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원주방향균열이 존재하는 원전 배관계통의 파괴거동에 관한 실험적 연구(I) - 직관부에서의 균열거동 평가 -
최영환,박윤원,Choi, Young-Hwan,Park, Youn-Won,Wilkowski, Gery 대한기계학회 1999 大韓機械學會論文集A Vol.23 No.7
The purpose of this study is to investigate experimentally the effects of both seismic loading and crack length on the fracture behavior of piping system with a circumferential crack in nuclear power plants. The experiments were performed using both large scale piping system facility and 4 points bending test machine under PWR operating conditions. The difference in the load carrying capacities between cracked piping and non-cracked piping was also investigated using the results from experiments and numerical calculations. The results obtained from the experiments and estimation are as follows : (1) The safety margin under seismic loading is larger than those under quasi static loading or simple cyclic loading. (2) There was no significant effect of crack length on tincture behavior of piping system with both a surface crack and a through-wall crack. (3) The load carrying capacity in cracked piping was reduced by factors of 7 to 46 compared to non-cracked piping.
고온관 누설에 의한 가압열충격 사고시 원자로 용기의 건전성 평가를 위한 결정론적 파괴역학 해석
이상민,최재붕,김영진,박윤원,정명조,Lee, Sang-Min,Choi, Jae-Boong,Kim, Young-Jin,Park, Youn-Won,Jhung, Myung-Jo 대한기계학회 2002 大韓機械學會論文集A Vol.26 No.11
In a nuclear power plant, reactor pressure vessel (RPV) is the primary pressure boundary component that must be protected against failure. The neutron irradiation on RPV in the beltline region, however, tends to cause localized damage accumulation, leading to crack initiation and propagation which raises RPV integrity issues. The objective of this paper is to estimate the integrity of RPV under hot leg leaking accident by applying the finite element analysis. In this paper, a parametric study was performed for various crack configurations based on 3-dimensional finite element models. The crack configuration, the crack orientation, the crack aspect ratio and the clad thickness were considered in the parametric study. The effect of these parameters on the maximum allowable nil-ductility transition reference temperature ($(RT_{NDT})$) was investigated on the basis of finite element analyses.
가동중 중수로 압력관의 외경과 두꼐 변화를 고려한 결함의 파손확률 예측
곽상록,이준성,김영진,박윤원,Kwak, Sang-Log,Lee, Joon-Seong,Kim, Young-Jin,Park, Youn-Won 대한기계학회 2002 大韓機械學會論文集A Vol.26 No.11
The pressure tube is a major component of the CANDU reactor, which supports nuclear fuel bundle and heavy water coolant. Pressure tubes are installed horizontally inside the reactor and only selected samples are periodically examined during in-service inspection. In this respect, a probabilistic safety assessment method is more appropriate fur the assessment of overall pressure tube safety. The failure behavior of CANDU pressure tubes, however, is governed by delayed hydride cracking which is the major difference from pipings and reactor pressure vessels. Since the delayed hydride cracking has more widely distributed governing parameters, it is impossible to apply a general PFM methodology directly. In this paper, a PFM methodology for the safety assessment of CANDU pressure tubes is introduced by applying Monte Carlo simulation in determining failure probability Initial hydrogen concentration, flaw shape and depth, axial and radial crack growth rate and fracture toughness were considered as probabilistic variables. Parametric study has been done under the base of pressure tube dimension and hydride precipitation temperature in calculating failure probability. Unstable fracture and plastic collapse are used for the failure assessment. The estimated failure probability showed about three-order difference with changing dimensions of pressure tube.
압력관의 확률론적평가에 타당한 파손평가선도 작성에 관한 연구
곽상록(Sang-log Kwak),왕종배(Jong-bae Wang),최영환(Young-hwan Choi),박윤원(Youn-won Park) 대한기계학회 2003 대한기계학회 춘추학술대회 Vol.2003 No.11
Pressure tubes are major component of nuclear reactor, but only selected samples are periodically<br/> examined due to numerous numbers of tubes. Current in-service inspection result show there is high<br/> probability of flaw existence at un-inspected pressure tube. Probabilistic analysis is applied in this study for<br/> the integrity assessment of un-inspected pressure tube. But all the current integrity evaluations procedures are<br/> based on conventional deterministic approaches. So many integrity evaluation parameters are not directly<br/> apply to probabilistic analysis. As a result of this study failure assessment diagram are proposed based on test<br/> data.
