Click on Package Name to get detailed information.
Click on Abstract to read the package abstract.
Click on RSICC Tapelist to view list of files distributed with package.
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: CRITICALITY CALCULATIONS |
| 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. |
ARC 11.2892
FEDC | Abstract | C00824 MNYCP 02 | Code System for Analysis of Nuclear Reactors. |
| BCG | Abstract | C00578 C0170 00 | A Code For Calculating Pointwise Neutron Spectra and Criticality in Fast Reactor Cells. |
| CITATION-LDI 2 | Abstract | C00643 PC386 02 | Nuclear Reactor Core Analysis Code System. |
| COG11.1 | Abstract | C00829 MNYCP 00 | Multiparticle Monte Carlo Code System for Shielding and Criticality Use. |
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. |
| MKENO-DAR | Abstract | C00513 FM380 00 | Direct Angular Representation Monte Carlo Code for Criticality Safety Analysis |
| MORSE-C | Abstract | C00431 C7600 00 | Monte Carlo Multigroup Neutron and Gamma-Ray Transport Code System. |
| MULTI-KENO2 | Abstract | C00492 FM380 00 | A Monte Carlo Code System for Criticality Safety Analysis. |
| MVP-GMVP II | Abstract | C00739 MNYCP 00 | General Purpose Monte Carlo Codes for Neutron and Photon Transport Calculations based on Continuous Energy and Multigroup Methods. |
| 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 |
| OMEGA | Abstract | C00433 BESM6 00 | Monte Carlo Criticality Code System. |
PERSENT 11.2892
FEDC | Abstract | C00823 MNYCP 02 | Perturbation and Sensitivity Code for Assembly Homogenized Multi-group Transport Problems |
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. |
| SIXTUS-3 | Abstract | C00609 MFMWS 00 | Three-Dimensional, Nodal, Neutron Diffusion Criticality Code System in Hex-Z Geometry. |
| SLIDERULE 1.0 | Abstract | C00704 PC586 01 | Nuclear Criticality Slide Rule. |
| TRIGAP | Abstract | C00600 IBMPC 00 | A Computer Code for TRIGA Type Reactors. |
TRIPOLI-4 8.1
OECD | Abstract | C00806 MNYCP 00 | Code System for Coupled Neutron, Photon, Electron, Positron, 3-D, Time Dependent, Monte-Carlo, Transport Calculations. |
TRIPOLI-4 9S
OECD | Abstract | C00815 MNYCP 00 | Code System for Coupled Neutron, Photon, Electron, Positron, 3-D, Time Dependent, Monte-Carlo, Transport Calculations. |
| VENTEASY | Abstract | C00776 PCX86 00 | Criticality search for a desired Keffective by adjusting dimensions, nuclide concentrations, or buckling |
| VIM 5.1 | Abstract | C00754 MNYWS 01 | Continuous Energy Neutron and Gamma-ray Transport Code System. |