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。     

核动力厂选址假想事故源项的探讨

介绍了美国核审管部门对核动力厂选址假想事故源项确定的历史演变以及相关研究进展。推荐了一套核动力厂选址假想事故源项,以及确定非居住区边界的评价分析方法和假设条件。计算了M310和AP1000核动力厂厂址的非居住区的最大径向距离。计算结果表明:(1)对于M310和AP1000核动力厂,采用确定论方法估算大气扩散因子,安全壳泄漏率(体积分数)取0.3%/d,非居住区最大径向距离分别不超过1.4 km和1.9 km;(2)在厂址选址阶段,考虑核动力厂的技术路线尚未明确或采用新技术路线,采用"不考虑安全壳喷淋系统等能动安全设施对核素的去除作用"的选址假想事故源项来确定厂址非居住区的边界基本上是可行的。     

美国核电厂风险评估的安全效益（三）

【美国《核新闻》2003年1月刊报道】 委托监管应用 美国核管会(NRC)在监管过程中积累了大量风险知识,并根据从实施概率风险分析(PRA)中获得的这些知识对监管作出了诸多改进。本章将对一些比较重要的风险通报应用进行概要介绍。 ATWS(未能紧急停堆的预期瞬态)规则 ATWS是反应堆事故保护停堆作用失败后的停堆事件。这个不太可能发生的事件将引起反应堆系统的高压,同时产生远远超出反应堆停堆散热能力的衰变热,因此反应堆必须停堆并保持在次临界状态。NRC在1983年发布了ATWS规则(10 CFR 50.62),通过以下措施降低ATWS风险: 降低预…     

ACP100非居住区边界剂量环境影响评估

介绍了ACP100发生选址假想事故后,对非居住区边界剂量环境影响的评估方法。基于RG1.183建立了ACP100选址假想事故后释放到环境的源项计算模型,并结合厂址的气象条件采用PAVAN程序得到非居住区边界剂量,研究了裂变产物释放方式、照射时间、自然去除、厂址气象条件和源项对非居住区边界剂量的影响。结果表明:上述因素对剂量均有较大的影响,确定ACP100非居住区的边界应考虑这些因素。     

美国核管会认为戴维斯—贝瑟核电厂事故本是可以预防的

【美国《核废物新闻》2002年10月10日刊报道】美国核管会(NRC)特别工作组得出结论:2002年3月,戴维斯-贝瑟核电厂反应堆压力容器封头因腐蚀而产生孔洞的事故(详见本刊2002年第3期和第7期相关文章)本来是可以预防的。目前该反应堆仍处于停堆状态,等待更换反应堆压力容器封头,并接受有关安全和实绩的综合审查。NRC在该电厂发生事故后,组建了一个特别工作组,以对NRC与戴维斯-贝瑟核电厂损坏有关的所有监管活动进行全面审查,其中包括检查和评估计划、该行业内的一般活动、研究和国际实践。该工作组由与戴维斯-贝瑟核电厂监管活动无关的NRC工…     

小型堆非居住区和规划限制区计算模型对比分析

为明确小型堆非居住区和规划限制区的计算方法，选择合适的计算模型，减少审评过程中存在的问题，本研究首先介绍了小型堆的定义，并比较了我国和国际上对小型堆定义的差别。同时说明了我国对小型堆非居住区和规划限制区设置的剂量准则和计算要求。并基于美国NRC发布的RG4.28草案，对比分析了其推荐的现实模型ARCON96与保守模型PAVAN之间的差别。ARCON96模型相比于PAVAN模型能够很好地考虑静风和尾流的影响，更真实地模拟了设施近距离的扩散情景，可有效的缩小非居住区和规划限制区的范围，进而提升小型堆的经济性。     

西安脉冲堆对环境的影响

应用厂址周围的人口与食谱调查资料以及反应堆参数 ,计算了该堆及其同位素生产线在正常运行及事故工况下厂址控制区边界的最大个人有效剂量当量和80km范围内的集体有效剂量当量。计算结果表明 :在正常运行时 ,厂址控制区边界最大个人有效剂量当量为6 0×10 -3mSv/a ,80km范围内的集体有效剂量当量为0 35人·mSv/a。反应堆最大假想事故事故下 ,所致厂址限制区外(500m)公众最大全身有效剂量当量为2 1×10-2mSv ,甲状腺剂量为3 8mSv ;事故持续30天后 ,80km范围内的集体有效剂量当量为0 14人·Sv(全身)和97人·Sv(甲状腺) ;正常运行工况和最大事故期间对本地区环境的影响都是可以接受的     

