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
Alphabetical Listing
Package NameAbstractRSICC TapelistTitle
1DB-2DB-3DBAbstractC00741 PC586 00One-Dimensional Diffusion Code System for Nuclear Reactor.
1DXAbstractP00096 U1108 00A One-Dimensional Diffusion Code System for Producing Energy Group Collapsed and Self-Shielded Cross Sections.
3DDTAbstractC00605 C6600 00Multigroup Diffusion Code System for Use in Fast Reactor Analysis.
ABAREXAbstractP00248 MNYCP 01Neutron Spherical Optical-Statistical Model Code System.
ABBN-90AbstractD00182 MNYCP 00Multigroup Constant Set for Calculation of Neutron and Photon Radiation Fields and Functionals, Including the CONSYST2 Program.
ABLEIT-TRANSAbstractP00247 C0175 00Error Propagation Analysis for Burnup Calculation.
ACAB-2008AbstractC00758 MNYCP 01Activation Abacus Inventory Code System for Nuclear Applications.
ACATAbstractP00257 FM380 00Monte Carlo Simulation of Atomic Collisions in Amorphous Targets in the Binary Collision Approximation.
ACDOS3AbstractC00442 C7600 00Calculation of Activities and Dose Rates Produced by Neutron Activation.
ACFAAbstractC00478 I3033 00A Versatile Activation Code for Coolant and Structural Materials.
ACHILLESAbstractM00019 MNYCP 00Heat Transfer in PWR Core During LOCA Reflood Phase.
ACOHAbstractC00191 I3675 00Aerojet COHORT Monte Carlo Code System.
ACORNSAbstractP00264 IBMPC 01Analysis of Correlations Used in Neutron Spectrometry.
ACRA-IIAbstractC00213 I0360 00Kernel Integration Code System for Estimation of Radiation Doses Caused by a Hypothetical Reactor Accident.
ACRA-TRITAbstractC00283 I0360 00The Tritium Version of ACRA-II, Estimation of Radiation Doses Caused by a Hypothetical Reactor Accident.
ACROAbstractC00354 I0360 00Calculation of Organ Dose from Acute or Chronic Inhalation and Ingestion of Radionuclides.
ACT-ARAAbstractC00372 CYXMP 00Code System for the Calculation of Changes in Radiological Source Terms with Time.
ACTIVAbstractP00590 I0370 00Sandwich Detector Activity from In-Pile Slowing-Down Spectra Experiment.
ACTIV87AbstractD00169 ALLCP 00Fast Neutron Activation Cross Section File.
ACTIV-PCAbstractP00287 IBMPC 00A Program to Process Gamma or X-ray Spectra.
ACTL82AbstractD00069 ALLCP 01Evaluated Neutron Activation Cross-Section Library.
ACTV-F/HAbstractD00155 ALLCP 00Neutron Activation Cross Section Library for Fusion Reactor Design.
ACTV-FUS/INTAbstractD00170 ALLCP 00International Library of Neutron Activation Cross-Section Data for Fusion Reactor Application.
ADASAGEAbstractP00426 IBMPC 00Ada Application Development System, Versions 4.02, 4.0 and 3.1.
ADEFTA 4.1AbstractP00543 MNYCP 01Atomic Densities for Transport Analysis Script.
ADENAAbstractP00190 C0000 00Code System for Application of Adjusted Data in Calculating Fission-Product Decay Energies and Spectra.
ADENAAbstractP00190 I3033 00Code System for Application of Adjusted Data in Calculating Fission-Product Decay Energies and Spectra.
ADJMOMAbstractC00212 I3675 00Adjoint Moments Method Gamma-Ray Transport Code System.
ADLER IIIAbstractP00058 I0360 00A Program to Calculate Cross Sections from Adler-Adler Resonance Parameters.
ADOAbstractC00189 I3675 00Aerojet Discrete Ordinates Calculational System.
ADS-LIB/V2.0AbstractD00250 MNYCP 00Test Library for Accelerator Driven Systems V2.0
ADVANTG 3.0.3AbstractC00831 MNYCP 03ADVANTG 3.0.3: AutomateD VAriaNce reducTion Generator
AGDATAAbstractD00127 I0360 00Two Agricultural Production Data Libraries (AGDATC and AGDATG) for Dose and Risk Assessment Models.
AIR DATAAbstractD00014 I0360 00Sample Input to ANISN for Calculation of Neutron and Secondary Gamma-Ray Transport in Air.
AIRBORNEAbstractC00263 I0360 00Airborne Contaminants Dispersion Code.
AIRDIFAbstractC00360 C6600 00A Two-Dimensional Atmospheric Radiation Diffusion Code.
AIRDOS-PCAbstractC00551 IBMPC 00Clean Air Act Compliance Software for Personal Computers. See C00542/CAP-88.
AIREKMOD-RRAbstractP00588 D0VAX 00A Point Kinetic Computer Code for Reactivity Transient Analysis in Nuclear Research Reactors AIREKMOD-RR User's Guide.
AIREKMOD-RRAbstractP00588 PCX86 01Reactivity Transients in Nuclear Research Reactors
AIREMAbstractC00242 I3691 00Calculation of Doses, Population Doses, and Ground Depositions Due to Atmospheric Emissions of Radionuclides.
AIRFEWGAbstractD00049 I0360 00Results of ANISN Multigroup Calculations of Gamma-Ray, Neutron, and Secondary Gamma-Ray Transport in Infinite Homogeneous Air Using DLC-31/(DPL-1/FEWG1) Cross Sections.
AIRGAMMAAbstractC00567 FM380 00A Program For The Calculation Of External Exposure To Gamma Rays From A Radioactive Cloud.
AIRSCATAbstractC00341 DP010 00Calculation of Dose Rate for Gamma-Rays Scattered in Air.
AIRTRANSAbstractC00110 I3675 00Monte Carlo Time and Energy-Dependent Three-Dimensional Radiation Transport Code.
AISITE IIAbstractC00286 I0360 00Reactor Siting Code System.
AKERNAbstractC00190 C0000 00Aerojet Point Kernel Integration Calculational System.
AKERNAbstractC00190 U1108 00Aerojet Point Kernel Integration Calculational System.
AKTIVAbstractC00339 I0360 00An Evaluation of Activity, Afterheat and Biological Hazard Potential of Stainless Steel Structures in Fusion Reactor Blankets.
ALARA 2.7.8AbstractC00723 MNYCP 00Code System for Analytic and Laplacian Adaptive Radioactivity Analysis.
ALARM-B2AbstractP00218 I0360 00A Computer Code System for Analysis of a Large Break LOCA of a BWR.
ALBEDO/ALBEZAbstractC00555 IBMPC 00Calculates Attenuation of Radiation in Single and Double Bends.
ALBEDO-DATAAbstractD00224 MNYCP 00KSU Neutron Albedo Data.
ALBEMOAbstractC00268 C6600 00Albedo Monte Carlo Code System.
ALDOSEAbstractC00577 IBMPC 00Dose Calculation for Alpha Disc Source.
ALEPH-LIB-JEFF3.1AbstractD00230 MNYCP 00ACE Format Neutron Cross Section Library based on JEFF3.1.
ALGAM-97AbstractC00152 I3675 00Monte Carlo Estimation of Internal Dose from Gamma-Ray Sources in a Phantom Man.
ALICE2017AbstractP00550 PCX86 06Statistical Model Code System to Calculate Particle Spectra from HMS Precompound Nucleus Decay.
ALKASYS-PCAbstractC00558 IBMPC 00A Computer Program For Studies of Rankine-Cycle Space Nuclear Power Systems.
ALPHA-MAbstractP00169 I0360 00Least-Squares Resolution of Gamma-Ray Spectra in Environmental Samples.
ALPHNAbstractC00612 IBMPC 00Code System for Calculating (alpha,n) Neutron Production in Canisters of High-Level Waste.
ALPSAbstractP00144 F2307 00Alpha Spectrum Analysis Code System.
AMARAAbstractP00079 I3675 00Nuclear Data Adjustment Using Lagrange's Multipliers Method.
AMCAbstractC00090 I3675 00Monte Carlo Albedo Code for Neutron and Capture Gamma-Ray Distributions in Rectangular Concrete Ducts.
AMPAbstractC00793 PCX86 00Advanced Multi-Physics.
AMPX01AbstractD00027 I3675 02126-Group Coupled Neutron and Gamma-Ray Transport Cross-Section Data Generated by AMPX.
AMPX-77AbstractP00315 ALLMF 01Modular Code System for Generating Coupled Multigroup Neutron-Gamma Libraries from ENDF/B.
AMUSEAbstractP00028 C6600 00Gamma-Ray Spectra Unfolding Code.
ANAAbstractP00356 IBMPC 00Code System for Gamma-Ray Spectra Analyses.
ANGELO-LAMBDAAbstractP00544 MNYCP 01Covariance Matrix Interpolation and Mathematical Verification.
ANIPLO D50AbstractP00213 I0360 00A Digital Computer Program for Plotting Results from Calculations with the Sn Computer Program ANISN.
ANISN-ORNLAbstractC00254 MNYCP 02One-Dimensional Discrete Ordinates Transport Code System with Anisotropic Scattering.
ANISN-PCAbstractC00514 IBMPC 00Multigroup One-Dimensional Discrete Ordinates Transport Code System with Anisotropic Scattering. RSIC recommends CCC-650/DOORS3.2a for most applications.
ANITA-2000AbstractC00693 MNYCP 00Code System to Calculate Isotope Inventories from Neutron Irradiation for Fusion Applications.
ANITA-4AbstractC00606 MNYCP 01Analysis of Neutron Induced Transmutation and Activation. See ANITA-2000 (CCC-693).
ANL-BPBAbstractM00004 MNYCP 00Argonne National Laboratory Code Center: Benchmark Problem Book.
ANS643AbstractD00129 IBMPC 02ANS-6.4.3 Geometric Progression Gamma-Ray Buildup Factor Coefficients.
ANSIFTAbstractP00077 C6600 00ANSI Standard Fortran Sifting Program.
ANSIFTAbstractP00077 I0360 00ANSI Standard Fortran Sifting Program.
ANSL-VAbstractD00154 ALLCP 01ENDF/B-V Based Multigroup Cross Section Libraries for Advanced Neutron Source (ANS) Reactor Studies.
ANTE 2AbstractC00131 I3675 00Adjoint Monte Carlo Time-Dependent Neutron Transport Code in Combinatorial Geometry.
APARNA-IIAbstractC00296 I0360 00Integral Transport Theory Code System Based on Discrete Ordinate Representation in Space and Direction-Slab Geometry.
APPLE-2AbstractP00111 FM200 00Plotter of Neutron and Gamma-Ray Spectra and Reaction Rates.
APPLE-2AbstractP00111 I3081 00Plotter of Neutron and Gamma-Ray Spectra and Reaction Rates.
APSAIAbstractP00065 I3691 00Activity Calculations and Plotting of Neutron or Gamma-Ray Spectra Generated by Discrete Ordinates Code System ANISN.
APUD 3.0AbstractC00637 IBMPC 00Code System for Analyzing, Predicting Consequences of, and Guiding the Response to Nuclear Emergencies.
ARCAbstractC00224 C6600 00Aircraft Radiation Transport Code System, Crew Dose Calculation.
ARC 11.0
FEDC
AbstractC00824 MNYWS 00Code System for Analysis of Nuclear Reactors.
ARCON96AbstractC00664 IBMPC 00Code System to Calculate Atmospheric Relative Concentrations in Building Wakes.
AREACAbstractC00438 I3033 00Radiological Emission Analysis Code System.
AREADAbstractP00088 I0360 00Input Data Processor for Transport Codes.
ARMYL-GAbstractC00297 U1106 00Calculation of Transmission Factors for Gamma Rays from Nuclear Explosions.
ARMYL-NAbstractC00298 U1106 00Calculation of Transmission Factors for Neutrons from Nuclear Explosions.
ARRRGAbstractC00404 U1100 00Calculation of Radiation Dose to Man from Radionuclides in the Environment.
ART MOD2AbstractP00611 PCX86 00Fission Product Migration in Primary System and Containment
ASFIT-VARIAbstractC00336 H0000 00Gamma-Ray Transport Code System for One-Dimensional Finite Systems.
ASFIT-VARIAbstractC00336 IBMPC 00Gamma-Ray Transport Code System for One-Dimensional Finite Systems.
ASOPAbstractC00126 IRISC 00Multigroup One-Dimensional Discrete Ordinates Transport Code System for Shield Optimization.
ASTROSAbstractC00073 I7090 00Calculation of Primary and Secondary Proton Dose Rates in Spheres and Slabs of Tissue.
AT123DAbstractC00417 I0360 00Analytical Transient One-, Two-, and Three-Dimensional Simulation of Waste Transport in an Aquifer System.
ATHENA_2DAbstractP00431 MNYCP 00Code System For Simulation Of Hypothetical Recriticality Accidents in a Thermal Neutron Spectrum.
ATM-TOXAbstractC00472 I3033 00An Atmospheric Transport Model for Toxic Substances.
ATTOW-KBAbstractC00132 I0370 00Multigroup Two-Dimensional Removal-Diffusion (Spinney Method) Shielding Code System.
AUS98AbstractC00519 MNYWS 01Modular System for Neutronics Calculations of Fission Reactors, Fusion Blankets, and Other Systems.
AUTOJOM-JOMREADAbstractP00008 C6600 00Computer Programs to Generate or Check Coefficients for Quadratic Equations Describing 3D Geometries.
AXMIXAbstractP00075 CYXMP 00ANISN Cross Section Code System.
AXMIXAbstractP00075 IRISC 01ANISN Cross Section Code System.
AXMIX-PCAbstractP00297 IBMPC 00ANISN Cross Section Mixing Code System.
BABELAbstractD00104 I3033 00Multi-Purpose Neutron and Gamma-Ray Cross Section Library for Fast Reactor Shielding Design.
BALTOROAbstractC00479 C6600 00Code for Coupling of Monte Carlo and Discrete Ordinates Radiation Transport Calculations.
BARC-35AbstractD00124 IBMMF 0035-Group Neutron Cross Sections and Resonance Self-Shielding Factors Generated in ISOTXS and BRKOXS Format from ENDF/B-IV Using MINX.
BASACFAbstractP00285 IBMPC 00Bayesian Approach to Spectrum Adjustment with Covariance Filter.
BAYESAbstractP00205 DP010 00User's Guide for A General-Purpose Computer Code System for Fitting a Functional Form to Experimental Data.
BCGAbstractC00578 C0170 00A Code For Calculating Pointwise Neutron Spectra and Criticality in Fast Reactor Cells.
BEACON MOD3AbstractP00402 CDCMF 00Code System for Thermal-Hydraulic Analysis of Nuclear Reactor Containments.
BEBCAbstractC00077 I7090 00Electron Bremsstrahlung Penetration Code for Space Vehicles.
BEDAbstractC00078 I7090 00Electron Penetration Code for Space Vehicles.
BERMUDAAbstractC00616 FV260 03Discrete Ordinates Code System for Shielding Analysis for Use with Fusion and Fission Reactors.
BEST-5AbstractP00591 I0370 00Power Reactor Fuel Cycle Optimization by Bellman Method.
BETA IIAbstractC00117 C6600 00Monte Carlo Bremsstrahlung and Electron Transport Analysis in Geometry.
BETA IIAbstractC00117 I0360 00Monte Carlo Bremsstrahlung and Electron Transport Analysis in Geometry.
BETA-S 6AbstractC00657 MNYCP 01Code System to Calculate Multigroup Beta-Ray Spectra.
BFR
USSO
AbstractP00449 C0176 00Code System for Common Cause Failure Data Analysis.
BIGGIAbstractC00066 I3675 00Numerical Gamma-Ray Transport Code for Plane or Spherical Multilayer Geometry, Versions 3P and 4T.
BIGGIAbstractC00066 U1108 00Numerical Gamma-Ray Transport Code for Plane or Spherical Multilayer Geometry, Versions 3P and 4T.
BIGGI-4TAbstractC00780 I0360 00Gamma Transport in Multi-Region Shield in Planar or Spherical Geometry.
BISON 1.5AbstractC00464 HM200 00A One-dimensional Discrete Ordinate Transport and Burnup Calculation Code. (Burnup of Isotopes and One-Dimensional Transport).
BISON-CAbstractC00659 MNYWS 00One-Dimensional Discrete Ordinate Transport and Burnup Calculation Code System.
BLOCKAGE V2.5RAbstractP00377 IBMPC 00Code System to Calculate Potential Plugging of Emergency Core Cooling Suction Strainers by Debris in a BWR.
BLT-FEMWATER
USSO
AbstractC00633 PC386 00Code System to Solve for Release and Transport of Contaminants through Saturated/Unsaturated Media.
BMC-MGAbstractC00291 C6600 00Multigroup Monte Carlo Neutron and Gamma-Ray Shielding Code System for Plutonium.
BOB-7 SERIESAbstractP00084 F2306 00Theory and Use of Gamma-Ray Spectrum Analysis Codes for Ge(Li) Detectors.
BOLD VENTURE IVAbstractC00459 I3033 00A Reactor Analysis Code System.
BONAbstractP00173 I0360 00A Code System for Unfolding Multisphere Spectrometer Neutron Measurements.
BOREHOLE-EB6.8-MGAbstractD00268 MNYCP 00Multi-Group Cross-Section Library for Deterministic and Monte Carlo Codes.
BOT3P-5.3AbstractP00530 MNYCP 02Code System for 2D and 3D Mesh Generation and Graphical Display of Geometry and Results for Radiation Transport Codes.
BOXERAbstractC00766 MNYWS 00Fine-flux Cross Section Condensation, 2D Few Group Diffusion and Transport Burnup Calculations
BPAbstractD00008 I0360 00Data for Selected Shielding Benchmark Problems Specified in ORNL-RSIC-25, Shielding Benchmark Problems.
BPPCAbstractC00076 I7090 00Proton Penetration Codes for Space Vehicles.
BREESE-IIAbstractP00143 I3033 00Auxiliary Routines for Implementing the Albedo Option in the MORSE Monte Carlo Code System.
BREMRADAbstractC00031 I7090 00External and Internal Bremsstrahlung Calculation Code.
BRHGAMAbstractC00350 I3033 00Monte Carlo Estimation of Absorbed Dose from X-Ray Sources in Phantom Man.
BRMSTKAbstractP00044 C6600 00CSEWG Integral Data Testing Shielding Experiment Code System.
BRMSTKAbstractP00044 I3691 00CSEWG Integral Data Testing Shielding Experiment Code System.
BSPRP2AbstractP00372 IRISC 00Code System to Process DORT Boundary-Flux Files.
BUCORSTAbstractP00339 PC386 00A Code to Prepare Burnup-Dependent Multigroup Nuclear Reactor Source Terms.
BUGENDF70.BOLIBAbstractD00262 PCX86 00ENDF/B-VII.0 Broad-Group Coupled Cross Section Library for LWR Shielding & Pressure Vessel Dosimetry Applications.
BUGJEFF311.BOLIBAbstractD00254 MNYCP 01JEFF-3.1.1 Broad-Group Coupled Cross Section Library For LWR Shielding & Pressure Vessel Dosimetry Applications.
BUGLE-80AbstractD00075 IBMPC 02Coupled 47 Neutron, 20 Gamma-Ray Group, P3, Cross Section Library for LWR Shielding Calculations by the ANS-6.1.2 Working Group on Multigroup Cross Sections. We recommend ENDF/B-VI derived data in DLC-185/BUGLE-96.
BUGLE-80AbstractD00075 IBMPC 03Coupled 47 Neutron, 20 Gamma-Ray Group, P3, Cross Section Library for LWR Shielding Calculations by the ANS-6.1.2 Working Group on Multigroup Cross Sections. We recommend ENDF/B-VI derived data in DLC-185/BUGLE-96.
BUGLE-80AbstractD00075 PC386 01Coupled 47 Neutron, 20 Gamma-Ray Group, P3, Cross Section Library for LWR Shielding Calculations by the ANS-6.1.2 Working Group on Multigroup Cross Sections. We recommend ENDF/B-VI derived data in DLC-185/BUGLE-96.
BUGLE-93AbstractD00175 ALLCP 01Coupled 47 Neutron, 20 Gamma-Ray Group Cross Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications.
BUGLE-96AbstractD00185 ALLCP 00Coupled 47 Neutron, 20 Gamma-Ray Group Cross Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications.
BULK_C-12AbstractC00738 PC586 00Code System to Estimate Neutron and Photon Effective Dose Rates from Medium Energy Protons or Carbon Ions Through Concrete or Concrete/Iron.
BULK-IAbstractP00574 PCX86 00Radiation Shielding Tool for Proton Accelerator Facilities.
BURDAbstractP00582 IBMPC 00Bayesian Estimation in Data Analysis of Probabilistic Safety Assessment.
BURP-2AbstractC00237 C6600 00Calculation of Buildup and Decay of Radioactive Fission Products.
