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).

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.

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.

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.

scholarly journals CANDU FIRE DATABASE

2019 ◽  
Vol 8 (2) ◽  
pp. 179-189 ◽  
Author(s):  
Hossam Shalabi ◽  
George Hadjisophocleous
Keyword(s):  
Risk Assessment ◽  
Time Series ◽  
Nuclear Power ◽  
Data Exchange ◽  
Power Plants ◽  
Nuclear Energy ◽  
Energy Agency ◽  
Candu Reactors ◽  
Weighting Factors ◽  
Future Potential

The Nuclear Energy Agency (NEA) is a specialized agency within the Organization for Economic Co-operation and Development (OECD). The International Fire Data Exchange Project (OECD FIRE) was designed by the NEA to encourage multilateral co-operation in the collection and analysis of data relating to fire events in nuclear power plants. We used Python advanced software to analyze the data related to CANDU reactor plants in Canada from the OECD FIRE Database, while providing weighting factors/percentage tables to be used in CANDU Fire probabilistic risk assessment analysis. We also used 5 different time-series methods to predict future potential fires in CANDU reactors, compared the results from different methods, and identified the best method to predict future fires in CANDU power plants.

OECD Pipe Failure Data Exchange Project (OPDE): 2003 Status Report

2004 ◽  
Author(s):  
Bengt Lydell ◽  
Eric Mathet ◽  
Karen Gott
Keyword(s):  
Data Exchange ◽  
Influence Factors ◽  
Status Report ◽  
First Year ◽  
Pipe Failure ◽  
Project Organization ◽  
Energy Agency ◽  
Reliability Parameters ◽  
Failure Data ◽  
Plant Operations

Established in 2002 as a three-year multilateral cooperation in the collection and analysis of data relating to pipe failure events, the OPDE project currently is supported by eighteen organizations from twelve member countries of the Nuclear Energy Agency of the Organisation for Economic Co-operation and Development (OECD-NEA). Objectives of the OPDE are to collect and analyze pipe failure data to promote a better understanding of underlying causes of failure, observed and potential impact on plant operations and safety, and prevention. The data analysis includes characterization of pipe failure events by reliability attributes and influence factors to facilitate the estimation of piping reliability parameters. With emphasis on data validity and quality, during the first year of operation a coding format has been developed to ensure consistent interpretation and applications. This paper represents an abbreviated 2003 status report and an overview of the project organization and objectives.

Design and Analysis of a Reliable Communication System in Nuclear Safety Instrument & Control System

2021 ◽  
Author(s):  
Le Li ◽  
Zhihui Zhang ◽  
Chao Gao ◽  
Fei Zhou ◽  
Guangqiang Ma
Keyword(s):  
Communication Networks ◽  
Communication System ◽  
Communication Systems ◽  
Nuclear Power ◽  
Data Exchange ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Safety ◽  
Safety Communication ◽  
Digital Control Systems

Abstract With the development of digital instrument and control technology for nuclear power plants in recent decades, communication networks have become an important part of safety digital control systems, which takes charge in data exchange between the various sub-systems, and extremely impact on the reliability and safety of the entire I&C system. Traditional communication systems where some special features, such as reliability, safety, real-time, certainty, and independence are not strictly required are various illustrated. However, how to implement a communication system in a safety I&C system is rarely stated in current research. In this research, a reliable safety communication system applied in nuclear power plants is designed and analyzed. The five key characteristics of nuclear safety communication networks are explained, followed by explanation of how to achieve these characteristics. The analysis and verification of the designed system are also stated in this paper, which contributes to proving that the designed nuclear safety communication system could applied in the nuclear power plants.

scholarly journals Study on Probabilistic Safety Goals for Multimodule High-Temperature Gas-Cooled Reactor Based on Chinese Societal Risks

2021 ◽  
Vol 2021 ◽  
pp. 1-10
Author(s):  
Jinghan Zhang ◽  
Jun Zhao ◽  
Jiejuan Tong
Keyword(s):  
High Temperature ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Regulatory Commission ◽  
Nuclear Safety ◽  
Energy Agency ◽  
Water Reactor ◽  
Regulatory Commission ◽  
High Temperature Gas ◽  
Probabilistic Safety

