44-BWR Waste Package Loading Curve Evaluation Rev 00A, ICN 00

August, 2004

CAL-DSU-NU-000008

1. PURPOSE
The objective of this calculation is to evaluate the required minimum burnup as a function of initial 
boiling water reactor (BWR) assembly enrichment that would permit loading of spent nuclear fuel 
into the 44 BWR waste package configuration as provided in Attachment IV. This calculation is an 
application of the methodology presented in 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 (wt%) U-235, and a burnup range of 0 through 40 GWd/MTU. 
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 BWR 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 of 100 percent of the current BWR projected waste stream being able to be 
disposed of in the 44-BWR waste package with Ni-Gd Alloy absorber plates is contingent 
upon the referenced waste stream being sufficiently similar to the waste stream received for 
disposal. 
 
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 2004i, 
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 for the loaded waste package configuration. The isotopic compositions for 
SNF were calculated in Wimmer (2004) and used as input to the MCNP code (CRWMS M&O 
1998b) 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 the Selection of MCNP Cross Section Libraries report (CRWMS M&O 1998c, 
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 a generic 
General Electric (GE) 7x7 assembly design with no burnable absorber rods, and dimensions from 
Larsen et.al. (1976, pp. A-1 to A-3), and waste package dimensions provided in Attachment IV from 
the following references: BSC 2003; BSC 2004a; BSC 2004b; BSC 2004c; BSC 2004d; BSC 2004e; 
and BSC 2004h. 
3. ASSUMPTIONS 
3.1 ASSEMBLY DESIGN 
Assumption: It was assumed that the GE 7x7 assembly design is the most limiting BWR fuel 
assembly design for reactivity calculations. 
Rationale: The basis for this assumption is that several assembly designs were evaluated in 
Attachment I and the results show the 7x7 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 is used in Sections 5 and 6. 
3.2 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 25C (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. 
3.3 CROSS SECTION SUBSTITUTION 
Assumption: Since the zinc cross section libraries are unavailable, it was assumed that representing 
the zinc material composition in aluminum 6061 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. 
Confirmation Status: This assumption requires no further confirmation based on the stated rationale. 
Use in the Calculation: This assumption is used in Section 5.2.2. 

3.4 IRON AND ALUMINUM OXIDE VOLUME EXPANSION 
Assumption: It was assumed that the volume expansion from the oxidation and hydration of the 
carbon steel and Al 6061 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. 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. 
Use in the Calculation: This assumption is used in Attachment IV. 
3.5 ASSEMBLY HARDWARE 
Assumption: It was assumed that representing the upper and lower tie plate regions, and the channel 
for each assembly design with the same materials produces the same reactivity effects. 
Rationale: The basis for this assumption is that the BWR assemblies are all designed to fit in the 
same area. Slight variances in the component volume fractions of the upper and lower tie plate 
regions are present as well as minor variances in the channel dimensions, but the basic hardware is 
the same. Variations in these components will produce negligible effects on system reactivity. As 
long as these components are represented, the desired effect on system reactivity is present. 
Confirmation Status: This assumption requires no further confirmation based on the stated rationale. 
Use in the Calculation: This assumption is used in Attachment I and IV. 
4. USE OF COMPUTER SOFTWARE 
4.1 MCNP 
The baselined MCNP code (CRWMS M&O 1998b) 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 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 LP-SI.11Q-BSC, Subsection 2, Software Management. 
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 results that 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 (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 
(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. 
5.1 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. 
5.2 MATERIALS 
This section provides an overview of the materials that were selected for use in the MCNP inputs. 
5.2.1 Tuff Material Description 
Waste package configurations were represented with a tuff reflector since this is the composition 
of the drift material. The tuff composition used for the loading curve determinations is presented 
in Table 1.

Table 1. Tuff Material Composition 
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 
5.2.2 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 2. 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 3. 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 4. The thermal shunts were represented as SB-209 A96061 T4 (aluminum 6061) 
as described in Table 5. The basket side and corner guides, and the basket stiffeners were 
represented as Grade 70 A 516 carbon steel as described in Table 6. Stiffeners were placed 
equidistant along the length of the basket in eight axial locations. Waste package basket material 
thicknesses were taken from the BWR 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 = 
A ( At% ) (Eq. 2)
ii 
I 
. A (At% )
ii 
i =1 
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 

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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 
Table 2. SB-575 N06022 Material Composition 
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. 
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 

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Table 3. Material Specifications for SA-240 S31600 
NOTES: a ZAID = MCNP material identifier 
b 
Carbon and nitrogen specifications are from ASM International (1987, p. 931) and remaining material 
compositions are from ASME (2001, Section II, SA-240, Table 1) 
c 
Density is for stainless steel 316 from ASTM (1999, G 1-90, p. 7, Table X1) 
Table 4. Material Specifications for Ni-Gd Alloy (UNS N06464) with 1.5 wt% Gdb 
Source: ASTM B 932-04 2004, Table 1 and Section 8 
a
NOTES: 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 
Element/Isotope ZAIDa Wt% Element/Isotope ZAID Wt% 
C-natb 6000.50c 0.0200 Fe-54 26054.60c 3.7053 
N-14b 7014.50c 0.0800 Fe-56 26056.60c 60.2691 
Si-nat 14000.50c 0.7500 Fe-57 26057.60c 1.4173 
P-31 15031.50c 0.0450 Fe-58 26058.60c 0.1905 
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 
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 

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Table 5. Material Specifications for Al 6061 
Source: ASM International 1990, p. 102 
a
NOTES: ZAID = MCNP material identifier. 
b 
Zn cross-section data unavailable; therefore, it was substituted as Al-27 (See Assumption 3.3). 
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 6. Grade 70 A516 Carbon Steel Composition 
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 
5.2.3 Fuel Assembly MCNP Material Descriptions 
The fuel assembly materials listed in this section refer to the upper and lower tie-plate 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-2 as presented in Table 7.
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 
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 

Table 7. Zircaloy-2 Material Composition 
Source: ASTM B 811-97 2000, p. 2, Table 2 
a
NOTES: ZAID = MCNP material identifier. 
b 
From ASM International 1967, p. 1 
The primary material components in the upper and lower tie-plate regions are SS304 (Table 9),
Zircaloy-2 (Table 7), Zircaloy-4 as represented in Table 8, and moderator (represented as water at 
1.0 g/cm3 density). Both the upper and lower tie-plate regions are represented with material 
compositions that represent the homogenization of all of the components in the regions. The 
homogenization of the base components into single homogenized material compositions is 
performed using Equations 4 through 6. The component material volume fractions were derived 
based on information from Larsen et. al. (1976) in the spreadsheet tie-plate-comp in Attachment IV 
and provided in Table 10. Table 11 presents the base case upper and lower tie-plate homogenized 
material compositions. 
Table 8. Zircaloy-4 Material Composition 
Source: ASTM B 811-97 2000, p. 2, Table 2 
NOTES: a ZAID = MCNP material identifier. 
b 
From ASM International 1990, p. 666, Table 6. 
M 
Density Material dHomogenize =.[()( Material dHomogenize in Fraction Volume )]
. 
mm 
m 
(Eq. 4) 
Element/Isotope ZAIDa Wt% Element/Isotope ZAID Wt% 
Cr-50 24050.60c 0.0042 Ni-58 28058.60c 0.0370 
Cr-52 24052.60c 0.0837 Ni-60 28060.60c 0.0147 
Cr-53 24053.60c 0.0097 Ni-61 28061.60c 0.0007 
Cr-54 24054.60c 0.0024 Ni-62 28062.60c 0.0021 
Fe-54 26054.60c 0.0076 Ni-64 28064.60c 0.0006 
Fe-56 26056.60c 0.1241 O-16 8016.50c 0.1250 
Fe-57 26057.60c 0.0029 Zr-nat 40000.60c 98.1350 
Fe-58 26058.60c 0.0004 Sn-nat 50000.35c 1.4500 
Densityb = 6.55 g/cm3 
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 Density = 6.56b g/cm3 

