超临界水堆与先进沸水堆安全特性差异性分析

超临界水堆和先进沸水堆在结构和原理上存在很多相似点,但在安全特性方面存在一定的差异性。本文通过对比分析超临界水堆和先进沸水堆两个堆型来研究它们安全特性的差异性,研究结果表明:超临界水堆由于不存在汽水两相,不会发生偏离泡核沸腾现象,又因为采用的是非能动和能动相结合系统,具有较好的安全特性;和相同功率的先进沸水堆相比,其比焓高,流量低,泵消耗功率低,压力下降速率更快,消耗的核燃料低;超临界水堆安全系统与先进沸水堆安全系统虽然都采用了非能动安全系统和能动系统,但在超临界水堆能动安全系统中,其压力控制系统控制速度较快,而非能动控制系统则较为稳定。     

中国实验快堆超功率瞬态分析研究

在核电厂的安全分析中,系统程序能够对多种事故瞬态进行瞬态分析。随着我国快堆技术的发展,在快堆系统程序的开发和应用等方面也做了一些工作。本文运用法国原子能委员开发的快堆系统分析程序DYN4G建立了中国实验快堆(CEFR)的系统模型,包括堆芯、一回路、二回路、汽水回路和事故余热排出系统,并计算了满功率下的稳态,与设计值进行了比对,同时完成了超功率事故的瞬态分析,并与《CEFR最终安全分析报告》中的计算进行了对比验证。计算结果表明程序能较好的模拟CEFR的稳态和超功率情况,为进一步开展CEFR的安全研究及钠冷快堆的安全分析打下了基础。     

CEFR和BN800系统运行工况的分析和比较

研究反应堆系统的运行工况,对于反应堆的安全运行是非常重要的,本文对我国实验快堆CEFR和俄罗斯BN800快堆的总体特征进行了比较,并对CEFR和BN800的系统运行分类情况进行了比较分析,在此基础上,对CEFR和BN800系统的主要运行工况进行了细致的比较和分析,详细地分析了两个反应堆系统在各种运行工况下运行的要求、工况参数和运行过程以及相应的各项安全要求,比较分析了它们之间的相同之处和区别所在。     

中国实验快堆堆容器冷却系统全厂断电工况温度场分析

堆容器冷却系统是中国实验快堆(CEFR)-回路系统中的重要辅助系统之一,用于在各种工况下对反应堆堆容器进行冷却.本文利用国际通用的计算流体力学软件STAR-CD对CEFR堆容器冷却系统进行三维数值模拟,得到了在全厂断电事故发展过程中堆容器冷却系统的温度场和流场的瞬态分析结果,为相应部件的力学分析以及其它工况的分析提供了数据,对快堆优化设计和安全分析提供了重要的理论支持.     

CEFR最终安全分析报告编写情况

CEFR最终安全分析报告(FSAR)是实验快堆执照申领应提交的技术文件之一。快堆工程部根据实验快堆工程二级网络进度计划提出了2005年完成报告书初稿编写的目标。CEFR最终安全分析报告主要参考和依据下列文件编制:“CEFR安全分析报告格式与内容”、“核电厂安全分析报告的标准格式     

CEFR在ULOF工况下的自然循环能力分析

自然循环能力是衡量钠冷快堆固有安全性的重要指标,堆芯布置、回路设计及工况参数等都会影响堆芯自然循环能力,因此不同堆型的自然循环能力有很大差异。为了保证堆芯事故得到有效缓解,中国实验快堆(CEFR)的设计中通过优化系统布置,重点考虑了堆芯自然循环。本文采用SAS4A程序对CEFR进行系统建模,分析了CEFR在无保护失流(ULOF)工况下的堆芯热工水力参数瞬态特性,验证了CEFR利用自身自然循环和负反馈设计进行事故缓解的能力,本文还对一回路流动阻力和二回路钠装量对堆芯自然循环的影响进行分析。计算结果表明,CEFR具有良好的自然循环特性,在ULOF工况下可以依靠其负反馈停堆,并能够建立起稳定的自然循环从而导出堆芯余热。     

CEFR事故余热排放系统管道热应变测量及结果分析

中国实验快堆(CEFR)以钠作为冷却剂。事故余热排放系统是CEFR快堆的专设安全设施,在反应堆出现地震、系统供电全部中断、全部蒸汽发生器给水中断的事故工况时,将堆芯余热通过空气冷却器非能动地排放到最终热阱。CEFR事故余热排出系统设计温度为550℃,运行温度最高为516℃,全部为双层管道,管道内运行介质为高温的液态金属钠。通过对事故余热排放系统进行热应变测量和数据分析,掌握系统管道的应力应变情况和监测系统运行状态的应力变化。     

