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.

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: HEAT TRANSFER
Package NameAbstractRSICC TapelistTitle
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.
BWR-LTASAbstractC00485 I3033 01A Boiling Water Reactor Long-Term Accident Simulation Code.
CASKCODESAbstractP00262 IBMPC 00CAPSIZE, SCOPE, AND KWIKDOSE for Shipping Cask Optimization, Dose Calculation, Parameter Evaluation, and Shielding Requirements.
COBRA-3C-RERTRAbstractP00606 I0370 00COBRA-3C-RERTR
COBRA-ENAbstractP00507 MNYCP 01Thermal-Hydraulic Transient Analysis of Reactor Cores.
COBRA-SFS VERSION 6.0AbstractP00614 MNYCP 02COBRA-SFS Thermal-Hydraulic Analysis of Multi-Assembly Spent Fuel Storage and Transportation Systems.
COMMIX-1B
USSO		AbstractP00393 DVX11 003-D Single-Phase Thermal Hydraulics
COMMIX-1B
USSO		AbstractP00393 I3033 003-D Single-Phase Thermal Hydraulics
COMMIX-1C
USSO		AbstractP00393 MNYCP 003-D Single-Phase Thermal Hydraulics
COMPBRN3AbstractP00389 PC386 00Code System for Modeling Compartment Fires.
CORTESAbstractP00404 I0360 00Code System for Thermal & Mechanical Analysis of Tees.
DYN3D/M2AbstractP00579 I3090 00Reactivity Transients in Light H2O Reactors with Hexagonal Geometry.
EXCURS-3-RRAbstractP00586 D0VAX 00Kinetics of Research Reactor Reactivity Transient Analysis.
FIRACAbstractP00444 CY000 00Nuclear Facilities Fire Accident Model
FLODISAbstractP00417 I0360 00Code System to Calculate Thermal Response of FSV HTGR Core.
FLOWPLOT IIAbstractP00234 I3033 00Fluid Dynamics and Heat Transfer Plotting Package.
FRANCOAbstractP00363 MNYCP 00Finite Element Fuel Rod Analysis Code System.
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.
GAPCON-THERMALAbstractP00499 C7600 00Code System to Calculate Fuel Steady State & Transient Behavior.
HASSANAbstractP00593 I0370 00Time-Dependent Temperature Distribution and Stress and Strain in HTR Fuel Pins.
HEATING 7.3AbstractP00199 MNYCP 06Multidimensional, Finite-Difference Heat Conduction Analysis Code System.
HORNAbstractC00568 I3083 00A Computer Code To Analyze The Gas-Phase Transport of Fission Products In Reactor Cooling System Under Severe Accidents.
JASMINE V.3AbstractC00795 MNYCP 00JAEA Simulator for Multiphase INteractions and Explosions.
KFIXAbstractP00409 C7600 00Code System to Calculate Transient 2-Dimensional 2-Fluid Flow Dynamics.
LAPUR6
USSO		AbstractP00395 PC586 02BWR Core Stability Measurements.
LTCAbstractP00329 IBMPC 00LMR Transient Calculation Code System.
MARCH2AbstractP00473 CDCMF 00Code System to Model LWR Meltdown Accident Response.
MINETAbstractP00490 CY000 00Momentum Integral Network Method for Thermal-Hydraulic Systems Analysis.
MOXY-MOD32AbstractP00385 I0360 00BWR Core Heat Transfer Code System.
NORMAAbstractP00471 PC586 00Code System to Solve Burnup Dependent Neutron Diffusion Equations in Two and Three Dimensions.
ORMDIN
USSO		AbstractP00399 I3033 002-D Nonlinear Inverse Heat Conduction.
ORTHIS-ORTHATAbstractP00569 I0360 00ORTHIS: Steady-State Heat Conduction in 2-D X-Y, R-Z and R-Theta Geometry; ORTHAT: Transient Heat Conduction in 2-D X-Y, R-Z and R-Theta Geometry.
ORTURBAbstractP00418 I0360 00HTGR Steam Turbine Dynamic Behavior.
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.
REFLUXAbstractP00403 I3033 00Code System to Predict LWR Reflood Heat Transfer.
RELAP5/MOD1/029_EXE
810		AbstractP00423 C0176 01Reactor System Transient Code.
SCDAP/RELAP5/MOD3.3-EXE
810		AbstractP00581 MNYCP 01A Best-Estimate Transient Simulation of Light Water Reactor Coolant Systems During a Severe Accident.
SCOPEAbstractP00210 I3033 00Computer Code System for Shipping Cask Optimization and Parametric Evaluation.
SCORCH-B2AbstractP00601 I0370 00BWR Core Heating During LOCA.
SCORE-EVETAbstractP00442 C7600 00Code System for Three-Dimensional Hydraulic Reactor Core Analysis.
SCRELAAbstractP00408 SUN05 00Code System for Supercritical Water Cooled Reactor LOCA Analysis.
SFHA
USSO		AbstractP00413 IBMPC 00Code System for Spent Fuel Heating Analysis.
SPIRT
USSO		AbstractP00476 C7600 00Code System to Calculate Stress-Strains from Transient Pressures.
SSC-L V3.3
USSO		AbstractP00400 I3090 00Transient Response in LMFBR System.
TEMPEST-BNWAbstractP00559 C7600 00Transient 3-D Thermohydraulics for FBR.
THACT-RRAbstractP00587 D0VAX 00Analysis of Thermal Hydraulics Transients in Research Reactor Core.
THYDE-B1/MOD2AbstractP00553 FM200 00Computer Code for PWR LOCA Thermohydraulic Transient Analysis.
TORACAbstractP00459 C0170 00Code System to Calculate Tornado-Induced Flow Material Transport.
TRISTAN-IJSAbstractP00537 IBMPC 00Multigroup Three-Dimensional Direct Integration Method Radiation Transport Analysis Code System.
TRUMPAbstractP00522 MNYCP 01Code System for Transient and Steady-State Temperature Distribution in Multidimensional Systems.
UHSAbstractP00390 IPS70 00Ultimate Heat Sink Cooling Pond and Spray Pond Analysis Models.
USINTAbstractP00415 MNYCP 00Code System to Calculate Heat and Mass Transfer In Concrete
UTSGAbstractP00379 I3033 00Code System for Calculating the Nonlinear Transient Behavior of a Natural Circulation U-Tube Steam Generator with Its Main Steam System.
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.