NUREG/CR-6990 

ONRL/TM-2008/047

Development and Testing of 

ENDF/B-VI.8 and 

ENDF/B-VII.0 Coupled 

Neutron-Gamma 

Libraries for SCALE 6

Office of Nuclear Material Safety and Safeguards 
 

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NUREG/CR-6990 

ORNL/TM-2008/047 

Development and Testing of 

ENDF/B-VI.8 and 

ENDF/B-VII.0 Coupled 

Neutron-Gamma 

Libraries for SCALE 6

Manuscript Completed: December 2008 

Date Published: February 2009 

Prepared by 

D. Wiarda, M.E. Dunn, D.E. Peplow, T.M. Miller and H. Akkurt 

Oak Ridge National Laboratory 

Managed by UT-Battelle, LLC 

Oak Ridge, TN 37831-6170 

A.B. Barto, NRC Project Manager 

NRC Job Code J5609 

Office of Nuclear Material Safety and Safeguards 
 
iii 

ABSTRACT 

New coupled neutron-gamma shielding libraries have been developed for use with the SCALE 

(Standardized Computer Analyses for Licensing Evaluation) code system. Two libraries have been 

generated based on the Evaluated Nuclear Data File/B Version VI (ENDF/B-VI) Release 8 

(ENDF/B-VI.8) and ENDF/B-VII Release 0 (ENDF/B-VII.0) cross-section evaluations. Both libraries 

were generated using 200 neutron and 47 gamma groups. The neutron group structure is identical to the 

199-group VITAMIN-B6 structure except an additional neutron group has been added to extend the top 

energy boundary to 20 MeV. In addition to the “fine-group” libraries, a “broad-group” ENDF/B-VII.0 

library has been generated using a 27 neutron and 19 gamma group structure. In addition to the coupled 

cross-section library development, new shielding codes Monaco and MAVRIC (Monaco with Automated 

Variance Reduction using Importance Calculations) have been developed for SCALE. This report 

documents the ENDF/B-VI.8 and ENDF/B-VII.0 cross-section library development and the benchmark 

testing of the new libraries with the new SCALE shielding transport codes Monaco and MAVRIC. 
 

iv 
v 

TABLE OF CONTENTS 

Page

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

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

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

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

ACRONYMS AND ABBREVIATIONS ................................................................................................... xv 

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

2 LIBRARY GENERATION ................................................................................................................ 3 

2.1 AMPX Input File Generation .................................................................................................... 3 

2.2 Processing .................................................................................................................................. 8 

2.2.1 Group structure ............................................................................................................. 8 

2.2.2 AMPX execution sequences for neutron and gamma-ray yield processing ................. 9 

2.2.2.1 Point-wise cross-section data ..................................................................... 13

2.2.2.2 Comparison with NJOY ............................................................................. 15 

2.2.2.3 Neutron and gamma yield processing ........................................................ 16 

2.2.2.4 Thermal neutron processing ....................................................................... 18 

2.2.2.5 Generating full-range Bondarenko factors and thermal data ..................... 20

2.2.2.6 Gamma processing ..................................................................................... 22 

2.2.2.7 Creating the master library ......................................................................... 26

2.2.3 Adding dose data ........................................................................................................ 28 

3 BENCHMARK TESTING ............................................................................................................... 37 

3.1 SAS4 and MAVRIC Benchmark Comparison Calculations with ENDF/B-VI.8 Library ...... 37 

3.1.1 Cask descriptions ........................................................................................................ 38 

3.1.1.1 CASTOR-V/21 .......................................................................................... 38 

3.1.1.2 Westinghouse MC-10 ................................................................................ 39 

3.1.1.3 TN-24P ....................................................................................................... 40 

3.1.2 Calculations ................................................................................................................ 40 

3.1.2.1 Computational results ................................................................................ 41 

3.1.2.2 Benchmark against experimental measurement ......................................... 50 

3.2 MAVRIC Benchmark Calculations Using ENDF/B-VI.8 and ENDF/B-VII.0 Libraries ....... 53

3.2.1 Neutron fields in three-section concrete labyrinth from 252Cf source ........................ 53 

3.2.1.1 Benchmark overview ................................................................................. 53 

3.2.1.2 Monaco/MAVRIC calculations ................................................................. 54 

3.2.1.3 Detector response functions ....................................................................... 55

3.2.1.4 Neutron transport through the labyrinth..................................................... 60 

3.2.2 Neutron and photon leakage spectra from 252Cf source at centers of six iron 

spheres of different diameters .................................................................................... 68 

3.2.2.1 Benchmark overview ................................................................................. 68 

3.2.2.2 Monaco calculations .................................................................................. 69 

3.2.3 Comparison of measured and calculated neutron transmission through heavy 

water for a 252Cf source placed in the center of a 30 cm diameter sphere .................. 76 

3.2.3.1 Experiment overview ................................................................................. 76 

3.2.3.2 Monaco calculations .................................................................................. 76 
 

vi 

TABLE OF CONTENTS (continued) 

Page

3.2.4 Comparison of measured and calculated neutron transmission through steel 

for a 252Cf source ........................................................................................................ 80 

3.2.4.1 Experiment overview ................................................................................. 80 

3.2.4.2 Monaco calculations .................................................................................. 80 

3.2.5 Systematic evaluation of neutron shielding effects for materials ............................... 82 

3.2.5.1 Experimental overview .............................................................................. 82 

3.2.5.2 MCNP simulations ..................................................................................... 83 

3.2.5.3 MAVRIC calculations ............................................................................... 83 

3.2.5.4 Results ........................................................................................................ 84 

4 SUMMARY ...................................................................................................................................... 87 

5 REFERENCES ................................................................................................................................. 89 

APPENDIX A: XML INPUT TEMPLATES .......................................................................................... A-1

APPENDIX B: EVALUATIONS INCLUDED IN ENDF/B-VI.8 LIBRARY ....................................... B-1 

APPENDIX C: EVALUATIONS INCLUDED IN ENDF/B-VII.0 LIBRARY ...................................... C-1 
vii 

LIST OF FIGURES 

Figure Page

2.1 Work flow for AMPX master library creation. ......................................................................... 3

2.2 AMPX sequence for producing neutron multigroup data. ...................................................... 10

2.3 AMPX sequence to generate the coupled gamma-neutron multigroup data. .......................... 11

2.4 Weighting spectrum generated by JERGENS. ........................................................................ 14

2.5 Point-wise capture cross section as processed with AMPX and NJOY for 140Ce. .................. 16

2.6 AMPX and NJOY scattering data for MT=504 in hydrogen. ................................................. 26

2.7 Interpolation for Claiborne-Trubey gamma dose function. ..................................................... 31

2.8 Interpolation for ICRU-57, Table A.21 (air) Kerma gamma dose function. ........................... 32

2.9 Interpolation for ambient dose equivalent ............................................................................... 33

2.10 Interpolation for effective dose ............................................................................................... 34

3.1 Views of cask for storage and transport of radioactive material (CASTOR). ......................... 38

3.2 Views of MC-10 cask. ............................................................................................................. 39

3.3 Views of TN-24P cask. ........................................................................................................... 40

3.4 The ratio of the computed TN-24P cask tallies using the ENDF/B-VI.8 and 

ENDF/B-IV shielding cross sections. ...................................................................................... 43

3.5 The ratios of the computed TN-24P cask tallies using MAVRIC and SAS4. ......................... 44

3.6 The ratio of the computed MC-10 cask values using the ENDF/B-VI.8 and 

ENDF/B-IV shielding cross-section libraries with MAVRIC. ................................................ 47

3.7 The ratio of the computed CASTOR cask values using the ENDF/B-VI.8 and 

ENDF/B-IV shielding cross-section libraries with MAVRIC. ................................................ 50

3.8 The ratio of the computed to measured values using MAVRIC with the 

ENDF/B-VI.8 (new) and ENDF/B-IV (old) libraries for the neutron dose rates. ................... 51

3.9 The ratio of the computed to measured values using MAVRIC with the 

ENDF/B-VI.8 (new) and ENDF/B-IV (old) shielding libraries for the gamma dose rates. .... 53

3.10 Cutaway view of labyrinth model ........................................................................................... 54

3.11 6

Li(n,t)4He cross section from the ENDF/B VI.8 200n/47g shielding library. ........................ 55

3.12 Response (counts per unit flux) for the 2-inch diameter Bonner sphere 

(with cadmium cover). ............................................................................................................ 57

3.13 Response (counts per unit flux) for the 3-inch diameter Bonner sphere 

(with cadmium cover). ............................................................................................................ 57

3.14 Response (counts per unit flux) for the 5-inch diameter Bonner sphere 

(with cadmium cover). ............................................................................................................ 58

3.15 Response (counts per unit flux) for the 5-inch diameter Bonner sphere 

(no cover). ............................................................................................................................... 58

3.16 Response (counts per unit flux) for the 8-inch diameter Bonner sphere 

(with cadmium cover). ............................................................................................................ 59

3.17 Response (counts per unit flux) for the 10-inch diameter Bonner sphere 

(with cadmium cover). ............................................................................................................ 59

3.18 Response (counts per unit flux) for the 12-inch diameter Bonner sphere 

(with cadmium cover). ............................................................................................................ 60

3.19 Neutron count rate C/E for the 2-inch Bonner sphere (with cadmium) and 

the uncovered source. .............................................................................................................. 61

3.20 Neutron count rate C/E for the 2-inch Bonner sphere (with cadmium) and 

the covered source. .................................................................................................................. 61 
 

viii 

LIST OF FIGURES (continued) 

Figure Page

3.21 Neutron count rate C/E for the 3-inch Bonner sphere (with cadmium) and 

the uncovered source. .............................................................................................................. 62

3.22 Neutron count rate C/E for the 3-inch Bonner sphere (with cadmium) and 

the covered source. .................................................................................................................. 62

3.23 Neutron count rate C/E for the 5-inch Bonner sphere (with cadmium) and 

the uncovered source. .............................................................................................................. 63

3.24 Neutron count rate C/E for the 5-inch Bonner sphere (with cadmium) and 

the covered source. .................................................................................................................. 63

3.25 Neutron count rate C/E for the 5-inch Bonner sphere (no cadmium) and the 

uncovered source. .................................................................................................................... 64

3.26 Neutron count rate C/E for the 5-inch Bonner sphere (no cadmium) and the 

covered source. ........................................................................................................................ 64

3.27 Neutron count rate C/E for the 8-inch Bonner sphere (with cadmium) and 

the uncovered source. .............................................................................................................. 65

3.28 Neutron count rate C/E for the 8-inch Bonner sphere (with cadmium) and 

the covered source. .................................................................................................................. 65

3.29 Neutron count rate C/E for the 10-inch Bonner sphere (with cadmium) and 

the uncovered source. .............................................................................................................. 66

3.30 Neutron count rate C/E for the 10-inch Bonner sphere (with cadmium) and 

the covered source. .................................................................................................................. 66

3.31 Neutron count rate C/E for the 12-inch Bonner sphere (with cadmium) and 

the uncovered source. .............................................................................................................. 67

3.32 Neutron count rate C/E for the 12-inch Bonner sphere (with cadmium) and 

the covered source. .................................................................................................................. 67

3.33 Iron sphere experiments. ......................................................................................................... 68

3.34 Comparison of the neutron calculation to the measurement of the bare source. ..................... 69

3.35 Comparison of the photon calculation to the measurement of the bare source. ...................... 70

3.36 Comparison of the neutron calculation to the measurement of the 20 cm shield. ................... 70

3.37 Comparison of the neutron calculation to the measurement of the 30 cm shield. ................... 71

3.38 Comparison of the neutron calculation to the measurement of the 40 cm shield. ................... 71

3.39 Comparison of the neutron calculation to the measurement of the 50 cm shield. ................... 72

3.40 Comparison of the neutron calculation to the measurement of the 60 cm shield. ................... 72

3.41 Comparison of the neutron calculation to the measurement of the 70 cm shield. ................... 73

3.42 Comparison of the photon calculation to the measurement of the 30 cm shield. .................... 73

3.43 Comparison of the photon calculation to the measurement of the 40 cm shield. .................... 74

3.44 Comparison of the photon calculation to the measurement of the 50 cm shield. .................... 74

3.45 Comparison of the photon calculation to the measurement of the 60 cm shield. .................... 75

3.46 Comparison of the photon calculation to the measurement of the 70 cm shield. .................... 75

3.47 Photon flux through the 70 cm shield as calculated by Monaco and MCNP5. ....................... 76

3.48 Monaco model of 252Cf source in heavy water experiment. .................................................... 77

3.49 Comparison of the Monaco (ENDF/B-VI.8) results and the measured data. .......................... 78

3.50 Ratios of the Monaco (ENDF/B-VI.8 fine-group library) results and the measured data. ...... 78

3.51 Comparison of the Monaco (ENDF/B-VII.0) results and the measured data. ......................... 79

3.52 Ratios of the Monaco (ENDF/B-VII.0 fine-group library) results and the measured data. .... 79

3.53 Monaco output (ENDF/B-VI.8 fine-group library) for total flux at 10 cm using 

two different tallies compared to the two sets of measured data. ............................................ 80
ix 

LIST OF FIGURES (continued) 

Figure Page

3.54 Ratios of the Monaco (ENDF/B-VI.8 fine-group library) point detector tally 

to the measured fluxes. ............................................................................................................ 81

3.55 Monaco output (ENDF/B-VII.0 fine-group library) for total flux at 10 cm using 

two different tallies compared to the two sets of measured data. ............................................ 81

3.56 Ratios of the Monaco (ENDF/B-VII.0 fine-group library) point detector tally to 

the measured fluxes. ................................................................................................................ 82

3.57 Neutron shielding experiment example ................................................................................... 83

3.58 Ratio of MAVRIC calculation of neutron detector dose rate to experimental 

measurement. ........................................................................................................................... 84

3.59 Type 2a results—25 cm of stainless steel first, then polyethylene. ......................................... 85

3.60 Type 2b results—polyethylene, then 25 cm of stainless steel. ................................................ 85

3.61 Type 3 results for polyethylene fixed at a thickness of 15 cm. ............................................... 85
 

x 
xi 

LIST OF TABLES 

Table Page

2.1 Substitution tags available for use in XML input templates ..................................................... 5

2.2 Template control elements that allow looping over the evaluations ......................................... 7

2.3 Energy boundaries for the 200 neutron group structure ............................................................ 8

2.4 Energy boundaries for the 47 gamma group structure .............................................................. 9

2.5 Energy boundaries for the 27 neutron group structure .............................................................. 9

2.6 Energy boundaries for the 19 gamma group structure .............................................................. 9

2.7 Evaluations excluded from the ENDF/B-VI.8 library ............................................................. 12

2.8 Available dose functions on the coupled neutron-gamma libraries ........................................ 28

3.1 TN-24P calculations input summary ....................................................................................... 41

3.2 The computational results for TN24P casks ............................................................................ 42

3.3 Westinghouse MC-10 calculations input summary ................................................................. 45

3.4 The computational results for MC-10 casks ............................................................................ 46

3.5 CASTOR-V/21 calculations input summary ........................................................................... 48

3.6 The computational results for CASTOR casks ....................................................................... 49

3.7 The measured and computed neutron dose rates and the ratio of 

computed to measured (C/M) dose rates for each cask model ................................................ 51

3.8 The measured and computed gamma dose rates and the ratio of 

the computed to measured (C/M) dose rates for each cask model .......................................... 52

B-1 Evaluations and IDs used for the ENDF/B-VII.8 library ...................................................... B-1 

C-1 Evaluations included in ENDF/B-VII.0 library ..................................................................... C-1 
 

xii 
xiii 

ACKNOWLEDGMENTS 

Sincere appreciation is expressed to Andrew B. Barto of the U.S. Nuclear Regulatory Commission Office 

of Nuclear Material Safety and Safeguards for supporting the development and testing of ENDF/B-VI.8 

and ENDF/B-VII.0 coupled neutron-gamma libraries for SCALE. In addition, the authors would like to 

thank Bryan Broadhead, Mark Williams, and Steve Bowman for their technical review of the report and 

guidance in the development and testing of the cross-section libraries. Also, special appreciation is 

expressed to Joyce B. Echols and Willena C. Carter for their efforts in preparing the final manuscript for 

publication. 
 

xiv 
xv 

ACRONYMS AND ABBREVIATIONS 

1-D one-dimensional 

2-D two-dimensional 

3-D three-dimensional 

CADIS Consistent Adjoint-Driven Importance Sampling 

CENTRM Continuous ENergy TRansport Module 

ENDF/B-IV Evaluated Nuclear Data File Version IV 

ENDF/B-VI.8 Evaluated Nuclear Data File Version VI Release 8 

ENDF/B-VII.0 Evaluated Nuclear Data File Version VII 

ID identifier 

MAVRIC Monaco with Automated Variance Reduction using Importance Calculations 

MORSE Multigroup Oak Ridge Stochastic Experiment 

NNDC National Nuclear Data Center 

ORNL Oak Ridge National Laboratory 

SAS4 SCALE Shielding Analysis Sequence Number 4 

SCALE Standardized Computer Analyses for Licensing Evaluation 

SGGP SCALE Generalized Geometry Package 

SLBW Single-Level Breit-Wigner 

XML Extensible Markup Language 
 

xvi 
1 

1 INTRODUCTION 

Before development of Standardized Computer Analyses for Licensing Evaluation (SCALE) Version 6, 

coupled neutron-gamma cross-section libraries distributed with the code system were limited to 

U.S. Evaluated Nuclear Data File/B Version IV (ENDF/B-IV) data.1

 The latest version of the ENDF/B 

evaluation database that can be obtained from the National Nuclear Data Center (NNDC) at Brookhaven 

National Laboratory (BNL) is ENDF/B-VII.0 (Ref. 2). With the release of ENDF/B-VII in December 

2006 (Ref. 3), the Oak Ridge National Laboratory (ORNL) initiated efforts to update the SCALE libraries 

for shielding applications to include both ENDF/B-VI Release 8 (ENDF/B-VI.8) and ENDF/B-VII.0 

cross-section data. This report documents the cross-section library generation effort and the subsequent 

benchmark testing of the data libraries with SCALE. 

In concert with the new cross-section library development activities, ORNL has developed new shielding 

capabilities for the SCALE code system. The Monaco shielding module is a new multigroup, fixedsource Monte Carlo transport code that can be used to perform shielding analyses for licensing 

applications. Monaco is the result of a modernization effort combining the multigroup neutron and 

photon physics of the well-known MORSE (Multigroup Oak Ridge Stochastic Experiment) Monte Carlo 

shielding code with the flexibility of the SCALE Generalized Geometry Package (SGGP) and 

cross-section processing. Monaco uses the same cross-section package as other SCALE modules. 

Available tallies in Monaco include point detectors, region-based flux tallies, and mesh tallies. Any 

Monaco tally can be folded with a response function, either user-entered or from a standard list available 

with each SCALE cross-section library. Mesh tally values and uncertainties can be viewed with a special 

Java viewer on all computer platforms that run SCALE. Monaco is implemented in SCALE through the 

MAVRIC (Monaco with Automated Variance Reduction using Importance Calculations) computational 

sequence. The MAVRIC sequence in SCALE combines the results of an adjoint calculation from the 

three-dimensional (3-D) deterministic code Denovo with Monaco. Both an importance map for weight 

windows and a biased source are automatically generated from the adjoint flux using the CADIS 

(Consistent Adjoint-Driven Importance Sampling) methodology. MAVRIC is completely automated—

from a simple user input, it creates the cross sections (forward and adjoint), calculates the first collision 

source, computes the adjoint fluxes, creates the importance map and biased source, and executes Monaco. 

Users can start and stop the calculation at various points so that progress can be monitored and 

importance maps can be reused for similar problems. For simple problems that do not require the 

advanced variance reduction, the MAVRIC sequence is an easy way to compute problem-dependent 

cross sections and execute Monaco with a common user input. The MAVRIC/Monaco capability 

represents a significant radiation transport advancement for SCALE. 

This report provides detailed computational results using the new shielding libraries and radiation 

transport capabilities of MAVRIC/Monaco. 
 

2 
 3 

2 LIBRARY GENERATION 

2.1 AMPX Input File Generation

A flow diagram is presented in Fig. 2.1 to illustrate the automated procedure to generate the AMPX 

master library from ENDF/B evaluations. 

Fig. 2.1 Work flow for AMPX master library creation. 

Input files are generated using an XML listing of the desired evaluations and XML templates. The XML 

listing of the desired evaluation is generated by a Java program that scans a directory tree of ENDF files 

and writes the relevant information to a file. A typical listing looks like the following. 

<Material tag="al27" 

 endf="1325" 

 za="13027" 

 tape="/home/dw8/ampx/running/endf6/rev8/First.endf" 

 mod="4" 

 rel="8" 

 eval="EVAL-FEB97" 

 neutron="yes" 

 author="M.B.CHADWICK & P.G.YOUNG " 

 lab="M.B.CHADWICK & P.G.YOUNG " 

 dist="DIST-NOV 1" 

 rdate="REV3- " 

 rev="6" 

 version="6" 

 awr="26.74975" 

 lis="0" 

 nis="1" 
4 

 scattering="1.34799E+00" 

 file2="yes" 

 file3="yes" 

 file4="yes" 

 file6="yes" 

 AWP0="yes" 

 file12="yes" 

 /> 

 <Material tag="graphite" 

 endf="31" 

 za="131" 

 tape="/home/dw8/ampx/running/endf6/rev8/Thermal.endf" 

 mod="1" 

 rel="3" 

 eval="EVAL-apr93" 

 neutron="yes" 

 author="MACFARLANE " 

 lab="MACFARLANE " 

 dist="DIST-DEC95" 

 rdate=" " 

 rev="6" 

 version="6" 

 awr="11.898" 

 lis="0" 

 scattering="0.0" 

 nis="0" 

 file7="yes" 

 file7temps="296.0 400.0 500.0 600.0 700.0 800.0 1000.0 1200.0 1600.0 2000.0" 

 /> 

XML templates used to create the AMPX input files are provided in Appendix A. The various templates 

can refer to the attribute values by name. For example writeData(endf) in the templates provided in 

Appendix A, will be substituted by the ENDF MAT number listed in the endf attribute. Table 2.1 lists 

some specialized additional values that can also be substituted. 

In addition, the template reader reads a configuration file. The configuration file lists metastable isotopes, 

thermal moderator evaluations, and some special nuclei. In most cases the identifier (ID) used in SCALE 

is the ZA value; however, in the cases of metastable isotopes, thermal moderator evaluations, and some 

special nuclei, the SCALE ID is different. The configuration file allows supplying those special SCALE 

ID values. For the thermal moderators, one or more fast evaluations, identified by their endf numbers, are 

also supplied. The following is the configuration file used for the ENDF 6 data. 

<ConfigFile> 

 <!-- metastable nuclei for which the scale id is different from the real za value --> 

 <metastable> 

 <nuclei endf="9547" realza="95242" scaleza="95601"/> <!-- Am242m --> 

 . . . 

 <nuclei endf="5253" realza="52129" scaleza="52611"/> <!-- Te129m --> 

 <nuclei endf="6153" realza="61148" scaleza="61601"/> <!-- Pm148m --> 

 </metastable> 

 <!-- nuclei for which the scale id is different from the real za value --> 

 <specialNuclei> 

 <nuclei endf="1001" realza="1001" scaleza="1801"/> 

 <nuclei endf="1002" realza="1001" scaleza="1802"/> 

 </specialNuclei> 

 <!-- thermal nuclei - need to check the temperature stuff again - these are the 

temperatures listed in file 7 --> 

 <thermal> 

 <nuclei endf="1" realza="101" scaleza="1001"> <!-- H_H20 --> 

 <fastMat endf="125"/> <!-- H --> 

 </nuclei> 
 

5 

 . . . 

 <nuclei endf="27" realza="127" scaleza="4509"> <!-- BEO --> 

 <fastMat endf="425"/> <!-- BE9 --> 

 <fastMat endf="825"/> <!-- 016 --> 

 </nuclei> 

 <nuclei endf="31" realza="131" scaleza="6312"> <!-- GRAPHITE --> 

 <fastMat endf="600"/> <!-- C --> 

 </nuclei> 

 </thermal> 

</ConfigFile> 

In addition to the attributes available directly from the above XML file, the tags listed in Table 2.1 can 

also be substituted. 

Table 2.1 Substitution tags available for use in XML input templates 

Attribute Description 

tag The tag name of the material 

version The version of the evaluation 

mod The modification of the evaluation 

rev The revision of the evaluation 

tape The tape location of the evaluation 

za The ZA ID of the evaluation 

a The mass of the evaluation 

z The z value of the evaluation 

scaleid or 

scaleidthermala1 

The SCALE ZA ID. This is: endf + 10000 * mod + 1000000*version. If this is a 

thermal material, the ENDF MAT ID of the first thermal evaluation is used. Similarly, 

the mod number of the first thermal evaluation is used. The version is always of the 

current material. 

scaleidthermal2 The SCALE ZA ID. This is: endf + 10000 * mod + 1000000*version. If this is a 

thermal material and a second thermal evaluation exists, the ENDF MAT ID of the 

second thermal evaluation is used. Similarly, the mod number of the first thermal 

evaluation is used. The version is always of the current material. 

thermalscaleid The same as scaleid but endf and mod value are from the current material, which is 

assumed to be a moderator. 

changedza or 

changedzathermal1 

For thermal, metastable, and some special material a changed ZA value is used in 

SCALE. This tag prints out the changed ZA value. If the ZA value does not need any 

changes, the original value is printed. 

changedzatherml2 For thermal, metastable, and some special material a changed ZA value is used in 

SCALE. This tag prints out the changed ZA value for the second thermal evaluation. 

