Core safety of Indian nuclear power plants (NPPs) under extreme conditions

Nuclear power is currently the fourth largest source of electricity production in India after thermal, hydro and renewable sources of electricity. Currently, India has 20 nuclear reactors in operation and seven other reactors are under construction. Most of these reactors are indigenously designed and built Heavy Water Reactors. In addition, a 300 MWe Advanced Heavy Water Reactor has already been designed and in the process of deployment in near future for demonstration of power production from Thorium apart from enhanced safety features by passive means. India has ambitious plans to enhance the share of electricity production from nuclear. The recent Fukushima accident has raised concerns of safety of Nuclear Power Plants worldwide. The Fukushima accident was caused by extreme events, i.e., large earthquake followed by gigantic Tsunami which are not expected to hit India’s coast considering the geography of India and historical records. Nevertheless, systematic investigations have been conducted by nuclear scientists in India to evaluate the safety of the current Nuclear Power Plants in case of occurrence of such extreme events in any nuclear site. This paper gives a brief outline of the safety features of Indian Heavy Water Reactors for prevention and mitigation of such extreme events. The probabilistic safety analysis revealed that the risk from Indian Heavy Water Reactors are negligibly small.     

Intelligent decision aids for abnormal events in nuclear power plants

German nuclear power plants are characterized by a high degree of automation, not only for normal operation but also for abnormal events. Therefore the role of the operating personnel is mainly a supervisory function. Nevertheless, for a spectrum of unexpected events the operating personnel have to react with manual recovery actions. In order to minimize human error in such recovery actions, different kinds of intelligent decision aid support the operators today. In this paper such aids are discussed and one of them is described in more detail.     

Thresholds and goals for safety performance indicators for nuclear power plants

Performance indicators have found a wide use in the nuclear industry. IAEA developed a safety performance indicators framework for use by NPPs. Karachi Nuclear Power Complex (KNPC) worked, under the auspices of IAEA, on developing guidelines for establishing goals and thresholds, aggregation to higher levels and trend indexes. This paper describes the work done by KNPC on goals and thresholds along with the recommended guidelines.     

Incorporating organizational factors into probabilistic safety assessment of nuclear power plants through canonical probabilistic models

The ω-factor approach is a method that explicitly incorporates organizational factors into Probabilistic safety assessment of nuclear power plants. Bayesian networks (BNs) are the underlying formalism used in this approach. They have a structural part formed by a graph whose nodes represent organizational variables, and a parametric part that consists of conditional probabilities, each of them quantifying organizational influences between one variable and its parents in the graph. The aim of this paper is twofold. First, we discuss some important limitations of current procedures in the ω-factor approach for either assessing conditional probabilities from experts or estimating them from data. We illustrate the discussion with an example that uses data from Licensee Events Reports of nuclear power plants for the estimation task. Second, we introduce significant improvements in the way BNs for the ω-factor approach can be constructed, so that parameter acquisition becomes easier and more intuitive. The improvements are based on the use of noisy-OR gates as model of multicausal interaction between each BN node and its parents.     

核电厂地震概率风险评估研究综述

随着概率地震危险性分析与分解理论与应用的发展,场地相关谱生成理论也得到了不断发展,基于中国场地相关谱的核工程等重要基础设施地震易损性与风险分析研究还较为匮乏。该文总结了标量型中国概率地震危险性分析与分解理论方法,提出了向量型中国概率地震危险性分析与分解、条件型中国概率地震危险性分析基本原理,给出了基于中国概率地震危险性分析与分解的我国场地一致危险谱、条件均值谱、广义条件均值谱和条件一致危险谱生成理论和方法,总结了基于中国场地相关谱的核电厂结构地震易损性与风险分析理论基础,以我国某核电厂厂址及核电厂安全壳结构为算例,生成算例厂址场地相关谱,计算不同场地相关谱条件下核电厂安全壳结构地震易损性与风险。分析结果表明：不同场地相关谱条件下,我国核电厂安全壳结构安全裕量都较大;基于条件均值谱计算得到的风险结果偏于不保守。

     

Seismic probabilistic risk assessment and seismic margins studies for nuclear power plants

This paper reviews the seismic probabilistic risk assessment and seismic margins studies for nuclear power plants in the United States. The techniques employed in these studies are briefly described. A few comments on the evaluation of the fragility of structures and equipment are discussed. Seismic PRA is a systematic process to evaluate the safety of nuclear power plants. In the process, it integrates all the elements such as seismic hazard, component fragility and plant system. Thus, it provides the overall view of the safety of an entire plant under a seismic event.

