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Packages with Keyword: NEUTRON CROSS SECTIONS
Package NameAbstractRSICC TapelistTitle
ABBN-90AbstractD00182 MNYCP 00Multigroup Constant Set for Calculation of Neutron and Photon Radiation Fields and Functionals, Including the CONSYST2 Program.
ACTL82AbstractD00069 ALLCP 01Evaluated Neutron Activation Cross-Section Library.
ACTV-F/HAbstractD00155 ALLCP 00Neutron Activation Cross Section Library for Fusion Reactor Design.
ALEPH-LIB-JEFF3.1AbstractD00230 MNYCP 00ACE Format Neutron Cross Section Library based on JEFF3.1.
ANSL-VAbstractD00154 ALLCP 01ENDF/B-V Based Multigroup Cross Section Libraries for Advanced Neutron Source (ANS) Reactor Studies.
BARC-35AbstractD00124 IBMMF 0035-Group Neutron Cross Sections and Resonance Self-Shielding Factors Generated in ISOTXS and BRKOXS Format from ENDF/B-IV Using MINX.
BUGLE-93AbstractD00175 ALLCP 01Coupled Neutron, Gamma-Ray Group Cross Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications.
BUGLE-96AbstractD00185 ALLCP 00Coupled Neutron, Gamma-Ray Group Cross Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications.
CANDULIB-AECLAbstractD00210 MNYCP 00Burnup-Dependent ORIGEN-S Cross Section Libraries for CANDU Reactor Fuel Characterization.
COBBAbstractD00016 I3675 01123-Group Neutron Cross Section Data Generated from ENDF/B-II Data for Use in the XSDRN Discrete Ordinates Spectral Averaging Code.
COVFILS-2AbstractD00137 ALLCP 00Neutron Data and Covariances for Sensitivity and Uncertainty Analysis.
CRYO-S(A,B)-ACE1AbstractD00253 MNYCP 00Scattering Law and Continuous Energy Cross Section Library of Materials at Cryogenic Temperatures.
DABL69AbstractD00130 I0360 01Defense Nuclear Applications Broad-Group Library based on ENDF/B-V in ANISN Format.
DDXLIBAbstractD00123 FM380 01125-Neutron Group Double Differential Cross Section Library.
DOSDAM77-81AbstractD00081 C6400 00Multigroup Cross Sections in SAND-II Format for Spectral, Integral, and Damage Analyses.
DOSDAM81-82AbstractD00097 C0000 00Multigroup Cross Sections in SAND-II Format for Spectral, Integral, and Damage Analyses.
DOSDAM84AbstractD00131 IBMMF 00Multigroup Cross Sections in SAND-II Format for Spectral, Integral, and Damage Analyses.
DPL-400 GEDT1AbstractD00031 I0360 08Defense Nuclear Agency 37 Neutron, 21 Gamma Ray Coupled, P3, Multigroup Library in ANISN Format.
DPL-401 NEDTAbstractD00031 I0360 09Defense Nuclear Agency 37 Neutron, 21 Gamma Ray Coupled, P3, Multigroup Library in ANISN Format.
DPL-402A/GPDT1AbstractD00031 I0360 10Defense Nuclear Agency 37 Neutron, 21 Gamma Ray Coupled, P3, Multigroup Library in ANISN Format.
DPL-402B/GPDT1AbstractD00031 I0360 11Defense Nuclear Agency 37 Neutron, 21 Gamma Ray Coupled, P3, Multigroup Library in ANISN Format.
E3LWRAbstractD00098 C0000 0045 Neutron, 16 Gamma-Ray and 15 Neutron, 5 Gamma-Ray Group LWR Cross Section Libraries Derived from EURLIB-III using the AGRUKO Optimized Collapsing Scheme.
ENDL82AbstractD00103 ALLCP 00Neutron Library in Transmittal Format.
EPRAbstractD00037 I3691 05Coupled 100-Group Neutron 21-Group Gamma-ray Cross Sections for EPR Neutronics.
EPR MASTERAbstractD00052 I3691 00100 Neutron Group Cross Sections in AMPX Master Library Format.
EURLIB-IIIAbstractD00035 I0360 01100 Neutron, 20 Gamma-Ray Group Cross Section Library for Use in the European Shielding Benchmark Program.
FEWG1-81AbstractD00031 I0370 06Defense Nuclear Agency 37 Neutron, 21 Gamma Ray Coupled, P3, Multigroup Library in ANISN Format.
FEWG1-85AbstractD00031 I0360 07Defense Nuclear Agency 37 Neutron, 21 Gamma Ray Coupled, P3, Multigroup Library in ANISN Format.
FSX96AbstractD00190 MNYWS 00Collection of Continuous Energy Cross Section Libraries for MCNP Based on JENDL 3.2, JENDL, Fusion File and Dosimetry File.
FSXJ32AbstractD00244 MNYCP 00A Continuous Energy Cross Section MCNP Nuclear Data Library Based on JENDL-3.2.
FSXLIB-J3AbstractD00165 ALLCP 00MCNP continuous energy neutron cross section library based on JENDL-3.
