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Packages with Keyword: MULTIGROUP CROSS SECTIONS
Package NameAbstractRSICC TapelistTitle
1DXAbstractP00096 U1108 00A One-Dimensional Diffusion Code System for Producing Energy Group Collapsed and Self-Shielded Cross Sections.
ABBN-90AbstractD00182 MNYCP 00Multigroup Constant Set for Calculation of Neutron and Photon Radiation Fields and Functionals, Including the CONSYST2 Program.
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.
AMPX-77AbstractP00315 ALLMF 01Modular Code System for Generating Coupled Multigroup Neutron-Gamma Libraries from ENDF/B.
ANSL-VAbstractD00154 ALLCP 01ENDF/B-V Based Multigroup Cross Section Libraries for Advanced Neutron Source (ANS) Reactor Studies.
AXMIX-PCAbstractP00297 IBMPC 00ANISN Cross Section Mixing Code System.
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.
CLEARAbstractD00042 I3691 00126 Neutron, 36 Gamma-Ray Cross Sections in AMPX and CCCC Interface Formats for LMFBR Neutronics Calculations.
COBBAbstractD00016 I3675 01123-Group Neutron Cross Section Data Generated from ENDF/B-II Data for Use in the XSDRN Discrete Ordinates Spectral Averaging Code.
CODAC (2)AbstractP00073 I0360 00For TIMOC 72, Monte Carlo Three-Dimensional Neutron Transport Code's Data Generator.
COMANDAbstractP00091 I0360 00A Multigroup ANISN Cross Section Data Library Collapsing Code System.
COMBINE-PCAbstractP00286 IBMPC 00Code System to Compute Neutron Spectra and ENDF/B Version 5 Based Multigroup Neutron Constants.
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.
DETAN 95AbstractP00361 MNYCP 00Code System to Calculate Spectrum-Averaged Cross Sections and Detector Responses in Neutron Spectra.
DINTAbstractP00049 C6600 00Multigroup Coherent-Incoherent Cross Section Data Generator for Photon Transport Calculations.
DINTAbstractP00049 I0360 00Multigroup Coherent-Incoherent Cross Section Data Generator for Photon Transport Calculations.
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.
ENBAL2AbstractP00160 I0370 00A Program to Generate Multigroup Neutron Kerma Factors.
ENDL82AbstractD00103 ALLCP 00Neutron Library in Transmittal Format.
ENTOSANAbstractP00188 C0175 00Code System for Calculating Fine-Group Dosimetry Cross Section Values from ENDF/B Data.
ENTOSANAbstractP00188 D8810 00Code System for Calculating Fine-Group Dosimetry Cross Section Values from ENDF/B Data.
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.
FDMXPCAbstractP00322 IPCAT 00Code System for Calculation of Neutron Transmission and Other Functionals from Evaluated Data in ENDF Format.
FEDGROUP-3AbstractP00123 I0360 00Code System for Processing Evaluated Nuclear in ENDF/B, KEDAK or UKNDL Formats into Constants for Reactor Physics Calculation.
FEDGROUPC86REV3AbstractP00194 MNYCP 01Code System for Processing Evaluated Nuclear in ENDF/B, KEDAK or UKNDL Formats into Constants for Reactor Physics Calculation.
FEDGROUP-RAbstractP00349 MNYCP 00Code System for Processing Evaluated Nuclear in ENDF/B, KEDAK or UKNDL Formats into Constants for Reactor Physics Calculation.
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.
FIGEROAbstractP00149 C0000 00Processing Codes for Generating Multigroup Neutron Cross Sections from ENDF/B for Use in Discrete Ordinates Calculations.
FITOCOAbstractP00189 C0175 00Converter of Fine-Group Flux Density and Cross Section Data to Coarse Group Values.
FLUNGAbstractD00086 I3033 00Coupled 35-Group Neutron and 21-Group Gamma Ray, P3 Cross Sections for Fusion Applications.
FORSENAbstractP00170 I0360 00A Multigroup Processing Code for Use with Sensitivity Profiles to Assess the Effect of Cross Section Changes.
GALAXY-6AbstractP00098 I0370 00Neutron Multigroup Cross Section Processor.
GAMLEG-75AbstractP00086 C7600 00Multigroup Cross Section Generator for Photon Transport Calculations.
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.
