• Nuclear Engineering and Technology
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• 제52권11호
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• pp.2638-2651
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• 2020

Nozzle corner cracks present at the intersection of reactor pressure vessels (RPVs) and inlet or outlet nozzles have been a persistent problem for a number of years. The fracture analysis of such nozzle corner cracks is very important and critical for the efficient design and assessment of the structural integrity of RPVs. This paper aims to perform an engineering critical assessment of RPVs with nozzle corner cracks subjected to several transients accompanied by pressurized thermal shocks. The critical crack size of the RPV model with nozzle corner cracks under transient loading is evaluated on failure assessment curve. In particular, the influence of cladding on the crack initiation of nozzle corner crack under thermal transients is studied. The influence of primary internal pressure and secondary thermal stress on the stress field at nozzle corner and SIF at crack front is analyzed. Finally, the influence of different crack size and crack shape on the final critical crack size is analyzed.

• 대한기계학회논문집A
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• 제25권2호
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• pp.275-282
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• 2001

The reactor pressure vessel is usually cladded with stainless steel to prevent corrosion and radiation embrittlement. Number of subclad cracks may be found during an in-service-inspection due to the presence of cladding. It is specified, in ASME Sec. XI, that a subclad crack is characterized as a surface crack when the thickness of the clad is less than 40% of the crack depth. This condition is provided to keep the crack integrity evaluation conservative. In order to refine the fracture assessment procedures for such subclad cracks under a pressurized thermal shock condition, three dimensional finite element analyses are applied for various subclad cracks existing under cladding. A total of 36 crack geometries are analyzed, and the results are compared with those for surface cracks. The resulting stress intensity factors for subclad cracks are 6 to 44% less than those for surface cracks. It is proven that the flaw characterization condition as specified in ASME Sec. XI can be overly conservative for some subclad cracks.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 춘계학술대회논문집A
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• pp.441-446
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• 2001

In the integral type reactor, SMART, all the major components such as steam generators, pressurizer and pumps are located inside the single reactor pressure vessel. The objective of this study is to evaluate the structural integrity for RPV of SMART under the postulated pressurized thermal shock by applying the finite element analysis. Input data for the finite element analysis were generated using the commercial code I-DEAS, and the fracture mechanics analysis was performed using the ABAQUS. The crack configurations, the crack aspect ratio and the clad thickness were considered in the parametric study. The effects of these parameters on the reference nil-ductility transition temperature were also investigated.

• Structural Engineering and Mechanics
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• 제8권1호
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• pp.103-118
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• 1999

The structural integrity of the reactor pressure vessel under pressurized thermal shock (PTS) is evaluated in this study. For given material properties and transient histories such as temperature and pressure, the stress distribution is found and stress intensity factors are obtained for a wide range of crack sizes. The stress intensity factors are compared with the fracture toughness to check if cracking is expected to occur during the transient. A round robin problem of the PTS during a small break loss of coolant transient has been analyzed as a part of the international comparative assessment study, and the evaluation results are discussed. The maximum allowable nil-ductility transition temperatures are determined for various crack sizes.

• Nuclear Engineering and Technology
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• 제32권2호
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• pp.99-107
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• 2000

A probabilistic integrity analysis method is presented for a reactor vessel under pressurized thermal shock(PTS) based on Monte Carlo simulation. This method can be applied to the structural integrity assessment of a reactor vessel subjected to pressurized thermal shock where the coolant temperature transient cannot be expressed explicitly as a time function. An axially or circumferentially oriented infinite length surface crack is assumed to be in the beltline weld region of the rector vessel's inside surface. The random variables are the initial crack depth, neutron fluence on the vessel's inside surface, the copper and nickel content of the vessel materials, R $T_{NDT}$ , $K_{IC}$ , and K/aub la/. The reliability of a sample reactor vessel under PTS is assessed quantitatively and the influence of the amount of neutron fluence is also examined by applying the present method.sent method.

