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Note: RESTRICTIONS APPLY TO SOME PACKAGES -
810 -- US DOE 10CFR810 Jurisdiction
FEDC -- US Government Agencies and Their Contractors Only
OECD -- Restricted/See Abstract
RUGA -- Restricted Use Government Authorized
USSO -- US Distribution Only
USUNV -- US Universities Only
| Packages with Keyword: DIFFUSION THEORY |
| Package Name | Abstract | RSICC Tapelist | Title |
| 1DB-2DB-3DB | Abstract | C00741 PC586 00 | One-Dimensional Diffusion Code System for Nuclear Reactor. |
| 3DDT | Abstract | C00605 C6600 00 | Multigroup Diffusion Code System for Use in Fast Reactor Analysis. |
| AIRDIF | Abstract | C00360 C6600 00 | A Two-Dimensional Atmospheric Radiation Diffusion Code. |
| AMP | Abstract | C00793 PCX86 00 | Advanced Multi-Physics. |
ARC 11.2892
FEDC | Abstract | C00824 MNYCP 02 | Code System for Analysis of Nuclear Reactors. |
| ATHENA_2D | Abstract | P00431 MNYCP 00 | Code System For Simulation Of Hypothetical Recriticality Accidents in a Thermal Neutron Spectrum. |
| AUS98 | Abstract | C00519 MNYWS 01 | Modular System for Neutronics Calculations of Fission Reactors, Fusion Blankets, and Other Systems. |
| BOLD VENTURE IV | Abstract | C00459 I3033 00 | A Reactor Analysis Code System. |
| BOXER | Abstract | C00766 MNYWS 00 | Fine-flux Cross Section Condensation, 2D Few Group Diffusion and Transport Burnup Calculations |
| CITATION-LDI 2 | Abstract | C00643 PC386 02 | Nuclear Reactor Core Analysis Code System. |
| DASH-FP | Abstract | C00366 C0000 00 | A One-Dimensional Analytic-Numerical Solution to the Problem of Multicomponent Time-Dependent Diffusion of Fission Products. |
DIF3D 11.2892
FEDC | Abstract | C00784 MNYCP 02 | Code System Using Variational Nodal Methods and Finite Difference Methods to Solve Neutron Diffusion and Transport Theory Problems. |
| DIFMOD | Abstract | C00572 I3083 00 | A Computer Program To Calculate The Leaching of Radionuclides and the Corrosion of Cemented Waste Forms in Water or Brine. |
| DLS | Abstract | C00264 C6600 00 | Two-Dimensional Shielding Calculational System with Diffusion Theory and Line-of-Sight Method. |
| EQUIVA-1.1 | Abstract | P00323 IMFPC 00 | Generation of Environment-Insensitive Equivalent Diffusion Theory Parameters for PWR Reflector Regions. |
| EQUIVA-2 | Abstract | P00324 IMFPC 00 | Generation of Environment-Insensitive Equivalent Diffusion Theory Parameters for PWR Reflector Regions. |
ERANOS 2.0
OECD | Abstract | C00745 MNYWS 00 | Modular Code and Data System for Fast Reactor Neutronics Analyses |
| FEM-2D | Abstract | C00260 C6600 00 | Two-Dimensional Diffusion Theory Code System Based on the Method of Finite Elements. |
| FEMB | Abstract | C00340 B6700 00 | A Two-Dimensional Diffusion Theory Finite Element Program. |
| FINELM | Abstract | C00483 MFMWS 00 | Multigroup Finite Element Diffusion Code System. |
| FPZD | Abstract | C00603 PC386 00 | Code System for Multigroup Neutron Diffusion/Depletion Calculations. |
| GENP-2 | Abstract | C00575 ALLMF 00 | Generalized Perturbation Theory Code System. |
| GNOMER | Abstract | C00625 MNYCP 01 | Multigroup 3-Dimensional Neutron Diffusion Nodal Code System with Thermohydraulic Feedbacks. |
| GRENADE | Abstract | C00516 C1787 00 | Green's Function Nodal Algorithm for the Diffusion Equation. |
| GRENADE | Abstract | C00516 D0780 00 | Green's Function Nodal Algorithm for the Diffusion Equation. |
| HEXAB-3D | Abstract | C00593 I0370 00 | Three-Dimensional Few-Group Coarse Mesh Diffusion Code for Neutron Physics Calculation of Reactor Core in Hexagonal Geometry. |