Date Prints the current date. The second argument optionally lists the date format. 
6 

The terms fast1, fast2, or gamma may be appended to all available attributes. If fast1 is appended, the 

values of the attributes for the first fast thermal evaluation are printed. For example, in the case of 

graphite, appending fast1 to the writeData attribute [writeData(tagfast1)] substitutes C, the tag for the 

first fast evaluation to be coupled with graphite. If fast2 is appended, the values of the attributes for the 

second fast thermal evaluation are printed. For example, writeData(endffast2) for the case of BeO 

substitutes 825, the ENDF material for the second fast evaluation to be coupled with BeO. If gamma is 

appended, the value of the attribute pertaining to the coupled gamma evaluation is printed. 

The writeData(attribute) function can optionally have two more parameters, the first of which gives the 

total length of the string to be printed and the second one the character to use for padding. Thus, in the 

case of graphite, the function writeData(endf,10,*) will print “*******600.” 

The templates are XML files with several special elements that allow looping over the various 

evaluations. The tags listed in Table 2.2 control the behavior of the templates. Templates for the input 

files used for the library creation effort are given in Appendix A. 

Each restriction can contain more than one value and those values can be preceded by a plus (+) or a 

minus (−). A plus indicates that the indicated condition must be met, whereas a minus negates the 

indicated condition. For example, restrict=“+gamma + file2(yes)” will only loop over evaluations that 

have a File 2 and have an associated gamma evaluation. However, restrict=“gamma file2(yes)” will 

only loop over evaluations that have a File 2 or have an associated gamma evaluation. The restriction 

restrict=“−gamma +file2(yes)” will only loop over evaluations that have a File 2 and no associated 

gamma evaluation. 

The templates used to generate the various input files are given in Appendix A and should contain enough 

comments to serve as examples on how to use the input generator. 

The template is parsed using a Java program where the program invocation is typically as follows. 

java -jar ampx.jar -l Endf6.xml -c Config.xml -l template.xml 

The argument for the −t option gives the file name of the template name, the argument for the −l option 

gives the file name containing the XML list of the ENDF evaluations, and the argument for the 

−c option lists the configuration file. 
 

7 

Table 2.2 Template control elements that allow looping over the evaluations 

Tag Description 

loop starts a loop over the various evaluations. It has the following optional attributes. 

restrict Restricts the loop to special cases. Available values are as follows. 

 thermal or 

thermal1 

Only loops over materials that are moderators. 

thermal2 Only loops over materials that are moderators and have 

second fast evaluations. 

 gamma or 

gammathermal1 

Only loops over materials that have an associated gamma 

evaluation. In the case of moderators, the gamma 

evaluations need to exist for the first fast evaluation. 

gammathermal2 Only loops over materials that are moderators and have at 

least two fast evaluations and the second fast evaluation 

has a gamma evaluation associated with it. 

Metastable Only loops over materials that are metastable. 

For any attribute listed in the XML file that has the associated value. For 

example restrict=“file2(yes),” only loops over evaluations that contain a File 2. 

start The index at which to start. Before the value is used, any writeData functions 

are substituted. 

stop The index at which to stop. Before the value is used, any writeData functions 

are substituted. 

step The index at which to stop the loop. Before the value is used, any writeData

functions are substituted 

number The number of times the loop should run. Before the value is used, any 

writeData functions are substituted. 

openFile With attribute name opens the indicated file for writing. The file name must be unique as it is 

used as the key. 

closeFile With attribute name closes a previously opened file. 

writeFile With attribute name writes data to the indicated file. Optionally it can contain an attribute 

restrict, which has the same function as for the loop element to restrict what gets written into the 

file. Any CDATA element and anything enclosed in a text element is written to the indicated 

file. Because a text element allows a restrict attribute with the same functionality as for the loop 

element, enclosure into this element allows writing different text depending on the particulars of 

the current evaluations. 

noLast Allows printing content if currently in a loop and if this is not the last time the loop will run. 

first Allows printing content if currently in a loop and if this is first time the loop will run. 

noFirst Allows printing content if currently in a loop and if this is not the first time the loop will run. 
8 

2.2 Processing

2.2.1 Group structure 

The neutron group structure used is the 199-group VITAMIN-B6 structure with one extra group added to 

bring the upper group boundary to 20 MeV. The neutron group structure is listed in Table 2.3 and the 

gamma group structure in Table 2.4. For the broad-group library, 27 neutron groups and 19 gamma 

groups were used. The neutron group boundaries are listed in Table 2.5 and the gamma group boundaries 

in Table 2.6. 

Table 2.3 Energy boundaries for the 200 neutron group structure 

Energy boundaries for the 200 neutron groups used for the fine-group library 

1 2.0000E+07 42 2.7253E+06 83 3.3373E+05 124 2.3579E+04 165 5.0435E+00 

2 1.9640E+07 43 2.5924E+06 84 3.0197E+05 125 2.1875E+04 166 3.9279E+00 

3 1.7332E+07 44 2.4660E+06 85 2.9849E+05 126 1.9305E+04 167 3.0590E+00 

4 1.6905E+07 45 2.3852E+06 86 2.9721E+05 127 1.5034E+04 168 2.3824E+00 

5 1.6487E+07 46 2.3653E+06 87 2.9452E+05 128 1.1709E+04 169 1.8554E+00 

6 1.5683E+07 47 2.3457E+06 88 2.8725E+05 129 1.0595E+04 170 1.4450E+00 

7 1.4918E+07 48 2.3069E+06 89 2.7324E+05 130 9.1188E+03 171 1.3000E+00 

8 1.4550E+07 49 2.2313E+06 90 2.4724E+05 131 7.1017E+03 172 1.1253E+00 

9 1.4191E+07 50 2.1225E+06 91 2.3518E+05 132 5.5308E+03 173 1.0800E+00 

10 1.3840E+07 51 2.0190E+06 92 2.2371E+05 133 4.3074E+03 174 1.0400E+00 

11 1.3499E+07 52 1.9205E+06 93 2.1280E+05 134 3.7074E+03 175 1.0000E+00 

12 1.2840E+07 53 1.8268E+06 94 2.0242E+05 135 3.3546E+03 176 8.7643E-01 

13 1.2523E+07 54 1.7377E+06 95 1.9255E+05 136 3.0354E+03 177 8.0000E-01 

14 1.2214E+07 55 1.6530E+06 96 1.8316E+05 137 2.7465E+03 178 6.8256E-01 

15 1.1618E+07 56 1.5724E+06 97 1.7422E+05 138 2.6126E+03 179 6.2506E-01 

16 1.1052E+07 57 1.4957E+06 98 1.6573E+05 139 2.4852E+03 180 5.3158E-01 

17 1.0513E+07 58 1.4227E+06 99 1.5764E+05 140 2.2487E+03 181 5.0000E-01 

18 1.0000E+07 59 1.3534E+06 100 1.4996E+05 141 2.0347E+03 182 4.1399E-01 

19 9.5123E+06 60 1.2874E+06 101 1.4264E+05 142 1.5846E+03 183 3.6680E-01 

20 9.0484E+06 61 1.2246E+06 102 1.3569E+05 143 1.2341E+03 184 3.2500E-01 

21 8.6071E+06 62 1.1648E+06 103 1.2907E+05 144 9.6112E+02 185 2.7500E-01 

22 8.1873E+06 63 1.1080E+06 104 1.2277E+05 145 7.4852E+02 186 2.2500E-01 

23 7.7880E+06 64 1.0026E+06 105 1.1679E+05 146 5.8295E+02 187 1.8400E-01 

24 7.4082E+06 65 9.6164E+05 106 1.1109E+05 147 4.5400E+02 188 1.5000E-01 

25 7.0469E+06 66 9.0718E+05 107 9.8037E+04 148 3.5357E+02 189 1.2500E-01 

26 6.7032E+06 67 8.6294E+05 108 8.6517E+04 149 2.7536E+02 190 1.0000E-01 

27 6.5924E+06 68 8.2085E+05 109 8.2503E+04 150 2.1445E+02 191 7.0000E-02 

28 6.3763E+06 69 7.8082E+05 110 7.9499E+04 151 1.6702E+02 192 5.0000E-02 

29 6.0653E+06 70 7.4274E+05 111 7.1998E+04 152 1.3007E+02 193 4.0000E-02 

30 5.7695E+06 71 7.0651E+05 112 6.7379E+04 153 1.0130E+02 194 3.0000E-02 

31 5.4881E+06 72 6.7206E+05 113 5.6562E+04 154 7.8893E+01 195 2.1000E-02 

32 5.2205E+06 73 6.3928E+05 114 5.2475E+04 155 6.1442E+01 196 1.4500E-02 

33 4.9659E+06 74 6.0810E+05 115 4.6309E+04 156 4.7851E+01 197 1.0000E-02 

34 4.7237E+06 75 5.7844E+05 116 4.0868E+04 157 3.7266E+01 198 5.0000E-03 

35 4.4933E+06 76 5.5023E+05 117 3.4307E+04 158 2.9023E+01 199 2.0000E-03 

36 4.0657E+06 77 5.2340E+05 118 3.1828E+04 159 2.2603E+01 200 5.0000E-04 

37 3.6788E+06 78 4.9787E+05 119 2.8501E+04 160 1.7604E+01 201 1.0000E-05 

38 3.3287E+06 79 4.5049E+05 120 2.7000E+04 161 1.3710E+01 

39 3.1664E+06 80 4.0762E+05 121 2.6058E+04 162 1.0677E+01 

40 3.0119E+06 81 3.8774E+05 122 2.4788E+04 163 8.3153E+00 

41 2.8651E+06 82 3.6883E+05 123 2.4176E+04 164 6.4760E+00 
 

9 

Table 2.4 Energy boundaries for the 47 gamma group structure

Energy boundaries for the 47 gamma groups used for the fine-group library 

1 2.0000E+07 11 5.0000E+06 21 1.8000E+06 31 7.0000E+05 41 1.0000E+05 

2 1.4000E+07 12 4.5000E+06 22 1.6600E+06 32 6.0000E+05 42 7.5000E+04 

3 1.2000E+07 13 4.0000E+06 23 1.5700E+06 33 5.1200E+05 43 7.0000E+04 

4 1.0000E+07 14 3.5000E+06 24 1.5000E+06 34 5.1000E+05 44 6.0000E+04 

5 8.0000E+06 15 3.0000E+06 25 1.4400E+06 35 4.5000E+05 45 4.5000E+04 

6 7.5000E+06 16 2.7500E+06 26 1.3300E+06 36 4.0000E+05 46 3.0000E+04 

7 7.0000E+06 17 2.5000E+06 27 1.2000E+06 37 3.0000E+05 47 2.0000E+04 

8 6.5000E+06 18 2.3500E+06 28 1.0000E+06 38 2.6000E+05 48 1.0000E+04 

9 6.0000E+06 19 2.1500E+06 29 9.0000E+05 39 2.0000E+05 

10 5.5000E+06 20 2.0000E+06 30 8.0000E+05 40 1.5000E+05 

Table 2.5 Energy boundaries for the 27 neutron group structure

Energy boundaries for the 27 neutron groups used for the broad-group library 

1 2.0000E+07 7 4.0762E+05 13 2.9023E+01 19 1.0000E+00 25 5.0000E-02 

2 6.3763E+06 8 1.1109E+05 14 1.0677E+01 20 8.0000E-01 26 3.0000E-02 

3 3.0119E+06 9 1.5034E+04 15 3.0590E+00 21 4.1399E-01 27 1.0000E-02 

4 1.8268E+06 10 3.0354E+03 16 1.8554E+00 22 3.2500E-01 28 1.0000E-05 

5 1.4227E+06 11 5.8295E+02 17 1.3000E+00 23 2.2500E-01 

6 9.0718E+05 12 1.0130E+02 18 1.1253E+00 24 1.0000E-01 

Table 2.6 Energy boundaries for the 19 gamma group structure

Energy boundaries for the 19 gamma groups used for the broad-group library 

1 2.0000E+07 5 5.0000E+06 9 2.0000E+06 13 8.0000E+05 17 2.0000E+05

2 1.0000E+07 6 4.0000E+06 10 1.6600E+06 14 6.0000E+05 18 1.0000E+05 

3 8.0000E+06 7 3.0000E+06 11 1.3300E+06 15 4.0000E+05 19 4.5000E+04 

4 6.5000E+06 8 2.5000E+06 12 1.0000E+06 16 3.0000E+05 20 1.0000E+4 

2.2.2 AMPX execution sequences for neutron and gamma-ray yield processing 

The ORNL AMPX4

 processing system contains more than 100 distinct modules that can be used to 

perform various nuclear data processing functions. For generating a cross-section library, the specific 

modules needed to process a cross-section evaluation depend upon the data specified in the nuclide 

evaluation. For example, if the nuclide is a resonance nuclide, it will require more modules to be 

executed relative to a nonresonance nuclide. If the evaluation is a thermal moderator material, the AMPX 

execution sequence will be different because the final cross-section data will be a mixture of data from 

one evaluation above the thermal range and from another evaluation in the thermal range. 

The general procedure to generate groupwise neutron cross-section data is outlined in Fig. 2.2. The flow 

chart to produce the coupled part of the library is depicted in Fig. 2.3. 
10 

Fig. 2.2 AMPX sequence for producing neutron multigroup data. 
11 

Fig. 2.3 AMPX sequence to generate the coupled gamma-neutron multigroup data. 
 

12 

The ENDF/B-VI.8 and ENDF/B-VII.0 cross-section evaluations were downloaded from the NNDC Web 

page.2

 Some of the ENDF/B-VI.8 evaluations contain incomplete evaluations preventing them from 

being used to generate a complete cross-section library for radiation transport calculations. These 

evaluations are excluded and are listed in Table 2.7. The ENDF/B-VI.8 evaluation for S (ENDF MAT 

1600) does not list MT=18, fission, in File 3; therefore, the cross-section data for this reaction only extend 

from 1 × 10−5

 eV to 1.0921 × 106 eV as derived from the resolved resonance data in File 2, which 

provides fission width data for some of the resonances. Clearly, the presence of fission width information 

for S in File 2 is not correct as S is not a fissionable nuclide. 

Table 2.7 Evaluations excluded from the ENDF/B-VI.8 library 

Evaluation ENDF 

material Reason for exclusion 

40Ar 1837 Evaluation is missing a file 3 and thus does not yield energy data above 

500000.0 eV (end of resolved resonance data). 

115In 4931 File 3 only gives cross-section data for MT=51 and 102, corresponding to 

(z,n1) and (z,γ). 

41K 1931 File 3 only gives cross-section data for MT=103, corresponding to (z,p). 

24Mg 1225 File 3 only gives cross-section data for MT=102, corresponding to (z,γ). 

100Mo 4249 File 3 only gives cross-section data for MT=102, corresponding to (z,γ). 

98Mo 4243 File only gives cross-section data for MT=102, corresponding to (z,γ). 

92Mo 4225 File 3 only gives cross-section data for MT=102, corresponding to (z,γ). 

59Ni 2828 File 3 only gives cross-section data for MT=102, 103, and 107, 

corresponding to (z,γ), (z,p), and (z,α). 

46Ti 2225 File 3 only gives cross-section data for MT=103. 

47Ti 2228 File 3 only gives cross-section data for MT=28 and 103, corresponding to 

(z,np) and (z,p). 

48Ti 2231 File 3 only gives cross-section data for MT=28, 103, and 107, 

corresponding to (z,np), (z,p), and (z,α). 

50Ti 2237 File 3 only gives data for MT=107 and (z,α). 

In the case of the ENDF/B-VII.0 library, the thermal evaluations for U(UO2) and O(UO2) are excluded as 

there are some inconsistencies in the ENDF evaluations that prevent the AMPX module Y12 from 

processing the data. 
13 

2.2.2.1 Point-wise cross-section data 

The following AMPX modules are used to create the point-wise neutron cross-section data. 

POLIDENT 

Whether cross sections are being prepared for continuous energy applications (CE-KENO or the 

CENTRM module in SCALE for resonance self-shielding) or for multigroup applications, the first 

module that will be executed is the POLIDENT module. For resonance nuclides, POLIDENT contains a 

procedure that determines a point energy mesh in the resolved resonance region and calculates the 

cross sections on the energy mesh. An analogous procedure is performed for nuclides having unresolved 

resonance data. For all nuclides, POLIDENT combines the resonance functions with the ENDF/B File 3 

point data files to produce linearized cross-section functions. The combined data functions are written out 

in a standard AMPX tabular data format which is based on the ENDF/B TAB1 format.5

 

TGEL and DCON 

One of the primary reasons to linearize the cross-section data is to produce internally consistent data sets. 

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 linearizing the data that are given in an evaluation does not lead 

to this consistency. To ensure consistency, the total cross section and other “redundant” reactions 

(e.g., fission, inelastic) are formed by summing the partial values. The TGEL or DCON modules are 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, need 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. The output data from BROADEN are exported in an AMPX tabular format. 

JERGENS 

An energy-dependent weighting function is needed to produce multigroup cross sections. The JERGENS 

module provides a large 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. To calculate the group-averaged infinite dilute cross sections, JERGENS 

was used to produce a “smooth” weighting spectrum that has the form of a Maxwellian—1/E—fission 

spectrum—1/E above 10 MeV. The weighting spectrum used for infinitely dilute group cross sections is 

shown in Fig. 2.4. The weighting spectrum is normalized with the factor 1/(σT + σ0), where σT is the total 

cross section and σ0 the background cross section. 
14 

Fig. 2.4 Weighting spectrum generated by JERGENS. 

Input template used 

The input was generated using an XML template. The XML template used is given in Appendix A. The 

following is the generated AMPX input file for 235U. 

=shell 

ln -sf /Users/dw8/ampx/running/endf7/rel0/n-092_U_235.endf ft11f001 

end 

=polident 

-1$$ 20000 

0$$ 31 32 e 1$$ 1 t 

2$$ 9228 11 2 6 e 

4** a5 0.001 e 

6$$ a3 0 15000 t 

end 

=shell 

cp ft31f001 ${RTNDIR}/../neutron/u235_polident 

cp ft32f001 ${RTNDIR}/../neutron/u235_polident_ft32 

end 

=tgel 

-1$$ 5000000 

0$$ 31 33 e t 

end 

=broaden 

limit=5000000 

logpt=33 logdp=34 

t= 0 300 600 900 1200 2400 

end 

=shell 

cp ft34f001 ${RTNDIR}/../neutron/u235_broaden 
15 

end 

=pickeze 

-1$$ 3000000 

0$$ 34 35 

1$$ 1 0 0 1 0 e t 

2$$ 9228 

5** 300.0 t 

end 

=jergens 

-1$$ a11 3000000 e 

0$$ 35 30 18 1$$ 2 t 

3$$ 1 0 4 t 

3$$ 1099 4 0 t 

 2099 0 read 1.0 -1 0 

 2099 0 save 0 0 0 

 1099 2099 div 1.0 9228 1 

 1099 0 save 0 0 0 

end 

=shell 

cp ft30f001 ${RTNDIR}/../neutron/u235_weight 

end 

Input files have been processed on a DEC Alpha Tru64 machine, a 32-bit Linux machine, and a 64-bit 

Intel Mac. The differences, between the point-wise data from the different computing systems, are mostly 

smaller than 10−3

, the precision to which the energy grid was determined. In a few instances 

(

148Nd, 108Cd, 239Pu) larger differences are found due to rounding differences between the three platforms. 

For the final library, the data generated on the 64-bit Intel Mac were used. 

The BROADEN module does not add additional points while performing Doppler broadening. 

This posed a problem at the boundary between the resolved and unresolved region and at the end of the 

unresolved range when generating Bondarenko factors. As the cross section in the unresolved region 

does not vary rapidly, normally relatively few energy points are needed to describe the point-wise 

cross-section data adequately. However, at the low energy end and the high energy range of the 

unresolved resonance region, the point-wise cross section may show discontinuities. This is an artifact of 

the evaluation. The ENDF formats and procedures5

 do not define the procedure for treating 

discontinuities. The BROADEN module incorrectly interprets these discontinuities as sharp resonances 

unless there is a sufficient number of energy points on either site of the discontinuity. Therefore, the 

module POLIDENT has been updated to add 10 additional energy points at both ends of the unresolved 

resonance range to correct the problem for BROADEN. 

2.2.2.2 Comparison with NJOY 

For verification purposes, the AMPX-processed data were compared to NJOYgenerated data.6

 To 

investigate differences between the two programs that are relevant to the coupled library productions, 

we converted the point-wise data into group averaged data using a flux of Maxwellian 1/E fission 

spectrum—1/E above 10 MeV. Group averaged data were compared at 300 K because this temperature is 

used for the coupled library. Because AMPX leaves redundant reactions in the master library, whereas 

NJOY deletes them, these reactions are excluded from the comparison. Overall, the agreement with 

NJOY is good. 

Differences are mostly due to the fact that the Doppler broadening module, BROADR, in NJOY adds 

additional data points whereas the AMPX BROADEN routine does not. As an example, we present the 

capture cross section in 140Ce in Fig. 2.5. 
16 

Fig. 2.5 Point-wise capture cross section as processed with AMPX and NJOY for 140Ce. 

2.2.2.3 Neutron and gamma yield processing 

Fast neutron evaluation 

Y12 

A feature unique to the AMPX system is the Y12 module. This module reads an ENDF/B evaluation and 

produces tabular double-differential collision kinematics data for all reactions. The functions it produces 

tell what particles are produced and at what energies and directions when a reaction is encountered at 

some source energy point. 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 factors and groupto-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 the AMPX master library 

format. 
17 

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, along with the point data files from 

POLIDENT and JERGENS are passed to X10 to produce gamma-ray yield transfer matrices. The gamma 

yield matrices tell where gamma rays are produced as a function of 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. 

Input template used 

The input was generated using an XML template. The XML template used is given in Appendix A. 

The following is the generated AMPX input file for 235U. 

=shell 

ln -sf /Users/dw8/ampx/running/endf7/rel0/n-092_U_235.endf ft11f001 

ln -sf ${RTNDIR}/../neutron/u235_polident ft31f001 

ln -sf ${RTNDIR}/../neutron/u235_broaden ft34f001 

ln -sf ${RTNDIR}/../neutron/u235_weight ft30f001 

end 

=y12 

0$$ 32 11 0 0 

1$$ 9228 

2$$ 2 6 

3$$ 32 8 8 8 8 5 t 

t 

end 

=pickeze 

-1$$ 3000000 

0$$ 34 35 

1$$ 1 0 0 1 0 e t 

2$$ 9228 

5** 300.0 t 

end 

=x10 

neutron 

-1$$ 3000000 

0$$ 1 30 35 32 

1$$ 9228 238 90 0 5 0 0 

2$$ 99 2099 9228 9228 3** 1.15860E+01 6$$ 99 2099 t 

t 

u235 9228 ENDFBV7 REL0 REV7 MOD7 AMPX 10/23/07 

end 

=rade 

1$$ 1 e t 

end 

=shell 

cp ft01f001 ${RTNDIR}/../neutron/u235_fast 

end 

=x10 

yield 

-1$$ 9000000 

0$$ 71 30 35 32 

1$$ 9228 200 0 47 5 2 5 

2$$ 99 2099 9228 9228 3** 1.15860E+01 6$$ 99 2099 t 

t 

u235 9228 ENDFBV7 REL0 REV7 MOD7 AMPX 02/11/08 

end 

=rade 

1$$ 71 e t 

end 

=shell 

cp ft71f001 ${RTNDIR}/..//neutron/u235_yield 

end 
18 

Input files have been processed on a DEC Alpha Tru64 machine, a 32-bit Linux machine, and a 64-bit 

Intel Mac. For the final library, the data generated on the 64-bit Intel Mac were used. 

2.2.2.4 Thermal neutron processing 

FLANGE6

If the nuclide is a nonmoderator material, the FLANGE6 module is used to produce thermal scattering 

matrices at 300, 600, 900, 1200, and 2100 K based on the free-gas scattering model. The matrices are 

exported in the AMPX master library format. 

JERGENS, Y12, and X10 

If the nuclide is a moderator material with ENDF S(α,β) data (e.g., water-bound hydrogen, graphite), the 

thermal ENDF/B evaluations are processed by the Y12 module to produce thermal kinematics data as a 

function of temperatures provided in the ENDF/B evaluation. The thermal scatter kernel for other 

materials is represented as free gas. In addition, the Y12 module produces thermal point-wise 

cross-section data. The resulting thermal kinematics file and thermal point-wise cross-section file are 

passed to the X10 module, along with the JERGENS weighting spectrum, to produce a special thermal 

AMPX master library. 

Special requirements for BeO 

In all of the ENDF/B thermal scattering laws except BeO, 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 having hydrogen present in a water 

molecule. The file then specifies that oxygen should simply be treated by the free-gas model. This is 

very convenient because it 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 BeO, it is not possible to separate the beryllium behavior 

(or oxygen) out of the data for the BeO molecule. The scattering law data are given for the molecule as a 

combined unit. There are several ways one can deal with this situation. The one that was chosen to be 

analogous to the water-bound 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 is prepared and called “O bound in BeO.” For the O in BeO case, the oxygen data are 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). 

Input for moderator data 

The input was generated using an XML template. The XML template used is given in Appendix A. 

The following is the generated AMPX input file for hydrogen in H2O. 

