| Peripheral Shielding Routines (shielding research tools) |
| Package Name | Abstract | RSICC Tapelist | Title |
| 1DX | Abstract | P00096 U1108 00 | A One-Dimensional Diffusion Code System for Producing Energy Group Collapsed and Self-Shielded Cross Sections. |
| ABAREX | Abstract | P00248 MNYCP 01 | Neutron Spherical Optical-Statistical Model Code System. |
| ABLEIT-TRANS | Abstract | P00247 C0175 00 | Error Propagation Analysis for Burnup Calculation. |
| ACAT | Abstract | P00257 FM380 00 | Monte Carlo Simulation of Atomic Collisions in Amorphous Targets in the Binary Collision Approximation. |
| ACORNS | Abstract | P00264 IBMPC 01 | Analysis of Correlations Used in Neutron Spectrometry. |
| ACTIV | Abstract | P00590 I0370 00 | Sandwich Detector Activity from In-Pile Slowing-Down Spectra Experiment. |
| ACTIV-PC | Abstract | P00287 IBMPC 00 | A Program to Process Gamma or X-ray Spectra. |
| ADASAGE | Abstract | P00426 IBMPC 00 | Ada Application Development System. |
| ADEFTA 4.1 | Abstract | P00543 MNYCP 01 | Atomic Densities for Transport Analysis Script. |
| ADENA | Abstract | P00190 C0000 00 | Code System for Application of Adjusted Data in Calculating Fission-Product Decay Energies and Spectra. |
| ADENA | Abstract | P00190 I3033 00 | Code System for Application of Adjusted Data in Calculating Fission-Product Decay Energies and Spectra. |
| ADLER III | Abstract | P00058 I0360 00 | A Program to Calculate Cross Sections from Adler-Adler Resonance Parameters. |
| AIREKMOD-RR | Abstract | P00588 D0VAX 00 | Reactivity Transients in Nuclear Research Reactors |
| AIREKMOD-RR | Abstract | P00588 PCX86 01 | Reactivity Transients in Nuclear Research Reactors |
| ALARM-B2 | Abstract | P00218 I0360 00 | A Computer Code System for Analysis of a Large Break LOCA of a BWR. |
| ALICE2017 | Abstract | P00550 PCX86 06 | Statistical Model Code System to Calculate Particle Spectra from HMS Precompound Nucleus Decay. |
| ALPHA-M | Abstract | P00169 I0360 00 | Least-Squares Resolution of Gamma-Ray Spectra in Environmental Samples. |
| AMARA | Abstract | P00079 I3675 00 | Nuclear Data Adjustment Using Lagrange's Multipliers Method. |
| AMPX-77 | Abstract | P00315 ALLMF 01 | Modular Code System for Generating Coupled Multigroup Neutron-Gamma Libraries from ENDF/B. |
| AMUSE | Abstract | P00028 C6600 00 | Gamma-Ray Spectra Unfolding Code. |
| ANA | Abstract | P00356 IBMPC 00 | Code System for Gamma-Ray Spectra Analyses. |
| ANGELO-LAMBDA | Abstract | P00544 MNYCP 01 | Covariance Matrix Interpolation and Mathematical Verification. |
| ANIPLO D50 | Abstract | P00213 I0360 00 | A Digital Computer Program for Plotting Results from Calculations with the Sn Computer Program ANISN. |
| ANSIFT | Abstract | P00077 C6600 00 | ANSI Standard Fortran Sifting Program. |
| ANSIFT | Abstract | P00077 I0360 00 | ANSI Standard Fortran Sifting Program. |
| APPLE-2 | Abstract | P00111 FM200 00 | Plotter of Neutron and Gamma-Ray Spectra and Reaction Rates. |
| APPLE-2 | Abstract | P00111 I3081 00 | Plotter of Neutron and Gamma-Ray Spectra and Reaction Rates. |
| APSAI | Abstract | P00065 I3691 00 | Activity Calculations and Plotting of Neutron or Gamma-Ray Spectra Generated by Discrete Ordinates Code System ANISN. |
| AREAD | Abstract | P00088 I0360 00 | Input Data Processor for Transport Codes. |
| ART MOD2 | Abstract | P00611 PCX86 00 | Fission Product Migration in Primary System and Containment |
| ATHENA_2D | Abstract | P00431 MNYCP 00 | Code System For Simulation Of Hypothetical Recriticality Accidents in a Thermal Neutron Spectrum. |
| AUTOJOM-JOMREAD | Abstract | P00008 C6600 00 | Computer Programs to Generate or Check Coefficients for Quadratic Equations Describing 3D Geometries. |
| AXMIX-PC | Abstract | P00297 IBMPC 00 | ANISN Cross Section Code System. |
| BASACF | Abstract | P00285 IBMPC 00 | Bayesian Approach to Spectrum Adjustment with Covariance Filter. |
| BAYES | Abstract | P00205 DP010 00 | User's Guide for A General-Purpose Computer Code System for Fitting a Functional Form to Experimental Data. |
| BEACON MOD3 | Abstract | P00402 CDCMF 00 | Code System for Thermal-Hydraulic Analysis of Nuclear Reactor Containments. |
| BEST-5 | Abstract | P00591 I0370 00 | Power Reactor Fuel Cycle Optimization by Bellman Method. |
BFR
USSO | Abstract | P00449 C0176 00 | Code System for Common Cause Failure Data Analysis. |
| BLOCKAGE V2.5R | Abstract | P00377 IBMPC 00 | Code System to Calculate Potential Plugging of Emergency Core Cooling Suction Strainers by Debris in a BWR. |
| BON | Abstract | P00173 I0360 00 | A Code System for Unfolding Multisphere Spectrometer Neutron Measurements. |
| BOT3P-5.3 | Abstract | P00530 MNYCP 02 | Code System for 2D and 3D Mesh Generation and Graphical Display of Geometry and Results for Radiation Transport Codes. |
| BREESE-II | Abstract | P00143 I3033 00 | Auxiliary Routines for Implementing the Albedo Option in the MORSE Monte Carlo Code System. |
| BRMSTK | Abstract | P00044 C6600 00 | CSEWG Integral Data Testing Shielding Experiment Code System. |
| BRMSTK | Abstract | P00044 I3691 00 | CSEWG Integral Data Testing Shielding Experiment Code System. |
| BSPRP2 | Abstract | P00372 IRISC 00 | Code System to Process DORT Boundary-Flux Files. |
| BUCORST | Abstract | P00339 PC386 00 | A Code to Prepare Burnup-Dependent Multigroup Nuclear Reactor Source Terms. |
| BULK-I | Abstract | P00574 PCX86 00 | Radiation Shielding Tool for Proton Accelerator Facilities. |
| BURD | Abstract | P00582 IBMPC 00 | Bayesian Estimation in Data Analysis of Probabilistic Safety Assessment. |
| CADE | Abstract | P00567 MNYCP 00 | Multiple Particle Emission Cross-Sections by Weisskopf-Ewing Theory. |
| CAFDATS | Abstract | P00549 MNYCP 00 | Converter of Angular Fluxes of DORT, ANISN and TORT Systems. |
CALENDF-2010
OECD | Abstract | P00578 PCX86 00 | Pointwise, Multigroup Neutron Cross-Sections and Probability Tables from ENDF/B Evaluations. |
| CARP-82 | Abstract | P00131 I3033 00 | Multigroup Albedo Data Using DOT Angular Flux Results. |
| CASKCODES | Abstract | P00262 IBMPC 00 | CAPSIZE, SCOPE, AND KWIKDOSE for Shipping Cask Optimization, Dose Calculation, Parameter Evaluation, and Shielding Requirements. |
| CASTHY | Abstract | P00316 FM000 00 | Statistical Model Calculation for Neutron Cross Sections and Capture Gamma-Ray Spectra. |
| CCRMN | Abstract | P00366 MNYCP 00 | Monte Carlo Simulation of the Coupled Transport of Electrons and Photons. |
| CEAR-PPU | Abstract | P00528 PC586 00 | Code System for Monte Carlo Simulation of Detector Pulse Pile Up. |
| CECP-BWR | Abstract | P00370 PC386 00 | Estimating Boiling Water Reactor Decomissioning Costs. |
| CECP-PWR | Abstract | P00371 PC386 00 | Estimating Pressurized Water Reactor Decomissioning Costs. |
| CEM03.03 | Abstract | P00532 MNYCP 01 | Monte-Carlo Code System to Calculate Nuclear Reactions in the Framework of Improved Cascade-Exciton Model. |
CEMENT 1.02
USSO | Abstract | P00412 IBMPC 00 | Computer Code System for the Estimation of Long-Term Performance of Cement-Based Materials. |
| CERPI-CEREL | Abstract | P00147 I0360 00 | Code Systems for Automatic Analysis of Gamma-Ray Spectra Obtained with Ge(Li) Detectors. |
| CGS 11.4 | Abstract | P00243 MFMWS 03 | Common Graphics System. |
| CHENDF 7.02 | Abstract | P00333 MNYCP 05 | Codes for Handling ENDF/B-V and ENDF/B-VI Data. |
| COAG-II | Abstract | P00070 I0360 00 | Calculation of the Westcott Epithermal Index and the Westcott 2200 m/s Neutron Flux. |
| COBRA-3C-RERTR | Abstract | P00606 I0370 00 | COBRA-3C-RERTR |
| COBRA4I | Abstract | P00419 MNYCP 00 | Code Sytem to Calculate Rod-Bundle and Core Thermal-Hydraulics. |
| COBRA-EN | Abstract | P00507 MNYCP 01 | Thermal-Hydraulic Transient Analysis of Reactor Cores. |
| COBRA-SFS VERSION 6.0 | Abstract | P00614 MNYCP 02 | COBRA-SFS Thermal-Hydraulic Analysis of Multi-Assembly Spent Fuel Storage and Transportation Systems. |
| CODAC (2) | Abstract | P00073 I0360 00 | For TIMOC 72, Monte Carlo Three-Dimensional Neutron Transport Code's Data Generator. |
