Production and Testing of 

the VITAMIN-B7 Fine-Group 

and BUGLE-B7 Broad-Group 

Coupled Neutron/Gamma 

Cross-Section Libraries 

Derived from ENDF/B-VII.0

Nuclear Data

Office of Nuclear Reactor Regulation

NUREG/CR-7045

ORNL/TM-2011/12
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Production and Testing of 

the VITAMIN-B7 Fine-Group 

and BUGLE-B7 Broad-Group 

Coupled Neutron/Gamma 

Cross-Section Libraries 

Derived from ENDF/B-VII.0 

Nuclear Data

Manuscript Completed: January 2011

Date Published: September 2011

Prepared by

J. M. Risner, D. Wiarda, M. E. Dunn, T. M. Miller

D. E. Peplow, and B. W. Patton

Oak Ridge National Laboratory

Managed by UT-Battelle, LLC

Oak Ridge, TN 37831-6170

Bernard L. Grenier, NRC Project Manager

NRC Job Code J4234

Office of Nuclear Reactor Regulation

NUREG/CR-7045

ORNL/TM-2011/12

iii 

ABSTRACT 

New coupled neutron-gamma cross-section libraries have been developed for use in light water 

reactor (LWR) shielding applications, including pressure vessel dosimetry calculations. The 

libraries, which were generated using Evaluated Nuclear Data File/B Version VII Release 0 

(ENDF/B-VII.0), use the same fine-group and broad-group energy structures as the 

VITAMIN-B6 and BUGLE-96 libraries. The processing methodology used to generate both 

libraries is based on the methods used to develop VITAMIN-B6 and BUGLE-96 and is 

consistent with ANSI/ANS 6.1.2. The ENDF data were first processed into the fine-group 

pseudo-problem-independent VITAMIN-B7 library and then collapsed into the broad-group 

BUGLE-B7 library. The VITAMIN-B7 library contains data for 391 nuclides. This represents a 

significant increase compared to the VITAMIN-B6 library, which contained data for 120 nuclides. 

The BUGLE-B7 library contains data for the same nuclides as BUGLE-96, and maintains the 

same numeric IDs for those nuclides. The broad-group data includes nuclides which are 

infinitely dilute and group collapsed using a concrete weighting spectrum, as well as nuclides 

which are self-shielded and group collapsed using weighting spectra representative of important 

regions of LWRs. The verification and validation of the new libraries includes a set of critical 

benchmark experiments, a set of regression tests that are used to evaluate multigroup crosssection libraries in the SCALE code system, and three pressure vessel dosimetry benchmarks. 

Results of these tests confirm that the new libraries are appropriate for use in LWR shielding 

analyses and meet the requirements of Regulatory Guide 1.190. 

v 

CONTENTS 

Page 

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

CONTENTS .................................................................................................................................. v

LIST OF FIGURES ...................................................................................................................... ix

LIST OF TABLES ......................................................................................................................... xi

ACKNOWLEDGMENTS ............................................................................................................. xiii

LIST OF ACRONYMS AND UNITS ............................................................................................ xiv

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

1.1 Background ....................................................................................................................... 1

1.1.1 VITAMIN Libraries ................................................................................................. 1

1.1.2 BUGLE Libraries ................................................................................................... 2

1.2 Evaluated Nuclear Data Files ............................................................................................ 3

1.3 Cross-Section Processing and Testing ............................................................................. 3

2 FINE-GROUP LIBRARY SPECIFICATIONS ........................................................................... 5

2.1 Name ................................................................................................................................. 5

2.2 Materials, Temperatures, and Background Cross Sections .............................................. 5

2.3 Energy Group Structure .................................................................................................... 6

2.4 Weighting Function ........................................................................................................... 6

2.5 Legendre Order of Scattering ............................................................................................ 7

2.6 Processing Codes and Procedures ................................................................................. 31

2.6.1 AMPX Execution Sequences for Neutron and Gamma-Ray Yield Processing ... 31

2.6.1.1 Processing of Pointwise Cross-Section Data ......................................... 31

2.6.1.2 Neutron and Gamma Yield Processing .................................................. 33

2.6.1.3 Thermal Neutron Processing ................................................................. 34

2.6.1.4 Generation of Full-Range Bondarenko Factors and Thermal Data ........ 34

2.6.1.5 Gamma Processing ................................................................................ 35

2.6.1.6 Creating the Master Library .................................................................... 36

3 BROAD-GROUP LIBRARY SPECIFICATIONS ..................................................................... 37

3.1 Name ............................................................................................................................... 37

3.2 Materials and Energy Group Structure ............................................................................ 37

3.3 Self-Shielding and Group Collapsing of the Broad-Group Library................................... 37

3.3.1 Self-Shielding Calculations .................................................................................. 37
vi 

CONTENTS (continued) 

Page 

3.3.1.1 PWR Fuel Cell Self-Shielding ................................................................ 40

3.3.1.2 BWR Fuel Cell Self-Shielding ................................................................ 41

3.3.1.3 Coolant Self-Shielding ............................................................................ 42

3.3.1.4 Carbon Steel and Stainless Steel Self-Shielding ................................... 42

3.3.1.5 Concrete Self-Shielding ......................................................................... 43

3.3.2 Fine-Group Spectra Used for Group Collapsing .................................................. 43

3.3.2.1 BWR XSDRNPM Model for Weighting Spectra ..................................... 44

3.3.2.2 PWR XSDRNPM Model for Weighting Spectra ..................................... 45

3.4 Processing Codes and Procedures ................................................................................. 54

3.5 Library Format and Content ............................................................................................ 54

3.6 Response functions ......................................................................................................... 62

4 LIBRARY VERIFICATION AND VALIDATION ....................................................................... 65

4.1 Processing Methods ........................................................................................................ 65

4.2 Unit Tests ........................................................................................................................ 65

4.3 Critical Benchmark Experiments ..................................................................................... 66

4.3.1 Results for the Criticality Benchmarks for Thermal Systems ............................... 68

4.3.2 Results for the Criticality Benchmarks for Fast Systems ..................................... 68

4.4 SCALE Regression Tests ............................................................................................... 69

4.4.1 Neutron Transmission through an Iron Sphere ................................................... 70

4.4.2 Neutron and Photon Leakage Spectra from a 252Cf Source at the Centers of 

Six Iron Spheres of Different Diameters (ALARM-CF-FE-SHIELD-001) ............ 71

4.4.3 D-T Neutrons through an Iron Sphere ................................................................. 79

4.4.4 Neutron Transmission through a Heavy Water Sphere ....................................... 81

4.4.5 Ueki Shielding Measurements ............................................................................. 82

4.4.6 XSDRN Results ................................................................................................... 86

4.4.7 Summary of SCALE Regression Tests ............................................................... 86

4.5 Light Water Reactor Pressure Vessel Dosimetry Benchmarks ....................................... 87

4.5.1 H.B. Robinson Unit 2 Pressure Vessel Benchmark ............................................. 87

4.5.2 Oak Ridge National Laboratory Pool Critical Assembly (PCA) ............................ 88

4.5.2.1 Comparison of Calculated Reaction Rates with Experimental Results .. 89

4.5.3 CEN/SK Laboratory VENUS-3 Benchmark ......................................................... 92
vii 

CONTENTS (continued) 

Page

4.6 Conclusions from Verification and Validation Tests ...................................................... 105

5 REFERENCES ..................................................................................................................... 106

ix 

LIST OF FIGURES 

Page 

Figure 2.1. Neutron weighting spectrum used to create the fine-group VITAMIN-B7 

neutron library. ....................................................................................................... 29

Figure 2.2. Gamma weighting spectrum used to create the fine-group VITAMIN-B7 

gamma library. ....................................................................................................... 30

Figure 2.3. AMPX sequence for producing neutron multigroup data. ...................................... 32

Figure 2.4. AMPX sequence to generate coupled neutron–gamma multigroup data. ............. 32

Figure 3.1. One-dimensional BWR reactor plant model used to generate weighting 

spectra for group collapsing BUGLE-B7 cross sections from VITAMIN-B7 

data (from Ref. [1]). ................................................................................................ 44

Figure 3.2. One-dimensional PWR reactor plant model used to generate weighting 

spectra for group collapsing BUGLE-B7 cross sections from VITAMIN-B7 

data (from Ref. [1]). ................................................................................................ 45

Figure 3.3. Fine-group neutron weighting spectra from the 1D BWR and PWR models. ........ 46

Figure 3.4. Fine-group gamma weighting spectra from the 1D BWR and PWR models. ........ 52

Figure 3.5. Procedure for self-shielding nuclide data and calculating BWR- and 

PWR-specific flux spectra. ..................................................................................... 56

Figure 3.6. Procedure for collapsing self-shielded fine-group cross sections into the 

BUGLE-B7 library. ................................................................................................. 57

Figure 3.7. Procedure for generating infinitely dilute BUGLE-B7 data with the concrete 

weighting spectrum. ............................................................................................... 58

Figure 4.1. Comparison of the Monaco neutron calculation with two sets of experimental 

data for the iron sphere benchmark. ...................................................................... 71

Figure 4.2. Benchmark geometry for the iron sphere leakage measurements. ....................... 72

Figure 4.3. Comparison of the Monaco neutron spectrum with the measured spectrum 

from the bare source of the ALARM-CF-FE-SHIELD-001 experiment. ................. 73

Figure 4.4. Comparison of the Monaco photon spectrum with the measured spectrum 

from the bare source of the ALARM-CF-FE-SHIELD-001 experiment. ................. 73

Figure 4.5. Comparison of the Monaco neutron spectrum with the measurement for the 

20-cm-diameter shield from the ALARM-CF-FE-SHIELD-001 experiment. ........... 74

Figure 4.6. Comparison of the Monaco neutron spectrum with the measurement for the 

30-cm-diameter shield from the ALARM-CF-FE-SHIELD-001 experiment. ........... 74

Figure 4.7. Comparison of the Monaco neutron spectrum with the measurement for the 

40-cm-diameter shield from the ALARM-CF-FE-SHIELD-001 experiment. ........... 75

Figure 4.8. Comparison of the Monaco neutron spectrum with the measurement for the 

50-cm-diameter shield from the ALARM-CF-FE-SHIELD-001 experiment. ........... 75

Figure 4.9. Comparison of the Monaco neutron spectrum with the measurement for the 

60-cm-diameter shield from the ALARM-CF-FE-SHIELD-001 experiment. ........... 76
x 

LIST OF FIGURES (continued) 

Page 

Figure 4.10. Comparison of the Monaco and MCNP neutron spectra with the 

measurement for the 70-cm-diameter shield from the 

ALARM-CF-FE-SHIELD-001 experiment. ............................................................. 76

Figure 4.11. Comparison of the Monaco photon spectrum with the measurement for the 

30-cm-diameter shield from the ALARM-CF-FE-SHIELD-001 experiment. ........... 77

Figure 4.12. Comparison of the Monaco photon spectrum with the measurement for the 

40-cm-diameter shield from the ALARM-CF-FE-SHIELD-001 experiment. ........... 77

Figure 4.13. Comparison of the Monaco photon spectrum with the measurement for the 

50-cm-diameter shield from the ALARM-CF-FE-SHIELD-001 experiment. ........... 78

Figure 4.14. Comparison of the Monaco photon spectrum with the measurement for the 

60-cm-diameter shield from the ALARM-CF-FE-SHIELD-001 experiment. ........... 78

Figure 4.15. Comparison of the Monaco and MCNP photon spectra with the 

measurement for the 70-cm-diameter shield from the 

ALARM-CF-FE-SHIELD-001 experiment. ............................................................. 79

Figure 4.16. Geometry of the Urbana D-T transmission experiment. ........................................ 79

Figure 4.17. Comparison of the Monaco neutron spectrum at 200 cm with the 

measured data for the Urbana experiment. ........................................................... 80

Figure 4.18. Geometry of the heavy water experiment. ............................................................. 81

Figure 4.19. Comparison of the Monaco calculation and measured data for the 

heavy water transmission benchmark. .................................................................. 82

Figure 4.20. Geometry of a type 3 configuration from the neutron shielding experiments 

of Ueki et al. Half of the paraffin block surrounding the source has been 

removed from the model to show the source and the conical cutout. .................... 83

Figure 4.21. MAVRIC C/E ratios for the neutron dose rates from the Ueki type 1 

experiments for single material shields of a given thickness. ................................ 85

Figure 4.22. MAVRIC C/E ratios for the neutron dose rates from the type 2a experiments, 

which consist of 25 cm of SS-304 followed by T cm of polyethylene..................... 85

Figure 4.23. MAVRIC C/E ratios for the neutron dose rates from the type 2b experiments, 

which consist of T cm of polyethylene followed by 25 cm of SS-304. .................... 85

Figure 4.24. MAVRIC C/E ratios for the neutron dose rates from the type 3 experiments, 

which consist of t cm of SS-304, 15 cm ofpolyethylene, 

and (25-t) cm of SS-304. ....................................................................................... 86

Figure 4.25. Close-up of the SCALE model of the PCA core. ................................................... 90

Figure 4.26. SCALE model of the pressure vessel wall benchmark facility (water omitted). ..... 90

Figure 4.27. Southwest quadrant of the VENUS-3 benchmark geometry as modeled in the 

SGGP (water omitted). .......................................................................................... 93
xi 

LIST OF TABLES 

Page 

Table 2.1. Nuclide data in the VITAMIN-B6 and VITAMIN-B7 libraries ..................................... 8

Table 2.2. Background cross-section values at which Bondarenko factors are 

tabulated in the Vitamin-B7 library ......................................................................... 18

Table 2.3. Neutron group energy boundaries for the VITAMIN-B7 library .............................. 27

Table 2.4. Photon group energy boundaries for the VITAMIN-B7 library ................................ 28

Table 3.1. Neutron group energy boundaries for the BUGLE-B7 library ................................. 38

Table 3.2. Gamma group energy boundaries for the BUGLE-B7 library ................................. 39

Table 3.3. Description of mixtures used for self-shielding nuclides in the PWR 

pin cell model ......................................................................................................... 40

Table 3.4. Description of mixtures used for self-shielding nuclides in the BWR 

pin cell model ......................................................................................................... 41

Table 3.5. Description of mixtures used for self-shielding nuclides in the coolant model ....... 42

Table 3.6. Description of mixtures used for self-shielding nuclides in the steel models .......... 42

Table 3.7. Description of mixtures used for self-shielding nuclides in the concrete model ..... 43

Table 3.8. Description of the homogenized core and coolant mixtures used in the 

1D BWR reactor plant model .................................................................................. 44

Table 3.9. Description of the homogenized core and coolant mixtures used in the 

1D PWR reactor plant model .................................................................................. 45

Table 3.10. Fine-group neutron weighting spectra from the 1D BWR and PWR models .......... 47

Table 3.11. Fine-group gamma weighting spectra from the 1D BWR and PWR models .......... 53

Table 3.12. AMPX modules used to self-shield and group collapse fine-group 

cross sections into the BUGLE-B7 library .............................................................. 55

Table 3.13. Nuclides in BUGLE-B7 that are infinitely dilute and collapsed with the 

concrete weighting spectrum .................................................................................. 59

Table 3.14. Nuclides in BUGLE-B7 that are self-shielded and collapsed using specific 

BWR and PWR weighting spectra .......................................................................... 60

Table 3.15. Neutron response functions included with the BUGLE-B7 ANISN-formatted 

library ...................................................................................................................... 63

Table 3.16. Row positions for neutron kerma factors with ANISN ID 8001 and gamma-ray 

kerma factors with ANISN ID 8003 ......................................................................... 64

Table 3.17. Row positions for neutron kerma factors with ANISN ID 8002 and gamma-ray 

kerma factors with ANISN ID 8004 ......................................................................... 64

Table 4.1. Critical benchmark experiments used for validation of the VITAMIN-B7 library .... 66

Table 4.2. Calculated eigenvalues for the thermal system critical benchmarks used for 

validation of the VITAMIN-B7 library ...................................................................... 68
xii 

LIST OF TABLES (continued) 

Page 

Table 4.3. Calculated eigenvalues for the fast system critical benchmarks used for 

validation of the VITAMIN-B7 library ...................................................................... 69

Table 4.4. Summary of the SCALE regression test problems used for verification and 

validation of the VITAMIN-B7 library ...................................................................... 70

Table 4.5. Summary of MAVRIC results for the Ueki type 1 experiments with a 

252Cf neutron source ............................................................................................... 84

Table 4.6. Calculated-to-experiment ratios (C/E) for dosimetry activities from 

H. B. Robinson Unit 2, Cycle 9 ............................................................................... 87

Table 4.7. 237Np(n,f)137Cs results for the PVWBF benchmark analysis ................................... 91

Table 4.8. 238U(n,f)137Cs results for the PVWBF benchmark analysis ..................................... 91

Table 4.9. 103Rh(n,n’)103mRh results for the PVWBF benchmark analysis ............................... 91

Table 4.10. 115In(n,n’)115mIn results for the PVWBF benchmark analysis .................................. 91

Table 4.11. 58Ni(n,p)58Co results for the PVWBF benchmark analysis ...................................... 92

Table 4.12. 27Al(n,)

23Na results for the PVWBF benchmark analysis...................................... 92

Table 4.13. Calculated-to-Measured (C/M) Ratios for the VENUS-3 Benchmark 

Experiment 58Ni(n,p)58Co Equivalent Fission Fluxes using TORT with 

BUGLE-96 and BUGLE-B7 Cross-Sections ........................................................... 94

Table 4.14. Calculated-to-Measured (C/M) Ratios for the VENUS-3 Benchmark 

Experiment 115In(n,n’)115mIn Equivalent Fission Fluxes using TORT with 

BUGLE-96 and BUGLE-B7 Cross-Sections ......................................................... 100

Table 4.15. Calculated-to-Measured (C/M) Ratios for the VENUS-3 Benchmark 

Experiment 27Al(n,)

24Na Equivalent Fission Fluxes using TORT with 

BUGLE-96 and BUGLE-B7 Cross-Sections ......................................................... 103

 
xiii 

ACKNOWLEDGMENTS 

This work was performed under contract with the U.S. Nuclear Regulatory Commission (NRC) 

Office of Nuclear Reactor Regulation. The authors thank Bernard Grenier, the NRC Project 

Manager, and Benjamin Parks, NRC Technical Monitor, for their support and guidance. The 

authors also thank John Wagner and Cecil Parks of the Oak Ridge National Laboratory (ORNL) 

for their vision and efforts to establish the project and financial support for the work, along with 

the valuable technical discussions and guidance they provided on issues related to the 

VITAMIN/BUGLE libraries needed for pressure vessel fluence calculations, and Igor Remec and 

Mark Williams of ORNL for their technical review of the report and their helpful suggestions 

during the course of the work. Special appreciation is expressed to Hannah Turpin for 

assistance in preparing the final manuscript for publication.
xiv 

LIST OF ACRONYMS AND UNITS 

1D one-dimensional 

ANS American Nuclear Society 

BWR boiling water reactor 

CSEWG Cross-Section Evaluation Working Group 

DORT discrete-ordinates radiation transport 

ENDF/B-VII Version VII of the U.S. Evaluated Nuclear Data File 

HBR-2 H. B. Robinson Unit 2 

ICSBEP International Handbook of Evaluated Criticality Safety Benchmark 

Experiments 

IRDF 

LWR 

International Reactor Dosimetry File 

light water reactor 

MTR material test reactor 

NNDC National Nuclear Data Center 

ORNL Oak Ridge National Laboratory 

PCA Pool Critical Assembly 

PVWBF pressure vessel wall benchmark facility 

PWR pressurized water reactor 

SCALE Standardized Computer Analyses for Licensing Evaluation 

SGGP SCALE Generalized Geometry Package 

SNLRMLS Sandia National Laboratories Radiation Metrology Laboratory 

 

 

 

 

 

 

 

 
1 

1 INTRODUCTION 

This report describes the production and testing of revised multigroup cross-section libraries for 

light water reactor (LWR) shielding applications, including pressure vessel dosimetry. The new 

libraries, which are updates of the widely used VITAMIN-B6 and BUGLE-96 libraries [1], are 

based on Release 0 of ENDF/B-VII data [2]. The fine-group library, which is designated 

VITAMIN-B7, was prepared by processing ENDF/B-VII data into the group structure used in the 

VITAMIN-B6 library (199 neutron groups and 47 gamma groups). The fine-group library was 

then collapsed into the broad-group BUGLE-B7 library using the same group structure as 

BUGLE-96 (47 neutron groups and 20 gamma groups) with fine-group weighting spectra 

representative of various regions of pressurized water reactors (PWRs) and boiling water 

reactors (BWRs). 

VITAMIN-B7 and BUGLE-B7 are available in AMPX and ANISN formats, respectively, for use in 

a range of multigroup deterministic and stochastic radiation transport codes. Although these 

libraries were developed primarily for use in LWR shielding applications, they may potentially be 

used for other types of analyses provided they are validated in an appropriate manner. 

1.1 Background 

Multigroup cross-section libraries used in radiation transport calculations can be generally 

categorized as fine group or broad group. Fine-group libraries include a sufficient number of 

groups such that differences between the flux spectrum used to produce the groupwise cross 

sections and the actual flux spectra in a given application have a negligible effect on the 

multigroup data. Fine-group libraries typically contain hundreds of energy groups. Broad-group 

libraries, which typically contain tens of groups, are produced by collapsing fine-group data 

using flux spectra that closely approximate the spectra that are encountered in a particular 

application. A broad-group library that is developed using appropriate weighting spectra for a 

given application can produce calculated fluxes more quickly with little loss in accuracy 

compared to a fine-group calculation. 

As computing resources become more powerful, it is possible to use fine-group libraries more 

routinely. However, broad-group libraries are still useful, especially as two-dimensional models 

become more detailed and three-dimensional models are more commonly used. Furthermore, 

broad-group libraries are well-suited to parametric studies, where a large number of calculations 

may be run for a given model. 

1.1.1 VITAMIN Libraries 

The VITAMIN (Versatile Integrated Techniques using AMPX and MINX for Investigating 

Neutronics) multigroup cross-section library concept was originally developed to provide users 

of radiation transport codes with fine-group cross sections from which application-specific 

libraries (typically with broader energy group structures) could be derived. Temperature and 

resonance self-shielding corrections were provided for by including Bondarenko shielding 

factors in the library over a range of background cross sections. 

The first VITAMIN library, VITAMIN-C [3], was produced in 1978 based on ENDF/B-IV data. 

The MINX [4] code system (which was the predecessor to NJOY [5]) was used to produce the 

neutron transport cross sections, and the AMPX [6] code system was used to produce the 

photon production and transport cross sections. The VITAMIN-C library contains 171 neutron 
2 

groups and 36 gamma groups. While the primary applications for which VITAMIN-C was 

developed were fusion and liquid-metal fast breeder reactor neutronics studies, the library was 

also applied to LWR shielding and RPV dosimetry applications, shipping cask design, shielding 

benchmarks, and air transport analyses. VITAMIN-C was released through the Radiation 

Shielding Information Computational Center (RSICC, which at the time was known as RSIC) as 

DLC-041. 

The next VITAMIN library was VITAMIN-E [7], with the “E” designation representing the 

alphabetic equivalent of the fifth version of ENDF/B (i.e., ENDF/B-V), which was the primary 

source of evaluated data used to produce VITAMIN-E. Like VITAMIN-C, VITAMIN-E was 

generated using MINX and AMPX. The energy group structure was a slight refinement of the 

VITAMIN-C structure, with 174 neutron groups and 38 gamma groups. VITAMIN-E was 

released through RSIC as DLC-113 in 1987. 

The VITAMIN-B6 library, based primarily on ENDF/B-VI Release 3, was released through 

RSICC as DLC-184 in 1994. The ENDF data were processed with NJOY and converted to 

AMPX master library format with the AMPX SMILER module. The VITAMIN-B6 group structure 

has 199 neutron groups and 47 photon groups. The 199 neutron groups are based on a 

combination of the 175 groups in VITAMIN-J (a European library based on the VITAMIN-C and 

VITAMIN-E structures) and the 27 groups in the SCALE [8] shielding library, with the VITAMIN-J 

boundaries at higher energies used when the energy values differed slightly between the 

VITAMIN-J and SCALE libraries. The thermal energy range, which extends to 5.043 eV, 

contains 36 neutron groups with upscattering. The 47 photon groups are based on a 

combination of the 42 photon groups in VITAMIN-J and the 18 groups in the SCALE shielding 

library. 

1.1.2 BUGLE Libraries 

The first broad-group library in the BUGLE (Broad User Group Library ENDF/B) series was 

created based on methodology that was being developed by the ANS 6.1 Working Group in the 

late 1970s. The BUGLE library, which contains 45 neutron groups and 16 gamma groups, was 

derived from VITAMIN-C and released through RSIC as DLC-047. Initial experience with the 

BUGLE library revealed that the energy group structure and the weighting technique for 

collapsing the fine-group library to the broad-group library were inadequate. 

A new library, BUGLE-80 [9], was then developed from the same fine-group data. BUGLE-80 

increased the number of energy groups to 47 neutron groups and 20 gamma groups. Specific 

changes were made for neutron energies above 10 MeV, in the vicinity of the 2.3-MeV minimum 

in the oxygen cross section and near the 25-keV minimum in the iron cross section. In addition, 

groups were added in the energy region below 5 eV, but no upscattering transfers were 

included in the thermal range. BUGLE-80 was released through RSIC as DLC-075. Following 

the release of BUGLE-80, the SAILOR (Shielded and Application Independent Libraries for 

Operating Reactors) library [10] was developed based on the analysis of pressure vessel 

neutron fluence levels in specific LWRs. SAILOR used the same group structure as BUGLE-80 

but employed five distinct weighting spectra rather than the single energy spectrum weighting 

approach of BUGLE and BUGLE-80. SAILOR was released through RSIC as DLC-076, and 

was widely used by industry for RPV fluence calculations. 

The BUGLE-96 library was developed from VITAMIN-B6 using the same multiple-weightingspectra technique that was applied in the SAILOR library. BUGLE-96 also added cross-section 

sets having upscatter data for four thermal groups. The addition of the upscatter data was 
3 

intended to improve the application of BUGLE-96 for problems which require a more accurate 

calculation of thermal fluxes, including thermal foils in RPV surveillance capsules. BUGLE-96 

also added new dosimetry response functions and kerma factors. 

1.2 Evaluated Nuclear Data Files 

Version VII of the U.S. Evaluated Nuclear Data File (ENDF/B-VII) was released in December 

2006. Reference [2] lists the principal improvements in ENDF/B-VII relative ENDF/B-VI. 

Among those improvements, the following are potentially significant with respect to cross 

sections used for LWR shielding applications: 

1. New evaluations for U, Pu, Th, Np, and Am 

2. More precise standard neutron reaction cross sections for H, 6

Li, 10B, and Au, and more 

precise fission cross sections for 235U and 238U 

3. Improved thermal neutron scattering 

4. An extensive set of neutron cross sections for fission products 

5. New light nucleus neutron reactions 

Along with these changes to the evaluated data, the formatting for the gamma yield data has 

changed substantially for many newer evaluations, with data being moved from Files 12 and 13 

to File 6. 

Although these changes are not expected to be as significant for LWR shielding applications as 

were the improvements to the iron evaluation in ENDF/B-VI, the availability of multigroup 

libraries based on ENDF/B-VII provides users of multigroup transport codes the most up-to-date 

data and enables adherence to Regulatory Guide 1.190 [11]. 

1.3 Cross-Section Processing and Testing 

The methodology used to create and test the VITAMIN-B7 and BUGLE-B7 cross-section 

libraries is consistent with ANSI/ANS-6.1.2 [12]. VITAMIN-B7 and BUGLE-B7 represent the first 

of the VITAMIN/BUGLE library series to be processed entirely (with the exception of neutron 

and gamma kerma factors) with the AMPX code system [13]. A detailed description of the 

processing of the fine-group library is provided in Chapter 2. Details of the processing of the 

broad-group BUGLE-B7 data from the fine-group library are provided in Chapter 3. Finally, the 

verification and validation studies that were completed for the new libraries are described in 

Chapter 4. 

 

5 

2 FINE-GROUP LIBRARY SPECIFICATIONS 

2.1 Name 

The fine-group pseudo-problem-independent library is designated as VITAMIN-B7. This 

continues the “VITAMIN” naming convention, which started with the ENDF/B-IV-based 

VITAMIN-C library and continued with the ENDF/B-V-based VITAMIN-E library and ENDF/B-VIbased VITAMIN-B6 library. 

2.2 Materials, Temperatures, and Background Cross 

Sections 

A set of 391 nuclides containing fast neutron data and 18 thermal evaluations were processed 

for the VITAMIN-B7 library. These nuclides are listed in Table 2.1 along with the corresponding 

ENDF MAT numbers and SCALE library IDs. For thermal evaluations the corresponding values 

for the coupled fast evaluations are also listed. The final library uses the SCALE IDs to be 

compatible with the SCALE code system. Note that although all the nuclides listed contain 

neutron interaction and photon interaction data, some evaluations do not contain gamma-ray 

production data. 

All data are based on ENDF formatted files in the ENDF/B-VII.0 library downloaded from the 

National Nuclear Data Center at Brookhaven National Laboratory. Table 2.1 indicates those 

nuclides that are contained in the VITAMIN-B6 library as well as those in VITAMIN-B7. It should 

be noted that ENDF/B-VII.0 provides isotopic data for some elements for which previous 

ENDF/B releases included only elemental data. For example, the new VITAMIN-B7 library does 

not include elemental titanium (which was present in ENDF/B-VI), but does include 46Ti, 47Ti, 

48Ti, 49Ti, and 50Ti. 

The Bondarenko (f-factor) method is used for resonance self-shielding and temperature effects. 

Bondarenko factors are produced for evaluations that have resolved and/or unresolved 

resonance parameters, unless the ENDF evaluation contains isotopic data in File 2 but only 

elemental data in File 3. For ENDF/B-VII, the only elements for which that combination occurs 

are vanadium (Z = 23) and zinc (Z = 30). 

All materials for which Bondarenko factors are produced were processed at temperatures of 

300, 600, 1000, and 2100 K. In most cases, background cross sections (which are typically 

referred to as 0) of 1010, 106

, 105, 104, 1000, 300, 100, 50, 10, and 1 barns were used. In some 

cases the cross sections calculated for low values of the background cross section can become 

negative. (See the discussion of the AMPX FABULOUS module in Section 2.6.1.4.) In these 

cases, the affected values (e.g., 147Pm with a background cross section 0 of 1 b) have been 

eliminated from the master library. The available values for the background cross section for 

each nuclide are listed in Table 2.2. 

In the case of thermal moderators with S() data, scattering matrices were produced at all 

temperatures given in the ENDF/B-VII.0 files. 
6 

2.3 Energy Group Structure 

The energy group structure of the VITAMIN-B7 library is identical to that of the VITAMIN-B6 

library, with 199 neutron groups and 42 gamma groups. 

The neutron energy group structure has an upper energy limit of 19.64 MeV. The thermal 

neutron energy range (i.e., the range of groups that include upscatter) has an upper energy 

boundary of 5.043 eV and includes 36 groups. The neutron groups typically have uniform 

lethargy widths ranging from 0.025 to 0.25 for energies above 1.445 eV, with additional 

boundaries to resolve resonance minima that are important for shielding calculations. 

The gamma energy group structure has an upper energy limit of 30 MeV. The energy range 

from 14 MeV to 30 MeV is covered by two groups. From 14 MeV to 8 MeV, the group widths are 

uniform at 2 MeV. From 8 MeV to 1 MeV, there are group boundaries every 500 keV, with 

some additional boundaries in the vicinity of important source energies. Below 1 MeV, the 

group widths decrease in a nearly monotonic fashion from 200 keV to 10 keV. The most narrow 

photon group is a 2-keV group that spans the positron annihilation photon energy of 511 keV. 

The group boundaries for the neutron and gamma energy group structures are given in 

Table 2.3 and Table 2.4, respectively. 

Note that additional problem-dependent libraries can be easily derived from VITAMIN-B7 

without having to repeat the fine-group averaging from the ENDF files. 

2.4 Weighting Function 

The weighting function used to create the fine-group neutron library is a typical combination of a 

fission spectrum, a 1/E slowing-down spectrum, and a Maxwellian thermal spectrum. The 

specific functional forms and energy ranges of these three constituents of the neutron weighting 

spectrum are given by the following: 

Energy range VITAMIN-B7 groups Functional form of weighting spectrum

820.8 keV–20.0 MeV 1–66

ܹሺܧሻ ൌ ܥଵ√ܧ݁ିா/ఏ

C1 = 2.5625 MeV-1.5 

 = 1.273 MeV 

0.125 eV–820.8 keV 67–87 ܹሺܧሻ ൌ 1/ܧ

10-5 eV–0.125 eV 187–199

ܹሺܧሻ ൌ ܥଷܧ݁ିா/௞்

C3 = 9498.4 eV-2 

kT = 0.025 eV 

The neutron weighting function is shown in Figure 2.1. 

The weighting spectrum used to create the fine-group photon library consists of a 1/E spectrum 

for energies between 100 keV and 10 MeV, with “roll-offs” below 100 keV to represent the effect 

of photoelectric absorption and above 10 MeV to represent the effect of pair production. The 

gamma weighting function is shown in Figure 2.2. 
7 

2.5 Legendre Order of Scattering 

The scattering cross sections for both neutrons and photons are provided as Legendre 

polynomial expansions with P7 order for nuclides with Z values up to 30 (Zn), and P5 for Z

values above 30. These expansion orders are consistent with the VITAMIN-B6 library. The 

lower expansion order for higher-Z nuclides is reasonable given the trend to reduced scattering 

anisotropy as Z values increase. Dosimetry calculations using discrete-ordinates transport 

codes often employ P5 or P3 scattering expansion. Thermal neutron scattering expansions are 

P7 for moderators with S() data, but are limited to P3 expansions for nuclides that use the 

free gas model.

