Online Catalog
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Click on Abstract to read the package abstract.
Click on RSICC Tapelist to view list of files distributed with package.

Note: RESTRICTIONS APPLY TO SOME PACKAGES -
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
Packages with Keyword: ENDF FORMAT
Package NameAbstractRSICC TapelistTitle
ADENAAbstractP00190 C0000 00Code System for Application of Adjusted Data in Calculating Fission-Product Decay Energies and Spectra.
ADENAAbstractP00190 I3033 00Code System for Application of Adjusted Data in Calculating Fission-Product Decay Energies and Spectra.
ADLER IIIAbstractP00058 I0360 00A Program to Calculate Cross Sections from Adler-Adler Resonance Parameters.
AMPX-77AbstractP00315 ALLMF 01Modular Code System for Generating Coupled Multigroup Neutron-Gamma Libraries from ENDF/B.
CALENDF-2010
OECD
AbstractP00578 PCX86 00Pointwise, Multigroup Neutron Cross-Sections and Probability Tables from ENDF/B Evaluations.
CHENDF 7.02AbstractP00333 MNYCP 05Codes for Handling ENDF/B-V and ENDF/B-VI Data.
CODAC (2)AbstractP00073 I0360 00For TIMOC 72, Monte Carlo Three-Dimensional Neutron Transport Code's Data Generator.
COMPLOTAbstractP00259 IBMMF 00Convert EXFOR Format Data to Computation Format and Plot Comparisons of EXFOR and ENDF/B Evaluated Data (Version 86-1).
CRECTJ5AbstractP00250 D0780 00A Computer Program for Compilation of Evaluated Nuclear Data in ENDF/B Format.
EDITORAbstractP00035 I0360 00Alters Mode, Copies, Merges, Punches, Edits, or Adds to ENDF/B-Formatted Data on Tapes or Cards.
ENDLIB-97AbstractD00179 MNYCP 01LLNL Libraries of Atomic Data, Electron Data, and Photon Data in Evaluated Nuclear Data Library (ENDL) Type Format.
ENDVER/GUIAbstractP00572 PCX86 00The ENDF File Verification Support Package.
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.
EVALPLOTAbstractP00211 I3081 00A Program to Plot Data in the Evaluated Nuclear Data File/Version B Format.
F5TABAbstractP00221 D0780 00Code System for Converting Energy Distribution Cross Section Data to Tabulated Data.
FDMXPCAbstractP00322 IPCAT 00Code System for Calculation of Neutron Transmission and Other Functionals from Evaluated Data in ENDF Format.
FEDGROUP-3AbstractP00123 I0360 00Program System for Processing Evaluated Nuclear Data in ENDF/B, KEDAK or UKNDL Format to Constants to be Used in Reactor Physics Calculation.
FEDGROUPC86REV3AbstractP00194 MNYCP 01Code System for Processing Evaluated Nuclear in ENDF/B, KEDAK or UKNDL Formats into Constants for Reactor Physics Calculation.
FIGEROAbstractP00149 C0000 00Processing Codes for Generating Multigroup Neutron Cross Sections from ENDF/B for Use in Discrete Ordinates Calculations.
FORSENAbstractP00170 I0360 00A Multigroup Processing Code for Use with Sensitivity Profiles to Assess the Effect of Cross Section Changes.
GECINXAbstractP00193 H6000 00A Code System for Collapsing Multigroup Cross Sections in CCCC Format.
GLUCSAbstractP00192 D0VAX 00A Generalized Least-Squares Code System for Updating Cross Section Evaluations with Correlated Data Sets.
GROUPXSAbstractP00246 C0740 00Processing of Double-Differential Cross Sections in the New ENDF-VI Format.
MACK-IVAbstractP00132 I3691 00Calculation of Nuclear Response Functions from Nuclear Data in ENDF Format.
MENDL-2PAbstractD00207 MNYCP 00Proton Reaction Data Library for Nuclear Activation (Medium Energy Nuclear Data Library.)
MICROX-2AbstractP00374 MNYCP 02Code System to Create Broad-Group Cross Sections with Resonance Interference and Self-Shielding from Fine-Group and Pointwise Cross Sections.
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.
MISSIONARYAbstractP00114 I0360 00ENDF/B to NDL Data Format Converter.
MIXENAbstractP00318 IRISC 00Code System to Replace Files 4 and 6 of ENDF-6 with Files 4 and 5 of ENDF/B-IV.
NANICKAbstractP00120 I0360 00Infinitely-Diluted Multigroup Cross-Section Generator - from ENDF/B.
NASIF-NARESAbstractP00121 I0360 00A Code System for Computing Shielding Factors from ENDF/B Tapes.
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.
NJOY97.0AbstractP00368 MNYCP 00Code System for Producing Pointwise and Multigroup Neutron and Photon Cross Sections from ENDF/B Data.
NJOY99.0AbstractP00480 MNYCP 00Code 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.
NPTXSAbstractP00090 I0360 00Data Generator: Neutron Point Cross Sections from ENDF/B Resolved and Unresolved Resonance Parameters.
NSLINKAbstractP00314 D0VAX 00NJOY SCALE LINK.
ORPLOT-PCAbstractP00328 PC386 00Plotting Package for Data Evaluation Intercomparison.
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.
RESENDDAbstractP00215 C0740 00A Code System for Reconstruction of Resonance Cross Sections from Evaluated Nuclear Data in ENDF/B Format.
RESENDDAbstractP00215 D0780 00A Code System for Reconstruction of Resonance Cross Sections from Evaluated Nuclear Data in ENDF/B Format.
REX2-87AbstractP00290 D8810 00A Code For Calculating Self-Shielded Multigroup Neutron Cross Sections and Self-Shielding Factors From Preprocessed ENDF/B Basic Data Files.
SAMMY-8AbstractP00158 MNYCP 12Code System for Multilevel R-Matrix Fits to Neutron and Charged-Particle Cross-Section Data Using Bayes' Equations.
STAY'SLAbstractP00113 DP010 00Least Squares Dosimetry Unfolding Code System.
STAYSL PNNLAbstractP00589 PCX86 00STAYSL PNNL Suite of Software Tools.
SUPERTOG III M2AbstractP00013 I3691 00Data Generator--Fine Group Constants and PN Scattering Matrices from ENDF/B.
SUPERTOG-4AbstractP00013 I0360 00Data Generator--Fine Group Constants and PN Scattering Matrices from ENDF/B.
SUPERTOG-LTTAbstractP00228 I0360 00A Modification of PSR-13/SUPERTOG-III Applied to Libraries with Tabulated Elastic Scattering and Anistropy Densities.
UNFAbstractP00521 PC586 00Code System to Calculate Multistep Compound Nucleus Neutron Cross-Sections and Spectra for Structural Materials.
VSOP94AbstractC00670 MNYWS 00Code System for Reactor Physics and Fuel Cycle Simulation.
WINDOWS IIAbstractP00161 I0370 00A Program for the Analysis of Spectral Data Foil-Activation Measurements.
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.