共查询到20条相似文献,搜索用时 390 毫秒
1.
R. T. Islamov A. A. Derevyankin I. V. Zhukov M. A. Berberova I. V. Glukhov D. R. Islamov 《Atomic Energy》2011,109(6):375-379
The objective of the present work is to develop recommendations for controlling the safety of nuclear power plants on the
basis of risk assessments and safety certification of nuclear power plants. The Kursk nuclear power plant is considered as
an example of a nuclear power plant with an RBMK reactor. The concept of risk assessment of a nuclear power plant consists
in constructing a set of scenarios of the appearance and development of possible accidents followed by an evaluation of the
realization frequency and determination of the scales of the consequences of each one. The result of an analysis is an evaluation
of a system of risk indicators in accordance with the requirements of the safety compliance certificate of the nuclear power
plant as well as the development of recommendations for increasing plant safety. In risk assessment, the consequences are
divided into categories of the seriousness of the damage, for which their probability is evaluated separately. The graphical
interpretation of risk due to any dangerous object consists of frequency–consequences curves. Recommendations are developed
on the basis of the results of risk analysis. 相似文献
2.
G. A. Ershov Yu. L. Ermakovich M. A. Kozlov M. A. Parfentiev A. I. Kalinkin A. A. Kalinkin 《Atomic Energy》2010,109(2):81-87
The application of the BARS program system for choosing the optimal parameters for checking the serviceability of the safety
systems of nuclear power plants is described. The types of checks as well as the mathematical models determining the reliability
of the components of the system for each type of check are described. Calculations of the failure probability of the safety
systems of nuclear power plants which incorporate components with different levels of reliability are presented. The values
of the optimal period between serviceability checks for each type of system are determined. An evaluation of the effect of
failures of a general type on the reliability of safety systems is presented for two forms of the α-factor model. 相似文献
3.
风险指引的安全裕度是近十年来核工业界提出的新的安全理念。本文阐述了基于离散动态事件树的风险指引的安全裕度分析方法,给出该方法下核燃料包壳失效概率均值和标准差的数学表达式。针对简化压水堆模型下的全厂断电事故,提出了基于离散动态事件树的风险指引的安全裕度计算流程,计算了两种离散动态事件树分支规则下燃料包壳失效的风险指引的安全裕度及其不确定性。计算结果表明,不同的分支规则、模型参数分布、系统程序最大时间步长对核燃料包壳失效概率均值和标准差均有显著影响。提出了一种改进的可变概率阈值的分支方法,以更好地平衡风险指引的安全裕度分析过程中计算精度与计算资源的匹配问题。 相似文献
4.
A seismic risk analysis has been performed to evaluate the seismic safety of a nuclear power plant for strong earthquakes beyond a design earthquake level. A site-specific median spectrum has generally been used for a seismic fragility analysis of structures and equipment in Korean nuclear power plants as a part of a probabilistic seismic risk assessment. The site-specific response spectrum, however, does not represent the same probability of an exceedance over entire frequency range of interest. The site-specific uniform hazard spectrum (UHS) is more appropriate for use in a seismic probabilistic risk assessment (SPRA) than the site-specific spectrum, since there are only a few strong motion data and seismological information for the nuclear plant sites in Korea. In this study, the uniform hazard spectra were developed using the available seismic hazard data for four Korean NPP sites. 相似文献
5.
概率安全分析的发展及应用展望 总被引:1,自引:0,他引:1
对于核动力厂,概率安全分析(PSA)是评价风险、认识风险和管理风险的有效工具.本文介绍了PSA技术在国内及国际上发展和应用情况,并结合我国实际对PSA的发展应用进行了一些展望. 相似文献
6.
Analysis of criticality in shipment and storage of fuel at a nuclear power plant with a VVÉR reactor
G. L. Ponomarenko 《Atomic Energy》1999,87(1):466-471
The substantiation of nuclear safety during shipment and storage of fresh and spent fuel at nuclear power plants with VVéR
reactors is examined in the light of the more stringent nuclear safety rules. Possible technical measures for satisfying the
safety criterion are examined, for example, the concept of subcritical fresh fuel. An example of the estimation of the probability
of the formation of a critical mass as result of fuel assemblies falling randomly out of a container is presented. Certain
characteristic features of the calculation of the neutron-physical characteristics of fuel in a cooling pond are presented,
for example, the nonconservative nature of a separate analysis in the infinite approximation. 4 figures, 5 references.
