LBB evaluation for a typical Japanese PWR primary loop by using the US NRC approved methods

Conference · Mon Mar 31 23:00:00 EST 1997

OSTI ID:489295

Swamy, S A; Bhowmick, D C; Prager, D E ^[1]

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1. Westinghouse Nuclear Technology Division, Pittsburgh, PA (United States)

The regulatory requirements for postulated pipe ruptures have changed significantly since the first nuclear plants were designed. The Leak-Before-Break (LBB) methodology is now accepted as a technically justifiable approach for eliminating postulation of double-ended guillotine breaks (DEGB) in high energy piping systems. The previous pipe rupture design requirements for nuclear power plant applications are responsible for all the numerous and massive pipe whip restraints and jet shields installed for each plant. This results in significant plant congestion, increased labor costs and radiation dosage for normal maintenance and inspection. Also the restraints increase the probability of interference between the piping and supporting structures during plant heatup, thereby potentially impacting overall plant reliability. The LBB approach to eliminate postulating ruptures in high energy piping systems is a significant improvement to former regulatory methodologies, and therefore, the LBB approach to design is gaining worldwide acceptance. However, the methods and criteria for LBB evaluation depend upon the policy of individual country and significant effort continues towards accomplishing uniformity on a global basis. In this paper the historical development of the U.S. LBB criteria will be traced and the results of an LBB evaluation for a typical Japanese PWR primary loop applying U.S. NRC approved methods will be presented. In addition, another approach using the Japanese LBB criteria will be shown and compared with the U.S. criteria. The comparison will be highlighted in this paper with detailed discussion.

• Research Organization: Nuclear Regulatory Commission, Washington, DC (United States). Div. of Engineering Technology; Electricite de France (EDF), 69 - Villeurbanne (France); Battelle, Columbus, OH (United States)
• OSTI ID: 489295
• Report Number(s): NUREG/CP--0155; CONF-9510432--; ON: TI97004806
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
22 GENERAL STUDIES OF NUCLEAR REACTORS
JAPANESE ORGANIZATIONS
LEAKS
PIPES
PRIMARY COOLANT CIRCUITS
PWR TYPE REACTORS
SAFETY ANALYSIS
US NRC