scholarly journals Non-Linear Design Evaluation of Class 1-3 Nuclear Power Piping

10.5772/20592 ◽  
2011 ◽  
Author(s):  
Lingfu Zeng ◽  
Lennart G. ◽  
Lars Dahlstrom
Keyword(s):  
Nuclear Power ◽  
Design Evaluation ◽  
Non Linear ◽  
Class 1

Non-Linear Design Verification of Nuclear Power Piping According to ASME III NB/NC

2008 ◽  
Author(s):  
Lingfu Zeng ◽  
Lennart G. Jansson
Keyword(s):  
Nuclear Power ◽  
Design Evaluation ◽  
Piping System ◽  
Design Verification ◽  
Intensive Study ◽  
Non Linear ◽  
Piping Systems ◽  
Design Requirements

A nuclear piping system which is found to be disqualified, i.e. overstressed, in design evaluation in accordance with ASME III, can still be qualified if further non-linear design requirements can be satisfied in refined non-linear analyses in which material plasticity and other non-linear conditions are taken into account. This paper attempts first to categorize the design verification according to ASME III into the linear design and non-linear design verifications. Thereafter, the corresponding design requirements, in particular, those non-linear design requirements, are reviewed and examined in detail. The emphasis is placed on our view on several formulations and design requirements in ASME III when applied to nuclear power piping systems that are currently under intensive study in Sweden.

More on Non-Linear Design Verification of Nuclear Power Piping According to ASME III NB/NC

2009 ◽  
Author(s):  
Lingfu Zeng ◽  
Lennart G. Jansson ◽  
Lars Dahlstro¨m
Keyword(s):  
Nuclear Power ◽  
Design Evaluation ◽  
Piping System ◽  
Design Verification ◽  
Design Rules ◽  
Non Linear ◽  
Design Requirements

A nuclear piping system which is found to be disqualified, i.e. overstressed, in design evaluation in accordance with ASME III, can still be qualified if further non-linear design requirements can be satisfied in refined non-linear analyses in which material plasticity and other non-linear conditions are taken into account. Our work presented earlier in ICONE16 categorized the design rules in ASME III into linear design and non-linear design rules and examined the corresponding design requirements. In this paper, a more in-depth review of these rules for service limit level D is conducted. In particular, several rules which are currently intensively applied but often found to be inconsistent and confusing in Sweden are studied and discussed in detail. Suggestions for improvements and guides for a reasonable application of these rules, which have been practiced in several ongoing projects, are given.

More on Design Evaluation of Nuclear Power Piping Subjected to Time-History Dynamic Loads According to ASME B&PV III

2013 ◽  
Author(s):  
Lingfu Zeng ◽  
Lennart G. Jansson
Keyword(s):  
Finite Element Analysis ◽  
Nuclear Power ◽  
Time History ◽  
Dynamic Loads ◽  
Design Evaluation ◽  
Piping System ◽  
Element Analysis ◽  
Design Rules ◽  
Non Linear ◽  
Design Requirements

A nuclear piping system which is found to be disqualified, i.e. overstressed, in design evaluation in accordance with ASME B&PV code, Section III, can still be qualified if further non-linear design requirements can be satisfied in refined non-linear analyses in which material plasticity and other non-linear conditions are taken into account. Our work presented earlier in ICONE16-20 categorized the design rules in the code into linear and non-linear design rules and attempted to clarify the corresponding design requirements. In this paper, an in-depth review of these rules and relevant requirements for time-history dynamic loads designated in Service Limit Levels C and D, such as those caused by e.g. water-hammer effects, is given. In particular, several changes of these rules in the current edition of the codes and their applications are addressed, and several rules which are intensively applied but often found to be inconsistent and confusing in Sweden are discussed. It is suggested that the design evaluation can be made directly through non-linear transient finite element analysis using a strain-based criterion for time-history dynamic loads of either reversing or non-reversing type.

Fatigue Assessment of Class 1 Nuclear Power Piping According to ASME BPV Code: Why Other Alternatives?

2014 ◽  
Author(s):  
Lingfu Zeng ◽  
Lennart G. Jansson
Keyword(s):  
Finite Element Analysis ◽  
Finite Element ◽  
Stress Intensity ◽  
Nuclear Power ◽  
Element Analysis ◽  
Fatigue Assessment ◽  
Elastic Plastic ◽  
Linear Finite Element ◽  
Non Linear ◽  
Class 1

In this paper, fatigue assessment of Class 1 nuclear power piping according to ASME Boiler and Pressure Vessel Code, Section III, NB-3600, is discussed. Key parameters involved in the fatigue assessment, i.e. the alternating stress intensity Salt, the penalty factor Ke and the cumulative damage factor U, are addressed. In particular, a so-called simplified elastic-plastic discontinuity analysis for alternative verification when basic fatigue requirements found unsatisfactory, and the procedures for evaluating the alternating stress intensity Salt, is reviewed. Factors that can significantly affect the reliability and accuracy of the fatigue assessment are examined. It is illustrated that there is a great need of other alternatives to this simplified elastic-plastic analysis procedure. An alternative based on non-linear finite element analysis is suggested. This paper is a continuation of our work presented in ICONE16-21, which attempted to categorize design rules in the code into linear and non-linear rules and to clarify corresponding requirements that can be used in combination with non-linear finite element analysis.

