COG Risk-Informed In-Service Inspection (RI-ISI) Project

2012 ◽  
Author(s):  
Patrick J. O’Regan ◽  
Mehdi Rezaie-Manesh ◽  
Scott T. Chesworth
Keyword(s):  
Operating Experience ◽  
Nuclear Safety ◽  
Degradation Mechanisms ◽  
Acceptance Criteria ◽  
Risk Levels ◽  
Periodic Inspection ◽  
Starting Point ◽  
Best Fit ◽  
Service Inspection

In 2009, the Canadian Nuclear Safety Commission (CNSC) prepared “Guidelines for Risk-Informed In-service Inspection for Piping,” which provided guidance and acceptance criteria for licensees to develop an RI-ISI program as an alternate to the current CSA N285.4 and augmented programs for piping inspection. A project was initiated by the CANDU Owners Group (COG) to develop a best-fit RI-ISI methodology for the CANDU design and evaluate plant risk levels between current and RI-ISI inspection programs. The traditional EPRI RI-ISI methodology was selected as a starting point, and four plant systems were evaluated. Both failure potential and consequence of failure were used to establish the risk significance for all in-scope piping components. Once the risk associated with each component was established, elements were selected for inspection based upon the sampling percentages of the EPRI RI-ISI methodology, and as a final check a comparison was made between plant risk under the current CSA N285.4 (and augmented) inspection program and under RI-ISI. The project was successfully concluded in 2011, and results confirmed that the EPRI RI-ISI methodology can be adapted to the CANDU design and the degradation mechanisms evaluated under RI-ISI are consistent with CANDU operating experience. The original CSA N285.4 basis for the CANDU Periodic Inspection Program (PIP) was validated, and potential improvements to the station inspection programs were identified.

CANDU Owners Group Risk-Informed In-Service Inspection (RI-ISI) Project

2013 ◽  
Author(s):  
Paul Lafreniere ◽  
Patrick O’Regan ◽  
Mehdi Rezaie-Manesh ◽  
Scott Chesworth
Keyword(s):  
Value Added ◽  
Nuclear Safety ◽  
Acceptance Criteria ◽  
Methodology Development ◽  
Starting Point ◽  
Nuclear Systems ◽  
Conventional Systems ◽  
Best Fit ◽  
Service Inspection

The Canadian Nuclear Safety Commission (CNSC) has prepared “Guidelines for Risk-Informed In-service Inspection (RI-ISI) for Piping.” This document provided instructions and acceptance criteria for licensees to develop an RI-ISI program integrating both risk insights and traditional analysis as an alternate to the current programs for piping inspection. In response to the CNSC guideline, a CANDU Owners Group (COG) project was initiated to develop a best-fit methodology for nuclear systems within the CANDU design. The traditional EPRI RI-ISI methodology was selected as a starting point for the CANDU best-fit methodology development. Since characterization of risk requires knowledge of both failure potential and the consequence of failure, these values were determined for in-scope piping welds. Once the risk associated with each component was established, the components were ranked accordingly, elements were selected based upon the sampling percentages of EPRI RI-ISI, and a comparison was made between plant risk under the current CSA N285.4 / augmented programs and RI-ISI. Application to a number of nuclear and non-nuclear systems has been completed and has shown where inspection allocation can be improved and where low-value added inspections can be reduced. Work is underway to extend the results of this project to conventional systems and components beyond piping welds.

Regulatory Perspective for the Definition of Probabilistic Acceptance Criteria for CANDU Pressure Tubes

2016 ◽  
Author(s):  
Sankar Laxman ◽  
Blair Carroll ◽  
John Jin
Keyword(s):  
Probabilistic Method ◽  
Operating Experience ◽  
Service Evaluation ◽  
Steering Committee ◽  
Canadian Standard Association ◽  
Degradation Mechanisms ◽  
Acceptance Criteria ◽  
Reactor Unit ◽  
Technical Requirements ◽  
Pressure Tubes