 
8 

Table 2.1. Nuclide data in the VITAMIN-B6 and VITAMIN-B7 libraries

Z Nuclidea VITAMIN-B6 

nuclide ID 

VITAMIN-B7 

nuclide ID MATb

 Scale ID Gamma-ray 

productionc

1 H-1 h1 125 1801 X 

 H-1 (H2O) h1(h2o) h1 / h_h2o 125 / 1 1001 X 

H-1 (CH2) h1(ch2) h1 / h_ch2 125 / 37 1901 X 

H-1 (ZrH) h1 / h_zrh 125 / 7 1701 X 

H-1 (benzine) h1 / benzine 125 / 40 1601 X 

H-1 (liq CH4) h1 / l_ch4 125 / 33 1101 X 

H-1 (solid CH4) h1 / s_ch4 125 / 34 1201 X 

H-1 (ortho H) h1 / ortho_h 125 / 3 1401 X 

H-1 (para H) h1 / para_h 125 / 2 1501 X 

H-2 h2 128 1802 X 

H-2 (D2O) h2(d2o) h2 / d_d2o 128 / 11 1002 X 

H-2 (ortho D) h2 / ortho_d 128 / 13 1402 X 

H-2 (para D) h2 / para_d 128 / 12 1502 X 

H-3 h3 h3 131 1003 

2 He-3 he3 he3 225 2003 

 He-4 he4 he4 228 2004 

3 Li-6 li6 li6 325 3006 X 

 Li-7 li7 li7 328 3007 X 

4 Be-7 be7 be7 419 4007 

 Be-9 be9 be9 425 4009 X 

Be-9 (Be 

metal) 

be9(th) be9 / be_metal 425 / 26 4309 X 

Be-9 (BeO) be9 / be_beo 425 / 27 4509 X 

5 B-10 b10 b10 525 5010 X 

 B-11 b11 b11 528 5011 X 

6 C c 600 6012 X 

 C c 600 6000 X 

 C (benzine) c / benzine 600 / 40 6500 X 

 C (graphite) c (gph) c / graphite 600 / 31 6312 X 

7 N-14 n14 n14 725 7014 X 

 N-15 n15 n15 728 7015 X 

8 O-16 o16 o16 825 8016 X 

 O-16 (BeO) o16 / o_beo 825 / 28 8516 X 

 O-17 o17 o17 828 8017 

9 F-19 f19 f19 925 9019 X 

11 Na-22 na22 1122 11022 

 Na-23 na23 na23 1125 11023 X 

12 Mg mg 1200 X 

Mg-24 mg24 1225 12025 X 

Mg-25 mg25 1228 12025 X 

13 Al-27 al27 al27 1325 13027 X 

 Al-27 (Al-27) al27 / al27 1325 / 45 13701 X 

14 Si si 1400 14000 

 Si-28 si28 1425 14028 X 

 Si-29 si29 1428 14029 X 

Si-30 si30 1431 14030 
9 

Table 2.1. Nuclide data in the VITAMIN-B6 and VITAMIN-B7 libraries

Z Nuclidea VITAMIN-B6 

nuclide ID 

VITAMIN-B7 

nuclide ID MATb

 Scale ID Gamma-ray 

productionc

15 P-31 p31 p31 1525 15031 X 

16 S s 1600 16000 X 

 S-32 s32 s32 1625 16032 X 

 S-33 s33 1628 16033 X 

S-34 s34 1631 16034 X 

S-36 s36 1637 16036 X 

17 Cl cl (nat) 1700 17000 X 

 Cl-35 cl35 1725 17035 X 

 Cl-37 cl37 1731 17037 X 

18 Ar-36 ar36 1825 18036 

 Ar-38 ar38 1831 18038 

Ar-40 ar40 1837 18040 

19 K k 1900 19000 X 

 K-39 k39 1925 19039 X 

 K-40 k40 1928 19040 X 

 K-41 k41 1931 19041 X 

20 Ca ca 2000 20000 X 

 Ca-40 ca40 2025 20040 X 

 Ca-42 ca42 2031 20042 X 

Ca-43 ca43 2034 20043 X 

Ca-44 ca44 2037 20044 X 

Ca-46 ca46 2043 20046 X 

Ca-48 ca48 2049 20048 X 

21 Sc-45 sc45 2125 21045 X 

22 Ti ti 2200 22000 X 

 Ti-46 ti46 2225 22046 X 

 Ti-47 ti47 2228 22047 X 

Ti-48 ti48 2231 22048 X 

Ti-49 ti49 2234 22049 X 

Ti-50 ti50 2237 22050 X 

23 V v v 2300 23000 X 

24 Cr-50 cr50 cr50 2425 24050 X 

 Cr-52 cr52 cr52 2431 24052 X 

Cr-53 cr53 cr53 2434 24053 X 

Cr-54 cr54 cr54 2437 24054 X 

25 Mn-55 mn55 mn55 2525 25055 X 

26 Fe-54 fe54 fe54 2625 26054 X 

 Fe-56 fe56 fe56 2631 26056 X 

Fe-56 (Fe-56) fe56 / fe56 2631 / 56 26701 X 

Fe-57 fe57 fe57 2634 26057 X 

Fe-58 fe58 fe58 2637 26058 X 

27 Co-58m co58m 2723 27601 

 
10 

Table 2.1. Nuclide data in the VITAMIN-B6 and VITAMIN-B7 libraries

Z Nuclidea VITAMIN-B6 

nuclide ID 

VITAMIN-B7 

nuclide ID MATb

 Scale ID Gamma-ray 

productionc

28 Ni-58 ni58 ni58 2825 28058 X 

 Ni-59 ni59 2828 28059 X 

Ni-60 ni60 ni60 2831 28060 X 

Ni-61 ni61 ni61 2834 28061 X 

Ni-62 ni62 ni62 2837 28062 X 

Ni-64 ni64 ni64 2843 28064 X 

29 Cu63 cu63 cu63 2925 29063 X 

 Cu65 cu65 cu65 2931 29065 X 

30 Zn zn 3000 30000 

31 Ga ga 3100 31000 X 

 Ga-69 ga69 

 Ga-71 ga71 3131 31071 

32 Ge-70 ge70 3225 32070 X 

 Ge-72 ge72 3231 32072 X 

Ge-73 ge73 3234 32073 X 

Ge-74 ge74 3237 32074 X 

Ge-76 ge76 3243 32076 X 

33 As-74 as74 3322 33074 X 

 As-75 as75 3325 33075 X 

34 Se-74 se74 3425 34074 

 Se-76 se76 3431 34076 

Se-77 se77 3434 34077 

Se-78 se78 3437 34078 

Se-79 se79 3440 34079 

Se-80 se80 3443 34080 

Se-82 se82 3449 34082 

35 Br-79 br79 3525 35079 

 Br-81 br81 3531 35081 

36 Kr-78 kr78 3625 36078 

 Kr-80 kr80 3631 36080 

Kr-82 kr82 3637 36082 

Kr-83 kr83 3640 36083 

Kr-84 kr84 3643 36084 

Kr-85 kr85 3646 36085 X 

Kr-86 kr86 3649 36086 

37 Rb-85 rb85 3725 37085 

 Rb-86 rb86 3728 37086 X 

Rb-87 rb87 3731 37087 

38 Sr-84 sr84 3825 38084 X 

 Sr-86 sr86 3831 38086 

Sr-87 sr87 3834 38087 

Sr-88 sr88 3837 38088 

Sr-89 sr89 3840 38089 

Sr-90 sr90 3843 38090 
11 

Table 2.1. Nuclide data in the VITAMIN-B6 and VITAMIN-B7 libraries

Z Nuclidea VITAMIN-B6 

nuclide ID 

VITAMIN-B7 

nuclide ID MATb

 Scale ID Gamma-ray 

productionc

39 Y-89 y89 y89 3925 39089 X 

 Y-90 y90 3928 39090 X 

Y-91 y91 3931 39091 

40 Zr zr 4000 40000 

 Zr-90 zr90 4025 40090 X 

 Zr-90 (ZrH) zr90 / zr_zrh 4025 / 58 40790 X 

Zr-91 zr91 4028 40091 X 

Zr-91 (ZrH) zr91 / zr_zrh 4028 / 58 40790 X 

Zr-92 zr92 4031 40092 X 

Zr-92 (ZrH) zr92 / zr_zrh 4031 / 58 40790 X 

Zr-93 zr93 4034 40093 

Zr-93 (ZrH) zr93 / zr_zrh 4034 / 58 40790 

Zr-94 zr94 4037 40094 X 

Zr-94 (ZrH) zr94 / zr_zrh 4037 / 58 40790 X 

Zr-95 zr95 4040 40095 

Zr-95 (ZrH) zr95 / zr_zrh 4040 / 58 40790 

Zr-96 zr96 4043 40096 X 

Zr-96 (ZrH) zr96 / zr_zrh 4043 / 58 40790 X 

41 Nb-93 nb93 nb93 4125 41093 X 

 Nb-94 nb94 4128 41094 

Nb-95 nb95 4131 41095 

42 Mo mo 4200 42000 X 

 Mo-92 mo92 4225 42092 X 

 Mo-94 mo94 4231 42094 X 

 Mo-95 mo95 4234 42095 X 

Mo-96 mo96 4237 42096 X 

Mo-97 mo97 4240 42097 X 

Mo-98 mo98 4243 42098 X 

Mo-99 mo99 4246 42099 

Mo-100 mo100 4249 42100 

43 Tc-99 tc99 4325 43099 X 

44 Ru-96 ru96 4425 44096 

 Ru-98 ru98 4431 44098 

Ru-99 ru99 4434 44099 

Ru-100 ru100 4437 44100 

Ru-101 ru101 4440 44101 X 

Ru-102 ru102 4443 44102 

Ru-103 ru103 4446 44103 

Ru-104 ru104 4449 44104 

Ru-105 ru105 4452 44105 

Ru-106 ru106 4455 44106 

45 Rh-103 rh103 4525 45103 X 

 Rh-105 rh105 4531 45105 

 

 
12 

Table 2.1. Nuclide data in the VITAMIN-B6 and VITAMIN-B7 libraries

Z Nuclidea VITAMIN-B6 

nuclide ID 

VITAMIN-B7 

nuclide ID MATb

 Scale ID Gamma-ray 

productionc

46 Pd-102 pd102 4625 46102 X 

 Pd-104 pd104 4631 46104 X 

Pd-105 pd105 4634 46105 X 

Pd-106 pd106 4637 46106 X 

Pd-107 pd107 4640 46107 

Pd-108 pd108 4643 46108 X 

Pd-110 pd110 4649 46110 X 

47 Ag-107 ag107 ag107 4725 47107 X 

 Ag-109 ag109 ag109 4731 47109 X 

Ag-110m ag110m 4735 47601 

Ag-111 ag111 4737 47111 X 

48 Cd cd (nat) 4800 48000 

 Cd-106 cd106 4825 48106 X 

 Cd-108 cd108 4831 48108 

Cd-110 cd110 4837 48110 

Cd-111 cd111 4840 48111 X 

Cd-112 cd112 4843 48112 

Cd-113 cd113 4846 48113 

Cd-114 cd114 4849 48114 

Cd-115m cd115m 4853 48601 X 

Cd-116 cd116 4855 48116 

49 In in (nat) 4900 49000 

 In-113 in113 4925 49113 

 In-115 in115 4931 49115 

50 Sn sn (nat) 5000 50000 

 Sn-112 sn112 5025 50112 

 Sn-113 sn113 5028 50113 X 

Sn-114 sn114 5031 50114 

Sn-115 sn115 5034 50115 

Sn-116 sn116 5037 50116 

Sn-117 sn117 5040 50117 

Sn-118 sn118 5043 50118 

Sn-119 sn119 5046 50119 

Sn-120 sn120 5049 50120 

Sn-122 sn122 5055 50122 

Sn-123 sn123 5058 50123 

Sn-124 sn124 5061 50124 

Sn-125 sn125 5064 50125 X 

Sn-126 sn126 5067 50126 

51 Sb-121 sb121 5125 51121 

 Sb-123 sb123 5131 51123 

Sb-124 sb124 5134 51124 

Sb-125 sb125 5137 51125 

Sb-126 sb126 5140 51126 X 

 

 
13 

Table 2.1. Nuclide data in the VITAMIN-B6 and VITAMIN-B7 libraries

Z Nuclidea VITAMIN-B6 

nuclide ID 

VITAMIN-B7 

nuclide ID MATb

 Scale ID Gamma-ray 

productionc

52 Te-120 te120 5225 52120 

 Te-122 te122 5231 52122 

Te-123 te123 5234 52123 

Te-124 te124 5237 52124 

Te-125 te125 5240 52125 

Te-126 te126 5243 52126 

Te-127m te127m 5247 52601 

Te-128 te128 5249 52128 

Te-129m te129m 5253 52611 

Te-130 te130 5255 52130 

Te-132 te132 5261 52132 X 

53 I-127 i127 5325 53127 X 

 I-129 i129 5331 53129 

I-130 i130 5334 53130 X 

I-131 i131 5337 53131 

I-135 i135 5349 53135 

54 Xe-123 xe123 5422 54123 

 Xe-124 xe124 5425 54124 

Xe-126 xe126 5431 54126 

Xe-128 xe128 5437 54128 

Xe-129 xe129 5440 54129 

Xe-130 xe130 5443 54130 

Xe-131 xe131 5446 54131 X 

Xe-132 xe132 5449 54132 

Xe-133 xe133 5452 54133 

Xe-134 xe134 5455 54134 

Xe-135 xe135 5458 54135 

Xe-136 xe136 5461 54136 

55 Cs-133 cs133 5525 55133 X 

 Cs-134 cs134 5528 55134 

Cs-135 cs135 5531 55135 

Cs-136 cs136 5534 55136 

Cs-137 cs137 5537 55137 

56 Ba-130 ba130 5625 56130 

Ba-132 ba132 5631 56132 

Ba-133 ba133 5634 56133 X 

Ba-134 ba134 5637 56134 

Ba-135 ba135 5640 56135 

Ba-136 ba136 5643 56136 

Ba-137 ba137 5646 56137 

Ba-138 ba138 ba138 5649 56138 

Ba-140 ba139 5655 56140 

57 La-138 la138 5725 57138 

 La-139 la139 5728 57139 

La-140 la140 5731 57140 X 

 
14 

Table 2.1. Nuclide data in the VITAMIN-B6 and VITAMIN-B7 libraries

Z Nuclidea VITAMIN-B6 

nuclide ID 

VITAMIN-B7 

nuclide ID MATb

 Scale ID Gamma-ray 

productionc

58 Ce-136 cd136 5825 58136 X 

 Ce-138 ce138 5831 58138 X 

Ce-139 ce139 5834 58139 X 

Ce-140 ce140 5837 58140 

Ce-141 ce141 5840 58141 

Ce-142 ce142 5843 58142 

Ce-143 ce143 5846 58143 X 

Ce-144 ce144 5849 58144 

59 Pr-141 pr141 5925 59141 X 

 Pr-142 pr142 5928 59142 X 

Pr-143 pr143 5931 59143 

60 Nd-142 nd142 6025 60142 X 

 Nd-143 nd143 6028 60143 X 

Nd-144 nd144 6031 60144 X 

Nd-145 nd145 6034 60145 X 

Nd-146 nd146 6037 60146 X 

Nd-147 nd147 6040 60147 X 

Nd-148 nd148 6043 60148 X 

Nd-150 nd150 6049 60150 X 

61 Pm-147 pm147 6149 61147 

 Pm-148 pm148 6152 61148 

Pm-148m pm148m 6153 61601 

Pm-149 pm149 6155 61149 

Pm-151 pm151 6161 61151 X 

62 Sm-144 sm144 6225 62144 X 

 Sm-147 sm147 6234 62147 X 

Sm-148 sm148 6237 62148 X 

Sm-149 sm149 6240 62149 X 

Sm-150 sm150 6243 62150 X 

Sm-151 sm151 6246 62151 X 

Sm-152 sm152 6249 62152 X 

Sm-153 sm153 6252 62153 X 

Sm-154 sm154 6255 62154 X 

63 Eu-151 eu151 eu151 6325 63151 

 Eu-152 eu152 eu152 6328 63152 

Eu-153 eu153 eu153 6331 63153 X 

Eu-154 eu154 eu154 6334 63154 

Eu-155 eu155 eu155 6337 63155 

Eu-156 eu156 6340 63156 

Eu-157 eu157 6343 63157 X 

64 Gd-152 gd152 6425 64152 X 

 Gd-153 gd153 6428 64153 X 

Gd-154 gd154 6431 64154 X 

Gd-155 gd155 6434 64155 X 

Gd-156 gd156 6437 64156 X 

Gd-157 gd157 6440 64157 X 

Gd-158 gd158 6443 64158 X 

Gd-160 gd159 6449 64160 X 
15 

Table 2.1. Nuclide data in the VITAMIN-B6 and VITAMIN-B7 libraries

Z Nuclidea VITAMIN-B6 

nuclide ID 

VITAMIN-B7 

nuclide ID MATb

 Scale ID Gamma-ray 

productionc

65 Tb-159 tb159 6525 65159 

 Tb-160 tb160 6528 65160 X 

66 Dy-156 dy156 6625 66156 X 

 Dy-158 dy158 6631 66158 X 

Dy-160 dy160 6637 66160 X 

Dy-161 dy161 6640 66161 X 

Dy-162 dy162 6643 66162 X 

Dy-163 dy163 6646 66163 X 

Dy-164 dy164 6649 66164 X 

67 Ho-165 ho165 6725 67165 X 

 Ho-166m ho166m 6729 67601 X 

68 Er-162 er162 6825 68162 X 

 Er-164 er164 6831 68164 X 

Er-166 er166 6837 68166 X 

Er-167 er167 6840 68167 X 

Er-168 er168 6843 68168 X 

Er-170 er170 6849 68170 X 

71 Lu-175 lu175 7125 71175 

 Lu-176 lu176 7128 71176 

72 Hf-174 hf174 hf174 7225 72174 

 Hf-176 hf176 hf176 7231 72176 

Hf-177 hf177 hf177 7234 72177 

Hf-178 hf178 hf178 7237 72178 

Hf-179 hf179 hf179 7240 72179 

Hf-180 hf180 hf180 7243 72180 

73 Ta-181 ta181 ta181 7328 73181 X 

 Ta-182 ta182 ta182 7331 73182 

74 W-182 w182 w182 7431 74182 X 

 W-183 w183 w183 7434 74183 X 

W-184 w184 w184 7437 74184 X 

W-186 w186 w186 7443 74186 X 

75 Re-185 re185 re185 7525 75185 

 Re-187 re187 re187 7531 75187 

77 Ir-191 ir191 7725 77191 X 

 Ir-193 ir193 7731 77193 X 

79 Au-197 au197 au197 7925 79197 X 

80 Hg-196 hg196 8025 80196 X 

 Hg-198 hg198 8031 80198 X 

Hg-199 hg199 8034 80199 X 

Hg-200 hg200 8037 80200 X 

Hg-201 hg201 8040 80201 X 

Hg-202 hg202 8043 80202 X 

Hg-204 hg204 8049 80204 X 

 
16 

Table 2.1. Nuclide data in the VITAMIN-B6 and VITAMIN-B7 libraries

Z Nuclidea VITAMIN-B6 

nuclide ID 

VITAMIN-B7 

nuclide ID MATb

 Scale ID Gamma-ray 

productionc

82 Pb-204 pb204 8225 82204 X 

 Pb-206 pb206 pb206 8231 82206 X 

Pb-207 pb207 pb207 8234 82207 X 

Pb-208 pb208 pb208 8237 82208 X 

83 Bi-209 bi209 bi209 8325 83209 X 

88 Ra-223 ra223 8825 88223 

 Ra-224 ra224 8828 88224 

Ra-225 ra225 8831 88225 

Ra-226 ra226 8834 88226 

89 Ac-225 ac225 8925 89225 

 Ac-226 ac226 8928 89226 

 Ac-227 ac227 8931 89227 

90 Th-227 th227 9025 90227 

 Th-228 th228 9028 90228 

Th-229 th229 9031 90229 

Th-230 th230 th230 9034 90230 

Th-232 th232 th232 9040 90232 X 

Th-233 th233 9043 90233 

Th-234 th234 9046 90234 

91 Pa-231 pa231 pa231 9131 91231 X 

 Pa-232 pa232 9134 91232 

Pa-233 pa233 pa233 9137 91233 X 

92 U-232 u232 u232 9219 92232 X 

 U-233 u233 u233 9222 92233 X 

U-234 u234 u234 9225 92234 X 

U-235 u235 u235 9228 92235 X 

U-236 u236 u236 9231 92236 X 

U-237 u237 u237 9234 92237 X 

U-238 u238 u238 9237 92238 X 

U-239 u239 9240 92239 X 

U-240 u240 9243 92240 X 

U-241 u241 9246 92241 X 

93 Np-235 np235 9340 93235 

 Np-236 np236 9343 93236 

Np-237 np237 np237 9346 93237 X 

Np-238 np238 np238 9349 93238 

Np-239 np239 np239 9352 93239 

94 Pu-236 pu236 pu236 9428 94236 

 Pu-237 pu237 pu237 9431 94237 

Pu-238 pu238 pu238 9434 94238 

Pu-239 pu239 pu239 9437 94239 X 

Pu-240 pu240 pu240 9440 94240 X 

Pu-241 pu241 pu241 9443 94241 X 

Pu-242 pu242 pu242 9446 94242 X 

Pu-243 pu243 pu243 9449 94243 X 

Pu-244 pu244 pu244 9452 94244 

Pu-246 pu246 9458 94246 

 
17 

Table 2.1. Nuclide data in the VITAMIN-B6 and VITAMIN-B7 libraries

Z Nuclidea VITAMIN-B6 

nuclide ID 

VITAMIN-B7 

nuclide ID MATb

 Scale ID Gamma-ray 

productionc

95 Am-241 am241 am241 9543 95241 X 

 Am-242 am242 am242 9546 95242 

Am-242m am242m am242m 9547 95601 

Am-243 am243 am243 9549 95243 X 

Am-244 am244 9552 95244 

Am-244m am244m 9553 95611 

96 Cm-241 cm241 cm241 9628 96241 

 Cm-242 cm242 cm242 9631 96242 X 

Cm-243 cm243 cm243 9634 96243 

Cm-244 cm244 cm244 9637 96244 

Cm-245 cm245 cm245 9640 96245 

Cm-246 cm246 cm246 9643 96246 

Cm-247 cm247 cm247 9646 96247 

Cm-248 cm248 cm248 9649 96248 X 

Cm-249 cm249 9652 96249 

Cm-250 cm250 9655 96250 

97 Bk-249 bk249 9752 97249 

 Bk-250 bk250 9755 97250 

98 Cf-249 cf249 9852 98249 

 Cf-250 cf250 9855 98250 X 

Cf-251 cf251 9858 98251 X 

Cf-252 cf252 9861 98252 X 

Cf-253 cf253 9864 98253 

Cf-254 cf254 9867 98254 

99 Es-253 es253 9913 99253 

 Es-254 es254 9914 99254 

Es-255 es255 9915 99255 

100 Fm-255 fm255 9936 100255 

a For nuclides with thermal moderator data, the nuclide ID for the fast evaluation is given, followed by 

the nuclide ID for the moderator. 

b For thermal moderators the ENDF MAT number for the moderator is given followed by the ENDF MAT 

number of the fast evaluation. c For nuclides with thermal moderator data, the nuclide ID for the fast evaluation is given, followed by 

the nuclide ID for the moderator. 

 
18 

Table 2.2. Background cross-section values at which Bondarenko factors are 

tabulated in the Vitamin-B7 library 

(All nuclides were processed at 300, 600, 1000 and 2100 K)

Nuclidea Background cross sections (barns) Legendre 

10 order 10 106 105 104 1000 300 100 50 10 1 

H-1 X X X X X X X X X X 7 

H-1 (H2O) X X X X X X X X X X 7 

H-1 (CH2) X X X X X X X X X X 7 

H-1 (ZrH) X X X X X X X X X X 7 

H-1 (benzine) X X X X X X X X X X 7 

H-1 (liq CH4) X X X X X X X X X X 7 

H-1 (solid CH4) X X X X X X X X X X 7 

H-1 (ortho H) X X X X X X X X X X 7 

H-1 (para H) X X X X X X X X X X 7 

H-2 X X X X X X X X X X 7 

H-2 (D2O) X X X X X X X X X X 7 

H-2 (ortho D) X X X X X X X X X X 7 

H-2 (para D) X X X X X X X X X X 7 

H-3 X X X X X X X X X X 7 

He-3 X X X X X X X X X X 7 

He-4 X X X X X X X X X X 7 

Li-6 X X X X X X X X X X 7 

Li-7 X X X X X X X X X X 7 

Be-7 X X X X X X X X X X 7 

Be-9 X X X X X X X X X X 7 

Be-9 (Be metal) X X X X X X X X X X 7 

Be-9 (BeO) X X X X X X X X X X 7 

B10 X X X X X X X X X X 7 

B11 X X X X X X X X X X 7 

C X X X X X X X X X X 7 

C (benzene) X X X X X X X X X X 7 

C (graphite) X X X X X X X X X X 7 

N-14 X X X X X X X X X X 7 

N-15 X X X X X X X X X X 7 

O-16 X X X X X X X X X X 7 

O-16 (BeO) X X X X X X X X X X 7 

O-17 X X X X X X X X X X 7 

F-19 X X X X X X X X X X 7 

Na-23 X X X X X X X X X X 7 

Mg-24 X X X X X X X X X X 7 

Mg-25 X X X X X X X X X X 7 

Mg-26 X X X X X X X X X X 7 

Al-27 X X X X X X X X X X 7 

Al-27 (Al-27) X X X X X X X X X X 7 

Si-28 X X X X X X X X X X 7 

Si-29 X X X X X X X X X X 7 

Si-30 X X X X X X X X X X 7 

P-31 X X X X X X X X X X 7 

S-32 X X X X X X X X X X 7 

S-33 X X X X X X X X X X 7 

S-34 X X X X X X X X X X 7 

S-36 X X X X X X X X X X 7 
19 

Table 2.2. Background cross-section values at which Bondarenko factors are 

tabulated in the Vitamin-B7 library 

(All nuclides were processed at 300, 600, 1000 and 2100 K)

Nuclidea Background cross sections (barns) Legendre 

10 order 10 106 105 104 1000 300 100 50 10 1 

Cl-35 X X X X X X X X X X 7 

Cl-37 X X X X X X X X X X 7 

Ar-36 X X X X X X X X X X 7 

Ar-38 X X X X X X X X X X 7 

Ar-40 X X X X X X X X X X 7 

K-39 X X X X X X X X X X 7 

K-40 X X X X X X X X X X 7 

K-41 X X X X X X X X X X 7 

Ca-40 X X X X X X X X X X 7 

Ca-42 X X X X X X X X X X 7 

Ca-43 X X X X X X X X X X 7 

Ca-44 X X X X X X X X X X 7 

Ca-46 X X X X X X X X X X 7 

Ca-48 X X X X X X X X X X 7 

Sc-45 X X X X X X X X X X 7 

Ti-46 X X X X X X X X X X 7 

Ti-47 X X X X X X X X X X 7 

Ti-48 X X X X X X X X X X 7 

Ti-49 X X X X X X X X X X 7 

Ti-50 X X X X X X X X X X 7 

V 7 

Cr-50 X X X X X X X X X X 7 

Cr-52 X X X X X X X X X X 7 

Cr-53 X X X X X X X X X X 7 

Cr-54 X X X X X X X X X X 7 

Mn-55 X X X X X X X X X X 7 

Fe-54 X X X X X X X X X X 7 

Fe-56 X X X X X X X X X X 7 

Fe-56 (Fe-56) X X X X X X X X X X 7 

Fe-57 X X X X X X X X X X 7 

Fe-58 X X X X X X X X X X 7 

Co-58m X X X X X X X X X X 7 

Co-59 X X X X X X X X X X 7 

Ni-58 X X X X X X X X X X 7 

Ni-59 X X X X X X X X X X 7 

Ni-60 X X X X X X X X X X 7 

Ni-61 X X X X X X X X X X 7 

Ni-62 X X X X X X X X X X 7 

Ni-64 X X X X X X X X X X 7 

Cu-63 X X X X X X X X X X 7 

Cu-65 X X X X X X X X X X 7 

Zn 7 

Ga-69 X X X X X X X X X X 5 

Ga-71 X X X X X X X X X X 5 

Ge-70 X X X X X X X X X X 5 

Ge-72 X X X X X X X X X X 5 

Ge-73 X X X X X X X X X X 5 
20 

Table 2.2. Background cross-section values at which Bondarenko factors are 

tabulated in the Vitamin-B7 library 

(All nuclides were processed at 300, 600, 1000 and 2100 K)

Nuclidea Background cross sections (barns) Legendre 

10 order 10 106 105 104 1000 300 100 50 10 1 

Ge-74 X X X X X X X X X X 5 

Ge-76 X X X X X X X X X X 5 

As-74 X X X X X X X X X X 5 

As-75 X X X X X X X X X X 5 

Se-74 X X X X X X X X X X 5 

Se-76 X X X X X X X X X X 5 

Se-77 X X X X X X X X X X 5 

Se-78 X X X X X X X X X X 5 

Se-79 X X X X X X X X X X 5 

Se-80 X X X X X X X X X X 5 

Se-82 X X X X X X X X X X 5 

Br-79 X X X X X X X X X X 5 

Br-81 X X X X X X X X X X 5 

Kr-78 X X X X X X X X X X 5 

Kr-80 X X X X X X X X X X 5 

Kr-82 X X X X X X X X X X 5 

Kr-83 X X X X X X X X X X 5 

Kr-84 X X X X X X X X X X 5 

Kr-85 X X X X X X X X X X 5 

Kr-86 X X X X X X X X X X 5 

Rb-85 X X X X X X X X X X 5 

Rb-86 X X X X X X X X X X 5 

Rb-87 X X X X X X X X X X 5 

Sr-84 X X X X X X X X X X 5 

Sr-86 X X X X X X X X X X 5 

Sr-87 X X X X X X X X X X 5 

Sr-88 X X X X X X X X X X 5 

Sr-89 X X X X X X X X X X 5 

Sr-90 X X X X X X X X X X 5 

Y-89 X X X X X X X X X X 5 

Y-90 X X X X X X X X X X 5 

Y-91 X X X X X X X X X X 5 

Zr-90 X X X X X X X X X X 5 

Zr-90 (ZrH) X X X X X X X X X X 5 

Zr-91 X X X X X X X X X X 5 

Zr-91 (ZrH) X X X X X X X X X X 5 

Zr-92 X X X X X X X X X X 5 

Zr-92 (ZrH) X X X X X X X X X X 5 

Zr-93 X X X X X X X X X X 5 

Zr-93 (ZrH) X X X X X X X X X X 5 

Zr-94 X X X X X X X X X X 5 

Zr-94 (ZrH) X X X X X X X X X X 5 

Zr-95 X X X X X X X X X X 5 

Zr-95 (ZrH) X X X X X X X X X X 5 

Zr-96 X X X X X X X X X X 5 

Zr-96 (ZrH) X X X X X X X X X X 5 

Nb-93 X X X X X X X X X X 5 
21 

Table 2.2. Background cross-section values at which Bondarenko factors are 

tabulated in the Vitamin-B7 library 

(All nuclides were processed at 300, 600, 1000 and 2100 K)

Nuclidea Background cross sections (barns) Legendre 

10 order 10 106 105 104 1000 300 100 50 10 1 

Nb-94 X X X X X X X X X X 5 

Nb-95 X X X X X X X X X X 5 

Mo-92 X X X X X X X X X X 5 

Mo-94 X X X X X X X X X X 5 

Mo-95 X X X X X X X X X X 5 

Mo-96 X X X X X X X X X X 5 

Mo-97 X X X X X X X X X X 5 

Mo-98 X X X X X X X X X X 5 

Mo-99 X X X X X X X X X X 5 

Mo-100 X X X X X X X X X X 5 

Tc-99 X X X X X X X X X X 5 

Ru-96 X X X X X X X X X X 5 

Ru-98 X X X X X X X X X X 5 

Ru-99 X X X X X X X X X X 5 

Ru-100 X X X X X X X X X X 5 

Ru-101 X X X X X X X X X X 5 

Ru-102 X X X X X X X X X X 5 

Ru-103 X X X X X X X X X X 5 

Ru-104 X X X X X X X X X X 5 

Ru-105 X X X X X X X X X X 5 

Ru-106 X X X X X X X X X X 5 

Rh-103 X X X X X X X X X X 5 

Rh-105 X X X X X X X X X X 5 

Pd-102 X X X X X X X X X X 5 

Pd-104 X X X X X X X X X X 5 

Pd-105 X X X X X X X X X X 5 

Pd-106 X X X X X X X X X X 5 

Pd-107 X X X X X X X X X X 5 

Pd-108 X X X X X X X X X X 5 

Pd-110 X X X X X X X X X X 5 

Ag-107 X X X X X X X X X X 5 

Ag-109 X X X X X X X X X X 5 

Ag-110m X X X X X X X X X X 5 

Ag-111 X X X X X X X X X X 5 

Cd-106 X X X X X X X X X X 5 

Cd-108 X X X X X X X X X X 5 

Cd-110 X X X X X X X X X X 5 

Cd-111 X X X X X X X X X X 5 

Cd-112 X X X X X X X X X X 5 

Cd-113 X X X X X X X X X X 5 

Cd-114 X X X X X X X X X X 5 

Cd-115m X X X X X X X X X X 5 

Cd-116 X X X X X X X X X X 5 

In-113 X X X X X X X X X X 5 

In-115 X X X X X X X X X X 5 

Sn-112 X X X X X X X X X X 5 

Sn-113 X X X X X X X X X X 5 
22 

Table 2.2. Background cross-section values at which Bondarenko factors are 

tabulated in the Vitamin-B7 library 

(All nuclides were processed at 300, 600, 1000 and 2100 K)

