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: THERMAL HYDRAULICS
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.
BLOCKAGE V2.5RAbstractP00377 IBMPC 00Code System to Calculate Potential Plugging of Emergency Core Cooling Suction Strainers by Debris in a BWR.
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
CONTEMPT4AbstractP00397 MNYCP 00Code System for PWR & BWR Multicompartment Containment Analysis.
CONTEMPT-LT28B
USSO		AbstractP00387 C7600 00Code System to Predict Containment Pressure-Temperature Response To a Loss-Of-Coolant Accident.
ENTREE 1.4.0AbstractP00519 MNYWS 00BWR Core Simulation System for Space and Time Dependent Coupled Phenomena.
KRAKENAbstractC00877 PCX86 00Computational Reactor Analysis Framework.
MARCH2AbstractP00473 CDCMF 00Code System to Model LWR Meltdown Accident Response.
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.
ORINC
USSO		AbstractP00439 I0360 00Code System for 1-D Implicit Heat Conduction Solution.
ORSMAC
USSO		AbstractP00437 I3033 00Code System to Calculate Fluid Circulation Patterns Near Jets.
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.
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.
SCORE-EVETAbstractP00442 C7600 00Code System for Three-Dimensional Hydraulic Reactor Core Analysis.
SQUIRT VER2
USSO		AbstractP00583 PCX86 00Code System to Predict Leakage Rate and Area of Crack Opening for Cracked Pipes in Nuclear Power Plants.
SSC-L V3.3
USSO		AbstractP00400 I3090 00Transient Response in LMFBR System.
THYDE-B1/MOD2AbstractP00553 FM200 00Computer Code for PWR LOCA Thermohydraulic Transient Analysis.
THYDE-P2AbstractP00554 FV100 00Computer Code for PWR LOCA Thermohydraulic Transient Analysis.
TRISTAN-IJSAbstractP00537 IBMPC 00Multigroup Three-Dimensional Direct Integration Method Radiation Transport Analysis Code System.
USINTAbstractP00415 MNYCP 00Code System to Calculate Heat and Mass Transfer In Concrete
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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.