Analysis of severe core damage accident progression for the heavy water reactor

Analysis of severe core damage accident progression for the heavy water reactor 1 corresp first-author Corresponding author. E-mail address: lltong@sjtu.edu.cn Corresponding author. E-mail address: lltong@sjtu.edu.cn lltong@sjtu.edu.cn 1 1 1 School of Mechanical Engineering, Shanghai Jiao Tong University, Shanghai 200240, China School of Mechanical Engineering, Shanghai Jiao Tong University, Shanghai 200240, China In this study, the severe accident progression analysis of generic Canadian deuterium uranium reactor 6 was preliminarily provided using an integrated severe accident analysis code. The selected accident sequences were multiple steam generator tube rupture and large break loss-of-coolant accidents because these led to severe core damage with an assumed unavailability for several critical safety systems. The progressions of severe accident included a set of failed safety systems normally operated at full power, and initiative events led to primary heat transport system inventory blow-down or boil off. The core heat-up and melting, steam generator response, fuel channel and calandria vessel failure were analyzed. The results showed that the progression of a severe core damage accident induced by steam generator tube rupture or large break loss-of-coolant accidents in a CANDU reactor was slow due to heat sinks in the calandria vessel and vault. Severe Accident steam generator tube rupture large break loss-of-coolant accidents References Meneley D K , Blahnik C , Rogers J T , et al . International seminar on mass and heat transfer in severe reactor accidents , Cesme, Turkey , May 22-26, 1995 . Petoukhov S M , Mathew P M . Technical meeting on severe accident and accident management , Toranomon Pastoral, Minatoku, Tokyo, Japan , Mar.14-16, 2006 . IAEA , IAEA-TECDOC-1594 , 2008 . IAEA-TECDOC-1594 Gheorghe N , Adrian M . Nucl Eng Design 1997 , 170 : 67 - 72 . Nucl Eng Design Kim D H , IAEA/AECL/CNSC Technical meeting on severe accident analyses, accident management and PSA applications for PHWRs , Mississauga, Ontario, Canada , Nov.7-10, 2005 . Mathew P M , White A J , Snell V G , et al . Transactions of the 17th international conference on structural mechanics in reactor technology (SMiRT 17) Prague , Czech Republic , Aug.17-22, 2003 . Mills P J , Sanderson D B , Haugen K A , et al . Proc 17th Ann conference of Canadian nuclear society , Fredericton, Canada , Jun. 9-12, 1996 . IAEA Safety Guide, No.NS-G-2.15 , Vienna , 2009 . IAEA Safety Guide, No.NS-G-2.15 10.13538/j.1001-8042/nst.21.251-256 2010-04-10 2010-08-20 2021-01-19 2010 Nuclear Science and Techniques 04 21