Reactor primary coolant system pipe rupture study. Progress report No. 35, January--March 1976. [BWR]

Technical Report · Sat May 01 00:00:00 EDT 1976

DOI:https://doi.org/10.2172/7252077· OSTI ID:7252077

The pipe rupture study is designed to extend the understanding of failure-causing mechanisms and to provide improved capability for evaluating reactor piping systems to minimize the probability of failures. Following a detailed review to determine the effort most needed to improve nuclear system piping (Phase I), analytical and experimental efforts (Phase II) were started in 1965. This progress report summarizes the recent accomplishments of a broad program in (a) basic fatigue crack growth rate studies focused on LWR primary piping materials in a simulated BWR primary coolant environment, and (b) studies directed at quantifying weld sensitization in Type-304 stainless steel.

• Research Organization: General Electric Co., San Jose, Calif. (USA). Boiling Water Reactor Systems Dept.
• DOE Contract Number: AT(04-3)-189
• OSTI ID: 7252077
• Report Number(s): GEAP-10207-35
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
BWR TYPE REACTORS
PRIMARY COOLANT CIRCUITS
PIPES
RUPTURES
CARBON STEELS
CRACKS
FATIGUE
INCONEL 600
STAINLESS STEEL-304
STAINLESS STEEL-304L
STRESS CORROSION
WELDED JOINTS
ALLOYS
CHEMICAL REACTIONS
CHROMIUM ALLOYS
CHROMIUM STEELS
CHROMIUM-NICKEL STEELS
COOLING SYSTEMS
CORROSION
CORROSION RESISTANT ALLOYS
FAILURES
HEAT RESISTING ALLOYS
INCONEL ALLOYS
IRON ALLOYS
IRON BASE ALLOYS
JOINTS
MECHANICAL PROPERTIES
NICKEL ALLOYS
NICKEL BASE ALLOYS
NIOBIUM ALLOYS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTORS
STAINLESS STEELS
STEELS
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210100 - Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled