Round robin analysis for probabilistic structural integrity of reactor pressure vessel under pressurized thermal shock

2005 ◽  
Vol 19 (2) ◽  
pp. 634-648 ◽  
Author(s):  
Myung Jo Jhung ◽  
Changheui Jang ◽  
Seok Hun Kim ◽  
Young Hwan Choi ◽  
Hho Jung Kim ◽  
...  
Keyword(s):  
Thermal Shock ◽  
Pressure Vessel ◽  
Structural Integrity ◽  
Round Robin ◽  
Reactor Pressure Vessel ◽  
Pressurized Thermal Shock ◽  
Reactor Pressure

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.

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.

Round robin analysis of pressurized thermal shock for reactor pressure vessel

2003 ◽  
Vol 226 (2) ◽  
pp. 141-154 ◽  
Author(s):  
Myung Jo Jhung ◽  
Seok Hun Kim ◽  
Jin Ho Lee ◽  
Youn Won Park
Keyword(s):  
Thermal Shock ◽  
Pressure Vessel ◽  
Round Robin ◽  
Reactor Pressure Vessel ◽  
Pressurized Thermal Shock ◽  
Reactor Pressure

Sensitive Analyses of Probabilistic Fracture Mechanics for Reactor Pressure Vessel during Pressurized Thermal Shock and Comparison with the Results of International Round Robin Analyses in Asian Countries

2011 ◽  
Vol 462-463 ◽  
pp. 878-883
Author(s):  
Yasuhiro Kanto ◽  
S. Yoshimura
Keyword(s):  
Fracture Mechanics ◽  
Thermal Shock ◽  
Pressure Vessel ◽  
Round Robin ◽  
Reactor Pressure Vessel ◽  
Asian Countries ◽  
Probabilistic Fracture Mechanics ◽  
Pressurized Thermal Shock ◽  
International Round ◽  
Reactor Pressure

This paper demonstrates sensitive analyses of probabilistic fracture mechanics (PFM) for reactor pressure vessel (RPV) during pressurized thermal shock (PTS) loading, and comparison of our calculation with the results of the international round robin (RR) analyses in Asian countries (Korea, Taiwan and Japan). The international Round Robin activity was performed in PFM sub-committees in the Atomic Energy Research Committee of Japan Welding Engineering Society (JWES) in conjunction with Korea and Taiwan research groups. The purposes of this program are to establish reliable procedures to evaluate fracture probability of reactor pressure vessels during pressurized thermal shock and to maintain the continuous cooperation among Asian institutes in the probabilistic approach to nuclear safety. Some parameters to RPV failure probabilities are chosen to evaluate their significance quantatively. The differences caused by selection of analyzing programs and some input parameters will be discussed.

Computational studies on pressurized thermal shock in reactor pressure vessel

2021 ◽  
Vol 152 ◽  
pp. 107987
Author(s):  
Rakesh Chouhan ◽  
Anuj Kumar Kansal ◽  
Naresh Kumar Maheshwari ◽  
Avaneesh Sharma
Keyword(s):  
Thermal Shock ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Computational Studies ◽  
Pressurized Thermal Shock ◽  
Reactor Pressure
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