Computer Programs

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nesc0325 | 2-DB, 2-D MultiGroup Diffusion, X-Y, R-Theta, Hexagonal Geometry Fast Reactor, Criticality Search |

nesc0567 | 3-DB, 3-D MultiGroup Diffusion, X-Y-Z, R-Theta-Z, Triangular-Z Geometry, Fast Reactor Burnup |

nea-0912 | ABLEIT-TRANS, Isotope Concentration and Sensitivities on Cross-Sections Data |

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

psr-0190 | ADENA, Fission Products Beta Spectra and Gamma Spectra in 19 Group from U235 Pu239 Mixture |

nea-0321 | ANDROMEDA, 1-D Burnup for Fuel Cycle Analysis of FBR |

nea-1638 | ANITA-2000, Isotope Inventories from Neutron Irradiation, for Fusion Applications |

nea-1343 | ANITA-4, Isotope Inventories from Neutron Irradiation, for Fusion Applications |

ccc-0519 | AUS, Neutron Transport and Gamma Transport System for Fission Reactors and Fusion Reactors |

nea-0373 | BEST-4, Fuel Cycle and Cost Optimization for Discrete Power Levels |

nea-0404 | BEST-5, Power Reactor Fuel Cycle Optimization by Bellman Method |

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

nea-0591 | BEVE, Isotope Buildup in LWR Fuel Pin with Self-Shielding in Pellet |

nea-0870 | BISON, 1-D Burnup and Transport in Slab, Cylindrical, Spherical Geometry |

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

nea-0236 | BOLERO, 2 Group Burnup for PWR and BWR in R-Z Geometry with Restart and Recycle |

nea-1187 | BOREAS, Nuclear and Thermohydraulic LWR Burnup Simulation |

nea-1523 | BOXER, Fine-flux cross section condensation, 2D few group diffusion and transport burnup calculations |

nea-0237 | BURNY, 5 Group BWR and PWR Burnup in X-Y Geometry by Diffusion Calculation |

nea-0350 | BURNY-SQUID, 2-D Burnup of UO2 and Mix UO2 PuO2 Fuel in X-Y or R-Z Geometry |

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

ests1071 | CECP, Decommissioning Costs for PWR and BWR |

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

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

nea-0451 | CICLON, Neutronics Calculation for PWR Transition Fuel Cycle Management |

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 |

nesc0540 | CLOTHO, Mass Flow Data Calculation for Program PACTOLUS |

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

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

psr-0614 | COBRA-SFS CYCLE 4, Code System for Thermal Hydraulic Analysis of Spent Fuel Casks |

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

nea-1578 | COMRAD96, Nuclear Fuel Burnup and Depletion Calculation System |

iaea0928 | COMTA, Ceramic Fuel Elements Stress Analysis |

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

nea-0427 | CONDOR-3, Local and Spectrum Dependent Burnup with Mesh-Wise Depletion |

iaea1226 | CORD, PWR Core Design and Fuel Management |

nea-0463 | CRACKLE, Fast Reactor Pu 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 |

nea-0151 | DANCOFF-3, Dancoff Correction for Cylindrical Fuel Rod at H2O Gaps and for Fuel Clusters |

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

nea-0664 | 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-0446 | DELIGHT-7, Point Reactivity Burnup for HTGR Lattice with P1 Neutron Scattering Approximation |

psr-0523 | DEPLETOR Version 2, provides depletion capability to the Purdue Advanced Reactor Core Simulator (PARCS) code |

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

nea-0298 | DISCOUNT-G, Nuclear Power Program with Cost Analysis and Pu Production Optimization |

nesc0579 | DWARF, 1-D Few-Group Neutron Diffusion with Thermal Feedback for Burnup and Xe Oscillation |

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-0341 | ERUPT, 2-D 2 Group Fuel Management in R-Z Geometry with Fuel Shuffling |

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

nea-0693 | FAPMAN-ORSIM, General Cost Optimization for System of Nuclear Power Plants |

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

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 |

nea-0235 | FLARE-JAERI, 3-D BWR and ATR Simulation |

ccc-0603 | FPZD, Reactor Burnup by MultiGroup Neutron Diffusion |

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

nea-0314 | FURNACE-J, 2-D Diffusion Burnup for Fast Reactors from JAERI Fast-Set |

nesc0223 | GAD-2, Fuel Cycle Depletion Calculation with Partial Refueling and Fuel Recycling |

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

nesc0711 | GEOCOST-BC, Geothermal Power Plant Electricity Generator Cost, Thermodynamics Calculation |

iaea1222 | HAMCIND, Cell Burnup with Fission Products Poisoning |

nea-0176 | HETERO, Flux and Power Distribution in Thermal Reactor by 3-D, 2 Group Line Source Sink Method |

nea-0100 | HYTHEST, Dependence of Fuel Fabrication Tolerances on Hydraulics of BWR, PWR |

nea-0353 | ICON, Reactor Operation Fission Products Inventory Calculation |

nea-1340 | INVENT-STUDSVIK, Fission Products Abundances in U235, U238, Pu239 Samples |

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

nea-0434 | ISOTEX-1, Time-Dependent Heavy Isotope and Fission Products Concentration in U Reactor or Pu Reactor |

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-0417 | KOSAK, Power Plant Cost Optimization with Pu Availability Option |

