| nesc0325 | 2-DB, 2-D MultiGroup Diffusion, X-Y, R-Theta, Hexagonal Geometry Fast Reactor, Criticality Search | |
| nea-0108 | ALCI, Homogenious 2-D MultiGroup Neutron Diffusion in X-Z, R-Z, R-Theta Geometry with Criticality Search | |
| ccc-0558 | ALKASYS, Rankine-Cycle Space Nuclear Power System | |
| psr-0315 | AMPX-77, Modular System for Coupled Neutron-Gamma MultiGroup Cross-Sections from ENDF/B-5 | |
| nea-0633 | ANIPLO-50, Plot of Scalar Flux and Dose Rates from ANISN Calculation | |
| ccc-0254 | ANISN, 1-D Neutron Transport and Gamma Transport in Slab, Cylindrical, Spherical Geometry with Anisotropic Scattering | |
| ccc-0082 | ANISN-E, 1-D Transport Program ANISN with Exponential Model | |
| nea-0363 | ANISN-FONTENAY, 1-D Planar, Spherical, Cylindrical Neutron Transport and Gamma Transport with Deep Penetration | |
| ccc-0255 | ANISN-W, 1-D Transport Calculation for Deep Penetration Problems | |
| ccc-0514 | ANISN/PC, MultiGroup 1-D Discrete Ordinates Transport with Anisotropic Scattering | |
| nea-0546 | APPLE, Plot of 1-D MultiGroup Neutron Flux and Gamma Flux and Reaction Rates from ANISN | |
| nea-0320 | ARGO, 1-D Neutron Diffusion in Slab, Cylindrical, Spherical Geometry from JAERI Fast-Set, ABBN, RCBN | |
| nea-1006 | ASDIC, Fast Reactor Hexagonal 3 Component Fuel Pin Diffusion Coefficient | |
| nea-0661 | ASNBILD, Generator of JCL and Data for Program ANISN on CDC Computer | |
| ccc-0519 | AUS, Neutron Transport and Gamma Transport System for Fission Reactors and Fusion Reactors | |
| nea-0179 | AXIFLUX, Cosine Function Fit of Experimental Axial Flux in Cylindrical Reactor | |
| ccc-0459 | BOLD/VENTURE-4, Reactor Analysis System with Sensitivity and Burnup | |
| nea-1187 | BOREAS, Nuclear and Thermohydraulic LWR Burnup Simulation | |
| nea-1678 | BOT3P5.3, 3D Mesh Generator and Graphical Display of Geometry for Radiation Transport Codes, Display of Results | |
| nea-1523 | BOXER, Fine-flux cross section condensation, 2D few group diffusion and transport burnup calculations | |
| iaea1190 | BRETISLAV,OLDRICH, Neutron Diffusion in Hexagonal Geometry for WWER Critical Fuel Assemblies | |
| nesc0270 | CAESAR-4, 1-D MultiGroup Diffusion in Slab, Cylindrical, Slab Geometry, Criticality Search | |
| nea-0649 | CARMEN-SYSTEM, Programs System for Thermal Neutron Diffusion and Burnup with Feedback | |
| iaea0920 | CEBIS, 1-D 2 Group Diffusion Code for Reactor Calculation | |
| nesc0387 | CITATION, 3-D MultiGroup Diffusion with 1st Order Perturbation and Criticality Search | |
| ccc-0643 | CITATION-LDI2, 2-D MultiGroup Diffusion, Perturbation, Criticality Search, for PC | |
| nea-0357 | CLUP-77C, Collision Probability and Neutron Flux Distribution in Multi-Region Fuel Cluster | |
| ccc-0726 | CNCSN, One, Two- and Three-Dimensional Coupled Neutral and Charged Particle Sn Code System | |
| ccc-0724 | COG10, Multiparticle Monte Carlo Code System for Shielding and Criticality Use | |
| iaea1226 | CORD, PWR Core Design and Fuel Management | |
| nea-0057 | CRAM-360, 1-D, 2-D Multi Group Diffusion with Keff Calculation or Criticality Search | |
| nea-1416 | D3D/D3E, 2-D, 3-D MultiGroup Neutron Diffusion in Rectangular, Cylindrical, Triangular-Z Geometry | |
| nea-0672 | DIAMANT-2, MultiGroup Neutron Transport with Anisotropic Scattering in Triangular Geometry | |
