The JEFF-3.1 Nuclear Data Library

Data Bank ISBN 92-64-02314-3

JEFF Report 21

Edited by Arjan Koning Robin Forrest Mark Kellett Robert Mills Hans Henriksson Yolanda Rugama

NUCLEAR ENERGY AGENCY ORGANISATION FOR ECONOMIC CO-OPERATION AND DEVELOPMENT

ORGANISATION FOR ECONOMIC CO-OPERATION AND DEVELOPMENT

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NUCLEAR ENERGY AGENCY

The OECD Nuclear Energy Agency (NEA) was established on 1st February 1958 under the name of the OEEC European Nuclear Energy Agency. It received its present designation on 20th April 1972, when Japan became its first non-European full member. NEA membership today consists of 28 OECD member countries: Australia, Austria, Belgium, Canada, the Czech Republic, Denmark, Finland, France, Germany, Greece, Hungary, Iceland, Ireland, Italy, Japan, Luxembourg, Mexico, the Netherlands, Norway, Portugal, Republic of Korea, the Slovak Republic, Spain, Sweden, Switzerland, Turkey, the United Kingdom and the United States. The Commission of the European Communities also takes part in the work of the Agency. The mission of the NEA is: to assist its member countries in maintaining and further developing, through international co-operation, the scientific, technological and legal bases required for a safe, environmentally friendly and economical use of nuclear energy for peaceful purposes, as well as to provide authoritative assessments and to forge common understandings on key issues, as input to government decisions on nuclear energy policy and to broader OECD policy analyses in areas such as energy and sustainable development. Specific areas of competence of the NEA include safety and regulation of nuclear activities, radioactive waste management, radiological protection, nuclear science, economic and technical analyses of the nuclear fuel cycle, nuclear law and liability, and public information. The NEA Data Bank provides nuclear data and computer program services for participating countries. In these and related tasks, the NEA works in close collaboration with the International Atomic Energy Agency in Vienna, with which it has a Co-operation Agreement, as well as with other international organisations in the nuclear field.

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FOREWORD

The Joint Evaluated Fission and Fusion (JEFF) Project is a collaborative effort among the member countries of the NEA Data Bank to develop a reference nuclear data library. The JEFF library contains sets of evaluated nuclear data, mainly for fission and fusion applications; it contains a number of different data types, including neutron and proton interaction data, radioactive decay data, fission yield data, thermal scattering law data and photo-atomic interaction data.

The latest version of the JEFF library, JEFF-3.1, was released by the NEA in May 2005. JEFF-3.1 combines the efforts of the JEFF and EFF/EAF (European Fusion File/European Activation File) working groups who have contributed to this combined fission and fusion library. The neutron general purpose library contains incident neutron data for 381 materials from 1H to 255Fm. The activation library (based on the European Activation File, EAF-2003) contains 774 different targets from 1H to 257Fm. The radioactive decay data library contains data for 3 852 isotopes, of which 226 are stable. The proton special purpose library contains incident proton data for 26 materials from 40Ca to 209Bi. The thermal scattering law library covers 9 materials, and the fission yield library covers 19 isotopes of neutron induced fission yield from 232Th to 245Cm and 3 isotopes with spontaneous fission yields (242Cm, 244Cm and 252Cf).

Acknowledgements

We acknowledge the support and dedication of many colleagues in the nuclear data community in Europe, Japan and the United States without whom this work would not have been possible, in particular: C. Chabert, J-P. Delaroche, H. Derrien, H. Duarte, B. Duchemin, I. Duhamel, A. Filatenkov, E. Fort, F. Froehner, J. Galy, C-S. Gil, W. Haeck, F-J. Hambsch, M. Herman, S. Hilaire, M. Honusek, T. Ivanova, D-H. Kim, N. Larson, P. de Leege, R. MacFarlane, A. Mengoni, L. Mercatali, C. Mihaelescu, W. Mannhart, M.C. Moxon, S. Pelloni, P. Ribon, D. Ridikas, G. Rimpault, J.L. Rowlands, E Simeckova, J. Tommasi, B. Zefran, M. Zmitko, and Harm Gruppelaar, who left us too soon to experience the impact of his work.

TABLE OF CONTENTS

Foreword ...... 3

Introduction ...... 7

The general-purpose neutron data library...... 8 The need for a new library...... 8 New evaluations...... 8

Thermal scattering law library...... 8

Expected performance relative to JEF-2.2 ...... 9

Benchmarking and quality assurance ...... 9 Present status ...... 9 The library format and contents...... 10 Outlook ...... 11

Activation library, JEFF-3.1/A...... 11 Summary...... 11 Background...... 11 Library description...... 11

Radioactive decay data library, JEFF-3.1/RDD ...... 12 Summary...... 12 Data sources available ...... 12

Fission yield libraries, JEFF-3.1/NFY and SFY...... 13

Proton special-purpose library...... 13

The JEFF Team ...... 15

References ...... 17

Working papers from JEFF meetings 2002-2005...... 19 JEFF documents...... 19 EFF documents ...... 26

5 Tables ...... 33 Table 1. Origin of evaluations in the JEFF-3.1 general-purpose neutron data library ...... 35 Table 2. Origin of evaluations in the JEFF-3.1 thermal scattering law library...... 48 Table 3. Contents of the JEFF-3.1 general-purpose neutron data library. List of available reactions (MTs) in files MF=3 to MF=6 ...... 49 Table 4. Contents of the JEFF-3.1 general-purpose neutron data library. List of available reactions (MTs) in photon production files MF=12 to MF=15...... 65 Table 5. Contents of the JEFF-3.1 general-purpose neutron data library. List of covariance data (MF=33to MF=36)...... 69 Table 6. Resonance data in the JEFF-3.1 general-purpose neutron data library ...... 71 Table 7. Contents of the JEFF-3.1 thermal scattering law library...... 84 Table 8. Contents of the JEFF-3.1 proton special-purpose library. List of available reactions (MTs) in files MF=3 to MF=6...... 85 Table 9. Simple integral neutron cross-section data from JEFF-3.1 calculated with the INTER computer code...... 86

Available on CD-ROM

Appendix 1. Selection of data for the JEFF-3.1 general-purpose neutron data file

Appendix 2. Performance reports on the JEFF-3.1 general-purpose neutron data library

Eeff calculations using JEFF-3.1 nuclear data Shielding benchmark calculations with MCNP-4c3 using JEFF-3.1 nuclear data Criticality safety benchmark calculations with MCNP-4c3 using JEFF-3.1 nuclear data

Appendix 3. List of definitions to the ENDF-6 file format for JEFF-3.1 Table A3.1. Definitions of file types (MF) Table A3.2. Definitions of reaction types (MT)

6 Introduction

The objective of the OECD Nuclear Energy Agency (NEA) co-ordinated Joint Evaluated Fission and Fusion (JEFF) data project is to develop and promote the use of high quality evaluated nuclear data sets in standard formats for a wide range of scientific and technical applications. The JEFF project assesses the needs for nuclear data improvements and brings together experts in different areas such as experiments, data evaluations, verification and compilation of the data under strict quality assurance procedures, file processing and benchmarking.

JEFF is a collaborative effort among NEA Data Bank member countries. The project maintains close links with other similar international efforts and projects aimed at producing evaluated nuclear data, for example through active participation in the NEA Working Party on International Nuclear Data Evaluation Co-operation (WPEC). While the objective of the JEF-2.2 library (1992) was to achieve improved performance for existing reactors and fuel cycles, its successor, the JEFF-3 project, aims at providing users with a more extensive set of data for a wider range of applications. While existing reactors and fuel cycles remain the essential application areas of the nuclear data library, innovative reactor concepts (Gen-IV systems), transmutation of radioactive waste, fusion, medical applications, and various non-energy related industrial applications are now also envisaged as scientific application areas that will make use of the JEFF data. The European Fusion File (EFF) also contributes to the JEFF project through evaluations and data validation.

Figure 1 illustrates the evolution of the JEFF project, including the latest release of JEFF-3.1. The JEFF-3.1 release consists of the following libraries:

x neutron general-purpose library;

x thermal scattering library;

x neutron activation library;

x decay data library;

x fission yield data library;

x proton special-purpose library.

Figure 1: The evolution of the JEFF project

7 The general-purpose neutron data library

The need for a new library

Following the release of the JEFF-3.0 library in April 2002, various benchmark tests have confirmed the expected performance improvements of this general-purpose neutron data library over JEF-2.2. However, they have also shown that the reactivity of small low-enriched uranium systems, relevant to LWR analyses, is underestimated by about 500 pcm when using JEFF-3.0 data. The 238U inelastic, scattering and capture cross-sections have been identified as the possible reasons for this 238 underestimation. A new U evaluation has been assembled, resulting in appreciably improved keff values. Subgroup 22 of WPEC has contributed significantly to these improvements. Very extensive automated benchmark tests have been conducted, probing the current quality of the new data library. These benchmark tests include for example criticality calculations performed with MCNP, TRIPOLI and APOLLO for an unprecedented set of benchmark cases. Current-day computer power enables future revisions of the JEFF library to be quickly tested with such benchmarking schemes.

It was decided to adopt an eight-time group representation for delayed neutron data, as suggested by Subgroup 6 of WPEC. It is considered that this representation has two advantages – the longest lived, dominant precursors are explicitly represented and the calculation convenience of having the same set of time constants for all fissioning isotopes.

A final justification for releasing JEFF-3.1 is that there was a desire from the nuclear industry (in particular in France) to start validating the new JEFF library in the fall of 2005.

New evaluations

Another reason to develop the JEFF-3.1 general-purpose neutron data library is that a significant number of isotopes have undergone a major revision over the last few years. This holds both for the general-purpose and the special-purpose libraries. This report contains a table with the justification for the choice of the various isotopes for JEFF-3.1 (see Appendix 1). New evaluations are included for the Ti isotopes (IRK Vienna), the Ca, Sc, Fe, Ge, Pb and Bi isotopes (NRG Petten), 103Rh, 127I, 129I, the Hf isotopes, 236-238U, and 241Am (CEA). For other isotopes, more recent evaluations from other libraries were adopted.

Revised thermal scattering data have been produced for all important moderator and structural materials and this was included in JEFF-3.1.

Various experimental nuclear data activities have taken place during the development of JEFF-3.1. The programme at the nTOF facility at CERN comprised capture and fission cross-section measurements including isotopes of relevance to the thorium fuel cycle and several transuranic isotopes. The measurement programme at IRMM Geel mainly covered neutron data related to waste transmutation and other innovative concepts, as well as basic nuclear physics and standards data.

Thermal scattering law library

The thermal scattering law library contains the following nine evaluations: hydrogen bound in 9 24 water, zirconium, polyethylene (CH2) and CaH2, deuterium bound on D2O, Be, graphite, Mg and 9 finally calcium bound in CaH2. All files are new evaluations, except Be and hydrogen in polyethylene, which are from the JEFF-3.0 library. Many of the evaluations are the result of an IAEA co-ordinated project on thermal neutron scattering. Calculations for a variety of temperatures were made with the

8 LEAPR module of NJOY to obtain thermal scattering data that are accurate over a wider range of energy and momentum transfer. To validate these files on a microscopic level, detailed comparisons with a significant number of measurements of differential and integral neutron cross-sections and other relevant data have been performed. The current models used for these files are able to describe the experimental data reasonably well. The generating and processing chain for the thermal neutron scattering files with NJOY was carefully investigated.

Expected performance relative to JEF-2.2

In the coming years, benchmarking studies will provide quantitative evidence of the JEFF-3.1 library performance with respect to the previous file versions, in particular JEF-2.2. The latter is a “natural” reference, as it has benefited from many years of validation and users’ feedback (see JEF Report 17). However, it is already possible to give some indications or trends concerning the JEFF-3.1 expected impact on fission reactor applications.

For LWR applications, the longstanding JEF-2.2 problem of 236U under-prediction in actinide inventory calculations has been corrected (increase of the 235U epithermal capture cross-section). Similarly, the underestimation of the 242Pu content and correlated 243Am build-up in spent fuel has been fixed (increase of 241Pu capture and decrease of fission near the 0.3 eV resonance). Reactivity predictions for fresh UOX-fuelled cores should be slightly lower than with JEF-2.2, in agreement with measurements. The reactivity of systems containing significant quantities of 27Al will also be improved (higher thermal capture). For MOX cores, however, JEFF-3.1 will yield larger values than JEF-2.2, thus over-predicting reactivity. Individual fission product poisoning effects should be better predicted with JEFF-3.1 as a result of improved capture cross-sections (103Rh, 133Cs, 95Mo, 143Nd, 154Eu, 155Eu, 149Sm). Important thermal fission product yields used for fission rate normalisation have been revised for a better agreement with measurements (increased 235U yield of 137Cs, decrease 239Pu yield of 148Nd).

For fast reactor applications, the reactivity of un-irradiated uranium systems should no longer be overestimated (larger 235U capture). On the other hand, the reactivity of plutonium-fuelled fast reactors should still be underestimated, being even lower than with JEF-2.2 as a result of the changes in the 238U inelastic scattering. 241Pu build-up should be improved (lower 240Pu capture cross-section above 100 eV), as well as the 238Pu, 242Pu, 242Cm and 242mAm produced from 241Am (branching ratio of 0.85, lower capture cross-section above 20 keV). Calculated sodium-void effects should no longer require a posteriori corrections (revised 23Na inelastic and elastic cross-sections). Finally, steel reflector effects should also be better predicted (improved 56Fe and 52Cr scattering cross-sections).

Compared with JEF-2.2, a better agreement between JEFF-3.1-based calculations and measurements is expected in neutron transmission applications containing the following materials: Fe, 16O, 9Be, W, Na.

Initial testing of the JEFF-3.1 file has also indicated that further improvements will be needed, in particular for 237Np capture, 239Pu fission and capture, 242Pu capture, 241Am capture, 56Fe scattering, fission product cross-sections (fast range), decay data and yields of short-lived fission products.

Benchmarking and quality assurance

Present status

Extensive benchmarking of the JEFF-3.1 data library is now under way for both the general-purpose and the special-purpose sub-libraries. In order to avoid further delays in the publication of the present

9 report, only results from the extensive benchmarking effort on criticality, effective delayed neutron parameters, and shielding using a Monte Carlo approach (MCNP) are reported on here (see Appendix 2). At a later stage, a wider validation of benchmark results using different methods (deterministic and Monte Carlo), codes and for various integral quantities, can be expected.

A quality assurance procedure has been developed at NEA for the assembly, basic testing and loading of the JEFF-3.1 library into a database. Verifications performed at the loading stage include format checks (CHECKR), physics checks (FIZCON and PSYCHE), processing using different versions of NJOY and graphical comparisons with other data libraries. Specific tools were developed within this framework (e.g. the JANIS data display program and database loading programs) and existing programs (e.g. FIZCON, INTER) were extended to enable more stringent consistency checks.

The library format and contents

Each material in the JEFF-3.1 library is given in the ENDF-6 format. This format stores nuclear data in various sections or files (MF). Each file contains data for different reactions (MT). Appendix 3 gives the list of MF and MT numbers and their meaning in the ENDF-6 format.

Table 1 provides the list of materials in the neutron general-purpose library and the origin of the file. A material corresponds to a nucleus of a single isotope or to a nucleus of a natural element. A nucleus is either in its ground state or in an isomeric state. In the first column of Table 1, the material is represented as ZZZ-SY-AAA where AAA and ZZZ are the atomic charge and mass given as three-digit integers. SY is the symbol of the material represented with two characters. Note that AAA is equal to zero when the material represents a natural element. In addition, when the nucleus is in its first isomeric state, the letter “M” is appended to the end of the column (e.g. 95-Am-242M). The materials are classified in ascending atomic charge order. The second column provides the material number used to identify the isotope or the element in the library. The third column gives the source library from where the data originates. The forth to sixth columns specify the laboratory that performed the evaluation, the date of the evaluation and the authors.

Table 2 provides the same information as Table 1, for neutron thermal scattering data in JEFF-3.1. The following compounds are included: hydrogen bound in water, zirconium, polyethylene (CH2) and 9 24 CaH2, deuterium bound on D2O, Be, graphite, Mg and finally calcium bound in CaH2.

Table 3 provides some details concerning the available cross-section data in the general-purpose neutron data library. The first two columns identify the material, and the third column lists the MTs found in file 1 (MF=1, general information). Reactions (MTs) for which cross-section data are available (MF=3, reaction cross-sections) are listed in the forth column. Columns 5 through 7 provide the list of reactions for which angular distributions, energy distributions or energy-angle distributions are given in MF=4, 5 and 6 respectively. Table 4 gives the list of materials for which photon production data is available (MF = 12 to 15) in the general-purpose neutron data library. Table 5 provides information on available covariance data in MF=31 to 36 in the general-purpose neutron data library, and Table 6 gives details on resonance data given in MF=2. The third column of Table 6 states whether resonance data is present, and the forth column gives the formalism used. The remaining columns give details on the resonance parameters in the resolved resonance range (RRR) and the unresolved resonance range (URR) with the energy of the first three s-wave resonances.

Table 7 lists the contents of MF=1 and MF=7 of the thermal scattering law library. Table 8 lists similarly the contents of the proton special purpose library, with information from MF=1, 3 and 6.

10 Table 9, finally, gives simple integral neutron cross-section data from the neutron general-purpose library of JEFF-3.1 calculated with the INTER-7.0 computer code. All input parameters are listed as well.

Outlook

For the next release, JEFF-3.2, we foresee more efforts on fission product evaluations, minor actinide evaluations, and inclusion of more covariance data in the files.

Activation library, JEFF-3.1/A

Summary

The activation data library in ENDF-6 format, JEFF-3.1/A, is based on the European Activation File (EAF-2003). The JEFF-3.1/A library contains 12 617 excitation functions involving 774 different targets from 1H to 257Fm, atomic numbers Z=1 to Z=100, in the energy range 10–5 eV to 20 MeV. Uniquely, an uncertainty file is also provided that quantifies the degree of confidence placed in the data for each reaction channel.

Background

Calculations of the activation of materials, arising from the operation of fusion devices producing large fluxes of neutrons from the D-T reaction, play an important role in fusion technology studies. In order to have confidence in the results of such calculations, it is necessary that the inventory code and the data libraries be thoroughly validated. This is accomplished by comparing the predictions of the code system with activation measurements made on materials relevant to fusion technology in well-characterised neutron fields.

The European Activation File (EAF) project has been an ongoing process performed through European and world-wide co-operation, leading to the creation of successive EAF versions. The EAF-2003 release [1] has benefited from the generation and maintenance of comprehensive activation files and the development of the new processing code SAFEPAQ-II [2]. Therefore, the JEFF project has adopted this version of the activation files for the JEFF-3.1/A library.

Library description

The EAF nuclear data library has been developed as part of the European Fusion Development Agreement (EFDA) Fusion Technology Programme. A series of measurements on fusion-relevant materials in several complementary neutron fields have been carried out over the last few years. In addition, analyses of measurements, carried out outside Europe and outside the fusion programme, have been undertaken.

Cross-section validation exercises, performed against both experimental data and systematics, enable a comprehensive assessment of the data. The SAFEPAQ-II software is used to apply a series of modifications to the original source data. A very important set of modifications concerns re-normalisation and branching using experimental or systematic data. A total of 3 225 reactions (26% of all the reactions) have thus been changed. This is a challenging task as the source contains non-threshold reactions with an energy-dependent branching ratio.

11 Validation using integral data has been performed by means of direct comparison with measurements of sample materials under fusion-relevant neutron spectra. Irradiations have been carried out at ENEA FNG, FZK Isochron-cyclotron, Sergiev Posad SNEG-13 and JAERI FNS and integral C/E comparisons made (C/E is the ratio of the library value to the experimental value). The results of these benchmark exercises, in concurrence with differential data, have indicated where modifications to the data should be applied.

The results from the experiments used to validate the activation data are presented in a series of tables in Ref. [3], including the pathways responsible for production of the various radionuclides. In cases where a single pathway dominates, it is possible to extract the effective cross-section for the reaction and compare this with the library value calculated in the same neutron spectrum. About 290 reactions from the JEFF-3.1/A library are plotted in the reference. The graphs include uncertainty estimates as well as experimental differential data from the EXFOR database. C/E graphs showing the available integral data are also given and, using these graphs, a statement is made indicating whether the EAF data are consistent with the experimental data or whether changes to data would be needed. It should be noted that most of the measurements were carried out with fusion-relevant materials, resulting in small uncertainties for activities produced in major nuclides, but larger uncertainties for activities produced in minor nuclides or impurities.

Radioactive decay data library, JEFF-3.1/RDD

Summary

Radioactive decay data forms an integral part of the nuclear data requirements for nuclear applications. The JEFF-3.1 decay data library contains 3 852 nuclides from neutron to 272Rg. The decay data sub-group of the JEFF Project decided that a completely new start should be made in the construction of this library compared to the previous versions. The JEFF-3.1/RDD library is described in detail in JEFF Report 20 [4]. Some background information concerning the evolution of the project is provided below.

Data sources available

The basis for the selection of isotopes were taken from the NUBASE-2003 file [5], which contains basic nuclear properties, hence all known isomers were identified from here. Three evaluated libraries, from within Europe, were used to replace individual NUBASE files, for almost 600 nuclei in all. Two libraries originated from the UK (UKPADD-6.4 and UKHEDD-2.4) [7] and the third from the Decay Data Evaluation Project (DDEP) [8] conducted by the Laboratoire National Henri Bequerel (LNHB) at Saclay, France.

Files were also selected from the ENSDF library [9] – these files contain decay data derived from basic nuclear structure data. Approximately 900 nuclei, containing sufficient data to allow calculation of their energy balance to a consistency of better than 1%, were selected from ENSDF.

The JEFF-3.1 library stores all data in the internationally-accepted ENDF-6 format [6]. Recent enhancements, allowing the storage of basic nuclear properties for stable nuclei (i.e. spin and parity), have been adopted in this library. The isotopic abundances for all appropriate nuclei (taken from NUBASE-2003) have been added as the last line in the comments section (MF=1, MT=451) whilst an unused field in the format is investigated.

12 Throughout the compilation process the library has followed NEA quality assurance procedures and been checked with the standard ENDF utility codes STANEF, CHECKR and FIZCON [10]. Significant enhancements to the FIZCON code have been added for the verification of decay data [11] and minor corrections to the latest versions (7.00/7.01) have also been implemented following problems noted in checking this JEFF library, most notably the checking of the symbol field for meta-stable states and the inclusion of stable nuclei. Users should be warned that earlier versions of the checking codes, e.g. version 6.11, should not be used in testing this library.

Fission yield libraries, JEFF-3.1/NFY and SFY

The fission product yield libraries include independent and cumulative yields for neutron-induced fission of 232Th, 233,234,235,236,238U, 237,238Np, 238,239,240,241,242Pu, 241,242m,243Am, 243,244,245Cm and spontaneous fission of 242,244Cm and 252Cf. The data represent a development of the UKFY3 file described in Ref. [12], which was itself developed from earlier UK evaluations described in Refs. [13-15].

The principle changes from earlier libraries include an extended experimental database, calculation of the cumulative yields using the JEFF-3.1 decay data, improved calculation of uncertainties in the yields, improved adjustment to the physical constraints and the inclusion of new ternary yield data for 3H and 4He from Serot, Wagemans, et al. [16].

Proton special-purpose library

The proton special purpose library consists of 26 evaluated isotopes. The data are based primarily on a theoretical analysis with the nuclear model code TALYS [17]. The nuclear model parameters of TALYS have been adjusted to reproduce the existing experimental data. For several materials that are of key importance to transmutation programs (ADS), valuable experimental data was provided by the EU FP5 HINDAS project. Together, this results in data files that provide a complete representation of nuclear data needed for transport, damage, heating, radioactivity and shielding applications over the incident proton energy range from 1 to 200 MeV. This collection of isotopic evaluations is created by running TALYS with input parameters that do not, or only slightly, deviate from the default values.

These isotopic evaluations are thus of comparable quality. The same set of nuclear models is used and, equally important, the same ENDF-6 formatting procedures for each isotope. These data files are complete in their description of reaction channels, and use a compact method to store the data.

All transport data for particles, photons and residual nuclides are filed using a combination of MF=1, MF=3 and MF=6. This includes cross-sections, angular distributions, double-differential spectra, photon production cross-sections, and residual production (activation) cross-sections. These evaluations can thus be used as both transport and activation libraries.

The files have been checked by the BNL checking codes CHECKR, FIZCON and PSYCHE and have been successfully processed into an MCNP library by the processing code NJOY.

THE JEFF TEAM

The following scientists and institutes have contributed to the successful release of the JEFF-3.1 nuclear data libraries.

Chairs O. Bersillon (Decay Data), R.A. Forrest (EFF) and A.J. Koning (JEFF)

Evaluation O. Bouland (CEA), A. Courcelle (CEA), M.C. Duijvestijn (NRG), E. Dupont (CEA), J. Kopecky (JUKO), D. Leichtle (FZK), F. Marie (CEA), M. Mattes (Stutt), E. Menapace (ENEA), B. Morillon (CEA), C. Mounier (CEA), G. Noguerre (CEA), P. Pereslavtsev (FZK), P. Romain (CEA), O. Serot (CEA), S. Simakov (FZK), S. Tagesen (IIK), H. Vonach (IIK)

Experimental nuclear data P. Batistoni (ENEA), P. Bem (Rez), F. Gunsing (CEA), M. Pillon (ENEA), A. Plompen (IRMM), P. Rullhusen (IRMM), K. Seidel (TUD)

Nuclear modelling and theory M. Avrigeanu (NIP), V, Avrigeanu (NIP), E. Bauge (CEA), H. Leeb (TUW), M.J. Lopez Jimenez (CEA)

Processing, benchmarking and validation D. Bernard (CEA), A. Bidaud (CEN), R. Dagan (FZK), C. Dean (Serco), P. Dos-Santos-Uzarralde (CEA), U. Fischer (FZK), A. Hogenbirk (NRG), R. Jacqmin (CEA), C. Jouanne (CEA), I. Kodeli (IAEA), J. Leppanen (VTT), S.C. van der Marck (NRG), R. Perel (RIP), R. Perry (Serco), M. Pescarini (ENEA), A. Santamarina (CEA), J-C. Sublet (CEA), A. Trkov (IAEA)

Decay data and fission yields M.M. Be (CEA), T.D. Huynh (CEA), M.A. Kellett (IAEA), R. Mills (Nexia), A. Nichols (IAEA)

NEA Data Bank H. Henriksson, C. Nordborg, A. Nouri, Y. Rugama, E. Sartori

Participating institutions Technische Universitaet Wien (TUW), Vienna, Austria Institut fuer Isotopenforschung und Kernphysik (IIK), Vienna, Austria Nuclear Physics Institute Rez (Rez), Czech Republic VTT Processes (VTT), Finland Commisariat à l’Énergie Atomique (CEA) Bruyères-le-Chatel, France Commisariat à l’Énergie Atomique (CEA) Cadarache, France Commisariat à l’Énergie Atomique (CEA) Saclay, France CEN de Bordeaux (CEN), France Forschungszentrum Karlsruhe (FZK), Germany Stuttgart University (Stutt), Germany Technische Universitaet Dresden (TUD), Germany Racah Institute of Physics (RIP), Jerusalem, Israel National Agency for New Technology, Energy and the Environment (ENEA), Italy Nuclear Research and Consultancy Group (NRG), Petten, the Netherlands JUKO Research (JUKO), Alkmaar, the Netherlands National Institute of Physics and Nuclear Engineering (NIP), Bucharest, Romania UK Atomic Energy Authority Fusion (UKAEA), Abingdon, United Kingdom Nexia Solutions Ltd., Seascale (Nexia), United Kingdom Serco Assurance (Serco), Winfrith, United Kingdom

International organisations Institute for Reference Materials and Measurements (JRC/IRMM) International Atomic Energy Agency (IAEA) Nuclear Energy Agency (OECD/NEA) Data Bank

16 REFERENCES

[1] Forrest, R.A., J. Kopecky, J-Ch. Sublet, The European Activation File: EAF-2003 Cross-section Library, UKAEA Report UKAEA FUS 486, Dec. 2002.

[2] Forrest, R.A., SAFEPAQ-II User Manual, UKAEA report UKAEA FUS 454, Issue 6, Jan. 2005.

[3] Forrest R.A., M. Pillon, U. von Möllendorff, K. Seidel, J. Kopecky, J-Ch. Sublet, Validation of EASY-2003 Using Integral Measurements, UKAEA report UKAEA FUS 500, Dec. 2003.

[4] JEFF Report 20, The JEFF-3.1 Decay Data and Fission Yields Libraries, OECD Nuclear Energy Agency, forthcoming.

[5] Audi, G., O. Bersillon, J. Blachot, A.H. Wapstra, “The NUBASE Evaluation of Nuclear and Decay Properties”, Nucl. Phys. A, 729, 3-128 (2003).

[6] McLane, V., ENDF-102: Data Formats and Procedures for the Evaluated Nuclear Data File ENDF-6, BNL-NCS-44945-01/04-Rev., April 2001.

[7] Nichols, A., UKHEDD-2.4 and UKPADD-6.4.

[8] Bé, M.M., E. Browne, V. Chechev, V. Chisté, R. Dersch, C. Dulieu, R.G. Helmer, N. Kuzmenco, A.L. Nichols, E. Schönfeld, Nucléide, Table de Radionucléides sur CD-Rom, Version 2-2004, http://www.nucleide.org/NucData.htm.

[9] Tuli, J.K., The Evaluated Nuclear Structure Data file (ENSDF), BNL-NCS-51655 Rev. 87, April 1987.

[10] Dunford, C., ENDF Utility Checking Codes, 7.01, Brookhaven National Laboratory, Upton, NY, USA.

[11] Perry, R., C.J. Dean, FIZCON Enhancements, UKNSF report.

[12] Mills, R.W., Fission Product Yield Evaluation, Thesis, University of Birmingham, UK (1995).

[13] James, M.F., R.W. Mills, D.R. Weaver, UKFY2 Part 1: Methods and Outline, AEAT report, AEA-TRS-1015 (1991).

[14] James, M.F., R.W. Mills, D.R. Weaver, UKFY2 Part 2: Tables of Measured Data, AEAT report, AEA-TRS-1018 (1991).

[15] James, M.F., R.W. Mills, D.R. Weaver, UKFY2 Part 3: Tables of Discrepant Data, AEAT report, AEA-TRS-1019 (1991).

17 [16] Serot, O., C. Wagemans, Jan Heyse “New Results on Helium and Tritium Gas Production From Ternary Fission”, Proceedings of the International Conference on Nuclear Data for Science and Technology, R.C. Haight, M.B. Chadwick, T. Kawano, P. Talou (eds.), American Institute of Physics, ISBN 0-7354-0254-X/05, CP769, 857 (2005).

[17] Koning, A.J., S. Hilaire, M.C. Duijvestijn, “TALYS: Comprehensive Nuclear Reaction Modeling”, Proceedings of the International Conference on Nuclear Data for Science and Technology, R.C. Haight, M.B. Chadwick, T. Kawano, P. Talou (eds.), American Institute of Physics, ISBN 0-7354-0254-X/05, CP769, 857 (2005).

18 WORKING PAPERS FROM JEFF MEETINGS 2002-2005

The papers listed below were presented at the JEFF working group meetings from April 2002 until November 2005. The papers are for internal use within the JEFF group, and can therefore only be requested from the NEA and distributed after permission from the authors. The papers are considered as non-official publications having draft status.

JEFF documents

JEF/DOC-1125 ALEPH-DLG 1.1.0 Creating Cross-section Libraries for MCNP(X) and ALEPH, W. Haeck, B. Verboomen JEF/DOC-1124 Thermal Neutron Scattering Data at Low Temperatures, M. Mattes JEF/DOC-1123 Sensitivities of keff Calculated with Monte Carlo Methods: Theory and First Results, R. Perel JEF/DOC-1122 Qualification of JEFF-3.1 Evaluations: LWR Reactivity and Fuel Inventory Prediction, D. Bernard JEF/DOC-1121 Assessing Uncertainties in the Continuum Region with the Backward-forward Monte Carlo Method: Example n+89Y Evaluation, E. Bauge JEF/DOC-1120 Documentation for the JEFF-3.1 Decay Data Library, M. Kellett JEF/DOC-1119 Recently Launched IAEA Nuclear Data Section Projects, M. Kellett JEF/DOC-1118 TRIPOLI-4.4 Benchmarks, C. Jouanne JEF/DOC-1117 Photonuclear Data Evaluation of Uranium-235, I. Raskinyte JEF/DOC-1116 Experimental Activities at n_TOF CERN, F. Gunsing JEF/DOC-1115 Fission Products for BUC Applications, L. Leal JEF/DOC-1114 Present Status of JEFF-3.1 Benchmarking with ERANOS-2.0, J. Tommasi JEF/DOC-1113 Status of 238U Resolved and Unresolved Evaluation and Impact of JEFF-3.1 in Burn-up Calculation, A. Courcelle JEF/DOC-1112 TRIPOLI-4.4 – JEFF-3.1 & ENDB/B-VII b1 with the ICSBEP Criticality Models, J-Ch. Sublet, C. Jouanne, Y. Peneliau, F-X. Hugot JEF/DOC-1111 ECCOLIB-JEFF-3.1 Libraries, J-Ch. Sublet, C.J. Dean, D. Plisson-Rieunier JEF/DOC-1110 Analysis of New Data for Hydrogen Bound in Zirconium Introduced in TRIGA Cores, R. Dagan JEF/DOC-1109 Validation of the JEFF-3.1 Fission Yields and Decay Data by Decay Heat and Fission Product Inventories, R. Mills JEF/DOC-1108 Benchmark Calculations for Fast Reactors Using JEFF-3.1, ENDF/B-VII and JENDL-3.3, D-H. Kim JEF/DOC-1107 Criticality Safety Benchmark Calculations with MCNP-4C3 Using JEFF-3.1 Nuclear Data, S. Van der Marck JEF/DOC-1106 Shielding Benchmark Calculations with MCNP-4C3 Using JEFF-3.1 Nuclear Data, S. Van der Marck JEF/DOC-1105 Beta-eff Calculations Using JEFF-3.1 Nuclear Data, S. Van der Marck JEF/DOC-1104 Experimental Activities at IRMM, P. Rullhusen

19 JEF/DOC-1103* Review of Spectral Data for Important Nuclides in JEFF-3.1 Decay Data Files, C. Dean JEF/DOC-1103 Review of Spectral Data for Important Nuclides in JEFF-3.1 Decay Data Files, C. Dean JEF/DOC-1102 Proposal for an Optimized European Multigroup Mesh: The SHEM-281g Mesh, A. Santamarina JEF/DOC-1101 The New OECD-NEA High Priority Request List for Nuclear Data, A. Plompen JEF/DOC-1100 Neutron (n,xn?) Cross-section Measurements for 52Cr, 209Bi and 206,207,208Pb from Threshold Up to 20 MeV, C. Mihailescu JEF/DOC-1099 Verification and Validation of a Multi-temperature JEFF-3.1 Library for MCNP(X), W. Haeck JEF/DOC-1098 Minutes of the Meeting of the Sub-group on Decay Data and Fission Yields, May 2005, Y. Rugama JEF/DOC-1097 Minutes JEFF Meeting, May 2005, Y. Rugama JEF/DOC-1096 Impact of Nuclear Data Uncertainties on a GEN IV-Thorium-reactor at Equilibrium. Proposition of an Update of the High Priority List, A. Bidaud JEF/DOC-1095 Analysis of MASURCA-1Ac/-1B Experiments, J. Tommasi, R. Jacqmin JEF/DOC-1094 Recent Activities on Innovative Radionuclide Production for Metabolic Radiotherapy and PET and on Relevant Experimental and Evaluated Nuclear Data, E. Menapace JEF/DOC-1093 JEFF: The Next Twenty Years?, M. Salvatores JEF/DOC-1092 Some Questions Related to Data Accuracies and Covariances, H. Weigmann JEF/DOC-1091 The Development of Nuclear Data Libraries in Western Europe and the JEF Project, J. Rowlands JEF/DOC-1090 JEFF-3.1 Decay Data and Pulse Fission calculations, Huynh T.D. JEF/DOC-1089 A New Neutron + 235U Evaluation from 2.25 keV Up to 30 MeV, P. Romain, B. Morillon JEF/DOC-1088 A New Neutron-deuterium Data Evaluation, R. Lazaukas JEF/DOC-1087 JENDL-3.3 and JEFF-3.1 NJOY99.90 Deterministic Code Module Updates, J-Ch. Sublet, K. Kosako JEF/DOC-1086 Revised Evaluation of the 241Am Isotope, O. Bouland, D. Bernard JEF/DOC-1085 Summary Record of the JEFF Scientific Co-ordination Group Meeting, NEA Secretariat JEF/DOC-1084 TRIPOLI-4.4 & JEFF-3.1, ENDF/B-VII b0 and JENDl-3.3 with the ICSBEP Criticality Models, J-Ch. Sublet, C. Jouanne, Y. Peneliau, F-X. Hugot JEF/DOC-1083 COGEND: A Code to Generate Nuclear Decay Scheme Data in ENDF-6 Format, R.J. Perry, C.J. Dean JEF/DOC-1082 Update of Previous LWR Cross-section Library Comparison Calculations at VTT, J. Leppanen JEF/DOC-1081 COGEND: Improvements to the Processing of Continuous Spectra, R.J. Perry JEF/DOC-1080 Comparisons Between JEFF31 and JEF-2, C. Mounier JEF/DOC-1079 Brief Report on 95Mo, 99Tc and 245Cm: JEFF-3.1 Status, O. Serot JEF/DOC-1078 Comparative Study of Certain Aspects of the JENDL3.3, JEF-2.2 (UKFY2) and JEFF-3.1 (UKFY3.6) Fission Yields files, O. Serot JEF/DOC-1077 Low Neutron Cross-sections of the Hafnium Isotopes, G. Noguere JEF/DOC-1076 Neutron Data Evaluation of Iodine-127 and Iodine-129 Isotopes, E. Dupont, G. Noguere, E. Bauge JEF/DOC-1075 ORNL-CEA Evaluation of 238U Resonance Parameters Up to 20 keV, A. Courcelle

* Presentation.

20 JEF/DOC-1074 New Evaluation of 238U Cross-section in the Unresolved Range (20 keV-150 keV), A. Courcelle JEF/DOC-1073 Delayed Neutron Yield and Relative Abundance Data for some Minor Isotopes, Missing from Current Compilations, J. Rowlands JEF/DOC-1072 Assessment and Evaluation of Decay Data for EAF-2004/05, A.L. Nichols, R.J. Perry JEF/DOC-1071 Comparison of Fission and Radiative Capture Cross-section Data for 235U and 238U, M. Venhart JEF/DOC-1070 Initial Bechmarking of the JEFF-3.1 Fission Yields and Decay Data, R. Mills JEF/DOC-1069 Test of JEFF-3.1 with MCNP: Criticality and Beta-effective Benchmarks, A. Hogenbirk JEF/DOC-1068 VALDUC with JEFF-3.1T4 235U and 238U, C. Dean, R.J. Perry JEF/DOC-1067 Status of the Photonuclear Activation File (PAF): Reaction Cross-sections, Fission Fragments and Delayed Neutron, D. Ridikas JEF/DOC-1066 Recent Activities of the IAEA Nuclear Data Section, M. Kellett JEF/DOC-1065 The JEFF-3.1 Decay Data Library, M. Kellett JEF/DOC-1064 Experimental Activities at IRMM, P. Rullhusen JEF/DOC-1063 Proposal for Nuclear Data Dispersion Matrix, G. Palmiotti, M. Salvatores, Argonne National Laboratory JEF/DOC-1062 Systematic Comparison of Evaluated Nuclear Data Files, J. Leppanen JEF/DOC-1061 Evaluation of the Resonance Region for 58Fe, M.C. Moxon JEF/DOC-1060 Theoretical Models and Relevant Calculations of Photon Production and Photonuclear Reaction Data, E. Menapace, G. Maino, F. Groppi JEF/DOC-1059 Comparison Between Theoretical Calculation and Experimental Results of Excitation Functions for Production of Relevant Biomedical Radionuclides, E. Menapace,C. Birattari, M.L. Bonardi, F. Groppi, S. Morzenti, C. Zona JEF/DOC-1058 Sensitivity and Uncertainty Analysis of a Thorium Molten Salt Reactor, A. Bidaud JEF/DOC-1057 Current Projects of the IAEA-NDS, A. Trkov JEF/DOC-1056 Short comments on the resolved range of 166Er and 167Er in JEFF-3.0: Proposals for JEFF-3.1, A. Courcelle, C. Chabert, O. Serot JEF/DOC-1055 Review of the Most Reactor Relevant Fission Product Cumulated Yields, O. Serot, B. Roque, A. Santamarina JEF/DOC-1054 Progress on 99Tc Evaluation, F. Gunsing, O. Serot JEF/DOC-1053 Thermal Neutron Scattering Cross-section for H in CaH2 and Ca in CaH2, O. Serot JEF/DOC-1052 56Fe Benchmarking Results, R. Jacqmin JEF/DOC-1051 Benchmark Results for 233U and Thorium, S. van der Marck JEF/DOC-1050 Proposal for a Revised 241Am Evaluation in JEFF-3.1, O. Bouland, A. Santamarina JEF/DOC-1049 Proposal for General Purpose Library JEFF-3.1T, A. Koning, J-Ch. Sublet JEF/DOC-1048 Assembly and Feedback of the JEFF-3.0T3 Radioactive Decay Data File, M. Kellett JEF/DOC-1047 A Survey on Nuclear Data Status and Needs for Special Purposes, Following Items Discussed in Previous JEFF Meetings and NEA-NSC Workshop on “R&D Needs”, E. Menapace JEF/DOC-1046 JEFF Meeting Presentation, I. Kodeli JEF/DOC-1045 VALDUC with JEFF-3.1 Hydrogen, C. Dean, S. Connolly JEF/DOC-1044 TRIPOLI4.3 Benchmarks, C. Jouanne JEF/DOC-1043 JEFF-3.0/JEF-2 Improvement on Fuel Inventory Prediction. Trends from Integral Experiments and Proposal for JEFF-3.1 Evaluations, D. Bernard JEF/DOC-1042 Summary Record of the JEFF Scientific Co-ordination Group Meeting, NEA Secretariat JEF/DOC-1041 Summary Record of the Joint JEFF/EFF/EAF Meeting, NEA Secretariat

21 JEF/DOC-1040 Summary Record of the Decay Data and Fission Yields Meeting, NEA Secretariat JEF/DOC-1039 LWR Cross-section Library Comparison Calculations at VTT, J. Leppanen JEF/DOC-1038 From JEFF-3.0 to JEFF-3.1T: The Way Forward, A. Koning, J-Ch. Sublet JEF/DOC-1037 Report from the CSEWG Meeting, A. Koning JEF/DOC-1036 Overview of the 238U Evaluation in the Resolved Resonance Range, H. Derrien, A. Courcelle, L. Leal, N. Larson JEF/DOC-1035 Experimental Activities at IRMM, P. Rullhusen JEF/DOC-1034 ICSBEP JEFF-3.1T Benchmarkings: Preliminary Results, J-Ch. Sublet, C. Jouanne, Y. Peneliau JEF/DOC-1033 On the Use of S(D,E) Tables for Isotopes with Pronounced Resonances, R. Dagan JEF/DOC-1032 Benchmarking of Uranium-238 Evaluations Against Spherical Transmission and (n,xn)-reaction Rxperimental Data, S. Simakov (presented at ND2004) JEF/DOC-1031 Status of the UKFY3 Fission Yield Evaluation December 2004, R. Mills JEF/DOC-1030 Use of JEFF-3T3 Decay Data with FISPACT-2003, R. Forrest JEF/DOC-1029 Status of UK Decay Data Evaluations, C. Dean JEF/DOC-1028 Comparison of Evaluated Data Libraries: “Chosen Physical Constants of Chosen Nuclides in Spent Fuel”, M. Venhart, V. Chrapciak, Comenius University Bratislava, Department of Nuclear Physics and Biophysics, Study in Diploma Thesis JEF/DOC-1027 The Testing of Recent JEF(F) Decay Data and Fission Product Yields Files for Irradiated Nuclear Fuel Decay Heat Calculations, Drs. R.W. Mills, D.R. Parker, BNFL Sellafield JEF/DOC-1026 Experimental Validation of the APOLLO2 Code for High Burn-up MOX Fuels. JEF-2.2 Results and JEFF-3.0 Improvements, D. Bernard, et al. JEF/DOC-1025 Summary Record of the Joint JEFF/EFF/EAF Meeting, 11-12 May 2004, A. Nouri JEF/DOC-1024 Summary Record of JEFF Scientific Co-ordination Group Meeting, 12 May 2004, A. Nouri JEF/DOC-1023 Recent Nuclear Data Activities at the IAEA Nuclear Data Section, A. Trkov JEF/DOC-1022 Summary Record of the Decay Data and Fission Yields Meeting, 10 May 2004, A. Nouri JEF/DOC-1021 Thermal Neutron Scattering Data, M. Mattes, J. Keinert JEF/DOC-1020 Working Group Experimental Activities, P. Rullhusen JEF/DOC-1019 Experimental Activities at IRMM, Activity Report May 2004, P. Rullhusen JEF/DOC-1018 Implementation of the Delayed Neutron Data Recommendation, A. Nouri JEF/DOC-1017 Benchmarking pre-JEFF-3.1 238U, S.C. van der Marck JEF/DOC-1016 New Evaluations for All Ca and Ge Isotopes, M.C. Duijvestijn, A.J. Koning JEF/DOC-1015 New Data Evaluations for the Fe isotopes, A.J. Koning, M.C. Duijvestijn JEF/DOC-1014 Evaluation of 238U Resonance Parameters from 0 to 20 keV, H. Derrien, A. Courcelle, et al. JEF/DOC-1013 238U Effective Capture Integral Calculations Using Recent Resonance Parameters Sets, A. Courcelle JEF/DOC-1012 JEFF-3.0 Based APOLL02 Library – Processing and Validation, S. Mengelle JEF/DOC-1011 JEFF-3.0 Feedbacks – Pre-JEFF-3.1 Proposals, J-Ch. Sublet, B. Roque, F. Laugier, J. Tommasi JEF/DOC-1010 TRIPOLI-4.3 JEFF-3.0, Pre-JEFF-3.1 Benchmarkings, J-Ch. Sublet, C. Jouanne, Y. Peneliau JEF/DOC-1009 Testing of BRC 238U Evaluations with Fast Benchmarks, E. Dupont, B. Morillon, P. Romain, J-Ch. Sublet JEF/DOC-1008 JEF-2.2 and JEFF-3.0 Comparison: Prediction of Isotopic Concentrations and Reactivity for Thermal Lattices Configurations, A. Courcelle, D. Bernard, A. Santamarina

22 JEF/DOC-1007 First Conclusions of the WPEC/Subgroup-22 Nuclear Data for Improved LEU-LWR Reactivity Predictions, A. Courcelle, et al. JEF/DOC-1006 Mg Thermal Neutron Scattering Cross-sections, C. Mounier JEF/DOC-1005 CaHx Thermal Neutron Scattering Cross-sections, O. Serot JEF/DOC-1004 Neutron Data Evaluation of Rhodium, E. Dupont, E. Bauge, S. Hilaire, A. Koning JEF/DOC-1003 On the Importance of the New Formula for the Double Differential Scattering Kernel, R. Dagan JEF/DOC-1002 Final results of the Evaluation of All Important Neutron Induced Reactions on 46,47,48,49,50Ti Including Full Covariance Information, S. Tagesen, H. Vonach JEF/DOC-1001 Motivation for New Measurements on Am Isotopes, O. Bouland JEF/DOC-1000 Comparative Study of the Independent Fission Yields in JEF-2.2 (UKFY2) and JEFF3T2 (UKFY 3.4), O. Serot, R. Mills JEF/DOC-999 Progress on the Production of UKFY3.5, R. Mills JEF/DOC-998 Modifications to COGEND to Process E-n and E-D Decay Data, R. Perry JEF/DOC-997 COGEND: A Code to Generate Nuclear Decay Scheme Data in ENDF-6 Format, R. Perry, C. Dean JEF/DOC-996 JEFF-3.0/DD T3 Library – Assembly and First Verification, M.A. Kellett, O. Bersillon JEF/DOC-995 Summary Record of the JEFF Scientific Co-ordination Group, 19 Nov. 2003, A. Nouri JEF/DOC-994 Summary Record of the Joint JEFF/EFF/EAF Meeting, 18-19 Nov. 2003, A. Nouri JEF/DOC-993 Summary Record of the JEFF Experimentalist Sub-group Meeting, 17 Nov. 2003, A. Nouri JEF/DOC-992 Summary Record of the Sub-group on Decay Data and Fission Yields, 17 Nov. 2003, A. Nouri JEFDOC-991 Status of Decay Data Evaluations, M.A. Kellett, C.J. Dean JEFDOC-990 Status of UK Decay Data Evaluations, C.J. Dean JEFDOC-989 Measurement Needs for Thermal Reactor Applications: A Short List, Nov. 2003, A. Courcelle JEFDOC-988 New Nuclear Data Libraries for Pb and Bi, November 2003, A.J. Koning, M.C. Duijvestijn, R. Klein Meulekamp, S. van der Marck, A. Hogenbirk JEFDOC-987 First Validation of the New BRC 238U Evaluated File, November 2003, M.J. Lopez Jimenez. JEFDOC-986 Iodine-127 and Iodine-129 Neutron Capture and Total Cross-sections from 0.5 eV to 100 keV for Nuclear Waste Transmutation Studies, New version from Jan. 2005, G. Noguère, A. Brusegan, A. Leprêtre, N. Hérault, E. Macavero, O. Bouland, G. Rudolf JEFDOC-985 Experimental Activities at IRMM, Nov. 2003, P. Rullhusen JEFDOC-984 Impressions from the CSEWG Meeting, Nov. 2003, A. Koning JEFDOC-983 ENDF-6 Formatted Neutron Activation File Improvements, November 2003, P. Dos-Santos-Uzarralde, J-Ch. Sublet JEFDOC-982 The JEFF-3.0/A Neutron Activation File, Nov. 2003, J-Ch. Sublet, A.J. Koning, R.A. Forrest, J. Kopecky JEFDOC-981 Milestones for the Development of the JEFF-3 Decay Data File, Nov. 2003, O. Bersillon JEFDOC-980 Experimental Test of the Crystal Lattice Model of the R-Matrix Code SAMMY, Nov. 2003, A. Courcelle, et al. JEFDOC-979 Summary Report of the WPEC/Subgroup-22 “Nuclear Data for LEU-LWR Reactivity Prediction”, Nov. 2003, A Courcelle, et al. JEFDOC-978 Improvements of Isotopic Ratios Prediction Through Takahama-3 Chemical Assays with the JEFF-3.0 Nuclear Data Library, Nov. 2003, A. Courcelle, et al.

23 JEFDOC-977 JEF-2.2 Nuclear Data Statistical Adjustment Using Post-irradiation Experiments, Nov. 2003, A. Courcelle, et al. JEFDOC-976 Delayed Neutron Data in 8 Time Groups, Nov. 2003, J.L. Rowlands JEFDOC-975 Spectral Index Calculations for Dimple, TRX, Proteus, Godiva, Jezebel and Skidoo, Nov. 2003, S.C. van der Marck, A. Hogenbirk JEFDOC-974 Criticality Results for Many Benchmark Cases – The Releases JEFF-3.0, ENDF-B/ VI.8, JENDL-3.3, and ENDF-B/VII-prelim, Nov. 2003, S.C. van der Marck, A. Hogenbirk JEFDOC-973 Benchmark Results for the Effective Delayed Neutron Fraction, Nov. 2003, S.C. van der Marck, R. Klein Meulekamp JEFDOC-972 The Thermal Activation Cross-section of 58Fe, May 2003, M. Moxon JEFDOC-971 Summary Record of the April 2003 JEFF SCG Meeting, A. Nouri JEFDOC-970 Summary Record of the April 2003 JEFF Meeting, A. Nouri JEFDOC-969 Summary Content of the JEFF-3.0 General Purpose Library, April 2003, A. Nouri JEFDOC-968 235U Evaluations Used in Molten Salt Project (MOST), April 2003, Kophazi, Szieberth, Fehér, de Leege JEFDOC-967 JEFF-3.0 Feedback: Status April 2003, A. Nouri JEFDOC-966 Comparison of Thermal Cross-sections and Resonance Integrals for Dosimetry Reactions, April 2003, A. Trkov JEFDOC-965 Investigation of 238U Inelastic Data from JEF-2.2, JEFF-3.0, ENDF/B-VI.2 and ENDF/B-VI.8, April 2003, O. Litaize, O. Serot, E. Dupont JEFDOC-964 Brief Report on 99Tc: Status and Recommendation, April 2003, O. Serot, F. Gunsing JEFDOC-963 Progress on 52Cr Inelastic Scattering and Zr Activation Cross-sections, April 2003, A. Plompen JEFDOC-962 Modifications to COGEND, April 2003, R. Perry JEFDOC-961 Further Analysis of JEFF-3T2 Decay Data Using FIZCON, April 2003, R. Perry, C. Dean JEFDOC-960 Installation and Testing of FIZCON6.13, April 2003, R. Perry JEFDOC-959 Specification and Maintenance Document for Decay Data Checking in FIZCON, April 2003, M. Grimstone, R. Perry, C. Dean JEFDOC-958 Preliminary Testing of the JEFF-3T2RDD Decay Data Starter File, April 2003, R. Forrest JEFDOC-957 Nuclides of Importance to Decay Heat Calculations, April 2003, R. Mills JEFDOC-956 Preliminary Analysis of JEFF-3.0 vs. JEF-2 Trends in Fast Spectrum Experiments, April 2003, E. Dupont JEFDOC-955 Criticality Benchmark Results – The New Releases JEFF-3.0, ENDF/B-VI.8 and JENDL-3.3, April 2003, S. van der Marck, A. Hogenbirk JEFDOC-954 Needs for 238U Cross-section Measurements to Improve LWR Reactivity Prediction, April 2003, A. Courcelle JEFDOC-953 Analysis of Hellstrand Experimental Correlations, April 2003, A. Courcelle, et al. JEFDOC-952 Summary Report of the WPEC/Subgroup 22, Nuclear Data for Improved LEU-LWR Reactivity Prediction, April 2003, A. Courcelle JEFDOC-951 Microscopic Evaluation of Eu Isotopes, April 2003, E. Bauge JEFDOC-950 MONK8B Analysis of the Hellstrand Experiment, April 2003, D. Hanlon, C.J. Dean JEF/DOC-949 The Generation of Beta Spectra from ENDF6 Evaluations, April 2003, C.J. Dean JEF/DOC-948 Measurements of Thermal Neutron Capture Cross-sections of 241Am, 242gsAm, 243Am and 242Pu, April 2003, F. Marie JEFDOC-947 Comparison of 133Cs and 153Eu Neutron Capture from New Evaluations to JEF-2.2 for a Range of Thermal Reactors, April 2003, R. Mills

24 JEFDOC-946 Evaluation of the Neutron Cross-section for 153Eu, April 2003, M.C. Moxon JEFDOC-945 Evaluation of the Neutron Cross-section for 133Cs, April 2003, M.C. Moxon JEFDOC-944 Summary Records of the JEFF Meeting, December 2002, A. Nouri JEFDOC-943 Calculation of the VALDUC Lattices with Revised 235U Data, April 2003, D. Hanlon JEFDOC-942 Nuclear Data Networking: Minutes of the Informal Meeting Between OECD/NEA Data Bank and EC/DG Research, 17 February 2003 JEFDOC-941 Summary Records of the JEFF/SCG Meeting, December 2002, A. Nouri JEFDOC-940 Cross-section Covariance Data Available in Recent Evaluated Nuclear Data Libraries, I. Kodeli JEFDOC-939 Status of IRDF-2002 Development, A. Nichols JEFDOC-938 Cross-section Evaluation for the System n + 238U, M.J. Lopez JEFDOC-937 Status of the CERN-nTOF Facility, H. Leeb JEFDOC-936 Status of 99Tc Evaluation, F. Gunsing JEFDOC-935 Nuclear Data Status and Needs for Medical Applications, E. Menapace JEFDOC-934 JEFF-3.0: NJOY99 and CALENDF-2002 Processing – Viewpoint, Prospect and Solutions, J-Ch. Sublet, P. Ribon JEFDOC-933 Validation of Depletion Calculation Through Takahama-3 Chemical Assays with JEFF-3.0, A. Courcelle, A. Santamarina JEFDOC-932 Brief Progress Report of the WPEC-SG22, A. Courcelle JEFDOC-931 Motivation for New 241Am Measurements, O. Bouland JEFDOC-930 Status of the JANIS Software, December 2002, A. Nouri JEFDOC-929 SAMQUA – A Program for Generating All Possible Combinations of Quantum Numbers Leading to the Same Compound Nucleus State in the Framework of the R-matrix Code SAMMY, O. Bouland, R. Babut, N.M. Larson JEFDOC-928 MYRRHA: A European Multipurpose ADS for R&D. Pre-design Phase Completion, April 2002, N. Messaoudi JEFDOC-927 HINDAS, A European Project on High- and Intermediate-energy Nuclear Data for Accelerator-driven Systems, April 2002, F. Haddad JEFDOC-926 Light-charged Particle Emission in Fast Neutron-induced Reactions, April 2002, J.P. Meulders, et al. JEFDOC-925 JEFF Proposals for ENDF-7 Format, P. Ribon, et al. (not available) JEFDOC-924 Current Status and Proposals Concerning Hf Evaluation in JEFF-3, April 2002, J-M. Palau JEFDOC-923 FP6 Integrated Projects: Networks of Excellence and Expression of Interest, April 2002 JEFDOC-922 The Difference Between Thermal and Fast Spectrum Values of Fission Product Yields – The Information Obtained in Measurements of Delayed Neutron Yields, April 2002, A. D’Angelo, J. Rowlands JEFDOC-921 JEFF-3.0 Trends Derived from the PROFIL Fast Spectrum Experiments, April 2002, E. Dupont, et al. JEFDOC-920 Conclusions Concerning the Delayed Neutron Data for the Major Actinides, April 2002, A. d’Angelo, J.L. Rowlands JEFDOC-919 New Evaluation of 239Pu Data, April 2002, P. Romain, et al. JEFDOC-918 Integral Validation of the 239Pu CEA Evaluation Adopted in JEFF-3.0, April 2002, E. Dupont, et al. JEFDOC-917 Re-evaluation of the 240Pu Cross-sections in the Unresolved Resonance Energy Range, April 2002, O. Bouland JEFDOC-916 Re-evaluation and Validation of the 241Pu Resonance Parameters in the Energy Range Thermal to 20 eV, April 2002, H. Derrien A. Courcelle A. Santamarina

25 JEFDOC-915 JEFF-3.0 235U and 238U Trends in Fast Spectrum Experiments, April 2002, E. Dupont, et al. JEFDOC-914 238U JEFF-3.0 Benchmarking, April 2002, A. Courcelle, E. Dupont, O. Litaize JEFDOC-913 Aluminium Data Measurements and Evaluation for Criticality Safety Applications, April 2002, L.C. Leal, et al. JEFDOC-912 Progress Report on Main Fission Products: Recommendations for 103Rh, 152Eu and 153Eu, April 2002, O. Serot, et al. JEFDOC-911 Comparative Validation Results of JEFF-3T2, JEF-2.2, ENDF/B-VI.5 and JENDL-3.2, April 2002, A. Hogenbirk JEFDOC-910 Data Sensitivity of Design Calculations for High Conversion D2O Reactors, April 2002, S. Pelloni JEFDOC-909 The UKFY3.4 Fission Product Yield Evaluation, April 2002, R.W. Mills JEFDOC-908 Possible Re-evaluations of Decay Data for JEFF (UK Sponsorship), April 2002, A. Nichols JEFDOC-907 Spectral Data in JEF-2.2 and JEFF-3T2, April 2002, R.J. Perry, C.J. Dean JEFDOC-906 Analysis of JEFF-3T2 Decay Data Using FIZCON, April 2002, R. Perrey, et al. JEFDOC-905 Experimental Activities at IRMM, April 2002, P. Rullhusen JEFDOC-904 Status of the nTOF Facility at CERN, April 2002, F. Gunsing JEFDOC-903 Measurement of the 232Th Cross-section in the Unresolved Region, April 2002, G. Lobo JEFDOC-902 Reserved to A. Nouri (JEFF minutes), April 2002 JEFDOC-901 Reserved to A. Nouri (JEFF/SCG minutes), April 2002 JEFDOC-900 Energy Release in 239Pu Fission, April 2002, O. Serot, et al.

EFF documents

EFFDOC-959 Requirements for Nuclear Data Development and Neutronic Analysis Efforts in the EU Fusion Technology Program 2007-2013 (FP7), P. Batistoni, U. Fischer, R. Forrest EFFDOC-958 Report of the IAEA TC Meeting on Nuclear Data for IFMIF, Karlsruhe, 4 October 2005, U. Fischer, R. Forrest, A. Mengoni EFFDOC-957 Theoretical Calculations of Covariance Matrices for Li Isotopes, H. Leeb, M.T. Pigni EFFDOC-956 Measurement of Tritium Production Rates and of Neutron and Gamma-ray Flux Spectra in the TBM Mock-up, K. Seidel EFFDOC-955 Validation Experiment of Gamma-activities of Tantalum Irradiated in Fusion Peak Neutron Field, K. Seidel EFFDOC-954 Monte-Carlo-based Transport and Sensitivity/Uncertainty Analyses of the Tritium Production in the HCPB Breeder Blanket Mock-up Experiment, U. Fischer, D. Leichtle, R. Perel EFFDOC-953 Evaluation of 181Ta Cross-sections for the EFF Data Library Up to 150 MeV, P. Pereslavtsev EFFDOC-952 First Benchmark Analyses of Ta Evaluated Data for Fusion Neutron Transport Calculations, S.P. Simakov, U. Fischer EFFDOC-951 Analyses of the NPI/Rez Validation Experiment for Ta Activation Cross-sections in an IFMIF-like Neutron Spectrum, S.P. Simakov, et al. EFFDOC-950 Deterministic 3-D Analysis of HCPB Breeder Blanket Mock-up Experiment, I. Kodeli EFFDOC-949 Comparison of Covariance Matrices Calculated by A. Koning (2005) and Covariance Matrices from Evaluation of Experimental Data for 56Fe, S. Tagesen, H. Vonach EFFDOC-948 Nuclear Decay Heat in Molybdenum Measurements and Analysis, M. Pillon

26 EFFDOC-947 Validation Report for EAF-2005 and Deuteron Data, R. Forrest EFFDOC-946 Validation Experiment on Tantalum Activation in the NPI p-D2O Neutron Field of IFMIF-like Spectrum, M. Honusek EFFDOC-945 Validation Experiment on Tungsten Activation in the NPI p-D2O Neutron Field of IFMIF-like Spectrum, E. Simeckova EFFDOC-944 Experimental Determination of the NPI p-D2O Neutron Source Spectrum for Activation Benchmark Tests, P. Bém EFFDOC-943 Comparison of new 48Ti Measurements with 2002 Evaluation, S. Tagesen EFFDOC-942 Evaluation of 182,183,184,186W Neutron Cross-sections Up to 20 MeV for Combination with the FZK Theoretical 150 MeV Evaluation, S. Tagesen EFFDOC-941 Progress on Deuteron-induced Activation Cross-section Evaluation by Means of Optical Model Potential Analysis, M. Avrigeanu EFFDOC-940 Progress on Calculation of the Activation Cross-sections for Ta and Hf up to 60 MeV, V. Avrigeanu EFFDOC-939 Release of EAF-2005.1 Library, J. Kopecky EFFDOC-938 Status of the TBM Neutronics Experiment, P. Batistoni EFFDOC-937 Summary Record of the EFF/EAF Monitor Group Meeting, May 2005, H. Henriksson EFFDOC-936 Evaluation of All Important Neutron Cross-sections for 184W Including Full Covariance Information, S. Tagesen EFFDOC-935 New Evaluation of d+Li Data Up to 50 MeV for IFMIF, P. Pereslavtsev EFFDOC-934 Progress Report of TU Dresden, K. Seidel EFFDOC-933 HCPB Breeder Blanket Mock-up Experiment Pre-analysis Using TORT Code, I. Kodeli EFFDOC-932 Benchmark Analyses of 9Be JEFF-3.1T Data for Fusion Applications, D. Leichtle EFFDOC-931 Testing of New W Data Evaluations for EFF-3, U. Fischer EFFDOC-930 Benchmark Analyses of Ti JEFF-3.1T Data for Fusion Applications, U. Fischer EFFDOC-929 Benchmark Analyses of Pb JEFF-3.1T Data for Fusion Applications, S. Simakov EFFDOC-928 Activation Cross-sections of the Hf and Ta Stable Isotopes and their Sensitivity to Nuclear Reaction Mechanisms, M. Avrigeanu, V. Avrigeanu, R.A. Forrest, F.L. Roman EFFDOC-927 The Release of EASY-2005 and Current Work, R.A. Forrest EFFDOC-926 Testing MCSEN Track-length Sensitivity Calculations for the TBM Mock-up Experiment, R. Perel EFFDOC-925 Status of Test Blanket Module (TBM) Neutronics Experiment, P. Batistoni EFFDOC-924 Validation of EAF-2005 – New Tools, J. Kopecky EFFDOC-923 Summary Record of the EFF/EAF Monitor Group Meeting, November 2004, H. Henriksson EFFDOC-922 Validation Analyses of IEAF-2001 Activation Cross-section Data for SS-316 and F82H Steels Irradiated in a White d-Li Neutron Field, S.P. Simakov, et al. (Presented at the ND2004 in Santa Fe, USA) EFFDOC-921 Evaluation of n+W Cross-section Data Up to 150 MeV Neutron Energy, P. Pereslavtsev, U. Fischer (Presented at the ND2004 in Santa Fe, USA) EFFDOC-920 Nuclear Data for Fusion Energy Technologies: Requests, Status and Development Needs, U. Fischer, et al. (Presented at the ND2004 in Santa Fe, USA) EFFDOC-919 TALYS, Monte Carlo and Covariances, A.J. Koning EFFDOC-918 HCPB Breeder Blanket Mock-up Experiment Analysis Using DORT and SUSD3D Codes, I. Kodeli EFFDOC-917 Nuclear Decay Heat in Tantalum – Measurements and Analysis, M. Pillon (presented by P. Batistoni)

27 EFFDOC-916 Validation Experiment of Gamma Activities of Pb Irradiated in Fusion Peak Neutron Field, R. Eichin, R.A. Forrest, H. Freiesleben, K. Seidel, S. Unholzer EFFDOC-915 Implementation and Testing of Sensitivity Calculations of Monte Carlo Track Length Estimators in MCSEN, R. Perel (presented by U. Fischer) EFFDOC-914 Analysis of Rez Data for Dosimetry Foils Using EAF-2004, R. Forrest EFFDOC-913 Analysis of the NPI Validation Experiment for Tungsten Activation Cross-sections in an IFMIF-like Neutron Spectrum, S. Simakov, et al. EFFDOC-912 Evaluation of n + 182,183,184,186W Cross-sections for EFF Up to 150 MeV, P. Pereslavtsev, U. Fischer EFFDOC-911 First Benchmark Analyses for Ti Data Evaluations, U. Fischer, S. Simakov EFFDOC-910 Activation Benchmark Tests at the NPI Fast Neutron Facility, P. Bém EFFDOC-909 Status of EAF-2005, J. Kopecky EFFDOC-908 Status of Neutronics Experiment on a Mock-up of a Test Blanket Module (TBM), P. Batistoni EFFDOC-907 Calculation of D-Li Cross-sections for Energies Up to 50 MeV Using Realistic Nucleon-nucleon Interactions, M. Avrigeanu EFFDOC-906 Development of Theoretical Tools and Their Use to Calculate Cross-sections Relevant to the EAF and EFF Files, V. Avrigeanu EFFDOC-905 Summary Record of the EFF/EAF Monitor Group Meeting 10 May 2004, R. Forrest EFFDOC-904 Radiation Shielding and Dosimetry Experiments Updates in the SINBAD Database, I. Kodeli, H. Hunter, E. Sartori EFFDOC-903 Nuclear Decay Heat in Yttrium, Revised Measurements and Analysis, M. Pillon EFFDOC-902 Decay Heat Estimate Based on EAF-2003 and FENDL/A-2 Nuclear Data Libraries for Fusion Relevant Materials in Comparison with FNS-JAERI Experiments, 10 May 2004, D. Cepraga, G. Cambi, M. Frisoni, E. Menapace EFFDOC-901 Coupled Channels Calculations for d + 6Li, M. Avrigeanu EFFDOC-900 Post-analysis of the NPI Irradiation Experiment on Eurofer Up to 35 MeV Using IEAF-2001 Activation Cross-sections, S. Simakov, et al. EFFDOC-899 Status of n + W Cross-section Evaluations for EFF Up to 150 MeV, P. Pereslavtsev, U. Fischer EFFDOC-898 Monte-Carlo-based Pre-analysis of the HCPB Breeder Blanket Mock-up Experiment: Neutron Spectra, Tritium Production and Sensitivities to Cross-sections, U. Fischer, R. Perel EFFDOC-897 Re-analysis of TUD Benchmark Experiment on Tungsten Using JENDL-3.3 Data, U. Fischer EFFDOC-896 Status of TBM Neutronics Experiment, 10 May 2004, P. Batistoni EFFDOC-895 International Benchmark Comparison on Tritium Measurements, 10 May 2004, P. Batistoni EFFDOC-894 Progress Report of TU Dresden, 10 May 2004, K. Seidel EFFDOC-893 Progress Report on Fast-neutron Reaction Analysis for Fusion Low-activation Materials, 10 May 2004, V. Avrigeanu EFFDOC-892 New Actions for EAF-2005, 10 May 2004, J. Kopecky EFFDOC-891 Summary Record of the EFF/EAF Monitor Group Meeting, 17 November 2003, M.A. Kellett EFFDOC-890 High Priority Request List: Fusion, A.J. Koning EFFDOC-889 Inelastic Scattering on 52Cr with the n,ncg Technique, A.J.M. Plompen EFFDOC-888 Covariance Determination for Evaluations Including Extensive Modelling, H. Leeb EFFDOC-887 Calculation of Deuterium-lithium Cross-sections for Energies Up to 50 MeV Using Realistic Nucleon-nucleon Interactions, M. Avrigeanu EFFDOC-886 Pre-analysis of TBM Experiment Using DORT and SUSD3D Codes, I. Kodeli

28 EFFDOC-885 Tungsten Benchmark Experiments: Re-analysis Using JENDL-3.3, I. Kodeli EFFDOC-884 TBM Neutronics Experiment: Design of the Measurements of Tritium Production and of Neutron and Gamma-ray Flux Spectra, K. Seidel EFFDOC-883 Validation Experiment of Gamma Activities of Yttrium Irradiated in Fusion Peak Neutron Field, K. Seidel EFFDOC-882 Validation of EASY-2003, R.A. Forrest EFFDOC-881 EASY-2004, R.A. Forrest, J. Kopecky EFFDOC-880 Progress Report on Fast-neutron Reaction Analysis for Fusion Low-activation Materials, V. Avrigeanu EFFDOC-879 Status of the Evaluation of all Ti Isotopes, S. Tagesen, H. Vonach EFFDOC-878 Modification to GLUCS to Avoid a Systematic Underestimation of Cross-sections (In Particular with Inconsistent and Highly Correlated Experimental Data), S. Tagesen, H. Vonach EFFDOC-877 Design of TBM Neutronics Experiment, P. Batistoni EFFDOC-876 IEAF-2001 Cross-section Data Library for Activation and Transmutation Analyses of Intermediate Energy Systems with the FISPACT Inventory Code, U. Fischer, et al. EFFDOC-875 Pre-analysis of the Validation Experiment on Eurofer Up to 55 MeV in an IFMIF- like Neutron Spectrum Benchmark, S. Simakov EFFDOC-874 Benchmark Analysis of the EFF-3.03 52Cr Data Evaluation, U. Fischer, S. Simakov, I. Schmuck EFFDOC-873 Sensitivity Calculation of Monte-Carlo Track-length Estimators for Cross-section Sensitivity and Uncertainty Analyses, R. Perel EFFDOC-872 Evaluation of n + 184W Cross-section Data for EFF Up to 150 MeV, P. Pereslavtsev EFFDOC-871 Measurement and Analysis of Nuclear Decay Heat in Yttrium, M. Pillon EFFDOC-870 IEAF-2001 Cross-section Data Libraries for Activation and Transmutation Analyses of Intermediate-energy Systems, U. Fischer, et al. (presented at the TRAMU Workshop, Darmstadt, 2-5 Sept. 2003) EFFDOC-869 Production of 62mCo in Copper Irradiated in the Karlsruhe d-Be Neutron Field, U. von Möllendorff, S. Simakov EFFDOC-868 Summary Record of the EFF/EAF Monitor Group Meeting, 29 April 2003, M.A. Kellett EFFDOC-867 Analysis of FNG Benchmark Experiment on Tungsten Using DORT, TWODANT and SUSD3D Deterministic Codes, I. Kodeli EFFDOC-866 Recent Progress in the SINBAD Project, I. Kodeli EFFDOC-865 Progress Towards EASY-2004, R.A. Forrest EFFDOC-864 Design of TBM Neutronics Experiment (2003), P. Batistoni EFFDOC-863 Final Report on the Tungsten Experiment (2002), P. Batistoni, M. Angelone, L. Petrizzi, M. Pillon EFFDOC-862 Preliminary Measurements and Analysis of Nuclear Decay Heat in Sc, Sm, Gd and Dy, M. Pillon EFFDOC-861 EAF-2003 Versus 252Cf Integral Data, J. Kopecky EFFDOC-860 Monte Carlo Transport and Sensitivity Analyses for the TUD Neutron Transport Benchmark Experiment on Tungsten, U. Fischer, et al. EFFDOC-859 Validation Analyses of IEAF-2001 Activation Cross-section Data for SS-316 Steel, S. Simakov, et al. EFFDOC-858 Measurement and Analysis of Induced Activation in CuCrZr Irradiated in Fusion Peak Neutron Field, R. Eichin, R.A. Forrest, H. Freiesleben, S.A. Goncharov, V.D. Kovalchuk, D.V. Markovskij, D.V. Maximov, K. Seidel, S. Unholzer EFFDOC-857 Measurement and Analysis of Neutron and Gamma-ray Flux Spectra in Tungsten, H. Freiesleben, C. Negoita, K. Seidel, S. Unholzer, U. Fischer, D. Leichtle, M Angelone, P. Batistoni, M. Pillon

29 EFFDOC-856 Comparison of Different Existing Cross-section Calculations for 46,47,49,50Ti, H. Vonach, S. Tagesen EFFDOC-855 Results of Inclusion of New Geel Data for 58Ni into the Recent Evaluation, H. Vonach, S. Tagesen EFFDOC-854 Progress Report on Fast-neutron Analysis for Fusion Low-activation Materials, V. Avrigeanu EFFDOC-853 Progress Report on Calculation of D-Li Cross-sections for Energies Up to 50 MeV, M. Avrigeanu EFFDOC-852 Nuclear Data for Design Analyses of the Test Blanket Modules in ITER: Review and Recommendations for EFF/JEFF Evaluations, U. Fischer EFFDOC-851 Summary Record of the EFF/EAF Monitor Group Meeting, 5-6 December 2002, M.A. Kellett EFFDOC-850 Neutronics and Nuclear Data: Status and Perspectives for Fusion Technology Applications (presented at the Dresden Workshop on Fast Neutron Physics in September 2002), U. Fischer EFFDOC-849 EFDA, G. Le Marois EFFDOC-848 Processed Covariance Matrices for 28Si, Evaluation EFF-3.0, I. Kodeli EFFDOC-847 Processed Covariance Matrices for 56Fe, Evaluation JEFF-3.0/EFF-3.1, I. Kodeli EFFDOC-846 ALARA/IEAF-2001: Validation and Application for IFMIF Activation and Transmutation Analyses, S. Simakov EFFDOC-845 Monte Carlo Sensitivity Analysis of FNS and KANT Be Experiments Using the 9Be EFF3.0/NMOD=5 Evaluation, R. Perel EFFDOC-844 Benchmark Analysis of EFF and FENDL Carbon Cross-section Data, D. Leichtle EFFDOC-843 Integral Data for EAF-2003, J. Kopecky EFFDOC-842 Status of EAF libraries, R.A. Forrest EFFDOC-841 A Complementary Report on Decay Heat Estimate for Fusion-relevant Materials Based on EAF-99 and FENDL/A-2 Libraries in Comparison with FNS-JAERI Experiments, E. Menapace EFFDOC-840 FNS Experimental Validation of the Decay Power Calculation Code and Nuclear Databases – FISPACT-2001 and EAF-99 & EAF-2001 & EAF-2001 Decay Libraries – 7 Hours Irradiation, J-Ch. Sublet, F. Maekawa EFFDOC-839 New FNS Experimental Validation of the Decay Power Calculation Code and Nuclear Databases – FISPACT-2001 and EAF-99 & EAF-2001 & EAF-2001 Decay Libraries – 5 Minutes Irradiation, J-Ch. Sublet, F. Maekawa EFFDOC-838 Measurement and Analysis of Nuclear Decay Heat in CuCrZr, M. Pillon EFFDOC-837 ZZ-VITAMIN-J/COVA/EFF-3 Cross-section Covariance Matrix Library, I. Kodeli EFFDOC-836 Progress Report of TUD, K. Seidel EFFDOC-835 Completion of Model Calculation of Fast Neutron Reaction Cross-sections for all Stable Isotopes of Ni and Mo, V. Avrigeanu, M. Avrigeanu, T. Glodariu EFFDOC-834 Inelastic Scattering of 58Ni and Progress with Activation Measurements, paper from PHYSOR’2002 supplied, A. Plompen EFFDOC-833 Status of the Complete Evaluation of all Important Neutron Reactions on 48Ti Including Full Covariance Information, H. Vonach, S. Tagesen EFFDOC-832 Final Full Documentation of the Evaluations of 58Ni, 60Ni and 28Si, H. Vonach, S. Tagesen EFFDOC-831 Progress Report on W Experiment at FNG, P. Batistoni EFFDOC-830 A Brief Overview of the European Fusion File (EFF) Project (presented at IAEA FEC 2002, Lyon), M.A. Kellett, R.A. Forrest, P. Batistoni, on behalf of the EFF Project members EFFDOC-829 Summary Record of the EFF/EAF Monitor Group Meeting, 25 April 2002, M.A. Kellett

30 EFFDOC-828 Needs for Nuclear Data Development and Improvement in the Fusion Technology Program in 2003-2006 and Beyond, P. Batistoni, U. Fischer, R.A. Forrest EFFDOC-827 Measurements on Neutron-irradiated Chromium (report sent to EFDA), M. Pillon, M. Angelone EFFDOC-826 Neutronics and Nuclear Data for the IFMIF Neutron Source (ISFNT-6 paper), U. Fischer, S.P. Simakov, A. Konobeev, P. Pereslavtsev, P. Wilson EFFDOC-825 Nuclear Data for the IFMIF Neutron Source – Status and Future Plans (presented to the JEFF IE group), U. Fischer EFFDOC-824 Status of the New Evaluation of Neutron Cross-sections for 12C and 16O, A. Mengoni EFFDOC-823 Progress with Ni Cross-section Measurements, A. Plompen EFFDOC-822 Measurement and Analysis of Neutron and Gamma-ray Flux Spectra in SiC, H. Freiesleben, C. Negoita, K. Seidel, S. Unholzer, Y. Chen, U. Fischer, M. Angelone, P. Batistoni, M. Pillon EFFDOC-821 Activation Experiment with Tungsten in Fusion Peak Neutron Field, K. Seidel, R.A. Forrest, H. Freiesleben, S.A. Goncharov, V.D. Kovalchuk, D.V. Markovskij, D.V. Maximovich, S. Unholzer, R. Weigel EFFDOC-820 Validation of the EASY-2001 Activation System (ISFNT-6 paper), R.A. Forrest, M. Pillon, U. von Möllendorff, K. Seidel EFFDOC-819 EAF-2001 Validation Against Integral Experiments, J. Kopecky EFFDOC-818 Deterministic Transport, Sensitivity and Uncertainty Analysis of SiC Benchmark Experiment Using EFF-3 and FENDL-2 Evaluations, I. Kodeli EFFDOC-817 First Results from Monte Carlo Sensitivity Calculations Using the 9Be EFF3.05 Evaluation, D. Leichtle EFFDOC-816 EAF-2001 in ENDF-6 Format and NJOY99-68 Processing, J-Ch. Sublet, A.J. Koning, R.E. MacFarlane EFFDOC-815 Monte Carlo Transport, Sensitivity and Uncertainty Analyses for the TUD Benchmark Experiment on SiC, U. Fischer, R. Perel, Y. Chen EFFDOC-814 Analysis of the Shutdown Dose Rate Experiment: Comparison of Results with EFF/EAF, JENDL and FENDL Libraries, P. Batistoni EFFDOC-813 Final Results of the Analysis of Benchmark Experiment on SiC, P. Batistoni EFFDOC-812 Measurements of Decay Heat and Validation of the European Activation Code System for Fusion Power Plant Applications, M. Pillon EFFDOC-811 Measurements on Neutron-irradiated Chromium, M. Pillon EFFDOC-810 Summary Record of the EFF/EAF Monitor Group Meeting, 20 November 2001, M.A. Kellett

TABLES

Table 1. Origin of evaluations in the JEFF-3.1 general-purpose neutron data library

Material MAT Data source Laboratory Evaluation Author(s) 1-H-1 125 ENDF/B-VI.8 LANL EVAL-FEB98 G.M. Hale, P.G. Young 1-H-2 128 ENDF/B-VI.8 LANL EVAL-FEB97 P.G. Young, G.M. Hale, M.B.Chadwick 1-H-3 131 JEFF-3.0 CNDC EVAL-JAN91 Zhuang Youxiang 2-He-3 225 JEFF-3.0 LANL EVAL-MAY90 G. Hale, D. Dodder, P. Young 2-He-4 228 JENDL-3.3 JAERI EVAL-FEB87 K. Shibata 3-Li-6 325 JEFF-3.0 LANL EVAL-APR89 G.M. Hale, P.G. Young 3-Li-7 328 JEFF-3.0 ECN EVAL-AUG90 Birmingham, Petten, Geel, LASL 4-Be-9 425 EFF3 MOD6 IRK, VIENNA EVAL-JAN97 Vienna, Obninsk 5-B-10 525 JEFF-3.0 LANL EVAL-NOV89 G.M. Hale, P.G. Young 5-B-11 528 ENDF/B-VI.8 LANL EVAL-MAY89 P.G. Young 6-C-0 600 ENDF/B-VI.8 LANL, ORNL EVAL-JUN96 M.B. Chadwick, P.G. Young, C.Y. Fu 7-N-14 725 ENDF/B-VI.8 LANL EVAL-JUN97 M.B. Chadwick, P.G. Young 7-N-15 728 JEFF-3.0 LANL EVAL-SEP83 E. Arthur, P. Young, G. Hale

35 8-O-16 825 ENDF/B-VI.8 LANL EVAL-APR01 Hale, Young, Chadwick, Caro, Lubitz 8-O-17 828 JEFF-3.0 NEA RCOM-JUN91 JEF SCG 9-F-19 925 Pre-ENDF/B-VII CNDC, ORNL EVAL-JUN90 Z.X. Zhao, C.Y. Fu, D.C. Larson 11-Na-22 1122 JEFF-3.0 NEA RCOM-JUN83 Scientific Co-ordination Group 11-Na-23 1125 JEFF-3.0 NEA EVAL-APR99 E. Fort, et al. 12-Mg-24 1225 JENDL-3.3 DEC, NEDAC EVAL-MAR87 M. Hatchya (DEC), T. Asami (NEDAC) 12-Mg-25 1228 JENDL-3.3 DEC, NEDAC EVAL-MAR87 M. Hatchya (DEC), T. Asami (NEDAC) 12-Mg-26 1231 JENDL-3.3 DEC, NEDAC EVAL-MAR87 M. Hatchya (DEC), T. Asami (NEDAC) 13-Al-27 1325 JEFF-3.0 LANL EVAL-FEB97 M.B. Chadwick, P.G. Young 14-Si-28 1425 JEFF-3.0 IRK EVAL-MAY96 S. Tagesen, H. Vonach, A. Trkov 14-Si-29 1428 JENDL-3.3 TIT, JAERI EVAL-MAR88 H. Kitazawa,Y. Harima, T. Fukahori 14-Si-30 1431 JENDL-3.3 TIT, JAERI EVAL-MAR88 H. Kitazawa,Y. Harima, T. Fukahori 15-P-31 1525 JENDL-3.3 FUJI E.C. EVAL-MAY87 H. Nakamura 16-S-32 1625 JENDL-3.3 FUJI E.C. EVAL-MAY87 H. Nakamura 16-S-33 1628 JENDL-3.3 FUJI E.C. EVAL-MAY87 H. Nakamura 16-S-34 1631 JENDL-3.3 FUJI E.C. EVAL-MAY87 H. Nakamura 16-S-36 1637 JENDL-3.3 FUJI E.C. EVAL-MAY87 H. Nakamura

Table 1. Origin of evaluations in the JEFF-3.1 general-purpose neutron data library (cont.)

Material MAT Data source Laboratory Evaluation Author(s) 17-Cl-35 1725 Pre-ENDF/B-VII ORNL EVAL-OCT03 R. Sayer, K. Guber, L. Leal, N. Larson 17-Cl-37 1731 Pre-ENDF/B-VII ORNL EVAL-OCT03 R. Sayer, K. Guber, L. Leal, N. Larson 18-Ar-36 1825 JEFF-3.0 NEA RCOM-JUN83 Scientific Co-ordination Group 18-Ar-38 1831 JEFF-3.0 NEA RCOM-JUN83 Scientific Co-ordination Group 18-Ar-40 1837 JENDL-3.3 KHI EVAL-MAR94 T. Watanabe 19-K-39 1925 JENDL-3.3 FUJI E.C. EVAL-MAY87 H. Nakamura 19-K-40 1928 JENDL-3.3 FUJI E.C. EVAL-MAY87 H. Nakamura 19-K-41 1931 JENDL-3.3 FUJI E.C. EVAL-MAY87 H. Nakamura 20-Ca-40 2025 New eval. NRG EVAL-OCT04 A.J. Koning 20-Ca-42 2031 New eval. NRG EVAL-OCT04 A.J. Koning 20-Ca-43 2034 New eval. NRG EVAL-OCT04 A.J. Koning 20-Ca-44 2037 New eval. NRG EVAL-OCT04 A.J. Koning 20-Ca-46 2043 New eval. NRG EVAL-OCT04 A.J. Koning

36 20-Ca-48 2049 New eval. NRG EVAL-OCT04 A.J. Koning 21-Sc-45 2125 New eval. NRG EVAL-OCT04 A.J. Koning 22-Ti-46 2225 New eval. IRK EVAL-JAN04 Vienna: S. Tagesen, H. Vonach 22-Ti-47 2228 New eval. IRK EVAL-JAN04 Vienna: S. Tagesen, H. Vonach 22-Ti-48 2231 New eval. IRK EVAL-JAN04 Vienna: S. Tagesen, H. Vonach 22-Ti-49 2234 New eval. IRK EVAL-JAN04 Vienna: S. Tagesen, H. Vonach 22-Ti-50 2237 New eval. IRK EVAL-JAN04 Vienna: S. Tagesen, H. Vonach 23-V-0 2300 JEFF-3.0 ECN EVAL-OCT91 Gruppelaar, van der Kamp, Kopecky, Nierop 24-Cr-50 2425 JEFF-3.0 ORNL EVAL-NOV89 D.C. Hetrick, N.M. Larson, Fu 24-Cr-52 2431 New eval. IRK, CEA EVAL-DEC95 S. Tagesen, H. Vonach, O. Bouland 24-Cr-53 2434 JEFF-3.0 ORNL EVAL-NOV89 Shibata, D.C. Hetrick, N.M. Larson, Fu 24-Cr-54 2437 JEFF-3.0 ORNL EVAL-NOV89 D.C. Hetrick, N.M. Larson, Fu 25-Mn-55 2525 JENDL-3.3 JAERI, MAPI EVAL-MAR87 K. Shibata,T. Hojuyama 26-Fe-54 2625 New eval. NRG EVAL-OCT04 A.J. Koning 26-Fe-56 2631 New eval. NRG EVAL-FEB04 European Joint Collaboration 26-Fe-57 2634 NRG-2004 NRG EVAL-OCT04 A.J. Koning 26-Fe-58 2637 NRG-2004 NRG EVAL-OCT04 A.J. Koning

Table 1. Origin of evaluations in the JEFF-3.1 general-purpose neutron data library (cont.)

Material MAT Data source Laboratory Evaluation Author(s) 27-Co-58 2722 JEFF-3.0 NEA RCOM-JUN83 Scientific Co-ordination Group 27-Co-58M 2723 JEFF-3.0 NEA RCOM-JUN82 Scientific Co-ordination Group 27-Co-59 2725 JEFF-3.0 ANL, ORNL EVAL-JUL89 A. Smith, G. Desaussure 28-Ni-58 2825 JEFF-3.0 IRK, IJS EVAL-AUG99 European Joint Collaboration 28-Ni-59 2828 JEFF-3.0 NEA, ECN RCOM-NOV87 Gruppelaar,van der Kamp, Kopecky, Nierop 28-Ni-60 2831 JEFF-3.0 IRK, IJS EVAL-JAN00 European Joint Collaboration 28-Ni-61 2834 ENDF/B-VI.8 LANL,ORNL EVAL-SEP97 S. Chiba, M.B. Chadwick, Hetrick 28-Ni-62 2837 ENDF/B-VI.8 LANL,ORNL EVAL-SEP97 S. Chiba, M.B. Chadwick, Hetrick 28-Ni-64 2843 ENDF/B-VI.8 LANL, ORNL EVAL-SEP97 S. Chiba, M.B. Chadwick, Hetrick 29-Cu-63 2925 ENDF/B-VI.8 LANL, ORNL EVAL-FEB98 A.J. Koning, M..B. Chadwick, Hetrick 29-Cu-65 2931 ENDF/B-VI.8 LANL, ORNL EVAL-FEB98 A.J. Koning, M..B. Chadwick, Hetrick 30-Zn-0 3000 JEFF-3.0 FEI EVAL-DEC89 M.N. Nikolaev, S.V. Zabrodskaja 31-Ga-0 3100 JEFF-3.0 KHI EVAL-MAR94 T. Watanabe

37 32-Ge-70 3225 New eval. NRG EVAL-DEC04 A.J. Koning 32-Ge-72 3231 New eval. NRG EVAL-DEC04 A.J. Koning 32-Ge-73 3234 New eval. NRG EVAL-DEC04 A.J. Koning 32-Ge-74 3237 New eval. NRG EVAL-OCT04 A.J. Koning 32-Ge-76 3243 New eval. NRG EVAL-OCT04 A.J. Koning 33-As-75 3325 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 34-Se-74 3425 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 34-Se-76 3431 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 34-Se-77 3434 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 34-Se-78 3437 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 34-Se-79 3440 JENDL-3.3 JNDC EVAL-MAR90 JNDC FP Nuclear Data WG 34-Se-80 3443 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 34-Se-82 3449 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 35-Br-79 3525 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 35-Br-81 3531 JEFF-3.0 NEA RCOM-JUL82 H. Gruppelaar, E. Menapace 36-Kr-78 3625 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 36-Kr-80 3631 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group

Table 1. Origin of evaluations in the JEFF-3.1 general-purpose neutron data library (cont.)

Material MAT Data source Laboratory Evaluation Author(s) 36-Kr-82 3637 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 36-Kr-83 3640 JEFF-3.0 NEA RCOM-JUL82 H. Gruppelaar, E. Menapace 36-Kr-84 3643 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 36-Kr-85 3646 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 36-Kr-86 3649 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 37-Rb-85 3725 JEFF-3.0 NEA RCOM-JUL82 H. Gruppelaar, E. Menapace 37-Rb-86 3728 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 37-Rb-87 3731 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 38-Sr-84 3825 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 38-Sr-86 3831 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 38-Sr-87 3834 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 38-Sr-88 3837 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 38-Sr-89 3840 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group

38 38-Sr-90 3843 JEFF-3.0 NEA RCOM-JUL82 H. Gruppelaar, E. Menapace 39-Y-89 3925 JEFF-3.0 ANL, LLNL EVAL-JAN86 R. Howerton (LLNL), A. Smith (ANL) 39-Y-90 3928 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 39-Y-91 3931 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 40-Zr-90 4025 JENDL-3.3 JNDC EVAL-AUG89 JNDC FP Nuclear Data WG 40-Zr-91 4028 JENDL-3.3 JNDC EVAL-AUG89 JNDC FP Nuclear Data WG 40-Zr-92 4031 JENDL-3.3 JNDC EVAL-AUG89 JNDC FP Nuclear Data WG 40-Zr-93 4034 JEFF-3.0 NEA RCOM-JUL82 H. Gruppelaar, E. Menapace 40-Zr-94 4037 JENDL-3.3 JNDC EVAL-AUG89 JNDC FP Nuclear Data WG 40-Zr-95 4040 JEFF-3.0 NEA RCOM-JUL83 H. Gruppelaar, E. Menapace 40-Zr-96 4043 JENDL-3.3 JNDC EVAL-AUG89 JNDC FP Nuclear Data WG 41-Nb-93 4125 ENDF/B-VI.8 LANL, ANL EVAL-DEC97 M. Chadwick, P. Young, D.L. Smith 41-Nb-94 4128 JENDL-3.3 JNDC EVAL-MAR90 JNDC FP Nuclear Data WG 41-Nb-95 4131 JENDL-3.3 JNDC EVAL-MAR90 JNDC FP Nuclear Data WG 42-Mo-92 4225 JENDL-3.3 JNDC EVAL-AUG89 JNDC FP Nuclear Data WG 42-Mo-94 4231 JENDL-3.3 JNDC EVAL-AUG89 JNDC FP Nuclear Data WG 42-Mo-95 4234 Pre-ENDF/B-VII BNL, KAERI EVAL-OCT03 Lee, Oh, Mughabghab, Oblozinsky

Table 1. Origin of evaluations in the JEFF-3.1 general-purpose neutron data library (cont.)

