Practical reliability engineering applications to nuclear safety

PRA studies have been successful in providing a quantitative perspective on the important contributions to risk and on the relative impact of potential hardware modifications and procedural changes in reducing public risk. In addition it is the expectation that regulatory agencies will also use this technology to justify a relaxation in requirements that are too strict and the elimination of requirements that are irrelevant or counter to safety. This paper outlines several on-going application oriented R&D efforts that will improve safety in operation, lead to a continuing demonstration that nuclear plants are achieving acceptably low risk, and achieve a higher plant productivity.     

A consistent approach to assess safety criteria for reactivity initiated accidents

In the context of more and more demanding reactor managements, the fuel assembly discharge burn-up increases and raises the question of the current safety criteria relevance. In order to assess new safety criteria for reactivity initiated accidents, the IRSN is developing a consistent and original approach to assess safety. This approach is based on:

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A thorough understanding of the physical mechanisms involved in each phase (PCMI and post-boiling phases) of the RIA, supported by the interpretation of the experimental database. This experimental data is constituted of global test outcomes, such as CABRI or Nuclear Safety Research Reactor (NSRR) experiments, and analytical program outcomes, such as PATRICIA tests, intending to understand some particular physical phenomena;

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The development of computing codes, modelling the physical phenomena. The physical phenomena observed during the tests mentioned above were modelled in the SCANAIR code. SCANAIR is a thermal-mechanical code calculating fuel and clad temperatures and strains during RIA. The CLARIS module is used as a post-calculation tool to evaluate the clad failure risk based on critical flaw depth. These computing codes were validated by global and analytical tests results;

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The development of a methodology. The first step of this methodology is the identification of all the parameters affecting the hydride rim depth. Besides, an envelope curve resulting from burst tests giving the hydride rim depth versus oxidation thickness is defined. After that, the critical flaw depth for a given energy pulse is calculated then compared to the hydride rim depth. This methodology results in an energy or enthalpy limit versus burn-up.

This approach is planned to be followed for each phase of the RIA. An example of application is presented to evaluate a PCMI limit for a zircaloy-4 cladding UO₂ rod at Hot Zero Power.     

A probability approach to estimating the safety of a reactor installation

Krasnaya Zvezda Scientific-Designer Bureau. Central Institute of Atomic Energy, All-Union Research Institute of Atomic Energy. Translated from Atomnaya Énergiya, Vol. 74, No. 1, pp. 38–42, January, 1993.     

Analyses of operational transients in the TRADE experimental program

The TRiga Accelerator Driven Experiment (TRADE) consisted of the coupling of an external proton accelerator to a target in a subcritical core configuration of the TRIGA reactor at the ENEA Casaccia Research Center. In the frame of the safety-related evaluations of the TRIGA reactor a large number of experimental transients induced by reactivity insertions were performed in a critical configuration. Power and fuel temperature experimental transients were analysed by several tools and calculated results showed a reasonable agreement with experiments. Meaningful discrepancies in the simulated temperature trends were found systematically in the 300–600 kW power range. In the paper experimental results in this power range are reported and analysed by means of the TIESTE-MINOSSE code. Investigations lead to suppose the existence of the clad superheating phenomenon at the boiling onset (i.e. boiling delay). On the basis of the studies performed in the past about this subject, the code was improved to take into account this phenomenon. The new simulation results showed a reasonable agreement with the experimental fuel temperature. Results also clearly indicate the local origin of this phenomenon.     

An integrated approach to the reliability of engineering systems

An integrated approach for the reliability of engineering systems consisting of interdependent structural, mechanical, and electrical subsystems is presented. The method is a blend of structural and systems reliability methods and is capable of accounting for physical or functional dependencies between subsystems and their components. Advantages of the method over the existing methods are discussed and some of the difficulties in applying it are pointed out. As an example application, a simple method for the reliability assessment of a general system subjected to seismic hazard is presented.     

Probabilistic economic approach to increasing nuclear plant safety

Editors' Note. During the editing of this article, various and at times conflicting opinions were expressed by members of the editorial board both about the article as a whole and about several ideas expressed in it — for example, on the use of the profit category, the creation of an insurance fund for nuclear plant accidents, etc. In view of this, the editors would be grateful for readers' responses to this article.     

