| psr-0315 | AMPX-77, Modular System for Coupled Neutron-Gamma MultiGroup Cross-Sections from ENDF/B-5 |
| ccc-0082 | ANISN-E, 1-D Transport Program ANISN with Exponential Model |
| nea-0363 | ANISN-FONTENAY, 1-D Planar, Spherical, Cylindrical Neutron Transport and Gamma Transport with Deep Penetration |
| ccc-0254 | ANISN-ORNL, 1-D Neutron Transport & Gamma Transport in Slab, Cylindrical, Spherical Geometry with Anisotropic Scattering |
| ccc-0255 | ANISN-W, 1-D Transport Calculation for Deep Penetration Problems |
| ccc-0514 | ANISN/PC, MultiGroup 1-D Discrete Ordinates Transport with Anisotropic Scattering |
| ccc-0459 | BOLD/VENTURE-4, Reactor Analysis System with Sensitivity and Burnup |
| nea-1678 | BOT3P5.4, 3D Mesh Generator and Graphical Display of Geometry for Radiation Transport Codes, Display of Results |
| nesc0387 | CITATION, 3-D MultiGroup Diffusion with 1st Order Perturbation and Criticality Search |
| ccc-0643 | CITATION-LDI2, 2-D MultiGroup Diffusion, Perturbation, Criticality Search, for PC |
| ccc-0726 | CNCSN 2009, One, Two- and Three-Dimensional Coupled Neutral and Charged Particle Sn Parallel Multi-Threaded Code System |
| ccc-0829 | COG11.1, Multiparticle Monte Carlo Code System for Shielding and Criticality Use |
| iaea1226 | CORD-2, PWR Core Design and Fuel Management |
| nea-1903 | CRISTAL V2.0.3, Criticality calculation package |
| ccc-0547 | DANTSYS3.0, 1-D, 2-D, 3-D MultiGroup Discrete Ordinate Method Transport |
| ccc-0649 | DIF3D 8.0/VARIANT8.0, 2-D 3-D Multigroup Diffusion/Transport Theory Nodal & Finite Difference Solver, Variational Method |
| ccc-0784 | DIF3D10.0, Variational Nodal Methods, Finite Difference Methods to Solve N diffusion & Transport Theory Problems |
| nea-0391 | DLS, 2-D Diffusion with Line-of-Sight Method for Cavities |
| ccc-0650 | DOORS3.2A, 1-,2-,3-dimensional discrete-ordinates system for deep-penetration neutron and photon transport |
| ccc-0276 | DOT-3.5, 2-D Neutron Transport, Gamma Transport Program DOT with New Space-Scaling |
| ccc-0320 | DOT-4.2, 2-D Neutron Transport, Gamma Transport with Space Dependent Mesh and Quadrature |
| uscd1234 | DRAGON 5.1, Reactor Cell Calculation System with Burnup |
| nea-0322 | DTF4-J, 1-D Neutron Transport with Anisotropic Scattering by Sn Method |
| nea-1683 | ERANOS 2.3N, Modular code and data system for fast reactor neutronics analyses |
| nesc0156 | EXTERMINATOR-2, 2-D MultiGroup Neutron Diffusion in X-Y R-Z or R-Theta Geometry |
| nea-0443 | FEM-2D, 2-D MultiGroup Diffusion in X-Y Geometry |
| nea-0545 | FEM-BABEL, 3-D MultiGroup Neutron Diffusion by Galerkin Method |
| nea-0896 | FINELM, MultiGroup Diffusion in 3-D by Finite Elements Method |
| nesc0380 | GATT, 3-D Few-Group Neutron Diffusion for Power Distribution in Hexagonal Reactor Core for HTGR |
| iaea1271 | GNOMER, Core Power Distribution by 1-D, 2-D, 3-D MultiGroup Neutron Diffusion |
| nesc0277 | HAMMER, 1-D MultiGroup Neutron Transport Infinite System Cell Calculation for Few-Group Diffusion Calculation |
| nea-0624 | JOSHUA, Neutronics, Hydraulics, Burnup, Refuelling of LWR |
| ccc-0548 | KENO5A-PC, Monte-Carlo Criticality with Supergrouping |
| uscd1241 | MCART, solve the time dependent neutron transport equation |
| nea-1643 | MCB1C, Monte-Carlo Continuous Energy Burnup Code |
| nea-1733 | MCNP4B-GN, Monte Carlo Code System for (gamma,n) production and transport in high-Z materials |
| iaea0889 | MCRAC/RBI, In Core Fuel Management, Program of PFMP System |
