Analyzing the alternative shutdown cooling behaviors for Chinshan Nuclear Power Plant using CFD simulation Yung-Shin Tseng a,⇑ , Chih-Hung Lin a , Yng-Ruey Yuann a , Jong-Rong Wang a , F. Peter Tsai b a Institute of Nuclear Energy Research, Taiwan, ROC b Cool-Tec Co., USA article info Article history: Received 27 September 2010 Received in revised form 15 July 2011 Accepted 15 July 2011 Available online 1 September 2011 Keywords: CSNPP BWR4 RHR Alternative shutdown cooling IVVI CFD abstract The Chinshan Nuclear Power Plant (CSNPP) is a GE-designed BWR4 plant, having two identical units with rated core thermal power of 1804 MWt each unit. Several alternative shutdown cooling methods driven by natural or mixed convection has been proposed by the plant for studying the core cooling capability when the Residual Heat Removal (RHR) systems are not available or the refueling tasks, such as the In Vessel Visual Inspection (IVVI) work etc., is being proceeded. One of the examples is to connect a pipe from the outlet of the new spent fuel heat exchanger to the reactor cavity. The design of the alternatives shall ensure that the maximum core fluid temperature is limited below the boiling temperature of water. In this study, a Computational Fluid Dynamics (CFD) model is developed to analyze the natural convec- tion phenomena during the shutdown period. Through a series of assumption, modeling and meshing processes, a calculation domain with approximate four million meshes including the RPV, reactor cavity and spent fuel pool, have been solved in this study. The analysis results showed that the passive alterna- tive shutdown cooling system could provide sufficient heat removal capability to maintain the maximum core fluid temperature below boiling temperature. The results also indicated that the alternative shut- down cooling system could be served as an appropriate solution for CSNPP when the RHR is inoperable. Ó 2011 Elsevier Ltd. All rights reserved. 1. Introduction The Chinshan Nuclear Power Plant (CSNPP) is a GE-designed BWR4 plant, having two identical units with rated core thermal power of 1804 MWt each unit. The two units of CSNPP have been commercially operated from 1978 and 1979, respectively. In the 30 years, the CSNPP has produced thousands of spend nuclear fuels (SNFs). Those SNFs are placed in the spend fuel pool (SFP) to remove the decay heat generated by the fission products and heavy elements in the SNF. However, the increased SNFs occupy a lot of space in the SFP, and result in the fuels existing in core zone cannot be temporarily placed in the SFP during refueling stage. For this reason, the CSNPP owner, Taipower Co. (TPC), attempts to program an in-core-shuffle process to replace the discharge and new fuels during refueling stage for the last cycle. During refueling stage, the in-core-shuffle process implies that the core continuously produces decay heat after reactor shutdown, due to a part of fuels existing in the reactor core. The decay heat in the core has to be removed by the Residual Heat Removal (RHR) system to maintain the temperature of cooling water lower than the operative requirement, such as the working items of fuel discharging and In Vessel Visual Inspection (IVVI) after reactor shuts down in 48 and 120 h, respectively. However, the RHR systems have to be ceased during refueling stage, when jointed pipes of RHR are for maintenance. Hence, three alternative shutdown cooling cases driven by natural or mixed convection have been proposed by the plant for the estimate of the core cooling capability, while the RHR system is in unavailable condition. The source of cooling water for the three alternative cooling cases comes from new SFP cooling systems (NSFPCSs), in which the capability for each NSFPCS is 5.56 MWt at 56 °C (Taipower Co., 2006) and the maximum capability for two parallel connection of NSFPCSs with 90% flow rate is about 10 MWt (90%  2  5.56 MWt = 10.008 MWt). In the three cases, the cool- ing water will be injected into reactor from (1) the cavity diffuser in the cavity, (2) the core spray in the reactor pressure vessel (RPV), and (3) the jet pumps from recirculation line, respectively. The de- sign of the alternatives is able to ensure that the maximum water temperature in the core and the bulk temperatures of cavity and upper tied plane are limited below the boiling temperature of water, 56 °C and 45 °C, respectively. As the natural convection or mixed flow situation (natural or forced convection) occurs, the original methodology based on a system code (e.g., RELAP, or TRACE) cannot be suitable for analyz- ing local temperature distribution in the reactor due to coarse 0306-4549/$ - see front matter Ó 2011 Elsevier Ltd. All rights reserved. doi:10.1016/j.anucene.2011.07.017 ⇑ Corresponding author. Address: Institute of Nuclear Energy Research, No. 1000, Wenhua Rd., Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan, ROC. Tel.: +886 3 4711400x6073; fax: +886 3 4711064. E-mail address: ystseng@iner.gov.tw (Y.-S. Tseng). Annals of Nuclear Energy 38 (2011) 2557–2568 Contents lists available at SciVerse ScienceDirect Annals of Nuclear Energy journal homepage: www.elsevier.com/locate/anucene