Computer Programs

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nea-1839 | ACAB-2008, ACtivation ABacus Code |

nea-1638 | ANITA-IEAF, Isotope Inventories from Intermediate Energy Neutron Irradiation for Fusion Applications |

ccc-0657 | BETA-S 6, Multi-Group Beta-Ray Spectra |

ccc-0459 | BOLD/VENTURE-4, Reactor Analysis System with Sensitivity and Burnup |

nea-1735 | CARL 2.3, radiotoxicity, activity, dose and decay power calculations for spent fuel |

ests1071 | CECP(BWR) CECP(PWR), Decommissioning Costs for PWR and BWR |

ccc-0544 | CEPXS ONELD, 1-D Coupled Electron Photon MultiGroup System |

ccc-0837 | CEPXS, Coupled Electron-Photon Cross Section |

ccc-0604 | CHAINS-PC, Decay Chain Atomic Densities |

ccc-0755 | CINDER 1.05, Actinide Transmutation Calculations Code |

nesc0313 | CINDER, Depletion and Decay Chain Calculation for Fission Products in Thermal Reactors |

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 |

iaea0883 | CLUB, Cell Calculation PF Candu PWR Fuel Clusters |

nesc0873 | COAST-4, Design and Cost of Tokamak Fusion Reactors |

psr-0614 | COBRA-SFS 6.0, Thermal-Hydraulic Analysis of Multi-Assembly Spent Fuel Storage and Transportation Systems |

ests0135 | COBRA-SFS CYCLE3, Thermal Hydraulic Analysis of Spent Fuel Casks |

nesc0498 | CONCEPT-5, Cost and Economics Analysis for Nuclear Fuel or Fossil Fuel Power Plant |

iaea1226 | CORD-2, PWR Core Design and Fuel Management |

iaea0873 | CRITIC, In-Core Fuel Management for CANDU PWR |

nea-1892 | CUMYIELD.MT, cumulative yields calculations of radioactive decay isotopes considering decay chain |

ccc-0640 | DCHAIN, Isotope Buildup and Isotope Decay by 1 Point Approximation |

nea-1603 | DCHAIN-SP 2001, Code System for Analyzing Decay and Build-up Characteristics of Spallation Products |

nea-1893 | DECAYHEAT.MT, decay heat calculations from radioactive isotopes |

nea-1887 | DESAE, develop prospective nuclear energy scenarios in a regional and global scale |

psr-0531 | EEDB, The Energy Economic Data Base |

nea-1683 | ERANOS 2.3N, Modular code and data system for fast reactor neutronics analyses |

nea-0617 | FAPMAN-IC, LWR Fuel Cost Analysis with Program ORSIM Interface |

nea-1080 | FEMAXI-6, Thermal and Mechanical Behaviour of LWR Fuel Rods |

nea-1901 | FINIX 1.19.12, thermal and mechanical behaviour of a nuclear fuel rod during steady-state and transient conditions |

nea-0897 | FISP-6, Fission Products Inventory and Energy Release in Irradiated Fuel |

nea-0706 | FISPIN, Isotope Buildup and Isotope Decay for Actinides, Fission Products, Structure Materials |

nesc0301 | FREVAP-6, Metal Fission Products Release from HTGR Fuel Elements |

nesc0576 | GEM-NESC, Fuel Cycle Cost and Economics for Thermal Reactor, Present Worth Analysis |

nea-1894 | INVENTDYN.MT, calculates the dynamics of the amount of isotope and its daughter nuclides with time stamps |

nea-0624 | JOSHUA, Neutronics, Hydraulics, Burnup, Refuelling of LWR |

nea-0288 | KERBREK, Fuel Cycle Cost Analysis for Power Reactor |

nea-1001 | KORIGEN, Isotope Inventory, Radiation Heat from PWR Burnup |

nea-0441 | KPD, Time-Dependent Fuel Cycle Cost Calculation for Various Reactor Types |

ccc-0631 | LWRARC-1.0, PWR and BWR Spent Fuel Decay Heat Generator |

nea-1643 | MCB1C, Monte-Carlo Continuous Energy Burnup Code |

iaea0889 | MCRAC/RBI, In Core Fuel Management, Program of PFMP System |

psr-0455 | MONTEBURNS 2.0, An Automated, Multi-Step Monte Carlo Burnup Code System |

nea-1845 | MURE v2 - SMURE, MCNP Utility for Reactor Evolution: couples Monte-Carlo transport with fuel burnup calculations |

iaea1411 | NAAPRO, Neutron Activation Analysis Prognosis and Optimization code |

nesc0683 | NUFUEL, Conditions for Power Production, U Fuel, Pu Recycle and Reprocessing |

nesc0588 | ORCOST-2, PWR, BWR, HTGR, Fossil Fuel Power Plant Cost and Economics |

nea-1324 | OREST, LWR Burnup Simulation Using Program HAMMER and ORIGEN |

ccc-0371 | ORIGEN-2.2, Isotope Generation and Depletion Code Matrix Exponential Method |

ccc-0702 | ORIGEN-ARP 2.00, Isotope Generation and Depletion Code System-Matrix Exponential Method with GUI and Graphics Capability |

nea-0622 | ORIGEN-JR, Radiation Source and Nuclide Transmutation with In-Core Burnup |

nea-1880 | ORIP-XXI, isotope transmutation simulations |

nea-1339 | PEPIN, Methodology for Computing Concentrations, Activities, Gamma-Ray Spectra, and Residual Heat from Fission Products. |

nea-1663 | PLUTON, Isotope Generation and Depletion in Highly Irradiated LWR Fuel Rods |

nea-1929 | POSY, Power system model for system costs evaluation - Will be released soon |

ccc-0708 | REBUS-PC 1.4, Code System for Analysis of Research Reactor Fuel Cycles |

ccc-0653 | REBUS3/VARIANT8.0, Code System for Analysis of Fast Reactor Fuel Cycles |

nesc1065 | REFCO83, Nuclear Fuel Cycle Cost Economics Using Discounted Cash Flow Analysis |

nea-1231 | REFREP, Near-Field Model for Spent Fuel Repository |

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 |

nea-1767 | SMAFS, Steady-state analysis Model for Advanced Fuelcycle Schemes |

nea-0450 | SOTHIS, PWR Fuel Cycle Equilibrium Cost Evaluation |

nea-0374 | SPES, Fuel Cycle Optimization for LWR |

nea-1151 | SUSD, Sensitivity and Uncertainty in Neutron Transport and Detector Response |

nea-1628 | SUSD3D, 1-, 2-, 3-Dimensional Cross Section Sensitivity and Uncertainty Code |

iaea1214 | TRIGAC, Flux and Power Distribution and Burnup for TRIGA Reactor |

nea-0415 | TRITON, 3-D Multi-Region Neutron Diffusion Burnup with Criticality Search |

iaea0884 | TRIVENI, 3-D Fuel Management for PHWR CANDU |

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 |

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 |