| 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-2005, Pointwise, Multigroup Neutron Cross-Sections and Probability Tables from ENDF/B Evaluations | |
| 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 | |
| iaea0952 | DIGA/NSL, 3 Regions Lattice Cell Neutron Flux Diffusion Calculation | |
| 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-2005.1, 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 | |
| nea-1038 | FAIR-DDX, Double Diffusion Cross-Sections Scattering Matrix Generated from ENDF/B-4 or JENDL-2 | |
| iaea0830 | FEDGROUP, Group Constant Library from ENDF/B, KEDAK, UKNDL | |
| iaea0903 | FEONAN, Flux Smoothing of Spectrometer System | |
| 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-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 | GROUPIE2007, 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-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 | |
| 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, Code System to Generate Multigroup Covariance Matrices from ENDF/B-VI Uncertain-ty 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-0732 | SCALE 5.1/ORIGEN-ARP5.1: Modular system for criticality, shielding, source term, fuel depletion/decay, reactor physics | |
| ccc-0405 | SENSIT, Integral Response Sensitivity from Neutron Cross-Sections, Gamma Cross-Sections Errors | |
| ccc-0661 | SOURCES-4C, Calculating Alpha, N, Fission, Delayed Neutron Sources and Spectra | |
| 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 | |
| iaea1252 | TEST, Sort, Delete, List ANISN and DOT Cross-Sections Library Data | |
| nea-0634 | THERLIB, Library Generated for THERMOS from FACEL 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 | |
| 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-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 | |
| 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 | |
| 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-DLS, Discrete Level Schemes and Gamma Branching Ratios 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 ENDL82, 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 | |
| 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 |