Fatigue Benchmark Comparison Effort Between Code_Aster and CNNC/NPIC Software: Part 2

2020 ◽  
Author(s):  
Hai Xie ◽  
Zichen Kong ◽  
Xuejiao Shao ◽  
Tanguy Mathieu ◽  
Furui Xiong
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Austenitic Stainless Steels ◽  
Combination Method ◽  
Practical Implementation ◽  
Phase 2 ◽  
Software Comparison ◽  
Fatigue Design ◽  
Nickel Base ◽  
Fatigue Evaluation

Abstract Fatigue is identified as a significant degradation mode that affects nuclear power plants world-wide. Recent research on the interaction between fatigue degradation and the influence of PWR environment has caused international concern and triggered numerous research programs [1]. In this context, several codes & standards, including the RCC-M code, have included some technical mandatory or non-mandatory sections to address the issue. In RCC-M, this is compiled in the Rules in Probation Phase 2 and 3 [2]. Due to the novelty of these rules, there is room for improvement for the specific and practical implementation of these rules. AFCEN has hence launched a benchmark exercise at the end of 2019 to help increase the quality of these rules. Part 1 of this paper [3] states that EDF and CNNC/NPIC have launched an effort to benchmark their respective codes on fatigue calculation including the EAF algorithm. In the second part of the benchmark, the two companies started the code comparison based on a benchmark case provided by AFCEN. As stated previously, the 2016 edition of RCC-M code integrates the modifications made to the Code in Probation Phase 2 and 3(RPP)[2], which respectively modify the fatigue design curve for austenitic stainless steels and Nickel base alloys, as well as integrate environmental effects in the fatigue evaluation for austenitic stainless steel components. In this paper, a comparison between RCC-M RPP and NUREG/CR-6909 rev.1 [3] is proposed. The comparison focuses on the technical details of the strain rate calculation and transient combination method. The cumulative fatigue usage factor with or without considering EAF according to RCC-M RPP – 2 and RPP – 3 is given by EDF, using code_aster and its POST_RCCM operator. CNNC/NPIC will provide multiple sets of results including cumulative fatigue usage factors according to RCC-M RPP and NUREG CR/6909 rev. 1 respectively using its own software. Comparison of selection for peak and valleys points, Sn and Fen values are also presented. Differences of the algorithms of the two codes are also discussed.

Modifications of the 2016 Edition of the RCC-M Code to Account for Environmentally Assisted Fatigue

2016 ◽  
Author(s):  
Stéphan Courtin ◽  
Thomas Métais ◽  
Manuela Triay ◽  
Eric Meister ◽  
Stéphane Marie
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Austenitic Stainless Steels ◽  
Steering Committee ◽  
Nuclear Industry ◽  
Life Extension ◽  
Asme Code ◽  
Nickel Base ◽  
Fatigue Evaluation ◽  
International Experts

The French nuclear industry has to face nowadays a series of challenges it did not have to face a decade ago. The most significant one is to ensure a reliable and safe operation of Nuclear Power Plants (NPP) in a context of both an ageing reactor fleet and new builds. The new constructions need rules that integrate a strong operation feedback while the older NPPs need rules that will guarantee the life extension beyond 40 years of operation. In this context, a new edition of the French RCC-M Code is planned for 2016. This new edition integrates the modifications made to the Code as a result of Requests for Modification (RM), which can be submitted by anyone and which help to continuously improve the quality and robustness of the Code. Concerning fatigue analyses, the RCC-M Code steering committee has acknowledged end of 2014 the reception of two RM to modify the fatigue design curve for austenitic stainless steels and Nickel base alloys, as well as to integrate environmental effects in the fatigue evaluation for austenitic stainless steel components. The contents of these two RM were based on the proposals presented in Reference [1]. AFCEN required a technical review of these two RM and this task was performed by a working group composed by French and international experts. This process concluded to the approval of these two RM to be integrated to the 2016 edition of the RCC-M Code. This paper offers a presentation of these two new Rules in Probation Phase (RPP), this format being quite similar to Code Cases proposed by ASME Code.

Fatigue Monitoring System and Post-Processing of Temperature Measurements: Surge Line Under Stratification Loading

2015 ◽  
Author(s):  
Benoit Jouan ◽  
Jürgen Rudolph ◽  
Steffen Bergholz
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Thermal Loading ◽  
Austenitic Stainless Steels ◽  
Nuclear Industry ◽  
Integral Approach ◽  
Safety Standards ◽  
Analysis Process ◽  
Surge Line ◽  
Fatigue Evaluation

