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 Subject: NUCLEAR CRITICALITY SAFETY |
| Package Name | Abstract | RSICC Tapelist | Title |
| BCG | Abstract | C00578 C0170 00 | A Code For Calculating Pointwise Neutron Spectra and Criticality in Fast Reactor Cells. |
| CARNAC | Abstract | C00238 I3691 00 | Calculation of Flux and Neutron Spectra in the Case of Criticality Accident. |
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 |
| MOCA | Abstract | C00590 IPCAT 00 | Monte Carlo Criticality Code System for Hexagonal Geometries. |
| 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. |
| NCSP-DAT | Abstract | M00002 MNYCP 01 | Nuclear Data in Support of the Nuclear Criticality Safety Program. |
| 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. |
| 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. |
| SWAT | Abstract | C00714 MNYCP 01 | Step-Wise Burnup Analysis Code System to Combine SRAC-95 Cell Calculation Code and ORIGEN2. |
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. |