| Packages with Keyword: NEUTRON |
| Package Name | Abstract | RSICC Tapelist | Title |
| AARE-V1.0 | Abstract | C00846 MNYCP 00 | Activation in Accelerator Radiation Environments |
| ABAREX | Abstract | P00248 MNYCP 01 | Neutron Spherical Optical-Statistical Model Code System. |
| ACAB-2008 | Abstract | C00758 MNYCP 01 | Activation Abacus Inventory Code System for Nuclear Applications. |
| ACDOS3 | Abstract | C00442 C7600 00 | Calculation of Activities and Dose Rates Produced by Neutron Activation. |
| ACOH | Abstract | C00191 I3675 00 | Aerojet COHORT Monte Carlo Code System. |
| 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 |
| AIRDIF | Abstract | C00360 C6600 00 | A Two-Dimensional Atmospheric Radiation Diffusion Code. |
| AIRTRANS | Abstract | C00110 I3675 00 | Monte Carlo Time and Energy-Dependent Three-Dimensional Radiation Transport Code. |
| AKERN | Abstract | C00190 C0000 00 | Aerojet Point Kernel Integration Calculational System. |
| AKERN | Abstract | C00190 U1108 00 | Aerojet Point Kernel Integration Calculational System. |
| ALARA 2.7.8 | Abstract | C00723 MNYCP 00 | Code System for Analytic and Laplacian Adaptive Radioactivity Analysis. |
| ALBEDO/ALBEZ | Abstract | C00555 IBMPC 00 | Calculates Attenuation of Radiation in Single and Double Bends. |
| ALBEMO | Abstract | C00268 C6600 00 | Albedo Monte Carlo Code System. |
| ALPHN | Abstract | C00612 IBMPC 00 | Code System for Calculating (alpha,n) Neutron Production in Canisters of High-Level Waste. |
| AMC | Abstract | C00090 I3675 00 | Monte Carlo Albedo Code for Neutron and Capture Gamma-Ray Distributions in Rectangular Concrete Ducts. |
| 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. |
| 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. |
| ARC | Abstract | C00224 C6600 00 | Aircraft Radiation Transport Code System, Crew Dose Calculation. |
| ARMYL-N | Abstract | C00298 U1106 00 | Calculation of Transmission Factors for Neutrons from Nuclear Explosions. |
| ASOP | Abstract | C00126 IRISC 00 | Multigroup One-Dimensional Discrete Ordinates Transport Code System for Shield Optimization. |
| AUS98 | Abstract | C00519 MNYWS 01 | Modular System for Neutronics Calculations of Fission Reactors, Fusion Blankets, and Other Systems. |
| BASACF | Abstract | P00285 IBMPC 00 | Bayesian Approach to Spectrum Adjustment with Covariance Filter. |
| BERMUDA | Abstract | C00616 FV260 03 | Discrete Ordinates Code System for Shielding Analysis for Use with Fusion and Fission Reactors. |
| 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. |
| 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. |
| 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 |
| BUCORST | Abstract | P00339 PC386 00 | A Code to Prepare Burnup-Dependent Multigroup Nuclear Reactor Source Terms. |
| 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. |
| CAFDATS | Abstract | P00549 MNYCP 00 | Converter of Angular Fluxes of DORT, ANISN and TORT Systems. |
| CALOR95 | Abstract | C00610 MNYWS 00 | Monte Carlo Code System for Design and Analysis of Calorimeter Systems, Spallation Neutron Source (SNS) Target Systems, etc. |
| 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. |
| CASIM | Abstract | C00265 I0360 00 | Monte Carlo Simulation of Transport of Hadron Cascades in Bulk Matter. |
| 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. |
| CINDER 1.05 | Abstract | C00755 PC586 00 | Code System for Actinide Transmutation Calculations |
| CLES | Abstract | D00233 MNYCP 00 | Cross Section Library of Moderator Materials for Low-Energy Neutron Sources. |
| CNCSN 2009 | Abstract | C00726 PCX86 01 | One, Two- and Three-Dimensional Coupled Neutral and Charged Particle SN Parallel Multi-Threaded Code System. |