=shell 

ln -sf /Users/dw8/ampx/running/endf7/rel0/tsl-HinH2O.endf ft11f001 

ln -sf ${RTNDIR}/../groupStructure ft47f001 

end 

=y12 

0$$ 37 11 38 0 

1$$ 1 

2$$ 2 6 

3$$ 32 8 8 8 8 5 t 

t 

end 
19 

=shell 

cp ft37f001 ${RTNDIR}/../thermal/h_h2o_kin 

cp ft38f001 ${RTNDIR}/../thermal/h_h2o_point 

end 

=jergens 

0$$ 11 30 18 

1$$ 1 t 

3$$ 1099 0 2 t 

end 

=x10 

neutron 

0$$ 4 30 38 37 

1$$ 1 200 34 

 0 

 5 2 0 

2$$ 99 1099 1 1 

3** 0 t 

t 

h_h2o 1 ENDFB V6 REL0 REV7 MOD0 AMPX 12/09/07 

end 

=shell 

cp ft04f001 ${RTNDIR}/../thermal/h_h2o_master 

end 

=rade 

1$$ 4 e t 

end 

=paleale 

0$$ 4 e 2$$ 1 0 e 5$$ f 200 t 

end 

Input files have been processed on a DEC Alpha Tru64 machine, a 32-bit Linux machine, and a 64-bit 

Intel Mac. For the final library, the data generated on the 64-bit Intel Mac were used. Differences 

between the results on the three platforms were within expected ranges. 

Input for nonmoderator data 

The input was generated using an XML template. The XML template used is given in Appendix A. 

The following is the generated AMPX input file for 235U. 

=shell 

ln -sf ${RTNDIR}/../../neutron/u235_weight ft30f001 

ln -sf ${RTNDIR}/../groupStructure ft47f001 

=flange6 

nl=3 za=92235 awr=233.0248 free=1.15860E+01 

mat=0 master=41 igm=200 neg=34 

t=300 t=600 t=900 t=1200 t=2100 

end 

=shell 

cp ft41f001 ${RTNDIR}/../neutron/u235_flange6 

end 

Input files have been processed on a DEC Alpha Tru64 machine, a 32-bit Linux machine, and a 64-bit 

Intel Mac. The cross-section data produced on the 64-bit Intel Mac are used in the final coupled library. 

Differences between the results on the three platforms exist, even if the Legendre coefficients are 

expanded and the real scattering matrices are compared. This is most likely due to rounding effects 

because mostly small scattering values are affected. The differences do not have significant impact on the 

library data. 
20 

2.2.2.5 Generating full-range Bondarenko factors and thermal data 

To facilitate self-shielding calculations, full-range Bondarenko factors for several temperatures and 

background cross sections are supplied on the coupled library. The Bondarenko method is basically a 

self-shielding method that parameterizes cross sections for a nuclide as a function of temperature, T, and 

the “background” cross section, σ0, of all the other nuclides mixed with the nuclide. Simplistically, given 

the temperature and background cross-section values, one determines self-shielded cross sections by 

interpolating in tables. Since self-shielding causes the “background” values that a nuclide sees to change, 

an iterative procedure involving all nuclides is used. 

PRUDE 

For nuclides with unresolved resonance data, the PRUDE module must be run 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. In the sequences that were prepared, 

temperatures of 300, 900, and 2100 K were chosen, along with background values of 10−6

, 1, 10, 100, 

1000, 104

, 105, 106, and 108. The cross sections are written out in an AMPX tabular file. This module can 

only be run for mono-isotopic evaluations. 

FABULOUS 

For all nuclides, the point data from PRUDE (if applicable) and POLIDENT are passed to the 

FABULOUS module with the weighting spectrum file from JERGENS. FABULOUS uses the same 

temperatures and background cross sections described for PRUDE to numerically Doppler broaden the 

functions exterior to the unresolved region, 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. 

Input files 

The input was generated using an XML template. The XML template used is given in Appendix A. 

The following is the generated AMPX input file for 235U. 

=shell 

ln -sf /Users/dw8/ampx/running/endf7/rel0/n-092_U_235.endf ft11f001 

ln -sf ${RTNDIR}/../../neutron/u235_polident ft31f001 

ln -sf ${RTNDIR}/../../neutron/u235_weight ft30f001 

ln -sf ${RTNDIR}/../groupStructure ft47f001 

ln -sf ${RTNDIR}/../../bon_test/neutron/u235_polident ft31f001 

ln -sf ${RTNDIR}/../../bon_test/neutron/u235_weight ft30f001 

end 

=prude 

0$$ 36 

1$$ 1 t 

2$$ 9228 9 3 11 2 t 

3** 1+8 1+6 1+5 1+4 1000 100 10 1 1-6 

4** 300.0 900.0 2000.0 e 

5** 0 1 e t 

end 

=shell 

cp ft36f001 ${RTNDIR}/../neutron/u235_prude 

end 
21 

=tgel 

-1$$ 9000000 

0$$ 31 32 e t 

end 

=fabulous 

Title Card for FABULOUS case 

-1$$ 2500000 

0$$ 2 32 30 36 

1$$ 9228 9228 3 9 200 

2$$ 99 2099 

5** 233.0248 0.0001 t 

8** 300.0 900.0 2000.0 

9** 1+8 1+6 1+5 1+4 1000 100 10 1 1-6 

t 

end 

=shell 

cp ft02f001 ${RTNDIR}/../neutron/u235_fabulous 

end 

Input files have been processed on a DEC Alpha Tru64 machine, a 32-bit Linux machine, and a 64-bit 

Intel Mac. The data generated on the 64-bit Intel Mac were used for the final library. 

During the course of the library generation, several issues were discovered and corrected in PRUDE and 

FABULOUS. The PRUDE module calculates the point-wise cross sections for the unresolved resonance 

range for different values of σ0, combining File 2 and File 3 data if necessary. The same calculation is 

done in the POLIDENT module for infinite dilution. The two modules produced inconsistent results 

because the combining was done in two different ways. The PRUDE module was updated to use the 

same methods as POLIDENT. The infinite diluted cross-section data are now consistent between the two 

modules. 

Initially the FABULOUS module could not process all evaluations because some of the unresolved 

cross-section data are negative for lower values of σ0 (i.e., highly self-shielded). This is most likely due 

to Single-Level Breit-Wigner (SLBW) approximation. Calculation of the “total” Bondarenko factor is 

given by 

 . (2.1) 

The value for ( ) σ T +σ 0 can become negative. The NJOY6

 manual describes how the calculation in the 

UNRES function is done, which is similar to the PRUDE calculation. The manual states that for the 

higher actinides the approximation used in calculating this quantity can produce negative shielded 

cross-section data for lower values of σ0. This should not be a problem as the higher actinides are rarely 

present in these quantities in a reactor. The affected evaluations are 238Np, 242mAm, 242Am, and 253Cf. 

In addition, an infinite loop was discovered in FABULOUS for 238Np because the code tried to use a 

step-size smaller than the available computer precision. This problem and some overstepped bounds 

issues have been fixed. 

Greater than expected differences were found between the Bondarenko factors calculated in Linux and on 

Tru64. These differences occurred because the Doppler broadening routine in FABULOUS converts the 

energy into velocity-like terms for the Bondarenko calculation. These terms are converted back to energy 

afterwards. On occasion, the end points of the unresolved resonance range may not be included in the 

( )

( ) ∫

∫

2

0

2

0

σσ

φ

σσ

φσ

+

+

T

T

T
22 

final point-wise cross-section data due to this conversion. The code has been updated to always include 

these end points. After these changes, good agreement is obtained for the Bondarenko factors calculated 

on the three platforms. 

2.2.2.6 Gamma processing 

The processing for gamma-ray interactions is much simpler in contrast to neutron interactions. Most of 

the important gamma-ray scattering reactions are due to interactions of the gamma ray with the electrons 

in a nucleus. Since the number of electrons in a nucleus is just the Z-value, the cross sections are 

calculated on a per electron basis and simply multiplied by Z to get the value for a nuclide. Since Z is the 

same for all isotopes of an element, one only has to calculate values for the elements. Furthermore, there 

are only two processes—incoherent (Klein-Nishina) and coherent (Rutherford) scattering—to be treated. 

There is only one exception to the number of processes, and this exception is the pair-production of two 

0.511 MeV photons. 

JERGENS 

The JERGENS module is used to produce a weighting spectrum. For gamma-ray processing, the 

spectrum is typically a 1/E spectrum with “roll-offs” that account for significant absorption of the gamma 

rays because of photoelectric absorption. As before, the weighting spectrum is written out in a point-wise 

AMPX tabular file based on the ENDF/B TAB1 format. 

Y12 

The Y12 module produces a tabular kinematics file that describes coherent and incoherent scattering and 

pair production. In addition, Y12 produces a point-wise data file for all of these processes, and the 

point-wise file is exported in an AMPX tabular format. 

X10 

The tabular files from JERGENS and Y12 are passed to X10 with the tabular kinematics file. X10 

produces a group-averaged AMPX master library for gamma-ray interactions. 

RADE and PALEALE 

The RADE module is used to perform internal consistency checks on the AMPX master library 

containing the gamma-ray interaction data. The PALEALE module is used to print the content of the 

library and allows data checking. 

Gamma interaction data used 

The Photo-Atomic interaction sublibrary of ENDF/B-VI.8 and ENDF/B-VII.0 were used in the library 

generation effort. In both cases data exist for the following 100 elements: 

H He Li Be B C N O F Ne Na Mg Al Si P S Cl Ar K Ca Sc Ti V Cr Mn Fe Co Ni Cu Zn Ga Ge 

As Se Br Kr Rb Sr Y Zr Nb Mo Tc Ru Rh Ce Pr Nd Pm Sm Eu Gd Tb Dy Ho Er Tm Yb Lu Hf 

Ta W Re Os Ir Pt Au Hg Tl Pb Bi Po At Rn Fr Ra Ac Th Pa U Np Pu Am Cm Bk Cf Es Fm 
23 

Input used to process gamma cross-section data 

The input was generated using an XML template. The XML template used is given in Appendix A. 

The following is the generated AMPX input file for uranium. 

=shell 

ln -sf /Users/dw8/ampx/running/endf7/rel0/photoat-092_U_000.endf ft11f001 

end 

=y12 

0$$ 32 11 31 0 

1$$ 9200 

2$$ 2 6 

3$$ 32 8 8 8 8 5 t 

t 

end 

=shell 

cp ft32f001 ${RTNDIR}/../gamma/u_kin 

cp ft31f001 ${RTNDIR}/../gamma/u_point 

end 

=jergens 

0$$ 0 30 18 

1$$ 1 

2** 1.0e3 3.0e7 t 

3$$ 1599 0 11 t 

end 

=x10 

gamma 

0$$ 1 30 31 32 

1$$ 9200 0 0 

 47 

 0 0 5 

2$$ 99 1599 9200 9200 

3** 0.0 t 

7** 2.0E+7 1.4E+7 1.2E+7 1.E+7 8.E+6 

 7.5E+6 7.E+6 6.5E+6 6.E+6 5.5E+6 

 5.0E+6 4.5E+6 4.E+6 3.5E+6 3.E+6 

 2.75E+6 2.5E+6 2.35E+6 2.15E+6 2.E+6 

 1.8E+6 1.66E+6 1.57E+6 1.5E+6 1.44E+6 

 1.33E+6 1.2E+6 1.E+6 9.0E+5 8.0E+5 

 7.0E+5 6.0E+5 5.12E+5 5.1E+5 4.5E+5 

 4.0E+5 3.0E+5 2.6E+5 2.0E+5 1.5E+5 

 1.0E+5 7.5E+4 7.0E+4 6.0E+4 4.5E+4 

 3.0E+4 2.0E+4 1.0E+4 e 

t 

u 9200 ENDFB V6 REL8 REV7 MOD1 AMPX 12/09/07 

end 

=shell 

cp ft01f001 ${RTNDIR}/../gamma/u_master 

end 

=rade 

1$$ 1 e t 

end 

=paleale 

0$$ 1 e 2$$ 0 1 e 4$$ 502 504 e 5$$ f 47 t 

end 

Input files have been processed on a DEC Alpha Tru64 machine, a 32-bit Linux machine, and a 64-bit 

Intel Mac. While the data processed on all three platforms were compared, no significant differences 

were found, and the final library uses the data generated on the 64-bit Intel Mac. 
24 

Processing of gamma data 

File 23 containing “smooth” photon interaction cross sections gives the data for MT=502, photon 

coherent scattering in the range from 1.0 eV to 1.0 × 1011 eV. These data need to be combined with form 

factors from File 27, MT=502 (form factor), MT=505 (imaginary scattering factor), and MT=506 

(real scattering factor), which are given over the ranges 0.0 eV to 1.0 × 109 eV, 1.0 eV to 1.0 × 107 eV, 

and 1.0 eV to 1.0 × 107 eV, respectively. The program then tries to interpolate for energies listed for 

MT=502 and warns that the interpolation is outside the range. In the case of tungsten, the two values 

bracketing the 1.0 × 107

 eV in MT=502 are 7.070513 × 106 eV and 2.135261 × 107 eV. The program 

gives a warning and assumes that the desired value is 0.0. 

While the number of energy points in File 23 for MT=502 change for the various elements, the data for 

MT=505 and MT=506 in File 27 come in the range from 1.0 eV to 1.0 × 107

 eV for all elements. Thus, 

the above problem is present for all elements used in the coupled library. 

On Tru64, Y12 fails for hydrogen as the last two interpolation points for File 27 and MT=502 are given 

as (1.0 × 106

; 8.18290 × 10−27) and (1.0 × 109; 8.18290 × 10−39). After the last two values are changed to 

0.0 in the ENDF file, the calculation succeeds. 

Data generated on Linux and on Tru64 show greater than expected differences in the scattering matrix for 

MT=504 for the lower energy groups. This is due to rounding effects in the calculations of the grid for 

the Klein-Nishina scattering, where Y12 reconstructs the point-wise scattering in terms of the 

scatter-cosine, which is given as a function of incident energy E and scattered energy E′ as 

e'

e = 1

1+ e( ) 1− μ , (2.2) 

where e and e′ are the energy in terms of the electron mass unit e= E / 0.5110034 . A function then 

calculates the cosine of the scattering angle from the above formula as 

μ=1+

1

e

− 1

e'

=1+

5.110034 ×105 × (E'−E)

E × E' , (2.3) 

using the first approach after first calculating e and e′, all in single precision. In single precision, the two 

forms of the equation can have two difference values. An example is E = 1.0 eV and E′ = 0.99999607. 

In this case, the first form yields μ = −1.0186844 and the second form μ = −1.0102476 . The later is 

consistent with the value calculated for both forms of the equation if calculation is done in double 

precision. This can lead to large differences. An example is silicon, where the largest difference is found 

scattering from group 41 to 41 in the fourth Legendre coefficient: single precision 4.0716E−02, double 

precision −1.6598E−01. For the final library, all calculations in the Y12 klein_nishina function were 

done internally in double precision and returned as single precision values. 

After these changes, good agreement is obtained for gamma cross-section data generated on the three 

platforms. 
25 

Comparison with NJOY 

For independent verification purposes, the AMPX-processed data were compared to NJOY-generated 

data.6

 The NJOY code was used to process the gamma evaluations. The groupwise ENDF files 

generated by GAMINR were converted to AMPX master files using SMILER. This module converts 

MT=522 to MT=527 and MT=525 to MT=1527—this feature is temporarily disabled to allow direct 

comparison. All NJOY files have the following one-dimensional (1-D) MT values: 501, 502, 504, 516, 

522, 525, and 1599. The MT=1599 gives the flux and MT=525 is not defined in the ENDF-VI or 

ENDF-V reaction list. The NJOY manual states that this is the heating cross section defined in ENDF-V 

as 621. 

AMPX writes out more reaction values because it keeps the redundant reactions. 

The flux written by GAMINR is not normalized to the sum of the flux over all energy groups as is done 

for gamma processes in X10. SMILER has been changed to write out the normalized flux to allow a 

direct comparison between the results from AMPX and NJOY. 

Summation over total and redundant reactions 

X10 first calculates the group averages for all reactions listed in the ENDF file as translated by Y12. 

The following reactions are than treated in a special manner. 

If MT=515 (pair production, electron field) or MT=517 (pair production, nuclear field) exists, the 

redundant reaction MT=516 (pair production) is created as the sum of the two, erasing all previous data in 

the reaction that might have been set by the ENDF evaluation. If MT=516 does not previously exist, it is 

NOT created, even if MT=515 or MT=517 exists. 

If any MT in the range MT=534–572 (subshell photoelectric or electro-atomic cross section for various 

atomic levels) exist, then calculate 

602

MT=522 Photoelectric absorption. (In version 5, this was MT=602.)

MT=602 Photoelectric absorption.

MT=527 ,

MT

MT

MT MT=

= x

= x

= x +x

∑

∑

∑

where x MT is the group averaged cross section for the indicated reaction and the sum extends over 

reaction numbers MT=534–572. The total cross section is then recalculated as 

 . 

(These calculations are done in the subroutine fixsums_g of X10). 

Not all evaluations considered for the coupled library include MT=527 or MT=602, which are added by 

X10 as outlined above. All evaluations include MT=515, 516, and 517, and X10 will recalculate 

MT=516. All evaluations include MT=522, which will be erased by X10 and recalculated as outlined 

above. All evaluations include MT=501, which will be erased by X10 and recalculated as outlined above. 

AMPX defines some special MT values. MT=527 is the sum of all gamma-ray absorption processes 

(in contrast to the ENDF defined value where it is defined as electro-atomic bremsstrahlung). MT=602 is 

still assumed to be photoelectric absorption, and the group-averaged cross-section data for MT=602 and 

502 504 516 522 501=MT =MT =MT =MT =MT x+x+x+x=
26 

MT=522 are identical in the final library. MT=1527 is the gamma energy absorption coefficient or kerma 

factor. NJOY writes this into MT=527 in GAMINR. 

One-dimensional processes 

AMPX and NJOY results are compared for MT=501, 502, 504, 516, and 522 because these cross sections 

are directly comparable between the two programs. No significant differences were found. 

Two-dimensional processes 

To compare the two-dimensional (2-D) processes, we first compared the Legendre coefficients. If there 

was a big difference between NJOY and AMPX for these, we expanded the Legendre series and 

compared the scattering matrices directly. An example of the observed differences can be seen in Fig. 2.6 

for MT=504 in hydrogen. Scattering is from group 1 to group 38. 

Fig. 2.6 AMPX and NJOY scattering data for MT=504 in hydrogen. Scattering 

is from group 1 to group 38. 

2.2.2.7 Creating the master library 

Three steps are needed to create the final master library. 

1. Create a master library for each evaluation containing the neutron data, Bondarenko shielding factors, 

and gamma data. In the case of moderator data, the fast evaluation also needs to be combined into 

this master library. The ENDF material ID needs to be changed to the SCALE ID. 

2. Combine the master libraries for various evaluations into one master library. 
27 

3. Add dose data to the master library. 

The steps are outlined in more detail below. 

Create master library for each evaluation 

The master library for nonmoderator data needs to contain gamma data, fast neutron data, neutron yield 

data, and Bondarenko data. In addition, several SCALE specific identifiers need to be set. Input files are 

generated from XML templates which are given in Appendix A. The following is the generated AMPX 

input for 235U. 

=shell 

ln -sf ${RTNDIR}/../neutron/u235_fast ft01f001 

ln -sf ${RTNDIR}/../neutron/u235_yield ft02f001 

ln -sf ${RTNDIR}/../neutron/u235_fabulous ft03f001 

ln -sf ${RTNDIR}/../neutron/u235_flange6 ft04f001 

ln -sf ${RTNDIR}/../gamma/u_master ft05f001 

end 

=simonize 

Identifier=92235 master=20 source=endf 

title= u235 9228 ENDFB7 REL0 REV7 MOD7 AMPX 02/11/08 

id45=7079228 id46=0 id47=7019200 id48=7079228 

neutron=1 id19=9228 

 yield=2 id19=9228 

 BONDARENKO=3 id19=9228 

2dn=4 id19=92235 

gamma=5 id19=9200 

end 

=rade 

1$$ 20 e t 

end 

=shell 

cp ft20f001 ${RTNDIR}/../master/u235_neutron 

end 

In the case of thermal moderators, the fast evaluations need to be added to the master. The full template 

used to generate the input files is given in Appendix A. The following is the generated AMPX input for 

hydrogen in H2O. 

=shell 

ln -sf ${RTNDIR}/../neutron/h1_fast ft01f001 

ln -sf ${RTNDIR}/../neutron/h1_yield ft02f001 

ln -sf ${RTNDIR}/../neutron/h1_fabulous ft03f001 

ln -sf ${RTNDIR}/../thermal/h_h2o_master ft04f001 

ln -sf ${RTNDIR}/../gamma/h_master ft05f001 

end 

=simonize 

Identifier=1001 master=20 source=endf 

title= Fast: h1 Thermal: h_h2o 1 ENDFB7 REL0 REV7 MOD0 AMPX 02/11/08 

id45=7050125 id46=7000001 id47=7010100 id48=0 

neutron=1 id19=125 

 yield=2 id19=125 

 BONDARENKO=3 id19=125 

2dn=4 id19=1 MT=0 MODERATOR 

gamma=5 id19=100 

end 

=shell 

cp ft20f001 ${RTNDIR}/../master/h_h2o_thermal 

end 
28 

2.2.3 Adding dose data 

The SCALE master library contains dose factor data as listed in Table 2.8. The same dose data are used 

for all three libraries (i.e., the ENDF/B-VI.8 fine-group library and the ENDF/B-VII.0 fine-group and 

broad-group libraries). 

Table 2.8 Available dose functions on the coupled neutron-gamma libraries 

MAT MT Description 

999 1 1/V absorber 

1/v function (normalized to 1.0 at 2200 m/s <==> 0.0253 eV) 

All reaction values contain the same data 

27 

101 

102 

900 9032 International Commission on Radiation Units and Measurements, Report 44 

(ICRU-44), Table B.3 (air) Kerma (Gy/h)/(neutron/cm2

/s) 

9033 ICRU-44, Table B.3 (air) Kerma (rad/h)/(neutron/cm2

/s) 

9034 Ambient dose equivalent (ICRU-57, Table A.42) (Sv/h)/(neutron/cm2

/s) 

9035 Ambient dose equivalent (ICRU-57, Table A.42) (rem/h)/(neutron/cm2

/s) 

9036 Effective dose (ICRU-57, Table A.41) (Sv/h)/(neutron/cm2

/s) 

9037 Effective dose (ICRU-57, Table A.41) (rem/h)/(neutron/cm2

/s) 

9029 American National Standards Institute (ANSI) standard (1977) neutron 

flux-to-dose-rate factors (rem/h)/(neutron/cm2

/s) 

9031 ANSI standard (1991) neutron flux-to-dose-rate factors 

(rem/h)/(neutron/cm2

/s) 

9503 Claiborne-Trubey conversion factors (rad/h)/(photons/cm2

/s) 

9504 ANSI standard (1977) gamma flux-to-dose-rate factors 

(rem/h)/(photons/cm2

/s) 

9505 ANSI standard (1991) gamma flux-to-dose-rate factors 

(rem/h)/(photons/cm2

/s) 

9506 ICRU-57 Table A.21 (air) Kerma (Gy/h)/(photons/cm2

/s) 

9507 ICRU-57 Table A.21 (air) Kerma (rad/h)/(photons/cm2

/s) 

9508 Ambient dose equivalent (ICRU-57 Table A.21) (Sv/h)/(photons/cm2

/s) 

9509 Ambient dose equivalent (ICRU-57 Table A.21) (rem/h)/(photons/cm2

/s) 

9510 Effective dose (ICRU-57 Table A.17) (Sv/h)/(photons/cm2

/s) 

9511 Effective dose (ICRU-57 Table A.17) (rem/h)/(photons/cm2

/s) 

Source: Table S1.4.8 in Ref. 1.
29 

The JERGENS module is used to generate the point-wise data, and the VEL module is used to create the 

group averages. 

Input is as follows for the MAT=900 input. 

