Online Catalog
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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: WORKSTATION
Package NameAbstractRSICC TapelistTitle
ANISN-ORNLAbstractC00254 MNYCP 02One-Dimensional Discrete Ordinates Transport Code System with Anisotropic Scattering.
BETA-S 6AbstractC00657 MNYCP 01Code System to Calculate Multigroup Beta-Ray Spectra.
BISON-CAbstractC00659 MNYWS 00One-Dimensional Discrete Ordinate Transport and Burnup Calculation Code System.
BOT3P-5.3AbstractP00530 MNYCP 02Code System for 2D and 3D Mesh Generation and Graphical Display of Geometry and Results for Radiation Transport Codes.
BSPRP2AbstractP00372 IRISC 00Code System to Process DORT Boundary-Flux Files.
CALOR95AbstractC00610 MNYWS 00Monte Carlo Code System for Design and Analysis of Calorimeter Systems, Spallation Neutron Source (SNS) Target Systems, etc.
CCRMNAbstractP00366 MNYCP 00Monte Carlo Simulation of the Coupled Transport of Electrons and Photons.
CGS 11.4AbstractP00243 MFMWS 03Common Graphics System.
CHENDF 7.02AbstractP00333 MNYCP 05Codes for Handling ENDF/B-V and ENDF/B-VI Data.
DANTSYS 3.0AbstractC00547 MFMWS 01One-, Two-, and Three-Dimensional, Multigroup, Discrete-Ordinates Transport Code System.
DCHAIN 1.3AbstractC00640 MNYCP 01Code System for Radioactive Decay and Reaction Chain Calculations.
DETAN 95AbstractP00361 MNYCP 00Code System to Calculate Spectrum-Averaged Cross Sections and Detector Responses in Neutron Spectra.
DOORS 3.2AAbstractC00650 MFMWS 04One, Two- and Three-Dimensional Discrete Ordinates Neutron/Photon Transport Code System.
DRAGON3.05DAbstractC00647 MNYWS 03Lattice Cell Code System.
EGS4AbstractC00331 MNYCP 00Monte Carlo Simulation of the Coupled Transport of Electrons and Photons.
EXPRESSAbstractC00622 MNYCP 00Exact Preparedness Supporting System.
GBANISNAbstractC00628 IRISC 00One-Dimensional Discrete Ordinates Transport Code System with Anisotropic Scattering with the GroupBand Option.
GICX40AbstractD00092 ALLCP 00Coupled 42-Neutron, 21-Gamma-Ray Group Cross Sections for 40 Elements in Group Independent Form for Fusion Reactor Calculations.
GMAAbstractP00367 MNYCP 00Code System for Calculation of Reactor Accident Consequences.
GNASH-FKKAbstractP00535 MNYCP 00Pre-equilibrium, Statistical Nuclear-Model Code System for Calculation Cross Sections and Emission Spectra.
GRESS 3.0AbstractP00231 MFMWS 02Gradient Enhanced Software System.
HEATING 7.3AbstractP00199 MNYCP 06Multidimensional, Finite-Difference Heat Conduction Analysis Code System.
LAHET 2.8AbstractC00696 MFMWS 00Code System for High Energy Particle Transport Calculations.
LEPRICONAbstractP00277 I3033 01PWR Pressure Vessel Surveillance Dosimetry Analysis System.
LEPRICONAbstractP00277 IRISC 00PWR Pressure Vessel Surveillance Dosimetry Analysis System.
MARLOWE 15BAbstractP00137 MNYCP 08Computer Simulation of Atomic Collisions in Crystalline Solids.
MCNP-DSP
810
AbstractC00699 MNYCP 00Monte Carlo N-Particle Transport Code System with Digital Signal Processing based on MCNP4A.
MCNP-DSP-EXE
810
AbstractC00699 MNYCP 01Monte Carlo N-Particle Transport Code System with Digital Signal Processing based on MCNP4A.
