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 S
Package NameAbstractRSICC TapelistTitle
S1CALCAbstractP00134 I0360 00A Multigroup Thermal Neutron Scattering Law Data Generator for Hydrogen and Deuterium.
S3AbstractC00322 C6600 00Kernel Integration Code System--Multigroup Gamma-Ray Scattering.
S3AbstractC00322 DVX11 00Kernel Integration Code System--Multigroup Gamma-Ray Scattering.
S3AbstractC00322 IBMPC 00Kernel Integration Code System--Multigroup Gamma-Ray Scattering.
SABINE-3AbstractC00121 C7600 00Spinney (Removal-Diffusion) Shielding Code System in One-Dimensional Geometry.
SABINE-3AbstractC00121 I0370 00Spinney (Removal-Diffusion) Shielding Code System in One-Dimensional Geometry.
SABINE-3AbstractC00121 U1106 00Spinney (Removal-Diffusion) Shielding Code System in One-Dimensional Geometry.
SABINE-PCAbstractC00121 IBMPC 00Spinney (Removal-Diffusion) Shielding Code System in One-Dimensional Geometry.
SACALC3AbstractC00802 PCX86 00Calculates the Average Solid Angle Subtended by a Volume.
SACHETAbstractC00571 D8810 00A Computer Program To Evaluate The Dynamic Fission Product Inventories in the Multiple Compartment System of PWR's.
SAEROSAAbstractP00573 MNYCP 00Single-Species Aerosol Coagulation and Deposition with Arbitrary Size Resolution.
SAFE-D/SAFE-RAbstractP00496 MNYCP 00Code System for the Analysis of Component Failure Data with a Compound Statistical Model.
SAHYB-2AbstractC00820 I0360 00Solution of Ordinary Differential Equation with User-Supplied Subroutine
SAILAbstractD00057 I0360 0023 Neutron, 17 Gamma-Ray Group ALBEDO DATA for Concrete and Steel, Based on DOT 1-1/2-D Calculations using DLC-31/FEWG1 Data.
SAILORAbstractD00076 I3033 00Coupled, Self-Shielded, 47 Neutron, 20 Gamma-Ray, P3, Cross Section Library for Light Water Reactors. We recommend ENDF/B-VI derived data in DLC-185/BUGLE-96.
SAILORAbstractD00076 PC386 01Coupled, Self-Shielded, 47 Neutron, 20 Gamma-Ray, P3, Cross Section Library for Light Water Reactors. We recommend ENDF/B-VI derived data in DLC-185/BUGLE-96.
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.
SALE3DAbstractP00443 CY000 00ICEd-ALE Treatment of 3-D Fluid Flow.
SAM-CEAbstractC00187 C6600 00Monte Carlo Time-Dependent Complex Geometry (Combinatorial) Code System for the Solution of the Forward Neutron and Forward and Adjoint Gamma-Ray Transport Equations.
SAM-CEAbstractC00187 I0360 00Monte Carlo Time-Dependent Complex Geometry (Combinatorial) Code System for the Solution of the Forward Neutron and Forward and Adjoint Gamma-Ray Transport Equations.
SAM-CEPAbstractC00192 C6600 00Monte Carlo Code System Correlated to the Simultaneous Solution of Multiple, Perturbed, Time-Dependent Neutron Transport Problems in Complex Three-Dimensional Geometry.
SAMCRAbstractP00487 U1100 00Code System for 2-D Elastodynamic Fracture Analysis.
SAMMY-8AbstractP00158 MNYCP 12Code System for Multilevel R-Matrix Fits to Neutron and Charged-Particle Cross-Section Data Using Bayes' Equations.
SAMPO80AbstractP00204 DGNOV 00Gamma-Ray Spectrum Analysis Method for Minicomputers.
SAMPO-LRCAbstractP00186 C6600 00Gamma-Ray Spectrum Analysis Code.
SAMSYAbstractC00315 C0073 00A One-Dimensional Multilayer Multigroup Neutron Removal-Diffusion and Gamma-Ray Point Kernel Calculator.
SAND-IIAbstractC00112 MNYCP 03Neutron Flux Spectra Determination by Multiple Foil Activation Method. We recommend PSR-345/SNL-SAND-II.
