A Stochastic Model for Piping Failure Frequency Analysis Using OPDE Data

2009 ◽  
Vol 131 (5) ◽  
Author(s):  
X.-X. Yuan ◽  
M. D. Pandey ◽  
J. Riznic
Keyword(s):  
Nuclear Power ◽  
Process Model ◽  
Power Plants ◽  
Probabilistic Models ◽  
Accurate Estimation ◽  
Model Parameters ◽  
Pipe Failure ◽  
Failure Frequency ◽  
Failure Data ◽  
Class 1

The accurate estimation of piping failure frequency is an important task to support the probabilistic risk assessment and risk-informed in-service inspection of nuclear power plants. Although probabilistic models have been reported in the literature to analyze the piping failure frequency, this paper proposes a stochastic point process model that incorporates both a time dependent trend and plant-specific (or cohort) effects on the failure rate. A likelihood based statistical method is proposed for estimating the model parameters. A case study is presented to analyze the Class 1 pipe failure data given in the OPDE Database.

A Point Process Model for Piping Failure Frequency Analysis Using OPDE Data

2008 ◽  
Author(s):  
X.-X. Yuan ◽  
M. D. Pandey ◽  
J. Riznic
Keyword(s):  
Point Process ◽  
Nuclear Power ◽  
Process Model ◽  
Power Plants ◽  
Probabilistic Models ◽  
Accurate Estimation ◽  
Model Parameters ◽  
Pipe Failure ◽  
Point Process Model ◽  
Failure Frequency

The accurate estimation of piping failure frequency is an important task to support the probabilistic risk assessment and risk-informed in-service inspection of nuclear power plants. Although probabilistic models have been reported in the literature to analyze the piping failure frequency, this paper proposes a stochastic point process model that incorporates both a time dependent trend and plant specific (or cohort) effects on the failure rate. A likelihood based statistical method is proposed for estimating the model parameters. A case study is presented to analyze the Class 1 pipe failure data given in the OPDE Database.

Progress With the International Pipe Failure Data Exchange Project

2007 ◽  
Author(s):  
Bengt Lydell ◽  
Alejandro Huerta ◽  
Karen Gott
Keyword(s):  
Nuclear Power ◽  
Data Exchange ◽  
Power Plants ◽  
Crack Depth ◽  
Pipe Failure ◽  
Energy Agency ◽  
Service Experience ◽  
Failure Data ◽  
Wall Thinning ◽  
Engineered Safeguards

Certain member countries of the Organisation for Economic Cooperation and Development (OECD) in 2002 established the OECD Pipe Failure Data Exchange Project (OPDE) to produce an international database on the piping service experience applicable to commercial nuclear power plants. OPDE is operated under the umbrella of the OECD Nuclear Energy Agency (NEA). The Project collects pipe failure data including service-induced wall thinning, part through-wall crack, pinhole leak, leak, and rupture/severance (i.e., events involving large leak rates up to and beyond the make-up capacity of engineered safeguards systems). The part through-wall events include degradation in excess of code allowable for pipe wall thinning or crack depth. OPDE also addresses such degradation that could have generic implications regarding the reliability of in-service inspection. At the end of 2006 the OPDE database included approximately 3,700 records on pipe failure affecting ASME Code Class 1 through 3 and non-Code piping. This paper summarizes the unique data quality considerations that are associated with the reporting and recording of piping component degradation and failure. The paper also summarizes the database content and puts it in perspective relative to past efforts to systematically collect and evaluate service experience data on piping performance.

The OECD Pipe Failure Data Exchange Project: Data Validation on Canadian Plants

2004 ◽  
Author(s):  
Alexandre Colligan ◽  
Robert Lojk ◽  
Jovica Riznic
Keyword(s):  
Data Collection ◽  
Nuclear Power ◽  
Data Exchange ◽  
Power Plants ◽  
Data Validation ◽  
Pipe Failure ◽  
Energy Agency ◽  
Failure Data ◽  
Project Data ◽  
Project Objectives

The Nuclear Energy Agency (NEA) of the Organization for Economic Co-Operation and Development (OECD) has initiated a project to establish an international pipe failure data collection and exchange program. This OECD Pipe Failure Data Exchange (OPDE) Project has been established to encourage multilateral co-operation in the collection and analysis of data relating to pipe failure events in commercial nuclear power plants. This paper presents a brief description of the ODPE project objectives and work scope, as well as the Canadian contribution on data validation with respect to development and application of the pipe failure data collection on which OPDE is based.