=jergens 

-1$$ 2 2 100 900 3 e 

0$$ 0 1 19 1$$ 18 t 

3$$ 9032 0 9032 e t 

3$$ 9033 0 9033 e t 

3$$ 9034 0 9034 e t 

3$$ 9035 0 9035 e t 

3$$ 9036 0 9036 e t 

3$$ 9037 0 9037 e t 

3$$ 9029 0 9029 e t 

3$$ 9031 0 9031 e t 

3$$ 9503 0 9503 e t 

3$$ 9504 0 9504 e t 

3$$ 9505 0 9505 e t 

3$$ 9506 0 9506 e t 

3$$ 9507 0 9507 e t 

3$$ 9508 0 9508 e t 

3$$ 9509 0 9509 e t 

3$$ 9510 0 9510 e t 

3$$ 9511 0 9511 e t 

3$$ 9502 0 9027 e t 

end 

=jergens 

0$$ 0 4 18 

1$$ 2 t 

3$$ 1099 0 4 e t 

3$$ 1599 0 11 e t 

end 

=shell 

ln -sf ${RTNDIR}/groupStructure ft47f001 

end 

=vel Do the 238 Group Neutron Case 

0$$ 2 1 4 1$$ 14 200 47 t 

t 

10$$ 900 9032 0 11$$ 900 9032 3 99 1099 3 t 

10$$ 900 9033 0 11$$ 900 9033 3 99 1099 3 t 

10$$ 900 9034 0 11$$ 900 9034 3 99 1099 3 t 

10$$ 900 9035 0 11$$ 900 9035 3 99 1099 3 t 

10$$ 900 9036 0 11$$ 900 9036 3 99 1099 3 t 

10$$ 900 9037 0 11$$ 900 9037 3 99 1099 3 t 

10$$ 900 9029 0 11$$ 900 9029 3 99 1099 3 t 

10$$ 900 9031 0 11$$ 900 9031 3 99 1099 3 t 

10$$ 900 9734 0 11$$ 900 9034 3 99 1099 3 t 

10$$ 900 9735 0 11$$ 900 9035 3 99 1099 3 t 

10$$ 900 9736 0 11$$ 900 9036 3 99 1099 3 t 

10$$ 900 9737 0 11$$ 900 9037 3 99 1099 3 t 

10$$ 900 9729 0 11$$ 900 9029 3 99 1099 3 t 

10$$ 900 9731 0 11$$ 900 9031 3 99 1099 3 t 

 900Dose Factors from ICRU Report 44 and 57 

end 

=vel Do the 238 Group Neutron Case 

0$$ 9 1 4 1$$ 16 200 47 2 t 

t 

10$$ 900 9502 1 11$$ 900 9502 3 99 1599 3 t 

10$$ 900 9503 1 11$$ 900 9503 3 99 1599 3 t 

10$$ 900 9504 1 11$$ 900 9504 3 99 1599 3 t 

10$$ 900 9505 1 11$$ 900 9505 3 99 1599 3 t 

10$$ 900 9506 1 11$$ 900 9506 3 99 1599 3 t 

10$$ 900 9507 1 11$$ 900 9507 3 99 1599 3 t 

10$$ 900 9508 1 11$$ 900 9508 3 99 1599 3 t 

10$$ 900 9509 1 11$$ 900 9509 3 99 1599 3 t 

10$$ 900 9510 1 11$$ 900 9510 3 99 1599 3 t 

10$$ 900 9511 1 11$$ 900 9511 3 99 1599 3 t 

10$$ 900 9734 1 11$$ 900 9505 3 99 1599 3 t 
30 

10$$ 900 9735 1 11$$ 900 9506 3 99 1599 3 t 

10$$ 900 9736 1 11$$ 900 9510 3 99 1599 3 t 

10$$ 900 9737 1 11$$ 900 9511 3 99 1599 3 t 

10$$ 900 9729 1 11$$ 900 9504 3 99 1599 3 t 

10$$ 900 9731 1 11$$ 900 9505 3 99 1599 3 t 

 900Dose Factors from ICRU Report 44 and 57 

end 

=simonize 

Identifier=900 master=3 source=endf 

id45=900 id46=0 id47=0 id48=0 

1dn= 2 id19=900 

1DG= 9 ID19=900 

title= Dose factors from ansl/ans 6.1.1 - 1977 

end 

=rade 

0$$ 3 e t 

end 

=paleale 

0$$ 3 0 2$$ 1 1 e t 

end 

=shell 

cp ft03f001 ${RTNDIR}/dose_900_master 

end 

Input is as follows for the MAT=999 input. 

=jergens 

-1$$ 2 2 100 999 3 e 

0$$ 0 1 18 1$$ 4 t 

3$$ 1 0 7 e t 

3$$ 27 0 7 e t 

3$$ 101 0 7 e t 

3$$ 102 0 7 e t 

end 

=jergens 

0$$ 0 4 18 

1$$ 1 t 

3$$ 1099 0 4 e t 

end 

=shell 

ln -sf ${RTNDIR}/groupStructure ft47f001 

end 

=prilosec 

tab1= 1 

logwt= 4 

matwt= 99 mtwt= 1099 IGM= 200 

master= 2 

end 

=simonize 

Identifier=999 master=3 source=endf 

id45=999 id46=0 id47=0 id48=0 iftg=0 

1dn= 2 id19=999 

title= 1/v function (normalized to 1.0 at 2200m/s<==>0.0253eV) 

end 

=rade 

0$$ 3 e t 

end 

=shell 

cp ft03f001 ${RTNDIR}/dose_999_master 

end 
31 

Some gamma dose factors did not extend to higher energies and the JERGENS module needed to be 

changed to extend the dose factors. In these cases an extrapolation to higher energies was used as 

follows. 

9503: Claiborne-Trubey conversion factors (rad/h)/(photons/cm2

/s) 

We used an interpolation with five coefficients and interpolated from 6 × 106

 eV to 1.48 × 107 eV. 

The interpolation was used for energies above 15 MeV. The following formula was used. 

y = C0 ∗exp C1+C2X +C3X2 +C4X3 +C5X4

( )

X = ln( ) E ,E in eV

C0 = 6.65922 ×10−11

C1 = 2.91298

C2 = 0.622646

C3 = −0.00467212

C4 = −0.00105295

C5 = 6.73012 ×10−5

The interpolation is shown in Fig. 2.7. 

Fig. 2.7 Interpolation for Claiborne-Trubey gamma dose function. 
32 

9506 and 9507: International Commission on Radiation Units and Measurements, Report 57 

(ICRU-57), Table A.21 (air) Kerma 

We used an interpolation with five coefficients and interpolated from 4 × 106

 eV to 1.0 × 107 eV. 

The interpolation was used for energies above 10 MeV. The formula is as follows. 

y = C0 ∗exp C1+C2X +C3X2 +C4X3 +C5X4

( )

X = ln( ) E ,E in eV

C0 = 4.422278 ×10−6

C1 = −0.898794

C2 = 1.4646

C3 = −0.0209916

C4 = −0.00152315

C5 = 6.80427 ×10−5

(The factor C0 takes into account that the JERGENS module scales the internal results to obtain the 

results in (Gy/h)/(photons/cm2

/s) or (rad/h)/(photons/cm2/s). The interpolation is shown in Fig. 2.8. 

Fig. 2.8 Interpolation for ICRU-57, Table A.21 (air) Kerma gamma dose function. 
33 

9508 and 9509: Ambient dose equivalent (ICRU-57, Table A.21) 

We used an interpolation with five coefficients and interpolated from 4 × 106

 eV to 1.0 × 107 eV. 

The interpolation was used for energies above 10 MeV. The formula is as follows. 

y = C0 ∗exp C1+C2X +C3X2 +C4X3 +C5X4

( )

X = ln( ) E ,E in eV

C0 = 8.043111×10−10

C1 = −11.8556

C2 = 4.52715

C3 = −0.0788323

C4 = −0.0110208

C5 = 0.000440013

(The factor C0 takes into account that the JERGENS module scales the internal result to obtain the results 

in (Sv/h)/(photons/cm2

/s) or (rem/h)/(photons/cm2/s). The interpolation is shown in Fig. 2.9. 

Fig. 2.9 Interpolation for ambient dose equivalent (ICRU-57, Table A.21). 
34 

9510 and 9511 Effective dose (ICRU-57, Table A.17) (Sv/h)/(photons/cm2

/s) 

We used an interpolation with five coefficients and interpolated from 4 × 106

 eV to 1.0 × 107 eV. 

The interpolation was used for energies above 10 MeV. The formula is as follows. 

y = C0 ∗exp C1+C2X +C3X2 +C4X3 +C5X4

( )

X = ln( ) E ,E in eV

C0 = 2.154528e - 5

C1 = 4.74445

C2 = 1.12057

C3 = −0.0487202

C4 = −0.00119816

C5 = 0.000129643

(The factor C0 takes into account that the JERGENS module scales the internal result to obtain the results 

in (Sv/h)/(photons/cm2

/s) or (rem/h)/(photons/cm2/s). The interpolation is depicted in Fig. 2.10. 

Fig. 2.10 Interpolation for effective dose (ICRU-57, Table A.17) . 
35 

Using AMPX, coupled neutron-gamma libraries were generated for ENDF/B-VI.8 and ENDF/B-VII.0. 

Complete listings of the ENDF/B-VI.8 and ENDF/B-VII.0 libraries are provided in Appendixes B and C, 

respectively. 
36 
 37 

3 BENCHMARK TESTING 

As noted in the Introduction, new shielding capabilities have been developed for SCALE along with the 

ENDF/B-VI.8 and ENDF/B-VII.0 coupled library development. To this end, benchmark testing was 

performed to test both the new shielding transport capabilities and the new data libraries. As a result, the 

initial series of benchmark calculations have been used to compare the previous SCALE 5.1 Shielding 

Analysis Sequence Number 4 (SAS4) shielding sequence with the new MAVRIC sequence in SCALE 6. 

The SAS4-MAVRIC comparisons are used to verify the new transport capability relative to the previous 

SAS4 capability. These comparison calculations were performed using the ENDF/B-VI.8 coupled 

library. Because of this, the SAS4-MAVRIC calculations serve to verify the new shielding capability 

with the ENDF/B-VI.8 library. After the SAS4-MAVRIC comparisons, the ENDF/B-VI.8 and 

ENDF/B-VII.0 libraries are used with MAVRIC for benchmark calculations to validate the performance 

of MAVRIC and the ENDF/B-VI.8 and ENDF/B-VII.0 libraries. 

3.1 SAS4 and MAVRIC Benchmark Comparison Calculations with ENDF/B-VI.8 

Library

The new ENDF/B-VI.8 200 neutron group and 47 gamma group shielding library was tested by 

benchmarking the computational results against the results obtained with the previous SCALE shielding 

library, which was ENDF/B-IV with 27 neutron and 18 gamma groups. The computational results with 

the new library were also benchmarked against experimental measurements. The computations were 

performed using both the SAS4 and MAVRIC sequences of SCALE. SAS4 uses MORSE to perform 

Monte Carlo radiation transport along with XSDRN to create 1-D biasing parameters. MAVRIC uses 

Monaco to perform the Monte Carlo radiation transport and Denovo to create 3-D biasing parameters. 

Three sets of experimental measurements were selected for comparison purposes. All three consist of 

dose measurements on the outer surface of fully loaded spent nuclear fuel storage casks. Where feasible, 

SAS4 in the SCALE 5.1 release was used to model these storage casks with the new ENDF/B-VI.8 and 

the old ENDF/B-IV shielding libraries. MAVRIC, which is to be released as part of the SCALE 6 

package, was also used to model each of these casks with the ENDF/B-VI.8 and ENDF/B-IV shielding 

libraries. Details about the storage casks, measurements, and previous validation calculations with 

SCALE 4.3 can be found in EPRI report TR-104329 (Ref. 7). A brief summary of the current verification 

and validation calculations for each cask and the computational results are presented in the following 

sections. 
38 

3.1.1 Cask descriptions 

3.1.1.1 CASTOR-V/21 

The CASTOR (CAsk for Storage and Transport Of Radioactive material) cask was developed by 

Gesellschaft für Nuklear Service. This cask consists of a cast-iron body, and is 4.9 m tall and 2.4 m in 

diameter. Two concentric rows of polyethylene rods are incorporated in the cask wall to provide neutron 

shielding. The external surface consists of heat transfer fins oriented circumferentially around the cask 

surface. The fuel basket is configured to hold 21 pressurized water reactor (PWR) spent fuel assemblies 

and is constructed of stainless steel and borated stainless steel. The vertical and horizontal 2-D slices of 

the CASTOR geometry, as represented in Monaco using the SCALE Generalized Geometry Package 

(SGGP) are shown in Fig. 3.1. The green and pink rectangles and squares in Fig. 3.1 represent the 

homogenized fuel assemblies. The red regions in Fig. 3.1 above and below the fuel assemblies represents 

the end fittings where the 60Co activation sources are located. 

Fig. 3.1 Views of cask for storage and transport of radioactive material (CASTOR). (Dimensions in 

centimeters.) 

Vertical Slice of CASTOR Horizontal Slice of CASTOR 

10.725 

R = 110 

R = 76.35

182.9

219.6248.6

212.8

203.9
39 

3.1.1.2 Westinghouse MC-10 

The Westinghouse MC-10 PWR spent fuel cask consists of a forged steel body. The cask is 4.8 m tall 

and 2.7 m in diameter. Neutron shielding is provided by a layer of BISCO NS-3 on the outside of the 

cask. The fuel basket is configured to hold 24 PWR spent fuel assemblies and is constructed of 

aluminum. Each of the 24 basket locations contains a removable stainless steel enclosure and neutron 

poison material for criticality control. Figure 3.2 shows 2-D slices of the MC-10 geometry as represented 

in Monaco using SGGP. The green and magenta rectangles and squares in Fig. 3.2 represent the 

homogenized fuel assemblies. The red regions above and below the fuel assemblies in Fig. 3.2 represents 

the end fittings where the 60Co activation sources are located. 

Vertical Slice of MC-10 Horizontal Slice of MC-10 

Fig. 3.2 Views of MC-10 cask. (Dimensions in centimeters.) 
40 

3.1.1.3 TN-24P 

The TN-24P cask was developed by Transnuclear, Inc. The TN-24P cask also has a forged steel body, 

but is surrounded by a thin layer of resin enclosed in a steel outer shell. The resin is for neutron shielding. 

The TN-24P cask is 5 m tall and 2.3 m in diameter. The fuel basket will also hold 24 PWR spent fuel 

assemblies and is constructed of aluminum and boron plates. The vertical and horizontal 2-D slices of the 

TN-24P geometry, as represented in Monaco using the standard KENO-VI geometry package, are shown 

in Fig. 3.3. The dark blue and aquamarine rectangles and squares in Fig. 3.3 represent the homogenized 

fuel assemblies. The light blue above and below the fuel assemblies in Fig. 3.3 represents the end fittings 

where the 60Co activation sources are located. 

Vertical Slice of TN-24P Horizontal Slice of TN-24P 

Fig. 3.3 Views of TN-24P cask. (Dimensions in centimeters.) 

3.1.2 Calculations 

Each cask was loaded with spent Westinghouse 15 × 15 fuel assemblies. The initial enrichment, burnup, 

and cooling times for all the fuel assemblies are described in Ref. 7. For the calculations with the 

ENDF/B-IV shielding library, the neutron and gamma sources from the spent fuel in the previous 

SCALE 4.3 SAS4 calculations were used. For the ENDF/B-VI.8 library, the data from Ref. 7, which 

described the life of the spent fuel assemblies, were entered into OrigenArp, and the neutron and photon 

sources from the spent fuel were calculated using the ENDF/B-VI.8 shielding library group structure. 

The version of OrigenArp that was used in this process is the version that was released with SCALE 5.1. 

The previous SCALE 4.3 SAS4 calculations most likely used the version of ORIGEN-S that was released 

with SCALE 4.3 to create the spent fuel sources, so this could be one source of differences in comparison 

of old and new calculated results. Similarly, the 60Co sources previously used in the SCALE 4.3 SAS4 

calculations were used again with the ENDF/B-IV library. For the ENDF/B-VI.8 shielding library, the 

27 neutron-18 gamma group structure was converted to the new 200 neutron-47 gamma group structure. 

In this conversion, energy was conserved rather than number of particles. 
41 

Dose measurements were made at several different points on the top, bottom, and radial (near the fuel 

midplane) surfaces of each cask. SAS4 calculates dose averaged over a surface that represents the 

average of the measurements made at the different points on the surfaces of each cask. However, in the 

current version, Monaco does not yet have the capability of tallying on a surface. Therefore, very thin 

(1 mm thick) volume tallies were used instead of surface tallies. For the side or radial doses, the 

averaging surface is an outside cask surface (either the fin tips or the cask body) with an extent 

corresponding to the middle half of the fuel height. This is similar to the measurements made in Ref. 7. 

The top and bottom doses are on the top and bottom surfaces respectively, with radii also similar to those 

used in Ref. 7. These similarities allow for a more direct comparison between the calculations and 

experimental measurements. 

SAS4 requires that the cask model be symmetric around the fuel midplane, so the geometries for the top 

and bottom doses vary slightly above and below the active fuel region. Since this variation is very small, 

either the top or bottom geometry could be used to calculate the radial doses. Therefore, the Monaco 

geometries were also modeled symmetric about the fuel midplane, instead of using the actual geometries, 

to be consistent with the SAS4 simulations. The source types referred to in Table 3.1, Table 3.3, and 

Table 3.5 are all spent fuel assemblies, except for Co source type, which is a 60Co activation source. The 

same nomenclature as in Ref. 7 is used in this report to designate source or spent fuel assembly types. 

The full details about each spent fuel assembly and 60Co source are given in Ref. 7, Section 3. 

3.1.2.1 Computational results 

The input summary for each cask and the computed results using MAVRIC and SAS4, where available, 

with the ENDF/B-VI.8 and ENDF/B-IV shielding cross sections are presented in the following tables. 

The computational results were benchmarked against measurements as well, and the results are presented 

at the end of this section. 

Table 3.1 summarizes the geometry, source type, tally location, and particle type tallied for all the 

different TN-24P cask calculations. 

Table 3.1 TN-24P calculations input summary 

Case 

number Filename Geometry Source type 

Tally 

location 

Particle 

tallied 

1 out22.t24 Bottom W Bottom Gamma 

2 out24.t24 Bottom V Bottom Gamma 

3 out12.t24 Bottom W Radial Gamma 

4 out14.t24 Bottom V Radial Gamma 

5 out5.t24 Bottom Co Bottom Gamma 

6 out7.t24 Top Co Top Gamma 

7 out8.t24 Top Co Top Gamma 

8 out16.t24 Top W Top Gamma 

9 out18.t24 Top V Top Gamma 

10 out21.t24 Bottom W Bottom Neutron 

11 out23.t24 Bottom V Bottom Neutron 

12 out11.t24 Bottom W Radial Neutron 

13 out13.t24 Bottom V Radial Neutron 

14 out15.t24 Top W Top Neutron 

15 out17.t24 Top V Top Neutron 
42 

The computational results for TN-24P casks, using SAS4 in SCALE 5.1 and MAVRIC, with the 

ENDF/B-VI.8 and ENDF/B-IV libraries are presented in Table 3.2. 

Table 3.2 The computational results for TN24P casks 

SAS4 SCALE 5.1 

(ENDF/B-IV) 

MAVRIC 

(ENDF/B-IV) 

SAS4 SCALE 5.1 

(ENDF/B-VI.8) 

MAVRIC 

(ENDF/B-VI.8) 

Case 

no. 

Dose rate 

(rem/h) 1σ

Dose rate 

(rem/h) 1σ

Dose rate 

(rem/h) 1σ

Dose rate 

(rem/h) 1σ

1 9.200E-03 0.017 9.800E-03 0.034 4.600E-03 0.017 4.900E-03 0.015 

2 3.900E-03 0.016 4.000E-03 0.034 1.900E-03 0.017 2.000E-03 0.025 

3 2.880E-02 0.017 2.840E-02 0.029 1.560E-02 0.020 1.420E-02 0.035 

4 6.400E-03 0.025 5.700E-03 0.027 3.000E-03 0.023 2.900E-03 0.031 

5 9.880E-02 0.007 9.690E-02 0.022 7.230E-02 0.008 7.180E-02 0.023 

6 3.140E-02 0.006 3.120E-02 0.019 2.240E-02 0.007 2.200E-02 0.024 

7 5.100E-03 0.009 5.100E-03 0.020 3.500E-03 0.010 3.500E-03 0.042 

8 1.400E-03 0.020 1.300E-03 0.028 7.000E-04 0.015 7.000E-04 0.017 

9 6.000E-04 0.039 5.000E-04 0.022 3.000E-04 0.018 3.000E-04 0.026 

10 4.520E-02 0.012 4.920E-02 0.005 5.410E-02 0.011 5.800E-02 0.007 

11 2.320E-02 0.012 2.490E-02 0.004 2.720E-02 0.012 2.950E-02 0.005 

12 1.100E-03 0.008 1.100E-03 0.007 1.400E-03 0.009 1.300E-03 0.007 

13 4.000E-04 0.008 4.000E-04 0.008 4.000E-04 0.008 4.000E-04 0.006 

14 2.020E-02 0.012 1.850E-02 0.008 2.230E-02 0.015 2.040E-02 0.017 

15 9.900E-03 0.013 8.900E-03 0.009 1.040E-02 0.015 9.500E-03 0.008 

The ratio of the computed values using the new (ENDF/B-VI.8, 200 neutron and 47 gamma groups) and 

old (ENDF/B-IV, 27 neutron and 18 gamma groups) shielding libraries for all the TN-24P cask 

calculations are shown in Fig. 3.4. The corresponding geometries and source types for the case numbers 

were described in Table 3.1. In these figures red and blue points represent the computed ratios using 

SAS4 in SCALE 5.1 and MAVRIC, respectively. The error bars represent two sigma values. 

As can be seen from the figures, the computed ratios using SAS4 in SCALE 5.1 and MAVRIC are in 

good agreement, within statistical fluctuations, for all tallies. Therefore, both codes show similar 

behavior when the same libraries are used. Furthermore, the ratios are above 1 (from a few percent up to 

30%) for all the neutron tallies and below 1 (20% to 50%) for all the gamma tallies for both codes in a 

very consistent way. However, there are substantial differences in the response values computed with the 

two different versions of ENDF data. 
43 

Fig. 3.4 The ratio of the computed TN-24P cask tallies using the ENDF/B-VI.8 and 

ENDF/B-IV shielding cross sections. The red and blue points represent the computed ratios using 

SAS4 in SCALE 5.1 and MAVRIC, respectively. The error bars represent two standard deviations. 
44 

For code comparison, the ratios of the computed values using MAVRIC and SAS4 in SCALE 5.1 with 

both the ENDF/B-VI.8 and ENDF/B-IV shielding libraries, for the results in Table 3.2, are shown in 

Fig. 3.5. As can be seen from the figures, the two codes are in agreement within ±20%. 

 

Fig. 3.5 The ratios of the computed TN-24P cask tallies using MAVRIC and SAS4. In this 

figure, the red and blue points represent the computed ratios using the ENDF/B-VI.8 (new) and 

ENDF/B-IV (old) cross sections, respectively. The error bars represent two standard deviations. 
45 

The geometry, source type, tally location, and particle type tallied for all the different MC-10 cask 

calculations are summarized in Table 3.3. 

Table 3.3 Westinghouse MC-10 calculations input summary 

Case 

number Filename Geometry Source type Tally 

location 

Particle 

tallied 

1 mcbtmg1 Bottom W Bottom Gamma 

2 mcbtmg2 Bottom R Bottom Gamma 

3 mcbtmg3 Bottom N Bottom Gamma 

4 mcradg1 Top W Radial Gamma 

5 mcradg2 Top R Radial Gamma 

6 mcradg3 Top N Radial Gamma 

7 mctopg1 Top W Top Gamma 

8 mctopg2 Top R Top Gamma 

9 mctopg3 Top N Top Gamma 

10 mcbtmc1 Bottom Co Bottom Gamma 

11 mcbtmc2 Bottom Co Bottom Gamma 

12 mcbtmc3 Bottom Co Bottom Gamma 

13 mctopc1 Top Co Top Gamma 

14 mctopc2 Top Co Top Gamma 

15 mcbtmn1 Bottom R Bottom Neutron 

16 mcbtmn2 Bottom N Bottom Neutron 

17 mcradn1 Top R Radial Neutron 

18 mcradn2 Top N Radial Neutron 

19 mctopn1 Top R Top Neutron 

20 mctopn2 Top N Top Neutron 
46 

The computational results for MC-10 casks using MAVRIC with the ENDF/B-IV and ENDF/B-VI.8 

shielding libraries and their ratios are presented in Table 3.4. 

Table 3.4 The computational results for MC-10 casks 

MAVRIC 

(ENDF/B-IV) 

MAVRIC 

(ENDF/B-VI.8) 

Case 

no. 

Dose rate 

(rem/h) 1σ

Dose rate 

(rem/h) 1σ Ratio 

1 2.278E-04 0.034 1.971E-04 0.04 0.865 

2 7.194E-04 0.031 4.689E-04 0.04 0.652 

3 2.076E-03 0.029 1.188E-03 0.02 0.572 

4 1.129E-02 0.063 1.008E-02 0.07 0.894 

5 2.038E-02 0.044 1.522E-02 0.07 0.747 

6 2.638E-03 0.059 1.537E-03 0.08 0.582 

7 8.393E-05 0.034 7.022E-05 0.03 0.837 

8 1.964E-04 0.024 1.281E-04 0.04 0.652 

9 3.448E-04 0.055 2.229E-04 0.03 0.647 

10 4.468E-02 0.013 3.249E-02 0.01 0.727 

11 7.308E-03 0.023 4.830E-03 0.03 0.661 

12 4.001E-02 0.009 2.854E-02 0.01 0.713 

13 3.675E-03 0.018 2.451E-03 0.02 0.667 

14 1.584E-02 0.021 1.177E-02 0.02 0.743 

15 1.732E-03 0.040 1.485E-03 0.01 0.857 

16 1.716E-03 0.034 1.558E-03 0.01 0.908 

17 1.507E-02 0.023 1.608E-02 0.01 1.067 

18 6.390E-03 0.019 6.308E-03 0.01 0.987 

19 1.766E-02 0.007 1.535E-02 0.01 0.869 

20 1.257E-02 0.008 1.084E-02 0.01 0.862 
47 

The ratio of the computed values using MAVRIC with the ENDF/B-VI.8 and ENDF/B-IV shielding 

libraries for all the MC-10 cask tallies are shown in Fig. 3.6. The corresponding geometries and source 

types for the case numbers were described in Table 3.3. In these figures, the error bars represent two 

sigma values. 

As can be seen from the figures, the ratios are below 1 (20% to 60%) for all the gamma tallies, as they 

were for the TN-24P casks. 

Fig. 3.6 The ratio of the computed MC-10 cask values using the ENDF/B-VI.8 and ENDF/B-IV 

shielding cross-section libraries with MAVRIC.
48 

The geometry, source type, tally location, and particle type tallied for all the different CASTOR cask 

calculations are summarized in Table 3.5. 

