陆上小型堆烟羽应急计划区划分初步研究

小型堆烟羽应急计划区(EPZ)大小作为其市场推广和应用的重要外部约束条件之一,意味着制定合适的划分准则和确立其大小范围具有十分重大的意义。结合现行大堆烟羽应急计划区(EPZ)的划分准则,本文分析了国内外小型堆烟羽应急计划区(EPZ)划分方法,提出陆上小型堆采用剂量/距离的划分方法。在研究中,基于MAAP程序对某小型堆进行建模计算,从中得出了较为合理的机理性应急源项;并通过大气扩散计算软件MACCS程序进行烟羽应急计划区(EPZ)计算;同时对厂址差异进行相关的灵敏性分析。     

小型堆应急计划区划分研究

利用一体化严重事故分析程序(MELCOR)对小型堆断电事故进行仿真分析,并将结果作为大气扩散计算软件MACCS的输入,计算分析某滨海地区放射性后果。结合建立的小型堆应急计划区划分准则,通过计算确定适用于小型堆的应急计划区大小。     

核电厂烟羽应急计划区划分方法研究

结合不同类型反应堆的安全特性,对不同的烟羽应急计划区(PEPZ)划分方法进行对比分析,然后依据反应堆类型进行系统归类,提出不同堆型所适用的烟羽应急计划区划分方法,最后以我国的模块式高温气冷堆示范电厂(HTR-PM)为例进行划分方法的初步应用。初步研究结果表明,HTR-PM在厂址边界处满足烟羽应急计划区的划分准则,相对于目前的大型轻水堆,可以明显减小其烟羽应急计划区。     

某压水堆烟羽应急计划区划分初步研究

结合某小型压水堆发生核事故的厂址特点,依据确定的辐射后果评价技术,估算了某200MW压水堆发生典型超设计基准事故周围公众的个人有效剂量和甲状腺剂量。借鉴国内、外现有应急计划区划分的相关经验,结合厂址区域的实际情况,将剂量估算结果与标准规定的干预水平值进行比对,初步确定了某小型压水堆的烟羽应急计划区范围。通过本研究,为小型堆烟羽应急计划区的划分提供了一种有效的方法,为核事故应急决策提供了科学依据。     

小型模块化压水反应堆应急计划区划分探讨

文章概述了小型模块化压水反应堆(以下简称小堆)厂址适宜性要求、国际上现行应急计划区的分区和大小,介绍了反应堆应急计划区的确定方法,同时对我国小堆应急计划区的划分提出可能的建议。     

核电厂烟羽应急计划区划分探讨

应急计划区划分是核电厂应急准备工作的一项重要内容。基于国家推荐的划分原则,分析某AP1000核电机组烟羽应急计划区计算方法及计算结果,并从电厂安全、事故后果、人口因素等方面讨论实际划分中遇到的问题,提出对后续机组烟羽应急计划区划分的建议。     

小型核反应堆应急计划区的研究

适用于大型核反应堆的应急计划区划分方法不能完全适用于小型核反应堆。应急计划区的划分需充分考虑公众可能受到的预期剂量和可防止剂量,将所得到的剂量数据与规定的干预水平相比较,确定应急计划区范围。以烟羽应急计划区范围外的预期剂量不超过10 m Sv为例,利用MACCS软件计算在指定应急计划区范围的限制条件下,对释放源项水平的要求,详细说明应急计划区的划分及计算过程。     

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。     

确定核动力厂应急计划区大小的准则和方法

在详细介绍了一些国家或地区确定核电厂应急计划区的准则和方法的基础上,将确定应急计划区的方法主要分为确定论和概率论两种,并进行了评论和比较。分析了应用美国确定应急计划区大小的准则时可能遇到的困难,提出了我国确定核电厂应急计划区大小准则的建议。最后用广东大亚湾核电站概率安全评价第三级事故后果的计算说明,应用本文所建议的准则,大亚湾核电站不论是否安装砂堆过滤器,采用10km作为烟羽应急计划区半径均是适宜的。     

小型堆选址相关标准的适用性讨论

小型堆由于具有良好的安全性能而日益受到国内外关注。小型堆的设计目标是尽可能地靠近用户，但是其选址相关标准的制定尚处于起步阶段。本文采用假设的小型堆选址假想事故对内陆厂址和滨海厂址的非居住区和规划限制区半径进行粗略估算，并对现行标准中关于确定非居住区、规划限制区、应急计划区半径，以及放射性三废排放要求的具体规定是否适用提出了讨论。     

Risk-informed extension of the inservice inspection interval for pressurized water reactor vessel from 10 to 20 years

The Pressurized Water Reactor Owners Group (PWROG) methodology for extending the inservice inspection interval for reactor vessel welds in pressurized water reactors is described in technical report, WCAP-16168-NP, Revision 1. This report presents a risk-informed basis for extending the interval between inspections from the current interval of 10-20 years. This paper discusses the background of the PWROG inservice inspection interval extension methodology and the results of pilot plant studies on typical Westinghouse, Combustion Engineering, and Babcock and Wilcox plant designs. These pilot plant studies showed that the change in risk associated with the proposed inspection interval extension was within the guidelines specified in the United States Nuclear Regulatory Commission (NRC) Regulatory Guide 1.174 for an acceptably small change in risk. This paper also discusses the application of the extended interval to other pressurized water reactors and provides examples of the application.     

核电厂风险指引管理研究

风险指引管理是确定论与概率论方法相结合的一种新的安全管理模式.为了促进我国这项工作的开展,有必要对国内外的相关法规、标准和实践进行全面和系统的研究.本文介绍了核电厂风险指引决策的基本原则、方法与风险接受准则,讨论了风险指引决策对概率安全评价(PSA)的要求,并对我国核电厂采用风险指引管理提出了建议.     

