Fracture Mechanics Analysis of Reactor Pressure Vessel Under Pressurized Thermal Shock-The Effect of Elastic-Plastic Behavior and Stainless Steel Cladding-

2002 ◽  
Vol 26 (1) ◽  
pp. 39-47
Author(s):  
Jae-Hwang Ju ◽  
Gi-Ju Gang ◽  
Myeong-Jo Jeong
Keyword(s):  
Stainless Steel ◽  
Fracture Mechanics ◽  
Thermal Shock ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Plastic Behavior ◽  
Elastic Plastic ◽  
Pressurized Thermal Shock ◽  
Mechanics Analysis ◽  
Reactor Pressure

Fracture-Mechanics Analysis of the Reactor Pressure Vessel Beznau 1 Based on Thermal-Hydraulics Input Data From KWU-MIX and CFD Analyses

2016 ◽  
Author(s):  
Tomas Nicak ◽  
Richard Trewin ◽  
Elisabeth Keim ◽  
Ingo Cremer ◽  
Sebastien Blasset ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Thermal Shock ◽  
Pressure Vessel ◽  
Brittle Failure ◽  
Input Data ◽  
Cold Water ◽  
Reactor Pressure Vessel ◽  
Mechanics Analysis ◽  
Cfd Analyses ◽  
Reactor Pressure

The integrity of a reactor pressure vessel (RPV) has to be ensured throughout its entire life in accordance with the applicable regulations. Typically an assessment of the RPV against brittle failure needs to be conducted by taking into account all possible loading cases. One of the most severe loading cases, which can potentially occur during the operating time, is the loss-of-coolant accident, where cold water is injected into the RPV nearly at operating conditions. High pressure in combination with a thermal shock of the ferritic pressure vessel wall caused by the injection of cold water leads to a considerable load at the belt-line area known as Pressurized Thermal Shock (PTS). Usually the assessment against brittle failure is based on a deterministic fracture-mechanics analysis, in which common parameters like J-integral or stress intensity factor are employed to calculate the load path for an assumed (postulated) flaw during the PTS event. The most important input data for the fracture-mechanics analysis is the transient thermal-hydraulics (TH) load of the RPV during the emergency cooling. Such data can be calculated by analytical fluid-mixing codes verified on experiments, such as KWU-MIX, or by numerical Computational Fluid Dynamics (CFD) tools after suitable validation. In KWU-MIX, which is the standard used for TH calculations within PTS analyses, rather conservative analytical models for the quantification of mixing and, depending on the water level, condensation processes in the downcomer (including simplified stripe and plume formations) are utilized. On the contrary, the numerical CFD tools can provide best-estimate results due to the possibility to consider more realistically the stripe and plume formations as well as the geometry of the RPV in detail. In order to quantify the safety margin inherent to the standard approach, two fracture-mechanics analyses of the RPV Beznau 1 based on thermal-hydraulic input data from KWU-MIX and CFD analyses were performed. Subsequently the resulting loading paths were compared between each other and with material properties obtained from the irradiation surveillance program of the RPV to demonstrate the exclusion of brittle-fracture initiation.

Influence Factors of Fracture Mechanics Analysis of Reactor Pressure Vessel under Pressurized Thermal Shock

2019 ◽  
Vol 795 ◽  
pp. 333-339
Author(s):  
Juan Luo ◽  
Jia Cheng Luo
Keyword(s):  
Fracture Toughness ◽  
Fracture Mechanics ◽  
Thermal Shock ◽  
Neutron Irradiation ◽  
Pressure Vessel ◽  
Influence Factors ◽  
Pressurized Thermal Shock ◽  
Inner Surface ◽  
Mechanics Analysis ◽  
Reactor Pressure

When the reactor pressure vessel (RPV) is subjected to pressurized thermal shock (PTS), the cooling water injected by the emergency core cooling system (ECCS) will generate a large temperature difference in the wall thickness of the pressure vessel. On the other hand, the fracture toughness of the RPV material decreases a lot under long-term neutron irradiation. Under this condition, the PTS transient may cause a rapid growth of defects in the inner surface of the vessel, resulting in failure of the pressure vessel. In this paper, the fracture mechanics analysis method of RPV under pressurized thermal shock is studied. The thermal analysis and structural analysis of the pressure vessel are performed by finite element method. The stress intensity factor and fracture toughness are obtained through calculation. At the same time, the influence factors of fracture mechanics analysis of RPV under PTS condition are analyzed. The effects of different crack size, crack type, load transient, and neutron irradiation flux on the PTS fracture mechanics analysis results are evaluated. Results show that the larger the ratio of length to depth for axial inner surface cracks, the easier RPV crack grows. Under small break condition, the circumferential cracks are safer than axial cracks. The longer the operating time, the more severe the embrittlement of RPV materials, which will lead to the failure of RPV more easily. For the two typical PTS transients studied in this paper, the re-pressurization condition is safer than the small break condition. The results can provide basis for structural integrity assessment of RPV under PTS condition.