| COG LIBMAKER | Abstract | P00607 MNYCP 00 | LIBMAKER |
| COGAP | Abstract | P00375 MNYCP 01 | Nuclear Power Plant Containment Hydrogen Control System Evaluation Code. |
| COMAND | Abstract | P00091 I0360 00 | A Multigroup ANISN Cross Section Data Library Collapsing Code System. |
| COMBINE-PC | Abstract | P00286 IBMPC 00 | Code System to Compute Neutron Spectra and ENDF/B Version 5 Based Multigroup Neutron Constants. |
| COMIDA | Abstract | P00343 MNYCP 00 | Radionuclide Food Chain Model for Acute Fallout Deposition. |
COMMIX-1B
USSO | Abstract | P00393 DVX11 00 | 3-D Single-Phase Thermal Hydraulics |
COMMIX-1B
USSO | Abstract | P00393 I3033 00 | 3-D Single-Phase Thermal Hydraulics |
COMMIX-1C
USSO | Abstract | P00393 MNYCP 00 | 3-D Single-Phase Thermal Hydraulics |
| COMNUC3B | Abstract | P00302 CYXMP 00 | A Compound Nucleus Analysis Program. |
| COMPAR | Abstract | P00240 C0170 00 | Compares Multigroup Cross Sections Generated by NJOY, GROUPIE, FLANGE-II, ETOG-3 and XLACS. |
| COMPARE-MOD1A | Abstract | P00410 C7600 00 | Code System to Calculate Transient Flow With Heat Sinks & Doors. |
| COMPARE-MOD1A | Abstract | P00410 I3033 00 | Code System to Calculate Transient Flow With Heat Sinks & Doors. |
| COMPASS 1.0.0 | Abstract | P00520 PC586 00 | Computerization of MARSSIM for Planning and Assessing Site Surveys. |
| COMPBRN3 | Abstract | P00389 PC386 00 | Code System for Modeling Compartment Fires. |
| COMPLOT | Abstract | P00259 IBMMF 00 | Convert EXFOR Format Data to Computation Format and Plot Comparisons of EXFOR and ENDF/B Evaluated Data (Version 86-1). |
| CONFOLD | Abstract | P00053 C6600 00 | Least-Structure Unfolding Code System for Measured Neutron and Gamma-Ray Spectra. |
| CONFOLD | Abstract | P00053 I0360 00 | Least-Structure Unfolding Code System for Measured Neutron and Gamma-Ray Spectra. |
| CONTEMPT4 | Abstract | P00397 MNYCP 00 | Code System for PWR & BWR Multicompartment Containment Analysis. |
CONTEMPT-LT28B
USSO | Abstract | P00387 C7600 00 | Code System to Predict Containment Pressure-Temperature Response To a Loss-Of-Coolant Accident. |
| CONVERT | Abstract | P00036 C6600 00 | An IBM-to-CDC Program Conversion Code. |
| COOL-C | Abstract | P00017 I0360 00 | Spectra Unfolding Codes. |
| CORTES | Abstract | P00404 I0360 00 | Code System for Thermal & Mechanical Analysis of Tees. |
| CRECTJ5 | Abstract | P00250 D0780 00 | A Computer Program for Compilation of Evaluated Nuclear Data in ENDF/B Format. |
| CRESO | Abstract | P00184 I3081 00 | Resonance Data-Handling Code System. |
| CUPED | Abstract | P00032 I3675 00 | Scintillation Spectrometer Polyenergetic Gamma Photon Experimental Distributions Unfolding Code. |
| D2O | Abstract | P00398 PC486 00 | Code System for Computing Thermodynamic and Transport Properties of D2O. |
| DANCOFF3 | Abstract | P00279 D8810 00 | Calculates Dancoff Correction. |
| DANCOFF-MC | Abstract | P00509 MNYCP 00 | Code System for Monte Carlo Calculation of Dancoff Factors in Irregular Geometries. |
DANESS V1.0
FEDC | Abstract | P00555 MNYCP 00 | Dynamic Analysis of Nuclear Energy System Strategies. |
| DANTE | Abstract | P00185 I0370 00 | Unfolding Code System for Energy Spectra Evaluation for Dosimetry Purposes. |
| DASQHE | Abstract | P00278 D8810 00 | Calculates Dancoff Corrections Factors. |
| DATINIT | Abstract | P00258 DGMV1 00 | Interactive Program To Access Photon Interaction Data. |
| DENIS | Abstract | P00082 I0360 00 | Monte Carlo Simulation of the Capture and Detection of Neutrons with Large Liquid Scintillators. |
| DEPLETOR | Abstract | P00523 MNYCP 00 | Code System to Provide Depletion Capability to the U.S. NRC PARCS Code |
DEPOSITION
FEDC | Abstract | P00420 IBMPC 00 | Code System to Calculate Particle Penetration Through Aerosol Transport Lines. |
| DETAN 95 | Abstract | P00361 MNYCP 00 | Code System to Calculate Spectrum-Averaged Cross Sections and Detector Responses in Neutron Spectra. |
| DIFBAS | Abstract | P00334 MNYCP 00 | A Bayesian Approach to Unfolding a Neutron Spectrum from a Spectrum of Recoiled Protons. |
| DIMEN | Abstract | P00341 IBMPC 00 | Code System for Isotope Identification by Gamma-Ray Analysis. |
| DINT | Abstract | P00049 C6600 00 | Multigroup Coherent-Incoherent Cross Section Data Generator for Photon Transport Calculations. |
| DINT | Abstract | P00049 I0360 00 | Multigroup Coherent-Incoherent Cross Section Data Generator for Photon Transport Calculations. |
| DOMINO | Abstract | P00064 I0360 00 | A General Purpose Code System for Coupling Discrete Ordinates and Monte Carlo Radiation Transport Calculations. |
| DOMINO-II | Abstract | P00162 I3033 00 | A General Purpose Code System for Coupling Discrete Ordinates and Monte Carlo Radiation Transport Calculations. |
| DOMUS | Abstract | P00301 IPCXT 00 | A Program for Decomposing A Two-Dimensional Spectrum. |
| DOQDP | Abstract | P00110 I0360 00 | Discrete Ordinates Quadrature Generator. |
| DORGLIB | Abstract | P00181 I0360 00 | An Interactive Program for Displaying Nuclide Decay and Generation Data Based on ORIGEN Data Library. |
| DORIAN | Abstract | P00425 IBMPC 00 | Code System to Implement Bayes Method for Plant Aging Risk Analysis. |
| DSNP | Abstract | P00592 I3033 00 | Dynamic Simulation Nuclear Power. |
| DSNQUAD | Abstract | P00251 IPCXT 00 | Calculates Angular Quadrature Weights and Cosines. |
| DUFOLD | Abstract | P00042 I0360 00 | Derivative Unfolding Code - Determination of Neutron Spectra from NE-213 Pulse Height Data. |
| DWBA07/DWBB07 | Abstract | P00338 MNYCP 01 | Code System for Inelastic and Elastic Scattering with Nucleon-Nucleon Potential |
| DWUCK-CHUCK | Abstract | P00546 MNYCP 00 | Nuclear Model Code System for Distorted Wave Born Approximation and Coupled Channel Calculations. |
| DYN3D/M2 | Abstract | P00579 I3090 00 | Reactivity Transients in Light H2O Reactors with Hexagonal Geometry. |
| ECIS-12 | Abstract | P00612 MNYCP 00 | Code System to Solve the Coupled Differential Equations Arising in Nuclear Model Calculations. |
| EDISTR | Abstract | P00191 I3033 00 | Prepares a Nuclear Decay Data Base for Internal Radiation Dosimetry Calculations. |
| EDITOR | Abstract | P00035 I0360 00 | Alters Mode, Copies, Merges, Punches, Edits, or Adds to ENDF/B-Formatted Data on Tapes or Cards. |
| EEDB | Abstract | P00531 MNYCP 00 | The Energy Economic Data Base. |
| ELAN | Abstract | P00141 ICL00 00 | Neutron Cross-Section Self-Shielding Code System. |
| ELIESE-3 | Abstract | P00003 I0370 00 | Analyses of Elastic and Inelastic Scattering Cross Sections. |
| EMPIRE-II | Abstract | P00497 PC586 01 | Comprehensive Nuclear Model Code, Nucleons, Ions Induced Cross-Sections. |
| ENBAL2 | Abstract | P00160 I0370 00 | A Program to Generate Multigroup Neutron Kerma Factors. |
| ENDVER/GUI | Abstract | P00572 PCX86 00 | The ENDF File Verification Support Package. |
| ENLOSS | Abstract | P00047 C6600 00 | Calculation of Energy Loss of Charged Particles. |
| ENTOSAN | Abstract | P00188 C0175 00 | Code System for Calculating Fine-Group Dosimetry Cross Section Values from ENDF/B Data. |
| ENTOSAN | Abstract | P00188 D8810 00 | Code System for Calculating Fine-Group Dosimetry Cross Section Values from ENDF/B Data. |
| ENTREE 1.4.0 | Abstract | P00519 MNYWS 00 | BWR Core Simulation System for Space and Time Dependent Coupled Phenomena. |
EPIPE
USSO | Abstract | P00485 CY000 00 | Code System for Static and Dynamic Piping System Analysis. |
| EQUIVA-1.1 | Abstract | P00323 IMFPC 00 | Generation of Environment-Insensitive Equivalent Diffusion Theory Parameters for PWR Reflector Regions. |
| EQUIVA-2 | Abstract | P00324 IMFPC 00 | Generation of Environment-Insensitive Equivalent Diffusion Theory Parameters for PWR Reflector Regions. |
| ERIC-2 | Abstract | P00119 I0360 00 | Calculator of Resonance Integral and Effective Capture and Fission Cross Sections for Fissile and Non-Fissile Nuclides - Thermal or Fast Reactors. |
