Toward risk-based control of nuclear power plant configurations

This paper presents an analysis of risks associated with component outage configurations during power operation of a nuclear power plant and discusses approaches and strategies for developing a risk-based configuration control system. A configuration, as used here, is a set of component states. The objective of risk-based configuration control is to detect and control plant configurations using a risk perspective.The configuration contributions to core-melt frequency and core-melt probability are studied for two plants. Some equipment configurations can cause large core-melt frequency increases and there are a number of such configurations that are not currently controlled by technical specifications. However, the expected frequency of occurrences of the impacting configurations is small and the actual core-melt probability contributions are also generally small. Effective strategies and criteria for controlling configuration risks are presented. Such control strategies take into consideration the risks associated with configurations, the nature and characteristics of the configuration risks, and also the practical considerations such as adequate repair times and/or options to transfer to low risk configurations. Alternative types of criteria are discussed that are not overly restrictive to result in unnecessary plant shutdown, but rather motivate effective test and maintenance practices that control risk-significant configurations to allow continued operation with an adequate margin to meet challenges to safety.     

Probabilistic risk assessment in nuclear power plant regulation

A specific program is recommended to utilize more effectively probabilistic risk assessment in nuclear power plant regulation. It is based upon the engineering insights from the Reactor Safety Study (WASH-1400) and some follow-on risk assessment research by USNRC. The Three Mile Island accident is briefly discussed from a risk viewpoint to illustrate a weakness in current practice. The development of a probabilistic safety goal is recommended with some suggestions on underlying principles. Some ongoing work on risk perception and the draft probabilistic safety goal being reviewed in Canada is described. Some suggestions are offered on further risk assessment research. Finally, some recent U.S. Nuclear Regulatory Commission actions are described.     

Use of PRA and safety goals in nuclear power plant regulation

Probabilistic risk assessments (PRAs) have been performed on a number of nuclear power plants, both by the NRC and industry. The NRC has used risk perspectives gained from PRAs, both in an absolute as well as a relative sense, as an aid in making decisions on plant-specific as well as generic safety issues. However, substantial uncertainties pervade present-day risk assessments, which makes the application of the results of such analyses difficult at best in the regulation of nuclear power. Nonetheless, the Commission approved in January 1983 a policy statement on safety goals for public comment and a two year evaluation period. These safety goals include quantitative design objectives which could serve in the future as risk benchmarks for use by the NRC as part of the decision making process on matters relating to nuclear safety. While the Commission's policy statement explicitly excludes the safety goals from use both in licensing cases and in regulation for the two year evaluation period, PRA will be used generically and on a plant-specific basis more and more to assess the importance of new safety issues, prioritize resources within the agency, and test the adequacy of (or in some instances the need for) NRC's regulations.     

Current steel forgings and their properties for steam generator of nuclear power plant

Current steel forgings used for the construction of steam generators (SGs) for a nuclear power plant (NPP) were introduced from the following three (3) features. (1) Integral type steel forgings such as (i) primary head integrated with nozzles, manways and supports, (ii) steam drum head integrated with nozzle and handhole, (iii) conical shell integrated with cylindrical sections and handholes, have been developed to enhance the structural integrity of the component and to make fabrication and inspection including in-service inspection easier. (2) The high strength steel such as SA508 Cl.3a has been adopted to decrease the weight of components, resulting in enhancement of aseismatic properties of SG. (3) Low Si SA508 Cl.3 steel with the addition of Al was investigated and was applied to a heavy thick tube sheet forging, to minimize the macro- and micro-segregations in the ingot.     

Aspects on decommissioning of the Greifswald nuclear power plant

The most important part of the nuclear facilities in the former German Democratic Republic is situated at Greifswald near the Baltic Sea. Shortly after reunification of the Germany states, a decision was taken to decommission all Russian pressurised water reactors. The dismantling of them will be the biggest decommissioning project of series reactors world-wide.The low level of radioactive contamination, especially in the primary circuit, makes recycling of much material after decontamination possible.     