BUSHAbstractC00333 I0360 00A Code to Calculate Radiation Doses Inside Buildings from Routine Releases of Radionuclides to the Atmosphere.
BUTTERCUPAbstractC00779 MNYCP 00A Dual-Layer Photon Buildup Factor Code.
BWR-LTASAbstractC00485 I3033 01Code System for Boiling Water Reactor Long-Term Accident Simulation.
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)
OECD
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.
CEPXSAbstractC00837 MNYCP 00Coupled Electron-Photon Cross Section
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-RERTRAbstractP00606 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 4AAbstractP00614 MNYCP 01Code 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.
COG LIBMAKERAbstractP00607 MNYCP 00LIBMAKER
COG SUPPLEMENTAL LIBRARIESAbstractD00271 MNYCP 00COG LibMaker – Data Conversion Utility
COG11.1AbstractC00829 MNYCP 00Multiparticle 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.
D2OAbstractP00398 PC486 00Code System for Computing Thermodynamic and Transport Properties of D2O.
DABL69AbstractD00130 I0360 01Defense Nuclear Applications Broad-Group Library based on ENDF/B-V in ANISN Format.
DACRINAbstractC00273 U1100 00Airborne Radionuclide Organ Dose Calculational System.
DANCOFF3AbstractP00279 D8810 00Calculates Dancoff Correction.
DANCOFF-MCAbstractP00509 MNYCP 00Code System for Monte Carlo Calculation of Dancoff Factors in Irregular Geometries.
DANESS V1.0
FEDC
AbstractP00555 MNYCP 00Dynamic Analysis of Nuclear Energy System Strategies.
DANTEAbstractP00185 I0370 00Unfolding Code System for Energy Spectra Evaluation for Dosimetry Purposes.
DANTSYS 3.0AbstractC00547 MFMWS 01One-, Two-, and Three-Dimensional, Multigroup, Discrete-Ordinates Transport Code System. See new release CCC-707/PARTISN.
DART-V.1AbstractC00830 MNYCP 00Displacement per Atom, Primary Knocked-on Atoms Produced in an Atomic Solid Target
DASH-FPAbstractC00366 C0000 00A One-Dimensional Analytic-Numerical Solution to the Problem of Multicomponent Time-Dependent Diffusion of Fission Products.
DASQHEAbstractP00278 D8810 00Calculates Dancoff Corrections Factors.
DATINITAbstractP00258 DGMV1 00Interactive Program To Access Photon Interaction Data.
DAVEAbstractC00166 I3675 00Monte Carlo Gamma-Ray Transport Code System in One-Dimensional Spherical Geometry.
DCHAIN 1.3AbstractC00640 MNYCP 01Code System for Radioactive Decay and Reaction Chain Calculations.
DCHAIN2AbstractC00370 PC486 00A Code System for Calculation of Transmutation of Nuclides.
DCHAIN-SP2001AbstractC00712 MNYWS 01Code System for Analyzing Decay and Build-up Characteristics of Spallation Products.
DCTDOSAbstractC00520 IBMPC 00Neutron and Gamma-Ray Penetration in Composite Duct Systems.
DDXCODESAbstractC00583 FM380 00One-, Two- and Three-Dimensional Transport Codes Using Multigroup Double-Differential Form Cross Sections.
DDXLIBAbstractD00123 FM380 01125-Neutron Group Double Differential Cross Section Library.
DECAYREMAbstractD00030 I0360 02Radioactive Decay Spectra in EXREM Format.
DECDC 1.0AbstractD00213 MNYCP 00Nucear Decay Data Files for Radiation Dosimetry Calculations.
DEISAbstractC00455 C6600 00Draft Environmental Impact Statement on Licensing Requirements for Land Disposal of Radioactive Waste.
DEMON & DEMON RAbstractC00181 I3675 00Demonstration Monte Carlo Code System in Slab Geometry.
DENISAbstractP00082 I0360 00Monte Carlo Simulation of the Capture and Detection of Neutrons with Large Liquid Scintillators.
DEPLETORAbstractP00523 MNYCP 00Code System to Provide Depletion Capability to the U.S. NRC PARCS Code
DEPOSITION
FEDC
AbstractP00420 IBMPC 00Code System to Calculate Particle Penetration Through Aerosol Transport Lines.
DETAN 95AbstractP00361 MNYCP 00Code System to Calculate Spectrum-Averaged Cross Sections and Detector Responses in Neutron Spectra.
DIAMANT2AbstractC00414 PC386 00Multigroup Two-Dimensional Discrete Ordinates Transport Code System for Triangular Geometry, Release 2.0.
DIF3D 11.0
FEDC
AbstractC00784 MNYCP 01Code System Using Variational Nodal Methods and Finite Difference Methods to Solve Neutron Diffusion and Transport Theory Problems.
DIFBASAbstractP00334 MNYCP 00A Bayesian Approach to Unfolding a Neutron Spectrum from a Spectrum of Recoiled Protons.
DIFMODAbstractC00572 I3083 00A Computer Program To Calculate The Leaching of Radionuclides and the Corrosion of Cemented Waste Forms in Water or Brine.
DIMENAbstractP00341 IBMPC 00Code System for Isotope Identification by Gamma-Ray Analysis.
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.
DINT-YAECAbstractC00306 ALLMF 00Evaluator of I1 and I2 Integrals as Used in Long-Term External Gamma-Ray Doses from Routine Atmospheric Releases.
DIPHOAbstractC00140 I3675 00Monte Carlo Gamma-Ray Code System-Infinite Medium, Mono-energetic and Isotropic Point Source.
DISDOSAbstractC00170 I0360 00Calculation of Dose Distribution in Human Phantoms Irradiated by External Photon Sources.
DISKTRANAbstractC00533 CYXMP 00Dose Calculations at Detectors from the End of a Cylinder Using DOT IV Scalar Flux Data.
DISKTRANAbstractC00533 I3033 00Dose Calculations at Detectors from the End of a Cylinder Using DOT IV Scalar Flux Data.
DISPERSAbstractC00454 MNYCP 00Mathematical Models for Dispersion of Radionuclides
DIXY-2AbstractC00812 I0370 002-D Homogeneous and Inhomogeneous Neutron Diffusion N X-Z, R-Z, R-Theta Geometry with Perturbation.
DKRAbstractC00323 CY000 00A Radioactivity and Dose Rate Calculation Code System for Fusion Reactors.
DLSAbstractC00264 C6600 00Two-Dimensional Shielding Calculational System with Diffusion Theory and Line-of-Sight Method.
DOMINOAbstractP00064 I0360 00A General Purpose Code System for Coupling Discrete Ordinates and Monte Carlo Radiation Transport Calculations.
DOMINO-IIAbstractP00162 I3033 00General Purpose Code System for Coupling DOT-IV Discrete Ordinates and Monte Carlo Radiation Transport Calculations.
DOMUSAbstractP00301 IPCXT 00A Program for Decomposing A Two-Dimensional Spectrum.
DOORS 3.2AAbstractC00650 MFMWS 04One, Two- and Three-Dimensional Discrete Ordinates Neutron/Photon Transport Code System.
DOPEXAbstractC00177 I3675 00Laminated Shield Weight Optimization Code System-Steepest Descent Calculational Model.
DOPEXAbstractC00177 U1108 00Laminated Shield Weight Optimization Code System-Steepest Descent Calculational Model.
DOPEX-1D2CAbstractC00214 I0360 00A One-Dimensional, Two-Constraint Radiation Shield Optimization Code System.
DOQDPAbstractP00110 I0360 00Discrete Ordinates Quadrature Generator.
DORGLIBAbstractP00181 I0360 00An Interactive Program for Displaying Nuclide Decay and Generation Data Based on ORIGEN Data Library.
DORIANAbstractP00425 IBMPC 00Code System to Implement Bayes Method for Plant Aging Risk Analysis.
DOSCOVAbstractD00090 I0360 0024-Group Covariance Data.
DOSDAM77-81AbstractD00081 C6400 00620 Group, SAND-II Formatted, Neutron Cross Sections Based on ENDF/B-IV and Other Sources for Spectral, Integral, and Damage Analyses.
DOSDAM81-82AbstractD00097 C0000 00Multigroup Cross Sections in SAND-II Format for Spectral, Integral, and Damage Analyses.
DOSDAM84AbstractD00131 IBMMF 00Multigroup Cross Sections in SAND-II Format for Spectral, Integral, and Damage Analyses.
DOSDAT II-81AbstractD00079 I0370 00Dose-Rate Conversion Factors for External Exposure to Photons and Electrons.
DOSDAT-DOEAbstractD00144 ALLMF 00Dose-Rate Conversion Factors for External Exposure to Photons and Electrons.
DOSDAT-DOEAbstractD00144 IBMPC 01Dose-Rate Conversion Factors for External Exposure to Photons and Electrons.
DOSE 1AbstractC00165 I3675 00Gamma-Radiation Dosimetry for Arbitrary Source and Target Geometry.
DOSE-SGTRAbstractC00624 IBMPC 00Code System to Calculate the Integrated Iodine Release to the Environment During a Steam Generator Tube Rupture in a PWR.
DOSFACTER IIAbstractC00400 D0750 00Calculation of Dose-Rate Conversion Factors for Exposure to Photons and Electrons.
DOSFACTER IIAbstractC00400 I0360 00Calculation of Dose-Rate Conversion Factors for Exposure to Photons and Electrons.
DOSFACTER-DOEAbstractC00536 I3033 00Calculation of Dose-Rate Conversion Factors for Exposure to Photons and Electrons.
DPCTAbstractC00580 CYXMP 00A Deterministic-Probabilistic Model For Contaminant Transport.
DPL-400 GEDT1AbstractD00031 I0360 08Defense Nuclear Agency 37 Neutron, 21 Gamma Ray Coupled, P3, Multigroup Library in ANISN Format.
DPL-401 NEDTAbstractD00031 I0360 09Defense Nuclear Agency 37 Neutron, 21 Gamma Ray Coupled, P3, Multigroup Library in ANISN Format.
DPL-402A/GPDT1AbstractD00031 I0360 10Defense Nuclear Agency 37 Neutron, 21 Gamma Ray Coupled, P3, Multigroup Library in ANISN Format.
DPL-402B/GPDT1AbstractD00031 I0360 11Defense Nuclear Agency 37 Neutron, 21 Gamma Ray Coupled, P3, Multigroup Library in ANISN Format.
DRAGON2PARTISNAbstractC00803 PCX86 00Cross-Sections Data Generation for PARTISN4.0.
DRAGON3.05DAbstractC00647 MNYWS 03Lattice Cell Code System.
DRALISTAbstractD00080 ALLCP 00Radioactive Decay Data for Application to Radiation Dosimetry and Radiological Assessments.
DSNPAbstractP00592 I3033 00Dynamic Simulation Nuclear Power.
DSNQUADAbstractP00251 IPCXT 00Calculates Angular Quadrature Weights and Cosines.
DTF-69AbstractC00130 C6600 00One-Dimensional Multigroup Photon Transport Discrete Ordinates Code System.
DTF-INDIAAbstractC00458 I0370 00Multigroup Neutron Transport Discrete Ordinates Code System with One-Dimensional, Anisotropic Scattering.
DTF-IVAbstractC00042 C6600 00Multigroup Neutron Transport Discrete Ordinates Code System with One-Dimensional, Anisotropic Scattering.
DTF-IV MODIFIEDAbstractC00042 I0370 00Multigroup Neutron Transport Discrete Ordinates Code System with One-Dimensional, Anisotropic Scattering.
DTF-TRACAAbstractC00412 U1100 00One-Dimensional Multigroup Neutron Transport Discrete Ordinates Code System.
DTKAbstractC00223 I3675 00One-Dimensional Multigroup Neutron Transport Code System.
DUFOLDAbstractP00042 I0360 00Derivative Unfolding Code - Determination of Neutron Spectra from NE-213 Pulse Height Data.
DUSTAbstractC00453 I3033 00Albedo Monte Carlo Simulation of Neutron Streaming Through Multilegged Ducts.
DUST-BNLAbstractC00634 PC386 00Disposal Unit Source Term by One-Dimensional, Transient, Finite-Difference, Subsurface Release and Transport of Contaminants.
DWBA07/DWBB07AbstractP00338 MNYCP 01Code System for Inelastic and Elastic Scattering with Nucleon-Nucleon Potential
DWNWNDAbstractC00383 DP010 00Interactive Gaussian Plume Atmospheric Transport Model.
DWUCK-CHUCKAbstractP00546 MNYCP 00Nuclear Model Code System for Distorted Wave Born Approximation and Coupled Channel Calculations.
DYN3D/M2AbstractP00579 I3090 00Reactivity Transients in Light H2O Reactors with Hexagonal Geometry.
E3LWRAbstractD00098 C0000 0045 Neutron, 16 Gamma-Ray and 15 Neutron, 5 Gamma-Ray Group LWR Cross Section Libraries Derived from EURLIB-III using the AGRUKO Optimized Collapsing Scheme.
EACRP-D2O-LATTICESAbstractD00264 MNYCP 00Compilation of Reactor Physics Measurements in HWRs Lattices.
EASY-QAD 2.0.1AbstractC00744 PC586 02A Visualization Code System for Gamma and Neutron Shielding Calculations, Version 2.0.1
ECIS-12AbstractP00612 MNYCP 00ECIS-12, Coupled Channel, Statistical Model, Schroedinger and Dirac Equation, Dispersion Relation
ECPL82AbstractD00106 ALLCP 00Evaluated Charged-Particle Data Library.
E-DEP-1AbstractC00275 D0VAX 00Heavy Ion Energy Deposition Code System.
EDISTRAbstractP00191 I3033 00Prepares a Nuclear Decay Data Base for Internal Radiation Dosimetry Calculations.
EDITORAbstractP00035 I0360 00Alters Mode, Copies, Merges, Punches, Edits, or Adds to ENDF/B-Formatted Data on Tapes or Cards.
EDMULT 6.4AbstractC00430 MNYCP 02Evaluates Electron Depth-Dose Distributions in Multilayer Slab Absorbers.
EDNAAbstractC00104 I7090 00Electron Dose and Number Analysis Code by Kernel Integration.
EDOAbstractC00489 U1110 00A Code System in Fortran V for the Evaluation of Dose During Normal Operation of a Nuclear Power Plant.
EDSFI
USSO
AbstractD00215 PC486 00Electrical Distribution System Functional Inspection Data Base.
EEDBAbstractP00531 MNYCP 00The Energy Economic Data Base.
EFDOSAbstractC00411 I0360 00Calculation of Effective Committed Dose Equivalents by Inhalation of Radioactive Materials Occurring in Routine Atmospheric Releases from Nuclear Fuel Cycle Facilities.
EGADAbstractC00206 I0360 00Calculation of Dose from External Gamma-Ray Emitters.
EGS4AbstractC00331 MNYCP 00Monte Carlo Simulation of the Coupled Transport of Electrons and Photons.
ELANAbstractP00141 ICL00 00Neutron Cross-Section Self-Shielding Code System.
ELAST2AbstractD00208 MNYCP 00Database of Cross Sections for the Elastic Scattering of Electrons and Positrons by Atoms.
ELBAAbstractC00119 I0360 00Electron and Bremsstrahlung Dose Rate Code.
ELECSPECAbstractD00100 DP010 00Electron Spectra from Decay of Fission Products.
ELEORBITAbstractC00751 PCX86 003-D Simulation of Electron Orbits in Magnetic Multipole Plasma Source.
ELFAbstractC00167 I0360 00Monte Carlo Neutron Transport Code System for Cylinders and Spheres.
ELGATLAbstractC00295 C6600 00Calculation of Energy Spectra from Coupled Electron-Photon Slowing Down.
ELIESE-3AbstractP00003 I0370 00Analyses of Elastic and Inelastic Scattering Cross Sections.
ELPHOAbstractC00301 I0360 00Three-Dimensional Monte Carlo Electromagnetic Transport Code System.
ELTRANAbstractC00155 C3600 00One-Dimensional Monte Carlo Electron Transport Code System.
EMERALDAbstractC00211 I0360 00Calculation of Activity Releases and Potential Doses from a Pressurized Water Reactor Plant.
EMERALD-NORMALAbstractC00250 I0370 00Calculation of Activity Releases and Potential Doses from the Normal Operation of a Pressurized Water Reactor Plant.
EMPIRE-IIAbstractP00497 PC586 01Comprehensive Nuclear Model Code, Nucleons, Ions Induced Cross-Sections.
ENBAL2AbstractP00160 I0370 00A Program to Generate Multigroup Neutron Kerma Factors.
ENDF UTIL. CODESAbstractM00008 MNYCP 00ENDF Checking and Utility Codes.
ENDL82AbstractD00103 ALLCP 00Neutron Library in Transmittal Format.
ENDLIB-97AbstractD00179 MNYCP 01LLNL Libraries of Atomic Data, Electron Data, and Photon Data in Evaluated Nuclear Data Library (ENDL) Type Format.
ENDVER/GUIAbstractP00572 PCX86 00The ENDF File Verification Support Package.
ENEDEPAbstractC00227 GE400 00Energy Deposition Code System for GE 265 Time-Sharing System.
ENLOSSAbstractP00047 C6600 00Calculation of Energy Loss of Charged Particles.
ENSL82-CDRL82AbstractD00110 ALLCP 00Evaluated Nuclear Structure Libraries.
ENTOSANAbstractP00188 C0175 00Code System for Calculating Fine-Group Dosimetry Cross Section Values from ENDF/B Data.
ENTOSANAbstractP00188 D8810 00Code System for Calculating Fine-Group Dosimetry Cross Section Values from ENDF/B Data.
ENTREE 1.4.0AbstractP00519 MNYWS 00BWR Core Simulation System for Space and Time Dependent Coupled Phenomena.
EPICS2014AbstractD00272 MNYCP 00Electron Photon Interaction Cross Sections
EPIPE
USSO
AbstractP00485 CY000 00Code System for Static and Dynamic Piping System Analysis.
EPRAbstractD00037 I3691 05Coupled 100-Group Neutron 21-Group Gamma-ray Cross Sections for EPR Neutronics.
EPR MASTERAbstractD00052 I3691 00100 Neutron Group Cross Sections in AMPX Master Library Format.
EPRI-CINDERAbstractC00309 C6600 00General Point-Depletion and Fission Product Code System and Four-Group Fission Product Neutron Absorption Chain Data Library Generated from ENDF/B-IV for Thermal Reactors.
EQUIVA-1.1AbstractP00323 IMFPC 00Generation of Few-Group Equivalent Diffusion Theory Parameters for PWR Reflector Regions.
EQUIVA-2AbstractP00324 IMFPC 00Generation of Environment-Insensitive Equivalent Diffusion Theory Parameters for PWR Reflector Regions.
ERANOS 2.0
OECD
AbstractC00745 MNYWS 00Modular Code and Data System for Fast Reactor Neutronics Analyses
ERIC-2AbstractP00119 I0360 00Calculator of Resonance Integral and Effective Capture and Fission Cross Sections for Fissile and Non-Fissile Nuclides - Thermal or Fast Reactors.
ERINNIAbstractP00219 I0360 00Optical Model Calculation of Multiple Cascading Particle Emissions.
ERPEXAbstractC00305 C0073 00Monte Carlo Distributions of Energetic Proton Ranges in Silicon.
ERRORJAbstractP00526 MNYCP 03Covariance Processing Code System, Version 2.3.
ESDORAAbstractC00183 U1108 00Fission Product Inventory and Gamma-Ray Dose Rate from a Radioactive Cloud System.
ESGAbstractD00065 I0360 0056-Group Cross Section Library Based on VITAMIN-C Generated by Using SPHINX and XSDRNPM to Collapse 171 Groups.
ESPAbstractC00193 I0360 00General Purpose Monte Carlo Neutron Transport Code System.
ESTIMAAbstractP00201 I3033 00A Code System for Calculating Average Parameters from Sets of Resolved Resonance Parameters.
ETHELAbstractP00217 I0360 00Code System for Generating Cross Sections for PSR-128/THERMOS.
ETOE-2AbstractP00585 I3033 00Cross-Sections Library for Program MC**2 Generator from ENDF/B.
ETRANAbstractC00107 I0360 00Monte Carlo Code System for Electron and Photon Through Extended Media.
EURCYLAbstractP00076 I0370 00Finite Element Three-Dimensional Mesh Generator for Cylinder - Cylinder Intersections.
EURLIB-IIIAbstractD00035 I0360 01100 Neutron, 20 Gamma-Ray Group Cross Section Library for Use in the European Shielding Benchmark Program.
EVALPLOTAbstractP00211 I3081 00A Program to Plot Data in the Evaluated Nuclear Data File/Version B Format.
EVAPAbstractP00010 I0360 00Calculation of Particle Evaporation from Excited Compound Nuclei.
EVNTREAbstractP00465 D0VAX 00Code System for Event Progression Analysis for PRA.
