Response of proposed passive safety injection system for an intermediate size break LOCA on CHASNUPP-1

A passive safety injection system (PSIS) is proposed for Chashma nuclear power plant-1 (CHASNUPP-1) type nuclear power plants, for the simplification of their safety systems. This system is based upon passive components and is proposed in place of the existing safety injection system, for safety enhancement. The functionality of the proposed system is analyzed using reactor simulation. For this purpose an intermediate size break LOCA is simulated using the simulation software APROS. For this transient, different thermal-hydraulic parameters of the proposed and other safety related systems are presented and discussed. The results obtained show that the proposed system works properly by performing its role in the transient, leading to cold shutdown conditions.     

IR-360 nuclear power plant safety functions and component classification

The IR-360 nuclear power plant as a 2-loop PWR of 360 MWe power generation capacity is under design in MASNA Company. For design of the IR-360 structures, systems and components (SSCs), the codes and standards and their design requirements must be determined. It is a prerequisite to classify the IR-360 safety functions and safety grade of structures, systems and components correctly for selecting and adopting the suitable design codes and standards. This paper refers to the IAEA nuclear safety codes and standards as well as USNRC standard system to determine the IR-360 safety functions and to formulate the principles of the IR-360 component classification in accordance with the safety philosophy and feature of the IR-360. By implementation of defined classification procedures for the IR-360 SSCs, the appropriate design codes and standards are specified. The requirements of specific codes and standards are used in design process of IR-360 SSCs by design engineers of MASNA Company. In this paper, individual determination of the IR-360 safety functions and definition of the classification procedures and roles are presented. Implementation of this work which is described with example ensures the safety and reliability of the IR-360 nuclear power plant.     

备用系统可靠性动态问题方法探究

随着核电安全性日益受到世界公众关注,用概率论的方法对核电厂系统进行可靠性分析也越来越显示出其重要作用.在进行系统可靠性分析时,故障树方法是国际上公认的一种简单、有效、经典的方法.但随着所要分析的工程系统日趋庞大,系统工作过程日趋复杂,对含有动态过程的系统可靠性分析成为故障树方法中的棘手问题.本文以可靠性分析中的典型动态问题--备用系统的可靠性为研究对象,运用蒙特卡罗方法编程探究如何运用故障树来求解可靠性动态问题,并在进行程序验证后对一个备用系统实例进行计算,给出计算结果.     

Seismic safety of nuclear power plants in eastern Europe

This paper summarizes the work performed by the International Atomic Energy Agency in the areas of safety review and applied research in support of programmes for the assessment and enhancement of seismic safety in Eastern Europe and in particular, WWER type nuclear power plants during the past seven years. Three major topics are discussed; engineering safety review services in relation to external events, technical guidelines for the assessment and upgrading of WWER type nuclear power plants, and the Coordinated Research Programme on "Benchmark study for the seismic analysis and testing of WWER type nuclear power plants". These topics are summarized in a way to provide an overview of the past and present safety situation in selected WWER type plants which are all located in Eastern European countries. The main conclusion of this paper is that even though there is now a thorough understanding of the seismic safety issues in these operating nuclear power plants, the implementation of seismic upgrades to structures, systems and components are lagging behind, particularly for those cases in which re-evaluation indicated the necessity to strengthen the safety related structures or install new safety systems.     

非能动系统可靠性分析方法在AP1000非能动ERVC系统中的应用

由于核电厂安全水平要求的逐渐提高,越来越多的非能动系统被用于先进反应堆堆型中,但对这些非能动系统可靠性评价的工作还处于初级阶段。本文根据非能动系统可靠性评价流程,通过RELAP5热工水力学程序模拟非能动系统物理过程,对AP1000反应堆压力容器外部冷却（ERVC）系统进行了可靠性评价。通过计算得到了压力容器下封头温度等参数的累积密度分布曲线,根据不同的成功准则即可获得AP1000 ERVC系统的可靠性。该非能动系统可靠性评价结果可用于核电厂PSA模型中,以更好地指导核电厂设计及提高核电厂的安全性。     

Seismic reliability of electrical power transmission systems

The reliability of electric power transmission systems is important for the probabilistic safety assessment of nuclear power plants under a given earthquake loading as it relates to the loss of off site power to the nuclear power plants. Here, a comprehensive model to evaluate the seismic reliability of electric power transmission systems is presented. The model provides probabilistic assessments of structural damage and abnormal power flow that can lead to power interruption in a transmission system under a given earthquake. With the proposed methodology seismic capacities of electrical. equipment are determined on the basis of available test data and simple modeling from which fragility functions of specific substations are developed. Earthquake ground motions are defined as stochastic processes. Probabilities of network disconnectivity and abnormal power flow are assessed through Monte Carlo simulations. The proposed model is applied to the electric power network in San Francisco and vicinity under the 1989 Loma Prieta earthquake, and the probabilities of power interruption are contrasted with the actual power failures observed during that earthquake.     