| ccc-0649 | DIF3D 8.0/VARIANT8.0, 2-D 3-D Multigroup Diffusion/Transport Theory Nodal & Finite Difference Solver, Variational Method | |
| nea-0808 | DIFFUSION-ACE, 3-D Neutron Diffusion by Leakage Iteration Method | |
| nea-0184 | DIXY-2, 2-D Homogeneous and Inhomogeneous Neutron Diffusion N X-Z, R-Z, R-Theta Geometry with Perturbation | |
| nea-0391 | DLS, 2-D Diffusion with Line-of-Sight Method for Cavities | |
| iaea1241 | DNTM/R2D, 2-D Transport in X-Y Geometry | |
| ccc-0650 | DOORS, 1-,2-,3-dimensional discrete-ordinates system for deep-penetration neutron and photon transport | |
| ccc-0543 | DORT, 1-D 2-D Discrete Ordinate Neutron and Photon Transport with Deep Penetration | |
| ccc-0276 | DOT-3.5, 2-D Neutron Transport, Gamma Transport Program DOT with New Space-Scaling | |
| ccc-0320 | DOT-4.2, 2-D Neutron Transport, Gamma Transport with Space Dependent Mesh and Quadrature | |
| nea-1506 | DPOL3D, 2 Group, 3-D Core Transients and Steady State | |
| uscd1234 | DRAGON 3.05D, Reactor Cell Calculation System with Burnup | |
| ccc-0647 | DRAGON, Reactor Cell Calculation System with Burnup | |
| nesc0209 | DTF-4, 1-D MultiGroup Time-Independent Boltzman Equation, Slab, Cylindrical, Spherical Geometry, Sn-Method, Pl-Method | |
| nea-0269 | DTF-G, Reactivity and Flux by 1-D Sn Method on Planar Cylindrical Spherical Geometry | |
| nea-0322 | DTF4-J, 1-D Neutron Transport with Anisotropic Scattering by Sn Method | |
| nea-1683 | ERANOS 2.0, Modular code and data system for fast reactor neutronics analyses | |
| nea-0534 | EREBUS, Burnup by 2-D MultiGroup Neutron Diffusion with Criticality Search | |
| nea-0449 | ESTOQ, 1st Collision Source Calculation for Program DOT in R-Z Geometry | |
| 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-0315 | EXPANDA-70, 1-D Neutron Diffusion in Slab, Cylindrical, Spherical Geometry with Criticality Search | |
| nesc0156 | EXTERMINATOR-2, 2-D MultiGroup Neutron Diffusion in X-Y R-Z or R-Theta Geometry | |
| ccc-0440 | EXTREME, 2-D Discrete Ordinate System with Exponential Space Expansion | |
| iaea0835 | FASVER, 2 Group 2-D Diffusion in X-Y Geometry and Adjoint Solution for PWR with Reflector | |
| nea-0443 | FEM-2D, 2-D MultiGroup Diffusion in X-Y Geometry | |
| nea-0545 | FEM-BABEL, 3-D MultiGroup Neutron Diffusion by Galerkin Method | |
| nea-0566 | FEM-RZ, 2-D MultiGroup Neutron Transport in R-Z Geometry, Eigenvalue and Fixed Source Problems | |
| nea-0478 | FEMB, 2-D Homogeneous Neutron Diffusion in X-Y Geometry with Keff Calculation, Dyadic Fission Matrix | |
| ests0486 | FESH, 2-D Multigroup Neutron Transport with Isotropic Scattering | |
| iaea1221 | FIGA, Source Distribution Conversion from X-Y to R-Theta Geometry | |
| nea-0896 | FINELM, MultiGroup Diffusion in 3-D by Finite Elements Method | |
| nesc0167 | FLARE, 3-D BWR Reactivity and Power Distribution Appraisal Calculation | |
| nea-0596 | FOCUS, Neutron Transport System for Complex Geometry Reactor Core and Shielding Problems by Monte-Carlo | |
| nesc0028 | FOG-1-2-3, 1-D Few-Group Diffusion in Slab Cylindrical Spherical Geometry, Criticality Search, Buckling | |
| ccc-0603 | FPZD, Reactor Burnup by MultiGroup Neutron Diffusion | |
