• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.633-638
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• 1997

중소형 일체형 원자로의 자연대류 실험 및 해석을 통해 일체형 원자로의 자연대류에 의한 잔열제거 기능의 특성 및 피동 안전성을 파악하였다. 이를 위해 일체형 원자로 축소 실험장치를 이용한 자연대류 실증 실험을 수행하였으며, 실험 결과를 RETRAN-03와 COMMIX-1B 코드 해석 결과와 비교, 검증하였다. 실험 결과 일부 증기발생기의 열제거 기능 상실이 발생한 경우에도 노심으로 유입되는 냉각재의 온도가 균일하게 분포하여 피동 잔열 제거가 원만히 일어남을 알 수 있었고 해석 결과와 일치하였다.

• Nuclear Engineering and Technology
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• v.26 no.4
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• pp.516-525
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• 1994

Since the TMI and Chernobyl accidents, many passive safety features are suggested in advanced reactors in order to enhance the safety in future nuclear power plants. In order to verify the effectiveness and provide the data for detailed design of passive cooling system, in the present work, the effects of air inlet position and external condition on the natural circulated air flow rate and the natural and forced convective heat transfer coefficient have been investigated for the one-side heated closed path such as the passive containment cooling system of the Westinghouse's AP-600. A series of experiments have been peformed with the 1/26th scaled segment type test facility of the AP-600 passive containment. Under natural and forced convection, the air velocities and temperatures are measured at several points of the air flow path. The experimental result are compared with a simple one-dimensional model and it shows a good agreement.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.604-609
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• 1996

원자력 발전소 부분 충수운전 중에 발생할 수 있는 정지냉각기능 상실사고인 과배수 사건에 대한 확률론적 안전성 평가를 수행하였다. 본 분석의 주된 목적은 과배수로 인한 정지냉각기능 상실사건에 대하여 노심손상 빈도를 계산하고 안전성 향상방안을 도출하는데 있다. 과배수 사건은 초기 부분 충수운전중 발생하는 것으로 가정하였으며 이 때의 발전소 배열(Plant Configuration)은 영광 3,4호기의 운전절차서 및 발전소 운전경험을 근거로 결정하여. 현재 운전상태에 대한 확률론적 안전성평가를 수행하였다. 분석결과 인간오류가 노심손상빈도에 가장크게 기여하는 인자로 나타났으며 인간오류를 줄 일수 있는 대체냉각 절차를 선정하여 재분석을 수행하였다. 고려된 대체냉각 수단은 피동적인 잔열제거 방법인 열규응축냉각(Reflux Cooling)과 정지냉각펌프의 대체계통으로 격납용기 살수펌프를 사용하는 경우의 두가지이다. 본 분석에서는 두가지 대체냉각수단을 모두 채택하는 것으로 가정하여 대체냉각 사용에 따른 효과를 비교하였는데 노심손상 빈도가 1/1000로 감소 하였다. 따라서 절차서 개정에 의한 대체 냉각수단확보는 부분 충수운전중 발전소 안전성 향상에 매우 효과가 큰 것으로 나타났다.

• Transactions of the KSME C: Technology and Education
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• v.3 no.4
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• pp.257-264
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• 2015

To maintain sustainability of nuclear energy as an important energy source, both safety problem and Spent Nuclear Fuels(SNFs) problem should be solved. In case of Gen-IV reactors such as fast reactor, SNFs can be used as fuels by using fast neutrons. It can be a suitable treatment method of high-level waste in near future. Liquid metals such as Sodium or Lead-Bismuth Eutectic (LBE) can be possibly used as a coolant to use fast neutrons. In this paper, it was described that natural circulation parameter studies, design analyses, material selections and a completion of facilities. To develop a natural circulation facility, thermal hydraulic analyses were performed. Installation technique of liquid metal natural circulation were secured.

