[en] Natural circulation is a heat removal process whereby Reactor Coolant System (RCS) flow is driven by temperature and density differences in the RCS fluid between the core and steam generators. The combination of core heat addition and steam generator heat removal would cause continuous flow to develop through the RCS and should provide enough heat removal to adequately cool the core. Natural Circulation Cooldown (NCC) capability was evaluated to ensure the safe shutdown function of the power up rated Kori 3 and 4 and Yonggwang 1 and 2 Nuclear Power Plants (NPPs). The evaluation was performed for the duration for normal operation to the conditions under which the initiation of the Residual Heat Removal (RHR) is permitted in accordance with the initiation of the Residual Heat Removal (RHR) is permitted in accordance with the US NRC Branch Technical Position (BTP) Reactor Systems Branch (RSB) 5-1. BTP RSB 5-1 requires the use of only safety-grade equipment and the assumptions of the concurrent loss of offsite power with a single failure. Steam Generator (SG) Power-Operated Relief Valve (PORV), Pressurizer PORV and Auxiliary Feedwater Pump were used as safety-grade means of the RCS cooldown and depressurization for NCC analysis of Kori 3 and 4 and Yonggwang 1 and 2 NPPs. The evaluation of NCC capability was done using a computer code, CENTS. The analysis result showed that the amount of safety-grade auxiliary feedwater required to cool the RCS down to RHR entry concluded that the power up rated Kori 3 and 4 and Yonggwang 1 and 2 NPPs can be cooled and de pressurized to RHR entry conditions by the natural circulation in conformance with BTP RSB 5-1 requirements

[en] Many regulatory requirements and recommendations following the Fukushima accident have been issued to cope with the extended station blackout (SBO) by the NRC, INPO, IAEA, ENSREG, WENRA, etc., and the nuclear safety improvement design features of each country have been enhanced to incorporate the lessons learned from the Fukushima accident. There have been many evaluations to cope with the extended loss of alternating current (AC) power (ELAP) event after the Fukushima accident. PWROG has developed the FLEX support guideline (FSG) that provides the guidance to mitigate the consequences of ELAP event based on the FLEX. The FSG is interfaced with emergency operating guidelines (EOGs) and severe accident management guidelines (SAMGs). However, the FSG developed by PWROG is not compatible with EOGs for both OPR1000 and APR1400 NPPs. Therefore, it is necessary to develop an extended station blackout recovery guideline (ESRG) to cope with an extended SBO event utilizing the newly adopted safety improvement design features against Fukushima accident for OPR1000 and APR1400 NPPs. The ESRG is also performed to satisfy all safety functions and to prevent from entering SAMGs during an extended SBO event. Therefore, this ESRG is entirely appropriate to cope with an extended SBO event by utilizing the newly adopted safety improvement design features following Fukushima accident for OPR1000 and APR1400 NPPs. This guideline will be considered in the establishment of accident management planning in near future.

[en] There have been many evaluations and recommendations for the extended Station Black Out (SBO) condition of the nuclear power plant. For example, the 'SECY-11-0093/0137', is a recommendation of NRC and the 'WCAP-17601-P' is an evaluation of the PWROG. The extended loss of AC power (ELAP) can be defined as same with the extended (or prolonged) SBO which has a Loss of Offsite Power (LOOP) condition and loss of all Emergency Diesel Generators (EDG), Alternative Alternating Current (AAC), but Direct Current (DC) source is available. This evaluation provides NSSS responses to an ELAP for the OPR1000 unit. And the results presented provide certain phenomena which occur during the ELAP, the maximum coping time until a core uncovery condition. It is assumed for this case that sufficient SG secondary makeup inventory exists or can be attained, so that the duration of the ELAP prior to core damage is dependent solely upon the loss of inventory from the RCS. Even with a limited RCS cooldown and depressurization, and conservatively high assumed RCP seal leakage, the plant can be sustained for over 65 hours prior to core uncovery

[en] The concept of Common-Cause Failure (CCF) first appeared in the aerospace industry several decades ago, and nuclear power industry actively adopted the concept to the nuclear power plant (NPP) system analysis after the TMI accident. Since digital Instrumentation and Control (I and C) systems were applied to the NPP design, the CCF issues once again drew attention from the nuclear power industry in 90's. Identification of CCF has not been considered as a challenging issue because of its simplicity. However, as the systems become more complex and interconnected, demands are increasing to analyze CCF in more detail, for example, CCF with multiple initiating events or supporting situation awareness of the operation crew. The newly suggested CCF propagation paths identification method, CCF-SIREn, is expected to resolve path identification issue more practically and efficiently. CCF-SIREn uses general diagrams so that the compatibility and usability can be hugely increased. It also offers up-to-date CCF information with a least analysis effort whenever the ordinary NPP design change processes are made. A back-propagation technique is still under development to find out root-causes from the suspiciously responding signals, alarms and components. The probabilistic approach is also under consideration to prioritize defined CCF.

[en] The Load Rejection to House Load test at 50% power was successfully performed during the UCN 3 PAT period. In this test, all plant control systems automatically controlled the plant from 50% power to house load operation mode. The KISPAC code, which was used in the performance analysis during the design process of UCN 3 and 4, predictions of the test agreed with the measured data demonstrating the validity of the code as well as completeness of the plant design