Validation of Constraint Based Methodology in Structural Integrity – VOCALIST: Experimental Programme

2002 ◽  
Author(s):  
Richard Bass ◽  
Ulrich Eisele ◽  
Elisabeth Keim ◽  
Heikki Keinanen ◽  
Ste´phane Marie ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Experimental Investigations ◽  
Integrity Assessment ◽  
Experimental Programme ◽  
Safety Margins ◽  
Materials Used ◽  
Shallow Crack

The aim of VOCALIST (Validation of Constraint-Based Assessment Methodology in Structural Integrity) is to develop and validate innovative procedures for assessing the level of, and possible changes to, constraint-related safety margins in ageing pressure boundary components [1]. An iterative process of experiment and analysis will address this overall objective. The experimental investigations within VOCALIST are performed on three different materials representing the as new state of materials used for components of nuclear power plants as well as a state representing an in service degraded state of RPV materials. Within the experimental programme fracture mechanics specimens with different constraint situations are tested in order to quantify the influence of the constraint on the specimens failure behaviour as a basis for the advanced components integrity assessment. The investigations are performed on small laboratory specimens as well as on biaxially loaded cruciform specimens and large piping components. Within this contribution the experimental programme of VOCALIST is introduced. The investigated materials are characterized in terms of their mechanical properties. Special consideration is given to results of fracture mechanics specimens highlighting the constraint effect via the shallow crack effect and its contribution to a shift of the master curve.

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.

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.

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.

Verification of Probabilistic Fracture Mechanics Analysis Code PASCAL

2017 ◽  
Author(s):  
Yinsheng Li ◽  
Genshichiro Katsumata ◽  
Koichi Masaki ◽  
Shotaro Hayashi ◽  
Yu Itabashi ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Working Group ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Irradiation Embrittlement ◽  
Energy Agency ◽  
Probabilistic Fracture Mechanics ◽  
Integrity Assessments

Probabilistic fracture mechanics (PFM) has been recognized as a promising methodology in structural integrity assessments of aged pressure boundary components of nuclear power plants because it can rationally represent the influencing parameters in their inherent probabilistic distributions without over conservativeness. In Japan, a PFM analysis code PASCAL (PFM Analysis of Structural Components in Aging LWR) has been developed by the Japan Atomic Energy Agency (JAEA) to evaluate the through-wall cracking frequencies of Japanese reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and pressurized thermal shock (PTS) transients. In addition, efforts have been made to strengthen the applicability of PASCAL to structural integrity assessments of domestic RPVs against non-ductile fracture. On the other hand, unlike deterministic analysis codes, the verification of PFM analysis codes is not easy. A series of activities has been performed to verify the applicability of PASCAL. In this study, as a part of the verification activities, a working group was established in Japan, with seven organizations from industry, universities and institutes voluntarily participating as members. Through one year activities, the applicability of PASCAL for structural integrity assessments of domestic RPVs was confirmed with great confidence. This paper presents the details of the verification activities of the working group including the verification plan, approaches and results.

Large EDF Tests on Aged Cast Duplex Stainless Steel Components: Part I — Reduced Scale Tests

2015 ◽  
Author(s):  
Patrick Le Delliou ◽  
Sébastien Saillet ◽  
Georges Bezdikian
Keyword(s):  
Stainless Steel ◽  
Duplex Stainless Steel ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Evolution ◽  
Thermal Ageing ◽  
Life Extension ◽  
Term Operation ◽  
Integrity Assessment ◽  
Safety Margins

Thermal ageing of cast duplex stainless steel primary loops components (elbows, pump casings and branch connections) is a concern for long-term operation of EDF nuclear power plants. The thermal ageing embrittlement results from the micro-structural evolution of the ferrite phase (spinodal decomposition), and can reduce the fracture toughness properties of the steel. In addition, it is necessary to consider manufacturing quality and the possible occurrence of casting defects such as shrinkage cavities. In a context of life extension, it is important to assess the safety margins to crack initiation and crack propagation instability. This paper presents several tests conducted by EDF on aged cast duplex stainless steel NPP components, respectively on two-third scale elbows and welded mock-ups. The main characteristics of the tests are recalled, the results are presented, and finally, the lessons drawn are summarized. These tests and their detailed analyses contribute to validate and justify the methodology used by EDF in the integrity assessment of in-service cast duplex stainless steel components.

Benchmark Analysis on Probabilistic Fracture Mechanics Analysis Codes Considering Multiple Cracks and Crack Initiation in Aged Piping of Nuclear Power Plants

2014 ◽  
Author(s):  
Yinsheng Li ◽  
Kazuya Osakabe ◽  
Genshichiro Katsumata ◽  
Jinya Katsuyama ◽  
Kunio Onizawa ◽  
...  
Keyword(s):  
Risk Assessment ◽  
Fracture Mechanics ◽  
Crack Initiation ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Structural Integrity ◽  
Multiple Cracks ◽  
Benchmark Analysis ◽  
Probabilistic Fracture Mechanics

In recent years, cracks have been detected in piping systems of nuclear power plants. Many of them are multiple cracks in the same welded joints. Therefore, structural integrity evaluation and risk assessment considering multiple cracks and crack initiation in aged piping have become increasingly important. Probabilistic fracture mechanics (PFM) is a rational methodology in structural integrity evaluation and risk assessment of aged piping in nuclear power plants. Two PFM codes, PASCAL-SP and PRAISE-JNES, have been improved or developed in Japan for the structural integrity evaluation and risk assessment considering the age related degradation mechanisms of pipes. Although the purposes to develop these two codes are different, both have almost the same basic functions to obtain the failure probabilities of pipes. In this paper, a benchmark analysis was conducted considering multiple cracks and crack initiation, in order to confirm their reliability and applicability. Based on the numerical investigation in consideration of important influence factors such as crack number, crack location, crack distribution and crack detection probability of in-service inspection, it was concluded that the analysis results of these two codes are in good agreement.

