TRAC-BF1/MOD1: An advanced best-estimate computer program for BWR accident analysis: User`s guide. Volume 2

Technical Report · Mon Jun 01 04:00:00 EDT 1992

OSTI ID:10160982

Rettig, W H; Wade, N L ^[1]

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1. eds.; EG and G Idaho, Inc., Idaho Falls, ID (United States)

The TRAC-BWR code development program at the Idaho National Engineering Laboratory has developed versions of the Transient Reactor Analysis Code (TRAC) for the US Nuclear Regulatory Commission and the public. The TRAC-BF1/MODI version of the computer code provides a best-estimate analysis capability for analyzing the full range of postulated accidents in boiling water reactor (BWR) systems and related facilities. This version provides a consistent and unified analysis capability for analyzing all areas of a large- or small-break loss-of-coolant accident (LOCA), beginning with the blowdown phase and continuing through heatup, reflood with quenching, and, finally, the refill phase of the accident. Also provided is a basic capability for the analysis of operational transients up to and including anticipated transients without scram (ATWS). The TRAC-BF1/MOD1 version produces results consistent with previous versions. Assessment calculations using the two TRAC-BFI versions show overall improvements in agreement with data and computation times as compared to earlier versions of the TRAC-BWR series of computer codes.

• Research Organization: Nuclear Regulatory Commission, Washington, DC (United States). Div. of Systems Research; EG and G Idaho, Inc., Idaho Falls, ID (United States)
• Sponsoring Organization: Nuclear Regulatory Commission, Washington, DC (United States)
• DOE Contract Number: AC07-76ID01570
• OSTI ID: 10160982
• Report Number(s): NUREG/CR--4356-Vol.2; EGG--2626-Vol.2; ON: TI92017873
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210100
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900
99 GENERAL AND MISCELLANEOUS
990200
ATWS
BLOWDOWN
BWR TYPE REACTORS
COMPUTER CALCULATIONS
HEAT TRANSFER
HYDRAULICS
LOSS OF COOLANT
MATHEMATICS AND COMPUTERS
POWER REACTORS
NONBREEDING
LIGHT-WATER MODERATED
BOILING WATER COOLED
REACTOR SAFETY
T CODES