EXCURS-3-RRAbstractP00586 D0VAX 00Kinetics of Research Reactor Reactivity Transient Analysis.
EXIFON2.0AbstractP00305 IPCXT 01A Model for Statistical Multistep Direct and Multistep Compound Reactions.
EXPALSAbstractC00787 C7600 00Least Square Fit of Linear Combination of Exponential Decay Function.
EXPRESSAbstractC00622 MNYCP 00Exact Preparedness Supporting System.
EXTREMEAbstractC00440 I3033 00Two-Dimensional Discrete-Ordinates Code System with Exponential Expansion of Spatial Variables.
EZVIDEOAbstractP00237 IBMPC 00Graphics Routines for the IBM PC.
F5TABAbstractP00221 D0780 00Code System for Converting Energy Distribution Cross Section Data to Tabulated Data.
FAMRECAbstractP00167 C7600 01Fuel Assembly Mechanical Response Code System.
FANACAbstractP00179 I3033 00A Shape Analysis Code Package for Resonance Parameter Extraction from Neutron Capture Data for Light- and Medium-Weight Nuclei.
FANALAbstractP00178 I3033 00A Least-Squares Shape Analysis Code System.
FANGAbstractP00140 C0000 00An Angular Folding Code System for Channel Theory Analysis.
FANGAbstractP00140 I0360 00An Angular Folding Code System for Channel Theory Analysis.
FANTOMAbstractC00375 BESM6 00Monte Carlo Calculation of the Response of an External Detector to a Photon Source in the Lungs of a Heterogeneous Phantom.
FASTER IIIAbstractC00168 U1108 00Monte Carlo Neutron and Photon Transport Code System in Complex Geometries.
FASTER-IIIAbstractC00168 I3675 00Monte Carlo Neutron and Photon Transport Code System in Complex Geometries.
FASTGRASSAbstractP00479 MNYCP 00Code System to Predict Fission Product Release in Ubase Fuels.
FASTPLOT 1.0AbstractP00354 IBMPC 00Interface to Microsoft FORTRAN Graphics.
FATDUDAbstractP00080 I0360 00Foil Activation Data Unfolding Code System.
FBSAMAbstractP00103 I0360 00User-Storage - Magnetic Disk Data Manipulator.
FCXSECAbstractD00085 PC386 0122 Neutron, 21 Gamma-Ray Group Cross Section Libraries in ANISN Format for Nuclear Fuel Cycle Shielding Calculations.
FDKRAbstractC00541 I4381 00Radioactivity and Dose Rate Calculation Code for Fission, Fusion and Hybrid Reactors.
FDMXPCAbstractP00322 IPCAT 00Code System for Calculation of Neutron Transmission and Other Functionals from Evaluated Data in ENDF Format.
FE3DGWAbstractC00531 D0780 00Code System for Finite-Element, Three-Dimensional Ground-Water Flow Analysis.
FEAST METALAbstractP00563 MNYCP 00Fuel Engineering and Structural Analysis Tool.
FEDGROUP-3AbstractP00123 I0360 00Program System for Processing Evaluated Nuclear Data in ENDF/B, KEDAK or UKNDL Format to Constants to be Used in Reactor Physics Calculation.
FEDGROUPC86REV3AbstractP00194 MNYCP 01Code System for Processing Evaluated Nuclear in ENDF/B, KEDAK or UKNDL Formats into Constants for Reactor Physics Calculation.
FEDGROUP-RAbstractP00349 MNYCP 00Multigroup Neutron Cross Section Processing System from Data in ENDF/B Format.
FEM-2DAbstractC00260 C6600 00Two-Dimensional Diffusion Theory Code System Based on the Method of Finite Elements.
FEMAXI 6 VER.1AbstractP00536 IBMPC 00Code System for Light Water Reactor Fuel Analysis.
FEMBAbstractC00340 B6700 00A Two-Dimensional Diffusion Theory Finite Element Program.
FEMRZAbstractC00342 F2307 00A Finite-Element Method Two-Dimensional Multigroup Neutron Transport Code System, (r,z) Geometry.
FEMWASTE/FEMWATERAbstractC00451 C7600 00A Finite-Element Model of Waste and Water Transport through Porous Saturated-Unsaturated Media.
FEMWASTE/FEMWATERAbstractC00451 PC386 00A Finite-Element Model of Waste and Water Transport through Porous Saturated-Unsaturated Media.
FENDL-2.0AbstractD00183 MNYCP 01Compendium of Reference and Processed Sub-libraries Derived from International Evaluated Nuclear Data Files for Fusion Applications.
FENDL-2.1AbstractD00222 MNYCP 00Compendium of Reference and Processed Sub-libraries Derived from International Evaluated Nuclear Data Files for Fusion Applications.
FEP 4.16AbstractP00440 IBMPC 00Fault-tree, Event tree, & P&ID Editors.
FERDO/FERDAbstractP00102 I3033 00Multichannel Neutron and Gamma-Ray Spectrum Matrix Unfolding Code Systems.
FERDORAbstractP00017 I7090 00Spectra Unfolding Codes.
FERDORAbstractP00017 U1108 00Spectra Unfolding Codes.
FERD-PCAbstractP00273 IBMPC 00Interactive Multichannel Neutron and Gamma-Ray Spectrum Matrix Unfolding Code System.
FERRETAbstractP00145 U0000 00Least-Squares Solution to Nuclear Data and Reactor Physics Problems.
FESHAbstractC00676 CDCMF 00X-Y Multigroup Neutron Transport Code System.
FEWA-FEMAAbstractC00477 I3033 00A Finite Element Model of Water and Other Material through Aquifers.
FEWG1-81AbstractD00031 I0370 06Defense Nuclear Agency 37 Neutron, 21 Gamma Ray Coupled, P3, Multigroup Library in ANISN Format.
FEWG1-85AbstractD00031 I0360 07Defense Nuclear Agency 37 Neutron, 21 Gamma Ray Coupled, P3, Multigroup Library in ANISN Format.
FGR-DOSEAbstractD00167 ALLCP 01Dose Coefficients from Federal Guidance Reports 11 and 12.
FGXRRSAbstractD00132 C0000 00Few Group Cross Section Library for Research Reactor Calculations.
FIGEROAbstractP00149 C0000 00Processing Codes for Generating Multigroup Neutron Cross Sections from ENDF/B for Use in Discrete Ordinates Calculations.
FINELMAbstractC00483 MFMWS 00Multigroup Finite Element Diffusion Code System.
FIPDIGAbstractC00251 I0360 00One-Dimensional Time-Dependent Fission Product Diffusion Code System.
FIRACAbstractP00444 CY000 00Nuclear Facilities Fire Accident Model
FIREDATAAbstractD00125 PC486 00Nuclear Power Plant Fire Data Base for Personal Computers.
FISP-6AbstractC00538 I3090 00An Enhanced Code for the Evaluation of Fission Product Inventories and Decay Heat.
FISPACT-IIAbstractC00836 MNYCP 01FISPACT-II 3.20 & TENDL-2015, -2014, ENDF/B-VII.1, JENDL-4.0, JEFF-3.2, CENDL-3.1 and GEFY libraries.
FISPINAbstractC00413 ICL00 00Nuclide Inventory Calculation System.
FIS-PRODAbstractD00152 ALLCP 00Chinese Evaluated Fission Product Yield Library in ENDF/B-V Format.
FISSP & CLOUDAbstractC00163 MNYCP 01Fission Product Inventory, Release, Transport and Dose Calculation.
FITOCOAbstractP00189 C0175 00Converter of Fine-Group Flux Density and Cross Section Data to Coarse Group Values.
FLANGE-ORNLAbstractP00566 I0360 00Flanged Pipe Joint Stress Analysis, Internal Pressure, Moment Loads, Temperature.
FLEPAbstractD00022 I3033 00Coefficients for the Analytic Representation of Nonelastic Cross Sections and Particle-Emission Spectra from Various Nucleon-Nucleus Collisions in the Energy Range 25 to 400 MeV.
FLODISAbstractP00417 I0360 00Code System to Calculate Thermal Response of FSV HTGR Core.
FLOWPLOT IIAbstractP00234 I3033 00Fluid Dynamics and Heat Transfer Plotting Package.
FLUKA05-PRE-LIBAbstractD00260 PCX86 00FLUKA05 Multi-Group, Multi-Purpose Nuclear Data Library, Neutrons, Photons, Charged Particles.
FLUKA-TRANKAAbstractC00207 C6600 00Three-Dimensional High-Energy Extranuclear Hadron Cascade Monte Carlo System for Cylindrical Backstop Geometries.
FLUNGAbstractD00086 I3033 00Coupled 35-Group Neutron and 21-Group Gamma Ray, P3 Cross Sections for Fusion Applications.
FLUSHAbstractP00043 C6600 00Spectral Unfolding Code - Stepwise Regression of System Response Functions.
FLYSPEC-SHORTSAbstractP00196 C7600 00Neutron Unfolding Code System for Reducing Proton-Recoil Pulse-Height Obtained with NE-213 Liquid Scintillator.
FOCUSAbstractC00390 I3033 00Adjoint Monte Carlo Neutron Transport Code System.
FONTAAbstractC00423 S4044 00Code System For Calculating Individual And Collective Doses From Reactor Accidents Using Pasquill's Plume Model.
FOODAbstractC00403 U1108 00Calculation of Radiation Dose to Man from Radionuclides in the Environment.
FORECAST V3.0AbstractP00384 IBMPC 00Forecast Regulatory Effects Cost Analysis Program.
FORISTAbstractP00092 C0000 00Neutron Spectrum Unfolding Code System - Iterative Smoothing Technique.
FORISTAbstractP00092 I0360 00Neutron Spectrum Unfolding Code System - Iterative Smoothing Technique.
FORSENAbstractP00170 I0360 00A Multigroup Processing Code for Use with Sensitivity Profiles to Assess the Effect of Cross Section Changes.
FORSIM VIAbstractP00078 C6600 00A Fortran-Oriented Simulation Package for the Automated Solution of Partial and Ordinary Differential Equation Systems.
FORSSAbstractC00334 C0000 00A Sensitivity and Uncertainty Analysis Code System.
FORSSAbstractC00334 I0360 00A Sensitivity and Uncertainty Analysis Code System.
FOTELP-2014AbstractC00581 MNYCP 04Photons, Electrons and Positrons Transport in 3D by Monte Carlo Techniques
FOURACESAbstractP00183 I0370 00Code System for Producing Spectrum Weighted, Group Averaged Cross Sections from ENDF/B, KEDAK, or UK Libraries.
FPDLAbstractD00066 I0360 00Fission Product Yields, Gamma Ray and Beta Spectra in ENDF-III Format for 235U, 238U, 239Pu, 232Th, and 233U.
FPGAMAbstractC00386 F2307 00Calculation of Fission-Product Gamma-Ray Spectra.
FPICAbstractC00028 I3675 00Fission Product Inventory Code.
FPIPAbstractC00162 C6600 00Fission Product Inventory Code System.
FPZDAbstractC00603 PC386 00Code System for Multigroup Neutron Diffusion/Depletion Calculations.
FRANCOAbstractP00363 MNYCP 00Finite Element Fuel Rod Analysis Code System.
FRANTIC3AbstractP00406 CDCMF 00Time-Dependent Reliability Analysis.
FRAPCON2AbstractP00517 MFMWS 00Fuel Rod Thermal-Mechanical Behavior, Versions FRAPCON2, FRAPCON2/VIM4, & FRAPCON2/VIM5.
FRAPT6/MOD1
USSO
AbstractP00436 C0176 00Code System for Transient Analysis of Fuel Rods.
FRAPT6/V21
USSO
AbstractP00436 C0176 01Code System for Transient Analysis of Fuel Rods.
FRCRL2AbstractC00231 C6400 00Calculation of Fission-Product Release in Reactor Accident Analyses.
FREEFORMAbstractP00081 I0360 00Free-Form Input Reading Routines.
FSCATTAbstractC00186 I3033 00Discrete Ordinates Gamma-Ray Transport Code System in Plane Geometry.
FSCATTAbstractC00186 U1108 00Discrete Ordinates Gamma-Ray Transport Code System in Plane Geometry.
F-SCOREAbstractP00617 PCX86 00F-Score Nuclide ID Scoring Applications (Version 1.0)
FSKY4CAbstractC00771 PCX86 00Gamma Ray Skyshine Analysis Code.
FSX96AbstractD00190 MNYWS 00Collection of Continuous Energy Cross Section Libraries for MCNP Based on JENDL 3.2, JENDL, Fusion File and Dosimetry File.
FSXJ32AbstractD00244 MNYCP 00A Continuous Energy Cross Section MCNP Nuclear Data Library Based on JENDL-3.2.
FSXLIB-J3AbstractD00165 ALLCP 00MCNP continuous energy neutron cross section library based on JENDL-3. See DLC-190/FSX96 based on JENDL3.2.
FSXLIB-J33AbstractD00223 MNYCP 01Continuous Energy Neutron Cross Section Library for MCNP Based on JENDL 3.3.
FTFAbstractD00056 I0360 00Multigroup Neutron and Gamma-Ray Dose Transmission Factors for Concrete Slabs.
FUELSDATAAbstractP00446 C7600 00Code System to Model Verification Fuel Rod Data.
FURNACEAbstractC00615 C0740 00Code System for Neutronic Calculations in Three Dimension Toroidal Geometry.
G33-GPAbstractC00494 IBMPC 01Kernel Integration Code System - Multigroup Gamma-Ray Scattering Using the GP Buildup Factor.
G3-6EDAbstractC00075 C6600 00Kernel Integration Code System - Multigroup Gamma Ray Scattering.
G3-6EDAbstractC00075 I3033 00Kernel Integration Code System - Multigroup Gamma Ray Scattering.
GABASAbstractP00175 U1108 00A Code System for Generating Composite Time-Dependent Fission Produce Spectra.
GADJETAbstractC00115 C6600 00Monte Carlo Gamma-Ray Adjoint Energy Transport Code in Complex Three-Dimensional Geometry.
GADRAS-DRF-18.7.6AbstractP00610 PCX86 03Gamma Detector Response and Analysis Software–Detector Response Function.
GAINCALBAbstractP00056 I0360 00Determination of the Gain Used with Organic Scintillation Detect.
GAKER-KIRAAbstractC00813 C3600 00Energy Transfer of Protons in H2O or Polyethylene and Deuterons in D2O.
GALAXY-6AbstractP00098 I0370 00Neutron Multigroup Cross Section Processor.
GALE BWRAbstractC00335 U1100 00Boiling Water and Pressurized Water Reactors Gaseous and Liquid Effluents Radiological Assessment Code System.
GALE PWR & BWRAbstractC00335 I3033 00Boiling Water and Pressurized Water Reactors Gaseous and Liquid Effluents Radiological Assessment Code System.
GALE86AbstractC00506 MNYCP 02Calculation of Routine Radioactive Releases in Gaseous and Liquid Effluents from Boiling Water and Pressurized Water Reactors.
GAMANAbstractP00083 DP010 00Qualitative and Quantitative Evaluation of Ge(Li) Gamma-Ray Spectra.
GAMANALAbstractP00506 D0VAX 00Code System for Computerized Quantitative Analysis By Gamma-Ray Spectrometry.
GAMDAT-78AbstractD00083 I0370 00Library of Gamma-Ray Decay Data for 2055 Radionuclides.
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.
GAMLEG-JRAbstractP00116 F2307 00Multigroup Cross-Section Generator for Photon Transport Calculations.
GAMLEG-JRAbstractP00116 I3033 00Multigroup Cross-Section Generator for Photon Transport Calculations.
GAMLIBAbstractD00006 I0360 0099-Group Neutron Cross Sections for Use in the GAM Portion of the GGC Multigroup Cross Section Code.
GAMMAAbstractP00095 I0360 00Monte Carlo Code System for Calculating Efficiencies and Response Functions of NaI(Tl) Crystals for Gamma Rays from Thick Disk Sources.
GAMMOMAbstractC00135 ALLMF 00Gamma-Ray Moments Method Codes--GRMM and SPENCER.
GAMMOM-IAbstractC00226 I0360 00Gamma-Ray Moments Method Code System.
GAMMONAbstractD00071 ALLCP 00Activation Library for Fusion Reaction Application and Other Design Studies.
GAMTABAbstractD00032 I0360 00Radioactive-Decay Gamma-Rays Ordered by Energy and Nuclide.
GAMTOT78AbstractD00109 CY00I 00Compilation of Radioactive Decay and Capture Gamma Rays.
GAMX1AbstractP00209 I0370 00A Computer Code System for Evaluating Spectra Peak Areas.
GANAPOL-ABNTTAbstractM00011 MNYCP 00Analytical Benchmarks for Nuclear Engineering Applications: Case Studies in Neutron Transport Theory.
GANDR/SEMOVEAbstractC00765 PCX86 00Program for Calculating Derivatives of Processed Multigroup Nuclear Data by Discrete Differences.
GAPCON-THERMALAbstractP00499 C7600 00Code System to Calculate Fuel Steady State & Transient Behavior.
GARGAbstractD00073 C0000 0027-Group Neutron Cross Sections in Discrete Ordinates Format Generated with FIGERO (PSR-149) from ENDF-B Data.
GARLIBAbstractD00013 I3565 01Multigroup Resonance-Region Cross Sections for Use in Shielding Calculations.
GARLIBAbstractD00013 I7090 00Multigroup Resonance-Region Cross Sections for Use in Shielding Calculations.
GAROLAbstractP00033 I7090 00Calculation of Resonance Neutron Absorption in Two-Region Problems.
GASPARAbstractC00463 I3033 01Calculates Radiation Exposure to Man from Routine Air Releases of Nuclear Reactor Effluents.
GASPAR IIAbstractC00463 D0780 00Calculates Radiation Exposure to Man from Routine Air Releases of Nuclear Reactor Effluents.
GASSAbstractC00080 I7090 00Monte Carlo Calculation of Self Shielding by Encapsulated Gamma-Ray Sources.
GAUSS VAbstractP00045 I0360 00A Code system for Analysis of Gamma-Ray Spectra from Ge(Li) Spectrometers.
GAUSS VIIAbstractP00045 C0000 00A Code system for Analysis of Gamma-Ray Spectra from Ge(Li) Spectrometers.
GBANISNAbstractC00628 IRISC 00One-Dimensional Discrete Ordinates Transport Code System with Anisotropic Scattering with the GroupBand Option.
GCIAbstractP00421 IBMPC 00Generic Communications Index
GEAF-1AbstractD00158 D8810 00100 Group Cross Sections for Neutron Activation.
GECINXAbstractP00193 H6000 00A Code System for Collapsing Multigroup Cross Sections in CCCC Format.
GEFAbstractP00564 PCX86 02A GEneral description of the Fission process.
GEFAbstractP00564 PCX86 03A GEneral description of the Fission process.
GELI2/SPAN2AbstractP00094 I0360 00Calculation of Nuclide Abundaces from Multichannel Gamma-ray Spectra.
GEMAbstractP00540 PC586 00Monte-Carlo Code for Simulating a Decaying Process of an Excited Nucleus.
GENII 2.10AbstractC00737 PCX86 02Environmental Radiation Dosimetry Software System.
GENII-LIN 2.1AbstractC00728 PC586 01GENII-LIN Multipurpose Health Physics Code System with a New Object-Oriented Interface, Release 2.0.
GENP-2AbstractC00575 ALLMF 00Generalized Perturbation Theory Code System.
GENRDAbstractP00040 C6600 00Free Format Card Input Processor.
GENRDAbstractP00040 I0360 00Free Format Card Input Processor.
GERESAbstractP00241 I0370 00A Code to Produce Cross-Section Libraries for ANISN Based on Heterogeneous Fast Reactor Cell Calculations Using MC2II Data.
GES_MCAbstractC00742 PC586 00Gamma-electron Efficiency Simulator, Version 3.1
GETOUTAbstractC00461 C0176 00A Computer Code System for Predicting One-Dimensional Radionuclide Decay Chain Transport through Geologic Media.
GFX-GAMIXAbstractC00397 I3033 00A Spherical Harmonics Code System for Evaluation of Terrestrial Gamma-Radiation Fields.
GGC-3AbstractP00012 I3565 00Multigroup Cross Section Code System for Use in Diffusion and Transport Codes.
GGC-3 & GGC-4AbstractP00012 I3675 00Multigroup Cross Section Code System for Use in Diffusion and Transport Codes.
GGC-4AbstractP00012 U1108 00Multigroup Cross Section Code System for Use in Diffusion and Transport Codes.
GGG-GPAbstractC00564 IBMPC 00Kernel Integration Code System - Multigroup Gamma-Ray Scattering Using the GP Buildup Factor.
GGTC-ENELAbstractP00128 I0360 00Code System for Producing Few-Group Neutron Cross Sections from Multigroup Data Libraries.
