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.

Incefa-Plus Project: Lessons Learned From the Project Data and Impact on Existing Fatigue Assessment Procedures

2020 ◽  
Vol 142 (6) ◽  
Author(s):  
Sam Cuvilliez ◽  
Alec McLennan ◽  
Kevin Mottershead ◽  
Jonathan Mann ◽  
Matthias Bruchhausen
Keyword(s):  
Surface Finish ◽  
Environmental Effects ◽  
Nuclear Power ◽  
Power Plants ◽  
Lessons Learned ◽  
Pressurized Water ◽  
Fatigue Assessment ◽  
Horizon 2020 ◽  
Project Data ◽  
Assessment Procedures

Abstract This work focuses on the analysis of the data generated during the INCEFA+ project (INcreasing safety in nuclear power plants (NPPs) by Covering gaps in Environmental Fatigue Assessment, a five-year project supported by the European Commission Horizon 2020 program). 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 subfactor 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 pressurized water reactor (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 environmentally assisted fatigue (EAF) assessment, but may also depend on the environmental correction Fen factor expression that is used to take environmental effects into account.

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.

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.

Incefa-Plus Project: Lessons Learned From the Project Data and Impact on Existing Fatigue Assessment Procedures

2021 ◽  
Author(s):  
Sam Cuvilliez ◽  
Alec Mclennan ◽  
Jonathan Mann ◽  
Matthias Bruchhausen ◽  
Kevin Mottershead
Keyword(s):  
Lessons Learned ◽  
Fatigue Assessment ◽  
Project Data ◽  
Assessment Procedures

Expectations for Inservice Testing Programs at New Nuclear Power Plants

2017 ◽  
Author(s):  
Thomas G. Scarbrough
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Reactor Core ◽  
Nuclear Regulatory Commission ◽  
Lessons Learned ◽  
Mechanical Equipment ◽  
American Society ◽  
Regulatory Commission ◽  
Safety Systems

In a series of Commission papers, the U.S. Nuclear Regulatory Commission (NRC) described its policy for inservice testing (IST) programs to be developed and implemented at nuclear power plants licensed under 10 CFR Part 52. This paper discusses the expectations for IST programs based on those Commission policy papers as applied in the NRC staff review of combined license (COL) applications for new reactors. For example, the design and qualification of pumps, valves, and dynamic restraints through implementation of American Society of Mechanical Engineers (ASME) Standard QME-1-2007, “Qualification of Active Mechanical Equipment Used in Nuclear Power Plants,” as accepted in NRC Regulatory Guide (RG) 1.100 (Revision 3), “Seismic Qualification of Electrical and Active Mechanical Equipment and Functional Qualification of Active Mechanical Equipment for Nuclear Power Plants,” will enable IST activities to assess the operational readiness of those components to perform their intended functions. ASME has updated the Operation and Maintenance of Nuclear Power Plants (OM Code) to improve the IST provisions for pumps, valves, and dynamic restraints that are incorporated by reference in the NRC regulations with applicable conditions. In addition, lessons learned from performance experience and testing of motor-operated valves (MOVs) will be implemented as part of the IST programs together with application of those lessons learned to other power-operated valves (POVs). Licensee programs for the Regulatory Treatment of Non-Safety Systems (RTNSS) will be implemented for components in active nonsafety-related systems that are the first line of defense in new reactors that rely on passive systems to provide reactor core and containment cooling in the event of a plant transient. This paper also discusses the overlapping testing provisions specified in ASME Standard QME-1-2007; plant-specific inspections, tests, analyses, and acceptance criteria; the applicable ASME OM Code as incorporated by reference in the NRC regulations; specific license conditions; and Initial Test Programs as described in the final safety analysis report and applicable RGs. Paper published with permission.

Information Technology and the Development of a Global Safety Culture

2011 ◽  
pp. 558-581
Author(s):  
Susan L. Rothwell
Keyword(s):  
Information Technology ◽  
Safety Culture ◽  
Nuclear Power ◽  
Power Plants ◽  
Health And Safety ◽  
Lessons Learned ◽  
Energy Independence ◽  
Power Safety ◽  
Under Construction ◽  
Culture Development

A nuclear power plant is one of the most complex sociotechnical systems ever created, with operation requiring multiple organizations, extensive interaction, and a mission to protect public health and safety. A strong global nuclear power safety culture is important, with over 400 nuclear power plants worldwide and more under construction to reduce fossil fuel dependency. We increasingly rely on technology, stressing our need for energy independence, security, reliability, education, and safety. Lessons learned from nuclear power safety culture development have a large potential audience. Unfortunately, the complexity of nuclear power and restricted access to operational data have limited outside research on and understanding of nuclear power safety culture. This chapter provides a conceptual, methodological, empirical, and operational perspective on the development of commercial nuclear power safety culture, focusing on the role of information technology (IT) in building, maintaining, and expanding global nuclear power safety culture.

scholarly journals Reactor performance, system reliability, instrumentation and control

2020 ◽  
Vol 6 ◽  
pp. 43
Author(s):  
Andreas Schumm ◽  
Madalina Rabung ◽  
Gregory Marque ◽  
Jary Hamalainen
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Lessons Learned ◽  
Software Systems ◽  
Reactor Performance ◽  
Horizon 2020 ◽  
Term Operation ◽  
In The Beginning ◽  
Electrical Cables

