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-1564	EASY-2010, European Neutron Activation System
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 2.0, 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-0449	ESTOQ, 1st Collision Source Calculation for Program DOT in R-Z Geometry
nea-0984	ETHEL, Thermos Cross-Sections Library Generator Program
nea-0394	ETOA, ABBN MultiGroup Constants from ENDF/B for Fast Reactors
nea-0893	EVGRP, Photo Production MultiGroup Cross-Sections Generated from ENDF/B-4
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-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
nea-0867	FOURACES, MultiGroup Cross-Sections, Resonance Calculation from ENDF/B, KEDAK, UKNDL
nesc0033	GAM, Slowing-Down Neutron Spectra in P1 and B1 Approximation, MultiGroup Constant
nesc0547	GAMB-1T, Group Constant Library from P1 or B1 Approximation Neutron Spectra in ANISN Format, DOT Format
ccc-0042	GAMLEG-JR, MultiGroup Gamma Cross-Sections, Energy Absorption Coefficient Generator for Transport Calculation
nesc0185	GAMTEC-2, MultiGroup Constant for Homogeneous or Heterogeneous Core
nesc0298	GGC-4, MultiGroup Neutron Spectra and Broad Group Cross-Sections Calculation, P1, B1, B2, B3 Approximation
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
iaea1330	HEATER, Reaction Rate Tables from Cross-Sections with Weighting
iaea1253	HOMO, Homogenization of ANISN and DOT Condensed Cross-Sections Output
nesc0467	HRG-3, Slowing-Down Neutron Spectra Using P1 and B1 Approximation with Average Cross-Sections Calculation
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-0035	MUPO, Critical 43 Group Spectra Calculation for Homogeneous Reactor
iaea0890	MURLI, 1-D Flux, Reaction Rate in Cylindrical Geometry Thermal Reactor Lattice by Transport
iaea0892	MURLI-CLUSTER, Lattice Calculation of PHWR with Rod Clustered Fuel
iaea0863	NANICK, Infinitely Dilute Group Constant and Scattering Matrix from ENDF/B
psr-0355	NJOY-94, General ENDF/B Processing System for Reactor Design Problems
psr-0368	NJOY-97, General ENDF/B Processing System for Reactor Design Problems
psr-0171	NJOY91, General ENDF/B Processing System for Reactor Design Problems
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
iaea0817	PROB, Transport Equation in Slab Geometry and Collision Probability by Overrelaxation Method
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
iaea0822	RAM-1, Thermal Flux Derivatives at Plane Geometry Control Rod Boundary by Monte-Carlo
nea-0262	REFLOS, Fuel Loading and Cost from Burnup and Heavy Atomic Mass Flow Calculation in HWR
iaea0935	REX1-87, MultiGroup Neutron Cross-Sections from ENDF/B
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-0785	SCALE 6.1.2, Modular system for criticality, shielding, source term, fuel depletion/decay, inventories, reactor physics
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
nea-0842	SRAC-95, Cell Calculation with Burnup, Fuel Management for Thermal Reactors
psr-0013	SUPERTOG, MultiGroup Cross-Sections Generator from ENDF/B for Programs GAM, ANISN, DOT
iaea0894	SUPERTOG-LTT, SUPERTOG with Tabular Elastic Scattering Anisotropy from ENDL
ccc-0204	SWANLAKE, Cross-Sections Sensitivity Analysis for 1-D Discrete Ordinate Calculation
nesc0050	TEMPEST-2, Thermalization Program for Neutron Spectra and MultiGroup Cross-Sections
nea-0634	THERLIB, Library Generated for THERMOS from FACEL Library
nea-0634	THERMLIB, Generator and Edit of Program THERMOS-OTA Library
nesc0184	THERMOS BRT-1, 1-D Integral Transport for Neutron Spectra, Slab and Cylinder
nea-0043	THERMOS, Space-Dependent Thermal Flux in 1-D Slab or Cylinder
nesc0184	THERMOS-ANL, 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
iaea0821	WELWING, Material Buckling for HWR with Annular Fuel Elements
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-0886	ZZ AMPX-2/123, 123-Group Neutron Cross-Section Library from ENDF/B-4 by AMPX-2
nea-0886	ZZ AMPX-2/219, 219-Group Neutron Cross-Section Library from ENDF/B-4 by AMPX-2
dlc-0154	ZZ ANSLV, Multigroup Cross Sections Library for ANS Reactor Design Studies
iaea1237	ZZ BARC-75, Coupled 50 Neutron-Group 25 Gamma-Group Cross-Section Library for 42 Nuclides
iaea0949	ZZ BROND, Evaluated Neutron Data Library in ENDF-6 Format
iaea1256	ZZ CENPL, Chinese Evaluated Nuclear Parameter Library
iaea1256	ZZ CENPL-DLS, Discrete Level Schemes and Gamma Branching Ratios Library
iaea1256	ZZ CENPL-FBP, Fission Barrier Paramater Library
iaea1256	ZZ CENPL-GDRP, Giant Dispole Resonance Parameter Library
iaea1256	ZZ CENPL-MCC, Nuclear Ground State Atomic Masses Library
iaea1256	ZZ CENPL-NLD, Nuclear Level Density Parameter Library
iaea1256	ZZ CENPL-OMP, Optical Model Parameter Library
iaea1297	ZZ CL50G, 50-Group Multigroup Library in AMPX Format for Fast Reactor Calculation
dlc-0077	ZZ COVERV, Multigroup Cross-Section Covariance Matrices in COVERX Format
dlc-0100	ZZ ELECSPEC, Electron Spectra Data Library from Fission Product Decay
dlc-0103	ZZ ENDL86, Evaluated Charged Particle, Neutron, Photon Cross-Section Library
nea-0878	ZZ GAMDAT-78, Gamma Decay Data of Radioisotopes
nea-1344	ZZ GROUPSTRUCTURES, VITAMIN-J, XMAS, ECCO-33, ECCO2000 Standard Group Structures
iaea1215	ZZ IRAN-LIB, Multigroup Neutron Gamma Cross-Section Library for 33 Elements in ANISN Format
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
dlc-0045	ZZ SENPRO/45C, Multigroup Sensitivity Library for Fast Reactors, Thermal Reactors
nea-1518	ZZ WIMKAL-88, 69-Group KAERI WIMS Library for Thermal Reactors
nea-0329	ZZ WIMS-TRIGA, ZZ-WIMSLIB/IJS, WIMS Data Libraries