• 대한용접접합학회:학술대회논문집
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• 대한용접접합학회 1996년도 특별강연 및 추계학술발표 개요집
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• pp.19-34
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• 1996

A review was made of fatigue failures of nuclear power plant components in Japan, which were experienced in service and during periodical inspection. No case has been recently reported of a service fatigue failure of a reactor pressure vessel itself, excluding nozzle corner cracks, that occurred many years ago. But, service fatigue failures have been occasionally experienced in piping systems, pumps, and valves, on which fatigue design seems to have been inadequately applied. The causes of fatigue failures can be divided into two categories: mechanical-vibration-induced fatigue and thermal-fluctuation-induced fatigue. Vibration-induced fatigue failure occurs more frequently than is generally thought. The lesson gleaned from the present survey is a recognition that a service fatigue failure may occur due to any one or a combination of the following factors: (1) lack of communication between designers and fabrication engineers, (2) lack of knowledge about a possibility of fatigue failure and poor consideration about the effects of residual stresses, (3) lack of consideration on possible vibration in the design and fabrication stages, and (4) lack of fusion or poor penetration in a welded joint.

• Nuclear Engineering and Technology
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• 제55권5호
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• pp.1616-1629
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• 2023

To quantify the conservatism of existing ASME strain-based evaluation methods for seismic loading, this paper presents very low cycle fatigue test data of elbows under various cyclic loading conditions and comparison of evaluation results with experimental failure cycles. For strain-based evaluation methods, the method presented in ASME BPVC CC N-900 and Sec. VIII are used. Predicted failure cycles are compared with experimental failure cycle to quantify the conservatism of evaluation methods. All methods give very conservative failure cycles. The CC N-900 method is the most conservative and prediction results are only ~0.5% of experimental data. For Sec. VIII method, the use of the option using code tensile properties gives ~3% of experimental data, and the use of the material-specific reduction of area can reduce conservatism but still gives ~15% of experimental data.

• 한국재난정보학회 논문집
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• 제14권4호
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• pp.450-457
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• 2018

연구목적: 2016년 9월 경주 이후 구조물 및 비구조물의 지진 안전성 및 내진성능에 관한 문제가 이슈화 되고 있으며, 특히 배관 시스템의 경우 구조요소 보다 지진 발생시 내진성능에 있어서 취약하다고 볼 수 있다. 스프링클러 배관 시스템과 같은 비구조적 구성요소의 손상으로 인해 지진 발생 및 이후에 상당한 경제적 손실이나 생명 손실을 초래 할 수 있다. 연구방법: 본 연구는 Triple Friction Pendulum Bearings (TPBs)을 설치한 건물 배관 시스템을 이용한 소방 배관 시스템의 내진성능평가를 제시한다. Kobe, Kocaeli, GyeongJu 지진을 고려하여 지반 운동의 불확실성을 고려하였다. 연구결과: 빌딩 시스템과 파이핑 시스템의 첫 번째 모드는 각각 약 5.8Hz와 약 2.742Hz로 나타났으며, 또한 TPBs 시스템이 적용된 배관 시스템의 최대 변위는 Kobe, Kocaeli 및 GyeongJu 지진의 경우 각각 49%, 14.4%, 21.5%가 감소한 것으로 나타났다. 결론: 따라서 건물 배관 시스템에서 지진 격리 시스템을 사용하면 지진이 심할 때 TPB가 없는 일반적으로 설치된 걸물 배관 시스템보다 지진 위험을 줄일 수 있다.

• Nuclear Engineering and Technology
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• 제31권3호
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• pp.267-275
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• 1999

This paper deals with the industrial problem of thermal striping damage on the French prototype fast breeder reactor, Phenix and it was studied in coordination with the research program of IAEA. The thermomechanical and fracture mechanics evaluation procedure of thermal striping damage on the tee-junction of the secondary piping using Green's function method and standard FEM is presented. The thermohydraulic(T/H) loading condition used in the present analysis is the random type thermal loads computed by T/H analysis on the turbulent mixing of the two flows with different temperatures. The thermomechanical fatigue damage was evaluated according to ASME code section 111 subsection NH. The results of the fatigue analysis showed that fatigue failure would occur at the welded joint within 90,000 hours of operation. The assessment for the fracture behavior of the welded joint showed that the crack would be initiated at an early stage in the operation. It took 42,698.9 hours for the crack to propagate up to 5 mm along the thickness direction. After then, however, the instability analysis, using tearing modulus, showed that the crack would be arrested, which was in agreement with the actual observation of the crack. An efficient analysis procedure using Green's function approach for the crack propagation problem under random type load was proposed in this study. The analysis results showed good agreement with those of the practical observations.

