Description of a study for the thermal anneal of neutron-embrittled reactor vessel materials

Journal Article · Sat Jul 01 04:00:00 EDT 1978 · J. Eng. Mater. Technol.; (United States)

DOI:https://doi.org/10.1115/1.3443490· OSTI ID:6561098

Mager, T R ^[1]; Marston, T U

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1. Westinghouse Electric Corp., Pittsburgh

A number of early light water reactor plants were constructed from materials having low initial Charpy upper shelf values and high copper and phosphorus content. As these elements have been shown to contribute the most to the radiation sensitivity of reactor pressure vessel material, there is a possibility that thermal annealing of the beltline regions of these vessels may become necessary to meet Nuclear Regulatory Commission requirements for continued operation. Recognizing the possibility that thermal annealing treatment of the reactor vessel may become a reality, a program was started to determine the kinetics and mechanisms of thermal annealing to restore preservice fracture toughness and to develop engineering procedures for ready application to large-scale nuclear pressure vessels. This paper presents the program scope required for establishing the feasibility of and methodology for an in situ thermal anneal of an embrittled reactor vessel.

• OSTI ID: 6561098
• Journal Information: J. Eng. Mater. Technol.; (United States), Journal Name: J. Eng. Mater. Technol.; (United States) Vol. 100:3; ISSN JEMTA
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210100* -- Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled
ALLOYS
ANNEALING
CONTAINERS
EMBRITTLEMENT
HEAT TREATMENTS
IRON ALLOYS
IRON BASE ALLOYS
PHYSICAL RADIATION EFFECTS
PRESSURE VESSELS
PWR TYPE REACTORS
RADIATION EFFECTS
REACTORS
STEELS
WATER COOLED REACTORS
WATER MODERATED REACTORS