Nuclidea Background cross sections (barns) Legendre 

10 order 10 106 105 104 1000 300 100 50 10 1 

Sn-114 X X X X X X X X X X 5 

Sn-115 X X X X X X X X X X 5 

Sn-116 X X X X X X X X X X 5 

Sn-117 X X X X X X X X X X 5 

Sn-118 X X X X X X X X X X 5 

Sn-119 X X X X X X X X X X 5 

Sn-120 X X X X X X X X X X 5 

Sn-122 X X X X X X X X X X 5 

Sn-123 X X X X X X X X X X 5 

Sn-124 X X X X X X X X X X 5 

Sn-125 X X X X X X X X X X 5 

Sn-126 X X X X X X X X X X 5 

Sb-121 X X X X X X X X X X 5 

Sb-123 X X X X X X X X X X 5 

Sb-124 X X X X X X X X X X 5 

Sb-125 X X X X X X X X X X 5 

Sb-126 X X X X X X X X X X 5 

Te-120 X X X X X X X X X X 5 

Te-122 X X X X X X X X X X 5 

Te-123 X X X X X X X X X X 5 

Te-124 X X X X X X X X X X 5 

Te-125 X X X X X X X X X X 5 

Te-126 X X X X X X X X X X 5 

Te-127m X X X X X X X X X X 5 

Te-128 X X X X X X X X X X 5 

Te-129m X X X X X X X X X X 5 

Te-130 X X X X X X X X X X 5 

Te-132 X X X X X X X X X X 5 

I-127 X X X X X X X X X X 5 

I-129 X X X X X X X X X X 5 

I-130 X X X X X X X X X X 5 

I-131 X X X X X X X X X X 5 

I-135 X X X X X X X X X X 5 

Xe-123 X X X X X X X X X X 5 

Xe-124 X X X X X X X X X X 5 

Xe-126 X X X X X X X X X X 5 

Xe-128 X X X X X X X X X X 5 

Xe-129 X X X X X X X X X X 5 

Xe-130 X X X X X X X X X X 5 

Xe-131 X X X X X X X X X X 5 

Xe-132 X X X X X X X X X X 5 

Xe-133 X X X X X X X X X X 5 

Xe-134 X X X X X X X X X X 5 

Xe-135 X X X X X X X X X X 5 

Xe-136 X X X X X X X X X X 5 

Cs-133 X X X X X X X X X X 5 

Cs-134 X X X X X X X X X X 5 
23 

Table 2.2. Background cross-section values at which Bondarenko factors are 

tabulated in the Vitamin-B7 library 

(All nuclides were processed at 300, 600, 1000 and 2100 K)

Nuclidea Background cross sections (barns) Legendre 

10 order 10 106 105 104 1000 300 100 50 10 1 

Cs-135 X X X X X X X X X X 5 

Cs-136 X X X X X X X X X X 5 

Cs-137 X X X X X X X X X X 5 

Ba-130 X X X X X X X X X X 5 

Ba-132 X X X X X X X X X X 5 

Ba-133 X X X X X X X X X X 5 

Ba-134 X X X X X X X X X X 5 

Ba-135 X X X X X X X X X X 5 

Ba-136 X X X X X X X X X X 5 

Ba-137 X X X X X X X X X X 5 

Ba-138 X X X X X X X X X X 5 

Ba-140 X X X X X X X X X X 5 

La-138 X X X X X X X X X X 5 

La-139 X X X X X X X X X X 5 

La-140 X X X X X X X X X X 5 

Ce-136 X X X X X X X X X X 5 

Ce-138 X X X X X X X X X X 5 

Ce-139 X X X X X X X X X X 5 

Ce-140 X X X X X X X X X X 5 

Ce-141 X X X X X X X X X X 5 

Ce-142 X X X X X X X X X X 5 

Ce-143 X X X X X X X X X X 5 

Ce-144 X X X X X X X X X X 5 

Pr-141 X X X X X X X X X X 5 

Pr-142 X X X X X X X X X X 5 

Pr-143 X X X X X X X X X X 5 

Nd-142 X X X X X X X X X X 5 

Nd-143 X X X X X X X X X X 5 

Nd-144 X X X X X X X X X X 5 

Nd-145 X X X X X X X X X X 5 

Nd-146 X X X X X X X X X X 5 

Nd-147 X X X X X X X X X X 5 

Nd-148 X X X X X X X X X X 5 

Nd-150 X X X X X X X X X X 5 

Pm-147 X X X X X X X X X 5 

Pm-148 X X X X X X X X 5 

Pm-148m X X X X X X X X X X 5 

Pm-149 X X X X X X X X X 5 

Pm-151 X X X X X X X X X X 5 

Sm-144 X X X X X X X X X X 5 

Sm-147 X X X X X X X X X X 5 

Sm-148 X X X X X X X X X X 5 

Sm-149 X X X X X X X X X X 5 

Sm-150 X X X X X X X X X X 5 

Sm-151 X X X X X X X X X X 5 

Sm-152 X X X X X X X X X X 5 

Sm-153 X X X X X X X X X X 5 
24 

Table 2.2. Background cross-section values at which Bondarenko factors are 

tabulated in the Vitamin-B7 library 

(All nuclides were processed at 300, 600, 1000 and 2100 K)

Nuclidea Background cross sections (barns) Legendre 

10 order 10 106 105 104 1000 300 100 50 10 1 

Sm-154 X X X X X X X X X X 5 

Eu-151 X X X X X X X X X 5 

Eu-152 X X X X X X X X X 5 

Eu-153 X X X X X X X X X X 5 

Eu-154 X X X X X X X X 5 

Eu-155 X X X X X X X X 5 

Eu-156 X X X X X X X X X X 5 

Eu-157 X X X X X X X X X X 5 

Gd-152 X X X X X X X X X X 5 

Gd-153 X X X X X X X X X X 5 

Gd-154 X X X X X X X X X X 5 

Gd-155 X X X X X X X X X X 5 

Gd-156 X X X X X X X X X X 5 

Gd-157 X X X X X X X X X X 5 

Gd-158 X X X X X X X X X X 5 

Gd-160 X X X X X X X X X X 5 

Tb-159 X X X X X X X X X X 5 

Tb-160 X X X X X X X X X X 5 

Dy-156 X X X X X X X X X 5 

Dy-158 X X X X X X X X X X 5 

Dy-160 X X X X X X X X X X 5 

Dy-161 X X X X X X X X X X 5 

Dy-162 X X X X X X X X X X 5 

Dy-163 X X X X X X X X X X 5 

Dy-164 X X X X X X X X X X 5 

Ho-165 X X X X X X X X X X 5 

Ho-166m X X X X X X X X X X 5 

Er-162 X X X X X X X X X X 5 

Er-164 X X X X X X X X X X 5 

Er-166 X X X X X X X X X X 5 

Er-167 X X X X X X X X X X 5 

Er-168 X X X X X X X X X X 5 

Er-170 X X X X X X X X X X 5 

Lu-175 X X X X X X X X X X 5 

Lu-176 X X X X X X X X X X 5 

Hf-174 X X X X X X X X X X 5 

Hf-176 X X X X X X X X X X 5 

Hf-177 X X X X X X X X X X 5 

Hf-178 X X X X X X X X X X 5 

Hf-179 X X X X X X X X X X 5 

Hf-180 X X X X X X X X X X 5 

Ta-181 X X X X X X X X X X 5 

Ta-182 X X X X X X X X X X 5 

W-182 X X X X X X X X X X 5 

W-183 X X X X X X X X X X 5 

W-184 X X X X X X X X X X 5 

W-186 X X X X X X X X X X 5 
25 

Table 2.2. Background cross-section values at which Bondarenko factors are 

tabulated in the Vitamin-B7 library 

(All nuclides were processed at 300, 600, 1000 and 2100 K)

Nuclidea Background cross sections (barns) Legendre 

10 order 10 106 105 104 1000 300 100 50 10 1 

Re-185 X X X X X X X X X X 5 

Re-187 X X X X X X X X X X 5 

Ir-191 X X X X X X X X X X 5 

Ir-193 X X X X X X X X X X 5 

Au-197 X X X X X X X X X X 5 

Hg-196 X X X X X X X X X X 5 

Hg-198 X X X X X X X X X X 5 

Hg-199 X X X X X X X X X X 5 

Hg-200 X X X X X X X X X X 5 

Hg-201 X X X X X X X X X X 5 

Hg-202 X X X X X X X X X X 5 

Hg-204 X X X X X X X X X X 5 

Pb-204 X X X X X X X X X X 5 

Pb-206 X X X X X X X X X X 5 

Pb-207 X X X X X X X X X X 5 

Pb-208 X X X X X X X X X X 5 

Bi-209 X X X X X X X X X X 5 

Ra-223 X X X X X X X X X X 5 

Ra-224 X X X X X X X X X X 5 

Ra-225 X X X X X X X X X X 5 

Ra-226 X X X X X X X X X X 5 

Ac-225 X X X X X X X X X X 5 

Ac-226 X X X X X X X X X X 5 

Ac-227 X X X X X X X X X X 5 

Th-227 X X X X X X X X X X 5 

Th-228 X X X X X X X X X X 5 

Th-229 X X X X X X X X X X 5 

Th-230 X X X X X X X X X X 5 

Th-232 X X X X X X X X X X 5 

Th-233 X X X X X X X X X X 5 

Th-234 X X X X X X X X X X 5 

Pa-231 X X X X X X X X X X 5 

Pa-232 X X X X X X X X X X 5 

Pa-233 X X X X X X X X X X 5 

U-232 X X X X X X X X X X 5 

U-233 X X X X X X X X X X 5 

U-234 X X X X X X X X X X 5 

U-235 X X X X X X X X X X 5 

U-236 X X X X X X X X X X 5 

U-237 X X X X X X X X X X 5 

U-238 X X X X X X X X X X 5 

U-239 X X X X X X X X X X 5 

U-240 X X X X X X X X X X 5 

U-241 X X X X X X X X X X 5 

Np-235 X X X X X X X X X X 5 

Np-236 X X X X X X X X X X 5 

Np-237 X X X X X X X X X X 5 
26 

Table 2.2. Background cross-section values at which Bondarenko factors are 

tabulated in the Vitamin-B7 library 

(All nuclides were processed at 300, 600, 1000 and 2100 K)

Nuclidea Background cross sections (barns) Legendre 

10 order 10 106 105 104 1000 300 100 50 10 1 

Np-238 X X X X X X X X X X 5 

Np-239 X X X X X X X X X X 5 

Pu-236 X X X X X X X X 5 

Pu-237 X X X X X X X X X X 5 

Pu-238 X X X X X X X X X X 5 

Pu-239 X X X X X X X X X X 5 

Pu-240 X X X X X X X X X X 5 

Pu-241 X X X X X X X X X X 5 

Pu-242 X X X X X X X X X X 5 

Pu-243 X X X X X X X X X X 5 

Pu-244 X X X X X X X X X X 5 

Pu-246 X X X X X X X X X X 5 

Am-241 X X X X X X X X X X 5 

Am-242 X X X X X X X X X X 5 

Am-242m X X X X X X X X 5 

Am-243 X X X X X X X X X X 5 

Am-244 X X X X X X X X X X 5 

Am-244m X X X X X X X X X X 5 

Cm-241 X X X X X X X X X X 5 

Cm-242 X X X X X X X X X X 5 

Cm-243 X X X X X X X X X X 5 

Cm-244 X X X X X X X X X X 5 

Cm-245 X X X X X X X X X X 5 

Cm-246 X X X X X X X X X X 5 

Cm-247 X X X X X X X X X 5 

Cm-248 X X X X X X X X X X 5 

Cm-249 X X X X X X X X X X 5 

Cm-250 X X X X X X X X X X 5 

Bk-249 X X X X X X X X X X 5 

Bk-250 X X X X X X X X X X 5 

Cf-249 X X X X X X X X X X 5 

Cf-250 X X X X X X X X X X 5 

Cf-251 X X X X X X X X X X 5 

Cf-252 X X X X X X X X X X 5 

Cf-253 X X X X X X X X 5 

Cf-254 X X X X X X X X X X 5 

Es-253 X X X X X X X X X X 5 

Es-254 X X X X X X X X X X 5 

Es-255 X X X X X X X X X X 5 

Fm-255 X X X X X X X X X X 5 

a For nuclides with thermal moderator data, the nuclide ID for the fast evaluation is given, 

followed by the nuclide ID for the moderator. 
27 

Table 2.3. Neutron group energy boundaries for the VITAMIN-B7 library

Grp 

no. 

Upper 

energy (eV) 

Lethargy 

width 

 Grp

no. 

Upper 

energy (eV) 

Lethargy 

width 

Grp

No. 

Upper 

energy (eV) 

Lethargy 

width 

1 1.9640E+07 0.1250 68 7.8082E+05 0.0500 135 3.0354E+03 0.1000 

2 1.7332E+07 0.0249 69 7.4274E+05 0.0500 136 2.7465E+03 0.0500 

3 1.6905E+07 0.0250 70 7.0651E+05 0.0500 137 2.6126E+03 0.0500 

4 1.6487E+07 0.0500 71 6.7206E+05 0.0500 138 2.4852E+03 0.1000 

5 1.5683E+07 0.0500 72 6.3928E+05 0.0500 139 2.2487E+03 0.1000 

6 1.4918E+07 0.0250 73 6.0810E+05 0.0500 140 2.0347E+03 0.2500 

7 1.4550E+07 0.0250 74 5.7844E+05 0.0500 141 1.5846E+03 0.2500 

8 1.4191E+07 0.0250 75 5.5023E+05 0.0500 142 1.2341E+03 0.2500 

9 1.3840E+07 0.0249 76 5.2340E+05 0.0500 143 9.6112E+02 0.2500 

10 1.3499E+07 0.0501 77 4.9787E+05 0.1000 144 7.4852E+02 0.2500 

11 1.2840E+07 0.0250 78 4.5049E+05 0.1000 145 5.8295E+02 0.2500 

12 1.2523E+07 0.0250 79 4.0762E+05 0.0500 146 4.5400E+02 0.2500 

13 1.2214E+07 0.0500 80 3.8774E+05 0.0500 147 3.5357E+02 0.2500 

14 1.1618E+07 0.0499 81 3.6883E+05 0.1000 148 2.7536E+02 0.2500 

15 1.1052E+07 0.0500 82 3.3373E+05 0.1000 149 2.1445E+02 0.2500 

16 1.0513E+07 0.0500 83 3.0197E+05 0.0116 150 1.6702E+02 0.2500 

17 1.0000E+07 0.0500 84 2.9849E+05 0.0043 151 1.3007E+02 0.2500 

18 9.5123E+06 0.0500 85 2.9721E+05 0.0091 152 1.0130E+02 0.2500 

19 9.0484E+06 0.0500 86 2.9452E+05 0.0250 153 7.8893E+01 0.2500 

20 8.6071E+06 0.0500 87 2.8725E+05 0.0500 154 6.1442E+01 0.2500 

21 8.1873E+06 0.0500 88 2.7324E+05 0.1000 155 4.7851E+01 0.2500 

22 7.7880E+06 0.0500 89 2.4724E+05 0.0500 156 3.7266E+01 0.2500 

23 7.4082E+06 0.0500 90 2.3518E+05 0.0500 157 2.9023E+01 0.2500 

24 7.0469E+06 0.0500 91 2.2371E+05 0.0500 158 2.2603E+01 0.2500 

25 6.7032E+06 0.0167 92 2.1280E+05 0.0500 159 1.7604E+01 0.2500 

26 6.5924E+06 0.0333 93 2.0242E+05 0.0500 160 1.3710E+01 0.2500 

27 6.3763E+06 0.0500 94 1.9255E+05 0.0500 161 1.0677E+01 0.2500 

28 6.0653E+06 0.0500 95 1.8316E+05 0.0500 162 8.3153E+00 0.2500 

29 5.7695E+06 0.0500 96 1.7422E+05 0.0500 163 6.4760E+00 0.2500 

30 5.4881E+06 0.0500 97 1.6573E+05 0.0500 164 5.0435E+00 0.2500 

31 5.2205E+06 0.0500 98 1.5764E+05 0.0499 165 3.9279E+00 0.2500 

32 4.9659E+06 0.0500 99 1.4996E+05 0.0500 166 3.0590E+00 0.2500 

33 4.7237E+06 0.0500 100 1.4264E+05 0.0500 167 2.3824E+00 0.2500 

34 4.4933E+06 0.1000 101 1.3569E+05 0.0500 168 1.8554E+00 0.2500 

35 4.0657E+06 0.1000 102 1.2907E+05 0.0500 169 1.4450E+00 0.1057 

36 3.6788E+06 0.1000 103 1.2277E+05 0.0499 170 1.3000E+00 0.1442 

37 3.3287E+06 0.0500 104 1.1679E+05 0.0500 171 1.1253E+00 0.0412 

38 3.1664E+06 0.0500 105 1.1109E+05 0.1250 172 1.0800E+00 0.0377 

39 3.0119E+06 0.0500 106 9.8037E+04 0.1250 173 1.0400E+00 0.0392 

40 2.8651E+06 0.0500 107 8.6517E+04 0.0475 174 1.0000E+00 0.1319 

41 2.7253E+06 0.0500 108 8.2503E+04 0.0371 175 8.7643E-01 0.0912 

42 2.5924E+06 0.0500 109 7.9499E+04 0.0991 176 8.0000E-01 0.1588 

43 2.4660E+06 0.0333 110 7.1998E+04 0.0663 177 6.8256E-01 0.0880 

44 2.3852E+06 0.0084 111 6.7379E+04 0.1750 178 6.2506E-01 0.1620 

45 2.3653E+06 0.0083 112 5.6562E+04 0.0750 179 5.3158E-01 0.0612 

46 2.3457E+06 0.0167 113 5.2475E+04 0.1250 180 5.0000E-01 0.1888 

47 2.3069E+06 0.0333 114 4.6309E+04 0.1250 181 4.1399E-01 0.1210 

48 2.2313E+06 0.0500 115 4.0868E+04 0.1750 182 3.6680E-01 0.1210 

49 2.1225E+06 0.0500 116 3.4307E+04 0.0750 183 3.2500E-01 0.1671 

50 2.0190E+06 0.0500 117 3.1828E+04 0.1104 184 2.7500E-01 0.2007 
28 

Table 2.3. Neutron group energy boundaries for the VITAMIN-B7 library

Grp 

no. 

Upper 

energy (eV) 

Lethargy 

width 

 Grp

no. 

Upper 

energy (eV) 

Lethargy 

width 

Grp

No. 

Upper 

energy (eV) 

Lethargy 

width 

51 1.9205E+06 0.0500 118 2.8501E+04 0.0541 185 2.2500E-01 0.2012 

52 1.8268E+06 0.0500 119 2.7000E+04 0.0355 186 1.8400E-01 0.2043 

53 1.7377E+06 0.0500 120 2.6058E+04 0.0500 187 1.5000E-01 0.1823 

54 1.6530E+06 0.0500 121 2.4788E+04 0.0250 188 1.2500E-01 0.2231 

55 1.5724E+06 0.0500 122 2.4176E+04 0.0250 189 1.0000E-01 0.3567 

56 1.4957E+06 0.0500 123 2.3579E+04 0.0750 190 7.0000E-02 0.3365 

57 1.4227E+06 0.0499 124 2.1875E+04 0.1250 191 5.0000E-02 0.2231 

58 1.3534E+06 0.0500 125 1.9305E+04 0.2500 192 4.0000E-02 0.2877 

59 1.2874E+06 0.0500 126 1.5034E+04 0.2500 193 3.0000E-02 0.3567 

60 1.2246E+06 0.0501 127 1.1709E+04 0.1000 194 2.1000E-02 0.3704 

61 1.1648E+06 0.0500 128 1.0595E+04 0.1500 195 1.4500E-02 0.3716 

62 1.1080E+06 0.1000 129 9.1188E+03 0.2500 196 1.0000E-02 0.6931 

63 1.0026E+06 0.0417 130 7.1017E+03 0.2500 197 5.0000E-03 0.9163 

64 9.6164E+05 0.0583 131 5.5308E+03 0.2500 198 2.0000E-03 1.3863 

65 9.0718E+05 0.0500 132 4.3074E+03 0.1500 199 5.0000E-04 3.9120 

66 8.6294E+05 0.0500 133 3.7074E+03 0.1000 1.0000E-05 

67 8.2085E+05 0.0500 134 3.3546E+03 0.1000 

Note: The groups shaded in gray (groups 164–199) are the thermal groups. 

Table 2.4. Photon group energy boundaries for the 

VITAMIN-B7 library 

Grp 

no. 

Upper 

energy 

(eV) 

Grp 

no. 

Upper 

energy 

(eV) 

Grp

no. 

Upper 

energy 

(eV) 

1 3.00E+07 16 3.00E+06 31 3.00E+05 

2 2.00E+07 17 2.50E+06 32 2.00E+05 

3 1.40E+07 18 2.00E+06 33 1.50E+05 

4 1.20E+07 19 1.66E+06 34 1.00E+05 

5 1.00E+07 20 1.50E+06 35 7.50E+04 

6 8.00E+06 21 1.34E+06 36 7.00E+04 

7 7.50E+06 22 1.33E+06 37 6.00E+04 

8 7.00E+06 23 1.00E+06 38 4.50E+04 

9 6.50E+06 24 8.00E+05 39 4.00E+04 

10 6.00E+06 25 7.00E+05 40 3.00E+04 

11 5.50E+06 26 6.00E+05 41 2.00E+04 

12 5.00E+06 27 5.12E+05 42 1.00E+04 

13 4.50E+06 28 5.10E+05 1.00E+03 

14 4.00E+06 29 4.50E+05 

15 3.50E+06 30 4.00E+05 
29 

Figure 2.1. Neutron weighting spectrum used to create the finegroup VITAMIN-B7 neutron library. 
30 

Figure 2.2. Gamma weighting spectrum used to create the finegroup VITAMIN-B7 gamma library. 
31 

2.6 Processing Codes and Procedures 

The AMPX code system was used to generate pointwise and multigroup data in AMPX master 

library format. Input files were generated using an XML listing of the desired evaluations and 

XML templates. The XML listings for each evaluation were generated by a Java program that 

scans a directory tree of ENDF files and writes the relevant information to a file. The input 

listings for each evaluation are included in the D00245MNYCP00 data package available from 

the Radiation Safety Information Computational Center (RSICC). 

Relevant XML templates were processed with the EXSITE program to produce input files for 

AMPX. The input files for all cases are provided on the accompanying CD. 

2.6.1 AMPX Execution Sequences for Neutron and Gamma-Ray Yield Processing 

The AMPX processing system contains more than 100 distinct modules that can be used to 

perform a wide range of nuclear data processing functions. The specific modules needed to 

process a cross-section evaluation and generate a pointwise or multigroup cross-section library 

depend upon the data specified in the nuclide evaluation. For example, a resonance nuclide will 

require more modules to be executed than will a non-resonance nuclide. If the evaluation is a 

thermal moderator material with S() data, the AMPX execution sequence will be different 

because the final cross-section data will be a mixture of data from one evaluation for energies 

above the thermal cutoff and from another evaluation in the thermal range. 

The general procedure to generate multigroup neutron cross-section data is outlined in 

Figure 2.3. The flow chart to produce the coupled part of the library is depicted in Figure 2.4. 

The ENDF/B-VII.0 cross-section evaluations were downloaded from the National Nuclear Data 

Center (NNDC) Web page [14]. The thermal evaluations for U(UO2) and O(UO2) were excluded 

because some inconsistencies in the ENDF evaluations prevent the AMPX Y12 module from 

processing the data. 

2.6.1.1 Processing of Pointwise Cross-Section Data 

POLIDENT 

Whether cross sections are being prepared for continuous energy or multigroup applications, 

the first module that is executed is POLIDENT. For resonance nuclides, POLIDENT determines 

a point energy mesh in the resolved resonance region and calculates the cross sections on that 

energy mesh. An analogous procedure is performed for nuclides having unresolved resonance 

data. For each nuclide, POLIDENT combines the resonance parameter data (if present) with 

the smooth cross-section data from the ENDF/B File 3 point data file to produce linearized 

cross-section functions. The combined data functions are written out in a standard AMPX 

tabular data format that is based on the ENDF/B TAB1 format.
32 

Figure 2.3. AMPX sequence for producing neutron multigroup data. 

Figure 2.4. AMPX sequence to generate coupled neutron–gamma multigroup data. 
33 

TGEL 

One of the primary reasons to linearize cross-section data is to produce internally consistent 

datasets. When all of the cross-section data are produced on a linear–linear grid, the sum of a 

group of linear functions is itself linear; however, simply applying the interpolation method (e.g., 

lin-log, log-log) given in an evaluation does not necessarily lead to this consistency. To ensure 

consistency, the total cross section and other “redundant” reactions (e.g., fission, inelastic 

scatter) are formed by summing the linearized partial values. The TGEL module is used to 

ensure all redundant reactions are equal to the sum of the partial reactions. The output is 

written in the same tabular format that is used by POLIDENT.

BROADEN 

Many applications, especially those dealing with reactors, require cross sections that vary as a 

function of temperature. The point cross sections can be Doppler broadened by a numerical 

treatment used in the BROADEN module. This module adds additional points as needed to 

describe the broadened resonance structure. Unnecessary grid points are omitted. The output 

data from BROADEN are exported in an AMPX tabular format. 

JERGENS 

The production of multigroup cross sections from pointwise data requires a weighting function to 

collapse the pointwise data to the multigroup energy structure. The JERGENS module provides 

a variety of options for generating weighting spectra. JERGENS has options for calculating 

weighting spectra that may use point cross sections from POLIDENT. The output is written in 

an AMPX tabular format. For the pointwise neutron cross-section data of the coupled library 

generation task, JERGENS was used to produce a “smooth” weighting spectrum that has the 

form of a Maxwellian —1/E— fission spectrum as described in Section 2.4. 

2.6.1.2 Neutron and Gamma Yield Processing 

Y12 

A unique feature of the AMPX system is the Y12 module, which reads an ENDF/B evaluation 

and produces tabular collision kinematics data for all reactions. The functions produced by Y12 

describe what particles are produced and at what energies and directions when a reaction is 

encountered at some source energy. The output from Y12 is an AMPX tabular kinematics file. 

X10 

The X10 module reads tabular point cross sections, a tabular weighting spectrum, and the 

tabular kinematics data file and uses these three files to produce group-averaged cross sections 

and group-to-group transfer matrices. This code contains no programming related to the 

physics of any reaction and is programmed in a manner that treats neutron data, gamma-ray 

data, or gamma-ray yield data. The output from X10 is a set of group cross sections that are 

written on a file in AMPX master library format. 

When a cross-section evaluation contains gamma-production data, either in ENDF File 6 or in 

the special Files 12, 13, 14 and 15, the kinematics data file produced by Y12 and the point data 

files from POLIDENT and JERGENS are passed to X10 to produce gamma-ray yield transfer 

matrices. The gamma yield matrices indicate where gamma rays are produced as a function of 
34 

neutron reaction. The yield data are exported from X10 in an AMPX master library format. After 

the gamma transport data are processed in a separate AMPX sequence, the gamma yield data 

are coupled with the multigroup neutron and gamma-ray data. 

2.6.1.3 Thermal Neutron Processing 

FLANGE6 

If a nuclide does not have S() data given in ENDF, the FLANGE6 module is used to produce 

thermal scattering matrices at the desired temperatures based on the free gas–scattering 

model. The matrices are exported in AMPX master library format. As noted in Section 2.5, 

thermal scattering matrices based on the free gas model are limited to a P3 expansion order. 

JERGENS, Y12, and X10 

If a nuclide is a bound moderator material (e.g., hydrogen bound in water, carbon in graphite), 

the ENDF/B thermal scattering law evaluations are processed by the Y12 module to produce 

thermal kinematics data as a function of temperatures provided in the ENDF/B evaluation. In 

addition, the Y12 module produces thermal pointwise cross-section data. The resulting thermal 

kinematics file and thermal pointwise data file are passed to the X10 module, along with the 

JERGENS weighting spectrum, to produce a special thermal AMPX master library. 

In many of the ENDF/B thermal evaluations, the cross sections for the principal isotope in a 

compound can be isolated. For example, the thermal data for water gives the scattering law 

data for hydrogen in a manner that describes the specific binding effects for hydrogen present in 

a water molecule. The file then specifies that oxygen should simply be treated by the free gas 

model. This treatment allows one to produce a set of full-range cross sections for water-bound 

hydrogen by collecting together the thermal hydrogen values and combining them with the 

values from the fast-neutron hydrogen evaluation. For other moderators, such as BeO, it is not 

possible to separate the elemental data (e.g., the Be in BeO) out of the data for the molecule. In 

those cases the scattering law data are given for the molecule as a combined unit. There are 

several ways to deal with this situation. The one that was chosen to be analogous to the waterbound hydrogen case was to couple the data from the fast neutron evaluation for Be with the 

thermal BeO data, and call this set “Be bound in BeO.” For the oxygen component, a special 

set was prepared and called “O bound in BeO.” For the case of O in BeO, the oxygen data 

were prepared from the fast neutron evaluation without any thermal data (i.e., no thermal 

scattering data as would normally be present from a free gas calculation by the FLANGE6 

module). 

2.6.1.4 Generation of Full-Range Bondarenko Factors and Thermal Data 

PRUDE 

For nuclides with unresolved resonance data, the PRUDE module is used to produce point 

“averaged” cross sections that are a function of temperature and background cross section. 

The background cross section is simply defined as the cross section per unit atom of a nuclide 

of the mixture (other than the nuclide itself) in which the nuclide is mixed. The cross sections 

are written out in an AMPX tabular file. This module can be run only for monoisotopic 

evaluations. In ENDF/B-VII, the only neutron evaluations that are not monoisotopic are those 

for vanadium (Z = 23) and zinc (Z = 30). Those evaluations contain elemental data in File 3 but 

have data for multiple isotopes in File 2. 
35 

FABULOUS 

For all nuclides, the point data from POLIDENT and from PRUDE (if applicable) are passed to 

the FABULOUS module, along with the weighting spectrum file from JERGENS. FABULOUS 

uses the same temperatures and background cross sections described for PRUDE to Doppler 

broaden the functions outside the unresolved region to the temperatures at which the nuclides 

were processed, after which the functions are spliced together with the appropriate unresolved 

functions. These spliced functions are then used in a numerical scheme that calculates 

Bondarenko factors for all nuclides for elastic scattering, capture, fission, and total 

cross sections. The Bondarenko factors can be used in resonance self-shielding calculations. 

For low values of the background cross section 0, the unresolved cross-section data can 

become negative. Calculation of the “total” Bondarenko factor is given by 

 (2.1) 

In some cases, the value of can become negative. The NJOY manual [5] describes 

how this calculation is performed in the UNRESR module, which is similar to the AMPX PRUDE 

module. Reference [5] states that for materials with strong resonance overlaps, the 

approximation used in calculating this quantity can produce negative shielded cross-section 

data for lower values of 0. If this happens, FABULOUS omits the relevant value of 0 from the 

AMPX master library. The nuclides for which this occurs are indicated in Table 2.2 by the 

absence of Bondarenko data for some 0 values. 

2.6.1.5 Gamma Processing 

The processing of gamma-ray cross-section data is much simpler than that of neutron data. 

The following processes are included in the processing of the gamma-ray data from the ENDF 

evaluations: incoherent (Compton) scattering, coherent (Rayleigh) scattering, pair production in 

the electric field of the nucleus, pair production in the electric field of an electron (also known as 

triplet production), and photoelectric absorption. Note that unlike neutron cross sections, photon 

cross sections are elemental rather than isotopic. 

JERGENS 

The JERGENS module is used to produce a weighting spectrum for use in generating 

multigroup gamma-ray cross sections. The gamma spectrum is typically a 1/E spectrum with 

“roll-offs” that account for significant absorption of the gamma rays due to photoelectric 

absorption at low energies and pair production at high energies. As with the neutron 

evaluations, the weighting spectrum is written out in a pointwise AMPX tabular file based on the 

ENDF/B TAB1 format. 

 T

   T + 0

 2



   T + 0

 2

   T + 0
36 

Y12 

The Y12 module produces a tabular kinematics file that describes coherent (Rayleigh) and 

incoherent (Compton) scattering along with pair production. In addition, Y12 produces a 

pointwise data file for all of these processes. The pointwise file is exported in an AMPX tabular 

format. 

X10 

The tabular files from JERGENS and Y12 are passed to X10, along with the tabular kinematics 

file. X10 produces a group-averaged AMPX master library for gamma-ray interactions. 

2.6.1.6 Creating the Master Library 

After all the neutron and gamma data have been processed as described above, a master 

library containing data for all the nuclides is created. The input files for this process are 

included in the D00245MNYCP00 data package available from RSICC.
37 

3 BROAD-GROUP LIBRARY SPECIFICATIONS 

3.1 Name 

The problem-dependent broad-group cross-section library that was developed from the 

VITAMIN-B7 pseudo-problem-independent fine-group library is designated as BUGLE-B7. This 

naming convention maintains consistency with the previous BUGLE libraries while using a suffix 

that is consistent with that of the fine-group library. 

3.2 Materials and Energy Group Structure 

The BUGLE-B7 library contains a subset of the nuclides present in the VITAMIN-B7 fine-group 

library (see Table 2.1). The contents of the BUGLE-B7 library are described in Section 3.5. 

Some nuclides appear multiple times (with distinct identifiers in each case) because of the use 

of different resonance self-shielding and distinct flux weighting spectra for specific regions of 

typical PWR and BWR reactor shielding applications. These self-shielding and spectral 

weighting effects are discussed in Section 3.3. 

The BUGLE-B7 neutron and gamma group structures are identical to those used in the 

BUGLE-96 library. The energy group boundaries for the 47 neutron groups are provided in 

Table 3.1. The energy group boundaries for the 20 gamma groups are provided in Table 3.2. 

3.3 Self-Shielding and Group Collapsing of the 

Broad-Group Library 

While broad-group libraries such as BUGLE-B7 allow the users of transport codes to run 

calculations more rapidly and with fewer computer resources compared with the use of finegroup libraries, the accuracy of the results is dependent upon using broad-group data that have 

been collapsed from the fine-group library using appropriate weighting spectra. As with 

previous BUGLE libraries, the BUGLE-B7 library was developed for in-vessel and reactor cavity 

analyses for LWRs (both PWRs and BWRs). The validity of the BUGLE-B7 data would need to 

be explicitly demonstrated to justify its use for any other application. Even for LWR applications, 

it is important that properly self-shielded and collapsed data sets are selected for specific 

regions of the reactor plant geometry in transport models. 

