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
USSO -- US Distribution Only
USUNV -- US Universities Only
Packages starting with C
Package NameAbstractRSICC TapelistTitle
CAACAbstractC00476 D0VAX 00Code System for Implementation of Atmospheric Dispersion Assessment Required by the Clear Air Act. See CCC-542/CAP-88.
CAACAbstractC00476 I3033 00Code System for Implementation of Atmospheric Dispersion Assessment Required by the Clear Air Act. See CCC-542/CAP-88.
CACA-2AbstractC00302 I0360 00Heavy Isotope and Fission-Product Concentration Calculation Code System.
CADAbstractD00059 I0360 0051 Neutron, 25 Gamma-Ray Group ALBEDO DATA Generated with DOT for Various Materials.
CADEAbstractP00567 MNYCP 00Multiple Particle Emission Cross-Sections by Weisskopf-Ewing Theory.
CAFDATSAbstractP00549 MNYCP 00Converter of Angular Fluxes of DORT, ANISN and TORT Systems.
CALENDF-2010
OECD
AbstractP00578 PCX86 00Pointwise, Multigroup Neutron Cross-Sections and Probability Tables from ENDF/B Evaluations.
CALKUXAbstractC00594 IBMPC 00Code System to Calculate Exposure Transmission of Medical X-ray Beams Through Barrier Materials.
CALOR95AbstractC00610 MNYWS 00Monte Carlo Code System for Design and Analysis of Calorimeter Systems, Spallation Neutron Source (SNS) Target Systems, etc.
CAMERAAbstractC00240 C0074 00Radiation Transport Analysis Code System and the Computerized Man (CAM) Model.
CAMERAAbstractC00240 IBMPC 01Radiation Transport Analysis Code System and the Computerized Man (CAM) Model.
CANDULIB-AECLAbstractD00210 MNYCP 00Burnup-Dependent ORIGEN-S Cross Section Libraries for CANDU Reactor Fuel Characterization.
CAP-88AbstractC00542 D0VAX 00Clean Air Act Assessment Package.
CAP-88AbstractC00542 I3090 00Clean Air Act Assessment Package.
CAP88-PCAbstractC00542 IBMPC 01Clean Air Act Assessment Package.
CAPS-2AbstractC00074 CDCMF 00Analysis of Structures for Fallout Radiation Shielding.
CARL 2.3AbstractC00743 PC586 01Code System to Calculate Radiotoxicity, Activity, Dose and Decay Power Calculations for Spent Fuel.
CARMEN SYSTEMAbstractC00487 U1110 00A Code System for Neutronics PWR Calculation by Diffusion Theory with Space-Dependent Feedback Effects.
CARNACAbstractC00238 I3691 00Calculation of Flux and Neutron Spectra in the Case of Criticality Accident.
CARP-82AbstractP00131 I3033 00Multigroup Albedo Data Using DOT Angular Flux Results.
CARSTEPAbstractC00024 I7090 00Trajectory and Environment Code-Electron and Proton Fluxes Impinging on Spacecraft in Orbit.
CASCADEAbstractC00176 C6600 00Monte Carlo Simulation of the Transport of High Energy Electrons and Photons in Matter.
CASCADEAbstractC00176 I0360 00Monte Carlo Simulation of the Transport of High Energy Electrons and Photons in Matter.
CASIMAbstractC00265 I0360 00Monte Carlo Simulation of Transport of Hadron Cascades in Bulk Matter.
CASKAbstractD00023 I3691 0422 Neutron, 18 Gamma-Ray Group, P3, Cross Sections for Shipping Cask Analysis.
CASK-81AbstractD00023 IBMPC 0622 Neutron, 18 Gamma-Ray Group, P3, Cross Sections for Shipping Cask Analysis.
CASK-81AbstractD00023 I0370 0522 Neutron, 18 Gamma-Ray Group, P3, Cross Sections for Shipping Cask Analysis.
