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
Peripheral Shielding Routines (shielding research tools)
Package NameAbstractRSICC TapelistTitle
1DXAbstractP00096 U1108 00A One-Dimensional Diffusion Code System for Producing Energy Group Collapsed and Self-Shielded Cross Sections.
ABAREXAbstractP00248 MNYCP 01Neutron Spherical Optical-Statistical Model Code System.
ABLEIT-TRANSAbstractP00247 C0175 00Error Propagation Analysis for Burnup Calculation.
ACATAbstractP00257 FM380 00Monte Carlo Simulation of Atomic Collisions in Amorphous Targets in the Binary Collision Approximation.
ACORNSAbstractP00264 IBMPC 01Analysis of Correlations Used in Neutron Spectrometry.
ACTIVAbstractP00590 I0370 00Sandwich Detector Activity from In-Pile Slowing-Down Spectra Experiment.
ACTIV-PCAbstractP00287 IBMPC 00A Program to Process Gamma or X-ray Spectra.
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.
ADLER IIIAbstractP00058 I0360 00A Program to Calculate Cross Sections from Adler-Adler Resonance Parameters.
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
ALARM-B2AbstractP00218 I0360 00A Computer Code System for Analysis of a Large Break LOCA of a BWR.
ALICE2017AbstractP00550 PCX86 06Statistical Model Code System to Calculate Particle Spectra from HMS Precompound Nucleus Decay.
ALPHA-MAbstractP00169 I0360 00Least-Squares Resolution of Gamma-Ray Spectra in Environmental Samples.
ALPSAbstractP00144 F2307 00Alpha Spectrum Analysis Code System.
AMARAAbstractP00079 I3675 00Nuclear Data Adjustment Using Lagrange's Multipliers Method.
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.
ANSIFTAbstractP00077 C6600 00ANSI Standard Fortran Sifting Program.
ANSIFTAbstractP00077 I0360 00ANSI Standard Fortran Sifting Program.
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.
AREADAbstractP00088 I0360 00Input Data Processor for Transport Codes.
ART MOD2AbstractP00611 PCX86 00Fission Product Migration in Primary System and Containment
ATHENA_2DAbstractP00431 MNYCP 00Code System For Simulation Of Hypothetical Recriticality Accidents in a Thermal Neutron Spectrum.
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.
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.
BEACON MOD3AbstractP00402 CDCMF 00Code System for Thermal-Hydraulic Analysis of Nuclear Reactor Containments.
BEST-5AbstractP00591 I0370 00Power Reactor Fuel Cycle Optimization by Bellman Method.
BFR
USSO
AbstractP00449 C0176 00Code System for Common Cause Failure Data Analysis.
BLOCKAGE V2.5RAbstractP00377 IBMPC 00Code System to Calculate Potential Plugging of Emergency Core Cooling Suction Strainers by Debris in a BWR.
BOB-7 SERIESAbstractP00084 F2306 00Theory and Use of Gamma-Ray Spectrum Analysis Codes for Ge(Li) Detectors.
BONAbstractP00173 I0360 00A Code System for Unfolding Multisphere Spectrometer Neutron Measurements.
BOT3P-5.3AbstractP00530 MNYCP 02Code System for 2D and 3D Mesh Generation and Graphical Display of Geometry and Results for Radiation Transport Codes.
BREESE-IIAbstractP00143 I3033 00Auxiliary Routines for Implementing the Albedo Option in the MORSE Monte Carlo Code System.
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.
BULK-IAbstractP00574 PCX86 00Radiation Shielding Tool for Proton Accelerator Facilities.
BURDAbstractP00582 IBMPC 00Bayesian Estimation in Data Analysis of Probabilistic Safety Assessment.
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.
CARP-82AbstractP00131 I3033 00Multigroup Albedo Data Using DOT Angular Flux Results.
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.
CCRMNAbstractP00366 MNYCP 00Monte Carlo Simulation of the Coupled Transport of Electrons and Photons.
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.
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.
CHENDF 7.02AbstractP00333 MNYCP 05Codes for Handling ENDF/B-V and ENDF/B-VI Data.
COAG-IIAbstractP00070 I0360 00Calculation of the Westcott Epithermal Index and the Westcott 2200 m/s Neutron Flux.
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
COGAPAbstractP00375 MNYCP 01Nuclear Power Plant Containment Hydrogen Control System Evaluation Code.
