Published 1993 | Version v1
Book Metadata-only

TRAC analysis of passive containment cooling system performance

Creators

  • 1. Toshiba Corp., Kawasaki (Japan). Nuclear Engineering Lab.
  • 2. Toshiba Corp., Yokohama (Japan). Isogo Engineering Center

Description

A passive containment cooling system (PCCS) is a promising concept to improve the reliability of decay heat removal during an accident. Toshiba has carried out analytical studies for PCCS development in addition to experimental studies, using a best estimate thermal hydraulic computer code TRAC. In order to establish an analytical model for the PCCS performance analysis, it is necessary for the analytical model to be qualified against experimental results and thoroughly address the phenomena important for PCCS performance analysis. In this paper, the TRAC qualification for PCCS application is reported. A TRAC model has been verified against a drain line break simulation test conducted at the PCCS integral test facility, GIRAFFE. The result shows that the TRAC model can accurately predict the major system response and the PCCS performance in the drain line break test. In addition, the results of several sensitivity analyses, showing various points concerning the modeling in the PCCS performance analysis, have been reported. The analyses have been carried out for the SBWR and the analytical points are closely related to important phenomena which can affect PCCS performance

Part of:
2nd ASME-JSME international conference on nuclear engineering -- 1993

Additional details

Publishing Information

Publisher
American Society of Mechanical Engineers.
Imprint Place
New York, NY (United States)
ISBN
0-7918-0636-7
Imprint Title
2nd ASME-JSME international conference on nuclear engineering -- 1993
Imprint Pagination
770 p.
Journal Page Range
p. 143-150.

Conference

Title
2. Japan Society of Mechanical Engineers (JSME)/American Society of Mechanical Engineers (ASME) joint international conference on nuclear engineering.
Dates
21-24 Mar 1993.
Place
San Francisco, CA (United States).

INIS

Country of Publication
United States
Country of Input or Organization
United States
INIS RN
25058683
Subject category
S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
Resource subtype / Literary indicator
Conference, Numerical Data
Descriptors DEI
AFTER-HEAT REMOVAL; BWR TYPE REACTORS; ENGINEERED SAFETY SYSTEMS; LOSS OF COOLANT; MATHEMATICAL MODELS; PERFORMANCE; REACTOR SAFETY; T CODES; TEST FACILITIES; THEORETICAL DATA
Descriptors DEC
ACCIDENTS; COMPUTER CODES; DATA; ENRICHED URANIUM REACTORS; INFORMATION; NUMERICAL DATA; POWER REACTORS; REACTOR ACCIDENTS; REACTORS; REMOVAL; SAFETY; THERMAL REACTORS; WATER COOLED REACTORS; WATER MODERATED REACTORS

Optional Information

Secondary number(s)
CONF-930352--.