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RESEARCH PAPERS

A Description of a Study for the Thermal Anneal of Neutron-Embrittled Reactor Vessel Materials

[+] Author and Article Information
T. R. Mager

Pressurized Water Reactor Systems Division, Westinghouse Electric Corporation, Pittsburgh, Pa.

T. U. Marston

Electric Power Research Institute, Palo Alto, Calif.

J. Eng. Mater. Technol 100(3), 272-278 (Jul 01, 1978) (7 pages) doi:10.1115/1.3443490 History: Received February 24, 1978; Revised May 04, 1978; Online August 17, 2010

Abstract

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 bellline regions of the 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.

Copyright © 1978 by ASME
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