure and contact with water/moisture. 
There are currently little data or information available concerning the speciation of the material 
released from the DSNF waste forms between solute and colloid forms. Some as yet incomplete 
information indicates that more colloids may form from the degradation of N-Reactor fuel 
(which will constitute approximately 80% of the DSNF inventory by weight) than for the CSNF 
and HLW glass waste forms under repository conditions (Gray and Einziger 1998, Section 4.2; 
Hurt 2000). It is currently planned to co-dispose all the N-reactor SNF canisters, called 
Multicanister Overpacks or MCOs, with HLW glass canisters in approximately 160 waste 
packages (two MCOs and two HLW glass canisters per waste package). Also, it is also currently 
planned that all DSNF (with the exception of the naval SNF and the immobilized Pu waste 
forms) canisters will be co-disposed with HLW glass in their waste packages (five HLW 
canisters with one DSNF canister in the center) (CRWMS M&O 2000d, Attachment I, Table 
I-1). Because of this co-disposal plan for DSNF and because the quantity of DSNF in the 
repository will only be a few percent by weight of the total SNF and HLW glass inventory, it is

 36  
recommended that for TSPA-SR, the colloid formation/migration dynamics recommended in 
CRWMS M&O (2000b), based on CSNF colloid behavior, be applied to the DSNF waste forms. 
In the event that the TSPA-SR becomes sensitive to this assumption, it must be validated or 
modified through further analysis or testing. Testing that can be used to support the validation of 
the use of the colloid model of CRWMS M&O (2000b) for DSNF is currently being performed 
at ANL (Mertz 2000). 
7.2.2 Immobilized Ceramic Pu Disposition Waste Form 
For the immobilized ceramic Pu disposal waste form, the upper-limit degradation model should 
be taken as the instantaneous release of all the Pu from the waste package upon postclosure 
contact with groundwater. Since this model represents a bounding upper case, it does not require 
validation for its usage. 
The conservative model for the degradation of the immobilized ceramic Pu disposition waste 
form should be taken as the dissolution rate of the borosilicate glass matrix with the assumption 
that the plutonium is homogeneously dispersed in the glass phase when contacted by 
groundwater. 
For the best-estimate case, the release of the immobilized ceramic Pu should be taken as the 
dissolution rate of the ceramic, taken as the high pH dissolution rate given in Shaw (1999) with 
an exposed surface area of the ceramic corresponding to the total geometric surface area of the 
Pu-containing ceramic disks in a waste canister. This estimate will still be somewhat 
conservative because it doesn’t take credit for the glass-matrix coverage of surface areas on the 
Pu ceramic disks. 
These conservative and best-estimate models for the immobilized ceramic Pu have not been 
validated. If used in the TSPA, they would require validation per AP-SIII.2Q and/or 
confirmatory testing per the requirements of ASTM C1174. There are tests currently being 
performed at ANL that could provide support for validation of these models if they are used in 
the TSPA.

 37  
Table 1b. DSNF, Naval SNF, Pu Disposition Release/Degradation Models 
Waste 
Type 
DSNF 
Group (1-1) 
Upper-limit Model Conservative Model Best-estimate Model 
Surrogate Model Ref. Surrogate Model Ref. Surrogate Model Ref. 
