Cleavage and Ductile Thermal Shock Fractures of Corner-Cracked Nozzles

1989 ◽  
Vol 111 (3) ◽  
pp. 241-246 ◽  
Author(s):  
G. Yagawa ◽  
K. Ishihara
Keyword(s):  
Thermal Shock ◽  
Pressure Vessel ◽  
Structural Integrity ◽  
Three Dimensional ◽  
Pressure Vessel Steel ◽  
Fracture Parameter ◽  
The Finite Element Method ◽  
Vessel Steel ◽  
Pressurized Thermal Shock ◽  
Reactor Pressure

In order to study the structural integrity of the reactor pressure vessel under pressurized thermal shock, both the cleavage and the ductile thermal shock fracture experiments using initially corner-cracked nozzle specimens made of Type A508 class 3 pressure vessel steel were performed. In both experiments, unstable fractures were realized, although the test conditions were very conservative compared to those of real plants. Finally, the three-dimensional and time-dependent fracture parameter obtained with the finite element method was successfully employed to discuss the fracture phenomenon.

Measurements of Stress During Thermal Shock in Clad Reactor Pressure Vessel Material Using Time-Resolved In-Situ Synchrotron X-Ray Diffraction

2018 ◽  
Author(s):  
Sam Oliver ◽  
Chris Simpson ◽  
Andrew James ◽  
Christina Reinhard ◽  
David Collins ◽  
...  
Keyword(s):  
Thermal Shock ◽  
Pressure Vessel ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Pressure Vessel Steel ◽  
X Ray Diffraction ◽  
Time Resolved ◽  
Vessel Steel ◽  
X Ray ◽  
Reactor Pressure

Nuclear reactor pressure vessels must be able to withstand thermal shock due to emergency cooling during a loss of coolant accident. Demonstrating structural integrity during thermal shock is difficult due to the complex interaction between thermal stress, residual stress, and stress caused by internal pressure. Finite element and analytic approaches exist to calculate the combined stress, but validation is limited. This study describes an experiment which aims to measure stress in a slice of clad reactor pressure vessel during thermal shock using time-resolved synchrotron X-ray diffraction. A test rig was designed to subject specimens to thermal shock, whilst simultaneously enabling synchrotron X-ray diffraction measurements of strain. The specimens were extracted from a block of SA508 Grade 4N reactor pressure vessel steel clad with Alloy 82 nickel-base alloy. Surface cracks were machined in the cladding. Electric heaters heat the specimens to 350°C and then the surface of the cladding is quenched in a bath of cold water, representing thermal shock. Six specimens were subjected to thermal shock on beamline I12 at Diamond Light Source, the UK’s national synchrotron X-ray facility. Time-resolved strain was measured during thermal shock at a single point close to the crack tip at a sample rate of 30 Hz. Hence, stress intensity factor vs time was calculated assuming K-controlled near-tip stress fields. This work describes the experimental method and presents some key results from a preliminary analysis of the data.

An Investigation of the Structural Integrity of a Reactor Pressure Vessel Using Three-Dimensional Computational Fluid Dynamics and Finite Element Method Based Probabilistic Pressurized Thermal Shock Analysis for Optimizing Maintenance Strategy

2018 ◽  
Vol 140 (5) ◽  
Author(s):  
Xiaoyong Ruan ◽  
Toshiki Nakasuji ◽  
Kazunori Morishita
Keyword(s):  
Fluid Dynamics ◽  
Finite Element Method ◽  
Computational Fluid Dynamics ◽  
Finite Element ◽  
Spatial Distribution ◽  
Pressure Vessel ◽  
Structural Integrity ◽  
Three Dimensional ◽  
Pressurized Thermal Shock ◽  
Reactor Pressure

The structural integrity of a reactor pressure vessel (RPV) is important for the safety of a nuclear power plant. When the emergency core cooling system (ECCS) is operated and the coolant water is injected into the RPV due to a loss-of-coolant accident (LOCA), the pressurized thermal shock (PTS) loading takes place. With the neutron irradiation, PTS loading may lead an RPV to fracture. Therefore, it is necessary to evaluate the performance of RPV during PTS loading to keep the reactor safety. In the present study, optimization of RPV maintenance is considered, where two different attempts are made to investigate the RPV integrity during PTS loading by employing the deterministic and probabilistic methodologies. For the deterministic integrity evaluation, three-dimensional computational fluid dynamics (3D-CFD) and finite element method (FEM) simulations are performed, and stress intensity factors (SIFs) are obtained as a function of crack position inside the RPV. As to the probabilistic integrity evaluation, on the other hand, a practically more useful spatial distribution of SIF on the RPV is calculated. By comparing the distribution thus obtained with the fracture toughness included as a part of the master curve, the dependence of conditional failure probabilities on the position inside the RPV is obtained. Using the spatial distribution of conditional failure probabilities in RPV, the priority of the inspection and maintenance is finally discussed.

