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.

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

Overview of Recent NUGENIA Activities Particularly in Relation to Structural Integrity

2015 ◽  
Author(s):  
John Sharples ◽  
Elisabeth Keim
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Nuclear Research ◽  
Joint Research ◽  
Non Profit ◽  
Integrity Assessment ◽  
Industry Research ◽  
Implementation And Dissemination ◽  
Technical Area

NUGENIA, an international non-profit association founded under Belgian legislation and launched in March 2012, is dedicated to nuclear research and development (R&D) with a focus on Generation II and III power plants. NUGENIA is the integrated framework between industry, research and safety organisations for safe, reliable and competitive nuclear power production, and is aimed at running an open innovation marketplace, to promote the emergence of joint research and to facilitate the implementation and dissemination of R&D results. The technical scope of NUGENIA consists of eight technical areas. One of these areas, Technical Area 4, is associated with the structural integrity assessment of systems, structures and components. A brief overview of recent NUGENIA activities in general is provided in this paper and a specific focus is given on developments in relation to Technical Area 4.

Advanced Structural Integrity Assessment Tools for Safe Long Term Operation: ATLAS+ Project — Status of the Activities of the WP3 on Modelling

2019 ◽  
Author(s):  
Stéphane Marie ◽  
Arnaud Blouin ◽  
Tomas Nicak ◽  
Dominique Moinereau ◽  
Anna Dahl ◽  
...  
Keyword(s):  
Large Scale ◽  
Structural Integrity ◽  
Assessment Tools ◽  
Local Approach ◽  
Coupled Models ◽  
Large Defect ◽  
Ductile Tearing ◽  
Term Operation ◽  
Integrity Assessment

Abstract The main objective and mission of the ATLAS+ project is to develop advanced structural assessment tools to address the remaining technology gaps for the safe and long term operation of nuclear reactor pressure coolant boundary systems. ATLAS+ WP3 focuses mainly on ductile tearing prediction for large defect in components: Several approaches have been developed to accurately model the ductile tearing process and to take into account phenomena such as the triaxiality effect, or the ability to predict large tearing in industrial components. These advanced models include local approach coupled models or advanced energetic approaches. Unfortunately, the application of these tools is today rather limited to R&D expertise. However, because of the continuous progress in the performance of the calculation tools and accumulated knowledge, in particular by members of ATLAS+, these models can now be considered as relevant for application in the context of engineering assessments. WP3 will therefore: • Illustrate the implementation of these models for industrial applications through the interpretation of large scale mock-ups (with cracks in weld joints for some of them), • Make recommendations for the implementation of the advanced models in engineering assessments, • Correct data from the conventional engineering approach by developing a methodology to produce J-Δa curve suitable case by case, based on local approach models, • Improve the tools, guidance and procedures for undertaking leak-before-break (LBB) assessments of piping components, particularly in relation to representing structural representative fracture toughness J-Resistance curves and the influence of weld residual stresses. To achieve these goals, WP3 is divided into 4 sub-WPs and this paper presents the progress of the work performed in each sub-WP after 24 months of activities.

The Experimental Work in Support to ATLAS+ Project

2018 ◽  
Author(s):  
Dominique Moinereau ◽  
Patrick Le Delliou ◽  
Anna Dahl ◽  
Yann Kayser ◽  
Szabolcs Szavai ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Experimental Work ◽  
Large Scale ◽  
Nuclear Reactor ◽  
Structural Integrity ◽  
Assessment Tools ◽  
Ductile Material ◽  
Small Scale ◽  
Term Operation ◽  
Integrity Assessment

The 4-years European project ATLAS+ project was launched in June 2017. Its main objective is to develop advanced structural assessment tools to address the remaining technology gaps for the safe and long term operation of nuclear reactor pressure coolant boundary systems. The transferability of ductile material properties from small scale fracture mechanics specimens to large scale components is one of the topics of the project. A large programme of experimental work is to be conducted in support of the development and validation of advanced tools for structural integrity assessment within the framework of the work-package 1 (WP 1): Design and execution of simulation oriented experiments to validate models at different scales. The experimental work is based on a full set of fracture mechanics experiments conducted on standard specimens and large scale components (several pipes and one mock-up), including a full materials characterization. Three materials are considered: • a ferritic steel 15NiCuMoNb5 (WB 36) • an aged austenitic stainless steel weld • a VVER (eastern PWR) dissimilar metal weld (DMW) The paper presents the WP 1, the experimental programme and summarizes the first results.

