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Packages with Keyword: DIFFUSION THEORY
Package NameAbstractRSICC TapelistTitle
1DB-2DB-3DBAbstractC00741 PC586 00One-Dimensional Diffusion Code System for Nuclear Reactor.
3DDTAbstractC00605 C6600 00Multigroup Diffusion Code System for Use in Fast Reactor Analysis.
AIRDIFAbstractC00360 C6600 00A Two-Dimensional Atmospheric Radiation Diffusion Code.
AMPAbstractC00793 PCX86 00Advanced Multi-Physics.
ARC 11.2892
AbstractC00824 MNYCP 02Code System for Analysis of Nuclear Reactors.
ATHENA_2DAbstractP00431 MNYCP 00Code System For Simulation Of Hypothetical Recriticality Accidents in a Thermal Neutron Spectrum.
AUS98AbstractC00519 MNYWS 01Modular System for Neutronics Calculations of Fission Reactors, Fusion Blankets, and Other Systems.
BOLD VENTURE IVAbstractC00459 I3033 00A Reactor Analysis Code System.
BOXERAbstractC00766 MNYWS 00Fine-flux Cross Section Condensation, 2D Few Group Diffusion and Transport Burnup Calculations
CITATION-LDI 2AbstractC00643 PC386 02Nuclear Reactor Core Analysis Code System.
DASH-FPAbstractC00366 C0000 00A One-Dimensional Analytic-Numerical Solution to the Problem of Multicomponent Time-Dependent Diffusion of Fission Products.
DIF3D 11.2892
AbstractC00784 MNYCP 02Code System Using Variational Nodal Methods and Finite Difference Methods to Solve Neutron Diffusion and Transport Theory Problems.
DIFMODAbstractC00572 I3083 00A Computer Program To Calculate The Leaching of Radionuclides and the Corrosion of Cemented Waste Forms in Water or Brine.
DLSAbstractC00264 C6600 00Two-Dimensional Shielding Calculational System with Diffusion Theory and Line-of-Sight Method.
EQUIVA-1.1AbstractP00323 IMFPC 00Generation of Environment-Insensitive Equivalent Diffusion Theory Parameters for PWR Reflector Regions.
EQUIVA-2AbstractP00324 IMFPC 00Generation of Environment-Insensitive Equivalent Diffusion Theory Parameters for PWR Reflector Regions.
AbstractC00745 MNYWS 00Modular Code and Data System for Fast Reactor Neutronics Analyses
FEM-2DAbstractC00260 C6600 00Two-Dimensional Diffusion Theory Code System Based on the Method of Finite Elements.
FEMBAbstractC00340 B6700 00A Two-Dimensional Diffusion Theory Finite Element Program.
FINELMAbstractC00483 MFMWS 00Multigroup Finite Element Diffusion Code System.
FPZDAbstractC00603 PC386 00Code System for Multigroup Neutron Diffusion/Depletion Calculations.
GENP-2AbstractC00575 ALLMF 00Generalized Perturbation Theory Code System.
GNOMERAbstractC00625 MNYCP 01Multigroup 3-Dimensional Neutron Diffusion Nodal Code System with Thermohydraulic Feedbacks.
GRENADEAbstractC00516 C1787 00Green's Function Nodal Algorithm for the Diffusion Equation.
GRENADEAbstractC00516 D0780 00Green's Function Nodal Algorithm for the Diffusion Equation.
HEXAB-3DAbstractC00593 I0370 00Three-Dimensional Few-Group Coarse Mesh Diffusion Code for Neutron Physics Calculation of Reactor Core in Hexagonal Geometry.
IONMIGAbstractC00526 ALLMF 00Code System for Radionuclide Migration Calculations.
LABAN-PELAbstractC00611 IMFPC 00A Two-Dimensional, Multigroup Diffusion, High-Order Response Matrix Code.
MARC-PNAbstractC00311 D8810 00A Neutron Diffusion Code System with Spherical Harmonics Option.
