Computer Programs

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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 |

iaea1251 | AMICO, Cross-Sections Data for ANISN, DOT from WIMS Library |

psr-0315 | AMPX-77, Modular System for Coupled Neutron-Gamma MultiGroup Cross-Sections from ENDF/B-5 |

nea-1235 | AND, Atomic Number Densities for Criticality Calculation |

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 |

iaea0883 | CLUB, Cell Calculation PF Candu PWR Fuel Clusters |

psr-0286 | COMBINE-PC, MultiGroup Neutron Cross-Sections in B1 or B3 Approximation from ENDF/B-5 |

nea-1516 | DANCOFF-MC, Dancoff Correlation for Arbitrary Lattices by Monte-Carlo |

uscd1234 | DRAGON 3.05D, Reactor Cell Calculation System with Burnup |

uscd1237 | DRAGON2PARTISN, Cross-Sections Data Generation for PARTISN4.0 |

iaea1202 | EQUIVA, Few-Group Diffusion Parameter for PWR Reflector Region by 1-D Transport Calculation |

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-0892 | ESTIMA, Neutron Width Level Spacing, Neutron Strength Function of S- Wave, P-Wave Resonances |

nea-1890 | FISPACT-II 5.0, Inventory Simulation Platform for Nuclear Observables and Materials Science |

nea-1907 | FRENDY V2, Nuclear Data Processing System for Evaluated Nuclear Data File |

nea-0543 | GGTC-ENEL, MultiGroup Neutron Spectra in P1, B1, B2, B3 Approximation and Thermos Calculation |

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 |

psr-0020 | LAPHAN0, P0 Gamma Production Matrices from ENDF/B |

nea-0124 | LGH, Gamma Streaming and Neutron Streaming for Duct |

psr-0233 | LSL-M2, Neutron Spectra Log Adjustment for Dosimetry Applications |

psr-0117 | MARS-ORNL, Processing Program Collection for AMPX, CCCC, ANISN, DOT, MORSE Format Library |

nea-1562 | MICROX-2, Group Constant Generator with Resonance Interference and Self-Shielding |

psr-0105 | MINX, MultiGroup Cross-Sections and Self-Shielding Factors from ENDF/B for Program SPHINX |

nea-1896 | MOSRA-SRAC, Lattice Calculation Module of the Modular Code System for Nuclear Reactor Analyses MOSRA |

psr-0480 | NJOY99, Data Processing System of Evaluated Nuclear Data Files ENDF Format |

iaea1389 | NRSC, Neutron Resonance Spectrum Calculation System |

nea-1347 | NSLINK, Coupling of NJOY Cross-Sections Generator Code to SCALE-3 System |

psr-0156 | PAPIN, Cross Section, Self-Shielding Factors for Fertile Isotopes in Unresolved Resonance Region |

nea-0169 | PROCOPE, Collision Probability in Pin Clusters and Infinite Rod Lattices |

psr-0534 | PUFF-IV 6.1.0, Code System to Generate Multigroup Covariance Matrices from ENDF/B-VI Uncertainty Files |

nesc0453 | RICE, Energy Exchange Matrix, Damage Cross-Sections, Recoil Energy Spectra from ENDF/B |

nea-0598 | RSYST, Modular System for Reactor Core and Shielding Problems |

ccc-0834 | SCALE 6.2.4, A Comprehensive Modeling and Simulation Suite for Nuclear Safety Analysis and Design |

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 V2.2.0 -R-, 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 |

ccc-0204 | SWANLAKE, Cross-Sections Sensitivity Analysis for 1-D Discrete Ordinate Calculation |

nesc0184 | THERMOS-BRT-1, 1-D Integral Transport for Neutron Spectra, Slab and Cylinder |

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 |

iaea1210 | WEDRO-1.1, Data Processing Routines for WIMS-D/4 WIMSE File |

iaea0946 | WILMA, WIMS Nuclear Data Library Maintenance |

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 |

iaea0887 | WIMSCORE-ENEA, 2-Group Constant from WIMS-D/4 for Programs TDB, TRITON, CITATION |

nea-1507 | WIMSD5, Deterministic Multigroup Reactor Lattice Calculations |

iaea1254 | WINTER, Interactive WIMS Input Preparation |

nesc0572 | XLACS, Fast Resonance and Thermal MultiGroup Cross-Sections from ENDF/B, Breit-Wigner, for Program XSDRN |

nea-1882 | XSUN-2017, 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 |