Authors & Affiliations

Buchatsky A.A.2, Vilensky O.Yu.1, Gulenko A.G.2, Kaidalov V.B.1, Pristrom S.A.1
1JSC “Afrikantov OKB Mechanical Engineering”, Nizhny Novgorod, Russia
2Federal State Unitary Enterprise Central Research Institute of Structural Materials “Prometey”, Saint Petersburg, Russia

Pristrom S.A. – Lead Design Engineer, Afrikantov OKB Mechanical Engineering. Contacts: 15 Burnakovsky proezd, Nizhny Novgorod, Russia, 603074. Tel.: (831)246-94-95; e-mail: This email address is being protected from spambots. You need JavaScript enabled to view it.
Vilensky O.Yu. – Dr. Sci. (Tech.), Head of Strengh Validation Department, Afrikantov OKB Mechanical Engineering.
Kaidalov V.B. – Dr. Sci. (Tech.), Chief Specialist, Afrikantov OKB Mechanical Engineering.
Buchatsky A.A. – Dr. Sci. (Tech.), Senior Researcher, Federal State Unitary Enterprise Central Research Institute of Structural Materials "Prometey".
Gulenko A.G. – Dr. Sci. (Tech.), Lead Researcher, Federal State Unitary Enterprise Central Research Institute of Structural Materials "Prometey".

Abstract

Study relevance for fast neutron reactor vessel strength analysis under beyond-design-basis accident is stipulated by importance of this component for safety requirements. This reactor feature is high temperatures both in normal operation conditions and beyond-design-basis accidents. In the aggregate with duration of these modes, these temperatures determine the requirements for the methodical approach and, especially, for the analytical and experimental studies of short- and long-term mechanical characteristics of vessel structural materials within temperature range of 600-800°C on a time basis of 1000 h.

As the result of the analysis, the time intervals of BN-600 reactor vessel integrity maintenance in the most loaded cross-sections for various increased temperatures within 600-800°C were obtained for the accident selected as the analytical beyond-design-basis one with the sufficient coolant temperature increase up to ~800°C. This analysis is required to evaluate the time sufficient to take measures for accident management and reactor vessel temperature decrease down to the acceptable one a ccording to the conditions of safe operation values.

The evaluation conservatism consisting in determination of crack initiation stage only, without the process of crack development up to the critical size assumes further updating of the used procedure as for development of evaluation model for the initiated crack growth based on creeping mechanism.

Keywords
reactor vessel, beyond-design-basis accident, short- and long-term mechanical characteristics, strain-stress state, long-term strength, crack initiation

Article Text (PDF, in Russian)

References

UDC 539.3

Problems of Atomic Science and Technology. Series: Nuclear and Reactor Constants, 2016, issue 4, 4:10