Introduction to the proceedings of the international seminar on structural reliability of mechanical components and subassemblies of Nuclear Power Plants

Although the integrity and safety of many mechanical components and subassemblies of nuclear power plants are demonstrated by the appropriate design codes and supplementary requirements, such procedures seldom provide guidance as to “how safe” the structures are. By combining the technologies of solid mechanics and probabilistic structural reliability methods, engineers are finding many and varied opportunities to demonstrate margins in terms of probabilities of failure.With reference to the large mechanical system components typical of nuclear power plants, reliability assessments are receiving more emphasis in recent years as evidenced by the increased attention to such reliability-related techniques as failure mode and effects analyses, fault tree analyses, common cause failure analyses, and single point failure analyses. These techniques are ordinarily applied at the outset in a qualitative manner, tracing the casual sequences of potential component failure. The results of these analyses serve as the foundation for more sophisticated probabilistic structural reliability analyses which have the objective of calculating the probability of failure (that is, unreliability) of the system or component in question.     

核电站安全分级对DCS系统设计影响分析

根据核电站设计总体要求,特别是对仪控系统可用性和可靠性的要求,通过分析核电站中系统、设备及其功能的安全分级,解析现代数字仪化控系统( DCS)的技术特点.结合实际在建核电站中不同DCS总体技术方案设计实施过程中的差异,从满足核电站安全运行以及安全评审相关法规标准的需求出发,阐明核电站中不同安全分级的系统和设备对DCS总...     

Safety features of the gas-cooled fast breeder reactor

The safety features of the gas-cooled fast breeder reactor (GCFR) are described in the context of the 300-MW(e) demonstration plant design. They are of two general types, inherent and design-related. The inherent features are principally associated with the helium coolant and the nuclear coefficients. Design-related features influencing safety include shutdown systems, residual heat removal systems, method of core support, and the prestressed concrete reactor vessel (PCRV). This paper discusses the safety-related aspects of each of these. Recently completed residual heat removal system reliability studies are also discussed. The probability of residual heat removal system failure in the GCFR is found to be lower than that described for light water reactors. The safety characteristics of larger plants are examined, and increases in size are found to improve GCFR safety margins.     

Power-generating unit with dual-loop sodium-cooled fast reactor and dual-wall steam generator

The results of a probabilistic analysis performed to validate the safety of AES-2006 designed for the site of the Novovoronezh nuclear power plant are presented. The requirements for the AES-2006 design are examined. The characteristic features of the AES-2006 design for the conditions at the Novovoronezh nuclear power plant site are described, including the diversity of the equipment and operating regime, passive systems, and scheduled maintenance of safety systems with the reactor operating at power. The scope of the probabilistic safety analysis performed at the development stage of the technical design is described. The important problems which must be solved in a probabilistic safety analysis for the designs of new nuclear power plants are discussed. Translated from Atomnaya énergiya,Vol. 106, No. 3, pp. 123–129, March, 2009.     

Nuclear Power Plant Compliance with the Principle of Power-System Indepedence from Safety Systems Operation

The purpose of this paper is to show that during the operation of safety systems at nuclear power plants the principle of independence from the power system, which is one of the basic principles inocrporated in the design of safety systems, is not satisfied and the power system, especially if it is deficient, cannot guarantee the required electricity and protection for safety systems from general failures. To satisfy the independence principle, guarantee the required quality of electricity, and protect the safety systems in nuclear power plants from general failures, it is proposed that the presently operative algorithm for starting up diesel generators be reexamined. When the safety systems at nuclear power plants perform their required functions, they should operate from autonomous diesel generators at the nuclear power plant, which are equipped with electricity quality regulators (frequency and voltage), and not from the power system. It is also suggested that the variant of the algorithm where diesel generators are started up as a preventative measure when the quality of the electricity in the power system drops below admissable limits be reexamined.     

有关核电厂主给水系统设计的探讨

介绍了目前核电厂主给水系统隔离的几种设计方案，从事故进程和核电厂运行事件两个方面阐明了每种设计方案的优劣，得出了符合核安全原则的设计方案，这一分析对核电厂的设计和改造有一定的借鉴作用。     

Ecological problems in the development of nuclear power

Conclusions The initial and middle stages of the nuclear fuel cycle, i.e., mining and reprocessing of ore, uranium enrichment, production of fuel elements, and the normal operation of a nuclear power plant, do not cause any serious danger to the environment. Comparisons show that the negative-effect coal-fired HEP is much greater.The probability for accidents involving the emission of a large quantity of radionuclides in modern nuclear power plants equipped with tested safety systems is significantly lower than the accident probability in other areas of industry. This conclusion is valid, however, if safety requirements, starting with the nuclear power plant, are satisfiedunscrupulously, if the strictest technological discipline, making sure that all the components have sufficient reliability, is followed, and if constant efforts are made to train personnel.It is as yet impossible to evaluate quantitatively the environmental effects of reprocessing plants.Czechoslovakian Technical University, Prague. Translated from Atomnaya Énergiya, Vol. 49, No. 6, pp. 352–357, December, 1980.     