A conceptual approach to the estimation of societal willingness-to-pay for nuclear safety programs

The design, refurbishment and future decommissioning of nuclear reactors are crucially concerned with reducing the risk of radiation exposure that can result in adverse health effects and potential loss of life. To address this concern, large financial investments have been made to ensure safety of operating nuclear power plants worldwide. The efficacy of the expenditures incurred to provide safety must be judged against the safety benefit to be gained from such investments. We have developed an approach that provides a defendable basis for making that judgement.If the costs of risk reduction are disproportionate to the safety benefits derived, then the expenditures are not optimal; in essence the societal resources are being diverted away from other critical areas such as health care, education and social services that also enhance the quality of life. Thus, the allocation of society’s resources devoted to nuclear safety must be continually appraised in light of competing needs, because there is a limit on the resources that any society can devote to extend life.The purpose of the paper is to present a simple and methodical approach to assessing the benefits of nuclear safety programs and regulations. The paper presents the Life-Quality Index (LQI) as a tool for the assessment of risk reduction initiatives that would support the public interest and enhance both safety and the quality of life. The LQI is formulated as a utility function consistent with the principles of rational decision analysis. The LQI is applied to quantify the societal willingness-to-pay (SWTP) for safety measures enacted to reduce of the risk of potential exposures to ionising radiation. The proposed approach provides essential support to help improve the cost–benefit analysis of engineering safety programs and safety regulations.     

A generalised model for PWR for use on safety and reliability studies

In order to carry out a priori studies in the related fields of safety and reliability, it is necessary to have a model on which the proposed site risks can be evaluated and plant and equipment reliability specified. Such a model must have a reasonable probability of ultimate attainment and demonstration.It is obviously necessary that any model proposed shall be capable of meeting all National and International safety standards, legislation and recommendations to which the UK is committed. These would include ICRP, EEC, HSE, NII, MRC and NRPB pronouncements as well as areas covered by specific legislation or regulation.     

A mathematical model and operational program complex for neutron physics processes in the BN-350 reactor

Translated from Atomnaya Énergiya, Vol. 69, No. 6, pp. 368–370, December, 1990     

Some comments on the probabilistic approach to safety analysis

A comparison of the deterministic and the probabilistic approach in structural analysis is outlined. A discussion of some of the criticisms of the probabilistic approach is presented and some of its advantages are indicated.     

On the structural reliability of reactor safety containments

A method for reliability based design of reactor safety containments is suggested, and a brief review of classical reliability analysis is presented. Seismic and climatic load occurrences are modeled by uniform Poisson processes. Extreme value distributions are assumed to represent the seismic, climatic, external and internal pressure load intensities. Reliabilities are calculated for various design loads and load combinations.     

Contributions of structural mechanics to safety and reliability of German nuclear power stations

Traditionally structural mechanics considerations have played a competent role in the design of German nuclear power stations and their fuel. Structural mechanics development and validation programs have set standards of “gooddesign practise” and established the proof of safety against catastrophic failure.     

Systems reliability/structural reliability — Interaction and differences in approach

The question of reliability technology using quantified techniques is considered for systems and structures. Systems reliability analysis has progressed to a viable and proven methodology whereas this has yet to be fully achieved for large scale structures.Structural loading variants over the life-time of the plant are considered to be more difficult to analyse than for systems, even though a relatively crude model may be a necessary starting point. Various reliability characteristics and environmental conditions are considered which enter this problem.The rare event situation is briefly mentioned together with aspects of proof testing and normal and upset loading conditions.     

Tritium sorption on concrete protective coatings and the impact on operational safety

Exposures of concrete and selected coating materials to tritiated atmospheres have shown that tritium sorption on these materials and subsequent desorption are important parameters in defining tritium sources within a tritium-handling facility. Exposure time, tritium concentration and humidity of the air atmosphere affected the amount of tritiated water vapor sorbed. Some of the selected coatings reduced the tritium sorbed to less than 1% of unprotected concrete samples.Work funded by AECL Research and the Canadian Fusion Fuels Technology Project (CFFTP).     

Results of reliability test program on light water reactor piping

The Japan Atomic Energy Research Institute has conducted a piping reliability test program to demonstrate the safety and reliability of light water reactor primary piping. In this program, pipe fatigue test, leak-before-break (LBB) verification test and pipe rupture test were carried out to examine the integrity of piping, to verify the LBB and to demonstrate the effectiveness of protective measures against jet impingement and pipe whip loads under a pipe rupture event.In the pipe fatigue test, a procedure to predict the fatigue crack growth was developed, and the integrity of piping during the plant service life was evaluated. In the LBB verification test, the pipe fracture test and the leak rate test were performed to verify the LBB in the primary piping.In the pipe rupture test, the influence of jet impingement on the target disk and the deformation behavior of whipping pipe and restraint were investigated. Using the test results, the jet impingement behavior and the effectiveness of pipe whip restraint were demonstrated.     

