Analyses of Fission Product Retention under ISLOCA using MELCOR for APR 1400

An Interfacing System Loss-of-Coolant Accident (ISLOCA) is one of bypass scenarios considered in Nuclear Power Plants (NPPs). This accident can occur due to an unisolated rupture (outside of containment) of a piping connected to Reactor Coolant System (RCS). Since there are not appropriate strategies to prevent core damage during the ISLOCA, it is assumed that the probability of the initial event is as same as the Core Damage Frequency (CDF) of the ISLOCA in Probabilistic Safety Assessment (PSA). Even though the CDF for the ISLOCA is considerably low, this can lead to a significant and direct release of fission products from the RCS into the environment. Therefore, it is important to estimate the behavior of the fission products during the ISLOCA. The U.S. Nuclear Regulatory Commission (NRC) has carried out the State-of-the-Art Reactor Consequence Analyses (SOARCA) project to develop best estimates of the offsite radiological health consequences for potential severe accidents [1]. In this project, various scenarios were analyzed with MELCOR [2] to estimate the accident progression and the behavior of the fission products. Especially, the state-of-the-art modeling approach was used to simulate the ISLOCA appropriately. Auxiliary buildings including the drain system and the ventilation system to influence the behavior of the fission products were modeled in detail. In addition, the mechanisms such as the aerosol deposition and the pool scrubbing during the ISLOCA piping were considered. The SOARCA project was performed for two pilot plants: Peach Bottom and Surry. Peach Bottom is a Boiling Water Reactor (BWR) and Surry is a Pressurized Water Reactor (PWR). Among the pilot plants, the ISLOCA scenarios were analyzed for the Surry NPP only. Although Surry is the Westinghouse PWR, differences between the Surry NPP and the NPPs in Korea exist in plant-specific design such as the structure of the auxiliary building. Because the results of the ISLOCA analysis of the SOARCA project are not generally applicable to the NPPs in Korea, the ISLOCA analyses were performed on the Advanced Power Reactor 1400 MWe (APR1400) in this study. Analysis methodology and modeling approach were established based on best practices of the SOARCA project in this study. 2. Modeling