Development of Advanced Fatigue Evaluation Methodology for Monitoring Major Components in Nuclear Power Plant

2010 ◽  
Author(s):  
W. Kim ◽  
Jongjooh Kwon ◽  
Hong Tae Kang ◽  
Gyeong-Hoi Koo ◽  
Tae-Ryong Kim
Keyword(s):  
Finite Element Method ◽  
Finite Element ◽  
Nuclear Power ◽  
Evaluation Methodology ◽  
Coolant Temperature ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Fatigue Monitoring ◽  
Improve Accuracy ◽  
Fatigue Evaluation

In an attempt to develop fatigue monitoring system, two improved fatigue evaluation schemes have been proposed to monitor fatigue degradation in major components and piping of the pressurized water reactor. Proposed methods are both aimed to obtain realistic fatigue usage factors for given plant transients. Developed schemes utilize plant operating signals such as coolant temperature, pressure and flow rate. Finite element method and an improved Green’s function approach were used to calculate stresses and fatigue usage. Case studies were performed to validate effectiveness of each proposed scheme. It has been confirmed that proposed schemes can effectively reduce excessive conservatism in estimating fatigue usage and improve accuracy in stress calculation.

scholarly journals Finite Element Based Full-Life Cyclic Stress Analysis of 316 Grade Nuclear Reactor Stainless Steel Under Constant, Variable, and Random Fatigue Loading

2018 ◽  
Vol 140 (5) ◽  
Author(s):  
Bipul Barua ◽  
Subhasish Mohanty ◽  
Joseph T. Listwan ◽  
Saurindranath Majumdar ◽  
Krishnamurti Natesan
Keyword(s):  
Finite Element ◽  
Stainless Steel ◽  
Stress Analysis ◽  
Nuclear Reactor ◽  
Three Dimensional ◽  
Fatigue Loading ◽  
Cyclic Stress ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Fatigue Evaluation

Although S∼N curve-based approaches are widely followed for fatigue evaluation of nuclear reactor components and other safety critical structural systems, there is a chance of large uncertainty in estimated fatigue lives. This uncertainty may be reduced by using a more mechanistic approach such as physics based three-dimensional (3D) finite element (FE) methods. In a recent paper (Barua et al., 2018, ASME J. Pressure Vessel Technol., 140(1), p. 011403), a fully mechanistic fatigue modeling approach which is based on time-dependent stress–strain evolution of material over the entire fatigue life was presented. Based on this approach, in this work, FE-based cyclic stress analysis was performed on 316 nuclear grade reactor stainless steel (SS) fatigue specimens, subjected to constant, variable, and random amplitude loading, for their entire fatigue lives. The simulated results are found to be in good agreement with experimental observation. An elastic-plastic analysis of a pressurized water reactor (PWR) surge line (SL) pipe under idealistic fatigue loading condition was performed and compared with experimental results.

A Study of Some Key Parameters Used in the Environmental Fatigue Evaluation of a Hot Leg Surge Nozzle

2009 ◽  
Author(s):  
Tim F. Wiley ◽  
Tim J. Pournaras ◽  
Chris T. Kupper ◽  
Mark A. Gray ◽  
Seth A. Swamy
Keyword(s):  
Nuclear Power ◽  
Sensor Data ◽  
Maximum Temperature ◽  
Pressurized Water Reactor ◽  
Maximum Temperature Difference ◽  
Pressurized Water ◽  
Coolant Pump ◽  
Surge Line ◽  
Fatigue Evaluation ◽  
Made In

When considering environmentally assisted fatigue (EAF) in the fatigue evaluation of nuclear power plant components, some assumptions made pertaining to plant operation in the design basis fatigue analyses have to be re-evaluated to accommodate potential increase in fatigue usage factors resulting from environmental effects. The surge line was identified in NUREG/CR-6260 [1] to be a representative component for the evaluation of EAF for Pressurized Water Reactor (PWR) plants. For some PWR plants, the hot leg surge nozzle is one of the components evaluated for environmentally assisted fatigue. The hot leg surge nozzle was chosen for this study because the results of the fatigue evaluation are highly dependent on several key parameters, such as maximum temperature difference between the pressurizer and hot leg piping during heatups and cooldowns, the amount of temperature sensor data available along the surge line, availability of thermal event cycle counting, and the frequency and timing of reactor coolant pump starts and stops during heatups and cooldowns. This paper assesses the impacts of the assumptions made in these key parameters on the environmental fatigue evaluation results for a typical hot leg surge nozzle.

Stress-Based Environmental Fatigue Monitoring of PWR Charging Nozzle

2013 ◽  
Author(s):  
Timothy Gilman ◽  
Archana Chinthapalli ◽  
Michael Hoehn
Keyword(s):  
Nuclear Power ◽  
Nuclear Regulatory Commission ◽  
Pressurized Water ◽  
Technical Report ◽  
Fatigue Monitoring ◽  
Energy Center ◽  
Plant Data ◽  
Water Reactors ◽  
Fatigue Evaluation ◽  
Regulatory Commission

This paper describes the techniques utilized to perform a stress-based environmentally-assisted fatigue evaluation of Westinghouse-designed charging branch nozzles on the reactor coolant loop of the Callaway Energy Center nuclear power plant. Analysis results from using idealized, design transient definitions are compared to those resulting from analysis of the actual plant data. Benchmarking analyses, performed to address Nuclear Regulatory Commission (NRC) concerns about simplified methodologies, are described. The simplified results are also compared to those produced using an advanced, multiaxial stress-based fatigue methodology defined in a recent EPRI technical report [3]. This paper concludes that stress-based fatigue monitoring using actual plant data is an effective way for a plant to manage environmentally-assisted fatigue of charging nozzles in pressurized water reactors (PWRs).