| ccc-0841 | MMS3D, Method of Manufactured Solutions for 3D one-group SN Equations with escalating order of non-smoothness |
| nea-0527 | MONK, Keff, Collision Rate, Flux Distribution in General Geometry from UKNDL by Monte-Carlo Method |
| nea-1905 | MORET 5.D.1, Monte Carlo simulation tool to solve transport equation for neutrons |
| nea-1633 | MOSRA-LIGHT, High Speed 3-D X-Y-Z Nodal Diffusion Code for Vector Computers |
| nea-1896 | MOSRA-SRAC, Lattice Calculation Module of the Modular Code System for Nuclear Reactor Analyses MOSRA |
| nea-1673 | MVP/GMVP V.3, MC Codes for Neutron & Photon Transport Calc. based on Continuous Energy and Multigroup Methods |
| ccc-0641 | NESTLE 5.2.1, Few-Group Neutron Diffusion for Steady-State and Transient Problems by Nodal Expansion Method (NEM) |
| nea-1591 | OMEGA, Subcritical and Critical Neutron Transport in General 3-D Geometry by Monte-Carlo |
| nea-1324 | OREST, LWR Burnup Simulation Using Program HAMMER and ORIGEN |
| ccc-0760 | PARTISN 5.97, 1-D, 2-D, 3-D Time-Dependent, Multigroup Deterministic Parallel Neutral Particle Transport Code |
| ccc-0842 | PARTISN 8.29, Time-Dependent, Parallel Neutral Particle Transport Code System |
| ccc-0708 | REBUS-PC 1.4, Code System for Analysis of Research Reactor Fuel Cycles |
| nea-1840 | SERPENT 1.1.7, 3-D continuous-energy Monte Carlo reactor physics burnup calculation, lattice physics applications |
| nea-1923 | SERPENT V2.2.X, 3-D continuous-energy Monte Carlo reactor physics burnup calculation, lattice physics applications |
| iaea1437 | SUPERMC 3.3.0, Super Monte Carlo simulation program for nuclear and radiation process |
| ccc-0204 | SWANLAKE, Cross-Sections Sensitivity Analysis for 1-D Discrete Ordinate Calculation |
| ccc-0638 | TART2022, 3D Coupled Neutron-Photon Combinatorial Geometry, Time Dependent, Monte Carlo Transport Code |
| nesc0558 | TASK, 1-D MultiGroup Diffusion or Transport Theory Reactor Kinetics with Delayed Neutron |
| ccc-0759 | TITAN 1.29, A Three-Dimensional Deterministic Radiation Transport Code System |
| ccc-0543 | TORT-DORT, 1-D 2-D 3-D Discrete Ordinate Neutron and Photon Transport with Deep Penetration |
| nea-1716 | TRIPOLI-4 VERS. 8.1, 3D general purpose continuous energy Monte Carlo Transport code |
| nea-1940 | TRIPOLI-4.12, Coupled Neutron, Photon, Electron, Positron 3-D, Time Dependent Monte-Carlo Transport Calculation |
| nea-1086 | TRISTAN, 3-D fixed source radiation transport |
| nea-0415 | TRITON, 3-D Multi-Region Neutron Diffusion Burnup with Criticality Search |
| uscd1239 | VENTEASY, Criticality Search for a Desired Keffective by Adjusting Dimensions, Nuclide Concentrations, or Buckling |
| ccc-0654 | VENTURE-PC 1.1, Reactor Analysis System with Sensitivity and Burnup |
| nea-1856 | VESTA 2.1&AURORA1.0, Monte Carlo depletion interface code and AURORA 1.0.0, Depletion analysis tool |
| ccc-0754 | VIM 5.1, Steady-State 3-D Neutron Transport Using ENDF/B or Multigroup Cross Sections |
| iaea0871 | VPI-NECM, Nuclear Engineering Program Collection for College Training |
| nea-0655 | VSOP, Neutron Spectra, 2-D Flux Synthesis, Fuel Management, Thermohydraulics Calculation |
| iaea1440 | VSOP99-11, Neutron Spectra, 2-D Flux Synthesis, Fuel Management, Thermohydraulics Calculation |
| nea-1882 | XSUN-2023, Windows interface environment for transport and sensitivity-uncertainty software TRANSX-2, PARTISN and SUSD3D |
| nea-1206 | ZZ MATXS70-JEFF87, 69+1 Group MATXS Library in WIMS BOXER Structure |
| nea-1264 | ZZ VITAMIN J/COVA, Covariance Matrix Data Library for Uncertainty Analysis |
