Online Catalog
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Note: RESTRICTIONS APPLY TO SOME PACKAGES -
810 -- US DOE 10CFR810 Jurisdiction
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OECD -- Restricted/See Abstract
RUGA -- Restricted Use Government Authorized
USSO -- US Distribution Only
USUNV -- US Universities Only
Packages with Subject: NUCLEAR SYSTEM SAFETY ANALYSIS
Package NameAbstractRSICC TapelistTitle
ACRA-IIAbstractC00213 I0360 00Kernel Integration Code System for Estimation of Radiation Doses Caused by a Hypothetical Reactor Accident.
ACRA-TRITAbstractC00283 I0360 00The Tritium Version of ACRA-II, Estimation of Radiation Doses Caused by a Hypothetical Reactor Accident.
ALARM-B2AbstractP00218 I0360 00A Computer Code System for Analysis of a Large Break LOCA of a BWR.
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.
BLOCKAGE V2.5RAbstractP00377 IBMPC 00Code System to Calculate Potential Plugging of Emergency Core Cooling Suction Strainers by Debris in a BWR.
BOT3P-5.3AbstractP00530 MNYCP 02Code System for 2D and 3D Mesh Generation and Graphical Display of Geometry and Results for Radiation Transport Codes.
BWR-LTASAbstractC00485 I3033 01A Boiling Water Reactor Long-Term Accident Simulation Code.
COBRA4IAbstractP00419 MNYCP 00Code Sytem to Calculate Rod-Bundle and Core Thermal-Hydraulics.
COGAPAbstractP00375 MNYCP 01Nuclear Power Plant Containment Hydrogen Control System Evaluation Code.
COMIDAAbstractP00343 MNYCP 00Radionuclide Food Chain Model for Acute Fallout Deposition.
COMPBRN3AbstractP00389 PC386 00Code System for Modeling Compartment Fires.
COMRADEX4AbstractC00332 I0360 00Evaluator of Potential Radiological Doses in the Near (< 10 km) Environment of Radioactive Release.
CRAC2AbstractC00419 C0000 00Code System for Calculating Reactor Accident Consequences.
CRAC2AbstractC00419 I3033 00Code System for Calculating Reactor Accident Consequences.
D2OAbstractP00398 PC486 00Code System for Computing Thermodynamic and Transport Properties of D2O.
DIF3D 11.2892
FEDC		AbstractC00784 MNYCP 02Code System Using Variational Nodal Methods and Finite Difference Methods to Solve Neutron Diffusion and Transport Theory Problems.
DORIANAbstractP00425 IBMPC 00Code System to Implement Bayes Method for Plant Aging Risk Analysis.
DOSE-SGTRAbstractC00624 IBMPC 00Code System to Calculate the Integrated Iodine Release to the Environment During a Steam Generator Tube Rupture in a PWR.
EDSFI
USSO		AbstractD00215 PC486 00Electrical Distribution System Functional Inspection Data Base.
EMERALDAbstractC00211 I0360 00Calculation of Activity Releases and Potential Doses from a Pressurized Water Reactor Plant.
ENTREE 1.4.0AbstractP00519 MNYWS 00BWR Core Simulation System for Space and Time Dependent Coupled Phenomena.
EVNTREAbstractP00465 D0VAX 00Code System for Event Progression Analysis for PRA.
FEP 4.16AbstractP00440 IBMPC 00Fault-tree, Event tree, & P&ID Editors.
FIRACAbstractP00444 CY000 00Nuclear Facilities Fire Accident Model
FIREDATAAbstractD00125 PC486 00Nuclear Power Plant Fire Data Base for Personal Computers.
FONTAAbstractC00423 S4044 00Code System For Calculating Individual And Collective Doses From Reactor Accidents Using Pasquill's Plume Model.
FRANTIC3AbstractP00406 CDCMF 00Time-Dependent Reliability Analysis.
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.
FRCRL2AbstractC00231 C6400 00Calculation of Fission-Product Release in Reactor Accident Analyses.
