Assessment of the Importance of Different Factors for the Delay in Reaching a Serious RBMK Core State with Complete Loss of Power to the Power Unit

Abstract

The computational results obtained for different scenarios of an accident caused by the loss of all power to a power-generating unit with RBMK and no possibility of restoring power from backup sources (dieselgenerators and others) are presented. The main problem is to determine the effect of different factors on the time to failure of physical safety barriers for the RBMK core. It is shown that passive water reserves in water tanks in the reactor's emergency cooling system and deaerators as well as means for lower the pressure in the circulation loop make it possible to slow down seal-failure onset to fuel elements and channel tubes to 20-24 h after an accident starts. The diagnostics indicators making it possible to judge the state of reactor facilities are analyzed.The accident at the NPP in Fukushima (NPP) has reconfirmed the importance of analyzing unanticipated accidents and the control actions taken by operators to mitigate them. In the present article, we present the results of additional calculations of an accident caused by complete loss of power to a power-generating unit with RBMK and the impossibility of restoring power by using backup sources (diesel generators and others). The main problem of these calculations is to determine the effect of different factors on the time to failure of physical safety barrier of the core.Calculations performed to validate the guides for controlling unanticipated accidents have shown that in the case of complete loss of power the time to core heat-up is approximately 2 h and to failure of its elements another 4 h [1, 2]. At the same time, replenishment of the reactor with water from sources which are accessible when all pumps and other equipment power by diesel generators are inoperative makes it possible to increase the time to core damage from 6 to 24 h. The analysis was performed for second-generation power units (Nos. 3, 4 Kursk NPP, Nos. 1, 2 Smolensk NPP) using the Relap5/mod3.2 code for four scenarios: 1) with failure of emergency reactor cooling systems (ERCS), 2) with water supplied from water tanks of ERCS in both circuits of the multiple forced circulation loop (MFCL), 3) with forced opening of one fast-acting reducing setup for dumping steam into the turbine condenser (BRU-K) with and without connecting a feed channel, and 4) with water supplied from ERCS water tanks with forced opening of one BRU-K valve and connecting a feed channel.The model developed for the RBMK power-generating unit included the main equipment of the MFCL, steam-line systems, including a turbine, condensate-feed channel (from deaerators to steam drums), the ERCS system, elements of the control system and pressure protection of the MFCL and others. The circulation loop was modeled with two circuits, one of which included three equivalent technological channels and two contained three equivalent group distributing headers with 12 equivalent channels (Fig. 1). A feature of the model was that the thermal radiation between the fuel elements and channel tube was taken into account.