21-PWR Waste Package with Absorber Plates Loading Curve Evaluation REV 00C, ICN 00

CAL-DSU-NU-000006

December 2004

1. PURPOSE
The objective of this calculation is to evaluate the required minimum burnup as a function of initial 
pressurized water reactor (PWR) assembly enrichment that would permit loading of spent nuclear 
fuel into the 21 PWR waste package with absorber plates design as provided in Attachment IV. This 
calculation is an example of the application of the methodology presented in the Disposal Criticality 
Analysis Methodology Topical Report (YMP 2003). The scope of this calculation covers a range 
of enrichments from 0 through 5.0 weight percent U-235, and a burnup range of 0 through 45 
GWd/MTU. Higher burnups were not necessary because 45 GWd/MTU was high enough for the 
loading curve determination. 
This activity supports the validation of the use of burnup credit for commercial spent nuclear fuel 
applications. The intended use of these results will be in establishing PWR waste package 
configuration loading specifications. 
Limitations of this evaluation are as follows: 
• 
The results are based on burnup credit for actinides and selected fission products as proposed 
in YMP (2003, Table 3-1) and referred to as the Principal Isotopes. Any change to the isotope 
listing will have a direct impact on the results of this report. 
• 
The results are based on 1.5 wt% Gd in the Ni-Gd Alloy material and having no tuff inside the 
waste package. If the Gd loading is reduced or a process to introduce tuff inside the waste 
package is defined, then this report would need to be reevaluated based on the alternative 
materials. 
This calculation is subject to the Quality Assurance Requirements and Description (QARD) (DOE 
2004) because it concerns engineered barriers that are included in the Q-List (BSC 2004k, 
Appendix A) as items important to safety and waste isolation. 

2. METHOD
The method used to perform the reactivity calculations involves the simulation of the burnup and 
decay of fuel assemblies, for various initial enrichments and spent nuclear fuel (SNF) burnups, and 
the calculation of keff (effective neutron multiplication factor) for the loaded waste package 
configuration. The isotopic compositions for SNF were calculated in Isotopic Generation and 
Confirmation of the PWR Application Model (BSC 2003b) and used as input to Software Code: 
MCNP (CRWMS M&O 1998e) to calculate keff for the waste package loaded with various 
burnup/enrichment pairs. The keff calculations are based on taking credit for burnup with a subset 
of the total isotopes present in commercial SNF known as the Principal Isotopes (YMP 2003, 
Table 3-1). 
The keff calculations were performed using continuous-energy neutron cross-section libraries as 
selected in Selection of MCNP Cross Section Libraries report (CRWMS M&O 1998b, pp. 61-68). 
The SNF from the various burnup/enrichment pairs were simulated, and the results reported from 
the MCNP calculations were the combined average values of keff from three estimates (collision, 
absorption, and track length) listed in the final generation summary in the MCNP output. Each of 
the waste package configurations was represented in detail using specifications for the Babcock & 
Wilcox (B&W) 15x15 assembly design (Summary Report of Commercial Reactor Criticality Data 
for Crystal River Unit 3 [Punatar 2001, Section 2]), and waste package dimensions provided in 
Attachment IV from the following references: Design and Engineering, 21 PWR A, B, D & E Fuel 
Plates (BSC 2004a), Design and Engineering, 21 PWR C Fuel Plate (BSC 2004b), Design and 
Engineering, 21 PWR Corner Guide (BSC 2004c), Design and Engineering, 21 PWR End Side 
Guide (BSC 2004d), Design and Engineering, 21 PWR Side Guide (BSC 2004e), Design and 
Engineering, 21-PWR Waste Package Configuration (BSC 2004f), Design and Engineering, 21PWR 
Waste Package Configuration (BSC 2004g), and Design and Engineering, Fuel Tube (BSC 
2004h). 
3. ASSUMPTIONS 
3.1 ASSEMBLY DESIGN 
Assumption: It is assumed that the B&W 15x15 assembly design is the most limiting PWR fuel 
assembly design. 
Rationale: The basis for this assumption is that a previous analysis for the BR-100 transportation 
cask established the B&W 15x15 fuel assembly as one of the most reactive fuel assembly designs 
(B&W Fuel Company 1991, p. II 6-6). In addition, several assembly designs were evaluated in 
Attachment II and the results show the B&W 15x15 design to be the most reactive. 
Confirmation Status: This assumption requires no further confirmation based on the stated rationale. 
Use in the Calculation: This assumption was used in Section 5. 

3.2 MATERIAL DEGRADATION PRODUCTS 
Assumption: It was assumed that all of the iron within the waste package internal components turns 
into hematite (Fe2O3) or goethite (FeOOH) as it oxidizes and hydrates, and all of the aluminum turns 
into gibbsite (Al[OH]3). 
Rationale: The basis for this assumption is that these are considered the primary minerals that iron 
and aluminum will form, and are substantiated by Geochemistry Model Abstraction and Sensitivity 
Studies for the 21 PWR CSNF Waste Packages (BSC 2004i, Sections 6.3.2 and 6.4.2). 
Confirmation Status: This assumption requires no further confirmation based on the stated rationale. 
Use in the Calculation: This assumption is used in the sensitivity studies in Attachment I and the 
postclosure bounding configuration cases in Section 5. 
3.3 RANGE OF APPLICABILITY VERSUS RANGE OF PARAMETERS 
Assumption: It is assumed that range of parameters (ROP) for the configurations evaluated in this 
report match the range of applicability (ROA) of the benchmarks used to establish the lower bound 
tolerance limits provided by Framatome ANP (2003). 
Rationale: The basis for this assumption is that the ROP concerning materials, geometry, and 
neutron spectrum for the given configurations are considered sufficiently similar to the benchmark 
experiment parameters. 
Confirmation Status: This assumption requires no further confirmation based on the stated rationale. 
Use in the Calculation: This assumption is used in Section 6 and Attachment I. 
3.4 CROSS SECTION SUBSTITUTION 
Assumption: Since the zinc cross section libraries are unavailable, it was assumed that representing 
the zinc material composition in SB-209 A96061 T4 as aluminum would maintain the same 
neutronic characteristics. 
Rationale: The rationale for this assumption is that the nuclear cross-sections for these two elements 
are sufficiently similar (Parrington et al. 1996, pp. 20 and 24). 
Confirmation Status: This assumption requires no further confirmation based on the stated rationale. 
Use in the Calculation: This assumption was used in Section 5.2.1. 

3.5 RADIAL PROFILE EFFECTS 
Assumption: It was assumed that the radial variation in burnup within the spent fuel assemblies is 
accounted for during the depletion calculations for BSC (2003b). 
Rationale: The basis for this assumption is that BSC (2003b) uses inserted burnable absorbers and 
other parameters during the depletion to harden the neutron spectrum and represent heterogeneous 
effects, which increases the buildup of plutonium and thus increases the fuel assembly's reactivity 
worth. This method of depletion is confirmed in BSC (2003b) to provide conservative results with 
comparisons against measured values. 
Confirmation Status: This assumption requires no further confirmation based on the stated rationale. 
Use in the Calculation: This assumption is used in Section 5. 
3.6 DUST MASS COLLECTED ON WASTE PACKAGE OUTER BARRIER 
Assumption: It is assumed that up to 20 kg of dust per waste package will collect on the outer 
barrier following closure of the repository. 
Rationale: During the ventilation period prior to sealing the primary entrances to the repository, dust 
will accumulate on the waste package outer barrier surfaces (Total Dust Settling on Naval Long 
Waste Packages in 100 Years [BSC 2004l, Section 1]). The source of a majority of this dust is 
expected to be from the outside air being used for the ventilation. Rock dust from construction 
activities is expected to be removed prior to waste package emplacement. Preliminary results 
BSC (2004l, Table 4) indicate that the maximum mass of dust per waste package accumulated in 100 
years is approximately 18.2 kg. Thus, the assumption is conservative. 
Confirmation Status: This assumption requires no further confirmation based on the stated rationale. 
Use in the Calculation: This assumption is used in Attachment I. 
3.7 HYDRAULIC FLUID COMPOSITION 
Assumption: It was assumed that the hydraulic fluid used as an alternative moderator material was 
a conventional silicone fluid (polysiloxane fluid) with a viscosity of 10cSt with a degree of 
polymerization of four (which is necessary for a viscosity of 10 cSt at 25°C (Gelest Inc. 2004, p.11). 
Rationale: The basis for this assumption is that this material is a common hydraulic fluid (Gelest 
Inc. 2004, p. 7). 
Confirmation Status: This assumption requires no further confirmation based on the stated rationale. 
Use in the Calculation: This assumption is used in Attachment I. 

3.8 IRON AND ALUMINUM OXIDE VOLUME EXPANSION 
Assumption: It was assumed that the volume expansion from the oxidation and hydration of the iron 
in carbon steel and aluminum in SB-209 A96061 T4 components followed the ratio of theoretical 
densities. 
Rationale: The basis for this assumption is that the internal components can degrade over time and 
the amount present and volume will vary as the exposed region oxidizes and hydrides before flaking 
off. This assumption is used in representations to capture these effects on system reactivity for 
various degrees of degradation with retention of various amounts of corrosion products. 
Confirmation Status: This assumption requires no further confirmation since an adequate range for 
various amounts of degradation and volume occupied have been evaluated in Attachment I. 
Use in the Calculation: This assumption is used in Attachment IV. 
3.9 DIMENSIONAL SCALING FACTORS 
Assumption: The following scaling factors were assumed for purposes of scaling dimensions from 
the drawings in Characteristics of Spent Fuel, High-Level Waste, and Other Radioactive Wastes 
Which May Require Long-Term Isolation (DOE 1987): 
1) 5 mm is equal to 0.937 in. for the Combustion Engineering (CE) 14x14 assembly design.
2) 12.5 mm is equal to 1.8125 in. for the CE 15x15 assembly design.
3) 5 mm is equal to 0.891 in. for the CE 16x16 assembly design.
Rationale: The basis for this assumption is that the dimensions were scaled from DOE (1987,
Figures 1-4, 1-1, and 1-9, respectively).
Confirmation Status: This assumption requires no further confirmation since minor variations in the
scaled dimensions will have a negligible effect on system reactivity. The effect is considered
negligible as long as the materials for the scaled regions are represented in some capacity.
Use in the Calculation: This assumption is used in Attachment II.
3.10 CE 15X15 ASSEMBLY DESIGN INSTRUMENT TUBE 
Assumption: It was assumed that the instrument tube outer diameter and cladding thickness in the 
CE 15x15 assembly design was the same as that listed for the CE 16x16 assembly design. 
Rationale: The basis for this assumption is that the designs are similar and manufactured by the 
same vendor. 
Confirmation Status: This assumption requires no further confirmation since minor variations in the 
dimensions will have a negligible effect on system reactivity. The effect is considered negligible 
as long as the instrument tube is represented. 

Use in the Calculation: This assumption is used in Attachment II. 
3.11 CE NICKEL ALLOY 
Assumption: It was assumed that CE nickel alloy is similar to Inconel 718. 
Rationale: The basis for this assumption is that fuel assembly design information for Babcock & 
Wilcox and Westinghouse assemblies typically use Inconel 718 in the end-fittings. 
Confirmation Status: This assumption requires no further confirmation since axial reflection from 
the end-fitting region has a negligible effect on system reactivity. The effect is considered negligible 
as long as a similar type of material is represented in this region. 
Use in the Calculation: This assumption is used in Attachment II. 
4. USE OF COMPUTER SOFTWARE AND MODELS 
4.1 MCNP 
The baselined MCNP code (CRWMS M&O 1998e) was used to calculate the neutron multiplication 
factor for the various spent fuel compositions. The software specifications are as follows: 
• 
Software Title: MCNP 
• 
Version/Revision Number: Version 4B2LV 
• 
Status/Operating System: Qualified/HP-UX B.10.20 
• 
Software Tracking Number: 30033 V4B2LV (Computer Software Configuration Item 
Number) 
• 
Computer Type: Hewlett Packard 9000 Series Workstations 
• 
Computer Processing Unit number: 700887 
The input and output files for the MCNP calculations are contained on a compact disc attachment 
to this calculation report (Attachment IV) as described in Sections 5 and 8, such that an independent 
repetition of the software use may be performed. The MCNP software used was (1) appropriate for 
the application of multiplication factor calculations, (2) used only within the range of validation as 
documented throughout Briesmeister (1997) and Software Qualification Report for MCNP Version 
4B2, A General Monte Carlo N-Particle Transport Code (CRWMS M&O 1998a), and (3) obtained 
from Software Configuration Management in accordance with appropriate procedures. 
4.2 EXCEL 
• 
Software Title: Excel 
• 
Version/Revision number: Microsoft® Excel 97 SR-2 
.
• Computer Environment: Software is installed on a DELL OptiPlex GX240 personal computer, 
Civilian Radioactive Waste Management System (CRWMS) Management and Operating 
Contractor (M&O) tag number 150527, running Microsoft Windows 2000, Service Pack 4. 

Microsoft Excel for Windows, Version 1997 SR-2, is used in calculations and analysis to manipulate 
the inputs using standard mathematical expressions and operations. It is also used to tabulate and 
chart results. The user-defined formulas, inputs, and results are documented in sufficient detail to 
allow an independent repetition of computations. Thus, Microsoft Excel is used only as a worksheet 
and not as a software routine. Microsoft Excel 1997 SR-2 is an exempt software product in 
accordance with Software Management (LP-SI.11Q-BSC, Subsection 2). 
The spreadsheet files for the Excel calculations are documented in Attachment IV. 
5. CALCULATION 
This report evaluates the minimum required burnup of an assembly, for a specific initial enrichment, 
at which the calculated keff is equal to the critical limit (CL). The CL is the value of keff at which 
the configuration is potentially critical, and accounts for the criticality analysis methodology bias 
and uncertainty. In equation notation the CL is represented as shown in Equation 1.
CL(x) = f(x) -.kEROA -.kISO -.km (Eq. 1) 
where 
x = a neutronic parameter used for trending 
f(x) = the lower bound tolerance limit function accounting for biases and uncertainties 
that cause the calculation results to deviate from the true value of keff for a critical 
experiment, as reflected over an appropriate set of critical experiments 
.kEROA = penalty for extending the range of applicability 
.kISO = penalty for isotopic composition bias and uncertainty 
.km = an arbitrary margin to ensure subcriticality for preclosure and turns the CL function 
into an upper subcritical limit (USL) function (it is not applicable for use in 
postclosure analyses because there is no risk associated with a subcritical event) 
A more detailed discussion of the CL calculation is provided in YMP (2003, Section 3.5.3). A series 
of computer calculations were performed in order to develop a set of curves which show keff versus 
burnup for different initial enrichments, and the minimum burnup required to reach the CL or USL. 
A burnup credit loading curve depicts the relationship between the initial enrichment of a fuel 
assembly and the required minimum burnup needed to suppress the reactivity of that fuel assembly 
sufficiently to allow it to be safely loaded into the waste package. Any assembly whose burnup 
exceeds the required minimum burnup, given the initial enrichment of the fuel assembly, may be 
loaded into the waste package. 
There are two time periods to consider for applicability of a loading curve - preclosure and 
postclosure. The preclosure time-period is the period before permanent closure of the repository and 
includes the operations involving handling, loading, and sealing of the waste packages. During the 
preclosure time period it is currently required that the system be designed such that the calculated 
keff be sufficiently below unity to show at least a five percent margin after allowance for the bias in 
the method of calculation, and the uncertainty in the experiments used to validate the method of 
calculation (Project Design Criteria Document [Doraswamy 2004, Section 4.9.2.2]). The 
postclosure time-period is the period after permanent closure of the repository throughout the 

10,000-year regulatory period (10 CFR 63.2). During the postclosure time-period a variety of 
conditions may affect the waste package internal configurations. A process to identify configuration 
classes that have the potential for criticality is provided in YMP (2003, Section 3.6). YMP (2003) 
is the source for the postclosure methodology (Project Requirements Document [Canori and Leitner 
2003, PRD-013/T-016 and PRD-013/T-038]). This report provides a limited search for potential 
configurations that provide the highest keff values. 
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Title: 21-PWR Waste Package with Absorber Plates Loading Curve Evaluation 
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5.1 PARAMETER DESCRIPTION 
5.1.1 Burnup Profiles 
A bounding profile is defined as one that would maximize fuel assembly reactivity. Thus, a truly 
bounding profile would be where the fuel has not been irradiated, which is referred to as the “fresh 
fuel” assumption. This “fresh fuel” assumption is very conservative in calculations of criticality 
potential. As fuel is burned in a reactor, the burnup of the fuel becomes distributed axially and the 
reactivity of the fuel decreases. The profile of this axial distribution attains a flattened cosine shape 
with time, although the exact profile will vary significantly with operating history and other effects 
unique to the individual reactor. An axial profile database has been composed for various PWR fuel 
assembly designs, which included variations in enrichment, burnup, and burnable absorbers 
(Cacciapouti and Van Volkinburg, 1997). To develop a waste package loading curve, which would 
encompass the isotopic axial variations caused by different assembly irradiation histories, requires 
the development of a limiting axial profile that takes credit for fuel burnup. 
The axial profiles used in this calculation were developed from an axial profile database 
(Cacciapouti and Van Volkinburg 1997). Limiting axial profiles were developed for a set of eight 
burnup groups in PWR Axial Burnup Profile Analysis (BSC 2003a, Table 32) as listed in Table 1. 
Radial profiles are accounted for during the fuel depletion calculations (see Assumption 3.5). 
Table 1. Seven-Node Limiting Axial Burnup Profiles by Group 
Axial 
Pos.a 
Group 1b 
10 = x < 15 
Group 2b 
15 = x < 20 
Group 3b 
20 = x < 25 
Group 4b 
25 = x < 30 
Group 5b 
30 = x < 35 
Group 6b 
35 = x < 40 
Group 7b 
40 = x < 45 
Group 8b 
45 = x 
0.028 0.497 0.554 0.525 0.587 0.599 0.619 0.635 0.640 
0.083 0.837 0.882 0.882 0.903 0.907 0.914 0.923 0.926 
0.139 1.009 1.021 0.986 1.016 1.028 1.027 1.024 1.023 
0.194 – 
0.806 1.150 1.126 1.122 1.113 1.104 1.094 1.094 1.086 
0.861 0.859 0.920 0.934 0.950 0.971 0.988 0.991 0.998 
0.917 0.664 0.694 0.766 0.738 0.778 0.822 0.800 0.841 
0.972 0.333 0.416 0.440 0.454 0.471 0.506 0.502 0.541 
NOTES: ab 
Axial Pos. is percent of core height from bottom to top 
Burnup ranges are in units of GWd/MTU 