치수변화를 고려한 CANDU 압력관의 축방향 결함에 대한 파손확률 예측
곽상록(Sang-Log Kwak),이준성(Joon-Seong Lee),김영진(Young-Jin Kim),박윤원(Youn-Won Park) 대한기계학회 2002 대한기계학회 춘추학술대회 Vol.2002 No.3
The pressure tube is a major component of the CANDU reactor, which supports nuclear fuel bundle and heavy water coolant. Pressure tubes are installed horizontally inside the reactors and only selected samples are periodically examined during in-service inspection(ISI) due to numerous numbers of tubes. ISI results shows half of the pressure tubes have flaw, and axial flaw is dominant rather than circumferential flaw in pressure tube. In this respect, a probabilistic safety assessment method is more appropriate for the assessment of overall pressure tube safety. The failure behavior of CANDU pressure tubes, however, is governed by delayed hydride cracking which is not considered in the application of PFM on pipings and vessels. Since the delayed hydride cracking has more widely distributed governing parameters, it is impossible to apply a general PFM methodology directly. In this paper, a PFM methodology for the safety assessment of CANDU pressure tubes is introduced by applying Monte Carlo simulation in determining failure probability. Initial hydrogen concentration, flaw shape and depth, crack growth rate and fracture toughness were considered as probabilistic variables. Parametric study has been done under the base of pressure tube dimensions, which are derived from ISI data. Applying both unstable fracture condition and plastic collapse condition set failure criteria. The estimated failure probability showed about three-order difference with changing dimensions of pressure tube.
다중 관통균열이 존재하는 증기발생기 세관의 최적 국부파손모델 결정
문성인(Seong In Moon),김윤재(Yun Jae Kim),김영진(Young Jin Kim),박윤원(Youn Won Park),송명호(Myung Ho Song),이진호(Jin Ho Lee) 대한기계학회 2002 대한기계학회 춘추학술대회 Vol.2002 No.3
It is commonly required that steam generator tubes wall-thinned in excess of 40% should be plugged. However, the plugging criteria are known to be too conservative for some locations and types of defects and its application is confined to a single crack. In the previous study, a crack coalescence model applicable to steam generator tubes with two collinear axial through-wall cracks was proposed and a coalescence evaluation diagram was developed which can be used to determine whether the adjacent cracks detected by NDE coalesce or not. In this paper, a total of 9 local failure models including flow stress model, necking base model, stress base model, reaction force model and plastic zone contact model were introduced to determine the optimum local failure model. Plastic collapse tests and FEA using the plate with two collinear through-wall cracks were performed for the selection of the optimum local failure model. By comparing the test results with the prediction results obtained from local failure models, reaction model and plastic zone contact model were determined as the optimum local failure models.
PTS 사고하에서 원자로 압력용기 표면균열에 대한 구속효과의 정량화 : J-Q 접근법의 적용
김진수(Jin-Su Kim),최재붕(Jae-Boong Choi),김영진(Young-Jin Kim),박윤원(Youn-Won Park) 대한기계학회 2001 대한기계학회 춘추학술대회 Vol.2001 No.8
In recent years, the integrity of Reactor Pressure Vessel(RPV) under pressurized thermal shock (PTS) accident has been treated as one of the most critical issues. Under PTS condition, the combination of thermal and mechanical stress by steep temperature gradient and internal pressure causes considerably high tensile stress at the inside of RPV wall. As a result, cracks on inner surface of RPV may experience elastic-plastic behavior which can be characterized by J-integral. In such a case, however, J-integral may possibly lose its validity due to the constraint effect. The degree of constraint effect is influenced by the loading mode, crack geometry and material properties. In this paper, in order to investigate the effect of clad thickness and crack geometry on constraint effect, three dimensional finite element analyses were performed for various surface cracks. Total of 27 crack geometries were analyzed and results were presented by a two-parameter characterization based on the J-integral and the Q-stress.
인장하중이 작용하는 평판에 존재하는 반타원 표면균열의 J-적분 계산식
심도준(Do-Jun Shim),김윤재(Yun-Jae Kim),최재붕(Jae-Boong Choi),김영진(Young-Jin Kim),박윤원(Youn-Won Park) 대한기계학회 2001 대한기계학회 춘추학술대회 Vol.2001 No.3
This paper provides a simplified engineering J estimation method for semi-elliptical surface cracked plates in tension, based on the reference stress approach. Note that the essential element of the reference stress approach is the plastic limit load in the definition of the reference stress. However, for surface cracks, the definition of the limit load is ambiguous (local or global limit load), and thus the most relevant limit load (and thus reference stress) for the J estimation should be determined. In the present work, such limit load solution is found by comparing reference stress based J results with those from extensive 3-D finite element analyses. Validation of the proposed equation against FE J results based on actual experimental tensile data of a 304 stainless steel shows excellent agreements not only for the J values at the deepest point but also for those at an arbitrary point along the crack front, including at the surface point. Thus the present results provide a good engineering tool for elastic-plastic fracture analyses of surface cracked plates in tension.