European Project on Structural Integrity STYLE: Project Status

2013 ◽  
Author(s):  
Elisabeth Keim ◽  
Tomas Nicak
Keyword(s):  
Time Management ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Life Time ◽  
Nuclear Industry ◽  
Current Status ◽  
Special Focus ◽  
Term Operation ◽  
Integrity Assessment ◽  
Plant Safety

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 designs and for the GEN-III designs. A special focus is placed on reactor coolant pressure boundary (RCPB) components, because its integrity and functionality from the time of first operation until end of life is required to ensure plant safety. The overall objective of STYLE is to assess, optimize and develop the use of advanced tools for the structural integrity assessment of RCPB components relevant to ageing and life time management and to support the integration of the knowledge created in the project into mainstream nuclear industry assessment codes. The project concept is based on carefully selected research topics, which thematically cover the complex multidisciplinary character of structural assessment of RCPB components. The prioritization of the work reflects the needs of industrial end-users and assessment of currently available techniques and data at European and international level. This paper describes the current status of the project STYLE and summarizes its main results achieved up to date (Feb 2013). The project is coordinated by AREVA GmbH with 20 partner organizations from Europe, one collaborator from USA and one collaborator from Russia: AREVA GmbH, Germany (coordination, WP2 leader) AREVA SAS, France Bay Zoltan, Hungary CEA, France (WP1 leader) EDF, France EDF Energy Ltd., UK (WP7 leader) EK, Hungary IdS, France INR, Romania IWM, Germany JRC, Netherlands (WP4 leader) NRI, Czech Republic NRG, Netherlands SCK-CEN, Belgium AMEC, UK (WP3 and WP5 leader) University of Bristol, UK University of Manchester, UK Tecnatom, Spain Vattenfall, Sweden (WP6 leader) VTT, Finland ORNL, USA NIKIET, Russia

STYLE: Project Overview

2010 ◽  
Author(s):  
Tomas Nicak ◽  
Elisabeth Keim
Keyword(s):  
Time Management ◽  
Nuclear Power ◽  
Power Plants ◽  
Large Scale ◽  
Best Practice ◽  
Structural Integrity ◽  
Life Time ◽  
Assessment Tools ◽  
Integrity Assessment ◽  
University Of Bristol

The purpose of this paper is to introduce a new EUROATOM project focusing on the structural integrity assessment of reactor coolant pressure boundary components (RCPB) relevant to ageing and life time management. The project started in January 2010 and will last 4 years. The project is coordinated by AREVA NP GmbH with 20 partner organizations from Europe, one collaborator from USA and one collaborator from Russia: AEKI, Hungary; AREVA NP GmbH, Germany (coordination, WP2 leader); AREVA NP SAS, France; Bay Zoltan, Hungary; British Energy Generation Ltd., UK (WP7 leader); CEA, France (WP1 leader); EDF, France; IdS, France; INR, Romania; IWM, Germany; JRC, Netherlands (WP4 leader); NRI, Czech Republik; NRG, Netherlands; SCK-CEN, Belgium; Serco Assurance Technical Services, UK (WP3 and WP5 leader); University of Bristol, UK; University of Manchester, UK; Technatom, Spain; Vattenfall, Sweden (WP6 leader); VTT, Finland. Within STYLE (Structural integrity for lifetime management – non-RPV components) realistic failure models for some of the key components will be identified. The range of assessment tools considered will include those for assessment of component failure by advanced fracture mechanics analyses validated on small and large scale experiments, quantification of weld residual stresses by numerical analysis and by measurements, stress corrosion crack initiation/ growth effects and assessment of RCPB components (excluding the reactor pressure vessel) under dynamic and seismic loading. Based on theoretical and experimental results, performance assessment and further development of simplified engineering assessment methods (EAM) will be carried out considering both deterministic and probabilistic approaches. Integrity assessment case studies and large scale demonstration experiments will be performed on Mock-ups of safety-relevant components. These will include a repair weld in an aged butt-welded austenitic pipe, a dissimilar narrow gap TIG weld (following the EPR design) and a cladded ferritic pipe. Moreover experiments on specimens and feature test pieces will be carried out to support the large scale Mock-up analyses. The end product of the project (“STYLE TOOLS”) will comprise best practice guidelines on the use of advanced tools, on improvement and qualification of EAM as a part of European Leak-before-break (LBB) procedures and on life time management of the integrity of RCPB components in European nuclear power plants. The project will interact with the European Network of Excellence NULIFE.

scholarly journals Prognostics and Health Management in Nuclear Power Plants: An Updated Method-Centric Review With Special Focus on Data-Driven Methods

2021 ◽  
Vol 9 ◽  
Author(s):  
Xingang Zhao ◽  
Junyung Kim ◽  
Kyle Warns ◽  
Xinyan Wang ◽  
Pradeep Ramuhalli ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Health Management ◽  
Nuclear Industry ◽  
Data Availability ◽  
Data Driven ◽  
Special Focus ◽  
Prognostics And Health Management ◽  
Term Operation