5.2 INPUT PARAMETER DESCRIPTION 
Sensitivity studies provided in Attachment I were used as the basis for the selection of parameters 
that maximize the resultant keff values. 
5.3 MATERIALS 
This section provides an overview of the materials that were selected for use in the MCNP inputs. 
5.3.1 Tuff Material Description 
Waste package configurations were represented with a tuff reflector since this is the composition 
of the drift material. The base components of the tuff composition used for the loading curve 
determinations is presented in Table 2 with the derived material specifications for the input files 
presented in Attachment IV (workbook Tuff composition.xls, sheet Latest_Tuff). 
Table 2. Tuff Material Composition 
Compound Wt% 
SiO2 76.29 
Al2O3 12.55 
FeO 0.14 
Fe2O3 0.97 
MgO 0.13 
CaO 0.5 
Na2O 3.52 
K2O 4.83 
TiO2 0.11 
P2O5 0.05 
MnO 0.07 
Source: DTN:GS000308313211.001, mean values from file zz_sep_249138.txt 
NOTE: Derived elemental/isotopic number densities for MCNP inputs are provided in Attachment IV, spreadsheet 
Tuff composition.xls, sheet Latest_Tuff 
5.3.2 Waste Package MCNP Material Descriptions 
The waste package representation for the MCNP calculations follows the description as that shown 
in Attachment IV. The outer barrier of the waste package was represented as SB-575 N06022, 
which is a specific type of nickel-based alloy as described in Table 3. The inner barrier was 
represented as SA-240 S31600, which is nuclear grade 316 stainless steel (SS) with tightened 
control on carbon and nitrogen content (ASM International 1987, p. 931; and ASME 2001, Section 
II, SA-240, Table 1) as described in Table 4. The fuel basket plates were represented as Ni-Gd 
Alloy (Unified Numbering System [UNS] designation is UNS N06464) with 1.5 wt% Gd as 
described in Table 5, and the thermal shunts were represented as SB-209 A96061 T4 (aluminum 
6061) as described in Table 6. The basket side and corner guides, and the basket stiffeners were 
represented as Grade 70 A 516 carbon steel as described in Table 7. Stiffeners were placed 

equidistant along the length of the basket in eight axial locations. Waste package basket material 
thicknesses were taken from the PWR drawings in Attachment IV. 
The chromium, nickel, and iron elemental weight percents obtained from the references were 
expanded into their constituent natural isotopic weight percents for use in MCNP. This expansion 
was performed by: (1) calculating a natural weight fraction of each isotope in the elemental state, 
and (2) multiplying the elemental weight percent in the material of interest by the natural weight 
fraction of the isotope in the elemental state to obtain the weight percent of the isotope in the 
material of interest. This process is described mathematically in Equations 2 and 3. The atomic 
mass values and atom percent of natural element values for these calculations are from Parrington 
et al. (1996). 
WFi = IA ( At% ) (Eq. 2)
ii 
. A (At% )
ii 
i =1 
where 
WFi = the weight fraction of isotopei in the natural element 
Ai = the atomic mass of isotopei 
At%i = the atom percent of isotopei in the natural element 
I = the total number of isotopes in the natural element 
Wt%i = WF (E ) (Eq. 3)
i wt % 
where 
Wt%i = the weight percent of isotopei in the material composition 
WFi = the weight fraction of isotopei from Equation 2 
Ewt% = the referenced weight percent of the element in the material composition 

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Table 3. SB-575 N06022 Material Composition 
Element/ 
Isotope ZAIDa Wt% 
Element/ 
Isotope ZAID Wt% 
C-nat 6000.50c 0.0150 Co-59 27059.50c 2.5000 
Mn-55 25055.50c 0.5000 W-182b 74182.55c 0.7877 
Si-nat 14000.50c 0.0800 W-183b 74183.55c 0.4278 
Cr-50 24050.60c 0.8879 W-184b 74184.55c 0.9209 
Cr-52 24052.60c 17.7863 W-186b 74186.55c 0.8636 
Cr-53 24053.60c 2.0554 V 23000.50c 0.3500 
Cr-54 24054.60c 0.5202 Fe-54 26054.60c 0.2260 
Ni-58 28058.60c 36.8024 Fe-56 26056.60c 3.6759 
Ni-60 28060.60c 14.6621 Fe-57 26057.60c 0.0865 
Ni-61 28061.60c 0.6481 Fe-58 26058.60c 0.0116 
Ni-62 28062.60c 2.0975 S-32 16032.50c 0.0200 
Ni-64 28064.60c 0.5547 P-31 15031.50c 0.0200 
Mo-nat 42000.50c 13.5000 Density = 8.69 g/cm3 
Source: DTN: MO0003RIB00071.000 
a
NOTES: ZAID = MCNP material identifier 
b 
W-180 cross section libraries are not available so the atom percents of the remaining isotopes were used 
to renormalize the elemental weight and derive isotopic weight percents excluding the negligible 0.120 atom 
percent in nature contribution from W-180. 
Table 4. Material Specifications for SA-240 S31600 
Element/Isotope ZAIDa Wt% Element/Isotope ZAID Wt% 
C-natb 6000.50c 0.0200 Fe-54 26054.60c 3.6911 
N-14b 7014.50c 0.0800 Fe-56 26056.60c 60.0322 
Si-nat 14000.50c 1.0000 Fe-57 26057.60c 1.4119 
P-31 15031.50c 0.0450 Fe-58 26058.60c 0.1897 
S-32 16032.50c 0.0300 Ni-58 28058.60c 8.0641 
Cr-50 24050.60c 0.7103 Ni-60 28060.60c 3.2127 
Cr-52 24052.60c 14.2291 Ni-61 28061.60c 0.1420 
Cr-53 24053.60c 1.6443 Ni-62 28062.60c 0.4596 
Cr-54 24054.60c 0.4162 Ni-64 28064.60c 0.1216 
Mn-55 25055.50c 2.0000 Mo-nat 42000.50c 2.5000 
Densityc = 7.98 g/cm3 
NOTES: a ZAID = MCNP material identifier 
b 
Carbon and nitrogen specifications are from ASM International (1987, p. 931) and remaining material 
compositions are from ASM International (1990b p. 843) 
c 
Density is for stainless steel 316 from ASTM (G 1-90, p. 7, Table X1) 

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Table 5. Material Specifications for Ni-Gd Alloy (UNS N06464) with 1.5 wt% Gdb 
Element/Isotope ZAIDa Wt% Element/Isotope ZAID Wt% 
C-nat 6000.50c 0.0100 Gd-152 64152.50c 0.0029 
Mn-55 25055.50c 0.5000 Gd-154 64154.50c 0.0320 
Si-nat 14000.50c 0.0800 Gd-155 64155.50c 0.2187 
Cr-50 24050.60c 0.6602 Gd-156 64156.50c 0.3045 
Cr-52 24052.60c 13.2247 Gd-157 64157.50c 0.2343 
Cr-53 24053.60c 1.5283 Gd-158 64158.50c 0.3742 
Cr-54 24054.60c 0.3868 Gd-160 64160.50c 0.3335 
Ni-58 28058.60c 43.3679 Fe-54 26054.60c 0.0565 
Ni-60 28060.60c 17.2778 Fe-56 26056.60c 0.9190 
Ni-61 28061.60c 0.7637 Fe-57 26057.60c 0.0216 
Ni-62 28062.60c 2.4717 Fe-58 26058.60c 0.0029 
Ni-64 28064.60c 0.6537 S-32 16032.50c 0.0050 
Mo-nat 42000.50c 14.5500 P-31 15031.50c 0.0050 
Co-59 27059.50c 2.0000 O-16 8016.50c 0.0050 
Density = 8.76 g/cm3 
Source: ASTM B 932-04, Table 1 and Section 8 
a
NOTE: ZAID = MCNP material identifier 
b 
1.5wt% Gd is based on typical value of 75% credit (NRC 2000, p. 8-4) allowed for fixed neutron absorbers 
and a nominal Gd loading of 2.0 wt% for Ni-Gd Alloy 
Table 6. Material Specifications for SB-209 A96061 T4 
Element/Isotope ZAIDa Wt% Element/Isotope ZAID Wt% 
Si-nat 14000.50c 0.6000 Mg-nat 12000.50c 1.0000 
Fe-54 26054.60c 0.0396 Cr-50 24050.60c 0.0081 
Fe-56 26056.60c 0.6433 Cr-52 24052.60c 0.1632 
Fe-57 26057.60c 0.0151 Cr-53 24053.60c 0.0189 
Fe-58 26058.60c 0.0020 Cr-54 24054.60c 0.0048 
Cu-63 29063.60c 0.1884 Ti-nat 22000.50c 0.1500 
Cu-65 29065.60c 0.0866 Al-27b 13027.50c 96.9300 
Mn-55 25055.50c 0.1500 Densityc = 2.7065 g/cm3 
Source: ASM International 1990a, p. 102 
a
NOTES: ZAID = MCNP material identifier. 
b 
Zn cross-section data unavailable; therefore, it was substituted as Al-27 (See Assumption 3.4). 
c 
ASTM G 1-90 1999, p. 7, Table X1 indicates 2.7 g/cm3; ASME 2001, Section II, Table NF-2 indicates a 
converted value from 0.098 lb/in3 of 2.713 g/cm3; therefore the midpoint was used. 

Table 7. Grade 70 A516 Carbon Steel Composition 
Element/Isotope ZAIDa Wt% Element/Isotope ZAID Wt% 
C-nat 6000.50c 0.2700 Fe-54 26054.60c 5.5558 
Mn-55 25055.50c 1.0450 Fe-56 26056.60c 90.3584 
P-31 15031.50c 0.0350 Fe-57 26057.60c 2.1252 
S-32 16032.50c 0.0350 Fe-58 26058.60c 0.2856 
Si-nat 14000.50c 0.2900 Density = 7.850 g/cm3 
Source: ASTM A 516/A 516M-90 1991, p. 2, Table 1; density from ASME 2001, Sec II, Part A, SA-20, Section 14.1 
NOTE: a ZAID = MCNP material identifier 
5.3.3 Fuel Assembly MCNP Material Descriptions 
The fuel assembly materials listed in this section refer to the upper and lower end-fitting materials, 
the cladding, and fuel plenum materials. In order to simplify the geometry the spacer grids were 
omitted from the MCNP representations. This is considered conservative with respect to criticality 
calculations for under-moderated lattices because there is less moderator displacement thereby 
increasing the moderator effectiveness where the spacer grids would normally be. 
The cladding composition was Zircaloy-4 as presented in Table 8.
Table 8. Zircaloy-4 Material Composition 
Element/Isotope ZAIDa Wt% Element/Isotope ZAID Wt% 
Cr-50 24050.60c 0.0042 Fe-57 26057.60c 0.0045 
Cr-52 24052.60c 0.0837 Fe-58 26058.60c 0.0006 
Cr-53 24053.60c 0.0097 O-16 8016.50c 0.1250 
Cr-54 24054.60c 0.0024 Zr-nat 40000.60c 98.1150 
Fe-54 26054.60c 0.0119 Sn-nat 50000.35c 1.4500 
Fe-56 26056.60c 0.1930 Densityb = 6.56 g/cm3 
Source: ASTM B 811-97 2000, p. 2, Table 2 
NOTES: a ZAID = MCNP material identifier. 
b From ASM International 1990a, p. 666, Table 6. 
The primary material components in the upper and lower end-fitting regions are SS304 (Table 9), 
Inconel (represented as Inconel-718 as shown in Table 10), Zircaloy-4 as represented in Table 8, and 
moderator (represented as water at 1.0 g/cm3 density). Both the upper and lower end-fitting regions 
are represented with material compositions that represent the homogenization of all of the 
components in the regions for the B&W 15x15 assembly design. The homogenization of the base 
components into single homogenized material compositions is performed using Equations 4 through 
6. Table 11 presents the component material volume fractions for the upper and lower end-fitting 
regions and Table 12 presents the base case upper and lower end-fitting homogenized material 
compositions. Table 13 presents the upper and lower fuel rod plenum material volume fractions and 
Table 14 presents the base case upper and lower fuel rod plenum homogenized material 
compositions. 

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M 
DensityMaterialdHomogenize =.[()( MaterialdHomogenizeinFractionVolume )]
. 
mm 
m 
(Eq. 4) 
where 
m = a single component material of the homogenized material 
M = the total number of component materials in the homogenized material 
. = the mass density of the component material. 
. ComponentofFractionMass . .(.) ( MaterialdHomogenizeinFractionVolume ). 
...
mm 
. MaterialdHomogenizeinMaterial .
.
= 
. DensityMaterialdHomogenize ..
(Eq. 5) 
. ofPercentWeight .
. .. ofFractionMass .. ComponentofPercentWeight . 
. MaterialComponent .... . 
.=. inMaterialComponent .. tConstituenMaterial . 
..
. .
. intConstituen 
.. 
. MaterialdHomogenize ..
.. 
MaterialComponentin .
. MaterialdHomogenize . 
(Eq. 6) 
Table 9. SS304 Material Composition 
Element Wt% Element Wt% 
Carbon 0.080 (0.030a) Chromium 19.000 
Nitrogen 0.100 Manganese 2.000 
Silicon 0.750 Iron Balance 68.745 (68.045a) 
Phosphorous 0.045 Nickel 9.250 (10.000a) 
Sulfur 0.030 Density = 7.94 g/cm3 
Source: ASME (2001, Section II, SA-240, Table 1); Density from ASTM (1999, G 1-90, p. 7, Table X1) 
a
NOTE: Carbon and Nickel composition corresponds to SS304L which yields a different iron balance 
Table 10. Inconel 718 Material Composition 
Element Wt% Element Wt% Element Wt% 
Nickel 52.500 Molybdenum 3.050 Silicon 0.180 
Chromium 19.000 Titanium 0.900 Carbon 0.040 
Iron 18.500 Aluminum 0.500 Sulfur 0.008 
Niobiuma 5.130 Manganese 0.180 Density = 8.19 g/cm3 
Source: Lynch 1989, p. 496 
a
NOTE: Reference identifies this material as “columbium,” which is actually the element niobium. 

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Table 11. End-Fitting Component Material Volume Fractions 
Assembly Design 
Stainless Steel 
Type 304 Inconel Zircaloy-4 Moderator 
Upper End-Fitting 0.2756 0.0441 0.0081 0.6722 
Lower End-Fitting 0.1656 0.0306 0.0125 0.7913 
Source: Punatar 2001, Table 2-3 
Table 12. End-Fitting Homogenized Material Compositions 
Element/ 
Isotope ZAIDa 
Upper End-Fitting 
Wt%b 
Lower End-Fitting 
Wt%b 
C-nat 6000.50c 0.0245 0.0203 
N-14 7014.50c 0.0668 0.0539 
Si-nat 14000.50c 0.5210 0.4229 
P-31 15031.50c 0.0301 0.0243 
S-32 16032.50c 0.0209 0.0170 
Cr-50 24050.60c 0.6181 0.5098 
Cr-52 24052.60c 12.3822 10.2114 
Cr-53 24053.60c 1.4309 1.1800 
Cr-54 24054.60c 0.3622 0.2987 
Mn-55 25055.50c 1.3563 1.0968 
Fe-54 26054.60c 2.6847 2.1808 
Fe-56 26056.60c 43.6633 35.4677 
Fe-57 26057.60c 1.0269 0.8342 
Fe-58 26058.60c 0.1380 0.1121 
Ni-58 28058.60c 8.3820 7.2490 
Ni-60 28060.60c 3.3394 2.8880 
Ni-61 28061.60c 0.1476 0.1277 
Ni-62 28062.60c 0.4777 0.4132 
Ni-64 28064.60c 0.1263 0.1093 
H-1 1001.50c 2.2972 3.6312 
O-16 8016.50c 18.2314 28.8196 
Al-27 13027.50c 0.0552 0.0514 
Ti-nat 22000.50c 0.0993 0.0925 
Nb-93 41093.50c 0.5659 0.5272 
Mo-nat 42000.50c 0.3364 0.3135 
Zr-nat 40000.60c 1.5920 3.2990 
Sn-nat 50000.35c 0.0235 0.0488 
Density (g/cm3) 3.2748 2.4388 
a
NOTES: ZAID = MCNP material identifier. 
b 
Homogenization used stainless steel 304L values for carbon, nickel, and iron 

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Table 13. Fuel Rod Plenum Material Volume Fractions 
Assembly Design Plenum Location 
Type 304 Stainless 
Steel 
Gas 
(modeled as void) Zircaloy-4 
Babcock & Wilcox Upper 0.0811 0.7793 0.1396 
15×15 Lower 0.1569 0.5973 0.2458 
Source: Punatar 2001, Table 2-9 and Figures 2-3 and 2-7 
NOTE: Volume fractions are renormalized to exclude the cladding, which is modeled explicitly in the input. 
Table 14. Fuel Rod Plenum Homogenized Material Compositions 
Element/Isotope ZAIDa 
Wt% of Element/Isotope in Material Composition 
Upper Fuel Rod Plenumb Lower Fuel Rod Plenumb 
C-nat 6000.50c 0.0124 0.0131 
N-14 7014.50c 0.0413 0.0436 
Si-nat 14000.50c 0.3096 0.3270 
P-31 15031.50c 0.0186 0.0196 
S-32 16032.50c 0.0124 0.0131 
Cr-50 24050.60c 0.3302 0.3485 
Cr-52 24052.60c 6.6148 6.9806 
Cr-53 24053.60c 0.7644 0.8067 
Cr-54 24054.60c 0.1935 0.2042 
Mn-55 25055.50c 0.8257 0.8720 
Fe-54 26054.60c 1.5943 1.6829 
Fe-56 26056.60c 25.9299 27.3712 
Fe-57 26057.60c 0.6099 0.6438 
Fe-58 26058.60c 0.0820 0.0865 
Ni-58 28058.60c 2.7744 2.9298 
Ni-60 28060.60c 1.1053 1.1672 
Ni-61 28061.60c 0.0489 0.0516 
Ni-62 28062.60c 0.1581 0.1670 
Ni-64 28064.60c 0.0418 0.0442 
O-16 8016.50c 0.0734 0.0705 
Zr-nat 40000.60c 57.6077 55.3392 
Sn-nat 50000.35c 0.8514 0.8178 
Density (g/cm3) 1.5597 2.8583 
a
NOTES: ZAID = MCNP material identifier. 
b 
Homogenization used stainless steel 304L values for carbon and nickel 
5.3.4 Fuel Material 
The following information provides the details needed to duplicate the input file specifications. The 
uranium dioxide fresh fuel compositions for each U-235 enrichment used in this evaluation are 
specified in Table 15, and were calculated using Equation 7 (Bowman et al. 1995, p. 20) for each 
isotope based on the U-235 wt%. 