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

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 |

ccc-0343 | LEOPARD-MICRO, Spectrum-Dependent Non-Spatial Fuel Depletion |

nea-0965 | LOLA-SYSTEM, JEN-UPM PWR Fuel Management System Burnup Code System |

nesc9449 | LPGC, Levelized Steam Electric Power Generator Cost |

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

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

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

nesc9479 | MGA, Pu Isotope Abundance from Multichannel Analyzer Gamma Spectra |

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

nesc0798 | MSF21/VTE21, Desalination Plant Heat, Mass Balance, Design, Cost Optimization |

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

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

nesc0146 | NPRFCCP, Fuel Cycle Cost and Economics for Multi-Region Reactor |

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 |

nesc0699 | ORSIM, Nuclear Fuel, Fossil Fuel Hydroelectric Power Plant Cost and Economics |

nesc0540 | PACTOLUS, Nuclear Power Plant Cost and Economics by Discounted Cash Flow Method |

nea-0521 | PAS-1, 2-D, 3-D Linear Static and Dynamic Stress Analysis with 2-D Steady-State Temperature Distribution |

iaea0819 | PELINOMIC, Power Plant Cost Optimization for Dispersed Load Centres |

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

nesc0454 | PHENIX, 2-D MultiGroup Diffusion Fast Reactor Burnup Calculation and Fuel Cycle Analysis |

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

nesc0340 | POWERCO, Nuclear Power Plant Electricity Cost and Economics |

nea-1675 | PPICA, Power Plant Investment Cost Analysis |

iaea0888 | PSU-LEOPARD, Program LEOPARD in PFMP System, Fast Neutron and Thermal Neutron Spectra Calculation |

nesc0441 | PWCOST, Fuel Cycle Cost and Economics by Present Worth Levelized Method |

ccc-0639 | RACC-PULSE, Neutron Activation in Fusion Reactor System |

ccc-0627 | RADAC, Radioactive Decay and Accumulation of Long Lived Isotopes |

nea-0475 | RASPA, Burnup with Fission Products Inventory, Gamma Spectra, Isotopic Power Density |

ccc-0443 | REAC*3, Isotope Activation and Transmutation in Fusion Reactors |

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 |

ests0176 | RECAP, Replacement Energy Cost for Short-Term Reactor Plant Shut-Down |

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

nea-0262 | REFLOS, Fuel Loading and Cost from Burnup and Heavy Atomic Mass Flow Calculation in HWR |

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

nea-0101 | REP-3, Time-Dependent Xe and Sm Poisoning from Space-Dependent Flux Distribution |

ccc-0137 | RIBD, Fission Products Inventory and Delay Heat in Fast Reactors, with Data Library |

ccc-0382 | RIBD-IRT, Isotope Buildup and Isotope Decay from Fission Source |

nea-0239 | RIBOT-5, 0-D Burnup for 5 Group BWR or PWR Lattice |

nea-0589 | RICE-CEGB, Long-Term Actinides and Fission Products Inventory of Irradiated Fuel |

nesc0831 | RO-75, Reverse Osmosis Plant Design Optimization and Cost Optimization |

nea-0598 | RSYST, Modular System for Reactor Core and Shielding Problems |

nea-1078 | SACHET, Dynamic Fission Products Inventory in PWR Multiple Compartment System |

nea-1779 | SAGEP-FR, Sensitivity Analysis of Fast Reactor Parameters |

ccc-0834 | SCALE 6.2, A Comprehensive Modeling and Simulation Suite for Nuclear Safety Analysis and Design |

iaea0913 | SCENARIOS, Simulation of Reactor Introduction and Operation Scenario Needs |

nea-0235 | SCOPERS-2, BWR and PWR Core Performance Simulation |

nea-1840 | SERPENT 1.1.7, 3-D continuous-energy Monte Carlo reactor physics burnup calculation, lattice physics applications |

iaea0925 | SHARDA, Thermal Reactor Isotope Irradiation Analysis |

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-0842 | SRAC-95, Cell Calculation with Burnup, Fuel Management for Thermal Reactors |

iaea0882 | STAR, Fuel Management of BWR |

iaea0900 | STOFFEL-1, Steady-State In-Pile Behaviour of Cylindrical H2O Cooled Oxide Fuel Rod |

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 |

nea-1698 | SWAT, Step-Wise Burnup Analysis Code System to Combine SRAC-95 Cell Calculation Code and ORIGEN2 |

iaea0872 | TACHY, BWR Fuel Management by 2-D Coarse Mesh Neutron Diffusion |

iaea1338 | TEMPUL, Temperature Distribution in Fuel Element after Pulse |

nea-0486 | TOTEM, Demand Assessment for Nuclear Power Plants and Conventional Power Plants |

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 2.1.5, 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 |

nea-1882 | XSUN-2013, Windows interface environment for transport and sensitivity-uncertainty software TRANSX-2, PARTISN and SUSD3D |

nea-0072 | ZADOC, 2 Group Time-Dependent Burnup in X-Y Geometry with Fuel Management |

iaea0912 | ZZ AMZ, 70-Group 40 Isotope Multigroup Library for Fast Reactor Calculation |

dlc-0089 | ZZ LUMP, Lumped Fission Product Cross-Section Library for Fast Reactor Analysis from ENDF/B-V |

dlc-0038 | ZZ ORYX-E/38B, Group Constant Library from ENDF/B Fission Product Data for ORIGEN Calculation |