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810 -- US DOE 10CFR810 Jurisdiction
FEDC -- US Government Agencies and Their Contractors Only
OECD -- Restricted/See Abstract
USSO -- US Distribution Only
USUNV -- US Universities Only
Data Library Collection (nuclear data libraries, etc.)
Package NameAbstractRSICC TapelistTitle
ABBN-90AbstractD00182 MNYCP 00Multigroup Constant Set for Calculation of Neutron and Photon Radiation Fields and Functionals, Including the CONSYST2 Program.
ACTIV87AbstractD00169 ALLCP 00Fast Neutron Activation Cross Section File.
ACTL82AbstractD00069 ALLCP 01Evaluated Neutron Activation Cross-Section Library.
ACTV-F/HAbstractD00155 ALLCP 00Neutron Activation Cross Section Library for Fusion Reactor Design.
ACTV-FUS/INTAbstractD00170 ALLCP 00International Library of Neutron Activation Cross-Section Data for Fusion Reactor Application.
ADS-LIB/V2.0AbstractD00250 MNYCP 00Test Library for Accelerator Driven Systems V2.0
AGDATAAbstractD00127 I0360 00Two Agricultural Production Data Libraries (AGDATC and AGDATG) for Dose and Risk Assessment Models.
AIR DATAAbstractD00014 I0360 00Sample Input to ANISN for Calculation of Neutron and Secondary Gamma-Ray Transport in Air.
AIRFEWGAbstractD00049 I0360 00Results of ANISN Multigroup Calculations of Gamma-Ray, Neutron, and Secondary Gamma-Ray Transport in Infinite Homogeneous Air Using DLC-31/(DPL-1/FEWG1) Cross Sections.
ALBEDO-DATAAbstractD00224 MNYCP 00KSU Neutron Albedo Data.
ALEPH-LIB-JEFF3.1AbstractD00230 MNYCP 00ACE Format Neutron Cross Section Library based on JEFF3.1.
AMPX01AbstractD00027 I3675 02126-Group Coupled Neutron and Gamma-Ray Transport Cross-Section Data Generated by AMPX.
ANS643AbstractD00129 IBMPC 02ANS-6.4.3 Geometric Progression Gamma-Ray Buildup Factor Coefficients.
ANSL-VAbstractD00154 ALLCP 01ENDF/B-V Based Multigroup Cross Section Libraries for Advanced Neutron Source (ANS) Reactor Studies.
BABELAbstractD00104 I3033 00Multi-Purpose Neutron and Gamma-Ray Cross Section Library for Fast Reactor Shielding Design.
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.
BOREHOLE-EB6.8-MGAbstractD00268 MNYCP 00Multi-Group Cross-Section Library for Deterministic and Monte Carlo Codes.
BPAbstractD00008 I0360 00Data for Selected Shielding Benchmark Problems Specified in ORNL-RSIC-25, Shielding Benchmark Problems.
BUGENDF70.BOLIBAbstractD00262 PCX86 00ENDF/B-VII.0 Broad-Group Coupled Cross Section Library for LWR Shielding & Pressure Vessel Dosimetry Applications.
BUGJEFF311.BOLIBAbstractD00254 MNYCP 01JEFF-3.1.1 Broad-Group Coupled Cross Section Library For LWR Shielding & Pressure Vessel Dosimetry Applications.
BUGLE-80AbstractD00075 IBMPC 02Coupled 47 Neutron, 20 Gamma-Ray Group, P3, Cross Section Library for LWR Shielding Calculations by the ANS-6.1.2 Working Group on Multigroup Cross Sections. We recommend ENDF/B-VI derived data in DLC-185/BUGLE-96.
BUGLE-80AbstractD00075 IBMPC 03Coupled 47 Neutron, 20 Gamma-Ray Group, P3, Cross Section Library for LWR Shielding Calculations by the ANS-6.1.2 Working Group on Multigroup Cross Sections. We recommend ENDF/B-VI derived data in DLC-185/BUGLE-96.
