Structural integrity assessment of the reactor pressure vessel under the pressurized thermal shock loading

Chen, Mingya; Lü, Feng; Wang, Rongshan; Ren, Aixia

doi:10.1016/j.nucengdes.2014.01.021

Cited by 35 publications

(8 citation statements)

References 11 publications

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“…In 2014, Chen et al [15], performed fracture mechanics analysis of the beltline region of a reactor pressure vessel under the event of pressurized thermal shock. The authors stated that analysis under such unanticipated events is the key element of the integrity evaluation.…”

Section: Literature Reviewmentioning

confidence: 99%

Fracture analysis of the set-in nozzle of a PWR reactor pressure vessel – Part 1: Determination of critical crack

Murtaza

Hyder

2018

Engineering Fracture Mechanics

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Section: Literature Reviewmentioning

confidence: 99%

Fracture analysis of the set-in nozzle of a PWR reactor pressure vessel – Part 1: Determination of critical crack

Murtaza

Hyder

2018

Engineering Fracture Mechanics

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“…Several studies on the integrity assessment of RPV subjected to PTS loadings are mainly on the PTS initiated by postulated accidents [3]. Chen et al have performed the structural integrity assessment of the RPV under the PTS loading initiated by a loss of coolant accident (LOCA) [7]. Also, Jhung et al have evaluated the structural integrity of RPV based on pressurized thermal shock caused by SBLOCA [8].…”

Section: Introductionmentioning

confidence: 99%

Integrity Evaluation of a Reactor Pressure Vessel Based on a Sequential Abaqus-FRANC3D Simulation Method

Annor-Nyarko

Xia

2021

Science and Technology of Nuclear Installations

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The safety-risk pressurized thermal shock (PTS) have on a reactor pressure vessel (RPV) is one of the most important studies for the lifetime ageing management of a reactor. Several studies have investigated PTS induced by postulated accidents and other anticipated transients. However, there is no study that analyzes the effect of PTS induced by one of the most frequent anticipated operational occurrences—inadvertent operation of the safety injection system. In this paper, a sequential Abaqus-FRANC3D simulation method is proposed to study the integrity status of an ageing pressurized water reactor subjected to PTS induced by inadvertent actuation of the safety injection system. A sequential thermal-mechanical coupling analysis is first performed using a three-dimensional reactor pressure vessel finite element model (3D-FEM). A linear elastic fracture mechanics submodel with a postulated semielliptical surface crack is then created from the 3D-FEM. Subsequently, the submodel is used to evaluate the stress intensity factors based on the M-integral approach coupled within the proposed simulation method. Finally, the stress intensity factors (SIFs) obtained using the proposed method are compared with the conventional extended finite element method approach, and the result shows a good agreement. The maximal thermomechanical stress concentration was observed at the inlet nozzle-inner wall intersection. In addition, The ASME fracture toughness of the reactor vessel’s steel compared with SIFs show that the PTS event and crack configuration analysed may not pose a risk to the integrity of the RPV. This work serves as a critical reference for the ageing management and fatigue life prediction of reactor pressure vessels.

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“…2 The failure criterion in the probabilistic assessment of RPV integrity employs conventional linear elastic fracture mechanics by comparison of crack driving forces (such as stress intensity factor K I ) calculated for assessed points along the crack front with its allowable value (fracture toughness K IC or K Ia ). 3,4 There are many uncertainties that should be considered including material chemical composition, mechanical properties, loads, crack sizes, and so on. Monte Carlo techniques are usually used to estimate the increase in failure probability as the vessel accumulates radiation damage over its operating life.…”

Section: Introductionmentioning

confidence: 99%

Weibull stress analysis for the corner crack in reactor pressure vessel nozzle

Jin

Wang

et al. 2019

Advances in Mechanical Engineering

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The nozzle region of a reactor pressure vessel experiences higher and more complex stresses than the remaining part of the reactor pressure vessel. If a corner crack is postulated in the nozzle region, it is necessary to consider the potential strong influence of constraint on fracture behavior due to inelastic deformations of crack tip. In accordance with the requirement of probabilistic fracture mechanics, Weibull stress in local approach to fracture is analyzed with the consideration of constraint effect. Conventional fracture analysis is also carried out using a three-dimensional crack model, and the fracture driving forces ( KI and J) and T-stress are obtained. Weibull stress along the crack tip is also computed by three-dimensional models. The modified boundary layer model with plastic correction is developed for a corner crack in the reactor pressure vessel nozzle. Under the J- T stress field from three-dimensional models, Weibull stress values obtained using the modified boundary layer model are compared and discussed with that by three-dimensional models. It is found that the modified boundary layer model can effectively predict the Weibull stress under the J- T stress field, which simplified the Weibull stress calculation process for complex structures.

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Structural integrity assessment of the reactor pressure vessel under the pressurized thermal shock loading

Cited by 35 publications

References 11 publications

Fracture analysis of the set-in nozzle of a PWR reactor pressure vessel – Part 1: Determination of critical crack

Fracture analysis of the set-in nozzle of a PWR reactor pressure vessel – Part 1: Determination of critical crack

Integrity Evaluation of a Reactor Pressure Vessel Based on a Sequential Abaqus-FRANC3D Simulation Method

Weibull stress analysis for the corner crack in reactor pressure vessel nozzle

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