Structural integrity of water reactor pressure boundary components. Progress report ending 29 February 1976. [ASTM-A533-B; ASTM-A508]

Technical Report · Thu Aug 26 04:00:00 EDT 1976

OSTI ID:7227376

Loss, F J ^[1]

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1. ed.

The report describes progress in the following areas: (a) fatigue crack propagation in reactor pressure vessel steels in an air environment, (b) dynamic fracture toughness of 1-in. (25-mm) and precracked Charpy-V bend specimens under impact loading, (c) postirradiation notch ductility and properties recovery in reactor vessel steels, (d) factors contributing to variable resistance of structural steels to radiation embrittlement, and (e) the initial program plan to investigate the phenomena of warm prestress and plastic net ligament in support of thermal shock studies.

• Research Organization: Naval Research Lab., Washington, DC (USA)
• OSTI ID: 7227376
• Report Number(s): NRL-Report-8006; NRL-NUREG-1
• Country of Publication: United States
• Language: English

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36 MATERIALS SCIENCE
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CHEMICAL COMPOSITION
CONTAINERS
COOLING SYSTEMS
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MATERIALS TESTING
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PHYSICAL RADIATION EFFECTS
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PRESSURE VESSELS
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RADIATION EFFECTS
REACTOR COMPONENTS
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REACTOR MATERIALS
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