BUGLE-80AbstractD00075 PC386 01Coupled 47 Neutron, 20 Gamma-Ray Group, P3, Cross Section Library for LWR Shielding Calculations by the ANS-6.1.2 Working Group on Multigroup Cross Sections. We recommend ENDF/B-VI derived data in DLC-185/BUGLE-96.
BUGLE-93AbstractD00175 ALLCP 01Coupled 47 Neutron, 20 Gamma-Ray Group Cross Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications.
BUGLE-96AbstractD00185 ALLCP 00Coupled 47 Neutron, 20 Gamma-Ray Group Cross Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications.
CADAbstractD00059 I0360 0051 Neutron, 25 Gamma-Ray Group ALBEDO DATA Generated with DOT for Various Materials.
CANDULIB-AECLAbstractD00210 MNYCP 00Burnup-Dependent ORIGEN-S Cross Section Libraries for CANDU Reactor Fuel Characterization.
CASKAbstractD00023 I3691 0422 Neutron, 18 Gamma-Ray Group, P3, Cross Sections for Shipping Cask Analysis.
CASK-81AbstractD00023 IBMPC 0622 Neutron, 18 Gamma-Ray Group, P3, Cross Sections for Shipping Cask Analysis.
CASK-81AbstractD00023 I0370 0522 Neutron, 18 Gamma-Ray Group, P3, Cross Sections for Shipping Cask Analysis.
CLAW-IVAbstractD00036 I0360 02Coupled 30 Neutron, 12 Gamma-Ray Group Cross Sections Based on ENDF/B-IV for Radiation Transport Calculations.
CLAW-IVAbstractD00036 I3033 03Coupled 30 Neutron, 12 Gamma-Ray Group Cross Sections Based on ENDF/B-IV for Radiation Transport Calculations.
CLEARAbstractD00042 I3691 00126 Neutron, 36 Gamma-Ray Cross Sections in AMPX and CCCC Interface Formats for LMFBR Neutronics Calculations.
CLESAbstractD00233 MNYCP 00Cross Section Library of Moderator Materials for Low-Energy Neutron Sources.
COBBAbstractD00016 I3675 01123-Group Neutron Cross Section Data Generated from ENDF/B-II Data for Use in the XSDRN Discrete Ordinates Spectral Averaging Code.
COG SUPPLEMENTAL LIBRARIESAbstractD00271 MNYCP 00COG LibMaker – Data Conversion Utility
COV-15GROUP-2006AbstractD00232 MNYCP 0015-Group Cross Section Covariance Matrix Library.
COVERVAbstractD00077 I0360 01Compilation of Multigroup Cross-section Covariance Matrices in COVERX Format for Several Important Materials (Generated from ENDF/B-V Data using PSR-093/PUFF2).
COVERXAbstractD00044 I0360 02Compilation of Multigroup Cross-Section Covariance Matrices in COVERX Format for Several Important Materials.
COVFILSAbstractD00091 I0360 00A 30-Group Covariance Library Based on ENDF/B-V.
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.
CTR DATAAbstractD00028 I3675 0173-Group P3 Coupled Neutron and Gamma-Ray Cross Sections for Fusion Reactor Calculations.
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.
DECAYREMAbstractD00030 I0360 02Radioactive Decay Spectra in EXREM Format.
DECDC 1.0AbstractD00213 MNYCP 00Nucear Decay Data Files for Radiation Dosimetry Calculations.
DOSCOVAbstractD00090 I0360 0024-Group Covariance Data.
DOSDAM77-81AbstractD00081 C6400 00620 Group, SAND-II Formatted, Neutron Cross Sections Based on ENDF/B-IV and Other Sources 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.
DOSDAT II-81AbstractD00079 I0370 00Dose-Rate Conversion Factors for External Exposure to Photons and Electrons.
