Nuclear Engineering and Design 260 (2013) 54–63 Contents lists available at SciVerse ScienceDirect Nuclear Engineering and Design j ourna l h om epa ge: www.elsevier.com/locate/nucengdes An experimental study on the validation of cooling capability for the Passive Auxiliary Feedwater System (PAFS) condensation heat exchanger Seok Kim a , Byoung-Uhn Bae a , Yun-Je Cho a , Yu-Sun Park a , Kyoung-Ho Kang a , Byong-Jo Yun b,∗ a Thermal Hydraulics Safety Research Division, Korea Atomic Energy Research Institute, 1045 Daedeok-daero, Yuseong-gu, Daejeon, 305-353, Republic of Korea b School of Mechanical Engineering, Pusan National University, 30 Jangjeon-dong, Geumjeong-gu, Busan, 609-735, Republic of Korea h i g h l i g h t s • PAFS is designed to replace a conventional active Auxiliary Feedwater System. • A SET facility is constructed for investigating the thermal-hydraulic behavior of the PAFS system. • Experimental results proved that the PCHX design satisfied the heat removal requirements. • Results of the MARS-KS code provided a conservative prediction of the heat transfer phenomena. a r t i c l e i n f o Article history: Received 4 May 2012 Received in revised form 25 February 2013 Accepted 13 March 2013 a b s t r a c t The Passive Auxiliary Feedwater System (PAFS) is one of the advanced safety features adopted in the Advanced Power Reactor Plus (APR+). PAFS is designed to replace a conventional active Auxiliary Feed- water System (AFWS). The PAFS cools down the steam generator secondary side and eventually removes the decay heat from the reactor core by a natural circulation mechanism, i.e., condensing steam in nearly horizontal U-tubes submerged inside a pool. A separate effect test facility was constructed with the aim of validating the cooling and operational performance of the PAFS. The PAFS Condensing Heat Removal Assessment Loop (PASCAL) was constructed by simulating a single Passive Condensation Heat Exchanger (PCHX) tube submerged in the Passive Condensation Cooling Tank (PCCT) according to the volumetric scaling methodology. Quasi-steady state (SS) test cases and PCCT level decrease (PL) were sequentially performed with the steam generator heater power set at 540 kW to investigate the thermal-hydraulic behavior of the PAFS system and the characteristics of the natural circulation in the loop. The experi- mental results proved that the current PCHX design satisfied the heat removal requirement for cooling down the reactor core during an accident condition. Therefore, the PAFS can replace a conventional active AFWS in the APR+ by utilizing the two-phase natural circulation flow. The Multi-dimensional Analysis of Reactor Safety, KINS Standard Version (MARS-KS), a thermal hydraulic system analysis code, was uti- lized to validate the present experimental data. The results of the MARS-KS code provided a conservative prediction of the heat transfer phenomena for the PCHX cooling performance. © 2013 Elsevier B.V. All rights reserved. 1. Introduction The Advanced Power Reactor Plus (APR+) is a Generation III+ nuclear power plant being developed in Korea. The Passive Auxil- iary Feedwater System (PAFS) is one of the advanced safety features being adopted in the APR+. PAFS is intended to completely replace the conventional active auxiliary feedwater system (Song et al., 2010; Cheon et al., 2010). PAFS can improve the reliability of the safety system and reduce operator error, which are the fundamen- tal weak points outlined in the Probability Safety Assessment (PSA). ∗ Corresponding author. Tel.: +82 51 510 2484. E-mail address: bjyun@pusan.ac.kr (B.-J. Yun). The PAFS cools down the secondary side of the steam generator (SG) and eventually removes the decay heat from the reactor core by adopting a natural circulation mechanism, i.e., condensing steam in the nearly-horizontal Passive Condensation Heat Exchanger (PCHX) tubes submerged inside the PCCT. When the water level in the steam generator becomes lower than 25% of the wide range of the water level transmitter during an accident situation, the actua- tion valve at the return-water line opens and then the PAFS begins its natural circulation flow. To satisfy a single failure criterion, the PAFS is composed of two independent trains, but a single PAFS train is capable of removing the whole decay heat from the reactor core during the anticipated accident transients. Fig. 1 shows a schematic diagram of the PAFS for the APR+. The diagram comprises a steam- supply line, a PCHX, a return-water line, and a passive condensation 0029-5493/$ – see front matter © 2013 Elsevier B.V. All rights reserved. http://dx.doi.org/10.1016/j.nucengdes.2013.03.016