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.

The Influence of Chemistry Concentration Uncertainty on the Probabilistic Fracture Mechanics Analysis

2011 ◽  
Author(s):  
Bo-Yi Chen ◽  
Chin-Cheng Huang ◽  
Hsuing-Wei Chou ◽  
Ru-Feng Liu ◽  
Hsien-Chou Lin
Keyword(s):  
Fracture Mechanics ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Fracture Probability ◽  
Reactor Pressure Vessel ◽  
Oak Ridge ◽  
Probabilistic Fracture Mechanics ◽  
Full Power ◽  
Reactor Pressure

The chemistry concentration uncertainty of cooper and nickel significantly affects the shift in reference nil-ductility transition temperature (ΔRTNDT). The uncertainty comes from the methods and equipments applied in measurements, the lack of specimen in surveillance capsule, and the non-homogeneous of material. The variations of ΔRTNDT result in the differences of failure probability of reactor pressure vessel. In this study, the structural integrity of Chinshan boiling water reactor RPV shell welds was evaluated by probabilistic fracture mechanics code-Fracture Analysis of Vessel – Oak Ridge (FAVOR). The influence of chemistry concentration uncertainty on the fracture probability of Chinshan nuclear power plant RPV with 32 and 64 effective full power years (EFPY) operation was discussed. The results of this work can be used to evaluate the structural integrity of the welds located at the RPV beltline region, and provide the aging assessment of reactor pressure vessel.

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.

Nuclear Reactor Pressure Vessel Integrity Assessment: Enhancement Provided by 3D Modeling Flaw Stability Calculations

2013 ◽  
Author(s):  
Adolfo Arrieta-Ruiz ◽  
Eric Meister ◽  
Henriette Churier
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Structural Integrity ◽  
Reactor Pressure Vessel ◽  
Integrity Assessment ◽  
Potential Damage ◽  
Vessel Integrity ◽  
Reactor Pressure

Structural integrity of the Reactor Pressure Vessel (RPV) is one of the main considerations regarding safety and lifetime of Nuclear Power Plants (NPP) since this component is considered as not reasonably replaceable. Brittle fracture risk associated with the embrittlement of RPV steel in irradiated areas is the main potential damage. In France, deterministic integrity assessment for RPV is based on the crack initiation stage. The stability of an under-clad postulated flaw in the core area is currently evaluated under a Pressurized Thermal Shock (PTS) through a fracture mechanics simplified method. One of the axes of EDF’s implemented strategy for NPP lifetime extension is the improvement of the deterministic approach with regards to the input data and methods so as to reduce conservatisms. In this context, 3D finite element elastic-plastic calculations with flaw modelling have been carried out recently in order to quantify the enhancement provided by a more realistic approach in the most severe events. The aim of this paper is to present both simplified and 3D modelling flaw stability evaluation methods and the results obtained by running a small break LOCA event.

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.

The Delivery of Technical Training Within the European Style Project

2013 ◽  
Author(s):  
Adam Toft ◽  
John Sharples
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Reactor Pressure Vessel ◽  
Nuclear Industry ◽  
Technical Training ◽  
Illustrative Case ◽  
Integrity Assessment ◽  
Leak Before Break ◽  
Reactor Pressure

The STYLE project considers structural integrity for lifetime management of non-reactor pressure vessel components of nuclear power plant. The project is funded under the seventh European Commission framework programme. A broad objective of the project is to assess, optimise and develop application of advanced tools for structural integrity assessment of reactor coolant pressure boundary components other than the reactor pressure vessel. One aspect of the STYLE project is intended to address the issue of succession planning within the European nuclear industry. With many key technical experts now approaching retirement it is essential to progress the technical expertise of those at an earlier stage of their career in the industry. The paper describes how technical training has been delivered as an integral part of the STYLE project to support retention of the current level of technical capability in future. Diverse aspects of training are described. These include participation in experimental work, numerical modelling and simulation, application of engineering assessment procedures, leak-before-break, probabilistic fracture mechanics and materials behaviour. An illustrative case study is described, in which trainees received practical instruction in the essential steps for technical justification of a leak-before-break argument.

