Modelling of reactor pressure vessel subjected to pressurized thermal shock using 3D-XFEM (original) (raw)
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Engineering Fracture Mechanics, 2020
This paper presents the integrity analyses of a model reactor pressure vessel (RPV) subjected to pressurized thermal shocks (PTS). The analyses are performed with a one-way multi-step strategy that includes the thermo-hydraulics, thermo-mechanical and fracture mechanics analyses to simulate three hypothetical loss of coolant accidents (LOCA). The thermo-hydraulics analyses are performed with the system code TRACE and a three-dimensional (3D) model of the RPV, providing the input for the structural analyses with the finite element code ABAQUS. These employ a sequential use of global model (entire RPV) and submodels (a portion of the RPV containing the crack), where the eXtended Finite Element (XFEM) approach is employed to compute the stress intensity factor (SIF) of a postulated crack in the RPV wall. The results first present a verification of the multi-step strategy with the FAVOR code for uniform temperature distribution in the RPV wall. Under uniform temperature and pressure load, a significant effect of the nozzle geometry on the asymmetric stress distribution is demonstrated. The second part of the results shows that the stresses and the SIFs are also sensitive to the nonuniform temperatures due to the presence of the cooling plumes. It is also confirmed that the analyses with ABAQUS and FAVOR provide very similar results for the medium and large break LOCA transients. For the small break LOCA, the FAVOR code underestimates the SIFs due to the missing nozzle geometry in combination with system pressure. Finally, the paper corroborates that the use of TRACE and XFEM, within the one-way multi-step simulation strategy, reduces the computational costs and the number of assumptions and approximations needed for feasible and relivable 3D fracture mechanics analyses of the RPV with consideration of the cooling plume effect.
Procedia Structural Integrity, 2022
The reactor pressure vessel (RPV) is subjected to many cyclic loadings critical of which is the integrity risk posed by pressurized thermal shock (PTS) loadings induced by one of the most frequent anticipated operational occurrences-inadvertent operation of the safety injection system (SIS). In this paper, the fracture mechanics analysis for an ageing pressurized water reactor subjected to PTS induced by inadvertent actuation of the SIS is performed using a proposed simplified Abaqus-FRANC3D co-simulation method. A three-dimensional (3-D) finite element half-symmetric model of a typical pressured water reactor RPV is used to perform thermal-mechanical stress coupling analysis. 3-D fracture mechanic submodel with an assumed surface crack is created for the computation of the transient integrity parameter-stress intensity factor (SIF), using M-integral approach coupled in the proposed co-simulation method. Subsequently ASME method is used to evaluate the vessel's material fracture toughness. Finally, the SIFs obtained with the simplified co-simulation method is compared with the conventional virtual crack-closure technique (VCCT), and the result show good agreement. This work serves as a useful reference for fast crack propagation and life prediction analysis of ageing PWR RPVs.
Potential impact of enhanced fracture-toughness data on pressurized-thermal-shock analysis
1990
The Heavy Section Steel Technology (HSST) Program is involved with the generation of enhanced'' fracture-initiation toughness and fracture-arrest toughness data of prototypic nuclear reactor vessel steels. These two sets of data are enhanced because they have distinguishing characteristics that could potentially impact PWR pressure vessel integrity assessments for the pressurized-thermal shock (PTS) loading condition which is a major plant-life extension issue to be confronted in the 1990's. Currently, the HSST Program is planning experiments to verify and quantify, for A533B steel, the distinguishing characteristic of elevated initiation-fracture toughness for shallow flaws which has been observed for other steels. Deterministic and probabilistic fracture-mechanics analyses were performed to examine the influence of the enhanced initiation and arrest fracture-toughness data on the cleavage fracture response of a nuclear reactor pressure vessel subjected to PTS loadin...