Table 15. Fresh Fuel Compositions 
Enrichment 
(wt% U-235) Wt% U-234 Wt% U-235 Wt% U-236 Wt% U-238 Wt% Oxygen 
1.5 0.0106 1.3222 0.0061 86.8098 11.8513 
2.0 0.0144 1.7630 0.0081 86.3625 11.8519 
2.5 0.0184 2.2037 0.0101 85.9152 11.8526 
3.0 0.0224 2.6444 0.0122 85.4677 11.8533 
3.5 0.0265 3.0851 0.0142 85.0202 11.8540 
4.0 0.0306 3.5258 0.0162 84.5727 11.8547 
4.5 0.0348 3.9665 0.0182 84.1251 11.8553 
5.0 0.0390 4.4072 0.0203 83.6775 11.8560
U 
234 235 wt% )1.0837
wt%=*(0.007731) (U 
U 
236 235
wt%=(0.0046)*(U wt%) (Eq. 7) 
U
238 234 235
wt%=-100 U - wt% U - wt% U 236 wt% 
The initial oxygen mass is calculated using Equations 8 through 10. In Equations 8 and 9 the atomic 
mass values (M) come from Audi and Wapstra (1995). 
. wt%i .-1 
.
Mass U =..100 .. 
(Eq. 8)
UO mol 2 . i Mi .
..
.. 
where the weight percentages (wt%i) of the uranium isotopes (U234, U236, and U238) in uranium for 
a given initial enrichment were calculated using Equation 7.
Mass O =()(
oxygen for M 2 ) (Eq. 9)
UO mol 2 
.Mass O .
.
UO in Mass O 2 = 
..
Mass U
UO mol 2 
..
(UO in Mass U ) (Eq. 10)
2
. 
. UO mol 2 . 
where 
U Mass in UO2 is the fresh fuel uranium mass 
The spent fuel isotopes used in the MCNP cases correspond to those of the Principal Isotope Set 
(YMP 2003, Table 3-1). The irradiated fuel material compositions were taken from BSC (2003b, 
Section 6). Each nodal-depleted fuel composition is contained in the MCNP input files provided 

in Attachment IV. The nodal fuel isotopic compositions are listed in the input files in terms of 
ZAIDs and atoms/b-cm. The spreadsheet used for deriving interpolated values and the axial profile 
isotopics is contained in Attachment IV (spreadsheet IDBinputs.xls). 
5.4 MCNP GEOMETRIC DESCRIPTIONS 
The drawing for the 21-PWR waste package with absorber plates dimensions is contained 
electronically in Attachment IV. The MCNP representation of the 21-PWR waste package follows 
the same description as that shown in Attachment IV for the initial (at time of loading) 
configuration. 
When developing a loading curve, a configuration that results in the highest keff should be used in 
order to set an upper bounding limit that encompasses all other configurations. Therefore, the 
selection of a bounding configuration follows a linear progression based upon the results of other 
cases. Several potential configurations that could occur in the repository over a 10,000 year 
regulatory period were evaluated to determine which result in the highest keff values. The 
configurations are intended to investigate the effects on system reactivity as the waste package 
internal components degrade and the geometry changes. A series of configurations were evaluated 
with the descriptions and results provided in Attachment I. Based on the results, a combination of 
parameters was selected which will produce the most reactive representation for the generation of 
the loading curve. 
Based on the results of configurations presented in Attachment I, the preclosure bounding 
configuration is the as loaded representation with dry tuff surrounding the waste package. Dry tuff 
is used over saturated tuff due to the thermal load within the repository for the first 300 years 
precluding water from the surrounding drift area. The postclosure bounding configuration is 
represented by the intact fuel rods with saturated tuff surrounding the waste package and waste 
package internal hardware represented as in case3, from Attachment I, Section I.6. These 
configurations are the representative configurations for the loading curve evaluations. 
5.4.1 Fuel Assembly 
The physical dimensions for the fuel assembly design represented in the MCNP inputs were 
obtained from Punatar (2001), and are presented in Table 16, and illustrated in Figures 1 and 2. 
Table 16. B&W 15x15 Fuel Assembly Specifications 
Assembly Component Specification 
Fuel Pellet Outer Diameter 0.93980 cm 
Fuel Rod Cladding Inner Diameter 0.95758 cm 
Fuel Rod Cladding Outer Diameter 1.09220 cm 
Guide Tube Inner Diameter 1.26492 cm 
Guide Tube Outer Diameter 1.34620 cm 
Instrument Tube Inner Diameter 1.12014 cm 
Instrument Tube Outer Diameter 1.38193 cm 
Source: Punatar 2001, p. 2-5 

Pin Pitch = 1.44272 cm 
IT 
GT 
GT GT 
GT 
GT 
GT 
GT 
GT 
GT 
GT 
GT 
GT 
GT GT 
GT 
GT 
Guide Tube Instrument Tube Fuel Pin 
This sketch is not to scale. 
Source: (Punatar 2001, p. 2-3) 
GT 
IT 
Figure 1. Fuel Pin, Guide Tube, and Instrument Tube Locations in Fuel Assembly 

Licensing Calculation 
Region 1: 14.12 cm 
Region 2: 3.18 cm CRGTa Flange 
Region 3: 7.62 cm 
Region 4: 5.08 cm 
Region 5: 8.731 cm 
Region 6; 8.573 cm 
Region 7: 14.741 cm 
38.44 cm 
b 
53.579 cm 
b 
53.576 cm 
b53.579 cm 
360.172 cm b 
53.578 cm 
b 
53.658 cm 
b 
53.762 cm 
Region 9: 16.723 cm 
Region 10: 5.08 cm 
Region 11: 12.70 cm 
Region 12: 12.22 cm 
Upper Plenum 
Upper Core Grid 
Upper Pad 
Upper End fitting 
End Spacer Grid 
Upper Fuel Plenum 
Intermediate Spacer GridIntermediate Spacer GridIntermediate Spacer Grid 
Region 8: 
Active Fuel 
Intermediate Spacer GridIntermediate Spacer GridIntermediate Spacer GridLower Fuel Plenum 
Grid and End Fitting 
Lower Pad 
Lower Core Grid 
Region Between Grid and Plate 
Source: Punatar 2001, p. 2-11 
NOTES: a Control Rod Guide Tube (CRGT)
b Intermediate spacer grid height 3.81 cm 
Figure 2. Mark-B4 Fuel Assembly Axial Dimensions by Region 

5.5 INPUT PARAMETER SUMMARY 
Based on the sensitivity studies presented in Attachment I, the following parameters listed in Table 
17 were selected for use in the loading curve generation:
Table 17. Loading Curve Parameters 
Parameter Value Selection Basis Applicability 
Fuel 
Material 
UO2 Typical commercial PWR fuel Preclosure and 
Postclosure 
Absorber 1.5 wt% Gd at nominal 
thickness 
Based on typical value of 75% credit (NRC 2000, p. 8-4) 
allowed for fixed neutron absorbers and a nominal Gd 
loading of 2.0 wt% for Ni-Gd Alloy. Thickness based on 
results from Attachment I, Section I.6 
Preclosure and 
Postclosure 
Moderator Water at density of 1.0 
g/cm3 
Attachment I, Section I.3 Preclosure and 
Postclosure 
Fuel 
Density 
10.741 g/cm3 [a] Attachment I, Section I.1 Preclosure and 
Postclosure 
Reflector Dry Tuff Attachment I, Section I.5 Preclosure 
100% Saturated Tuff Attachment I, Section I.5 Postclosure 
Geometry Lattice array in 
Standard Vertical 
configuration (See 
Attachment I, Figure 36) 
Attachment I, Section I.3 Preclosure 
Lattice array in 
Standard Vertical 
configuration (See 
Attachment I, Figure 36) 
with aluminum and iron 
in plates oxidized 
Attachment I, Section I.6 (case3) and Section I.8 Postclosure 
a
NOTE: Calculated based on 98% theoretical density value of 10.96 g/cm3 for UO2 (Todreas and Kazimi 1990, 
p. 296)
6. RESULTS 
The loading curves for the 21 PWR waste package are presented in this section. The keff results 
represent the average combined collision, absorption, and track-length estimator from the MCNP 
calculations. The standard deviation (s) represents the standard deviation of keff about the average 
combined collision, absorption, and track-length estimate due to the Monte Carlo calculation 
statistics. It should be noted that in the following sections, any reference to enrichment refers to 
assembly average initial enrichment, and burnup refers to assembly average burnup. 
The corresponding MCNP input and output files for the cases used in this evaluation are provided 
electronically in Attachment IV. 

6.1 MAXIMUM FRESH FUEL ENRICHMENT 
This section presents the results of the maximum fresh fuel enrichments that can be loaded into the 
waste package with no burnup required. The determination of the maximum fresh fuel enrichment 
limit for the 21-PWR waste package with Ni-Gd Alloy absorber plates is determined by calculating 
keff for a range of initial enrichments and plotting them against the initial enrichments. The keff 
values plotted include a two-s allowance for computational uncertainty. The intersection of this 
curve and a line representing the critical limit (or USL) shows where the waste package has a 
potential for criticality. The results of the fresh fuel calculations are presented in Table 18 for the 
preclosure and postclosure bounding configurations, and are illustrated in Figure 3.
Table 18. Fresh Fuel keff Results 
Configuration Preclosure Configuration Postclosure Configuration 
Enrichment (Wt 
% U-235) keff s keff + 2s keff s keff + 2s 
1.5 0.85742 0.00049 0.85840 0.85504 0.00047 0.85598 
2 0.94279 0.00051 0.94381 0.93989 0.0005 0.94089 
2.5 1.00382 0.00048 1.00478 1.00146 0.00054 1.00254 
3 1.04892 0.00051 1.04994 1.04666 0.00055 1.04776 
3.5 1.08529 0.00053 1.08635 1.08269 0.00049 1.08367 
4 1.11396 0.00055 1.11506 1.11283 0.00052 1.11387 
4.5 1.13984 0.00046 1.14076 1.13819 0.00061 1.13941 
5 1.16027 0.00055 1.16137 1.15837 0.00059 1.15955 
Figure 3 shows that the maximum fresh fuel enrichment that would meet the loading curve criteria 
are 2.393 wt% U-235 for the postclosure bounding configuration, and 1.974 wt% U-235 for the 
preclosure bounding configuration. A CL = 0.9894 was taken from Framatome ANP (2003, Table 
11) based on laboratory critical experiments. This CL was chosen because the fresh fuel 
configuration range of parameters (i.e., materials, geometry, spectrum) is subsumed by the 
laboratory critical experiment parameters, and since no burned fuel is present in these 
configurations, the commercial reactor critical benchmarks are not directly applicable. Using this 
CL with the five percent margin as described in Section 5 yields a preclosure USL = 0.9894. 

keff 
1.2 
1.15 
1.1 
1.05 
1 
0.95 
0.9 
0.85 
0.8 
iisfilisfilllMax. enrchment that 
sates Precosure 
USL = 1.974 
Max. enrichment that 
sates Postcosure 
CL = 2.393 
Postcosure CL = 0.9894 
Precosure USL = 0.9394 
0123456 
Fuel Enrichment (Wt% U-235) 
lPreclosure Configuration Postcosure Configuration 
Figure 3. Fresh Fuel keff Results 
6.2 BURNED FUEL 
6.2.1 Preclosure 
The results for spent fuel with five-year decay time (the five-year decay time is based on the 
minimum cooling time required for the fuel assemblies to be classified as standard fuel [10 CFR 
961.11]) isotopic compositions are presented in Table 20. During the preclosure time period, each 
waste package is to remain at or below the USL. The minimum burnup required for each initial 
enrichment is determined by plotting the calculated keff versus the burnup. The burnup value of the 
intersection of the plotted curve with the USL is the required minimum burnup and are illustrated 
in Figures 4 through 17 and summarized in Table 21 and Figure 18. The keff values plotted include 
a two-s allowance for computational uncertainty. Any burnup value greater than this will result in 
a keff less than the USL, and is acceptable to be loaded into the waste package. In Figures 4 through17, the USL is presented as a function of average energy of a neutron causing fission (AENCF). The 
reported USL intercept values correspond to those for the USL as a function of the most 
conservative trend parameter as prescribed by YMP 2003 (Section 3.5.3.2.6). The USL equation 
is provided in Table 19.

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Table 19. Spent Nuclear Fuel Upper Subcritical Limit Function 
Trend Parameter 
USL Equation 
AENCF 
USL (AENCF) = -0.06262*AENCF + 0.9920 - 0.05 
NOTE: Framatome ANP (2003, Table 11) provides the CL function which is transformed into an USL function using 
.km = 0.05 (see Section 5 for details of USL transformation) 
Table 20. Preclosure Spent Nuclear Fuel keff Results 
Initial 7 Node Fuel Zone 1 Node Fuel Zone 
.(keff + 2s)a 
Enrichment 
(Wt% U-235) 
(GWd/ 
MTU) 
Burnup 
keff s 
AENCF 
(MeV) keff s 
AENCF 
(MeV) 
2.0 10 0.88508 0.0005 0.2096 0.8917 0.00049 0.2082 -0.0066 
15 0.85593 0.00053 0.2204 0.86227 0.00051 0.2208 -0.0063 
20 0.8308 0.00051 0.2304 0.83471 0.00049 0.2325 -0.0039 
25 0.80964 0.00047 0.2388 0.81034 0.00045 0.2435 -0.0007 
2.5 10 0.93053 0.0005 0.1967 0.93733 0.00053 0.1961 -0.0069 
15 0.90045 0.00055 0.2068 0.90731 0.00047 0.2071 -0.0067 
20 0.8717 0.00053 0.2170 0.87521 0.00047 0.2187 -0.0034 
25 0.84898 0.0005 0.2248 0.84775 0.00052 0.2289 0.0012 
30 0.82512 0.00046 0.2334 0.82298 0.0005 0.2390 0.0021 
3.0 10 0.97102 0.00054 0.1874 0.97732 0.00053 0.1867 -0.0063 
15 0.93966 0.00053 0.1969 0.94458 0.00053 0.1970 -0.0049 
20 0.91068 0.00054 0.2050 0.91454 0.00049 0.2074 -0.0038 
25 0.8859 0.00053 0.2131 0.88489 0.00057 0.2167 0.0009 
30 0.86123 0.0005 0.2212 0.85886 0.00045 0.2270 0.0025 
35 0.83547 0.00048 0.2300 0.83313 0.00043 0.2363 0.0024 
40 0.81789 0.00051 0.2362 0.81343 0.00045 0.2451 0.0046 
3.5 10 1.00571 0.00051 0.1802 1.01182 0.00055 0.1804 -0.0062 
15 0.9743 0.00053 0.1898 0.97925 0.00053 0.1888 -0.0049 
20 0.94354 0.0005 0.1973 0.94982 0.00051 0.1979 -0.0063 
25 0.91901 0.00056 0.2035 0.92021 0.00049 0.2075 -0.0011 
30 0.89495 0.00055 0.2097 0.89216 0.00054 0.2177 0.0028 
35 0.86804 0.0005 0.2206 0.86561 0.00049 0.2259 0.0025 
40 0.84834 0.00056 0.2253 0.84127 0.00051 0.2344 0.0072 
45 0.82412 0.00053 0.2353 0.81916 0.00048 0.2439 0.0051 
4.0 10 1.03478 0.00058 0.1758 1.0404 0.00055 0.1754 -0.0056 
15 1.00411 0.00056 0.1832 1.00895 0.00048 0.1832 -0.0047 
20 0.97336 0.00057 0.1904 0.97971 0.00051 0.1917 -0.0062 
25 0.95 0.00058 0.1966 0.95062 0.00053 0.1999 -0.0005 
30 0.92288 0.0006 0.2035 0.92289 0.0005 0.2073 0.0002 
35 0.8966 0.00057 0.2117 0.89614 0.0005 0.2170 0.0006 
40 0.87871 0.00053 0.2171 0.87118 0.00048 0.2247 0.0076 
45 0.85171 0.00058 0.2251 0.84819 0.00048 0.2332 0.0037 