OASIS程序的开发与应用

全面描述了对来自法国原子能委员会 (简称CEA)的快堆系统安全分析程序OASIS的引进和开发工作 ,并在此基础上介绍了该程序在中国实验快堆 (ChinaExperimentalFastReactor,简称CEFR)初步安全分析报告中对主给水管道断裂事故的分析计算。     

中国实验快堆主冷却系统建模与仿真研究

中国实验快堆(CEFR)是钠冷快中子反应堆,其一、二回路的运行特性对反应堆的安全运行具有重要的影响。使用JTopmeret软件建立CEFR一、二回路主冷却系统和蒸汽发生器(SG)的仿真模型,用于计算系统任意一点的流量、压力、温度等运行参数。在稳态及瞬态工况下,系统主要参数仿真值与设计值的误差均小于2%,满足系统仿真的精度要求。     

美日先进沸水堆的研究工作进入最后阶段

【《瑞士原子能协会通报》1984年10月第17期报道】美国通用电气公司同日本的合作者签订了进行先进沸水堆第三阶段的研究协定。参加这项研究工作的日本公司有:东京电力公司、东京芝浦电气公司和日立公司。第三阶段的研究工作将进行8个月。研究成功后,先进沸水堆将比原有的沸水堆系统在技术上和经济上有明显的改善。由于保留了沸水堆安全性和可靠性等优点,并作了进一步的改进,先进沸水堆的投     

AB1600—Progress of ABWR technology toward next generation ABWR

Advanced boiling water reactor (ABWR) plants have achieved an excellent operating performance since the first ABWR plant started its commercial operation in 1996. Based on the ABWR technology, progress has been made towards a next generation ABWR, AB1600 in Toshiba. The AB1600 plant aims at meeting the demand for the replacement of the current BWRs, which is expected to be realized by 2020 and beyond. In the AB1600 design, therefore, further improvements in economic and reliability aspects have been pursued by incorporating several new technologies. The reactor power is uprated to 1600 MWe from 1350 MWe of the first ABWR plant in order to benefit from economy of scale. A large fuel assembly with high power density is adopted in order to reduce both of the capital cost and maintenance cost associated with refueling by decreasing the number of fuel assemblies and control rod drives. The AB1600 safety system design employs a hybrid safety system, which consists of both active and passive systems for the design basis and the beyond-design basis accidents, to enhance the safety of the plant. As a countermeasure against severe accidents, the passive systems for the functions of decay heat removal, coolant injection and molten core debris cooling are incorporated.     

Modern Technology Applied in Advanced BWR (ABWR)

The Advanced Boiling Water Reactor (ABWR) has been developed by an international team of BWR manufacturers to respond to worldwide utility needs in the 1990's. Major objectives of the ABWR program are design simplification; improved safety and reliability; reduced construction, fuel and operating costs; improved maneuverability; and reduced radiation exposure and radwaste. The ABWR is the result of the continuing evolution of the BWR, incorporating state-of-the-art technology and improvements based on worldwide experience, and extensive design and test and development programs. The ABWR incorporates the best proven features from BWR designs in Japan, the United States and Europe. The many new features are seen to provide superiority in terms of performance characteristics and economics relative to current LWR designs. The Tokyo Electric Power Co., Inc. recently announced the selection of General Electric Co., Hitachi, Ltd. and Toshiba Corp. to design and construct two lead Advanced Boiling Water Reactors as Unit 6 and 7 at the Kashiwazaki Kariwa Nuclear Power Station. Construction is scheduled for the early 1990's, and commercial operation planned for 1996 for Unit 6 and 1998 for Unit 7.     

中国实验快堆一级概率安全评价--事件树的建立

基于前期对初因事件的确定和归集研究，根据初因事件组的特征及对各初因事件序列的初步分析，确定了中国实验快堆(CEFR)一级概率安全评价(PSA)报告所要建立的事件树数目、各事件树的题头事件、事件序列后果的分类等。最后，根据CEFR具体安全设计特征创建了完整的事件树，为后续事件序列的深入分析奠定了重要基础。     

论解决核电安全问题的途径

从现有水冷反应堆核电厂存在堆芯熔化危险这一安全问题的焦点出发,分析了改进型反应堆AP-600、SIR、非能动安全反应堆PIUS和具有固有安全的模块高温气冷堆MHTGR等的安全特性.按照下一代水冷反应堆的设计要求和用户要求,提出了解决水堆核电厂安全问题的新概念——自安全铀氢锆反应堆,该堆型可能成为世界水堆核电发展的一个方问。中国核动力研究设计院正在探讨这种堆型。     

