WWER-1000 reactors embrittlement

Abstract This chapter describes the embrittlement processes in WWER reactor pressure vessel (RPV) materials during operation - radiation embrittlement and thermal ageing. Current trend curves for both types of WWER RPV materials are given and explained. Surveillance specimen programmes are shown, as their results are used for RPV integrity and lifetime evaluation. Finally, anneahng of the RPV is proposed as the most efficient mitigation measure. 

Key words WWER reactor pressure vessel, radiation embrittlement,... 

Characteristics of embrittlement of WWER reactor pressure vessel (RPV) materials... 

Irradiation Embrittlement of Reactor Pressure Vessels (RPVs) Table 3.4 Requirements for chemical composition of WWER RPV materials... 

Embrittlement of reactor pressure vessels (RPVs) in WWER-type reactors... 

Embrittlement of RPVs in WWER-type reactors 111 Table 5.1 Values of irradiation embrittlement of steel 15Kh2MFA... 

IAEA, Guidelines for Prediction of Irradiation Embrittlement of Operating WWER-440 Reactor Pressure Vessels, IAEA TECDOC 1442, International Atomic Energy Agency, Vienna, 2005. 

The Russian predictive embrittlement method is based primarily on test reactor data and utilizes the shift in CVN impact energy properties essentially at the 47 J temperature (PNAE, 1989). The effects of Cu, P and O (E > 0.5 MeV) are included in the method. A new local approach called the Unified Curve (similar to the Master Curve method) for assessing vessel integrity has recently has been added for structural analysis purposes (Margolin et ai, 2007). Additionally, an IAEA activity on embrittlement prediction for WWER-440 RPVs was completed (IAEA, 2005), with the published IAEA report providing recommended guidelines based on a larger database than previously available. 

Amayev A D, Kryukov A M and Sokolov M A (1993), Recovery of the transition temperature of irradiated WWER 440 vessel metal by annealing, pp. 369-379 in Radiation Embrittlement of Nuclear Reactor Pressure Vessel Steels An International Review (Fourth Volume), ASTM STP1170, L E Steele, ed., American Society for Testing and Materials, Philadelphia, PA. 

Research work has been undertaken to determine the dependence of irradiation embrittlement on material chemistry, heat treatment and service factors (irradiation conditions, temperature, coolant chemistry, etc.) in Russian research reactors and WWER-440 reactors. The irradiations were carried out mainly on surveillance specimens irradiated in nuclear power plants in locations where the inlet temperature was 270 °C (for WWER-440). The intent was to provide carefully controlled irradiation conditions in terms of temperature and neutron spectrum. The resulting major conclusion was that a substantial body of data established an irradiation-induced increase of the brittle fracture temperature, (similar to Tnj), of the general form of (CF) (FF) ... 

Regarding WWER-1000 reactors, Kryukov et al define the regulation for embrittlement as ... 

---