The deterministic structural integrity assessment of reactor pressure vessels under pressurized thermal shock loading

Chen, Mingya; Lü, Feng; Wang, Rongshan; Huang, Ping; Liu, Xiangbin; Zhang, Guodong; Xu, Chaoliang

doi:10.1016/j.nucengdes.2015.03.008

Cited by 13 publications

(4 citation statements)

References 8 publications

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“…The elastic-plastic fracture parameter results shown in Figure 11 demonstrate good agreement with previous analyses of NESC-1 and correctly predict the location of fracture initiation that was observed experimentally. The occurrence of the critical location for fracture initiation during PTS at the HAZ/base interface has recently been reported for a similar case by Chen et al [31].…”

Section: Modelling Of Ptssupporting

confidence: 68%

Parametric design of scaled-down pressurised thermal shock test specimens using inelastic analysis

Coules

Orrock²,

Truman³

2017

Engineering Fracture Mechanics

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General rightsThis document is made available in accordance with publisher policies. Please cite only the published version using the reference above. AbstractPressurised thermal shock tests that simulate fracture conditions relevant to nuclear reactor pressure vessels are difficult to perform and require very large specimens. This study examines the feasibility of scaled-down specimens to allow easier testing. A series of finite element models was used to design scaled-down specimens that produce a crack-driving force, crack tip temperature trajectory and constraint conditions very similar to those which occurred in a previous large-scale spinning cylinder experiment known as NESC-1. It is shown that equivalent conditions can be achieved in much smaller specimens than NESC-1 using a practical set of testing parameters.Keywords: Thermal shock, finite element analysis, pressure vessel, spinning cylinder, J-integral Graphical abstractHighlights  Thermal shock tests for nuclear reactor pressure vessels can be miniaturised.  Models show that scaled-down tests can give equivalent fracture initiation conditions.  Specimens equivalent to the NESC-1 thermal shock test have been designed.

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Section: Modelling Of Ptssupporting

confidence: 68%

Parametric design of scaled-down pressurised thermal shock test specimens using inelastic analysis

Coules

Orrock²,

Truman³

2017

Engineering Fracture Mechanics

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“…The results show that the crack depth, crack type, plastic effect, and cladding thickness change the safety margin (SM) significantly, and the SM at the deepest point of the crack is not always smaller than that of the surface point, indicating that both the deepest and surface points of the crack front should be considered. The deterministic structural integrity assessment of the reactor pressure vessels was carried out by Chen under pressurized thermal shock loading (Chen et al, 2015). The results have shown that the critical part along the crack front is always the clad-based metal interface point rather than the deepest point for either the crack initiation assessment or the crack arrest assessment under the thermal load.…”

Section: Introductionmentioning

confidence: 99%

Study on Thermal Shock Failure Characteristics of RPV During IVR System Start-Up for Typical Advanced PWR

et al. 2022

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When the large pressurized water reactor in-vessel retention (IVR) system is put into operation, the outer wall of the reactor pressure vessel (RPV) will experience severe temperature fluctuations and be subjected to high internal pressure loads at the same time. In order to ensure the structural integrity of the RPV under such conditions, first, through the severe accident system program, the case that has the greatest impact on the pressure-bearing thermal shock of the outer wall of the reactor pressure vessel was selected under the station blackout (SBO) accident. Then, based on this case, the fracture mechanics finite element method was used to calculate and evaluate the pressure thermal shock (PTS) of the RPV, and the final crack size at the end of the life of the core barrel and the lower head was obtained by fatigue crack expansion calculation. The maximum ratio of the stress intensity factor correction value and corresponding limit value under the PTS transient load is about 0.874, which meets the requirements of RCC-M specification. The results of the study indicate that the RPV will not experience fracture failure when the IVR system is put into service during the station blackout accident for the HPR1000 nuclear power plant.

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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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The deterministic structural integrity assessment of reactor pressure vessels under pressurized thermal shock loading

Cited by 13 publications

References 8 publications

Parametric design of scaled-down pressurised thermal shock test specimens using inelastic analysis

Parametric design of scaled-down pressurised thermal shock test specimens using inelastic analysis

Study on Thermal Shock Failure Characteristics of RPV During IVR System Start-Up for Typical Advanced PWR

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

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