DOSDAT-DOEAbstractD00144 ALLMF 00Dose-Rate Conversion Factors for External Exposure to Photons and Electrons.
DOSDAT-DOEAbstractD00144 IBMPC 01Dose-Rate Conversion Factors for External Exposure to Photons and Electrons.
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.
DRALISTAbstractD00080 ALLCP 00Radioactive Decay Data for Application to Radiation Dosimetry and Radiological Assessments.
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.
EACRP-D2O-LATTICESAbstractD00264 MNYCP 00Compilation of Reactor Physics Measurements in HWRs Lattices.
ECPL82AbstractD00106 ALLCP 00Evaluated Charged-Particle Data Library.
EDSFI
USSO
AbstractD00215 PC486 00Electrical Distribution System Functional Inspection Data Base.
ELAST2AbstractD00208 MNYCP 00Database of Cross Sections for the Elastic Scattering of Electrons and Positrons by Atoms.
ELECSPECAbstractD00100 DP010 00Electron Spectra from Decay of Fission Products.
ENDL82AbstractD00103 ALLCP 00Neutron Library in Transmittal Format.
ENDLIB-97AbstractD00179 MNYCP 01LLNL Libraries of Atomic Data, Electron Data, and Photon Data in Evaluated Nuclear Data Library (ENDL) Type Format.
ENSL82-CDRL82AbstractD00110 ALLCP 00Evaluated Nuclear Structure Libraries.
EPICS2014AbstractD00272 MNYCP 00Electron Photon Interaction Cross Sections
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.
ESGAbstractD00065 I0360 0056-Group Cross Section Library Based on VITAMIN-C Generated by Using SPHINX and XSDRNPM to Collapse 171 Groups.
EURLIB-IIIAbstractD00035 I0360 01100 Neutron, 20 Gamma-Ray Group Cross Section Library for Use in the European Shielding Benchmark Program.
FCXSECAbstractD00085 PC386 0122 Neutron, 21 Gamma-Ray Group Cross Section Libraries in ANISN Format for Nuclear Fuel Cycle Shielding Calculations.
FENDL-2.0AbstractD00183 MNYCP 01Compendium of Reference and Processed Sub-libraries Derived from International Evaluated Nuclear Data Files for Fusion Applications.
FENDL-2.1AbstractD00222 MNYCP 00Compendium of Reference and Processed Sub-libraries Derived from International Evaluated Nuclear Data Files for Fusion Applications.
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.
FGR-DOSEAbstractD00167 ALLCP 01Dose Coefficients from Federal Guidance Reports 11 and 12.
FGXRRSAbstractD00132 C0000 00Few Group Cross Section Library for Research Reactor Calculations.
FIREDATAAbstractD00125 PC486 00Nuclear Power Plant Fire Data Base for Personal Computers.
FIS-PRODAbstractD00152 ALLCP 00Chinese Evaluated Fission Product Yield Library in ENDF/B-V Format.
FLEPAbstractD00022 I3033 00Coefficients for the Analytic Representation of Nonelastic Cross Sections and Particle-Emission Spectra from Various Nucleon-Nucleus Collisions in the Energy Range 25 to 400 MeV.
FLUKA05-PRE-LIBAbstractD00260 PCX86 00FLUKA05 Multi-Group, Multi-Purpose Nuclear Data Library, Neutrons, Photons, Charged Particles.
FLUNGAbstractD00086 I3033 00Coupled 35-Group Neutron and 21-Group Gamma Ray, P3 Cross Sections for Fusion Applications.
FPDLAbstractD00066 I0360 00Fission Product Yields, Gamma Ray and Beta Spectra in ENDF-III Format for 235U, 238U, 239Pu, 232Th, and 233U.
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. See DLC-190/FSX96 based on JENDL3.2.
FSXLIB-J33AbstractD00223 MNYCP 01Continuous Energy Neutron Cross Section Library for MCNP Based on JENDL 3.3.