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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. 
MaterialdHomogenizeinFractionVolume
m
(
)
.
(
.
)
m
ComponentofFractionMass
.
= 
.
.
..
..
.
.
.
.
MaterialdHomogenizeinMaterial 
DensityMaterialdHomogenize 
(Eq. 5) 
.
..... 
. 
ofPercentWeight 
MaterialComponent 
intConstituen 
MaterialdHomogenize 
.
..... 
. 
ofFractionMass 
ComponentofPercentWeight 
tConstituenMaterial 
.
.
... 
.. 
. 
... 
= 
. 
... 
. 
... 
inMaterialComponent 
MaterialdHomogenize 
MaterialComponentin
.
(Eq. 6) 
Table 9. SS304 Material Composition 
Source: ASME (2001, Section II, SA-240, Table 1); Density from ASTM (1999, G 1-90, p. 7, Table X1) 
Table 10. Tie-Plate Component Material Volume Fractions for BWR Assembly Design 
Source: Derived in Attachment IV (spreadsheet GEassembly_Vol.xls) 
Element Wt% Element Wt% 
Carbon 0.080 Chromium 19.000 
Nitrogen 0.100 Manganese 2.000 
Silicon 0.750 Iron 68.745 
Phosphorous 0.045 Nickel 9.250 
Sulfur 0.030 Density = 7.94 g/cm3 
Assembly Design 
Stainless Steel 
Type 304 Zircaloy-2 Zircaloy-4 Moderator 
Upper Tie-Plate 0.0528 0.0641 0.0458 0.8373 
Lower Tie-Plate 0.1528 0.0344 0.0225 0.7903 

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Source: Derived in Attachment IV (spreadsheet GEassembly_Vol.xls) 
a
NOTES: ZAID = MCNP material identifier. 
b 
Values slightly differ in Attachment I sensitivity cases, but since they were kept constant in the sensitivity 
cases, there is no impact on the selected bounding representations 
5.2.4 Fuel Material 
The following information provides the details needed to duplicate the input file specifications. The 
uranium dioxide fresh fuel compositions for each uranium-235 enrichment used in this evaluation 
are specified in Table 12, and were calculated using Equation 7 (Bowman et al. 1995, p. 20) for each 
isotope based on the U-235 wt%. 
Table 11. Tie-Plate Homogenized Material Compositions for BWR 
Fuel Assembly Design 
Element/ 
Isotope ZAIDa 
Upper Tie-Plate 
Wt% 
Lower Tie-Plate 
Wt% 
Flooded Dry Flooded Dry 
H-1 1001.50c 4.7396 3.7212 
C 6000.50c 0.0170 0.0294 0.0408 0.0612 
N-14 7014.50c 0.0212 0.0368 0.0511 0.0765 
O-16 8016.50c 37.6623 0.0790 29.5532 0.0294 
Si 14000.50c 0.1591 0.2759 0.3829 0.5737 
P-31 15031.50c 0.0095 0.0166 0.0230 0.0344 
S 16000.60c 0.0064 0.0110 0.0153 0.0229 
Cr-50 24050.60c 0.1699 0.2947 0.4060 0.6082 
Cr-52 24052.60c 3.4032 5.9038 8.1318 12.1834 
Cr-53 24053.60c 0.3933 0.6823 0.9397 1.4079 
Cr-54 24054.60c 0.0995 0.1727 0.2379 0.3564 
Mn-55 25055.50c 0.4242 0.7358 1.0210 1.5297 
Fe-54 26054.60c 0.8272 1.4350 1.9845 2.9732 
Fe-56 26056.60c 13.4536 23.3393 32.2754 48.3561 
Fe-57 26057.60c 0.3164 0.5489 0.7591 1.1373 
Fe-58 26058.60c 0.0425 0.0738 0.1020 0.1528 
Ni-58 28058.60c 1.3261 2.3006 3.1769 4.7597 
Ni-60 28060.60c 0.5283 0.9165 1.2657 1.8963 
Ni-61 28061.60c 0.0234 0.0405 0.0559 0.0838 
Ni-62 28062.60c 0.0756 0.1311 0.1811 0.2713 
Ni-64 28064.60c 0.0200 0.0347 0.0479 0.0717 
Zr 40000.60c 35.7552 62.028 15.3981 23.0699 
Sn 50000.35c 0.5283 0.9166 0.2275 0.3409 
Density (g/cm3) 1.9768b 1.1395 2.3765b 1.5862 

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 irradiated fuel material compositions were taken from Wimmer (2004, Section 6) and are listed 
in the corresponding inputs in atoms/b-cm. 
Table 12. 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

5.3 MCNP GEOMETRIC DESCRIPTIONS 
The drawing for the 44-BWR waste package configuration is contained electronically in 
Attachment IV. The MCNP representation of the 44-BWR waste package configuration 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. 
5.3.1 Bounding Configuration 
Based on an evaluation of the sensitivity results presented in Attachment I, the internal configuration 
class that results in the highest keff values for the BWR waste package during postclosure was 
selected for the bounding representation. This configuration is based on the waste package internals 
degrading slower than the waste form, with the waste form remaining in place and is representative 
of configuration class IP-1a from YMP (2003, Figure 3-2). 
The geometric representation was based on an igneous event occurring causing the waste package 
and internals to heat up allowing cladding over-pressurization (and thus SNF oxidation) and thermal 
creep of the entire waste package. The magma from the igneous event causes the waste package to 
slump and become porous allowing moisture inside. The configuration is not considered to be able 
to hold a bathtub configuration of water therefore the system is considered dry with the exception 
of having enough moisture present to cause the SNF to hydrate. 
Once the fuel cladding is breached, oxidation of the fuel material can occur and cause clad breach 
propagation (unzipping). 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 2004f, Attachment I, file data0 files.zip, file data0.ymf). Schoepite is more reactive 
than comparable enriched UO2 (See Attachment I) and is therefore used to establish the minimum 
fresh fuel enrichment that can be loaded into the waste package to meet postclosure objectives. The 
derivation of this set of cases was made in Attachment IV (spreadsheet schoepite.xls, sheet Fresh). 
In order to make a proper comparison, the fresh fuel uranium mass was used as the basis for the 
amount of fresh schoepite that could form. Based on this the equivalent schoepite density over the 
active fuel region within the channel would be greater than theoretical (4.874 g/cm3 [BSC 2004f, 
Attachment I, file data0 files.zip, file data0.ymf]). Therefore, it was represented at theoretical 
density with the active fuel length adjusted to what is necessary to conserve uranium mass. 