The BUGLE-B7 library was produced by self-shielding nuclides from the VITAMIN-B7 library in 

representative materials, adjusting to the appropriate temperature, and then group collapsing 

using weighting spectra from key regions of one-dimensional (1D) PWR and BWR models. 

3.3.1 Self-Shielding Calculations 

The self-shielding was performed using the BONAMI code in SCALE. In the core region, 

nuclides in the fuel, cladding, and coolant were self-shielded and temperature corrected using 

PWR and BWR fuel-clad-moderator pin cell models. In the coolant regions outside the core, 

hydrogen, oxygen, and boron were self-shielded in an infinite water medium. The constituents 

of carbon steel and stainless steel were self-shielded in infinite media of carbon steel (A533-B) 

and stainless steel (SS-304). The constituents of concrete were self-shielded in an infinite 

medium of Type 04 concrete. Details of the BONAMI parameters for these calculations are 

provided in Sections 3.3.1.1–3.3.1.5.
38 

Table 3.1. Neutron group energy boundaries for the 

BUGLE-B7 library 

Grp 

no. Upper energy (eV) Lethargy

width 

VITAMIN-B7 

groups 

1 1.7332E+07 0.1999 2-7 

2 1.4191E+07 0.1500 8-12 

3 1.2214E+07 0.2000 13-16 

4 1.0000E+07 0.1500 17-19 

5 8.6071E+06 0.1500 20-22 

6 7.4082E+06 0.2000 23-27 

7 6.0653E+06 0.2000 28-31 

8 4.9658E+06 0.3000 32-35 

9 3.6788E+06 0.2000 36-38 

10 3.0119E+06 0.1000 39-40 

11 2.7253E+06 0.1000 41-42 

12 2.4660E+06 0.0417 43-44 

13 2.3652E+06 0.0083 45 

14 2.3457E+06 0.0500 46-47 

15 2.2313E+06 0.1500 48-50 

16 1.9205E+06 0.1500 51-53 

17 1.6530E+06 0.2000 54-57 

18 1.3534E+06 0.3000 58-62 

19 1.0026E+06 0.2000 63-66 

20 8.2085E+05 0.1000 67-68 

21 7.4274E+05 0.2000 69-72 

22 6.0810E+05 0.2000 73-76 

23 4.9787E+05 0.3000 77-80 

24 3.6883E+05 0.2159 81-84 

25 2.9721E+05 0.4841 85-94 

26 1.8316E+05 0.5000 95-104 

27 1.1109E+05 0.5000 105-110 

28 6.7379E+04 0.5000 111-114 

29 4.0868E+04 0.2500 115-116 

30 3.1828E+04 0.2000 117-119 

31 2.6058E+04 0.0750 120-121 

32 2.4176E+04 0.1000 122-123 

33 2.1875E+04 0.3750 124-125 

34 1.5034E+04 0.7500 126-129 

35 7.1017E+03 0.7500 130-133 

36 3.3546E+03 0.7500 134-140 

37 1.5846E+03 1.2500 141-145 

38 4.5400E+02 0.7500 146-148 

39 2.1445E+02 0.7500 149-151 

40 1.0130E+02 1.0000 152-155 

41 3.7267E+01 1.2500 156-160 

42 1.0677E+01 0.7500 161-163 

43 5.0435E+00 1.0000 164-167 

44 1.8554E+00 0.7500 168-174 

45 8.7643E-01 0.7500 175-180 

46 4.1399E-01 1.4207 181-188 

47 1.0000E-01 9.2103 189-199 

1.0000E-05 

Note: The groups shaded in gray are the thermal groups. See Sections 

3.4 and 3.5 for a discussion of upscatter in the thermal energy range. 
39 

Table 3.2. Gamma group energy 

boundaries for the BUGLE-B7 library

Grp 

No. 

Upper Energy 

(eV) 

VITAMIN-B7 

Groups 

1 1.40E+07 3-4 

2 1.00E+07 5 

3 8.00E+06 6-7 

4 7.00E+06 8-9 

5 6.00E+06 10-11 

6 5.00E+06 12-13 

7 4.00E+06 14-15 

8 3.00E+06 16-17 

9 2.00E+06 18-19 

10 1.50E+06 20-22 

11 1.00E+06 23 

12 8.00E+05 24 

13 7.00E+05 25 

14 6.00E+05 26-29 

15 4.00E+05 30-31 

16 2.00E+05 32-33 

17 1.00E+05 34-36 

18 6.00E+04 37-39 

19 3.00E+04 40 

20 2.00E+04 41 

 1.00E+04 

 
40 

3.3.1.1 PWR Fuel Cell Self-Shielding 

The PWR pin cell model, as described in Table 3.3, is based on the description in Table 3.6 of 

[1]. The coolant temperature was corrected to 583 K, which is incorrectly listed as the BWR 

coolant temperature in Table 3.6 of [1]. The Zircaloy-4 composition was modified to be 

consistent with the composition of Zircaloy-4 in the SCALE standard composition library, with 

the addition of a trace (0.01 wt %) of nickel, which is not present in Zircaloy-4. The nickel was 

added to provide cross-section data for nickel self-shielded in cladding to maintain consistency 

with the self-shielded nuclides provided in the BUGLE-96 library. 

PWR pin cell calculations run with Zirlo cladding rather than Zircaloy-4 resulted in almost no 

difference in the self-shielded cross sections for the major isotopes in the cladding. Thus, the 

VITAMIN-B7 and BUGLE-B7 libraries are considered to be appropriate for PWR cores with ZrNb cladding (such as ZIRLO and M5) as well as Zircaloy-4. 

Geometry: Cylinder 

Number of zones: 3 

Mixing table length: 40 

Dancoff factor: Cylindrical cell (Sauer’s method) square lattice with clad 

Temperature: 1000 K for zone 1, 600 K for zones 2 and 3 

Table 3.3. Description of mixtures used for self-shielding nuclides 

in the PWR pin cell model 

Zone 

no. 

Mixture 

no. 

Nuclide Concentration

(atoms/b-cm) 

Outer radius

(cm) 

Temperature 

(K) 

Zone type 

1 1 

U-235 

U-238 

O-16 

 6.325 × 10-4

 2.166 × 10-2

 4.465 × 10-2

0.41783 921 Fuel 

2 2a

Zr 

Sn 

Fe 

Cr 

Ni 

Hf 

 4.254 × 10-2

 4.825 × 10-4 

 1.486 × 10-4 

 7.598 × 10-5 

 6.731 × 10-6 

 2.213 × 10-6

0.47498 672 Cladding 

3 3 

H-1b

O-16 

B-10 

 4.714 × 10-2

 

 2.357 × 10-2

 4.200 × 10-6

0.71079 583 Moderator 

a The isotopic number densities in the BONAMI input are based on the element number 

densities listed here and isotopic abundances from [15]. 

b

 H-1 in H2O. 

 
41 

3.3.1.2 BWR Fuel Cell Self-Shielding 

The BWR pin cell model, as described in Table 3.4, is based on the description in Table 3.6 of 

[1]. The coolant temperature was corrected to 551 K, which is incorrectly listed as the PWR 

coolant temperature in Table 3.6 of [1]. The Zircaloy-4 composition was modified to be 

consistent with the composition of Zircaloy-4 in the SCALE standard composition library, with 

the addition of a trace (0.01 wt %) of nickel, which is not present in Zircaloy-4. The nickel was 

added to provide cross-section data for nickel self-shielded in cladding to maintain consistency 

with the self-shielded nuclides provided in the BUGLE-96 library. 

Geometry: Cylinder 

Number of zones: 3 

Mixing table length: 39 

Dancoff factor: Cylindrical cell (Sauer’s method) square lattice with clad 

Temperature: 1000 K for zone 1, 600 K for zones 2 and 3 

Table 3.4. Description of mixtures used for self-shielding nuclides 

in the BWR pin cell model 

Zone 

no. 

Mixture 

no.

Nuclide Concentration

(atoms/b-cm) 

Outer radius

(cm) 

Temperature 

(K) 

Zone type 

1 1 

U-235 

U-238 

O-16 

 4.959 × 10-4

 2.177 × 10-2

 4.455 × 10-2

0.53213 921 Fuel 

2 2a

Zr 

Sn 

Fe 

Cr 

Ni 

Hf 

 4.254 × 10-2

 4.825 × 10-4 

 1.486 × 10-4 

 7.598 × 10-5 

 6.731 × 10-6 

 2.213 × 10-6

0.6134 672 Cladding 

3 3 H-1b

O-16 

 2.475 × 10-2 

 1.2375 × 10-2 0.9174 551 Moderator 

a

 The isotopic number densities in the BONAMI input are based on the element number 

densities listed here and isotopic abundances from [15]. 

b

 H-1 in H2O. 

 
42 

3.3.1.3 Coolant Self-Shielding 

The coolant self-shielding model, as described in Table 3.5, is based on the PWR coolant 

composition in Table 3.3 of [1]. 

Geometry: Slab 

Number of zones: 2 

Mixing table length: 3 

Dancoff factor: Homogeneous (e = 0.0) 

Temperature: 600 K 

Table 3.5. Description of mixtures used for self-shielding nuclides in the coolant model 

Zone 

no. 

Mixture 

no. 

Nuclide Concentration

(atoms/b-cm) 

Outer radius

(cm) 

Temperature 

(K) Zone type 

1 1 

H-1a

O-16 

B-10 

 4.714 × 10-2

 2.357 × 10-2

 4.2 × 10-6

10.0 600 Moderator 

a H-1 in H2O. 

3.3.1.4 Carbon Steel and Stainless Steel Self-Shielding 

The carbon steel (A533-B) and stainless steel (SS-304) self-shielding model, as described in 

Table 3.6, is based on the description in Table 3.3 of [1]. Note that this includes the silicon in 

the steel, which was present in the XSDRNPM models of [1] but not in the BONAMI input. The 

silicon has been included in the BONAMI input for consistency between the BONAMI and 

XSDRNPM models. 

Geometry: Slab 

Number of zones: 2 

Mixing table length: 36 

Dancoff factor: Homogeneous (e = 0.0) 

Temperature: 600 K 

Table 3.6. Description of mixtures used for self-shielding nuclides in the steel models 

Zone 

no. 

Mixture 

no. 

Nuclide Concentration

(atoms/b-cm) 

Outer radius

(cm) 

Temperature 

(K) 

Zone type 

1 (A533-

B) 1a

Fe 

Cr 

Ni 

Mn 

C 

Si 

 8.19 × 10-2

 1.27 × 10-4

 4.44 × 10-4 

 1.12 × 10-3

 9.81 × 10-4

 3.74 × 10-4

10.0 600 Moderator 

2 (SS304) 2a

Fe 

Cr 

Ni 

Mn 

C 

Si 

 5.83 × 10-2

 1.74 × 10-2

 8.55 × 10-3 

 1.52 × 10-3

 2.37 × 10-4

 8.93 × 10-4

20.0 600 Moderator 

a The isotopic number densities in the BONAMI input are based on the element number 

densities listed here and isotopic abundances from [15]. 
43 

3.3.1.5 Concrete Self-Shielding 

The concrete self-shielding model, as described in Table 3.7, is based on the composition of 

Type 04 concrete from [16]. A trace (0.01 wt %) of carbon, which is not present in Type 04 

concrete, was added to the mixture to provide cross-section data for carbon self-shielded in 

concrete. This addition maintains consistency with the self-shielded nuclides provided in the 

BUGLE-96 library. 

Geometry: Slab 

Number of zones: 1 

Mixing table length: 28 

Dancoff factor: Homogeneous (e = 0.0) 

Temperature: 300 K 

Table 3.7. Description of mixtures used for self-shielding nuclides in the concrete model 

Zone 

no. 

Mixture 

no. 

Nuclide Concentration

(atoms/b-cm) 

Outer radius 

(cm) 

Temperature 

(K) Zone type 

1 1b

H-1a

O-16 

Si 

Ca 

Na 

Mg 

Al 

S 

K 

Fe 

C 

7.767 × 10-3

4.408 × 10-2

1.591 × 10-2 

2.915 × 10-3 

1.048 × 10-3 

1.487 × 10-4 

2.388 × 10-3 

5.634 × 10-5 

6.931 × 10-4 

3.127 × 10-4 

1.178 × 10-5

100.0 300 Moderator 

a

 H-1 in H2O. 

b

 The isotopic number densities in the BONAMI input are based on the element 

number densities listed here and isotopic abundances from [15]. 

3.3.2 Fine-Group Spectra Used for Group Collapsing 

The self-shielded fine-group cross sections were group collapsed to the BUGLE group structure 

using weighting spectra at the following locations of the 1D PWR and BWR reactor plant 

models: (1) off-center in the core region of the BWR model, (2) off-center in the core region of 

the PWR model, (3) the downcomer region in the PWR model, (4) within the PWR pressure 

vessel at one-fourth the vessel thickness, and (5) within the PWR concrete biological shield. 

The weighting spectra were generated using the 1D XSDRNPM discrete-ordinates transport 

code in SCALE. Details of the 1D PWR and BWR models are provided in Sections 3.3.2.1 and 

3.3.2.2. 

The calculated neutron flux spectra are shown in Figure 3.3 and listed in Table 3.10. The 

gamma spectra are shown in Figure 3.4 and listed in Table 3.11. In addition to these selfshielded nuclides with distinct weighting spectra, many of the nuclides in the VITAMIN-B7 library 

were collapsed to the BUGLE group structure using the flux spectrum calculated within the 

concrete shield. These nuclides are infinitely dilute (i.e., they are not self-shielded). 
44 

3.3.2.1 BWR XSDRNPM Model for Weighting Spectra 

The geometry of the 1D BWR reactor plant model is shown in Figure 3.1. The compositions of 

the homogenized core region and the coolant regions outside the core are provided in Table 

3.8. The compositions of the remaining regions (SS-304, A533-B, and concrete) are the same 

as those used in the BONAMI cases. The nuclides used in each region are those that were 

self-shielded in the equivalent BONAMI cases. For the core region, nuclide identifiers ending in 

“21” are self-shielded in the fuel, those ending in “22” are self-shielded in the cladding, and 

those ending in “23” are self-shielded in the coolant. 

Figure 3.1. One-dimensional BWR reactor plant model used to generate 

weighting spectra for group collapsing BUGLE-B7 cross sections from 

VITAMIN-B7 data (from Ref. [1]). 

Table 3.8. Description of the homogenized core and coolant mixtures 

used in the 1D BWR reactor plant model 

(The remaining mixtures are equivalent to those in the 

BONAMI descriptions of Sections 3.3.1.4 and 3.3.1.5)

Mixture no. Description Nuclide Nuclide 

identifier 

Concentration 

(atoms/b-cm) 

1 Homogenized

BWR core 

H-1 

O-16 

Zr-90 

Zr-91 

Zr-92 

Zr-94 

Zr-96 

Fe-54 

Fe-56 

Fe-57 

Fe-58 

U-235 

U-238 

O-16 

100123 

801623 

4009022 

4009122 

4009222 

4009422 

4009622 

2605422 

2605622 

2605722 

2605822 

9223521 

9223821 

801621 

1.5354 × 10-2

7.677 × 10-3 

2.99609 × 10-3 

6.46833 × 10-4

9.88697 × 10-4

1.00196 × 10-3

1.61420 × 10-4

1.1818 × 10-6

1.83715 × 10-5 

4.206 × 10-7 

5.61 × 10-8

1.2125 × 10-4

5.322 × 10-3

1.0884 × 10-2

2 Coolant H-1 

O-16 

100131 

801631 

4.950 × 10-2 

2.475 × 10-2
45 

3.3.2.2 PWR XSDRNPM Model for Weighting Spectra 

The geometry of the 1D PWR reactor plant model is shown in Figure 3.2. The compositions of 

the homogenized core region and the coolant regions outside the core are provided in 

Table 3.9. The compositions of the remaining regions (SS-304, A533-B, and concrete) are the 

same as those used in the BONAMI cases. The nuclides used in each region are those that 

were self-shielded in the equivalent BONAMI cases. For the core region, nuclide identifiers 

ending in “11” are self-shielded in the fuel, those ending in “12”are self-shielded in the cladding, 

and those ending in “13” are self-shielded in the coolant. 

Figure 3.2. One-dimensional PWR reactor plant model used to generate 

weighting spectra for group collapsing BUGLE-B7 cross sections from 

VITAMIN-B7 data (from Ref. [1]). 

Table 3.9. Description of the homogenized core and coolant mixtures 

used in the 1D PWR reactor plant model 

(The remaining mixtures are equivalent to those in the 

BONAMI descriptions of Sections 3.3.1.4 and 3.3.1.5)

Mixture no. Description Nuclide Nuclide 

identifier 

Concentration 

(atoms/b-cm) 

