Proceedings of the seminar on leak before break in reactor piping and vessels

Conference · Tue Apr 01 04:00:00 EST 1997

OSTI ID:468617

Faidy, C ^[1]; Gilles, P ^[2]

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1. ed.; Electricite de France, Villeurbanne (France)
2. ed.; Framatome, Paris (France)

The objective of the seminar was to present the current state of the art in Leak-Before-Break (LBB) methodology development, validation, and application in an international forum. With particular emphasis on industrial applications and regulatory policies, the seminar provided an opportunity to compare approaches, experiences, and codifications developed by different countries. The seminar was organized into four topic areas: status of LBB applications; technical issues in LBB methodology; complementary requirements (leak detection and inspection); LBB assessment and margins. As a result of this seminar, an improved understanding of LBB gained through sharing of different viewpoints from different countries, permits consideration of: simplified pipe support design and possible elimination of loss-of-coolant-accident (LOCA) mechanical consequences for specific cases; defense-in-depth type of applications without support modifications; support of safety cases for plants designed without the LOCA hypothesis. In support of these activities, better estimates of the limits to the LBB approach should follow, as well as an improvement in codifying methodologies. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

• 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)
• Sponsoring Organization: Nuclear Regulatory Commission, Washington, DC (United States); CEA, 75 - Paris (France); International Atomic Energy Agency, Vienna (Austria); Organization for Economic Cooperation and Development, 75 - Paris (France); Societe Francaise d`Energie Nucleaire (SFEN), 75 - Paris (France)
• OSTI ID: 468617
• 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
CRACK PROPAGATION
ECCS
FRACTURE MECHANICS
IN-SERVICE INSPECTION
LEADING ABSTRACT
LEAK TESTING
MEETINGS
NUCLEAR POWER PLANTS
PIPES
REACTOR COOLING SYSTEMS
REACTOR MATERIALS
REACTOR SAFETY
REGULATIONS