Risk–benefit analyses of SG tube maintenance based on probabilistic fracture mechanics

As an application of probabilistic fracture mechanics (PFM), a risk–benefit analysis was performed for the purpose of optimizing maintenance activities of steam generator (SG) tubes used in pressurized water reactors (PWRs). The probabilities of the SG tube leakage and rupture are defined as risks in this study. A model was made modifying pc-PRAISE (Piping Reliability Analysis Including Seismic Events) to evaluate the risks during 60 year operations due to stress corrosion cracking (SCC) of the tubes under various maintenance strategies for SG tubes.In the risk analysis, parameters such as inspection accuracy, inspection interval, sampling inspection and crack propagation law were selected for sensitivity analysis. Based on the risk analysis, a risk–benefit analysis was conducted when implementing two maintenance strategies taking both costs and revenues for 60 year operations into account. In the risk–benefit analysis, the expected cost of leakage or rupture was calculated by multiplying ‘probability of leakage or rupture’ by ‘expected loss of leakage or rupture accident’. To justify whether it is worthwhile implementing the maintenance strategies or not, the net present value (NPV) was calculated as an index, which is one of the most fundamental financial indices for decision-making based on the discounted cash flow (DCF) method.The results demonstrated that in the risk analysis, the risks are influenced significantly by the crack propagation law, accuracy of inspection and sampling inspection. In the risk–benefit analysis, it was suggested that investment to improve inspection accuracy would reduce the total costs of 60 year operations significantly and increase the NPV.Although the analysis was mainly conducted for SG tubes made of Inconel 600 mill anneal (MA) material, the analysis was also carried out for Inconel 690 thermal treatment (TT) material, making assumptions on its crack initiation and crack propagation law. In addition, the effect of introducing maintenance criteria, namely, operation with a crack justified by certain criteria, on NPV was evaluated.     

The price of risk reduction: Optimization of test and maintenance integrating risk and cost

Testing and maintenance (T&M) improve the reliability of safety systems and components in nuclear power plants, which is of special importance for standby systems. Early optimizations of single component test intervals were based on minimizing the risk, e.g. the time-average unavailability, without cost considerations. However, the appropriate development of T&M strategy depends not only on the T&M intervals but also on the resources (human and material) available to implement such strategies. Since these testing and maintenance activities are associated with substantial cost, they present an important domain, where risk reduction and costs can be balanced.The objective of this paper focuses on assessing how costs and component ageing may affect the T&M optimization in terms of minimal system risk. The costs are expressed as a function of the selected risk measure. The time-averaged function of the selected risk measure is obtained from probabilistic safety assessment, i.e. the fault tree analysis at the system level, extended with inclusion of time parameters related to T&M activities. Additionally, component ageing is taken into account while developing the system reliability model presented in this paper. The testing strategy is also addressed. Sequential and staggered testing strategies are compared. The developed approach is applied on a standard test system and the obtained results are presented. The results show that the risk-informed surveillance requirements differ from existing ones in technical specifications, which are deterministically based. The presented approach achieves a significant reduction in system unavailability accompanied with relatively small changes in total T&M costs.     

以可靠性为中心的维修在核电站维修优化中的应用与创新

以可靠性为中心的维修(简称RCM)是目前国外广泛采用的维修优化方法。本文介绍了RCM分析方法以及该方法在大亚湾核电站的应用情况,并介绍了对RCM决断逻辑进行优化和改进的情况,最后通过实例说明RCM改进后所取得的效果。     

Probability-based assessment and maintenance of welded joints damaged by fatigue

This paper presents a probabilistic reliability assessment procedure for steel components damaged by fatigue. The study combines the structural reliability theory with a maintenance strategy. The fatigue assessment model is based on a modelisation of the fatigue phenomenon issued from the principles of fracture mechanics theory. The safety margin includes the crack growth propagation and allows to treat fatigue damage in a general manner. Damaging cycles and non damaging cycles are distinguished. The sensitivity study of the different parameters shows that some variables can be taken as deterministic. Applications are made on a welded joint ‘stiffener/bottom-plate' of a typical steel bridge. The model is then used for taking into account inspection results. Non destructive inspection (NDI) techniques are also used for updating failure probabilities. The results show their ability to be inserted in a maintenance strategy for optimizing the next inspection time. This has led to define cost functions related to the total maintenance cost; this cost is then minimized for determining the optimal next inspection time. An example of welded joint cracked by fatigue highlights the different concepts. The approach presented in the paper is not only restrained to fatigue problems, but can be applied to a wide variety of degrading phenomena.     