Licensing Calculation 
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Table 20. Preclosure Spent Nuclear Fuel keff Results 
Initial 7 Node Fuel Zone 1 Node Fuel Zone 
.(keff + 2s)a 
Enrichment 
(Wt% U-235) 
(GWd/ 
MTU) 
Burnup 
keff s 
AENCF 
(MeV) keff s 
AENCF 
(MeV) 
4.5 10 1.05854 0.00055 0.1720 1.06668 0.00055 0.1713 -0.0081 
15 1.0289 0.0006 0.1786 1.03462 0.00059 0.1780 -0.0057 
20 1.00001 0.00056 0.1855 1.00663 0.0005 0.1855 -0.0065 
25 0.97449 0.00061 0.1918 0.97767 0.00053 0.1940 -0.0030 
30 0.94911 0.00061 0.1977 0.95165 0.00053 0.2004 -0.0024 
35 0.92334 0.00053 0.2044 0.92505 0.00058 0.2092 -0.0018 
40 0.90304 0.00054 0.2104 0.89899 0.00049 0.2167 0.0041 
45 0.87742 0.00065 0.2188 0.8739 0.00052 0.2249 0.0038 
5.0 10 1.08132 0.00055 0.1689 1.08738 0.00058 0.1681 -0.0061 
15 1.05225 0.00056 0.1741 1.05803 0.00052 0.1751 -0.0057 
20 1.02368 0.00055 0.1812 1.02929 0.00053 0.1810 -0.0056 
25 1.00005 0.00052 0.1864 1.00312 0.00054 0.1884 -0.0031 
30 0.9739 0.00056 0.1928 0.97666 0.00052 0.1954 -0.0027 
35 0.94695 0.00055 0.1994 0.94965 0.00052 0.2026 -0.0026 
40 0.9273 0.00046 0.2036 0.92413 0.00054 0.2105 0.0030 
45 0.90526 0.00057 0.2101 0.89889 0.00046 0.2179 0.0066 
a
NOTE: .(keff + 2s) = (keff + 2s)7 Node - (keff + 2s)1 Node 
keff+2s 
0.94 
0.92 
0.9 
0.88 
0.86 
0.84 
0.82 
0.8 
0 5 1015202530 
Burnup (GWd/MTU) 
itilit ()Upper Subcrca LimAENCF
Figure 4. Preclosure 7 Node Spent Nuclear Fuel keff Results for 2.0 Wt% U-235 Initial Enrichment 

keff+2s 
0.8 
0.82 
0.84 
0.86 
0.88 
0.9 
0.92 
0.94 
0 5 1015202530
Burnup (GWd/MTU) 
itilit ()Upper Subcrca LimAENCF
Figure 5. Preclosure 1 Node Spent Nuclear Fuel keff Results for 2.0 Wt% U-235 Initial Enrichment 
keff+2s 
Mi/) 
0.8 
0.82 
0.84 
0.86 
0.88 
0.9 
0.92 
0.94 
0.96 
0.98 
n. Burnup (GWdMTU7 Node = 10.315 
0 5 101520253035
Burnup (GWd/MTU) 
itiit ()Upper Subcrcal LimAENCF
Figure 6. Preclosure 7 Node Spent Nuclear Fuel keff Results for 2.5 Wt% U-235 Initial Enrichment 

keff+2s 
Mi/) 
0.8 
0.82 
0.84 
0.86 
0.88 
0.9 
0.92 
0.94 
0.96 
0.98 
n. Burnup (GWdMTU1 Node = 11.472 
0 5 101520253035 
Burnup (GWd/MTU) 
itiit ()Upper Subcrcal LimAENCF
Figure 7. Preclosure 1 Node Spent Nuclear Fuel keff Results for 2.5 Wt% U-235 Initial Enrichment 
1 
0.98 
Min. Burnup (GWd/MTU) 
0.96 7 Node = 16.941 
0.94 
0.92 
s 
0.9 
+ 20.88 
keff0.86 
0.84 
0.82 
0.8 
0 5 10 15 20 25 30 35 40 45 
Burnup (GWd/MTU) 
itiit ()Upper Subcrcal LimAENCF
Figure 8. Preclosure 7 Node Spent Nuclear Fuel keff Results for 3.0 Wt% U-235 Initial Enrichment 

Figure 9. Preclosure 1 Node Spent Nuclear Fuel keff Results for 3.0 Wt% U-235 Initial Enrichment 
1 
0.98 Min. Burnup (GWd/MTU) 
1 Node = 17.710 
0.96 
0.94 
0.92 
s 
0.9 
eff+20.88 
k0.86 
0.84 
0.82 
0.8 
0 5 10 15 20 25 30 35 40 45 
Burnup (GWd/MTU) 
itiit ()Upper Subcrcal LimAENCF
1 
keff + 2s 
Mi/) 
0.8 
0.85 
0.9 
0.95 
1.05 
n. Burnup (GWdMTU7 Node = 23.100 
0 5 101520253035404550 
Burnup (GWd/MTU) 
itiit ()Upper Subcrcal LimAENCF
Figure 10. Preclosure 7 Node Spent Nuclear Fuel keff Results for 3.5 Wt% U-235 Initial Enrichment 

1 
keff+2s 
Mi/) 
0.8 
0.85 
0.9 
0.95 
1.05 
n. Burnup (GWdMTU1 Node = 23.655 
0 5 101520253035404550
Burnup (GWd/MTU) 
itiit ()Upper Subcrcal LimAENCF
Figure 11. Preclosure 1 Node Spent Nuclear Fuel keff Results for 3.5 Wt% U-235 Initial Enrichment 
1 
keff + 2s 
Mi/) 
0.8 
0.85 
0.9 
0.95 
1.05 
n. Burnup (GWdMTU7 Node = 29.028 
0 5 101520253035404550
Burnup (GWd/MTU) 
itil Limit ()Upper SubcrcaAENCF
Figure 12. Preclosure 7 Node Spent Nuclear Fuel keff Results for 4.0 Wt% U-235 Initial Enrichment 

1 
keff+2s 
Mi (/) 
0.8 
0.85 
0.9 
0.95 
1.05 
1.1 
n. BurnupGWdMTU1 Node = 29.062 
0 5 101520253035404550
Burnup (GWd/MTU) 
itil Limit ()Upper SubcrcaAENCF
Figure 13. Preclosure 1 Node Spent Nuclear Fuel keff Results for 4.0 Wt% U-235 Initial Enrichment 
1 
keff+2s 
Mi(/) 
0.8 
0.85 
0.9 
0.95 
1.05 
1.1 
n. Burnup GWdMTU7 Node = 34.059 
0 5 101520253035404550
Burnup (GWd/MTU) 
itil Limit ()Upper SubcrcaAENCF
Figure 14. Preclosure 7 Node Spent Nuclear Fuel keff Results for 4.5 Wt% U-235 Initial Enrichment 

1 
keff+2s 
Mi(/) 
0.8 
0.85 
0.9 
0.95 
1.05 
1.1 
n. Burnup GWdMTU1 Node = 34.482 
0 5 101520253035404550
Burnup (GWd/MTU) 
itil Limit ()Upper SubcrcaAENCF
Figure 15. Preclosure 1 Node Spent Nuclear Fuel keff Results for 4.5 Wt% U-235 Initial Enrichment 
1 
keff+2s 
Mi() 
0.8 
0.85 
0.9 
0.95 
1.05 
1.1 
n. Burnup GWd/MTU7 Node = 39.737 
0 5 101520253035404550
Burnup (GWd/MTU) 
itil Limit ()Upper SubcrcaAENCF
Figure 16. Preclosure 7 Node Spent Nuclear Fuel keff Results for 5.0 Wt% U-235 Initial Enrichment 

1 
keff+2s 
Mi/) 
0.8 
0.85 
0.9 
0.95 
1.05 
1.1 
1.15 
n. Burnup (GWdMTU1 Node = 39.278 
0 5 101520253035404550 
Burnup (GWd/MTU) 
itil Limit ()Upper SubcrcaAENCF
Figure 17. Preclosure 1 Node Spent Nuclear Fuel keff Results for 5.0 Wt% U-235 Initial Enrichment 
Table 21. Minimum Required Burnups for Intercept of Upper Subcritical Limit 
Initial Enrichment 7 Node Fuel Zone 1 Node Fuel Zone 
(Wt% U-235) Burnup (GWd/MTU) 5% BU Unc.a Burnup (GWd/MTU) 5% BU Unc.a 
1.974 0 0 0 0 
2.5 10.315 10.831 11.472 12.046 
3 16.941 17.788 17.710 18.596 
3.5 23.100 24.255 23.655 24.838 
4 29.028 30.479 29.062 30.515 
4.5 34.059 35.762 34.482 36.206 
5 39.737 41.724 39.278 41.242 
a
NOTE: Required minimum burnup including 5% uncertainty associated with assembly burnup records 
BSC (2003b, Section 6.2.2) states that the SNF isotopics (BSC 2003b, Section 6) in a single axial 
zone representation bound commercial PWR fuel assemblies at the same initial enrichment and 
burnup. Therefore, using a limiting axial profile in conjunction with the bounding isotopics would 
be adding additional conservatism that has already been bounded and is unnecessary. The seven 
axial zone representation results were provided for comparison purposes only, and to illustrate that 
for the selected bounding configurations, axial profile effects are negligible. 

) 
0 
5 
10 
15 
20 
25 
30 
35 
40 
45 
(Required Minimum BurnupGWd/MTU
0123456 
Initial Enrichment (Wt% U-235) 
i1 Node Loading Curve 7 Node Loadng Curve 
Figure 18. Required Minimum Burnups for Intercept of Upper Subcritical Limit 
6.2.2 Postclosure 
The results for spent fuel with five-year decay time (the five-year decay time is based on the 
minimum cooling time required for the fuel assemblies to be classified as standard fuel [10 CFR 
961.11]) isotopic compositions are presented in Table 23. The five-year decay time isotopics are 
used because they bound fuel that is cooled longer with respect to reactivity. The minimum burnup 
required for each initial enrichment is determined by plotting the calculated keff versus the burnup. 
The burnup value of the intersection of the plotted curve with the CL is the required minimum 
burnup and are illustrated in Figures 19 through 32 and summarized in Table 24 and Figure 33. The 
keff values plotted include a two-s allowance for computational uncertainty. Any burnup value 
greater than this will result in a keff less than the CL, and is acceptable to be loaded into the waste 
package. In Figures 19 through 32, the CL is presented as a function of AENCF. The reported CL 
intercept values correspond to those for the CL as a function of the most conservative trend 
parameter as prescribed by YMP 2003 (Section 3.5.3.2.6). The CL equation is provided in Table 22. 
Table 22. Spent Nuclear Fuel Critical Limit Function 
Trend Parameter 
CL Equation 
AENCF 
CL (AENCF) = -0.06262*AENCF + 0.9920 
Source: Framatome ANP (2003, Table 11) 

Licensing Calculation 
Table 23. Postclosure Spent Nuclear Fuel keff Results 
Initial 
Enrichment( 
Wt% U-235) 
Burnup 
(GWd/ 
MTU) 
7 Node Fuel Zone 1 Node Fuel Zone 
.keff + 2sakeff s 
AENCF 
(MeV) keff s 
AENCF 
(MeV) 
2.0 10 0.88261 0.0005 0.2106 0.88941 0.00044 0.2091 -0.0067 
15 0.85428 0.0005 0.2215 0.85981 0.00051 0.2224 -0.0055 
20 0.82789 0.00044 0.2326 0.83066 0.00052 0.2340 -0.0029 
25 0.80912 0.00054 0.2395 0.80823 0.00048 0.2446 0.0010 
2.5 10 0.92833 0.00051 0.1974 0.93462 0.00049 0.1968 -0.0062 
15 0.89825 0.00057 0.2091 0.90268 0.00049 0.2095 -0.0043 
20 0.87083 0.00057 0.2173 0.87386 0.00054 0.2198 -0.0030 
25 0.84687 0.00049 0.2261 0.84572 0.00044 0.2300 0.0012 
30 0.82326 0.00053 0.2331 0.82157 0.00049 0.2405 0.0018 
3.0 10 0.96898 0.00053 0.1885 0.97382 0.00053 0.1880 -0.0048 
15 0.93858 0.00058 0.1983 0.9431 0.00052 0.1981 -0.0044 
20 0.90847 0.00051 0.2067 0.91238 0.00054 0.2080 -0.0040 
25 0.88502 0.00053 0.2136 0.88339 0.00048 0.2182 0.0017 
30 0.86038 0.00054 0.2215 0.85625 0.00051 0.2288 0.0042 
35 0.83389 0.00052 0.2310 0.83187 0.00048 0.2384 0.0021 
40 0.81595 0.00048 0.2379 0.80985 0.00047 0.2469 0.0061 
3.5 10 1.00293 0.00053 0.1811 1.00762 0.00051 0.1809 -0.0046 
15 0.97181 0.00055 0.1898 0.97687 0.00049 0.1902 -0.0049 
20 0.94191 0.00055 0.1978 0.94679 0.00054 0.2003 -0.0049 
25 0.9176 0.00058 0.2056 0.91784 0.00056 0.2086 -0.0002 
30 0.89178 0.00051 0.2126 0.8896 0.00053 0.2178 0.0021 
35 0.86684 0.00051 0.2205 0.86361 0.00048 0.2271 0.0033 
40 0.84817 0.00054 0.2259 0.84012 0.00051 0.2360 0.0081 
45 0.82228 0.00053 0.2366 0.81709 0.00049 0.2447 0.0053 
4.0 10 1.03059 0.00054 0.1763 1.03738 0.00052 0.1760 -0.0068 
15 1.00061 0.00056 0.1841 1.00734 0.00054 0.1843 -0.0067 
20 0.97152 0.00058 0.1915 0.97721 0.00054 0.1922 -0.0056 
25 0.94649 0.00051 0.1979 0.94867 0.00053 0.2009 -0.0022 
30 0.92166 0.00055 0.2044 0.92043 0.00045 0.2091 0.0014 
35 0.89509 0.00052 0.2126 0.89356 0.00053 0.2180 0.0015 
40 0.87697 0.00053 0.2180 0.86873 0.0005 0.2262 0.0083 
45 0.85105 0.00055 0.2257 0.8449 0.00048 0.2337 0.0063 
4.5 10 1.05776 0.00051 0.1727 1.06256 0.00061 0.1722 -0.0050 
15 1.02722 0.00055 0.1793 1.03281 0.00053 0.1798 -0.0055 
20 0.99887 0.00052 0.1863 1.00513 0.0006 0.1867 -0.0064 
25 0.97387 0.00054 0.1918 0.97668 0.00052 0.1945 -0.0028 
30 0.94795 0.00054 0.1982 0.94905 0.00055 0.2018 -0.0011 
35 0.92133 0.00055 0.2067 0.92157 0.00047 0.2093 -0.0001 
40 0.9022 0.00054 0.2107 0.89641 0.00055 0.2177 0.0058 

Licensing Calculation 
Title: 21-PWR Waste Package with Absorber Plates Loading Curve Evaluation 
Document Identifier: CAL-DSU-NU-000006 REV 00B Page 43 of 60 
Table 23. Postclosure Spent Nuclear Fuel keff Results 
Initial 
Enrichment( 
Wt% U-235) 
Burnup 
(GWd/ 
MTU) 
7 Node Fuel Zone 1 Node Fuel Zone 
.keff + 2sakeff s 
AENCF 
(MeV) keff s 
AENCF 
(MeV) 
45 0.87778 0.00059 0.2187 0.87247 0.00051 0.2259 0.0055 
5.0 10 1.07938 0.00058 0.1689 1.08386 0.00055 0.1689 -0.0044 
15 1.05009 0.00058 0.1755 1.05647 0.00053 0.1755 -0.0063 
20 1.01987 0.00054 0.1824 1.02724 0.00053 0.1829 -0.0073 
25 0.99812 0.00058 0.1874 1.00018 0.00054 0.1890 -0.0020 
30 0.97272 0.00063 0.1936 0.97296 0.00055 0.1964 -0.0001 
35 0.94438 0.00056 0.2004 0.94592 0.00049 0.2041 -0.0014 
40 0.9287 0.0006 0.2047 0.92162 0.00051 0.2106 0.0073 
45 0.90086 0.00053 0.2125 0.89623 0.00046 0.2178 0.0048 
a
NOTE: .keff + 2s = (keff + 2s)7 Node - (keff + 2s)1 Node 
keff+2s 
1 
0.98 
0.96 
0.94 
0.92 
0.9 
0.88 
0.86 
0.84 
0.82 
0.8 
0 5 1015202530 
Burnup (GWd/MTU) 
itiit ()Upper Subcrcal LimAENCF
Figure 19. Postclosure 7 Node Spent Nuclear Fuel keff Results for 2.0 Wt% U-235 Initial Enrichment 

1 
keff+2s 
0.8 
0.82 
0.84 
0.86 
0.88 
0.9 
0.92 
0.94 
0.96 
0.98 
0 5 1015202530
Burnup (GWd/MTU) 
itiit ()Upper Subcrcal LimAENCF
Figure 20. Postclosure 1 Node Spent Nuclear Fuel keff Results for 2.0 Wt% U-235 Initial Enrichment 
1 
keff+2s 
0.8 
0.82 
0.84 
0.86 
0.88 
0.9 
0.92 
0.94 
0.96 
0.98 
0 5 101520253035
Burnup (GWd/MTU) 
itiit ()Upper Subcrcal LimAENCF
Figure 21. Postclosure 7 Node Spent Nuclear Fuel keff Results for 2.5 Wt% U-235 Initial Enrichment 

1 
keff+2s 
0.8 
0.82 
0.84 
0.86 
0.88 
0.9 
0.92 
0.94 
0.96 
0.98 
0 5 101520253035
Burnup (GWd/MTU) 
itiit ()Upper Subcrcal LimAENCF
Figure 22. Postclosure 1 Node Spent Nuclear Fuel keff Results for 2.5 Wt% U-235 Initial Enrichment 
1 
keff + 2s 
0.8 
0.82 
0.84 
0.86 
0.88 
0.9 
0.92 
0.94 
0.96 
0.98 
0 5 1015202530354045
Burnup (GWd/MTU) 
itiit ()Upper Subcrcal LimAENCF
Figure 23. Postclosure 7 Node Spent Nuclear Fuel keff Results for 3.0 Wt% U-235 Initial Enrichment 

