Improved Safety Injection Flow Model Associated with Target Depressurization during Severe Accident Management

Abstract

In Korea, Severe Accident Management Guidance (SAMG) for Optimized Power Reactor (OPR1000) has been developed through the research results related to the WOG SAMG [1], in which the operation phase shifts from the Emergency Operation Procedure (EOP) to the SAMG when the core exit temperature (CET) reaches 923K. There are seven mitigation strategies and each strategy is performed using several measurable parameters which can confirm the status of a nuclear power plant (NPP). Accordingly, operators have to monitor the variation of these parameters very closely. Among these parameters, the reactor core water level is one of the most important information to make provisions against a worsening situation. The water level can be a complement to the core temperature measurement because it can assess coolability of the core. Thus it helps to determine whether or not to carry out the third SAMG strategy, injecting coolant in to reactor coolant system (RCS) [2]. However, the accuracy of the core water level measured through the level meter is not fully guaranteed during severe accident. It is due to the volatile boiling conditions in the reactor core caused by the residual heat and oxidation heat. These uncertainties underscore the need for a supplementary methodology to predict the water level. Although many researches have been conducted to develop this methodology, there is still lack of studies for improving SAMG [3]. Therefore, the objective of this study is to develop a methodology for prediction of core water level using core exit parameters. It means the methodology using other variables whose reliability of measurement are higher than the level meter. The other variables are the CET which is already widely used as reliable parameter and core exit pressure. Generally, the measurement in the exit are thought to be more accurate than inside of the core. The combination of these two variables makes it possible to predict the thermal properties such as latent heat of water in core and enthalpy of evaporated steam. Also the small break loss of coolant accident (SBLOCA) without safety injection (SI) was selected as an initial event as this sequence is the main severe accident of OPR1000 [4]. The performance of the developed methodology is verified by using MELCOR 1.8.6 code which has been developed by Sandia National Laboratory and has been selected by the NRC as a legitimate tool [5].