| ERINNI | Abstract | P00219 I0360 00 | Optical Model Calculation of Multiple Cascading Particle Emissions. |
| ERRORJ | Abstract | P00526 MNYCP 03 | Multigroup Covariance Matrices Generation from ENDF/B-6 Format. |
| ESTIMA | Abstract | P00201 I3033 00 | A Code System for Calculating Average Parameters from Sets of Resolved Resonance Parameters. |
| ETHEL | Abstract | P00217 I0360 00 | Code System for Generating Cross Sections for PSR-128/THERMOS. |
| ETOE-2 | Abstract | P00585 I3033 00 | Cross-Sections Library for Program MC**2 Generator from ENDF/B. |
| EURCYL | Abstract | P00076 I0370 00 | Finite Element Three-Dimensional Mesh Generator for Cylinder - Cylinder Intersections. |
| EVALPLOT | Abstract | P00211 I3081 00 | A Program to Plot Data in the Evaluated Nuclear Data File/Version B Format. |
| EVAP | Abstract | P00010 I0360 00 | Calculation of Particle Evaporation from Excited Compound Nuclei. |
| EVNTRE | Abstract | P00465 D0VAX 00 | Code System for Event Progression Analysis for PRA. |
| EXCURS-3-RR | Abstract | P00586 D0VAX 00 | Kinetics of Research Reactor Reactivity Transient Analysis. |
| EXIFON2.0 | Abstract | P00305 IPCXT 01 | A Model for Statistical Multistep Direct and Multistep Compound Reactions. |
| EZVIDEO | Abstract | P00237 IBMPC 00 | Graphics Routines for the IBM PC. |
| F5TAB | Abstract | P00221 D0780 00 | Code System for Converting Energy Distribution Cross Section Data to Tabulated Data. |
| FAMREC | Abstract | P00167 C7600 01 | Fuel Assembly Mechanical Response Code System. |
| FANAC | Abstract | P00179 I3033 00 | A Shape Analysis Code Package for Resonance Parameter Extraction from Neutron Capture Data for Light- and Medium-Weight Nuclei. |
| FANAL | Abstract | P00178 I3033 00 | A Least-Squares Shape Analysis Code System. |
| FANG | Abstract | P00140 C0000 00 | An Angular Folding Code System for Channel Theory Analysis. |
| FANG | Abstract | P00140 I0360 00 | An Angular Folding Code System for Channel Theory Analysis. |
| FASTGRASS | Abstract | P00479 MNYCP 00 | Code System to Predict Fission Product Release in Ubase Fuels. |
| FASTPLOT 1.0 | Abstract | P00354 IBMPC 00 | Interface to Microsoft FORTRAN Graphics. |
| FATDUD | Abstract | P00080 I0360 00 | Foil Activation Data Unfolding Code System. |
| FBSAM | Abstract | P00103 I0360 00 | User-Storage - Magnetic Disk Data Manipulator. |
| FDMXPC | Abstract | P00322 IPCAT 00 | Code System for Calculation of Neutron Transmission and Other Functionals from Evaluated Data in ENDF Format. |
| FEAST METAL | Abstract | P00563 MNYCP 00 | Fuel Engineering and Structural Analysis Tool. |
| FEDGROUP-3 | Abstract | P00123 I0360 00 | Code System for Processing Evaluated Nuclear in ENDF/B, KEDAK or UKNDL Formats into Constants for Reactor Physics Calculation. |
| FEDGROUPC86REV3 | Abstract | P00194 MNYCP 01 | Code System for Processing Evaluated Nuclear in ENDF/B, KEDAK or UKNDL Formats into Constants for Reactor Physics Calculation. |
| FEDGROUP-R | Abstract | P00349 MNYCP 00 | Code System for Processing Evaluated Nuclear in ENDF/B, KEDAK or UKNDL Formats into Constants for Reactor Physics Calculation. |
| FEMAXI 6 VER.1 | Abstract | P00536 IBMPC 00 | Code System for Light Water Reactor Fuel Analysis. |
| FEP 4.16 | Abstract | P00440 IBMPC 00 | Fault-tree, Event tree, & P&ID Editors. |
| FERDO/FERD | Abstract | P00102 I3033 00 | Multichannel Neutron and Gamma-Ray Spectrum Matrix Unfolding Code Systems. |
| FERDOR | Abstract | P00017 I7090 00 | Spectra Unfolding Codes. |
| FERDOR | Abstract | P00017 U1108 00 | Spectra Unfolding Codes. |
| FERD-PC | Abstract | P00273 IBMPC 00 | Interactive Multichannel Neutron and Gamma-Ray Spectrum Matrix Unfolding Code System. |
| FERRET | Abstract | P00145 U0000 00 | Least-Squares Solution to Nuclear Data and Reactor Physics Problems. |
| FIGERO | Abstract | P00149 C0000 00 | Processing Codes for Generating Multigroup Neutron Cross Sections from ENDF/B for Use in Discrete Ordinates Calculations. |
| FIRAC | Abstract | P00444 CY000 00 | Nuclear Facilities Fire Accident Model |
| FITOCO | Abstract | P00189 C0175 00 | Converter of Fine-Group Flux Density and Cross Section Data to Coarse Group Values. |
| FLANGE-ORNL | Abstract | P00566 I0360 00 | Flanged Pipe Joint Stress Analysis, Internal Pressure, Moment Loads, Temperature. |
| FLODIS | Abstract | P00417 I0360 00 | Code System to Calculate Thermal Response of FSV HTGR Core. |
| FLOWPLOT II | Abstract | P00234 I3033 00 | Fluid Dynamics and Heat Transfer Plotting Package. |
| FLUSH | Abstract | P00043 C6600 00 | Spectral Unfolding Code - Stepwise Regression of System Response Functions. |
| FLYSPEC-SHORTS | Abstract | P00196 C7600 00 | Neutron Unfolding Code System for Reducing Proton-Recoil Pulse-Height Obtained with NE-213 Liquid Scintillator. |
| FORECAST V3.0 | Abstract | P00384 IBMPC 00 | Forecast Regulatory Effects Cost Analysis Program. |
| FORIST | Abstract | P00092 C0000 00 | Neutron Spectrum Unfolding Code System - Iterative Smoothing Technique. |
| FORIST | Abstract | P00092 I0360 00 | Neutron Spectrum Unfolding Code System - Iterative Smoothing Technique. |
| FORSEN | Abstract | P00170 I0360 00 | A Multigroup Processing Code for Use with Sensitivity Profiles to Assess the Effect of Cross Section Changes. |
| FORSIM VI | Abstract | P00078 C6600 00 | A Fortran-Oriented Simulation Package for the Automated Solution of Partial and Ordinary Differential Equation Systems. |
| FOURACES | Abstract | P00183 I0370 00 | Code System for Producing Spectrum Weighted, Group Averaged Cross Sections from ENDF/B, KEDAK, or UK Libraries. |
| FRANCO | Abstract | P00363 MNYCP 00 | Finite Element Fuel Rod Analysis Code System. |
| FRANTIC3 | Abstract | P00406 CDCMF 00 | Time-Dependent Reliability Analysis. |
| FRAPCON2 | Abstract | P00517 MFMWS 00 | Fuel Rod Thermal-Mechanical Behavior. |
FRAPT6/MOD1
USSO | Abstract | P00436 C0176 00 | Code System for Transient Analysis of Fuel Rods. |
FRAPT6/V21
USSO | Abstract | P00436 C0176 01 | Code System for Transient Analysis of Fuel Rods. |
| FREEFORM | Abstract | P00081 I0360 00 | Free-Form Input Reading Routines. |
| F-SCORE | Abstract | P00617 PCX86 00 | F-Score Nuclide ID Scoring Applications |
| FUELSDATA | Abstract | P00446 C7600 00 | Code System to Model Verification Fuel Rod Data. |
| GABAS | Abstract | P00175 U1108 00 | A Code System for Generating Composite Time-Dependent Fission Produce Spectra. |
| GADRAS-DRF-19.4.0 | Abstract | P00610 PCX86 06 | Gamma Detector Response and Analysis Software–Detector Response Function. |
| GAINCALB | Abstract | P00056 I0360 00 | Determination of the Gain Used with Organic Scintillation Detect. |
| GALAXY-6 | Abstract | P00098 I0370 00 | Neutron Multigroup Cross Section Processor. |
| GAMAN | Abstract | P00083 DP010 00 | Qualitative and Quantitative Evaluation of Ge(Li) Gamma-Ray Spectra. |
| GAMANAL | Abstract | P00506 D0VAX 00 | Code System for Computerized Quantitative Analysis By Gamma-Ray Spectrometry. |
| GAMIDENT | Abstract | P00154 C0000 00 | A Program to Aid in the Identification of Unknown Materials by Gamma-ray Spectroscopy. |
| GAMLEG-75 | Abstract | P00086 C7600 00 | Multigroup Cross Section Generator for Photon Transport Calculations. |
| GAMMA | Abstract | P00095 I0360 00 | Monte Carlo Code System for Calculating Efficiencies and Response Functions of NaI(Tl) Crystals for Gamma Rays from Thick Disk Sources. |
| GAMX1 | Abstract | P00209 I0370 00 | A Computer Code System for Evaluating Spectra Peak Areas. |
| GAPCON-THERMAL | Abstract | P00499 C7600 00 | Code System to Calculate Fuel Steady State & Transient Behavior. |
| GAROL | Abstract | P00033 I7090 00 | Calculation of Resonance Neutron Absorption in Two-Region Problems. |
| GAUSS V | Abstract | P00045 I0360 00 | A Code system for Analysis of Gamma-Ray Spectra from Ge(Li) Spectrometers. |
| GAUSS VII | Abstract | P00045 C0000 00 | A Code system for Analysis of Gamma-Ray Spectra from Ge(Li) Spectrometers. |
| GCI | Abstract | P00421 IBMPC 00 | Generic Communications Index |