SAND-II-SNLAbstractP00345 SUN04 00Neutron Flux Spectra Determination by Multiple Foil Activation - Iterative Method.
SANDORAbstractC00364 C7600 00Isotope Generation and Depletion Code Matrix Exponential Method.
SANDYLAbstractC00361 C0000 00A Monte Carlo Three-Dimensional Code System for Calculating Combined Photon-Electron Transport in Complex Systems.
SAP N-GAbstractC00092 I7094 00Neutron and Gamma-Ray Albedo Model Scatter Shield Analysis Code System.
SAPHIRE 7.27AbstractP00405 PCX86 05Systems Analysis Programs for Hand-On Integrated Reliability Evaluations.
SARA 4.16
USSO
AbstractP00484 IBMPC 00System Analysis and Risk Assessment System.
SATURNAbstractP00057 I3675 00P1 or Transport Corrected Multigroup Neutron Cross Section Data Processor.
SC2N3NAbstractP00309 D0VAX 00Systematics of (n,2n) and (n,3n) Cross Sections.
SCALE 6.1AbstractC00785 MNYCP 00A Comprehensive Modeling and Simulation Suite for Nuclear Safety Analysis and Design; Includes ORIGEN (Source & Executables).
SCALE 6.1-EXEAbstractC00785 MNYCP 01A Comprehensive Modeling and Simulation Suite for Nuclear Safety Analysis and Design; Includes ORIGEN (Executables - No Source).
SCAMPIAbstractP00352 MNYWS 01SCAMPI: Collection of Codes for Manipulating Multigroup Cross Section Libraries in AMPX Format.
SCANSAbstractP00029 I3675 00Spectra Calculation from Activated Nuclide Sets.
SCANS 1AAbstractP00373 PC386 01Shipping Cask Design Review Analysis.
SCAP-82AbstractC00418 C7600 00Single Scatter, Albedo Scatter, or Point Kernel Analysis Code System in Complex Geometry.
SCAT-2AbstractP00294 MNYCP 03Code System for Calculating Total and Elastic Scattering Cross Sections Based on an Optical Model of the Spherical Nucleus, Versions SCAT-2 and SCAT-2B.
SCDAP/RELAP5/MOD3.3
810
AbstractP00581 MNYCP 00A Best-Estimate Transient Simulation of Light Water Reactor Coolant Systems During a Severe Accident.
SCDAP/RELAP5/MOD3.3-EXE
810
AbstractP00581 MNYCP 01A Best-Estimate Transient Simulation of Light Water Reactor Coolant Systems During a Severe Accident.
SCEPTRE 1.1
FEDC
AbstractC00807 PCX86 00Sandia Computational Engine for Particle Transport for Radiation Effects
SCINFULAbstractP00267 CY0MP 00Scintillator Full Response to Neutron Detection.
SCINFULAbstractP00267 D8600 00Scintillator Full Response to Neutron Detection.
SCIP V1.1AbstractC00749 PCX86 00Radioactive Surface Contamination Investigation Program.
SCOPEAbstractP00210 I3033 00Computer Code System for Shipping Cask Optimization and Parametric Evaluation.
SCORCH-B2AbstractP00601 I0370 00BWR Core Heating During LOCA.
SCORE-4AbstractC00234 I0370 00Two-Dimensional Multigroup Removal-Diffusion Shielding Code System.
SCORE-EVETAbstractP00442 C7600 00Code System for Three-Dimensional Hydraulic Reactor Core Analysis.
SCRELAAbstractP00408 SUN05 00Code System for Supercritical Water Cooled Reactor LOCA Analysis.
SDCAbstractC00060 I3675 00Kernel Integration Shield Design Code for Radioactive Fuel Handling Facilities.
SECAAbstractP00104 I0360 00Evaluator of Angular Bounds for a Two-Dimensional Symmetric Gaussian Quadrature Set.
SEDONEAbstractC00345 I0360 00A Simulator of Tidal Transient Hydrodynamic Sediment Concentrations Conditions in Controlled Rivers and Estuaries.
SEISIM1AbstractP00453 C7600 00Code System for Seismic Probabilistic Risk Assessment.
SELFS-3AbstractP00551 C6600 00Self-Shielding Correlation of Foil Activation Neutron Spectra Analysis by SAND-II.
