| Alphabetical Listing |
| Package Name | Abstract | RSICC Tapelist | Title |
| 1DB-2DB-3DB | Abstract | C00741 PC586 00 | One-Dimensional Diffusion Code System for Nuclear Reactor. |
| 1DX | Abstract | P00096 U1108 00 | A One-Dimensional Diffusion Code System for Producing Energy Group Collapsed and Self-Shielded Cross Sections. |
| 3DDT | Abstract | C00605 C6600 00 | Multigroup Diffusion Code System for Use in Fast Reactor Analysis. |
| AARE-V1.0 | Abstract | C00846 MNYCP 00 | Activation in Accelerator Radiation Environments |
| ABAREX | Abstract | P00248 MNYCP 01 | Neutron Spherical Optical-Statistical Model Code System. |
| ABBN-90 | Abstract | D00182 MNYCP 00 | Multigroup Constant Set for Calculation of Neutron and Photon Radiation Fields and Functionals, Including the CONSYST2 Program. |
| ABLEIT-TRANS | Abstract | P00247 C0175 00 | Error Propagation Analysis for Burnup Calculation. |
| ACAB-2008 | Abstract | C00758 MNYCP 01 | Activation Abacus Inventory Code System for Nuclear Applications. |
| ACAT | Abstract | P00257 FM380 00 | Monte Carlo Simulation of Atomic Collisions in Amorphous Targets in the Binary Collision Approximation. |
| ACDOS3 | Abstract | C00442 C7600 00 | Calculation of Activities and Dose Rates Produced by Neutron Activation. |
| ACFA | Abstract | C00478 I3033 00 | A Versatile Activation Code for Coolant and Structural Materials. |
| ACOH | Abstract | C00191 I3675 00 | Aerojet COHORT Monte Carlo Code System. |
| ACORNS | Abstract | P00264 IBMPC 01 | Analysis of Correlations Used in Neutron Spectrometry. |
| ACRA-II | Abstract | C00213 I0360 00 | Kernel Integration Code System for Estimation of Radiation Doses Caused by a Hypothetical Reactor Accident. |
| ACRA-TRIT | Abstract | C00283 I0360 00 | The Tritium Version of ACRA-II, Estimation of Radiation Doses Caused by a Hypothetical Reactor Accident. |
| ACT-ARA | Abstract | C00372 CYXMP 00 | Code System for the Calculation of Changes in Radiological Source Terms with Time. |
| ACTIV | Abstract | P00590 I0370 00 | Sandwich Detector Activity from In-Pile Slowing-Down Spectra Experiment. |
| ACTIV87 | Abstract | D00169 ALLCP 00 | Fast Neutron Activation Cross Section File. |
| ACTIV-PC | Abstract | P00287 IBMPC 00 | A Program to Process Gamma or X-ray Spectra. |
| ACTL82 | Abstract | D00069 ALLCP 01 | Evaluated Neutron Activation Cross-Section Library. |
| ACTV-F/H | Abstract | D00155 ALLCP 00 | Neutron Activation Cross Section Library for Fusion Reactor Design. |
| ACTV-FUS/INT | Abstract | D00170 ALLCP 00 | International Library of Neutron Activation Cross-Section Data for Fusion Reactor Application. |
| 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. |
| ADJMOM | Abstract | C00212 I3675 00 | Adjoint Moments Method Gamma-Ray Transport Code System. |
| ADLER III | Abstract | P00058 I0360 00 | A Program to Calculate Cross Sections from Adler-Adler Resonance Parameters. |
| ADO | Abstract | C00189 I3675 00 | Aerojet Discrete Ordinates Calculational System. |
| ADS-LIB/V2.0 | Abstract | D00250 MNYCP 00 | Test Library for Accelerator Driven Systems V2.0 |
ADVANTG 3.2.0
810 | Abstract | C00854 PCX86 00 | AutomateD VAriaNce reducTion Generator |
| ADVANTG 3.2.1 | Abstract | C00854 PCX86 01 | AutomateD VAriaNce reducTion Generator |
| AGDATA | Abstract | D00127 I0360 00 | Two Agricultural Production Data Libraries (AGDATC and AGDATG) for Dose and Risk Assessment Models. |
| AIR DATA | Abstract | D00014 I0360 00 | Sample Input to ANISN for Calculation of Neutron and Secondary Gamma-Ray Transport in Air. |
| AIRBORNE | Abstract | C00263 I0360 00 | Airborne Contaminants Dispersion Code. |
| AIRDIF | Abstract | C00360 C6600 00 | A Two-Dimensional Atmospheric Radiation Diffusion Code. |
| AIRDOS-PC | Abstract | C00551 IBMPC 00 | Clean Air Act Compliance Software for Personal Computers. |
| AIREKMOD-RR | Abstract | P00588 D0VAX 00 | Reactivity Transients in Nuclear Research Reactors |
| AIREKMOD-RR | Abstract | P00588 PCX86 01 | Reactivity Transients in Nuclear Research Reactors |
| AIREM | Abstract | C00242 I3691 00 | Calculation of Doses, Population Doses, and Ground Depositions Due to Atmospheric Emissions of Radionuclides. |
| AIRFEWG | Abstract | D00049 I0360 00 | Results of ANISN Multigroup Calculations of Gamma-Ray, Neutron, and Secondary Gamma-Ray Transport in Infinite Homogeneous Air Using DLC-31/(DPL-1/FEWG1) Cross Sections. |
| AIRGAMMA | Abstract | C00567 FM380 00 | A Program For The Calculation Of External Exposure To Gamma Rays From A Radioactive Cloud. |
| AIRSCAT | Abstract | C00341 DP010 00 | Calculation of Dose Rate for Gamma-Rays Scattered in Air. |
| AIRTRANS | Abstract | C00110 I3675 00 | Monte Carlo Time and Energy-Dependent Three-Dimensional Radiation Transport Code. |
| AISITE II | Abstract | C00286 I0360 00 | Reactor Siting Code System. |
| AKERN | Abstract | C00190 C0000 00 | Aerojet Point Kernel Integration Calculational System. |
| AKERN | Abstract | C00190 U1108 00 | Aerojet Point Kernel Integration Calculational System. |
| AKTIV | Abstract | C00339 I0360 00 | An Evaluation of Activity, Afterheat and Biological Hazard Potential of Stainless Steel Structures in Fusion Reactor Blankets. |
| ALARA 2.7.8 | Abstract | C00723 MNYCP 00 | Code System for Analytic and Laplacian Adaptive Radioactivity Analysis. |
| ALARM-B2 | Abstract | P00218 I0360 00 | A Computer Code System for Analysis of a Large Break LOCA of a BWR. |
| ALBEDO/ALBEZ | Abstract | C00555 IBMPC 00 | Calculates Attenuation of Radiation in Single and Double Bends. |
| ALBEDO-DATA | Abstract | D00224 MNYCP 00 | KSU Neutron Albedo Data. |
| ALBEMO | Abstract | C00268 C6600 00 | Albedo Monte Carlo Code System. |
| ALDOSE | Abstract | C00577 IBMPC 00 | Dose Calculation for Alpha Disc Source. |
| ALEPH-LIB-JEFF3.1 | Abstract | D00230 MNYCP 00 | ACE Format Neutron Cross Section Library based on JEFF3.1. |
| ALGAM-97 | Abstract | C00152 I3675 00 | Monte Carlo Estimation of Internal Dose from Gamma-Ray Sources in a Phantom Man. |
| ALICE2017 | Abstract | P00550 PCX86 06 | Statistical Model Code System to Calculate Particle Spectra from HMS Precompound Nucleus Decay. |
| ALKASYS-PC | Abstract | C00558 IBMPC 00 | A Computer Program For Studies of Rankine-Cycle Space Nuclear Power Systems. |
| ALPHA-M | Abstract | P00169 I0360 00 | Least-Squares Resolution of Gamma-Ray Spectra in Environmental Samples. |
| ALPHN | Abstract | C00612 IBMPC 00 | Code System for Calculating (alpha,n) Neutron Production in Canisters of High-Level Waste. |
| AMARA | Abstract | P00079 I3675 00 | Nuclear Data Adjustment Using Lagrange's Multipliers Method. |
| AMC | Abstract | C00090 I3675 00 | Monte Carlo Albedo Code for Neutron and Capture Gamma-Ray Distributions in Rectangular Concrete Ducts. |
| AMP | Abstract | C00793 PCX86 00 | Advanced Multi-Physics. |
| AMPX01 | Abstract | D00027 I3675 02 | Modular Code System for Generating Coupled Multigroup Neutron-Gamma Libraries from ENDF/B. |
| 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. |
| ANISN-ORNL | Abstract | C00254 MNYCP 02 | One-Dimensional Discrete Ordinates Transport Code System with Anisotropic Scattering. |
| ANISN-PC | Abstract | C00514 IBMPC 00 | Multigroup One-Dimensional Discrete Ordinates Transport Code System with Anisotropic Scattering. |
| ANITA-2000 | Abstract | C00693 MNYCP 00 | Analysis of Neutron Induced Transmutation and Activation. |
| ANITA-4 | Abstract | C00606 MNYCP 01 | Analysis of Neutron Induced Transmutation and Activation. |
| ANL-BPB | Abstract | M00004 MNYCP 00 | Argonne National Laboratory Code Center: Benchmark Problem Book. |
| ANS643 | Abstract | D00129 IBMPC 02 | Geometric Progression Gamma-Ray Buildup Factor Coefficients. |
| ANSIFT | Abstract | P00077 C6600 00 | ANSI Standard Fortran Sifting Program. |
| ANSIFT | Abstract | P00077 I0360 00 | ANSI Standard Fortran Sifting Program. |
| ANSL-V | Abstract | D00154 ALLCP 01 | ENDF/B-V Based Multigroup Cross Section Libraries for Advanced Neutron Source (ANS) Reactor Studies. |
| ANTE 2 | Abstract | C00131 I3675 00 | Adjoint Monte Carlo Time-Dependent Neutron Transport Code in Combinatorial Geometry. |
| APARNA-II | Abstract | C00296 I0360 00 | Integral Transport Theory Code System Based on Discrete Ordinate Representation in Space and Direction-Slab Geometry. |
| 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. |
| APUD 3.0 | Abstract | C00637 IBMPC 00 | Code System for Analyzing, Predicting Consequences of, and Guiding the Response to Nuclear Emergencies. |
| ARC | Abstract | C00224 C6600 00 | Aircraft Radiation Transport Code System, Crew Dose Calculation. |
ARC 11.2892
FEDC | Abstract | C00824 MNYCP 02 | Code System for Analysis of Nuclear Reactors. |
| ARCON96 | Abstract | C00664 IBMPC 00 | Code System to Calculate Atmospheric Relative Concentrations in Building Wakes. |
| AREAC | Abstract | C00438 I3033 00 | Radiological Emission Analysis Code System. |
| AREAD | Abstract | P00088 I0360 00 | Input Data Processor for Transport Codes. |
| ARMYL-G | Abstract | C00297 U1106 00 | Calculation of Transmission Factors for Gamma Rays from Nuclear Explosions. |
| ARMYL-N | Abstract | C00298 U1106 00 | Calculation of Transmission Factors for Neutrons from Nuclear Explosions. |
| ARRRG | Abstract | C00404 U1100 00 | Calculation of Radiation Dose to Man from Radionuclides in the Environment. |
| ART MOD2 | Abstract | P00611 PCX86 00 | Fission Product Migration in Primary System and Containment |
| ASFIT-VARI | Abstract | C00336 H0000 00 | Gamma-Ray Transport Code System for One-Dimensional Finite Systems. |
| ASFIT-VARI | Abstract | C00336 IBMPC 00 | Gamma-Ray Transport Code System for One-Dimensional Finite Systems. |
| ASOP | Abstract | C00126 IRISC 00 | Multigroup One-Dimensional Discrete Ordinates Transport Code System for Shield Optimization. |
| ASTROS | Abstract | C00073 I7090 00 | Calculation of Primary and Secondary Proton Dose Rates in Spheres and Slabs of Tissue. |
| AT123D | Abstract | C00417 I0360 00 | Analytical Transient One-, Two-, and Three-Dimensional Simulation of Waste Transport in an Aquifer System. |
| ATHENA_2D | Abstract | P00431 MNYCP 00 | Code System For Simulation Of Hypothetical Recriticality Accidents in a Thermal Neutron Spectrum. |
| ATM-TOX | Abstract | C00472 I3033 00 | An Atmospheric Transport Model for Toxic Substances. |
| ATTOW-KB | Abstract | C00132 I0370 00 | Multigroup Two-Dimensional Removal-Diffusion (Spinney Method) Shielding Code System. |
| AUS98 | Abstract | C00519 MNYWS 01 | Modular System for Neutronics Calculations of Fission Reactors, Fusion Blankets, and Other Systems. |
| 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. |
| BABEL | Abstract | D00104 I3033 00 | Multi-Purpose Neutron and Gamma-Ray Cross Section Library for Fast Reactor Shielding Design. |
| BALTORO | Abstract | C00479 C6600 00 | Code for Coupling of Monte Carlo and Discrete Ordinates Radiation Transport Calculations. |
| BARC-35 | Abstract | D00124 IBMMF 00 | 35-Group Neutron Cross Sections and Resonance Self-Shielding Factors Generated in ISOTXS and BRKOXS Format from ENDF/B-IV Using MINX. |
| 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. |
| BCG | Abstract | C00578 C0170 00 | A Code For Calculating Pointwise Neutron Spectra and Criticality in Fast Reactor Cells. |
| BEACON MOD3 | Abstract | P00402 CDCMF 00 | Code System for Thermal-Hydraulic Analysis of Nuclear Reactor Containments. |
| BEBC | Abstract | C00077 I7090 00 | Electron Bremsstrahlung Penetration Code for Space Vehicles. |
| BED | Abstract | C00078 I7090 00 | Electron Penetration Code for Space Vehicles. |
| BERMUDA | Abstract | C00616 FV260 03 | Discrete Ordinates Code System for Shielding Analysis for Use with Fusion and Fission Reactors. |
| BEST-5 | Abstract | P00591 I0370 00 | Power Reactor Fuel Cycle Optimization by Bellman Method. |
| BETA II | Abstract | C00117 C6600 00 | Monte Carlo Bremsstrahlung and Electron Transport Analysis in Geometry. |
| BETA II | Abstract | C00117 I0360 00 | Monte Carlo Bremsstrahlung and Electron Transport Analysis in Geometry. |
| BETA-S 6 | Abstract | C00657 MNYCP 01 | Code System to Calculate Multigroup Beta-Ray Spectra. |
BFR
USSO | Abstract | P00449 C0176 00 | Code System for Common Cause Failure Data Analysis. |
| BIGGI | Abstract | C00066 I3675 00 | Numerical Gamma-Ray Transport Code for Plane or Spherical Multilayer Geometry. |
| BIGGI | Abstract | C00066 U1108 00 | Numerical Gamma-Ray Transport Code for Plane or Spherical Multilayer Geometry. |
| BIGGI-4T | Abstract | C00780 I0360 00 | Gamma Transport in Multi-Region Shield in Planar or Spherical Geometry. |
| BISON 1.5 | Abstract | C00464 HM200 00 | One-Dimensional Discrete Ordinate Transport and Burnup Calculation Code System. |
| BISON-C | Abstract | C00659 MNYWS 00 | One-Dimensional Discrete Ordinate Transport and Burnup Calculation Code System. |
| BLOCKAGE V2.5R | Abstract | P00377 IBMPC 00 | Code System to Calculate Potential Plugging of Emergency Core Cooling Suction Strainers by Debris in a BWR. |
BLT-FEMWATER
USSO | Abstract | C00633 PC386 00 | Code System to Solve for Release and Transport of Contaminants through Saturated/Unsaturated Media. |
| BMC-MG | Abstract | C00291 C6600 00 | Multigroup Monte Carlo Neutron and Gamma-Ray Shielding Code System for Plutonium. |
| BOLD VENTURE IV | Abstract | C00459 I3033 00 | A Reactor Analysis Code System. |
| BON | Abstract | P00173 I0360 00 | A Code System for Unfolding Multisphere Spectrometer Neutron Measurements. |
| BOREHOLE-EB6.8-MG | Abstract | D00268 MNYCP 00 | Multi-Group Cross-Section Library for Deterministic and Monte Carlo Codes. |
| 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. |
| BOXER | Abstract | C00766 MNYWS 00 | Fine-flux Cross Section Condensation, 2D Few Group Diffusion and Transport Burnup Calculations |
| BP | Abstract | D00008 I0360 00 | Data for Selected Shielding Benchmark Problems Specified in ORNL-RSIC-25, Shielding Benchmark Problems. |
| BPPC | Abstract | C00076 I7090 00 | Proton Penetration Codes for Space Vehicles. |
| BREESE-II | Abstract | P00143 I3033 00 | Auxiliary Routines for Implementing the Albedo Option in the MORSE Monte Carlo Code System. |
| BREMRAD | Abstract | C00031 I7090 00 | External and Internal Bremsstrahlung Calculation Code. |
| BRHGAM | Abstract | C00350 I3033 00 | Monte Carlo Estimation of Absorbed Dose from X-Ray Sources in Phantom Man. |
| 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. |
| BUGENDF70.BOLIB | Abstract | D00262 PCX86 00 | ENDF/B-VII.0 Broad-Group Coupled Cross Section Library for LWR Shielding & Pressure Vessel Dosimetry Applications. |
| BUGJEFF311.BOLIB | Abstract | D00254 MNYCP 01 | JEFF-3.1.1 Broad-Group Coupled Cross Section Library For LWR Shielding & Pressure Vessel Dosimetry Applications. |
| BUGLE-80 | Abstract | D00075 IBMPC 03 | Coupled Neutron, Gamma-Ray Group Cross Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications. |
| BUGLE-80 | Abstract | D00075 PC386 01 | Coupled Neutron, Gamma-Ray Group Cross Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications. |
| BUGLE-93 | Abstract | D00175 ALLCP 01 | Coupled Neutron, Gamma-Ray Group Cross Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications. |
| BUGLE-96 | Abstract | D00185 ALLCP 00 | Coupled Neutron, Gamma-Ray Group Cross Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications. |
| BULK_C-12 | Abstract | C00738 PC586 00 | Code System to Estimate Neutron and Photon Effective Dose Rates from Medium Energy Protons or Carbon Ions Through Concrete or Concrete/Iron. |
| 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. |
| BURP-2 | Abstract | C00237 C6600 00 | Calculation of Buildup and Decay of Radioactive Fission Products. |
| BUSH | Abstract | C00333 I0360 00 | A Code to Calculate Radiation Doses Inside Buildings from Routine Releases of Radionuclides to the Atmosphere. |
| BUTTERCUP | Abstract | C00779 MNYCP 00 | A Dual-Layer Photon Buildup Factor Code. |
| BWR-LTAS | Abstract | C00485 I3033 01 | A Boiling Water Reactor Long-Term Accident Simulation Code. |
| CAAC | Abstract | C00476 D0VAX 00 | Code System for Implementation of Atmospheric Dispersion Assessment Required by the Clear Air Act. |
| CAAC | Abstract | C00476 I3033 00 | Code System for Implementation of Atmospheric Dispersion Assessment Required by the Clear Air Act. |
| CACA-2 | Abstract | C00302 I0360 00 | Heavy Isotope and Fission-Product Concentration Calculation Code System. |
| CAD | Abstract | D00059 I0360 00 | 51 Neutron, 25 Gamma-Ray Group ALBEDO DATA Generated with DOT for Various Materials. |
| 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. |
| CALKUX | Abstract | C00594 IBMPC 00 | Code System to Calculate Exposure Transmission of Medical X-ray Beams Through Barrier Materials. |
| CALOR95 | Abstract | C00610 MNYWS 00 | Monte Carlo Code System for Design and Analysis of Calorimeter Systems, Spallation Neutron Source (SNS) Target Systems, etc. |
| CAMERA | Abstract | C00240 C0074 00 | Radiation Transport Analysis Code System and the Computerized Man (CAM) Model. |
| CAMERA | Abstract | C00240 IBMPC 01 | Radiation Transport Analysis Code System and the Computerized Man (CAM) Model. |
| CANDULIB-AECL | Abstract | D00210 MNYCP 00 | Burnup-Dependent ORIGEN-S Cross Section Libraries for CANDU Reactor Fuel Characterization. |
| CAP-88 | Abstract | C00542 D0VAX 00 | Clean Air Act Assessment Package. |
| CAP-88 | Abstract | C00542 I3090 00 | Clean Air Act Assessment Package. |
| CAP-88 | Abstract | C00542 IBMPC 00 | Clean Air Act Assessment Package. |
| CAP88-PC | Abstract | C00542 IBMPC 01 | Clean Air Act Assessment Package. |
| CAPS-2 | Abstract | C00074 CDCMF 00 | Analysis of Structures for Fallout Radiation Shielding. |
| CARL 2.3 | Abstract | C00743 PC586 01 | Code System to Calculate Radiotoxicity, Activity, Dose and Decay Power Calculations for Spent Fuel. |
| CARMEN SYSTEM | Abstract | C00487 U1110 00 | A Code System for Neutronics PWR Calculation by Diffusion Theory with Space-Dependent Feedback Effects. |
| CARNAC | Abstract | C00238 I3691 00 | Calculation of Flux and Neutron Spectra in the Case of Criticality Accident. |
| CARP-82 | Abstract | P00131 I3033 00 | Multigroup Albedo Data Using DOT Angular Flux Results. |
| CARSTEP | Abstract | C00024 I7090 00 | Trajectory and Environment Code-Electron and Proton Fluxes Impinging on Spacecraft in Orbit. |
| CASCADE | Abstract | C00176 C6600 00 | Monte Carlo Simulation of the Transport of High Energy Electrons and Photons in Matter. |