The Importance of a Strong Training Element Within the European STYLE Project

2012 ◽  
Author(s):  
Adam Toft ◽  
John Sharples
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Reactor Pressure Vessel ◽  
Nuclear Industry ◽  
Framework Programme ◽  
Integrity Assessment ◽  
Future Viability ◽  
European Nuclear ◽  
Reactor Pressure

With many key technical experts within the European nuclear industry now approaching retirement, the continued training and professional development of less experienced people is vital for the future viability of the industry. Consequently, European framework programme projects are including a strong training element within their work packages. The STYLE project considers structural integrity for lifetime management of non-reactor pressure vessel components of nuclear power plant. The project is funded under the seventh European Commission framework programme. The objective of the project is to assess, optimise and develop application of advanced tools for structural integrity assessment of reactor coolant pressure boundary components other than the reactor pressure vessel.

Probabilistic Pressurized Thermal Shocks Analysis for a Reactor Pressure Vessel

2014 ◽  
Author(s):  
Guian Qian ◽  
Markus Niffenegger
Keyword(s):  
Aspect Ratio ◽  
Pressure Vessel ◽  
Crack Depth ◽  
Crack Density ◽  
Reactor Pressure Vessel ◽  
Probabilistic Methods ◽  
Critical Crack ◽  
Research User ◽  
Reactor Pressure ◽  
Thermal Shocks

Both deterministic and probabilistic methods are used to analyze the integrity of a reactor pressure vessel (RPV) subjected to pressurized thermal shocks (PTSs). The FAVOR code is applied to calculate the probabilities for crack initiation and failure of the RPV subjected to two transients, by considering crack distributions based on cracks observed in the Shoreham and pressure vessel research user facility (PVRUF) RPVs. The crack parameters, i.e. crack density, depth, aspect ratio, orientation and location are assumed as random variables following different distributions. KI of the cracks with the same depth increases with its aspect ratio. Both KI and KIC at the crack tip increase with crack depth, which is the reason why a deeper crack does not necessarily lead to a higher failure probability. The underclad crack is the most critical crack and the deeper crack is the least critical one in this study. Considering uncertainties of the transients results in higher failure probabilities.

scholarly journals ANALYSIS OF 3D SEMI-ELLIPTICAL CRACK ON REACTOR PRESSURE VESSEL WALL WITH LOAD STRESS AND CRACK RATIO

2019 ◽  
Vol 21 (1) ◽  
pp. 33
Author(s):  
Mike Susmikanti ◽  
Roziq Himawan ◽  
Entin Hartini ◽  
Rokhmadi Rokhmadi
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Goodness Of Fit ◽  
Structural Integrity ◽  
Reactor Pressure Vessel ◽  
Elliptic Crack ◽  
J Integral ◽  
Integrity Assessment ◽  
Neutron Exposure ◽  
Reactor Pressure

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, and geometry. The modeling of SIF value and goodness of fit are using MINITAB. The fracture condition could be predicted in comparison to the SIF value and fracture toughness. For the load stress 70 MPa and 80 MPa, with a crack ratio 0.25, 0.33 and 0.5,  the material on RPV wall will in fracture condition.Keywords: Semi elliptic surface crack, 3-dimension, reactor pressure vessel, elastic-plastic fracture mechanics, J-integral

Probabilistic Pressurized Thermal Shocks Analyses for a Reactor Pressure Vessel

2015 ◽  
Vol 137 (6) ◽  
Author(s):  
Guian Qian ◽  
Markus Niffenegger
Keyword(s):  
Aspect Ratio ◽  
Pressure Vessel ◽  
Crack Depth ◽  
Crack Density ◽  
Reactor Pressure Vessel ◽  
Probabilistic Methods ◽  
Critical Crack ◽  
Research User ◽  
Reactor Pressure ◽  
Thermal Shocks

Both deterministic and probabilistic methods are used to assess the integrity of a reactor pressure vessel (RPV) subjected to pressurized thermal shocks (PTSs). The FAVOR code is applied to calculate the probabilities for crack initiation and failure of the RPV subjected to two transients, by considering crack distributions based on cracks observed in the Shoreham and pressure vessel research user facility (PVRUF) RPVs. The crack parameters, i.e., crack density, depth, aspect ratio, orientation, and location are assumed as random variables following different distributions. KI of the cracks with the same depth increases with its aspect ratio. Both KI and KIc at the crack tip increase with crack depth, which is the reason why a deeper crack does not necessarily lead to a higher failure probability. The underclad crack is the most critical crack and the deeper crack is the least critical one in this study. Considering uncertainties of the transients results in higher failure probabilities.

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