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.

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.

Nuclear PWR Plants: Managing the Expertises of Cast Components Removed From Reactor Coolant Circuit

2008 ◽  
Author(s):  
Georges Bezdikian
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Coolant Circuit ◽  
Integrity Assessment ◽  
Reactor Coolant ◽  
Maintenance Program ◽  
Nuclear Plants ◽  
Toughness Evaluation ◽  
Circuit Components ◽  
Prediction Criteria

The French utility has organized the aging degradation assessment for nuclear plants in operation in maximum safety rules. Also the utility decided to engaged large program of expertises, for each degradations identified, based on research and development activities and prediction criteria program of Nuclear Plants in function of several actions: expertises, data bases on characteristics, etc. This paper shows the expertise results and thermal aging for cast duplex stainless steel studies on reactor coolant circuit components engaged to evaluate and to monitor the toughness evaluation and increasing factor. Also this paper presents the strategy associated. This paper shows the applications were evolved for 3-loop PWR plants. The monitoring was mainly oriented on evaluation of the ratio computation and measurement. This integrity assessment and expertises results values available periodically were performed considering: • the life evaluation of the plants and alternative maintenance actions, • the large database from cast reactor coolant component assessed after removed from nuclear power plants, • the identification of degradation for different components and prediction criteria proposed. The results obtained are updated in the periodic maintenance program and in volume of expertise database for life management.

Application of Metallographic Methods for Thermal Ageing Evaluation on Samples from Primary Piping of NPP

2017 ◽  
Vol 891 ◽  
pp. 60-66
Author(s):  
Jana Petzová ◽  
Martin Březina ◽  
Miloš Baľák ◽  
Mária Dománková ◽  
Ľudovít Kupča
Keyword(s):  
Material Properties ◽  
Nuclear Power ◽  
Power Plants ◽  
Operating Temperature ◽  
Thermal Ageing ◽  
Primary Circuit ◽  
Degradation Mechanisms ◽  
Reliable Operation ◽  
Term Operation

During a long-term operation of nuclear power plants (NPP), the changes of structural material properties occur. To ensure the safe and reliable operation, it is necessary to monitor and evaluate these changes mainly on components from primary circuit of NPPs. One of the dominant ageing mechanisms of NPP components besides the radiation embrittlement and the fatigue loads is the thermal ageing. The thermal ageing is the temperature, material and time dependent degradation mechanisms due to long-term exposure at the operating temperature of 570 K.This paper describes the project for thermal ageing monitoring at primary piping in NPP Bohunice Unit 3. There are summarized the results obtained from evaluation of original primary piping material.

Guided Waves As an Online Monitoring Technology for Long Term Operation in NPPs: New Experimental Results on a Steam Discharge Pipe

2017 ◽  
Author(s):  
Francesco Bertoncini ◽  
Mauro Cappelli ◽  
Francesco Cordella ◽  
Marco Raugi
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Guided Waves ◽  
Complex Structure ◽  
Theoretical Background ◽  
Guided Wave ◽  
Experimental Results ◽  
Primary Circuit ◽  
Term Operation

Guided Wave (GW) testing is regularly used for finding defect locations through long range screening using low-frequency waves (from 5 to 250 kHz) [1]-[3]. Magnetostrictive sensors can overcome some issues, which usually limit the application to Nuclear Power Plants (NPPs) [4], like for example, high temperatures [5]-[6], high wall thickness of components in the primary circuit, and characteristic defect typologies. The authors have already shown the basic theoretical background, some simulations and some first experimental results concerning a real steel pipe, used for steam discharge, having a complex structure. Collecting more experimental data with a novel test campaign on the same pipe its complex structure results as a useful benchmark for the application of GWs as Non Destructive Techniques (NDT). Experimental measures using a symmetrical probe and a local probe in different configurations (pulse-echo and pitch-catch) indicate that GW testing with magnetostrictive sensors can be reliably applied to long-term monitoring of NPP components.

Project STYLE: Post-Test Analysis of Mock-Up3

2014 ◽  
Author(s):  
Tomas Nicak ◽  
Herbert Schendzielorz ◽  
Elisabeth Keim ◽  
Gottfried Meier ◽  
Dominique Moinereau ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Large Scale ◽  
Structural Integrity ◽  
Small Scale ◽  
Primary Circuit ◽  
Term Operation ◽  
Integrity Assessment ◽  
Work Related ◽  
Post Test ◽  
Circuit Components

The safety and reliability of all systems has to be maintained throughout the lifetime of a nuclear power plant. Continuous R&D work is needed in targeted areas to meet the challenges of long term operation of existing and new plants designs. The European project STYLE aims to develop and validate advanced methods of structural integrity assessment applicable in the ageing and lifetime management of primary circuit components. There are three large scale mock-up tests in STYLE each of them dedicated to investigate specific effects. This paper presents the work related to Mock-up3, which is dedicated to investigate influence of cladding on the crack initiation and propagation as well as the transferability of material properties from small scale specimens to a large scale component. The performed post-test analyses focus on both the further understanding and interpretation of the Mock-up3 test and on the effect of cladding on structural integrity and LBB behavior of reactor coolant pressure boundary components.