1 Homogenized

PWR core 

H-1 

O-16 

B-10 

Zr-90 

Zr-91 

Zr-92 

Zr-94 

Zr-96 

Fe-54 

Fe-56 

Fe-57 

Fe-58 

U-235 

U-238 

O-16 

100113 

801613 

501013 

4009012 

4009112 

4009212 

4009412 

4009612 

2605412 

2605612 

2605712 

2605812 

9223511 

9223811 

801611 

2.768 × 10-2

1.384 × 10-2

2.466 × 10-6 

2.19023 × 10-3

4.77635 × 10-4

7.30075 × 10-4

7.39867 × 10-4

1.19196 × 10-4

8.5196 × 10-7 

1.32444 × 10-5 

3.0324 × 10-7

4.0432 × 10-8

1.903 × 10-4

6.515 × 10-3

1.343 × 10-2

2 Coolant H-1 

O-16 

B-10 

100131 

801631 

501031 

4.714 × 10-2 

2.357 × 10-2

4.200 × 10-6
46 

Figure 3.3. Fine-group neutron weighting spectra from the 1D BWR 

and PWR models. 
47 

Table 3.10. Fine-group neutron weighting spectra from the 1D BWR and PWR models

Fine 

group 

BWR 

core 

PWR 

core 

PWR 

downcomer 

PWR 1/4T 

pressure 

vessel 

PWR 

concrete 

1 7.4034E+05 1.1652E+06 1.3689E+04 1.9840E+03 7.2613E+01 

2 2.8348E+08 4.3737E+08 3.4877E+06 4.1141E+05 1.1445E+04 

3 3.1895E+08 4.9821E+08 5.1197E+06 6.7461E+05 2.1537E+04 

4 1.0541E+09 1.6354E+09 1.5503E+07 2.0159E+06 6.3624E+04 

5 1.7355E+09 2.6795E+09 2.3267E+07 2.8704E+06 8.4169E+04 

6 1.2327E+09 1.8984E+09 1.5693E+07 1.8803E+06 5.3037E+04 

7 1.5787E+09 2.4352E+09 2.0224E+07 2.4039E+06 6.6197E+04 

8 1.9935E+09 3.0750E+09 2.5339E+07 2.9915E+06 8.2124E+04 

9 2.4650E+09 3.7954E+09 3.0405E+07 3.5454E+06 9.6953E+04 

10 6.7912E+09 1.0425E+10 7.9712E+07 9.1150E+06 2.4370E+05 

11 5.0383E+09 7.7160E+09 5.6083E+07 6.2334E+06 1.6565E+05 

12 5.2018E+09 7.9778E+09 6.0638E+07 6.8372E+06 1.8448E+05 

13 1.4361E+10 2.1786E+10 1.4042E+08 1.4953E+07 3.8116E+05 

14 2.0364E+10 3.0770E+10 1.8482E+08 1.8765E+07 4.6688E+05 

15 3.0735E+10 4.6945E+10 3.1298E+08 3.2407E+07 8.4803E+05 

16 4.0259E+10 6.0834E+10 3.5899E+08 3.6893E+07 9.0410E+05 

17 5.5065E+10 8.3267E+10 4.8276E+08 4.8719E+07 1.1951E+06 

18 7.4767E+10 1.1329E+11 6.5538E+08 6.4877E+07 1.6252E+06 

19 9.4952E+10 1.4288E+11 7.5819E+08 7.3429E+07 1.7946E+06 

20 1.1929E+11 1.7846E+11 8.6205E+08 8.0158E+07 1.8780E+06 

21 1.5525E+11 2.3294E+11 1.1258E+09 1.0075E+08 2.3070E+06 

22 1.9219E+11 2.8800E+11 1.3218E+09 1.1627E+08 2.5898E+06 

23 2.3238E+11 3.4640E+11 1.4744E+09 1.2641E+08 2.7759E+06 

24 2.9512E+11 4.4199E+11 1.9114E+09 1.6081E+08 3.5913E+06 

25 1.1260E+11 1.6889E+11 7.1988E+08 6.0067E+07 1.3928E+06 

26 2.5152E+11 3.7720E+11 1.6337E+09 1.3582E+08 3.1327E+06 

27 4.2894E+11 6.4243E+11 2.6646E+09 2.2157E+08 5.1116E+06 

28 4.7381E+11 7.0131E+11 2.5788E+09 2.0881E+08 4.6578E+06 

29 5.3265E+11 7.8521E+11 2.6482E+09 2.0629E+08 4.6384E+06 

30 6.4612E+11 9.6079E+11 3.3779E+09 2.5305E+08 6.2228E+06 

31 7.0308E+11 1.0388E+12 3.2904E+09 2.4088E+08 5.8598E+06 

32 8.1088E+11 1.2002E+12 3.7656E+09 2.6499E+08 6.5489E+06 

33 8.5463E+11 1.2551E+12 3.6693E+09 2.6636E+08 6.4273E+06 

34 1.9370E+12 2.8282E+12 7.5263E+09 5.3062E+08 1.3642E+07 

35 2.0759E+12 2.9834E+12 6.7056E+09 4.9100E+08 1.1938E+07 

36 2.2960E+12 3.2612E+12 6.4363E+09 4.8393E+08 1.1400E+07 

37 1.3611E+12 1.9341E+12 3.7710E+09 2.6612E+08 7.0454E+06 

38 1.6110E+12 2.3020E+12 4.8869E+09 3.3192E+08 9.4936E+06 

39 1.6246E+12 2.3028E+12 4.9872E+09 3.8484E+08 1.0667E+07 

40 1.7513E+12 2.4746E+12 5.2588E+09 4.1583E+08 1.2380E+07 

41 1.9182E+12 2.7033E+12 5.7562E+09 4.7781E+08 1.5014E+07 

42 1.9311E+12 2.7061E+12 5.6114E+09 4.5397E+08 1.5058E+07 

43 1.3380E+12 1.8738E+12 4.2353E+09 3.5762E+08 1.3683E+07 

44 3.6292E+11 5.1067E+11 1.3065E+09 1.1141E+08 4.2002E+06 
48 

Table 3.10. Fine-group neutron weighting spectra from the 1D BWR and PWR models

Fine 

group 

BWR 

core 

PWR 

core 

PWR 

downcomer 

PWR 1/4T 

pressure 

vessel 

PWR 

concrete 

45 3.7059E+11 5.2125E+11 1.4303E+09 1.2501E+08 5.4672E+06 

46 7.2871E+11 1.0175E+12 2.4588E+09 2.0222E+08 8.9979E+06 

47 1.3277E+12 1.8268E+12 4.3055E+09 4.3190E+08 1.5657E+07 

48 1.8683E+12 2.5485E+12 5.6681E+09 6.0272E+08 1.9632E+07 

49 1.8160E+12 2.4622E+12 5.3208E+09 5.7392E+08 1.8294E+07 

50 1.7709E+12 2.3880E+12 5.1263E+09 5.5508E+08 1.5305E+07 

51 1.7033E+12 2.2811E+12 4.6866E+09 5.5973E+08 1.3896E+07 

52 2.0165E+12 2.6985E+12 5.7552E+09 6.9216E+08 2.4030E+07 

53 1.9678E+12 2.6029E+12 5.4552E+09 7.4819E+08 2.1839E+07 

54 1.9144E+12 2.5209E+12 5.1811E+09 7.1039E+08 1.9868E+07 

55 2.0032E+12 2.6273E+12 5.3326E+09 7.4358E+08 2.4080E+07 

56 2.1002E+12 2.7369E+12 5.3660E+09 7.1873E+08 2.6627E+07 

57 2.0562E+12 2.6588E+12 5.2589E+09 7.6846E+08 2.7606E+07 

58 1.7172E+12 2.2071E+12 4.1276E+09 8.0072E+08 1.9045E+07 

59 2.1402E+12 2.7532E+12 5.2048E+09 7.2908E+08 2.8223E+07 

60 2.1494E+12 2.7339E+12 5.3453E+09 1.0972E+09 3.3087E+07 

61 1.9991E+12 2.5273E+12 4.8444E+09 1.0321E+09 3.2701E+07 

62 3.2889E+12 4.1520E+12 7.6215E+09 1.7021E+09 4.0242E+07 

63 1.2881E+12 1.6229E+12 2.8549E+09 6.7632E+08 1.3089E+07 

64 2.1573E+12 2.6952E+12 4.8049E+09 1.2719E+09 2.6598E+07 

65 2.0823E+12 2.5687E+12 4.4237E+09 9.0783E+08 3.2951E+07 

66 2.4022E+12 2.9289E+12 4.8882E+09 9.9454E+08 3.8374E+07 

67 2.6672E+12 3.2138E+12 5.0901E+09 8.8520E+08 4.6651E+07 

68 2.5545E+12 3.0408E+12 4.7773E+09 9.0662E+08 5.1140E+07 

69 2.3821E+12 2.8187E+12 4.5885E+09 1.2637E+09 5.6060E+07 

70 2.2967E+12 2.7104E+12 4.4566E+09 1.4812E+09 6.2259E+07 

71 2.2357E+12 2.6308E+12 4.4009E+09 1.9916E+09 7.1873E+07 

72 2.1751E+12 2.5536E+12 4.2913E+09 2.1566E+09 8.2835E+07 

73 2.1151E+12 2.4794E+12 4.2139E+09 1.5470E+09 8.4246E+07 

74 2.0546E+12 2.4056E+12 4.0724E+09 1.3153E+09 7.7597E+07 

75 1.9945E+12 2.3333E+12 3.9410E+09 1.2173E+09 8.4693E+07 

76 1.9157E+12 2.2411E+12 3.7900E+09 1.0836E+09 8.0072E+07 

77 3.1915E+12 3.7632E+12 6.5006E+09 2.3478E+09 1.2367E+08 

78 2.4525E+12 2.9360E+12 5.2166E+09 1.5580E+09 7.3697E+07 

79 1.4408E+12 1.7268E+12 3.0279E+09 6.8903E+08 4.3161E+07 

80 1.4439E+12 1.7207E+12 3.0149E+09 1.6381E+09 4.5030E+07 

81 3.2869E+12 3.8611E+12 6.4222E+09 3.2368E+09 1.1971E+08 

82 3.3507E+12 3.8896E+12 6.3385E+09 3.3818E+09 1.4396E+08 

83 3.7462E+11 4.3239E+11 7.0691E+08 4.4117E+08 1.6897E+07 

84 1.3773E+11 1.5895E+11 2.6040E+08 1.4508E+08 6.3249E+06 

85 2.8933E+11 3.3400E+11 5.4775E+08 2.3122E+08 1.2996E+07 

86 7.8566E+11 9.0719E+11 1.4859E+09 3.6386E+08 3.4904E+07 

87 1.5319E+12 1.7677E+12 2.8983E+09 6.9973E+08 7.2281E+07 

88 2.9140E+12 3.3594E+12 5.5549E+09 1.7596E+09 1.4144E+08 

89 1.3910E+12 1.6027E+12 2.6734E+09 1.0177E+09 6.9762E+07 
49 

Table 3.10. Fine-group neutron weighting spectra from the 1D BWR and PWR models

Fine 

group 

BWR 

core 

PWR 

core 

PWR 

downcomer 

PWR 1/4T 

pressure 

vessel 

PWR 

concrete 

90 1.3457E+12 1.5530E+12 2.6105E+09 5.8077E+08 6.9327E+07 

91 1.3093E+12 1.5126E+12 2.5600E+09 9.2774E+08 6.2300E+07 

92 1.2783E+12 1.4780E+12 2.5134E+09 5.9753E+08 6.0962E+07 

93 1.2461E+12 1.4421E+12 2.4633E+09 3.5654E+08 5.7686E+07 

94 1.2121E+12 1.4034E+12 2.4090E+09 7.9281E+08 5.5178E+07 

95 1.1826E+12 1.3700E+12 2.3641E+09 1.1297E+09 6.3833E+07 

96 1.1517E+12 1.3354E+12 2.3124E+09 8.5419E+08 7.8954E+07 

97 1.1272E+12 1.3078E+12 2.2742E+09 8.3562E+08 8.9362E+07 

98 1.0988E+12 1.2763E+12 2.2311E+09 4.9196E+08 8.6548E+07 

99 1.0742E+12 1.2490E+12 2.1908E+09 2.5833E+08 7.9384E+07 

100 1.0485E+12 1.2188E+12 2.1438E+09 1.0576E+09 8.5262E+07 

101 1.0257E+12 1.1928E+12 2.1067E+09 5.8177E+08 7.8929E+07 

102 1.0037E+12 1.1681E+12 2.0706E+09 9.6342E+08 8.9268E+07 

103 9.8082E+11 1.1422E+12 2.0307E+09 6.2906E+08 7.8103E+07 

104 9.6313E+11 1.1224E+12 2.0017E+09 5.8502E+08 7.9068E+07 

105 2.3263E+12 2.7157E+12 4.8721E+09 9.8982E+08 1.8685E+08 

106 2.2137E+12 2.5932E+12 4.6998E+09 7.1494E+08 1.6321E+08 

107 8.2928E+11 9.6456E+11 1.7430E+09 1.9605E+08 6.1852E+07 

108 6.1885E+11 7.3178E+11 1.3525E+09 8.9514E+08 5.4065E+07 

109 1.6602E+12 1.9371E+12 3.5295E+09 7.1066E+08 1.3947E+08 

110 1.0546E+12 1.2482E+12 2.3140E+09 8.8534E+08 9.1425E+07 

111 2.7534E+12 3.2181E+12 5.9313E+09 1.2755E+09 2.1530E+08 

112 1.1314E+12 1.3295E+12 2.4708E+09 3.8644E+08 9.0173E+07 

113 1.8223E+12 2.1533E+12 4.0405E+09 8.8147E+08 1.7182E+08 

114 1.7415E+12 2.0794E+12 3.9487E+09 6.2868E+08 1.5480E+08 

115 2.4648E+12 2.8758E+12 5.3928E+09 6.4109E+08 1.9616E+08 

116 1.0162E+12 1.1953E+12 2.2672E+09 2.1258E+08 9.0193E+07 

117 1.4692E+12 1.7302E+12 3.2921E+09 1.0792E+08 1.3436E+08 

118 7.0469E+11 8.3395E+11 1.5981E+09 3.2966E+07 6.2731E+07 

119 4.6046E+11 5.4385E+11 1.0517E+09 1.7637E+08 4.3321E+07 

120 6.4545E+11 7.6143E+11 1.4763E+09 1.2172E+09 6.1346E+07 

121 3.1618E+11 3.7644E+11 7.3564E+08 5.9378E+08 3.1251E+07 

122 3.1397E+11 3.7475E+11 7.3110E+08 3.1150E+08 3.0834E+07 

123 9.5796E+11 1.1290E+12 2.1799E+09 6.9480E+08 9.1506E+07 

124 1.5321E+12 1.8347E+12 3.5907E+09 6.5949E+08 1.4999E+08 

125 3.0774E+12 3.6337E+12 7.0826E+09 9.3190E+08 2.9548E+08 

126 2.9742E+12 3.5351E+12 7.0031E+09 8.7233E+08 2.9222E+08 

127 1.1678E+12 1.3921E+12 2.7821E+09 3.3123E+08 1.1590E+08 

128 1.7224E+12 2.0628E+12 4.1520E+09 3.6833E+08 1.7098E+08 

129 2.8150E+12 3.3881E+12 6.8822E+09 3.7791E+08 2.8689E+08 

130 2.8204E+12 3.3649E+12 6.8697E+09 9.8607E+08 2.8322E+08 

131 2.7559E+12 3.3112E+12 6.8618E+09 1.0295E+09 2.8562E+08 

132 1.6343E+12 1.9691E+12 4.1184E+09 5.5869E+08 1.6840E+08 

133 1.0920E+12 1.3130E+12 2.7490E+09 4.0615E+08 1.0833E+08 

134 1.0745E+12 1.2996E+12 2.7507E+09 3.8060E+08 9.6542E+07 
50 

Table 3.10. Fine-group neutron weighting spectra from the 1D BWR and PWR models

Fine 

group 

BWR 

core 

PWR 

core 

PWR 

downcomer 

PWR 1/4T 

pressure 

vessel 

PWR 

concrete 

135 1.0471E+12 1.2827E+12 2.7526E+09 3.3591E+08 7.2082E+07 

136 5.0545E+11 6.2941E+11 1.3758E+09 1.4741E+08 4.3914E+07 

137 5.8435E+11 6.7638E+11 1.3758E+09 1.2083E+08 5.7623E+07 

138 1.0466E+12 1.2716E+12 2.7534E+09 1.7560E+08 1.2982E+08 

139 1.0779E+12 1.2988E+12 2.7637E+09 2.5952E+08 1.2528E+08 

140 2.6025E+12 3.1702E+12 6.9325E+09 8.5679E+08 3.0272E+08 

141 2.5938E+12 3.1603E+12 6.9582E+09 8.6975E+08 2.9377E+08 

142 2.5886E+12 3.1562E+12 6.9856E+09 7.0789E+08 2.9019E+08 

143 2.5382E+12 3.1102E+12 7.0227E+09 6.9958E+08 2.8935E+08 

144 2.5032E+12 3.0779E+12 7.0590E+09 6.8924E+08 2.8915E+08 

145 2.4749E+12 3.0530E+12 7.0969E+09 6.6231E+08 2.8910E+08 

146 2.4533E+12 3.0362E+12 7.1279E+09 3.9652E+08 2.8857E+08 

147 2.4255E+12 3.0138E+12 7.1828E+09 5.9109E+08 2.8856E+08 

148 2.3763E+12 2.9627E+12 7.2245E+09 6.2887E+08 2.8843E+08 

149 2.3656E+12 2.9537E+12 7.2650E+09 6.4364E+08 2.8818E+08 

150 2.3261E+12 2.9196E+12 7.3094E+09 6.5075E+08 2.8798E+08 

151 2.1287E+12 2.7186E+12 7.3504E+09 6.5300E+08 2.8752E+08 

152 2.3217E+12 2.9142E+12 7.3936E+09 6.5287E+08 2.8706E+08 

153 2.2008E+12 2.7881E+12 7.4367E+09 6.5064E+08 2.8651E+08 

154 2.2039E+12 2.7967E+12 7.4797E+09 6.4605E+08 2.8588E+08 

155 2.0972E+12 2.6976E+12 7.5227E+09 6.3879E+08 2.8518E+08 

156 2.0546E+12 2.6313E+12 7.5644E+09 6.2900E+08 2.8437E+08 

157 2.0795E+12 2.6752E+12 7.6067E+09 6.1637E+08 2.8351E+08 

158 1.7398E+12 2.2970E+12 7.6466E+09 6.0102E+08 2.8250E+08 

159 1.9647E+12 2.5578E+12 7.6886E+09 5.8279E+08 2.8150E+08 

160 1.9410E+12 2.5249E+12 7.7261E+09 5.6292E+08 2.8036E+08 

161 1.8821E+12 2.4617E+12 7.7562E+09 5.3813E+08 2.7900E+08 

162 1.3821E+12 1.8895E+12 7.7880E+09 5.1095E+08 2.7757E+08 

163 1.5255E+12 2.0917E+12 7.8204E+09 4.8177E+08 2.7607E+08 

164 1.7361E+12 2.3269E+12 8.0473E+09 4.5827E+08 2.8036E+08 

165 1.7358E+12 2.3237E+12 8.1219E+09 4.2504E+08 2.7772E+08 

166 1.7669E+12 2.3641E+12 8.2442E+09 3.9069E+08 2.7859E+08 

167 1.7774E+12 2.3817E+12 8.3982E+09 3.5591E+08 2.7968E+08 

168 1.8023E+12 2.4170E+12 8.5658E+09 3.2486E+08 2.8115E+08 

169 7.6709E+11 1.0319E+12 3.7014E+09 1.2885E+08 1.1962E+08 

170 1.0409E+12 1.4008E+12 5.1174E+09 1.6523E+08 1.6428E+08 

171 2.9056E+11 3.9221E+11 1.4623E+09 4.4951E+07 4.6847E+07 

172 2.6869E+11 3.6279E+11 1.3487E+09 4.0260E+07 4.3050E+07 

173 2.8371E+11 3.8373E+11 1.4290E+09 4.0745E+07 4.4855E+07 

174 9.6750E+11 1.3053E+12 4.8306E+09 1.3067E+08 1.5173E+08 

175 6.8130E+11 9.2215E+11 3.4488E+09 8.5179E+07 1.0591E+08 

176 1.2083E+12 1.6330E+12 6.1481E+09 1.3911E+08 1.8629E+08 

177 6.7008E+11 9.0790E+11 3.5097E+09 7.2979E+07 1.0393E+08 

178 1.2859E+12 1.7436E+12 7.1731E+09 1.2836E+08 1.9454E+08 

179 5.0940E+11 6.9543E+11 3.2058E+09 4.7441E+07 7.5270E+07 
51 

Table 3.10. Fine-group neutron weighting spectra from the 1D BWR and PWR models

Fine 

group 

BWR 

core 

PWR 

core 

PWR 

downcomer 

PWR 1/4T 

pressure 

vessel 

PWR 

concrete 

180 1.6203E+12 2.2133E+12 1.2427E+10 1.4692E+08 2.3248E+08 

181 1.1136E+12 1.5368E+12 1.1960E+10 9.8701E+07 1.5140E+08 

182 1.1896E+12 1.6559E+12 1.7076E+10 1.0605E+08 1.5126E+08 

183 1.8814E+12 2.6572E+12 3.7321E+10 1.6323E+08 2.1925E+08 

184 2.7266E+12 3.9199E+12 7.3722E+10 2.1999E+08 2.9307E+08 

185 3.2258E+12 4.7111E+12 1.0916E+11 2.3430E+08 3.5321E+08 

186 3.6594E+12 5.4140E+12 1.4474E+11 2.3376E+08 4.6719E+08 

187 3.4197E+12 5.1174E+12 1.5120E+11 1.8940E+08 5.6715E+08 

188 4.1202E+12 6.2160E+12 1.9807E+11 2.0177E+08 9.6181E+08 

189 5.8834E+12 8.9272E+12 3.0547E+11 2.5952E+08 2.1568E+09 

190 4.2281E+12 6.4532E+12 2.3451E+11 1.6286E+08 2.3951E+09 

191 2.0300E+12 3.1130E+12 1.1748E+11 6.5982E+07 1.5131E+09 

192 1.8542E+12 2.8539E+12 1.1072E+11 4.9697E+07 1.6982E+09 

193 1.4149E+12 2.1843E+12 8.6955E+10 2.9774E+07 1.5654E+09 

194 8.0702E+11 1.2493E+12 5.0856E+10 1.2906E+07 1.0466E+09 

195 4.2120E+11 6.5357E+11 2.7046E+10 5.1424E+06 6.1392E+08 

196 3.0781E+11 4.7853E+11 2.0073E+10 2.7927E+06 4.9546E+08 

197 9.1821E+10 1.4306E+11 6.0675E+09 5.8151E+05 1.6107E+08 

198 1.7008E+10 2.6548E+10 1.1289E+09 7.9657E+04 3.1250E+07 

199 1.1410E+09 1.7814E+09 7.5436E+07 4.4364E+03 2.1357E+06 

 
52 

Figure 3.4. Fine-group gamma weighting spectra from the 1D BWR 

and PWR models. 
53 

Table 3.11. Fine-group gamma weighting spectra from the 1D BWR and PWR models 

Fine 

group 

BWR 

core 

PWR 

core 

PWR 

downcomer 

PWR 1/4T 

pressure 

vessel 

PWR 

concrete 

1 6.5625E+02 6.0659E+02 1.2340E+04 2.0883E+03 4.5662E+01

2 3.8697E+05 3.8075E+05 7.0004E+05 8.6233E+04 1.4901E+03

3 3.0811E+06 4.5177E+06 3.6753E+06 4.0796E+05 5.7632E+03

4 7.3879E+07 1.1705E+08 2.5340E+08 2.4284E+07 1.5185E+06

5 4.2255E+09 2.6380E+09 1.0215E+11 8.8743E+09 9.7966E+07

6 2.0652E+10 2.3818E+10 1.5940E+11 1.4294E+10 3.3993E+08

7 6.3363E+10 9.1133E+10 4.8905E+10 5.0100E+09 1.1066E+08

8 9.2640E+10 1.4474E+11 3.3782E+10 3.8429E+09 6.4522E+07

9 1.9306E+11 3.2867E+11 5.4526E+10 5.7934E+09 1.7301E+08

10 2.7373E+11 5.1008E+11 5.9217E+10 6.4439E+09 1.8634E+08

11 3.6621E+11 6.7757E+11 3.9188E+10 5.1374E+09 1.7496E+08

12 4.8172E+11 8.6989E+11 4.7531E+10 6.0754E+09 3.7454E+08

13 1.3998E+12 2.3963E+12 5.9742E+10 7.4576E+09 1.8038E+08

14 1.9455E+12 3.4019E+12 6.5233E+10 8.4190E+09 4.7152E+08

15 2.9489E+12 5.2058E+12 8.5363E+10 1.0486E+10 2.6937E+08

16 5.4454E+12 9.5606E+12 1.0778E+11 1.3055E+10 3.1449E+08

17 1.0896E+13 1.8731E+13 5.5368E+11 2.6988E+10 7.6379E+08

18 9.5278E+12 1.6421E+13 1.8974E+11 1.9285E+10 3.4869E+08

19 5.2995E+12 9.3788E+12 1.0037E+11 1.0331E+10 1.9397E+08

20 6.2922E+12 1.1200E+13 1.1766E+11 1.1653E+10 1.9011E+08

21 4.2595E+11 7.6561E+11 8.2820E+09 7.7874E+08 1.2177E+07

22 1.6054E+13 2.8722E+13 2.8397E+11 2.9300E+10 4.8943E+08

23 1.2606E+13 2.1996E+13 2.5588E+11 2.6068E+10 4.1545E+08

24 6.8086E+12 1.2120E+13 1.4718E+11 1.4343E+10 2.3374E+08

25 7.6130E+12 1.3389E+13 1.7329E+11 1.6832E+10 2.6329E+08

26 6.6842E+12 1.1575E+13 1.8426E+11 1.7673E+10 2.7884E+08

27 3.9421E+12 6.7731E+12 1.1253E+11 2.3384E+10 2.8219E+08

28 4.4756E+12 8.3124E+12 3.8183E+11 1.9738E+10 3.2562E+08

29 3.3769E+12 5.8446E+12 2.0854E+11 1.8669E+10 2.9277E+08

30 5.5828E+12 9.5714E+12 5.4382E+11 4.6990E+10 7.8313E+08

31 3.4384E+12 5.7836E+12 9.1767E+11 7.1222E+10 1.3853E+09

32 6.4196E+11 1.0683E+12 7.4570E+11 4.5103E+10 1.1372E+09

33 2.8565E+11 4.5359E+11 1.0878E+12 3.2230E+10 1.7310E+09

34 1.2480E+11 1.9858E+11 8.1060E+11 5.6950E+09 1.1695E+09

35 1.2878E+10 2.0266E+10 1.7867E+11 3.2536E+08 2.1199E+08

36 1.6623E+10 2.5910E+10 3.4271E+11 2.4270E+08 3.2498E+08

37 2.8897E+10 3.9736E+10 5.1089E+11 2.6484E+07 2.0151E+08

38 1.1308E+10 1.4875E+10 1.1802E+11 4.2032E+05 1.7484E+07

39 3.6878E+09 5.4675E+09 1.1007E+11 1.7072E+05 5.6426E+06

40 1.6152E+09 2.4150E+09 1.0967E+10 2.0718E+05 2.6908E+05

41 1.0354E+09 1.5679E+09 8.1821E+07 3.2258E+05 3.7104E+04

42 3.7997E+07 6.5722E+07 4.1407E+03 5.3263E+03 2.3838E+03
54 

3.4 Processing Codes and Procedures 

All the processing of the fine-group VITAMIN-B7 library into the broad-group BUGLE-B7 library 

was performed using modules of the AMPX code system. The names and a brief description of 

each module are provided in Table 3.12. Figure 3.5 illustrates the steps that were performed to 

self-shield the fine-group data and create the fine-group weighting spectra. Figure 3.6 illustrates 

the steps that were used to collapse the fine-group data using the PWR- and BWR-specific flux 

spectra. 

The treatment of upscatter in the thermal groups was handled in two ways. In many shielding 

applications, accurate calculation of the thermal flux is relatively unimportant, because neutron 

transport is dominated by higher-energy neutrons. When the thermal flux is unimportant, the 

transport calculations can be run more rapidly by ignoring upscatter and hence eliminating the 

need to perform outer iterations to converge the flux. 

The method used to remove the upscatter terms in the BUGLE-B7 library is consistent with that 

used in BUGLE-96. In the MALOCS processing step illustrated in Figure 3.6, the upscatter 

between two groups is set to zero and the downscatter is reduced by an equivalent amount to 

preserve the net transfer rate between the two groups. The in-group scattering terms for both 

groups are increased by a corresponding amount to preserve the total scattering rate. This 

approach is referred to as the “ANISN upscatter” approximation. Although this adjustment 

provides an acceptable solution in most circumstances, it can lead to negative downscatter 

terms if the upscatter is greater than the downscatter between two groups. 

The final processing sequence used in creating the BUGLE-B7, shown in Figure 3.7, was the 

creation of a full set of infinitely dilute cross sections collapsed with the concrete weighting 

spectrum. These nuclides can be used for general purposes where the problem-specific selfshielded and weighted data sets are not appropriate. 

3.5 Library Format and Content 

The BUGLE-B7 data library contains two major parts: 

1. BUGLE-B7: a complete replacement for all the data contained in BUGLE-96. This 

library uses the “ANISN upscatter” approximation to remove the upscatter cross sections 

2. BUGLE-B7T: the same cross sections as BUGLE-B7 with the inclusion of the thermal 

upscatter cross sections 

The BUGLE-B7 library package is available in ANISN card image format and AMPX working 

library format. The table width of the ANISN-formatted files for BUGLE-B7 is 67, and the table 

length is 70. For BUGLE-B7T the table width is 67, and the table length is 74. 

Table positions 1 through 3 are defined in the normal manner as follows: 

1. absorption cross section (a) 

2. fission cross section multiplied by the average number of neutrons per fission (f) 

3. total cross section (t) 

For the BUGLE-B7 data, position 4 is the in-group scattering cross section, and positions 5 

through 70 are the downscatter cross sections. For the BUGLE-B7T, data positions 4 through 7 
55 

are the upscatter cross sections, position 8 is the in-group scattering cross section, and 

positions 9 through 74 are the downscatter cross sections. 

Listings of the library contents for BUGLE-B7 are given in Table 3.13 and Table 3.14. Table 

3.13 lists the materials that were processed as infinitely dilute (i.e., with no self-shielding) and 

collapsed using the concrete weighting spectrum. Table 3.14 lists the materials that were selfshielded and collapsed using the PWR and BWR spectra described in Section 3.3. 

Table 3.12. AMPX modules used to self-shield and group collapse fine-group 

cross sections into the BUGLE-B7 library 

Module name Function 

AIM Converts AMPX master libraries between binary and BCD formats 

AJAX Merges and deletes nuclides from AMPX master libraries 

ALPO Produces an ANISN library from an AMPX working library 

BONAMI Performs interpolation on Bondarenko factors to self-shield reaction cross 

sections 

MALOCS Collapses data in an AMPX master library to a coarser group structure 

based on a specified weighting spectrum 

NITAWL Converts a master library to working library format 

PALEALE Lists data contained in AMPX master and working libraries 

WORKER Prepares working libraries for use in transport calculations. Interpolates 

thermal scattering matrices, which is particularly useful for nuclides with 

upscattering 

XSDRNPM Performs 1D discrete-ordinates transport calculations using cross sections 

from an AMPX working library 
56 

Figure 3.5. Procedure for self-shielding nuclide data and calculating 

BWR- and PWR-specific flux spectra. 
57 

Figure 3.6. Procedure for collapsing self-shielded fine-group cross 

sections into the BUGLE-B7 library. 

 
58 

Figure 3.7. Procedure for generating infinitely dilute BUGLE-B7 data 

with the concrete weighting spectrum. 
59 

Table 3.13. Nuclides in BUGLE-B7 that are infinitely dilute and 

collapsed with the concrete weighting spectrum 

Nuclide ANISN ID Nuclide ANISN ID Nuclide ANISN ID 

Ag-107 1-6 Fe-54 275-282 Pa-233 567-572 

Ag-109 7-12 Fe-56 283-290 Pb-206 573-578 

Al-27 13-20 Fe-57 291-298 Pb-207 579-584 

Am-241 21-26 Fe-58 299-306 Pb-208 585-590 

Am-242 27-32 Ga 307-312 Pu-236 591-596 

Am-242m 33-38 H-1 (H2O) 313-320 Pu-237 597-602 

Am-243 39-44 H-1 (CH2) 321-328 Pu-238 603-608 

Au-197 45-50 H-2 (D2O) 329-336 Pu-239 609-614 

B-10 51-58 H-3 337-344 Pu-240 615-620 

B-11 59-66 He-3 345-352 Pu-241 621-626 

Ba-138 67-72 He-4 353-360 Pu-242 627-632 

Be-9 73-80 Hf-174 361-366 Pu-243 633-638 

Be-9 

(Thermal) 

81-88 Hf-176 367-372 Pu-244 639-644 

Bi-209 89-94 Hf-177 373-378 Re-185 645-650 

C 95-102 Hf-178 379-784 Re-187 651-656 

C (Graphite) 103-110 Hf-179 385-390 S 657-664 

Ca 111-118 Hf-180 391-396 S-32 665-672 

Cd-Nat 119-124 In-Nat 397-402 Si 673-680 

Cl-Nat 125-132 K 403-410 Sn-Nat 681-686 

Cm-241 133-138 Li-6 411-418 Ta-181 687-692 

Cm-242 139-144 Li-7 419-426 Ta-182 693-698 

Cm-243 145-150 Mg 427-434 Th-230 699-704 

Cm-244 151-156 Mn-55 435-442 Th-232 705-710 

Cm-245 157-162 Mo 443-448 Ti 711-718 

Cm-246 163-168 N-14 449-456 U-232 719-724 

Cm-247 169-174 N-15 457-464 U-233 725-730 

Cm-248 175-180 Na-23 465-472 U-234 731-736 

Co-59 181-188 Nb-93 473-478 U-235 737-742 

Cr-50 189-196 Ni-58 479-486 U-236 743-748 

Cr-52 197-204 Ni-60 487-494 U-237 749-754 

Cr-53 205-212 Ni-61 495-502 U-238 755-760 

Cr-54 213-220 Ni-62 503-510 V 761-768 

Cu-63 221-228 Ni-64 511-518 W-Nat 769-774 

Cu-65 229-236 Np-237 519-524 W-182 775-780 

Eu-151 237-242 Np-238 525-530 W-183 781-786 

Eu-152 243-248 Np-239 531-536 W-184 787-792 

Eu-153 249-254 O-16 537-544 W-186 793-798 

Eu-154 255-260 O-17 545-552 Y-89 799-804 

Eu-155 261-266 P-31 553-560 Zr 805-810 

F-19 267-274 Pa-231 561-566 Zr (Zirc-2) 811-816 

 
60 

Table 3.14. Nuclides in BUGLE-B7 that are self-shielded and 

collapsed using specific BWR and PWR weighting spectra

Nuclide ANISN ID Description 

B-10 1001-1008 PWR core coolant 

Cr-50 1009-1016 PWR core clad 

Cr-52 1017-1024 PWR core clad 

Cr-53 1025-1032 PWR core clad 

Cr-54 1033-1040 PWR core clad 

Fe-54 1041-1048 PWR core clad 

Fe-56 1049-1056 PWR core clad 

Fe-57 1057-1064 PWR core clad 

Fe-58 1065-1072 PWR core clad 

H-1 (H2O) 1073-1080 PWR core coolant 

Ni-58 1081-1088 PWR core clad 

Ni-60 1089-1096 PWR core clad 

Ni-61 1097-1104 PWR core clad 

Ni-62 1105-1112 PWR core clad 

Ni-64 1113-1120 PWR core clad 

O-16 1121-1128 PWR core coolant 

U-235 1129-1136 PWR core fuel 

U-238 1137-1142 PWR core fuel 

Zr 1149-1154 PWR core clad 

 

H-1 (H2O) 2001-2008 PWR downcomer 

O-16 2009-2016 PWR downcomer 

C 2017-2024 PWR downcomer 

Cr-50 2025-2032 PWR downcomer 

Cr-52 2033-2040 PWR downcomer 

Cr-53 2041-2048 PWR downcomer 

Cr-54 2049-2056 PWR downcomer 

Fe-54 2057-2064 PWR downcomer 

Fe-56 2065-2072 PWR downcomer 

Fe-57 2073-2080 PWR downcomer 

Fe-58 2081-2088 PWR downcomer 

Mn-55 2089-2096 PWR downcomer 

Ni-58 2097-2104 PWR downcomer 

Ni-60 2105-2112 PWR downcomer 

Ni-61 2113-2120 PWR downcomer 

Ni-62 2121-2128 PWR downcomer 

Ni-64 2129-2136 PWR downcomer 

 

C 3001-3008 PWR 1⁄4 T in pressure vessel 

Cr-50 3009-3016 PWR 1⁄4 T in pressure vessel 

Cr-52 3017-3024 PWR 1⁄4 T in pressure vessel 

Cr-53 3025-3032 PWR 1⁄4 T in pressure vessel 

Cr-54 3033-3040 PWR 1⁄4 T in pressure vessel 

Fe-54 3041-3048 PWR 1⁄4 T in pressure vessel 

Fe-56 3049-3056 PWR 1⁄4 T in pressure vessel 

Fe-57 3057-3064 PWR 1⁄4 T in pressure vessel 

Fe-58 3065-3072 PWR 1⁄4 T in pressure vessel 

Mn-55 3073-3080 PWR 1⁄4 T in pressure vessel 

Ni-58 3081-3088 PWR 1⁄4 T in pressure vessel 

Ni-60 3089-3096 PWR 1⁄4 T in pressure vessel 
61 

Table 3.14. Nuclides in BUGLE-B7 that are self-shielded and 

collapsed using specific BWR and PWR weighting spectra

Nuclide ANISN ID Description 

Ni-61 3097-3104 PWR 1⁄4 T in pressure vessel 

Ni-62 3105-3112 PWR 1⁄4 T in pressure vessel 

Ni-64 3113-3120 PWR 1⁄4 T in pressure vessel 

 

Al-27 4001-4008 Concrete type 04 

C 4009-4016 Concrete type 04 

Ca 4017-4024 Concrete type 04 

Fe-54 4025-4032 Concrete type 04 

Fe-56 4033-4040 Concrete type 04 

Fe-57 4041-4048 Concrete type 04 

Fe-58 4049-4056 Concrete type 04 

H-1 (H2O) 4057-4064 Concrete type 04 

K 4065-4072 Concrete type 04 

Mg 4073-4080 Concrete type 04 

Na-23 4081-4088 Concrete type 04 

O-16 4089-4096 Concrete type 04 

Si 4097-4104 Concrete type 04 

 

C 5001-5008 Carbon steel 

C 5009-5016 Stainless steel 

Cr-50 5017-5024 Carbon steel 

Cr-50 5025-5032 Stainless steel 

Cr-52 5033-5040 Carbon steel 

Cr-52 5041-5048 Stainless steel 

Cr-53 5049-5056 Carbon steel 

Cr-53 5057-5064 Stainless steel 

Cr-54 5065-5072 Carbon steel 

Cr-54 5073-5080 Stainless steel 

Fe-54 5081-5088 Carbon steel 

Fe-54 5089-5096 Stainless steel 

Fe-56 5097-5104 Carbon steel 

Fe-56 5105-5112 Stainless steel 

Fe-57 5113-5120 Carbon steel 

Fe-57 5121-5128 Stainless steel 

Fe-58 5129-5136 Carbon steel 

Fe-58 5137-5144 Stainless steel 

Mn-55 5145-5152 Carbon steel 

Mn-55 5153-5160 Stainless steel 

Ni-58 5161-5168 Carbon steel 

Ni-58 5169-5176 Stainless steel 

Ni-60 5177-5185 Carbon steel 

Ni-60 5186-5192 Stainless steel 

Ni-61 5193-5200 Carbon steel 

Ni-61 5201-5208 Stainless steel 

Ni-62 5209-5216 Carbon steel 

Ni-62 5217-5224 Stainless steel 

Ni-64 5225-5232 Carbon steel 

Ni-64 5233-5240 Stainless steel 

Fe-54 6001-6008 BWR core clad 

Fe-56 6009-6016 BWR core clad 
62 

Table 3.14. Nuclides in BUGLE-B7 that are self-shielded and 

collapsed using specific BWR and PWR weighting spectra

Nuclide ANISN ID Description 

Fe-57 6017-6024 BWR core clad 

Fe-58 6025-6032 BWR core clad 

H-1 (H2O) 6033-6040 BWR core coolant 

O-16 6041-6048 BWR core coolant 

O-16 6049-6056 BWR core fuel 

U-235 6057-6062 BWR core fuel 

U-238 6063-6068 BWR core fuel 

Zr 6069-6074 BWR core clad 

3.6 Response functions 

The BUGLE-B7 ANISN libraries include response function data consistent with that in BUGLE96. The following response functions are included: 

1. Neutron cross sections for 43 reactions. These data are based primarily on ENDF/B-VII 

data. Some reaction data are based on the International Reactor Dosimetry File IRDF2002 [17] and the SNLRML compendium of dosimetry data [18]. 

2. Fission spectra () and neutrons per fission () data for nine nuclides. 

3. Silicon displacement kerma data from SNLRML. 

4. Total kerma factors for neutrons and photons. Those factors were generated using the 

HEATR (neutron) and GAMINR (photon) modules in NJOY. 

5. Several parameters related to the neutron group structure. 

6. Multipliers for total neutron flux and for neutron flux with E > 1.0 MeV, E > 0.1 MeV, and 

E < 0.414 eV. 

A list of the response functions is provided in Table 3.15. The broad-group responses were 

collapsed from fine-group data using two weighting spectra: a flat (uniform) spectrum and the 

fine-group spectrum calculated at the 1⁄4-thickness location in the pressure vessel. The 

response functions are included in the library with the same structure as the cross-section data, 

that is, 67-by-70 cross-section matrices (or 67-by-74 for a library with upscatter groups). Within 

a matrix each row represents a different response function. The row positions of the response 

functions are listed in Table 3.15. The ANISN IDs of the response tables are 7001 and 7002 for 

the flat-weighted responses and 7003 and 7004 for the pressure-vessel-weighted responses. 

Note that the In-115 and Rh-103 inelastic scattering cross sections (Rows 28 and 46, 

respectively) are actually the cross sections for the production of the metastable isomers for 

these two isotopes. In the ENDF/B data, inelastic scattering cross sections are provided to 

discrete levels (MT = 51–90) and to the continuum (MT = 91). For dosimetry purposes, the 

cross section of interest is the “inelastic to metastable” cross section, which incorporates the 

production of the metastable state based on the appropriate scattering levels and branching 

ratios for the decay of each excited state. In IRDF-2002, the “inelastic to metastable” pointwise 

cross-section data are found in File 10 rather than in File 3 (see page 12 of [19]). 

The neutron and photon kerma factors are also provided in the format of cross-section tables 

with ANISN IDs of 8001–8004. The kerma factors were collapsed from the VITAMIN-B7 data 

using the concrete weighting spectra for neutrons and photons. The 8001 and 8002 sets 
63 

contain the neutron kerma factors for the 120 nuclides that were included in BUGLE-96. The 

8003 and 8004 sets contain the photon kerma factors. Note that photon kermas are in many 

cases provided for multiple isotopes of a given element, even though photon kerma values are 

identical for all isotopes of a given element. This duplication is retained to provide consistent 

row ordering of the neutron and photon kerma data. The row positions for the kerma data are 

listed in Table 3.16 and Table 3.17. 

Table 3.15. Neutron response functions included with the BUGLE-B7 ANISN-formatted 

library 

Row Response data Source 

of data Row Response data Source of data 

Data sets 7001 (flat weighted) and 7003 (1/4-thickness pressure vessel weighted) 

1 Group Upper Energy (MeV) N/A 29 I-127 (n, 2n) ENDF 

2 U-235 fission spectrum () ENDF 30 Sc-45 (n, ) ENDF 

3 Li-6 (n, x) He-4 SNLRML 31 Na-23 (n, ) ENDF 

4 B-10 (n, ) ENDF 32 Fe-58 (n, ) ENDF 

5 Th-232 (n, f) ENDF 33 Co-59 (n, ) ENDF 

6 U-235 (n, f) ENDF 34 Cu-63 (n, ) ENDF 

7 U-238 (n, f) ENDF 35 In-115 (n, ) ENDF 

8 Np-237 (n, f) ENDF 36 Au-197 (n, ) ENDF 

9 Pu-239 (n, f) ENDF 37 Th-232 (n, ) ENDF 

10 Al-27 (n, p) ENDF 38 U-238 (n, ) ENDF 

11 Al-27 (n, ) ENDF 39 Square root of Emid (MeV1/2) N/A 

12 S-32 (n, p) SNLRML 40 Total neutron flux N/A 

13 Ti-46 (n, p) ENDF 41 U-234 (n, f) ENDF 

14 Ti-47 (n, p) ENDF 42 U-236 (n, f) ENDF 

15 Ti-47 (n, np) ENDF 43 Pu-240 (n, f) ENDF 

16 Ti-48 (n, p) ENDF 44 Pu-241 (n, f) ENDF 

17 Ti-48 (n, np) ENDF 45 Pu-242 (n, f) ENDF 

18 Mn-55 (n, 2n) ENDF 46 Rh-103 (n, n’) IRDF-2002 

19 Fe-54 (n, p) ENDF 47 Si displacement kerma (eVb) SNLRML 

20 Fe-56 (n, p) ENDF 48 U-238 fission spectrum () ENDF 

21 Co-59 (n, 2n) ENDF 49 Pu-239 fission spectrum () ENDF 

22 Co-59 (n, ) ENDF 50 Neutron flux with E > 1.0 MeV N/A 

23 Ni-58 (n, p) ENDF 51 Neutron flux with E > 0.1 MeV N/A 

24 Ni-58 (n, 2n) ENDF 52 Neutron flux with E < 0.414 eV N/A 

25 Ni-60 (n, p) ENDF 53 Midpoint energy (MeV) N/A 

26 Cu-63 (n, ) ENDF 54 Group energy width (MeV) N/A 

27 Cu-65 (n, 2n) ENDF 55 Group lethargy width N/A 

28 In-115 (n, n’) IRDF-2002 

Data Sets 7002 (flat weighted) and 7004 (1/4-thickness pressure vessel weighted) 

1 Pu-238 (n, f) ENDF 9 Pu-241 neutrons/fission () ENDF 

2 U-234 neutrons/fission () ENDF 10 Pu-242 neutrons/fission () ENDF 

3 U-235 neutrons/fission () ENDF 11 U-234 fission spectrum () ENDF 

4 U-236 neutrons/fission () ENDF 12 U-236 fission spectrum () ENDF 

5 U-238 neutrons/fission () ENDF 13 Pu-238 fission spectrum () ENDF 

6 Pu-238 neutrons/fission () ENDF 14 Pu-240 fission spectrum () ENDF 

7 Pu-239 neutrons/fission () ENDF 15 Pu-241 fission spectrum () ENDF 

8 Pu-240 neutrons/fission () ENDF 16 Pu-242 fission spectrum () ENDF 
64 

Table 3.16. Row positions for neutron kerma factors with ANISN ID 

8001 and gamma-ray kerma factors with ANISN ID 8003 

(All kerma values are in units of eVb) 

Row Nuclide Row Nuclide Row Nuclide 

1 Ag-107 21 Cm-242 41 Fe-54 

2 Ag-109 22 Cm-243 42 Fe-56 

3 Al-27 23 Cm-244 43 Fe-57 

4 Am-241 24 Cm-245 44 Fe-58 

5 Am-242 25 Cm-246 45 Ga 

6 Am-242m 26 Cm-247 46 H-1 (H2O) 

7 Am-243 27 Cm-248 47 H-1 (CH2) 

8 Au-197 28 Co-59 48 H-2 (D2O) 

9 B-10 29 Cr-50 49 H-3 

10 B-11 30 Cr-52 50 He-3 

11 Ba-138 31 Cr-53 51 He-4 

12 Be-9 32 Cr-54 52 Hf-174 

13 Be-9 (thermal) 33 Cu-63 53 Hf-176 

14 Bi-209 34 Cu-65 54 Hf-177 

15 C 35 Eu-151 55 Hf-178 

16 C (graphite) 36 Eu-152 56 Hf-179 

17 Ca 37 Eu-153 57 Hf-180 

18 Cd 38 Eu-154 58 In 

19 Cl 39 Eu-155 59 K 

20 Cm-241 40 F-19 60 Li-6 

Table 3.17. Row positions for neutron kerma factors with ANISN ID 

8002 and gamma-ray kerma factors with ANISN ID 8004 

(All kerma values are in units of eVb)

Row Nuclide Row Nuclide Row Nuclide 

1 Li-7 21 Pa-233 41 Ta-182 

2 Mg 22 Pb-206 42 Th-230 

3 Mn-55 23 Pb-207 43 Th-232 

4 Mo 24 Pb-208 44 Ti 

5 N-14 25 Pu-236 45 U-232 

6 N-15 26 Pu-237 46 U-233 

7 Na-23 27 Pu-238 47 U-234 

8 Nb-93 28 Pu-239 48 U-235 

9 Ni-58 29 Pu-240 49 U-236 

10 Ni-60 30 Pu-241 50 U-237 

11 Ni-61 31 Pu-242 51 U-238 

12 Ni-62 32 Pu-243 52 V 

13 Ni-64 33 Pu-244 53 W-Nat 

14 Np-237 34 Re-185 54 W-182 

15 Np-238 35 Re-187 55 W-183 

16 Np-239 36 S 56 W-184 

17 O-16 37 S-32 57 W-186 

18 O-17 38 Si 58 Y-89 

19 P-31 39 Sn-Nat 59 Zr 

20 Pa-231 40 Ta-181 60 Zr (Zirc-2) 
65 

4 LIBRARY VERIFICATION AND VALIDATION 

The VITAMIN-B7 and BUGLE-B7 libraries were verified and validated using three levels of 

testing: (1) automatic diagnostic software routines to check for internal consistency of the data 

files, (2) “unit tests” of the processed cross sections for each element, and (3) computational 

evaluation of criticality and shielding benchmark problems. 

Verification of the libraries was accomplished using the first two levels of testing. These steps 

were an important part of the processing methodology, as changes that are made to processing 

codes to resolve a problem may cause unexpected changes in other portions of the processed 

data files. This was particularly important for verification of the gamma yield matrices, as the 

formatting for the yield data has changed substantially for many newer evaluations, with data 

being moved from Files 12 and 13 to File 6. Details of the verification steps are provided in 

Sections 4.1 and 4.2. 

The benchmark evaluations provide an additional level of verification of the processing of the 

data. In addition, the shielding benchmarks provide validation for the use of the libraries for 

LWR shielding applications. The benchmarks were analyzed using deterministic transport 

calculations, Monte Carlo calculations, and the hybrid MAVRIC sequence in SCALE. Details of 

the benchmark evaluations are provided in Sections 4.3 through 4.5. 

