Environmental Assisted Fatigue and EDF 900 MWe PWRs Fleet: Towards an Exemption of Environmental Effects Consideration for Secondary Circuit Components

2018 ◽  
Author(s):  
Sam Cuvilliez ◽  
Gaëlle Léopold ◽  
Thomas Métais
Keyword(s):  
Environmental Effects ◽  
Nuclear Power ◽  
Power Plants ◽  
Fatigue Loading ◽  
Degradation Process ◽  
Fatigue Curve ◽  
Primary Circuit ◽  
Secondary Circuit ◽  
Technical Specifications ◽  
Circuit Components

Environmentally Assisted Fatigue (EAF) is receiving nowadays an increased level of attention for existing Nuclear Power Plants (NPPs) as utilities are now working to extend their life. In the wake of numerous experimental fatigue tests carried out in air and also in a PWR environment, the French RCC-M code [1] has recently been amended (in its 2016 edition) with two Rules in Probatory Phase (RPP), equivalent to ASME code-cases, “RPP-2” and “RPP-3” [2] [3]. RPP-2 consists of an update of the design fatigue curve in air for stainless steels (SSs) and nickel-based alloys, and is also associated with RPP-3 which provides guidelines for incorporating the environmental penalty “Fen” factor in fatigue usage factor calculations. Alongside this codification effort, an EAF screening has recently been carried out within EDF DT [4] on various areas of the primary circuit of the 900 MWe plants of the EDF fleet. This screening led to the identification of a list of 35 “sentinel locations” which are defined as areas most prone to EAF degradation process. These locations will be subjected to detailed EAF analysis in the stress report calculations (according to the above-mentioned RCC-M code cases) for the fourth decennial inspection of the 900 MWe (VD4 900 MWe) power plants. The potential impact of EAF on the secondary circuit components is another question to address in anticipation of the VD4 900 MWe, as they may be considered as class 1 or class 2 equipment for RCC-M application according to the equipment specification. This paper presents the approach proposed by EDF towards an exemption of environmental effects consideration for secondary circuit components. The argument is first based on a review of experimental campaigns led in Japan and France (respectively on fatigue test specimens and at the component scale) which indicate a Dissolved Oxygen (DO) content threshold below which environmental effects are almost inexistent. The (conservative) value of 40 ppb has been selected consistently with NUREG/CR-6909 revision 0 [5]. The second part of the argument is built, on the one hand, on the analysis of the EDF Technical Specifications for Operation (STE) which narrows the scope of the study only to unit outages, and, on the other hand, on the analysis of 5 years of operations of all 900 MWe plants of the EDF fleet (equivalent to 170 reactor-years). It has been shown that the DO content rarely exceeded the 40 ppb threshold in the secondary coolant, and that in this case, the considered locations were not submitted to any fatigue loading.

Long Term Fatigue Evaluation in Primary Circuit Components Within Nuclear Power Plants Operated by E.ON

2013 ◽  
Author(s):  
Sven H. Reese ◽  
Dietmar Klucke
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Primary Circuit ◽  
Integrity Assessment ◽  
Comprehensive Information ◽  
Substantial Progress ◽  
Circuit Components ◽  
Numerical Finite Element ◽  
Fatigue Evaluation

Temperature-measuring thermocouples have been applied to various positions on primary circuit piping where most significant thermal loads were expected. Measuring positions were monitored and evaluated, leading to comprehensive information of existing thermal loadings like stratification and thermo shock events. During design of NPP (nuclear power plant) predicted cumulative fatigue usage factors (CUF) were defined based on specified transients. Conservative assumptions are part of this predicted end of life CUF. In comparison to detailed analysis based on real measured values, these predictions based on specified loads are leading to more conservative results in general. Evaluations underline the conservatism of design predictions in general and result in substantial progress in component integrity assessment knowledge. The range of methods to calculate component specific fatigue usage factors goes from conservative approaches based on the evaluation of the stress range of the specific events up to numerical Finite Element simulations. Based on the level of detail the conservatism decreases while the complexity of the model increases. An overview of monitoring measures of passive piping components in terms of thermal fatigue assessment is being applied in NPPs operated by E.ON Kernkraft GmbH. Evaluation methods will be discussed in detail and differences between these methods will be presented.