AP1000核电厂烟羽应急计划区划分初步研究

应急计划区的划分是核电厂应急计划制定中的重要内容之一。第三代核电机组AP1000应急计划区的划分研究,对其应急计划和应急准备工作具有十分重要的意义。首先介绍了应急计划区划分的一般方法以及我国相关法规的要求;然后以某滨海厂址为例,根据现阶段AP1000事故源项的研究结果和该厂址的气象观测数据,使用PAVAN和MACCS程序对相应的事故后果进行计算;最后结合相应的准则对计算结果进行分析和评价。初步的研究结果表明,AP1000核电厂取半径3km范围作为烟羽应急计划区的内区、半径7km范围作为烟羽应急计划区的外区是合适的。     

风险指引的CPR1000核电厂LBLOCA分析方法初步研究

随着核电厂安全分析方法的不断发展,结合传统确定论分析与概率风险评价（PSA）的风险指引型安全分析方法逐渐引起安审当局和核电业主的广泛关注。本文基于国际上风险指引型分析方法在其他领域的应用现状,提出了风险指引的大破口失水事故（LBLOCA）分析方法,并重新评估了CPR1000核电厂的堆芯燃料包壳峰值温度（PCT）裕量。在PSA分析中,识别并量化了LBLOCA发生后可能发生的162个事件序列,并采用确定论现实分析方法（DRM）对筛选出的18个概率较大的事件序列进行了计算分析。然后通过期望值评估法和特定序列覆盖法对LBLOCA的PCT裕量进行了评估。结果表明,本文方法下LBLOCA的PCT裕量约为36~55 ℃,相比于传统的DRM裕量提升了16~35 ℃。     

Risk-informed inservice inspection program

The Nuclear Regulatory Commission (NRC) has developed draft guidance for power reactor licenses on acceptable methods for using probabilistic risk assessment (PRA) information and insights in support of plant-specific applications to change the current licensing basis (CLB) for inservice inspection (ISI) of piping. This process is also known as risk-informed inservice inspection programs (RI-ISI). The risk-informed inservice inspection process for operating nuclear power plants provides an alternative method for selecting and categorizing piping components that are inspected for the purposes of meeting the requirements of ASME Section XI. A RI-ISI approach will incorporate probabilistic techniques to help define the scope, type and frequency of inservice inspection. The risk-informed process may support a decrease in the number of inspection and inspection intervals but will also identify areas where increased resources should be allocated to enhance safety. The approach discussed in this paper follows the method developed by NRC staff.     

Flexible conversion ratio fast reactors: Overview

Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. The performance achievable by the unity conversion ratio cores of these reactors was compared to an existing supercritical carbon dioxide-cooled (S-CO₂) fast reactor design and an uprated version of an existing sodium-cooled fast reactor. All concepts have cores rated at 2400 MWt. The cores of the liquid-cooled reactors are placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchangers (IHXs) coupling a primary coolant to a compact and efficient supercritical CO₂ Brayton cycle power conversion system. The S-CO₂ reactor is directly coupled to the S-CO₂ Brayton cycle power conversion system. Decay heat is removed passively using an enhanced reactor vessel auxiliary cooling system (RVACS) and a passive secondary auxiliary cooling system (PSACS). The selection of the water-cooled versus air-cooled heat sink for the PSACS as well as the analysis of the probability that the PSACS may fail to complete its mission was performed using risk-informed methodology. In addition to these features, all reactors were designed to be self-controllable. Further, the liquid-cooled reactors utilized common passive decay heat removal systems whereas the S-CO₂ uses reliable battery powered blowers for post-LOCA decay heat removal to provide flow in well defined regimes and to accommodate inadvertent bypass flows. The multiple design limits and challenges which constrained the execution of the four fast reactor concepts are elaborated. These include principally neutronics and materials challenges. The neutronic challenges are the large positive coolant reactivity feedback, small fuel temperature coefficient, small effective delayed neutron fraction, large reactivity swing and the transition between different conversion ratio cores. The burnup, temperature and fluence constraints on fuels, cladding and vessel materials are elaborated for three categories of material - materials currently available, available on a relatively short time scale and available only with significant development effort. The selected fuels are the metallic U-TRU-Zr (10% Zr) for unity conversion ratio and TRU-Zr (75% Zr) for zero conversion ratio. The principal selected cladding and vessel materials are HT-9 and A533 or A508, respectively, for current availability, T-91 and 9Cr-1Mo steel for relatively short-term availability and oxide dispersion strengthened ferritic steel (ODS) available only with significant development.     

压水堆核电站应急源项的选定和应急计划区的划分

从我国已建压水准核电站的具体实例出发，结合国外的先进经验，对核电站应急源项的选定及应急计划区的划分工作进行研究和总结，对新建核电站的应急源项的选定和应急计划区的划分提出一些个人的看法。     

国际原子能机构应急计划区的发展研究

为了保证在发生核事故时能及时有效地采取防护行动以保护公众和环境,需要事先在核设施周围建立应急计划区,并在该区域内需要制定应急计划并做好应急准备。本文系统地介绍了国际原子能机构（IAEA）有关应急计划区发展历程及预防行动区（PAZ）和紧急防护行动计划区（UPZ）的特征,详细阐述了在福岛核事故后IAEA有关应急计划区的最新变化发展,重点分析了Ⅰ类威胁核设施的PAZ和UPZ的变化,为完善我国核电厂核应急工作提供启示与参考。