Bias Estimates Used in Lieu of 

Validation of Fission Products 

and Minor Actinides in MCNP 

Keff Calculations for PWR 

Burnup Credit Casks

 

Office of Nuclear Material Safety and Safeguards

NUREG/CR-7205

ORNL/TM-2012/544
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Bias Estimates Used in Lieu of 

Validation of Fission Products 

and Minor Actinides in MCNP 

Keff Calculations for PWR

Burnup Credit Casks

Manuscript Completed: May 2014

Date Published: September 2015

Prepared by:

D. E. Mueller, W. J. Marshall, D. G. Bowen, and J. C. Wagner

Oak Ridge National Laboratory

Managed by UT-Battelle, LLC

Oak Ridge, TN 37831-6170

A. Barto, Project Manager

NRC Job Code J5728

Office of Nuclear Material Safety and Safeguards

NUREG/CR-7205

ORNL/TM-2012/544

iii

ABSTRACT

The U.S. Nuclear Regulatory Commission Division of Spent Fuel Management issued Interim 

Staff Guidance (ISG) 8, Revision 3 in September 2012. This ISG provides guidance for NRC 

Staff review of burnup credit analyses supporting transport and dry storage of pressurized water 

reactor spent nuclear fuel (SNF) in casks. ISG-8, Rev. 3 includes, among other things, guidance 

for addressing validation of criticality (keff) calculations crediting the presence of a limited set of 

fission products and minor actinides (FP&MA). Based on previous work documented in 

NUREG/CR-7109, ISG-8, Rev. 3 includes a recommendation to accept use of 1.5 or 3% of the 

FP&MA worth, depending on the criticality code and cross-section data used, to conservatively 

account for the bias and bias uncertainty associated with the selected insufficiently validated 

FP&MAs. This bias is applied in addition to the bias and bias uncertainty resulting from 

validation of keff calculations for the major actinides in SNF. The work described in this report

involves comparison of FP&MA worths calculated using SCALE and Monte Carlo N-Particle 

(MCNP) with Evaluated Nuclear Data Files, Part B–V, –VI, and –VII based nuclear data. The 

comparison supports use of the 1.5% FP&MA worth bias when either SCALE or MCNP codes 

are used for criticality calculations, provided the other conditions of ISG-8, Rev. 3,

Recommendation 4 are met. The additional conditions include that the cask design is similar to 

the hypothetical generic burnup credit-32 cask model and that the credited FP&MA worth is no 

more than 0.1 ∆keff.

v

TABLE OF CONTENTS

Section Page

ABSTRACT................................................................................................................................iii

LIST OF FIGURES ...................................................................................................................vii

LIST OF TABLES.......................................................................................................................ix

ACKNOWLEDGMENTS.............................................................................................................xi

ACRONYMS............................................................................................................................ xiii

1. INTRODUCTION................................................................................................................... 1

1.1 Calculation Method......................................................................................................... 3

2. REFERENCE MODEL DESCRIPTION ................................................................................. 5

2.1 GBC-32 Cask Model....................................................................................................... 5

2.2 PWR Fuel Assembly Specification.................................................................................11

2.3 Reference Model Fuel Compositions.............................................................................11

3. REFERENCE FP&MA WORTHS .........................................................................................15

4. MCNP RESULTS .................................................................................................................17

5. SUMMARY AND CONCLUSIONS........................................................................................19

6. REFERENCES.....................................................................................................................21

APPENDIX A. SCALE 6.1 CSAS5 MODEL ...........................................................................A-1

APPENDIX B. MCNP MODEL ...............................................................................................B-1

APPENDIX C. DETAILED FP&MA WORTH RESULTS FOR EACH CODE AND 

NUCLEAR DATA COMBINATION..................................................................C-1

vii

LIST OF FIGURES

Figure Page

Figure 2.1. Radial cross section of one quarter of the GBC-32 cask. ....................................... 8

Figure 2.2. Cross-sectional view of an assembly cell in KENO V.a model for the 

GBC-32 cask.......................................................................................................... 9

Figure 2.3. Isometric view of GBC-32 cask with top and front halves removed........................10

Figure 2.4. Burnup credit loading curves for the GBC-32 application models. .........................12

ix

LIST OF TABLES

Table Page

Table 1.1. Fission products and minor actinides....................................................................... 2

Table 2.1. Physical dimensions for the GBC-32 cask ............................................................... 6

Table 2.2. Material compositions for GBC-32 cask model ........................................................ 7

Table 2.3. PWR fuel assembly specifications..........................................................................11

Table 2.4. Burned fuel compositions for reference FP&MA worth calculations ........................13

Table 3.1. Reference FP&MA worth results using SCALE 6.1 and the ENDF/B–VII 

 238 group nuclear data library ...............................................................................16

Table 4.1. Variation in calculated FP&MA worths using various codes and data.....................17

Table C.1. FP&MA worths using SCALE 6.1 and ENDF/B–VII continuous-energy data........C-1

Table C.2. FP&MA worths using SCALE 6.1 and ENDF/B–VI 238 energy group data ..........C-2

Table C.3. FP&MA worths using SCALE 6.1 and ENDF/B–V 238 energy group data ...........C-3

Table C.4. FP&MA worths using SCALE 6.1 and ENDF/B–V 44 energy group data .............C-4

Table C.5. FP&MA worths using MCNP5 (version 1.60) and ENDF/B–VII data.....................C-5

Table C.6. FP&MA worths using MCNP5 (version 1.60) and ENDF/B–VI data......................C-6

Table C.7. FP&MA worths using MCNP5 (version 1.60) and ENDF/B–V data.......................C-7

Table C.8. FP&MA worths using MCNP6 (new release) and ENDF/B–V data.......................C-8

Table C.9. FP&MA worths using MCNP6 (new release) and ENDF/B–VI data......................C-9

Table C.10. FP&MA worths using MCNP6 (new release) and ENDF/B–VII data...................C-10

Table C.11. FP&MA worths using MCNP6 (new release) and ENDF/B–VII.1 data................C-11

xi

ACKNOWLEDGMENTS

This work was performed under contract with the Division of Spent Fuel Management, U.S. 

Nuclear Regulatory Commission. The authors thank A. Barto, the NRC Project Manager, and Z. 

Li and V. Wilson of the Office of Nuclear Material Safety and Safeguards for their support and 

guidance. The careful reviews of the draft manuscript by J. M. Scaglione and J. B. Clarity are

very much appreciated. Finally, the authors are thankful to D. J. Weaver for her preparation of 

the final report.

xiii

ACRONYMS

BUC burnup credit

CENTRM continuous energy transport module

CSAS Criticality Safety Analysis Sequence

ENDF/B Evaluated Nuclear Data Files, Part B

FP fission products

FP&MA fission products and minor actinides

GBC generic burnup credit cask; as in GBC-32, a generic burnup credit cask 

model with a 32 PWR fuel assembly capacity

GWd/MTU unit of nuclear fuel burnup; gigawatt-days per initial metric ton of uranium

ISG interim staff guidance

keff effective neutron multiplication factor

MA minor actinides

MCNP Monte Carlo N-Particle

NRC U.S. Nuclear Regulatory Commission

NUREG/CR Publications prepared by NRC Contractors

OFA optimized fuel assembly; as in Westinghouse 17 × 17 OFA

ORNL Oak Ridge National Laboratory

PWR pressurized water reactor

SNF spent nuclear fuel

1

1 INTRODUCTION

The concept of taking credit for the reduction in reactivity resulting from the net consumption of 

fissile nuclides and creation of neutron absorbing actinides and fission products during reactor 

operation is commonly referred to as “burnup credit” (BUC). The analysis supporting BUC 

requires two different types of computer calculations. First, the analyst uses a computer code to 

generate spent nuclear fuel (SNF) compositions based on conditions experienced by the fuel 

assemblies in the reactor. The spent fuel compositions are then typically used in threedimensional Monte Carlo method neutron transport calculations to determine the keff values for 

the system loaded with SNF. The combination of the computer codes, nuclear data,

computational input options used, and modeling approximations used is referred to as the 

“computational method.”

Computational method validation is used to quantify the relationship between calculated values 

and reality and is performed by comparing results, calculated using the computational method, 

to measured or expected results. For example, the computational method would be used to 

model a set of laboratory critical experiments, yielding calculated estimates for the neutron 

multiplication factor (keff). Appropriate statistical techniques are applied to the calculated and 

expected keff values to determine the bias and bias uncertainty associated with the 

computational method.

Use of the same computational method in both the safety analysis and the validation study 

enables the analyst to quantify the bias and bias uncertainty associated with the computational 

method. However, the accuracy of the calculated bias and bias uncertainty is also dependent 

on the degree of similarity between the safety analysis models and the critical experiments. If 

materials are present in either the safety analysis models or critical experiments and not in the 

other, any bias associated with these materials may not be captured correctly by the validation 

study. Bias errors also can be introduced when a material is present in both but has a different

effect on keff. The same nuclear data errors then affect the keff of the safety analysis models and 

critical experiments differently, resulting in different keff biases. Such differences may result from 

factors such as shifts in the neutron energy spectrum to higher or lower energies or because of

spatial variation of the neutron flux.

Burnup credit analyses for storage and transportation casks frequently seek to take credit for

some fission products (FP) and actinides for which little or no appropriate critical experiment 

data are available for use in validation studies. At the U.S. Nuclear Regulatory Commission’s 

(NRC’s) request, Oak Ridge National Laboratory (ORNL) staff prepared report NUREG/CR7109, An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety 

Analyses—Criticality (keff) Predictions (Ref. 1), which included recommendations for addressing 

validation of fission products and minor actinides (FP&MA) in keff calculations. Based in part on 

the uncertainty analysis work presented in NUREG/CR-7109, the NRC Division of Spent Fuel 

Management provided the following guidance in Recommendation 4 of Interim Staff Guidance 

(ISG) 8, Rev. 3 (Ref. 2) (some supplementary explanation is provided within brackets in the 

following text):

“Fission product and minor actinide credit

The applicant may credit the minor actinide and fission product nuclides listed in Table 2

[of ISG-8, Rev. 3; Table 2 lists the FP&MA shown in Table 1.1 in this report], provided 

the bias and bias uncertainty associated with the major actinides is determined as 
2

described above [ISG-8, Rev. 3 provides guidance for validation of major actinides]. One 

point five percent (1.5%) of the worth of the minor actinides and fission products 

conservatively covers the bias due to these isotopes. Due to the conservatism in this 

value, no additional uncertainty in the bias needs to be applied. This estimate is 

appropriate provided the applicant: 

• uses the SCALE code system with the ENDF/B–V [Evaluated Nuclear Data Files, 

Part B–V], ENDF/B–VI, or ENDF/B–VII cross section libraries,

• can justify that its design is similar to the hypothetical GBC-32 [generic burnup credit

cask–32] system design used as the basis for the NUREG/CR-7109 criticality 

validation, and 

• demonstrates that the credited minor actinide and fission product worth is no greater 

than 0.1 in keff. 

For well qualified, industry standard code systems other than SCALE with the 

ENDF/B-V, ENDF/B–VI, or ENDF/B–VII cross section libraries, a conservative estimate 

for the bias associated with minor actinide and fission product nuclides of 3.0% of their 

worth may be used. Use of a minor actinide and fission product bias less than 3.0% 

should be accompanied by additional justification that the lower value is an appropriate 

estimate of the bias associated with that code system.”

The FP&MA nuclides covered by the guidance are listed in Table 2 of ISG-8, Rev. 3, and below 

in Table 1.1.

Table 1.1. Fission products and minor actinides

Fission products

95Mo 99Tc 101Ru 103Rh

109Ag 133Cs 143Nd 145Nd

147Sm 149Sm 150Sm 151Sm

152Sm 151Eu 153Eu 155Gd

Minor actinides

236U 237Np 243Am

This report provides a complete description of a GBC-32 model loaded with pressurized water 

reactor (PWR) SNF. Single-axial zone burned fuel compositions are provided at fuel burnups of 

20 and 40 GWd/MTU for post-irradiation cooling times of 5 and 40 years. Reference FP&MA 

worths are provided for the four burnup and cooling time combinations for several of the nuclear 

data libraries available in SCALE 6.1 (Ref. 3). Comparisons of SCALE 6.1 and Monte Carlo 

N-Particle (MCNP) (Ref. 4) FP&MA worths are also provided.