For the assurance of fitness-for-service of CANDU Pressure Tubes (PTs), guidelines and acceptance criteria are provided in Canadian Standard Association (CSA) N285.8-15, Technical requirements for in-service evaluation of zirconium alloy pressure Tubes in CANDU reactors. With respect to the assessment of risk of operation associated with degradation mechanisms and aging of the PTs in the entire core of a given reactor Unit, Clause 7 of CSA N285.8 allows Licensee’s to use either a deterministic or probabilistic method to assess the likelihood of PT failures. When a probabilistic method is used, the Licensee is obligated to demonstrate that the combined frequency of PT failure(s) over the evaluation period, due to the various potential degradation mechanisms, is less than the maximum acceptable frequency provided in Table C.1 of CSA N285.8-15. The maximum acceptable frequency provided in Table C.1 of CSA N285.8-15 was developed in the early-1990’s based on reactor operating experience and knowledge at that time, Station Siting Guides and Consultative Regulatory Guide C-006 (Revision 1). A task group was established by the CSA N285.8 Technical Steering Committee to re-evaluate the allowable failure frequencies to confirm that they remain relevant given the current state of knowledge and the additional evaluation tools available. This paper provides Canadian Nuclear Safety Commission staff views regarding the technical basis for revisions to the allowable frequencies based upon current industry practices in conducting probabilistic core assessments.

Derivation of Design Fatigue Curves for Austenitic Stainless Steel Grades 1.4541 and 1.4550 Within the German Nuclear Safety Standard KTA 3201.2

2013 ◽  
Author(s):  
Xaver Schuler ◽  
Karl-Heinz Herter ◽  
Jürgen Rudolph
Keyword(s):  
Austenitic Stainless Steel ◽  
Stainless Steels ◽  
Room Temperature ◽  
Elevated Temperatures ◽  
Austenitic Stainless Steels ◽  
Nuclear Safety ◽  
Safety Standard ◽  
Fatigue Design ◽  
Steel Grades ◽  
Best Fit

Titanium and niobium stabilized austenitic stainless steels X6CrNiTi18-10S (material number 1.4541, correspondent to Alloy 321) respectively X6CrNiNb18-10S (material number 1.4550, correspondent to Alloy 347) are widely applied materials in German nuclear power plant components. Related requirements are defined in Nuclear Safety Standard KTA 3201.1. Fatigue design analysis is based on Nuclear Safety Standard KTA 3201.2. The fatigue design curve for austenitic stainless steels in the current valid edition of KTA 3201.2 is essentially identical with the design curve included in ASME-BPVC III, App I (ed. 2007, add. July 2008 respectively back editions). In the current code revision activities of KTA 3201.2 the compatibility of latest in air fatigue data for austenitic stainless steels with the above mentioned grades were examined in detail. The examinations were based on statistical evaluations of 149 strain controlled test data at room temperature and 129 data at elevated temperatures to derive best-fit mean data curves. Results of two additional load controlled test series (at room temperature and 288°C) in the high cycle regime were used to determine a technical endurance limit at 107 cycles. The related strain amplitudes were determined by consideration of the cyclic stress strain curve. The available fatigue data for the two austenitic materials at room temperature and elevated temperatures showed a clear temperature dependence in the high cycle regime demanding for two different best-fit curves. The correlation of the technical endurance limit(s) at room temperature and elevated temperatures with the ultimate strength of the materials is discussed. Design fatigue curves were derived by application of the well known factors to the best-fit curves. A factor of SN = 12 was applied to load cycles correspondent to the NUREG/CR-6909 approach covering influences of data scatter, surface roughness, size and sequence. In terms of strain respectively stress amplitudes in the high cycle regime, for elevated temperatures (>80°C) a factor of Sσ = 1.79 was applied considering and combining in detail the partial influences of data scatter surface roughness, size and mean stress. For room temperature a factor of Sσ = 1.88 shall be applied. As a result, new design fatigue curves for austenitic stainless steel grades 1.4541 and 1.4550 will be available within the German Nuclear Safety Standard KTA 3201.2. The fatigue design rules for all other austenitic stainless steel grades will be based on the new ASME-BPVC III, App I (ed. 2010) design curve.

