NEUTRON EMBRITTLEMENT OF REACTOR PRESSURE VESSEL STEELS

Abstract
This report presents the status of observations at the U. S. Naval Research Laboratory on the embrittlement of steels which are commonly used for the primary pressure containment vessels of nuclear power plants. The demonstrated criterion of nil ductility transition (NDT) temperature provides the basis for meaningful analysis of neutron-induced embrittlement in reactor steels. Results to date indicate that the degree of embrittlement depends upon the material, the nuetron exposure, and the temperature during irradiation. These same variables also affect the degree of notch ductility recovery effected by postirradiation heat treatment. In addition, the time and temperature of heat treatment have been shown to play an important role in establishing the recovery pattern. The validity of these experimental observations are being tested through correlations with data from reactor surveillance programs and from specimens of the SL-1 reactor pressure vessel. Preliminary data from dosimetry in the SM-1A reactor permit the extension of experimental data to predict the increase in NDT of the reactor pressure vessel.

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