以可靠性为中心的核电厂厂用水系统维修分析

通过对AP1000核电厂厂用水系统进行故障模式及影响分析(FMEA)以及逻辑决断分析(LTA),深入了解该系统工艺及相关设备的功能故障、故障模式和影响,进而建立优化的维修决策.与现行的维修策略相比,经由可靠性为中心的维修( RCM)优化所得的维修策略对显性故障更多的是选择状态监测/定期维护,而对隐蔽性故障则采用定期试验...     

Using Risk-Informed Asset Management for Feedwater System Preventative Maintenance Optimization

The initial development of a South Texas Project Nuclear Operating Company process for supporting preventative maintenance optimization by applying the Balance-Of-Plant model and Risk-Informed Asset Management alpha-level software applications is presented. Preventative maintenance activities are evaluated in the South Texas Project Risk- Informed Asset Management software while the plant maintains or improves upon high levels of nuclear safety. In the Balance-Of-Plant availability application, the level of detail in the feedwater system is enhanced to support plant decision-making at the component failure mode and human error mode level of indenture by elaborating on the current model at the super-component level of indenture. The enhanced model and modeling techniques are presented. Results of case studies in feedwater system preventative maintenance optimization using plant-specific data are also presented.     

Some thoughts on allocation of reliability

The technical feasibility of allocating reliability to reactor systems, subsystems, components, and structures is discussed in this paper. The basic premise for this analysis is that a set of objective functions or safety variables has been defined on a global basis for a class of nuclear power plants. The decision variables, which represent the system, subsystem, component, and structural reliabilities are related to the global objective functions by a risk model obtained from an existing plant-specific probabilistic risk assessment (PRA). A multiobjective optimization technique is employed to obtain the set of decision variables which optimize (minimize) all of the objective functions. A cost function is introduced (and incorporated in the optimization scheme) which measures the cost of increasing reliability. Illustrative calculations were performed for a boiling water reactor with an existing PRA.     

Prospects for Using Mixed Oxide Fuel in Fast Sodium-Cooled Reactors

When designing new fast reactors, it is desirable to increase as much as possible the breeding occurring in the core in order to ensure the minimum excess reactivity for burnup on the one hand and a closed fuel cycle without replenishment with external plutonium and without separating plutonium from uranium during chemical reprocessing of irradiated fuel on the other. The latter requirement greatly decreases the risk of plutonium proliferation in such a fuel cycle. This requires a core breeding ratio 1.05–1.08. Such values can be achieved by using technologically perfected and tested oxide fuel with its volume fraction in the core increased to 55–60%. The results of computational-theoretical studies on the selection and optimization of cores with high fuel fractions for BN-1600 and BN-800 reactors are presented in this article. It is shown that such cores can be built in principle.     

Techniques for CAD reconstruction of `as-built'' environments and application to preparing for dismantling of plants

Electricité de France is using CAD-generated numeric geometrical models to simulate maintenance operations and enable optimizing maintenance procedures. These models are also used to program the machines or robots for certain servicing procedures. They are used in the operator interfaces for robot control, and provide the operator with virtual cameras or enable generating specific information (such as virtual force feedback). Even more recently, CAD models have been integrated in what is known as ‘virtual reality' software, giving the operators a sensation of ‘immersion' in a virtual universe. Depending on the need and on the type of results expected from the simulations, one needs more or less precise models of the environment in which work will be performed. EDF is using several techniques to get ‘as-built' models of the environments. This article describes the SOISIC system, which is a 3D laser sensor widely used for environment data acquisition, associated with 3Dipsos software, for CAD model reconstruction. These techniques, and the applications subsequently developed for maintenance applications, can be used in preparing and carrying out dismantling operations: ‘as-built' CAD modeling of the installation can be used in the preparatory phase, providing plans, simulating the various steps, calculating waste volumes, helping in optimization of waste management, etc. These models can also be used during the actual dismantling process, to program the machines or robots used, or in the robot or machine supervisory system. Some of the presented techniques have been used in a room in the Brennilis plant, which is currently being dismantled.     