Operational Lessons Learned Based on Comprehensive Temperature Measurements at Primary Circuit Components

2015 ◽  
Author(s):  
Sven H. Reese ◽  
Johannes Seichter ◽  
Dietmar Klucke ◽  
H. Ertugrul Karabaki
Keyword(s):  
Temperature Measurement ◽  
Nuclear Power ◽  
Power Plants ◽  
Numerical Models ◽  
Plug Flow ◽  
Evaluation Process ◽  
Lessons Learned ◽  
Evaluation Procedure ◽  
Primary Circuit ◽  
Circuit Components

In nuclear power plants operated by E.ON, thermocouples have been installed since commissioning of the plants at fatigue relevant locations, especially at primary circuit components. Temperature measurement planes have been retro-fitted, further developed and the positions of the measurement planes have been reassigned and optimized continuously based on operational lessons learned. Comprehensive surveillance activities yield to a significant amount of information which is used to analyze the component specific health status. Additionally this information can be used to optimize plant’s behavior in the context of operational excellence. An evaluation of temperature measurement is needed from the regulatory point of view and reported periodically in the context of long term fatigue evaluation being a significant part of the German ageing management process and break preclusion concept. Beyond that, the detailed information of temperature transients, gained by these measurements, allows the engineer to analyze thermal loadings of monitored components. Subsequently this information is used to optimize operation of the plants by minimizing fatigue relevant transients. The more detailed the temperature transient information is the more complex analytical and numerical models have to be in order to comprehensively consider relevant effects. Therefore numerical Finite-Element models of primary circuit components have been developed allowing the engineer to analyze temperature loading (e.g. stratification or plug-flow events) in detail and to draw conclusions being used to optimize the plant. In the context of this publication examples of pressurized light water reactors will be discussed in detail showing the ability of the detailed evaluation process and the effectiveness of the evaluation procedure: • Operational lessons learned from temperature measurement evaluation. • Minimizing stratification events in the surge line by optimizing the point in time when main coolant pumps are turned off. • Temperature loading of the auxiliary spray line during start-up phase.

scholarly journals Demonstration of Fatigue for LTO License of NPP Borssele

2015 ◽  
Author(s):  
M. H. C. Hannink ◽  
F. J. Blom ◽  
P. W. B. Quist ◽  
A. E. de Jong ◽  
W. Besuijen
Keyword(s):  
Environmental Effects ◽  
Nuclear Power ◽  
Power Plants ◽  
Integrated Approach ◽  
Temperature Measurements ◽  
Fatigue Assessment ◽  
Term Operation ◽  
Load Monitoring ◽  
Number Of Cycles ◽  
Ageing Management

Long Term Operation (LTO) of nuclear power plants (NPPs) requires an ageing management review and a revalidation of Time Limited Ageing Analyses (TLAAs) of structures and components important for nuclear safety. An important ageing effect to manage is fatigue. Generally, the basis for this is formed by the fatigue analyses of the safety relevant components. In this paper, the methodology for the revalidation of fatigue TLAAs is demonstrated for LTO of NPP Borssele in the Netherlands. The LTO demonstration starts with a scoping survey to determine the components and locations having relevant fatigue loadings. The scope was defined by assessment against international practice and guidelines and engineering judgment. Next, a methodical review was performed of all existing fatigue TLAAs. This also includes the latest international developments regarding environmental effects. In order to reduce conservatism, a comparison was made between the number of cycles in the analyses and the number of cycles projected to the end of the intended LTO period. The projected number of cycles is based on transient counting. The loading conditions used in the analyses were assessed by means of temperature measurements by the fatigue monitoring system (FAMOS). As a result of the review, further fatigue assessment or assessment of environmental effects was necessary for certain locations. New analyses were performed using state-of-the-art calculation and assessment methods. The methodology is demonstrated by means of an example of the surge line. The model includes the piping, as well as the nozzles on the pressurizer and the main coolant line. The thermal loadings for the fatigue analysis are based on temperature measurements. Fatigue management of the NPP is ensured by means of the fatigue concept where load monitoring, transient counting and fatigue assessment are coupled through an integrated approach during the entire period of LTO.

