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Catalog of Programs in Category C

C. Static Design Studies

psr-0315 AMPX-77, Modular System for Coupled Neutron-Gamma MultiGroup Cross-Sections from ENDF/B-5
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-0254 ANISN-ORNL, 1-D Neutron Transport & Gamma Transport in Slab, Cylindrical, Spherical Geometry with Anisotropic Scattering
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
ccc-0459 BOLD/VENTURE-4, Reactor Analysis System with Sensitivity and Burnup
nea-1678 BOT3P5.4, 3D Mesh Generator and Graphical Display of Geometry for Radiation Transport Codes, Display of Results
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
ccc-0726 CNCSN 2009, One, Two- and Three-Dimensional Coupled Neutral and Charged Particle Sn Parallel Multi-Threaded Code System
ccc-0829 COG11.1, Multiparticle Monte Carlo Code System for Shielding and Criticality Use
iaea1226 CORD-2, PWR Core Design and Fuel Management
nea-1903 CRISTAL V2.0.3, Criticality calculation package
ccc-0547 DANTSYS3.0, 1-D, 2-D, 3-D MultiGroup Discrete Ordinate Method Transport
ccc-0649 DIF3D 8.0/VARIANT8.0, 2-D 3-D Multigroup Diffusion/Transport Theory Nodal & Finite Difference Solver, Variational Method
ccc-0784 DIF3D10.0, Variational Nodal Methods, Finite Difference Methods to Solve N diffusion & Transport Theory Problems
nea-0391 DLS, 2-D Diffusion with Line-of-Sight Method for Cavities
ccc-0650 DOORS3.2A, 1-,2-,3-dimensional discrete-ordinates system for deep-penetration neutron and photon transport
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
uscd1234 DRAGON 3.05D, Reactor Cell Calculation System with Burnup
nea-0322 DTF4-J, 1-D Neutron Transport with Anisotropic Scattering by Sn Method
nea-1683 ERANOS 2.3N, Modular code and data system for fast reactor neutronics analyses
nesc0156 EXTERMINATOR-2, 2-D MultiGroup Neutron Diffusion in X-Y R-Z or R-Theta Geometry
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-0896 FINELM, MultiGroup Diffusion in 3-D by Finite Elements Method
nesc0380 GATT, 3-D Few-Group Neutron Diffusion for Power Distribution in Hexagonal Reactor Core for HTGR
iaea1271 GNOMER, Core Power Distribution by 1-D, 2-D, 3-D MultiGroup Neutron Diffusion
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
ccc-0510 KENO-IV(RG), KENO-IV with Random Geometry
ccc-0548 KENO5A-PC, Monte-Carlo Criticality with Supergrouping
uscd1241 MCART, solve the time dependent neutron transport equation
nea-1643 MCB1C, Monte-Carlo Continuous Energy Burnup Code
nea-1733 MCNP4B-GN, Monte Carlo Code System for (gamma,n) production and transport in high-Z materials
iaea0889 MCRAC/RBI, In Core Fuel Management, Program of PFMP System
ccc-0841 MMS3D, Method of Manufactured Solutions for 3D one-group SN Equations with escalating order of non-smoothness
nea-0527 MONK, Keff, Collision Rate, Flux Distribution in General Geometry from UKNDL by Monte-Carlo Method
nea-1905 MORET 5.D.1, Monte Carlo simulation tool to solve transport equation for neutrons
nea-1633 MOSRA-LIGHT, High Speed 3-D X-Y-Z Nodal Diffusion Code for Vector Computers
nea-1896 MOSRA-SRAC, Lattice Calculation Module of the Modular Code System for Nuclear Reactor Analyses MOSRA
nea-1673 MVP/GMVP V.3, MC Codes for Neutron & Photon Transport Calc. based on Continuous Energy and Multigroup Methods
ccc-0641 NESTLE 5.2.1, Few-Group Neutron Diffusion for Steady-State and Transient Problems by Nodal Expansion Method (NEM)
nea-1591 OMEGA, Subcritical and Critical Neutron Transport in General 3-D Geometry by Monte-Carlo
nea-1324 OREST, LWR Burnup Simulation Using Program HAMMER and ORIGEN
ccc-0760 PARTISN 5.97, 1-D, 2-D, 3-D Time-Dependent, Multigroup Deterministic Parallel Neutral Particle Transport Code
ccc-0842 PARTISN 8.29, Time-Dependent, Parallel Neutral Particle Transport Code System
ccc-0708 REBUS-PC 1.4, Code System for Analysis of Research Reactor Fuel Cycles
nea-1840 SERPENT 1.1.7, 3-D continuous-energy Monte Carlo reactor physics burnup calculation, lattice physics applications
nea-1923 SERPENT V2.2.X, 3-D continuous-energy Monte Carlo reactor physics burnup calculation, lattice physics applications
iaea1437 SUPERMC 3.3.0, Super Monte Carlo simulation program for nuclear and radiation process
ccc-0204 SWANLAKE, Cross-Sections Sensitivity Analysis for 1-D Discrete Ordinate Calculation
ccc-0638 TART2022, 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
ccc-0759 TITAN 1.29, A Three-Dimensional Deterministic Radiation Transport Code System
ccc-0543 TORT-DORT, 1-D 2-D 3-D Discrete Ordinate Neutron and Photon Transport with Deep Penetration
nea-1716 TRIPOLI-4 VERS. 8.1, 3D general purpose continuous energy Monte Carlo Transport code
nea-1878 TRIPOLI-4 version 9S, Coupled Neutron, Photon, Electron, Positron 3-D, Time Dependent Monte-Carlo Transport Calculation
nea-1086 TRISTAN, 3-D fixed source radiation transport
nea-0415 TRITON, 3-D Multi-Region Neutron Diffusion Burnup with Criticality Search
uscd1239 VENTEASY, Criticality Search for a Desired Keffective by Adjusting Dimensions, Nuclide Concentrations, or Buckling
ccc-0654 VENTURE-PC 1.1, Reactor Analysis System with Sensitivity and Burnup
nea-1856 VESTA 2.1&AURORA1.0, Monte Carlo depletion interface code and AURORA 1.0.0, Depletion analysis tool
ccc-0754 VIM 5.1, Steady-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
iaea1440 VSOP99-11, Neutron Spectra, 2-D Flux Synthesis, Fuel Management, Thermohydraulics Calculation
nea-1882 XSUN-2023, Windows interface environment for transport and sensitivity-uncertainty software TRANSX-2, PARTISN and SUSD3D
nea-1206 ZZ MATXS70-JEFF87, 69+1 Group MATXS Library in WIMS BOXER Structure
nea-1264 ZZ VITAMIN J/COVA, Covariance Matrix Data Library for Uncertainty Analysis