核电机组电气设计分析

近年来,我国核电事业得到快速发展,一大批核电项目陆续开工建设,其中大部分为引进技术的二代改进型和三代核电机组。由于技术输出国的标准规范与我国现有的核电设计标准不一致,以及考虑厂址适应性等问题,我国对引进的核电机组存在逐步消化、吸收并改进的过程。本文对核电机组的电气设计进行探讨,希望对今后同类工程具有参考作用。     

核电厂在役检查探讨

在役检查是指核电厂承压边界重要核安全相关部件的定期检查,是核电厂整个寿期内的重要活动之一。本文从核电厂营运单位的角度论述了核电厂在役检查的组织体系、文件体系、不同阶段的重点工作和在役检查人员配置等。     

秦山核电厂的老化及寿期管理

介绍了核电厂老化及寿期管理的相关背景以及国外核电厂在延寿方面采取的两种主要模式,即执照更新模式和PSR模式。结合目前秦山核电厂开展的主要老化管理工作,提出了秦山核电厂延寿的设想,并对核电厂寿期管理中存在的问题进行了讨论。     

核电厂的安全文化

以DNMC(大亚湾核电站)管理干部安全文化培训教材为基础,简要地说明核电厂的安全、设计、管理、组织文化、安全文化这一系列概念的基本内容和相互之间的关系。重点解释了安全文化体系的构成要素;政策层、管理层和员工个人三层承诺的基本要求;并对典型的安全文化事项,如透明的文化、习惯性违规等进行了分析和论断。人三层承诺的基本要求;并对典型的安全文化事项,如透明的文化、习惯性违规等进行了分析和论断。     

核电运行人才的培养

以秦山第二核电厂为例,介绍了系统性培训方法SAT和运行人才培养体系、培训管理组织机构和责任、核电厂运行人才培养的逐级培训体系、在岗培训制度、运行处的岗位授权制度、人才培养的激励机制等,为核电厂运行人才培养提供参考。     

Coolant stratification in nuclear power plant pipes

Translated from Atomnaya Énergiya, Vol. 65, No. 1, pp. 68–69, July, 1988     

核电厂运行人员防人因失误管理

要确保核电厂安全稳定运行,做好设备和人员管理是两项重要工作,本文结合秦山第三核电有限公司在防止人因失误方面的一些尝试,探讨了核电厂运行过程中如何有效控制和防止人因失误方面的思路和构想。     

Irradiation experiments in the testing nuclear power plant VAK

Within the German Research Programme “Integrity of Components” the first two capsules were irradiated in the Testing Nuclear Power Reactor VAK. The materials are of the 22 NiMoCr 3 7 and 20 MnMoNi 5 5 types and represent the lower bound of the base material regarding upper shelf energy and chemical composition (Cu, S, P), as well as a state of material which does not meet both chemical and toughness requirements (low upper shelf test melt). Tensile, Charpy, drop-weight, and fracture mechanics specimens were irradiated up to a range of 1.5 to 2 × 10¹⁹ cm⁻² (E > 1 MeV). Despite the materials being at or beyond the specification limits, the results show irradiation sensitivity which can be predicted from the US Reg. Guide Trend Curves (1.99) and KWU Trend Curves in a conservative manner. The procedure to determine the adjusted reference temperature RT_NDT (adj.) on the basis of ΔT_41J (following ASTM E 185) could also be confirmed as conservative by comparing the different criteria derived from Charpy and drop weight tests in the unirradiated and irradiated condition.The results of fracture mechanics testing in the linear elastic range show a remarkable temperature margin to the K_Ic-curve of ASME XI.Prestrained compact tension specimens CT 40 mm made of 22 NiMoCr 3 7 material with an upper shelf energy of approx. 100 J were wedge loaded in a range up to 30 MPa m and exposed to the water environment during radiation. Macroscopic examination gave no indications of stress corrosion cracking.From tests of these specimens in the linear elastic range, a fracture toughness K_Ic^*, which was not affected by the prestrain and environment history, was found depending only on the overload applied during the prestraining procedure.     

核电站预防性维修的重要性

日本福岛核事故后,我国核电站认真吸取教训,在总结经验教训的基础上,结合核电站设计和运行的关系说明了核电站预防性维修的重要性。文章介绍了预防性维修的内容和制定预防性维修大纲的基本技术方法,阐述了核电站预防性维修的方法及其所起到的安全保障作用。