Seismic analysis — What goal?

The seismic analysis of nuclear components is characterized today by extensive engineering computer calculations in order to satisfy both the component standard codes such as ASME III as well as federal regulations and guides. The current nuclear seismic design procudure has envolved in a fragmented fashion and continues to change its elements as improved technology leads to changing standards and guides. The dominant trend is a monotonic increase in the overall conservatism with time causing a similar trend in costs of nuclear power plants. Ironically our improvements in the state of art are feeding a process which is eroding the very incentives that attracted us to nuclear power in the first place. This paper examines the cause of this process and suggests that what is needed is a realistic goal which appropriately addresses the overall incertainty of the seismic design process.     

压水堆核电厂换料物理启动试验优化研究和应用实践

核电厂设计和运行相关核安全法规、导则要求核电厂换料后必须进行物理启动试验。随着堆芯换料设计日趋成熟,试验程序和试验方法得到充分检验。为提升运行经济性,各核电厂设计和运行人员不同程度地开展了换料后物理启动试验优化的研究与实施。本文基于压水堆核电厂监管要求和核电厂运行要求分析,针对物理启动试验优化提出了定性评价、物理分析和试验验证的系统性论证方法,并以秦山核电厂320 MWe机组为例,进行了完善的研究与可行性论证。实施物理启动试验优化后,核电厂换料大修时间大幅缩短,相比以往可提前约2天进入满功率运行,显著提高了核电厂运行负荷因子,提升了运行经济性。     

Seismic qualification of systems, structures, equipment and components

The purpose of this paper is to give an overview of the various qualification procedures available to the vendors of nuclear power plants and equipment for hopefully achieving NRC (Nuclear Regulatory Commission) plant licensing and overall guaranteed safe operation. These procedures usually involve computer-aided analyses for large systems and structures, but trend toward shaking table tests for small equipment and components.The dynamic analysis and testing required for seismic qualification can be covered in a practical manner by reference to several pertinent Regulatory Guides and Standards. They have been issued by the NRC on specific subjects, but often represent a consensus of more general standards prepared by ASME, IEEE, ASCE, ANSI and NEMA. These documents cover such diverse subjects as (a) reactor site criteria, (b) seismic design limits and loading combinations, (c) system damping values, and (d) recommended vibration test practices.The author has been directly concerned with IEEE Std 344 on seismic qualification practices and has therefore included the latest ideas and suggestions for revising this document. In general, there has been a continuing escalation in the g-level of seismic requirements. This present overview indicates a need for R&D work and re-examination of published documents to counterbalance unwarranted conservatism.     

美国核电厂监督管理简况及其启示

美国核管会新的反应堆监督检查程序将监督管理力度集中在反应堆安全，辐射安全，电厂安全保卫3个领域，具体落实在初始事件，缓解系统，屏障完整性，应急准备，职业辐射安全，公众辐射安全，实体保卫这7项基本点上，并配合一系列的检查活动，达到更有效，客观，及时地评价核电厂运行安全水平的目的，结合我国核安全监督管理的实践及现状，我们应该吸取美国的经验，对有限的资源进行优化，把监管重点放在风险较大的问题上，以预防事故的发生，并适当地引入美国新的反应堆监督检查程序的一些思想及方法，发展和完善国家核安全局的监督管理模式。     

Technical and economical incentives behind strain limitation in piping

An inspection of conventional industrial plants subsequent to severe earthquakes showed that nothing or next to nothing had been damaged in these plants, although - assessed to the codes and standards for nuclear power plants - they were not designed to. Beside a lot of conservatism, an analysis revealed that structures subject to plasticization exhibit a much more favourable behaviour than anticipated on the basis of the design calculations. Thus, the introduction of the strain limitation approach promised service and cost advantages, above all for the expensive nuclear power plant piping.However, at the current state of the art it is possible to design piping sufficiently flexible to obtain satisfactory operational stresses. A cost analysis showed that - on the basis of today's dimensioning regulations, strain limitation is only economic in special cases.Strain limitation is nevertheless the adequate procedure in terms of engineering in those cases where today safeguarding against accidents is based on stresses of Stress Category D. It is therefore recommended to develop rules for admissible strains and economic methods for strain assessment. The efforts and expense should, however, be in line with the economic benefits.     

