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.

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.

Advances in Residual Stress Modeling and Measurement for the Structural Integrity Assessment of Welded Thermal Power Plant

2008 ◽  
Vol 41-42 ◽  
pp. 391-400 ◽  
Author(s):  
Lyndon Edwards ◽  
Mike C. Smith ◽  
Mark Turski ◽  
Michael E. Fitzpatrick ◽  
P. John Bouchard
Keyword(s):  
Residual Stress ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Residual Stresses ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Economic Life ◽  
Life Assessment ◽  
Integrity Assessment

The safe operation of both thermal and nuclear power plant is increasingly dependent upon structural integrity assessment of pressure vessels and piping. Furthermore, structural failures most commonly occur at welds so the accurate design and remnant life assessment of welded plant is critical. The residual stress distribution assumed in defect assessments often has a deciding influence on the analysis outcome, and in the absence of accurate and reliable knowledge of the weld residual stresses, the design codes and procedures use assumptions that yield very conservative assessments that can severely limit the economic life of some plant. However, recent advances in both the modeling and measurement of residual stresses in welded structures and components open up the possibility of characterising weld residual stresses in operating plant using state-of–the–art fully validated Finite Element simulations. This paper describes research undertaken to predict residual stresses in stainless steel welds in order to provide validated reliable, accurate Structural Integrity assessment of nuclear power plant components

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.

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.

Structural Integrity Assessment of Doel 3 and Tihange 2 RPVs: ASME B&PV Code Section III Primary Stress Re-Evaluation Accounting for the Presence of Hydrogen Flakes

2017 ◽  
Author(s):  
Pierre Dulieu ◽  
Valéry Lacroix
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Local Area ◽  
Pressure Vessels ◽  
Collapse Pressure ◽  
Acceptance Criterion ◽  
Integrity Assessment ◽  
Primary Stress ◽  
Extensive Evaluation

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 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 evaluation 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(d)(2) of ASME B&PV Code Section XI, one of the parts that have to be addressed through the Flaw Acceptability Assessment is the Primary Stress Re-Evaluation assuming local area reductions of the pressure retaining membrane i.e., of the core shells, due to presence of the flaws. This is performed using the limit analysis provided by Article NB-3228.1 of the ASME B&PV Code Section III. Results are compared to those using the plastic analysis of Article NB-3228.3. The acceptance criterion that needs to be verified is that the calculated collapse pressure should be more than 1.5 times the design pressure. The paper presents the 2D conservative approach developed in order to carry out this analysis dealing with large number and high density of flaws. Furthermore, the paper validates this 2D conservative methodology through detailed 3D XFEM elastic-plastic calculations.

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.

Evaluation of the Crack Interaction and Failure Behavior of Components With Crack Fields Using a Damage Mechanical Approach

2019 ◽  
Author(s):  
Christian Swacek ◽  
Patrick Gauder ◽  
Michael Seidenfuss
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Base Material ◽  
Pressure Vessels ◽  
Failure Behavior ◽  
Multiple Cracks ◽  
Critical Crack ◽  
Ongoing Research ◽  
Integrity Assessment ◽  
Crack Tips

Abstract In 2012 non-destructive testing measurements (NDT) of the reactor pressure vessels (RPV) in the Belgian Nuclear Power Plants Doel 3 and Tihange 2 revealed a high quantity of indications in the upper and lower core shells. The most likely explanation is that the measured indications are hydrogen flakes positioned in segregated zones in the base material of the pressure vessels. These hydrogen flakes have a laminar and quasi-laminar orientation with an inclination up to 15° to the pressure retaining surface. Under internal pressure, the crack tips undergo predominantly mixed mode loading conditions, where the induced stress and strain fields of the single crack tips influence each other. The safety assessment of crack afflicted pressurized components is performed by fracture mechanical approaches. For the evaluation of multiple cracks in crack fields, state of the art codes and standards apply interaction criteria and grouping methods, to determine a representative crack, which has to be evaluated. In this paper, the interaction of cracks in crack fields is numerically and experimentally evaluated. Damage mechanical models based on the Rousselier- and the Beremin model are used to investigate numerically the interaction of cracks in crack fields. Experimental data from ferritic flat tensile specimens afflicted with cracks are used to verify the numerical results. The damage mechanical calculations reveal critical crack arrangements due to coalescence behavior and cleavage fracture probability. These results and ongoing research intends the derivation of interaction criteria for cracks in crack fields. The interaction criteria will be used for the definition of a representative flaw for a conservative integrity assessment of crack afflicted components.

Constraint-Dependent J-R Curves of a Dissimilar Metal Welded Joint for Connecting Pipe-Nozzle of Nuclear Pressure Vessel

2015 ◽  
Vol 137 (2) ◽  
Author(s):  
J. Wang ◽  
G. Z. Wang ◽  
F. Z. Xuan ◽  
S. T. Tu
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Welded Joint ◽  
Pressure Vessels ◽  
Stress Constraint ◽  
Dissimilar Metal ◽  
Local Strength ◽  
T Stress

In this paper, the J-R curves of two cracks (A508 HAZ crack 2 and A508/Alloy52Mb interface crack 3) located at the weakest region in an Alloy52M dissimilar metal welded joint (DMWJ) for connecting pipe-nozzle of nuclear pressure vessel have been measured by using single edge-notched bend (SENB) specimens with different crack depths a/W (different constraint). Based on the modified T-stress constraint parameter τ*, the equations of constraint-dependent J-R curves for the crack 2 and crack 3 were obtained. The predicted J-R curves using different constraint equations derived from the three pairs of crack growth amount all agree with the experimental J-R curves. The results show that the modified T-stress approach for obtaining constraint-dependent J-R curves of homogeneous materials can also be used for the DMWJs with highly heterogeneous mechanical properties (local strength mismatches) in nuclear power plants. The use of the constraint-dependent J-R curves may increase the accuracy of structural integrity design and assessment for the DMWJs of nuclear pressure vessels.

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