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냉각기능상실사고 시 사용후핵연료 저장조의 열수력적 거동 특성
전병국(Byong Guk Jeon),김기환(Ki Hwan Kim),김병재(Byung Jae Kim),김종록(Jong Rok Kim),박종국(Jong Kuk Park),문상기(Sang Ki Moon) 대한기계학회 2016 대한기계학회 춘추학술대회 Vol.2016 No.12
After the Fukushima nuclear power plant accident, several efforts have been made to ensure integrity of spent fuel pools as well as nuclear reactors under prolonged station blackout. Regardless of low individual decay heat level, release of radioactive materials from spent fuel pools can be tragic because of a large number of fuel assemblies. In this research, thermal hydraulic behavior and safety of fuel rods inside a subchannel of nuclear fuel under loss of cooling accidents were evaluated through a 5 by 5 heater rods - experiment. We chose three cases representing decay heat at 1 day, 1 week, and 1 month after shutdown, corresponding to 0.38 kW/rod, 0.21 kW/rod, and 0.11 kW/rod, respectively. At higher decay heat level, mixture water level was much escalated because of large void generation while the maximum rod temperature was increased fast. The vessel module, COBRA-TF, in the MARS code was assessed by comparing with the experiment result. The water mixture level as well as the wall temperature profile was not well represented by the code. The models are needed to be further improved.
MARS 코드를 활용한 냉각기능상실사고 시 사용후핵연료 저장조의 열수력적 거동 예비 해석
전병국(Byong Guk Jeon),김병재(Byung Jae Kim),김기환(Ki Hwan Kim),김종록(Jong Rok Kim),문상기(Sang Ki Moon) 대한기계학회 2015 대한기계학회 춘추학술대회 Vol.2015 No.11
In a loss of cooling accident caused by a loss of off-site power, similar to the Fukushima accident, water inside a spent fuel storage pool heats up and evaporates by decay heat. That leads to uncovery and failure of spent fuels and subsequent release of radioactive materials. To avoid such severe consequence, it is important to identify the thermalhydraulic behavior of spent fuels under the accident. We are planning to conduct experiments with the ATHER facility. Prior to the experiments, for a loss of cooling accident, peak temperature of fuel claddings as well as water level is calculated using the best estimate code, MARS, under different decay powers and discussions have been made.
FINCLS 자연 순환 실험에 대한 MARS-KS 코드 및 상관식 분석
박지환(Ji-Hwan Park),전병국(Byung-Guk Jeon),윤은구(Eunkoo Yun),박현식(Hyun-Sik Park) 대한기계학회 2018 대한기계학회 춘추학술대회 Vol.2018 No.12
FINCLS is a simplified test loop to understand thermal-hydraulic phenomena which occurs during natural circulation tests. Experiment was conducted at 1 MPa using the FINCLS facility. The single-phase natural circulation flow rate increases when the heater power increases. The calculation results from the MARS-KS code and existing correlation were compared with experimental values of temperature distribution and natural circulation flow rate. MARS-KS considered pressure drop model such as sudden contraction and sudden expansion and heat transport with air. Correlation and MARS-KS can predict single-phase natural circulation (SPNC) flow rate applying momentum and energy conservation. The MARS-KS calculation result was compared with the experiment when the pressure drop model was applied and when the heat loss model was applied to the atmosphere. The SPNC flow rate was compared with the modified height by calculating an average temperature.