| CASCADE | Abstract | C00176 I0360 00 | Monte Carlo Simulation of the Transport of High Energy Electrons and Photons in Matter. |
| CASIM | Abstract | C00265 I0360 00 | Monte Carlo Simulation of Transport of Hadron Cascades in Bulk Matter. |
| CASK | Abstract | D00023 I3691 04 | 22 Neutron, 18 Gamma-Ray Group, P3, Cross Sections for Shipping Cask Analysis. |
| CASK-81 | Abstract | D00023 IBMPC 06 | 22 Neutron, 18 Gamma-Ray Group, P3, Cross Sections for Shipping Cask Analysis. |
| CASK-81 | Abstract | D00023 I0370 05 | 22 Neutron, 18 Gamma-Ray Group, P3, Cross Sections for Shipping Cask Analysis. |
| 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. |
| CAVEAT | Abstract | C00169 I3675 00 | General Purpose Monte Carlo Time-Dependent Radiation Transport Code in Complex Geometry. |
| CCRMN | Abstract | P00366 MNYCP 00 | Monte Carlo Simulation of the Coupled Transport of Electrons and Photons. |
| CDR | Abstract | C00182 C6600 00 | A Constant Dose Range Code System, Using the LANL-NWEF Neutron-Gamma-Ray Air Flux Tape. |
| CDR | Abstract | C00182 I0360 00 | A Constant Dose Range Code System, Using the LANL-NWEF Neutron-Gamma-Ray Air Flux Tape. |
| 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. |
| CEPXS | Abstract | C00837 MNYCP 00 | Coupled Electron-Photon Cross Section |
| CEPXS/ONELD 1.0 | Abstract | C00544 MNYCP 02 | One-Dimensional Coupled Electron-Photon Multigroup Discrete Ordinates Code System. |
| 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. |
| CHAINS-PC | Abstract | C00604 IBMPC 00 | Code System to Compute Atom Density of Members of a Single Decay Chain. |
| CHAINT-MC | Abstract | C00584 CYXMP 00 | A Two-Dimensional Model for the Analysis of Contaminant Transport in a Fractured Porous Medium. |
| CHARGE II | Abstract | C00070 C6500 00 | Space Radiation Shielding Code - Proton and Electron Penetration of Multilayered Slabs and Spheres. |
| CHARGE II | Abstract | C00070 I3675 00 | Space Radiation Shielding Code - Proton and Electron Penetration of Multilayered Slabs and Spheres. |
| CHARGE-PC | Abstract | C00070 IBMPC 00 | Space Radiation Shielding Code - Proton and Electron Penetration of Multilayered Slabs and Spheres. |
| CHENDF 7.02 | Abstract | P00333 MNYCP 05 | Codes for Handling ENDF/B-V and ENDF/B-VI Data. |
| CHNSED | Abstract | C00671 I0360 00 | Code System to Model Sediment & Containment Transport. |
| CINDER 1.05 | Abstract | C00755 PC586 00 | Code System for Actinide Transmutation Calculations |
| CITATION-LDI 2 | Abstract | C00643 PC386 02 | Nuclear Reactor Core Analysis Code System. |
| CLAW-IV | Abstract | D00036 I0360 02 | Coupled 30 Neutron, 12 Gamma-Ray Group Cross Sections Based on ENDF/B-IV for Radiation Transport Calculations. |
| CLAW-IV | Abstract | D00036 I3033 03 | Coupled 30 Neutron, 12 Gamma-Ray Group Cross Sections Based on ENDF/B-IV for Radiation Transport Calculations. |
| CLEAR | Abstract | D00042 I3691 00 | 126 Neutron, 36 Gamma-Ray Cross Sections in AMPX and CCCC Interface Formats for LMFBR Neutronics Calculations. |
| CLES | Abstract | D00233 MNYCP 00 | Cross Section Library of Moderator Materials for Low-Energy Neutron Sources. |
| CLOUD-M | Abstract | C00032 I3565 00 | Gamma-Ray Dose Rate from a Radioactive Cloud-Kernel Integration Code. |
| CNCSN 2009 | Abstract | C00726 PCX86 01 | One, Two- and Three-Dimensional Coupled Neutral and Charged Particle SN Parallel Multi-Threaded Code System. |
| COAG-II | Abstract | P00070 I0360 00 | Calculation of the Westcott Epithermal Index and the Westcott 2200 m/s Neutron Flux. |
| COBB | Abstract | D00016 I3675 01 | 123-Group Neutron Cross Section Data Generated from ENDF/B-II Data for Use in the XSDRN Discrete Ordinates Spectral Averaging Code. |
| 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 |
| COG SUPPLEMENTAL LIBRARIES | Abstract | D00271 MNYCP 00 | COG LibMaker – Data Conversion Utility |
| COG11.1 | Abstract | C00829 MNYCP 00 | Multiparticle Monte Carlo Code System for Shielding and Criticality Use. |
| COGAP | Abstract | P00375 MNYCP 01 | Nuclear Power Plant Containment Hydrogen Control System Evaluation Code. |
| COHORT-II | Abstract | C00198 I7094 00 | General Purpose Monte Carlo Radiation Transport Code System. |
| COLLIMATOR | Abstract | C00136 I7090 00 | Monte Carlo Calculation of the Spectrum of Gamma Radiation from a Collimated Co-60 Source. |
| COLUMN2 | Abstract | C00534 ALLMF 00 | Calculation of Effects of Physicochemical Processes on Migration. |
| 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). |
| COMPRASH | Abstract | C00072 I3675 00 | Spinney Removal-Diffusion Shielding Code. |
| COMRADEX4 | Abstract | C00332 I0360 00 | Evaluator of Potential Radiological Doses in the Near (< 10 km) Environment of Radioactive Release. |
| CONDOR-3 | Abstract | C00811 I0370 00 | Two-Dimensional Reactor Program with Local and Spectrum Dependent Burnup. |
| CONDOS-II | Abstract | C00416 I0360 00 | Code for Estimating Radiation Doses from Radionuclide-Containing Consumer Products. |
| 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. |
| CONSTRIP V | Abstract | C00139 I3675 00 | Vertical Barrier-Finite Source Plane Gamma-Ray Penetration Code System. |
| 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. |
| COV-15GROUP-2006 | Abstract | D00232 MNYCP 00 | 15-Group Cross Section Covariance Matrix Library. |
| COVERV | Abstract | D00077 I0360 01 | Compilation of Multigroup Cross-Section Covariance Matrices in COVERX Format for Several Important Materials. |
| COVERX | Abstract | D00044 I0360 02 | Compilation of Multigroup Cross-Section Covariance Matrices in COVERX Format for Several Important Materials. |
| COVFILS | Abstract | D00091 I0360 00 | Neutron Data and Covariances for Sensitivity and Uncertainty Analysis. |
| COVFILS-2 | Abstract | D00137 ALLCP 00 | Neutron Data and Covariances for Sensitivity and Uncertainty Analysis. |
| CRAC2 | Abstract | C00419 C0000 00 | Code System for Calculating Reactor Accident Consequences. |
| CRAC2 | Abstract | C00419 I3033 00 | Code System for Calculating Reactor Accident Consequences. |
| 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. |
| CRRIS | Abstract | C00518 I3033 00 | Computerized Radiological Risk Investigation System for Assessing Doses and Health Risks from Atmospheric Releases of Radionuclides. |
| CRRIS | Abstract | C00518 PC586 00 | Computerized Radiological Risk Investigation System for Assessing Doses and Health Risks from Atmospheric Releases of Radionuclides. |
| CRYO-S(A,B)-ACE1 | Abstract | D00253 MNYCP 00 | Scattering Law and Continuous Energy Cross Section Library of Materials at Cryogenic Temperatures. |
| CRYSTAL BALL | Abstract | C00233 C6600 00 | Code System for Determining Neutron Spectra from Activation Measurements. |
| CRYSTAL BALL | Abstract | C00233 I0360 00 | Code System for Determining Neutron Spectra from Activation Measurements. |
| CTF | Abstract | C00871 PCX86 00 | CTF is the modernized version of the legacy subchannel thermal hydraulics (TH) code COBRA-TF |
| CTR DATA | Abstract | D00028 I3675 01 | 73-Group P3 Coupled Neutron and Gamma-Ray Cross Sections for Fusion Reactor Calculations. |
| CUPED | Abstract | P00032 I3675 00 | Scintillation Spectrometer Polyenergetic Gamma Photon Experimental Distributions Unfolding Code. |
| CYGAS | Abstract | C00317 I3033 00 | A Gamma-Ray Attenuation Code System for Large Gamma-Ray Sources Shielded by Coaxial Cylinders. |
| CYGNUS-C SPHERE | Abstract | C00232 I0360 00 | Monte Carlo Neutron Transport Code System in Spherical Geometry. |
| CYLDOS | Abstract | C00389 I0360 00 | A Cylindrical Geometry Gamma-Ray Flux Attenuation Code System. |
| D2O | Abstract | P00398 PC486 00 | Code System for Computing Thermodynamic and Transport Properties of D2O. |
| DABL69 | Abstract | D00130 I0360 01 | Defense Nuclear Applications Broad-Group Library based on ENDF/B-V in ANISN Format. |
| DACRIN | Abstract | C00273 U1100 00 | Airborne Radionuclide Organ Dose Calculational System. |
| 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. |
| DANTSYS 3.0 | Abstract | C00547 MFMWS 01 | One-, Two-, and Three-Dimensional, Multigroup, Discrete-Ordinates Transport Code System. |
| DART-V.1 | Abstract | C00830 MNYCP 00 | Displacement per Atom, Primary Knocked-on Atoms Produced in an Atomic Solid Target |
| DASH-FP | Abstract | C00366 C0000 00 | A One-Dimensional Analytic-Numerical Solution to the Problem of Multicomponent Time-Dependent Diffusion of Fission Products. |
| DASQHE | Abstract | P00278 D8810 00 | Calculates Dancoff Corrections Factors. |
| DATINIT | Abstract | P00258 DGMV1 00 | Interactive Program To Access Photon Interaction Data. |
| DAVE | Abstract | C00166 I3675 00 | Monte Carlo Gamma-Ray Transport Code System in One-Dimensional Spherical Geometry. |
| DCHAIN 1.3 | Abstract | C00640 MNYCP 01 | Code System for Radioactive Decay and Reaction Chain Calculations. |
| DCHAIN-SP2001 | Abstract | C00712 MNYWS 01 | Code System for Analyzing Decay and Build-up Characteristics of Spallation Products. |
| DCTDOS | Abstract | C00520 IBMPC 00 | Neutron and Gamma-Ray Penetration in Composite Duct Systems. |
| DDXCODES | Abstract | C00583 FM380 00 | One-, Two- and Three-Dimensional Transport Codes Using Multigroup Double-Differential Form Cross Sections. |
| DDXLIB | Abstract | D00123 FM380 01 | 125-Neutron Group Double Differential Cross Section Library. |
| DECAYREM | Abstract | D00030 I0360 02 | Radioactive Decay Spectra in EXREM Format. |
| DECDC 1.0 | Abstract | D00213 MNYCP 00 | Nucear Decay Data Files for Radiation Dosimetry Calculations. |
| DEIS | Abstract | C00455 C6600 00 | Draft Environmental Impact Statement on Licensing Requirements for Land Disposal of Radioactive Waste. |
| DEMON & DEMON R | Abstract | C00181 I3675 00 | Demonstration Monte Carlo Code System in Slab Geometry. |
| 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. |
| DIAMANT2 | Abstract | C00414 PC386 00 | Multigroup Two-Dimensional Discrete Ordinates Transport Code System for Triangular Geometry, Release 2.0. |
DIF3D 11.2892
FEDC | Abstract | C00784 MNYCP 02 | Code System Using Variational Nodal Methods and Finite Difference Methods to Solve Neutron Diffusion and Transport Theory Problems. |
| DIFBAS | Abstract | P00334 MNYCP 00 | A Bayesian Approach to Unfolding a Neutron Spectrum from a Spectrum of Recoiled Protons. |
| DIFMOD | Abstract | C00572 I3083 00 | A Computer Program To Calculate The Leaching of Radionuclides and the Corrosion of Cemented Waste Forms in Water or Brine. |
| 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. |
| DINT-YAEC | Abstract | C00306 ALLMF 00 | Evaluator of I1 and I2 Integrals as Used in Long-Term External Gamma-Ray Doses from Routine Atmospheric Releases. |
| DIPHO | Abstract | C00140 I3675 00 | Monte Carlo Gamma-Ray Code System-Infinite Medium, Mono-energetic and Isotropic Point Source. |
| DISDOS | Abstract | C00170 I0360 00 | Calculation of Dose Distribution in Human Phantoms Irradiated by External Photon Sources. |
| DISKTRAN | Abstract | C00533 CYXMP 00 | Dose Calculations at Detectors from the End of a Cylinder Using DOT IV Scalar Flux Data. |
| DISKTRAN | Abstract | C00533 I3033 00 | Dose Calculations at Detectors from the End of a Cylinder Using DOT IV Scalar Flux Data. |
| DISPERS | Abstract | C00454 MNYCP 00 | Collection of Mathematical Models for Dispersion in Surface Water and Groundwater. |
| DIXY-2 | Abstract | C00812 I0370 00 | 2-D Homogeneous and Inhomogeneous Neutron Diffusion N X-Z, R-Z, R-Theta Geometry with Perturbation. |
| DKR | Abstract | C00323 CY000 00 | A Radioactivity and Dose Rate Calculation Code System for Fusion Reactors. |
| DLS | Abstract | C00264 C6600 00 | Two-Dimensional Shielding Calculational System with Diffusion Theory and Line-of-Sight Method. |
| 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. |
| DOORS 3.2A | Abstract | C00650 MFMWS 04 | One, Two- and Three-Dimensional Discrete Ordinates Neutron/Photon Transport Code System. |
| DOPEX | Abstract | C00177 I3675 00 | Laminated Shield Weight Optimization Code System-Steepest Descent Calculational Model. |
| DOPEX | Abstract | C00177 U1108 00 | Laminated Shield Weight Optimization Code System-Steepest Descent Calculational Model. |
| DOPEX-1D2C | Abstract | C00214 I0360 00 | A One-Dimensional, Two-Constraint Radiation Shield Optimization Code System. |
| 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. |
| DOSCOV | Abstract | D00090 I0360 00 | 24-Group Covariance Data. |
| DOSDAM77-81 | Abstract | D00081 C6400 00 | Multigroup Cross Sections in SAND-II Format for Spectral, Integral, and Damage Analyses. |
| DOSDAM81-82 | Abstract | D00097 C0000 00 | Multigroup Cross Sections in SAND-II Format for Spectral, Integral, and Damage Analyses. |
| DOSDAM84 | Abstract | D00131 IBMMF 00 | Multigroup Cross Sections in SAND-II Format for Spectral, Integral, and Damage Analyses. |
| DOSDAT II-81 | Abstract | D00079 I0370 00 | Dose-Rate Conversion Factors for External Exposure to Photons and Electrons. |
| DOSDAT-DOE | Abstract | D00144 ALLMF 00 | Dose-Rate Conversion Factors for External Exposure to Photons and Electrons. |
| DOSDAT-DOE | Abstract | D00144 IBMPC 01 | Dose-Rate Conversion Factors for External Exposure to Photons and Electrons. |
| DOSE 1 | Abstract | C00165 I3675 00 | Gamma-Radiation Dosimetry for Arbitrary Source and Target Geometry. |
| DOSE-SGTR | Abstract | C00624 IBMPC 00 | Code System to Calculate the Integrated Iodine Release to the Environment During a Steam Generator Tube Rupture in a PWR. |
| DOSFACTER II | Abstract | C00400 D0750 00 | Calculation of Dose-Rate Conversion Factors for Exposure to Photons and Electrons. |
| DOSFACTER II | Abstract | C00400 I0360 00 | Calculation of Dose-Rate Conversion Factors for Exposure to Photons and Electrons. |
| DOSFACTER-DOE | Abstract | C00536 I3033 00 | Calculation of Dose-Rate Conversion Factors for Exposure to Photons and Electrons. |
| DOT 3.5 | Abstract | C00276 I0360 00 | Two-Dimensional Discrete Ordinates Radiation Transport Code System. |
| DPCT | Abstract | C00580 CYXMP 00 | A Deterministic-Probabilistic Model For Contaminant Transport. |
| DPL-400 GEDT1 | Abstract | D00031 I0360 08 | Defense Nuclear Agency 37 Neutron, 21 Gamma Ray Coupled, P3, Multigroup Library in ANISN Format. |
| DPL-401 NEDT | Abstract | D00031 I0360 09 | Defense Nuclear Agency 37 Neutron, 21 Gamma Ray Coupled, P3, Multigroup Library in ANISN Format. |
| DPL-402A/GPDT1 | Abstract | D00031 I0360 10 | Defense Nuclear Agency 37 Neutron, 21 Gamma Ray Coupled, P3, Multigroup Library in ANISN Format. |
| DPL-402B/GPDT1 | Abstract | D00031 I0360 11 | Defense Nuclear Agency 37 Neutron, 21 Gamma Ray Coupled, P3, Multigroup Library in ANISN Format. |
| DRAGON2PARTISN | Abstract | C00803 PCX86 00 | Cross-Sections Data Generation for PARTISN4.0. |
| DRAGON3.05D | Abstract | C00647 MNYWS 03 | Lattice Cell Code System. |
| DRALIST | Abstract | D00080 ALLCP 00 | Radioactive Decay Data for Application to Radiation Dosimetry and Radiological Assessments. |
| DSNP | Abstract | P00592 I3033 00 | Dynamic Simulation Nuclear Power. |
| DSNQUAD | Abstract | P00251 IPCXT 00 | Calculates Angular Quadrature Weights and Cosines. |
| DTF-69 | Abstract | C00130 C6600 00 | Multigroup Neutron Transport Discrete Ordinates Code System with One-Dimensional, Anisotropic Scattering. |
| DTF-INDIA | Abstract | C00458 I0370 00 | Multigroup Neutron Transport Discrete Ordinates Code System with One-Dimensional, Anisotropic Scattering. |
| DTF-IV | Abstract | C00042 C6600 00 | Multigroup Neutron Transport Discrete Ordinates Code System with One-Dimensional, Anisotropic Scattering. |
| DTF-IV MODIFIED | Abstract | C00042 I0370 00 | Multigroup Neutron Transport Discrete Ordinates Code System with One-Dimensional, Anisotropic Scattering. |
| DTF-TRACA | Abstract | C00412 U1100 00 | Multigroup Neutron Transport Discrete Ordinates Code System with One-Dimensional, Anisotropic Scattering. |
| DTK | Abstract | C00223 I3675 00 | One-Dimensional Multigroup Neutron Transport Code System. |
| DUFOLD | Abstract | P00042 I0360 00 | Derivative Unfolding Code - Determination of Neutron Spectra from NE-213 Pulse Height Data. |
| DUST | Abstract | C00453 I3033 00 | Albedo Monte Carlo Simulation of Neutron Streaming Through Multilegged Ducts. |
| DUST-BNL | Abstract | C00634 PC386 00 | Disposal Unit Source Term by One-Dimensional, Transient, Finite-Difference, Subsurface Release and Transport of Contaminants. |
| DWBA07/DWBB07 | Abstract | P00338 MNYCP 01 | Code System for Inelastic and Elastic Scattering with Nucleon-Nucleon Potential |
| DWNWND | Abstract | C00383 DP010 00 | Interactive Gaussian Plume Atmospheric Transport Model. |
| 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. |
| E3LWR | Abstract | D00098 C0000 00 | 45 Neutron, 16 Gamma-Ray and 15 Neutron, 5 Gamma-Ray Group LWR Cross Section Libraries Derived from EURLIB-III using the AGRUKO Optimized Collapsing Scheme. |
| EACRP-D2O-LATTICES | Abstract | D00264 MNYCP 00 | Compilation of Reactor Physics Measurements in HWRs Lattices. |
| EASY-QAD 2.0.1 | Abstract | C00744 PC586 02 | A Visualization Code System for Gamma and Neutron Shielding Calculations. |
| ECIS-12 | Abstract | P00612 MNYCP 00 | Code System to Solve the Coupled Differential Equations Arising in Nuclear Model Calculations. |
| ECPL82 | Abstract | D00106 ALLCP 00 | Evaluated Charged-Particle Data Library. |
| E-DEP-1 | Abstract | C00275 D0VAX 00 | Heavy Ion Energy Deposition Code System. |
| 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. |
| EDMULT 6.4 | Abstract | C00430 MNYCP 02 | Evaluates Electron Depth-Dose Distributions in Multilayer Slab Absorbers. |
| EDNA | Abstract | C00104 I7090 00 | Electron Dose and Number Analysis Code by Kernel Integration. |
| EDO | Abstract | C00489 U1110 00 | A Code System in Fortran V for the Evaluation of Dose During Normal Operation of a Nuclear Power Plant. |
EDSFI
USSO | Abstract | D00215 PC486 00 | Electrical Distribution System Functional Inspection Data Base. |
| EEDB | Abstract | P00531 MNYCP 00 | The Energy Economic Data Base. |
| EFDOS | Abstract | C00411 I0360 00 | Calculation of Effective Committed Dose Equivalents by Inhalation of Radioactive Materials Occurring in Routine Atmospheric Releases from Nuclear Fuel Cycle Facilities. |
| EGAD | Abstract | C00206 I0360 00 | Calculation of Dose from External Gamma-Ray Emitters. |
| EGS4 | Abstract | C00331 MNYCP 00 | Monte Carlo Simulation of the Coupled Transport of Electrons and Photons. |