DSNF 1. Naval See 
Section 
6.3.1 
See Section 
6.3.1 
(1-7) See 
Section 
6.3.1 
See Section 6.3.1 (1-7) See 
Section 
6.3.1 
See Section 6.3.1 (1-7) 
DSNF 2. Pu /U 
Alloy 
Fermi Full release 
over TSPA time 
step 
N/A U – 6% Mo (bounding) 
10 X Unirradiated U-metal 
best-estimate 
Wet oxic conditions: 
R (kg/m2-s) = 1.88 x 104 exp 
(-7970/TK) 
Humid oxic conditions: 
R (kg/m2-s) = 0.27 x 103 exp 
(-7240/TK) 
(1-9) U - 8% Mo (semi-empirical) 
Unirradiated U-metal bestestimate 
Wet oxic conditions: 
R (kg/m2-s) = 1.88 x 103 
exp (-7970/TK) 
Humid oxic conditions: 
R (kg/m2-s) = 0.27 x 102 
exp (-7240/TK) 
(1-9) 
DSNF 3. Pu/U 
Carbide 
FFTF Full release 
over TSPA time 
step 
N/A UC2 (bounding) 
100 X Unirradiated U-metal 
best-estimate 
R (kg/m2-s) = 100 x [1.88 x 
103 exp (-7970/TK)] 
(1-8) UC2 (bounding) 
100 X Unirradiated U-metal 
best-estimate: 
R (kg/m2-s) = 100 x [1.88 x 
103 exp (-7970/TK)] 
(1-8) 
DSNF 4. MOX 
and Pu 
Oxide 
LWR SNF Full release 
over TSPA time 
step 
N/A LWR SNF (bounding) 
100 X Intact U Oxide bestestimate 
(1-6) UO2 (semi-empirical) 
U Oxide best-estimate 
model 
(1-5) 
DSNF 5. Th/U 
Carbide 
UC2 Full release 
over TSPA time 
step 
(1-4) UC2 (bounding) 
10 X Unirradiated U-metal 
best-estimate 
(1-4) SiC (semi-empirical) 
R (kg/m2-s) = 0.6 x 10-12 
(1-4) 
DSNF 6. Th/U 
Oxide 
FSV Full release 
over TSPA time 
step 
N/A ThO2 - 2 
wt% UO2 
(bounding) 
1000 X Best-estimate 
(1-3) Synroc (semi-empirical) 
k (mg/m2-d) = 82.0 x 10 (- 
1000/TK) 
(1-3) 
DSNF 7. UMetal- 
Based 
N-Reactor Full release 
over TSPA time 
step 
N/A N-reactor (semi-empirical) 
1.26 x 1011exp(-66400/RTK) 
mg/cm2-hr 
R = 8.314 J/mol-K 
(3) N-reactor (semi-empirical) 
2.52 x 1010exp(- 
66400/RTK) 
mg/cm2-hr 
R = 8.314 J/mol-K 
(2)

 38  
Waste 
Type 
DSNF 
Group (1-1) 
Upper-limit Model Conservative Model Best-estimate Model 
Surrogate Model Ref. Surrogate Model Ref. Surrogate Model Ref. 
DSNF 8a. Intact 
U Oxide 
LWR SNF Full release 
over TSPA time 
step 
N/A LWR SNF (bounding) 
100 X Intact U Oxide bestestimate 
(1-6) LWR SNF (semi-empirical) 
Commercial LWR SNF 
dissolution model(a) 
(1-6) 
DSNF 8b. 
Damaged 
U Oxide 
TMI Full release 
over TSPA time 
step 
TMI (bounding) 
100 X Intact U Oxide bestestimate 
TMI (bounding) 
100 X Intact U Oxide bestestimate 
DSNF 9. Albased 
FRR, ATR Full release 
over TSPA time 
step 
N/A U-Al alloy in 
bicarbonate 
solution 
(bounding) 
36 mgU/m2-d @ 25°C 
360 mgU/m2-d @ 90°C 
(5) SRS U-Al 
SNF in J- 
13 well 
water 
(empirical) 
0.22 mgU/m2-d @ 25°C 
2.20 mgU/m2-d @ 90°C 
(5) 
DSNF 10. 