Since the FP&MA worths calculated using MCNP are either lower than or statistically the same 

as the SCALE results (see Table 4.1), these comparisons demonstrate that it is appropriate to 

apply the 1.5% of the FP&MA worth bias to MCNP calculations using ENDF/B–V, –VI, –VII and 

–VII.1 data. This additional bias term supplements the MCNP keff calculation validation, thereby 

addressing the lack of keff validation for the FP&MAs.
3

1.1 CALCULATION METHOD

The calculation method described in this section demonstrates that MCNP used with its 

ENDF/B–V, –VI, –VII or –VII.1 based data sets produces FP&MA worth results that are similar 

to results generated using the SCALE 6.1 code system with ENDF/B–V, –VI and –VII based 

nuclear data libraries.

The calculation method used is the following:

1. SCALE Criticality Safety Analysis Sequence–5 (CSAS5) and MCNP criticality 

calculations are performed for the GBC-32 cask (Ref. 5) loaded with burned PWR fuel at 

20 and 40 GWd/MTU burnups and with post-irradiation cooling times of 5 and 40 years.

Complete descriptions of the cask and fuel assembly geometry and materials are 

provided in Section 2. Single axial zone burned fuel compositions are provided at burnup 

values of 20 and 40 GWd/MTU for 5- and 40-year post-irradiation cooling times. SCALE 

CSAS5 and MCNP cask models and code-specific (i.e., in code-specific input format)

burned fuel composition inputs are provided in Appendixes A and B of this report.

To simplify the calculations, single axial zone models are used for the determination of 

FP&MA worth. This is done because the goal is to show that given the same models and 

compositions, the MCNP FP&MA worths will be similar to the FP&MA worths calculated 

using SCALE. Comparisons are made at 20 and 40 GWd/MTU burnups following 5- and 

40-year post-irradiation cooling times to demonstrate that the comparison is valid over a 

range of safety analysis model parameters.

2. Results from criticality calculations performed with and without the FP&MAs are used to 

calculate the FP&MA worth for each burnup and cooling time combination. The FP&MA 

worths are calculated by subtracting the keff calculated with FP&MA removed from the keff

calculated with all FP&MAs present.

3. The calculated FP&MA worth results are then compared to FP&MA reference results

presented in Table 3.1 in Section 3. The reference results were calculated, consistent 

with the computational method used to generate NUREG/CR-7109, using SCALE 6.1 

and its ENDF/B–VII 238 energy group library. Detailed nuclide-by-nuclide keff worth 

results for other nuclear data libraries and for MCNP are provided in Appendix C.

Under prediction of FP&MA worth leads to the calculation of a smaller FP&MA bias term, but the 

reduced FP&MA worth has a significantly larger and more conservative impact on the maximum 

keff through the calculated keff value than does the reduction in the uncertainty associated with 

the 1.5% of FP&MA worth bias term. Consequently, variation of FP&MA worth below the 

reference FP&MA reference values is reasonable because the calculated FP&MA worths do not 

exceed the reference results FP&MA worths by more than 1.5%.

Subject to additional considerations prescribed by ISG-8, Rev. 3 and described earlier in this 

section, it is recommended that the 1.5% of the FP&MA worth uncertainty be used in MCNP 

based keff analyses to cover the poor validation of FP&MA nuclides in MCNP keff calculations.

5

2 REFERENCE MODEL DESCRIPTION

The following subsections describe the reference model used to generate the FP&MA worths 

presented in Sections 3 and 4 and in Appendix C. The cask and SNF model used is the same 

as is described in NUREG/CR-6747 (Ref. 5). For the reader’s convenience, relevant model 

information is provided in Sections 2.1 and 2.2. The burned PWR fuel compositions used in the 

reference model are described in Section 2.3. Input listings for the SCALE 6.1 CSAS5 and 

MCNP models of the GBC-32 cask are provided in Appendixes A and B, respectively.

2.1 GBC-32 CASK MODEL

As is discussed in NUREG/CR-6747, the GBC-32 cask model was developed as a 

computational benchmark that could be used for the estimation of the additional reactivity 

associated with FP&MA. The model was intended to capture the important features of transport 

and storage casks without use of non-essential details or proprietary information associated with 

actual cask designs. The cask holds 32 PWR assemblies in storage cells with 0.75-cm-thick 

steel walls as well as one BoralTM plate between each pair of assemblies and on the external 

faces of the basket.

The water-filled cask dimensions and materials are described in Tables Table 2.1 and Table

2.2. Figures Figure 2.1, Figure 2.2, and Figure 2.3 show features of the GBC-32 cask model.

Tables Table 2.1 and Table 2.2, and Figures 2.1 and Figure 2.2 were reproduced from 

NUREG/CR-6747.

In addition to the information provided in Tables 2.1 and Table 2.2, the BoralTM panel present on 

the outside of each external fuel assembly storage cell is covered by a 0.75-cm-thick type 304 

stainless steel plate. The width of each external vertical steel plate shown in Figure 2.1 matches 

the storage cell inside dimension of 22.0 cm. The width of each external horizontal steel plate 

matches the storage cell outside dimension of 23.5 cm.
6

Table 2.1. Physical dimensions for the GBC-32 cask

Parameter Inches Centimeters

Cell inside dimension 8.6614 22.0000

Cell outside dimension 9.2520 23.5000

Cell wall thickness 0.2953 0.7500

BoralTM panel thicknessa 0.1010 0.2565

BoralTM center thickness 0.0810 0.2057

BoralTM Al plate thickness 0.0100 0.0254

Cell pitch 9.3530 23.7565

Boral panel width 7.5000 19.0500

Cell heightb 144.0000 365.76

BoralTM panel heightb 144.0000 365.76

Cask inside diameter 68.8976 175.0000

Cask outside diameter 84.6457 215.0000

Cask radial thickness 7.8740 20.0000

Base plate thickness 11.8110 30.0000

Cask lid thickness 11.8110 30.0000

Cask inside height 161.7165 410.7600

Active fuel heightb 144.0000 365.76

Bottom assembly hardware thickness 5.9055 15.0000

Top assembly hardware thickness 11.8110 30.0000

a BoralTM is a clad composite of aluminum and boron carbide. A BoralTM panel or 

plate consists of three distinct layers. The outer layers are aluminum cladding, which 

form a sandwich with a central layer that consists of a uniform aggregate of fine 

boron carbide particles within an aluminum alloy matrix. b The cell height, BoralTM panel height, and active fuel height are all equivalent, and 

their lower boundaries are coincident, 15 cm above the base plate.
7

Table 2.2. Material compositions for GBC-32 cask model

Isotope Atom density

(atoms/barn-cm) Weight percent

Water (density = 0.9983 g/cm3)

Hydrogen (H) 0.06674 11.19

Oxygen (O) 0.03337 88.81

Total 0.10011 100.0

Stainless steel 304 (density = 7.92 g/cm3) (Ref. 6)

Chromium (Cr) 0.01743 19.0

Manganese (Mn) 0.00174 2.0

Iron (Fe) 0.05936 69.5

Nickel (Ni) 0.00772 9.5

Total 0.08625 100.0

BoralTM panel aluminum cladding (density = 2.699 g/cm3)

Aluminum (Al) 0.0602 (Ref. 7) 100.0

Total 0.0602 100.0

BoralTM panel central layer (0.0225 g 10B/cm2)

a

Boron-10 (10B) 6.5794E-03 4.13

Boron-11 (11B) 2.7260E-02 18.81

Carbon (C) 8.4547E-03 6.37

Aluminum (Al) 4.1795E-02 70.69

Total 8.4089E-02 100.0 a Note: 0.030 g 10B/cm2 is the loading from the manufacturer (AAR8

) that corresponds to the 

modeled Boral panel thickness of 0.101 inches. However, current NRC guidance9

recommends only 75% credit for fixed neutron absorbers, and thus 75% of 0.030, or 0.0225 g 10B/cm2 is used.
8

Figure 2.1. Radial cross section of one-quarter of the GBC-32 cask
9

Figure 2.2. Cross-sectional view of an assembly cell in KENO V.a model for 

the GBC-32 cask
10

Figure 2.3. Isometric view of GBC-32 cask with top and front halves removed
11

2.2 PWR FUEL ASSEMBLY SPECIFICATION

The reference PWR fuel assembly design used in NUREG/CR-6747 and in NUREG/CR-7109 is 

representative of a Westinghouse 17 × 17 Optimized Fuel Assembly (OFA). To be consistent 

with prior work, the same assembly design is used in this report for calculation of the reference 

FP&MA worths. The simplified model used here and in NUREG/CR-6747 replaces the assembly 

grids and assembly structures above and below the active fuel length with water. Thus the 

simplified fuel assembly model is a 365.76-cm-long 17 × 17 array of fuel pins and guide tubes. 

Table 2.3, reproduced here from Table 3 of NUREG/CR-6747, provides the fuel assembly 

specifications.

The spent fuel pellet compositions will be addressed in the next section. The fuel rod clad, guide 

tubes, and instrument tube are all modeled as 100% zirconium, having an atom density of 

0.0423 atoms per barn-cm (Ref. Error! Bookmark not defined.).

Table 2.3. PWR fuel assembly specifications

Parameter Inches Centimeters

Fuel outside diameter 0.3088 0.7844

Cladding inside diameter 0.3150 0.8001

Cladding outside diameter 0.3600 0.9144

Cladding radial thickness 0.0225 0.0572

Rod pitch 0.4960 1.2598

Guide tube/thimble inside diameter 0.4420 1.1227

Guide tube/thimble outside diameter 0.4740 1.2040

Thimble radial thickness 0.0160 0.0406

Instrument tube inside diameter 0.4420 1.1227

Instrument tube outside diameter 0.4740 1.2040

Instrument tube radial thickness 0.0160 0.0406

Active fuel length 144 365.76

Array size 17 × 17

Number of fuel rods 264

Number of guide tubes/thimbles 24

Number of instrument tubes 1

2.3 REFERENCE MODEL FUEL COMPOSITIONS

For the purposes of the reference FP&MA worth calculations, fuel compositions were calculated 

for two initial enrichment/final burnup pairs along a representative loading curve determined for 

the analysis documented in NUREG/CR-7109 (Ref. 1). This loading curve is the “Actinides & 16

FP” loading curve shown below in Figure 2.4, which was reproduced from Figure 4.4 of 

NUREG/CR-7109. The models used to generate the loading curves shown in the figure included 
12

an 18-axial zone burnup distribution and a 5-year post-irradiation cooling time. The background 

in this figure shows all PWR fuel inventory as of the end of 2002 according to the RW-859

(2002) data (Ref. 10).

Figure 2.4. Burnup credit loading curves for the GBC-32 application models

In particular, the fuel compositions used in the reference FP&MA worth calculations are for fuel 

with initial enrichments of 3.14 and 4.68 wt % 235U burned to 20 and 40 GWd/MTU, respectively.

Fuel compositions were generated for each combination with both a 5-year and 40-year postirradiation cooling time. Table 2.4 provides the burned fuel compositions used in the calculation 

of the reference FP&MA worth values. These irradiated fuel compositions were generated using 

the same computational methods described in NUREG/CR-7109 (Ref. 1).

