A numerical procedure for computing the fragility of NPP components under random seismic excitation

A numerical procedure is proposed in the paper for computing seismic fragility functions for equipment components in Nuclear Power Plants. The procedure is based on the hypothesis, which is typical when seismic excitation of components is addressed, of linear behaviour of the building. Given the large size of the FE element models adopted for the building, which makes direct Monte Carlo simulation impossible, the response surface methodology is used to model the influence of the random variables on the dynamic response. To account for stochastic loading the latter is estimated by means of a simulation procedure. Once the response surfaces defining the statistical properties of the response are available, the Monte Carlo method is used to compute the failure probability. A procedure for refining the RS estimation is also proposed, based on the evaluation of risk for a prototype site.A validation example is given, regarding the simplified modelling of a reactor building resting on a base-isolation system; results obtained by plain Monte Carlo analysis are compared to those computed via the proposed procedure The latter is finally applied to a real life case, taken from the preliminary design of the auxiliary building within the IRIS international project.     

Uncertainty analysis of system fragility for seismic safety evaluation of NPP

In this study, a Seismic Probabilistic Safety Assessment (SPSA) methodology considering the uncertainty of fragilities was studied. A system fragility curve is estimated by combining component fragilities expressed by two variance sources, inherent randomness and modeling uncertainty. The sampling based methods, Monte Carlo Simulation (MCS) and Latin Hypercube Sampling (LHS), were used to quantify the uncertainties of the system fragility. The SPSA of an existing nuclear power plant (NPP) was performed to compare the two uncertainty analysis methods. Convergence of the uncertainty analysis for the system fragility was estimated by calculating High Confidence Low Probability of Failure (HCLPF) capacity. Alternate HCLPF capacity by composite standard deviation was also verified. The annual failure frequency of the NPP was estimated and the result was discussed with that from the other researches. As a result, the criteria of the uncertainty analysis and its effect was investigated.     

Numerical computation of fragility curves for NPP equipment

The seismic probabilistic risk assessment (PRA) methodology is a popular approach for evaluating the risk of failure of engineering structures due to earthquake. In this framework, fragility curves express the conditional probability of failure of a structure or component for a given seismic input motion parameter A, such as peak ground acceleration (PGA) or spectral acceleration. The failure probability due to a seismic event is obtained by convolution of fragility curves with seismic hazard curves. In general, a log-normal model is used in order to estimate fragilities. In nuclear engineering practice, these fragilities are determined using safety factors with respect to design earthquake. This approach allows to determine fragility curves based on design study but largely draws on expert judgement and simplifying assumptions. When a more realistic assessment of seismic fragility is needed, simulation-based statistical estimation of fragility curves is more appropriate. In this paper, we will discuss statistical estimation of parameters of fragility curves and present results obtained for a reactor coolant system of nuclear power plant. We have performed non-linear dynamic response analyses using artificially generated strong motion time histories. Uncertainties due to seismic loads as well as model uncertainties are taken into account and propagated using Monte Carlo simulation.     

地震概率安全分析中地震易损度不确定性分布的研究

核电厂地震概率安全分析(PSA)中,构筑物和设备的地震易损度是在给定地面运动强度条件下的条件失效概率。地震易损度的不确定性分布较为复杂,在地震PSA定量化过程中难于处理。本文针对地震易损度的数学模型进行研究,采用数值方法求解地震易损度的均值和方差。在均值和方差相等的条件下,以几种常见的不确定性分布类型近似地震易损度的不确定性分布。通过比较可以看出,Beta分布可以较为准确地描述地震易损度的不确定性分布。     

