Consolidation of scientific and technological expertise to assess the reliability of reactor pressure vessel embrittlement prediction in particular for the arctic area plant (COBRA)

Ballesteros A., Bros J., Debarberis L., Sevini F., Erak D., Gezashchenko S., Kryukov A., Shtrombakh Y., Goloschapov S., Ionov, Pytkin Y., Anikeev Y., Banyuk G., Plusch A.5, Gillemot F., Tatar L.6,Petrosyan V.

Abstract:
The CORBA project was designed to solve the problem of irradiation conditions of the surveillance specimens of the reactor pressure vessel (RPV). The evaluation and prognosis of RPV material embrittlement in WWER were carried out and the allowable period of their safe operations were performed on the basis of impact test results of irradiated surveillance specimens. Thermocouples were installed throughout the instrumentation channels of the vessel head to measure directly the irradiation temperature in the surveillance position during reactor operation. The results show temperature correction was not required, when data of surveillance specimens were used for embrittlement assessment of WWER-440/213 reactor pressure vessel materials.

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