Transactions of the Korean Nuclear Society Virtual spring Meeting May 13-14, 2021 Test result on a small break loss-of-coolant accident simulation for the control rod driving mechanism nozzle rupture with a failure of safety injection pump using the ATLAS facility

The safety issues related to the structural integrity of the upper head penetration nozzle of the nuclear reactor were raised in 2002 and 2012 as problems with the RPV wall thinning around the control rod drive mechanism (CRDM) penetration nozzle in the Davis Besse nuclear power plant in the United States and the micro-cracking on the control element drive mechanism of Younggwang Unit 3 [1, 2]. Recently, the safety issue on an upper head penetration nozzle of a reactor pressure vessel (RPV) was raised again as 35 of the 84 welds in which stress corrosion cracks were found during the preventive maintenance of Hanbit Unit 5 were disapproved [3]. Circumferential cracking of the penetration nozzle can lead to a small break loss-ofcoolant accident (SBLOCA) at the RPV upper head. In general, when an SBLOCA occurs at the upper head of a RPV, the characteristic of the thermalhydraulic phenomenon is that the break flow is mainly discharged in a vapor phase compared to the other located SBLOCAs such as a break at hot legs, cold legs, and direct vessel injection lines. Therefore, it was reported that the coolant inventory in the primary system was relatively well preserved under a safety injection pump (SIP) operation, resulting in a later core heat-up [4]. However, as a multiple failure accident scenario, an SBLOCA at the RPV upper head with a failure of SIP can damage a reactor core if a proper accident management (AM) action is not implemented. In this study, a multiple failure accident of an SBLOCA due to a break of two CRDM nozzles with a failure of SIP was simulated by using a thermalhydraulic integral effect test facility, ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) [5]. Regarding an operator’s AM action, 50 % opening of an atmospheric dump valve (ADV) of steam generator was taken according to the pre-specified maximum heater rod surface temperature in the core. In addition, the auxiliary feedwater (AFW) was supplied when the secondary side collapsed water level reached a wide-range of 25 % in both steam generators. The main objective of this study is to investigate the thermal-hydraulic phenomena during an SBLOCA at RPV upper head with a failure of SIP focused especially on the break flow behavior related to the water level in the RPV upper head and the loop seal clearing (LSC) behavior associated with the pressure build-up through the RPV bypass line. Considering the confidential problem, all the data in this paper are presented with the normalized value using random numbers.