1 
keff+2s 
0.8 
0.82 
0.84 
0.86 
0.88 
0.9 
0.92 
0.94 
0.96 
0.98 
0 5 1015202530354045
Burnup (GWd/MTU) 
itiit ()Upper Subcrcal LimAENCF
Figure 24. Postclosure 1 Node Spent Nuclear Fuel keff Results for 3.0 Wt% U-235 Initial Enrichment 
1 
keff + 2s 
Mi(/) 
0.8 
0.85 
0.9 
0.95 
1.05 
n. Burnup GWdMTU7 Node = 13.820 
0 5 101520253035404550
Burnup (GWd/MTU) 
itiit ()Upper Subcrcal LimAENCF
Figure 25. Postclosure 7 Node Spent Nuclear Fuel keff Results for 3.5 Wt% U-235 Initial Enrichment 

1 
0.95 
0.9 
0.85 
0.8 
keff+2s 
Mi/) 
1.05 
n. Burnup (GWdMTU1 Node = 14.629 
0 5 101520253035404550 
Burnup (GWd/MTU) 
itiit ()Upper Subcrcal LimAENCF
Figure 26. Postclosure 1 Node Spent Nuclear Fuel keff Results for 3.5 Wt% U-235 Initial Enrichment 
1 
keff + 2s 
Mi/) 
0.8 
0.85 
0.9 
0.95 
1.05 
n. Burnup (GWdMTU7 Node = 18.718 
0 5 101520253035404550 
Burnup (GWd/MTU) 
itil Limit ()Upper SubcrcaAENCF
Figure 27. Postclosure 7 Node Spent Nuclear Fuel keff Results for 4.0 Wt% U-235 Initial Enrichment 

1 
keff+2s 
Mi(/) 
0.8 
0.85 
0.9 
0.95 
1.05 
1.1 
n. Burnup GWdMTU1 Node = 19.717 
0 5 101520253035404550
Burnup (GWd/MTU) 
itil Limit ()Upper SubcrcaAENCF
Figure 28. Postclosure 1 Node Spent Nuclear Fuel keff Results for 4.0 Wt% U-235 Initial Enrichment 
1 
keff+2s 
Mi(/) 
0.8 
0.85 
0.9 
0.95 
1.05 
1.1 
n. Burnup GWdMTU7 Node = 23.976 
0 5 101520253035404550
Burnup (GWd/MTU) 
itil Limit ()Upper SubcrcaAENCF
Figure 29. Postclosure 7 Node Spent Nuclear Fuel keff Results for 4.5 Wt% U-235 Initial Enrichment 

1 
keff+2s 
Mi(/) 
0.8 
0.85 
0.9 
0.95 
1.05 
1.1 
n. Burnup GWdMTU1 Node = 24.627 
0 5 101520253035404550
Burnup (GWd/MTU) 
itil Limit ()Upper SubcrcaAENCF
Figure 30. Postclosure 1 Node Spent Nuclear Fuel keff Results for 4.5 Wt% U-235 Initial Enrichment 
1 
keff+2s 
Mi(/) 
0.8 
0.85 
0.9 
0.95 
1.05 
1.1 
n. Burnup GWdMTU7 Node = 28.817 
0 5 101520253035404550
Burnup (GWd/MTU) 
itil Limit ()Upper SubcrcaAENCF
Figure 31. Postclosure 7 Node Spent Nuclear Fuel keff Results for 5.0 Wt% U-235 Initial Enrichment 

1 
keff+2s 
Mi/) 
0.8 
0.85 
0.9 
0.95 
1.05 
1.1 
n. Burnup (GWdMTU1 Node = 28.945 
0 5 101520253035404550 
Burnup (GWd/MTU) 
itil Limit ()Upper SubcrcaAENCF
Figure 32. Postclosure 1 Node Spent Nuclear Fuel keff Results for 5.0 Wt% U-235 Initial Enrichment 
Table 24. Minimum Required Burnups for Intercept of Critical Limit 
Initial Enrichment 
(Wt% U-235) 
7 Node Fuel Zone 1 Node Fuel Zone 
Burnup (GWd/MTU) 5% BU Unc.a Burnup (GWd/MTU) 5% BU Unc.a 
2.393 0.000 0.000 0.000 0.000 
2.5b 1.398 1.468 2.924 3.070 
3.0b 8.290 8.705 9.112 9.568 
3.5 13.820 14.511 14.629 15.360 
4.0 18.718 19.654 19.717 20.703 
4.5 23.976 25.175 24.627 25.858 
5.0 28.817 30.258 28.945 30.392 
a
NOTES: Required minimum burnup including 5% uncertainty associated with assembly burnup records 
b 
Where the intercept was below 10 GW/MTU the values were extrapolated 
BSC (2003b) states that the SNF isotopics (BSC 2003b, Section 6) in a single axial zone 
representation bound commercial PWR fuel assemblies at the same initial enrichment and burnup. 
Therefore, using a limiting axial profile in conjunction with the bounding isotopics would be adding 
additional conservatism that has already been bounded and is unnecessary. The seven axial zone 
representation results were provided for comparison purposes only, and to illustrate that for the 
selected bounding configurations, axial profile effects are negligible. 

0 
5 
10 
15 
20 
25 
30 
35 
0 1 2 3 4 5 6 
(/)Required Minimum BurnupGWdMTU
Initial Enrichment (Wt% U-235) 
1 Node Loading Curve 7 Node Loading Curve 
Figure 33. Required Minimum Burnups for Intercept of Critical Limit 
6.2.3 Waste Stream Comparison 
The waste stream inventory in terms of number of fuel assemblies at given burnups and enrichments 
was taken from Waste Packages and Source Terms for the Commercial 1999 Design Basis Waste 
Streams (CRWMS M&O 2000, Attachment III) using the "Case A" arrival forecast. "Case A" refers 
to 10-year-old youngest fuel first for 63,000 MTU. This arrival forecast was selected based on 
Licensing Position LP-009, Waste Stream Parameters (Williams 2003, Section 3.4). The results of 
the loading curve compared against the waste stream inventory are presented in Figure 34. This 
curve is based on the selection of the minimum required burnups using a single axial zone 
representation including a 5% burnup uncertainty associated with assembly burnup records. The 
squares in the legend indicate number groupings of assemblies at a particular burnup and enrichment 
(e.g., 100-199 indicates that there are 100 to 199 assemblies at a listed burnup and enrichment); Base 
LC is the loading curve based on the nominal required minimum burnup; and Fitted Polynomial is 
the quintic polynomial fitted to the loading curve adjusted for a five percent uncertainty associated 
with the reported assembly burnups. The waste stream information that was extracted and sorted 
is provided in Attachment IV as the workbook wstreamplot.xls. 

Burnup (GWd/MTU)
80 
70 
60 
50 
40 
30 
20 
10 
0 
0123 456 
y5432R2 = 1 
Filial 
A 
= 0.451x - 8.7061x + 66.281x - 249.22x + 475.35x - 358.38 
1-99 
100-199 
200-299 
300+ 
Base LC 
tted Poynomcceptable 
Unacceptable 
Initial Enrichment (Wt% U-235) 
NOTE: The fitted polynomial represents the loading curve based on a five percent burnup uncertainty of the 
bounding curve (Base LC) 
Figure 34. Loading Curve and Projected Waste Stream 
Waste stream population comparisons are provided in Table 25 for the 21-PWR waste package with 
absorber plates. Assemblies that are considered "unacceptable" do not meet the requirements for 
loading into the 21-PWR waste package with absorber plates, and will need to be disposed of in a 
different waste package design. 
Table 25. Loading Curve Waste Stream Acceptability Comparison 
Loading Curve # Assemblies Below 
Configuration Minimum Required Burnup Acceptable (%) Unacceptable (%) 
Base LC 1194 98.73 1.27 
Fitted Polynomial 1524 98.37 1.63 
6.2.4 Misloaded Assembly 
Since there is a possibility of inserting a single SNF assembly into a waste package during loading 
that does not meet the requirements of the design basis loading curve (Commercial Spent Nuclear 
Fuel Waste Package Misload Analysis [BSC 2003c, p. 26]), the impact of a misloaded assembly was 
evaluated. A comparison for the burnup/enrichment pairs that make up the loading curve was made 
against configurations involving a misloaded assembly. For the comparisons, the misloaded 
assembly used was an assembly that was 10 GWd/MTU and 20 GWd/MTU below the design basis 
required burnup for a given enrichment. The isotopic material compositions were interpolated from 
BSC (2003b, Section 6). For conservative purposes with respect to criticality, the misloaded 
assembly was placed in the most reactive position within the waste package (center position). The 

Licensing Calculation 
Title: 21-PWR Waste Package with Absorber Plates Loading Curve Evaluation 
Document Identifier: CAL-DSU-NU-000006 REV 00B Page 53 of 60 
results of the comparisons are provided in Tables 26 and 27 and illustrated in Figure 35, with the 
nominal cases using the design basis required burnup based on the equation presented in Figure 34. 
Table 26. Misloaded Assembly Results for 10 GWd/MTU Underburned Assembly 
Enrichment Misloaded Waste Package Nominally Loaded Waste Package 
(Wt% U-235)/ 
Design Basis 
Required Burnup AENCF AENCF 
(GWd/MTU) keff s (MeV) keff s (MeV) .keff + 2s 
2.5 / 11.472 0.93871 0.00050 0.1951 0.92799 0.00054 0.2000 0.01064 
3.0 / 17.710 0.93884 0.00049 0.1999 0.92807 0.00046 0.2033 0.01083 
3.5 / 23.655 0.93769 0.00049 0.2021 0.92808 0.00052 0.2047 0.00955 
4.0 / 29.062 0.93760 0.00052 0.2041 0.92839 0.00050 0.2062 0.00925 
4.5 / 34.482 0.93517 0.00051 0.2061 0.92663 0.00055 0.2079 0.00846 
5.0 / 39.278 0.93673 0.00058 0.2065 0.92796 0.00054 0.2087 0.00885 
Table 27. Misloaded Assembly Results for 20 GWd/MTU Underburned Assembly 
Enrichment 
(Wt% U-235)/ 
Design Basis 
Required Burnup 
(GWd/MTU) 
Misloaded Waste Package Nominally Loaded Waste Package 
keff s 
AENCF 
(MeV) keff s 
AENCF 
(MeV) .keff + 2s 
3.5 / 23.953 0.95487 0.00053 0.1983 0.92808 0.00052 0.2047 0.02681 
4.0 / 29.511 0.95327 0.00050 0.2005 0.92839 0.00050 0.2062 0.02488 
4.5 / 34.645 0.94999 0.00053 0.2022 0.92663 0.00055 0.2079 0.02332 
5.0 / 39.567 0.94812 0.00057 0.2037 0.92796 0.00054 0.2087 0.02022 
0 .k + 2s 
.km0.01 
0.02 
0.03 
0.04 
0.05 
0.06 
0.07 
0.08 
0.09 
0.1 
eff 
= 0.05 
0123456 
Initial Enrichment (Wt% U-235) 
//10 GWdMTU Below 20 GWdMTU Below 
Figure 35. Misloaded Assembly Results 

Figure 35 shows that the preclosure reactivity effect of misloading underburned commercial SNF 
into an intact, flooded 21-PWR waste package, with absorber plates, is about one percent for fuel 
underburned by 10 GWd/MTU, and two to three percent for SNF underburned by 20 GWd/MTU. 
Since the arbitrary margin imposed to ensure subcriticality (.km) is five percent, the misloading of 
even grossly underburned SNF will not actually result in a critical configuration. Note that no credit 
is being taken for the fact that not all SNF will be at the design basis required minimum burnup 
value - most SNF will actually be at higher burnups that will reduce the reactivity of the waste 
package and compensate for the positive reactivity caused by a misloaded, underburned SNF 
assembly. 
6.3 SUMMARY OF RESULTS 
Results presented in Attachment I (Table 32) illustrate that the 21-PWR waste package with 
absorber plates will not go critical with commercial SNF if there is no moderator present within the 
waste package (i.e., dry criticality is not possible). A misloaded, underburned assembly is 
considered probable (BSC 2003c, p. 26), but since the preclosure administrative margin is large 
enough to subsume the increase in reactivity associated with a substantially underburned assembly, 
the waste package will not go critical during the preclosure time period. Note that these calculations 
are conservatively based on a fully flooded waste package, although the waste packages are expected 
to exclude water moderation. It is expected that loaded waste packages will have enough assemblies 
sufficiently above the design basis minimum required burnup to compensate for a misloaded 
assembly during the preclosure or postclosure time period. 
This report recommends that the 21-PWR waste package with Ni-Gd Alloy absorber plates be 
loaded using the loading curve with the five percent uncertainty assessed to the burnup values. The 
loading curve is described by the quintic polynomial presented in Equation 11.
y(x) = 0.451x5 - 8.7061x4 + 66.281x3 - 249.22x2 + 475.35x - 358.38 (Eq. 11) 
where 
x = the initial enrichment of the fuel assembly in wt% U-235 
y(x) = the required minimum burnup (GWd/MTU) for an assembly with enrichment x 
This recommendation is made based on the expectation that the probability of selecting enough 
assemblies with sufficient burnups above the design basis value to compensate for a single 
misloaded assembly is adequate for keeping the waste package criticality probability below the 
threshold criterion. Using this loading curve allows 98.28 percent of the current PWR projected 
waste stream to be disposed of in the 21-PWR waste package with Ni-Gd Alloy absorber plates 
waste package. A different waste package design could be used for disposing of the remaining 1.72 
percent of the projected PWR waste stream or predetermined loading patterns could be used to blend 
assemblies that are below the design basis minimum burnup with assemblies that have high enough 
burnups to compensate. 
All outputs are reasonable compared to the inputs and the results of this calculation are suitable for 
their intended use. 

7. REFERENCES 
7.1 DOCUMENTS CITED 
ASM International. 1987. Corrosion. Volume 13 of Metals Handbook. 9th Edition. Metals Park, 
Ohio: ASM International. TIC: 209807. 
ASM International 1990a. Properties and Selection: Nonferrous Alloys and Special-Purpose 
Materials. Volume 2 of ASM Handbook. Formerly 10th Edition, Metals Handbook. 5th Printing 
1998. [Materials Park, Ohio]: ASM International. TIC: 241059. 
ASM International 1990b. Properties and Selection: Irons, Steels, and High-Performance Alloys. 
Volume 1 of Metals Handbook. 10th Edition. Materials Park, Ohio: ASM International. TIC: 
245666. 
Audi, G. and Wapstra, A.H. 1995. Atomic Mass Adjustment, Mass List for Analysis. [Upton, New 
York: Brookhaven National Laboratory, National Nuclear Data Center]. TIC: 242718. 
B&W Fuel Company 1991. Final Design Package Babcock & Wilcox BR-100 100 Ton Rail/Barge 
Spent Fuel Shipping Cask. Volume 2. 51-1203400-01. DBABE0000-00272-1000-00014 REV 00. 
Lynchburg, Virginia: B&W Fuel Company. ACC: MOV.19960802.0083. 
Bowman, S.M.; Hermann, O.W.; and Brady, M.C. 1995. Sequoyah Unit 2 Cycle 3. Volume 2 of 
Scale-4 Analysis of Pressurized Water Reactor Critical Configurations. ORNL/TM-12294/V2. Oak 
Ridge, Tennessee: Oak Ridge National Laboratory. TIC: 244397. 
Briesmeister, J.F., ed. 1997. MCNP-A General Monte Carlo N-Particle Transport Code. LA-12625-
M, Version 4B. Los Alamos, New Mexico: Los Alamos National Laboratory. ACC: 
MOL.19980624.0328. 
BSC (Bechtel SAIC Company) 2003a. PWR Axial Burnup Profile Analysis. CAL-DSU-NU-000012 
REV 00A. Las Vegas, Nevada: Bechtel SAIC Company. ACC: DOC.20031002.0002. 
BSC (Bechtel SAIC Company) 2003b. Isotopic Generation and Confirmation of the PWR 
Application Model. CAL-DSU-NU-000004 REV 00A. Las Vegas, Nevada: Bechtel SAIC Company. 
ACC: DOC.20031110.0003. 
BSC (Bechtel SAIC Company) 2003c. Commercial Spent Nuclear Fuel Waste Package Misload 
Analysis. CAL-WHS-MD-000003 REV 00A. Las Vegas, Nevada: Bechtel SAIC Company. ACC: 
DOC.20031002.0005. 
BSC (Bechtel SAIC Company) 2004a. 21 PWR A, B, D & E Fuel Plates. 000-MW0-DSU0-01201-
000-00B. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20040803.0004. 
BSC (Bechtel SAIC Company) 2004b. 21 PWR C Fuel Plate. 000-MW0-DSU0-01301-000-00B. 
Las Vegas, Nevada: M&O. ACC: ENG.20040804.0001. 