GICX40AbstractD00092 ALLCP 00Coupled 42-Neutron, 21-Gamma-Ray Group Cross Sections for 40 Elements in Group Independent Form for Fusion Reactor Calculations.
GIFTAbstractP00124 C0076 00A Combinatorial Geometry Code System with Model Testing Routines.
GIFTAbstractP00124 D0VAX 00A Combinatorial Geometry Code System with Model Testing Routines.
GIFTAbstractP00124 U0000 00A Combinatorial Geometry Code System with Model Testing Routines.
GIPAbstractP00229 IBMPC 00Group-Organized Cross-Section Input Program.
GIRAFFEAbstractP00304 I3033 00General Isotope Release Analysis For Failed Elements.
GLUCSAbstractP00192 D0VAX 00A Generalized Least-Squares Code System for Updating Cross Section Evaluations with Correlated Data Sets.
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, Version gn9cp8.
GNOMERAbstractC00625 MNYCP 01Multigroup 3-Dimensional Neutron Diffusion Nodal Code System with Thermohydraulic Feedbacks.
GOFRRAbstractP00127 I0360 00Generator of Graphical Output of DOT and ANISN Fluxes and Reaction Rates.
GRACE-IIAbstractC00026 I3675 00Gamma Ray Kernel Integration Dose Rate and Heating Code-Cylinders and Spheres.
GRASS-SSTAbstractP00489 MNYCP 00Code System to Predict Fission-Gas Release & Fuel Swelling.
GREAT-GRASSAbstractC00143 I3675 00Monte Carlo Radiation Transport Code Systems for Fallout Shielding.
GRENADEAbstractC00516 C1787 00Green's Function Nodal Algorithm for the Diffusion Equation.
GRENADEAbstractC00516 D0780 00Green's Function Nodal Algorithm for the Diffusion Equation.
GRESS 3.0AbstractP00231 MFMWS 02Gradient Enhanced Software System.
GRETELAbstractP00100 I0370 00Analyzer and Processor of Ge(Li) Gamma-Ray Spectrometric Data.
GRFPAKAbstractP00478 I0360 00Code System to Plot CORTES FEM Results.
GROUP STRUCTUREAbstractD00156 ALLCP 00Standard Energy Group Structures Of Cross Section Libraries For Reactor Shielding, Reactor Cell Fusion Neutronics Applications: VITAMIN-J, ECC0-33, ECC0-2000.
GROUPSTRUCTURESAbstractD00274 MNYCP 00GROUPSTRUCTURES, VITAMIN-J, XMAS, ECCO-33, ECCO2000 Standard Group Structures
GROUPXSAbstractP00246 C0740 00Processing of Double-Differential Cross Sections in the New ENDF-VI Format.
GRPANLAbstractP00321 D0VAX 00Code System for Analyzing Ge and Alpha-Particle Detector Spectra.
GRSACAbstractC00774 PCX86 00Graphite Reactor Severe Accident Code.
GRTUNCL3DAbstractC00721 MNYCP 01Code to Calculate Semi-Analytic First Collision Source and Uncollided Flux.
GRUCONAbstractP00615 MNYCP 00Data Processing for Evaluated Working libraries (transport and shielding)
GT2R2AbstractP00483 ALLMF 00Code System to Calculate Fuel Rod Thermal Performance.
GUI2QAD-3DAbstractC00697 PC586 01Point Kernel Code System for Neutron and Gamma-Ray Shielding Calculations in Complex Geometry, Including a Graphical User Interface.
HAARM-3AbstractP00401 CDCMF 00Aerosol Behavior Log-Normal Distribution Model.
HABIT 1.1AbstractC00665 IBMPC 01Code System for Evaluation of Control Room Habitability.
HADOCAbstractC00452 U1100 00Calculates External and Inhalation Doses from Acute Radionuclide Releases on the Hanford Site.
HALLMARKAbstractD00005 I0360 00Discrete Ordinates and Monte Carlo Results of Neutron and Secondary Gamma-Ray Transport in Air-Over-Ground Geometry.
HAMAbstractC00267 U1108 00Monte Carlo Multigroup Neutron and Photon High Altitude Transport Code System.
HARADAbstractC00387 I0360 00Calculation of Daughter Concentrations in Air Following the Atmospheric Release of a Parent Radionuclide.
HASSANAbstractP00593 I0370 00Time-Dependent Temperature Distribution and Stress and Strain in HTR Fuel Pins.
HATCHES-19AbstractD00206 PC586 02Database for Radiochemical Modelling.
HAUSER*5AbstractP00152 U0000 00Code System for Calculating Nuclear Cross Sections.
HEATING 7.3AbstractP00199 MNYCP 06Multidimensional, Finite-Difference Heat Conduction Analysis Code System, Versions 7.2i and 7.3.
HEATKAUAbstractC00805 PCX86 00HEATKAU Program.
HECTR 1.5+
USSO
AbstractP00457 CY000 00Hydrogen Event Containment Response Code System.
HEITLERAbstractP00004 I7030 00Cross Section Generator.
HELLOAbstractD00058 I0360 0047 Neutron, 21 Gamma-Ray Group Cross Sections for Radiation Transport for Neutron Energies up to 60 MeV.
HEPROWAbstractC00799 MNYCP 00Unfolding of Pulse Height Spectra Using Bayes Theorem and Maximum Entropy Method.
HERADAbstractC00444 CY00I 00Three-Dimensional Monte Carlo Computer Code System for Calculating Radiation Damage from Ion Beams.
HERMES-KFAAbstractC00687 MNYWS 00Monte Carlo Code System for High-Energy Radiation Transport Calculations.
HEXAB-3DAbstractC00593 I0370 00Three-Dimensional Few-Group Coarse Mesh Diffusion Code for Neutron Physics Calculation of Reactor Core in Hexagonal Geometry.
HGSYSTEMAbstractM00021 MNYCP 00Atmospheric Dispersion for Ideal Gases and Hydrogen Fluoride (HF)
HGSYSTEMUF6AbstractC00832 MNYCP 00Model for Simulating Dispersion due to Atmospheric Release of UF6.
HIC-1AbstractC00249 I0360 00Monte Carlo Code System for Calculating Heavy Ion Reactions at Energies > 50 MeV/Nucleon.
HILOAbstractD00087 I0370 0066 Neutron, 21 Gamma-Ray Group Cross Sections for Radiation Transport for Neutron Energies up to 400 MeV.
HILO2KAbstractD00220 MNYCP 00Coupled 83 Neutron, 22 Photon Group Cross Sections for Neutron Energies Up to 2 GeV.
HILO86AbstractD00119 I0360 0066 Neutron, 22 Gamma-Ray Group Cross Sections for Radiation Transport for Neutron Energies Up to 400 MeV.
HILO86AbstractD00119 PC386 0166 Neutron, 22 Gamma-Ray Group Cross Sections for Radiation Transport for Neutron Energies Up to 400 MeV.
HILO86RAbstractD00187 ALLCP 0066 Neutron, 22 Gamma-Ray Group Cross Sections for Radiation Transport for Neutron Energies Up to 400 MeV.
HIMACAbstractM00001 MNYCP 02Experimental Data of Neutron Yields from Thick Targets Bombarded by 100 to 800 MeV / Nucleon Heavy Ions.
HORNAbstractC00568 I3083 00A Computer Code To Analyze The Gas-Phase Transport of Fission Products In Reactor Cooling System Under Severe Accidents.
HOTSPOT 3.0.2AbstractM00009 IBMPC 03Health Physics Code System for Evaluating Accidents Involving Radioactive Materials.
HPICEAbstractD00007 I0360 05Evaluated Photon Interaction Library, ENDF/B File 23 Format.
HPPOS 1.5AbstractD00173 IBMPC 00Health Physics Position Database.
HPPOS V2AbstractD00173 IBMPC 01Health Physics Positions (HPPOS) Data Base Based on Current 10 CFR 20.
HSI-DRGAbstractP00435 IBMPC 00Code System for Use with Human System Interface Design Review Guidelines.
HUGOAbstractD00099 I3033 00Photon Interaction Data in ENDF/B-V Format.
HUGO VIAbstractD00146 I3033 00Photon Interaction Data in ENDF/B-VI Format. PHOTB6 in DLC-179/ENDLIB-97 is an updated version of these data.
HYACINTHAbstractC00294 I0360 00Fast Heavy Isotope Point Burnup and Decay Code System - Analytical Solution.
HYPERMETAbstractP00101 C3800 00Gamma-Ray Spectra Analyzer Germanium Detector.
HYPERMETAbstractP00101 F150F 00Gamma-Ray Spectra Analyzer Germanium Detector.
HYPERMETAbstractP00101 I0360 00Gamma-Ray Spectra Analyzer Germanium Detector.
ICARAbstractP00291 IPCAT 00A Code For Combinatorial Calculation of Level Densities.
ICOMAbstractC00651 PC386 00Code System for Calculating Ion Track Condensed Collision Model.
IDCAbstractC00384 I0360 00ICRP Dosimetric Calculational System.
IEAF-2001AbstractD00217 MNYCP 00Intermediate Energy Activation File - 2001.
IERAbstractP00024 I3675 00A Gauss-based Quadrature Formula Applied to Sievert's Integral. An Exponential Integral Routine.
IMPACTS-BRC2.1AbstractC00666 IBMPC 00Code System for Analysis of Potential Radiological Impacts.
IMPORTANCEAbstractP00407 I0370 00FTA Basic Event & Cut Set Ranking.
INAPAbstractC00235 U1108 00Improved Neutron Activation Prediction Code Systems.
INDOSAbstractC00236 DP010 00Conversational Computer Code Systems to Implement ICRP-10-10A Models for Estimation of Internal Radiation Dose to Man.
INDOSE V2.1.1AbstractC00720 PC586 00Internal Dosimetry Code System Using Biokinetics Models
INDRAAbstractC00303 I0360 00A Modular System for Calculating the Neutronics and Photonics Characteristics of a Fusion Reactor Blanket.
INFLTBAbstractP00313 ALLCP 00Gamma-Ray Absorption Coefficient Calculation.
INGDOSAbstractC00408 DP010 00A Conversational Code System Designed to Implement NRC Reg-Guide 1.109 Models for Estimation of Annual Doses from Ingestion of Atmospherically Released Radionuclides in Foods.
INGENAbstractP00207 C0000 00A General-Purpose Mesh Generator for Finite Element Codes.
INREM IIAbstractC00392 I3033 00Computer Implementation of Recent Models for Estimating the Dose Equivalent to Organs of Man from an Inhaled or Ingested Radionuclide.
INREM/EXREMAbstractC00185 I0360 00Beta and Gamma Radiation Environmental Dose Code Systems.
INTERTRAN IAbstractC00473 ALLMF 00A Code System for Assessing the Impact from Transporting Radioactive Material.
INTRIGUE-IIAbstractP00054 I0360 00Logarithmic and Semilogarithmic CALCOMP Plot Routines.
INTRUDE-ANSAbstractC00539 D8810 00A Repository Intrusion Risk Evaluation Code.
INVENTAbstractC00540 D8810 00A Radionuclide Inventory and Hazard Index Code.
IODESAbstractC00365 I0360 00A Code System for Calculating the Estimation of Dose to the World Population from Releases of Iodine-129 to the Environment.
IONMIGAbstractC00526 ALLMF 00Code System for Radionuclide Migration Calculations.
IRAN-LIBAbstractD00159 IBMPC 00A P-3 Coupled Neutron-Gamma Cross Section Library in ISOTXS For Use with ANISN/PC (CCC-514).
IRDAMAbstractC00524 IPCXT 00Interactive Rapid Dose Assessment Model.
IRDF-2002AbstractD00229 MNYCP 01The International Reactor Dosimetry File.
IRDF82AbstractD00094 I0360 00International Reactor Dosimetry Data.
IRDF-90AbstractD00161 ALLCP 01The International Reactor Dosimetry File.
I-R-MANAbstractD00050 ALLCP 00Photon Interaction Data on ICRP Reference Man.
IRPHE-VENUS-RECYCLEAbstractD00263 MNYCP 00Plutonium Recycling Physics Project Critical Experiments.
IRRAS 4.16
USSO
AbstractP00386 IBMPC 04Code System to Calculate Integrated Reliability and Risk Analysis.
ISOGEN IIAbstractC00055 I3675 00Radioisotope Generator Code.
ISO-PC 2.1AbstractC00636 IBMPC 01Kernel Integration Code System for General Purpose Isotope Shielding Analyses.
ITER-2AbstractP00148 C0000 00Codes for Unfolding Activation Detector Data and Pulse Height Spectra.
ITS6
FEDC
AbstractC00792 PCX86 00Integrated TIGER Series of Coupled Electron/Photon Monte Carlo Transport Codes System.
JASMINE V.3AbstractC00795 MNYCP 00JAEA Simulator for Multiphase INteractions and Explosions.
JDL-IMPORTANCEAbstractM00005 MNYCP 00Adjoint Function: Physical Basis of Variational & Perturbation Theory in Transport & Diffusion Problems.
JDL-REACTOR-KINAbstractM00006 MNYCP 00Nuclear Reactor Kinetics and Control.
JDL-THERMODYNAMAbstractM00007 MNYCP 00Thermodynamics: Frontiers and Foundations.
JENDL/D-99AbstractD00204 MNYCP 00JENDL Dosimetry File 99.
JENDL-1AbstractD00070 ALLCP 00Japanese Evaluated Nuclear Data Library.
JENDL-2AbstractD00122 FM380 00Japanese Evaluated Neutron Cross Section Data in ENDF/B-IV Format.
JFSAbstractD00111 I3033 0070 Group Neutron Fast Reactor Cross Section Set and 25 Group Neutron Fast Reactor Cross Section Set.
JFS3J2AbstractD00108 FM200 0070 Group Neutron Fast Reactor Cross Section Set Based on JENDL-2B.
JIMCOFAbstractD00078 F2307 00Multigroup Constants fFle Based on ENDF/B IV.
JN-METD 2&1AbstractC00208 I0370 00Neutron Transport Code System with Isotropic Scattering, Bare Slabs and Homogeneous Slabs (JN Method 1), Multilayer Slabs (JN Method 2).
JPHYDROAbstractP00594 I0360 00Voids and Flow Velocity in Steady-State BWR System.
K009AbstractC00062 I7090 00Solid Angle Integration Charged Particle Penetration Code.
K019AbstractC00100 I0360 00Shield Thickness Calculation Program for Space Vehicles.
KAMCCOAbstractC00325 I0370 00Three-Dimensional Time Dependent Monte Carlo Code System for Fast Neutron Physics Problems.
KAOS/LIB-VAbstractD00160 CY000 00A Library of Nuclear Response Functions Generated by KAOS-V Code From ENDF/B-V and Other Data Files.
KAOS-VAbstractP00306 CY000 00An Evaluation Tool For Neutron Kerma Factors and Other Nuclear Responses.
KAP-VIAbstractC00094 U1108 00Kernel Integration Code System in Complex Geometry.
KASYAbstractC00814 I0370 003-D Homogeneous Neutron Diffusion in X-Y-Z, R-Theta, Hexagonal-Z Geometry by Synthesis Method.
KCUTAbstractP00584 IBMPC 00Code to Generate Minimal Cut Sets for Fault Trees.
KDDKAbstractD00061 I0360 00Measured Results of Delayed Beta- and Gamma-Ray Spectra due to Thermal-Neutron Fission of U-235.
KDLIBEAbstractC00124 I3675 00Kernel-Diffusion Shielding Analysis System.
KEDAK3AbstractD00141 I0370 00Evaluated Neutron Nuclear Data for Reactor Physics Calculations.
KENO2MCNPAbstractP00541 PC586 00Conversion of Input Data between KENO V.a and MCNP File Formats, Version 5L.
KERMALAbstractD00142 ALLCP 00Neutron and Gamma-Ray Kerma Factors Based on LLNL Nuclear Data Files.
KERNELAbstractC00672 IBMPC 00Monte Carlo Code System for Electron (Positron) Dose Kernel Calculations.
KFIXAbstractP00409 C7600 00Code System to Calculate Transient 2-Dimensional 2-Fluid Flow Dynamics.
KFIX 3DAbstractP00383 C7600 00Code System to Calculate Three-Dimensional Extension Two-Phase Flow Dynamics.
KICHE 1.3AbstractC00796 PCX86 00Kinetics of Iodine Chemistry in the Containment of LWRs under Severe Accident Conditions.
KIMAbstractC00376 I3033 00A Two-Dimensional Monte Carlo Code System for Linear Neutron Transport Calculations.
KORIGENAbstractC00457 I3033 00A Modification of the Isotope Generation and Depletion Code System ORIGEN. CCC-702/ORIGEN-ARP is recommended for new ORIGEN users.
KRONICAbstractC00229 I0360 00Calculation of Annual Average External (Beta and Gamma Radiation) Doses from Chronic Atmospheric Releases of Radionuclides.
KRONICAbstractC00229 U1108 00Calculation of Annual Average External (Beta and Gamma Radiation) Doses from Chronic Atmospheric Releases of Radionuclides.
KUXAbstractC00515 ALLCP 00Medical X-Ray Shielding Calculation.
KX-RAYAbstractD00021 I0360 00Evaluated X-ray Cross Section Library.
L26P3S34AbstractD00112 IBMMF 00ENDL 26-Group up to P3 Library Prepared by SUPERTOG for 34 Materials.
LA100AbstractD00168 ALLCP 00Evaluated Nuclear Data Library for Transport Calculations Involving Incident Neutrons and Protons of Energy Up to 100 MeV.
LABAN-PELAbstractC00611 IMFPC 00A Two-Dimensional, Multigroup Diffusion, High-Order Response Matrix Code.
LADTAP IIAbstractC00363 C7600 00Code System for Calculating Radiation Exposure to Man from Routine Release of Nuclear Reactor Liquid Effluents.
LADTAP IIAbstractC00363 D0780 00Code System for Calculating Radiation Exposure to Man from Routine Release of Nuclear Reactor Liquid Effluents.
LADTAP IIAbstractC00363 I3033 00Code System for Calculating Radiation Exposure to Man from Routine Release of Nuclear Reactor Liquid Effluents.
LAFPX-VAbstractD00054 C0000 01A Multigroup Reaction Cross-Section Collapsing Code and Library of 154-Group Fission-Product Cross Sections.
LAFPX-VAbstractD00054 C0000 02A Multigroup Reaction Cross-Section Collapsing Code and Library of 154-Group Fission-Product Cross Sections.
LAHET 2.8AbstractC00696 MFMWS 00Code System for High Energy Particle Transport Calculations.
LAHIMACKAbstractD00128 I0360 00A Multigroup Library of Neutron and Gamma Cross Sections and Response Functions in the Energy Range up to 800 MeV.
LAPHANOAbstractP00020 C6600 00PO Multigroup Photon Production Matrix and Source Vector Code for ENDF Data.
LAPHANOAbstractP00020 I0360 00PO Multigroup Photon Production Matrix and Source Vector Code for ENDF Data.
LAPUR6
USSO
AbstractP00395 PC586 02BWR Core Stability Measurements.
LAS CRUCES
USSO
AbstractD00194 ALLCP 00Las Cruces Trench Site Database, Vadose Model.
LASERAbstractC00344 I0360 00A One-Dimensional, Neutron-Thermalization, Lattice-Cell Program Based on MUFT and THERMOS.
LAZYAbstractP00595 I0360 00General Experimental Data Processing Program.
LEAFAbstractC00312 C6600 00Fission Product Release Calculator-From a Reactor Containment Building for Arbitrary Radioactive Decay Chains.
LEAP-ADDELTAbstractP00138 I0360 00Multigroup Thermal Neutron Scattering Data Generator for Hydrogen in Light Water and Deuterium in Heavy Water.
LEBCAbstractC00052 I7090 00Electron Bremsstrahlung Code.
LEGENDRE FUNCTIAbstractP00108 I0360 00Legendre Functions of the First Kind and Legendre Polynomials.
LENDLAbstractD00034 I0360 02Livermore Evaluated Neutron and Secondary Gamma-Ray Production Cross-Section Library in ENDF/B-IV Format.
LENDL VAbstractD00120 I0360 00Lawrence Livermore National Laboratory Evaluated Nuclear Data Library in ENDF-V Format.
LEOPARDAbstractC00343 C0000 00A Spectrum-Dependent Non-Spatial Fuel Depletion Code System.
LEOPARDAbstractC00343 IBMPC 00A Spectrum-Dependent Non-Spatial Fuel Depletion Code System.
LEPAbstractD00001 I0360 02Cascade and Evaporation Particle Results from Low-Energy Intranuclear Cascade Calculations.
LEPRICONAbstractP00277 I3033 01PWR Pressure Vessel Surveillance Dosimetry Analysis System.
LEPRICONAbstractP00277 IRISC 00PWR Pressure Vessel Surveillance Dosimetry Analysis System.
LG-HAbstractC00087 I7090 00Ray Analysis Cylindrical Duct Kernel Code for Neutrons and Gamma Rays.