• Nuclear Engineering and Technology
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• 제40권7호
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• pp.607-614
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• 2008

Since pipes with wall-thinning defects can collapse at fluid pressure that are lower than expected, the collapse moment of wall-thinned pipes should be determined accurately for the safety of nuclear power plants. Wall-thinning defects, which are mostly found in pipe bends and elbows, are mainly caused by flow-accelerated corrosion. This lowers the failure pressure, load-carrying capacity, deformation ability, and fatigue resistance of pipe bends and elbows. This paper offers a support vector regression (SVR) model further enhanced with a fuzzy algorithm for calculation of the collapse moment and for evaluating the integrity of wall-thinned piping systems. The fuzzy support vector regression (FSVR) model is applied to numerical data obtained from finite element analyses of piping systems with wall-thinning defects. In this paper, three FSVR models are developed, respectively, for three data sets divided into extrados, intrados, and crown defects corresponding to three different defect locations. It is known that FSVR models are sufficiently accurate for an integrity evaluation of piping systems from laser or ultrasonic measurements of wall-thinning defects.

• 한국압력기기공학회 논문집
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• 제15권1호
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• pp.40-45
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• 2019

Unlike the welded pipes in the primary system of light water nuclear power plants being periodically inspected with in-Service inspection program, relatively small pipes with the outer diameter less than 2 inch have not been regularly inspected to date. However, after several failure reports on the occurrence of critical crack-like defects in small pipes, inspection for the small pipes has been more demanded because it could cause the provisional outage of nuclear power plants. Nevertheless, there's no particular method to examine the small pipes having access limitations for inspection due to various reasons; inaccessible area, excessive radiation exposure, hazardous surrounding, and etc. This study is to develop a reliable inspection technique using torsional and flexural modes of guided wave to detect defects that could occur in inaccessible area. The attribute of guided wave that can travel a long distance enables to inspect even isolated range of the pipe from accessible location. This paper presents a case study of the evaluation test on 3/4" small-bore pipes with guide wave method. The test result demonstrates the crack signal behavior and assures possibility to detect the crack signal in a flexural mode, which is clearly distinguishable from the symmetric structure signal in a torsional mode.

• 대한기계학회논문집A
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• 제31권3호
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• pp.304-312
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• 2007

The objective of this study is to develop a local failure criterion for a wall thinning defect of piping components. For this purpose, a series of tensile tests was performed using several types of simulated specimens with different stress states, including smooth round bar, notched round bar (five different notch radii), and grooved plate (three different groove radii). In addition, finite element (FE) simulations were performed on the simulated specimen tests and the results were compared with the test results. From the comparisons, the equivalent stress and strain corresponding to maximum load and final failure of notched specimens were proposed as failure criteria under tensile load. The criteria were verified by employing them to the estimation of failure of grooved plate specimens that simulate the wall thinning defect. It showed that the proposed criteria accurately estimate the maximum load and final failure of grooved plate specimen tests.

• 물과 미래
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• 제22권2호
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• pp.223-231
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• 1989

하천제방의 붕괴로부터 발생하는 제내지의 침수현상을 월류(overtopping), 제체의 붕괴(breaking), 파이핑(piping) 등의 세경우로 구분하고, 각 경우에 대하여 제내지로의 유입량을 예측하는 해석모형을 제시하였다. 가상수치실험을 통한 월류 및 제방제체의 붕괴에 의한 하도-제내지간의 흐름의 특징을 해석한 결과 붕괴부의 폭이 유출량에 미치는 영향이 가장 큰 것으로 나타났으며, 붕괴시간이나 유량계수 등의 영향을 작게 나타났다.

• 대한기계학회논문집A
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• 제29권5호
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• pp.734-742
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• 2005

The present study proposes the integrity evaluation model for a straight pipe with local wall thinning defect, which reflects the characteristics of training shape and loading condition in the Piping of nuclear power plant. For this purpose, a series of finite element analyses are performed under various defect geometries and loading conditions, and real pipe experiment data performed previously is employed. The model includes the effect of thinning length as well as thinning depth and width, and also it considers the combined loading effect between internal pressure and bending moment. The proposed model has been validated using the results of finite element analysis and pipe experiment data. The results indicate that the proposed model provides more reliable predictions of pipe failure than the current existing model, in terms of accuracy, consistency, and conservativeness of results.

• 한국압력기기공학회 논문집
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• 제16권1호
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• pp.56-64
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• 2020

The safety related heat exchangers have been evaluated for their performance during the operation of the nuclear power plant. The evaluation program for the safety related heat exchanger was developed in 2010 and used by KHNP based on EPRI TR-10739 algorithms. The spend fuel pool heat exchanger is one of the safety related heat exchanger in the nuclear power plant and also evaluated for their performance. Recently the performance evaluation for the spend fuel pool heat exchanger was not available because of the decreased heat in the spend fuel pool due to the long term overhaul. This paper analyzes the main cause of evaluation failure in the evaluation process and suggests the criteria for the heat exchanger performance evaluation during the long term overhaul.