BSC (Bechtel SAIC Company) 2004c. Design and Engineering, 21 PWR Corner Guide. 000-MW0-
DSU0-01001-000-00A. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20040305.0005. 
BSC (Bechtel SAIC Company) 2004d. Design and Engineering, 21 PWR End Side Guide. 000-
MW0-DSU0-00901-000-00A. Las Vegas, Nevada: Bechtel SAIC Company. ACC: 
ENG.20040305.0010. 
BSC (Bechtel SAIC Company) 2004e. Design and Engineering, 21 PWR Side Guide. 000-MW0-
DSU0-00801-000-00A. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20040305.0002. 
BSC (Bechtel SAIC Company) 2004f. Design and Engineering, 21-PWR Waste Package 
Configuration. 000-MW0-DSU0-00402-000-00B. Las Vegas, Nevada: Bechtel SAIC Company. 
ACC: ENG.20040119.0004. 
BSC (Bechtel SAIC Company) 2004g. 21-PWR Waste Package Configuration. 000-MW0-DSU0-
00403-000-00D. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20040708.0005. 
BSC (Bechtel SAIC Company) 2004h. Design and Engineering, Fuel Tube. 000-MW0-DSU0-
02001-000-00A. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20040305.0012. 
BSC (Bechtel SAIC Company) 2004i. Geochemistry Model Abstraction and Sensitivity Studies for 
the 21 PWR CSNF Waste Packages. MDL-DSU-MD-000001 REV 00 [Errata 001]. Las Vegas, 
Nevada: Bechtel SAIC Company. ACC: MOL.20021107.0154; DOC.20040225.0005. 
BSC (Bechtel SAIC Company) 2004j. Seismic Consequence Abstraction. MDL-WIS-PA-000003 
REV 0 Errata 1. Las Vegas, Nevada: Bechtel SAIC Company. ACC: DOC.20030818.0006; 
DOC.20040218.0002. 
BSC (Bechtel SAIC Company) 2004k. Q-List. 000-30R-MGR0-00500-000-000 REV 00. Las Vegas, 
Nevada: Bechtel SAIC Company. ACC: ENG.20040721.0007. 
BSC (Bechtel SAIC Company) 2004l. Total Dust Settling on Naval Long Waste Packages in 100 
Years. 800-M0C-VU00-00900-000-00A. Las Vegas, Nevada: Bechtel SAIC Company. ACC: 
ENG.20040913.0001. 
BSC (Bechtel SAIC Company) 2004m. Waste Package Masses. 000-00C-MGR0-01100-000-00B. 
Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20040603.0008. 
Cacciapouti, R.J. and Van Volkinburg, S. 1997. Axial Burnup Profile Database for Pressurized 
Water Reactors. YAEC-1937. Bolton, Massachusetts: Yankee Atomic Electric Company. ACC: 
MOL.19980209.0184. 
Canori, G.F. and Leitner, M.M. 2003. Project Requirements Document. TER-MGR-MD-000001 
REV 02. Las Vegas, Nevada: Bechtel SAIC Company. ACC: DOC.20031222.0006. 

CRWMS M&O 1998a. Software Qualification Report for MCNP Version 4B2, A General Monte 
Carlo N-Particle Transport Code. CSCI: 30033 V4B2LV. DI: 30033-2003, Rev. 01. Las Vegas, 
Nevada: CRWMS M&O. ACC: MOL.19980622.0637. 
CRWMS M&O 1998b. Selection of MCNP Cross Section Libraries. B00000000-01717-5705-00099 
REV 00. Las Vegas, Nevada: CRWMS M&O. ACC: MOL.19980722.0042. 
CRWMS M&O 1998c. Summary Report of Commercial Reactor Criticality Data for Sequoyah Unit 
2. B00000000-01717-5705-00064 REV 01. Las Vegas, Nevada: CRWMS M&O. ACC: 
MOL.19980716.0015. 
CRWMS M&O 1998d. Summary Report of Commercial Reactor Criticality Data for McGuire Unit 
1. B00000000-01717-5705-00063 REV 01. Las Vegas, Nevada: CRWMS M&O. ACC: 
MOL.19980622.0079. 
CRWMS M&O 2000. Waste Packages and Source Terms for the Commercial 1999 Design Basis 
Waste Streams. CAL-MGR-MD-000001 REV 00. Las Vegas, Nevada: CRWMS M&O. ACC: 
MOL.20000214.0479. 
DOE (U.S. Department of Energy) 1987. Appendix 2A. Physical Descriptions of LWR Fuel 
Assemblies. Volume 3 of Characteristics of Spent Fuel, High-Level Waste, and Other Radioactive 
Wastes Which May Require Long-Term Isolation. DOE/RW-0184. Washington, D.C.: U.S. 
Department of Energy, Office of Civilian Radioactive Waste Management. ACC: 
HQX.19880405.0024. 
DOE (U.S. Department of Energy) 2004. Quality Assurance Requirements and Description. 
DOE/RW-0333P, Rev. 16. Washington, D.C.: U.S. Department of Energy, Office of Civilian 
Radioactive Waste Management. ACC: DOC.20040907.0002. 
Doraswamy, N. 2004. Project Design Criteria Document. 000-3DR-MGR0-00100-000-002. Las 
Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20040721.0003. 
Einziger, R.E. 1991. "Effects of an Oxidizing Atmosphere in a Spent Fuel Packaging Facility." 
Proceedings of the Topical Meeting on Nuclear Waste Packaging, FOCUS '91, September 29– 
October 2, 1991, Las Vegas, Nevada. Pages 88-99. La Grange Park, Illinois: American Nuclear 
Society. TIC: 231173. 
Framatome ANP 2003. Critical Limit Development for 21 PWR Waste Package. 32-5029773-00. 
[Lynchburg, Virginia]: Framatome ANP. ACC: DOC.20031212.0004. 
Gelest, Inc. 2004. Gelest Silicone Fluids: Stable, Inert Media. Morrisville, Pennsylvania: Gelest, Inc. 
TIC: 256122. 
Inco Alloys International. 1998. INCONEL Alloy 625. Huntington, West Virginia: Inco Alloys 
International. TIC: 240361. 

Lide, D.R., ed. 2002. CRC Handbook of Chemistry and Physics. 83rd Edition. Boca Raton, Florida: 
CRC Press. TIC: 253582 
Lynch, C.T., ed. 1989. Practical Handbook of Materials Science. Boca Raton, Florida: CRC Press. 
TIC: 240577. 
Parrington, J.R.; Knox, H.D.; Breneman, S.L.; Baum, E.M.; and Feiner, F. 1996. Nuclides and 
Isotopes, Chart of the Nuclides. 15th Edition. San Jose, California: General Electric Company and 
KAPL, Inc. TIC: 233705. 
Punatar, M.K. 2001. Summary Report of Commercial Reactor Criticality Data for Crystal River Unit 
3. TDR-UDC-NU-000001 REV 02. Las Vegas, Nevada: Bechtel SAIC Company. ACC: 
MOL.20010702.0087. 
Roberts, W.L.; Campbell, T.J.; and Rapp, G.R., Jr. 1990. Encyclopedia of Minerals. 2nd Edition. 
New York, New York: Van Nostrand Reinhold. TIC: 242976. 
Stout, R.B. and Leider, H.R., eds. 1997. Waste Form Characteristics Report Revision 1. UCRL-ID-
108314. Version 1.2. Livermore, California: Lawrence Livermore National Laboratory. ACC: 
MOL.19980512.0133. 
Todreas, N.E. and Kazimi, M.S. 1990. Nuclear Systems I, Thermal Hydraulic Fundamentals. New 
York, New York: Hemisphere Publishing. TIC: 226511. 
Williams, N.H. 2003. "Contract No. DE-AC28-01RW12101 - Licensing Position-009, Waste Stream 
Parameters." Letter from N.H. Williams (BSC) to J.D. Ziegler (DOE/ORD), November 13, 2003, 
1105039412, with enclosure. ACC: MOL.20031215.0076. 
Williams, N.H. 2004. Decision Proposal, Technical Decision, Statement for Consideration: Change 
the Current Neutron Absorber Material in the CSNF Waste Packages from Borated Stainless Steel 
to a Nickel-Gadolinium Alloy. Tracking No. TMRB-2004-009. [Las Vegas, Nevada: Bechtel SAIC 
Company]. ACC: MOL.20040622.0307. 
YMP (Yucca Mountain Site Characterization Project) 2003. Disposal Criticality Analysis 
Methodology Topical Report. YMP/TR-004Q, Rev. 02. Las Vegas, Nevada: Yucca Mountain Site 
Characterization Office. ACC: DOC.20031110.0005. 
7.2 CODES, REGULATIONS, STANDARDS , AND PROCEDURES 
10 CFR (Code of Federal Regulations) 63. Energy: Disposal of High-Level Radioactive Wastes in 
a Geologic Repository at Yucca Mountain, Nevada. Readily available. 
10 CFR (Code of Federal Regulations) 961. Energy: Standard Contract for Disposal of Spent 
Nuclear Fuel and/or High-Level Radioactive Waste. Readily Available. 

ASME (American Society of Mechanical Engineers) 2001. 2001 ASME Boiler and Pressure Vessel 
Code (includes 2002 addenda). New York, New York: American Society of Mechanical Engineers. 
TIC: 251425. 
ASTM A 516/A 516M-90. 1991. Standard Specification for Pressure Vessel Plates, Carbon Steel, 
for Moderate-and Lower-Temperature Service. West Conshohocken, Pennsylvania: American 
Society for Testing and Materials. TIC: 240032. 
ASTM B 811-97. 2000. Standard Specification for Wrought Zirconium Alloy Seamless Tubes for 
Nuclear Reactor Fuel Cladding. West Conshohocken, Pennsylvania: American Society for Testing 
and Materials. TIC: 247245. 
ASTM B 932-04. 2004. Standard Specification for Low-Carbon Nickel-Chromium-Molybdenum-
Gadolinium Alloy Plate, Sheet, and Strip. West Conshohocken, Pennsylvania: American Society for 
Testing and Materials. TIC: 255846. 
ASTM G 1-90 (Reapproved 1999). 1999. Standard Practice for Preparing, Cleaning, and 
Evaluating Corrosion Test Specimens. West Conshohocken, Pennsylvania: American Society for 
Testing and Materials. TIC: 238771. 
LP-SI.11Q-BSC, Rev. 0, ICN 0. Software Management. Washington, D.C.: U.S. Department of 
Energy, Office of Civilian Radioactive Waste Management. ACC: DOC.20040225.0007. 
NRC (U.S. Nuclear Regulatory Commission) 2000. Standard Review Plan for Spent Fuel Dry 
Storage Facilities. NUREG-1567. Washington, D.C.: U.S. Nuclear Regulatory Commission. TIC: 
247929. 
7.3 SOURCE DATA LISTED BY DATA TRACKING NUMBER 
GS000308313211.001. Geochemistry of Repository Block. Submittal date: 03/27/2000. 
MO0003RIB00071.000. Physical and Chemical Characteristics of Alloy 22. Submittal date: 
03/13/2000. 
MO0109HYMXPROP.001. Matrix Hydrologic Properties Data. Submittal date: 09/17/2001. 
LB990501233129.001. Fracture Properties for the UZ Model Grids and Uncalibrated Fracture and 
Matrix Properties for the UZ Model Layers for AMR U0090, "Analysis of Hydrologic Properties 
Data". Submittal date: 08/25/1999. 
7.4 SOFTWARE CODES 
CRWMS M&O 1998e. Software Code: MCNP. V4B2LV. HP, HPUX 9.07 and 10.20; PC, Windows 
95; Sun, Solaris 2.6. 30033 V4B2LV. 

8. ATTACHMENTS 
Table 28 presents the attachment specifications for this calculation file. 
Table 28. Attachment Listing 
Attachment # # of Pages Date Created Description 
I 12 N/A Sensitivity Studies 
II 6 N/A PWR Assembly Lattice Design Sensitivity 
III 2 N/A Listing of contents on Attachment IV 
IV N/A 09/20/2004 Compact Disc attachment containing information 
listed in Attachment III 

Attachment I: Sensitivity Studies 
I. DESCRIPTION AND RESULTS 
Sensitivity studies were performed to observe the waste package as it behaves over time in the 
repository and to determine which material characteristics result in the highest keff values. A brief 
description of the sensitivity studies performed and their results is provided in the following 
sections. 
In each of the sensitivity cases, the waste package dimensions correspond to those provided in 
Attachment IV (21PWRWP.zip) with any irradiated fuel compositions taken from BSC (2003b, 
Section 6). Each configuration is represented with dry tuff surrounding the waste package. The dry 
tuff composition for each of the sensitivity cases is represented as listed in Table 29.
Table 29. Tuff Composition for Sensitivity Cases 
Compound Wt% 
SiO2 76.29 
Al2O3 12.55 
FeO 0.14 
Fe2O3 0.97 
MgO 0.13 
CaO 0.5 
Na2O 3.52 
K2O 4.83 
TiO2 0.11 
P2O5 0.05 
MnO 0.07 
Source: DTN:GS000308313211.001, mean values from file zz_sep_249138.txt. 
NOTE: Derived elemental/isotopic number densities for MCNP inputs are provided in Attachment IV, workbook Tuff 
composition.xls, sheet Latest_Tuff 
I.1 FUEL DENSITY EFFECTS 
Variations in fuel density were evaluated. This set of cases was performed in order to assess the fuel 
density that would result in the highest keff values given the fixed lattice dimensions of the B&W 
15x15 fuel assembly. Fuel density values were varied from 9.8 g/cm3 through 10.8 g/cm3 for a 
representative assembly with 3.0 wt% U-235 fresh fuel and 3.0 wt% U-235 initial enriched fuel with 
a burnup of 30 GWd/MTU. The results of this set of cases are presented in Table 30.

Table 30. Spent Nuclear Fuel Density keff Resultsa 
Fuel Density (g/cm3) 
Fresh Burned 
keff s Filenameb keff s Filenameb 
9.8 1.03949 0.0005 case1 0.852 0.00053 case1 
10 1.0411 0.00053 case2 0.85357 0.00057 case2 
10.2 1.04304 0.00059 case3 0.85398 0.00051 case3 
10.4 1.04491 0.00051 case4 0.8568 0.00048 case4 
10.6 1.04783 0.00053 case5 0.85785 0.00049 case5 
10.8 1.04955 0.00052 case6 0.85906 0.00048 case6 
a
NOTES: Homogenized end-fitting regions used a water density of 0.1 g/cm3 in the homogenization. Since these 
cases only evaluate the trend of keff with different fuel densities, and the end-fitting regions are outside the 
active fuel region and constant in all the cases, there is no impact on the trend. 
b 
Filenames are the same but are contained in a unique directory structure in Attachment IV as explained in 
Attachment III. 
The results show that an increase in fuel density, which effectively increases the fissile mass in a 
fixed geometry, causes keff to increase. 
I.2 WASTE PACKAGE FUEL ASSEMBLY GEOMETRY 
Variations in fuel assembly geometries were evaluated. This set of cases investigated the effects of 
different positioning of the fuel assemblies within the waste package. Three configurations where 
evaluated. One where the assembly was centered within the waste package basket cell as could 
occur when the waste package is in a vertical position during loading operations, and two where the 
fuel assemblies are resting against the basket plates which occurs when the waste package is in a 
horizontal position. The three different geometric representations are provided in Figures 36through 38 and the results presented in Table 31. Base case compositions correspond to a fuel 
assembly with 3.0 wt% U-235 fresh fuel and density of 10.741 g/cm3. 
Figure 36. Standard Vertical Position Waste Package Geometry 

Figure 37. Standard Horizontal Position Waste Package Geometry 
Figure 38. Rotated Horizontal Position Waste Package Geometry 

Table 31. Fuel Assembly Geometry keff Results 
Configuration keff s Base Filename 
Standard Vertical 1.04892 0.00051 vert 
Standard Horizontal 1.04100 0.00053 horiz1 
Rotated Horizontal 1.03346 0.00049 horiz2 
The results show that a standard vertical geometry representation produces the highest keff. 
I.3 OPTIMUM MODERATOR DENSITY 
A search for optimum moderator density was performed. This set of cases was used to show that 
the fuel assemblies placed into a waste package configuration is an under-moderated system. 
Moderator density values were varied from 0.0 g/cm3 through 1.0 g/cm3. Base case values 
correspond to a fresh fuel assembly with 3.0 wt% U-235 initial enrichment. The results of this set 
of cases are presented in Table 32 and are illustrated in Figure 39. 
Table 32. Moderator Density Sensitivity Results 
Moderator Density 
(g/cm3) keff s Filename 
0.0a 0.45169 0.00021 Case0a 
0.0 0.34835 0.00020 Case0 
0.1 0.56110 0.00039 Case1 
0.2 0.68843 0.00045 Case2 
0.3 0.77881 0.00051 Case3 
0.4 0.85103 0.00054 Case4 
0.5 0.90441 0.00053 Case5 
0.6 0.94797 0.00054 Case6 
0.7 0.98167 0.00050 Case7 
0.8 1.00983 0.00053 Case8 
0.9 1.03076 0.00061 Case9 
1.0 1.04892 0.00051 vert 
NOTE: a Case used 5.0 wt% U-235 enriched fuel to evaluate maximum dry fuel reactivity 

keff 
1.2 
1.0 
0.8 
0.6 
0.4 
0.2 
0.0 
0.0 0.2 0.4 0.6 0.8 1.0 1.2 
Water Density (g/cm3) 
Figure 39. Moderator Density Sensitivity Results 
Another moderator material instead of water was also evaluated. Hydraulic fluid/oil that may leak 
from a handling crane. Estimates indicate that a 100 ton or 200 ton crane, which could be utilized 
by the handling facility, would contain approximately 100 to 135 gallons of fluid/oil. The 
representative hydraulic fluid follows the description provided in Gelest, inc. (2004) and the material 
safety data sheet contained in Attachment IV (0130223.pdf) and has a chemical form as follows: 
CH3CH3 CH3 
SiO SiCH3
CH3 SiO 
CH3 CH3 
CH3
4 
Source: Gelest Inc. 2004, p. 11
For the event of this fluid/oil getting into the waste package several cases were evaluated to observe 
the effects on system reactivity. The cases were as follows:
Case1 - the entire waste package is filled with fluid/oil at a density of 0.9 g/cm3 (quantity indicates
this is non-mechanistic, but will bound all other configurations (keff = 0.97892, s = 0.00053)
In order to assess whether the hydraulic fluid results in an over- or under-moderation of the system,
the density of the material was lowered in these cases to observe the effects on keff:
Case2 - used a density of 0.85 g/cm3 (keff = 0.96779, s = 0.00051)
Case3 - used a density of 0.8 g/cm3 (keff = 0.95510, s = 0.00054) 

Since the resulting keff values decreased with decreased density, the system is under-moderated and 
no further evaluations are warranted. Water is a better moderator than the representative hydraulic 
fluid. 
I.4 SIMPLIFIED GEOMETRY 
Several geometric simplifications were evaluated in order to determine the impact on system 
reactivity. An external trunion region and basket stiffeners are present on the drawings presented 
in Attachment IV (21PWRWP.zip). A comparison was made between cases with and without the 
trunion region with or without the stiffeners being represented, as well as with the trunions, 
stiffeners, and representative hardware material on the axial ends (accounts for various for barrier 
lids). The configurations were evaluated with 3.0 wt% U-235 fresh fuel. The results of this set of 
cases are presented in Table 33.
Table 33. Simplified Geometry Results 
Case Description keff s Filename 
Base case with no trunion region and basket 
stiffeners present 1.04892 0.00051 vert 
Base case with trunion region, basket stiffeners 
present, and axial hardware outside inner 
region represented 
1.04892 0.00051 wtsends 
Base case with no trunion region and no 
stiffeners present 1.04921 0.00055 wostif 
Base case with trunion region and stiffeners 
present 1.04892 0.00051 wtands 
These results indicate that the presence of trunions, stiffeners, or both in the representation have an 
insignificant impact on system reactivity. 
I.5 TUFF EVALUATIONS 
Variations for tuff present in and around the waste package were evaluated. This set of cases was 
performed in order to assess the impact tuff could have on system reactivity. Various levels of 
saturation were evaluated, as well as geometric arrangements. Configurations were evaluated using 
the 15x15 assembly design with 3.0 wt% U-235 fresh fuel. The results for the external reflector tuff 
cases are presented in Table 34. Cases where the tuff material was uniformly dispersed within the 
waste package are presented in Table 35. This set of cases is very unlikely since the basket plates 
would serve as a barrier to prevent tuff from getting into the internal regions of the basket geometry. 
Cases where the tuff would most likely be able to enter the waste package and be mobile would be 
in solution, and were evaluated with the results presented in Table 36. Two solution sets were 
evaluated, one where the solution is in all void regions, and one where it accumulates external to the 
fuel basket tubes over the active fuel region. Derivations of the tuff material compositions were 
performed in Attachment IV (spreadsheet Tuff composition.xls). 