LGH-GAbstractC00239 I0360 00Calculation of Gamma Radiation through Partially Shielded Gaps (Buildup Factor Method in Taylors Approximation).
LHSAbstractP00394 PC386 00Code System to Generate Latin Hypercube and Random Samples.
LHSAbstractP00394 SUN05 00Code System to Generate Latin Hypercube and Random Samples.
LIB123AbstractD00153 ALLCP 00AMPX-II P3 123-Group Neutron Cross Section Master Interface Library.
LIBMAKAbstractP00087 I0360 00ANISN-Type Binary Data Processing Code System.
LIE-PNAbstractC00816 I0360 00Pn Neutron Transport in Radial Geometry Cell with Source Problems Calculation.
LINEDOSEAbstractC00468 IBMPC 00A Line Source Shielding Code for Personal Computers.
LINSEDAbstractC00673 I0360 001D Multireach Sediment Transport Model
LIONSAbstractC00247 I0360 00Calculation of Fission Product Inventory, Gamma-Ray Dose Rates and Gamma-Ray Doses by Kernel Integration.
LOGNORMLAbstractP00307 IPCAT 00Lognormal Probability Analysis Code System for Estimating Doses in Epidemiologic Studies.
LOOM-PAbstractP00153 F2307 00A Finite Element Mesh Generation Code System with On-Line Graphic Display.
LOUHI82AbstractP00236 U1108 00General Purpose Unfolding Program with Linear and Nonlinear Regularizations.
LPGSAbstractC00385 I3033 00Code System for Calculating Radiation Exposure Resulting from Accidental Radioactive Releases to the Hydrosphere.
LPPCAbstractC00051 I7090 00Proton Penetration Code.
LPSCAbstractC00064 I7090 00Proton Penetration Code - Multilayer Slab Geometry.
LPTAUAbstractP00340 MNYCP 00Quasi-Random Sequence Generators.
LRSPCAbstractC00050 I7090 00Range and Stopping Power Calculator.
LSHINSEAbstractC00554 IBMPC 00Calculates Flux and Dose Rate from the Scattering of Radiation in Air.
LSL-M2AbstractP00233 D6220 00Least-Squares Logarithmic Adjustment of Neutron Spectra.
LSL-M2AbstractP00233 IBMPC 00Least-Squares Logarithmic Adjustment of Neutron Spectra.
LSMOD-GLSMODAbstractP00342 IBMPC 00A Least-Squares Computational Tool Kit.
LSVDCAbstractC00053 I7090 00Space Vehicle Dose Calculation.
LSVDCAbstractC00053 I7090 01Space Vehicle Dose Calculation.
LTCAbstractP00329 IBMPC 00LMR Transient Calculation Code System (version 5).
LUIN-IIAbstractC00220 C6600 00Analytical Straight-Ahead Transport Code System-Calculation of Cosmic-Ray Spectra, Fluxes and Ionization in the Earth's Atmosphere.
LUMPAbstractD00089 I0360 00Evaluated Lumped Fission Product Cross Sections for Fast Reactor Analysis--Based on ENDF/B-V Data.
MACK-IVAbstractP00132 I3691 00Calculation of Nuclear Response Functions from Nuclear Data in ENDF Format.
MACKLIBAbstractD00029 I3675 00100 Group Neutron Kerma Factors and Reaction Cross Sections Generated by MACK from Data in ENDF Format.
MACKLIB-IV-82AbstractD00060 I0360 01A Library of Nuclear Response Functions Generated with the MACK-V Computer Program from ENDF/B-IV.
MADONNAAbstractC00425 I0370 00Two-dimensional Neutron Streaming Coupled Removal-Diffusion-Albedo-Transport Code System.
MAEROSAbstractP00466 C7600 00Code System for Multicomponent Aerosol Time Evolution.
MAGIKAbstractC00359 I0360 00A Monte Carlo Code System for Computing Induced Residual Activation Dose Rates.
MAGNAAbstractC00158 C3600 00Multi-Source Gamma-Ray Kernel Integration Code System.
MAINTAINAbstractP00067 I0360 00Code System for Use in Maintaining and Revising Card Image Files on Tape.
MANYFILEAbstractP00068 I0360 00Utility Routine - Manipulation of Data Sets Between Various I-O Devices.
MAPAbstractC00150 I3675 00Kernel Integration Code System in Complex Geometry with Special Application to Surface Sources Determined by Discrete Ordinates Calculations.
MARCH2AbstractP00473 CDCMF 00Code System to Model LWR Meltdown Accident Response.
MARCOPOLOAbstractP00225 I0360 00Code System for Calculating the Radial and Axial Neutron Diffusion Coefficients in One-Group and Multigroup Theory.
MARC-PNAbstractC00311 D8810 00A Neutron Diffusion Code System with Spherical Harmonics Option.
MARD 4.16AbstractP00448 IBMPC 00Models And Results Database System.
MARIA SYSTEMAbstractP00359 D6000 00Code System to Calculate Cross Sections for PWR Fuel Assembly Calculations.
MARINRADAbstractC00503 C1785 00Code System Model for Assessing the Consequences of Release of Radioactive Material into the Oceans.
MARLOWE 15BAbstractP00137 MNYCP 08Computer Simulation of Atomic Collisions in Crystalline Solids (Version 15).
MARMERAbstractC00579 D8350 00A Flexible Point-Kernel Shielding Code System.
MARMERAbstractC00579 PC486 00A Flexible Point-Kernal Shielding Code System.
MARSAbstractP00117 I0360 00Collection of Computer Codes for Manipulating Multigroup Cross Section Libraries in AMPX or CCCC Formats.
MARTHAAbstractP00232 I0360 00Monte Carlo Response Function Calculation for Sodium Iodide Photon Detectors.
MARVIKEN-JIT
OECD
AbstractD00269 MNYCP 00Marviken Full Scale Jet Impingement Tests Experiments.
MASSAbstractD00025 I0360 01Atomic Mass Evaluation.
MATADORAbstractC00689 CDCMF 00Radionuclide Behavior in Containments.
MATEXPAbstractP00059 I0360 00Matrix Exponential Method Applied to Systems of Ordinary Differential Equations.
MATJEFF31.BOLIBAbstractD00242 MNYCP 00Fine-Group Cross Section Library Based on JEFF3.1 for Nuclear Fission Applications.
MATXS1AbstractD00114 C0000 0030-Group Neutron, 12-Group Photon Cross Sections from ENDF/B-IV in MATXS Format.
MATXS10AbstractD00176 ALLCP 0030-Group Neutron, 12-Group Photon Cross Sections from ENDF/B-VI in MATXS Format.
MATXS11AbstractD00177 ALLCP 0080-Group Neutron, 24-Group Photon Cross Sections from ENDF/B-VI in MATXS Format.
MATXS175/42-JEAbstractD00151 D8810 00JEF/EFF Based VITAMIN-J 175 Neutron, 42 Photon Multigroup Data Library in MATXS Format.
MATXS5AAbstractD00115 C0000 0030-Group Neutron, 12-Group Photon Cross Sections from ENDF/B-V in MATSX Format.
MATXS6AAbstractD00116 C0000 0080-Group Neutron, 24-Group Photon Fast-Reactor Cross Section from ENDF/B-V in MATXS Format.
MATXS70-JEF87AbstractD00148 D8810 00JEF/EFF Based 70 Group Neutron Data Library in MATXS Format.
MATXS7AAbstractD00117 C0000 0069-Group Thermal-Reactor Neutron Cross Section Data from ENDF/B-V in MATXS Format.
MATXSLIBJ33AbstractD00258 MNYCP 01JENDL-3.3 Based, 175 N-42 Photon Groups (VITAMIN-J) MATXS Library for Discrete Ordinates Multi-Group.
MATXUFAbstractP00130 I0360 00On-Line Derivative Method, Spectrum Unfolding Code System for NE-213 Liquid Fast Scintillation Proton Recoil Data.
MAVRACAbstractC00023 I7090 00Model Astronaut and Vehicle Radiation Analysis Code.
MAX-XTREMEAbstractP00001 C0000 00Generalized Several-Constraint LaGrange Multiplier.
MAZE IIAbstractP00041 U1108 00Spectral Unfolding Code.
MAZE-1AbstractP00041 C6600 00Spectral Unfolding Code.
MC**2-2AbstractP00350 SUN05 01Code System for Calculating Fast Neutron Spectra and Multigroup Cross-sections from ENDF/B Data (November 2000 Version).
MC**2-3AbstractP00577 MNYCP 00Multigroup Cross Section Generation Code for Fast Reactor Analysis.
MC**2-3 EXEAbstractP00577 MNYCP 01Multigroup Cross Section Generation Code for Fast Reactor Analysis.
MCART
USUNV
AbstractC00809 PCX86 00Solve the Time-Dependent Neutron Transport Equation.
MCB1CAbstractC00719 MNYWS 00Monte-Carlo Continuous Energy Burnup Code System.
MCB63NEA.BOLIBAbstractD00216 MNYCP 00ENDF/B-VI Release 3 Cross Section Library for Use with the MCNP Monte Carlo Code.
MCFLAREAbstractC00093 I7090 00Monte Carlo Code to Simulate Solar Flare Events and Estimate Probable Doses Encountered on Interplanetary Missions.
MCJEF22NEA.BOLIBAbstractD00203 MNYCP 01JEF 2.2 Cross Section Library for the MCNP Monte Carlo Code.
MCJEFF3.1NEAAbstractD00228 MNYCP 00Neutron Cross Section Library Based on JEFF3.1 for Use with MCNP.
MCNP6.2
810
AbstractC00850 MNYCP 00Monte Carlo N–Particle Transport Code System and Data Libraries.
MCNP6.2-EXE
810
AbstractC00850 MNYCP 01Monte Carlo N–Particle Transport Code System and Data Libraries.
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 01Transport Code System with Digital Signal Processing based on MCNP4A EXE only.
MCNPX-POLIMI
810
AbstractC00791 MNYCP 00Monte Carlo N-Particle Transport Code System To Simulate Time-Analysis Quantities.
MCNPX-POLIMI-EXE
810
AbstractC00791 MNYCP 01Monte Carlo N-Particle Transport Code System To Simulate Time-Analysis Quantities EXE Only
MCRACAbstractC00562 IBMPC 00Multiple Cycle Reactor Analysis Code.
MCRTOFAbstractC00435 FM200 00Monte Carlo Code System for Calculation of Multiple Scattering of Neutrons in the Resonance Region.
MCRTOFAbstractC00435 I0360 00Monte Carlo Code System for Calculation of Multiple Scattering of Neutrons in the Resonance Region.
MCUNED
810
AbstractC00804 PCX86 00MCNPX Extension for Using Light Ion Evaluated Nuclear Data Library.
MCVIEWAbstractP00202 FM780 00View Factor Calculation for Three-Dimensional Geometries.
MECC-7AbstractC00156 I0360 00Medium-Energy Intranuclear Cascade Code System.
MEDUSA-IBAbstractC00505 HM200 00One-Dimensional Lagrangian Code for Plasma Hydrodynamic Analysis of a Fusion Pellet Driven by Ion Beams.
MEDUSA-PIJAbstractC00349 F2307 00One-Dimensional Laser Fusion Analyzer (Including Neutron Heating Effect) Collision Probability Method.
MEGAAbstractC00839 MNYCP 00MEGA: Mechanistic and Engineering Fission Gas Release Prediction Model for UO2 Fuel
MENDL-2PAbstractD00207 MNYCP 00Proton Reaction Data Library for Nuclear Activation (Medium Energy Nuclear Data Library.)
MENSLIBAbstractD00084 I0370 0060 Group, P5, Cross Sections in DTF-IV for Transport Calculations for Neutrons with Energies Up to 60 MeV.
MERCURE 4-82AbstractC00142 I3033 00Three-Dimensional Code System for Integrating Multigroup Line-of-Sight Attenuation Kernels by Monte Carlo Techniques.
MESAAbstractP00223 I3033 00Non-Linear Least Squares Spectral Analysis.
MESODIF-IIAbstractC00498 D0780 00A Variable Trajectory Plume Segment Model to Assess Ground-Level Air Concentrations and Depositions of Routine Effluent Releases from Nuclear Power Facilities.
MESOIAbstractC00497 D0780 00Interactive Mesoscale Lagrangian Puff Dispersion Model with Deposition and Decay. See CCC-677/MESORAD.
MESORAD 1.4AbstractC00677 D0VAX 00Code System for Emergency Response Dose Assessment.
MESYSTAbstractC00706 MNYWS 00Code System to Simulate 3D Tracer Dispersion in Atmosphere.
METDAbstractP00197 DGMV1 00Computer Code Systems for Use with Meteorological Data.
METDAbstractP00197 I3033 00Computer Code Systems for Use with Meteorological Data.
MEVDPAbstractC00157 C6600 00Primary Radiation Transport Code System - Complex Geometry - Computerized Anatomical Model Man.
MGA8AbstractP00542 MNYCP 00Code System to Determine Pu Isotope Abundances from Multichannel Analyzer Gamma Spectra.
MGCLIBAbstractD00118 FM380 00137 and 26 Neutron Multigroup Cross Section Library with the Bondarenko Type Shielding Table.
MICAPAbstractP00261 I3033 00A Monte Carlo Code System for Analysis of Ionization Chamber Responses.
MICROX-2AbstractP00374 MNYCP 02Code System to Create Broad-Group Cross Sections with Resonance Interference and Self-Shielding from Fine-Group and Pointwise Cross Sections.
MIGROS3AbstractP00265 I0370 00A Code for the Generation of Group Constants for Reactor Calculations from Neutron Nuclear Data in KEDAK Format.
MILDOSAbstractC00398 C0000 00Calculation of Radiation Doses from Uranium Recovery Operations.
MILDOS-AREAAbstractC00608 IBMPC 00Calculation of Radiation Dose from Uranium Recovery Operations for Large-Area Sources.
MINETAbstractP00490 CY000 00Momentum Integral Network Method for Thermal-Hydraulic Systems Analysis.
MINIGALAbstractP00180 I3033 00Neutron Cross Section Processing System for Calculating Average Values from Data in the Standard United Kingdom Nuclear Data Library Format.
MINTEQAbstractP00494 DVX11 00Code System to Model Aqueous Geochemical Equilibria.
MINXAbstractP00105 C6600 00Multigroup Interpretation of Nuclear X-Sections from ENDF/B Standard CCCC-III Interface Formats.
MINXAbstractP00105 I0360 00Multigroup Interpretation of Nuclear X-Sections from ENDF/B Standard CCCC-III Interface Formats.
MISSIONARYAbstractP00114 I0360 00ENDF/B to NDL Data Format Converter.
MIXENAbstractP00318 IRISC 00Code System to Replace Files 4 and 6 of ENDF-6 with Files 4 and 5 of ENDF/B-IV.
MKENO-DARAbstractC00513 FM380 00Direct Angular Representation Monte Carlo Code for Criticality Safety Analysis
MMCRAbstractC00441 FM200 00Multigroup Monte Carlo Neutron and Photon Transport Code.
MMRWAbstractM00018 MNYCP 00Canadian and Early British Energy Reports on Nuclear Reactor Theory (1940-1946).
MMRW-BOOKSAbstractM00020 MNYCP 00MMRW-BOOKS: Legacy books on slowing down, thermalization, particle transport theory, random processes in reactors.
MOCAAbstractC00590 IPCAT 00Monte Carlo Criticality Code System for Hexagonal Geometries.
MOCUPAbstractP00365 DALPU 00MCNP/ORIGEN Coupling Utility Programs.
MODELAbstractC00329 I3033 00Models of Trapped Proton and Electron Environments for Solar Maximum and Minimum.
MOMENT IAbstractC00188 U1108 00Moments Method Neutron Transport Code System.
MOMGEM-MOMDISAbstractC00085 I7090 00Moments Method Reconstruction of Scattered Gamma-Ray Distributions.
MONK 6.3
FEDC
AbstractC00393 I3033 00A General Purpose Monte Carlo Neutronics Code System.
MONTEBURNS 2.0AbstractP00455 MNYCP 02An Automated, Multi-Step Monte Carlo Burnup Code System.
MONTUK-80AbstractD00072 ALLCP 01UKCTR III Transmutation and Activation Data, 100-Group Neutron Activation Cross-Section Data for Fusion Reactor Structure and Coolant Materials.
MORECAAbstractP00411 PC386 00Computer Code System for Simulating Modular High-Temperature Gas Cooled Reactor Core Heatup.
MORNAbstractP00062 I0360 00Calculation of the Response of Sodium Iodide Crystals to Gamma Rays.
MORSE-ALBAbstractC00394 FM200 00Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System, Albedo Version. We recommend C00474/ALLCP/02 MORSE-CGA.
MORSE-ANSI STD.AbstractC00127 I3675 00A General Purpose Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System. We recommend C00474/ALLCP/02 MORSE-CGA.
MORSE-BAbstractC00368 I0370 00General Purpose Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System. We recommend C00474/ALLCP/02 MORSE-CGA.
MORSE-CAbstractC00431 C7600 00Monte Carlo Multigroup Neutron Code System for the Solution of Criticality Problems. We recommend C00474/ALLCP/02 MORSE-CGA.
MORSE-CGAbstractC00203 C0000 00A General Purpose Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System with Combinatorial Geometry. We recommend C00474/ALLCP/02 MORSE-CGA.
MORSE-CGAbstractC00203 CY000 00A General Purpose Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System with Combinatorial Geometry. We recommend C00474/ALLCP/02 MORSE-CGA.
MORSE-CGAbstractC00203 D0VAX 00A General Purpose Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System with Combinatorial Geometry. We recommend C00474/ALLCP/02 MORSE-CGA.
MORSE-CGAbstractC00203 I0360 00A General Purpose Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System with Combinatorial Geometry. We recommend C00474/ALLCP/02 MORSE-CGA.
MORSE-CGAbstractC00203 U0000 00A General Purpose Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System with Combinatorial Geometry. We recommend C00474/ALLCP/02 MORSE-CGA.
MORSE-CGAAbstractC00474 ALLCP 03A General Purpose Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System with Array Geometry Capability, Version 2.
MORSEC-SP2AbstractP00142 H6000 00A Multigroup Cross Section Module for the MORSE Monte Carlo Computer Code System.
MORSE-CVAbstractC00535 HM280 00Multigroup Neutron and Gamma-Ray Monte Carlo Transport Code with Covariance Calculation. We recommend C00474/ALLCP/02 MORSE-CGA.
MORSE-EAbstractC00258 I0360 00Special Purpose Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System. We recommend C00474/ALLCP/02 MORSE-CGA.
MORSE-EMPAbstractC00588 IBMPC 00General Purpose Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System with Array Geometry Capability. We recommend C00474/ALLCP/02 MORSE-CGA.
MORSE-HAbstractC00471 I3081 00A Revised Version of the MORSE Monte Carlo Radiation Transport Code System. We recommend C00474/ALLCP/02 MORSE-CGA.
MORSE-LAbstractC00261 C6600 00Multigroup Neutron and Gamma-Ray Transport Code System for the Solution of Penetration Problems. We recommend C00474/ALLCP/02 MORSE-CGA.
MORSE-SGCAbstractC00277 C7600 00A Super Grouped Cross Section Version of the MORSE Code System. We recommend either C00474/ALLCP/02 MORSE-CGA, or C00545/IRISC/01 SCALE 4.2.
MOSRA-LIGHTAbstractP00505 MNYWS 00High-Speed Three-Dimensional Nodal Diffusion Code System.
MOXY-MOD32AbstractP00385 I0360 00BWR Core Heat Transfer Code System.
MRIPP 1.0
810
AbstractC00655 PC386 00Magnetic Resonance Image Phantom Code System to Calibrate in vivo Measurement Systems.
MRSPAKAbstractP00212 DVX11 00A Code System To Generate a Text File Containing Combinatorial Geometry Data Corresponding to PADL2 Geometry.
MSM-SOURCEAbstractP00369 MNYCP 00Code System for Generation of Input Data for MCNP.
MTR_PC 2.6AbstractC00674 PC386 00Modular Code System for Neutronics, Thermalhydraulics and Shielding Calculations.
MULTI-KENO2AbstractC00492 FM380 00A Monte Carlo Code System for Criticality Safety Analysis.
MUP2AbstractP00289 I3090 00A Program to Calculate Fast Neutron Data for Medium-Heavy Nuclei.
MUREAbstractC00764 MNYWS 00MCNP Utility for Reactor Evolution.
MURLIAbstractC00378 DP011 00Integral Transport Theory Code System for Thermal Reactor Lattice Cell Calculation.
MUSCATAbstractC00281 I0360 00Calculation of Neutron Currents in Spherical and Cylindrical Cavities by Means of View Factors.
MUSPALBAbstractC00171 ICL00 00Albedo Calculation of Multigroup Spectra of Neutrons Transmitted Through Multilayer Slab Shielding.
