• Nuclear Engineering and Technology
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• 제28권3호
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• pp.257-264
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• 1996

The conventional Equipment Environment Qualification (EEQ) envelope is developed based on the containment responses during the design basis events. The Safety Depressurization System (SDS) design without In-containment Refueling Water Storage Tank (IRWST) adopted in the Ulchin 3&4 challenges the conventional EEQ envelope during long term Feed and Bleed (F&B) operation due to the direct discharge of high mass and energy into the containment. Therefore, it is necessary to confirm that the containment pressure and temperature history during the long term F&B operation does not violate the conventional EEQ envelope. However, this subject has never been quantitatively assessed before. To investigate the success path of long term F&B operation this paper analyzes the thermal hydraulic response of the containment and Reactor Coolant System (RCS) until the completion of depressurization and cooldown of RCS into Shutdown Cooling System (SCS) entry condition. It is found that the SCS entry condition can be reached within 6 hours without violating the EEQ curve by proper operation of SDS valves, High Pressure Safety Injection (HPSI) pumps and active Containment Heat Removal System (CHRS). The suggested strategy not only demonstrates the feasibility of long term F&B operation but also can be utilized in the preparation of Emergency Procedure Guidelines (EPGs)

• 한국추진공학회:학술대회논문집
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• 한국추진공학회 2011년도 제36회 춘계학술대회논문집
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• pp.414-417
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• 2011

본 논문에서 Compressor 및 turbine 에서의 Blade failure등의 내부파손이 이를 둘러싸고 있는 케이스 내부에 머무르게 하는 엔진설계의 방법에 대한 이론 및 Simulation 등을 기술 하였다. 가장 무거운 부품 중에 하나인 케이스의 두께 최적화는, 항공기의 안정성뿐만 아니라 항공 효율을 높이기 위한 경량화의 목적을 위해서도 매우 중요한 설계목표 이다. 이러한 목적을 위하여 이론적 접근방법으로 에너지 밸런스 방법을 사용하였으며, 파손된 블레이드의 거동특성 및 영향성 평가를 위한 유한요소해석을 위하여 LS-DYNA가 사용 되어졌다.

• Nuclear Engineering and Technology
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• 제49권5호
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• pp.941-952
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• 2017

Using the Generation of Thermal-Hydraulic Information for Containments (GOTHIC) code, thermal-hydraulic phenomena that occur inside the containment have been investigated, along with the preliminary design of the passive containment cooling system (PCCS) of an innovative pressurized water reactor (PWR). A GOTHIC containment model was constructed with reference to the design data of the Advanced Power Reactor 1400, and report related PCCS. The effects of the design parameters were evaluated for passive containment cooling tank (PCCT) geometry, PCCS heat exchanger (PCCX) location, and surface area. The analyzed results, obtained using the single PCCT, showed that repressurization and reheating phenomena had occurred. To resolve these problems, a coupled PCCT concept was suggested and was found to continually decrease the containment pressure and temperature without repressurization and reheating. If the installation level of the PCCX is higher than that of the PCCT, it may affect the PCCS performance. Additionally, it was confirmed that various means of increasing the external surface area of the PCCX, such as fins, could help improve the energy removal performance of the PCCS. To improve the PCCS design and investigate its performance, further studies are needed.

• Nuclear Engineering and Technology
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• 제54권1호
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• pp.177-185
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• 2022

Hydrogen risk in the containment filtered venting system (CFVS) vessel was analyzed, considering operation pressure and modes with the effect of PAR and accident scenarios. The CFVS is to depressurize the containment by venting the containment atmosphere through the filtering system. The CFVS could be subject to hydrogen risk due to the change of atmospheric conditions while the containment atmosphere passes through the CFVS. It was found that hydrogen risk increased as the CFVS opening pressure was set higher because more combustible gases generated by Molten Core Concrete Interaction flowed into the CFVS. Hydrogen risk was independent of operation modes and found only at the early phase of venting both for continuous and cyclic operation modes. With PAR, hydrogen risk appeared only at the 0.9 MPa opening pressure for Station Black-Out accidents. Without PAR, however, hydrogen risk appeared even with the CFVS opening set-point of 0.5 MPa. In a slow accident like SBO, hydrogen risk was more threatening than a fast accident like Large Break Loss-of-Coolant Accident. Through this study, it is recommended to set the CFVS opening pressure lower than 0.9 MPa and to operate it in the cyclic mode to keep the CFVS available as long as possible.