| ELAN | Abstract | P00141 ICL00 00 | Neutron Cross-Section Self-Shielding Code System. |
| ELAST2 | Abstract | D00208 MNYCP 00 | Database of Cross Sections for the Elastic Scattering of Electrons and Positrons by Atoms. |
| ELBA | Abstract | C00119 I0360 00 | Electron and Bremsstrahlung Dose Rate Code. |
| ELECSPEC | Abstract | D00100 DP010 00 | Electron Spectra from Decay of Fission Products. |
| ELEORBIT | Abstract | C00751 PCX86 00 | 3-D Simulation of Electron Orbits in Magnetic Multipole Plasma Source. |
| ELF | Abstract | C00167 I0360 00 | Monte Carlo Neutron Transport Code System for Cylinders and Spheres. |
| ELGATL | Abstract | C00295 C6600 00 | Calculation of Energy Spectra from Coupled Electron-Photon Slowing Down. |
| ELIESE-3 | Abstract | P00003 I0370 00 | Analyses of Elastic and Inelastic Scattering Cross Sections. |
| ELPHO | Abstract | C00301 I0360 00 | Three-Dimensional Monte Carlo Electromagnetic Transport Code System. |
| ELTRAN | Abstract | C00155 C3600 00 | One-Dimensional Monte Carlo Electron Transport Code System. |
| EMERALD | Abstract | C00211 I0360 00 | Calculation of Activity Releases and Potential Doses from a Pressurized Water Reactor Plant. |
| EMERALD-NORMAL | Abstract | C00250 I0370 00 | Calculation of Activity Releases and Potential Doses from the Normal Operation of a Pressurized Water Reactor Plant. |
| 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. |
| ENDF UTIL. CODES | Abstract | M00008 MNYCP 00 | ENDF Checking and Utility Codes. |
| ENDL82 | Abstract | D00103 ALLCP 00 | Neutron Library in Transmittal Format. |
| ENDLIB-97 | Abstract | D00179 MNYCP 01 | LLNL Libraries of Atomic Data, Electron Data, and Photon Data in Evaluated Nuclear Data Library (ENDL) Type Format. |
| ENDVER/GUI | Abstract | P00572 PCX86 00 | The ENDF File Verification Support Package. |
| ENEDEP | Abstract | C00227 GE400 00 | Energy Deposition Code System for GE 265 Time-Sharing System. |
| ENLOSS | Abstract | P00047 C6600 00 | Calculation of Energy Loss of Charged Particles. |
| ENSL82-CDRL82 | Abstract | D00110 ALLCP 00 | Evaluated Nuclear Structure Libraries. |
| 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. |
| EPICS2014 | Abstract | D00272 MNYCP 00 | Electron Photon Interaction Cross Sections |
| EPICS2017 | Abstract | D00272 MNYCP 01 | Electron Photon Interaction Cross Sections |
EPIPE
USSO | Abstract | P00485 CY000 00 | Code System for Static and Dynamic Piping System Analysis. |
| EPR | Abstract | D00037 I3691 05 | Coupled 100-Group Neutron 21-Group Gamma-ray Cross Sections for EPR Neutronics. |
| EPR MASTER | Abstract | D00052 I3691 00 | 100 Neutron Group Cross Sections in AMPX Master Library Format. |
| EPRI-CINDER | Abstract | C00309 C6600 00 | General Point-Depletion and Fission Product Code System and Four-Group Fission Product Neutron Absorption Chain Data Library Generated from ENDF/B-IV for Thermal Reactors. |
| EQUIVA-1.1 | 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. |
ERANOS 2.0
OECD | Abstract | C00745 MNYWS 00 | Modular Code and Data System for Fast Reactor Neutronics Analyses |
| 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. |
| ERPEX | Abstract | C00305 C0073 00 | Monte Carlo Distributions of Energetic Proton Ranges in Silicon. |
| ERRORJ | Abstract | P00526 MNYCP 03 | Multigroup Covariance Matrices Generation from ENDF/B-6 Format. |
| ESDORA | Abstract | C00183 U1108 00 | Fission Product Inventory and Gamma-Ray Dose Rate from a Radioactive Cloud System. |
| ESG | Abstract | D00065 I0360 00 | 56-Group Cross Section Library Based on VITAMIN-C Generated by Using SPHINX and XSDRNPM to Collapse 171 Groups. |
| ESP | Abstract | C00193 I0360 00 | General Purpose Monte Carlo Neutron Transport Code System. |
| 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. |
| ETRAN | Abstract | C00107 I0360 00 | Monte Carlo Code System for Electron and Photon Through Extended Media. |
| EURCYL | Abstract | P00076 I0370 00 | Finite Element Three-Dimensional Mesh Generator for Cylinder - Cylinder Intersections. |
| EURLIB-III | Abstract | D00035 I0360 01 | 100 Neutron, 20 Gamma-Ray Group Cross Section Library for Use in the European Shielding Benchmark Program. |
| 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. |
| EXPALS | Abstract | C00787 C7600 00 | Least Square Fit of Linear Combination of Exponential Decay Function. |
| EXPRESS | Abstract | C00622 MNYCP 00 | Exact Preparedness Supporting System. |
| EXTREME | Abstract | C00440 I3033 00 | Two-Dimensional Discrete-Ordinates Code System with Exponential Expansion of Spatial Variables. |
| 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. |
| FANTOM | Abstract | C00375 BESM6 00 | Monte Carlo Calculation of the Response of an External Detector to a Photon Source in the Lungs of a Heterogeneous Phantom. |
| FASTER III | Abstract | C00168 U1108 00 | Monte Carlo Neutron and Photon Transport Code System in Complex Geometries. |
| FASTER-III | Abstract | C00168 I3675 00 | Monte Carlo Neutron and Photon Transport Code System in Complex Geometries. |
| 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. |
| FCXSEC | Abstract | D00085 PC386 01 | 22 Neutron, 21 Gamma-Ray Group Cross Section Libraries in ANISN Format for Nuclear Fuel Cycle Shielding Calculations. |
| FDKR | Abstract | C00541 I4381 00 | Radioactivity and Dose Rate Calculation Code for Fission, Fusion and Hybrid Reactors. |
| FDMXPC | Abstract | P00322 IPCAT 00 | Code System for Calculation of Neutron Transmission and Other Functionals from Evaluated Data in ENDF Format. |
| FE3DGW | Abstract | C00531 D0780 00 | Code System for Finite-Element, Three-Dimensional Ground-Water Flow Analysis. |
| 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. |
| FEM-2D | Abstract | C00260 C6600 00 | Two-Dimensional Diffusion Theory Code System Based on the Method of Finite Elements. |
| FEMAXI 6 VER.1 | Abstract | P00536 IBMPC 00 | Code System for Light Water Reactor Fuel Analysis. |
| FEMB | Abstract | C00340 B6700 00 | A Two-Dimensional Diffusion Theory Finite Element Program. |
| FEMRZ | Abstract | C00342 F2307 00 | A Finite-Element Method Two-Dimensional Multigroup Neutron Transport Code System, (r,z) Geometry. |
| FEMWASTE/FEMWATER | Abstract | C00451 C7600 00 | A Finite-Element Model of Waste and Water Transport through Porous Saturated-Unsaturated Media. |
| FEMWASTE/FEMWATER | Abstract | C00451 PC386 00 | A Finite-Element Model of Waste and Water Transport through Porous Saturated-Unsaturated Media. |
| FENDL-2.0 | Abstract | D00183 MNYCP 01 | Compendium of Reference and Processed Sub-libraries Derived from International Evaluated Nuclear Data Files for Fusion Applications. |
| FENDL-2.1 | Abstract | D00222 MNYCP 00 | Compendium of Reference and Processed Sub-libraries Derived from International Evaluated Nuclear Data Files for Fusion Applications. |
| 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. |
| FESH | Abstract | C00676 CDCMF 00 | X-Y Multigroup Neutron Transport Code System. |
| FEWA-FEMA | Abstract | C00477 I3033 00 | A Finite Element Model of Water and Other Material through Aquifers. |
| FEWG1-81 | Abstract | D00031 I0370 06 | Defense Nuclear Agency 37 Neutron, 21 Gamma Ray Coupled, P3, Multigroup Library in ANISN Format. |
| FEWG1-85 | Abstract | D00031 I0360 07 | Defense Nuclear Agency 37 Neutron, 21 Gamma Ray Coupled, P3, Multigroup Library in ANISN Format. |
| FGR-DOSE | Abstract | D00167 ALLCP 01 | Dose Coefficients from Federal Guidance Reports 11 and 12. |
| FGXRRS | Abstract | D00132 C0000 00 | Few Group Cross Section Library for Research Reactor Calculations. |
| FIGERO | Abstract | P00149 C0000 00 | Processing Codes for Generating Multigroup Neutron Cross Sections from ENDF/B for Use in Discrete Ordinates Calculations. |
| FINELM | Abstract | C00483 MFMWS 00 | Multigroup Finite Element Diffusion Code System. |
| FIPDIG | Abstract | C00251 I0360 00 | One-Dimensional Time-Dependent Fission Product Diffusion Code System. |
| FIRAC | Abstract | P00444 CY000 00 | Nuclear Facilities Fire Accident Model |
| FIREDATA | Abstract | D00125 PC486 00 | Nuclear Power Plant Fire Data Base for Personal Computers. |
| FISP-6 | Abstract | C00538 I3090 00 | An Enhanced Code for the Evaluation of Fission Product Inventories and Decay Heat. |
| FISPACT-II 5.0 | Abstract | C00836 MNYCP 03 | Inventory Simulation Platform for Nuclear Observables and Materials Science. |
| FISPIN | Abstract | C00413 ICL00 00 | Nuclide Inventory Calculation System. |
| FIS-PROD | Abstract | D00152 ALLCP 00 | Chinese Evaluated Fission Product Yield Library in ENDF/B-V Format. |
| FISSP & CLOUD | Abstract | C00163 MNYCP 01 | Fission Product Inventory, Release, Transport and Dose Calculation. |
| 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. |
| FLEP | Abstract | D00022 I3033 00 | Coefficients for the Analytic Representation of Nonelastic Cross Sections and Particle-Emission Spectra from Various Nucleon-Nucleus Collisions in the Energy Range 25 to 400 MeV. |
| 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. |
| FLUKA05-PRE-LIB | Abstract | D00260 PCX86 00 | FLUKA05 Multi-Group, Multi-Purpose Nuclear Data Library, Neutrons, Photons, Charged Particles. |
| FLUKA-TRANKA | Abstract | C00207 C6600 00 | Three-Dimensional High-Energy Extranuclear Hadron Cascade Monte Carlo System for Cylindrical Backstop Geometries. |
| FLUNG | Abstract | D00086 I3033 00 | Coupled 35-Group Neutron and 21-Group Gamma Ray, P3 Cross Sections for Fusion Applications. |
| 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. |
| FOCUS | Abstract | C00390 I3033 00 | Adjoint Monte Carlo Neutron Transport Code System. |
| FONTA | Abstract | C00423 S4044 00 | Code System For Calculating Individual And Collective Doses From Reactor Accidents Using Pasquill's Plume Model. |
| FOOD | Abstract | C00403 U1108 00 | Calculation of Radiation Dose to Man from Radionuclides in the Environment. |
| 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. |
| FORSS | Abstract | C00334 C0000 00 | A Sensitivity and Uncertainty Analysis Code System. |
| FORSS | Abstract | C00334 I0360 00 | A Sensitivity and Uncertainty Analysis Code System. |
| FOTELP-2014 | Abstract | C00581 MNYCP 04 | Monte Carlo Simulation of Photons, Electrons and Positrons Transport. |
| FOURACES | Abstract | P00183 I0370 00 | Code System for Producing Spectrum Weighted, Group Averaged Cross Sections from ENDF/B, KEDAK, or UK Libraries. |
| FPDL | Abstract | D00066 I0360 00 | Fission Product Yields, Gamma Ray and Beta Spectra in ENDF-III Format for 235U, 238U, 239Pu, 232Th, and 233U. |
| FPGAM | Abstract | C00386 F2307 00 | Calculation of Fission-Product Gamma-Ray Spectra. |
| FPIC | Abstract | C00028 I3675 00 | Fission Product Inventory Code. |
| FPIP | Abstract | C00162 C6600 00 | Fission Product Inventory Code System. |
| FPZD | Abstract | C00603 PC386 00 | Code System for Multigroup Neutron Diffusion/Depletion Calculations. |
| 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. |
| FRCRL2 | Abstract | C00231 C6400 00 | Calculation of Fission-Product Release in Reactor Accident Analyses. |
| FREEFORM | Abstract | P00081 I0360 00 | Free-Form Input Reading Routines. |
| FSCATT | Abstract | C00186 I3033 00 | Discrete Ordinates Gamma-Ray Transport Code System in Plane Geometry. |
| FSCATT | Abstract | C00186 U1108 00 | Discrete Ordinates Gamma-Ray Transport Code System in Plane Geometry. |
| F-SCORE | Abstract | P00617 PCX86 00 | F-Score Nuclide ID Scoring Applications |
| FSKY4C | Abstract | C00771 PCX86 00 | Gamma Ray Skyshine Analysis Code. |
| FSX96 | Abstract | D00190 MNYWS 00 | Collection of Continuous Energy Cross Section Libraries for MCNP Based on JENDL 3.2, JENDL, Fusion File and Dosimetry File. |
| FSXJ32 | Abstract | D00244 MNYCP 00 | A Continuous Energy Cross Section MCNP Nuclear Data Library Based on JENDL-3.2. |
| FSXLIB-J3 | Abstract | D00165 ALLCP 00 | MCNP continuous energy neutron cross section library based on JENDL-3. |
| FSXLIB-J33 | Abstract | D00223 MNYCP 01 | Continuous Energy Neutron Cross Section Library for MCNP Based on JENDL 3.3. |
| FTF | Abstract | D00056 I0360 00 | Multigroup Neutron and Gamma-Ray Dose Transmission Factors for Concrete Slabs. |
| FUELSDATA | Abstract | P00446 C7600 00 | Code System to Model Verification Fuel Rod Data. |
| FURNACE | Abstract | C00615 C0740 00 | Code System for Neutronic Calculations in Three Dimension Toroidal Geometry. |
| G3-6ED | Abstract | C00075 C6600 00 | Kernel Integration Code System - Multigroup Gamma Ray Scattering. |
| G3-6ED | Abstract | C00075 I3033 00 | Kernel Integration Code System - Multigroup Gamma Ray Scattering. |
| GABAS | Abstract | P00175 U1108 00 | A Code System for Generating Composite Time-Dependent Fission Produce Spectra. |
| GADJET | Abstract | C00115 C6600 00 | Monte Carlo Gamma-Ray Adjoint Energy Transport Code in Complex Three-Dimensional Geometry. |
| 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. |
| GAKER-KIRA | Abstract | C00813 C3600 00 | Energy Transfer of Protons in H2O or Polyethylene and Deuterons in D2O. |
| GALAXY-6 | Abstract | P00098 I0370 00 | Neutron Multigroup Cross Section Processor. |
| GALE BWR | Abstract | C00335 U1100 00 | Boiling Water and Pressurized Water Reactors Gaseous and Liquid Effluents Radiological Assessment Code System. |
| GALE PWR & BWR | Abstract | C00335 I3033 00 | Boiling Water and Pressurized Water Reactors Gaseous and Liquid Effluents Radiological Assessment Code System. |
| GALE86 | Abstract | C00506 MNYCP 02 | Calculation of Routine Radioactive Releases in Gaseous and Liquid Effluents from Boiling Water and Pressurized Water Reactors. |
| 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. |
| GAMDAT-78 | Abstract | D00083 I0370 00 | Library of Gamma-Ray Decay Data for 2055 Radionuclides. |
| 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. |
| GAMLIB | Abstract | D00006 I0360 00 | 99-Group Neutron Cross Sections for Use in the GAM Portion of the GGC Multigroup Cross Section Code. |
| 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. |
| GAMMOM | Abstract | C00135 ALLMF 00 | Gamma-Ray Moments Method Code System. |
| GAMMOM-I | Abstract | C00226 I0360 00 | Gamma-Ray Moments Method Code System. |
| GAMMON | Abstract | D00071 ALLCP 00 | Gamma-Ray Moments Method Code System. |
| GAMTAB | Abstract | D00032 I0360 00 | Radioactive-Decay Gamma-Rays Ordered by Energy and Nuclide. |
| GAMTOT78 | Abstract | D00109 CY00I 00 | Compilation of Radioactive Decay and Capture Gamma Rays. |
| GAMX1 | Abstract | P00209 I0370 00 | A Computer Code System for Evaluating Spectra Peak Areas. |
| GANAPOL-ABNTT | Abstract | M00011 MNYCP 00 | Analytical Benchmarks; Case Studies in Neutron Transport Theory. |
| GANDR/SEMOVE | Abstract | C00765 PCX86 00 | Program for Calculating Derivatives of Processed Multigroup Nuclear Data by Discrete Differences. |
| GAPCON-THERMAL | Abstract | P00499 C7600 00 | Code System to Calculate Fuel Steady State & Transient Behavior. |
| GARG | Abstract | D00073 C0000 00 | 27-Group Neutron Cross Sections in Discrete Ordinates Format Generated with FIGERO (PSR-149) from ENDF-B Data. |
| GARLIB | Abstract | D00013 I3565 01 | Multigroup Resonance-Region Cross Sections for Use in Shielding Calculations. |
| GARLIB | Abstract | D00013 I7090 00 | Multigroup Resonance-Region Cross Sections for Use in Shielding Calculations. |
| GAROL | Abstract | P00033 I7090 00 | Calculation of Resonance Neutron Absorption in Two-Region Problems. |
| GASPAR | Abstract | C00463 I3033 01 | Calculates Radiation Exposure to Man from Routine Air Releases of Nuclear Reactor Effluents. |
| GASPAR II | Abstract | C00463 D0780 00 | Calculates Radiation Exposure to Man from Routine Air Releases of Nuclear Reactor Effluents. |
| GASS | Abstract | C00080 I7090 00 | Monte Carlo Calculation of Self Shielding by Encapsulated Gamma-Ray Sources. |
| 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. |
| GBANISN | Abstract | C00628 IRISC 00 | One-Dimensional Discrete Ordinates Transport Code System with Anisotropic Scattering with the GroupBand Option. |
| GCI | Abstract | P00421 IBMPC 00 | Generic Communications Index |
| GEAF-1 | Abstract | D00158 D8810 00 | 100 Group Cross Sections for Neutron Activation. |
| 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. |
| GENII 2.10 | Abstract | C00737 PCX86 02 | Environmental Radiation Dosimetry Software System. |
| GENII-LIN 2.1 | Abstract | C00728 PC586 01 | GENII-LIN Multipurpose Health Physics Code System with a New Object-Oriented Interface, Release 2.0. |
| GENP-2 | Abstract | C00575 ALLMF 00 | Generalized Perturbation Theory Code System. |
| 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. |
| GES_MC | Abstract | C00742 PC586 00 | Gamma-electron Efficiency Simulator. |
| GETOUT | Abstract | C00461 C0176 00 | A Computer Code System for Predicting One-Dimensional Radionuclide Decay Chain Transport through Geologic Media. |
| GFX-GAMIX | Abstract | C00397 I3033 00 | A Spherical Harmonics Code System for Evaluation of Terrestrial Gamma-Radiation Fields. |
| 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. |
| GGG-GP | Abstract | C00564 IBMPC 00 | Kernel Integration Code System - Multigroup Gamma-Ray Scattering Using the GP Buildup Factor. |
| GGTC-ENEL | Abstract | P00128 I0360 00 | Code System for Producing Few-Group Neutron Cross Sections from Multigroup Data Libraries. |
| GICX40 | Abstract | D00092 ALLCP 00 | Coupled 42-Neutron, 21-Gamma-Ray Group Cross Sections for 40 Elements in Group Independent Form for Fusion Reactor Calculations. |
| 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. |
| GNOMER | Abstract | C00625 MNYCP 01 | Multigroup 3-Dimensional Neutron Diffusion Nodal Code System with Thermohydraulic Feedbacks. |
| GOFRR | Abstract | P00127 I0360 00 | Generator of Graphical Output of DOT and ANISN Fluxes and Reaction Rates. |
| GRACE-II | Abstract | C00026 I3675 00 | Gamma Ray Kernel Integration Dose Rate and Heating Code-Cylinders and Spheres. |
| GRASS-SST | Abstract | P00489 MNYCP 00 | Code System to Predict Fission-Gas Release & Fuel Swelling. |
| GREAT-GRASS | Abstract | C00143 I3675 00 | Monte Carlo Radiation Transport Code Systems for Fallout Shielding. |
| GRENADE | Abstract | C00516 C1787 00 | Green's Function Nodal Algorithm for the Diffusion Equation. |
| GRENADE | Abstract | C00516 D0780 00 | Green's Function Nodal Algorithm for the Diffusion Equation. |
| 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. |
