Seismic PSA method for multiple nuclear power plants in a site

The maximum number of nuclear power plants in a site is eight and about 50% of power plants are built in sites with three or more plants in the world. Such nuclear sites have potential risks of simultaneous multiple plant damages especially at external events. Seismic probabilistic safety assessment method (Level-1 PSA) for multi-unit sites with up to 9 units has been developed. The models include Fault-tree linked Monte Carlo computation, taking into consideration multivariate correlations of components and systems from partial to complete, inside and across units. The models were programmed as a computer program CORAL reef. Sample analysis and sensitivity studies were performed to verify the models and algorithms and to understand some of risk insights and risk metrics, such as site core damage frequency (CDF per site-year) for multiple reactor plants. This study will contribute to realistic state of art seismic PSA, taking consideration of multiple reactor power plants, and to enhancement of seismic safety.     

一管多机布置抽水蓄能电站非线性动力学建模与瞬态分析

一管多机布置抽水蓄能电站存在共用管段引水系统,其中共用管段的水力耦合相互作用是抽水蓄能电站系统动力学建模的关键。考虑共用管段水力系统耦合作用,将其分解成多个仅依赖于各子管段流量的单变量函数,通过结合水泵水轮机组动态特性,建立一管多机布置抽水蓄能电站非线性动力学模型。利用数值模拟探究了两机在额定工况并列运行时,一机突甩全负荷在不同导叶关闭规律且另一机组在调速器PID控制下,对各支管内机组流量、水头、出力动态特性的影响规律。仿真结果表明正常运行机组出力和流量动态响应具有正相关关系,机组水头动态响应与其流量和出力具有负相关关系,且波动周期相近。突甩负荷机组水头与流量的动态响应与导叶关闭规律密切相关,突甩负荷机组水头波动程度始终高于正常运行机组水头,且两机组水头波动周期相近;研究结果为探究一管多机抽水蓄能电站系统瞬态动力学建模与稳定性调控提供理论参考。     

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.     

Risk optimization: siting of nuclear power electricity generating units

Radiological doses from nuclear power plant accidents can extend to great distances away from the site. These technical and perceived risks of nuclear power generation are addressed in an optimization of the siting of additional nuclear units primarily in the middle Atlantic and northeastern United States of America. This study relies on a constrained optimization of long-term latent cancer and early fatality effects from probable severe accident releases to the public. The results provide policy perspectives that would benefit from nuclear power, but provide an inherent safety margin in the selection of sites for future generating units of electricity. The selection of sites from among the best is aided by these calculations for minimizing fatality risks at 17 existing sites. The mathematical model minimizes a linear function represented as the number of 1000 MWe units multiplied by the estimated severe accident site risk at existing sites assuming constraints on: the allowed percentage increase in total societal latent cancer risk as well as allowed percentage increase in early fatality risk; lower bounds for existing units at the sites; and an upper bound of doubling the units and risks. The early fatality risk was also varied for several levels of increase in latent cancer risk.     

Formulation of a MIMO linear power system stabiliser for identical generating units at a common busbar

The fundamental formulation, analysis and sample simulation studies reported show how the linear power system stabiliser (PSS) can be evolved so as to accommodate more than one identical power generating unit operating at a common busbar. The presentation augments the more conventional approach by which the PSS is formulated for a single-machine-infinite-busbar system. In preference to single-generator installations, practical generating stations frequently configure multiple identical units to operate in parallel. Although this reduces dependence on a single unit, a multi-unit configuration is susceptible to oscillations between two or more units. Such oscillations are classified as intraplant dynamics. The proposed PSS formulation includes rigorous consideration of intraplant dynamics at the design stage. Simulation studies clearly substantiate the superiority of the proposed multi-input-multi-output PSS over the more conventional single-input-single-output version.     