Licensing Calculation 
Title: 44-BWR Waste Package Loading Curve Evaluation 
Document Identifier: CAL-DSU-NU-000008 REV 00A Page 23 of 34 
5.3.2 Fuel Assembly 
The physical dimensions for the fuel assembly design represented in the MCNP inputs are presented 
in Table 13.
Source: Larsen et. al. 1976, pp. A-1 to A-3 
a
NOTES: The representations use a smeared pellet density over the inner clad diameter. 
b 
Values in parentheses are from the source reference. 
c 
Channel dimensions were taken from Punatar 2001, Table 2-1. 
5.4 INPUT PARAMETER SUMMARY 
Based on the sensitivity studies presented in Attachment I, the following parameters listed in Table 
14 were selected for use in the loading curve generation:
Table 13. 7x7 BWR Fuel Assembly Specifications 
Assembly Component Specificationb 
Lattice 7x7 
Fuel Pellet Outer Diametera 1.23952 cm (0.488 in.) 
Fuel Rod Cladding Thickness 0.08128 cm (0.032 in.) 
Fuel Rod Cladding Inner Diameter 1.26746 cm 
Fuel Rod Cladding Outer Diameter 1.43002 cm (0.563 in.) 
Active Fuel Length 365.76 cm (144 in.) 
Clad Material Zircaloy-2 
Channel Material Zircaloy-2 
Pin Pitch 1.87452 cm (0.738 in.) 
Channel Inner Widthc 13.246 cm 
Channel Thicknessc 0.3048 cm 
Table 14. Loading Curve Parameters 
Parameter Value Selection Basis Applicability 
Fuel Material UO2 Fresh fuel is more reactive than burned fuel Preclosure 
UO3(H2O)2 Attachment I, Section I.8 Postclosure 
Absorber 1.5 wt% Gd Based on typical value of 75% credit (NRC 2000, p. 8-4) Preclosure 
allowed for fixed neutron absorbers and a nominal Gd and 
loading of 2.0 wt% for Ni-Gd Alloy Postclosure 
Moderator Water at density of 1.0 
g/cm3 
Attachment I, Section I.4 Preclosure 
Void Less moderation between cavities of schoepite make the Postclosure 
Ni-Gd alloy less effective as a neutron absorber due to 
decreased thermalization 
Fuel Density 10.741 g/cm3 [a] Attachment I, Section I.2 Preclosure 
4.874 g/cm3 Section 5.3.1 Postclosure 
Reflector 100% Attachment I, Section I.6 Preclosure 
Saturated Tuff 
100% Saturated Tuff Attachment I, Section I.6 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 44 BWR 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 44-BWR 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 15 for the 
preclosure and postclosure bounding configurations, and are illustrated in Figure 1.
Table 14. Loading Curve Parameters 
Parameter Value Selection Basis Applicability 
Geometry Lattice array in 
Standard Vertical 
Attachment I, Section I.3 Preclosure 
configuration (See 
Attachment I, Figure 6) 
Homogeneous fuel Attachment I, Section I.8 Postclosure 
mixture separated by 
steel, aluminum, and Ni-
Gd alloy plates 
Table 15. Fresh Fuel keff Results 
Configuration Preclosure Configuration Postclosure Configuration 
Enrichment (Wt 
% U-235) keff s keff + 2s keff s keff + 2s 
3.0 0.90045 0.00048 0.9017 
3.5 0.90426 0.00051 0.90528 0.94246 0.00049 0.94405 
4 0.93099 0.00057 0.93213 0.97904 0.00050 0.97751 
4.5 0.95484 0.00055 0.95594 1.00672 0.00051 1.00726 
5 0.97522 0.00052 0.97626 1.03234 0.00048 1.03289 

A CL (BU) = 0.980 - 0.0003*BU was taken from Moscalu (2004, p. 30) which is considered 
applicable to configuration class IP-1a from YMP (2003, Figure 3-2). Figure 1 illustrates that the 
maximum fresh fuel enrichment that would meet the loading curve criteria are 3.960 wt% U-235 for 
the preclosure configuration, and 3.999 wt% U-235 for the postclosure configuration. 
1.07 
1.02 
0.97 
0.92 
0.87 
0.82 
Fuel Enrichment (Wt% U-235) 
Figure 1. Fresh Fuel keff Results 
6.2 BURNED FUEL 
These cases used bounding spent nuclear fuel compositions that ensure that the .kISO term from 
Equation 1 can be set to zero. The process for deriving spent fuel isotopics that set .kISO to zero is 
discussed in BSC (2004j). 
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 16. The postclosure irradiated fuel 
compositions were derived in Attachment IV (workbook schoepite.xls, sheet SNF). The minimum 
burnup required for each initial enrichment to meet the CL or USL 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. Any burnup value greater than this will result in a keff less 
than the CL or USL, and is acceptable to be loaded into the waste package. The results for the 
preclosure configuration 5.0 wt% U-235 initial enrichment cases are illustrated in Figure 2 whichhas an extrapolated minimum required burnup to meet the USL. In Figure 2 the USL is presented 
as a function of burnup with the USL equation provided in Table 17. The preclosure 3.5 to 4.5 wt% 
U-235 initial enriched cases were not plotted because they are below the USL at the minimum 
burnup that ensures .kISO is set to zero. The postclosure cases were not plotted because they are 
keff+2s 
iCL = 0.98 
USL = 0.93 
Max. fresh fuel 
enrichment to 
satsfy USL = 
3.960 wt% U-235 
Max. fresh fuel 
enrichment to 
satisfy CL = 
3.999 wt% U-235 
2 2.53 3.54 4.55 5.5 
lPreclosure Configuration Postcosure Configuration 

Licensing Calculation 
Title: 44-BWR Waste Package Loading Curve Evaluation 
Document Identifier: CAL-DSU-NU-000008 REV 00A Page 26 of 34 
below the CL at the minimum burnup that ensures .kISO is set to zero. The minimum burnups that 
ensure .kISO is set to zero come from Wimmer (2004, Figure 34), and are provided in Table 18. 
Table 17. Spent Nuclear Fuel Upper Limit Functions 
NOTE: Moscalu (2004, p. 30) provides the CL function which is transformed into an USL function using .km = 0.05 
(see Section 5 for details of USL transformation) 
Table 16. Spent Nuclear Fuel keff Results 
Initial Enrichment Burnup Preclosure Postclosure 
(Wt% U-235) (GWd/ MTU) keff s keff+2s keff s keff+2s 
3.5 10 0.87524 0.00051 0.87626 
15 0.86536 0.00051 0.86638 
20 0.84816 0.00052 0.84920 
25 0.83142 0.00048 0.83238 
30 0.81621 0.00048 0.81717 
35 0.80296 0.00056 0.80408 
40 0.78970 0.00052 0.79074 
4.0 10 0.89266 0.00055 0.89376 0.92519 0.00045 0.92609 
15 0.88031 0.00055 0.88141 0.90865 0.00051 0.90967 
20 0.86251 0.00052 0.86355 0.88493 0.00049 0.88591 
25 0.84574 0.00056 0.84686 
30 0.82907 0.00049 0.83005 
35 0.81445 0.00047 0.81539 
40 0.80024 0.00053 0.80130 
4.5 10 0.90795 0.00056 0.90907 0.94453 0.00049 0.94551 
15 0.89377 0.00055 0.89487 0.92687 0.00045 0.92777 
20 0.87522 0.00051 0.87624 0.90481 0.00049 0.90579 
25 0.85943 0.00056 0.86055 
30 0.84085 0.00049 0.84183 
35 0.82633 0.00044 0.82721 
40 0.81118 0.00050 0.81218 
5.0 10 0.92162 0.00052 0.92266 0.96366 0.00047 0.96460 
15 0.90784 0.00054 0.90892 0.94440 0.00045 0.94530 
20 0.88907 0.00049 0.89005 0.92100 0.00050 0.92200 
25 0.87044 0.00052 0.87148 
30 0.85496 0.00052 0.85600 
35 0.83772 0.00052 0.83876 
40 0.82230 0.00050 0.82330 
Repository Period Trend Parameter Equation 
Postclosure Burnup CL (BU) = 0.980 - 0.0003*BU 
Preclosure Burnup USL (BU) = 0.980 - 0.0003*BU - 0.05 