| COHORT-II | Abstract | C00198 I7094 00 | General Purpose Monte Carlo Radiation Transport Code System. |
| COMBINE-PC | Abstract | P00286 IBMPC 00 | Code System to Compute Neutron Spectra and ENDF/B Version 5 Based Multigroup Neutron Constants. |
| 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. |
| COMPRASH | Abstract | C00072 I3675 00 | Spinney Removal-Diffusion Shielding Code. |
| CRESO | Abstract | P00184 I3081 00 | Resonance Data-Handling Code System. |
| 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. |
| CYGNUS-C SPHERE | Abstract | C00232 I0360 00 | Monte Carlo Neutron Transport Code System in Spherical Geometry. |
| DANTSYS 3.0 | Abstract | C00547 MFMWS 01 | One-, Two-, and Three-Dimensional, Multigroup, Discrete-Ordinates Transport Code System. |
| 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. |
| DEMON & DEMON R | Abstract | C00181 I3675 00 | Demonstration Monte Carlo Code System in Slab Geometry. |
| 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. |
| 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. |
| 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. |
| DOORS 3.2A | Abstract | C00650 MFMWS 04 | One, Two- and Three-Dimensional Discrete Ordinates Neutron/Photon Transport Code System. |
| DOT 3.5 | Abstract | C00276 I0360 00 | Two-Dimensional Discrete Ordinates Radiation Transport Code System. |
| DRAGON3.05D | Abstract | C00647 MNYWS 03 | Lattice Cell Code System. |
| 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. |
| EASY-QAD 2.0.1 | Abstract | C00744 PC586 02 | A Visualization Code System for Gamma and Neutron Shielding Calculations. |
| ELF | Abstract | C00167 I0360 00 | Monte Carlo Neutron Transport Code System for Cylinders and Spheres. |
| EMPIRE-II | Abstract | P00497 PC586 01 | Comprehensive Nuclear Model Code, Nucleons, Ions Induced Cross-Sections. |
| ESP | Abstract | C00193 I0360 00 | General Purpose Monte Carlo Neutron Transport Code System. |
| EXIFON2.0 | Abstract | P00305 IPCXT 01 | A Model for Statistical Multistep Direct and Multistep Compound Reactions. |
| EXTREME | Abstract | C00440 I3033 00 | Two-Dimensional Discrete-Ordinates Code System with Exponential Expansion of Spatial Variables. |
| 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. |
| FDKR | Abstract | C00541 I4381 00 | Radioactivity and Dose Rate Calculation Code for Fission, Fusion and Hybrid Reactors. |
| FEM-2D | Abstract | C00260 C6600 00 | Two-Dimensional Diffusion Theory Code System Based on the Method of Finite Elements. |
| 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. |
| FLUKA05-PRE-LIB | Abstract | D00260 PCX86 00 | FLUKA05 Multi-Group, Multi-Purpose Nuclear Data Library, Neutrons, Photons, Charged Particles. |
| 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. |
| FPZD | Abstract | C00603 PC386 00 | Code System for Multigroup Neutron Diffusion/Depletion Calculations. |
| FURNACE | Abstract | C00615 C0740 00 | Code System for Neutronic Calculations in Three Dimension Toroidal Geometry. |
| GBANISN | Abstract | C00628 IRISC 00 | One-Dimensional Discrete Ordinates Transport Code System with Anisotropic Scattering with the GroupBand Option. |
| GENP-2 | Abstract | C00575 ALLMF 00 | Generalized Perturbation Theory Code System. |
| GIRAFFE | Abstract | P00304 I3033 00 | General Isotope Release Analysis For Failed Elements. |
| 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. |
| GROUPSTRUCTURES | Abstract | D00274 MNYCP 00 | GROUPSTRUCTURES, VITAMIN-J, XMAS, ECCO-33, ECCO2000 Standard Group Structures |
| 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. |
| HAM | Abstract | C00267 U1108 00 | Monte Carlo Multigroup Neutron and Photon High Altitude Transport Code System. |
| HEPROW | Abstract | C00799 MNYCP 00 | Unfolding of Pulse Height Spectra Using Bayes Theorem and Maximum Entropy Method. |