CASKCODESAbstractP00262 IBMPC 00CAPSIZE, SCOPE, AND KWIKDOSE for Shipping Cask Optimization, Dose Calculation, Parameter Evaluation, and Shielding Requirements.
CASTHYAbstractP00316 FM000 00Statistical Model Calculation for Neutron Cross Sections and Capture Gamma-Ray Spectra.
CAVEATAbstractC00169 I3675 00General Purpose Monte Carlo Time-Dependent Radiation Transport Code in Complex Geometry.
CCRMNAbstractP00366 MNYCP 00Monte Carlo Simulation of the Coupled Transport of Electrons and Photons.
CCVM DATABASE (OCTOBER 2010)AbstractM00016 MNYCP 00CSNI Code Validation Matrix of Thermo-Hydraulic Codes for LWR LOCA and Transients.
CDRAbstractC00182 C6600 00A Constant Dose Range Code System, Using the LANL-NWEF Neutron-Gamma-Ray Air Flux Tape.
CDRAbstractC00182 I0360 00A Constant Dose Range Code System, Using the LANL-NWEF Neutron-Gamma-Ray Air Flux Tape.
CEAR-PPUAbstractP00528 PC586 00Code System for Monte Carlo Simulation of Detector Pulse Pile Up.
CECP-BWRAbstractP00370 PC386 00Estimating Boiling Water Reactor Decomissioning Costs.
CECP-PWRAbstractP00371 PC386 00Estimating Pressurized Water Reactor Decomissioning Costs.
CEM03.03AbstractP00532 MNYCP 01Monte-Carlo Code System to Calculate Nuclear Reactions in the Framework of the Improved Cascade-Exciton Model.
CEMENT 1.02
USSO
AbstractP00412 IBMPC 00Computer Code System for the Estimation of Long-Term Performance of Cement-Based Materials.
CEPXS/ONELD 1.0AbstractC00544 MNYCP 02One-Dimensional Coupled Electron-Photon Multigroup Discrete Ordinates Code System.
CERPI-CERELAbstractP00147 I0360 00Code Systems for Automatic Analysis of Gamma-Ray Spectra Obtained with Ge(Li) Detectors.
CGS 11.4AbstractP00243 MFMWS 03Common Graphics System, Version 11.4.
CHAINS-PCAbstractC00604 IBMPC 00Code System to Compute Atom Density of Members of a Single Decay Chain.
CHAINT-MCAbstractC00584 CYXMP 00A Two-Dimensional Model for the Analysis of Contaminant Transport in a Fractured Porous Medium.
CHARGE IIAbstractC00070 C6500 00Space Radiation Shielding Code - Proton and Electron Penetration of Multilayered Slabs and Spheres.
CHARGE IIAbstractC00070 I3675 00Space Radiation Shielding Code - Proton and Electron Penetration of Multilayered Slabs and Spheres.
CHARGE-PCAbstractC00070 IBMPC 00Space Radiation Shielding Code - Proton and Electron Penetration of Multilayered Slabs and Spheres.
CHENDF 7.02AbstractP00333 MNYCP 05Codes for Handling ENDF/B-V and ENDF/B-VI Data.
CHNSEDAbstractC00671 I0360 00Code System to Model Sediment & Containment Transport.
CINDER 1.05AbstractC00755 PC586 00Code System for Actinide Transmutation Calculations
CITATION-LDI 2AbstractC00643 PC386 02Nuclear Reactor Core Analysis Code System.
CLAW-IVAbstractD00036 I0360 02Coupled 30 Neutron, 12 Gamma-Ray Group Cross Sections Based on ENDF/B-IV for Radiation Transport Calculations.
CLAW-IVAbstractD00036 I3033 03Coupled 30 Neutron, 12 Gamma-Ray Group Cross Sections Based on ENDF/B-IV for Radiation Transport Calculations.
CLEARAbstractD00042 I3691 00126 Neutron, 36 Gamma-Ray Cross Sections in AMPX and CCCC Interface Formats for LMFBR Neutronics Calculations.