Nuclear safety goal is the basic standard for limiting the operational risks of nuclear power plants. The statistics of societal risks are the basis for nuclear safety goals. Core damage frequency (CDF) and large early release frequency (LERF) are typical probabilistic safety goals that are used in the regulation of water-cooled reactors currently. In fact, Chinese current probabilistic safety goals refer to the Nuclear Regulatory Commission (NRC) and the International Atomic Energy Agency (IAEA), and they are not based on Chinese societal risks. And the CDF and LERF proposed for water reactor are not suitable for high-temperature gas-cooled reactors (HTGR), because the design of HTGR is very different from that of water reactor. And current nuclear safety goals are established for single reactor rather than unit or site. Therefore, in this paper, the development of the safety goal of NRC was investigated firstly; then, the societal risks in China were investigated in order to establish the correlation between the probabilistic safety goal of multimodule HTGR and Chinese societal risks. In the end, some other matters about multireactor site were discussed in detail.

INCEFA-PLUS Project: Lessons Learned From the Project Data and Impact on Existing Fatigue Assessment Procedures

2020 ◽  
Author(s):  
Sam Cuvilliez ◽  
Alec McLennan ◽  
Kevin Mottershead ◽  
Jonathan Mann ◽  
Matthias Bruchhausen
Keyword(s):  
Surface Finish ◽  
Environmental Effects ◽  
Nuclear Power ◽  
Power Plants ◽  
Lessons Learned ◽  
Fatigue Assessment ◽  
Data Points ◽  
Project Data ◽  
Assessment Procedures ◽  
Surface Finish Effect

Abstract The INCEFA+ project (INcreasing Safety in nuclear power plants by Covering gaps in Environmental Fatigue Assessment) is a five year project supported by the European Commission HORIZON2020 programme, which will conclude in June 2020. This project aims to generate and analyse Environmental Assisted Fatigue (EAF) experimental data (approximately 230 fatigue data points generated on austenitic stainless steel), and focuses on the effect of several key parameters such as mean strain, hold times and surface finish, and how they interact with environmental effects (air or PWR environment). This work focuses on the analysis of the data obtained during the INCEFA+ project. More specifically, this paper discusses how the outcome of this analysis can be used to evaluate existing fatigue assessment procedures that incorporate environmental effects in a similar way to NUREG/CR-6909. A key difference between these approaches and the NUREG/CR-6909 is the reduction of conservatisms resulting from the joint implementation of the adjustment sub-factor related to surface finish effect (as quantified in the design air curve derivation) and a Fen penalization factor for fatigue assessment of a location subjected to a PWR primary environment. The analysis presented in this paper indicates that the adjustment (sub-)factor on life associated with the effect of surface finish in air (as described in the derivation of the design air curve in NUREG/CR-6909) leads to substantial conservatisms when it is used to predict fatigue lifetimes in PWR environments for rough specimens. The corresponding margins can be explicitly quantified against the design air curve used for EAF assessment, but may also depend on the environmental correction Fen factor expression that is used to take environmental effects into account.

Probabilistic Fracture Mechanics Benchmarking Study Involving the xLPR and PASCAL-SP Codes: Analysis by PASCAL-SP

2020 ◽  
Author(s):  
Akihiro Mano ◽  
Jinya Katsuyama ◽  
Yinsheng Li
Keyword(s):  
Fracture Mechanics ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Nuclear Power ◽  
Power Plants ◽  
Corrosion Cracking ◽  
Energy Agency ◽  
Pressurized Water ◽  
Water Reactor ◽  
Probabilistic Fracture Mechanics

Abstract A probabilistic fracture mechanics (PFM) analysis code, PASCAL-SP, has been developed by Japan Atomic Energy Agency (JAEA) to evaluate the failure probability of piping within nuclear power plants considering aged-related degradations such as stress corrosion cracking and fatigue for both pressurized water reactor and boiling water reactor environments. To strengthen the applicability of PASCAL-SP, a benchmarking study is being performed with a PFM analysis code, xLPR, which has been developed by U.S.NRC in collaboration with EPRI. In this benchmarking study, deterministic and probabilistic analyses are undertaken on primary water stress corrosion cracking using the common analysis conditions. A deterministic analysis on the weld residual stress distributions is also considered. These analyses are carried out by U.S.NRC and JAEA independently using their own codes. Currently, the deterministic analyses by both xLPR and PASCAL-SP codes have been finished and probabilistic analyses are underway. This paper presents the details of conditions and comparisons of the results between the two aforementioned codes for the deterministic analyses. Both codes were found to provide almost the same results including the values of stress intensity factor. The conditions and results of the probabilistic analysis obtained from PASCAL-SP are also discussed.

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