Engineering Analysis with Boundary Elements, 2012
The failure probabilities of the reactor pressure vessel (RPV) for low temperature overpressurization (LTOP) and cool-down transients are calculated in this study. For the cooldown transient, a pressureetemperature limit curve is generated in accordance with Section XI, Appendix G of the American Society of Mechanical Engineers (ASME) code, from which safety margin factors are deliberately removed for the probabilistic fracture mechanics analysis. Then, sensitivity analyses are conducted to understand the effects of some input parameters. For the LTOP transient, the failure of the RPV mostly occurs during the period of the abrupt pressure rise. For the cool-down transient, the decrease of the fracture toughness with temperature and time plays a main role in RPV failure at the end of the cool-down process. As expected, the failure probability increases with increasing fluence, Cu and Ni contents, and initial reference temperature-nil ductility transition (RT NDT). The effect of warm prestressing on the vessel failure probability for LTOP is not significant because most of the failures happen before the stress intensity factor reaches the peak value while its effect reduces the failure probability by more than one order of magnitude for the cool-down transient.
Fracture toughness prediction of a valve body: Numerical analysis
Engineering Failure Analysis, 2010
The crack repair technique by ''crack grinding", due to its efficiency, has been widely used to extend the service life of the cracked structural components. The ''crack grinding" technique is carried out by removing the materials which contains the crack. So it is of great importance to study the effects of this repair technique. In this paper, the finite element method is used to analyse the behavior of repaired cracks in mode I by computing the stress-intensity factors at the crack tip. 3D numerical simulation was carried out to model the steam governor valve in a coal-fired thermal power plant with a semi-elliptic crack. Since no extra auxiliary solutions are required for the computation, the proposed method appears to be visual and can be used for evaluation of the associated stress-intensity factors. The plot of the stress-intensity factors according to the crack length in mode I, shows that the stress-intensity factor exhibits an asymptotic behavior as the crack length increases. It's to be noted that a transition zone allows connecting the refined specific crack tip mesh to the global structural discretization. So this technique can be used for complex industrial structures.
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA
Reactor Pressure Vessel (RPV) wall is an important component in the Nuclear Power Plant (NPP). During reactor operation, RPV is subjected to high temperature, pressure, and neutron exposure. This condition could lead to RPV structure failure. In order to assure the integrity of RPV during the reactor lifetime, it is mandatory to perform a structural integrity assessment of RPV by evaluating postulated crack in RPV. In the previous study, the crack has evaluated in 2-D. However, 3-D analysis of semi-elliptic crack shape in the surface of the thick plate for RPV wall using SA 508 Steel is yet to be analyzed. The objective of this study is to analyze and modeling the evaluation in variation crack ratio with some load stress in 3-D. The Stress Intensity Factor (SIF) and J-integral are used as crack parameter. The J-Integral were calculated using MSC MARC MENTAT based on Finite Element Method (FEM) for obtaining the SIF value. The inputs are a crack ratio, load stress, material property,...
Unstable crack propagation under severe accident scenario conditions in a pressurized water reactor
2009
In the case of a severe accident scenario of a pressurized water reactor which includes cracking of the vessel bottom head, it is crucial to predict the leak rate and hence the crack size for the ex-vessel accident management. We present an experimental framework to analyze the crack propagation under such severe conditions for different 16MND5 French nuclear steel grades. An original experimental setup has been designed in order to perform bi-axial tests (tensile load independent of internal pressure) at high temperatures (1180K-1280K) on tubular test specimens. The temperature loading and the mechanical loading can be set to reproduce the stress distribution of the hemispherical vessel bottom head submitted to an internal pressure. Moreover, the test was designed to be easily transposable to the real structure in terms of crack propagation and depressurization thanks to an energy based scaling methodology. We observed the crack initiation and propagation with two high speed digital cameras. Force, internal pressure, displacement and temperature fields were also measured and synchronized with the optical measurements. The different creep stages are observed and characterized. The crack propagation and opening history have been measured. During crack initiation and propagation stages, the depressurization can be correlated with the crack geometry. Finally, the setup has been designed in order to validate future numerical analysis.
Fracture mechanical evaluation of an in-vessel melt retention scenario
Annals of Nuclear Energy, 2008
This paper presents methods to compute J-integral values for cracks in two-and three-dimensional thermo-mechanical loaded structures using the finite element code ANSYS. The developed methods are used to evaluate the behavior of a crack on the outside of an emergency cooled reactor pressure vessel (RPV) during a severe core melt down accident. It will be shown, that water cooling of the outer surface of a RPV during a core melt down accident can prevent vessel failure due to creep and ductile rupture. Further on, we present J-integral values for an assumed crack at the outside of the lower plenum of the RPV, at its most stressed location for an emergency cooling (thermal shock) scenario.