大型集成概率安全分析软件系统的研究与发展

FDS团队在广泛调研和深入分析国际现有概率安全分析软件及其关键技术的基础上,研发了具有自主知识产权的大型概率安全分析软件系统RiskA.该软件提供了系统建模、故障树分析、事件树分析、不确定性分析、可靠性数据管理与分析、敏感性分析和重要度分析等概率安伞分析所需的基本功能.介绍RiskA的设计思想、总体结构、主要功能、技术特点和相关测试与应用等.     

Fuzzy probability on reliability study of nuclear power plant probabilistic safety assessment: A review

Fault tree analysis (FTA) is a graphical model which has been widely used as a deductive tool for nuclear power plant (NPP) probabilistic safety assessment (PSA). The conventional one assumes that basic events of fault trees always have precise failure probabilities or failure rates. However, in real-world applications, this assumption is still arguable. For example, there is a case where an extremely hazardous accident has never happened or occurs infrequently. Therefore, reasonable historical failure data are unavailable or insufficient to be used for statistically estimating the reliability characteristics of their components. To deal with this problem, fuzzy probability approaches have been proposed and implemented. However, those existing approaches still have limitations, such as lack of fuzzy gate representations and incapability to generate probabilities greater than 1.0E-3. Therefore, a review on the current implementations of fuzzy probabilities in the NPP PSA is necessary. This study has categorized two types of fuzzy probability approaches, i.e. fuzzy based FTA and fuzzy hybrid FTA. This study also confirms that the fuzzy based FTA should be used when the uncertainties are the main focus of the FTA. Meanwhile, the fuzzy hybrid FTA should be used when the reliability of basic events of fault trees can only be expressed by qualitative linguistic terms rather than numerical values.     

A quantitative evaluation of reliability of passive systems within probabilistic safety assessment framework for VHTR

This article presents a quantitative evaluation of the reliability of passive systems (RoPS) within the probabilistic safety assessment (PSA) framework for very high temperature reactors (VHTR). VHTRs have unfavorable features in regard to defining a robust failure state. From the viewpoint of PSA, the evaluation of the RoPS as a part of VHTR’s PSA should carefully consider the correct status of a passive system in order to resolve these unfavorable features. This article focuses on the application of multiple states criteria to determine the status of a passive system. Two approaches, i.e., the exceedance probability (EP) model and the stress–strength interference (SSI) model were proposed for the multiple states of the system. A feasibility study has examined the basic features of the proposed approaches by using the reactor cavity cooling system (RCCS) for Korean VHTR. The primary condition for the usefulness of the proposed approaches is that sufficient information should be provided in order to determine the system strengths for the multiple states. With regard to the engineering practice, the EP approach for the multiple states can provide a practical solution concerning the evaluation of the RoPS for VHTR’s PSA.     

A method for quantifying the reliability of instrumentation for engineered safety features to improve accident sequence recognition in reactor control rooms

This work adapts fault trees from plant-specific probabilistic risk analyses (PRAs) to quantitatively evaluate the reliability of the instrumentation for engineered safety features (ESFs). The purpose is to help improve reactor operator recognition and identification of potential accident sequences. The PRA system fault trees provide a framework for assessing the plant indicators so that the plant conditions are made more readily apparent to plant personnel through the conversion of system fault trees to alarm trees. In the alarm tree, possible states of each instrumented alarm are identified as “true” or “false”. In addition, a “warning” status is also defined and integrated into the alarm analysis routine. The impact of this additional status condition on the Boolean laws used to evaluate the alarm trees is examined. An application is described for a BWR high pressure coolant injection system (HPCI) that would be utilized during many severe reactor accidents.     

Increasing the operational reliability of the collectors of PGV-1000,-1000M steam generators

The possibilities for lowering the tensile stresses acting in the zone of weld seam No. 111 are studied. Stresses above the yield point can arise in this zone during operation because of the presence of stress concentrators and deep ribs and cuts, formed when the steam generator is manufactured, on the inner surface of pockets in the collectors. The tensile stresses can be decreased without decreasing the pressure of the medium in the second loop by using mechanical compressing devices. A variant for producing a compressive stress on branch pipe Du 1200 using a collapsible ring is examined. The reliability of the unit attaching the coolant collector to the branch pipes of PGV-1000 in the zone of weld seam No. 111 can be increased by effectively removing deposits from the pockets. It is suggested that a mechanical compressing device be used to decrease the stresses on the inner surface of a pocket in the collector and to decrease the cycling damage in the region of weld seam No. 111. __________ Translated from Atomnaya énergiya, Vol. 108, No. 1, pp. 9–12, January, 2008.