• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2004년도 요약집 춘계학술발표대회
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• pp.388.1-388.1
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• 2004

• Nuclear Engineering and Technology
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• 제53권12호
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• pp.4179-4188
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• 2021

This study identifies efficient earthquake intensity measures (IMs) for seismic performances and fragility evaluations of the reactor containment building (RCB) in the advanced power reactor 1400 (APR1400) nuclear power plant (NPP). The computational model of RCB is constructed using the beam-truss model (BTM) for nonlinear analyses. A total of 90 ground motion records and 20 different IMs are employed for numerical analyses. A series of nonlinear time-history analyses are performed to monitor maximum floor displacements and accelerations of RCB. Then, probabilistic seismic demand models of RCB are developed for each IM. Statistical parameters including coefficient of determination (R²), dispersion (i.e. standard deviation), practicality, and proficiency are calculated to recognize strongly correlated IMs with the seismic performance of the NPP structure. The numerical results show that the optimal IMs are spectral acceleration, spectral velocity, spectral displacement at the fundamental period, acceleration spectrum intensity, effective peak acceleration, peak ground acceleration, A95, and sustained maximum acceleration. Moreover, weakly related IMs to the seismic performance of RCB are peak ground displacement, root-mean-square of displacement, specific energy density, root-mean-square of velocity, peak ground velocity, Housner intensity, velocity spectrum intensity, and sustained maximum velocity. Finally, a set of fragility curves of RCB are developed for optimal IMs.

• 한국전산구조공학회논문집
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• 제26권2호
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• pp.157-164
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• 2013

본 연구에서는 여객용 항공기 충돌 시 강섬유보강콘크리트를 사용한 철근콘크리트 원전 격납건물의 구조적 거동을 유한요소해석을 이용하여 고찰한다. 항공기 충돌에 의해 원전 격납건물에 가해지는 하중은 Riera 충격하중 시간함수와 충돌 시 접촉면적을 이용하여 모델링하였다. 강섬유보강콘크리트의 재료모델은 CSCM Concrete Model을 사용하였다. 기존에 제안된 강섬유보강콘크리트의 강도예상모델을 이용하여 재료모델의 입력변수를 결정하였다. 강섬유의 함유량에 따른 원전 격납 건물의 항공기 충돌에 대한 구조적 거동을 상용 유한요소 해석 프로그램인 LS-DYNA를 이용하여 해석하였다. 해석결과를 바탕으로 항공기 충돌에 대한 저항성을 평가하였으며, 보수적인 안전성이 요구되는 원전 격납건물에 강섬유보강콘크리트를 적용할 경우 항공기 충돌에 대한 저항성 증대 효과를 기대할 수 있는 것으로 고찰되었다.

• 콘크리트학회논문집
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• 제17권4호
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• pp.521-526
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• 2005

원전 격납건물의 축대칭 모델은 해석상의 간편성으로 인하여 널리 사용된다. 하지만, 일반적인 돔 텐던의 배치는 축대칭 형상이 아니며 곡률을 가진 돔에 임의로 배치된 관계로 축대칭 근사화시 좀 더 엄밀한 수학적 유도가 요구된다. 본 연구에서는 국내의 CANDU형 및 한국형 격납건물 돔에 비축대칭으로 배치된 텐던을 축대칭 모델에 적용하기 위한 합리적인 변환 절차를 제안하였다. 텐던 강성의 모델링에서는 실제 3차원으로 배치된 돔 텐던의 자오선방향 및 원환방향으로의 강성 기여를 고려할 수 있도록 텐던을 등가의 층으로 근사화하였다. 프리스트레싱의 효과는 등가하중법 및 초기응력법 관점에서 고찰하였으며, 축대칭 모델의 방법론에 적합하도록 등가하중 및 초기응력을 유도하였다. 후속 논문에서는 제안된 모델을 적용한 수치 예제들을 범용구조해석 프로그램으로 해석하고 타당성을 검증하였다.

• Journal of Advanced Marine Engineering and Technology
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• 제39권5호
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• pp.516-521
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• 2015

This study is about a hydrogen mitigation system in a containment building like an offshore or a nuclear plant. A hydrogen explosion is possibly happened after condensation of steam if hydrogen releases with steam in a containment buildings. Passive autocatalytic recombiner is the one of the measures, but the performance of this equipment is not sure because the distribution of hydrogen is very irregular and is not predicted correctly. This study proposes a new approach for improving the hydrogen removing performance with hydrogen-guiding property. The steam is simulated and analysed. The results show that the shallow air containment reduced over 55% of the released hydrogen and the deep air containment type reduces over 80% of released hydrogen.