Review of the Configuration Risk Management Methodologies

2021 ◽  
Author(s):  
Yuhang Zhang ◽  
Zhijian Zhang ◽  
He Wang ◽  
Lixuan Zhang ◽  
Dabin Sun
Keyword(s):  
Risk Management ◽  
Real Time ◽  
Nuclear Power ◽  
Safety Assessment ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Probabilistic Safety Assessment ◽  
Technical Specifications ◽  
Risk Monitor ◽  
Probabilistic Safety

Abstract To ensure nuclear safety and prevent or mitigate the consequences of accidents, many safety systems have been set up in nuclear power plants to limit the consequences of accidents. Even though technical specifications based on deterministic safety analysis are applied to avoid serious accidents, they are too poor to handle multi-device managements compared with configuration risk management which computes risks in nuclear power plants based on probabilistic safety assessment according to on-going configurations. In general, there are two methodologies employed in configuration risk management: living probabilistic safety assessment (LPSA) and risk monitor (RM). And average reliability databases during a time of interest are employed in living probabilistic safety assessment, which may be naturally applied to make long-term or regular management projects. While transient risk databases are involved in risk monitor to measure transient risks in nuclear power plants, which may be more appropriate to monitor the real-time risks in nuclear power plants and provide scientific real-time suggestions to operators compared with living probabilistic safety assessment. And this paper concentrates on the applications and developments of living probabilistic safety assessment and risk monitor which are the mainly foundation of the configuration risk management to manage nuclear power plants within safe threshold and avoid serious accidents.

An Efficient, Flexible Arrangement to Generate High Quality Process or Domestic Heat in Recuperative Gas Cycle Power Plants: Application to Helium Direct Cycle Nuclear Power Plants

1979 ◽  
Vol 101 (1) ◽  
pp. 130-140 ◽  
Author(s):  
Z. P. Tilliette ◽  
B. Pierre ◽  
P. F. Jude
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Hot Water ◽  
Primary Circuit ◽  
Inlet Temperature ◽  
Total Efficiency ◽  
High Quality ◽  
Quality Process ◽  
Direct Cycle

The advantages of gas turbine power plants in general and closed cycle systems under gas pressure in particular for waste heat recovery are well known. A satisfactory efficiency for electric power generation and good conditions to obtain a significant amount of hot water above 100°C lead to a high fuel utilization. However, as in most of projects, it is not much possible to produce high temperature steam or water without significantly decreasing the electricity production. A new method for an additional generation of high quality process or domestic heat is proposed. The basic feature of this method lies in arranging one or two steam generators or preheaters in parallel with the low pressure side of the recuperator. The high total efficiency and the noteworthy flexibility of this system are emphasized. This arrangement is suitable for any kind of heat source, but the applications presented in this paper are related to helium direct cycle nuclear power plants the main features of which are a single 600 MW(e) turbomachine, a turbine inlet temperature of 775°C, no or one intermediate cooling and a primary circuit fully integrated in a pre-stressed concrete reactor vessel.