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).
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.
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.
CRECTJ5AbstractP00250 D0780 00A Computer Program for Compilation of Evaluated Nuclear Data in ENDF/B Format.
CRESOAbstractP00184 I3081 00Resonance Data-Handling Code System.
CUPEDAbstractP00032 I3675 00Scintillation Spectrometer Polyenergetic Gamma Photon Experimental Distributions Unfolding Code.
D2OAbstractP00398 PC486 00Code System for Computing Thermodynamic and Transport Properties of D2O.
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.
DASQHEAbstractP00278 D8810 00Calculates Dancoff Corrections Factors.
DATINITAbstractP00258 DGMV1 00Interactive Program To Access Photon Interaction Data.
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.
DIFBASAbstractP00334 MNYCP 00A Bayesian Approach to Unfolding a Neutron Spectrum from a Spectrum of Recoiled Protons.
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.
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.
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.
DSNPAbstractP00592 I3033 00Dynamic Simulation Nuclear Power.
DSNQUADAbstractP00251 IPCXT 00Calculates Angular Quadrature Weights and Cosines.
DUFOLDAbstractP00042 I0360 00Derivative Unfolding Code - Determination of Neutron Spectra from NE-213 Pulse Height Data.
DWBA07/DWBB07AbstractP00338 MNYCP 01Code System for Inelastic and Elastic Scattering with Nucleon-Nucleon Potential
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.
ECIS-12AbstractP00612 MNYCP 00ECIS-12, Coupled Channel, Statistical Model, Schroedinger and Dirac Equation, Dispersion Relation
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.
EEDBAbstractP00531 MNYCP 00The Energy Economic Data Base.
ELANAbstractP00141 ICL00 00Neutron Cross-Section Self-Shielding Code System.
ELIESE-3AbstractP00003 I0370 00Analyses of Elastic and Inelastic Scattering Cross Sections.
EMPIRE-IIAbstractP00497 PC586 01Comprehensive Nuclear Model Code, Nucleons, Ions Induced Cross-Sections.
ENBAL2AbstractP00160 I0370 00A Program to Generate Multigroup Neutron Kerma Factors.
ENDVER/GUIAbstractP00572 PCX86 00The ENDF File Verification Support Package.
ENLOSSAbstractP00047 C6600 00Calculation of Energy Loss of Charged Particles.
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.
EPIPE
USSO
AbstractP00485 CY000 00Code System for Static and Dynamic Piping System Analysis.
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.
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.
ERRORJAbstractP00526 MNYCP 03Covariance Processing Code System, Version 2.3.
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.
EURCYLAbstractP00076 I0370 00Finite Element Three-Dimensional Mesh Generator for Cylinder - Cylinder Intersections.
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.
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.
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.
FDMXPCAbstractP00322 IPCAT 00Code System for Calculation of Neutron Transmission and Other Functionals from Evaluated Data in ENDF Format.
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.
FEMAXI 6 VER.1AbstractP00536 IBMPC 00Code System for Light Water Reactor Fuel Analysis.
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.
FIGEROAbstractP00149 C0000 00Processing Codes for Generating Multigroup Neutron Cross Sections from ENDF/B for Use in Discrete Ordinates Calculations.
FIRACAbstractP00444 CY000 00Nuclear Facilities Fire Accident Model
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.
FLODISAbstractP00417 I0360 00Code System to Calculate Thermal Response of FSV HTGR Core.
FLOWPLOT IIAbstractP00234 I3033 00Fluid Dynamics and Heat Transfer Plotting Package.
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.
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.
FOURACESAbstractP00183 I0370 00Code System for Producing Spectrum Weighted, Group Averaged Cross Sections from ENDF/B, KEDAK, or UK Libraries.
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.
FREEFORMAbstractP00081 I0360 00Free-Form Input Reading Routines.
FUELSDATAAbstractP00446 C7600 00Code System to Model Verification Fuel Rod Data.
GABASAbstractP00175 U1108 00A Code System for Generating Composite Time-Dependent Fission Produce Spectra.
GADRAS-DRF-18.7.2AbstractP00610 PCX86 02Gamma Detector Response and Analysis Software–Detector Response Function.
GAINCALBAbstractP00056 I0360 00Determination of the Gain Used with Organic Scintillation Detect.