Unknown 
N/A Full release 
over TSPA time 
step 
N/A N/A (bounding) 
10 X Unirradiated U-metal 
best-estimate 
(1-11) N/A (empirical) 
Unirradiated U-metal bestestimate 
(1-11) 
DSNF 11. U-Zr- 
Hx 
TRIGA Full release 
over TSPA time 
step 
N/A U-Zr-Hx (bounding) 
0.1 X U-oxide best-estimate 
(1-12) 
and 
(7) 
U-Zr-Hx (empirical) 
0.1 x U-oxide bestestimate 
(1-12) 
and 
(7) 
Pu N/A N/A Full release 
over TSPA time 
step 
N/A Multi-phase 
titanate 
ceramic 
10 times the best-estimate 
ceramic model (b) 
(4) 
and 
(6) 
Multiphase 
titanate 
ceramic 
(semi-empirical) 
(b) 
pH < 7 
Log10 R = -0.167 (pH) – 
4.66 
pH ³ 7 
Log10 R = 0.25 (pH) – 8 
R in g/m2-d 
(4) 
and 
(6) 
Notes: (bounding)–indicates a bounding model per the materials behavior modeling logic of ASTM C 1174-97 
(empirical)–indicates an empirical model per the materials behavior modeling logic of ASTM C 1174-97 
(semi-empirical)–indicates a semi-empirical model per the materials behavior modeling logic of ASTM C 1174-97 
(mechanistic)–indicates a mechanistic model per the materials behavior logic of ASTM C 1174-97 
(a) Commercial LWR SNF Dissolution Model 
Source: CRWMS M&O (2000a, Section 6.2.2.2, Equation 11 and Table 14) 
Log10 R (mg/m2-d) = 5.479057 + [-2457.050662 (1/TK)] + [1.5110878 (-log10(CO3))] + [-1.729906 (-log10O2)] + [0.234718 pH] + [-0.799526 log10 BU] + [400.755947 (-log10 
O2) (1/TK)] + [780.806133 (log10 BU) (1/TK)] + [0.172305 (log10 BU) (-log10(CO3))] + [0.174428 (log10BU) (-log10 O2)] + [-0.271203 (log10BU) (pH)] + [- 
0.339535 (-log10(CO3))2]

 39  
(b) The effect of enhanced surface area (over the geometric surface area of the ceramic disks) and radiation effects in the ceramic matrix is to increase the 
effective surface area for dissolution by a factor of 30 (Shaw 1999, Section 6,2) 
Sources/references: 
(1) DOE (1999) 
(1-1) Table 5-2 
(1-2) Not used 
(1-3) Section 6.6 and Figure 6-4 
(1-4) Section 6.5 and Figure 6-3 
(1-5) Section 6.4 
(1-6) Sections 6.4 and 6.8 
(1-7) Section 6.1 
(1-8) Section 6.3 
(1-9) Section 6.2 and Figure 6.1 
(1-10) Not used 
(1-11) Section 6.10 
(1-12) Section 6.11 
(2) Gray and Einziger (1998, Section 4.1, Equation 3) 
(3) Gray and Einziger (1998, Figures 3-13) 
(4) Shaw (1999) 
(5) Wiersma and Mickalonis (1998, Tables 3 and 4) 
(6) CRWMS M&O (2000c) 
(7) LMITCO (1997, Section 2.1.8)

 40  
Table 2. Summary and Comparison of N-Reactor SNF Stage 1 and Stage 2 Dissolution Rates (mg/m2-d) 
T(oC) CO3
-- 
(molar) 
pH HNO3 
(molar) 
J-13 well 
water 
Unirradiated UMetal 
Rate(6) 
Measured(1) 
Stage 1 Rate 
(mg/m2-d) 
Calculated(2) 
Stage 1 Rate 
(mg/m2-d) 
Measured (3) 
Maximum Stage 2 
Rate 
(mg/m2-d) 
25 2 x 10-2 5 -- -- 394 63 79 (5) 
25 2 x 10-2 8 -- -- 394 160 145 220 
25 2 x 10-4 10 -- -- 394 50 44 100 
75 2 x 10-2 10 -- -- 18379 2100 1851 4000 
75 2 x 10-4 5 -- -- 18379 200 142 1000 
75 2 x 10-4 8 -- -- 18379 150 250 180,000 
25 N/A 5 1 x 10-5 -- 394 38 -- (5) 
25 N/A 3 1 x 10-3 -- 394 130 -- (5) 
25 2 x 10-3 8.5 -- X 394 (4) 72 13000 
75 2 x 10-3 8.5 -- X 18379 (4) 614 55000 