SENPROAbstractD00045 I3691 02Compilation of Multigroup Sensitivity Profiles in SENPRO Format for Fast Reactor Core and Shield Benchmarks and Thermal Reactor Benchmarks.
SENSITAbstractC00405 C7600 00One-Dimensional, Multigroup Cross Section and Design Sensitivity and Uncertainty Analysis Code System - Generalized Perturbation Theory.
SERA-1C1AbstractC00729 MNYCP 01Simulation Environment for Radiotherapy Applications.
SERPENT
OECD
AbstractC00757 MNYWS 00Continuous Energy Monte Carlo Reactor Physics Burnup Calculation Code.
SERPENT117-ACELIBAbstractD00249 MNYCP 00Continuous-Energy X-Sec Library, Radioactive Decay, Fission Yield Data for SERPENT in ACE.
SESOILAbstractC00629 IBMPC 03Code System to Calculate One-Dimensional Vertical Transport for the Unsaturated Soil Zone.
SETSAbstractP00380 CDCMF 00Set Equation Transformation System.
SFACTORAbstractC00310 I0360 00Dose Equivalent to a Target Organ Calculator.
SFAKAbstractC00437 I3033 00Code System for Calculation of the Self-Absorption of Unscattered Gamma Radiation from Fuel Assemblies.
SFHA
USSO
AbstractP00413 IBMPC 00Code System for Spent Fuel Heating Analysis.
SHADOKAbstractC00216 C6600 00Transport Code Systems, P1 Scattering in Infinite Cylindrical and Spherical Geometries by Polynomial Approximation.
SHADRAC(G-30)AbstractC00084 I7090 00Kernel Integration Code - Shield Heating and Dose Rate Calculation in Complex Geometry.
SHAMSIAbstractD00135 I3033 0048 Group Cross-Section Library for Fusion Nucleonics Analysis.
SHARDAAbstractC00521 C0740 00Sample Heat, Activity, Reactivity, and Dose Analysis for Safety Analysis of Irradiations in a Research Reactor.
SHC
USSO
AbstractP00493 CY000 00Seismic/Hazard Characterization in the Eastern U.S.
SHIELDAbstractC00667 MNYCP 01Monte-Carlo Code for Simulating Interaction of High Energy Hadrons with Complex Macroscopic Targets.
SHIELDOSEAbstractC00379 ALLMF 00Code System for Space Shielding Radiation Dose Calculations.
SHIELDOSE-PCAbstractC00379 IBMPC 00Code System for Space Shielding Radiation Dose Calculations.
SHREDIAbstractC00284 I0360 00Multigroup Two-Dimensional (x-y, r-o geometry) Neutron Removal-Diffusion (Spinney Method) Shielding Code System.
SIGMA IIAbstractC00118 C6000 00Space Radiation Dose Analysis Within Complex Configurations.
SIGMA IIAbstractC00118 PC486 00Space Radiation Dose Analysis Within Complex Configurations.
SIGMA-AAbstractD00139 ALLMF 00Photon Interaction and Absorption Cross Sections.
SIGMA-AAbstractD00139 IBMPC 00Photon Interaction and Absorption Cross Sections.
SIGPIAbstractP00475 D0785 00Fault Tree Cut Set System Performance.
SIMMER II
USSO
AbstractC00691 MFMWS 00Code System for Two-Dinensional Sn-Neutronics and Fluid Dynamics.
SINBAD SEARCH TOOLAbstractP00580 MNYCP 00SINBAD Search Tool
SINBAD2013.12AbstractD00237 MNYCP 03Shielding Integral Benchmark Archive and Database, Version December 2013
SIOBAbstractP00139 I0360 00Calculation of Least-Squares Shape Fitting Several Neutron Transmission Measurements Using the Breit-Wigner Multilevel Formula.
SIR-3AbstractP00055 C6400 00Sievert's Integral Routine-Computer Evaluation.
SIR-3AbstractP00055 I3675 00Sievert's Integral Routine-Computer Evaluation.
SIXTUS-3AbstractC00609 MFMWS 00Three-Dimensional, Nodal, Neutron Diffusion Criticality Code System in Hex-Z Geometry.