Over Limit Oxygen Effect on Crack Propagation of WWER 440 RPV Steel in Primary Circuit Water

2009 ◽  
Author(s):  
Miroslava Ernestova ◽  
Anna Hojna
Keyword(s):  
Crack Propagation ◽  
Nuclear Power ◽  
Power Plants ◽  
Water Environment ◽  
Nuclear Research ◽  
Compact Tension ◽  
Oxygen Effect ◽  
Primary Circuit ◽  
Nuclear Research Institute ◽  
Primary Water

Experience with operating nuclear power plants worldwide reveals that many failures may be attributed to fatigue associated with mechanical loading due to vibration and with corrosion effect due to exposure to high-temperature environment. In order to clarify the simultaneous influence on reactor pressure vessel (RPV) material testing of ferritic steel 15Ch2MFA used for RPV of WWER 440 was performed at Nuclear Research Institute (NRI) autoclaves. Cyclic and constant loadings were applied to Compact Tension (CT) specimens in WWER primary water environment at 290°C and simultaneous effect of different oxygen levels (< 20 ppb, 200 ppb, 2000 ppb) on crack propagation has been evaluated. Obtained crack growth rates are compared with ASME XI Code and VERLIFE curves and crack behaviour is discussed.

INCEFA-PLUS Project: Lessons Learned From the Project Data and Impact on Existing Fatigue Assessment Procedures

2020 ◽  
Author(s):  
Sam Cuvilliez ◽  
Alec McLennan ◽  
Kevin Mottershead ◽  
Jonathan Mann ◽  
Matthias Bruchhausen
Keyword(s):  
Surface Finish ◽  
Environmental Effects ◽  
Nuclear Power ◽  
Power Plants ◽  
Lessons Learned ◽  
Fatigue Assessment ◽  
Data Points ◽  
Project Data ◽  
Assessment Procedures ◽  
Surface Finish Effect

Abstract The INCEFA+ project (INcreasing Safety in nuclear power plants by Covering gaps in Environmental Fatigue Assessment) is a five year project supported by the European Commission HORIZON2020 programme, which will conclude in June 2020. This project aims to generate and analyse Environmental Assisted Fatigue (EAF) experimental data (approximately 230 fatigue data points generated on austenitic stainless steel), and focuses on the effect of several key parameters such as mean strain, hold times and surface finish, and how they interact with environmental effects (air or PWR environment). This work focuses on the analysis of the data obtained during the INCEFA+ project. More specifically, this paper discusses how the outcome of this analysis can be used to evaluate existing fatigue assessment procedures that incorporate environmental effects in a similar way to NUREG/CR-6909. A key difference between these approaches and the NUREG/CR-6909 is the reduction of conservatisms resulting from the joint implementation of the adjustment sub-factor related to surface finish effect (as quantified in the design air curve derivation) and a Fen penalization factor for fatigue assessment of a location subjected to a PWR primary environment. The analysis presented in this paper indicates that the adjustment (sub-)factor on life associated with the effect of surface finish in air (as described in the derivation of the design air curve in NUREG/CR-6909) leads to substantial conservatisms when it is used to predict fatigue lifetimes in PWR environments for rough specimens. The corresponding margins can be explicitly quantified against the design air curve used for EAF assessment, but may also depend on the environmental correction Fen factor expression that is used to take environmental effects into account.

Research on the Micro-Mechanical State at Tip of Environmentally Assisted Cracking Based on Latin Hypercube Sampling Method

2019 ◽  
Vol 795 ◽  
pp. 74-78
Author(s):  
Kuan Zhao ◽  
He Xue ◽  
Ling Yan Zhao
Keyword(s):  
Oxide Film ◽  
Material Properties ◽  
Nuclear Power ◽  
Power Plants ◽  
Sampling Method ◽  
Latin Hypercube Sampling ◽  
Primary Circuit ◽  
Mechanical State ◽  
Latin Hypercube ◽  
Environmentally Assisted Cracking

Environmentally assisted cracking (EAC) of nickel-based alloys is one of the most significant potential safety hazards in the primary circuit of nuclear power plants. To understand the influence of randomness on micro-mechanical state at tip of EAC, Latin hypercube sampling method is applied to analyze the uncertainty of stress-strain in the oxide film at the EAC tip considering the uncertainties of load and material properties of base metal and oxide film. Meanwhile, to improve the efficiency of numerical analysis, MATLAB is employed in the secondary development for ABAQUS. With the help of finite element numerical simulation and Latin hypercube sampling method, the uncertainty of mechanical properties at tip of EAC in one-inch compact tension specimen is simulated and analyzed in this study. The results show that the randomness of material properties and load markedly affect the uncertainty of micro-mechanical state. Among the variables, The randomness of load has the greatest influence on uncertainty of strain, and Poisson`s ratio of oxide film is the smallest effect.