In a carbon-constrained world, future uses of nuclear power technologies can contribute to climate change mitigation as the installed electricity generating capacity and range of applications could be much greater and more diverse than with the current plants. To preserve the nuclear industry competitiveness in the global energy market, prognostics and health management (PHM) of plant assets is expected to be important for supporting and sustaining improvements in the economics associated with operating nuclear power plants (NPPs) while maintaining their high availability. Of interest are long-term operation of the legacy fleet to 80 years through subsequent license renewals and economic operation of new builds of either light water reactors or advanced reactor designs. Recent advances in data-driven analysis methods—largely represented by those in artificial intelligence and machine learning—have enhanced applications ranging from robust anomaly detection to automated control and autonomous operation of complex systems. The NPP equipment PHM is one area where the application of these algorithmic advances can significantly improve the ability to perform asset management. This paper provides an updated method-centric review of the full PHM suite in NPPs focusing on data-driven methods and advances since the last major survey article was published in 2015. The main approaches and the state of practice are described, including those for the tasks of data acquisition, condition monitoring, diagnostics, prognostics, and planning and decision-making. Research advances in non-nuclear power applications are also included to assess findings that may be applicable to the nuclear industry, along with the opportunities and challenges when adapting these developments to NPPs. Finally, this paper identifies key research needs in regard to data availability and quality, verification and validation, and uncertainty quantification.

The Delivery of Technical Training Within the European Style Project

2013 ◽  
Author(s):  
Adam Toft ◽  
John Sharples
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Reactor Pressure Vessel ◽  
Nuclear Industry ◽  
Technical Training ◽  
Illustrative Case ◽  
Integrity Assessment ◽  
Leak Before Break ◽  
Reactor Pressure

The STYLE project considers structural integrity for lifetime management of non-reactor pressure vessel components of nuclear power plant. The project is funded under the seventh European Commission framework programme. A broad objective of the project is to assess, optimise and develop application of advanced tools for structural integrity assessment of reactor coolant pressure boundary components other than the reactor pressure vessel. One aspect of the STYLE project is intended to address the issue of succession planning within the European nuclear industry. With many key technical experts now approaching retirement it is essential to progress the technical expertise of those at an earlier stage of their career in the industry. The paper describes how technical training has been delivered as an integral part of the STYLE project to support retention of the current level of technical capability in future. Diverse aspects of training are described. These include participation in experimental work, numerical modelling and simulation, application of engineering assessment procedures, leak-before-break, probabilistic fracture mechanics and materials behaviour. An illustrative case study is described, in which trainees received practical instruction in the essential steps for technical justification of a leak-before-break argument.

The Importance of a Strong Training Element Within the European STYLE Project

2012 ◽  
Author(s):  
Adam Toft ◽  
John Sharples
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Reactor Pressure Vessel ◽  
Nuclear Industry ◽  
Framework Programme ◽  
Integrity Assessment ◽  
Future Viability ◽  
European Nuclear ◽  
Reactor Pressure

With many key technical experts within the European nuclear industry now approaching retirement, the continued training and professional development of less experienced people is vital for the future viability of the industry. Consequently, European framework programme projects are including a strong training element within their work packages. The STYLE project considers structural integrity for lifetime management of non-reactor pressure vessel components of nuclear power plant. The project is funded under the seventh European Commission framework programme. The objective of the project is to assess, optimise and develop application of advanced tools for structural integrity assessment of reactor coolant pressure boundary components other than the reactor pressure vessel.

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.

Achievements of the European Style Project Technical Training Programme

2014 ◽  
Author(s):  
Adam Toft ◽  
John Sharples
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Training Programme ◽  
Reactor Pressure Vessel ◽  
Nuclear Industry ◽  
Integrity Assessment ◽  
European Nuclear ◽  
Broad Objective ◽  
Reactor Pressure

The STYLE project considers structural integrity for lifetime management of non-reactor pressure vessel components of nuclear power plant. The project is funded under the seventh European Commission framework programme. A broad objective of the project is to assess, optimise and develop application of advanced tools for structural integrity assessment of reactor coolant pressure boundary components other than the reactor pressure vessel. One aspect of the STYLE project is intended to address the issue of succession planning within the European nuclear industry. With many key technical experts now approaching retirement it is essential to progress the technical expertise of those at an earlier stage of their career in the industry.

Effect of Creep on the Residual Stresses in Tube-to-Tubesheet Joints

2010 ◽  
Author(s):  
Nor Eddine Laghzale ◽  
Abdel-Hakim Bouzid
Keyword(s):  
Contact Pressure ◽  
Analytical Model ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Environmental Safety ◽  
Maintenance Cost ◽  
Joint Interface ◽  
Plant Safety ◽  
Material Creep ◽  
Nuclear Power Plant Safety

Steam generators are the subject of major concern in nuclear power plant safety. Within these generators, in addition to the structural integrity, the gross tightness barrier, which separates the primary and secondary circuits, is primarily ensured by the presence of a residual contact pressure at the tube-to-tubesheet joint interface. Any leakage is unacceptable, and its consequences are very heavy in terms of the human and environmental safety as well as maintenance cost. Some studies have been conducted to understand the main reasons for such a failure. However, no analytical model able to predict the attenuation of the residual contact pressure under the effect of material creep relaxation behavior. The development of a simple analytical model able to predict the change of the residual contact pressure as a function of time is laid out in this paper. The results from the analytical model are checked and compared with those of finite elements.