A study of a commercial MHD power plant scheme

Power engineering specialists are currently interested in electrical power stations with magnetohydrodynamic generators. This interest has been generated by the fact that fossil fuels are becoming increasingly costlier, and with the exploitation of remote and practically inaccessible deposits, a more rational utilisation of fuel has become necessary. Research on magnetohydrodynamic generators is being conducted in many countries at present. In the USSR a composite pilot plant with an MHD generator whose output exceeds 20 MW has already been operative for several thousand hours. However the pilot plant has to be considerably modified to serve as a model for the 500–1000 MW industrial power unit. This paper is devoted to an investigation of one of the possible process flow diagrams of MHD electrical power plants. The structure of MHD electrical power plants, the interrelation between the aggregates, issues concerning the starting of the plant and the working of the power unit under various partial load conditions are discussed. With the availability of new theoretical and experimental data, the process flow diagrams of industrial MHD electrical power plants will naturally undergo changes. However, the methodical approach and the investigations, described in this paper should retain their validity for all process flow diagrams of electrical power plants with MHD generators.     

Methods and practices used in incident analysis in the Finnish nuclear power industry

According to the Finnish Nuclear Energy Act it is licensee's responsibility to ensure safe use of nuclear energy. Radiation and Nuclear Safety Authority (STUK) is the regulatory body responsible for the state supervision of the safe use of nuclear power in Finland. One essential prerequisite for the safe and reliable operation of nuclear power plants is that lessons are learned from the operational experience. It is utility's prime responsibility to assess the operational events and implement appropriate corrective actions. STUK controls licensees' operational experience feedback arrangements and implementation as part of its inspection activities. In addition to this in Finland, the regulatory body performs its own assessment of the operational experience. Review and investigation of operational events is a part of the regulatory oversight of operational safety. Review of operational events is done by STUK basically at three different levels. First step is to perform a general review of all operational events, transients and reactor scram reports, which the licensees submit for information to STUK. The second level activities are related to the clarification of events at site and entering of events' specific data into the event register database of STUK. This is done for events which meet the set criteria for the operator to submit a special report to STUK for approval. Safety significance of operational events is determined using probabilistic safety assessment (PSA) techniques. Risk significance of events and the number of safety significant events are followed by STUK indicators. The final step in operational event assessment performed by STUK is to assign STUK's own investigation team for events deemed to have special importance, especially when the licensee's organisation has not operated as planned. STUK launches its own detail investigation once a year on average. An analysis and evaluation of event investigation methods applied at STUK, and at the two Finnish nuclear power plant operators Teollisuuden Voima Oy (TVO) and Fortum Power and Heat Oy (Fortum) was carried out by the Technical Research Centre (VTT) on request of STUK at the end of 1990s. The study aimed at providing a broad overview and suggestions for improvement of the whole organisational framework to support event investigation practices at the regulatory body and at the utilities. The main objective of the research was to evaluate the adequacy and reliability of event investigation analysis methods and practices in the Finnish nuclear power industry and based on the results to further develop them. The results and suggestions of the research are reviewed in the paper and the corrective actions implemented in event investigation and operating experience procedures both at STUK and at utilities are discussed as well. STUK has developed its own procedure for the risk-informed analysis of nuclear power plant events. The PSA based event analysis method is used to assess the safety significance and importance measures associated with the unavailability of components and systems subject to Technical Specifications. The insights from recently performed PSA based analyses are also briefly discussed in the paper.     

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.     

基于PSHA的核电厂近断层抗震设计谱构建方法

核电是一种高效、清洁的能源,随着核电厂未来向内陆区的发展,其可能会遭遇到近断层地震动的影响,但是目前我国核电厂抗震规范设计谱并未考虑近断层地震动。该文首先基于大量实际近断层脉冲型和相应无脉冲地震动记录,研究了脉冲对反应谱的放大效应,建立了修正的近断层脉冲放大系数模型;继而将地震动脉冲效应引入到近断层概率地震危险性分析中,并基于设定断层模型,给出了不同场地类型的一致危险性反应谱;通过对地震危险性结果的分解,分析了对场地最危险震级和距离,并将结果引入地震动衰减关系中得到设计谱,最后通过近断层脉冲放大系数对设计谱进行修正,得到考虑近断层脉冲效应的核电厂抗震设计谱。通过研究,建立了一种基于概率地震危险性分析框架下,考虑近断层脉冲型地震动的工程场地核电厂抗震设计谱的构建方法。     

Human factors, human reliability and risk assessment in license renewal of a nuclear power plant

