• 한국안전학회지
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• 제16권4호
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• pp.29-38
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• 2001

The purpose of this paper is to evaluate the structural integrity of a reactor pressure vessel subjected to the pressurized thermal shock(PTS) during the transient events, such as main steam line break(MSLB) and small break loss of coolant accident(SBLOCA). For postulated surface or subsurface cracks, variation curves of stress intensity factor are obtained by using the three different methods, including ASME section XI code anlysis, the finite element alternating method and the finite element method. From the stress intensity factor curves, the maximum allowable nil-ductility transition temperatures(RT/NDT/) are determined by the tangent criterion and the maximum criterion for various crack configurations and two initial transient events. As a result of the analysis, it is noted that axial cracks have smaller maximum allowable RT$_{NDT}$ values than same-sized circumferential cracks for both the transient events in the case of the tangent criterion. Axial cracks have smaller RT$_{NDT}$ values than same-sized circumferential cracks for MSLB and circumferential cracks have smaller values than axial cracks for SBLOCA in the case of the maximum criterion.

• 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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• 제45권3호
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• pp.277-294
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• 2013

In the early 1980s, attention focused on the possibility that pressurized thermal shock (PTS) events could challenge the integrity of a nuclear reactor pressure vessel (RPV) because operational experience suggested that overcooling events, while not common, did occur, and because the results of in-reactor materials surveillance programs showed that RPV steels and welds, particularly those having high copper content, experience a loss of toughness with time due to neutron irradiation embrittlement. These recognitions motivated analysis of PTS and the development of toughness limits for safe operation. It is now widely recognized that state of knowledge and data limitations from this time necessitated conservative treatment of several key parameters and models used in the probabilistic calculations that provided the technical of the PTS Rule, 10 CFR 50.61. To remove the unnecessary burden imposed by these conservatisms, and to improve the NRC's efficiency in processing exemption and license exemption requests, the NRC undertook the PTS re-evaluation project. This paper provides a synopsis of the results of that project, and the resulting Alternate PTS rule, 10 CFR 50.61a.

• 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.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2000년도 추계학술대회논문집B
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• pp.228-233
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• 2000

The Korean Next Generation Reactor(KNGR) is a Pressurized Water Reactor adopting direct vessel injection(DVI) to optimize the performance of emergency core cooling system(ECCS). In a certain accident, however, pressurized thermal shock(PTS) of the vessel due to the sudden contact with the injected cold water is expected. In this paper, an accident of Main Steam Line Break(MSLB) has been numerically investigated with direct vessel injections and an increased volume flow rate in some cold legs. Using FLUENT code, temperature distributions of the fluid in the downcomer and of reactor vessel including the core region have been calculated, together with the distribution of convective heat transfer coefficient(CHTC) at the cladding surface of the reactor vessel. The result shows that some parts of the core region of the reactor vessel have higher temperature gradient expressing higher thermal stress.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1999년도 춘계학술발표회요약집
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• pp.120-120
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• 1999

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.635-640
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• 1998

Thermal fluid mixing behavior during a postulated medium-size hot leg break loss of coolant accident is analyzed for the international comparative assessment study on pressurized thermal shock (PTS-ICAS) proposed by OECD-NEA. The applicability of RELAP5 code to analyze thermal fluid mixing behavior is evaluated through a simple modeling relevant to the problem constraints. Based on the calculation result, the onset of Thermal stratification is investigated using Theofanous's empirical correlation. Sensitivity calculations using a fine node model and crossflow model are also performed to evaluate the modeling capability on multi-dimensional characteristics related to thermal fluid mixing.

• 대한기계학회논문집A
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• 제28권6호
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• pp.693-700
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• 2004

The probabilistic fracture mechanics(PFM) is a useful analytical tool to assess the integrity of reactor pressure vessel(RPV) at the event of pressurized thermal shock(PTS). In PFM, the probabilities of flaw initiation and propagation are estimated by comparing the applied stress intensity factor with the fracture toughness calculated by the simulation of various stochastic variables. It is known that the results of PFM analyses are dependent on the choice of the stochastic parameters and assumptions. Of the various variables and assumptions, we investigated the effects of the RT$_{NDT}$ shift equations, fracture toughness curves, and flaw distributions on the PFM results for the three PTS transients. The results showed that the combined effects of the RT$_{NDT}$ shift equations and fracture toughness curves are complicated and dependent on the characteristics of the transients, the chemistry of the materials, the fast neutron fluence, and so on.

• 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.

• 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.