FTFAbstractD00056 I0360 00Multigroup Neutron and Gamma-Ray Dose Transmission Factors for Concrete Slabs.
GAMDAT-78AbstractD00083 I0370 00Library of Gamma-Ray Decay Data for 2055 Radionuclides.
GAMLIBAbstractD00006 I0360 0099-Group Neutron Cross Sections for Use in the GAM Portion of the GGC Multigroup Cross Section Code.
GAMMONAbstractD00071 ALLCP 00Activation Library for Fusion Reaction Application and Other Design Studies.
GAMTABAbstractD00032 I0360 00Radioactive-Decay Gamma-Rays Ordered by Energy and Nuclide.
GAMTOT78AbstractD00109 CY00I 00Compilation of Radioactive Decay and Capture Gamma Rays.
GARGAbstractD00073 C0000 0027-Group Neutron Cross Sections in Discrete Ordinates Format Generated with FIGERO (PSR-149) from ENDF-B Data.
GARLIBAbstractD00013 I3565 01Multigroup Resonance-Region Cross Sections for Use in Shielding Calculations.
GARLIBAbstractD00013 I7090 00Multigroup Resonance-Region Cross Sections for Use in Shielding Calculations.
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.
GROUP STRUCTUREAbstractD00156 ALLCP 00Standard Energy Group Structures Of Cross Section Libraries For Reactor Shielding, Reactor Cell Fusion Neutronics Applications: VITAMIN-J, ECC0-33, ECC0-2000.
GROUPSTRUCTURESAbstractD00274 MNYCP 00GROUPSTRUCTURES, VITAMIN-J, XMAS, ECCO-33, ECCO2000 Standard Group Structures
HALLMARKAbstractD00005 I0360 00Discrete Ordinates and Monte Carlo Results of Neutron and Secondary Gamma-Ray Transport in Air-Over-Ground Geometry.
HATCHES-19AbstractD00206 PC586 02Database for Radiochemical Modelling.
HELLOAbstractD00058 I0360 0047 Neutron, 21 Gamma-Ray Group Cross Sections for Radiation Transport for Neutron Energies up to 60 MeV.
HILOAbstractD00087 I0370 0066 Neutron, 21 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.
HPICEAbstractD00007 I0360 05Evaluated Photon Interaction Library, ENDF/B File 23 Format.
HPPOS 1.5AbstractD00173 IBMPC 00Health Physics Position Database.
HPPOS V2AbstractD00173 IBMPC 01Health Physics Positions (HPPOS) Data Base Based on Current 10 CFR 20.
HUGOAbstractD00099 I3033 00Photon Interaction Data in ENDF/B-V Format.
HUGO VIAbstractD00146 I3033 00Photon Interaction Data in ENDF/B-VI Format. PHOTB6 in DLC-179/ENDLIB-97 is an updated version of these data.
IEAF-2001AbstractD00217 MNYCP 00Intermediate Energy Activation File - 2001.
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 00International Reactor Dosimetry Data.
IRDF-90AbstractD00161 ALLCP 01The International Reactor Dosimetry File.
I-R-MANAbstractD00050 ALLCP 00Photon Interaction Data on ICRP Reference Man.
IRPHE-VENUS-RECYCLEAbstractD00263 MNYCP 00Plutonium Recycling Physics Project Critical Experiments.
JENDL/D-99AbstractD00204 MNYCP 00JENDL Dosimetry File 99.
JENDL-1AbstractD00070 ALLCP 00Japanese Evaluated Nuclear Data Library.
JENDL-2AbstractD00122 FM380 00Japanese Evaluated Neutron Cross Section Data in ENDF/B-IV Format.
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.
KAOS/LIB-VAbstractD00160 CY000 00A Library of Nuclear Response Functions Generated by KAOS-V Code From ENDF/B-V and Other Data Files.