GALAXY-6AbstractP00098 I0370 00Neutron Multigroup Cross Section Processor.
GAMANAbstractP00083 DP010 00Qualitative and Quantitative Evaluation of Ge(Li) Gamma-Ray Spectra.
GAMANALAbstractP00506 D0VAX 00Code System for Computerized Quantitative Analysis By Gamma-Ray Spectrometry.
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.
GAMMAAbstractP00095 I0360 00Monte Carlo Code System for Calculating Efficiencies and Response Functions of NaI(Tl) Crystals for Gamma Rays from Thick Disk Sources.
GAMX1AbstractP00209 I0370 00A Computer Code System for Evaluating Spectra Peak Areas.
GAPCON-THERMALAbstractP00499 C7600 00Code System to Calculate Fuel Steady State & Transient Behavior.
GAROLAbstractP00033 I7090 00Calculation of Resonance Neutron Absorption in Two-Region Problems.
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.
GCIAbstractP00421 IBMPC 00Generic Communications Index
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.
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.
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.
GGTC-ENELAbstractP00128 I0360 00Code System for Producing Few-Group Neutron Cross Sections from Multigroup Data Libraries.
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.
GOFRRAbstractP00127 I0360 00Generator of Graphical Output of DOT and ANISN Fluxes and Reaction Rates.
GRASS-SSTAbstractP00489 MNYCP 00Code System to Predict Fission-Gas Release & Fuel Swelling.
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.
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.
GRUCONAbstractP00615 MNYCP 00Data Processing for Evaluated Working libraries (transport and shielding)
GT2R2AbstractP00483 ALLMF 00Code System to Calculate Fuel Rod Thermal Performance.
HAARM-3AbstractP00401 CDCMF 00Aerosol Behavior Log-Normal Distribution Model.
HASSANAbstractP00593 I0370 00Time-Dependent Temperature Distribution and Stress and Strain in HTR Fuel Pins.
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.
HECTR 1.5+
USSO
AbstractP00457 CY000 00Hydrogen Event Containment Response Code System.
HEITLERAbstractP00004 I7030 00Cross Section Generator.
HSI-DRGAbstractP00435 IBMPC 00Code System for Use with Human System Interface Design Review Guidelines.
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.
IERAbstractP00024 I3675 00A Gauss-based Quadrature Formula Applied to Sievert's Integral. An Exponential Integral Routine.
IMPORTANCEAbstractP00407 I0370 00FTA Basic Event & Cut Set Ranking.
INFLTBAbstractP00313 ALLCP 00Gamma-Ray Absorption Coefficient Calculation.
INGENAbstractP00207 C0000 00A General-Purpose Mesh Generator for Finite Element Codes.
INTRIGUE-IIAbstractP00054 I0360 00Logarithmic and Semilogarithmic CALCOMP Plot Routines.
IRRAS 4.16
USSO
AbstractP00386 IBMPC 04Code System to Calculate Integrated Reliability and Risk Analysis.
ITER-2AbstractP00148 C0000 00Codes for Unfolding Activation Detector Data and Pulse Height Spectra.
JPHYDROAbstractP00594 I0360 00Voids and Flow Velocity in Steady-State BWR System.
KAOS-VAbstractP00306 CY000 00An Evaluation Tool For Neutron Kerma Factors and Other Nuclear Responses.
KCUTAbstractP00584 IBMPC 00Code to Generate Minimal Cut Sets for Fault Trees.
KENO2MCNPAbstractP00541 PC586 00Conversion of Input Data between KENO V.a and MCNP File Formats, Version 5L.
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.
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.
LAZYAbstractP00595 I0360 00General Experimental Data Processing Program.
LEAP-ADDELTAbstractP00138 I0360 00Multigroup Thermal Neutron Scattering Data Generator for Hydrogen in Light Water and Deuterium in Heavy Water.
LEGENDRE FUNCTIAbstractP00108 I0360 00Legendre Functions of the First Kind and Legendre Polynomials.
LEPRICONAbstractP00277 I3033 01PWR Pressure Vessel Surveillance Dosimetry Analysis System.
LEPRICONAbstractP00277 IRISC 00PWR Pressure Vessel Surveillance Dosimetry Analysis System.
LHSAbstractP00394 PC386 00Code System to Generate Latin Hypercube and Random Samples.