0

10

20

30

40

50

60

70

1.5 2 2.5 3 3.5 4 4.5 5

Final Assembly Average Burnup (GWd/MTU)

Initial Enrichment (wt % 235U)

Actinides-Only

Actinides & 16 FP

Actinides & all FP

Acceptable

Not Acceptable
13

Table 2.4. Burned fuel compositions for reference FP&MA worth calculations

Initial 

enrichment 

and

final burnup

3.1 wt % and 20 GWd/MTU 4.7 wt % and 40 GWd/MTU

Cooling time 5 years 40 years 5 years 40 years

Nuclide Atom densities

(atoms/barn-cm)

Major 

actinides

234U 7.3901E-08 3.2108E-07 2.6245E-07 1.3559E-06

235U 3.6046E-04 3.6060E-04 3.6741E-04 3.6758E-04

238U 2.2345E-02 2.2345E-02 2.1683E-02 2.1683E-02

238Pu 1.0223E-06 7.7547E-07 4.5238E-06 3.4316E-06

239Pu 1.3719E-04 1.3705E-04 1.7413E-04 1.7396E-04

240Pu 3.9569E-05 3.9477E-05 6.3031E-05 6.3301E-05

241Pu 1.6819E-05 3.0797E-06 3.2475E-05 5.9464E-06

242Pu 3.8879E-06 3.8877E-06 1.1666E-05 1.1665E-05

241Am 4.8440E-06 1.7840E-05 9.7110E-06 3.4784E-05

Minor 

actinides

236U 7.0106E-05 7.0252E-05 1.3508E-04 1.3532E-04

237Np 5.5048E-06 6.2485E-06 1.3907E-05 1.5362E-05

243Am 5.1541E-07 5.1371E-07 2.5940E-06 2.5854E-06

Fission 

products

95Mo 2.8435E-05 2.8435E-05 5.4717E-05 5.4717E-05

99Tc 2.9026E-05 2.9023E-05 5.5148E-05 5.5142E-05

101Ru 2.6347E-05 2.6347E-05 5.1618E-05 5.1618E-05

103Rh 1.8264E-05 1.8264E-05 3.1900E-05 3.1900E-05

109Ag 2.1375E-06 2.1375E-06 4.3961E-06 4.3961E-06

133Cs 3.0552E-05 3.0552E-05 5.7163E-05 5.7163E-05

143Nd 2.2819E-05 2.2819E-05 4.1309E-05 4.1309E-05

145Nd 1.6695E-05 1.6695E-05 3.1395E-05 3.1395E-05

147Sm 5.6742E-06 7.4723E-06 9.3467E-06 1.1950E-05

149Sm 1.7982E-07 1.7982E-07 2.2875E-07 2.2875E-07

150Sm 6.4540E-06 6.4540E-06 1.3517E-05 1.3517E-05

151Sm 4.1578E-07 3.1754E-07 6.4777E-07 4.9471E-07

152Sm 2.6638E-06 2.6639E-06 4.2868E-06 4.2871E-06

151Eu 1.6613E-08 1.1485E-07 2.5993E-08 1.7905E-07

153Eu 2.2518E-06 2.2518E-06 4.9812E-06 4.9812E-06

155Gd 7.0661E-08 1.3548E-07 1.7686E-07 3.3841E-07

16O 4.6943E-02 4.6943E-02 4.6949E-02 4.6949E-02

Total 7.0120E-02 7.0122E-02 6.9808E-02 6.9811E-02
14

As has been shown in prior work (Ref. 11), accurate keff determination for PWR spent fuel 

systems requires that the axial burnup distribution be modeled. The calculations performed for 

this report used a single axially uniform fuel composition for each burnup and post-irradiation 

cooling time. This was done to simplify the calculations and is appropriate because the objective 

of the calculations is limited to showing that the other computational methods (e.g., MCNP with 

ENDF/B–V, –VI, –VII, and –VII.1 nuclear data) generate FP&MA worth results that are similar to 

the reference results. Calculations and comparisons were made at two burnup/enrichment 

points and for two post-irradiation cooling times to demonstrate that the comparisons made and 

conclusions reached are valid over the range of burned fuel compositions typically observed in 

burnup credit criticality analyses for transportation and dry storage operations.
15

3 REFERENCE FP&MA WORTHS

The nuclear data uncertainty analysis results, which were presented in NUREG/CR-7109 and 

served as the basis for ISG-8, Rev. 3, Recommendation 4, were generated using the SCALE 

6.1 CSAS5 sequence and the ENDF/B–VII-based 238 neutron energy group library (v7-238). In 

that work, the CSAS5 sequence used the CENTRM module to perform resolved resonance 

calculations and performed KENO V.a Monte Carlo method neutron transport calculations.

The same computational method was used to generate the reference FP&MA worth results, 

which are presented in Table 3.1. The models described in Section 2, including the burned fuel 

composition information provided in Table 2.4, were used to calculate the reference FP&MA 

worths. Nominal calculations were performed with all of the nuclides present. Individual 

calculations were also performed with each of the FP and MA removed, with all FP&MA

removed, with all MA removed, and with all FP removed. The group or individual nuclide worth 

values were then calculated as the change in keff because of the presence of the nuclide or 

group of nuclides using the following equation:

worthnuclide or group = knominal – knuclide or group removed

The results presented in the “All FP&MA,” “All MA,” and “All FP” rows of Table 3.1 were 

calculated with the group constituents simultaneously removed. The reference FP&MA worth 

results to be used for comparison with results from other computational methods are the worths 

listed in the “All FP&MA” row. The other worth values are provided for reference and to facilitate 

more detailed comparisons.
16

Table 3.1. Reference FP&MA worth results using SCALE 6.1 and

 the ENDF/B–VII 238 group nuclear data library

Initial enrichment and

final burnup

3.1 wt % and 20 GWd/MTU 4.7 wt % and 40 GWd/MTU

Cooling time

(years) 5 40 5 40

Reactivity worth values

(∆k)

a

All FP&MA

(reference worth) -0.06742 -0.06921 -0.10203 -0.10442

All MA -0.00602 -0.00588 -0.01154 -0.01093

All FP -0.06058 -0.06206 -0.08820 -0.09085

236U -0.00373 -0.00359 -0.00579 -0.00526

237Np -0.00193 -0.00210 -0.00488 -0.00459

243Am -0.00022 -0.00020 -0.00128 -0.00108

95Mo -0.00123 -0.00078 -0.00207 -0.00158

99Tc -0.00278 -0.00254 -0.00439 -0.00370

101Ru -0.00088 -0.00072 -0.00165 -0.00155

103Rh -0.00681 -0.00616 -0.01023 -0.00934

109Ag -0.00105 -0.00058 -0.00180 -0.00154

133Cs -0.00324 -0.00309 -0.00551 -0.00490

143Nd -0.00895 -0.00864 -0.01348 -0.01269

145Nd -0.00187 -0.00163 -0.00332 -0.00279

147Sm -0.00139 -0.00157 -0.00223 -0.00241

149Sm -0.01570 -0.01502 -0.01646 -0.01541

150Sm -0.00095 -0.00092 -0.00165 -0.00159

151Sm -0.00588 -0.00440 -0.00760 -0.00562

152Sm -0.00233 -0.00199 -0.00323 -0.00286

151Eu -0.00027 -0.00107 -0.00043 -0.00157

153Eu -0.00150 -0.00133 -0.00309 -0.00260

155Gd -0.00359 -0.00659 -0.00738 -0.01330

a Monte Carlo uncertainty (1 σ) is 0.00014 ∆k and is not included in the FP&MA worths.
17

4 MCNP RESULTS

FP&MA reactivity worth calculations were performed using MCNP5 version 1.60 and MCNP6

with available continuous-energy nuclear data libraries including ENDF/B–V, –VI, –VII

and -VII.1. FP&MA worth results for these data sets and codes are provided in Table 4.1.

Table 4.1. Variation in calculated FP&MA worths using various codes and data

Initial enrichment and final burnup 3.1 wt % and 20 

GWd/MTU

4.7 wt % and 40 

GWd/MTU

Cooling time (years) 5 40 5 40

Code Nuclear

dataa FP&MA worths (Δ)

b and % change from reference

SCALE 6.1 (CENTRM) 

(reference worth) v7-238 -0.06742 -0.06921 -0.10203 -0.10442

MCNP5 (1.60) ENDF/B-VII -0.06748 -0.06969 -0.10213 -0.10456

(% change)c +0.1 +0.7 +0.1 +0.1

MCNP6 ENDF/B-VII -0.06762 -0.06961 -0.10222 -0.10450

(% change)c +0.3 +0.6 +0.2 +0.1

MCNP6 ENDF/B-VII.1 -0.06663 -0.0684 -0.10099 -0.10268

(% change)c -1.2 -1.2 -1.0 -1.7

MCNP5 (1.60) ENDF/B-V -0.06688 -0.06913 -0.10154 -0.10402

(% change)c -0.8 -0.1 -0.5 -0.4

MCNP6 ENDF/B-V -0.06705 -0.06906 -0.10136 -0.10395

(% change)c -0.6 -0.2 -0.7 -0.5

SCALE 6.1 (CENTRM) 

(reference worth)d v6-238 -0.04518 -0.04923 -0.06939 -0.07535

MCNP5 (1.60)d ENDF/B-VI -0.04529 -0.04918 -0.0691 -0.07487

(% change)c +0.2 -0.1 -0.4 -0.6

MCNP6d ENDF/B-VI -0.04525 -0.04908 -0.0692 -0.07511

(% change)c +0.2 -0.3 -0.3 -0.3

a The SCALE nuclear data libraries are described in Section M4 of Ref. Error! Bookmark not defined..

The MCNP nuclear data libraries are described in Ref. 4. b The Monte Carlo one-standard deviation uncertainty associated with all reported FP&MA worths is no

greater than 0.00015 ∆k. This uncertainty is not included in the FP&MA worths. c Percent change from reference worth.

d Compositions excluded 95Mo, 101Ru, 143Nd, 145Nd, 150Sm, 151Sm, and 152Sm, for which ENDF/B–VI data

were not available in MCNP5 (v1.60) and MCNP6.

The 1.5% of the FP&MA worth bias term specified in Recommendation 4 of ISG-8, Rev. 3 was 

based in general on the work reported in NUREG/CR-7109 (Ref. 1) and more specifically on the 

work reported in Section 7.4.3 of Ref. 1. In that work, the uncertainty information associated with 

the nuclear data was combined with the nuclide-, reaction-, and energy-dependent sensitivity of 

keff to nuclear data variation for representative spent fuel pool and burnup credit cask models to

generate the uncertainty in keff due to the uncertainty in nuclear data. The uncertainties in keff

due to FP&MA were then compared to their worths, yielding an estimate that 1.5% of the 
18

FP&MA worth would yield a bounding estimate for the biases associated with the FP&MAs in 

the GBC-32 model. Information concerning the analysis technique and supporting the 

determination of the 1.5% value is presented in NUREG/CR-7109.

Without exception, the results obtained using MCNP vary from the reference values by less than 

1.7% of the FP&MA worth value. This is expected because the variation between the results is 

solely due to nuclear data variation or differences in the implementation of the Monte Carlo 

calculation in SCALE CSAS5 versus MCNP.

Under prediction of FP&MA worth leads to the calculation of a smaller FP&MA bias term, but the 

reduced FP&MA worth has a significantly larger and more conservative impact on the maximum 

keff through the calculated keff value than does the reduction in the 1.5% of FP&MA worth bias

term. Consequently, variation from the reference values is tolerable, provided the calculated 

FP&MA worths do not exceed the reference results by more than 1.5% of the FP&MA worths.

These results support the use the 1.5% of FP&MA worth bias term described in ISG-8, Rev. 3,

Recommendation 4, when MCNP and the nuclear data sets described in Table 4.1 are used in 

safety analyses. It may be possible to use the same method and comparisons to justify 

application of the 1.5% or 3.0% of FP&MA worth biases to results generated using other codes 

and/or nuclear data.

Some of the reference worths provided in Table 4.1 appear to exceed the restriction provided in 

Recommendation 4 of ISG-8, Rev. 3, that the credited minor actinide and fission product worth 

not exceed 0.1 ∆keff. Some of the FP&MA worths reported in this work are higher than those 

reported in NUREG/CR-7109 because of the use of the single-axial zone model, which amplifies 

the importance of the higher levels of FP&MAs in the center of the fuel. Further, when the 

number of significant figures is considered, the 0.10442 ∆keff value does not exceed 0.1. Thus, if 

the credited FP&MA worth is slightly higher than 0.1 ∆keff, the credited worth should be

considered to be no greater than 0.1 ∆keff, meeting the ISG-8, Rev. 3, criterion.

Detailed FP&MA worth results for each code and nuclear data combination are provided in 

Appendix C.

Note that all MCNP calculations were performed with the MCNP5, v1.60, and MCNP6 codes 

and nuclear data distributed by the MCNP developers. Since the neutron absorption reaction in 

FP&MA is the only significant interaction of neutrons with FP&MA nuclides and the simulation of 

neutron capture is straightforward, the FP&MA worth results are not affected by which MCNP 

version is used. Review of the data in Table 4.1 reveals that FP&MA worths calculated using 

MCNP5 and MCNP6 using the same nuclear data set are statistically the same (i.e. vary by less 

than one or two standard deviations). Consequently, the conclusion concerning the application 

of the SCALE FP&MA uncertainty information to MCNP calculation keff values is not MCNP 

version specific. However, the conclusions may not be applicable to special MCNP versions 

and/or data that were not generated by the MCNP developers.
19

5 SUMMARY AND CONCLUSIONS

The criticality safety of SNF in transportation or storage systems relies on the accurate 

calculation of the keff values. Validation studies are used to establish the relationship between

the actual and calculated keff values. Unfortunately, insufficient critical experiment data is 

available to support use of the conventional validation approach for fission products and minor 

actinides in BUC criticality safety evaluations for PWR SNF casks. Work documented in 

NUREG/CR-7109 (Ref. 1) supports the use of nuclear data uncertainty to provide a bounding 

estimate of the potential bias associated with taking credit for FP&MA in burnup credit criticality 

analyses. Based on that work, NRC Division of Spent Fuel Management Interim Staff Guidance 

8, Rev. 3, Recommendation 4, provides guidance for adoption of a bias term equal to 1.5% or 

3.0% of the FP&MA worth, depending on the code and nuclear data used. 

