A Preliminary Safety Analysis for the Prototype Gen IV Sodium-Cooled Fast Reactor

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• 作者：Kwi Lim Lee ; ^{kllee@kaeri.re.kr" class="auth_mail" title="E-mail the corresponding author} ; Kwi-Seok Ha ; Jae-Ho Jeong ; Chi-Woong Choi ; Taekyeong Jeong ; Sang June Ahn ; Seung Won Lee ; Won-Pyo Chang ; Seok Hun Kang ; Jaewoon Yoo
• 关键词：Anticipated Transient Without Scram ; Design Basis Accidents ; Design Extended Conditions ; MARS-LMR ; MATRA-LMR-FB ; Prototype Gen-IV Sodium-Cooled Fast Reactor ; Source Term
• 刊名：Nuclear Engineering and Technology
• 出版年：2016
• 出版时间：October 2016
• 年：2016
• 卷：48
• 期：5
• 页码：1071-1082
• 全文大小：2614 K

文摘

Korea Atomic Energy Research Institute has been developing a pool-type sodium-cooled fast reactor of the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR). To assess the effectiveness of the inherent safety features of the PGSFR, the system transients during design basis accidents and design extended conditions are analyzed with MARS-LMR and the subchannel blockage events are analyzed with MATRA-LMR-FB. In addition, the in-vessel source term is calculated based on the super-safe, small, and simple reactor methodology. The results show that the PGSFR meets safety acceptance criteria with a sufficient margin during the events and keeps accidents from deteriorating into more severe accidents.