• 한국안전학회지
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• 제21권1호
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• pp.28-34
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• 2006

A procedure of estimating failure probability is demonstrated for a pressurized pipe of CrMo steel used at $538^{\circ}C$. Probabilistic fracture mechanics were employed considering variations of pressure loading, material properties and geometry. Probability density functions of major material variables were determined by statistical analyses of implemented data obtained by previous experiments. Distributions of the major variables were reflected in Monte Carlo simulation and failure probability as a function of operating time was determined. The creep crack growth life assessed by conventional deterministic approach was shown to be conservative compared with those obtained by probabilistic one. Sensitivity analysis for each input variable was also conducted to understand the most influencing variables to the residual life analysis. Internal pressure, creep crack growth coefficient and creep coefficient were more sensitive to failure probability than other variables.

• 한국압력기기공학회 논문집
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• 제15권2호
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• pp.40-49
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• 2019

In this paper, failure probabilities of the OPR1000 reactor vessel under pressurized thermal shock (PTS) were estimated using the probabilistic fracture mechanics code, R-PIE. Input variables of initial crack distribution, crack size, copper contents, and upper shelf toughness were selected for the sensitivity analyses. A wide range of the input data were considered. Through-wall cracking frequencies determined by the product of the vessel failure probability and the corresponding occurrence frequency of the transient were also compared to the acceptance criterion. The results showed that transient history had the most significant impact on the vessel failure probability. Moreover, conservative assumptions resulted in extremely high through-wall cracking frequencies.

• 전산구조공학
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• 제11권1호
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• pp.153-160
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• 1998

본 연구에서는 가압열충격 사고로 소형 냉각재 상실사고를 가정하여 냉각재의 온도와 압력의 이력으로 부터 용기 벽의 온도분포를 구하고, 이로 부터 열응력과 압응력을 해석적으로 구하였다. 또 균열 선단에서의 응력강도계수와 파괴인성치를 ASME코드의 방법을 이용하여 구하였고, 이들을 시간에 따라 비교하여 균열의 진전여부를 평가하였다. 원자로 용기 벽에 존재하는 여러 형태의 균열이 견딜 수 있는 최대 기준무연성천이온도를 결정하였으며 평가 결과에 대하여 고찰하였다.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2004년도 추계학술대회
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• pp.90-95
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• 2004

The probabilistic fracture analysis is used to determine the effects of uncertainties involved in material properties, location and size of flaws, etc, which can not be addressed using a deterministic approach. In this paper the probabilistic fracture analysis is applied for evaluating the RV(Reactor Vessel) under PTS(Pressurised Thermal Shock). A semi-elliptical axial crack is assumed in the inside surface of RV. The selected random parameters are initial crack depth, neutron fluence, chemical composition of material (copper, nickel and phosphorous) and $RT_{NDT}$. The deterministically calculated $K_I$ and crack tip temperature are used for the probabilistic calculation. Using Monte Carlo simulation, the crack initiation probability for fixed flaw and PNNL(Pacific Northwest National Laboratory) flaw distribution is calculated. As the results show initiation probability of fixed flaw is much higher than that of PNNL distribution, the postulated crack sizes of 1/10t in this paper and 1/4t of ASME are evaluated to be very conservative.

• 대한기계학회논문집A
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• 제25권7호
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• pp.1139-1146
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• 2001

The reactor pressure vessel(RPV) is usually cladded with stainless steel to prevent corrosion and radiation embrittlement, and a number of subclad cracks have been found during an in-service-inspection. These subclad cracks should be assured for a safe operation under normal conditions and faulted conditions such as pressurized thermal shock(PTS). Currently available integrity assessment procedure for an RPV, ASME Code Sec. XI, are built on the basis of linear fracture mechanics (LEFM). In PTS condition, however, thermal stress and mechanical stress give rise to high tensile stress at the cladding and elastic-plastic behavior is expected in this area. Therfore, ASME Code Sec. XI is overly conservative in assessing the structural integrity under PTS condition. In this paper, the fracture parameter (stress intensity factor, K, and RT(sub)NDT) from elastic analysis using ASME Sec. XI and finite element method were validated against 3-D elastic-plastic finite element analyses. The difference between elastic and elastic-plastic analysis became significant with increasing crack depth. Therfore, it is recommended to perform elastic-plastic analysis for the accurate assessment of subclad cracks under TPS which causes plastic deformation at the cladding.