The work documented in this report provides justification for use of 1.5% of the FP&MA worth as 

a bias in BUC criticality safety evaluations using MCNP with the ENDF/B–V, –VI, –VII or –VII.1

nuclear data distributed with those code systems by the MCNP development team. For other 

code systems or nuclear data sets, it may be possible to use the same method to confirm that 

their computational method yields FP&MA worths similar to those calculated using the SCALE 

6.1 CSAS5 sequence and the ENDF/B–VII 238 neutron energy group nuclear data library.

Section 2 of this report provides complete descriptions of the reference cask model, including 

dimensions, materials, and SNF compositions. These models were used to calculate FP&MA 

worths using MCNP and its ENDF/B–V, –VI, –VII and –VII.1 data. These FP&MA worths are 

compared with the reference SCALE results to show that they are similar.

The results generated using MCNP with multiple sets of nuclear data are presented in 

Section 4. All nuclear data sets examined yielded FP&MA worths that were within 1.7% of the 

reference FP&MA worths. The MCNP results were no greater than 0.7% larger than the 

reference FP&MA worths for any of the burnup, decay time, and nuclear data libraries 

considered. Consequently, use of the 1.5 % of FP&MA worth bias to account for poor validation 

of FP&MAs in criticality calculations performed using MCNP5 or MCNP6 with ENDF/B–V, –VI, 

–VII or –VII.1 data is recommended.

21

6 REFERENCES

1. J. M. Scaglione, et al., An Approach for Validating Actinide and Fission Product Burnup 

Credit Criticality Analyses – Criticality (keff) Predictions, NUREG/CR-7109 (ORNL/TM2011/514), prepared for the U.S. Nuclear Regulatory Commission by Oak Ridge National 

Laboratory, Oak Ridge, Tenn., April 2012.

2. Division of Spent Fuel Storage and Transportation, Interim Staff Guidance-8, Rev. 3, 

Burnup Credit in the Criticality Analyses of PWR Spent Fuel in Transportation and Storage 

Casks, U.S. Nuclear Regulatory Commission, September 27, 2012.

3. Scale: A Comprehensive Modeling and Simulation Suite for Nuclear Safety Analysis and 

Design, ORNL/TM-2005/39, Version 6.1, Oak Ridge National Laboratory, Oak Ridge, 

Tenn. June 2011. Available from Radiation Safety Information Computational Center at 

Oak Ridge National Laboratory as CCC-785.

4. MCNP – A General Monte Carlo N-Particle Transport Code, Version 5, LA-UR-03-1987, 

Los Alamos National Laboratory, Los Alamos, N.M., February 1, 2008. Available from 

Radiation Safety Information and Computational Center at Oak Ridge National Laboratory 

as CCC-740.

5. J. C. Wagner, Computational Benchmark for Estimation of Reactivity Margin from Fission 

Products and Minor Actinides in PWR Burnup Credit, NUREG/CR-6747 (ORNL/TM2000/306), prepared for the U.S. Nuclear Regulatory Commission by Oak Ridge National 

Laboratory, Oak Ridge, Tenn., October 2001.

6. C. D. Harmon II, et al., Criticality Calculations with MCNP: A Primer, LA-12827-M, Los 

Alamos National Laboratory, Los Alamos, N.M., 1994.

7. J. J. Duderstadt and L. J. Hamilton, Nuclear Reactor Analysis, John Wiley & Sons, 1976.

8. Boral–The Proven Neutron Absorber, AAR Advanced Structures, Livonia, Michigan.

9. Standard Review Plan for Spent Fuel Dry Storage Systems at a General License Facility, 

NUREG-1536, Revision 1, U.S. Nuclear Regulatory Commission, Washington, D.C., July 

2010.

10. RW-859 Nuclear Fuel Data, Energy Information Administration, Washington, D.C., 

October 2004.

11. J. C. Wagner, M. D. DeHart, and C. V. Parks, Recommendations for Addressing Axial 

Burnup in PWR Burnup Credit Analyses, NUREG/CR-6801 (ORNL/TM-2001/273) 

prepared for the U.S. Nuclear Regulatory Commission by Oak Ridge National Laboratory, 

Oak Ridge, Tenn., March 2003.

A-1

APPENDIX A. SCALE 6.1 CSAS5 MODEL

This appendix includes a SCALE 6.1 CSAS5 model and spent nuclear fuel composition 

information in the SCALE input format for fuel burned to 20 GWd/MTU, following a 5-

year post-irradiation cooling time. Composition inputs for other burnup/cooling time 

combinations are provided following this CSAS5 input file.

=csas5 

gbc-32 FP&MA reference worth model

v7-238 

read comp

' 3.14 wt%, 20 gwd/mtu, 5 year cooling time

 u-234 101 0 7.3901E-08 293.0 end

 u-235 101 0 3.6046E-04 293.0 end

 u-236 101 0 7.0106E-05 293.0 end

 u-238 101 0 2.2345E-02 293.0 end

np-237 101 0 5.5048E-06 293.0 end

pu-238 101 0 1.0223E-06 293.0 end

pu-239 101 0 1.3719E-04 293.0 end

pu-240 101 0 3.9569E-05 293.0 end

pu-241 101 0 1.6819E-05 293.0 end

pu-242 101 0 3.8879E-06 293.0 end

am-241 101 0 4.8440E-06 293.0 end

am-243 101 0 5.1541E-07 293.0 end

mo-95 101 0 2.8435E-05 293.0 end

tc-99 101 0 2.9026E-05 293.0 end

ru-101 101 0 2.6347E-05 293.0 end

rh-103 101 0 1.8264E-05 293.0 end

ag-109 101 0 2.1375E-06 293.0 end

cs-133 101 0 3.0552E-05 293.0 end

nd-143 101 0 2.2819E-05 293.0 end

nd-145 101 0 1.6695E-05 293.0 end

sm-147 101 0 5.6742E-06 293.0 end

sm-149 101 0 1.7982E-07 293.0 end

sm-150 101 0 6.4540E-06 293.0 end

sm-151 101 0 4.1578E-07 293.0 end

eu-151 101 0 1.6613E-08 293.0 end

sm-152 101 0 2.6638E-06 293.0 end

eu-153 101 0 2.2518E-06 293.0 end

gd-155 101 0 7.0661E-08 293.0 end

 o-16 101 0 4.6943E-02 293.0 end

zr 2 0 0.0423 293.0 end

o-16 3 0 0.03337 293.0 end

h-1 3 0 0.06674 293.0 end

o-16 4 0 0.03337 293.0 end

h-1 4 0 0.06674 293.0 end

cr 5 0 0.01743 293.0 end

mn 5 0 0.00174 293.0 end

fe 5 0 0.05936 293.0 end

ni 5 0 0.00772 293.0 end

b-10 6 0 6.5794E-03 293.0 end

b-11 6 0 2.7260E-02 293.0 end

c 6 0 8.4547E-03 293.0 end
A-2

al-27 6 0 4.1795E-02 293.0 end

cr 7 0 0.01743 293.0 end

mn 7 0 0.00174 293.0 end

fe 7 0 0.05936 293.0 end

ni 7 0 0.00772 293.0 end

al 8 0 0.0602 293.0 end

o-16 9 0 0.03337 293.0 end

h-1 9 0 0.06674 293.0 end

o-16 10 0 0.03337 293.0 end

h-1 10 0 0.06674 293.0 end

zr 11 0 0.0423 293.0 end

end comp

read celldata

latticecell squarepitch 

 fueld= 0.7844 101 

 gapd= 0.8001 4 

 cladd= 0.9144 2

 pitch= 1.2598 3 end

end celldata

read param

res=1000000 tme=10000 gen=10100 nsk=100 npg=10000 sig=0.0001 

htm=no

end param

read geom

unit 1

 cylinder 101 1 0.3922 365.76 0

 cylinder 4 1 0.40005 365.76 0

 cylinder 2 1 0.4572 365.76 0

 cuboid 3 1 4p0.6299 365.76 0

'

' Guide Thimble/Instrument Tube (assumed to be same)

unit 2

 cylinder 10 1 0.56135 365.76 0

 cylinder 11 1 0.6020 365.76 0

 cuboid 10 1 4p0.6299 365.76 0

'

' Top Half Horizontal Boral Panel

unit 40

 cuboid 8 1 9.525 -9.525 0.0254 0 365.76 0

 cuboid 6 1 9.525 -9.525 0.12827 0 365.76 0

'

' Right-Hand Side Half Vertical Boral Panel

unit 50

 cuboid 8 1 0.0254 0 9.525 -9.525 365.76 0

 cuboid 6 1 0.12827 0 9.525 -9.525 365.76 0

'

' Bottom Half Horizontal Boral Panel

unit 60

 cuboid 8 1 9.525 -9.525 0 -0.0254 365.76 0

 cuboid 6 1 9.525 -9.525 0 -0.12827 365.76 0

'

' Left-Hand Side Half Vertical Boral Panel

unit 70
A-3

 cuboid 8 1 0 -0.0254 9.525 -9.525 365.76 0

 cuboid 6 1 0 -0.12827 9.525 -9.525 365.76 0

'

' Assembly Basket Cell

unit 101

 array 1 -10.7083 -10.7083 0

 cuboid 9 1 4p11 365.76 0

 cuboid 5 1 4p11.75 365.76 0

 cuboid 9 1 4p11.87827 365.76 0

 hole 40 0 11.75 0

 hole 50 11.75 0 0

 hole 60 0 -11.75 0

 hole 70 -11.75 0 0

'

' Top Boral/Basket Plate

unit 110

 cuboid 6 1 9.525 -9.525 0.10287 0 365.76 0

 cuboid 8 1 9.525 -9.525 0.12827 0 365.76 0

 cuboid 9 1 11.75 -11.75 0.12827 0 365.76 0

 cuboid 5 1 11.75 -11.75 0.87827 0 365.76 0

'

' Left-Hand Side Boral/Basket Plate

unit 112

 cuboid 6 1 0 -0.10287 9.525 -9.525 365.76 0

 cuboid 8 1 0 -0.12827 9.525 -9.525 365.76 0

 cuboid 9 1 0 -0.12827 10.9999 -10.9999 365.76 0

 cuboid 5 1 0 -0.87827 10.9999 -10.9999 365.76 0

'

' Right-Hand Side Boral/Basket Plate

unit 113

 cuboid 6 1 0.10287 0 9.525 -9.525 365.76 0

 cuboid 8 1 0.12827 0 9.525 -9.525 365.76 0

 cuboid 9 1 0.12827 0 10.9999 -10.9999 365.76 0

 cuboid 5 1 0.87827 0 10.9999 -10.9999 365.76 0

unit 114

 array 3 -47.51308 0 0

'

' Cask Inner Volume

global unit 200

 array 2 -71.26962 0 0

 zhemicyl+y 9 1 87.5 395.76 -15

 hole 114 0 47.51309 0

'

' Exterior Half Boral Panels

' Top Plates

 hole 110 -35.63481 71.26964 0

 hole 110 -11.87827 71.26964 0

 hole 110 11.87827 71.26964 0

 hole 110 35.63481 71.26964 0

 hole 110 59.39135 47.51309 0

 hole 110 -59.39135 47.51309 0

' Left-Hand Side Plates

 hole 112 -47.51310 59.39135 0

 hole 112 -71.26964 35.63481 0

 hole 112 -71.26964 11.87827 0

' Right-Hand Side Plates

 hole 113 47.51310 59.39135 0
A-4

 hole 113 71.26964 35.63481 0

 hole 113 71.26964 11.87827 0

' Steel Cask/Overpack

 zhemicyl+y 7 1 107.5 425.76 -45

'

' Cuboid Surrounding Cask

 cuboid 0 1 108 -108 108 0 425.76 -45

end geom

'

'

' Assembly Type: Westinghouse 17x17 OFA/V5

read array

 ara= 1 nux= 17 nuy= 17 nuz= 1

 fill

 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1

 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1

 1 1 1 1 1 2 1 1 2 1 1 2 1 1 1 1 1

 1 1 1 2 1 1 1 1 1 1 1 1 1 2 1 1 1

 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1

 1 1 2 1 1 2 1 1 2 1 1 2 1 1 2 1 1

 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1

 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1

 1 1 2 1 1 2 1 1 2 1 1 2 1 1 2 1 1

 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1

 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1

 1 1 2 1 1 2 1 1 2 1 1 2 1 1 2 1 1

 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1

 1 1 1 2 1 1 1 1 1 1 1 1 1 2 1 1 1

 1 1 1 1 1 2 1 1 2 1 1 2 1 1 1 1 1

 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1

 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1

 end fill

 ara=2 nux=6 nuy=2 nuz=1 fill 12r101 end fill

 ara=3 nux=4 nuy=1 nuz=1 fill 4r101 end fill

end array

read bounds

 +xb=specular

 -xb=specular

 +yb=specular

 +zb=void

 -zb=void

' the -yb must be kept at specular because only half of cask is modeled

 -yb=specular

end bounds

end data

end
A-5

Composition input for fuel burned to 20 GWd/MTU, following a 40-year cooling 

time.