Nuclear Reactor Pressure Vessel Integrity Assessment: Enhancement Provided by 3D Modeling Flaw Stability Calculations

2013 ◽  
Author(s):  
Adolfo Arrieta-Ruiz ◽  
Eric Meister ◽  
Henriette Churier
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Structural Integrity ◽  
Reactor Pressure Vessel ◽  
Integrity Assessment ◽  
Potential Damage ◽  
Vessel Integrity ◽  
Reactor Pressure

Structural integrity of the Reactor Pressure Vessel (RPV) is one of the main considerations regarding safety and lifetime of Nuclear Power Plants (NPP) since this component is considered as not reasonably replaceable. Brittle fracture risk associated with the embrittlement of RPV steel in irradiated areas is the main potential damage. In France, deterministic integrity assessment for RPV is based on the crack initiation stage. The stability of an under-clad postulated flaw in the core area is currently evaluated under a Pressurized Thermal Shock (PTS) through a fracture mechanics simplified method. One of the axes of EDF’s implemented strategy for NPP lifetime extension is the improvement of the deterministic approach with regards to the input data and methods so as to reduce conservatisms. In this context, 3D finite element elastic-plastic calculations with flaw modelling have been carried out recently in order to quantify the enhancement provided by a more realistic approach in the most severe events. The aim of this paper is to present both simplified and 3D modelling flaw stability evaluation methods and the results obtained by running a small break LOCA event.

Life Time Management: A Practical Approach of Nuclear Power Plants

2002 ◽  
Author(s):  
Jaroslav Bartonicek ◽  
Klaus-Juergen Metzner ◽  
Friedrich Schoeckle
Keyword(s):  
Time Management ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Mechanical Systems ◽  
Life Time ◽  
Practical Approach ◽  
Degradation Mechanisms ◽  
High Level ◽  
Quality Status

A comprehensive life time management has to take care of all safety and availability relevant components in nuclear power plants, with different intensity, of course. For instance, mechanical systems and components can be basically classified/ranked into three different groups: (1): The quality status of the components in this group has to be guaranteed on a pre-defined (high) level. (2): The quality status of the components in this group has to be maintained on its actual level. (3): Other components with no specific quality demands. Regarding the first group, integrity has to be guaranteed. Therefore it is necessary to monitor the possible root causes of degradation mechanisms during plant operation; thus the degradation effects can be assessed and — more important — controlled to maintain the safety standard on the demanded high level without any compromise. The monitoring of consequences of degradation mechanisms is being performed as an additional redundant measure. The requirements to maintain the quality status of the second group of components can be fulfilled by monitoring of the consequences of operational degradation mechanisms to be performed by preventive maintenance activities, in terms of tests, inspections and repairs, using either time dependant procedures or component condition orientated methods. For the third group of components, no preventive action is necessary. However, failures and malfunctions have to be assessed statistically to avoid a reduction of the required basic component quality. In the first two groups all safety relevant components and systems are included. Generally, aging management programs cover these two groups of components; life time management covers all of above groups. This paper concentrates on mechanical systems and components; it summarizes the practical approach to life time management as it is realized in German nuclear power plants. The application is discussed using dedicated examples.

Structural Integrity Assessment of Doel 3 and Tihange 2 RPVs Affected by Hydrogen Flakes: Refined X-FEM Analyses

2016 ◽  
Author(s):  
Pierre Dulieu ◽  
Valéry Lacroix
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Extended Finite Element Method ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Extended Finite Element ◽  
Integrity Assessment ◽  
Safety Case ◽  
Component Manufacturing ◽  
Reactor Pressure

During the 2012 outage at Doel 3 and Tihange 2 Nuclear Power Plants, specific ultrasonic in-service inspections revealed a large number of quasi-laminar indications in the base metal of the reactor pressure vessels, mainly in the lower and upper core shells. The observed indications could subsequently be attributed to hydrogen flaking induced during the component manufacturing process. As a consequence, a Flaw Acceptability Assessment had to be performed as a part of the Safety Case demonstrating the fitness-for-service of these units. In that framework, detailed analyses using eXtended Finite Element Method were conducted to model the specific character of hydrogen flakes. Their quasi-laminar orientation as well as their high density required setting up 3D multi-flaws model accounting for flaw interaction. These calculations highlighted that even the most penalizing flaw configurations are harmless in terms of structural integrity despite the consideration of higher degradation of irradiated material toughness.

scholarly journals European Project ATLAS+: Small and Large Scale Ductile Tearing Experiments on Ferritic Steel WB36 to Study Transferability of Material Ductile Properties

2019 ◽  
Author(s):  
Anna Dahl ◽  
Dominique Moinereau ◽  
Patrick Le Delliou ◽  
Willy Vincent
Keyword(s):  
Ferritic Steel ◽  
Large Scale ◽  
Structural Integrity ◽  
Assessment Tools ◽  
Test Facility ◽  
Small Scale ◽  
Feed Water ◽  
Ductile Tearing ◽  
Integrity Assessment ◽  
European Project