The ageing management of power plants is nowadays a main issue for all nuclear industry actors: states, regulatory agencies, operators, designers or suppliers. Consequently, lots of operators have to deal with demanding safety requirements to ensure the operation of power plants particularly in the context of lifetime extension. With regard of the fatigue assessment of nuclear components, stringent safety standards are synonymous of new parameters to take into account in the fatigue analysis process such as for instance: new design of fatigue curves particularly for austenitic stainless steels, the consideration of environmentally assisted fatigue (EAF) and stratification effects. In this context AREVA developed within the integral approach AREVA Fatigue Concept (AFC) new tools and methods to live up to operators expectations. The last mentioned stratification issue will be focused on in the framework of this dedicated paper. Based on measured thermal loads, the Fast Fatigue Evaluation (FFE) process allows for highly-automated and reliable data processing to evaluate time-dependent cumulative usage factors of mechanical components. This method has recently been extended to the consideration of stratification loading with surge line application. The paper presents the latest AREVA research and development activities on the FFE method applied to a surge line under stratification thermal loading. An additional CFD analysis was performed in order to calculate realistic thermal loadings during start-up conditions of nuclear power plant conditions. The FFE methodology was used to calculate thermal stress at all relevant locations. This approach opens the possibility of a realistic CUF calculation. The methodology, the principle results and benefits are presented in the paper.

Development of a Fatigue Design Curve for Austenitic Stainless Steels in LWR Environments: A Review

2002 ◽  
Author(s):  
Omesh K. Chopra
Keyword(s):  
Stainless Steels ◽  
Pressure Vessel ◽  
Type Strain ◽  
Nuclear Power ◽  
Austenitic Stainless Steels ◽  
Environmental Parameters ◽  
Light Water Reactor ◽  
Code Design ◽  
Fatigue Design ◽  
Asme Code

The ASME Boiler and Pressure Vessel Code provides rules for the construction of nuclear power plant components and specifies fatigue design curves for structural materials. However, the effects of light water reactor (LWR) coolant environments are not explicitly addressed by the Code design curves. Existing fatigue strain–vs.–life (ε–N) data illustrate potentially significant effects of LWR coolant environments on the fatigue resistance of pressure vessel and piping steels. This paper reviews the existing fatigue ε–N data for austenitic stainless steels in LWR coolant environments. The effects of key material, loading, and environmental parameters, such as steel type, strain amplitude, strain rate, temperature, dissolved oxygen level in water, and flow rate, on the fatigue lives of these steels are summarized. Statistical models are presented for estimating the fatigue ε–N curves for austenitic stainless steels as a function of the material, loading, and environmental parameters. Two methods for incorporating environmental effects into the ASME Code fatigue evaluations are presented. Data available in the literature have been reviewed to evaluate the conservatism in the existing ASME Code fatigue design curves.

Corrosion Resistant Cladding by YAG Laser Welding in Underwater Environment

2002 ◽  
Author(s):  
Tsutomu Kochi ◽  
Toshio Kojima ◽  
Suemi Hirata ◽  
Ichiro Morita ◽  
Katsura Ohwaki
Keyword(s):  
Laser Welding ◽  
Water Level ◽  
Nuclear Power ◽  
Power Plants ◽  
Drainage Water ◽  
Yag Laser ◽  
Corrosion Resistant ◽  
Underwater Environment ◽  
Nickel Base ◽  
Cladding Layers

It is known that stress-corrosion cracking (SCC) will occur in nickel-base alloys used in Reactor Pressure Vessel (RPV) and Internals of nuclear power plants. A SCC sensitivity has been evaluated by IHI in each part of RPV and Internals. There are several water level instrumentation nozzles installed in domestic BWR RPV. In water level instrumentation nozzles, 182 type nickel-base alloys were used for the welding joint to RPV. It is estimated the SCC potential is high in this joint because of a higher residual stress than the yield strength (about 400MPa). This report will describe a preventive maintenance method to these nozzles Heat Affected Zone (HAZ) and welds by a corrosion resistant cladding (CRC) by YAG Laser in underwater environment (without draining a reactor water). There are many kinds of countermeasures for SCC, for example, Induction Heating Stress Improvement (IHSI), Mechanical Stress Improvement Process (MSIP) and so on. A YAG laser CRC is one of them. In this technology a laser beam is used for heat source and irradiated through an optical fiber to a base metal and SCC resistant material is used for welding wires. After cladding the HAZ and welds are coated by the corrosion resistant materials so their surfaces are improved. A CRC by gas tungsten arc welding (GTAW) in an airenvironment had been developed and already applied to a couple of operating plants (16 Nozzles). This method was of course good but it spent much time to perform because of an installation of some water-proof working boxes to make a TIG-weldability environment. CRC by YAG laser welding in underwater environment has superior features comparing to this conventional TIG method as follows. At the viewpoint of underwater environment, (1) an outage term reduction (no drainage water). (2) a radioactive exposure dose reduction for personnel. At that of YAG laser welding, (1) A narrower HAZ. (2) A smaller distortion. (3) A few cladding layers. A YAG laser CRC test in underwater environment was carried out in the different welding position, horizontal, vertical upward and downward. The soundness of cladding layers (about 3 mm) is confirmed in visual and penetration test, and cross section observation. In the application to the actual plants, it is preferable to reduce the start and end point numbers of beads with which a defect is easy to cause. Therefore a special welding equipment for a YAG laser CRC that could weld continuously was developed.

