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TH_Code_Descriptions - Reactor Thermal Hydraulics/Heat...

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Reactor Thermal Hydraulics/Heat Transfer/Safety Code Packages from RSICC RELAP5/MOD1 : LWR Loss of Coolant Analysis. (RSICC CODE PACKAGE PSR-423) RELAP5 was developed to describe the behavior of a light water reactor (LWR) subjected to postulated transients such as loss of coolant from large or small pipe breaks, pump failures, etc. RELAP5 calculates fluid conditions such as velocities, pressures, densities, qualities, temperatures; thermal conditions such as surface temperatures, temperature distributions, heat fluxes; pump conditions; trip conditions; reactor power and reactivity from point reactor kinetics; and control system variables. In addition to reactor applications, the program can be applied to transient analysis of other thermal-hydraulic systems with water as the fluid. This package contains RELAP5/MOD1 for CDC, VAX or IBM mainframe computers. COBRA-EN: Code System for Thermal-Hydraulic Transient Analysis of Light Water Reactor Fuel Assemblies and Cores. (RSICC CODE PACKAGE PSR-507) Starting from a steady-state condition in a LWR core or fuel element, COBRA-EN can be used to simulate the thermal-hydraulic transient response to user-supplied changes of the total power, of the outlet pressure and of the inlet enthalpy and mass flow rate. The COBRA-EN code was developed in the eighties to be used as the thermal-hydraulic section in the successive versions of the NORMA [Brega 1995], QUARK [Alloggio 1994] and NORMA-FP [Brega 1991] computer programs, which were all designed for application to light water power reactors. The first was designed as a long-term reactivity simulator, the second as a core dynamics analyzer and the last one to unfold the flux and power fine structure in the large homogenized nodes generally used by the first two. COBRA4I: Code System to Calculate Rod-Bundle and Core Thermal-Hydraulics. (RSICC CODE PACKAGE PSR-419) COBRA4I performs steady-state and transient thermal-hydraulic analysis of rod bundle nuclear fuel elements and cores via the subchannel analysis method. TRAC-PF1: Best-Estimate Analysis PWR LOCA. (RSICC CODE PACKAGE PSR-481)
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TRAC-PF1 performs best estimate analyses of loss of coolant accidents and other transients in pressurized light water reactors. The program can also be used to model a wide range of thermal hydraulic experiments in reduced scale facilities. Models employed include reflood, multi-dimensional two-phase flow, nonequilibrium thermodynamics, generalized heat transfer, and reactor kinetics. Automatic steady-state and dump/restart capabilities are provided. The changes reported in TRACNEWS issues through Number 7 are incorporated in this release. TRAC-PF1 was developed on a CDC computer at Los Alamos National Laboratory. The PC version of TRAC-PF1 was converted at CNEN in 1989 and has not been updated since that time.
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  • Spring '11
  • Schubring
  • Pressurized water reactor, Light water reactor, Nuclear fuel, U.S. Nuclear Regulatory Commission, RSICC CODE PACKAGE

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