| GECINX | Abstract | P00193 H6000 00 | A Code System for Collapsing Multigroup Cross Sections in CCCC Format. |
| GEF | Abstract | P00564 PCX86 03 | A GEneral description of the Fission process. |
| GELI2/SPAN2 | Abstract | P00094 I0360 00 | Calculation of Nuclide Abundaces from Multichannel Gamma-ray Spectra. |
| GEM | Abstract | P00540 PC586 00 | Monte-Carlo Code for Simulating a Decaying Process of an Excited Nucleus. |
| GENRD | Abstract | P00040 C6600 00 | Free Format Card Input Processor. |
| GENRD | Abstract | P00040 I0360 00 | Free Format Card Input Processor. |
| GERES | Abstract | P00241 I0370 00 | A Code to Produce Cross-Section Libraries for ANISN Based on Heterogeneous Fast Reactor Cell Calculations Using MC2II Data. |
| GGC-3 | Abstract | P00012 I3565 00 | Multigroup Cross Section Code System for Use in Diffusion and Transport Codes. |
| GGC-3 & GGC-4 | Abstract | P00012 I3675 00 | Multigroup Cross Section Code System for Use in Diffusion and Transport Codes. |
| GGC-4 | Abstract | P00012 U1108 00 | Multigroup Cross Section Code System for Use in Diffusion and Transport Codes. |
| GGTC-ENEL | Abstract | P00128 I0360 00 | Code System for Producing Few-Group Neutron Cross Sections from Multigroup Data Libraries. |
| GIFT | Abstract | P00124 C0076 00 | A Combinatorial Geometry Code System with Model Testing Routines. |
| GIFT | Abstract | P00124 D0VAX 00 | A Combinatorial Geometry Code System with Model Testing Routines. |
| GIFT | Abstract | P00124 U0000 00 | A Combinatorial Geometry Code System with Model Testing Routines. |
| GIP | Abstract | P00229 IBMPC 00 | Group-Organized Cross-Section Input Program. |
| GIRAFFE | Abstract | P00304 I3033 00 | General Isotope Release Analysis For Failed Elements. |
| GLUCS | Abstract | P00192 D0VAX 00 | A Generalized Least-Squares Code System for Updating Cross Section Evaluations with Correlated Data Sets. |
| GMA | Abstract | P00367 MNYCP 00 | Code System for Calculation of Reactor Accident Consequences. |
| GNASH-FKK | Abstract | P00535 MNYCP 00 | Pre-equilibrium, Statistical Nuclear-Model Code System for Calculation Cross Sections and Emission Spectra. |
| GOFRR | Abstract | P00127 I0360 00 | Generator of Graphical Output of DOT and ANISN Fluxes and Reaction Rates. |
| GRASS-SST | Abstract | P00489 MNYCP 00 | Code System to Predict Fission-Gas Release & Fuel Swelling. |
| GRESS 3.0 | Abstract | P00231 MFMWS 02 | Gradient Enhanced Software System. |
| GRETEL | Abstract | P00100 I0370 00 | Analyzer and Processor of Ge(Li) Gamma-Ray Spectrometric Data. |
| GRFPAK | Abstract | P00478 I0360 00 | Code System to Plot CORTES FEM Results. |
| GROUPXS | Abstract | P00246 C0740 00 | Processing of Double-Differential Cross Sections in the New ENDF-VI Format. |
| GRPANL | Abstract | P00321 D0VAX 00 | Code System for Analyzing Ge and Alpha-Particle Detector Spectra. |
| GRUCON | Abstract | P00615 MNYCP 00 | Data Processing for Evaluated Working libraries (transport and shielding) |
| GT2R2 | Abstract | P00483 ALLMF 00 | Code System to Calculate Fuel Rod Thermal Performance. |
| HAARM-3 | Abstract | P00401 CDCMF 00 | Aerosol Behavior Log-Normal Distribution Model. |
| HASSAN | Abstract | P00593 I0370 00 | Time-Dependent Temperature Distribution and Stress and Strain in HTR Fuel Pins. |
| HAUSER*5 | Abstract | P00152 U0000 00 | Code System for Calculating Nuclear Cross Sections. |
| HEATING 7.3 | Abstract | P00199 MNYCP 06 | Multidimensional, Finite-Difference Heat Conduction Analysis Code System. |
HECTR 1.5+
USSO | Abstract | P00457 CY000 00 | Hydrogen Event Containment Response Code System. |
| HEITLER | Abstract | P00004 I7030 00 | Cross Section Generator. |
| HSI-DRG | Abstract | P00435 IBMPC 00 | Code System for Use with Human System Interface Design Review Guidelines. |
| HYPERMET | Abstract | P00101 C3800 00 | Gamma-Ray Spectra Analyzer Germanium Detector. |
| HYPERMET | Abstract | P00101 F150F 00 | Gamma-Ray Spectra Analyzer Germanium Detector. |
| HYPERMET | Abstract | P00101 I0360 00 | Gamma-Ray Spectra Analyzer Germanium Detector. |
| ICAR | Abstract | P00291 IPCAT 00 | A Code For Combinatorial Calculation of Level Densities. |
| IER | Abstract | P00024 I3675 00 | A Gauss-based Quadrature Formula Applied to Sievert's Integral. An Exponential Integral Routine. |
| IMPORTANCE | Abstract | P00407 I0370 00 | FTA Basic Event & Cut Set Ranking. |
| INFLTB | Abstract | P00313 ALLCP 00 | Gamma-Ray Absorption Coefficient Calculation. |
| INGEN | Abstract | P00207 C0000 00 | A General-Purpose Mesh Generator for Finite Element Codes. |
| INTRIGUE-II | Abstract | P00054 I0360 00 | Logarithmic and Semilogarithmic CALCOMP Plot Routines. |
| IRRAS 4.16 | Abstract | P00386 IBMPC 04 | Code System to Calculate Integrated Reliability and Risk Analysis. |
| ITER-2 | Abstract | P00148 C0000 00 | Codes for Unfolding Activation Detector Data and Pulse Height Spectra. |
| KAOS-V | Abstract | P00306 CY000 00 | An Evaluation Tool For Neutron Kerma Factors and Other Nuclear Responses. |
| KCUT | Abstract | P00584 IBMPC 00 | Code to Generate Minimal Cut Sets for Fault Trees. |
| KENO2MCNP | Abstract | P00541 PC586 00 | Conversion of Input Data between KENO V.a and MCNP File Formats. |
| KFIX | Abstract | P00409 C7600 00 | Code System to Calculate Transient 2-Dimensional 2-Fluid Flow Dynamics. |
| KFIX 3D | Abstract | P00383 C7600 00 | Code System to Calculate Three-Dimensional Extension Two-Phase Flow Dynamics. |
| LAPHANO | Abstract | P00020 C6600 00 | PO Multigroup Photon Production Matrix and Source Vector Code for ENDF Data. |
| LAPHANO | Abstract | P00020 I0360 00 | PO Multigroup Photon Production Matrix and Source Vector Code for ENDF Data. |
LAPUR6
USSO | Abstract | P00395 PC586 02 | BWR Core Stability Measurements. |
| LAZY | Abstract | P00595 I0360 00 | General Experimental Data Processing Program. |
| LEAP-ADDELT | Abstract | P00138 I0360 00 | Multigroup Thermal Neutron Scattering Data Generator for Hydrogen in Light Water and Deuterium in Heavy Water. |
| LEGENDRE FUNCTI | Abstract | P00108 I0360 00 | Legendre Functions of the First Kind and Legendre Polynomials. |
| LEPRICON | Abstract | P00277 I3033 01 | PWR Pressure Vessel Surveillance Dosimetry Analysis System. |
| LEPRICON | Abstract | P00277 IRISC 00 | PWR Pressure Vessel Surveillance Dosimetry Analysis System. |
| LHS | Abstract | P00394 PC386 00 | Code System to Generate Latin Hypercube and Random Samples. |
| LHS | Abstract | P00394 SUN05 00 | Code System to Generate Latin Hypercube and Random Samples. |
| LIBMAK | Abstract | P00087 I0360 00 | ANISN-Type Binary Data Processing Code System. |
| LOGNORML | Abstract | P00307 IPCAT 00 | Lognormal Probability Analysis Code System for Estimating Doses in Epidemiologic Studies. |
| LOOM-P | Abstract | P00153 F2307 00 | A Finite Element Mesh Generation Code System with On-Line Graphic Display. |
| LOUHI82 | Abstract | P00236 U1108 00 | General Purpose Unfolding Program with Linear and Nonlinear Regularizations. |
| LPTAU | Abstract | P00340 MNYCP 00 | Quasi-Random Sequence Generators. |
| LSL-M2 | Abstract | P00233 D6220 00 | Least-Squares Logarithmic Adjustment of Neutron Spectra. |
| LSL-M2 | Abstract | P00233 IBMPC 00 | Least-Squares Logarithmic Adjustment of Neutron Spectra. |
| LSMOD-GLSMOD | Abstract | P00342 IBMPC 00 | A Least-Squares Computational Tool Kit. |
| LTC | Abstract | P00329 IBMPC 00 | LMR Transient Calculation Code System. |
| MACK-IV | Abstract | P00132 I3691 00 | Calculation of Nuclear Response Functions from Nuclear Data in ENDF Format. |
| MAEROS | Abstract | P00466 C7600 00 | Code System for Multicomponent Aerosol Time Evolution. |
| MAINTAIN | Abstract | P00067 I0360 00 | Code System for Use in Maintaining and Revising Card Image Files on Tape. |
| MANYFILE | Abstract | P00068 I0360 00 | Utility Routine - Manipulation of Data Sets Between Various I-O Devices. |
| MARCH2 | Abstract | P00473 CDCMF 00 | Code System to Model LWR Meltdown Accident Response. |