华龙一号调试首堆试验研究与设计

介绍了国家核安全局（NNSA）、国际原子能机构（IAEA）和美国核管会（NRC）对核电厂调试首堆试验的相关要求,结合核电厂运行经验反馈和同类型核电机组工程实践确定了华龙一号调试首堆试验的设计原则。同时,通过分析华龙一号核电机组采用的新设计理念和新设计特点,研究并确定了华龙一号调试首堆试验的项目。分析了各首堆试验项目的试验条件、试验内容和验收准则,以便于华龙一号调试首堆试验的开展。      

Japanese regulatory practices on licensing, inspection, operational supervision and related standards

Japanese view on the safety of nuclear power plants is based on the concept that the primary responsibility for securing safety lies on electric power companies, installers of reactors.Under this concept, the Ministry of International Trade and Industry (MITI), in the course of designing and construction, has been performed an examination of the basic design and the detailed design of nuclear power plants, and in each stage of construction, a pre-operational inspection process. In addition, MITI, in operating stage, has been made throughgoing investigations on the causes of troubles and incidents as well as accidents that may affect operation, forcing utilities to take measures to prevent recurrence, and implementing safety regulation based on the “preventive maintenance” including elaborate checkings and overhaulings at the periodical inspections conducted for a period of three to four months after every 12-month operation cycle under the laws and regulations.This paper discusses the current status of nuclear power development in Japan, safety regulatory systems, views on safety and future prospects of securing safety.     

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.     

秦山三期CANDU-6核电机组技术规格书的修改

描述了制定秦山三期CANDU 6核电机组技术规格书 (TS)的核安全法规依据 ,简要介绍了按照中国核安全法规的要求、参照美国NRC制定的压水堆核电站成熟的标准技术规格书的格式和应用加拿大CANDU机组采用的运行方针和政策所积累的成熟的运行经验 ,对AECL提交的TS进行修改的情况以及对TS中一些重要技术问题的修改内容 ,说明了修改后的TS基本满足了中国核安全法规的要求 ,可以在核电站的运行中使用     

Development of a consensus standard for verification and validation of nuclear system thermal-fluids software

With the resurgence of nuclear power and increased interest in advanced nuclear reactors as an option to supply abundant energy without the associated greenhouse gas emissions of the more conventional fossil fuel energy sources, there is a need to establish internationally recognized standards for the verification and validation (V&V) of software used to calculate the thermal–hydraulic behavior of advanced reactor designs for both normal operation and hypothetical accident conditions. To address this need, ASME (American Society of Mechanical Engineers) Standards and Certification has established the V&V 30 Committee, under the jurisdiction of the V&V Standards Committee, to develop a consensus standard for verification and validation of software used for design and analysis of advanced reactor systems. The initial focus of this committee will be on the V&V of system analysis and computational fluid dynamics (CFD) software for nuclear applications. To limit the scope of the effort, the committee will further limit its focus to software to be used in the licensing of High-Temperature Gas-Cooled Reactors. Although software verification will be an important and necessary part of the standard, much of the initial effort of the committee will be focused on the validation of existing software and new models that could be used in the licensing process. In this framework, the Standard should conform to Nuclear Regulatory Commission (NRC) and other regulatory practices, procedures and methods for licensing of nuclear power plants as embodied in the United States (U.S.) Code of Federal Regulations and other pertinent documents such as Regulatory Guide 1.203, “Transient and Accident Analysis Methods” and NUREG-0800, “NRC Standard Review Plan”. In addition, the Standard should be consistent with applicable sections of ASME NQA-1-2008 “Quality Assurance Requirements for Nuclear Facility Applications (QA)”. This paper describes the general requirements for the proposed V&V 30 Standard, which includes: (a) applicable NRC and other regulatory requirements for defining the operational and accident domain of a nuclear system that must be considered if the system is to be licensed, (b) the corresponding calculation domain of the software that should encompass the nuclear operational and accident domain to be used to study the system behavior for licensing purposes, (c) the definition of the scaled experimental data set required to provide the basis for validating the software, (d) the ensemble of experimental data sets required to populate the validation matrix for the software in question, and (e) the practices and procedures to be used when applying a validation standard. Although this initial effort will focus on software for licensing of High-Temperature Gas-Cooled Reactors, it is anticipated that the practices and procedures developed for this Standard can eventually be extended to other nuclear and non-nuclear applications.     

对滨河研究堆堆址设计基准洪水的探讨

通过对有关标准条文的理解以及对现有水文计算技术和方法、研究堆水灾事故危害程度的分析,结合工程设计实例,对滨河研究堆堆址设计基准洪水提出了明确的标准,即较高功率研究堆应该具有与核电厂相同的防洪基准.     

浮动核电站设计中的安全准则研究

浮动核电站作为船海工程与核电工程的结合,属于核能工程的新领域,国内尚缺少相应的安全设计准则。结合海洋核动力平台示范工程实际设计需求,基于对陆上压水堆核电厂、海上移动式平台、核动力舰船规范的分析,从浮动核电站总体设计、平台设计以及核安全3个层面分别提出了相应的安全设计准则。研究表明,浮动核电站的安全设计应围绕3项基本安全功能进行;平台设计应考虑布置、结构、辅助系统、电力、通信、消防6个因素;核安全设计应充分考虑其孤岛运行和海洋应用场景对核动力装置系统设备设计、运行的制约影响。      

