反应堆主设备安装工艺及专用工具的设计研究

通过对国内外反应堆主设备吊运安装技术特点、主设备安装要求、反应堆厂房布置的研究,借鉴以往主设备安装工艺的优化与改进经验,确定一套适用于M310堆型的反应堆主设备安装工艺,并研制出满足实际工程建设所需的主设备安装专用工具.     

风险管理在核岛主设备采购中的应用浅析

核岛主设备是核电厂建造期间需要重点关注的采购物项,其交付进度和质量保证对核电厂工程建设目标顺利达成至关重要。本文通过对核岛主设备采购中的范围管理、进度控制、质量控制、沟通协调等环节的风险进行识别评估,有针对性地提出风险防范和应对措施,探讨了风险管理在核岛主设备采购中的应用,为提升设备采购水平提供参考。     

秦山三期(重水堆)核电工程调试计划管理和进度控制

概述了秦山三期核电工程调试管理、调试进度计划和管理、主要节点进度计划和影响进度的主要技术因素 ,最后介绍了从管理上加强进度控制的体会     

华龙一号核岛主设备复杂路径转运设备技术研究

基于对华龙一号堆型的土建布置和专用转运设备结构设计要求的分析,对核岛主设备转运技术进行研究,设计了核岛主设备转运设备,对其总体结构及主要构件进行分析,形成了华龙一号核岛主设备复杂路径转运技术。该技术能够满足华龙一号堆型的主设备转运的各项指标和要求,且具有自主知识产权,解决了我国第三代核电的自主研収过程中主设备转运安装等亟待解决的重大问题。     

法国乔治·贝斯Ⅱ浓缩厂的建设进入一个新阶段

法国阿海珐集团（Areva）宣布，该集团已向浓缩技术公司（ETC）移交了乔治·贝斯Ⅱ（Georges BesseⅡ）浓缩厂离心机厂房的钥匙。这意味着乔治·贝斯Ⅱ浓缩厂的建设进度己满足了项目建设进度表的要求，该厂的建设即将进入下一个阶段，即离心机的安装阶段。     

压水堆核电站主设备安装关键路径影响因素分析

压水堆核电站主设备安装影响因素是客观存在的,特别是主设备的设计、供货等方面因素,已经对国内部分压水堆核电站主设备安装产生一定影响。本文通过对压水堆核电站主设备安装关键路径、关键影响因素的分析,提出了相应的应对措施。     

三代压水堆核电站核岛主设备安全特征分析

为了满足"三代"核电技术要求,三代压水堆核电站核岛主设备在容量、设计寿命和安全裕量等方面均进行了优化设计。本文对AP1000和EPR等机型核岛主设备的技术特征、改进目标和改进方法等进行了分析,为核电厂的设备运行和改造、新电厂的设计提供参考。     

压水堆一回路主设备材质分析

压水堆核电站主设备主要包括反应堆压力容器、主管道、蒸汽发生器、稳压器以及主泵泵壳等。压力容器构成了一回路的压力边界,均是质保1级、安全1级、抗震1类的,在高温高压和中子辐照作用下,为了确保主设备能够长期稳定运行,对主设备的材质及其焊接性能要求非常高。根据ASME B&P规范,文章对主设备材质及其焊接性能进行分析和讨论,为我国核电厂主设备的材质选择提供参考作用。     

电气厂房结构设计中叠合梁（板）结构的设计与分析

岭澳核电站二期3#机组电气厂房（3LX），结构上的一大特点就是叠合梁（板）结构的大量使用，本文主要从力学角度分析了叠合梁（板）的受力特点，从施工和功能上分析了电气厂房采用叠合梁（板）结构的原因。     

秦山核电二期工程反应堆冷却剂系统主设备力学分析

介绍了秦山核电二期工程反应堆冷却剂系统主设备设计中所涉及的力学分析工作.主设备包括反应堆压力容器(包含控制棒驱动机构、堆内构件和燃料组件)、蒸汽发生器、主泵、稳压器.主要涉及的内容包含每个设备部件的应力、疲劳、热棘轮、断裂力学分析和内部构件流致振动分析以及试验验证等.设计者已掌握了主设备设计中所涉及的力学分析技术,取得了大量的成果.但是,仍有部分工作是与国外的设计单位合作完成的,我们还需做更深入的研究.     