| HEXAB-3D | Abstract | C00593 I0370 00 | Three-Dimensional Few-Group Coarse Mesh Diffusion Code for Neutron Physics Calculation of Reactor Core in Hexagonal Geometry. |
| INAP | Abstract | C00235 U1108 00 | Improved Neutron Activation Prediction Code Systems. |
| INDRA | Abstract | C00303 I0360 00 | A Modular System for Calculating the Neutronics and Photonics Characteristics of a Fusion Reactor Blanket. |
| 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). |
| KAMCCO | Abstract | C00325 I0370 00 | Three-Dimensional Time Dependent Monte Carlo Code System for Fast Neutron Physics Problems. |
| 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. |
| KDLIBE | Abstract | C00124 I3675 00 | Kernel-Diffusion Shielding Analysis System. |
| 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. |
| LAHET 2.8 | Abstract | C00696 MFMWS 00 | Code System for High Energy Particle Transport Calculations. |
| LASER | Abstract | C00344 I0360 00 | A One-Dimensional, Neutron-Thermalization, Lattice-Cell Program Based on MUFT and THERMOS. |
| 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. |
| LG-H | Abstract | C00087 I7090 00 | Ray Analysis Cylindrical Duct Kernel Code for Neutrons and Gamma Rays. |
| MADONNA | Abstract | C00425 I0370 00 | Two-dimensional Neutron Streaming Coupled Removal-Diffusion-Albedo-Transport Code System. |
| MAGIK | Abstract | C00359 I0360 00 | A Monte Carlo Code System for Computing Induced Residual Activation Dose Rates. |
| MAP | Abstract | C00150 I3675 00 | Kernel Integration Code System in Complex Geometry with Special Application to Surface Sources Determined by Discrete Ordinates Calculations. |
| MARC-PN | Abstract | C00311 D8810 00 | A Neutron Diffusion Code System with Spherical Harmonics Option. |
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. |
| MEDUSA-PIJ | Abstract | C00349 F2307 00 | One-Dimensional Lagrangian Code for Plasma Hydrodynamic Analysis of a Fusion Pellet Driven by Ion Beams. |
| MERCURE 4-82 | Abstract | C00142 I3033 00 | Three-Dimensional Code System for Integrating Multigroup Line-of-Sight Attenuation Kernels by Monte Carlo Techniques. |
| 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. |
| MOCA | Abstract | C00590 IPCAT 00 | Monte Carlo Criticality Code System for Hexagonal Geometries. |
| MOCUP | Abstract | P00365 DALPU 00 | MCNP/ORIGEN Coupling Utility Programs. |
| MOMENT I | Abstract | C00188 U1108 00 | Moments Method Neutron Transport Code System. |
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. |
| 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. |
| 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. |
| 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. |
| 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. |
| NAAPRO | Abstract | C00722 PC586 00 | Neutron Activation Analysis PRognosis and Optimization Code System. |
| NACT | Abstract | C00502 U1100 00 | Screening Program for Neutron Activation Products. |
| NAP | Abstract | C00101 I7090 00 | Multigroup Time-Dependent Neutron Activation Prediction Code. |
| NITRAN | Abstract | C00582 FM380 00 | Neutron Transport Code System Based On Anisotropic Scattering. |
| NMTC/JAM | Abstract | C00717 PC586 00 | High Energy Particle Transport Code System. |
| NRN | Abstract | C00054 C6600 00 | Multigroup Removal-Diffusion Code System for Planes, Cylinders and Spheres. |
| 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. |
| NUCWIZ | Abstract | P00616 PCX86 00 | NucWiz |
| NUFACE | Abstract | P00284 CYXMP 00 | An Interface Code For The Calculation of Nuclear Responses. |
| NX1-NX2 | Abstract | P00310 D0VAX 00 | Code System to Calculate Excitation Functions for (n,charged particle) Reactions. |