(1) As given in Table 1 of Gray and Einziger (1998) 
(2) Calculated using Equation 1: log R (mg/m2-d) = 8.52 + 0.347 log[CO3
--] + 0.088 pH – 1929/TK 
(3) As estimated from the Cs release plots from Figures 7 through 15 of Gray and Einziger (1998) 
(4) These samples showed only Stage 2 dissolution behavior 
(5) These samples showed only Stage 1 dissolution behavior 
(6) Calculated from the expression for wet oxic condition corrosion: R = 1.88 x 103 exp (-7970/TK) kg/m2-s = 1.62 x 1014 exp (-7970/TK) mg/m2-d from DOE 
(1999)

 41  
Table 3a. Dissolution Rates at pH 8.5, 0.002M CO3
--, and 20% Oxygen Calculated from Proposed Models 
DSNF GROUP Approximate Fuel 
Matrix Density,r (b) 
in g/cm3 
(r/rGroup 7) 
Inventory 
(MTHM)(a) 
Best-estimate 
Corrosion 
Rate(c) @ 50oC 
(mg/m2-d) 
Best-estimate 
Corrosion 
Rate(c) @100oC 
(mg/m2-d) 
Conservative 
Corrosion 
Rate(c) @ 50oC 
(mg/m2-d) 
Conservative 
Corrosion 
Rate(c) @ 
100oC (mg/m2- 
d) 
Group 1–Naval SNF (see 6.3.1) 65 (see 6.3.1) (see 6.3.1) (see 6.3.1) (see 6.3.1) 
Group 2–Pu/U Alloy 19.05 (1) 8.5 3.1 x 103 8.5 x 104 3.1 x 104 8.5 x 105 
Group 3–Pu/U Carbide 11.28 (0.59) 0.1 1.84 x 105 5.04 x 106 1.84 x 105 5.04 x 106 
Group 4–MOX and PU Oxide 11.03 (0.58) 11.59 6.21 35.51 621 3551 
Group 5–Th/U Carbide 9.35 (0.49) 24.52 0.025 0.025 1.52 x 104 4.17 x 105 
Group 6–Th/U Oxide 9.87 (0.52) 46.98 0.034 0.088 34 88 
Group 7–U Metal-Based 19.05 (1) 1984.81 1.1 x 105 3.0 x 106 5.5 x 105 1.5 x 107 
Group 8a–Intact U Oxide(e) 9.86 (0.52) 166.2 (8a + 8b) 9.32 22.8 932 2280 
Group 8b–Damaged U Oxide 9.86 (0.52) -- 932 2280 932 2280 
Group 9–Al-Based 2.70 (0.14) 19.54 ~0.03 ~0.3 ~5 ~50 
Group 10–Unknown SNF 19.05(d) (1) 4.24 3.1 x 103 8.5 x 104 3.1 x 105 8.5 x 106 
Group 11–U-Zr-Hx 6.89 (0.36) 1.51 0.40 2.2 39 220 
Pu Immobilized Ceramic 5.5 (0.29) (g) N/A 0.011 (f) 0.31 (f) 0.11 (h) 3.1(h) 
Source:(a) DOE (1999, Appendix D, Radionuclide Inventory Summary of DOE SNF and HLW) 
(b) DOE (1999, Appendix C, Attachment A) 
(c) R = Rgroup7 x (D/Dgroup 7) 
(d) Conservatively taken to be that of metallic uranium 
(e) Equation 3, nominal burnup of 40 MWd/kgU 
(f) Shaw (1999, Section 6.1, Table 6) 
(g) CRWMS M&O (1998a, Section 2.2.1); note also that the estimated Pu content of the ceramic is 10.5% per Section 2.2.3 
(h) 10 times best estimate; Section 6.3.12

 42  
Table 3b. Fractional DSNF Waste Form Dissolution Rates at 50oC, pH 8.5, 0.002M CO3
--, and 20% Oxygen Calculated for Best-estimate Models 
DSNF GROUP Best-estimate Release 
Rate 
(mg/m2-d) 
Exposed Specific 
Surface Area 
(m2/g) (a) 
Exposed Total 
Surface Area 
(m2) (c) 
Fractional 
Corrosion Rate 
(d-1) 
Group 1–Naval SNF (see 6.3.1) See Section 6.3.1 (see 6.3.1) (see 6.3.1) 
Group 2–Pu/U Alloy l 3.1 x 103 1.2 x 10-3 2.04 x 104 3.7 x 10-3 
Group 3–Pu/U Carbide 1.84 x 105 2.7 x 10-3 5.40 x 102 0.84 
Group 4–MOX and PU Oxide 6.21 4.0 x 10-3 9.27 x 104 4.3 x 10-5 
Group 5–Th/U Carbide 0.025 2.2 x 10-2 1.08 x 106 1.1 x 10-6 
Group 6–Th/U Oxide 0.034 3.6 x 10-4 3.38 x 104 2.4 x 10-8 
Group 7–U Metal-Based 1.1 x 105 7.0 x 10-5 2.86 x 105 0.7 x 10-2 