Abstract The 4-years European project ATLAS+ (Advanced Structural Integrity Assessment Tools for Safe long Term Operation) has been launched in June 2017. One of its objectives is to study the transferability of material ductile properties from small scale specimens to large scale components and validate some advanced tools for structural integrity assessment. The study of properties transferability is based on a wide experimental programme which includes a full set of fracture experiments conducted on conventional fracture specimens and large scale components (mainly pipes). Three materials are considered in the programme : a ferritic steel WB36 typical from secondary feed water line in German PWR reactors, an aged stainless steel austenitic weld representative of EPR design and a typical VVER austenitic dissimilar weld (DMW). This paper describes the experimental work conducted on the ferritic steel WB 36 (15NiCuMoNb5) and summarizes the experimental results available after 2 years of work. Numerous mechanical tests have been conducted on a wide panel of fracture mechanics specimens for a full characterization of the ferritic steel: Tensile properties, Hardness, Charpy Energy, pre-cracked Charpy PCC, Master curve on CT and SENT specimens, ductile tearing properties on CT and SENT specimens. In parallel, it is planned to test three 4PB large scale tests on pipings (FP1, FP2 and FP3) at room temperature on the EDF test facility with 3 configurations (shape, size and location) of cracks: through wall crack (TWC), internal and external ½ elliptical cracks. Progress of these large scale experiments is described including first results.

Advanced Structural Integrity Assessment Tools for Safe Long Term Operation: ATLAS+ Project — Status of the Activities of the WP3 on Modelling in 2020

2020 ◽  
Author(s):  
Arnaud Blouin ◽  
Stéphane Marie ◽  
Tomas Nicak ◽  
Antti Timperi ◽  
Peter Gill
Keyword(s):  
Large Scale ◽  
Structural Integrity ◽  
Assessment Tools ◽  
Local Approach ◽  
Coupled Models ◽  
Large Defect ◽  
Ductile Tearing ◽  
Term Operation ◽  
Integrity Assessment

Abstract The main objective and mission of the ATLAS+ project is to develop advanced structural assessment tools to address the remaining technology gaps for the safe and long term operation of nuclear reactor pressure coolant boundary systems. ATLAS+ WP3 focuses mainly on ductile tearing prediction for large defect in piping and associated components: Several approaches have been developed to accurately model the ductile tearing process and to take into account phenomena such as triaxiality effects, or the ability to predict large tearing in industrial components. These advanced models include local approach coupled models or advanced energetic approaches. Unfortunately, the application of these tools is currently rather limited to R&D expertise. However, because of the continuous progress in the performance of calculation tools and accumulated knowledge, in particular by members of the ATLAS+ consortium, these models can now be considered as relevant for application in the context of engineering assessments. WP3 has been planned to: • Illustrate the implementation of these models for industrial applications through the interpretation of large scale mock-ups (with cracks in weld joints for some of them), • Make recommendations for the implementation of the advanced models in engineering assessments, • Correct data from the conventional engineering approach by developing a methodology to produce J-Δa curve suitable case by case, based on local approach models, • Improve the tools, guidance and procedures for undertaking leak-before-break (LBB) assessments of piping components, particularly in relation to representing structural representative fracture toughness J-Resistance curves and the influence of weld residual stresses. To achieve these goals, WP3 is divided into 4 sub-WPs and this paper presents the progress of the work performed in each sub-WP after 36 months of activities.

FSI-Based Assessment of FAC-Caused Wall Thinned Piping

2009 ◽  
Author(s):  
Sun-Hye Kim ◽  
Yoon-Suk Chang ◽  
Young-Jin Kim
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Flow Characteristics ◽  
Limit Load ◽  
Bend Angle ◽  
Piping System ◽  
Integrity Assessment ◽  
Wall Thinning ◽  
Analysis Scheme

Lots of investigations on failures of wall thinned piping have been carried out since the accident of Surry unit 2 in USA. From these preceding efforts, flow accelerated corrosion (FAC) which is a kind of wall thinning phenomenon is revealed main factor of failure of pipes in nuclear power plants. However, there are a few researches which directly take into account of flow characteristics and geometric changes for stress assessment of FAC-caused wall thinned piping. In this paper, structural integrity assessment employing a fluid-structure interaction (FSI) analysis scheme is performed on pipes representing secondary piping system of PWR which consists of straight pipes and elbows of various bend angles. Prior to the assessment, CFD analyses are conducted to predict plausible wall thinning location by considering flow and geometric parameters such as bend angle and radius of elbow. Then, for typical pipe geometry, detailed limit load analyses are performed to calculate maximum stress caused by turbulence and velocity of flow near the wall thinned part. Through these kinds of detailed parametric analyses, effects of FSI were observed, which should be considered for assessment of FAC-caused wall thinned piping.

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