| MARCOPOLO | Abstract | P00225 I0360 00 | Code System for Calculating the Radial and Axial Neutron Diffusion Coefficients in One-Group and Multigroup Theory. |
| MARD 4.16 | Abstract | P00448 IBMPC 00 | Models And Results Database System. |
| MARIA SYSTEM | Abstract | P00359 D6000 00 | Code System to Calculate Cross Sections for PWR Fuel Assembly Calculations. |
| MARLOWE 15B | Abstract | P00137 MNYCP 08 | Computer Simulation of Atomic Collisions in Crystalline Solids. |
| MARS | Abstract | P00117 I0360 00 | Collection of Computer Codes for Manipulating Multigroup Cross Section Libraries in AMPX or CCCC Formats. |
| MATEXP | Abstract | P00059 I0360 00 | Matrix Exponential Method Applied to Systems of Ordinary Differential Equations. |
| MATXUF | Abstract | P00130 I0360 00 | On-Line Derivative Method, Spectrum Unfolding Code System for NE-213 Liquid Fast Scintillation Proton Recoil Data. |
| MAX-XTREME | Abstract | P00001 C0000 00 | Generalized Several-Constraint LaGrange Multiplier. |
| MAZE II | Abstract | P00041 U1108 00 | Spectral Unfolding Code. |
| MAZE-1 | Abstract | P00041 C6600 00 | Spectral Unfolding Code. |
| MC**2-2 | Abstract | P00350 SUN05 01 | Multigroup Cross Section Generation Code for Fast Reactor Analysis. |
| MC**2-3 | Abstract | P00577 MNYCP 00 | Multigroup Cross Section Generation Code for Fast Reactor Analysis. |
| MC**2-3 EXE | Abstract | P00577 MNYCP 01 | Multigroup Cross Section Generation Code for Fast Reactor Analysis. |
| MCVIEW | Abstract | P00202 FM780 00 | View Factor Calculation for Three-Dimensional Geometries. |
| MESA | Abstract | P00223 I3033 00 | Non-Linear Least Squares Spectral Analysis. |
| METD | Abstract | P00197 DGMV1 00 | Computer Code Systems for Use with Meteorological Data. |
| METD | Abstract | P00197 I3033 00 | Computer Code Systems for Use with Meteorological Data. |
| MGA8 | Abstract | P00542 MNYCP 00 | Code System to Determine Pu Isotope Abundances from Multichannel Analyzer Gamma Spectra. |
| MICAP | Abstract | P00261 I3033 00 | A Monte Carlo Code System for Analysis of Ionization Chamber Responses. |
| MICROX-2 | Abstract | P00374 MNYCP 02 | Code System to Create Broad-Group Cross Sections with Resonance Interference and Self-Shielding from Fine-Group and Pointwise Cross Sections. |
| MIGROS3 | Abstract | P00265 I0370 00 | A Code for the Generation of Group Constants for Reactor Calculations from Neutron Nuclear Data in KEDAK Format. |
| MINET | Abstract | P00490 CY000 00 | Momentum Integral Network Method for Thermal-Hydraulic Systems Analysis. |
| MINIGAL | Abstract | P00180 I3033 00 | Neutron Cross Section Processing System for Calculating Average Values from Data in the Standard United Kingdom Nuclear Data Library Format. |
| MINTEQ | Abstract | P00494 DVX11 00 | Code System to Model Aqueous Geochemical Equilibria. |
| MINX | Abstract | P00105 C6600 00 | Multigroup Interpretation of Nuclear X-Sections from ENDF/B Standard CCCC-III Interface Formats. |
| MINX | Abstract | P00105 I0360 00 | Multigroup Interpretation of Nuclear X-Sections from ENDF/B Standard CCCC-III Interface Formats. |
| MISSIONARY | Abstract | P00114 I0360 00 | ENDF/B to NDL Data Format Converter. |
| MIXEN | Abstract | P00318 IRISC 00 | Code System to Replace Files 4 and 6 of ENDF-6 with Files 4 and 5 of ENDF/B-IV. |
| MOCUP | Abstract | P00365 DALPU 00 | MCNP/ORIGEN Coupling Utility Programs. |
| MONTEBURNS 2.0 | Abstract | P00455 MNYCP 02 | Automated, Multi-Step Monte Carlo Burnup Code System. |
| MORECA | Abstract | P00411 PC386 00 | Computer Code System for Simulating Modular High-Temperature Gas Cooled Reactor Core Heatup. |
| MORN | Abstract | P00062 I0360 00 | Calculation of the Response of Sodium Iodide Crystals to Gamma Rays. |
| MORSEC-SP2 | Abstract | P00142 H6000 00 | Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System. |
| MOSRA-LIGHT | Abstract | P00505 MNYWS 00 | High-Speed Three-Dimensional Nodal Diffusion Code System. |
| MOXY-MOD32 | Abstract | P00385 I0360 00 | BWR Core Heat Transfer Code System. |
| MRSPAK | Abstract | P00212 DVX11 00 | A Code System To Generate a Text File Containing Combinatorial Geometry Data Corresponding to PADL2 Geometry. |
| MSM-SOURCE | Abstract | P00369 MNYCP 00 | Code System for Generation of Input Data for MCNP. |
| MUP2 | Abstract | P00289 I3090 00 | A Program to Calculate Fast Neutron Data for Medium-Heavy Nuclei. |
| MUXS | Abstract | P00187 I3033 00 | Generator of Multigroup Cross Sections for Charged Particle Transport Problems. |
| NAISAP | Abstract | P00085 F2306 00 | Theory and Use of Gamma-Ray Spectrum Analysis Codes for NaI(Tl) Detectors. |
| NANICK | Abstract | P00120 I0360 00 | Infinitely-Diluted Multigroup Cross-Section Generator - from ENDF/B. |
| NASIF-NARES | Abstract | P00121 I0360 00 | A Code System for Computing Shielding Factors from ENDF/B Tapes. |
| NAUA-MOD5 NAUA-MOD5/M | Abstract | P00556 MNYCP 00 | Aerosols in Reactor Containment During Meltdown. |
| NEUPAC | Abstract | P00177 FM200 00 | Neutron Unfolding Code System for Calculating Neutron Flux Spectra from Activation Data of Dosimeter Foils. |
| NEVEMOR | Abstract | P00026 I3675 00 | Multigroup-Multiregion Calculation of Flux Spectra and Energy Deposition for Fast Neutrons. |
| NJOY91.119 | Abstract | P00171 MFMWS 04 | Code System for Producing Pointwise and Multigroup Neutron and Photon Cross Sections from ENDF/B Data. |
| NJOY94.61 | Abstract | P00355 MFMWS 03 | Code System for Producing Pointwise and Multigroup Neutron and Photon Cross Sections from ENDF/B Data. |
| NJOY97.0 | Abstract | P00368 MNYCP 00 | Code System for Producing Pointwise and Multigroup Neutron and Photon Cross Sections from ENDF/B Data. |
| NJOY99.0 | Abstract | P00480 MNYCP 00 | Code System for Producing Pointwise and Multigroup Neutron and Photon Cross Sections from ENDF/B Data. |
| NJOY-UTIL-EIR | Abstract | P00296 C0825 00 | Utilities For the NJOY (6/83) Nuclear Data Processing System. |
| NONSAP-C | Abstract | P00458 C7600 00 | Code System for Analysis of 3-D Reinforced Concrete Structures. |
| NORMA | Abstract | P00471 PC586 00 | Code System to Solve Burnup Dependent Neutron Diffusion Equations in Two and Three Dimensions. |
| NORMA-FP | Abstract | P00470 PC586 00 | Code System to Perform Neutronic and Thermal-Hydraulic Subchannel Analysis from Converged Coarse-Mesh Nodal Solutions. |
| NPTXS | Abstract | P00090 I0360 00 | Data Generator: Neutron Point Cross Sections from ENDF/B Resolved and Unresolved Resonance Parameters. |
| NRCPAGE | Abstract | P00491 DVX11 00 | Code System to Detect Recurring Loss of Special Nuclear Materials. |
| NRCPIPES 2.0A | Abstract | P00429 IBMPC 00 | Code System for Fracture Mechanics Analysis of Circumferential Surface Cracks in Pipes. |
| NSLINK | Abstract | P00314 D0VAX 00 | NJOY SCALE LINK. |
| NUCHART | Abstract | P00545 IBMPC 00 | Nuclear Properties and Decay Data Chart of Nuclides. |
| NUCWIZ | Abstract | P00616 PCX86 00 | NucWiz |
| NUFACE | Abstract | P00284 CYXMP 00 | An Interface Code For The Calculation of Nuclear Responses. |
| NX1-NX2 | Abstract | P00310 D0VAX 00 | Code System to Calculate Excitation Functions for (n,charged particle) Reactions. |
| O5S | Abstract | P00014 DP010 00 | Response Function Generator--An O5R Monte Carlo Code for Calculating Pulse Height Distributions Due to Monoenergetic Neutrons Incident on Organic Scintillators. |
| O5S | Abstract | P00014 I3675 00 | Response Function Generator--An O5R Monte Carlo Code for Calculating Pulse Height Distributions Due to Monoenergetic Neutrons Incident on Organic Scintillators. |
| OCA-P | Abstract | P00392 I3033 00 | Pressure Vessel Fracture-Mechanics Code System. |
| OCA-P | Abstract | P00392 IBMPC 00 | Pressure Vessel Fracture-Mechanics Code System. |
| OCTAVIA | Abstract | P00460 I0370 00 | Code System to Calculate Pressure Vessel Failure Probabilities. |
| OMCOST | Abstract | P00381 I3033 00 | Code System for Non-fuel O & M Cost Estimation for Large Steam-Electric Power Plants. |