The OECD Pipe Failure Data Exchange Project: Validation of Canadian Data

2006 ◽  
Author(s):  
Tom Viglaski ◽  
Andrei Blahoianu ◽  
Bengt Lydell ◽  
Jovica Riznic
Keyword(s):  
Data Collection ◽  
Nuclear Power ◽  
Data Exchange ◽  
Power Plants ◽  
Structural Integrity ◽  
Passive Component ◽  
Pipe Failure ◽  
Energy Agency ◽  
Failure Data ◽  
Project Objectives

Structural integrity of piping systems is important to plant safety and operability. Information on degradation and failure of piping components and systems is collected and evaluated by regulatory agencies, international organizations (e.g., OECD/NEA and IAEA) and industry organizations worldwide to establish systematic feedback to reactor regulation and research and development programs associated with non-destructive examination (NDE) technology, in-service inspection (ISI) programs, leak-before-break evaluations, risk-informed ISI, and probabilistic safety assessment (PSA) applications involving passive component reliability. In 2002, the Nuclear Energy Agency (NEA) of the Organization for Economic Co-Operation and Development (OECD) has initiated an international pipe failure data collection and exchange project. The OECD Pipe Failure Data Exchange (OPDE) Project has been established to encourage multilateral co-operation in the collection and analysis of data relating to pipe failure events in commercial nuclear power plants. At present, the database contains 3644 records to which twelve participating countries contributed. This paper presents a brief description of the ODPE project objectives and work scope, as well as the Canadian contribution on data validation with respect to development and application of the pipe failure data collection on which OPDE is based. It gives a number of tables and figures that can be obtained from these records, with selected data ranging from a very broad (i.e. level of participation in the database from each member country), to very specific (i.e. plant operational state at time of pipe failure discovery for CANDU reactors).

Probabilistic Models of Reliability of Cast Austenitic Stainless Steel Piping

2011 ◽  
Author(s):  
Haiyang Qian ◽  
David Harris ◽  
Timothy J. Griesbach
Keyword(s):  
Fracture Toughness ◽  
Stainless Steel ◽  
Austenitic Stainless Steel ◽  
Tensile Properties ◽  
Nuclear Power ◽  
Power Plants ◽  
Probabilistic Models ◽  
Probabilistic Approach ◽  
Delta Ferrite ◽  
Steel Piping

Thermal embrittlement of cast austenitic stainless steel piping is of growing concern as nuclear power plants age. The difficulty of inspecting these components adds to the concerns regarding their reliability, and an added concern is the presence of known defects introduced during the casting fabrication process. The possible presence of defects and difficulty of inspection complicate the development of programs to manage the risk contributed by these embrittled components. Much work has been done in the past to characterize changes in tensile properties and fracture toughness as functions of time, temperature, composition, and delta ferrite content, but this work has shown a great deal of scatter in relationships between the important variables. The scatter in material correlations, difficulty of inspection and presence of initial defects calls for a probabilistic approach to the problem. The purpose of this study is to describe a probabilistic fracture mechanics analysis of the maximum allowable flaw sizes in cast austenitic stainless steel piping in commercial power reactors. Attention is focused on fully embrittled CF8M material, and the probability of failure for a given crack size, load and composition is predicted considering scatter in tensile properties and fracture toughness (fracture toughness is expressed as a crack growth resistance relation in terms of J-Δa). Random loads can also be included in the analysis, with results generated by Monte Carlo simulation. This paper presents preliminary results for CF8M to demonstrate the sensitivity of key input variables. The outcome of this study is the flaw sizes (length and depth) that will fail with a given probability when a given load is applied.

Evaluation of Flange Calculations Using Strain-Based Acceptance Criteria

2019 ◽  
Author(s):  
Alexander Mutz ◽  
Manfred Schaaf
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Elastic Stress ◽  
Calculation Methods ◽  
Acceptance Criteria ◽  
Class 1 ◽  
Torque Moment ◽  
Calculation Algorithms ◽  
Calculation Code

Abstract The Nuclear Power Plant KKG in Gösgen, Switzerland was designed according to the ASME Boiler and Pressure Vessel Code. The ASME BPVC, Section III, Appendix 11 regulates the flange calculation for class 2 and 3 components, it is also used for class 1 flanges. A standard for the determination of the required gasket characteristics is not well established which leads to a lack of clarity. As a hint different y and m values for different kinds of gasket are invented in ASME BPVC Section III [1]. The KTA 3201.2[2] and KTA 3211.2[3] regulate the calculation of bolted flanged joints in German nuclear power plants. The gasket characteristics required for these calculation methods are based on DIN 28090-1[4], they can be determined experimentally. In Europe, the calculation code EN 1591-1 [5] and the gasket characteristics according to EN 13555[6] are used for flange calculations. Because these calculation algorithms provide not only a stress analysis but also a tightness proof, it would be preferable to use them also in the NPP’s in Switzerland. Additionally, for regulatory approval also the requirements of the ASME BPVC must be fullfilled. For determining the bolting up torque moment of flanges several tables for different nominal diameters of flanges using different gaskets and different combinations of bolt and flange material were established. As leading criteria for an allowable state, the gasket surface pressure, the allowable elastic stress of the bolts and the strain in the flange should be a good and conservative basis for determining allowable torque moments. The herein established tables show only a small part according to a previous paper [7] where different calculation methods for determining bolting up moments were compared to each other. In this paper the bolting-up torque moments determined with the European standard EN 1591-1 for the flange, are assessed on the strain-based acceptance criteria in ASME BPVC, Section III, Appendices EE and FF. The assessment of the torque moment of the bolts remains elastically which should lead to a more conservative insight of the behavior of the flanges.