Table 3.5 CASTOR-V/21 calculations input summary

Case 

number Filename Geometry Source 

type 

Tally 

location 

Particle 

tallied 

1 c21bg1 Bottom V-15 Bottom Gamma 

2 c21bg2 Bottom T Bottom Gamma 

3 c21rg1 Top V-15 Radial Gamma 

4 c21rg2 Top T Radial Gamma 

5 c21tg1 Top V-15 Top Gamma 

6 c21tg2 Top T Top Gamma 

7 c21bc1 Bottom Co Bottom Gamma 

8 c21tc1 Top Co Top Gamma 

9 c21tc2 Top Co Top Gamma 

10 c21bn2 Bottom T Bottom Neutron 

11 c21rn1 Top V-15 Radial Neutron 

12 c21rn2 Top T Radial Neutron 

13 c21tn1 Top V-15 Top Neutron 

14 c21tn2 Top T Top Neutron 
49 

The computational results for CASTOR casks using MAVRIC with the ENDF/B-VI.8 and ENDF/B-IV 

shielding libraries and their ratios are presented in Table 3.6. 

Table 3.6 The computational results for CASTOR casks 

MAVRIC 

(ENDF/B-IV) 

MAVRIC 

(ENDF/B-VI.8) 

Case 

no. 

Dose rate 

(rem/h) 1σ

Dose rate 

(rem/h) 1σ Ratio 

1 1.368E-03 0.038 8.309E-04 0.024 0.607 

2 4.309E-03 0.026 2.399E-03 0.016 0.557 

3 5.710E-02 0.041 3.528E-02 0.026 0.618 

4 1.350E-02 0.035 6.941E-03 0.020 0.514 

5 6.306E-04 0.050 3.931E-04 0.032 0.623 

6 1.839E-03 0.043 1.058E-03 0.026 0.575 

7 1.738E-02 0.024 1.164E-02 0.026 0.669 

8 2.951E-03 0.049 2.022E-03 0.057 0.685 

9 1.654E-02 0.039 1.230E-02 0.047 0.744 

10 2.964E-02 0.020 3.402E-02 0.027 1.148 

11 3.744E-03 0.012 6.763E-03 0.010 1.807 

12 4.020E-03 0.012 5.985E-03 0.011 1.489 

13 2.065E-02 0.045 3.009E-02 0.029 1.457 

14 2.396E-02 0.024 3.027E-02 0.030 1.263 

The ratio of the computed values using MAVRIC with the ENDF/B-VI.8 and ENDF/B-IV shielding 

libraries for all the CASTOR cask tallies are shown in Fig. 3.7. The corresponding geometries and source 

types for the case numbers were described in Table 3.5. In these figures, the error bars represent 

two sigma values. 

As can be seen from the figures, the ratios are above 1 (up to 80%) for all the neutron tallies and below 1 

(up to 50%) for all the gamma tallies. 
50 

Fig. 3.7 The ratio of the computed CASTOR cask values using the ENDF/B-VI.8 and 

ENDF/B-IV shielding cross-section libraries with MAVRIC.

3.1.2.2 Benchmark against experimental measurement 

The computational results using MAVRIC with both shielding libraries were benchmarked against the 

measured data points. The measured, computed (using both the ENDF/B-VI.8 and ENDF/B-IV libraries), 

and the ratio of computed to measured (C/M) data for neutron cases are presented in Table 3.7. The C/M 

ratios for the neutron cases for both the ENDF/B-VI.8 and ENDF/B-IV shielding libraries are plotted in 

Fig. 3.8. 

As shown in Table 3.7 and Fig. 3.8, overall the computed values using the ENDF/B-VI.8 shielding library 

are higher than the values computed using the ENDF/B-IV shielding library. There is better agreement 

with the measured values using the ENDF/B-VI.8 library for cases 8 and 9. Since the uncertainties in the 

measured values are not available, it is not clear whether the change in the ratios is within the statistical 

uncertainty. 
51 

Table 3.7 The measured and computed neutron dose rates and the ratio of 

computed to measured (C/M) dose rates for each cask model 

Case 

no. 

Cask 

type Geometry Meas. (mrem/h) 

ENDF/B-IV 

(mrem/h) 

ENDF/B-VI.8 

(mrem/h) 

C/M 

ENDF/B-IV 

C/M 

ENDF/B-VI.8 

1 TN24 Bottom 57.90 74.09 87.48 1.28 1.51 

2 Side 2.80 1.45 1.76 0.52 0.63 

3 Top 28.50 27.37 29.93 0.96 1.05 

4 MC10 Bottom 4.60 3.45 3.04 0.75 0.66 

5 Side 19.60 21.46 22.39 1.09 1.14 

6 Top 56.70 40.22 46.18 0.71 0.81 

7 CASTOR Bottom 51.30 56.65 68.70 1.10 1.34 

8 Side 11.40 7.76 12.75 0.68 1.12 

9 Top 51.50 44.61 60.35 0.87 1.17 

Fig. 3.8 The ratio of the computed to measured values using MAVRIC with the 

ENDF/B-VI.8 (new) and ENDF/B-IV (old) libraries for the neutron dose rates. 
52 

The measured, computed (using both the ENDF/B-VI.8 and ENDF/B-IV shielding libraries), and C/M 

ratio data for the gamma cases are presented in Table 3.8. The C/M ratios for these cases for both the 

ENDF/B-VI.8 and ENDF/B-IV shielding libraries are plotted in Fig. 3.9. As presented in the table and 

figure, for the gamma cases, the computed values using the ENDF/B-VI.8 shielding library are lower than 

the values computed using the ENDF/B-IV shielding library. This trend was investigated further and 

found to be consistent with a neglect of coherent scattering at relatively low photon energies in the older 

library. Indeed it appears that the older library used gamma data that preceded ENDF/B-IV. The newer 

library has a much stronger theoretical basis. There is better agreement with the measured values using 

the ENDF/B-VI.8 library for case 6, and for cases 2, 5, and 8 even though the discrepancies for these 

cases are still between 40 and 50%. However for cases 1, 3, 4, 7, and 9, the agreement with the measured 

data was better with the ENDF/B-IV shielding library. Since the uncertainties in the measured values are 

not available, it is not clear whether the change in the ratios is within the experimental uncertainty. 

Table 3.8 The measured and computed gamma dose rates and the ratio of 

the computed to measured (C/M) dose rates for each cask model 

Case 

no. 

Cask 

type Geometry Meas. (mrem/h) 

ENDF/B-IV 

(mrem/h) 

ENDF/B-VI.8 

(mrem/h) 

C/M 

ENDF/B-IV 

C/M 

ENDF/B-VI.8 

1 TN24 Bottom 117.0 110.68 88.73 0.95 0.76 

2 Side 12.3 34.11 17.08 2.77 1.39 

3 Top 37.9 38.17 26.44 1.01 0.70 

4 MC10 Bottom 62.0 55.03 67.70 0.89 1.09 

5 Side 21.4 34.30 26.85 1.60 1.25 

6 Top 14.6 20.14 14.64 1.38 1.00 

7 CASTOR Bottom 24.5 23.06 14.86 0.94 0.61 

8 Side 30.2 70.60 42.22 2.34 1.40 

9 Top 38.4 21.96 15.77 0.57 0.41 
53 

Fig. 3.9 The ratio of the computed to measured values using MAVRIC with the 

ENDF/B-VI.8 (new) and ENDF/B-IV (old) shielding libraries for the gamma dose rates.

3.2 MAVRIC Benchmark Calculations Using ENDF/B-VI.8 and ENDF/B-VII.0 Libraries

3.2.1 Neutron fields in three-section concrete labyrinth from 252Cf source 

3.2.1.1 Benchmark overview 

Experimental measurements of the neutron flux from a 252Cf source through an 18 m concrete labyrinth 

were made in Russia in 1982 to obtain benchmark data for code validation.8

 

Two sources were used: (1) a bare 252Cf neutron source and (2) a 252Cf neutron source encased in a 

polyethylene sphere. 

Six labyrinths were constructed. 

1. Empty, three-section labyrinth. 

2. First corner is lined with polyethylene plates covered with cadmium. 

3. First corner is lined with polyethylene plates without cadmium. 

4. Two corners are lined with borated concrete. 

5. Two polyethylene plates are installed in the second leg of the labyrinth (bare source only). 
54 

6. Dead end at the end of the first leg of the labyrinth. 

Seven Bonner sphere detectors [with 6

LiI(Eu) crystal] were used. 

1. 2 in. diameter, cadmium covered 

2. 3 in. diameter, cadmium covered 

3. 5 in. diameter, cadmium covered 

4. 5 in. diameter, not covered 

5. 8 in. diameter, cadmium covered 

6. 10 in. diameter, cadmium covered 

7. 12 in. diameter, cadmium covered. 

The geometry of the labyrinth is shown in Fig. 3.10. 

Fig. 3.10 Cutaway view of labyrinth model (ceiling removed). 

As part of the study in Ref. 8, benchmark calculations were performed with Monte Carlo N-particle 

(MCNP) code. Calculations were done in two steps: (1) the determination of the detector response in a 

free-field and (2) the transport of neutrons through the labyrinth, where computed fluxes were convolved 

with the response functions. 

3.2.1.2 Monaco/MAVRIC calculations 

Monaco and MAVRIC and the new ENDF/B-VI.8 200 neutron-47 gamma shielding library were used to 

compute the detector count rates measured by the various detectors for the first labyrinth design (empty, 

three section) using both the covered and uncovered californium (252Cf) source. Calculations followed the 

benchmark instructions and were done in two steps. 

1. Determination of the detector response function using Monaco—for each energy group, a Monaco 

calculation was done to find the integrated product of the 6

Li(n,t)4He cross section and neutron flux 
55 

inside the crystal of the detector (Fig. 3.11). Monaco calculations were done for each energy group 

(200) incident on each detector (7). 

2. MAVRIC, using the FW-CADIS method, was used to calculate neutron fluxes along the path through 

the labyrinth. FW-CADIS performs a quick forward discrete ordinates calculation and a quick adjoint 

discrete ordinates calculation to create an importance map and biased source distribution that, when 

used with Monaco, will give results for multiple tallies with nearly uniform statistical uncertainty. 

Each FW-CADIS calculation is optimized for a specific source and response, so calculations had to 

be done for each source (2) using each detector size (7). 

1.00E‐02

1.00E‐01

1.00E+00

1.00E+01

1.00E+02

1.00E+03

1.00E+04

1.00E‐10 1.00E‐08 1.00E‐06 1.00E‐04 1.00E‐02 1.00E+00 1.00E+02

6Li(n,t)4He cross section (barns)

Energy (MeV)

Fig. 3.11 6Li(n,t)4He cross section from the ENDF/B VI.8 200n/47g shielding library. 

3.2.1.3 Detector response functions 

Seven Bonner spheres were used to measure the neutron flux through the labyrinths. Sizes were 2-, 3-, 5-, 

8-, 10- and 12-inch diameters. Each was made of polyethylene (ρ=0.92±0.03 g/cm3

) with an 0.08-cmthick cadmium layer (ρ= 8.648 g/cm3

). One other detector, 5 inches in diameter, had no cadmium cover. 

The crystal was a 6

LiI(Eu) cylinder, 1 cm in diameter and 1 cm high (ρ=4.08 g/cm3), with a lithium 

enrichment of 90 at. %. 

The benchmark model used by the experimental team for making detector response functions left out the 

europium content (since it was less than 0.3 wt %.). Source particles were started on the surface of a 

sphere with radius=160 cm, pointed inward striking the outer sphere of the detector model. The model of 

the detector is shown in Figure 20 of Ref. 8. 
56 

The Monaco model of the detector was a spherically symmetric model and included the following layers 

(outward, from center): 

• 0.57235712 cm of LiI (to maintain volume of 0.7853982 cm3), 

• 0.05 cm of TiO2, 

• 0.03 cm of aluminum D16, 

• 0.07 cm of void, 

• 0.10 cm of aluminum D16, 

• 0.05 cm of void, and 

• polyethylene of the detector and its optional 0.08 cm cadmium cover. 

For a detector response function, what is needed is the total number of 6

Li(n,t)4He reactions occurring in 

the crystal (radius Rc) per unit flux for monoenergetic source neutrons. The count rate C(E) for incident 

neutrons of energy E is 

 

C(E) = N6LiV σ n,t ∫ (E')φ(E')dE' , (3.1) 

for a source strength S scaled such that the flux in an empty sphere the same size as the Bonner sphere 

(radius RD) is unity. (This gives S = πRD

2 .) 

C(E) = 1.66072 ×10−2

b cm

⎛

⎝ ⎜ 

⎞

⎠ ⎟ 4π

3

RC

3 ⎛ 

⎝ ⎜ ⎞ ⎠ ⎟ σ n,t ∫ (E')φ(E')dE' . (3.2) 

The volume of the crystal is 4π

3

RC

3 = 0.7853982 cm

3

. 

For each detector (7), a series of 200 Monaco calculations were done (about 18 minutes each), and the 

responses for each incident neutron energy were determined as described above. As can be seen in 

Figs. 3.12–3.18, the computed detector response functions were similar to the response functions 

computed by MCNP in the benchmark report. 
57 

0.0

0.2

0.4

0.6

0.8

1.0

1.2

1.4

1.E-08 1.E-06 1.E-04 1.E-02 1.E+00 1.E+02

Count rate per unit flux

Energy (MeV)

benchmark

Monaco

Fig. 3.12 Response (counts per unit flux) for the 2-inch diameter Bonner sphere 

(with cadmium cover). (Benchmark is computed function from Ref. 8.) 

0.0

0.2

0.4

0.6

0.8

1.0

1.2

1.4

1.E-08 1.E-06 1.E-04 1.E-02 1.E+00 1.E+02

Count rate per unit flux

Energy (MeV)

benchmark

Monaco

Fig. 3.13 Response (counts per unit flux) for the 3-inch diameter Bonner sphere 

(with cadmium cover). (Benchmark is computed function from Ref. 8.) 
58 

0.0

0.2

0.4

0.6

0.8

1.0

1.2

1.4

1.E-08 1.E-06 1.E-04 1.E-02 1.E+00 1.E+02

Count rate per unit flux

Energy (MeV)

benchmark

Monaco

Fig. 3.14 Response (counts per unit flux) for the 5-inch diameter Bonner sphere 

(with cadmium cover). (Benchmark is computed function from Ref. 8.) 

0.0

0.2

0.4

0.6

0.8

1.0

1.2

1.4

1.E-08 1.E-06 1.E-04 1.E-02 1.E+00 1.E+02

Count rate per unit flux

Energy (MeV)

benchmark

Monaco

Fig. 3.15 Response (counts per unit flux) for the 5-inch diameter Bonner sphere 

(no cover). (Benchmark is computed function from Ref. 8.) 
59 

0.0

0.2

0.4

0.6

0.8

1.0

1.2

1.4

1.E-08 1.E-06 1.E-04 1.E-02 1.E+00 1.E+02

Count rate per unit flux

Energy (MeV)

benchmark

Monaco

Fig. 3.16 Response (counts per unit flux) for the 8-inch diameter Bonner sphere 

(with cadmium cover). (Benchmark is computed function from Ref. 8.) 

0.0

0.2

0.4

0.6

0.8

1.0

1.2

1.4

1.E-08 1.E-06 1.E-04 1.E-02 1.E+00 1.E+02

Count rate per unit flux

Energy (MeV)

benchmark

Monaco

Fig. 3.17 Response (counts per unit flux) for the 10-inch diameter Bonner sphere 

(with cadmium cover). (Benchmark is computed function from Ref. 8.) 
60 

0.0

0.2

0.4

0.6

0.8

1.0

1.2

1.4

1.E-08 1.E-06 1.E-04 1.E-02 1.E+00 1.E+02

Count rate per unit flux

Energy (MeV)

benchmark

Monaco

Fig. 3.18 Response (counts per unit flux) for the 12-inch diameter Bonner sphere 

(with cadmium cover). (Benchmark is computed function from Ref. 8.) 

The Monaco-computed detector response functions compared well to the response functions computed by 

the benchmark team using MCNP. The results for the larger Bonner spheres show statistical noise in the 

Monaco results and could be reduced by longer run times. For these detectors, the Monaco responses are 

also a bit lower than the MCNP responses and could give lower total detector count rates. 

3.2.1.4 Neutron transport through the labyrinth 

The materials and labyrinth dimensions were taken from the benchmark report.8

 For each size detector 

(7) and for each type of source (2), MAVRIC and the FW-CADIS method were used to compute the 

detector count rates at all 10 detector locations for the first labyrinth design (empty, three-section 

labyrinth). Neutron flux spectra were computed at each location and then folded with the different 

detector response functions. 

In the benchmark report, the authors computed the detector count rates with MCNP. For the 12-inch 

detector, the stated relative uncertainties ranged from 0.5% for the closest detector location to 2% for the 

furthest detector location. 

The ratios of the calculated count rates to the experimentally measured count rates are shown for both the 

benchmark team using MCNP and MAVRIC in Figs. 3.19–3.32. The error bars in the C/E ratio are due 

almost entirely to the uncertainty in the experimental measurements. MAVRIC relative uncertainties 

were in the 1–2% range for the bare source and in the 2–3% range for the covered source. 
61 

0.00

0.25

0.50

0.75

1.00

1.25

1.50

1.75

2.00

0 1 2 3 4 5 6 7 8 9 10 11

C/E Ratio

Measurement Position

benchmark

MAVRIC

Fig. 3.19 Neutron count rate C/E for the 2-inch Bonner sphere (with cadmium) 

and the uncovered source. (Benchmark is computed function from Ref. 8.) 

0.00

0.25

0.50

0.75

1.00

1.25

1.50

1.75

2.00

0 1 2 3 4 5 6 7 8 9 10 11

C/E Ratio

Measurement Position

benchmark

MAVRIC

Fig. 3.20 Neutron count rate C/E for the 2-inch Bonner sphere (with cadmium) 

and the covered source. (Benchmark is computed function from Ref. 8.) 
62 

0.00

0.25

0.50

0.75

1.00

1.25

1.50

1.75

2.00

0 1 2 3 4 5 6 7 8 9 10 11

C/E Ratio

Measurement Position

benchmark

MAVRIC

Fig. 3.21 Neutron count rate C/E for the 3-inch Bonner sphere (with cadmium) 

and the uncovered source. (Benchmark is computed function from Ref. 8.) 

0.00

0.25

0.50

0.75

1.00

1.25

1.50

1.75

2.00

0 1 2 3 4 5 6 7 8 9 10 11

C/E Ratio

Measurement Position

benchmark

MAVRIC

Fig. 3.22 Neutron count rate C/E for the 3-inch Bonner sphere (with cadmium) and the 

covered source. (Benchmark is computed function from Ref. 8.)
63 

0.00

0.25

0.50

0.75

1.00

1.25

1.50

1.75

2.00

0 1 2 3 4 5 6 7 8 9 10 11

C/E Ratio

Measurement Position

benchmark

MAVRIC

Fig. 3.23 Neutron count rate C/E for the 5-inch Bonner sphere (with cadmium) and the 

uncovered source. (Benchmark is computed function from Ref. 8.) 

0.00

0.25

0.50

0.75

1.00

1.25

1.50

1.75

2.00

0 1 2 3 4 5 6 7 8 9 10 11

C/E Ratio

Measurement Position

benchmark

MAVRIC

Fig. 3.24 Neutron count rate C/E for the 5-inch Bonner sphere (with cadmium) and the 

covered source. (Benchmark is computed function from Ref. 8.) 
64 

0.00

0.25

0.50

0.75

1.00

1.25

1.50

1.75

2.00

0 1 2 3 4 5 6 7 8 9 10 11

C/E Ratio

Measurement Position

benchmark

MAVRIC

Fig. 3.25 Neutron count rate C/E for the 5-inch Bonner sphere (no cadmium) and the 

uncovered source. (Benchmark is computed function from Ref. 8.)

0.00

0.25

0.50

0.75

1.00

1.25

1.50

1.75

2.00

0 1 2 3 4 5 6 7 8 9 10 11

C/E Ratio

Measurement Position

benchmark

MAVRIC

Fig. 3.26 Neutron count rate C/E for the 5-inch Bonner sphere (no cadmium) and the 

covered source. (Benchmark is computed function from Ref. 8.) 
65 

0.00

0.25

0.50

0.75

1.00

1.25

1.50

1.75

2.00

0 1 2 3 4 5 6 7 8 9 10 11

C/E Ratio

Measurement Position

benchmark

MAVRIC

Fig. 3.27 Neutron count rate C/E for the 8-inch Bonner sphere (with cadmium) and the 

uncovered source. (Benchmark is computer function from Ref. 8.)

0.00

0.25

0.50

0.75

1.00

1.25

1.50

1.75

2.00

0 1 2 3 4 5 6 7 8 9 10 11

C/E Ratio

Measurement Position

benchmark

MAVRIC

Fig. 3.28 Neutron count rate C/E for the 8-inch Bonner sphere (with cadmium) and the 

covered source. (Benchmark is computer function from Ref. 8.) 
66 

0.00

0.25

0.50

0.75

1.00

1.25

1.50

1.75

2.00

0 1 2 3 4 5 6 7 8 9 10 11

C/E Ratio

Measurement Position

benchmark

MAVRIC

Fig. 3.29 Neutron count rate C/E for the 10-inch Bonner sphere (with cadmium) and the 

uncovered source. (Benchmark is computer function from Ref. 8.)

0.00

0.25

0.50

0.75

1.00

1.25

1.50

1.75

2.00

0 1 2 3 4 5 6 7 8 9 10 11

C/E Ratio

Measurement Position

benchmark

MAVRIC

Fig. 3.30 Neutron count rate C/E for the 10-inch Bonner sphere (with cadmium) and the 

covered source. (Benchmark is computer function from Ref. 8.)
67 

0.00

0.25

0.50

0.75

1.00

1.25

1.50

1.75

2.00

0 1 2 3 4 5 6 7 8 9 10 11

C/E Ratio

Measurement Position

benchmark

MAVRIC

Fig. 3.31 Neutron count rate C/E for the 12-inch Bonner sphere (with cadmium) and the 

uncovered source. (Benchmark is computer function from Ref. 8.)

0.00

0.25

0.50

0.75

1.00

1.25

1.50

1.75

2.00

0 1 2 3 4 5 6 7 8 9 10 11

C/E Ratio

Measurement Position

benchmark

MAVRIC

Fig. 3.32 Neutron count rate C/E for the 12-inch Bonner sphere (with cadmium) and 

the covered source. (Benchmark is computed function from Ref. 8.) 
68 

The MAVRIC count rates are all a bit lower than the benchmark MCNP count rates (due to lower detector 

response functions) but follow the MCNP C/E well for both the covered and uncovered sources for all 

seven detectors. 

3.2.2 Neutron and photon leakage spectra from 252Cf source at centers of six iron spheres of 

different diameters 

3.2.2.1 Benchmark overview 

Neutron and photon leakage spectra were measured from a 252Cf source through iron spheres of various 

diameters 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.9

 

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 (20 cm, 30 cm, 40 cm, 50 cm, 60 cm, and 70 cm). Note that no 

photon measurement was made for the 20 cm shield. Each data set was the difference of two 

measurements—one with and one without a cone shaped shield between the source and detector. 

The experimental geometry is shown in Fig. 3.33. 

Fig. 3.33 Iron sphere experiments.

steel capsule: 

R2,outer = 0.53 cm, 

R2,inner = 0.21 cm

source Cf-252 (point source) 

located at center of spheres 

R = 3 Router

spherical 

detector 

surface 

white regions are void 

Iron sphere: 

Router = 10, 15, 20, 25, 30, and 35 cm, 

Rinner = 0.93 cm

Cu shell: 

R1,outer = 0.84 cm, 

R1,inner = 0.61 cm 
69 

3.2.2.2 Monaco calculations 

Monaco (SCALE 6) and the new ENDF/B-VI.8 fine-group shielding library were used to compute 

leakage spectra of both neutrons and photons according to the benchmark instructions. 

For the Monaco neutron source, the experimentally measured neutron flux spectrum of the bare source 

was converted into a 200-group source specification. For the Monaco photon source, the experimentally 

measured photon flux spectrum of the bare source was converted into a 47-group source specification. 

Note that the “bare source” did include 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. 

For every neutron emitted from the 252Cf source, 3.82 photons were emitted. Monaco combined the 

sources into a single coupled neutron-photon source. 

Benchmark results are listed in the above document in terms of 4πR2

Φ(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%, so the experimental values are shown on 

the following plots as two lines, representing the values ± 1σ. 

Since the Monaco source was created from the bare source measurements, the Monaco simulation for 

these should match the experiment well, as shown in Figs. 3.34 and 3.35. 

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

Energy (M eV)

4πR 2

Φ (E)/Q (/M eV )

M onaco

measured + σ

measured - σ

Fig. 3.34 Comparison of the neutron calculation to the measurement of the bare source. 
70 

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

Energy (M eV)

4πR 2

Φ (E)/Q (/M eV )

M onaco

measured + σ

measured - σ

Fig. 3.35 Comparison of the photon calculation to the measurement of the bare source. 

Neutron flux results for various shield thicknesses are shown in Figs. 3.36–3.41. 

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

Energy (M eV)

4πR 2

Φ (E)/Q (/M eV )

M onaco

measured + σ

measured - σ

Fig. 3.36 Comparison of the neutron calculation to the measurement of the 20 cm shield. 
71 

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

Energy (M eV)

4πR 2

Φ (E)/Q (/M eV )

M onaco

measured + σ

measured - σ

Fig. 3.37 Comparison of the neutron calculation to the measurement of the 30 cm shield. 

