AP1000核电厂反应堆冷却剂系统泄漏分析

第三代AP1000非能动核电厂的主要特征是采用非能动安全原理,使核电厂的系统、设备、构筑物大幅度简化,安全性、可靠性、经济性大幅度提高,以满足美国先进轻水堆业主要求文件的基本要求。本文针对美国业主要求文件(Utility Requirements Document,URD)第三卷第五章《专设安全系统》中对非能动先进轻水堆核电厂反应堆冷却剂系统压力控制功能的要求:在很小的反应堆冷却剂系统(Reactor Coolant System,RCS)净泄漏率(不大于2.27 m3·h-1)条件下,具有足够的系统冷却剂装量及补水能力,以保证在8 h(28 800 s)内不会触发自动降压系统而进行计算分析,本分析采用安全分析报告小破口失水事故(Loss of coolant accident,LOCA)分析采用的NOTRUMP程序,分析结果表明AP1000核电厂可满足上述美国URD要求。     

非能动核电厂非安全系统实施监管时的若干问题

美国核管会(U.S.Nuclear Regulatory Commission,简称NRC)在对非能动核电厂AP600和AP1000进行安全审查过程中,提出对非安全系统监管(Regulatory Treatment of Non-Safety System,简称RTNSS)的安全要求,这是NRC对非能动核电厂监管的重要特点之一。本文介绍了NRC提出RTNSS的历程、监管要求和实施程序,并研究了我国非能动核电厂的非安全相关构筑物、系统和部件的监管方面可能存在的问题,最后对于RTNSS相关安全要求与我国最新发布的核安全法规的一致性,作了评估说明。     

AP1000核电厂对安全有贡献的非安全物项质量保证要求

在AP1000核电厂中,由于全面采用非能动安全系统来缓解事故,除有限的提供安全相关隔离功能(如安全壳隔离)的系统外,所有能动系统均设计为非安全级的。但这些非安全级设备、系统和构筑物有提供电厂安全纵深防御和补充反应堆冷却剂及导出衰变热的功能。因此,这些需要额外监管的对安全有贡献的非安全物项都应满足10 CFR 50中附录B关于QA的18项要求,通过将这些要求与ISO 9001:2008《质量管理体系要求》进行对比,得出了针对安全有重要贡献的非安全物项,除一般的非安全级别物项的ISO 9001外附加的核质量保证要求,并提出了需要业主和供应商关注的事项和采取的措施。     

对AP1000核电厂简化应急计划的探讨

本文介绍了AP1000核电厂的非能动安全系统设计特性和美国核管理委员会对先进轻水堆简化应急计划的见解。针对AP1000的事故预防和缓解的主要特点,对其场外应急计划的简化问题进行探讨,提出了简化应急计划需要关注的三个主要问题：（1）缓解应急响应的紧迫性,（2）适当缩小应急计划区,（3）修订场外应急防护措施。     

AP600的设计证明合格、准备建造

【美国《核新闻》2000年9月刊报道】 AP 600是一个设计相对简单、非常安全的核电厂,它在经济性上能与其它发电系统相竞争。AP 600是由西屋公司、西屋公司的分包商和捐款人、美国能源部(DOE)、电力研究设计院(EPRI)共同开发的600 MW先进压力轻水堆核电厂。 该反应堆起主要作用的非能动式安全系统依赖的仅仅是自然力诸如有重力、天然循环、对流、蒸发和冷凝,而不是靠交流电 18源和电动部件。非能动系统和使用有经验基础的部件大大增加了安全性、加大了公众对核电的接受程度,并且使申请许可证变得容易了,它们还简化了电厂系统、设备、以…     

非能动系统可靠性分析方法在AP1000非能动ERVC系统中的应用

由于核电厂安全水平要求的逐渐提高,越来越多的非能动系统被用于先进反应堆堆型中,但对这些非能动系统可靠性评价的工作还处于初级阶段。本文根据非能动系统可靠性评价流程,通过RELAP5热工水力学程序模拟非能动系统物理过程,对AP1000反应堆压力容器外部冷却（ERVC）系统进行了可靠性评价。通过计算得到了压力容器下封头温度等参数的累积密度分布曲线,根据不同的成功准则即可获得AP1000 ERVC系统的可靠性。该非能动系统可靠性评价结果可用于核电厂PSA模型中,以更好地指导核电厂设计及提高核电厂的安全性。     