MUXSAbstractP00187 I3033 00Generator of Multigroup Cross Sections for Charged Particle Transport Problems.
MVP-GMVP IIAbstractC00739 MNYCP 00General Purpose Monte Carlo Codes for Neutron and Photon Transport Calculations based on Continuous Energy and Multigroup Methods.
MYRAAbstractC00056 C0000 00Calculation of Shipping Costs and Cask Designs for Irradiated Fuel Elements.
MYRAAbstractC00056 I7090 00Calculation of Shipping Costs and Cask Designs for Irradiated Fuel Elements.
NAAPROAbstractC00722 PC586 00Neutron Activation Analysis PRognosis and Optimization Code System.
NABAbstractD00018 I0360 00100-Group, P3, Neutron Cross Section Data for Sodium and Aluminum.
NACAbstractC00164 C0000 00Neutron Activation Analysis and Product Isotope Inventory Code System.
NACAbstractC00164 IBMMF 00Neutron Activation Analysis and Product Isotope Inventory Code System.
NAC-PCAbstractC00164 IBMPC 00Neutron Activation Analysis and Product Isotope Inventory Code System.
NACTAbstractC00502 U1100 00Screening Program for Neutron Activation Products.
NAISAPAbstractP00085 F2306 00Theory and Use of Gamma-Ray Spectrum Analysis Codes for NaI(Tl) Detectors.
NANICKAbstractP00120 I0360 00Infinitely-Diluted Multigroup Cross-Section Generator - from ENDF/B.
NAPAbstractC00101 I7090 00Multigroup Time-Dependent Neutron Activation Prediction Code.
NASIF-NARESAbstractP00121 I0360 00A Code System for Computing Shielding Factors from ENDF/B Tapes.
NAUA-MOD5 NAUA-MOD5/MAbstractP00556 MNYCP 00Aerosols in Reactor Containment During Meltdown.
NCRP49AbstractC00462 IBMPC 00X-Ray Shield Calculation System.
NCSP-DATAbstractM00002 MNYCP 01Nuclear Data in Support of the Nuclear Criticality Safety Program.
NEACRP-H2O-LATTICESAbstractD00265 MNYCP 00Compilation of Reactor Physics Measurements in LWRs Lattices.
NE-SPECAbstractP00150 F2307 00A Code System for Unfolding a Pulse Height Distribution of Neutrons Measured by an NE-213 Organic Scintillator.
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
NEUPACAbstractP00177 FM200 00Neutron Unfolding Code System for Calculating Neutron Flux Spectra from Activation Data of Dosimeter Foils.
NEVEMORAbstractP00026 I3675 00Multigroup-Multiregion Calculation of Flux Spectra and Energy Deposition for Fast Neutrons.
NITRANAbstractC00582 FM380 00Neutron Transport Code System Based On Anisotropic Scattering.
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.
NMTC/JAERI97AbstractC00694 SUN05 00Monte Carlo Nucleon Meson Transport Code System.
NMTC/JAMAbstractC00717 PC586 00High Energy Particle Transport Code System.
NONSAP-CAbstractP00458 C7600 00Code System for Analysis of 3-D Reinforced Concrete Structures.
NORMAAbstractP00471 PC586 00Code System to Solve Burnup Dependent Neutron Diffusion Equations in Two and Three Dimensions.
NORMA-FPAbstractP00470 PC586 00Code System to Perform Neutronic and Thermal-Hydraulic Subchannel Analysis from Converged Coarse-Mesh Nodal Solutions.
NOXAbstractD00017 I0360 00199-Group, P5, Coupled Neutron and Secondary Gamma-Ray Cross Section Data for Nitrogen and Oxygen.
NPCSL-81AbstractD00082 I0370 00Point Neutron Cross Sections Generated from ENDF/B-IV with the NPTXS Modules of PSR-63/AMPX-II.
NPTXSAbstractP00090 I0360 00Data Generator: Neutron Point Cross Sections from ENDF/B Resolved and Unresolved Resonance Parameters.
NRCDOSE 2.3.20AbstractC00684 PC586 14Code System for Evaluating Routine Radioactive Effluents from Nuclear Power Plants with Windows Interface.
NRCDOSE72V1.2.3AbstractC00768 PCX86 03Code System for Evaluating Routine Radioactive Effluents from Nuclear Power Plants with Windows Interface.
NRCPAGEAbstractP00491 DVX11 00Code System to Detect Recurring Loss of Special Nuclear Materials.
NRCPIPES 2.0AAbstractP00429 IBMPC 00Code System for Fracture Mechanics Analysis of Circumferential Surface Cracks in Pipes.
NRNAbstractC00054 C6600 00Multigroup Removal-Diffusion Code System for Planes, Cylinders and Spheres.
NSLINKAbstractP00314 D0VAX 00NJOY SCALE LINK.
NUCCONAbstractC00439 S7800 00A Code System for Calculation of Time-Dependent Nuclide Concentrations, Activity, Gamma-Ray Dose Rate and Biological Hazard Potential of Fusion Reactor Materials Due to Neutron Irradiation.
NUCDECAYAbstractD00172 PC386 01Nuclear Decay Data for Radiation Dosimetry Calculations for ICRP and MIRD.
NUCDECAYCALCAbstractD00202 PC586 00Nuclear Decay Data for Radiation Dosimetry Calculations for ICRP. See newer version in RASCAL (CCC-553).
NUCHARTAbstractP00545 IBMPC 00Nuclear Properties and Decay Data Chart of Nuclides.
NUCWIZAbstractP00616 PCX86 00NucWiz
NUFACEAbstractP00284 CYXMP 00An Interface Code For The Calculation of Nuclear Responses.
NUGAM 2&3 SSLABAbstractC00210 I0360 00Monte Carlo Prediction of Photon Transport Distributions.
NUTRANAbstractC00675 I0370 00Code System for Long-Term Repository Safety Analysis.
NX1-NX2AbstractP00310 D0VAX 00Code System to Calculate Excitation Functions for (n,charged particle) Reactions.
O5RAbstractC00017 I3675 00A General-Purpose Monte Carlo Neutron Transport Code System.
O5SAbstractP00014 DP010 00Response Function Generator--An O5R Monte Carlo Code for Calculating Pulse Height Distributions Due to Monoenergetic Neutrons Incident on Organic Scintillators.
O5SAbstractP00014 I3675 00Response Function Generator--An O5R Monte Carlo Code for Calculating Pulse Height Distributions Due to Monoenergetic Neutrons Incident on Organic Scintillators.
O6RAbstractC00128 I3675 00A General-Purpose Monte Carlo Transport Code System.
OCA-PAbstractP00392 I3033 00Pressure Vessel Fracture-Mechanics Code System.
OCA-PAbstractP00392 IBMPC 00Pressure Vessel Fracture-Mechanics Code System.
OCTAVIAAbstractP00460 I0370 00Code System to Calculate Pressure Vessel Failure Probabilities.
OGREAbstractC00046 I3675 00A General-Purpose Monte Carlo Gamma-Ray Transport Code System.
OGRE-MINAbstractC00409 DGECL 00A General-Purpose Monte Carlo Gamma-Ray Transport Code System for Minicomputers.
OMCOSTAbstractP00381 I3033 00Code System for Non-fuel O & M Cost Estimation for Large Steam-Electric Power Plants.
OMEGAAbstractC00433 BESM6 00Monte Carlo Criticality Code System.
ONETRANAbstractC00266 C7600 00A One-Dimensional Multigroup Discrete Ordinates Finite Element Transport Code System. We recommend CCC-547/TWODANT-SYS.
ONETRANAbstractC00266 CY000 00A One-Dimensional Multigroup Discrete Ordinates Finite Element Transport Code System. We recommend CCC-547/TWODANT-SYS.
ONETRANAbstractC00266 I3033 00A One-Dimensional Multigroup Discrete Ordinates Finite Element Transport Code System. We recommend CCC-547/TWODANT-SYS.
OOSIIAbstractC00324 C0000 00Calculation of Isotropic Scattering by Particles for One-Dimensional and Three-Dimensional Transport in Slabs by Invariant Imbedding, Orders-of-Scattering Method, Including Check Calculations by Integral Transport Theory and Monte Carlo.
OPEX-IIAbstractC00103 I7090 00Radiation Shield Optimization Code.
OPTIMAbstractC00817 I0370 00Minimization of Band-Width of Finite Elements Problems.
ORCENT-2AbstractP00474 I3033 00Code System for Analysis of Steam Turbine Cycles Supplied by Light Water Reactors.
ORESUNDAbstractD00267 MNYCP 00Nordic Mesoscale Dispersion Experiments over Land-Water-Land.
ORIGEN2.2AbstractC00371 ALLCP 03Isotope Generation and Depletion Code - Matrix Exponential Method. New ORIGEN users are advised to get CCC-750/SCALE6 and run the ORIGEN-ARP code system in that package.
ORIGEN-JENDL32AbstractC00703 MNYWS 00Isotope Generation and Depletion Code with Libraries Based on JENDL3.2. New ORIGEN users are advised to get CCC-750/SCALE6 and run the ORIGEN-ARP code system in that package.
ORINC
USSO
AbstractP00439 I0360 00Code System for 1-D Implicit Heat Conduction Solution.
ORION-IIAbstractC00491 FM780 00A Computer Code to Estimate Environmental Concentration and Dose Due to Airborne Release of Radioactive Material.
ORIP_XXIAbstractC00731 PC586 02Computer Programs for Isotope Transmutation Simulations.
ORLIBJ32AbstractD00255 MNYCP 00ORIGEN2 LIBRARIES BASED ON JENDL-3.2.
ORMDIN
USSO
AbstractP00399 I3033 002-D Nonlinear Inverse Heat Conduction.
ORMGEN3DAbstractP00430 CY0MP 00Mesh Generator for 3-D Crack Geometries.
ORMONTEAbstractP00275 IBMPC 00Uncertainty Analysis Code System for Use with User-Developed Systems Models.
ORPHEE VIAbstractC00159 I3675 00Kernel Integration Code System - Attenuation of Fast Neutrons in Cylindrical Layers of Water and Dense Material.
ORPLOT-PCAbstractP00328 PC386 00Plotting Package for Data Evaluation Intercomparison.
ORSMAC
USSO
AbstractP00437 I3033 00Code System to Calculate Fluid Circulation Patterns Near Jets.
ORTHIS-ORTHATAbstractP00569 I0360 00ORTHIS: Steady-State Heat Conduction in 2-D X-Y, R-Z and R-Theta Geometry; ORTHAT: Transient Heat Conduction in 2-D X-Y, R-Z and R-Theta Geometry.
ORTURBAbstractP00418 I0360 00HTGR Steam Turbine Dynamic Behavior.
ORYX-EAbstractD00038 I0360 00ORIGEN Yields and Cross Sections Nuclear Transmutation and Decay Data from ENDF/B-IV.
ORYX-EAbstractD00038 I0360 01ORIGEN Yields and Cross Sections Nuclear Transmutation and Decay Data from ENDF/B-IV.
OZMAAbstractC00406 I0370 00Calculation of Resonance Reaction Rates in Reactor Lattices Using Resonance Profile Tabulations.
PABLMAbstractC00402 U1100 00Calculation of Accumulated Radiation Doses to Man from Radionuclides Found in Food Products and from Radionuclides in the Environment.
PADF-2007AbstractD00259 PCX86 00Proton Activation Data File in ENDF-6 Format.
PADLOCAbstractC00330 U0000 00A One-Dimensional, Time-Dependent Program for Calculating Coolant and Plateout Fission Product Concentrations in a Network of Pipes.
PAGANAbstractC00621 IBMPC 00Code System for Performance Assessment Ground-water Analysis for Low-level Nuclear Waste.
PALLAS-1D(VII)AbstractC00380 FM380 00Multigroup Time-Independent Neutron Transport Code System for Plane or Spherical Geometry.
PALLAS-2DCY-FXAbstractC00391 FM380 00Multigroup Neutron/Gamma-Ray Direct Integration Transport Code System for Two-Dimensional Cylindrical Geometry.
PAPER 1AbstractP00097 C6600 00Monte Carlo Calculation of Solid Angle and Self-Absorption Factors for an Inclined Cylindrical Source Viewed by a Cylindrical Detector.
PAPINAbstractP00156 I0370 00A Code System to Calculate Cross Section Probability Tables, Bondarenko and Transmission Self-Shielding Factors for Fertile Isotopes in the Unresolved Resonance Region.
PARET-ANLAbstractP00516 MNYCP 00Code System to Predict Consequences of Nondestructive Accidents in Research and Test Reactor Cores.
PARET-ANL(NESC)AbstractP00565 MNYCP 00Code System to Predict Consequences of Nondestructive Accidents in Research and Test Reactor Cores.
PART61AbstractC00499 IBMPC 01Low-Level Radioactive Waste Impacts Analysis System.
PARTISN 5.97AbstractC00760 MNYCP 00Time-Dependent, Parallel Neutral Particle Transport Code System.
PATCH-7AbstractC00243 C0074 00Three-Dimensional Kernel Integration Code-Explicit Single Scattering Option.
PAVANAbstractC00445 I3033 00Atmospheric Dispersion Code System for Evaluating Accidental Radioactivity Releases from Nuclear Power Stations.
P-CARESAbstractP00538 PC586 00Probabilistic Computer Analysis for Rapid Evaluation of Structures.
PC-BATLEAbstractP00451 IBMPC 00Code System to Calculate Brief Adversary Threat Loss Estimate.
PCC/SRCAbstractP00456 D0VAX 00Code System to Calculate Correlation & Regression Coefficients.
PC-PRAISEAbstractP00391 IBMPC 00Code System for Analysis of Piping Reliability Including Seismic Events.
PEFPYDAbstractD00096 ALLMF 02Aggregate Fission-Product Decay Data Based on ENDF/B-IV and -V.
PEGASAbstractP00336 IBMPC 00Pre-Equilibrium-Equilibrium Gamma-and-Spin Code System.
PELE-1CAbstractP00461 C7600 00Code System for Fluid-Structure Interaction Analysis.
PELINOMIC-3AAbstractP00596 I0370 00Power Plant Cost Optimization for Dispersed Load Centers.
PELINSCAAbstractP00168 I0360 00A Code System for Nuclear Elastic and Inelastic Scattering Calculations.
PELSHIEAbstractC00202 C0000 00General Purpose Kernel Integration Shielding Code System-Point and Extended Gamma-Ray Sources.
PELSHIE3AbstractC00202 IBMMF 00General Purpose Kernel Integration Shielding Code System-Point and Extended Gamma-Ray Sources.
PENELOPE2014AbstractC00782 PCX86 01Code System for Monte Carlo Simulation of Electron and Photon Transport.
PENELOPE-MPIAbstractC00713 IBMSP 00Code System to Perform Monte Carlo Simulation of Electron Gamma-Ray Showers in Arbitrary Marerials.
PENGEOMAbstractC00840 MNYCP 00Tools for Handling Complex Quadric Geometries in Monte Carlo Simulations of Radiation Transport
PEPINAbstractC00285 I0360 00Methodology for Computing Concentrations, Activities, Gamma-Ray Spectra, and Residual Heat from Fission Products.
PEQAG-2AbstractP00293 IPCAT 00A Pre-equilibrium Computer Code With Gamma Emission.
PERSENT11.0
FEDC
AbstractC00823 MNYWS 00Perturbation and Sensitivity Code for Assembly Homogenized Multi-group Transport Problems
PF-COMPAbstractC00106 C3600 00Building Fallout Radiation Protection Factor Analysis.
PFPLAbstractC00607 D0VAX 00Puff-Plume Atmospheric Deposition Model.
PGAA-IAEAAbstractD00234 MNYCP 00Databsae for Prompt Gamma-Ray Neutron Activation Analysis.
PHAZE
USSO
AbstractP00432 IBMPC 00Parametric Hazard Function Estimation.
PHITS-2.88AbstractC00778 MNYCP 05PHITS-2.88, Particle and Heavy Ion Transport code System
PHOBIAAbstractD00236 PCX86 00Photon buildup factors to account for angular incidence on shield walls.
PHOEL-2AbstractC00327 I0360 00A Monte Carlo Calculation of Initial Energy of Photoelectrons and Compton Electrons Produced by Photons in Water.
PHOTXAbstractD00136 D0VAX 01Photon Interaction Cross Section Library.
PHOTXAbstractD00136 IBMPC 00Photon Interaction Cross Section Library.
PICAAbstractC00160 D0VAX 00Monte Carlo Medium-Energy Photon-Induced Intranuclear Cascade Anal Code System.
PICAAbstractC00160 I0360 00Monte Carlo Medium-Energy Photon-Induced Intranuclear Cascade Anal Code System.
PICESAbstractP00568 I3033 00Probabilistic Investigation of Capacity and Energy Shortages.
PICFEEAbstractC00175 I3675 00Fission Product Inventory Code System.
PICTUREAbstractP00238 IBMPC 00Combinatorial Geometry Printer Plotting.
PIEDECAbstractC00566 FM380 00A Practical Internal Exposure Dose Evaluation Code.
PIGGAbstractC00138 C3600 00A Multigroup One-Dimensional P-1 Radiation Transport Code System.
PIPEAbstractC00219 I0360 00Numerical Gamma-Ray Transport Code System for Plane/Spherical Geometry.
PIXE2010AbstractD00246 MNYCP 00Proton/alpha Ionization (K, L, M shell), Tabulated Cross Section Library.
PIXSEAbstractP00133 I0360 00A Generator of Multigroup and Multipoint Cross Sections for Thermal Reactor Calculations.
PKIAbstractC00573 C0830 00A Point Kernel Integration Code For Radiation Shielding of Loop System.
PLACIDAbstractC00381 I0370 00Monte Carlo Simulation of Gamma Streaming Through Straight Cylindrical Ducts.
PLASMXAbstractP00106 C6600 00A Multigroup Ionization and Charge Exchange Cross-Section Code System for Neutral Hydrogen Transport in Plasmas.
PLOTENDFAbstractP00214 I3033 00A Program for Producing Graphical Output.
PLOTFBAbstractP00018 I3675 00ENDF/B Data Plotting Code.
PLOTNFITAbstractP00382 IBMPC 00Code System for Data Plotting and Curve Fitting.
PLOT-SAbstractP00552 PC586 00Plotting Program with Special Features for Windows Environment.
PLOTTAB-89.1AbstractP00274 ALLCP 00Plot Continuous Curves or Discrete Points.
PLUDOSAbstractC00313 I0360 00Calculator of Ground Level External Gamma-Ray Dose from a Radioactive Plume.
PLUMEXAbstractC00356 I0360 00A Computer Program to Evaluate External Exposures to a Gaussian Plume by Point Kernel Integration.
PMK2-VVER440-REPORTSAbstractM00012 MNYCP 00Results of the Experiments Performed in the PMK-2 Facility for VVER Safety Studies.
PNAbstractC00818 I0370 00MultiGroup Neutron Transport.
PNESDAbstractD00166 PC386 00Proton Nucleus Elastic Scattering Data.
POINT2015AbstractD00273 MNYCP 00POINT 2015: ENDF/B-VII.1 Final Temperature Dependent Cross Section Library
POISSXAbstractC00819 I0370 00Poisson Equation on Rectangle with Various Boundary Conditions.
POLLAAbstractP00208 I3033 00A Fortran Program to Convert R-MATRIX-Type Multilevel Resonance Parameters for Fissile Nuclei into Equivalent KAPUR-PEIERLS-Type Parameters.
POLYRESAbstractP00438 MNYCP 00Richards Equation Solver; Rectangular Finite Volume Flux Updating Solution.
POPLIBAbstractD00012 I0360 03A Compendium of Neutron-Induced Secondary Gamma-Ray Yield and Cross Section Data.
POPOP4AbstractP00011 I3675 00Converter of Gamma-Ray Spectra to Secondary Gamma-Ray Production Cross Sections.
POWERAbstractP00069 C7600 00Source Distribution Input Data Generator for ANISN Code.
PREANGAbstractP00166 C0175 00Calculation of Pre-equilibrium Angular Distributions with the Exciton Model.
PRE-ANISNAbstractP00332 PC386 00A Preprocessing Code for ANISN and Other Radiation Transport Codes.
PRECO2006AbstractP00226 MNYCP 02Exciton Model Code System for Calculating Preequilibrium and Direct Double Differential Cross Sections.
PR-EDBAbstractD00196 IBMPC 03Power Reactor Embrittlement Data Base, Version 3.
PREDEX-1AbstractP00597 I0370 00U, Pu, Nitric Acid Distribution in Counter Current Solvent Extraction.
PREMAbstractP00224 I0360 00Code System for Pre-equilibrium Process with Multiple Nucleon Emission.
PREMORAbstractC00369 I0360 00A Point Reactor Exposure Code System for Survey Nuclear Analysis of Power Plant Performance.
PREP/SPOPAbstractC00772 MNYCP 00Uncertainty and Sensitivity Analysis Monte Carlo Program and Input Preparation.