| GROUP STRUCTURE | Abstract | D00156 ALLCP 00 | Standard Energy Group Structures Of Cross Section Libraries For Reactor Shielding, Reactor Cell Fusion Neutronics Applications: VITAMIN-J, ECC0-33, ECC0-2000. |
| GROUPSTRUCTURES | Abstract | D00274 MNYCP 00 | GROUPSTRUCTURES, VITAMIN-J, XMAS, ECCO-33, ECCO2000 Standard Group Structures |
| 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. |
| GRSAC | Abstract | C00774 PCX86 00 | Graphite Reactor Severe Accident Code. |
| GRTUNCL3D | Abstract | C00721 MNYCP 01 | Code to Calculate Semi-Analytic First Collision Source and Uncollided Flux. |
| 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. |
| GUI2QAD-3D | Abstract | C00697 PC586 01 | A Graphical User Interface for QAD-CGPIC, a Point Kernal Code for Neutron and Gamma-Ray Shielding Calculations in Complex Geometry. |
| HAARM-3 | Abstract | P00401 CDCMF 00 | Aerosol Behavior Log-Normal Distribution Model. |
| HABIT 1.1 | Abstract | C00665 IBMPC 01 | Code System for Evaluation of Control Room Habitability. |
| HADOC | Abstract | C00452 U1100 00 | Calculates External and Inhalation Doses from Acute Radionuclide Releases on the Hanford Site. |
| HALLMARK | Abstract | D00005 I0360 00 | Discrete Ordinates and Monte Carlo Results of Neutron and Secondary Gamma-Ray Transport in Air-Over-Ground Geometry. |
| HAM | Abstract | C00267 U1108 00 | Monte Carlo Multigroup Neutron and Photon High Altitude Transport Code System. |
| HARAD | Abstract | C00387 I0360 00 | Calculation of Daughter Concentrations in Air Following the Atmospheric Release of a Parent Radionuclide. |
| HASSAN | Abstract | P00593 I0370 00 | Time-Dependent Temperature Distribution and Stress and Strain in HTR Fuel Pins. |
| HATCHES-19 | Abstract | D00206 PC586 02 | Thermodynamic Database for Radiochemical Modelling. |
| 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. |
| HEATKAU | Abstract | C00805 PCX86 00 | HEATKAU Program. |
HECTR 1.5+
USSO | Abstract | P00457 CY000 00 | Hydrogen Event Containment Response Code System. |
| HEITLER | Abstract | P00004 I7030 00 | Cross Section Generator. |
| HELLO | Abstract | D00058 I0360 00 | 47 Neutron, 21 Gamma-Ray Group Cross Sections for Radiation Transport for Neutron Energies up to 60 MeV. |
| HEPROW | Abstract | C00799 MNYCP 00 | Unfolding of Pulse Height Spectra Using Bayes Theorem and Maximum Entropy Method. |
| HERAD | Abstract | C00444 CY00I 00 | Three-Dimensional Monte Carlo Computer Code System for Calculating Radiation Damage from Ion Beams. |
| HERMES-KFA | Abstract | C00687 MNYWS 00 | Monte Carlo Code System for High-Energy Radiation Transport Calculations. |
| HEXAB-3D | Abstract | C00593 I0370 00 | Three-Dimensional Few-Group Coarse Mesh Diffusion Code for Neutron Physics Calculation of Reactor Core in Hexagonal Geometry. |
| HGSYSTEM | Abstract | M00021 MNYCP 00 | Atmospheric Dispersion for Ideal Gases and Hydrogen Fluoride (HF) |
| HGSYSTEMUF6 | Abstract | C00832 MNYCP 00 | Model for Simulating Dispersion due to Atmospheric Release of UF6. |
| HIC-1 | Abstract | C00249 I0360 00 | Monte Carlo Code System for Calculating Heavy Ion Reactions at Energies > 50 MeV/Nucleon. |
| HILO | Abstract | D00087 I0370 00 | Group Cross Sections for Radiation Transport |
| HILO2K | Abstract | D00220 MNYCP 00 | Group Cross Sections for Radiation Transport |
| HILO86 | Abstract | D00119 I0360 00 | Group Cross Sections for Radiation Transport |
| HILO86 | Abstract | D00119 PC386 01 | Group Cross Sections for Radiation Transport |
| HILO86R | Abstract | D00187 ALLCP 00 | Group Cross Sections for Radiation Transport |
| HIMAC | Abstract | M00001 MNYCP 02 | Experimental Data of Neutron Yields from Thick Targets Bombarded by 100 to 800 MeV / Nucleon Heavy Ions. |
| HORN | Abstract | C00568 I3083 00 | A Computer Code To Analyze The Gas-Phase Transport of Fission Products In Reactor Cooling System Under Severe Accidents. |
| HOTSPOT 3.0.2 | Abstract | M00009 IBMPC 03 | Health Physics Code System for Evaluating Accidents Involving Radioactive Materials. |
| HPICE | Abstract | D00007 I0360 05 | Evaluated Photon Interaction Library, ENDF/B File 23 Format. |
| HPPOS 1.5 | Abstract | D00173 IBMPC 00 | Health Physics Position Database. |
| HPPOS V2 | Abstract | D00173 IBMPC 01 | Health Physics Position Database. |
| HSI-DRG | Abstract | P00435 IBMPC 00 | Code System for Use with Human System Interface Design Review Guidelines. |
| HUGO | Abstract | D00099 I3033 00 | Photon Interaction Data in ENDF/B Format. |
| HUGO VI | Abstract | D00146 I3033 00 | Photon Interaction Data in ENDF/B Format. |
| HYACINTH | Abstract | C00294 I0360 00 | Fast Heavy Isotope Point Burnup and Decay Code System - Analytical Solution. |
| 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. |
| ICOM | Abstract | C00651 PC386 00 | Code System for Calculating Ion Track Condensed Collision Model. |
| IDC | Abstract | C00384 I0360 00 | ICRP Dosimetric Calculational System. |
| IEAF-2001 | Abstract | D00217 MNYCP 00 | Intermediate Energy Activation File - 2001. |
| IER | Abstract | P00024 I3675 00 | A Gauss-based Quadrature Formula Applied to Sievert's Integral. An Exponential Integral Routine. |
| IMPACTS-BRC2.1 | Abstract | C00666 IBMPC 00 | Code System for Analysis of Potential Radiological Impacts. |
| IMPORTANCE | Abstract | P00407 I0370 00 | FTA Basic Event & Cut Set Ranking. |
| INAP | Abstract | C00235 U1108 00 | Improved Neutron Activation Prediction Code Systems. |
| INDOS | Abstract | C00236 DP010 00 | Conversational Computer Code Systems to Implement ICRP-10-10A Models for Estimation of Internal Radiation Dose to Man. |
| INDOSE V2.1.1 | Abstract | C00720 PC586 00 | Internal Dosimetry Code System Using Biokinetics Models |
| INDRA | Abstract | C00303 I0360 00 | A Modular System for Calculating the Neutronics and Photonics Characteristics of a Fusion Reactor Blanket. |
| INFLTB | Abstract | P00313 ALLCP 00 | Gamma-Ray Absorption Coefficient Calculation. |
| INGDOS | Abstract | C00408 DP010 00 | A Conversational Code System Designed to Implement NRC Reg-Guide 1.109 Models for Estimation of Annual Doses from Ingestion of Atmospherically Released Radionuclides in Foods. |
| INGEN | Abstract | P00207 C0000 00 | A General-Purpose Mesh Generator for Finite Element Codes. |
| INREM II | Abstract | C00392 I3033 00 | Computer Implementation of Recent Models for Estimating the Dose Equivalent to Organs of Man from an Inhaled or Ingested Radionuclide. |
| INREM/EXREM | Abstract | C00185 I0360 00 | Beta and Gamma Radiation Environmental Dose Code Systems. |
| INTERTRAN I | Abstract | C00473 ALLMF 00 | A Code System for Assessing the Impact from Transporting Radioactive Material. |
| INTRIGUE-II | Abstract | P00054 I0360 00 | Logarithmic and Semilogarithmic CALCOMP Plot Routines. |
| INTRUDE-ANS | Abstract | C00539 D8810 00 | A Repository Intrusion Risk Evaluation Code. |
| INVENT | Abstract | C00540 D8810 00 | A Radionuclide Inventory and Hazard Index Code. |
| IODES | Abstract | C00365 I0360 00 | A Code System for Calculating the Estimation of Dose to the World Population from Releases of Iodine-129 to the Environment. |
| IONMIG | Abstract | C00526 ALLMF 00 | Code System for Radionuclide Migration Calculations. |
| IRAN-LIB | Abstract | D00159 IBMPC 00 | A P-3 Coupled Neutron-Gamma Cross Section Library in ISOTXS For Use with ANISN/PC (CCC-514). |
| IRDAM | Abstract | C00524 IPCXT 00 | Interactive Rapid Dose Assessment Model. |
| IRDF-2002 | Abstract | D00229 MNYCP 01 | The International Reactor Dosimetry File. |
| IRDF82 | Abstract | D00094 I0360 00 | The International Reactor Dosimetry File. |
| IRDF-90 | Abstract | D00161 ALLCP 01 | The International Reactor Dosimetry File. |
| I-R-MAN | Abstract | D00050 ALLCP 00 | Photon Interaction Data on ICRP Reference Man. |
| IRPHE-VENUS-RECYCLE | Abstract | D00263 MNYCP 00 | Plutonium Recycling Physics Project Critical Experiments. |
| IRRAS 4.16 | Abstract | P00386 IBMPC 04 | Code System to Calculate Integrated Reliability and Risk Analysis. |
| ISOGEN II | Abstract | C00055 I3675 00 | Radioisotope Generator Code. |
| ISO-PC 2.1 | Abstract | C00636 IBMPC 01 | Kernel Integration Code System for General Purpose Isotope Shielding Analyses. |
| ITER-2 | Abstract | P00148 C0000 00 | Codes for Unfolding Activation Detector Data and Pulse Height Spectra. |
ITS6
FEDC | Abstract | C00792 PCX86 00 | Integrated TIGER Series of Coupled Electron/Photon Monte Carlo Transport Codes System. |
| JASMINE V.3 | Abstract | C00795 MNYCP 00 | JAEA Simulator for Multiphase INteractions and Explosions. |
| JDL-IMPORTANCE | Abstract | M00005 MNYCP 00 | Adjoint Function: Physical Basis of Variational & Perturbation Theory in Transport & Diffusion Problems. |
| JDL-REACTOR-KIN | Abstract | M00006 MNYCP 00 | Nuclear Reactor Kinetics and Control. |
| JDL-THERMODYNAM | Abstract | M00007 MNYCP 00 | Thermodynamics: Frontiers and Foundations. |
| JENDL/D-99 | Abstract | D00204 MNYCP 00 | JENDL Dosimetry File 99. |
| JENDL-1 | Abstract | D00070 ALLCP 00 | Japanese Evaluated Neutron Cross Section Data in ENDF/B-IV Format. |
| JENDL-2 | Abstract | D00122 FM380 00 | Japanese Evaluated Neutron Cross Section Data in ENDF/B-IV Format. |
| JFS | Abstract | D00111 I3033 00 | 70 Group Neutron Fast Reactor Cross Section Set and 25 Group Neutron Fast Reactor Cross Section Set. |
| JFS3J2 | Abstract | D00108 FM200 00 | 70 Group Neutron Fast Reactor Cross Section Set Based on JENDL-2B. |
| JIMCOF | Abstract | D00078 F2307 00 | Multigroup Constants fFle Based on ENDF/B IV. |
| JN-METD 2&1 | Abstract | C00208 I0370 00 | Neutron Transport Code System with Isotropic Scattering, Bare Slabs and Homogeneous Slabs (JN Method 1), Multilayer Slabs (JN Method 2). |
| K009 | Abstract | C00062 I7090 00 | Solid Angle Integration Charged Particle Penetration Code. |
| K019 | Abstract | C00100 I0360 00 | Shield Thickness Calculation Program for Space Vehicles. |
| KAMCCO | Abstract | C00325 I0370 00 | Three-Dimensional Time Dependent Monte Carlo Code System for Fast Neutron Physics Problems. |
| KAOS/LIB-V | Abstract | D00160 CY000 00 | A Library of Nuclear Response Functions Generated by KAOS-V Code From ENDF/B-V and Other Data Files. |
| KAOS-V | Abstract | P00306 CY000 00 | An Evaluation Tool For Neutron Kerma Factors and Other Nuclear Responses. |
| KAP-VI | Abstract | C00094 U1108 00 | Kernel Integration Code System in Complex Geometry. |
| KASY | Abstract | C00814 I0370 00 | 3-D Homogeneous Neutron Diffusion in X-Y-Z, R-Theta, Hexagonal-Z Geometry by Synthesis Method. |
| KCUT | Abstract | P00584 IBMPC 00 | Code to Generate Minimal Cut Sets for Fault Trees. |
| KDDK | Abstract | D00061 I0360 00 | Measured Results of Delayed Beta- and Gamma-Ray Spectra due to Thermal-Neutron Fission of U-235. |
| KDLIBE | Abstract | C00124 I3675 00 | Kernel-Diffusion Shielding Analysis System. |
| KEDAK3 | Abstract | D00141 I0370 00 | Evaluated Neutron Nuclear Data for Reactor Physics Calculations. |
| KENO2MCNP | Abstract | P00541 PC586 00 | Conversion of Input Data between KENO V.a and MCNP File Formats. |
| KERMAL | Abstract | D00142 ALLCP 00 | Neutron and Gamma-Ray Kerma Factors Based on LLNL Nuclear Data Files. |
| KERNEL | Abstract | C00672 IBMPC 00 | Monte Carlo Code System for Electron (Positron) Dose Kernel Calculations. |
| 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. |
| KICHE 1.3 | Abstract | C00796 PCX86 00 | Kinetics of Iodine Chemistry in the Containment of LWRs under Severe Accident Conditions. |
| KIM | Abstract | C00376 I3033 00 | A Two-Dimensional Monte Carlo Code System for Linear Neutron Transport Calculations. |
| KORIGEN | Abstract | C00457 I3033 00 | A Modification of the Isotope Generation and Depletion Code System ORIGEN. |
| KRAKEN | Abstract | C00877 PCX86 00 | Computational Reactor Analysis Framework. |
| KRONIC | Abstract | C00229 I0360 00 | Calculation of Annual Average External (Beta and Gamma Radiation) Doses from Chronic Atmospheric Releases of Radionuclides. |
| KRONIC | Abstract | C00229 U1108 00 | Calculation of Annual Average External (Beta and Gamma Radiation) Doses from Chronic Atmospheric Releases of Radionuclides. |
| KUX | Abstract | C00515 ALLCP 00 | Medical X-Ray Shielding Calculation. |
| KX-RAY | Abstract | D00021 I0360 00 | Evaluated X-ray Cross Section Library. |
| L26P3S34 | Abstract | D00112 IBMMF 00 | ENDL 26-Group up to P3 Library Prepared by SUPERTOG for 34 Materials. |
| LA100 | Abstract | D00168 ALLCP 00 | Evaluated Nuclear Data Library for Transport Calculations Involving Incident Neutrons and Protons of Energy Up to 100 MeV. |
| LABAN-PEL | Abstract | C00611 IMFPC 00 | A Two-Dimensional, Multigroup Diffusion, High-Order Response Matrix Code. |
| LADTAP II | Abstract | C00363 C7600 00 | Code System for Calculating Radiation Exposure to Man from Routine Release of Nuclear Reactor Liquid Effluents. |
| LADTAP II | Abstract | C00363 D0780 00 | Code System for Calculating Radiation Exposure to Man from Routine Release of Nuclear Reactor Liquid Effluents. |
| LADTAP II | Abstract | C00363 I3033 00 | Code System for Calculating Radiation Exposure to Man from Routine Release of Nuclear Reactor Liquid Effluents. |
| LAFPX-V | Abstract | D00054 C0000 01 | A Multigroup Reaction Cross-Section Collapsing Code and Library of 154-Group Fission-Product Cross Sections. |
| LAFPX-V | Abstract | D00054 C0000 02 | A Multigroup Reaction Cross-Section Collapsing Code and Library of 154-Group Fission-Product Cross Sections. |
| LAHET 2.8 | Abstract | C00696 MFMWS 00 | Code System for High Energy Particle Transport Calculations. |
| LAHIMACK | Abstract | D00128 I0360 00 | A Multigroup Library of Neutron and Gamma Cross Sections and Response Functions in the Energy Range up to 800 MeV. |
| 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. |
LAS CRUCES
USSO | Abstract | D00194 ALLCP 00 | Las Cruces Trench Site Database, Vadose Model. |
| LASER | Abstract | C00344 I0360 00 | A One-Dimensional, Neutron-Thermalization, Lattice-Cell Program Based on MUFT and THERMOS. |
| LAZY | Abstract | P00595 I0360 00 | General Experimental Data Processing Program. |
| LEAF | Abstract | C00312 C6600 00 | Fission Product Release Calculator-From a Reactor Containment Building for Arbitrary Radioactive Decay Chains. |
| LEAP-ADDELT | Abstract | P00138 I0360 00 | Multigroup Thermal Neutron Scattering Data Generator for Hydrogen in Light Water and Deuterium in Heavy Water. |
| LEBC | Abstract | C00052 I7090 00 | Electron Bremsstrahlung Code. |
| LEGENDRE FUNCTI | Abstract | P00108 I0360 00 | Legendre Functions of the First Kind and Legendre Polynomials. |
| LENDL | Abstract | D00034 I0360 02 | Livermore Evaluated Neutron and Secondary Gamma-Ray Production Cross-Section Library in ENDF/B-IV Format. |
| LENDL V | Abstract | D00120 I0360 00 | Lawrence Livermore National Laboratory Evaluated Nuclear Data Library in ENDF-V Format. |
| LEOPARD | Abstract | C00343 C0000 00 | A Spectrum-Dependent Non-Spatial Fuel Depletion Code System. |
| LEOPARD | Abstract | C00343 IBMPC 00 | A Spectrum-Dependent Non-Spatial Fuel Depletion Code System. |
| LEP | Abstract | D00001 I0360 02 | Cascade and Evaporation Particle Results from Low-Energy Intranuclear Cascade Calculations. |
| LEPRICON | Abstract | P00277 I3033 01 | PWR Pressure Vessel Surveillance Dosimetry Analysis System. |
| LEPRICON | Abstract | P00277 IRISC 00 | PWR Pressure Vessel Surveillance Dosimetry Analysis System. |
| LG-H | Abstract | C00087 I7090 00 | Ray Analysis Cylindrical Duct Kernel Code for Neutrons and Gamma Rays. |
| LGH-G | Abstract | C00239 I0360 00 | Calculation of Gamma Radiation through Partially Shielded Gaps (Buildup Factor Method in Taylors Approximation). |
| 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. |
| LIB123 | Abstract | D00153 ALLCP 00 | AMPX-II P3 123-Group Neutron Cross Section Master Interface Library. |
| LIBMAK | Abstract | P00087 I0360 00 | ANISN-Type Binary Data Processing Code System. |
| LIE-PN | Abstract | C00816 I0360 00 | Pn Neutron Transport in Radial Geometry Cell with Source Problems Calculation. |
| LINEDOSE | Abstract | C00468 IBMPC 00 | A Line Source Shielding Code for Personal Computers. |
| LINSED | Abstract | C00673 I0360 00 | 1D Multireach Sediment Transport Model |
| LIONS | Abstract | C00247 I0360 00 | Calculation of Fission Product Inventory, Gamma-Ray Dose Rates and Gamma-Ray Doses by Kernel Integration. |
| 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. |
| LPGS | Abstract | C00385 I3033 00 | Code System for Calculating Radiation Exposure Resulting from Accidental Radioactive Releases to the Hydrosphere. |
| LPPC | Abstract | C00051 I7090 00 | Proton Penetration Code. |
| LPSC | Abstract | C00064 I7090 00 | Proton Penetration Code - Multilayer Slab Geometry. |
| LPTAU | Abstract | P00340 MNYCP 00 | Quasi-Random Sequence Generators. |
| LRSPC | Abstract | C00050 I7090 00 | Range and Stopping Power Calculator. |
| LSHINSE | Abstract | C00554 IBMPC 00 | Calculates Flux and Dose Rate from the Scattering of Radiation in Air. |
| 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. |
| LSVDC | Abstract | C00053 I7090 00 | Space Vehicle Dose Calculation. |
| LSVDC | Abstract | C00053 I7090 01 | Space Vehicle Dose Calculation. |
| LTC | Abstract | P00329 IBMPC 00 | LMR Transient Calculation Code System. |
| LUIN-II | Abstract | C00220 C6600 00 | Analytical Straight-Ahead Transport Code System-Calculation of Cosmic-Ray Spectra, Fluxes and Ionization in the Earth's Atmosphere. |
| LUMP | Abstract | D00089 I0360 00 | Evaluated Lumped Fission Product Cross Sections for Fast Reactor Analysis--Based on ENDF/B-V Data. |
| MACK-IV | Abstract | P00132 I3691 00 | Calculation of Nuclear Response Functions from Nuclear Data in ENDF Format. |
| MACKLIB | Abstract | D00029 I3675 00 | A Library of Nuclear Response Functions Generated with the MACK-V Computer Program from ENDF/B-IV. |
| MACKLIB-IV-82 | Abstract | D00060 I0360 01 | A Library of Nuclear Response Functions Generated with the MACK-V Computer Program from ENDF/B-IV. |
| MADONNA | Abstract | C00425 I0370 00 | Two-dimensional Neutron Streaming Coupled Removal-Diffusion-Albedo-Transport Code System. |
| MAEROS | Abstract | P00466 C7600 00 | Code System for Multicomponent Aerosol Time Evolution. |
| MAGIK | Abstract | C00359 I0360 00 | A Monte Carlo Code System for Computing Induced Residual Activation Dose Rates. |
| MAGNA | Abstract | C00158 C3600 00 | Multi-Source Gamma-Ray Kernel Integration Code System. |
| 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. |