Table 34. External Waste Package Reflector Evaluation Results 
Case Description keff s Filename 
Dry tuff outside waste package 1.04892 0.00051 vert 
100% saturated tuff outside 
waste package 1.04895 0.00055 tuffsat 
Water outside waste package 1.04962 0.00055 water 
Void outside waste package 1.04882 0.00055 void 
Table 35. Waste Package Tuff Internal Configuration Solid Results 
Case Description keff s Filename 
Dry tuff in all void regions 0.39419 0.00021 t0all 
100% saturated tuff in all void 
regions 0.59427 0.00039 t100all 
Table 36. Waste Package Tuff Internal Configuration Solution Results 
Case Description Solution in All Void Solution in Void Outside FBTa Filenameb 
keff s keff s 
10 volume percent of available 
saturated tuff in solution 1.04873 0.00053 1.04914 0.00052 t10 
20 volume percent of available 
saturated tuff in solution 1.04896 0.00058 1.04956 0.00047 t20 
30 volume percent of available 
saturated tuff in solution 1.04863 0.00056 1.04808 0.00048 t30 
40 volume percent of available 
saturated tuff in solution 1.04897 0.00051 1.04902 0.00053 t40 
50 volume percent of available 
saturated tuff in solution 1.05047 0.00052 1.04972 0.00057 t50 
100 volume percent of available 
saturated tuff in solution 1.04880 0.00053 1.04882 0.00048 t100 
a
NOTES: FBT = fuel basket tube 
b 
Filenames are the same but are contained in a unique directory structure in Attachment IV as explained in 
Attachment III. 
These results indicate that the 50 volume percent case distributing the tuff mixture over all void 
regions produces the highest keff value. This is based on having 50% of the original assumed 20 kg 
of tuff (see Assumption 3.6) that could get into the waste package entering and being in solution. 
Since this is an assumed amount of tuff that can get inside the waste package, and the results (varied 
from 10% though 100%) are statistically equivalent keff values at the 95 percent confidence level as 
using pure water for moderator and reflector, having 20 kg of tuff or less in the waste package is 
considered to produce equivalent reactivity levels as pure water. 
I.6 INTERNAL COMPONENT DEGRADATION 
A set of sensitivity cases was performed to evaluate effects of changing conditions as waste package 
internal components degrade. These cases are based on the waste package internal structures 
degrading before the waste form and is representative of configuration class IP-3 from YMP (2003, 

Figure 3-2). The cases used a 3.0 wt% U-235 fresh fuel composition for each of the runs. 
Variations were made to the amount of corrosion product retained within the basket cells to observe 
the sensitivity of neutron spectrum. The corrosion product mixture composition was derived in 
Attachment IV (spreadsheet misc.xls, sheet degraded_sens) based on the amounts of iron and 
aluminum contained within the basket cells. Iron was assumed to form the mineral Hematite (Fe2O3) 
or Goethite (FeOOH), and aluminum was assumed to form into the mineral Gibbsite (Al[OH]3). The 
results and a brief description of the cases are provided in Table 37.
Table 37. keff Results for Degraded Internal Component Cases 
Filename Case Description keff s 
case1 Nominal with carbon steel as Hematite and Al as Gibbsite 
in original locations, fully flooded 
1.03613 0.00046 
case2 Nominal with carbon steel as Goethite and Al as Gibbsite in 
original locations, fully flooded 
1.02445 0.00053 
case3 Same as case1 but oxidized plates in expanded volume 1.04614 0.00055 
case4 Same as case2 but oxidized plates in expanded volume 1.02547 0.00051 
case5 Al and steel removed from system, fully flooded 1.03121 0.00056 
case6g All corrosion product (CP) distributed uniformly throughout 
package (all Fe represented as goethite), fully flooded 
0.9156 0.0005 
case6h All CP distributed uniformly throughout package (all Fe 
represented as hematite), fully flooded 
0.90662 0.00053 
case7g Same as case6g but at 33% CP retention level 0.99090 0.00051 
case7h Same as case6h but at 33% CP retention level 0.98819 0.00048 
case8g Same as case6g but at 66% CP retention level 0.95247 0.00051 
case8h Same as case6h but at 66% CP retention level 0.94640 0.00053 
case9 Same as case3 but absorber plates at 5mm thickness 1.04351 0.00053 
The results indicate that a fully flooded system with the oxidized aluminum and iron remaining in 
place (case3) produces the highest system reactivity. The iron mineral hematite produces higher keff 
values for this configuration. Corrosion products mixed in solution displace moderator, which 
reduces neutron thermalization within the assemblies and thus results in lower keff values. 

I.7 ABSORBER PLATE DEGRADATION 
A series of cases were evaluated in order to determine the effects of the absorber plate as it degrades 
over time. Variations were made to the amount of corrosion product retained within the basket cells 
to observe the sensitivity of neutron spectrum. The corrosion product mixture composition was 
derived in Attachment IV (spreadsheet misc.xls, sheet degraded_sens) based on the amounts of iron 
and aluminum contained within the basket cells. Iron was assumed to form the mineral Hematite 
(Fe2O3), and aluminum was assumed to form into the mineral Gibbsite (Al[OH]3). 
Current degradation rate information (Williams 2004) for the Ni-Gd alloy indicate that the maximum 
amount of degradation will be less than 1 mm per side, resulting in a minimum of 5 mm of Ni-Gd 
absorber. Since geochemistry calculations for this material and waste package configuration are not 
available, varied amounts of estimated corrosion product composition were evaluated. The amounts 
varied from 0% to 100%. The results for this set of cases are presented in Table 38.
Table 38. keff Results for 5mm Thick Absorber Plate Cases 
Corrosion Product Retained keff s Filename 
0% 1.03421 0.00057 5m0cp 
33% 0.99101 0.00053 5m33cp 
66% 0.95058 0.00046 5m66cp 
100% 0.91264 0.00053 5m100cp 
These results indicate that the configuration is more reactive without corrosion product composition 
represented in the basket cells, as was also indicated in Table 37.
Table 39 illustrates results as a function of absorber plate (Ni-Gd Alloy) thickness, with 0% and 
100% corrosion product retention. 
Table 39. keff Results as a function of Absorber Plate Thickness 
% Absorber Plate 
Removed 
Remaining Plate 
Thickness (mm) 
100% Corrosion 
Product Retention 
0% Corrosion 
Product Retention 
Filenameakeff s keff s 
0 7.25 0.90687 0.0005 1.03130 0.00061 0d 
10 6.525 0.90669 0.00052 1.03043 0.00057 10d 
20 5.8 0.90855 0.00051 1.03248 0.00057 20d 
30 5.075 0.9117 0.00055 1.03382 0.00051 30d 
40 4.35 0.91411 0.00053 1.03451 0.00055 40d 
a
NOTE: Filenames are the same but are contained in a unique directory structure in Attachment IV as explained in 
Attachment III 
These results indicate that the 0% corrosion product retention cases produce higher keff values than 
the 100% corrosion product retention cases, and the system keff increases for a fully flooded system 
as the absorber plates become thinner. 

I.8 COMPROMISED FUEL RODS 
This configuration class is based on the waste form degrading before or at the same rate as the waste 
package internal structures and is representative of configuration classes IP-1 and IP-2 from YMP 
(2003, Figure 3-2). Seismic studies have determined that a peak ground velocity of 1.067 m/s or 
greater results in fuel cladding failure (BSC 2004j, Table 30) which results in exposure of the spent 
nuclear fuel to an oxidizing atmosphere. Once the fuel cladding is breached, oxidation of the fuel 
material can occur and cause clad breach propagation (unzipping). Therefore, a set of cases was 
evaluated that involved oxidized fuel. The thermodynamically stable state for oxidized uranium is 
UO3 (Einziger 1991, p. 88). If moisture is present in the atmosphere hydration may also occur 
(Einziger 1991, p. 88) and form the compound UO3(H2O)2 (Einziger 1991, Figure 1) otherwise 
known as the mineral schoepite (BSC 2004i, Attachment I, file data0 files.zip, file data0.ymf). 
A set of sensitivity studies was performed in order to evaluate various configurations. The cases 
used a 3.0 wt% U-235 fresh fuel schoepite composition for each of the runs except case0 and 
case0nw which are provided for fresh fuel base case k8 values to compare against. The 
compositions were derived in Attachment IV (spreadsheet misc.xls, sheet schoepite) for the fuel 
material. The results and a brief description of the cases are provided in Table 40. It should be 
noted that the sensitivity runs were for an infinite two-dimensional lattice configuration in a 
representative waste package basket geometry, which is the reason for the high eigenvalues. 
Table 40. Compromised Fuel Assembly Sensitivity Cases 
Filename Case Description k8 s 
case0 Fresh fuel base case for comparison, fully flooded in 
nominal geometry. 1.08436 0.00093 
case0nw Same as case0 but dry conditions. 0.46317 0.00037 
case1 Schoepite expanded around clad; nominal standard vertical 
geometry; dry conditions; Schoepite block at theoretical 
density. 
1.02388 0.00095 
case2 Schoepite expanded around clad; aluminum shunts and 
fuel basket tube have corroded into gibbsite and goethite, 
respectively, occupying original volumes; nominal standard 
vertical geometry, dry conditions, Schoepite block at 
theoretical density. 
0.95172 0.00111 
case3 Same as Case2 but the gibbsite and goethite volume is 
expanded and applied to the thickness along the active fuel 
length. 
0.92806 0.00104 
case4 Same as Case1 but the system is compressed to the point 
where the basket plates are in contact with the fuel basket 
tubes. 
1.02551 0.00099 
case5 Same configuration as Case4 but the basket plate 
materials and fuel basket tube have oxidized into gibbsite 
and goethite. 
0.95743 0.00098 
case6 Same as Case5 but the waste package compression is in 
the vertical direction with expansion in the x direction. 0.95389 0.00105 
case7 Like Case2 but fuel basket tube is represented as hematite 
instead of goethite. 0.96238 0.00099 
case8 Same as Case5 but the fuel basket tube is hematite 
instead of goethite. 0.96785 0.00107 

Engineered Systems Calculation 
Title: 21-PWR Waste Package with Absorber Plates Loading Curve Evaluation 
Document Identifier: CAL-DSU-NU-000006 REV 00B Page I-11 of 12 
Table 40. Compromised Fuel Assembly Sensitivity Cases 
Filename Case Description k8 s 
case9 Same conditions as Case1 but system is fully collapsed so 
there is no spacing between fuel basket tube, schoepite, 
and basket. Nominal pin pitch used. 
1.02997 0.00096 
case9a Same conditions as Case1 but system is fully collapsed so 
there is no spacing between fuel basket tube and 
schoepite. Pin pitch adjusted so schoepite has no spaces 
around it in unit cell. Also, surrounds guide tubes and 
instrument tubes. 
1.03559 0.00094 
case9b Same as case9a with schoepite at theoretical density with 
all clad, guide tubes, and instrument tubes removed from 
system. Basket materials collapsed around schoepite 
block. 
1.0776 0.00092 
case10 Same as Case1 but water fills all void space. 1.02817 0.00097 
case11 Same as Case3 but the fuel basket tube corrosion product 0.94719 0.00102 
g and h composition has settled within the basket cell and 
assembly is in standard horizontal position geometry (See 
Figure 37); g designates goethite as the primary iron 
mineral and h denotes hematite as the primary iron 
mineral, CP in solution. 
0.95350 0.00099 
case12 Same as Case11 but CP oxide is uniformly distributed 0.94269 0.00097 
g and h throughout basket cell; g designates goethite as the 
primary iron mineral and h denotes hematite as the primary 
iron mineral. 
0.94577 0.00103 
case13 Same as Case12 but only water in basket cell. 0.98394 0.001 
case14 Same as Case10 but the water and schoepite are 
homogenized throughout FBT region. 1.08035 0.00097 
case15 
g and h 
Like case12 but CP and schoepite are uniformly mixed as 
dry CP occupying space throughout basket cell; g 
designates goethite as the primary iron mineral and h 
denotes hematite as the primary iron mineral, CP in 
solution. 
1.00210 0.00080 
0.98491 0.00086 
The results indicate that a nominal configuration with schoepite uniformly dispersed in solution in 
a fully flooded system that can result from a seismic event produces the highest reactivity (case14) 
for ruptured fuel rods. The case0 configuration resulted in a higher reactivity, and is representative 
of the preclosure intact fuel rod nominal configuration. 
I.9 WASTE PACKAGE INTERACTION 
A set of cases was evaluated to assess the impact of package-to-package interaction during 
preclosure operations. Variations were made in the spacing and interstitial material between 
packages to observe the sensitivity to such parameters. The interstitial material was represented as 
void and water, and the spacing was varied from 0 to 1 cm. A brief description of each case and the 
results are presented in Table 41.

Engineered Systems Calculation 
Table 41. Waste Package Interaction Results 
Case Description keff s Filename 
Single WP with no others, filled with water and 
void outside 
1.04882 0.00055 c0 
Infinite array of WP touching water inside and 
void outside 
1.04947 0.00052 c1 
Infinite array of WP touching void inside and 
water outside 
0.33384 0.00022 c2 
Infinite array of WP touching water inside and 
water outside 
1.04808 0.00051 c3 
Infinite array of WP 1 cm spacing water inside 
and void outside 
1.04930 0.00054 c4 
Infinite array of WP 1 cm spacing void inside 
and water outside 
0.33111 0.00022 c5 
Infinite array of WP 1 cm spacing water inside 
and water outside 
1.04779 0.00051 c6 
These results indicate that waste packages have a negligible neutronic influence on other waste 
packages. Without water, the waste packages have no criticality concern, and with water 
represented in different areas, the eigenvalues are within two sigma of the single waste package. 

Licensing Calculation 
Title: 21-PWR Waste Package with Absorber Plates Loading Curve Evaluation 
Document Identifier: CAL-DSU-NU-000006 REV 00B Page II-1 of 6 
Attachment II: PWR Assembly Lattice Design Sensitivity 
Variations in fuel assembly lattice design were evaluated. This set of cases was performed in order 
to assess which fuel assembly lattice design would result in the highest keff values when loaded in 
a waste package configuration and confirm Assumption 3.1. Fuel assembly lattices were varied 
using Babcock & Wilcox (B&W), Westinghouse (W), and Combustion Engineering (CE) geometric 
arrangements in pure water. These cases were evaluated using a fresh fuel enrichment of 5.0 wt% 
U-235 in a nominal waste package configuration and 5.0 wt% U-235 initial fresh fuel enrichment 
at 30 GWd/MTU burnup isotopic compositions. Pertinent assembly design parameters for the 
representations are provided in Table 42. Although spacer grid information is presented in Table 
42, the MCNP representations did not represent them in order to maximize system reactivity. 
Throughout this section STD refers to standard, and OFA refers to optimized fuel assembly designs. 
Table 42. Fuel Assembly Parameters 
Assembly Design 
/ Parametera 
B&W 
15x15b 
W 17x17 
(STD)c 
W 17x17 
(OFA)d 
CE 
14x14e 
CE 
15x15e 
CE 
16x16e 
W 
15x15e 
W 15x15 
(OFA)e 
Rod pitch 1.44272 1.25984 1.25984 1.4732 1.397 1.28524 1.43002 1.43002 
(cm [in.]) (0.580) (0.550) (0.506) (0.563) (0.563) 
Assembly pitch 21.81098 21.50364 21.50364 20.574 20.828 20.574 21.42236 21.39696 
(cm [in.]) (8.1) (8.2) (8.1) (8.434) (8.424) 
Rod outer 1.0922 0.94996 0.91440 1.1176 1.06172 0.97028 1.07188 1.07188 
diameter (OD) (cm (0.44) (0.418) (0.382) (0.422) (0.422) 
[in.]) 
Cladding thickness 0.06731 0.05715 0.05715 0.07112 0.06604 0.0635 0.061468 0.061468 
(cm [in.]) (0.028) (0.026) (0.025) (0.0242) (0.0242) 
Rod length 390.366 385.1534 384.704 373.38 355.6 408.94 385.7752 385.699 
(cm [in.]) (147) (140) (161) (151.88) (151.85) 
Active fuel length 360.172 365.76 365.76 347.98 335.28 381 363.22 365.76 
(cm [in.]) (137) (132) (150) (143) (144) 
U mass per 463.63 458.88 423.12 386 kg 412.769 426 kg 469 kg 462.7 kg 
assembly kg kg kg (0.386 kg (910 (0.426 (0.469 (0.4627 
MT) lbs) MT) MT) MT) 
Plenum spring SS304 SS302f SS302f SS302 SS302 SS302 SS302 SS302 
material 
Plenum spring N/A N/A N/A 45.359 22.6796 45.359 18.5973 11.3398 
mass per (0.10) (0.050) (0.10) (0.041) (0.025) 
assembly (g [lb]) 
Plenum length 28.766 17.9654 17.516 21.2725 N/A 24.19858 20.828 20.828 
(cm [in.]) (8.375) (9.527) (8.2) (8.2) 
Upper end-fiting 8.731 15.506 15.506 16.8402 7.9756 24.69642 8.8773 9.017 
length (cm [in.]) (6.63) (3.140) (9.723) (3.495) (3.550) 
Lower end-fitting 16.723 11.951 11.951 7.9375 8.2296 9.68248 6.95452 6.95452 
length (cm [in.]) (3.125) (3.24) (3.812) (2.738) (2.738) 
Intermediate Inconel-Inconelg Zircaloy Zircaloy-Zircaloy-Zircaloy-Inconel-Zircaloyspacer 
grid 718 4 4 4 718 4 
material 
Upper spacer grid Inconel-Inconelg Inconelg Zircaloy-Zircaloy-Zircaloy-Inconel-Inconelmaterial 
718 4 4 4 718 718 
Bottom spacer grid Inconel-Inconelg Inconelg Inconel-Inconel-Inconel-N/A Inconel