AP1000非能动安全系统初步应用研究

对于堆芯失去冷却能力和安全壳升温升压事故，AP1000非能动安全系统在设计上仅凭重力和气体压力等非能动源来实现缓解各种事故功能。本报告介绍了AP1000非能动安全系统各分系统和子系统的设计以及相关特点。在此基础上，对将AP1000非能动安全系统应用于环路式先进堆进行了初步探讨和研究。     

向商业化迈进的AP600

【《欧洲核能世界浏览》1990年11-12月号报道】经过4年的开发,美国西屋公司设计的先进非能动安全600Mw。反应堆(AP 600),达到了美国能源部(DOE)要求的目标:一种简化、固有安全和新型的先进轻水堆。     

概率截断值对先进轻水堆核电厂应急计划区划分的影响

1997年,美国核管会（NRC）在对被动与改进型先进轻水堆的应急计划进行评估后指出,在现有的技术框架下先进轻水堆的应急计划应当保持不变,但也表明如果考虑到严重事故发生概率更低,事故的延迟时间更长,则有可能简化对先进轻水堆的应急计划要求,减小应急计划区。这意味着,如果在事故选择时不考虑低于某一概率截断值的事故,则有可能对先进轻水堆核电厂应急计划区的划分产生较大的影响。本文以AP1000核电机组为例,参考美国NUREG-0396的方法,使用MACCS程序对选取不同事故概率截断值可能产生的影响进行研究。研究结果表明,只有当概率截断值高于某些相对概率较大、而后果较为严重的事故的发生概率时,才会对先进轻水堆应急计划区的划分产生较大影响。     

AP1000核电厂主给水管道断裂事故瞬态特性分析

AP1000是目前国际上典型的“三代”非能动核电厂,基于最佳估算程序RELAP5/MOD3.3,对AP1000核电厂系统进行了详细的建模分析,获得了主给水管道断裂事故下AP1000核电厂关键参数的瞬态特性和非能动系统响应特性。结果表明,事故过程中一、二回路的压力和温度呈现波动变化,一回路压力最大值为17.13 MPa,低于设计压力的91%,主蒸汽系统的压力也低于设计值的91%,满足验收准则的要求。     

Further development of the load coefficient method, lcm, for rational seismic design of nuclear facilities

The purpose of this paper is to present the results of a study conducted to compare the results of the Load Coefficient Method, LCM, proposed for seismic load determination, to modal analysis and the equivalent static load methods as defined in Section 3.7.2 of the U.S. Nuclear Regulatory Commission Standard Review Plan. The comparison is conducted using a number of nuclear power plant piping systems which used response spectra modal analysis input in their original design.The real piping systems studied are considered to be representative of ASME Section III nuclear Class 2 and 3 piping systems required to be designed to resist currently defined seismic loadings. Section 2 of this paper provides numerical comparisons of the application of LCM, Response Spectrum and Equivalent Static Load Methods.     

核电厂设备抗震设计标准功率谱密度的生成方法

对于核电厂设备抗震设计的输入地震波,通常要求其同时包络目标反应谱（RRS）和标准功率谱密度（PSD）,然而目前国内外对标准PSD缺少统一的算法。在美国核管会标准审查大纲（SRP）3.7.1建议的标准PSD生成方法基础上,优化了迭代过程,提出了一个改进的标准PSD合成方法,并在2个核电设备RRS算例上实现了该方法。结果显示改进的标准PSD生成方法与RRS匹配程度较高,同时计算快速、简便,收敛精度与基于随机振动理论方法计算的结果相似,此法可以作为核电厂设备抗震设计输入人工地震波的标准PSD检验依据。      

湖南某核技术利用项目辐射防护设计与辐射环境监测验证

为了满足《电离辐射防护与辐射源安全基本标准》(GB 18871—2002)的要求,对湖南某核技术利用项目进行了辐射防护屏蔽设计并在设计的基础上进行了辐射环境监测验证。监测结果表明,该项目辐射防护设计符合相关标准或规范要求。     

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.     