| MAP | Abstract | C00150 I3675 00 | Kernel Integration Code System in Complex Geometry with Special Application to Surface Sources Determined by Discrete Ordinates Calculations. |
| 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. |
| MARC-PN | Abstract | C00311 D8810 00 | A Neutron Diffusion Code System with Spherical Harmonics Option. |
| 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. |
| MARINRAD | Abstract | C00503 C1785 00 | Code System Model for Assessing the Consequences of Release of Radioactive Material into the Oceans. |
| MARLOWE 15B | Abstract | P00137 MNYCP 08 | Computer Simulation of Atomic Collisions in Crystalline Solids. |
| MARMER | Abstract | C00579 D8350 00 | A Flexible Point-Kernel Shielding Code System. |
| MARMER | Abstract | C00579 PC486 00 | A Flexible Point-Kernel Shielding Code System. |
| MARS | Abstract | P00117 I0360 00 | Collection of Computer Codes for Manipulating Multigroup Cross Section Libraries in AMPX or CCCC Formats. |
| MASS | Abstract | D00025 I0360 01 | Atomic Mass Evaluation. |
| MATADOR | Abstract | C00689 CDCMF 00 | Radionuclide Behavior in Containments. |
| MATEXP | Abstract | P00059 I0360 00 | Matrix Exponential Method Applied to Systems of Ordinary Differential Equations. |
| MATJEFF31.BOLIB | Abstract | D00242 MNYCP 00 | Fine-Group Cross Section Library Based on JEFF3.1 for Nuclear Fission Applications. |
| MATXS1 | Abstract | D00114 C0000 00 | Neutron, Photon Cross Sections from ENDF/B-IV in MATXS Format. |
| MATXS10 | Abstract | D00176 ALLCP 00 | Neutron, Photon Cross Sections from ENDF/B-IV in MATXS Format. |
| MATXS11 | Abstract | D00177 ALLCP 00 | Neutron, Photon Cross Sections from ENDF/B-IV in MATXS Format. |
| MATXS175/42-JE | Abstract | D00151 D8810 00 | JEF/EFF Based VITAMIN-J 175 Neutron, 42 Photon Multigroup Data Library in MATXS Format. |
| MATXS5A | Abstract | D00115 C0000 00 | Neutron, Photon Cross Sections from ENDF/B-IV in MATXS Format. |
| MATXS6A | Abstract | D00116 C0000 00 | Neutron, Photon Cross Sections from ENDF/B-IV in MATXS Format. |
| MATXS70-JEF87 | Abstract | D00148 D8810 00 | JEF/EFF Based 70 Group Neutron Data Library in MATXS Format. |
| MATXS7A | Abstract | D00117 C0000 00 | Neutron, Photon Cross Sections from ENDF/B-IV in MATXS Format. |
| MATXSLIBJ33 | Abstract | D00258 MNYCP 01 | JENDL-3.3 Based, 175 Neutron-42 Photon Groups (VITAMIN-J) MATXS Library for Discrete Ordinates Multi-Group Transport Codes. |
| MATXUF | Abstract | P00130 I0360 00 | On-Line Derivative Method, Spectrum Unfolding Code System for NE-213 Liquid Fast Scintillation Proton Recoil Data. |
| MAVRAC | Abstract | C00023 I7090 00 | Model Astronaut and Vehicle Radiation Analysis Code. |
| 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. |
MCART
USUNV | Abstract | C00809 PCX86 00 | Solve the Time-Dependent Neutron Transport Equation. |
| MCB1C | Abstract | C00719 MNYWS 00 | Monte-Carlo Continuous Energy Burnup Code System. |
| MCB63NEA.BOLIB | Abstract | D00216 MNYCP 00 | ENDF/B-VI Release 3 Cross Section Library for Use with the MCNP Monte Carlo Code. |
| MCFLARE | Abstract | C00093 I7090 00 | Monte Carlo Code to Simulate Solar Flare Events and Estimate Probable Doses Encountered on Interplanetary Missions. |
| MCJEF22NEA.BOLIB | Abstract | D00203 MNYCP 01 | JEF 2.2 Cross Section Library for the MCNP Monte Carlo Code. |
| MCJEFF3.1NEA | Abstract | D00228 MNYCP 00 | Neutron Cross Section Library Based on JEFF3.1 for Use with MCNP. |
MCNP6.3-EXE
810 | Abstract | C00870 MNYCP 01 | Monte Carlo N-Particle Transport Code System. |
MCNP-DSP-EXE
810 | Abstract | C00699 MNYCP 01 | Monte Carlo N-Particle Transport Code System with Digital Signal Processing based on MCNP4A. |
MCNPX-POLIMI-EXE
810 | Abstract | C00791 MNYCP 01 | Monte Carlo N-Particle Transport Code System To Simulate Time-Analysis Quantities. |
| MCRAC | Abstract | C00562 IBMPC 00 | Multiple Cycle Reactor Analysis Code. |
| MCRTOF | Abstract | C00435 FM200 00 | Monte Carlo Code System for Calculation of Multiple Scattering of Neutrons in the Resonance Region. |
| MCRTOF | Abstract | C00435 I0360 00 | Monte Carlo Code System for Calculation of Multiple Scattering of Neutrons in the Resonance Region. |
MCUNED
810 | Abstract | C00804 PCX86 00 | MCNPX Extension for Using Light Ion Evaluated Nuclear Data Library. |
| MCVIEW | Abstract | P00202 FM780 00 | View Factor Calculation for Three-Dimensional Geometries. |
| MECC-7 | Abstract | C00156 I0360 00 | Medium-Energy Intranuclear Cascade Code System. |
| MEDUSA-PIJ | Abstract | C00349 F2307 00 | One-Dimensional Lagrangian Code for Plasma Hydrodynamic Analysis of a Fusion Pellet Driven by Ion Beams. |
| MEGA | Abstract | C00839 MNYCP 00 | MEGA: Mechanistic and Engineering Fission Gas Release Prediction Model for UO2 Fuel |
| MENDL-2P | Abstract | D00207 MNYCP 00 | Proton Reaction Data Library for Nuclear Activation (Medium Energy Nuclear Data Library.) |
| MENSLIB | Abstract | D00084 I0370 00 | 60 Group, P5, Cross Sections in DTF-IV for Transport Calculations for Neutrons with Energies Up to 60 MeV. |
| MERCURE 4-82 | Abstract | C00142 I3033 00 | Three-Dimensional Code System for Integrating Multigroup Line-of-Sight Attenuation Kernels by Monte Carlo Techniques. |
| MESA | Abstract | P00223 I3033 00 | Non-Linear Least Squares Spectral Analysis. |
| MESODIF-II | Abstract | C00498 D0780 00 | A Variable Trajectory Plume Segment Model to Assess Ground-Level Air Concentrations and Depositions of Routine Effluent Releases from Nuclear Power Facilities. |
| MESOI | Abstract | C00497 D0780 00 | Interactive Mesoscale Lagrangian Puff Dispersion Model with Deposition and Decay. |
| MESORAD 1.4 | Abstract | C00677 D0VAX 00 | Code System for Emergency Response Dose Assessment. |
| MESYST | Abstract | C00706 MNYWS 00 | Code System to Simulate 3D Tracer Dispersion in Atmosphere. |
| 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. |
| MEVDP | Abstract | C00157 C6600 00 | Primary Radiation Transport Code System - Complex Geometry - Computerized Anatomical Model Man. |
| MGA8 | Abstract | P00542 MNYCP 00 | Code System to Determine Pu Isotope Abundances from Multichannel Analyzer Gamma Spectra. |
| MGCLIB | Abstract | D00118 FM380 00 | 137 and 26 Neutron Multigroup Cross Section Library with the Bondarenko Type Shielding Table. |
| 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. |
| MILDOS | Abstract | C00398 C0000 00 | Calculation of Radiation Doses from Uranium Recovery Operations. |
| MILDOS-AREA | Abstract | C00608 IBMPC 00 | Calculation of Radiation Dose from Uranium Recovery Operations for Large-Area Sources. |
| 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. |
| MKENO-DAR | Abstract | C00513 FM380 00 | Direct Angular Representation Monte Carlo Code for Criticality Safety Analysis |
| MMCR | Abstract | C00441 FM200 00 | Multigroup Monte Carlo Neutron and Photon Transport Code. |
| MMRW | Abstract | M00018 MNYCP 00 | Canadian and Early British Energy Reports on Nuclear Reactor Theory (1940-1946). |
| MMRW-BOOKS | Abstract | M00020 MNYCP 00 | MMRW-BOOKS: Legacy books on slowing down, thermalization, particle transport theory, random processes in reactors. |
| MMS3D | Abstract | C00841 MNYCP 00 | Method of Manufactured Solutions for 3D one-group SN Equations with escalating order of non-smoothness |
| MOCA | Abstract | C00590 IPCAT 00 | Monte Carlo Criticality Code System for Hexagonal Geometries. |
| MOCUP | Abstract | P00365 DALPU 00 | MCNP/ORIGEN Coupling Utility Programs. |
| MODEL | Abstract | C00329 I3033 00 | Models of Trapped Proton and Electron Environments for Solar Maximum and Minimum. |
| MOMENT I | Abstract | C00188 U1108 00 | Moments Method Neutron Transport Code System. |
| MOMGEM-MOMDIS | Abstract | C00085 I7090 00 | Moments Method Reconstruction of Scattered Gamma-Ray Distributions. |
MONK 6.3
FEDC | Abstract | C00393 I3033 00 | A General Purpose Monte Carlo Neutronics Code System. |
| MONTEBURNS 2.0 | Abstract | P00455 MNYCP 02 | Automated, Multi-Step Monte Carlo Burnup Code System. |
| MONTUK-80 | Abstract | D00072 ALLCP 01 | UKCTR III Transmutation and Activation Data, 100-Group Neutron Activation Cross-Section Data for Fusion Reactor Structure and Coolant Materials. |
| 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. |
| MORSE-ALB | Abstract | C00394 FM200 00 | Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System. |
| MORSE-ANSI STD. | Abstract | C00127 I3675 00 | Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System. |
| MORSE-B | Abstract | C00368 I0370 00 | Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System. |
| MORSE-C | Abstract | C00431 C7600 00 | Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System. |
| MORSE-CG | Abstract | C00203 C0000 00 | Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System. |
| MORSE-CG | Abstract | C00203 CY000 00 | Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System. |
| MORSE-CG | Abstract | C00203 D0VAX 00 | Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System. |
| MORSE-CG | Abstract | C00203 I0360 00 | Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System. |
| MORSE-CG | Abstract | C00203 U0000 00 | Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System. |
| MORSE-CGA | Abstract | C00474 ALLCP 03 | Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System. |
| MORSEC-SP2 | Abstract | P00142 H6000 00 | Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System. |
| MORSE-CV | Abstract | C00535 HM280 00 | Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System. |
| MORSE-E | Abstract | C00258 I0360 00 | Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System. |
| MORSE-EMP | Abstract | C00588 IBMPC 00 | Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System. |
| MORSE-H | Abstract | C00471 I3081 00 | Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System. |
| MORSE-L | Abstract | C00261 C6600 00 | Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System. |
| MORSE-SGC | Abstract | C00277 C7600 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. |
MRIPP 1.0
810 | Abstract | C00655 PC386 00 | Magnetic Resonance Image Phantom Code System to Calibrate in vivo Measurement Systems. |
| 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. |
| MTR_PC 2.6 | Abstract | C00674 PC386 00 | Modular Code System for Neutronics, Thermalhydraulics and Shielding Calculations. |
| MULTI-KENO2 | Abstract | C00492 FM380 00 | A Monte Carlo Code System for Criticality Safety Analysis. |
| MUP2 | Abstract | P00289 I3090 00 | A Program to Calculate Fast Neutron Data for Medium-Heavy Nuclei. |
| MURE V2-SMURE | Abstract | C00764 MNYWS 01 | Serpent - MCNP Utility for Reactor Evolution. |
| MURLI | Abstract | C00378 DP011 00 | Integral Transport Theory Code System for Thermal Reactor Lattice Cell Calculation. |
| MUSCAT | Abstract | C00281 I0360 00 | Calculation of Neutron Currents in Spherical and Cylindrical Cavities by Means of View Factors. |
| MUSPALB | Abstract | C00171 ICL00 00 | Albedo Calculation of Multigroup Spectra of Neutrons Transmitted Through Multilayer Slab Shielding. |
| MUXS | Abstract | P00187 I3033 00 | Generator of Multigroup Cross Sections for Charged Particle Transport Problems. |
| MVP-GMVP II | Abstract | C00739 MNYCP 00 | General Purpose Monte Carlo Codes for Neutron and Photon Transport Calculations based on Continuous Energy and Multigroup Methods. |
| MYRA | Abstract | C00056 C0000 00 | Calculation of Shipping Costs and Cask Designs for Irradiated Fuel Elements. |
| MYRA | Abstract | C00056 I7090 00 | Calculation of Shipping Costs and Cask Designs for Irradiated Fuel Elements. |
| NAAPRO | Abstract | C00722 PC586 00 | Neutron Activation Analysis PRognosis and Optimization Code System. |
| NAB | Abstract | D00018 I0360 00 | 100-Group, P3, Neutron Cross Section Data for Sodium and Aluminum. |
| NAC | Abstract | C00164 C0000 00 | Neutron Activation Analysis and Product Isotope Inventory Code System. |
| NAC | Abstract | C00164 IBMMF 00 | Neutron Activation Analysis and Product Isotope Inventory Code System. |
| NAC-PC | Abstract | C00164 IBMPC 00 | Neutron Activation Analysis and Product Isotope Inventory Code System. |
| NACT | Abstract | C00502 U1100 00 | Screening Program for Neutron Activation Products. |
| 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. |
| NAP | Abstract | C00101 I7090 00 | Multigroup Time-Dependent Neutron Activation Prediction Code. |
| 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. |
| NCRP49 | Abstract | C00462 IBMPC 00 | X-Ray Shield Calculation System. |
| NCSP-DAT | Abstract | M00002 MNYCP 01 | Nuclear Data in Support of the Nuclear Criticality Safety Program. |
| NEACRP-H2O-LATTICES | Abstract | D00265 MNYCP 00 | Compilation of Reactor Physics Measurements in LWRs Lattices. |
| NESTLE 5.2.1 | Abstract | C00641 MNYCP 04 | Code System to Solve the Few-Group Neutron Diffusion Equation Utilizing the Nodal Expansion Method (NEM) for Eigenvalue, Adjoint, and Fixed-Source |
| 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. |
| NITRAN | Abstract | C00582 FM380 00 | Neutron Transport Code System Based On Anisotropic Scattering. |
| 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. |
| NMTC/JAERI97 | Abstract | C00694 SUN05 00 | Monte Carlo Nucleon Meson Transport Code System. |
| NMTC/JAM | Abstract | C00717 PC586 00 | High Energy Particle Transport Code 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. |
| NOX | Abstract | D00017 I0360 00 | 199-Group, P5, Coupled Neutron and Secondary Gamma-Ray Cross Section Data for Nitrogen and Oxygen. |
| NPCSL-81 | Abstract | D00082 I0370 00 | Point Neutron Cross Sections Generated from ENDF/B-IV with the NPTXS Modules of PSR-63/AMPX-II. |
| NPTXS | Abstract | P00090 I0360 00 | Data Generator: Neutron Point Cross Sections from ENDF/B Resolved and Unresolved Resonance Parameters. |
| NRCDOSE 2.3.20 | Abstract | C00684 PC586 14 | Code System for Evaluating Routine Radioactive Effluents from Nuclear Power Plants with a Windows Interface. |
| NRCDOSE72V1.2.3 | Abstract | C00768 PCX86 03 | Code System for Evaluating Routine Radioactive Effluents from Nuclear Power Plants with a Windows Interface. |
| 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. |
| NRN | Abstract | C00054 C6600 00 | Multigroup Removal-Diffusion Code System for Planes, Cylinders and Spheres. |
| NSLINK | Abstract | P00314 D0VAX 00 | NJOY SCALE LINK. |
| NUCCON | Abstract | C00439 S7800 00 | A Code System for Calculation of Time-Dependent Nuclide Concentrations, Activity, Gamma-Ray Dose Rate and Biological Hazard Potential of Fusion Reactor Materials Due to Neutron Irradiation. |
| NUCDECAY | Abstract | D00172 PC386 01 | Nuclear Decay Data for Radiation Dosimetry Calculations for ICRP and MIRD. |
| NUCDECAYCALC | Abstract | D00202 PC586 00 | Nuclear Decay Data for Radiation Dosimetry Calculations for ICRP. |
| 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. |
| NUGAM 2&3 SSLAB | Abstract | C00210 I0360 00 | Monte Carlo Prediction of Photon Transport Distributions. |
| NUTRAN | Abstract | C00675 I0370 00 | Code System for Long-Term Repository Safety Analysis. |
| NX1-NX2 | Abstract | P00310 D0VAX 00 | Code System to Calculate Excitation Functions for (n,charged particle) Reactions. |
| O5R | Abstract | C00017 I3675 00 | A General-Purpose Monte Carlo Neutron Transport Code System. |
| 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. |
| O6R | Abstract | C00128 I3675 00 | A General-Purpose Monte Carlo Transport Code System. |
| 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. |
| OGRE | Abstract | C00046 I3675 00 | A General-Purpose Monte Carlo Gamma-Ray Transport Code System. |
| OGRE-MIN | Abstract | C00409 DGECL 00 | A General-Purpose Monte Carlo Gamma-Ray Transport Code System for Minicomputers. |
| OMCOST | Abstract | P00381 I3033 00 | Code System for Non-fuel O & M Cost Estimation for Large Steam-Electric Power Plants. |
| OMEGA | Abstract | C00433 BESM6 00 | Monte Carlo Criticality Code System. |
| ONETRAN | Abstract | C00266 C7600 00 | A One-Dimensional Multigroup Discrete Ordinates Finite Element Transport Code System. |
| ONETRAN | Abstract | C00266 CY000 00 | A One-Dimensional Multigroup Discrete Ordinates Finite Element Transport Code System. |
| ONETRAN | Abstract | C00266 I3033 00 | A One-Dimensional Multigroup Discrete Ordinates Finite Element Transport Code System. |
| OOSII | Abstract | C00324 C0000 00 | Calculation of Isotropic Scattering by Particles for One-Dimensional and Three-Dimensional Transport in Slabs by Invariant Imbedding, Orders-of-Scattering Method, Including Check Calculations by Integral Transport Theory and Monte Carlo. |
| OPERATIONAL MONTE CARLO GUI | Abstract | P00619 PCX86 01 | Operational Monte Carlo GUI (OMG) |
| OPEX-II | Abstract | C00103 I7090 00 | Radiation Shield Optimization Code. |
| OPTIM | Abstract | C00817 I0370 00 | Minimization of Band-Width of Finite Elements Problems. |
| ORCENT-2 | Abstract | P00474 I3033 00 | Code System for Analysis of Steam Turbine Cycles Supplied by Light Water Reactors. |
| ORESUND | Abstract | D00267 MNYCP 00 | Nordic Mesoscale Dispersion Experiments over Land-Water-Land. |
| ORIGEN2.2 | Abstract | C00371 ALLCP 03 | Isotope Generation and Depletion Code - Matrix Exponential Method. |
| ORIGEN-JENDL32 | Abstract | C00703 MNYWS 00 | Isotope Generation and Depletion Code - Matrix Exponential Method. |
ORINC
USSO | Abstract | P00439 I0360 00 | Code System for 1-D Implicit Heat Conduction Solution. |
| ORION-II | Abstract | C00491 FM780 00 | A Computer Code to Estimate Environmental Concentration and Dose Due to Airborne Release of Radioactive Material. |
| ORIP_XXI | Abstract | C00731 PC586 02 | Computer Programs for Isotope Transmutation Simulations. |
| ORLIBJ32 | Abstract | D00255 MNYCP 00 | ORIGEN2 Libraries Based on JENDL-3.2. |
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. |
| ORPHEE VI | Abstract | C00159 I3675 00 | Kernel Integration Code System - Attenuation of Fast Neutrons in Cylindrical Layers of Water and Dense Material. |
| 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. |
| ORYX-E | Abstract | D00038 I0360 00 | ORIGEN Yields and Cross Sections Nuclear Transmutation and Decay Data from ENDF/B-IV. |
| ORYX-E | Abstract | D00038 I0360 01 | ORIGEN Yields and Cross Sections Nuclear Transmutation and Decay Data from ENDF/B-IV. |
| OZMA | Abstract | C00406 I0370 00 | Calculation of Resonance Reaction Rates in Reactor Lattices Using Resonance Profile Tabulations. |