4.1 Processing Methods 

The first step in verifying the VITAMIN-B7 and BUGLE-B7 libraries made use of diagnostic 

checks in the AMPX code system. The RADE module was used to check all the AMPXformatted libraries. RADE performs the tests listed below: 

1. t = a + s

2. in = in(partial) 

3. a = c + f

4. c = ng + nnp + nd

5. el(g) = g'el,0(g’) (for all processes with a scattering matrix) 

6. 0(gg’) > 0 

7. t, a, f,n,np, ... > 0 

8. fl,min ≤ fl(gg’) ≤ 1.0, 

where ݂ሺ݃→݃ᇱ

ሻ ൌ 	 ఙ೗ሺ௚→௚ᇱሻ

ሺଶ௟ାଵሻఙబሺ௚→௚ᇱሻ

 and fl,min = 1.0 for all odd l, and is given by the following values for even l: 

 l = 2: 0.5 

 l = 4: 0.433 

 l = 6: 0.419 

 l = 8: 0.414 

4.2 Unit Tests 

A significant difference between the VITAMIN-B6 and VITAMIN-B7 fine-group libraries is the 

processing techniques that were used to generate the libraries. When the VITAMIN-B6 library 
66 

was generated, the NJOY code system was the only cross-section processing code able to 

process ENDF/B-VI data. NJOY was thus used to generate the fine-group VITAMIN-B6 library, 

while the AMPX-77 code system was used to process VITAMIN-B6 into the broad-group 

BUGLE-96 library. 

The current version of AMPX has the ability to process ENDF/B-VI and ENDF/B-VII data. As 

one means of verifying the processing of the ENDF evaluations by the various AMPX modules, 

a set of “unit tests” was performed for the fine-group data. The unit tests, which were performed 

for each element in the VITAMIN-B7 library, were used to compare neutron and photon fluxes 

calculated using XSDRNPM and the VITAMIN-B7 library with equivalent MCNP5 [20] 

calculations with continuous energy cross-section data based on ENDF/B-VII. These tests 

provided an assessment of the fine-group data based on the calculated fluxes and also 

confirmed that the AMPX processing of the fine-group library was consistent with the NJOY 

processing of the continuous energy MCNP data. This was particularly important for gamma 

yield matrices, as the formatting for the yield data has changed substantially for many newer 

evaluations, with data being moved from Files 12 and 13 to File 6. 

4.3 Critical Benchmark Experiments 

A set of criticality benchmark experiments was evaluated as the first step of validating the finegroup VITAMIN-B7 library. The benchmark critical configurations provide an important test of 

the cross-section processing methods and resulting libraries for fuel and moderator materials 

and some structural materials. The benchmarks, which are listed in Section 4.3.1, were 

selected from the International Handbook of Evaluated Criticality Safety Benchmark 

Experiments [21]. The selected benchmarks are also indicated by their CSEWG (Cross-Section 

Evaluation Working Group) identifiers. 

Each of the benchmark experiments listed in Table 4.1 was modeled using the SCALE 

generalized geometry package and calculated using the KENO-VI Monte Carlo transport code 

in SCALE. 

Table 4.1. Critical benchmark experiments used for validation of the VITAMIN-B7 library

 ICSBEP identifier CSEWG 

identifier Description 

Thermal System Benchmarks 

HEU-SOL-THERM013, 

Cases 1-4 

ORNL-1 

 through 

ORNL-4 

Unreflected spheres of highly enriched uranium 

(93.2 wt % 235U) as uranyl nitrate in H2O. Some 

solutions were poisoned with boric acid. The H/235U 

ratios ranged from 972 to 1378. The spherical shell, 

which was constructed of aluminum, had a diameter of 

approximately 69.2 cm. 

HEU-SOL-THERM-032 ORNL-10 Unreflected sphere of highly enriched uranium (93.2 

wt % 235U) as uranyl nitrate in H2O. The H/235U ratio 

was 1835. The spherical shell, which was constructed 

of aluminum, had a diameter of approximately 122 cm. 

PU-SOL-THERM-021, 

Cases 7,3,2, and 5 

PNL-1, PNL-6, 

PNL-8, PNL-12 

Unreflected and water-reflected spheres of plutonium 

(4.57 wt % 240Pu) as plutonium nitrate in H2O. The 

spherical shell, which was made of stainless steel, had 

a diameter of approximately 38.6 cm. 

PU-SOL-THERM-011, 

Cases 18-1, 18-6, and 

16-5 

PNL-3, PNL-4, 

PNL-5 

Unreflected spheres of plutonium (4.17– 4.20 wt % 

240Pu) as plutonium nitrate in H2O. The spherical shells, 

which were made of stainless steel, had diameters of 

approximately 40.6 and 45.7 cm. 
67 

Table 4.1. Critical benchmark experiments used for validation of the VITAMIN-B7 library

 ICSBEP identifier CSEWG 

identifier Description 

PU-SOL-THERM-004 PNL-7 Unreflected sphere of plutonium (0.54 wt % 240Pu) as 

plutonium nitrate in H2O. The spherical shell, which 

was made of stainless steel, had a diameter of 

approximately 35.6 cm. 

Fast System Benchmarks 

Jezebel PU-METFAST-001 Unreflected 6.3849 cm sphere of Pu (98.98 wt %) and 

Ga (1.02 wt %). The atom percentages of 239Pu, 240Pu, 

and 241Pu are 95.2, 4.5, and 0.3, respectively. 

Jezebel-Pu-240 PU-METFAST-002 Unreflected 6.6595 cm sphere of Pu (98.99 wt %) and 

Ga (1.01 wt %). The atom percentages of 239Pu, 240Pu, 

and 241Pu are 76.4, 20.1, and 0.4, respectively. 

Jezebel-U-233 U233-METFAST-001 Unreflected 5.9838 cm sphere of U. The atom 

percentages of 233U, 234U, 235U, and 238U are 98.13, 

1.24, 0.03, and 0.6, respectively. 

Godiva HEU-METFAST-001, 

Case 1 

Unreflected spherical geometry with six spherical shells 

of highly enriched uranium. The enrichment of each 

shell is slightly different, with weight percentages of 

234U, 235U, and 238U of approximately 1, 94, and 5, 

respectively. The outer shell radius is 8.7499 cm. 

Godiva HEU-METFAST-001, 

Case 2 

Unreflected sphere of highly enriched uranium. The 

weight percentages of 234U, 235U, and 238U are 1.02, 

93.71, and 5.27, respectively. The spherical radius is 

8.7407 cm. 

FLATTOP-U-235 HEU-METFAST-028 A highly enriched uranium spherical core (6.1156-cm 

radius) with a large sphere (24.1242-cm radius) of 

natural uranium as a reflector. 

FLATTOP-PU PU-METFAST-006 A spherical plutonium core (4.5532-cm radius) with 

weight percentages of 93.8, 4.8, 0.3, and 1.1 of 239Pu, 

240Pu, 241Pu, and Ga, respectively. The core is 

surrounded by a natural uranium reflector with an outer 

radius of 24.1242 cm. 

FLATTOP-U-233 U233-METFAST-005, 

Case 1 

A 5.0444-cm uranium core surrounded by a beryllium 

reflector with an outer radius of 7.0891 cm. The weight 

percentages of 233U, 234U, and 238U in the core were 

98.2, 1.1, and 0.7, respectively. 

FLATTOP-U-233 U233-METFAST-005, 

Case 2 

A 4.5999-cm uranium core surrounded by a beryllium 

reflector with an outer radius of 8.7960 cm. The weight 

percentages of 233U, 234U, and 238U in the core were 

98.2, 1.1, and 0.7, respectively. 

BIG TEN IEU-METFAST-007 A reflected cylinder of uranium that is 10.0634 wt % 

235U. The cylindrical core is surrounded by inner and 

outer reflectors. The inner reflector consists of 

alternating plates of natural uranium and highly 

enriched uranium. The outer reflector is depleted 

uranium. 

ZPR-6-6A IEU-COMPFAST-001 A highly enriched uranium oxide plate fuel critical 

assembly with a depleted uranium reflector blanket and 

an outer reflector of structural materials. 

ZPR-6-7 MIX-COMPFAST-001 A highly enriched critical assembly with a mixture of 

uranium oxide and plutonium oxide plates surrounded 

by a depleted uranium reflector blanket and an outer 

reflector of structural materials. 
68 

4.3.1 Results for the Criticality Benchmarks for Thermal Systems 

The results of the validation tests for the thermal system benchmarks are summarized in Table 

4.2. The eigenvalues calculated using VITAMIN-B7 are all within 2 sigma of the experimental 

keff values. 

Table 4.2. Calculated eigenvalues for the thermal system critical benchmarks 

used for validation of the VITAMIN-B7 library 

(All calculations were performed using KENO VI. Uncertainties are 1-sigma values) 

ICSBEP identifier CSEWG 

identifier

Experimental 

keff

Cross-section library used in KENO-VI 

VITAMIN-B6 VITAMIN-B7 

Continuous 

energy 

ENDF/B-VII 

HEU-SOL-THERM-013, Case 1 ORNL-1 1.0012 ± 0.0026 0.9989 ± 0.0002 0.9987 ± 0.0002 0.9980 ± 0.0002

HEU-SOL-THERM-013, Case 2 ORNL-2 1.0007 ± 0.0036 0.9980 ± 0.0002 0.9978 ± 0.0002 0.9973 ± 0.0002

HEU-SOL-THERM-013, Case 3 ORNL-3 1.0009 ± 0.0036 0.9947 ± 0.0002 0.9944 ± 0.0002 0.9940 ± 0.0002

HEU-SOL-THERM-013, Case 4 ORNL-4 1.0003 ± 0.0036 0.9963 ± 0.0002 0.9961 ± 0.0002 0.9955 ± 0.0002

HEU-SOL-THERM-032 ORNL-10 1.0013 ± 0.0026 0.9989 ± 0.0002 1.0001 ± 0.0002 0.9992 ± 0.0002

PU-SOL-THERM-021, Case 7 PNL-1 1.0000 ± 0.0032 1.0077 ± 0.0002 1.0063 ± 0.0002 1.0044 ± 0.0002

PU-SOL-THERM-021, Case 3 PNL-6 1.0000 ± 0.0065 1.0035 ± 0.0002 1.0056 ± 0.0002 0.9982 ± 0.0002

PU-SOL-THERM-021, Case 2 PNL-8 1.0000 ± 0.0032 1.0068 ± 0.0002 1.0059 ± 0.0002 1.0035 ± 0.0002

PU-SOL-THERM-021, Case 5 PNL-12 1.0000 ± 0.0032 1.0044 ± 0.0002 1.0044 ± 0.0002 1.0033 ± 0.0002

PU-SOL-THERM-011, Case 18-1 PNL-3 1.0000 ± 0.0052 0.9960 ± 0.0002 0.9946 ± 0.0002 0.9936 ± 0.0002

PU-SOL-THERM-011, Case 18-6 PNL-4 1.0000 ± 0.0052 1.0019 ± 0.0002 1.0007 ± 0.0002 0.9990 ± 0.0002

PU-SOL-THERM-011, Case 16-5 PNL-5 1.0000 ± 0.0052 1.0071 ± 0.0002 1.0066 ± 0.0002 1.0037 ± 0.0002

PU-SOL-THERM-004 PNL-7 1.0000 ± 0.0047 1.0043 ± 0.0002 1.0044 ± 0.0002 1.0038 ± 0.0002

4.3.2 Results for the Criticality Benchmarks for Fast Systems 

The results of the validation tests for the fast system benchmarks are summarized in Table 4.3. 

The eigenvalues calculated using VITAMIN-B7 are within 2 sigma of the experimental keff values 

for seven of the fast benchmarks but not for the FLATTOP-U233, BIG TEN, or ZPR 

benchmarks. The eigenvalues calculated using KENO-VI with continuous energy cross 

sections based on ENDF/B-VII were also outside the 2-sigma range for the same cases. 

Additional calculations were run using KENO-VI with continuous energy ENDF/B-VI cross 

sections and also using MCNP5 with ENDF/B-VI and ENDF/B-VII data. For the FLATTOPU233, BIG TEN, and ZPR benchmarks, none of those calculations was within 2 sigma of the 

experimental values other than the second FLATTOP case using either KENO-VI or MCNP5 

with ENDF/B-VI cross sections. 
69 

Table 4.3. Calculated eigenvalues for the fast system critical benchmarks 

used for validation of the VITAMIN-B7 library 

(All calculations were performed using KENO VI. Uncertainties are 1-sigma values)

ICSBEP identifier CSEWG 

identifier

Experimental 

keff

Cross-section library used in KENO-VI 

VITAMIN-B6 VITAMIN-B7 

Continuous 

energy ENDF/BVII 

PU-MET-FAST-001 Jezebel 1.0000 ± 0.0020 0.9960 ± 0.0002 1.0003 ± 0.0002 0.9991 ± 0.0002

PU-MET-FAST-002 Jezebel-Pu-240 1.0000 ± 0.0020 0.9966 ± 0.0002 1.0005 ± 0.0002 0.9996 ± 0.0002

U233-MET-FAST-001 Jezebel-U-233 1.0000 ± 0.0010 0.9924 ± 0.0002 0.9994 ± 0.0002 0.9993 ± 0.0002

HEU-MET-FAST-001, Case 1 Godiva 1.0000 ± 0.0010 0.9951 ± 0.0002 1.0007 ± 0.0002 0.9999 ± 0.0002

HEU-MET-FAST-001, Case 2 Godiva 1.0000 ± 0.0010 0.9950 ± 0.0002 1.0005 ± 0.0002 0.9996 ± 0.0002

HEU-MET-FAST-028 FLATTOP-U-2351.0000 ± 0.0030 1.0010 ± 0.0002 1.0055 ± 0.0002 1.0029 ± 0.0002

PU-MET-FAST-006 FLATTOP-PU 1.0000 ± 0.0030 1.0019 ± 0.0002 1.0028 ± 0.0002 1.0009 ± 0.0002

U233-MET-FAST-005, Case 1 FLATTOP-U-2331.0000 ± 0.0030 0.9948 ± 0.0002 0.9931 ± 0.0002 0.9924 ± 0.0002

U233-MET-FAST-005, Case 2 FLATTOP-U-2331.0000 ± 0.0030 0.9974 ± 0.0002 0.9907 ± 0.0002 0.9902 ± 0.0002

IEU-MET-FAST-007 BIG TEN 1.0062 ± 0.0003 1.0170 ± 0.0002 1.0097 ± 0.0002 1.0048 ± 0.0002

IEU-COMP-FAST-001 ZPR-6-6A 1.0017 ± 0.0009 1.0003 ± 0.0002 0.9986 ± 0.0002 0.9957 ± 0.0002

MIX-COMP-FAST-001 ZPR-6-7 1.0005 ± 0.0009 1.0000 ± 0.0002 0.9935 ± 0.0002 0.9891 ± 0.0002

4.4 SCALE Regression Tests 

The next set of benchmark experiments are referred to as the “SCALE regression tests.” They 

have been used to test the Monaco Monte Carlo transport code in SCALE and to verify the 

shielding multigroup libraries in SCALE: the 200-group neutron/47-group gamma ENDF/B-VI 

and ENDF/B-VII libraries, as well as the 27-group neutron/19-group gamma ENDF/B-VI and 

ENDF/B-VII libraries. These calculations were run using the configuration-controlled version of 

SCALE at ORNL, which will be released as SCALE 6.1. 

An additional benchmark, the measurement of D-T neutrons through an iron sphere, was added 

to the SCALE regression tests. This experiment was added to provide a more rigorous test of 

the iron cross sections at high energies compared with the experiments using 252Cf sources. 

One experiment from the original regression test set (doses in a three-section concrete 

labyrinth) was deleted. Simulations for the labyrinth experiment are very detailed and timeconsuming and are not very relevant for testing an LWR cross-section library. A summary of 

the five problems that were modeled is provided in Table 4.4. 

For the problems that could be modeled in one dimension, calculations were also run using 

XSDRNPM. 

 
70 

Table 4.4. Summary of the SCALE regression test problems used for 

verification and validation of the VITAMIN-B7 library 

Problem and description Source Materials (trace) Geometry Measured 

1 252Cf leakage spectrum 

 through iron sphere 

252Cf 

n 

Fe (C, S) Sphere n Flux 

 

2 252Cf leakage from 

different iron spheres 

(ALARM-CF-FE-SHIELD-001) 

252Cf 

n, γ

Fe (Mn, C, Cu) Sphere n Flux 

γ flux 

 

3 D-T source transmission 

 through iron sphere 

D-T/D-D 

n 

Fe (Mn, S, P, C) Sphere n Flux 

 

4 252Cf transmission 

through D2O sphere 

252Cf 

n 

D2O (H2O), B, 

concrete, 

polyethylene 

Sphere with 

shields 

n Flux 

 

5 252Cf through various 

shield types and thicknesses 

252Cf 

n 

Paraffin, 

polyethylene, NS-4-

FR, Resin-F, 

KRAFTON-HB, 

Type 304SS 

Slab shields n Dose equiv. 

rate 

4.4.1 Neutron Transmission through an Iron Sphere 

In the early 1990s, several experiments measuring the transmission of 252Cf neutrons through a 

sphere of iron were performed to benchmark ENDF/B-VI cross-section data [22]. Two sets of 

measurements were made—one by the Czechoslovakian National Research Institute and the 

other by the Skoda Company. 

The Monaco model consists of a point source and a spherical shell of iron. A region tally (for a 

thin spherical shell at radius =100 cm) was used to collect the flux information. For the source 

energy distribution ݌ሺܧሻ, the Watt spectrum with constants a = 1.025 MeV and b = 2.926/MeV 

was used. 

݌ሺܧሻ ൌ ܿ݁ିா/௔sinh൫√ܾܧ൯

The measurements listed in [22] were normalized such that the total flux at the detector (at 

r=100 cm) without the iron shield would have been 1 n/cm2

/s. In Monaco, the source strength 

was set to a value of 4(100)2

 = 125664 n/s, which would yield exactly 1 n/cm2/s at 100 cm for 

an unshielded source. No biasing was used, other than implicit capture. The Monaco results 

are shown in Figure 4.1. 
71 

Figure 4.1. Comparison of the Monaco neutron calculation with two 

sets of experimental data for the iron sphere benchmark. 

4.4.2 Neutron and Photon Leakage Spectra from a 252Cf Source at the 

Centers of Six Iron Spheres of Different Diameters 

(ALARM-CF-FE-SHIELD-001) 

Neutron and photon leakage spectra from a 252Cf source surrounded by iron spheres of various 

diameters were measured in Russia during the 1980s. These data and instructions for 

benchmark calculations were presented in 2007 in an Organization for Economic Cooperation 

and Development report [23] with the identification number ALARM-CF-FE-SHIELD-001. 

Thirteen sets of experimental data are presented in the report—one for the bare source and one 

for each of six different diameters of iron shielding. Photon measurements are presented for the 

bare source and five different diameter spheres. Each measured data set was actually the 

difference of two measurements—one with and one without a solid conical shield between the 

source and detector. The conical shields, which were made of borated polyethylene for 

measurements of neutron spectra and lead for measurements of gamma spectra, were 

constructed to completely shadow the detector from radiation coming directly from each 

spherical geometry. The difference in the two sets of measurements effectively removed any 

detector response that was due to room return, and was used as the comparison basis for the 

calculations in [23] and in the present analysis. 

The benchmark geometry consists of only the sphere without any room walls, as shown in 

Figure 4.2. 

1.E-05

1.E-04

1.E-03

1.E-02

1.E-01

1.E+00

1.E+01

1.E-03 1.E-02 1.E-01 1.E+00 1.E+01 1.E+02

n/cm^2/sec/MeV

energy (MeV)

Monaco

NRI

Skoda
72 

Figure 4.2. Benchmark geometry for the iron sphere leakage measurements. 

Monaco was used with the VITAMIN-B7 library to compute the leakage spectra of both neutrons 

and photons according to the benchmark instructions. The benchmark model is onedimensional and does not include the shield that was used to correct for room return. 

The experimentally measured neutron flux spectrum of the bare source was used for the 

Monaco neutron source description. For the Monaco photon source, the experimentally 

measured photon flux spectrum of the bare source was used. Note that the “bare source” 

included copper and steel canisters, so a small fraction of the measured photon flux could have 

come from secondary gammas from neutrons interacting in the source capsule. The two 

strengths of the sources were set using the fact that for every neutron emitted from the 252Cf 

source, 3.82 photons were emitted. 

Benchmark results are listed in the reference document in terms of 4R2

(E)/Q, where R is the 

radius of the detector (three times the radius of the shield, or 60 cm for the bare source) and Q 

is the neutron source strength. Experimental uncertainties ranged from 5% to 30%. The 

experimental values are shown on the following plots (Figure 4.3 through Figure 4.15) as two 

lines, representing the experimental values ± 1σ. 

Since the Monaco source was created from the bare source measurements, the Monaco 

simulation for these should match the experiment well. This agreement is shown in Figure 4.3 

and Figure 4.4. Comparisons of calculated and measured neutron transmission are shown in 

Figure 4.5 through Figure 4.10. Photon comparisons are shown in Figure 4.11 through Figure 

4.15. For the 70-cm-diameter shield, MCNP simulations were also performed. The MCNP 

results are shown in Figure 4.10 and Figure 4.15. 
73 

Figure 4.3. Comparison of the Monaco neutron spectrum with the 

measured spectrum from the bare source of the ALARM-CF-FE-SHIELD001 experiment. 

Figure 4.4. Comparison of the Monaco photon spectrum with the 

measured spectrum from the bare source of the ALARM-CF-FE-SHIELD001 experiment. 

1.E-07

1.E-06

1.E-05

1.E-04

1.E-03

1.E-02

1.E-01

1.E+00

1.E+01

0.001 0.01 0.1 1 10 100

4

πR2

Φ(E)/Q (/MeV)

Energy (MeV)

Monaco

measured + σ

measured - σ

1.E-06

1.E-05

1.E-04

1.E-03

1.E-02

1.E-01

1.E+00

1.E+01

1.E+02

0 2 4 6 8 10 12

4

πR2

Φ(E)/Q (/MeV)

Energy (MeV)

Monaco

measured + σ

measured - σ
74 

.

Figure 4.5. Comparison of the Monaco neutron spectrum with the 

measurement for the 20-cm-diameter shield from the ALARM-CF-FESHIELD-001 experiment. 

Figure 4.6. Comparison of the Monaco neutron spectrum with the 

measurement for the 30-cm-diameter shield from the ALARM-CF-FESHIELD-001 experiment. 

1.E-07

1.E-06

1.E-05

1.E-04

1.E-03

1.E-02

1.E-01

1.E+00

1.E+01

0.001 0.01 0.1 1 10 100

4

πR2

Φ(E)/Q (/MeV)

Energy (MeV)

Monaco

measured + σ

measured - σ

1.E-07

1.E-06

1.E-05

1.E-04

1.E-03

1.E-02

1.E-01

1.E+00

1.E+01

0.001 0.01 0.1 1 10 100

4

πR2

Φ(E)/Q (/MeV)

Energy (MeV)

Monaco

measured + σ

measured - σ
75 

Figure 4.7. Comparison of the Monaco neutron spectrum with the 

measurement for the 40-cm-diameter shield from the ALARM-CF-FESHIELD-001 experiment. 

Figure 4.8. Comparison of the Monaco neutron spectrum with the 

measurement for the 50-cm-diameter shield from the ALARM-CF-FESHIELD-001 experiment. 

1.E-07

1.E-06

1.E-05

1.E-04

1.E-03

1.E-02

1.E-01

1.E+00

1.E+01

0.001 0.01 0.1 1 10 100

4

πR2

Φ(E)/Q (/MeV)

Energy (MeV)

Monaco

measured + σ

measured - σ

1.E-07

1.E-06

1.E-05

1.E-04

1.E-03

1.E-02

1.E-01

1.E+00

1.E+01

0.001 0.01 0.1 1 10 100

4

πR2

Φ(E)/Q (/MeV)

Energy (MeV)

Monaco

measured + σ

measured - σ
76 

Figure 4.9. Comparison of the Monaco neutron spectrum with the 

measurement for the 60-cm-diameter shield from the ALARM-CF-FESHIELD-001 experiment. 

Figure 4.10. Comparison of the Monaco and MCNP neutron spectra 

with the measurement for the 70-cm-diameter shield from the ALARMCF-FE-SHIELD-001 experiment. 

1.E-07

1.E-06

1.E-05

1.E-04

1.E-03

1.E-02

1.E-01

1.E+00

1.E+01

0.001 0.01 0.1 1 10 100

4

πR2

Φ(E)/Q (/MeV)

Energy (MeV)

Monaco

measured + σ

measured - σ

1.E-07

1.E-06

1.E-05

1.E-04

1.E-03

1.E-02

1.E-01

1.E+00

1.E+01

0.001 0.01 0.1 1 10 100

4

πR2

Φ(E)/Q (/MeV)

Energy (MeV)

Monaco

measured + σ

measured - σ

MCNP CE
77 

Figure 4.11. Comparison of the Monaco photon spectrum with the 

measurement for the 30-cm-diameter shield from the ALARM-CF-FESHIELD-001 experiment. 

Figure 4.12. Comparison of the Monaco photon spectrum with the 

measurement for the 40-cm-diameter shield from the ALARM-CF-FESHIELD-001 experiment. 

1.E-06

1.E-05

1.E-04

1.E-03

1.E-02

1.E-01

1.E+00

0 2 4 6 8 10 12

4

πR2

Φ(E)/Q (/MeV)

Energy (MeV)

Monaco

measured + σ

measured - σ

1.E-06

1.E-05

1.E-04

1.E-03

1.E-02

1.E-01

1.E+00

0 2 4 6 8 10 12

4

πR2

Φ(E)/Q (/MeV)

Energy (MeV)

Monaco

measured + σ

measured - σ
78 

Figure 4.13. Comparison of the Monaco photon spectrum with the 

measurement for the 50-cm-diameter shield from the ALARM-CF-FESHIELD-001 experiment. 

Figure 4.14. Comparison of the Monaco photon spectrum with the 

measurement for the 60-cm-diameter shield from the ALARM-CF-FESHIELD-001 experiment. 

1.E-06

1.E-05

1.E-04

1.E-03

1.E-02

1.E-01

1.E+00

0 2 4 6 8 10 12

4

πR2

Φ(E)/Q (/MeV)

Energy (MeV)

Monaco

measured + σ

measured - σ

1.E-06

1.E-05

1.E-04

1.E-03

1.E-02

1.E-01

1.E+00

0 2 4 6 8 10 12

4

πR2

Φ(E)/Q (/MeV)

Energy (MeV)

Monaco

measured + σ

measured - σ
79 

Figure 4.15. Comparison of the Monaco and MCNP photon spectra 

with the measurement for the 70-cm-diameter shield from the ALARMCF-FE-SHIELD-001 experiment. 

4.4.3 D-T Neutrons through an Iron Sphere 

Integral measurements were made of the neutron flux from a D-T (with 5% D-D) neutron source 

transmitted through an iron spherical shell with a thickness of 30.45 cm [24]. This experiment at 

the University of Illinois in Urbana was also performed with a 252Cf source and was included as a 

SINBAD benchmark experiment [25]. Some energy-dependent neutron flux data and computer 

simulation data were also presented in these references. The idealized problem geometry is 

shown in Figure 4.16. 

Figure 4.16. Geometry of the Urbana D-T transmission experiment. 

1.E-06

1.E-05

1.E-04

1.E-03

1.E-02

1.E-01

1.E+00

0 2 4 6 8 10 12

4

πR2

Φ(E)/Q (/MeV)

Energy (MeV)

Monaco

measured + σ

measured - σ

MCNP CE
80 

The D-T (with 5% D-D) neutron source spectrum was given and used as a source in Monaco. 

The measurements reported in [22] were normalized such that the total flux at 200 cm without 

the iron sphere would have been 1 n/cm2

/s. The Monaco source strength was set to 

4(200)2

 = 502654 n/s so that results at 200 cm would be 1 n/cm2/s for an unshielded 

measurement. A region tally was used to collect flux information at 200 cm. No variance 

reduction, other than implicit capture, was used. The comparison with the measured values is 

shown in Figure 4.17. At high energies, calculation and the measurement agree well. At 

energies less than 10 MeV, the simulation underpredicts the measurement. This 

underprediction is consistent with the results of the simulations in the original references. 

Figure 4.17. Comparison of the Monaco neutron spectrum at 200 cm 

with the measured data for the Urbana experiment. 

 

0.0001

0.001

0.01

0.1

1

1 10

Leakage/source/MeV

neutron energy (MeV)

Experiment

MAVRIC
81 

4.4.4 Neutron Transmission through a Heavy Water Sphere 

The transmission of 252Cf neutrons through a sphere filled with heavy water was measured in 

Prague in the mid-1990s [26]. For each experiment, two measurements were made: one with 

an iron/polyethylene shield and one without the shield. The purpose of the shield was to block 

all direct particle transmission from the sphere to the detector. The difference in these two 

measurements accounts for scatter from the floor, walls, and ceiling (which is about a 5% effect 

for energies above 10 keV). A great amount of detail for the materials and geometry of the 

source holder, insertion tube, and detectors is provided in [26]. Two different assemblies 

holding the californium source were measured. The geometry, with the shield in place, is shown 

in Figure 4.18. 

Figure 4.18. Geometry of the heavy water experiment. 

The Monaco model for this benchmark consists of two nested spheres for the source and the 

heavy water. For the experiments where the iron/polyethylene shield is present, three borated 

polyethylene cylinders and one iron cylinder are placed between the sphere and the detector 

position at (75,0,0), where the origin of the coordinate system corresponds to the source at the 

center of the heavy water sphere. The experiment sat 2 m above the floor of an experimental 

hall that measured 10 × 13 × 25 m. Neither of the two source holder assemblies was modeled 

in any detail. 

For the source energy distribution, the Watt spectrum with constants a =1.025 MeV and 

b =2.926/MeV was used. 

݌ሺܧሻ ൌ ܿ݁ିா/௔sinh൫√ܾܧ൯

The measurements listed in the paper were normalized such that the total flux at the detector 

(at r = 75 cm) without the D2O sphere or the shields would have been 1 n/cm2

/s. In Monaco, the 

source strength was set to a strength of 4(75)2

 = 70686 n/s. No biasing was used, other than 

implicit capture. 
82 

A point detector tally at the detector location was used to calculate the energy-dependent 

neutron flux. Results are shown in Figure 4.19 for the Monaco simulation compared with the 

measurements using the two different source assemblies. 

Figure 4.19. Comparison of the Monaco calculation and measured 

data for the heavy water transmission benchmark. 

4.4.5 Ueki Shielding Measurements 

K. Ueki et al. of the Nuclear Technology Division, Ship Research Division, in Japan performed a 

series of simple studies on a variety of shielding materials layered in different combinations [27]. 

These measurements used both neutron (252Cf) and photon (60Co) sources to investigate the 

shielding effectiveness of steel, graphite, and many hydrogenous materials as single shields or 

in combinations. The detector used was an Aloka moderator-type survey meter, made of 

polyethylene, which is supposed to mimic human dose response. The data were corrected for 

room scatter. Three types of experiments were performed: a single shield material (type 1), 

two-layer steel/polyethylene combinations (type 2), and three-layer steel/polyethylene 

combinations (type 3). The geometry for a type 3 experiment is shown in Figure 4.20. 

MAVRIC was used to simulate the experiments that used the neutron source. For penetration 

through thick shields, the MAVRIC sequence and the CADIS methodology [28] can be used to 

calculate results with low relative uncertainties in short times. For each of the 46 shield 

configurations that Ueki measured, an input file was constructed that utilized the advanced 

variance reduction. For these simulations, an adjoint calculation using Denovo was performed 

using a broad-group version of the VITAMIN-B7 library. An importance map and a biased 

1.E-04

1.E-03

1.E-02

1.E-01

1.E+00

1.E+01

1.E-02 1.E-01 1.E+00 1.E+01

Flux per unit Energy

Energy (MeV)

Assembly A8

Assembly A25

Monaco
83 

source distribution were then calculated by MAVRIC from the adjoint fluxes. The Monaco 

simulation used the VITAMIN-B7 fine-group shielding library, the broad-group importance map, 

and a biased source distribution (constructed from a combination of the fine-group 

representation of the original source and the broad-group importances). 

Figure 4.20. Geometry of a type 3 configuration from the neutron 

shielding experiments of Ueki et al. Half of the paraffin block surrounding 

the source has been removed from the model to show the source and the 

conical cutout. 

The MAVRIC calculations for the type 1 experiments (single shield) are summarized in Table 

4.5, and the ratios of the calculated to measured values are shown in Figure 4.21. All of the 

calculated values are within ±20% of the measured data, with the exception of polyethylene. 

Reported results with MCNP also show good agreement with the findings of Ueki et al. for all 

materials except polyethylene. It is not clear if this mismatch is related to thermal scattering 

cross sections common to both MAVRIC and MCNP or if may be related to the way in which 

Ueki processed his experimental results. 

Comparisons of the MAVRIC calculations and measured data for the type 2 and type 3 

experiments are shown in Figure 4.22, Figure 4.23, and Figure 4.24. The poor agreement is 

probably due to the polyethylene component that MAVRIC could not match in the type 1 

experiments. 

 
84 

Table 4.5. Summary of MAVRIC results for the Ueki type 1 

experiments with a 252Cf neutron source 

 

 MAVRIC 

shield Ueki Runtime (min) 

Neutron DoseEquivalent 

(μSv/h) 

mat cm (μSv/h) 

Adj. 