Replaces composition 101 provided in the above generic burnup credit–32 (GBC-32)

cask model input file.

' 3.14 wt%, 20 gwd/mtu, 40 year cooling time

 u-234 101 0 3.2108E-07 293.0 end

 u-235 101 0 3.6060E-04 293.0 end

 u-236 101 0 7.0252E-05 293.0 end

 u-238 101 0 2.2345E-02 293.0 end

np-237 101 0 6.2485E-06 293.0 end

pu-238 101 0 7.7547E-07 293.0 end

pu-239 101 0 1.3705E-04 293.0 end

pu-240 101 0 3.9477E-05 293.0 end

pu-241 101 0 3.0797E-06 293.0 end

pu-242 101 0 3.8877E-06 293.0 end

am-241 101 0 1.7840E-05 293.0 end

am-243 101 0 5.1371E-07 293.0 end

mo-95 101 0 2.8435E-05 293.0 end

tc-99 101 0 2.9023E-05 293.0 end

ru-101 101 0 2.6347E-05 293.0 end

rh-103 101 0 1.8264E-05 293.0 end

ag-109 101 0 2.1375E-06 293.0 end

cs-133 101 0 3.0552E-05 293.0 end

nd-143 101 0 2.2819E-05 293.0 end

nd-145 101 0 1.6695E-05 293.0 end

sm-147 101 0 7.4723E-06 293.0 end

sm-149 101 0 1.7982E-07 293.0 end

sm-150 101 0 6.4540E-06 293.0 end

sm-151 101 0 3.1754E-07 293.0 end

eu-151 101 0 1.1485E-07 293.0 end

sm-152 101 0 2.6639E-06 293.0 end

eu-153 101 0 2.2518E-06 293.0 end

gd-155 101 0 1.3548E-07 293.0 end

 o-16 101 0 4.6943E-02 293.0 end
A-6

Composition input for fuel burned to 40 GWd/MTU, following a 5-year cooling 

time.

Replaces composition 101 provided in the above GBC-32 input file.

' 4.68 wt%, 40 gwd/mtu, 5 year cooling time

 u-234 101 0 2.6245E-07 293.0 end

 u-235 101 0 3.6741E-04 293.0 end

 u-236 101 0 1.3508E-04 293.0 end

 u-238 101 0 2.1683E-02 293.0 end

np-237 101 0 1.3907E-05 293.0 end

pu-238 101 0 4.5238E-06 293.0 end

pu-239 101 0 1.7413E-04 293.0 end

pu-240 101 0 6.3031E-05 293.0 end

pu-241 101 0 3.2475E-05 293.0 end

pu-242 101 0 1.1666E-05 293.0 end

am-241 101 0 9.7110E-06 293.0 end

am-243 101 0 2.5940E-06 293.0 end

mo-95 101 0 5.4717E-05 293.0 end

tc-99 101 0 5.5148E-05 293.0 end

ru-101 101 0 5.1618E-05 293.0 end

rh-103 101 0 3.1900E-05 293.0 end

ag-109 101 0 4.3961E-06 293.0 end

cs-133 101 0 5.7163E-05 293.0 end

nd-143 101 0 4.1309E-05 293.0 end

nd-145 101 0 3.1395E-05 293.0 end

sm-147 101 0 9.3467E-06 293.0 end

sm-149 101 0 2.2875E-07 293.0 end

sm-150 101 0 1.3517E-05 293.0 end

sm-151 101 0 6.4777E-07 293.0 end

eu-151 101 0 2.5993E-08 293.0 end

sm-152 101 0 4.2868E-06 293.0 end

eu-153 101 0 4.9812E-06 293.0 end

gd-155 101 0 1.7686E-07 293.0 end

 o-16 101 0 4.6949E-02 293.0 end
A-7

Composition input for fuel burned to 40 GWd/MTU, following a 40-year cooling 

time.

Replaces composition 101 provided in the above GBC-32 input file.

' 4.68 wt%, 40 gwd/mtu, 40 year cooling time

 u-234 101 0 1.3559E-06 293.0 end

 u-235 101 0 3.6758E-04 293.0 end

 u-236 101 0 1.3532E-04 293.0 end

 u-238 101 0 2.1683E-02 293.0 end

np-237 101 0 1.5362E-05 293.0 end

pu-238 101 0 3.4316E-06 293.0 end

pu-239 101 0 1.7396E-04 293.0 end

pu-240 101 0 6.3301E-05 293.0 end

pu-241 101 0 5.9464E-06 293.0 end

pu-242 101 0 1.1665E-05 293.0 end

am-241 101 0 3.4784E-05 293.0 end

am-243 101 0 2.5854E-06 293.0 end

mo-95 101 0 5.4717E-05 293.0 end

tc-99 101 0 5.5142E-05 293.0 end

ru-101 101 0 5.1618E-05 293.0 end

rh-103 101 0 3.1900E-05 293.0 end

ag-109 101 0 4.3961E-06 293.0 end

cs-133 101 0 5.7163E-05 293.0 end

nd-143 101 0 4.1309E-05 293.0 end

nd-145 101 0 3.1395E-05 293.0 end

sm-147 101 0 1.1950E-05 293.0 end

sm-149 101 0 2.2875E-07 293.0 end

sm-150 101 0 1.3517E-05 293.0 end

sm-151 101 0 4.9471E-07 293.0 end

eu-151 101 0 1.7905E-07 293.0 end

sm-152 101 0 4.2871E-06 293.0 end

eu-153 101 0 4.9812E-06 293.0 end

gd-155 101 0 3.3841E-07 293.0 end

 o-16 101 0 4.6949E-02 293.0 end

B-1

APPENDIX B. MCNP MODEL

This appendix includes a MCNP model and spent nuclear fuel composition information in the 

MCNP input format for fuel burned to 20 GWd/MTU, following a 5-year post-irradiation cooling 

time. Composition inputs for other burnup/cooling time combinations are provided following this 

MCNP input file.

GBC-32, FP&MA reference worth model

c

c *******************************************************

c * GBC-32 cask (NUREG/CR-6747)