The major tasks of a seismic PRA such as the evaluation of hazard curves, component fragility and plant system are also present in probabilistic analyses of nonnuclear facilities. The concept and technique embodied in seismic PRA for nuclear power plants can be applied to other types of engineering facilities.     

核电站安全的地震威胁及对策分析

核能是一种清洁高效的无碳能源,在人类文明进程及世界现代化发展中具有重要意义,但也存在核泄漏和核废料等问题。对世界核能利用情况及各国核电站建设与规划情况进行了回顾与展望,总结了历史上历次核电站事故发生的根本原因,着重探讨了地震对核电站的威胁。以2011年日本大地震造成的福岛核电站事故为例,分析了核电站事故对社会、经济及生态环境造成的影响,同时还提出了一些建议:对已建核电站进行抗震加固和延寿评估,做好未来核电站建设规划,改善管理机制,做好地震预警及核电站安全监测,优化应急处置机制,深化核能概率安全分析,普及群众安全教育,定期开展灾难演习,强化国际合作等举措,为人类在探索正确使用核能的进程中少受核能副作用的伤害而发挥重要作用。     

基于贝叶斯网络的核电站应急电力系统安全评价

在故障树分析法(FTA)基础上提出了一种基于贝叶斯网络(BN)的核电站应急电力系统安全评价方法,比较了FTA和BN在建立安全评价模型和评价能力上的不同.该方法在应对众多影响因素上有很大优势,能进行更多有意义的分析:既能进行前向的预测推理,又能进行后向的诊断推理,可以找出影响故障的组合模式,从而能够找出系统的薄弱环节.同时采用基于Matlab的BNT软件包,大大简化了计算过程.通过对10MW高温气冷堆(HTR-10)应急电力系统的安全评价实例的分析,证明该方法是对传统的基于故障树分析的安全评价方法的有益改进.     

Development of human resources for Indian nuclear power programme

The continuing research and development on nuclear technology by research establishments in the country and maturing of Indian industry have brought the nuclear energy programme in India to a stage where it is poised to take a quantum leap forward. The vision of expansion of nuclear power also requires a well-structured specialized human resource development programme. This paper discusses the requirements of the human resource development programme for nuclear energy, the challenges in the way of its realization, its national and international status and traces the history of nuclear education in the country. It brings out the linkage of human resource development programme with the nuclear energy programme in the country. It also describes the initiatives by the university system in the area of nuclear education and support provided by the Department of Atomic Energy to the university system by way of extra-mural funding and by providing access to research facilities.     

Issues related to structural aging in probabilistic risk assessment of nuclear power plants

Structural components and systems have an important safety function in nuclear power plants. Although they are essentially passive under normal operating conditions, they play a key role in mitigating the impact of extreme environmental events such as earthquakes, winds, fire and floods on plant safety. Moreover, the importance of structural components and systems in accident mitigation is amplified by common-cause effects. Reinforced concrete structural components and systems in NPPs are subject to a phenomenon known as aging, leading to time-dependent changes in strength and stiffness that may impact their ability to withstand various challenges during their service lives from operation, the environment and accidents. Time-dependent changes in structural properties as well as challenges to the system are random in nature. Accordingly, condition assessment of existing structures should be performed within a probabilistic framework. The mathematical formalism of a probabilistic risk assessment (PRA) provides a means for identifying aging structural components that may play a significant role in mitigating plant risk. Structural condition assessments supporting a decision regarding continued service can be rendered more efficient if guided by the logic of a PRA.     