| OPERATIONAL MONTE CARLO GUI | Abstract | P00619 PCX86 01 | Operational Monte Carlo GUI (OMG) |
| ORCENT-2 | Abstract | P00474 I3033 00 | Code System for Analysis of Steam Turbine Cycles Supplied by Light Water Reactors. |
ORINC
USSO | Abstract | P00439 I0360 00 | Code System for 1-D Implicit Heat Conduction Solution. |
ORMDIN
USSO | Abstract | P00399 I3033 00 | 2-D Nonlinear Inverse Heat Conduction. |
| ORMGEN3D | Abstract | P00430 CY0MP 00 | Mesh Generator for 3-D Crack Geometries. |
| ORMONTE | Abstract | P00275 IBMPC 00 | Uncertainty Analysis Code System for Use with User-Developed Systems Models. |
| ORPLOT-PC | Abstract | P00328 PC386 00 | Plotting Package for Data Evaluation Intercomparison. |
ORSMAC
USSO | Abstract | P00437 I3033 00 | Code System to Calculate Fluid Circulation Patterns Near Jets. |
| ORTHIS-ORTHAT | Abstract | P00569 I0360 00 | ORTHIS: 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. |
| ORTURB | Abstract | P00418 I0360 00 | HTGR Steam Turbine Dynamic Behavior. |
| PAPER 1 | Abstract | P00097 C6600 00 | Monte Carlo Calculation of Solid Angle and Self-Absorption Factors for an Inclined Cylindrical Source Viewed by a Cylindrical Detector. |
| PAPIN | Abstract | P00156 I0370 00 | A Code System to Calculate Cross Section Probability Tables, Bondarenko and Transmission Self-Shielding Factors for Fertile Isotopes in the Unresolved Resonance Region. |
| PARET-ANL | Abstract | P00516 MNYCP 00 | Code System to Predict Consequences of Nondestructive Accidents in Research and Test Reactor Cores. |
| PARET-ANL(NESC) | Abstract | P00565 MNYCP 00 | Code System to Predict Consequences of Nondestructive Accidents in Research and Test Reactor Cores. |
| P-CARES | Abstract | P00538 PC586 00 | Probabilistic Computer Analysis for Rapid Evaluation of Structures. |
| PC-BATLE | Abstract | P00451 IBMPC 00 | Code System to Calculate Brief Adversary Threat Loss Estimate. |
| PCC/SRC | Abstract | P00456 D0VAX 00 | Code System to Calculate Correlation & Regression Coefficients. |
| PC-PRAISE | Abstract | P00391 IBMPC 00 | Code System for Analysis of Piping Reliability Including Seismic Events. |
| PEGAS | Abstract | P00336 IBMPC 00 | Pre-Equilibrium-Equilibrium Gamma-and-Spin Code System. |
| PELE-1C | Abstract | P00461 C7600 00 | Code System for Fluid-Structure Interaction Analysis. |
| PELINOMIC-3A | Abstract | P00596 I0370 00 | Power Plant Cost Optimization for Dispersed Load Centers. |
| PELINSCA | Abstract | P00168 I0360 00 | A Code System for Nuclear Elastic and Inelastic Scattering Calculations. |
| PEQAG-2 | Abstract | P00293 IPCAT 00 | A Pre-equilibrium Computer Code With Gamma Emission. |
PHAZE
USSO | Abstract | P00432 IBMPC 00 | Parametric Hazard Function Estimation. |
| PICES | Abstract | P00568 I3033 00 | Probabilistic Investigation of Capacity and Energy Shortages. |
| PICTURE | Abstract | P00238 IBMPC 00 | Combinatorial Geometry Printer Plotting. |
| PIXSE | Abstract | P00133 I0360 00 | A Generator of Multigroup and Multipoint Cross Sections for Thermal Reactor Calculations. |
| PLASMX | Abstract | P00106 C6600 00 | A Multigroup Ionization and Charge Exchange Cross-Section Code System for Neutral Hydrogen Transport in Plasmas. |
| PLOTENDF | Abstract | P00214 I3033 00 | A Program for Producing Graphical Output. |
| PLOTFB | Abstract | P00018 I3675 00 | ENDF/B Data Plotting Code. |
| PLOTNFIT | Abstract | P00382 IBMPC 00 | Code System for Data Plotting and Curve Fitting. |
| PLOT-S | Abstract | P00552 PC586 00 | Plotting Program with Special Features for Windows Environment. |
| PLOTTAB-89.1 | Abstract | P00274 ALLCP 00 | Plot Continuous Curves or Discrete Points. |
| POLLA | Abstract | P00208 I3033 00 | A Fortran Program to Convert R-MATRIX-Type Multilevel Resonance Parameters for Fissile Nuclei into Equivalent KAPUR-PEIERLS-Type Parameters. |
| POLYRES | Abstract | P00438 MNYCP 00 | Richards Equation Solver; Rectangular Finite Volume Flux Updating Solution. |
| POPOP4 | Abstract | P00011 I3675 00 | Converter of Gamma-Ray Spectra to Secondary Gamma-Ray Production Cross Sections. |
| POWER | Abstract | P00069 C7600 00 | Source Distribution Input Data Generator for ANISN Code. |
| PREANG | Abstract | P00166 C0175 00 | Calculation of Pre-equilibrium Angular Distributions with the Exciton Model. |
| PRE-ANISN | Abstract | P00332 PC386 00 | A Preprocessing Code for ANISN and Other Radiation Transport Codes. |
| PRECO2006 | Abstract | P00226 MNYCP 02 | Exciton Model Code System for Calculating Preequilibrium and Direct Double Differential Cross Sections. |
| PREDEX-1 | Abstract | P00597 I0370 00 | U, Pu, Nitric Acid Distribution in Counter Current Solvent Extraction. |
| PREM | Abstract | P00224 I0360 00 | Code System for Pre-equilibrium Process with Multiple Nucleon Emission. |
| PREPRO2019 | Abstract | P00351 MNYCP 10 | Pre-Processing Code System for Data in ENDF/B Format. |
| PSAPACK-4.2 | Abstract | P00613 PCX86 00 | Probabilistic Safety Analysis with Fault Event Trees. |
| PSDREC | Abstract | P00441 DP011 00 | Code System for Power Spectral Density Recognition Continuous On-line Reactor Surveillance. |
| PUFF-IV | Abstract | P00534 MNYCP 01 | Determination of Multigroup Covariance Matrices from ENDF/B-V Uncertainty Files. |
| Q&A | Abstract | P00428 IBMPC 00 | Questions and Answers Based on Revised 10 CFR Part 20 |
| QUARK | Abstract | P00492 PC586 00 | Code System for 2-Group, 3D Neutronic Kinetics Calculations Coupled to Core Thermal Hydraulics. |
| RADAK | Abstract | P00122 I0360 00 | Flux Spectra Unfolding Code System - Neutron or Gamma-Ray Detectors. |
| RADCOMPT 2.10L | Abstract | P00348 IBMPC 00 | Sample Analysis Code System for the Dual Channel Counter. |
| RCSLK9 | Abstract | P00452 IBMPC 00 | Code System to Calculate Reactor Coolant System Leak Rate. |
| RDMM | Abstract | P00598 I0360 00 | Flux Spectra from In-Pile Fast Neutron Activation Experiment. |
| REACTION | Abstract | P00347 AL000 00 | Code System to Calculate Integral Parameters with Reaction Rates from WIMS Output. |
| REACTION | Abstract | P00347 IBMPC 00 | Code System to Calculate Integral Parameters with Reaction Rates from WIMS Output. |
| RECAP | Abstract | P00414 IBMPC 00 | Replacement Energy Cost Analysis Package. |
| RECAP | Abstract | P00414 IBMPC 01 | Replacement Energy Cost Analysis Package. |
| REEX-1 | Abstract | P00599 I0370 00 | U, Pu, Nitric Acid Distribution in Counter Current Pluristage Stripping. |
| REFCO83 | Abstract | P00447 I3033 00 | Nuclear Fuel Cycle Cost Economics Code System. |
| REFERDOU | Abstract | P00249 FM380 00 | Code System for NE-213 Unfolding of Neutron Spectra up to 100 MeV with Response Function Error Propagation. |
| REFLUX | Abstract | P00403 I3033 00 | Code System to Predict LWR Reflood Heat Transfer. |
| REFUM-BROAD | Abstract | P00039 F2307 00 | Monte Carlo Codes for Calculating Efficiencies and Response Functions of NaI(Tl) Crystals for Thick Disk Gamma-Ray Sources. |
| REGN | Abstract | P00165 I0360 00 | Code System for Solving Nonlinear Systems of Equations via the Gauss-Newton Method. |
RELAP5/MOD1/029_EXE
810 | Abstract | P00423 C0176 01 | Reactor System Transient Code. |
| REMIT 5.1 | Abstract | P00482 IBMPC 01 | Radiation Exposure Monitoring and Information Transmittal System. |
| REPC | Abstract | P00195 C0000 00 | Estimation of Nuclear Reaction Effects in Proton-Tissue-Dose Calculations. |
| RESENDD | Abstract | P00215 C0740 00 | A Code System for Reconstruction of Resonance Cross Sections from Evaluated Nuclear Data in ENDF/B Format. |
| RESENDD | Abstract | P00215 D0780 00 | A Code System for Reconstruction of Resonance Cross Sections from Evaluated Nuclear Data in ENDF/B Format. |
| RESPMG | Abstract | P00060 I0360 00 | Response Matrix Generation Code System. |
| REX2-87 | Abstract | P00290 D8810 00 | A Code For Calculating Self-Shielded Multigroup Neutron Cross Sections and Self-Shielding Factors From Preprocessed ENDF/B Basic Data Files. |
| RFSP-JUL | Abstract | P00126 I0360 00 | Unfolding Code System for Neutron Spectra Evaluation from Activation Data. |