Material MAT Data source Laboratory Evaluation Author(s) 42-Mo-96 4237 JENDL-3.3 JNDC EVAL-AUG89 JNDC FP Nuclear Data WG 42-Mo-97 4240 JENDL-3.3 JNDC EVAL-AUG89 JNDC FP Nuclear Data WG 42-Mo-98 4243 JENDL-3.3 JNDC EVAL-AUG89 JNDC FP Nuclear Data WG 42-Mo-99 4246 JENDL-3.3 JNDC EVAL-MAR90 JNDC FP Nuclear Data WG 42-Mo-100 4249 JENDL-3.3 JNDC EVAL-AUG89 JNDC FP Nuclear Data WG 43-Tc-99 4331 New eval. CEA, NRG EVAL-FEB05 Gunsing, Serot, Koning 44-Ru-96 4425 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 44-Ru-98 4431 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 44-Ru-99 4434 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 44-Ru-100 4437 JEFF-3.0 NEA RCOM-JUL82 H. Gruppelaar, E. Menapace 44-Ru-101 4440 JEFF-3.0 NEA RCOM-JUL86 H. Gruppelaar 44-Ru-102 4443 JEFF-3.0 NEA RCOM-JUL82 H. Gruppelaar, E. Menapace 44-Ru-103 4446 JEFF-3.0 NEA RCOM-JUL82 H. Gruppelaar, E. Menapace

39 44-Ru-104 4449 JEFF-3.0 NEA RCOM-JUL82 H. Gruppelaar, E. Menapace 44-Ru-105 4452 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 44-Ru-106 4455 JEFF-3.0 NEA RCOM-JUL82 H. Gruppelaar, E. Menapace 45-Rh-103 4525 New eval. CAD, BRC, + EVAL-FEB05 E. Dupont, E. Bauge, M.C. Moxon 45-Rh-105 4531 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 46-Pd-102 4625 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 46-Pd-104 4631 JEFF-3.0 NEA RCOM-JUL83 H. Gruppelaar, E. Menapace 46-Pd-105 4634 JEFF-3.0 NEA RCOM-JUL83 H. Gruppelaar, E. Menapace 46-Pd-106 4637 JEFF-3.0 NEA RCOM-JUL83 H. Gruppelaar, E. Menapace 46-Pd-107 4640 JEFF-3.0 NEA RCOM-JUL82 H. Gruppelaar, E. Menapace 46-Pd-108 4643 JEFF-3.0 NEA RCOM-JUL83 H. Gruppelaar, E. Menapace 46-Pd-110 4649 JEFF-3.0 NEA RCOM-JUL83 H. Gruppelaar, E. Menapace 47-Ag-107 4725 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 47-Ag-109 4731 JEFF-3.0 NEA RCOM-JUL83 H. Gruppelaar, E. Menapace 47-Ag-110m 4735 JENDL-3.3 JNDC EVAL-MAR90 JNDC FP Nuclear Data WG 47-Ag-111 4737 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 48-Cd-106 4825 JEFF-3.0 UA, ANL, + EVAL-AUG94 J. McCabe, A.B. Smith, +

Table 1. Origin of evaluations in the JEFF-3.1 general-purpose neutron data library (cont.)

Material MAT Data source Laboratory Evaluation Author(s) 48-Cd-108 4831 JEFF-3.0 UA, ANL, + EVAL-AUG94 J. McCabe, A.B. Smith, + 48-Cd-110 4837 JEFF-3.0 UA, ANL EVAL-AUG94 J. McCabe, A.B. Smith 48-Cd-111 4840 JEFF-3.0 UA, ANL EVAL-AUG94 J. McCabe, A.B. Smith 48-Cd-112 4843 JEFF-3.0 UA, ANL, + EVAL-AUG94 J. McCabe, A.B. Smith, + 48-Cd-113 4846 JEFF-3.0 UA, ANL EVAL-AUG94 J. McCabe, A.B. Smith 48-Cd-114 4849 JEFF-3.0 UA, ANL, + EVAL-AUG94 J. McCabe, A.B. Smith, + 48-Cd-115m 4853 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 48-Cd-116 4855 JEFF-3.0 UA, ANL, + EVAL-AUG94 J. McCabe, A.B. Smith, + 49-In-113 4925 JENDL-3.3 JNDC EVAL-MAR90 JNDC FP Nuclear Data WG 49-In-115 4931 JENDL-3.3 JNDC EVAL-MAR90 JNDC FP Nuclear Data WG 50-Sn-112 5025 JENDL-3.3 JNDC EVAL-MAR90 JNDC FP Nuclear Data WG 50-Sn-114 5031 JENDL-3.3 JNDC EVAL-MAR90 JNDC FP Nuclear Data WG 50-Sn-115 5034 JENDL-3.3 JNDC EVAL-MAR90 JNDC FP Nuclear Data WG

40 50-Sn-116 5037 JENDL-3.3 JNDC EVAL-MAR90 JNDC FP Nuclear Data WG 50-Sn-117 5040 JENDL-3.3 JNDC EVAL-MAR90 JNDC FP Nuclear Data WG 50-Sn-118 5043 JENDL-3.3 JNDC EVAL-MAR90 JNDC FP Nuclear Data WG 50-Sn-119 5046 JENDL-3.3 JNDC EVAL-MAR90 JNDC FP Nuclear Data WG 50-Sn-120 5049 JENDL-3.3 JNDC EVAL-MAR90 JNDC FP Nuclear Data WG 50-Sn-122 5055 JENDL-3.3 JNDC EVAL-MAR90 JNDC FP Nuclear Data WG 50-Sn-123 5058 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 50-Sn-124 5061 JENDL-3.3 JNDC EVAL-MAR90 JNDC FP Nuclear Data WG 50-Sn-125 5064 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 50-Sn-126 5067 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 51-Sb-121 5125 JENDL-3.3 JNDC EVAL-AUG89 JNDC FP Nuclear Data WG 51-Sb-123 5131 JENDL-3.3 JNDC EVAL-AUG89 JNDC FP Nuclear Data WG 51-Sb-124 5134 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 51-Sb-125 5137 JEFF-3.0 NEA RCOM-JUL82 H. Gruppelaar, E. Menapace 51-Sb-126 5140 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 52-Te-120 5225 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 52-Te-122 5231 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group

Table 1. Origin of evaluations in the JEFF-3.1 general-purpose neutron data library (cont.)

Material MAT Data source Laboratory Evaluation Author(s) 52-Te-123 5234 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 52-Te-124 5237 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 52-Te-125 5240 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 52-Te-126 5243 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 52-Te-127m 5247 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 52-Te-128 5249 JEFF-3.0 NEA RCOM-JUL83 H. Gruppelaar, E. Menapace 52-Te-129m 5253 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 52-Te-130 5255 JEFF-3.0 NEA RCOM-JUL82 H. Gruppelaar, E. Menapace 52-Te-132 5261 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 53-I-127 5325 New eval. CAD, BRC, + EVAL-FEB05 G. Noguere, E. Dupont, E. Bauge 53-I-129 5331 New eval. CAD, BRC, + EVAL-FEB05 G. Noguere, E. Dupont, E. Bauge 53-I-130 5334 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 53-I-131 5337 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group

41 53-I-135 5349 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 54-Xe-124 5425 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 54-Xe-126 5431 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 54-Xe-128 5437 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 54-Xe-129 5440 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 54-Xe-130 5443 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 54-Xe-131 5446 JEFF-3.0 NEA RCOM-JUL82 H. Gruppelaar 54-Xe-132 5449 JEFF-3.0 NEA RCOM-JUL82 H. Gruppelaar, E. Menapace 54-Xe-133 5452 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 54-Xe-134 5455 JEFF-3.0 NEA RCOM-JUL82 H. Gruppelaar, E. Menapace 54-Xe-135 5458 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 54-Xe-136 5461 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 55-Cs-133 5525 JEFF-3.0 BNL, KAERI, + EVAL-AUG99 S.Y. Oh, S. Mughabghab, R. Schenter 55-Cs-134 5528 JEFF-3.0 NEA RCOM-JUL82 H. Gruppelaar, E. Menapace 55-Cs-135 5531 JEFF-3.0 NEA RCOM-JUL82 H. Gruppelaar, E. Menapace 55-Cs-136 5534 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 55-Cs-137 5537 JEFF-3.0 NEA RCOM-JUL82 H. Gruppelaar, E. Menapace

Table 1. Origin of evaluations in the JEFF-3.1 general-purpose neutron data library (cont.)

Material MAT Data source Laboratory Evaluation Author(s) 56-Ba-130 5625 JENDL-3.3 JNDC EVAL-MAR90 JNDC FP Nuclear Data WG 56-Ba-132 5631 JENDL-3.3 JNDC EVAL-MAR90 JNDC FP Nuclear Data WG 56-Ba-134 5637 JENDL-3.3 JNDC EVAL-MAR90 JNDC FP Nuclear Data WG 56-Ba-135 5640 JENDL-3.3 JNDC EVAL-MAR90 JNDC FP Nuclear Data WG 56-Ba-136 5643 JENDL-3.3 JNDC EVAL-MAR90 JNDC FP Nuclear Data WG 56-Ba-137 5646 JENDL-3.3 JNDC EVAL-MAR90 JNDC FP Nuclear Data WG 56-Ba-138 5649 JENDL-3.3 JNDC EVAL-MAR90 JNDC FP Nuclear Data WG 56-Ba-140 5655 JEFF-3.0 NEA RCOM-JUL82 H. Gruppelaar, E. Menapace 57-La-138 5725 JENDL-3.3 JNDC EVAL-MAR90 JNDC FP Nuclear Data WG 57-La-139 5728 JEFF-3.0 NEA RCOM-JUL83 H. Gruppelaar, E. Menapace 57-La-140 5731 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 58-Ce-140 5837 JEFF-3.0 NEA RCOM-JUL82 H. Gruppelaar 58-Ce-141 5840 JEFF-3.0 NEA RCOM-JUL83 H. Gruppelaar, E. Menapace

42 58-Ce-142 5843 JEFF-3.0 NEA RCOM-JUL83 H. Gruppelaar, E. Menapace 58-Ce-143 5846 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 58-Ce-144 5849 JEFF-3.0 NEA RCOM-JUL83 H. Gruppelaar, E. Menapace 59-Pr-141 5925 JEFF-3.0 NEA RCOM-JUL82 H. Gruppelaar, E. Menapace 59-Pr-142 5928 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 59-Pr-143 5931 JEFF-3.0 NEA RCOM-JUL82 H. Gruppelaar, E. Menapace 60-Nd-142 6025 JEFF-3.0 NEA RCOM-JUL82 Scientific Co-ordination Group 60-Nd-143 6028 JEFF-3.0 NEA RCOM-JUL82 H. Gruppelaar, E. Menapace 60-Nd-144 6031 JEFF-3.0 NEA RCOM-JUL83 H. Gruppelaar, E. Menapace 60-Nd-145 6034 JEFF-3.0 NEA RCOM-JUL83 H. Gruppelaar, E. Menapace 60-Nd-146 6037 JEFF-3.0 NEA RCOM-JUL83 H. Gruppelaar, E. Menapace 60-Nd-147 6040 JEFF-3.0 NEA RCOM-JUL82 H. Gruppelaar, E. Menapace 60-Nd-148 6043 JEFF-3.0 NEA RCOM-JUL82 H. Gruppelaar, E. Menapace 60-Nd-150 6049 JEFF-3.0 NEA RCOM-JUN82 H. Gruppelaar, E. Menapace 61-Pm-147 6149 JEFF-3.0 NEA RCOM-JUN83 H. Gruppelaar, E. Menapace 61-Pm-148 6152 JEFF-3.0 NEA RCOM-JUN82 Scientific Co-ordination Group 61-Pm-148m 6153 JEFF-3.0 NEA RCOM-JUN82 H. Gruppelaar, E. Menapace

Table 1. Origin of evaluations in the JEFF-3.1 general-purpose neutron data library (cont.)

Material MAT Data source Laboratory Evaluation Author(s) 61-Pm-149 6155 JEFF-3.0 NEA RCOM-JUN82 H. Gruppelaar, E. Menapace 61-Pm-151 6161 JEFF-3.0 NEA RCOM-JUN82 Scientific Co-ordination Group 62-Sm-144 6225 JEFF-3.0 NEA RCOM-JUN82 Scientific Co-ordination Group 62-Sm-147 6234 JEFF-3.0 NEA RCOM-JUN83 H. Gruppelaar, E. Menapace 62-Sm-148 6237 JEFF-3.0 NEA RCOM-JUN82 H. Gruppelaar, E. Menapace 62-Sm-149 6240 JEFF-3.0 BNL+KAERI EVAL-AUG99 J.H.Chang and S.F.Mughabghab 62-Sm-150 6243 JEFF-3.0 NEA RCOM-JUN82 H. Gruppelaar, E. Menapace 62-Sm-151 6246 JEFF-3.0 NEA RCOM-JUN82 H. Gruppelaar, E. Menapace 62-Sm-152 6249 JEFF-3.0 NEA RCOM-JUN83 H. Gruppelaar, E. Menapace 62-Sm-153 6252 JEFF-3.0 NEA RCOM-JUN82 Scientific Co-ordination Group 62-Sm-154 6255 JEFF-3.0 NEA RCOM-JUN82 H. Gruppelaar, E. Menapace 63-Eu-151 6325 JEFF-3.0 NEA RCOM-JUN82 H. Gruppelaar, H.CH. Rieffe 63-Eu-152 6328 JEFF-3.0 JNDC EVAL-DEC90 JNDC FP Nuclear Data WG

43 63-Eu-153 6331 JEFF-3.0 JEARI, JNDC EVAL-MAR89 T. Asami, JNDC FP ND WG 63-Eu-154 6334 JEFF-3.0 ORNL, BNL EVAL-MAY89 R.Q. Wright, H. Takahashi 63-Eu-155 6337 JEFF-3.0 ORNL, HEDL EVAL-DEC88 Wright, Prince, Schenter 63-Eu-156 6340 JEFF-3.0 NEA RCOM-JUN82 Scientific Co-ordination Group 63-Eu-157 6343 JEFF-3.0 NEA RCOM-JUN82 Scientific Co-ordination Group 64-Gd-152 6425 JEFF-3.0 JNDC EVAL-MAR90 JNDC FP Nuclear Data WG 64-Gd-154 6431 JEFF-3.0 NEA RCOM-JUN82 E. Menapace, et al. 64-Gd-155 6434 JEFF-3.0 NEA RCOM-JUL86 Scientific Co-ordination Group 64-Gd-156 6437 JEFF-3.0 NEA RCOM-JUN82 E. Menapace, et al. 64-Gd-157 6440 JEFF-3.0 NEA RCOM-JUN82 E. Menapace, et al. 64-Gd-158 6443 JEFF-3.0 NEA RCOM-JUN82 E. Menapace, et al. 64-Gd-160 6449 JEFF-3.0 NEA RCOM-JUN82 E. Menapace, et al. 65-Tb-159 6525 JEFF-3.0 NEA RCOM-JUN83 H. Gruppelaar, E. Menapace 65-Tb-160 6528 JEFF-3.0 NEA RCOM-JUN82 Scientific Co-ordination Group 66-Dy-160 6637 JEFF-3.0 NEA RCOM-JUN82 Scientific Co-ordination Group 66-Dy-161 6640 ENDF/B-VI.8 ORNL, HEDL EVAL-OCT74 Wright, Schenter, Schmittroth 66-Dy-162 6643 ENDF/B-VI.8 ORNL, HEDL EVAL-OCT74 Wright, Schenter, Schmittroth

Table 1. Origin of evaluations in the JEFF-3.1 general-purpose neutron data library (cont.)

Material MAT Data source Laboratory Evaluation Author(s) 66-Dy-163 6646 ENDF/B-VI.8 ORNL, HEDL EVAL-OCT74 Wright, Schenter, Schmittroth 66-Dy-164 6649 ENDF/B-VI.8 ORNL, BNW EVAL-JUN67 Wright, Leonard, Stewart 67-Ho-165 6725 ENDF/B-VI.8 LANL EVAL-APR88 P.G. Young, E.D. Arthur 68-Er-162 6825 JENDL-3.3 TIT EVAL-SEP00 A.K.M. Harun, A.R. Rashid, + 68-Er-164 6831 JENDL-3.3 TIT EVAL-SEP00 A.K.M. Harun, A.R. Rashid, + 68-Er-166 6837 JENDL-3.3 TIT EVAL-SEP00 A.K.M. Harun, A.R. Rashid, + 68-Er-167 6840 JENDL-3.3 TIT EVAL-SEP00 A.K.M. Harun, A.R. Rashid, + 68-Er-168 6843 JENDL-3.3 TIT EVAL-SEP00 A.K.M. Harun, A.R. Rashid, + 68-Er-170 6849 JENDL-3.3 TIT EVAL-SEP00 A.K.M. Harun, A.R. Rashid, + 71-Lu-175 7125 ENDF/B-VI.8 ORNL, BNW EVAL-MAR98 R.Q. Wright, Leonard-Stewart 71-Lu-176 7128 ENDF/B-VI.8 ORNL, BNW EVAL-MAR98 R.Q. Wright, Leonard-Stewart 72-Hf-174 7225 JENDL-3.3 + New eval. NAIG+ EVAL-JAN05 G. Noguere (CAD) 72-Hf-176 7231 JENDL-3.3 + New eval. NAIG+ EVAL-JAN05 G. Noguere (CAD)

44 72-Hf-177 7234 JENDL-3.3 + New eval. NAIG+ EVAL-JAN05 G. Noguere (CAD) 72-Hf-178 7237 JENDL-3.3 + New eval. NAIG+ EVAL-JAN05 G. Noguere (CAD) 72-Hf-179 7240 JENDL-3.3 + New eval. NAIG+ EVAL-JAN05 G. Noguere (CAD) 72-Hf-180 7243 JENDL-3.3 + New eval. NAIG+ EVAL-JAN05 G. Noguere (CAD) 73-Ta-181 7328 JENDL-3.3 NAIG EVAL-MAR87 N. Yamamuro 73-Ta-182 7331 JEFF-3.0 NEA RCOM-JUN82 Scientific Co-ordination Group 74-W-182 7431 JENDL-3.3 KHI, NEDAC EVAL-MAR87 T. Watanabe (KHI), T. Asami (NEDAC) 74-W-183 7434 JENDL-3.3 KHI, NEDAC EVAL-MAR87 T. Watanabe (KHI), T. Asami (NEDAC) 74-W-184 7437 JENDL-3.3 KHI, NEDAC EVAL-MAR87 T. Watanabe (KHI), T. Asami (NEDAC) 74-W-186 7443 JENDL-3.3 KHI, NEDAC EVAL-MAR87 T. Watanabe (KHI), T. Asami (NEDAC) 75-Re-185 7525 JEFF-3.0 ORNL, LANL EVAL-MAR90 L.W. Weston, P.G. Young 75-Re-187 7531 JEFF-3.0 ORNL, LANL EVAL-MAR90 L.W. Weston, P.G. Young 76-Os-0 7600 JEFF-3.0 + Corrections FEI EVAL-JAN90 M.N. Nikolaev 77-Ir-191 7725 ENDF/B-VI.8 ORNL EVAL-MAR95 R.Q. Wright, R.R. Spencer 77-Ir-193 7731 ENDF/B-VI.8 ORNL EVAL-MAR95 R.Q. Wright, R.R. Spencer 78-Pt-0 7800 JEFF-3.0 + Corrections NEA, LLNL EVAL-OCT83 R.J. Howerton 79-Au-197 7925 JEFF-3.0 LANL EVAL-JAN84 P.G. Young

Table 1. Origin of evaluations in the JEFF-3.1 general-purpose neutron data library (cont.)

Material MAT Data source Laboratory Evaluation Author(s) 80-Hg-196 8025 JENDL-3.3 JNDC EVAL-SEP97 K. Shibata, T. Fukahori, S. Chiba, + 80-Hg-198 8031 JENDL-3.3 JNDC EVAL-SEP97 K. Shibata, T. Fukahori, S. Chiba, + 80-Hg-199 8034 JENDL-3.3 JNDC EVAL-SEP97 K. Shibata, T. Fukahori, S. Chiba, + 80-Hg-200 8037 JENDL-3.3 JNDC EVAL-SEP97 K. Shibata, T. Fukahori, S. Chiba, + 80-Hg-201 8040 JENDL-3.3 JNDC EVAL-SEP97 K. Shibata, T. Fukahori, S. Chiba, + 80-Hg-202 8043 JENDL-3.3 JNDC EVAL-SEP97 K. Shibata, T. Fukahori, S. Chiba, + 80-Hg-204 8049 JENDL-3.3 JNDC EVAL-SEP97 K. Shibata, T. Fukahori, S. Chiba, + 81-Tl-0 8100 JEFF-3.0 SIU EVAL-DEC93 Zhou Yiming, Ma Gonggui 82-Pb-204 8225 New eval. NRG EVAL-DEC04 A.J. Koning 82-Pb-206 8231 New eval. NRG EVAL-DEC04 A.J. Koning 82-Pb-207 8234 New eval. NRG EVAL-DEC04 A.J. Koning 82-Pb-208 8237 New eval. NRG EVAL-DEC04 A.J. Koning 83-Bi-209 8325 New eval. NRG EVAL-DEC04 A.J. Koning

45 88-Ra-223 8825 JENDL-3.3 TIT EVAL-AUG88 N. Takagi 88-Ra-224 8828 JENDL-3.3 TIT EVAL-AUG88 N. Takagi 88-Ra-225 8831 JENDL-3.3 TIT EVAL-AUG88 N. Takagi 88-Ra-226 8834 JENDL-3.3 TIT EVAL-AUG88 N. Takagi 89-Ac-225 8925 JENDL-3.3 TIT EVAL-AUG88 N. Takagi 89-Ac-226 8928 JENDL-3.3 TIT EVAL-AUG88 N. Takagi 89-Ac-227 8931 JENDL-3.3 TIT EVAL-AUG88 N. Takagi 90-Th-227 9025 JENDL-3.3 TIT EVAL-AUG88 N. Takagi 90-Th-228 9028 JENDL-3.3 KINKI U. EVAL-JUN87 T. Ohsawa 90-Th-229 9031 JEFF-3.0 TIT, NEA EVAL-AUG88 N. Takagi 90-Th-230 9034 JEFF-3.0 NEA RCOM-JUN82 Scientific Co-ordination Group 90-Th-232 9040 ENDF/B-VI.8 + New eval. MINSK EVAL-JUN01 V.M. Maslov, et al. 90-Th-233 9043 JENDL-3.3 KINKI U. EVAL-JUL87 T. Ohsawa 90-Th-234 9046 JENDL-3.3 KINKI U. EVAL-JUL87 T. Ohsawa 91-Pa-231 9131 JEFF-3.0 NEA RCOM-JUN82 Scientific Co-ordination Group 91-Pa-232 9134 ENDF/B-VI.8 ORNL, TIT EVAL-DEC99 R.Q. Wright, N. Takagi 91-Pa-233 9137 JEFF-3.0 NEA RCOM-JUN82 Scientific Co-ordination Group

Table 1. Origin of evaluations in the JEFF-3.1 general-purpose neutron data library (cont.)

Material MAT Data source Laboratory Evaluation Author(s) 92-U-232 9219 JEFF-3.0 NEA RCOM-JUN82 Scientific Co-ordination Group 92-U-233 9222 NDL 3.3 + ENDF/B-VI.4 SAEI, + Eval-MAR00 T. Mutsunobu, T. Kawano 92-U-234 9225 New eval. MINSK, + EVAL-SEP02 V.M. Maslov, et al. 92-U-235 9228 JEFF-3.0 ORNL, LANL, + EVAL-NOV89 Weston, Young, Poenitz, Lubitz 92-U-236 9231 New eval. BRC, + EVAL-NOV04 Lopez, Jimenez, Morillon, Romain 92-U-237 9234 New eval. BRC ,+ EVAL-NOV04 Lopez, Jimenez, Morillon, Romain 92-U-238 9237 New eval. BRC, ORNL, + EVAL-OCT04 Lopez, Jimenez, Morillon, Romain 93-Np-235 9340 JENDL-3.3 JAERI EVAL-MAR95 T. Nakagawa 93-Np-236 9343 JENDL-3.3 JAERI EVAL-MAR93 T. Nakagawa 93-Np-237 9346 JENDL 3.3 JAERI EVAL-Jan01 T. Nakagawa, O. Iwamoto 93-Np-238 9349 JEFF-3.0 NEA RCOM-JUN82 Scientific Co-ordination Group 93-Np-239 9352 JEFF-3.0 ORNL EVAL-DEC88 R.Q. Wright 94-Pu-236 9428 JENDL-3.3 JAERI EVAL-FEB02 O. Iwamoto

46 94-Pu-237 9431 JEFF-3.0 NEA RCOM-JUN82 Scientific Co-ordination Group 94-Pu-238 9434 JEFF-3.0 MAPI, JAERI EVAL-MAR89 T. Kawakita, T. Nakagawa 94-Pu-239 9437 JEFF-3.0 BRC, CAD, + EVAL: ROMA In, Morillon, Dos Santos, Uzarralde 94-Pu-240 9440 JEFF-3.0 + New eval. BRC, CAD EVAL-JUL04 Bouland, Derrien, Morillon, Romain 94-Pu-241 9443 JEFF-3.0 JAERI EVAL-OCT87 Y. Kikuchi, N. Sekine,T. Nakagawa 94-Pu-242 9446 JEFF-3.0 ENEA, IJS EVAL-OCT98 A. Ventura, S. Masetti, A. Trkov 94-Pu-243 9449 JEFF-3.0 NEA RCOM-JUN82 Scientific Co-ordination Group 94-Pu-244 9452 JEFF-3.0 NEA RCOM-JUN82 Scientific Co-ordination Group 94-Pu-246 9458 JENDL-3.3 JAERI EVAL-MAR95 T. Nakagawa 95-Am-241 9543 VALUATION CAD, NEA, + EVAL-APR05 Bouland, Bernard, Rugama, et al. 95-Am-242g 9546 JENDL-3.3 MINSK, + EVAL-FEB97 V.M. Maslov, et al. 95-Am-242m 9547 JENDL-3.3 MINSK, BYEL EVAL-DEC96 V.M. Maslov, et al. 95-Am-243 9549 JENDL-3.3 MINSK, JAERI EVAL-JAN02 V.M. Maslov, +, T. Nakagawa 95-Am-244 9552 JENDL-3.3 JAERI EVAL-MAR88 T. Nakagawa 95-Am-244m 9553 JENDL-3.3 JAERI EVAL-MAR88 T. Nakagawa 96-Cm-240 9625 JENDL-3.3 JAERI EVAL-OCT95 T. Nakagawa, T. Liu 96-Cm-241 9628 JEFF-3.0 NEA RCOM-JUN82 Scientific Co-ordination Group

Table 1. Origin of evaluations in the JEFF-3.1 general-purpose neutron data library (cont.)

Material MAT Data source Laboratory Evaluation Author(s) 96-Cm-242 9631 JEFF-3.0 NEA RCOM-JUN82 E. Menapace 96-Cm-243 9634 JEFF-3.0 JAERI EVAL-MAR89 T. Nakagawa 96-Cm-244 9637 ENDF/B-VI.8 HEDL, SRL, + EVAL-APR78 Mann, Benjamin, Howerton, et al. 96-Cm-245 9640 JENDL-3.3 MINSK,BYEL EVAL-NOV95 V.M. Maslov, et al. 96-Cm-246 9643 JEFF-3.0 MINSK BYEL EVAL-FEB96 96-Cm-247 9646 JEFF-3.0 JAERI EVAL-MAR89 T. Nakagawa, Y. Kikuchi 96-Cm-248 9649 JEFF-3.0 JAERI EVAL-MAR84 Y. Kikuchi, T. Nakagawa 96-Cm-249 9652 JENDL-3.3 JAERI EVAL-OCT95 T. Nakagawa, T. Liu 96-Cm-250 9655 JENDL-3.3 JAERI EVAL-OCT95 T. Nakagawa, T. Liu 97-Bk-247 9746 JENDL-3.3 JAERI EVAL-JUN95 T. Nakagawa, T. Liu 97-Bk-249 9752 JENDL-3.3 JAERI EVAL-MAR85 Y. Kikuchi, T. Nakagawa 97-Bk-250 9755 JEFF-3.0 JAERI EVAL-MAR87 T. Nakagawa 98-Cf-249 9852 JENDL-3.3 JAERI EVAL-JUL95 T. Nakagawa, T. Liu

47 98-Cf-250 9855 JENDL-3.3 JAERI EVAL-MAR86 T. Nakagawa 98-Cf-251 9858 JEFF-3.0 NEA RCOM-JUN82 Scientific Co-ordination Group 98-Cf-252 9861 JEFF-3.0 NEA RCOM-JUN82 Scientific Co-ordination Group 98-Cf-254 9867 JEFF-3.0 TIT EVAL-AUG87 N. Takagi 99-Es-253 9913 JEF-2.2 NEA RCOM-JUN82 Scientific Co-ordination Group 99-Es-254 9914 JENDL-3.3 TIT EVAL-AUG87 N. Takagi 99-Es-255 9916 JENDL-3.3 TIT EVAL-AUG87 N. Takagi 100-Fm-255 9936 JENDL-3.3 TIT EVAL-AUG87 N. Takagi

Table 2. Origin of evaluations in the JEFF-3.1 thermal scattering law library

Material MAT Data source Laboratory Evaluation Author(s)

H(H2O) 1 New eval. IKE EVAL-JAN04 J. Keinert, M. Mattes H(ZrH) 7 New eval. IKE EVAL-JAN05 J. Keinert

H(CaH2) 8 New eval. CEA, ILL EVAL-OCT04 O. Serot (CEA Cadarache)

D(D2O) 11 New eval. IKE EVAL-FEB04 J. Keinert, Mattes 4Be 26 JEFF-3.0 IKE EVAL-DEC89 J. Keinert, M. Mattes Graphite 31 New eval. IKE EVAL-JAN05 J. Keinert, M. Mattes

H(CH2) 37 JEFF-3.0 IKE EVAL-SEP81 J. Keinert, M. Mattes 24Mg 52 New eval. CEA EVAL-SEP03 C. Mounier

Ca(CaH2) 59 New eval. CEA, ILL EVAL-OCT04 O. Serot (CEA Cadarache)

Table 3. Contents of the JEFF-3.1 general-purpose neutron data library. List of available reactions (MTs) in files MF=3 to MF=6.

Z Sym-A MF=1 MF=3 MF=4 MF=5 MF=6 1 1H 451 1, 2, 102 2 1 2H 451 1-3, 16, 102 2 16 1 3H 451 1, 2, 16, 17 2, 16 16 17 2 3He 451 1, 2, 102-104 2 2 4He 451 1, 2 2 3 6Li 451 1, 2, 4, 24, 51-81, 102, 103, 105 2, 24, 51-81, 105 24 3 7Li 451 1, 2, 4, 16, 24, 25, 28, 51-53, 91, 102, 104 2, 51, 52 16, 24, 25, 28, 53, 91 4 9Be 451 1, 2, 102-105, 107, 875-890 2 107, 875-890 5 10B 451 1, 2, 4, 51-85, 102-104, 107, 113, 600-605, 800, 2, 51-85 801 5 11B 451 1, 2, 4, 16, 22, 28, 51-60, 91, 102, 103, 105, 107 2, 51-60 16, 22, 28, 91, 103, 107 6 natC 451 1-5, 28, 51-62, 91, 102-104, 107 2, 28, 51-62, 91 28, 91 5 7 14N 451 1, 2, 4, 5, 16, 51-77, 102-105, 107, 108, 600-606, 2, 51-77 5, 16 650-653, 700, 701, 800-810 15 49 7 N 451 1, 2, 4, 16, 22, 28, 51-57, 91, 102-105, 107 2, 16, 22, 28, 51-57, 91 16, 22, 28, 91 8 16O 451 1-5, 16, 22, 23, 28, 32, 41, 44, 45, 51-57, 91, 2, 51-57, 600-603, 650-669, 5, 16, 22, 23, 28, 32, 41, 44, 45, 91, 108, 102-105, 107, 108, 112, 600-603, 650-669, 700-709, 800-803 112, 749 700-709, 749, 800-803 8 17O 451 1, 2, 4, 16, 22, 28, 51-62, 91, 102-104, 107 2, 16, 22, 28, 51-62, 91 16, 22, 28, 91 9 19F 451 1-4, 16, 22, 28, 51-71, 91, 102-105, 107 2, 51-71 16, 22, 28, 91, 103, 107 11 22Na 451 1, 2, 4, 16, 22, 28, 51-61, 91, 102-104, 107, 108, 2, 16, 22, 28, 51-61, 91 16, 22, 28, 91 251 11 23Na 451 1, 2, 4, 16, 22, 28, 51-77, 91, 102, 103, 107, 251 2, 16, 22, 28, 51-77, 91 16, 22, 28, 91 12 24Mg 451 1, 2, 4, 16, 22, 28, 51-61, 91, 102, 103, 107 2, 16, 22, 28, 51-61, 91 16, 22, 28, 91 12 25Mg 451 1, 2, 4, 16, 22, 28, 51-67, 91, 102, 103, 107 2, 16, 22, 28, 51-67, 91 16, 22, 28, 91 12 26Mg 451 1, 2, 4, 16, 22, 28, 51-63, 91, 102, 103, 107 2, 16, 22, 28, 51-63, 91 16, 22, 28, 91 13 27Al 451 1-5, 16, 22, 28, 32, 33, 45, 51-89, 91, 102-105, 2 5, 16, 22, 28, 32, 33, 45, 51-89, 91, 108, 111, 107, 108, 111, 112, 117, 600-619, 649-669, 112, 117, 600-619, 649-669, 699-710, 749, 699-710, 749, 800-819, 849 800-819, 849 14 28Si 451 1-4, 16, 22, 28, 51-67, 91, 102-107, 111, 600-613, 2 16, 22, 28, 51-67, 91, 600-613, 649, 649, 800-815, 849 800-815, 849 14 29Si 451 1, 2, 4, 16, 22, 28, 51-79, 91, 102, 103, 107, 111, 2, 51-79 16, 22, 28, 91, 203, 207 203, 207 14 30Si 451 1, 2, 4, 16, 22, 28, 51-69, 91, 102, 103, 107, 111, 2, 51-69 16, 22, 28, 91, 203, 207 203, 207

Table 3. Contents of the JEFF-3.1 general-purpose neutron data library. List of available reactions (MTs) in files MF=3 to MF=6 (cont.).

Z Sym-A MF=1 MF=3 MF=4 MF=5 MF=6 15 31P 451 1, 2, 4, 16, 22, 28, 51-56, 91, 102, 103, 107 2, 16, 22, 28, 51-56, 91 16, 22, 28, 91 16 32S 451 1, 2, 4, 16, 22, 28, 51-56, 91, 102, 103, 107 2, 16, 22, 28, 51-56, 91 16, 22, 28, 91 16 33S 451 1, 2, 4, 16, 22, 28, 51-57, 91, 102, 103, 107 2, 16, 22, 28, 51-57, 91 16, 22, 28, 91 16 34S 451 1, 2, 4, 16, 22, 28, 51-55, 91, 102, 103, 107 2, 16, 22, 28, 51-55, 91 16, 22, 28, 91 16 36S 451 1, 2, 4, 16, 22, 28, 51-55, 91, 102, 103, 107 2, 16, 22, 28, 51-55, 91 16, 22, 28, 91 17 35Cl 451 1, 2, 4, 16, 22, 28, 32, 51-80, 91, 102-105, 107, 2 16, 22, 28, 32, 51-80, 91, 111, 112, 600-629, 111, 112, 600-629, 649-680, 699-730, 749, 649-680, 699-730, 749, 800-820, 849 800-820, 849 17 37Cl 451 1, 2, 4, 16, 17, 22, 28, 32, 51-77, 91, 102-105, 107, 2 16, 17, 22, 28, 32, 51-77, 91, 649-661, 649-661, 699-715, 749, 800-805, 849 699-715, 749, 800-805, 849 18 36Ar 451 1, 2, 4, 16, 22, 28, 51-71, 91, 102-104, 107, 108, 2, 16, 22, 28, 51-71, 91 16, 22, 28, 91 251, 252 18 38Ar 451 1, 2, 4, 16, 22, 28, 51-69, 91, 102-104, 107, 108, 2, 16, 22, 28, 51-69, 91 16, 22, 28, 91 251, 252 18 40Ar 451 1, 2, 4, 16, 17, 22, 28, 51-75, 91, 102-107 2, 16, 17, 22, 28, 51-75, 91 16, 17, 22, 28, 91 19 39K 451 1, 2, 4, 16, 22, 28, 51-54, 91, 102, 103, 107 2, 16, 22, 28, 51-54, 91 16, 22, 28, 91 50 19 40K 451 1, 2, 4, 16, 22, 28, 51-55, 91, 102, 103, 107 2, 16, 22, 28, 51-55, 91 16, 22, 28, 91 19 41K 451 1, 2, 4, 16, 22, 28, 51, 52, 91, 102, 103, 107 2, 16, 22, 28, 51, 52, 91 16, 22, 28, 91 20 40Ca 451 1-5, 16, 22, 28, 29, 44, 45, 51-70, 91, 102-108, 2 5, 16, 22, 28, 29, 44, 45, 51-70, 91, 102, 108, 111, 112, 115, 117, 600-610, 649-655, 699-705, 111, 112, 115, 117, 600-610, 649-655, 749-755, 799-810, 849 699-705, 749-755, 799-810, 849, 20 42Ca 451 1-5, 16, 22, 28, 29, 32, 51-70, 91, 102-108, 111, 2 5, 16, 22, 28, 29, 32, 51-70, 91, 102, 108, 112, 115, 117, 600-610, 649-655, 699-705, 111, 112, 115, 117, 600-610, 649-655, 749-755, 799-810, 849 699-705, 749-755, 799-810, 849 20 43Ca 451 1-5, 16, 17, 22, 24, 28, 32, 41, 51-70, 91, 102-108, 2 5, 16, 17, 22, 24, 28, 32, 41, 51-70, 91, 102, 111, 112, 600-610, 649-655, 699-705, 749-755, 108, 111, 112, 600-610, 649-655, 699-705, 799-810, 849 749-755, 799-810, 849 20 44Ca 451 1-5, 16, 17, 22, 28, 51-70, 91, 102-108, 111, 2 5, 16, 17, 22, 28, 51-70, 91, 102, 108, 111, 600-610, 649-655, 699-705, 749-752, 800-810, 600-610, 649-655, 699-705, 749-752, 849 800-810, 849 20 46Ca 451 1-5, 16, 17, 22, 28, 51-70, 91, 102-105, 107, 2 5, 16, 17, 22, 28, 51-70, 91, 102, 600-610, 600-610, 649-655, 699-705, 749, 800-804, 849 649-655, 699-705, 749, 800-804, 849 20 48Ca 451 1-5, 16, 17, 22, 28, 51-70, 91, 102-105, 107, 2 5, 16, 17, 22, 28, 51-70, 91, 102, 600-604, 600-604, 649-655, 699-705, 749, 800-803, 849 649-655, 699-705, 749, 800-803, 849 21 45Sc 451 1-5, 16, 22, 28, 32, 33, 41, 51-70, 91, 102-108, 2 5, 16, 22, 28, 32, 33, 41, 51-70, 91, 102, 108, 111, 112, 117, 600-610, 649-655, 699-705, 111, 112, 117, 600-610, 649-655, 699-705, 749-755, 799-810, 849 749-755, 799-810, 849

Table 3. Contents of the JEFF-3.1 general-purpose neutron data library. List of available reactions (MTs) in files MF=3 to MF=6 (cont.).

Z Sym-A MF=1 MF=3 MF=4 MF=5 MF=6 22 46Ti 451 1-4, 16, 22, 28, 51-70, 91, 102-107, 111, 600-610, 2 16, 22, 28, 51-70, 91, 102, 111, 600-610, 649-655, 699-705, 749-755, 799-810, 849 649-655, 699-705, 749-755, 799-810, 849 22 47Ti 451 1-4, 16, 22, 28, 51-70, 91, 102-107, 111, 600-610, 2 16, 22, 28, 51-70, 91, 102, 111, 600-610, 649-655, 699-705, 749-755, 799-810, 849 649-655, 699-705, 749-755, 799-810, 849 22 48Ti 451 1-4, 16, 22, 28, 51-70, 91, 102-107, 111, 600-610, 2 16, 22, 28, 51-70, 91, 102, 111, 600-610, 649-655, 699-705, 749-755, 799-810, 849 649-655, 699-705, 749-755, 799-810, 849 22 49Ti 451 1-4, 16, 22, 28, 51-70, 91, 102-107, 111, 600-610, 2 16, 22, 28, 51-70, 91, 102, 111, 600-610, 649-655, 699-705, 749-753, 799-810, 849 649-655, 699-705, 749-753, 799-810, 849 22 50Ti 451 1-4, 16, 22, 28, 51-70, 91, 102-107, 600-610, 2 16, 22, 28, 51-70, 91, 102, 600-610, 649-655, 699-705, 749-751, 800-810, 849 649-655, 699-705, 749-751, 800-810, 849 23 natV 451 1, 2, 4, 16, 22, 28, 32, 51-74, 91, 102-107, 111, 2, 32, 51-74 32 16, 22, 28, 91, 103, 107 112 24 50Cr 451 1-4, 16, 22, 28, 51-56, 91, 102-104, 107 2 16, 22, 28, 51-56, 91, 103, 107 24 52Cr 451 1-4, 16, 22, 28, 51-60, 91, 102-107 2 16, 22, 28, 51-60, 91, 103, 107 24 53Cr 451 1-4, 16, 22, 28, 51-63, 91, 102, 103, 107 2 16, 22, 28, 51-63, 91, 103, 107 24 54Cr 451 1-4, 16, 51-54, 91, 102, 103, 107 2 16, 51-54, 91, 103, 107 25 55Mn 451 1, 2, 4, 16, 22, 28, 51-66, 91, 102-107, 203-207 2, 51-66 16, 22, 28, 91, 203-207 51 26 54Fe 451 1-5, 16, 22, 28, 44, 51-70, 91, 102-108, 111, 112, 2 5, 16, 22, 28, 44, 51-70, 91, 102, 108, 111, 115, 600-610, 649-655, 699-705, 749-755, 112, 115, 600-610, 649-655, 699-705, 799-810, 849 749-755, 799-810, 849 26 56Fe 451 1-5, 16, 22, 28, 51-82, 91, 102-107, 600-613, 649, 2, 51-82, 600-613, 800-810 5, 16, 22, 28, 91, 649, 849 800-810, 849 26 57Fe 451 1-5, 16, 17, 22, 24, 28, 32, 41, 51-70, 91, 102-108, 2 5, 16, 17, 22, 24, 28, 32, 41, 51-70, 91, 102, 111, 112, 600-610, 649-655, 699-705, 749-755, 108, 111, 112, 600-610, 649-655, 699-705, 799-810, 849 749-755, 799-810, 849 26 58Fe 451 1-5, 16, 17, 22, 28, 51-70, 91, 102-108, 600-610, 2 5, 16, 17, 22, 28, 51-70, 91, 102, 108, 649-655, 699-705, 749-752, 800-810, 849 600-610, 649-655, 699-705, 749-752, 800-810, 849 27 58Co 451 1, 2, 4, 16, 22, 28, 51-62, 91, 102-104, 107, 251 2, 16, 22, 28, 51-62, 91 16, 22, 28, 91 27 58mCo 451 1, 2, 4, 16, 22, 28, 51-62, 91, 102-104, 107, 251 2, 16, 22, 28, 51-62, 91 16, 22, 28, 91 27 59Co 451 1, 2, 4, 16, 22, 28, 32, 33, 51-69, 91, 102-108, 111, 2, 16, 22, 28, 32, 33, 51-69, 91 16, 22, 28, 32, 33, 91 112 28 58Ni 451 1-4, 16, 22, 28, 51-58, 91, 102-107, 112 2 16, 22, 28, 51-58, 91, 103, 107 28 59Ni 451 1-3, 16, 22, 28, 32, 34, 102-107, 111 2, 16, 22, 28, 32, 34 16, 22, 28, 32, 34 28 60Ni 451 1-4, 16, 22, 28, 51-61, 91, 102-107 2 16, 22, 28, 51-61, 91, 103, 107 28 61Ni 451 1-5, 16, 28, 51-58, 91, 102, 103, 107, 111 2 5, 16, 28, 51-58, 91, 103, 107 28 62Ni 451 1-5, 16, 22, 28, 51-54, 91, 102-104, 107 2 5, 16, 22, 28, 51-54, 91, 103, 107

Table 3. Contents of the JEFF-3.1 general-purpose neutron data library. List of available reactions (MTs) in files MF=3 to MF=6 (cont.).

Z Sym-A MF=1 MF=3 MF=4 MF=5 MF=6 28 64Ni 451 1-5, 16, 22, 28, 51, 52, 91, 102-104, 107 2 5, 16, 22, 28, 51, 52, 91, 103, 107 29 63Cu 451 1-5, 16, 22, 28, 51-72, 91, 102-104, 106, 107 2 5, 16, 22, 28, 51-72, 91, 103, 107 29 65Cu 451 1-5, 16, 22, 28, 51-63, 91, 102-107 2 5, 16, 22, 28, 51-63, 91, 103, 107 30 natZn 451 1, 2, 4, 16, 17, 22, 28, 51, 91, 102, 103, 105, 107, 2, 16, 17, 22, 28, 51, 91 16, 17, 22, 28, 91 251-253 31 natGa 451 1, 2, 4, 16, 17, 22, 28, 32, 51-88, 91, 102-107, 111, 2, 16, 17, 22, 28, 32, 51-88, 91 16, 17, 22, 28, 32, 91 251 32 70Ge 451 1-5, 16, 22, 24, 28, 29, 32, 44, 45, 51-70, 91, 2 5, 16, 22, 24, 28, 29, 32, 44, 45, 51-70, 91, 102-108, 111, 112, 115, 117, 600-610, 649-655, 102, 108, 111, 112, 115, 117, 600-610, 699-705, 749-755, 799-810, 849 649-655, 699-705, 749-755, 799-810, 849 32 72Ge 451 1-5, 16, 17, 22, 24, 28, 32, 51-70, 91, 102-108, 2 5, 16, 17, 22, 24, 28, 32, 51-70, 91, 102, 108, 111, 112, 600-610, 649-655, 699-705, 749-755, 111, 112, 600-610, 649-655, 699-705, 799-810, 849 749-755, 799-810, 849 32 73Ge 451 1-5, 16, 17, 22, 24, 28, 32, 33, 41, 51-70, 91, 2 5, 16, 17, 22, 24, 28, 32, 33, 41, 51-70, 91, 102-108, 111, 112, 600-610, 649-655, 699-705, 102, 108, 111, 112, 600-610, 649-655, 749-755, 799-810, 849 699-705, 749-755, 799-810, 849 32 74Ge 451 1-5, 16, 17, 22, 28, 32, 51-70, 91, 102-107, 111, 2 5, 16, 17, 22, 28, 32, 51-70, 91, 102, 111,

52 600-610, 649-655, 699-705, 749-754, 800-810, 600-610, 649-655, 699-705, 749-754, 849 800-810, 849 32 76Ge 451 1-5, 16, 17, 22, 28, 51-70, 91, 102-105, 107, 2 5, 16, 17, 22, 28, 51-70, 91, 102, 600-610, 600-610, 649-655, 699-705, 749, 800-804, 849 649-655, 699-705, 749, 800-804, 849 33 75As 451 1, 2, 4, 51-63, 91, 102-107, 111, 251-253 2, 51-63, 91 91 34 74Se 451 1, 2, 4, 91, 102-107, 111, 251-253 2, 91 91 34 76Se 451 1, 2, 4, 51-54, 91, 102-107, 111, 251-253 2, 51-54, 91 91 34 77Se 451 1, 2, 4, 51-60, 91, 102-107, 111, 251-253 2, 51-60, 91 91 34 78Se 451 1, 2, 4, 51-59, 91, 102-107, 111, 251-253 2, 51-59, 91 91 34 79Se 451 1, 2, 4, 16, 17, 22, 28, 32, 51-79, 91, 102-107 2, 16, 17, 22, 28, 32, 51-79, 91 16, 17, 22, 28, 32, 91 34 80Se 451 1, 2, 4, 51-58, 91, 102-107, 111, 251-253 2, 51-58, 91 91 34 82Se 451 1, 2, 4, 51-53, 91, 102-107, 111, 251-253 2, 51-53, 91 91 35 79Br 451 1, 2, 4, 51-61, 91, 102-107, 111, 251-253 2, 51-61, 91 91 35 81Br 451 1, 2, 4, 51-54, 91, 102-107, 111, 251-253 2, 51-54, 91 91 36 78Kr 451 1, 2, 4, 16, 51-53, 91, 102-107, 111, 251-253 2, 16, 51-53, 91 16, 91 36 80Kr 451 1, 2, 4, 16, 51-56, 91, 102-107, 111, 251-253 2, 16, 51-56, 91 16, 91 36 82Kr 451 1, 2, 4, 16, 51-59, 91, 102-107, 111, 251-253 2, 16, 51-59, 91 16, 91 36 83Kr 451 1, 2, 4, 16, 17, 51-56, 91, 102-107, 111, 251-253 2, 16, 17, 51-56, 91 16, 17, 91

Table 3. Contents of the JEFF-3.1 general-purpose neutron data library. List of available reactions (MTs) in files MF=3 to MF=6 (cont.).

Z Sym-A MF=1 MF=3 MF=4 MF=5 MF=6 36 84Kr 451 1, 2, 4, 16, 51-63, 91, 102-107, 111, 251-253 2, 16, 51-63, 91 16, 91 36 85Kr 451 1, 2, 4, 51, 91, 102-107, 111 2, 51, 91 91 36 86Kr 451 1, 2, 4, 16, 17, 51-64, 91, 102-107, 111, 251-253 2, 16, 17, 51-64, 91 16, 17, 91 37 85Rb 451 1, 2, 4, 51-53, 91, 102-107, 111, 251-253 2, 51-53, 91 91 37 86Rb 451 1, 2, 4, 51, 91, 102-107, 111 2, 51, 91 91 37 87Rb 451 1, 2, 4, 51, 52, 91, 102-107, 111, 251-253 2, 51, 52, 91 91 38 84Sr 451 1, 2, 4, 91, 102-107, 111 2, 91 91 38 86Sr 451 1, 2, 4, 51-57, 91, 102-107, 111, 251-253 2, 51-57, 91 91 38 87Sr 451 1, 2, 4, 51-57, 91, 102-107, 111, 251-253 2, 51-57, 91 91 38 88Sr 451 1, 2, 4, 51-55, 91, 102-107, 111, 251-253 2, 51-55, 91 91 38 89Sr 451 1, 2, 4, 51-62, 91, 102-107, 111 2, 51-62, 91 91 38 90Sr 451 1, 2, 4, 51-54, 91, 102-107, 111 2, 51-54, 91 91 39 89Y 451 1, 2, 4, 16, 22, 28, 51-62, 91, 102-107 2, 16, 22, 28, 51-62, 91 16, 22, 28, 91 90 53 39 Y 451 1, 2, 4, 51, 52, 91, 102-107, 111 2, 51, 52, 91 91 39 91Y 451 1, 2, 4, 51-58, 91, 102, 104-107, 111 2, 51-58, 91 91 40 90Zr 451 1-4, 16, 22, 28, 32, 51-57, 91, 102-107, 111, 2, 51-57 16, 22, 28, 32, 91, 203-207 203-207 40 91Zr 451 1-4, 16, 17, 22, 28, 32, 51-64, 91, 102-107, 2, 51-64 16, 17, 22, 28, 32, 91, 203-207 203-207 40 92Zr 451 1-4, 16, 17, 22, 28, 32, 33, 51-67, 91, 102-105, 2, 51-67 16, 17, 22, 28, 32, 33, 91, 203-205, 207 107, 203-205, 207 40 93Zr 451 1, 2, 4, 16, 22, 28, 51-62, 91, 102-107, 111, 251, 2, 16, 22, 28, 51-62, 91 16, 22, 28, 91 252 40 94Zr 451 1-4, 16, 17, 22, 28, 32, 51-64, 91, 102-105, 107, 2, 51-64 16, 17, 22, 28, 32, 91, 203-205, 207 203-205, 207 40 95Zr 451 1, 2, 4, 16, 51-64, 91, 102-107, 111, 251, 252 2, 16, 51-64, 91 16, 91 40 96Zr 451 1-4, 16, 17, 22, 28, 51-57, 91, 102-105, 107, 2, 51-57 16, 17, 22, 28, 91, 203-205, 207 203-205, 207 41 93Nb 451 1, 2, 4, 5, 16, 17, 22, 28, 32, 33, 51-73, 91, 2, 16, 17, 22, 28, 32, 33, 51-73, 91 16, 17, 22, 28, 32, 33, 91 5 102-105, 107 41 94Nb 451 1, 2, 4, 16, 17, 22, 28, 32, 33, 51-79, 91, 102-107 2, 16, 17, 22, 28, 32, 33, 51-79, 91 16, 17, 22, 28, 32, 33, 91 41 95Nb 451 1, 2, 4, 16, 17, 22, 28, 32, 33, 51-70, 91, 102-107 2, 16, 17, 22, 28, 32, 33, 51-70, 91 16, 17, 22, 28, 32, 33, 91 42 100Mo 451 1, 2, 4, 16, 17, 22, 28, 51-54, 91, 102-105, 107, 2, 51-54 16, 17, 22, 28, 91, 203-205, 207 203-205, 207

Table 3. Contents of the JEFF-3.1 general-purpose neutron data library. List of available reactions (MTs) in files MF=3 to MF=6 (cont.).

Z Sym-A MF=1 MF=3 MF=4 MF=5 MF=6 42 92Mo 451 1, 2, 4, 16, 22, 28, 51-66, 91, 102-107, 111, 2, 51-66 16, 22, 28, 91, 203-207 203-207 42 94Mo 451 1, 2, 4, 16, 17, 22, 28, 51-69, 91, 102-107, 111, 2, 51-69 16, 17, 22, 28, 91, 203-207 203-207 42 95Mo 451 1, 2, 4, 16, 17, 22, 28, 51-54, 91, 102, 103, 107, 2, 51-54, 600, 601, 800-805 16, 17, 22, 28, 91, 102, 649, 849 112, 600, 601, 649, 800-805, 849 42 96Mo 451 1, 2, 4, 16, 17, 22, 28, 51-67, 91, 102-105, 107, 2, 51-67 16, 17, 22, 28, 91, 203-205, 207 203-205, 207 42 97Mo 451 1, 2, 4, 16, 17, 22, 28, 51-63, 91, 102-107, 203-207 2, 51-63 16, 17, 22, 28, 91, 203-207 42 98Mo 451 1, 2, 4, 16, 17, 22, 28, 51-64, 91, 102-105, 107, 2, 51-64 16, 17, 22, 28, 91, 203-205, 207 203-205, 207 42 99Mo 451 1, 2, 4, 16, 17, 22, 28, 32, 33, 51-66, 91, 102-105, 2, 16, 17, 22, 28, 32, 33, 51-66, 91 16, 17, 22, 28, 32, 33, 91 107 43 99Tc 451 1-5, 16, 17, 22, 24, 28, 32, 33, 41, 51-70, 91, 2 5, 16, 17, 22, 24, 28, 32, 33, 41, 51-70, 91, 102-108, 111, 112, 600-610, 649-655, 699-705, 102, 108, 111, 112, 600-610, 649-655, 749-755, 799-810, 849 699-705, 749-755, 799-810, 849 44 100Ru 451 1, 2, 4, 51-57, 91, 102-107, 111, 251-253 2, 51-57, 91 91 101

54 44 Ru 451 1, 2, 4, 16, 17, 22, 28, 51-65, 91, 102, 103, 107, 2, 16, 17, 22, 28, 51-65, 91 16, 17, 22, 28, 91 251-253 44 102Ru 451 1, 2, 4, 51-69, 91, 102-107, 111 2, 51-69, 91 91 44 103Ru 451 1, 2, 4, 16, 51-89, 91, 102-107, 111, 251, 252 2, 16, 51-89, 91 16, 91 44 104Ru 451 1, 2, 4, 51, 52, 91, 102-107, 111, 251-253 2, 51, 52, 91 91 44 105Ru 451 1, 2, 4, 51-59, 91, 102-107, 111 2, 51-59, 91 91 44 106Ru 451 1, 2, 4, 51-54, 91, 102-107, 111 2, 51-54, 91 91 44 96Ru 451 1, 2, 4, 91, 102-107, 111, 251-253 2, 91 91 44 98Ru 451 1, 2, 4, 91, 102-107, 111, 251-253 2, 91 91 44 99Ru 451 1, 2, 4, 51-62, 91, 102-107, 111, 251-253 2, 51-62, 91 91 45 103Rh 451 1-5, 11, 16, 17, 22, 24, 25, 28, 29, 32-34, 37, 41, 2 5, 11, 16, 17, 22, 24, 25, 28, 29, 32-34, 37, 42, 44, 45, 51-65, 91, 102-108, 111, 112, 115-117, 41, 42, 44, 45, 51-65, 91, 108, 111, 112, 600-615, 649-665, 699-715, 749-765, 799-815, 115-117, 600-615, 649-665, 699-715, 849 749-765, 799-815, 849 45 105Rh 451 1, 2, 4, 51-53, 91, 102 2, 51-53, 91 91 46 102Pd 451 1, 2, 4, 91, 102-107, 111, 251-253 2, 91 91 46 104Pd 451 1, 2, 4, 16, 51-64, 91, 102-107, 111, 251, 252 2, 16, 51-64, 91 16, 91 46 105Pd 451 1, 2, 4, 16, 51-91, 102-107, 111, 251, 252 2, 16, 51-91 16, 91 46 106Pd 451 1, 2, 4, 16, 51-70, 91, 102-107, 111, 251, 252 2, 16, 51-70, 91 16, 91

Table 3. Contents of the JEFF-3.1 general-purpose neutron data library. List of available reactions (MTs) in files MF=3 to MF=6 (cont.).

Z Sym-A MF=1 MF=3 MF=4 MF=5 MF=6 46 107Pd 451 1, 2, 4, 16, 51-67, 91, 102-107, 111, 251, 252 2, 16, 51-67, 91 16, 91 46 108Pd 451 1, 2, 4, 16, 51-60, 91, 102-107, 111, 251, 252 2, 16, 51-60, 91 16, 91 46 110Pd 451 1, 2, 4, 16, 51-60, 91, 102-107, 111, 251, 252 2, 16, 51-60, 91 16, 91 47 107Ag 451 1, 2, 4, 16, 51-56, 91, 102-107, 111, 251-253 2, 16, 51-56, 91 16, 91 47 109Ag 451 1, 2, 4, 16, 51-64, 91, 102-107, 111, 251, 252 2, 16, 51-64, 91 16, 91 47 110mAg 451 1, 2, 4, 16, 17, 22, 28, 32, 33, 51-79, 91, 102-107 2, 16, 17, 22, 28, 32, 33, 51-79, 91 16, 17, 22, 28, 32, 33, 91 47 111Ag 451 1, 2, 4, 51-60, 91, 102-107, 111 2, 51-60, 91 91 48 106Cd 451 1, 2, 4, 16, 22, 28, 32, 33, 51-55, 91, 102-107 2, 16, 22, 28, 32, 33, 51-55, 91 16, 22, 28, 32, 33, 91 48 108Cd 451 1, 2, 4, 16, 17, 22, 28, 32, 33, 51-60, 91, 102-107 2, 16, 17, 22, 28, 32, 33, 51-60, 91 16, 17, 22, 28, 32, 33, 91 48 110Cd 451 1, 2, 4, 16, 17, 22, 28, 32, 33, 51-64, 91, 102-107 2, 16, 17, 22, 28, 32, 33, 51-64, 91 16, 17, 22, 28, 32, 33, 91 48 111Cd 451 1, 2, 4, 16, 17, 22, 28, 32, 33, 51-61, 91, 102-107 2, 16, 17, 22, 28, 32, 33, 51-61, 91 16, 17, 22, 28, 32, 33, 91 48 112Cd 451 1, 2, 4, 16, 17, 22, 28, 32, 33, 51-58, 91, 102-107 2, 16, 17, 22, 28, 32, 33, 51-58, 91 16, 17, 22, 28, 32, 33, 91 48 113Cd 451 1, 2, 4, 16, 17, 22, 28, 32, 33, 51-62, 91, 102-107 2, 16, 17, 22, 28, 32, 33, 51-62, 91 16, 17, 22, 28, 32, 33, 91 114 55 48 Cd 451 1, 2, 4, 16, 17, 22, 28, 32, 33, 51-60, 91, 102-107 2, 16, 17, 22, 28, 32, 33, 51-60, 91 16, 17, 22, 28, 32, 33, 91 48 115mCd 451 1, 2, 4, 91, 102-107, 111 2, 91 91 48 116Cd 451 1, 2, 4, 16, 17, 22, 28, 32, 33, 51-56, 91, 102-107 2, 16, 17, 22, 28, 32, 33, 51-56, 91 16, 17, 22, 28, 32, 33, 91 49 113In 451 1, 2, 4, 16, 17, 22, 28, 32, 33, 51-74, 91, 102-107 2, 16, 17, 22, 28, 32, 33, 51-74, 91 16, 17, 22, 28, 32, 33, 91 49 115In 451 1, 2, 4, 16, 17, 22, 28, 32, 33, 51-63, 91, 102-105, 2, 16, 17, 22, 28, 32, 33, 51-63, 91 16, 17, 22, 28, 32, 33, 91 107 50 112Sn 451 1, 2, 4, 16, 17, 22, 28, 51-73, 91, 102-107, 111 2, 16, 17, 22, 28, 51-73, 91 16, 17, 22, 28, 91 50 114Sn 451 1, 2, 4, 16, 17, 22, 28, 51-76, 91, 102-107 2, 16, 17, 22, 28, 51-76, 91 16, 17, 22, 28, 91 50 115Sn 451 1, 2, 4, 16, 17, 22, 28, 32, 51-66, 91, 102-107 2, 16, 17, 22, 28, 32, 51-66, 91 16, 17, 22, 28, 32, 91 50 116Sn 451 1, 2, 4, 16, 17, 22, 28, 51-60, 91, 102-105, 107 2, 16, 17, 22, 28, 51-60, 91 16, 17, 22, 28, 91 50 117Sn 451 1, 2, 4, 16, 17, 22, 28, 32, 51-62, 91, 102-105, 107 2, 16, 17, 22, 28, 32, 51-62, 91 16, 17, 22, 28, 32, 91 50 118Sn 451 1, 2, 4, 16, 17, 22, 28, 51-57, 91, 102-105, 107 2, 16, 17, 22, 28, 51-57, 91 16, 17, 22, 28, 91 50 119Sn 451 1, 2, 4, 16, 17, 22, 28, 32, 51-70, 91, 102-105, 107 2, 16, 17, 22, 28, 32, 51-70, 91 16, 17, 22, 28, 32, 91 50 120Sn 451 1, 2, 4, 16, 17, 22, 28, 51-66, 91, 102-105, 107 2, 16, 17, 22, 28, 51-66, 91 16, 17, 22, 28, 91 50 122Sn 451 1, 2, 4, 16, 17, 22, 28, 51-59, 91, 102-105, 107 2, 16, 17, 22, 28, 51-59, 91 16, 17, 22, 28, 91 50 123Sn 451 1, 2, 4, 51-59, 91, 102-107, 111 2, 51-59, 91 91 50 124Sn 451 1, 2, 4, 16, 17, 22, 28, 51-70, 91, 102-105, 107 2, 16, 17, 22, 28, 51-70, 91 16, 17, 22, 28, 91 50 125Sn 451 1, 2, 4, 51-53, 91, 102-107, 111 2, 51-53, 91 91

Table 3. Contents of the JEFF-3.1 general-purpose neutron data library. List of available reactions (MTs) in files MF=3 to MF=6 (cont.).

Z Sym-A MF=1 MF=3 MF=4 MF=5 MF=6 50 126Sn 451 1, 2, 4, 51-60, 91, 102-107, 111 2, 51-60, 91 91 51 121Sb 451 1, 2, 4, 16, 17, 22, 28, 51-63, 91, 102-105, 107, 2, 51-63 16, 17, 22, 28, 91, 203-205, 207 203-205, 207 51 123Sb 451 1, 2, 4, 16, 17, 22, 28, 51-58, 91, 102-105, 107, 2, 51-58 16, 17, 22, 28, 91, 203-205, 207 203-205, 207 51 124Sb 451 1, 2, 4, 51-58, 91, 102-105, 107, 111 2, 51-58, 91 91 51 125Sb 451 1, 2, 4, 51-69, 91, 102-107, 111 2, 51-69, 91 91 51 126Sb 451 1, 2, 4, 51-55, 91, 102-107, 111 2, 51-55, 91 91 52 120Te 451 1, 2, 4, 91, 102-107, 111, 251-253 2, 91 91 52 122Te 451 1, 2, 4, 51-54, 91, 102-107, 111, 251-253 2, 51-54, 91 91 52 123Te 451 1, 2, 4, 51-58, 91, 102-107, 111, 251-253 2, 51-58, 91 91 52 124Te 451 1, 2, 4, 51-61, 91, 102-107, 111, 251-253 2, 51-61, 91 91 52 125Te 451 1, 2, 4, 51-56, 91, 102-107, 111, 251-253 2, 51-56, 91 91 52 126Te 451 1, 2, 4, 51-63, 91, 102-107, 111, 251-253 2, 51-63, 91 91 127m 56 52 Te 451 1, 2, 4, 91, 102 2, 91 91 52 128Te 451 1, 2, 4, 51-59, 91, 102-107, 111, 251 2, 51-59, 91 91 52 129mTe 451 1, 2, 4, 91, 102-107, 111 2, 91 91 52 130Te 451 1, 2, 4, 51-60, 91, 102-107, 111, 251-253 2, 51-60, 91 91 52 132Te 451 1, 2, 4, 51-56, 91, 102-107, 111 2, 51-56, 91 91 53 127I 451 1-5, 11, 16, 17, 22, 24, 25, 28, 32-34, 37, 41, 42, 2 5, 11, 16, 17, 22, 24, 25, 28, 32-34, 37, 41, 44, 45, 51-65, 91, 102-108, 111, 112, 115-117, 42, 44, 45, 51-65, 91, 108, 111, 112, 600-615, 649-665, 699-715, 749-765, 799-815, 115-117, 600-615, 649-665, 699-715, 849 749-765, 799-815, 849 53 129I 451 1-5, 11, 16, 17, 22, 24, 25, 28, 32-34, 37, 41, 42, 2 5, 11, 16, 17, 22, 24, 25, 28, 32-34, 37, 41, 44, 45, 51-65, 91, 102-107, 111, 112, 115-117, 42, 44, 45, 51-65, 91, 111, 112, 115-117, 600-615, 649-665, 699-715, 749-765, 799-805, 600-615, 649-665, 699-715, 749-765, 849 799-805, 849 53 130I 451 1, 2, 4, 91, 102-107, 111 2, 91 91 53 131I 451 1, 2, 4, 51-54, 91, 102-107, 111 2, 51-54, 91 91 53 135I 451 1, 2, 4, 91, 102 2, 91 91 54 124Xe 451 1, 2, 4, 16, 17, 51-54, 91, 102-107, 111 2, 16, 17, 51-54, 91 16, 17, 91 54 126Xe 451 1, 2, 4, 16, 17, 51-54, 91, 102-107, 111 2, 16, 17, 51-54, 91 16, 17, 91 54 128Xe 451 1, 2, 4, 16, 17, 51-53, 91, 102-107, 111 2, 16, 17, 51-53, 91 16, 17, 91 54 129Xe 451 1, 2, 4, 16, 17, 51-56, 91, 102-107, 111 2, 16, 17, 51-56, 91 16, 17, 91

Table 3. Contents of the JEFF-3.1 general-purpose neutron data library. List of available reactions (MTs) in files MF=3 to MF=6 (cont.).

Z Sym-A MF=1 MF=3 MF=4 MF=5 MF=6 54 130Xe 451 1, 2, 4, 16, 17, 51-56, 91, 102-107, 111 2, 16, 17, 51-56, 91 16, 17, 91 54 131Xe 451 1, 2, 4, 16, 17, 51-56, 91, 102-107, 111, 251-253 2, 16, 17, 51-56, 91 16, 17, 91 54 132Xe 451 1, 2, 4, 16, 17, 51-54, 91, 102-107, 111, 251-253 2, 16, 17, 51-54, 91 16, 17, 91 54 133Xe 451 1, 2, 4, 51, 91, 102-107, 111 2, 51, 91 91 54 134Xe 451 1, 2, 4, 16, 17, 51-53, 91, 102-107, 111, 251-253 2, 16, 17, 51-53, 91 16, 17, 91 54 135Xe 451 1, 2, 4, 51, 91, 102 2, 51, 91 91 54 136Xe 451 1, 2, 4, 16, 17, 51-53, 91, 102-107, 111 2, 16, 17, 51-53, 91 16, 17, 91 55 133Cs 451 1, 2, 4, 16, 51-55, 91, 102, 103, 107 2, 16, 51-55, 91 16, 91 55 134Cs 451 1, 2, 4, 51-55, 91, 102-107, 111, 251-253 2, 51-55, 91 91 55 135Cs 451 1, 2, 4, 16, 51-68, 91, 102-107, 111, 251, 252 2, 16, 51-68, 91 16, 91 55 136Cs 451 1, 2, 4, 91, 102-107, 111, 251-253 2, 91 91 55 137Cs 451 1, 2, 4, 16, 51-60, 91, 102-107, 111, 251, 252 2, 16, 51-60, 91 16, 91 56 130Ba 451 1, 2, 4, 16, 17, 22, 28, 32, 51-57, 91, 102-107, 111 2, 16, 17, 22, 28, 32, 51-57, 91 16, 17, 22, 28, 32, 91 132 57 56 Ba 451 1, 2, 4, 16, 17, 22, 28, 51-60, 91, 102-107 2, 16, 17, 22, 28, 51-60, 91 16, 17, 22, 28, 91 56 134Ba 451 1, 2, 4, 16, 17, 22, 28, 51-63, 91, 102-105, 107 2, 16, 17, 22, 28, 51-63, 91 16, 17, 22, 28, 91 56 135Ba 451 1, 2, 4, 16, 17, 22, 28, 32, 51-57, 91, 102-107 2, 16, 17, 22, 28, 32, 51-57, 91 16, 17, 22, 28, 32, 91 56 136Ba 451 1, 2, 4, 16, 17, 22, 28, 51-60, 91, 102-105, 107 2, 16, 17, 22, 28, 51-60, 91 16, 17, 22, 28, 91 56 137Ba 451 1, 2, 4, 16, 17, 22, 28, 32, 51-58, 91, 102-105, 107 2, 16, 17, 22, 28, 32, 51-58, 91 16, 17, 22, 28, 32, 91 56 138Ba 451 1, 2, 4, 16, 17, 22, 28, 51-69, 91, 102-105, 107 2, 16, 17, 22, 28, 51-69, 91 16, 17, 22, 28, 91 56 140Ba 451 1, 2, 4, 16, 17, 91, 102-107, 111, 251, 252 2, 16, 17, 91 16, 17, 91 57 138La 451 1, 2, 4, 16, 17, 22, 28, 32, 33, 51-65, 91, 102-107 2, 16, 17, 22, 28, 32, 33, 51-65, 91 16, 17, 22, 28, 32, 33, 91 57 139La 451 1, 2, 4, 16, 51-73, 91, 102-107, 111, 251, 252 2, 16, 51-73, 91 16, 91 57 140La 451 1, 2, 4, 51-56, 91, 102, 104-107, 111 2, 51-56, 91 91 58 140Ce 451 1, 2, 4, 51-58, 91, 102-107, 111 2, 51-58, 91 91 58 141Ce 451 1, 2, 4, 16, 51-62, 91, 102-107, 111, 251, 252 2, 16, 51-62, 91 16, 91 58 142Ce 451 1, 2, 4, 16, 17, 51-56, 91, 102-107, 111, 251, 252 2, 16, 17, 51-56, 91 16, 17, 91 58 143Ce 451 1, 2, 4, 91, 102-107, 111 2, 91 91 58 144Ce 451 1, 2, 4, 16, 17, 51-68, 91, 102-107, 111, 251, 252 2, 16, 17, 51-68, 91 16, 17, 91 59 141Pr 451 1, 2, 4, 16, 51-74, 91, 102-107, 111, 251, 252 2, 16, 51-74, 91 16, 91 59 142Pr 451 1, 2, 4, 91, 102-107, 111 2, 91 91

Table 3. Contents of the JEFF-3.1 general-purpose neutron data library. List of available reactions (MTs) in files MF=3 to MF=6 (cont.).

Z Sym-A MF=1 MF=3 MF=4 MF=5 MF=6 59 143Pr 451 1, 2, 4, 51-56, 91, 102-107, 111 2, 51-56, 91 91 60 142Nd 451 1, 2, 4, 51-58, 91, 102-107, 111, 251-253 2, 51-58, 91 91 60 143Nd 451 1, 2, 4, 16, 51-76, 91, 102-107, 111, 251, 252 2, 16, 51-76, 91 16, 91 60 144Nd 451 1, 2, 4, 16, 51-64, 91, 102-107, 111, 251, 252 2, 16, 51-64, 91 16, 91 60 145Nd 451 1, 2, 4, 16, 51-64, 91, 102-107, 111, 251, 252 2, 16, 51-64, 91 16, 91 60 146Nd 451 1, 2, 4, 16, 51-63, 91, 102-107, 111, 251, 252 2, 16, 51-63, 91 16, 91 60 147Nd 451 1, 2, 4, 51-53, 91, 102-107, 111, 251-253 2, 51-53, 91 91 60 148Nd 451 1, 2, 4, 16, 51-57, 91, 102-107, 111, 251, 252 2, 16, 51-57, 91 16, 91 60 150Nd 451 1, 2, 4, 16, 17, 22, 28, 51-59, 91, 102-107, 111, 2, 16, 17, 22, 28, 51-59, 91 16, 17, 22, 28, 91 251-253 61 147Pm 451 1, 2, 4, 16, 51-61, 91, 102-107, 111, 251, 252 2, 16, 51-61, 91 16, 91 61 148Pm 451 1, 2, 4, 91, 102 2, 91 91 61 148mPm 451 1, 2, 4, 91, 102, 251-253 2, 91 91 61 149Pm 451 1, 2, 4, 51-56, 91, 102-107, 111 2, 51-56, 91 91 58 61 151Pm 451 1, 2, 4, 51-56, 91, 102-107, 111 2, 51-56, 91 91 62 144Sm 451 1, 2, 4, 91, 102-107, 111, 251-253 2, 91 91 62 147Sm 451 1, 2, 4, 16, 51-64, 91, 102-107, 111, 251, 252 2, 16, 51-64, 91 16, 91 62 148Sm 451 1, 2, 4, 51-66, 91, 102-107, 111 2, 51-66, 91 91 62 149Sm 451 1, 2, 4, 16, 17, 51-60, 91, 102, 103, 107 2, 16, 17, 51-60, 91 16, 17, 91 62 150Sm 451 1, 2, 4, 51-66, 91, 102-107, 111, 251-253 2, 51-66, 91 91 62 151Sm 451 1, 2, 4, 16, 51-91, 102-107, 111, 251, 252 2, 16, 51-91 16, 91 62 152Sm 451 1, 2, 4, 16, 51-63, 91, 102-107, 111, 251, 252 2, 16, 51-63, 91 16, 91 62 153Sm 451 1, 2, 4, 51-62, 91, 102-107, 111 2, 51-62, 91 91 62 154Sm 451 1, 2, 4, 51-69, 91, 102-107, 111, 251-253 2, 51-69, 91 91 63 151Eu 451 1, 2, 4, 16, 17, 22, 28, 51-59, 91, 102-107, 111 2, 16, 17, 22, 28, 51-59, 91 16, 17, 22, 28, 91 63 152Eu 451 1, 2, 4, 16, 17, 22, 28, 32, 33, 51-61, 91, 102-107, 2, 16, 17, 22, 28, 32, 33, 51-61, 91 16, 17, 22, 28, 32, 33, 91 251 63 153Eu 451 1, 2, 4, 16, 17, 22, 28, 51-61, 91, 102, 103, 107, 2, 16, 17, 22, 28, 51-61, 91 16, 17, 22, 28, 91 251 63 154Eu 451 1, 2, 4, 16, 17, 22, 28, 51-55, 91, 102-107 2, 16, 17, 22, 28, 51-55, 91 16, 17, 22, 28, 91 63 155Eu 451 1, 2, 4, 16, 17, 22, 28, 51-59, 91, 102-107 2, 16, 17, 22, 28, 51-59, 91 16, 17, 22, 28, 91 63 156Eu 451 1, 2, 4, 51-55, 91, 102-107, 111 2, 51-55, 91 91

Table 3. Contents of the JEFF-3.1 general-purpose neutron data library. List of available reactions (MTs) in files MF=3 to MF=6 (cont.).

Z Sym-A MF=1 MF=3 MF=4 MF=5 MF=6 63 157Eu 451 1, 2, 4, 91, 102 2, 91 91 64 152Gd 451 1, 2, 4, 16, 17, 22, 28, 32, 33, 51-79, 91, 102-107, 2, 16, 17, 22, 28, 32, 33, 51-79, 91 16, 17, 22, 28, 32, 33, 91 251 64 154Gd 451 1, 2, 4, 51-65, 91, 102-107, 111 2, 51-65, 91 91 64 155Gd 451 1, 2, 4, 51-79, 91, 102-107, 111, 251 2, 51-79, 91 91 64 156Gd 451 1, 2, 4, 16, 51-64, 91, 102-107, 111 2, 16, 51-64, 91 16, 91 64 157Gd 451 1, 2, 4, 51-78, 91, 102-107, 111, 251 2, 51-78, 91 91 64 158Gd 451 1, 2, 4, 51-61, 91, 102-107, 111 2, 51-61, 91 91 64 160Gd 451 1, 2, 4, 51-56, 91, 102-105, 107 2, 51-56, 91 91 65 159Tb 451 1, 2, 4, 16, 51-66, 91, 102, 251, 252 2, 16, 51-66, 91 16, 91 65 160Tb 451 1, 2, 4, 91, 102, 251-253 2, 91 91 66 160Dy 451 1, 2, 4, 51-64, 91, 102, 251-253 2, 51-64, 91 91 66 161Dy 451 1, 2, 4, 51-57, 91, 102 2, 51-57, 91 91 66 162Dy 451 1, 2, 4, 51-56, 91, 102 2, 51-56, 91 91 59 66 163Dy 451 1, 2, 4, 51-56, 91, 102 2, 51-56, 91 91 66 164Dy 451 1, 2, 4, 16, 17, 51-62, 91, 102, 103, 107 2, 16, 17, 51-62, 91 16, 17, 91 67 165Ho 451 1, 2, 4, 16, 17, 37, 51-63, 91, 102 2, 16, 17, 37, 51-63, 91 16, 17, 37, 91 68 162Er 451 1, 2, 4, 16, 17, 22, 28, 51-67, 91, 102-105, 107 2, 16, 17, 22, 28, 51-67, 91 16, 17, 22, 28, 91 68 164Er 451 1, 2, 4, 16, 17, 22, 28, 51-66, 91, 102-105, 107 2, 16, 17, 22, 28, 51-66, 91 16, 17, 22, 28, 91 68 166Er 451 1, 2, 4, 16, 17, 22, 28, 32, 51-74, 91, 102-105, 107 2, 16, 17, 22, 28, 32, 51-74, 91 16, 17, 22, 28, 32, 91 68 167Er 451 1, 2, 4, 16, 17, 22, 28, 32, 51-73, 91, 102-105, 107 2, 16, 17, 22, 28, 32, 51-73, 91 16, 17, 22, 28, 32, 91 68 168Er 451 1, 2, 4, 16, 17, 22, 28, 32, 51-79, 91, 102-105, 107 2, 16, 17, 22, 28, 32, 51-79, 91 16, 17, 22, 28, 32, 91 68 170Er 451 1, 2, 4, 16, 17, 22, 28, 32, 51-64, 91, 102-105, 107 2, 16, 17, 22, 28, 32, 51-64, 91 16, 17, 22, 28, 32, 91 71 175Lu 451 1, 2, 4, 16, 17, 51-58, 91, 102, 103, 107 2, 16, 17, 51-58, 91 16, 17, 91 71 176Lu 451 1, 2, 4, 16, 17, 51-58, 91, 102, 103, 107 2, 16, 17, 51-58, 91 16, 17, 91 72 174Hf 451 1, 2, 4, 16, 17, 51-68, 91, 102 2, 16, 17, 51-68, 91 16, 17, 91 72 176Hf 451 1, 2, 4, 16, 17, 51-73, 91, 102, 103, 107 2, 16, 17, 51-73, 91 16, 17, 91 72 177Hf 451 1, 2, 4, 16, 17, 51-66, 91, 102, 103, 107 2, 16, 17, 51-66, 91 16, 17, 91 72 178Hf 451 1, 2, 4, 16, 17, 51-71, 91, 102, 103, 107 2, 16, 17, 51-71, 91 16, 17, 91 72 179Hf 451 1, 2, 4, 16, 17, 51-62, 91, 102, 103, 107 2, 16, 17, 51-62, 91 16, 17, 91 72 180Hf 451 1, 2, 4, 16, 17, 51-61, 91, 102, 103, 107 2, 16, 17, 51-61, 91 16, 17, 91

Table 3. Contents of the JEFF-3.1 general-purpose neutron data library. List of available reactions (MTs) in files MF=3 to MF=6 (cont.).

Z Sym-A MF=1 MF=3 MF=4 MF=5 MF=6 73 181Ta 451 1, 2, 4, 16, 17, 28, 51-64, 91, 102, 103, 107 2, 16, 17, 28, 51-64, 91 16, 17, 28, 91 73 182Ta 451 1, 2, 4, 16, 17, 51-58, 91, 102, 107, 251-253 2, 16, 17, 51-58, 91 16, 17, 91 74 182W 451 1, 2, 4, 16, 17, 22, 28, 51-70, 91, 102-104, 107, 2, 51-70 16, 17, 22, 28, 91, 203, 204, 207 203, 204, 207 74 183W 451 1, 2, 4, 16, 17, 22, 28, 51-60, 91, 102-104, 107, 2, 51-60 16, 17, 22, 28, 91, 203, 204, 207 203, 204, 207 74 184W 451 1, 2, 4, 16, 17, 22, 28, 51-60, 91, 102-104, 107, 2, 51-60 16, 17, 22, 28, 91, 203, 204, 207 203, 204, 207 74 186W 451 1, 2, 4, 16, 17, 22, 28, 51-62, 91, 102-104, 107, 2, 51-62 16, 17, 22, 28, 91, 203, 204, 207 203, 204, 207 75 185Re 451 1, 2, 4, 16, 17, 51-59, 91, 102 2, 16, 17, 51-59, 91 16, 17, 91 75 187Re 451 1, 2, 4, 16, 17, 51-66, 91, 102 2, 16, 17, 51-66, 91 16, 17, 91 76 natOs 451 1, 2, 4, 16, 17, 51-67, 91, 102, 103, 107, 251-253 2, 16, 17, 51-67, 91 16, 17, 91 77 191Ir 451 1, 2, 4, 16, 17, 51-54, 91, 102, 103, 107, 251-253 2, 16, 17, 51-54, 91 16, 17, 91 77 193Ir 451 1, 2, 4, 16, 17, 51-54, 91, 102, 103, 107, 251-253 2, 16, 17, 51-54, 91 16, 17, 91 nat

60 78 Pt 451 1, 2, 4, 16, 17, 91, 102, 251-253 2, 16, 17, 91 16, 17, 91 79 197Au 451 1, 2, 4, 16, 17, 37, 51-63, 91, 102, 103, 107 2, 16, 17, 37, 51-63, 91 16, 17, 37, 91 80 196Hg 451 1, 2, 4, 16, 17, 22, 28, 51-70, 91, 102-104, 107 2, 16, 17, 22, 28, 51-70, 91 16, 17, 22, 28, 91 80 198Hg 451 1, 2, 4, 16, 17, 22, 28, 51-69, 91, 102-104, 107 2, 16, 17, 22, 28, 51-69, 91 16, 17, 22, 28, 91 80 199Hg 451 1, 2, 4, 16, 17, 22, 28, 51-63, 91, 102-104, 107 2, 16, 17, 22, 28, 51-63, 91 16, 17, 22, 28, 91 80 200Hg 451 1, 2, 4, 16, 17, 22, 28, 51-84, 91, 102-104, 107 2, 16, 17, 22, 28, 51-84, 91 16, 17, 22, 28, 91 80 201Hg 451 1, 2, 4, 16, 17, 22, 28, 51-63, 91, 102-104, 107 2, 16, 17, 22, 28, 51-63, 91 16, 17, 22, 28, 91 80 202Hg 451 1, 2, 4, 16, 17, 22, 28, 51-69, 91, 102-104, 107 2, 16, 17, 22, 28, 51-69, 91 16, 17, 22, 28, 91 80 204Hg 451 1, 2, 4, 16, 17, 22, 28, 51-68, 91, 102-104, 107 2, 16, 17, 22, 28, 51-68, 91 16, 17, 22, 28, 91 81 natTl 451 1-4, 16, 17, 22, 28, 51-66, 91, 102-105, 107 2, 51-66 16, 17, 22, 28, 91 82 204Pb 451 1-5, 16, 17, 22, 24, 28, 32, 33, 41, 51-71, 91, 2 5, 16, 17, 22, 24, 28, 32, 33, 41, 51-71, 91, 102-107, 111, 600-610, 649-655, 699-705, 102, 111, 600-610, 649-655, 699-705, 749-753, 799-810, 849 749-753, 799-810, 849 82 206Pb 451 1-5, 16, 17, 22, 24, 28, 32, 33, 41, 51-75, 91, 2 5, 16, 17, 22, 24, 28, 32, 33, 41, 51-75, 91, 102-105, 107, 108, 600-610, 649-655, 699-705, 102, 108, 600-610, 649-655, 699-705, 749, 749, 800-810, 849 800-810, 849 82 207Pb 451 1-5, 16, 17, 22, 24, 28, 32, 33, 41, 51-70, 91, 2 5, 16, 17, 22, 24, 28, 32, 33, 41, 51-70, 91, 102-105, 107, 600-610, 649-655, 699-705, 749, 102, 600-610, 649-655, 699-705, 749, 800-810, 849 800-810, 849

Table 3. Contents of the JEFF-3.1 general-purpose neutron data library. List of available reactions (MTs) in files MF=3 to MF=6 (cont.).

Z Sym-A MF=1 MF=3 MF=4 MF=5 MF=6 82 208Pb 451 1-5, 16, 17, 22, 24, 28, 32, 33, 41, 51-70, 91, 2 5, 16, 17, 22, 24, 28, 32, 33, 41, 51-70, 91, 102-105, 107, 600-607, 649-655, 699-705, 749, 102, 600-607, 649-655, 699-705, 749, 800-810, 849 800-810, 849 83 209Bi 451 1-5, 16, 17, 22, 24, 28, 32, 33, 41, 51-70, 91, 2 5, 16, 17, 22, 24, 28, 32, 33, 41, 51-70, 91, 102-108, 112, 600-610, 649-655, 699-705, 102, 108, 112, 600-610, 649-655, 699-705, 749-754, 799-810, 849 749-754, 799-810, 849 88 223Ra 451, 452 1, 2, 4, 16-18, 37, 51-64, 91, 102 2, 16-18, 37, 51-64, 91 16-18, 37, 91 88 224Ra 451 1, 2, 4, 16, 17, 37, 51-61, 91, 102 2, 16, 17, 37, 51-61, 91 16, 17, 37, 91 88 225Ra 451 1, 2, 4, 16, 17, 37, 51-56, 91, 102 2, 16, 17, 37, 51-56, 91 16, 17, 37, 91 88 226Ra 451, 452 1, 2, 4, 16-18, 37, 51-66, 91, 102 2, 16-18, 37, 51-66, 91 16-18, 37, 91 89 225Ac 451 1, 2, 4, 16, 17, 37, 51, 91, 102 2, 16, 17, 37, 51, 91 16, 17, 37, 91 89 226Ac 451 1, 2, 4, 16, 17, 37, 91, 102 2, 16, 17, 37, 91 16, 17, 37, 91 , 89 227Ac 451, 452 1, 2, 4, 16-18, 37, 51-59, 91, 102 2, 16-18, 37, 51-59, 91 16-18, 37, 91 90 227Th 451, 452, 455, 1, 2, 4, 16-18, 37, 91, 102 2, 16-18, 37, 91 16-18, 37, 91, 455 456 90 228Th 451, 452, 455, 1, 2, 4, 16-18, 51-62, 91, 102 2, 16-18, 51-62, 91 16-18, 91 61 456 90 229Th 451, 452, 455, 1, 2, 4, 16-18, 37, 51-54, 91, 102, 251 2, 16-18, 37, 51-54, 91 16-18, 37, 91, 455 456 90 230Th 451, 452, 458 1, 2, 4, 16-18, 51-67, 91, 102, 251-253 2, 16-18, 51-67, 91 16-18, 91 90 232Th 451, 452, 455, 1, 2, 4, 16-21, 37, 38, 51-81, 91, 102 2, 16-21, 37, 38, 51-81, 91 16-21, 37, 38, 91, 455 456, 458 90 233Th 451, 452, 455, 1, 2, 4, 16-18, 51-65, 91, 102 2, 16-18, 51-65, 91 16-18, 91 456 90 234Th 451, 452, 455, 1, 2, 4, 16-18, 51-67, 91, 102 2, 16-18, 51-67, 91 16-18, 91 456 91 231Pa 451, 452, 455, 1, 2, 4, 16-18, 51-62, 91, 102, 251-253 2, 16-18, 51-62, 91 16-18, 91, 455 456 91 232Pa 451, 452, 455, 1, 2, 4, 16-18, 37, 91, 102 2, 16-18, 37, 91 16-18, 37, 91 456 91 233Pa 451, 452, 458 1, 2, 4, 16-18, 51-55, 91, 102, 251-253 2, 16-18, 51-55, 91 16-18, 91 92 232U 451, 452, 455, 1, 2, 4, 16-18, 51-63, 91, 102, 251-253 2, 16-18, 51-63, 91 16-18, 91, 455 456 92 233U 451, 452, 455, 1, 2, 4, 16-18, 51-69, 91, 102 2, 16-18, 51-69, 91 16-18, 91, 455 456, 458 92 234U 451, 452, 455, 1, 2, 4, 16-21, 38, 51-86, 91, 102 2, 16-21, 38, 51-86, 91 16-21, 38, 91, 455 456, 458

Table 3. Contents of the JEFF-3.1 general-purpose neutron data library. List of available reactions (MTs) in files MF=3 to MF=6 (cont.).

Z Sym-A MF=1 MF=3 MF=4 MF=5 MF=6 92 235U 451, 452, 455, 1, 2, 4, 16-21, 37, 38, 51-84, 91, 102 2, 18, 51-84 18, 455 16, 17, 37, 91 456, 458 92 236U 451, 452, 455, 1-4, 16-21, 37, 38, 51-81, 91, 102 2, 18, 51-81 18, 455 16, 17, 37, 91 456, 458 92 237U 451, 452, 455, 1-4, 16-21, 37, 38, 51-84, 91, 102 2, 18, 51-84 18, 455 16, 17, 37, 91 456 92 238U 451, 452, 455, 1-4, 16-21, 37, 38, 51-91, 102 2, 18, 51-90 18, 455 16, 17, 37, 91 456, 458 93 235Np 451, 452, 455, 1, 2, 4, 16-18, 51-66, 91, 102 2, 16-18, 51-66, 91 16-18, 91 456 93 236Np 451, 452, 455, 1, 2, 4, 16-18, 51-54, 91, 102 2, 16-18, 51-54, 91 16-18, 91 456 93 237Np 451, 452, 455, 1, 2, 4, 16-18, 37, 51-82, 91, 102 2, 18, 51-82 18, 455 16, 17, 37, 91 456, 458 93 238Np 451, 452, 455, 1, 2, 4, 16-18, 91, 102 2, 16-18, 91 16-18, 91, 455 456 93 239Np 451, 452, 455, 1, 2, 4, 16-18, 51-58, 91, 102 2, 16-18, 51-58, 91 16-18, 91 456

62 94 236Pu 451, 452, 455, 1, 2, 4, 16-18, 51-54, 91, 102 2, 16-18, 51-54, 91 16-18, 91 456, 458 94 237Pu 451, 452, 458 1, 2, 4, 16-20, 51-60, 91, 102, 251-253 2, 16-20, 51-60, 91 16-20, 91 94 238Pu 451, 452, 455, 1, 2, 4, 16-18, 51-78, 91, 102, 251 2, 16-18, 51-78, 91 16-18, 91, 455 456, 458 94 239Pu 451, 452, 455, 1-4, 16-18, 37, 51-77, 91, 102 2, 16-18, 37, 51-77, 91 16-18, 37, 91, 455 456, 458 94 240Pu 451, 452, 455, 1, 2, 4, 16-18, 37, 51-74, 91, 102 2, 16-18, 37, 51-74, 91 16-18, 37, 91, 455 456, 458 94 241Pu 451, 452, 455, 1, 2, 4, 16-18, 37, 51-61, 91, 102, 251 2, 16-18, 37, 51-61, 91 16-18, 37, 91, 455 456, 458 94 242Pu 451, 452, 455, 1, 2, 4, 16-18, 37, 51-70, 91, 102, 251 2, 16-18, 37, 51-70, 91 16-18, 37, 91, 455 456, 458 94 243Pu 451, 452 1, 2, 4, 16-18, 37, 91, 102, 251-253 2, 16-18, 37, 91 16-18, 37, 91 94 244Pu 451, 452, 458 1, 2, 4, 16-20, 37, 51-55, 91, 102, 251-253 2, 16-20, 37, 51-55, 91 16-20, 37, 91 94 246Pu 451, 452, 455, 1, 2, 4, 16-18, 51, 52, 91, 102 2, 16-18, 51, 52, 91 16-18, 91 456 95 241Am 451, 452, 455, 1, 2, 4, 16-18, 51-60, 91, 102 2, 16-18, 51-60, 91 16-18, 91, 455 456, 458 95 242Am 451, 452, 455, 1, 2, 4, 16-21, 51-75, 91, 102 2, 16-21, 51-75, 91 16-21, 91 456

Table 3. Contents of the JEFF-3.1 general-purpose neutron data library. List of available reactions (MTs) in files MF=3 to MF=6 (cont.).