• 전산구조공학
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• 제9권2호
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• pp.153-161
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• 1996

본 논문에서는 가압열충격의 파괴역학적 해석에 필요한 이론을 조사하였고 원자로용기의 구조건전성을 평가하기 위하여 해석과정을 전산화하였다. 우선 사고 transient에 대하여 원자로용기내의 압력과 주입되는 냉각재의 온도변화가 주어지면 이들로 부터 시간에 따른 용기에서의 온도와 응력분포를 구하고, 중성자 조사량과 용기 재질의 화학성분으로 부터 기준무연성천이온도의 분포가 구해지며 이로부터 파괴인성치 K/sub IA/와 K/sub IC/의 분포가 얻어진다. 또한 응력분포로 부터 균열의 크기 및 형상에 따라 응력확대계수 K/sub I/이 구해지므로 이를 K/sub IA/및 K/sub IC/와 비교함으로써 균열의 성장거동을 예측할 수 있다. 지금까지 보고된 가압열충격을 유발할 수 있는 대표적인 사고 transient가 국내 발전소에 발생할 경우를 가정하여 해석을 수행하였고 그 결과에 대하여 검토하였다.

• 한국전산구조공학회:학술대회논문집
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• 한국전산구조공학회 2004년도 봄 학술발표회 논문집
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• pp.309-316
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• 2004

A probabilistic structural integrity assessment is performed for a reactor pressure vessel under PTS(Pressurized Thermal Shock). A semi-elliptical finite axial crack is assumed to he in the beltline region(either base metal or weld meta)1 of the reactor vessel inside surface. The selected random variables are initial crack depth, neutron fluence on the vessel inside surface, copper, nickel, and phosphorus content of the vessel material, and RT/sub NDT/. The probabilities of crack initiation or vessel failure where the crack is propagated through vessel wall are calculated. The probabilities obtained with random crack size are compared to these obtained with deterministic us. Since the failure function cannot to explicitly by selected by selected random variables, Monte Carlo Simulation is applied to perform probabilistic analysis The influence of the amount of neutron fluence is also examined to assess the structural reliability for vessel life time.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2000년도 추계학술대회논문집A
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• pp.286-291
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• 2000

The reactor pressure vessel is usually cladded with stainless steel to prevent corrosion and radiation embrittlement, and number of subclad cracks have been found during an in-service-inspection. Therefore assessment for subclad cracks should be made for normal operating conditions and faulted conditions such as PTS. Thus, in order to find the optimum fracture assessment procedures for subclad cracks under a pressurized thermal shock condition, in this paper, three different analyses were performed, ASME Sec. XI code analysis, an LEFM(Liner elastic fracture mechanics) analysis and an EPFM(Elastic plastic fracture mechanics) analysis. The stress intensity factor and the Maximum $RT_{NDT}$ were used for characterizing. Analysis based on ASME Sec. XI code does not completely consider the actual stress distribution of the crack surface, so the resulting Maximum allowable $RT_{NDTS}$ can be non-conservative, especially for deep cracks. LEFM analysis, which does not consider elastic-plastic behavior of the clad material, is much more non-conservative than EPFM analysis. Therefore, It is necessary to perform EPFM analysis for the assessment of subclad cracks under PTS.

• Nuclear Engineering and Technology
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• 제49권4호
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• pp.752-759
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• 2017

The fatigue crack growth behavior of Type 347 (S347) and Type 347N (S347N) stainless steel was evaluated under the operating conditions of a pressurized water reactor (PWR). These two materials showed different fatigue crack growth rates (FCGRs) according to the changes in dissolved oxygen content and frequency. Under the simulated PWR conditions for normal operation, the FCGR of S347N was lower than that of S347 and insensitive to the changes in PWR water conditions. The higher yield strength and better corrosion resistance of the nitrogen-alloyed Type 347 stainless steel might be a main cause of slower FCGR and more stable properties against changes in environmental conditions.

• Nuclear Engineering and Technology
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• 제30권5호
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• pp.470-484
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• 1998

An analysis program for the evaluation of pressure vessel integrity under pressurized thermal shock (PTS) is developed. For given material properties and transient history such as temperature and pressure, the stress distribution is calculated and then stress intensity factors are obtained for a wide range of crack sizes. The stress intensity factors are compared with the fracture toughness to check if cracking is expected to occur during the transient. Using this program a round robin problem of PTS during a small break loss of coolant transient has been analyzed as a part of the international comparative assessment study. The allowable maximum reference nil-ductility transition temperatures are determined for various crack sizes.