Sn MC value 

rel 

unc C/E 

none 0 6.90E+02 1.3 10.3 6.19E+02 0.09% 0.90 Polyethylene 

5 2.98E+02 13.0 42.0 3.01E+02 0.59% 1.01 

10 1.12E+02 11.3 62.1 1.27E+02 0.99% 1.14 

15 4.38E+01 12.6 91.9 5.48E+01 1.14% 1.25 

20 1.81E+01 38.9 181.3 2.37E+01 0.99% 1.31 

25 7.98E+00 39.0 241.4 1.11E+01 0.99% 1.40 

30 3.46E+00 38.3 241.2 5.32E+00 1.12% 1.54 

NS-4-FR 

5 3.43E+02 1.2 10.1 3.14E+02 0.40% 0.91 

10 1.42E+02 1.3 10.6 1.40E+02 0.58% 0.98 

15 6.09E+01 1.3 10.7 6.20E+01 0.67% 1.02 

20 2.79E+01 1.3 10.7 2.83E+01 0.71% 1.02 

25 1.27E+01 1.2 10.2 1.28E+01 0.74% 1.00 

30 1.2 10.8 6.13E+00 0.74% 

Resin-F 

5 3.43E+02 1.1 10.2 3.32E+02 0.38% 0.97 

10 1.51E+02 1.2 10.3 1.54E+02 0.56% 1.02 

15 6.65E+01 1.3 10.7 7.22E+01 0.66% 1.09 

20 3.04E+01 1.3 10.6 3.42E+01 0.71% 1.12 

25 1.38E+01 1.3 10.8 1.63E+01 0.72% 1.18 

30 6.89E+00 1.3 10.4 7.94E+00 0.74% 1.15 

KRAFTON-HB 

5.3 3.04E+02 1.2 10.4 2.96E+02 0.44% 0.97 

11 1.27E+02 1.2 10.2 1.26E+02 0.64% 0.99 

16 5.58E+01 1.2 10.4 5.35E+01 0.84% 0.96 

21 2.55E+01 1.2 10.3 2.38E+01 0.87% 0.93 

27 1.22E+01 1.3 10.2 1.05E+01 0.86% 0.86 

32 6.09E+00 1.3 10.8 5.04E+00 0.85% 0.83 

SS-304 

5 4.87E+02 1.4 10.4 4.73E+02 0.33% 0.97 

10 3.62E+02 1.5 10.5 3.49E+02 0.57% 0.96 

15 2.67E+02 1.6 10.2 2.52E+02 0.76% 0.95 

20 1.95E+02 1.7 10.1 1.77E+02 0.89% 0.91 

25 1.42E+02 1.8 10.1 1.22E+02 0.98% 0.86 

30 1.03E+02 2.0 15.4 8.47E+01 0.86% 0.82 
85 

Figure 4.21. MAVRIC C/E ratios for the neutron dose rates from the 

Ueki type 1 experiments for single material shields of a given thickness.

 Figure 4.22. MAVRIC C/E ratios 

for the neutron dose rates from the 

type 2a experiments, which consist 

of 25 cm of SS-304 followed by T cm 

of polyethylene.

Figure 4.23. MAVRIC C/E ratios for 

the neutron dose rates from the type 

2b experiments, which consist of T

cm of polyethylene followed by 25 cm 

of SS-304.

0.80

0.90

1.00

1.10

1.20

1.30

1.40

1.50

1.60

0 5 10 15 20 25 30 35

MAVRIC/Ueki C/E

Thickness of Shielding Material (cm)

Polyethylene

NS-4-FR

Resin-F

KRAFTON-HB

SUS-304

0.30

0.40

0.50

0.60

0.70

0.80

0.90

0 5 10 15 20 25

MAVRIC/Ueki C/E

Thickness T in cm

0.80

0.90

1.00

1.10

1.20

1.30

1.40

0 5 10 15 20 25

MAVRIC/Ueki C/E

Thickness T in cm
86 

Figure 4.24. MAVRIC C/E ratios for the neutron dose rates from the 

type 3 experiments, which consist of t cm of SS-304, 15 cm 

ofpolyethylene, and (25-t) cm of SS-304. 

4.4.6 XSDRN Results 

The SCALE regression tests described in Sections 4.4.1 through 4.4.3 are essentially one 

dimensional. XSDRN inputs were constructed for these experiments and compared with the 3D 

MAVRIC results. In general, the results from the XSDRN calculations were essentially identical 

to the MAVRIC results. 

4.4.7 Summary of SCALE Regression Tests 

The VITAMIN-B7 library performed well for nearly all the XSDRN and MAVRIC simulations of 

the various experiments. The only exceptions were for the calculated photon spectra for several 

of the iron spheres from the ALARM-CF-FE-SHIELD-001 benchmark and the Ueki experiments 

that contained polyethylene in the shields. 

0.60

0.70

0.80

0.90

1.00

1.10

1.20

0 5 10 15 20 25

MAVRIC/Ueki C/E

Thickness t in cm
87 

4.5 Light Water Reactor Pressure Vessel Dosimetry 

Benchmarks 

4.5.1 H.B. Robinson Unit 2 Pressure Vessel Benchmark 

H. B. Robinson Unit 2 (HBR-2) is a 2300 MWt PWR designed by Westinghouse and placed into 

operation in March 1971. The HBR-2 core consists of 157 fuel elements with a 15 × 15 lattice. 

The core is enclosed by a core baffle, core barrel, thermal shield, pressure vessel, and 

biological shield. 

In [29] dosimetry activities were calculated for a thermal shield surveillance capsule at an 

azimuthal angle of 20o and a reactor cavity location at 0o based on power distribution data for 

fuel cycle 9. The DORT discrete-ordinates radiation transport code was used to calculate 

one-dimensional (1D) and two-dimensional (2D) neutron flux distributions. A flux synthesis 

method was used to construct a three-dimensional (3D) flux distribution from the 1D and 2D 

transport solutions [30]. The synthesized flux was used with dosimetry cross sections from the 

CROSS-95 dosimetry library [31] to calculate reaction rates for the following dosimetry 

reactions: 237Np(n,f)137Cs, 238U(n,f)137Cs, 58Ni(n,p)58Co, 54Fe(n,p)54Mn, 46Ti(n,p)46Sc, and 

63Cu(n,)60Co. Details of the analysis are provided in [29]. 

For this benchmark comparison, the calculations described in [29] were updated using the 

BUGLE-B7 library. The calculated activities increased by approximately 2.5% to 4.5% for the 

threshold [(n,p) and n,)] reactions. The activities for the fission reactions increased slightly at 

the capsule location and were essentially unchanged at the reactor cavity location. In all cases, 

the increases provide improved agreement with the measured data. The calculated-toexperiment (C/E) values for calculations using the BUGLE-96 and BUGLE-B7 libraries are 

presented in Table 4.6. 

Table 4.6. Calculated-to-experiment ratios (C/E) for dosimetry activities from 

H. B. Robinson Unit 2, Cycle 9 

Dosimetry 

location and 

cross-section 

library 

Dosimetry reaction 

Average 237 a Np(n,f) 

137Cs 

238U(n,f)

137Cs 

58Ni(n,p)

58Co 

54Fe(n,p)

54Mn 

46Ti(n,p)

46Sc 

63Cu(n,) 60Co 

Capsule 

BUGLE-96 0.90 

(0.92)b

0.85 

(0.89) 

0.96 0.93 0.85 0.90 

(0.93) 

0.90 ± 0.04 

(0.91 ± 0.04)

BUGLE-B7 0.91 

(0.93) 

0.86 

(0.91) 

0.98 0.96 0.88 0.94 

(0.96) 

0.92 ± 0.05 

(0.94 ± 0.04)

 

Cavity 

BUGLE-96 0.58 

(0.61) 

0.74 

(0.82) 

0.97 0.96 0.90 0.93 

(0.96) 

0.90 ± 0.09 

(0.92 ± 0.06)

BUGLE-B7 0.58 

(0.61) 

0.74 

(0.83) 

1.00 0.99 0.94 0.97 

(1.00) 

0.93 ± 0.11 

(0.95 ± 0.07) a The average C/M and standard deviation values exclude the 237Np reaction at the cavity location, which is 

suspected of having an inaccurate measurement. This exclusion is consistent with the data reduction performed 

in [29]. b The ratio values in parentheses for the 237Np, 238U, and 63Cu reactions include the effect of adjustments to the 

measured data. These adjustments, described on page 11 of [29], are intended to compensate for photofission in 

the 237Np and 238U dosimeters and for Co impurities in the 63Cu dosimeter. The measured 137Cs activities in the 237Np dosimeters were reduced by 2.5% and 5% at the capsule and cavity locations, respectively. The 

corresponding reduction factors for the 238U dosimeters were 5% and 10%. The measured 60Co activities in the 63Cu dosimeters were reduced by 2.5% at both locations.
88 

4.5.2 Oak Ridge National Laboratory Pool Critical Assembly (PCA) 

The ORNL Pool Critical Assembly (ORNL PCA) pressure vessel wall benchmark facility 

(PVWBF) is one of the most widely used benchmarks for the qualification of radiation transport 

methods for LWR applications. The purpose of this benchmark was to validate the capabilities 

of calculational methods to predict the reaction rates in the region outside a reactor core when 

the neutron source, material compositions, and geometry are well defined. Analysis of the PCA 

benchmark can be used for partial fufillment of the requirements for the qualification of the 

methodology for pressure vessel neutron fluence calculations, as required by Regulatory Guide 

1.190. Details of the benchmark can be found in [32] and will not be repeated here except to 

point out differences with previous models and benchmarks. 

The PCA is a small critical assembly composed of material test reactor (MTR) fuel elements 

(approximately 3” by 3” plates) in a water pool. The benchmark experiment consisted of the 

PCA reactor core (nominally 10 kW) and the components used to mock up the core-to-cavity 

region in LWRs. These components were the thermal shield, the pressure vessel simulator, 

and the void box, which simulated the reactor cavity. An overall view of the PCA core as 

modeled in KENO-VI geometry is shown in Figure 4.25, while an overall view of the entire 

PVWBF in KENO-VI geometry is shown in Figure 4.26. An aluminum plate, referred to as the 

reactor window simulator, was added to the facility for operational reasons. The thickness of 

the water gaps between the aluminum window and thermal shield and that between the thermal 

shield and the pressure vessel steel block was varied by the experimental run. 

The set of measurements analyzed in this evaluation is the 12/13 (12 cm between the aluminum 

window and the thermal shield and 13 cm between the thermal shield and the pressure vessel 

steel block). The PCA 12/13 configuration is geometrically similar to the thermal shield–

downcomer–pressure vessel design that is typical of many PWRs. In particular, the 12-cm 

water gap on the core side of the thermal shield, the 13-cm water gap between the thermal 

shield and the pressure vessel simulator, the 6-cm-thick stainless steel thermal shield, the 

22.5-cm-thick low-alloy-steel pressure vessel, and the simulated reactor cavity (void box) 

positioned behind the pressure vessel mockup retain the characteristics of a PWR. 

The PCA source was composed of 25 MTR new fuel elements with 93% 235U enrichment. The 

power distribution of the fuel elements was obtained from fission chamber measurements and 

calculations and was provided as a two-dimensional distribution function as a part of the 

benchmark specification. These source data values were not utilized in the present analysis, 

because the computational sequence described below is capable of generating the critical 

reactor source term. 

During the PCA experiments, measurements were taken at several locations within the mockup 

via vertically oriented experimental access tubes. The tube locations, which are described in 

detail in [32], provide transverse data extending from the reactor core outward through the 

pressure vessel simulator and into the void box. All of the foil measurements were conducted at 

locations in the experiment tubes at the core midplane elevation. 

The measurements provide sufficient data to generate comparisons of measurements to 

calculations throughout the entire 12/13 configuration. Data from experiment tube locations A4, 

A5, and A6 establish the means for verification of calculated flux gradients within the pressure 

vessel wall itself. Since measurements at operating reactors can provide data only in the 

downcomer region internal to the pressure vessel or in the reactor cavity external to the vessel 

wall, the PCA data points located inside the thick-walled vessel establish a key set of 

comparisons to aid in the accurate determination of flux gradients within the pressure vessel 

wall. 
89 

Figure 4.25 shows a view of the MTR fuel assemblies used in the PCA core, with the safety and 

regulation rods fully inserted. The standard elements (shown in blue) consist of 18 MTR fuel 

plates. The elements shown in green contain 19 fuel plates, which was the standard fuel 

element design in the Oak Ridge Research Reactor. Note that the individual plates with the fuel 

meat interior region are modeled. This eliminates the need for using homogenized assemblies 

as were used in previous benchmark studies. 

Figure 4.26 shows the complete SCALE Generalized Geometry Package (SGGP) model used 

in the KENO-VI criticality source calculations and the MAVRIC/Monaco shielding calculations. 

The pool water has been omitted from the picture for illustration purposes. The vertical 

experiment tubes centered along the midpoint of the PCA core are numbered A1–A8, with A1 

being the tube closest to the core and A8 designating the largest tube, which is outside the void 

box. These tubes contained the foils and fission detectors used to experimentally measure the 

fluxes at the A1–A7 locations as given in Table 1.6 of [32]. 

The procedure for calculating the fluxes and reaction rates in this evaluation was as follows: 

1. An initial KENO-VI calculation was run to determine the critical rod positions. 

2. The KENO-VI calculation was rerun specifying a mesh to tally the spatial- and energydependent source particle distributions for subsequent transport using the Monaco code 

in the MAVRIC sequence. This calculation used a much larger number of particles than 

a standard keff calculation, because the objective was to obtain a well-converged source 

distribution to use in subsequent MONACO/Mavric calculations. 

3. The MAVRIC sequence was run using the source mesh file for the source particles in 

conjunction with an importance map and biased source generated with adjoint fluxes 

calculated by the Denovo discrete-ordinates code. This automated variance reduction 

process provides consistent weight windows and biased source distributions for 

transport of the core neutrons to the desired locations (experimental tubes A1–A7). 

Calculation of the reaction rates of interest is discussed in Section 4.5.2.1. 

4.5.2.1 Comparison of Calculated Reaction Rates with Experimental Results 

A number of activation foils were used in the PVWBF experiments to calculate the flux. To 

minimize potential errors and provide a common basis for the benchmark exercises, the 

measured reaction rates were converted to equivalent fission fluxes, which are defined as 

235

( )( )

() ()

i

fe

i

E E dE

E E dE

  

   



where i(E) = the dosimetry cross section for the reaction being considered, 

 (E) = the neutron flux at the dosimetry location, and 

 235(E) = the 235U fission spectrum 

 

The benchmark evaluation lists the foils that were used for radiometric measurements of 

equivalent fission fluxes. The reactions are 237Np(n,f)137Cs, 238U(n,f)137Cs, 103Rh(n,n’)103mRh, 

115In(n,n’)115mIn, 58Ni(n,p)58Co, and 27Al(n,)24Na. For the 103Rh and 115In reactions, the reaction 

rates in the numerator were calculated using dosimetry cross sections from [17] for the 

response function in the MAVRIC sequence. Note that for those two inelastic scattering 

reactions, there is no single ENDF reaction type (MT value) which can be used to compute the 

production rate of the metastable isomer. For the other four reactions, the appropriate reaction 

cross sections from the transport library (either VITAMIN-B6 or VITAMIN-B7) were used as the 

reponse functions. The fission-spectrum-averaged cross sections in the denominator were 
90 

taken from Table 1.6 of [32]. These values were also calculated using the reaction cross 

sections used in the numerator and the 235U fission spectra from the VITAMIN-B6 and 

VITAMIN-B7 libraries, and shown to be in very good agreement with the values listed in [32]. 

The measured and calculated fission equivalent fluxes for the six reactions listed above are 

provided in Table 4.7 through Table 4.12. The calculated values have relative uncertainties of 

less than ~1% at the 1 level for the fission and inelastic scattering reactions, and less than 

~2% for the 58Ni and 27Al reactions. 

Figure 4.25. Close-up of the SCALE model of the PCA core. 

Figure 4.26. SCALE model of the pressure vessel wall benchmark 

facility (water omitted). 
91 

Table 4.7. 237Np(n,f)137Cs results for the PVWBF benchmark analysis

Experiment 

tube 

position 

Distance 

from 

core 

(cm) 

Experiment

al 

fission 

equivalent 

flux 

Monaco/MAVRIC

fission equivalent 

flux using 

VITAMIN-B6 

Library 

VITAMIN-B6 

C/E ratio 

Monaco/MAVRIC 

fission equivalent 

flux using 

VITAMIN-B7 

Library 

VITAMIN-B7 

C/E ratio 

A1 12.0 6.64E-06 6.55E-06 0.99 6.92E-06 1.04 

A3 29.7 2.27E-07 2.51E-07 1.11 2.66E-07 1.17 

A4 39.5 9.27E-08 9.04E-08 0.98 9.69E-08 1.05 

A5 44.7 5.18E-08 5.04E-08 0.97 5.37E-08 1.04 

A6 50.1 2.70E-08 2.59E-08 0.96 2.78E-08 1.03 

A7 59.1 7.25E-09 7.51E-09 1.04 8.12E-09 1.12 

Table 4.8. 238U(n,f)137Cs results for the PVWBF benchmark analysis 

Experiment 

tube no. 

Distance 

from 

core (cm) 

Experimenta

l 

fission 

equivalent 

flux 

Monaco/MAVRIC

fission equivalent 

flux using 

VITAMIN-B6 

library 

VITAMIN-B6

C/E ratio 

Monaco/MAVRIC 

fission equivalent 

flux using 

VITAMIN-B7 

library 

VITAMIN-B7

C/E ratio 

A4 39.5 6.11E-08 5.88E-08 0.96 6.07E-08 0.99 

A5 44.7 2.74E-08 2.60E-08 0.95 2.64E-08 0.96 

A6 50.1 1.12E-08 1.06E-08 0.95 1.07E-08 0.96 

Table 4.9. 103Rh(n,n’)103mRh results for the PVWBF benchmark analysis 

Experiment 

tube 

position 

Distance 

from 

core (cm) 

Experimental 

fission 

equivalent 

flux 

Monaco/MAVRIC

fission equivalent 

flux using 

VITAMIN-B6 

library 

VITAMIN-B6 

C/E ratio 

Monaco/MAVRIC 

fission equivalent 

flux using 

VITAMIN-B7 

library 

VITAMIN-B7

C/E ratio 

A1 12.0 5.54E-06 5.91E-06 1.07 6.01E-06 1.09 

A4 39.5 7.74E-08 8.34E-08 1.08 8.58E-08 1.11 

A5 44.7 4.35E-08 4.54E-08 1.04 4.63E-08 1.06 

A6 50.1 2.19E-08 2.29E-08 1.05 2.35E-08 1.07 

Table 4.10. 115In(n,n’)115mIn results for the PVWBF benchmark analysis 

Experiment 

tube 

position 

Distance 

from 

core (cm) 

Experimental 

fission 

equivalent 

flux 

Monaco/MAVRIC

fission equivalent 

flux using 

VITAMIN-B6 

library 

VITAMIN-B6

C/E ratio 

Monaco/MAVRIC 

fission equivalent 

flux using 

VITAMIN-B7 

library 

VITAMIN-B7

C/E ratio 

A1 12.0 5.61E-06 5.80E-06 1.03 5.92E-06 1.06 

A2 23.8 6.06E-07 6.43E-07 1.06 6.49E-07 1.07 

A3 29.7 1.99E-07 2.22E-07 1.11 2.26E-07 1.14 

A4 39.5 5.87E-08 6.48E-08 1.10 6.63E-08 1.13 

A5 44.7 2.76E-08 3.03E-08 1.10 3.08E-08 1.12 

A6 50.1 1.17E-08 1.33E-08 1.13 1.34E-08 1.15 
92 

Table 4.11. 58Ni(n,p)58Co results for the PVWBF benchmark analysis 

Experiment 

tube 

position 

Distance 

from 

core (cm) 

Experimental 

fission 

equivalent 

flux 

Monaco/MAVRIC

fission equivalent 

flux using 

VITAMIN-B6 

library 

VITAMIN-B6 

C/E ratio 

Monaco/MAVRIC 

fission equivalent 

flux using 

VITAMIN-B7 

library 

VITAMIN-B7

C/E ratio 

A1 12.0 5.83E-06 5.78E-06 0.99 5.89E-06 1.01 

A2 23.8 6.18E-07 6.31E-07 1.02 6.39E-07 1.03 

A3 29.7 2.31E-07 2.48E-07 1.07 2.53E-07 1.10 

A4 39.5 5.30E-08 5.34E-08 1.01 5.50E-08 1.04 

A5 44.7 2.09E-08 2.15E-08 1.03 2.19E-08 1.05 

A6 50.1 7.43E-09 8.22E-09 1.11 8.26E-09 1.11 

Table 4.12. 27Al(n,)

23Na results for the PVWBF benchmark analysis 

Experiment 

tube 

position 

Distance 

from 

core (cm) 

Experimental 

fission 

equivalent 

flux 

Monaco/MAVRIC

fission equivalent 

flux using 

VITAMIN-B6 

library 

VITAMIN-B6

C/E ratio 

Monaco/MAVRIC 

fission equivalent 

flux using 

VITAMIN-B7 

library 

VITAMIN-B7

C/E ratio 

A1 12.0 7.87E-06 7.73E-06 0.98 7.74E-06 0.98 

A2 23.8 1.02E-06 1.00E-06 0.98 1.03E-06 1.01 

A3 29.7 4.48E-07 4.56E-07 1.02 4.60E-07 1.03 

A4 39.5 1.02E-07 9.97E-08 0.98 1.01E-07 0.99 

A5 44.7 4.10E-08 4.08E-08 0.99 4.19E-08 1.02 

A6 50.1 1.54E-08 1.61E-08 1.05 1.65E-08 1.07 

4.5.3 CEN/SK Laboratory VENUS-3 Benchmark 

The VENUS Critical Facility is a low power (approximately 650 W) reactor located at SCK/CEN 

in Belgium. The VENUS-3 benchmark experiment [33 and 34] was a joint effort between ORNL, 

under the direction of the NRC, and the Belgian regulatory authority. The objective was to study 

the applicability of 1D and 2D flux synthesis methods for configurations in which partial length 

shield assemblies (PLSAs) are used in a reactor. In the VENUS-3 PLSAs, some of the UO2 fuel 

was replaced with stainless steel in fuel rods near the periphery of the core to reduce the 

neutron fluence at the pressure vessel. 

The VENUS-3 benchmark used fuel assemblies with 4% and 3.3% enriched 235U UO2 fuel rods, 

with some of the 3.3% enriched fuel replaced by PLSAs. The standard fuel rods had a fuel 

length of 50 cm. In the PLSAs, the lower 25 cm of fuel was replaced by 25 cm of stainless 

steel. The fuel rods were spaced with a 1.26 cm pitch, which is typical of commonly used 17 × 

17 PWR fuel assemblies. 

In the southwest quadrant of the VENUS-3 experiment, a mockup of the internal structure of a 

reactor vessel was placed next to the VENUS reactor. The mockup structure, which started at 

the fuel assemblies and extended outward toward the pressure vessel, consisted of an outer 

core baffle, core barrel, water gap, neutron pad, jacket, and pressure vessel. A picture of the 

southwest quadrant of the VENUS benchmark geometry, as constructed with the SGGP, is 

shown in Figure 4.27. (All water has been omitted from the figure for clarity.) In Figure 4.27, 

the PLSAs are the rods where the fuel in the upper half of the rods is colored bright green and 

the stainless steel in the lower half of the rods is colored blue. To the right of the PLSAs in 

Figure 4.27 are the outer core baffle (dark blue), the core barrel (white), the neutron pad (red), 

and the jacket and pressure vessel (magenta). 
93 

During the VENUS-3 benchmark, reaction rates were measured at 30 different radial locations. 

At each of these radial locations there were up to 14 different measurements in the axial 

direction. At each of these axial positions up to 3 activation reactions were measured. These 

activation reactions were measured inside the inner baffle, outer baffle, core barrel, water gap, 

PLSA rods, and 3.3% enriched fuel rods in the southwest quadrant of the reactor. The 

activation reactions that were measured are: 

1. 58Ni(n,p)58Co 

2. 115In(n,n’)115mIn 

3. 27Al(n,α)

24Na 

The benchmark consists of 38 27Al(n,α)

24Na measurements, 104 115In(n,n’)115mIn measurements, 

and 242 58Ni(n,p)58Co measurements. The measured reaction rates were not reported in the 

benchmark. Instead the equivalent fission fluxes (see Section 4.5.2.1) were reported. 

For the present evaluation, the TORT [35] discrete ordinates radiation transport code was used 

to model the VENUS-3 benchmark. Calculations were run using the BUGLE-96 and BUGLE-B7 

libraries. For each dosimetry reaction, equivalent fission fluxes were calculated using the flatweighted response function data in the BUGLE libraries. The fission-spectrum-averaged cross 

sections were taken from [34]. 

Figure 4.27. Southwest quadrant of the VENUS-3 benchmark geometry 

as modeled in the SGGP (water omitted). 

 
94 

Table 4.13. Calculated-to-Measured (C/M) Ratios for the VENUS-3 Benchmark Experiment 58Ni(n,p)58Co Equivalent Fission Fluxes using TORT with BUGLE-96 and BUGLE-B7 Cross-Sections 

 Measured TORT with BUGLE-96 TORT with BUGLE-B7 

Measurement Position (cm) Equivalent Equivalent Equivalent 

Zone X Y Z* Fission Flux** Fission Flux C/M Fission Flux C/M 

 

3.3% Fuel -29.61 -0.63 106.05 9.316E+08 8.82E+08 0.95 8.85E+08 0.95 

 110.15 1.282E+09 1.28E+09 1.00 1.29E+09 1.00 

 114.25 1.644E+09 1.56E+09 0.95 1.57E+09 0.95 

 118.35 1.986E+09 1.88E+09 0.95 1.89E+09 0.95 

 122.45 2.276E+09 2.27E+09 1.00 2.27E+09 1.00 

 125.55 2.494E+09 2.45E+09 0.98 2.46E+09 0.99 

 128.65 2.683E+09 2.68E+09 1.00 2.69E+09 1.00 

 131.75 2.860E+09 2.81E+09 0.98 2.82E+09 0.99 

 134.85 2.925E+09 2.87E+09 0.98 2.87E+09 0.98 

 137.95 2.855E+09 2.82E+09 0.99 2.83E+09 0.99 

 141.05 2.744E+09 2.69E+09 0.98 2.69E+09 0.98 

 145.15 2.406E+09 2.38E+09 0.99 2.38E+09 0.99 

 149.25 1.967E+09 1.95E+09 0.99 1.95E+09 0.99 

 153.35 1.473E+09 1.57E+09 1.06 1.57E+09 1.07 

 

3.3% Fuel -27.09 0.63 106.05 1.141E+09 1.09E+09 0.96 1.10E+09 0.96 

 110.15 1.589E+09 1.60E+09 1.01 1.60E+09 1.01 

 114.25 2.014E+09 1.93E+09 0.96 1.94E+09 0.96 

 118.35 2.424E+09 2.33E+09 0.96 2.34E+09 0.97 

 122.45 2.774E+09 2.77E+09 1.00 2.78E+09 1.00 

 125.55 2.988E+09 2.97E+09 1.00 2.99E+09 1.00 

 128.65 3.160E+09 3.17E+09 1.00 3.19E+09 1.01 

 131.75 3.267E+09 3.25E+09 1.00 3.27E+09 1.00 

 134.85 3.295E+09 3.28E+09 0.99 3.29E+09 1.00 

 137.95 3.218E+09 3.21E+09 1.00 3.22E+09 1.00 

 141.05 3.046E+09 3.03E+09 1.00 3.04E+09 1.00 

 145.15 2.679E+09 2.66E+09 0.99 2.67E+09 1.00 

 149.25 2.195E+09 2.18E+09 0.99 2.19E+09 1.00 

 153.35 1.651E+09 1.75E+09 1.06 1.76E+09 1.07 

 

3.3% Fuel -30.87 -3.15 106.05 7.661E+08 7.21E+08 0.94 7.25E+08 0.95 

 110.15 1.057E+09 1.04E+09 0.98 1.04E+09 0.99 

 114.25 1.350E+09 1.27E+09 0.94 1.27E+09 0.94 

 118.35 1.620E+09 1.53E+09 0.95 1.54E+09 0.95 

 122.45 1.876E+09 1.84E+09 0.98 1.85E+09 0.99 

 125.55 2.067E+09 2.03E+09 0.98 2.03E+09 0.98 

 128.65 2.280E+09 2.30E+09 1.01 2.31E+09 1.01 

 131.75 2.527E+09 2.51E+09 0.99 2.51E+09 0.99 

 134.85 2.628E+09 2.60E+09 0.99 2.61E+09 0.99 

 137.95 2.622E+09 2.58E+09 0.99 2.59E+09 0.99 

 141.05 2.495E+09 2.46E+09 0.99 2.47E+09 0.99 

 145.15 2.209E+09 2.19E+09 0.99 2.19E+09 0.99 

 149.25 1.829E+09 1.79E+09 0.98 1.79E+09 0.98 

 153.35 1.376E+09 1.44E+09 1.05 1.45E+09 1.05 

 
95 

Table 4.13. Calculated-to-Measured (C/M) Ratios for the VENUS-3 Benchmark Experiment 58Ni(n,p)58Co Equivalent Fission Fluxes using TORT with BUGLE-96 and BUGLE-B7 Cross-Sections 

 Measured TORT with BUGLE-96 TORT with BUGLE-B7 

Measurement Position (cm) Equivalent Equivalent Equivalent 

Zone X Y Z* Fission Flux** Fission Flux C/M Fission Flux C/M 

 

3.3% Fuel -0.63 -32.13 106.05 1.285E+09 1.25E+09 0.97 1.25E+09 0.97 

 110.15 1.768E+09 1.80E+09 1.02 1.80E+09 1.02 

 114.25 2.210E+09 2.16E+09 0.98 2.17E+09 0.98 

 118.35 2.606E+09 2.56E+09 0.98 2.56E+09 0.98 

 122.45 2.898E+09 2.94E+09 1.02 2.94E+09 1.02 

 125.55 3.045E+09 3.09E+09 1.02 3.10E+09 1.02 

 128.65 3.120E+09 3.15E+09 1.01 3.16E+09 1.01 

 131.75 3.118E+09 3.14E+09 1.01 3.15E+09 1.01 

 134.85 3.047E+09 3.08E+09 1.01 3.08E+09 1.01 

 137.95 2.894E+09 2.95E+09 1.02 2.95E+09 1.02 

 141.05 2.706E+09 2.74E+09 1.01 2.74E+09 1.01 

 145.15 2.333E+09 2.38E+09 1.02 2.38E+09 1.02 

 149.25 1.888E+09 1.92E+09 1.02 1.92E+09 1.02 

 153.35 1.403E+09 1.57E+09 1.12 1.58E+09 1.12 

 

PLSA -32.13 -0.63 131.05 2.280E+09 2.23E+09 0.98 2.23E+09 0.98 

 134.15 2.426E+09 2.40E+09 0.99 2.40E+09 0.99 

 137.25 2.443E+09 2.41E+09 0.99 2.41E+09 0.99 

 141.35 2.309E+09 2.28E+09 0.99 2.28E+09 0.99 

 145.45 2.030E+09 2.01E+09 0.99 2.01E+09 0.99 

 149.55 1.656E+09 1.65E+09 1.00 1.66E+09 1.00 

 153.65 1.250E+09 1.04E+09 0.83 1.04E+09 0.83 

 

PLSA -32.13 0.63 106.5 5.115E+08 5.17E+08 1.01 5.20E+08 1.02 

 110.5 7.161E+08 7.55E+08 1.05 7.59E+08 1.06 

 114.5 9.113E+08 9.48E+08 1.04 9.53E+08 1.05 

 118.5 1.094E+09 1.13E+09 1.04 1.14E+09 1.04 

 122.5 1.264E+09 1.31E+09 1.04 1.32E+09 1.04 

 125.5 1.400E+09 1.45E+09 1.03 1.45E+09 1.04 

 128.5 1.624E+09 1.65E+09 1.01 1.65E+09 1.02 

 

PLSA -34.65 0.63 106.5 2.713E+08 2.72E+08 1.00 2.74E+08 1.01 

 110.5 3.843E+08 4.02E+08 1.05 4.06E+08 1.06 

 114.5 4.896E+08 5.03E+08 1.03 5.07E+08 1.04 

 118.5 5.899E+08 6.02E+08 1.02 6.07E+08 1.03 

 122.5 6.947E+08 7.16E+08 1.03 7.21E+08 1.04 

 125.5 7.980E+08 8.20E+08 1.03 8.26E+08 1.03 

 128.5 1.025E+09 1.04E+09 1.02 1.05E+09 1.02 

 131.05 1.617E+09 1.62E+09 1.00 1.63E+09 1.01 

 134.15 1.846E+09 1.85E+09 1.00 1.85E+09 1.00 

 137.25 1.897E+09 1.88E+09 0.99 1.88E+09 0.99 

 141.35 1.846E+09 1.80E+09 0.97 1.80E+09 0.97 

 145.45 1.628E+09 1.59E+09 0.98 1.60E+09 0.98 

 149.55 1.354E+09 1.31E+09 0.97 1.31E+09 0.97 

 153.65 1.033E+09 8.31E+08 0.80 8.32E+08 0.81 

 
96 

Table 4.13. Calculated-to-Measured (C/M) Ratios for the VENUS-3 Benchmark Experiment 58Ni(n,p)58Co Equivalent Fission Fluxes using TORT with BUGLE-96 and BUGLE-B7 Cross-Sections 

 Measured TORT with BUGLE-96 TORT with BUGLE-B7 

Measurement Position (cm) Equivalent Equivalent Equivalent 

Zone X Y Z* Fission Flux** Fission Flux C/M Fission Flux C/M 

 

PLSA -37.17 -0.63 131.05 1.084E+09 1.05E+09 0.97 1.05E+09 0.97 

 134.15 1.234E+09 1.21E+09 0.98 1.21E+09 0.98 

 137.25 1.280E+09 1.25E+09 0.98 1.25E+09 0.98 

 141.35 1.227E+09 1.19E+09 0.97 1.19E+09 0.97 

 145.45 1.108E+09 1.07E+09 0.96 1.07E+09 0.96 

 149.55 9.125E+08 8.83E+08 0.97 8.85E+08 0.97 

 153.65 7.026E+08 5.66E+08 0.81 5.67E+08 0.81 

 