FUELSDATAAbstractP00446 C7600 00Code System to Model Verification Fuel Rod Data.
GIRAFFEAbstractP00304 I3033 00General Isotope Release Analysis For Failed Elements.
GMAAbstractP00367 MNYCP 00Code System for Calculation of Reactor Accident Consequences.
GRSACAbstractC00774 PCX86 00Graphite Reactor Severe Accident Code.
HAARM-3AbstractP00401 CDCMF 00Aerosol Behavior Log-Normal Distribution Model.
HORNAbstractC00568 I3083 00A Computer Code To Analyze The Gas-Phase Transport of Fission Products In Reactor Cooling System Under Severe Accidents.
HOTSPOT 3.0.2AbstractM00009 IBMPC 03Health Physics Code System for Evaluating Accidents Involving Radioactive Materials.
IMPACTS-BRC2.1AbstractC00666 IBMPC 00Code System for Analysis of Potential Radiological Impacts.
IMPORTANCEAbstractP00407 I0370 00FTA Basic Event & Cut Set Ranking.
IRRAS 4.16AbstractP00386 IBMPC 04Code System to Calculate Integrated Reliability and Risk Analysis.
KFIXAbstractP00409 C7600 00Code System to Calculate Transient 2-Dimensional 2-Fluid Flow Dynamics.
MARCH2AbstractP00473 CDCMF 00Code System to Model LWR Meltdown Accident Response.
MARD 4.16AbstractP00448 IBMPC 00Models And Results Database System.
MATADORAbstractC00689 CDCMF 00Radionuclide Behavior in Containments.
MORECAAbstractP00411 PC386 00Computer Code System for Simulating Modular High-Temperature Gas Cooled Reactor Core Heatup.
MURE V2-SMUREAbstractC00764 MNYWS 01Serpent - MCNP Utility for Reactor Evolution.
NRCDOSE 2.3.20AbstractC00684 PC586 14Code System for Evaluating Routine Radioactive Effluents from Nuclear Power Plants with a Windows Interface.
NRCDOSE72V1.2.3AbstractC00768 PCX86 03Code System for Evaluating Routine Radioactive Effluents from Nuclear Power Plants with a Windows Interface.
NRCPAGEAbstractP00491 DVX11 00Code System to Detect Recurring Loss of Special Nuclear Materials.
NRCPIPES 2.0AAbstractP00429 IBMPC 00Code System for Fracture Mechanics Analysis of Circumferential Surface Cracks in Pipes.
NUTRANAbstractC00675 I0370 00Code System for Long-Term Repository Safety Analysis.
OCTAVIAAbstractP00460 I0370 00Code System to Calculate Pressure Vessel Failure Probabilities.
ORINC
USSO		AbstractP00439 I0360 00Code System for 1-D Implicit Heat Conduction Solution.
ORMDIN
USSO		AbstractP00399 I3033 002-D Nonlinear Inverse Heat Conduction.
ORMGEN3DAbstractP00430 CY0MP 00Mesh Generator for 3-D Crack Geometries.
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.
PAVANAbstractC00445 I3033 00Atmospheric Dispersion Code System for Evaluating Accidental Radioactivity Releases from Nuclear Power Stations.
P-CARESAbstractP00538 PC586 00Probabilistic Computer Analysis for Rapid Evaluation of Structures.
PELE-1CAbstractP00461 C7600 00Code System for Fluid-Structure Interaction Analysis.
PRESTAbstractC00355 I0360 00Calculator of Pressure and Temperature Transient in Containment Studies.
PRISIMAbstractC00574 IBMPC 00Plant Risk Status Information Management System.
PSDRECAbstractP00441 DP011 00Code System for Power Spectral Density Recognition Continuous On-line Reactor Surveillance.
QUARKAbstractP00492 PC586 00Code System for 2-Group, 3D Neutronic Kinetics Calculations Coupled to Core Thermal Hydraulics.
