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 Keyword: REACTOR SAFETY
Package NameAbstractRSICC TapelistTitle
ATHENA_2DAbstractP00431 MNYCP 00Code System For Simulation Of Hypothetical Recriticality Accidents in a Thermal Neutron Spectrum.
BEACON MOD3AbstractP00402 CDCMF 00Code System for Thermal-Hydraulic Analysis of Nuclear Reactor Containments.
COBRA-3C-RERTRAbstractP00606 I0370 00COBRA-3C-RERTR
COMPBRN3AbstractP00389 PC386 00Code System for Modeling Compartment Fires.
DYN3D/M2AbstractP00579 I3090 00Reactivity Transients in Light H2O Reactors with Hexagonal Geometry.
EDSFI
USSO
AbstractD00215 PC486 00Electrical Distribution System Functional Inspection Data Base.
ENTREE 1.4.0AbstractP00519 MNYWS 00BWR Core Simulation System for Space and Time Dependent Coupled Phenomena.
EXPRESSAbstractC00622 MNYCP 00Exact Preparedness Supporting System.
FIRACAbstractP00444 CY000 00Nuclear Facilities Fire Accident Model
FIREDATAAbstractD00125 PC486 00Nuclear Power Plant Fire Data Base for Personal Computers.
FRANTIC3AbstractP00406 CDCMF 00Time-Dependent Reliability Analysis.
FRAPCON2AbstractP00517 MFMWS 00Fuel Rod Thermal-Mechanical Behavior.
FRAPT6/MOD1
USSO
AbstractP00436 C0176 00Code System for Transient Analysis of Fuel Rods.
FRAPT6/V21
USSO
AbstractP00436 C0176 01Code System for Transient Analysis of Fuel Rods.
FUELSDATAAbstractP00446 C7600 00Code System to Model Verification Fuel Rod Data.
GRSACAbstractC00774 PCX86 00Graphite Reactor Severe Accident Code.
HECTR 1.5+
USSO
AbstractP00457 CY000 00Hydrogen Event Containment Response Code System.
HSI-DRGAbstractP00435 IBMPC 00Code System for Use with Human System Interface Design Review Guidelines.
JDL-REACTOR-KINAbstractM00006 MNYCP 00Nuclear Reactor Kinetics and Control.
KFIX 3DAbstractP00383 C7600 00Code System to Calculate Three-Dimensional Extension Two-Phase Flow Dynamics.
MAEROSAbstractP00466 C7600 00Code System for Multicomponent Aerosol Time Evolution.
MARCH2AbstractP00473 CDCMF 00Code System to Model LWR Meltdown Accident Response.
MATADORAbstractC00689 CDCMF 00Radionuclide Behavior in Containments.
MINETAbstractP00490 CY000 00Momentum Integral Network Method for Thermal-Hydraulic Systems Analysis.
MORECAAbstractP00411 PC386 00Computer Code System for Simulating Modular High-Temperature Gas Cooled Reactor Core Heatup.
NAUA-MOD5 NAUA-MOD5/MAbstractP00556 MNYCP 00Aerosols in Reactor Containment During Meltdown.
NORMAAbstractP00471 PC586 00Code System to Solve Burnup Dependent Neutron Diffusion Equations in Two and Three Dimensions.
NORMA-FPAbstractP00470 PC586 00Code System to Perform Neutronic and Thermal-Hydraulic Subchannel Analysis from Converged Coarse-Mesh Nodal Solutions.
OCTAVIAAbstractP00460 I0370 00Code System to Calculate Pressure Vessel Failure Probabilities.
PARET-ANLAbstractP00516 MNYCP 00Code System to Predict Consequences of Nondestructive Accidents in Research and Test Reactor Cores.
PARET-ANL(NESC)AbstractP00565 MNYCP 00Code System to Predict Consequences of Nondestructive Accidents in Research and Test Reactor Cores.
PMK2-VVER440-REPORTSAbstractM00012 MNYCP 00Results of the Experiments Performed in the PMK-2 Facility for VVER Safety Studies.
PRISIMAbstractC00574 IBMPC 00Plant Risk Status Information Management System.
REACTORSHIELDING-NMSAbstractM00014 MNYCP 00REACTORSHIELDING-NMS, Reactor Shielding for Nuclear Engineers.
RELAP3B/MOD110
810
AbstractP00422 C7600 00Reactor System Transient Code.
SAMCRAbstractP00487 U1100 00Code System for 2-D Elastodynamic Fracture Analysis.
SEISIM1AbstractP00453 C7600 00Code System for Seismic Probabilistic Risk Assessment.
SETSAbstractP00380 CDCMF 00Set Equation Transformation System.
SHC
USSO
AbstractP00493 CY000 00Seismic/Hazard Characterization in the Eastern U.S.
SIMMER II
USSO
AbstractC00691 MFMWS 00Code System for Two-Dinensional Sn-Neutronics and Fluid Dynamics.
SOFIRE-2AbstractP00570 I0370 00Containment Temperature and Pressure During Na Pool Fire, 1-Cell or 2 Cell.
SOLA-LOOPAbstractP00464 C7600 00Nonequilibrium, Drift-Flux Code System for Two-Phase Flow Network Analysis
TOXRISKAbstractC00692 CDCMF 00Code System for Toxic Gas Accident Analysis.
USINTAbstractP00415 MNYCP 00Code System to Calculate Heat and Mass Transfer In Concrete
VISA2AbstractP00445 MNYCP 00Code System to Calculate Probability of Reactor Vessel Failure.
WAKEAbstractP00605 I0370 00Navier Stokes Equation with 2-D Turbulence, Stream Function, Vorticity.
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.