A method for identifying instrument faults in nuclear power plants possibly leading to wrong situation assessment

In this paper, we propose a method for identifying instrument faults that could potentially affect an operators’ situation assessment capability in nuclear power plants (NPPs), an issue which has received a lot of attention recently. In the proposed method, patterns of selected plant parameter trends in selected plant states and NPP operators’ patterns of selected plant parameter trends in selected plant states are analyzed, and a comparison between the two kinds of patterns is performed to identify instrument faults which could potentially affect a NPP operators’ capability to correctly assess a plant's conditions. An example application is presented to demonstrate how the proposed method can be used to identify the possibilities of operators’ developing a wrong situation assessment because of instrument faults and to identify the corresponding safety concerns. We conclude that in order to get more accurate results, an analysis with a full-scope NPP simulator and interviews with NPP operators will be necessary.     

Safety features in nuclear power plants to eliminate the need of emergency planning in public domain

Following the Fukushima accident, the safety features of Nuclear Power Plants (NPP) are being re-examined worldwide including India to demonstrate capabilities to cope with severe accidents. In order to restore public confidence and support for nuclear power, it is felt necessary to design future NPPs with near zero impact outside the plant boundary and thus enabling elimination of emergency planning in public domain. Authors have identified a set of safety features which are needed to be incorporated in advanced reactors to achieve this goal. These features enabling prevention, termination, mitigation and containment of radioactivity for beyond design basis accidents arising from extreme natural events are essential for achieving the goal of elimination of emergency planning in public domain. Inherent safety characteristics, passive and engineered safety features to achieve these functions are discussed in this paper. Present trends and future developments in this direction are also described briefly.     

Environmental prediction, risk assessment and extreme events: adaptation strategies for the developing world

The uncertainty associated with predicting extreme weather events has serious implications for the developing world, owing to the greater societal vulnerability to such events. Continual exposure to unanticipated extreme events is a contributing factor for the descent into perpetual and structural rural poverty. We provide two examples of how probabilistic environmental prediction of extreme weather events can support dynamic adaptation. In the current climate era, we describe how short-term flood forecasts have been developed and implemented in Bangladesh. Forecasts of impending floods with horizons of 10 days are used to change agricultural practices and planning, store food and household items and evacuate those in peril. For the first time in Bangladesh, floods were anticipated in 2007 and 2008, with broad actions taking place in advance of the floods, grossing agricultural and household savings measured in units of annual income. We argue that probabilistic environmental forecasts disseminated to an informed user community can reduce poverty caused by exposure to unanticipated extreme events. Second, it is also realized that not all decisions in the future can be made at the village level and that grand plans for water resource management require extensive planning and funding. Based on imperfect models and scenarios of economic and population growth, we further suggest that flood frequency and intensity will increase in the Ganges, Brahmaputra and Yangtze catchments as greenhouse-gas concentrations increase. However, irrespective of the climate-change scenario chosen, the availability of fresh water in the latter half of the twenty-first century seems to be dominated by population increases that far outweigh climate-change effects. Paradoxically, fresh water availability may become more critical if there is no climate change.     

The development and application of the accident dynamic simulator for dynamic probabilistic risk assessment of nuclear power plants

This paper describes the principal modelling concepts, practical aspects, and an application of the Accident Dynamic Simulator (ADS) developed for full scale dynamic probabilistic risk assessment (DPRA) of nuclear power plants. Full scale refers not only to the size of the models, but also to the number of potential sequences which should be studied. Plant thermal-hydraulics behaviour, safety systems response, and operator interactions are explicitly accounted for as integrated active parts in the development of accident scenarios. ADS uses discrete dynamic event trees (D-DET) as the main accident scenario modelling approach, and introduces computational techniques to minimize the computer memory requirement and expedite the simulation. An operator model (including procedure-based behaviour and several types of omission and commission errors) and a thermal-hydraulic model with a PC run time more than 300 times faster than real accident time are among the main modules of ADS. To demonstrate the capabilities of ADS, a dynamic PRA of the Steam Generator Tube Rupture event of a US nuclear power plant is analyzed.     