核电厂仪表与控制系统设计在英国GDA过程中的若干问题分析

为了承接英国核电项目建设,核电项目的核安全审查和执照申请须首先通过英国核监管办公室（ONR）的通用设计审查（GDA）。GDA主要基于ONR发布的审查导则—安全评估原则/技术评估指南（SAP/TAG）及相关的国际原子能机构（IAEA）、西欧核安全监管协会（WENRA）导则和?国际电工委员会（IEC）标准开展,通过分析GDA对仪表与控制（I&C）系统设计主要关注问题及相关导则和标准要求,并结合英国欧洲压水堆（UK-EPR）和英国先进非能动压水堆（UK-AP1000）的审查经验,提出了I&C系统总体设计应对策略。可为后续项目通过GDA提供参考借鉴。      

Seismic risk analysis and decisions for nuclear power plants

Seismic risk analysis and associated sensitivity studies constitute a part of the Seismic Safety Margins Research Program being conducted by the Lawrence Livermore National Laboratory for the US Nuclear Regulatory Commission. Although seismic risks are an important contributor to the total nuclear risk, the occurrence of earthquake-related seismic phenomena is low. Safety decisions involving seismic hazards must be made, however. This paper briefly described several categories of decisions that can be made using seismic risk analysis. While risk analysis does not provide all the information required for these decisions, it is a useful tool in that it provides additional information for the decision-making process. We anticipate a growing interest in the use of seismic risk analysis in nuclear safety evaluations.     

利用HHT产生与多阻尼楼板谱匹配的地震波

美国原子能管理委员会（USNRC）规范规定了用于核电厂抗震分析和设计的地震波要求。在抗震分析和设计中，采用的地震波可与多阻尼目标反应谱匹配，也可与单阻尼目标反应谱匹配。然而，在对核电设备和部件进行动力时程分析时，则需要与多阻尼目标楼板谱匹配的地震波。基于此问题，提出利用希尔伯特-黄变换（HHT）方法，通过修改种子地震波的频率和振幅信息，使之与多阻尼目标楼板谱匹配，且完全符合USNRC规范的匹配标准，从而为核电设备和部件的地震安全评估提供合适的地震激励。      

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.     

Licensing documentation and licensing process for dismantling and decontamination projects in Lithuania

One of the main tasks of any decommissioning project is the licensing process, which allows implementation of developed strategies in the real nuclear power plant. The Lithuanian laws on nuclear energy and radioactive waste management require that dismantling and decontamination projects shall be licensed by the Lithuanian State Nuclear Power Safety Inspectorate and other Authorities. Licensing is an inseparable part of the Lithuanian regulatory and supervisory system for safety of nuclear facilities. The licensing process starts when the nuclear power plant submits the first licensing document(s) to the Authorities. Usually the licensing documents cover Basic Design, Safety Assessment, Environmental Impact Assessment and Civil Design reports. Safety Assessment Report is a major document, which justifies that the proposed activities will be implemented in compliance with design and regulatory requirements. Licensing process is completed when all the licensing documents are approved by the Authorities and authorization to start dismantling and decontamination activities is received by the nuclear power plant.Current paper discusses the main steps of the licensing process adopted for the first dismantling and decontamination projects at Ignalina Nuclear Power Plant and provides examples of safety assessment in the case of bounding initiating events that can be caused by dismantling and decommissioning activities.     

Availability of nuclear standards information

A brief summary of the nuclear standards organizations in the United States and a discussion as to how such standards may be located are presented. Some 25 non-governmental organizations which may prepare nuclear standards are identified. The most important organization is the American National Standards Institute which not only prepares such standards but co-ordinates the standards work of others. The increasing activity of several government agencies, specifically the Atomic Energy Commission, Environmental Protection Agency and the National Bureau of Standards is also noted. The annual Compilation of Nuclear Standards (Vols. 1 and 2) prepared by the Nuclear Safety Information Center is the best single source of such information, although many other sources — including relevant information centers, as well as published documents — are identified.