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 Subject: THERMODYNAMICS AND FLUID DYNAMICS
Package NameAbstractRSICC TapelistTitle
ALARM-B2AbstractP00218 I0360 00A Computer Code System for Analysis of a Large Break LOCA of a BWR.
BEACON MOD3AbstractP00402 CDCMF 00Code System for Thermal-Hydraulic Analysis of Nuclear Reactor Containments.
BWR-LTASAbstractC00485 I3033 01A Boiling Water Reactor Long-Term Accident Simulation Code.
COBRA4IAbstractP00419 MNYCP 00Code Sytem to Calculate Rod-Bundle and Core Thermal-Hydraulics.
COBRA-ENAbstractP00507 MNYCP 01Thermal-Hydraulic Transient Analysis of Reactor Cores.
CORTESAbstractP00404 I0360 00Code System for Thermal & Mechanical Analysis of Tees.
D2OAbstractP00398 PC486 00Code System for Computing Thermodynamic and Transport Properties of D2O.
DYN3D/M2AbstractP00579 I3090 00Reactivity Transients in Light H2O Reactors with Hexagonal Geometry.
FEMAXI 6 VER.1AbstractP00536 IBMPC 00Code System for Light Water Reactor Fuel Analysis.
FLODISAbstractP00417 I0360 00Code System to Calculate Thermal Response of FSV HTGR Core.
FRANCOAbstractP00363 MNYCP 00Finite Element Fuel Rod Analysis Code System.
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.
GAPCON-THERMALAbstractP00499 C7600 00Code System to Calculate Fuel Steady State & Transient Behavior.
GRACE-IIAbstractC00026 I3675 00Gamma Ray Kernel Integration Dose Rate and Heating Code-Cylinders and Spheres.
GT2R2AbstractP00483 ALLMF 00Code System to Calculate Fuel Rod Thermal Performance.
HATCHES-19AbstractD00206 PC586 02Thermodynamic Database for Radiochemical Modelling.
HEATING 7.3AbstractP00199 MNYCP 06Multidimensional, Finite-Difference Heat Conduction Analysis Code System.
JDL-THERMODYNAMAbstractM00007 MNYCP 00Thermodynamics: Frontiers and Foundations.
KFIXAbstractP00409 C7600 00Code System to Calculate Transient 2-Dimensional 2-Fluid Flow Dynamics.
KFIX 3DAbstractP00383 C7600 00Code System to Calculate Three-Dimensional Extension Two-Phase Flow Dynamics.
LAPUR6
USSO		AbstractP00395 PC586 02BWR Core Stability Measurements.
MAEROSAbstractP00466 C7600 00Code System for Multicomponent Aerosol Time Evolution.
MARCH2AbstractP00473 CDCMF 00Code System to Model LWR Meltdown Accident Response.
MCVIEWAbstractP00202 FM780 00View Factor Calculation for Three-Dimensional Geometries.
MINETAbstractP00490 CY000 00Momentum Integral Network Method for Thermal-Hydraulic Systems Analysis.
MINTEQAbstractP00494 DVX11 00Code System to Model Aqueous Geochemical Equilibria.
MORECAAbstractP00411 PC386 00Computer Code System for Simulating Modular High-Temperature Gas Cooled Reactor Core Heatup.
MOXY-MOD32AbstractP00385 I0360 00BWR Core Heat Transfer Code System.
MTR_PC 2.6AbstractC00674 PC386 00Modular Code System for Neutronics, Thermalhydraulics and Shielding Calculations.
ORCENT-2AbstractP00474 I3033 00Code System for Analysis of Steam Turbine Cycles Supplied by Light Water Reactors.
ORINC
USSO		AbstractP00439 I0360 00Code System for 1-D Implicit Heat Conduction Solution.
ORMDIN
USSO		AbstractP00399 I3033 002-D Nonlinear Inverse Heat Conduction.
ORSMAC
USSO		AbstractP00437 I3033 00Code System to Calculate Fluid Circulation Patterns Near Jets.
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.
PELE-1CAbstractP00461 C7600 00Code System for Fluid-Structure Interaction Analysis.
POLYRESAbstractP00438 MNYCP 00Richards Equation Solver; Rectangular Finite Volume Flux Updating Solution.
RCSLK9AbstractP00452 IBMPC 00Code System to Calculate Reactor Coolant System Leak Rate.
REFLUXAbstractP00403 I3033 00Code System to Predict LWR Reflood Heat Transfer.
SALE3DAbstractP00443 CY000 00ICEd-ALE Treatment of 3-D Fluid Flow.
SCORE-EVETAbstractP00442 C7600 00Code System for Three-Dimensional Hydraulic Reactor Core Analysis.
SCRELAAbstractP00408 SUN05 00Code System for Supercritical Water Cooled Reactor LOCA Analysis.
SIMMER II
USSO		AbstractC00691 MFMWS 00Code System for Two-Dinensional Sn-Neutronics and Fluid Dynamics.
SOLA-DFAbstractP00454 C7600 00Code System to Calculate Transient 2-Dimensional 2-Phase Flow.
SOLA-LOOPAbstractP00464 C7600 00Nonequilibrium, Drift-Flux Code System for Two-Phase Flow Network Analysis
SOPHIAAbstractC00857 MNYCP 00A Lagrangian-based computational fluid dynamics code for nuclear thermal hydraulics and safety applications.
SPIRT
USSO		AbstractP00476 C7600 00Code System to Calculate Stress-Strains from Transient Pressures.
STERNOAbstractC00057 C0000 00Two Dimensional Gamma-Ray Heating Kernel Integration Code.
TEMPEST-BNWAbstractP00559 C7600 00Transient 3-D Thermohydraulics for FBR.
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.
USINTAbstractP00415 MNYCP 00Code System to Calculate Heat and Mass Transfer In Concrete
USRHYDAbstractC00197 I3675 00Electron and X-Ray Energy Deposition and Hydrodynamics Code System.
UTSGAbstractP00379 I3033 00Code System for Calculating the Nonlinear Transient Behavior of a Natural Circulation U-Tube Steam Generator with Its Main Steam System.
WREM-TOODEE2AbstractP00469 ALLMF 002-D Time-Dependent Fuel Element, Thermal Analysis Code System.
123ABCDEFGHIJKLMNOPQRSTUVWXYZ
FAQ | ORNL Home | DOE | Site Map | Security Notice
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.