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: GAMMA-RAY CROSS SECTION PROCESSING
Package NameAbstractRSICC TapelistTitle
AMPX-77AbstractP00315 ALLMF 01Modular Code System for Generating Coupled Multigroup Neutron-Gamma Libraries from ENDF/B.
AXMIX-PCAbstractP00297 IBMPC 00ANISN Cross Section Code System.
DATINITAbstractP00258 DGMV1 00Interactive Program To Access Photon Interaction Data.
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.
ENBAL2AbstractP00160 I0370 00A Program to Generate Multigroup Neutron Kerma Factors.
FOURACESAbstractP00183 I0370 00Code System for Producing Spectrum Weighted, Group Averaged Cross Sections from ENDF/B, KEDAK, or UK Libraries.
GAMIDENTAbstractP00154 C0000 00A Program to Aid in the Identification of Unknown Materials by Gamma-ray Spectroscopy.
GAMLEG-75AbstractP00086 C7600 00Multigroup Cross Section Generator for Photon Transport Calculations.
HEITLERAbstractP00004 I7030 00Cross Section Generator.
INFLTBAbstractP00313 ALLCP 00Gamma-Ray Absorption Coefficient Calculation.
LIBMAKAbstractP00087 I0360 00ANISN-Type Binary Data Processing Code System.
MICAPAbstractP00261 I3033 00A Monte Carlo Code System for Analysis of Ionization Chamber Responses.
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.
SATURNAbstractP00057 I3675 00P1 or Transport Corrected Multigroup Neutron Cross Section Data Processor.
STAPREFAbstractP00498 PC586 00Code System to Calculate Nuclear Reaction Cross Sections by Evaporation Model.
STAPRE-H95AbstractP00325 MNYCP 01Code System to Calculate Energy-Averaged Cross Sections of Particle Induced Nuclear Reactions.
TECALCAbstractP00074 DP010 00Interactive Calculation of Compton Coherent and Photoelectric Mass Attenuation Coefficients for Photons (E<1 MeV), and the Mass Absorption Coefficient for Known Materials.
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The Radiation Safety Information Computational Center (RSICC) collects, analyzes, maintains, and distributes software in the areas of radiation transport and safety. RSICC resides in the Nuclear Energy and Fuel Cycle Division (NEFCD) at Oak Ridge National Laboratory.