The SCORPIO-VVER Core Monitoring and Surveillance System for VVER Type of Reactors

2014 ◽  
Author(s):  
Jozef Molnar ◽  
Radim Vocka
Keyword(s):  
Surveillance System ◽  
Nuclear Power ◽  
Power Plants ◽  
Slovak Republic ◽  
Technical Specification ◽  
Training Simulator ◽  
History Of ◽  
Monitoring And Surveillance ◽  
Technical Specifications ◽  
European Nuclear

The SCORPIO-VVER core monitoring and surveillance system has proved since the first installation at Dukovany NPP in 1999 to be a valuable tool for the reactor operators and reactor physicists. It is now installed on four units of Dukovany NPP (EDU, Czech Republic), on two units of Bohunice NPP (EBO, Slovak Republic) replacing the original Russian VK3 system and on the full scale plant training simulator at the Centre for training and education of the reactor operators and reactor physicist in Trnava (Slovak Republic). By both Czech and Slovak nuclear regulatory bodies the system was licensed as a Technical Specification Surveillance tool. Since it’s first installation, the development of SCORPIO-VVER system continues along with the changes in VVER reactors operation. The system is being adapted according the utility needs and several notable improvements in physical modules of the system were introduced. The most significant changes were done in support of the latest optimized Gd bearing fuel assemblies, improvements in the area of core design (neutron physics, core thermal hydraulics and fuel thermal mechanics), adaptation of the system to up-rated unit conditions (uprated power up to 107%), in design and methodology of the limit and technical specifications checking and improvements in the predictive part of the system. After the currently finished upgrades the SCORPIO-VVER is still in focus of Central European nuclear power plants with the roadmap of upgrades and modifications up to 2016. This paper shortly describes the system’s main functions, the history of implementation at the VVER-440 type of reactors and deals with the system’s future upgrades and plans to meet the latest requirements of efficient and safety NPP operation.

Comparison of Strain Range Measures and Environmental Fatigue Calculation Methodologies for the Stepped Pipe Tests

2008 ◽  
Author(s):  
Robert Gurdal ◽  
Steven X. Xu
Keyword(s):  
Finite Element ◽  
Environmental Effects ◽  
Nuclear Power ◽  
Power Plants ◽  
Strain Range ◽  
The United States ◽  
Fatigue Tests ◽  
Finite Element Analyses ◽  
Strain Measure ◽  
Component Design

Various strain measure formulas exist at this time for the calculation of the strain amplitude required for fatigue calculations, and various methodologies have been suggested in the years 2005 through 2008 to take into account — in general — the environmental effects on fatigue (EAF = Environmentally-Assisted Fatigue). The purpose of this technical paper is to compare these strain measure formulas and these EAF methodologies for the case of the thermal fatigue tests of a stainless steel stepped pipe for which results have been published in the Proceedings of the 2004 PVP Conference [1]. Thermal transient finite element analyses and cyclic elasto-plastic finite element analyses were performed to obtain the thermal gradients through the pipe thickness and the resulting strain ranges. These strain ranges are based on the various strain measure definitions presented at the 2001/2005 PVP Conferences (see Ref. [2] and [3]). These various strain measure definitions were compared. Using one of the stepped pipe inside surface locations and using one of the strain range values (out of the various strain measure definitions), the allowable number of design cycles has been calculated, based on the currently mandated methodologies for the environmental effects on fatigue (EAF). These methodologies are the EAF methodologies to be applied in the United States for future fatigue calculations, either for license renewal of the currently operating nuclear power plants or for the design of new plants. The fatigue results are compared and discussed for their implication in component design.

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