• Nuclear Engineering and Technology
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• 제49권3호
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• pp.459-465
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• 2017

The goal of this study is to evaluate the effective thermal conductivity and diffusivity of containment walls as heat sinks or passive cooling systems during nuclear power plant (NPP) accidents. Containment walls consist of steel reinforced concrete, steel liners, and tendons, and provide the main thermal resistance of the heat sinks, which varies with the volume fraction and geometric alignment of the rebar and tendons, as well as the temperature and chemical composition. The target geometry for the containment walls of this work is the standard Korean NPP OPR1000. Sample tests and numerical simulations are conducted to verify the correlations for models with different densities of concrete, volume fractions, and alignments of steel. Estimation of the effective thermal conductivity and diffusivity of the containment wall models is proposed. The Maxwell model and modified Rayleigh volume fraction model employed in the present work predict the experiment and finite volume method (FVM) results well. The effective thermal conductivity and diffusivity of the containment walls are summarized as functions of density, temperature, and the volume fraction of steel for the analysis of the NPP accidents.

• Nuclear Engineering and Technology
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• 제55권6호
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• pp.2069-2087
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• 2023

In nuclear power plant (NPP) accidents, the containment is subject to high temperatures and high internal pressures, which may further trigger serious chain accidents such as core meltdown and hydrogen explosion, resulting in a significantly higher accident level. Therefore, studying the mechanical performance of a containment under high temperature and high internal pressure is relevant to the safety of NPPs. Based on similarity principles, the 1:3.2 scale model of a prestressed concrete containment vessel (PCCV) of a NPP was designed. The loading method, which considers the thermal-pressure coupling conditions, was used. The mechanical response of the PCCV was investigated with a simultaneous increase in internal pressure and temperature, and the failure mechanism of the PCCV under thermal-pressure coupling conditions was revealed.

• 한국콘크리트학회:학술대회논문집
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• 한국콘크리트학회 1998년도 가을 학술발표대회 논문집(III)
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• pp.649-655
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• 1998

In this study, the personal-computer program was developed to predict prestressing losses containment structures of Nuclear Power Plants by concrete creep and shrinkage. This program is constituted of three parts, which are pre-processor, calculation module and post-processor. Input data for his program are : material properties of concrete, rebar, liner and duct, test results of concrete creep and shrinkage, relative humidity, dimension of containment structures, and the number of prestressing tendon related on containment structures. To obtain better results, this program was made to reflect the prestressing losses due to influence that occurred after prestressing each tendon, thus it can predict prestressing losses and allowable prestressing forces of each tendon. As a case study, this program was applied to containment structures of Youngwang 3 & 4 NPP's and analytical result was compared with test results in In-service Inspection of containment structures. From this comparison, it was proved that this program could well predict prestressing losses by concrete creep and shrinkage.

• Nuclear Engineering and Technology
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• 제51권6호
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• pp.1669-1680
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• 2019

Seismic safety is considered to be one of the key design objectives of AP1000 nuclear power plant (NPP) in strong earthquakes. Dynamic behavior, damage development and aggravation effect are studied in this study for the three main components of AP1000 NPP, namely reinforced concrete shield building (RCSB), steel vessel containment (SVC) and reinforced concrete auxiliary building (RCAB). Characteristics including nonlinear concrete tension and compressive constitutions with plastic damage are employed to establish the numerical model, which is further validated by existing studies. The author investigates three earthquakes and eight input levels with the maximum magnitude of 2.4 g and the results show that the concrete material of both RCSB and RCAB have suffered serious damage in intense earthquakes. Considering RCAB in the whole NPP, significant damage aggravation effect can be detected, which is mainly concentrated at the upper intersection between RCSB and RCAB. SVC and reinforcing bar demonstrate excellent seismic performance with no obvious damage.

• Nuclear Engineering and Technology
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• 제54권1호
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• pp.386-400
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• 2022

Seismic fragility analysis, a part of seismic probabilistic risk assessment (SPRA), is commonly used to establish the relationship between a representative property of earthquakes and the failure probability of a structure, component, or system. Current guidelines on the SPRA of nuclear power plants (NPPs) used worldwide mainly reflect the earthquake characteristics of the western United States. However, different earthquake characteristics may have a significant impact on the seismic fragility of a structure. Given the concern, this study aimed to investigate the effects of earthquake characteristics on the seismic fragility of concrete containments housing the OPR-1000 reactor. Earthquake time histories were created from 30 ground motions (including those of the 2016 Gyeongju earthquake) by spectral matching to the site-specific response spectrum of Hanbit nuclear power plants in South Korea. Fragility curves of the containment structure were determined under the linear response history analysis using a lumped-mass stick model and 30 ground motions, and were compared in terms of earthquake characteristics. The results showed that the median capacity and high confidence of low probability of failure (HCLPF) tended to highly depend on the sustained maximum acceleration (SMA), and increase when using the time histories which have lower SMA compared with the others.

• 한국지진공학회논문집
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• 제10권1호
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• pp.63-74
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• 2006

원전 구조물 및 주요기기의 지진 안전성 평가에서는 내진성능을 정량화하는 방법으로 취약도 분석이 사용되고 있다. 지진취약도 분석은 격납건물의 설계 시 반영된 보수성을 배제한 실질적인 내진성능을 평가하는 것으로 이러한 보수성을 성능 및 응답에 관련된 확률론적 변수로 고려하여 평가하게 된다. 본 연구에서는 비선형 지진 해석으로부터 얻은 구조물의 변위응답을 기초로 한 지진취약도 분석 방법을 제시하였다. 또한 원전부지에서 선정된 발생가능한 근거리지진, 원거리지진, 설계지진 및 확률론적 시나리오지진을 시나리오지진으로 선정하고 이들 지진동에 대한 비선형 지진해석을 통하여 한국 표준형 원전 격납건물의 지진취약도를 평가하였다.