c * Assembly Type: Westinghouse 17x17 OFA/V5

c *

c *******************************************************

c ---------- ASSEMBLY ----------

c fuel rod

1 1 7.0120e-02 -1 u=2 $ fuel

2 3 0.10011 1 -2 u=2 $ gap

3 2 0.0423 2 -3 u=2 $ Zr Clad

4 3 0.10011 3 u=2 $ water in fuel 

cell

c guide tube

5 3 0.10011 -4:5 u=3 $ water in guide 

tubes

6 2 0.0423 4 -5 u=3 $ guide tubes

c assembly

7 3 0.10011 -6 7 -8 9 u=1

 lat=1 fill= -9:9 -9:9 0:0

 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1

 1 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 1

 1 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 1

 1 2 2 2 2 2 3 2 2 3 2 2 3 2 2 2 2 2 1

 1 2 2 2 3 2 2 2 2 2 2 2 2 2 3 2 2 2 1

 1 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 1

 1 2 2 3 2 2 3 2 2 3 2 2 3 2 2 3 2 2 1

 1 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 1

 1 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 1

 1 2 2 3 2 2 3 2 2 3 2 2 3 2 2 3 2 2 1

 1 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 1

 1 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 1

 1 2 2 3 2 2 3 2 2 3 2 2 3 2 2 3 2 2 1

 1 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 1

 1 2 2 2 3 2 2 2 2 2 2 2 2 2 3 2 2 2 1

 1 2 2 2 2 2 3 2 2 3 2 2 3 2 2 2 2 2 1

 1 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 1

 1 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 1

 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1

c

c

c ---------- STANDARD BASKET ----------

c

c Universe 4 = standard symetric cell with assembly

10 0 20 -21 22 -23 fill=1 u=4

11 4 0.08625 30 -31 32 -33 (-20:21:-22:23) u=4

c right
B-2

20 3 0.10011 31 #21 #22 u=4 $ Water

21 7 0.0602 31 -41 12 -13 u=4 $ Al Boral Clad

22 5 0.084089 41 12 -13 u=4 $ Boral Center

c left

30 3 0.10011 -30 #31 #32 u=4 $ Water

31 7 0.0602 -30 40 12 -13 u=4 $ Al Boral Clad

32 5 0.084089 -40 12 -13 u=4 $ Boral Center

c top

40 3 0.10011 33 30 -31 #41 #42 u=4 $ Water

41 7 0.0602 33 -43 10 -11 u=4 $ Al Boral Clad

42 5 0.084089 43 10 -11 u=4 $ Boral Center

c bottom

50 3 0.10011 -32 30 -31 #51 #52 u=4 $ Water

51 7 0.0602 -32 42 10 -11 u=4 $ Al Boral Clad

52 5 0.084089 -42 10 -11 u=4 $ Boral Center

c

c

c ---------- PERIPHERY BASKET/POISON PLATES ----------

c

c Universe 5 = top cell, which has boral on bottom

100 3 0.10011 22 u=5

101 3 0.10011 -22 #102 #103 #104 u=5

102 4 0.08625 30 -31 -22 32 u=5

103 7 0.0602 -32 42 10 -11 u=5 $ Al Boral Clad

104 5 0.084089 -42 10 -11 u=5 $ Boral Center

c

c Universe 6 = bottom cell, which has boral on top

110 3 0.10011 -23 u=6

111 3 0.10011 23 #112 #113 #114 u=6

112 4 0.08625 30 -31 23 -33 u=6

113 7 0.0602 33 -43 10 -11 u=6 $ Al Boral Clad

114 5 0.084089 43 10 -11 u=6 $ Boral Center

c

c Universe 7 = right-hand side cell, which has boral on left-hand side

120 3 0.10011 20 u=7

121 3 0.10011 -20 #122 #123 #124 u=7

122 4 0.08625 22 -23 -20 30 u=7

123 7 0.0602 -30 40 12 -13 u=7 $ Al Boral Clad

124 5 0.084089 -40 12 -13 u=7 $ Boral Center

c

c Universe 8 = left-hand side cell, which has boral on right-hand side

130 3 0.10011 -21 u=8

131 3 0.10011 21 #132 #133 #134 u=8

132 4 0.08625 22 -23 21 -31 u=8

133 7 0.0602 31 -41 12 -13 u=8 $ Al Boral Clad

134 5 0.084089 41 12 -13 u=8 $ Boral Center

c

c

c ---------- BASKET CORNERS ----------

c

c Universe 10 = top-left corner cell, which has boral on bottom-right

150 3 0.10011 -21 22 u=10

151 4 0.08625 -31 32 (21:-22) #152 #153 u=10

152 3 0.10011 21 -31 23 -53 u=10

153 3 0.10011 50 -30 32 -22 u=10

c right

155 3 0.10011 31 #156 #157 u=10 $ Water
B-3

156 7 0.0602 31 -41 12 -13 u=10 $ Al Boral Clad

157 5 0.084089 41 12 -13 u=10 $ Boral Center

c bottom

160 3 0.10011 -32 -31 #161 #162 u=10 $ Water

161 7 0.0602 -32 42 10 -11 u=10 $ Al Boral Clad

162 5 0.084089 -42 10 -11 u=10 $ Boral Center

c

c

c Universe 11 = top-right corner cell, which has boral on bottom-left

170 3 0.10011 20 22 u=11

171 4 0.08625 30 32 (-20:-22) #172 #173 u=11

172 3 0.10011 30 -20 23 -53 u=11

173 3 0.10011 31 -51 32 -22 u=11

c left

175 3 0.10011 -30 #176 #177 u=11 $ Water

176 7 0.0602 -30 40 12 -13 u=11 $ Al Boral Clad

177 5 0.084089 -40 12 -13 u=11 $ Boral Center

c bottom

180 3 0.10011 -32 30 #181 #182 u=11 $ Water

181 7 0.0602 -32 42 10 -11 u=11 $ Al Boral Clad

182 5 0.084089 -42 10 -11 u=11 $ Boral Center

c

c

c Universe 12 = bottom-left corner cell, which has boral on top-right

190 3 0.10011 -21 -23 u=12

191 4 0.08625 -31 -33 (21:23) #192 #193 u=12

192 3 0.10011 50 -30 23 -33 u=12

193 3 0.10011 21 -31 52 -22 u=12

c right

195 3 0.10011 31 #196 #197 u=12 $ Water

196 7 0.0602 31 -41 12 -13 u=12 $ Al Boral Clad

197 5 0.084089 41 12 -13 u=12 $ Boral Center

c top

200 3 0.10011 33 -31 #201 #202 u=12 $ Water

201 7 0.0602 33 -43 10 -11 u=12 $ Al Boral Clad

202 5 0.084089 43 10 -11 u=12 $ Boral Center

c

c

c Universe 13 = bottom-right corner cell, which has boral on top-left

210 3 0.10011 20 -23 u=13

211 4 0.08625 30 -33 (-20:23) #212 #213 u=13

212 3 0.10011 31 -51 23 -33 u=13

213 3 0.10011 30 -20 52 -22 u=13

c left

215 3 0.10011 -30 #216 #217 u=13 $ Water

216 7 0.0602 -30 40 12 -13 u=13 $ Al Boral Clad

217 5 0.084089 -40 12 -13 u=13 $ Boral Center

c top

220 3 0.10011 33 30 #221 #222 u=13 $ Water

221 7 0.0602 33 -43 10 -11 u=13 $ Al Boral Clad

222 5 0.084089 43 10 -11 u=13 $ Boral Center

c

c

c ---------- CASK ----------

c

c Cask Loading Pattern

500 3 0.10011 50 -51 52 -53 u=20 lat=1
B-4

 fill=-5:4 -5:4 0:0

 20 20 20 20 20 20 20 20 20 20

 20 20 20 5 5 5 5 20 20 20

 20 20 11 4 4 4 4 10 20 20

 20 7 4 4 4 4 4 4 8 20

 20 7 4 4 4 4 4 4 8 20

 20 7 4 4 4 4 4 4 8 20

 20 7 4 4 4 4 4 4 8 20

 20 20 13 4 4 4 4 12 20 20

 20 20 20 6 6 6 6 20 20 20

 20 20 20 20 20 20 20 20 20 20

c rhs lhs

c

600 0 -200 101 -102 fill=20 (-11.87827 -11.87827 0.0)

601 3 0.10011 -200 100 -101 $ water above baseplate and below fuel

602 6 0.08625 -201 99 -100 $ steel baseplate

603 3 0.10011 -200 102 -103 $ water above fuel and below cask lid

604 6 0.08625 -201 103 -104 $ steel cask lid

605 6 0.08625 200 -201 100 -103 $ cask

700 0 300 -301 302 -303 99 -104 201 $ enclosing void

701 0 (-300:301:-302:303:-99:104) $ outside universe

1 cz 0.3922 $ Pellet O.R.

2 cz 0.40005 $ Clad I.R.

3 cz 0.4572 $ Clad O.R.

4 cz 0.56135 $ Water rod I.R.

5 cz 0.6020 $ Water rod O.R.

6 px 0.6299 $ Fuel rod half-pitch

7 px -0.6299 $ "

8 py 0.6299 $ "

9 py -0.6299 $ "

10 px -9.525 $ boral plate half-width

11 px 9.525 $ "

12 py -9.525 $ "

13 py 9.525 $ "

20 px -11.0000 $ SS Box cell I.D., 8.6614"

21 px 11.0000 $ "

22 py -11.0000 $ "

23 py 11.0000 $ "

30 px -11.7500 $ SS Box cell O.D., thkns=0.2953"

31 px 11.7500 $ "

32 py -11.7500 $ "

33 py 11.7500 $ "

40 px -11.7754 $ Al clad thickness outside Boral

41 px 11.7754 $ "

42 py -11.7754 $ "

43 py 11.7754 $ "

50 px -11.87825 $ cell half-pitch (to center of Boral plate)

51 px 11.87825 $ "

52 py -11.87825 $ "

53 py 11.87825 $ "

c

c axial planes

99 pz -45. $ Bottom of Cask Baseplate

100 pz -15. $ Top of Cask Baseplate

101 pz 0. $ Bottom of fuel

102 pz 365.76 $ Top of fuel
B-5

103 pz 395.76 $ Bottom of Cask Lid

104 pz 425.76 $ Top of Cask Lid

c

200 cz 87.5 $ Cask body inner radius

201 cz 107.5 $ Cask body outer radius

c

*300 px -108.0 $ Outer Enclosing Box

*301 px 108.0

*302 py -108.0

*303 py 108.0

imp:n 1 88r 0

kcode 10000 0.95 100 10000 0 0 200 1

c NSRCK RKK IKZ KCT MSRK KNRM MRKP KC8

c

c

c Initial Source = Uniform throughout assemblies

c

sdef par=1 erg=d1 axs=0 0 1 x=d4 y=fx d5 z=d3

c

sp1 -2 1.2895

c

si3 h 0 365.76

sp3 0 1

c

c

si4 s 12 13 14 15

 11 12 13 14 15 16

 11 12 13 14 15 16

 11 12 13 14 15 16

 11 12 13 14 15 16

 12 13 14 15

sp4 1 31r

c

ds5 s 26 26 26 26

 25 25 25 25 25 25

 24 24 24 24 24 24

 23 23 23 23 23 23

 22 22 22 22 22 22

 21 21 21 21

c

si11 -66.3500 -44.9327

si12 -44.0935 -22.6762

si13 -21.8369 -0.4196

si14 0.4196 21.8369

si15 22.6762 44.0935

si16 44.9327 66.3500

c

si21 -66.3500 -44.9327

si22 -44.0935 -22.6762

si23 -21.8369 -0.4196

si24 0.4196 21.8369

si25 22.6762 44.0935

si26 44.9327 66.3500

c

sp11 0 1

sp12 0 1
B-6

sp13 0 1

sp14 0 1

sp15 0 1

sp16 0 1

sp21 0 1

sp22 0 1

sp23 0 1

sp24 0 1

sp25 0 1

sp26 0 1

c

c

c

c Material Cards

c

c Fuel 20 GWd/MTU & 5 year cooling time

c (total = 7.0120e-02)

m1 92234.70c 7.3901e-08

 92235.70c 3.6046e-04

 92236.70c 7.0106e-05

 92238.70c 2.2345e-02

 93237.70c 5.5048e-06

 94238.70c 1.0223e-06

 94239.70c 1.3719e-04

 94240.70c 3.9569e-05

 94241.70c 1.6819e-05

 94242.70c 3.8879e-06

 95241.70c 4.8440e-06

 95243.70c 5.1541e-07

 42095.70c 2.8435e-05

 43099.70c 2.9026e-05

 44101.70c 2.6347e-05

 45103.70c 1.8264e-05

 47109.70c 2.1375e-06

 55133.70c 3.0552e-05

 60143.70c 2.2819e-05

 60145.70c 1.6695e-05

 62147.70c 5.6742e-06

 62149.70c 1.7982e-07

 62150.70c 6.4540e-06

 62151.70c 4.1578e-07

 63151.70c 1.6613e-08

 62152.70c 2.6638e-06

 63153.70c 2.2518e-06

 64155.70c 7.0661e-08

 8016.70c 4.6943e-02

c

c Zr clad

m2 40090.70c 2.1763e-02

 40091.70c 4.7461e-03

 40092.70c 7.2545e-03

 40094.70c 7.3517e-03

 40096.70c 1.1844e-03

c Water

m3 8016.70c 3.3370e-02

 1001.70c 6.6740e-02

mt3 lwtr.10t
B-7

c

c Basket Material - Stainless Steel Type 304

m4 24050.70c 7.5821e-04

 24052.70c 1.4605e-02

 24053.70c 1.6559e-03

 24054.70c 4.1135e-04

 25055.70c 1.7400e-03

 26054.70c 3.4726e-03

 26056.70c 5.4463e-02

 26057.70c 1.2584e-03

 26058.70c 1.6621e-04

 28058.70c 5.2558e-03

 28060.70c 2.0242e-03

 28061.70c 8.8008e-05

 28062.70c 2.8024e-04

 28064.70c 7.1796e-05

c

c Boral Central Section @ 0.0225 g/cmsq

m5 5010.70c 6.5794e-03

 5011.70c 2.7260e-02

 6000.70c 8.4547e-03

 13027.70c 4.1795e-02

c

c Cask Material - Stainless Steel Type 304

m6 24050.70c 7.5821e-04

 24052.70c 1.4605e-02

 24053.70c 1.6559e-03

 24054.70c 4.1135e-04

 25055.70c 1.7400e-03

 26054.70c 3.4726e-03

 26056.70c 5.4463e-02

 26057.70c 1.2584e-03

 26058.70c 1.6621e-04

 28058.70c 5.2558e-03

 28060.70c 2.0242e-03

 28061.70c 8.8008e-05

 28062.70c 2.8024e-04

 28064.70c 7.1796e-05

c

c Aluminum Clad for Boral

m7 13027.70c 6.0200e-02

c

prdmp j -60 j 2
B-8

Composition input for fuel burned to 20 GWd/MTU, following a 40-year cooling 

time.

The total number density in cell 1 is changed to 7.0122e-02

Replace composition m1 provided in the above generic burnup credit–32 (GBC-32)

cask model input file.

c Fuel 20 GWd/MTU & 40 year cooling time

c (total = 7.0122e-02)

m1 92234.70c 3.2108e-07

 92235.70c 3.6060e-04

 92236.70c 7.0252e-05

 92238.70c 2.2345e-02

 93237.70c 6.2485e-06

 94238.70c 7.7547e-07

 94239.70c 1.3705e-04

 94240.70c 3.9477e-05

 94241.70c 3.0797e-06

 94242.70c 3.8877e-06

 95241.70c 1.7840e-05

 95243.70c 5.1371e-07

 42095.70c 2.8435e-05

 43099.70c 2.9023e-05

 44101.70c 2.6347e-05

 45103.70c 1.8264e-05

 47109.70c 2.1375e-06

 55133.70c 3.0552e-05

 60143.70c 2.2819e-05

 60145.70c 1.6695e-05

 62147.70c 7.4723e-06

 62149.70c 1.7982e-07

 62150.70c 6.4540e-06

 62151.70c 3.1754e-07

 63151.70c 1.1485e-07

 62152.70c 2.6639e-06

 63153.70c 2.2518e-06

 64155.70c 1.3548e-07

 8016.70c 4.6943e-02
B-9

Composition input for fuel burned to 40 GWd/MTU, following a 5-year cooling 

time.

The total number density in cell 1 is changed to 6.9808e-02

Replace composition m1 provided in the above GBC-32 input file.

c Fuel 40 GWd/MTU & 5 year cooling time

c (total = 6.9808e-02)

m1 92234.70c 2.6245e-07

 92235.70c 3.6741e-04

 92236.70c 1.3508e-04

 92238.70c 2.1683e-02

 93237.70c 1.3907e-05

 94238.70c 4.5238e-06

 94239.70c 1.7413e-04

 94240.70c 6.3031e-05

 94241.70c 3.2475e-05

 94242.70c 1.1666e-05

 95241.70c 9.7110e-06

 95243.70c 2.5940e-06

 42095.70c 5.4717e-05

 43099.70c 5.5148e-05

 44101.70c 5.1618e-05

 45103.70c 3.1900e-05

 47109.70c 4.3961e-06

 55133.70c 5.7163e-05

 60143.70c 4.1309e-05

 60145.70c 3.1395e-05

 62147.70c 9.3467e-06

 62149.70c 2.2875e-07

 62150.70c 1.3517e-05

 62151.70c 6.4777e-07

 63151.70c 2.5993e-08

 62152.70c 4.2868e-06

 63153.70c 4.9812e-06

 64155.70c 1.7686e-07

 8016.70c 4.6949e-02
B-10

Composition input for fuel burned to 40 GWd/MTU, following a 40-year cooling 

time.

