• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.14 no.2
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• pp.41-48
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• 2018

In this study, the dynamic absorber and the damper are applied to improve the seismic performance of the piping system, and their quantitative effects on the piping system performance are examined. For this purpose, the response performances of piping system applied with the dynamic absorber/damper are compared with those of the original piping system. Firstly, the frequency response analyses of the piping system with the presence or the absence of dynamic absorber/damper are performed and these results are compared. It has been shown that the maximum acceleration response per the frequency of the piping system is considerably reduced by installing the dynamic absorber and the damper. Secondly, the seismic responses of the piping systems with and without dynamic absorber/damper are compared. As a result of the numerical analyses, it is confirmed that key responses are reduced by 17%-63% due to the installation of the dynamic absorber and damper. Finally, as a result of the seismic performance evaluation, it is confirmed that the HCLPF (High Confidence of Low Probability of Failure) seismic performances are increased by 1.22 to 2.70 times with respect to the failure modes with an aid of the dynamic absorber and damper.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.2
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• pp.25-33
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• 2016

The thermal stratification phenomenon in the nuclear power plant can cause abnormal deformation of the piping, contact with the support, damage to the support system. Repetition of the thermal stratification phenomenon or variation of the thermal boundary layer can cause thermal fatigue. Thermal stratification phenomenon in nuclear power plants is still an ongoing issue and active research has been carried out. In this paper, the current situation in Korean nuclear power plants is described, followed by the status of research and the future problems on the thermal stratification phenomenon in Korea.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.1
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• pp.56-61
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• 2016

After Japanese Fukushima nuclear power plant accident caused by the beyond design basis earthquake and tsunami, it has turned to be a major challenge for nuclear safety. IAEA, US NRC and EU have provided new safety design standards for beyond design basis event, Domestic regulatory bodies have also enacted guidances for licensees and applicants on additional methods related to beyond design basis events. This paper describes several evaluation methods for applying to nuclear power plants piping for beyond design basis earthquake. As a results, energy method based on the absorbed energy on nuclear power plant, deterministic method following design code and theory, experience method considering past earthquake data and information and probabilistic methods similar to probabilistic risk assessment were reviewed.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.11 no.2
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• pp.25-30
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• 2015

In the present study, complex leakage probabilities of nuclear pipes due to fatigue and stress corrosion cracking are evaluated by using the PINTIN(Piping INTegrity INner flaws) that is developed based on the existing PRAISE(Piping Reliability Analysis Including Seismic Events) program. With regard to the aging and crack instability, small leak and big leak probabilities are calculated for several pipes in a reactor coolant system of domestic nuclear plant. Moreover, sensitivity analysis is also performed to find out the effect of parameters for the leakage of pipes, which shows the coolant temperature is the most influencing parameter.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.16 no.2
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• pp.25-34
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• 2020

As leakage in nuclear power plants could cause a variety of problems, it is very critical to monitor leakage from the safety point of view. Accordingly, a new type of leak detection system is currently being developed and flow characteristics of the sampling and transportation system are investigated by using numerical analysis as a part of the development process in this study. The results showed that the steam mass fraction varied according to the effect of the gap between the insulation and piping component, transportation velocity, and material properties of porous media during the sampling and transportation process. The results of this study should be useful for understanding flow characteristics of the sampling and transportation system and its design and application.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.25 no.9
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• pp.599-605
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• 2015

The revised U.S. Nuclear Regulatory Commission(NRC), Regulatory Guide(RG) 1.20, rev.3 requires the evaluation of the potential adverse effects from pressure fluctuations and vibrations on piping and components for the reactor coolant, steam, feedwater, and condensate systems. Detailed vibration analyses for the systems attached to the steam generator are very difficult, because these piping systems are very complicated. This paper suggests a screening method for the flow-induced vibration of acoustic resonances and pump-induced vibration of the piping systems attached to the steam generator in order to conduct the APR1400 comprehensive vibration assessment program. This paper seeks to address the areas such as potential vibration sources, and methods to prevent the occurrence of acoustic resonances and pump-induced vibration of piping systems attached to the steam generator, for conducting the APR1400 comprehensive vibration assessment program. The screening method in this paper will be used to estimate the flow-induced vibration of the piping systems attached to the steam generator for the APR1400.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.37 no.9
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• pp.1175-1182
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• 2013