模块式高温气冷堆核电站设计特点

模块式高温气冷堆具有安全、灵活、可靠、经济性好的优点,受到核技术先进国家的重视。本文着重介绍了美国新近推出的模块式高温气冷堆核电站的设计特点和安全特性。     

Future directions in boiling water reactor design

The Advanced Boiling Water Reactor (ABWR) is being developed by an international team of BWR manufacturers to respond to worldwide utility needs in the 1990s. Major objectives of the ABWR program are design simplification; improved safety and reliability; reduced construction, fuel and operating costs; improved maneuverability; and reduced occupational exposure and radwaste.The ABWR incorporates the best proved features from BWR designs in Europe, Japan, and the United States and application of leading edge technology. Key features of the ABWR are internal recirculation pumps; fine-motion, electro-hydraulic control rod drives; digital control and instrumentation; multiplexed, fiber optic cabling network; pressure suppression containment with horizontal vents; cylindrical reinforced concrete containment; structural integration of the containment and reactor building; severe accident capability; state-of-the-art fuel; advanced turbine/generator with 52 in. last stage buckets; and advanced radwaste technology.The ABWR is being developed as the next generation Japan standard BWR under the guidance and leadership of the Tokyo Electric Power Company, Inc. and a group of Japanese BWR utilities. During 1987, the Tokyo Electric Power Company, Inc. announced its decision to proceed with two ABWR units at its Kashiwazaki-Kariwa Nuclear Power Station, with commercial operation of the first unit in 1996 and the second unit in 1998. The units will be supplied by a joint venture of General Electric, Hitachi and Toshiba, with General Electric selected to supply the nuclear steam supply systems, fuel and turbine/generators. In the United States it is being adapted to the needs of U.S. utilities through the Electric Power Research Institute's Advanced LWR Requirements Program, and is being reviewed by the U.S. Nuclear Regulatory Commission for certification as a preapproved U.S. Standard BWR under the U.S. Department of Energy's ALWR Design Verification Program. These cooperative Japanese and U.S. Programs are expected to establish the ABWR as a world class BWR for the 1990s.International cooperative efforts are also underway aimed at development of a simplified BWR employing natural circulation and passive safety systems. This BWR concept, while only in the conceptual design stage, shows significant technical and economic promise.     

China Experimental Fast Reactor（CEFR）：Engineering Progress of CEFR Project in 2005

2005年，工程建设取得新的进展，工程质量处于受控状态，项目管理水平进一步提升。     

Model extension and improvement for simulator-based software safety analysis

One of the major concerns when employing digital I&C system in nuclear power plant is digital system may introduce new failure mode, which differs with previous analog I&C system. Various techniques are under developing to analyze the hazard originated from software faults in digital systems. Preliminary hazard analysis, failure modes and effects analysis, and fault tree analysis are the most extensive used techniques. However, these techniques are static analysis methods, cannot perform dynamic analysis and the interactions among systems. This research utilizes “simulator/plant model testing” technique classified in (IEEE Std 7-4.3.2-2003, 2003. IEEE Standard for Digital Computers in Safety Systems of Nuclear Power Generating Stations) to identify hazards which might be induced by nuclear I&C software defects. The recirculation flow system, control rod system, feedwater system, steam line model, dynamic power-core flow map, and related control systems of PCTran–ABWR model were successfully extended and improved. The benchmark against ABWR SAR proves this modified model is capable to accomplish dynamic system level software safety analysis and better than the static methods. This improved plant simulation can then futher be applied to hazard analysis for operator/digital I&C interface interaction failure study, and the hardware-in-the-loop fault injection study.     

Safety and licensing of MHTGR

The Modular High-Temperature Gas-Cooled Reactor (MHTGR) design meets stringent top-level regulatory and user safety requirements that require that the normal and off-normal operation of the plant not disturb the public's day-to-day activities. Quantitative, top-level regulatory criteria have been specified from US NRC and EPA sources to guide the design. The user/utility group has further specified that these criteria be met at the plant's exclusion area boundary (EAB). The focus of the safety approach has then been centered on retaining the radionuclide inventory within the fuel by removing core heat, controlling chemical attack, and by controlling heat generation. The MHTGR is shown to passively meet the stringent requirements with margin. No operator action is required and the plant is insensitive to operator error.