| O5R | Abstract | C00017 I3675 00 | A General-Purpose Monte Carlo Neutron Transport Code System. |
| O6R | Abstract | C00128 I3675 00 | A General-Purpose Monte Carlo Transport Code System. |
| 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. |
| 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. |
| ORIP_XXI | Abstract | C00731 PC586 02 | Computer Programs for Isotope Transmutation Simulations. |
| ORPHEE VI | Abstract | C00159 I3675 00 | Kernel Integration Code System - Attenuation of Fast Neutrons in Cylindrical Layers of Water and Dense Material. |
| OZMA | Abstract | C00406 I0370 00 | Calculation of Resonance Reaction Rates in Reactor Lattices Using Resonance Profile Tabulations. |
| 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. |
| PATCH-7 | Abstract | C00243 C0074 00 | Three-Dimensional Kernel Integration Code-Explicit Single Scattering Option. |
| PEQAG-2 | Abstract | P00293 IPCAT 00 | A Pre-equilibrium Computer Code With Gamma Emission. |
| PIGG | Abstract | C00138 C3600 00 | A Multigroup One-Dimensional P-1 Radiation Transport Code System. |
| PREMOR | Abstract | C00369 I0360 00 | A Point Reactor Exposure Code System for Survey Nuclear Analysis of Power Plant Performance. |
| PRIMEDANA-2 | Abstract | C00490 I3081 00 | Collapses Multigroup Cross Sections and Obtains Reaction Parameters by Solving Transport or Diffusion Equations. |
| PROB | Abstract | C00287 I0370 00 | Multigroup One-Dimensional Transport Code System, Collision Probability Method. |
| PSU-LEOPARD/RBI | Abstract | C00563 IBMPC 01 | A Spectrum Dependent Non-Spatial Depletion Code. |
| 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. |
| QAD-P5 | Abstract | C00048 C6400 00 | Kernel Integration Code System. |
| QAD-UE | Abstract | C00448 H6000 00 | Kernel Integration Code System. |
| 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. |
| RADHEAT-V4 | Abstract | C00300 FM380 00 | A Code System To Generate Multigroup Constants and Analyze Radiation Transport for Shielding Safety Evaluation. |
| 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. |
| RASC-2D | Abstract | C00318 I0370 00 | Two-Dimensional Removal Diffusion Code Reactor Shielding Design Code System. |
| REACTORSHIELDING-NMS | Abstract | M00014 MNYCP 00 | REACTORSHIELDING-NMS, Reactor Shielding for Nuclear Engineers. |
| REDIFFUSION | Abstract | C00347 I0360 00 | One-Dimensional Neutron Removal-Diffusion and Gamma-Ray Kernel Integration or Diffusion Theory Calculator. |
| REFIT-2009 | Abstract | C00775 PCX86 00 | Multilevel Resonance Parameter Least Square Fit of Neutron Transmission, Capture, Fission & Self Indication Data. |
| 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. |
| 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. |
| RICANT | Abstract | C00569 D8810 00 | A Computer Code for 2-D Transport Calculations in x-y Geometry Using the Interface Current Method. |
| RMET21 | Abstract | C00597 D0VAX 00 | Detailed Space and Energy Treatment of Neutron Resonances for Homogeneous Mixtures and Cylinderized Reactor Cells. |
| ROCKWELL-RSDM | Abstract | M00017 MNYCP 00 | Reactor Shielding Design Manual by Rockwell T. III. |
| RSYST | Abstract | C00269 I0360 00 | Integrated Modular Code System for Shielding and Reactor Physics Calculations. |
| 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. |
| 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. |
| 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. |
| SAP N-G | Abstract | C00092 I7094 00 | Neutron and Gamma-Ray Albedo Model Scatter Shield Analysis Code System. |
| SC2N3N | Abstract | P00309 D0VAX 00 | Systematics of (n,2n) and (n,3n) Cross Sections. |
| 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. |
