Leak-before-break analysis of Type 304 stainless steel piping

Conference · Sun Jan 01 04:00:00 EST 1989

OSTI ID:5960909

Awadalla, N. G.; Sindelar, R. L.; Daugherty, W. L.; Mehta, H. S.; Ranganath, S.

The nuclear materials production reactors at the Savannah River Plant (SRP) were designed and built in the 1950's and have operated successfully since that time. Unlike commercial power reactors, the production reactors are moderated and cooled by heavy water and are operated at moderately low temperatures and internal pressures. In addition, the entire primary coolant pressure boundary is constructed of Type 304 stainless steel or its cast equivalent, CF-8, except for seals, gaskets and other serviceable parts. Due to the low applied stresses coupled with high toughness, the primary coolant piping is highly tolerant of defects. In the operational history of the plant, several instances of minor leakage from stress corrosion cracks have occurred in the piping, thus exemplifying a Leak-Before-Break (LBB) capacity of the system. Fundamentally, LBB capability provides the assurance that a postulated through-wall crack could be detected by the resulting leakage before the onset of crack instability and the ensuing pipe failure. The Leak-Before-Break demonstration of the SRP primary coolant piping has been formalized through a detailed fracture assessment of postulated flaws in piping with the result that the through-wall crack length at instability conditions is well in excess of the crack length corresponding to the minimum detectable leak rate of the primary coolant system. Additional elements supporting the demonstration of SRP piping integrity include the in-service inspection program and the moderator leak detection system. The extent, frequency and method of non-destructive inspection for SRP piping conforms in general with Section XI of the ASME code and NUREG-0313 requirements for commercial Boiling Water Reactors.

• Research Organization: Du Pont de Nemours (E.I.) and Co., Aiken, SC (USA). Savannah River Lab.
• DOE Contract Number: AC09-76SR00001
• OSTI ID: 5960909
• Report Number(s): DP-MS-88-146; CONF-890855-20; ON: DE89010460
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
220200* -- Nuclear Reactor Technology-- Components & Accessories
42 ENGINEERING
420500 -- Engineering-- Materials Testing
ALLOYS
AUSTENITIC STEELS
CHEMICAL ANALYSIS
CHEMICAL REACTIONS
CHROMIUM ALLOYS
CHROMIUM-NICKEL STEELS
CORROSION
CORROSION RESISTANT ALLOYS
FAILURES
FRACTURES
HEAT RESISTANT MATERIALS
HEAT RESISTING ALLOYS
HIGH ALLOY STEELS
IRON ALLOYS
IRON BASE ALLOYS
LEAKS
MATERIALS
NATIONAL ORGANIZATIONS
NICKEL ALLOYS
NONDESTRUCTIVE ANALYSIS
PIPES
SAVANNAH RIVER PLANT
STAINLESS STEEL-304
STAINLESS STEELS
STEEL-CR19NI10
STEELS
STRESS ANALYSIS
STRESS CORROSION
US AEC
US DOE
US ERDA
US ORGANIZATIONS