0 5 1015202530354045 
Burnup (GWd/MTU) 
Figure 2. Spent Nuclear Fuel keff Results for 5.0 Wt% U-235 Initial Enrichment 
For generating the loading curve initial enrichments greater than 3.960 wt% U-235 (see Figure 1)
will require burnup credit in order to remain at or below the USL, and enrichments greater than 
3.999 for the CL. Wimmer (2004, Figure 34) illustrates that the minimum burnup required to set 
.kISO to zero is 10.3 GWd/MTU for 4.0 wt% U-235 initial enriched fuel assemblies, which can also 
be used as the minimum for enrichments greater than 3.960 wt% U-235. Based on the results 
presented in Table 16, the minimum burnups from Wimmer (2004, Figure 34) will result in keff 
values below the USL and CL and will be used in establishing the loading curve. 
The minimum required burnups as a function of initial enrichment are presented in Table 18. Also, 
an additional margin for a burnup uncertainty of five percent is included. 
a
NOTES: Values are minimums from Wimmer 2004, Figure 34, to ensure conservative isotopic compositions 
b 
Required minimum burnup including 5% uncertainty associated with assembly burnup records 
c 
Interpolated from 4.0 and 5.0 wt% U-235 minimum values 
keff + 2s 
i0.8 
0.82 
0.84 
0.86 
0.88 
0.9 
0.92 
0.94 
0.96 
Extrapolated Mn. Burnup required to 
meet USL = 8.227 GWd/MTU 
USL(BU) Extrapolation 
Table 18. Minimum Required Burnups for Intercept of Upper Subcritical Limit 
Initial Enrichment (Wt% U-235) Burnup (GWd/MTU)a 5% BU Unc.b 
3.960 0 0 
4.0 10.3 10.82 
4.5 18.45c 19.37 
5.0 26.6 27.93 

LicensingCalculationTitle: 44-BWR Waste Package Loading Curve EvaluationDocument Identifier: CAL-DSU-NU-000008 REV 00APage 28 of 346.3 WASTE STREAM COMPARISONThe waste stream inventory in terms of number of fuel assemblies at given burnups and enrichmentswas taken from 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 wasselected based on Licensing Position LP-009, Waste Stream Parameters (Williams 2003). Theresults of the loading curve compared against the waste stream inventory are presented in Figure 3. 
The squares in the legend indicate number groupings of assemblies at a particular burnup andenrichment (e.g., 100-199 indicates that there are 100 to 199 assemblies at a listed burnup andenrichment); Nominal LC is the loading curve based on the nominal required minimum burnup; and5% BU Unc. LC is the loading curve adjusted to accommodate five percent uncertainty associatedwith the reported assembly burnups. The waste stream information that was extracted and sortedis provided in Attachment IV as the workbook wstreamplot.xls.
0102030405060700.01.02.03.04.05.0Initial Enrichment (Wt% U-235)
Burnup (GWd/MTU)
1-99100-199200-299300+
Nominal LC5% BU Unc. LCAcceptableUnacceptableFigure 3. Loading Curve and Projected Waste Stream6.4 SUMMARY OF RESULTSResults presented in Attachment I (Table 26) illustrate that the 44-BWR waste package with Ni-GdAlloy absorber plates will not go critical with commercial fuel assemblies if there is no moderatorpresent within the waste package. 
In equation notation the loading curve is as shown in Equation 11:
BU = 270.375*E - 1070.69 (3.960 wt% U-235 < E = 4.0 wt% U-235)
BU = 17.115*E - 57.645 (4.0 wt% U-235 < E = 5.0 wt% U-235)(Eq. 11)

The results of this report allow 100 percent of the current BWR projected waste stream to be 
disposed of in the 44-BWR waste package with Ni-Gd Alloy absorber plates. 
All outputs are reasonable compared to the inputs and the results of this calculation are suitable for 
their intended use. 

7. REFERENCES
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. 
ASM International 1967. "Zircaloy-2, Nuclear Reactor Alloy, Filing Code:Zr-3 Zirconium Alloy." 
Alloy Digest, (July), . Materials Park, Ohio: ASM International. TIC: 239929. 
ASM International. 1987. Corrosion. Volume 13 of Metals Handbook. 9th Edition. Metals Park, 
Ohio: ASM International. TIC: 209807. 
ASM International 1990. 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. 
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. 
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. 
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) 2002. External Criticality Calculation for DOE SNF Codisposal 
Waste Packages. CAL-DSD-NU-000001 REV A. Las Vegas, Nevada: Bechtel SAIC Company. 
ACC: MOL.20021105.0162. 
BSC (Bechtel SAIC Company) 2003. Design and Engineering Organization, 44-BWR Waste 
Package Configuration. 000-MW0-DSU0-00502-000-00A. Las Vegas, Nevada: Bechtel SAIC 
Company. ACC: ENG.20031021.0004. 
BSC (Bechtel SAIC Company) 2004a. Design and Engineering, 44 BWR Side Guide. 000-MW0-
DSU0-01501-000-00A. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20040305.0004. 
BSC (Bechtel SAIC Company) 2004b. Design and Engineering, 44 BWR Corner Guide. 000-MW0-
DSU0-01601-000-00A. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20040303.0052. 
BSC (Bechtel SAIC Company) 2004c. 44 BWR A, B, D, F & G Fuel Plates. 000-MW0-DSU0-
01801-000-00B. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20040804.0002. 
BSC (Bechtel SAIC Company) 2004d. 44 BWR C & E Fuel Plates. 000-MW0-DSU0-01901-000-
00B. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20040804.0003. 
BSC (Bechtel SAIC Company) 2004e. 44-BWR Waste Package Configuration. 000-MW0-DSU0-
00503-000-00C. Las Vegas, Nevada: Bechtel SAIC Company. ACC: ENG.20040708.0007. 
BSC (Bechtel SAIC Company) 2004f. 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) 2004g. 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) 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. Q-List. 000-30R-MGR0-00500-000-000 REV 00. Las Vegas, 
Nevada: Bechtel SAIC Company. ACC: ENG.20040721.0007. 
BSC (Bechtel SAIC Company) 2004j. Isotopic Model for Commercial SNF Burnup Credit. CAL-
DSU-NU-000007 REV 00A. Las Vegas, Nevada: Bechtel SAIC Company. ACC: 
DOC.20040818.0001. 

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. Software Code: MCNP. V4B2LV. HP, HPUX 9.07 and 10.20; PC, Windows 
95; Sun, Solaris 2.6. 30033 V4B2LV. 
CRWMS M&O 1998c. 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 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) 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.20040823.0004. 
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. 
Gelest, Inc. 2004. Gelest Silicone Fluids : Stable, Inert Media. Morrisville, Pennsylvania: Gelest, 
Inc. TIC: 256122. 
GS000308313211.001. Geochemistry of Repository Block. Submittal date: 03/27/2000. 
Harwell, J.W. 2003. Commercial Reactor Reactivity Analysis for Grand Gulf, Unit 1. 32-5029393-
00. [Lynchburg, Virginia]: Framatome ANP. ACC: DOC.20040109.0003. 
Larsen, N.H.; Parkos, G.R.; and Raza, O. 1976. Core Design and Operating Data for Cycles 1 and 
2 of Quad Cities 1. EPRI NP-240. Palo Alto, California: Electric Power Research Institute. TIC: 
237267. 
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. 
Lide, D.R., ed. 2002. CRC Handbook of Chemistry and Physics. 83rd Edition. Boca Raton, Florida: 
CRC Press. TIC: 253582. 