OKB “Gidropress”. Translated from Atomnaya éneriya, Vol. 87, No. 1, pp. 11–16, July, 1999. 相似文献
7.
《Journal of Nuclear Science and Technology》2012,49(1):121-132
ABSTRACTHuman-induced initiators (category-B actions) are the initiators that are caused by human errors and are rarely explicitly identified and modeled in probabilistic safety assessments (PSAs). The current concern over the safety of multi-unit nuclear power sites is also a motivation for this research. This study proposes a novel process for identifying and quantifying category-B actions and ultimately, how to derive a human-induced initiating event frequency in a multi-unit scenario. Hence, this study fundamentally applies a scenario–system–action search scheme using maintenance and testing procedures, quantifies the human error probability by using the cause-based decision tree and technique for human error rate prediction method, models category-B human actions in the developed fault trees, and derives the human-induced initiating event frequency. The procedure, which is used in this approach, essentially involves system analysis, fault tree development, human error identification, screening, and quantification. The multi-unit loss of offsite power is used as an example accident situation which illustrates the application of the suggested method. Hence, the human-induced initiating event frequency for the loss of off-site power scenario for two units is derived. The application of this method would advance the efforts concerning multi-unit nuclear power plant (NPP) site risk analysis. 相似文献
8.
9.
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. 相似文献
10.
传统意义上核电厂数字化仪控系统主要依靠提升设备的可靠性来满足电厂安全目标。随着监管要求的逐步提高,在提升设备可靠性基础上,基于概率论技术的设计手段逐步成为核电厂安全设计新的研究方向。本文应用概率安全评价(PSA)技术,对典型电厂始发事件进行分析及研究,之后对仪控设计方案整体进行PSA建模,再将其置于电厂PSA模型中,通过定量评估分析,识别薄弱环节,给出优化改进措施。在此基础上提出了一套确保核电厂仪控系统满足整体安全目标的可靠性设计流程。 相似文献
11.
Conclusions The simple graphic method proposed for assessment of the operational safety of operating power-generating units of nuclear
power plants makes it possible to:
take into account the safety history of each observed power-generating unit by constructing a safety plot the rating values
over a fixed time interval;
analyze trends associated with the rating approaching the safety limit;
characterize the state of the first and second shielding level of the physical barriers;
reveal efficiently failures of normal operation which increase the probability of a serious accident and which require analysis
of accident precursors;
use as the control limit on the safety plot the toerance limitR
u, which makes it possible to regard, with high probability, any group of events for which the rating exceeds the limitR
u as a precursor of a serious accident, since most rating values must be concentrated in the interval [0,R
u]; and
use this approach, together with other methods for assessment of the operational safety of power-generating units, in the
practice of safety assessment for licensing nuclear power plants.
Russian Science and Technology Center Gosatomnadzora. Translated from Atomnaya énergiya, Vol. 76, No. 1, pp. 77–84, January,
1994. 相似文献
12.
风险指引的安全裕度是近十年来核电行业提出的新的安全理念。本文研究风险指引的安全裕度的计算框架和蒙特卡罗抽样方法下的风险指引的安全裕度定量化技术,并重点研究蒙特卡罗抽样方法下的核电站全厂断电(SBO)事故下的风险指引的安全裕度定量化技术。借鉴蒙特卡罗抽样次数估算方法和基于蒙特卡罗的可靠度计算方法,根据蒙特卡罗抽样方法下的风险指引的安全裕度的不确定度计算方法以及蒙特卡罗抽样次数的估算流程,计算得出在绝对误差小于001或相对误差小于5%时,两种不同误差方法选择时SBO事故的风险指引的安全裕度计算的抽样次数,并分别完成两个抽样次数下核燃料包壳失效概率均值和标准差定量化计算。计算结果表明,不同的抽样方法、不同的正态分布对核燃料包壳失效概率均值和标准差均有显著影响。 相似文献
13.
本文评述苏联从1990年7月1日起执行的核动力厂新的安全法规ОПБ-88。新的法规有许多新内容和新要求,本文评述其中有重大发展的原则,这些原则包括纵深防御原则;超设计基准事故;定量安全目标;概率分析要求;安全素养;质量保证;设备的核安全分级;对营运单位的要求;安全许可证制度;以及设计的基本安全原则。新法规的贯彻执行将对苏联核动力厂的安全提高到国际水平有重大推动作用。 相似文献
14.
