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Catalog of Programs in Category B

B. Spectrum Calculations, Generation of Group Constants and Cell Problems


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