This paper describes human factors and human reliability assessments carried out as a part of operating license renewal of a nuclear power plant. The structure and contents of human factors assessments, the source material and the role of probabilistic safety assessment are described. Similar evaluations are recommended as an integral part of periodic safety reviews of regulated industrial facilities.The qualitative part of the human factors review is structured according to an international guide. The assessments are here enhanced with operating experience evaluations, measured by quantitative statistical data obtained from inspections and assessments made by plant safety and quality assurance personnel, by regulatory authorities and by peer reviews.The quantitative assessment is based on the roles and contributions of human errors in the accident risk of the target plant. The assessment uses importance measures quantified in probabilistic risk assessment. The scope and the quality of the risk assessment and the scope and the quality of human reliability assessments are also taken into account. Furthermore, the assessment describes how risk assessment can be used to reduce errors and improve human factors. The results tend to be very plant-specific, and the errors have very different importances in different operating states and for different initiating event categories. The results are useful for planning preventive actions, i.e. for preventing errors by developing and prioritizing human factors improvement activities.     

Development of a dynamical systems model of plant programmatic performance on nuclear power plant safety risk

Application of probabilistic risk assessment (PRA) techniques to model nuclear power plant accident sequences has provided a significant contribution to understanding the potential initiating events, equipment failures and operator errors that can lead to core damage accidents. Application of the lessons learned from these analyses has resulted in significant improvements in plant operation and safety. However, this approach has not been nearly as successful in addressing the impact of plant processes and management effectiveness on the risks of plant operation. The research described in this paper presents an alternative approach to addressing this issue. In this paper we propose a dynamical systems model that describes the interaction of important plant processes on nuclear safety risk. We discuss development of the mathematical model including the identification and interpretation of significant inter-process interactions. Next, we review the techniques applicable to analysis of nonlinear dynamical systems that are utilized in the characterization of the model. This is followed by a preliminary analysis of the model that demonstrates that its dynamical evolution displays features that have been observed at commercially operating plants. From this analysis, several significant insights are presented with respect to the effective control of nuclear safety risk. As an important example, analysis of the model dynamics indicates that significant benefits in effectively managing risk are obtained by integrating the plant operation and work management processes such that decisions are made utilizing a multidisciplinary and collaborative approach. We note that although the model was developed specifically to be applicable to nuclear power plants, many of the insights and conclusions obtained are likely applicable to other process industries.     

Determination of the source rate released into the environment from a nuclear power plant

A study was carried out to determine the source rate from a nuclear power plant using the tracer experimental data conducted at the Yeoung-Kwang nuclear site. The least squares method optimises the agreement between the released source rate and the calculated source rate by minimising the errors between the measured concentrations and the calculated ones using the Gaussian plume model. The least squares estimator generally estimates the source rate to be within a factor of 2. The forecasting ability of the source rate is improved by applying the modified dispersion coefficients that are calculated using the experimental data. Determination of the source rate in an early phase nuclear emergency will be helpful for the decision making when taking appropriate and prompt countermeasures in the case of a radiological emergency.     

A quantitative analysis of a risk impact due to a starting time extension of the emergency diesel generator in optimized power reactor-1000

An emergency diesel generator (EDG) is the ultimate electric power supply source for the operation of emergency engineered safety features when a nuclear power plant experiences a loss of off-site power (LOOP). If a loss of coolant accident (LOCA) with a simultaneous LOOP occurs, the EDG should be in the state of a full power within 10 s, which is a prescribed regulatory requirement in the technical specifications (TS) of the Optimized Power Reactor-1000 (OPR-1000).Recently, the US nuclear regulatory commission (NRC) has been preparing a new risk-informed emergency core cooling system (ECCS) rule called 10 CFR 50.46. The new rule redefines the size for the design basis LOCA and it relaxes some of the requirements such as the single failure criteria, simultaneous LOOP, and the methods of analysis. The revision of the ECCS rule will provide flexibility for plant changes if the plant risks are checked and balanced with the specified criteria.The present study performed a quantitative analysis of the plant risk impact due to the EDG starting time extension given that the new rule will be applied to OPR-1000. The thermal-hydraulic analysis and OPR-1000 probabilistic safety assessment (PSA) model were combined to estimate the whole plant risk impact. Also, sensitivity analyses were implemented for the important uncertainty parameters.     