Z Sym-A MF=1 MF=3 MF=4 MF=5 MF=6 95 242mAm 451, 452, 455, 1, 2, 4, 16-18, 51-75, 91, 102 2, 16-18, 51-75, 91 16-18, 91, 455 456, 458 95 243Am 451, 452, 455, 1, 2, 4, 16-18, 51-60, 91, 102 2, 16-18, 51-60, 91 16-18, 91, 455 456, 458 95 244Am 451, 452, 455, 1, 2, 4, 16-18, 37, 51-75, 91, 102 2, 16-18, 37, 51-75, 91 16-18, 37, 91 456 95 244mAm 451, 452, 455, 1, 2, 4, 16-18, 37, 51-75, 91, 102 2, 16-18, 37, 51-75, 91 16-18, 37, 91 456 96 240Cm 451, 452, 455, 1, 2, 4, 16-18, 51, 91, 102 2, 16-18, 51, 91 16-18, 91 456 96 241Cm 451, 452, 458 1, 2, 4, 16-20, 51-54, 91, 102, 251-253 2, 16-20, 51-54, 91 16-20, 91 96 242Cm 451, 452, 455, 1, 2, 4, 16-18, 51-53, 91, 102, 251 2, 16-18, 51-53, 91 16-18, 91, 455 456, 458 96 243Cm 451, 452, 455, 1, 2, 4, 16-18, 37, 51-62, 91, 102, 251 2, 16-18, 37, 51-62, 91 16-18, 37, 91, 455 456, 458 96 244Cm 451, 452, 455, 1, 2, 4, 16-20, 51-53, 91, 102 2, 16-20, 51-53, 91 16-20, 91, 455 456, 458 96 245Cm 451, 452, 455, 1, 2, 4, 16-18, 51-67, 91, 102 2, 16-18, 51-67, 91 16-18, 91, 455 63 456 96 246Cm 451, 452, 455, 1, 2, 4, 16-21, 51-70, 91, 102 2, 16-21, 51-70, 91 16-21, 91, 455 456 96 247Cm 451, 452, 455, 1, 2, 4, 16-18, 37, 51-65, 91, 102, 251 2, 16-18, 37, 51-65, 91 16-18, 37, 91 456 96 248Cm 451, 452, 455, 1, 2, 4, 16-18, 37, 51-58, 91, 102, 251 2, 16-18, 37, 51-58, 91 16-18, 37, 91, 455 456, 458 96 249Cm 451, 452, 455, 1, 2, 4, 16-18, 51-69, 91, 102 2, 16-18, 51-69, 91 16-18, 91 456 96 250Cm 451, 452, 455, 1, 2, 4, 16-18, 51, 52, 91, 102 2, 16-18, 51, 52, 91 16-18, 91 456 97 247Bk 451, 452, 455, 1, 2, 4, 16-18, 51-71, 91, 102 2, 16-18, 51-71, 91 16-18, 91 456 97 249Bk 451, 452, 455, 1, 2, 4, 16-18, 37, 51-68, 91, 102 2, 16-18, 37, 51-68, 91 16-18, 37, 91 456 97 250Bk 451, 452, 455, 1, 2, 4, 16-18, 37, 51-68, 91, 102, 251 2, 16-18, 37, 51-68, 91 16-18, 37, 91 456 98 249Cf 451, 452, 455, 1, 2, 4, 16-18, 51-78, 91, 102 2, 16-18, 51-78, 91 16-18, 91, 455 456 98 250Cf 451, 452, 455, 1, 2, 4, 16-18, 51-79, 91, 102 2, 16-18, 51-79, 91 16-18, 91 456 98 251Cf 451, 452, 455, 1, 2, 4, 16-18, 37, 91, 102, 251-253 2, 16-18, 37, 91 16-18, 37, 91, 455 456

Table 3. Contents of the JEFF-3.1 general-purpose neutron data library. List of available reactions (MTs) in files MF=3 to MF=6 (cont.).

Z Sym-A MF=1 MF=3 MF=4 MF=5 MF=6 98 252Cf 451, 452 1, 2, 4, 16-18, 37, 91, 102, 251-253 2, 16-18, 37, 91 16-18, 37, 91 98 254Cf 451, 452, 455, 1, 2, 4, 16-18, 37, 51, 91, 102, 251 2, 16-18, 37, 51, 91 16-18, 37, 91 456 99 253Es 451 1, 2, 102, 251-253 99 254Es 451, 452, 455, 1, 2, 4, 16-18, 37, 51, 91, 102 2, 16-18, 37, 51, 91 16-18, 37, 91, 455 456 99 255Es 451, 452, 455, 1, 2, 4, 16-18, 37, 51-53, 91, 102 2, 16-18, 37, 51-53, 91 16-18, 37, 91 456 100 255Fm 451, 452, 455, 1, 2, 4, 16-18, 37, 51, 91, 102 2, 16-18, 37, 51, 91 16-18, 37, 91 456 64

Table 4. Contents of the JEFF-3.1 general-purpose neutron data library. List of available reactions (MTs) in photon production files MF=12 to MF=15.

Z Sym-A MF=12 MF=13 MF=14 MF=15 1 1H 102 102 1 2H 102 102 3 6Li 57, 102 57, 102 3 7Li 51, 102 51, 102 4 9Be 102 102 5 10B 102, 801 4, 103 4, 102, 103, 801 5 11B 102 4 4, 102 6 natC 51, 102 51, 102 7 14N 102 4, 28, 32, 103-105, 107 4, 28, 32, 102-105, 107 102 7 15N 4, 16, 22, 28, 103-105, 107 4, 16, 22, 28, 103-105, 107 4, 16, 22, 28, 103-105, 107 8 16O 102 4, 16, 22, 103-105, 107 4, 16, 22, 102-105, 107 9 19F 51-71, 102 51-71, 102 102 23 65 11 Na 51-61, 102 3 3, 51-61, 102 3, 102 12 24Mg 16, 22, 28, 51-61, 91, 102, 103, 107 16, 22, 28, 51-61, 91, 102, 103, 107 16, 22, 28, 91, 102, 103, 107 12 25Mg 16, 22, 28, 51-67, 91, 102, 103, 107 16, 22, 28, 51-67, 91, 102, 103, 107 16, 22, 28, 91, 102, 103, 107 12 26Mg 16, 22, 28, 51-63, 91, 102, 103, 107 16, 22, 28, 51-63, 91, 102, 103, 107 16, 22, 28, 91, 102, 103, 107 13 27Al 102 102 102 14 28Si 51-67, 102, 601-613, 801-815 51-67, 102, 601-613, 801-815 14 29Si 51-69, 102, 103, 107 3 3, 51-69, 102, 103, 107 3, 102, 103, 107 14 30Si 51-64, 102, 107 3 3, 51-64, 102, 107 3, 102, 107 15 31P 102, 103, 107 3, 4 3, 4, 102, 103, 107 3, 102, 103, 107 16 32S 16, 22, 28, 51-56, 91, 102, 103, 107 16, 22, 28, 51-56, 91, 102, 103, 107 16, 22, 28, 91, 102, 103, 107 16 33S 16, 22, 28, 51-57, 91, 102, 103, 107 16, 22, 28, 51-57, 91, 102, 103, 107 16, 22, 28, 91, 102, 103, 107 16 34S 16, 22, 28, 51-55, 91, 102, 103, 107 16, 22, 28, 51-55, 91, 102, 103, 107 16, 22, 28, 91, 102, 103, 107 16 36S 16, 22, 28, 51-55, 91, 102, 103, 107 16, 22, 28, 51-55, 91, 102, 103, 107 16, 22, 28, 91, 102, 103, 107 17 35Cl 102 102 102 17 37Cl 102 102 102 19 39K 16, 22, 28, 51-54, 91, 102, 103, 107 16, 22, 28, 51-54, 91, 102, 103, 107 16, 22, 28, 91, 102, 103, 107

Table 4. Contents of the JEFF-3.1 general-purpose neutron data library. List of available reactions (MTs) in photon production files MF=12 to MF=15 (cont.).

Z Sym-A MF=12 MF=13 MF=14 MF=15 19 40K 16, 22, 28, 51-55, 91, 102, 103, 107 16, 22, 28, 51-55, 91, 102, 103, 107 16, 22, 28, 91, 102, 103, 107 19 41K 16, 22, 28, 51, 52, 91, 102, 103, 107 16, 22, 28, 51, 52, 91, 102, 103, 107 16, 22, 28, 91, 102, 103, 107 23 natV 102 3 3, 102 3, 102 24 50Cr 51-56, 102 51-56, 102 102 24 52Cr 51-60, 102 51-60, 102 102 24 53Cr 51-63, 102 51-63, 102 102 24 54Cr 51-54, 102 51-54, 102 102 25 55Mn 4, 16, 22, 28, 102, 103, 107 4, 16, 22, 28, 102, 103, 107 4, 16, 22, 28, 102, 103, 107 26 56Fe 51-82, 102, 601-613, 801-810 51-82, 102, 601-613, 801-810 102 27 59Co 102 3 3, 102 3, 102 28 58Ni 51-58, 102 51-58, 102 102 28 59Ni 3, 102 3, 102 3, 102 60 66 28 Ni 51-61, 102 51-61, 102 102 28 61Ni 51-58, 102 51-58, 102 102 28 62Ni 51-54, 102 51-54, 102 102 28 64Ni 51, 52, 102 51, 52, 102 102 29 63Cu 51-72, 102 51-72, 102 102 29 65Cu 51-63, 102 51-63, 102 102 31 natGa 102 3 3, 102 3, 102 39 89Y 102 3 3, 102 3, 102 40 90Zr 16, 22, 28, 51-57, 91, 102, 103, 107 16, 22, 28, 51-57, 91, 102, 103, 107 16, 22, 28, 91, 102, 103, 107 40 91Zr 16, 22, 28, 51-64, 91, 102, 103, 107 16, 22, 28, 51-64, 91, 102, 103, 107 16, 22, 28, 91, 102, 103, 107 40 92Zr 16, 17, 22, 28, 51-67, 91, 102, 103, 107 16, 17, 22, 28, 51-67, 91, 102, 103, 107 16, 17, 22, 28, 91, 102, 103, 107 40 94Zr 16, 17, 22, 28, 51-64, 91, 102, 103, 107 16, 17, 22, 28, 51-64, 91, 102, 103, 107 16, 17, 22, 28, 91, 102, 103, 107 40 96Zr 16, 17, 22, 28, 51-57, 91, 102, 103, 107 16, 17, 22, 28, 51-57, 91, 102, 103, 107 16, 17, 22, 28, 91, 102, 103, 107 41 93Nb 102 3 3, 102 3, 102 42 100Mo 102 3 3, 102 3, 102 42 92Mo 102 3 3, 102 3, 102 42 94Mo 102 3 3, 102 3, 102

Table 4. Contents of the JEFF-3.1 general-purpose neutron data library. List of available reactions (MTs) in photon production files MF=12 to MF=15 (cont.).

Z Sym-A MF=12 MF=13 MF=14 MF=15 42 95Mo 51-54, 601, 801-805 51-54, 601, 801-805 42 96Mo 102 3 3, 102 3, 102 42 97Mo 102 3 3, 102 3, 102 42 98Mo 102 3 3, 102 3, 102 63 151Eu 51-59, 102 3 3, 51-59, 102 3, 102 67 165Ho 102 4, 16, 17, 37 4, 16, 17, 37, 102 4, 16, 17, 37, 102 68 162Er 16, 17, 51-67, 91, 102, 103, 107 16, 17, 51-67, 91, 102, 103, 107 16, 17, 91, 102, 103, 107 68 164Er 16, 17, 51-66, 91, 102, 103, 107 16, 17, 51-66, 91, 102, 103, 107 16, 17, 91, 102, 103, 107 68 166Er 16, 17, 22, 28, 51-74, 91, 102-104, 107 16, 17, 22, 28, 51-74, 91, 102-104, 107 16, 17, 22, 28, 91, 102-104, 107 68 167Er 16, 17, 22, 28, 51-73, 91, 102-104, 107 16, 17, 22, 28, 51-73, 91, 102-104, 107 16, 17, 22, 28, 91, 102-104, 107 68 168Er 16, 17, 22, 28, 51-79, 91, 102-104, 107 16, 17, 22, 28, 51-79, 91, 102-104, 107 16, 17, 22, 28, 91, 102-104, 107 68 170Er 16, 17, 22, 28, 51-64, 91, 102-104, 107 16, 17, 22, 28, 51-64, 91, 102-104, 107 16, 17, 22, 28, 91, 102-104, 107 174 67 72 Hf 16, 17, 51-68, 91, 102 16, 17, 51-68, 91, 102 16, 17, 91, 102 72 176Hf 16, 17, 51-73, 91, 102, 103, 107 16, 17, 51-73, 91, 102, 103, 107 16, 17, 91, 102, 103, 107 72 177Hf 16, 17, 51-66, 91, 102, 103, 107 16, 17, 51-66, 91, 102, 103, 107 16, 17, 91, 102, 103, 107 72 178Hf 16, 17, 51-71, 91, 102, 103, 107 16, 17, 51-71, 91, 102, 103, 107 16, 17, 91, 102, 103, 107 72 179Hf 16, 17, 51-62, 91, 102, 103, 107 16, 17, 51-62, 91, 102, 103, 107 16, 17, 91, 102, 103, 107 72 180Hf 16, 17, 51-61, 91, 102, 103, 107 16, 17, 51-61, 91, 102, 103, 107 16, 17, 91, 102, 103, 107 73 181Ta 16, 17, 28, 51-64, 91, 102, 103, 107 16, 17, 28, 51-64, 91, 102, 103, 107 16, 17, 28, 91, 102, 103, 107 74 182W 22, 51-70, 102, 103, 107 3 3, 22, 51-70, 102, 103, 107 3, 22, 102, 103, 107 74 183W 22, 51-60, 102, 103, 107 3 3, 22, 51-60, 102, 103, 107 3, 22, 102, 103, 107 74 184W 51-60, 102, 107 3 3, 51-60, 102, 107 3, 102, 107 74 186W 22, 51-62, 102, 107 3 3, 22, 51-62, 102, 107 3, 22, 102, 107 76 natOs 51-67, 91, 102 51-67, 91, 102 91, 102 77 191Ir 51, 52, 54, 91, 102, 103, 107 51, 52, 54, 91, 102, 103, 107 91, 102, 103, 107 77 193Ir 51, 53, 54, 91, 102, 103, 107 51, 53, 54, 91, 102, 103, 107 91, 102, 103, 107 78 natPt 102 3 3, 102 3, 102 79 197Au 102 4, 16, 17, 37 4, 16, 17, 37, 102 4, 16, 17, 37, 102 80 196Hg 16, 17, 22, 28, 51-70, 91, 102-104, 107 16, 17, 22, 28, 51-70, 91, 102-104, 107 16, 17, 22, 28, 91, 102-104, 107

Table 4. Contents of the JEFF-3.1 general-purpose neutron data library. List of available reactions (MTs) in photon production files MF=12 to MF=15 (cont.).

Z Sym-A MF=12 MF=13 MF=14 MF=15 80 198Hg 16, 17, 22, 28, 51-69, 91, 102-104, 107 16, 17, 22, 28, 51-69, 91, 102-104, 107 16, 17, 22, 28, 91, 102-104, 107 80 199Hg 16, 17, 22, 28, 51-56, 58-63, 91, 102-104, 107 16, 17, 22, 28, 51-56, 58-63, 91, 102-104, 107 16, 17, 22, 28, 91, 102-104, 107 80 200Hg 16, 17, 22, 28, 51-84, 91, 102-104, 107 16, 17, 22, 28, 51-84, 91, 102-104, 107 16, 17, 22, 28, 91, 102-104, 107 80 201Hg 16, 17, 22, 28, 51-63, 91, 102-104, 107 16, 17, 22, 28, 51-63, 91, 102-104, 107 16, 17, 22, 28, 91, 102-104, 107 80 202Hg 16, 17, 22, 28, 51-69, 91, 102-104, 107 16, 17, 22, 28, 51-69, 91, 102-104, 107 16, 17, 22, 28, 91, 102-104, 107 80 204Hg 16, 17, 22, 28, 51-68, 91, 102-104, 107 16, 17, 22, 28, 51-68, 91, 102-104, 107 16, 17, 22, 28, 91, 102-104, 107 90 232Th 18, 102 3 3, 18, 102 3, 18, 102 92 233U 4, 18, 102 3 3, 4, 18, 102 3, 4, 18, 102 92 235U 4, 18, 102 3 3, 4, 18, 102 3, 18, 102 92 236U 18, 102 3 3, 18, 102 3, 18, 102 92 237U 18, 102 3 3, 18, 102 3, 18, 102 92 238U 18, 102 3 3, 18, 102 3, 18, 102 240 68 94 Pu 4, 18, 102 3 3, 4, 18, 102 3, 18, 102 94 242Pu 16, 17, 51-70, 91, 102 16, 17, 51-70, 91, 102 16, 17, 91, 102 94 243Pu 18, 102 3 3, 18, 102 3, 18, 102 96 242Cm 18, 102 3 3, 18, 102 3, 18, 102 96 243Cm 18, 102 3 3, 18, 102 3, 18, 102 96 244Cm 18, 102 3 3, 18, 102 3, 18, 102 96 246Cm 18, 102 3 3, 18, 102 3, 18, 102 96 247Cm 18, 102 3 3, 18, 102 3, 18, 102 96 248Cm 18, 102 3 3, 18, 102 3, 18, 102 98 251Cf 18, 102 3 3, 18, 102 3, 18, 102 98 252Cf 18, 102 3 3, 18, 102 3, 18, 102

Table 5. Contents of the JEFF-3.1 general-purpose neutron data library. List of covariance data (MF=33 to MF=36).

Z Sym-A MF=31 MF=32 MF=33 MF=34 MF=35 MF=36 MF=40 1 3H 1, 2, 16, 17 4 9Be 1, 2, 102-105, 107, 875-890 6 natC 1-5, 28, 51-62, 91, 102-104, 107 9 19F 1-4, 16, 22, 28, 51, 52, 102-105, 107 14 28Si 1-4, 16, 22, 28, 51-67, 91, 102-107, 111, 600-613, 649, 800-815, 849, 851, 852 22 46Ti 1-4, 16, 22, 28, 51-70, 91, 102-107, 111, 854 22 47Ti 1-4, 16, 22, 28, 51-70, 91, 102-107, 111, 852, 854 22 48Ti 1-4, 16, 22, 28, 51-70, 91, 102-107, 111, 851, 852 22 49Ti 1-4, 16, 22, 28, 51-70, 91, 102-107, 111, 852, 854 22 50Ti 1-4, 16, 22, 28, 51-70, 91, 102-107, 854 23 natV 1, 16, 103, 107 24 50Cr 1-4, 16, 22, 28, 51-56, 91, 102-104, 107 24 52Cr 1-4, 16, 22, 28, 51-60, 91, 102-107, 851, 852 2 53 69 24 Cr 1-4, 16, 22, 28, 51-63, 91, 102, 103, 107 24 54Cr 1-4, 16, 51-54, 91, 102, 103, 107 25 55Mn 151 1, 2, 4, 16, 22, 28, 51-66, 91, 102-107 26 54Fe 1-4, 16, 22, 28, 51-57, 91, 102-104, 107 26 56Fe 1-4, 16, 22, 28, 51-82, 91, 102-107, 600-613, 649, 800-810, 849, 851-853 2 26 57Fe 1-4, 16, 22, 28, 51-55, 91, 102, 103, 107 26 58Fe 1-4, 16, 22, 28, 51, 52, 91, 102, 103, 107 27 59Co 1, 16, 103, 107 28 58Ni 1-4, 16, 22, 28, 51-58, 91, 102-107, 112, 854 2 28 60Ni 1-4, 16, 22, 28, 51-61, 91, 102-107, 851-853 2 28 61Ni 1-4, 16, 28, 51-58, 91, 102, 103, 107, 111 28 62Ni 1-4, 16, 22, 28, 51-54, 91, 102-104, 107 28 64Ni 1-4, 16, 22, 28, 51, 52, 91, 102-104, 107 29 63Cu 1-4, 16, 22, 28, 51-72, 91, 102-104, 106, 107 29 65Cu 1-4, 16, 22, 28, 51-63, 91, 102-107 39 89Y 1, 2, 4, 16, 91, 102, 103, 107 40 90Zr 4, 16, 102

Table 5. Contents of the JEFF-3.1 general-purpose neutron data library. List of covariance data (MF=33 to MF=36) (cont.).

Z Sym-A MF=31 MF=32 MF=33 MF=34 MF=35 MF=36 MF=40 41 93Nb 1, 2, 4, 16, 17, 102 4 75 185Re 102 75 187Re 102 79 197Au 1 92 233U 452, 455, 456 151 1, 2, 4, 16-18, 102 2 18 92 235U 452, 456 95 241Am 102 70

Table 6. Resonance data in the JEFF-3.1 general-purpose neutron data library

Z Isotope Res data Formalism Upper limit RRR Av param in URR Upper limit URR Energy of 3 s-wave res. (eV) 1 1H No 1 2H No 1 3H No 2 3He No 2 4He No 3 6Li No 3 7Li No 4 9Be No 5 10B No 5 11B No 6 natC No 7 14N No 7 15N No 8 16O No 71 8 17O No 9 19F No 11 22Na Yes ML Breit-Wigner 1.50E+04 Yes 1.00E+05 1.45E+02 11 23Na Yes ML Breit-Wigner 3.50E+05 No 2.85E+03 2.99E+05 5.38E+05 12 24Mg Yes ML Breit-Wigner 5.20E+05 No 6.59E+05 1.04E+06 1.08E+06 12 25Mg Yes ML Breit-Wigner 2.20E+05 No 7.33E+04 12 26Mg Yes ML Breit-Wigner 4.50E+05 No 13 27Al Yes Reich-Moore 8.45E+05 No 3.48E+04 8.63E+04 1.43E+05 14 28Si Yes Reich-Moore 1.75E+06 No 5.57E+04 1.82E+05 3.01E+05 14 29Si Yes ML Breit-Wigner 2.00E+05 No 14 30Si Yes ML Breit-Wigner 5.00E+05 No 1.83E+05 15 31P Yes ML Breit-Wigner 5.44E+05 No 2.68E+04 8.40E+04 9.27E+04 16 32S Yes ML Breit-Wigner 1.57E+06 No 1.03E+05 1.45E+05 1.60E+05 16 33S Yes ML Breit-Wigner 2.60E+05 No 1.35E+04 2.39E+04 5.20E+04 16 34S Yes ML Breit-Wigner 4.80E+05 No 3.01E+05 3.57E+05 4.71E+05 16 36S No 17 35Cl No 17 37Cl No

Table 6. Resonance data in the JEFF-3.1 general-purpose neutron data library (cont.)

Z Isotope Res data Formalism Upper limit RRR Av param in URR Upper limit URR Energy of 3 s-wave res. (eV) 18 36Ar Yes ML Breit-Wigner 4.65E+04 Yes 1.20E+06 2.00E+04 18 38Ar Yes ML Breit-Wigner 3.00E+05 Yes 1.20E+06 1.35E+05 18 40Ar Yes Reich-Moore 1.50E+06 No 7.65E+04 1.72E+05 2.20E+05 19 39K Yes ML Breit-Wigner 2.00E+05 No 9.37E+03 2.55E+04 4.27E+04 19 40K No 19 41K Yes ML Breit-Wigner 1.25E+05 No 2.03E+03 5.51E+03 1.67E+04 20 40Ca Yes ML Breit-Wigner 5.00E+05 No 2.04E+04 4.21E+04 8.89E+04 20 42Ca Yes ML Breit-Wigner 3.00E+05 No 9.35E+03 2.27E+04 2.63E+04 20 43Ca Yes ML Breit-Wigner 4.00E+04 No 3.33E+03 4.32E+03 5.19E+03 20 44Ca Yes ML Breit-Wigner 5.00E+05 No 1.51E+04 2.31E+04 2.72E+04 20 46Ca No 20 48Ca Yes ML Breit-Wigner 5.00E+05 No 1.61E+05 21 45Sc Yes ML Breit-Wigner 1.00E+05 No 3.30E+03 4.33E+03 6.69E+03 22 46Ti No 72 22 47Ti No 22 48Ti No 22 49Ti No 22 50Ti No 23 natV Yes ML Breit-Wigner 1.00E+05 No 1.67E+02 1.42E+03 6.20E+03 23 natV Yes ML Breit-Wigner 1.00E+05 No 4.15E+03 6.80E+03 1.16E+04 24 50Cr Yes Reich-Moore 7.92E+05 No 5.50E+03 2.84E+04 3.73E+04 24 52Cr Yes Reich-Moore 1.20E+06 No 3.17E+04 5.03E+04 9.67E+04 24 53Cr Yes Reich-Moore 2.00E+05 No 4.21E+03 5.67E+03 6.75E+03 24 54Cr Yes Reich-Moore 9.00E+05 No 2.26E+04 1.19E+05 1.31E+05 25 55Mn Yes ML Breit-Wigner 1.00E+05 No 3.37E+02 1.10E+03 1.66E+03 26 54Fe Yes Reich-Moore 7.00E+05 No 7.79E+03 5.29E+04 7.19E+04 26 56Fe Yes Reich-Moore 8.50E+05 No 2.78E+04 7.40E+04 8.36E+04 26 57Fe Yes Reich-Moore 2.00E+05 No 3.95E+03 6.26E+03 2.90E+04 26 58Fe Yes Reich-Moore 3.50E+05 Yes 3.00E+06 1.04E+04 3.77E+04 4.38E+04 27 58Co Yes ML Breit-Wigner 5.96E+02 No 9.00E+00 27 58mCo Yes ML Breit-Wigner 5.00E+02 No 1.09E+00 27 59Co Yes Reich-Moore 1.00E+05 No 1.32E+02 4.32E+03 5.02E+03

Table 6. Resonance data in the JEFF-3.1 general-purpose neutron data library (cont.)

Z Isotope Res data Formalism Upper limit RRR Av param in URR Upper limit URR Energy of 3 s-wave res. (eV) 28 58Ni Yes Reich-Moore 8.12E+05 No 1.53E+04 3.61E+04 6.33E+04 28 59Ni Yes ML Breit-Wigner 1.00E+04 No 2.03E+02 4.21E+03 6.28E+03 28 60Ni Yes Reich-Moore 4.50E+05 No 1.25E+04 2.87E+04 4.31E+04 28 61Ni Yes Reich-Moore 7.00E+04 No 7.15E+03 7.55E+03 8.75E+03 28 62Ni Yes Reich-Moore 6.00E+05 No 4.54E+03 4.29E+04 9.47E+04 28 64Ni Yes Reich-Moore 6.00E+05 No 1.43E+04 3.38E+04 1.29E+05 29 63Cu Yes Reich-Moore 9.95E+04 No 5.79E+02 6.50E+02 8.07E+02 29 65Cu Yes Reich-Moore 9.95E+04 No 2.30E+02 1.36E+03 2.53E+03 30 natZn Yes ML Breit-Wigner 1.00E+05 No 2.82E+02 2.63E+03 4.17E+03 30 natZn Yes ML Breit-Wigner 1.00E+05 No 3.24E+02 4.05E+03 1.04E+04 30 natZn Yes ML Breit-Wigner 1.00E+05 No 2.23E+02 4.48E+02 5.88E+02 30 natZn Yes ML Breit-Wigner 1.00E+05 No 5.14E+02 5.04E+03 1.34E+04 30 natZn Yes ML Breit-Wigner 1.00E+05 No 1.80E+04 2.30E+04 3.19E+04 31 natGa Yes ML Breit-Wigner 5.90E+03 No 1.11E+02 3.34E+02 4.73E+02 73 31 natGa Yes ML Breit-Wigner 5.60E+03 No 9.57E+01 2.87E+02 3.76E+02 32 70Ge Yes ML Breit-Wigner 1.50E+04 No 1.12E+03 1.47E+03 1.94E+03 32 72Ge Yes ML Breit-Wigner 4.00E+04 No 2.61E+03 2.74E+03 3.65E+03 32 73Ge Yes ML Breit-Wigner 8.53E+03 No 1.02E+02 2.04E+02 2.25E+02 32 74Ge Yes ML Breit-Wigner 6.20E+04 No 2.85E+03 3.04E+03 4.99E+03 32 76Ge Yes ML Breit-Wigner 5.00E+04 No 5.50E+02 4.76E+03 1.39E+04 33 75As Yes ML Breit-Wigner 2.41E+03 No 4.70E+01 9.24E+01 2.53E+02 34 74Se Yes ML Breit-Wigner 2.40E+03 No 2.71E+01 2.72E+02 1.03E+03 34 76Se Yes ML Breit-Wigner 7.49E+03 No 3.77E+02 8.62E+02 2.12E+03 34 77Se Yes ML Breit-Wigner 2.72E+03 No 1.12E+02 2.12E+02 2.91E+02 34 78Se Yes ML Breit-Wigner 1.21E+04 No 3.83E+02 3.23E+03 6.17E+03 34 79Se No Yes 1.00E+05 34 80Se Yes ML Breit-Wigner 6.71E+03 No 1.97E+03 4.27E+03 4.72E+03 34 82Se Yes ML Breit-Wigner 3.11E+04 No 6.58E+03 1.38E+04 2.27E+04 35 79Br Yes ML Breit-Wigner 4.12E+02 No 3.58E+01 5.38E+01 1.58E+02 35 81Br Yes ML Breit-Wigner 3.63E+03 No 1.01E+02 1.36E+02 2.05E+02 36 78Kr Yes ML Breit-Wigner 8.65E+02 No 1.08E+02 4.51E+02 6.40E+02 36 80Kr Yes ML Breit-Wigner 1.00E+03 No 8.92E+01 1.06E+02 6.40E+02

Table 6. Resonance data in the JEFF-3.1 general-purpose neutron data library (cont.)

Z Isotope Res data Formalism Upper limit RRR Av param in URR Upper limit URR Energy of 3 s-wave res. (eV) 36 82Kr Yes ML Breit-Wigner 1.00E+02 No 3.98E+01 36 83Kr Yes ML Breit-Wigner 5.21E+02 No 2.79E+01 2.33E+02 36 84Kr Yes ML Breit-Wigner 2.00E+03 No 5.19E+02 5.80E+02 1.63E+03 36 85Kr No 36 86Kr Yes ML Breit-Wigner 1.30E+04 No 2.73E+03 37 85Rb Yes ML Breit-Wigner 9.23E+03 No 4.74E+02 5.26E+02 6.55E+02 37 86Rb No 37 87Rb Yes ML Breit-Wigner 1.59E+04 No 3.78E+02 3.84E+03 5.12E+03 38 84Sr Yes ML Breit-Wigner 3.50E+03 No 3.65E+02 5.17E+02 6.35E+02 38 86Sr Yes ML Breit-Wigner 1.99E+04 No 5.88E+02 3.25E+03 1.02E+04 38 87Sr Yes ML Breit-Wigner 2.89E+03 No 3.54E+00 6.48E+02 6.76E+02 38 88Sr Yes ML Breit-Wigner 1.66E+05 No 1.39E+04 9.25E+04 1.06E+05 38 89Sr No 38 90Sr No 74 39 89Y Yes SL Breit-Wigner 1.50E+05 No 2.60E+03 7.49E+03 9.73E+03 39 90Y No 39 91Y No 40 90Zr Yes ML Breit-Wigner 1.71E+05 No 3.86E+03 1.34E+04 1.73E+04 40 91Zr Yes ML Breit-Wigner 3.02E+04 Yes 1.00E+05 2.92E+02 6.81E+02 1.53E+03 40 92Zr Yes ML Breit-Wigner 7.10E+04 Yes 1.00E+05 2.69E+03 4.64E+03 6.81E+03 40 93Zr Yes ML Breit-Wigner 1.17E+02 Yes 5.00E+04 1.10E+02 40 94Zr Yes ML Breit-Wigner 5.35E+04 Yes 1.00E+05 2.24E+03 5.76E+03 7.07E+03 40 95Zr Yes ML Breit-Wigner 3.00E+03 Yes 5.00E+04 1.72E+02 4.43E+02 7.14E+02 40 96Zr Yes ML Breit-Wigner 1.00E+05 No 5.44E+03 1.54E+04 2.47E+04 41 93Nb Yes SL Breit-Wigner 7.35E+03 No 1.06E+02 1.19E+02 1.93E+02 41 94Nb Yes ML Breit-Wigner 2.80E+01 Yes 1.00E+05 1.16E+01 2.26E+01 41 95Nb No Yes 1.00E+05 42 100Mo Yes ML Breit-Wigner 2.60E+04 Yes 1.00E+05 3.64E+02 1.41E+03 1.94E+03 42 92Mo Yes ML Breit-Wigner 5.00E+04 Yes 1.00E+05 3.47E+02 3.18E+03 6.82E+03 42 94Mo Yes ML Breit-Wigner 2.00E+04 Yes 1.00E+05 1.54E+03 3.60E+03 4.62E+03 42 95Mo Yes ML Breit-Wigner 2.14E+03 Yes 2.06E+05 4.49E+01 1.60E+02 3.58E+02 42 96Mo Yes ML Breit-Wigner 1.90E+04 Yes 1.00E+05 1.31E+02 2.37E+03 3.29E+03

Table 6. Resonance data in the JEFF-3.1 general-purpose neutron data library (cont.)

Z Isotope Res data Formalism Upper limit RRR Av param in URR Upper limit URR Energy of 3 s-wave res. (eV) 42 97Mo Yes ML Breit-Wigner 1.80E+03 Yes 1.00E+05 7.09E+01 2.28E+02 2.68E+02 42 98Mo Yes ML Breit-Wigner 3.20E+04 Yes 1.00E+05 4.68E+02 1.53E+03 2.55E+03 42 99Mo No Yes 1.00E+05 43 99Tc Yes Reich-Moore 6.00E+03 Yes 1.00E+05 5.58E+00 2.03E+01 3.98E+01 44 100Ru Yes ML Breit-Wigner 5.02E+02 No 2.31E+02 44 101Ru Yes ML Breit-Wigner 1.04E+03 Yes 2.50E+04 1.57E+01 4.23E+01 5.21E+01 44 102Ru Yes SL Breit-Wigner 1.61E+03 No 1.30E+03 44 103Ru Yes ML Breit-Wigner 4.69E+01 Yes 3.00E+03 3.10E+00 9.30E+00 1.55E+01 44 104Ru Yes ML Breit-Wigner 1.51E+03 No 2.24E+02 3.81E+02 6.27E+02 44 105Ru No 44 106Ru No 44 96Ru No 44 98Ru No 44 99Ru Yes ML Breit-Wigner 1.01E+03 No 1.01E+01 2.52E+01 5.71E+01 75 45 103Rh Yes ML Breit-Wigner 4.12E+03 Yes 4.01E+04 1.26E+00 4.68E+01 6.83E+01 45 105Rh No 46 102Pd Yes ML Breit-Wigner 3.97E+02 No 1.91E+02 46 104Pd Yes ML Breit-Wigner 3.20E+03 Yes 1.00E+05 1.83E+02 2.95E+02 3.47E+02 46 105Pd Yes ML Breit-Wigner 2.06E+03 Yes 2.00E+04 1.18E+01 1.32E+01 2.52E+01 46 106Pd Yes ML Breit-Wigner 3.20E+03 Yes 1.00E+05 2.82E+02 4.08E+02 4.63E+02 46 107Pd Yes ML Breit-Wigner 6.60E+02 Yes 5.00E+04 3.92E+00 5.20E+00 6.83E+00 46 108Pd Yes ML Breit-Wigner 3.20E+03 Yes 1.00E+05 3.32E+01 9.10E+01 1.13E+02 46 110Pd Yes ML Breit-Wigner 2.90E+03 Yes 1.00E+05 2.63E+02 8.58E+02 9.00E+02 47 107Ag Yes ML Breit-Wigner 1.06E+03 No 1.63E+01 4.15E+01 4.48E+01 47 109Ag Yes ML Breit-Wigner 9.84E+02 Yes 8.00E+04 5.19E+00 3.04E+01 4.01E+01 47 110mAg Yes ML Breit-Wigner 1.25E+02 Yes 1.00E+05 1.15E+01 1.95E+01 2.12E+01 47 111Ag No 48 106Cd Yes ML Breit-Wigner 6.00E+03 Yes 1.00E+05 2.32E+02 4.56E+02 4.97E+02 48 108Cd Yes ML Breit-Wigner 6.10E+03 Yes 1.00E+05 5.42E+01 2.34E+02 3.12E+02 48 110Cd Yes ML Breit-Wigner 7.18E+03 Yes 1.00E+05 8.95E+01 2.31E+02 3.70E+02 48 111Cd Yes SL Breit-Wigner 2.40E+03 No 2.75E+01 6.94E+01 8.61E+01 48 112Cd Yes ML Breit-Wigner 7.35E+03 Yes 1.00E+05 6.68E+01 2.26E+02 4.43E+02

Table 6. Resonance data in the JEFF-3.1 general-purpose neutron data library (cont.)

Z Isotope Res data Formalism Upper limit RRR Av param in URR Upper limit URR Energy of 3 s-wave res. (eV) 48 113Cd Yes SL Breit-Wigner 2.25E+03 No 1.78E-01 1.84E+01 6.37E+01 48 114Cd Yes ML Breit-Wigner 8.00E+03 Yes 1.00E+05 1.20E+02 2.27E+02 3.92E+02 48 115mCd No 48 116Cd Yes ML Breit-Wigner 9.00E+03 Yes 1.00E+05 2.90E+01 6.76E+02 8.89E+02 49 113In Yes ML Breit-Wigner 8.30E+02 Yes 1.00E+05 1.80E+00 4.70E+00 1.46E+01 49 115In Yes ML Breit-Wigner 2.00E+03 Yes 1.00E+05 1.46E+00 3.82E+00 9.07E+00 50 112Sn Yes ML Breit-Wigner 1.50E+03 Yes 1.00E+05 6.47E+01 9.48E+01 1.05E+02 50 114Sn Yes ML Breit-Wigner 2.50E+03 Yes 1.00E+05 2.78E+02 6.66E+02 9.84E+02 50 115Sn Yes ML Breit-Wigner 9.50E+02 Yes 1.00E+05 2.93E+02 4.48E+02 4.94E+02 50 116Sn Yes ML Breit-Wigner 2.00E+03 Yes 1.00E+05 1.11E+02 1.34E+03 1.56E+03 50 117Sn Yes ML Breit-Wigner 2.35E+03 Yes 1.00E+05 3.88E+01 1.20E+02 1.24E+02 50 118Sn Yes ML Breit-Wigner 4.80E+03 Yes 1.00E+05 3.59E+02 7.71E+02 1.58E+03 50 119Sn Yes ML Breit-Wigner 1.30E+03 Yes 1.00E+05 1.15E+02 1.41E+02 2.23E+02 50 120Sn Yes ML Breit-Wigner 7.00E+04 Yes 1.00E+05 9.52E+02 3.12E+03 4.36E+03 76 50 122Sn Yes ML Breit-Wigner 2.90E+04 Yes 1.00E+05 1.75E+03 5.40E+03 6.82E+03 50 123Sn No 50 124Sn Yes ML Breit-Wigner 1.07E+04 Yes 1.00E+05 50 125Sn No 50 126Sn No 51 121Sb Yes ML Breit-Wigner 2.00E+03 Yes 1.00E+05 6.22E+00 1.54E+01 2.96E+01 51 123Sb Yes ML Breit-Wigner 2.50E+03 Yes 1.00E+05 2.14E+01 5.05E+01 7.67E+01 51 124Sb No 51 125Sb No 51 126Sb No 52 120Te No 52 122Te Yes ML Breit-Wigner 3.79E+03 No 7.28E+01 2.31E+02 3.33E+02 52 123Te Yes ML Breit-Wigner 6.33E+02 No 2.33E+00 2.41E+01 3.59E+01 52 124Te Yes ML Breit-Wigner 5.70E+03 No 3.52E+02 5.41E+02 6.79E+02 52 125Te Yes ML Breit-Wigner 1.34E+03 No 2.63E+01 1.06E+02 1.35E+02 52 126Te Yes ML Breit-Wigner 6.35E+03 No 2.01E+02 1.75E+03 2.15E+03 52 127mTe No 52 128Te Yes ML Breit-Wigner 1.29E+04 No 3.48E+02 4.24E+02 1.32E+03

Table 6. Resonance data in the JEFF-3.1 general-purpose neutron data library (cont.)

Z Isotope Res data Formalism Upper limit RRR Av param in URR Upper limit URR Energy of 3 s-wave res. (eV) 52 129mTe No 52 130Te Yes ML Breit-Wigner 3.43E+04 No 7.02E+03 1.00E+04 2.42E+04 52 132Te No 53 127I Yes Reich-Moore 5.20E+03 Yes 5.76E+04 2.04E+01 3.12E+01 3.77E+01 53 129I Yes Reich-Moore 5.10E+03 Yes 2.78E+04 4.14E+01 7.21E+01 7.61E+01 53 130I No 53 131I No 53 135I No 54 124Xe Yes ML Breit-Wigner 3.75E+02 No 5.16E+00 9.88E+00 2.52E+02 54 126Xe Yes ML Breit-Wigner 6.50E+02 No 8.66E+01 1.00E+02 4.60E+02 54 128Xe Yes ML Breit-Wigner 4.00E+03 No 2.38E+02 3.71E+02 5.55E+02 54 129Xe Yes ML Breit-Wigner 4.20E+03 No 9.44E+00 9.05E+01 9.23E+01 54 130Xe Yes ML Breit-Wigner 4.00E+03 No 4.30E+02 9.41E+02 1.48E+03 54 131Xe Yes ML Breit-Wigner 2.19E+03 No 1.44E+01 4.60E+01 7.56E+01 77 54 132Xe Yes ML Breit-Wigner 4.26E+03 No 6.43E+02 2.13E+03 2.85E+03 54 133Xe No 54 134Xe Yes ML Breit-Wigner 2.00E+03 No 1.00E+03 54 135Xe No 54 136Xe No 55 133Cs Yes ML Breit-Wigner 3.99E+03 Yes 8.16E+04 5.86E+00 2.25E+01 4.76E+01 55 134Cs Yes ML Breit-Wigner 1.87E+02 No 1.23E+01 3.90E+01 4.21E+01 55 135Cs Yes ML Breit-Wigner 4.63E+02 Yes 5.00E+04 5.00E+01 1.32E+02 2.14E+02 55 136Cs Yes ML Breit-Wigner 2.97E+02 No 4.23E+01 6.63E+01 1.36E+02 55 137Cs Yes ML Breit-Wigner 1.06E+04 Yes 1.00E+05 7.20E+02 1.52E+03 2.07E+03 56 130Ba Yes ML Breit-Wigner 2.53E+03 Yes 1.00E+05 4.63E+01 5.79E+01 1.36E+02 56 132Ba No Yes 1.00E+05 56 134Ba Yes ML Breit-Wigner 1.06E+04 Yes 1.00E+05 1.02E+02 2.63E+02 3.29E+02 56 135Ba Yes ML Breit-Wigner 5.96E+03 Yes 1.00E+05 2.42E+01 8.15E+01 8.72E+01 56 136Ba Yes ML Breit-Wigner 3.45E+04 Yes 1.00E+05 4.49E+02 5.11E+02 2.16E+03 56 137Ba Yes ML Breit-Wigner 1.19E+04 Yes 1.00E+05 1.27E+02 4.18E+02 5.77E+02 56 138Ba Yes ML Breit-Wigner 1.00E+05 No 7.88E+03 3.09E+04 4.03E+04 56 140Ba Yes ML Breit-Wigner 2.30E+04 Yes 1.00E+05 1.22E+02 1.07E+03 2.87E+03

Table 6. Resonance data in the JEFF-3.1 general-purpose neutron data library (cont.)

Z Isotope Res data Formalism Upper limit RRR Av param in URR Upper limit URR Energy of 3 s-wave res. (eV) 57 138La Yes ML Breit-Wigner 3.30E+02 Yes 1.00E+05 2.99E+00 2.10E+01 6.71E+01 57 139La Yes ML Breit-Wigner 1.08E+04 Yes 2.50E+05 7.22E+01 8.76E+02 9.63E+02 57 140La No 58 140Ce Yes ML Breit-Wigner 2.00E+04 No 2.54E+03 6.01E+03 9.57E+03 58 141Ce Yes ML Breit-Wigner 5.00E+02 Yes 1.00E+04 2.30E+01 5.90E+01 9.50E+01 58 142Ce Yes ML Breit-Wigner 4.88E+03 Yes 1.00E+04 1.29E+03 1.64E+03 2.74E+03 58 143Ce No 58 144Ce Yes ML Breit-Wigner 1.04E+04 Yes 2.00E+04 3.60E+02 1.16E+03 1.96E+03 59 141Pr Yes ML Breit-Wigner 5.77E+03 Yes 3.00E+05 8.52E+01 1.12E+02 2.19E+02 59 142Pr No 59 143Pr No 60 142Nd Yes ML Breit-Wigner 1.11E+04 No 2.19E+02 2.35E+02 6.36E+02 60 143Nd Yes ML Breit-Wigner 5.52E+03 Yes 1.00E+04 5.53E+01 1.27E+02 1.36E+02 60 144Nd Yes ML Breit-Wigner 1.20E+04 Yes 5.00E+04 3.74E+02 7.34E+02 1.28E+03 78 60 145Nd Yes ML Breit-Wigner 2.17E+03 Yes 5.00E+04 4.35E+00 4.25E+01 8.56E+01 60 146Nd Yes ML Breit-Wigner 1.00E+04 Yes 5.00E+04 3.60E+02 7.23E+02 8.10E+02 60 147Nd Yes ML Breit-Wigner 3.57E+01 No 4.79E+00 6.26E+00 1.03E+01 60 148Nd Yes ML Breit-Wigner 7.70E+03 Yes 5.00E+04 9.49E+01 1.55E+02 2.88E+02 60 150Nd Yes ML Breit-Wigner 5.12E+03 No 7.88E+01 2.66E+02 3.13E+02 61 147Pm Yes ML Breit-Wigner 3.19E+02 Yes 5.00E+04 5.36E+00 6.57E+00 6.92E+00 61 148Pm No 61 148mPm Yes ML Breit-Wigner 1.00E+00 No 1.69E-01 61 149Pm No 61 151Pm No 62 144Sm No 62 147Sm Yes ML Breit-Wigner 7.69E+02 Yes 7.00E+04 3.40E+00 1.83E+01 2.71E+01 62 148Sm No 62 149Sm Yes ML Breit-Wigner 5.20E+02 Yes 1.00E+05 9.73E-02 8.72E-01 4.95E+00 62 150Sm Yes ML Breit-Wigner 5.93E+02 No 2.07E+01 4.81E+01 1.11E+02 62 151Sm Yes ML Breit-Wigner 2.97E+02 Yes 4.00E+03 4.56E-01 1.09E+00 1.70E+00 62 152Sm Yes ML Breit-Wigner 3.69E+03 Yes 7.00E+04 8.05E+00 6.22E+01 8.77E+01 62 153Sm No

Table 6. Resonance data in the JEFF-3.1 general-purpose neutron data library (cont.)

Z Isotope Res data Formalism Upper limit RRR Av param in URR Upper limit URR Energy of 3 s-wave res. (eV) 62 154Sm Yes ML Breit-Wigner 3.10E+03 No 9.30E+01 2.61E+02 3.42E+02 63 151Eu Yes ML Breit-Wigner 9.88E+01 Yes 1.00E+04 3.21E-01 4.61E-01 1.06E+00 63 152Eu Yes ML Breit-Wigner 6.55E+00 Yes 1.00E+05 8.84E-01 1.34E+00 1.63E+00 63 153Eu Yes ML Breit-Wigner 9.72E+01 Yes 1.00E+05 1.73E+00 2.46E+00 3.29E+00 63 154Eu Yes ML Breit-Wigner 6.30E+01 Yes 1.00E+04 1.90E-01 8.57E-01 1.37E+00 63 155Eu Yes ML Breit-Wigner 3.75E+01 Yes 1.00E+04 6.03E-01 2.04E+00 7.19E+00 63 156Eu No 63 157Eu No 64 152Gd Yes ML Breit-Wigner 2.66E+03 Yes 1.00E+05 3.31E+00 8.00E+00 9.55E+00 64 154Gd Yes ML Breit-Wigner 2.76E+02 No 9.41E+00 1.16E+01 2.23E+01 64 155Gd Yes ML Breit-Wigner 1.82E+02 Yes 1.00E+05 2.68E-02 2.01E+00 2.57E+00 64 156Gd Yes ML Breit-Wigner 1.58E+03 Yes 5.00E+04 3.31E+01 8.02E+01 1.50E+02 64 157Gd Yes ML Breit-Wigner 2.15E+02 No 3.14E-02 2.83E+00 1.62E+01 64 158Gd Yes ML Breit-Wigner 6.04E+03 No 2.23E+01 1.01E+02 2.43E+02 79 64 160Gd Yes ML Breit-Wigner 2.88E+03 No 2.22E+02 4.22E+02 4.48E+02 65 159Tb Yes ML Breit-Wigner 1.70E+02 Yes 1.00E+04 3.34E+00 4.98E+00 1.11E+01 65 160Tb No 66 160Dy Yes ML Breit-Wigner 2.44E+01 No 1.88E+00 1.04E+01 2.05E+01 66 161Dy Yes ML Breit-Wigner 1.00E+03 Yes 1.00E+04 2.71E+00 3.68E+00 4.33E+00 66 162Dy Yes ML Breit-Wigner 1.22E+04 Yes 5.00E+04 5.44E+00 7.11E+01 1.17E+02 66 163Dy Yes ML Breit-Wigner 1.00E+03 No 1.71E+00 5.81E+00 1.62E+01 66 164Dy Yes ML Breit-Wigner 1.60E+04 Yes 5.00E+04 1.47E+02 4.50E+02 5.36E+02 67 165Ho Yes SL Breit-Wigner 1.52E+02 No 3.92E+00 8.16E+00 1.28E+01 68 162Er Yes ML Breit-Wigner 2.50E+02 No 5.48E+00 7.60E+00 1.46E+01 68 164Er Yes ML Breit-Wigner 8.00E+02 No 7.92E+00 3.05E+01 4.97E+01 68 166Er Yes ML Breit-Wigner 3.00E+03 No 1.56E+01 7.38E+01 8.17E+01 68 167Er Yes ML Breit-Wigner 5.91E+02 Yes 1.00E+04 4.60E-01 5.83E-01 5.99E+00 68 168Er Yes ML Breit-Wigner 3.50E+03 No 7.97E+01 1.89E+02 2.44E+02 68 170Er Yes ML Breit-Wigner 3.00E+03 No 9.51E+01 2.84E+02 4.97E+02 71 175Lu Yes ML Breit-Wigner 4.11E+02 Yes 1.00E+04 2.59E+00 4.75E+00 5.22E+00 71 176Lu Yes ML Breit-Wigner 1.02E+02 Yes 1.00E+04 1.41E-01 1.57E+00 4.36E+00 72 174Hf Yes Reich-Moore 2.20E+02 Yes 5.00E+04 4.06E+00 1.34E+01 3.00E+01

Table 6. Resonance data in the JEFF-3.1 general-purpose neutron data library (cont.)

Z Isotope Res data Formalism Upper limit RRR Av param in URR Upper limit URR Energy of 3 s-wave res. (eV) 72 176Hf Yes Reich-Moore 7.00E+02 Yes 5.00E+04 7.89E+00 4.83E+01 5.33E+01 72 177Hf Yes Reich-Moore 2.50E+02 Yes 5.00E+04 1.10E+00 2.39E+00 5.90E+00 72 178Hf Yes Reich-Moore 1.50E+03 Yes 5.00E+04 7.79E+00 1.05E+02 1.65E+02 72 179Hf Yes Reich-Moore 2.50E+02 Yes 5.00E+04 5.69E+00 1.77E+01 1.91E+01 72 180Hf Yes Reich-Moore 2.50E+03 Yes 5.00E+04 7.25E+01 1.72E+02 4.47E+02 73 181Ta Yes ML Breit-Wigner 2.40E+03 Yes 1.00E+05 4.28E+00 1.04E+01 1.40E+01 73 182Ta Yes ML Breit-Wigner 3.50E+01 Yes 1.00E+04 1.47E-01 1.82E+00 5.98E+00 74 182W Yes ML Breit-Wigner 1.20E+04 No 4.15E+00 2.11E+01 1.14E+02 74 183W Yes ML Breit-Wigner 2.20E+03 No 7.64E+00 2.70E+01 4.06E+01 74 184W Yes ML Breit-Wigner 1.50E+04 No 1.02E+02 1.69E+02 1.84E+02 74 186W Yes ML Breit-Wigner 1.50E+04 No 1.88E+01 1.71E+02 2.18E+02 75 185Re Yes ML Breit-Wigner 2.00E+03 Yes 3.50E+04 2.16E+00 5.92E+00 7.22E+00 75 187Re Yes ML Breit-Wigner 2.00E+03 Yes 3.50E+04 4.42E+00 1.11E+01 1.60E+01 76 natOs Yes ML Breit-Wigner 1.60E+02 No 80 76 natOs Yes ML Breit-Wigner 3.36E+03 No 2.24E+01 4.47E+01 6.63E+01 76 natOs Yes ML Breit-Wigner 9.90E+02 No 9.47E+00 1.27E+01 2.02E+01 76 natOs Yes ML Breit-Wigner 5.00E+03 No 3.87E+01 7.86E+01 1.50E+02 76 natOs Yes ML Breit-Wigner 7.70E+01 No 6.71E+00 8.96E+00 1.03E+01 76 natOs Yes ML Breit-Wigner 1.50E+02 No 9.16E+01 1.45E+02 2.02E+02 76 natOs Yes ML Breit-Wigner 1.50E+02 No 2.05E+01 1.28E+02 2.00E+02 77 191Ir Yes ML Breit-Wigner 1.60E+02 Yes 1.00E+04 6.53E-01 5.36E+00 6.13E+00 77 193Ir Yes ML Breit-Wigner 3.00E+02 Yes 1.00E+04 1.30E+00 9.07E+00 2.45E+01 78 natPt No 79 197Au Yes ML Breit-Wigner 5.00E+03 No 4.91E+00 4.65E+01 5.81E+01 80 196Hg Yes ML Breit-Wigner 1.03E+02 No 9.35E+01 80 198Hg Yes ML Breit-Wigner 4.59E+02 No 2.31E+01 8.98E+01 3.03E+02 80 199Hg Yes ML Breit-Wigner 9.68E+02 No 3.35E+01 1.30E+02 1.75E+02 80 200Hg Yes ML Breit-Wigner 8.58E+03 No 1.33E+03 2.71E+03 5.06E+03 80 201Hg Yes ML Breit-Wigner 7.54E+02 No 4.30E+01 7.09E+01 2.10E+02 80 202Hg Yes ML Breit-Wigner 4.52E+03 No 1.71E+03 4.11E+03 80 204Hg No 81 natTl Yes SL Breit-Wigner 5.00E+04 No 2.38E+02 8.42E+02 1.13E+03

Table 6. Resonance data in the JEFF-3.1 general-purpose neutron data library (cont.)

Z Isotope Res data Formalism Upper limit RRR Av param in URR Upper limit URR Energy of 3 s-wave res. (eV) 81 natTl Yes SL Breit-Wigner 5.00E+04 No 2.79E+03 3.04E+03 5.10E+03 82 204Pb Yes ML Breit-Wigner 5.00E+04 No 1.69E+03 2.48E+03 6.70E+03 82 206Pb Yes Reich-Moore 9.00E+05 No 1.64E+04 6.60E+04 9.26E+04 82 207Pb Yes Reich-Moore 4.75E+05 No 4.13E+04 1.02E+05 1.82E+05 82 208Pb Yes Reich-Moore 1.00E+06 No 5.06E+05 8.88E+05 9.98E+05 83 209Bi Yes ML Breit-Wigner 2.00E+05 No 8.00E+02 2.31E+03 5.11E+03 88 223Ra No 88 224Ra No 88 225Ra No 88 226Ra Yes ML Breit-Wigner 1.00E+03 No 5.37E-01 2.43E+01 3.91E+01 89 225Ac No 89 226Ac No 89 227Ac No 90 227Th No 81 90 228Th Yes ML Breit-Wigner 7.80E+00 No 1.90E+00 7.55E+00 90 229Th Yes SL Breit-Wigner 9.50E+00 No 6.09E-01 1.26E+00 1.44E+00 90 230Th Yes ML Breit-Wigner 2.51E+02 No 1.43E+00 1.73E+01 2.38E+01 90 232Th Yes ML Breit-Wigner 4.00E+03 Yes 1.50E+05 2.18E+01 2.35E+01 5.95E+01 90 233Th No 90 234Th No 91 231Pa Yes ML Breit-Wigner 1.43E+01 Yes 1.00E+03 3.96E-01 4.94E-01 7.43E-01 91 232Pa Yes ML Breit-Wigner 1.00E+01 No 3.30E-01 6.90E-01 1.40E+00 91 233Pa Yes ML Breit-Wigner 3.85E+01 Yes 1.00E+04 7.95E-01 1.34E+00 1.64E+00 92 232U Yes ML Breit-Wigner 5.30E+01 Yes 1.00E+03 5.98E+00 1.27E+01 2.10E+01 92 233U Yes Reich-Moore 1.50E+02 Yes 3.00E+04 1.70E-01 4.40E-01 1.43E+00 92 234U Yes ML Breit-Wigner 1.50E+03 Yes 1.40E+05 5.16E+00 2.27E+01 2.34E+01 92 235U Yes Reich-Moore 2.25E+03 Yes 2.50E+04 3.66E-05 2.74E-01 1.13E+00 92 236U Yes ML Breit-Wigner 1.50E+03 No 5.45E+00 2.97E+01 3.41E+01 92 237U Yes ML Breit-Wigner 2.00E+02 Yes 1.00E+04 1.50E+00 5.00E+00 8.50E+00 92 238U Yes Reich-Moore 2.00E+04 Yes 3.00E+05 6.67E+00 2.09E+01 3.67E+01 93 235Np No 93 236Np No

Table 6. Resonance data in the JEFF-3.1 general-purpose neutron data library (cont.)

Z Isotope Res data Formalism Upper limit RRR Av param in URR Upper limit URR Energy of 3 s-wave res. (eV) 93 237Np Yes ML Breit-Wigner 5.00E+02 Yes 3.50E+04 4.90E-01 1.32E+00 1.48E+00 93 238Np Yes ML Breit-Wigner 1.00E+02 Yes 1.00E+04 1.52E+00 2.57E+00 3.62E+00 93 239Np No 94 236Pu Yes ML Breit-Wigner 1.00E+01 Yes 3.00E+04 3.16E+00 6.30E+00 1.20E+01 94 237Pu No 94 238Pu Yes ML Breit-Wigner 5.00E+02 Yes 1.50E+05 2.90E+00 9.98E+00 1.86E+01 94 239Pu Yes Reich-Moore 2.50E+03 Yes 3.00E+04 2.96E-01 7.82E+00 1.09E+01 94 240Pu Yes Reich-Moore 5.70E+03 Yes 4.00E+04 1.06E+00 2.04E+01 3.83E+01 94 241Pu Yes Reich-Moore 3.00E+02 Yes 3.00E+04 1.50E-01 2.64E-01 1.73E+00 94 242Pu Yes ML Breit-Wigner 1.15E+03 Yes 4.00E+04 2.67E+00 1.46E+01 2.26E+01 94 243Pu Yes ML Breit-Wigner 1.02E+02 Yes 1.00E+04 1.66E+00 4.16E+00 6.66E+00 94 244Pu Yes ML Breit-Wigner 2.49E+02 Yes 1.00E+04 2.12E+01 3.26E+01 4.40E+01 94 246Pu No 95 241Am Yes ML Breit-Wigner 1.50E+02 Yes 4.00E+04 3.07E-01 5.76E-01 1.27E+00 82 95 242Am Yes ML Breit-Wigner 1.00E+02 Yes 4.43E+04 3.50E-01 2.19E+00 2.31E+00 95 242mAm Yes ML Breit-Wigner 4.30E+01 Yes 2.73E+04 1.78E-01 6.15E-01 1.10E+00 95 243Am Yes ML Breit-Wigner 2.50E+02 Yes 4.24E+04 4.19E-01 9.83E-01 1.36E+00 95 244Am No 95 244mAm No 96 240Cm Yes ML Breit-Wigner 1.50E+02 Yes 3.00E+04 3.39E+00 1.02E+01 1.69E+01 96 241Cm No 96 242Cm Yes ML Breit-Wigner 2.75E+02 Yes 4.00E+04 1.36E+01 3.03E+01 3.75E+01 96 243Cm Yes SL Breit-Wigner 7.00E+01 Yes 4.00E+04 6.71E-01 1.14E+00 1.47E+00 96 244Cm Yes SL Breit-Wigner 5.25E+02 Yes 1.00E+04 7.67E+00 1.68E+01 2.28E+01 96 245Cm Yes ML Breit-Wigner 1.00E+02 Yes 5.50E+04 8.50E-01 1.98E+00 2.15E+00 96 246Cm Yes ML Breit-Wigner 4.00E+02 Yes 4.30E+04 4.32E+00 1.53E+01 3.30E+01 96 247Cm Yes SL Breit-Wigner 6.00E+01 Yes 3.00E+04 1.25E+00 2.92E+00 3.19E+00 96 248Cm Yes ML Breit-Wigner 1.50E+03 Yes 3.00E+04 7.25E+00 2.69E+01 3.50E+01 96 249Cm Yes ML Breit-Wigner 1.50E+02 Yes 2.50E+04 1.50E+01 3.00E+01 5.00E+01 96 250Cm Yes ML Breit-Wigner 1.50E+02 Yes 3.00E+04 2.00E+01 2.70E+02 4.50E+02 97 247Bk Yes SL Breit-Wigner 1.50E+02 No 1.01E+00 3.02E+00 5.03E+00 97 249Bk Yes ML Breit-Wigner 6.00E+01 Yes 3.00E+04 1.95E-01 1.34E+00 1.60E+00

Table 6. Resonance data in the JEFF-3.1 general-purpose neutron data library (cont.)

Z Isotope Res data Formalism Upper limit RRR Av param in URR Upper limit URR Energy of 3 s-wave res. (eV) 97 250Bk Yes ML Breit-Wigner 1.00E+02 Yes 3.00E+04 1.82E+00 3.91E+00 6.00E+00 98 249Cf Yes ML Breit-Wigner 7.00E+01 Yes 3.00E+04 7.00E-01 3.88E+00 5.07E+00 98 250Cf Yes SL Breit-Wigner 1.50E+02 Yes 3.00E+04 1.00E+00 1.70E+01 3.30E+01 98 251Cf Yes ML Breit-Wigner 1.64E+02 Yes 1.00E+04 7.12E-01 8.87E+00 1.70E+01 98 252Cf Yes ML Breit-Wigner 3.67E+02 Yes 1.00E+04 1.71E+01 3.46E+01 5.21E+01 98 254Cf No 99 253Es Yes ML Breit-Wigner 1.01E+02 Yes 1.00E+04 5.86E-01 4.31E+00 8.04E+00 99 254Es No 99 255Es No 100 255Fm No 83

Table 7. Contents of the JEFF-3.1 thermal scattering law library

Material MAT no. MF1 MF7

H(H2O) 1 451 4 H(ZrH) 7 451 2, 4

H(CaH2) 8 451 2, 4

D(D2O) 11 451 4 4Be 26 451 2, 4 Graphite 31 451 2, 4

H(CH2) 37 451 2, 4 24Mg 52 451 2, 4

Ca(CaH2) 59 451 2, 4

Table 8. Contents of the JEFF-3.1 proton special-purpose library. List of available reactions (MTs) in files MF=3 to MF=6.

Z Sym MF1 MF3 MF6 20 40Ca 451 2, 5 2, 5 20 42Ca 451 2, 5 2, 5 20 43Ca 451 2, 5 2, 5 20 44Ca 451 2, 5 2, 5 20 46Ca 451 2, 5 2, 5 20 48Ca 451 2, 5 2, 5 21 45Sc 451 2, 5 2, 5 22 46Ti 451 2, 5 2, 5 22 47Ti 451 2, 5 2, 5 22 48Ti 451 2, 5 2, 5 22 49Ti 451 2, 5 2, 5 22 50Ti 451 2, 5 2, 5 26 54Fe 451 2, 5 2, 5

85 26 56Fe 451 2, 5 2, 5 26 57Fe 451 2, 5 2, 5 26 58Fe 451 2, 5 2, 5 32 70Ge 451 2, 5 2, 5 32 72Ge 451 2, 5 2, 5 32 73Ge 451 2, 5 2, 5 32 74Ge 451 2, 5 2, 5 32 76Ge 451 2, 5 2, 5 82 204Pb 451 2, 5 2, 5 82 206Pb 451 2, 5 2, 5 82 207Pb 451 2, 5 2, 5 82 208Pb 451 2, 5 2, 5 83 209Bi 451 2, 5 2, 5

Table 9. Simple integral neutron cross-section data from JEFF-3.1 calculated with the INTER computer code

V Avg V G fact Avg res integ Avg V fiss V (E14) MT Reaction (2200 m/s) [1E-5 to 10] eV (Avg V)/V(E0) [5E-1 to 1E+5] eV [1E+3 to 2E+7] eV 1.4E+7 eV Material(MAT)=125:H-1 001 Total 3.04548E+01 3.30070E+01 1.08381 2.40810E+02 4.00392E+00 6.92003E-01 002 Elastic 3.01228E+01 3.26750E+01 1.08473 2.40661E+02 4.00390E+00 6.91974E-01 102 n,gamma 3.31982E-01 3.32070E-01 1.00026 1.49128E-01 3.95867E-05 2.97350E-05 Material(MAT)=128:H-2 001 Total 4.23413E+00 4.69340E+00 1.10848 4.13548E+01 2.54100E+00 8.12870E-01 002 Elastic 4.23362E+00 4.69290E+00 1.10849 4.13546E+01 2.53520E+00 6.46437E-01 102 n,gamma 5.05707E-04 5.05780E-04 1.00015 2.28669E-04 6.99577E-06 9.53295E-06 Material(MAT)=131:H-3 001 Total 1.98141E+00 2.21350E+00 1.11712 2.05035E+01 1.95013E+00 9.73764E-01 002 Elastic 1.98141E+00 2.21350E+00 1.11712 2.05035E+01 1.94993E+00 9.23564E-01 Material(MAT)=225:He-3 001 Total 5.31964E+03 5.32140E+03 1.00032 2.41544E+03 2.96585E+00 1.16941E+00 002 Elastic 3.86100E+00 4.31490E+00 1.11757 3.86455E+01 2.14796E+00 9.71905E-01 102 n,gamma 3.10003E-05 3.10090E-05 1.00027 1.39449E-05 4.80149E-09 1.32057E-09

86 Material(MAT)=228:He-4 001 Total 8.54660E-01 9.58730E-01 1.12176 9.28162E+00 3.66169E+00 1.05708E+00 002 Elastic 8.54660E-01 9.58730E-01 1.12176 9.28162E+00 3.66169E+00 1.05708E+00 Material(MAT)=325:Li-6 001 Total 9.41742E+02 9.42080E+02 1.00036 4.31319E+02 1.92581E+00 1.45307E+00 002 Elastic 7.27754E-01 8.18980E-01 1.12535 8.18293E+00 1.44486E+00 8.65891E-01 102 n,gamma 3.85004E-02 3.85110E-02 1.00027 1.73185E-02 1.17405E-05 1.02286E-05 Material(MAT)=328:Li-7 001 Total 1.08499E+00 1.21640E+00 1.12110 1.25033E+01 1.81894E+00 1.44555E+00 002 Elastic 1.03958E+00 1.17100E+00 1.12638 1.24829E+01 1.64065E+00 1.01451E+00 102 n,gamma 4.54021E-02 4.54160E-02 1.00030 2.04148E-02 7.19468E-06 4.12092E-06 Material(MAT)=425:Be-9 001 Total 6.51084E+00 7.33650E+00 1.12682 7.41846E+01 2.70883E+00 1.01556E+00 002 Elastic 6.50206E+00 7.32780E+00 1.12699 7.41806E+01 2.67479E+00 9.87723E-01 102 n,gamma 8.77353E-03 8.77590E-03 1.00027 3.95827E-03 1.16436E-05 1.24715E-06 Material(MAT)=525:B-10 001 Total 3.84215E+03 3.84330E+03 1.00030 1.75052E+03 2.64081E+00 1.46866E+00 002 Elastic 2.25048E+00 2.53740E+00 1.12749 2.63854E+01 2.08994E+00 9.08833E-01 102 n,gamma 4.99872E-01 4.99840E-01 0.99994 2.23447E-01 3.32211E-05 0.00000E+00

Table 9. Simple integral neutron cross-section data from JEFF-3.1 calculated with the INTER computer code (cont.)

V Avg V G fact Avg res integ Avg V fiss V (E14) MT Reaction (2200 m/s) [1E-5 to 10] eV (Avg V)/V(E0) [5E-1 to 1E+5] eV [1E+3 to 2E+7] eV 1.4E+7 eV Material(MAT)=528:B-11 001 Total 5.06677E+00 5.71290E+00 1.12753 5.93102E+01 2.44559E+00 1.41534E+00 002 Elastic 5.06127E+00 5.70740E+00 1.12766 5.93074E+01 2.41371E+00 7.91187E-01 102 n,gamma 5.50006E-03 5.50200E-03 1.00035 2.74777E-03 2.78593E-06 2.14331E-07 Material(MAT)=600:C-Nat 001 Total 4.94136E+00 5.57250E+00 1.12774 5.75769E+01 2.38916E+00 1.30237E+00 002 Elastic 4.93799E+00 5.56920E+00 1.12782 5.75754E+01 2.37556E+00 8.19286E-01 102 n,gamma 3.36705E-03 3.37740E-03 1.00307 1.52817E-03 9.48826E-06 1.15189E-04 Material(MAT)=725:N-14 001 Total 1.21698E+01 1.34850E+01 1.10808 1.11668E+02 2.00777E+00 1.56850E+00 002 Elastic 1.02676E+01 1.15820E+01 1.12806 1.10813E+02 1.86926E+00 9.69143E-01 102 n,gamma 7.49915E-02 7.49970E-02 1.00008 3.36688E-02 3.65302E-05 1.72232E-05 Material(MAT)=728:N-15 001 Total 4.56325E+00 5.13890E+00 1.12615 5.21077E+01 2.61947E+00 1.65255E+00 002 Elastic 4.56322E+00 5.13890E+00 1.12615 5.21077E+01 2.60717E+00 9.61092E-01

87 102 n,gamma 2.40099E-05 2.40330E-05 1.00096 3.20621E-05 1.87229E-05 8.60002E-06 Material(MAT)=825:O-16 001 Total 3.97321E+00 4.48260E+00 1.12822 4.68501E+01 2.80403E+00 1.59271E+00 002 Elastic 3.97302E+00 4.48250E+00 1.12822 4.68500E+01 2.78686E+00 9.57242E-01 102 n,gamma 1.90002E-04 1.90050E-04 1.00028 8.54667E-05 2.94284E-08 8.09378E-09 Material(MAT)=828:O-17 001 Total 4.08957E+00 4.58360E+00 1.12079 4.49406E+01 1.52525E+00 1.67803E+00 002 Elastic 3.85073E+00 4.34470E+00 1.12827 4.48329E+01 1.36951E+00 7.20000E-01 102 n,gamma 3.82863E-03 3.82720E-03 0.99962 1.71412E-03 1.93617E-04 2.08823E-04 Material(MAT)=925:F-19 001 Total 3.74669E+00 4.22520E+00 1.12771 5.14824E+01 3.63682E+00 1.75887E+00 002 Elastic 3.73711E+00 4.21560E+00 1.12804 5.14672E+01 2.64410E+00 9.45527E-01 102 n,gamma 9.57845E-03 9.58110E-03 1.00028 1.51906E-02 2.08080E-04 3.53925E-05 Material(MAT)=1122:Na-22 001 Total 2.90779E+04 2.91360E+04 1.00202 1.57862E+04 2.94234E+00 1.61298E+00 002 Elastic 7.41575E+01 8.36890E+01 1.12853 9.63692E+02 2.04851E+00 7.29026E-01 102 n,gamma 2.52100E+02 2.52580E+02 1.00192 1.28624E+02 7.72932E-04 2.10470E-04 Material(MAT)=1125:Na-23 001 Total 3.62117E+00 4.01830E+00 1.10968 1.23147E+02 3.42022E+00 1.80909E+00 002 Elastic 3.08980E+00 3.48680E+00 1.12849 1.22836E+02 3.04276E+00 1.01904E+00 102 n,gamma 5.31372E-01 5.31550E-01 1.00034 3.10635E-01 2.30165E-04 1.48817E-04

Table 9. Simple integral neutron cross-section data from JEFF-3.1 calculated with the INTER computer code (cont.)

V Avg V G fact Avg res integ Avg V fiss V (E14) MT Reaction (2200 m/s) [1E-5 to 10] eV (Avg V)/V(E0) [5E-1 to 1E+5] eV [1E+3 to 2E+7] eV 1.4E+7 eV Material(MAT)=1225:Mg-24 001 Total 3.87745E+00 4.36920E+00 1.12683 5.40101E+01 3.55191E+00 1.74764E+00 002 Elastic 3.82716E+00 4.31890E+00 1.12849 5.39809E+01 3.20560E+00 5.24937E-01 102 n,gamma 5.02895E-02 5.03060E-02 1.00033 2.91613E-02 3.88811E-04 3.44609E-05 Material(MAT)=1228:Mg-25 001 Total 2.83737E+00 3.17760E+00 1.11992 4.13022E+01 3.10292E+00 1.82570E+00 002 Elastic 2.64699E+00 2.98720E+00 1.12852 4.12050E+01 2.73706E+00 6.85733E-01 102 n,gamma 1.90374E-01 1.90440E-01 1.00033 9.73437E-02 3.57814E-04 6.98834E-06 Material(MAT)=1231:Mg-26 001 Total 2.92345E+00 3.29430E+00 1.12685 3.39465E+01 3.28270E+00 1.85076E+00 002 Elastic 2.88514E+00 3.25600E+00 1.12853 3.39293E+01 3.04993E+00 6.81588E-01 102 n,gamma 3.83094E-02 3.83220E-02 1.00033 1.72200E-02 3.22012E-04 2.12583E-05 Material(MAT)=1325:Al-27 001 Total 1.68510E+00 1.87180E+00 1.11079 2.38144E+01 3.24197E+00 1.74426E+00 002 Elastic 1.45164E+00 1.63820E+00 1.12855 2.36868E+01 2.92519E+00 7.79437E-01

88 102 n,gamma 2.33463E-01 2.33540E-01 1.00033 1.27603E-01 5.46344E-04 6.27689E-04 Material(MAT)=1425:Si-28 001 Total 2.16030E+00 2.41630E+00 1.11851 2.38949E+01 3.16056E+00 1.73039E+00 002 Elastic 1.99117E+00 2.24710E+00 1.12855 2.38130E+01 2.95997E+00 6.96955E-01 102 n,gamma 1.69135E-01 1.69190E-01 1.00033 8.18819E-02 6.91704E-04 6.32557E-04 Material(MAT)=1428:Si-29 001 Total 2.99378E+00 3.36580E+00 1.12428 3.36688E+01 3.61851E+00 1.75658E+00 002 Elastic 2.89228E+00 3.26420E+00 1.12858 3.36041E+01 3.30146E+00 7.70796E-01 102 n,gamma 1.01499E-01 1.01690E-01 1.00188 6.47622E-02 2.99550E-04 3.25080E-04 Material(MAT)=1431:Si-30 001 Total 2.64032E+00 2.96650E+00 1.12355 3.53750E+01 3.59449E+00 1.77936E+00 002 Elastic 2.53256E+00 2.85820E+00 1.12858 3.46707E+01 3.41825E+00 7.37451E-01 102 n,gamma 1.07758E-01 1.08350E-01 1.00550 7.04389E-01 6.78705E-04 4.79509E-04 Material(MAT)=1525:P-31 001 Total 3.35111E+00 3.76070E+00 1.12223 3.72489E+01 2.88188E+00 1.83244E+00 002 Elastic 3.18493E+00 3.59450E+00 1.12859 3.71731E+01 2.55479E+00 1.16143E+00 102 n,gamma 1.66183E-01 1.66240E-01 1.00033 7.58402E-02 1.01302E-03 9.89520E-06 Material(MAT)=1625:S-32 001 Total 1.51395E+00 1.64090E+00 1.08385 1.22652E+01 2.77745E+00 1.73826E+00 002 Elastic 9.78639E-01 1.10450E+00 1.12860 1.19346E+01 2.53441E+00 7.64114E-01 102 n,gamma 5.28204E-01 5.28370E-01 1.00032 2.45208E-01 5.91342E-04 3.23500E-06

Table 9. Simple integral neutron cross-section data from JEFF-3.1 calculated with the INTER computer code (cont.)

V Avg V G fact Avg res integ Avg V fiss V (E14) MT Reaction (2200 m/s) [1E-5 to 10] eV (Avg V)/V(E0) [5E-1 to 1E+5] eV [1E+3 to 2E+7] eV 1.4E+7 eV Material(MAT)=1628:S-33 001 Total 3.41074E+00 3.80450E+00 1.11543 3.45035E+01 2.70933E+00 1.76369E+00 002 Elastic 2.88785E+00 3.25920E+00 1.12860 3.22619E+01 2.32269E+00 7.29608E-01 102 n,gamma 3.50050E-01 3.50160E-01 1.00032 1.63419E-01 2.06035E-04 7.15924E-07 Material(MAT)=1631:S-34 001 Total 2.33180E+00 2.60300E+00 1.11631 2.39443E+01 2.64806E+00 1.77993E+00 002 Elastic 2.10820E+00 2.37930E+00 1.12861 2.38451E+01 2.51616E+00 7.48048E-01 102 n,gamma 2.23602E-01 2.23670E-01 1.00032 9.92256E-02 2.77325E-04 3.66670E-06 Material(MAT)=1637:S-36 001 Total 2.36577E+00 2.65280E+00 1.12133 2.67993E+01 2.64465E+00 1.82909E+00 002 Elastic 2.21531E+00 2.50030E+00 1.12862 2.66792E+01 2.56528E+00 7.93447E-01 102 n,gamma 1.50455E-01 1.52560E-01 1.01398 1.20122E-01 2.49779E-04 1.36924E-06 Material(MAT)=1725:Cl-35 001 Total 6.50682E+01 6.77940E+01 1.04189 1.37649E+02 2.82680E+00 2.02639E+00 002 Elastic 2.09690E+01 2.36610E+01 1.12838 1.19210E+02 2.47621E+00 8.90589E-01

89 102 n,gamma 4.36185E+01 4.36520E+01 1.00077 1.79651E+01 8.80508E-04 4.78074E-04 Material(MAT)=1731:Cl-37 001 Total 1.59639E+00 1.74630E+00 1.09393 1.89685E+01 2.61868E+00 2.11029E+00 002 Elastic 1.16328E+00 1.31290E+00 1.12862 1.87656E+01 2.45384E+00 1.00833E+00 102 n,gamma 4.33111E-01 4.33430E-01 1.00074 2.02939E-01 4.36191E-04 2.86825E-04 Material(MAT)=1825:Ar-36 001 Total 7.96350E+01 8.93080E+01 1.12147 6.69243E+02 2.86333E+00 1.82368E+00 002 Elastic 7.45911E+01 8.41850E+01 1.12862 6.66982E+02 2.51921E+00 7.92078E-01 102 n,gamma 5.04389E+00 5.12340E+00 1.01576 2.26121E+00 3.41829E-03 4.38002E-04 Material(MAT)=1831:Ar-38 001 Total 9.74215E+00 1.09250E+01 1.12138 7.89270E+01 2.75368E+00 1.94775E+00 002 Elastic 8.94034E+00 1.00900E+01 1.12863 7.85736E+01 2.55119E+00 8.35933E-01 102 n,gamma 8.01803E-01 8.34300E-01 1.04053 3.53343E-01 7.90696E-04 1.36301E-04 Material(MAT)=1837:Ar-40 001 Total 1.31519E+00 1.39970E+00 1.06422 1.07072E+01 2.97259E+00 2.29042E+00 002 Elastic 6.55161E-01 7.39440E-01 1.12865 1.04184E+01 2.55313E+00 7.88040E-01 102 n,gamma 6.60032E-01 6.60210E-01 1.00027 2.88811E-01 1.01837E-03 3.46454E-04 Material(MAT)=1925:K-39 001 Total 4.19098E+00 4.46030E+00 1.06426 2.27656E+01 2.79479E+00 2.01647E+00 002 Elastic 2.08870E+00 2.35740E+00 1.12863 2.16905E+01 2.58524E+00 9.59432E-01 102 n,gamma 2.09798E+00 2.09860E+00 1.00028 1.07007E+00 2.69600E-03 7.42755E-06

Table 9. Simple integral neutron cross-section data from JEFF-3.1 calculated with the INTER computer code (cont.)

V Avg V G fact Avg res integ Avg V fiss V (E14) MT Reaction (2200 m/s) [1E-5 to 10] eV (Avg V)/V(E0) [5E-1 to 1E+5] eV [1E+3 to 2E+7] eV 1.4E+7 eV Material(MAT)=1928:K-40 001 Total 3.75618E+01 3.79020E+01 1.00905 4.94325E+01 2.84401E+00 2.05040E+00 002 Elastic 2.78538E+00 3.14370E+00 1.12864 3.35851E+01 2.29664E+00 9.81356E-01 102 n,gamma 3.00006E+01 3.00090E+01 1.00028 1.34665E+01 1.58449E-03 4.10919E-06 Material(MAT)=1931:K-41 001 Total 4.05690E+00 4.39150E+00 1.08249 3.64086E+01 2.86509E+00 2.08421E+00 002 Elastic 2.59773E+00 2.93190E+00 1.12866 3.48571E+01 2.51293E+00 1.00108E+00 102 n,gamma 1.45918E+00 1.45960E+00 1.00029 1.55151E+00 5.95575E-03 1.73407E-05 Material(MAT)=2025:Ca-40 001 Total 3.46997E+00 3.86380E+00 1.11349 3.31713E+01 2.83215E+00 2.15785E+00 002 Elastic 3.06004E+00 3.45370E+00 1.12865 3.29644E+01 2.63279E+00 9.13613E-01 102 n,gamma 4.07531E-01 4.07650E-01 1.00030 2.05769E-01 1.51393E-03 4.12903E-04 Material(MAT)=2031:Ca-42 001 Total 1.91957E+00 2.07880E+00 1.08296 1.56843E+01 3.37225E+00 2.15703E+00 002 Elastic 1.23650E+00 1.39560E+00 1.12865 1.53210E+01 2.93185E+00 8.93025E-01

90 102 n,gamma 6.83070E-01 6.83240E-01 1.00024 3.63336E-01 2.66602E-03 1.03534E-03 Material(MAT)=2034:Ca-43 001 Total 1.58735E+01 1.64180E+01 1.03428 1.36430E+02 3.36798E+00 2.17612E+00 002 Elastic 4.20915E+00 4.75050E+00 1.12863 1.30649E+02 2.68961E+00 9.04232E-01 102 n,gamma 1.16642E+01 1.16670E+01 1.00023 5.78042E+00 2.81581E-03 1.84957E-03 Material(MAT)=2037:Ca-44 001 Total 4.24620E+00 4.67850E+00 1.10181 3.58669E+01 3.26208E+00 2.20125E+00 002 Elastic 3.35776E+00 3.78980E+00 1.12867 3.54473E+01 2.72656E+00 9.21344E-01 102 n,gamma 8.88447E-01 8.88690E-01 1.00027 4.19537E-01 1.18977E-03 1.36568E-03 Material(MAT)=2043:Ca-46 001 Total 7.00018E-01 7.00210E-01 1.00027 9.09167E+01 3.45203E+00 2.24524E+00 002 Elastic 0.00000E+00 0.00000E+00 0.00000 9.05982E+01 2.99143E+00 9.48626E-01 102 n,gamma 7.00018E-01 7.00210E-01 1.00027 3.18522E-01 5.34989E-04 1.39966E-03 Material(MAT)=2049:Ca-48 001 Total 4.84844E+00 5.33190E+00 1.09972 4.04005E+01 3.25557E+00 2.29332E+00 002 Elastic 3.75553E+00 4.23880E+00 1.12868 3.99157E+01 3.16175E+00 9.80209E-01 102 n,gamma 1.09291E+00 1.09310E+00 1.00020 4.84856E-01 2.04214E-04 8.38941E-04 Material(MAT)=2125:Sc-45 001 Total 4.98776E+01 5.28040E+01 1.05868 2.33484E+02 3.38297E+00 2.22608E+00 002 Elastic 2.27341E+01 2.56560E+01 1.12852 2.21605E+02 2.51285E+00 9.93542E-01 102 n,gamma 2.71415E+01 2.71470E+01 1.00019 1.18163E+01 4.82458E-03 1.50655E-03

Table 9. Simple integral neutron cross-section data from JEFF-3.1 calculated with the INTER computer code (cont.)

V Avg V G fact Avg res integ Avg V fiss V (E14) MT Reaction (2200 m/s) [1E-5 to 10] eV (Avg V)/V(E0) [5E-1 to 1E+5] eV [1E+3 to 2E+7] eV 1.4E+7 eV Material(MAT)=2225:Ti-46 001 Total 1.05718E+01 1.11390E+01 1.05361 1.38817E+02 3.31866E+00 2.31722E+00 002 Elastic 4.45899E+00 5.02650E+00 1.12726 1.36383E+02 2.63476E+00 1.01755E+00 102 n,gamma 6.11286E+00 6.11210E+00 0.99988 2.43356E+00 4.87896E-03 5.93338E-04 Material(MAT)=2228:Ti-47 001 Total 1.05709E+01 1.11380E+01 1.05360 1.38818E+02 3.31858E+00 2.31611E+00 002 Elastic 4.45800E+00 5.02540E+00 1.12727 1.36385E+02 2.36112E+00 9.81610E-01 102 n,gamma 6.11286E+00 6.11210E+00 0.99988 2.43356E+00 5.90897E-03 6.06647E-04 Material(MAT)=2231:Ti-48 001 Total 1.05613E+01 1.11230E+01 1.05317 1.38743E+02 3.31837E+00 2.32241E+00 002 Elastic 4.44847E+00 5.01080E+00 1.12641 1.36309E+02 2.71137E+00 1.02634E+00 102 n,gamma 6.11287E+00 6.11210E+00 0.99988 2.43356E+00 2.26615E-03 5.94579E-04 Material(MAT)=2234:Ti-49 001 Total 1.05690E+01 1.11350E+01 1.05358 1.38817E+02 3.31871E+00 2.32236E+00 002 Elastic 4.45616E+00 5.02320E+00 1.12725 1.36384E+02 2.77596E+00 9.90561E-01

91 102 n,gamma 6.11287E+00 6.11210E+00 0.99988 2.43356E+00 3.78821E-03 6.22830E-04 Material(MAT)=2237:Ti-50 001 Total 1.05682E+01 1.11340E+01 1.05357 1.38818E+02 3.31869E+00 2.32262E+00 002 Elastic 4.45529E+00 5.02220E+00 1.12724 1.36385E+02 2.89697E+00 9.60212E-01 102 n,gamma 6.11288E+00 6.11220E+00 0.99988 2.43356E+00 8.48028E-04 4.79753E-04 Material(MAT)=2300:V-Nat 001 Total 1.02507E+01 1.09190E+01 1.06518 2.65086E+02 3.85758E+00 2.35807E+00 002 Elastic 5.20656E+00 5.87650E+00 1.12868 2.62444E+02 3.19304E+00 1.10429E+00 102 n,gamma 5.04419E+00 5.04240E+00 0.99964 2.64298E+00 1.90136E-03 6.39995E-04 Material(MAT)=2425:Cr-50 001 Total 1.82987E+01 1.86080E+01 1.01689 2.60271E+02 3.66852E+00 2.41614E+00 002 Elastic 2.37167E+00 2.67700E+00 1.12873 2.52802E+02 2.97399E+00 1.09707E+00 102 n,gamma 1.59270E+01 1.59310E+01 1.00023 7.46971E+00 4.41724E-03 1.08000E-03 Material(MAT)=2431:Cr-52 001 Total 3.71429E+00 4.09460E+00 1.10239 3.69913E+01 3.39292E+00 2.44218E+00 002 Elastic 2.95420E+00 3.33440E+00 1.12868 3.65391E+01 2.94790E+00 1.08355E+00 102 n,gamma 7.60092E-01 7.60250E-01 1.00020 4.52165E-01 2.11168E-03 7.23925E-04 Material(MAT)=2434:Cr-53 001 Total 2.59891E+01 2.70120E+01 1.03937 3.42880E+02 3.78221E+00 2.41771E+00 002 Elastic 7.92349E+00 8.94340E+00 1.12872 3.34449E+02 2.99690E+00 1.01604E+00 102 n,gamma 1.80656E+01 1.80690E+01 1.00018 8.43178E+00 2.07872E-03 9.39396E-04

Table 9. Simple integral neutron cross-section data from JEFF-3.1 calculated with the INTER computer code (cont.)

V Avg V G fact Avg res integ Avg V fiss V (E14) MT Reaction (2200 m/s) [1E-5 to 10] eV (Avg V)/V(E0) [5E-1 to 1E+5] eV [1E+3 to 2E+7] eV 1.4E+7 eV Material(MAT)=2437:Cr-54 001 Total 2.92391E+00 3.25390E+00 1.11286 4.25940E+01 3.67328E+00 2.41770E+00 002 Elastic 2.56388E+00 2.89380E+00 1.12868 4.24031E+01 2.98286E+00 1.00132E+00 102 n,gamma 3.60034E-01 3.60120E-01 1.00024 1.90908E-01 1.59417E-03 5.45397E-04 Material(MAT)=2525:Mn-55 001 Total 1.56008E+01 1.58860E+01 1.01828 6.15513E+02 3.67946E+00 2.57127E+00 002 Elastic 2.18647E+00 2.46800E+00 1.12875 6.03749E+02 2.80002E+00 1.21794E+00 102 n,gamma 1.34144E+01 1.34180E+01 1.00027 1.17637E+01 2.94457E-03 6.63000E-04 Material(MAT)=2625:Fe-54 001 Total 4.45333E+00 4.73710E+00 1.06372 1.07745E+02 3.64968E+00 2.60816E+00 002 Elastic 2.20161E+00 2.48490E+00 1.12867 1.06563E+02 3.12611E+00 1.18706E+00 102 n,gamma 2.25172E+00 2.25220E+00 1.00022 1.18148E+00 4.88778E-03 9.35309E-04 Material(MAT)=2631:Fe-56 001 Total 1.47933E+01 1.63650E+01 1.10623 1.21968E+02 3.24797E+00 2.58777E+00 002 Elastic 1.22073E+01 1.37780E+01 1.12869 1.20639E+02 2.62649E+00 1.16278E+00

92 102 n,gamma 2.58593E+00 2.58650E+00 1.00021 1.32903E+00 2.73132E-03 7.67939E-04 Material(MAT)=2634:Fe-57 001 Total 2.66648E+00 2.69340E+00 1.01009 8.30276E+01 4.08150E+00 2.69920E+00 002 Elastic 2.03915E-01 2.30150E-01 1.12866 7.85114E+01 2.57456E+00 1.25805E+00 102 n,gamma 2.46257E+00 2.46320E+00 1.00027 1.41952E+00 1.76058E-03 8.92490E-04 Material(MAT)=2637:Fe-58 001 Total 8.85573E+00 9.82640E+00 1.10961 9.67920E+01 4.79543E+00 2.71964E+00 002 Elastic 7.54121E+00 8.51160E+00 1.12868 9.55300E+01 3.98352E+00 1.27552E+00 102 n,gamma 1.31452E+00 1.31480E+00 1.00024 1.26200E+00 2.10066E-03 9.75096E-04 Material(MAT)=2722:Co-58 001 Total 1.95377E+03 1.97300E+03 1.00982 1.18551E+04 3.91418E+00 2.71649E+00 002 Elastic 7.36061E+01 8.36040E+01 1.13583 5.54887E+03 2.62372E+00 1.24890E+00 102 n,gamma 1.64607E+02 1.67750E+02 1.01907 2.19955E+02 2.62689E-03 7.16790E-04 Material(MAT)=2723:Co-58M 001 Total 1.45046E+05 1.49590E+05 1.03132 2.35506E+05 3.91247E+00 2.71944E+00 002 Elastic 7.06280E+03 8.35190E+03 1.18251 5.90540E+04 2.71667E+00 1.24895E+00 102 n,gamma 1.00868E+05 1.03440E+05 1.02551 1.27801E+05 4.30483E-03 7.52090E-04 Material(MAT)=2725:Co-59 001 Total 4.32091E+01 4.40000E+01 1.01829 8.55772E+02 3.73341E+00 2.71319E+00 002 Elastic 6.03175E+00 6.80780E+00 1.12866 7.79911E+02 3.00024E+00 1.38970E+00 102 n,gamma 3.71774E+01 3.71920E+01 1.00038 7.58617E+01 5.00514E-03 8.99996E-04

Table 9. Simple integral neutron cross-section data from JEFF-3.1 calculated with the INTER computer code (cont.)

