Online Catalog
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Click on Abstract to read the package abstract.
Click on RSICC Tapelist to view list of files distributed with package.

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: BURNUP
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.
ABLEIT-TRANSAbstractP00247 C0175 00Error Propagation Analysis for Burnup Calculation.
ARC 11.2892
AbstractC00824 MNYCP 02Code System for Analysis of Nuclear Reactors.
BISON 1.5AbstractC00464 HM200 00One-Dimensional Discrete Ordinate Transport and Burnup Calculation Code System.
BISON-CAbstractC00659 MNYWS 00One-Dimensional Discrete Ordinate Transport and Burnup Calculation Code System.
BOLD VENTURE IVAbstractC00459 I3033 00A Reactor Analysis Code System.
BOXERAbstractC00766 MNYWS 00Fine-flux Cross Section Condensation, 2D Few Group Diffusion and Transport Burnup Calculations
BUCORSTAbstractP00339 PC386 00A Code to Prepare Burnup-Dependent Multigroup Nuclear Reactor Source Terms.
CANDULIB-AECLAbstractD00210 MNYCP 00Burnup-Dependent ORIGEN-S Cross Section Libraries for CANDU Reactor Fuel Characterization.
CONDOR-3AbstractC00811 I0370 00Two-Dimensional Reactor Program with Local and Spectrum Dependent Burnup.
DCHAIN-SP2001AbstractC00712 MNYWS 01Code System for Analyzing Decay and Build-up Characteristics of Spallation Products.
DEPLETORAbstractP00523 MNYCP 00Code System to Provide Depletion Capability to the U.S. NRC PARCS Code
DRAGON3.05DAbstractC00647 MNYWS 03Lattice Cell Code System.
AbstractC00745 MNYWS 00Modular Code and Data System for Fast Reactor Neutronics Analyses
FEMAXI 6 VER.1AbstractP00536 IBMPC 00Code System for Light Water Reactor Fuel Analysis.
FPZDAbstractC00603 PC386 00Code System for Multigroup Neutron Diffusion/Depletion Calculations.
LASERAbstractC00344 I0360 00A One-Dimensional, Neutron-Thermalization, Lattice-Cell Program Based on MUFT and THERMOS.
LEOPARDAbstractC00343 C0000 00A Spectrum-Dependent Non-Spatial Fuel Depletion Code System.
LEOPARDAbstractC00343 IBMPC 00A Spectrum-Dependent Non-Spatial Fuel Depletion Code System.
MARIA SYSTEMAbstractP00359 D6000 00Code System to Calculate Cross Sections for PWR Fuel Assembly Calculations.
MOCUPAbstractP00365 DALPU 00MCNP/ORIGEN Coupling Utility Programs.
MONTEBURNS 2.0AbstractP00455 MNYCP 02Automated, Multi-Step Monte Carlo Burnup Code System.
ORIP_XXIAbstractC00731 PC586 02Computer Programs for Isotope Transmutation Simulations.
ORLIBJ32AbstractD00255 MNYCP 00ORIGEN2 Libraries Based on JENDL-3.2.
PREMORAbstractC00369 I0360 00A Point Reactor Exposure Code System for Survey Nuclear Analysis of Power Plant Performance.
PSU-LEOPARD/RBIAbstractC00563 IBMPC 01A Spectrum Dependent Non-Spatial Depletion Code.
PWR-AXBUPRO-GKNAbstractD00209 MNYCP 00Measured Axial Burnup Profiles for NeckarWesthiem PWR Reactors.
PWR-AXBUPRO-SNLAbstractD00201 MNYCP 00Axial Burnup Profile Database for Pressurized Water Reactors.
RAPIDAbstractC00797 PCX86 00RAdial Power and Burnup Prediction by Following Fissile Isotope Distribution in the Pellet.
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.
RICECCCAbstractC00348 I0360 00A Reactor Nuclide Inventory Code for Calculating Actinides and Fission Products.
SERPENTAbstractC00757 MNYWS 00Continuous Energy Monte Carlo Reactor Physics Burnup Calculation Code.
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.
TRIGAPAbstractC00600 IBMPC 00A Computer Code for TRIGA Type Reactors.
TRIGLAVAbstractP00495 PC586 00Code System to Calculate Mixed Cores in TRIGA Mark II Research Reactor.
VENTURE-PCAbstractC00654 PC586 02A Reactor Analysis Code System.
WIMS-ANL 4.0AbstractC00698 MNYCP 00Deterministic Code System for Reactor Lattice Calculation.
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.