| ccc-0612 | ALPHN, (Alpha, N) Neutron Production in High-Level Waste Canisters |
| nea-0403 | AMARA, Correlation of Nuclear Data to Integral Experiment by Lagrange Multipliers |
| psr-0315 | AMPX-77, Modular System for Coupled Neutron-Gamma MultiGroup Cross-Sections from ENDF/B-5 |
| nea-1798 | ANGELO-LAMBDA, Covariance matrix interpolation and mathematical verification |
| ccc-0657 | BETA-S 6, Multi-Group Beta-Ray Spectra |
| nea-1278 | CALENDF-2010, Pointwise, Multigroup Neutron Cross-Sections and Probability Tables from ENDF/B Evaluations |
| nea-1516 | DANCOFF-MC, Dancoff Correlation for Arbitrary Lattices by Monte-Carlo |
| uscd1234 | DRAGON 5.1, Reactor Cell Calculation System with Burnup |
| uscd1237 | DRAGON2PARTISN, Cross-Sections Data Generation for PARTISN4.0 |
| nea-1683 | ERANOS 2.3N, Modular code and data system for fast reactor neutronics analyses |
| nea-1676 | ERRORJ-2.3, Multigroup covariance matrices generation from ENDF-6 format |
| nea-1890 | FISPACT-II 5.X, Inventory Simulation Platform for Nuclear Observables and Materials Science |
| nea-1907 | FRENDY V2, Nuclear Data Processing System for Evaluated Nuclear Data File |
| nea-1896 | MOSRA-SRAC, Lattice Calculation Module of the Modular Code System for Nuclear Reactor Analyses MOSRA |
| psr-0480 | NJOY99.24, Data Processing System of Evaluated Nuclear Data Files ENDF Format |
| psr-0534 | PUFF-IV 6.1.0, Code System to Generate Multigroup Covariance Matrices from ENDF/B-VI Uncertainty Files |
| ccc-0826 | SCEPTRE 1.7, Sandia Computational Engine for Particle Transport for Radiation Effects |
| 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 |
| ccc-0661 | SOURCES-4C, Calculating Alpha, N, Fission, Delayed Neutron Sources and Spectra |
| psr-0317 | TRANSX-2.15, Neutron Gamma Particle Transport Tables from MATXS Format Cross-Sections |
| nea-0655 | VSOP, Neutron Spectra, 2-D Flux Synthesis, Fuel Management, Thermohydraulics Calculation |
| ccc-0698 | WIMS-ANL 4.0, Deterministic Code System for Lattice Calculation |
| nea-0329 | WIMS-D/4, MultiGroup Reactor Lattice Calculation for Thermal Reactor and Fast Reactor |
| nea-1507 | WIMSD5, Deterministic Multigroup Reactor Lattice Calculations |
| nea-1882 | XSUN-2023, Windows interface environment for transport and sensitivity-uncertainty software TRANSX-2, PARTISN and SUSD3D |
| nea-0878 | ZZ GAMDAT-78, Gamma Decay Data of Radioisotopes |
| nea-0796 | ZZ JFS-V2., Cross-Sections Library 25-Groups ABBN and 70-Group JFS for Fast Reactor Calculation |