FSXLIB-J33AbstractD00223 MNYCP 01Continuous Energy Neutron Cross Section Library for MCNP Based on JENDL 3.3.
GAMLIBAbstractD00006 I0360 0099-Group Neutron Cross Sections for Use in the GAM Portion of the GGC Multigroup Cross Section Code.
GARGAbstractD00073 C0000 0027-Group Neutron Cross Sections in Discrete Ordinates Format Generated with FIGERO (PSR-149) from ENDF-B Data.
GEAF-1AbstractD00158 D8810 00100 Group Cross Sections for Neutron Activation.
GICX40AbstractD00092 ALLCP 00Coupled 42-Neutron, 21-Gamma-Ray Group Cross Sections for 40 Elements in Group Independent Form for Fusion Reactor Calculations.
HILOAbstractD00087 I0370 0066 Neutron, 22 Gamma-Ray Group Cross Sections for Radiation Transport for Neutron Energies up to 400 MeV.
HILO2KAbstractD00220 MNYCP 00Coupled 83 Neutron, 22 Photon Group Cross Sections for Neutron Energies Up to 2 GeV.
HILO86AbstractD00119 I0360 0066 Neutron, 22 Gamma-Ray Group Cross Sections for Radiation Transport for Neutron Energies Up to 400 MeV.
HILO86AbstractD00119 PC386 0166 Neutron, 22 Gamma-Ray Group Cross Sections for Radiation Transport for Neutron Energies Up to 400 MeV.
HILO86RAbstractD00187 ALLCP 0066 Neutron, 22 Gamma-Ray Group Cross Sections for Radiation Transport for Neutron Energies Up to 400 MeV.
IRAN-LIBAbstractD00159 IBMPC 00A P-3 Coupled Neutron-Gamma Cross Section Library in ISOTXS For Use with ANISN/PC (CCC-514).
IRDF-2002AbstractD00229 MNYCP 01The International Reactor Dosimetry File.
IRDF82AbstractD00094 I0360 00The International Reactor Dosimetry File.
JFSAbstractD00111 I3033 0070 Group Neutron Fast Reactor Cross Section Set and 25 Group Neutron Fast Reactor Cross Section Set.
JFS3J2AbstractD00108 FM200 0070 Group Neutron Fast Reactor Cross Section Set Based on JENDL-2B.
JIMCOFAbstractD00078 F2307 00Multigroup Constants fFle Based on ENDF/B IV.
KEDAK3AbstractD00141 I0370 00Evaluated Neutron Nuclear Data for Reactor Physics Calculations.
L26P3S34AbstractD00112 IBMMF 00ENDL 26-Group up to P3 Library Prepared by SUPERTOG for 34 Materials.
LAFPX-VAbstractD00054 C0000 01A Multigroup Reaction Cross-Section Collapsing Code and Library of 154-Group Fission-Product Cross Sections.
LAFPX-VAbstractD00054 C0000 02A Multigroup Reaction Cross-Section Collapsing Code and Library of 154-Group Fission-Product Cross Sections.
LENDLAbstractD00034 I0360 02Livermore Evaluated Neutron and Secondary Gamma-Ray Production Cross-Section Library in ENDF/B-IV Format.
LIB123AbstractD00153 ALLCP 00AMPX-II P3 123-Group Neutron Cross Section Master Interface Library.
MACKLIBAbstractD00029 I3675 00A Library of Nuclear Response Functions Generated with the MACK-V Computer Program from ENDF/B-IV.
MACKLIB-IV-82AbstractD00060 I0360 01A Library of Nuclear Response Functions Generated with the MACK-V Computer Program from ENDF/B-IV.
MATXS1AbstractD00114 C0000 00Neutron, Photon Cross Sections from ENDF/B-IV in MATXS Format.
MATXS10AbstractD00176 ALLCP 00Neutron, Photon Cross Sections from ENDF/B-IV in MATXS Format.
MATXS11AbstractD00177 ALLCP 00Neutron, Photon Cross Sections from ENDF/B-IV in MATXS Format.
MATXS5AAbstractD00115 C0000 00Neutron, Photon Cross Sections from ENDF/B-IV in MATXS Format.
MATXS6AAbstractD00116 C0000 00Neutron, Photon Cross Sections from ENDF/B-IV in MATXS Format.
MATXS70-JEF87AbstractD00148 D8810 00JEF/EFF Based 70 Group Neutron Data Library in MATXS Format.
MATXS7AAbstractD00117 C0000 00Neutron, Photon Cross Sections from ENDF/B-IV in MATXS Format.
MCJEFF3.1NEAAbstractD00228 MNYCP 00Neutron Cross Section Library Based on JEFF3.1 for Use with MCNP.
MENSLIBAbstractD00084 I0370 0060 Group, P5, Cross Sections in DTF-IV for Transport Calculations for Neutrons with Energies Up to 60 MeV.
MGCLIBAbstractD00118 FM380 00137 and 26 Neutron Multigroup Cross Section Library with the Bondarenko Type Shielding Table.
NPCSL-81AbstractD00082 I0370 00Point Neutron Cross Sections Generated from ENDF/B-IV with the NPTXS Modules of PSR-63/AMPX-II.