PREPRO2017AbstractP00351 MNYCP 09Pre-Processing Code System for Data in ENDF/B Format.
PRESTAbstractC00355 I0360 00Calculator of Pressure and Temperature Transient in Containment Studies.
PRESTOAbstractC00549 D8810 00Point Kernel Calculation for Complex and Time-Dependent Gamma-Ray Source Spectra.
PRESTO-IIAbstractC00504 I0360 00Code System for Low-Level Waste Environmental Transport and Risk Assessment.
PRIMEDANA-2AbstractC00490 I3081 00Collapses Multigroup Cross Sections and Obtains Reaction Parameters by Solving Transport or Diffusion Equations.
PRISIMAbstractC00574 IBMPC 00Plant Risk Status Information Management System.
PROBAbstractC00287 I0370 00Multigroup One-Dimensional Transport Code System, Collision Probability Method.
PROCIVAbstractC00488 U1110 00A Code System for Calculating the Protection Factors Against Radioactive Fallout for Apartment Buildings.
PSAPACK-4.2AbstractP00613 PCX86 00Probabilistic Safety Analysis with Fault Event Trees.
PSDRECAbstractP00441 DP011 00Code System for Power Spectral Density Recognition Continuous On-line Reactor Surveillance.
PSU-LEOPARD/RBIAbstractC00563 IBMPC 01A Spectrum Dependent Non-Spatial Depletion Code.
PTRANAbstractC00618 PC386 00Proton Monte Carlo Transport Program for the PC.
PUCORAbstractD00067 I3691 0084 Group Neutron Cross Sections for Uranium-Plutonium Cycle LWR and PWR Models in AMPX Master Library Format.
PUDKAbstractD00074 I0360 00Measured Results of Delayed Beta- and Gamma-Ray Spectra Due to Thermal-Neutron Fission of Pu239 and Pu241.
PUFF-IVAbstractP00534 MNYCP 01Code System to Generate Multigroup Covariance Matrices from ENDF/B-VI Uncertainty Files, Version 6.0.1.
PURSEAbstractC00338 C6600 00A Plutonium Radiation Source Code System.
PUSHLDAbstractC00271 C0074 00Gamma-Ray Three-Dimensional Calculation of Dose Rates from Plutonium in Various Geometries.
PUTZ 2.1AbstractC00595 IBMPC 00A Point-Kernel Photon Shielding Code.
PVCAbstractD00048 I3691 0036 Group, P5, Photon Interaction Cross Sections for 38 Materials in ANISN Format.
PVEAbstractD00126 I3033 0038 Group, P8, Photon Interaction Cross Sections in ANISN Format from VITAMIN-E.
PVIS-4AbstractC00798 MNYCP 00Pressure Vessel Irradiation Source.
PWR-AXBUPRO-GKNAbstractD00209 MNYCP 00Measured Axial Burnup Profiles for NeckarWesthiem PWR Reactors.
PWR-AXBUPRO-SNLAbstractD00201 MNYCP 00Axial Burnup Profile Database for Pressurized Water Reactors.
Q&AAbstractP00428 IBMPC 00Questions and Answers Based on Revised 10 CFR Part 20
QADAbstractC00048 I0360 00Kernel Integration Code System.
QAD-BSAAbstractC00346 C0000 00Point-Kernel Shielding Code System.
QAD-CGGP-AAbstractC00645 MNYCP 00Point Kernel Code System for Neutron and Gamma-Ray Shielding Calculations Using the GP Buildup Factor.
QADMOD-GAbstractC00396 I3033 00Point Kernel Gamma-Ray Shielding Code.
QADMOD-GPAbstractC00565 IBMPC 00Point Kernel Gamma-Ray Shielding Code With Geometric Progression Buildup Factors.
QAD-P5AbstractC00048 C6400 00Kernel Integration Code System.
QAD-QCAbstractC00401 C0000 00Three-Dimensional Point Kernel Gamma-Ray Shielding Code.
QAD-QCAbstractC00401 I0360 00Three-Dimensional Point Kernel Gamma-Ray Shielding Code.
QAD-UEAbstractC00448 H6000 00A Revised Numerical Integration Option for Gamma-Ray Volume Source Problems in the QAD-CG Point Kernel Shielding Code.
QBFAbstractC00617 PC386 00Code System to Calculate Radiation Dose Rates Relative to Spent Fuel Shipping Casks.
QBSHIELDAbstractC00599 IBMPC 00Spherical Shield Design for Gamma-Ray Sources Using the Buildup Factor Method.
QUARKAbstractP00492 PC586 00Code System for 2-Group, 3D Neutronic Kinetics Calculations Coupled to Core Thermal Hydraulics.
QUINCE-PCAbstractC00556 IBMPC 00Calculates Absorbed Dose From Skin Contamination.
RABFIN PARTSAbstractC00668 IBMPC 00Code System for Calculating Gaseous Effluent Dose Parameters.
RACCAbstractC00388 CY000 00A Code System for Computing Radioactivity-Related Parameters for Fusion Reactor Systems.
RACCAbstractC00388 I3033 00A Code System for Computing Radioactivity-Related Parameters for Fusion Reactor Systems.
RACC-PULSEAbstractC00639 MNYWS 00RACC Code System for Computing Radioactivity-Related Parameters for Fusion Reactor Systems Modified for Pulsed/Intermittent Activation Analysis.
RACERAbstractC00174 U1108 00Calculation of Potential External Dose from Airborne Fission Products Following Postulated Reactor Accident.
RAD 2AbstractC00122 I7090 00Fission Product Radioactivities Calculation.
RADACAbstractC00627 PC486 02Code System for Calculating Radioactive Decay and Accumulation of Decayed Products Using Integer-Array Arithmetic for Precise Evaluation of the Bateman Equations.
RADAKAbstractP00122 I0360 00Flux Spectra Unfolding Code System - Neutron or Gamma-Ray Detectors.
RADCOMPT 2.10LAbstractP00348 IBMPC 00Sample Analysis Code System for the Dual Channel Counter.
RADDECAY 4.02AbstractD00134 IBMPC 03Radioactive Decay Data for Radiological Assessments.
RADHEAT-V4AbstractC00300 FM380 00A Code System To Generate Multigroup Constants and Analyze Radiation Transport for Shielding Safety Evaluation.
RADOSAbstractC00088 I3675 00Gamma-Ray Dose Estimation from Cloud of Radioactive Gases by Kernel Integration.
RADRISKAbstractC00422 DGMV1 00Estimates Radiation Doses and Health Effects from Inhalation or Ingestion of Radionuclides. See C00476/CAAC.
RADRISKAbstractC00422 I3033 00Estimates Radiation Doses and Health Effects from Inhalation or Ingestion of Radionuclides. See C00476/CAAC.
RADSHIP-2AbstractC00523 FM200 00Code System To Analyze Radiological Impact From Radwaste Transportation.
RADSYSAbstractC00530 I3033 00Code System for Radioactivity Buildup and Radioactive Waste Generation Calculations.
RADTRAD 3.03AbstractC00800 IBMPC 00A Simplified Model for RADionuclide Transport and Removal And Dose Estimation.
RADTRAD 3.03-EXEAbstractC00800 IBMPC 01A Simplified Model for RADionuclide Transport and Removal And Dose Estimation.
RAFFLE/2AbstractC00279 C0176 00A General Purpose Monte Carlo Code System for Neutron Transport with Mixed Zone Geometry Option.
RAFFLE/2 MOD 2AbstractC00279 I0360 00A General Purpose Monte Carlo Code System for Neutron Transport with Mixed Zone Geometry Option.
RAIDAbstractC00083 I7090 00Monte Carlo Multibend Duct Shielding Code.
RANCHMDAbstractC00589 D8810 00Radionuclide Chain Transport with Matrix Diffusion.
RAPIDAbstractC00797 PCX86 00RAdial Power and Burnup Prediction by Following Fissile Isotope Distribution in the Pellet.
RASC-2DAbstractC00318 I0370 00Two-Dimensional Removal Diffusion Code Reactor Shielding Design Code System.
RASCAL 4.3AbstractC00783 PCX86 02Radiological Assessment Systems for Consequence AnaLysis.
RASPAAbstractC00352 C7600 00A Code for the Calculation of Buildup and Decay of Fission Products and Actinides.
RATAFAbstractC00681 IMFPC 01Code System for the Radioactive Liquid Tank Failure Study.
RBDAbstractC00632 IBMPC 00U.S. Army Radiological Bioassay and Dosimetry.
RCSLK9AbstractP00452 IBMPC 00Code System to Calculate Reactor Coolant System Leak Rate.
RDMMAbstractP00598 I0360 00Flux Spectra from In-Pile Fast Neutron Activation Experiment.
REAC*3AbstractC00443 IBMPC 00Computer Code System for Activation and Transmutation.
REAC*3AbstractC00443 MFMWS 00Computer Code System for Activation and Transmutation.
REACTIONAbstractP00347 AL000 00Code System to Calculate Integral Parameters with Reaction Rates from WIMS Output.
REACTIONAbstractP00347 IBMPC 00Code System to Calculate Integral Parameters with Reaction Rates from WIMS Output.
REACTORSHIELDING-NMSAbstractM00014 MNYCP 00REACTORSHIELDING-NMS, Reactor Shielding for Nuclear Engineers.
REBEL 3AbstractC00299 I0360 00Adjoint Monte Carlo Calculation of Radiation Doses to Human Organs in Dwelling Rooms.
REBEL-2AbstractC00299 ICL00 00Adjoint Monte Carlo Calculation of Radiation Doses to Human Organs in Dwelling Rooms.
REBEL-2AbstractC00299 C6600 00Adjoint Monte Carlo Calculation of Radiation Doses to Human Organs in Dwelling Rooms.
REBUS 11.0
FEDC
AbstractC00822 MNYWS 00Code System for Analysis of Fast Reactor Fuel Cycles.
REBUS 11.0 EXE_ONLYAbstractC00822 MNYWS 01Code System for Analysis of Fast Reactor Fuel Cycles.
REBUS3/VARIANT8AbstractC00653 MNYWS 01Code System for Analysis of Fast Reactor Fuel Cycles.
REBUS-PC 1.4AbstractC00708 PC586 00Code System for Analysis of Research Reactor Fuel Cycles.
RECAPAbstractP00414 IBMPC 00Replacement Energy Cost Analysis Package.
RECAPAbstractP00414 IBMPC 01Replacement Energy Cost Analysis Package.
RECOILAbstractD00055 I3033 01Multigroup Primary Recoil Spectra, Displacement Rates and Gas-Production Rates for Radiation Damage Studies.
REDIFFUSIONAbstractC00347 I0360 00One-Dimensional Neutron Removal-Diffusion and Gamma-Ray Kernel Integration or Diffusion Theory Calculator.
REEX-1AbstractP00599 I0370 00U, Pu, Nitric Acid Distribution in Counter Current Pluristage Stripping.
REFCO83AbstractP00447 I3033 00Nuclear Fuel Cycle Cost Economics Code System.
REFERDOUAbstractP00249 FM380 00Code System for NE-213 Unfolding of Neutron Spectra up to 100 MeV with Response Function Error Propagation.
REFIT-2009AbstractC00775 PCX86 00Multilevel Resonance Parameter Least Square Fit of Neutron Transmission, Capture, Fission & Self Indication Data.
REFLUXAbstractP00403 I3033 00Code System to Predict LWR Reflood Heat Transfer.
REFREPAbstractC00570 D8810 00A Near-Field Model For A Spent Fuel Repository.
REFUM-BROADAbstractP00039 F2307 00Monte Carlo Codes for Calculating Efficiencies and Response Functions of NaI(Tl) Crystals for Thick Disk Gamma-Ray Sources.
REGNAbstractP00165 I0360 00Code System for Solving Nonlinear Systems of Equations via the Gauss-Newton Method.
RELAP3B/MOD110
810
AbstractP00422 C7600 00Reactor System Transient Code.
RELAP4/MOD7/101
810
AbstractP00416 C0176 00Best Estimate Code System to Calculate Thermal & Hydraulic Phenomena in a Nuclear Reactor or Related System.
RELAP5/MOD1/029
810
AbstractP00423 C0176 00Thermal Hydraulic Computer Code System.
RELAP5/MOD1/029_EXE
810
AbstractP00423 C0176 01RELAP5/MOD1/029_EXE_only
REMIT 5.1AbstractP00482 IBMPC 01Radiation Exposure Monitoring and Information Transmittal System.
REPCAbstractP00195 C0000 00Estimation of Nuclear Reaction Effects in Proton-Tissue-Dose Calculations.
REPRISK PC 1.02AbstractC00586 PC386 01Repository Risk Assessment Software for Personal Computers.
RESENDDAbstractP00215 C0740 00A Code System for Reconstruction of Resonance Cross Sections from Evaluated Nuclear Data in ENDF/B Format.
RESENDDAbstractP00215 D0780 00A Code System for Reconstruction of Resonance Cross Sections from Evaluated Nuclear Data in ENDF/B Format.
RESPMGAbstractP00060 I0360 00Response Matrix Generation Code System.
RESRAD 6.5AbstractC00786 PCX86 00Code System to Implement Residual Radioactive Material Guidelines, and RESRAD-BUILD 3.0.
REST 1;2;3AbstractC00225 I0360 00Fission Product Inventory Code System with Fission Product Escape Model.
RETRACAbstractC00635 D0VAX 00Code System for the Analysis of Material Test Reactor (MTR) Cores.
RETRANSAbstractC00669 SUN05 00Code System For Calculating Reactivity Transients In a LWR.
REX2-87AbstractP00290 D8810 00A Code For Calculating Self-Shielded Multigroup Neutron Cross Sections and Self-Shielding Factors From Preprocessed ENDF/B Basic Data Files.
RFSP-JULAbstractP00126 I0360 00Unfolding Code System for Neutron Spectra Evaluation from Activation Data.
RFUNCAbstractP00312 D0VAX 00Code System to Analyze Differential Scattering Data.
RGENDFAbstractP00239 C0170 00Format Translation from NJOY GENDF Format to ENDF/B-V and Other Formats.
RHEINAbstractC00585 I3090 00Reactor Code System for Neutron Physics Calculation.
RIBD-IIAbstractC00137 C6600 00Radioisotope Buildup and Decay Code System.
RIBD-IIAbstractC00137 I0360 00Radioisotope Buildup and Decay Code System.
RIBD-IRTAbstractC00382 U1100 00Radioisotope Buildup and Decay Code System.
RICANTAbstractC00569 D8810 00A Computer Code for 2-D Transport Calculations in x-y Geometry Using the Interface Current Method.
RICEAbstractP00022 I0360 00A Program to Calculate Primary Recoil Atom Spectra from ENDF/B Data.
RICECCCAbstractC00348 I0360 00A Reactor Nuclide Inventory Code for Calculating Actinides and Fission Products.
RICMAbstractP00600 I0360 00Resonance Absorption in Multi-Region Slab or Square or Hexagonal Lattice.
RIPPLEAbstractP00571 CYXMP 00A Computer Program for Incompressible Fluid Dynamics with Free Surfaces.
RISKAPAbstractC00486 I3033 00Analysis of Increased Risk to Arbitrary Populations.
RISKIND 2.0
FEDC
AbstractC00623 IBMPC 02Radiological Risk Assessment Code System for Spent Nuclear Fuel Transportation.
RITTSAbstractD00011 I0360 00121-Group Coupled Neutron and Gamma-Ray Cross-Section Data for Transport Codes.
RIVER-RADAbstractC00626 MNYCP 00Code System for Simulating the Transport of Radionuclides in Rivers.
RMET21AbstractC00597 D0VAX 00Detailed Space and Energy Treatment of Neutron Resonances for Homogeneous Mixtures and Cylinderized Reactor Cells.
RNGPAbstractP00066 I3675 00Random Number Generator Package.
ROCKWELL-RSDMAbstractM00017 MNYCP 00Reactor Shielding Design Manual by Rockwell T. III.
ROLAIDS-CPMAbstractP00353 SUN04 00Code System to Calculate Group-Averaged Cross Sections Using the Collision Probability Method.
RRRAbstractC00196 I0360 00Radiation Transport in Air-Analysis of Routine Releases of Short-Lived Radioactive Nuclides.
RSAC6.2AbstractC00125 PC586 03Radiological Safety Analysis Code System.
RSAC-7.2AbstractC00761 PC586 01Radiological Safety Analysis Code System.
RSYSTAbstractC00269 I0360 00Integrated Modular Code System for Shielding and Reactor Physics Calculations.
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 8.1.0AbstractP00158 MNYCP 13Code 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 8.0.9AbstractP00608 PCX86 00Systems Analysis Programs for Hands-On Integrated Reliability Evaluations.
SARA 4.16AbstractP00484 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 6.2.1AbstractC00834 MNYCP 02A Comprehensive Modeling and Simulation Suite for Nuclear Safety Analysis and Design; Includes ORIGEN and AMPX.
SCALE 6.2.1-EXEAbstractC00834 MNYCP 03A Comprehensive Modeling and Simulation Suite for Nuclear Safety Analysis and Design; Includes ORIGEN and AMPX - EXE Only
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
SCEPTRE 1.7AbstractC00826 PCX86 01Sandia 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.
SERPENTAbstractC00757 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 2017.12AbstractD00237 MNYCP 05Shielding Integral Benchmark Archive and Database, Version December 2017
SINBAD SEARCH TOOLAbstractP00580 MNYCP 00SINBAD Search Tool
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 6.0AbstractC00767 MNYCP 06SoftWare for Optimization of Radiation Detectors, SWORD Version 6.0
SYVAC-D/2AbstractC00690 D0VAX 00Code System For Risk Assessment From Underground Radioactive Waste Disposal In the United Kingdom.
TACT-IIIAbstractC00447 I3033 00Calculation of the Transport of Radioactivity from a Reactor Core.
TALYS-1.2AbstractP00548 PC586 01Nuclear Model Code System for Analysis and Prediction of Nuclear Reactions and Generation of Nuclear Data.
TAM3AbstractP00308 IBMPC 00Demonstrates Monte Carlo Sensitivity and Uncertainty Analysis.
TART2016AbstractC00638 MNYCP 08Coupled Neutron-Photon, 3-D, Combinatorial Geometry, Time Dependent, Monte Carlo Transport Code System.
TASKAbstractC00184 I0360 00Generalized One-Dimensional Radiation Transport and Diffusion Kinetics Code System.
TDAAbstractC00180 MNYWS 01A Time-Dependent, Multigroup, One-Dimensional, Discrete Ordinates Transport Code System.
TDFAbstractD00162 ALLCP 00Thermonuclear Data File.
TDOWN-IVAbstractP00172 H6000 00A Code System to Generate Composition- and Spatially-Dependent Neutron Cross Sections for Multigroup Neutronics Analysis.
TDTAbstractC00256 I0360 00Generalized One-Dimensional Multigroup Time-Dependent Transport and Diffusion Kinetic Code System.
TDTORTAbstractC00709 MNYWS 00Time-Dependent, 3-D, Discrete Ordinates, Neutron Transport Code System.
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.
TEMACAbstractP00468 D0VAX 00Top Event Matrix Analysis Code System.
TEMPEST-2AbstractP00558 I0360 00Thermalization Program for Neutron Spectra and MultiGroup Cross-Sections.
TEMPEST-BNWAbstractP00559 C7600 00Transient 3-D Thermohydraulics for FBR.
TENDL-2008-ACEAbstractD00243 MNYCP 00TALYS-Based Cross Section Library for Use with MCNP(X).
TENDL-2010-ACEAbstractD00248 MNYCP 00TALYS-Based Cross Section Library for Use with MCNP(X).
TENDL-2011-ACEAbstractD00252 MNYCP 00TALYS-Based Cross Section Library for Use with MCNP(X).
TENDL-2012-ACEAbstractD00266 MNYCP 00TALYS-Based Cross Section Library for Use with MCNPX.
TERFOC-NAbstractC00596 MFMWS 00Terrestrial Food-Chain Model for Normal Operations.
TESSAbstractC00215 C3600 00Multigroup Discrete Ordinates Code System for Slab and Spherical Geometries.
THACT-RRAbstractP00587 D0VAX 00Analysis of Thermal Hydraulics Transients in Research Reactor Core.
THERMGAMAbstractD00140 ALLCP 00Prompt Gamma Rays from Thermal-Neutron Capture.
THERMOS-OTAAbstractP00107 C0173 00Multigroup Integral Transport Code System for Thermal Lattice Calculations using Collision Probability Method for Slabs and Cylinders.
THERMOS-OTAAbstractP00107 C0740 00Multigroup Integral Transport Code System for Thermal Lattice Calculations using Collision Probability Method for Slabs and Cylinders.
THERMOS-OTAAbstractP00107 U1108 00Multigroup Integral Transport Code System for Thermal Lattice Calculations using Collision Probability Method for Slabs and Cylinders.