LHSAbstractP00394 SUN05 00Code System to Generate Latin Hypercube and Random Samples.
LIBMAKAbstractP00087 I0360 00ANISN-Type Binary Data Processing Code System.
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.
LPTAUAbstractP00340 MNYCP 00Quasi-Random Sequence Generators.
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.
LTCAbstractP00329 IBMPC 00LMR Transient Calculation Code System (version 5).
MACK-IVAbstractP00132 I3691 00Calculation of Nuclear Response Functions from Nuclear Data in ENDF Format.
MAEROSAbstractP00466 C7600 00Code System for Multicomponent Aerosol Time Evolution.
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.
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.
MARD 4.16AbstractP00448 IBMPC 00Models And Results Database System.
MARIA SYSTEMAbstractP00359 D6000 00Code System to Calculate Cross Sections for PWR Fuel Assembly Calculations.
MARLOWE 15BAbstractP00137 MNYCP 08Computer Simulation of Atomic Collisions in Crystalline Solids (Version 15).
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.
MATEXPAbstractP00059 I0360 00Matrix Exponential Method Applied to Systems of Ordinary Differential Equations.
MATXUFAbstractP00130 I0360 00On-Line Derivative Method, Spectrum Unfolding Code System for NE-213 Liquid Fast Scintillation Proton Recoil Data.
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.
MCVIEWAbstractP00202 FM780 00View Factor Calculation for Three-Dimensional Geometries.
MESAAbstractP00223 I3033 00Non-Linear Least Squares Spectral Analysis.
METDAbstractP00197 DGMV1 00Computer Code Systems for Use with Meteorological Data.
METDAbstractP00197 I3033 00Computer Code Systems for Use with Meteorological Data.
MGA8AbstractP00542 MNYCP 00Code System to Determine Pu Isotope Abundances from Multichannel Analyzer Gamma Spectra.
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.
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.
MOCUPAbstractP00365 DALPU 00MCNP/ORIGEN Coupling Utility Programs.
MONTEBURNS 2.0AbstractP00455 MNYCP 02An Automated, Multi-Step Monte Carlo Burnup Code System.
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.
MORSEC-SP2AbstractP00142 H6000 00A Multigroup Cross Section Module for the MORSE Monte Carlo Computer Code System.
MOSRA-LIGHTAbstractP00505 MNYWS 00High-Speed Three-Dimensional Nodal Diffusion Code System.
MOXY-MOD32AbstractP00385 I0360 00BWR Core Heat Transfer Code System.
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.
MUP2AbstractP00289 I3090 00A Program to Calculate Fast Neutron Data for Medium-Heavy Nuclei.
MUXSAbstractP00187 I3033 00Generator of Multigroup Cross Sections for Charged Particle Transport Problems.
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.
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.
NE-SPECAbstractP00150 F2307 00A Code System for Unfolding a Pulse Height Distribution of Neutrons Measured by an NE-213 Organic Scintillator.
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.
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.
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.
NPTXSAbstractP00090 I0360 00Data Generator: Neutron Point Cross Sections from ENDF/B Resolved and Unresolved Resonance Parameters.
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.
NSLINKAbstractP00314 D0VAX 00NJOY SCALE LINK.
NUCHARTAbstractP00545 IBMPC 00Nuclear Properties and Decay Data Chart of Nuclides.
NUCWIZAbstractP00616 PCX86 00NucWiz
NUFACEAbstractP00284 CYXMP 00An Interface Code For The Calculation of Nuclear Responses.
NX1-NX2AbstractP00310 D0VAX 00Code System to Calculate Excitation Functions for (n,charged particle) Reactions.
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.
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.
OMCOSTAbstractP00381 I3033 00Code System for Non-fuel O & M Cost Estimation for Large Steam-Electric Power Plants.
ORCENT-2AbstractP00474 I3033 00Code System for Analysis of Steam Turbine Cycles Supplied by Light Water Reactors.
ORINC
USSO
AbstractP00439 I0360 00Code System for 1-D Implicit Heat Conduction Solution.
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.
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.
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.
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.
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.
PEQAG-2AbstractP00293 IPCAT 00A Pre-equilibrium Computer Code With Gamma Emission.
PHAZE
USSO
AbstractP00432 IBMPC 00Parametric Hazard Function Estimation.
PICESAbstractP00568 I3033 00Probabilistic Investigation of Capacity and Energy Shortages.