• Nuclear Engineering and Technology
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• 제54권11호
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• pp.4159-4169
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• 2022

Hydrogen mitigation using Passive Autocatalytic Recombiners (PARs) has been widely accepted methodology inside reactor containment of accident struck Nuclear Power Plants. They reduce hydrogen concentration inside reactor containment by recombining it with oxygen from containment air on catalyst surfaces at ambient temperatures. Exothermic heat of reaction drives the product steam upwards, establishing natural convection around PAR, thus invoking homogenisation inside containment. CFD models resolving individual catalyst plate channels of PAR provide good insight about temperature and hydrogen recombination. But very thin catalyst plates compared to large dimensions of the enclosures involved result in intensive calculations. Hence, empirical correlations specific to PARs being modelled are often used in integral containment studies. In this work, an experimentally validated CFD model of PAR has been employed for developing an empirical correlation for Indian PAR. For this purpose, detailed parametric study involving different gas mixture variables at PAR inlet has been performed. For each case, respective values of gas mixture variables at recombiner outlet have been tabulated. The obtained data matrix has then been processed using regression analysis to obtain a set of correlations between inlet and outlet variables. The empirical correlation thus developed, can be easily plugged into commercially available CFD software.

• Nuclear Engineering and Technology
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• 제56권4호
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• pp.1513-1525
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• 2024

An improved mitigation system for thermally-induced steam generator tube rupture accidents was introduced to prevent direct environmental release of fission products bypassing the containment in the OPR1000. This involves injecting bypassed steam into the containment, cooling, and decontaminating it using a water coolant tank. To evaluate its performance, a severe accident analysis was performed using the MELCOR 2.2 code for OPR1000. Simulation results show that the proposed system sufficiently prevented the release of radioactive nuclides (RNs) into the environment via containment injection. The pool scrubbing system effectively decontaminated the injected RN and consequently reduced the aerosol mass in the containment atmosphere. However, the decay heat of the collected RNs causes re-vaporization. To restrict the re-vaporization, an external water source was considered, where the decontamination performance was significantly improved, and the RNs were effectively isolated. However, due to the continuous evaporation of the feed water caused by decay heat, a substantial amount of steam is released into the containment. Despite the slight pressurization inside the containment by the injected and evaporated steam, the steam decreased the hydrogen mole fraction, thereby reducing the possibility of ignition.

• Nuclear Engineering and Technology
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• 제28권2호
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• pp.129-136
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• 1996

영향도를 이용하여 영광 3, 4호기의 격납용기 성능해석을 수행하였다. 기존의 사상수목기법을 응용한 격납용기 성능해석은 사건들 사이의 의존 관계를 명확히 나타내기 어렵고, 사고진행사상수목(APET) 에서 알 수 있듯이, 격납용기와 같은 복잡한 계통에 적용할 경우 그 의존 관계를 그림으로조차 나타낼 수가 없으며, 또한, 의사결정문제를 다루는 데에도 많은 한계점을 지니고 있다. 이러한 문제점들을 해결하기 위하여 새로이 개발된 방법론인 영향도를 영광 3, 4호기 격납용기 성능해석과 사고관리방안을 평가하는 데에 적용하여 보았다. 본 연구에서 얻은 계산 결과와 기존의 사상수목 기법을 이용하여 계산한 결과와 비교한 결과, 거의 일치하는 계산 결과를 얻을 수 있으면서도 전체 격납용기 계통을 한 눈에 알기 쉽게 그림으로 나타낼 수 있었다. 또한, 향도가 의사결정문제를 일반적으로 다룰 수 있음을 보이기 위하여 본 방법론을 사고관리방안을 평가하는 데에 이용하여, 원자로 냉각계통 감압과 원자로공동 범람 방안, 두 가지 사고관리방안을 평가하여 보았다. 모두 초기 격납용기 파손에는 나쁜 영향을 주는 것으로 나 타났으나, 후기 격납용기 파손이나 중기발생기 세관파손에는 원자로공동범람과 일차계통 감압이 각각 어느 정도 긍정적인 영향을 미치는 것으로 나타났다. 본 연구를 통하여, 영향도를 이용한 격납응기 성능 해석은 사상수목기법을 이용한 분석에 비해, 진행되는 사건들 사이의 의존관계를 보다 명확히 나타낼 수 있고, 또한 영향도는 운전자의 의사결정을 잘 나타낼 수 있으므로 사고관리기법을 평가하는 데에도 쉽게 적용할 수 있음을 알 수 있다. 결론적으로, 본 연구에서는 영향도가 사상수목기법이 지니고 있는 여러 한계점들을 쉽게 극복하며 격납용기 성능해석에 적용할 수 있음을 보였다.

