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Catalog of Programs in Category B

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

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