As a consequence of a total loss of AC power supply (station blackout) at a VVER-1000 leading to unavailability of major active safety systems, the safety criteria ensuring the safe operation of the nuclear power plant would be violated and core heat-up with possible core degradation could occur. A dedicated accident management measure (primary side depressurization) can be applied to reduce the primary pressure and to activate the injection from the passive emergency core cooling systems (accumulators). The analyses presented in this paper are aiming at both a detailed investigation of the accident sequence, taking into account the depressurization of the primary circuit, and the possibilities to prevent or at least to mitigate a damage of the reactor core so as to gain additional time for taking necessary countermeasures. The analyses are performed using the codes ASTEC and ATHLET developed by IRSN (Institut de Radioprotection et de Sûreté Nucléaire) and GRS (Gesellschaft für Anlagen- und Reaktorsicherheit mbH).

The studies presented are aiming at a detailed investigation of the behaviour of a VVER-1000/V-320 reactor and the containment structures during a postulated severe accident, including the ways and means by which these accidents may be prevented or mitigated. A hypothetical station blackout scenario (loss of the offsite electric power system concurrent with a turbine trip and unavailability of the emergency AC power system), belonging to the typical beyond design basis accidents, has been investigated. Station blackout results in reactor shut down, loss of feed water and trip of all reactor coolant pumps. Continuous evaporation of the secondary side leads to steam generators’ depletion followed by heating up of the core. In case of unavailability of essential safety systems the core will be severely damaged and finally the reactor pressure vessel (RPV) might fail. The analyses are performed using the integral code ASTEC commonly developed by IRSN (Institut de Radioprotection et de Suˆrete´ Nucle´aire) and GRS (Gesellschaft fu¨r Anlagen- und Reaktorsicherheit mbH). Code-to-code comparative analyses for the early thermal-hydraulic phase have been performed with the GRS code ATHLET. A large number of sensitivity calculations have been done regarding the axial core power distribution, heat losses, and RPV lower head modelling. The analyses have shown that, despite the considerable differences in the codes themselves, the calculation results are similar in terms of thermal hydraulic response. There are discrepancies in timings of phenomena, which are within the limitations of the physical models and the applied nodalizations. It was one objective of this investigation to evaluate the Severe Accident Management (SAM) procedures for VVER-1000 reactors, by for instance estimating the time available for taking appropriate decisions and preparing counter-measures. To evaluate the effect of possible operator actions, a SAM procedure (primary side depressurization) is included into the simulation. Without SAMs, the simulation provides plastic rupture of the RPV after approximately 4.3 h, while with SAMs, a prolongation of the vessel failure time is obtained by approximately 90 minutes. Currently, the late phase of the accident is investigated in more detail by comparing the lower head behaviour as simulated by ASTEC with results from dedicated finite element calculations. The work contributes to the reliability of the ASTEC code by means of plant applications.

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The research field focussing on the investigations and the analyses of severe accidents is an important part of the nuclear safety. To maintain the safety barriers as long as possible and to retain the radioactivity within the airtight premises or the containment, to avoid or mitigate the consequences of such events and to assess the risk, thorough studies are needed. On the one side, it is the aim of the severe accident research to understand the complex phenomena during the in-and ex-vessel phase, involving reactor-physics, thermal-hydraulics, physicochemical and mechanical processes. On the other side the investigations strive for effective severe accident management measures. This paper is focused on the possibilities for accident management measures in case of severe accidents. The reactor pressure vessel is the last barrier to keep the molten materials inside the reactor, and thus to prevent higher loads to the containment. To assess the behaviour of a nuclear power plant during transient or accident conditions, computer codes are widely used, which have to be validated against experiments or benchmarked against other codes. The analyses performed with the integral code ASTEC cover two accident sequences which could lead to a severe accident: a small break loss of coolant accident and a station blackout. The results have shown that in case of unavailability of major active safety systems the reactor pressure vessel would ultimately fail. The discussed issues concern the main phenomena during the early and late in-vessel phase of the accident, the time to core heat-up, the hydrogen production, the mass of corium in the reactor pressure vessel lower plenum and the failure of the reactor pressure vessel. Additionally, possible operator's actions and countermeasures in the preventive or mitigative domain are addressed. The presented investigations contribute to the validation of the European integral severe accidents code ASTEC for VVER-1000 type of reactors.