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

Energy (M eV)

4πR 2

Φ (E)/Q (/M eV )

M onaco

measured + σ

measured - σ

Fig. 3.38 Comparison of the neutron calculation to the measurement of the 40 cm shield. 
72 

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

Energy (M eV)

4πR 2

Φ (E)/Q (/M eV )

M onaco

measured + σ

measured - σ

Fig. 3.39 Comparison of the neutron calculation to the measurement of the 50 cm shield.

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

Energy (M eV)

4πR 2

Φ (E)/Q (/M eV )

M onaco

measured + σ

measured - σ

Fig. 3.40 Comparison of the neutron calculation to the measurement of the 60 cm shield. 
73 

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

Energy (M eV)

4πR 2

Φ (E)/Q (/M eV )

M onaco

measured + σ

measured - σ

Fig. 3.41 Comparison of the neutron calculation to the measurement of the 70 cm shield.

Photon flux results for various shield thicknesses are given in Figs. 3.42–3.46. 

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

Energy (M eV)

4πR 2

Φ (E)/Q (/M eV )

M onaco

measured + σ

measured - σ

Fig. 3.42 Comparison of the photon calculation to the measurement of the 30 cm shield. 
74 

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

Energy (M eV)

4πR 2

Φ (E)/Q (/M eV )

M onaco

measured + σ

measured - σ

Fig. 3.43 Comparison of the photon calculation to the measurement of the 40 cm shield.

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

Energy (M eV)

4πR 2

Φ (E)/Q (/M eV )

M onaco

measured + σ

measured - σ

Fig. 3.44 Comparison of the photon calculation to the measurement of the 50 cm shield. 
75 

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

Energy (M eV)

4πR 2

Φ (E)/Q (/M eV )

M onaco

measured + σ

measured - σ

Fig. 3.45 Comparison of the photon calculation to the measurement of the 60 cm shield.

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

Energy (M eV)

4πR 2

Φ (E)/Q (/M eV )

M onaco

measured + σ

measured - σ

Fig. 3.46 Comparison of the photon calculation to the measurement of the 70 cm shield.
76 

Photon fluxes are underpredicted by Monaco and MCNP5 as is depicted in Fig. 3.47. 

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

Energy (M eV)

4πR 2

Φ (E)/Q (/M eV )

M onaco

measured + σ

measured - σ

MCNP CE

Fig. 3.47 Photon flux through the 70 cm shield as calculated by Monaco and MCNP5.

MCNP results in the benchmark study were similar to those shown in Fig. 3.47 (see Fig. 20 of Ref. 9). 

3.2.3 Comparison of measured and calculated neutron transmission through heavy water for a 

252Cf source placed in the center of a 30 cm diameter sphere

3.2.3.1 Experiment overview 

The transmission of 252Cf neutrons through a sphere filled with heavy water was measured in Prague in 

the mid 1990s (Ref. 10). 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 is given for the materials and geometry of the source holder, insertion tube, and 

detectors in Ref. 10. Two different assemblies holding the californium source were measured. 

3.2.3.2 Monaco calculations 

The Monaco SGGP geometry model 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 (75,0,0). The experiment 

sat 2 m above the floor of an experimental hall that measured 10 × 13 × 25 m. Here, the origin 

corresponds to the source at the center of the heavy water sphere. Neither of the two source holder 

assemblies was modeled in any detail. The geometry model is shown in Fig. 3.48. 
77 

Fig. 3.48 Monaco model of 252Cf source in heavy water experiment. 

A second Monaco input file was created for the geometry without the 4 cylindrical shields in place. 

For the source, the Watt spectrum with constants α =1.025 MeV, b =2.926/MeV was used. 

p(E) = ce−E /α sinh( bE ) . (3.3) 

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 4π(75)2

 = 70686 n/s. No biasing was used, other than implicit capture. 

A point detector tally at the detector location was used to calculate the energy dependent neutron flux. 

Results for the ENDF/B-VI.8 library are depicted in Figs 3.49 and 3.50. Results for ENDF/B-VII.0 are 

shown in Figs. 3.51 and 3.52. 

D2O 

Cf 

Detector 

Iron Borated

polyethylene (3%) 
78 

1.E-04

1.E-03

1.E-02

1.E-01

1.E+00

1.E-02 1.E-01 1.E+00 1.E+01

Flux per unit Energy

Energy (MeV)

Assembly A8

Assembly A25

Monaco

Fig. 3.49 Comparison of the Monaco (ENDF/B-VI.8) results and the measured data. 

0

0.5

1

1.5

2

2.5

0.01 0.1 1 10

C/E ratios

Energy (MeV)

assembly A8

assembly A25

Fig. 3.50 Ratios of the Monaco (ENDF/B-VI.8 fine-group library) results and the measured data. 
79 

1.E-04

1.E-03

1.E-02

1.E-01

1.E+00

1.E-02 1.E-01 1.E+00 1.E+01

Flux

Energy (MeV)

Assembly A8

Assembly A25

Monaco

Fig. 3.51 Comparison of the Monaco (ENDF/B-VII.0) results and the measured data. 

0

0.5

1

1.5

2

2.5

3

0.01 0.1 1 10

C/E ratios

Energy (MeV)

assembly A8

assembly A25

Fig. 3.52 Ratios of the Monaco (ENDF/B-VII.0 fine-group library) results and the measured data.
80 

3.2.4 Comparison of measured and calculated neutron transmission through steel for a 252Cf 

source 

3.2.4.1 Experiment overview 

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. Two sets of measurements were 

made—one by the Czechoslovakian National Research Institute and the other by the Skoda Company.11 

3.2.4.2 Monaco calculations 

The Monaco model consists of just a point source and a spherical shell of iron. Two different tallies are 

used which should all give the same result: a region tally (for a thin spherical shell at radius=100 cm); a 

point detector tally at (x,y,z)=(100,0,0). For the source, the Watt spectrum with constants 

α = 1.025 MeV, b = 2.926/MeV was used. 

p(E) = ce−E /α sinh( bE ) . (3.4) 

The measurements listed in the paper 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 strength of 4π(100)2

 = 125664 n/s. No biasing was used, other than implicit capture. The geometry 

is shown in Fig. 1 of Ref. 11. 

Results for ENDF/B-VI.8 are depicted in Figs. 3.53 and 3.54. Results for ENDF/B-VII.0 are shown in 

Figs. 3.55 and 3.56. 

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

Flux per unit energy

Energy (MeV)

NRI

Skoda

pt det

region

Fig. 3.53 Monaco output (ENDF/B-VI.8 fine-group library) for total flux at 10 cm 

using two different tallies compared to the two sets of measured data. 
81 

0

1

2

3

0.01 0.1 1 10

C/E Ratios

Energy (MeV)

C/E NRI

C/E Skoda

Fig. 3.54 Ratios of the Monaco (ENDF/B-VI.8 fine-group library) point detector tally 

to the measured fluxes. 

1.E-05

1.E-04

1.E-03

1.E-02

1.E-01

1.E+00

1.E-03 1.E-02 1.E-01 1.E+00 1.E+01

Flux per unit energy

Energy (MeV)

NRI

Skoda

pt det

region

Fig. 3.55 Monaco output (ENDF/B-VII.0 fine-group library) for total flux at 10 cm 

using two different tallies compared to the two sets of measured data. 
82 

0

1

2

3

0.01 0.1 1 10

C/E Ratios

Energy (MeV)

C/E NRI

C/E Skoda

Fig. 3.56 Ratios of the Monaco (ENDF/B-VII.0 fine-group library) point detector tally 

to the measured fluxes.

3.2.5 Systematic evaluation of neutron shielding effects for materials 

3.2.5.1 Experimental overview 

Transmission of neutrons from a 252Cf source through different thicknesses of five different materials was 

measured to determine their shielding effectiveness.12 Combinations of steel and polyethylene were also 

measured to determine the change in shielding effectiveness on the order of the materials. The detector 

used was an Aloka moderator-type survey meter, made of polyethylene, which is supposed to mimic the 

human dose response. The data presented by Ueki et al.12 was corrected for room scatter. An example of 

one experiment is shown in Fig. 3.57. 
83 

Fig. 3.57 Neutron shielding experiment example (half of paraffin removed to show source inside cone).

3.2.5.2 MCNP simulations 

Polyethylene proved problematic for MCNP simulations. Reported results show good correlation with 

the findings of Ueki et al. for all materials except polyethylene.13–1

15 

3.2.5.3 MAVRIC calculations 

For penetration through thick shields, the MAVRIC sequence and the CADIS methodology were used. 

For each experiment, an adjoint calculation using GTRUNC/Denovo was performed using the new 

ENDF/B-VII.0 broad-group library. An importance map and a biased source distribution were calculated 

by MAVRIC from the adjoint fluxes. The Monaco simulation used the new ENDF/B-VI.8 fine-group 

shielding library and the importance map, but not the biased source distribution, so that the 252Cf spectral 

shape would be better preserved with the fine-group library. Adjoint calculations typically took 

6–10 minutes and the Monaco calculations typically took 10–13 minutes each, and all had less than 4% 

statistical uncertainties. 

252Cf source 

Paraffin 

SS-304 

Shield 

SS-304 

Detector 
84 

3.2.5.4 Results 

MAVRIC calculations match the type 1 experiments (single shield) within ±20%, except for 

polyethylene. MAVRIC does not match the type 2 or type 3 experiments, probably due to the 

polyethylene component. Similar difficulties were observed with MCNP calculations. 

Type 1 Experiment: Single material. Experiment geometries are shown in Fig. 1 of Ref. 12. Results are 

shown in Fig. 3.58. 

0.80

0.90

1.00

1.10

1.20

1.30

1.40

1.50

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

Fig. 3.58 Ratio of MAVRIC calculation of neutron detector dose rate to 

experimental measurement.

Type 2 Experiment: Stainless steel and polyethylene. Experiment geometries are shown in Ref. 12. 

Results are shown in Figs. 3.59 and . 3.60. 
85 

0.30

0.40

0.50

0.60

0.70

0.80

0.90

1.00

0 5 10 15 20 25

MAVRIC/Ueki C/E

Thickness T in cm

Fig. 3.59 Type 2a results—25 cm of stainless 

steel first, then polyethylene.

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

Fig. 3.60 Type 2b results—polyethylene, 

then 25 cm of stainless steel.

Type 3 Experiment: Polyethylene sandwiched in stainless steel. Experiment geometries are shown in 

Ref. 12. Results are depicted in Fig. 3.61. 

0.60

0.70

0.80

0.90

1.00

1.10

1.20

1.30

0 5 10 15 20 25

MAVRIC/Ueki C/E

Thickness t in cm

Fig. 3.61 Type 3 results for polyethylene fixed at a 

thickness of 15 cm. 
86 
 87 

4 SUMMARY 

Coupled neutron-gamma libraries have been developed for SCALE based on ENDF/B-VI.8 and 

ENDF/B-VII.0 cross-section evaluations. The AMPX cross-section processing system was used for all 

library generation tasks for SCALE. ENDF/B-VI.8 and ENDF/B-VII.0 fine-group libraries were 

generated using 200 neutron and 47 gamma groups. The neutron group structure is identical to the 

199-group VITAMIN-B6 structure except an additional neutron group has been added to extend the top 

energy boundary to 20 MeV. In addition to the fine-group libraries, a broad-group ENDF/B-VII.0 library 

has been generated using a 27 neutron and 19 gamma group structure. The broad-group library is 

primarily used to perform adjoint discrete ordinates calculations needed to prepare importance maps and a 

biased source distribution for biasing the forward Monte Carlo calculations. The SCALE shielding 

libraries have the following features. 

• The multigroup neutron data were generated using a weighting function that has the form of a 

Maxwellian—1/E—fission spectrum—1/E above 10 MeV. This spectrum is consistent with the 

238-group ENDF/B-VI library that was distributed with SCALE 5.1. 

• Full-range Bondarenko factors are provided for all nuclides, thereby enabling faster neutron 

resonance self-shielding calculations using the narrow resonance approximation. 

• Response functions consistent with previous SCALE libraries are maintained in the new shielding 

libraries. 

As part of rigorous differential data testing activities, the AMPX code system was used to process all 

ENDF/B evaluations on three computing platforms: DEC Alpha Tru64, 32-bit Linux, and 64-bit Intel 

Mac. Because of the differences between compilers, experience has shown that different results can be 

obtained with different computing platforms. The use of the three platforms provided confirmatory 

checks of the cross-section processing calculations. In addition, independent verifications were 

performed between the AMPX results and results obtained with the NJOY code system for 1-D and 2-D 

processes. In general, the AMPX-NJOY comparisons show good agreement for the 1-D and 2-D 

processes. Although some differences have been observed between the independent processing systems, 

the overall agreement is good. The independent processing code comparisons provide additional 

verification checks for the differential cross-section data. 

To complete the library testing effort, radiation transport calculations were performed with the SCALE 

shielding codes Monaco and MAVRIC using the coupled libraries to simulate several shielding 

benchmark calculations. Monaco/MAVRIC and the new ENDF/B-VI.8 and ENDF/B-VII.0 fine-group, 

coupled libraries matched experimental measurements well — in terms of transmission/attenuation values 

and spectral results. The ENDF/B-VII.0 broad-group library was not used in forward calculations but 

was used in creating importance maps for Monaco. Overall, the results obtained with MAVRIC/Monaco 

using the coupled libraries are good and demonstrate the effectiveness of the new SCALE methods and 

data capabilities for shielding applications. 
88 
 89 

5 REFERENCES 

1. SCALE: A Modular Code System for Performing Standardized Computer Analyses for Licensing 

Evaluation, ORNL/TM-2005/39, Version 5.1, Vols. 1–III (November 2006). Available from 

Radiation Safety Information Computational Center at Oak Ridge National Laboratory as CCC732. 

2. National Nuclear Data Center (NNDC) at Brookhaven National Laboratory (BNL), Website 

http://www.nndc.bnl.gov

3. M. B. Chadwick et al., “ENDF/B-VII.0 Next Generation Evaluated Nuclear Data Library for 

Nuclear Science and Technology,” Special Issue on Evaluated Nuclear Data File ENDF/B-VII.0 

Nuclear Data Sheets 107(12), 2931–3059 (December 2006). 

4. M. E. Dunn and N. M. Greene, “AMPX-2000: A Cross-Section Processing System for Generating 

Nuclear Data for Criticality Safety Applications,” Trans. Am. Nucl. Soc. 86, 118–119 (2002). 

5. ENDF-102 Data Formats and Procedures for the Evaluated Nuclear Data File ENDF-6, 

BNL-NCS-44945, Rev. 10/91 (ENDF/B-VI), Brookhaven National Laboratory (October 1991). 

6. R. E. MacFarlane and D.W. Muir, NJOY99.0 Code System for Producing Pointwise and 

Multigroup Neutron and Photon Cross Sections from ENDF/B Data, PSR-480/NJOY99.0, 

Los Alamos National Laboratory, March 2000. 

7. B. L. Broadhead et al., “Evaluation of Shielding Analysis Methods in Spent Fuel Cask 

Environments,” EPRI TR-104329, Electric Power Research Institute, Palo Alto, California, 

May 1995. 

8. M. Nikolaev, N. Prokhorova, and T. Ivanova, “Neutron Fields in Three-Section Concrete 

Labyrinth from Cf-252 Source” (ALARM-CF-AIR-LAB-001), International Handbook of 

Evaluated Criticality Safety Benchmark Experiments, Volume VIII — Criticality Alarm/Shielding 

Benchmarks, NEA/NSC/DOC/(95)03, Organization for Economic Co-operation and Development 

— Nuclear Energy Agency (OECD-NEA) (September 2007). 

9. G. Manturov, Y. Rozhikhin, and L. Trykov, “Neutron and Photon Leakage Spectra from Cf-252 

Source at Centers of Six Iron Spheres of Different Diameters” (ALARM-CF-FE-SHIELD-001), 

International Handbook of Evaluated Criticality Safety Benchmark Experiments, Volume VIII — 

Criticality Alarm/Shielding Benchmarks, NEA/NSC/DOC/(95)03, Organization for Economic Cooperation and Development — Nuclear Energy Agency (OECD-NEA) (September 2007). 

10. B. Jansky, Z. Turik, E. Novak, J. Kyncl, F. Cvachovec, and P. Tiller, “Comparison of Measured 

and Calculatied Neutron Transmission Through Heavy Water for 252Cf Source Placed in the Center 

of 30 cm Diameter Sphere,” Annals of Nuclear Energy 24(15), 1189–1212 (1997). 

11. E. Sajo, M. L. Williams, and M. Asgari, “Comparison of Measured and Calculated Neutron 

Transmission Through Steel for a 252Cf Source,” Annals of Nuclear Energy 20(9), 585–604 (1993). 
90 

12. K. Ueki, A. Ohashi, N. Nariyama, S. Nagayama, T. Fujita, K. Hattori, and Y. Anayama, 

“Systematic Evaluation of Neutron Shielding Effects for Materials,” Nucl. Sci. Eng. 124, 

455–464 (1996). 

13. D. A. Torres, R. D. Mosteller, and J. E. Sweezy, “Comparison of MCNP5 and Experimental 

Results on Neutron Shielding Effects for Materials,” ANS Annual Meeting, Pittsburgh, PA, 

June 2004. Trans. Am. Nucl. Soc. 90, 484–486 (2004). 

14. D. A. Torres, B. Hehr, A. Sood, and R. D. Mosteller, “Comparison of MCNP5 Results with 

Experimental Measurements for Neutron-Shielding Materials,” in Proceedings of the American 

Nuclear Society’s 14th Biennial Topical Meeting of the Radiation Protection and Shielding 

Division, pp. 93–96, Carlsbad, New Mexico, April 3–6, 2006. 

15. C. J. Solomon, A. Sood, and R. D. Mosteller, “A Study of A Polyethylene and Stainless Steel 

Composite Deep Shield Benchmark of MCNP,” in Proceedings of the American Nuclear Society’s 

14th Biennial Topical Meeting of the Radiation Protection and Shielding Division, pp. 97–100, 

Carlsbad, New Mexico, April 3–6, 2006.
A-1 

APPENDIX A 

XML INPUT TEMPLATES 

The input files used to run the AMPX program sequences are generated using an XML template. The 

XML templates used for the various tasks are reproduced here. 

Input template to generate point-wise cross-section data 

<!DOCTYPE InputData [ 

<!ENTITY polident_files "${RTNDIR}/../neutron/writeData(tag)_polident"> 

<!ENTITY broaden_file "${RTNDIR}/../neutron/writeData(tag)_broaden"> 

<!ENTITY weighting_file "${RTNDIR}/../neutron/writeData(tag)_weight"> 

<!ENTITY home "input_neutron"> 

<!ENTITY case "&home;/writeData(tag)"> 

<!ENTITY batch "&home;/batch"> 

<!ENTITY ampx "/Users/dw8/ampx/mac/install/cmds/batcha"> 

]> 

<InputData> 

 <openFile name="&batch;"/> 

 <writeFile name="&batch;"> 

<text>#!/bin/bash -x 

</text> 

</writeFile> 

<loop restrict="+file3(yes) +neutron(yes)"> 

 <openFile name="&case;"/> 

 <writeFile name="&case;"> 

<text>=shell 

ln -sf writeData(tape) ft11f001 <!-- link ENDF tape in --> 

end 

=polident <!-- create point wise data --> 

-1$$ 20000 

0$$ 31 32 e 1$$ 1 t <!-- tab1 output on unit 31, file header info on unit 32 

--> 

2$$ writeData(endf) 11 2 6 e 

4** a5 0.001 e <!-- generate grid with precision of epsilon=0.001 --> 

6$$ a3 0 15000 t 

end 

=shell 

cp ft31f001 &polident_files; 

cp ft32f001 &polident_files;_ft32 

end 

=tgel <!-- tgel to fix total cross section data --> 

-1$$ 5000000 

0$$ 31 33 e t <!-- new file is on logical unit 33 --> 

end 

=broaden <!-- broaden the data --> 

limit=5000000 

logpt=33 logdp=34 <!-- broadened data are on logical unit 34 --> 

t= 0 300 600 900 1200 2400 

end 

=shell 

cp ft34f001 &broaden_file; 

end 

=pickeze <!-- pickeze --> 

-1$$ 3000000 

0$$ 34 35 <!-- pick T=300 from broadened file at unit=34 and write to unit 35 --

> 

1$$ 1 0 0 1 0 e t 

2$$ writeData(endf) 
 

A-2 

5** 300.0 t 

end 

=jergens <!-- jergens to generate desired flux --> 

-1$$ a11 3000000 e 

0$$ 35 30 18 1$$ 2 t 

3$$ 1 0 4 t 

3$$ 1099 4 0 t 

 2099 0 read 1.0 -1 0 

 2099 0 save 0 0 0 

 1099 2099 div 1.0 writeData(endf) 1 

 1099 0 save 0 0 0 

end 

=shell 

cp ft30f001 &weighting_file; <!-- save weighting spectrum --> 

end 

</text> 

 </writeFile> 

 <closeFile name="&case;"/> 

 <writeFile name="&batch;"> 

<text>&x; writeData(tag) 

</text> 

 </writeFile> 

 </loop> 

 <closeFile name="&batch;"/> 

</InputData> 

Input template to generate multigroup neutron and gamma cross-section data 

<?xml version="1.0"?> 

<!DOCTYPE InputData [ 

<!ENTITY polident_files "${RTNDIR}/../../neutron/writeData(tag)_polident"> 

<!ENTITY broaden_file "${RTNDIR}/../../neutron/writeData(tag)_broaden"> 

<!ENTITY weighting_file "${RTNDIR}/../../neutron/writeData(tag)_weight"> 

<!ENTITY x10_fast "${RTNDIR}/../neutron/writeData(tag)_fast"> 

<!ENTITY x10_yield "${RTNDIR}/../neutron/writeData(tag)_yield"> 

<!ENTITY group "${RTNDIR}/../groupStructure"> 

<!ENTITY mmel1 "32"> <!-- In y12 Lobatto quadrature angles for elastic scattering matrices 

A<=180 --> 

<!ENTITY mmel2 "8"> <!-- In y12 Lobatto quadrature angles for elastic scattering matrices 

A>180 --> 

<!ENTITY mmin1 "8"> <!-- In y12 Lobatto quadrature angles for discrete level inelastic 

scattering matrices --> 

<!ENTITY mmdf "8"> <!-- In y12 Number of Lobatto quadrature angles --> 

<!ENTITY mmin2 "8"> <!-- In y12 Lobatto quadrature angles for discrete level inelastic 

scattering matrices --> 

<!ENTITY nl "5"> <!-- In y12 Number of legendre polynomilas to use --> 

<!ENTITY igm "200"> <!-- number of neutron groups --> 

<!ENTITY ipm "47"> <!-- number of gamma groups --> 

<!ENTITY nlneutron "5"> <!-- The maximum order of Legendre fit to neutron scattering 

processes --> 

<!ENTITY nlyield "2"> <!-- The maximum order of Legendre fit to gamma-ray yield 

processes --> 

<!ENTITY nlgamma "5"> <!-- The maximum order of Legendre fit to gamma-ray interaction 

scattering --> 

<!ENTITY title "writeData(tag) writeData(endf) ENDFBV7 RELwriteData(rel) REVwriteData(rev) 

MODwriteData(mod) AMPX writeData(date)"> 

<!ENTITY home "input_fast"> 

<!ENTITY case "&home;/writeData(tag)"> 
A-3 

<!ENTITY batch "&home;/batch"> 

<!ENTITY ampx "~/ampx/mac/install/cmds/batcha"> 

]> 

<InputData> 

 <openFile name="&batch;"/> 

 <writeFile name="&batch;"> 

<text>#!/bin/bash -x 

</text>a 

</writeFile> 

<loop restrict="+file3(yes) +neutron(yes) +neutron(yes)"> 

 <openFile name="&case;"/> 

 <writeFile name="&case;"> 

<text>=shell 

ln -sf writeData(tape) ft11f001 

ln -sf &polident_files; ft31f001 

ln -sf &broaden_file; ft34f001 

ln -sf &weighting_file; ft30f001 

ln -sf &group; ft47f001 

</text> 

<text>end 

=y12 <!-- determine kinematic data --> 

0$$ 32 11 0 0 <!-- kinematic data on unit 32, ENDF on 11. Generate no thermal neutron or 

gamma data --> 

1$$ writeData(endf) 

2$$ 2 6 <!-- Endf file is BCD and ENDF version 6 --> 

3$$ &mmel1; &mmel2; &mmin1; &mmdf; &mmin2; &nl; t 

t 

end 

=pickeze <!-- pickeze --> 

-1$$ 9000000 

0$$ 34 35 <!-- pick T=300 from broadened file at unit=34 and write to unit 35 --

> 

1$$ 1 0 0 1 0 e t 

2$$ writeData(endf) 

5** 300.0 t 

end 

</text> 

<text>=x10 

neutron <!-- x10 for yield --> 

-1$$ 9000000 

0$$ 1 30 35 32 <!-- Master on 1, weighting spectrum on 30, point data for T=300 on 35, 

kinematic data in 32 --> 

1$$ writeData(endf) &igm; 0 &ipm; &nlneutron; &nlyield; &nlgamma; 

2$$ 99 2099 writeData(endf) writeData(endf) </text> <!-- flux data --> 

<text restrict="file2(yes)">3** writeData(scattering)</text> 

<text restrict="-file2(yes)">3** 0.0</text> 

<text> 6$$ 99 2099 t 

t 

&title; 

end 

=rade <!-- use rade to check results --> 

1$$ 1 e t 

end 

=shell 

cp ft01f001 &x10_fast; 

end 

</text> 

<text>=x10 

yield 

-1$$ 9000000 

0$$ 71 30 35 32 <!-- Master on 71, weighting spectrum on 30, point data for T=300 on 35, 

kinematic data in 32 --> 

1$$ writeData(endf) &igm; 0 &ipm; &nlneutron; &nlyield; &nlgamma; 
 

