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.

Structural Reliability Evaluation on the Pressurized Water Reactor Pressure Vessel Under Pressurized Thermal Shock Events

2014 ◽  
Author(s):  
Hsoung-Wei Chou ◽  
Chin-Cheng Huang
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Thermal Shock ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Reactor Pressure Vessel ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Pressurized Thermal Shock ◽  
Reactor Pressure

The failure probability of the pressurized water reactor pressure vessel for a domestic nuclear power plant in Taiwan has been evaluated according to the technical basis of the USNRC’s new pressurized thermal shock (PTS) screening criteria. The ORNL’s FAVOR code and the PNNL’s flaw models are employed to perform the probabilistic fracture mechanics analysis based on the plant specific parameters of the domestic reactor pressure vessel. Meanwhile, the PTS thermal hydraulic and the probabilistic risk assessment data analyzed from a similar nuclear power plant in the United States for establishing the new PTS rule are applied as the loading condition. Besides, an RT-based regression formula derived by the USNRC is also utilized to verify the through-wall cracking frequencies. It is found that the through-wall cracking of the analyzed reactor pressure vessel only occurs during the PTS events resulted from the stuck-open primary safety relief valves that later reclose, but with only an insignificant failure risk. The results indicate that the Taiwan domestic PWR reactor pressure vessel has sufficient structural margin for the PTS attack until either the end-of-license or for the proposed extended operation.

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.

Integrity Evaluation of Korean Nuclear Reactor Pressure Vessel Under Pressurized Thermal Shock Conditions According to JEAC

2013 ◽  
Author(s):  
Se-Chang Kim ◽  
Jae-Boong Choi ◽  
Doo-Ho Cho ◽  
Sang-Min Lee ◽  
Yong-Beum Kim ◽  
...  
Keyword(s):  
Finite Element ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Thermal Shock ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Reactor Pressure Vessel ◽  
Operation Conditions ◽  
Pressurized Thermal Shock ◽  
Reactor Pressure

In nuclear power plant, reactor pressure vessel (RPV) is the primary equipment that contains reactor cores and coolant. The RPV integrity should be evaluated in consideration with transient operation conditions and material deterioration. Especially, the pressurized thermal shock (PTS) has been considered as one of the most important issues regarding the RPV integrity since Rancho Seco nuclear power plant accident in1978. In this paper, integrity evaluation of Korean RPV was performed by using finite element analysis. PTS conditions like small break loss of coolant accident (SBLOCA) and Turkey Point steam line break (TP-SLB) were applied as loading conditions. Neutron fluence data of actual RPV operated over 30 years was used to determine fracture toughness of RPV material. The 3-dimensional finite element model including circumferential surface crack was generated for fracture mechanics analysis. The RPV integrity was evaluated according to Japan Electric Association Code (JEAC).

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.

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

Development of an Integrity Evaluation System on the Basis of Cooperative Virtual Reality Environment for Nuclear Power Plant

2004 ◽  
Author(s):  
J. C. Kim ◽  
J. B. Choi ◽  
Y. H. Choi
Keyword(s):  
Virtual Reality ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Pressure Vessel ◽  
Evaluation System ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Reactor Pressure Vessel ◽  
Virtual Reality Environment ◽  
Reactor Pressure

Since early 1950’s fracture mechanics has brought significant impact on structural integrity assessment in a wide range of industries such as power, transportation, civil and petrochemical industries, especially in nuclear power plant industries. For the last two decades, significant efforts have been devoted in developing defect assessment procedures, from which various fitness-for-purpose or fitness-for-service codes have been developed. From another aspect, recent advances in IT (Information Technologies) bring rapid changes in various engineering fields. IT enables people to share information through network and thus provides concurrent working environment without limitations of working places. For this reason, a network system based on internet or intranet has been appeared in various fields of business. Evaluating the integrity of structures is one of the most critical issues in nuclear industry. In order to evaluate the integrity of structures, a complicated and collaborative procedure is required including regular in-service inspection, fracture mechanics analysis, etc. And thus, experts in different fields have to cooperate to resolve the integrity problem. In this paper, an integrity evaluation system on the basis of cooperative virtual reality environment for reactor pressure vessel which adapts IT into a structural integrity evaluation procedure for reactor pressure vessel is introduced. The proposed system uses Virtual Reality (VR) technique, Virtual Network Computing (VNC) and knowledge based programs. This system is able to support 3-dimensional virtual reality environment and to provide experts to cooperate by accessing related data through internet. The proposed system is expected to provide a more efficient integrity evaluation for reactor pressure vessel.

Application of Flaw Updating Process on Probabilistic Structural Evaluation for a Reactor Pressure Vessel Under Pressurized Thermal Shocks

2018 ◽  
Author(s):  
Pin-Chiun Huang ◽  
Hsoung-Wei Chou ◽  
Yuh-Ming Ferng
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Fracture Probability ◽  
Reactor Pressure Vessel ◽  
Structural Evaluation ◽  
User Facility ◽  
Research User ◽  
Reactor Pressure ◽  
Thermal Shocks

The structural integrity of a reactor pressure vessel (RPV) is one of the most important issues for the operation of nuclear power plant. Nowadays, the probabilistic fracture mechanics (PFM) technique is widely used in evaluating the structural integrity of RPVs. However, the flaw characteristics used for PFM analysis are mainly derived from the Pressure Vessel Research User Facility (PVRUF) and Shoreham vessel inspection database, which may not be able to truly represent the vessel-specific condition of the analyzed RPV. In this work, the NUREG-2163 procedure which modifies the flaw characteristic parameters is employed. The Bayesian updating process which combines the prior flaw data with non-destructive examination (NDE) results as well as uncertainties is used to develop the posterior vessel-specific flaw distributions. Subsequently, the updated flaw files are used for PFM analysis to investigate the effects of NDE updated flaw characteristics on the fracture probability of RPV subjected to pressurized thermal shocks. Considering the updated flaws based on the NDE data, the analyzed results could be more plant-specific to predict the fracture risks of RPVs during operation.

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