| PABLM | Abstract | C00402 U1100 00 | Calculation of Accumulated Radiation Doses to Man from Radionuclides Found in Food Products and from Radionuclides in the Environment. |
| PADF-2007 | Abstract | D00259 PCX86 00 | Proton Activation Data File in ENDF-6 Format. |
| PADLOC | Abstract | C00330 U0000 00 | A One-Dimensional, Time-Dependent Program for Calculating Coolant and Plateout Fission Product Concentrations in a Network of Pipes. |
| PAGAN | Abstract | C00621 IBMPC 00 | Code System for Performance Assessment Ground-water Analysis for Low-level Nuclear Waste. |
| PALLAS-1D(VII) | Abstract | C00380 FM380 00 | Multigroup Time-Independent Neutron Transport Code System for Plane or Spherical Geometry. |
| PALLAS-2DCY-FX | Abstract | C00391 FM380 00 | Multigroup Time-Independent Neutron Transport Code System for Plane or Spherical Geometry. |
| 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. |
| PART61 | Abstract | C00499 IBMPC 01 | Low-Level Radioactive Waste Impacts Analysis System. |
| PARTISN 8.29 | Abstract | C00842 MNYCP 00 | Time-Dependent, Parallel Neutral Particle Transport Code System. |
| PATCH-7 | Abstract | C00243 C0074 00 | Three-Dimensional Kernel Integration Code-Explicit Single Scattering Option. |
| PAVAN | Abstract | C00445 I3033 00 | Atmospheric Dispersion Code System for Evaluating Accidental Radioactivity Releases from Nuclear Power Stations. |
| 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. |
| PEFPYD | Abstract | D00096 ALLMF 02 | Aggregate Fission-Product Decay Data Based on ENDF/B-IV and -V. |
| 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. |
| PELSHIE | Abstract | C00202 C0000 00 | General Purpose Kernel Integration Shielding Code System-Point and Extended Gamma-Ray Sources. |
| PELSHIE3 | Abstract | C00202 IBMMF 00 | General Purpose Kernel Integration Shielding Code System-Point and Extended Gamma-Ray Sources. |
| PENELOPE2014 | Abstract | C00782 PCX86 01 | Code System for Monte Carlo Simulation of Electron and Photon Transport. |
| PENELOPE-MPI | Abstract | C00713 IBMSP 00 | Code System for Monte Carlo Simulation of Electron and Photon Transport. |
| PENGEOM | Abstract | C00840 MNYCP 00 | Tools for Handling Complex Quadric Geometries in Monte Carlo Simulations of Radiation Transport |
| PEPIN | Abstract | C00285 I0360 00 | Methodology for Computing Concentrations, Activities, Gamma-Ray Spectra, and Residual Heat from Fission Products. |
| PEQAG-2 | Abstract | P00293 IPCAT 00 | A Pre-equilibrium Computer Code With Gamma Emission. |
PERSENT 11.2892
FEDC | Abstract | C00823 MNYCP 02 | Perturbation and Sensitivity Code for Assembly Homogenized Multi-group Transport Problems |
| PF-COMP | Abstract | C00106 C3600 00 | Building Fallout Radiation Protection Factor Analysis. |
| PFPL | Abstract | C00607 D0VAX 00 | Puff-Plume Atmospheric Deposition Model. |
| PGAA-IAEA | Abstract | D00234 MNYCP 00 | Databsae for Prompt Gamma-Ray Neutron Activation Analysis. |
PHAZE
USSO | Abstract | P00432 IBMPC 00 | Parametric Hazard Function Estimation. |
| PHITS-2.88 | Abstract | C00778 MNYCP 05 | Particle and Heavy Ion Transport code System. |
| PHOBIA | Abstract | D00236 PCX86 00 | Photon buildup factors to account for angular incidence on shield walls. |
| PHOEL-2 | Abstract | C00327 I0360 00 | A Monte Carlo Calculation of Initial Energy of Photoelectrons and Compton Electrons Produced by Photons in Water. |
| PHOTX | Abstract | D00136 D0VAX 01 | Photon Interaction Cross Section Library. |
| PHOTX | Abstract | D00136 IBMPC 00 | Photon Interaction Cross Section Library. |
| PICA | Abstract | C00160 D0VAX 00 | Monte Carlo Medium-Energy Photon-Induced Intranuclear Cascade Anal Code System. |
| PICA | Abstract | C00160 I0360 00 | Monte Carlo Medium-Energy Photon-Induced Intranuclear Cascade Anal Code System. |
| PICES | Abstract | P00568 I3033 00 | Probabilistic Investigation of Capacity and Energy Shortages. |
| PICFEE | Abstract | C00175 I3675 00 | Fission Product Inventory Code System. |
| PICTURE | Abstract | P00238 IBMPC 00 | Combinatorial Geometry Printer Plotting. |
| PIEDEC | Abstract | C00566 FM380 00 | A Practical Internal Exposure Dose Evaluation Code. |
| PIGG | Abstract | C00138 C3600 00 | A Multigroup One-Dimensional P-1 Radiation Transport Code System. |
| PIPE | Abstract | C00219 I0360 00 | Numerical Gamma-Ray Transport Code System for Plane/Spherical Geometry. |
| PIXE2010 | Abstract | D00246 MNYCP 00 | Proton/alpha Ionization (K, L, M shell), Tabulated Cross Section Library. |
| PIXSE | Abstract | P00133 I0360 00 | A Generator of Multigroup and Multipoint Cross Sections for Thermal Reactor Calculations. |
| PKI | Abstract | C00573 C0830 00 | A Point Kernel Integration Code For Radiation Shielding of Loop System. |
| PLACID | Abstract | C00381 I0370 00 | Monte Carlo Simulation of Gamma Streaming Through Straight Cylindrical Ducts. |
| 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. |
| PLUDOS | Abstract | C00313 I0360 00 | Calculator of Ground Level External Gamma-Ray Dose from a Radioactive Plume. |
| PLUMEX | Abstract | C00356 I0360 00 | A Computer Program to Evaluate External Exposures to a Gaussian Plume by Point Kernel Integration. |
| PMK2-VVER440-REPORTS | Abstract | M00012 MNYCP 00 | Results of the Experiments Performed in the PMK-2 Facility for VVER Safety Studies. |
| PN | Abstract | C00818 I0370 00 | MultiGroup Neutron Transport. |
| PNESD | Abstract | D00166 PC386 00 | Proton Nucleus Elastic Scattering Data. |
| POINT2015 | Abstract | D00273 MNYCP 00 | A Temperature-Dependent Linearly Interpolable, Tabulated Cross Section Library Based on ENDF/B |
| POISSX | Abstract | C00819 I0370 00 | Poisson Equation on Rectangle with Various Boundary Conditions. |
| 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. |
| POPLIB | Abstract | D00012 I0360 03 | A Compendium of Neutron-Induced Secondary Gamma-Ray Yield and Cross Section Data. |
| 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. |
| PR-EDB | Abstract | D00196 IBMPC 03 | Power Reactor Embrittlement Data Base. |
| 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. |
| PREMOR | Abstract | C00369 I0360 00 | A Point Reactor Exposure Code System for Survey Nuclear Analysis of Power Plant Performance. |
| PREP/SPOP | Abstract | C00772 MNYCP 00 | Uncertainty and Sensitivity Analysis Monte Carlo Program and Input Preparation. |
| PREPRO2019 | Abstract | P00351 MNYCP 10 | Pre-Processing Code System for Data in ENDF/B Format. |
| PREST | Abstract | C00355 I0360 00 | Calculator of Pressure and Temperature Transient in Containment Studies. |
| PRESTO | Abstract | C00549 D8810 00 | Point Kernel Calculation for Complex and Time-Dependent Gamma-Ray Source Spectra. |
| PRESTO-II | Abstract | C00504 I0360 00 | Code System for Low-Level Waste Environmental Transport and Risk Assessment. |
| PRIMEDANA-2 | Abstract | C00490 I3081 00 | Collapses Multigroup Cross Sections and Obtains Reaction Parameters by Solving Transport or Diffusion Equations. |
| PRISIM | Abstract | C00574 IBMPC 00 | Plant Risk Status Information Management System. |
| PROB | Abstract | C00287 I0370 00 | Multigroup One-Dimensional Transport Code System, Collision Probability Method. |
| PROCIV | Abstract | C00488 U1110 00 | A Code System for Calculating the Protection Factors Against Radioactive Fallout for Apartment Buildings. |
| 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. |
| PSU-LEOPARD/RBI | Abstract | C00563 IBMPC 01 | A Spectrum Dependent Non-Spatial Depletion Code. |
| PTRAN | Abstract | C00618 PC386 00 | Proton Monte Carlo Transport Program for the PC. |
| PUCOR | Abstract | D00067 I3691 00 | 84 Group Neutron Cross Sections for Uranium-Plutonium Cycle LWR and PWR Models in AMPX Master Library Format. |
| PUDK | Abstract | D00074 I0360 00 | Measured Results of Delayed Beta- and Gamma-Ray Spectra Due to Thermal-Neutron Fission of Pu239 and Pu241. |
| PUFF-IV | Abstract | P00534 MNYCP 01 | Determination of Multigroup Covariance Matrices from ENDF/B-V Uncertainty Files. |
| PURSE | Abstract | C00338 C6600 00 | A Plutonium Radiation Source Code System. |
| PUSHLD | Abstract | C00271 C0074 00 | Gamma-Ray Three-Dimensional Calculation of Dose Rates from Plutonium in Various Geometries. |
| PUTZ 2.1 | Abstract | C00595 IBMPC 00 | A Point-Kernel Photon Shielding Code. |
| PVC | Abstract | D00048 I3691 00 | 36 Group, P5, Photon Interaction Cross Sections for 38 Materials in ANISN Format. |
| PVE | Abstract | D00126 I3033 00 | 38 Group, P8, Photon Interaction Cross Sections in ANISN Format from VITAMIN-E. |
| PVIS-4 | Abstract | C00798 MNYCP 00 | Pressure Vessel Irradiation Source. |
| PWR-AXBUPRO-GKN | Abstract | D00209 MNYCP 00 | Measured Axial Burnup Profiles for NeckarWesthiem PWR Reactors. |
| PWR-AXBUPRO-SNL | Abstract | D00201 MNYCP 00 | Axial Burnup Profile Database for Pressurized Water Reactors. |
| Q&A | Abstract | P00428 IBMPC 00 | Questions and Answers Based on Revised 10 CFR Part 20 |
| QAD | Abstract | C00048 I0360 00 | Kernel Integration Code System. |
| QAD-BSA | Abstract | C00346 C0000 00 | Kernel Integration Code System. |
| QAD-CGGP-A | Abstract | C00645 MNYCP 00 | Kernel Integration Code System. |
| QADMOD-G | Abstract | C00396 I3033 00 | Point Kernel Gamma-Ray Shielding Code With Geometric Progression Buildup Factors. |
| QADMOD-GP | Abstract | C00565 IBMPC 00 | Point Kernel Gamma-Ray Shielding Code With Geometric Progression Buildup Factors. |
| QAD-P5 | Abstract | C00048 C6400 00 | Kernel Integration Code System. |
| QAD-QC | Abstract | C00401 C0000 00 | Kernel Integration Code System. |
| QAD-QC | Abstract | C00401 I0360 00 | Kernel Integration Code System. |
| QAD-UE | Abstract | C00448 H6000 00 | Kernel Integration Code System. |
| QBF | Abstract | C00617 PC386 00 | Code System to Calculate Radiation Dose Rates Relative to Spent Fuel Shipping Casks. |
| QBSHIELD | Abstract | C00599 IBMPC 00 | Spherical Shield Design for Gamma-Ray Sources Using the Buildup Factor Method. |
| QUARK | Abstract | P00492 PC586 00 | Code System for 2-Group, 3D Neutronic Kinetics Calculations Coupled to Core Thermal Hydraulics. |
| QUINCE-PC | Abstract | C00556 IBMPC 00 | Calculates Absorbed Dose From Skin Contamination. |
| RABFIN PARTS | Abstract | C00668 IBMPC 00 | Code System for Calculating Gaseous Effluent Dose Parameters. |
| RACC | Abstract | C00388 CY000 00 | A Code System for Computing Radioactivity-Related Parameters for Fusion Reactor Systems. |
| RACC | Abstract | C00388 I3033 00 | A Code System for Computing Radioactivity-Related Parameters for Fusion Reactor Systems. |
| RACC-PULSE | Abstract | C00639 MNYWS 00 | RACC Code System for Computing Radioactivity-Related Parameters for Fusion Reactor Systems Modified for Pulsed/Intermittent Activation Analysis. |
| RACER | Abstract | C00174 U1108 00 | Calculation of Potential External Dose from Airborne Fission Products Following Postulated Reactor Accident. |
| RAD 2 | Abstract | C00122 I7090 00 | Fission Product Radioactivities Calculation. |
| RADAC | Abstract | C00627 PC486 02 | Code System for Calculating Radioactive Decay and Accumulation of Decayed Products Using Integer-Array Arithmetic for Precise Evaluation of the Bateman Equations. |
| 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. |
| RADHEAT-V4 | Abstract | C00300 FM380 00 | A Code System To Generate Multigroup Constants and Analyze Radiation Transport for Shielding Safety Evaluation. |
| RADOS | Abstract | C00088 I3675 00 | Gamma-Ray Dose Estimation from Cloud of Radioactive Gases by Kernel Integration. |
| RADRISK | Abstract | C00422 DGMV1 00 | Estimates Radiation Doses and Health Effects from Inhalation or Ingestion of Radionuclides. |
| RADRISK | Abstract | C00422 I3033 00 | Estimates Radiation Doses and Health Effects from Inhalation or Ingestion of Radionuclides. |
| RADSHIP-2 | Abstract | C00523 FM200 00 | Code System To Analyze Radiological Impact From Radwaste Transportation. |
| RADSYS | Abstract | C00530 I3033 00 | Code System for Radioactivity Buildup and Radioactive Waste Generation Calculations. |
| RADTRAD 3.03 | Abstract | C00800 IBMPC 00 | A Simplified Model for RADionuclide Transport and Removal And Dose Estimation. |
| RADTRAD 3.03-EXE | Abstract | C00800 IBMPC 01 | A Simplified Model for RADionuclide Transport and Removal And Dose Estimation. |
| RAFFLE/2 | Abstract | C00279 C0176 00 | A General Purpose Monte Carlo Code System for Neutron Transport with Mixed Zone Geometry Option. |
| RAFFLE/2 MOD 2 | Abstract | C00279 I0360 00 | A General Purpose Monte Carlo Code System for Neutron Transport with Mixed Zone Geometry Option. |
| RAID | Abstract | C00083 I7090 00 | Monte Carlo Multibend Duct Shielding Code. |
| RANCHMD | Abstract | C00589 D8810 00 | Radionuclide Chain Transport with Matrix Diffusion. |
| RAPID | Abstract | C00797 PCX86 00 | RAdial Power and Burnup Prediction by Following Fissile Isotope Distribution in the Pellet. |
| RASC-2D | Abstract | C00318 I0370 00 | Two-Dimensional Removal Diffusion Code Reactor Shielding Design Code System. |
| RASCAL 4.3 | Abstract | C00783 PCX86 02 | Radiological Assessment for Consequence Analysis. |
| RASPA | Abstract | C00352 C7600 00 | A Code for the Calculation of Buildup and Decay of Fission Products and Actinides. |
| RATAF | Abstract | C00681 IMFPC 01 | Code System for the Radioactive Liquid Tank Failure Study. |
| RBD | Abstract | C00632 IBMPC 00 | U.S. Army Radiological Bioassay and Dosimetry. |
| 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. |
| REAC*3 | Abstract | C00443 IBMPC 00 | Computer Code System for Activation and Transmutation. |
| REAC*3 | Abstract | C00443 MFMWS 00 | Computer Code System for Activation and Transmutation. |
| 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. |
| REACTORSHIELDING-NMS | Abstract | M00014 MNYCP 00 | REACTORSHIELDING-NMS, Reactor Shielding for Nuclear Engineers. |
| REBEL 3 | Abstract | C00299 I0360 00 | Adjoint Monte Carlo Calculation of Radiation Doses to Human Organs in Dwelling Rooms. |
| REBEL-2 | Abstract | C00299 ICL00 00 | Adjoint Monte Carlo Calculation of Radiation Doses to Human Organs in Dwelling Rooms. |
| REBEL-2 | Abstract | C00299 C6600 00 | Adjoint Monte Carlo Calculation of Radiation Doses to Human Organs in Dwelling Rooms. |
REBUS 11.0 EXE_ONLY
FEDC | Abstract | C00822 MNYWS 01 | Code System for Analysis of Fast Reactor Fuel Cycles. |
REBUS 11.2892
FEDC | Abstract | C00822 MNYCP 02 | Code System for Analysis of Fast Reactor Fuel Cycles. |
| REBUS3/VARIANT8 | Abstract | C00653 MNYWS 01 | Code System for Analysis of Fast Reactor Fuel Cycles. |
| REBUS-PC 1.4 | Abstract | C00708 PC586 00 | Code System for Analysis of Research Reactor Fuel Cycles. |
| RECAP | Abstract | P00414 IBMPC 00 | Replacement Energy Cost Analysis Package. |
| RECAP | Abstract | P00414 IBMPC 01 | Replacement Energy Cost Analysis Package. |
| RECOIL | Abstract | D00055 I3033 01 | Multigroup Primary Recoil Spectra, Displacement Rates and Gas-Production Rates for Radiation Damage Studies. |
| REDIFFUSION | Abstract | C00347 I0360 00 | One-Dimensional Neutron Removal-Diffusion and Gamma-Ray Kernel Integration or Diffusion Theory Calculator. |
| 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. |
| REFIT-2009 | Abstract | C00775 PCX86 00 | Multilevel Resonance Parameter Least Square Fit of Neutron Transmission, Capture, Fission & Self Indication Data. |
| REFLUX | Abstract | P00403 I3033 00 | Code System to Predict LWR Reflood Heat Transfer. |
| REFREP | Abstract | C00570 D8810 00 | A Near-Field Model For A Spent Fuel Repository. |
| 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. |
| REPRISK PC 1.02 | Abstract | C00586 PC386 01 | Repository Risk Assessment Software for Personal Computers. |
| 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. |
| REST 1;2;3 | Abstract | C00225 I0360 00 | Fission Product Inventory Code System with Fission Product Escape Model. |
| RETRAC | Abstract | C00635 D0VAX 00 | Code System for the Analysis of Material Test Reactor (MTR) Cores. |
| RETRANS | Abstract | C00669 SUN05 00 | Code System For Calculating Reactivity Transients In a LWR. |
| 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. |
| RHEIN | Abstract | C00585 I3090 00 | Reactor Code System for Neutron Physics Calculation. |
| RIBD-II | Abstract | C00137 C6600 00 | Radioisotope Buildup and Decay Code System. |
| RIBD-II | Abstract | C00137 I0360 00 | Radioisotope Buildup and Decay Code System. |
| RIBD-IRT | Abstract | C00382 U1100 00 | Radioisotope Buildup and Decay Code System. |
| RICANT | Abstract | C00569 D8810 00 | A Computer Code for 2-D Transport Calculations in x-y Geometry Using the Interface Current Method. |
| RICE | Abstract | P00022 I0360 00 | A Program to Calculate Primary Recoil Atom Spectra from ENDF/B Data. |
| RICECCC | Abstract | C00348 I0360 00 | A Reactor Nuclide Inventory Code for Calculating Actinides and Fission Products. |
| RIPPLE | Abstract | P00571 CYXMP 00 | A Computer Program for Incompressible Fluid Dynamics with Free Surfaces. |
| RISKAP | Abstract | C00486 I3033 00 | Analysis of Increased Risk to Arbitrary Populations. |
RISKIND 2.0
FEDC | Abstract | C00623 IBMPC 02 | Radiological Risk Assessment Code System for Spent Nuclear Fuel Transportation. |
| RITTS | Abstract | D00011 I0360 00 | 121-Group Coupled Neutron and Gamma-Ray Cross-Section Data for Transport Codes. |
| RIVER-RAD | Abstract | C00626 MNYCP 00 | Code System for Simulating the Transport of Radionuclides in Rivers. |
| RMET21 | Abstract | C00597 D0VAX 00 | Detailed Space and Energy Treatment of Neutron Resonances for Homogeneous Mixtures and Cylinderized Reactor Cells. |
| RNGP | Abstract | P00066 I3675 00 | Random Number Generator Package. |
| ROCKWELL-RSDM | Abstract | M00017 MNYCP 00 | Reactor Shielding Design Manual by Rockwell T. III. |
| ROLAIDS-CPM | Abstract | P00353 SUN04 00 | Code System to Calculate Group-Averaged Cross Sections Using the Collision Probability Method. |
| RRR | Abstract | C00196 I0360 00 | Radiation Transport in Air-Analysis of Routine Releases of Short-Lived Radioactive Nuclides. |
| RSAC6.2 | Abstract | C00125 PC586 03 | Radiological Safety Analysis. |
| RSAC-7.2 | Abstract | C00761 PC586 01 | Radiological Safety Analysis. |
| RSYST | Abstract | C00269 I0360 00 | Integrated Modular Code System for Shielding and Reactor Physics Calculations. |
| S1CALC | Abstract | P00134 I0360 00 | A Multigroup Thermal Neutron Scattering Law Data Generator for Hydrogen and Deuterium. |