Maximal network reliability for a stochastic power transmission network

Many studies regarded a power transmission network as a binary-state network and constructed it with several arcs and vertices to evaluate network reliability. In practice, the power transmission network should be stochastic because each arc (transmission line) combined with several physical lines is multistate. Network reliability is the probability that the network can transmit d units of electric power from a power plant (source) to a high voltage substation at a specific area (sink). This study focuses on searching for the optimal transmission line assignment to the power transmission network such that network reliability is maximized. A genetic algorithm based method integrating the minimal paths and the Recursive Sum of Disjoint Products is developed to solve this assignment problem. A real power transmission network is adopted to demonstrate the computational efficiency of the proposed method while comparing with the random solution generation approach.     

填料-高压微雾复合式两级直接蒸发冷却空调机组的研制

介绍核电站常规岛用两级直接蒸发冷却空调机组的优点、工作原理及设计过程,并对实验样机进行测试。实验数据表明,在南通中湿度地区,当室外空气的干湿球温差为4~5℃时,该组合式空调机组对室内空气降4~5℃,使之接近室外空气的湿球温度。指出两级直接蒸发冷却空调机组在核电站常规岛中具有广泛的应用前景。     

Procedures and applications to enlarge the level 1+ PSA to internal fires in German nuclear power plants

Investigations have shown that the consequences from fires in nuclear power plants can be significant. Methodologies considering fire in probabilistic safety analyses have been evolving in the last few years. In order to provide a basis for further discussions on benefits and limits of such an analysis in Germany, current methods are investigated. As a result a qualitative screening process is proposed to identify critical fire zones followed by a quantitative event tree analysis in which the fire caused frequency of initiating events and different core damage states will be determined. The models and data proposed for a probabilistic fire risk analysis have been successfully applied in complete and partial fire risk assessments in German nuclear power plants.     

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.     

Calculating conditional core damage probabilities for nuclear power plant operations

A part of managing nuclear power plant operations is the control of plant risk over time as components are taken out of service or plant upsets are caused by initiating events. Unfortunately, measuring risk over time proves to be challenging, even with modern probabilistic risk analyses (PRAs) and PRA tools. In general, the process of measuring the operational risk would satisfy three desires: (1) the measurement would provide the risk magnitude for a particular event or over a period of time; (2) the risk results could be summed for a period of time to obtain a cumulative risk profile; and (3) the measurement process would be tractable while still using the current modeling techniques and tools. This paper demonstrates the calculation of the conditional core damage probability (CCDP) for the two cases of component outages and initiating events. In addition, two potential complications were identified that must be addressed when performing a CCDP calculation. The first complication, determining the appropriate nonrecovery probabilities to be applied to an inoperable component or initiating event, addresses the possibility of the plant operators preventing damage to the plant from their actions. The second complication, adjusting common-cause probabilities specific to the plant configuration, accounts for the fact that the PRA common-cause probabilities built into the model are applicable only during nominal conditions. The examples presented in the paper illustrate the potential under-estimation in CCDP when modifications to common-cause probabilities are ignored. These underestimation errors ranged from a factor of two to over a factor of six underestimation in CCDP.     

A framework for evaluating hydrogen control and management

The present paper presents a new framework for assessing accident management strategies using decision trees. The containment event tree (CET) model considers characteristics associated with the implementation of each strategy. It is constructed and quantified using data obtained from NUREG-1150, other probabilistic risk assessments, and the MAAP4 calculations. The proposed framework for evaluating hydrogen control strategies is based on the concept of a measure using a risk triplet. Ulchin units of nuclear power plants 3 and 4 are used as the reference plant. On the basis of best-estimate assessment, it is shown that it is beneficial to execute hydrogen igniters rather than to do nothing with respect to expected value of hydrogen concentration in the containment during an accident. The proposed approach is shown to be flexible in that it can be applied to various accident management strategies based on the timing of mitigation. The advantage of using the CET for assessing an accident management strategy lies with its capability for modeling both the positive and negative aspects associated with progression of the accident, which may in turn affect the containment failure mode.