V Avg V G fact Avg res integ Avg V fiss V (E14) MT Reaction (2200 m/s) [1E-5 to 10] eV (Avg V)/V(E0) [5E-1 to 1E+5] eV [1E+3 to 2E+7] eV 1.4E+7 eV Material(MAT)=2825:Ni-58 001 Total 2.96383E+01 3.28590E+01 1.10865 2.67948E+02 3.65247E+00 2.71372E+00 002 Elastic 2.50176E+01 2.82370E+01 1.12868 2.65780E+02 3.16310E+00 1.29537E+00 102 n,gamma 4.62067E+00 4.62170E+00 1.00023 2.16790E+00 7.81934E-03 4.99542E-04 Material(MAT)=2828:Ni-59 001 Total 9.83465E+01 9.86710E+01 1.00330 5.17906E+02 1.19612E+01 5.82713E+00 002 Elastic 2.35427E+00 2.65750E+00 1.12879 3.67551E+02 1.18167E+01 4.60821E+00 102 n,gamma 8.07520E+01 8.07820E+01 1.00037 1.26737E+02 4.85057E-03 5.60002E-04 Material(MAT)=2831:Ni-60 001 Total 3.74894E+00 3.87620E+00 1.03394 8.02122E+01 3.46259E+00 2.72305E+00 002 Elastic 9.84800E-01 1.11150E+00 1.12870 7.88391E+01 2.93598E+00 1.27690E+00 102 n,gamma 2.76414E+00 2.76460E+00 1.00018 1.37324E+00 6.28746E-03 7.56393E-04 Material(MAT)=2834:Ni-61 001 Total 1.04566E+01 1.14810E+01 1.09800 1.06184E+02 3.58225E+00 2.76093E+00 002 Elastic 7.95807E+00 8.98220E+00 1.12869 1.04567E+02 2.34507E+00 1.30772E+00

93 102 n,gamma 2.49853E+00 2.49910E+00 1.00023 1.52555E+00 4.91953E-03 4.22401E-04 Material(MAT)=2837:Ni-62 001 Total 2.43887E+01 2.56720E+01 1.05264 4.62763E+02 3.64198E+00 2.75836E+00 002 Elastic 9.97227E+00 1.12560E+01 1.12875 4.56789E+02 3.06982E+00 1.28132E+00 102 n,gamma 1.44164E+01 1.44160E+01 0.99999 5.97539E+00 6.71034E-03 2.39399E-04 Material(MAT)=2843:Ni-64 001 Total 1.53452E+00 1.53700E+00 1.00161 8.56927E+01 4.07532E+00 2.75888E+00 002 Elastic 1.63720E-02 1.84800E-02 1.12874 8.49405E+01 3.57206E+00 1.27457E+00 102 n,gamma 1.51815E+00 1.51850E+00 1.00024 7.52291E-01 4.68378E-03 1.93599E-04 Material(MAT)=2925:Cu-63 001 Total 9.61284E+00 1.02750E+01 1.06891 1.01779E+02 3.66096E+00 2.91220E+00 002 Elastic 5.14324E+00 5.80490E+00 1.12865 9.68503E+01 2.87827E+00 1.47499E+00 102 n,gamma 4.46960E+00 4.47030E+00 1.00016 4.92903E+00 1.07519E-02 2.72000E-03 Material(MAT)=2931:Cu-65 001 Total 1.60409E+01 1.78290E+01 1.11148 1.52485E+02 3.74049E+00 2.95880E+00 002 Elastic 1.38921E+01 1.56800E+01 1.12869 1.50307E+02 3.04140E+00 1.40274E+00 102 n,gamma 2.14880E+00 2.14930E+00 1.00023 2.17899E+00 7.10443E-03 5.20002E-04 Material(MAT)=3000:Zn-Nat 001 Total 5.30484E+00 5.84900E+00 1.10257 9.97970E+01 3.80368E+00 2.95000E+00 002 Elastic 4.22602E+00 4.76990E+00 1.12870 9.73000E+01 3.15852E+00 1.69709E+00 102 n,gamma 1.07882E+00 1.07910E+00 1.00022 2.49716E+00 1.58928E-02 2.48699E-03

Table 9. Simple integral neutron cross-section data from JEFF-3.1 calculated with the INTER computer code (cont.)

V Avg V G fact Avg res integ Avg V fiss V (E14) MT Reaction (2200 m/s) [1E-5 to 10] eV (Avg V)/V(E0) [5E-1 to 1E+5] eV [1E+3 to 2E+7] eV 1.4E+7 eV Material(MAT)=3100:Ga-Nat 001 Total 9.84797E+00 1.07550E+01 1.09211 2.26047E+02 3.80480E+00 3.23069E+00 002 Elastic 7.04515E+00 7.95160E+00 1.12866 2.02347E+02 2.95062E+00 1.62693E+00 102 n,gamma 2.80283E+00 2.80350E+00 1.00025 2.36993E+01 1.53131E-02 7.61364E-04 Material(MAT)=3225:Ge-70 001 Total 1.55257E+01 1.71460E+01 1.10434 1.43586E+02 4.39187E+00 3.28016E+00 002 Elastic 1.25907E+01 1.42100E+01 1.12863 1.41220E+02 3.51303E+00 1.73323E+00 102 n,gamma 2.93501E+00 2.93550E+00 1.00016 2.36557E+00 2.17554E-02 6.86584E-04 Material(MAT)=3231:Ge-72 001 Total 9.97383E+00 1.11530E+01 1.11824 1.04549E+02 4.57000E+00 3.31673E+00 002 Elastic 9.16369E+00 1.03430E+01 1.12868 1.03801E+02 3.49410E+00 1.74081E+00 102 n,gamma 8.10138E-01 8.10300E-01 1.00020 7.47966E-01 1.74680E-02 8.69823E-04 Material(MAT)=3234:Ge-73 001 Total 1.92357E+01 1.98610E+01 1.03250 4.61939E+02 4.11702E+00 3.31387E+00 002 Elastic 4.83311E+00 5.45440E+00 1.12856 3.99620E+02 2.78815E+00 1.71276E+00

95 94 102 n,gamma 1.44026E+01 1.44060E+01 1.00027 6.22392E+01 1.40008E-02 7.05540E-04 Material(MAT)=3237:Ge-74 001 Total 7.26637E+00 8.14750E+00 1.12126 9.73499E+01 4.44431E+00 3.34262E+00 002 Elastic 6.84628E+00 7.72740E+00 1.12869 9.69320E+01 3.30254E+00 1.71645E+00 102 n,gamma 4.20090E-01 4.20160E-01 1.00017 4.18082E-01 6.08952E-03 8.12471E-04 Material(MAT)=3243:Ge-76 001 Total 7.91114E+00 8.91010E+00 1.12628 1.00008E+02 4.56345E+00 3.39814E+00 002 Elastic 7.76112E+00 8.76010E+00 1.12871 9.87022E+01 3.40410E+00 1.77726E+00 102 n,gamma 1.50021E-01 1.50060E-01 1.00025 1.30591E+00 2.63586E-03 9.78019E-04 Material(MAT)=3325:As-75 001 Total 6.19044E+00 6.43540E+00 1.03958 1.83644E+02 4.07585E+00 3.73274E+00 002 Elastic 1.88929E+00 2.13230E+00 1.12863 1.23742E+02 2.83635E+00 2.45463E+00 102 n,gamma 4.30115E+00 4.30310E+00 1.00046 5.99015E+01 4.59555E-02 1.53781E-03 Material(MAT)=3425:Se-74 001 Total 5.16882E+01 5.17410E+01 1.00101 1.32547E+03 4.67325E+00 6.23995E+00 002 Elastic 1.01594E-02 1.12420E-02 1.10657 7.45767E+02 3.65815E+00 4.47900E+00 102 n,gamma 5.16780E+01 5.17290E+01 1.00099 5.79705E+02 2.94543E-02 1.64001E-03 Material(MAT)=3431:Se-76 001 Total 8.69202E+01 8.71900E+01 1.00310 1.33606E+02 4.13803E+00 3.76858E+00 002 Elastic 1.91713E+00 2.16390E+00 1.12870 8.92164E+01 3.37087E+00 2.50056E+00 102 n,gamma 8.50031E+01 8.50260E+01 1.00027 4.43899E+01 3.64057E-02 1.62327E-03

Table 9. Simple integral neutron cross-section data from JEFF-3.1 calculated with the INTER computer code (cont.)

V Avg V G fact Avg res integ Avg V fiss V (E14) MT Reaction (2200 m/s) [1E-5 to 10] eV (Avg V)/V(E0) [5E-1 to 1E+5] eV [1E+3 to 2E+7] eV 1.4E+7 eV Material(MAT)=3434:Se-77 001 Total 4.44167E+01 4.47390E+01 1.00726 1.44659E+02 4.19914E+00 3.79736E+00 002 Elastic 2.41508E+00 2.72590E+00 1.12869 1.08710E+02 2.92080E+00 2.53773E+00 102 n,gamma 4.20016E+01 4.20130E+01 1.00028 3.59489E+01 4.11323E-02 2.06045E-03 Material(MAT)=3437:Se-78 001 Total 2.95097E+00 3.27940E+00 1.11129 6.84704E+01 4.26703E+00 3.83190E+00 002 Elastic 2.55092E+00 2.87920E+00 1.12871 6.39552E+01 3.47895E+00 2.60438E+00 102 n,gamma 4.00054E-01 4.00160E-01 1.00027 4.51521E+00 1.50675E-02 1.52333E-03 Material(MAT)=3440:Se-79 001 Total 5.63420E+01 5.71860E+01 1.01498 1.82151E+02 4.98592E+00 3.84400E+00 002 Elastic 6.34025E+00 7.17110E+00 1.13105 1.21472E+02 3.62728E+00 2.51693E+00 102 n,gamma 5.00018E+01 5.00150E+01 1.00027 6.06797E+01 3.09790E-02 1.00054E-03 Material(MAT)=3443:Se-80 001 Total 1.68050E+00 1.81840E+00 1.08207 1.13629E+02 4.41486E+00 3.89107E+00 002 Elastic 1.07037E+00 1.20810E+00 1.12869 1.12628E+02 3.65168E+00 2.69381E+00

95 102 n,gamma 6.10128E-01 6.10300E-01 1.00028 1.00143E+00 1.51340E-02 1.45032E-03 Material(MAT)=3449:Se-82 001 Total 2.96072E+00 3.33600E+00 1.12677 4.91176E+01 4.54578E+00 3.95270E+00 002 Elastic 2.91571E+00 3.29100E+00 1.12872 4.90507E+01 3.82441E+00 2.79755E+00 102 n,gamma 4.50079E-02 4.50190E-02 1.00026 6.70126E-02 4.61867E-03 1.32207E-03 Material(MAT)=3525:Br-79 001 Total 1.35428E+01 1.38640E+01 1.02369 2.51831E+02 4.33777E+00 3.86129E+00 002 Elastic 2.44021E+00 2.75390E+00 1.12857 1.15850E+02 3.00756E+00 2.62747E+00 102 n,gamma 1.11026E+01 1.11100E+01 1.00063 1.35981E+02 8.73364E-02 1.72088E-03 Material(MAT)=3531:Br-81 001 Total 6.94178E+00 7.47900E+00 1.07738 1.88174E+02 4.47974E+00 3.91687E+00 002 Elastic 4.16600E+00 4.70200E+00 1.12866 1.33904E+02 3.44433E+00 2.68545E+00 102 n,gamma 2.77578E+00 2.77700E+00 1.00043 5.42704E+01 5.89765E-02 1.63915E-03 Material(MAT)=3625:Kr-78 001 Total 1.36121E+01 1.47400E+01 1.08285 1.68139E+02 5.07189E+00 3.78486E+00 002 Elastic 8.76549E+00 9.89300E+00 1.12863 1.44785E+02 3.51542E+00 1.92660E+00 102 n,gamma 4.84661E+00 4.84690E+00 1.00005 2.33538E+01 7.15888E-02 1.76983E-04 Material(MAT)=3631:Kr-80 001 Total 2.06734E+01 2.18240E+01 1.05565 3.13919E+02 5.23669E+00 3.65418E+00 002 Elastic 8.93530E+00 1.00840E+01 1.12851 2.45525E+02 3.81265E+00 1.93756E+00 102 n,gamma 1.17381E+01 1.17400E+01 1.00019 6.83936E+01 4.28568E-02 1.04505E-03

Table 9. Simple integral neutron cross-section data from JEFF-3.1 calculated with the INTER computer code (cont.)

V Avg V G fact Avg res integ Avg V fiss V (E14) MT Reaction (2200 m/s) [1E-5 to 10] eV (Avg V)/V(E0) [5E-1 to 1E+5] eV [1E+3 to 2E+7] eV 1.4E+7 eV Material(MAT)=3637:Kr-82 001 Total 3.99432E+01 4.11960E+01 1.03137 3.85861E+02 5.01738E+00 3.78332E+00 002 Elastic 9.77004E+00 1.10230E+01 1.12826 2.02677E+02 3.89878E+00 2.04440E+00 102 n,gamma 3.01731E+01 3.01730E+01 1.00000 1.83184E+02 1.85246E-02 3.36355E-05 Material(MAT)=3640:Kr-83 001 Total 2.18072E+02 2.18180E+02 1.00052 3.35645E+02 5.18901E+00 3.50009E+00 002 Elastic 1.04112E+01 1.17120E+01 1.12497 1.43011E+02 3.69292E+00 2.07549E+00 102 n,gamma 2.07661E+02 2.06470E+02 0.99428 1.91687E+02 3.95263E-02 4.05716E-05 Material(MAT)=3643:Kr-84 001 Total 6.83268E+00 7.70160E+00 1.12717 1.06202E+02 5.24608E+00 3.84927E+00 002 Elastic 6.74979E+00 7.61870E+00 1.12873 1.02752E+02 4.16871E+00 2.10486E+00 102 n,gamma 8.28891E-02 8.29130E-02 1.00029 3.45012E+00 7.31558E-03 2.16013E-05 Material(MAT)=3646:Kr-85 001 Total 5.22436E+00 5.69530E+00 1.09014 7.94170E+01 4.76214E+00 4.03083E+00 002 Elastic 3.56430E+00 4.03480E+00 1.13200 7.77738E+01 4.35050E+00 2.86496E+00

96 102 n,gamma 1.66006E+00 1.66050E+00 1.00027 1.64321E+00 5.58426E-03 2.00911E-03 Material(MAT)=3649:Kr-86 001 Total 6.10820E+00 6.88650E+00 1.12741 1.01108E+02 5.54416E+00 3.91874E+00 002 Elastic 6.04673E+00 6.82500E+00 1.12871 1.00972E+02 4.68397E+00 2.16218E+00 102 n,gamma 6.14645E-02 6.14780E-02 1.00023 1.36106E-01 1.29145E-03 5.31394E-06 Material(MAT)=3725:Rb-85 001 Total 6.81099E+00 7.62630E+00 1.11970 1.17884E+02 4.77926E+00 4.03089E+00 002 Elastic 6.33454E+00 7.14980E+00 1.12869 1.11665E+02 3.85162E+00 2.83936E+00 102 n,gamma 4.76452E-01 4.76550E-01 1.00020 6.21950E+00 2.81558E-02 7.37329E-04 Material(MAT)=3728:Rb-86 001 Total 8.49793E+00 8.97430E+00 1.05606 1.01906E+02 4.83008E+00 4.05370E+00 002 Elastic 3.59784E+00 4.07330E+00 1.13216 7.79952E+01 3.74978E+00 2.86052E+00 102 n,gamma 4.90009E+00 4.90100E+00 1.00018 2.39106E+01 1.08713E-02 2.16141E-03 Material(MAT)=3731:Rb-87 001 Total 2.47264E+00 2.77550E+00 1.12247 7.25427E+01 4.91192E+00 4.07523E+00 002 Elastic 2.35264E+00 2.65540E+00 1.12870 7.04712E+01 4.00207E+00 2.89831E+00 102 n,gamma 1.20001E-01 1.20030E-01 1.00028 2.07165E+00 2.86419E-03 7.64339E-05 Material(MAT)=3825:Sr-84 001 Total 4.74353E+00 5.24980E+00 1.10672 1.24233E+02 5.06453E+00 6.26200E+00 002 Elastic 3.93111E+00 4.43700E+00 1.12870 1.14041E+02 4.18298E+00 4.66472E+00 102 n,gamma 8.12427E-01 8.12730E-01 1.00038 1.01928E+01 6.81472E-02 2.02001E-03

Table 9. Simple integral neutron cross-section data from JEFF-3.1 calculated with the INTER computer code (cont.)

V Avg V G fact Avg res integ Avg V fiss V (E14) MT Reaction (2200 m/s) [1E-5 to 10] eV (Avg V)/V(E0) [5E-1 to 1E+5] eV [1E+3 to 2E+7] eV 1.4E+7 eV Material(MAT)=3831:Sr-86 001 Total 5.07848E+00 5.36730E+00 1.05687 1.17934E+02 4.83674E+00 4.05035E+00 002 Elastic 2.23810E+00 2.52610E+00 1.12868 1.12836E+02 4.32875E+00 2.83831E+00 102 n,gamma 2.84038E+00 2.84120E+00 1.00028 5.09801E+00 1.79912E-02 1.95121E-03 Material(MAT)=3834:Sr-87 001 Total 1.92555E+01 1.97680E+01 1.02661 1.86436E+02 4.90578E+00 4.07492E+00 002 Elastic 3.24763E+00 3.66450E+00 1.12836 6.80123E+01 4.37010E+00 2.89534E+00 102 n,gamma 1.60079E+01 1.61030E+01 1.00597 1.18424E+02 1.95297E-02 2.38936E-03 Material(MAT)=3837:Sr-88 001 Total 3.64252E+00 4.11060E+00 1.12851 4.40012E+01 4.83970E+00 4.09786E+00 002 Elastic 3.63672E+00 4.10480E+00 1.12872 4.39955E+01 4.52077E+00 2.94661E+00 102 n,gamma 5.80024E-03 5.80180E-03 1.00027 5.76642E-03 1.27540E-03 1.59881E-03 Material(MAT)=3840:Sr-89 001 Total 4.08464E+00 4.55650E+00 1.11551 7.05337E+01 5.03845E+00 4.11974E+00 002 Elastic 3.66463E+00 4.13630E+00 1.12872 7.00423E+01 4.74596E+00 2.96262E+00

97 102 n,gamma 4.20015E-01 4.20130E-01 1.00027 4.91440E-01 8.19075E-03 2.15555E-03 Material(MAT)=3843:Sr-90 001 Total 4.59500E+00 5.08400E+00 1.10643 5.82125E+01 5.08317E+00 4.14078E+00 002 Elastic 3.69497E+00 4.18380E+00 1.13229 5.77320E+01 4.40653E+00 2.96998E+00 102 n,gamma 9.00032E-01 9.00280E-01 1.00027 4.80433E-01 5.94259E-03 1.50930E-03 Material(MAT)=3925:Y-89 001 Total 9.01661E+00 9.86960E+00 1.09460 9.48938E+01 5.63556E+00 3.85693E+00 002 Elastic 7.73094E+00 8.58360E+00 1.11029 9.39966E+01 4.91895E+00 2.24392E+00 102 n,gamma 1.28567E+00 1.28600E+00 1.00023 8.97240E-01 4.76080E-03 6.97734E-04 Material(MAT)=3928:Y-90 001 Total 7.20535E+00 7.69970E+00 1.06861 8.82582E+01 5.08578E+00 4.13991E+00 002 Elastic 3.70523E+00 4.19870E+00 1.13317 8.36077E+01 3.96739E+00 2.94810E+00 102 n,gamma 3.50013E+00 3.50110E+00 1.00027 4.65043E+00 1.76875E-02 1.92237E-03 Material(MAT)=3931:Y-91 001 Total 5.12270E+00 5.61200E+00 1.09552 7.74608E+01 5.14846E+00 4.16195E+00 002 Elastic 3.72265E+00 4.21160E+00 1.13134 7.58331E+01 4.14353E+00 2.85036E+00 102 n,gamma 1.40005E+00 1.40040E+00 1.00027 1.62767E+00 6.77333E-03 1.65636E-03 Material(MAT)=4025:Zr-90 001 Total 5.40629E+00 6.10080E+00 1.12846 7.73259E+01 5.73633E+00 3.83550E+00 002 Elastic 5.39517E+00 6.08970E+00 1.12872 7.71854E+01 5.25592E+00 2.17244E+00 102 n,gamma 1.11263E-02 1.11290E-02 1.00021 1.40601E-01 6.28599E-03 1.03383E-03

Table 9. Simple integral neutron cross-section data from JEFF-3.1 calculated with the INTER computer code (cont.)

V Avg V G fact Avg res integ Avg V fiss V (E14) MT Reaction (2200 m/s) [1E-5 to 10] eV (Avg V)/V(E0) [5E-1 to 1E+5] eV [1E+3 to 2E+7] eV 1.4E+7 eV Material(MAT)=4028:Zr-91 001 Total 1.18921E+01 1.32620E+01 1.11519 1.50977E+02 5.74780E+00 3.84553E+00 002 Elastic 1.06444E+01 1.20140E+01 1.12867 1.44089E+02 5.06028E+00 2.13985E+00 102 n,gamma 1.24766E+00 1.24790E+00 1.00017 6.88861E+00 9.20165E-03 1.00200E-03 Material(MAT)=4031:Zr-92 001 Total 7.37830E+00 8.29840E+00 1.12471 1.13649E+02 5.76145E+00 3.84583E+00 002 Elastic 7.14908E+00 8.06920E+00 1.12870 1.13029E+02 4.76913E+00 2.10824E+00 102 n,gamma 2.29214E-01 2.29260E-01 1.00021 6.20763E-01 1.52913E-02 1.02401E-03 Material(MAT)=4034:Zr-93 001 Total 7.68851E+00 8.44970E+00 1.09900 1.81202E+02 5.62826E+00 4.11875E+00 002 Elastic 5.90789E+00 6.66820E+00 1.12869 1.48166E+02 4.55459E+00 2.38077E+00 102 n,gamma 1.78062E+00 1.78150E+00 1.00049 3.30361E+01 1.46075E-02 1.00828E-03 Material(MAT)=4037:Zr-94 001 Total 6.23477E+00 7.03090E+00 1.12769 8.94050E+01 5.75911E+00 3.84357E+00 002 Elastic 6.18495E+00 6.98100E+00 1.12871 8.91451E+01 4.76208E+00 2.13748E+00

98 102 n,gamma 4.98188E-02 4.98280E-02 1.00019 2.60082E-01 9.77974E-03 1.00184E-03 Material(MAT)=4040:Zr-95 001 Total 5.64375E+00 6.34130E+00 1.12359 1.32316E+02 5.72506E+00 4.19620E+00 002 Elastic 5.41876E+00 6.11620E+00 1.12871 1.26561E+02 4.94779E+00 2.42718E+00 102 n,gamma 2.24993E-01 2.25080E-01 1.00038 5.75529E+00 1.35755E-02 2.00725E-03 Material(MAT)=4043:Zr-96 001 Total 6.18599E+00 6.97930E+00 1.12825 1.04150E+02 5.79846E+00 3.84628E+00 002 Elastic 6.16319E+00 6.95650E+00 1.12872 9.83162E+01 5.15893E+00 2.11567E+00 102 n,gamma 2.28057E-02 2.28180E-02 1.00054 5.83434E+00 5.74046E-03 1.00053E-03 Material(MAT)=4125:Nb-93 001 Total 7.51996E+00 8.26500E+00 1.09907 9.53486E+01 5.79033E+00 3.96935E+00 002 Elastic 6.36466E+00 7.10760E+00 1.11673 8.56012E+01 4.53168E+00 2.10918E+00 102 n,gamma 1.15530E+00 1.15730E+00 1.00176 9.74724E+00 2.78472E-02 5.60002E-04 Material(MAT)=4128:Nb-94 001 Total 2.22549E+01 2.30720E+01 1.03671 2.36241E+02 5.73091E+00 4.32530E+00 002 Elastic 6.48708E+00 7.31950E+00 1.12832 1.10772E+02 4.14325E+00 2.81424E+00 102 n,gamma 1.57678E+01 1.57520E+01 0.99903 1.25363E+02 1.79928E-02 1.00541E-03 Material(MAT)=4131:Nb-95 001 Total 1.27305E+01 1.34810E+01 1.05894 1.39891E+02 5.71855E+00 4.32526E+00 002 Elastic 5.73028E+00 6.47870E+00 1.13061 9.83884E+01 4.84114E+00 2.81694E+00 102 n,gamma 7.00025E+00 7.00220E+00 1.00028 4.15029E+01 6.84151E-02 1.00995E-03

Table 9. Simple integral neutron cross-section data from JEFF-3.1 calculated with the INTER computer code (cont.)

V Avg V G fact Avg res integ Avg V fiss V (E14) MT Reaction (2200 m/s) [1E-5 to 10] eV (Avg V)/V(E0) [5E-1 to 1E+5] eV [1E+3 to 2E+7] eV 1.4E+7 eV Material(MAT)=4225:Mo-92 001 Total 5.59637E+00 6.31400E+00 1.12823 7.98423E+01 5.89305E+00 4.06088E+00 002 Elastic 5.57562E+00 6.29330E+00 1.12871 7.89887E+01 5.24197E+00 2.27589E+00 102 n,gamma 2.07557E-02 2.07610E-02 1.00027 8.53757E-01 2.79986E-02 1.04482E-03 Material(MAT)=4231:Mo-94 001 Total 6.04323E+00 6.81940E+00 1.12843 8.84864E+01 5.88853E+00 4.06141E+00 002 Elastic 6.03011E+00 6.80620E+00 1.12871 8.72506E+01 4.86944E+00 2.27669E+00 102 n,gamma 1.31124E-02 1.31150E-02 1.00021 1.23582E+00 3.56549E-02 1.00786E-03 Material(MAT)=4234:Mo-95 001 Total 1.99816E+01 2.08020E+01 1.04103 3.12353E+02 5.89045E+00 4.36250E+00 002 Elastic 6.41386E+00 7.23670E+00 1.12829 2.01753E+02 4.51781E+00 2.84294E+00 102 n,gamma 1.35678E+01 1.35650E+01 0.99978 1.10600E+02 4.80498E-02 1.40815E-03 Material(MAT)=4237:Mo-96 001 Total 5.34722E+00 5.95890E+00 1.11440 1.62142E+02 5.88934E+00 4.06777E+00 002 Elastic 4.75171E+00 5.36320E+00 1.12868 1.44744E+02 4.74392E+00 2.28840E+00

99 102 n,gamma 5.95512E-01 5.95750E-01 1.00040 1.73980E+01 2.77164E-02 1.00993E-03 Material(MAT)=4240:Mo-97 001 Total 7.98732E+00 8.74280E+00 1.09458 1.05408E+02 5.88976E+00 4.06094E+00 002 Elastic 5.88701E+00 6.64440E+00 1.12865 8.85513E+01 4.34875E+00 2.27434E+00 102 n,gamma 2.10031E+00 2.09840E+00 0.99909 1.68566E+01 4.27044E-02 1.00104E-03 Material(MAT)=4243:Mo-98 001 Total 5.80088E+00 6.53090E+00 1.12584 1.01793E+02 5.88452E+00 4.06868E+00 002 Elastic 5.67088E+00 6.40080E+00 1.12871 9.53742E+01 4.72501E+00 2.29452E+00 102 n,gamma 1.29997E-01 1.30090E-01 1.00075 6.41857E+00 2.68333E-02 1.01157E-03 Material(MAT)=4246:Mo-99 001 Total 1.37293E+01 1.44790E+01 1.05462 1.41503E+02 5.72682E+00 4.32713E+00 002 Elastic 5.72905E+00 6.47690E+00 1.13053 1.00108E+02 4.01619E+00 2.81875E+00 102 n,gamma 8.00027E+00 8.00240E+00 1.00026 4.13948E+01 3.54163E-02 1.00065E-03 Material(MAT)=4249:Mo-100 001 Total 5.52608E+00 6.21170E+00 1.12407 1.00774E+02 5.88707E+00 4.07231E+00 002 Elastic 5.32703E+00 6.01260E+00 1.12870 9.69834E+01 4.56263E+00 2.23357E+00 102 n,gamma 1.99057E-01 1.99090E-01 1.00019 3.79090E+00 2.12326E-02 1.00826E-03 Material(MAT)=4331:Tc-99 001 Total 2.81477E+01 2.89020E+01 1.02679 4.16632E+02 5.92070E+00 4.20027E+00 002 Elastic 5.32558E+00 6.00700E+00 1.12796 9.38016E+01 4.16177E+00 2.24836E+00 102 n,gamma 2.28221E+01 2.28950E+01 1.00319 3.22830E+02 8.86188E-02 1.33780E-03

Table 9. Simple integral neutron cross-section data from JEFF-3.1 calculated with the INTER computer code (cont.)

V Avg V G fact Avg res integ Avg V fiss V (E14) MT Reaction (2200 m/s) [1E-5 to 10] eV (Avg V)/V(E0) [5E-1 to 1E+5] eV [1E+3 to 2E+7] eV 1.4E+7 eV Material(MAT)=4425:Ru-96 001 Total 6.32602E+00 7.20630E+00 1.13915 1.23624E+02 5.92249E+00 6.87140E+00 002 Elastic 6.07745E+00 6.95730E+00 1.14478 1.11399E+02 4.78217E+00 4.89204E+00 102 n,gamma 2.48575E-01 2.48920E-01 1.00137 1.22245E+01 1.42408E-01 1.40000E-03 Material(MAT)=4431:Ru-98 001 Total 1.41701E+01 1.57600E+01 1.11221 1.23164E+02 5.95393E+00 6.58240E+00 002 Elastic 6.21459E+00 7.78780E+00 1.25316 1.11368E+02 4.87250E+00 4.90643E+00 102 n,gamma 7.95549E+00 7.97220E+00 1.00210 1.17966E+01 2.61225E-02 2.38001E-03 Material(MAT)=4434:Ru-99 001 Total 1.20752E+01 1.27300E+01 1.05427 2.50073E+02 5.43102E+00 4.32872E+00 002 Elastic 4.97303E+00 5.61220E+00 1.12853 8.76754E+01 4.00633E+00 3.03825E+00 102 n,gamma 7.10217E+00 7.11830E+00 1.00227 1.62395E+02 6.18145E-02 2.51477E-03 Material(MAT)=4437:Ru-100 001 Total 9.73363E+00 1.02410E+01 1.05218 7.92562E+01 5.45025E+00 4.34760E+00 002 Elastic 3.93346E+00 4.43970E+00 1.12871 7.13407E+01 4.55168E+00 3.09271E+00

100 102 n,gamma 5.80017E+00 5.80180E+00 1.00027 7.91550E+00 5.72047E-02 2.52898E-03 Material(MAT)=4440:Ru-101 001 Total 8.49337E+00 9.15120E+00 1.07745 2.00157E+02 5.93843E+00 4.19486E+00 002 Elastic 5.07868E+00 5.73190E+00 1.12862 8.95327E+01 4.06711E+00 2.17407E+00 102 n,gamma 3.41468E+00 3.41930E+00 1.00136 1.10625E+02 6.63402E-02 1.00912E-03 Material(MAT)=4443:Ru-102 001 Total 5.26876E+00 5.78000E+00 1.09704 6.60695E+01 5.48814E+00 4.38905E+00 002 Elastic 3.96878E+00 4.47960E+00 1.12872 6.31203E+01 4.51361E+00 3.25241E+00 102 n,gamma 1.29997E+00 1.30040E+00 1.00034 2.94918E+00 5.94090E-02 1.58180E-03 Material(MAT)=4446:Ru-103 001 Total 7.18727E+01 7.25730E+01 1.00975 6.89172E+02 5.83095E+00 4.34622E+00 002 Elastic 5.04629E+00 5.68460E+00 1.12649 9.42649E+01 3.66742E+00 2.45141E+00 102 n,gamma 6.68264E+01 6.68890E+01 1.00093 5.93666E+02 1.44340E-01 2.12022E-03 Material(MAT)=4449:Ru-104 001 Total 8.42116E+00 9.46690E+00 1.12418 1.06903E+02 5.51583E+00 4.43238E+00 002 Elastic 8.12393E+00 9.16960E+00 1.12871 1.01259E+02 4.63153E+00 3.17236E+00 102 n,gamma 2.97232E-01 2.97300E-01 1.00024 5.64401E+00 3.19619E-02 9.09251E-04 Material(MAT)=4452:Ru-105 001 Total 4.28853E+00 4.81490E+00 1.12274 7.16876E+01 5.50783E+00 4.45482E+00 002 Elastic 4.08852E+00 4.61480E+00 1.12873 6.46193E+01 3.89924E+00 3.08549E+00 102 n,gamma 2.00007E-01 2.00060E-01 1.00027 7.06702E+00 4.07980E-02 2.14829E-03

Table 9. Simple integral neutron cross-section data from JEFF-3.1 calculated with the INTER computer code (cont.)

V Avg V G fact Avg res integ Avg V fiss V (E14) MT Reaction (2200 m/s) [1E-5 to 10] eV (Avg V)/V(E0) [5E-1 to 1E+5] eV [1E+3 to 2E+7] eV 1.4E+7 eV Material(MAT)=4455:Ru-106 001 Total 4.26047E+00 4.79010E+00 1.12432 6.45412E+01 5.52688E+00 4.47375E+00 002 Elastic 4.11446E+00 4.64410E+00 1.12873 6.25115E+01 4.45662E+00 3.11283E+00 102 n,gamma 1.46005E-01 1.46040E-01 1.00027 2.02978E+00 1.12556E-02 1.62707E-03 Material(MAT)=4525:Rh-103 001 Total 1.46614E+02 1.50360E+02 1.02553 1.05330E+03 5.76648E+00 4.30114E+00 002 Elastic 3.88416E+00 4.36370E+00 1.12347 8.49141E+01 3.98869E+00 2.26306E+00 102 n,gamma 1.42730E+02 1.45990E+02 1.02286 9.68383E+02 8.48846E-02 2.20335E-03 Material(MAT)=4531:Rh-105 001 Total 1.53371E+04 1.24920E+04 0.81447 1.57628E+04 5.49583E+00 4.45468E+00 002 Elastic 4.08852E+00 4.61480E+00 1.12873 7.00159E+01 4.45188E+00 3.13806E+00 102 n,gamma 1.53330E+04 1.24870E+04 0.81439 1.56927E+04 1.19453E-01 1.89058E-03 Material(MAT)=4625:Pd-102 001 Total 9.07312E+00 9.80010E+00 1.08012 1.18784E+02 6.08934E+00 6.91461E+00 002 Elastic 5.70836E+00 6.44230E+00 1.12858 1.05829E+02 5.02142E+00 5.18782E+00

101 102 n,gamma 3.36476E+00 3.35770E+00 0.99791 1.29559E+01 4.72718E-02 3.19001E-03 Material(MAT)=4631:Pd-104 001 Total 4.77052E+00 5.34410E+00 1.12023 1.26663E+02 5.81452E+00 4.31016E+00 002 Elastic 4.45540E+00 5.02880E+00 1.12871 1.12851E+02 4.64168E+00 2.21209E+00 102 n,gamma 3.15122E-01 3.15230E-01 1.00035 1.38117E+01 4.61101E-02 1.09600E-03 Material(MAT)=4634:Pd-105 001 Total 2.70706E+01 2.76440E+01 1.02117 1.75282E+02 5.81456E+00 4.41289E+00 002 Elastic 5.24512E+00 5.91670E+00 1.12803 8.21835E+01 3.89326E+00 2.49280E+00 102 n,gamma 2.18255E+01 2.17270E+01 0.99549 9.30981E+01 1.35719E-01 2.26276E-03 Material(MAT)=4637:Pd-106 001 Total 4.88900E+00 5.47950E+00 1.12078 1.06946E+02 5.82203E+00 4.32932E+00 002 Elastic 4.58859E+00 5.17910E+00 1.12869 1.01033E+02 4.59024E+00 2.24127E+00 102 n,gamma 3.00411E-01 3.00390E-01 0.99991 5.91346E+00 5.45932E-02 9.73198E-04 Material(MAT)=4640:Pd-107 001 Total 6.25655E+00 6.83140E+00 1.09187 1.94711E+02 5.83960E+00 4.34360E+00 002 Elastic 4.45130E+00 5.02390E+00 1.12864 8.98341E+01 4.03367E+00 2.25590E+00 102 n,gamma 1.80525E+00 1.80750E+00 1.00122 1.04877E+02 1.20911E-01 9.65398E-04 Material(MAT)=4643:Pd-108 001 Total 1.07868E+01 1.12300E+01 1.04109 6.30731E+02 5.83260E+00 4.35178E+00 002 Elastic 3.43065E+00 3.86940E+00 1.12790 4.57693E+02 4.52463E+00 2.27031E+00 102 n,gamma 7.35616E+00 7.36060E+00 1.00060 1.73037E+02 4.29796E-02 8.78498E-04

Table 9. Simple integral neutron cross-section data from JEFF-3.1 calculated with the INTER computer code (cont.)

V Avg V G fact Avg res integ Avg V fiss V (E14) MT Reaction (2200 m/s) [1E-5 to 10] eV (Avg V)/V(E0) [5E-1 to 1E+5] eV [1E+3 to 2E+7] eV 1.4E+7 eV Material(MAT)=4649:Pd-110 001 Total 5.29551E+00 5.94800E+00 1.12321 7.13228E+01 5.83264E+00 4.38634E+00 002 Elastic 5.06850E+00 5.72090E+00 1.12871 6.90350E+01 4.47027E+00 2.29977E+00 102 n,gamma 2.27009E-01 2.27070E-01 1.00027 2.28776E+00 2.17900E-02 8.99294E-04 Material(MAT)=4725:Ag-107 001 Total 4.51804E+01 4.60790E+01 1.01989 2.10128E+02 6.03096E+00 4.25875E+00 002 Elastic 7.56918E+00 8.53490E+00 1.12759 1.03198E+02 4.36589E+00 2.69870E+00 102 n,gamma 3.76113E+01 3.75440E+01 0.99821 1.06930E+02 1.01761E-01 8.78603E-07 Material(MAT)=4731:Ag-109 001 Total 9.30482E+01 9.38040E+01 1.00812 1.69090E+03 5.68305E+00 4.33369E+00 002 Elastic 2.30898E+00 2.59690E+00 1.12468 2.16415E+02 3.90897E+00 2.54116E+00 102 n,gamma 9.07392E+01 9.12070E+01 1.00515 1.47448E+03 1.33509E-01 1.01500E-03 Material(MAT)=4735:Ag-110M 001 Total 8.84909E+01 8.83790E+01 0.99874 1.56209E+02 5.59999E+00 4.46900E+00 002 Elastic 6.49750E+00 7.31770E+00 1.12623 6.58520E+01 3.83631E+00 2.64754E+00

104 102 102 n,gamma 8.19934E+01 8.10620E+01 0.98864 9.03415E+01 5.78990E-01 1.04266E-03 Material(MAT)=4737:Ag-111 001 Total 7.24813E+00 7.79810E+00 1.07588 1.69565E+02 5.53005E+00 4.57077E+00 002 Elastic 4.24802E+00 4.79720E+00 1.12927 6.57698E+01 4.21776E+00 3.32494E+00 102 n,gamma 3.00011E+00 3.00090E+00 1.00027 1.03795E+02 5.22261E-02 1.93898E-03 Material(MAT)=4825:Cd-106 001 Total 4.72711E+00 5.19150E+00 1.09824 9.63023E+01 5.80243E+00 4.33052E+00 002 Elastic 3.60878E+00 4.07320E+00 1.12868 8.24914E+01 4.37956E+00 2.48666E+00 102 n,gamma 1.11833E+00 1.11830E+00 1.00002 1.38109E+01 1.67909E-01 6.00076E-04 Material(MAT)=4831:Cd-108 001 Total 5.16197E+00 5.68460E+00 1.10125 1.31720E+02 5.84760E+00 4.37039E+00 002 Elastic 4.06086E+00 4.58330E+00 1.12866 1.15074E+02 4.41561E+00 2.52544E+00 102 n,gamma 1.10111E+00 1.10130E+00 1.00019 1.66459E+01 1.25051E-01 6.00020E-04 Material(MAT)=4837:Cd-110 001 Total 2.93270E+01 3.16800E+01 1.08023 2.23103E+02 5.86601E+00 4.40867E+00 002 Elastic 1.83146E+01 2.06670E+01 1.12844 1.81835E+02 4.46586E+00 2.56489E+00 102 n,gamma 1.10124E+01 1.10130E+01 1.00004 4.12676E+01 7.39418E-02 6.00025E-04 Material(MAT)=4840:Cd-111 001 Total 2.86759E+01 2.91310E+01 1.01586 1.20763E+02 5.89117E+00 4.44297E+00 002 Elastic 4.64849E+00 5.23930E+00 1.12709 7.11984E+01 3.98385E+00 2.62094E+00 102 n,gamma 2.40274E+01 2.38910E+01 0.99434 4.95647E+01 1.02943E-01 6.00252E-05

Table 9. Simple integral neutron cross-section data from JEFF-3.1 calculated with the INTER computer code (cont.)

V Avg V G fact Avg res integ Avg V fiss V (E14) MT Reaction (2200 m/s) [1E-5 to 10] eV (Avg V)/V(E0) [5E-1 to 1E+5] eV [1E+3 to 2E+7] eV 1.4E+7 eV Material(MAT)=4843:Cd-112 001 Total 8.53545E+00 9.35110E+00 1.09556 9.53259E+01 5.90564E+00 4.44747E+00 002 Elastic 6.34021E+00 7.15570E+00 1.12862 8.21914E+01 4.45649E+00 2.60444E+00 102 n,gamma 2.19524E+00 2.19540E+00 1.00007 1.31346E+01 7.40631E-02 6.00318E-05 Material(MAT)=4846:Cd-113 001 Total 2.07451E+04 2.77690E+04 1.33860 4.93075E+02 5.93147E+00 4.47681E+00 002 Elastic 2.44312E+01 5.69090E+01 2.32936 1.01465E+02 4.14484E+00 2.52500E+00 102 n,gamma 2.07207E+04 2.77130E+04 1.33744 3.91610E+02 5.79242E-02 6.00005E-04 Material(MAT)=4849:Cd-114 001 Total 4.81643E+00 5.39310E+00 1.11973 1.17565E+02 5.95002E+00 4.48368E+00 002 Elastic 4.48037E+00 5.05690E+00 1.12869 1.04595E+02 4.46541E+00 2.64271E+00 102 n,gamma 3.36055E-01 3.36150E-01 1.00029 1.29701E+01 3.29201E-02 9.99993E-04 Material(MAT)=4853:Cd-115M 001 Total 3.53850E+01 3.60230E+01 1.01804 2.67164E+02 5.52741E+00 4.61923E+00 002 Elastic 4.38391E+00 5.01360E+00 1.14363 7.06723E+01 4.58814E+00 3.10513E+00

103 102 n,gamma 3.10011E+01 3.10100E+01 1.00027 1.96491E+02 4.68143E-02 2.39885E-03 Material(MAT)=4855:Cd-116 001 Total 5.90629E+00 6.65680E+00 1.12707 7.82829E+01 5.98643E+00 4.51997E+00 002 Elastic 5.83131E+00 6.58180E+00 1.12871 7.67081E+01 4.50324E+00 2.68197E+00 102 n,gamma 7.49863E-02 7.50030E-02 1.00022 1.57494E+00 1.72456E-02 6.00042E-05 Material(MAT)=4925:In-113 001 Total 1.57690E+01 1.63940E+01 1.03963 4.08641E+02 5.53705E+00 4.62764E+00 002 Elastic 3.69538E+00 4.16750E+00 1.12777 8.44001E+01 4.45197E+00 2.49870E+00 102 n,gamma 1.20736E+01 1.22260E+01 1.01265 3.24241E+02 2.00994E-01 1.00552E-03 Material(MAT)=4931:In-115 001 Total 2.03651E+02 2.07860E+02 1.02065 3.40638E+03 5.53556E+00 4.63543E+00 002 Elastic 2.53642E+00 2.83810E+00 1.11892 1.96647E+02 4.44722E+00 2.94201E+00 102 n,gamma 2.01114E+02 2.05020E+02 1.01941 3.20974E+03 1.62835E-01 1.02122E-03 Material(MAT)=5025:Sn-112 001 Total 5.61792E+00 6.21080E+00 1.10553 1.13334E+02 5.53681E+00 4.63999E+00 002 Elastic 4.60876E+00 5.20190E+00 1.12869 8.32629E+01 4.78510E+00 2.93033E+00 102 n,gamma 1.00915E+00 1.00890E+00 0.99977 3.00715E+01 1.09058E-01 1.11473E-03 Material(MAT)=5031:Sn-114 001 Total 4.69113E+00 5.27880E+00 1.12528 8.78257E+01 5.53279E+00 4.63944E+00 002 Elastic 4.56583E+00 5.15350E+00 1.12871 8.14919E+01 4.79072E+00 2.91372E+00 102 n,gamma 1.25298E-01 1.25340E-01 1.00030 6.33397E+00 8.94539E-02 1.02469E-03

Table 9. Simple integral neutron cross-section data from JEFF-3.1 calculated with the INTER computer code (cont.)

V Avg V G fact Avg res integ Avg V fiss V (E14) MT Reaction (2200 m/s) [1E-5 to 10] eV (Avg V)/V(E0) [5E-1 to 1E+5] eV [1E+3 to 2E+7] eV 1.4E+7 eV Material(MAT)=5034:Sn-115 001 Total 3.82588E+01 3.92620E+01 1.02623 8.93714E+01 5.53896E+00 4.62614E+00 002 Elastic 8.40626E+00 9.47690E+00 1.12736 7.59777E+01 4.36278E+00 2.90432E+00 102 n,gamma 2.98525E+01 2.97850E+01 0.99775 1.33936E+01 8.06733E-02 1.00708E-03 Material(MAT)=5037:Sn-116 001 Total 4.45806E+00 5.01540E+00 1.12502 9.56221E+01 5.53888E+00 4.63939E+00 002 Elastic 4.33033E+00 4.88760E+00 1.12870 8.34537E+01 4.80487E+00 2.92097E+00 102 n,gamma 1.27727E-01 1.27780E-01 1.00042 1.21684E+01 5.61089E-02 1.01422E-03 Material(MAT)=5040:Sn-117 001 Total 7.29946E+00 7.95850E+00 1.09028 8.63472E+01 5.53961E+00 4.63577E+00 002 Elastic 5.12599E+00 5.78500E+00 1.12857 6.84831E+01 4.23211E+00 2.92142E+00 102 n,gamma 2.17347E+00 2.17350E+00 1.00000 1.78641E+01 5.46134E-02 1.00649E-03 Material(MAT)=5043:Sn-118 001 Total 4.45761E+00 5.00330E+00 1.12243 7.95210E+01 5.53118E+00 4.63893E+00 002 Elastic 4.23974E+00 4.78540E+00 1.12871 7.43083E+01 4.79760E+00 2.92815E+00

104 102 n,gamma 2.17871E-01 2.17890E-01 1.00009 5.21280E+00 3.19105E-02 1.00205E-03 Material(MAT)=5046:Sn-119 001 Total 6.87177E+00 7.47070E+00 1.08716 7.30315E+01 5.53652E+00 4.62728E+00 002 Elastic 4.69542E+00 5.29930E+00 1.12860 6.72380E+01 4.13630E+00 2.91981E+00 102 n,gamma 2.17635E+00 2.17150E+00 0.99776 5.15433E+00 3.45344E-02 1.00439E-03 Material(MAT)=5049:Sn-120 001 Total 5.50149E+00 6.19170E+00 1.12547 6.56693E+01 5.52982E+00 4.63889E+00 002 Elastic 5.36225E+00 6.05250E+00 1.12872 6.45238E+01 4.78888E+00 2.92807E+00 102 n,gamma 1.39242E-01 1.39260E-01 1.00015 1.14561E+00 1.78833E-02 1.00503E-03 Material(MAT)=5055:Sn-122 001 Total 3.99467E+00 4.48480E+00 1.12270 5.80979E+01 5.53266E+00 4.63851E+00 002 Elastic 3.81097E+00 4.30150E+00 1.12871 5.72121E+01 4.81736E+00 2.92983E+00 102 n,gamma 1.83698E-01 1.83330E-01 0.99799 8.85467E-01 1.03793E-02 1.00590E-03 Material(MAT)=5058:Sn-123 001 Total 4.57959E+00 5.18280E+00 1.13172 7.07646E+01 5.59164E+00 4.65191E+00 002 Elastic 4.54659E+00 5.14980E+00 1.13268 6.82345E+01 4.95542E+00 3.51682E+00 102 n,gamma 3.30011E-02 3.30100E-02 1.00027 2.52952E+00 3.90166E-02 2.38770E-03 Material(MAT)=5061:Sn-124 001 Total 4.55323E+00 5.12170E+00 1.12485 6.79956E+01 5.53200E+00 4.63822E+00 002 Elastic 4.41775E+00 4.98630E+00 1.12870 6.01740E+01 4.78562E+00 2.92906E+00 102 n,gamma 1.35485E-01 1.35390E-01 0.99931 7.82164E+00 6.34190E-03 1.00510E-03

Table 9. Simple integral neutron cross-section data from JEFF-3.1 calculated with the INTER computer code (cont.)

V Avg V G fact Avg res integ Avg V fiss V (E14) MT Reaction (2200 m/s) [1E-5 to 10] eV (Avg V)/V(E0) [5E-1 to 1E+5] eV [1E+3 to 2E+7] eV 1.4E+7 eV Material(MAT)=5064:Sn-125 001 Total 5.13944E+00 5.73030E+00 1.11498 8.70434E+01 5.61205E+00 4.65260E+00 002 Elastic 4.58942E+00 5.18020E+00 1.12872 7.22121E+01 5.07366E+00 3.71436E+00 102 n,gamma 5.50019E-01 5.50170E-01 1.00027 1.48307E+01 1.11216E-02 2.27306E-03 Material(MAT)=5067:Sn-126 001 Total 4.91338E+00 5.50730E+00 1.12087 5.95673E+01 5.61668E+00 4.65612E+00 002 Elastic 4.61337E+00 5.20720E+00 1.12871 5.94075E+01 5.14027E+00 3.14595E+00 102 n,gamma 3.00010E-01 3.00090E-01 1.00027 1.59773E-01 5.62493E-03 1.81098E-03 Material(MAT)=5125:Sb-121 001 Total 9.59805E+00 1.00820E+01 1.05044 2.82351E+02 5.77910E+00 4.70525E+00 002 Elastic 3.60524E+00 4.06790E+00 1.12833 6.84493E+01 4.34178E+00 2.37173E+00 102 n,gamma 5.99281E+00 6.01420E+00 1.00358 2.13663E+02 8.32566E-02 1.00408E-03 Material(MAT)=5131:Sb-123 001 Total 8.10223E+00 8.60630E+00 1.06222 2.18346E+02 5.79676E+00 4.73292E+00 002 Elastic 3.91453E+00 4.41740E+00 1.12845 9.62580E+01 4.53908E+00 2.51949E+00

105 102 n,gamma 4.18770E+00 4.18900E+00 1.00030 1.22088E+02 6.17992E-02 1.00514E-03 Material(MAT)=5134:Sb-124 001 Total 1.10748E+01 1.16830E+01 1.05490 8.77617E+01 5.60579E+00 4.65832E+00 002 Elastic 4.57453E+00 5.18070E+00 1.13252 6.14278E+01 3.88097E+00 3.11079E+00 102 n,gamma 6.50022E+00 6.50200E+00 1.00027 2.59782E+01 5.99563E-02 2.50741E-03 Material(MAT)=5137:Sb-125 001 Total 5.58945E+00 6.18080E+00 1.10579 9.97985E+01 5.62480E+00 4.65900E+00 002 Elastic 4.58942E+00 5.18050E+00 1.12879 8.17445E+01 4.62274E+00 3.14072E+00 102 n,gamma 1.00003E+00 1.00030E+00 1.00027 1.80541E+01 7.80453E-02 2.01261E-03 Material(MAT)=5140:Sb-126 001 Total 1.04288E+01 1.10440E+01 1.05899 1.29683E+02 5.62322E+00 4.65719E+00 002 Elastic 4.62865E+00 5.24230E+00 1.13258 8.40401E+01 4.57356E+00 3.22177E+00 102 n,gamma 5.80020E+00 5.80180E+00 1.00027 4.55309E+01 3.80566E-02 2.37553E-03 Material(MAT)=5225:Te-120 001 Total 7.17874E+00 8.19280E+00 1.14126 1.13139E+02 6.37980E+00 7.52705E+00 002 Elastic 4.89231E+00 5.90550E+00 1.20710 9.37612E+01 5.07946E+00 5.72557E+00 102 n,gamma 2.28643E+00 2.28730E+00 1.00036 1.93774E+01 1.41316E-01 1.40000E-03 Material(MAT)=5231:Te-122 001 Total 5.87157E+00 6.26810E+00 1.06754 1.88297E+02 5.58529E+00 4.64921E+00 002 Elastic 3.07046E+00 3.46550E+00 1.12864 1.14742E+02 4.62643E+00 3.15709E+00 102 n,gamma 2.80110E+00 2.80270E+00 1.00057 7.35552E+01 1.23053E-01 2.26881E-03

Table 9. Simple integral neutron cross-section data from JEFF-3.1 calculated with the INTER computer code (cont.)

V Avg V G fact Avg res integ Avg V fiss V (E14) MT Reaction (2200 m/s) [1E-5 to 10] eV (Avg V)/V(E0) [5E-1 to 1E+5] eV [1E+3 to 2E+7] eV 1.4E+7 eV Material(MAT)=5234:Te-123 001 Total 4.10594E+02 4.15520E+02 1.01200 6.19940E+03 5.59430E+00 4.65302E+00 002 Elastic 5.99734E-01 6.97630E-01 1.16323 6.54393E+02 4.21056E+00 3.11535E+00 102 n,gamma 4.09994E+02 4.14820E+02 1.01177 5.54502E+03 8.69948E-02 2.77100E-03 Material(MAT)=5237:Te-124 001 Total 1.06573E+01 1.11560E+01 1.04676 7.81580E+01 5.60728E+00 4.65583E+00 002 Elastic 3.85721E+00 4.35370E+00 1.12871 7.02285E+01 4.65325E+00 3.23078E+00 102 n,gamma 6.80014E+00 6.80200E+00 1.00027 7.92969E+00 1.09897E-01 2.16671E-03 Material(MAT)=5240:Te-125 001 Total 5.54761E+00 6.06260E+00 1.09283 1.01451E+02 5.62152E+00 4.65885E+00 002 Elastic 3.99743E+00 4.51190E+00 1.12870 7.82553E+01 4.19839E+00 3.15715E+00 102 n,gamma 1.55018E+00 1.55070E+00 1.00033 2.29080E+01 4.30885E-02 2.65827E-03 Material(MAT)=5243:Te-126 001 Total 5.16884E+00 5.70120E+00 1.10299 9.07430E+01 5.63152E+00 4.65544E+00 002 Elastic 4.13374E+00 4.66580E+00 1.12871 8.05515E+01 4.71990E+00 3.15764E+00

106 102 n,gamma 1.03511E+00 1.03540E+00 1.00030 1.01916E+01 3.94435E-02 2.07803E-03 Material(MAT)=5247:Te-127M 001 Total 1.40386E+01 1.46380E+01 1.04271 1.37271E+02 5.63338E+00 4.66034E+00 002 Elastic 4.63832E+00 5.23540E+00 1.12872 9.50185E+01 4.75448E+00 3.14254E+00 102 n,gamma 9.40032E+00 9.40290E+00 1.00027 4.22528E+01 6.62689E-02 2.52642E-03 Material(MAT)=5249:Te-128 001 Total 4.08658E+00 4.58500E+00 1.12196 5.52733E+01 5.88803E+00 4.82995E+00 002 Elastic 3.87165E+00 4.37000E+00 1.12872 5.35679E+01 4.91837E+00 2.87478E+00 102 n,gamma 2.14933E-01 2.14990E-01 1.00028 1.70540E+00 9.29373E-03 7.82185E-06 Material(MAT)=5253:Te-129M 001 Total 5.78626E+00 6.39000E+00 1.10434 9.59121E+01 5.66465E+00 4.65768E+00 002 Elastic 4.68623E+00 5.28970E+00 1.12877 8.99261E+01 4.71901E+00 3.12917E+00 102 n,gamma 1.10004E+00 1.10030E+00 1.00027 5.98609E+00 1.34254E-02 2.43217E-03 Material(MAT)=5255:Te-130 001 Total 4.93680E+00 5.53510E+00 1.12119 5.90384E+01 5.69073E+00 4.65728E+00 002 Elastic 4.64679E+00 5.24500E+00 1.12873 5.87021E+01 4.93479E+00 3.14781E+00 102 n,gamma 2.90012E-01 2.90090E-01 1.00028 3.36403E-01 4.48075E-03 1.95092E-03 Material(MAT)=5261:Te-132 001 Total 4.76009E+00 5.37260E+00 1.12868 5.78688E+01 5.70793E+00 4.65681E+00 002 Elastic 4.75809E+00 5.37060E+00 1.12874 5.78679E+01 4.97577E+00 3.19629E+00 102 n,gamma 2.00007E-03 2.00060E-03 1.00028 8.99657E-04 6.36469E-04 1.90163E-03

Table 9. Simple integral neutron cross-section data from JEFF-3.1 calculated with the INTER computer code (cont.)

V Avg V G fact Avg res integ Avg V fiss V (E14) MT Reaction (2200 m/s) [1E-5 to 10] eV (Avg V)/V(E0) [5E-1 to 1E+5] eV [1E+3 to 2E+7] eV 1.4E+7 eV Material(MAT)=5325:I-127 001 Total 8.94822E+00 9.35370E+00 1.04531 2.58412E+02 5.89173E+00 4.72889E+00 002 Elastic 3.14800E+00 3.55200E+00 1.12835 1.05026E+02 4.34962E+00 2.54034E+00 102 n,gamma 5.80022E+00 5.80160E+00 1.00024 1.53326E+02 7.27073E-02 1.56014E-03 Material(MAT)=5331:I-129 001 Total 3.69180E+01 3.74860E+01 1.01539 1.02265E+02 5.86811E+00 4.73354E+00 002 Elastic 4.38895E+00 4.95370E+00 1.12867 6.84940E+01 4.52568E+00 2.54660E+00 102 n,gamma 3.25291E+01 3.25330E+01 1.00011 3.35656E+01 4.46819E-02 1.35268E-03 Material(MAT)=5334:I-130 001 Total 2.27318E+01 2.33730E+01 1.02819 3.00436E+02 5.69450E+00 4.65731E+00 002 Elastic 4.73122E+00 5.36710E+00 1.13440 1.20084E+02 4.15314E+00 3.12862E+00 102 n,gamma 1.80006E+01 1.80060E+01 1.00027 1.80314E+02 3.37972E-02 2.51133E-03 Material(MAT)=5337:I-131 001 Total 5.43416E+00 6.04400E+00 1.11223 1.03201E+02 5.71537E+00 4.65677E+00 002 Elastic 4.73413E+00 5.34380E+00 1.12878 9.52881E+01 4.55314E+00 3.14336E+00

107 102 n,gamma 7.00024E-01 7.00220E-01 1.00027 7.91291E+00 1.32668E-02 2.10204E-03 Material(MAT)=5349:I-135 001 Total 4.84995E+00 5.47160E+00 1.12818 6.31293E+01 5.77496E+00 4.65687E+00 002 Elastic 4.82995E+00 5.45160E+00 1.12871 6.31172E+01 5.29670E+00 3.24567E+00 102 n,gamma 2.00007E-02 2.00060E-02 1.00028 1.20092E-02 4.67003E-04 1.63037E-03 Material(MAT)=5425:Xe-124 001 Total 1.68867E+02 1.70050E+02 1.00703 3.81479E+03 5.87956E+00 4.80510E+00 002 Elastic 4.31719E+00 4.86720E+00 1.12741 7.61085E+02 4.39985E+00 3.09581E+00 102 n,gamma 1.64549E+02 1.65190E+02 1.00387 3.05370E+03 1.06822E-01 0.00000E+00 Material(MAT)=5431:Xe-126 001 Total 6.52843E+00 7.08520E+00 1.08528 1.85649E+02 5.90447E+00 4.84850E+00 002 Elastic 4.32800E+00 4.88450E+00 1.12859 1.41847E+02 4.51435E+00 2.84298E+00 102 n,gamma 2.20042E+00 2.20060E+00 1.00009 4.38015E+01 7.26523E-02 0.00000E+00 Material(MAT)=5437:Xe-128 001 Total 9.67778E+00 1.02210E+01 1.05610 1.29311E+02 5.93538E+00 4.88760E+00 002 Elastic 4.31687E+00 4.87130E+00 1.12843 1.18334E+02 4.54191E+00 2.65503E+00 102 n,gamma 5.36091E+00 5.34940E+00 0.99786 1.09762E+01 4.93088E-02 0.00000E+00 Material(MAT)=5440:Xe-129 001 Total 2.23205E+01 2.28880E+01 1.02541 3.90332E+02 5.94671E+00 4.90550E+00 002 Elastic 4.31662E+00 4.86870E+00 1.12790 1.34672E+02 4.13125E+00 2.92495E+00 102 n,gamma 1.80039E+01 1.80190E+01 1.00084 2.55221E+02 6.54917E-02 0.00000E+00

Table 9. Simple integral neutron cross-section data from JEFF-3.1 calculated with the INTER computer code (cont.)

V Avg V G fact Avg res integ Avg V fiss V (E14) MT Reaction (2200 m/s) [1E-5 to 10] eV (Avg V)/V(E0) [5E-1 to 1E+5] eV [1E+3 to 2E+7] eV 1.4E+7 eV Material(MAT)=5443:Xe-130 001 Total 1.04985E+01 1.10370E+01 1.05125 8.51622E+01 5.97361E+00 4.90941E+00 002 Elastic 4.29767E+00 4.84950E+00 1.12840 8.08630E+01 4.62610E+00 2.94017E+00 102 n,gamma 6.20080E+00 6.18700E+00 0.99778 4.29919E+00 4.04340E-02 0.00000E+00 Material(MAT)=5446:Xe-131 001 Total 1.09431E+02 1.12710E+02 1.02994 2.96005E+03 5.99052E+00 4.93780E+00 002 Elastic 2.44340E+01 2.75790E+01 1.12873 2.06925E+03 4.17231E+00 2.94901E+00 102 n,gamma 8.49965E+01 8.51270E+01 1.00154 8.90777E+02 2.58103E-02 1.00010E-08 Material(MAT)=5449:Xe-132 001 Total 4.21487E+00 4.69960E+00 1.11500 8.99479E+01 6.02482E+00 4.94166E+00 002 Elastic 3.76618E+00 4.25080E+00 1.12869 8.84602E+01 4.76976E+00 2.97373E+00 102 n,gamma 4.48690E-01 4.48740E-01 1.00011 1.48772E+00 2.27543E-02 1.00000E-08 Material(MAT)=5452:Xe-133 001 Total 1.94976E+02 1.96510E+02 1.00786 4.55057E+02 5.75415E+00 4.65695E+00 002 Elastic 4.97157E+00 6.46680E+00 1.30076 9.78572E+01 4.61572E+00 3.15423E+00

108 102 n,gamma 1.90004E+02 1.90040E+02 1.00020 3.57200E+02 9.01661E-03 2.53768E-03 Material(MAT)=5455:Xe-134 001 Total 4.61203E+00 5.17330E+00 1.12169 8.58919E+01 6.08424E+00 4.97715E+00 002 Elastic 4.36260E+00 4.92410E+00 1.12870 8.53230E+01 4.99502E+00 2.96539E+00 102 n,gamma 2.49421E-01 2.49220E-01 0.99921 5.68949E-01 1.24418E-02 1.00000E-08 Material(MAT)=5458:Xe-135 001 Total 2.95223E+06 3.50200E+06 1.18624 1.28620E+04 5.82116E+00 4.65543E+00 002 Elastic 2.99586E+05 4.33220E+05 1.44606 5.21044E+03 5.04031E+00 3.43864E+00 102 n,gamma 2.65264E+06 3.06880E+06 1.15689 7.65162E+03 7.63153E-04 2.37589E-03 Material(MAT)=5461:Xe-136 001 Total 4.47647E+00 5.03210E+00 1.12412 1.06937E+02 6.13927E+00 4.98479E+00 002 Elastic 4.31647E+00 4.87200E+00 1.12871 1.06818E+02 5.30598E+00 2.96136E+00 102 n,gamma 1.60005E-01 1.60050E-01 1.00027 1.19365E-01 1.05566E-03 0.00000E+00 Material(MAT)=5525:Cs-133 001 Total 3.29728E+01 3.35600E+01 1.01781 5.34791E+02 5.97643E+00 4.84785E+00 002 Elastic 3.97058E+00 4.47730E+00 1.12761 1.13950E+02 4.15171E+00 3.07312E+00 102 n,gamma 2.90023E+01 2.90830E+01 1.00278 4.20812E+02 6.33581E-02 1.49523E-03 Material(MAT)=5528:Cs-134 001 Total 1.62111E+02 1.64730E+02 1.01613 2.19635E+02 5.77371E+00 4.65735E+00 002 Elastic 2.24271E+01 2.52550E+01 1.12607 1.40126E+02 3.98875E+00 3.15799E+00 102 n,gamma 1.39684E+02 1.39470E+02 0.99848 7.87740E+01 3.31913E-02 2.70517E-03

Table 9. Simple integral neutron cross-section data from JEFF-3.1 calculated with the INTER computer code (cont.)

V Avg V G fact Avg res integ Avg V fiss V (E14) MT Reaction (2200 m/s) [1E-5 to 10] eV (Avg V)/V(E0) [5E-1 to 1E+5] eV [1E+3 to 2E+7] eV 1.4E+7 eV Material(MAT)=5531:Cs-135 001 Total 1.40851E+01 1.47300E+01 1.04579 1.88934E+02 6.01834E+00 4.93401E+00 002 Elastic 5.07879E+00 5.73050E+00 1.12831 1.27890E+02 4.54217E+00 2.89811E+00 102 n,gamma 9.00631E+00 8.99960E+00 0.99926 6.10443E+01 2.77431E-02 9.20974E-04 Material(MAT)=5534:Cs-136 001 Total 5.21561E+00 5.71990E+00 1.09670 2.18762E+02 5.82570E+00 4.65802E+00 002 Elastic 3.91525E+00 4.41900E+00 1.12866 1.80349E+02 4.76313E+00 3.17955E+00 102 n,gamma 1.30036E+00 1.30090E+00 1.00045 3.84132E+01 2.77777E-02 2.55253E-03 Material(MAT)=5537:Cs-137 001 Total 2.65855E+00 2.98250E+00 1.12187 8.98314E+01 6.06730E+00 4.97952E+00 002 Elastic 2.51709E+00 2.84110E+00 1.12871 8.92325E+01 5.07321E+00 2.92165E+00 102 n,gamma 1.41456E-01 1.41480E-01 1.00019 5.99005E-01 6.83079E-03 8.95581E-04 Material(MAT)=5625:Ba-130 001 Total 1.44323E+01 1.48380E+01 1.02810 4.82943E+02 6.23443E+00 5.07875E+00 002 Elastic 3.13784E+00 3.54010E+00 1.12820 3.07809E+02 4.64154E+00 2.67829E+00

109 102 n,gamma 1.12944E+01 1.12980E+01 1.00029 1.75133E+02 5.24792E-01 1.64038E-02 Material(MAT)=5631:Ba-132 001 Total 1.03128E+01 1.07450E+01 1.04195 9.60675E+01 6.23172E+00 5.07830E+00 002 Elastic 3.31261E+00 3.74330E+00 1.13003 6.59367E+01 4.86493E+00 2.74511E+00 102 n,gamma 7.00023E+00 7.00210E+00 1.00027 3.01307E+01 2.73084E-01 3.81271E-03 Material(MAT)=5637:Ba-134 001 Total 5.44145E+00 5.88400E+00 1.08133 1.15133E+02 6.24948E+00 5.07826E+00 002 Elastic 3.43933E+00 3.88160E+00 1.12860 9.09261E+01 5.06410E+00 2.77496E+00 102 n,gamma 2.00213E+00 2.00240E+00 1.00013 2.42066E+01 1.37749E-01 1.50557E-03 Material(MAT)=5640:Ba-135 001 Total 7.60122E+00 7.83470E+00 1.03072 2.88052E+02 6.23955E+00 5.06948E+00 002 Elastic 1.80439E+00 2.03590E+00 1.12830 1.57286E+02 4.96942E+00 2.90088E+00 102 n,gamma 5.79684E+00 5.79890E+00 1.00035 1.30766E+02 9.71956E-02 1.00650E-03 Material(MAT)=5643:Ba-136 001 Total 2.97403E+00 3.30530E+00 1.11140 5.95491E+01 6.24670E+00 5.07790E+00 002 Elastic 2.57388E+00 2.90510E+00 1.12869 5.76422E+01 5.29536E+00 2.83966E+00 102 n,gamma 4.00153E-01 4.00220E-01 1.00017 1.90697E+00 3.67517E-02 1.01614E-03 Material(MAT)=5646:Ba-137 001 Total 9.01835E+00 9.51460E+00 1.05503 7.26831E+01 6.24176E+00 5.06345E+00 002 Elastic 3.89579E+00 4.39570E+00 1.12832 6.80078E+01 5.37689E+00 2.89473E+00 102 n,gamma 5.12256E+00 5.11890E+00 0.99928 4.67542E+00 1.48479E-02 1.00497E-03

Table 9. Simple integral neutron cross-section data from JEFF-3.1 calculated with the INTER computer code (cont.)

V Avg V G fact Avg res integ Avg V fiss V (E14) MT Reaction (2200 m/s) [1E-5 to 10] eV (Avg V)/V(E0) [5E-1 to 1E+5] eV [1E+3 to 2E+7] eV 1.4E+7 eV Material(MAT)=5649:Ba-138 001 Total 5.92926E+00 6.64640E+00 1.12094 6.03901E+01 6.23190E+00 5.07772E+00 002 Elastic 5.57014E+00 6.28710E+00 1.12872 6.01371E+01 5.57406E+00 2.88892E+00 102 n,gamma 3.59127E-01 3.59220E-01 1.00026 2.53072E-01 2.97335E-03 1.35041E-03 Material(MAT)=5655:Ba-140 001 Total 3.49423E+00 3.73890E+00 1.07003 4.01657E+02 6.14570E+00 5.03570E+00 002 Elastic 1.89951E+00 2.14350E+00 1.12847 3.87326E+02 5.28868E+00 2.93905E+00 102 n,gamma 1.59472E+00 1.59540E+00 1.00043 1.43308E+01 6.54869E-03 9.70605E-04 Material(MAT)=5725:La-138 001 Total 7.01184E+01 7.18860E+01 1.02521 4.91670E+02 6.17162E+00 5.04681E+00 002 Elastic 1.30237E+01 1.46750E+01 1.12680 1.27378E+02 4.86201E+00 2.90534E+00 102 n,gamma 5.70947E+01 5.72110E+01 1.00204 3.64288E+02 3.44466E-02 1.00535E-03 Material(MAT)=5728:La-139 001 Total 1.94823E+01 2.08320E+01 1.06925 1.01627E+02 6.07183E+00 5.04607E+00 002 Elastic 1.05499E+01 1.19010E+01 1.12809 8.97186E+01 4.84001E+00 2.96603E+00

113 110 102 n,gamma 8.93241E+00 8.93040E+00 0.99977 1.19081E+01 5.47889E-03 7.08895E-04 Material(MAT)=5731:La-140 001 Total 7.65319E+00 8.29450E+00 1.08380 2.90795E+02 5.93506E+00 4.67091E+00 002 Elastic 4.95310E+00 5.59370E+00 1.12933 2.25353E+02 4.17089E+00 2.99612E+00 102 n,gamma 2.70009E+00 2.70080E+00 1.00028 6.51025E+01 3.37906E-02 2.33503E-03 Material(MAT)=5837:Ce-140 001 Total 3.85005E+00 4.27150E+00 1.10945 4.89394E+01 5.93011E+00 4.66965E+00 002 Elastic 3.27545E+00 3.69700E+00 1.12869 4.86577E+01 5.39983E+00 3.22949E+00 102 n,gamma 5.74596E-01 5.74480E-01 0.99979 2.81670E-01 1.47860E-02 2.14120E-03 Material(MAT)=5840:Ce-141 001 Total 3.61496E+01 3.65950E+01 1.01233 3.87723E+02 6.16796E+00 5.05050E+00 002 Elastic 3.41214E+00 3.85140E+00 1.12873 2.22890E+02 5.11639E+00 2.93692E+00 102 n,gamma 3.27374E+01 3.27440E+01 1.00020 1.64833E+02 7.77741E-02 1.47177E-03 Material(MAT)=5843:Ce-142 001 Total 4.68385E+00 5.17100E+00 1.10401 2.00383E+02 6.19879E+00 5.06962E+00 002 Elastic 3.78337E+00 4.27040E+00 1.12872 1.99452E+02 4.97569E+00 2.94353E+00 102 n,gamma 9.00478E-01 9.00650E-01 1.00019 9.30422E-01 1.73754E-02 1.05978E-03 Material(MAT)=5846:Ce-143 001 Total 1.10338E+01 1.17090E+01 1.06119 2.75137E+02 6.00877E+00 4.67563E+00 002 Elastic 5.03360E+00 5.70710E+00 1.13380 2.33564E+02 5.07663E+00 3.24080E+00 102 n,gamma 6.00020E+00 6.00180E+00 1.00027 4.15728E+01 1.28590E-02 2.41014E-03

Table 9. Simple integral neutron cross-section data from JEFF-3.1 calculated with the INTER computer code (cont.)

V Avg V G fact Avg res integ Avg V fiss V (E14) MT Reaction (2200 m/s) [1E-5 to 10] eV (Avg V)/V(E0) [5E-1 to 1E+5] eV [1E+3 to 2E+7] eV 1.4E+7 eV Material(MAT)=5849:Ce-144 001 Total 4.53046E+00 4.96850E+00 1.09670 2.75337E+02 6.29028E+00 5.10378E+00 002 Elastic 3.40186E+00 3.83970E+00 1.12871 2.72534E+02 4.46034E+00 2.95810E+00 102 n,gamma 1.12860E+00 1.12880E+00 1.00020 2.80294E+00 7.89206E-03 9.46987E-04 Material(MAT)=5925:Pr-141 001 Total 1.40797E+01 1.44060E+01 1.02319 2.33638E+02 6.11865E+00 5.06028E+00 002 Elastic 2.59649E+00 2.92790E+00 1.12766 2.15765E+02 4.71569E+00 2.96974E+00 102 n,gamma 1.14832E+01 1.14780E+01 0.99957 1.78731E+01 1.54860E-02 2.72698E-03 Material(MAT)=5928:Pr-142 001 Total 2.50162E+01 2.56930E+01 1.02707 4.99554E+02 5.98729E+00 4.67107E+00 002 Elastic 5.01555E+00 5.68720E+00 1.13392 3.54487E+02 4.81663E+00 3.23383E+00 102 n,gamma 2.00007E+01 2.00060E+01 1.00027 1.45068E+02 2.91781E-02 2.57024E-03 Material(MAT)=5931:Pr-143 001 Total 9.40948E+01 9.50000E+01 1.00962 5.92578E+02 6.02221E+00 4.67591E+00 002 Elastic 5.09180E+00 5.97290E+00 1.17304 4.02927E+02 4.82591E+00 3.59646E+00

111 102 n,gamma 8.90030E+01 8.90280E+01 1.00028 1.89542E+02 5.67475E-02 2.24774E-03 Material(MAT)=6025:Nd-142 001 Total 2.71825E+01 2.82520E+01 1.03936 8.67610E+01 5.98653E+00 4.67566E+00 002 Elastic 8.54608E+00 9.63670E+00 1.12762 8.06803E+01 5.39060E+00 3.29115E+00 102 n,gamma 1.86364E+01 1.86160E+01 0.99889 6.08074E+00 2.43200E-02 2.41992E-03 Material(MAT)=6028:Nd-143 001 Total 4.22821E+02 4.32060E+02 1.02186 7.40712E+02 6.33797E+00 4.94350E+00 002 Elastic 8.68805E+01 9.73450E+01 1.12044 6.08148E+02 5.17213E+00 2.77188E+00 102 n,gamma 3.35941E+02 3.34720E+02 0.99636 1.32564E+02 7.75738E-02 1.07240E-03 Material(MAT)=6031:Nd-144 001 Total 3.84361E+00 3.87670E+00 1.00861 6.49448E+02 6.30889E+00 5.06023E+00 002 Elastic 2.60325E-01 2.93400E-01 1.12706 6.45368E+02 5.05959E+00 2.96461E+00 102 n,gamma 3.58329E+00 3.58330E+00 1.00000 4.07978E+00 2.60662E-02 9.13997E-04 Material(MAT)=6034:Nd-145 001 Total 6.01845E+01 6.25000E+01 1.03847 7.97907E+02 6.33924E+00 5.05685E+00 002 Elastic 1.83242E+01 2.06410E+01 1.12646 5.66711E+02 4.57894E+00 2.96619E+00 102 n,gamma 4.18603E+01 4.18580E+01 0.99995 2.31147E+02 7.39299E-02 1.00200E-03 Material(MAT)=6037:Nd-146 001 Total 5.78161E+00 6.34460E+00 1.09737 1.67751E+02 6.36724E+00 5.07162E+00 002 Elastic 4.38137E+00 4.94460E+00 1.12856 1.65087E+02 4.85727E+00 2.98000E+00 102 n,gamma 1.40023E+00 1.39990E+00 0.99979 2.66392E+00 2.92585E-02 8.94296E-04

Table 9. Simple integral neutron cross-section data from JEFF-3.1 calculated with the INTER computer code (cont.)

V Avg V G fact Avg res integ Avg V fiss V (E14) MT Reaction (2200 m/s) [1E-5 to 10] eV (Avg V)/V(E0) [5E-1 to 1E+5] eV [1E+3 to 2E+7] eV 1.4E+7 eV Material(MAT)=6040:Nd-147 001 Total 5.45976E+02 5.56630E+02 1.01951 1.13767E+03 6.14100E+00 4.69762E+00 002 Elastic 1.06029E+02 1.18670E+02 1.11927 5.31769E+02 4.35901E+00 3.27372E+00 102 n,gamma 4.39947E+02 4.37950E+02 0.99546 6.05838E+02 7.11493E-02 2.59846E-03 Material(MAT)=6043:Nd-148 001 Total 3.16554E+00 3.25180E+00 1.02724 5.77600E+02 6.45512E+00 5.24066E+00 002 Elastic 6.65140E-01 7.50430E-01 1.12823 5.57886E+02 4.83577E+00 3.08600E+00 102 n,gamma 2.50040E+00 2.50130E+00 1.00037 1.97140E+01 3.57942E-02 8.73797E-04 Material(MAT)=6049:Nd-150 001 Total 5.94153E+00 6.55300E+00 1.10292 1.85212E+02 6.81981E+00 5.24139E+00 002 Elastic 4.75467E+00 5.36630E+00 1.12863 1.69589E+02 4.66529E+00 2.72256E+00 102 n,gamma 1.18685E+00 1.18670E+00 0.99988 1.56227E+01 3.15871E-02 8.76310E-04 Material(MAT)=6149:Pm-147 001 Total 1.82545E+02 1.82020E+02 0.99711 3.30675E+03 6.46712E+00 5.17263E+00 002 Elastic 1.96463E+00 2.21570E+00 1.12777 1.16419E+03 4.51550E+00 3.03016E+00

112 102 n,gamma 1.80581E+02 1.79800E+02 0.99568 2.14255E+03 1.87729E-01 9.56997E-04 Material(MAT)=6152:Pm-148 001 Total 2.00520E+03 2.00650E+03 1.00067 4.00033E+04 6.16248E+00 4.70235E+00 002 Elastic 5.13441E+00 5.79530E+00 1.12872 1.44303E+02 3.68168E+00 3.28807E+00 102 n,gamma 2.00007E+03 2.00070E+03 1.00034 3.98590E+04 6.27577E-01 2.64869E-03 Material(MAT)=6153:Pm-148M 001 Total 1.07024E+04 1.58550E+04 1.48142 3.80209E+03 6.16253E+00 4.70235E+00 002 Elastic 3.57047E+01 7.12990E+01 1.99691 1.57003E+02 3.68174E+00 3.28807E+00 102 n,gamma 1.06666E+04 1.57830E+04 1.47970 3.64509E+03 6.27577E-01 2.64869E-03 Material(MAT)=6155:Pm-149 001 Total 1.40542E+03 1.40760E+03 1.00156 9.97992E+02 6.19523E+00 4.70681E+00 002 Elastic 5.37613E+00 7.18760E+00 1.33694 2.00290E+02 3.96386E+00 3.31971E+00 102 n,gamma 1.40005E+03 1.40040E+03 1.00027 7.97703E+02 6.12849E-01 2.26518E-03 Material(MAT)=6161:Pm-151 001 Total 7.05502E+02 7.08040E+02 1.00360 2.14743E+03 6.25801E+00 4.71701E+00 002 Elastic 5.48119E+00 7.85500E+00 1.43308 1.38791E+02 4.61059E+00 3.27933E+00 102 n,gamma 7.00020E+02 7.00190E+02 1.00024 2.00864E+03 1.91793E-03 2.32632E-03 Material(MAT)=6225:Sm-144 001 Total 5.46030E+00 6.26620E+00 1.14759 2.46208E+02 6.75530E+00 7.88210E+00 002 Elastic 4.76451E+00 5.57060E+00 1.16919 2.27331E+02 5.86841E+00 5.99188E+00 102 n,gamma 6.95795E-01 6.95580E-01 0.99969 1.88769E+01 2.12040E-01 2.07000E-03

Table 9. Simple integral neutron cross-section data from JEFF-3.1 calculated with the INTER computer code (cont.)

V Avg V G fact Avg res integ Avg V fiss V (E14) MT Reaction (2200 m/s) [1E-5 to 10] eV (Avg V)/V(E0) [5E-1 to 1E+5] eV [1E+3 to 2E+7] eV 1.4E+7 eV Material(MAT)=6234:Sm-147 001 Total 9.72067E+01 1.02360E+02 1.05301 1.63884E+03 6.92849E+00 5.03607E+00 002 Elastic 4.00190E+01 4.51040E+01 1.12707 8.45200E+02 4.76666E+00 2.95475E+00 102 n,gamma 5.71877E+01 5.72550E+01 1.00118 7.93641E+02 2.07862E-01 1.19600E-03 Material(MAT)=6237:Sm-148 001 Total 7.83882E+00 8.50420E+00 1.08489 3.71927E+02 6.17080E+00 4.70036E+00 002 Elastic 5.13872E+00 5.80340E+00 1.12935 3.45001E+02 4.92006E+00 3.35836E+00 102 n,gamma 2.70009E+00 2.70080E+00 1.00028 2.69267E+01 1.94588E-01 2.42606E-03 Material(MAT)=6240:Sm-149 001 Total 4.19315E+04 7.21320E+04 1.72023 4.02483E+03 7.07552E+00 5.20066E+00 002 Elastic 2.05356E+02 4.92560E+02 2.39858 5.38733E+02 4.47441E+00 3.06700E+00 102 n,gamma 4.17261E+04 7.16390E+04 1.71689 3.48571E+03 1.91141E-01 1.65881E-03 Material(MAT)=6243:Sm-150 001 Total 1.24602E+02 1.27050E+02 1.01963 8.40551E+02 6.23173E+00 4.71058E+00 002 Elastic 2.12547E+01 2.39030E+01 1.12460 5.01624E+02 4.66629E+00 3.35046E+00

113 102 n,gamma 1.03347E+02 1.03140E+02 0.99804 3.38928E+02 1.42895E-01 2.33405E-03 Material(MAT)=6246:Sm-151 001 Total 1.52220E+04 1.41140E+04 0.92724 3.76610E+03 6.77706E+00 5.00281E+00 002 Elastic 3.92486E+01 3.65600E+01 0.93151 2.97566E+02 3.81745E+00 2.77746E+00 102 n,gamma 1.51828E+04 1.40780E+04 0.92723 3.46510E+03 1.67794E-01 1.62690E-03 Material(MAT)=6249:Sm-152 001 Total 2.08945E+02 2.10080E+02 1.00544 9.05378E+03 7.03423E+00 5.01463E+00 002 Elastic 2.94540E+00 3.38580E+00 1.14951 6.07396E+03 4.69268E+00 2.95200E+00 102 n,gamma 2.06000E+02 2.06700E+02 1.00338 2.97981E+03 9.10177E-02 9.29597E-04 Material(MAT)=6252:Sm-153 001 Total 3.35459E+02 3.37240E+02 1.00531 3.11941E+03 6.31137E+00 4.72756E+00 002 Elastic 5.44481E+00 7.11190E+00 1.30618 2.47906E+02 4.17776E+00 3.28170E+00 102 n,gamma 3.30014E+02 3.30130E+02 1.00035 2.86892E+03 5.94565E-04 2.78529E-03 Material(MAT)=6255:Sm-154 001 Total 1.94372E+01 2.08530E+01 1.07283 2.69253E+02 6.34279E+00 4.73264E+00 002 Elastic 1.10358E+01 1.24530E+01 1.12841 2.35970E+02 4.73673E+00 3.33461E+00 102 n,gamma 8.40141E+00 8.40010E+00 0.99984 3.32715E+01 6.01075E-02 2.31791E-03 Material(MAT)=6325:Eu-151 001 Total 9.16842E+03 8.25860E+03 0.90076 3.60364E+03 7.03903E+00 5.12993E+00 002 Elastic 3.39374E+00 3.18050E+00 0.93716 2.51581E+02 4.45061E+00 2.84738E+00 102 n,gamma 9.16502E+03 8.25540E+03 0.90075 3.35117E+03 4.61639E-01 1.40000E-03

Table 9. Simple integral neutron cross-section data from JEFF-3.1 calculated with the INTER computer code (cont.)

V Avg V G fact Avg res integ Avg V fiss V (E14) MT Reaction (2200 m/s) [1E-5 to 10] eV (Avg V)/V(E0) [5E-1 to 1E+5] eV [1E+3 to 2E+7] eV 1.4E+7 eV Material(MAT)=6328:Eu-152 001 Total 1.27777E+04 1.18250E+04 0.92544 2.67100E+03 6.54564E+00 5.05046E+00 002 Elastic 2.90856E+01 2.93620E+01 1.00949 5.03433E+02 4.24169E+00 3.28122E+00 102 n,gamma 1.27486E+04 1.17960E+04 0.92524 2.16755E+03 5.27810E-01 1.00593E-03 Material(MAT)=6331:Eu-153 001 Total 3.22975E+02 3.20080E+02 0.99102 1.60556E+03 6.87232E+00 5.21414E+00 002 Elastic 1.03621E+01 1.15480E+01 1.11449 1.96332E+02 4.57751E+00 2.87930E+00 102 n,gamma 3.12613E+02 3.08530E+02 0.98693 1.40921E+03 2.63953E-01 1.00476E-03 Material(MAT)=6334:Eu-154 001 Total 1.85151E+03 2.20840E+03 1.19273 1.52618E+03 6.57906E+00 4.96817E+00 002 Elastic 6.64621E+00 7.42200E+00 1.11672 1.68460E+02 3.76144E+00 2.61035E+00 102 n,gamma 1.84487E+03 2.20090E+03 1.19300 1.35770E+03 2.21942E-01 2.60257E-03 Material(MAT)=6337:Eu-155 001 Total 3.77531E+03 3.87080E+03 1.02530 1.60405E+04 6.84272E+00 5.39196E+00 002 Elastic 1.76763E+01 2.04890E+01 1.15912 7.58882E+02 4.59072E+00 2.99332E+00

114 102 n,gamma 3.75764E+03 3.85040E+03 1.02467 1.52815E+04 2.02071E-01 1.26829E-03 Material(MAT)=6340:Eu-156 001 Total 4.87636E+02 4.90370E+02 1.00561 1.64423E+03 6.38714E+00 4.74149E+00 002 Elastic 5.59323E+00 7.99790E+00 1.42993 1.50304E+02 4.17578E+00 3.28122E+00 102 n,gamma 4.82043E+02 4.82380E+02 1.00069 1.49225E+03 2.02713E-03 2.78996E-03 Material(MAT)=6343:Eu-157 001 Total 1.95328E+02 1.95930E+02 1.00310 1.42875E+03 6.40340E+00 4.74570E+00 002 Elastic 5.33907E+00 6.02620E+00 1.12870 1.26464E+02 4.97423E+00 3.27677E+00 102 n,gamma 1.89989E+02 1.89910E+02 0.99957 1.30225E+03 4.35298E-03 2.33843E-03 Material(MAT)=6425:Gd-152 001 Total 1.06992E+03 1.06940E+03 0.99949 1.58152E+03 7.01419E+00 5.68473E+00 002 Elastic 1.39620E+01 1.56950E+01 1.12415 5.93444E+02 5.15118E+00 3.40192E+00 102 n,gamma 1.05595E+03 1.05370E+03 0.99784 9.88018E+02 2.94815E-01 1.02599E-03 Material(MAT)=6431:Gd-154 001 Total 8.85890E+01 8.90750E+01 1.00549 5.36009E+02 6.33383E+00 4.73239E+00 002 Elastic 3.58502E+00 4.04570E+00 1.12849 2.89576E+02 4.63566E+00 3.36103E+00 102 n,gamma 8.50040E+01 8.50300E+01 1.00030 2.46433E+02 3.07980E-01 2.61100E-03 Material(MAT)=6434:Gd-155 001 Total 6.07915E+04 5.13060E+04 0.84396 1.70604E+03 6.98683E+00 5.68006E+00 002 Elastic 5.90253E+01 5.73340E+01 0.97134 1.61822E+02 4.51608E+00 3.24257E+00 102 n,gamma 6.07325E+04 5.12480E+04 0.84384 1.54415E+03 3.48948E-01 2.07595E-05

Table 9. Simple integral neutron cross-section data from JEFF-3.1 calculated with the INTER computer code (cont.)

V Avg V G fact Avg res integ Avg V fiss V (E14) MT Reaction (2200 m/s) [1E-5 to 10] eV (Avg V)/V(E0) [5E-1 to 1E+5] eV [1E+3 to 2E+7] eV 1.4E+7 eV Material(MAT)=6437:Gd-156 001 Total 7.10870E+00 7.82650E+00 1.10097 3.08778E+02 6.71988E+00 5.38375E+00 002 Elastic 5.57268E+00 6.28940E+00 1.12861 2.08567E+02 4.76107E+00 3.07230E+00 102 n,gamma 1.53602E+00 1.53710E+00 1.00071 1.00206E+02 1.19318E-01 1.77620E-03 Material(MAT)=6440:Gd-157 001 Total 2.55015E+05 2.17300E+05 0.85209 9.44337E+02 6.70060E+00 5.39500E+00 002 Elastic 1.76028E+03 1.60260E+03 0.91042 1.82054E+02 4.47803E+00 3.08984E+00 102 n,gamma 2.53255E+05 2.15690E+05 0.85169 7.61984E+02 2.04190E-01 7.61219E-05 Material(MAT)=6443:Gd-158 001 Total 6.01037E+00 6.46360E+00 1.07541 1.95442E+02 6.42782E+00 4.74784E+00 002 Elastic 3.50975E+00 3.96120E+00 1.12862 1.32648E+02 4.83559E+00 3.29698E+00 102 n,gamma 2.50062E+00 2.50240E+00 1.00072 6.27807E+01 5.57585E-02 2.88859E-03 Material(MAT)=6449:Gd-160 001 Total 4.49666E+00 4.97650E+00 1.10670 1.65552E+02 6.46297E+00 4.75316E+00 002 Elastic 3.72656E+00 4.20610E+00 1.12870 1.57311E+02 4.89067E+00 3.28209E+00

115 102 n,gamma 7.70102E-01 7.70310E-01 1.00027 8.22007E+00 8.00632E-02 2.34045E-03 Material(MAT)=6525:Tb-159 001 Total 2.97831E+01 3.06330E+01 1.02855 6.06881E+02 6.71619E+00 5.00964E+00 002 Elastic 6.57074E+00 7.41640E+00 1.12871 1.93682E+02 4.03216E+00 2.49924E+00 102 n,gamma 2.32124E+01 2.32170E+01 1.00019 4.12930E+02 1.59049E-01 1.77106E-03 Material(MAT)=6528:Tb-160 001 Total 5.30385E+02 5.30920E+02 1.00101 1.31478E+03 6.46066E+00 4.75429E+00 002 Elastic 5.40696E+00 6.10290E+00 1.12871 1.80098E+02 4.33765E+00 3.25410E+00 102 n,gamma 5.24978E+02 5.24820E+02 0.99969 1.13468E+03 4.41010E-03 2.92251E-03 Material(MAT)=6637:Dy-160 001 Total 6.33488E+01 6.41790E+01 1.01311 2.31679E+03 6.45683E+00 4.75434E+00 002 Elastic 2.31584E+00 2.60320E+00 1.12411 6.40656E+02 4.67200E+00 3.65687E+00 102 n,gamma 6.10330E+01 6.15760E+01 1.00889 1.67613E+03 7.66040E-01 2.67380E-03 Material(MAT)=6640:Dy-161 001 Total 6.16906E+02 6.12950E+02 0.99359 1.24958E+03 7.13719E+00 5.18371E+00 002 Elastic 1.65957E+01 1.84680E+01 1.11279 1.62276E+02 5.33577E+00 3.70596E+00 102 n,gamma 6.00310E+02 5.94480E+02 0.99030 1.08681E+03 1.35669E-01 3.08394E-03 Material(MAT)=6643:Dy-162 001 Total 1.93849E+02 1.94790E+02 1.00488 3.51660E+03 7.14106E+00 5.18370E+00 002 Elastic 1.23686E-02 1.24760E-02 1.00866 7.68876E+02 5.72222E+00 3.75696E+00 102 n,gamma 1.93836E+02 1.94780E+02 1.00488 2.74771E+03 7.15063E-02 1.33693E-03

Table 9. Simple integral neutron cross-section data from JEFF-3.1 calculated with the INTER computer code (cont.)