• 한국전산구조공학회논문집
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• 제23권4호
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• pp.445-452
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• 2010

CANDU형 격납건물에 대하여 극한내압하중에 대한 확률론적 취약도 평가를 수행하였다. 격납건물 성능의 불확실성은 가동중 검사 결과를 통해 얻어진 재료 물성치 중앙값과 텐던 긴장력 중앙값을 적용하여 고려하였다. 격납건물은 개구부를 고려하여 3차원 유한요소로 모델링하였으며, 확률론적 취약도 평가를 위하여 대규모의 비선형 유한요소해석 모델을 적용하기에 적합한 효율적인 취약도 평가기법을 개발하였다. 월성 1호기 격납건물에 대한 물성치를 사용하였다. 개발된 새로운 취약도 평가기법을 도입하여 각각의 파괴모드에 대한 취약도 평가를 수행하였으며, 파괴모드 별, 신뢰도 수준별 취약도 곡선을 도출하였다. 벽체 중단부가 극한내압발생으로 인한 방사능물질 누출에 가장 취약한 것으로 나타났다.

• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 2009년 추계학술대회논문집
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• pp.121-128
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• 2009

During a hypothetical high-pressure accident in a nuclear power plant (NPP), molten corium can be ejected through a breach of a reactor pressure vessel (RPV) and dispersed by a following jet of a high-pressure steam in the RPV. The dispersed corium is fragmented into smaller droplets in a reactor cavity of the NPP by the steam jet and released into other compartments of the NPP by a overpressure in the cavity. The fragments of the corium transfer thermal energy to the ambient air in the containment or interact chemically with steam and generate hydrogen which may be burnt in the containment. The thermal loads from the ejected molten corium on the containment which is called direct containment heating (DCH) can threaten the integrity of the containment. DCH in a NPP containment is related to many physical phenomena such as multi-phase hydrodynamics, thermodynamics and chemical process. In the evaluation of the DCH load, the melt dispersion rates depending on the RPV pressure are the most important parameter. Mostly, DCH was evaluated by using lumped-analysis codes with some correlations obtained from experiments for the dispersion rates. In this study, MC3D code was used to evaluate the dispersion rates in the APR1400 NPP during the high-pressure accidents. MC3D is a two-phase analysis code based on Eulerian four-fields for melt jet, melt droplets, gas and water. The dispersion rates of the corium melt depending on the RPV pressure were obtained from the MC3D analyses and the values specific to the APR1400 cavity geometry were compared to a currently available correlation.

• 한국전산유체공학회지
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• 제10권3호
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• pp.9-18
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• 2005

During a hypothetical severe accident in a nuclear power plant (NPP), hydrogen is generated by the active reaction of fuel-cladding and steam in the reactor pressure vessel and released with steam into the containment. In order to mitigate hydrogen hazards possibly occurred in the NPP containment, hydrogen mitigation system (HMS) is usually adopted. The design of the next generation NPP (APR1400) designed in Korea specifies 26 passive autocatalytic recombiners and 10 igniters installed in the containment for the hydrogen mitigation. in this study, the analysis of the hydrogen and steam behavior during a total lose of feed water (TLOFW) accident in the APR1400 containment has been conducted by using the CFD code GASFLOW. During the accident, a huge amount of hot water, steam, and hydrogen is released in the in-containment refueling water storage tank (IRWST). The current design of the APR1400 includes flap-type dampers at the IRWST vents which are operated depending on the pressure difference between inside and outside of the IRWST. it was found that the flaps strongly affects the flow structure of the steam and hydrogen in the containment. The possibilities of a flame acceleration and transition from deflagration to detonation (DDT) were evaluated by using Sigma-Lambda criteria. Numerical results indicate the DDT possibility could be heavily reduced in the IRWST compartment when the flaps are installed.