GARLIBAbstractD00013 I3565 01Multigroup Resonance-Region Cross Sections for Use in Shielding Calculations.
GARLIBAbstractD00013 I7090 00Multigroup Resonance-Region Cross Sections for Use in Shielding Calculations.
GAROLAbstractP00033 I7090 00Calculation of Resonance Neutron Absorption in Two-Region Problems.
GEAF-1AbstractD00158 D8810 00100 Group Cross Sections for Neutron Activation.
GECINXAbstractP00193 H6000 00A Code System for Collapsing Multigroup Cross Sections in CCCC Format.
GGC-3AbstractP00012 I3565 00Multigroup Cross Section Code System for Use in Diffusion and Transport Codes.
GGC-3 & GGC-4AbstractP00012 I3675 00Multigroup Cross Section Code System for Use in Diffusion and Transport Codes.
GGC-4AbstractP00012 U1108 00Multigroup Cross Section Code System for Use in Diffusion and Transport Codes.
GGTC-ENELAbstractP00128 I0360 00Code System for Producing Few-Group Neutron Cross Sections from Multigroup Data Libraries.
GICX40AbstractD00092 ALLCP 00Coupled 42-Neutron, 21-Gamma-Ray Group Cross Sections for 40 Elements in Group Independent Form for Fusion Reactor Calculations.
GIPAbstractP00229 IBMPC 00Group-Organized Cross-Section Input Program.
GROUPXSAbstractP00246 C0740 00Processing of Double-Differential Cross Sections in the New ENDF-VI Format.
HELLOAbstractD00058 I0360 0047 Neutron, 21 Gamma-Ray Group Cross Sections for Radiation Transport for Neutron Energies up to 60 MeV.
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.
IEAF-2001AbstractD00217 MNYCP 00Intermediate Energy Activation File - 2001.
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.
LAHIMACKAbstractD00128 I0360 00A Multigroup Library of Neutron and Gamma Cross Sections and Response Functions in the Energy Range up to 800 MeV.
LEAP-ADDELTAbstractP00138 I0360 00Multigroup Thermal Neutron Scattering Data Generator for Hydrogen in Light Water and Deuterium in Heavy Water.
LIB123AbstractD00153 ALLCP 00AMPX-II P3 123-Group Neutron Cross Section Master Interface Library.
LIBMAKAbstractP00087 I0360 00ANISN-Type Binary Data Processing Code System.
MACK-IVAbstractP00132 I3691 00Calculation of Nuclear Response Functions from Nuclear Data in ENDF Format.
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.
MARSAbstractP00117 I0360 00Collection of Computer Codes for Manipulating Multigroup Cross Section Libraries in AMPX or CCCC Formats.
MATJEFF31.BOLIBAbstractD00242 MNYCP 00Fine-Group Cross Section Library Based on JEFF3.1 for Nuclear Fission Applications.
MATXS1AbstractD00114 C0000 00Neutron, Photon Cross Sections from ENDF/B-IV in MATXS Format.
MATXS175/42-JEAbstractD00151 D8810 00JEF/EFF Based VITAMIN-J 175 Neutron, 42 Photon Multigroup Data Library 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.
MENSLIBAbstractD00084 I0370 0060 Group, P5, Cross Sections in DTF-IV for Transport Calculations for Neutrons with Energies Up to 60 MeV.
MINIGALAbstractP00180 I3033 00Neutron Cross Section Processing System for Calculating Average Values from Data in the Standard United Kingdom Nuclear Data Library Format.
MINXAbstractP00105 C6600 00Multigroup Interpretation of Nuclear X-Sections from ENDF/B Standard CCCC-III Interface Formats.
MINXAbstractP00105 I0360 00Multigroup Interpretation of Nuclear X-Sections from ENDF/B Standard CCCC-III Interface Formats.
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.
NANICKAbstractP00120 I0360 00Infinitely-Diluted Multigroup Cross-Section Generator - from ENDF/B.
NJOY91.119AbstractP00171 MFMWS 04Code System for Producing Pointwise and Multigroup Neutron and Photon Cross Sections from ENDF/B Data.
NJOY94.61AbstractP00355 MFMWS 03Code System for Producing Pointwise and Multigroup Neutron and Photon Cross Sections from ENDF/B Data.
NJOY-UTIL-EIRAbstractP00296 C0825 00Utilities For the NJOY (6/83) Nuclear Data Processing System.