V Avg V G fact Avg res integ Avg V fiss V (E14) MT Reaction (2200 m/s) [1E-5 to 10] eV (Avg V)/V(E0) [5E-1 to 1E+5] eV [1E+3 to 2E+7] eV 1.4E+7 eV Material(MAT)=6646:Dy-163 001 Total 1.27615E+02 1.29430E+02 1.01425 1.71747E+03 7.13464E+00 5.18371E+00 002 Elastic 3.38905E+00 3.80460E+00 1.12262 2.27802E+02 5.14334E+00 3.65376E+00 102 n,gamma 1.24226E+02 1.25630E+02 1.01130 1.48945E+03 9.51243E-02 2.90953E-03 Material(MAT)=6649:Dy-164 001 Total 2.97751E+03 2.97920E+03 1.00057 8.86309E+02 7.22037E+00 5.26036E+00 002 Elastic 3.26899E+02 3.60740E+02 1.10352 5.43651E+02 4.88048E+00 3.01316E+00 102 n,gamma 2.65061E+03 2.61850E+03 0.98788 3.42589E+02 2.99838E-02 1.34147E-03 Material(MAT)=6725:Ho-165 001 Total 7.22400E+01 7.34850E+01 1.01723 9.15277E+02 6.88753E+00 5.32534E+00 002 Elastic 8.79578E+00 9.92200E+00 1.12804 2.40350E+02 4.23773E+00 2.87544E+00 102 n,gamma 6.34442E+01 6.35630E+01 1.00187 6.74922E+02 1.22018E-01 1.27051E-03 Material(MAT)=6825:Er-162 001 Total 2.69420E+01 2.79900E+01 1.03891 6.94011E+02 6.74843E+00 5.42452E+00 002 Elastic 8.02722E+00 9.05440E+00 1.12796 2.42991E+02 4.80704E+00 2.67860E+00

116 102 n,gamma 1.89148E+01 1.89360E+01 1.00112 4.51020E+02 1.81869E-01 1.63762E-03 Material(MAT)=6831:Er-164 001 Total 2.20816E+01 2.32510E+01 1.05296 3.40948E+02 6.78852E+00 5.45805E+00 002 Elastic 9.12779E+00 1.02980E+01 1.12823 1.74418E+02 4.67453E+00 2.70048E+00 102 n,gamma 1.29538E+01 1.29530E+01 0.99992 1.66528E+02 3.75123E-01 1.63762E-03 Material(MAT)=6837:Er-166 001 Total 2.92535E+01 3.08500E+01 1.05458 3.65356E+02 6.82014E+00 5.50210E+00 002 Elastic 1.24858E+01 1.40860E+01 1.12812 2.53647E+02 4.65353E+00 2.75532E+00 102 n,gamma 1.67677E+01 1.67650E+01 0.99982 1.11693E+02 9.57374E-02 1.93330E-03 Material(MAT)=6840:Er-167 001 Total 6.44406E+02 6.88930E+02 1.06909 3.45779E+03 6.83830E+00 5.51464E+00 002 Elastic 1.33758E+00 1.41300E+00 1.05639 4.84767E+02 4.32610E+00 2.71959E+00 102 n,gamma 6.43069E+02 6.87520E+02 1.06912 2.97299E+03 9.42873E-02 1.97240E-03 Material(MAT)=6843:Er-168 001 Total 1.17142E+01 1.28700E+01 1.09865 2.36914E+02 6.85496E+00 5.53530E+00 002 Elastic 8.98476E+00 1.01400E+01 1.12858 1.98795E+02 4.67989E+00 2.76975E+00 102 n,gamma 2.72949E+00 2.72990E+00 1.00014 3.81001E+01 3.59995E-02 1.89858E-03 Material(MAT)=6849:Er-170 001 Total 1.77795E+01 1.93240E+01 1.08684 6.15132E+02 6.81984E+00 5.50199E+00 002 Elastic 1.20024E+01 1.35440E+01 1.12847 5.69983E+02 4.60897E+00 2.72149E+00 102 n,gamma 5.77708E+00 5.77920E+00 1.00036 4.51291E+01 2.29469E-02 1.86103E-03

Table 9. Simple integral neutron cross-section data from JEFF-3.1 calculated with the INTER computer code (cont.)

V Avg V G fact Avg res integ Avg V fiss V (E14) MT Reaction (2200 m/s) [1E-5 to 10] eV (Avg V)/V(E0) [5E-1 to 1E+5] eV [1E+3 to 2E+7] eV 1.4E+7 eV Material(MAT)=7125:Lu-175 001 Total 2.95134E+01 3.03920E+01 1.02976 8.02165E+02 7.22216E+00 5.25740E+00 002 Elastic 6.43598E+00 7.25530E+00 1.12730 1.81486E+02 5.07417E+00 3.02075E+00 102 n,gamma 2.30775E+01 2.31360E+01 1.00255 6.20680E+02 1.63859E-01 2.25000E-03 Material(MAT)=7128:Lu-176 001 Total 2.09953E+03 3.59380E+03 1.71170 1.07081E+03 7.22920E+00 5.26000E+00 002 Elastic 2.51529E+00 4.20640E+00 1.67232 1.51063E+02 5.04403E+00 2.95252E+00 102 n,gamma 2.09701E+03 3.58960E+03 1.71175 9.19744E+02 2.05325E-01 3.08394E-03 Material(MAT)=7225:Hf-174 001 Total 5.97764E+02 5.95910E+02 0.99690 1.11876E+03 7.14368E+00 5.32588E+00 002 Elastic 4.81196E+01 5.37080E+01 1.11614 6.78027E+02 4.60652E+00 2.74155E+00 102 n,gamma 5.49645E+02 5.42200E+02 0.98646 4.40732E+02 2.76722E-01 4.18124E-06 Material(MAT)=7231:Hf-176 001 Total 2.68569E+01 2.76080E+01 1.02798 1.12105E+03 7.14374E+00 5.36465E+00 002 Elastic 5.52947E+00 6.23160E+00 1.12698 4.27084E+02 4.68666E+00 2.75094E+00

117 102 n,gamma 2.13274E+01 2.13770E+01 1.00231 6.93962E+02 9.37184E-02 4.21718E-07 Material(MAT)=7234:Hf-177 001 Total 3.72003E+02 3.79470E+02 1.02006 8.01142E+03 7.12749E+00 5.37822E+00 002 Elastic 3.61527E-02 4.34800E-02 1.20268 7.99304E+02 4.56377E+00 2.88401E+00 102 n,gamma 3.71967E+02 3.79420E+02 1.02004 7.21212E+03 1.63819E-01 1.93670E-07 Material(MAT)=7237:Hf-178 001 Total 9.05826E+01 9.16380E+01 1.01165 3.93222E+03 7.14114E+00 5.50544E+00 002 Elastic 6.62578E+00 7.43310E+00 1.12184 2.06033E+03 4.67231E+00 3.07960E+00 102 n,gamma 8.39568E+01 8.42050E+01 1.00295 1.87188E+03 5.55173E-02 7.33680E-10 Material(MAT)=7240:Hf-179 001 Total 4.85817E+01 4.94760E+01 1.01841 7.78365E+02 7.12497E+00 5.40086E+00 002 Elastic 7.79444E+00 8.77390E+00 1.12567 2.70093E+02 4.64117E+00 2.93616E+00 102 n,gamma 4.07873E+01 4.07020E+01 0.99791 5.08273E+02 1.05363E-01 7.51594E-08 Material(MAT)=7243:Hf-180 001 Total 3.54024E+01 3.82670E+01 1.08091 2.90435E+02 7.14400E+00 5.47607E+00 002 Elastic 2.23048E+01 2.51690E+01 1.12840 2.60839E+02 4.71468E+00 2.79554E+00 102 n,gamma 1.30976E+01 1.30980E+01 1.00003 2.95910E+01 3.28680E-02 4.12239E-06 Material(MAT)=7328:Ta-181 001 Total 2.63431E+01 2.71260E+01 1.02973 8.90488E+02 6.99496E+00 5.35709E+00 002 Elastic 5.66599E+00 6.38500E+00 1.12690 2.29957E+02 4.62119E+00 2.97818E+00 102 n,gamma 2.06771E+01 2.07410E+01 1.00310 6.59489E+02 8.74364E-02 2.05527E-07

Table 9. Simple integral neutron cross-section data from JEFF-3.1 calculated with the INTER computer code (cont.)

V Avg V G fact Avg res integ Avg V fiss V (E14) MT Reaction (2200 m/s) [1E-5 to 10] eV (Avg V)/V(E0) [5E-1 to 1E+5] eV [1E+3 to 2E+7] eV 1.4E+7 eV Material(MAT)=7331:Ta-182 001 Total 8.31941E+03 1.36640E+04 1.64241 1.32945E+03 6.78987E+00 5.36137E+00 002 Elastic 3.11518E+01 5.39930E+01 1.73321 2.86348E+02 3.76921E+00 2.74027E+00 102 n,gamma 8.28826E+03 1.36100E+04 1.64207 1.04310E+03 6.79110E-02 4.19999E-03 Material(MAT)=7431:W-182 001 Total 2.95817E+01 3.07720E+01 1.04022 1.10207E+03 6.80930E+00 5.19820E+00 002 Elastic 8.86836E+00 9.99330E+00 1.12685 4.73903E+02 4.40062E+00 2.54248E+00 102 n,gamma 2.07133E+01 2.07780E+01 1.00314 6.28168E+02 6.62525E-02 1.00365E-03 Material(MAT)=7434:W-183 001 Total 1.25024E+01 1.28210E+01 1.02548 8.00412E+02 6.82503E+00 5.20403E+00 002 Elastic 2.38810E+00 2.69340E+00 1.12785 4.65478E+02 3.93781E+00 2.84059E+00 102 n,gamma 1.01143E+01 1.01280E+01 1.00131 3.34524E+02 5.02757E-02 1.00000E-03 Material(MAT)=7437:W-184 001 Total 9.07071E+00 1.00190E+01 1.10457 3.07775E+02 6.73050E+00 5.21498E+00 002 Elastic 7.37288E+00 8.32090E+00 1.12859 2.91747E+02 4.28077E+00 2.53799E+00

122 118 102 n,gamma 1.69783E+00 1.69830E+00 1.00028 1.60286E+01 3.75970E-02 1.00382E-03 Material(MAT)=7443:W-186 001 Total 4.03905E+01 4.05810E+01 1.00472 3.57154E+03 6.72157E+00 5.21980E+00 002 Elastic 9.33799E-01 1.06260E+00 1.13791 3.04267E+03 4.28046E+00 2.80972E+00 102 n,gamma 3.94567E+01 3.95190E+01 1.00157 5.28861E+02 3.59600E-02 1.00418E-03 Material(MAT)=7525:Re-185 001 Total 1.21064E+02 1.22680E+02 1.01337 2.00825E+03 6.78826E+00 5.31250E+00 002 Elastic 8.89129E+00 9.95690E+00 1.11985 2.78889E+02 4.23946E+00 2.67870E+00 102 n,gamma 1.12172E+02 1.12720E+02 1.00492 1.72936E+03 1.81562E-01 1.10000E-03 Material(MAT)=7531:Re-187 001 Total 8.66662E+01 8.74780E+01 1.00937 5.30616E+02 6.78742E+00 5.32780E+00 002 Elastic 9.96233E+00 1.11900E+01 1.12321 2.37180E+02 4.18181E+00 2.69852E+00 102 n,gamma 7.67039E+01 7.62880E+01 0.99458 2.93436E+02 1.17055E-01 1.08200E-03 Material(MAT)=7600:Os-Nat 001 Total 3.13972E+01 3.32760E+01 1.05984 5.18558E+02 6.86504E+00 5.12511E+00 002 Elastic 1.54613E+01 1.74420E+01 1.12812 3.37977E+02 4.56026E+00 2.68762E+00 102 n,gamma 1.59359E+01 1.58340E+01 0.99359 1.80509E+02 7.70746E-02 1.00000E-03 Material(MAT)=7725:Ir-191 001 Total 9.68343E+02 9.66440E+02 0.99803 3.86285E+03 6.66963E+00 5.21126E+00 002 Elastic 1.39551E+01 1.52040E+01 1.08949 3.06755E+02 4.01167E+00 2.79307E+00 102 n,gamma 9.54388E+02 9.51240E+02 0.99670 3.55608E+03 1.85196E-01 6.08002E-03

Table 9. Simple integral neutron cross-section data from JEFF-3.1 calculated with the INTER computer code (cont.)

V Avg V G fact Avg res integ Avg V fiss V (E14) MT Reaction (2200 m/s) [1E-5 to 10] eV (Avg V)/V(E0) [5E-1 to 1E+5] eV [1E+3 to 2E+7] eV 1.4E+7 eV Material(MAT)=7731:Ir-193 001 Total 1.31430E+02 1.35740E+02 1.03279 1.67601E+03 6.67738E+00 5.21157E+00 002 Elastic 1.93851E+01 2.18050E+01 1.12482 3.00700E+02 4.13909E+00 2.79672E+00 102 n,gamma 1.12044E+02 1.13930E+02 1.01687 1.37527E+03 9.15473E-02 2.75201E-03 Material(MAT)=7800:Pt-Nat 001 Total 2.00324E+01 2.15580E+01 1.07617 2.69995E+02 6.26322E+00 5.36706E+00 002 Elastic 1.20310E+01 1.35790E+01 1.12870 1.58395E+02 4.56523E+00 3.00000E+00 102 n,gamma 8.00146E+00 7.97890E+00 0.99718 1.11600E+02 6.77970E-02 3.96946E-04 Material(MAT)=7925:Au-197 001 Total 1.05592E+02 1.06960E+02 1.01293 1.93816E+03 6.65283E+00 5.30290E+00 002 Elastic 6.85598E+00 7.71680E+00 1.12556 3.75048E+02 4.65321E+00 2.63617E+00 102 n,gamma 9.87362E+01 9.92400E+01 1.00511 1.56309E+03 7.79598E-02 1.32637E-03 Material(MAT)=8025:Hg-196 001 Total 3.18798E+03 3.16310E+03 0.99218 8.55166E+02 6.64303E+00 5.20526E+00 002 Elastic 1.09816E+02 1.21810E+02 1.10922 4.34881E+02 5.16363E+00 2.74304E+00

119 102 n,gamma 3.07817E+03 3.04120E+03 0.98801 4.20284E+02 6.15329E-03 1.45443E-08 Material(MAT)=8031:Hg-198 001 Total 1.47998E+01 1.64500E+01 1.11153 3.17602E+02 6.64053E+00 5.22630E+00 002 Elastic 1.28141E+01 1.44630E+01 1.12865 2.43293E+02 5.11554E+00 2.78936E+00 102 n,gamma 1.98569E+00 1.98780E+00 1.00105 7.43088E+01 3.18854E-02 1.21163E-07 Material(MAT)=8034:Hg-199 001 Total 2.21631E+03 2.20040E+03 0.99280 8.02529E+02 6.68114E+00 5.24483E+00 002 Elastic 6.66964E+01 7.40670E+01 1.11051 3.63836E+02 4.64541E+00 2.77126E+00 102 n,gamma 2.14962E+03 2.12630E+03 0.98915 4.38692E+02 3.05981E-02 1.96322E-07 Material(MAT)=8037:Hg-200 001 Total 1.60035E+01 1.78780E+01 1.11711 2.06389E+02 6.65760E+00 5.25734E+00 002 Elastic 1.45605E+01 1.64340E+01 1.12870 2.03938E+02 5.08095E+00 2.75192E+00 102 n,gamma 1.44302E+00 1.44330E+00 1.00021 2.45097E+00 2.15159E-02 2.57438E-07 Material(MAT)=8040:Hg-201 001 Total 2.22817E+01 2.41460E+01 1.08367 2.58054E+02 6.72394E+00 5.26926E+00 002 Elastic 1.44977E+01 1.63630E+01 1.12864 2.21754E+02 4.60320E+00 2.78757E+00 102 n,gamma 7.78400E+00 7.78330E+00 0.99992 3.43704E+01 2.02538E-02 4.21484E-07 Material(MAT)=8043:Hg-202 001 Total 1.95614E+01 2.14410E+01 1.09611 1.83390E+02 6.66376E+00 5.27444E+00 002 Elastic 1.46063E+01 1.64860E+01 1.12868 1.80269E+02 5.11331E+00 2.81855E+00 102 n,gamma 4.95501E+00 4.95560E+00 1.00012 3.12131E+00 1.36221E-02 4.76889E-07

Table 9. Simple integral neutron cross-section data from JEFF-3.1 calculated with the INTER computer code (cont.)

V Avg V G fact Avg res integ Avg V fiss V (E14) MT Reaction (2200 m/s) [1E-5 to 10] eV (Avg V)/V(E0) [5E-1 to 1E+5] eV [1E+3 to 2E+7] eV 1.4E+7 eV Material(MAT)=8049:Hg-204 001 Total 2.98988E+01 3.37440E+01 1.12859 3.11622E+02 6.67551E+00 5.30762E+00 002 Elastic 2.94673E+01 3.32650E+01 1.12888 3.08942E+02 5.32461E+00 2.84798E+00 102 n,gamma 4.31436E-01 4.78530E-01 1.10916 2.67972E+00 8.64935E-03 2.91008E-07 Material(MAT)=8100:Tl-Nat 001 Total 1.36267E+01 1.56630E+01 1.14944 2.14424E+02 6.58845E+00 5.17003E+00 002 Elastic 1.00346E+01 1.13260E+01 1.12870 2.00676E+02 4.86679E+00 2.58451E+00 102 n,gamma 3.59208E+00 4.33710E+00 1.20740 1.37484E+01 1.25110E-02 9.00357E-04 Material(MAT)=8225:Pb-204 001 Total 1.18849E+01 1.33300E+01 1.12156 1.45504E+02 6.36148E+00 5.23565E+00 002 Elastic 1.12242E+01 1.26690E+01 1.12870 1.43727E+02 5.17991E+00 2.72486E+00 102 n,gamma 6.60683E-01 6.60870E-01 1.00028 1.77749E+00 1.21825E-02 2.26710E-03 Material(MAT)=8231:Pb-206 001 Total 1.12938E+01 1.27430E+01 1.12836 1.29231E+02 6.18956E+00 5.26210E+00 002 Elastic 1.12640E+01 1.27140E+01 1.12870 1.29140E+02 5.04027E+00 2.74370E+00

120 102 n,gamma 2.97897E-02 2.97960E-02 1.00022 9.12451E-02 3.00133E-03 1.10198E-03 Material(MAT)=8234:Pb-207 001 Total 1.14855E+01 1.28720E+01 1.12074 1.32088E+02 6.48572E+00 5.27379E+00 002 Elastic 1.07735E+01 1.21600E+01 1.12870 1.31710E+02 5.41224E+00 2.75159E+00 102 n,gamma 7.11997E-01 7.12170E-01 1.00025 3.77978E-01 2.36486E-03 1.33946E-03 Material(MAT)=8237:Pb-208 001 Total 1.13994E+01 1.28660E+01 1.12869 1.38569E+02 6.48581E+00 5.39161E+00 002 Elastic 1.13989E+01 1.28660E+01 1.12870 1.38568E+02 6.11461E+00 2.82321E+00 102 n,gamma 4.92088E-04 4.92200E-04 1.00024 1.10300E-03 8.65541E-04 1.19036E-03 Material(MAT)=8325:Bi-209 001 Total 9.35393E+00 1.05530E+01 1.12822 1.44197E+02 6.50480E+00 5.33433E+00 002 Elastic 9.31994E+00 1.05190E+01 1.12869 1.44007E+02 5.78545E+00 2.73943E+00 102 n,gamma 3.38443E-02 3.38520E-02 1.00023 1.91082E-01 3.12345E-03 1.60094E-03 Material(MAT)=8825:Ra-223 001 Total 1.43132E+02 1.44770E+02 1.01143 8.70886E+02 7.55826E+00 6.17677E+00 002 Elastic 1.24280E+01 1.40270E+01 1.12870 4.33546E+02 5.24637E+00 3.11688E+00 018 Fission 7.00023E-01 7.00210E-01 1.00027 7.66372E-01 5.56997E-02 5.57000E-02 102 n,gamma 1.30004E+02 1.30040E+02 1.00028 4.36318E+02 4.61810E-02 3.75182E-06

Table 9. Simple integral neutron cross-section data from JEFF-3.1 calculated with the INTER computer code (cont.)

V Avg V G fact Avg res integ Avg V fiss V (E14) MT Reaction (2200 m/s) [1E-5 to 10] eV (Avg V)/V(E0) [5E-1 to 1E+5] eV [1E+3 to 2E+7] eV 1.4E+7 eV Material(MAT)=8828:Ra-224 001 Total 2.45285E+01 2.61440E+01 1.06587 3.52674E+02 7.55845E+00 6.13193E+00 002 Elastic 1.25281E+01 1.41400E+01 1.12870 3.23764E+02 5.51840E+00 3.12072E+00 102 n,gamma 1.20004E+01 1.20040E+01 1.00027 2.88986E+01 3.15041E-02 2.74345E-06 Material(MAT)=8831:Ra-225 001 Total 1.12426E+02 1.14010E+02 1.01406 9.84561E+02 7.57423E+00 6.14293E+00 002 Elastic 1.24277E+01 1.40270E+01 1.12870 3.89302E+02 5.27561E+00 3.12460E+00 102 n,gamma 9.99983E+01 9.99800E+01 0.99981 5.94560E+02 5.06573E-02 1.57332E-06 Material(MAT)=8834:Ra-226 001 Total 2.26138E+01 2.46960E+01 1.09207 4.95187E+02 7.57081E+00 6.15378E+00 002 Elastic 9.82243E+00 1.10660E+01 1.12658 2.13272E+02 5.51589E+00 3.12828E+00 018 Fission 6.99996E-06 6.99940E-06 0.99992 5.66293E-06 3.49115E-04 1.15001E-02 102 n,gamma 1.27913E+01 1.36300E+01 1.06558 2.81874E+02 1.08481E-01 3.21174E-05 Material(MAT)=8925:Ac-225 001 Total 1.01246E+03 1.01430E+03 1.00183 1.86205E+03 7.56480E+00 6.14287E+00

121 002 Elastic 1.24277E+01 1.40270E+01 1.12870 2.71348E+02 5.23432E+00 3.12460E+00 102 n,gamma 1.00003E+03 1.00030E+03 1.00025 1.59042E+03 1.71229E-01 3.76429E-06 Material(MAT)=8928:Ac-226 001 Total 1.12431E+02 1.14060E+02 1.01450 1.91552E+03 7.56141E+00 5.61669E+00 002 Elastic 1.24276E+01 1.40270E+01 1.12869 2.33384E+02 5.21479E+00 3.17445E+00 102 n,gamma 1.00003E+02 1.00030E+02 1.00031 1.68213E+03 1.75222E-01 5.72558E-09 Material(MAT)=8931:Ac-227 001 Total 9.14049E+02 1.04340E+03 1.14148 1.98721E+03 7.62148E+00 6.17021E+00 002 Elastic 1.24275E+01 1.40270E+01 1.12870 3.28913E+02 5.31166E+00 3.13747E+00 018 Fission 2.90010E-04 2.90100E-04 1.00030 1.76322E-04 1.26572E-02 1.24000E-01 102 n,gamma 9.01621E+02 1.02930E+03 1.14166 1.65734E+03 3.55656E-03 1.38926E-09 Material(MAT)=9025:Th-227 001 Total 1.74949E+03 1.75160E+03 1.00120 1.84870E+03 7.58441E+00 5.62352E+00 002 Elastic 1.24275E+01 1.40270E+01 1.12869 2.24342E+02 5.23137E+00 3.17291E+00 018 Fission 2.02007E+02 2.02060E+02 1.00029 2.06325E+02 4.83964E-01 2.05000E+00 102 n,gamma 1.53505E+03 1.53550E+03 1.00028 1.41797E+03 1.97199E-01 5.01486E-08 Material(MAT)=9028:Th-228 001 Total 1.33008E+02 1.35040E+02 1.01526 1.58426E+03 7.58646E+00 5.64353E+00 002 Elastic 1.28332E+01 1.44550E+01 1.12639 4.14912E+02 5.47683E+00 3.18439E+00 018 Fission 3.00009E-01 3.00080E-01 1.00025 1.33314E-01 1.05035E-01 5.49001E-01 102 n,gamma 1.19875E+02 1.20280E+02 1.00340 1.16914E+03 8.03674E-02 2.40360E-10

Table 9. Simple integral neutron cross-section data from JEFF-3.1 calculated with the INTER computer code (cont.)

V Avg V G fact Avg res integ Avg V fiss V (E14) MT Reaction (2200 m/s) [1E-5 to 10] eV (Avg V)/V(E0) [5E-1 to 1E+5] eV [1E+3 to 2E+7] eV 1.4E+7 eV Material(MAT)=9031:Th-229 001 Total 1.04164E+02 1.08670E+02 1.04326 1.85545E+03 7.62080E+00 6.18694E+00 002 Elastic 9.95927E+00 1.11750E+01 1.12203 1.77560E+02 5.15630E+00 3.13947E+00 018 Fission 3.08209E+01 3.13550E+01 1.01733 4.40460E+02 4.83964E-01 2.05000E+00 102 n,gamma 6.33838E+01 6.61400E+01 1.04349 1.23525E+03 2.09081E-01 7.86926E-06 Material(MAT)=9034:Th-230 001 Total 2.91775E+01 3.00750E+01 1.03075 1.22375E+03 8.45268E+00 6.30469E+00 002 Elastic 6.09280E+00 6.85070E+00 1.12440 3.72177E+02 5.39743E+00 3.52056E+00 018 Fission 0.00000E+00 0.00000E+00 0.00000 4.80897E-03 1.79644E-01 6.49999E-01 102 n,gamma 2.30847E+01 2.32240E+01 1.00603 8.51349E+02 1.99949E-02 1.10289E-02 Material(MAT)=9040:Th-232 001 Total 2.03780E+01 2.19990E+01 1.07955 3.03115E+02 7.77239E+00 5.84382E+00 002 Elastic 1.29746E+01 1.46070E+01 1.12582 2.17304E+02 4.81953E+00 2.75871E+00 018 Fission 0.00000E+00 0.00000E+00 0.00000 3.30946E-06 7.79584E-02 3.45946E-01 102 n,gamma 7.40345E+00 7.39220E+00 0.99848 8.56465E+01 9.01425E-02 1.00002E-03

122 Material(MAT)=9043:Th-233 001 Total 1.47808E+03 1.48020E+03 1.00145 8.61895E+02 7.63938E+00 5.67011E+00 002 Elastic 1.30281E+01 1.47080E+01 1.12891 2.07301E+02 5.33693E+00 3.16731E+00 018 Fission 1.50042E+01 1.50600E+01 1.00373 1.02684E+01 1.11388E-01 4.09999E-01 102 n,gamma 1.45005E+03 1.45050E+03 1.00028 6.43523E+02 8.86580E-02 6.07111E-09 Material(MAT)=9046:Th-234 001 Total 1.47780E+01 1.64550E+01 1.11350 3.86105E+02 7.67125E+00 5.67955E+00 002 Elastic 1.30280E+01 1.47050E+01 1.12873 2.92502E+02 5.54452E+00 3.16799E+00 018 Fission 0.00000E+00 0.00000E+00 0.00000 0.00000E+00 3.62950E-02 1.50000E-01 102 n,gamma 1.75003E+00 1.75030E+00 1.00014 9.34817E+01 1.02234E-01 1.54057E-07 Material(MAT)=9131:Pa-231 001 Total 2.35368E+02 2.35050E+02 0.99864 7.17568E+02 8.45151E+00 6.42764E+00 002 Elastic 8.45702E+00 9.39680E+00 1.11112 1.22088E+02 4.77976E+00 2.77152E+00 018 Fission 1.03869E-02 1.05460E-02 1.01533 1.20888E-01 9.81066E-01 1.97174E+00 102 n,gamma 2.26901E+02 2.25640E+02 0.99445 5.94653E+02 3.93544E-01 1.12225E-01 Material(MAT)=9134:Pa-232 001 Total 1.76216E+03 1.83220E+03 1.03976 1.21376E+03 7.68521E+00 5.67482E+00 002 Elastic 3.27448E+01 3.66970E+01 1.12069 2.03149E+02 5.10827E+00 3.18163E+00 018 Fission 1.51732E+03 1.57180E+03 1.03594 8.64233E+02 1.10250E+00 1.38100E+00 102 n,gamma 2.12095E+02 2.23680E+02 1.05464 1.46324E+02 7.41860E-02 5.62267E-09

Table 9. Simple integral neutron cross-section data from JEFF-3.1 calculated with the INTER computer code (cont.)

V Avg V G fact Avg res integ Avg V fiss V (E14) MT Reaction (2200 m/s) [1E-5 to 10] eV (Avg V)/V(E0) [5E-1 to 1E+5] eV [1E+3 to 2E+7] eV 1.4E+7 eV Material(MAT)=9137:Pa-233 001 Total 4.98816E+01 5.00990E+01 1.00436 9.89572E+02 7.32459E+00 5.42088E+00 002 Elastic 8.43156E+00 9.49140E+00 1.12570 1.32624E+02 4.50493E+00 3.01434E+00 018 Fission 0.00000E+00 0.00000E+00 0.00000 0.00000E+00 4.60094E-01 1.58187E+00 102 n,gamma 4.14500E+01 4.06080E+01 0.97968 8.54733E+02 1.86990E-01 6.66667E-03 Material(MAT)=9219:U-232 001 Total 1.57196E+02 1.54300E+02 0.98159 8.75547E+02 8.58339E+00 6.38921E+00 002 Elastic 7.54020E+00 8.46640E+00 1.12283 1.43146E+02 4.71590E+00 3.01517E+00 018 Fission 7.70858E+01 7.52960E+01 0.97679 4.15894E+02 2.48405E+00 1.23814E+00 102 n,gamma 7.25705E+01 7.05390E+01 0.97201 3.16405E+02 1.04713E-01 5.60544E-02 Material(MAT)=9222:U-233 001 Total 5.88496E+02 5.89330E+02 1.00141 1.04193E+03 7.68530E+00 5.91376E+00 002 Elastic 1.19929E+01 1.35040E+01 1.12599 1.41505E+02 4.51252E+00 3.01951E+00 018 Fission 5.31236E+02 5.29160E+02 0.99609 7.62395E+02 1.91137E+00 2.31117E+00 102 n,gamma 4.52673E+01 4.66630E+01 1.03084 1.37969E+02 7.27508E-02 7.47467E-04

123 Material(MAT)=9225:U-234 001 Total 1.19273E+02 1.20580E+02 1.01100 9.39733E+02 7.80521E+00 5.86053E+00 002 Elastic 1.94487E+01 2.17580E+01 1.11876 3.07807E+02 4.70804E+00 2.82329E+00 018 Fission 6.69642E-02 6.63140E-02 0.99030 8.21712E-01 1.17046E+00 1.97837E+00 102 n,gamma 9.97569E+01 9.87590E+01 0.99000 6.30902E+02 1.11551E-01 1.00005E-03 Material(MAT)=9228:U-235 001 Total 6.98719E+02 6.85930E+02 0.98169 5.52906E+02 7.65918E+00 5.87234E+00 002 Elastic 1.51153E+01 1.69400E+01 1.12069 1.44079E+02 4.41418E+00 2.85230E+00 018 Fission 5.84932E+02 5.71190E+02 0.97651 2.68845E+02 1.21851E+00 2.06000E+00 102 n,gamma 9.86724E+01 9.77930E+01 0.99109 1.39845E+02 9.53408E-02 1.20890E-03 Material(MAT)=9231:U-236 001 Total 1.37120E+01 1.47920E+01 1.07873 6.02388E+02 7.69863E+00 5.86721E+00 002 Elastic 8.35453E+00 9.42350E+00 1.12795 2.53163E+02 4.82887E+00 2.78350E+00 018 Fission 6.12983E-02 6.14330E-02 1.00220 4.33526E+00 5.94222E-01 1.63115E+00 102 n,gamma 5.29616E+00 5.30660E+00 1.00198 3.44749E+02 7.61704E-02 2.55198E-03 Material(MAT)=9234:U-237 001 Total 4.78448E+02 4.70180E+02 0.98272 1.32922E+03 8.11983E+00 5.83283E+00 002 Elastic 2.44354E+01 2.68690E+01 1.09958 2.00806E+02 4.59660E+00 2.70663E+00 018 Fission 1.70230E+00 1.67320E+00 0.98292 4.43703E+01 8.86875E-01 1.53306E+00 102 n,gamma 4.52310E+02 4.41640E+02 0.97641 1.08304E+03 5.63979E-02 2.53830E-03

Table 9. Simple integral neutron cross-section data from JEFF-3.1 calculated with the INTER computer code (cont.)

V Avg V G fact Avg res integ Avg V fiss V (E14) MT Reaction (2200 m/s) [1E-5 to 10] eV (Avg V)/V(E0) [5E-1 to 1E+5] eV [1E+3 to 2E+7] eV 1.4E+7 eV Material(MAT)=9237:U-238 001 Total 1.21209E+01 1.33370E+01 1.10033 5.92421E+02 7.90121E+00 5.95861E+00 002 Elastic 9.43722E+00 1.06480E+01 1.12828 3.17285E+02 4.92815E+00 2.77514E+00 018 Fission 2.65124E-05 2.65390E-05 1.00100 1.21666E-03 3.01253E-01 1.17879E+00 102 n,gamma 2.68363E+00 2.68910E+00 1.00205 2.74955E+02 6.98823E-02 2.51665E-03 Material(MAT)=9340:Np-235 001 Total 1.81433E+02 1.82980E+02 1.00854 1.05418E+03 7.42042E+00 5.92372E+00 002 Elastic 1.14244E+01 1.28950E+01 1.12872 1.65720E+02 4.39513E+00 2.97160E+00 018 Fission 1.99985E+01 1.99840E+01 0.99926 3.68405E+01 1.88716E+00 2.44871E+00 102 n,gamma 1.50010E+02 1.50100E+02 1.00063 8.51438E+02 1.21911E-01 7.37811E-04 Material(MAT)=9343:Np-236 001 Total 3.48303E+03 3.48320E+03 1.00004 1.42780E+03 7.76557E+00 5.84944E+00 002 Elastic 1.22963E+01 1.38790E+01 1.12872 1.45609E+02 5.08571E+00 3.14800E+00 018 Fission 2.76978E+03 2.76860E+03 0.99959 1.02340E+03 1.92231E+00 2.43181E+00 102 n,gamma 7.00944E+02 7.00640E+02 0.99956 2.58791E+02 1.89506E-01 1.00377E-03

129 124 Material(MAT)=9346:Np-237 001 Total 1.75802E+02 1.75010E+02 0.99547 8.31491E+02 7.85367E+00 5.78427E+00 002 Elastic 1.40851E+01 1.57320E+01 1.11691 1.69672E+02 4.65189E+00 2.78498E+00 018 Fission 2.03650E-02 1.98750E-02 0.97593 6.36694E-01 1.33431E+00 2.15590E+00 102 n,gamma 1.61697E+02 1.59250E+02 0.98490 6.60804E+02 1.58595E-01 1.07209E-03 Material(MAT)=9349:Np-238 001 Total 2.25073E+03 2.23270E+03 0.99197 1.15393E+03 7.84587E+00 6.05519E+00 002 Elastic 2.08548E+01 2.32930E+01 1.11691 1.42947E+02 4.86251E+00 2.91457E+00 018 Fission 2.02703E+03 2.00830E+03 0.99078 9.10283E+02 1.48823E+00 1.85000E+00 102 n,gamma 2.02846E+02 2.01020E+02 0.99099 1.00612E+02 3.39726E-02 1.06197E-02 Material(MAT)=9352:Np-239 001 Total 8.75253E+01 8.89010E+01 1.01571 6.17232E+02 7.85505E+00 5.53495E+00 002 Elastic 1.05221E+01 1.18760E+01 1.12871 1.60415E+02 4.75175E+00 2.65197E+00 018 Fission 0.00000E+00 0.00000E+00 0.00000 4.12331E-02 1.44754E+00 2.42000E+00 102 n,gamma 7.70032E+01 7.70240E+01 1.00027 4.56264E+02 2.02180E-01 1.00005E-05 Material(MAT)=9428:Pu-236 001 Total 2.05227E+02 2.05470E+02 1.00118 1.37044E+03 7.88166E+00 5.95089E+00 002 Elastic 9.17836E+00 1.03360E+01 1.12613 1.54037E+02 4.47911E+00 2.75396E+00 018 Fission 1.64821E+02 1.63890E+02 0.99438 9.49574E+02 2.09937E+00 2.54773E+00 102 n,gamma 3.12278E+01 3.12390E+01 1.00037 2.66711E+02 6.22997E-02 8.02168E-04

Table 9. Simple integral neutron cross-section data from JEFF-3.1 calculated with the INTER computer code (cont.)

V Avg V G fact Avg res integ Avg V fiss V (E14) MT Reaction (2200 m/s) [1E-5 to 10] eV (Avg V)/V(E0) [5E-1 to 1E+5] eV [1E+3 to 2E+7] eV 1.4E+7 eV Material(MAT)=9431:Pu-237 001 Total 2.65184E+03 2.65410E+03 1.00084 1.40075E+03 7.56532E+00 6.42911E+00 002 Elastic 1.17589E+01 1.32730E+01 1.12872 1.39822E+02 4.25632E+00 3.32070E+00 018 Fission 2.10007E+03 2.10060E+03 1.00027 1.07056E+03 3.00804E+00 2.69321E+00 102 n,gamma 5.40018E+02 5.40170E+02 1.00028 1.90351E+02 3.18665E-02 2.40001E-03 Material(MAT)=9434:Pu-238 001 Total 5.86500E+02 5.65140E+02 0.96358 4.11853E+02 8.81707E+00 6.86876E+00 002 Elastic 2.85777E+01 3.14730E+01 1.10132 2.34842E+02 6.01833E+00 3.58387E+00 018 Fission 1.78811E+01 1.71190E+01 0.95736 2.74921E+01 1.98604E+00 2.71459E+00 102 n,gamma 5.40041E+02 5.16550E+02 0.95650 1.49374E+02 9.82352E-02 5.86100E-09 Material(MAT)=9437:Pu-239 001 Total 1.02646E+03 1.10730E+03 1.07879 6.28125E+02 7.80185E+00 5.79457E+00 002 Elastic 7.99042E+00 8.80680E+00 1.10217 1.53245E+02 4.61805E+00 2.75043E+00 018 Fission 7.47785E+02 7.88570E+02 1.05454 2.92961E+02 1.79099E+00 2.41800E+00 102 n,gamma 2.70686E+02 3.09960E+02 1.14509 1.81119E+02 5.50898E-02 2.00097E-03

125 Material(MAT)=9440:Pu-240 001 Total 2.88552E+02 2.96290E+02 1.02681 9.45309E+03 7.83441E+00 5.79421E+00 002 Elastic 2.66692E+00 2.82780E+00 1.06034 9.65766E+02 4.78578E+00 2.68371E+00 018 Fission 5.91596E-02 6.10930E-02 1.03269 3.35686E+00 1.34866E+00 2.14480E+00 102 n,gamma 2.85826E+02 2.93400E+02 1.02650 8.48383E+03 8.33478E-02 2.06826E-03 Material(MAT)=9443:Pu-241 001 Total 1.38638E+03 1.44550E+03 1.04265 8.93360E+02 7.86018E+00 5.81392E+00 002 Elastic 1.12591E+01 1.24770E+01 1.10815 1.52373E+02 5.19542E+00 3.30462E+00 018 Fission 1.01214E+03 1.05950E+03 1.04677 5.60857E+02 1.65052E+00 2.29800E+00 102 n,gamma 3.62975E+02 3.73540E+02 1.02912 1.80110E+02 1.20515E-01 1.04967E-07 Material(MAT)=9446:Pu-242 001 Total 2.71347E+01 2.83680E+01 1.04544 1.45376E+03 7.95644E+00 5.95370E+00 002 Elastic 8.33694E+00 9.38480E+00 1.12568 3.22976E+02 4.81846E+00 2.79060E+00 018 Fission 2.55742E-03 2.57220E-03 1.00579 2.53511E-01 1.15468E+00 2.08860E+00 102 n,gamma 1.87952E+01 1.89800E+01 1.00985 1.13023E+03 8.92200E-02 7.99999E-04 Material(MAT)=9449:Pu-243 001 Total 2.89010E+02 2.89070E+02 1.00021 9.88402E+02 7.86099E+00 5.69267E+00 002 Elastic 1.94913E+01 2.17790E+01 1.11740 1.60741E+02 4.64947E+00 2.59107E+00 018 Fission 1.81406E+02 1.79910E+02 0.99173 5.53202E+02 1.06513E+00 1.90000E+00 102 n,gamma 8.81122E+01 8.73850E+01 0.99175 2.74439E+02 4.39346E-02 1.60000E-03

Table 9. Simple integral neutron cross-section data from JEFF-3.1 calculated with the INTER computer code (cont.)

V Avg V G fact Avg res integ Avg V fiss V (E14) MT Reaction (2200 m/s) [1E-5 to 10] eV (Avg V)/V(E0) [5E-1 to 1E+5] eV [1E+3 to 2E+7] eV 1.4E+7 eV Material(MAT)=9452:Pu-244 001 Total 1.14745E+01 1.27080E+01 1.10747 2.95806E+02 7.83117E+00 5.43671E+00 002 Elastic 9.64444E+00 1.08820E+01 1.12833 1.90117E+02 4.89269E+00 3.07582E+00 018 Fission 0.00000E+00 0.00000E+00 0.00000 6.59113E-02 9.92188E-01 1.68905E+00 102 n,gamma 1.83003E+00 1.82550E+00 0.99752 1.05320E+02 1.76183E-02 2.80001E-03 Material(MAT)=9458:Pu-246 001 Total 8.10580E+02 8.09600E+02 0.99880 8.40216E+02 7.91779E+00 5.99026E+00 002 Elastic 1.08221E+01 1.22150E+01 1.12870 1.47986E+02 4.00224E+00 2.97763E+00 018 Fission 0.00000E+00 0.00000E+00 0.00000 6.48026E-04 7.34052E-01 1.64370E+00 102 n,gamma 7.99758E+02 7.97390E+02 0.99704 6.92103E+02 1.14833E+00 9.34958E-04 Material(MAT)=9543:Am-241 001 Total 6.62526E+02 6.73610E+02 1.01673 1.68875E+03 7.81430E+00 5.74149E+00 002 Elastic 1.22568E+01 1.32800E+01 1.08350 1.53352E+02 4.59900E+00 2.51013E+00 018 Fission 3.15054E+00 3.33380E+00 1.05816 1.07808E+01 1.36486E+00 2.67000E+00 102 n,gamma 6.47119E+02 6.57000E+02 1.01527 1.52449E+03 2.57275E-01 7.01920E-04

126 Material(MAT)=9546:Am-242 001 Total 2.32143E+03 2.43740E+03 1.04997 1.30915E+03 7.75437E+00 5.71219E+00 002 Elastic 7.67958E+00 8.66930E+00 1.12887 1.36205E+02 4.45838E+00 2.84035E+00 018 Fission 2.09476E+03 2.19950E+03 1.05001 9.86371E+02 1.75829E+00 2.17538E+00 102 n,gamma 2.18985E+02 2.29250E+02 1.04688 1.86528E+02 4.74753E-02 7.33552E-04 Material(MAT)=9547:Am-242M 001 Total 7.63436E+03 8.36870E+03 1.09619 1.89817E+03 7.75507E+00 5.72723E+00 002 Elastic 5.26670E+00 6.57790E+00 1.24895 1.26681E+02 4.35948E+00 2.51956E+00 018 Fission 6.39825E+03 7.00790E+03 1.09529 1.53266E+03 1.84471E+00 2.54984E+00 102 n,gamma 1.23084E+03 1.35420E+03 1.10024 2.38793E+02 7.53542E-02 7.33552E-04 Material(MAT)=9549:Am-243 001 Total 8.42904E+01 8.62350E+01 1.02307 1.95021E+03 7.72909E+00 5.70066E+00 002 Elastic 7.47918E+00 8.38580E+00 1.12122 1.61350E+02 4.47966E+00 2.81774E+00 018 Fission 8.13240E-02 8.23540E-02 1.01267 2.19959E+00 1.06590E+00 2.14333E+00 102 n,gamma 7.67299E+01 7.77670E+01 1.01352 1.78650E+03 2.26210E-01 8.84191E-04 Material(MAT)=9552:Am-244 001 Total 2.91156E+03 2.99270E+03 1.02785 1.70151E+03 7.82589E+00 5.93596E+00 002 Elastic 1.16463E+01 1.31450E+01 1.12870 1.37607E+02 4.56686E+00 2.81081E+00 018 Fission 2.29994E+03 2.36580E+03 1.02864 1.24956E+03 1.76323E+00 2.43779E+00 102 n,gamma 5.99978E+02 6.13710E+02 1.02289 3.14342E+02 3.39147E-01 1.77551E-07

Table 9. Simple integral neutron cross-section data from JEFF-3.1 calculated with the INTER computer code (cont.)

V Avg V G fact Avg res integ Avg V fiss V (E14) MT Reaction (2200 m/s) [1E-5 to 10] eV (Avg V)/V(E0) [5E-1 to 1E+5] eV [1E+3 to 2E+7] eV 1.4E+7 eV Material(MAT)=9553:Am-244M 001 Total 2.01169E+03 2.21740E+03 1.10226 1.70155E+03 7.82538E+00 5.93596E+00 002 Elastic 1.16457E+01 1.31450E+01 1.12870 1.37565E+02 4.52984E+00 2.81081E+00 018 Fission 1.60006E+03 1.76290E+03 1.10180 1.24961E+03 1.76323E+00 2.43779E+00 102 n,gamma 3.99984E+02 4.41320E+02 1.10334 3.14319E+02 3.70872E-01 1.12652E-06 Material(MAT)=9625:Cm-240 001 Total 1.98403E+02 1.95230E+02 0.98400 8.93147E+02 7.62454E+00 5.97487E+00 002 Elastic 1.31128E+01 1.46210E+01 1.11501 1.91895E+02 4.94067E+00 3.05862E+00 018 Fission 9.75210E+00 9.50570E+00 0.97473 3.68128E+01 1.70213E+00 2.56284E+00 102 n,gamma 1.75538E+02 1.71100E+02 0.97473 6.64223E+02 2.83989E-01 7.74123E-04 Material(MAT)=9628:Cm-241 001 Total 2.86193E+03 2.86420E+03 1.00080 1.42094E+03 7.49123E+00 6.00732E+00 002 Elastic 1.21052E+01 1.36630E+01 1.12870 1.41496E+02 4.41660E+00 3.15932E+00 018 Fission 2.59952E+03 2.60020E+03 1.00025 1.16694E+03 2.68056E+00 2.83800E+00 102 n,gamma 2.50311E+02 2.50390E+02 1.00030 1.12277E+02 2.35780E-02 0.00000E+00

127 Material(MAT)=9631:Cm-242 001 Total 3.25962E+01 3.39580E+01 1.04179 3.02031E+02 8.09505E+00 6.14814E+00 002 Elastic 1.16331E+01 1.31160E+01 1.12744 1.82767E+02 4.62613E+00 2.75081E+00 018 Fission 5.06396E+00 5.04420E+00 0.99611 1.11382E+01 1.64784E+00 2.67226E+00 102 n,gamma 1.58991E+01 1.57990E+01 0.99368 1.07896E+02 3.23300E-02 3.84774E-05 Material(MAT)=9634:Cm-243 001 Total 7.57660E+02 7.59050E+02 1.00183 1.89927E+03 7.99333E+00 6.27963E+00 002 Elastic 9.94587E+00 1.11020E+01 1.11622 1.47045E+02 4.44528E+00 2.89415E+00 018 Fission 6.17563E+02 6.20180E+02 1.00424 1.55416E+03 2.17436E+00 2.75658E+00 102 n,gamma 1.30151E+02 1.27770E+02 0.98168 1.98026E+02 1.93723E-02 6.95557E-05 Material(MAT)=9637:Cm-244 001 Total 1.79673E+01 1.88440E+01 1.04880 9.22156E+02 7.56185E+00 6.69383E+00 002 Elastic 6.99615E+00 7.88500E+00 1.12705 3.18499E+02 4.84835E+00 3.07582E+00 018 Fission 6.03671E-01 5.96880E-01 0.98875 1.08169E+01 1.59232E+00 3.08001E+00 102 n,gamma 1.03675E+01 1.03620E+01 0.99950 5.92593E+02 1.26891E-01 3.00001E-03 Material(MAT)=9640:Cm-245 001 Total 2.51088E+03 2.37760E+03 0.94693 1.03822E+03 7.88030E+00 5.77180E+00 002 Elastic 1.08990E+01 1.22000E+01 1.11934 1.38582E+02 4.63005E+00 2.72536E+00 018 Fission 2.14110E+03 2.02860E+03 0.94748 7.94141E+02 1.73924E+00 2.31345E+00 102 n,gamma 3.58885E+02 3.36780E+02 0.93841 1.05451E+02 8.55642E-02 9.59492E-04

Table 9. Simple integral neutron cross-section data from JEFF-3.1 calculated with the INTER computer code (cont.)

V Avg V G fact Avg res integ Avg V fiss V (E14) MT Reaction (2200 m/s) [1E-5 to 10] eV (Avg V)/V(E0) [5E-1 to 1E+5] eV [1E+3 to 2E+7] eV 1.4E+7 eV Material(MAT)=9643:Cm-246 001 Total 1.06829E+01 1.18760E+01 1.11172 2.91042E+02 7.92288E+00 5.71424E+00 002 Elastic 9.22731E+00 1.04130E+01 1.12854 1.71938E+02 4.84050E+00 2.72817E+00 018 Fission 1.44196E-01 1.44550E-01 1.00246 4.72170E+00 1.21329E+00 1.89824E+00 102 n,gamma 1.31135E+00 1.31830E+00 1.00533 1.13989E+02 1.07675E-01 1.00001E-03 Material(MAT)=9646:Cm-247 001 Total 1.47773E+02 1.46410E+02 0.99075 1.27586E+03 8.04082E+00 5.80739E+00 002 Elastic 8.79213E+00 9.87380E+00 1.12303 1.40990E+02 4.87505E+00 2.75132E+00 018 Fission 8.17765E+01 7.98630E+01 0.97661 6.00584E+02 1.90942E+00 2.53913E+00 102 n,gamma 5.72043E+01 5.66680E+01 0.99063 5.34238E+02 7.69887E-02 1.48236E-08 Material(MAT)=9649:Cm-248 001 Total 9.46845E+00 1.03120E+01 1.08913 5.87232E+02 8.07564E+00 5.82899E+00 002 Elastic 6.52770E+00 7.36510E+00 1.12829 3.15874E+02 5.37627E+00 2.76539E+00 018 Fission 3.70031E-01 3.70410E-01 1.00101 1.12763E+01 1.23979E+00 2.68000E+00 102 n,gamma 2.57072E+00 2.57690E+00 1.00239 2.59727E+02 5.44384E-02 1.04940E-09

128 Material(MAT)=9652:Cm-249 001 Total 2.11936E+01 2.23320E+01 1.05373 3.85445E+02 8.09977E+00 6.00847E+00 002 Elastic 9.20178E+00 1.03850E+01 1.12854 1.67732E+02 4.29340E+00 2.67397E+00 018 Fission 1.02332E+01 1.01890E+01 0.99570 1.55007E+02 2.10511E+00 2.36769E+00 102 n,gamma 1.75865E+00 1.75850E+00 0.99990 6.22590E+01 1.37328E-02 8.75960E-04 Material(MAT)=9655:Cm-250 001 Total 1.24948E+02 1.29730E+02 1.03828 1.11081E+03 8.14933E+00 6.00927E+00 002 Elastic 3.95898E+01 4.45380E+01 1.12498 8.06468E+02 4.93756E+00 2.71391E+00 018 Fission 2.08910E-03 2.09140E-03 1.00112 7.65082E-02 1.52016E+00 2.16040E+00 102 n,gamma 8.53565E+01 8.51910E+01 0.99806 3.03844E+02 2.36550E-02 4.75167E-04 Material(MAT)=9746:Bk-247 001 Total 2.88158E+02 2.81690E+02 0.97755 1.50215E+03 7.98469E+00 6.01006E+00 002 Elastic 1.23480E+01 1.36370E+01 1.10438 1.58482E+02 4.64499E+00 2.80160E+00 018 Fission 9.19367E+01 8.93510E+01 0.97187 4.47787E+02 9.97879E-01 2.25354E+00 102 n,gamma 1.83873E+02 1.78700E+02 0.97187 8.95275E+02 1.11018E-01 8.49091E-04 Material(MAT)=9752:Bk-249 001 Total 7.18874E+02 1.07870E+03 1.50049 1.29914E+03 8.10993E+00 5.81701E+00 002 Elastic 3.93473E+00 4.66360E+00 1.18525 1.68565E+02 5.07220E+00 2.74684E+00 018 Fission 3.97027E+00 5.95950E+00 1.50102 6.49433E+00 9.55483E-01 2.67000E+00 102 n,gamma 7.10969E+02 1.06800E+03 1.50224 1.12336E+03 1.58534E-01 2.29838E-09

Table 9. Simple integral neutron cross-section data from JEFF-3.1 calculated with the INTER computer code (cont.)

V Avg V G fact Avg res integ Avg V fiss V (E14) MT Reaction (2200 m/s) [1E-5 to 10] eV (Avg V)/V(E0) [5E-1 to 1E+5] eV [1E+3 to 2E+7] eV 1.4E+7 eV Material(MAT)=9755:Bk-250 001 Total 1.32407E+03 1.25280E+03 0.94615 8.45482E+02 8.15285E+00 5.82051E+00 002 Elastic 1.22451E+01 1.35090E+01 1.10324 1.40576E+02 4.83577E+00 2.75066E+00 018 Fission 9.58642E+02 9.05610E+02 0.94468 5.05696E+02 2.04094E+00 2.07649E+00 102 n,gamma 3.53184E+02 3.33650E+02 0.94468 1.98892E+02 3.24960E-02 8.02854E-09 Material(MAT)=9852:Cf-249 001 Total 2.17580E+03 2.10850E+03 0.96909 3.03201E+03 8.09670E+00 6.01349E+00 002 Elastic 6.23064E+00 6.76760E+00 1.08618 1.29952E+02 4.73579E+00 2.92032E+00 018 Fission 1.66528E+03 1.61280E+03 0.96846 2.20828E+03 1.73867E+00 2.28922E+00 102 n,gamma 5.04289E+02 4.89010E+02 0.96970 6.93734E+02 1.66645E-01 8.55210E-04 Material(MAT)=9855:Cf-250 001 Total 1.95119E+03 1.96020E+03 1.00463 9.18706E+03 8.15380E+00 5.83854E+00 002 Elastic 1.67689E+02 1.84650E+02 1.10115 7.49318E+02 5.28430E+00 2.75106E+00 018 Fission 4.09059E+00 4.07240E+00 0.99555 1.93633E+01 1.87399E+00 2.72133E+00 102 n,gamma 1.77941E+03 1.77150E+03 0.99555 8.41801E+03 9.96477E-02 4.96617E-07

129 Material(MAT)=9858:Cf-251 001 Total 8.19537E+03 8.17050E+03 0.99697 6.80460E+03 7.91474E+00 5.77377E+00 002 Elastic 1.21116E+01 1.38540E+01 1.14385 3.03977E+02 4.58164E+00 2.60142E+00 018 Fission 5.32159E+03 5.31190E+03 0.99817 4.88709E+03 1.71653E+00 2.40000E+00 102 n,gamma 2.86167E+03 2.84480E+03 0.99412 1.61349E+03 7.80842E-02 1.00000E-03 Material(MAT)=9861:Cf-252 001 Total 6.37779E+01 6.46940E+01 1.01437 3.29341E+02 7.84384E+00 5.74373E+00 002 Elastic 1.11084E+01 1.25070E+01 1.12591 1.71241E+02 4.72008E+00 2.60000E+00 018 Fission 3.21755E+01 3.18840E+01 0.99093 1.11029E+02 1.89635E+00 2.52929E+00 102 n,gamma 2.04940E+01 2.03040E+01 0.99072 4.70546E+01 4.18569E-02 8.00008E-04 Material(MAT)=9867:Cf-254 001 Total 1.71212E+01 1.84900E+01 1.07996 1.97600E+02 8.30810E+00 5.85843E+00 002 Elastic 1.06210E+01 1.19880E+01 1.12873 1.76926E+02 5.26532E+00 2.76645E+00 018 Fission 2.00005E+00 2.00050E+00 1.00022 1.38344E+01 2.14566E+00 2.64781E+00 102 n,gamma 4.50015E+00 4.50140E+00 1.00028 6.48603E+00 6.65454E-03 2.33501E-09 Material(MAT)=9913:Es-253 001 Total 2.10995E+02 2.22680E+02 1.05539 6.99419E+03 9.76273E+00 9.76000E+00 002 Elastic 9.81451E+00 1.10180E+01 1.12261 6.81419E+02 9.76231E+00 9.76000E+00 102 n,gamma 2.01180E+02 2.11660E+02 1.05211 6.31277E+03 4.23303E-04 0.00000E+00

Table 9. Simple integral neutron cross-section data from JEFF-3.1 calculated with the INTER computer code (cont.)

V Avg V G fact Avg res integ Avg V fiss V (E14) MT Reaction (2200 m/s) [1E-5 to 10] eV (Avg V)/V(E0) [5E-1 to 1E+5] eV [1E+3 to 2E+7] eV 1.4E+7 eV Material(MAT)=9914:Es-254 001 Total 2.00485E+03 2.00550E+03 1.00033 1.37294E+03 8.31076E+00 5.86267E+00 002 Elastic 1.06210E+01 1.19880E+01 1.12873 1.48113E+02 5.17089E+00 2.76645E+00 018 Fission 1.96606E+03 1.96660E+03 1.00028 1.20746E+03 2.18469E+00 2.22184E+00 102 n,gamma 2.81654E+01 2.69110E+01 0.95546 1.73685E+01 6.92355E-02 1.86848E-09 Material(MAT)=9916:Es-255 001 Total 7.90532E+01 8.04400E+01 1.01754 5.56108E+02 8.34678E+00 5.86913E+00 002 Elastic 1.06209E+01 1.19880E+01 1.12872 1.95205E+02 5.27618E+00 2.77075E+00 018 Fission 1.34305E+01 1.34350E+01 1.00032 8.25400E+01 2.23506E+00 2.75813E+00 102 n,gamma 5.50018E+01 5.50170E+01 1.00028 2.78050E+02 4.02055E-02 1.77280E-09 Material(MAT)=9936:Fm-255 001 Total 3.39673E+03 3.39910E+03 1.00069 1.43394E+03 8.35407E+00 5.87337E+00 002 Elastic 1.06209E+01 1.19880E+01 1.12872 1.69072E+02 5.13490E+00 2.77075E+00 018 Fission 3.36011E+03 3.36110E+03 1.00029 1.16353E+03 2.32759E+00 2.36716E+00 102 n,gamma 2.60008E+01 2.60080E+01 1.00028 1.01252E+02 2.03360E-02 4.04429E-09

136 130

APPENDIX 1

The choice of evaluation for the JEFF-3.1 Neutron General Purpose Library with comparison of the origin of the JEFF-3.0 evaluations is presented below, including some background information about the JEFF-3.1 project. This document was prepared by A.J. Koning, NRG Petten, during the compilation of the library.

Contents of General Purpose Library JEFF-3.1

This section discusses the selection of isotopes for the JEFF-3.1 neutron general purpose file. The procedure for this is reminiscent to that of the JEFF-3.0 library. The contents of the JEFF- 3.0T starter file were proposed in 1996 in a document by John Rowlands [row96]. For each nuclide for which evaluations, in any data library, were available, he proposed a best evaluation for JEFF-3.0, keeping in mind the most important application of each nuclide. According to John Rowlands, ”the judgement about ”best” is based on the date of the evaluation and its completeness”. The resulting selection was then discussed at the JEFF meeting and after several changes the JEFF-3.0T starter file was ready. What followed was a few years of solving format problems, testing and benchmarking, leading to the release of JEFF-3.0 in April 2002. A similar procedure has now been followed for the JEFF-3.1 data library. In most cases, the reasons outlined in [row96] to prefer one evaluation over the other are still valid. In other cases, new evaluations have appeared, and the choice was reconsidered. This section contains a table which specifies the data file for each isotope in JEFF-3.1, together with its justification. For the selection process, basically 4 rules were put into practice:

1. Include new European evaluations that have emerged since the release of JEFF-3.0.

2. As the default upgrade, adopt newer versions of evaluations that have been selected for JEFF-3.0: if for JEFF-3.0 an isotope was selected from e.g. JENDL-3.2, then adopt JENDL-3.3 for the same isotope in JEFF-3.1 (provided the JENDL-3.3 evaluation is an improvement!).

3. Choose source evaluations for JEFF-3.1 other than those of JEFF-3.0 if there are reasons to do so. This would overrule 2.

4. There were still 17 evaluated isotopes in other libraries that are not included in JEFF-3.0 (e.g. Nb-94,Fm-255). They are adopted.

The existing available data libraries in the world, JEFF-3.0, ENDF/B-VI.8, ENDF/B-VIIb0 (preliminary version), JENDL-3.3, EFF-2.4, EFF-3.0, BROND-2.2, and CENDL-2.1 have been scanned to come to the choice detailed in the table below. As additional guidelines, use has been made of the following references: row96 J. Rowlands, Proposed selection of data for JEFF-3, JEF/DOC-657; see also JEF/DOC- 664.695,741 dui02 M.C. Duijvestijn and A. Hogenbirk, MCNP libraries: Contents of the JEFF-3.0, ENDF/B- VI.8. and JENDL-3.3 based neutron cross-section libraries for MCNP, NRG report 02.50466 (2002)

1 ﬁs03 U. Fischer, Nuclear data for design analyses of the test blanket modules in ITER: Review and recommendations for EFF/JEFF evaluations, EFF-DOC-852 (2003). sub04 J.C. Sublet et al, JEFF-3.0 feedbacks and preJEFF-3.1 proposals, JEF/DOC-1011 bou04 O. Bouland, Motivation for new measurements on Am isotopes, JEF/DOC-1001. nou04 A. Nouri ”The JEFF-3.0 library”, JEFF Report 19.

The table below displays the resulting selection for the 381 materials (374 nuclides and 7 elements) that are included in JEFF-3.1.

Explanation of the table

The first column ’Isotope’ lists all isotopes included in JEFF-3.1. Note that it includes, as a sort of warning, the stable isotopes that do not exist in JEFF-3.1, nor in any other library, such as e.g. the Ne and Yb isotopes. The second column ’JEFF-3.0’ specifies the source of the evaluation adopted in JEFF-3.0. This is basically equal to ”Rowlands’ list” [row96]. The third column ’JEFF-3.1’ specifies the evaluation adopted in JEFF-3.1. If this column contains ’JEFF-3.0’ there is thus no change. The fourth column ’S’ specifies whether an isotope is a stable nuclide (S) or not (blank). The fifth column ’E’ specifies the maximum energy of the evaluation in MeV. The sixth column ’Justification/Comments’ gives a short justification for the upgrade for JEFF- 3.1 or other comments. The seventh column ’Change’ contains a ’*’ if a different source library than that of JEFF-3.0 is adopted (rule 3 above), an ’n’ if the isotope did not exist in JEFF-3.0 (rule 4) and an ’N’ if a new evaluation for JEFF-3.1 is made (rule 1). ¿From As (Z=33) to Er (Z=65), the choice is often accompanied the comment ”More recent evaluations available”. This generally concerns fission products. We still use JEFF-3.0 for these, although other evaluations (usually JENDL-3.3) are much more complete and more recent. Many of the FP evaluations of JEFF-3.0 (mostly taken from JEF-2.2, which are often based on ENDF/B-V) are however validated or even post-adjusted to various integral experiments (e.g. reactivity worth) such as STEK. Note that for a few FP’s (e.g. Cs-133), detailed studies have already led to clear recommendations, as indicated in the table below. For future releases of the JEFF library, the results of WPEC subgroup 23 on fission product evaluationsll serve as a helpful guideline.

2 Isotope JEFF-3.0 JEFF-3.1 S E Justiﬁcation/Comments Change H -001 ENDF/B-VI ENDF/B-VI.8 S 150 Extension to 150 MeV DIST-SEP91 Thermal capture cross section revised H -002 ENDF/B-VI.3 ENDF/B-VI.8 S 150 Extension to 150 MeV New evaluation H -003 CENDL-2 JEFF-3.0 S 20 - He-003 ENDF/B-VI.3 JEFF-3.0 S 20 - He-004 JENDL-3.2 JENDL-3.3 S 20 T-matrix deleted Li-006 ENDF/B-VI.3 JEFF-3.0 S 20 - Li-007 EFF-2.4 JEFF-3.0 S 20 Be-009 EFF-3.0 EFF-3.0 MOD6 S 20 Recent revision included B -010 ENDF/B-VI.3 JEFF-3.0 S 20 - B -011 ENDF/B-VI.3 ENDF/B-VI.8 S 20 Minor threshold revision compared to ENDF/B-VI.3 C -000 ENDF/B-VI.3 ENDF/B-VI.8 S 150 Extension to 150 MeV N -014 ENDF/B-VI.3 ENDF/B-VI.8 S 150 Gamma-ray spectra corrected Extension to 150 MeV N -015 ENDF/B-VI.3 JEFF-3.0 S 20 - O -016 JEF-2.2 ENDF/B-VI.8 S 150 Well validated, many new gamma data * New R-matrix, extension to 150 MeV O -017 JEF-2.2 JEFF-3.0 S 20 - O -018 - - S - - F -019 ENDF/B-VI.3 ENDF/B-VIIb0 S 20 Gamma production added negative probabilities removed New resonance evaluation Ne-020 - - S - - Ne-021 - - S - - Ne-022 - - S - - Na-022 JEF-2.2 JEFF-3.0 20 MF5 correction included for JEFF-3.1 Na-023 JEF-2.2 JEFF-3.0 S 20

3 Isotope JEFF-3.0 JEFF-3.1 S E Justiﬁcation/Comments Change Mg-024 JENDL-3.2:Mg-000 JENDL-3.3 S 20 Isotopic evaluation Mg-025 JENDL-3.2:Mg-000 JENDL-3.3 S 20 Isotopic evaluation Mg-026 JENDL-3.2:Mg-000 JENDL-3.3 S 20 Isotopic evaluation Al-027 ENDF/B-VI JEFF-3.0 S 150 - DIST-AUG99 + mod. Si-028 EFF-3.0 JEFF-3.0 S 20 - Si-029 JENDL-3.2 JENDL-3.3 S 20 Processing problems for ENDF/B-VI Si-030 JENDL-3.2 JENDL-3.3 S 20 Processing problems for ENDF/B-VI P -031 JENDL-3.2 JENDL-3.3 S 20 New calculated MT1,2 MT103 from JENDL-DF99 S -032 JENDL-3.2:S -000 JENDL-3.3 S 20 Isotopic evaluation S -033 JENDL-3.2:S -000 JENDL-3.3 S 20 Isotopic evaluation S -034 JENDL-3.2:S -000 JENDL-3.3 S 20 Isotopic evaluation S -036 JENDL-3.2:S -000 JENDL-3.3 S 20 Isotopic evaluation Cl-035 JEF-2.2:Cl-000 ENDF/B-VIIb0 S 20 Recent isotopic evaluation * New resonance parameters (ORNL) Cl-037 JEF-2.2:Cl-000 ENDF/B-VIIb0 S 20 Recent isotopic evaluation * New resonance parameters (ORNL) Ar-036 JEF-2.2 JEFF-3.0 S 20 MF5 correction included for JEFF-3.1 Ar-038 JEF-2.2 JEFF-3.0 S 20 MF5 correction included for JEFF-3.1 Ar-040 JEF-2.2 JENDL-3.3 S 20 New resonance data * More complete K -039 JENDL-3.2:K -000 JENDL-3.3 S 20 Isotopic evaluation Gamma production added K -040 JENDL-3.2:K -000 JENDL-3.3 S 20 Isotopic evaluation Gamma production added K -041 JENDL-3.2:K -000 JENDL-3.3 S 20 Isotopic evaluation Gamma production added Ca-040 JENDL-3.2:Ca-000 New evaluation S 200 NRG-2004 N Ca-042 JENDL-3.2:Ca-000 New evaluation S 200 NRG-2004 N Ca-043 JENDL-3.2:Ca-000 New evaluation S 200 NRG-2004 N Ca-044 JENDL-3.2:Ca-000 New evaluation S 200 NRG-2004 N Ca-046 JENDL-3.2:Ca-000 New evaluation S 200 NRG-2004 N Ca-048 JENDL-3.2:Ca-000 New evaluation S 200 NRG-2004 N Sc-045 ENDF/B-VI.3 New evaluation S 200 NRG-2004 N

4 Isotope JEFF-3.0 JEFF-3.1 S E Justiﬁcation/Comments Change Ti-046 JENDL-3.2:Ti-000 New evaluation S 20 IRK Vienna N Ti-047 JENDL-3.2:Ti-000 New evaluation S 20 IRK Vienna N Ti-048 JENDL-3.2:Ti-000 New evaluation S 20 IRK Vienna N Ti-049 JENDL-3.2:Ti-000 New evaluation S 20 IRK Vienna N Ti-050 JENDL-3.2:Ti-000 New evaluation S 20 IRK Vienna N V -000 EFF-3.0 JEFF-3.0 S 20 - Cr-050 ENDF/B-VI.3 JEFF-3.0 S 20 Cr-052 EFF-3.0 EFF-3.05 S 20 Modiﬁcations for continuum spectrum Cr-053 ENDF/B-VI.3 JEFF-3.0 S 20 Cr-054 ENDF/B-VI.3 JEFF-3.0 S 20 Mn-055 JENDL-3.2 JENDL-3.3 S 20 MF4,5 replaced by MF6 from JENDL-FF Photon production added Fe-054 ENDF/B-VI.3 New evaluation S 200 NRG-2004 N Fe-056 EFF-3.1 New evaluation S 200 EFF-3.1 ( < 20 MeV) + NRG-2004 N Fe-057 JEF-2.2 New evaluation S 200 NRG-2004 N Fe-058 JEF-2.2 New evaluation S 200 NRG-2004 + Moxon/Trkov res. data N Co-058 JEF-2.2 JEFF-3.0 20 MF5 correction included for JEFF-3.1 Co-058m JEF-2.2 JEFF-3.0 20 MF5 correction included for JEFF-3.1 Co-059 ENDF/B-VI.3 JEFF-3.0 S 20 - Ni-058 EFF-3.1 JEFF-3.0+corr. S 20 - Ni-059 JEF-2.2 JEFF-3.0+corr. 20 MF5 correction included for JEFF-3.1 Ni-060 EFF-3.1 JEFF-3.0+corr. S 20 Processing problem solved Ni-061 ENDF/B-VI.3 ENDF/B-VI.8 S 20 Ni-61,62,64 now up to 150 MeV Ni-062 ENDF/B-VI.3 ENDF/B-VI.8 S 20 - Ni-064 ENDF/B-VI.3 ENDF/B-VI.8 S 20 - Cu-063 ENDF/B-VI.3 ENDF/B-VI.8 S 150 Extension to 150 MeV Gamma-ray spectrum updated Cu-065 ENDF/B-VI.3 ENDF/B-VI.8 S 150 Extension to 150 MeV Gamma-ray spectrum updated

5 Isotope JEFF-3.0 JEFF-3.1 S E Justiﬁcation/Comments Change Zn-000 BROND-2.2 JEFF-3.0 S 20 - Ga-000 JENDL-3.2 JEFF-3.0 S 20 JENDL-3.3 isotopic evaluations exist, + JEF-2.2 but without gamma production Ge-070 JEF-2.2 New evaluation S 200 NRG-2004 N Ge-072 JEF-2.2 New evaluation S 200 NRG-2004 N Ge-073 JEF-2.2 New evaluation S 200 NRG-2004 N Ge-074 JEF-2.2 New evaluation S 200 NRG-2004 N Ge-076 JEF-2.2 New evaluation S 200 NRG-2004 N As-075 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Se-074 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Se-076 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Se-077 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Se-078 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Se-079 JENDL-3.2 JENDL-3.3 20 T-matrix deleted Se-080 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Se-082 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Br-079 JEF-2.2 JEFF-3.0 S 20 - Br-081 JEF-2.2 JEFF-3.0 S 20 - Kr-078 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Kr-080 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Kr-082 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Kr-083 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Kr-084 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Kr-085 JEF-2.2 JEFF-3.0 20 More recent evaluations available Kr-086 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Rb-085 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Rb-086 JEF-2.2 JEFF-3.0 20 - Rb-087 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Sr-084 JEF-2.2 JEFF-3.0 S 20 - Sr-086 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Sr-087 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Sr-088 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Sr-089 JEF-2.2 JEFF-3.0 20 More recent evaluations available Sr-090 JEF-2.2 JEFF-3.0 20 More recent evaluations available Y -089 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Y -090 JEF-2.2 JEFF-3.0 20 - Y -091 JEF-2.2 JEFF-3.0 20 More recent evaluations available

6 Isotope JEFF-3.0 JEFF-3.1 S E Justiﬁcation/Comments Change Zr-090 JENDL-3.2 JENDL-3.3 S 20 MF4,5 replaced by MF6 from JENDL-FF Photon prod. added, MT1,2,102 revised Zr-091 JENDL-3.2 JENDL-3.3 S 20 MF4,5 replaced by MF6 from JENDL-FF Photon prod. added, MT1,2,102 revised Zr-092 JENDL-3.2 JENDL-3.3 S 20 MF4,5 replaced by MF6 from JENDL-FF Photon prod. added, MT1,2,102 revised Zr-093 JEF-2.2 JEFF-3.0 20 More complete JENDL-3.3 available Zr-094 JENDL-3.2 JENDL-3.3 S 20 MF4,5 replaced by MF6 from JENDL-FF Photon prod. added, MT1,2,102 revised Zr-095 JEF-2.2 JEFF-3.0 20 More complete JENDL-3.3 available Zr-096 JENDL-3.2 JENDL-3.3 S 20 MF4,5 replaced by MF6 from JENDL-FF Photon prod. added, MT1,2,102 revised Nb-093 ENDF/B-VI.3 ENDF/B-VI.8 S 150 Extension to 150 MeV Nb-094 - JENDL-3.3 20 More complete than ENDF/B-V (JEF-2.2) n Nb-095 - JENDL-3.3 20 More complete than ENDF/B-V (JEF-2.2) n Mo-092 JENDL-3.2 JENDL-3.3 S 20 T-matrix deleted Gamma production data added Mo-094 JENDL-3.2 JENDL-3.3 S 20 T-matrix deleted Gamma production data added Mo-095 JENDL-3.2 ENDF/B-VIIb0 S 20 Best resonance integral * Mo-096 JENDL-3.2 JENDL-3.3 S 20 T-matrix deleted Gamma production data added Mo-097 JENDL-3.2 JENDL-3.3 S 20 T-matrix deleted Gamma production data added Mo-098 JENDL-3.2 JENDL-3.3 S 20 T-matrix deleted Gamma production data added Mo-099 JEF-2.2 JENDL-3.3 S 20 Much more complete evaluation * Consistent with other isotopes Mo-100 JENDL-3.2 JENDL-3.3 S 20 T-matrix deleted Gamma production data added

7 Isotope JEFF-3.0 JEFF-3.1 S E Justiﬁcation/Comments Change Tc-099 JENDL-3.2 New evaluation 20 CEA/NRG evaluation N Ru-096 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Ru-098 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Ru-099 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Ru-100 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Ru-101 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Ru-102 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Ru-103 JEF-2.2 JEFF-3.0 20 More recent evaluations available Ru-104 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Ru-105 JEF-2.2 JEFF-3.0 20 More recent evaluations available Ru-106 JEF-2.2 JEFF-3.0 20 More recent evaluations available Rh-103 ENDF/B-VI.8 New evaluation S 20 CEA N Rh-105 JEF-2.2 JEFF-3.0 20 More recent evaluations available Pd-102 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Pd-104 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Pd-105 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Pd-106 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Pd-107 JEF-2.2 JEFF-3.0 20 More recent evaluations available Pd-108 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Pd-110 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Ag-107 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Ag-109 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Ag-110m - JENDL-3.3 20 - n Ag-111 JEF-2.2 JEFF-3.0 20 -

8 Isotope JEFF-3.0 JEFF-3.1 S E Justiﬁcation/Comments Change Cd-106 ENDF/B-VI.4 JEFF-3.0 S 20 - Cd-108 ENDF/B-VI.4 JEFF-3.0 S 20 - Cd-110 ENDF/B-VI.4 JEFF-3.0 S 20 - Cd-111 ENDF/B-VI.4 JEFF-3.0 S 20 - Cd-112 ENDF/B-VI.4 JEFF-3.0 S 20 - Cd-113 ENDF/B-VI.4 JEFF-3.0 S 20 - Cd-114 ENDF/B-VI.4 JEFF-3.0 S 20 - Cd-115m ENDF/B-VI.4 JEFF-3.0 20 - Cd-116 ENDF/B-VI.4 JEFF-3.0 S 20 - In-113 JENDL-3.2 JENDL-3.3 S 20 T-matrix deleted Small adjustments In-115 JENDL-3.2 JENDL-3.3 S 20 T-matrix deleted Small adjustments Sn-112 EFF-2.4 = JENDL-3 JENDL-3.3 S 20 T-matrix deleted New energy distributions Sn-114 EFF-2.4 = JENDL-3 JENDL-3.3 S 20 T-matrix deleted New energy distributions Sn-115 EFF-2.4 = JENDL-3 JENDL-3.3 S 20 T-matrix deleted New energy distributions Sn-116 EFF-2.4 = JENDL-3 JENDL-3.3 S 20 T-matrix deleted New energy distributions Sn-117 EFF-2.4 = JENDL-3 JENDL-3.3 S 20 T-matrix deleted New energy distributions Sn-118 EFF-2.4 = JENDL-3 JENDL-3.3 S 20 T-matrix deleted New energy distributions Sn-119 EFF-2.4 = JENDL-3 JENDL-3.3 S 20 T-matrix deleted New energy distributions Sn-120 EFF-2.4 = JENDL-3 JENDL-3.3 S 20 T-matrix deleted New energy distributions Sn-122 EFF-2.4 = JENDL-3 JENDL-3.3 S 20 T-matrix deleted New energy distributions Sn-123 JEF-2.2 JEFF-3.0 20 More recent evaluations available Sn-124 EFF-2.4 = JENDL-3 JENDL-3.3 S 20 T-matrix deleted New energy distributions Sn-125 JEF-2.2 JEFF-3.0 20 More recent evaluations available Sn-126 JEF-2.2 JEFF-3.0 20 -

9 Isotope JEFF-3.0 JEFF-3.1 S E Justiﬁcation/Comments Change Sb-121 JENDL-3.2 JENDL-3.3 S 20 MF4,5 replaced by MF6 from JENDL-FF, T-matrix deleted Sb-123 JENDL-3.2 JENDL-3.3 S 20 MF4,5 replaced by MF6 from JENDL-FF, T-matrix deleted Sb-124 JEF-2.2 JEFF-3.0 20 More recent evaluations available Sb-125 JEF-2.2 JEFF-3.0 20 More recent evaluations available Sb-126 JEF-2.2 JEFF-3.0 20 More recent evaluations available Te-120 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Te-122 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Te-123 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Te-124 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Te-125 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Te-126 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Te-127m JEF-2.2 JEFF-3.0 20 More recent evaluations available Te-128 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Te-129m JEF-2.2 JEFF-3.0 20 More recent evaluations available Te-130 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Te-132 JEF-2.2 JEFF-3.0 20 More recent evaluations available I -127 JEF-2.2 New evaluation S 20 CEA evaluation N I -129 JEF-2.2 New evaluation 20 CEA evaluation N I -130 JEF-2.2 JEFF-3.0 20 - I -131 JEF-2.2 JEFF-3.0 20 More recent evaluations available I -135 JEF-2.2 JEFF-3.0 20 - Xe-124 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Xe-126 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Xe-128 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Xe-129 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Xe-130 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Xe-131 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Xe-132 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Xe-133 JEF-2.2 JEFF-3.0 20 More recent evaluations available Xe-134 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Xe-135 JEF-2.2 JEFF-3.0 20 More recent evaluations available Xe-136 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Cs-133 ENDF/B-VI.7 JEFF-3.0 S 20 Recommended in JEF/DOC-874 Cs-134 JEF-2.2 JEFF-3.0 20 More recent evaluations available Cs-135 JEF-2.2 JEFF-3.0 20 More recent evaluations available Cs-136 JEF-2.2 JEFF-3.0 20 More recent evaluations available Cs-137 JEF-2.2 JEFF-3.0 20 More recent evaluations available

10 Isotope JEFF-3.0 JEFF-3.1 S E Justiﬁcation/Comments Change Ba-130 JENDL-3.2 JENDL-3.3 S 20 T-matrix deleted Small adjustments Ba-132 JENDL-3.2 JENDL-3.3 S 20 T-matrix deleted Small adjustments Ba-134 JENDL-3.2 JENDL-3.3 S 20 T-matrix deleted Small adjustments Ba-135 JENDL-3.2 JENDL-3.3 S 20 T-matrix deleted Small adjustments Ba-136 JENDL-3.2 JENDL-3.3 S 20 T-matrix deleted Small adjustments Ba-137 JENDL-3.2 JENDL-3.3 S 20 T-matrix deleted Small adjustments Ba-138 JENDL-3.2 JENDL-3.3 S 20 T-matrix deleted Small adjustments Ba-140 JEF-2.2 JEFF-3.0 20 More recent evaluations available La-138 JENDL-3.2 JENDL-3.3 S 20 T-matrix deleted Small adjustments La-139 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available La-140 JEF-2.2 JEFF-3.0 20 - Ce-136 - - S - - Ce-138 - - S - - Ce-140 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Ce-141 JEF-2.2 JEFF-3.0 20 More recent evaluations available Ce-142 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Ce-143 JEF-2.2 JEFF-3.0 20 - Ce-144 JEF-2.2 JEFF-3.0 20 More recent evaluations available Pr-141 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Pr-142 JEF-2.2 JEFF-3.0 20 - Pr-143 JEF-2.2 JEFF-3.0 20 More recent evaluations available

11 Isotope JEFF-3.0 JEFF-3.1 S E Justiﬁcation/Comments Change Nd-142 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Nd-143 JEF-2.2 JEFF-3.0 S 20 Recommended in JEF/DOC-885 Nd-144 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Nd-145 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Nd-146 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Nd-147 JEF-2.2 JEFF-3.0 20 More recent evaluations available Nd-148 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Nd-150 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Pm-147 JEF-2.2 JEFF-3.0 20 More recent evaluations available Pm-148 JEF-2.2 JEFF-3.0 20 More recent evaluations available Pm-148m JEF-2.2 JEFF-3.0 20 More recent evaluations available Pm-149 JEF-2.2 JEFF-3.0 20 More recent evaluations available Pm-151 JEF-2.2 JEFF-3.0 20 - Sm-144 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Sm-147 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Sm-148 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Sm-149 JEF-2.2 JEFF-3.0 S 20 Recommended in JEF/DOC-885 Sm-150 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Sm-151 JEF-2.2 JEFF-3.0 20 More recent evaluations available Sm-152 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Sm-153 JEF-2.2 JEFF-3.0 20 More recent evaluations available Sm-154 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Eu-151 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Eu-152 JENDL-3.2 JEFF-3.0 20 Corrections of JEFF-3.0 not in JENDL-3.3 Recommended in JEF/DOC-912 Eu-153 JENDL-3.2 JEFF-3.0 S 20 Corrections of JEFF-3.0 not in JENDL-3.3 Recommended in JEF/DOC-912 Eu-154 ENDF/B-VI.7 JEFF-3.0 20 Recommended in JEF/DOC-885 Eu-155 ENDF/B-VI.7 JEFF-3.0 20 Recommended in JEF/DOC-885 Eu-156 JEF-2.2 JEFF-3.0 20 More recent evaluations available Eu-157 JEF-2.2 JEFF-3.0 20 More recent evaluations available

12 Isotope JEFF-3.0 JEFF-3.1 S E Justiﬁcation/Comments Change Gd-152 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Gd-154 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Gd-155 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Gd-156 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Gd-157 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Gd-158 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Gd-160 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Tb-159 JEF-2.2 JEFF-3.0 S 20 More recent evaluations available Tb-160 JEF-2.2 JEFF-3.0 20 - Dy-156 - - S - - Dy-158 - - S - - Dy-160 JEF-2.2 JEFF-3.0 S 20 Processing problems for ENDF/B-VI.8 * Dy-161 JEF-2.2 ENDF/B-VI.8 S 20 Low energy data revised by Wright * Dy-162 JEF-2.2 ENDF/B-VI.8 S 20 Low energy data revised by Wright * Dy-163 JEF-2.2 ENDF/B-VI.8 S 20 Low energy data revised by Wright * Dy-164 JEF-2.2 ENDF/B-VI.8 S 20 Low energy data revised by Wright * Ho-165 ENDF/B-VI.4 ENDF/B-VI.8 S 30 Corrected total and neutron width and Emax Er-162 BROND-2.2 JENDL-3.3 S 20 New evaluation (2000), more complete * New resonance parameters, Mughabghab 1984 Er-164 BROND-2.2 JENDL-3.3 S 20 New evaluation (2000), more complete * New resonance parameters, Mughabghab 1984 Er-166 BROND-2.2 JENDL-3.3 S 20 New evaluation (2000), more complete * New resonance parameters, Mughabghab 1984 Er-167 BROND-2.2 JENDL-3.3 S 20 New evaluation (2000), more complete * New resonance parameters, Mughabghab 1984 Er-168 BROND-2.2 JENDL-3.3 S 20 New evaluation (2000), more complete * New resonance parameters, Mughabghab 1984 Er-170 BROND-2.2 JENDL-3.3 S 20 New evaluation (2000), more complete * New resonance parameters, Mughabghab 1984 Tm-169 - - S - -

13 Isotope JEFF-3.0 JEFF-3.1 S E Justiﬁcation/Comments Change Yb-168 - - S - - Yb-170 - - S - - Yb-171 - - S - - Yb-172 - - S - - Yb-173 - - S - - Yb-174 - - S - - Yb-176 - - S - - Lu-175 JEF-2.2 ENDF/B-VI.8 S 20 New evaluation from R.Q. Wright (1998): * MLBW, Mugh. 1984 instead of SLBW (1967) Lu-176 JEF-2.2 ENDF/B-VI.8 S 20 New evaluation from R.Q. Wright (1998): * MLBW, Mugh. 1984 instead of SLBW (1967) Hf-174 JENDL-3.2 New evaluation S 20 CEA/RPI (JENDL-3.3 revised) N Hf-176 JENDL-3.2 New evaluation S 20 CEA/RPI (JENDL-3.3 revised) N Hf-177 JENDL-3.2 New evaluation S 20 CEA/RPI (JENDL-3.3 revised) N Hf-178 JENDL-3.2 New evaluation S 20 CEA/RPI (JENDL-3.3 revised) N Hf-179 JENDL-3.2 New evaluation S 20 CEA/RPI (JENDL-3.3 revised) N Hf-180 JENDL-3.2 New evaluation S 20 CEA/RPI (JENDL-3.3 revised) N

14 Isotope JEFF-3.0 JEFF-3.1 S E Justification/Comments Change Ta-180 - - S - - Ta-181 JENDL-3.2 JENDL-3.3 S 20 MT1,2,107 revised Ta-182 JEF-2.2 JEFF-3.0 20 - W -180 - - S - - W -182 JENDL-3.2 JENDL-3.3 S 20 MF4,5 repl. by MF6 (JENDL-FF) Photon prod. added, MT2 revised W -183 JENDL-3.2 JENDL-3.3 S 20 MF4,5 repl. by MF6 (JENDL-FF) Photon prod. added, MT2 revised W -184 JENDL-3.2 JENDL-3.3 S 20 MF4,5 repl. by MF6 (JENDL-FF) Photon prod. added, MT2 revised W -186 JENDL-3.2 JENDL-3.3 S 20 MF4,5 repl. by MF6 (JENDL-FF) Photon prod. added, MT2 revised Re-185 ENDF/B-VI.4 JEFF-3.0 S 20 = ENDF/B-VI.8 Re-187 ENDF/B-VI.4 JEFF-3.0 S 20 = ENDF/B-VI.8 Os-000 BROND-2.2 JEFF-3.0 + corr. S 20 Decay scheme modified Ir-191 BROND-2.2 for Ir-000 ENDF/B-VI.8 S 20 Newer isotopic evaluation (1995) * partly based on BROND-2.2 Ir-193 BROND-2.2 for Ir-000 ENDF/B-VI.8 S 20 Newer isotopic evaluation (1995) * partly based on BROND-2.2 Pt-000 ENDL-78 REV L JEFF-3.0 + corr. S 20 MF4 skeleton added Au-197 ENDF/B-VI (1984) JEFF-3.0 S 20 = ENDF/B-VI.8 Hg-196 CENDL for Hg-000 JENDL-3.3 S 20 Isotopic evaluation * Hg-198 CENDL for Hg-000 JENDL-3.3 S 20 Isotopic evaluation * Hg-199 CENDL for Hg-000 JENDL-3.3 S 20 Isotopic evaluation * Hg-200 CENDL for Hg-000 JENDL-3.3 S 20 Isotopic evaluation * Hg-201 CENDL for Hg-000 JENDL-3.3 S 20 Isotopic evaluation * Hg-202 CENDL for Hg-000 JENDL-3.3 S 20 Isotopic evaluation * Hg-204 CENDL for Hg-000 JENDL-3.3 S 20 Isotopic evaluation * Tl-000 CENDL JEFF-3.0 S 20 - + modification Pb-204 JENDL-3.2 New evaluation S 200 NRG-2004 N Pb-206 JENDL-3.2 New evaluation S 200 NRG-2004 N Pb-207 JENDL-3.2 New evaluation S 200 NRG-2004 N Pb-208 JENDL-3.2 New evaluation S 200 NRG-2004 N Bi-209 ENDF/B-VI.4 New evaluation S 200 NRG-2004 N

15 Isotope JEFF-3.0 JEFF-3.1 S E Justification/Comments Change Ra-223 - JENDL-3.3 20 - n Ra-224 - JENDL-3.3 20 - n Ra-225 - JENDL-3.3 20 - n Ra-226 - JENDL-3.3 20 - n Ac-225 - JENDL-3.3 20 - n Ac-226 - JENDL-3.3 20 - n Ac-227 - JENDL-3.3 20 - n Th-227 JENDL-3.2 JENDL-3.3 20 T-matrix deleted, small adj. 8 time groups Th-228 JENDL-3.2 JENDL-3.3 20 T-matrix deleted Small adjustments Th-229 JENDL-3.2 JEFF-3.0 20 8 time groups + modifications Th-230 JEF-2.2 JEFF-3.0 20 - Th-232 JENDL-3.2+ENDF/B-VI.4 Maslov, Minsk S 20 + revisions, 8 time groups * Th-233 JENDL-3.2 JENDL-3.3 20 T-matrix deleted Small adjustments Th-234 JENDL-3.2 JENDL-3.3 20 T-matrix deleted Small adjustments Pa-231 JEF-2.2 JEFF-3.0 20 8 time groups Pa-232 - ENDF/B-VI.8 20 More recent res parameters n than JENDL-3.3 Pa-233 JEF-2.2 JEFF-3.0 20 - U -232 JEF-2.2+ENDF/B-VI.4 JEFF-3.0 20 8 time groups U -233 JENDL-3.2+ENDF/B-VI.4 JENDL-3.3+ENDF/B-VI.4 20 +NRG correction, fuel cycle * +modifications 8 time groups U -234 JEF-2.2+ENDL-78 Maslov, Minsk S 20 fuel cycle feedback * 8 time groups U -235 ENDF/B-VI.6+ JEFF-3.0 S 30 New evaluation of CEA + modifications under study, 8 time groups U -236 JENDL-3.2+ENDF/B-VI.4 New evaluation 30 CEA, 8 time groups N U -237 JENDL-3.2+JEF-2.2 New evaluation 30 CEA, 8 time groups N + modifications U -238 JEF-2.2+JENDL-3.2 New evaluation S 30 CEA, 8 time groups N +ENDF/B-VI.4+modific.

16 Isotope JEFF-3.0 JEFF-3.1 S E Justification/Comments Change Np-235 - JENDL-3.3 20 - n Np-236 JENDL-3.2 JENDL-3.3 20 T-matrix deleted Small adjustments Np-237 ENDF/B-VI.4+JEF-2.2 JENDL-3.3 20 8 time groups + modifications Np-238 JEF-2.2+ENDF/B-VI.4 JEFF-3.0 20 8 time groups Np-239 ENDF/B-VI.4+JENDL-3.2 JEFF-3.0 20 - + modifications Pu-236 JEF-2.2 JENDL-3.3 20 Newer resonance eval.