KDDKAbstractD00061 I0360 00Measured Results of Delayed Beta- and Gamma-Ray Spectra due to Thermal-Neutron Fission of U-235.
KEDAK3AbstractD00141 I0370 00Evaluated Neutron Nuclear Data for Reactor Physics Calculations.
KERMALAbstractD00142 ALLCP 00Neutron and Gamma-Ray Kerma Factors Based on LLNL Nuclear Data Files.
KX-RAYAbstractD00021 I0360 00Evaluated X-ray Cross Section Library.
L26P3S34AbstractD00112 IBMMF 00ENDL 26-Group up to P3 Library Prepared by SUPERTOG for 34 Materials.
LA100AbstractD00168 ALLCP 00Evaluated Nuclear Data Library for Transport Calculations Involving Incident Neutrons and Protons of Energy Up to 100 MeV.
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.
LAHIMACKAbstractD00128 I0360 00A Multigroup Library of Neutron and Gamma Cross Sections and Response Functions in the Energy Range up to 800 MeV.
LAS CRUCES
USSO
AbstractD00194 ALLCP 00Las Cruces Trench Site Database, Vadose Model.
LENDLAbstractD00034 I0360 02Livermore Evaluated Neutron and Secondary Gamma-Ray Production Cross-Section Library in ENDF/B-IV Format.
LENDL VAbstractD00120 I0360 00Lawrence Livermore National Laboratory Evaluated Nuclear Data Library in ENDF-V Format.
LEPAbstractD00001 I0360 02Cascade and Evaporation Particle Results from Low-Energy Intranuclear Cascade Calculations.
LIB123AbstractD00153 ALLCP 00AMPX-II P3 123-Group Neutron Cross Section Master Interface Library.
LUMPAbstractD00089 I0360 00Evaluated Lumped Fission Product Cross Sections for Fast Reactor Analysis--Based on ENDF/B-V Data.
MACKLIBAbstractD00029 I3675 00100 Group Neutron Kerma Factors and Reaction Cross Sections Generated by MACK from Data in ENDF Format.
MACKLIB-IV-82AbstractD00060 I0360 01A Library of Nuclear Response Functions Generated with the MACK-V Computer Program from ENDF/B-IV.
MARVIKEN-JIT
OECD
AbstractD00269 MNYCP 00Marviken Full Scale Jet Impingement Tests Experiments.
MASSAbstractD00025 I0360 01Atomic Mass Evaluation.
MATJEFF31.BOLIBAbstractD00242 MNYCP 00Fine-Group Cross Section Library Based on JEFF3.1 for Nuclear Fission Applications.
MATXS1AbstractD00114 C0000 0030-Group Neutron, 12-Group Photon Cross Sections from ENDF/B-IV in MATXS Format.
MATXS10AbstractD00176 ALLCP 0030-Group Neutron, 12-Group Photon Cross Sections from ENDF/B-VI in MATXS Format.
MATXS11AbstractD00177 ALLCP 0080-Group Neutron, 24-Group Photon Cross Sections from ENDF/B-VI in MATXS Format.
MATXS175/42-JEAbstractD00151 D8810 00JEF/EFF Based VITAMIN-J 175 Neutron, 42 Photon Multigroup Data Library in MATXS Format.
MATXS5AAbstractD00115 C0000 0030-Group Neutron, 12-Group Photon Cross Sections from ENDF/B-V in MATSX Format.
MATXS6AAbstractD00116 C0000 0080-Group Neutron, 24-Group Photon Fast-Reactor Cross Section from ENDF/B-V in MATXS Format.
MATXS70-JEF87AbstractD00148 D8810 00JEF/EFF Based 70 Group Neutron Data Library in MATXS Format.
MATXS7AAbstractD00117 C0000 0069-Group Thermal-Reactor Neutron Cross Section Data from ENDF/B-V in MATXS Format.
MATXSLIBJ33AbstractD00258 MNYCP 01JENDL-3.3 Based, 175 N-42 Photon Groups (VITAMIN-J) MATXS Library for Discrete Ordinates Multi-Group.
