• 한국지진공학회:학술대회논문집
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• 한국지진공학회 2005년도 학술발표회 논문집
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• pp.453-460
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• 2005

In this study, the seismic fragility analysis of a base isolated condensate storage tank installed in the nuclear power plant. The condensate storage tank is safety related structure in a nuclear power plant. The failure of this tank affect significantly to the core damage frequency of the nuclear power plants. The seismic analysis of the liquid storage tank was performed by the simple calculation method and the dynamic time storage analysis method. The convective and impulsive fluid mass is modeled as added masses proposed by several researchers. To evaluate the effectiveness of the isolation system, the comparison of HCLPF and core damage frequencies in non-isolated and isolated cases are carried out. It can be found from the results that the seismic isolation system increases the seismic capacity of a condensate storage tank and decreases the core damage frequency significantly.

• Nuclear Engineering and Technology
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• 제48권5호
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• pp.1252-1263
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• 2016

This paper presents a numerical method to simulate ductile tearing in cracked components under high strain rates using finite element damage analysis. The strain rate dependence on tensile properties and multiaxial fracture strain is characterized by the model developed by Johnson and Cook. The damage model is then defined based on the ductility exhaustion concept using the strain rate dependent multiaxial fracture strain concept. The proposed model is applied to simulate previously published three cracked pipe bending test results under two different test speed conditions. Simulated results show overall good agreement with experimental results.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2004년도 추계학술발표회 발표논문집
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• pp.674-674
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• 2004

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 춘계학술발표회
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• pp.960-961
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• 2005

• Nuclear Engineering and Technology
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• 제53권9호
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• pp.3068-3084
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• 2021

Investigations of large commercial aircraft impact effect on nuclear power plant (NPP) buildings have been drawing extensive attentions, particularly after the 9/11 event, and this paper aims to experimentally assess the damage and vibrations of NPP buildings subjected to aircraft crash. In present Part I, two shots of reduce-scaled model test of aircraft impacting on NPP building were carried out. Firstly, the 1:15 aircraft model (weighs 135 kg) and RC NPP model (weighs about 70 t) are designed and prepared. Then, based on the large rocket sled loading test platform, the aircraft models were accelerated to impact perpendicularly on the two sides of NPP model, i.e., containment and auxiliary buildings, with a velocity of about 170 m/s. The strain-time histories of rebars within the impact area and acceleration-time histories of each floor of NPP model are derived from the pre-arranged twenty-one strain gauges and twenty tri-axial accelerometers, and the whole impact processes were recorded by three high-speed cameras. The local penetration and perforation failure modes occurred respectively in the collision scenarios of containment and auxiliary buildings, and some suggestions for the NPP design are given. The maximum acceleration in the 1:15 scaled tests is 1785.73 g, and thus the corresponding maximum resultant acceleration in a prototype impact might be about 119 g, which poses a potential threat to the nuclear equipment. Furthermore, it was found that the nonlinear decrease of vibrations along the height was well reflected by the variations of both the maximum resultant vibrations and Cumulative Absolute Velocity (CAV). The present experimental work on the damage and dynamic responses of NPP structure under aircraft impact is firstly presented, which could provide a benchmark basis for further safety assessments of prototype NPP structure as well as inner systems and components against aircraft crash.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.887-892
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• 1998

A risk informed approach to relax AOTs and STIs of RPS/ESPAS in technical specifications of kori units 3, 4 was performed in this paper. With the proposed AOTs and STIs, system unavailabilities and core damage frequency were quantified using PSA model. The results shoe that core damage frequency is slightly increased by extending AOTs and STIs but negligible. As considering the benefits such as reduction of plant transients and man power for test and maintenance, the relaxation of AOTs/STIs of RPS/ESFAS is justified

• Nuclear Engineering and Technology
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• 제51권2호
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• pp.561-572
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• 2019