| nea-1021 | FURNACE, Neutronic Calculation in 3-D Toroidal Geometry | |
| nea-1827 | GANAPOL-ABNTT, Analytical Benchmarks for Nuclear Engineering Applications, Case Studies in Neutron Transport Theory | |
| nesc0606 | GAPER-1D, 1-D MultiGroup 1st Order Perturbation Transport Theory for Reactivity Coefficient | |
| nesc0380 | GATT, 3-D Few-Group Neutron Diffusion for Power Distribution in Hexagonal Reactor Core for HTGR | |
| ccc-0628 | GBANISN, ANISN Like 1-D Neutron and Gamma Transport with Group Band Fluxes | |
| nea-0605 | GENP-2, Program System for Integral Reactor Perturbation | |
| psr-0304 | GIRAFFE, Isotope Release Analysis in LMFBR Fuel Elements Failure | |
| iaea1271 | GNOMER, Core Power Distribution by 1-D, 2-D, 3-D MultiGroup Neutron Diffusion | |
| iaea0908 | GRENADE, Green's Function Nodal Algorithm for Diffusion Equation | |
| nesc0277 | HAMMER, 1-D MultiGroup Neutron Transport Infinite System Cell Calculation for Few-Group Diffusion Calculation | |
| nesc0136 | HERESY, 2-D Few-Group Static Eigenvalues Calculation for Thermal Reactor | |
| nea-0176 | HETERO, Flux and Power Distribution in Thermal Reactor by 3-D, 2 Group Line Source Sink Method | |
| iaea1240 | HEXAB-3D, 3-D Few-Group Diffusion for Hexagonal Core Geometry | |
| iaea0914 | HEXNOD23, 2-D, 3-D Coarse Mesh Solution of Steady State Diffusion Equation in Hexagonal Geometry | |
| nea-0624 | JOSHUA, Neutronics, Hydraulics, Burnup, Refuelling of LWR | |
| nea-0343 | KASY, 3-D Homogeneous Neutron Diffusion in X-Y-Z, R-Theta, Hexagonal-Z Geometry by Synthesis Method | |
| nesc0450 | KENO, MultiGroup P1 Scattering Monte-Carlo Transport Calculation for Criticality, Keff, Flux in 3-D | |
| ccc-0510 | KENO-4(RG), KENO-4 with Random Geometry | |
| ccc-0436 | KENO-4/CRC, MultiGroup 3-D P1 Scattering Monte-Carlo Transport Calculation with Neutron Balance Edit | |
| nea-1467 | KENO-VA-PVM KENO-VA-SM, KENO5A for Parallel Processors | |
| ccc-0548 | KENO5A-PC, Monte-Carlo Criticality with Supergrouping | |
| nea-0616 | KIM, Steady-State Transport for Fixed Source in 2-D Thermal Reactor by Monte-Carlo | |
| iaea1232 | LABAN-PEL, 2-D MultiGroup Neutron Diffusion in X-Y Geometry by Response Matrix Eigenvalue | |
| nea-0167 | LIE-PN, Pn Neutron Transport in Radial Geometry Cell with Source Problems Calculation | |
| nea-0836 | MADONNA, Neutron Flux with Void Region by Removal Diffusion Method | |
| nea-0528 | MARC, MultiGroup Diffusion in R-Z, X-Y, R-Theta, Slab, Cylindrical, Spherical, Hexagonal, Triangular Geometry | |
| nea-0926 | MARIA-SYSTEM, PWR Fuel Assembly Calculation for Program CARMEN-System and Simulation | |
| nea-1643 | MCB1C, Monte-Carlo Continuous Energy Burnup Code | |
| ccc-0699 | MCNP-DSP, Monte Carlo Neutron-Particle Transport Code with Digital Signal Processing | |
| ccc-0730 | MCNP/MCNPX, Monte Carlo Particle Transport Code System Including MCNP5 1.40, MCNPX 2.5.0, VISED19L and Data Libraries | |
| nea-1733 | MCNP4B-GN, Monte Carlo Code System for (gamma,n) production and transport in high-Z materials | |
| ccc-0701 | MCNP4C2, Coupled Neutron, Electron Gamma 3-D Time-Dependent Monte Carlo Transport Calculations | |
| ccc-0705 | MCNPX 2.3.0, Monte Carlo Code System for Multiparticle and High Energy Applications | |
