Online Catalog
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810 -- US DOE 10CFR810 Jurisdiction
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Packages with Keyword: ANISN FORMAT
Package NameAbstractRSICC TapelistTitle
AIR DATAAbstractD00014 I0360 00Sample Input to ANISN for Calculation of Neutron and Secondary Gamma-Ray Transport in Air.
AMPX01AbstractD00027 I3675 02Modular Code System for Generating Coupled Multigroup Neutron-Gamma Libraries from ENDF/B.
APPLE-2AbstractP00111 FM200 00Plotter of Neutron and Gamma-Ray Spectra and Reaction Rates.
APPLE-2AbstractP00111 I3081 00Plotter of Neutron and Gamma-Ray Spectra and Reaction Rates.
BABELAbstractD00104 I3033 00Multi-Purpose Neutron and Gamma-Ray Cross Section Library for Fast Reactor Shielding Design.
BPAbstractD00008 I0360 00Data for Selected Shielding Benchmark Problems Specified in ORNL-RSIC-25, Shielding Benchmark Problems.
BUGLE-80AbstractD00075 IBMPC 03Coupled Neutron, Gamma-Ray Group Cross Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications.
BUGLE-80AbstractD00075 PC386 01Coupled Neutron, Gamma-Ray Group Cross Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications.
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.
CASKAbstractD00023 I3691 0422 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.
CASK-81AbstractD00023 IBMPC 0622 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.
COMANDAbstractP00091 I0360 00A Multigroup ANISN Cross Section Data Library Collapsing Code System.
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.
DOQDPAbstractP00110 I0360 00Discrete Ordinates Quadrature Generator.
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.
EPRAbstractD00037 I3691 05Coupled 100-Group Neutron 21-Group Gamma-ray Cross Sections for EPR Neutronics.
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.
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.
FGXRRSAbstractD00132 C0000 00Few Group Cross Section Library for Research Reactor Calculations.
FIGEROAbstractP00149 C0000 00Processing Codes for Generating Multigroup Neutron Cross Sections from ENDF/B for Use in Discrete Ordinates Calculations.
GARGAbstractD00073 C0000 0027-Group Neutron Cross Sections in Discrete Ordinates Format Generated with FIGERO (PSR-149) from ENDF-B Data.
GIPAbstractP00229 IBMPC 00Group-Organized Cross-Section Input Program.
HALLMARKAbstractD00005 I0360 00Discrete Ordinates and Monte Carlo Results of Neutron and Secondary Gamma-Ray Transport in Air-Over-Ground Geometry.
HELLOAbstractD00058 I0360 0047 Neutron, 21 Gamma-Ray Group Cross Sections for Radiation Transport for Neutron Energies up to 60 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.
LAHIMACKAbstractD00128 I0360 00A Multigroup Library of Neutron and Gamma Cross Sections and Response Functions in the Energy Range up to 800 MeV.
LIBMAKAbstractP00087 I0360 00ANISN-Type Binary Data Processing Code System.
MACK-IVAbstractP00132 I3691 00Calculation of Nuclear Response Functions from Nuclear 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.
MGCLIBAbstractD00118 FM380 00137 and 26 Neutron Multigroup Cross Section Library with the Bondarenko Type Shielding Table.
NABAbstractD00018 I0360 00100-Group, P3, Neutron Cross Section Data for Sodium and Aluminum.
NOXAbstractD00017 I0360 00199-Group, P5, Coupled Neutron and Secondary Gamma-Ray Cross Section Data for Nitrogen and Oxygen.
POWERAbstractP00069 C7600 00Source Distribution Input Data Generator for ANISN Code.
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.
RITTSAbstractD00011 I0360 00121-Group Coupled Neutron and Gamma-Ray Cross-Section Data for Transport Codes.
SAILORAbstractD00076 PC386 01Coupled, Self-Shielded, 47 Neutron, 20 Gamma-Ray, P3, Cross Section Library for Light Water Reactors.
SATURNAbstractP00057 I3675 00P1 or Transport Corrected Multigroup Neutron Cross Section Data Processor.
SHAMSIAbstractD00135 I3033 0048 Group Cross-Section Library for Fusion Nucleonics Analysis.
SPHINXAbstractP00129 C7600 00A One-Dimensional Diffusion and Transport Nuclear Cross Section Processing Code System.
SPHINXAbstractP00129 I0360 00A One-Dimensional Diffusion and Transport Nuclear Cross Section Processing Code System.
UKCTRI-81AbstractD00064 I0370 0146-Group Neutron Cross Sections and Kerma Factors for Fusion Reactor Calculations.
VELMAbstractD00133 I0360 00Multigroup Cross-Section Libraries Based on ENDF/B-V Data for Sodium-Cooled Reactor Shield Analysis.
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.
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.
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.