内陆代表性厂址事故扩散因子不同估算方法的比较

针对内陆核电选址中小风、静风频率较高的厂址,分别采用美国核管会(NRC)导则推荐方法和运用三维客观诊断风场与Lagrangian烟团模型模拟整年8 760小时逐时排放方法,计算了湖南桃花江厂址事故工况下的大气扩散因子,探讨复杂条件下大气扩散模型的适宜性。研究表明:在非居住区边界概率论方法计算的最大小时事故扩散因子较烟团模型计算方法保守;Lagrangian烟团模型计算的小时事故扩散因子在某些方位大于概率论方法结果,某些远距离子区的扩散因子大于近距离子区;对于释放时间相对较长的情形,导则方法估算结果仍偏小。由此可见,导则推荐方法得到的扩散因子存在不保守的情形,建议在计算复杂地形、小静风频率较高的内陆厂址事故扩散因子时慎重选择扩散模型。     

美国核管会收到首份小堆厂址申请

正【世界核新闻网站2016年5月16日报道】为了能够在美国田纳西州克林奇河(Clinch River)建设小型模块堆,田纳西流域管理局(TVA)近期向美国核管会(NRC)提交早期厂址许可(ESP)申请。田管局首席核官员Joe Grimes表示,这是核管会迄今收到的首份小堆厂址申请。     

小型自然循环铅基快堆固有安全性分析研究

自然循环反应堆一回路运行不需要设置驱动泵,具有结构简单、经济性好、固有安全性高等特点,是开发高安全性反应堆的重要发展方向。铅基冷却剂(铅或铅铋合金)的密度是水10倍以上,在相同温差下,铅基冷却剂的密度差比水更大,具有更好的自然循环能力,是设计自然循环反应堆的理想冷却剂。目前,国内外学者关于小型自然循环铅基快堆的研究主要集中于概念设计研究,关于该堆型的固有安全性研究较少,相关事故演化机理尚未明晰。本文在系统介绍小型自然循环铅基快堆的技术特点和研究现状的基础上,开展100 MW_(th)级小型自然循环铅基快堆无保护事故分析,深入探讨在极端假设事故工况下小型自然循环铅基快堆的固有安全性,为相关设计研究提供参考。     

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。     

Economic potential of modular reactor nuclear power plants based on the Chinese HTR-PM project

Modular reactors with improved safety features have been developed after the Three-Mile Island accident. Economics of small modular reactors compared to large light water reactors whose power output is 10 times higher is the major issue for these kind of reactors to be introduced into the market. Based on the Chinese high temperature gas-cooled reactor pebble-bed module (HTR-PM) project, this paper analyzes economical potentials of modular reactor nuclear power plants. The reactor plant equipments are divided into 6 categories such as RPV and reactor internals, other NSSS components and so on. The economic impact of these equipments is analyzed. It is found that the major difference between an HTR-PM plant and a PWR is the capital costs of the RPV and the reactor internals. The fact, however, that RPV and reactor internals costs account for only 2% of the total plant costs in PWR plants demonstrates the limited influence of this difference. On the premise of multiple NSSS modules forming a nuclear power plant with a plant capacity equivalent to a typical PWR plant, an upper value and a target value of the total plant capital costs are estimated. A comparison is made for two design proposals of the Chinese HTR-PM project. It is estimated that the specific costs of a ready-to-build 2 × 250 MW_th modular plant will be only 5% higher than the specific costs of one 458 MW_th plant. When considering the technical uncertainties of the latter, a 2 × 250 MW_th modular plant seems to be more attractive. Finally, four main points are listed for MHTGRs to achieve economic viability.     

Activities for 4S USNRC licensing

4S (Super-Safe, Small and Simple) is a small sized sodium-cooled fast reactor being developed for the electricity supply in remote areas, high-temperature steam supply more than 400 °C, seawater desalination, and hydrogen production. The system design of power output of 10 MWe (30 MWt) has been completed. The main feature is that it does not have to be refueled for a long period (i.e. 30 years for 10 MWe version), and enable the reactor closure sealed during plant operation. Furthermore, the small size of the reactor makes the reactor building suitable for below grade installing. These two features can provide resolutions for the issues relevant to safety, security, and safeguard, which become much more serious matter internationally these days.4S is a pool-type reactor which contains the whole primary cooling system in a vessel. For the purpose of reducing the maintenance requirements with the reactor, (1) reflectors to compensate for fuel burn-up instead of control rods, (2) electromagnetic pump (EMP) which has no rotating parts, and (3) residual heat removal system by natural circulation and natural air draft are adopted. Therefore, exchange of the reactor components is not required during plant operation, in addition to no needs for refueling.Toshiba has initiated the U.S. Nuclear Regulatory Commission (NRC) pre-application review of 10 MWe version for the purpose of applying for design approval (DA). A series of public meetings with NRC has been held four times, and five technical reports have been submitted to NRC in preparation for DA application. Topics discussed in these meetings included, plant design, metallic fuel, safety design philosophy, safety analysis, measures against severe accident, phenomena identification and ranking table (PIRT), etc. Some useful comments and questions on the issues regarding the specific feature of 4S as well as sodium-cooled fast reactor were raised by NRC at the public meetings. Among them, those items which are applicable to general sodium-cooled fast reactors, e.g. principal design criteria, guideline for safety analysis, validation and verification for safety analysis code, quality requirements, severe accident, and emergency planning are presented in this paper.     