Managing Environmental Fatigue of the Real Components in a PWR

2010 ◽  
Author(s):  
Horst Rothenho¨fer ◽  
Gu¨nter Ko¨nig
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Life Time ◽  
Monitoring Data ◽  
Design Phase ◽  
Term Operation ◽  
Penalty Factor ◽  
Fatigue Evaluation ◽  
Operational Modes

In this paper the relevance of environmental fatigue for real components of a PWR is discussed based on the ANL procedure according to NUREG/CR-6909 using the penalty factor Fen. Monitoring data of several German PWRs for the surge lines and spray lines have been evaluated and assessed for five years of operation each. Monitored loads are compared to specified loads and lab loads and the fatigue evaluation is assessed comparing traditional fatigue evaluation to the ANL approach. Taking into account environmental fatigue the calculated life time of a component in LWR environments can be reduced by a factor of 10 or even more compared to common fatigue evaluation. These tremendous reductions are based on lab results whereas experience from existent nuclear power plants has not yet shown any relevance of this influence. The conclusions of evaluations and assessments are summarized in a concept for long term operation (LTO). This concept contains procedures how to manage environmental fatigue. In the design phase conservatism in specified loads, especially numbers of load cycles, can be reduced based on the duty to monitor real loads and assess fatigue during operation. In the operating phase monitoring data should be used to record loads, evaluate and assess fatigue and subsequently reduce fatigue by optimizing operational modes.

Evaluation of Environmental Fatigue in PWR PLM Activities

2005 ◽  
Author(s):  
Masanobu Iwasaki ◽  
Yasukazu Takada ◽  
Takao Nakamura
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Japan Society ◽  
Know How ◽  
Engineering Society ◽  
Fatigue Initiation ◽  
Plant Life ◽  
Fatigue Evaluation ◽  
Term Maintenance

It is important to evaluate environmental fatigue for establishing long-term maintenance plans as part of Plant Life Management (PLM) activities for nuclear power plants. In Japan, the former MITI requested the utilities in 2000 to use “The Guidelines for Evaluating Fatigue Initiation Life Reduction in LWR Environment (MITI guidelines)” for PLM evaluation. In 2002, Thermal and Nuclear Power Engineering Society (TENPES) issued the guidelines for applying the evaluation formulas of MITI guidelines to actual plants. At present, fatigue evaluations taking into account environmental effects as part of PLM activities are conducted in accordance with these guidelines. This paper describes how a typical PWR plant conducts such an evaluation. The Japan Society of Mechanical Engineers (JSME) is now drawing up a code for environmental fatigue evaluation, incorporating the latest data on fatigue experiments and know-how on fatigue evaluation. After being issued, this code will be used to evaluate environmental fatigue in PLM activities.

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 Simulation of MASPn Experiments in MISTRA Test Facility with COCOSYS Code

2008 ◽  
Vol 2008 ◽  
pp. 1-7 ◽  
Author(s):  
Mantas Povilaitis ◽  
Egidijus Urbonavičius
Keyword(s):  
Boundary Conditions ◽  
Nuclear Power ◽  
Power Plants ◽  
Volume Fraction ◽  
Characteristic Length ◽  
Initial Conditions ◽  
Test Facility ◽  
Work Package ◽  
Acceptable Agreement

An issue of the stratified atmospheres in the containments of nuclear power plants is still unresolved; different experiments are performed in the test facilities like TOSQAN and MISTRA. MASPn experiments belong to the spray benchmark, initiated in the containment atmosphere mixing work package of the SARNET network. The benchmark consisted of MASP0, MASP1 and MASP2 experiments. Only the measured depressurisation rates during MASPn were available for the comparison with calculations. When the analysis was performed, the boundary conditions were not clearly defined therefore most of the attention was concentrated on MASP0 simulation in order to develop the nodalisation scheme and define the initial and boundary conditions. After achieving acceptable agreement with measured depressurisation rate, simulations of MASP1 and MASP2 experiments were performed to check the influence of sprays. The paper presents developed nodalisation scheme of MISTRA for the COCOSYS code and the results of analyses. In the performed analyses, several parameters were considered: initial conditions, loss coefficient of the junctions, initial gradients of temperature and steam volume fraction, and characteristic length of structures. Parametric analysis shows that in the simulation the heat losses through the external walls behind the lower condenser installed in the MISTRA facility determine the long-term depressurisation rate.

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