Piping Service Life Experience in Commercial Nuclear Power Plants: Progress With the OECD Pipe Failure Data Exchange Project

2004 ◽  
Author(s):  
Bengt Lydell ◽  
Eric Mathet ◽  
Karen Gott
Keyword(s):  
Data Exchange ◽  
Power Plants ◽  
Nuclear Energy ◽  
Structural Integrity ◽  
Life Experience ◽  
Crack Depth ◽  
Pipe Failure ◽  
Energy Agency ◽  
Failure Data ◽  
Wall Thinning

An extension of a 1994–98 R&D project established in 2002 by certain member countries of the Organisation for Economic Cooperation (OECD), the OECD Pipe Failure Data Exchange (OPDE) Project has produced a major database on the piping service experience applicable to commercial nuclear plants. The 3-year project is operated under the umbrella of the OECD Nuclear Energy Agency (NEA) and organizations producing or regulating more than 80% of nuclear energy generation worldwide are contributing data to the OPDE Project. The Project considers pipe failure data including service-induced wall thinning, part through-wall crack, pinhole leak, leak, and rupture/severance (i.e., events involving large leak rates up to and beyond the make-up capacity of engineered safeguards systems). The part through-wall events include degradation in excess of code allowable for pipe wall thinning or crack depth. OPDE also addresses such degradation that could have generic implications regarding the reliability of in-service inspection. At the end of 2003 the OPDE database included approximately 4,400 records on pipe failure affecting ASME Code Class 1 through 3 and non-Code piping. The database also included an additional 450 records on water hammer events where the structural integrity of piping was challenged but did not fail. This paper summarizes the unique data quality considerations that are associated with piping components. The paper also summarizes the database content.

Comparison of German KTA and ASME Nuclear Design Codes for Class 1, 2, 3 Components and Piping

2011 ◽  
Author(s):  
Daniel Hofer ◽  
Henry Schau ◽  
Hu¨seyin Ertugrul Karabaki ◽  
Ralph Hill
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Design Codes ◽  
Nuclear Facility ◽  
Design Standards ◽  
Design Rules ◽  
Design Specifications ◽  
Division 1 ◽  
Class 1 ◽  
Standard Organization

This paper compares the design rules of the ASME Boiler and Pressure Vessel Code, Section III, Division 1, Rules for Construction of Nuclear Facility Components, with German nuclear design standards for Class 1, 2, 3 components and piping. The paper is focused on a comparison of the equations for Design by Analysis and on Piping equations. The ASME Section III Code has been used in combination with design specifications for design of German nuclear power plants. Together with manufacturers, inspectors and power plant owners, the German regulatory authority decided to develop their own nuclear design standards. The current versions being used are from 1992 and 1996. New versions of KTA design standards for pressure retaining components (KTA 3201.2 and KTA 3211.2) are currently under development. This comparison will cover the major differences between the design rules for ASME Section III, Div. 1 and KTA standards 3201.2 and 3211.2 as well as code or standard organization by sections, paragraphs, articles and code development.

Design Issues and Considerations During Reactor Vessel Head Replacement, Steam Generator Replacement, and Steam Generator Snubber Elimination at Nuclear Power Plants

2004 ◽  
Author(s):  
Dilip Bhavnani ◽  
James Annett
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Supply System ◽  
Reactor Vessel ◽  
Original Design ◽  
Pressurized Water ◽  
Class 1 ◽  
Current Design ◽  
Successful Design

One of the key maintenance activities in a nuclear power plant is the replacement of major components in the Nuclear Steam Supply System. In order to achieve significant operational improvements, the replacement components are not an exact replacement of the existing components. The replacement of components in the nuclear steam supply system in many Pressurized Water Reactor plants may include steam generators, replacement of reactor vessel heads with integrated head assemblies, and elimination of steam generator snubbers. The replacement components may not be supplied and/or designed by the original supplier. The changes in the components have to be compared to a plant’s current design and licensing bases and regulatory commitments. The qualification of these components involves non-linear, Nuclear Class 1 analyses, where portions of the configuration and analyses are proprietary, and there is a coupling of the response between the containment structure and the components. Ultimately, the qualification of the reactor coolant system and reactor vessel internals must be demonstrated, not just the qualification of the replacement components. A key element for the successful completion of these component replacements is the method by which the design and licensing bases is maintained and the work of the various groups involved in the design coordinated. This paper outlines how in a typical two unit PWR plant, major component replacements can impact original design bases and issues that should be considered in creating successful design and configuration documents. Design interface issues, configuration combinations, and coordination requirements are identified.

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