Managing CANDU Plant Ageing Using a Risk Informed Engineering Approach

Author(s):  
Keith S. Dinnie

A common challenge to operators of plants nearing the end of design life or undergoing life extension is to maintain safe and economic operation where multiple components are degrading simultaneously due to ageing effects. Typically, the management of ageing is carried out on a component-by-component basis but the real challenge is to ensure that the collective impacts of degradation are controlled such that the risk posed by continued operation of the plant remains acceptably small. The strategy being proposed to the Canadian industry is to use a risk-informed approach that derives failure frequency targets for individual components in a manner that ensures that the total risk remains within established limits. These frequency limits can then be embodied in fitness for service guidance to manage component reliability. The approach is to use the component importance measures in the plant PSA to derive the failure frequency that would result in a risk contribution of 1E−06 or less per reactor-year to the severe core damage frequency. Given that the safety goal limit for existing plants is 1E−04 per reactor-year, this would allow a number of components to be managed in this way without a significant increase in severe core damage frequency relative to the limit, where a cumulative increase of 1E−05 per reactor-year or more would be considered “significant”. A limit is placed on the derived “allowable” failure frequency for any individual component by deterministic considerations, in that the frequency is not permitted to exceed the maximum for the event class for which it was licensed in Canada. The frequency is also reviewed for economic and operability implications to ensure such risks are not unreasonably high. This approach helps to achieve a balanced allocation of inspection and maintenance resources as well as maintaining an adequate safety margin. The paper summarizes some of the challenges facing the current CANDU fleet, and provides examples of how the proposed approach could be applied to selected components. It should be noted that the approach is under consideration by the Canadian industry but is not committed at the present time.

2016 ◽  
Vol 42 (05) ◽  
pp. 518-525 ◽  
Author(s):  
Erik Berntorp ◽  
Nadine Andersson

There are two main bioengineering approaches to extending the half-life of factor (F)VIII or FIX products used for hemophilia replacement therapy. These are fusion to Fc-immunoglobulin G (FVIII and FIX) or to albumin (FIX) or pegylation/glycopegylation (FVIII and FIX). Four FVIII and three FIX products are in clinical development or have recently been licensed in regions of the world. The reported half-life extension is approximately 1.5-fold for FVIII and 2.5-fold, or even longer, for FIX. Clinical trials have shown promising results with respect to extension of dose intervals and efficacy in the treatment and prevention of bleeding events. The role of these products in clinical practice has been discussed in terms of either improving convenience and adherence through prolongation of the interval between infusions or maintaining current intervals thereby increasing trough levels and the safety margin against bleeds. This review of extended half-life products addresses the possibilities and problems of their introduction in hemophilia treatment.


Author(s):  
Gangyang Zheng ◽  
Yu Gong ◽  
Zhijian Zhang ◽  
Zibin Liu

With “theory of nuclear safety (TONS)”, this paper intends to explain the Core Damage (CD) scenario of a Nuclear Power Plant (NPP) with the systematic methodology, many notions introduced here can be extended to other types of nuclear installations, as well. This systematic methodology combines the Risk-Informed Safety Margin Characterization (RISMC) Metatheory of TONS, and the basic reliability theory. A “metatheory” of such theories, here, is a theory to analyze the Theory of Nuclear Safety (TONS); in its own theory system, it is designed to summarize the safety of a NPP. Meanwhile, the basic reliability theory, which is decided by the authors, is focus on the mission reliability model (a model can be established by Reliability Block Diagram (RBD)); then the related basic concepts, is simple and clear, and quite mature in NPP field. The present work outlines the traditional reliability theory and the RISMC-based Metatheory, and these two concepts here are taken as the appropriate TONS to analyze the CD Scenario, after that, a renovate or renew TONS, from these two sides, can be introduced to analyze the fundamental safety of NPP.


2014 ◽  
Vol 2014 ◽  
pp. 1-9
Author(s):  
Wenxue Qian ◽  
Xiaowei Yin ◽  
Liyang Xie

A component with multiple weak sites is widely used in practical engineering and the existence of multiple weak sites can significantly decrease the component reliability. On the other hand, only a few components bear static loading and most components bear dynamic loading. In this paper, a reliability model of isomorphic component with multiple weak sites is built based on an order statistics model and the influences of strength decentrality and loading decentrality on isomorphic component with multiple weak sites are discussed. Furthermore the influence of loading times is studied in detail. The results show that unlike a component with only one weak site, not only does the failure of a component with multiple weak sites have a relationship with the safety margin, but there also exist relationships with the number of weak sites, the loading roughness, and loading times. The work in this paper is of some guiding significance in reliability design and assessment of a component with multiple weak sites under complex loading.


Author(s):  
Gangyang Zheng ◽  
Paul Nelson ◽  
Vera Moiseytseva ◽  
Ernie Kee ◽  
Fatma Yilmaz

The U.S. Nuclear Regulatory Commission (NRC) is mandated to ensure “adequate protection” to the public health and safety, regardless of cost. It also has steadfastly declined to specify precisely what constitutes “adequate protection,” except that it does not mean “zero risk.” Rather it judges on a case-by-case basis whether the “adequate protection” standard has been met. NRC also seems to reserve the right to require an even higher level of protection, when that can be achieved in a manner that it judges to meet similarly imprecisely specified criteria such as “practicality” and “reasonableness.” In Regulatory Guide 1.174 NRC comes close to a concrete specification of “adequate protection,” albeit one that depends upon the historical licensing basis for a specific plant. And the technical portion of this paper begins with a description of how the approach of Regulatory Guide 1.174 can be viewed from the perspective of Risk-Informed Safety Margin Characterization. Meanwhile, in this research, in order to better understand the role of regulation, a microeconomic model of a price-taking nuclear power plant is constructed, particularly of the cost (C) of achieving any specified level of core damage frequency (CDF). Solution of this model reveals an economic optimum, at a point that balances plant value against risk of losing the plant via an accident involving core damage. For CDFs slightly smaller than this economic optimum there is scope for a regulatory mandate of even smaller CDF, should that be deemed either necessary to attain “adequate protection,” or reasonably attainable in order to achieve greater than adequate protection of the public health and safety. It is argued that regulatory bodies must have scope for discretionary decisions, because the information necessary to formulate a reasonable approximation to the cost curve C (fortunately) does not exist.