| SCORE-4 | Abstract | C00234 I0370 00 | Two-Dimensional Multigroup Removal-Diffusion Shielding Code System. |
| SELFS-3 | Abstract | P00551 C6600 00 | Self-Shielding Correlation of Foil Activation Neutron Spectra Analysis by SAND-II. |
| 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. |
| SHADOK | Abstract | C00216 C6600 00 | Transport Code Systems, P1 Scattering in Infinite Cylindrical and Spherical Geometries by Polynomial Approximation. |
| SHREDI | Abstract | C00284 I0360 00 | Multigroup Two-Dimensional (x-y, r-o geometry) Neutron Removal-Diffusion (Spinney Method) Shielding Code System. |
| SIXTUS-3 | Abstract | C00609 MFMWS 00 | Three-Dimensional, Nodal, Neutron Diffusion Criticality Code System in Hex-Z Geometry. |
| 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. |
| 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. |
| SMAUG-13 | Abstract | C00194 C6600 00 | Calculation of Neutron and Prompt Gamma-Ray Doses Resulting from an Atmospheric Nuclear Detonation. |
| 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. |
| SNOW | Abstract | C00282 I0360 00 | Two-Dimensional Discrete Ordinates Multigroup Transport Code System in Plane and Cylindrical Geometry with Isotropic and Anisotropic Scattering. |
| 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. |
| 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. |
| STRAINT | Abstract | C00259 I0360 00 | One-Dimensional Multigroup Neutron Transport Discrete Ordinates Code System. |
| 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. |
| TALYS-1.2 | Abstract | P00548 PC586 01 | Nuclear Model Code System for Analysis and Prediction of Nuclear Reactions and Generation of Nuclear Data. |
| 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. |
| TEMPEST-2 | Abstract | P00558 I0360 00 | Thermalization Program for Neutron Spectra and MultiGroup Cross-Sections. |
| TESS | Abstract | C00215 C3600 00 | Multigroup Discrete Ordinates Code System for Slab and Spherical Geometries. |
| THIDA-2 | Abstract | C00410 FM380 00 | Code System for the Calculation of Transmutation, Activation, Decay Heat and Dose Rate in Fusion Reactors. |
| 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. |
| TNG1 | Abstract | P00298 D6220 00 | A Multistep Statistical Model Based on the Hauser-Feshbach Theory For The Evaluation Of Neutron Data. |
| 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. |
| 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. |
| TRANZIT | Abstract | C00172 C7600 00 | Multigroup Time-Dependent Discrete Ordinates Radiation Transport Code System in (rho,z) Cylindrical Geometry. |
| 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. |
| TREEDE | Abstract | C00326 C0000 00 | Monte Carlo Neutron Transport Code System Based on the Track Rotation Estimator. |
| 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. |
| 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. |
| 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. |
| UMG 3.3 | Abstract | P00529 PC586 00 | Unfolding with Maxed and Gravel. |
| UNIFY-ECN | Abstract | P00288 C0170 00 | A Program to Calculate Fast Neutron Data for Structural Materials. |
| 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. |
| VCS | Abstract | C00262 I0360 00 | Coupled Discrete Ordinates-Adjoint Monte Carlo Calculation of Radiation Protection Factors in Vehicles. |
| VENTURE-PC | Abstract | C00654 PC586 02 | A Reactor Analysis Code 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. |
| VSOP94 | Abstract | C00670 MNYWS 00 | Computer Code System for Reactor Physics and Fuel Cycle Simulation. |
| WIMKAL-88 | Abstract | D00193 MNYCP 00 | 69 Energy Group, Neutron Cross Section Library For Thermal Reactor Calculations in WIMSD Format. |
| 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. |