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. 
Moscalu, D.R. 2004. Range of Parameters for BWR SNF in a 44-BWR Waste Package. 32-5041139-
00. Las Vegas, Nevada: Areva. ACC: DOC.20040623.0004.
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. 
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. 
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 Grand Gulf Unit 
1. TDR-UDC-NU-000002 REV 00. Las Vegas, Nevada: Bechtel SAIC Company. ACC: 
MOL.20011008.0008. 
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. 
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. 
Wimmer, L.B. 2004. Isotopic Generation and Confirmation of the BWR Appl. Model. 32-5035847-
01. Las Vegas, Nevada: Areva. ACC: DOC.20040630.0007.
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. 

8. ATTACHMENTS 
Table 19 presents the attachment specifications for this calculation file. 
Table 19. Attachment Listing 
Attachment # # of Pages Date Created Description 
I 14 N/A Sensitivity studies 
II 6 N/A Alternative neutron absorber evaluation 
III 2 N/A Listing of contents on Attachment IV 
IV N/A 06/15/2004 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 (44BWRWP.zip). Each configuration is represented with dry tuff surrounding the 
waste package. The dry tuff composition for each of the sensitivity cases is represented as having 
a porosity of 15.7 % (BSC 2002, p. 10) and a tuff composition as listed in Table 20.
Source: BSC 2002, Table 5-3. 
NOTE: Derived elemental/isotopic number densities for MCNP inputs are provided in Attachment IV, spreadsheet 
Tuff composition.xls, sheet Sens_tuff 
I.1 ASSEMBLY LATTICE 
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 7x7, 8x8, and 9x9 geometric arrangements in pure water. The 8x8 and 9x9 base assembly 
design parameters were taken from Punatar (2001, Sections 2 and 3). The 8x8 design was evaluated 
with 2 water rods, 4 water rods, 4 lattice cells containing only water (which is consistent with 1 large 
water rod), and the 9x9 design was evaluated with 5 water rods. The 7x7 fuel assembly design 
parameters are from Larsen et. al. (1976). These cases were evaluated using a fresh fuel enrichment 
of 5.0 wt% U-235 in a nominal waste package configuration. The results of this set of cases are 
presented in Table 23 and illustrated in Figure 4. The channel and upper and lower tie plate region 
compositions as derived in Attachment IV (spreadsheet GEassembly_Vol.xls) were maintained for 
each of the designs, as well as the fuel pellet density. 
Table 20. Tuff Composition for Sensitivity Cases 
Compound Wt% 
SiO2 76.83 
Al2O3 12.74 
FeO 0.84 
MgO 0.25 
CaO 0.56 
Na2O 3.59 
K2O 4.93 
TiO2 0.1 
P2O5 0.02 
MnO 0.07 

Licensing Calculation 
Title: 44-BWR Waste Package Loading Curve Evaluation 
Document Identifier: CAL-DSU-NU-000008 REV 00A Page I-2 of 14 
Source: Punatar 2001, Section 2 
a
NOTES: The base case representations use a smeared pellet density over the inner clad diameter. 
b 
Values were averaged and taken directly from Harwell (2003, case mcnp20, assembly K2) 
Table 22. 7x7 BWR Fuel Assembly Specifications 
Source: Larsen et. al. 1976, pp. A-1 to A-3 
a
NOTES: The base case representations use a smeared pellet density over the inner clad diameter. 
b 
Values in parentheses are from the source reference. 
c 
Channel dimensions were taken from Punatar 2001, Table 2-1. 
Table 23. Fuel Assembly Lattice Design keff Results 
Table 21. 8x8 and 9x9 BWR Fuel Assembly Specifications 
Assembly Component Specification 
Lattice 8x8 9x9 
Fuel Pellet Outer Diametera 1.02997 cm N/A 
Fuel Rod Cladding Inner Diameter 1.05156 cm 0.95008 cmb 
Fuel Rod Cladding Outer Diameter 1.22936 cm 1.10136 cmb 
Number of Water Rods 2 5 
Channel Inner Width 13.246 cm 13.246 cm 
Channel Thickness 0.3048 cm 0.3048 cm 
Active Fuel Length 381 cm 381 cm 
Clad Material Zircaloy-2 Zircaloy-2 
Channel Material Zircaloy-2 Zircaloy-2 
Pin Pitch 1.61544 cm 1.43002 cm 
Assembly Component Specification 
Lattice 7x7 
Fuel Pellet Outer Diametera 1.23952 cm (0.488 in.)b 
Fuel Rod Cladding Thickness 0.08128 cm (0.032 in.) 
Fuel Rod Cladding Inner Diameter 1.26746 cm 
Fuel Rod Cladding Outer Diameter 1.43002 cm (0.563 in.) 
Active Fuel Length 365.76 cm (144 in.) 
Clad Material Zircaloy-2 
Channel Material Zircaloy-2 
Pin Pitch 1.87452 cm (0.738 in.) 
Channel Inner Widthc 13.246 cm 
Channel Thicknessc 0.3048 cm 
Assembly Description Designation Filename keff s 
8x8 with 2 water rods 8x8 2WR 8x8wr2 0.95385 0.00056 
7x7 7x7 7x7 0.96562 0.00057 
8x8 with 4 water rods 8x8 4WR 8x8wr4 0.95447 0.00053 
9x9 with 5 water rods 9x9 5WR 9x9 0.95597 0.0005 
8x8 with 4 empty cells in center for water 
(representitive of 1 big water rod) 8x8 1LGWR 8x8wr1B 0.95838 0.00057 

keff +2s 
0.968
0.966
0.964
0.962
0.96
0.958
0.956
0.954
0.952
0.95 
0.948 
8x8 2WR 7x7 8x8 4WR 9x9 5WR 8x8 1LGWR 
Assembly Type 
Figure 4. Fuel Assembly Lattice Design keff Results 
Based on the results presented in Table 23 and Figure 4, the representation for the 7x7 assembly 
produces the highest keff. 
I.2 FUEL DENSITY EFFECTS 
Variations in fuel density were evaluated. Two sets of cases were evaluated with the 8x8 assembly 
design with 2 water rods - smeared and unsmeared. 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 fuel assembly and maintaining the assembly mass. Fuel density values were varied from 9.78 
g/cm3 through 10.696 g/cm3 for a representative assembly with 5.0 wt% U-235 fresh fuel. The 
results of this set of cases are presented in Table 24 and illustrated in Figure 5. 
Table 24. Fuel Density keff Results 
Smeared Unsmeared 
Fuel Density 
(g/cm3) keff s Filename 
Fuel Density 
(g/cm3) keff s Filename 
9.78a 0.95385 0.00056 8x8wr2 9.896 0.95157 0.00052 d9 
9.88 0.95549 0.00057 d2 9.996 0.95143 0.00052 d8 
9.98 0.95684 0.00059 d3 10.096 0.95200 0.00055 d7 
10.08 0.95892 0.00053 d4 10.196a 0.95325 0.00051 d1 
10.18 0.95877 0.00053 d5 10.296 0.95513 0.00053 d2 
10.28 0.95945 0.00056 d6 10.396 0.95689 0.00053 d3 