Errors in evaluated neutron cross sections

The underestimation of the errors in the recommended neutron cross sections calculated on the basis of an analysis of measurement results using strict statistical methods is discussed. The basic reasons for the underestimation of the errors in the cross sections are described. A statistical model is proposed — a constant bias model — for taking into account effectively the component of the measurement error that is unknown to the experimentor. The results of a statistical analysis of the measurements of the ratio of the fission cross sections of ²³⁸U and ²³⁵U on the basis of the constant bias model are presented. It is noted that the experimental errors are much higher (on the average by a factor of 2) than the declared errors. The consequences of using the generally accepted method for correcting the errors in the evaluated cross sections are examined for an exactly solvable model example. It is shown that such a procedure does reconstruct the real errors in the evaluated cross sections.__________Translated from Atomnaya Énergiya, Vol. 98, No. 1, pp. 3–12, January, 2005.     

Development of nonequilibrium pressurizer model with noncondensable gas

To investigate thermal–hydraulic characteristics of a steam–gas pressurizer in the integral type reactor, the steam–gas pressurizer model based on the two-region nonequilibrium concept was developed and introduced into RETRAN-3D/INT code. The model includes an explicit solution method for the one-dimensional governing equations and the equation of the state solution method to determine the thermal–hydraulic state of the steam–gas pressurizer volume. In addition, the wall condensation model based on the diffusion layer modeling was included to consider the effect of the noncondensable gas. The developed model was verified with the results from the pressurizer insurge experiment conducted at Massachusetts Institute of Technology. From the verification results, it was concluded that the developed steam–gas pressurizer model can sufficiently predict the pressurizer transient and it can be used as a component model of the one-dimensional system code based on the homogeneous equilibrium model.     

Evaluation of the parameters of an experimental tokamak fusion-reactor and the influence of the indeterminacy of the plasmaphysical data

Conclusion Analysis using a simplified technicoeconomic model of a tokamak reactor and the procedures of graphic processing of the results allow the regions of the optimum parameters of an experimental minimum-cost reactor with various limitations to be determined. When refined and ignition scalings are taken into account simultaneously and an appropriate optimization is made we get the parameters of the experimental reactor, which differ slightly from the basic variant (in particular, a lower neutron load on the first wall). The reactor cost in this case should increase by 15–25% in comparison with the basic variant and remains virtually constant when the most favorable ignition scaling (ASDEX) is realized.Translated from Atomnaya Énergiya, Vol. 62, No. 1, pp. 22–28, January, 1987.     

A study on the free drop impact of a cask using commercial FEA codes

The package used to transport radioactive materials, which is called a cask, must be designed to keep its contents safe under normal and hypothetical accident conditions. The design requirements of the cask are verified by test or finite element analysis (FEA). Comparing evaluation procedures for the safety of a new cask, the cost of FEA is generally much less than that test. Therefore, FEA is mainly used to verify safety of a cask under the considered conditions. However, one commercial FEA code may show different results from another FEA code for the same problem due to the modeler's several assumptions for simplifying actual states into the FE model and due to modeling technique. Materials of the components of a cask display elastic–plastic or elastic–perfectly plastic behavior under the considered conditions in which large deformation, impact and contact mechanism are included. The behavior is simulated with difficulty and may have different results depending on FEA codes. In this paper, finite element analysis is carried out for the 9-m free drop and the puncture condition under the hypothetical accident condition by using LS-DYNA3D and ABAQUS/Explicit. Energy and effective stress on each component are presented and compared between the two FEA codes, where the effective stress designates the maximum von Mises stress on inner and outer shells.     

Nuclear-Hydrogen Power

Methods for obtaining hydrogen and using hydrogen in power engineering, transportation, and industry, and methods for handling hydrogen (storage and safety) are examined.The concept of nuclear-hydrogen power – using the energy generated by nuclear reactors to produce hydrogen and using this hydrogen in power engineering and industry – is presented. The development of nuclear-hydrogen power will contribute to global energy security and decrease the demand for fuels which affect climate change on our planet.The technologies needed for nuclear-hydrogen power to become a reality – high-temperature nuclear reactors, apparatus for the efficient production of hydrogen from water, hydrogen fuel cells, chemothermal converters, and hydrogen storage and shipment technology – are analyzed.     