PLSA -37.17 0.63 106.5 1.648E+08 1.68E+08 1.02 1.70E+08 1.03 

 110.5 2.296E+08 2.48E+08 1.08 2.51E+08 1.09 

 114.5 2.947E+08 3.14E+08 1.06 3.17E+08 1.08 

 118.5 3.548E+08 3.83E+08 1.08 3.87E+08 1.09 

 122.5 4.275E+08 4.56E+08 1.07 4.60E+08 1.08 

 125.5 5.062E+08 5.37E+08 1.06 5.41E+08 1.07 

 128.5 6.752E+08 6.92E+08 1.03 6.96E+08 1.03 

 

PLSA -37.17 -14.49 106.5 1.141E+08 1.17E+08 1.03 1.19E+08 1.04 

 110.5 1.480E+08 1.72E+08 1.16 1.74E+08 1.17 

 114.5 1.793E+08 2.17E+08 1.21 2.20E+08 1.23 

 118.5 2.212E+08 2.64E+08 1.19 2.67E+08 1.21 

 122.5 2.646E+08 3.12E+08 1.18 3.15E+08 1.19 

 125.5 3.204E+08 3.63E+08 1.13 3.66E+08 1.14 

 128.5 4.479E+08 4.62E+08 1.03 4.65E+08 1.04 

 

Inner Baffle -4.41 -4.41 114.5 1.474E+09 1.51E+09 1.02 1.52E+09 1.03 

 131.5 2.091E+09 2.12E+09 1.01 2.14E+09 1.02 

 145.5 1.523E+09 1.57E+09 1.03 1.59E+09 1.04 

Inner Baffle -4.41 -0.63 114.5 1.233E+09 1.25E+09 1.01 1.26E+09 1.02 

 131.5 1.752E+09 1.76E+09 1.00 1.78E+09 1.01 

 145.5 1.248E+09 1.31E+09 1.05 1.32E+09 1.06 

Outer Baffle -39.69 -18.27 114.5 9.002E+07 1.00E+08 1.11 1.01E+08 1.13 

 131.5 2.211E+08 2.29E+08 1.04 2.31E+08 1.04 

 145.5 2.167E+08 2.21E+08 1.02 2.22E+08 1.02 

Outer Baffle -39.69 -11.97 114.5 1.362E+08 1.50E+08 1.10 1.52E+08 1.11 

 131.5 3.803E+08 3.88E+08 1.02 3.90E+08 1.02 

 145.5 3.778E+08 3.84E+08 1.02 3.85E+08 1.02 

Outer Baffle -39.69 -5.67 114.5 1.636E+08 1.82E+08 1.11 1.83E+08 1.12 

 131.5 4.639E+08 4.79E+08 1.03 4.81E+08 1.04 

 145.5 4.638E+08 4.77E+08 1.03 4.78E+08 1.03 
97 

Table 4.13. Calculated-to-Measured (C/M) Ratios for the VENUS-3 Benchmark Experiment 58Ni(n,p)58Co Equivalent Fission Fluxes using TORT with BUGLE-96 and BUGLE-B7 Cross-Sections 

 Measured TORT with BUGLE-96 TORT with BUGLE-B7 

Measurement Position (cm) Equivalent Equivalent Equivalent 

Zone X Y Z* Fission Flux** Fission Flux C/M Fission Flux C/M 

 

Outer Baffle -39.69 -0.63 106.5 9.802E+07 1.03E+08 1.05 1.04E+08 1.07 

 110.5 1.360E+08 1.52E+08 1.12 1.53E+08 1.13 

 114.5 1.750E+08 1.90E+08 1.09 1.92E+08 1.10 

 118.5 2.106E+08 2.32E+08 1.10 2.34E+08 1.11 

 122.5 2.576E+08 2.79E+08 1.08 2.81E+08 1.09 

 125.5 3.085E+08 3.29E+08 1.07 3.31E+08 1.07 

 128.5 3.880E+08 4.12E+08 1.06 4.14E+08 1.07 

 131.5 5.003E+08 5.06E+08 1.01 5.08E+08 1.01 

 134.5 5.571E+08 5.68E+08 1.02 5.69E+08 1.02 

 137.5 5.731E+08 5.88E+08 1.03 5.89E+08 1.03 

 141.5 5.544E+08 5.65E+08 1.02 5.66E+08 1.02 

 145.5 4.945E+08 5.05E+08 1.02 5.06E+08 1.02 

 149.5 4.052E+08 4.20E+08 1.04 4.21E+08 1.04 

 153.5 2.912E+08 2.89E+08 0.99 2.89E+08 0.99 

 

Outer Baffle -37.17 -20.79 114.5 1.059E+08 1.15E+08 1.09 1.16E+08 1.10 

 131.5 2.395E+08 2.44E+08 1.02 2.45E+08 1.02 

 145.5 2.220E+08 2.29E+08 1.03 2.30E+08 1.04 

Outer Baffle -30.87 -20.79 114.5 3.080E+08 3.11E+08 1.01 3.13E+08 1.02 

 131.5 5.194E+08 5.33E+08 1.03 5.36E+08 1.03 

 145.5 4.341E+08 4.52E+08 1.04 4.55E+08 1.05 

Outer Baffle -24.57 -20.79 114.5 6.242E+08 6.39E+08 1.02 6.44E+08 1.03 

 131.5 9.238E+08 9.44E+08 1.02 9.51E+08 1.03 

 145.5 7.106E+08 7.34E+08 1.03 7.39E+08 1.04 

Outer Baffle -44.73 -0.63 106.5 3.736E+07 3.77E+07 1.01 3.86E+07 1.03 

 110.5 5.027E+07 5.23E+07 1.04 5.34E+07 1.06 

 114.5 6.366E+07 6.62E+07 1.04 6.75E+07 1.06 

 118.5 7.684E+07 8.14E+07 1.06 8.30E+07 1.08 

 122.5 9.566E+07 1.01E+08 1.05 1.02E+08 1.07 

 125.5 1.124E+08 1.14E+08 1.01 1.16E+08 1.03 

 128.5 1.330E+08 1.36E+08 1.02 1.38E+08 1.04 

 131.5 1.509E+08 1.49E+08 0.99 1.51E+08 1.00 

 134.5 1.646E+08 1.67E+08 1.01 1.69E+08 1.03 

 137.5 1.680E+08 1.70E+08 1.01 1.72E+08 1.02 

 141.5 1.636E+08 1.66E+08 1.01 1.68E+08 1.03 

 145.5 1.448E+08 1.49E+08 1.03 1.51E+08 1.04 

 149.5 1.207E+08 1.26E+08 1.04 1.27E+08 1.05 

 153.5 8.650E+07 8.71E+07 1.01 8.83E+07 1.02 

 

Core Barrel -49.77 -9.45 114.5 2.486E+07 2.73E+07 1.10 2.79E+07 1.12 

 131.5 5.210E+07 5.41E+07 1.04 5.49E+07 1.05 

 145.5 5.039E+07 5.24E+07 1.04 5.31E+07 1.05 
98 

Table 4.13. Calculated-to-Measured (C/M) Ratios for the VENUS-3 Benchmark Experiment 58Ni(n,p)58Co Equivalent Fission Fluxes using TORT with BUGLE-96 and BUGLE-B7 Cross-Sections 

 Measured TORT with BUGLE-96 TORT with BUGLE-B7 

Measurement Position (cm) Equivalent Equivalent Equivalent 

Zone X Y Z* Fission Flux** Fission Flux C/M Fission Flux C/M 

 

Core Barrel -49.77 -0.63 106.5 1.694E+07 1.80E+07 1.06 1.84E+07 1.08 

 110.5 2.221E+07 2.47E+07 1.11 2.52E+07 1.13 

 114.5 2.824E+07 3.18E+07 1.13 3.24E+07 1.15 

 118.5 3.597E+07 3.88E+07 1.08 3.95E+07 1.10 

 122.5 4.388E+07 4.58E+07 1.04 4.66E+07 1.06 

 125.5 4.936E+07 5.25E+07 1.06 5.34E+07 1.08 

 128.5 5.539E+07 5.99E+07 1.08 6.09E+07 1.10 

 131.5 6.090E+07 6.35E+07 1.04 6.45E+07 1.06 

 134.5 6.453E+07 6.91E+07 1.07 7.01E+07 1.09 

 137.5 6.676E+07 7.00E+07 1.05 7.10E+07 1.06 

 141.5 6.438E+07 6.80E+07 1.06 6.90E+07 1.07 

 145.5 5.873E+07 6.18E+07 1.05 6.27E+07 1.07 

 149.5 4.932E+07 5.27E+07 1.07 5.35E+07 1.08 

 153.5 3.757E+07 3.87E+07 1.03 3.93E+07 1.05 

 

Core Barrel -47.25 -18.27 106.5 1.415E+07 1.54E+07 1.09 1.57E+07 1.11 

 110.5 1.923E+07 2.10E+07 1.09 2.14E+07 1.11 

 114.5 2.383E+07 2.69E+07 1.13 2.73E+07 1.15 

 118.5 2.960E+07 3.25E+07 1.10 3.30E+07 1.12 

 122.5 3.559E+07 3.87E+07 1.09 3.93E+07 1.10 

 125.5 4.042E+07 4.35E+07 1.08 4.41E+07 1.09 

 128.5 4.540E+07 4.92E+07 1.08 4.99E+07 1.10 

 131.5 5.030E+07 5.27E+07 1.05 5.34E+07 1.06 

 134.5 5.411E+07 5.69E+07 1.05 5.76E+07 1.06 

 137.5 5.520E+07 5.74E+07 1.04 5.81E+07 1.05 

 141.5 5.348E+07 5.56E+07 1.04 5.62E+07 1.05 

 145.5 4.780E+07 5.02E+07 1.05 5.08E+07 1.06 

 149.5 4.042E+07 4.23E+07 1.05 4.28E+07 1.06 

 153.5 3.071E+07 3.06E+07 1.00 3.11E+07 1.01 

 

Core Barrel -45.99 -22.05 114.5 2.187E+07 2.45E+07 1.12 2.50E+07 1.14 

 131.5 4.348E+07 4.57E+07 1.05 4.63E+07 1.07 

 145.5 4.092E+07 4.27E+07 1.04 4.31E+07 1.05 

Core Barrel -44.73 -24.57 114.5 2.108E+07 2.32E+07 1.10 2.36E+07 1.12 

 131.5 3.942E+07 4.14E+07 1.05 4.20E+07 1.07 

 145.5 3.673E+07 3.81E+07 1.04 3.86E+07 1.05 

Core Barrel -42.21 -28.35 114.5 2.880E+07 2.58E+07 0.89 2.63E+07 0.91 

 131.5 4.480E+07 4.14E+07 0.92 4.21E+07 0.94 

 145.5 3.927E+07 3.60E+07 0.92 3.66E+07 0.93 

Core Barrel -38.43 -33.39 114.5 2.534E+07 2.59E+07 1.02 2.65E+07 1.05 

 131.5 3.764E+07 3.83E+07 1.02 3.92E+07 1.04 

 145.5 3.059E+07 3.24E+07 1.06 3.31E+07 1.08 
99 

Table 4.13. Calculated-to-Measured (C/M) Ratios for the VENUS-3 Benchmark Experiment 58Ni(n,p)58Co Equivalent Fission Fluxes using TORT with BUGLE-96 and BUGLE-B7 Cross-Sections 

 Measured TORT with BUGLE-96 TORT with BUGLE-B7 

Measurement Position (cm) Equivalent Equivalent Equivalent 

Zone X Y Z* Fission Flux** Fission Flux C/M Fission Flux C/M 

 

Core Barrel -35.91 -35.91 106.5 1.749E+07 1.77E+07 1.01 1.82E+07 1.04 

 110.5 2.249E+07 2.33E+07 1.04 2.38E+07 1.06 

 114.5 2.743E+07 2.91E+07 1.06 2.98E+07 1.09 

 118.5 3.162E+07 3.30E+07 1.04 3.38E+07 1.07 

 122.5 3.529E+07 3.66E+07 1.04 3.75E+07 1.06 

 125.5 3.743E+07 3.88E+07 1.04 3.97E+07 1.06 

 128.5 3.848E+07 4.02E+07 1.04 4.11E+07 1.07 

 131.5 3.864E+07 4.07E+07 1.05 4.16E+07 1.08 

 134.5 3.843E+07 4.05E+07 1.05 4.13E+07 1.08 

 137.5 3.811E+07 3.94E+07 1.03 4.03E+07 1.06 

 141.5 3.508E+07 3.65E+07 1.04 3.73E+07 1.06 

 145.5 3.090E+07 3.30E+07 1.07 3.37E+07 1.09 

 149.5 2.545E+07 2.66E+07 1.05 2.72E+07 1.07 

 153.5 1.974E+07 2.05E+07 1.04 2.10E+07 1.07 

 

Core Barrel -18.27 -47.25 106.5 3.421E+07 3.45E+07 1.01 3.49E+07 1.02 

 110.5 4.652E+07 4.87E+07 1.05 4.92E+07 1.06 

 114.5 5.720E+07 5.97E+07 1.04 6.03E+07 1.05 

 118.5 6.502E+07 6.87E+07 1.06 6.95E+07 1.07 

 122.5 7.095E+07 7.58E+07 1.07 7.67E+07 1.08 

 125.5 7.478E+07 7.92E+07 1.06 8.01E+07 1.07 

 128.5 7.515E+07 8.11E+07 1.08 8.20E+07 1.09 

 131.5 7.681E+07 8.14E+07 1.06 8.22E+07 1.07 

 134.5 7.491E+07 7.98E+07 1.07 8.07E+07 1.08 

 137.5 7.236E+07 7.68E+07 1.06 7.76E+07 1.07 

 141.5 6.641E+07 7.02E+07 1.06 7.09E+07 1.07 

 145.5 5.715E+07 6.11E+07 1.07 6.18E+07 1.08 

 149.5 4.667E+07 5.04E+07 1.08 5.10E+07 1.09 

 153.5 3.445E+07 3.59E+07 1.04 3.63E+07 1.06 

 

Core Barrel -0.63 -49.77 106.5 4.321E+07 4.46E+07 1.03 4.53E+07 1.05 

 110.5 5.814E+07 6.26E+07 1.08 6.35E+07 1.09 

 114.5 7.080E+07 7.58E+07 1.07 7.69E+07 1.09 

 118.5 8.203E+07 8.73E+07 1.06 8.85E+07 1.08 

 122.5 8.938E+07 9.63E+07 1.08 9.77E+07 1.09 

 125.5 9.403E+07 1.01E+08 1.07 1.02E+08 1.09 

 128.5 9.540E+07 1.03E+08 1.08 1.04E+08 1.09 

 131.5 9.618E+07 1.03E+08 1.08 1.05E+08 1.09 

 134.5 9.429E+07 1.01E+08 1.07 1.03E+08 1.09 

 137.5 9.057E+07 9.72E+07 1.07 9.86E+07 1.09 

 141.5 8.207E+07 8.88E+07 1.08 9.00E+07 1.10 

 145.5 7.174E+07 7.72E+07 1.08 7.83E+07 1.09 

 149.5 5.807E+07 6.41E+07 1.10 6.50E+07 1.12 

 153.5 4.317E+07 4.62E+07 1.07 4.69E+07 1.09 

* - The axial extent of the fuel in the TORT model is from 105 cm to 155 cm. 

** - The measurement uncertainty for all the 58Ni(n,p)58Co was 5.7% at the one-sigma level. 

 
100 

Table 4.14. Calculated-to-Measured (C/M) Ratios for the VENUS-3 Benchmark Experiment 115In(n,n’)115mIn Equivalent Fission Fluxes using TORT with BUGLE-96 and BUGLE-B7 Cross-Sections

 Measured TORT with BUGLE-96 TORT with BUGLE-B7 

Measurement Position (cm) Equivalent Equivalent Equivalent 

Zone X Y Z* Fission Flux** Fission Flux C/M Fission Flux C/M 

 

Inner Baffle -4.41 -4.41 131.5 2.713E+09 2.65E+09 0.98 2.64E+09 0.97 

Inner Baffle -4.41 -0.63 131.5 2.249E+09 2.22E+09 0.99 2.21E+09 0.98 

Outer Baffle -39.69 -0.63 106.5 1.189E+08 1.22E+08 1.03 1.22E+08 1.03 

 110.5 1.698E+08 1.81E+08 1.07 1.82E+08 1.07 

 114.5 2.181E+08 2.30E+08 1.06 2.31E+08 1.06 

 118.5 2.657E+08 2.82E+08 1.06 2.82E+08 1.06 

 122.5 3.256E+08 3.40E+08 1.04 3.40E+08 1.04 

 125.5 3.871E+08 4.02E+08 1.04 4.01E+08 1.04 

 128.5 4.859E+08 4.99E+08 1.03 4.97E+08 1.02 

 131.5 6.112E+08 6.10E+08 1.00 6.06E+08 0.99 

 134.5 6.915E+08 6.84E+08 0.99 6.79E+08 0.98 

 137.5 7.196E+08 7.08E+08 0.98 7.03E+08 0.98 

 141.5 6.894E+08 6.82E+08 0.99 6.76E+08 0.98 

 145.5 6.182E+08 6.09E+08 0.99 6.04E+08 0.98 

 149.5 5.062E+08 5.04E+08 0.99 5.00E+08 0.99 

 153.5 3.614E+08 3.44E+08 0.95 3.41E+08 0.94 

 

Water Gap -44.73 -0.63 106.5 3.767E+07 3.83E+07 1.02 3.86E+07 1.02 

 110.5 5.187E+0 5.39E+07 1.04 5.42E+07 1.05 

 114.5 6.610E+07 6.89E+07 1.04 6.93E+07 1.05 

 118.5 8.268E+07 8.50E+07 1.03 8.54E+07 1.03 

 122.5 9.990E+07 1.05E+08 1.05 1.05E+08 1.06 

 125.5 1.187E+08 1.20E+08 1.01 1.20E+08 1.01 

 128.5 1.405E+08 1.43E+08 1.01 1.43E+08 1.02 

 131.5 1.553E+08 1.57E+08 1.01 1.57E+08 1.01 

 134.5 1.725E+08 1.76E+08 1.02 1.75E+08 1.02 

 137.5 1.835E+08 1.79E+08 0.98 1.79E+08 0.97 

 141.5 1.786E+08 1.75E+08 0.98 1.75E+08 0.98 

 145.5 1.596E+08 1.57E+08 0.99 1.57E+08 0.98 

 149.5 1.344E+08 1.32E+08 0.98 1.31E+08 0.98 

 153.5 1.003E+08 9.09E+07 0.91 9.08E+07 0.91 

 

Core Barrel -49.77 -9.45 114.5 3.114E+07 3.21E+07 1.03 3.24E+07 1.04 

 131.5 6.647E+07 6.54E+07 0.98 6.57E+07 0.99 

 145.5 6.490E+07 6.30E+07 0.97 6.31E+07 0.97 
101 

Table 4.14. Calculated-to-Measured (C/M) Ratios for the VENUS-3 Benchmark Experiment 115In(n,n’)115mIn Equivalent Fission Fluxes using TORT with BUGLE-96 and BUGLE-B7 Cross-Sections

 Measured TORT with BUGLE-96 TORT with BUGLE-B7 

Measurement Position (cm) Equivalent Equivalent Equivalent 

Zone X Y Z* Fission Flux** Fission Flux C/M Fission Flux C/M 

 

Core Barrel -49.77 -0.63 106.5 1.898E+07 1.97E+07 1.04 1.99E+07 1.05 

 110.5 2.632E+07 2.78E+07 1.06 2.81E+07 1.07 

 114.5 3.364E+07 3.60E+07 1.07 3.64E+07 1.08 

 118.5 4.168E+07 4.43E+07 1.06 4.47E+07 1.07 

 122.5 5.081E+07 5.29E+07 1.04 5.34E+07 1.05 

 125.5 5.862E+07 6.08E+07 1.04 6.12E+07 1.04 

 128.5 6.611E+07 6.91E+07 1.05 6.95E+07 1.05 

 131.5 7.527E+07 7.42E+07 0.99 7.46E+07 0.99 

 134.5 7.857E+07 8.03E+07 1.02 8.06E+07 1.03 

 137.5 8.181E+07 8.17E+07 1.00 8.20E+07 1.00 

 141.5 7.946E+07 7.94E+07 1.00 7.97E+07 1.00 

 145.5 7.187E+07 7.19E+07 1.00 7.21E+07 1.00 

 149.5 6.074E+07 6.06E+07 1.00 6.08E+07 1.00 

 153.5 4.538E+07 4.34E+07 0.96 4.36E+07 0.96 

 

Core Barrel -47.25 -18.27 114.5 3.145E+07 3.32E+07 1.06 3.34E+07 1.06 

 131.5 6.527E+07 6.64E+07 1.02 6.65E+07 1.02 

 145.5 6.317E+07 6.29E+07 1.00 6.29E+07 1.00 

Core Barrel -45.99 -22.05 114.5 2.916E+07 3.10E+07 1.06 3.11E+07 1.07 

 131.5 5.782E+07 5.90E+07 1.02 5.91E+07 1.02 

 145.5 5.406E+07 5.47E+07 1.01 5.47E+07 1.01 

Core Barrel -44.73 -24.57 106.5 1.616E+07 1.65E+07 1.02 1.66E+07 1.03 

 110.5 2.251E+07 2.30E+07 1.02 2.32E+07 1.03 

 114.5 2.843E+07 2.93E+07 1.03 2.95E+07 1.04 

 118.5 3.422E+07 3.51E+07 1.03 3.53E+07 1.03 

 122.5 4.037E+07 4.12E+07 1.02 4.14E+07 1.03 

 125.5 4.450E+07 4.58E+07 1.03 4.60E+07 1.03 

 128.5 4.936E+07 5.06E+07 1.03 5.08E+07 1.03 

 131.5 5.326E+07 5.37E+07 1.01 5.38E+07 1.01 

 134.5 5.578E+07 5.68E+07 1.02 5.68E+07 1.02 

 137.5 5.706E+07 5.71E+07 1.00 5.71E+07 1.00 

 141.5 5.482E+07 5.47E+07 1.00 5.47E+07 1.00 

 145.5 4.892E+07 4.90E+07 1.00 4.91E+07 1.00 

 149.5 4.325E+07 4.04E+07 0.94 4.05E+07 0.94 

 153.5 3.041E+07 2.86E+07 0.94 2.87E+07 0.94 

 
102 

Table 4.14. Calculated-to-Measured (C/M) Ratios for the VENUS-3 Benchmark Experiment 115In(n,n’)115mIn Equivalent Fission Fluxes using TORT with BUGLE-96 and BUGLE-B7 Cross-Sections

 Measured TORT with BUGLE-96 TORT with BUGLE-B7 

Measurement Position (cm) Equivalent Equivalent Equivalent 

Zone X Y Z* Fission Flux** Fission Flux C/M Fission Flux C/M 

 

Core Barrel -42.21 -28.35 106.5 1.696E+07 1.71E+07 1.01 1.74E+07 1.02 

 110.5 2.323E+07 2.38E+07 1.02 2.41E+07 1.04 

 114.5 2.884E+07 3.02E+07 1.05 3.06E+07 1.06 

 118.5 3.446E+07 3.57E+07 1.04 3.62E+07 1.05 

 122.5 3.955E+07 4.10E+07 1.04 4.15E+07 1.05 

 125.5 4.336E+07 4.47E+07 1.03 4.53E+07 1.04 

 128.5 4.667E+07 4.81E+07 1.03 4.87E+07 1.04 

 131.5 4.954E+07 5.02E+07 1.01 5.07E+07 1.02 

 134.5 5.079E+07 5.18E+07 1.02 5.23E+07 1.03 

 137.5 5.121E+07 5.15E+07 1.01 5.20E+07 1.02 

 141.5 4.861E+07 4.88E+07 1.00 4.92E+07 1.01 

 145.5 4.393E+07 4.37E+07 0.99 4.41E+07 1.00 

 149.5 3.846E+07 3.59E+07 0.93 3.63E+07 0.94 

 153.5 2.700E+07 2.59E+07 0.96 2.62E+07 0.97 

 

Core Barrel -38.43 -33.39 114.5 2.928E+07 2.92E+07 1.00 2.97E+07 1.01 

 131.5 4.457E+07 4.40E+07 0.99 4.46E+07 1.00 

 145.5 3.611E+07 3.66E+07 1.01 3.72E+07 1.03 

Core Barrel -35.91 -35.91 114.5 3.187E+07 3.23E+07 1.01 3.28E+07 1.03 

 131.5 4.601E+07 4.61E+07 1.00 4.68E+07 1.02 

 145.5 3.651E+07 3.69E+07 1.01 3.75E+07 1.03 

Core Barrel -18.27 -47.25 114.5 7.292E+07 7.40E+07 1.01 7.43E+07 1.02 

 131.5 1.001E+08 1.01E+08 1.01 1.02E+08 1.02 

 145.5 7.480E+07 7.57E+07 1.01 7.60E+07 1.02 

Core Barrel -0.63 -49.77 106.5 5.576E+07 5.01E+07 0.90 5.05E+07 0.91 

 110.5 7.180E+07 7.17E+07 1.00 7.22E+07 1.01 

 114.5 8.627E+07 8.79E+07 1.02 8.85E+07 1.03 

 118.5 9.990E+07 1.02E+08 1.02 1.02E+08 1.02 

 122.5 1.101E+08 1.12E+08 1.02 1.13E+08 1.03 

 125.5 1.140E+08 1.17E+08 1.03 1.18E+08 1.04 

 128.5 1.168E+08 1.20E+08 1.03 1.21E+08 1.04 

 131.5 1.159E+08 1.21E+08 1.04 1.21E+08 1.05 

 134.5 1.154E+08 1.18E+08 1.02 1.19E+08 1.03 

 137.5 1.111E+08 1.13E+08 1.02 1.14E+08 1.03 

 141.5 1.006E+08 1.03E+08 1.03 1.04E+08 1.03 

 145.5 8.685E+07 8.96E+07 1.03 9.02E+07 1.04 

 149.5 7.624E+07 7.35E+07 0.96 7.40E+07 0.97 

 153.5 5.050E+07 5.18E+07 1.03 5.22E+07 1.03 

* - The axial extent of the fuel in the TORT model is from 105 cm to 155 cm. 

** - The measurement uncertainty for all the 115In(n,n’)115mIn was 5.5% at the one-sigma level. 

 
103 

Table 4.15. Calculated-to-Measured (C/M) Ratios for the VENUS-3 Benchmark Experiment 27Al(n,)

24Na Equivalent Fission Fluxes using TORT with BUGLE-96 and BUGLE-B7 Cross-Sections 

 Measured TORT with BUGLE-96 TORT with BUGLE-B7 

Measurement Position (cm) Equivalent Equivalent Equivalent 

Zone X Y Z* Fission Flux** Fission Flux C/M Fission Flux C/M 

 

Inner Baffle -4.41 -4.41 131.5 1.561E+09 1.89E+09 1.21 1.94E+09 1.24 

Inner Baffle -4.41 -0.63 131.5 1.927E+09 1.58E+09 0.82 1.62E+09 0.84 

Outer Baffle -39.69 -0.63 114.5 1.852E+08 1.94E+08 1.05 1.99E+08 1.08 

 131.5 4.788E+08 4.76E+08 0.99 4.87E+08 1.02 

 145.5 4.765E+08 4.68E+08 0.98 4.78E+08 1.00 

Water Gap -44.73 -0.63 114.5 8.376E+07 8.28E+07 0.99 8.54E+07 1.02 

 131.5 1.871E+08 1.71E+08 0.91 1.76E+08 0.94 

 145.5 1.782E+08 1.70E+08 0.95 1.74E+08 0.98 

Core Barrel -49.77 -9.45 114.5 3.197E+07 3.49E+07 1.09 3.60E+07 1.13 

 131.5 6.316E+07 6.17E+07 0.98 6.36E+07 1.01 

 145.5 6.026E+07 5.95E+07 0.99 6.13E+07 1.02 

Core Barrel -49.77 -0.63 114.5 4.110E+07 4.10E+07 1.00 4.24E+07 1.03 

 131.5 7.669E+07 7.33E+07 0.96 7.55E+07 0.98 

 145.5 7.270E+07 7.11E+07 0.98 7.32E+07 1.01 

Core Barrel -47.25 -18.27 114.5 3.159E+07 3.25E+07 1.03 3.35E+07 1.06 

 131.5 6.022E+07 5.76E+07 0.96 5.94E+07 0.99 

 145.5 5.705E+07 5.45E+07 0.95 5.61E+07 0.98 

Core Barrel -45.99 -22.05 114.5 3.051E+07 2.98E+07 0.98 3.08E+07 1.01 

 131.5 5.328E+07 5.03E+07 0.94 5.18E+07 0.97 

 145.5 4.801E+07 4.66E+07 0.97 4.80E+07 1.00 

Core Barrel -44.73 -24.57 114.5 2.753E+07 2.77E+07 1.01 2.86E+07 1.04 

 131.5 4.977E+07 4.51E+07 0.91 4.65E+07 0.93 

 145.5 4.330E+07 4.14E+07 0.96 4.26E+07 0.98 

Core Barrel -42.21 -28.35 114.5 3.168E+07 3.36E+07 1.06 3.47E+07 1.10 

 131.5 5.149E+07 4.96E+07 0.96 5.13E+07 1.00 

 145.5 4.350E+07 4.31E+07 0.99 4.45E+07 1.02 

Core Barrel -38.43 -33.39 114.5 3.683E+07 3.45E+07 0.94 3.57E+07 0.97 

 131.5 5.481E+07 4.86E+07 0.89 5.02E+07 0.92 

 145.5 4.279E+07 4.24E+07 0.99 4.38E+07 1.02 

Core Barrel -35.91 -35.91 114.5 3.807E+07 3.96E+07 1.04 4.09E+07 1.08 

 131.5 5.532E+07 5.27E+07 0.95 5.45E+07 0.98 

 145.5 4.231E+07 4.39E+07 1.04 4.54E+07 1.07 
104 

Table 4.15. Calculated-to-Measured (C/M) Ratios for the VENUS-3 Benchmark Experiment 27Al(n,)

24Na Equivalent Fission Fluxes using TORT with BUGLE-96 and BUGLE-B7 Cross-Sections 

 Measured TORT with BUGLE-96 TORT with BUGLE-B7 

Measurement Position (cm) Equivalent Equivalent Equivalent 

Zone X Y Z* Fission Flux** Fission Flux C/M Fission Flux C/M 

 

Core Barrel -18.27 -47.25 114.5 6.806E+07 6.48E+07 0.95 6.67E+07 0.98 

 131.5 8.915E+07 8.70E+07 0.98 8.95E+07 1.00 

 145.5 6.547E+07 6.64E+07 1.01 6.83E+07 1.04 

Core Barrel -0.63 -49.77 114.5 9.554E+07 8.70E+07 0.91 8.96E+07 0.94 

 131.5 1.202E+08 1.17E+08 0.97 1.20E+08 1.00 

 145.5 8.848E+07 8.86E+07 1.00 9.12E+07 1.03 

* - The axial extent of the fuel in the TORT model is from 105 cm to 155 cm. 

** - The measurement uncertainty for all the 27Al(n,)

24Na was 7.2% at the one-sigma level. 

 
105 

4.6 Conclusions from Verification and Validation Tests 

The VITAMIN-B7 and BUGLE-B7 libraries have been verified and validated by an extensive set 

of tests and by comparison to benchmark experiments. Although these libraries were not 

developed for criticality applications, the results of the critical benchmark analyses help to verify 

the VITAMIN-B7 library by demonstrating agreement with experimental data that is consistent 

with results obtained using the VITAMIN-B6 library and with results using continuous energy 

data based on ENDF/B-VII. 

The primary application for which the VITAMIN-B7 and BUGLE-B7 libraries were developed is 

use in LWR shielding analyses, including pressure vessel dosimetry calculations. The shielding 

benchmarks in Sections 4.4 and 4.5 provide an extensive set of tests of both the VITAMIN-B7 

and BUGLE-B7 libraries. These tests include Monte Carlo calculations, two-dimensional and 

three-dimensional deterministic transport calculations, and the hybrid MAVRIC sequence in 

SCALE. The results of the shielding benchmarks demonstrate that the VITAMIN-B7 and 

BUGLE-B7 libraries are appropriate for use in LWR shielding applications. 

The calculated dosimetry responses for the pressure vessel dosimetry benchmarks in Section 

4.5 using the VITAMIN-B7 and BUGLE-B7 libraries are typically equal to or slightly higher (by 

up to about 4%) than the responses calculated using VITAMIN-B6 and BUGLE-96. In all the 

benchmark analyses, the calculated reaction rates are within 20% of the measured data, which 

is consistent with the calculational uncertainty criterion in Regulatory Guide 1.190. 

The differences in the calculated responses for the shielding benchmarks using the 

VITAMIN-B7 and BUGLE-B7 libraries compared to VITAMIN-B6 and BUGLE-96 are not as 

pronounced as the changes that occurred when the VITAMIN-B6 and BUGLE-96 libraries were 

developed. VITAMIN-B6 and BUGLE-96 included significant changes in some ENDF 

evaluations (particularly iron) which had a considerable effect on pressure vessel dosimetry 

applications. In comparison, there were no substantial changes to the cross-section data for 

many of the structural elements (e.g., Fe, Cr, and Ni) in ENDF/B-VII.0 compared to 

ENDF/B-VI.3. 

The results of the benchmark evaluations suggest that VITAMIN-B6, VITAMIN-B7, BUGLE-96, 

and BUGLE-B7 are all suitable for LWR shielding analyses. While the new libraries do not 

provide a demonstrable improvement for LWR shielding applications, they provide validated 

libraries based on the latest version of the Evaluated Nuclear Data File, which enables 

adherence to Regulatory Guide 1.190. 
106 

5 REFERENCES 

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Neutron/ Photon Transport Code, ORNL/TM-13221, October 1997. 



UNITED STATES 

NUCLEAR REGULATORY COMMISSION

WASHINGTON, DC 20555-0001

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NUREG/CR-7045 Production and Testing of the VITAMIN-B7 Fine-Group and BUGLE-B7

Broad-Group Coupled Neutron/Gamma Cross-Section Libraries Derived

from ENDF/B-VII.0 Nuclear Data