Evaluation on Constraint Effect of Reactor Pressure Vessel Under Pressurized Thermal Shock

2013 ◽  
Author(s):  
Naoki Ogawa ◽  
Kentaro Yoshimoto ◽  
Takatoshi Hirota ◽  
Shohei Sakaguchi ◽  
Toru Oumaya
Keyword(s):  
Fracture Toughness ◽  
Thermal Shock ◽  
Surface Crack ◽  
Pressure Vessel ◽  
Crack Tip ◽  
Reactor Pressure Vessel ◽  
Constraint Effect ◽  
Elastic Plastic ◽  
Pressurized Thermal Shock ◽  
Reactor Pressure

In recent years, the integrity of reactor pressure vessel (RPV) under pressurized thermal shock (PTS) accident has become controversial issue since the larger shift of RTNDT in some higher fluence surveillance data raised a concern on RPV integrity. Under PTS condition, the combination of thermal stress due to a temperature gradient and mechanical stress due to internal pressure causes considerable tensile stress inside the wall of RPV. Currently, RPV integrity is assessed by comparing stress intensity factor on a crack tip under PTS condition and a reference toughness curve based on the fracture toughness data of irradiated compact specimens. Since PTS loading is large enough to cause plastic deformation, a crack tip behavior on the inner surface of RPV can be explained by elastic-plastic fracture mechanics using the J-integral. In this study, 3D elastic plastic finite element analyses were performed to assess the crack tip behavior on surface of a RPV under Loss of coolant Accident, which causes one of the most severe PTS condition. In order to quantify the constraint effect on a surface crack, J-Q approach was applied. The constraint effect of a surface crack was compared with a compact specimen and its influence on the fracture toughness was assessed. As a result, the difference of constraint effect was clearly obtained. And it is recommended to consider constraint effects in the evaluation of structural integrity of RPV under PTS.

Further Evaluation of the Structural Integrity of the Chinese Qinshan 300-MWe Reactor Pressure Vessel Under Pressurized Thermal Shock Using the Master Curve Method

2016 ◽  
Author(s):  
Yupeng Cao ◽  
Yinbiao He ◽  
Hui Hu ◽  
Hui Li
Keyword(s):  
Fracture Mechanics ◽  
Thermal Shock ◽  
Pressure Vessel ◽  
Master Curve ◽  
Structural Integrity ◽  
Reactor Pressure Vessel ◽  
Master Curve Method ◽  
Pressurized Thermal Shock ◽  
Curve Method ◽  
Reactor Pressure

Pressurized thermal shock (PTS) is a potential major threat to the structural integrity of the reactor pressure vessel (RPV) in a nuclear power plant. An earlier work on the PTS analysis of the Chinese Qinshan 300-MWe RPV was performed with the single parameter fracture mechanics method by Shanghai nuclear engineering research and design institute (SNERDI). The integrity analysis of this RPV under PTS was re-evaluated using the Master Curve method later in the paper PVP2015-45577[1]. The objective of this paper is to expand on the previous work, covering more crack geometries and transients to discuss the differences in the use of Master curve based and single parameter linear elastic fracture mechanics based method for PTS analysis. Attempts are made to consider additional size adjustment to the long crack front, which yields more reasonable maximum allowable transition temperature.

XFEM and Standard Fracture-Mechanics Analyses of the Reactor Pressure Vessel Gösgen Based on Thermal-Hydraulics Input Data From KWU-MIX and CFD Analyses

2019 ◽  
Author(s):  
Alexander Mutz ◽  
Tomas Nicak ◽  
Richard Trewin ◽  
Ingo Cremer
Keyword(s):  
Fracture Mechanics ◽  
Thermal Shock ◽  
Pressure Vessel ◽  
Brittle Failure ◽  
Input Data ◽  
Cold Water ◽  
Reactor Pressure Vessel ◽  
Thermal Hydraulics ◽  
Mechanics Analysis ◽  
Reactor Pressure

Abstract The integrity of a reactor pressure vessel (RPV) has to be ensured throughout its entire life in accordance with the applicable regulations. Typically an assessment of the RPV against brittle failure needs to be conducted by taking into account all possible loading cases. One of the most severe loading cases, which can potentially occur during the operating time, is the loss-of-coolant accident, where cold water is injected into the RPV at operating conditions. High pressure in combination with a thermal shock of the ferritic pressure vessel wall caused by the injection of cold water leads to a considerable load at the belt-line area known as Pressurized Thermal Shock (PTS). Usually the assessment against brittle failure is based on a deterministic fracture-mechanics analysis, in which common parameters like J-integral or stress intensity factor are employed to calculate the load path for an assumed (postulated) flaw during the PTS event. As an alternative to this standard approach a fracture mechanics assessments based on eXtended Finite Element Method (XFEM) approach can be performed. The most important input data for the fracture-mechanics analysis is the transient thermal-hydraulics (TH) load of the RPV during the emergency cooling. Such data can be calculated by analytical fluid-mixing codes verified on experiments, such as KWU-MIX, or by numerical Computational Fluid Dynamics (CFD) tools after suitable validation. In KWU-MIX, which is the standard used for TH calculations within PTS analyses, rather conservative analytical models for the quantification of mixing and, depending on the water level, condensation processes in the downcomer (including simplified stripe and plume formations) are utilized. On the contrary, the numerical CFD tools can provide best-estimate results due to the possibility to consider more realistically the stripe and plume formations as well as the geometry of the RPV in detail. In a previous paper [1] results of standard and XFEM analyses of the RPV Gösgen 1 based on thermal-hydraulics input data from KWU-MIX were presented. This paper presents new results based on thermal-hydraulics input data from CFD. The new results are compared with those from [1] in order to show additional safety margins obtained by using thermal-hydraulics input data from CFD.

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