| ccc-0715 | MCNPX 2.4.0, Monte Carlo N-Particle Transport Code System for Multiparticle and High Energy Applications. | |
| ccc-0746 | MCNPX 2.6.0, Monte Carlo All-Particle Transport Code System including MCNPDATA and VISED Version X_22S | |
| iaea0889 | MCRAC, In Core Fuel Management, Program of PFMP System | |
| nea-1005 | MOBIDIC, Fast Reactor Hexagonal Infinite Lattice 2 Component Fuel Pin Diffusion Coefficient | |
| iaea1238 | MOCA, Criticality of VVER Reactor Hexagonal Fuel Assemblies | |
| nea-1279 | MODICO, 1-D Time-Dependent 1 Group, 2 Group Neutron Diffusion with Delayed Neutron Precursors | |
| nea-0527 | MONK, Keff, Collision Rate, Flux Distribution in General Geometry from UKNDL by Monte-Carlo Method | |
| nea-1747 | MONTE-CARLO-WS-2005, Proceedings of Monte Carlo Criticality Calculations & TRIPOLI-IV Workshops 2005 | |
| ccc-0127 | MORSE, MultiGroup Neutron Transport and Gamma Transport for Complex Geometry Shields by Monte-Carlo | |
| ccc-0431 | MORSE-C, Neutron Transport, Gamma Transport for Criticality Calculation by Monte-Carlo Method | |
| ccc-0474 | MORSE-CGA, Monte-Carlo Neutron and Gamma MultiGroup Transport with Array Geometry | |
| nea-1181 | MORSE-DD, Monte-Carlo Neutron Transport with Combinatorial Geometry Using DDL Cross-Sections Library | |
| ccc-0588 | MORSE-EMP, Monte-Carlo Neutron and Gamma MultiGroup Transport with Array Geometry, for PC | |
| nea-1633 | MOSRA-LIGHT, High Speed 3-D X-Y-Z Nodal Diffusion Code for Vector Computers | |
| nea-0933 | MULTI-KENO, Criticality Safety Analysis by Monte-Carlo | |
| 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 | |
| nea-1673 | MVP/GMVP II, MC Codes for Neutron & Photon Transport Calc. based on Continuous Energy and Multigroup Methods | |
| iaea1173 | NEHEX-3D, 3-D Neutron Diffusion for Fast Reactors and WWER in Hexagonal Geometry | |
| ccc-0641 | NESTLE, Few-Group Neutron Diffusion for Steady-State and Transient Problems by Nodal Expansion Method (NEM) | |
| psr-0355 | NJOY-94, General ENDF/B Processing System for Reactor Design Problems | |
| psr-0171 | NJOY91, General ENDF/B Processing System for Reactor Design Problems | |
| iaea1171 | NOTRAN/3D, 3-D Neutron Transport in X-Y-Z Geometry by Discrete Nodal Transport Method | |
| nea-1591 | OMEGA, Subcritical and Critical Neutron Transport in General 3-D Geometry by Monte-Carlo | |
| ccc-0266 | ONETRAN, 1-D Transport in Planar, Cylindrical, Spherical Geom. for Homogeneous, Inhomogeneous Probl., Anisotropic Source | |
| nea-1324 | OREST, LWR Burnup Simulation Using Program HAMMER and ORIGEN | |
| nea-0702 | PALLAS-2DY, 2-D Neutron Transport, Gamma Transport with Anisotropic Scattering for Fixed Source | |
| ccc-0707 | PARTISN 4.00: 1-D, 2-D, 3-D Time-Dependent, Multigroup Deterministic Parallel Neutral Particle Transport Code | |
| nea-1238 | PASC-1, Petten AMPX-II/SCALE-3 Code System for Reactor Neutronics Calculation | |
| nea-0464 | PN, 1-D, 2-D, 3-D MultiGroup Neutron Transport | |
| nea-0181 | RADIFLUX, Bessel Function Fit of Radial Flux Data in Cylindrical Reactor | |
| nesc0631 | RAFFLE-2/MOD2, 3-D Steady-State Monte-Carlo Neutron Transport, Cell Averaged Scattering Cross-Sections Calculation | |
| ccc-0279 | RAFFLE-V, General Geometry Neutron Transport by Monte-Carlo | |
| ccc-0708 | REBUS-PC 1.4, Code System for Analysis of Research Reactor Fuel Cycles | |