Pu-237 JEF-2.2 JEFF-3.0 20 - Pu-238 JENDL-3.2+BROND-2.2 JEFF-3.0 20 8 time groups +ENDF/B-VI+modific. Pu-239 JEFF-3.0 JEFF-3.0 updated 30 8 time groups Pu-240 JEFF-3.0 New evaluation 30 CEA, 8 time groups Pu-241 JENDL-3.2+modific. JEFF-3.0 20 8 time groups Pu-242 JEFF-3.0 JEFF-3.0 20 8 time groups Pu-243 JEF-2.2 JEFF-3.0 20 - Pu-244 JEF-2.2 JEFF-3.0 20 - Pu-246 - JENDL-3.3 20 - n Am-241 JEF-2.2 New evaluation 20 CEA, JENDL-3.3 revised, N +modifications new RR, 8 time groups Am-242g ENDF/B-VI.2+ENDL-78 JENDL-3.3 20 * Am-242m JEF-2.2+ENDF/B-VI JENDL-3.3 20 8 time groups * +modifications Am-243 JEF-2.2 JENDL-3.3 20 8 time groups * +modifications Am-244 JENDL-3.2 JENDL-3.3 20 T-matrix deleted Small adjustments Am-244m JENDL-3.2 JENDL-3.3 20 T-matrix deleted Small adjustments

17 Isotope JEFF-3.0 JEFF-3.1 S E Justification/Comments Change Cm-240 - JENDL-3.3 20 - n Cm-241 JEF-2.2 JEFF-3.0 20 - +modifications Cm-242 JEF-2.2+JENDL-3.2 JEFF-3.0 20 8 time groups +ENDF/B-VI+modific. Cm-243 JEF-2.2+JENDL-3.2 JEFF-3.0 20 8 time groups +ENDF/B-VI+modific. Cm-244 JEF-2.2 ENDF/B-VI.8 20 8 time groups * +modifications Cm-245 JENDL-3.2 JENDL-3.3 20 8 time groups Cm-246 JEFF-3.0 JEFF-3.0 20 8 time groups Cm-247 JENDL-3.2+JEF-2.2 JEFF-3.0 20 - + modifications Cm-248 JENDL-3.2+JEF-2.2 JEFF-3.0 20 8 time groups + modifications Cm-249 JENDL-3.2 JENDL-3.3 20 New evaluation Cm-250 JENDL-3.2 JENDL-3.3 20 New evaluation Bk-247 - JENDL-3.3 20 - n Bk-249 ENDF/B-VI.4 + JEF-2.2 JENDL-3.3 20 Performs better Bk-250 JENDL-3.2 JEFF-3.0 20 - + modification Cf-249 ENDF/B-VI.4 + JEF-2.2 JENDL-3.3 20 Processing problems for JEFF-3.0 and ENDF/B-VI.8, 8 time groups Cf-250 JEF-2.2 JENDL-3.3 20 Performs better Cf-251 JEF-2.2 JEFF-3.0 20 8 time groups Cf-252 JEF-2.2 JEFF-3.0 20 - Cf-254 JENDL-3.2 JEFF-3.0 20 - + modification Es-253 - JEF-2.2 20 More recent than ENDF/B-V n Es-254 JENDL-3.2 JENDL-3.3 20 T-matrix deleted, 8 time groups Es-255 JENDL-3.2 JENDL-3.3 20 T-matrix deleted Fm-255 - JENDL-3.3 20 - n

18 APPENDIX 2 Three reports on the performance of the JEFF-3.1 neutron general purpose library are included below. They were all prepared by S. Van der Marck, NRG Petten in October 2005, see JEF/DOC-1105-1107 in the JEFF working paper list above. The three reports are entitled:

• βeff calculations using JEFF-3.1 nuclear data,

• Shielding benchmarck calculations with MCNP-4c3 using JEFF-3.1 nuclear data,

• Criticality Safety benchmarck calculations with MCNP-4c3 using JEFF-3.1 nuclear data.

Abstract Calculations were performed with MCNP-4C3 to validate the delayed neutron data of the new JEFF-3.1 nuclear data library.

Keywords MCNP-4C3 JEFF-3.1 nuclear data effective delayed neutron fraction

2-16 21616/05.69454/P Contents

1 Introduction 5

2 Benchmark systems 6

3 Results 9

References 11

Appendix A JEFF-3.1 results for þeff and þ0 in 8 groups 13

21616/05.69454/P 3-16 4-16 21616/05.69454/P 1. Introduction

Recently the new nuclear data library JEFF-3.1 has been made available by the NEA. The delayed neutron data in this library is new. For the ﬁrst time, 8 time groups have been used to represent these data, instead of the customary 6 time groups [1]. To validate the results that can be obtained with the new delayed neutron data, we have calculated þeff for many systems for which measurements have been reported in the open literature. The results of these calculations are reported here. The results in this report have been obtained using MCNP-4C3 [2], extended with a method to calculate þeff [3]. The standard versions of MCNP-4C3, MCNP-5, and MCNPX can all work with delayed neutron data in 8 time groups. However, because these codes do not calculate a value for þeff, we included our own þeff method into MCNP-4C3. This method has been used earlier to calculate þeff for other nuclear data libraries, viz. JEFF-3.0, ENDF/B-VI.8, and JENDL-3.3 [4].

21616/05.69454/P 5-16 2. Benchmark systems

We have searched in the literature for measurements of the effective delayed neutron fraction, the result of which is listed below. We will use these experiments as benchmarks for the calculation of þeff on the basis of JEFF-3.1 nuclear data. For some systems we have found experimental values for the parameter Þ, which is linked to þeff through Þ D [k.1 þeff/ 1]=l, where l is the prompt neutron life time. All systems described below are at delayed criticality, so that the parameter we can compare with is the value Þdc D Þ.k D 1/ D þeff=l. When not stated explicitly, the MCNP [2] model for the experiment was taken, without modifications, from the ICSBEP data [5]. Where possible, the ICSBEP identification is given in brackets after the benchmark name. ž Godiva (heu-met-fast-001) A bare sphere of highly enriched (94 wt%) uranium. ž Jezebel (pu-met-fast-001) A bare sphere of plutonium (95 at% Pu-239). ž Skidoo (u233-met-fast-001) A bare sphere of uranium, of which 98 at% U-233. ž Topsy (Flattop 25, heu-met-fast-028) A highly enriched (93 wt%) uranium sphere surrounded by a thick reflector of normal uranium. Experimental results are given in Ref. [6]. ž Popsy (Flattop-Pu, pu-met-fast-006) A plutonium (94 wt% Pu-239) sphere surrounded by a thick reflector of normal uranium. Ex- perimental results are given in Ref. [6]. ž Flattop 23 (u233-met-fast-006) A uranium (98 at% U-233) sphere surrounded by a thick reflector of normal uranium. Experi- mental results are given in Ref. [6]. ž Big Ten (ieu-met-fast-007) A large, mixed-uranium-metal cylindrical core with 10% average U-235 enrichment, surrounded by a thick reflector of depleted uranium [7]. ž ZPR (heu-met-inter-001, ieu-met-fast-010, mix-met-fast-011 case 1, pu-met-inter-002) Four cores in the Zero Power Reactor at ANL. The first one is a highly enriched uranium/iron benchmark, reflected by steel. The second is a heterogeneous cylindrical core of uranium (average enrichment 9%). The third has plutonium/uranium/zirconium fuel, reflected by graphite. The last core had heterogeneous plutonium metal fuel with carbon/stainless steel dilutions, and a steel reflector. Measured values for þeff are given in e.g. Ref. [8]. ž SNEAK (cores 7A, 7B, 9C1, and 9C2) Measurements of þeff in four unmoderated PuO2-UO2 cores, surrounded by a depleted uranium reflector [9]. One core, 9C1, had only uranium as fuel. The 9C2 core was diluted with sodium. MCNP models were built based on the R-Z model descriptions in Ref. [9]. ž Masurca (cores R2 and ZONA2) Measurements of þeff by several international groups in two unmoderated cores, viz. R2 and ZONA2 [10]. Core R2 had of ¾ 30% enriched uranium as fuel, whereas ZONA2 had both

plutonium and depleted uranium. Both cores were surrounded by a 50-50 UO2-Na mixture blanket, and by steel shielding. MCNP models were built based on the R-Z model descriptions

6-16 21616/05.69454/P in Ref. [10]. ž FCA (cores XIX-1, XIX-2, and XIX-3) Measurements of þeff by several international groups in three unmoderated cores in the Fast Critical Assembly [10]. One core had highly enriched uranium, one had plutonium and natural uranium, and the third one had plutonium as fuel. The cores were surrounded by two blanket regions, one with depleted uranium oxide and sodium, and another one with only depleted uranium metal. MCNP models were built based on the R-Z model descriptions in Ref. [10]. ž TCA (related to leu-comp-therm-006)

A light water moderated low-enriched UO2 core in the Tank-type Critical Assembly. From the description of this experiment in Ref. [11] it is clear that this experiment is closely related to benchmark leu-comp-therm-006 [5]. We have taken the MCNP input decks given in Ref. [5], and changed the loading pattern, the water height and lattice pitch. ž IPEN/MB-01 (related to leu-comp-therm-077) Measurement of þeff in the research reactor IPEN/MB-01, with a core consisting of 28 ð 26 UO2 (4.3% enriched) fuel rods inside a light water filled tank [12]. An MCNP input deck was made available by the authors of Ref. [12]. ž Winco slab tanks (related to heu-sol-therm-038 case 5) Measurement of Þ in the Westinghouse Idaho Nuclear Company Slab Tank Assembly. The experiment consisted of two thin coaxial slab tanks with 93% enriched uranyl nitrate solution. From the description of this experiment in Ref.[13] it is clear that this experiment is closely related to heu-sol-therm-038, case 5 [5]. We have taken the MCNP input deck given in Ref. [5], and removed the stainless steel absorber between the two slab tanks. ž Stacy (leu-sol-therm-004, -007, -016, -021) Measurements of þeff=l in uranyl nitrate solution (10 % enrichment) in several cores in the STACY facility. From the description of these experiments in Ref. [14], one can identify several experiments that have been included in the criticality benchmark collection [5]. ž Sheba (core II) Measurement of þeff=l in a critical assembly vessel, filled with 5% enriched uranyl fluoride, UO2F2, the Solution High-Energy Burst Assembly [15]. The vessel had a cylindrical shape, and there was no reflector. An MCNP model was built from scratch. ž SHE-8 Measurement of þeff=l in a split table type critical assembly called Semi-Homogeneous Assem- bly [16]. The core was shaped in a hexagonal prism, with graphite matrix tubes and graphite rods. There was no axial reflector. The central region in core 8 consisted of 73 fuel rods with

2.9% enriched UO2 dispersed in graphite. An MCNP model was built from scratch. ž Proteus (core 5) Measurement of þeff=l in a graphite reﬂected pebble bed reactor, containing uranium-carbon fuel pebbles (16.7% enrichment) and graphite moderator pebbles. As the reactor was operated below 1 kW, no coolant was needed. An MCNP model was made available by the authors of Ref. [17]. For Godiva, Jezebel, and Skidoo, both Keepin [18] and Paxton [6] give experimental values for þeff. Although these values are not identical, the differences are small and have no signiﬁ- cant impact on the conclusions drawn in this paper. Therefore we will use the numbers given by Keepin, because this is the commonly used reference. For Big Ten, an experimental value Þ D .1:17š0:01/ð105 s1 is given by Paxton [6]. Based on a

21616/05.69454/P 7-16 8 calculation of þeff D 720 pcm, he estimates the prompt neutron life time to be 6:15ð10 s, which is consistent with our calculation of 6:15 ð 108 s within a fraction of a percent. Therefore we feel it is justified in this case to compare with the value of þeff D 720 pcm as if it were determined by experiment. For the Stacy, Winco, SHE-8, Sheba-II, and Proteus experiments, we will compare with the Þ values given in the respective references, because that is the measured quantity. Also, for the Winco experiment, there is some uncertainty in deriving a value for þeff from the measured Þ. The comparison in the next section is done by dividing the calculated þeff by the prompt neutron fission life time. This life time is calculated by MCNP by default, and is given in the output as the ’fission lifespan’ (see the discussion of life time estimation in section 2.VIII.B of the manual [2]).

8-16 21616/05.69454/P 3. Results

The results for the Sheba-II, SHE-8, TCA and Winco experiments should be viewed with some caution, since the preparation of the MCNP models for these experiments involved interpretation on our part, based on the references given earlier. However, since the computational results for these cases are close to the experimental values, we judge the models to be appropriate for calculating þeff and Þ.

Experiment Calculation C/E (JEFF-3.1) JEFF-3.0 JEFF-3.1 TCA 771š17 817š9 793.4š1.4 1.029š0.002 IPEN 742š7 788š4 771.7š4.1 1.040š0.005 Masurca R2 721š11 735š7 728.9š6.9 1.011š0.009 Masurca Z2 349š6 358š5 356.2š4.6 1.021š0.013 FCA-XIX-1 742š24 767š8 749.7š7.5 1.010š0.010 FCA-XIX-2 364š9 387š5 383.7š4.9 1.054š0.013 FCA-XIX-3 251š4 255š4 250.3š4.1 0.997š0.016 Sneak-9C1 758š24 757š7 743.8š7.0 0.981š0.009 Sneak-7A 395š12 381š5 377.3š4.8 0.955š0.013 Sneak-7B 429š13 447š5 434.2š5.0 1.012š0.012 Sneak-9C2 426š19 395š5 388.0š4.9 0.911š0.013 Zpr-Heu 667š15 696š9 685.8š8.8 1.028š0.013 Zpr-U9 725š17 769š9 747.7š8.3 1.031š0.011 Zpr-Mox 381š9 373š6 369.8š5.1 0.971š0.014 Zpr-Pu 222š5 243š5 241.3š5.2 1.087š0.022 BigTen 720š7 765š7 735.3š6.4 1.021š0.009 Godiva 659š10 677š8 669.4š4.0 1.016š0.006 Topsy 665š13 666š8 659.5š3.9 0.992š0.006 Jezebel 194š10 200š5 201.8š2.2 1.040š0.011 Popsy 276š7 283š5 282.5š2.5 1.024š0.009 Skidoo 290š10 284š5 309.1š2.7 1.066š0.009 Flattop 360š9 343š6 361.9š2.8 1.005š0.008 Table 3.1 The experimental and calculated þeff (in pcm). The uncertainty for C/E includes only the statistical uncertainty of the calculation. This corresponds to the error bar in Fig. 3.1. The experimental uncertainty range is shown with dashed lines in Fig. 3.1.

Concerning the Winco slab tank experiment, another remark is in order. Our calculation of þeff yields a value of 845 š 4 pcm, contradicting the value 1500 š 120 calculated in Ref. [13], based on the experimental Þ-value and other experimental information, not involving the prompt neutron ﬁssion life time. However, in Ref. [13] it is noted that its calculated value is higher than the expected value of roughly 900 pcm, the reason for which was not well understood. Our value for Þ is close to the experimental value.

21616/05.69454/P 9-16 Experiment Calculation C/E JEFF-3.0 JEFF-3.1 Proteus 3.60š0.02 3.82š0.05 3.70š0.02 1.028š0.005 SHE-8 6.53š0.34 6.36š0.08 6.21š0.15 0.951š0.024 Sheba-II 200.3š3.6 203.79š1.38 1.017š0.007 Stacy-029 122.7š4.1 122.2š2.6 120.76š0.75 0.984š0.006 Stacy-033 116.7š3.9 129.5š2.4 114.11š0.72 0.978š0.006 Stacy-046 106.2š3.7 109.3š2.2 104.77š0.66 0.987š0.006 Stacy-030 126.8š2.9 129.8š2.7 129.17š0.81 1.019š0.006 Stacy-125 152.8š2.6 162.3š2.4 156.38š1.00 1.023š0.006 Stacy-215 109.2š1.8 114.6š2.3 109.28š0.68 1.001š0.006 Winco 1109.3š0.3 1169.š13. 1134.98š5.54 1.023š0.005 Table 3.2 The experimental and calculated Þ (in s1). The uncertainty for C/E includes only the statistical uncertainty of the calculation. This corresponds to the error bar in Fig. 3.1. The experimental uncertainty range is shown with dashed lines in Fig. 3.1.

1.2 exp. uncertainty JEFF-3.0 JEFF-3.1 1.15 thermal fast

1.1

1.05

0.95

0.9

Stacy FCA Sneak ZPR

0.85 TCA Topsy Winco Popsy SHE-8 Skidoo Flattop BigTen Godiva Jezebel Proteus Sheba-II Masurca IPEN/MB01

0.8

Figure 3.1 C/E for þeff (or þeff=l, see text) for many benchmark systems. The systems are roughly ordered with respect to the average energy at which ﬁssion takes place, from low energy (left) to high (right).

10-16 21616/05.69454/P References

[1] J.L. Rowlands, Delayed neutron data in 8 time groups, JEFDOC-976 (2003) J.L. Rowlands, Delayed neutron yield and relative abundance data for some minor isotopes, missing from current compilations, JEFDOC-1073 (2005) [2] J.F. Briesmeister (Ed.), MCNP - A General Monte Carlo N-Particle Transport Code, Version 4C, Technical Report LA-13709-M, Los Alamos National Laboratory, USA (2000); J.S. Hen- dricks, MCNP4C3, Report X-5:RN(U)-JSH-01-17, Los Alamos National Laboratory, USA (2001) [3] S.C. van der Marck and R. Klein Meulekamp, Calculation of the effective delayed neutron fraction by Monte Carlo, Proc. Physor-2004, Chicago, USA (2004); R. Klein Meulekamp and S.C. van der Marck, Calculation of the effective delayed neutron fraction by Monte Carlo, to be published in Nucl. Sci. Eng. (2006) [4] S.C. van der Marck, R. Klein Meulekamp, A. Hogenbirk, and A.J. Koning, Benchmark results for delayed neutron data, Proc. ND-2004, Santa Fe, USA (2004) pp. 531–534 [5] J. Blair Briggs (Ed.), International Handbook of Evaluated Criticality Safety Benchmark Ex- periments, NEA/NSC/DOC(95)03/I, Nuclear Energy Agency, Paris (September 2003 Edi- tion) [6] H.C. Paxton, Fast Critical Experiments, Progress in Nuclear Energy 7 (1981) 151–174 [7] L.J. Sapir, H.H. Helmick, and J.D. Orndoff, Big Ten, a 10% Enriched Uranium Critical As- sembly: Kinetic Studies, Trans. Am. Nucl. Soc. 15, 1 (1972) 312 [8] E. Fort, V. Zammit-Averlant, M. Salvatores, and A. Filip, Recommended values of the delayed neutron yield for U-235, U-238, and Pu-239, JEFDOC-820 [9] E.A. Fischer, Integral measurements of the effective delayed neutron fractions in the fast critical assembly SNEAK, Nucl. Sci. Eng. 62 (1977) 105–116 [10] S. Okajima, T. Sakurai, J.F. Lebrat, V.Z. Averlant, and M. Martini, Summary on International Benchmark Experiments for Effective Delayed Neutron Fraction, Progress in Nuclear Energy, 41 (2002) 285–301 [11] K. Nakajima, Re-evaluation of the Effective Delayed Neutron Fraction Measured by the Sub- stitution Technique for a Light Water Moderated Low-Enriched Uranium Core, J. Nucl. Sci. Technol. 38 (2001) 1120–1125 [12] A. dos Santos, R. Diniz, L.C.C.B. Fanaro, R. Jerez, G.S. de Andrade e Silva, and M. Yam- aguchi, The experimental determination of the effective delayed neutron parameters of the IPEN/MB-01 reactor, Proc. Physor-2004, Chicago [13] G.D. Spriggs, A Measurement of the Effective Delayed Neutron Fraction of the Westinghouse Idaho Nuclear Company Slab Tank Assembly using Rossi-Þ Techniques, Nucl. Sci. Eng. 115 (1993) 76–80 [14] K. Tonoike, Y. Miyoshi, T. Kikuchi, and T. Yamamoto, Kinetic Parameter þeff=l Measure- ment on Low Enriched Uranyl Nitrate Solution with Single Unit Cores (600, 280T, 800) of STACY, J. Nucl. Sci. Technol. 39 (2002) 1227–1236 [15] K.B. Butterﬁeld, The SHEBA experiment, Trans. Am. Nucl. Soc. 74 (1994) 199–200; R. Sanchez and P. Jaegers, Prompt neutron decay constants in SHEBA, Trans. Am. Nucl. Soc. 76 (1997) 363–364 [16] M. Takano, M. Hirano, R. Shindo, and T. Doi, Analysis of SHE critical experiments by neutronic design codes for experimental Very High Temperature Reactor, J. Nucl. Sci. Techn. 22 (1985) 358–370;

21616/05.69454/P 11-16 Y. Kaneko, F. Akino, and T. Yamane, Evaluation of delayed neutron data for thermal ﬁs- sion of U-235 based on integral experiments at Semi-Homogeneous Experiment, J. Nucl. Sci. Techn. 25 (1988) 673–681 [17] T. Williams, M. Rosselet, and W. Scherer (Eds.), Critical Experiments and Reactor Physics Calculations for Low-Enriched High Temperature Gas Cooled Reactors, IAEA Tecdoc 1249 (advance electronic version, 2001) [18] G.R. Keepin, Physics of Nuclear Kinetics, Addison-Wesley, Reading, USA, 1965

12-16 21616/05.69454/P APPENDIX A. JEFF-3.1 results for þeff and þ0 in 8 groups

In the JEFF-3.1 nuclear data library the delayed neutron data have been represented using 8 time groups. Since this is the ﬁrst time this representation has been used, it is deemed useful to list the results of our calculations per time group, for reference.

Godiva Jezebel Skidoo

group g þeff;g þ0;g þeff;g þ0;g þeff;g þ0;g 1 24.2š0.8 20.9š0.5 6.5š0.4 6.4š0.3 26.6š0.8 22.4š0.5 2 95.4š1.5 93.0š1.0 45.4š1.0 47.2š0.7 46.1š1.0 44.1š0.7 3 69.4š1.3 61.5š0.8 18.6š0.7 19.2š0.4 44.4š1.0 38.0š0.6 4 132.6š1.8 124.6š1.1 30.9š0.8 31.3š0.6 62.3š1.2 56.4š0.8 5 208.9š2.2 198.5š1.4 68.9š1.3 71.0š0.8 93.6š1.5 86.1š0.9 6 62.7š1.2 59.5š0.8 8.4š0.4 8.5š0.3 12.0š0.5 10.9š0.3 7 59.3š1.2 55.5š0.7 19.2š0.7 20.1š0.5 19.4š0.7 17.0š0.4 8 17.0š0.7 16.0š0.4 3.8š0.3 3.7š0.2 4.8š0.4 3.9š0.2 Table A.1 The partial þeff and þ0 for Godiva, Jezebel and Skidoo (in pcm).

Topsy Popsy Flattop-23

group g þeff;g þ0;g þeff;g þ0;g þeff;g þ0;g 1 21.1š0.7 20.8š0.5 6.1š0.4 9.0š0.3 21.3š0.7 20.6š0.5 2 85.5š1.4 107.6š1.0 47.3š1.0 76.9š0.9 47.5š1.0 73.2š0.9 3 60.1š1.2 61.2š0.8 19.9š0.7 31.0š0.6 39.4š1.0 44.4š0.7 4 125.8š1.7 147.7š1.2 40.6š0.9 79.9š0.9 66.4š1.2 99.3š1.0 5 202.7š2.2 249.9š1.6 90.6š1.4 170.7š1.3 107.8š1.5 175.1š1.3 6 75.3š1.3 107.9š1.0 30.5š0.8 81.5š0.9 33.9š0.8 79.7š0.9 7 64.2š1.2 84.3š0.9 33.1š0.8 65.8š0.8 31.0š0.8 60.5š0.8 8 24.8š0.7 40.4š0.6 14.3š0.5 37.7š0.6 14.7š0.6 36.9š0.6 Table A.2 The partial þeff and þ0 for Topsy, Popsy, and Flattop-23 (in pcm).

group g þeff;g þ0;g 1 15.4š0.9 20.1š0.7 2 97.8š2.3 120.5š1.8 3 52.0š1.7 63.2š1.3 4 125.7š2.6 155.7š2.1 5 221.7š3.5 279.1š2.8 6 101.4š2.3 130.7š1.9 7 77.8š2.1 98.6š1.7 8 43.3š1.6 54.6š1.2 Table A.3 The partial þeff and þ0 for Big Ten (in pcm).

21616/05.69454/P 13-16 Pu Mox HEU U9

g þeff;g þ0;g þeff;g þ0;g þeff;g þ0;g þeff;g þ0;g 1 8.2š1.0 7.7š0.6 7.9š0.7 9.0š0.5 25.9š1.7 21.7š1.0 15.7š1.2 19.5š0.9 2 52.5š2.4 50.2š1.5 63.2š2.1 62.4š1.3 98.2š3.4 98.7š2.1 98.1š3.0 118.0š2.3 3 22.3š1.6 21.8š1.0 30.6š1.4 29.6š0.9 69.9š2.8 65.0š1.7 53.6š2.3 65.8š1.7 4 34.9š2.0 33.4š1.2 59.3š2.0 58.5š1.3 136.6š3.9 133.3š2.4 124.7š3.3 158.7š2.7 5 86.3š3.1 83.1š1.9 118.3š2.8 119.2š1.8 213.9š4.9 207.2š3.0 226.4š4.5 277.7š3.5 6 7.9š0.9 8.9š0.6 37.9š1.7 35.3š1.0 65.6š2.8 62.0š1.7 105.3š3.1 128.8š2.4 7 25.2š1.7 21.4š1.0 39.4š1.7 36.7š1.0 57.5š2.6 56.6š1.6 80.5š2.7 96.8š2.1 8 3.9š0.7 3.9š0.4 13.3š1.0 13.9š0.6 18.4š1.5 16.1š0.8 43.5š2.0 52.3š1.5 Table A.4 The partial þeff and þ0 for ZPR-Pu, ZPR-Mox, ZPR-HEU, and ZPR-U9 (in pcm).

TCA IPEN/MB-01

group g þeff;g þ0;g þeff;g þ0;g 1 24.3š0.2 21.4š0.1 26.3š0.8 22.9š0.5 2 115.1š0.5 106.0š0.3 111.3š1.6 103.7š1.0 3 68.7š0.4 61.0š0.2 70.1š1.2 61.5š0.8 4 150.5š0.6 136.9š0.4 148.4š1.8 136.2š1.2 5 256.0š0.8 234.3š0.5 250.0š2.3 229.0š1.5 6 83.6š0.4 75.6š0.3 76.5š1.3 69.7š0.9 7 69.7š0.4 63.3š0.3 66.5š1.2 60.8š0.8 8 25.6š0.2 23.5š0.2 22.5š0.7 21.1š0.5 Table A.5 The partial þeff and þ0 for TCA and IPEN/MB-01 (in pcm).

R2 ZONA2

group g þeff;g þ0;g þeff;g þ0;g 1 21.9š1.2 21.3š0.8 6.5š0.6 8.3š0.5 2 102.1š2.6 110.1š1.8 57.9š1.8 71.6š1.4 3 63.5š2.1 64.8š1.3 21.8š1.1 27.9š0.9 4 132.6š2.9 142.2š2.0 50.1š1.7 62.9š1.3 5 222.7š3.8 241.6š2.6 114.9š2.6 146.6š2.0 6 87.8š2.4 96.8š1.6 44.0š1.6 60.7š1.3 7 69.8š2.1 75.7š1.5 40.3š1.5 55.8š1.2 8 28.4š1.4 33.5š1.0 20.8š1.2 25.9š0.8 Table A.6 The partial þeff and þ0 for Masurca-R2 and Masurca-ZONA2 (in pcm).

14-16 21616/05.69454/P FCA-XIX-1 FCA-XIX-2 FCA-XIX-3

group g þeff;g þ0;g þeff;g þ0;g þeff;g þ0;g 1 24.4š1.4 20.9š0.8 7.2š0.7 8.2š0.5 6.0š0.6 8.1š0.5 2 106.4š2.9 101.9š1.7 60.2š1.9 69.8š1.4 50.0š1.8 56.7š1.3 3 75.0š2.3 68.2š1.4 26.5š1.3 30.9š0.9 22.3š1.3 24.4š0.8 4 147.5š3.3 136.7š2.0 59.5š2.0 68.9š1.4 39.9š1.6 47.3š1.1 5 234.2š4.2 227.4š2.5 122.5š2.8 146.8š2.0 82.8š2.4 99.4š1.7 6 73.5š2.4 77.5š1.5 45.7š1.7 56.6š1.3 16.7š1.0 26.3š0.9 7 68.1š2.3 67.5š1.4 42.1š1.6 52.6š1.2 26.4š1.3 31.9š0.9 8 20.6š1.2 23.4š0.8 19.9š1.1 26.5š0.9 6.3š0.6 10.4š0.5 Table A.7 The partial þeff and þ0 for FCA-XIX-1, FCA-XIX-2, and FCA-XIX-3 (in pcm).

7A 7B 9C2 9C1

g þeff;g þ0;g þeff;g þ0;g þeff;g þ0;g þeff;g þ0;g 1 6.5š0.6 8.8š0.5 7.6š0.6 10.9š0.5 6.9š0.7 9.4š0.5 20.8š1.2 22.3š0.8 2 59.6š1.9 75.4š1.4 64.6š1.9 80.8š1.5 60.7š1.9 77.7š1.5 101.7š2.6 110.5š1.8 3 23.7š1.2 31.9š0.9 25.8š1.2 33.1š1.0 26.3š1.3 32.3š0.9 62.8š2.1 62.3š1.3 4 56.7š1.9 76.9š1.5 63.6š2.0 82.7š1.5 56.4š1.9 77.1š1.5 133.7š2.9 147.6š2.0 5 120.4š2.7 166.1š2.2 136.4š2.8 178.0š2.2 123.9š2.8 165.7š2.1 225.6š3.8 251.6š2.6 6 47.0š1.7 71.8š1.4 60.5š1.9 81.3š1.5 49.3š1.8 71.0š1.4 93.8š2.5 105.9š1.7 7 41.8š1.6 57.6š1.3 49.4š1.7 65.2š1.3 42.8š1.6 60.3š1.3 73.4š2.2 82.3š1.5 8 21.6š1.1 33.2š1.0 26.4š1.2 36.1š1.0 21.6š1.2 32.2š0.9 31.9š1.4 39.3š1.0 Table A.8 The partial þeff and þ0 for Sneak-7A, Sneak-7B, Sneak-9C2, and Sneak-9C1 (in pcm).

029 033 046

group g þeff;g þ0;g þeff;g þ0;g þeff;g þ0;g 1 25.4š0.9 22.1š0.5 24.5š0.8 21.7š0.5 24.5š0.8 22.5š0.5 2 113.5š1.8 102.8š1.0 111.3š1.8 104.0š1.0 107.1š1.7 102.3š1.0 3 68.7š1.4 61.3š0.8 64.6š1.4 60.0š0.8 66.1š1.4 61.0š0.8 4 143.6š2.0 130.8š1.2 143.6š2.0 131.0š1.2 141.7š2.0 131.5š1.2 5 241.2š2.7 220.0š1.5 243.7š2.7 221.5š1.5 237.0š2.6 219.8š1.5 6 70.2š1.4 61.4š0.8 66.3š1.4 61.4š0.8 65.6š1.3 60.9š0.8 7 61.7š1.3 54.9š0.7 59.7š1.3 54.8š0.7 59.2š1.3 53.9š0.7 8 17.4š0.7 16.4š0.4 17.7š0.7 16.3š0.4 17.3š0.7 15.8š0.4 Table A.9 The partial þeff and þ0 for Stacy runs 029, 033, and 046 (in pcm).

21616/05.69454/P 15-16 030 125 215

group g þeff;g þ0;g þeff;g þ0;g þeff;g þ0;g 1 25.0š0.9 22.5š0.5 25.9š0.9 22.2š0.5 24.4š0.9 21.9š0.5 2 116.4š1.8 103.3š1.0 114.7š1.9 101.8š1.0 109.4š1.8 101.7š1.0 3 69.9š1.4 61.5š0.8 71.6š1.5 62.5š0.8 67.7š1.4 62.1š0.8 4 145.8š2.1 130.5š1.2 149.7š2.1 131.7š1.2 141.7š2.0 130.9š1.2 5 248.6š2.7 221.2š1.5 251.3š2.8 220.2š1.5 240.5š2.6 218.5š1.5 6 67.2š1.4 61.4š0.8 68.7š1.5 61.8š0.8 66.4š1.4 60.4š0.8 7 59.8š1.3 54.5š0.7 64.5š1.4 55.7š0.8 57.2š1.3 53.4š0.7 8 17.7š0.7 15.5š0.4 19.0š0.8 16.5š0.4 18.2š0.7 16.2š0.4 Table A.10 The partial þeff and þ0 for Stacy runs 030, 125, and 215 (in pcm).

Winco Sheba-II

group g þeff;g þ0;g þeff;g þ0;g 1 29.9š0.8 21.8š0.3 24.9š0.9 21.3š0.5 2 124.5š1.6 101.3š0.7 117.4š2.1 102.2š1.1 3 82.1š1.3 61.1š0.6 72.7š1.6 62.5š0.8 4 169.2š1.9 130.6š0.8 151.8š2.3 133.2š1.2 5 273.1š2.3 215.8š1.0 262.0š3.1 227.8š1.6 6 78.0š1.3 60.6š0.6 72.5š1.6 64.3š0.8 7 68.7š1.2 54.2š0.5 62.4š1.5 56.7š0.8 8 19.7š0.6 15.8š0.3 19.2š0.8 17.6š0.4 Table A.11 The partial þeff and þ0 for the Winco slab tanks and Sheba-II (in pcm).

Proteus SHE-8

group g þeff;g þ0;g þeff;g þ0;g 1 23.2š0.7 22.0š0.5 22.9š3.0 20.8š1.5 2 109.0š1.5 102.5š1.0 106.1š6.3 104.6š3.4 3 67.1š1.2 61.4š0.8 74.0š5.8 62.1š2.6 4 140.7š1.7 132.3š1.2 140.7š7.6 129.4š3.8 5 231.3š2.2 217.5š1.5 249.1š10.1 220.1š5.0 6 64.5š1.2 59.8š0.8 61.9š5.0 59.3š2.6 7 57.1š1.1 54.4š0.7 55.6š4.6 55.3š2.5 8 15.9š0.6 15.4š0.4 15.8š2.4 16.6š1.4 Table A.12 The partial þeff and þ0 for Proteus and SHE-8 (in pcm).

16-16 21616/05.69454/P

Abstract Shielding benchmark calculations were performed with MCNP-4C3 using the new JEFF-3.1 nuclear data library. The benchmarks used were Oktavian (for Al, Co, Cr, Cu, LiF, Mn, Mo, Si, Ti, W, Zr), FNS (for Be, C, Fe, and Pb), LLNL Pulsed Spheres (for concrete), and NIST Water Spheres (for water and Cd). The results of these calculations are reported as graphs of the neutron spectra, and in some cases also the photon spectrum, as well as in graphs and tables of calculated over experimental values of the neutron spectra.

Keywords MCNP-4C3 JEFF-3.1 nuclear data shielding benchmarks

2-52 21616/05.69455/P Contents

1 Introduction 5

2 Brief description of the shielding benchmarks 6 2.1 Oktavian 6 2.2 FNS 7 2.3 LLNL Pulsed Spheres 8 2.4 NIST Water Spheres 9

3 Results 10 3.1 Oktavian 10 3.1.1 Oktavian, Al 10 3.1.2 Oktavian, Co 11 3.1.3 Oktavian, Cr 12 3.1.4 Oktavian, Cu 13 3.1.5 Oktavian, LiF 14 3.1.6 Oktavian, Mn 15 3.1.7 Oktavian, Mo 16 3.1.8 Oktavian, Si 17 3.1.9 Oktavian, Ti 18 3.1.10 Oktavian, W 19 3.1.11 Oktavian, Zr 20 3.2 FNS 21 3.2.1 Calculational methods 21 3.2.2 FNS, Be, 5 cm 21 3.2.3 FNS, Be, 15 cm 23 3.2.4 FNS, C, 5 cm 25 3.2.5 FNS, C, 20 cm 27 3.2.6 FNS, C, 40 cm 29 3.2.7 FNS, N, 20 cm 31 3.2.8 FNS, O, 20 cm 33 3.2.9 FNS, Fe, 5 cm 35 3.2.10 FNS, Fe, 20 cm 37 3.2.11 FNS, Fe, 40 cm 39 3.2.12 FNS, Fe, 60 cm 41 3.2.13 FNS, Pb, 5 cm 43 3.2.14 FNS, Pb, 20 cm 45 3.2.15 FNS, Pb, 40 cm 47 3.3 LLNL Pulsed Spheres 49 3.3.1 LLNL Pulsed Spheres, Concrete, 21 cm 49 3.3.2 LLNL Pulsed Spheres, Concrete, 35.5 cm 49 3.4 NIST Water Spheres 50

4 Summary 51

21616/05.69455/P 3-52 References 52

4-52 21616/05.69455/P 1. Introduction

21616/05.69455/P 5-52 2. Brief description of the shielding benchmarks

2.1 Oktavian

The Oktavian benchmark speciﬁcations are given on the IAEA web site [3]. The measured quantity

0.5 Zr

2.5 0.3

5.0

radius 30.0

Figure 2.1 Schematic drawing of the Oktavian geometry for the Zr benchmark (not to scale). All dimensions are in cm. was the leakage current spectrum from the outer surface of a spherical pile of a material. At the center of a pile a 14 MeV D-T neutron source was located. The leakage spectrum is given in units of neutrons per MeV per source neutron. The materials for which benchmark calculations have been performed are list in Table 2.1. A schematic drawing of one of the models is given in Fig. 2.1 (see Ref. [4]).

material outer diameter aluminum Al 40 cm cobalt Co 40 cm chromium Cr 40 cm copper Cu 61 cm lithium ﬂuoride LiF 61 cm molybdenum Mo 61 cm manganese Mn 61 cm silicon Si 60 cm titanium Ti 40 cm tungsten W 40 cm zirconium Zr 61 cm Table 2.1 The materials used in the Oktavian shielding benchmarks

6-52 21616/05.69455/P 2.2 FNS

The FNS Time-of-Flight benchmark speciﬁcations are given on the IAEA web site [5]. The neu-

detector tally

detector tally zero importance region

detector tally

detector tally d

source detector tally

graphite

Figure 2.2 Schematic drawing of the MCNP model for the FNS benchmark for graphite (not to scale). tron spectrum emrging from slabs of material of varying thickness, was measured at ﬁve different angles. The slabs were placed at 20 cm distance from a 14 MeV D-T neutron source. The materials for which benchmark calculations have been performed are listed in Table 2.2. A schematic drawing of the MCNP model for this benchmark is given in Fig. 2.2 [6].

material slab thickness angles beryllium Be 5 cm 0Ž, 24.9Ž, 42.8Ž, 66.8Ž 15 cm 0Ž, 12.2Ž, 24.9Ž, 42.8Ž, 66.8Ž graphite C 5 cm 0Ž, 24.9Ž, 42.8Ž, 66.8Ž 20 cm 0Ž, 12.2Ž, 24.9Ž, 42.8Ž, 66.8Ž 40 cm 0Ž, 12.2Ž, 24.9Ž, 42.8Ž, 66.8Ž nitrogen N 20 cm 0Ž, 12.2Ž, 24.9Ž, 42.8Ž, 66.8Ž oxygen O 20 cm 0Ž, 12.2Ž, 24.9Ž, 42.8Ž, 66.8Ž iron Fe 5 cm 0Ž, 24.9Ž, 42.8Ž, 66.8Ž 20 cm 0Ž, 12.2Ž, 24.9Ž, 42.8Ž, 66.8Ž 40 cm 0Ž, 12.2Ž, 24.9Ž, 42.8Ž, 66.8Ž 60 cm 0Ž, 12.2Ž, 24.9Ž, 42.8Ž lead Pb 5 cm 0Ž, 24.9Ž, 42.8Ž, 66.8Ž 20 cm 0Ž, 12.2Ž, 24.9Ž, 42.8Ž, 66.8Ž 40 cm 0Ž, 12.2Ž, 24.9Ž, 42.8Ž, 66.8Ž Table 2.2 The materials used in the FNS shielding benchmarks

21616/05.69455/P 7-52 2.3 LLNL Pulsed Spheres

The description of the experiments performed in the Livermore Pulsed Sphere Program is given in Ref. [7]. Time-of-ﬂight measurements were performed for neutrons passing through spherical shells of varying thickness, containing different materials. The source was a 14 MeV D-T neutron source.

5.1

3.8 6

21.0

concrete

Figure 2.3 Schematic drawing of the MCNP model for the LLNL benchmark for concrete (not to scale). All dimensions are in cm.

Calculations were performed for two cases with concrete, one with inner radius of 5.1 cm, outer radius 21.0 cm (see Fig. 2.3), and the other with the same inner radius, but an outer radius of 35.5 cm. The composition of the concrete is given in Table 2.3; the density of the concrete was 2.35 g/cm3. The MCNP model used in the present work was taken over from Ref. [8].

Material at% O 55.7 H 15.1 Si 14.9 Ca 3.6 Al 3.2 C 3.1 Mg 1.8 Na 1.3 other < 1:0 each Table 2.3 The composition of the concrete used in the LLNL Pulsed Sphere experiments.

8-52 21616/05.69455/P 2.4 NIST Water Spheres

A description of the experiments performed at NIST with water and cadmium shielding is given in Ref. [10]. A Cf source was placed at the center of the experiment, and fission foils were placed at two, diametrically opposed positions. Between the source and the fission foils were several combinations of shielding materials, see Table 2.4. The fission foils used were 235U, 239Pu, 238U, 237Np.

Cd Cd

Cf source fission fission foils foils

water

Figure 2.4 Schematic drawing of the MCNP model for the NIST benchmark (not to scale).

radius material (inch) 1.5 2.0 1.5 Cd 2.0 Cd

1.5 H2O 2.0 H2O 2.5 H2O

1.5 H2O+Cd 2.0 H2O+Cd 2.5 H2O+Cd Table 2.4 The radius and material combinations for which experiments have been performed.

21616/05.69455/P 9-52 3. Results

In this section we report on the results of our calculations. These results are reported as graphs of the neutron spectra, and in some cases also the photon spectrum, as well as in graphs and tables of calculated over experimental values of the neutron spectra.

3.1 Oktavian

3.1.1 Oktavian, Al

Benchmark Oktavian, Al Benchmark Oktavian, Al: C/E 10 2 benchmark; neutrons benchmark uncertainty calculation; neutrons calculation; neutrons

1.5 1 source-neutron) ] 2

0.1 C/E [-]

0.5

0.01 Leakage current/lethargy [1/(cm

0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.1 The result for the neutron leakage spectrum for the Oktavian Al benchmark.

Benchmark Oktavian, Al 10 benchmark; photons calculation; photons

1 MeV source-neutron) ] 2

0.1

0.01

Leakage current/lethargy/energy [1/(cm 0.1 1 10 E [MeV]

Figure 3.2 The result for the photon leakage spectrum for the Oktavian Al benchmark.

energy range C/E Al 0.1 1.0 1:01 š 0:01 Al 1.0 5.0 0:90 š 0:01 Al 5.0 10.0 0:90 š 0:01 Al 10.0 20.0 1:06 š 0:01 Table 3.1 C/E values for the neutron spectrum of the Oktavian Al benchmark.

10-52 21616/05.69455/P 3.1.2 Oktavian, Co

Benchmark Oktavian, Co Benchmark Oktavian, Co: C/E 10 2 benchmark; neutrons benchmark uncertainty calculation; neutrons calculation; neutrons

1.5 1 source-neutron) ] 2

0.1 C/E [-]

0.5

0.01 Leakage current/lethargy [1/(cm

0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.3 The result for the neutron leakage spectrum for the Oktavian Co benchmark.

Benchmark Oktavian, Co 10 benchmark; photons calculation; photons

1 MeV source-neutron) ] 2

0.1

0.01

Leakage current/lethargy/energy [1/(cm 0.1 1 10 E [MeV]

Figure 3.4 The result for the photon leakage spectrum for the Oktavian Co benchmark.

energy range C/E Co 0.1 1.0 0:64 š 0:01 Co 1.0 5.0 0:66 š 0:01 Co 5.0 10.0 0:54 š 0:01 Co 10.0 20.0 0:98 š 0:01 Table 3.2 C/E values for the neutron spectrum of the Oktavian Co benchmark.

21616/05.69455/P 11-52 3.1.3 Oktavian, Cr

Benchmark Oktavian, Cr Benchmark Oktavian, Cr: C/E 10 2 benchmark; neutrons benchmark uncertainty calculation; neutrons calculation; neutrons

1.5 1 source-neutron) ] 2

0.1 C/E [-]

0.5

0.01 Leakage current/lethargy [1/(cm

0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.5 The result for the neutron leakage spectrum for the Oktavian Cr benchmark.

Benchmark Oktavian, Cr 10 benchmark; photons calculation; photons

1 MeV source-neutron) ] 2

0.1

0.01

Leakage current/lethargy/energy [1/(cm 0.1 1 10 E [MeV]

Figure 3.6 The result for the photon leakage spectrum for the Oktavian Cr benchmark.

energy range C/E Cr 0.1 1.0 1:09 š 0:02 Cr 1.0 5.0 1:19 š 0:01 Cr 5.0 10.0 1:10 š 0:01 Cr 10.0 20.0 0:95 š 0:01 Table 3.3 C/E values for the neutron spectrum of the Oktavian Cr benchmark.

12-52 21616/05.69455/P 3.1.4 Oktavian, Cu

Benchmark Oktavian, Cu Benchmark Oktavian, Cu: C/E 10 2 benchmark; neutrons benchmark uncertainty calculation; neutrons calculation; neutrons

1.5 1 source-neutron) ] 2

0.1 C/E [-]

0.5

0.01 Leakage current/lethargy [1/(cm

0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.7 The result for the neutron leakage spectrum for the Oktavian Cu benchmark.

Benchmark Oktavian, Cu 10 benchmark; photons calculation; photons

1 MeV source-neutron) ] 2

0.1

0.01

Leakage current/lethargy/energy [1/(cm 0.1 1 10 E [MeV]

Figure 3.8 The result for the photon leakage spectrum for the Oktavian Cu benchmark.

energy range C/E Cu 0.0 0.1 0:27 š 0:05 Cu 0.1 1.0 1:07 š 0:01 Cu 1.0 5.0 1:11 š 0:01 Cu 5.0 10.0 1:36 š 0:02 Cu 10.0 20.0 1:12 š 0:01 Table 3.4 C/E values for the neutron spectrum of the Oktavian Cu benchmark.

21616/05.69455/P 13-52 3.1.5 Oktavian, LiF

Benchmark Oktavian, LiF Benchmark Oktavian, LiF: C/E 10 2 benchmark; neutrons benchmark uncertainty calculation; neutrons calculation; neutrons

1.5 1 source-neutron) ] 2

0.1 C/E [-]

0.5

0.01 Leakage current/lethargy [1/(cm

0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.9 The result for the neutron leakage spectrum for the Oktavian LiF benchmark.

Benchmark Oktavian, LiF 10 benchmark; photons calculation; photons

1 MeV source-neutron) ] 2

0.1

0.01

Leakage current/lethargy/energy [1/(cm 0.1 1 10 E [MeV]

Figure 3.10 The result for the photon leakage spectrum for the Oktavian LiF benchmark.

energy range C/E LiF 0.1 1.0 0:76 š 0:01 LiF 1.0 5.0 0:89 š 0:01 LiF 5.0 10.0 0:88 š 0:01 LiF 10.0 20.0 1:09 š 0:01 Table 3.5 C/E values for the neutron spectrum of the Oktavian LiF benchmark.

14-52 21616/05.69455/P 3.1.6 Oktavian, Mn

Benchmark Oktavian, Mn Benchmark Oktavian, Mn: C/E 10 2 benchmark; neutrons benchmark uncertainty calculation; neutrons calculation; neutrons

1.5 1 source-neutron) ] 2

0.1 C/E [-]

0.5

0.01 Leakage current/lethargy [1/(cm

0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.11 The result for the neutron leakage spectrum for the Oktavian Mn benchmark.

Benchmark Oktavian, Mn 10 benchmark; photons calculation; photons

1 MeV source-neutron) ] 2

0.1

0.01

Leakage current/lethargy/energy [1/(cm 0.1 1 10 E [MeV]

Figure 3.12 The result for the photon leakage spectrum for the Oktavian Mn benchmark.

energy range C/E Mn 0.0 0.1 0:38 š 0:05 Mn 0.1 1.0 1:05 š 0:01 Mn 1.0 5.0 1:09 š 0:01 Mn 5.0 10.0 1:12 š 0:01 Mn 10.0 20.0 1:00 š 0:01 Table 3.6 C/E values for the neutron spectrum of the Oktavian Mn benchmark.

21616/05.69455/P 15-52 3.1.7 Oktavian, Mo

Benchmark Oktavian, Mo Benchmark Oktavian, Mo: C/E 10 2 benchmark; neutrons benchmark uncertainty calculation; neutrons calculation; neutrons

1.5 1 source-neutron) ] 2

0.1 C/E [-]

0.5

0.01 Leakage current/lethargy [1/(cm

0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.13 The result for the neutron leakage spectrum for the Oktavian Mo benchmark.

energy range C/E Mo 0.0 0.1 0:27 š 0:19 Mo 0.1 1.0 0:73 š 0:01 Mo 1.0 5.0 0:81 š 0:01 Mo 5.0 10.0 0:84 š 0:01 Mo 10.0 20.0 1:02 š 0:01 Table 3.7 C/E values for the neutron spectrum of the Oktavian Mo benchmark.

16-52 21616/05.69455/P 3.1.8 Oktavian, Si

Benchmark Oktavian, Si Benchmark Oktavian, Si: C/E 10 2 benchmark; neutrons benchmark uncertainty calculation; neutrons calculation; neutrons

1.5 1 source-neutron) ] 2

0.1 C/E [-]

0.5

0.01 Leakage current/lethargy [1/(cm

0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.14 The result for the neutron leakage spectrum for the Oktavian Si benchmark.

Benchmark Oktavian, Si 10 benchmark; photons calculation; photons

1 MeV source-neutron) ] 2

0.1

0.01

Leakage current/lethargy/energy [1/(cm 0.1 1 10 E [MeV]

Figure 3.15 The result for the photon leakage spectrum for the Oktavian Si benchmark.

energy range C/E Si 0.1 1.0 1:09 š 0:02 Si 1.0 5.0 1:00 š 0:01 Si 5.0 10.0 0:85 š 0:01 Si 10.0 20.0 1:09 š 0:01 Table 3.8 C/E values for the neutron spectrum of the Oktavian Si benchmark.

21616/05.69455/P 17-52 3.1.9 Oktavian, Ti

Benchmark Oktavian, Ti Benchmark Oktavian, Ti: C/E 10 2 benchmark; neutrons benchmark uncertainty calculation; neutrons calculation; neutrons

1.5 1 source-neutron) ] 2

0.1 C/E [-]

0.5

0.01 Leakage current/lethargy [1/(cm

0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.16 The result for the neutron leakage spectrum for the Oktavian Ti benchmark.

Benchmark Oktavian, Ti 10 benchmark; photons calculation; photons

1 MeV source-neutron) ] 2

0.1

0.01

Leakage current/lethargy/energy [1/(cm 0.1 1 10 E [MeV]

Figure 3.17 The result for the photon leakage spectrum for the Oktavian Ti benchmark.

energy range C/E Ti 0.0 0.1 1:17 š 0:11 Ti 0.1 1.0 1:33 š 0:01 Ti 1.0 5.0 1:20 š 0:01 Ti 5.0 10.0 1:14 š 0:01 Ti 10.0 20.0 1:18 š 0:01 Table 3.9 C/E values for the neutron spectrum of the Oktavian Ti benchmark.

18-52 21616/05.69455/P 3.1.10 Oktavian, W

Benchmark Oktavian, W Benchmark Oktavian, W: C/E 10 2 benchmark; neutrons benchmark uncertainty calculation; neutrons calculation; neutrons

1.5 1 source-neutron) ] 2

0.1 C/E [-]

0.5

0.01 Leakage current/lethargy [1/(cm

0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.18 The result for the neutron leakage spectrum for the Oktavian W benchmark.

Benchmark Oktavian, W 10 benchmark; photons calculation; photons

1 MeV source-neutron) ] 2

0.1

0.01

Leakage current/lethargy/energy [1/(cm 0.1 1 10 E [MeV]

Figure 3.19 The result for the photon leakage spectrum for the Oktavian W benchmark.

energy range C/E W 0.0 0.1 0:30 š 0:16 W 0.1 1.0 0:97 š 0:01 W 1.0 5.0 0:76 š 0:01 W 5.0 10.0 0:64 š 0:01 W 10.0 20.0 0:91 š 0:01 Table 3.10 C/E values for the neutron spectrum of the Oktavian W benchmark.

21616/05.69455/P 19-52 3.1.11 Oktavian, Zr

Benchmark Oktavian, Zr Benchmark Oktavian, Zr: C/E 10 2 benchmark; neutrons benchmark uncertainty calculation; neutrons calculation; neutrons

1.5 1 source-neutron) ] 2

0.1 C/E [-]

0.5

0.01 Leakage current/lethargy [1/(cm

0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.20 The result for the neutron leakage spectrum for the Oktavian Zr benchmark.

energy range C/E Zr 0.0 0.1 0:97 š 0:04 Zr 0.1 1.0 1:23 š 0:01 Zr 1.0 5.0 1:06 š 0:01 Zr 5.0 10.0 1:02 š 0:01 Zr 10.0 20.0 1:16 š 0:01 Table 3.11 C/E values for the neutron spectrum of the Oktavian Zr benchmark.

20-52 21616/05.69455/P 3.2 FNS

3.2.1 Calculational methods For each of the experiments mentiond in Section 2 the neutron source spectrum was taken from the underlying referene. Hence, in Oktavian and FNS analyses an isotropic neutron distribution was assumed. However, it is known that the neutron source has a peak energy of 14.8 MeV in the forward direction, and a peak at 13.3 MeV in the backward direction (see Ref. [9]). The assumption of isotropy will slightly inﬂuence the calculated angular spectra for the FNS benchmarks. This inﬂuence is expected to be of the order of several percent only.

3.2.2 FNS, Be, 5 cm

FNS, material: Be, thickness: 5.08 cm, angle: 0 FNS, material: Be, thickness: 5.08 cm, angle: 0 0.01 2

Experimental uncertainty Calculation 0.001 Experiment Calculation 1.5

0.0001 source-neutron) ] 2 1 C/E [-] 1e-05

0.5 1e-06 Flux/lethargy [1/(cm

1e-07 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.21 Neutron spectrum for the FNS, Be, d=5cm benchmark at 0Ž angle.

FNS, material: Be, thickness: 5.08 cm, angle: 24.9 FNS, material: Be, thickness: 5.08 cm, angle: 24.9 0.01 2

Experimental uncertainty Calculation 0.001 Experiment Calculation 1.5

0.0001 source-neutron) ] 2 1 C/E [-] 1e-05

0.5 1e-06 Flux/lethargy [1/(cm

1e-07 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.22 Neutron spectrum for the FNS, Be, d=5cm benchmark at 24.9Ž angle.

21616/05.69455/P 21-52 FNS, material: Be, thickness: 5.08 cm, angle: 41.8 FNS, material: Be, thickness: 5.08 cm, angle: 41.8 0.01 2

Experimental uncertainty Calculation 0.001 Experiment Calculation 1.5

0.0001 source-neutron) ] 2 1 C/E [-] 1e-05

0.5 1e-06 Flux/lethargy [1/(cm

1e-07 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.23 Neutron spectrum for the FNS, Be, d=5cm benchmark at 42.8Ž angle.

FNS, material: Be, thickness: 5.08 cm, angle: 66.8 FNS, material: Be, thickness: 5.08 cm, angle: 66.8 0.01 2

Experimental uncertainty Calculation 0.001 Experiment Calculation 1.5

0.0001 source-neutron) ] 2 1 C/E [-] 1e-05

0.5 1e-06 Flux/lethargy [1/(cm

1e-07 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.24 Neutron spectrum for the FNS, Be, d=5cm benchmark at 66.8Ž angle.

energy range 0Ž 12.2Ž 24.9Ž 42.8Ž 66.8Ž 0.1 1.0 0.95 š 0.01 0.93 š 0.01 0.95 š 0.01 0.87 š 0.01 1.0 5.0 0.90 š 0.01 0.93 š 0.01 0.91 š 0.01 0.84 š 0.01 5.0 10.0 0.81 š 0.01 0.78 š 0.01 0.81 š 0.01 0.80 š 0.01 10.0 20.0 0.96 š 0.01 0.96 š 0.01 0.94 š 0.01 0.69 š 0.01 Table 3.12 C/E values for the neutron spectrum of the FNS Be 5cm benchmark.

22-52 21616/05.69455/P 3.2.3 FNS, Be, 15 cm

FNS, material: Be, thickness: 15.24 cm, angle: 0 FNS, material: Be, thickness: 15.24 cm, angle: 0 2 0.001 Experimental uncertainty Experiment Calculation Calculation 1.5 0.0001 source-neutron) ] 2 1

1e-05 C/E [-]

1e-06 0.5 Flux/lethargy [1/(cm

1e-07 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.25 Neutron spectrum for the FNS, Be, d=15cm benchmark at 0Ž angle.

FNS, material: Be, thickness: 15.24 cm, angle: 12.2 FNS, material: Be, thickness: 15.24 cm, angle: 12.2 2 0.001 Experimental uncertainty Experiment Calculation Calculation 1.5 0.0001 source-neutron) ] 2 1

1e-05 C/E [-]

1e-06 0.5 Flux/lethargy [1/(cm

1e-07 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.26 Neutron spectrum for the FNS, Be, d=15cm benchmark at 12.2Ž angle.

FNS, material: Be, thickness: 15.24 cm, angle: 24.9 FNS, material: Be, thickness: 15.24 cm, angle: 24.9 2 0.001 Experimental uncertainty Experiment Calculation Calculation 1.5 0.0001 source-neutron) ] 2 1

1e-05 C/E [-]

1e-06 0.5 Flux/lethargy [1/(cm

1e-07 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.27 Neutron spectrum for the FNS, Be, d=15cm benchmark at 24.9Ž angle.

21616/05.69455/P 23-52 FNS, material: Be, thickness: 15.24 cm, angle: 41.8 FNS, material: Be, thickness: 15.24 cm, angle: 41.8 2 0.001 Experimental uncertainty Experiment Calculation Calculation 1.5 0.0001 source-neutron) ] 2 1

1e-05 C/E [-]

1e-06 0.5 Flux/lethargy [1/(cm

1e-07 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.28 Neutron spectrum for the FNS, Be, d=15cm benchmark at 42.8Ž angle.

FNS, material: Be, thickness: 15.24 cm, angle: 66.8 FNS, material: Be, thickness: 15.24 cm, angle: 66.8 2 0.001 Experimental uncertainty Experiment Calculation Calculation 1.5 0.0001 source-neutron) ] 2 1

1e-05 C/E [-]

1e-06 0.5 Flux/lethargy [1/(cm

1e-07 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.29 Neutron spectrum for the FNS, Be, d=15cm benchmark at 66.8Ž angle.

energy range 0Ž 12.2Ž 24.9Ž 42.8Ž 66.8Ž 0.1 1.0 0.85 š 0.01 0.97 š 0.01 0.95 š 0.01 0.94 š 0.01 0.89 š 0.01 1.0 5.0 0.87 š 0.01 0.93 š 0.01 0.91 š 0.01 0.87 š 0.01 0.84 š 0.01 5.0 10.0 0.75 š 0.01 0.81 š 0.01 0.83 š 0.01 0.81 š 0.01 0.78 š 0.02 10.0 20.0 0.98 š 0.01 0.81 š 0.01 1.01 š 0.01 0.83 š 0.01 0.80 š 0.02 Table 3.13 C/E values for the neutron spectrum of the FNS Be 15cm benchmark.

24-52 21616/05.69455/P 3.2.4 FNS, C, 5 cm

FNS, material: C, thickness: 5.06 cm, angle: 0 FNS, material: C, thickness: 5.06 cm, angle: 0 0.01 2

Experimental uncertainty Calculation 0.001 Experiment Calculation 1.5

0.0001 source-neutron) ] 2 1 C/E [-] 1e-05

0.5 1e-06 Flux/lethargy [1/(cm

1e-07 0 1 10 1 10 E [MeV] E [MeV]

Figure 3.30 Neutron spectrum for the FNS, C, d=5cm benchmark at 0Ž angle.

FNS, material: C, thickness: 5.06 cm, angle: 24.9 FNS, material: C, thickness: 5.06 cm, angle: 24.9 0.01 2

Experimental uncertainty Calculation 0.001 Experiment Calculation 1.5

0.0001 source-neutron) ] 2 1 C/E [-] 1e-05

0.5 1e-06 Flux/lethargy [1/(cm

1e-07 0 1 10 1 10 E [MeV] E [MeV]

Figure 3.31 Neutron spectrum for the FNS, C, d=5cm benchmark at 24.9Ž angle.

FNS, material: C, thickness: 5.06 cm, angle: 41.8 FNS, material: C, thickness: 5.06 cm, angle: 41.8 0.01 2

Experimental uncertainty Calculation 0.001 Experiment Calculation 1.5

0.0001 source-neutron) ] 2 1 C/E [-] 1e-05

0.5 1e-06 Flux/lethargy [1/(cm

1e-07 0 1 10 1 10 E [MeV] E [MeV]

Figure 3.32 Neutron spectrum for the FNS, C, d=5cm benchmark at 42.8Ž angle.

21616/05.69455/P 25-52 FNS, material: C, thickness: 5.06 cm, angle: 66.8 FNS, material: C, thickness: 5.06 cm, angle: 66.8 0.01 2

Experimental uncertainty Calculation 0.001 Experiment Calculation 1.5

0.0001 source-neutron) ] 2 1 C/E [-] 1e-05

0.5 1e-06 Flux/lethargy [1/(cm

1e-07 0 1 10 1 10 E [MeV] E [MeV]

Figure 3.33 Neutron spectrum for the FNS, C, d=5cm benchmark at 66.8Ž angle.

energy range 0Ž 12.2Ž 24.9Ž 42.8Ž 66.8Ž 0.1 1.0 0.81 š 0.01 0.84 š 0.01 0.83 š 0.01 0.77 š 0.01 1.0 5.0 0.94 š 0.01 0.73 š 0.01 0.79 š 0.01 0.76 š 0.01 5.0 10.0 1.01 š 0.01 0.83 š 0.01 0.93 š 0.01 0.95 š 0.01 10.0 20.0 1.01 š 0.01 1.23 š 0.01 1.39 š 0.01 1.01 š 0.01 Table 3.14 C/E values for the neutron spectrum of the FNS C 5cm benchmark.

26-52 21616/05.69455/P 3.2.5 FNS, C, 20 cm

FNS, material: C, thickness: 20.24 cm, angle: 0 FNS, material: C, thickness: 20.24 cm, angle: 0 0.01 2

Experimental uncertainty Calculation 0.001 Experiment Calculation 1.5

0.0001 source-neutron) ] 2 1 C/E [-] 1e-05

0.5 1e-06 Flux/lethargy [1/(cm

1e-07 0 1 10 1 10 E [MeV] E [MeV]

Figure 3.34 Neutron spectrum for the FNS, C, d=20cm benchmark at 0Ž angle.

FNS, material: C, thickness: 20.24 cm, angle: 12.2 FNS, material: C, thickness: 20.24 cm, angle: 12.2 0.01 2

Experimental uncertainty Calculation 0.001 Experiment Calculation 1.5

0.0001 source-neutron) ] 2 1 C/E [-] 1e-05

0.5 1e-06 Flux/lethargy [1/(cm

1e-07 0 1 10 1 10 E [MeV] E [MeV]

Figure 3.35 Neutron spectrum for the FNS, C, d=20cm benchmark at 12.2Ž angle.

FNS, material: C, thickness: 20.24 cm, angle: 24.9 FNS, material: C, thickness: 20.24 cm, angle: 24.9 0.01 2

Experimental uncertainty Calculation 0.001 Experiment Calculation 1.5

0.0001 source-neutron) ] 2 1 C/E [-] 1e-05

0.5 1e-06 Flux/lethargy [1/(cm

1e-07 0 1 10 1 10 E [MeV] E [MeV]

Figure 3.36 Neutron spectrum for the FNS, C, d=20cm benchmark at 24.9Ž angle.

21616/05.69455/P 27-52 FNS, material: C, thickness: 20.24 cm, angle: 41.8 FNS, material: C, thickness: 20.24 cm, angle: 41.8 0.01 2

Experimental uncertainty Calculation 0.001 Experiment Calculation 1.5

0.0001 source-neutron) ] 2 1 C/E [-] 1e-05

0.5 1e-06 Flux/lethargy [1/(cm

1e-07 0 1 10 1 10 E [MeV] E [MeV]

Figure 3.37 Neutron spectrum for the FNS, C, d=20cm benchmark at 42.8Ž angle.

FNS, material: C, thickness: 20.24 cm, angle: 66.8 FNS, material: C, thickness: 20.24 cm, angle: 66.8 0.01 2

Experimental uncertainty Calculation 0.001 Experiment Calculation 1.5

0.0001 source-neutron) ] 2 1 C/E [-] 1e-05

0.5 1e-06 Flux/lethargy [1/(cm

1e-07 0 1 10 1 10 E [MeV] E [MeV]

Figure 3.38 Neutron spectrum for the FNS, C, d=20cm benchmark at 66.8Ž angle.

energy range 0Ž 12.2Ž 24.9Ž 42.8Ž 66.8Ž 0.1 1.0 0.72 š 0.01 0.82 š 0.01 0.76 š 0.01 0.72 š 0.01 0.75 š 0.01 1.0 5.0 0.77 š 0.01 0.81 š 0.01 0.81 š 0.01 0.84 š 0.01 0.83 š 0.01 5.0 10.0 0.90 š 0.01 1.01 š 0.01 0.89 š 0.01 1.01 š 0.01 1.04 š 0.01 10.0 20.0 0.98 š 0.01 1.07 š 0.01 1.20 š 0.01 1.24 š 0.01 1.07 š 0.01 Table 3.15 C/E values for the neutron spectrum of the FNS C 20cm benchmark.

28-52 21616/05.69455/P 3.2.6 FNS, C, 40 cm

FNS, material: C, thickness: 40.48 cm, angle: 0 FNS, material: C, thickness: 40.48 cm, angle: 0 0.001 2

Experimental uncertainty Experiment Calculation Calculation 0.0001 1.5

1e-05 source-neutron) ] 2 1 C/E [-] 1e-06

0.5 1e-07 Flux/lethargy [1/(cm

1e-08 0 1 10 1 10 E [MeV] E [MeV]

Figure 3.39 Neutron spectrum for the FNS, C, d=40cm benchmark at 0Ž angle.

FNS, material: C, thickness: 40.48 cm, angle: 12.2 FNS, material: C, thickness: 40.48 cm, angle: 12.2 0.001 2

Experimental uncertainty Experiment Calculation Calculation 0.0001 1.5

1e-05 source-neutron) ] 2 1 C/E [-] 1e-06

0.5 1e-07 Flux/lethargy [1/(cm

1e-08 0 1 10 1 10 E [MeV] E [MeV]

Figure 3.40 Neutron spectrum for the FNS, C, d=40cm benchmark at 12.2Ž angle.

FNS, material: C, thickness: 40.48 cm, angle: 24.9 FNS, material: C, thickness: 40.48 cm, angle: 24.9 0.001 2

Experimental uncertainty Experiment Calculation Calculation 0.0001 1.5

1e-05 source-neutron) ] 2 1 C/E [-] 1e-06

0.5 1e-07 Flux/lethargy [1/(cm

1e-08 0 1 10 1 10 E [MeV] E [MeV]

Figure 3.41 Neutron spectrum for the FNS, C, d=40cm benchmark at 24.9Ž angle.

21616/05.69455/P 29-52 FNS, material: C, thickness: 40.48 cm, angle: 41.8 FNS, material: C, thickness: 40.48 cm, angle: 41.8 0.001 2

Experimental uncertainty Experiment Calculation Calculation 0.0001 1.5

1e-05 source-neutron) ] 2 1 C/E [-] 1e-06

0.5 1e-07 Flux/lethargy [1/(cm

1e-08 0 1 10 1 10 E [MeV] E [MeV]

Figure 3.42 Neutron spectrum for the FNS, C, d=40cm benchmark at 42.8Ž angle.

FNS, material: C, thickness: 40.48 cm, angle: 66.8 FNS, material: C, thickness: 40.48 cm, angle: 66.8 0.001 2

Experimental uncertainty Experiment Calculation Calculation 0.0001 1.5

1e-05 source-neutron) ] 2 1 C/E [-] 1e-06

0.5 1e-07 Flux/lethargy [1/(cm

1e-08 0 1 10 1 10 E [MeV] E [MeV]

Figure 3.43 Neutron spectrum for the FNS, C, d=40cm benchmark at 66.8Ž angle.

energy range 0Ž 12.2Ž 24.9Ž 42.8Ž 66.8Ž 0.1 1.0 0.71 š 0.01 0.70 š 0.01 0.80 š 0.01 0.60 š 0.02 0.94 š 0.02 1.0 5.0 0.78 š 0.01 0.90 š 0.01 0.91 š 0.01 0.86 š 0.01 0.91 š 0.01 5.0 10.0 0.85 š 0.01 1.02 š 0.01 1.01 š 0.01 1.05 š 0.02 1.12 š 0.02 10.0 20.0 1.03 š 0.01 1.08 š 0.01 1.09 š 0.01 1.14 š 0.02 0.94 š 0.04 Table 3.16 C/E values for the neutron spectrum of the FNS C 40cm benchmark.

30-52 21616/05.69455/P 3.2.7 FNS, N, 20 cm

FNS, material: N, thickness: 20.0 cm, angle: 0 FNS, material: N, thickness: 20.0 cm, angle: 0 0.001 2

Experimental uncertainty Calculation 0.0001 Experiment Calculation 1.5

1e-05 source-neutron) ] 2 1 C/E [-] 1e-06

0.5 1e-07 Flux/lethargy [1/(cm

1e-08 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.44 Neutron spectrum for the FNS, N, d=20cm benchmark at 0Ž angle.

FNS, material: N, thickness: 20.0 cm, angle: 12.2 FNS, material: N, thickness: 20.0 cm, angle: 12.2 0.001 2

Experimental uncertainty Calculation 0.0001 Experiment Calculation 1.5

1e-05 source-neutron) ] 2 1 C/E [-] 1e-06

0.5 1e-07 Flux/lethargy [1/(cm

1e-08 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.45 Neutron spectrum for the FNS, N, d=20cm benchmark at 12.2Ž angle.

FNS, material: N, thickness: 20.0 cm, angle: 24.9 FNS, material: N, thickness: 20.0 cm, angle: 24.9 0.001 2

Experimental uncertainty Calculation 0.0001 Experiment Calculation 1.5

1e-05 source-neutron) ] 2 1 C/E [-] 1e-06

0.5 1e-07 Flux/lethargy [1/(cm

1e-08 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.46 Neutron spectrum for the FNS, N, d=20cm benchmark at 24.9Ž angle.

21616/05.69455/P 31-52 FNS, material: N, thickness: 20.0 cm, angle: 41.8 FNS, material: N, thickness: 20.0 cm, angle: 41.8 0.001 2

Experimental uncertainty Calculation 0.0001 Experiment Calculation 1.5

1e-05 source-neutron) ] 2 1 C/E [-] 1e-06

0.5 1e-07 Flux/lethargy [1/(cm

1e-08 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.47 Neutron spectrum for the FNS, N, d=20cm benchmark at 42.8Ž angle.

FNS, material: N, thickness: 20.0 cm, angle: 66.8 FNS, material: N, thickness: 20.0 cm, angle: 66.8 0.001 2

Experimental uncertainty Calculation 0.0001 Experiment Calculation 1.5

1e-05 source-neutron) ] 2 1 C/E [-] 1e-06

0.5 1e-07 Flux/lethargy [1/(cm

1e-08 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.48 Neutron spectrum for the FNS, N, d=20cm benchmark at 66.8Ž angle.

energy range 0Ž 12.2Ž 24.9Ž 42.8Ž 66.8Ž 0.1 1.0 0.80 š 0.01 0.73 š 0.01 0.69 š 0.01 0.71 š 0.01 0.56 š 0.01 1.0 5.0 0.86 š 0.01 0.79 š 0.01 0.75 š 0.01 0.77 š 0.01 0.64 š 0.01 5.0 10.0 0.92 š 0.01 0.89 š 0.01 0.80 š 0.01 0.77 š 0.01 0.57 š 0.02 10.0 20.0 0.85 š 0.01 0.76 š 0.01 0.57 š 0.01 0.50 š 0.01 0.41 š 0.01 Table 3.17 C/E values for the neutron spectrum of the FNS N 20cm benchmark.

32-52 21616/05.69455/P 3.2.8 FNS, O, 20 cm

FNS, material: O, thickness: 20.0 cm, angle: 0 FNS, material: O, thickness: 20.0 cm, angle: 0 0.001 2

Experimental uncertainty Calculation 0.0001 Experiment Calculation 1.5

1e-05 source-neutron) ] 2 1 C/E [-] 1e-06

0.5 1e-07 Flux/lethargy [1/(cm

1e-08 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.49 Neutron spectrum for the FNS, O, d=20cm benchmark at 0Ž angle.

FNS, material: O, thickness: 20.0 cm, angle: 12.2 FNS, material: O, thickness: 20.0 cm, angle: 12.2 0.001 2

Experimental uncertainty Calculation 0.0001 Experiment Calculation 1.5

1e-05 source-neutron) ] 2 1 C/E [-] 1e-06

0.5 1e-07 Flux/lethargy [1/(cm

1e-08 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.50 Neutron spectrum for the FNS, O, d=20cm benchmark at 12.2Ž angle.

FNS, material: O, thickness: 20.0 cm, angle: 24.9 FNS, material: O, thickness: 20.0 cm, angle: 24.9 0.001 2

Experimental uncertainty Calculation 0.0001 Experiment Calculation 1.5

1e-05 source-neutron) ] 2 1 C/E [-] 1e-06

0.5 1e-07 Flux/lethargy [1/(cm

1e-08 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.51 Neutron spectrum for the FNS, O, d=20cm benchmark at 24.9Ž angle.

21616/05.69455/P 33-52 FNS, material: O, thickness: 20.0 cm, angle: 41.8 FNS, material: O, thickness: 20.0 cm, angle: 41.8 0.001 2

Experimental uncertainty Calculation 0.0001 Experiment Calculation 1.5

1e-05 source-neutron) ] 2 1 C/E [-] 1e-06

0.5 1e-07 Flux/lethargy [1/(cm

1e-08 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.52 Neutron spectrum for the FNS, O, d=20cm benchmark at 42.8Ž angle.

FNS, material: O, thickness: 20.0 cm, angle: 66.8 FNS, material: O, thickness: 20.0 cm, angle: 66.8 0.001 2

Experimental uncertainty Calculation 0.0001 Experiment Calculation 1.5

1e-05 source-neutron) ] 2 1 C/E [-] 1e-06

0.5 1e-07 Flux/lethargy [1/(cm

1e-08 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.53 Neutron spectrum for the FNS, O, d=20cm benchmark at 66.8Ž angle.

energy range 0Ž 12.2Ž 24.9Ž 42.8Ž 66.8Ž 0.1 1.0 0.80 š 0.01 0.91 š 0.01 0.79 š 0.01 0.70 š 0.01 0.63 š 0.01 1.0 5.0 0.84 š 0.01 1.03 š 0.01 0.92 š 0.01 0.78 š 0.01 0.58 š 0.01 5.0 10.0 1.06 š 0.01 1.30 š 0.01 0.93 š 0.01 0.72 š 0.01 0.54 š 0.02 10.0 20.0 0.89 š 0.01 0.81 š 0.01 0.55 š 0.01 0.50 š 0.01 0.40 š 0.01 Table 3.18 C/E values for the neutron spectrum of the FNS O 20cm benchmark.

34-52 21616/05.69455/P 3.2.9 FNS, Fe, 5 cm

FNS, material: Fe, thickness: 5.00 cm, angle: 0 FNS, material: Fe, thickness: 5.00 cm, angle: 0 0.01 2

Experimental uncertainty Calculation 0.001 Experiment Calculation 1.5

0.0001 source-neutron) ] 2 1 C/E [-] 1e-05

0.5 1e-06 Flux/lethargy [1/(cm

1e-07 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.54 Neutron spectrum for the FNS, Fe, d=5cm benchmark at 0Ž angle.

FNS, material: Fe, thickness: 5.00 cm, angle: 24.9 FNS, material: Fe, thickness: 5.00 cm, angle: 24.9 0.01 2

Experimental uncertainty Calculation 0.001 Experiment Calculation 1.5

0.0001 source-neutron) ] 2 1 C/E [-] 1e-05

0.5 1e-06 Flux/lethargy [1/(cm

1e-07 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.55 Neutron spectrum for the FNS, Fe, d=5cm benchmark at 24.9Ž angle.

FNS, material: Fe, thickness: 5.00 cm, angle: 41.8 FNS, material: Fe, thickness: 5.00 cm, angle: 41.8 0.01 2

Experimental uncertainty Calculation 0.001 Experiment Calculation 1.5

0.0001 source-neutron) ] 2 1 C/E [-] 1e-05

0.5 1e-06 Flux/lethargy [1/(cm

1e-07 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.56 Neutron spectrum for the FNS, Fe, d=5cm benchmark at 42.8Ž angle.

21616/05.69455/P 35-52 FNS, material: Fe, thickness: 5.00 cm, angle: 66.8 FNS, material: Fe, thickness: 5.00 cm, angle: 66.8 0.01 2

Experimental uncertainty Calculation 0.001 Experiment Calculation 1.5

0.0001 source-neutron) ] 2 1 C/E [-] 1e-05

0.5 1e-06 Flux/lethargy [1/(cm

1e-07 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.57 Neutron spectrum for the FNS, Fe, d=5cm benchmark at 66.8Ž angle.

energy range 0Ž 12.2Ž 24.9Ž 42.8Ž 66.8Ž 0.1 1.0 0.83 š 0.01 0.89 š 0.01 0.88 š 0.01 0.78 š 0.01 1.0 5.0 0.92 š 0.01 0.90 š 0.01 0.87 š 0.01 0.85 š 0.01 5.0 10.0 0.98 š 0.01 0.98 š 0.02 0.85 š 0.02 0.72 š 0.02 10.0 20.0 0.92 š 0.01 1.27 š 0.01 0.84 š 0.01 0.58 š 0.02 Table 3.19 C/E values for the neutron spectrum of the FNS Fe 5cm benchmark.

36-52 21616/05.69455/P 3.2.10 FNS, Fe, 20 cm

FNS, material: Fe, thickness: 20.0 cm, angle: 0 FNS, material: Fe, thickness: 20.0 cm, angle: 0 0.001 2

Experimental uncertainty Calculation 0.0001 Experiment Calculation 1.5

1e-05 source-neutron) ] 2 1 C/E [-] 1e-06

0.5 1e-07 Flux/lethargy [1/(cm

1e-08 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.58 Neutron spectrum for the FNS, Fe, d=20cm benchmark at 0Ž angle.

FNS, material: Fe, thickness: 20.0 cm, angle: 12.2 FNS, material: Fe, thickness: 20.0 cm, angle: 12.2 0.001 2

Experimental uncertainty Calculation 0.0001 Experiment Calculation 1.5

1e-05 source-neutron) ] 2 1 C/E [-] 1e-06

0.5 1e-07 Flux/lethargy [1/(cm

1e-08 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.59 Neutron spectrum for the FNS, Fe, d=20cm benchmark at 12.2Ž angle.

FNS, material: Fe, thickness: 20.0 cm, angle: 24.9 FNS, material: Fe, thickness: 20.0 cm, angle: 24.9 0.001 2

Experimental uncertainty Calculation 0.0001 Experiment Calculation 1.5

1e-05 source-neutron) ] 2 1 C/E [-] 1e-06

0.5 1e-07 Flux/lethargy [1/(cm

1e-08 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.60 Neutron spectrum for the FNS, Fe, d=20cm benchmark at 24.9Ž angle.

21616/05.69455/P 37-52 FNS, material: Fe, thickness: 20.0 cm, angle: 41.8 FNS, material: Fe, thickness: 20.0 cm, angle: 41.8 0.001 2

Experimental uncertainty Calculation 0.0001 Experiment Calculation 1.5

1e-05 source-neutron) ] 2 1 C/E [-] 1e-06

0.5 1e-07 Flux/lethargy [1/(cm

1e-08 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.61 Neutron spectrum for the FNS, Fe, d=20cm benchmark at 42.8Ž angle.

FNS, material: Fe, thickness: 20.0 cm, angle: 66.8 FNS, material: Fe, thickness: 20.0 cm, angle: 66.8 0.001 2

Experimental uncertainty Calculation 0.0001 Experiment Calculation 1.5

1e-05 source-neutron) ] 2 1 C/E [-] 1e-06

0.5 1e-07 Flux/lethargy [1/(cm

1e-08 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.62 Neutron spectrum for the FNS, Fe, d=20cm benchmark at 66.8Ž angle.

energy range 0Ž 12.2Ž 24.9Ž 42.8Ž 66.8Ž 0.1 1.0 0.84 š 0.01 0.93 š 0.01 0.93 š 0.01 0.89 š 0.01 0.87 š 0.01 1.0 5.0 0.91 š 0.01 0.93 š 0.01 0.91 š 0.01 0.90 š 0.01 0.86 š 0.01 5.0 10.0 0.94 š 0.03 1.01 š 0.03 0.93 š 0.04 0.91 š 0.04 0.92 š 0.07 10.0 20.0 0.85 š 0.01 0.95 š 0.01 1.06 š 0.02 0.91 š 0.03 0.67 š 0.05 Table 3.20 C/E values for the neutron spectrum of the FNS Fe 20cm benchmark.

38-52 21616/05.69455/P 3.2.11 FNS, Fe, 40 cm

FNS, material: Fe, thickness: 40.0 cm, angle: 0 FNS, material: Fe, thickness: 40.0 cm, angle: 0 0.0001 2

Experimental uncertainty Calculation 1e-05 Experiment Calculation 1.5

1e-06 source-neutron) ] 2 1 C/E [-] 1e-07

0.5 1e-08 Flux/lethargy [1/(cm

1e-09 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.63 Neutron spectrum for the FNS, Fe, d=40cm benchmark at 0Ž angle.

FNS, material: Fe, thickness: 40.0 cm, angle: 12.2 FNS, material: Fe, thickness: 40.0 cm, angle: 12.2 0.0001 2

Experimental uncertainty Calculation 1e-05 Experiment Calculation 1.5

1e-06 source-neutron) ] 2 1 C/E [-] 1e-07

0.5 1e-08 Flux/lethargy [1/(cm

1e-09 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.64 Neutron spectrum for the FNS, Fe, d=40cm benchmark at 12.2Ž angle.

FNS, material: Fe, thickness: 40.0 cm, angle: 24.9 FNS, material: Fe, thickness: 40.0 cm, angle: 24.9 0.0001 2

Experimental uncertainty Calculation 1e-05 Experiment Calculation 1.5

1e-06 source-neutron) ] 2 1 C/E [-] 1e-07

0.5 1e-08 Flux/lethargy [1/(cm

1e-09 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.65 Neutron spectrum for the FNS, Fe, d=40cm benchmark at 24.9Ž angle.

21616/05.69455/P 39-52 FNS, material: Fe, thickness: 40.0 cm, angle: 41.8 FNS, material: Fe, thickness: 40.0 cm, angle: 41.8 0.0001 2

Experimental uncertainty Calculation 1e-05 Experiment Calculation 1.5

1e-06 source-neutron) ] 2 1 C/E [-] 1e-07

0.5 1e-08 Flux/lethargy [1/(cm

1e-09 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.66 Neutron spectrum for the FNS, Fe, d=40cm benchmark at 42.8Ž angle.

FNS, material: Fe, thickness: 40.0 cm, angle: 66.8 FNS, material: Fe, thickness: 40.0 cm, angle: 66.8 0.0001 2

Experimental uncertainty Calculation 1e-05 Experiment Calculation 1.5

1e-06 source-neutron) ] 2 1 C/E [-] 1e-07

0.5 1e-08 Flux/lethargy [1/(cm

1e-09 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.67 Neutron spectrum for the FNS, Fe, d=40cm benchmark at 66.8Ž angle.

energy range 0Ž 12.2Ž 24.9Ž 42.8Ž 66.8Ž 0.1 1.0 0.91 š 0.01 0.98 š 0.01 0.92 š 0.01 0.87 š 0.01 0.81 š 0.01 1.0 5.0 0.97 š 0.01 1.03 š 0.01 1.02 š 0.02 0.92 š 0.02 0.80 š 0.03 5.0 10.0 0.85 š 0.13 1.06 š 0.13 0.91 š 0.15 0.54 š 0.16 10.0 20.0 0.81 š 0.02 0.77 š 0.05 0.79 š 0.08 0.68 š 0.14 0.33 š 0.33 Table 3.21 C/E values for the neutron spectrum of the FNS Fe 40cm benchmark.

40-52 21616/05.69455/P 3.2.12 FNS, Fe, 60 cm

FNS, material: Fe, thickness: 60.0 cm, angle: 0 FNS, material: Fe, thickness: 60.0 cm, angle: 0 0.0001 2

Experimental uncertainty Calculation 1e-05 Experiment Calculation 1.5

1e-06 source-neutron) ] 2 1 C/E [-] 1e-07

0.5 1e-08 Flux/lethargy [1/(cm

1e-09 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.68 Neutron spectrum for the FNS, Fe, d=60cm benchmark at 0Ž angle.

FNS, material: Fe, thickness: 60.0 cm, angle: 12.2 FNS, material: Fe, thickness: 60.0 cm, angle: 12.2 0.0001 2

Experimental uncertainty Calculation 1e-05 Experiment Calculation 1.5

1e-06 source-neutron) ] 2 1 C/E [-] 1e-07

0.5 1e-08 Flux/lethargy [1/(cm

1e-09 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.69 Neutron spectrum for the FNS, Fe, d=60cm benchmark at 12.2Ž angle.

FNS, material: Fe, thickness: 60.0 cm, angle: 24.9 FNS, material: Fe, thickness: 60.0 cm, angle: 24.9 0.0001 2

Experimental uncertainty Calculation 1e-05 Experiment Calculation 1.5

1e-06 source-neutron) ] 2 1 C/E [-] 1e-07

0.5 1e-08 Flux/lethargy [1/(cm

1e-09 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.70 Neutron spectrum for the FNS, Fe, d=60cm benchmark at 24.9Ž angle.

21616/05.69455/P 41-52 FNS, material: Fe, thickness: 60.0 cm, angle: 41.8 FNS, material: Fe, thickness: 60.0 cm, angle: 41.8 0.0001 2

Experimental uncertainty Calculation 1e-05 Experiment Calculation 1.5

1e-06 source-neutron) ] 2 1 C/E [-] 1e-07

0.5 1e-08 Flux/lethargy [1/(cm

1e-09 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.71 Neutron spectrum for the FNS, Fe, d=60cm benchmark at 42.8Ž angle.

energy range 0Ž 12.2Ž 24.9Ž 42.8Ž 0.1 1.0 0.86 š 0.01 0.95 š 0.01 0.88 š 0.01 0.91 š 0.01 1.0 5.0 0.73 š 0.03 0.84 š 0.04 0.98 š 0.05 0.96 š 0.07 5.0 10.0 0.36 š 0.31 0.11 š 0.37 0.03 š 0.22 10.0 20.0 0.62 š 0.11 0.73 š 0.26 0.94 š 0.44 3.50 š 0.57 Table 3.22 C/E values for the neutron spectrum of the FNS Fe 60cm benchmark.

42-52 21616/05.69455/P 3.2.13 FNS, Pb, 5 cm

FNS, material: Pb, thickness: 5.08 cm, angle: 0 FNS, material: Pb, thickness: 5.08 cm, angle: 0 0.01 2

Experimental uncertainty Calculation 0.001 Experiment Calculation 1.5

0.0001 source-neutron) ] 2 1 C/E [-] 1e-05

0.5 1e-06 Flux/lethargy [1/(cm

1e-07 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.72 Neutron spectrum for the FNS, Pb, d=5cm benchmark at 0Ž angle.

FNS, material: Pb, thickness: 5.08 cm, angle: 24.9 FNS, material: Pb, thickness: 5.08 cm, angle: 24.9 0.01 2

Experimental uncertainty Calculation 0.001 Experiment Calculation 1.5

0.0001 source-neutron) ] 2 1 C/E [-] 1e-05

0.5 1e-06 Flux/lethargy [1/(cm

1e-07 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.73 Neutron spectrum for the FNS, Pb, d=5cm benchmark at 24.9Ž angle.

FNS, material: Pb, thickness: 5.08 cm, angle: 41.8 FNS, material: Pb, thickness: 5.08 cm, angle: 41.8 0.01 2

Experimental uncertainty Calculation 0.001 Experiment Calculation 1.5

0.0001 source-neutron) ] 2 1 C/E [-] 1e-05

0.5 1e-06 Flux/lethargy [1/(cm

1e-07 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.74 Neutron spectrum for the FNS, Pb, d=5cm benchmark at 42.8Ž angle.

21616/05.69455/P 43-52 FNS, material: Pb, thickness: 5.08 cm, angle: 66.8 FNS, material: Pb, thickness: 5.08 cm, angle: 66.8 0.01 2

Experimental uncertainty Calculation 0.001 Experiment Calculation 1.5

0.0001 source-neutron) ] 2 1 C/E [-] 1e-05

0.5 1e-06 Flux/lethargy [1/(cm

1e-07 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.75 Neutron spectrum for the FNS, Pb, d=5cm benchmark at 66.8Ž angle.

energy range 0Ž 12.2Ž 24.9Ž 42.8Ž 66.8Ž 0.1 1.0 0.84 š 0.01 0.92 š 0.01 0.89 š 0.01 0.84 š 0.01 1.0 5.0 0.94 š 0.01 0.93 š 0.01 0.87 š 0.01 0.84 š 0.01 5.0 10.0 0.91 š 0.01 0.99 š 0.02 0.84 š 0.02 0.87 š 0.02 10.0 20.0 0.93 š 0.01 1.16 š 0.01 0.81 š 0.01 0.67 š 0.02 Table 3.23 C/E values for the neutron spectrum of the FNS Pb 5cm benchmark.

44-52 21616/05.69455/P 3.2.14 FNS, Pb, 20 cm

FNS, material: Pb, thickness: 20.3 cm, angle: 0 FNS, material: Pb, thickness: 20.3 cm, angle: 0 0.01 2

Experimental uncertainty Calculation 0.001 Experiment Calculation 1.5

0.0001 source-neutron) ] 2 1 C/E [-] 1e-05

0.5 1e-06 Flux/lethargy [1/(cm

1e-07 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.76 Neutron spectrum for the FNS, Pb, d=20cm benchmark at 0Ž angle.

FNS, material: Pb, thickness: 20.3 cm, angle: 12.2 FNS, material: Pb, thickness: 20.3 cm, angle: 12.2 0.01 2

Experimental uncertainty Calculation 0.001 Experiment Calculation 1.5

0.0001 source-neutron) ] 2 1 C/E [-] 1e-05

0.5 1e-06 Flux/lethargy [1/(cm

1e-07 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.77 Neutron spectrum for the FNS, Pb, d=20cm benchmark at 12.2Ž angle.

FNS, material: Pb, thickness: 20.3 cm, angle: 24.9 FNS, material: Pb, thickness: 20.3 cm, angle: 24.9 0.01 2

Experimental uncertainty Calculation 0.001 Experiment Calculation 1.5

0.0001 source-neutron) ] 2 1 C/E [-] 1e-05

0.5 1e-06 Flux/lethargy [1/(cm

1e-07 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.78 Neutron spectrum for the FNS, Pb, d=20cm benchmark at 24.9Ž angle.

21616/05.69455/P 45-52 FNS, material: Pb, thickness: 20.3 cm, angle: 41.8 FNS, material: Pb, thickness: 20.3 cm, angle: 41.8 0.01 2

Experimental uncertainty Calculation 0.001 Experiment Calculation 1.5

0.0001 source-neutron) ] 2 1 C/E [-] 1e-05

0.5 1e-06 Flux/lethargy [1/(cm

1e-07 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.79 Neutron spectrum for the FNS, Pb, d=20cm benchmark at 42.8Ž angle.

FNS, material: Pb, thickness: 20.3 cm, angle: 66.8 FNS, material: Pb, thickness: 20.3 cm, angle: 66.8 0.01 2

Experimental uncertainty Calculation 0.001 Experiment Calculation 1.5

0.0001 source-neutron) ] 2 1 C/E [-] 1e-05

0.5 1e-06 Flux/lethargy [1/(cm

1e-07 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.80 Neutron spectrum for the FNS, Pb, d=20cm benchmark at 66.8Ž angle.

energy range 0Ž 12.2Ž 24.9Ž 42.8Ž 66.8Ž 0.1 1.0 0.88 š 0.01 0.95 š 0.01 0.93 š 0.01 0.86 š 0.01 0.82 š 0.01 1.0 5.0 0.89 š 0.01 0.94 š 0.01 0.91 š 0.01 0.85 š 0.01 0.82 š 0.01 5.0 10.0 0.82 š 0.03 0.93 š 0.03 0.89 š 0.04 0.79 š 0.04 0.72 š 0.06 10.0 20.0 0.95 š 0.01 0.88 š 0.01 0.83 š 0.02 0.73 š 0.03 0.63 š 0.05 Table 3.24 C/E values for the neutron spectrum of the FNS Pb 20cm benchmark.