The seismic reliability of VVER-1000 NPP prestressed containment building

The failure probability assessment of the containment building is an essential feature of the Level 2 PSA studies of nuclear power plants. The primary purpose of this paper is to demonstrate the methodology of evaluating containment seismic induced probability of failure without containment pressurization. The Loviisa, Finland site is one of the most seismically stable in the world and the numerically evaluated seismic induced failure probabilities are not representative for other sites. In addition, the containment concept described in this paper is not the typical Russian design which uses helical tendons in the cylindrical part of the structure and has a ring girder at the spring line of the structure. So the conclusions reached are applicable only to the containment configuration described in the paper. The geometry of the containment was determined by its preliminary design. The seismic hazard of the plant site was assessed during Level 1 PSA of the Loviisa plant. The initial information for seismic fragility analysis of the containment is the seismic response of the structure. The structural model for response analysis was the stick model. The stress analysis of the containment was carried out using the shell element model. The fragility evaluation of the containment was performed with the PROSAN-program. The structure was modeled as a parallel system consisting of the most heavily stressed elements. The resulting fragility curve gives the conditional probability of failure as a function of peak ground acceleration. The seismic hazard and the fragility were convolved to obtain the annual nonexceedance probability distribution for the collapse frequency of the structure.     

Improved response factor methods for seismic fragility of reactor building

A seismic probabilistic risk assessment (PRA) method has been applied to evaluate the safety of nuclear reactor buildings during earthquakes. Improvement was made to two methods (based on linear response and based on non-linear response) of fragility analysis in seismic PRA. The conventional method, which is based on linear response, considers increases of seismic capacity implicitly, using the non-linear behaviour of the structure. We described how to evaluate the capacity increase factor for the linear response method. Secondly, we proposed a method based on the non-linear response and a stratified two-point estimation method which can efficiently evaluate the variability of non-linear responses. We applied the two method to a PWR-type nuclear reactor building and ascertained that these method are useful and effective.     

Response of a NPP reactor building under seismic action with regard to different soil properties

The object of this investigation is the response of a reactor building on seismic action with systematic variation of the soil stiffness. A thin-walled orthotropic containment shell on varying heavy and rigid foundations is regarded as calculation model. The soil stiffness is simulated by means of spring elements for horizontal translation and for rocking motions of the building. By the response spectra method the loads of the containment shell are calculated for a horizontal seismic excitation. The investigation is aimed at determining the influence of differentiated soil stiffnesses on the containment action effects and at recognizing the causes for the occurring effects.The results are thoroughly represented by selected quantities of the building's response, the effects from the soil-structure interaction are discussed and the causes of the effects clearly explained. A possibility is provided for determining critical soil stiffnesses which cause a significant intensification effect.The results of the investigations show that both the soil stiffness and structural configuration of the reactor building, particularly in case of the substructure being heavy and rigid, exert a decisive influence on the loading of the superstructure.     

Evaluation of seismic fragility of structures—a case study

The present paper attempts to evaluate the seismic fragility for a typical elevated water-retaining structure. The structure is analysed for two cases: (i) empty tank; and (ii) tank filled with water. The various parameters that could affect the seismic structural response include material strength of concrete and reinforcing steel, effective prestress available in the tank, ductility ratio and structural damping available within the structure, normalised ground motion response spectral shape, foundation and surrounding soil parameters and the total height of water available in the tank. Based on this case study, the seismic fragility of the structure is developed. The results are presented as families of conditional probability curves plotted against peak ground acceleration (PGA) at two critical locations. The procedure adopted, incorporates the various randomness and uncertainty associated with the parameters under consideration.     

An advanced seismic analysis of an NPP powerful turbogenerator on an isolation pedestal

This paper presents a detailed seismic analysis of a powerful high-speed Russian turbine within a Nuclear Power Plant. Although dozens of these turbines have worked reliably since the 1970s worldwide, until the last decade, only simplified structural analyses were available due to the turbines’ complicated overall structure and internal design. The current analysis considers the detailed geometry of the turbine itself and the vibration and seismic isolation system within the turbine's pedestal and the full range of operational, accident and seismic loads like high pressure, outside loads induced by pipelines and so on.To solve the problem of the turbine seismic qualification, the following steps have been taken. The first step was to create detailed finite element models of the turbine's high and low pressure parts and rotor system with bearings. Using such models, corresponding simplified models were developed to be included into the coupled model of the system: “Building–Vibroisolation Pedestal–Turbine” (BVT). The second step was the analysis of that coupled system. Soil–structure interaction was considered using actual soil conditions. Three components of time history acceleration were used to define seismic excitation. As the result of BVT system analysis, a full picture of time history displacements and loads was determined. At the same time, a problem of rotor gaps was solved. In the final step, determined loads were applied to the detailed models of the turbine's parts for seismic qualification of the whole structure.     