EPR机组首堆试验项目选取与实施

为满足国家核安全局（NNSA）、国际原子能机构（IAEA）、美国核管会（NRC）、英国核安全局（ONR）和法国核安全局（ASN）对新型压水堆核电厂首堆试验的相关要求,本文通过分析欧洲先进压水堆（EPR）机组的新设计理念和新设计特点并结合已建成压水堆核电机组的工程实践,提出了采用控制变量确定选取原则并通过五步选取流程确定首堆试验项目的方法。实践证明,该方法不仅可确保首堆试验选取与确定工作顺利有效的开展,还能使新概念设计和具有新特性物项的性能得到充分和完整的验证,保证了新堆型核电站后期安全稳定地运行,该方法也适用于华龙一号在内的其他压水堆核电技术路线。      

浅谈我国核电厂重要厂用水系统换热器隔离阀门与放射性监测仪表配置优化

对于运行核电厂来说,重要厂用水系统与质量和安全密切相关.核电厂重要厂用水系统用于导出设备冷却水系统所传输的热量,将其输送到海水中,因此是核岛的最终热阱.本文描述了当前我国大部分核电厂重要厂用水系统换热器隔离阀门与放射性监测仪表的配置现状,分析了包括美国、法国以及国际原子能机构对于重要厂用水系统设计要求的相同点与不同点,...     

核电厂老化和寿命管理现状与进展

探讨了国内外核电厂老化和寿命管理方法、法规和实践。总体来说,IAEA的法规和导则指导性强,美国的法规和实践可操作性强,法国的法规和实践则较为系统化和标准化。针对这些特点,结合我国老化管理现状,给出了我国在进行核电厂老化和寿命管理实践以及建立相应的法规导则体系时应注意的问题。     

核电厂Profibus-DP技术设计准则及应用

现场总线技术在当今核电行业应用日益广泛,文章在解析现场总线技术特点和介绍Profibus-DP技术的基础上,依照第三代非能动核电厂中仪控系统分层要求,对核电厂Profibus-DP技术应用情况进行了阐述和分析;最后根据Profibus-DP总线技术的内部原理、特点和标准规范,提出了其在核电厂中基于工程应用的各项设计准则,进而满足项目的实际需求。     

美国联邦法规10 CFR 50附录Ⅰ的修订方案及对我国辐射防护审管的启示

本文详细介绍了美国核管理委员会(NRC)对轻水堆的设计目标基准的修订方案和策略,并在此基础上,考虑到我国核电厂址向内陆发展所致公众照射途径的变化,提出了需要明确核动力厂设计目标值的建议,以及应用现行辐射防护相关标准需要关注的问题:(1)ICRP第103号出版物从以前基于过程的实践和干预的方法发展为基于辐射照射情况性质(计划照射、应急照射和现存照射)的方法,应当注意区分不同的照射情况;(2)ICRP第103号出版物在数值上更新了当量剂量和有效剂量的辐射权重因子和组织权重因子,因此,实施剂量评估所采用的剂量转换因子也需要更新。     

Safety design consideration for HTGR coupling with hydrogen production plant

Safety requirements and design considerations are examined for a nuclear hydrogen production system that consists of High Temperature Gas-cooled Reactor (HTGR) and a hydrogen production plant by thermochemical water splitting iodine–sulfur process (IS process). Requirements in order to construct hydrogen production plants under conventional chemical plant regulation are identified in order to take into account a fundamental difference in safety philosophy between the nuclear facility and chemical plant and meet requests from the potential users of nuclear heat. In addition, safety requirements for the collocation of the nuclear facility and hydrogen production plant utilizing IS process (IS plant) are investigated. Furthermore, design considerations to comply with the requirements are suggested and the technical feasibility of the design considerations is evaluated. The evaluation results for a reference plant showed that safe distance determined by the chemical plant regulation against combustible gas and hazardous chemical leakages comply with the plant layout design. Furthermore, the results demonstrated the feasibility of IS plant construction under non-nuclear regulation by showing that the tritium concentration in IS plant can be maintained below the regulation limit and reactor normal operation can be achieved during abnormal conditions in the IS plant. These results clarified that design considerations suggested for coupling the IS plant to HTGR are reasonably practicable. The proposed criteria can be used not only for coupling hydrogen production plants but also for other chemical plants such as steam reforming plants, etc.     

评苏联核动力厂安全总则ОПБ-88

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。     

福岛核事故后核安全改进行动及安全要求研究

2011年“3·11”日本福岛第一核电厂严重核事故给世界核工业界造成了巨大影响。本文总结了从福岛核事故中汲取的教训,介绍了主要核电国家,如美国、日本、法国以及中国在福岛核事故后十年来实施的一系列核安全改进行动和核安全法规标准修订。阐述了核安全要求和核安全理念在中国的实施现状及实践,包括实际消除早期或大量放射性释放、事故工况划分、纵深防御概念、移动设施配置等,对后续核安全发展方向进行探讨并提出建议。