ORYX-EAbstractD00038 I0360 00ORIGEN Yields and Cross Sections Nuclear Transmutation and Decay Data from ENDF/B-IV.
ORYX-EAbstractD00038 I0360 01ORIGEN Yields and Cross Sections Nuclear Transmutation and Decay Data from ENDF/B-IV.
REFIT-2009AbstractC00775 PCX86 00Multilevel Resonance Parameter Least Square Fit of Neutron Transmission, Capture, Fission & Self Indication Data.
SNLRMLAbstractD00178 ALLCP 00Recommended Dosimetry Cross Section Compendium.
SUGGELAbstractP00508 MNYWS 00Program Suggesting the Orbital Angular Momentum of a Neutron Resonance From the Magnitude Of Its Neutron Width.
TEMPEST-2AbstractP00558 I0360 00Thermalization Program for Neutron Spectra and MultiGroup Cross-Sections.
TENDL-2008-ACEAbstractD00243 MNYCP 00TALYS-Based Cross Section Library for Use with MCNP(X).
TENDL-2010-ACEAbstractD00248 MNYCP 00TALYS-Based Cross Section Library for Use with MCNP(X).
TENDL-2011-ACEAbstractD00252 MNYCP 00TALYS-Based Cross Section Library for Use with MCNP(X).
TENDL-2012-ACEAbstractD00266 MNYCP 00TALYS-Based Cross Section Library for Use with MCNP(X).
TSL-ACE/2013AbstractD00270 ALLCP 00TSL-ACE/2013
UKCTRI-81AbstractD00064 I0370 0146-Group Neutron Cross Sections and Kerma Factors for Fusion Reactor Calculations.
UKNDLAbstractD00039 I0370 00United Kingdom Evaluated Neutron Cross-Section Data Library.
UKNDL-81AbstractD00107 I3033 00The Aldermaston Nuclear Data Library.
VELMAbstractD00133 I0360 00Multigroup Cross-Section Libraries Based on ENDF/B-V Data for Sodium-Cooled Reactor Shield Analysis.
VITAMIN-4CAbstractD00053 I3691 00Broad-Group and Fine-Group and Coupled Neutron/Gamma Cross-Section Libraries Derived from ENDF/B-VII.0 Nuclear Data.
VITAMIN-B6AbstractD00184 ALLCP 00Broad-Group and Fine-Group and Coupled Neutron/Gamma Cross-Section Libraries Derived from ENDF/B-VII.0 Nuclear Data.
VITAMIN-B7/BUGLE-B7AbstractD00245 MNYCP 01Broad-Group and Fine-Group and Coupled Neutron/Gamma Cross-Section Libraries Derived from ENDF/B-VII.0 Nuclear Data.
VITAMIN-CAbstractD00041 I0360 02Broad-Group and Fine-Group and Coupled Neutron/Gamma Cross-Section Libraries Derived from ENDF/B-VII.0 Nuclear Data.
VITAMIN-EAbstractD00113 I3033 02Broad-Group and Fine-Group and Coupled Neutron/Gamma Cross-Section Libraries Derived from ENDF/B-VII.0 Nuclear Data.
VITAMIN-J/COVAAbstractD00157 D8810 00Broad-Group and Fine-Group and Coupled Neutron/Gamma Cross-Section Libraries Derived from ENDF/B-VII.0 Nuclear Data.
VITAMIN-J/COVA/EFFAbstractD00197 ALLCP 00Broad-Group and Fine-Group and Coupled Neutron/Gamma Cross-Section Libraries Derived from ENDF/B-VII.0 Nuclear Data.
VITJEFF31.BOLIBAbstractD00235 MNYCP 00A JEFF-3.1 Multigr Coupled (199n + 42gamma) X-Section Lib. in AMPX Fmt for Nuclear Fission Applications.
WIMSLIB-IJS0AbstractD00147 D8810 00Extended Version of the WIMS 69-group Library.
WIMSLIB-IJS1AbstractD00147 D8810 01Extended Version of the WIMS 69-group Library.
WIMSLIB-JEF87AbstractD00095 D0VAX 00Extended Version of the WIMS 69-group Library.
WLUP 3.0AbstractD00231 MNYCP 0169- and 172- Group Cross Section Libraries for WIMS.
W-M-NRSMAbstractD00026 U1108 00WANL-MSFC Nuclear Rocket Shielding Methods Data Generator (GAMLEG-W, APPROPOS, NAGS, and SATURN) and Multigroup Neutron and Gamma-ray Cross Section Libraries 1-6.
YUMMYAbstractD00221 MNYCP 00Multi-temperature, Neutron Cross Section Library Based on ENDF/B-V and ENDF/B-VI for use with MCNP.
The Radiation Safety Information Computational Center (RSICC) is a Department of Energy Specialized Information Analysis Center (SIAC) authorized to collect, analyze, maintain, and distribute computer software and data sets in the areas of radiation transport and safety. RSICC resides in the Reactor and Nuclear Systems Division (RNSD) at Oak Ridge National Laboratory.