Problems and Experience in Guaranteeing the Nuclear Safety of the Storage of Spent Fuel from Nuclear Power Plants

Nuclear-safety problems are examined and the results of investigations of nuclear safety of storage sites are presented for spent nuclear fuel from nuclear power plants. The initial events of anticipated and unanticipated accidents, methods and errors in the calculation of k _eff taking account of burnup to ensure nuclear safety, the possibility of measuring k _eff of storage sites experimentally, and new forms of fuel with a consummable absorber are calculated.     

新一代CANDU—9重水堆核电厂简介

ＣＡＮＤＵ－９是电功率为９００ＭＷ级的重水堆核电厂,其设计基于达灵顿和布鲁斯Ｂ多机组核电厂,并融入了一些最新的工程设计和研究成果,除了继续采用成熟的系统和部件外,在安全性,地可靠性和可维护性方面作了重要改进。ＣＡＮＤＵ－９综合考虑了安全审评和执照申请过程中发现的问题,产使其体现在安全设计理念中,特别是对慢化剂系统,端屏蔽冷却系统,系统和应急堆芯冷却系统进行了改进。     

Availability of redundant safety systems with common-mode and undetected failures

The engineered safety features of nuclear power plants contain redundant standby safety systems that are tested periodically to ensure their operability should an emergency occur. Analytical equations are developed in this paper for the unavailability of common m-out-of-n systems, taking into account common-cause failures and failures that are not detectable by regular periodic tests. Consecutive, random and staggered testing schemes are considered. The unavailability model for individual components includes separate downtime contributions due to testing, repair, hardware failures, human testing and repair errors as well as failures caused by true demands. Computer codes have been developed to incorporate these failure modes in the unavailability analysis. The parameters of the model are estimated from documented field data, and typical systems are analyzed to illustrate the methods and the significance of common-mode failures.     

In-service monitoring of piping and vessels in nuclear power plants: report on BMU Study SR 2218

The BMU-Study SR 2218 is made with a view to describing and assessing fatigue monitoring as commonly applied today to piping and vessels in nuclear power plants. First, the fundamentals with regard to strain, instrumentation and calculation of fatigue are compiled and the current regulations listed. With reference to the available literature, own experience and a survey conducted among operators and manufacturers of the now common types of installations, the points of measurement, the methods of measuring, the acquisition systems for measured data and the methods of analysis are described, as are the results obtained from the measurements so far. After a careful review of these results, proposals are presented for improving the acquisition of measured data and, most important of all, for analysis and assessment. In the nuclear power plants in Germany those areas which are of relevance to fatigue have been identified by means of temperature measurements taken in the recent past in the various types of installations. Many thorough analyses have been made on the basis of transients measured in order to assess the relevance of the areas in terms of fatigue. The degrees of fatigue established are normally D=0.2–0.3, that is <D=0.5 for loads occurring in service. It can be stated, in summary, that the degrees of fatigue obtained from fatigue monitoring systems are more realistic than those obtained from the approach adopted in the design stage. As the requirements of the nuclear power plants are specific and varied, there can be no universal and flexible system that is adequate for all applications. That is why it will always be necessary to find solutions for each individual case. This paper gives an overview about the content and the results of the study.     

A practical system for isolating nuclear power plants from earthquake attack

The earthquake-generated forces and deformations of the main substructures of a nuclear power plant can be reduced by a factor of about ten times by mounting the power plant building on a recently developed base-isolation system. The very high forces which the ‘resonant appendage’ effect may induce in some critical components (such as fuel elements and control rods) may be reduced by a factor of 40 or more by the isolation system. This system combines recently developed hysteretic dampers with components which support the structure while providing high flexibility for horizontal motion. These dampers utilize the plastic deformation of solid steel beams, while a convenient support system with adequate horizontal flexibility may be provided by laminated rubber bearings of the type frequently used to support bridge decks. The earthquake attack on a nuclear power plant is particularly hazardous since it attacks simultaneously all the plant components, including the control and safety systems. Undetected deterioration of a set of components may further increase the probability of multiple failure during earthquakes. Hence the large reduction in earthquake-induced forces and deformations, which may be achieved with the base-isolation system described, will greatly reduce the likelihood of earthquake-induced accident or damage in those nuclear power plants located in regions of high seismicity.