The lock-in mode infrared thermography (IRT) technique has been developed to improve the detection capability of defects in materials with high thermal conductivity, and it has been shown to provide better detection capability than conventional active IRT. Therefore, to investigate the application of this technique to nuclear piping components, lock-in mode IRT tests were conducted on pipe specimens containing simulated wall-thinning defects. Phase images of the wall-thinning defects were obtained from the tests, and they were compared with thermal images obtained from conventional active IRT tests. It showed that the ability to size the detected wall-thinning defects in piping components was improved by using lock-in mode IRT. The improvement was especially apparent when detecting short and narrow defects and defects with slanted edges. However, the detection capability for shallow wall-thinning defects did not improve much when using lock-in mode IRT.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.11 no.1
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• pp.69-73
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• 2015

The study investigated effects of residual stresses and Charpy impact energy on brittle fractures of the butt weld between the valve and the piping, and of the valve body in nuclear power plants via a linear elastic fracture mechanics approach in the ASME B&PV Code, Sec.XI and finite element analysis. Weld residual stress in a butt weld between close check valve and piping, and residual stress in the valve due to casting process were assumed to be proportional to yield strength of base metal. Operating stresses in the butt weld and the valve body were calculated using approximate engineering formulae and finite element analysis, respectively. Applied stress intensity factors were calculated by assuming postulated cracks with specific sizes and then by substituting the residual stresses and the operating stresses into engineering formulae presented in the ASME B&PV Code, Sec.III. Plane strain fracture toughness was derived by using a correlation between Charpy V-notch impact energy and fracture toughness. Structural integrity of the weld and the body against brittle fracture was assessed by using the applied stress intensity factors, plane strain fracture toughness and the linear elastic fracture mechanics approach. As a result, it was identified that the structural integrity was maintained with decreasing the residual stress levels and increasing the Charpy V-notch impact energy.

• Proceedings of the Korean Information Science Society Conference
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• 2001.10b
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• pp.238-240
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• 2001

원자력발전소에서는 기기, 배관 및 각종 지지구조물 등 설비에 대하여 시간의 경과에 따른 취약화 정도를 측정하기 위하여 대략 15개월을 주기로 호기별 비파괴검사로 감시 및 평가하는 가동중검사를 실시한다. 증기발생기, 주복수기와 같은 세관으로 구성된 설비는 와전류탐상검사를 수행하여 신호데이터를 취득하고 건전성 여부를 평가한 다음 그 결과를 Optical Disk에 신호데이터와 함께 저장한다. 본 논문에서는 저장된 방대한 양의 검사 결과 파일을 추출하여 데이터베이스로 구축하고, 행열 수량, 모양, 방향 및 열번호 부여방법이 상이한 다양한 배열 형태의 세관 Map을 편집하여 사용자 요구에 따라 검사정보를 색상 Tube로 Mapping 처리하여 세관의 상태, 검사이력, 결함성장률 및 변화추이 분석을 시각적으로 파악할 수 프로그램 구현 사례를 소개한다.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.2
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• pp.66-70
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• 2016

Following the 2011 Fukushima Nuclear Power Plant accident, the IAEA has issued a revised version of the Nuclear Safety Standard for beyond design basis earthquake to consider the core meltdown accident. In Korea, relevant laws and regulations were also revised to consider beyond design basis earthquake to nuclear components. In this paper, CAV, an seismic damage factor that determines the restart of nuclear power plant after operating breakdown earthquake, is proposed for extension to the beyond design basis earthquake. For pipings not satisfying the beyond design basis earthquake condition, several evaluation methods are suggested, such as strain-based evaluation methods, simple nonlinear analysis method and cumulative damage evaluation method.