秦山CANDU堆功率测量校正和控制改进

本文介绍了秦山三核CANDU6堆功率测量、控制设备的分区布置,论述了反应堆功率控制信号的计算校正和反应堆的区域功率控制,从CANDU6核功率控制设备、堆物理角度浅析其实现分区精细控制的机理,并阐述了为了提高反应堆功率控制系统可靠性和安全性而进行的主要设计改进     

The VVERs at KudanKulam

A Nuclear Power Project is being set-up at KudanKulam in the state of Tamil Nadu, India in collaboration with the Russian Federation. The project comprises of two units each of 1000 MWe VVER type reactor. The design of the plant and supply of all the major equipment is in the scope of the Russian Federation while development of infrastructure and project construction is in Indian scope of works. The VVER (Version V-412) reactor that is under construction at KudanKulam site is an advanced PWR, which incorporates all the features of a modern PWR as per the current Russian, Western and IAEA standards. The KudanKulam site in the southern Indian state of Tamil Nadu was one among the several sites evaluated by the Site Selection Committee, which cleared KudanKulam site for setting up an installed capacity up to 6000 MWe. The design, construction and operation of the plant meets the regulatory and licensing requirements of Russian regulatory body “RTN” as also India's Atomic Energy Regulatory Board. The supply of the equipment from the Russian Federation is on schedule and the project construction work by various Indian agencies is also ahead of schedule. The two units of KudanKulam Nuclear Power Project (KKNPP) are scheduled to achieve first criticality in the year 2007–2008. The paper discusses various design features, project construction and management aspects.     

数字化功率测量保护装置的研制

介绍了自行研制的以单片机为核心的300#反应堆功率测量保护装置及其运行情况,详细说明了装置的工作原理、硬件组成及软件部分的控制思想。将核仪器测量参数进行数字化处理是反应堆控制保护系统计算机化的前提,也是保护系统“冗余和多样性原则”在核测仪器中的应用。     

核电工程投资与进度集成控制统一编码系统

介绍了国际原子能机构和美国能源经济数据库的核电投资编码系统。在此基础上,为自行开发的核电工程投资与进度的集成控制系统设计了一套统一编码系统。在２００ＭＷ低温供热堆概算系统和投资与进度集成控制系统中的应用表明,该编码系统是可行的而且是有效的。该编码系统还可以进一步用于工程设计,设备材料和文档等方面和管理。     

Knowledge-directed characterization of nuclear power plant reactor core parameters

A major life-limiting factor of the UK's Advanced Gas-Cooled Reactors (AGRs) is the condition of the graphite core. Installation of new measurement equipment is difficult and expensive, therefore maximizing the information gained from existing equipment is highly desirable. The main approach to determining the health of an AGR core is through periodic inspections undertaken during planned outages. However, there is the desire to supplement this inspection activity through the analysis of data gathered as part of routine plant operation. One such source of data is measurements taken during refueling and this paper describes knowledge-directed characterization of this refueling data, both spatially across the reactor core and temporally across the operational lifetime of the core. Characterization provides information relating to the current condition of the reactor core and allows suspected ageing trends to be visualized and confirmed. A standard approach for characterizing reactor core data is presented and applied to a variety of different reactor core parameters. The benefit of this approach is that it allows engineers to distill large volumes of refueling data into a readily understandable format in a short period of time. It also allows hypothesized trends relating to the ageing process within the core to be tested and provides supporting evidence for these hypotheses. The trending data is also valuable as it can form the basis of a predictive model of ageing of the reactor core. The ageing process of nuclear graphite is understood from theoretical and experimental viewpoints and this empirical data, gathered from operating reactors, further supports this understanding. This paper represents the initial exploration of using refueling data to construct a predictive model of AGR reactor core ageing.     

核电站维修的三维数字化动态管理

详细介绍了数字电厂技术在大亚湾核电站反应堆厂房内设备的转运和空间布置动态管理中应用的全过程,论述了大亚湾三维数字化动态管理的建立及其在核电站维修项目中应用的方法和主要步骤。本项目利用外部数据库对电厂维修的模型状态进行保存,避免了对原有三维竣工模型的破坏和变动;紧密结合核电厂维修工作的主要进度步骤,对核电厂维修工作的空间布置和进度计划进行了全过程的仿真和优化,在核电厂维修工作中,较好地解决了在有限空间内进行维修空间计划安排和布置的仿真和优化问题。并在大亚湾核电厂2号机组更换反应堆顶盖中成功地进行了应用,缩短关键路径工期16小时,总工期缩短92．5小时。     

