Online Catalog
Click on Package Name to get detailed information.
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
RUGA -- Restricted Use Government Authorized
USSO -- US Distribution Only
USUNV -- US Universities Only
Packages with Keyword: NEUTRON CROSS SECTION PROCESSING
Package NameAbstractRSICC TapelistTitle
1DXAbstractP00096 U1108 00A One-Dimensional Diffusion Code System for Producing Energy Group Collapsed and Self-Shielded Cross Sections.
ADLER IIIAbstractP00058 I0360 00A Program to Calculate Cross Sections from Adler-Adler Resonance Parameters.
AMARAAbstractP00079 I3675 00Nuclear Data Adjustment Using Lagrange's Multipliers Method.
AMPX-77AbstractP00315 ALLMF 01Modular Code System for Generating Coupled Multigroup Neutron-Gamma Libraries from ENDF/B.
AXMIX-PCAbstractP00297 IBMPC 00ANISN Cross Section Mixing Code System.
BREESE-IIAbstractP00143 I3033 00Auxiliary Routines for Implementing the Albedo Option in the MORSE Monte Carlo Code System.
CALENDF-2010
OECD
AbstractP00578 PCX86 00Pointwise, Multigroup Neutron Cross-Sections and Probability Tables from ENDF/B Evaluations.
CODAC (2)AbstractP00073 I0360 00For TIMOC 72, Monte Carlo Three-Dimensional Neutron Transport Code's Data Generator.
COMBINE-PCAbstractP00286 IBMPC 00Code System to Compute Neutron Spectra and ENDF/B Version 5 Based Multigroup Neutron Constants.
EDITORAbstractP00035 I0360 00Alters Mode, Copies, Merges, Punches, Edits, or Adds to ENDF/B-Formatted Data on Tapes or Cards.
ELANAbstractP00141 ICL00 00Neutron Cross-Section Self-Shielding Code System.
ELIESE-3AbstractP00003 I0370 00Analyses of Elastic and Inelastic Scattering Cross Sections.
ENBAL2AbstractP00160 I0370 00A Program to Generate Multigroup Neutron Kerma Factors.
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.
ERIC-2AbstractP00119 I0360 00Calculator of Resonance Integral and Effective Capture and Fission Cross Sections for Fissile and Non-Fissile Nuclides - Thermal or Fast Reactors.
ETHELAbstractP00217 I0360 00Code System for Generating Cross Sections for PSR-128/THERMOS.
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 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.
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.
FOURACESAbstractP00183 I0370 00Code System for Producing Spectrum Weighted, Group Averaged Cross Sections from ENDF/B, KEDAK, or UK Libraries.
GALAXY-6AbstractP00098 I0370 00Neutron Multigroup Cross Section Processor.
GAROLAbstractP00033 I7090 00Calculation of Resonance Neutron Absorption in Two-Region Problems.
GECINXAbstractP00193 H6000 00A Code System for Collapsing Multigroup Cross Sections in CCCC Format.
GERESAbstractP00241 I0370 00A Code to Produce Cross-Section Libraries for ANISN Based on Heterogeneous Fast Reactor Cell Calculations Using MC2II Data.
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.
GIPAbstractP00229 IBMPC 00Group-Organized Cross-Section Input Program.
GLUCSAbstractP00192 D0VAX 00A Generalized Least-Squares Code System for Updating Cross Section Evaluations with Correlated Data Sets.
LEAP-ADDELTAbstractP00138 I0360 00Multigroup Thermal Neutron Scattering Data Generator for Hydrogen in Light Water and Deuterium in Heavy Water.
LIBMAKAbstractP00087 I0360 00ANISN-Type Binary Data Processing Code System.
LSL-M2AbstractP00233 D6220 00Least-Squares Logarithmic Adjustment of Neutron Spectra.
LSL-M2AbstractP00233 IBMPC 00Least-Squares Logarithmic Adjustment of Neutron Spectra.
MARCOPOLOAbstractP00225 I0360 00Code System for Calculating the Radial and Axial Neutron Diffusion Coefficients in One-Group and Multigroup Theory.
MICAPAbstractP00261 I3033 00A Monte Carlo Code System for Analysis of Ionization Chamber Responses.
MIGROS3AbstractP00265 I0370 00A Code for the Generation of Group Constants for Reactor Calculations from Neutron Nuclear Data in KEDAK Format.
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.
MISSIONARYAbstractP00114 I0360 00ENDF/B to NDL Data Format Converter.
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.
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.
PUFF-IVAbstractP00534 MNYCP 01Determination of Multigroup Covariance Matrices from ENDF/B-V Uncertainty Files.
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.
RICEAbstractP00022 I0360 00A Program to Calculate Primary Recoil Atom Spectra from ENDF/B Data.
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.
SAIPSAbstractP00203 E1040 00Information Processing System for Calculating Neutron Spectra from Measured Reaction Rates.
SAIPS-PCAbstractP00295 IBMPC 00Information Processing System for Calculating Neutron Spectra from Measured Reaction Rates.
SATURNAbstractP00057 I3675 00P1 or Transport Corrected Multigroup Neutron Cross Section Data Processor.
SCAT-2AbstractP00294 MNYCP 03Code System for Calculating Total and Elastic Scattering Cross Sections Based on an Optical Model of the Spherical Nucleus.
SLAROMAbstractP00244 FM380 00A Code to Produce Cell Averaged Cross Sections for Fast Critical Assemblies and Fast Power Reactors.
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.
UKE-IIIAbstractP00015 I3691 00Cross Section Format Translator - UKNDL to ENDF/B.
URRAbstractP00281 D6220 00Calculates Resonance Neutron Cross-Section Probability Tables, Bondarenko Self-Shielding Factors and Self-Indication Ratios for Fissile and Fertile Nuclides.
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.