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.

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.

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 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 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

INCEFA-PLUS: Increasing Safety in NPPs by Covering Gaps in Environmental Fatigue Assessment

2018 ◽  
Author(s):  
Kevin Mottershead ◽  
Matthias Bruchhausen ◽  
Sam Cuvilliez ◽  
Sergio Cicero
Keyword(s):  
Strain Amplitude ◽  
Surface Finish ◽  
Nuclear Power ◽  
Power Plants ◽  
Phase 1 ◽  
Hold Time ◽  
Austenitic Stainless Steels ◽  
Common Interest ◽  
Horizon 2020 ◽  
Mean Strain

INCEFA-PLUS is a five-year project supported by the European Commission HORIZON 2020 programme. The project commenced in mid-2015. Sixteen organisations 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. Within INCEFA-PLUS, the effects of mean strain and stress, hold time, strain amplitude and surface finish on fatigue endurance of austenitic stainless steels in light water reactor environments are being studied experimentally, these being issues of common interest to all participants. The data obtained is being collected and standardised in an online environmental fatigue database, implemented with the assistance of an INCEFA-PLUS led CEN (European Committee for Standardization) workshop on this aspect. Later in the project it is planned that INCEFA-PLUS will develop and disseminate methods for including the new data into assessment approaches for environmental fatigue degradation. This paper provides an overall update to project developments since it was last presented at PVP2017 [[1]]. As well as being a standalone paper, the paper will also serve as an introduction to more detailed papers also being submitted covering 4 specific aspects of the project. In particular, this paper summarises: • The results for Phase 1 testing. • The agreed plans for Phase 2 testing • A summary of emerging sensitivities to, and inter-dependencies between: ○ Mean strain and stress, surface finish, strain amplitude and hold time. ○ Environment ○ Material ○ Laboratory • Latest thinking on direction for the project in its last two years. • The latest thoughts on how the project results will be used to advance development of improved assessment guidelines. • A summary of dissemination achieved and planned for the forthcoming year.

Lessons Learned in Implementation of Guidelines for Quantifying Support System Initiating Event Frequencies

2008 ◽  
Author(s):  
Rupert A. Weston ◽  
Ashley J. Mossa
Keyword(s):  
Support System ◽  
Nuclear Power ◽  
Power Plants ◽  
Fault Tree ◽  
Lessons Learned ◽  
Pressurized Water ◽  
Common Cause Failures ◽  
Areas Of Concern ◽  
Pilot Implementation ◽  
American Society

The pilot implementation results for Regulatory Guide 1.200 identified four probabilistic risk assessment (PRA) technical elements that required additional guidance. One of these elements involved the use of fault tree technique to quantify the frequencies of support system initiating events (SSIEs). To address this technical element, guidelines were developed by the Electric Power Research Institute (EPRI) to provide a common industry approach for addressing the identification and quantification of SSIEs. The EPRI guidelines were issued as an interim report to allow trial use and pilot implementation by the industry prior to finalizing the guidelines. These interim guidelines provide an industry-consensus approach for addressing areas of concern in the development of support system initiating event models to ensure that the associated supporting requirements of the American Society of Mechanical Engineers (ASME) PRA Standard for internally initiated events are satisfied. A Pressurized Water Reactor Owners Group (PWROG) pilot implementation of the EPRI interim guidelines was conducted to determine whether the pilot participants have adequately addressed all areas of concern in the development of SSIE models. To determine this, a SSIE model currently used was selected by each of four the pilot participants and subject to detail review to demonstrate whether these models meet the expectations of the EPRI interim guidelines. The EPRI interim guidelines identified the areas of concern to be addressed in using fault tree technique to develop and quantify SSIE models. The guidelines addressed several areas of concern including the treatment of passive failures, the assignment of an appropriate mission time for primary and secondary failures, treatment of common cause failures (CCFs) between running and standby equipment, and consideration of all combinations of CCFs. The PWROG pilot implementation of the interim guidelines summarized the lessons learned and provided feedback to EPRI for consideration in finalizing the guidelines. In addition to the compilation of lessons learned, the PWROG implementation of the EPRI interim guidelines identified existing practices used to develop fault tree models for quantifying SSIE frequencies. Such practices did not necessarily follow a common approach and did not fully meet the expectations of the interim guidelines. Detailed reviews of the SSIE models currently in use at nuclear power plants (NPPs) for the pilot participants demonstrated that the elements of evaluation described in the interim guidelines were not addressed consistently among the PWROG pilot participants. Recommended improvements were identified and incorporated in the SSIE models to meet the expectations of the EPRI interim guidelines. The re-quantification of SSIE frequencies based on the recommended improvements, demonstrated that by not adequately addressing all elements in the evaluation, the SSIE frequency may be under-estimated.

scholarly journals The COLIBRI experimental program in the CROCUS reactor: characterization of the fuel rods oscillator

2020 ◽  
Vol 225 ◽  
pp. 04020 ◽  
Author(s):  
Vincent Lamirand ◽  
Pavel Frajtag ◽  
Daniel Godat ◽  
Oskari Pakari ◽  
Axel Laureau ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Oscillation Amplitude ◽  
Experimental Program ◽  
Pressurized Water ◽  
Fuel Rods ◽  
Horizon 2020 ◽  
Neutron Noise ◽  
Water Reactors