核电站反应堆保护机柜失电缺省值分析研究 总被引:1,自引:0,他引:1
为了降低反应堆保护机柜(RPC)失电引入的安全风险,红沿河核电站开展了针对RPC失电的缺省值分析工作,论文在简要介绍红沿河核电站数字化仪控系统(DCS)平台的基础上,对RPC失电相关的缺省值分析范围进行了界定,通过实例对其分析原则进行了介绍,对其实现方式及应用进行了说明。该研究对提升DCS本身的可靠性、电站的安全水平和可用性有重要意义。 相似文献
15.
秦山三期重水堆核电站风险监测器研发进展 总被引:3,自引:1,他引:2
吴宜灿 胡丽琴 李亚洲 罗月童 袁润 王芳 王家群 顾晓慧 汪进 陈珊琦 王强龙 黄群英 汪建业 张振华 陈明军 曾春 宋明海 苏长松 彭晓春 张刚平 《核科学与工程》2011,31(1):68-74,85
核电站风险监测系统(Risk Monitor)可对核电站的运行风险进行实时监测和预测,是概率安全评价(PSA)技术的高级应用之一.FDS团队广泛调研了国际现有核电站风险监测系统的研发现状,深入研究了风险监测系统涉及的各种关键算法并探索了相关实现技术,基于前期自主研发的大型集成概率安全分析软件RiskA发展了通用核电站风... 相似文献
16.
Modeling of spurious activations in safety instrumented systems has been studied for over a decade. The spurious activation of a plant protection system in nuclear power plants (NPPs) leads to increased electricity generation cost. An in-depth view on spurious activation of digital plant protection systems of NPPs for human errors in maintenance tasks is presented in this paper. A new model which considers human errors in maintenance and periodic tests to predict component failure rates is presented. The model has been applied to OPR-1000 reactor protection system for quantification of spurious trip frequency by fault-tree analysis. The major causes of spurious activation in a nuclear reactor protection system are identified. A set of case studies has been performed with the variation of magnitudes of human errors probability and maintenance strategies, in which, the human errors in maintenance are found to significantly influence reactor spurious trip frequency. This study is expected to provide a useful mean to designers as well as maintainers of the digital reactor protection system to improve plant availability and safety. 相似文献
17.
重要厂用水系统是核电厂重要的安全系统之一,其失效概率通常由系统可靠性分析获得。而地震情况下设备的失效概率是地震动峰值加速度的函数,且地震的发生又具有随机性,目前概率安全评价中传统的故障树分析方法对此种情况缺乏足够的处理能力。本文采用蒙特卡罗模拟方法解决条件概率的问题,针对地震情况系统可靠性分析,提出了评价模型,并对核电厂重要厂用水系统进行了分析计算,得到地震情况下重要厂用水系统的年失效概率为1.46×10-4。计算结果与设备抗震性能数据符合,验证了分析模型的合理性。 相似文献
18.
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。 相似文献
19.
R. V. Arutyunyan L. A. Bolshov L. M. Vorobiova E. K. Khandogina S. M. Novikov T. A. Shashina N. S. Skvortsova M. I. Chubirko N. M. Pichuzhkina 《Atomic Energy》2010,109(2):137-142
Risk analysis applied to public health is used to evaluate the ecological safety of the territory of Voronezh Oblast, where
the Novovoronezh nuclear power plant is located and a new plant is under construction. It is found that for the population
of the territory of Voronezh the individual risk of developing cancer because of the contamination of air by chemical carcinogens
and the risk of death due to pollutants are 1000–10000 times higher than the risk of developing cancer from the additional
irradiation of the population of Novovoronezh associated with the operation of the nuclear power plant. The results obtained
showed that modern nuclear technologies have no effect on the public health as compared with sources of chemical risk. 相似文献
20.
核电站严重事故后果概率安全评价(PSA)是采用概率论的方法对核电站放射性后果进行分析,并定量给出放射性物质对核电站周围公众的健康效应影响。以国内某压水堆核电站为参考厂址,建立合适的场外后果分析模型。采用分层抽样方法对参考厂址1a的气象数据进行抽样,源项和释放特征等数据取自二级PSA的研究结果。利用事故后果评价程序对核电站严重事故后果进行计算,并用概率论方法对结果进行评估。通过计算将各事故和事故谱的场外个人剂量表示为CCDF曲线和总频率-剂量曲线,再用概率论方法得到不同距离处个人剂量超过指定剂量的条件概率;也可用此方法对确定烟羽应急计划区的安全准则中所描述的"大多数严重事故序列"进行量化。 相似文献