| IONMIG | Abstract | C00526 ALLMF 00 | Code System for Radionuclide Migration Calculations. |
| LABAN-PEL | Abstract | C00611 IMFPC 00 | A Two-Dimensional, Multigroup Diffusion, High-Order Response Matrix Code. |
| MARC-PN | Abstract | C00311 D8810 00 | A Neutron Diffusion Code System with Spherical Harmonics Option. |
| MARIA SYSTEM | Abstract | P00359 D6000 00 | Code System to Calculate Cross Sections for PWR Fuel Assembly Calculations. |
| MCRAC | Abstract | C00562 IBMPC 00 | Multiple Cycle Reactor Analysis Code. |
| MOSRA-LIGHT | Abstract | P00505 MNYWS 00 | High-Speed Three-Dimensional Nodal Diffusion Code System. |
| NESTLE 5.2.1 | Abstract | C00641 MNYCP 04 | Code System to Solve the Few-Group Neutron Diffusion Equation Utilizing the Nodal Expansion Method (NEM) for Eigenvalue, Adjoint, and Fixed-Source |
REBUS 11.0 EXE_ONLY
FEDC | Abstract | C00822 MNYWS 01 | Code System for Analysis of Fast Reactor Fuel Cycles. |
REBUS 11.2892
FEDC | Abstract | C00822 MNYCP 02 | Code System for Analysis of Fast Reactor Fuel Cycles. |
| REBUS3/VARIANT8 | Abstract | C00653 MNYWS 01 | Code System for Analysis of Fast Reactor Fuel Cycles. |
| REBUS-PC 1.4 | Abstract | C00708 PC586 00 | Code System for Analysis of Research Reactor Fuel Cycles. |
| REDIFFUSION | Abstract | C00347 I0360 00 | One-Dimensional Neutron Removal-Diffusion and Gamma-Ray Kernel Integration or Diffusion Theory Calculator. |
| RETRANS | Abstract | C00669 SUN05 00 | Code System For Calculating Reactivity Transients In a LWR. |
| RHEIN | Abstract | C00585 I3090 00 | Reactor Code System for Neutron Physics Calculation. |
| SIXTUS-3 | Abstract | C00609 MFMWS 00 | Three-Dimensional, Nodal, Neutron Diffusion Criticality Code System in Hex-Z Geometry. |
| SNAP-3D | Abstract | C00434 MNYCP 01 | Multigroup Complex Geometry Neutron Diffusion Code System. |
| SOLTRAN | Abstract | C00763 PCX86 00 | Solving Multi-Dimensional Simplified P2 Transport and Diffusion Problems of Hexagonal Geometry in Fast Reactors. |
| TASK | Abstract | C00184 I0360 00 | Generalized One-Dimensional Radiation Transport and Diffusion Kinetics Code System. |
| TEMPEST-BNW | Abstract | P00559 C7600 00 | Transient 3-D Thermohydraulics for FBR. |
| THT | Abstract | C00480 I0360 00 | Three-Dimensional Neutron Coarse Mesh Code System to Evaluate Average Bundle Fluxes and Power in Light Water Reactors. |
| TRIGLAV | Abstract | P00495 PC586 00 | Code System to Calculate Mixed Cores in TRIGA Mark II Research Reactor. |
| TRIGON | Abstract | C00290 U1108 00 | Two-Dimensional Multigroup Diffusion Code System-Trigonal or Hexagonal Mesh. |
| TRITAC | Abstract | C00560 D8810 00 | A Three-Dimensional Transport Code For Eigenvalue Problems Using The Diffusion Synthetic Acceleration Method. |
| UNIMUG3 | Abstract | C00407 C0170 00 | Solves Multigroup Diffusion Equations in One-Dimensional Systems. |
| VALE 1.1 | Abstract | C00613 IRISC 01 | A Multigroup Diffusion Theory Neutronics Code System for Solving Two- and Three-Dimensional Problems for Triagonal Geometries. |
| VALE 1.1 | Abstract | C00613 PC386 01 | A Multigroup Diffusion Theory Neutronics Code System for Solving Two- and Three-Dimensional Problems for Triagonal Geometries. |
| VENTEASY | Abstract | C00776 PCX86 00 | Criticality search for a desired Keffective by adjusting dimensions, nuclide concentrations, or buckling |
| VENTURE-PC | Abstract | C00654 PC586 02 | A Reactor Analysis Code System. |