RACERAbstractC00174 U1108 00Calculation of Potential External Dose from Airborne Fission Products Following Postulated Reactor Accident.
RASCAL 4.3AbstractC00783 PCX86 02Radiological Assessment for Consequence Analysis.
RCSLK9AbstractP00452 IBMPC 00Code System to Calculate Reactor Coolant System Leak Rate.
REFLUXAbstractP00403 I3033 00Code System to Predict LWR Reflood Heat Transfer.
REPRISK PC 1.02AbstractC00586 PC386 01Repository Risk Assessment Software for Personal Computers.
RETRANSAbstractC00669 SUN05 00Code System For Calculating Reactivity Transients In a LWR.
RISKAPAbstractC00486 I3033 00Analysis of Increased Risk to Arbitrary Populations.
RISKIND 2.0
FEDC		AbstractC00623 IBMPC 02Radiological Risk Assessment Code System for Spent Nuclear Fuel Transportation.
SAFE-D/SAFE-RAbstractP00496 MNYCP 00Code System for the Analysis of Component Failure Data with a Compound Statistical Model.
SAMCRAbstractP00487 U1100 00Code System for 2-D Elastodynamic Fracture Analysis.
SARA 4.16AbstractP00484 IBMPC 00System Analysis and Risk Assessment System.
SCANS 1AAbstractP00373 PC386 01Shipping Cask Design Review Analysis.
SCRELAAbstractP00408 SUN05 00Code System for Supercritical Water Cooled Reactor LOCA 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.
SIGPIAbstractP00475 D0785 00Fault Tree Cut Set System Performance.
SLIDERULE 1.0AbstractC00704 PC586 01Nuclear Criticality Slide Rule.
SMACSAbstractP00396 C7600 01Probabilistic Seismic Analysis Code System.
SOLA-DFAbstractP00454 C7600 00Code System to Calculate Transient 2-Dimensional 2-Phase Flow.
SOPHIAAbstractC00857 MNYCP 00A Lagrangian-based computational fluid dynamics code for nuclear thermal hydraulics and safety applications.
SPEEDIAbstractC00507 FM180 00Code System for Real-Time Prediction of Radiation Dose to the Public Due to an Accidental Release from a Nuclear Power Plant.
SPIRT-NRC
USSO		AbstractP00198 I3033 01Computerized Mathematical Models of Spray Washout of Airborne Contaminants (Radioactivity) in Containment Vessels.
SQUIRT VER2
USSO		AbstractP00583 PCX86 00Code System to Predict Leakage Rate and Area of Crack Opening for Cracked Pipes in Nuclear Power Plants.
SUBDOSA-IIAbstractC00270 U1100 00Calculation of External Gamma-Ray and Beta-Ray Doses from Accidental Atmospheric Releases of Radionuclides.
TACT-IIIAbstractC00447 I3033 00Calculation of the Transport of Radioactivity from a Reactor Core.
TEMACAbstractP00468 D0VAX 00Top Event Matrix Analysis Code System.
THYDE-B1/MOD2AbstractP00553 FM200 00Computer Code for PWR LOCA Thermohydraulic Transient Analysis.
THYDE-P2AbstractP00554 FV100 00Computer Code for PWR LOCA Thermohydraulic Transient Analysis.
TMAP 7AbstractC00858 PCX86 00Tritium Migration Analysis Program
TORACAbstractP00459 C0170 00Code System to Calculate Tornado-Induced Flow Material Transport.
TOXRISKAbstractC00692 CDCMF 00Code System for Toxic Gas Accident Analysis.
TSORTAbstractP00486 IBMPC 00Automated Technique for Nuclear Plant Training Task Assignment.
UHSAbstractP00390 IPS70 00Ultimate Heat Sink Cooling Pond and Spray Pond Analysis Models.
VISA2AbstractP00445 MNYCP 00Code System to Calculate Probability of Reactor Vessel Failure.
WREM-TOODEE2AbstractP00469 ALLMF 002-D Time-Dependent Fuel Element, Thermal Analysis Code System.
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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.