Underground nuclear power plants with surface turbine generators

This study has produced conceptual designs for light-water reactor underground nuclear power plants with portions of the plant at the surface. The investment cost penalty for underground excavation and cavity lining less interest during construction and escalation attributed to the underground portion of the plant at a favorable geologic site was estimated at between 3 and 4% above a comparable surface plant. The power plant turbine-generator was assumed to be at the surface and the reactor placed underground. Other equipment was located underground on the basis of a hazards analysis or by a functional relationship to the reactor. A guideline was adopted that the nuclear steam supply system and other principal subsystems should not require major redesign. Although some component relocation was allowable, it was permitted only if system performance and component operation were thought to be essentially unaffected. Plant size was stipulated as a single unit of 1000 MW(e) net which was not varied to determine an optimum size or number of units. Condenser cooling methods considered included natural convection wet cooling towers as well as once-through condenser cooling.In both the boiling-water (BWR) and pressurized-water (PWR) plant designs, four underground chambers are proposed to efficiently house the underground equipment. One of the four chambers is needed for the reactor and steam supply system. It is conceivable, if only reactor hazard protection is desired, that the equipment in two of the chambers (the nuclear auxiliaries and relay and switching) could be installed within a surface building. The location of the emergency core cooling system (ECCS) components as presently configured in surface plants relative to the nuclear steam supply system is critical to the satisfactory operation of the ECCS. In some situations the ECCS must draw water from the reactor area. The need for a positive head from this area to the appropriate pumps implies the pumps must be placed lower in elevation than the source of water. Consequently, it was appropriate to locate these ECCS components in a separate underground chamber near the reactor chamber.     

Dismantling design for the loop rooms on the MR reactor

The recently completed international co-operation project was aimed at planning for decommissioning the MR reactor identified as a pilot plant for the decommissioning of the other shutdown reactors on the site. The MR reactor was a pool-type, materials testing reactor with the total thermal power of 50 MW which incorporated pressure tubes containing fuel under test. The MR facility includes the reactor with its nine loop rig rooms containing pumps, heat exchangers and experimental equipment as well as systems and equipment located in other buildings in the complex. The objective of the MR reactor decommissioning project was to identify dismantling equipment and the decommissioning methodology for the reactor, loop rooms and redundant services to permit the refit and re-use of the building for a different nuclear related purpose. The dismantling design comprises two separate, but combined, tasks, namely, the dismantling of reactor installation itself and dismantling of experimental loops. The techniques proposed to undertake the dismantling operations within the loop rooms are described. Two options have been developed for removing contaminated equipment from the high radiation field loop rooms and packaging the waste into approved waste containers. The benefits and detriments of both methods have been identified, which allows implementing the safe, timely and cost-effective decommissioning.     

Testing of improved algorithms controlling energy release in VVÉR-1000 in maneuvering regimes at the Tianwan nuclear power plant (China)

The salient features of improved algorithms adopted at the Tianwan nuclear power plant for controlling the energy released in a VVéR-1000 core are examined. The optimal configuration of the controlling groups is chosen in the first two power-generating units, the reactor power can be varied automatically under the control of an automatic power regulator, the boron regulation system makes it possible to determine the first-loop makeup automatically, and a modern version of the Imitator Reactora program has been installed. The results of testing the algorithms in the No. 1 unit in regimes with single and cyclic (daily) power maneuvers are presented. The tests of single power maneuvers were combined with dynamic tests of equipment. The operation of the power-generating unit in a daily load schedule was tested separately. Five daily load-change cycles were conducted during these tests. __________ Translated from Atomnaya énergiya, Vol. 103, No. 5, pp. 277–282.     

粘钠设备水蒸气氮气清洗工艺研究

快堆电站一、二回路的设备因为粘有冷却剂—钠,在维修和退役前必须进行除钠处理。水蒸气氮气清洗是快堆电站设备清洗系统所采用的清洗除钠工艺。本文首先对该工艺的清洗原理进行了分析,然后对工艺研究的试验装置进行了论述,最后通过两种特征的粘钠设备的清洗,深入分析清洗过程中所出现的各种现象,为后续的快堆电站粘钠设备清洗技术的研究提供了保证,同时获得了可直接用于中国实验快堆(CEFR)设备清洗系统的运行经验。