SKETCH-N 1.0AbstractC00808 MNYCP 00Solve Neutron Diffusion Equations of Steady-State and Kinetics Problems.
SKEWGAUSAbstractP00089 I0360 00Skewed-Gaussian Line Peak Fitting Code - Multichannel Analyzer (MCA) Spectra - Ge(Li) and Semiconductor Detectors.
SKYDATA-KSUAbstractD00188 IBMPC 00Parameters for Approximate Neutron and Gamma-Ray Skyshine Response Functions and Ground Correction Factors.
SKYIII-PCAbstractC00289 IBMPC 01Calculation of the Effects of Structure Design on Neutron, Primary Gamma-Ray and Secondary Gamma-Ray Dose Rates in Air.
SKYPORTAbstractD00093 IBMPC 00Skyshine Importance Functions for Neutrons and Gamma Rays.
SKYSHINE-IIIAbstractC00289 D0VAX 00Calculation of the Effects of Structure Design on Neutron, Primary Gamma-Ray and Secondary Gamma-Ray Dose Rates in Air.
SKYSHINE-KSUAbstractC00646 IBMPC 03Code System to Calculate Neutron and Gamma-Ray Skyshine Doses Using the Integral Line-Beam Method.
SLAROMAbstractP00244 FM380 00A Code to Produce Cell Averaged Cross Sections for Fast Critical Assemblies and Fast Power Reactors.
SLDNAbstractC00221 A1000 00Code System for Shielding Calculations by the Method of Invariant Imbedding.
SLDNAbstractC00221 F2307 00Code System for Shielding Calculations by the Method of Invariant Imbedding.
SLDNAbstractC00221 FM200 00Code System for Shielding Calculations by the Method of Invariant Imbedding.
SLDNAbstractC00221 GE625 00Code System for Shielding Calculations by the Method of Invariant Imbedding.
SLDNAbstractC00221 I0360 00Code System for Shielding Calculations by the Method of Invariant Imbedding.
SLIDERULE 1.0AbstractC00704 PC586 01Nuclear Criticality Slide Rule.
SMACSAbstractP00396 C7600 01Probabilistic Seismic Analysis Code System.
SMAFSAbstractP00547 PC586 00Steady-State Analysis Model for Advanced Fuel Cycle Schemes.
SMARTAbstractC00602 ALLCP 00Code System for Calculating Early Offsite Consequences from Nuclear Reactor Accidents.
SMART/MANYCASKAbstractC00482 FM200 00A Program for Calculating Radiation Dose Rates.
SMAUG-13AbstractC00194 C6600 00Calculation of Neutron and Prompt Gamma-Ray Doses Resulting from an Atmospheric Nuclear Detonation.
SMOGAbstractP00216 I3033 00Code System for Neutron Cross Section Evaluation (Optical Method).
SNAKEAbstractP00135 I0360 00A Solid Angle Calculational System.
SNAP-3DAbstractC00434 MNYCP 01Multigroup Complex Geometry Neutron Diffusion Code System.
SNEXAbstractC00353 C0000 00A One-Dimensional Single Group Discrete Ordinates Transport Code System.
SNLRMLAbstractD00178 ALLCP 00Recommended Dosimetry Cross Section Compendium.
SNOWAbstractC00282 I0360 00Two-Dimensional Discrete Ordinates Multigroup Transport Code System in Plane and Cylindrical Geometry with Isotropic and Anisotropic Scattering.
SOFIPAbstractC00358 I3033 00Evaluator of Space Radiation Environment Encountered by Geocentric Satellites.
SOFIRE-2AbstractP00570 I0370 00Containment Temperature and Pressure During Na Pool Fire, 1-Cell or 2 Cell.
SOLA-DFAbstractP00454 C7600 00Code System to Calculate Transient 2-Dimensional 2-Phase Flow.
SOLA-LOOPAbstractP00464 C7600 00Nonequilibrium, Drift-Flux Code System for Two-Phase Flow Network Analysis
SOLTRANAbstractC00763 PCX86 00Solving Multi-Dimensional Simplified P2 Transport and Diffusion Problems of Hexagonal Geometry in Fast Reactors.
SORAAbstractP00174 I0360 00A Code System for Storage and Retrieval of Data from Radionuclide Analyses.