a
NOTES: Cases maintained equal mass of UO2
b 
MCNP representation had water filling gap between fuel and clad, this configuration is considered 
nonmechanistic for any extended period of time since the fuel material would oxidize and hydrate forming 
configurations similar to Section I.8 
keff 
0.966 
0.964 
0.962 
0.96 
0.958 
0.956 
0.954 
0.952 
0.95 
0.948 
9.6 9.8 10 10.2 10.4 10.6 10.8 
Fuel Density (g/cm3) 
Figure 5. Fuel Density keff Results 
I.3 WASTE PACKAGE FUEL ASSEMBLY GEOMETRY 
Variations in waste package fuel assembly geometry were evaluated. This set of cases investigated 
the effects of different positioning of the fuel assemblies within the waste package. Three 
configurations where evaluated using the 7x7 assembly design. 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 6 through 8 and the results presented in Table 25. 
Base case fresh fuel compositions correspond to a fuel assembly with 5.0 wt% U-235 initial 
enrichment. 
Table 24. Fuel Density keff Results 
Smeared Unsmeared 
Fuel Density 
(g/cm3) keff s Filename 
Fuel Density 
(g/cm3) keff s Filename 
10.38 0.96121 0.00059 d7 10.496 0.95781 0.00048 d4 
10.48 0.96269 0.00052 d8 10.596 0.95862 0.00057 d5 
10.58 0.96228 0.00058 d9 10.696 0.95909 0.00057 d6 
10.196a, b 0.95855 0.00056 d0 
Smeared Unsmeared 

Figure 6. Standard Vertical Position Waste Package Geometry 
Figure 7. Standard Horizontal Position Waste Package Geometry 

Figure 8. Rotated Horizontal Position Waste Package Geometry 
Table 25. Fuel Assembly Geometry keff Results 
Based on the results presented in Table 25 the standard vertical waste package geometry results in 
the highest keff. 
I.4 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 5.0 wt% U-235 initial enrichment. The results of this set 
of cases are presented in Table 26 and are illustrated in Figure 9. 
Configuration keff s Filename 
Standard Vertical 0.96562 0.00057 7x7 
Standard Horizontal 0.95214 0.00054 horiz1 
Rotated Horizontal 0.95162 0.00056 h2full 

keff 
1.1
1
0.9 
0.8 
0.7 
0.6 
0.5 
0.4 
0.3 
0.2 
0 0.2 0.4 0.6 0.8 1 1.2 
Moderator Density (g/cm3) 
Figure 9. Moderator Density Sensitivity Results 
Another moderator material was evaluated instead of water. 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: 
Table 26. Moderator Density Sensitivity Results 
Moderator Density 
(g/cm3) keff s Filename 
0.0 0.40131 0.00018 case0 
0.1 0.59716 0.00037 case1 
0.2 0.69781 0.00041 case2 
0.3 0.76799 0.00044 case3 
0.4 0.82015 0.0005 case4 
0.5 0.86029 0.00052 case5 
0.6 0.88941 0.00051 case6 
0.7 0.91526 0.00054 case7 
0.8 0.93532 0.00051 case8 
0.9 0.95149 0.00056 case9 
1.0 0.96562 0.00057 7x7 

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 (limited quantity indicates this is non-
mechanistic, but will bound all other configurations (keff = 0.91226, s = 0.00054) 
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 = 0.90315, s = 0.00050) 
Case3 - used a density of 0.8 g/cm3 (keff = 0.89414, s = 0.00052) 
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.5 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 (44BWRWP.zip). A comparison was made between cases with and without the 
trunion region as well as with and without the stiffeners being represented. Configurations were 
evaluated using the 7x7 assembly design with 5.0 wt% U-235 fresh fuel. The results of this set of 
cases are presented in Table 27.
These results indicate that the presence of trunions or stiffeners in the representation have an 
insignificant impact on system reactivity. 
Table 27. Simplified Geometry Results 
Case Description keff s Filename 
Base case with no trunion region and basket 
stiffeners present 
0.96562 0.00057 7x7 
Base case with trunion region and basket 
stiffeners present 
0.96524 0.00053 7x7wt 
Base case with no trunion region and no 
stiffeners present 
0.96731 0.00055 7x7wo 

I.6 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 7x7 assembly design with 5.0 wt% U-235 fresh fuel. The results for the external tuff set of cases 
are presented in Table 28. Cases where the tuff material was uniformly dispersed within the waste 
package, but external to the channel are presented in Table 29. 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, and there is no mechanism to keep water only in the inside of the channels. 
Also, if tuff could migrate to the inside of the basket geometry, then it would be able to get into the 
channel area, where it would exclude moderator, and thus reduce keff. Cases where the tuff would 
most likely accumulate (in the external regions of the basket geometry) were evaluated and the 
results are presented in Table 30.
Table 29. Waste Package Tuff Internal Configuration 1 Evaluation Results 
The final four cases in Table 29 are non-mechanistic but are presented for illustrative purposes only. 
Table 28. External Waste Package Tuff Evaluation Results 
Case Description keff s Filename 
Dry tuff outside waste package 0.96562 0.00057 7x7 
100% saturated tuff outside 
waste package 0.96606 0.00052 tuffsat 
Water outside waste package 0.96611 0.00054 7x7wr 
Void outside waste package 0.96688 0.00054 7x7vr 
Case Description keff s Filename 
Dry tuff in all void regions 0.47835 0.00021 t0all 
100% saturated tuff in all void 
regions 0.69185 0.00041 t100all 
Dry tuff in all void regions 
external to channel which 
contains pure water 
1.09382 0.00053 tuff 
25% saturated tuff in all void 
regions external to channel which 
contains pure water 
1.08553 0.00059 tuff25 
50% saturated tuff in all void 
regions external to channel which 
contains pure water 
1.07713 0.00051 tuff50 
75% saturated tuff in all void 
regions external to channel which 
contains pure water 
1.07181 0.00045 tuff75 
100% saturated tuff in all void 
regions external to channel which 
contains pure water 
1.06412 0.00050 tuff100 

These results indicate that a dry tuff reflector inside the package and around the fuel assemblies with 
water inside the channels produces the highest keff. This case is non-mechanistic as that tuff that 
accumulates in the waste package would be mobile and reach a configuration similar to that 
described by case t100all, which results in a very low keff. Therefore, tuff accumulation inside the 
waste package is considered non-mechanistic. 
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 MCNP_BWR_Geometries.xls, sheet Thinning plates [Ni-
Gd]) 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 3 mm of Ni-Gd 
absorber remaining. Since geochemistry calculations for this material and waste package 
configuration are not available, varied amounts of estimated corrosion product composition were 
evaluated. The corrosion products represented come from the fuel basket tubes and thermal shunts, 
with the amounts varied from 0% to 100%. The Ni-Gd absorber corrosion products were 
represented as being removed from the system since this would decrease the amount of neutron 
absorber present (therefore increasing system reactivity). The configurations were evaluated using 
the 7x7 assembly design with 5.0 wt% U-235 fresh fuel. The results for this set of cases are 
presented in Table 31.
Table 30. Waste Package Tuff Internal Configuration 2 Evaluation Results 
Case Description keff s Filename 
Dry tuff in all outer regions, pure 
water in inner regions 0.97752 0.00052 tout1 
25% saturated tuff in all outer 
regions, pure water in inner 
regions 
0.97435 0.00054 tout2 
50% saturated tuff in all outer 
regions, pure water in inner 
regions 
0.97381 0.00055 tout3 
75% saturated tuff in all outer 
regions, pure water in inner 
regions 
0.97225 0.00056 tout4 
100% saturated tuff in all outer 
regions, pure water in inner 
regions 
0.97279 0.00054 tout5 