PICTUREAbstractP00238 IBMPC 00Combinatorial Geometry Printer Plotting.
PIXSEAbstractP00133 I0360 00A Generator of Multigroup and Multipoint Cross Sections for Thermal Reactor Calculations.
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.
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.
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.
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.
PREPRO2017AbstractP00351 MNYCP 09Pre-Processing Code System for Data in ENDF/B Format.
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.
PUFF-IVAbstractP00534 MNYCP 01Code System to Generate Multigroup Covariance Matrices from ENDF/B-VI Uncertainty Files, Version 6.0.1.
Q&AAbstractP00428 IBMPC 00Questions and Answers Based on Revised 10 CFR Part 20
QUARKAbstractP00492 PC586 00Code System for 2-Group, 3D Neutronic Kinetics Calculations Coupled to Core Thermal Hydraulics.
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.
RCSLK9AbstractP00452 IBMPC 00Code System to Calculate Reactor Coolant System Leak Rate.
RDMMAbstractP00598 I0360 00Flux Spectra from In-Pile Fast Neutron Activation Experiment.
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.
RECAPAbstractP00414 IBMPC 00Replacement Energy Cost Analysis Package.
RECAPAbstractP00414 IBMPC 01Replacement Energy Cost Analysis Package.
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.
REFLUXAbstractP00403 I3033 00Code System to Predict LWR Reflood Heat Transfer.
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.
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.
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.
RICEAbstractP00022 I0360 00A Program to Calculate Primary Recoil Atom Spectra from ENDF/B Data.
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.
RNGPAbstractP00066 I3675 00Random Number Generator Package.
ROLAIDS-CPMAbstractP00353 SUN04 00Code System to Calculate Group-Averaged Cross Sections Using the Collision Probability Method.
S1CALCAbstractP00134 I0360 00A Multigroup Thermal Neutron Scattering Law Data Generator for Hydrogen and Deuterium.
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.
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.
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.
SAND-II-SNLAbstractP00345 SUN04 00Neutron Flux Spectra Determination by Multiple Foil Activation - Iterative Method.
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.
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.
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.
SCINFULAbstractP00267 CY0MP 00Scintillator Full Response to Neutron Detection.
SCINFULAbstractP00267 D8600 00Scintillator Full Response to Neutron Detection.
SCOPEAbstractP00210 I3033 00Computer Code System for Shipping Cask Optimization and Parametric Evaluation.
SCORCH-B2AbstractP00601 I0370 00BWR Core Heating During LOCA.
SCORE-EVETAbstractP00442 C7600 00Code System for Three-Dimensional Hydraulic Reactor Core Analysis.
SCRELAAbstractP00408 SUN05 00Code System for Supercritical Water Cooled Reactor LOCA Analysis.
SECAAbstractP00104 I0360 00Evaluator of Angular Bounds for a Two-Dimensional Symmetric Gaussian Quadrature Set.
SEISIM1AbstractP00453 C7600 00Code System for Seismic Probabilistic Risk Assessment.
SELFS-3AbstractP00551 C6600 00Self-Shielding Correlation of Foil Activation Neutron Spectra Analysis by SAND-II.
SETSAbstractP00380 CDCMF 00Set Equation Transformation System.
SFHA
USSO
AbstractP00413 IBMPC 00Code System for Spent Fuel Heating Analysis.
SHC
USSO
AbstractP00493 CY000 00Seismic/Hazard Characterization in the Eastern U.S.
SIGPIAbstractP00475 D0785 00Fault Tree Cut Set System Performance.
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.
SKEWGAUSAbstractP00089 I0360 00Skewed-Gaussian Line Peak Fitting Code - Multichannel Analyzer (MCA) Spectra - Ge(Li) and Semiconductor Detectors.
SLAROMAbstractP00244 FM380 00A Code to Produce Cell Averaged Cross Sections for Fast Critical Assemblies and Fast Power Reactors.
SMACSAbstractP00396 C7600 01Probabilistic Seismic Analysis Code System.
SMAFSAbstractP00547 PC586 00Steady-State Analysis Model for Advanced Fuel Cycle Schemes.
SMOGAbstractP00216 I3033 00Code System for Neutron Cross Section Evaluation (Optical Method).
SNAKEAbstractP00135 I0360 00A Solid Angle Calculational System.