Licensing Calculation 
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Table 42. Fuel Assembly Parameters 
Assembly Design 
/ Parametera 
B&W 
15x15b 
W 17x17 
(STD)c 
W 17x17 
(OFA)d 
CE 
14x14e 
CE 
15x15e 
CE 
16x16e 
W 
15x15e 
W 15x15 
(OFA)e 
material 718 625 625 625 718 
Intermediate grid 
length (cm) 
3.81 3.35788 5.71500 4.284h 2.946h 5.432h 3.81 5.715 
Upper grid length 8.573 14.656 14.656 4.284h 2.946h 5.432h 3.81 3.81 
(cm) 
Bottom grid length N/A 3.35788 3.35788 9.044h 6.63h 10.3188 N/A 3.81 
(cm) 
Total number of 7 8 8 9 10 11 7 7 
spacer grids 
Number of guide 16 24 24 5 8k 5 20 20 
tubes 
Guide tube OD 
(cm) 
1.3462 1.22428 
1.08966 
1.20396 
1.08966 
2.832i 1.1978 2.832i 1.382i 1.382i 
Guide tube wall 
thickness (cm [in.]) 
0.04064 0.04064 0.04064 0.091i N/A 0.091i 0.043i 0.043i 
Instrument tube 
OD (cm) 
1.38193 1.22428 1.20396 2.832i 1.059j 1.059i 1.382i 1.382i 
Instrument tube 0.130895 0.04064 0.04064 0.091i 0.069j 0.069i 0.043i 0.043i 
wall thickness 
(cm) 
NOTES: a 
b 
c 
d 
e 
Referenced dimensions in inches provided in parentheses and converted to cm where applicable. 
Parameters from Punatar (2001, Section 2 and Table 3-1). 
Parameters from CRWMS-M&O (1998c, Section 2 and Table 3-1). 
Parameters from CRWMS-M&O (1998d, Section 2 and Table 3-1). 
Parameters from DOE (1987, pp. 2A-55 to 2A-58 and Figure 1-4 for the CE 14x14 design; pp. 2A-67 to 2A70 
and Figure 1-1 for the CE 15x15 design; pp. 2A-73 to 2A-76 and Figure 1-9 for the CE 16x16 design; 
pp. 2A-319 to 2A-322 and drawing 1598E32 for the W 15x15 design; and pp. 2A-325 to 2A-328 and drawing 
f 
g 
h 
i 
1607E93 for the W 15x15 OFA design. 
Material from DOE (1987, pp. 2A-352 and 2A-346) 
References did not specify type, therefore Inconel-718 was used in the representations. 
Values based on Assumption 3.9 
Parameters from Stout and Leider (1997, pp. 2.1.2.2-2 and 2.1.2.2-3); not used in representation due to lack 
j 
k 
of information regarding location. 
Based on Assumption 3.10 
There are eight Zircaloy-4 guide bars (DOE 1987, p. 2A-68) in this design which were not represented in 
order to maximize reactivity 
The assembly materials listed here refer to the upper and lower end-fitting materials and the spacer 
grid materials. The primary material components in the upper and lower end-fitting regions are 
SS304 (see Table9), Inconel (represented as Inconel-718 as shown in Table 10 or Inconel-625 asshown in Table 43), Zircaloy-4 as represented in Table 8, and moderator (represented as water at 
1.0 g/cm3 density). Both the upper and lower end-fitting regions are represented with material 
compositions that represent the homogenization of the components in the regions for each assembly 
design. The homogenization of the base components into single homogenized material compositions 
is performed using Equations 4 through 6. Table 43 presents the material composition for Inconel625 
spacer grids constituent natural isotopic weight percents for use in MCNP using Equations 2 
and 3. Each of the homogenized material compositions is derived in Attachment IV (Workbook 
homog_mats.xls). 

Licensing Calculation 
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Table 43. Inconel 625 Material Composition 
Element/ 
Isotope ZAIDa Wt% 
Element/ 
Isotope ZAIDa Wt% 
C-nat 6000.50c 0.1000 Fe-58b 26058.60c 0.0145 
Si-nat 14000.50c 0.5000 Ni-58b 28058.60c 39.0889 
P-31 15031.50c 0.0150 Ni-60b 28060.60c 15.4562 
S-32 16032.50c 0.0150 Ni-61b 28061.60c 0.6835 
Cr-50b 24050.60c 0.8973 Ni-62b 28062.60c 2.1946 
Cr-52b 24052.60c 17.9955 Ni-64b 28064.60c 0.5769 
Cr-53b 24053.60c 2.0798 Ti-nat 22000.50c 0.4000 
Cr-54b 24054.60c 0.5273 Al-27 13027.50c 0.4000 
Mn-55 25055.50c 0.5000 Co-59 27059.50c 1.0000 
Fe-54b 26054.60c 0.2850 Mo-nat 42000.50c 9.0000 
Fe-56b 26056.60c 4.5934 Nbc 73181.50c 3.6500 
Fe-57b 26057.60c 0.1071 Density = 8.44 g/cm3 
Source: Inco Alloys International 1998, pp. 1 and 2 
a
NOTE: ZAID = MCNP material identifier. 
b 
Expanded constituent natural isotopic weight percents derived using Equations 2 and 3.
c 
Reference identifies this material as “columbium,” which is actually the element niobium. 
Table 44 presents the assembly hardware component masses and Tables 45 and 46 present the 
component material volume fractions for the upper and lower end-fitting regions. Since the spacer 
grids are not being represented in the spacer grid homogenization parameters are not being listed. 
Each of the homogenized material compositions is derived in Attachment IV (Workbook 
homog_mats.xls) along with the volume fractions for components that were not available. 
Table 44. Assembly End-Fitting Hardware Component Masses 
Hardware Part Name 
Upper End-Fitting 
CE 14x14 CE 15x15 CE 16x16 W 15x15 W 15x15 OFA 
Locking posts (kg/assembly) 2.63 (SS304) N/A 7.3 (SS304) N/A N/A 
Hold-down spring 
(kg/assembly) 
1.1 
(Inconel 718)a 
N/A 4.5 
(Inconel 718)a 
1.14 
(Inconel 718)a 
0.96 
(Inconel 718)a 
Flow plate (kg/assembly) 1.45 (SS304) N/A 3.2 (SS304) N/A N/A 
Hold-down plate 1.0 (SS304) N/A 1.8 (SS304) N/A N/A 
(kg/assembly) 
Top nozzle (kg/assembly) N/A 4.5 (SS304) N/A 10.7 (SS304) 6.89 (SS304) 
Lower End-Fitting 
Bottom nozzle (kg/assembly) 5.0 (SS304) 5.4 (SS304) 5.4 (SS304) 5.44 (SS304) 5.44 (SS304) 
Source: DOE 1987, pp. 2A-56, 2A-68, 2A-74, 2A-320, and 2A-326 
a
NOTE: DOE 1987 lists this material as CE nickel alloy. No information is available for this material so it was 
represented as Inconel 718 (See Assumption 3.11). 

Licensing Calculation 
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Table 45. Upper End-Fitting Component Material Volume Fractions 
Assembly Design 
Volume Fractions in Upper End-Fitting Region 
SS304 Inconel Zircaloy-4 Moderator 
B&W 15x15a 0.2756 0.0441 0.0081 0.6722 
W 17x17 STDb 0.1243 0.0168 0.0 0.8589 
W 17x17 OFAc 0.1303 0.0178 0.0051 0.8469 
CE 14x14d 9.0209E-02 1.884E-02 N/A 0.8910 
CE 15x15d 0.1646 N/A N/A 0.8354 
CE 16x16d 0.1489 0.0526 N/A 0.7985 
W 15x15d 0.3325 0.0342 N/A 0.6333 
W 15x15 OFAd 0.2113 0.0284 N/A 0.7603 
a
NOTES: Values from Punatar (2001, Table 2-6) 
b 
Values from CRWMS M&O (1998c, p. 9) 
C 
Values from CRWMS M&O (1998d, p. 15) 
d 
Values derived from information provided in Tables 42 and 44 based on conservation of mass and volume 
Table 46. Lower End-Fitting Component Material Volume Fractions 
Assembly Design 
Volume Fractions in Lower End-Fitting Region 
SS304 Inconel Zircaloy-4 Moderator 
B&W 15x15a 0.1656 0.0306 0.0125 0.7913 
W 17x17 STDb 0.1625 0.0 0.0 0.8375 
W 17x17 OFAc 0.1439 0.0 0.0137 0.8424 
CE 14x14d 0.1884 0.0 0.0 0.8116 
CE 15x15d 0.1915 0.0 0.0 0.8085 
CE 16x16d 0.1668 0.0 0.0 0.8332 
W 15x15d 0.2158 0.0 0.0 0.7842 
W 15x15 OFAd 0.2194 0.0 0.0 0.7806 
a
NOTES: Values from Punatar (2001, Table 2-3) 
b 
Values from CRWMS-M&O (1998c, p. 9) 
C 
Values from CRWMS-M&O (1998d, p. 15) 
d 
Values derived from information provided in Tables 42 and 44 based on conservation of mass and volume 
Table 47 presents the upper and lower fuel rod plenum material volume fractions. 

Licensing Calculation 
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Table 47. Fuel Rod Plenum Material Volume Fractions 
Assembly Design Plenum Location Stainless Steel 
Gas (represented 
as void) Zircaloy-4 
B&W 15x15a Upper 0.0811 0.7793 0.1396 
Lower 0.1569 0.5973 0.2458 
W 17x17 STDa Upper 0.0976 0.8368 0.0655 
Lower 0.1532 0.6389 0.2080 
W 17x17 OFAa Upper 0.1753 0.8247 0.0000 
Lower 0.0000 0.0000 1.0000 
CE 14x14b Upper 0.0162 0.9838 0.0000 
Lowerc 0.0000 0.0000 0.0000 
CE 15x15b Upper 0.0012 0.9988 0.0000 
Lower 0.0000 0.0000 0.0000 
CE 16x16b Upper 0.0142 0.9858 0.0000 
Lower 0.0000 0.0000 0.0000 
W 15x15b Upper 0.0006 0.9994 0.0000 
Lower 0.0000 0.0000 1.0000 
W 15x15 OFAb Upper 0.0003 0.9997 0.0000 
Lower 0.0000 0.0000 1.0000 
a
NOTES: Values derived in Attachment IV, workbook homog_mats.xls
b 
Upper plenum region represented as spring and void based on conservation of mass and volume from 
dimensions listed in Table 42 
c 
Represented as solid Al2O3 spacer at density of 3.97 g/cm3 (Lide, p. 4-39) 
Each of the fuel assembly designs were represented in MCNP using a fuel density of 10.741 g/cm3 
in order to minimize on linear mass loading differences, which are inherent based on the design 
differences. Based on this density and the referenced active fuel lengths, the total mass per assembly 
is greater than its nominal loading, and is based on the design parameters. Therefore trying to keep 
the total mass per assembly constant is not appropriate here. The results for the PWR assembly 
lattice design sensitivity cases are presented in Table 48 and indicate that the B&W 15x15 assembly 
design is the most reactive, but is statistically equivalent at the 95% confidence limit to the W 17x17 
OFA. 
Table 48. Fuel Assembly Lattice Design keff Results 
Assembly Filename keff s 
B&W 15x15 BW15 1.16167 0.00055 
W 17x17 OFA W17OFA 1.16147 0.00055 
W 17x17 STD W17STD 1.15613 0.00058 
W 15x15 OFA W15OFA 1.1593 0.00064 
W 15x15 W15 1.16082 0.00057 
CE 14x14 CE14 1.11257 0.00058 
CE 15x15 CE15 1.11695 0.00059 
CE 16x16 CE16 1.11273 0.00056 

INTENTIONALLY LEFT BLANK

Attachment III: Attachment CD Listing 
This attachment contains a listing and description of the files contained on the attachment CD of this 
report (Attachment IV). The CD was written using ROXIO Easy CD Creator 5 Basic installed on 
CRWMS M&O tag number 150527 central processing unit, and can be viewed on most standard 
CD-ROM drives. The zip archive was created using WINZIP 8.1. The file attributes on the CD are 
as follows: 
Filename File Size 
(bytes) 
File Date File Time Description 
21PWRWP.zip 1,204,743 8/24/2004 08:24a 21-PWR Waste Package Configuration Design 
Drawings 
0130223.pdf 68,126 6/10/2004 10:03a MSDS for representative hydraulic fluid 
cases.zip 51,010,067 9/20/2004 10:58a Archive containing MCNP files 
misc.xls 1,114,112 9/10/2004 08:24a Excel spreadsheet containing various 
geometry and material derivations 
Tuff composition.xls 59,904 9/09/2004 01:02p Excel spreadsheet containing tuff composition 
derivations 
homog_mats.xls 438,784 9/01/2004 11:17a Excel spreadsheet containing fuel assembly 
hardware component derivations 
wstreamplot.xls 710,656 9/20/2004 12:59p Excel spreadsheet containing sorted waste 
stream information 
IDBinputs.xls 265,728 9/10/2004 03:43p Excel spreadsheet containing irradiated fuel 
isotopic compositions 
There are 8 total files for the archive file 21PWRWP.zip with no particular naming system. The files 
contain the dimensions for the 21-PWR waste package configuration. 
There are 672 total files (not including folders) contained in a unique directory structure for the 
archive file cases.zip. Files without on "o" at the end are input files, and files with an "o" at the end 
are output files. The following extracted directory structure corresponds as follows: 
/temp3/PWRLC01CD/temp/*: 
where * corresponds as follows: 
/I.1/ - Contains files listed in Attachment I, Section I.1 with subdirectories burned and fresh 
corresponding to burned fuel isotopic composition and fresh fuel isotopic composition cases, 
respectively. 
/I.2/ -Contains files listed in Attachment I, Section I.2 
/I.3/ -Contains files listed in Attachment I, Section I.3 with subdirectories Hyd_Fluid and 
Water_density corresponding to hydraulic fluid and water density cases, respectively. 
/I.4/ -Contains files listed in Attachment I, Section I.4 
/I.5/ -Contains files listed in Attachment I, Section I.5 with subdirectories Tuff_External, 
InWPVoid, ExtFBT, and Tsol corresponding to the tabulated results in Attachment I, Tables 
34, 35, and 36, respectively.
/I.6/ -Contains files listed in Attachment I, Section I.6 

/I.7/ -Contains files listed in Attachment I, Section I.7 with subdirectories 5mm, 100cp, and Water 
corresponding to the tabulated results in Attachment I, Tables 38 and 39, respectively.
/I.8/ -Contains files listed in Attachment I, Section I.8 
/I.9/ -Contains files listed in Attachment I, Section I.9 
/Fresh_Fuel/ - Contains files used for the determination of the maximum fresh fuel enrichment that 
can be loaded into the waste package with two lower level directories denoted /Post_config/ and 
/Pre_config/ for postclosure and preclosure bounding configurations, respectively. The file naming 
system is as follows: X.X that represents the enrichment in wt% U-235 (ranging from 1.5 to 5.0). 
/LC/ - Contains files used for developing the loading curve as a function of burnup, with four lower 
level directories /7Post/, /7Pre/, /Post/, and /Pre/ representing the postclosure bounding 
representation using a seven zone axial burnup profile, the preclosure bounding representation using 
a seven zone axial burnup profile, the postclosure bounding representation using a single zone axial 
burnup profile, and the preclosure bounding representation using a single zone axial burnup profile, 
respectively. The file naming system is as follows: where the X.X represents the initial enrichment 
in wt% U-235 (i.e., 3.5 is 3.5 wt% U-235 [range from 2.0 to 5.0 wt% U-235]), the YY represents the 
burnup in GWd/MTU (range from 10 to 45 GWd/MTU), and the Z is either a 1 or a 7 denoting a 
single zone axial burnup profile or a seven zone axial burnup profile, respectively. 
/Misload/ - which contains the subdirectories /Nominal/, /Sub10/, and /Sub20/ representing cases 
loaded with fuel assemblies at the design basis loading curve value, cases with a 10 GWd/MTU 
underburned assembly from the design basis required burnup in the central basket location, and 
cases with a 20 GWd/MTU underburned assembly from the design basis required burnup in the 
central basket location. The file naming system is as follows: X.X for the /Nominal/ subdirectory 
cases representing the initial enrichment in wt% U-235 (i.e., 3.5 is 3.5 wt% U-235 [range from 2.5 
to 5.0 wt% U-235]); and X.XmYY in /Sub10/ and /Sub20/ subdirectories with X.X representing the 
initial enrichment in wt% U-235, and the YY is either a 10 or a 20 representing a 10 or 20 
GWd/MTU underburned assembly. 
/II/ - Contains files listed in Attachment II.