In the design criterion for the nuclear power plant piping system, the limit state of the piping against an earthquake is assumed to be plastic collapse. The failure of a common piping system, however, means the leakage caused by the cracks. Therefore, for the seismic fragility analysis of a nuclear power plant, a method capable of quantitatively expressing the failure of an actual piping system is required. In this study, it was conducted to propose a quantitative failure criterion for piping system, which is required for the seismic fragility analysis of nuclear power plants against critical accidents. The in-plane cyclic loading test was conducted to propose a quantitative failure criterion for steel pipe elbows in the nuclear power plant piping system. Nonlinear analysis was conducted using a finite element model, and the results were compared with the test results to verify the effectiveness of the finite element model. The collapse load point derived from the experiment and analysis results and the damage index based on the stress-strain relationship were defined as failure criteria, and seismic fragility analysis was conducted for the piping system of the BNL (Brookhaven National Laboratory) - NRC (Nuclear Regulatory Commission) benchmark model.

• Nuclear Engineering and Technology
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• 제54권8호
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• pp.2989-2998
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• 2022

Line contact structures, such as the contact between graphite brick and graphite tenon, widely exist in high-temperature gas-cooled reactors. Due to the stress concentration effect, the line contact area is one of the dangerous positions prone to failure in the nuclear reactor core. In this paper, the failure mechanism of line contact structures composed of IG11 nuclear graphite column and brick were investigated by means of experiment and finite element simulation. It was found that the failure process mainly includes three stages: firstly, the damage accumulation in nuclear graphite material led to the characteristic yielding of the line contact structure, but no macroscopic failure can be observed at this stage; secondly, the stresses near the contact area met Mohr failure criterion, and a crack initiated and propagated laterally in the contact zone, that is, local macroscopic failure occurred at this stage; finally, a second crack initiated in the contact area and developed in to a Y-shape, resulting in the final failure of the structure. This study lays a foundation for the structural design and safety assessment of high-temperature gas-cooled reactors.

• Biomolecules & Therapeutics
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• 제22권2호
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• pp.136-142
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• 2014

We investigated the protective effects of chlorogenic acid (CGA), a polyphenol compound, on oxidative damage induced by UVB exposure on human HaCaT cells. In a cell-free system, CGA scavenged 1,1-diphenyl-2-picrylhydrazyl radicals, superoxide anions, hydroxyl radicals, and intracellular reactive oxygen species (ROS) generated by hydrogen peroxide and ultraviolet B (UVB). Furthermore, CGA absorbed electromagnetic radiation in the UVB range (280-320 nm). UVB exposure resulted in damage to cellular DNA, as demonstrated in a comet assay; pre-treatment of cells with CGA prior to UVB irradiation prevented DNA damage and increased cell viability. Furthermore, CGA pre-treatment prevented or ameliorated apoptosis-related changes in UVB-exposed cells, including the formation of apoptotic bodies, disruption of mitochondrial membrane potential, and alterations in the levels of the apoptosis-related proteins Bcl-2, Bax, and caspase-3. Our findings suggest that CGA protects cells from oxidative stress induced by UVB radiation.

• 한국압력기기공학회 논문집
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• 제18권2호
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• pp.43-49
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• 2022

The purpose of this study is to evaluate the radiation effect on damage when the external shield of the spent nuclear fuel transport cask is damaged due to impact as the cause of an unexpected accident. The neutron and gamma-ray intensities and spectra are calculated using the ORIGEN-Arp module in the SCALE 6.2.4 code package⁽¹⁾ and then using MCNP6.2⁽²⁾ code calculate the dose rate. In order to evaluate the radiation dose according to the size of damage caused by external impact, various sized holes of 0.3~13.7% are assumed in the outer shield of the cask to evaluate the sensitivity to the dose. In the case of radiation source leakage, damage to the nuclear fuel assembly is assumed to be up to 6% based on overseas test cases. When only the outer shield is damaged, the maximum surface dose is calculated as 3.12E+03 mSv/hr. However, if the radiation source is leaked due to damage to the nuclear fuel assembly, it becomes 7.00E+05 mSv/hr which is about 200 times greater than the former case.