MARIA SYSTEMAbstractP00359 D6000 00Code System to Calculate Cross Sections for PWR Fuel Assembly Calculations.
MCRACAbstractC00562 IBMPC 00Multiple Cycle Reactor Analysis Code.
MOSRA-LIGHTAbstractP00505 MNYWS 00High-Speed Three-Dimensional Nodal Diffusion Code System.
NESTLE 5.2.1AbstractC00641 MNYCP 04Code System to Solve the Few-Group Neutron Diffusion Equation Utilizing the Nodal Expansion Method (NEM) for Eigenvalue, Adjoint, and Fixed-Source
AbstractC00822 MNYWS 01Code System for Analysis of Fast Reactor Fuel Cycles.
REBUS 11.2892
AbstractC00822 MNYCP 02Code System for Analysis of Fast Reactor Fuel Cycles.
REBUS3/VARIANT8AbstractC00653 MNYWS 01Code System for Analysis of Fast Reactor Fuel Cycles.
REBUS-PC 1.4AbstractC00708 PC586 00Code System for Analysis of Research Reactor Fuel Cycles.
REDIFFUSIONAbstractC00347 I0360 00One-Dimensional Neutron Removal-Diffusion and Gamma-Ray Kernel Integration or Diffusion Theory Calculator.
RETRANSAbstractC00669 SUN05 00Code System For Calculating Reactivity Transients In a LWR.
RHEINAbstractC00585 I3090 00Reactor Code System for Neutron Physics Calculation.
SIXTUS-3AbstractC00609 MFMWS 00Three-Dimensional, Nodal, Neutron Diffusion Criticality Code System in Hex-Z Geometry.
SNAP-3DAbstractC00434 MNYCP 01Multigroup Complex Geometry Neutron Diffusion Code System.
SOLTRANAbstractC00763 PCX86 00Solving Multi-Dimensional Simplified P2 Transport and Diffusion Problems of Hexagonal Geometry in Fast Reactors.
TASKAbstractC00184 I0360 00Generalized One-Dimensional Radiation Transport and Diffusion Kinetics Code System.
TEMPEST-BNWAbstractP00559 C7600 00Transient 3-D Thermohydraulics for FBR.
THTAbstractC00480 I0360 00Three-Dimensional Neutron Coarse Mesh Code System to Evaluate Average Bundle Fluxes and Power in Light Water Reactors.
TRIGLAVAbstractP00495 PC586 00Code System to Calculate Mixed Cores in TRIGA Mark II Research Reactor.
TRIGONAbstractC00290 U1108 00Two-Dimensional Multigroup Diffusion Code System-Trigonal or Hexagonal Mesh.
TRITACAbstractC00560 D8810 00A Three-Dimensional Transport Code For Eigenvalue Problems Using The Diffusion Synthetic Acceleration Method.
UNIMUG3AbstractC00407 C0170 00Solves Multigroup Diffusion Equations in One-Dimensional Systems.
VALE 1.1AbstractC00613 IRISC 01A Multigroup Diffusion Theory Neutronics Code System for Solving Two- and Three-Dimensional Problems for Triagonal Geometries.
VALE 1.1AbstractC00613 PC386 01A Multigroup Diffusion Theory Neutronics Code System for Solving Two- and Three-Dimensional Problems for Triagonal Geometries.
VENTEASYAbstractC00776 PCX86 00Criticality search for a desired Keffective by adjusting dimensions, nuclide concentrations, or buckling
VENTURE-PCAbstractC00654 PC586 02A Reactor Analysis Code System.
The Radiation Safety Information Computational Center (RSICC) is a Department of Energy Specialized Information Analysis Center (SIAC) authorized to collect, analyze, maintain, and distribute computer software and data sets in the areas of radiation transport and safety. RSICC resides in the Reactor and Nuclear Systems Division (RNSD) at Oak Ridge National Laboratory.