These results indicate that the configuration is more reactive without corrosion product composition 
represented in the basket cells. 
Table 32 illustrates results as a function of absorber plate (Ni-Gd Alloy) thickness, with 100% 
corrosion product retention. 
Results indicate that with 100% corrosion product retention, results are within one sigma (within 
statistical uncertainty of calculation) for 0 to 20% plate removal, and within two sigma through 40% 
plate removal. Therefore, the plates have the same relative effectiveness at 5 mm and 3 mm 
thickness. 
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 2004g, 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 2004f, 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 5.0 wt% U-235 fresh fuel schoepite composition for each of the runs. The compositions were 
derived in Attachment IV (spreadsheet schoepite.xls, sheet scoping) for the fuel material in 
conjunction with spreadsheet MCNP_BWR_Geometries.xls, sheet Sch kinf. Mineral densities used 
Table 31. keff Results for 3mm Thick Absorber Plate Cases 
Corrosion Product Retained keff s Filename 
0% 0.94116 0.00056 u0cp 
33% 0.93027 0.00053 u33cp 
66% 0.92652 0.00053 u66cp 
100% 0.92770 0.00049 u100cp 
Table 32. keff Results as a function of Absorber Plate Thickness 
% Absorber Plate 
Removed 
Remaining Plate 
Thickness (mm) keff s Filename 
0 5 0.92662 0.00051 d0u 
10 4.5 0.92724 0.00052 d10u 
20 4.0 0.92713 0.00054 d20u 
30 3.5 0.92742 0.00056 d30u 
40 3.0 0.92770 0.00049 d40u 

Licensing Calculation 
Title: 44-BWR Waste Package Loading Curve Evaluation 
Document Identifier: CAL-DSU-NU-000008 REV 00A Page I-12 of 14 
in the spreadsheets were taken from Roberts (et. al 1990). The results and a brief description of the 
cases are provided in Table 33.
a
NOTES: Typical mineral oxidized aluminum forms into (Al[OH]3)
b 
Typical mineral iron forms into (FeOOH) 
c 
Typical mineral iron forms into (Fe2O3) 
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 34.
Table 33. Compromised Fuel Assembly Sensitivity Cases 
Filename Case Description k8 s 
Case0 Fresh fuel base case for comparison 1.01259 0.00129 
Case1 Schoepite expanded around clad; nominal standard vertical 
geometry; dry conditions 1.07354 0.00107 
Case2 
Schoepite expanded around clad; aluminum shunts and 
fuel basket tube have corroded into gibbsitea and goethiteb , 
respectively, occupying original volumes; nominal standard 
vertical geometry, dry conditions 
1.00583 0.00119 
Case3 Same as Case2 but the gibbsite and goethite volume 
expansion is represented along active fuel length 0.9781 0.00119 
Case4 
Same as Case1 but the system is compressed to the point 
where the basket plates are in contact with the fuel basket 
tubes 
1.07381 0.00102 
Case5 
Same configuration as Case4 but the basket plate 
materials and fuel basket tube have oxidized into gibbsite 
and goethite 
1.01587 0.0012 
Case6 Same as Case5 but the waste package compression is in 
the vertical direction with expansion in the x direction 1.015 0.00117 
Case7 Like Case2 but fuel basket tube is represented as 
hematitec instead of goethite 1.03268 0.00117 
Case8 Same as Case5 but the fuel basket tube is hematite 
instead of goethite 1.03903 0.00118 
Case9 
Same conditions as Case1 but system is fully collapsed so 
there is no spacing betweeen channel, fuel basket tube, 
and basket 
1.08936 0.00107 
Case10 
Same as Case3 but the fuel basket tube corrosion product 
composition has settled within the basket cell and 
assembly is in standard horizontal position geometry (See 
Figure 7) 
1.02600 0.00119 
Case11 Same as Case1 but water fills all void space 0.98175 0.00115 

Licensing Calculation 
These results indicate that waste packages have a negligible neutronic influence on other waste 
packages. 
Table 34. Waste Package Interaction Results 
Case Description keff s Filename 
Single waste package with no others, filled with 
water and void outside 
0.96688 0.00054 7x7vr 
Infinite array of waste packages touching, 
water inside and void outside 
0.96507 0.00052 c1 
Infinite array of waste packages touching, void 
inside and water outside 
0.38419 0.00021 c2 
Infinite array of waste packages touching, 
water inside and water outside 
0.96592 0.00056 c3 
Infinite array of waste packages with 1 cm 
spacing, water inside and void outside 
0.96656 0.00058 c4 
Infinite array of waste packages with 1 cm 
spacing, void inside and water outside 
0.38018 0.00024 c5 
Infinite array of waste packages with 1 cm 
spacing, water inside and water outside 
0.96673 0.00056 c6 

INTENTIONALLY LEFT BLANK

Attachment II: Alternative Neutron Absorber Evaluation 
Since the loss of neutron absorber precludes the ability to load the 44-BWR waste package, 
alternative means for criticality control were evaluated. These cases were just a set of scoping 
studies performed in order to evaluate the effects of alternative materials on a representative waste 
package configuration. Candidate materials were evaluated in order to determine the relative impact 
on criticality control for the waste package. The materials were selected based on the assumption 
that their corrosion resistance is adequate for 10,000 years. Due to their superior corrosion 
resistance, it is assumed that the absorber basket maintains structural integrity in these evaluations. 
The candidate materials selected and evaluated are as follows: 
 
Ni-Gd Alloy - Gd loadings of 2.1, 1.9, 1.5, 1.0, and 0.5 wt% 
 
Alloy 22 (SB-575 N06022) with compartmentalized absorber blocks - ZrB2 and B4C were 
evaluated, both at the following B10 loadings: 18.4, 50.0, 99.0 wt% 
An intact basket configuration corresponding to that used for the preclosure configuration was 
evaluated to determine the effects on system reactivity of changing the material composition of the 
absorber material. The base geometric configuration is illustrated below in Figure 10. For the 
compartmentalized absorber cases, the geometric arrangement is illustrated in Figures 11 and 12.
Figure 10. Base Geometric Arrangement 
Channel 
Absorber 
Shunt 
Fuel Rod 
Water Rod 
Fuel Basket 
Tube 
Thermal 

Title
: 44-BWR Waste Package Loading Curve Evaluation
Document Identifier: CAL-DSU-NU-000008 REV 00A Page II-2 of 6
Figure 11. Top View of Absorber Blocks Encased in Alloy 22Figure 12. Side View of Absorber Blocks Encased in Alloy 22
Absorber 
Block 
Alloy 22 
Absorber 
Block 
Alloy 22 