A-4 

2$$ 99 2099 writeData(endf) writeData(endf) </text> <!-- flux data --> 

<text restrict="+file2(yes)>3** writeData(scattering)</text> 

<text restrict="-file2(yes)>3** 0.0</text> 

<text> 6$$ 99 2099 t 

t 

&title; 

end 

=rade <!-- use rade to check results --> 

1$$ 71 e t 

end 

=shell 

cp ft71f001 &x10_yield; 

end 

</text> 

 </writeFile> 

 <closeFile name="&case;"/> 

 <writeFile name="&batch;"> 

<text>&x; writeData(tag) 

</text> 

 </writeFile> 

 </loop> 

 <closeFile name="&batch;"/> 

</InputData> 

Input template to generate thermal cross-section data for moderators 

<?xml version="1.0"?> 

<!DOCTYPE InputData [ 

<!-- place where to put the kinematic data --> 

<!ENTITY thermal_kinematics "${RTNDIR}/../thermal/writeData(tag)_kin"> 

<!-- place where to put the point-wise data --> 

<!ENTITY thermal_point "${RTNDIR}/../thermal/writeData(tag)_point"> 

<!-- place for the master library --> 

<!ENTITY thermal_master "${RTNDIR}/../thermal/writeData(tag)_master"> 

<!-- title to use for the library --> 

<!ENTITY title "writeData(tag) writeData(endf) ENDFB V6 RELwriteData(rel) REVwriteData(rev) 

MODwriteData(mod) AMPX writeData(date)"> 

<!ENTITY group "${RTNDIR}/../groupStructure"> 

<!ENTITY home "input_thermal"> 

<!ENTITY case "&home;/writeData(tag)"> 

<!ENTITY batch "&home;/batch"> 

<!ENTITY ampx "~/ampx/mac/install/cmds/batcha"> 

]> 

<InputData> 

 <openFile name="&batch;"/> 

 <writeFile name="&batch;"> 

<text>#!/bin/bash -x 

</text> 

</writeFile> 

<loop restrict="file7(yes)"> <!-- only for evaluations that have file 7, i.e. have 

thermal data --> 

 <openFile name="&case;"/> 

 <writeFile name="&case;"> 

<text>=shell 

ln -sf writeData(tape) ft11f001 
A-5 

ln -sf &group; ft47f001 

end 

=y12 

0$$ 37 11 38 0 <!-- kinematics data will be on 37, ENDF is on 11, output for point-wise 

data is 38 --> 

1$$ writeData(endf) <!-- material to process --> 

2$$ 2 6 <!-- endf format is BCD and ENDF version is 6 --> 

3$$ 32 8 8 8 8 5 t <!-- Number of lobatto quadrutor: 32 for A <= 180, 8 otherwise; 

Order of legendre fit is 5 --> 

t 

end 

=shell <!-- save the kinematics data and the point-wise data --> 

cp ft37f001 &thermal_kinematics; 

cp ft38f001 &thermal_point; 

end 

=jergens <!-- generate weighting function --> 

0$$ 11 30 18 <!-- external is endf , output is on unit 30 --> 

1$$ 1 t <!-- Generate one function --> 

3$$ 1099 0 2 t <!-- mt value of weighting function is 1099, and we generate Maxwellian - 

1/E - fission spectrum --> 

end 

=x10 

neutron <!-- calculate neutron interaction --> 

0$$ 4 30 38 37 <!-- final on 4, jergens output on 30, y12 point-wise on 38, y12 

kinematics output on 37 --> 

1$$ writeData(endf) 200 34 <!-- new id, 200 neutron groups from that thermal 

groups --> 

 0 <!-- no gamma groups --> 

 5 2 0 <!-- order of legendre polynomials --> 

2$$ 99 1099 writeData(endf) writeData(endf) <!-- id of weighing function, kinematic data 

and point-wise data --> 

3** 0 t <!-- Potential Scattering Cross Section - 0 for all thermal runs since no file 

2 --> 

t 

&title; <!-- title of the library --> 

end 

=shell 

cp ft04f001 &thermal_master; <!-- copy master --> 

end 

=rade <!-- use rade to check results --> 

1$$ 4 e t 

end 

=paleale <!-- use paleale to print results in ASCII --> 

0$$ 4 e 2$$ 1 0 e 5$$ f 200 t 

end 

</text> 

 </writeFile> 

 <closeFile name="&case;"/> 

 <writeFile name="&batch;"> 

<text>&x; writeData(tag) 

</text> 

 </writeFile> 

 </loop> 

 <closeFile name="&batch;"/> 

</InputData> 
 

A-6 

Input template to generate thermal cross-section data for nonmoderators 

<?xml version="1.0"?> 

<!DOCTYPE InputData [ 

<!ENTITY weighting_file "${RTNDIR}/../neutron/writeData(tag)_weight"> 

<!ENTITY flange6 "${RTNDIR}/../neutron/writeData(tag)_flange6"> 

<!ENTITY group "${RTNDIR}/../groupStructure"> 

<!ENTITY igm "200"> <!-- number of neutron groups --> 

<!ENTITY neg "34"> <!-- number of thermal groups --> 

<!ENTITY home "input_bon"> 

<!ENTITY case "&home;/writeData(tag)"> 

<!ENTITY batch "&home;/batch"> 

<!ENTITY ampx "/home/dw8/ampx/build/install/cmds/batcha"> 

]> 

<InputData> 

 <openFile name="&batch;"/> 

 <writeFile name="&batch;"> 

<text>#!/bin/bash -x 

</text>a 

</writeFile> 

<loop restrict="+file3(yes) +neutron(yes)"> 

 <openFile name="&case;"/> 

 <writeFile name="&case;"> 

<text>=shell 

ln -sf &weighting_file; ft30f001 

ln -sf &group; ft47f001 

end 

</text> 

<text>=flange6 <!-- run flange6 --> 

nl=3 za=writeData(za) awr=writeData(awr) </text> 

<text restrict="file2(yes)">free=writeData(scattering) </text><text restrict="-

file2(yes)">free=0.0 </text> 

<text> 

mat=0 master=41 igm=200 neg=&neg; 

t=300 t=600 t=900 t=1200 t=2100 

end 

=shell 

cp ft41f001 &flange6; 

end 

</text> 

 </writeFile> 

 <closeFile name="&case;"/> 

 <writeFile name="&batch;"> 

<text>&x; writeData(tag) 

</text> 

 </writeFile> 

 </loop> 

 <closeFile name="&batch;"/> 

</InputData> 
A-7 

Input template used to generate full-range Bondarenko data 

<?xml version="1.0"?> 

<!DOCTYPE InputData [ 

<!ENTITY polident_files "${RTNDIR}/../../neutron/writeData(tag)_polident"> 

<!ENTITY weighting_file "${RTNDIR}/../../neutron/writeData(tag)_weight"> 

<!ENTITY prude "${RTNDIR}/../neutron/writeData(tag)_prude"> 

<!ENTITY fabulous "${RTNDIR}/../neutron/writeData(tag)_fabulous"> 

<!ENTITY group "${RTNDIR}/../groupStructure"> 

<!ENTITY igm "200"> <!-- number of neutron groups --> 

<!ENTITY neg "34"> <!-- number of thermal groups --> 

<!ENTITY home "input_bon"> 

<!ENTITY case "&home;/writeData(tag)"> 

<!ENTITY batch "&home;/batch"> 

<!ENTITY ampx "~/ampx/mac/install/cmds/batcha"> 

]> 

<InputData> 

 <openFile name="&batch;"/> 

 <writeFile name="&batch;"> 

<text>#!/bin/bash -x 

</text>a 

</writeFile> 

<loop restrict="+file3(yes) +neutron(yes)"> 

 <openFile name="&case;"/> 

 <writeFile name="&case;"> 

<text>=shell 

ln -sf writeData(tape) ft11f001 

ln -sf &polident_files; ft31f001 

ln -sf &weighting_file; ft30f001 

ln -sf &group; ft47f001 

end 

</text> 

<!-- run prude if unresolved --> 

<text restrict="unresolved(yes)">=prude 

0$$ 36 <!-- output will be on unit 36 --> 

1$$ 1 t <!-- and we process one material --> 

2$$ writeData(endf) 9 3 11 2 t <!-- process 9 sigma values, 3 temperatures. Endf file is on 

11 and BCD --> 

3** 1+8 1+6 1+5 1+4 1000 100 10 1 1-6 <!-- the 9 sigma values --> 

4** 300.0 900.0 2000.0 e <!-- the 3 temperatures --> 

5** 0 1 e t 

end 

=shell 

cp ft36f001 &prude; 

end 

</text> 

<text>=tgel 

-1$$ 9000000 

0$$ 31 32 e t 

end 

</text> 

<text>=fabulous <!-- generate full range bondarenko factors --> 

Title Card for FABULOUS case <!-- title for master --> 

-1$$ 2500000 

0$$ 2 32 30 </text> <!-- master output=2, point data = 37, flux = 30 --> 

<!-- unit of unresolved data if present is 36, 0 otherwise --> 

<text restrict="unresolved(yes)">36 </text> 
 

A-8 

<text>e 

1$$ writeData(endf) writeData(endf) 3 9 &igm; <!-- id on master, endf material to process, 

number of temps, number of sigs --> 

2$$ 99 2099 <!-- where to find the weighting function --> 

5** writeData(awr) 0.0001 t <!-- mass ratio and to what precision to calculate --> 

8** 300.0 900.0 2000.0 <!-- the 3 temperatures --> 

9** 1+8 1+6 1+5 1+4 1000 100 10 1 1-6 <!-- the 9 sigma values --> 

t 

end 

=shell 

cp ft02f001 &fabulous; 

end 

</text> 

 </writeFile> 

 <closeFile name="&case;"/> 

 <writeFile name="&batch;"> 

<text>&x; writeData(tag) 

</text> 

 </writeFile> 

 </loop> 

 <closeFile name="&batch;"/> 

</InputData> 

Input template used to process gamma cross-section data 

<?xml version="1.0"?> 

<!DOCTYPE InputData [ 

<!-- place where to put the kinematic data --> 

<!ENTITY gamma_kinematics "${RTNDIR}/../gamma/writeData(tag)_kin"> 

<!-- place where to put the point-wise data --> 

<!ENTITY gamma_point "${RTNDIR}/../gamma/writeData(tag)_point"> 

<!-- place for the master library --> 

<!ENTITY gamma_master "${RTNDIR}/../gamma/writeData(tag)_master"> 

<!-- title to use for the library --> 

<!ENTITY title "writeData(tag) writeData(endf) ENDFB V6 RELwriteData(rel) REVwriteData(rev) 

MODwriteData(mod) AMPX writeData(date)"> 

<!ENTITY home "input_gamma"> 

<!ENTITY case "&home;/writeData(tag)"> 

<!ENTITY batch "&home;/batch"> 

<!ENTITY ampx "~/ampx/mac/install/cmds/batcha"> 

]> 

<InputData> 

 <openFile name="&batch;"/> 

 <writeFile name="&batch;"> 

<text>#!/bin/bash -x 

</text> 

</writeFile> 

<loop restrict="-neutron(yes)"> 

 <openFile name="&case;"/> 

 <writeFile name="&case;"> 

<text>=shell 

ln -sf writeData(tape) ft11f001 

end 

=y12 
A-9 

0$$ 32 11 31 0 <!-- kinematics data will be on 32, ENDF is on 11, output for point-wise 

data is 31 --> 

1$$ writeData(endf) <!-- material to process --> 

2$$ 2 6 <!-- endf format is BCD and ENDF version is 6 --> 

3$$ 32 8 8 8 8 5 t <!-- Number of lobatto quadrutor: 32 for A <= 180, 8 otherwise; 

Order of legendre fit is 5 --> 

t 

end 

=shell <!-- save the kinematics data and the point-wise data --> 

cp ft32f001 &gamma_kinematics; 

cp ft31f001 &gamma_point; 

end 

=jergens <!-- generate weighting function --> 

0$$ 0 30 18 <!-- no external data needed, output is on unit 30 --> 

1$$ 1 <!-- Generate one function --> 

2** 1.0e3 3.0e7 t <!-- generate weighting function from 1e-3 to 3e7 --> 

3$$ 1599 0 11 t <!-- mt value of weighting function is 1599, and we generate 1/E with 

"rolloff" gamma weighting --> 

end 

=x10 

gamma <!-- calculate gamma-ray interaction --> 

0$$ 1 30 31 32 <!-- final on 1, jergens output on 30, y12 point-wise on 31 y12 kinematics 

output on 32 --> 

1$$ writeData(endf) 0 0 <!-- new id, no neutron or thermal groups --> 

 47 <!-- 47 gamma groups --> 

 0 0 5 <!-- order of legendre polynomials --> 

2$$ 99 1599 writeData(endf) writeData(endf) <!-- id of weighing function, kinematic data 

and point-wise data --> 

3** 0.0 t <!-- Potential Scattering Cross Section --> 

7** 2.0E+7 1.4E+7 1.2E+7 1.E+7 8.E+6 <!-- group 

structure --> 

 7.5E+6 7.E+6 6.5E+6 6.E+6 5.5E+6 

 5.0E+6 4.5E+6 4.E+6 3.5E+6 3.E+6 

 2.75E+6 2.5E+6 2.35E+6 2.15E+6 2.E+6 

 1.8E+6 1.66E+6 1.57E+6 1.5E+6 1.44E+6 

 1.33E+6 1.2E+6 1.E+6 9.0E+5 8.0E+5 

 7.0E+5 6.0E+5 5.12E+5 5.1E+5 4.5E+5 

 4.0E+5 3.0E+5 2.6E+5 2.0E+5 1.5E+5 

 1.0E+5 7.5E+4 7.0E+4 6.0E+4 4.5E+4 

 3.0E+4 2.0E+4 1.0E+4 e 

t 

&title; <!-- title of the library --> 

end 

=shell 

cp ft01f001 &gamma_master; <!-- copy master --> 

end 

=rade <!-- use rade to check results --> 

1$$ 1 e t 

end 

=paleale <!-- use paleale to print results in ASCII --> 

0$$ 1 e 2$$ 0 1 e 4$$ 502 504 e 5$$ f 47 t 

end 

</text> 

 </writeFile> 

 <closeFile name="&case;"/> 

 <writeFile name="&batch;"> 

<text>&x; writeData(tag) 

</text> 

 </writeFile> 

 </loop> 

 <closeFile name="&batch;"/> 

</InputData> 
 

A-10 

Input template to generate the AMPX master for nonmoderator evaluations 

<?xml version="1.0"?> 

<!DOCTYPE InputData [ 

<!ENTITY fast "${RTNDIR}/../neutron/writeData(tag)_fast"> 

<!ENTITY yield "${RTNDIR}/../neutron/writeData(tag)_yield"> 

<!ENTITY fabulous "${RTNDIR}/../neutron/writeData(tag)_fabulous"> 

<!ENTITY flange6 "${RTNDIR}/../neutron/writeData(tag)_flange6"> 

<!ENTITY gamma "${RTNDIR}/../gamma/writeData(taggamma)_master"> 

<!ENTITY title "writeData(tag) writeData(endf) ENDFB6 RELwriteData(rel) REVwriteData(rev) 

MODwriteData(mod) AMPX writeData(date)"> 

<!ENTITY neutron_final "${RTNDIR}/../master/writeData(tag)_neutron"> 

<!ENTITY home "input_stitch"> 

<!ENTITY case "&home;/writeData(tag)"> 

<!ENTITY batch "&home;/batch"> 

<!ENTITY ampx "~/ampx/mac/install/cmds/batcha"> 

]> 

<InputData> 

 <openFile name="&batch;"/> 

 <writeFile name="&batch;"> 

<text>#!/bin/bash -x 

</text> 

</writeFile> 

<loop restrict="+file3(yes) +neutron(yes) -file7(yes)"> <!-- restrict to non-moderator 

data --> 

 <openFile name="&case;"/> 

 <writeFile name="&case;"> 

<text>=shell 

ln -sf &fast; ft01f001 <!-- fast neutron data --> 

ln -sf &yield; ft02f001 <!-- gamma yield data --> 

ln -sf &fabulous; ft03f001 <!-- bondarenko data --> 

ln -sf &flange6; ft04f001 <!-- thermal scattering data --> 

</text> 

<text restrict="gamma">ln -sf γ ft05f001 <!-- add gamma evaluation if present --> 

</text> 

<text>end 

</text> 

<!-- run simonize to combine data --> 

<text>=simonize 

Identifier=writeData(changedza) master=20 source=endf <!-- change to --> 

title= &title; 

id45=writeData(scaleid) id46=0 </text> 

<!-- add gamma values if available --> 

<text restrict="gamma"> id47=writeData(scaleidgamma) </text> 

<text restrict="-gamma"> id47=0 </text> 

<text restrict="AWP0(yes) file12(yes) file13(yes)"> id48=writeData(scaleid) </text> 

<text restrict="-AWP0(yes) -file12(yes) -file13(yes)"> id48=0 </text> 

<!-- make sure simonize has at least one neutron flag to move over id data --> 

<text> 

neutron=1 id19=writeData(endf) <!-- fast neutron data - always present --> 

</text> 

<text> yield=2 id19=writeData(endf) <!-- yield data if present --> 
A-11 

</text> 

<text restrict="nis(0) nis(1)"> BONDARENKO=3 id19=writeData(endf) <!-- bondarenko data if 

present --> 

</text> 

<text>2dn=4 id19=writeData(za) <!-- thermal scattering data - always present --> 

</text> 

<text restrict="gamma">gamma=5 id19=writeData(endfgamma) <!-- add the gamma data if present 

--> 

</text> 

<text>end 

</text> 

<!-- copy final result --> 

<text> 

=rade 

1$$ 20 e t 

end 

=shell 

cp ft20f001 &neutron_final; 

end 

</text> 

 </writeFile> 

 <closeFile name="&case;"/> 

 <writeFile name="&batch;"> 

<text>&x; writeData(tag) 

</text> 

 </writeFile> 

 </loop> 

 <closeFile name="&batch;"/>ß 

</InputData> 

Input template to generate the AMPX master for moderator evaluations 

<?xml version="1.0"?> 

<!DOCTYPE InputData [ 

<!ENTITY fast "${RTNDIR}/../neutron/writeData(tag)_fast"> 

<!ENTITY yield "${RTNDIR}/../neutron/writeData(tag)_yield"> 

<!ENTITY fabulous "${RTNDIR}/../neutron/writeData(tag)_fabulous"> 

<!ENTITY flange6 "${RTNDIR}/../neutron/writeData(tag)_flange6"> 

<!ENTITY gamma "${RTNDIR}/../gamma/writeData(taggamma)_master"> 

<!ENTITY title "writeData(tag) writeData(endf) ENDFB6 RELwriteData(rel) REVwriteData(rev) 

MODwriteData(mod) AMPX writeData(date)"> 

<!ENTITY neutron_final "${RTNDIR}/../master/writeData(tag)_neutron"> 

<!ENTITY home "input_stitch"> 

<!ENTITY case "&home;/writeData(tag)"> 

<!ENTITY batch "&home;/batch"> 

<!ENTITY ampx "~/ampx/mac/install/cmds/batcha"> 

]> 

<InputData> 

 <openFile name="&batch;"/> 

 <writeFile name="&batch;"> 

<text>#!/bin/bash -x 
 

A-12 

</text> 

</writeFile> 

<loop restrict="+file3(yes) +neutron(yes) -file7(yes)"> <!-- restrict to non-moderator 

data --> 

 <openFile name="&case;"/> 

 <writeFile name="&case;"> 

<text>=shell 

ln -sf &fast; ft01f001 <!-- fast neutron data --> 

ln -sf &yield; ft02f001 <!-- gamma yield data --> 

ln -sf &fabulous; ft03f001 <!-- bondarenko data --> 

ln -sf &flange6; ft04f001 <!-- thermal scattering data --> 

</text> 

<text restrict="gamma">ln -sf γ ft05f001 <!-- add gamma evaluation if present --> 

</text> 

<text>end 

</text> 

<!-- run simonize to combine data --> 

<text>=simonize 

Identifier=writeData(changedza) master=20 source=endf <!-- change to --> 

title= &title; 

id45=writeData(scaleid) id46=0 </text> 

<!-- add gamma values if available --> 

<text restrict="gamma"> id47=writeData(scaleidgamma) </text> 

<text restrict="-gamma"> id47=0 </text> 

<text restrict="AWP0(yes) file12(yes) file13(yes)"> id48=writeData(scaleid) </text> 

<text restrict="-AWP0(yes) -file12(yes) -file13(yes)"> id48=0 </text> 

<!-- make sure simonize has at least one neutron flag to move over id data --> 

<text> 

neutron=1 id19=writeData(endf) <!-- fast neutron data - always present --> 

</text> 

<text> yield=2 id19=writeData(endf) <!-- yield data if present --> 

</text> 

<text restrict="nis(0) nis(1)"> BONDARENKO=3 id19=writeData(endf) <!-- bondarenko data if 

present --> 

</text> 

<text>2dn=4 id19=writeData(za) <!-- thermal scattering data - always present --> 

</text> 

<text restrict="gamma">gamma=5 id19=writeData(endfgamma) <!-- add the gamma data if present 

--> 

</text> 

<text>end 

</text> 

<!-- copy final result --> 

<text> 

=rade 

1$$ 20 e t 

end 

=shell 

cp ft20f001 &neutron_final; 

end 

</text> 

 </writeFile> 

 <closeFile name="&case;"/> 

 <writeFile name="&batch;"> 

<text>&x; writeData(tag) 

</text> 

 </writeFile> 

 </loop> 
A-13 

 <closeFile name="&batch;"/> 

</InputData> 
 

A-14 
B-1 

APPENDIX B 

EVALUATIONS INCLUDED IN ENDF/B-VI.8 LIBRARY 

Evaluations included in the multigroup library for ENDF/B-VI.8 are listed in Table B.1. 