CLESAbstractD00233 MNYCP 00Cross Section Library of Moderator Materials for Low-Energy Neutron Sources.
CLOUD-MAbstractC00032 I3565 00Gamma-Ray Dose Rate from a Radioactive Cloud-Kernel Integration Code.
CNCSN 2009AbstractC00726 PCX86 01One, Two- and Three-Dimensional Coupled Neutral and Charged Particle SN Parallel Multi-Threaded Code System.
COAG-IIAbstractP00070 I0360 00Calculation of the Westcott Epithermal Index and the Westcott 2200 m/s Neutron Flux.
COBBAbstractD00016 I3675 01123-Group Neutron Cross Section Data Generated from ENDF/B-II Data for Use in the XSDRN Discrete Ordinates Spectral Averaging Code.
COBRA-3C-RERTR
OECD
AbstractP00606 I0370 00COBRA-3C-RERTR
COBRA4IAbstractP00419 MNYCP 00Code Sytem to Calculate Rod-Bundle and Core Thermal-Hydraulics.
COBRA-ENAbstractP00507 MNYCP 01Thermal-Hydraulic Transient Analysis of Reactor Cores.
COBRA-SFS CYCLE 3AbstractP00472 MNYCP 00Code System for Thermal Hydraulic Analysis of Spent Fuel Casks.
CODAC (2)AbstractP00073 I0360 00For TIMOC 72, Monte Carlo Three-Dimensional Neutron Transport Code's Data Generator.
COG11.1 BETA2AbstractC00777 MNYCP 01Multiparticle Monte Carlo Code System for Shielding and Criticality Use.
COGAPAbstractP00375 MNYCP 01Nuclear Power Plant Containment Hydrogen Control System Evaluation Code.
COHORT-IIAbstractC00198 I7094 00General Purpose Monte Carlo Radiation Transport Code System.
COLLIMATORAbstractC00136 I7090 00Monte Carlo Calculation of the Spectrum of Gamma Radiation from a Collimated Co-60 Source.
COLUMN2AbstractC00534 ALLMF 00Calculation of Effects of Physicochemical Processes on Migration.
COMANDAbstractP00091 I0360 00A Multigroup ANISN Cross Section Data Library Collapsing Code System.
COMBINE-PCAbstractP00286 IBMPC 00Code System to Compute Neutron Spectra and ENDF/B Version 5 Based Multigroup Neutron Constants.
COMIDAAbstractP00343 MNYCP 00Radionuclide Food Chain Model for Acute Fallout Deposition.
COMMIX-1B
USSO
AbstractP00393 DVX11 003-D Single-Phase Thermal Hydraulics
COMMIX-1B
USSO
AbstractP00393 I3033 003-D Single-Phase Thermal Hydraulics
COMMIX-1C
USSO
AbstractP00393 MNYCP 003-D Single-Phase Thermal Hydraulics
COMNUC3BAbstractP00302 CYXMP 00A Compound Nucleus Analysis Program.
COMPARAbstractP00240 C0170 00Compares Multigroup Cross Sections Generated by NJOY, GROUPIE, FLANGE-II, ETOG-3 and XLACS.
COMPARE-MOD1AAbstractP00410 C7600 00Transient Flow W/Sinks & Doors
COMPARE-MOD1AAbstractP00410 I3033 00Code System to Calculate Transient Flow With Heat Sinks & Doors.
COMPASS 1.0.0AbstractP00520 PC586 00Computerization of MARSSIM for Planning and Assessing Site Surveys.
COMPBRN3AbstractP00389 PC386 00Code System for Modeling Compartment Fires.
COMPLOTAbstractP00259 IBMMF 00Convert EXFOR Format Data to Computation Format and Plot Comparisons of EXFOR and ENDF/B Evaluated Data (Version 86-1).
COMPRASHAbstractC00072 I3675 00Spinney Removal-Diffusion Shielding Code.
COMRADEX4AbstractC00332 I0360 00Evaluator of Potential Radiological Doses in the Near (< 10 km) Environment of Radioactive Release.