2011 ◽  
Vol 58-60 ◽  
pp. 529-534 ◽  
Author(s):  
Xin Qi ◽  
De C. Zuo ◽  
Zhan Zhang ◽  
Xiao Zong Yang

Importance measures are widely used to characterize the contribution of components to the system performance such as reliability, availability, risk, etc, and thus give great help in identifying system weaknesses and prioritizing system improvement activities. Although much work has been carried out on component importance analysis, most studies only concern the consistent states of components within which components exhibit consistent performance until state changes happen. Unfortunately, field data shows that many transient faults in components may result in severe consequences without causing any state changes, and, this can lead to a misunderstanding of component importance. This paper focuses on the reliability importance analysis in presence of transient faults, and proposes a composite measure for evaluation. A sample series parallel system is analyzed to illustrate the use of this measure.


Author(s):  
Rogelio Hernández Callejas ◽  
A. Liliana Medina-Almazán ◽  
Fco. Javier Merino Caballero ◽  
Salvador Vázquez Belmont

Irradiation embrittlement is a limiting condition for the long-term safety operation of a nuclear Reactor Pressure Vessel (RPV). When a Boiling Water Reactor (BWR) is approaching its initial licensing, in order to operate the reactor for another 20 years and more, it should be demonstrated that the irradiation embrittlement of the reactor vessel materials will be adequately managed by ensuring that the fracture toughness properties are above a certain level of the required safety margin. In this work the Charpy specimens recovered from two surveillance capsules of two BWRs (fluence 3.58×1017 – 9.03×1017 n/cm2) were impact tested at temperatures selected to establish the toughness transition and upper shelf of the irradiated RPV materials. The measured transition temperature shifts (ΔRTNDT) and the Upper Shelf Energy (USE) for the plate and weld materials were compared to the predictions calculated according to Regulatory Guide 1.99 Rev.2. The credibility of surveillance data were analyzed according with the five criteria established in the Regulatory Guide 1.99, Revision 2. The Master Curve (MC) approach and the instrumented impact tests using pre-cracked Charpy specimens were implemented in order to fully validate this techniques that can be used for embrittlement monitoring during life extension periods.


Author(s):  
Stephen M. Parker ◽  
Nathan A. Palm ◽  
Paul R. Stevenson ◽  
Bruce A. Bishop

Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code specifies a 10-year interval between reactor vessel (RV) nozzle weld inspections. The industry has expended significant cost and man-rem exposure performing inspections that have found no service-induced flaws in ASME Section XI Category B-F or B-J RV nozzle welds that do not contain Alloy 82/182. Furthermore, many plants have implemented a 20-year inspection interval for the RV shell-to-shell and shell-to-nozzle welds in accordance with WCAP-16168-NP-A, Revision 2. For many of these plants, continuing to inspect the RV nozzle welds on a 10-year interval presents a significant hardship without a corresponding increase in safety from performing the inspections. This paper will provide a summary of the technical basis and methodology developed by Westinghouse for extending the Section XI inspection interval from the current 10 years to 20 years for Category B-F and B-J RV nozzle-to-safe-end and safe-end-to-pipe welds that are not fabricated with Alloy 82/182 materials. Bounding change-in-failure-frequency values have been calculated for use in plant-specific implementation of the extended inspection interval. Plant-specific pilot studies have been performed and the results show that the change in risk associated with extending the interval from 10 to 20 years after the initial 10-year inservice inspection (ISI) satisfies the guidelines specified in Regulatory Guide 1.174 for an acceptably low change in risk for core damage frequency (CDF) and large early release frequency (LERF). Further, the pilot-plant results show that the effect of the extended inspection interval on the plant’s risk-informed inservice inspection (RI-ISI) program for piping, if any, would also be acceptable.


2020 ◽  
Vol 10 (4) ◽  
pp. 1432
Author(s):  
Yao Li ◽  
Caichao Zhu ◽  
Zi Wang

Due to the high cost of failures of wind turbines, redundancy designs are commonly applied in wind turbines for improving the reliability and availability of systems. For this reason, replacing failed components with other working components of the same type in redundant systems is becoming an attractive option of maintenance strategies towards more resilient systems. To quantitatively evaluate system’s reliability, this paper focuses on the reliability analysis of redundant systems of offshore wind turbines based on swapping existing components. The survival signature-based component swapping method is introduced to describe the new structure-function of the system upon swapping. Furthermore, the reliability model of redundant systems is established using the fault tree and survival signature. Following this, the influences of component swapping on component reliability importance measure (marginal reliability importance and joint reliability importance) without and with considerations of the imprecision of failure rates are explored. Finally, a 5MW offshore wind turbine is presented to show the applicability of the proposed approach for redundant systems, and the results show that the proposed approach can obtain realistic reliability assessment of redundant systems and considering component swapping can significantly improve system reliability.


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