46-52 21616/05.69455/P 3.2.15 FNS, Pb, 40 cm

FNS, material: Pb, thickness: 40.6 cm, angle: 0 FNS, material: Pb, thickness: 40.6 cm, angle: 0 0.001 2

Experimental uncertainty Calculation 0.0001 Experiment Calculation 1.5

1e-05 source-neutron) ] 2 1 C/E [-] 1e-06

0.5 1e-07 Flux/lethargy [1/(cm

1e-08 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.81 Neutron spectrum for the FNS, Pb, d=40cm benchmark at 0Ž angle.

FNS, material: Pb, thickness: 40.6 cm, angle: 12.2 FNS, material: Pb, thickness: 40.6 cm, angle: 12.2 0.001 2

Experimental uncertainty Calculation 0.0001 Experiment Calculation 1.5

1e-05 source-neutron) ] 2 1 C/E [-] 1e-06

0.5 1e-07 Flux/lethargy [1/(cm

1e-08 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.82 Neutron spectrum for the FNS, Pb, d=40cm benchmark at 12.2Ž angle.

FNS, material: Pb, thickness: 40.6 cm, angle: 24.9 FNS, material: Pb, thickness: 40.6 cm, angle: 24.9 0.001 2

Experimental uncertainty Calculation 0.0001 Experiment Calculation 1.5

1e-05 source-neutron) ] 2 1 C/E [-] 1e-06

0.5 1e-07 Flux/lethargy [1/(cm

1e-08 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.83 Neutron spectrum for the FNS, Pb, d=40cm benchmark at 24.9Ž angle.

21616/05.69455/P 47-52 FNS, material: Pb, thickness: 40.6 cm, angle: 41.8 FNS, material: Pb, thickness: 40.6 cm, angle: 41.8 0.001 2

Experimental uncertainty Calculation 0.0001 Experiment Calculation 1.5

1e-05 source-neutron) ] 2 1 C/E [-] 1e-06

0.5 1e-07 Flux/lethargy [1/(cm

1e-08 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.84 Neutron spectrum for the FNS, Pb, d=40cm benchmark at 42.8Ž angle.

FNS, material: Pb, thickness: 40.6 cm, angle: 66.8 FNS, material: Pb, thickness: 40.6 cm, angle: 66.8 0.001 2

Experimental uncertainty Calculation 0.0001 Experiment Calculation 1.5

1e-05 source-neutron) ] 2 1 C/E [-] 1e-06

0.5 1e-07 Flux/lethargy [1/(cm

1e-08 0 0.1 1 10 0.1 1 10 E [MeV] E [MeV]

Figure 3.85 Neutron spectrum for the FNS, Pb, d=40cm benchmark at 66.8Ž angle.

energy range 0Ž 12.2Ž 24.9Ž 42.8Ž 66.8Ž 0.1 1.0 0.86 š 0.01 0.92 š 0.01 0.82 š 0.01 0.81 š 0.01 0.85 š 0.01 1.0 5.0 0.84 š 0.01 0.94 š 0.01 0.87 š 0.01 0.86 š 0.01 0.80 š 0.01 5.0 10.0 0.69 š 0.09 0.97 š 0.11 0.76 š 0.15 0.71 š 0.14 0.12 š 0.21 10.0 20.0 0.89 š 0.02 0.80 š 0.04 0.73 š 0.06 0.77 š 0.11 0.12 š 0.13 Table 3.25 C/E values for the neutron spectrum of the FNS Pb 40cm benchmark.

48-52 21616/05.69455/P 3.3 LLNL Pulsed Spheres

3.3.1 LLNL Pulsed Spheres, Concrete, 21 cm

LLNL Pulsed Sphere; Concrete; 21 cm (2.0 mfp) LLNL Pulsed Sphere; Concrete; 21 cm (2.0 mfp) 0.1 2 experiment Experimental uncertainty mcnp Calculation

1.5 0.01

1 C/E [-] 0.001 Counts/ns/source

0.5

0.0001

0 15 20 25 30 35 40 45 50 15 20 25 30 35 40 45 50 Time [10-8 s] Time [10-8 s]

Figure 3.86 Neutron spectrum for the LLNL Pulsed Sphere, concrete, 21 cm (2.0 mfp) benchmark.

3.3.2 LLNL Pulsed Spheres, Concrete, 35.5 cm

LLNL Pulsed Sphere; Concrete; 35.5 cm (3.8 mfp) LLNL Pulsed Sphere; Concrete; 35.5 cm (3.8 mfp) 0.1 2 experiment Experimental uncertainty mcnp Calculation

1.5 0.01

1 C/E [-] 0.001 Counts/ns/source

0.5

0.0001

0 15 20 25 30 35 40 45 50 15 20 25 30 35 40 45 50 Time [10-8 s] Time [10-8 s]

Figure 3.87 Neutron spectrum for the LLNL Pulsed Sphere, concrete, 35.5 cm (3.8 mfp) benchmark.

21616/05.69455/P 49-52 3.4 NIST Water Spheres

NIST shielding benchmark (water, Cd) 1.1 Experimental uncertainty U-235 JEFF-3.1 Pu-239 U-238 Np-237 1.05

1 C/E for fission rates [-] 0.95

1.5" 2.0" 1.5" 2.0" 1.5" 2.0" 2.5" 1.5" 2.0" 2.5"

0.9 bare Cd water Cd+water

Figure 3.88 C/E values for the ﬁssion rate in ﬁssion foils in the NIST experiments.

material radius U-235 Pu-239 U-238 Np-237 bare 1.5” 0.992š0.004 0.971š0.004 0.960š0.005 0.978š0.004 bare 2.0” 0.992š0.004 0.968š0.004 0.954š0.005 0.978š0.004 Cd 1.5” 0.991š0.004 0.969š0.004 0.950š0.005 0.975š0.004 Cd 2.0” 0.988š0.004 0.970š0.004 0.947š0.005 0.975š0.004 ” H2O 1.5 1.044š0.001 1.032š0.002 0.956š0.003 0.988š0.004 ” H2O 2.0 1.023š0.003 1.009š0.003 0.951š0.004 0.970š0.004 ” H2O 2.5 1.013š0.003 1.002š0.003 0.967š0.006 0.967š0.006 ” Cd+H2O 1.5 1.017š0.004 0.987š0.005 0.945š0.002 0.956š0.001 ” Cd+H2O 2.0 1.008š0.005 0.977š0.007 0.945š0.004 0.961š0.003 ” Cd+H2O 2.5 0.996š0.006 0.954š0.008 0.942š0.005 0.967š0.004 Table 3.26 C/E values for the ﬁssion rates in the NIST benchmark.

50-52 21616/05.69455/P 4. Summary

The JEFF-3.1 general purpose library has been validated, using MCNP-4C3, with the shielding benchmarks Oktavian, FNS, LLNL Pusled Spheres, and NIST Water spheres. The ﬁrst three of these are 14 MeV experiments. This exercise provides insight in the global quality of a data library and also in particular problems of nuclear data for certain elements. Comparison with other data libraries and detailed inspection of the data ﬁles could reveal the origin of observed discrepancies. However, this is outside the scope of the present study.

21616/05.69455/P 51-52 References

[1] J.F. Briesmeister (Ed.), MCNP - A General Monte Carlo N-Particle Transport Code, Version 4C, Technical Report LA-13709-M, Los Alamos National Laboratory, USA (2000) J.S. Hendricks, MCNP4C3, Report X-5:RN(U)-JSH-01-17, Los Alamos National Laboratory, USA (2001) [2] M.C. Duijvestijn, A. Hogenbirk, S.C. van der Marck, ENDF/B-VI.8-PBMR - Contents of the ENDF/B-VI.8 based MCNP neutron transport cross section libraries for PBMR, NRG Report 21526/05.64901/C FAI/MD/AK (2005) [3] www-nds.iaea.or.at/fendl2/validation/benchmarks/jaerim94014/oktavian/n-leak [4] C. Ichihara, I. Kimura, S.A. Hayashi, J. Yamamoto, and A. Takahashi, Measurement of leakage neutron spectra from a spherical pile of zirconium irradiated with 14 MeV neutrons and validation of its nuclear data, J. Nucl. Sci. Techn. 40 (2003) 417–422 [5] www-nds.iaea.or.at/fendl2/validation/benchmarks/jaerim94014/fns-tof [6] Y. Oyama, S. Yamaguchi, and H. Maekawa Measruements and analyses of angular neutron ﬂux spectra on graphite and lithium-oxide slabs irradiated with 14.8 MeV neutrons, J. Nucl. Sci. Techn. 25 (1988) 419–428 [7] C. Wong, J.D. Anderson, P. Brown, L.F. Hansen, J.L. Kammerdiener, C. Logan, and B. Pohl, Livermore Pulsed Sphere Program: Program Summary through July 1971, UCRL-51144, Rev. 1 (1972) [8] J. D. Court, R.C. Brockhoff, and J.S. Hendricks Lawrence Livermore Pulsed Sphere Bench- mark Analysis of MCNP ENDF/B-VI, LA-12885 (1994) [9] M.Z. Youssef, A. Kumar, M.A. Abdou, Ch. Konno, F. Maekawa, M. Wada, Y. Oyama, H. Maekawa, and Y. Ikeda, Veriﬁcation of ITER shielding capability and FENDL data benchmarking through analysis of bulk shielding experiment on large SS316/water assembly bom- barded with 14 MeV neutrons, Fusion Engineering and Design 42 (1998) 235–245 [10] Sinbad data base, NEA (2004), http://www.nea.fr/html/science/shielding/sinbad/nist h2o/nist- abs.htm

52-52 21616/05.69455/P

Abstract Criticality safety benchmark calculations were performed with MCNP-4C3 using the new JEFF-3.1 nuclear data library. The benchmarks used were almost all taken from the International Criticality Safety Benchmark Evaluation Project (ICSBEP). Additional benchmarks were a Proteus core, a Dimple core, two Kritz cores, and two TRX cores. The results of these calculations are reported in tabular and graphical form.

Keywords MCNP-4C3 JEFF-3.1 nuclear data criticality safety benchmarks

2-54 21616/05.69456/P Contents

1 Introduction 5

2 Brief description of the criticality benchmarks 6 2.1 HEU benchmarks 6 2.1.1 Fast spectrum 6 2.1.2 Intermediate spectrum 8 2.1.3 Thermal spectrum 8 2.1.4 Mixed spectrum 12 2.2 IEU benchmarks 12 2.2.1 Fast spectrum 12 2.2.2 Intermediate spectrum 13 2.2.3 Thermal spectrum 14 2.3 LEU benchmarks 14 2.3.1 Fast spectrum 14 2.3.2 Intermediate spectrum 14 2.3.3 Thermal spectrum 14 2.4 PU benchmarks 20 2.4.1 Fast spectrum 20 2.4.2 Intermediate spectrum 20 2.4.3 Thermal spectrum 21 2.4.4 Mixed spectrum 22 2.5 MIX benchmarks 23 2.5.1 Fast spectrum 23 2.5.2 Intermediate spectrum 23 2.5.3 Thermal spectrum 23 2.6 U233 benchmarks 24 2.6.1 Fast spectrum 24 2.6.2 Intermediate spectrum 24 2.6.3 Thermal spectrum 24 2.7 Occurrence of elements 24

3 Results of validation calculations 26 3.1 HEU results 26 3.1.1 Fast spectrum 26 3.1.2 Intermediate spectrum 28 3.1.3 Thermal spectrum 28 3.1.4 Mixed spectrum 32 3.2 IEU results 32 3.2.1 Fast spectrum 32 3.2.2 Intermediate spectrum 33 3.2.3 Thermal spectrum 34 3.3 LEU results 34 3.3.1 Thermal spectrum 34

21616/05.69456/P 3-54 3.4 PU results 44 3.4.1 Fast spectrum 44 3.4.2 Intermediate spectrum 44 3.4.3 Thermal spectrum 45 3.4.4 Mixed spectrum 48 3.5 MIX results 48 3.5.1 Fast spectrum 48 3.5.2 Thermal spectrum 48 3.6 U233 results 50 3.6.1 Fast spectrum 50 3.6.2 Thermal spectrum 51 3.7 Summary 53

References 54

4-54 21616/05.69456/P 1. Introduction

In May 2005 a new version of the JEFF general purpose nuclear data library was released: JEFF- 3.1. Over the last several years there has been a concerted effort by many people in several countries to bring the quality of this nuclear data library to a new, higher level. Especially the underprediction of keff for many criticality safety benchmark with a thermal spectrum has received much attention. In this report, the results of many criticality safety benchmark runs are presented. The results show, among other things, that when using JEFF-3.1 nuclear data, the prediction of keff for the thermal spectrum benchmarks is much closer to the benchmark value. All results reported in this report were obtained by use of the Monte Carlo neutronics code MCNP, version 4C3 [1].

21616/05.69456/P 5-54 2. Brief description of the criticality benchmarks

Almost all benchmarks were taken from the International Handbook of Evaluated Criticality Safety Benchmark Experiments from the OECD-NEA project ICSBEP [2]. In Ref. [2], these benchmarks are subdivided in main categories according to three criteria. 1. The main fissionable isotope. The systems containing uranium-235 are subdivided according to the enrichment in 235U: there is low enriched uranium (LEU: wt% 235U < 10), intermediate enriched (IEU: 10 < wt% 235U < 60) and high enriched (HEU: wt% 235U > 60). There are also plutonium systems (PU), mixed uranium/plutonium systems (MIX), and 233U systems (U233). 2. The physical form of the fissile material: there are metal systems (MET), compound (COM), solution (SOL) and miscellaneous systems. 3. The neutron spectrum: if more than half of the fissions occurs for neutrons with energy below 0.625 eV the spectrum is thermal (THERM), if more than half occurs between 0.625 eV and 100 keV it is intermediate (INTER), and if more than half occurs over 100 keV it is fast (FAST). If none of these applies, the spectrum is classified as MIXED. The benchmark series identifier consists of the three components mentioned above, plus an additional sequence number. For each of the benchmark series for which calculations have been performed, a brief description is given below.

2.1 HEU benchmarks

2.1.1 Fast spectrum heu-met-fast-001 (1 case, ’Godiva’) ž A bare sphere of highly enriched uranium (LANL, 1950s). ž The uranium enrichment was 94% 235U. ž The isotopes in this benchmark model are U-f234,235,238g. heu-met-fast-005 (6 cases) ž Cores of uranium metal alloy and reflectors of molybdenum and beryllium (Obninsk, 1987). The amount of reflector material (Be-9) and the amount of molybdenum between fuel and reflector was varied. ž The uranium enrichment was 90% 235U. ž The isotopes in these benchmark models are C-nat, Al-27, Si-f2830g, Ti-f4650g, Cr-f50,5254g, Mn-55, Fe-f54,5658g, Ni-f58,6062,64g, Mo-f92,9498,100g, and U- f234,235,236,238g. In cases 2 6 Be-9 is also present. heu-met-fast-007 (43 cases) ž Uranium metal slabs moderated with polyethylene, plexiglas and teflon (ORNL, 1960s). Some of the polyethylene moderated experiments also had a polyethylene reflector. The H/235U ratio was varied between 0 and 5. ž The uranium enrichment was 93% 235U. ž The isotopes in these benchmark models are H-1, C-nat, U-f234,235,236,238g. In cases 2731 O-16 is also used, and in cases 3234 F-19 is also used.

6-54 21616/05.69456/P heu-met-fast-022 (1 case) ž A sphere of highly enriched uranium with C, Fe and W impurities, reflected by duraluminum (VNIIEF, Russia, 1962). ž The uranium enrichment was 90% 235U. ž The isotopes in this benchmark model are C-nat, Al-27, Fe-f54,5658g, Cu-f63,65g, W- f182184,186g, U-f234,235,238g. heu-met-fast-027 (1 case) ž A sphere of highly enriched uranium with C, Fe and W impurities, reflected by lead (VNIIEF, Russia, 1962). ž The uranium enrichment was 90% 235U. ž The isotopes in this benchmark model are C-nat, Fe-f54,5658g, W-f182184,186g, Pb- f206208g, U-f234,235,238g. heu-met-fast-028 (1 case, ’Topsy’, ’Flattop-25’) ž A sphere of highly enriched uranium reflected by normal uranium (LANL, 1960s). ž The uranium enrichment was 93% 235U. ž The isotopes in this benchmark model are U-f234,235,238g. heu-met-fast-057 (6 cases) ž Uranium metal spheres and cylinders, reflected by lead (LLNL, 1950s). ž The uranium enrichment was 93% 235U. ž The isotopes in these benchmark models are Be-9, Ca-f40,4244,46,48g, Fe-f54,5658g, Ni-f58,6062,64g, Pb-f204,206208g, U-f234,235,238g. heu-met-fast-060 (1 case) ž A cylindrical assembly of uranium metal and tungsten, with aluminum reflectors (ZPR-9 assembly 4, ANL, 1960s). ž The uranium enrichment was 93% 235U. ž The isotopes in this benchmark model are H-1, C-nat, F-19, Mg-f2426g, Al-27, Si-f2830g, Cl-f35,37g, Ti-f4650g, Cr-f50,5254g, Mn-55, Fe-f54,5658g, Ni-f58,6062,64g, Cu- f63,65g, Mo-f92,9498,100g, W-f182184,186g, U-f234,235,236,238g. heu-met-fast-064 (3 cases) ž Three cylinders of lead reflected highly enriched uranium (VNIITF, Russia, 1991). ž The uranium enrichment was 96% 235U. ž The isotopes in these benchmark models are C-nat, Si-f2830g, Cr-f50,5254g, Mn-55, Fe- f54,5658g, W-f182184,186g, Pb-f204,206208g, U-f234,235,238g. heu-met-fast-067 (2 cases) ž Cylindrical assemblies of uranium metal with tungsten, graphite (assembly 5) and aluminum assembly 6), reflected by dense aluminum. (ZPR-9 assemblies 5 and 6, ANL, 1960s). ž The uranium enrichment was 93% 235U. ž The isotopes in these benchmark models are H-1, C-nat, F-19, Mg-f2426g, Al-27, Si- f2830g, Cl-f35,37g, Ti-f4650g, Cr-f50,5254g, Mn-55, Fe-f54,5658g, Cu-f63,65g, Ni- f58,6062,64g, Mo-f92,9498,100g, W-f182184,186g, U-f234,235,236,238g. heu-met-fast-072 (1 case, ’Zeus’) ž An iron/HEU core surrounded by a copper reflector in the Los Alamos Critical Experiment Facility (LANL, 2002). In this report only configuration 1 (out of two) was computed.

21616/05.69456/P 7-54 ž The uranium enrichment was 93% 235U. ž The isotopes in these benchmark models are B-f10,11g, C-nat, N-f14,15g, Mg-f2426g, Al-27, Si-f2830g, P-31, S-f3234,36g, Ti-f4650g, V-nat, Cr-f50,5254g, Mn-55, Fe- f54,5658g, Ni-f58,6062,64g, Cu-f63,65g, Zn-nat, Ga-nat, Mo-f92,9498,100g, Pb- f204,206208g, U-f234,235,236,238g.

2.1.2 Intermediate spectrum heu-comp-inter-004 (1 case)

ž A k1 experiment at the HECTOR reactor (Winfrith, UK, 1960s): a graphite moderated uranium oxide core. The benchmark model is an inﬁnite medium with a material composition appropriate to the interpolated boron/235U ratio. ž The uranium enrichment was 92% 235U. ž The isotopes in these benchmark models are H-1, B-f10,11g, C-nat, O-16, Ca- f40,4244,46,48g, U-f234,235,236,238g. heu-comp-inter-005 (5 cases)

ž Several k1 experiments at the COBRA facility (Obninsk, Russia, 1980s, 1990s), containing uranium and nickel (KBR-7 assembly), stainless steel (KBR-9), stainless steel and molybdenum (KBR-10), chromium (KBR-15), and zirconium (KBR-16). ž The uranium enrichment was 90% 235U. ž The isotopes in these benchmark models are C-nat, O-16, Si-f2830g, Ti-f4650g, Cr- f50,5254g, Mn-55, Fe-f54,5658g, Ni-f58,6062,64g, U-f235,238g. Also used are Co-59 (case 1), Mo-f92,9498,100g (case 3) and H-1, Zr-f9092,94,96g, Hf-f174,176180g (case 5). heu-met-inter-001 (1 case) ž A cylindrical assembly if uranium and iron, reﬂected by stainless steel (ZPR-9 assemby 34, ANL, 1979). ž The uranium enrichment was 93% 235U. ž The isotopes in this benchmark model are H-1, C-nat, F-19, Mg-f2426g, Al-27, Si-f2830g, Cl-f35,37g, Cr-f50,5254g, Mn-55, Fe-f54,5658g, Ni-f58,6062,64g, Cu-f63,65g, Mo- f92,9498,100g, U-f234,235,236,238g. heu-met-inter-006 (4 cases, ’Zeus’) ž The inital set of Zeus experiments at the Los Alamos Critical Experiments Facility (LANL, 19992001): uranium metal interspersed with graphite plates in a cylindrical stack, surrounded by copper reﬂector. The C/235U ratio varied between 51.2 to 13.6. ž The uranium enrichment was 93% 235U. ž The isotopes in these benchmark models are C-nat, Mg-f2426g, Al-27, Si-f2830g, Ti- f4650g, Cr-f50,5254g, Mn-55, Fe-f54,5658g, Cu-f63,65g, U-f234,235,236,238g.

2.1.3 Thermal spectrum heu-met-therm-001 (2 cases) ž A polyethylene moderated and reﬂected uranium system with silicon (LANL, 1999). ž The uranium enrichment was 93% 235U. ž The isotopes in these benchmark models are H-1, C-nat, O-16, Si-f2830g, U- f234,235,236,238g.

8-54 21616/05.69456/P Thermal scattering data for H in CH2 is used. heu-met-therm-003 (7 cases) ž Lattices of oralloy cubes in water (LANL, 1955). The number of cubes and their spacing were varied. ž The uranium enrichment was 94% 235U. ž The isotopes in these benchmark models are H-1, C-nat, O-16, U-f234,235,238g.

Thermal scattering data for H in H2O is used. heu-met-therm-006 (23 cases) ž Lattices of SPERT-D fuel elements, reported by ORNL, 1964-65. The fuel elements consisted of plates of a uranium-aluminium alloy. The experiments were water-moderated and water- reﬂected. In some of the cases uranyl nitrate solution (with and without boron) was substituted for the water. The lattice conﬁguration was varied. ž The uranium enrichment was 93% 235U. ž The isotopes in these benchmark models are H-1, O-16, Mg-f2426g, Al-27, Si-f2830g, Ti- f4650g, Cr-f50,5254g, Mn-55, Fe-f54,5658g, Cu-f63,65g, and U-f234,235,238g. Ther-

mal scattering data for H in H2O is used. For cases 17 and 18 Cd-f106,108,110114,116g is also present, while for cases 4 and 19 23 C-nat, N-14, P-31, S-f3234,36g, and Ni-f58,6062,64g are present too. Besides, cases 20 23 have B-f10,11g. heu-met-therm-008 (2 cases) ž A polyethylene moderated and reﬂected uranium system with aluminium (LANL) ž The uranium enrichment was 93% 235U. ž The isotopes in these benchmark models are H-1, C-nat, Mg-f2426g, Al-27, Si-f2830g, Ti-f4650g, Cr-f50,5254g, Mn-55, Fe-f54,5658g, Cu-f63,65g, U-f234,235,236,238g.

Thermal scattering data for H in CH2 is used. heu-met-therm-009 (1 case) ž A polyethylene moderated and reﬂected uranium system with magnesium oxide (LANL, 2001) ž The uranium enrichment was 93% 235U. ž The isotopes in these benchmark models are H-1, Li-f6, B-10, C-nat, O-16, Mg- f2426g, Al-27, K-f39,40,41g, Ca-f40,4244,46,48g, Mn-55, Fe-f54,5658g, Cd- f106,108,110114,116g, U-f234,235,236,238g.

Thermal scattering data for H in CH2 is used. heu-met-therm-010 (2 cases) ž A polyethylene moderated and reﬂected uranium system with gadolinium (LANL, 2001). Two thicknesses of Gd foil were used. ž The uranium enrichment was 93% 235U. ž The isotopes in these benchmark models are H-1, C-nat, Gd-f152,154158,160g, U- f234,235,236,238g.

Thermal scattering data for H in CH2 is used. heu-met-therm-013 (2 cases) ž A polyethylene moderated and reﬂected uranium system with iron (LANL, ¾2002). Two thicknesses of iron plates were used.

21616/05.69456/P 9-54 ž The uranium enrichment was 93% 235U. ž The isotopes in these benchmark models are H-1, C-nat, Fe-f54,5658g, U- f234,235,236,238g.

Thermal scattering data for H in CH2 is used. heu-met-therm-014 (1 case) ž 2 ð 2 ð 23 array of uranium with silicon oxide, moderated and reﬂected by polyethylene (LANL, ¾2002) ž The uranium enrichment was 93% 235U. ž The isotopes in these benchmark models are H-1, C-nat, O-16, Si-f2830g, U- f234,235,236,238g.

Thermal scattering data for H in CH2 is used. heu-met-therm-016 (2 cases) ž 2ð2ð23 array of uranium with Ni-Cr-Mo-Gd alloy, moderated and reﬂected by polyethylene (LANL) ž The uranium enrichment was 93% 235U. ž The isotopes in these benchmark models are H-1, B-f10,11g, C-nat, O-16, Mg- f2426g, Al-27, Si-f2830g, P-31, Ca-f40,4244,46,48g, Ti-f4650g, V-nat, Cr- f50,5254g, Mn-55, Fe-f54,5658g, Co-59, Ni-f58,6062,64g, Cu-f63,65g, Br-79, Zr- f9092,94,96g, Nb-93, Mo-f92,9498,100g, Ag-f107,109g, Cd-f106,108,110114,116g, Gd-f152,154158,160g, Hf-f174,176,177,178,179,180g, Ta-181, W-f182184,186g, Pb- f204,206208g, U-f234,235,236,238g.

Thermal scattering data for H in CH2 is used. heu-met-therm-018 (2 cases) ž A polyethylene moderated and reﬂected uranium system with concrete (LANL, 2003). ž The uranium enrichment was 93% 235U. ž The isotopes in these benchmark models are H-1, Li-f6,7g, B-f10,11g, C-nat, O-16, Na-23, Mg-f2426g, Al-27, Si-f2830g, P-31, S-32, K-f39,40,41g, Ca-f40,4244,46,48g, Sc-45, Ti- f4650g, Fe-f54,5658g, Rh-103, Cd-f106,108,110114,116g, In-f113,115g, Eu-f151,153g, Gd-f152,154158,160g, U-f234,235,236,238g.

Thermal scattering data for H in CH2 is used. heu-sol-therm-001 (10 cases) ž Cylindrical reﬂected tank with uranyl nitrate, minimally reﬂected. The uranium concentration was varied and the critical height was determined. ž The uranium enrichment was 93% 235U. ž The isotopes in these benchmark models are H-1, C-nat, N-14, O-16, Al-27, Si-f2830g, P-31, S-f3234,36g, Cr-f50,5254g, Mn-55, Fe-f54,5658g, Ni-f58,6062,64g, Mo-

f92,9498,100g, U-f234,235,236,238g. Thermal scattering data for H in H2O is used. heu-sol-therm-002 (14 cases) ž Cylindrical reﬂected tank with uranyl nitrate, with a concrete reﬂector. The uranium concentration was varied and the critical height was determined. ž The uranium enrichment was 93% 235U. ž The isotopes in these benchmark models are H-1, C-nat, N-14, O-16, Na-23, Mg-f2426g, Al-27, Si-f2830g, S-f3234,36g, K-f3941g, Ca-f40,4244,46,48g, Ti-f4650g, Cr-

10-54 21616/05.69456/P f50,5254g, Mn-55, Fe-f54,5658g, and U-f234,235,236,238g. Thermal scattering data for

H in H2O is used. In cases 5 14 there is also Cu-f63,65g, in cases 1 4 P-31, Ni-f58,6062,64g, and Mo- f92,9498,100g. heu-sol-therm-004 (6 cases) ž Reﬂected uranyl-ﬂuoride solutions in heavy water (LANL, 1950s). The D/235U ratio varied from 34 to 431. ž The uranium enrichment was 94% 235U. ž The isotopes in these benchmark models are H-f1,2g, O-16, F-19, Si-f2830g, Cr- f50,5254g, Mn-55, Fe-f54,5658g, Ni-f58,6062,64g, U-f234,235,238g.

Thermal scattering data for D in D2O is used. heu-sol-therm-009 (4 cases)

ž Water reﬂected 6.4-liter spheres of highly enriched uranium oxyﬂuoride (UO2F2) solutions (ORNL, 1950s). The fuel concentration was varied. ž The uranium enrichment was 93% 235U. ž The isotopes in these benchmark models are H-1, O-16, F-19, Al-27, Si-f2830g, Mn-55,

Cu-f63,65g, U-f234,235,236,238g. Thermal scattering data for H in H2O is used. heu-sol-therm-010 (4 cases)

ž Water reﬂected 9.7-liter spheres of highly enriched uranium oxyﬂuoride (UO2F2) solutions (ORNL, 1950s). The fuel concentration and the temprature were varied. ž The uranium enrichment was 93% 235U. ž The isotopes in these benchmark models are H-1, O-16, F-19, Al-27, Si-f2830g, Mn-55,

Cu-f63,65g, U-f234,235,236,238g. Thermal scattering data for H in H2O is used. heu-sol-therm-013 (4 cases, ’ORNL-1’, : : :, ’ORNL-4’) ž Uranyl nitrate solutions poisoned with boric acid in an unreﬂected sphere (at ORNL, 1950s). ž The uranium enrichment was 93% 235U. ž The isotopes in these benchmark models are H-1, N-14, O-16, Al-27, Si-f2830g, Mn-55,

Cu-f63,65g, Uf234,235,236,238g. Thermal scattering data for H in H2O is used. For cases 1 3 B-f10,11g are also present. heu-sol-therm-032 (1 case) ž Unreﬂected 48-inch sphere of highly enriched uranyl nitrate solution (ORNL, 1950s). ž The uranium enrichment was 93% 235U. ž The isotopes in this benchmark model are H-1, N-14, O-16, Al-27, Si-f2830g, Mn-55, Fe-

f54,5658g, Cu-f63,65g, U-f233,234,235,236,238g. Thermal scattering data for H in H2O is used. heu-sol-therm-038 (30 cases) ž Two interacting slab tanks with highly enriched uranyl nitrate solution with various absorber- reflector plates (LANL, 1988). The absorber-reflector plates were made of poly-ethylene, borated poly-ethylene, boraflex, pyrex, stainless steel, hot rolled steel, lead, beryllium, depleted uranium, or cadmium. ž The uranium enrichment was 93% 235U. ž The isotopes in these benchmark models are H-1, Be-9, B-f10,11g, C-nat, N-14, O- 16, Na-23, Mg-f2426g, Al-27, Si-f2830g, P-31, S-f3234,36g, Ti-f4650g, Cr-

21616/05.69456/P 11-54 f50,5254g, Mn-55, Fe-f54,5658g, Ni-f58,6062,64g, Cu-f63,65g, Mo-f92,9498,100g, Cd-f106,108,110114,116g, Pb-f206208g, U-f234,235,236,238g. Thermal scattering data

for H in H2O and for H in CH2 is used. heu-sol-therm-039 (6 cases)

ž Mixture of highly enriched uranium hexafluoride (UF6) and hydrofluoric acid (HF), low H/U ratio, in a hot-water reflected tank (Valduc, 1960s). The uranium concentration was varied. ž The uranium enrichment was 93% 235U. ž The isotopes in these benchmark models are H-1, C-nat, O-16, F-19, Si-f2830g, Mn-55, Fe- f54,5658g, Ni-f58,6062,64g, Cu-f63,65g, U-f234,235,236,238g. Thermal scattering data

for H in H2O is used. heu-sol-therm-042 (8 cases) ž Unreﬂected cylinders (5 ft and 9 ft diameter) of highly enriched uranyl nitrate solution. The uranium concentration was varied. ž The uranium enrichment was 93% 235U. ž The isotopes in these benchmark models are H-1, C-nat, N-14, O-16, Cr-f50,5254g, Fe- f54,5658g, Ni-f58,6062,64g, Mo-f92,9498,100g, U-f234,235,236,238g. Thermal scat-

tering data for H in H2O is used.

2.1.4 Mixed spectrum heu-comp-mixed-003 (1 case) ž Heterogeneous assemblies of uranium dioxide fuel elements in a zirconium hydride moderator block, with beryllium reﬂector (RCC Kurchatov, 19921993). ž The uranium enrichment was 96% 235U. ž The isotopes in these benchmark models are H-1, Be-9, B-f10,11g, C-nat, O-16, Al-27, Si-f2830g, Ti-f4650g, Cr-f50,5254g, Mn-55, Fe-f54,5658g, Ni-f58,6062,64g, Zr-f9092,94,96g, Nb-93, Mo-f92,9498,100g, Hf-f174,176,177,178,179,180g, W-f182184,186g, U-f234,235,236,238g. Thermal scattering data for H in ZrH is used. heu-met-mixed-005 (5 cases) ž Critical experiments with heterogeneous compositions of highly enriched uranium, silicon dioxide and polyethylene (Obninsk, 1999). ž The uranium enrichment was 90% 235U. ž The isotopes in these benchmark models are H-1, Li-f6,7g, B-f10,11g, C-nat, N-14, O- 16, Na-23, Mg-f2426g, Al-27, Si-f2830g, S-32, K-f39,40,41g, Ca-f40,4244,46,48g, Ti-f4650g, V-nat, Cr-f50,5254g, Mn-55, Fe-f54,5658g, Ni-f58,6062,64g, Cu-f63,65g, Mo-f92,9498,100g, Cd-f106,108,110114,116g, Pb-f206208g, U-f234,235,236,238g.

Thermal scattering data for H in H2O is used.

2.2 IEU benchmarks

2.2.1 Fast spectrum ieu-met-fast-001 (4 cases, ’Jemima’) ž Bare cylindrical conﬁgurations of enriched and natural uranium (LANL, 19521954). ž The core average uranium enrichment was 3655% 235U.

12-54 21616/05.69456/P ž The isotopes in these benchmark models are Mg-f2426g, Al-27, Cr-f50,5254g, Mn-55, Fe-f54,5658g, Ni-f58,6062,64g, Cu-f63,65g, U-f234,235,238g. ieu-met-fast-002 (1 case) ž Cylindrical assembly with a core of alternating plates of enriched and natural uranium surrounded by a natural uranium reflector (LANL, 1956). ž The core average uranium enrichment was 16% 235U. ž The isotopes in this benchmark model are U-f234,235,238g. ieu-met-fast-007 (3 cases, ’Big Ten’) ž Three models of a large mixed-uranium-metal cylindrical core with 10% average uranium enrichment, surrounded by a thick depleted uranium reflector (LANL, 1971). The total uranium mass was ten metric tons. This core is modeled in three different benchmark models: detailed, simple and two-zone. ž The core average uranium enrichment was 10% 235U. ž The isotopes in these benchmark models are Si-f2830g, Cr-f50,5254g, Mn-55, Fe- f54,5658g, Ni-f58,6062,64g, Nb-93, U-f234,235,236,238g. ieu-met-fast-010 (1 case) ž A cylindrical assembly of uranium metal, 9% average enrichment, with a thick depleted uranium reflector (ANL, ZPR-9, U9 benchmark assembly, 1980). ž The core average uranium enrichment was 9% 235U. ž The isotopes in these benchmark models are H-1, C-nat, F-19, Al-27, Si-f2830g, Cl-f35,37g, Cr-f50,5254g, Mn-55, Fe-f54,5658g, Ni-f58,6062,64g, Cu-f63,65g, Mo- f92,9498,100g, U-f234,235,236,238g. ieu-met-fast-012 (1 case) ž A critical assembly of uranium metal, aluminium and steel, reflected by depleted uranium (ZPR-3 assembly 41, 1962). ž The core average uranium enrichment was 16% 235U. ž The isotopes in this benchmark model are C-nat, Al-27, Si-f2830g, Cr-f50,5254g, Mn-55, Fe-f54,5658g, Ni-f58,6062,64g, and U-f234,235,236,238g. ieu-met-fast-014 (2 cases) ž Cylindrical assemblies of uranium metal and tungsten with aluminum reflectors (ANL, ZPR-9 assemblies 2 and 3, 1964). Case 2 (assembly 3) had more tungsten than case 1 (assembly 1). ž The core average uranium enrichment was 16% 235U (case 1) and 21% 235U (case 2). ž The isotopes in these benchmark models are H-1, C-nat, F-19, Mg-f2426g, Al- 27, Si-f2830g, Cl-f35,37g, Ti-f4650g, Cr-f50,5254g, Mn-55, Fe-f54,5658g, Ni- f58,6062,64g, Cu-f63,65g, Mo-f92,9498,100g, W-f182184,186g, U-f234,235,236,238g.

2.2.2 Intermediate spectrum ieu-comp-inter-001 (4 cases)

ž For k1 experiments for combinations of uranium, thorium metal and varying amounts of polyethylene (IPPE, Obninsk, 19901994). The H/235U ratio varied from 0 to 70. ž Various combinations of 36% 235U and 90% 235U were used. ž The isotopes in these benchmark models are H-1, C-nat, O-16, Al-27, Si-f2830g, Ti- f4650g, Cr-f50,5254g, Mn-55, Fe-f54,5658g, Ni-f58,6062,64g, Th-232, U-f235,238g

21616/05.69456/P 13-54 2.2.3 Thermal spectrum ieu-comp-therm-002 (3 cases)

ž Stainless steel clad UO2 fuel rods in a water ﬁlled tank (MATR, Russia, 1970-1973). The fuel rods were arranged in a hexagonal lattice, with gadolinium absorber, cadmium absorber or no absorber. Cases 1, 3 and 5 were at room temperature, case 2 was at T=218.4ŽC, and cases 4 and 6 were at T=150.8ŽC. ž The uranium enrichment was 17% 235U. ž The isotopes in these benchmark models are H-1, C-nat, O-16, Si-f2830g, Ti-f4650g, Cr-f50,5254g, Mn-55, Fe-f54,5658g, Ni-f58,6062,64g, and U-f234,235,238g. Thermal

scattering data for H in H2O is used. For case 3 Al-27 and Gd-f152,154158,160g are also present, for case 5 Al-27 and Cd- f106,108,110114,116g. ieu-comp-therm-003 (2 cases) ž Zirconium hydride fuel rods in water, with a graphite reﬂector (Triga Mark II reactor, Ljubl- jana, 1991). The two cases differ only in the position of the 7 outermost fuel elements. ž The uranium enrichment was 20% 235U. ž The isotopes in these benchmark models are H-1, B-f10,11g, C-nat, N-14, O-16, Al-27, Si- f2830g, P-31, S-f3234,36g, Cr-f50,5254g, Mn-55, Fe-f54,5658g, Ni-f58,6062,64g, Zr-f9092,94,96g, Mo-f92,9498,100g, U-f235,238g. Thermal scattering data for graphite,

for H in ZrH and for H in H2O is used. Proteus (1 case) ž Core 5 of Proteus, a graphite reﬂected pebble bed reactor, containing uranium-carbon fuel pebbles and graphite moderator pebbles (PSI, 19951996) [3]. ž The uranium enrichment was 16.7% 235U. ž The isotopes in these benchmark models are H-1, Li-f6,7g, B-f10,11g, C-nat, N-14, O-16, Mg-f2426g, Al-27, Si-f2830g, S-f3234,36g, Cl-f35,37g, Ca-f40,4244,46,48g, Mn-55, Fe-f54,5658g, Cu-f63,65g, Ga-nat, Cd-f106,108,110114,116g, Gd-f154158,160g, Sm- f144,147150,152,154g, U-f234,235,236,238g. Thermal scattering data for graphite and for

H in H2O is used.

2.3 LEU benchmarks

2.3.1 Fast spectrum No benchmark models in this category are available from the ICSBEP handbook.

2.3.2 Intermediate spectrum No benchmark models in this category are available from the ICSBEP handbook.

2.3.3 Thermal spectrum leu-comp-therm-001 (8 cases)

ž Clusters of aluminium clad UO2 fuel rods in a water ﬁlled tank. The clusters were square, with no absorber plates, reﬂecting walls, dissolved poison or gadolinium impurities. The number of clusters and the separation between them was varied.

14-54 21616/05.69456/P ž The uranium enrichment was 2.4% 235U. ž The isotopes in these benchmark models are H-1, C-nat, O-16, Mg-f2426g, Al-27, Si-f2830g, Ti-f4650g, Cr-f50,5254g, Mn-55, Fe-f54,5658g, Cu-f63,65g, and U-

f234,235,236,238g. Thermal scattering data for H in H2O and for H in CH2 is used. leu-comp-therm-002 (5 cases)

ž Water moderated UO2 fuel rods, aluminium clad, in square arrays (pitch 2.54 cm, Paciﬁc Northwest Laboratories). The shape of the rod cluster was varied, as well as the distance between clusters (for cases 4 and 5). ž The uranium enrichment was 4.31% 235U. ž The isotopes in these benchmark models are H-1, C-nat, O-16, Mg-f2426g, Al-27, Si- f2830g, S-f3234,36g, Ca-f40,4244,46,48g, Ti-f4650g, Cr-f50,5254g, Mn-55, Fe-

f54,5658g, Cu-f63,65g, U-f234,235,236,238g. Thermal scattering data for H in H2O and for H in CH2 is used. leu-comp-therm-003 (22 cases)

ž Water moderated UO2 fuel rods, aluminium clad, in square arrays (pitch 1.684 cm), with gadolinium impurity in water (Paciﬁc Northwest Laboratories). The shape of the rod clusters was varied, as well as the distance between clusters. ž The uranium enrichment was 2.35% 235U. ž The isotopes in these benchmark models are H-1, C-nat, O-16, Mg-f2426g, Al- 27, Si-f2830g, Ti-f4650g, Cr-f50,5254g, Mn-55, Fe-f54,5658g, Cu-f63,65g, Gd-

f152,154158,160g, U-f234,235,236,238g. Thermal scattering data for H in H2O and for H in CH2 is used. leu-comp-therm-005 (16 cases)

ž UO2 fuel rods, aluminium clad, in water containing dissolved gadolinium (Paciﬁc Northwest Laboratories, early 1980s). The lattice pitch and the gadolinium concentration were varied. ž The uranium enrichment was 2.35% 235U (cases 113) and 4.31% 235U (cases 1416). ž The isotopes in these benchmark models are H-1, C-nat, N-14, O-16, Mg-f2426g, Al-27, Si-f2830g, S-f3234,36g Ca-f40,4244,46,48g, Ti-f4650g, Cr-f50,5254g, Mn-55, Fe- f54,5658g, Cu-f63,65g, Gd-f152,154158,160g, U-f234,235,236,238g. Thermal scattering

data for H in H2O and for H in CH2 is used. leu-comp-therm-006 (18 cases)

ž Arrays of UO2 fuel rods with water-to-fuel ratios from 1.5 to 3.0 in a Tank-type Critical Assembly (TCA, Japan, 19631975). The lattice pitch and the number of fuel rods were varied. ž The uranium enrichment was 2.6% 235U. ž The isotopes in these benchmark models are H-1, O-16, Al-27, U-f234,235,238g. Thermal

scattering data for H in H2O is used. leu-comp-therm-007 (10 cases)

ž Water moderated UO2 fuel rod arrays (Valduc, 1978). In four cases the arrays were square, with varying pitch. In the other cases the arrays had a triangular pitch, with either hexagonal pr pseudo-cylindrical section, and with varying pitch. ž The uranium enrichment was 4.738% 235U. ž The isotopes in these benchmark models are H-1, B-f10,11g, C-nat, N-14, O-16, Mg-

21616/05.69456/P 15-54 f2426g, Al-27, Si-f2830g, P-31, S-f3234,36g, Cr-f50,5254g, Mn-55, Fe-f54,5658g, Ni-f58,6062,64g, Zn-nat, U-f234,235,236,238g.

Thermal scattering data for H in H2O is used. leu-comp-therm-009 (27 cases)

ž Clusters of aluminium-clad UO2 fuel rods in a large water-ﬁlled tank (Paciﬁc Northwest Labo- ratories, 1977). There were three square-pitched clusters with two absorber plates in between. These plates were stainless steel, borated stainless steel, boral, copper, copper with 1% cadmium, cadmium, aluminium or zircaloy-4. ž The uranium enrichment was 4.3% 235U. ž The isotopes in these benchmark models are H-1, C-nat, O-16, Na-23, Mg-f2426g, Al-27, Si-f2830g, S-f3234,36g, Ca-f40,4244,46,48g, Ti-f4650g, Cr-f50,5254g, Mn-55, Fe-

f54,5658g, Cu-f63,65g, and U-f234,235,236,238g. Thermal scattering data for H in H2O and for H in CH2 is used. In some cases Na-23 (case 9 13) is present, in others B-f10,11g (case 5 9, 14, 15), Cd-f106,108,110114,116g (case 14 23), Ni-f58,6062,64g (cases 1 9, 14, 15), Mo- f92,9498,100g (cases 1 8), Sn-f112,114120,122,124g (cases 14, 15, 26, 27), and Zr- f9092,94,96g (cases 26, 27). leu-comp-therm-010 (30 cases)

ž Rectangular clusters of water moderated UO2 fuel rods, reﬂected by two lead, uranium or steel walls. Also the pitch and the number of fuel rods were varied. ž The uranium enrichment was 4.31% 235U. ž The isotopes in these benchmark models are H-1, C-nat, O-16, Mg-f2426g, Al-27, Si- f2830g, P-31, S-f3234,36g, Ca-f40,4244,46,48g, Ti-f4650g, Cr-f50,5254g, Mn-55, Fe-f54,5658g, Ni-f58,6062,64g, Cu-f63,65g, Mo-f92,9498,100g, Pb-f206208g, U-

f234,235,236,238g. Thermal scattering data for H in H2O and for H in CH2 is used. leu-comp-therm-016 (32 cases) ž Low enriched uranium pin assemblies placed in water and separated by absorber plates. The absorber material was steel, borated steel, boral, copper, copper with cadmium, cadmium, aluminium or zircaloy-4. ž The uranium enrichment was 93% 235U. ž The isotopes in these benchmark models are H-1, C-nat, O-16, Mg-f2426g, Al-27, Si-f2830g, Ti-f4650g, Cr-f50,5254g, Mn-55, Fe-f54,5658g, Cu-f63,65g, and U-

f234,235,236,238g. Thermal scattering data for H in H2O and for H in CH2 is used. In some cases B-f10,11g is also present (cases 8 14, 20), in others Na-23 (cases 12 19), S- f3234,36g (cases 1219, 2830), Ni-f58,6062,64g (cases 114, 20), Zr-f9092,94,96g (cases 31, 32), Mo-f92,9498,100g (cases 1 11), Cd-f106,108,110114,116g (cases 20 27), and Sn-f112,114120,122,124g (cases 20, 31, 32). leu-comp-therm-017 (19 cases) ž Low enriched uranium pin assemblies placed in water and reflected by steel or lead reflector plates. The separation between the clusters and the distance between the fuel and the reflector were varied. ž The uranium enrichment was 2.35% 235U. ž The isotopes in these benchmark models are H-1, C-nat, O-16, Mg-f2426g, Al-27 Si-f2830g, Ti-f4650g, Cr-f50,5254g, Mn-55, Fe-f54,5658g, Cu-f63,65g and U-

16-54 21616/05.69456/P f234,235,236,238g. Thermal scattering data for H in H2O and for H in CH2 is used. In cases 13, 2325 there is also Pb-f206208g, in the other cases there P-31, S-f3234,36g, Ni-f58,6062,64g, and Mo-f92,9498,100g. leu-comp-therm-019 (3 cases) ž Water-moderated hexagonally pitched lattices with low enriched cylindrical fuel rods with stainless steel cladding (RRC Kurchatov Institute, 1961). The pitch of the lattice was varied. ž The uranium enrichment was 5% 235U. ž The isotopes in these benchmark models are H-1, C-nat, O-16, Mg-f2426g, Al-27, Si-f2830g, Ti-f4650g, Cr-f50,5254g, Mn-55, Fe-f54,5658g, Ni-f58,6062,64g, Cu-

f63,65g, and U-f234,235,236,238g. Thermal scattering data for H in H2O is used. leu-comp-therm-039 (17 cases)

ž Incomplete arrays of water moderated and reﬂected UO2 fuel rods (Valduc, 1978). All arrays were square, with a varying number of positions not ﬁlled. ž The uranium enrichment was 4.738% 235U. ž The isotopes in these benchmark models are H-1, B-f10,11g, C-nat, N-14, O-16, Mg- f2426g, Al-27, Si-f2830g, P-31, S-f3234,36g, Cr-f50,5254g, Mn-55, Fe-f54,5658g, Ni-f58,6062,64g, Zn-nat, U-f234,235,236,238g.

Thermal scattering data for H in H2O is used. leu-comp-therm-051 (19 cases) ž Aluminium-clad uranium oxide in 9 assemblies of 14ð14 fuel rods. (Babcock-Wilcox Lynch- berg Research Center, 1978-79). The moderator and reﬂector was borated water, and the absorbers were stainless steel and boron. The loading pattern, water height and boron concentration were varied. ž The uranium enrichment was 2.5% 235U. ž The isotopes in these benchmark models are H-1, B-10, B-11, O-16, Mg-f2426g, Al- 27, Si-f2830g, Ti-f4650g, Cr-f50,5254g, Mn-55, Fe-f54,5658g, Cu-f63,65g, and U-

f234,235,236,238g. Thermal scattering data for H in H2O is used. For cases 2 9 there are also: C-nat, P-31, S-f3234,36g, Co-59, Ni-f58,6062,64g, and Mo-f92,9498,100g. leu-comp-therm-060 (28 cases)

ž Conﬁgurations of UO2 fuel assemblies in an RBMK-type graphite moderated reactor. Some conﬁgurations had empty channels, some had water in the fuel channels, and some had boron or thorium absorber rods. Also the enrichment was varied. ž The uranium enrichment was 1.8% 235U (cases 1 and 2), 2.4% 235U (cases 5 and 6) or 2.0% 235U (all other cases). ž The isotopes in these benchmark models are H-1, B-f10,11g, C-nat, N-14, O-16, F-19, Mg- f2426g, Al-27, Si-f2830g, P-31, S-f3234,36g, Ca-f40,4244,46,48g, Ti-f4650g, V-nat, Cr-f50,5254g, Mn-55, Fe-f54,5658g, Ni-f58,6062,64g, Cu-f63,65g, As-75, Zr-f9092,94,96g, Nb-93, Mo-f92,9498,100g, Ag- f107,109g, Cd-f106,108,110114,116g, Sn-f112,114120,122,124g, Hf-f174,176180g, W- f182184,186g, Au-197, Pb-f206208g, Bi-209, Th-f230,232g, U-f234,235,236,238g. Ther-

mal scattering data for graphite and for H in H2O is used. Dimple (1 case)

21616/05.69456/P 17-54 ž Assembly S01A, a cylindrical arrangement of uranium dioxide fuel pins on a square pitch of 1.32 cm, water moderated and reﬂected (Winfrith, UK) [4]. ž The uranium enrichment was 3% 235U. ž The isotopes in these benchmark models are H-f1,2g, C-nat, N-14 O-16 Mg-f2426g, Al- 27, Si-f2830g, P-31 S-f3234,36g, Cl-f35,37g, Ti-f4650g, V-nat, Cr-f50,5254g, Mn- 55 Fe-f54,5658g, Co-nat, Ni-f58,6062,64g, Cu-f63,65g, Sr-f84,8688g, Nb-93 Mo- f92,9498,100g, Sn-f112,114120,122,124g, U-f234,235,236,238g. Thermal scattering data

for graphite and for H in H2O and for H in CH2 is used. TRX (2 cases)

ž Light-water moderated UO2 pins with aluminium cladding in a hexagonal lattice [5]. The pitch was 1.806 cm (case 1) or 2.174 cm (case 2). ž The uranium enrichment was 1.3% 235U. ž The isotopes in these benchmark models are H-1, O-16, Al-27, Fe-f54,5658g, U-f235,238g.

Thermal scattering data for graphite and for H in H2O is used. leu-met-therm-001 (1 case) ž A natural uranium, heavy water moderated critical assembly (Yugoslavia, 1958). ž The uranium enrichment was 0.72% 235U. ž The isotopes in this benchmark model are H-f1,2g, B-f10,11g, C-nat, N-14, O-16, Mg- f2426g, Al-27, Si-f2830g, Ti-f4650g, Cr-f50,5254g, Mn-55, Fe-f54,5658g, Ni- f58,6062,64g, Cu-f63,65g, Cd-f106,108,110114,116g, U-f234,235,238g. Thermal scat-

tering data for D in D2O and for H in H2O is used. leu-sol-therm-001 (1 case, ’Sheba-II’)

ž An unreﬂected UO2F2+H2O cylindrical assembly (LANL). ž The uranium enrichment was 5% 235U. ž The isotopes in this benchmark model are H-1, N-14, O-16, F-19, Cr-f50,5254g, Mn-55, Fe-f54,5658g, Ni-f58,6062,64g, U-f234,235,236,238g

Thermal scattering data for H in H2O is used. leu-sol-therm-003 (9 cases) ž Full and truncated bare spheres of 10% enriched uranyl nitrate water solutions (Obninsk, 1965). The uranium concentration and the sphere diameters were varied. ž The uranium enrichment was 10% 235U. ž The isotopes in these benchmark models are H-1, N-14, O-16, Si-f2830g, Ti-f4650g, Cr- f50,5254g, Mn-55, Fe-f54,5658g, Ni-f58,6062,64g, U-f234,235,238g. Thermal scatter-

ing data for H in H2O is used. leu-sol-therm-004 (7 cases) ž Water reﬂected uranyl nitrate solution in a 60 cm cylindrical water tank (STACY, Japan, 1995). The uranium concentration was varied. ž The uranium enrichment was 10% 235U. ž The isotopes in these benchmark models are H-1, C-nat, N-14, O-16, Si-f2830g, P-31, S-f3234,36g, Cr-f50,5254g, Mn-55, Fe-f54,5658g, Ni-f58,6062,64g, U-

f234,235,236,238g. Thermal scattering data for H in H2O is used. leu-sol-therm-007 (5 cases) ž Unreﬂected uranyl nitrate solution in a 60 cm cylindrical water tank (STACY, Japan, 1995).

18-54 21616/05.69456/P The uranium concentration was varied. ž The uranium enrichment was 10% 235U. ž The isotopes in these benchmark models are H-1, C-nat, N-14, O-16, Si-f2830g, P-31, S-f3234,36g, Cr-f50,5254g, Mn-55, Fe-f54,5658g, Ni-f58,6062,64g, U-

f234,235,236,238g. Thermal scattering data for H in H2O is used. leu-sol-therm-016 (7 cases) ž Water reﬂected slabs (28 cm) of uranyl nitrate solutions (STACY, Japan, 1997). The uranium concentration was varied. ž The uranium enrichment was 10% 235U. ž The isotopes in these benchmark models are H-1, C-nat, N-14, O-16, Si-f2830g, P-31, S-f3234,36g, Cr-f50,5254g, Mn-55, Fe-f54,5658g, Ni-f58,6062,64g, U-

f234,235,236,238g. Thermal scattering data for H in H2O is used. leu-sol-therm-017 (6 cases) ž Unreﬂected slabs (28 cm) of uranyl nitrate solutions (STACY, Japan, 1997). The uranium concentration was varied. ž The uranium enrichment was 10% 235U. ž The isotopes in these benchmark models are H-1, C-nat, N-14, O-16, Si-f2830g, P-31, S-f3234,36g, Cr-f50,5254g, Mn-55, Fe-f54,5658g, Ni-f58,6062,64g, U-

f234,235,236,238g. Thermal scattering data for H in H2O is used. leu-sol-therm-018 (6 cases) ž Concrete reﬂected slabs (28 cm) of uranyl nitrate solutions (STACY, Japan, 1997). The thickness of the reﬂector was varied. ž The uranium enrichment was 10% 235U. ž The isotopes in these benchmark models are H-1, C-nat, N-14, O-16, Na-23, Mg-f2426g, Al-27, Si-f2830g, P-31, S-f3234,36g, Cl-f35,37g, K-f3941g, Ca- f40,4244,46,48g, Cr-f50,5254g, Mn-55, Fe-f54,5658g,Ni-f58,6062,64g, Cu-f63,65g,

and U-f234,235,236,238g. Thermal scattering data for H in H2O is used. leu-sol-therm-020 (4 cases) ž Water reﬂected uranyl nitrate solution in a 80 cm cylindrical water tank (STACY, Japan, 19981999). The uranium concentration was varied. ž The uranium enrichment was 10% 235U. ž The isotopes in these benchmark models are H-1, C-nat, N-14, O-16, Si-f2830g, P-31, S-f3234,36g, Cr-f50,5254g, Mn-55, Fe-f54,5658g, Ni-f58,6062,64g, U-

f234,235,236,238g. Thermal scattering data for H in H2O is used. leu-sol-therm-021 (4 cases) ž Unreﬂected uranyl nitrate solution in a 80 cm cylindrical water tank (STACY, Japan, 19981999). The uranium concentration was varied. ž The uranium enrichment was 10% 235U. ž The isotopes in these benchmark models are H-1, C-nat, N-14, O-16, Si-f2830g, P-31, S-f3234,36g, Cr-f50,5254g, Mn-55, Fe-f54,5658g, Ni-f58,6062,64g, U-

f234,235,236,238g. Thermal scattering data for H in H2O is used.

21616/05.69456/P 19-54 2.4 PU benchmarks

2.4.1 Fast spectrum pu-met-fast-001 (1 case, ’Jezebel’) ž A bare sphere of plutonium (LANL, 1950s). ž The plutonium-239 enrichment was 95.2% 239Pu. ž The isotopes in this benchmark model are Ga-nat, Pu-f239,240,241g. pu-met-fast-002 (1 case, ’Jezebel-240’) ž A bare sphere of plutonium (LANL, 1964). ž The plutonium enrichment was 76.4% 239Pu and 20.1% 240Pu. ž The isotopes in this benchmark model are Ga-nat, Pu-f239,240,241,242g. pu-met-fast-005 (1 case) ž A critical experiment of a plutonium sphere reflected by tungsten (LANL, 1958). ž The isotopes in these benchmark models are Ni-f58,6062,64g, Cu-f63,65g, Ga-nat, Zr- f9092,94,96g, W-f182184,186g, U-f235,238g, Pu-f239,240,241g. pu-met-fast-006 (1 case, ’Popsy’, ’Flattop-Pu’) ž A sphere of plutonium reflected by normal uranium (LANL, 1960s). ž The plutonium enrichment was 94.9% 239Pu. ž The isotopes in this benchmark model are Ga-nat, U-f234,235,238g, Pu-f239,240,241g. pu-met-fast-008 (1 case, ’Thor’) ž A sphere of plutonium reflected by thorium (LANL, 1961). ž The plutonium enrichment was 94.9% 239Pu. ž The isotopes in this benchmark model are Ga-nat, Th-232, Pu-f239,240,241g. pu-met-fast-012 (1 case) ž A cylidrical arrangement of short, close-packed stainless steel clad rods of plutonium metal (97.6 at.%239Pu), reflected on all sides by thick depleted uranium (IPPE, Obninsk, 1956). ž The isotopes in these benchmark models are C-nat, Si-f2830g, Ti-f4650g, Cr- f50,5254g, Mn-55, Fe-f54,5658g, Ni-f58,6062,64g, Cu-f63,65g, Ga-nat, U-f235,238g, Pu-f238,239,240,241g. pu-met-fast-013 (1 case) ž ž A cylidrical arrangement of short, close-packed stainless steel clad rods of plutonium metal (97.6 at.%239Pu), reflected on all sides by thick copper reflector (IPPE, Obninsk, 1960). ž The isotopes in these benchmark models are C-nat, Si-f2830g, Ti-f4650g, Cr- f50,5254g, Mn-55, Fe-f54,5658g, Ni-f58,6062,64g, Cu-f63,65g, Ga-nat, U-f235,238g, Pu-f238,239,240,241g.

2.4.2 Intermediate spectrum pu-comp-inter-001 (1 case)

ž A k1 experiment at the HECTOR reactor (Winfrith, UK, 1960s): a graphite moderated plutonium oxide core (5% 240Pu). The benchmark model is an inﬁnite medium with a material composition appropriate to the interpolated boron/239Pu ratio. ž The isotopes in these benchmark models are H-1, B-f10,11g, C-nat, O-16, Ca-

20-54 21616/05.69456/P f40,4244,46,48g, U-f235,238g, Pu-f239,240,241,242g. Thermal scattering data for graphite is used. pu-met-inter-002 (1 case) ž A cylindrical assembly containing plutonium, carbon and stainless steel, reﬂected by stainless ateel and iron (ANL, ZPR-6 assembly 10, 19811982). ž The plutonium enrichment was 95.3% 239Pu. ž The isotopes in these benchmark models are C-nat, Al-27, Si-f2830g, Cr-f50,5254g, Co-nat, Mn-55, Fe-f54,5658g, Ni-f58,6062,64g, Cu-f63,65g, Mo-f92,9498,100g, Pu- f238242g, Am-241.

2.4.3 Thermal spectrum pu-sol-therm-001 (6 cases) ž Water reﬂected 11.5 inch diameter spheres of plutonium nitrate solution (Paciﬁc Northwest Laboratories, 1960s). The plutonium concentration was varied. ž The plutonium enrichment was 95% 239Pu. ž The isotopes in these benchmark models are H-1, N-14, O-16, Cr-f50,5254g, Mn-55, Fe- f54,5658g, Ni-f58,6062,64g, Pu-f238,239,240,241,242g. Thermal scattering data for H in

H2O is used. pu-sol-therm-002 (7 cases) ž Water reﬂected 12 inch diameter spheres of plutonium nitrate solution (1950s). The plutonium concentration was varied. ž The plutonium enrichment was 96.9% 239Pu. ž The isotopes in these benchmark models are H-1, N-14, O-16, Cr-f50,5254g, Fe-

f54,5658g, Ni-f58,6062,64g, Pu-f239,240g. Thermal scattering data for H in H2O is used. pu-sol-therm-003 (8 cases) ž Water reﬂected 13 inch diameter spheres of plutonium nitrate solution (1950s). The plutonium concentration was varied. ž The plutonium enrichment was 98.3% 239Pu (cases 1 and 2) and 96.9% 239Pu (cases 37). ž The isotopes in these benchmark models are H-1, N-14, O-16, Al-27, Si-f2830g, Cr- f50,5254g, Mn-55, Fe-f54,5658g, Ni-f58,6062,64g, Cu-f63,65g, Pu-f239,240g. Thermal

scattering data for H in H2O is used. pu-sol-therm-004 (13 cases) ž Water reﬂected 14 inch diameter spheres of plutonium nitrate solution (1950s). The plutonium concentration was varied. ž The plutonium enrichment was 99.5% 239Pu (cases 14), 98.25% 239Pu (case 5), and 96.9% 239Pu (cases 613). ž The isotopes in these benchmark models are H-1, N-14, O-16, Cr-f50,5254g, Fe-

f54,5658g, Ni-f58,6062,64g, Pu-f239,240g. Thermal scattering data for H in H2O is used. pu-sol-therm-005 (9 cases) ž Water reﬂected 14 inch diameter spheres of plutonium nitrate solution (1950s). The plutonium concentration was varied. ž The plutonium enrichment was 96.0% 239Pu (cases 17), and 95.6% 239Pu (cases 8 and 9). ž The isotopes in these benchmark models are H-1, N-14, O-16, Cr-f50,5254g, Fe-

21616/05.69456/P 21-54 f54,5658g, Ni-f58,6062,64g, Pu-f239,240g. Thermal scattering data for H in H2O is used. pu-sol-therm-006 (3 cases) ž Water reﬂected 15 inch diameter spheres of plutonium nitrate solution (1950s). The plutonium concentration was varied. ž The plutonium enrichment was 96.9% 239Pu. ž The isotopes in these benchmark models are H-1, N-14, O-16, Cr-f50,5254g, Fe-

f54,5658g, Ni-f58,6062,64g, Pu-f239,240g. Thermal scattering data for H in H2O is used. pu-sol-therm-007 (8 cases) ž Water reflected 11.5 inch diameter spheres partly filled with plutonium nitrate solution (Pacific Northwest Laboratories, 1960s). The plutonium concentration was varied. ž The plutonium enrichment was 95% 239Pu. ž The isotopes in these benchmark models are H-1, N-14, O-16, Cr-f50,5254g, Mn-55, Fe- f54,5658g, Ni-f58,6062,64g, Pu-f238,239,240,241,242g. Thermal scattering data for H in

H2O is used. pu-sol-therm-008 (29 cases) ž Concrete reflected 14 inch diameter spheres of plutonium nitrate solution (Pacific Northwest Laboratories, 19612). The geometry and the thickness of the concrete reflector was varied. Some cases had an extra shell of stainless steel outside the solution tank, inside the reflector. some others an extra shell of cadmium at that place. ž The plutonium enrichment was 95% 239Pu. ž The isotopes in these benchmark models are H-1, N-14, O-16, Na-23, Mg-f2426g, Al-27, Si-f2830g, K-f3941g, Ca-f40,4244,46,48g, Cr-f50,5254g, Mn-55, Fe-f54,5658g, Ni- f58,6062,64g, Cd-f106,108,110114,116g, Pu-f238,239,240,241,242g. Thermal scattering

data for H in H2O is used. pu-sol-therm-012 (22 cases) ž Plutonium nitrate solution in a large tank, with and without water reflector (Valduc, 1974). The tank was a right parallelepiped of dimension 130 ð 130 ð 100 cm3. The water reflector was either on six sides (cases 25), on five sides (cases 613), or on no sides (cases 1423). The plutonium concentration was varied. ž The plutonium enrichment was 74% 239Pu, and 19% 240Pu. ž The isotopes in these benchmark models are H-1, B-10, C-nat, N-14, O-16, Al- 27, Si-f2830g, Cl-f35,37g, Ca-f40,4244,46,48g, Cr-f50,5254g, Fe-f54,5658g, Ni-

f58,6062,64g, Pu-f239,240,241,242g, Am-241. Thermal scattering data for H in H2O and for H in CH2 is used.

2.4.4 Mixed spectrum pu-met-mixed-001 (6 cases) ž Heterogeneous conﬁgurations of plutonium, silicon dioxide, and polyethylene (IPPE, Ob- ninsk, 19992000). The cores were conﬁgured to cover a broad range in neutron spectra. ž The isotopes in these benchmark models are H-1, Li-f6,7g, B-f10,11g, C-nat, N-14, O- 16, Na-23, Mg-f2426g, Al-27, Si-f2830g, S-32, K-f39,40,41g, Ca-f40,4244,46,48g, Ti-f4650g, V-nat, Cr-f50,5254g, Mn-55, Fe-f54,5658g, Ni-f58,6062,64g, Cu-f63,65g, Ga-nat, Mo-f92,9498,100g, Cd-f106,108,110114,116g, Pb-f206208g, U-f235,238g, Pu-

22-54 21616/05.69456/P f239,240,241g, Am-241.

Thermal scattering data for H in H2O is used.

2.5 MIX benchmarks

2.5.1 Fast spectrum mix-met-fast-011 (4 cases) ž Cylindrical assemblies of mixed ﬁssile plutonium and uranium metal, reﬂected by graphite (ANL, ZPPR-21 phases BE, 1990). The ratio of plutonium to uranium was varied. ž The isotopes in these benchmark models are H-1, Li-f6,7g, B-f10,11g, C-nat, N-14, O-16, Na-23, Al-27, Si-f2830g, Cr-f50,5254g, Mn-55, Fe-f54,5658g, Co-nat, Ni- f58,6062,64g, Cu-f63,65g, Zr-f9092,94,96g, Mo-f92,9498,100g, U-f234,235,236,238g, Pu-f238,239,240,241,242g, Am-241.

2.5.2 Intermediate spectrum No calculations for benchmarks in this category were performed.

2.5.3 Thermal spectrum mix-comp-therm-012 (33 cases) ž Rectangular parallelepipeds of homogeneous plutonium uranium mixed oxide polystyrene (Pacific Northwest Laboratories, 1970-1972). The cores were either unreflected (cases 2022 and 3133) or plexiglas reflected (cases 119 and 2330). The amount of plutonium in the MOX, and the plutonium vector were varied. ž The isotopes in these benchmark models are H-1, C-nat, O-16, U-f235,238g, Pu-

f238,239,240,241,242g, Am-241. Thermal scattering data for H in CH2 is used. Kritz (2 cases) ž Core 2:19, consisting of light water moderated and reﬂected square lattices with mixed oxide fuel rods (Studsvik, Sweden, 1970s) [6]. Criticality was obtained at room temperature and at 235.9ŽC, by adjusting the boron content of the water and by adjusting the water height. The only differences between the cold and the hot case are the densities, the water level, and some slight dimensional changes of the core components. ž The plutonium enrichment was 91% 239Pu. ž The isotopes in these benchmark models are H-1, B-f10,11g, N-14, O-16, Cr-f50,5254g, Fe-f54,5658g, Ni-f58,6062,64g, Zr-f9092,94,96g, Sn-f112,114120,122,124g, U-

f235,238g, Pu-f239,240,241,242g, Am-241. Thermal scattering data for H in H2O is used. mix-sol-therm-001 (3 cases) ž Critical experiments with mixed plutonium and uranium nitrate solutions in a large cylindrical geometry (PNL, 1980s). The concentration of the solution and the U/Pu ratio was varied. ž The isotopes in these benchmark models are H-1, Li-6, B-10, N-14, O-16, Cr- f50,5254g, Mn-55, Fe-f54,5658g, Ni-f58,6062,64g, Cd-f106,108,110114,116g, Gd- f152,154158,160g, U-f234,235,236,238g, Pu-f238,239,240,241,242g, Am-241.

Thermal scattering data for H in H2O is used.

21616/05.69456/P 23-54 2.6 U233 benchmarks

2.6.1 Fast spectrum u233-met-fast-001 (1 case, ’Skidoo’, ’Jezebel-233’) ž A bare sphere of highly enriched uranium-233 metal (LANL, 1961). ž The uranium enrichment was 98% 233U. ž The isotopes in these benchmark models are U-f233,234,235,238g. u233-met-fast-005 (2 cases) ž Highly enriched uranium-233 spheres, reﬂected by beryllium (LANL, 1958). The mass of the uranium-233 core was 10 kg (case 1) and 7.6 kg (case 2). ž The uranium enrichment was 98% 233U. ž The isotopes in these benchmark models are Be-9, O-16, U-f233,234,238g. u233-met-fast-006 (1 case, ’Flattop-23’) ž A highly enriched uranium-233 sphere, reﬂected by normal uranium (LANL, 1964). ž The uranium enrichment was 98% 233U. ž The isotopes in these benchmark models are U-f233,234,235,238g.

2.6.2 Intermediate spectrum No calculations for benchmarks in this category were performed.

2.6.3 Thermal spectrum u233-comp-therm-001 (8 cases) 235 233 233 ž Cores of UO2ZrO2 and cores of UO2ZrO2, with blankets of either UO2 or ThO2 (BAPL, 1960s). The moderator was light water. Five assemblies were rectangular (cases 15) and three hexagonal (cases 68). ž The uranium enrichment was 97% 233U for cases 2, 3, 4, 7, and 8, and 93% 235U for cases 1, 5, and 6. ž The isotopes in these benchmark models are H-1, B-10, C-nat, O-16, Cr-f50,5254g, Mn- 55, Fe-f54,5658g, Ni-f58,6062,64g, Zr-f9092,94,96g, Sn-f112,114120,122,124g, Gd-

f152,154158,160g, Th-232, U-f233,234,235,238g. Thermal scattering data for H in H2O and for H in CH2 is used. u233-sol-001 (5 cases) ž Unreﬂected spheres of uranium-233 nitrate solutions (ORNL, 1950s). The amount of boron poison and the uranium concentration was varied. ž The uranium enrichment was 98% 233U. ž The isotopes in these benchmark models are H-1, B-f10,11g, N-14, O-16, Al-27, Si-f2830g, Mn-55, Fe-f54,5658g, Cu-f63,65g, Th-232, U-f233,234,235,238g. Thermal scattering data

for H in H2O is used.

2.7 Occurrence of elements

In Table 2.1 are listed those elements that are present in a material of a benchmark series, either with more than 1 wt% or with more than 1e-4 atoms per barn-cm. Elements that do not show up in the table are either not at all present in the benchmark models, or only in minor fractions.

24-54 21616/05.69456/P H in H2O (59 benchmark series) D in D2O lmt01, hst04 Be hcm03 Be in BeO hcm03 C in graphite lct60, proteus, ict03, hci04, hmm05, pci01, pmm01 H in ZrH ict03, hcm03

H in CH2 (26 benchmark series) H (75 benchmark series) Li mmf11 Be hcm03, hst38, hmf05, hmf057, u3mf05 B (13 benchmark series) C (58 benchmark series) N (33 benchmark series) O (72 benchmark series) F hst04, hst09, hst10, hst39, hmf07, lst01 Na lst18, hmm05, hmt18, hst02, hst38, pmm01, pst08, mcf01 Mg (30 benchmark series) Al (53 benchmark series) Si (59 benchmark series) P hst02 S hst02, lct02, lct05, lct09, lct10, lst18 Cl pst12 K lst18, hmt18, hmm05, hst02, pmm01, pst08 Ca (15 benchmark series) Ti (14 benchmark series) Cr (59 benchmark series) Mn (54 benchmark series) Fe (77 benchmark series) Ni (62 benchmark series) Co lct51 Cu (18 benchmark series) Ga pmf01, pmf02, pmf05, pmf06, pmf08, pmf12, pmf13, pmm01 Zr lct09, lct16, lct60, ict03, hci05, hcm03, pmf05, kritz, mmf11, u3ct01 Nb lct60, imf07, hcm03 Mo ict03, hmt16, hst42, hci05, hcm03, hmf05 Cd lct09, lct16, ict02, hst38, hmt06, hmm05, pst08, pmm01 Sn lct09, kritz, u3ct01 Gd ict02, hmt10, hmt16 W imf14, hmf60, hmf67, hcm03, pmf05 Pb lct10, lct17, hst38, hmf27, hmf57, hmf64 Th lct60, ici01, pmf08, u3ct01 U (87 benchmark series) Pu (24 benchmark series) Table 2.1 A list of elements and the benchmark series in which these elements are present in a material, either with more than 1 wt% or with more than 1e-4 atoms per barn-cm.

21616/05.69456/P 25-54 3. Results of validation calculations

In this Section we report all the keff results of the calculations. In the following subsections, the results are given in graphical and tabular form. The columns contain the following items.

1. The benchmark value for keff, and its uncertainty in pcm between brackets. These values were obtained from Refs [2, 3, 4, 5, 6]. 2. Results from calculations based on JEFF-3.0 3. The results of the present work, based on JEFF-3.1. 4. The values for the third column divided by the ﬁrst column (including uncertainty). 5. The benchmark name

3.1 HEU results

3.1.1 Fast spectrum

1.04

JEFF-3.0 JEFF-3.1 benchmark uncertainty 1.03

1.02

eff 1.01 C/E for k

0.99

0.98 hmf05 hmf07 hmf57 Zeus hci04 hci05 hmi01 Topsy hmf22 hmf27 hmf64 hmf72 hcm03 Godiva hmm05 ZPR9-4 ZPR9-5,6

Figure 3.1 Results for the HEU benchmarks with a fast or intermediate spectrum

benchmark JEFF-3.0 JEFF-3.1 C/E (JEFF-3.1) name 1.00000(100) 0.99596( 40) 0.99644( 19) 0.99644(102) heu-met-fast-001 bare-sphere 1.00000(360) 0.99153( 69) 0.99242( 72) 0.99242(367) heu-met-fast-005 case-1 1.00070(360) 0.99515( 60) 0.99586( 70) 0.99516(367) heu-met-fast-005 case-2 0.99960(360) 0.99767( 80) 0.99721( 79) 0.99761(369) heu-met-fast-005 case-3 0.99890(360) 0.99103( 69) 0.99087( 87) 0.99196(371) heu-met-fast-005 case-4 0.99800(360) 0.99470( 80) 0.99540( 78) 0.99739(369) heu-met-fast-005 case-5 continued on next page

26-54 21616/05.69456/P benchmark JEFF-3.0 JEFF-3.1 C/E (JEFF-3.1) name 0.99870(360) 0.99384( 70) 0.99389( 71) 0.99518(367) heu-met-fast-005 case-6 0.99710( 10) 0.98927( 66) 0.99045( 64) 0.99333( 65) heu-met-fast-007 case-1 0.99860( 10) 0.99408( 63) 0.99500( 71) 0.99639( 72) heu-met-fast-007 case-2 1.00120( 10) 0.99680( 74) 0.99643( 74) 0.99524( 75) heu-met-fast-007 case-3 0.99700( 10) 0.99551( 74) 0.99460( 66) 0.99759( 67) heu-met-fast-007 case-4 1.00000( 10) 0.99517( 70) 0.99768( 70) 0.99768( 71) heu-met-fast-007 case-5 1.00280( 10) 1.00346( 74) 1.00215( 70) 0.99935( 71) heu-met-fast-007 case-6 0.99960( 10) 0.99781( 84) 0.99944( 75) 0.99984( 76) heu-met-fast-007 case-7 0.99920( 10) 0.99607( 79) 0.99750( 73) 0.99830( 74) heu-met-fast-007 case-8 1.00170( 80) 0.99880( 74) 0.99888( 76) 0.99718(110) heu-met-fast-007 case-9 1.00000( 10) 0.99587( 90) 0.99497( 79) 0.99497( 80) heu-met-fast-007 case-10 0.99820( 10) 0.99586( 89) 0.99487( 87) 0.99666( 88) heu-met-fast-007 case-11 0.99510( 10) 0.99045( 76) 0.98820( 90) 0.99307( 92) heu-met-fast-007 case-12 1.00090( 10) 0.99756( 91) 0.99912( 89) 0.99822( 90) heu-met-fast-007 case-13 0.99830( 10) 0.99532( 87) 0.99444( 92) 0.99613( 93) heu-met-fast-007 case-14 0.99780( 10) 0.99462( 89) 0.99604( 95) 0.99824( 96) heu-met-fast-007 case-15 0.99880( 10) 0.99389( 85) 0.99530( 94) 0.99650( 95) heu-met-fast-007 case-16 0.99720( 10) 0.99349( 90) 0.99429( 90) 0.99708( 91) heu-met-fast-007 case-17 0.99910( 10) 0.99507( 96) 0.99825( 98) 0.99915( 99) heu-met-fast-007 case-18 0.99830( 10) 0.99301( 64) 0.99405( 68) 0.99574( 69) heu-met-fast-007 case-19 0.99810( 10) 0.99426( 79) 0.99339( 67) 0.99528( 68) heu-met-fast-007 case-20 0.99870( 10) 0.99641( 70) 0.99535( 79) 0.99665( 80) heu-met-fast-007 case-21 0.99940( 10) 0.99502( 74) 0.99537( 73) 0.99597( 74) heu-met-fast-007 case-22 0.99930( 10) 0.99564( 70) 0.99711( 85) 0.99781( 86) heu-met-fast-007 case-23 1.00010( 10) 0.99567( 74) 0.99537( 72) 0.99527( 73) heu-met-fast-007 case-24 0.99900( 10) 0.99703( 80) 0.99572( 79) 0.99672( 80) heu-met-fast-007 case-25 0.99970( 10) 0.99651( 85) 0.99586( 97) 0.99616( 98) heu-met-fast-007 case-26 0.99650( 20) 0.99515( 69) 0.99323( 76) 0.99672( 79) heu-met-fast-007 case-27 0.99870( 20) 0.99471( 75) 0.99457( 75) 0.99586( 78) heu-met-fast-007 case-28 0.99780( 20) 0.99513( 72) 0.99567( 83) 0.99787( 86) heu-met-fast-007 case-29 0.99810( 20) 0.99395( 86) 0.99374( 83) 0.99563( 86) heu-met-fast-007 case-30 1.00130( 20) 0.99737( 86) 0.99702( 82) 0.99573( 85) heu-met-fast-007 case-31 0.99590( 10) 1.00345( 61) 1.00051( 62) 1.00463( 63) heu-met-fast-007 case-32 0.99950( 10) 1.01180( 78) 1.01017( 62) 1.01068( 62) heu-met-fast-007 case-33 0.99770( 10) 1.01727( 68) 1.01380( 75) 1.01614( 75) heu-met-fast-007 case-34 1.00110( 10) 0.99098( 88) 0.99198( 82) 0.99089( 83) heu-met-fast-007 case-35 0.99990( 10) 0.99988( 78) 0.99986( 89) 0.99996( 90) heu-met-fast-007 case-36 0.99880( 10) 0.99828( 95) 0.99826( 92) 0.99946( 93) heu-met-fast-007 case-37 1.00000( 10) 1.00037( 99) 0.99950( 93) 0.99950( 94) heu-met-fast-007 case-38 1.00180( 10) 1.00060( 81) 1.00407( 78) 1.00227( 78) heu-met-fast-007 case-39 continued on next page

21616/05.69456/P 27-54 benchmark JEFF-3.0 JEFF-3.1 C/E (JEFF-3.1) name 1.00130( 10) 1.00229( 88) 1.00370( 88) 1.00240( 88) heu-met-fast-007 case-40 0.99940( 10) 0.99773( 81) 0.99881(108) 0.99941(109) heu-met-fast-007 case-41 1.00160( 10) 0.99931( 88) 0.99987( 78) 0.99827( 79) heu-met-fast-007 case-42 0.99980( 10) 0.99828( 93) 0.99862( 88) 0.99882( 89) heu-met-fast-007 case-43 1.00000(190) 0.99466( 45) 0.99428( 42) 0.99428(195) heu-met-fast-022 1.00000(250) 1.00359( 40) 1.00199( 40) 1.00199(253) heu-met-fast-027 1.00000(300) 0.99379( 43) 1.00275( 51) 1.00275(304) heu-met-fast-028 1.00000(200) 0.97977( 76) 0.97977(215) heu-met-fast-057 case-1 1.00000(230) 0.99069( 65) 0.99069(239) heu-met-fast-057 case-2 1.00000(320) 1.00645( 69) 1.00645(327) heu-met-fast-057 case-3 1.00000(400) 0.98084( 59) 0.98084(404) heu-met-fast-057 case-4 1.00000(190) 1.01016( 63) 1.01016(200) heu-met-fast-057 case-5 1.00000(290) 0.98755( 71) 0.98755(299) heu-met-fast-057 case-6 0.99550(240) 1.01549( 50) 1.02008(246) heu-met-fast-060 0.99960( 80) 1.00075( 69) 1.00115(106) heu-met-fast-064 case-1 0.99960(100) 1.00240( 79) 1.00280(127) heu-met-fast-064 case-2 0.99960( 90) 0.99964( 73) 1.00004(116) heu-met-fast-064 case-3 0.99590(240) 1.02962( 55) 1.03386(247) heu-met-fast-067 case-1 0.99380(240) 1.00714( 58) 1.01342(248) heu-met-fast-067 case-2 0.99910(240) 1.01169( 74) 1.01260(251) heu-met-fast-072 case-1 Table 3.1 The results for HEU benchmarks with a fast spectrum

3.1.2 Intermediate spectrum benchmark JEFF-3.0 JEFF-3.1 C/E (JEFF-3.1) name 1.00000(400) 1.00934( 52) 1.01087( 48) 1.01087(403) heu-comp-inter-004 1.03200(400) 1.02078( 39) 1.01756( 33) 0.98601(389) heu-comp-inter-005 case-1 1.05000(800) 1.09699( 44) 1.10783( 44) 1.05508(763) heu-comp-inter-005 case-2 1.03000(600) 1.04833( 39) 1.05784( 42) 1.02703(584) heu-comp-inter-005 case-3 1.06400(1800) 1.15813( 46) 1.15926( 48) 1.08953(1692) heu-comp-inter-005 case-4 0.99700(1300) 0.91850( 42) 0.93460( 39) 0.93741(1305) heu-comp-inter-005 case-5 1.00100(600) 1.00766( 46) 1.01742( 78) 1.01640(604) heu-met-inter-001 case-1 0.99770( 80) 0.99287( 83) 0.99130( 78) 0.99359(112) heu-met-inter-006 case-1 1.00010( 80) 0.99394( 90) 0.99540( 84) 0.99530(116) heu-met-inter-006 case-2 1.00150( 90) 0.99771( 87) 0.99780( 75) 0.99631(117) heu-met-inter-006 case-3 1.00160( 80) 1.00362( 80) 1.00299( 76) 1.00139(110) heu-met-inter-006 case-4 Table 3.2 The results for HEU benchmarks with an intermediate spectrum

3.1.3 Thermal spectrum benchmark JEFF-3.0 JEFF-3.1 C/E (JEFF-3.1) name 1.00100(600) 1.00672( 88) 1.00636( 84) 1.00535(605) heu-met-therm-001 simple 1.00100(600) 1.01575( 81) 1.00881( 90) 1.00780(606) heu-met-therm-001 detail 1.00000(100) 1.00116( 79) 1.00060( 78) 1.00060(127) heu-met-therm-003 case-1 continued on next page

28-54 21616/05.69456/P 1.04

JEFF-3.0 JEFF-3.1 1.03 benchmark uncertainty

1.02

1.01 eff

1 C/E for k

0.99

0.98 hmt03 hmt06 hst01 hst02 hst04 hst38 hst39 hst42 hst09 hst10 hst13 hst32 hmt01 hmt08 hmt09 hmt10 hmt13 hmt14 hmt16 hmt18

0.97

Figure 3.2 Results for the HEU benchmarks with a thermal spectrum

benchmark JEFF-3.0 JEFF-3.1 C/E (JEFF-3.1) name 0.99100(300) 0.98122( 79) 0.98201( 82) 0.99093(314) heu-met-therm-003 case-2 0.98260(600) 0.97567( 91) 0.97550( 79) 0.99277(616) heu-met-therm-003 case-3 0.98760(400) 0.98071( 78) 0.98309( 92) 0.99543(416) heu-met-therm-003 case-4 0.99300(300) 0.99867( 90) 0.99847( 91) 1.00551(316) heu-met-therm-003 case-5 0.98890(300) 0.96585( 94) 0.96769( 85) 0.97855(316) heu-met-therm-003 case-6 0.99190(300) 0.98277( 85) 0.98360( 87) 0.99163(315) heu-met-therm-003 case-7 1.00000(440) 0.99450(100) 0.99451( 94) 0.99451(450) heu-met-therm-006 case-1 1.00000(400) 0.99522( 90) 0.99492( 92) 0.99492(411) heu-met-therm-006 case-2 1.00000(400) 0.99858( 80) 0.99879( 92) 0.99879(410) heu-met-therm-006 case-3 1.00000(400) 0.99376( 89) 0.99245( 90) 0.99245(410) heu-met-therm-006 case-4 1.00000(400) 0.99194( 89) 0.99324( 85) 0.99324(409) heu-met-therm-006 case-5 1.00000(400) 0.98783( 79) 0.99177( 83) 0.99177(409) heu-met-therm-006 case-6 1.00000(400) 0.98712( 79) 0.98891( 75) 0.98891(407) heu-met-therm-006 case-7 1.00000(400) 0.98251( 79) 0.98424( 72) 0.98424(407) heu-met-therm-006 case-8 1.00000(400) 0.98543( 69) 0.98618( 77) 0.98618(408) heu-met-therm-006 case-9 1.00000(400) 1.00221( 90) 1.00216( 72) 1.00216(406) heu-met-therm-006 case-10 1.00000(400) 0.99223( 89) 0.99279( 86) 0.99279(409) heu-met-therm-006 case-11 1.00000(400) 0.99368( 80) 0.99486( 78) 0.99486(408) heu-met-therm-006 case-12 1.00000(610) 1.01568(102) 1.01365( 89) 1.01365(616) heu-met-therm-006 case-13 1.00000(400) 0.98394( 89) 0.98754( 80) 0.98754(408) heu-met-therm-006 case-14 continued on next page

21616/05.69456/P 29-54 benchmark JEFF-3.0 JEFF-3.1 C/E (JEFF-3.1) name 1.00000(400) 0.98100( 78) 0.98268( 88) 0.98268(410) heu-met-therm-006 case-15 1.00000(400) 0.99494( 80) 0.99444( 68) 0.99444(406) heu-met-therm-006 case-16 1.00000(400) 0.99082( 89) 0.98973( 94) 0.98973(411) heu-met-therm-006 case-17 1.00000(400) 0.99494( 90) 0.99355( 92) 0.99355(411) heu-met-therm-006 case-18 1.00000(400) 0.99402( 70) 0.99474( 77) 0.99474(407) heu-met-therm-006 case-19 1.00000(400) 0.99360( 80) 0.99459( 87) 0.99459(409) heu-met-therm-006 case-20 1.00000(400) 0.99774( 90) 0.99796( 77) 0.99796(407) heu-met-therm-006 case-21 1.00000(400) 0.99746( 90) 0.99612( 87) 0.99612(409) heu-met-therm-006 case-22 1.00000(400) 0.99961( 90) 0.99933( 90) 0.99933(410) heu-met-therm-006 case-23 1.00090(520) 1.00995( 85) 1.00942( 90) 1.00851(527) heu-met-therm-008 detail 1.00320(630) 1.00243( 92) 1.00373( 82) 1.00053(633) heu-met-therm-009 simple 1.00300(720) 1.01110( 78) 1.00933( 87) 1.00631(723) heu-met-therm-010 7.5mil 1.00260(720) 1.00813( 87) 1.00926( 92) 1.00664(724) heu-met-therm-010 15mil 1.00210(220) 1.00563( 78) 1.00691( 75) 1.00480(232) heu-met-therm-013 625in 0.99830(200) 1.00771( 81) 1.00851( 91) 1.01023(220) heu-met-therm-013 15mil 0.99390(150) 1.00804( 99) 1.00609(103) 1.01226(182) heu-met-therm-014 simple 1.00170(160) 0.99850(101) 1.00084( 90) 0.99914(183) heu-met-therm-016 detail 1.00380(410) 1.00115( 83) 1.00067( 91) 0.99688(418) heu-met-therm-018 simple 1.00380(410) 0.99943( 88) 1.00065( 84) 0.99686(417) heu-met-therm-018 detail 1.00040(600) 0.99877(118) 0.99645(107) 0.99605(609) heu-sol-therm-001 case-1 1.00210(720) 0.99592(112) 0.99426(117) 0.99218(728) heu-sol-therm-001 case-2 1.00030(350) 1.00484(128) 1.00241(114) 1.00211(368) heu-sol-therm-001 case-3 1.00080(530) 1.00030(123) 0.99690(106) 0.99610(540) heu-sol-therm-001 case-4 1.00010(490) 1.00125(100) 1.00008( 84) 0.99998(497) heu-sol-therm-001 case-5 1.00020(460) 1.00552( 90) 1.00374(100) 1.00354(471) heu-sol-therm-001 case-6 1.00080(400) 0.99810(110) 0.99886(104) 0.99806(413) heu-sol-therm-001 case-7 0.99980(380) 1.00027(121) 0.99799(113) 0.99819(397) heu-sol-therm-001 case-8 1.00080(540) 0.99530(119) 0.99513(107) 0.99433(550) heu-sol-therm-001 case-9 0.99930(540) 0.99205( 99) 0.99274(550) heu-sol-therm-001 case-10 1.00250(580) 1.00297(110) 1.00412(118) 1.00162(590) heu-sol-therm-002 case-1 1.00280(580) 1.00888(101) 1.00819(117) 1.00537(590) heu-sol-therm-002 case-2 1.00330(680) 1.00132(100) 0.99814(127) 0.99486(690) heu-sol-therm-002 case-3 1.00340(690) 1.00489(111) 1.00507(103) 1.00166(695) heu-sol-therm-002 case-4 1.00180(440) 1.00772(121) 1.00389( 92) 1.00209(449) heu-sol-therm-002 case-5 1.00230(410) 1.01010(111) 1.00948(108) 1.00716(423) heu-sol-therm-002 case-6 1.00250(500) 1.00275(120) 1.00097(103) 0.99847(509) heu-sol-therm-002 case-7 1.00300(550) 1.00715(101) 1.00451(106) 1.00151(558) heu-sol-therm-002 case-8 1.00120(460) 1.00438( 90) 1.00263( 92) 1.00143(469) heu-sol-therm-002 case-9 1.00240(500) 1.00669( 91) 1.00726( 94) 1.00485(507) heu-sol-therm-002 case-10 1.00170(380) 1.00592( 90) 1.00478(101) 1.00307(392) heu-sol-therm-002 case-11 continued on next page