THIDA-2AbstractC00410 FM380 00Code System for the Calculation of Transmutation, Activation, Decay Heat and Dose Rate in Fusion Reactors.
THRUSHAbstractP00276 CYXMP 00Calculates Thermal Neutron Scattering Kernel.
THTAbstractC00480 I0360 00Three-Dimensional Neutron Coarse Mesh Code System to Evaluate Average Bundle Fluxes and Power in Light Water Reactors.
THYDE-B1/MOD2AbstractP00553 FM200 00Computer Code for the Analysis of Small-Break Loss-of-Coolant Accident of Boiling Water Reactors.
THYDE-P2AbstractP00554 FV100 00Computer Code for PWR LOCA Thermohydraulic Transient Analysis.
TIBSOAbstractC00512 MNYCP 00Code System to Calculate Production and Migration of Radionuclides in Nuclear Reactor Systems.
TIMEDAbstractC00292 I0360 00Calculation of Cumulated Activity of a Radionuclide in the Organs of the Human Body at a Given Time After Deposition.
TIMEXAbstractC00274 C7600 00One Dimensional, Time Dependent Multigroup Explicit Discrete Ordinates Radiation Transport Code System with Anisotropic Scattering.
TIMEXAbstractC00274 CY000 00One Dimensional, Time Dependent Multigroup Explicit Discrete Ordinates Radiation Transport Code System with Anisotropic Scattering.
TIMEXAbstractC00274 U1106 00One Dimensional, Time Dependent Multigroup Explicit Discrete Ordinates Radiation Transport Code System with Anisotropic Scattering.
TIMOC-72AbstractC00144 I0370 00Monte Carlo Three-Dimensional Neutron Transport Code System.
TIMOC-ESPAbstractC00432 U1110 00System for Generating and Analyzing Time Dependent Radiation Transport Results by Monte Carlo.
TIMS-1AbstractP00163 D0780 00Processing Code System for Production of Group Constants of Heavy Resonant Nuclei.
TIMS-1AbstractP00163 FM200 00Processing Code System for Production of Group Constants of Heavy Resonant Nuclei.
TIRION 4AbstractC00395 I3033 00A Program for Calculating Consequences of a Release of Radioactive Material to the Atmosphere.
TITAN 1.29AbstractC00759 PCX86 04A Three-Dimensional Deterministic Radiation Transport Code System.
TMMSAbstractC00246 I0360 00Gamma-Ray Penetration Shielding Code System, Transmission Matrix Method.
TNG1AbstractP00298 D6220 00A Multistep Statistical Model Based on the Hauser-Feshbach Theory For The Evaluation Of Neutron Data.
TORACAbstractP00459 C0170 00Code System to Calculate Tornado-Induced Flow Material Transport.
TOTEM-3AbstractP00603 I0370 00Demand Assessment for Nuclear Power Plants and Conventional Power Plants.
TOXRISKAbstractC00692 CDCMF 00Code System for Toxic Gas Accident Analysis.
TP1AbstractC00465 I3033 00A Computer Code System for the Calculation of Reactivity and Kinetic Parameters by One-Dimensional Neutron Transport Perturbation Theory.
TP2AbstractC00470 I3033 00A Computer Program for the Calculation of Reactivity and Kinetic Parameters by Two-Dimensional Neutron Transport Perturbation Theory.
TPASGAM 85AbstractD00088 ALLCP 04Radioactive Decay Library of Gamma-Ray Energies, Branching Ratios, and Cross Sections.
TPASSAbstractP00164 DP010 00A Gamma-Ray Spectral Data-Reduction and Analysis Code System.
TPHEXAbstractC00421 C0173 00Transmission Probability Code System for Calculating Neutron Flux Distributions in Hexagonal Geometry.
TPHEXAbstractC00421 CYXMP 00Transmission Probability Code System for Calculating Neutron Flux Distributions in Hexagonal Geometry.
TPTRIAAbstractC00550 I3083 00A Computer Program for the Reactivity and Kinetic Parameters for Two-Dimensional Triangular Geometry by Transport Perturbation Theory.
TRANSHEXAbstractC00449 U1108 00Two-dimensional Multigroup Collision Probability Code System for Hexagonal Geometry.
TRANSMITAbstractD00020 I0360 00Experimental Neutron Transmission Data Used to Test Total Cross Sections.
TRANSPORTAbstractC00244 C6600 00Charged Particle Beam Transport Systems Design Code System (First- and Second-Order Matrix Multiplication).
TRANSPORTAbstractC00244 I0360 00Charged Particle Beam Transport Systems Design Code System (First- and Second-Order Matrix Multiplication).
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.
TRANSX-CTRAbstractP00206 CY000 00Interfaces MATXS Cross-Section Libraries to Nuclear Transport Codes for Fusion Systems Analysis.
TRANZITAbstractC00172 C7600 00Multigroup Time-Dependent Discrete Ordinates Radiation Transport Code System in (rho,z) Cylindrical Geometry.
TRAPPAbstractC00205 I3691 00Transport of Alpha Particles and Protons with all Nuclear Reaction Products Neglected.
TRAXAbstractP00280 C0720 00A Program For Optics of Curved Crystal Neutron Spectrometers.
TRD-3AbstractC00362 I3033 00Two-Dimensional Removal-Diffusion Neutron Shielding Code System.
TRECOAbstractC00116 I3675 00An Orbital Integration Estimation of Trapped Radiation.
TR-EDBAbstractD00198 IBMPC 00Test Reactor Embrittlement Data Base, Version 1.
TREEDEAbstractC00326 C0000 00Monte Carlo Neutron Transport Code System Based on the Track Rotation Estimator.
TRG-SGDAbstractC00025 C0000 00Calculation of Secondary Gamma-Ray Dose Rate from a Nuclear Weapon Detonation-Monte Carlo Method.
TRIDENTAbstractC00293 C7600 00Two-Dimensional Multigroup Discrete Ordinates Transport Code System-(x,y) and (r,z) Geometries.
TRIDENTAbstractC00293 I0360 00Two-Dimensional Multigroup Discrete Ordinates Transport Code System-(x,y) and (r,z) Geometries.
TRIDENT-CTRAbstractC00377 C0000 00Two-Dimensional x-y and r-z Geometry Multigroup Transport Code System for Large Toroidal Reactors.
TRIGAPAbstractC00600 IBMPC 00A Computer Code for TRIGA Type Reactors.
TRIGLAVAbstractP00495 PC586 00Code System to Calculate Mixed Cores in TRIGA Mark II Research Reactor.
TRIGONAbstractC00290 U1108 00Two-Dimensional Multigroup Diffusion Code System-Trigonal or Hexagonal Mesh.
TRIPLETAbstractC00230 C6600 00Two-Dimensional, Multigroup, Triangular Mesh, Planar Geometry, Explicit Discrete Ordinates Code System.
TRIPLETAbstractC00230 C7600 00Two-Dimensional, Multigroup, Triangular Mesh, Planar Geometry, Explicit Discrete Ordinates Code System.
TRIPLETAbstractC00230 I0360 00Two-Dimensional, Multigroup, Triangular Mesh, Planar Geometry, Explicit Discrete Ordinates Code System.
TRIPOLI-4 8.1
OECD
AbstractC00806 MNYCP 003D General Purpose Continuous Energy Monte Carlo Transport Code.
TRIPOLI-4 9S
OECD
AbstractC00815 MNYCP 00Coupled Neutron, Photon, Electron, Positron 3-D, Time Dependent Monte-Carlo Transport Calculation.
TRIPOSAbstractC00537 CY00I 00Monte Carlo Ion Transport Analysis Code.
TRISTANAbstractC00511 HM280 00Multigroup Three-Dimensional Direct Integration Method Radiation Transport Analysis Code System.
TRISTAN-IJSAbstractP00537 IBMPC 00Steady-State Axial Temperature and Flow Velocity in Triga Channel.
TRITACAbstractC00560 D8810 00A Three-Dimensional Transport Code For Eigenvalue Problems Using The Diffusion Synthetic Acceleration Method.
TRUMPAbstractP00522 MNYCP 01Code System for Transient and Steady-State Temperature Distribution in Multidimensional Systems.
TSL-ACE/2013AbstractD00270 ALLCP 00TSL-ACE/2013
TSORTAbstractP00486 IBMPC 00Automated Technique for Nuclear Plant Training Task Assignment.
TURBINAAbstractP00604 I0370 00Reheat Steam Turbine Generator Design with Preheater and Condenser.
TWOTRANAbstractC00195 C6600 00Two-Dimensional Discrete Ordinates. We recommend CCC-547/TWODANT-SYS.
TWOTRAN IIAbstractC00222 C7600 00Two-Dimensional Multigroup Discrete Ordinates Transport C System in (x,y), (r,theta), and (r,z) Geometries.
TWOTRAN IIAbstractC00222 I3691 00Two-Dimensional Multigroup Discrete Ordinates Transport C System in (x,y), (r,theta), and (r,z) Geometries.
TWOTRAN-SPHEREAbstractC00129 C6600 00Multigroup Two-Dimensional Discrete Ordinates Transport Code System in Spherical Geometry.
UDAD IXAbstractC00685 I0370 00Uranium Dispersion & Dosimetry Model.
UHSAbstractP00390 IPS70 00Ultimate Heat Sink Cooling Pond and Spray Pond Analysis Models.
UKCTRI-81AbstractD00064 I0370 0146-Group Neutron Cross Sections and Kerma Factors for Fusion Reactor Calculations.
UKE-IIIAbstractP00015 I3691 00Cross Section Format Translator - UKNDL to ENDF/B.
UKFY2AbstractD00171 IBMPC 00UK Fission Product Yield Library, Version 2.
UKNDLAbstractD00039 I0370 00United Kingdom Evaluated Neutron Cross-Section Data Library.
UKNDL-81AbstractD00107 I3033 00The Aldermaston Nuclear Data Library.
UMG 3.3AbstractP00529 PC586 00Unfolding with Maxed and Gravel.
UMIBIOAbstractC00680 I3033 00Code System to Model Uranium Mills Bioassay Dosimetry.
UNFAbstractP00521 PC586 00Code System to Calculate Multistep Compound Nucleus Neutron Cross-Sections and Spectra for Structural Materials.
UNGERAbstractD00164 PC386 00Effective Dose Equivalent for Specific Radionuclides.
UNIFY-ECNAbstractP00288 C0170 00A Program to Calculate Fast Neutron Data for Structural Materials.
UNIMUG3AbstractC00407 C0170 00Solves Multigroup Diffusion Equations in One-Dimensional Systems.
UPDATEAbstractP00270 DGMV1 00Program to Update Fortran Source Files.
UPDATEAbstractP00270 I3081 00Program to Update Fortran Source Files.
UPEAKAbstractP00300 IPCXT 00A Program for Decomposing A One-Dimensional Spectrum.
UPEML 3.0AbstractP00245 ALLCP 01A Machine-Portable CDC UPDATE Emulator.
URRAbstractP00281 D6220 00Calculates Resonance Neutron Cross-Section Probability Tables, Bondarenko Self-Shielding Factors and Self-Indication Ratios for Fissile and Fertile Nuclides.
USINTAbstractP00415 MNYCP 00Code System to Calculate Heat and Mass Transfer In Concrete
USRHYDAbstractC00197 I3675 00Electron and X-Ray Energy Deposition and Hydrodynamics Code System.
UTMTOXAbstractC00500 D8600 00Unified Transport Model for Toxic Materials.
UTSGAbstractP00379 I3033 00Code System for Calculating the Nonlinear Transient Behavior of a Natural Circulation U-Tube Steam Generator with Its Main Steam System.
UTXS6AbstractD00211 MNYCP 00MCNP Continuous-Energy Neutron Cross Section Libraries for Temperatures from 300 to 1365K.
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.
VARSKIN 4AbstractC00781 PCX86 00Computer Code System to Assess Skin Dose from Skin Contamination
VCSAbstractC00262 I0360 00Coupled Discrete Ordinates-Adjoint Monte Carlo Calculation of Radiation Protection Factors in Vehicles.
VELMAbstractD00133 I0360 00Multigroup Cross-Section Libraries Based on ENDF/B-V Data for Sodium-Cooled Reactor Shield Analysis.
VENTEASYAbstractC00776 PCX86 00Criticality search for a desired Keffective by adjusting dimensions, nuclide concentrations, or buckling
VENTURE-PCAbstractC00654 PC586 02A Reactor Analysis Code System.
VESTA 2.1.5-AURORA 1.0AbstractC00769 PCX86 01VESTA 2.1.5 - Monte Carlo Depletion Interface Code; AURORA 1.0.0 - Depletion Analysis Tool.
VIDEO-PCAbstractP00311 IBMPC 00Super VGA Primitives Graphics System.
VIEWCXSAbstractP00514 PC586 00Interactive Graphic User Interface to View Neutron and Gamma-Ray Interaction Cross Sections.
VIM 5.1AbstractC00754 MNYWS 01Continuous Energy Neutron and Photon Transport Code System, April 2009 Release.
VIM_NCAbstractC00794 PCX86 00VIM Color Syntax for Nuclear Codes: NJOY, DRAGON, PARTISN, TORT, MONK, and MCNP.
VIP-MANAbstractD00256 MNYCP 00Computational Phantom.
VISA2AbstractP00445 MNYCP 00Code System to Calculate Probability of Reactor Vessel Failure.
VITAMIN-4CAbstractD00053 I3691 00171 Neutron Group Cross Sections and Bondarenko Factors in CCCC Interface Formats for Fusion and LMFBR Neutronics.
VITAMIN-B6AbstractD00184 ALLCP 00A Fine-Group Cross Section Library Based on ENDF/B-VI Release 3 for Radiation Transport Applications.
VITAMIN-B7/BUGLE-B7AbstractD00245 MNYCP 01Broad-Group and Fine-Group and Coupled Neutron/Gamma Cross-Section Libraries Derived from ENDF/B-VII.0 Nuclear Data.
VITAMIN-CAbstractD00041 I0360 02171 Neutron, 36 Gamma-Ray Group Cross Sections in AMPX and CCCC Interface Formats for Fusion and LMFBR Neutronics.
VITAMIN-EAbstractD00113 I3033 02174n, 38g Cross-Section Library in AMPX Format.
VITAMIN-J/COVAAbstractD00157 D8810 00Neutron Cross-Section Covariance Data in Multigroup Form.
VITAMIN-J/COVA/EFFAbstractD00197 ALLCP 00Neutron Cross-Section Covariance Data in Multigroup Form.
VITAMIN-J/KERMAAbstractD00150 I3090 00VITAMIN-J 175-Neutron and 38-Photon Kerma And Gas Production Cross Sections.
VITENDF70.BOLIBAbstractD00261 PCX86 00ENDF/B-VII.0 Multi-Group Coupled (199n +42gamma) Cross Section Library in AMPX Format for Nuclear Fission Applications.
VITENEA-EAbstractD00240 MNYCP 00VITENEA-E, AMPX 174-N,38-Gamma Multigroup X-sec. Library for Multidimensional Radiation Transport and Dose Evaluation.
VITENEA-JAbstractD00238 MNYCP 00AMPX 175-n,42-g Multigroup X-section Library for Nuclear Fusion Applications.
VITJEF22.BOLIBAbstractD00241 MNYCP 00JEF-2.2 Multigroup Coupled (199n + 42?) Cross-Section Library in AMPX Format for Nuclear Fission Applications.
VITJEFF31.BOLIBAbstractD00235 MNYCP 00A JEFF-3.1 Multigr Coupled (199n + 42gamma) X-Section Lib. in AMPX Fmt for Nuclear Fission Applications.
VITJEFF311.BOLIBAbstractD00257 MNYCP 01JEFF-3.1.1 Multi-Group Coupled (199n + 42gamma) X-Section Library in AMPX Format for Nuclear Fission Applications.
VIXENAbstractP00030 C6600 00A Code to Check Physical Consistency of Photon-Production Data in Revised ENDF Format.
VIXENAbstractP00030 I0360 00A Code to Check Physical Consistency of Photon-Production Data in Revised ENDF Format.
VPI-NECMAbstractC00481 C0740 00Nuclear Engineering Computer Models for In-Core Fuel Management Analysis.
VPI-NECMAbstractC00481 D0VAX 00Nuclear Engineering Computer Models for In-Core Fuel Management Analysis.
VPI-NECMAbstractC00481 PC486 00Nuclear Engineering Computer Models for In-Core Fuel Management Analysis.
VSOP94AbstractC00670 MNYWS 00Code System for Reactor Physics and Fuel Cycle Simulation.
VVER-BENCHMARKSAbstractM00003 MNYCP 00Collection of Neutronic VVER Reactor Benchmarks.
WAKEAbstractP00605 I0370 00Navier Stokes Equation with 2-D Turbulence, Stream Function, Vorticity.
WEERIEAbstractC00426 I3033 00Code System for Assessing the Radiological Consequences of Airborne Effluents from Nuclear Installations.
WHATIF-AQAbstractC00561 B7800 00A Computer Program For Speciation Calculation.
WILITAbstractP00344 MNYCP 00A Utility Program for WIMS Libraries.
WIMKAL-88AbstractD00193 MNYCP 0069 Energy Group, Neutron Cross Section Library For Thermal Reactor Calculations in WIMSD Format.
WIMS-ANL 4.0AbstractC00698 MNYCP 00Deterministic Code System for Reactor Lattice Calculation.
WIMSCORE-ENEAAbstractP00319 I3090 00Code System to Process WIMSD4 Interface Output Files and Generate Two-Group Data for Reactor Calculations.
WIMSD-5B.12AbstractC00656 MNYCP 02Deterministic Code System for Reactor Lattice Calculation
WIMSLIB-IJS0AbstractD00147 D8810 00Extended Version of the WIMS 69-group Library.
WIMSLIB-IJS1AbstractD00147 D8810 01Extended Version of the WIMS 69-group Library.
WIMSLIB-JEF87AbstractD00095 D0VAX 00JEF-1 Based 69 Group Neutron Data Library.
WINDOWSAbstractP00136 I0360 00A Program for the Analysis of Spectral Data Foil Activation Measurements.
WINDOWS IIAbstractP00161 I0370 00A Program for the Analysis of Spectral Data Foil-Activation Measurements.
WLUP 3.0AbstractD00231 MNYCP 0169- and 172- Group Cross Section Libraries for WIMS.
W-M-NRSMAbstractD00026 U1108 00WANL-MSFC Nuclear Rocket Shielding Methods Data Generator (GAMLEG-W, APPROPOS, NAGS, and SATURN) and Multigroup Neutron and Gamma-ray Cross Section Libraries 1-6.
WRAITHAbstractC00427 U1100 00Code System for Calculating Internal and External Doses Resulting from an Atmospheric Release of Radioactive Material.
WREM-TOODEE2AbstractP00469 ALLMF 002-D Time-Dependent Fuel Element, Thermal Analysis Code System.
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.
XCOMAbstractD00174 IBMPC 00Photon Cross Sections on a Personal Computer, Versions 1.2 and 1.3.
XG-IAEAAbstractD00163 IBMPC 00X-ray and Gamma-ray Standards For Detector Calibration.
XLACS-IIAAbstractP00182 I3033 00A Modified Version of XLACS-II for Processing ENDF Data into Multigroup Neutron Cross Sections in AMPX Master Library Format.
XOQDOQ-82AbstractC00316 DGMV1 00Radiological Assessment Code System - Meteorological Evaluation of Routine Effluent Releases at Nuclear Power Stations.
XOQDOQ-82AbstractC00316 I3033 00Radiological Assessment Code System - Meteorological Evaluation of Routine Effluent Releases at Nuclear Power Stations.
XOQDOQ-82AbstractC00316 IPCAT 00Radiological Assessment Code System - Meteorological Evaluation of Routine Effluent Releases at Nuclear Power Stations.
XPORT-PCAbstractC00559 IBMPC 00An Approximation For Black Body X-Ray Transport in Air.
XRAY_AACAbstractC00525 D0750 00X-ray Attenuation and Absorption Calculations.
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.
XSHLDAbstractC00495 IBMPC 00Diagnostic X-Ray Shielding Calculation.
XSUN-2013AbstractC00825 PCX86 00Windows interface environment for transport and sensitivity-uncertainty software TRANSX-2, PARTISN and SUSD3D
YUMMYAbstractD00221 MNYCP 00Multi-temperature, Neutron Cross Section Library Based on ENDF/B-V and ENDF/B-VI for use with MCNP.
ZOTT99AbstractP00272 ALLCP 02Zero-in On The Truth; Evaluation of Correlated Data Using Partitioned Least Squares.
ZYLIND-PCAbstractC00557 IBMPC 00An Interactive Point Kernel Program For Photon Dose Rate Prediction of Cylindrical Source/Shield Arrangements.
ZZ-PWR-MSLBAbstractD00275 MNYCP 00ZZ PWR-MSLB, PWR Main Steam-Line Break Benchmarks, Coupled Neutronics Thermal-Hydraulics
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.