MCB63NEA.BOLIBAbstractD00216 MNYCP 00ENDF/B-VI Release 3 Cross Section Library for Use with the MCNP Monte Carlo Code.
MCJEF22NEA.BOLIBAbstractD00203 MNYCP 01JEF 2.2 Cross Section Library for the MCNP Monte Carlo Code.
MCJEFF3.1NEAAbstractD00228 MNYCP 00Neutron Cross Section Library Based on JEFF3.1 for Use with MCNP.
MENDL-2PAbstractD00207 MNYCP 00Proton Reaction Data Library for Nuclear Activation (Medium Energy Nuclear Data Library.)
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.
MONTUK-80AbstractD00072 ALLCP 01UKCTR III Transmutation and Activation Data, 100-Group Neutron Activation Cross-Section Data for Fusion Reactor Structure and Coolant Materials.
NABAbstractD00018 I0360 00100-Group, P3, Neutron Cross Section Data for Sodium and Aluminum.
NEACRP-H2O-LATTICESAbstractD00265 MNYCP 00Compilation of Reactor Physics Measurements in LWRs Lattices.
NOXAbstractD00017 I0360 00199-Group, P5, Coupled Neutron and Secondary Gamma-Ray Cross Section Data for Nitrogen and Oxygen.
NPCSL-81AbstractD00082 I0370 00Point Neutron Cross Sections Generated from ENDF/B-IV with the NPTXS Modules of PSR-63/AMPX-II.
NUCDECAYAbstractD00172 PC386 01Nuclear Decay Data for Radiation Dosimetry Calculations for ICRP and MIRD.
NUCDECAYCALCAbstractD00202 PC586 00Nuclear Decay Data for Radiation Dosimetry Calculations for ICRP. See newer version in RASCAL (CCC-553).
ORESUNDAbstractD00267 MNYCP 00Nordic Mesoscale Dispersion Experiments over Land-Water-Land.
ORLIBJ32AbstractD00255 MNYCP 00ORIGEN2 LIBRARIES BASED ON JENDL-3.2.
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.
PADF-2007AbstractD00259 PCX86 00Proton Activation Data File in ENDF-6 Format.
PEFPYDAbstractD00096 ALLMF 02Aggregate Fission-Product Decay Data Based on ENDF/B-IV and -V.
PGAA-IAEAAbstractD00234 MNYCP 00Databsae for Prompt Gamma-Ray Neutron Activation Analysis.
PHOBIAAbstractD00236 PCX86 00Photon buildup factors to account for angular incidence on shield walls.
PHOTXAbstractD00136 D0VAX 01Photon Interaction Cross Section Library.
PHOTXAbstractD00136 IBMPC 00Photon Interaction Cross Section Library.
PIXE2010AbstractD00246 MNYCP 00Proton/alpha Ionization (K, L, M shell), Tabulated Cross Section Library.
PNESDAbstractD00166 PC386 00Proton Nucleus Elastic Scattering Data.
POINT2015AbstractD00273 MNYCP 00POINT 2015: ENDF/B-VII.1 Final Temperature Dependent Cross Section Library
POPLIBAbstractD00012 I0360 03A Compendium of Neutron-Induced Secondary Gamma-Ray Yield and Cross Section Data.
PR-EDBAbstractD00196 IBMPC 03Power Reactor Embrittlement Data Base, Version 3.
PUCORAbstractD00067 I3691 0084 Group Neutron Cross Sections for Uranium-Plutonium Cycle LWR and PWR Models in AMPX Master Library Format.
PUDKAbstractD00074 I0360 00Measured Results of Delayed Beta- and Gamma-Ray Spectra Due to Thermal-Neutron Fission of Pu239 and Pu241.
PVCAbstractD00048 I3691 0036 Group, P5, Photon Interaction Cross Sections for 38 Materials in ANISN Format.
