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: CROSS SECTION PROCESSING
Package NameAbstractRSICC TapelistTitle
CCRMNAbstractP00366 MNYCP 00Monte Carlo Simulation of the Coupled Transport of Electrons and Photons.
CEPXS/ONELD 1.0AbstractC00544 MNYCP 02One-Dimensional Coupled Electron-Photon Multigroup Discrete Ordinates Code System.
CNCSN 2009AbstractC00726 PCX86 01One, Two- and Three-Dimensional Coupled Neutral and Charged Particle SN Parallel Multi-Threaded Code System.
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.
DEPLETORAbstractP00523 MNYCP 00Code System to Provide Depletion Capability to the U.S. NRC PARCS Code
DTF-TRACAAbstractC00412 U1100 00Multigroup Neutron Transport Discrete Ordinates Code System with One-Dimensional, Anisotropic Scattering.
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.
GNASH-FKKAbstractP00535 MNYCP 00Pre-equilibrium, Statistical Nuclear-Model Code System for Calculation Cross Sections and Emission Spectra.
INDRAAbstractC00303 I0360 00A Modular System for Calculating the Neutronics and Photonics Characteristics of a Fusion Reactor Blanket.
MARIA SYSTEMAbstractP00359 D6000 00Code System to Calculate Cross Sections for PWR Fuel Assembly Calculations.
MICROX-2AbstractP00374 MNYCP 02Code System to Create Broad-Group Cross Sections with Resonance Interference and Self-Shielding from Fine-Group and Pointwise Cross Sections.
MORSE-CAbstractC00431 C7600 00Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System.
MUP2AbstractP00289 I3090 00A Program to Calculate Fast Neutron Data for Medium-Heavy Nuclei.
NITRANAbstractC00582 FM380 00Neutron Transport Code System Based On Anisotropic Scattering.
NSLINKAbstractP00314 D0VAX 00NJOY SCALE LINK.
PRIMEDANA-2AbstractC00490 I3081 00Collapses Multigroup Cross Sections and Obtains Reaction Parameters by Solving Transport or Diffusion Equations.
RADHEAT-V4AbstractC00300 FM380 00A Code System To Generate Multigroup Constants and Analyze Radiation Transport for Shielding Safety Evaluation.
REX2-87AbstractP00290 D8810 00A Code For Calculating Self-Shielded Multigroup Neutron Cross Sections and Self-Shielding Factors From Preprocessed ENDF/B Basic Data Files.
RHEINAbstractC00585 I3090 00Reactor Code System for Neutron Physics Calculation.
RSYSTAbstractC00269 I0360 00Integrated Modular Code System for Shielding and Reactor Physics Calculations.
SCAT-2AbstractP00294 MNYCP 03Code System for Calculating Total and Elastic Scattering Cross Sections Based on an Optical Model of the Spherical Nucleus.
SELFS-3AbstractP00551 C6600 00Self-Shielding Correlation of Foil Activation Neutron Spectra Analysis by SAND-II.
TALYS-1.2AbstractP00548 PC586 01Nuclear Model Code System for Analysis and Prediction of Nuclear Reactions and Generation of Nuclear Data.
UNFAbstractP00521 PC586 00Code System to Calculate Multistep Compound Nucleus Neutron Cross-Sections and Spectra for Structural Materials.
UNIFY-ECNAbstractP00288 C0170 00A Program to Calculate Fast Neutron Data for Structural Materials.
WIMS-ANL 4.0AbstractC00698 MNYCP 00Deterministic Code System for Reactor Lattice Calculation.
WIMSD-5B.12AbstractC00656 MNYCP 02Deterministic Code System for Reactor Lattice Calculation
X4ECSAbstractP00220 D0780 00A Code System to Combine Cross Section Data in EXFOR and/or ENDF/B-IV Format.
X4RAbstractP00222 DVX11 00Code System for Retrieving EXFOR Cross Section Data According to a Given Target Nucleus.
XSDRNAbstractC00123 C0073 00Multigroup One-Dimensional Discrete Ordinates Spectral Averaging N Transport Code System.
XSDRNAbstractC00123 I0360 00Multigroup One-Dimensional Discrete Ordinates Spectral Averaging N Transport Code System.
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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.