Fracture Toughness Predictive Models Within the SOTERIA EU Project

2018 ◽  
Author(s):  
P. M. James ◽  
M. Berveiller
Keyword(s):  
Cleavage Fracture ◽  
Nuclear Power ◽  
Power Plants ◽  
Radiation Effects ◽  
Austenitic Stainless Steels ◽  
Irradiation Damage ◽  
Tensile Behaviour ◽  
Horizon 2020 ◽  
Term Operation

SOTERIA is focused on the ‘safe long term operation of light water reactors’. This will be achieved through an improved understanding of radiation effects in nuclear structural materials. This project has received funding from the European Union’s Horizon 2020 research and innovation programme under agreement No 661913. The overall aim of the SOTERIA project is to improve the understanding of the ageing phenomena occurring in ferritic reactor pressure vessel steels and in the austenitic internals in order to provide crucial information to regulators and operators to ensure safe long-term operation (LTO) of existing European nuclear power plants (NPPs). SOTERIA has set up a collaborative research consortium which gathers the main European research centers and industrial partners who will combine advanced modelling tools with the exploitation of experimental data to focus on two major objectives: i) to identify ageing mechanisms when materials face environmental degradation (such as e.g. irradiation and corrosion) and ii) to provide a single platform containing data and tools for reassessment of structural components during NPPs lifetime. This paper provides an overview of the ongoing activities within the SOTERIA Project that are contained within the analytical work-package (WP5.3). These fracture aspects are focused on the estimates of fracture in both ferritic steels and irradiation assisted stress corrosion cracking (IASCC) in austenitic stainless steels, under irradiated conditions. This analytical development is supported by analytical estimates of irradiation damage and the resulting changes in tensile behaviour of the steels elsewhere in SOTERIA, as well as a wider number of experimental programmes. Cleavage fracture estimates are being considered by a range of modelling estimates including the Beremin, Microstructurally Informed Brittle Fracture Model (MIBF), JFJ and Bordet Models with efforts being made to understand the influence of heterogeneity on the predicted toughness’s. Efforts are also being considered to better understand ductile void evolution and the effect of plasticity on the cleavage fracture predictions. IASCC is being modelled through the INITEAC code previously developed within the predecessor project Perform 60 which is being updated to incorporate recent developments from within SOTERIA and elsewhere.

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.

Study on a New Design Fatigue Evaluation Method

2015 ◽  
Author(s):  
Seiji Asada ◽  
Takashi Hirano ◽  
Takehiko Sera
Keyword(s):  
Evaluation Method ◽  
Austenitic Stainless Steels ◽  
Carbon Steels ◽  
Design Evaluation ◽  
Low Alloy Steels ◽  
Mean Stress ◽  
Research Committee ◽  
Fatigue Design ◽  
Fatigue Evaluation ◽  
Design Factors

In order to develop new design fatigue curves for austenitic stainless steels, carbon steels and low alloy steels and a new design fatigue evaluation method that is rational and has a clear design basis, the Design Fatigue Curve (DFC) subcommittee was established in the Atomic Energy Research Committee in the Japan Welding Engineering Society. Tentative design fatigue curves were developed and studies on the effects of mean stress and design factors are ongoing. Design fatigue curves, including the effects of mean stress and design factors, are needed to establish a new fatigue design evaluation method. This paper describes the study on the new fatigue design evaluation method.

scholarly journals INCEFA-PLUS (Increasing Safety in Nuclear Power Plants by Covering Gaps in Environmental Fatigue Assessment)

2016 ◽  
Author(s):  
Kevin Mottershead ◽  
Matthias Bruchhausen ◽  
Thomas Métais ◽  
Sergio Cicero ◽  
David Tice ◽  
...  
Keyword(s):  
Surface Finish ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Hold Time ◽  
Austenitic Stainless Steels ◽  
Mean Stress ◽  
Stress Strain ◽  
Assessment Procedures ◽  
Strain Hold

INCEFA-PLUS is a major new five year project supported by the European Commission HORIZON2020 program. The project commenced in mid 2015. 16 organizations from across Europe have combined forces to deliver new experimental data which will support the development of improved guidelines for assessment of environmental fatigue damage to ensure safe operation of nuclear power plants. Prior to the start of INCEFA-PLUS, an in-kind study was undertaken by several European organizations with the aim of developing the current state of the art for this technical area. In addition to stress/strain amplitude, this study identified three additional experimental variables which required further study in order to support improved assessment methodology for environmental fatigue, namely the effects of mean stress/strain, hold time and surface finish. Within INCEFA-PLUS, the effects of these three variables on fatigue endurance of austenitic stainless steels in light water reactor environments are therefore being studied experimentally. The data obtained will be collected and standardized in an online environmental fatigue database. A dedicated CEN workshop will deliver a harmonized data format facilitating the exchange of data within the project but also beyond. Based on the data generated and the resulting improvement in understanding, it is planned that INCEFA-PLUS will develop and disseminate methods for including the new data into assessment procedures for environmental fatigue degradation. This will take better account of the effects of mean stress/strain, hold time and surface finish. This paper will describe the background to the project and will explain the expectations for it.

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