• 대한기계학회논문집A
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• 제21권7호
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• pp.1089-1096
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• 1997

In this paper, Rancho Seco transient which is reported as a typical pressurized thermal shock event is postulated to be occuring in the Kori unit 1 plant, the oldest nuclear power plant in Korea. For the given material properties, transient history such as temperature and pressure, and postulated flaw, the stress distribution is obtained to calculate stress intensities for a wide range of assumed crack sizes. The stress intensities are compared with the fracture toughness, which is determined using the material properties and the distribution of the nil ductility transition temperature, to determine if cracking is expected to occur during the transient. The allowable operating year for the transient is determined and the evaluation results are discussed.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2000년도 춘계학술대회논문집A
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• pp.908-913
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• 2000

In the process of integrity evaluation for nuclear power plant components, a series of fracture mechanics evaluation on surface cracks in reactor pressure vessel(RPV) must be conducted. These fracture mechanics evaluations are based on stress intensity factor, K. However, under pressurized thermal shock(PTS) conditions, 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. Besides, the internal pressure during the normal operation produces high tensile stress at the RPV wall. As a result cracks on inner surface of RPVs may experience elastic-plastic behavior which can be explained with J-integral. In such a case, however, J-integral may possibly lose its validity due to constraint effect. In this paper, in order to verify the suitability of J-integral, two dimensional finite element analyses were applied for various surface crack. Total of 18 crack geometries were analyzed, and Q stresses were obtained by comparing resulting HRR stress distribution with corresponding actual stress distributions. In conclusion, HRR stress fields were found to overestimate the actual crack-tin stress field due to constraint effect.

• 대한기계학회논문집A
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• 제26권1호
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• pp.105-112
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• 2002

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 vapidity 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 f-stress.

• Nuclear Engineering and Technology
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• 제54권5호
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• pp.1726-1746
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• 2022

This paper presents fracture mechanics analysis results for various cracks located at pressurized water reactor pressure vessel nozzle crotch corners taking into consideration constraint effect. Technical documents such as the ASME B&PV Code, Sec.XI were reviewed and then a fracture mechanics analysis procedure was proposed for structural integrity assessment of various nozzle crotch corner cracks under normal operation conditions considering the constraint effect. Linear elastic fracture mechanics analysis was performed by conducting finite element analysis with the proposed analysis procedure. Based on the evaluation results, elastic-plastic fracture mechanics analysis taking into account the constraint effect was performed only for the axial surface crack of the reactor pressure vessel outlet nozzle with cladding. The fracture mechanics analysis result shows that only the axial surface crack in the reactor pressure vessel outlet nozzle has the stress intensity factor exceeding the low bound of upper-shelf fracture toughness irrespectively of considering the constraint effect. It is confirmed that the J-integral for the axial crack of the outlet nozzle does not exceed the ductile crack initiation toughness. Hence, it can be ensured that the structural integrity of all the cracks is maintained during the normal operation.

• Nuclear Engineering and Technology
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• 제39권5호
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• pp.647-654
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• 2007

Postulation of flaws, one of the most important areas in RPV integrity assessment, significantly affects the results. In the present work, several parameters, such as orientation, underclad vs. surface cracking, crack depth and shape, etc., are postulated and parametric studies are performed to investigate the influence of the flaw parameters on the structural integrity assessment of the reactor pressure vessel during pressurized thermal shock. The influence of individual parameters describing the crack is evaluated based on sensitivity study results.