Computer Programs

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nesc0374 | 1-DX, 1-D Diffusion for Fast Reactor MultiGroup Cross-Sections, Group Constant Collapsing |

psr-0190 | ADENA, Fission Products Beta Spectra and Gamma Spectra in 19 Group from U235 Pu239 Mixture |

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-0519 | AUS, Neutron Transport and Gamma Transport System for Fission Reactors and Fusion Reactors |

nea-0636 | BASKER, Isotropic Scattering Kernel Calculation Using VIWI |

ccc-0657 | BETA-S, Multi-Group Beta-Ray Spectra |

psr-0117 | BINX, MINX Utility and SPHINX Utility, BCD to BIN Library Conversion |

nea-1523 | BOXER, Fine-flux cross section condensation, 2D few group diffusion and transport burnup calculations |

nea-1278 | CALENDF-2010, Pointwise, Multigroup Neutron Cross-Sections and Probability Tables from ENDF/B Evaluations |

psr-0117 | CINX, MINX Utility and SPHINX Utility, Library Data Collapsing |

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

nea-0357 | CLUP-77C, Collision Probability and Neutron Flux Distribution in Multi-Region Fuel Cluster |

nea-0294 | CODAC, MultiGroup Cross-Sections Generation from ENDF/B for Monte-Carlo Program TIMOC |

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

nea-0325 | CONDENSE, Conversion of JAERI Fast-Set to ABBN Format with Self-Shielding |

nea-0151 | DANCOFF-3, Dancoff Correction for Cylindrical Fuel Rod at H2O Gaps and for Fuel Clusters |

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

nea-0646 | DASQHE, Dancoff Correlation for Infinite Circular Rod Assembly in Square or Hexagonal Lattice |

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

ccc-0647 | DRAGON, Reactor Cell Calculation System with Burnup |

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

nea-0817 | ENTOSAN, 640 Group Constant Calculation with Resonance from ENDF/B |

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

nea-1683 | ERANOS 2.3, Modular code and data system for fast reactor neutronics analyses |

nea-1676 | ERRORJ, 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-0394 | ETOA, ABBN MultiGroup Constants from ENDF/B for Fast Reactors |

nea-0311 | EXPANDA-4, 1-D Neutron Diffusion in Slab, Spherical Geometry from JAERI Fast-Set with Criticality Calculation |

nea-0312 | EXPANDA-5, 1-D Neutron Diffusion in Slab, Cylindrical, Spherical Geometry for 2 Region Fast Reactor Criticality |

nea-0313 | EXPANDA-6, 1-D Neutron Diffusion in Slab, Cylindrical, Spherical Geometry from JAERI Fast-Set with PuO2 UO2 Mixture |

iaea0830 | FEDGROUP, Group Constant Library from ENDF/B, KEDAK, UKNDL |

nea-1890 | FISPACT-II 4.0, Inventory Simulation Platform for Nuclear Observables and Materials Science |

nea-0844 | FISPET, MultiGroup Fission Spectra Calculation from ENDF/B |

nea-0894 | FITOCO, Fine Group to Coarse Group Neutron Flux Conversion for Spectra Unfolding |

nea-0636 | FLAKER, Legendre Moments from Scattering Law Tables |

nea-0810 | FORM-OTA, MultiGroup Constant for Epithermal Neutron Slowing-Down in Homogeneous Media |

nesc0033 | GAM, Slowing-Down Neutron Spectra in P1 and B1 Approximation, MultiGroup Constant |

ccc-0042 | GAMLEG-JR, MultiGroup Gamma Cross-Sections, Energy Absorption Coefficient Generator for Transport Calculation |

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

iaea0849 | GROUPIE2010, Bondarenko Self-Shielded Cross Sections from ENDF/B |

iaea1222 | HAMCIND, Cell Burnup with Fission Products Poisoning |

nesc0277 | HAMMER, 1-D MultiGroup Neutron Transport Infinite System Cell Calculation for Few-Group Diffusion Calculation |

iaea1253 | HOMO, Homogenization of ANISN and DOT Condensed Cross-Sections Output |

nea-0329 | ICAROG, WIMS-D/4 Library Utility |

nea-0744 | INTEGR, Escape Transmission Probability in 1-D Cylindrical Geometry for Program FACEL |

nea-0513 | IRESINT-3, Resonance Absorption in Square or Hexagonal Lattice by Single-Level Breit-Wigner |

nea-0317 | JFUSER, JAERI Fast-Set Group Constant Collapsing and Data Conversion for Program LTFR-70 |

nea-0624 | JOSHUA, Neutronics, Hydraulics, Burnup, Refuelling of LWR |

nea-0578 | KEMA, KEDAK Utility, Data Update |

nea-0616 | KIM, Steady-State Transport for Fixed Source in 2-D Thermal Reactor by Monte-Carlo |

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

nesc0249 | LASER, Slowing-Down Neutron Spectra and Burnup for Thermal Reactors, Neutron Transport Theory |

nea-0573 | LASER-PNC, Neutron Spectra in Uniform Lattice with Burnup Calculation |

nesc0279 | LEOPARD, Fast and Thermal Neutron Spectra from Temperature and Geometry with Depletion Calculation |