The total number density in cell 1 is changed to 6.9811e-02

Replace composition m1 provided in the above GBC-32 input file.

c Fuel 40 GWd/MTU & 40 year cooling time

c (total = 6.9811e-02)

m1 92234.70c 1.3559e-06

 92235.70c 3.6758e-04

 92236.70c 1.3532e-04

 92238.70c 2.1683e-02

 93237.70c 1.5362e-05

 94238.70c 3.4316e-06

 94239.70c 1.7396e-04

 94240.70c 6.3301e-05

 94241.70c 5.9464e-06

 94242.70c 1.1665e-05

 95241.70c 3.4784e-05

 95243.70c 2.5854e-06

 42095.70c 5.4717e-05

 43099.70c 5.5142e-05

 44101.70c 5.1618e-05

 45103.70c 3.1900e-05

 47109.70c 4.3961e-06

 55133.70c 5.7163e-05

 60143.70c 4.1309e-05

 60145.70c 3.1395e-05

 62147.70c 1.1950e-05

 62149.70c 2.2875e-07

 62150.70c 1.3517e-05

 62151.70c 4.9471e-07

 63151.70c 1.7905e-07

 62152.70c 4.2871e-06

 63153.70c 4.9812e-06

 64155.70c 3.3841e-07

 8016.70c 4.6949e-02
C-1

APPENDIX C. DETAILED FP&MA WORTH RESULTS FOR EACH 

CODE AND NUCLEAR DATA COMBINATION

This appendix provides detailed fission products and minor actinides (FP&MA) worth 

results for each code and nuclear data combination. Results for SCALE 6.1 using the 

ENDF/B–VII 238 group library are provided in Table 3.1.

Table C.1. FP&MA worths using SCALE 6.1 and ENDF/B–VII continuous-energy data

Initial enrichment and

final burnup

3.1 wt % and 20 GWd/MTU 4.7 wt % and 40 GWd/MTU

Cooling time

(years) 5 40 5 40

FP&MA worths

(∆k)a

All FP&MA -0.06720 -0.06945 -0.10149 -0.10399

All MA -0.00609 -0.00657 -0.01120 -0.01096

All FP -0.06018 -0.06255 -0.08745 -0.09037

236U -0.00367 -0.00390 -0.00544 -0.00512

237Np -0.00209 -0.00249 -0.00435 -0.00472

243Am -0.00021 -0.00058 -0.00094 -0.00099

95Mo -0.00120 -0.00135 -0.00160 -0.00170

99Tc -0.00231 -0.00270 -0.00384 -0.00365

101Ru -0.00073 -0.00121 -0.00146 -0.00134

103Rh -0.00695 -0.00707 -0.01049 -0.00968

109Ag -0.00074 -0.00123 -0.00145 -0.00128

133Cs -0.00334 -0.00352 -0.00521 -0.00497

143Nd -0.00866 -0.00898 -0.01299 -0.01249

145Nd -0.00165 -0.00207 -0.00279 -0.00293

147Sm -0.00128 -0.00212 -0.00199 -0.00239

149Sm -0.01555 -0.01532 -0.01597 -0.01532

150Sm -0.00071 -0.00114 -0.00128 -0.00123

151Sm -0.00584 -0.00465 -0.00701 -0.00522

152Sm -0.00198 -0.00256 -0.00279 -0.00284

151Eu -0.00004 -0.00156 0.00002 -0.00150

153Eu -0.00165 -0.00171 -0.00270 -0.00288

155Gd -0.00357 -0.00712 -0.00667 -0.01294

a Monte Carlo uncertainty is 0.00014 ∆k for FP&MA worths.
C-2

Table C.2. FP&MA worths using SCALE 6.1 and ENDF/B–VI 238 energy group data

Initial enrichment and 

final burnup

3.1 wt % and 20 GWd/MTU 4.7 wt % and 40 GWd/MTU

Cooling time

(years) 5 40 5 40

FP&MA worths

(∆k)a

All FP&MA -0.06755 -0.06888 -0.10152 -0.10374

All MA -0.00673 -0.00618 -0.01158 -0.01087

All FP -0.06022 -0.06198 -0.08743 -0.09012

236U -0.00396 -0.00345 -0.00570 -0.00507

237Np -0.00225 -0.00238 -0.00471 -0.00471

243Am -0.00023 -0.00007 -0.00098 -0.00080

95Mo -0.00132 -0.00100 -0.00179 -0.00164

99Tc -0.00211 -0.00198 -0.00336 -0.00321

101Ru -0.00069 -0.00078 -0.00141 -0.00111

103Rh -0.00704 -0.00658 -0.01029 -0.00928

109Ag -0.00109 -0.00071 -0.00154 -0.00157

133Cs -0.00341 -0.00296 -0.00531 -0.00487

143Nd -0.00912 -0.00870 -0.01319 -0.01229

145Nd -0.00183 -0.00161 -0.00271 -0.00270

147Sm -0.00143 -0.00155 -0.00185 -0.00252

149Sm -0.01655 -0.01574 -0.01688 -0.01619

150Sm -0.00099 -0.00050 -0.00148 -0.00126

151Sm -0.00629 -0.00419 -0.00723 -0.00534

152Sm -0.00222 -0.00190 -0.00309 -0.00272

151Eu -0.00024 -0.00105 -0.00018 -0.00134

153Eu -0.00170 -0.00123 -0.00286 -0.00270

155Gd -0.00372 -0.00650 -0.00730 -0.01302

a Monte Carlo uncertainty is 0.00014 ∆k for FP&MA worths.
C-3

Table C.3. FP&MA worths using SCALE 6.1 and ENDF/B–V 238 energy group data

Initial enrichment and 

final burnup

3.1 wt % and 20 GWd/MTU 4.7 wt % and 40 GWd/MTU

Cooling time

(years) 5 40 5 40

FP&MA worths

(∆k)a

All FP&MA -0.06585 -0.06808 -0.09966 -0.10229

All MA -0.00615 -0.00632 -0.01159 -0.01109

All FP -0.05856 -0.06113 -0.08544 -0.08835

236U -0.00349 -0.00375 -0.00545 -0.00518

237Np -0.00197 -0.00247 -0.00484 -0.00496

243Am -0.00016 -0.00049 -0.00110 -0.00111

95Mo -0.00114 -0.00116 -0.00182 -0.00176

99Tc -0.00204 -0.00221 -0.00376 -0.00350

101Ru -0.00056 -0.00077 -0.00118 -0.00137

103Rh -0.00678 -0.00668 -0.01041 -0.00959

109Ag -0.00067 -0.00108 -0.00171 -0.00138

133Cs -0.00272 -0.00299 -0.00463 -0.00450

143Nd -0.00885 -0.00886 -0.01305 -0.01268

145Nd -0.00149 -0.00170 -0.00269 -0.00257

147Sm -0.00134 -0.00186 -0.00210 -0.00270

149Sm -0.01547 -0.01538 -0.01634 -0.01567

150Sm -0.00077 -0.00089 -0.00173 -0.00159

151Sm -0.00552 -0.00445 -0.00700 -0.00537

152Sm -0.00186 -0.00215 -0.00323 -0.00305

151Eu 0.00019 -0.00146 -0.00026 -0.00176

153Eu -0.00118 -0.00145 -0.00290 -0.00271

155Gd -0.00316 -0.00662 -0.00709 -0.01287

a Monte Carlo uncertainty is 0.00014 ∆k for FP&MA worths.
C-4

Table C.4. FP&MA worths using SCALE 6.1 and ENDF/B–V 44 energy group data

Initial enrichment and 

final burnup

3.1 wt % and 20 GWd/MTU 4.7 wt % and 40 GWd/MTU

Cooling time

(years) 5 40 5 40

FP&MA worths

(∆k)a

All FP&MA -0.06635 -0.06836 -0.10000 -0.10288

All MA -0.00622 -0.00639 -0.01186 -0.01126

All FP -0.05915 -0.06132 -0.08593 -0.08896

236U -0.00402 -0.00404 -0.00606 -0.00542

237Np -0.00221 -0.00249 -0.00489 -0.00480

243Am -0.00015 -0.00037 -0.00113 -0.00091

95Mo -0.00131 -0.00131 -0.00204 -0.00158

99Tc -0.00276 -0.00253 -0.00420 -0.00379

101Ru -0.00074 -0.00089 -0.00146 -0.00118

103Rh -0.00699 -0.00674 -0.01041 -0.00923

109Ag -0.00084 -0.00117 -0.00166 -0.00137

133Cs -0.00316 -0.00317 -0.00505 -0.00488

143Nd -0.00897 -0.00882 -0.01332 -0.01286

145Nd -0.00154 -0.00186 -0.00283 -0.00255

147Sm -0.00144 -0.00161 -0.00215 -0.00247

149Sm -0.01571 -0.01557 -0.01668 -0.01569

150Sm -0.00133 -0.00139 -0.00209 -0.00206

151Sm -0.00581 -0.00427 -0.00705 -0.00517

152Sm -0.00205 -0.00187 -0.00309 -0.00256

151Eu -0.00033 -0.00144 -0.00038 -0.00128

153Eu -0.00143 -0.00145 -0.00285 -0.00264

155Gd -0.00359 -0.00687 -0.00696 -0.01301

a Monte Carlo uncertainty is 0.00014 ∆k for FP&MA worths.
C-5

Table C.5. FP&MA worths using MCNP5 (version 1.60) and ENDF/B–VII data

Initial enrichment and 

final burnup

3.1 wt % and 20 GWd/MTU 4.7 wt % and 40 GWd/MTU

Cooling time

(years) 5 40 5 40

FP&MA worths

(∆k)a

All FP&MA -0.06748 -0.06969 -0.10213 -0.10456

All MA -0.00618 -0.00619 -0.01127 -0.01073

All FP -0.06042 -0.06244 -0.08820 -0.09096

236U (92236.70c) -0.00376 -0.00374 -0.00580 -0.00523

237Np (93237.70c) -0.00216 -0.00241 -0.00443 -0.00460

243Am (95243.70c) -0.00002 -0.00038 -0.00117 -0.00104

95Mo (42095.70c) -0.00101 -0.00105 -0.00176 -0.00164

99Tc (43099.70c) -0.00256 -0.00259 -0.00388 -0.00365

101Ru (44101.70c) -0.00076 -0.00093 -0.00135 -0.00142

103Rh (45103.70c) -0.00688 -0.00675 -0.01026 -0.00944

109Ag (47109.70c) -0.00087 -0.00080 -0.00160 -0.00139

133Cs (55133.70c) -0.00334 -0.00328 -0.00546 -0.00511

143Nd (60143.70c) -0.00877 -0.00887 -0.01331 -0.01259

145Nd (60145.70c) -0.00176 -0.00187 -0.00282 -0.00276

147Sm (62147.70c) -0.00137 -0.00191 -0.00214 -0.00240

149Sm (62149.70c) -0.01570 -0.01530 -0.01637 -0.01564

150Sm (62150.70c) -0.00081 -0.00097 -0.00151 -0.00136

151Sm (62151.70c) -0.00612 -0.00461 -0.00734 -0.00555

152Sm (62152.70c) -0.00211 -0.00220 -0.00298 -0.00285

151Eu (63151.70c) -0.00015 -0.00142 -0.00020 -0.00158

153Eu (63153.70c) -0.00167 -0.00156 -0.00270 -0.00271

155Gd (64155.70c) -0.00354 -0.00689 -0.00723 -0.01301

Note: MCNP = Monte Carlo N-Particle. a Monte Carlo uncertainty is less than 0.00012 ∆k for FP&MA worths.
C-6