• Nuclear Engineering and Technology
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• 제55권10호
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• pp.3677-3691
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• 2023

Since containment building is the major shielding structure to ensure safety of nuclear power plant, the structural behavior and ultimate pressure capacity of containments must be studied in depth. This paper addresses ambiguous issue of determining free-field position for liner failure by suggesting an expanded free-field region and comparing internal pressure capacities obtained by test data, conservative assumption and suggested free-field region. For this purpose, a practical approach to determine the free-field position for the evaluation of liner tearing is carried out. The maximum principal strain histories versus internal pressure capacities among different free-field positions at various azimuths and elevations are compared with those at the equipment hatch as a conservative assumption. The comparison shows that there are considerable differences in the internal pressure capacity at liner failure within the expanded free-field region compared to the vicinity of the equipment hatch. Additionally, this study proposes an approximate correlation with conservative factors by considering the expanded free-field ranges and material characteristics to determine realistic failure criteria for liner. The applicability of the proposed correlation is demonstrated by comparing the internal pressure capacities of full-scale containment buildings following liner failure criteria according to RG 1.216 and an approximate correlation.

• 한국콘크리트학회:학술대회논문집
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• 한국콘크리트학회 2004년도 추계 학술발표회 제16권2호
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• pp.81-84
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• 2004

Prestressing tendons in a nuclear containment building dome are non-axisymmetrically arranged in most cases. However, simple axisymmetric modeling of the containment has been often employed in practice, which requires the axisymmetric approximation of the actual tendon arrangements in the dome. A procedure was previously proposed that can implement the actual 3D tendon stiffness and prestressing effect into the axisymmetric model for CANDU type. This paper further verifies and compares some methodologies adopted in the proposed scheme through some numerical examples.

• 한국건축시공학회:학술대회논문집
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• 한국건축시공학회 2014년도 추계 학술논문 발표대회
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• pp.212-213
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• 2014

The purpose of this study is to suggest analysis model of RCB construction in nuclear power plant. For the objective, This study drew the risk factors of RCB construction from existing literature. The results of the study proposed analysis model made hierarchy in rebar, form, and concrete work. These will be baseline data for risk management in construction project of nuclear power plant.

• Nuclear Engineering and Technology
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• 제47권1호
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• pp.11-25
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• 2015

The accident at Japan's Fukushima Daiichi nuclear power plant in March 2011, caused by an earthquake and a subsequent tsunami, resulted in a failure of the power systems that are needed to cool the reactors at the plant. The accident progression in the absence of heat removal systems caused Units 1-3 to undergo fuel melting. Containment pressurization and hydrogen explosions ultimately resulted in the escape of radioactivity from reactor containments into the atmosphere and ocean. Problems in containment venting operation, leakage from primary containment boundary to the reactor building, improper functioning of standby gas treatment system (SGTS), unmitigated hydrogen accumulation in the reactor building were identified as some of the reasons those added-up in the severity of the accident. The Fukushima accident not only initiated worldwide demand for installation of adequate control and mitigation measures to minimize the potential source term to the environment but also advocated assessment of the existing mitigation systems performance behavior under a wide range of postulated accident scenarios. The uncertainty in estimating the released fraction of the radionuclides due to the Fukushima accident also underlined the need for comprehensive understanding of fission product behavior as a function of the thermal hydraulic conditions and the type of gaseous, aqueous, and solid materials available for interaction, e.g., gas components, decontamination paint, aerosols, and water pools. In the light of the Fukushima accident, additional experimental needs identified for hydrogen and fission product issues need to be investigated in an integrated and optimized way. Additionally, as more and more passive safety systems, such as passive autocatalytic recombiners and filtered containment venting systems are being retrofitted in current reactors and also planned for future reactors, identified hydrogen and fission product issues will need to be coupled with the operation of passive safety systems in phenomena oriented and coupled effects experiments. In the present paper, potential hydrogen and fission product issues raised by the Fukushima accident are discussed. The discussion focuses on hydrogen and fission product behavior inside nuclear power plant containments under severe accident conditions. The relevant experimental investigations conducted in the technical scale containment THAI (thermal hydraulics, hydrogen, aerosols, and iodine) test facility (9.2 m high, 3.2 m in diameter, and $60m^3$ volume) are discussed in the light of the Fukushima accident.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 추계학술발표회
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• pp.801-802
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• 2005