• Nuclear Engineering and Technology
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• v.32 no.2
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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.

• Proceedings of the KSME Conference
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• 2000.11b
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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.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.27 no.6
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• pp.987-996
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• 2003

The structural integrity of the reactor vessel with the approaching end of life must be assured for pressurized thermal shock. The regulation specifies the screening criteria for this and requires that specific analysis be performed for the reactor vessel which is anticipated to exceed the screening criteria at the end of plant life. In case the screening criteria is exceeded by the deterministic analysis, probabilistic analysis must be performed to show that failure probability Is within the limit. In this study, probabilistic fracture mechanics analysis of the reactor vessel for pressurized thermal shock is performed and the effects of residual stress and master curve on the failure probability are investigated.

• Proceedings of the KSME Conference
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• 2008.11a
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• pp.727-728
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• 2008

The objective of this study is to evaluate the integrity for a reactor pressure vessel under the pressurized thermal shock by applying the probability fracture mechanics. A semi-elliptical axial crack is assumed to be in the beltline region of the reactor pressure vessel. The selected random variables are the neutron fluence on the vessel inside surface, the content of copper, nickel, and phosphorus in the reactor pressure vessel material, and initial RTNDT. The probabilistic integrity analysis was performed using the Monte Carlo simulation.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.15 no.2
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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.

• Proceedings of the Computational Structural Engineering Institute Conference
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• 2004.04a
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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.

• Structural Engineering and Mechanics
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• v.37 no.3
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• pp.257-265
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• 2011

Plant-specific analyses of 5 types of domestic reactors in Korea are performed to assure the structural integrity of the reactor pressure vessel (RPV) during transients which are expected to initiate pressurized thermal shock (PTS) events. The failure probability of the RPV due to PTS is obtained by performing probabilistic fracture mechanics analysis. The through-wall cracking frequency is calculated and compared to the acceptance criterion. Considering the fluence at the end of life expected by surveillance test, the sufficient safety margin is expected for the structural integrity of all reactor pressure vessels except for the oldest one during the pressurized thermal shock events. If the flaw with aspect ratio of 1/12 is considered to eliminate the conservatism, the acceptance criteria is not exceeded for all plants until the fluence level of $8{\times}10^{19}\;n/cm^2$, generating sufficient margin beyond the design life.

• Nuclear Engineering and Technology
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• v.33 no.2
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• pp.132-144
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• 2001

Fracture mechanics is one of the major areas of the pressurized thermal shock (PTS) evaluation. To evaluate the reactor pressure vessel integrity associated with PTS, PFM methodology demands precise calculation of temperature, stress, and stress intensity factor for the variety of PTS transients. However, the existence of stainless steel cladding, with different thermal, physical, and mechanical property, at the inner surface of reactor pressure vessel complicates the fracture mechanics analysis. In this paper, treatment schemes to evaluate stress and resulting stress intensity factor for RPV with stainless steel clad are introduced. For a reference transient, the effects of clad thermal conductivity and thermal expansion coefficients on deterministic fracture mechanics analysis are examined.

• Computational Structural Engineering
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• v.11 no.1
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• pp.153-160
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• 1998

A small break loss of coolant accident is postulated as a pressurized thermal shock accident in this study. From the temperature and pressure histories of coolant, distributions of the temperature and stress in a vessel wall are analytically calculated. The stress intensity factor and fracture toughness of the vessel wall are determined at the crack tip using the ASME code method and they are compared to check if cracking is expected to occur during the transient postulated. The maximum allowable reference nil-ductility transition temperatures are determined for various crack sizes and the results are discussed.

• Structural Engineering and Mechanics
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• v.8 no.1
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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.