PVEAbstractD00126 I3033 0038 Group, P8, Photon Interaction Cross Sections in ANISN Format from VITAMIN-E.
PWR-AXBUPRO-GKNAbstractD00209 MNYCP 00Measured Axial Burnup Profiles for NeckarWesthiem PWR Reactors.
PWR-AXBUPRO-SNLAbstractD00201 MNYCP 00Axial Burnup Profile Database for Pressurized Water Reactors.
RADDECAY 4.02AbstractD00134 IBMPC 03Radioactive Decay Data for Radiological Assessments.
RECOILAbstractD00055 I3033 01Multigroup Primary Recoil Spectra, Displacement Rates and Gas-Production Rates for Radiation Damage Studies.
RITTSAbstractD00011 I0360 00121-Group Coupled Neutron and Gamma-Ray Cross-Section Data for Transport Codes.
SAILAbstractD00057 I0360 0023 Neutron, 17 Gamma-Ray Group ALBEDO DATA for Concrete and Steel, Based on DOT 1-1/2-D Calculations using DLC-31/FEWG1 Data.
SAILORAbstractD00076 I3033 00Coupled, Self-Shielded, 47 Neutron, 20 Gamma-Ray, P3, Cross Section Library for Light Water Reactors. We recommend ENDF/B-VI derived data in DLC-185/BUGLE-96.
SAILORAbstractD00076 PC386 01Coupled, Self-Shielded, 47 Neutron, 20 Gamma-Ray, P3, Cross Section Library for Light Water Reactors. We recommend ENDF/B-VI derived data in DLC-185/BUGLE-96.
SENPROAbstractD00045 I3691 02Compilation of Multigroup Sensitivity Profiles in SENPRO Format for Fast Reactor Core and Shield Benchmarks and Thermal Reactor Benchmarks.
SERPENT117-ACELIBAbstractD00249 MNYCP 00Continuous-Energy X-Sec Library, Radioactive Decay, Fission Yield Data for SERPENT in ACE.
SHAMSIAbstractD00135 I3033 0048 Group Cross-Section Library for Fusion Nucleonics Analysis.
SIGMA-AAbstractD00139 ALLMF 00Photon Interaction and Absorption Cross Sections.
SIGMA-AAbstractD00139 IBMPC 00Photon Interaction and Absorption Cross Sections.
SINBAD 2017.12AbstractD00237 MNYCP 05Shielding Integral Benchmark Archive and Database, Version December 2017
SKYDATA-KSUAbstractD00188 IBMPC 00Parameters for Approximate Neutron and Gamma-Ray Skyshine Response Functions and Ground Correction Factors.
SKYPORTAbstractD00093 IBMPC 00Skyshine Importance Functions for Neutrons and Gamma Rays.
SNLRMLAbstractD00178 ALLCP 00Recommended Dosimetry Cross Section Compendium.
STORM-ISRAELAbstractD00015 I0360 01Evaluated Photon Interaction Library, ENDF/B File 23 Format.
TDFAbstractD00162 ALLCP 00Thermonuclear Data File.
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 MCNPX.
THERMGAMAbstractD00140 ALLCP 00Prompt Gamma Rays from Thermal-Neutron Capture.
TPASGAM 85AbstractD00088 ALLCP 04Radioactive Decay Library of Gamma-Ray Energies, Branching Ratios, and Cross Sections.
TRANSMITAbstractD00020 I0360 00Experimental Neutron Transmission Data Used to Test Total Cross Sections.
TR-EDBAbstractD00198 IBMPC 00Test Reactor Embrittlement Data Base, Version 1.
TSL-ACE/2013AbstractD00270 ALLCP 00TSL-ACE/2013
UKCTRI-81AbstractD00064 I0370 0146-Group Neutron Cross Sections and Kerma Factors for Fusion Reactor Calculations.
UKFY2AbstractD00171 IBMPC 00UK Fission Product Yield Library, Version 2.