SOSUMAbstractC00109 I3675 00Multigroup Beta and Gamma-Ray Energy Sources from Activities.
SOURCES-4CAbstractC00661 MNYCP 04Code System for Calculating Alpha, N; Spontaneous Fission; and Delayed Neutron Sources and Spectra.
SPACETRAN 1;2;3AbstractC00120 I3675 00Dose Calculations at Detectors at Various Distances from the Surface of a Cylinder.
SPARAbstractC00228 C6600 00Calculation of Stopping Powers and Ranges for Muons, Charged Pions, Protons and Heavy Ions.
SPARAbstractC00228 I0360 00Calculation of Stopping Powers and Ranges for Muons, Charged Pions, Protons and Heavy Ions.
SPARESAbstractC00148 I3675 00Space Radiation Environment and Shielding Code System.
SPEC-4AbstractP00099 I0360 00Calculated Recoil Proton Energy Distributions from Monoenergetic and Continuous Spectrum Neutrons.
SPECTERAbstractP00023 I3565 00Calculation of Energy Distribution of Nuclear Reaction Products.
SPECTER-ANLAbstractP00263 D0VAX 00Neutron Damage Calculations for Materials Irradiations.
SPECTRAAbstractC00108 C0000 00Determination of Neutron Spectra from Activation.
SPECTRAAbstractC00108 C0073 00Determination of Neutron Spectra from Activation.
SPECTRAAbstractC00108 C3600 00Determination of Neutron Spectra from Activation.
SPECTRANS-2AbstractP00071 ICL00 00Neutron Spectrum Library Generation.
SPEEDIAbstractC00507 FM180 00Code System for Real-Time Prediction of Radiation Dose to the Public Due to an Accidental Release from a Nuclear Power Plant.
SPESAbstractP00602 I0370 00Fuel Cycle Optimization for LWR.
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.
SPIRT
USSO
AbstractP00476 C7600 00Code System to Calculate Stress-Strains from Transient Pressures.
SPIRT-NRC
USSO
AbstractP00198 I3033 01Computerized Mathematical Models of Spray Washout of Airborne Contaminants (Radioactivity) in Containment Vessels.
SPOORAbstractC00278 C7600 00Monte Carlo Simulation of the Turbulent Transport of Airborne Contaminants.
SPOT1AbstractC00460 I3033 00Shielding Problem Code Based on Methods of Ono and Tsuruo.
SPUNITAbstractP00266 D8600 00Spectrum Unfolding Using Information Theory.
SQUIRT VER2
USSO
AbstractP00583 PCX86 00Code System to Predict Leakage Rate and Area of Crack Opening for Cracked Pipes in Nuclear Power Plants.
SRAC95AbstractC00716 MNYWS 00Thermal Reactor Code System for Reactor Design and Analysis.
SRNA-2K5AbstractC00789 PCX86 00Proton Transport Simulation by Monte Carlo Techniques.
SRVAL
USSO
AbstractP00467 I3033 00Stock-Recruitment Model Validation Code System.
SSC-L V3.3
USSO
AbstractP00400 I3090 00Transient Response in LMFBR System.
STABA,STAGT,STEGT,STIG,STIGMAAbstractP00575 MNYCP 00Stress Analysis of Dragon HTR Graphite Structure.
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.
STAR CODESAbstractP00330 IBMPC 00Code System for Calculating Stopping-Power and Range Tables for Electrons, Protons, and Helium Ions.
STAX-2AbstractC00821 I0360 00Neutron Scattering Cross-Sections by Optical Model and Moldauer Theory with Hauser-Feshbach.
STAY'SLAbstractP00113 DP010 00Least Squares Dosimetry Unfolding Code System.
STAYSL PNNLAbstractP00589 PCX86 00STAYSL PNNL Suite of Software Tools.
STERNOAbstractC00057 C0000 00Two Dimensional Gamma-Ray Heating Kernel Integration Code.
STEX IIAbstractM00010 MNYCP 00International Steam Explosion Experimental Data Base.
STOPOW88AbstractC00790 MNYCP 00Stopping Power of Fast Ions in Matter.
STORMAbstractC00067 I7090 00Solar Flare Radiation Hazard to Earth Orbiting Vehicles.