优化设计输入和分析方法以提高核电厂抗震安全性

日本福岛事故后核电厂抵御极端自然灾害能力受到广泛关注,世界上各个国家都积极开展相关研究,而地震一直是核电厂工程安全问题的主要威胁之一,因此核电厂的抗震安全性更是成为业界分析研究的重点。楼层反应谱作为核电厂系统、结构和部件抗震设计的输入,其计算分析是核电厂抗震分析的重要环节,其结果对于核电厂的抗震安全水平起着举足轻重的作用。本文以中国核电工程有限公司自主研发的三代机型ACP1000标准设计为例,通过介绍楼层反应谱的输入、分析过程和方法,来阐述合理确定符合国情的地震输入、采用先进的建模和分析方法,对完善核电厂的抗震设计、提高核电厂的抗震安全性具有重要意义。     

Integrated assessment of pedestrian head impact protection in testing secondary safety and autonomous emergency braking

Pedestrian impact testing is used to provide information to the public about the relative level of protection provided by different vehicles to a struck pedestrian. Autonomous Emergency Braking (AEB) is a relatively new technology that aims to reduce the impact speed of such crashes. It is expected that vehicles with AEB will pose less harm to pedestrians, and that the benefit will come about through reductions in the number of collisions and a change in the severity of impacts that will still occur. In this paper, an integration of the assessment of AEB performance and impact performance is proposed based on average injury risk. Average injury risk is calculated using the result of an impact test and a previously published distribution of real world crash speeds. A second published speed distribution is used that accounts for the effects of AEB, and reduced average risks are implied. This principle allows the effects of AEB systems and secondary safety performance to be integrated into a single measure of safety. The results are used to examine the effect of AEB on Euro NCAP and ANCAP assessments using previously published results on the likely effect of AEB. The results show that, given certain assumptions about AEB performance, the addition of AEB is approximately the equivalent of increasing Euro NCAP test performance by one band, which corresponds to an increase in the score of 25% of the maximum.     

Environmental impact assessment process substitution: experiences of public participants

This paper explores policy action taken by the federal government in Canada to test the potential for substitution of the federal process to regulatory bodies as a means of encouraging environmental impact assessment (EIA) efficiency. Our purpose is to present the experiences of people who participated in the EIA of the Emera Brunswick Pipeline in order to gauge the impact of such process substitutions on meaningful public participation. Our approach included document reviews and a focus group session with some of the public participants in the Emera Pipeline hearings. We find that the National Energy Board hearing process substituted in the case did not meet many of the key requirements of meaningful participation and left some public participants feeling disrespected and marginalized.     

Application of genetic algorithm for reliability allocation in nuclear power plants

Reliability allocation is an optimization process of minimizing the total plant costs subject to the overall plant safety goal constraints. Reliability allocation was applied to determine the reliability characteristics of reactor systems, subsystems, major components and plant procedures that are consistent with a set of top-level performance goals; the core melt frequency, acute fatalities and latent fatalities. Reliability allocation can be performed to improve the design, operation and safety of new and/or existing nuclear power plants. Reliability allocation is a kind of a difficult multi-objective optimization problem as well as a global optimization problem. The genetic algorithm, known as one of the most powerful tools for most optimization problems, is applied to the reliability allocation problem of a typical pressurized water reactor in this article. One of the main problems of reliability allocation is defining realistic objective functions. Hence, in order to optimize the reliability of the system, the cost for improving and/or degrading the reliability of the system should be included in the reliability allocation process. We used techniques derived from the value impact analysis to define the realistic objective function in this article.     

Reliability analysis of microcomputer circuit modules and computer based control systems important to safety of nuclear power plants

Computer based safety related control and instrumentation (C&I) systems are being employed in Indian Nuclear Power Plants (NPPs). These systems are designed around standardized family of microcomputer based circuit modules, which are qualified to the stringent requirements of the nuclear industry. Reliability analysis of standardized microcomputer circuit modules, used in the safety related C&I systems, were carried out using the analysis package based on the methodology and database of MIL-STD-217-F1. The circuit modules are the main building blocks of the safety related C&I systems in the forthcoming Indian NPPs. The article presents reliability analysis results of microcomputer and related circuit modules and a representative safety C&I system-Programmable Digital Comparator System (PDCS). Comparison of reliability values for prototype PDCS using commercial grade components, and for upgraded version PDCS using MIL grade or equivalent screened components was made. The estimated failure rate values of standardized microcomputer circuit modules will be useful, for reliability assessment of various other safety related C&I systems developed around these modules, for ongoing and future Indian NPPs.