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
SORAAbstractP00174 I0360 00A Code System for Storage and Retrieval of Data from Radionuclide Analyses.
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.
SPECTRANS-2AbstractP00071 ICL00 00Neutron Spectrum Library Generation.
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.
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.
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.
STAY'SLAbstractP00113 DP010 00Least Squares Dosimetry Unfolding Code System.
STAYSL PNNLAbstractP00589 PCX86 00STAYSL PNNL Suite of Software Tools.
STRADEAbstractP00252 I3081 00Stratified Random Design.
SUGGELAbstractP00508 MNYWS 00Program Suggesting the Orbital Angular Momentum of a Neutron Resonance From the Magnitude Of Its Neutron Width.
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.
SWIFTAbstractP00031 C6600 00Monte Carlo Neutron Spectra Unfolding Code.
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.
TDOWN-IVAbstractP00172 H6000 00A Code System to Generate Composition- and Spatially-Dependent Neutron Cross Sections for Multigroup Neutronics Analysis.
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.
THACT-RRAbstractP00587 D0VAX 00Analysis of Thermal Hydraulics Transients in Research Reactor Core.
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.
THRUSHAbstractP00276 CYXMP 00Calculates Thermal Neutron Scattering Kernel.
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.
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.
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.
TPASSAbstractP00164 DP010 00A Gamma-Ray Spectral Data-Reduction and Analysis Code System.
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.
TRAXAbstractP00280 C0720 00A Program For Optics of Curved Crystal Neutron Spectrometers.
TRIGLAVAbstractP00495 PC586 00Code System to Calculate Mixed Cores in TRIGA Mark II Research Reactor.
TRISTAN-IJSAbstractP00537 IBMPC 00Steady-State Axial Temperature and Flow Velocity in Triga Channel.
TRUMPAbstractP00522 MNYCP 01Code System for Transient and Steady-State Temperature Distribution in Multidimensional Systems.
TSORTAbstractP00486 IBMPC 00Automated Technique for Nuclear Plant Training Task Assignment.
TURBINAAbstractP00604 I0370 00Reheat Steam Turbine Generator Design with Preheater and Condenser.
UHSAbstractP00390 IPS70 00Ultimate Heat Sink Cooling Pond and Spray Pond Analysis Models.
UKE-IIIAbstractP00015 I3691 00Cross Section Format Translator - UKNDL to ENDF/B.
UMG 3.3AbstractP00529 PC586 00Unfolding with Maxed and Gravel.
UNFAbstractP00521 PC586 00Code System to Calculate Multistep Compound Nucleus Neutron Cross-Sections and Spectra for Structural Materials.
UNIFY-ECNAbstractP00288 C0170 00A Program to Calculate Fast Neutron Data for Structural Materials.
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
UTSGAbstractP00379 I3033 00Code System for Calculating the Nonlinear Transient Behavior of a Natural Circulation U-Tube Steam Generator with Its Main Steam System.
VIDEO-PCAbstractP00311 IBMPC 00Super VGA Primitives Graphics System.
VIEWCXSAbstractP00514 PC586 00Interactive Graphic User Interface to View Neutron and Gamma-Ray Interaction Cross Sections.
VISA2AbstractP00445 MNYCP 00Code System to Calculate Probability of Reactor Vessel Failure.
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.
WAKEAbstractP00605 I0370 00Navier Stokes Equation with 2-D Turbulence, Stream Function, Vorticity.
WILITAbstractP00344 MNYCP 00A Utility Program for WIMS Libraries.
WIMSCORE-ENEAAbstractP00319 I3090 00Code System to Process WIMSD4 Interface Output Files and Generate Two-Group Data for Reactor Calculations.
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.
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.
XLACS-IIAAbstractP00182 I3033 00A Modified Version of XLACS-II for Processing ENDF Data into Multigroup Neutron Cross Sections in AMPX Master Library Format.
ZOTT99AbstractP00272 ALLCP 02Zero-in On The Truth; Evaluation of Correlated Data Using Partitioned Least Squares.
The Radiation Safety Information Computational Center (RSICC) is a Department of Energy Specialized Information Analysis Center (SIAC) authorized to collect, analyze, maintain, and distribute computer software and data sets in the areas of radiation transport and safety. RSICC resides in the Reactor and Nuclear Systems Division (RNSD) at Oak Ridge National Laboratory.