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.

Structural Integrity Assessment of Doel 3 and Tihange 2 RPVs Affected by Hydrogen Flakes: Methodology and Results

2016 ◽  
Author(s):  
Valéry Lacroix ◽  
Pierre Dulieu
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Integrity Assessment ◽  
Operating Modes ◽  
Innovative Methodology ◽  
Safety Case ◽  
Component Manufacturing ◽  
Accident Conditions

During the 2012 outage at Doel 3 and Tihange 2 Nuclear Power Plants, a large number of quasi-laminar indications were detected in 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, both units remained core unloaded pending the elaboration of an extensive Safety Case demonstrating that they can be safely operated. The Structural Integrity Assessment of the RPVs, through the Flaw Acceptability Analysis, aimed at demonstrating that the identified indications do not jeopardize the integrity of the reactor vessel in all operating modes, transients and accident conditions. This demonstration, presented in this paper, has been done on the basis of a specific innovative methodology inspired by the ASME XI procedure but adapted to the nature and number of indications found in the Doel 3 and Tihange 2 RPVs.

Quantification of the Conservatisms in the Flaw Acceptability Assessment of Doel 3 and Tihange 2 RPVs Containing Hydrogen Flakes

2019 ◽  
Author(s):  
Valéry Lacroix ◽  
Pierre Dulieu
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Pressure Vessels ◽  
Assessment Methodology ◽  
Operating Modes ◽  
Safety Case ◽  
Asme Code ◽  
Component Manufacturing ◽  
Accident Conditions ◽  
Reactor Pressure

Abstract During the 2012 outage at Doel 3 and Tihange 2 Nuclear Power Plants, a large number of quasi-laminar indications were detected in the lower and upper core shells of the reactor pressure vessels (RPVs). The observed indications could subsequently be attributed to hydrogen flaking induced during the component manufacturing process. As a consequence, both units remained core unloaded pending the elaboration of an extensive Safety Case demonstrating that they can be safely operated. One of the most challenging parts of this demonstration was the Flaw Acceptability Assessment, aiming at demonstrating that the identified indications do not jeopardize the integrity of the reactor vessel in all operating modes, transients and accident conditions. This analysis was done by using a methodology: innovative, in line with existing ASME Code Section XI requirements, specific, sufficiently wide to be accepted and, first and foremost, conservative. Through a brief reminder of the Flaw Acceptability Assessment methodology, the paper presents the main hypotheses done for the calculation and quantifies the conservatism related to each of them. This quantification clearly highlights the reliability of final result i.e., the demonstration of the Fitness-for-Service for continued operation of both Doel 3 and Tihange 2 RPVs.

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.

Application of Probabilistic Fracture Mechanics to Reactor Pressure Vessel Using PASCAL4 Code

2020 ◽  
Vol 143 (2) ◽  
Author(s):  
Kai Lu ◽  
Jinya Katsuyama ◽  
Yinsheng Li ◽  
Shinobu Yoshimura
Keyword(s):  
Fracture Mechanics ◽  
Power Plants ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Pressurized Water ◽  
Integrity Assessment ◽  
Thermal Transients ◽  
Probabilistic Fracture Mechanics ◽  
Integrity Assessments ◽  
Reactor Pressure

Abstract Probabilistic fracture mechanics (PFM) is considered to be a promising methodology in structural integrity assessments of pressure-boundary components in nuclear power plants since it can rationally represent the inherent probabilistic distributions for influence parameters without over-conservativeness. To strengthen the applicability of PFM methodology in Japan, Japan Atomic Energy Agency has developed a PFM analysis code PASCAL4 which enables the failure frequency evaluation of reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and thermal transients. PASCAL4 is expected to make a significant contribution to the probabilistic integrity assessment of Japanese RPVs. In this study, PFM analysis for a Japanese model RPV in a pressurized water reactor (PWR) was conducted using PASCAL4, and the effects of nondestructive examination (NDE) and neutron flux reduction on failure frequencies of the RPV were quantitatively evaluated. From the analysis results, it is concluded that PASCAL4 is useful for probabilistic integrity assessments of embrittled RPVs and can enhance the applicability of PFM methodology.

RPV Integrity Assessment: Comparison Between 3 Loops and 4 Loops Studies

2004 ◽  
Author(s):  
Georges Bezdikian ◽  
Dominique Moinereau ◽  
Claude Faidy
Keyword(s):  
Brittle Fracture ◽  
Nuclear Power ◽  
Power Plants ◽  
Electric Energy ◽  
Pressure Vessels ◽  
Surveillance Program ◽  
Integrity Assessment ◽  
Safety Studies ◽  
Comparison Of The Results ◽  
Reactor Pressure