The derivation of the parameters for the input files was made in Attachment IV (spreadsheet 
MCNP_BWR_Geometries.xls, sheets plug geo and special mats). In these specifications, the 
absorber material compositions were represented at 75% of the theoretical density values from Lide 
(2002, pp. 4-47 and 4-96). The 75% value is typically the maximum amount of credit granted by 
the NRC for fixed neutron absorbers (NRC 2000, p. 8-4). 
Table 35 provides a listing of the results of the sensitivity studies in terms of their relationship to 
the base preclosure case (taken from Section 6). Each case evaluated used 5.0 wt% U-235 fresh fuel 
in a 44-BWR waste package configuration. 
Table 35. BWR Waste Package Configuration Neutron Absorber Sensitivity Results 
Material Description keff s .keff 
a 
Equivalent 
Enrichment 
offset from LC 
Ni-Gd alloy with 2.1 wt% Gd 0.97027 0.00054 0.0050 -0.099 
Ni-Gd alloy with 1.9 wt% Gd 0.97218 0.00050 0.0030 -0.059 
Ni-Gd alloy with 1.5 wt% Gd 0.97522 0.00052 0.0000 0.000 
Ni-Gd alloy with 1.0 wt% Gd 0.97984 0.00059 -0.0046 0.086 
Ni-Gd alloy with 0.5 wt% Gd 0.98845 0.00055 -0.0132 0.246 
Compartmentalized Alloy-22 with 
B4C (nat B) 0.91234 0.00057 0.0629 -1.449 
Compartmentalized Alloy-22 with 
B4C (50 wt% B10) 0.87434 0.00053 0.1009 -2.384 
Compartmentalized Alloy-22 with 
B4C (99 wt% B10) 0.84716 0.00053 0.1281 -3.053 
Compartmentalized Alloy-22 with 
ZrB2 (nat B) 0.92895 0.00058 0.0463 -1.040 
Compartmentalized Alloy-22 with 
ZrB2 (50 wt% B10) 0.89308 0.00058 0.0821 -1.923 
Compartmentalized Alloy-22 with 
ZrB2 (99 wt% B10) 0.86659 0.00058 0.1086 -2.575 
NOTE: a .keff = loading curve material case keff minus new material keff 

0.0060 
0.0040 
0.0020 
.keff.keff 
0.0000
-0.0020
-0.0040
-0.0060
-0.0080
-0.0100
-0.0120
-0.0140
-0.0160
2.1 1.9 1.5 1 0.5 
Gd Loading (Wt%) 
Figure 13. Effects on keff as a function of Gd loading 
0.1400 
0.1200 
0.1000 
0.0800 
0.0600 
0.0400 
0.0200 
0.0000 
0 20 40 60 80 100 120 
B4C 
ZrB2 
B10 Loading (Wt%) 
Figure 14. Effects on keff as a Function of B10 Loading 
Below shows the relative impact on the 44-BWR waste package loading curve taken from Section 
6 from using the alternative neutron absorber materials. These results are only intended to illustrate 
the relative effects of the different absorber materials. The results are based on taking .keff using 
the alternative material in the preclosure configuration with a 5.0 wt% U-235 fresh fuel enrichment 
against the 1.5 wt% Gd, Ni-Gd Alloy with 5.0 wt% U-235 fresh fuel case used in Section 6.1. 

LicensingCalculationTitle: 44-BWR Waste Package Loading Curve EvaluationDocument Identifier: CAL-DSU-NU-000008 REV 00APage II-5 of 60102030405060700.01.02.03.04.05.06.0Initial Enrichment (Wt% U-235)
Burnup (GWd/MTU)
1-99100-199200-299300+
Ni-Gd (0.5)
Base LCNi-Gd (2.1)
A22ZrB2Figure 15. Loading Curve and Waste Stream Comparison

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 
44BWRWP.zip 1,465,133 8/19/2004 05:03p 44-BWR Waste Package Configuration Design 
Drawings 
0130223.pdf 68,126 6/10/2004 10:03a MSDS for representative hydraulic fluid 
cases.zip 12,241,204 8/19/2004 04:31p Archive containing MCNP files 
GEassembly_Vol.xls 39,936 7/29/2004 10:44a Excel spreadsheet containing tie plate material 
derivations 
MCNP_BWR_ 
Geometries.xls 
797,696 8/13/2004 10:31a 
Excel spreadsheet containing various 
geometry derivations 
schoepite.xls 48,128 8/18/2004 01:36p Excel spreadsheet containing schoepite 
material derivations 
Tuff composition.xls 49,152 8/10/2004 10:53a Excel spreadsheet containing tuff composition 
derivations 
wstreamplot.xls 295,936 8/04/2004 03:45p Excel spreadsheet containing sorted waste 
stream information 
There are 8 total files for the archive file 44BWRWP.zip with no particular naming system. The files 
contain the dimensions for the 44-BWR waste package configuration. 
There are 276 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: 
/Documents and Settings/scaglionej/My Documents/temp/*: 
where * corresponds as follows: 
/I.1/ -Contains files listed in Attachment I, Section I.1 
/I.2/ -Contains files listed in Attachment I, Section I.2 with subdirectories smeared and unsmeared 
corresponding to smeared and unsmeared pellet density cases, respectively. 
/I.3/ -Contains files listed in Attachment I, Section I.3 
/I.4/ -Contains files listed in Attachment I, Section I.4 with subdirectories Hyd_Fluid and 
Water_density corresponding to hydraulic fluid and water density cases, respectively. 
/I.5/ -Contains files listed in Attachment I, Section I.5 
/I.6/ -Contains files listed in Attachment I, Section I.6 with subdirectories Tuff External, Tuff 
Internal Config 1, and Tuff Internal Config 2 corresponding to the tabulated results in 
Attachment I, Tables 28, 29, and 30, respectively.
/I.7/ -Contains files listed in Attachment I, Section I.7 with subdirectories 3mm and Thickness 
corresponding to the tabulated results in Attachment I, Tables 31 and 32, respectively.

/I.8/ -Contains files listed in Attachment I, Section I.8 
/I.9/ -Contains files listed in Attachment I, Section I.9 
/Pre_LC/ - Contains files used for the preclosure loading curve configuration as a function of burnup 
where the naming system is as follows: XXYY where the XX represents the initial enrichment in wt% 
U-235 (i.e., 35 is 3.5 wt% U-235 [range from 3.5 to 5.0 wt% U-235]) and the YY represents the 
burnup in GWd/MTU (range from 10 to 40 GWd/MTU). A lower level directory denoted /Fresh/ 
contains the fresh fuel cases with a naming system as follows: X.X that represents the enrichment 
in wt% U-235 (ranging from 3.0 to 5.0). 
/Post_LC/ - Contains files used for the postclosure loading curve configuration as a function of 
burnup where the naming system is as follows: XXYY where the XX represents the initial enrichment 
in wt% U-235 (i.e., 35 is 3.5 wt% U-235 [range from 4.0 to 5.0 wt% U-235]) and the YY represents 
the burnup in GWd/MTU (range from 10 to 20 GWd/MTU). A lower level directory denoted 
/Fresh/ contains the fresh fuel cases with a naming system as follows: X.X that represents the 
enrichment in wt% U-235 (ranging from 3.0 to 5.0). 
/II/ - Contains the files from Attachment II with a naming system as follows: aXp7YYY where the 
X is either a 2 or a 3 denoting B4C or ZrB2 as the absorber, respectively; and the YYY is either n, e50, 
or e99 denoting the B10 loading as nominal, 50 wt%, or 99 wt%, respectively. Within this directory 
is a lower level directory denoted /Gd Enr/ which contains the files where the Gd loading in the Ni-
Gd Alloy was varied. The naming system corresponds as follows: X.X which represents the Gd 
loading weight percent (i.e., 0.5 = 0.5 wt% Gd).