scholarly journals Investigation of 2 Models to Set and Evaluate Quality Targets for Hb A1c: Biological Variation and Sigma-Metrics

2015 ◽  
Vol 61 (5) ◽  
pp. 752-759 ◽  
Author(s):  
Cas Weykamp ◽  
Garry John ◽  
Philippe Gillery ◽  
Emma English ◽  
Linong Ji ◽  
...  
Keyword(s):  
Task Force ◽  
Performance Criterion ◽  
Biological Variation ◽  
Risk Levels ◽  
Generic Models ◽  
Hb A1c ◽  
Starting Point ◽  
Allowable Error ◽  
The Individual ◽  
Metrics Model

Abstract BACKGROUND A major objective of the IFCC Task Force on Implementation of HbA1c Standardization is to develop a model to define quality targets for glycated hemoglobin (Hb A1c). METHODS Two generic models, biological variation and sigma-metrics, are investigated. We selected variables in the models for Hb A1c and used data of external quality assurance/proficiency testing programs to evaluate the suitability of the models to set and evaluate quality targets within and between laboratories. RESULTS In the biological variation model, 48% of individual laboratories and none of the 26 instrument groups met the minimum performance criterion. In the sigma-metrics model, with a total allowable error (TAE) set at 5 mmol/mol (0.46% NGSP), 77% of the individual laboratories and 12 of 26 instrument groups met the 2σ criterion. CONCLUSIONS The biological variation and sigma-metrics models were demonstrated to be suitable for setting and evaluating quality targets within and between laboratories. The sigma-metrics model is more flexible, as both the TAE and the risk of failure can be adjusted to the situation—for example, requirements related to diagnosis/monitoring or international authorities. With the aim of reaching (inter)national consensus on advice regarding quality targets for Hb A1c, the Task Force suggests the sigma-metrics model as the model of choice, with default values of 5 mmol/mol (0.46%) for TAE and risk levels of 2σ and 4σ for routine laboratories and laboratories performing clinical trials, respectively. These goals should serve as a starting point for discussion with international stakeholders in the field of diabetes.

scholarly journals Hysteresis Modelling of Mechanical Systems at Nonstationary Vibrations

2018 ◽  
Vol 2018 ◽  
pp. 1-15
Author(s):  
A. N. Danilin ◽  
A. D. Shalashilin
Keyword(s):  
Differential Equation ◽  
Phenomenological Models ◽  
Low Frequency ◽  
Hysteresis Cycle ◽  
Nonstationary Vibrations ◽  
Kinematic Parameters ◽  
First Order ◽  
Starting Point ◽  
The Right ◽  
Best Fit

This paper considers and reviews a number of known phenomenological models, used to describe hysteretic effects of various natures. Such models consider hysteresis system as a “black box” with experimentally known input and output, related via formal mathematical dependence to parameters obtained from the best fit to experimental data. In particular, we focus on the broadly used Bouc-Wen and similar phenomenological models. The current paper shows the conditions which the Bouc-Wen model must meet. An alternative mathematical model is suggested where the force and kinematic parameters are related by a first-order differential equation. In contrast to the Bouc-Wen model, the right hand side is a polynomial with two variables representing hysteresis trajectories in the process diagram. This approach ensures correct asymptotic approximation of the solution to the enclosing hysteresis cycle curves. The coefficients in the right side are also determined experimentally from the hysteresis cycle data during stable oscillations. The proposed approach allows us to describe hysteretic trajectory with an arbitrary starting point within the enclosed cycle using only one differential equation. The model is applied to the description of forced vibrations of a low-frequency pendulum damper.