• Nuclear Engineering and Technology
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• 제54권8호
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• pp.2960-2973
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• 2022

Since Fukushima nuclear power plant (NPP) accident in 2011, the importance of research on various severe accident phenomena has been emphasized. Particularly, detailed analysis of combustion risk is necessary following the containment damage caused by combustion in the Fukushima accident. Many studies have been conducted to evaluate the risk of local hydrogen concentration increases and flame propagation using computational code. In particular, the potential for combustion by local hydrogen concentration in specific areas within the containment has been emphasized. In this study, the process of flame propagation generated inside a reactor core to containment during a loss of coolant accident (LOCA) was analyzed using MELCOR 2.1 code. Later in the LOCA scenario, it was expected that hydrogen combustion occurred inside the reactor core owing to oxygen inflow through the cold leg break area. The main driving force of the oxygen intrusion is the elevated containment pressure due to the molten corium-concrete interaction. The thermal and mechanical loads caused by the flame threaten the integrity of the containment. Additionally, the containment spray system effectiveness in this situation was evaluated because changes in pressure gradient and concentrations of flammable gases greatly affect the overall behavior of ignition and subsequent containment integrity.

• 한국전산구조공학회논문집
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• 제25권5호
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• pp.381-388
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• 2012

본 연구에서는 Steel Fiber를 원전 격납건물에 적용하기 위한 적용성 평가를 위해서 Steel Fiber가 삽입된 격납건물에 대한 지진위험도 평가를 수행하였다. Steel Fiber를 콘크리트에 삽입함으로써 콘크리트의 구조적 성능에서 취약점인 인장성능을 향상시킬 수 있고, 압축강도 및 전단강도도 증가시킬 수 있는 장점이 있기 때문이다. 그러나 아직까지 원전 격납건물에 Steel Fiber를 적용하기 위한 노력은 진행되고 있지 않다. 재료적 우수성에도 불구하고 원전에 적용하기 위해서는 좀 더 많은 사용경험과 성능검증이 이루어져야 가능할 것이다. 따라서 본 연구에서는 원자력발전소 격납건물에 Steel Fiber를 사용하였을 경우, 격납건물의 지진안전성의 변화를 살펴보기 위하여 기존의 실험자료를 이용하여 취약도 평가를 수행하였다. 분석결과 Steel Fiber의 함유로 인하여 전단성능과 연성능력이 증가하여 지진취약도의 향상으로 나타났다. Steel Fiber함유량이 1.0%인 경우 지진내력이 10%가량 증가하는 효과를 얻을 수 있었다. 그러나 본 연구의 결과는 제한된 기존의 실험결과를 이용한 예비해석이므로 Steel Fiber의 실제 적용성을 적확하게 분석하기 위해서는 Steel Fiber가 함유된 다양한 콘크리트 부재실험을 통하여 그 물성의 변화를 파악하여야 할 것이다.

• Nuclear Engineering and Technology
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• 제52권5호
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• pp.1051-1065
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• 2020

Safety analysis of nuclear power plant (NPP) especially in accident conditions is a basic and necessary issue for applications and commercialization of reactors. Many previous researches and development works have been conducted. However, most achievements focused on the safety reliability of primary pressure system vessels. Few literatures studied the structural safety of huge concrete structures surrounding primary pressure system, especially for the fourth generation NPP which allows existing of through cracks. In this paper, structural safety reliability of concrete structures of HTR-PM in accidental double-ended break of hot gas ducts was studied by Exceedance Probability Method. It was calculated by Monte Carlo approaches applying numerical simulations by Abaqus. Damage parameters were proposed and used to define the property of concrete, which can perfectly describe the crack state of concrete structures. Calculation results indicated that functional failure determined by deterministic safety analysis was decided by the crack resistance capability of containment buildings, whereas the bearing capacity of concrete structures possess a high safety margin. The failure probability of concrete structures during an accident of double-ended break of hot gas ducts will be 31.18%. Adding the consideration the contingency occurrence probability of the accident, probability of functional failure is sufficiently low.

• Nuclear Engineering and Technology
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• 제49권1호
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• pp.224-233
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• 2017

In Korea, attempts have been made to develop high strength concrete for the safety and design life improvement of nuclear power plants. In this study, the cracking potentials of nuclear power plant-high strength concretes (NPP-HSCs) containing industrial by-products with W/B 0.34 and W/B 0.28, which are being reviewed for their application in the construction of containment structures, were evaluated through autogenous shrinkage, unrestrained drying shrinkage, and restrained drying shrinkage experiments. The cracking potentials of the NPP-HSCs with W/B 0.34 and W/B 0.28 were in the order of 0.34FA25 > 0.34FA25BFS25 > 0.34BFS50 > 0.34BFS65SF5 and 0.28FA25SF5 >> 0.28BFS65SF5 > 0.28BFS45SF5 > 0.28 FA20BFS25SF5, respectively. The cracking potentials of the seven mix proportions excluding 0.28FA25SF5 were lower than that of the existing nuclear power plant concrete; thus, the durability of a nuclear power plant against shrinkage cracking could be improved by applying the seven mix proportions with low cracking potentials.