Group 8a–Intact U Oxide(e) 9.37 4.0 x 10-1 1.33 x 108 4.3 x 10-3 
Group 8b–Damaged U Oxide 937 4.0 x 10-1 -- 4.3 x 10-1 
Group 9–Al-Based ~0.03 6.5 x 10-3 2.54 x 105 1.4 x 10-4 
Group 10–Unknown SNF 3.1 x 103 4.0 x 10-1 3.39 x 106 1.24 
Group 11–U-Zr-Hx 0.40 1.0 x 10-4 3.02 x 102 1.1 x 10-7 
Pu Immobilized Ceramic 1.33 x 10-3 2.5 x 10-5 (b) N/A 3.3 x 10-11 
Total l N/A N/A 1.38 x 108 N/A 
Source: (a) DOE (1999, Appendix B, Attachment A) 
(b) CRWMS M&O (1998a, Sections 2.2.1 and 2.2.2): 
surface area per disk = 
1 inch thick and 2.625 inches in diameter = [2p (2.625/2)2 + p(2.625)(1)] in2/disk [0.0254 m/in]2 = 0.0123 m2/disk 
weight per disk = (9755 g/can)/ (20 disks/can) = 487.75 g/disk 
so, specific surface area = (0.0123 m2/disk)/(487.75 g/disk) = 2.5 x 10-5 m2/g 
(c) Obtained by multiplying the specific surface area given in column 3 in m2 per gram by the inventory of the DSNF group 
given in Table 3a; using 1 MTHM = 2000 kg = 2,000,000 g. The total MTHM from column 3 of Table 3a is approximately 2333 
MTHM or 4.67 x 109 g. Thus the weighted average specific surface area is 1.38 x 108 m2/ 4.67 x 109 g @ 0.03 m2/g

 43  
8. INPUTS AND REFERENCES 
8.1 DOCUMENTS CITED 
Abrefah, J.; Buchanan, H.C.; Gerry, W.M.; Gray, W.J.; and Marschman, S.C. 1998. Dry Air 
Oxidation Kinetics of K-Basin Spent Nuclear Fuel. PNNL-11786. [Richland, Washington]: 
Pacific Northwest National Laboratory. TIC: 243094. 
Abrefah, J.; Gray, W.J.; Ketner, G.L.; Marschman, S.C.; Pyecha, T.D.; and Thornton, T.A. 1995. 
K-Basin Spent Nuclear Fuel Characterization Data Report. PNL-10778. Richland, Washington: 
Pacific Northwest National Laboratory. TIC: 243197. 
Abrefah, J.; Huang, F.H.; Gerry, W.M.; Gray, W.J.; Marschman, S.C.; and Thornton, T.A. 1999. 
Analysis of Ignition Testing on K-West Basin Fuel. PNNL-11816. Richland, Washington: Pacific 
Northwest National Laboratory. TIC: 248558. 
Batt, D.L. 1999. Minutes, Release Rate Test Team Meeting Minutes from November 4, 1999. 
CCN00-001277. Idaho Falls, Idaho: Idaho National Engineering and Environmental Laboratory. 
TIC: 248427. 
Bourcier, W.L. 1999. Interim Report on Development of a Model to Predict Dissolution 
Behavior of the Titanate Waste Form in a Repository and Compilation of Data from SPFT 
Ceramic Dissolution Tests. PIP-00-003. Livermore, California: Lawrence Livermore National 
Laboratory. TIC: 248554. 
CRWMS M&O (Civilian Radioactive Waste Management System Management and Operating 
Contractor) 1998a. Immobilized in Ceramic. Volume II of Report on Intact and Degraded 
Criticality for Selected Plutonium Waste Forms in a Geologic Repository. BBA000000-01717- 
5705-00020 REV 01. Las Vegas, Nevada: CRWMS M&O. ACC: MOL.19981217.0112. 
CRWMS M&O 1998b. "Waste Form Degradation, Radionuclide Mobilization, and Transport 
Through the Engineered Barrier System." Chapter 6 of Total System Performance Assessment- 
Viability Assessment (TSPA-VA) Analyses Technical Basis Document. B00000000-01717-4301- 
00006 REV 01. Las Vegas, Nevada: CRWMS M&O. ACC: MOL.19981008.0006. 
CRWMS M&O 1999a. Not Used. 