| RFUNC | Abstract | P00312 D0VAX 00 | Code System to Analyze Differential Scattering Data. |
| RGENDF | Abstract | P00239 C0170 00 | Format Translation from NJOY GENDF Format to ENDF/B-V and Other Formats. |
| RICE | Abstract | P00022 I0360 00 | A Program to Calculate Primary Recoil Atom Spectra from ENDF/B Data. |
| RIPPLE | Abstract | P00571 CYXMP 00 | A Computer Program for Incompressible Fluid Dynamics with Free Surfaces. |
| RNGP | Abstract | P00066 I3675 00 | Random Number Generator Package. |
| ROLAIDS-CPM | Abstract | P00353 SUN04 00 | Code System to Calculate Group-Averaged Cross Sections Using the Collision Probability Method. |
| S1CALC | Abstract | P00134 I0360 00 | A Multigroup Thermal Neutron Scattering Law Data Generator for Hydrogen and Deuterium. |
| SAEROSA | Abstract | P00573 MNYCP 00 | Single-Species Aerosol Coagulation and Deposition with Arbitrary Size Resolution. |
| SAFE-D/SAFE-R | Abstract | P00496 MNYCP 00 | Code System for the Analysis of Component Failure Data with a Compound Statistical Model. |
| SAIPS | Abstract | P00203 E1040 00 | Information Processing System for Calculating Neutron Spectra from Measured Reaction Rates. |
| SAIPS-PC | Abstract | P00295 IBMPC 00 | Information Processing System for Calculating Neutron Spectra from Measured Reaction Rates. |
| SALE3D | Abstract | P00443 CY000 00 | ICEd-ALE Treatment of 3-D Fluid Flow. |
| SAMCR | Abstract | P00487 U1100 00 | Code System for 2-D Elastodynamic Fracture Analysis. |
| SAMMY 8.1.0 | Abstract | P00158 MNYCP 13 | Code System for Multilevel R-Matrix Fits to Neutron and Charged-Particle Cross-Section Data Using Bayes' Equations. |
| SAMPO80 | Abstract | P00204 DGNOV 00 | Gamma-Ray Spectrum Analysis Method for Minicomputers. |
| SAMPO-LRC | Abstract | P00186 C6600 00 | Gamma-Ray Spectrum Analysis Code. |
| SAND-II-SNL | Abstract | P00345 SUN04 00 | Neutron Flux Spectra Determination by Multiple Foil Activation Method. |
| SAPHIRE 8.0.9 | Abstract | P00608 PCX86 00 | Systems Analysis Programs for Hands-On Integrated Reliability Evaluations. |
| SARA 4.16 | Abstract | P00484 IBMPC 00 | System Analysis and Risk Assessment System. |
| SATURN | Abstract | P00057 I3675 00 | P1 or Transport Corrected Multigroup Neutron Cross Section Data Processor. |
| SC2N3N | Abstract | P00309 D0VAX 00 | Systematics of (n,2n) and (n,3n) Cross Sections. |
| SCAMPI | Abstract | P00352 MNYWS 01 | Collection of Codes for Manipulating Multigroup Cross Section Libraries in AMPX Format. |
| SCANS | Abstract | P00029 I3675 00 | Spectra Calculation from Activated Nuclide Sets. |
| SCANS 1A | Abstract | P00373 PC386 01 | Shipping Cask Design Review Analysis. |
| SCAT-2 | Abstract | P00294 MNYCP 03 | Code System for Calculating Total and Elastic Scattering Cross Sections Based on an Optical Model of the Spherical Nucleus. |
SCDAP/RELAP5/MOD3.3-EXE
810 | Abstract | P00581 MNYCP 01 | A Best-Estimate Transient Simulation of Light Water Reactor Coolant Systems During a Severe Accident. |
| SCINFUL | Abstract | P00267 CY0MP 00 | Scintillator Full Response to Neutron Detection. |
| SCINFUL | Abstract | P00267 D8600 00 | Scintillator Full Response to Neutron Detection. |
| SCOPE | Abstract | P00210 I3033 00 | Computer Code System for Shipping Cask Optimization and Parametric Evaluation. |
| SCORCH-B2 | Abstract | P00601 I0370 00 | BWR Core Heating During LOCA. |
| SCORE-EVET | Abstract | P00442 C7600 00 | Code System for Three-Dimensional Hydraulic Reactor Core Analysis. |
| SCRELA | Abstract | P00408 SUN05 00 | Code System for Supercritical Water Cooled Reactor LOCA Analysis. |
| SECA | Abstract | P00104 I0360 00 | Evaluator of Angular Bounds for a Two-Dimensional Symmetric Gaussian Quadrature Set. |
| SEISIM1 | Abstract | P00453 C7600 00 | Code System for Seismic Probabilistic Risk Assessment. |
| SELFS-3 | Abstract | P00551 C6600 00 | Self-Shielding Correlation of Foil Activation Neutron Spectra Analysis by SAND-II. |
| SETS | Abstract | P00380 CDCMF 00 | Set Equation Transformation System. |
SFHA
USSO | Abstract | P00413 IBMPC 00 | Code System for Spent Fuel Heating Analysis. |
SHC
USSO | Abstract | P00493 CY000 00 | Seismic/Hazard Characterization in the Eastern U.S. |
| SIGPI | Abstract | P00475 D0785 00 | Fault Tree Cut Set System Performance. |
| SINBAD SEARCH TOOL | Abstract | P00580 MNYCP 00 | SINBAD Search Tool |
| SIOB | Abstract | P00139 I0360 00 | Calculation of Least-Squares Shape Fitting Several Neutron Transmission Measurements Using the Breit-Wigner Multilevel Formula. |
| SIR-3 | Abstract | P00055 C6400 00 | Sievert's Integral Routine-Computer Evaluation. |
| SIR-3 | Abstract | P00055 I3675 00 | Sievert's Integral Routine-Computer Evaluation. |
| SKEWGAUS | Abstract | P00089 I0360 00 | Skewed-Gaussian Line Peak Fitting Code - Multichannel Analyzer (MCA) Spectra - Ge(Li) and Semiconductor Detectors. |
| SLAROM | Abstract | P00244 FM380 00 | A Code to Produce Cell Averaged Cross Sections for Fast Critical Assemblies and Fast Power Reactors. |
| SMACS | Abstract | P00396 C7600 01 | Probabilistic Seismic Analysis Code System. |
| SMAFS | Abstract | P00547 PC586 00 | Steady-State Analysis Model for Advanced Fuel Cycle Schemes. |
| SMOG | Abstract | P00216 I3033 00 | Code System for Neutron Cross Section Evaluation (Optical Method). |
| SNAKE | Abstract | P00135 I0360 00 | A Solid Angle Calculational System. |
| SOFIRE-2 | Abstract | P00570 I0370 00 | Containment Temperature and Pressure During Na Pool Fire, 1-Cell or 2 Cell. |
| SOLA-DF | Abstract | P00454 C7600 00 | Code System to Calculate Transient 2-Dimensional 2-Phase Flow. |
| SOLA-LOOP | Abstract | P00464 C7600 00 | Nonequilibrium, Drift-Flux Code System for Two-Phase Flow Network Analysis |
| SORA | Abstract | P00174 I0360 00 | A Code System for Storage and Retrieval of Data from Radionuclide Analyses. |
| SPEC-4 | Abstract | P00099 I0360 00 | Calculated Recoil Proton Energy Distributions from Monoenergetic and Continuous Spectrum Neutrons. |
| SPECTER | Abstract | P00023 I3565 00 | Calculation of Energy Distribution of Nuclear Reaction Products. |
| SPECTER-ANL | Abstract | P00263 D0VAX 00 | Neutron Damage Calculations for Materials Irradiations. |
| SPECTRANS-2 | Abstract | P00071 ICL00 00 | Neutron Spectrum Library Generation. |
| SPES | Abstract | P00602 I0370 00 | Fuel Cycle Optimization for LWR. |
| SPHINX | Abstract | P00129 C7600 00 | A One-Dimensional Diffusion and Transport Nuclear Cross Section Processing Code System. |
| SPHINX | Abstract | P00129 I0360 00 | A One-Dimensional Diffusion and Transport Nuclear Cross Section Processing Code System. |
SPIRT
USSO | Abstract | P00476 C7600 00 | Code System to Calculate Stress-Strains from Transient Pressures. |
SPIRT-NRC
USSO | Abstract | P00198 I3033 01 | Computerized Mathematical Models of Spray Washout of Airborne Contaminants (Radioactivity) in Containment Vessels. |
| SPUNIT | Abstract | P00266 D8600 00 | Spectrum Unfolding Using Information Theory. |
SQUIRT VER2
USSO | Abstract | P00583 PCX86 00 | Code System to Predict Leakage Rate and Area of Crack Opening for Cracked Pipes in Nuclear Power Plants. |
SRVAL
USSO | Abstract | P00467 I3033 00 | Stock-Recruitment Model Validation Code System. |
SSC-L V3.3
USSO | Abstract | P00400 I3090 00 | Transient Response in LMFBR System. |
| STABA,STAGT,STEGT,STIG,STIGMA | Abstract | P00575 MNYCP 00 | Stress Analysis of Dragon HTR Graphite Structure. |
| STAPREF | Abstract | P00498 PC586 00 | Code System to Calculate Nuclear Reaction Cross Sections by Evaporation Model. |
| STAPRE-H95 | Abstract | P00325 MNYCP 01 | Code System to Calculate Energy-Averaged Cross Sections of Particle Induced Nuclear Reactions. |
| STAR CODES | Abstract | P00330 IBMPC 00 | Code System for Calculating Stopping-Power and Range Tables for Electrons, Protons, and Helium Ions. |
| STAY'SL | Abstract | P00113 DP010 00 | Least Squares Dosimetry Unfolding Code System. |
| STAYSL PNNL | Abstract | P00589 PCX86 00 | STAYSL PNNL Suite of Software Tools. |