| S3 | Abstract | C00322 C6600 00 | Kernel Integration Code System--Multigroup Gamma-Ray Scattering. |
| S3 | Abstract | C00322 DVX11 00 | Kernel Integration Code System--Multigroup Gamma-Ray Scattering. |
| S3 | Abstract | C00322 IBMPC 00 | Kernel Integration Code System--Multigroup Gamma-Ray Scattering. |
| SABINE-3 | Abstract | C00121 C7600 00 | Spinney (Removal-Diffusion) Shielding Code System in One-Dimensional Geometry. |
| SABINE-3 | Abstract | C00121 I0370 00 | Spinney (Removal-Diffusion) Shielding Code System in One-Dimensional Geometry. |
| SABINE-3 | Abstract | C00121 U1106 00 | Spinney (Removal-Diffusion) Shielding Code System in One-Dimensional Geometry. |
| SABINE-PC | Abstract | C00121 IBMPC 00 | Spinney (Removal-Diffusion) Shielding Code System in One-Dimensional Geometry. |
| SACALC3 | Abstract | C00802 PCX86 00 | Calculates the Average Solid Angle Subtended by a Volume. |
| SACHET | Abstract | C00571 D8810 00 | A Computer Program To Evaluate The Dynamic Fission Product Inventories in the Multiple Compartment System of PWR's. |
| 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. |
| SAHYB-2 | Abstract | C00820 I0360 00 | Solution of Ordinary Differential Equation with User-Supplied Subroutine |
| SAIL | Abstract | D00057 I0360 00 | 23 Neutron, 17 Gamma-Ray Group ALBEDO DATA for Concrete and Steel, Based on DOT 1-1/2-D Calculations using DLC-31/FEWG1 Data. |
| SAILOR | Abstract | D00076 PC386 01 | Coupled, Self-Shielded, 47 Neutron, 20 Gamma-Ray, P3, Cross Section Library for Light Water Reactors. |
| 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. |
| SAM-CE | Abstract | C00187 C6600 00 | Monte Carlo Time-Dependent Complex Geometry (Combinatorial) Code System for the Solution of the Forward Neutron and Forward and Adjoint Gamma-Ray Transport Equations. |
| SAM-CE | Abstract | C00187 I0360 00 | Monte Carlo Time-Dependent Complex Geometry (Combinatorial) Code System for the Solution of the Forward Neutron and Forward and Adjoint Gamma-Ray Transport Equations. |
| SAM-CEP | Abstract | C00192 C6600 00 | Monte Carlo Code System Correlated to the Simultaneous Solution of Multiple, Perturbed, Time-Dependent Neutron Transport Problems in Complex Three-Dimensional Geometry. |
| 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. |
| SAMSY | Abstract | C00315 C0073 00 | A One-Dimensional Multilayer Multigroup Neutron Removal-Diffusion and Gamma-Ray Point Kernel Calculator. |
| SAND-II | Abstract | C00112 MNYCP 03 | Neutron Flux Spectra Determination by Multiple Foil Activation Method. |
| SAND-II-SNL | Abstract | P00345 SUN04 00 | Neutron Flux Spectra Determination by Multiple Foil Activation Method. |
| SANDOR | Abstract | C00364 C7600 00 | Isotope Generation and Depletion Code Matrix Exponential Method. |
| SANDYL | Abstract | C00361 C0000 00 | A Monte Carlo Three-Dimensional Code System for Calculating Combined Photon-Electron Transport in Complex Systems. |
| SAP N-G | Abstract | C00092 I7094 00 | Neutron and Gamma-Ray Albedo Model Scatter Shield Analysis Code System. |
| 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. |
SCALE 6.3.1-EXE
810 | Abstract | C00860 MNYCP 03 | A Comprehensive Modeling and Simulation Suite for Nuclear Safety Analysis and Design |
| 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. |
| SCAP-82 | Abstract | C00418 C7600 00 | Single Scatter, Albedo Scatter, or Point Kernel Analysis Code System in Complex Geometry. |
| 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. |
SCEPTRE 1.1
FEDC | Abstract | C00807 PCX86 00 | Sandia Computational Engine for Particle Transport for Radiation Effects. |
SCEPTRE 1.7
FEDC | Abstract | C00826 PCX86 01 | Sandia Computational Engine for Particle Transport for Radiation Effects. |
| SCINFUL | Abstract | P00267 CY0MP 00 | Scintillator Full Response to Neutron Detection. |
| SCINFUL | Abstract | P00267 D8600 00 | Scintillator Full Response to Neutron Detection. |
| SCIP V1.1 | Abstract | C00749 PCX86 00 | Radioactive Surface Contamination Investigation Program. |
| 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-4 | Abstract | C00234 I0370 00 | Two-Dimensional Multigroup Removal-Diffusion Shielding Code System. |
| 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. |
| SDC | Abstract | C00060 I3675 00 | Kernel Integration Shield Design Code for Radioactive Fuel Handling Facilities. |
| SECA | Abstract | P00104 I0360 00 | Evaluator of Angular Bounds for a Two-Dimensional Symmetric Gaussian Quadrature Set. |
| SEDONE | Abstract | C00345 I0360 00 | A Simulator of Tidal Transient Hydrodynamic Sediment Concentrations Conditions in Controlled Rivers and Estuaries. |
| 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. |
| SENPRO | Abstract | D00045 I3691 02 | Compilation of Multigroup Sensitivity Profiles in SENPRO Format for Fast Reactor Core and Shield Benchmarks and Thermal Reactor Benchmarks. |
| SENSIT | Abstract | C00405 C7600 00 | One-Dimensional, Multigroup Cross Section and Design Sensitivity and Uncertainty Analysis Code System - Generalized Perturbation Theory. |
| SERA-1C1 | Abstract | C00729 MNYCP 01 | Simulation Environment for Radiotherapy Applications. |
| SERPENT2.2.1 | Abstract | C00872 MNYWS 01 | Continuous Energy Monte Carlo Reactor Physics Burnup Calculation Code. |
| SESOIL | Abstract | C00629 IBMPC 03 | Code System to Calculate One-Dimensional Vertical Transport for the Unsaturated Soil Zone. |
| SETS | Abstract | P00380 CDCMF 00 | Set Equation Transformation System. |
| SFACTOR | Abstract | C00310 I0360 00 | Dose Equivalent to a Target Organ Calculator. |
| SFAK | Abstract | C00437 I3033 00 | Code System for Calculation of the Self-Absorption of Unscattered Gamma Radiation from Fuel Assemblies. |
SFHA
USSO | Abstract | P00413 IBMPC 00 | Code System for Spent Fuel Heating Analysis. |
| SHADOK | Abstract | C00216 C6600 00 | Transport Code Systems, P1 Scattering in Infinite Cylindrical and Spherical Geometries by Polynomial Approximation. |
| SHADRAC(G-30) | Abstract | C00084 I7090 00 | Kernel Integration Code - Shield Heating and Dose Rate Calculation in Complex Geometry. |
| SHAMSI | Abstract | D00135 I3033 00 | 48 Group Cross-Section Library for Fusion Nucleonics Analysis. |
| SHARDA | Abstract | C00521 C0740 00 | Sample Heat, Activity, Reactivity, and Dose Analysis for Safety Analysis of Irradiations in a Research Reactor. |
SHC
USSO | Abstract | P00493 CY000 00 | Seismic/Hazard Characterization in the Eastern U.S. |
| SHIELD | Abstract | C00667 MNYCP 01 | Monte-Carlo Code System to Simulate Interaction of High Energy Hadrons with Complex Macroscopic Targets. |
| SHIELDOSE | Abstract | C00379 ALLMF 00 | Code System for Space Shielding Radiation Dose Calculations. |
| SHIELDOSE-PC | Abstract | C00379 IBMPC 00 | Code System for Space Shielding Radiation Dose Calculations. |
| SHREDI | Abstract | C00284 I0360 00 | Multigroup Two-Dimensional (x-y, r-o geometry) Neutron Removal-Diffusion (Spinney Method) Shielding Code System. |
| SIGMA II | Abstract | C00118 C6000 00 | Space Radiation Dose Analysis Within Complex Configurations. |
| SIGMA II | Abstract | C00118 PC486 00 | Space Radiation Dose Analysis Within Complex Configurations. |
| SIGMA-A | Abstract | D00139 ALLMF 00 | Photon Interaction and Absorption Cross Sections. |
| SIGMA-A | Abstract | D00139 IBMPC 00 | Photon Interaction and Absorption Cross Sections. |
| SIGPI | Abstract | P00475 D0785 00 | Fault Tree Cut Set System Performance. |
SIMMER II
USSO | Abstract | C00691 MFMWS 00 | Code System for Two-Dinensional Sn-Neutronics and Fluid Dynamics. |
| SINBAD SEARCH TOOL | Abstract | P00580 MNYCP 00 | SINBAD Search Tool |
| SINBAD-2022 | Abstract | D00237 MNYCP 06 | Shielding Integral Benchmark Archive and Database. |
| 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. |
| SIXTUS-3 | Abstract | C00609 MFMWS 00 | Three-Dimensional, Nodal, Neutron Diffusion Criticality Code System in Hex-Z Geometry. |
| SKETCH-N 1.0 | Abstract | C00808 MNYCP 00 | Solve Neutron Diffusion Equations of Steady-State and Kinetics Problems. |
| SKEWGAUS | Abstract | P00089 I0360 00 | Skewed-Gaussian Line Peak Fitting Code - Multichannel Analyzer (MCA) Spectra - Ge(Li) and Semiconductor Detectors. |
| SKYDATA-KSU | Abstract | D00188 IBMPC 00 | Parameters for Approximate Neutron and Gamma-Ray Skyshine Response Functions and Ground Correction Factors. |
| SKYIII-PC | Abstract | C00289 IBMPC 01 | Calculation of the Effects of Structure Design on Neutron, Primary Gamma-Ray and Secondary Gamma-Ray Dose Rates in Air. |
| SKYPORT | Abstract | D00093 IBMPC 00 | Skyshine Importance Functions for Neutrons and Gamma Rays. |
| SKYSHINE-III | Abstract | C00289 D0VAX 00 | Calculation of the Effects of Structure Design on Neutron, Primary Gamma-Ray and Secondary Gamma-Ray Dose Rates in Air. |
| SKYSHINE-KSU | Abstract | C00646 IBMPC 03 | Calculation of the Effects of Structure Design on Neutron, Primary Gamma-Ray and Secondary Gamma-Ray Dose Rates in Air. |
| SLAROM | Abstract | P00244 FM380 00 | A Code to Produce Cell Averaged Cross Sections for Fast Critical Assemblies and Fast Power Reactors. |
| SLDN | Abstract | C00221 A1000 00 | Code System for Shielding Calculations by the Method of Invariant Imbedding. |
| SLDN | Abstract | C00221 F2307 00 | Code System for Shielding Calculations by the Method of Invariant Imbedding. |
| SLDN | Abstract | C00221 FM200 00 | Code System for Shielding Calculations by the Method of Invariant Imbedding. |
| SLDN | Abstract | C00221 GE625 00 | Code System for Shielding Calculations by the Method of Invariant Imbedding. |
| SLDN | Abstract | C00221 I0360 00 | Code System for Shielding Calculations by the Method of Invariant Imbedding. |
| SLIDERULE 1.0 | Abstract | C00704 PC586 01 | Nuclear Criticality Slide Rule. |
| SMACS | Abstract | P00396 C7600 01 | Probabilistic Seismic Analysis Code System. |
| SMAFS | Abstract | P00547 PC586 00 | Steady-State Analysis Model for Advanced Fuel Cycle Schemes. |
| SMART | Abstract | C00602 ALLCP 00 | Code System for Calculating Early Offsite Consequences from Nuclear Reactor Accidents. |
| SMART/MANYCASK | Abstract | C00482 FM200 00 | A Program for Calculating Radiation Dose Rates. |
| SMAUG-13 | Abstract | C00194 C6600 00 | Calculation of Neutron and Prompt Gamma-Ray Doses Resulting from an Atmospheric Nuclear Detonation. |
| SMOG | Abstract | P00216 I3033 00 | Code System for Neutron Cross Section Evaluation (Optical Method). |
| SNAKE | Abstract | P00135 I0360 00 | A Solid Angle Calculational System. |
| SNAP-3D | Abstract | C00434 MNYCP 01 | Multigroup Complex Geometry Neutron Diffusion Code System. |
| SNEX | Abstract | C00353 C0000 00 | A One-Dimensional Single Group Discrete Ordinates Transport Code System. |
| SNLRML | Abstract | D00178 ALLCP 00 | Recommended Dosimetry Cross Section Compendium. |
| SNOW | Abstract | C00282 I0360 00 | Two-Dimensional Discrete Ordinates Multigroup Transport Code System in Plane and Cylindrical Geometry with Isotropic and Anisotropic Scattering. |
| SOFIP | Abstract | C00358 I3033 00 | Evaluator of Space Radiation Environment Encountered by Geocentric Satellites. |
| 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 |
| SOLTRAN | Abstract | C00763 PCX86 00 | Solving Multi-Dimensional Simplified P2 Transport and Diffusion Problems of Hexagonal Geometry in Fast Reactors. |
| SOPHIA | Abstract | C00857 MNYCP 00 | A Lagrangian-based computational fluid dynamics code for nuclear thermal hydraulics and safety applications. |
| SORA | Abstract | P00174 I0360 00 | A Code System for Storage and Retrieval of Data from Radionuclide Analyses. |
| SOSUM | Abstract | C00109 I3675 00 | Multigroup Beta and Gamma-Ray Energy Sources from Activities. |
| SOURCES-4C | Abstract | C00661 MNYCP 04 | Code System for Calculating (alpha,n), Spontaneous Fission, and Delayed Neutron Sources and Spectra. |
| SPACETRAN 1;2;3 | Abstract | C00120 I3675 00 | Dose Calculations at Detectors at Various Distances from the Surface of a Cylinder. |
| SPAR | Abstract | C00228 C6600 00 | Calculation of Stopping Powers and Ranges for Muons, Charged Pions, Protons and Heavy Ions. |
| SPAR | Abstract | C00228 I0360 00 | Calculation of Stopping Powers and Ranges for Muons, Charged Pions, Protons and Heavy Ions. |
| SPARES | Abstract | C00148 I3675 00 | Space Radiation Environment and Shielding Code System. |
| 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. |
| SPECTRA | Abstract | C00108 C0000 00 | Determination of Neutron Spectra from Activation. |
| SPECTRA | Abstract | C00108 C0073 00 | Determination of Neutron Spectra from Activation. |
| SPECTRA | Abstract | C00108 C3600 00 | Determination of Neutron Spectra from Activation. |
| SPECTRANS-2 | Abstract | P00071 ICL00 00 | Neutron Spectrum Library Generation. |
| SPEEDI | Abstract | C00507 FM180 00 | Code System for Real-Time Prediction of Radiation Dose to the Public Due to an Accidental Release from a Nuclear Power Plant. |
| 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. |
| SPOOR | Abstract | C00278 C7600 00 | Monte Carlo Simulation of the Turbulent Transport of Airborne Contaminants. |
| SPOT1 | Abstract | C00460 I3033 00 | Shielding Problem Code Based on Methods of Ono and Tsuruo. |
| 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. |
| SRAC95 | Abstract | C00716 MNYWS 00 | Thermal Reactor Code System for Reactor Design and Analysis. |
| SRNA-2K5 | Abstract | C00789 PCX86 00 | Proton Transport Simulation by Monte Carlo Techniques. |
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. |
| STACY | Abstract | C00859 PCX86 00 | Source Term Analysis Code System. |
STANDARDS 5.0.1
RUGA | Abstract | C00873 PCX86 01 | Storage Transportation and Disposal Analysis Resource and Data System. |
| 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. |
| STERNO | Abstract | C00057 C0000 00 | Two Dimensional Gamma-Ray Heating Kernel Integration Code. |
| STEX II | Abstract | M00010 MNYCP 00 | International Steam Explosion Experimental Data Base. |
| STOPOW88 | Abstract | C00790 MNYCP 00 | Stopping Power of Fast Ions in Matter. |
| STORM | Abstract | C00067 I7090 00 | Solar Flare Radiation Hazard to Earth Orbiting Vehicles. |
| STORM-ISRAEL | Abstract | D00015 I0360 01 | Evaluated Photon Interaction Library, ENDF/B File 23 Format. |
| STRADE | Abstract | P00252 I3081 00 | Stratified Random Design. |
| STRAGL | Abstract | C00201 C6600 00 | Calculation of Energy Loss Straggling of Heavy Charged Particles. |
| STRAINT | Abstract | C00259 I0360 00 | One-Dimensional Multigroup Neutron Transport Discrete Ordinates Code System. |
| STREAM | Abstract | C00321 C7600 00 | A Three-Dimensional Cylindrical-Geometry Monte Carlo Ray Tracing Code for Computing Light Transmission. |
| SUBDOSA-II | Abstract | C00270 U1100 00 | Calculation of External Gamma-Ray and Beta-Ray Doses from Accidental Atmospheric Releases of Radionuclides. |
| SUGGEL | Abstract | P00508 MNYWS 00 | Program Suggesting the Orbital Angular Momentum of a Neutron Resonance From the Magnitude Of Its Neutron Width. |
| SULSA | Abstract | M00015 MNYCP 00 | A Solution for the Neutron Spectrum Unfolding Problem Without Using Input Spectrum (Report Only). |
| 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. |
| SURF | Abstract | C00102 I3675 00 | Conical and Plane Surface Single Scattering Code. |
| SUSD | Abstract | C00501 HM150 00 | Cross Section Sensitivity and Uncertainty Analysis Including Secondary Neutron Energy and Angular Distributions. |
| SUSD | Abstract | C00501 I3090 00 | Cross Section Sensitivity and Uncertainty Analysis Including Secondary Neutron Energy and Angular Distributions. |
| SUSD3D | Abstract | C00695 MNYCP 01 | Multi-Dimensional, Discrete-Ordinates Based Cross Section Sensitivity and Uncertainty Analysis Code System. |
| SWAN | Abstract | C00248 C0000 00 | Code System for Analysis and Optimization of Fusion Reactor Nucleonic Characteristics. |
| SWAN | Abstract | C00248 CY000 00 | Code System for Analysis and Optimization of Fusion Reactor Nucleonic Characteristics. |
| SWAN | Abstract | C00248 I0360 00 | Code System for Analysis and Optimization of Fusion Reactor Nucleonic Characteristics. |
| SWANLAKE | Abstract | C00204 C6600 00 | Cross Section Sensitivity Analysis Code System for One-Dimensional Discrete Ordinates Calculations. |
| SWANLAKE | Abstract | C00204 I3033 00 | Cross Section Sensitivity Analysis Code System for One-Dimensional Discrete Ordinates Calculations. |
| SWAP-9 | Abstract | C00788 C0740 00 | 1-D Stress Analysis for Hydrostatic and Elastic Plastic Materials. |
| SWAT | Abstract | C00714 MNYCP 01 | Step-Wise Burnup Analysis Code System to Combine SRAC-95 Cell Calculation Code and ORIGEN2. |
| SWIFT | Abstract | C00679 C7600 00 | Code System to Calculate Waste-Isolation Flow and Transport. |
| SWIFT | Abstract | P00031 C6600 00 | Monte Carlo Neutron Spectra Unfolding Code. |
SWIFT2
USSO | Abstract | C00686 MNYCP 00 | Code System to Calculate Waste-Isolation Flow and Transport. |
| SWORD 7.0 | Abstract | C00767 MNYCP 07 | SoftWare for Optimization of Radiation Detectors. |
| SYVAC-D/2 | Abstract | C00690 D0VAX 00 | Code System For Risk Assessment From Underground Radioactive Waste Disposal In the United Kingdom. |
| TACT-III | Abstract | C00447 I3033 00 | Calculation of the Transport of Radioactivity from a Reactor Core. |
| 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. |
| TART2022 | Abstract | C00638 MNYCP 09 | Coupled Neutron-Photon, 3-D, Combinatorial Geometry, Time Dependent, Monte Carlo Transport Code System. |
| TASK | Abstract | C00184 I0360 00 | Generalized One-Dimensional Radiation Transport and Diffusion Kinetics Code System. |
| TDA | Abstract | C00180 MNYWS 01 | A Time-Dependent, Multigroup, One-Dimensional, Discrete Ordinates Transport Code System. |
| TDF | Abstract | D00162 ALLCP 00 | Thermonuclear Data File. |
| TDOWN-IV | Abstract | P00172 H6000 00 | A Code System to Generate Composition- and Spatially-Dependent Neutron Cross Sections for Multigroup Neutronics Analysis. |
| TDT | Abstract | C00256 I0360 00 | Generalized One-Dimensional Multigroup Time-Dependent Transport and Diffusion Kinetic Code System. |
| TDTORT | Abstract | C00709 MNYWS 00 | Time-Dependent, 3-D, Discrete Ordinates, Neutron Transport Code System. |
| 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. |
| TENDL-2008-ACE | Abstract | D00243 MNYCP 00 | TALYS-Based Cross Section Library for Use with MCNP(X). |
| TENDL-2010-ACE | Abstract | D00248 MNYCP 00 | TALYS-Based Cross Section Library for Use with MCNP(X). |