PM-HIP Research for Pressure Retaining Applications Within the Electric Power Industry

2014 ◽  
Author(s):  
David Gandy ◽  
John Siefert ◽  
Lou Lherbier ◽  
David Novotnak
Keyword(s):  
Electric Power ◽  
Nuclear Power ◽  
Hot Isostatic Pressing ◽  
Electric Power Industry ◽  
Power Industry ◽  
Plate Rolling ◽  
Alloy 690 ◽  
Class 1 ◽  
Water Reactors ◽  
Grade 91

For more than 60 years now, the nuclear power industry has relied on structural and pressure retaining materials generated via established manufacturing practices such as casting, plate rolling-and-welding, forging, drawing, and/or extrusion. During the past three years, EPRI has been leading the development and introduction of another established process, powder metallurgy and hot isostatic pressing (PM/HIP), for pressure retaining applications in the electric power industry. The research includes assessment of two primary alloys: 316L stainless steel and Grade 91 creep-strength enhanced ferritic steels, for introduction into the ASME Boiler and Pressure Vessel Code. Continuing DOE and EPRI research on other structural/pressure retaining alloys such as Alloy 690, SA 508 Class 1, Alloy 625, hard-facing materials, and others are also underway. This research will have a tremendous impact as we move forward over the next few decades on the selection of new alloys and components for advanced light water reactors and small modular reactors. Furthermore, fabrication of high alloy materials/components may require the use of new manufacturing processes to achieve acceptable properties for higher temperature applications such as those in Generation IV applications. Current research by EPRI and DOE will be reviewed and emphasis will be targeted at advanced applications where PM/HIP may be applied in the future.

Non-Linear BWR Stability Analysis

2008 ◽  
Author(s):  
F. Wehle ◽  
A. Schmidt ◽  
S. Opel ◽  
R. Velten
Keyword(s):  
Stability Analysis ◽  
Nuclear Power ◽  
Reverse Flow ◽  
High Amplitude ◽  
Boiling Water Reactors ◽  
Linear Regime ◽  
Power Oscillations ◽  
Non Linear ◽  
Water Reactors ◽  
Physical Reasons

Power oscillations associated with density waves in boiling water reactors (BWRs) have been studied widely. Industrial research in this area is active since the invention of the first BWR. Stability measurements have been performed in various plants already during commissioning phase but especially the magnitude and divergent nature of the oscillations during the LaSalle Unit 2 nuclear power plant event on March 9, 1988, renewed concern about the state of knowledge oN BWR instabilities. The appropriate representation of the physical processes in the non-linear regime requires typically time domain stability analysis. The objective of this paper is to present a physical model, applicable for stability analysis in the non-linear regime, which extends to high amplitude oscillations where inlet reverse flow occurs. The application of this model gives a deeper insight into the physical reasons for the prevention of the uncontrolled divergence of BWR oscillations. The mechanisms that have a stabilizing effect are demonstrated.

Pressure Blow Out Assessment for Class 1 Piping With Local Wall Thinning

2006 ◽  
Author(s):  
Kunio Hasegawa ◽  
Katsuyuki Shibata
Keyword(s):  
Internal Pressure ◽  
Nuclear Power ◽  
Tensile Load ◽  
Evaluation Methodology ◽  
Evaluation Procedure ◽  
Seismic Load ◽  
Analytical Evaluation ◽  
Wall Thinning ◽  
Bending Load ◽  
Class 1

Wall thinning caused by the flow of water in power piping systems became a major concern to the nuclear power industries. ASME Code Case N-597-3, “Requirements for Analytical Evaluation of Pipe Wall Thinning,” provides procedures and criteria for Code Class 2 and 3 piping for the evaluation of wall thinning. However, analytical evaluation procedure for Class 1 piping is not provideed in the Code Case. Recent full-scale experiments on locally thinned pipes have supported the development of more contemporary failure strength evaluation methodology for Class 1 piping. These evaluation methodologies are applicable for the loading type of bending, tensile or cyclic bending load. Prior to the failure by bending moment, tensile load or cyclic/seismic load, locally wall thinned pipes shall be considered pressure blow out by the internal pressure itself. This paper introduces the failure of a uniformly thinned cylinder by an internal pressure and describes limitation on wall thinning depth to avoid pressure blow out for Class 1 piping.

New Trends in Accurate Simulations for the Verification of Safety Margins in the Nuclear Power Plant Industry: Application of Virtual Performance Solution™ for the Response of Immersed Structures Subjected to Earthquakes

2011 ◽  
Author(s):  
Alain Tramec¸on ◽  
Jorg Kuhnert ◽  
Laurent Mouchette ◽  
Morgane Perrin
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Dynamic Simulations ◽  
Finite Point ◽  
Fluid Structure ◽  
Finite Element Software ◽  
Mesh Free ◽  
Safety Margins ◽  
Non Linear

Constraints on the safety of nuclear power plant components have increased recently along with the necessity to extend the lifespan of existing plants. For example, the acceleration levels to be sustained by the plant equipment during an earthquake have been increased many folds by the safety regulation agencies. Industrial and economic requirements plead for a verification of unknown safety margins, by accurate and physics based models taking into account all non-linear effects (for example contacts and fluid structure interaction). These effects are only approximately represented by standard linear analysis tools. Virtual Performance Solution (VPS), developed by ESI Group, includes (among other capabilities) a structural finite element software for non-linear, high velocity, dynamic simulations (PAM-CRASH), as well as a coupled, mesh free CFD module, FPM (Finite Point Method), developed in partnership with Fraunhofer ITWM. This solution accurately predicts fluid structure interactions, taking into account non-linear structural effects (contacts, friction, damping…) as well as complex fluid influences.

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