UKNDLAbstractD00039 I0370 00United Kingdom Evaluated Neutron Cross-Section Data Library.
UKNDL-81AbstractD00107 I3033 00The Aldermaston Nuclear Data Library.
UNGERAbstractD00164 PC386 00Effective Dose Equivalent for Specific Radionuclides.
UTXS6AbstractD00211 MNYCP 00MCNP Continuous-Energy Neutron Cross Section Libraries for Temperatures from 300 to 1365K.
VELMAbstractD00133 I0360 00Multigroup Cross-Section Libraries Based on ENDF/B-V Data for Sodium-Cooled Reactor Shield Analysis.
VIP-MANAbstractD00256 MNYCP 00Computational Phantom.
VITAMIN-4CAbstractD00053 I3691 00171 Neutron Group Cross Sections and Bondarenko Factors in CCCC Interface Formats for Fusion and LMFBR Neutronics.
VITAMIN-B6AbstractD00184 ALLCP 00A Fine-Group Cross Section Library Based on ENDF/B-VI Release 3 for Radiation Transport Applications.
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 02171 Neutron, 36 Gamma-Ray Group Cross Sections in AMPX and CCCC Interface Formats for Fusion and LMFBR Neutronics.
VITAMIN-EAbstractD00113 I3033 02174n, 38g Cross-Section Library in AMPX Format.
VITAMIN-J/COVAAbstractD00157 D8810 00Neutron Cross-Section Covariance Data in Multigroup Form.
VITAMIN-J/COVA/EFFAbstractD00197 ALLCP 00Neutron Cross-Section Covariance Data in Multigroup Form.
VITAMIN-J/KERMAAbstractD00150 I3090 00VITAMIN-J 175-Neutron and 38-Photon Kerma And Gas Production Cross Sections.
VITENDF70.BOLIBAbstractD00261 PCX86 00ENDF/B-VII.0 Multi-Group Coupled (199n +42gamma) Cross Section Library in AMPX Format for Nuclear Fission Applications.
VITENEA-EAbstractD00240 MNYCP 00VITENEA-E, AMPX 174-N,38-Gamma Multigroup X-sec. Library for Multidimensional Radiation Transport and Dose Evaluation.
VITENEA-JAbstractD00238 MNYCP 00AMPX 175-n,42-g Multigroup X-section Library for Nuclear Fusion Applications.
VITJEF22.BOLIBAbstractD00241 MNYCP 00JEF-2.2 Multigroup Coupled (199n + 42?) Cross-Section Library in AMPX Format for Nuclear Fission Applications.
VITJEFF31.BOLIBAbstractD00235 MNYCP 00A JEFF-3.1 Multigr Coupled (199n + 42gamma) X-Section Lib. in AMPX Fmt for Nuclear Fission Applications.
VITJEFF311.BOLIBAbstractD00257 MNYCP 01JEFF-3.1.1 Multi-Group Coupled (199n + 42gamma) X-Section Library in AMPX Format for Nuclear Fission Applications.
WIMKAL-88AbstractD00193 MNYCP 0069 Energy Group, Neutron Cross Section Library For Thermal Reactor Calculations in WIMSD Format.
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 00JEF-1 Based 69 Group Neutron Data 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.
XCOMAbstractD00174 IBMPC 00Photon Cross Sections on a Personal Computer, Versions 1.2 and 1.3.
XG-IAEAAbstractD00163 IBMPC 00X-ray and Gamma-ray Standards For Detector Calibration.
YUMMYAbstractD00221 MNYCP 00Multi-temperature, Neutron Cross Section Library Based on ENDF/B-V and ENDF/B-VI for use with MCNP.
ZZ-PWR-MSLBAbstractD00275 MNYCP 00ZZ PWR-MSLB, PWR Main Steam-Line Break Benchmarks, Coupled Neutronics Thermal-Hydraulics
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.