Investigation of main coolant pump trip problem in case of SB LOCA for Kozloduy Nuclear Power Plant, WWER-440/V230. (February 2015)

• Record Type: Journal Article
• Authors: Groudev, Pavlin; Andreeva, Marina; Pavlova, Malinka
• Is Part Of: Annals of nuclear energy. Volume 76(2015:Feb.)
• Page Start: 137
• Page End: 145
• Publication Date: 2015-02
• Subjects: Nuclear energy -- Safety analysis -- RELAP5/MOD3.2 -- WWER type reactor -- Symptom Based Emergency Operating Procedures; Nuclear energy -- Periodicals; Nuclear engineering -- Periodicals; 621.4805
• DOI: 10.1016/j.anucene.2014.09.050
• Languages: English
• ISSNs: 0306-4549

Abstract

Highlights: In this study we investigated scenarios with trip of MCP in case of SB LOCA. The reference power plant for the analyses is Unit 4 at Kozloduy NPP. The RELAP/MOD 3.2 computer code is used in performing the analyses. The results are done in support of development of SB EOPs. Abstract: This paper presents the results of thermal-hydraulic calculation of accident scenarios that involve the trip of main coolant pump (MCP) in case of Small break loss of coolant accident (SB LOCA) for WWER-440/V230 units at Kozloduy Nuclear Power Plant (KNPP), done in support of the development of Symptom Based Emergency Operating Procedures (SB EOPs) for this plant. The main purpose of these analyses is to show how the different time of MCP switching off results in primary inventory depletion in case of SB LOCA and it is reflect on peak cladding temperature. According to this, the SB LOCA scenario is regarded from the point of view of an inadequate core cooling. Therefore, the primary concern is Critical Safety Function (CSF) "Core cooling" and "Primary inventory". High core residual heat, minimal safety injection flow and other initial conditions challenging the mentioned CSFs are the main particularities of the accepted scenarios. The RELAP5/MOD3.2 computer code has been used to perform the analyses in a WWER-440 Nuclear Power Plant (NPP) model. A model of WWER-440 based on Unit 4 of Kozloduy NPP has been developed for the system's thermal-hydraulics code RELAP5/MOD3.2 at the Institute for Nuclear Research and Nuclear Energy – Bulgarian Academy of Sciences (INRNE-BAS), Sofia.