CRWMS M&O 1999b. Not Used. 
CRWMS M&O 1999c. Not Used. 
CRWMS M&O 1999d. Not Used. 
CRWMS M&O 1999e. Classification of the MGR Uncanistered Spent Nuclear Fuel Disposal 
Container System. ANL-UDC-SE-000001 REV 00. Las Vegas, Nevada: CRWMS M&O. ACC: 
MOL.19990928.0216.

 44  
CRWMS M&O 1999f. Not Used. 
CRWMS M&O 1999g. 1101213FM3 Waste Form Analyses & Models - PMR. Activity 
Evaluation, December 14, 1999. Las Vegas, Nevada: CRWMS M&O. ACC: 
MOL.19991217.0048. 
CRWMS M&O 2000a. CSNF Waste Form Degradation: Summary Abstraction. ANL-EBSMD-
000015 REV 00. Las Vegas, Nevada: CRWMS M&O. ACC: MOL.20000121.0161. 
CRWMS M&O 2000b. Waste Form Colloid-Associated Concentrations Limits: Abstraction and 
Summary. ANL-WIS-MD-000012 REV 00. Las Vegas, Nevada: CRWMS M&O. ACC: 
MOL.20000525.0397. 
CRWMS M&O 2000c. Defense High Level Waste Glass Degradation. ANL-EBS-MD-000016 
REV 00. Las Vegas, Nevada: CRWMS M&O. ACC: MOL.20000329.1183. 
CRWMS M&O 2000d. Inventory Abstraction. ANL-WIS-MD-000006 REV 00. Las Vegas, 
Nevada: CRWMS M&O. ACC: MOL.20000414.0643. 
CRWMS M&O 2000e. Waste Package Materials Department Analysis and Modeling Reports 
Supporting the Waste Form PMR. Development Plan TDP-EBS-MD-000005 REV 01. Las 
Vegas, Nevada: CRWMS M&O. ACC: MOL.20000202.0173. 
CRWMS M&O 2000f. Performance Assessment Sensitivity Analyses of Selected U.S. 
Department of Energy Spent Fuels. ANL-WIS-MD-000001 REV 00. Las Vegas, Nevada: 
CRWMS M&O. ACC: MOL.20000710.0530. 
CRWMS M&O 2000g. Waste Package Related Impacts of Plutonium Disposition Waste Forms 
in a Geologic Repository. TDR-EBS-MD-000003 REV 01 ICN 01. Las Vegas, Nevada: 
CRWMS M&O. ACC: MOL.20000510.0163. 
DOE (U.S. Department of Energy) 1998. NSNFP Release-Rate Test Program Plan. 
DOE/SNF/REP-022 Rev. 0. Washington, D.C.: Department of Energy. TIC: 245353. 
DOE (U.S. Department of Energy) 1999. DOE Spent Nuclear Fuel Information in Support of 
TSPA-SR. DOE/SNF/REP-0047, Rev. 0. Washington, D.C.: Department of Energy. TIC: 
245482. 
DOE (U.S. Department of Energy) 2000a. Not Used. 
DOE (U.S. Department of Energy) 2000b. DOE Spent Nuclear Fuel Grouping in Support of 
Criticality, DBE, TSPA-LA. DOE/SNF/REP-046 Rev.0, Idaho Falls, Idaho: U.S. Department of 
Energy, Idaho Operations Office. TIC: 248046

 45  
DOE (U.S. Department of Energy) 2000c. Quality Assurance Requirements and Description. 
DOE/RW-0333P, Rev. 10. Washington, D.C.: U.S. Department of Energy, Office of Civilian 
Radioactive Waste Management. ACC: MOL.20000427.0422 
Gray, W.J. 2000. Dissolution Rates of Aluminum-Based Spent Fuels Relevant to Geologic 
Disposal. PNNL-11979. Richland, Washington: Pacific Northwest National Laboratory. TIC: 
248761. 
Gray, W.J. and Einziger, R.E. 1998. Initial Results from Dissolution Rate Testing of N-Reactor 
Spent Fuel Over a Range of Potential Geologic Repository Aqueous Conditions. DOE/SNF/REP- 
022, Rev. 0. Washington, D.C.: U.S. Department of Energy. TIC: 245380. 
