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Catalog of Programs in Category F

F. Space - Time Kinetics, Coupled Neutronics - Hydrodynamics - Thermodynamics

ccc-0793 AMP, Advanced Multi-Physics
ccc-0459 BOLD/VENTURE-4, Reactor Analysis System with Sensitivity and Burnup
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-1614 COBRA-EN, Thermal-Hydraulic Transient Analysis of Reactor Cores
nea-1375 COSIMA, BWR Core Performance Simulator
nea-0425 COSTANZA-XE, 2-D Pebble-Bed or Prismatic Fuel Elements HTR Dynamic in Cylindrical Geometry
nea-1734 CRISSUE-S, Neutronics/Thermal-hydraulics Coupling in LWR Technology
nea-1411 DYN3D/M2, Reactivity Transients in Light H2O Reactors with Hexagonal Geometry
nea-1686 ENTREE 1.4.0, BWR Core Simulation System for Space and Time Dependent Coupled Phenomena
nea-0228 EXCURS-3, Reactor Kinetics and Heat Transfer in Cylindrical Channel During Accident
nesc0862 FX2-TH, 2-D MultiGroup Neutron Diffusion in X-Y, R-Z and R-Theta Geometry with Thermal Feedback
nesc0310 GAKIN-2, 1-D MultiGroup Time-Dependent Neutron Diffusion, Finite Difference Method
nea-0547 HASSAN, Time-Dependent Temperature Distribution and Stress and Strain in HTR Fuel Pins
uscd1241 MCART, solve the time dependent neutron transport equation
nesc0901 PAD, Coupled Neutronics, Thermohydraulics in 1-D Spherical, Cylindrical, Planar Geometry
iaea1228 PULSTRI-1, Mixed Core Triga Reactor Pulse Calculation
nea-1600 QUARK, 2-Group 3-D Neutronic Kinetics Coupled to Core Thermalhydraulics
nea-1867 RAPID, RAdial power and burnup Prediction by following fissile Isotope Distribution in the pellet
ccc-0361 SANDYL, 3-D Time-Dependent and Space-Dependent Gamma Electron Cascade Transport by Monte-Carlo
nea-1577 SKETCH-N 1.0, Solve Neutron Diffusion Equations of Steady-State and Kinetics Problems
nea-1911 SOPHIA, a Lagrangian-based CFD code for nuclear thermal-hydraulics and safety applications.
nea-0468 SPARK, Time-Dependent 1-D, 2-D, 3-D Diffusion with Heat Transfer and Feedback
nesc0558 TASK, 1-D MultiGroup Diffusion or Transport Theory Reactor Kinetics with Delayed Neutron
ccc-0180 TDA, Time-Dependent 1-D Neutron Transport, Gamma Transport by ANISN Method in Slab, Spherical, Cylindrical Geometry
ccc-0709 TDTORT, Time-Dependent, 3-D, Discrete Ordinates, Neutron Transport Code System with Delayed Neutrons
nea-1593 TRAC-PF1/EN MOD 3, Best Estimate Coupled 3-D Neutronics-Thermalhydraulics
ccc-0654 VENTURE-PC 1.1, Reactor Analysis System with Sensitivity and Burnup
nesc0511 VENUS-2, Reactor Kinetics with Feedback, 2-D LMFBR Disassembly Excursions
iaea0871 VPI-NECM, Nuclear Engineering Program Collection for College Training