Advances and Research Needs in Risk-Informed In-Service Inspection Methodology

2010 ◽  
Author(s):  
X.-X. Yuan
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Industry ◽  
Small Scale ◽  
Regulatory Body ◽  
Successful Implementation ◽  
Research Needs ◽  
Pilot Studies ◽  
Periodic Inspection ◽  
Service Inspection

While risk-informed in-service inspection (RI-ISI) program has been applied in several countries to enhance the traditional periodic inspection program (PIP), many other countries are waiting for more successful implementation experiences to be accumulated. Canadian Nuclear Safety Commission (CNSC), the regulatory body of nuclear industry in Canada, became increasingly interested in the risk-informed decision making methodology. Several small-scale pilot studies on RI-ISI have been initiated by Canadian utilities during the past few years. Nevertheless, a RI-ISI methodology appropriate for the CANDU technology that can be accepted by the stakeholders has yet to be developed. The development of the RI-ISI methodologies derived from the PWR/BWR operating experiences is first reviewed, followed by an examination of Canadian periodic inspection standard CSA N285.4 and its evolution from a RI-ISI perspective. Finally several key technical issues and research needs in developing an advanced RI-ISI methodology for nuclear power plants are identified.

Building the Technical Case for Long-Term Operation

2015 ◽  
Author(s):  
Richard Tilley ◽  
Robin Dyle
Keyword(s):  
Operating Experience ◽  
Academic Community ◽  
Degradation Mechanisms ◽  
Comprehensive Understanding ◽  
Term Operation ◽  
Societal Benefits ◽  
Actual Operating ◽  
Nuclear Plants ◽  
The Us

United States (US) and International utilities are actively engaged in assessing the economic and societal benefits of operating nuclear plants beyond their initial license periods. Nuclear plant generated electricity is still the largest contributor to non-carbon dioxide emitting generation. In the US, a majority of operating plants has already received approval for an additional 20 years of operation, and soon it is expected that utilities will begin the process to seek a second 20 year renewal. The keys to successful renewal are to maintain safe and reliable operations by building a sound technical case through the following activities: • Develop comprehensive understanding of aging degradation issues for systems, structures and components (SSCs) • Implement specific plant aging management programs to address aging degradation • Confirm behavior of degradation mechanisms for the entire period of operation This paper will step through the above elements to illustrate how a strong technical case may be created for safe and reliable long-term operation. Examples or case studies will be provided to clearly link the fundamental science of materials degradation to the inspection, testing and evaluation efforts implemented at a plant and to the confirmatory data that is provided by both actual operating experience and the extensive research and development projects pursued by industry, governments, and the academic community.

New Classification Assessment Rule for Defects Fitness

2015 ◽  
Author(s):  
Zhao-he Chen
Keyword(s):  
New Classification ◽  
Acceptance Criteria ◽  
Service Inspection

Based upon three typical materials and metrology stipulated in GB/T 19624, a new classification assessment rule for defects which exceed the acceptance criteria in construction code is given, which includes three levels depending upon the information acquired during assessment. Following its implementation into in-service inspection projects, the compliance with normal assessment can be attained without much effort.

Life Prediction and Monitoring of Nuclear Power Plant Components for Service-Related Degradation

2000 ◽  
Vol 123 (1) ◽  
pp. 58-64 ◽  
Author(s):  
Fredric A. Simonen ◽  
Stephen R. Gosselin
Keyword(s):  
Life Prediction ◽  
Nuclear Power ◽  
Fatigue Crack Initiation ◽  
Operating Experience ◽  
Degradation Mechanisms ◽  
Original Design ◽  
Structural Degradation ◽  
Plant Safety ◽  
Degradation Data ◽  
Design Life

This paper describes industry programs to manage structural degradation and to justify continued operation of nuclear components when unexpected degradation has been encountered due to design materials and/or operational problems. Other issues have been related to operation of components beyond their original design life in cases where there is no evidence of fatigue crack initiation or other forms of structural degradation. Data from plant operating experience have been applied in combination with inservice inspections and degradation management programs to ensure that the degradation mechanisms do not adversely impact plant safety. Probabilistic fracture mechanics calculations are presented to demonstrate how component failure probabilities can be managed through augmented inservice inspection programs.

Sign in / Sign up

Export Citation Format

Share Document