Hurt, W.L. 2000. May 11, 2000, Release Rate Test Review Meeting Minutes. CCN 00-012071. 
Idaho Falls, Idaho: INEEL, Bechtel Bwxt. TIC: 248756. 
LMITCO (Lockheed Martin Idaho Technologies Company) 1997. Uranium-Zirconium Hydride 
Fuels for TRIGA Reactors. UZR-28. Idaho Falls, Idaho: Idaho National Engineering and 
Environmental Laboratory. TIC: 245412. 
Mertz, C.J. 2000. Test Plan for the Characterization of Spent Nuclear Fuel Colloids. SNF-3A- 
121, Revision 0. [Argonne, Illinois]: Argonne National Laboratory. TIC: 247671. 
Mowbray, G.E. 2000. Revised Source Term for Naval Spent Nuclear Fuel and Special Case 
Waste Packages. Letter from G.E. Mowbray (DOD) to Dr. J.R. Dyer (DOE/YMSCO), 
September 6, 2000, with attachments. ACC: MOL.20000911.0354. 
Pasupathi, P. 2000. Metallic Uranium Dissolution Correlation from National Spent Nuclear Fuel 
Program Draft Report on the Oxidation Of Metallic Spent Fuel. Input Transmittal 00334.T. Las 
Vegas, Nevada: CRWMS M&O. ACC: MOL.20000831.0029. 
Shaw, H., ed. 1999. Plutonium Immobilization Project Input for Yucca Mountain Total Systems 
Performance Assessment. PIP-99-107. Livermore, California: Lawrence Livermore National 
Laboratory. TIC: 245437. 
Thornton, T.A. 1998. HPPP Issue 1; Preliminary TSPA for a Pyrophoric Event Involving NReactor 
SNF Waste Packages Interoffice correspondence from Thornton, T.A. (CRWMS M&O) 
to Clouet, J.S., Sareen S.S., and David Stahl, September 21, 1998 ACC: MOL.19981019.0001. 
Thornton, T.A. 1998. HPPP Issue 1; Preliminary TSPA for N-Reactor SNF Waste Packages. 
Interoffice correspondence from T.A. Thornton (CRWMS M&O) to J.S. Clouet, S.S. Sareen, and 
D. Stahl, June 1, 1998, LV.WP.TAT.05/98-111, with attachment. ACC: MOL.19980730.0142. 
Welsh, T.L.; Baker, R.B.; Hoppe, E.W.; Schmidt, A.J.; Abrefan, J.; Tingey, J.M.; Bredt, P.R.; 
Golcar, G.R.; and Makenas, B.J. 1997. Analysis of Sludge from Hanford K East Basin Canisters. 
HNF-SP-1201. [Richland, Washington]: Duke Engineering & Services Hanford. TIC: 248549.

 46  
Wiersma, B.J. and Mickalonis, J.I. 1998. Preliminary Report on the Dissolution Rate and 
Degradation of Aluminum Spent Nuclear Fuels in Repository Environments. WSRC-TR-98- 
00290. Aiken, South Carolina: Westinghouse Savannah River. TIC: 245382. 
8.2 CODES, STANDARDS, REGULATIONS, AND PROCEDURES 
AP-3.10Q, Rev. 2, ICN 3. Analyses and Models. Washington, D.C.: U.S. Department of Energy, 
Office of Civilian Radioactive Waste Management. ACC: MOL.20000918.0282. 
AP-3.15Q, Rev.1, ICN 2. Managing Technical Product Inputs. Washington, D.C.: U.S. 
Department of Energy, Office of Civilian Radioactive Waste Management. ACC: 
MOL.20000713.0363. 
AP-SIII.2Q, Rev. 0, ICN 2. Qualification of Unqualified Data and the Documentation of 
Rationale for Accepted Data. Las Vegas, Nevada: CRWMS M&O. ACC: 
MOL.19991214.0625. 
ASTM C 1174-97. 1997. Standard Practice for Prediction of the Long-Term Behavior of 
Materials, Including Waste Forms, Used in Engineered Barrier Systems (EBS) for Geological 
Disposal of High-Level Radioactive Waste. West Conshohocken, Pennsylvania: American 
Society for Testing and Materials. TIC: 246015.