Improvement of Probabilistic Fracture Mechanics Analysis Code PASCAL-SP With Regardto PWSCC

2019 ◽  
Vol 5 (3) ◽  
Author(s):  
Akihiro Mano ◽  
Yoshihito Yamaguchi ◽  
Jinya Katsuyama ◽  
Yinsheng Li
Keyword(s):  
Fracture Mechanics ◽  
Failure Probability ◽  
Power Plants ◽  
Crack Detection ◽  
Structural Integrity ◽  
Energy Agency ◽  
Pressurized Water ◽  
Integrity Assessment ◽  
Probabilistic Fracture Mechanics ◽  
Failure Probabilities

Probabilistic fracture mechanics (PFM) analysis is expected to be a rational method for structural integrity assessment because it can consider the uncertainties of various influence factors and evaluate the quantitative values such as failure probability of a cracked component as the solution. In the Japan Atomic Energy Agency, a PFM analysis code PASCAL-SP has been developed for structural integrity assessment of piping welds in nuclear power plants (NPP). In the past few decades, a number of cracks due to primary water stress corrosion cracking (PWSCC) have been detected in nickel-based alloy welds in the primary piping of pressurized water reactors (PWRs). Thus, structural integrity assessments considering PWSCC have become important. In this study, PASCAL-SP was improved considering PWSCC by introducing several analytical functions such as the models for evaluation of crack initiation time, crack growth rate (CGR), and probability of crack detection. By using the improved version of PASCAL-SP, the failure probabilities of pipes with a circumferential crack or an axial crack due to PWSCC were numerically evaluated. Moreover, the influence of leak detection and nondestructive examination (NDE) on failure probabilities was detected. Based on the obtained numerical results, it was concluded that the improved version of PASCAL-SP is useful for evaluating the failure probability of a pipe considering PWSCC.

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.

The Formulation of Material Characteristics of Austenitic Stainless Steels at Extremely High Temperature

2017 ◽  
Author(s):  
Kenta Shimomura ◽  
Takashi Onizawa ◽  
Shoichi Kato ◽  
Masanori Ando ◽  
Takashi Wakai
Keyword(s):  
High Temperature ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Structural Integrity ◽  
Austenitic Stainless Steels ◽  
Severe Accident ◽  
Integrity Assessment ◽  
Severe Accidents ◽  
Material Characteristics

This paper describes the formulation of material characteristics of austenitic stainless steels at extremely high temperature which meets in some kinds of severe accidents of nuclear power plants. After the severe accident in Fukushima dai-ichi nuclear power plants, it has been supposed to be very important not only to prevent the occurrence of abnormal conditions, i.e. from the first to the third layer safety, but also to prevent the expansion of the accident conditions, i.e. the fourth layer safety[1] [2]. In order to evaluate the structural integrity under the severe accident condition, material characteristics which can be used in the numerical analyses, such as finite element analysis, were required [3] [4]. However, there were no material characteristics applicable to the structural integrity assessment at extremely high temperature. Therefore, a series of tensile and creep tests was performed for austenitic stainless at extremely high temperature which meets in some kinds of severe accidents of nuclear power plants, namely up to 1000 °C. Based on the acquired data from the tests, monotonic stress-strain equation and creep rupture equation applicable to the structural analysis at extremely high temperature, up to 1000 °C were formulated. As a result, these formulae make it possible to conduct the structural integrity assessment using numerical analysis techniques, such as finite element method.

Development of Stress Intensity Factors for Surface Cracks With Large Aspect Ratio in Plates

2015 ◽  
Vol 137 (5) ◽  
Author(s):  
Yinsheng Li ◽  
Kunio Hasegawa ◽  
Genshichiro Katsumata ◽  
Kazuya Osakabe ◽  
Hiroshi Okada
Keyword(s):  
Stress Intensity ◽  
Aspect Ratio ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Crack Depth ◽  
Surface Cracks ◽  
Large Aspect Ratio ◽  
Element Analysis ◽  
Integrity Assessment

A number of surface cracks with large aspect ratio have been detected in components of nuclear power plants (NPPs) in recent years. The depths of these cracks are even larger than the half of crack lengths. When a crack is detected during in-service inspections, methods provided in ASME Boiler and Pressure Vessel Code Section XI or JSME Rules on fitness-for-service for NPPs can be used to assess the structural integrity of cracked components. The solution of the stress intensity factor (SIF) is very important in the structural integrity assessment. However, in the current codes, the solutions of the SIF are provided for semi-elliptical surface cracks with a limitation of a/ℓ ≤ 0.5, where a is the crack depth, and ℓ is the crack length. In this study, the solutions of the SIF were calculated using finite element analysis (FEA) with quadratic hexahedron elements for semi-elliptical surface cracks with large aspect ratio in plates. The crack dimensions were focused on the range of a/ℓ = 0.5–4.0 and a/t = 0.0–0.8, where t is the wall thickness. Solutions were provided at both the deepest and the surface points of the surface cracks. Furthermore, some of solutions were compared with the available existing results as well as with solutions obtained using FEA with quadratic tetrahedral elements and the virtual crack closure-integral method (VCCM). Finally, it was concluded that the solutions proposed in this paper are applicable in engineering applications.

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