TRAC code improvements
Description
The Transient Reactor Analysis Code (TRAC) is an advanced best-estimate systems code for analyzing LWR accidents. It is being developed at the Los Alamos National Laboratory under the sponsorship of the Reactor Safety Research Division of the US Nuclear Regulatory Commission. A preliminary TRAC version consisting of only one-dimensional components was completed in December 1976. The first publicly released version was TRAC-P1, completed in December 1977. It is described in the Los Alamos report LA-7279-MS. The TRAC-P1 program was designed primarily for the analysis of large-break loss-of-coolant accidents (LOCAs) in pressurized water reactors (PWRs). Because of its versatility, however, it can be applied directly to many analyses ranging from blowdowns in simple pipes to integral LOCA tests in multiloop facilities. A refined version, called TRAC-P1A, was released to the National Energy Software Center (NESC) in March 1979. It is described in the Los Alamos report LA-7777-MS. TRAC-PD2 contains improved reflood and heat-transfer models and improvements in the numerical solution methods. TRAC-PF1 was designed to improve the ability of TRAC-PD2 to handle small-break LOCAs and other transients. TRAC-PF1/MOD1 was designed to provide full balance-of-plant modeling capabilities. Some distinguishing characteristics of TRAC-PF1/MOD1 are summarized
Additional details
Publishing Information
- Imprint Title
- Proceedings of the eleventh water reactor safety research information meeting. Volume 2
- Journal Page Range
- p. 200-206.
- Report number
- NUREG/CP--0048-Vol.2
Conference
- Title
- 11. NRC water reactor safety research information meeting.
- Dates
- 14-24 Oct 1983.
- Place
- Gaithersburg, MD (USA).
INIS
- Country of Publication
- United States
- Country of Input or Organization
- United States
- INIS RN
- 16007000
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Resource subtype / Literary indicator
- Conference
- Descriptors DEI
- BLOWDOWN; BWR TYPE REACTORS; COMPUTER CODES; ECCS; HEAT TRANSFER; HYDRODYNAMICS; LOSS OF COOLANT; PRIMARY COOLANT CIRCUITS; PWR TYPE REACTORS; REACTOR ACCIDENTS; REACTOR COOLING SYSTEMS; REACTOR SAFETY; RESEARCH PROGRAMS; SCALE MODELS; SECONDARY COOLANT CIRCUITS; TRANSIENTS; TWO-PHASE FLOW; US NRC
- Descriptors DEC
- ACCIDENTS; COOLING SYSTEMS; ENERGY TRANSFER; ENRICHED URANIUM REACTORS; FLUID FLOW; FLUID MECHANICS; MECHANICS; NATIONAL ORGANIZATIONS; POWER REACTORS; REACTOR COMPONENTS; REACTOR PROTECTION SYSTEMS; REACTORS; SAFETY; STRUCTURAL MODELS; THERMAL REACTORS; US ORGANIZATIONS; WATER COOLED REACTORS; WATER MODERATED REACTORS