| nea-0262 | REFLOS, Fuel Loading and Cost from Burnup and Heavy Atomic Mass Flow Calculation in HWR | |
| iaea0929 | RICANT, Neutron Transport in X-Y Geometry for Isotropic Scattering | |
| nea-0598 | RSYST, Modular System for Reactor Core and Shielding Problems | |
| nea-1779 | SAGEP-FR, Sensitivity Analysis of Fast Reactor Parameters | |
| ccc-0732 | SCALE 5.1/ORIGEN-ARP5.1: Modular system for criticality, shielding, source term, fuel depletion/decay, reactor physics | |
| nea-0370 | SHADOK-3-6, Transport Equation with Anisotropic Diffusion in P1 Approximation for Spherical and Cylindrical Geometry | |
| nea-0319 | SIMPLED-4, 1-D Neutron Diffusion in Spherical, Cylindrical, Planar Geometry from JAERI Fast-Set and ABBN | |
| nea-0905 | SIXTUS-2, 2-D MultiGroup Diffusion in Hexagonal Geometry with Intranodal Solution | |
| nea-1426 | SIXTUS-3, 3-D Nodal Neutron Diffusion Criticality in Hexagonal Geometry | |
| nea-1081 | SLAROM, Neutron Flux Distribution and Spectra in Lattice Cell | |
| nea-0430 | SNAP, MultiGroup 3-D Neutron Diffusion in X-Z, R-Theta-Z, Hexagonal-Z, Triangular-Z Geometry | |
| nea-1826 | SOLTRAN, solving multi-dimensional simplified P2 transport and diffusion problems of hexagonal geometry in fast reactors | |
| nea-0414 | SQUID-360, 2-D Neutron Diffusion in X-Y and R-Z Geometry with Criticality Search and Constant Neutron Source | |
| nea-0703 | STEADY-ACE, 3-D Neutronics and Multichannel Thermohydraulics Analysis of BWR | |
| ccc-0248 | SWAN-PPL, Fusion Reactor 1-D Particle Transport Optimization | |
| ccc-0204 | SWANLAKE, Cross-Sections Sensitivity Analysis for 1-D Discrete Ordinate Calculation | |
| nesc0713 | SYN-3D, 2-D and 3-D Neutron Diffusion Static Eigenvalues, Single Channel Spatial Flux Synthesis | |
| iaea1383 | SYRCO-1, 1-D, 2 Groups Multi-Zone Interactive Diffusion Code | |
| ccc-0638 | TART2005, 3D Coupled Neutron-Photon Combinatorial Geometry, Time Dependent, Monte Carlo Transport Code | |
| nesc0558 | TASK, 1-D MultiGroup Diffusion or Transport Theory Reactor Kinetics with Delayed Neutron | |
| nea-0997 | THT, 3-D Coarse Mesh LWR Bundle Fluxes and Power with Disconinuity Factors | |
| nea-1024 | TP2, Calculation of Reactivity and Kinetic Parameter by 2-D Neutron Transport. Perturbation Theory | |
| nea-0900 | TPHEX, MultiGroup Neutron Flux in Homogeneous Hexagonal LWR Cells | |
| nea-0953 | TRANSHEX, 2-D Thermal Neutron Flux Distribution from EpiThermal Flux in Hexagonal Geometry | |
| nea-0117 | TRAWS-4, Axial Flux Distribution for Control Rod Variations | |
| ccc-0293 | TRIDENT, 2-D Neutron Transport for Homogeneous and Inhomogeneous Problems in X-Z, R-Z Geometry, Anisotropic Scattering | |
| nea-0384 | TRIGON, 2-D Homogeneous and Inhomogeneous Fixed Source Neutron Diffusion for Triangular or Hexagonal Mesh | |
| nea-1716 | TRIPOLI-4.3.3 & 4.4, Coupled Neutron, Photon, Electron, Positron 3-D, Time Dependent Monte-Carlo, Transport Calculation | |
| nea-1086 | TRISTAN, 3-D fixed source radiation transport | |
| nea-1087 | TRITAC, 3-D Transport by Discrete Ordinate Method in X-Y-Z Geometry | |
| nea-0415 | TRITON, 3-D Multi-Region Neutron Diffusion Burnup with Criticality Search | |
| nea-0471 | TVEDIM, 2-D Homogeneous and Inhomogeneous Neutron Diffusion for X-Y, R-Z, R-Theta Geometry | |