For the French utility (Electricite de France–EDF), Nuclear Energy represents 75% of generation of the total electric energy in France. Total nuclear electricity were generated mainly from Nuclear Power plants stations, 34 PWR NPPs 3-loop Reactors- 900 MWe, 20 PWR NPPs 4-loop Reactors- 1300 MWe and 4 PWR NPPs 4-loop Reactors- 1450 MWe. The 3-loop Reactor Pressure Vessel (RPV) integrity assessment, applied on 34 PWR NPPs Reactors, involved the verification of the integrity of the component under the most severe conditions of situation, and the result obtained was the justification of the 900 MWe RPV life management for at least 40 years and to prepare the projection beyond 40 years. Since 2000, in the continuity of these results, the studies were carried out on the 20 PWR NPPs 4-loop 1300 MWe Reactor Pressure Vessels, and the recent results obtained show the demonstration of the integrity of the RPV, in the most severe conditions of loading in relation with RTNDT (Reference Nil Ductility Transition Temperature), and other major parameters. This approach is based on specific mechanical safety studies on the RPV to demonstrate the absence of the risk of failure by brittle fracture. For these mechanical studies the major input data are necessary: 1 - the fluence distribution and the values of 3-loop and 4-loop RPV, 2 - RTNDT during the lifetime in operation, 3 - the temperature distribution values in the downcomer and the PTS evaluation. The main results must show significant margins against initiation of brittle fracture for all of 3-loop and 4-loop RPV. The flaws considered in this approach are shallow flaws beneath the cladding (subclad flaws) or in the first layer of cladding. The major tasks and expertises engaged by EDF are: • more precise assessment of the fluence calculations, • better knowledge of the vessel material properties, including the effect of radiation, • the NDE inspection program on the core zone. The comparison of the results are developed in this paper: • for the fluence evaluation and the optimisation of the fuel management, • the data gathered from radiation specimen capsules, removed from the vessels (radiation surveillance program), • and the thermal-hydraulic and mechanical calculations based on finite element thermal-hydraulic and 3D elastic-plastic mechanical computations.

Alternative Characterization Rules for Quasi-Laminar Flaws Based on 3D X-FEM Calculations

2015 ◽  
Author(s):  
Valéry Lacroix ◽  
Pierre Dulieu ◽  
Anne-Sophie Bogaert
Keyword(s):  
Finite Element Method ◽  
Nuclear Power ◽  
Power Plants ◽  
Extended Finite Element Method ◽  
Structural Integrity ◽  
Division I ◽  
Extended Finite Element ◽  
Alternative Characterization ◽  
Asme Code ◽  
Alternative Methodology

During the 2012 outage at Doel 3 and Tihange 2 Nuclear Power Plants, a large number of quasi-laminar indications were detected, mainly in the lower and upper core shells of the RPVs. In the frame of the Structural Integrity demonstration of these RPVs according to ASME XI principles, ASME XI IWB-3300 article requires combining closely spaced flaws in order to account for their mechanical interactions. However, it appeared early that the characterization rules were adapted neither to quasi-laminar flaws nor to such densities of flaws. Therefore, an alternative methodology to derive characterization rules for quasi-laminar flaws has been developed, implemented and validated. This work, based on 2D eXtended Finite Element Method (X-FEM) calculations and presented during ASME PVP 2014, has led to the proposed ASME Code Case N-848 “Alternative characterization rules for quasi-laminar flaws – Section XI, Division I”. This 2D approach, even though better suited to quasi-laminar flaws, results however in very conservative proximity rules. Therefore, it appeared that more realistic — although still conservative — proximity rules based on 3D X-FEM calculations could be developed.

Application of Probabilistic Fracture Mechanics Methodology for Japanese Reactor Pressure Vessels Using PASCAL4

2019 ◽  
Author(s):  
Kai Lu ◽  
Jinya Katsuyama ◽  
Yinsheng Li ◽  
Shinobu Yoshimura
Keyword(s):  
Fracture Mechanics ◽  
Power Plants ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Irradiation Embrittlement ◽  
Energy Agency ◽  
Integrity Assessment ◽  
Probabilistic Fracture Mechanics ◽  
Failure Frequency ◽  
Reactor Pressure

Abstract Probabilistic fracture mechanics (PFM) methodology, which represents the influence parameters in their inherent probabilistic distributions, is deemed to be promising in the structural integrity assessment of pressure-boundary components in nuclear power plants. To strengthen the applicability of PFM methodology in Japan, Japan Atomic Energy Agency has developed a PFM analysis code called PASCAL4 (PFM Analysis of Structural Components in Aging LWRs, Version 4) which can be used to evaluate the failure frequency of reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and pressurized thermal shock events. PASCAL4 is expected to make a significant contribution to the probabilistic integrity assessment of Japanese RPVs. In this study, PFM analyses are performed for a Japanese model RPV using PASCAL4, and the effects of non-destructive examination and neutron fluence mitigation on failure frequency of RPV are quantitatively evaluated. From the analysis results, it is concluded that PASCAL4 is useful for the structural integrity assessment of RPVs and can enhance the applicability of PFM methodology.