Table C.6. FP&MA worths using MCNP5 (version 1.60) and ENDF/B–VI data

Initial enrichment and 

final burnup

3.1 wt % and 20 GWd/MTU 4.7 wt % and 40 GWd/MTU

Cooling time

(years) 5 40 5 40

FP&MA worths

(∆k)a

All FP&MA -0.04529 -0.04918 -0.06910 -0.07487

All MA -0.00659 -0.00616 -0.01194 -0.01158

All FP -0.03797 -0.04222 -0.05474 -0.06139

236U (92236.69c) -0.00403 -0.00400 -0.00611 -0.00516

237Np (93237.69c) -0.00214 -0.00242 -0.00486 -0.00500

243Am (95243.69c) -0.00024 -0.00026 -0.00087 -0.00079

95Mo b b b b

99Tc (43099.66c) -0.00236 -0.00213 -0.00361 -0.00328

101Ru b b b b

103Rh (45103.66c) -0.00728 -0.00673 -0.01080 -0.01007

109Ag (47109.66c) -0.00097 -0.00031 -0.00144 -0.00132

133Cs (55133.66c) -0.00309 -0.00326 -0.00540 -0.00475

143Nd b b b b

145Nd b b b b

147Sm (62147.66c) -0.00142 -0.00183 -0.00169 -0.00240

149Sm (62149.66c) -0.01671 -0.01582 -0.01800 -0.01679

150Sm b b b b

151Sm b b b b

152Sm b b b b

151Eu (63151.66c) -0.00041 -0.00102 -0.00027 -0.00123

153Eu (63153.66c) -0.00149 -0.00134 -0.00292 -0.00271

155Gd (64155.66c) -0.00381 -0.00699 -0.00771 -0.01424

Note: MCNP = Monte Carlo N-Particle. a Monte Carlo uncertainty is less than 0.00014 ∆k for FP&MA worths. b 95Mo, 101Ru, 143Nd, 145Nd, 150Sm, 151Sm, and 152Sm excluded because ENDF/B–VI data for these 

nuclides was not distributed with MCNP5 (v1.60).
C-7

Table C.7. FP&MA worths using MCNP5 (version 1.60) and ENDF/B–V data

Initial enrichment and 

final burnup

3.1 wt % and 20 GWd/MTU 4.7 wt % and 40 GWd/MTU

Cooling time

(years) 5 40 5 40

FP&MA worths

(∆k)a

All FP&MA -0.06688 -0.06913 -0.10154 -0.10402

All MA -0.00678 -0.00696 -0.01303 -0.01238

All FP -0.05922 -0.06133 -0.08593 -0.08895

236U (92236.51c) -0.00452 -0.00428 -0.00698 -0.00627

237Np (93237.55c) -0.00209 -0.00270 -0.00485 -0.00492

243Am (95243.51c) -0.00018 -0.00046 -0.00122 -0.00091

95Mo (42095.50c) -0.00095 -0.00130 -0.00215 -0.00155

99Tc (43099.50c) -0.00198 -0.00233 -0.00370 -0.00339

101Ru (44101.50c) -0.00051 -0.00072 -0.00129 -0.00130

103Rh (45103.50c) -0.00702 -0.00692 -0.01042 -0.00939

109Ag (47109.50c) -0.00057 -0.00096 -0.00152 -0.00171

133Cs (55133.55c) -0.00295 -0.00312 -0.00513 -0.00496

143Nd (60143.50c) -0.00891 -0.00897 -0.01318 -0.01253

145Nd (60145.50c) -0.00143 -0.00146 -0.00267 -0.00265

147Sm (62147.50c) -0.00120 -0.00162 -0.00213 -0.00241

149Sm (62149.50c) -0.01580 -0.01557 -0.01623 -0.01588

150Sm (62150.50c) -0.00092 -0.00093 -0.00160 -0.00122

151Sm (62151.50c) -0.00532 -0.00436 -0.00688 -0.00502

152Sm (62152.50c) -0.00201 -0.00218 -0.00317 -0.00311

151Eu (63151.55c) -0.00011 -0.00120 -0.00022 -0.00142

153Eu (63153.55c) -0.00128 -0.00141 -0.00263 -0.00261

155Gd (64155.50c) -0.00340 -0.00672 -0.00709 -0.01304

Note: MCNP = Monte Carlo N-Particle. a Monte Carlo uncertainty is less than 0.00013 ∆k for FP&MA worths.
C-8

Table C.8. FP&MA worths using MCNP6 (new release) and ENDF/B–V data

Initial enrichment and 

final burnup

3.1 wt % and 20 GWd/MTU 4.7 wt % and 40 GWd/MTU

Cooling time

(years) 5 40 5 40

FP&MA worths

(∆k)a

All FP&MA -0.06705 -0.06906 -0.10136 -0.10395

All MA -0.0069 -0.00673 -0.01271 -0.01233

All FP -0.05915 -0.06108 -0.08581 -0.08898

236U (92236.51c) -0.00458 -0.00418 -0.00697 -0.0062

237Np (93237.55c) -0.00211 -0.00233 -0.00477 -0.0051

243Am (95243.51c) -0.00016 -0.00028 -0.00114 -0.00096

95Mo (42095.50c) -0.00106 -0.00107 -0.0019 -0.00184

99Tc (43099.50c) -0.0023 -0.00207 -0.00356 -0.00342

101Ru (44101.50c) -0.00075 -0.0005 -0.00113 -0.00119

103Rh (45103.50c) -0.00701 -0.0066 -0.01033 -0.00942

109Ag (47109.50c) -0.00088 -0.00076 -0.0013 -0.00139

133Cs (55133.55c) -0.00315 -0.00288 -0.00512 -0.00479

143Nd (60143.50c) -0.00885 -0.00847 -0.0132 -0.01262

145Nd (60145.50c) -0.0017 -0.00147 -0.00259 -0.00247

147Sm (62147.50c) -0.00145 -0.00162 -0.00203 -0.00252

149Sm (62149.50c) -0.01593 -0.01532 -0.01642 -0.01571

150Sm (62150.50c) -0.00063 -0.00077 -0.0015 -0.00144

151Sm (62151.50c) -0.00552 -0.00395 -0.00686 -0.00499

152Sm (62152.50c) -0.00227 -0.00211 -0.00309 -0.00283

151Eu (63151.55c) -0.00008 -0.00109 -0.0002 -0.00149

153Eu (63153.55c) -0.00142 -0.00129 -0.00288 -0.0025

155Gd (64155.50c) -0.00357 -0.00661 -0.00706 -0.01316

Note: MCNP = Monte Carlo N-Particle. a Monte Carlo uncertainty is less than 0.00013 ∆k for FP&MA worths.
C-9

Table C.9. FP&MA worths using MCNP6 (new release) and ENDF/B–VI data

Initial enrichment and 

final burnup

3.1 wt % and 20 GWd/MTU 4.7 wt % and 40 GWd/MTU

Cooling time

(years) 5 40 5 40

FP&MA worths

(∆k)a

All FP&MA -0.04525 -0.04908 -0.0692 -0.07511

All MA -0.00658 -0.00641 -0.01215 -0.0118

All FP -0.03828 -0.04205 -0.05505 -0.06144

236U (92236.69c) -0.00409 -0.00375 -0.00591 -0.00552

237Np (93237.69c) -0.00225 -0.0023 -0.00495 -0.00527

243Am (95243.69c) -0.00007 -0.00028 -0.00101 -0.00106

99Tc (43099.66c) -0.00235 -0.00206 -0.00377 -0.00352

103Rh (45103.66c) -0.0072 -0.00687 -0.01072 -0.01007

109Ag (47109.66c) -0.00087 -0.00076 -0.00138 -0.00145

133Cs (55133.66c) -0.00305 -0.0031 -0.00521 -0.0050

147Sm (62147.66c) -0.00136 -0.00174 -0.00198 -0.00264

149Sm (62149.66c) -0.01680 -0.01613 -0.01775 -0.01692

151Eu (63151.66c) -0.00016 -0.00117 -0.00001 -0.00154

153Eu (63153.66c) -0.00152 -0.00131 -0.00291 -0.00283

155Gd (64155.66c) -0.00363 -0.00707 -0.00772 -0.01431

Note: MCNP = Monte Carlo N-Particle. a Monte Carlo uncertainty is less than 0.00013 ∆k for FP&MA worths.
C-10

Table C.10. FP&MA worths using MCNP6 (new release) and ENDF/B–VII data

Initial enrichment and 

final burnup

3.1 wt % and 20 GWd/MTU 4.7 wt % and 40 GWd/MTU

Cooling time

(years) 5 40 5 40

FP&MA worths

(∆k)a

All FP&MA -0.06762 -0.06961 -0.10222 -0.1045

All MA -0.00606 -0.00608 -0.01126 -0.01089

All FP -0.06053 -0.06269 -0.08838 -0.09093

236U (92236.70c) -0.00393 -0.00368 -0.00579 -0.00514

237Np (93237.70c) -0.0019 -0.00232 -0.00447 -0.00448

243Am (95243.70c) -0.00021 -0.00032 -0.0011 -0.00081

95Mo (42095.70c) -0.00115 -0.00116 -0.00204 -0.0016

99Tc (43099.70c) -0.00253 -0.0025 -0.00403 -0.00351

101Ru (44101.70c) -0.00063 -0.00075 -0.00152 -0.0012

103Rh (45103.70c) -0.00677 -0.00664 -0.01045 -0.00929

109Ag (47109.70c) -0.00084 -0.00085 -0.00182 -0.0014

133Cs (55133.70c) -0.00332 -0.00323 -0.00536 -0.00481

143Nd (60143.70c) -0.00888 -0.00885 -0.01331 -0.0125

145Nd (60145.70c) -0.0018 -0.00166 -0.00297 -0.00255

147Sm (62147.70c) -0.00143 -0.00185 -0.00192 -0.00233

149Sm (62149.70c) -0.01558 -0.01531 -0.01649 -0.01559

150Sm (62150.70c) -0.00078 -0.0008 -0.00164 -0.00129

151Sm (62151.70c) -0.00584 -0.00431 -0.00756 -0.00514

152Sm (62152.70c) -0.00211 -0.00211 -0.0033 -0.00272

151Eu (63151.70c) -0.00027 -0.00142 -0.00023 -0.00142

153Eu (63153.70c) -0.00152 -0.00149 -0.00294 -0.00261

155Gd (64155.70c) -0.00352 -0.00681 -0.00727 -0.01304

Note: MCNP = Monte Carlo N-Particle. a Monte Carlo uncertainty is less than 0.00013 ∆k for FP&MA worths.
C-11

Table C.11. FP&MA worths using MCNP6 (new release) and ENDF/B–VII.1 data

Initial enrichment and 

final burnup

3.1 wt % and 20 GWd/MTU 4.7 wt % and 40 GWd/MTU

Cooling time

(years) 5 40 5 40

FP&MA worths

(∆k)a

All FP&MA -0.06663 -0.0684 -0.10099 -0.10268

All MA -0.00629 -0.00615 -0.01182 -0.01101

All FP -0.05954 -0.06127 -0.08662 -0.08883

236U (92236.80c) -0.00391 -0.0037 -0.00576 -0.00503

237Np (93237.80c) -0.00208 -0.00237 -0.00456 -0.00468

243Am (95243.80c) -0.00022 -0.00025 -0.00117 -0.00098

95Mo (42095.80c) -0.00094 -0.00095 -0.0018 -0.00149

99Tc (43099.80c) -0.00211 -0.00214 -0.00367 -0.0033

101Ru (44101.80c) -0.00086 -0.00068 -0.00132 -0.00125

103Rh (45103.80c) -0.00645 -0.00629 -0.00976 -0.00878

109Ag (47109.80c) -0.001 -0.0007 -0.0016 -0.00135

133Cs (55133.80c) -0.00328 -0.00306 -0.00535 -0.00475

143Nd (60143.80c) -0.00887 -0.00854 -0.01338 -0.01256

145Nd (60145.80c) -0.00153 -0.00144 -0.00254 -0.00235

147Sm (62147.80c) -0.00109 -0.00177 -0.00193 -0.00228

149Sm (62149.80c) -0.01568 -0.01515 -0.01648 -0.0152

150Sm (62150.80c) -0.00085 -0.00073 -0.00152 -0.00135

151Sm (62151.80c) -0.00597 -0.00451 -0.00735 -0.00519

152Sm (62152.80c) -0.00198 -0.00195 -0.00303 -0.00279

151Eu (63151.80c) -0.00015 -0.00117 -0.00025 -0.00143

153Eu (63153.80c) -0.00158 -0.00166 -0.00305 -0.00277

155Gd (64155.80c) -0.00352 -0.00669 -0.00724 -0.01316

Note: MCNP = Monte Carlo N-Particle. a Monte Carlo uncertainty is less than 0.00013 ∆k for FP&MA worths.





NUREG/CR-7205 Bias Estimates Used in Lieu of Validation of Fission Products and Minor 

Actinides in MCNP Keff Calculations for PWR Burnup Credit Casks