RSICC Catalog

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 Name	Abstract	RSICC Tapelist	Title
ATHENA_2D	Abstract	P00431 MNYCP 00	Code System For Simulation Of Hypothetical Recriticality Accidents in a Thermal Neutron Spectrum.
BEACON MOD3	Abstract	P00402 CDCMF 00	Code System for Thermal-Hydraulic Analysis of Nuclear Reactor Containments.
BLOCKAGE V2.5R	Abstract	P00377 IBMPC 00	Code System to Calculate Potential Plugging of Emergency Core Cooling Suction Strainers by Debris in a BWR.
COBRA-EN	Abstract	P00507 MNYCP 01	Thermal-Hydraulic Transient Analysis of Reactor Cores.
COBRA-SFS VERSION 6.0	Abstract	P00614 MNYCP 02	COBRA-SFS Thermal-Hydraulic Analysis of Multi-Assembly Spent Fuel Storage and Transportation Systems.
COMMIX-1B USSO	Abstract	P00393 DVX11 00	3-D Single-Phase Thermal Hydraulics
COMMIX-1B USSO	Abstract	P00393 I3033 00	3-D Single-Phase Thermal Hydraulics
COMMIX-1C USSO	Abstract	P00393 MNYCP 00	3-D Single-Phase Thermal Hydraulics
CONTEMPT4	Abstract	P00397 MNYCP 00	Code System for PWR & BWR Multicompartment Containment Analysis.
CONTEMPT-LT28B USSO	Abstract	P00387 C7600 00	Code System to Predict Containment Pressure-Temperature Response To a Loss-Of-Coolant Accident.
ENTREE 1.4.0	Abstract	P00519 MNYWS 00	BWR Core Simulation System for Space and Time Dependent Coupled Phenomena.
MARCH2	Abstract	P00473 CDCMF 00	Code System to Model LWR Meltdown Accident Response.
NORMA	Abstract	P00471 PC586 00	Code System to Solve Burnup Dependent Neutron Diffusion Equations in Two and Three Dimensions.
NORMA-FP	Abstract	P00470 PC586 00	Code System to Perform Neutronic and Thermal-Hydraulic Subchannel Analysis from Converged Coarse-Mesh Nodal Solutions.
ORINC USSO	Abstract	P00439 I0360 00	Code System for 1-D Implicit Heat Conduction Solution.
ORSMAC USSO	Abstract	P00437 I3033 00	Code System to Calculate Fluid Circulation Patterns Near Jets.
PARET-ANL	Abstract	P00516 MNYCP 00	Code System to Predict Consequences of Nondestructive Accidents in Research and Test Reactor Cores.
PARET-ANL(NESC)	Abstract	P00565 MNYCP 00	Code System to Predict Consequences of Nondestructive Accidents in Research and Test Reactor Cores.
PMK2-VVER440-REPORTS	Abstract	M00012 MNYCP 00	Results of the Experiments Performed in the PMK-2 Facility for VVER Safety Studies.
RELAP5/MOD1/029_EXE 810	Abstract	P00423 C0176 01	Reactor System Transient Code.
SCDAP/RELAP5/MOD3.3-EXE 810	Abstract	P00581 MNYCP 01	A Best-Estimate Transient Simulation of Light Water Reactor Coolant Systems During a Severe Accident.
SCORE-EVET	Abstract	P00442 C7600 00	Code System for Three-Dimensional Hydraulic Reactor Core Analysis.
SQUIRT VER2 USSO	Abstract	P00583 PCX86 00	Code System to Predict Leakage Rate and Area of Crack Opening for Cracked Pipes in Nuclear Power Plants.
SSC-L V3.3 USSO	Abstract	P00400 I3090 00	Transient Response in LMFBR System.
THYDE-B1/MOD2	Abstract	P00553 FM200 00	Computer Code for PWR LOCA Thermohydraulic Transient Analysis.
THYDE-P2	Abstract	P00554 FV100 00	Computer Code for PWR LOCA Thermohydraulic Transient Analysis.
TRISTAN-IJS	Abstract	P00537 IBMPC 00	Multigroup Three-Dimensional Direct Integration Method Radiation Transport Analysis Code System.
USINT	Abstract	P00415 MNYCP 00	Code System to Calculate Heat and Mass Transfer In Concrete

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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.