STORM-ISRAELAbstractD00015 I0360 01Evaluated Photon Interaction Library, ENDF/B File 23 Format.
STRADEAbstractP00252 I3081 00Stratified Random Design.
STRAGLAbstractC00201 C6600 00Calculation of Energy Loss Straggling of Heavy Charged Particles.
STRAINTAbstractC00259 I0360 00One-Dimensional Multigroup Neutron Transport Discrete Ordinates Code System.
STREAMAbstractC00321 C7600 00A Three-Dimensional Cylindrical-Geometry Monte Carlo Ray Tracing Code for Computing Light Transmission.
SUBDOSA-IIAbstractC00270 U1100 00Calculation of External Gamma-Ray and Beta-Ray Doses from Accidental Atmospheric Releases of Radionuclides.
SUGGELAbstractP00508 MNYWS 00Program Suggesting the Orbital Angular Momentum of a Neutron Resonance From the Magnitude Of Its Neutron Width.
SULSAAbstractM00015 MNYCP 00A Solution for the Neutron Spectrum Unfolding Problem Without Using Input Spectrum (Report Only).
SUPERDAN-PCAbstractP00282 IBMPC 00Calculates Dancoff Factor of Spheres, Cylinders and Slabs.
SUPERTOG III M2AbstractP00013 I3691 00Data Generator--Fine Group Constants and PN Scattering Matrices from ENDF/B.
SUPERTOG-4AbstractP00013 I0360 00Data Generator--Fine Group Constants and PN Scattering Matrices from ENDF/B.
SUPERTOG-JR.AbstractP00115 F2307 00A Code System for Generating Transport Group Constants, Energy Deposition Coefficients and Atomic Displacement Constants with ENDF/B.
SUPERTOG-JR.AbstractP00115 I0360 00A Code System for Generating Transport Group Constants, Energy Deposition Coefficients and Atomic Displacement Constants with ENDF/B.
SUPERTOG-LTTAbstractP00228 I0360 00A Modification of PSR-13/SUPERTOG-III Applied to Libraries with Tabulated Elastic Scattering and Anistropy Densities.
SURFAbstractC00102 I3675 00Conical and Plane Surface Single Scattering Code.
SUSDAbstractC00501 HM150 00Cross Section Sensitivity and Uncertainty Analysis Including Secondary Neutron Energy and Angular Distributions.
SUSDAbstractC00501 I3090 00Cross Section Sensitivity and Uncertainty Analysis Including Secondary Neutron Energy and Angular Distributions.
SUSD3DAbstractC00695 MNYCP 01Multi-Dimensional, Discrete-Ordinates Based Cross Section Sensitivity and Uncertainty Analysis Code System.
SWANAbstractC00248 C0000 00Code System for Analysis and Optimization of Fusion Reactor Nucleonic Characteristics.
SWANAbstractC00248 CY000 00Code System for Analysis and Optimization of Fusion Reactor Nucleonic Characteristics.
SWANAbstractC00248 I0360 00Code System for Analysis and Optimization of Fusion Reactor Nucleonic Characteristics.
SWANLAKEAbstractC00204 C6600 00Cross Section Sensitivity Analysis Code System for One-Dimensional Discrete Ordinates Calculations.
SWANLAKEAbstractC00204 I3033 00Cross Section Sensitivity Analysis Code System for One-Dimensional Discrete Ordinates Calculations.
SWAP-9AbstractC00788 C0740 001-D Stress Analysis for Hydrostatic and Elastic Plastic Materials.
SWATAbstractC00714 MNYCP 01Step-Wise Burnup Analysis Code System to Combine SRAC-95 Cell Calculation Code and ORIGEN2.
SWIFTAbstractC00679 C7600 00Code System to Calculate Waste-Isolation Flow and Transport.
SWIFTAbstractP00031 C6600 00Monte Carlo Neutron Spectra Unfolding Code.
SWIFT2
USSO
AbstractC00686 MNYCP 00Code System to Calculate Waste-Isolation Flow andTransport.
SWORD 5.0AbstractC00767 MNYCP 05SoftWare for Optimization of Radiation Detectors, SWORD Version 5.0.
SYVAC-D/2AbstractC00690 D0VAX 00Code System For Risk Assessment From Underground Radioactive Waste Disposal In the United Kingdom.
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.