• Nuclear Engineering and Technology
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• v.25 no.4
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• pp.515-528
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• 1993

In a Passive PWR, the successful actuation of Automatic Depressurization System (ADS) is essentially required so that no core damage is occurred following small LOCA. But it has been shown in the previous studies that Core Damage Frequency (CDF) from small LOCA is significantly caused by unavailability of ADS. In this study, the design vulnerabilities impacting the ADS unavailability have been identified and the design improvement items have been proposed through the system reliability assessment using the fault tree methodology The impacts on CDF according to the change of system unavailability have also been analyzed. In addition, small LOCA simulation using RELAP5/MOD3 code has been performed to show the thermal-hydraulic feasibility of the suggested design enhancements.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.40 no.3
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• pp.165-172
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• 2016

An experimental study of the thermal-hydraulic characteristics of passive safety systems (PSSs) was conducted using a system-integrated modular advanced reactor-integral test loop (SMART-ITL). The present passive safety injection system for the SMART-ITL consists of one train with the core makeup tank (CMT), the safety injection tank, and the automatic depressurization system. The objective of this study is to investigate the injection effect of the PSS on the small-break loss-of-coolant accident (SBLOCA) scenario for a 0.4 inch line break in the safety-injection system (SIS). The steady-state condition was maintained for 746 seconds before the break. When the major parameters of the target value and test results were compared, most of the thermal-hydraulic parameters agreed closely with each other. The water level of the reactor pressure vessel (RPV) was maintained higher than that of the fuel assembly plate during the transient, for the present CMT and safety injection tank (SIT) flow rate conditions. It can be seen that the capability of an emergency core cooling system is sufficient during the transient with SMART passive SISs.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 2002.11a
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• pp.111-116
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• 2002

고온의 소듐을 냉각재로 사용하는 풀형 액체금속로인KALIMER[1]는 기존의 경수로에 비해 고온에서 운전되므로, 잔열제거계통은 공기 자연순환에 의해 격납용기(CV) 외벽을 직접 냉각하는 방식의 PSDRS(Passive Safety-grade Decay Heat Removal System)[1]를 사용하며, 특히 고온의 구조물 표면온도에 의해 대류전열 외에도 복사전열과정이 활발히 일어난다.(중략)

• Journal of Energy Engineering
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• v.24 no.3
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• pp.96-108
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• 2015

A passive containment cooling system has been designed to remove the heat inside a containment during accidents without external power supply. In this work, the PCCS was introduced in the APR1400 plant to replace the containment spray system and, then, the thermal-hydraulic performance of the PCCS was analyzed using the system thermal-hydraulic computer code, MARS. A double-ended cold-leg break accident, which is known to induce the maximum pressure in the containment, is simulated, where the thermal hydraulics of the PCCS, the reactor coolant system, and the containment are simultaneously simulated. The results of the calculations showed that the PCCS can replace the existing spray system and that the containment building and its internal structure also play a very important role for the heat removal during the accident. Some sensitivity calculations were carried out to evaluate the model uncertainty and the effects of design parameters. The limitations of the PCCS are also discussed.

• Proceedings of the Materials Research Society of Korea Conference
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• 2008.05a
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• pp.40.1-40.1
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• 2008

• Journal of Energy Engineering
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• v.27 no.4
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• pp.27-35
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• 2018

The passive containment cooling system (PCCS) has been designed to remove the released decay heat during the accident by means of the condensation heat transfer phenomenon to guarantee the safety of the nuclear power plant. The heat removal performance of the PCCS is mainly governed by the condensation heat transfer of the steam-air mixture. In this study, the heat removal performance of the PCCS was evaluated by using the MARS-KS code with a new empirical correlation for steam condensation in the presence of a noncondensable gas. A new empirical correlation implemented into the MARS-KS code was developed as a function of parameters that affect the condensation heat transfer coefficient, such as the pressure, the wall subcooling, the noncondensable gas mass fraction and the aspect ratio of the condenser tube. The empirical correlation was applied to the MARS-KS code to replace the default Colburn-Hougen model. The various thermal-hydraulic parameters during the operation of the PCCS follonwing a large-break loss-of-coolant-accident were analyzed. The transient pressure behavior inside the containment from the MARS-KS with the empirical correlation was compared with calculated with the Colburn-Hougen model.