The present article presents the mechanical characterization of the fuel rods oscillator developed for the purposes of the COLIBRI experimental program in CROCUS. COLIBRI aims at investigating the radiation noise related to fuel vibrations. The main motivation is the increased amplitudes in the neutron noise distributions recorded in ex- and in-core detectors that have been observed in recent years in Siemens pre-Konvoi type of pressurized water reactors. Several potential explanations have been put forward, but no definitive conclusions could yet be drawn. Among others, changes in fuel assembly or pin vibration patterns, due to recent modifications of assembly structural designs, were pointed out as a possible cause. Computational dynamic tools are currently developed within the Horizon 2020 European project CORTEX, to help with understanding the additional noise amplitude. The COLIBRI program is used for their validation. An in-core device was designed, tested, and licensed between 2015 and 2019 for fuel rods oscillation in CROCUS, in successive steps from out-of-pile tests with dummy fuel rods to critical in-core tests. The characterization of its mechanical behavior is presented, in air and in water, and as a function of the load, for safety and experimental purposes. The device allows simultaneously oscillating up to 18 fuel rods. The maximum oscillation amplitude is 5 mm, while the maximum allowed frequency is 2 Hz, i.e. in the frequency range in which the induced neutron flux fluctuations are most pronounced in nuclear power plants.

Post-Fukushima Assessment of the AP1000® Plant

2014 ◽  
Author(s):  
Terry L. Schulz ◽  
Thomas A. Kindred ◽  
Bryan N. Friedman ◽  
John E. Glavin ◽  
Adam D. Malinowski ◽  
...  
Keyword(s):  
Nuclear Power Station ◽  
Nuclear Power ◽  
Power Plants ◽  
Lessons Learned ◽  
Fuel Oil ◽  
Passive Safety ◽  
Final Report ◽  
Power Station ◽  
Pressurized Water ◽  
Passive Coping

The AP1000 plant is an 1100-MWe class pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance, and safety with reduced plant costs. The AP1000 passive nuclear power plant is uniquely equipped to withstand an extended station blackout scenario such as the events following the earthquake and tsunami at the Fukushima Daiichi nuclear power station without compromising core and containment integrity. Without AC power, using passive safety technology, the AP1000 plant provides cooling for the core, containment and spent fuel pool for more than 3 days without the need for operator actions. Following this passive coping period, minimal operator actions are needed to extend the operation of the passive features to 7 days using installed equipment. With the re-supply of fuel oil the coping time may be extended for an indefinite time. Connections for a few, small, easily transportable components provide a diverse backup means of extending passive system operation after the first 3 days. As a result, the AP1000 design provides very robust protection of public safety and the utility investment. Following the accident at the Fukushima Dai-ichi nuclear power station in Japan, several initiatives were launched worldwide to assess the lessons learned. These include, but are not limited to, the European Nuclear Safety Regulators Group (ENSREG) stress tests, the Office for Nuclear Regulation (ONR) Final Report, the International Atomic Energy Agency (IAEA) Expert Mission Report, and the U.S. NRC Near-Term Task Force Recommendations. The AP1000 design has been assessed against these initiatives and lessons learned. The purpose of this paper is to describe: • How the accident at the Fukushima Dai-ichi nuclear power station was evaluated and translated into conclusions and recommendations for nuclear power plants worldwide • How the AP1000 plant was evaluated in light of the recommendations resulting from the various post-Fukushima assessments • The key conclusions resulting from the post-Fukushima evaluation of the AP1000 design

Lessons Learned in Reviewing CCF Events

2009 ◽  
Author(s):  
Ashley Mossa ◽  
Rupert Weston
Keyword(s):  
Information Exchange ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Lessons Learned ◽  
Level Of Detail ◽  
Data Sets ◽  
Pressurized Water ◽  
Data Set ◽  
Multiple Data Sets

The U.S. Nuclear Regulatory Commission (NRC) has an ongoing Common Cause Failure (CCF) data analysis program that periodically collects and evaluates information on component failures at U.S. commercial Nuclear Power Plants (NPPs). The primary information sources include the Licensee Event Reports (LER) and records from the Equipment Performance Information Exchange (EPIX) program. Once the information is collected, the failure records are evaluated to identify potential CCF events. CCF events are then coded, reviewed, and loaded into the NRC’s database. Verification of the CCF events is performed with the intended purpose of ensuring that events entered into the CCF database are indeed CCF events and that the event coding is consistent and correct. To ensure technical accuracy and correctness of the events loaded into the CCF database, the NRC requested the Pressurized Water Reactors Owners Group (PWROG) support in reviewing these events. Reviews of multiple data sets of CCF events were conducted on behalf of the PWROG. The data sets included CCF events that have occurred at U.S. commercial nuclear power plants. CCF events that occurred during 2006 through 2007 were included in the most recent data set that was reviewed. The level of information provided for reported CCF events varies from utility-to-utility. Without utility participation or input, the lack of consistency and varying level of detail can lead to incorrect interpretation and classification of a CCF event regarding its Probabilistic Risk Assessment (PRA) impact. This paper offers lessons learned from the reviews that were conducted. Insights for improving the consistency and level of detail related to the PRA information are summarized in this paper. The leading causes of initial misclassification of CCF events and patterns observed in conducting the reviews are discussed. The resolutions of misclassified CCF events are also discussed as part of the evaluation process to enhance the pedigree of the CCF database.

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