| STRADE | Abstract | P00252 I3081 00 | Stratified Random Design. |
| SUGGEL | Abstract | P00508 MNYWS 00 | Program Suggesting the Orbital Angular Momentum of a Neutron Resonance From the Magnitude Of Its Neutron Width. |
| SUPERDAN-PC | Abstract | P00282 IBMPC 00 | Calculates Dancoff Factor of Spheres, Cylinders and Slabs. |
| SUPERTOG III M2 | Abstract | P00013 I3691 00 | Data Generator--Fine Group Constants and PN Scattering Matrices from ENDF/B. |
| SUPERTOG-4 | Abstract | P00013 I0360 00 | Data Generator--Fine Group Constants and PN Scattering Matrices from ENDF/B. |
| SUPERTOG-JR. | Abstract | P00115 F2307 00 | Data Generator--Fine Group Constants and PN Scattering Matrices from ENDF/B. |
| SUPERTOG-JR. | Abstract | P00115 I0360 00 | Data Generator--Fine Group Constants and PN Scattering Matrices from ENDF/B. |
| SUPERTOG-LTT | Abstract | P00228 I0360 00 | Data Generator--Fine Group Constants and PN Scattering Matrices from ENDF/B. |
| SWIFT | Abstract | P00031 C6600 00 | Monte Carlo Neutron Spectra Unfolding Code. |
| TALYS-1.2 | Abstract | P00548 PC586 01 | Nuclear Model Code System for Analysis and Prediction of Nuclear Reactions and Generation of Nuclear Data. |
| TAM3 | Abstract | P00308 IBMPC 00 | Demonstrates Monte Carlo Sensitivity and Uncertainty Analysis. |
| TDOWN-IV | Abstract | P00172 H6000 00 | A Code System to Generate Composition- and Spatially-Dependent Neutron Cross Sections for Multigroup Neutronics Analysis. |
| TECALC | Abstract | P00074 DP010 00 | Interactive Calculation of Compton Coherent and Photoelectric Mass Attenuation Coefficients for Photons (E<1 MeV), and the Mass Absorption Coefficient for Known Materials. |
| TEMAC | Abstract | P00468 D0VAX 00 | Top Event Matrix Analysis Code System. |
| TEMPEST-2 | Abstract | P00558 I0360 00 | Thermalization Program for Neutron Spectra and MultiGroup Cross-Sections. |
| TEMPEST-BNW | Abstract | P00559 C7600 00 | Transient 3-D Thermohydraulics for FBR. |
| THACT-RR | Abstract | P00587 D0VAX 00 | Analysis of Thermal Hydraulics Transients in Research Reactor Core. |
| THERMOS-OTA | Abstract | P00107 C0173 00 | Multigroup Integral Transport Code System for Thermal Lattice Calculations using Collision Probability Method for Slabs and Cylinders. |
| THERMOS-OTA | Abstract | P00107 C0740 00 | Multigroup Integral Transport Code System for Thermal Lattice Calculations using Collision Probability Method for Slabs and Cylinders. |
| THERMOS-OTA | Abstract | P00107 U1108 00 | Multigroup Integral Transport Code System for Thermal Lattice Calculations using Collision Probability Method for Slabs and Cylinders. |
| THRUSH | Abstract | P00276 CYXMP 00 | Calculates Thermal Neutron Scattering Kernel. |
| THYDE-B1/MOD2 | Abstract | P00553 FM200 00 | Computer Code for PWR LOCA Thermohydraulic Transient Analysis. |
| THYDE-P2 | Abstract | P00554 FV100 00 | Computer Code for PWR LOCA Thermohydraulic Transient Analysis. |
| TIMS-1 | Abstract | P00163 D0780 00 | Processing Code System for Production of Group Constants of Heavy Resonant Nuclei. |
| TIMS-1 | Abstract | P00163 FM200 00 | Processing Code System for Production of Group Constants of Heavy Resonant Nuclei. |
| TNG1 | Abstract | P00298 D6220 00 | A Multistep Statistical Model Based on the Hauser-Feshbach Theory For The Evaluation Of Neutron Data. |
| TORAC | Abstract | P00459 C0170 00 | Code System to Calculate Tornado-Induced Flow Material Transport. |
| TOTEM-3 | Abstract | P00603 I0370 00 | Demand Assessment for Nuclear Power Plants and Conventional Power Plants. |
| TPASS | Abstract | P00164 DP010 00 | A Gamma-Ray Spectral Data-Reduction and Analysis Code System. |
| TRANSX 2.15 | Abstract | P00317 MFMWS 01 | Code system to produce neutron, photon, and particle transport tables for discrete-ordinates and diffusion codes from cross sections in MATXS format. |
| TRANSX-CTR | Abstract | P00206 CY000 00 | Interfaces MATXS Cross-Section Libraries to Nuclear Transport Codes for Fusion Systems Analysis. |
| TRAX | Abstract | P00280 C0720 00 | A Program For Optics of Curved Crystal Neutron Spectrometers. |
| TRIGLAV | Abstract | P00495 PC586 00 | Code System to Calculate Mixed Cores in TRIGA Mark II Research Reactor. |
| TRISTAN-IJS | Abstract | P00537 IBMPC 00 | Multigroup Three-Dimensional Direct Integration Method Radiation Transport Analysis Code System. |
| TRUMP | Abstract | P00522 MNYCP 01 | Code System for Transient and Steady-State Temperature Distribution in Multidimensional Systems. |
| TSORT | Abstract | P00486 IBMPC 00 | Automated Technique for Nuclear Plant Training Task Assignment. |
| TURBINA | Abstract | P00604 I0370 00 | Reheat Steam Turbine Generator Design with Preheater and Condenser. |
| UHS | Abstract | P00390 IPS70 00 | Ultimate Heat Sink Cooling Pond and Spray Pond Analysis Models. |
| UKE-III | Abstract | P00015 I3691 00 | Cross Section Format Translator - UKNDL to ENDF/B. |
| UMG 3.3 | Abstract | P00529 PC586 00 | Unfolding with Maxed and Gravel. |
| UNF | Abstract | P00521 PC586 00 | Code System to Calculate Multistep Compound Nucleus Neutron Cross-Sections and Spectra for Structural Materials. |
| UNIFY-ECN | Abstract | P00288 C0170 00 | A Program to Calculate Fast Neutron Data for Structural Materials. |
| UPDATE | Abstract | P00270 DGMV1 00 | Program to Update Fortran Source Files. |
| UPDATE | Abstract | P00270 I3081 00 | Program to Update Fortran Source Files. |
| UPEAK | Abstract | P00300 IPCXT 00 | A Program for Decomposing A One-Dimensional Spectrum. |
| UPEML 3.0 | Abstract | P00245 ALLCP 01 | A Machine-Portable CDC UPDATE Emulator. |
| URR | Abstract | P00281 D6220 00 | Calculates Resonance Neutron Cross-Section Probability Tables, Bondarenko Self-Shielding Factors and Self-Indication Ratios for Fissile and Fertile Nuclides. |
| USINT | Abstract | P00415 MNYCP 00 | Code System to Calculate Heat and Mass Transfer In Concrete |
| UTSG | Abstract | P00379 I3033 00 | Code System for Calculating the Nonlinear Transient Behavior of a Natural Circulation U-Tube Steam Generator with Its Main Steam System. |
| VIDEO-PC | Abstract | P00311 IBMPC 00 | Super VGA Primitives Graphics System. |
| VIEWCXS | Abstract | P00514 PC586 00 | Interactive Graphic User Interface to View Neutron and Gamma-Ray Interaction Cross Sections. |
| VISA2 | Abstract | P00445 MNYCP 00 | Code System to Calculate Probability of Reactor Vessel Failure. |
| VISUAL EDITOR 61 | Abstract | P00618 PCX86 00 | MCNPX/6 Visual Editor Computer Code 61 |
| VIXEN | Abstract | P00030 C6600 00 | A Code to Check Physical Consistency of Photon-Production Data in Revised ENDF Format. |
| VIXEN | Abstract | P00030 I0360 00 | A Code to Check Physical Consistency of Photon-Production Data in Revised ENDF Format. |
| WAKE | Abstract | P00605 I0370 00 | Navier Stokes Equation with 2-D Turbulence, Stream Function, Vorticity. |
| WILIT | Abstract | P00344 MNYCP 00 | A Utility Program for WIMS Libraries. |
| WIMSCORE-ENEA | Abstract | P00319 I3090 00 | Code System to Process WIMSD4 Interface Output Files and Generate Two-Group Data for Reactor Calculations. |
| WINDOWS | Abstract | P00136 I0360 00 | A Program for the Analysis of Spectral Data Foil Activation Measurements. |
| WINDOWS II | Abstract | P00161 I0370 00 | A Program for the Analysis of Spectral Data Foil Activation Measurements. |
| WREM-TOODEE2 | Abstract | P00469 ALLMF 00 | 2-D Time-Dependent Fuel Element, Thermal Analysis Code System. |
| X4ECS | Abstract | P00220 D0780 00 | A Code System to Combine Cross Section Data in EXFOR and/or ENDF/B-IV Format. |
| X4R | Abstract | P00222 DVX11 00 | Code System for Retrieving EXFOR Cross Section Data According to a Given Target Nucleus. |
| XLACS-IIA | Abstract | P00182 I3033 00 | A Modified Version of XLACS-II for Processing ENDF Data into Multigroup Neutron Cross Sections in AMPX Master Library Format. |
| ZOTT99 | Abstract | P00272 ALLCP 02 | Zero-in On The Truth; Evaluation of Correlated Data Using Partitioned Least Squares. |