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

psr-0117 | LINX, MINX Library Utility, Data Merge |

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

psr-0132 | MACK, Fluence to Kerma Generator from ENDF/B |

nea-0528 | MARC, MultiGroup Diffusion in R-Z, X-Y, R-Theta, Slab, Cylindrical, Spherical, Hexagonal, Triangular Geometry |

nea-1017 | MARCOPOLO, Radial and Axial Diffusion Coefficient for Cylindrical Wigner-Seitz Reactor Cell |

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

nesc0355 | MC**2-2, MultiGroup Neutron Spectra, Slowing-Down Calculation Using ENDF/B, P1 and B1 Approximation |

nea-0452 | MCDATA, MC**2 Cross-Sections Conversion for Programs CITATION, ANISN, DOT |

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

nea-0388 | MIGROS, Group Cross-Sections, Self-Shielding Factors, Scattering Transfer, Fission from KEDAK |

nea-0639 | MINIGAL, Average Thermal Cross-Sections, Epithermal Cross-Sections, Fission Cross-Sections from UKNDL |

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

psr-0142 | MORSEC-SP, Step Function Angular Distribution for Cross-Sections Calculation by Program MORSE |

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

nea-0035 | MUPO, Critical 43 Group Spectra Calculation for Homogeneous Reactor |

iaea0863 | NANICK, Infinitely Dilute Group Constant and Scattering Matrix from ENDF/B |

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-1238 | PASC-1, Petten AMPX-II/SCALE-3 Code System for Reactor Neutronics Calculation |

psr-0106 | PLASMX, MultiGroup Neutral Particle Transport in Tokamak CTR Plasma |

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

nea-1170 | PROF-DD, Generator of MultiGroup Cross-Sections Library DDX for MORSE-DD, ANISN-DD, DOT-DD |

iaea0888 | PSU-LEOPARD, Program LEOPARD in PFMP System, Fast Neutron and Thermal Neutron Spectra Calculation |

psr-0157 | PUFF-2, MultiGroup Covariance Matrices from ENDF/B-5 Error Files |

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

nesc0281 | RABBLE, Cross-Sections from Single-Level Resonance Parameter, Homogeneous or Heterogeneous Infinite System |

nea-0262 | REFLOS, Fuel Loading and Cost from Burnup and Heavy Atomic Mass Flow Calculation in HWR |

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

nea-0234 | RICM, Resonance Absorption in Multi-Region Slab or Square or Hexagonal Lattice |

nesc0213 | RIFF-RAFF, Resonance Integrals in 2 Region Cell, Isotropic Flux and Isotropic Scattering |

nea-1449 | ROLAIDS-CPM, 1-D Slowing-Down by Collision Problems Method |

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

ccc-0834 | SCALE 6.2.3, 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 |

ccc-0405 | SENSIT, Integral Response Sensitivity from Neutron Cross-Sections, Gamma Cross-Sections Errors |

nea-1840 | SERPENT 1.1.7, 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 |

nesc0279 | SPOTS, Library Generator for Program LEOPARD from Cross-Sections Data |

psr-0013 | SUPERTOG, MultiGroup Cross-Sections Generator from ENDF/B for Programs GAM, ANISN, DOT |

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

nea-0634 | THERLIB, Library Generated for THERMOS from FACEL Library |

nea-0634 | THERMLIB, Generator and Edit of Program THERMOS-OTA Library |

nea-0043 | THERMOL, Space-Dependent Thermal Flux in 1-D Slab or Cylinder |

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

nea-0628 | THERMOS-OTA, Thermal Flux by Integral Transport |

nea-0804 | TIMS-1, MultiGroup Cross-Sections of Heavy Isotope Mixture with Resonance from ENDF/B |

psr-0317 | TRANSX-2.15, Neutron Gamma Particle Transport Tables from MATXS Format Cross-Sections |

nea-0636 | VIWI, Neutron Speeds and Weights for Scattering Kernel Calculation |

nea-0655 | VSOP, Neutron Spectra, 2-D Flux Synthesis, Fuel Management, Thermohydraulics Calculation |

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

iaea0946 | WILMA, WIMS Nuclear Data Library Maintenance |

nea-0329 | WIMS, MultiGroup Reactor Lattice Calculation for Thermal Reactor and Fast Reactor |

ccc-0698 | WIMS-ANL 4.0, Deterministic Code System for Lattice Calculation |

iaea0887 | WIMSCORE, 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 |

nesc0393 | XSDRN, MultiGroup Cross-Sections from Resonance Data Library, Neutron Spectra and Group Constant Collapsing |

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-1, Cross-Sections Library 25-Groups ABBN and 70-Group JFS for Fast Reactor Calculation |

nea-0796 | ZZ JFS-2, 25 Group (ABBN) and 70 Group JFS Cross Sections Library for Fast Reactors |

nea-0796 | ZZ JFS-3/J2, 70 Group 30 Isotopes Cross Section Library for Fast Reactors |

iaea1235 | ZZ PNESD, Diffusion Elastic Scattering Cross-Section of 3 MeV to 1000 MeV Proton on Natural Isotopes |

nea-1518 | ZZ WIMKAL-88, 69-Group KAERI WIMS Library for Thermal Reactors |

nea-0329 | ZZ WIMS-TRIGA, ZZ-WIMSLIB/IJS, WIMS Data Libraries |