Structural Integrity for Nuclear Power Plant Against Excessive Load on Design Extension Condition

2013 ◽  
Author(s):  
Daigo Watanabe ◽  
Kiminobu Hojo
Keyword(s):  
Nuclear Power ◽  
Structural Integrity ◽  
Safety Factors ◽  
Design Code ◽  
Acceptance Criteria ◽  
Integrity Assessment ◽  
Current Design ◽  
Factor Design ◽  
Excessive Load ◽  
Load And Resistance Factor

This paper introduces an example of structural integrity evaluation for Light Water Reactor (LWR) against excessive loads on the Design Extension Condition (DEC). In order to assess the design acceptance level of DEC, three acceptance criteria which are the stress basis limit of the current design code, the strain basis limit of the current design code and the strain basis limit by using Load and Resistance Factor Design (LRFD) method were applied. As a result the allowable stress was increased by changing the acceptance criteria from the stress basis limit to the strain basis limit. It is shown that the practical margin of the LWR’s components still keeps even on DEC by introducing an appropriate criterion for integrity assessment and safety factors.

Structural Integrity Assessment of Steam Generator Tubes Using a New EPRI Statistical Approach

2002 ◽  
Author(s):  
Carlos Alexandre de Jesus Miranda ◽  
Miguel Mattar Neto
Keyword(s):  
Steam Generator ◽  
Wall Thickness ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Tube Wall ◽  
New Approach ◽  
Inconel 600 ◽  
Integrity Assessment ◽  
Tube Wall Thickness ◽  
Uniform Tube

A fundamental step in tube plugging management of a Steam Generator (SG), in a Nuclear Power Plant (NPP), is the tube structural integrity evaluation. The degradation of SG tubes may be considered one of the most serious problems found in PWRs operation, mainly when the tube material is the Inconel 600. The first repair criterion was based on the degradation mode where a uniform tube wall thickness corrosion thinning occurred. Thus, a requirement of a maximum depth of 40% of the tube wall thickness was imposed for any type of tube damage. A new approach considers different defects arising from different degradation modes, which comes from the in-service inspections (NDE) and how to consider the involved uncertainties. It is based on experimental results, using statistics to consider the involved uncertainties, to assess structural limits of PWR SG tubes. In any case, the obtained results, critical defect dimensions, are within the regulatory limits. In this paper this new approach will be discussed and it will be applied to two cases (two defects) using typical data of SG tubes of one Westinghouse NPP. The obtained results are compared with ‘historical’ approaches and some comments are addressed from the results and their comparison.

scholarly journals Inclusion of Warm Pre-Stress Concept in French RPV Structural Integrity Assessment

2016 ◽  
Author(s):  
Dominique Moinereau ◽  
Malik Ait-Bachir ◽  
Stéphane Chapuliot ◽  
Stéphane Marie ◽  
Clémentine Jacquemoud ◽  
...  
Keyword(s):  
Brittle Fracture ◽  
Fracture Resistance ◽  
Large Scale ◽  
Structural Integrity ◽  
Stress Effect ◽  
Local Approach ◽  
Experimental Database ◽  
Term Operation ◽  
Integrity Assessment ◽  
Brittle Fracture Resistance

Evaluation of the fracture resistance of nuclear reactor pressure vessel (RPV) regarding the risk of brittle fracture is a key point in the structural integrity assessment of the component (RPV). Such approach is codified in French RSE-M code, based on a very conservative methodology. With respect to long term operation, an improvement of the present methodology is necessary and in progress to reduce this conservatism. One possible significant improvement is the inclusion of the warm pre-stress (WPS) concept in the assessment. After a short description of the WPS concept, the process engaged in France to allow inclusion of WPS in the integrity assessment is presented. In a first step, experimental and numerical studies have been conducted in France by EDF, CEA and AREVA (also including international collaborations and projects) to demonstrate and validate the beneficial effect of WPS on the brittle fracture resistance of RPV steels. A large panel of experimental results and data is now available obtained on small, medium and large scale specimens on representative RPV steels (including highly irradiated RPV materials). These data have been included in a specific WPS experimental database. Main experiments have been interpreted by refined computations, based on elastic plastic analyses and local approach to cleavage fracture. In a second step, a new criterion (ACE criterion) has been proposed by French organizations (AREVA, CEA and EDF) for an easy simplified evaluation of warm pre-stress effect on the brittle fracture resistance of RPV steels. Accuracy and conservatism of the criterion is verified by comparison to experimental data results and numerical analyses. Finally, implementation of the WPS effect in the French RSE-M code (for in service assessment) is in progress, based on the ACE criterion. The present paper summarizes all these steps leading to codification of WPS in RSE-M code.

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