MOCUPAbstractP00365 DALPU 00MCNP/ORIGEN Coupling Utility Programs.
MONTEBURNS 2.0AbstractP00455 MNYCP 02Automated, Multi-Step Monte Carlo Burnup Code System.
MORSE-CGAAbstractC00474 ALLCP 03Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System.
MSM-SOURCEAbstractP00369 MNYCP 00Code System for Generation of Input Data for MCNP.
NESTLE 5.2.1AbstractC00641 MNYCP 04Code System to Solve the Few-Group Neutron Diffusion Equation Utilizing the Nodal Expansion Method (NEM) for Eigenvalue, Adjoint, and Fixed-Source
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.
PARET-ANLAbstractP00516 MNYCP 00Code System to Predict Consequences of Nondestructive Accidents in Research and Test Reactor Cores.
PENELOPE-MPIAbstractC00713 IBMSP 00Code System for Monte Carlo Simulation of Electron and Photon Transport.
PREPRO2019AbstractP00351 MNYCP 10Pre-Processing Code System for Data in ENDF/B Format.
QAD-CGGP-AAbstractC00645 MNYCP 00Kernel Integration Code System.
RACC-PULSEAbstractC00639 MNYWS 00RACC Code System for Computing Radioactivity-Related Parameters for Fusion Reactor Systems Modified for Pulsed/Intermittent Activation Analysis.
REAC*3AbstractC00443 IBMPC 00Computer Code System for Activation and Transmutation.
REAC*3AbstractC00443 MFMWS 00Computer Code System for Activation and Transmutation.
REBUS 11.0 EXE_ONLY
FEDC
AbstractC00822 MNYWS 01Code System for Analysis of Fast Reactor Fuel Cycles.
REBUS 11.2892
FEDC
AbstractC00822 MNYCP 02Code System for Analysis of Fast Reactor Fuel Cycles.
REBUS3/VARIANT8AbstractC00653 MNYWS 01Code System for Analysis of Fast Reactor Fuel Cycles.
ROLAIDS-CPMAbstractP00353 SUN04 00Code System to Calculate Group-Averaged Cross Sections Using the Collision Probability Method.
SAND-II-SNLAbstractP00345 SUN04 00Neutron Flux Spectra Determination by Multiple Foil Activation Method.
SCAMPIAbstractP00352 MNYWS 01Collection of Codes for Manipulating Multigroup Cross Section Libraries in AMPX Format.
SCRELAAbstractP00408 SUN05 00Code System for Supercritical Water Cooled Reactor LOCA Analysis.
SOURCES-4CAbstractC00661 MNYCP 04Code System for Calculating (alpha,n), Spontaneous Fission, and Delayed Neutron Sources and Spectra.
TART2016AbstractC00638 MNYCP 08Coupled Neutron-Photon, 3-D, Combinatorial Geometry, Time Dependent, Monte Carlo Transport Code System.
TIBSOAbstractC00512 MNYCP 00Code System to Calculate Production and Migration of Radionuclides in Nuclear Reactor Systems.
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.
VALE 1.1AbstractC00613 IRISC 01A Multigroup Diffusion Theory Neutronics Code System for Solving Two- and Three-Dimensional Problems for Triagonal Geometries.
VALE 1.1AbstractC00613 PC386 01A Multigroup Diffusion Theory Neutronics Code System for Solving Two- and Three-Dimensional Problems for Triagonal Geometries.
VIM 5.1AbstractC00754 MNYWS 01Continuous Energy Neutron and Gamma-ray Transport Code System.
WIMKAL-88AbstractD00193 MNYCP 0069 Energy Group, Neutron Cross Section Library For Thermal Reactor Calculations in WIMSD Format.
WIMSD-5B.12AbstractC00656 MNYCP 02Deterministic Code System for Reactor Lattice Calculation
ZOTT99AbstractP00272 ALLCP 02Zero-in On The Truth; Evaluation of Correlated Data Using Partitioned Least Squares.
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.