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Title of article :
Analytical studies of the heat removal capability of a passive auxiliary feedwater system (PAFS)
Author/Authors :
Cho، نويسنده , , Yun-Je and Bae، نويسنده , , Sung-Won and Bae، نويسنده , , Byoung-Uhn and Kim، نويسنده , , Seok and Kang، نويسنده , , Kyoung-Ho and Yun، نويسنده , , Byong Jo Yun، نويسنده ,
Issue Information :
روزنامه با شماره پیاپی سال 2012
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Abstract :
As passive safety features for nuclear power plants receive increasing attention, various studies have been conducted to develop safety systems for third-generation (GEN-III) nuclear power plants that are driven by passive systems, such as natural circulation, gravity, and resistance to high temperatures. Thus, South Korea has designed the Advanced Power Reactor Plus (APR+) with a two-loop PWR and 1500 MWe by adding passive safety features to the Advanced Power Reactor 1400 MWe (APR1400). The Passive Auxiliary Feedwater System (PAFS) is one of several passive safety systems being designed for the APR+, and extensive studies are being conducted to complete its design and to verify its feasibility. Because the PAFS removes decay heat from the reactor core under transient and accident conditions, it is necessary to evaluate the heat removal capability of the PAFS under hypothetical accident conditions. Therefore, in this paper, after introducing the characteristics of the PAFS and its design requirements, a performance analysis of the PAFS is performed for two accident cases: Loss of Condenser Vacuum (LOCV) and Feedwater Line Break (FLB). For the analysis, the Multi-dimensional Analysis of Reactor Safety (MARS) code is used, and a MARS model is developed by adding the PAFS model to the existing APR1400 model. The analysis results show that the PAFS has enough capacity to remove decay heat under the postulated accident conditions. In addition, the adequacy of modified control logic for main steam isolation valve (MSIV) is validated by comparing the traditional control logic.
Journal title :
Nuclear Engineering and Design Eslah
Journal title :
Nuclear Engineering and Design Eslah
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