双参数威布尔分布在核电站数据处理中的应用

核电站设备可靠性数据的处理是电站进行以可靠性为中心的维修（RCM）和寿期管理（LCM）的基础。在核电站失效数据的实际处理过程中,常会面临失效样本少、维修导致数据分布不独立等问题。为解决上述问题,本文提出以双参数威布尔分布作为寿命模型、采用贝叶斯方法来处理小样本失效数据的方法,并结合核电站运行数据进行验证。结果表明,本方法在处理样本较少以及存在维修老化问题时,具有更好的适用性和准确度。     

A knowledge-based multi-dimension discrete common cause failure model

Common cause failure (CCF) is analyzed as a manifestation of the probabilistic characteristics of component failure rate/probability stemming from stochastic environment load. CCF revolved concepts, such as ‘root cause’ and ‘coupling mechanism’ are interpreted mathematically from the viewpoint of random environment load bringing about failure dependency. Opinions about ‘inherent CCF’ and ‘additional CCF’, ‘absolute CCF’ and ‘relative CCF’ are presented and discussed. An easy-to-use CCF model is developed through multi-dimension environment load-component strength interference analysis and knowledge based parameter discretization. Owing to its strict statistical foundation, such a model has the ability of estimating component failure rate/probability and common cause failure rates/probabilities consistently, dealing with low redundancy system CCF and high redundancy system CCF uniformly, and predicting high multiplicity failure rate/probability based on low multiplicity failure data satisfactorily.     

Basic factors to forecast maintenance cost and failure processes for nuclear power plants

Two types of maintenance interventions are usually administered at nuclear power plants: planned and corrective. The cost incurred includes the labor (manpower) cost, cost for new parts, or emergency order of expensive items. At the plant management level there is a budgeted amount of money to be spent every year for such operations. It is very important to have a good forecast for this cost since unexpected events can trigger it to a very high level. In this research we present a statistical factor model to forecast the maintenance cost for the incoming month. One of the factors is the expected number of unplanned (due to failure) maintenance interventions. We introduce a Bayesian model for the failure rate of the equipment, which is input to the cost forecasting model. The importance of equipment reliability and prediction in the commercial nuclear power plant is presented along with applicable governmental and industry organization requirements. A detailed statistical analysis is performed on a set of maintenance cost and failure data gathered at the South Texas Project Nuclear Operating Company (STPNOC) in Bay City, Texas, USA.     

Homogenisation method for the modal analysis of a nuclear reactor with internal structures modelling and fluid–structure-interaction coupling

A homogenisation method is presented and validated in order to perform the dynamic analysis of a nuclear pressure vessel with a “reduced” numerical model accounting for inertial fluid–structure coupling and describing the geometrical details of the internal structures, periodically embedded within the nuclear reactor. Homogenisation techniques have been widely used in nuclear engineering to model confinement effects in reactor cores or tubes bundles. Application of such techniques to rector internals is investigated in the present paper. The theory bases of the method are first recalled. Adaptation of the homogenisation approach to the case of rector internals is then exposed: it is shown that in such case, confinement effects can de modelled by a suitable modification of classical fluid–structure symmetric formulation. The method is then validated by comparison of 3D and 2D calculations. In the latter, a “reduced” model with homogenised fluid is used, whereas in the former, a full finite element model of the nuclear pressure vessel with internal structures is elaborated. The homogenisation approach is proved to be efficient from the numerical point of view and accurate from the physical point of view. Confinement effects in the industrial case can then be highlighted.     

Molten corium oxidation model

In order to model oxidation of Zr–O and U–Zr–O melts, post-test appearance of refrozen oxidised melts in the CORA and QUENCH bundle tests performed at the Research Centre Karlsruhe (FZK) are analysed. Furthermore, data from new separate effect tests on ZrO₂ crucible dissolution by molten Zry, specially designed for investigation of long-term behaviour during the melt oxidation stage, are taken into consideration. On this base, a new model on oxidation of molten Zr–O and U–Zr–O mixtures in steam was developed, which allows interpretation of melt oxidation and hydrogen production observed in various bundle tests. The complete formulation of the analytical model, development of the numerical model and its validation against the crucible tests are presented.     

Safety management in NPPs using an evolutionary algorithm technique

The general goal of safety management in Nuclear Power Plants (NPPs) is to make requirements and activities more risk effective and less costly. The technical specification and maintenance (TS&M) activities in a plant are associated with controlling risk or with satisfying requirements, and are candidates to be evaluated for their resource effectiveness in risk-informed applications. Accordingly, the risk-based analysis of technical specification (RBTS) is being considered in evaluating current TS. The multi-objective optimization of the TS&M requirements of a NPP based on risk and cost, gives the pareto-optimal solutions, from which the utility can pick its decision variables suiting its interest. In this paper, a multi-objective evolutionary algorithm technique has been used to make a trade-off between risk and cost both at the system level and at the plant level for loss of coolant accident (LOCA) and main steam line break (MSLB) as initiating events.