| ccc-0547 | TWODANT-SYS, DANTSYS3.0, 1-D, 2-D, 3-D MultiGroup Discrete Ordinate Method Transport | |
| nesc0358 | TWOTRAN-2, 2-D MultiGroup Transport in X-Y, R-Z, R-Theta Geometry with Anisotropic Scattering | |
| ccc-0195 | TWOTRAN-GG-FC, General Geometry 2-D Transport with 1st Collision Source Calculation | |
| ccc-0613 | VALE-1, 2-D, 3-D MultiGroup Neutron Diffusion for Triangular Problems | |
| nesc0264 | VARI-QUIR-3, 2-D MultiGroup Steady-State Neutron Diffusion in X-Y R-Z or R-Theta Geometry | |
| ccc-0654 | VENTURE-PC 1.1, Reactor Analysis System with Sensitivity and Burnup | |
| nesc0510 | VIM, 3-D Monte-Carlo Analysis of Fast Critical Assemblies Using Point Cross-Sections | |
| ccc-0658 | VIM4.0, Stead-State 3-D Neutron Transport Using ENDF/B or Multigroup Cross Sections | |
| iaea0871 | VPI-NECM, Nuclear Engineering Program Collection for College Training | |
| nea-0655 | VSOP, Neutron Spectra, 2-D Flux Synthesis, Fuel Management, Thermohydraulics Calculation | |
| 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 | |
| dlc-0042 | ZZ CLEAR/42B, 126 Neutron-Group, 36 Gamma-Group Coupled Cross-Section in AMPX, CCCC Format, for LMFBR | |
| dlc-0011 | ZZ DLC-11 RITTS, 121-Group Coupled Cross-Section for ANISN, DOT, MORSE | |
| dlc-0016 | ZZ DLC-16 COBB, 123 Neutron-Group Cross-Section Library from ENDF/B for XSDRN Calculation | |
| dlc-0018 | ZZ DLC-18 NAB, 100 Neutron-Group Cross-Section Library of Na and Al for ANISN, DOT, MORSE Neutron Transport | |
| dlc-0002 | ZZ DLC-2D/100G, 100 Neutron-Group Cross-Section Library by SUPERTOG Calculation for ANISN, DOT | |
| dlc-0006 | ZZ DLC-6 GAMLIB, 99-Group Cross-Section Library by SUPERTOG Calculation from ENDF/B | |
| dlc-0187 | ZZ HILO86R, 66 Neutron, 22 Gamma Group Cross-Section for 400 MeV Neutron, 20 MeV Gamma | |
| dlc-0168 | ZZ LA100, ENDF Format Data Library for Neutron and Protons Up to 100 MeV | |
| dlc-0054 | ZZ LAFPX-V, Multigroup Fission Product Data Library from ENDF/B-V by Program NJOY | |
| nesc0532 | ZZ LASL-XSECS, Fast and Thermal Multigroup Cross-Section Library in LANL Transport Format | |
| dlc-0040 | ZZ LIB-IV, 50-Group Cross-Section Library in CCCC-III Format from ENDF/B-IV for Fast Reactors | |
| nea-1205 | ZZ MATX175/42-JEFF87, 172 Neutron-Group, 42 Gamma-Group MATXS Library in VITAMIN-J Structure | |
| dlc-0176 | ZZ MATXS10, 30-Group Neutron, 12-Group Gamma Cross-Sections in MATXS Format from ENDF/B-VI | |
| dlc-0177 | ZZ MATXS11, 80-Group Neutron, 24-Group Gamma Cross-Section in MATXS Format from ENDF/B-VI | |
| nea-1206 | ZZ MATXS70-JEFF87, 69+1 Group MATXS Library in WIMS BOXER Structure | |
| dlc-0076 | ZZ SAILOR, 47 Neutron-20 Gamma-Group Coupled Cross-Section Library from VITAMIN-C by AMPX | |
| dlc-0024 | ZZ SINEX, 100 Neutron-Group Neutron Reaction Cross-Section Library from ENDF/B by SUPERTOG for ANISN | |
| iaea0865 | ZZ TEMPEST/MUFT, Thermal Neutron and Fast Neutron Multigroup Cross-Section Library for Program LEOPARD | |
| nea-1264 | ZZ VITAMIN J/COVA, Covariance Matrix Data Library for Uncertainty Analysis | |
| dlc-0041 | ZZ VITAMIN-C/B, 171 Neutron-Group, 36 Gamma-Group Coupled Cross-Section for Fusion, LMFBR Calculations | |
| nea-1207 | ZZ WIMS-LIB/JEF87, 69+1 Group WIMS-D Library from JEF-1 |