Online Catalog
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 NameAbstractRSICC TapelistTitle
BCGAbstractC00578 C0170 00A Code For Calculating Pointwise Neutron Spectra and Criticality in Fast Reactor Cells.
CARNACAbstractC00238 I3691 00Calculation of Flux and Neutron Spectra in the Case of Criticality Accident.
DIF3D 11.2892
FEDC
AbstractC00784 MNYCP 02Code System Using Variational Nodal Methods and Finite Difference Methods to Solve Neutron Diffusion and Transport Theory Problems.
MKENO-DARAbstractC00513 FM380 00Direct Angular Representation Monte Carlo Code for Criticality Safety Analysis
MOCAAbstractC00590 IPCAT 00Monte Carlo Criticality Code System for Hexagonal Geometries.
MULTI-KENO2AbstractC00492 FM380 00A Monte Carlo Code System for Criticality Safety Analysis.
MVP-GMVP IIAbstractC00739 MNYCP 00General Purpose Monte Carlo Codes for Neutron and Photon Transport Calculations based on Continuous Energy and Multigroup Methods.
NCSP-DATAbstractM00002 MNYCP 01Nuclear Data in Support of the Nuclear Criticality Safety Program.
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
REBUS 11.0 EXE_ONLY
FEDC
AbstractC00822 MNYWS 01Code System for Analysis of Fast Reactor Fuel Cycles.
REBUS 11.2892
FEDC
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.
SIXTUS-3AbstractC00609 MFMWS 00Three-Dimensional, Nodal, Neutron Diffusion Criticality Code System in Hex-Z Geometry.
SLIDERULE 1.0AbstractC00704 PC586 01Nuclear Criticality Slide Rule.
SRAC95AbstractC00716 MNYWS 00Thermal Reactor Code System for Reactor Design and Analysis.
SWATAbstractC00714 MNYCP 01Step-Wise Burnup Analysis Code System to Combine SRAC-95 Cell Calculation Code and ORIGEN2.
TRIPOLI-4 8.1
OECD
AbstractC00806 MNYCP 00Code System for Coupled Neutron, Photon, Electron, Positron, 3-D, Time Dependent, Monte-Carlo, Transport Calculations.
The Radiation Safety Information Computational Center (RSICC) collects, analyzes, maintains, and distributes software in the areas of radiation transport and safety. RSICC resides in the Nuclear Energy and Fuel Cycle Division (NEFCD) at Oak Ridge National Laboratory.