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The purpose of this paper is to consider a procedure of water-water energetic reactor (WWER) reactor pressure vessel (RPV) lifetime prediction at the stages of design and lifetime extension using the standard irradiation embrittlement parameters as defined in regulatory documents. A comparison is made of the brittle fracture resistance (BFR) values evaluated using two criteria: shift in the critical brittleness temperature ΔT_c or shift in the brittle-to-ductile transition temperature ΔT_p and without shifts (T_c and T_p).

The radiation resistance was determined using the following three approaches: calculation based on standard values ΔT_c and T_c0 or ΔT_p and T_p0 (a level of excessive conservatism); calculation based on standard value ΔT_c and actual value T_c0 or actual values ΔT_p and T_p0 (the level of realistic conservatism); or calculation based on actual values of T_c and T_c0 or T_p and T_p0 (the level of actual conservatism). The BFR was evaluated based on the results of testing the specimens subjected to irradiation in research reactors as well as surveillance specimens subjected to irradiation immediately under operating conditions.

The excessive conservatism in determining the actual lifetime of nuclear reactor vessel materials can be eliminated by using the immediate values of critical brittleness temperature and ductile-to-brittle transition temperature.

Obtained results can be applied to extend WWER vessel operating time at the stages of designing and operation due to substantiated decrease in conservatism. And it will allow carrying out a statistical substantiated assessment of the resistance to brittle fracture of the RPV steels.