Table B.1 Evaluations and IDs used for the ENDF/B-VI.8 library

Nuclide ID Photon 

yield data 

Full range 

self-shielding 

factors 

Corresponding 

photon data 

h-1 1801 yes yes h 

h-2 1802 yes yes h 

h-3 1003 yes h 

he-3 2003 yes he 

he-4 2004 yes he 

li-6 3006 yes yes li 

li-7 3007 yes yes li 

be-9 4009 yes yes be 

b-10 5010 yes yes b 

b-11 5011 yes yes b 

c 6000 yes yes c 

n-14 7014 yes yes n 

n-15 7015 yes yes n 

o-16 8016 yes yes o 

o-17 8017 yes o 

f-19 9019 yes yes f 

na-23 11023 yes yes na 

mg 12000 yes yes mg 

al-27 13027 yes yes al 

si 1 4000 yes yes si 

si-28 14028 yes yes si 

si-29 14029 yes yes si 

si-30 14030 yes yes si 

p-31 15031 yes yes p 

s 16000 yes s 

s-32 16032 yes yes s 

cl 17000 yes yes cl 

cl-35 17035 yes yes cl 

cl-37 17037 yes yes cl 

k 19000 yes yes k 

ca 20000 yes yes ca 

sc-45 21045 yes yes sc 

ti 22000 yes yes ti 

v 23000 yes v 

cr-50 24050 yes yes cr 

cr-52 24052 yes yes cr 

cr-53 24053 yes yes cr 

cr-54 24054 yes yes cr 

mn-55 25055 yes yes mn 

fe-54 26054 yes yes fe 

fe-56 26056 yes yes fe 

fe-57 26057 yes yes fe 

fe-58 26058 yes yes fe 
B-2 

Table B.1 Evaluations and IDs used for the ENDF/B-VI.8 library (continued) 

Nuclide ID Photon 

yield data 

Full range 

self-shielding 

factors 

Corresponding 

photon data 

co-59 27059 yes yes co 

ni-58 28058 yes yes ni 

ni-60 28060 yes yes ni 

ni-61 28061 yes yes ni 

ni-62 28062 yes yes ni 

ni-64 28064 yes yes ni 

cu-63 29063 yes yes cu 

cu-65 29065 yes yes cu 

ga 31000 yes yes ga 

ge-72 32072 yes ge 

ge-73 32073 yes ge 

ge-74 32074 yes ge 

ge-76 32076 yes ge 

as-75 33075 yes as 

se-74 34074 yes se 

se-76 34076 yes se 

se-77 34077 yes se 

se-78 34078 yes se 

se-80 34080 yes se 

se-82 34082 yes se 

br-79 35079 yes br 

br-81 35081 yes br 

kr-78 36078 yes kr 

kr-80 36080 yes kr 

kr-82 36082 yes kr 

kr-83 36083 yes kr 

kr-84 36084 yes kr 

kr-85 36085 yes kr 

kr-86 36086 yes kr 

rb-85 37085 yes rb 

rb-86 37086 yes rb 

rb-87 37087 yes rb 

sr-84 38084 yes sr 

sr-86 38086 yes sr 

sr-87 38087 yes sr 

sr-88 38088 yes sr 

sr-89 38089 yes sr 

sr-90 38090 yes sr 

y-89 39089 yes yes y 

y-90 39090 yes y 

y-91 39091 yes y 

zr 40000 zr 

zr-90 40090 yes zr 

zr-91 40091 yes zr 

zr-92 40092 yes zr 

zr-93 40093 yes zr 

zr-94 40094 yes zr 

zr-95 40095 yes zr 

zr-96 40096 yes zr 

nb-93 41093 yes yes nb 
B-3 

Table B.1 Evaluations and IDs used for the ENDF/B-VI.8 library (continued) 

Nuclide ID Photon 

yield data 

Full range 

self-shielding 

factors 

Corresponding 

photon data 

nb-94 41094 yes nb 

nb-95 41095 yes nb 

mo 42000 yes mo 

mo-94 42094 yes mo 

mo-95 42095 yes mo 

mo-96 42096 yes mo 

mo-97 42097 yes mo 

mo-99 42099 yes mo 

tc-99 43099 yes tc 

ru-96 44096 yes ru 

ru-98 44098 yes ru 

ru-99 44099 yes ru 

ru-100 44100 yes ru 

ru-101 44101 yes ru 

ru-102 44102 yes ru 

ru-103 44103 yes ru 

ru-104 44104 yes ru 

ru-105 44105 yes ru 

ru-106 44106 yes ru 

rh-103 45103 yes rh 

rh-105 45105 yes rh 

pd-102 46102 yes yes pd 

pd-104 46104 yes yes pd 

pd-105 46105 yes yes pd 

pd-106 46106 yes yes pd 

pd-107 46107 yes pd 

pd-108 46108 yes yes pd 

pd-110 46110 yes yes pd 

ag-107 47107 yes ag 

ag-109 47109 yes ag 

ag-111 47111 yes ag 

cd 48000 yes cd 

cd-106 48106 yes cd 

cd-108 48108 yes cd 

cd-110 48110 yes cd 

cd-111 48111 yes cd 

cd-112 48112 yes cd 

cd-113 48113 yes cd 

cd-114 48114 yes cd 

cd-115 48601 yes cd 

cd-116 48116 yes cd 

in 49000 yes in 

in-113 49113 yes in 

sn-112 50112 yes sn 

sn-114 50114 yes sn 

sn-115 50115 yes sn 

sn-116 50116 yes sn 

sn-117 50117 yes sn 

sn-118 50118 yes sn 

sn-119 50119 yes sn 

sn-120 50120 yes sn 
B-4 

Table B.1 Evaluations and IDs used for the ENDF/B-VI.8 library (continued) 

Nuclide ID Photon 

yield data 

Full range 

self-shielding 

factors 

Corresponding 

photon data 

sn-122 50122 yes sn 

sn-123 50123 yes sn 

sn-124 50124 yes sn 

sn-125 50125 yes sn 

sn-126 50126 yes sn 

sb-121 51121 yes sb 

sb-123 51123 yes sb 

sb-124 51124 yes sb 

sb-125 51125 yes sb 

sb-126 51126 yes sb 

te-120 52120 yes te 

te-122 52122 yes te 

te-123 52123 yes te 

te-124 52124 yes te 

te-125 52125 yes te 

te-126 52126 yes te 

te-127 52601 yes te 

te-128 52128 yes te 

te-129 52611 yes te 

te-130 52130 yes te 

te-132 52132 yes te 

i-127 53127 yes yes i 

i-129 53129 yes i 

i-130 53130 yes i 

i-131 53131 yes i 

i-135 53135 yes i 

xe-124 54124 yes xe 

xe-126 54126 yes xe 

xe-128 54128 yes xe 

xe-129 54129 yes xe 

xe-130 54130 yes xe 

xe-131 54131 yes xe 

xe-132 54132 yes xe 

xe-133 54133 yes xe 

xe-134 54134 yes xe 

xe-135 54135 yes xe 

xe-136 54136 yes xe 

cs-133 55133 yes cs 

cs-134 55134 yes cs 

cs-135 55135 yes cs 

cs-136 55136 yes cs 

cs-137 55137 yes cs 

ba-134 56134 yes ba 

ba-135 56135 yes ba 

ba-136 56136 yes ba 

ba-137 56137 yes ba 

ba-138 56138 yes yes ba 

ba-140 56140 yes ba 

la-139 57139 yes la 

la-140 57140 yes la 

ce-140 58140 yes ce 
B-5 

Table B.1 Evaluations and IDs used for the ENDF/B-VI.8 library (continued) 

Nuclide ID Photon 

yield data 

Full range 

self-shielding 

factors 

Corresponding 

photon data 

ce-141 58141 yes ce 

ce-142 58142 yes ce 

ce-143 58143 yes ce 

ce-144 58144 yes ce 

pr-141 59141 yes pr 

pr-142 59142 yes pr 

pr-143 59143 yes pr 

nd-142 60142 yes nd 

nd-143 60143 yes nd 

nd-144 60144 yes nd 

nd-145 60145 yes nd 

nd-146 60146 yes nd 

nd-147 60147 yes nd 

nd-148 60148 yes nd 

nd-150 60150 yes nd 

pm-147 61147 yes pm 

pm-148 61148 yes pm 

pm-148 61601 yes pm 

pm-149 61149 yes pm 

pm-151 61151 yes pm 

sm-144 62144 yes sm 

sm-147 62147 yes sm 

sm-148 62148 yes sm 

sm-149 62149 yes sm 

sm-150 62150 yes sm 

sm-151 62151 yes sm 

sm-152 62152 yes sm 

sm-153 62153 yes sm 

sm-154 62154 yes sm 

eu-151 63151 yes yes eu 

eu-152 63152 yes eu 

eu-153 63153 yes yes eu 

eu-154 63154 yes eu 

eu-155 63155 yes eu 

eu-156 63156 yes eu 

eu-157 63157 yes eu 

gd-152 64152 yes gd 

gd-154 64154 yes gd 

gd-155 64155 yes gd 

gd-156 64156 yes gd 

gd-157 64157 yes gd 

gd-158 64158 yes gd 

gd-160 64160 yes gd 

tb-159 65159 yes tb 

tb-160 65160 yes tb 

dy-160 66160 yes dy 

dy-161 66161 yes dy 

dy-162 66162 yes dy 

dy-163 66163 yes dy 

dy-164 66164 yes dy 

ho-165 67165 yes yes ho 
B-6 

Table B.1 Evaluations and IDs used for the ENDF/B-VI.8 library (continued) 

Nuclide ID Photon 

yield data 

Full range 

self-shielding 

factors 

Corresponding 

photon data 

er-166 68166 yes er 

er-167 68167 yes er 

lu-175 71175 yes lu 

lu-176 71176 yes lu 

hf 72000 hf 

hf-174 72174 yes hf 

hf-176 72176 yes hf 

hf-177 72177 yes hf 

hf-178 72178 yes hf 

hf-179 72179 yes hf 

hf-180 72180 yes hf 

ta-181 73181 yes yes ta 

ta-182 73182 yes ta 

w 74000 yes w 

w-182 74182 yes yes w 

w-183 74183 yes yes w 

w-184 74184 yes yes w 

w-186 74186 yes yes w 

re-185 75185 yes re 

re-187 75187 yes re 

ir-191 77191 yes yes ir 

ir-193 77193 yes yes ir 

au-197 79197 yes yes au 

pb-206 82206 yes yes pb 

pb-207 82207 yes yes pb 

pb-208 82208 yes yes pb 

bi-209 83209 yes yes bi 

th-230 90230 yes th 

th-232 90232 yes yes th 

pa-231 91231 yes pa 

pa-232 91232 yes pa 

pa-233 91233 yes pa 

u-232 92232 yes u 

u-233 92233 yes yes u 

u-234 92234 yes u 

u-235 92235 yes yes u 

u-236 92236 yes u 

u-237 92237 yes yes u 

u-238 92238 yes yes u 

np-236 93236 yes np 

np-237 93237 yes yes np 

np-238 93238 yes np 

np-239 93239 yes np 

pu-236 94236 yes pu 

pu-237 94237 yes pu 

pu-238 94238 yes pu 

pu-239 94239 yes yes pu 

pu-240 94240 yes yes pu 

pu-241 94241 yes yes pu 

pu-242 94242 yes yes pu 

pu-243 94243 yes yes pu 
B-7 

Table B.1 Evaluations and IDs used for the ENDF/B-VI.8 library (continued) 

Nuclide ID Photon 

yield data 

Full range 

self-shielding 

factors 

Corresponding 

photon data 

pu-244 94244 yes pu 

am-241 95241 yes yes am 

am-242 95242 yes am 

am-242 95601 yes yes am 

am-243 95243 yes yes am 

cm-241 96241 yes cm 

cm-242 96242 yes yes cm 

cm-243 96243 yes cm 

cm-244 96244 yes yes cm 

cm-245 96245 yes cm 

cm-246 96246 yes cm 

cm-247 96247 yes yes cm 

cm-248 96248 yes yes cm 

bk-249 97249 yes bk 

cf-249 98249 yes cf 

cf-250 98250 yes yes cf 

cf-251 98251 yes yes cf 

cf-252 98252 yes yes cf 

cf-253 98253 yes cf 

es-253 99253 yes es 

h_h2o 1001 yes h 

para_h 1501 yes h 

ortho_h 1401 yes h 

h_zrh 1701 yes h 

d_d2o 1002 yes h 

para_d 1502 yes h 

ortho_d 1402 yes h 

be_metal 4309 yes yes be 

beo 4509 yes yes be 

graphite 6312 yes yes c 

l_ch4 1101 yes h 

s_ch4 1201 yes h 

h_ch2 1901 yes h 

benzine 1601 yes h 

zr_zrh 40701 zr 
B-8 
APPENDIX C 

EVALUATIONS INCLUDED IN ENDF/B-VII.0 LIBRARY 

Evaluations included in the multigroup library for ENDF/B-VII.0 are listed in Table C.1. 

Table C.1 Evaluations included in ENDF/B-VII.0 library

Nuclide ID Photon 

yield data 

Full range 

self-shielding 

factors 

Corresponding 

photon data 

h-1 1801 yes h 

h-2 1802 yes yes h 

h-3 1003 yes h 

he-3 2003 yes he 

he-4 2004 yes he 

li-6 3006 yes yes li 

li-7 3007 yes yes li 

be-7 4007 yes be 

be-9 4009 yes yes be 

b-10 5010 yes yes b 

b-11 5011 yes yes b 

c 6 000 yes yes c 

n-14 7014 yes yes n 

n-15 7015 yes yes n 

o-16 8016 yes yes o 

o-17 8017 yes o 

f-19 9019 yes yes f 

na-22 11022 yes na 

na-23 11023 yes yes na 

mg-24 12024 yes yes mg 

mg-25 12025 yes yes mg 

mg-26 12026 yes yes mg 

al-27 13027 yes yes al 

si-28 14028 yes yes si 

si-29 14029 yes yes si 

si-30 14030 yes yes si 

p-31 15031 yes yes p 

s-32 16032 yes yes s 

s-33 16033 yes yes s 

s-34 16034 yes yes s 

s-36 16036 yes yes s 

cl-35 17035 yes yes cl 

cl-37 17037 yes yes cl 

ar-36 18036 yes ar 

ar-38 18038 yes ar 

ar-40 18040 yes ar 

k-39 19039 yes yes k 

k-40 19040 yes yes k 

k-41 19041 yes yes k 

ca-40 20040 yes yes ca 

ca-42 20042 yes yes ca 

ca-43 20043 yes yes ca 

 C-1 
C-2 

Table C.1 Evaluations included in ENDF/B-VII.0 library (continued) 

Nuclide ID Photon 

yield data 

Full range 

self-shielding 

factors 

Corresponding 

photon data 

ca-44 20044 yes yes ca 

ca-46 20046 yes yes ca 

ca-48 20048 yes yes ca 

sc-45 21045 yes yes sc 

ti-46 22046 yes yes ti 

ti-47 22047 yes yes ti 

ti-48 22048 yes yes ti 

ti-49 22049 yes yes ti 

ti-50 22050 yes yes ti 

v 2 3000 yes no v 

cr-50 24050 yes yes cr 

cr-52 24052 yes yes cr 

cr-53 24053 yes yes cr 

cr-54 24054 yes yes cr 

mn-55 25055 yes yes mn 

fe-54 26054 yes yes fe 

fe-56 26056 yes yes fe 

fe-57 26057 yes yes fe 

fe-58 26058 yes yes fe 

co-58 27058 yes co 

co-58 27601 yes co 

co-59 27059 yes yes co 

ni-58 28058 yes yes ni 

ni-59 28059 yes yes ni 

ni-60 28060 yes yes ni 

ni-61 28061 yes yes ni 

ni-62 28062 yes yes ni 

ni-64 28064 yes yes ni 

cu-63 29063 yes yes cu 

cu-65 29065 yes yes cu 

zn 3 0000 no zn 

ga-69 31069 yes ga 

ga-71 31071 yes ga 

ge-70 32070 yes yes ge 

ge-72 32072 yes yes ge 

ge-73 32073 yes yes ge 

ge-74 32074 yes yes ge 

ge-76 32076 yes yes ge 

as-74 33074 yes yes as 

as-75 33075 yes yes as 

se-74 34074 yes se 

se-76 34076 yes se 

se-77 34077 yes se 

se-78 34078 yes se 

se-79 34079 yes se 

se-80 34080 yes se 

se-82 34082 yes se 

br-79 35079 yes br 

br-81 35081 yes br 

kr-78 36078 yes kr 
C-3 

Table C.1 Evaluations included in ENDF/B-VII.0 library (continued) 

Nuclide ID Photon 

yield data 

Full range 

self-shielding 

factors 

Corresponding 

photon data 

kr-80 36080 yes kr 

kr-82 36082 yes kr 

kr-83 36083 yes kr 

kr-84 36084 yes kr 

kr-85 36085 yes yes kr 

kr-86 36086 yes kr 

rb-85 37085 yes rb 

rb-86 37086 yes yes rb 

rb-87 37087 yes rb 

sr-84 38084 yes yes sr 

sr-86 38086 yes sr 

sr-87 38087 yes sr 

sr-88 38088 yes sr 

sr-89 38089 yes sr 

sr-90 38090 yes sr 

y-89 39089 yes yes y 

y-90 39090 yes yes y 

y-91 39091 yes y 

zr-90 40090 yes yes zr 

zr-91 40091 yes yes zr 

zr-92 40092 yes yes zr 

zr-93 40093 yes zr 

zr-94 40094 yes yes zr 

zr-95 40095 yes zr 

zr-96 40096 yes yes zr 

nb-93 41093 yes yes nb 

nb-94 41094 yes nb 

nb-95 41095 yes nb 

mo-92 42092 yes yes mo 

mo-94 42094 yes yes mo 

mo-95 42095 yes yes mo 

mo-96 42096 yes yes mo 

mo-97 42097 yes yes mo 

mo-98 42098 yes yes mo 

mo-99 42099 yes mo 

mo-100 42100 yes mo 

tc-99 43099 yes yes tc 

ru-96 44096 yes ru 

ru-98 44098 yes ru 

ru-99 44099 yes ru 

ru-100 44100 yes ru 

ru-101 44101 yes yes ru 

ru-102 44102 yes ru 

ru-103 44103 yes ru 

ru-104 44104 yes ru 

ru-105 44105 yes ru 

ru-106 44106 yes ru 

rh-103 45103 yes yes rh 

rh-105 45105 yes rh 

pd-102 46102 yes yes pd 

pd-104 46104 yes yes pd 
C-4 

Table C.1 Evaluations included in ENDF/B-VII.0 library (continued) 

Nuclide ID Photon 

yield data 

Full range 

self-shielding 

factors 

Corresponding 

photon data 

pd-105 46105 yes yes pd 

pd-106 46106 yes yes pd 

pd-107 46107 yes pd 

pd-108 46108 yes yes pd 

pd-110 46110 yes yes pd 

ag-107 47107 yes yes ag 

ag-109 47109 yes yes ag 

ag-110 47601 yes ag 

ag-111 47111 yes yes ag 

cd-106 48106 yes yes cd 

cd-108 48108 yes cd 

cd-110 48110 yes cd 

cd-111 48111 yes yes cd 

cd-112 48112 yes cd 

cd-113 48113 yes cd 

cd-114 48114 yes cd 

cd-115 48601 yes yes cd 

cd-116 48116 yes cd 

in-113 49113 yes in 

in-115 49115 yes in 

sn-112 50112 yes sn 

sn-113 50113 yes yes sn 

sn-114 50114 yes sn 

sn-115 50115 yes sn 

sn-116 50116 yes sn 

sn-117 50117 yes sn 

sn-118 50118 yes sn 

sn-119 50119 yes sn 

sn-120 50120 yes sn 

sn-122 50122 yes sn 

sn-123 50123 yes sn 

sn-124 50124 yes sn 

sn-125 50125 yes yes sn 

sn-126 50126 yes sn 

sb-121 51121 yes sb 

sb-123 51123 yes sb 

sb-124 51124 yes sb 

sb-125 51125 yes sb 

sb-126 51126 yes yes sb 

te-120 52120 yes te 

te-122 52122 yes te 

te-123 52123 yes te 

te-124 52124 yes te 

te-125 52125 yes te 

te-126 52126 yes te 

te-127 52601 yes te 

te-128 52128 yes te 

te-129 52611 yes te 

te-130 52130 yes te 

te-132 52132 yes yes te 

i-127 53127 yes yes i 
C-5 

Table C.1 Evaluations included in ENDF/B-VII.0 library (continued) 

Nuclide ID Photon 

yield data 

Full range 

self-shielding 

factors 

Corresponding 

photon data 

i-129 53129 yes i 

i-130 53130 yes yes i 

i-131 53131 yes i 

i-135 53135 yes i 

xe-123 54123 yes xe 

xe-124 54124 yes xe 

xe-126 54126 yes xe 

xe-128 54128 yes xe 

xe-129 54129 yes xe 

xe-130 54130 yes xe 

xe-131 54131 yes yes xe 

xe-132 54132 yes xe 

xe-133 54133 yes xe 

xe-134 54134 yes xe 

xe-135 54135 yes xe 

xe-136 54136 yes xe 

cs-133 55133 yes yes cs 

cs-134 55134 yes cs 

cs-135 55135 yes cs 

cs-136 55136 yes cs 

cs-137 55137 yes cs 

ba-130 56130 yes ba 

ba-132 56132 yes ba 

ba-133 56133 yes yes ba 

ba-134 56134 yes ba 

ba-135 56135 yes ba 

ba-136 56136 yes ba 

ba-137 56137 yes ba 

ba-138 56138 yes ba 

ba-140 56140 yes ba 

la-138 57138 yes la 

la-139 57139 yes la 

la-140 57140 yes yes la 

ce-136 58136 yes yes ce 

ce-138 58138 yes yes ce 

ce-139 58139 yes yes ce 

ce-140 58140 yes ce 

ce-141 58141 yes ce 

ce-142 58142 yes ce 

ce-143 58143 yes yes ce 

ce-144 58144 yes ce 

pr-141 59141 yes yes pr 

pr-142 59142 yes yes pr 

pr-143 59143 yes pr 

nd-142 60142 yes yes nd 

nd-143 60143 yes yes nd 

nd-144 60144 yes yes nd 

nd-145 60145 yes yes nd 

nd-146 60146 yes yes nd 

nd-147 60147 yes yes nd 

nd-148 60148 yes yes nd 
C-6 

Table C.1 Evaluations included in ENDF/B-VII.0 library (continued) 

Nuclide ID Photon 

yield data 

Full range 

self-shielding 

factors 

Corresponding 

photon data 

nd-150 60150 yes yes nd 

pm-147 61147 yes pm 

pm-148 61148 yes pm 

pm-148 61601 yes pm 

pm-149 61149 yes pm 

pm-151 61151 yes yes pm 

sm-144 62144 yes yes sm 

sm-147 62147 yes yes sm 

sm-148 62148 yes yes sm 

sm-149 62149 yes yes sm 

sm-150 62150 yes yes sm 

sm-151 62151 yes yes sm 

sm-152 62152 yes yes sm 

sm-153 62153 yes yes sm 

sm-154 62154 yes yes sm 

eu-151 63151 yes eu 

eu-152 63152 yes eu 

eu-153 63153 yes yes eu 

eu-154 63154 yes eu 

eu-155 63155 yes eu 

eu-156 63156 yes eu 

eu-157 63157 yes yes eu 

gd-152 64152 yes yes gd 

gd-153 64153 yes yes gd 

gd-154 64154 yes yes gd 

gd-155 64155 yes yes gd 

gd-156 64156 yes yes gd 

gd-157 64157 yes yes gd 

gd-158 64158 yes yes gd 

gd-160 64160 yes yes gd 

tb-159 65159 yes tb 

tb-160 65160 yes yes tb 

dy-156 66156 yes yes dy 

dy-158 66158 yes yes dy 

dy-160 66160 yes yes dy 

dy-161 66161 yes yes dy 

dy-162 66162 yes yes dy 

dy-163 66163 yes yes dy 

dy-164 66164 yes yes dy 

ho-165 67165 yes yes ho 

ho-166 67601 yes yes ho 

er-162 68162 yes yes er 

er-164 68164 yes yes er 

er-166 68166 yes yes er 

er-167 68167 yes yes er 

er-168 68168 yes yes er 

er-170 68170 yes yes er 

lu-175 71175 yes lu 

lu-176 71176 yes lu 

hf-174 72174 yes hf 

hf-176 72176 yes hf 
C-7 

Table C.1 Evaluations included in ENDF/B-VII.0 library (continued) 

Nuclide ID Photon 

yield data 

Full range 

self-shielding 

factors 

Corresponding 

photon data 

hf-177 72177 yes hf 

hf-178 72178 yes hf 

hf-179 72179 yes hf 

hf-180 72180 yes hf 

ta-181 73181 yes yes ta 

ta-182 73182 yes ta 

w-182 74182 yes yes w 

w-183 74183 yes yes w 

w-184 74184 yes yes w 

w-186 74186 yes yes w 

re-185 75185 yes re 

re-187 75187 yes re 

ir-191 77191 yes yes ir 

ir-193 77193 yes yes ir 

au-197 79197 yes yes au 

hg-196 80196 yes yes hg 

hg-198 80198 yes yes hg 

hg-199 80199 yes yes hg 

hg-200 80200 yes yes hg 

hg-201 80201 yes yes hg 

hg-202 80202 yes yes hg 

hg-204 80204 yes yes hg 

pb-204 82204 yes yes pb 

pb-206 82206 yes yes pb 

pb-207 82207 yes yes pb 

pb-208 82208 yes yes pb 

bi-209 83209 yes yes bi 

ra-223 88223 yes ra 

ra-224 88224 yes ra 

ra-225 88225 yes ra 

ra-226 88226 yes ra 

ac-225 89225 yes ac 

ac-226 89226 yes ac 

ac-227 89227 yes ac 

th-227 90227 yes th 

th-228 90228 yes th 

th-229 90229 yes th 

th-230 90230 yes th 

th-232 90232 yes yes th 

th-233 90233 yes th 

th-234 90234 yes th 

pa-231 91231 yes yes pa 

pa-232 91232 yes pa 

pa-233 91233 yes yes pa 

u-232 92232 yes yes u 

u-233 92233 yes yes u 

u-234 92234 yes yes u 

u-235 92235 yes yes u 

u-236 92236 yes yes u 

u-237 92237 yes yes u 

u-238 92238 yes yes u 
C-8 

Table C.1 Evaluations included in ENDF/B-VII.0 library (continued) 

Nuclide ID Photon 

yield data 

Full range 

self-shielding 

factors 

Corresponding 

photon data 

u-239 92239 yes yes u 

u-240 92240 yes yes u 

u-241 92241 yes yes u 

np-235 93235 yes np 

np-236 93236 yes np 

np-237 93237 yes yes np 

np-238 93238 yes np 

np-239 93239 yes np 

pu-236 94236 yes pu 

pu-237 94237 yes pu 

pu-238 94238 yes pu 

pu-239 94239 yes yes pu 

pu-240 94240 yes yes pu 

pu-241 94241 yes yes pu 

pu-242 94242 yes yes pu 

pu-243 94243 yes yes pu 

pu-244 94244 yes pu 

pu-246 94246 yes pu 

am-241 95241 yes yes am 

am-242 95242 yes am 

am-242 95601 yes am 

am-243 95243 yes yes am 

am-244 95244 yes am 

am-244 95611 yes am 

cm-241 96241 yes cm 

cm-242 96242 yes yes cm 

cm-243 96243 yes cm 

cm-244 96244 yes cm 

cm-245 96245 yes cm 

cm-246 96246 yes cm 

cm-247 96247 yes cm 

cm-248 96248 yes yes cm 

cm-249 96249 yes cm 

cm-250 96250 yes cm 

bk-249 97249 yes bk 

bk-250 97250 yes bk 

cf-249 98249 yes cf 

cf-250 98250 yes yes cf 

cf-251 98251 yes yes cf 

cf-252 98252 yes yes cf 

cf-253 98253 yes cf 

cf-254 98254 yes cf 

es-253 99253 yes es 

es-254 99254 yes es 

es-255 99255 yes es 

fm-255 100255 yes fm 

al27 13701 yes yes al 

fe56 26701 yes yes fe 

be_metal 4309 yes yes be 

be_beo 4509 yes yes be 

d_d2o 1002 yes h 
C-9 

Table C.1 Evaluations included in ENDF/B-VII.0 library (continued) 

Nuclide ID Photon 

yield data 

Full range 

self-shielding 

factors 

Corresponding 

photon data 

h_ch2 1901 yes h 

h_h2o 1001 yes h 

h_zrh 1701 yes h 

o_beo 8516 yes yes o 

zr_zrh 40701 yes yes zr 

benzine 1601 yes h 

graphite 6312 yes yes c 

l_ch4 1101 yes h 

ortho_d 1402 yes h 

ortho_h 1401 yes h 

para_d 1502 yes h 

para_h 1501 yes h 

s_ch4 1201 yes h 
NUREG/ 

ORNL/TM-2008/047