NOXAbstractD00017 I0360 00199-Group, P5, Coupled Neutron and Secondary Gamma-Ray Cross Section Data for Nitrogen and Oxygen.
NSLINKAbstractP00314 D0VAX 00NJOY SCALE LINK.
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.
PAPINAbstractP00156 I0370 00A Code System to Calculate Cross Section Probability Tables, Bondarenko and Transmission Self-Shielding Factors for Fertile Isotopes in the Unresolved Resonance Region.
PIXSEAbstractP00133 I0360 00A Generator of Multigroup and Multipoint Cross Sections for Thermal Reactor Calculations.
PLASMXAbstractP00106 C6600 00A Multigroup Ionization and Charge Exchange Cross-Section Code System for Neutral Hydrogen Transport in Plasmas.
POPOP4AbstractP00011 I3675 00Converter of Gamma-Ray Spectra to Secondary Gamma-Ray Production Cross Sections.
PUCORAbstractD00067 I3691 0084 Group Neutron Cross Sections for Uranium-Plutonium Cycle LWR and PWR Models in AMPX Master Library Format.
PVCAbstractD00048 I3691 0036 Group, P5, Photon Interaction Cross Sections for 38 Materials in ANISN Format.
RITTSAbstractD00011 I0360 00121-Group Coupled Neutron and Gamma-Ray Cross-Section Data for Transport Codes.
ROLAIDS-CPMAbstractP00353 SUN04 00Code System to Calculate Group-Averaged Cross Sections Using the Collision Probability Method.
S1CALCAbstractP00134 I0360 00A Multigroup Thermal Neutron Scattering Law Data Generator for Hydrogen and Deuterium.
SATURNAbstractP00057 I3675 00P1 or Transport Corrected Multigroup Neutron Cross Section Data Processor.
SCAMPIAbstractP00352 MNYWS 01Collection of Codes for Manipulating Multigroup Cross Section Libraries in AMPX Format.
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.
SUPERTOG-JR.AbstractP00115 F2307 00Data Generator--Fine Group Constants and PN Scattering Matrices from ENDF/B.
SUPERTOG-JR.AbstractP00115 I0360 00Data Generator--Fine Group Constants and PN Scattering Matrices from ENDF/B.
TDOWN-IVAbstractP00172 H6000 00A Code System to Generate Composition- and Spatially-Dependent Neutron Cross Sections for Multigroup Neutronics Analysis.
THERMOS-OTAAbstractP00107 C0173 00Multigroup Integral Transport Code System for Thermal Lattice Calculations using Collision Probability Method for Slabs and Cylinders.
THERMOS-OTAAbstractP00107 C0740 00Multigroup Integral Transport Code System for Thermal Lattice Calculations using Collision Probability Method for Slabs and Cylinders.
THERMOS-OTAAbstractP00107 U1108 00Multigroup Integral Transport Code System for Thermal Lattice Calculations using Collision Probability Method for Slabs and Cylinders.
TIMS-1AbstractP00163 D0780 00Processing Code System for Production of Group Constants of Heavy Resonant Nuclei.
TIMS-1AbstractP00163 FM200 00Processing Code System for Production of Group Constants of Heavy Resonant Nuclei.
TRANSX 2.15AbstractP00317 MFMWS 01Code system to produce neutron, photon, and particle transport tables for discrete-ordinates and diffusion codes from cross sections in MATXS format.
TRANSX-CTRAbstractP00206 CY000 00Interfaces MATXS Cross-Section Libraries to Nuclear Transport Codes for Fusion Systems Analysis.
UKCTRI-81AbstractD00064 I0370 0146-Group Neutron Cross Sections and Kerma Factors for Fusion Reactor Calculations.
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/KERMAAbstractD00150 I3090 00Broad-Group and Fine-Group and Coupled Neutron/Gamma Cross-Section Libraries Derived from ENDF/B-VII.0 Nuclear Data.
VITENEA-JAbstractD00238 MNYCP 00AMPX 175-n,42-g Multigroup X-section Library for Nuclear Fusion Applications.
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.
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.
XLACS-IIAAbstractP00182 I3033 00A Modified Version of XLACS-II for Processing ENDF Data into Multigroup Neutron Cross Sections in AMPX Master Library Format.
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.