CONDOR-3AbstractC00811 I0370 00Two-Dimensional Reactor Program with Local and Spectrum Dependent Burnup.
CONDOS-IIAbstractC00416 I0360 00Code for Estimating Radiation Doses from Radionuclide-Containing Consumer Products.
CONFOLDAbstractP00053 C6600 00Least-Structure Unfolding Code System for Measured Neutron and Gamma-Ray Spectra.
CONFOLDAbstractP00053 I0360 00Least-Structure Unfolding Code System for Measured Neutron and Gamma-Ray Spectra.
CONSTRIP VAbstractC00139 I3675 00Vertical Barrier-Finite Source Plane Gamma-Ray Penetration Code System.
CONTEMPT4AbstractP00397 MNYCP 00Code System for PWR & BWR Multicompartment Containment Analysis, Versions MOD5 & MOD6.
CONTEMPT-LT28B
USSO
AbstractP00387 C7600 00Code System to Predict Containment Pressure-Temperature Response To a Loss-Of-Coolant Accident.
CONVERTAbstractP00036 C6600 00An IBM-to-CDC Program Conversion Code.
COOL-CAbstractP00017 I0360 00Spectra Unfolding Codes.
CORTESAbstractP00404 I0360 00Code System for Thermal & Mechanical Analysis of Tees.
COV-15GROUP-2006AbstractD00232 MNYCP 0015-Group Cross Section Covariance Matrix Library.
COVERVAbstractD00077 I0360 01Compilation of Multigroup Cross-section Covariance Matrices in COVERX Format for Several Important Materials (Generated from ENDF/B-V Data using PSR-093/PUFF2).
COVERXAbstractD00044 I0360 02Compilation of Multigroup Cross-Section Covariance Matrices in COVERX Format for Several Important Materials.
COVFILSAbstractD00091 I0360 00A 30-Group Covariance Library Based on ENDF/B-V.
COVFILS-2AbstractD00137 ALLCP 00Neutron Data and Covariances for Sensitivity and Uncertainty Analysis.
CRAC2AbstractC00419 C0000 00Code System for Calculating Reactor Accident Consequences.
CRAC2AbstractC00419 I3033 00Code System for Calculating Reactor Accident Consequences.
CRECTJ5AbstractP00250 D0780 00A Computer Program for Compilation of Evaluated Nuclear Data in ENDF/B Format.
CRESOAbstractP00184 I3081 00Resonance Data-Handling Code System.
CRRISAbstractC00518 I3033 00Computerized Radiological Risk Investigation System for Assessing Doses and Health Risks from Atmospheric Releases of Radionuclides.
CRRISAbstractC00518 PC586 00Computerized Radiological Risk Investigation System for Assessing Doses and Health Risks from Atmospheric Releases of Radionuclides.
CRYO-S(A,B)-ACE1AbstractD00253 MNYCP 00Scattering Law and Continuous Energy Cross Section Library of Materials at Cryogenic Temperatures.
CRYSTAL BALLAbstractC00233 C6600 00Code System for Determining Neutron Spectra from Activation Measurements.
CRYSTAL BALLAbstractC00233 I0360 00Code System for Determining Neutron Spectra from Activation Measurements.
CTR DATAAbstractD00028 I3675 0173-Group P3 Coupled Neutron and Gamma-Ray Cross Sections for Fusion Reactor Calculations.
CUPEDAbstractP00032 I3675 00Scintillation Spectrometer Polyenergetic Gamma Photon Experimental Distributions Unfolding Code.
CYGASAbstractC00317 I3033 00A Gamma-Ray Attenuation Code System for Large Gamma-Ray Sources Shielded by Coaxial Cylinders.
CYGNUS-C SPHEREAbstractC00232 I0360 00Monte Carlo Neutron Transport Code System in Spherical Geometry.
CYLDOSAbstractC00389 I0360 00A Cylindrical Geometry Gamma-Ray Flux Attenuation Code System.
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.