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.

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