Optimizing Maintenance Strategy of a Reactor Pressure Vessel Using 3D-CFD and FEM Based Probabilistic Pressurized Thermal Shock Analysis

2017 ◽  
Author(s):  
Xiaoyong Ruan ◽  
Toshiki Nakasuji ◽  
Kazunori Morishita
Keyword(s):  
Spatial Distribution ◽  
Thermal Shock ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Cooling System ◽  
Reactor Pressure Vessel ◽  
Pressurized Thermal Shock ◽  
Core Cooling ◽  
Reactor Pressure

The structural integrity of a reactor pressure vessel (RPV) is important for the safety of a nuclear power plant. When the emergency core cooling system (ECCS) is operated and the coolant water is injected into the RPV due to a loss-of-coolant accident (LOCA), the pressurized thermal shock (PTS) loading takes place. With the neutron irradiation, PTS loading may lead a RPV to fracture. Therefore, it is necessary to evaluate the performance of RPV during PTS loading to keep the reactor safety. In the present study, optimization of RPV maintenance is considered, where two different attempts are made to investigate the RPV integrity during PTS loading by employing the deterministic and probabilistic methodologies. For the deterministic integrity evaluation, 3D-CFD and finite element method (FEM) simulations are performed, and stress intensity factors (SIFs) are obtained as a function of crack position inside the RPV. As to the probabilistic integrity evaluation, on the other hand, a more accurate spatial distribution of SIF on the RPV is calculated. By comparing the distribution thus obtained with the fracture toughness included as a part of the master curve, the dependence of fracture probabilities on the position inside the RPV is obtained. Using the spatial distribution of fracture probabilities in RPV, the priority of the inspection and maintenance is finally discussed.

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

2015 ◽  
Author(s):  
Yupeng Cao ◽  
Yinbiao He ◽  
Hu Hui ◽  
Hui Li ◽  
Fuzhen Xuan
Keyword(s):  
Thermal Shock ◽  
Pressure Vessel ◽  
Master Curve ◽  
Structural Integrity ◽  
Master Curve Method ◽  
Loss Of Coolant Accident ◽  
Pressurized Thermal Shock ◽  
Loss Of Coolant ◽  
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. A comprehensive structural integrity analysis of the Chinese Qinshan 300-MWe RPV subjected to PTS events including the small break loss-of-coolant accident (SB-LOCA) and large break loss-of-coolant accident (LB-LOCA) transients was performed by Shanghai nuclear engineering and design institute (SNERDI). The J-integral values at the deepest and the near cladding-base interface points of the crack were calculated with the linear elastic material model. And the RTPTS values were determined by the tangent approach. In the case that the RTNDT at or beyond the RPV design life may exceed the RTPTS according to the previous analysis procedure, the objective of this paper is to apply the Master Curve method to the re-evaluation of the integrity of this RPV, taking account of constraint and crack length effects. The over-conservatism in the previous evaluation is identified by comparing the new calculation with the previous one. The new RTPTS values are increased to varied extents for the different loading transients.

scholarly journals Probabilistic analysis of PWR Reactor Pressure Vessel under Pressurized Thermal Shock

2021 ◽  
Vol 8 (1) ◽  
pp. 1-9
Author(s):  
Kuen Ting ◽  
Anh Tuan Nguyen ◽  
Kuen Tsann Chen ◽  
Li Hwa Wang ◽  
Yuan Chih Li ◽  
...  
Keyword(s):  
Crack Initiation ◽  
Thermal Shock ◽  
Conditional Probability ◽  
Pressure Vessel ◽  
Structural Integrity ◽  
Reactor Pressure Vessel ◽  
Term Operation ◽  
Oak Ridge ◽  
Pressurized Thermal Shock ◽  
Reactor Pressure

The beltline region is the most important part of the reactor pressure vessel, become embrittlement due to neutron irradiation at high temperature after long-term operation. Pressurized thermal shock is one of the potential threats to the integrity of beltline region also the reactor pressure vessel structural integrity. Hence, to maintain the integrity of RPV, this paper describes the benchmark study for deterministic and probabilistic fracture mechanics analyzing the beltline region under PTS by using FAVOR code developed by Oak Ridge National Laboratory. The Monte Carlo method was employed in FAVOR code to calculate the conditional probability of crack initiation. Three problems from Probabilistic Structural Integrity of a PWR Reactor Pressure Vessel (PROSIR) round-robin analysis were selected to analyze, the present results showed a good agreement with the Korean participants’ results on the conditional probability of crack initiation.

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.

Elastic-Plastic Fracture Analysis of a Reactor Pressure Vessel Subjected to Pressurized Thermal Shock

2017 ◽  
Author(s):  
Huajing Guo ◽  
Zhongxian Wang ◽  
Poh-Sang Lam
Keyword(s):  
Thermal Shock ◽  
Pressure Vessel ◽  
Three Dimensional ◽  
Reactor Pressure Vessel ◽  
Pressurized Thermal Shock ◽  
Dimensional Finite Element ◽  
Fracture Theory ◽  
Crack Tip Constraint ◽  
Three Dimensional Finite Element ◽  
Reactor Pressure

Three-dimensional finite element models are used to analyze a reactor pressure vessel with an axial semi-elliptical surface crack subjected to pressurized thermal shock. During the thermal shock event, the J-A2 two-parameter fracture theory is used to investigate the temperature-dependent constraint effect at the deepest point and the surface point of the crack. Using the R6 methodology, a series of constraint-based crack failure assessment curves during the thermal shock can be obtained. It was found that the crack tip constraint should be considered for developing a more realistic failure criterion.

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