An Assessment of Mechanical Properties of A508-3 Steel Used in Chinese Nuclear Reactor Pressure Vessels

2015 ◽  
Vol 137 (3) ◽  
Author(s):  
Meifang Yu ◽  
Y. J. Chao ◽  
Zhen Luo
Keyword(s):  
Mechanical Properties ◽  
Nuclear Power ◽  
Power Plants ◽  
Fossil Fuels ◽  
Nuclear Reactor ◽  
Structural Integrity ◽  
Construction Materials ◽  
Pressure Vessels ◽  
Under Construction ◽  
Reactor Pressure

China has very ambitious goals of expanding its commercial nuclear power by 30 GW within the decade and wishes to phase out fossil fuels emissions by 40–45% by 2020 (from 2005 levels). With over 50 new nuclear power plants under construction or planned and a design life of 60 years, any discussions on structural integrity become very timely. Although China adopted its nuclear technology from France or USA at present time, e.g., AP1000 of Westinghouse, the construction materials are primarily “Made in China.” Among all issues, both the accumulation of the knowledge base of the materials and structures used for the power plant and the technical capability of engineering personnel are imminent. This paper attempts to compile and assess the mechanical properties, Charpy V-notch impact energy, and fracture toughness of A508-3 steel used in Chinese nuclear reactor pressure vessels (RPVs). All data are collected from open literature and by no means complete. However, it provides a glimpse into how this domestically produced steel compares with western RPV steels such as USA A533B and Euro 20MnMoNi55.

Structural Integrity Assessment of Doel 3 and Tihange 2 RPVs: Fatigue Crack Growth Analysis of Hydrogen Flakes

2017 ◽  
Author(s):  
Valéry Lacroix ◽  
Pierre Dulieu
Keyword(s):  
Fatigue Crack ◽  
Fatigue Crack Growth ◽  
Crack Growth ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Simplified Approach ◽  
Integrity Assessment ◽  
The Core

During the 2012 outages at Doel 3 and Tihange 2 Nuclear Power Plants, a large number of quasi-laminar indications were detected, mainly in the lower and upper core shells of the Reactor Pressure Vessels (RPVs). The observed indications could subsequently be attributed to hydrogen flaking induced during the component manufacturing process. As a consequence, both units remained core unloaded pending the elaboration of an extensive Safety Case demonstrating that they can be safely operated. The Structural Integrity Assessment of the RPVs, through the Flaw Acceptability Assessment, aimed at demonstrating that the identified indications do not jeopardize the integrity of the reactor vessel in all operating modes, transients and accident conditions. This demonstration has been done on the basis of a specific methodology inspired by the ASME B&PV Code Section XI procedure but adapted to the nature and the number of indications found in the Doel 3 and Tihange 2 RPVs. As requested by Article IWB-3610(a) of ASME B&PV Code Section XI, one of the parts that have to be addressed through the Flaw Acceptability Assessment is the Fatigue Crack Growth (FCG) Analysis of the flaws in the core shells until the end-of-service lifetime of the RPVs. Due to the large number of flaws in the core shells, a specific methodology has been developed in order not to perform the FCG Analysis of each flaw separately. The paper describes this simplified approach aiming at distributing the flaws according to their inclination and at defining envelope flaws covering the actual flaws to carry out FCG Analysis. Furthermore, the paper highlights and quantifies the conservatisms of this analysis leading finally to demonstrate that the FCG of hydrogen flakes is not a concern in Doel 3 and Tihange 2 RPVs.

scholarly journals Quick Evaluation Method for Defect Exceeding the Allowable Flaw Size in Pressure Vessel of Nuclear Reactor for Power Plant and Space Heating

2021 ◽  
Vol 9 ◽  
Author(s):  
Pan Liu ◽  
Yuebing Li ◽  
Ting Jin ◽  
Dasheng Wang
Keyword(s):  
Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Atmospheric Pollutants ◽  
Space Heating ◽  
Integrity Assessment ◽  
Evaluation Approach

Nuclear power can be used for power generation, space heating, and other fields, producing a limited level of greenhouse gases and no atmospheric pollutants. However, the safety of nuclear reactors is always a public concern. The reactor pressure vessels (RPVs) play an important role in the safe operation of a nuclear power plant. When a defect is inspected in the RPV, complex analytical evaluation procedures, including fatigue analysis and fracture assessment, are necessary to ensure the structural integrity of the defective component. Based on the RSE-M, a quick evaluation approach for RPVs with defects exceeding acceptance standards is proposed in this work to reduce the computational complexity and analysis time. The flaw evaluation is simplified by adjusting the inspection period based on the analysis of fatigue crack growth. The new method was applied to the RPVs with embedded defects and underclad semi-elliptical defects, respectively. The proposed evaluation approach was verified by the case of a typical RPV cylinder containing an embedded crack, where all possible transients during the operation of nuclear power plants are considered. During the allowable residual life obtained of 5-years, failure would not occur in the defective component via the conventional method, which gives confidence to the availability of the new approach. Consequently, the proposed method can be a valid reference for the structural integrity assessment of nuclear reactor components with defects exceeding acceptance standards.

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