30-54 21616/05.69456/P benchmark JEFF-3.0 JEFF-3.1 C/E (JEFF-3.1) name 1.00270(500) 1.01237(101) 1.01150(107) 1.00878(510) heu-sol-therm-002 case-12 1.00250(550) 0.99990(120) 0.99949(105) 0.99700(559) heu-sol-therm-002 case-13 1.00310(660) 1.00818(101) 1.01028(117) 1.00716(668) heu-sol-therm-002 case-14 1.00000(330) 0.98387( 97) 0.98387(344) heu-sol-therm-004 case-1 1.00000(360) 0.98053(111) 0.98053(377) heu-sol-therm-004 case-2 1.00000(390) 0.98696(103) 0.98696(404) heu-sol-therm-004 case-3 1.00000(460) 0.99029(108) 0.99029(473) heu-sol-therm-004 case-4 1.00000(520) 0.98902(121) 0.98902(534) heu-sol-therm-004 case-5 1.00000(590) 0.98441(114) 0.98441(601) heu-sol-therm-004 case-6 0.99900(430) 0.99963(111) 1.00063(100) 1.00163(442) heu-sol-therm-009 case-1 1.00000(390) 0.99790(108) 1.00068( 96) 1.00068(402) heu-sol-therm-009 case-2 1.00000(360) 0.99874( 97) 0.99955(107) 0.99955(376) heu-sol-therm-009 case-3 0.99860(350) 0.99432( 94) 0.99647(101) 0.99787(365) heu-sol-therm-009 case-4 1.00000(290) 0.99994( 88) 1.00173( 98) 1.00173(306) heu-sol-therm-010 case-1 1.00000(290) 1.00247( 91) 0.99978(100) 0.99978(307) heu-sol-therm-010 case-2 1.00000(290) 0.99949( 89) 0.99666( 99) 0.99666(307) heu-sol-therm-010 case-3 0.99920(290) 0.99780( 85) 0.99653(104) 0.99733(308) heu-sol-therm-010 case-4 1.00120(260) 0.99888( 60) 0.99757( 59) 0.99637(266) heu-sol-therm-013 case-1 1.00070(360) 0.99861( 60) 0.99896( 57) 0.99826(364) heu-sol-therm-013 case-2 1.00090(360) 0.99448( 70) 0.99406( 58) 0.99317(364) heu-sol-therm-013 case-3 1.00030(360) 0.99666( 70) 0.99622( 68) 0.99592(366) heu-sol-therm-013 case-4 1.00150(260) 0.99855( 37) 0.99705(262) heu-sol-therm-032 case-1 1.00000(250) 0.99514( 50) 0.99475( 52) 0.99475(255) heu-sol-therm-038 case-1 1.00000(250) 0.99513( 45) 0.99530( 51) 0.99530(255) heu-sol-therm-038 case-2 1.00000(250) 0.99625( 48) 0.99696( 52) 0.99696(255) heu-sol-therm-038 case-3 1.00000(250) 0.99551( 50) 0.99568( 56) 0.99568(256) heu-sol-therm-038 case-4 1.00000(250) 0.99595( 60) 0.99326( 61) 0.99326(257) heu-sol-therm-038 case-5 1.00000(250) 0.99538( 55) 0.99518( 55) 0.99518(256) heu-sol-therm-038 case-6 1.00000(320) 0.99697( 53) 0.99716( 52) 0.99716(324) heu-sol-therm-038 case-7 1.00000(260) 0.99777( 54) 0.99772( 47) 0.99772(264) heu-sol-therm-038 case-8 1.00000(330) 0.99746( 48) 0.99715( 53) 0.99715(334) heu-sol-therm-038 case-9 1.00000(260) 0.99782( 59) 0.99593( 52) 0.99593(265) heu-sol-therm-038 case-10 1.00000(250) 0.99656( 48) 0.99620( 56) 0.99620(256) heu-sol-therm-038 case-11 1.00000(250) 0.99715( 59) 0.99658( 63) 0.99658(258) heu-sol-therm-038 case-12 1.00000(500) 1.00081( 56) 1.00117( 52) 1.00117(503) heu-sol-therm-038 case-13 1.00000(500) 1.00280( 50) 1.00136( 61) 1.00136(504) heu-sol-therm-038 case-14 1.00000(500) 1.00094( 50) 1.00153( 48) 1.00153(502) heu-sol-therm-038 case-15 1.00000(500) 1.00049( 54) 0.99994( 51) 0.99994(503) heu-sol-therm-038 case-16 1.00000(260) 0.99705( 58) 0.99638( 57) 0.99638(266) heu-sol-therm-038 case-17 1.00000(320) 0.99655( 56) 0.99667( 56) 0.99667(325) heu-sol-therm-038 case-18 continued on next page

21616/05.69456/P 31-54 benchmark JEFF-3.0 JEFF-3.1 C/E (JEFF-3.1) name 1.00000(320) 0.99766( 64) 0.99662( 54) 0.99662(325) heu-sol-therm-038 case-19 1.00000(320) 0.99702( 54) 0.99826( 56) 0.99826(325) heu-sol-therm-038 case-20 1.00000(250) 0.99553( 47) 0.99547( 50) 0.99547(255) heu-sol-therm-038 case-21 1.00000(270) 0.99739( 52) 0.99811( 52) 0.99811(275) heu-sol-therm-038 case-22 1.00000(270) 0.99718( 48) 0.99715( 57) 0.99715(276) heu-sol-therm-038 case-23 1.00000(260) 0.99730( 54) 0.99673( 46) 0.99673(264) heu-sol-therm-038 case-24 1.00000(320) 0.99848( 49) 0.99763( 44) 0.99763(323) heu-sol-therm-038 case-25 1.00000(320) 0.99661( 49) 0.99839( 50) 0.99839(324) heu-sol-therm-038 case-26 1.00000(320) 0.99900( 50) 0.99751( 53) 0.99751(324) heu-sol-therm-038 case-27 1.00000(250) 0.99290( 58) 0.99290(257) heu-sol-therm-038 case-28 1.00000(250) 1.03988( 55) 1.04047( 50) 1.04047(255) heu-sol-therm-038 case-29 1.00000(270) 1.00572( 52) 1.00694( 52) 1.00694(275) heu-sol-therm-038 case-30 1.00000(570) 1.04080( 97) 1.03778(100) 1.03778(578) heu-sol-therm-039 case-1 1.00000(510) 1.03073( 93) 1.02986(101) 1.02986(519) heu-sol-therm-039 case-2 1.00120(540) 1.02917(106) 1.02780( 94) 1.02657(547) heu-sol-therm-039 case-3 1.00180(610) 1.03329( 92) 1.03237( 97) 1.03052(616) heu-sol-therm-039 case-4 1.00180(550) 1.04789( 94) 1.04286( 99) 1.04099(557) heu-sol-therm-039 case-5 1.00250(450) 1.02990( 90) 1.02762( 99) 1.02506(459) heu-sol-therm-039 case-6 0.99570(390) 0.99617( 48) 0.99683( 47) 1.00113(395) heu-sol-therm-042 case-1 0.99650(360) 0.99687( 53) 0.99690( 39) 1.00040(363) heu-sol-therm-042 case-2 0.99940(280) 1.00016( 41) 1.00007( 41) 1.00067(283) heu-sol-therm-042 case-3 1.00000(340) 1.00117( 36) 1.00163( 40) 1.00163(342) heu-sol-therm-042 case-4 1.00000(340) 0.99877( 29) 0.99916( 28) 0.99916(341) heu-sol-therm-042 case-5 1.00000(370) 0.99948( 32) 0.99834( 35) 0.99834(372) heu-sol-therm-042 case-6 1.00000(360) 0.99874( 28) 1.00054( 23) 1.00054(361) heu-sol-therm-042 case-7 1.00000(350) 1.00154( 24) 1.00082( 22) 1.00082(351) heu-sol-therm-042 case-8 Table 3.3 The results for HEU benchmarks with a thermal spectrum

3.1.4 Mixed spectrum benchmark JEFF-3.0 JEFF-3.1 C/E (JEFF-3.1) name 1.00000( 10) 0.99874( 77) 1.00217( 75) 1.00217( 76) heu-comp-mixed-003 case-5 1.00070(270) 1.00175( 88) 1.00105(284) heu-met-mixed-005 case-1 1.00030(280) 1.01274( 85) 1.01244(292) heu-met-mixed-005 case-2 1.00120(290) 1.01230( 90) 1.01109(303) heu-met-mixed-005 case-3 1.00160(300) 1.00569( 81) 1.00408(310) heu-met-mixed-005 case-4 1.00050(400) 0.99826( 86) 0.99776(409) heu-met-mixed-005 case-5

3.2 IEU results

3.2.1 Fast spectrum benchmark JEFF-3.0 JEFF-3.1 C/E (JEFF-3.1) name 0.99880( 90) 0.99636( 67) 0.99678( 60) 0.99798(108) ieu-met-fast-001 case-1 continued on next page

32-54 21616/05.69456/P 1.04 JEFF-3.0 JEFF-3.1 benchmark uncertainty

1.03

1.02

1.01 eff

1 C/E for k

0.99

0.98 Jemima ici01 imf02 BigTen ZPR-U9 ZPR3/41 ZPR9-2,3

0.97

Figure 3.3 Results for the IEU benchmarks with a fast or intermediate spectrum

benchmark JEFF-3.0 JEFF-3.1 C/E (JEFF-3.1) name 0.99880( 90) 0.99615( 62) 0.99861( 64) 0.99981(111) ieu-met-fast-001 case-2 0.99900( 30) 0.99523( 71) 0.99830( 63) 0.99930( 70) ieu-met-fast-001 case-3 0.99900( 30) 0.99576( 60) 0.99851( 70) 0.99951( 76) ieu-met-fast-001 case-4 0.99890( 90) 0.99521( 72) 0.99773( 55) 0.99883(106) ieu-met-fast-001 case-1i 0.99970( 90) 0.99622( 67) 0.99775( 63) 0.99805(110) ieu-met-fast-001 case-2i 0.99930( 30) 0.99644( 59) 0.99828( 63) 0.99898( 70) ieu-met-fast-001 case-3i 1.00020( 30) 0.99532( 70) 0.99843( 59) 0.99823( 66) ieu-met-fast-001 case-4i 1.00000(300) 0.99053( 59) 0.99156( 29) 0.99156(301) ieu-met-fast-002 case-1 1.00450( 70) 1.00058( 28) 0.99798( 28) 0.99351( 75) ieu-met-fast-007 detail 1.00450( 70) 0.99777( 27) 0.99330( 75) ieu-met-fast-007 simple 0.99480(130) 0.98774( 25) 0.99290(133) ieu-met-fast-007 twozone 0.99540(240) 0.99411( 31) 0.98942( 42) 0.99399(245) ieu-met-fast-010 case-1 1.00070(270) 0.99842( 60) 1.00137( 63) 1.00067(277) ieu-met-fast-012 case-1 0.99580(220) 1.00074( 27) 0.99868( 53) 1.00289(227) ieu-met-fast-014 case-1 0.99270(220) 1.00206( 26) 1.00153( 57) 1.00889(229) ieu-met-fast-014 case-2 Table 3.4 The results for IEU benchmarks with a fast spectrum

3.2.2 Intermediate spectrum benchmark JEFF-3.0 JEFF-3.1 C/E (JEFF-3.1) name 0.96900(500) 1.01364( 29) 0.99624( 35) 1.02811(517) ieu-comp-inter-001 case-1 0.98000(300) 1.02175( 41) 0.98696( 42) 1.00710(309) ieu-comp-inter-001 case-2 continued on next page

21616/05.69456/P 33-54 benchmark JEFF-3.0 JEFF-3.1 C/E (JEFF-3.1) name 1.01400(600) 1.04810( 56) 1.00623( 51) 0.99234(594) ieu-comp-inter-001 case-3 0.96400(1200) 0.94622( 60) 0.92252( 64) 0.95697(1247) ieu-comp-inter-001 case-4 Table 3.5 The results for IEU benchmarks with an intermediate spectrum

3.2.3 Thermal spectrum

1.02 JEFF-3.0 JEFF-3.1 benchmark uncertainty 1.015

1.01

1.005

eff 1

C/E for k 0.995

0.99

0.985

0.98 ict02 ict03 Proteus

0.975

Figure 3.4 Results for the IEU benchmarks with a thermal spectrum

benchmark JEFF-3.0 JEFF-3.1 C/E (JEFF-3.1) name 1.00140(390) 0.99364( 80) 0.99461( 83) 0.99322(398) ieu-comp-therm-002 case-1 1.00190(400) 0.99644( 70) 0.99723( 26) 0.99534(400) ieu-comp-therm-002 case-2 1.00170(440) 0.99987( 80) 0.99941( 82) 0.99771(447) ieu-comp-therm-002 case-3 1.00190(440) 0.99845( 90) 0.99932( 76) 0.99742(446) ieu-comp-therm-002 case-4 1.00140(430) 0.99356( 80) 0.99441( 78) 0.99302(437) ieu-comp-therm-002 case-5 1.00160(440) 0.99241( 99) 0.99323( 77) 0.99164(446) ieu-comp-therm-002 case-6 1.00060(560) 1.00130( 85) 1.00178( 85) 1.00118(566) ieu-comp-therm-003 core-132 1.00460(560) 1.00704( 83) 1.00701( 87) 1.00240(564) ieu-comp-therm-003 core-133 1.01120( 50) 1.00760( 90) 1.00695( 26) 0.99580( 56) ieu-comp-therm-proteus Table 3.6 The results for IEU benchmarks with a thermal spectrum

3.3 LEU results

3.3.1 Thermal spectrum

34-54 21616/05.69456/P 1.02 JEFF-3.0 JEFF-3.1 benchmark uncertainty 1.015

1.01

1.005

eff 1

C/E for k 0.995

0.99

0.985

0.98 lct01 lct03 lct05 lct06 lct07 lct09 lct10 lct02

0.975

Figure 3.5 Results for the LEU benchmarks with a thermal spectrum

1.02 JEFF-3.0 JEFF-3.1 benchmark uncertainty 1.015

1.01

1.005

eff 1

C/E for k 0.995

0.99

0.985

0.98 lct16 lct17 lct39 lct51 lct60 TRX lct19 lmt01 Dimple

0.975

Figure 3.6 Results for the LEU benchmarks with a thermal spectrum (continued)

benchmark JEFF-3.0 JEFF-3.1 C/E (JEFF-3.1) name 1.00000(310) 0.99617( 70) 0.99913( 76) 0.99913(319) leu-comp-therm-001 case-1 continued on next page

21616/05.69456/P 35-54 1.02 JEFF-3.0 JEFF-3.1 benchmark uncertainty 1.015

1.01

1.005

eff 1

C/E for k 0.995

0.99

0.985

0.98 lst03 lst04 lst16 lst17 lst18 lst01 lst07 lst20 lst21

0.975

Figure 3.7 Results for the LEU benchmarks with a thermal spectrum (continued)

benchmark JEFF-3.0 JEFF-3.1 C/E (JEFF-3.1) name 0.99980(300) 0.99443( 70) 0.99652( 67) 0.99672(308) leu-comp-therm-001 case-2 0.99980(300) 0.99355( 70) 0.99873( 64) 0.99893(307) leu-comp-therm-001 case-3 0.99980(300) 0.99400( 60) 0.99815( 68) 0.99835(308) leu-comp-therm-001 case-4 0.99980(300) 0.99304( 70) 0.99712( 62) 0.99732(306) leu-comp-therm-001 case-5 0.99980(300) 0.99437( 70) 0.99760( 71) 0.99780(308) leu-comp-therm-001 case-6 0.99980(310) 0.99472( 70) 0.99644( 67) 0.99664(317) leu-comp-therm-001 case-7 0.99980(300) 0.99295( 70) 0.99704( 68) 0.99724(308) leu-comp-therm-001 case-8 0.99970(200) 0.99478( 90) 0.99854( 88) 0.99884(219) leu-comp-therm-002 case-1 0.99970(200) 0.99394( 92) 0.99941( 81) 0.99971(216) leu-comp-therm-002 case-2 0.99970(200) 0.99430( 82) 0.99784( 83) 0.99814(217) leu-comp-therm-002 case-3 0.99970(180) 0.99409( 73) 0.99818( 82) 0.99848(198) leu-comp-therm-002 case-4 0.99970(190) 0.99130( 80) 0.99533( 78) 0.99563(206) leu-comp-therm-002 case-5 1.00000(390) 0.98456( 80) 0.98878( 66) 0.98878(396) leu-comp-therm-003 case-1 1.00000(390) 0.98554( 74) 0.98932( 83) 0.98932(399) leu-comp-therm-003 case-2 1.00000(390) 0.98636( 75) 0.99102( 72) 0.99102(397) leu-comp-therm-003 case-3 1.00000(390) 0.98537( 77) 0.99083( 73) 0.99083(397) leu-comp-therm-003 case-4 1.00000(390) 0.98656( 76) 0.98885( 77) 0.98885(398) leu-comp-therm-003 case-5 1.00000(390) 0.98353( 73) 0.98844( 78) 0.98844(398) leu-comp-therm-003 case-6 1.00000(390) 0.98712( 76) 0.99073( 74) 0.99073(397) leu-comp-therm-003 case-7 1.00000(390) 0.98820( 71) 0.99250( 79) 0.99250(398) leu-comp-therm-003 case-8 continued on next page

36-54 21616/05.69456/P benchmark JEFF-3.0 JEFF-3.1 C/E (JEFF-3.1) name 1.00000(390) 0.98071( 74) 0.98532( 70) 0.98532(396) leu-comp-therm-003 case-9 1.00000(390) 0.98090( 90) 0.98443( 74) 0.98443(397) leu-comp-therm-003 case-10 1.00000(390) 0.98055( 74) 0.98455( 83) 0.98455(399) leu-comp-therm-003 case-11 1.00000(390) 0.97860( 65) 0.98495( 60) 0.98495(395) leu-comp-therm-003 case-12 1.00000(390) 0.98200( 74) 0.98888( 71) 0.98888(397) leu-comp-therm-003 case-13 1.00000(390) 0.98177( 81) 0.98698( 80) 0.98698(398) leu-comp-therm-003 case-14 1.00000(390) 0.98395( 63) 0.98758( 76) 0.98758(398) leu-comp-therm-003 case-15 1.00000(390) 0.98295( 69) 0.98570( 69) 0.98570(396) leu-comp-therm-003 case-16 1.00000(390) 0.98218( 77) 0.98641( 80) 0.98641(398) leu-comp-therm-003 case-17 1.00000(390) 0.98065( 68) 0.98550( 66) 0.98550(396) leu-comp-therm-003 case-18 1.00000(390) 0.98235( 78) 0.98780( 74) 0.98780(397) leu-comp-therm-003 case-19 1.00000(390) 0.98377( 77) 0.98747( 73) 0.98747(397) leu-comp-therm-003 case-20 1.00000(390) 0.98051( 76) 0.98497( 67) 0.98497(396) leu-comp-therm-003 case-21 1.00000(390) 0.99300( 78) 0.99683( 69) 0.99683(396) leu-comp-therm-003 case-22 1.00000(230) 1.00009( 25) 1.00384( 87) 1.00384(246) leu-comp-therm-005 case-1 1.00000(210) 0.99843( 23) 1.00098( 80) 1.00098(225) leu-comp-therm-005 case-2 1.00000(290) 0.99981( 24) 1.00359( 74) 1.00359(299) leu-comp-therm-005 case-3 1.00000(250) 0.99914( 23) 1.00237( 85) 1.00237(264) leu-comp-therm-005 case-4 1.00000(470) 1.00011( 26) 1.00501( 71) 1.00501(475) leu-comp-therm-005 case-5 1.00000(420) 1.00161( 27) 1.00514( 93) 1.00514(430) leu-comp-therm-005 case-6 1.00000(430) 0.99908( 27) 1.00323( 85) 1.00323(438) leu-comp-therm-005 case-7 1.00000(210) 1.00105( 26) 1.00530( 73) 1.00530(222) leu-comp-therm-005 case-8 1.00000(400) 1.00211( 23) 1.00568( 85) 1.00568(409) leu-comp-therm-005 case-9 1.00000(280) 1.00139( 24) 1.00389( 92) 1.00389(295) leu-comp-therm-005 case-10 1.00000(430) 1.00243( 24) 1.00619( 88) 1.00619(439) leu-comp-therm-005 case-11 1.00000(660) 1.00047( 27) 1.00451( 76) 1.00451(664) leu-comp-therm-005 case-12 1.00000(640) 1.00700( 25) 1.01153( 78) 1.01153(645) leu-comp-therm-005 case-13 1.00000(200) 0.99425( 21) 0.99871( 78) 0.99871(215) leu-comp-therm-005 case-14 1.00000(200) 1.01426( 24) 1.01881( 80) 1.01881(215) leu-comp-therm-005 case-15 1.00000(320) 1.00709( 23) 1.01202( 69) 1.01202(327) leu-comp-therm-005 case-16 1.00000(200) 0.99627( 75) 0.99973( 23) 0.99973(201) leu-comp-therm-006 case-1 1.00000(200) 0.99588( 75) 0.99971( 24) 0.99971(201) leu-comp-therm-006 case-2 1.00000(200) 0.99564( 78) 0.99945( 24) 0.99945(201) leu-comp-therm-006 case-3 1.00000(200) 0.99726( 71) 0.99959( 25) 0.99959(202) leu-comp-therm-006 case-4 1.00000(200) 0.99579( 68) 0.99985( 24) 0.99985(201) leu-comp-therm-006 case-5 1.00000(200) 0.99506( 79) 1.00051( 26) 1.00051(202) leu-comp-therm-006 case-6 1.00000(200) 0.99555( 75) 1.00096( 23) 1.00096(201) leu-comp-therm-006 case-7 1.00000(200) 0.99440( 88) 1.00067( 26) 1.00067(202) leu-comp-therm-006 case-8 1.00000(200) 0.99468( 82) 0.99991( 23) 0.99991(201) leu-comp-therm-006 case-9 1.00000(200) 0.99685( 68) 1.00000( 22) 1.00000(201) leu-comp-therm-006 case-10 continued on next page

21616/05.69456/P 37-54 benchmark JEFF-3.0 JEFF-3.1 C/E (JEFF-3.1) name 1.00000(200) 0.99534( 71) 0.99968( 24) 0.99968(201) leu-comp-therm-006 case-11 1.00000(200) 0.99613( 74) 0.99935( 22) 0.99935(201) leu-comp-therm-006 case-12 1.00000(200) 0.99499( 73) 0.99959( 24) 0.99959(201) leu-comp-therm-006 case-13 1.00000(200) 0.99409( 74) 0.99976( 25) 0.99976(202) leu-comp-therm-006 case-14 1.00000(200) 0.99584( 73) 0.99956( 23) 0.99956(201) leu-comp-therm-006 case-15 1.00000(200) 0.99490( 70) 0.99946( 20) 0.99946(201) leu-comp-therm-006 case-16 1.00000(200) 0.99513( 74) 0.99985( 24) 0.99985(201) leu-comp-therm-006 case-17 1.00000(200) 0.99740( 73) 0.99977( 22) 0.99977(201) leu-comp-therm-006 case-18 1.00000(160) 0.99113( 88) 0.99776( 41) 0.99776(165) leu-comp-therm-007 case-1 1.00000(160) 0.99604( 94) 0.99991( 42) 0.99991(165) leu-comp-therm-007 case-2 1.00000(160) 0.99561( 78) 0.99728( 39) 0.99728(165) leu-comp-therm-007 case-3 1.00000(160) 0.99484( 66) 0.99722( 33) 0.99722(163) leu-comp-therm-007 case-4 1.00000(160) 0.99411( 87) 0.99642( 42) 0.99642(165) leu-comp-therm-007 case-5 1.00000(160) 0.99652( 88) 0.99918( 45) 0.99918(166) leu-comp-therm-007 case-6 1.00000(160) 0.99674( 84) 0.99830( 40) 0.99830(165) leu-comp-therm-007 case-7 1.00000(160) 0.99424( 87) 0.99806( 45) 0.99806(166) leu-comp-therm-007 case-8 1.00000(160) 0.99537( 89) 0.99870( 38) 0.99870(164) leu-comp-therm-007 case-9 1.00000(160) 0.99747( 81) 0.99829( 42) 0.99829(165) leu-comp-therm-007 case-10 1.00000(210) 0.99443( 80) 0.99790( 76) 0.99790(223) leu-comp-therm-009 case-1 1.00000(210) 0.99304( 79) 0.99799( 71) 0.99799(222) leu-comp-therm-009 case-2 1.00000(210) 0.99304( 79) 0.99866( 87) 0.99866(227) leu-comp-therm-009 case-3 1.00000(210) 0.99292( 79) 0.99947( 70) 0.99947(221) leu-comp-therm-009 case-4 1.00000(210) 0.99538( 80) 0.99945( 79) 0.99945(224) leu-comp-therm-009 case-5 1.00000(210) 0.99515( 80) 0.99782( 81) 0.99782(225) leu-comp-therm-009 case-6 1.00000(210) 0.99735( 80) 0.99801( 81) 0.99801(225) leu-comp-therm-009 case-7 1.00000(210) 0.99472( 80) 0.99886( 83) 0.99886(226) leu-comp-therm-009 case-8 1.00000(210) 0.99589( 85) 0.99898( 75) 0.99898(223) leu-comp-therm-009 case-9 1.00000(210) 0.99397( 90) 0.99768( 88) 0.99768(228) leu-comp-therm-009 case-10 1.00000(210) 0.99433( 80) 0.99811( 79) 0.99811(224) leu-comp-therm-009 case-11 1.00000(210) 0.99535( 70) 1.00028( 84) 1.00028(226) leu-comp-therm-009 case-12 1.00000(210) 0.99558( 80) 0.99780( 70) 0.99780(221) leu-comp-therm-009 case-13 1.00000(210) 0.99213( 79) 0.99760( 77) 0.99760(224) leu-comp-therm-009 case-14 1.00000(210) 0.99603( 80) 0.99920( 77) 0.99920(224) leu-comp-therm-009 case-15 1.00000(210) 0.99379( 89) 0.99728( 78) 0.99728(224) leu-comp-therm-009 case-16 1.00000(210) 0.99409( 80) 0.99880( 72) 0.99880(222) leu-comp-therm-009 case-17 1.00000(210) 0.99548( 70) 0.99712( 74) 0.99712(223) leu-comp-therm-009 case-18 1.00000(210) 0.99530( 80) 0.99868( 73) 0.99868(222) leu-comp-therm-009 case-19 1.00000(210) 0.99413( 90) 0.99776( 84) 0.99776(226) leu-comp-therm-009 case-20 1.00000(210) 0.99596( 80) 0.99797( 72) 0.99797(222) leu-comp-therm-009 case-21 1.00000(210) 0.99436( 80) 0.99851( 76) 0.99851(223) leu-comp-therm-009 case-22 continued on next page

38-54 21616/05.69456/P benchmark JEFF-3.0 JEFF-3.1 C/E (JEFF-3.1) name 1.00000(210) 0.99552( 70) 0.99906( 84) 0.99906(226) leu-comp-therm-009 case-23 1.00000(210) 0.99443( 70) 0.99726( 80) 0.99726(225) leu-comp-therm-009 case-24 1.00000(210) 0.99468( 70) 0.99824( 73) 0.99824(222) leu-comp-therm-009 case-25 1.00000(210) 0.99550( 70) 0.99809( 80) 0.99809(225) leu-comp-therm-009 case-26 1.00000(210) 0.99598( 70) 0.99764( 78) 0.99764(224) leu-comp-therm-009 case-27 1.00000(210) 1.00538( 82) 1.00598( 75) 1.00598(223) leu-comp-therm-010 case-1 1.00000(210) 1.00620( 84) 1.00532( 73) 1.00532(222) leu-comp-therm-010 case-2 1.00000(210) 1.00325( 75) 1.00570( 79) 1.00570(224) leu-comp-therm-010 case-3 1.00000(210) 0.99286( 85) 0.99459( 77) 0.99459(224) leu-comp-therm-010 case-4 1.00000(210) 0.99654( 71) 0.99926( 79) 0.99926(224) leu-comp-therm-010 case-5 1.00000(210) 0.99584( 70) 0.99861( 72) 0.99861(222) leu-comp-therm-010 case-6 1.00000(210) 0.99606( 74) 1.00018( 68) 1.00018(221) leu-comp-therm-010 case-7 1.00000(210) 0.99330( 72) 0.99666( 81) 0.99666(225) leu-comp-therm-010 case-8 1.00000(210) 0.99743( 76) 1.00025( 82) 1.00025(225) leu-comp-therm-010 case-9 1.00000(210) 0.99708( 75) 0.99944( 80) 0.99944(225) leu-comp-therm-010 case-10 1.00000(210) 0.99705( 74) 1.00007( 84) 1.00007(226) leu-comp-therm-010 case-11 1.00000(210) 0.99698( 79) 0.99878( 76) 0.99878(223) leu-comp-therm-010 case-12 1.00000(210) 0.99415( 74) 0.99597( 75) 0.99597(223) leu-comp-therm-010 case-13 1.00000(280) 0.99659( 88) 1.00105( 84) 1.00105(292) leu-comp-therm-010 case-14 1.00000(280) 0.99823( 90) 1.00249( 77) 1.00249(290) leu-comp-therm-010 case-15 1.00000(280) 0.99962( 86) 0.99999( 71) 0.99999(289) leu-comp-therm-010 case-16 1.00000(280) 0.99847( 78) 1.00224( 79) 1.00224(291) leu-comp-therm-010 case-17 1.00000(280) 0.99752( 79) 1.00041( 84) 1.00041(292) leu-comp-therm-010 case-18 1.00000(280) 0.99595( 78) 1.00076(106) 1.00076(299) leu-comp-therm-010 case-19 1.00000(280) 1.00199( 79) 1.00473( 90) 1.00473(294) leu-comp-therm-010 case-20 1.00000(280) 1.00091( 69) 1.00437( 87) 1.00437(293) leu-comp-therm-010 case-21 1.00000(280) 0.99913( 80) 1.00374( 74) 1.00374(290) leu-comp-therm-010 case-22 1.00000(280) 0.99581( 71) 1.00185( 81) 1.00185(291) leu-comp-therm-010 case-23 1.00000(280) 0.99484( 67) 0.99784( 92) 0.99784(295) leu-comp-therm-010 case-24 1.00000(280) 0.99593( 78) 1.00068( 76) 1.00068(290) leu-comp-therm-010 case-25 1.00000(280) 0.99624( 75) 1.00216( 86) 1.00216(293) leu-comp-therm-010 case-26 1.00000(280) 0.99407( 80) 1.00101( 78) 1.00101(291) leu-comp-therm-010 case-27 1.00000(280) 0.99601( 83) 1.00253( 79) 1.00253(291) leu-comp-therm-010 case-28 1.00000(280) 0.99691( 86) 0.99881( 74) 0.99881(290) leu-comp-therm-010 case-29 1.00000(280) 0.99452( 76) 0.99787( 83) 0.99787(292) leu-comp-therm-010 case-30 1.00000(310) 0.99509( 70) 0.99891( 78) 0.99891(320) leu-comp-therm-016 case-1 1.00000(310) 0.99341( 60) 0.99736( 70) 0.99736(318) leu-comp-therm-016 case-2 1.00000(310) 0.99578( 60) 0.99824( 77) 0.99824(319) leu-comp-therm-016 case-3 1.00000(310) 0.99316( 60) 0.99548( 85) 0.99548(322) leu-comp-therm-016 case-4 1.00000(310) 0.99626( 80) 0.99824( 70) 0.99824(318) leu-comp-therm-016 case-5 continued on next page

21616/05.69456/P 39-54 benchmark JEFF-3.0 JEFF-3.1 C/E (JEFF-3.1) name 1.00000(310) 0.99319( 60) 0.99788( 73) 0.99788(319) leu-comp-therm-016 case-6 1.00000(310) 0.99297( 70) 0.99833( 71) 0.99833(318) leu-comp-therm-016 case-7 1.00000(310) 0.99439( 70) 0.99779( 62) 0.99779(316) leu-comp-therm-016 case-8 1.00000(310) 0.99513( 60) 0.99748( 68) 0.99748(317) leu-comp-therm-016 case-9 1.00000(310) 0.99430( 70) 0.99872( 78) 0.99872(320) leu-comp-therm-016 case-10 1.00000(310) 0.99537( 80) 0.99808( 63) 0.99808(316) leu-comp-therm-016 case-11 1.00000(310) 0.99430( 80) 0.99798( 66) 0.99798(317) leu-comp-therm-016 case-12 1.00000(310) 0.99597( 60) 0.99811( 73) 0.99811(319) leu-comp-therm-016 case-13 1.00000(310) 0.99477( 70) 0.99835( 65) 0.99835(317) leu-comp-therm-016 case-14 1.00000(310) 0.99422( 60) 0.99816( 64) 0.99816(317) leu-comp-therm-016 case-15 1.00000(310) 0.99213( 79) 0.99556( 72) 0.99556(318) leu-comp-therm-016 case-16 1.00000(310) 0.99217( 70) 0.99719( 71) 0.99719(318) leu-comp-therm-016 case-17 1.00000(310) 0.99413( 60) 0.99828( 76) 0.99828(319) leu-comp-therm-016 case-18 1.00000(310) 0.99711( 70) 0.99810( 68) 0.99810(317) leu-comp-therm-016 case-19 1.00000(310) 0.99603( 80) 0.99745( 73) 0.99745(319) leu-comp-therm-016 case-20 1.00000(310) 0.99546( 80) 0.99821( 75) 0.99821(319) leu-comp-therm-016 case-21 1.00000(310) 0.99613( 70) 0.99947( 72) 0.99947(318) leu-comp-therm-016 case-22 1.00000(310) 0.99525( 80) 0.99948( 78) 0.99948(320) leu-comp-therm-016 case-23 1.00000(310) 0.99336( 70) 0.99910( 71) 0.99910(318) leu-comp-therm-016 case-24 1.00000(310) 0.99550( 70) 0.99918( 71) 0.99918(318) leu-comp-therm-016 case-25 1.00000(310) 0.99662( 70) 0.99818( 75) 0.99818(319) leu-comp-therm-016 case-26 1.00000(310) 0.99524( 70) 0.99907( 69) 0.99907(318) leu-comp-therm-016 case-27 1.00000(310) 0.99462( 60) 0.99734( 64) 0.99734(317) leu-comp-therm-016 case-28 1.00000(310) 0.99359( 70) 0.99671( 66) 0.99671(317) leu-comp-therm-016 case-29 1.00000(310) 0.99514( 80) 0.99820( 56) 0.99820(315) leu-comp-therm-016 case-30 1.00000(310) 0.99308( 79) 0.99891( 64) 0.99891(317) leu-comp-therm-016 case-31 1.00000(310) 0.99458( 70) 0.99928( 68) 0.99928(317) leu-comp-therm-016 case-32 1.00000(310) 0.99834( 70) 1.00197( 65) 1.00197(317) leu-comp-therm-017 case-1 1.00000(310) 0.99653( 80) 1.00137( 68) 1.00137(317) leu-comp-therm-017 case-2 1.00000(310) 0.99612( 60) 0.99968( 81) 0.99968(320) leu-comp-therm-017 case-3 1.00000(310) 0.99589( 70) 0.99912( 63) 0.99912(316) leu-comp-therm-017 case-10 1.00000(310) 0.99638( 80) 0.99894( 73) 0.99894(318) leu-comp-therm-017 case-11 1.00000(310) 0.99447( 60) 0.99677( 69) 0.99677(318) leu-comp-therm-017 case-12 1.00000(310) 0.99546( 60) 0.99866( 68) 0.99866(317) leu-comp-therm-017 case-13 1.00000(310) 0.99574( 70) 0.99764( 74) 0.99764(319) leu-comp-therm-017 case-14 1.00000(280) 0.99257( 70) 0.99648( 64) 0.99648(287) leu-comp-therm-017 case-15 1.00000(280) 0.99365( 70) 0.99758( 67) 0.99758(288) leu-comp-therm-017 case-16 1.00000(280) 0.99459( 80) 1.00075( 61) 1.00075(287) leu-comp-therm-017 case-17 1.00000(280) 0.99453( 70) 0.99863( 69) 0.99863(288) leu-comp-therm-017 case-18 1.00000(280) 0.99488( 70) 0.99922( 83) 0.99922(292) leu-comp-therm-017 case-19 continued on next page

40-54 21616/05.69456/P benchmark JEFF-3.0 JEFF-3.1 C/E (JEFF-3.1) name 1.00000(280) 0.99398( 70) 1.00031( 83) 1.00031(292) leu-comp-therm-017 case-20 1.00000(280) 0.99351( 70) 0.99737( 80) 0.99737(291) leu-comp-therm-017 case-21 1.00000(280) 0.99135( 79) 0.99543( 71) 0.99543(289) leu-comp-therm-017 case-22 1.00000(280) 0.99655( 70) 1.00106( 79) 1.00106(291) leu-comp-therm-017 case-23 1.00000(280) 0.99629( 70) 1.00087( 65) 1.00087(287) leu-comp-therm-017 case-24 1.00000(280) 0.99303( 60) 0.99945( 77) 0.99945(290) leu-comp-therm-017 case-25 1.00000(630) 1.01172( 81) 1.01393( 72) 1.01393(634) leu-comp-therm-019 case-1 1.00000(580) 1.00698( 81) 1.01165( 85) 1.01165(586) leu-comp-therm-019 case-2 1.00000(610) 1.00493( 50) 1.00521( 66) 1.00521(614) leu-comp-therm-019 case-3 1.00000(140) 0.99391( 88) 0.99652( 40) 0.99652(146) leu-comp-therm-039 case-1 1.00000(140) 0.99382( 90) 0.99775( 48) 0.99775(148) leu-comp-therm-039 case-2 1.00000(140) 0.99307( 93) 0.99628( 35) 0.99628(144) leu-comp-therm-039 case-3 1.00000(140) 0.99367( 82) 0.99558( 37) 0.99558(145) leu-comp-therm-039 case-4 1.00000(140) 0.99301( 82) 0.99753( 46) 0.99753(147) leu-comp-therm-039 case-5 1.00000(140) 0.99429( 86) 0.99712( 46) 0.99712(147) leu-comp-therm-039 case-6 1.00000(140) 0.99166( 82) 0.99676( 44) 0.99676(147) leu-comp-therm-039 case-7 1.00000(140) 0.99225( 80) 0.99652( 40) 0.99652(146) leu-comp-therm-039 case-8 1.00000(140) 0.99320( 80) 0.99633( 41) 0.99633(146) leu-comp-therm-039 case-9 1.00000(140) 0.99353( 80) 0.99728( 43) 0.99728(146) leu-comp-therm-039 case-10 1.00000(140) 0.99178( 83) 0.99618( 41) 0.99618(146) leu-comp-therm-039 case-11 1.00000(140) 0.99285( 83) 0.99640( 47) 0.99640(148) leu-comp-therm-039 case-12 1.00000(140) 0.99384( 78) 0.99581( 39) 0.99581(145) leu-comp-therm-039 case-13 1.00000(140) 0.99254( 86) 0.99644( 42) 0.99644(146) leu-comp-therm-039 case-14 1.00000(140) 0.99102( 95) 0.99608( 42) 0.99608(146) leu-comp-therm-039 case-15 1.00000(140) 0.99294( 92) 0.99725( 42) 0.99725(146) leu-comp-therm-039 case-16 1.00000(140) 0.99350( 68) 0.99669( 39) 0.99669(145) leu-comp-therm-039 case-17 1.00100(200) 0.99174( 69) 0.99464( 84) 0.99365(217) leu-comp-therm-051 case-1 1.00100(240) 0.99800( 70) 0.99868( 74) 0.99768(251) leu-comp-therm-051 case-2 1.00100(240) 0.99697( 80) 1.00009( 69) 0.99909(249) leu-comp-therm-051 case-3 1.00100(240) 0.99686( 70) 0.99894( 78) 0.99794(252) leu-comp-therm-051 case-4 1.00100(240) 0.99666( 80) 0.99905( 78) 0.99805(252) leu-comp-therm-051 case-5 1.00100(240) 0.99518( 70) 1.00069( 70) 0.99969(250) leu-comp-therm-051 case-6 1.00100(240) 0.99434( 70) 1.00051( 69) 0.99951(249) leu-comp-therm-051 case-7 1.00100(240) 0.99724( 80) 0.99982( 72) 0.99882(250) leu-comp-therm-051 case-8 1.00100(190) 0.99232( 70) 0.99550( 79) 0.99451(206) leu-comp-therm-051 case-9 1.00100(190) 0.99561( 70) 0.99703( 79) 0.99603(206) leu-comp-therm-051 case-10 1.00100(190) 0.99327( 70) 0.99565( 71) 0.99466(203) leu-comp-therm-051 case-11 1.00100(190) 0.99040( 69) 0.99635( 74) 0.99535(204) leu-comp-therm-051 case-12 1.00100(220) 0.98828( 79) 0.99273( 72) 0.99174(231) leu-comp-therm-051 case-13 1.00100(190) 0.98779( 69) 0.99086( 77) 0.98987(205) leu-comp-therm-051 case-14 continued on next page

21616/05.69456/P 41-54 benchmark JEFF-3.0 JEFF-3.1 C/E (JEFF-3.1) name 1.00100(240) 0.99085( 69) 0.99346( 72) 0.99247(250) leu-comp-therm-051 case-15 1.00100(200) 0.99006( 79) 0.99348( 78) 0.99249(215) leu-comp-therm-051 case-16 1.00100(270) 0.99284( 70) 0.99546( 66) 0.99447(278) leu-comp-therm-051 case-17 1.00100(210) 0.98934( 69) 0.99314( 71) 0.99215(222) leu-comp-therm-051 case-18 1.00100(190) 0.98842( 69) 0.99202( 66) 0.99103(201) leu-comp-therm-051 case-19 0.99900(260) 1.00390( 59) 1.00490(267) leu-comp-therm-060 case-1 0.99770(260) 1.00612( 65) 1.00844(268) leu-comp-therm-060 case-2 1.00010(260) 1.00554( 73) 1.00544(270) leu-comp-therm-060 case-3 1.00170(260) 1.01071( 78) 1.00899(271) leu-comp-therm-060 case-4 1.00090(260) 1.00687( 64) 1.00596(267) leu-comp-therm-060 case-5 0.98940(270) 0.99784( 81) 1.00853(285) leu-comp-therm-060 case-6 1.00280(260) 1.00976( 61) 1.00694(266) leu-comp-therm-060 case-7 1.00390(260) 1.01494( 79) 1.01100(270) leu-comp-therm-060 case-8 1.00430(260) 1.01176( 61) 1.00743(266) leu-comp-therm-060 case-9 1.00140(260) 1.01105( 72) 1.00964(269) leu-comp-therm-060 case-10 1.00010(260) 1.00604( 68) 1.00594(269) leu-comp-therm-060 case-11 1.00090(260) 1.00557( 64) 1.00467(267) leu-comp-therm-060 case-12 1.00100(260) 1.00512( 76) 1.00412(271) leu-comp-therm-060 case-13 1.00150(260) 1.00715( 79) 1.00564(271) leu-comp-therm-060 case-14 1.00120(260) 1.00395( 68) 1.00275(268) leu-comp-therm-060 case-15 1.00070(260) 1.00272( 67) 1.00202(268) leu-comp-therm-060 case-16 1.00120(260) 1.00623( 70) 1.00502(269) leu-comp-therm-060 case-17 1.00060(260) 1.00911( 73) 1.00850(270) leu-comp-therm-060 case-18 1.00070(260) 1.00850( 71) 1.00779(269) leu-comp-therm-060 case-19 1.00390(260) 1.01382( 68) 1.00988(268) leu-comp-therm-060 case-20 1.00030(260) 1.00714( 65) 1.00684(268) leu-comp-therm-060 case-21 1.00360(260) 1.01122( 73) 1.00759(269) leu-comp-therm-060 case-22 1.00170(260) 1.00773( 58) 1.00602(266) leu-comp-therm-060 case-23 1.00250(260) 1.00906( 64) 1.00654(267) leu-comp-therm-060 case-24 1.00160(260) 1.00724( 62) 1.00563(267) leu-comp-therm-060 case-25 0.99960(260) 1.00949( 77) 1.00989(271) leu-comp-therm-060 case-26 1.00000(200) 0.99278( 17) 0.99906( 60) 0.99906(209) leu-comp-therm-trx case-1 3d 1.00000(200) 0.99233( 18) 0.99809( 52) 0.99809(207) leu-comp-therm-trx case-2 3d 0.99830(170) 0.99510( 50) 1.00007( 50) 1.00177(177) leu-comp-therm-dimple 0.99900(570) 0.99474( 60) 0.99892( 58) 0.99992(574) leu-met-therm-001 case-1 0.99910(290) 1.01120( 88) 1.01252( 86) 1.01343(302) leu-sol-therm-001 0.99970(390) 0.99595( 81) 0.99706( 65) 0.99736(396) leu-sol-therm-003 case-1 0.99930(420) 0.99477( 70) 0.99676( 69) 0.99746(426) leu-sol-therm-003 case-2 0.99950(420) 0.99905( 69) 1.00143( 76) 1.00193(427) leu-sol-therm-003 case-3 0.99950(420) 0.99510( 69) 0.99352( 75) 0.99402(427) leu-sol-therm-003 case-4 continued on next page

42-54 21616/05.69456/P benchmark JEFF-3.0 JEFF-3.1 C/E (JEFF-3.1) name 0.99970(480) 0.99730( 64) 0.99731( 54) 0.99761(483) leu-sol-therm-003 case-5 0.99990(490) 0.99798( 54) 0.99979( 63) 0.99989(494) leu-sol-therm-003 case-6 0.99940(490) 0.99540( 54) 0.99672( 53) 0.99732(493) leu-sol-therm-003 case-7 0.99930(520) 0.99839( 49) 1.00021( 47) 1.00091(522) leu-sol-therm-003 case-8 0.99960(520) 0.99677( 43) 0.99668( 45) 0.99708(522) leu-sol-therm-003 case-9 0.99940( 80) 1.00098( 68) 1.00046( 76) 1.00106(110) leu-sol-therm-004 case-001 0.99990( 90) 1.00005( 69) 1.00354( 65) 1.00364(111) leu-sol-therm-004 case-029 0.99990( 90) 0.99887( 58) 0.99961( 70) 0.99971(114) leu-sol-therm-004 case-033 0.99990(100) 1.00106( 59) 1.00157( 72) 1.00167(123) leu-sol-therm-004 case-034 0.99990(100) 1.00159( 67) 1.00111( 59) 1.00121(116) leu-sol-therm-004 case-046 0.99940(110) 1.00054( 55) 1.00015( 54) 1.00075(123) leu-sol-therm-004 case-051 0.99960(110) 0.99919( 65) 1.00078( 62) 1.00118(126) leu-sol-therm-004 case-054 0.99610( 90) 0.99384( 72) 0.99561( 66) 0.99951(112) leu-sol-therm-007 case-14 0.99730( 90) 0.99850( 63) 0.99732( 79) 1.00002(120) leu-sol-therm-007 case-30 0.99850(100) 0.99648( 64) 0.99551( 61) 0.99701(117) leu-sol-therm-007 case-32 0.99880(110) 0.99920( 56) 1.00109( 67) 1.00229(129) leu-sol-therm-007 case-36 0.99830(110) 0.99699( 66) 0.99678( 64) 0.99848(128) leu-sol-therm-007 case-49 0.99960(130) 1.00525( 84) 1.00697( 99) 1.00737(163) leu-sol-therm-016 case-105 0.99990(130) 1.00507( 83) 1.00514( 83) 1.00524(154) leu-sol-therm-016 case-113 0.99940(140) 1.00564( 68) 1.00497( 81) 1.00557(162) leu-sol-therm-016 case-125 0.99960(140) 1.00344( 70) 1.00593( 76) 1.00633(159) leu-sol-therm-016 case-129 0.99950(140) 1.00297( 71) 1.00489( 69) 1.00539(156) leu-sol-therm-016 case-131 0.99920(150) 1.00090( 66) 1.00184( 68) 1.00264(165) leu-sol-therm-016 case-140 0.99940(150) 1.00254( 68) 1.00477( 79) 1.00537(169) leu-sol-therm-016 case-196 0.99810(130) 1.00242( 90) 1.00345( 86) 1.00536(156) leu-sol-therm-017 case-104 0.99860(130) 1.00420( 78) 1.00244( 78) 1.00385(152) leu-sol-therm-017 case-122 0.99890(140) 1.00081( 82) 1.00208( 83) 1.00318(163) leu-sol-therm-017 case-123 0.99920(140) 1.00174( 73) 1.00227( 79) 1.00307(161) leu-sol-therm-017 case-126 0.99870(150) 1.00268( 82) 0.99996( 71) 1.00126(166) leu-sol-therm-017 case-130 0.99960(150) 0.99886( 68) 1.00207( 66) 1.00247(164) leu-sol-therm-017 case-147 0.99920(100) 1.00028( 70) 1.00129( 66) 1.00209(120) leu-sol-therm-018 case-1 0.99960(100) 1.00057( 70) 1.00368( 64) 1.00408(119) leu-sol-therm-018 case-2 0.99960(100) 1.00164( 80) 1.00153( 78) 1.00193(127) leu-sol-therm-018 case-3 0.99970(100) 1.00209( 70) 1.00385( 71) 1.00415(123) leu-sol-therm-018 case-4 0.99920(100) 1.00215( 80) 1.00278( 64) 1.00358(119) leu-sol-therm-018 case-5 0.99960(100) 1.00343( 70) 1.00051( 76) 1.00091(126) leu-sol-therm-018 case-6 0.99950(100) 0.99944( 66) 0.99885( 62) 0.99935(118) leu-sol-therm-020 case-216 0.99960(100) 0.99944( 56) 0.99977( 62) 1.00017(118) leu-sol-therm-020 case-217 0.99970(120) 0.99835( 59) 0.99676( 46) 0.99706(129) leu-sol-therm-020 case-220 0.99980(120) 0.99768( 52) 0.99945( 46) 0.99965(129) leu-sol-therm-020 case-226 continued on next page

21616/05.69456/P 43-54 benchmark JEFF-3.0 JEFF-3.1 C/E (JEFF-3.1) name 0.99830( 90) 0.99600( 64) 0.99765( 65) 0.99935(111) leu-sol-therm-021 case-215 0.99850(100) 0.99680( 66) 0.99909( 59) 1.00059(116) leu-sol-therm-021 case-218 0.99890(110) 0.99658( 58) 0.99686( 47) 0.99796(120) leu-sol-therm-021 case-221 0.99930(120) 0.99794( 54) 0.99906( 58) 0.99976(133) leu-sol-therm-021 case-223 Table 3.7 The results for LEU benchmarks with a thermal spectrum

3.4 PU results

3.4.1 Fast spectrum

1.04 JEFF-3.0 JEFF-3.1 benchmark uncertainty

1.03

1.02 eff

C/E for k 1.01

0.99 pmm01 pci01 pmi02 pmf02 pmf05 Popsy pmf08 pmf12 pmf13 Jezebel

Figure 3.8 Results for the PU benchmarks with a fast, intermediate, or mixed spectrum

benchmark JEFF-3.0 JEFF-3.1 C/E (JEFF-3.1) name 1.00000(200) 1.00052( 40) 1.00010( 19) 1.00010(201) pu-met-fast-001 bare-sphere 1.00000(200) 1.00428( 67) 1.00428(211) pu-met-fast-002 1.00000(130) 1.00417( 73) 1.00304( 70) 1.00304(148) pu-met-fast-005 1.00000(300) 0.99492( 44) 1.00267( 43) 1.00267(303) pu-met-fast-006 1.00000( 60) 1.00937( 45) 1.00305( 35) 1.00305( 69) pu-met-fast-008 case-1 1.00090(210) 0.99756( 71) 1.00407( 67) 1.00317(220) pu-met-fast-012 1.00340(230) 1.00545( 72) 1.00466( 80) 1.00126(243) pu-met-fast-013 Table 3.8 The results for PU benchmarks with a fast spectrum

3.4.2 Intermediate spectrum

44-54 21616/05.69456/P benchmark JEFF-3.0 JEFF-3.1 C/E (JEFF-3.1) name 1.00000(1100) 1.00608( 57) 1.00716( 44) 1.00716(1101) pu-comp-inter-001 0.98690(260) 1.02150( 47) 1.02532( 73) 1.03893(273) pu-met-inter-002 case-1 Table 3.9 The results for PU benchmarks with an intermediate spectrum

3.4.3 Thermal spectrum

1.03 JEFF-3.0 JEFF-3.1 benchmark uncertainty 1.025

1.02

1.015

eff 1.01

C/E for k 1.005

0.995

0.99 pst01 pst02 pst03 pst04 pst05 pst07 pst08 pst12 pst06

0.985

Figure 3.9 Results for the PU benchmarks with a thermal spectrum

benchmark JEFF-3.0 JEFF-3.1 C/E (JEFF-3.1) name 1.00000(500) 1.00373( 95) 1.00138( 95) 1.00138(509) pu-sol-therm-001 case-1 1.00000(500) 1.00599(108) 1.00423( 95) 1.00423(509) pu-sol-therm-001 case-2 1.00000(500) 1.00637(102) 1.00760( 87) 1.00760(507) pu-sol-therm-001 case-3 1.00000(500) 1.00102( 96) 1.00164(104) 1.00164(511) pu-sol-therm-001 case-4 1.00000(500) 1.00529( 94) 1.00484(106) 1.00484(511) pu-sol-therm-001 case-5 1.00000(500) 1.00534( 95) 1.00779(102) 1.00779(510) pu-sol-therm-001 case-6 1.00000(470) 1.00399( 87) 1.00167( 97) 1.00167(480) pu-sol-therm-002 case-1 1.00000(470) 1.00416( 91) 1.00352( 87) 1.00352(478) pu-sol-therm-002 case-2 1.00000(470) 1.00326( 90) 1.00274( 95) 1.00274(479) pu-sol-therm-002 case-3 1.00000(470) 1.00532(101) 1.00476( 86) 1.00476(478) pu-sol-therm-002 case-4 1.00000(470) 1.00765( 91) 1.00719( 92) 1.00719(479) pu-sol-therm-002 case-5 1.00000(470) 1.00182( 92) 1.00221( 93) 1.00221(479) pu-sol-therm-002 case-6 1.00000(470) 1.00409(106) 1.00762( 97) 1.00762(480) pu-sol-therm-002 case-7 1.00000(470) 1.00296( 86) 1.00237( 92) 1.00237(479) pu-sol-therm-003 case-1 continued on next page

21616/05.69456/P 45-54 benchmark JEFF-3.0 JEFF-3.1 C/E (JEFF-3.1) name 1.00000(470) 1.00101( 89) 1.00102( 93) 1.00102(479) pu-sol-therm-003 case-2 1.00000(470) 1.00587( 80) 1.00217( 91) 1.00217(479) pu-sol-therm-003 case-3 1.00000(470) 1.00500( 85) 1.00378( 88) 1.00378(478) pu-sol-therm-003 case-4 1.00000(470) 1.00503( 90) 1.00519(103) 1.00519(481) pu-sol-therm-003 case-5 1.00000(470) 1.00522( 89) 1.00374( 96) 1.00374(480) pu-sol-therm-003 case-6 1.00000(470) 1.00695( 91) 1.00421( 87) 1.00421(478) pu-sol-therm-003 case-7 1.00000(470) 1.00558( 93) 1.00648( 87) 1.00648(478) pu-sol-therm-003 case-8 1.00000(470) 1.00328( 89) 1.00349( 74) 1.00349(476) pu-sol-therm-004 case-1 1.00000(470) 0.99903( 82) 0.99871( 86) 0.99871(478) pu-sol-therm-004 case-2 1.00000(470) 1.00220( 85) 1.00123( 70) 1.00123(475) pu-sol-therm-004 case-3 1.00000(470) 0.99791( 75) 0.99778( 87) 0.99778(478) pu-sol-therm-004 case-4 1.00000(470) 0.99898( 82) 0.99908( 88) 0.99908(478) pu-sol-therm-004 case-5 1.00000(470) 1.00175( 79) 1.00202( 94) 1.00202(479) pu-sol-therm-004 case-6 1.00000(470) 1.00539( 87) 1.00775( 84) 1.00775(477) pu-sol-therm-004 case-7 1.00000(470) 1.00101( 71) 1.00200( 75) 1.00200(476) pu-sol-therm-004 case-8 1.00000(470) 1.00094( 85) 0.99993( 90) 0.99993(479) pu-sol-therm-004 case-9 1.00000(470) 1.00164( 90) 1.00113( 86) 1.00113(478) pu-sol-therm-004 case-10 1.00000(470) 1.00052( 90) 0.99947( 91) 0.99947(479) pu-sol-therm-004 case-11 1.00000(470) 1.00331( 80) 1.00436( 78) 1.00436(476) pu-sol-therm-004 case-12 1.00000(470) 0.99914( 84) 1.00093( 77) 1.00093(476) pu-sol-therm-004 case-13 1.00000(470) 1.00128( 79) 1.00219( 75) 1.00219(476) pu-sol-therm-005 case-1 1.00000(470) 1.00182( 83) 1.00158( 81) 1.00158(477) pu-sol-therm-005 case-2 1.00000(470) 1.00349( 84) 1.00230( 99) 1.00230(480) pu-sol-therm-005 case-3 1.00000(470) 1.00242( 89) 1.00238( 93) 1.00238(479) pu-sol-therm-005 case-4 1.00000(470) 1.00548( 85) 1.00428( 89) 1.00428(478) pu-sol-therm-005 case-5 1.00000(470) 1.00454(101) 1.00327( 85) 1.00327(478) pu-sol-therm-005 case-6 1.00000(470) 1.00292( 86) 1.00198( 69) 1.00198(475) pu-sol-therm-005 case-7 1.00000(470) 0.99993( 84) 0.99808( 84) 0.99808(477) pu-sol-therm-005 case-8 1.00000(470) 1.00164( 89) 1.00313( 92) 1.00313(479) pu-sol-therm-005 case-9 1.00000(350) 0.99971( 89) 1.00058( 81) 1.00058(359) pu-sol-therm-006 case-1 1.00000(350) 1.00142( 84) 1.00076( 86) 1.00076(360) pu-sol-therm-006 case-2 1.00000(350) 1.00278( 74) 1.00319( 75) 1.00319(358) pu-sol-therm-006 case-3 1.00000(470) 1.00513( 92) 1.00526( 95) 1.00526(479) pu-sol-therm-007 case-2 1.00000(470) 1.00100( 97) 1.00114( 90) 1.00114(479) pu-sol-therm-007 case-3 1.00000(470) 1.00658( 99) 1.00776(103) 1.00776(481) pu-sol-therm-007 case-5 1.00000(470) 1.00107(104) 0.99940( 91) 0.99940(479) pu-sol-therm-007 case-6 1.00000(470) 1.00138( 90) 1.00374( 85) 1.00374(478) pu-sol-therm-007 case-7 1.00000(470) 0.99700(101) 0.99521( 94) 0.99521(479) pu-sol-therm-007 case-8 1.00000(470) 0.99683( 99) 0.99686( 93) 0.99686(479) pu-sol-therm-007 case-9 1.00000(470) 0.99778(107) 0.99803( 89) 0.99803(478) pu-sol-therm-007 case-10 continued on next page

46-54 21616/05.69456/P benchmark JEFF-3.0 JEFF-3.1 C/E (JEFF-3.1) name 1.00000(330) 1.00335( 91) 1.00211( 95) 1.00211(343) pu-sol-therm-008 case-1 1.00000(400) 1.00507( 86) 1.00292( 96) 1.00292(411) pu-sol-therm-008 case-2 1.00000(400) 1.01181(100) 1.01212( 88) 1.01212(409) pu-sol-therm-008 case-3 1.00000(400) 1.01577( 85) 1.01519(103) 1.01519(413) pu-sol-therm-008 case-4 1.00000(280) 1.01685( 82) 1.01849( 81) 1.01849(291) pu-sol-therm-008 case-5 1.00000(280) 1.01131( 80) 1.00925( 90) 1.00925(294) pu-sol-therm-008 case-7 1.00000(400) 0.99962( 86) 0.99549( 89) 0.99549(410) pu-sol-therm-008 case-8 1.00000(610) 1.00432(102) 1.00664(106) 1.00664(619) pu-sol-therm-008 case-9 1.00000(370) 1.02215( 90) 1.02138( 85) 1.02138(379) pu-sol-therm-008 case-10 1.00000(310) 1.01585( 83) 1.01651( 83) 1.01651(321) pu-sol-therm-008 case-11 1.00000(410) 1.01581(101) 1.01306(111) 1.01306(424) pu-sol-therm-008 case-12 1.00000(410) 1.02492( 98) 1.02222( 90) 1.02222(419) pu-sol-therm-008 case-13 1.00000(420) 1.00417( 93) 1.00590(100) 1.00590(432) pu-sol-therm-008 case-14 1.00000(410) 1.00996( 87) 1.00758(108) 1.00758(424) pu-sol-therm-008 case-15 1.00000(330) 1.00937( 97) 1.00791( 72) 1.00791(338) pu-sol-therm-008 case-16 1.00000(400) 1.01076( 81) 1.01187( 81) 1.01187(408) pu-sol-therm-008 case-17 1.00000(400) 1.01546( 95) 1.01525( 99) 1.01525(412) pu-sol-therm-008 case-18 1.00000(280) 1.01000( 94) 1.01023( 92) 1.01023(294) pu-sol-therm-008 case-19 1.00000(280) 1.01058( 92) 1.01005( 82) 1.01005(292) pu-sol-therm-008 case-20 1.00000(400) 0.99615( 93) 0.99649( 91) 0.99649(410) pu-sol-therm-008 case-21 1.00000(610) 0.99831( 97) 0.99535(113) 0.99535(620) pu-sol-therm-008 case-22 1.00000(370) 1.01573( 81) 1.01638( 87) 1.01638(380) pu-sol-therm-008 case-23 1.00000(310) 1.00799( 90) 1.00852( 88) 1.00852(322) pu-sol-therm-008 case-24 1.00000(420) 1.01664(100) 1.01388(105) 1.01388(433) pu-sol-therm-008 case-25 1.00000(410) 1.01091(102) 1.01140(104) 1.01140(423) pu-sol-therm-008 case-26 1.00000(420) 1.01116( 95) 1.01192(114) 1.01192(435) pu-sol-therm-008 case-27 1.00000(420) 1.00406(105) 1.00540( 90) 1.00540(429) pu-sol-therm-008 case-28 1.00000(410) 1.01321(110) 1.01387( 71) 1.01387(416) pu-sol-therm-008 case-29 1.00000(410) 1.01335( 86) 1.01085( 97) 1.01085(421) pu-sol-therm-008 case-30 1.00000(430) 1.00521( 63) 1.00632( 50) 1.00632(433) pu-sol-therm-012 case-2 1.00000(580) 1.00809( 60) 1.00738( 56) 1.00738(583) pu-sol-therm-012 case-3 1.00000(580) 1.00611( 52) 1.00863( 51) 1.00863(582) pu-sol-therm-012 case-4 1.00000(580) 1.00863( 49) 1.00944( 54) 1.00944(582) pu-sol-therm-012 case-5 1.00000( 70) 1.00622( 97) 1.00854(100) 1.00854(121) pu-sol-therm-012 case-6 1.00000(130) 1.00560( 94) 1.00806(100) 1.00806(164) pu-sol-therm-012 case-7 1.00000(130) 1.00689( 89) 1.00454( 97) 1.00454(162) pu-sol-therm-012 case-8 1.00000(430) 1.01228( 87) 1.01032( 83) 1.01032(438) pu-sol-therm-012 case-9 1.00000(430) 1.00579( 78) 1.00561( 78) 1.00561(437) pu-sol-therm-012 case-10 1.00000(430) 1.00818( 74) 1.00786( 72) 1.00786(436) pu-sol-therm-012 case-11 1.00000(430) 1.00765( 64) 1.00840( 65) 1.00840(435) pu-sol-therm-012 case-12 continued on next page

21616/05.69456/P 47-54 benchmark JEFF-3.0 JEFF-3.1 C/E (JEFF-3.1) name 1.00000(580) 1.00876( 48) 1.01028( 54) 1.01028(582) pu-sol-therm-012 case-13 1.00000(130) 1.00200(100) 1.00280( 85) 1.00280(155) pu-sol-therm-012 case-14 1.00000(430) 1.00922( 81) 1.01001( 86) 1.01001(438) pu-sol-therm-012 case-15 1.00000(430) 1.00353( 70) 1.00601( 75) 1.00601(436) pu-sol-therm-012 case-16 1.00000(430) 1.00769( 88) 1.00744( 76) 1.00744(437) pu-sol-therm-012 case-17 1.00000(430) 1.00666( 69) 1.00750( 60) 1.00750(434) pu-sol-therm-012 case-18 1.00000(430) 1.00766( 59) 1.00945( 68) 1.00945(435) pu-sol-therm-012 case-19 1.00000(580) 1.00739( 55) 1.00874( 51) 1.00874(582) pu-sol-therm-012 case-20 1.00000(580) 1.00832( 56) 1.00877( 52) 1.00877(582) pu-sol-therm-012 case-21 1.00000(580) 1.00956( 49) 1.00947( 41) 1.00947(581) pu-sol-therm-012 case-22 1.00000(580) 1.00841( 49) 1.00787( 44) 1.00787(582) pu-sol-therm-012 case-23 Table 3.10 The results for PU benchmarks with a thermal spectrum

3.4.4 Mixed spectrum benchmark JEFF-3.0 JEFF-3.1 C/E (JEFF-3.1) name 1.00020(370) 1.00441( 99) 1.00421(383) pu-met-mixed-001 case-1 1.00020(320) 0.99839( 83) 0.99819(331) pu-met-mixed-001 case-2 1.00050(250) 1.00406( 98) 1.00356(268) pu-met-mixed-001 case-3 1.00000(250) 1.00720( 91) 1.00720(266) pu-met-mixed-001 case-4 1.00010(250) 1.00823( 83) 1.00813(263) pu-met-mixed-001 case-5 1.00030(250) 1.00761( 87) 1.00731(264) pu-met-mixed-001 case-6

3.5 MIX results

3.5.1 Fast spectrum benchmark JEFF-3.0 JEFF-3.1 C/E (JEFF-3.1) name 0.98660(230) 0.98388( 49) 0.98997( 47) 1.00342(238) mix-comp-fast-001 0.98970(230) 0.98783( 48) 0.99002( 70) 1.00032(243) mix-met-fast-011 case-1 0.99980(230) 0.99274( 44) 0.99539( 46) 0.99559(235) mix-met-fast-011 case-2 1.00180(240) 0.99749( 49) 0.99852( 47) 0.99673(244) mix-met-fast-011 case-3 1.00120(240) 0.99842( 53) 1.00128( 42) 1.00008(243) mix-met-fast-011 case-4 Table 3.11 The results for MIX benchmarks with a fast spectrum

3.5.2 Thermal spectrum benchmark JEFF-3.0 JEFF-3.1 C/E (JEFF-3.1) name 1.00520(670) 0.97599( 51) 0.97597( 53) 0.97092(669) mix-comp-therm-012 case-1 1.00520(670) 0.97668( 58) 0.97812( 60) 0.97306(669) mix-comp-therm-012 case-2 1.00520(670) 0.97475( 65) 0.97389( 52) 0.96885(669) mix-comp-therm-012 case-3 1.00520(670) 0.97976( 50) 0.97874( 51) 0.97368(669) mix-comp-therm-012 case-4 1.00520(670) 0.97589( 53) 0.97640( 62) 0.97135(670) mix-comp-therm-012 case-5 1.00520(750) 0.98124( 59) 0.98007( 60) 0.97500(749) mix-comp-therm-012 case-6 continued on next page

48-54 21616/05.69456/P 1.02 JEFF-3.0 JEFF-3.1 benchmark uncertainty 1.015

1.01

1.005

eff 1

C/E for k 0.995

0.99

0.985

0.98 mcf01 ZPR-21

0.975

Figure 3.10 Results for the MIX benchmarks with a fast spectrum

1.04 JEFF-3.0 JEFF-3.1 benchmark uncertainty

1.03

1.02

1.01 eff

1 C/E for k

0.99

0.98 mct12 Kritz mst02

0.97

Figure 3.11 Results for the MIX benchmarks with a thermal spectrum

benchmark JEFF-3.0 JEFF-3.1 C/E (JEFF-3.1) name 1.00530(1340) 1.03196( 51) 1.03261( 53) 1.02717(1334) mix-comp-therm-012 case-7 continued on next page

21616/05.69456/P 49-54 benchmark JEFF-3.0 JEFF-3.1 C/E (JEFF-3.1) name 1.00530(550) 1.02686( 56) 1.02685( 49) 1.02144(549) mix-comp-therm-012 case-8 1.00530(550) 1.02358( 55) 1.02491( 46) 1.01951(549) mix-comp-therm-012 case-9 1.00530(550) 1.02479( 54) 1.02482( 58) 1.01942(550) mix-comp-therm-012 case-10 1.00530(550) 1.02224( 55) 1.02227( 51) 1.01688(549) mix-comp-therm-012 case-11 1.00530(550) 1.02506( 50) 1.02574( 56) 1.02033(550) mix-comp-therm-012 case-12 1.00530(1580) 1.03273( 55) 1.03410( 52) 1.02865(1572) mix-comp-therm-012 case-13 1.00120(860) 1.01907( 52) 1.02087( 67) 1.01965(861) mix-comp-therm-012 case-14 1.00120(860) 1.01942( 62) 1.02087( 65) 1.01965(861) mix-comp-therm-012 case-15 1.00120(860) 1.01687( 56) 1.01779( 53) 1.01657(861) mix-comp-therm-012 case-16 1.00120(860) 1.01591( 63) 1.01628( 57) 1.01506(861) mix-comp-therm-012 case-17 1.00120(860) 1.01467( 62) 1.01575( 58) 1.01453(861) mix-comp-therm-012 case-18 1.00120(860) 1.01467( 57) 1.01524( 56) 1.01402(861) mix-comp-therm-012 case-19 1.00140(1140) 1.01567( 55) 1.01664( 70) 1.01522(1140) mix-comp-therm-012 case-20 1.00140(1200) 1.01466( 71) 1.01517( 68) 1.01375(1200) mix-comp-therm-012 case-21 1.00140(1520) 1.01281( 63) 1.01292( 67) 1.01150(1519) mix-comp-therm-012 case-22 1.00070(1140) 1.00969( 56) 1.00890( 64) 1.00819(1141) mix-comp-therm-012 case-23 1.00070(1140) 1.01113( 68) 1.01082( 64) 1.01011(1141) mix-comp-therm-012 case-24 1.00070(1140) 1.00849( 64) 1.00973( 61) 1.00902(1141) mix-comp-therm-012 case-25 1.00070(1140) 1.00724( 57) 1.00795( 55) 1.00724(1141) mix-comp-therm-012 case-26 1.00070(1140) 1.00754( 64) 1.00784( 48) 1.00714(1140) mix-comp-therm-012 case-27 1.00070(1140) 1.01006( 66) 1.00978( 60) 1.00907(1141) mix-comp-therm-012 case-28 1.00070(1140) 1.00690( 68) 1.00958( 64) 1.00887(1141) mix-comp-therm-012 case-29 1.00070(1140) 1.00738( 62) 1.00773( 52) 1.00703(1140) mix-comp-therm-012 case-30 1.00170(1510) 0.99723( 78) 0.99655( 68) 0.99486(1509) mix-comp-therm-012 case-31 1.00170(1510) 0.99643( 61) 0.99804( 64) 0.99635(1509) mix-comp-therm-012 case-32 1.00170(1510) 0.99421( 73) 0.99452( 66) 0.99283(1509) mix-comp-therm-012 case-33 1.00000( 0) 0.99718( 22) 1.00105( 22) 1.00105( 22) mix-comp-therm-kritz core-2.19cold 1.00000( 0) 0.99713( 7) 1.00129( 22) 1.00129( 22) mix-comp-therm-kritz core-2.19hot 1.00000(240) 1.00209( 58) 1.00003( 47) 1.00003(245) mix-sol-therm-002 exp-58 1.00000(240) 1.00167( 61) 1.00092( 53) 1.00092(246) mix-sol-therm-002 exp-59 1.00000(240) 1.00026( 45) 1.00038( 49) 1.00038(245) mix-sol-therm-002 exp-61 Table 3.12 The results for MIX benchmarks with a thermal spectrum. † The benchmark value and its uncertainty is still under investigation for Kritz.

3.6 U233 results

3.6.1 Fast spectrum benchmark JEFF-3.0 JEFF-3.1 C/E (JEFF-3.1) name 1.00000(100) 1.01359( 38) 1.00445( 37) 1.00445(107) u233-met-fast-001 1.00000( 90) 1.00895( 41) 1.00099( 36) 1.00099( 97) u233-met-fast-005 case-1 1.00000( 60) 1.00710( 49) 1.00033( 48) 1.00033( 77) u233-met-fast-005 case-2 1.00000(140) 1.00391( 53) 1.00608( 44) 1.00608(147) u233-met-fast-006 Table 3.13 The results for U233 benchmarks with a fast spectrum

50-54 21616/05.69456/P 1.02 JEFF-3.0 JEFF-3.1 benchmark uncertainty 1.015

1.01

1.005

eff 1

C/E for k 0.995

0.99

0.985

0.98 Skidoo u3mf05 u3mf06

0.975

Figure 3.12 Results for the U233 benchmarks with a fast spectrum

3.6.2 Thermal spectrum

1.02 JEFF-3.0 JEFF-3.1 benchmark uncertainty 1.015

1.01

1.005

eff 1

C/E for k 0.995

0.99

0.985

0.98 u3ct01 u3st01

0.975

Figure 3.13 Results for the U233 benchmarks with a thermal spectrum

21616/05.69456/P 51-54 benchmark JEFF-3.0 JEFF-3.1 C/E (JEFF-3.1) name 1.00060(270) 0.99980(123) 0.99619( 98) 0.99559(287) u233-comp-therm-001 case-1 1.00150(250) 1.00335(106) 0.99884( 99) 0.99734(269) u233-comp-therm-001 case-2 1.00000(240) 0.99848(112) 0.99761(112) 0.99761(265) u233-comp-therm-001 case-3 1.00070(250) 0.99920( 85) 0.99569( 80) 0.99499(262) u233-comp-therm-001 case-4 1.00150(260) 1.00114( 76) 0.99538( 89) 0.99389(275) u233-comp-therm-001 case-5 1.00150(280) 0.99731(102) 0.99582(298) u233-comp-therm-001 case-6 0.99950(270) 1.00811(112) 0.99892(102) 0.99942(289) u233-comp-therm-001 case-7 1.00040(280) 1.00744(112) 0.99532( 90) 0.99492(294) u233-comp-therm-001 case-8 1.00000(310) 1.00205( 61) 0.99771( 53) 0.99771(315) u233-sol-therm-001 case-1 1.00050(330) 1.00084( 62) 0.99927( 80) 0.99877(339) u233-sol-therm-001 case-2 1.00060(330) 1.00039( 63) 0.99629( 58) 0.99569(335) u233-sol-therm-001 case-3 0.99980(330) 0.99987( 59) 0.99786( 69) 0.99806(337) u233-sol-therm-001 case-4 0.99990(330) 0.99972( 62) 0.99718( 60) 0.99728(335) u233-sol-therm-001 case-5 Table 3.14 The results for U233 benchmarks with a thermal spectrum

52-54 21616/05.69456/P 3.7 Summary

In the previous subsections, results for many benchmarks were presented. The number of benchmarks in each of the ICSBEP main categories is summarized in Table 3.15.

COMP MET SOL total therm / inter / fast / mixed therm / inter / fast / mixed therm LEU 257 / / / 1 / / / 49 307 IEU 9 / 4 / / / / 16 / 29 HEU / 6 / / 42 / 5 / 66 / 5 87 211 MIX 35 / / 1 / 1 / / 4 / 3 44 PU / 1 / / / 1 / 7 / 6 105 120 U233 8 / / / / / 4 / 5 17 total 309 / 11 / 1 / 1 43 / 6 / 97 / 11 249 728 Table 3.15 The number of benchmarks per main ICSBEP category for thermal/intermediate/fast neutron spectrum.

For a speciﬁc situation, it can be useful to compute averages of those benchmark cases that are most applicable to that situation. Here, however, we only give average values for C=E 1 for the main ICSBEP categories, see Table 3.16. Closer inspection of individual benchmarks wil reveal deviations from the average. For various of these cases it may well be that the nuclear data library can, at a later date, still be improved further to yield better results. A detailed discussion of these problems, and the possibilities to solve them, is beyond the scope of the present paper.

COMP MET SOL therm / inter / fast / mixed therm / inter / fast / mixed therm LEU 98 / / / 7 / / / 145 IEU 359 / 376 / / / / 197 / HEU / 1915 / / 217 308 / 60 / 116 / 528 66 MIX 512 / / 336 / / / 182 / 44 PU / 716 / / / 3842 / 251 / 476 568 U233 380 / / / / / 296 / 249 Table 3.16 The average values for C=E 1 (in pcm) per main ICSBEP benchmark category.

21616/05.69456/P 53-54 References

[5] J. Hardy, D. Klein and J. J. Volpe, A study of physics parameters in several H2O-moderated lattices of slightly enriched and natural uranium, Nuc. Sci. Eng. 40 (1970) 101; Cross section evaluation working group (CSEWG), Benchmark Speciﬁcations, ENDF-202 (1986) [6] I. Remec, J. C. Gehin, and R. J. Ellis, KRITZ-2:19 experiment on regular H2O/fuel pin lattices with mixed oxide fuel at temperatures up to 245ŽC, A Study (Draft) by the OECD/NEA Working Party on Scientiﬁc Issues of Reactor Systems (WPRS), To be published as part of the International Reactor Physics Benchmark Project. [7] S.C. van der Marck, A. Hogenbirk, and R. Klein Meulekamp, New temperature interpolation in MCNP, Proc. M&C-2005, Avignon (France), 2005

54-54 21616/05.69456/P APPENDIX 3 Below follows a list of definitions to the ENDF-6 file format for JEFF-3.1. The different file numbers (MFs) listed in Table A3.1 followed by the reaction types (MTs) in Table A3.2. For a complete description of the ENDF-6 format, see reference V. McLane, ENDF-102, Data Formats and Procedures for the Evaluated Nuclear Data File ENDF-6, BNL-NCS-44945-01/04-Rev., April 2001.

Table A3.1. Definitions of file types (MF) MF Description 01 General information 02 Resonance parameter data 03 Reaction cross-sections 04 Angular distributions for emitted particles 05 Energy distributions for emitted particles 06 Energy-angle distributions for emitted particles 07 Thermal neutron scattering law data 08 Radioactivity and fission product yield data 09 Multiplicities for radioactive nuclide production 10 Cross-sections for radioactive nuclide production 12 Multiplicities for photon production 13 Cross-sections for photon production 14 Angular distributions for photon production 15 Energy distributions for photon production 23 Photo-atomic interaction cross-sections 27 Atomic form factors or scattering functions for photo-atomic interactions 30 Data covariances obtained from parameter covariances and sensitivities 31 Data covariances for nubar 32 Data covariances for resonance parameters 33 Data covariances for reaction cross-sections 34 Data covariances for angular distributions 35 Data covariances for energy distributions 39 Data covariances for radionuclide production yields 40 Data covariances for radionuclide production cross-sections Table A3.2. Definitions of reaction types (MT) MT reaction Description Comments 1 (n,total) Neutron total cross-sections. Sum of Redundant. Undefined for incident MT=2, 4, 5, 11, 16-18, 22-26, 28-37, charged particles. 41-42, 44-45, 102-117.

2 (z,z0) Elastic scattering cross-section for incident particles. 3 (z,nonelastic) Nonelastic neutron cross-section. Redundant. For photon production Sum of MT=4, 5, 11, 16-18, 22-26, only. 28-37, 41-42, 44-45, 102-117. 4 (z,n) Production of one neutron in the exit Redundant. For incident neutrons, channel. Sum of the MT=50-91. this is inelastic scattering (MT=50 is undefined). 5 (z,anything) Sum of all reactions not given Each particle can be identified and its explicitly in another MT number. multiplicity given in File 6. Not This is a partial reaction to be added allowed in Files 4, 5. to obtain MT=1. 6-9 Not allowed in version 6. 9Be(n,2n) in version 5. 10 (z,continuum) Total continuum reaction; includes Redundant; to be used for derived all continuum reactions and excludes files only. all discrete reactions. 11 (z,2nd) Production of two neutrons and a deuteron, plus a residual. 12-15 Unassigned. 16 (z,2n) Production of two neutrons and a residual1. Sum of MT=875-891, if they are present. 17 (z,3n) 18 (z,fission) 19 (n,f) 20 (n,nf) Second-chance fission2. 21 (n,2nf) Third-chance fission2. 22 (z,nα) Production of a neutron and an alpha particle, plus a residual. 23 (n,n3α) Production of a neutron and three alpha particles, plus a residual.

1 The “residual” is the remainder after the reaction specified by MT has taken place (for example, A-1 after an n,2n reaction on target A). This “residual” may break up further if LR > 0. 2 Note that the partial fission cross-sections are not defined for incident charged particles. Table A3.2. Definitions of reaction types (MT) (cont.)

MT reaction Description Comments 24 (z,2nα) Production of two neutrons and an alpha particle, plus a residual. 25 (z,3nα) Production of three neutrons and an alpha particle, plus a residual. 26 Not allowed in version 6. Version 5: (n,2n) isomeric state; used in File 8 and 6, 9, or 10. 27 (n,abs) Absorption; sum of MT=18 and Rarely used. MT=102 through MT=117. 28 (z,np) Production of a neutron and a proton, plus a residual. 29 (z,n2α) Production of a neutron and two alpha particles, plus a residual. 30 (z,2n2α) Production of two neutrons and two alpha particles, plus a residual. 31 Not allowed for version 6. Used only as an LR flag. 32 (z,nd) Production of a neutron and a deuteron, plus a residual. 33 (z,nt) Production of a neutron and a triton, plus a residual. 34 (z,n3He) Production of a neutron and a 3He particle, plus a residual. 35 (z,nd2α) Production of a neutron, a deuteron and 2 alpha particles, plus a residual. 36 (z,nt2α) Production of a neutron, a triton and 2 alpha particles, plus a residual. 37 (z,4n) Production of 4 neutrons, plus a residual. 38 (n,3nf) Fourth-chance fission cross-section2. 39 Not allowed for version 6. Used only as an LR flag. 40 Not allowed for version 6. Used only as an LR flag. 41 (z,2np) Production of 2 neutrons and a proton, plus a residual. 42 (z,3np) Production of 3 neutrons and a proton, plus a residual. 43 (Unassigned) 44 (z,n2p) Production of a neutron and 2 protons, plus a residual. 45 (z,npα) Production of a neutron, a proton and an alpha particle, plus a residual. 46-49 Not allowed in Version 6. Version 5: description of 2nd neutron from 9Be(n,2n) reactions to excited states. 50 (y,n0) Production of a neutron, leaving the Not allowed for incident neutrons; residual nucleus in the ground state. use MT=2. 51 (z,n1) Production of a neutron, with residual in the 1st excited state Table A3.2. Definitions of reaction types (MT) (cont.)

MT reaction Description Comments

52 (z,n2) Production of a neutron, with residual in the 2nd excited state. … … 90 (z,n40) Production of a neutron, with residual in the 40th excited state. 91 (z,nc) Production of a neutron in the continuum not included in the above discrete representation. 92-100 (Unassigned) 101 (n,disap) Neutron disappearance; equal to sum Rarely used. of MT=102-117. 102 (z,γ) Radiative capture. 103 (z,p) Production of a proton, plus a For incident protons, this is inelastic residual. Sum of MT=600-649, if scattering (MT=600 is undefined). they are present. 104 (z,d) Production of a deuteron, plus a For incident deuterons, this is residual. Sum of MT=650-699, if inelastic scattering (MT=650 is they are present. undefined). 105 (z,t) Production of a triton, plus a residual. For incident tritons, this is inelastic Sum of MT=700-749, if they are scattering (MT=700 is undefined). present. 106 (z,3He) Production of a 3He particle plus a For incident 3He particles, this is residual. Sum of MT=750-799, if inelastic scattering (MT=750 is they are present. undefined). 107 (z,α) Production of an alpha particle, plus For incident alpha particles, this is a residual. Sum of MT=800-849, if inelastic scattering (MT=800 is they are present. undefined). 108 (z,2α) Production of 2 alpha particles, plus a residual. 109 (z,3α) Production of 3 alpha particles, plus a residual. 110 (Unassigned) 111 (z,2p) Production of 2 protons, plus a residual. 112 (z,pα) Production of a proton and an alpha particle, plus a residual. 113 (z,t2α) Production of a triton and 2 alpha particles, plus a residual. 114 (z,d2α) Production of a deuteron and 2 alpha particles, plus a residual. 115 (z,pd) Production of proton and a deuteron, plus a residual. 116 (z,pt) Production of proton and a triton, plus a residual. Table A3.2. Definitions of reaction types (MT) (cont.)

MT reaction Description Comments 117 (z,dα) Production of deuteron and an alpha particle, plus a residual. 118-119 (Unassigned) 120 Not allowed for version 6. Version 5: target destruction – nonelastic minus total (n,n′γ) 121-150 (Unassigned) 151 (n,RES) Resonance parameters that can be Incident neutrons only. used to calculate cross-sections at different temperatures in the resolved and unresolved energy regions. 152-200 (Unassigned) 201 (z,Xn) Total neutron production. Redundant; use in derived files only. 202 (z,Xγ) Total gamma production. Redundant; use in derived files only. 203 (z,Xp) Total proton production. Redundant; use in derived files only. 204 (z,Xd) Total deuteron production. Redundant; use in derived files only. 205 (z,Xt) Total triton production. Redundant; use in derived files only. 206 (z,X3He) Total 3He production. Redundant; use in derived files only. 207 (z,Xα) Total alpha particle production. Redundant; use in derived files only. 208 (z,Xπ+) Total π+ production. For use in high-energy evaluations. 209 (z,Xπ0) Total π0 production. For use in high-energy evaluations. 210 (z,Xπ–) Total π– production. For use in high-energy evaluations. 211 (z,Xµ+) Total µ+ production. For use in high-energy evaluations. 212 (z,Xµ–) Total µ– production. For use in high-energy evaluations. 213 (z,Xκ+) Total κ+ production. For use in high-energy evaluations. 0 0 214 (z,Xκ (long)) Total κ (long) production. For use in high-energy evaluations. 0 0 215 (z,Xκ (short)) Total κ (short) production. For use in high-energy evaluations. 216 (z,Xκ–) Total κ– production. For use in high-energy evaluations. 217 (z,Xp) Total anti-proton production. For use in high-energy evaluations. 218 (z,Xn) Total anti-neutron production. For use in high-energy evaluations. 219-250 (Unassigned)

251 (n,…) µ L , average cosine of the scattering Derived files only. angle (laboratory system) for elastic scattering of neutrons. 252 (n,…) ξ, average logarithmic energy Derived files only. decrement for elastic scattering of neutrons. 253 (n,…) γ, average of the square of the Derived files only. logarithmic energy decrement divided by twice the average logarithmic energy decrement, for elastic scattering of neutrons. 254-300 (Unassigned) Table A3.2. Definitions of reaction types (MT) (cont.)

MT reaction Description Comments 301-450 (z,…) Energy release parameters, E , σ , Derived files only. for total and partial cross-sections; MT= 300 plus the reaction MT number, e.g. MT=302 is the elastic scattering kerma. 451 (z,...) Heading or title information; given in File 1 only.

452 (z,...) νT , average total (prompt plus delayed) number of neutrons released per fission event. 453 (Unassigned) 454 (z,...) Independent fission product yield data.

455 (z,...) ν d , average number of delayed neutrons released per fission event.

456 (z,...) ν p , average number of prompt neutrons released per fission event. 457 (z,...) Radioactive decay data. 458 (n,...) Energy release in fission for incident neutrons. 459 (z,...) Cumulative fission product yield data. 460-464 (Unassigned) 465-466 Not allowed in version 6. Version 5: delayed and prompt neutrons from spontaneous fission. 467-499 (Unassigned) 500 Total charged-particle stopping power. 501 Total photon interaction. 502 Photon coherent scattering. 503 (Unassigned) 504 Photon incoherent scattering. 505 Imaginary scattering factor. 506 Real scattering factor. 507-514 (Unassigned) 515 Pair production, electron field. 516 Pair production; sum of MT=515, Redundant. 517. 517 Pair production, nuclear field. 518 Not allowed in version 6. 519-521 (Unassigned) 522 Photoelectric absorption. Version 5: MT=602. 523 Photo-excitation cross-section. 524-525 (Unassigned) 526 Electro-atomic scattering. Table A3.2. Definitions of reaction types (MT) (cont.)

MT reaction Description Comments 527 Electro-atomic bremsstrahlung. 528 Electro-atomic excitation cross-section. 529-531 (Unassigned) 532 Not allowed in version 6. Version 5: (γ,n). 533 Atomic relaxation data. Version 5: total photonuclear. 534 K (1s1/2) subshell photoelectric or electro-atomic cross-section. 535 L1 (2s1/2) subshell photoelectric or elctro-atomic cross-section. 536 L2 (2p1/2) subshell photoelectric or elctro-atomic cross-section. 537 L3 (2p3/2) subshell photoelectric or elctro-atomic cross-section. 538 M1 (3s1/2) subshell photoelectric or elctro-atomic cross-section. 539 M2 (3p1/2) subshell photoelectric or elctro-atomic cross-section. 540 M3 (3p3/2) subshell photoelectric or elctro-atomic cross-section. 541 M4 (3d3/2) subshell photoelectric or elctro-atomic cross-section. 542 M5 (3d5/2) subshell photoelectric or elctro-atomic cross-section. 543 N1 (4s1/2) subshell photoelectric or elctro-atomic cross-section. 544 N2 (4p1/2) subshell photoelectric or elctro-atomic cross-section. 545 N3 (4p3/2) subshell photoelectric or elctro-atomic cross-section. 546 N4 (4dp3/2) subshell photoelectric or elctro-atomic cross-section. 547 N5 (4d5/2) subshell photoelectric or elctro-atomic cross-section. 548 N6 (4f5/2) subshell photoelectric or elctro-atomic cross-section. 549 N7 (4f7/2) subshell photoelectric or elctro-atomic cross-section. 550 O1 (5s1/2) subshell photoelectric or elctro-atomic cross-section. 551 O2 (5p1/2) subshell photoelectric or elctro-atomic cross-section. 552 O3 (5p3/2) subshell photoelectric or elctro-atomic cross-section. Table A3.2. Definitions of reaction types (MT) (cont.)

MT reaction Description Comments 553 O4 (5d3/2) subshell photoelectric or elctro-atomic cross-section. 554 O5 (5d5/2) subshell photoelectric or elctro-atomic cross-section. 555 O6 (5f5/2) subshell photoelectric or elctro-atomic cross-section. 556 O7 (5f7/2) subshell photoelectric or elctro-atomic cross-section. 557 O8 (5g7/2) subshell photoelectric or elctro-atomic cross-section. 558 O9 (5g9/2) subshell photoelectric or elctro-atomic cross-section. 559 P1 (6s1/2) subshell photoelectric or elctro-atomic cross-section. 560 P2 (6p1/2) subshell photoelectric or elctro-atomic cross-section. 561 P3 (6p3/2) subshell photoelectric or elctro-atomic cross-section. 562 P4 (6d3/2) subshell photoelectric or elctro-atomic cross-section. 563 P5 (6d5/2) subshell photoelectric or elctro-atomic cross-section. 564 P6 (6f5/2) subshell photoelectric or elctro-atomic cross-section. 565 P7 (6f7/2) subshell photoelectric or elctro-atomic cross-section. 566 P8 (6g7/2) subshell photoelectric or elctro-atomic cross-section. 567 P9 (6g9/2) subshell photoelectric or elctro-atomic cross-section. 568 P10 (6h9/2) subshell photoelectric or elctro-atomic cross-section. 569 P11 (6h11/2) subshell photoelectric or elctro-atomic cross-section. 570 Q1 (7s1/2) subshell photoelectric or elctro-atomic cross-section. 571 Q2 (7p1/2) subshell photoelectric or elctro-atomic cross-section. 572 Q3 (7p3/2) subshell photoelectric or elctro-atomic cross-section. 573-599 (Unassigned) 600 (z,p0) Production of a proton leaving the Not allowed for incident protons; use residual nucleus in the ground state. MT=2. 601 (z,p1) Production of a proton, with residual in the 1st excited state. Table A3.2. Definitions of reaction types (MT) (cont.)

MT reaction Description Comments

602 (z,p2) Production of a proton, with residual Version 5: photoelectric absorption; in the 2nd excited state. see MT=522.

603 (z,p3) Production of a proton, with residual in the 3rd excited state. 604 (z,p4) Production of a proton, with residual in the 4th excited state. … … 649 (z,pc) Production of a proton in the continuum not included in the above discrete representation. 650 (z,d0) Production of a deuteron leaving the residual nucleus in the ground state. 651 (z,d1) Production of a deuteron, with the residual in the 1st excited state. 652 (z,d2) Production of a deuteron, with the residual in the 2nd excited state. … … 699 (z,dc) Production of a deuteron in the continuum not included in the above discrete representation. 700 (z,t0) Production of a triton leaving the residual nucleus in the ground state. 701 (z,t1) Production of a triton, with residual in the 1st excited state. 702 (z,t2) Production of a triton, with residual in the 2nd excited state. … … 749 (z,tc) Production of a triton in the continuum not included in the above discrete representation. 3 3 750 (n, He0) Production of a He particle leaving the residual nucleus in the ground state. 3 3 751 (n, He1) Production of a He, with residual in the 1st excited state. … … 3 3 799 (n, Hec) Production of a He in the continuum not included in the above discrete representation.

800 (z,α0) Production of an alpha particle leaving the residual nucleus in the ground state. Table A3.2. Definitions of reaction types (MT) (cont.)

MT reaction Description Comments

801 (z,α1) Production of an alpha particle, with residual in the 1st excited state. … …

849 (z,αc) Production of an alpha particle in the continuum not included in the above discrete representation. 850 (Unassigned) 851-870 Lumped reaction covariances. 871-874 (Unassigned)

875 (z,2n0) Production of 2 neutrons with residual in the ground state. 876 (z,2n1) Production of 2 neutrons with residual in the 1st excited state. … 891 (z,2nc) Production of 2 neutrons in the continuum not included in the above discrete representation. 892-999 (Unassigned)