Reanalysis and evaluation of the seismic capacity of the CANDU 700 MW NPP CERNAVODA

The dynamic response of structures of the CANDU 700 MW NPP due to seismic loadings was conventionally analyzed in the time domain using modal substructure procedures. The frequency-independent parameters were tuned to the main frequencies of the soil-structure system. This is a common procedure widely used in the preliminary design of power plant structures and provides conservative results. However, parallel to the rapid progress being made in upgrading the capability of data processing systems, methods and software tools have become available which work in the frequency domain using complex (soil-structure) mathematical models or models in which the soil is represented by frequency-dependent impedances. In order to demonstrate the reserves existing in the design of the CANDU 700 reactor building, frequency-domain calculations were additionally prepared. The analyses were based on appropriate mathematical 3D-models of the coupled vibrating structures of the reactor building and as the soil represented by frequency-dependent impedances. The results obtained by using the time and frequency domain methods were compared and the safety margins of the CANDU design discussed.     

Methods of non-linear stochastic dynamics for the assessment of structural fragility

Some of the external events which can significantly contribute to the overall risk of a nuclear power plant, give rise to a dynamic excitation of the structural components which form the plant. The computation of the risk associated with these external events requires an investigation of the behaviour of the structural components beyond the elastic limit. The stochastic nature of the excitation, then, leads one to deal with a non-linear stochastic dynamic problem.

No general method of solution exists for such a problem when large structural systems are considered, although classical methods of propagating uncertainty have been successfully employed. This paper investigates the possibility of formulating an approach founded on a suitable equivalent linearization technique. In particular the authors make operative a new method of fragility analysis to be applied directly to the linearized system.

The numerical example considers a framed structural component: its aim is to show the degree of accuracy that can be reached by the approach formulated in the paper.     

Estimates on seismic amplification factors of reactor components

Employing an averaging technique we obtain estimates on seismic amplification factors for different components in nuclear reactors.     

Methods of testing HTGR fuel-pin components

Luch Scientific Production Association. Translated from Atomnaya Énergiya, Vol. 73, No. 2, pp. 102-109, August, 1991.     

Improved approach for obtaining rotational components of seismic motion

The rotational component of seismic strong-motion is attracting attention since it is becoming evident that it may contribute considerably to the overall response of structures to earthquake motions. This paper presents an improved method for calculating the time histories of torsional and rocking components of ground motion corresponding to a set of three recorded orthogonal translational components. The mathematical model is based on a detailed representation of soil impedance and contributions of body waves. The dependence of the angle of wave incidence on the frequency of wave is properly given in the calculation of rotational components with consideration of critical incident angles. Numerical results of the torsion and rocking obtained from a set of three recorded translational components are also presented.     

用于核电厂模拟器的点堆动力学方程的改进解法

本文叙述了点堆动力学方程的两种改进解法,即反应性限制法和综合法。这两种方法分别比赫米特(Hermite)插值多项式法和国外全尺寸模拟器上所采用的变量隐含积分(Variablelmplicit Integration)法具有更高的精确度和更短的计算时间。     

Applications of multi-directional seismic inputs in the design of components

A method is presented to convert the existing component qualification level (acceleration in g's) for two-directional earthquake input to an equivalent qualification level for three-directional earthquake input, and vice-versa. This exact and conservative method is applicable to all simple equipment which uses static acceleration as the basis for design, such as auxiliary pumps, valves, tanks, heat exchangers, filters, and demineralizers.     

宁德核电站控制区智能配发系统

传统出入核电站控制区是采用“用证件换钥匙取剂量计”的人工流程,该流程需要人工24小时进行配发,工作量大、效率低。为了优化进出控制区流程,提高效率,实现无人化管理,宁德核电于2018年上线控制区智能配发系统。本文介绍该系统的组成及功能,结合现场调试经验,提出了系统存在的问题,并对后续改进工作提出了建议。