| TENDL-2011-ACE | Abstract | D00252 MNYCP 00 | TALYS-Based Cross Section Library for Use with MCNP(X). |
| TENDL-2012-ACE | Abstract | D00266 MNYCP 00 | TALYS-Based Cross Section Library for Use with MCNP(X). |
| TERFOC-N | Abstract | C00596 MFMWS 00 | Terrestrial Food-Chain Model for Normal Operations. |
| TESS | Abstract | C00215 C3600 00 | Multigroup Discrete Ordinates Code System for Slab and Spherical Geometries. |
| THACT-RR | Abstract | P00587 D0VAX 00 | Analysis of Thermal Hydraulics Transients in Research Reactor Core. |
| THERMGAM | Abstract | D00140 ALLCP 00 | Prompt Gamma Rays from Thermal-Neutron Capture. |
| 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. |
| THIDA-2 | Abstract | C00410 FM380 00 | Code System for the Calculation of Transmutation, Activation, Decay Heat and Dose Rate in Fusion Reactors. |
| THRUSH | Abstract | P00276 CYXMP 00 | Calculates Thermal Neutron Scattering Kernel. |
| THT | Abstract | C00480 I0360 00 | Three-Dimensional Neutron Coarse Mesh Code System to Evaluate Average Bundle Fluxes and Power in Light Water Reactors. |
| 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. |
| TIBSO | Abstract | C00512 MNYCP 00 | Code System to Calculate Production and Migration of Radionuclides in Nuclear Reactor Systems. |
| TIMED | Abstract | C00292 I0360 00 | Calculation of Cumulated Activity of a Radionuclide in the Organs of the Human Body at a Given Time After Deposition. |
| TIMEX | Abstract | C00274 C7600 00 | One Dimensional, Time Dependent Multigroup Explicit Discrete Ordinates Radiation Transport Code System with Anisotropic Scattering. |
| TIMEX | Abstract | C00274 CY000 00 | One Dimensional, Time Dependent Multigroup Explicit Discrete Ordinates Radiation Transport Code System with Anisotropic Scattering. |
| TIMEX | Abstract | C00274 U1106 00 | One Dimensional, Time Dependent Multigroup Explicit Discrete Ordinates Radiation Transport Code System with Anisotropic Scattering. |
| TIMOC-72 | Abstract | C00144 I0370 00 | Monte Carlo Three-Dimensional Neutron Transport Code System. |
| TIMOC-ESP | Abstract | C00432 U1110 00 | System for Generating and Analyzing Time Dependent Radiation Transport Results by Monte Carlo. |
| 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. |
| TIRION 4 | Abstract | C00395 I3033 00 | A Program for Calculating Consequences of a Release of Radioactive Material to the Atmosphere. |
| TITAN 1.29 | Abstract | C00759 PCX86 04 | A Three-Dimensional Deterministic Radiation Transport Code System. |
| TMAP 7 | Abstract | C00858 PCX86 00 | Tritium Migration Analysis Program |
| TMMS | Abstract | C00246 I0360 00 | Gamma-Ray Penetration Shielding Code System, Transmission Matrix Method. |
| 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. |
| TOXRISK | Abstract | C00692 CDCMF 00 | Code System for Toxic Gas Accident Analysis. |
| TP1 | Abstract | C00465 I3033 00 | A Computer Code System for the Calculation of Reactivity and Kinetic Parameters by One-Dimensional Neutron Transport Perturbation Theory. |
| TP2 | Abstract | C00470 I3033 00 | A Computer Code System for the Calculation of Reactivity and Kinetic Parameters by One-Dimensional Neutron Transport Perturbation Theory. |
| TPASGAM 85 | Abstract | D00088 ALLCP 04 | Radioactive Decay Library of Gamma-Ray Energies, Branching Ratios, and Cross Sections. |
| TPASS | Abstract | P00164 DP010 00 | A Gamma-Ray Spectral Data-Reduction and Analysis Code System. |
| TPHEX | Abstract | C00421 C0173 00 | Transmission Probability Code System for Calculating Neutron Flux Distributions in Hexagonal Geometry. |
| TPHEX | Abstract | C00421 CYXMP 00 | Transmission Probability Code System for Calculating Neutron Flux Distributions in Hexagonal Geometry. |
| TPTRIA | Abstract | C00550 I3083 00 | A Computer Program for the Reactivity and Kinetic Parameters for Two-Dimensional Triangular Geometry by Transport Perturbation Theory. |
| TRANSHEX | Abstract | C00449 U1108 00 | Two-dimensional Multigroup Collision Probability Code System for Hexagonal Geometry. |
| TRANSMIT | Abstract | D00020 I0360 00 | Experimental Neutron Transmission Data Used to Test Total Cross Sections. |
| TRANSPORT | Abstract | C00244 C6600 00 | Charged Particle Beam Transport Systems Design Code System (First- and Second-Order Matrix Multiplication). |
| TRANSPORT | Abstract | C00244 I0360 00 | Charged Particle Beam Transport Systems Design Code System (First- and Second-Order Matrix Multiplication). |
| 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. |
| TRANZIT | Abstract | C00172 C7600 00 | Multigroup Time-Dependent Discrete Ordinates Radiation Transport Code System in (rho,z) Cylindrical Geometry. |
| TRAPP | Abstract | C00205 I3691 00 | Transport of Alpha Particles and Protons with all Nuclear Reaction Products Neglected. |
| TRAX | Abstract | P00280 C0720 00 | A Program For Optics of Curved Crystal Neutron Spectrometers. |
| TRD-3 | Abstract | C00362 I3033 00 | Two-Dimensional Removal-Diffusion Neutron Shielding Code System. |
| TRECO | Abstract | C00116 I3675 00 | An Orbital Integration Estimation of Trapped Radiation. |
| TR-EDB | Abstract | D00198 IBMPC 00 | Test Reactor Embrittlement Data Base. |
| TREEDE | Abstract | C00326 C0000 00 | Monte Carlo Neutron Transport Code System Based on the Track Rotation Estimator. |
| TRG-SGD | Abstract | C00025 C0000 00 | Calculation of Secondary Gamma-Ray Dose Rate from a Nuclear Weapon Detonation-Monte Carlo Method. |
| TRIDENT | Abstract | C00293 C7600 00 | Two-Dimensional Multigroup Discrete Ordinates Transport Code System-(x,y) and (r,z) Geometries. |
| TRIDENT | Abstract | C00293 I0360 00 | Two-Dimensional Multigroup Discrete Ordinates Transport Code System-(x,y) and (r,z) Geometries. |
| TRIDENT-CTR | Abstract | C00377 C0000 00 | Two-Dimensional x-y and r-z Geometry Multigroup Transport Code System for Large Toroidal Reactors. |
| TRIGAP | Abstract | C00600 IBMPC 00 | A Computer Code for TRIGA Type Reactors. |
| TRIGLAV | Abstract | P00495 PC586 00 | Code System to Calculate Mixed Cores in TRIGA Mark II Research Reactor. |
| TRIGON | Abstract | C00290 U1108 00 | Two-Dimensional Multigroup Diffusion Code System-Trigonal or Hexagonal Mesh. |
| TRIPLET | Abstract | C00230 C6600 00 | Two-Dimensional, Multigroup, Triangular Mesh, Planar Geometry, Explicit Discrete Ordinates Code System. |
| TRIPLET | Abstract | C00230 C7600 00 | Two-Dimensional, Multigroup, Triangular Mesh, Planar Geometry, Explicit Discrete Ordinates Code System. |
| TRIPLET | Abstract | C00230 I0360 00 | Two-Dimensional, Multigroup, Triangular Mesh, Planar Geometry, Explicit Discrete Ordinates Code System. |
TRIPOLI-4 8.1
OECD | Abstract | C00806 MNYCP 00 | Code System for Coupled Neutron, Photon, Electron, Positron, 3-D, Time Dependent, Monte-Carlo, Transport Calculations. |
TRIPOLI-4 9S
OECD | Abstract | C00815 MNYCP 00 | Code System for Coupled Neutron, Photon, Electron, Positron, 3-D, Time Dependent, Monte-Carlo, Transport Calculations. |
| TRIPOS | Abstract | C00537 CY00I 00 | Monte Carlo Ion Transport Analysis Code. |
| TRISTAN | Abstract | C00511 HM280 00 | Multigroup Three-Dimensional Direct Integration Method Radiation Transport Analysis Code System. |
| TRISTAN-IJS | Abstract | P00537 IBMPC 00 | Multigroup Three-Dimensional Direct Integration Method Radiation Transport Analysis Code System. |
| TRITAC | Abstract | C00560 D8810 00 | A Three-Dimensional Transport Code For Eigenvalue Problems Using The Diffusion Synthetic Acceleration Method. |
| TRUMP | Abstract | P00522 MNYCP 01 | Code System for Transient and Steady-State Temperature Distribution in Multidimensional Systems. |
| TSL-ACE/2013 | Abstract | D00270 ALLCP 00 | TSL-ACE/2013 |
| 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. |
| TWOTRAN | Abstract | C00195 C6600 00 | Two-Dimensional Multigroup Discrete Ordinates Transport C System in (x,y), (r,theta), and (r,z) Geometries. |
| TWOTRAN II | Abstract | C00222 C7600 00 | Two-Dimensional Multigroup Discrete Ordinates Transport C System in (x,y), (r,theta), and (r,z) Geometries. |
| TWOTRAN II | Abstract | C00222 I3691 00 | Two-Dimensional Multigroup Discrete Ordinates Transport C System in (x,y), (r,theta), and (r,z) Geometries. |
| TWOTRAN-SPHERE | Abstract | C00129 C6600 00 | Multigroup Two-Dimensional Discrete Ordinates Transport Code System in Spherical Geometry. |
| UDAD IX | Abstract | C00685 I0370 00 | Uranium Dispersion & Dosimetry Model. |
| UHS | Abstract | P00390 IPS70 00 | Ultimate Heat Sink Cooling Pond and Spray Pond Analysis Models. |
| UKCTRI-81 | Abstract | D00064 I0370 01 | 46-Group Neutron Cross Sections and Kerma Factors for Fusion Reactor Calculations. |
| UKE-III | Abstract | P00015 I3691 00 | Cross Section Format Translator - UKNDL to ENDF/B. |
| UKFY2 | Abstract | D00171 IBMPC 00 | UK Fission Product Yield Library. |
| UKNDL | Abstract | D00039 I0370 00 | United Kingdom Evaluated Neutron Cross-Section Data Library. |
| UKNDL-81 | Abstract | D00107 I3033 00 | The Aldermaston Nuclear Data Library. |
| UMG 3.3 | Abstract | P00529 PC586 00 | Unfolding with Maxed and Gravel. |
| UMIBIO | Abstract | C00680 I3033 00 | Code System to Model Uranium Mills Bioassay Dosimetry. |
| UNF | Abstract | P00521 PC586 00 | Code System to Calculate Multistep Compound Nucleus Neutron Cross-Sections and Spectra for Structural Materials. |
| UNGER | Abstract | D00164 PC386 00 | Effective Dose Equivalent for Specific Radionuclides. |
| UNIFY-ECN | Abstract | P00288 C0170 00 | A Program to Calculate Fast Neutron Data for Structural Materials. |
| UNIMUG3 | Abstract | C00407 C0170 00 | Solves Multigroup Diffusion Equations in One-Dimensional Systems. |
| 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 |
| USRHYD | Abstract | C00197 I3675 00 | Electron and X-Ray Energy Deposition and Hydrodynamics Code System. |
| UTMTOX | Abstract | C00500 D8600 00 | Unified Transport Model for Toxic Materials. |
| 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. |
| UTXS6 | Abstract | D00211 MNYCP 00 | MCNP Continuous-Energy Neutron Cross Section Libraries for Temperatures from 300 to 1365K. |
| VALE 1.1 | Abstract | C00613 IRISC 01 | A Multigroup Diffusion Theory Neutronics Code System for Solving Two- and Three-Dimensional Problems for Triagonal Geometries. |
| VALE 1.1 | Abstract | C00613 PC386 01 | A Multigroup Diffusion Theory Neutronics Code System for Solving Two- and Three-Dimensional Problems for Triagonal Geometries. |
| VARSKIN 4 | Abstract | C00781 PCX86 00 | Code System for Assessing Skin Dose from Skin Contamination. |
| VCS | Abstract | C00262 I0360 00 | Coupled Discrete Ordinates-Adjoint Monte Carlo Calculation of Radiation Protection Factors in Vehicles. |
| VELM | Abstract | D00133 I0360 00 | Multigroup Cross-Section Libraries Based on ENDF/B-V Data for Sodium-Cooled Reactor Shield Analysis. |
| VENTEASY | Abstract | C00776 PCX86 00 | Criticality search for a desired Keffective by adjusting dimensions, nuclide concentrations, or buckling |
| VENTURE-PC | Abstract | C00654 PC586 02 | A Reactor Analysis Code System. |
VERA 4.3-EXE
810 | Abstract | C00855 PCX86 05 | Virtual Environment for Reactor Applications |
| VESTA 2.1.5-AURORA 1.0 | Abstract | C00769 PCX86 01 | A Generic Monte Carlo Code and Depletion Module Interface. |
| 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. |
| VIM 5.1 | Abstract | C00754 MNYWS 01 | Continuous Energy Neutron and Gamma-ray Transport Code System. |
| VIM_NC | Abstract | C00794 PCX86 00 | VIM Color Syntax for Nuclear Codes: NJOY, DRAGON, PARTISN, TORT, MONK, and MCNP. |
| VIP-MAN | Abstract | D00256 MNYCP 00 | Computational Phantom. |
| 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 |
| VITAMIN-4C | Abstract | D00053 I3691 00 | Broad-Group and Fine-Group and Coupled Neutron/Gamma Cross-Section Libraries Derived from ENDF/B-VII.0 Nuclear Data. |
| VITAMIN-B6 | Abstract | D00184 ALLCP 00 | Broad-Group and Fine-Group and Coupled Neutron/Gamma Cross-Section Libraries Derived from ENDF/B-VII.0 Nuclear Data. |
| VITAMIN-B7/BUGLE-B7 | Abstract | D00245 MNYCP 01 | Broad-Group and Fine-Group and Coupled Neutron/Gamma Cross-Section Libraries Derived from ENDF/B-VII.0 Nuclear Data. |
| VITAMIN-C | Abstract | D00041 I0360 02 | Broad-Group and Fine-Group and Coupled Neutron/Gamma Cross-Section Libraries Derived from ENDF/B-VII.0 Nuclear Data. |
| VITAMIN-E | Abstract | D00113 I3033 02 | Broad-Group and Fine-Group and Coupled Neutron/Gamma Cross-Section Libraries Derived from ENDF/B-VII.0 Nuclear Data. |
| VITAMIN-J/COVA | Abstract | D00157 D8810 00 | Broad-Group and Fine-Group and Coupled Neutron/Gamma Cross-Section Libraries Derived from ENDF/B-VII.0 Nuclear Data. |
| VITAMIN-J/COVA/EFF | Abstract | D00197 ALLCP 00 | Broad-Group and Fine-Group and Coupled Neutron/Gamma Cross-Section Libraries Derived from ENDF/B-VII.0 Nuclear Data. |
| VITAMIN-J/KERMA | Abstract | D00150 I3090 00 | Broad-Group and Fine-Group and Coupled Neutron/Gamma Cross-Section Libraries Derived from ENDF/B-VII.0 Nuclear Data. |
| VITENDF70.BOLIB | Abstract | D00261 PCX86 00 | ENDF/B-VII.0 Multi-Group Coupled (199n +42gamma) Cross Section Library in AMPX Format for Nuclear Fission Applications. |
| VITENEA-E | Abstract | D00240 MNYCP 00 | AMPX 175-n,42-g Multigroup X-section Library for Nuclear Fusion Applications. |
| VITENEA-J | Abstract | D00238 MNYCP 00 | AMPX 175-n,42-g Multigroup X-section Library for Nuclear Fusion Applications. |
| VITJEF22.BOLIB | Abstract | D00241 MNYCP 00 | JEF-2.2 Multigroup Coupled (199n + 42?) Cross-Section Library in AMPX Format for Nuclear Fission Applications. |
| VITJEFF31.BOLIB | Abstract | D00235 MNYCP 00 | A JEFF-3.1 Multigr Coupled (199n + 42gamma) X-Section Lib. in AMPX Fmt for Nuclear Fission Applications. |
| VITJEFF311.BOLIB | Abstract | D00257 MNYCP 01 | JEFF-3.1.1 Multi-Group Coupled (199n + 42gamma) X-Section Library in AMPX Format for Nuclear Fission Applications. |
| 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. |
| VPI-NECM | Abstract | C00481 C0740 00 | Nuclear Engineering Computer Models for In-Core Fuel Management Analysis. |
| VPI-NECM | Abstract | C00481 D0VAX 00 | Nuclear Engineering Computer Models for In-Core Fuel Management Analysis. |
| VPI-NECM | Abstract | C00481 PC486 00 | Nuclear Engineering Computer Models for In-Core Fuel Management Analysis. |
| VSOP94 | Abstract | C00670 MNYWS 00 | Computer Code System for Reactor Physics and Fuel Cycle Simulation. |
| VVER-BENCHMARKS | Abstract | M00003 MNYCP 00 | Collection of Neutronic VVER Reactor Benchmarks. |
| WAKE | Abstract | P00605 I0370 00 | Navier Stokes Equation with 2-D Turbulence, Stream Function, Vorticity. |
| WEERIE | Abstract | C00426 I3033 00 | Code System for Assessing the Radiological Consequences of Airborne Effluents from Nuclear Installations. |
| WHATIF-AQ | Abstract | C00561 B7800 00 | A Computer Program For Speciation Calculation. |
| WILIT | Abstract | P00344 MNYCP 00 | A Utility Program for WIMS Libraries. |
| WIMKAL-88 | Abstract | D00193 MNYCP 00 | 69 Energy Group, Neutron Cross Section Library For Thermal Reactor Calculations in WIMSD Format. |
| WIMS-ANL 4.0 | Abstract | C00698 MNYCP 00 | Deterministic Code System for Reactor Lattice Calculation. |
| WIMSCORE-ENEA | Abstract | P00319 I3090 00 | Code System to Process WIMSD4 Interface Output Files and Generate Two-Group Data for Reactor Calculations. |
| WIMSD-5B.12 | Abstract | C00656 MNYCP 02 | Deterministic Code System for Reactor Lattice Calculation |
| WIMSLIB-IJS0 | Abstract | D00147 D8810 00 | Extended Version of the WIMS 69-group Library. |
| WIMSLIB-IJS1 | Abstract | D00147 D8810 01 | Extended Version of the WIMS 69-group Library. |
| WIMSLIB-JEF87 | Abstract | D00095 D0VAX 00 | Extended Version of the WIMS 69-group Library. |
| 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. |
| WLUP 3.0 | Abstract | D00231 MNYCP 01 | 69- and 172- Group Cross Section Libraries for WIMS. |
| W-M-NRSM | Abstract | D00026 U1108 00 | WANL-MSFC Nuclear Rocket Shielding Methods Data Generator (GAMLEG-W, APPROPOS, NAGS, and SATURN) and Multigroup Neutron and Gamma-ray Cross Section Libraries 1-6. |
| WRAITH | Abstract | C00427 U1100 00 | Code System for Calculating Internal and External Doses Resulting from an Atmospheric Release of Radioactive Material. |
| 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. |
| XCOM | Abstract | D00174 IBMPC 00 | Photon Cross Sections on a Personal Computer. |
| XG-IAEA | Abstract | D00163 IBMPC 00 | X-ray and Gamma-ray Standards For Detector Calibration. |
| 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. |
| XOQDOQ-82 | Abstract | C00316 DGMV1 00 | Radiological Assessment Code System - Meteorological Evaluation of Routine Effluent Releases at Nuclear Power Stations. |
| XOQDOQ-82 | Abstract | C00316 I3033 00 | Radiological Assessment Code System - Meteorological Evaluation of Routine Effluent Releases at Nuclear Power Stations. |
| XOQDOQ-82 | Abstract | C00316 IPCAT 00 | Radiological Assessment Code System - Meteorological Evaluation of Routine Effluent Releases at Nuclear Power Stations. |
| XOQGAM | Abstract | C00847 MNYCP 00 | Methodology and Software Routines for Computation of Gamma Radiation Exposures from Finite-cloud Gaussian Plumes |
| XPORT-PC | Abstract | C00559 IBMPC 00 | An Approximation For Black Body X-Ray Transport in Air. |
| XRAY_AAC | Abstract | C00525 D0750 00 | X-ray Attenuation and Absorption Calculations. |
| XSDRN | Abstract | C00123 C0073 00 | Multigroup One-Dimensional Discrete Ordinates Spectral Averaging N Transport Code System. |
| XSDRN | Abstract | C00123 I0360 00 | Multigroup One-Dimensional Discrete Ordinates Spectral Averaging N Transport Code System. |
| XSHLD | Abstract | C00495 IBMPC 00 | Diagnostic X-Ray Shielding Calculation. |
| XSUN-2013 | Abstract | C00825 PCX86 00 | Windows interface environment for transport and sensitivity-uncertainty software TRANSX-2, PARTISN and SUSD3D |
| YUMMY | Abstract | D00221 MNYCP 00 | Multi-temperature, Neutron Cross Section Library Based on ENDF/B-V and ENDF/B-VI for use with MCNP. |
| ZOTT99 | Abstract | P00272 ALLCP 02 | Zero-in On The Truth; Evaluation of Correlated Data Using Partitioned Least Squares. |
| ZYLIND-PC | Abstract | C00557 IBMPC 00 | An Interactive Point Kernel Program For Photon Dose Rate Prediction of Cylindrical Source/Shield Arrangements. |
| ZZ-PWR-MSLB | Abstract | D00275 MNYCP 00 | ZZ PWR-MSLB, PWR Main Steam-Line Break Benchmarks, Coupled Neutronics Thermal-Hydraulics |