atw - International Journal for Nuclear Power | 01.2020

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atw Vol. 65 (2020) | Issue 1 ı January RESEARCH AND INNOVATION 28 | Fig. 8. Pressures of the primary and secondary system (Case 3). | Fig. 11. Peak cladding temperatures. pro viding the borated water to the RCS. Figure 7 shows the mass flow rates of SIP and SIT in Case 2. Case 3 is the additional scenario in this case study as described in previous section. Figure 8 shows the pressures of the primary and secondary systems in Case 3. The pressure of the primary system increases to the set point of the PSV and then oscillates. Thermal hydraulic behavior of Case 3 is similar to that of Case 2 until SDS valve opens. In Figure 9, one train of SDS opens 30 minutes after the PSV is first opened. The pressures of the primary and secondary systems decrease as the primary inventory is discharged through the SDS. The SIP is operated by the low pressurizer pressure signal. The primary pressure continues to decrease and reaches the injection pressure of SITs. Figure 10 shows the mass flow rates of the SIP and SIT. Peak cladding temperatures of all cases are shown in Figure 11. Case 1 is 325 °C , while both Case 2 and Case 3 are 354 °C . Injection of the SIP cools the core down immediately. The fuel cladding temperatures of all cases don’t exceed the maximum allowable fuel cladding temperature, 1,204 °C (2,200 °F ). Which means the core cooling capabilities are sufficient in all cases. | Fig. 9. Mass flow rates of PSV and SDS (Case 3). 4 Conclusions In this study, we present thermalhydraulic analysis for the TLOFW event in OPR1000 using the SPACE 3.0 code. Three different cases according to operations of the SDS and safety injection system were analyzed to examine the effectiveness of the RCS cool down strategy through the feed and bleed operations to mitigate the TLOFW event. The feed and bleed operations start with manually opening of the SDS valve after the PSV opening in accordance with the EOP. The simulation scenarios of Case 1 and Case 2 were based upon design requirements of the SDS described in the FSAR of Hanul units 3&4. Case 3 was the additional scenario in this comparative study. The peak cladding temperatures of all cases did not exceed 1,204 °C (2,200 °F) which is the maximum allowable fuel cladding temperature. The RCS cool down strategy through the feed and bleed operations can guarantee the core cooling capabilities during the TLOFW event. The earlier feed and bleed operation was more effective strategy for removing the decay heat. We also confirmed that the SPACE code is very useful code for analyzing the multiple failure accidents in the PWR. Acknowledgments This work was supported by the Korea Institute of Energy Technology Evaluation and Planning (KETEP) grant funded by the Korea government (MOTIE) (No. 20161510101840, Development of Design Extension Conditions Analysis and Management Technology for Prevention of Severe Accident). References 1. Korea Hydro and Nuclear Power Co. Ltd., “Development of Design Extension Conditions Analysis and Management Technology for Prevention of Severe Accident Report”, September, 2017. | Fig. 10. Mass flow rates of SIP and SIT (Case 3). 2. Kwon, Y.M. et al., “Comparative simulation of feed and bleed operation during the total loss of feedwater event by RELAP5:MOD3 and CEFLASH-4AS:REM computer codes, Nuclear Technology, Vol. 112, pp. 181– 193, 1995. 3. Kwon, Y.M., Song, J.H., “Feasibility of long term feed and bleed operation for total loss of feedwater event”, Journal of Korean Nuclear Society, Vol. 28 (3), pp. 257–264, 1996. 4. Park, R.J. et al., “Detailed evaluation of coolant injection into the reactor vessel with RCS depressurization for high pressure sequences”, Nuclear Engineering and Design, Vol. 239, pp. 2484–2490, 2009. 5. Pochard, R. et al., “Analysis of a feed and bleed procedure sensitivity study performed with the SIPACT simulator on a French 900 MWe NPP”, Nuclear Engineering, Des. 215, pp. 1–14, 2002. 6. Reventós, F. et al., “Analysis of the feed & bleed procedure for the Ascó NPP first approach study for operation support”, Nuclear Engineering, Des. 237, pp. 2006–2013, 2007. 7. S. J. Ha et al., “Development of the SPACE Code for Nuclear Power Plants,” Nuclear Engineering & Technology, Vol. 43, No. 1, pp. 45, 2011. 8. Final Safety Analysis Report Hanul 3,4, KHNP. Authors MinJeong Kim Minhee Kim Junkyu Song Bongsik Chu Central Research Institute Korea Hydro and Nuclear Power Co., Ltd. Deajeon, 34101 Rep. of Korea Jae-Seung Suh Hyunjin Lee System Engineering and Technology Co., Ltd. Daejeon, 34324 Rep. of Korea Research and Innovation Thermal-Hydraulic Analysis for Total Loss of Feedwater Event in PWR using SPACE Code ı MinJeong Kim, Minhee Kim, Junkyu Song, Bongsik Chu, Jae-Seung Suh and Hyunjin Lee
atw Vol. 65 (2020) | Issue 1 ı January CFD Simulation of Flow Characteristics and Thermal Performance in Circular Plate and Shell Oil Coolers Shen Ya-jie, Gao Yong-heng and Zhan Yong-jie Circular plate and shell heat exchangers were gradually applied as oil coolers. Hence, it was necessary to investigate their performance at low Reynolds number with high viscous oil. This paper provided a CFD simulation of flow characteristic and thermal performance in circular plate and shell oil cooler with different plate parameters, such as plate angle β, ratio of plate pitch to height p/h and corrugation styles. The fiction factor f and Colburn factor j were investigated for the various plate parameters. The numerical results showed that f increased with increasing β, and both increased as p/h decreased. When β
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