福岛核事故引发的若干思考

福岛核事故是一场损失惨重的核灾难,它提出了一些值得深入分析和思考的问题。文章对今后核电发展中遇到的选址（包括在选址后的几年内,定期观测和仔细研究核电厂周围区域的地震和地质条件的变化发展）、机组延寿、反应堆压力容器可能开裂和一址多堆等问题进行了简短的讨论。     

基于RG 1.183的先进小堆选址源项模型初步研究

在《用于评估核动力反应堆设计基准事故的替代放射性源项》RG 1.183所述的假想事故场景情况下,考虑目前大多数的先进小型压水堆地上-地下布置的设计特点,对传统大型压水堆选址源项计算模型做了改进:在原安全壳内放射性物质守恒方程的基础上,考虑辅助厂房的阻滞作用,建立辅助厂房内放射性物质守恒方程。并以某先进小型反应堆核电厂为例,利用新模型计算了代表核素的释放,与现有模型进行了对比。     

模块化小型核反应堆自动卸压系统分析研究

模块化小型核反应堆（SMR）与传统大型压水堆在结构上存在很大差异,导致两者的严重事故进程存在较大差异。因此本文结合SMR自身设计特点,建立反应堆严重事故分析模型,对SMR的典型事故瞬态进行模拟计算,并对严重事故进程、热工水力现象和系统安全进行研究。在此基础上提出了SMR自动卸压系统优化改进方案,通过对自动卸压系统各级卸压管线的位置和阀门有效面积进行深入研究,并对相关参数进行敏感性分析,提出符合反应堆自身特点的卸压阀门有效面积的优化设计方案,为小型核反应堆的严重事故预防和缓解提供有效的依据和参考。     

Desirability of small reactors, HTGR in particular

Small reactors of about 100–300 MWe, High Temperature Gas Cooled Reactors (HTGRs) in particular, are considered desirable in future, based on the following ways of thinking;

Global scale enhancement of nuclear energy is considered necessary from reduction of environment impact point of view.

Small reactors are desirable, due to (a) enhanced safety in terms of fuel inventory and inherent safety, then (b) easier plant siting, near populated or industrial area, (c) more flexible development, planning and construction than larger reactors by finer adjustment with demand, and (d) economic competitiveness attainable by means of adoption of more rationalized systems categorization, simpler and modularized design, mass production in factory, less work at construction site, and marketability including that of developing countries.

In such ways, small reactors can be economically designed, constructed and operated, by conquering “scale de-merit”, in contrast with scale merit of larger reactors as seen in current LWRs.

Small HTGRs, in particular, are mostly desirable and promising, from view points of wider applications, such as electricity use, wide range of heat uses and/or cogenaration, Pu burning with high efficiency, wider fuel cycle options, using U, Pu and/or Th, with or without reprocessing, and development stages, where not only test reactor programs for development & demonstration but also realization programs are already going on towards commercial operation start in 2005–2010.

Development programs on small HTGR, related global activities and cost evaluations by developers and JAPC are shown, and steps towards their global scale commercializations are proposed.     

Insights from the AP600 probabilistic safety study

A full Probabilistic Safety Study (PSS), including Levels 1, 2 and 3, has been performed for the AP600, a Westinghouse design midsize, advanced passive light-water reactor plant. This study was part of an interactive process in the development of the AP600 design. As a result of frequent interaction between PSS analysts and design engineers, many potential safety concerns were addressed early in the evolution of the AP600 design that led to the incorporation of beneficial design modifications procedures and analyses. As a consequence of this process, the AP600 design that was submitted to the U.S. Nuclear Regulatory Commission (NRC) for certification presents a very low core damage frequency and severe accident frequency. Very effective severe accident mitigating features are also demonstrated. The frequency of exceeding the Protective Action Guideline limit of 1 rem effective dose equivalent at the site boundary within 24 hours of the onset of core damage is negligibly small. Therefore, simplification of emergency planning required for the AP600 is justified.     

日本小型核动力反应堆及其技术特点

日本原子能研究所研制了包括一体化船用堆(MRX)在内的几种小型核反应堆．MRX采用容器内置式控制棒驱动机构、水淹式安全壳、非能动余热排出系统;MR-100G和MR-1G是专门为区域供热和冷却系统提供能源,一回路系统自加压的全自然循环一体化压水堆．其排放物活性较低,小型化、模块式结构．可直接建于城市,甚至办公大楼的地下．,水下探测器用小型潜水反应堆(SCR)的设计思路与MRX基本相同．但一回路为全自然循环,日本小型核反应堆发展的技术思路清晰,注重用途的拓展,具有战略发展远见．在将我国大型核动力反应堆研制经验及其相应技术的推广方面,日本小型反应堆的发展思路值得借鉴。