We present a cross-cutting review of three on-going Horizon 2020 projects (ADVISE, NOMAD, TEAM CABLES) and one already finished FP7 project (HARMONICS), which address the reliability of safety-relevant components and systems in nuclear power plants, with a scope ranging from the pressure vessel and primary loop to safety-critical software systems and electrical cables. The paper discusses scientific challenges faced in the beginning and achievements made throughout the projects, including the industrial impact and lessons learned. Two particular aspects highlighted concern the way the projects sought contact with end users, and the balance between industrial and academic partners. The paper concludes with an outlook on follow-up issues related to the long term operation of nuclear power plants.

scholarly journals “True Believers” or Numerical Terrorism at the Nuclear Power Plant

2011 ◽  
Vol 231 (5-6) ◽  
Author(s):  
Walter Krämer ◽  
Gerhard Arminger
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Childhood Cancer ◽  
Political Party ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
The Political ◽  
Heated Debate ◽  
Data Points

SummaryFor decades, there has been a heated debate about whether or not nuclear power plants contribute to childhood cancer in their respective neighbourhoods, with statisticians testifying on both sides. The present paper points to some flaws in the pro-arguments, taking a recent study prepared for the political party “Bündnis 90 /Grüne” as a specimen. Typical mistakes include an understatement of the size of tests of significance, disregard of important covariates and extreme reliance on very few selected data points.

Research of Screening and Grading Method of Ageing Sensitive Objectives in Nuclear Power Plant

2014 ◽  
Vol 543-547 ◽  
pp. 858-861
Author(s):  
Xiao Tian Liu ◽  
Yong Wang ◽  
Shao Rui Niu ◽  
Yan Zhao Zhang ◽  
Zhen Hao Shi ◽  
...  
Keyword(s):  
Fuzzy Logic ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Lessons Learned ◽  
Nonlinear Coupling ◽  
Highly Nonlinear ◽  
Ageing Management ◽  
First Time

This first step of ageing management in nuclear power plant is to determine the objectives and their priorities. The characteristics of the objectives are complex and highly nonlinear coupling. A fuzzy logic based screening and grading method have been developed in this research for the first time which combined the genetic ageing lessons learned and field expert experience to resolve the problem. The method have been approved of highly applicability and applied to ageing management in multiple nuclear power plants.

Post-Fukushima Research and Development Strategies and Priorities for Water Cooled Reactor Technology Development

2016 ◽  
Author(s):  
Katsumi Yamada ◽  
Abdallah Amri ◽  
Lyndon Bevington ◽  
Pal Vincze
Keyword(s):  
Research And Development ◽  
International Organizations ◽  
Nuclear Power ◽  
Power Plants ◽  
Development Strategies ◽  
Lessons Learned ◽  
Member States ◽  
North East ◽  
Severe Accidents ◽  
Fukushima Daiichi

The Great East Japan Earthquake and the subsequent tsunami on 11 March 2011 initiated accident conditions at several nuclear power plants (NPPs) on the north-east coast of Japan and developed into a severe accident at the Fukushima Daiichi NPP, which highlighted a number of nuclear safety issues. After the Fukushima Daiichi accident, new research and development (R&D) activities have been undertaken by many countries and international organizations relating to severe accidents at NPPs. The IAEA held, in cooperation with the OECD/NEA, the International Experts’ Meeting (IEM) on “Strengthening Research and Development Effectiveness in the Light of the Accident at the Fukushima Daiichi Nuclear Power Plant” at IAEA Headquarters in Vienna, Austria, 16–20 February 2015. The objective of the IEM was to facilitate the exchange of information on these R&D activities and to further strengthen international collaboration among Member States and international organizations. One of the main conclusions of the IEM was that the Fukushima Daiichi accident had not identified completely new phenomena to be addressed, but that the existing strategies and priorities for R&D should be reconsidered. Significant R&D activities had been already performed regarding severe accidents of water cooled reactors (WCRs) before the accident, and the information was very useful for predicting and understanding the accident progression. However, the Fukushima Daiichi accident highlighted several challenges that should be addressed by reconsidering R&D strategies and priorities. Following this IEM, the IAEA invited several consultants to IAEA Headquarters, Vienna, Austria, 12–14 May 2015, and held a meeting in order to discuss proposals on possible IAEA activities to facilitate international R&D collaboration in relation to severe accidents and how to effectively disseminate the information obtained at the IEM. The IAEA also held Technical Meeting (TM) on “Post-Fukushima Research and Development Strategies and Priorities” at IAEA Headquarters, Vienna, Austria, 15–18 December 2015. The objective of this meeting was to provide a platform for experts from Member States and international organizations to exchange perspectives and information on strategies and priorities for R&D regarding the Fukushima Daiichi accident and severe accidents in general. The experts discussed R&D topic areas that need further attention and the benefits of possible international cooperation. This paper discusses lessons learned from the Fukushima Daiichi accident based on the presentations and discussions at the meetings mentioned above, and identifies the needs for further R&D activities to develop WCR technologies to cope with Fukushima Daiichi-type accidents.

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