Verification Analyses for Newly Developed Automatic 3D Finite Element Crack Propagation System

2014 ◽  
Author(s):  
Hiroaki Doi ◽  
Hitoshi Nakamura ◽  
Wenwei Gu ◽  
Do-Jun Shim ◽  
Gery Wilkowski
Keyword(s):  
Finite Element ◽  
Crack Propagation ◽  
Nuclear Power ◽  
Power Plants ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
3D Finite Element ◽  
Primary Water ◽  
Axial Crack ◽  
Analysis System

In order to calculate the crack propagation in complicated-shaped locations in components such as weld in penetration structures of reactor pressure vessel of nuclear power plants, an automatic 3D finite element crack propagation system (CRACK-FEM) has been developed by the Nuclear Regulation Authority (NRA, Japan). To confirm the accuracy and effectiveness of this analysis system, a verification analysis was performed. The program used for comparison is PipeFracCAE developed by Engineering Mechanics Corporation of Columbus, which has been used for many crack propagation analyses in various applications. In this paper, the axial crack propagation analysis for primary water stress corrosion cracking (PWSCC) in a steam generator inlet nozzle of a pressurized water reactor (PWR) plant is presented. The results demonstrate that the two codes are in good agreement. The contents of this paper were conducted as a research project of the Japan Nuclear Energy Safety Organization (JNES) when one of the authors (Doi) belongs to JNES. After this project, JNES was abolished and its staff and task were absorbed into NRA on March 1, 2014.

The Failure Analysis and Processing of Digital Reactor Protection System

2013 ◽  
Author(s):  
Jun Zhao ◽  
Xing Zhou ◽  
Jin Hu ◽  
Yanling Yu
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Phase 1 ◽  
Protection System ◽  
Pressurized Water Reactor ◽  
Communication Failure ◽  
Pressurized Water ◽  
Power Failure ◽  
Self Test ◽  
Reactor Protection System

The Qinshan Nuclear Power Plant phase 1 unit (QNPP-1) has a power rating of 320 MWe generated by a pressurized water reactor that was designed and constructed by China National Nuclear Corporation (CNNC). The TELEPERM XS I&C system (TXS) is to be implemented to transform analog reactor protection system (RPS) in QNPP-1. The paper mainly describes the function, structure and characteristic of RPS in QNPP-1. It focuses on the outstanding features of digital I&C, such as strong online self-test capability, the degradation of the voting logic processing, interface improvements and CPU security. There are some typical failures during the operation of reactor protection system in QNPP-1. The way to analyze and process the failures is different from analog I&C. The paper summarizes typical failures of the digital RPS in the following types: CPU failure, communication failure, power failure, Input and output (IO) failure. It discusses the cause, risk and mainly processing points of typical failure, especially CPU and communication failures of the digital RPS. It is helpful for the maintenance of the system. The paper covers measures to improve the reliability of related components which has been put forward effective in Digital reactor protection system in QNPP-1. It will be valuable in nuclear community to improve the reliability of important components of nuclear power plants.

Construction of an Elementary Model for the Dynamic Analysis of a Pressurized Water Reactor

2013 ◽  
Vol 284-287 ◽  
pp. 652-656 ◽  
Author(s):  
Chiung Wen Tsai ◽  
Chun Kuan Shih ◽  
Jong Rong Wang
Keyword(s):  
Transfer Functions ◽  
Feedback Loops ◽  
Dynamic Responses ◽  
Coolant Temperature ◽  
Inlet Temperature ◽  
Pressurized Water Reactor ◽  
Elementary Model ◽  
Pressurized Water ◽  
Water Reactor ◽  
Neutron Kinetics

A lumped-parameter numerical model was constructed based on the conservation laws of mass and energy and the point neutron kinetics with 6 groups of delayed neutron to represent the dynamics of primary loop of a pressurized water reactor (PWR) core. On the viewpoint of control theory, the coupled phenomenon of neutron kinetics and thermohydraulics can be recognized as a dynamic system with feedback loops which is caused by the Doppler effect and the coolant temperature difference. Scilab was implemented to representing the equivalent transfer functions and associated feedback loops of a PWR core. The dynamic responses were performed by the perturbations of coolant inlet flow, coolant inlet temperature, and reactivity insertion.

Development of a Thermal Fatigue Screening and Evaluation Methodology for Pressurized Water Reactor Plants

2004 ◽  
Author(s):  
J. D. Keller ◽  
A. J. Bilanin ◽  
S. T. Rosinski
Keyword(s):  
Thermal Cycling ◽  
Development Program ◽  
Fatigue Cracking ◽  
Evaluation Methodology ◽  
Experimental Investigations ◽  
Pressurized Water Reactor ◽  
Primary Coolant ◽  
Modeling Tools ◽  
Pressurized Water ◽  
Water Reactor

Thermal cycling has been identified as a mechanism that can potentially lead to fatigue cracking in un-isolable branch lines attached to pressurized water reactor (PWR) primary coolant piping. A significant research and development program has been undertaken to understand the mechanisms causing thermal cycling and to develop models for predicting the thermal-hydraulic boundary conditions for use in piping structural and fatigue analysis. A combination of first-principles engineering modeling and scaled experimental investigations has been used to formulate improved thermal cycling modeling tools. This paper will provide an overview of the model development program, a summary of the supporting test program, and a description of the thermal cycling model structure. Benchmarking of the thermal cycling model against several PWR plant configurations is presented, demonstrating favorable comparison with cases where thermal stratification and cycling has been previously observed.

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