基于多专业耦合分析的钍基熔盐堆反应堆本体设计研究

基于固态燃料钍基熔盐堆（TMSR-SF1）的特点,提出了基于多专业耦合的反应堆本体设计方法。参考现有成熟的设计规范,结合固态燃料钍基熔盐堆反应堆本体的结构和功能要求,完成了反应堆本体结构设计方案,并进行了反应堆本体屏蔽设计分析、堆容器上顶盖传热与温度场分析、反应堆结构力学分析,最终通过本体结构设计与多专业分析的反复分析迭代,初步实现了TMSR-SF1反应堆本体设计,满足TMSR-SF1功能要求。此外,通过反应堆结构选材论证和制造可行性分析,确保了结构设计的工程可实施性。      

反应堆压力容器退役包装屏蔽方案初探北大核心CSCD

为保证核电反应堆压力容器安全退役,本文以国内最早运行的秦山一期反应堆压力容器源项为参考,模拟设计保持压力容器完整和切割压力容器两种包装屏蔽方案,通过估算两种方案下废物体积、包装成本、运输及处置成本,对比分析发现切割压力容器方案更佳。研究成果可为核电站退役工作提供支持。     

反应堆厂房环形吊车临时吊梁安装工艺改进

大亚湾核电站2号机组环形吊车齿轮箱更换时,临时吊梁原安装工艺存在技术缺陷,提出改进方案.现场实施结果表明,改进后的新工艺不仅保证了所有的安装要求,而且缩短了工期,并最大程度地保证了临时吊梁的安装质量和施工安全,确保了反应堆厂房环形吊车更换齿轮箱项目任务的圆满完成.     

大亚湾核电站延伸运行技术的研究与实施

延伸运行(SO)是指当一回路的硼浓度接近0mg/L时,通过降温和降功率引入反应性,以保证反应堆加深燃耗继续保持功率运行,世界上许多核电厂采用SO.大亚湾核电站成功实施了我国的首次延伸运行.延伸运行作为一个特定的运行模式,需要进行相关的设计论证和安全分析.由于连续或阶跃式的降负荷和降温度,核测量系统和反应堆控制保护系统的参数需要进行特殊设置.按照发电计划的安排,大亚湾核电站的第一次延伸运行于2003年3月11日到3月20日实施,顺利实现了错开大亚湾核电站209大修和100大修的目的.     

中低放废物近地表处置顶盖设计与审评的计算机程序（HELP）的验证与应用

本文介绍了对美国环保局用于中低放废物近地表处置顶盖设计和审评的计算机程序ＨＥＬＰ程序的验证与应用。用本院包气带水分运移现场试验数据对ＨＥＬＰ程序进行了验证分析，结果表明ＨＥＬＰ程序的预测结果是合理的。用ＨＥＬＰ程序分析了我国西南地区条件下顶盖表面层厚度和表面层状况对顶盖中水分分布的影响，在此基础上对典型顶盖性能进行了模拟分析。模拟结果表明，顶盖表面植被对顶盖中水分分布影响很大，在顶盖设计中应充分重视；在潮湿地区，废物处置的安全性必须考虑处置系统的化学屏障作用。建议在今后的顶盖研究中，加入工程经济的内容，以实现顶盖设计的优化。     

秦山核电二期工程反应堆及反应堆冷却剂系统的仪表和控制

秦山核电二期工程反应堆及反应堆冷却剂系统的仪表和控制设计参考了大亚湾核电站的设计,但作了冷却剂系统三环路改二环路的适应性修改.本文总结了秦山核电二期工程反应堆及反应堆冷却剂系统仪表和控制的设计、重要仪表控制设备的研制.具体介绍了反应堆保护系统保护变量的选取、反应堆控制系统对堆芯的控制和监测以及提高核电厂可利用率的设计,并着重介绍了重要仪表控制设备的国产化研制过程.1号机组的成功运行证明设计和研制是非常成功的.     

大亚湾核电站小支管振动测量结果分析评定

大亚湾核电站在运行中,有部分辅助系统的小支管(管径不大于5.08cm)的振动较大,并有少数小支管出现振裂的情况,给核电站的安全运行带来不利影响.在大亚湾核电站十年安全审评时对辅助给水系统、安全壳喷淋系统、反应堆换料水池和乏燃料水池冷却和处理系统、化学和容积控制系统、反应堆硼和水补给系统、余热排出系统、安全注入系统和设备冷却水系统的潜在敏感管进行了现场振动测量.本文按相关要求对测量结果进行了分析评定,给出敏感管清单及改造建议.     

热电偶柱及其支架结构完整性分析

热电偶柱及其支架结构安装于反应堆压力容器上腔室内的堆内构件压紧顶帽上,其完整性影响到反应堆的运行安全。本文对热电偶柱及其支架结构进行了自重载荷下静态分析、流致振动影响分析、模态分析、地震载荷和LOCA载荷下的反应分析。结果显示,原设计的螺栓无法承受含有LOCA载荷的D级工况。本文给出了结构受力规律和修改方案。修改设计后的结构安装于反应堆压力容器顶盖内,已投入正常运行。     

核电站控制保护及辅助系统仿真方法研究

分析了大亚湾核电站控制保护及辅助系统仿真模型,采用面向对象编程方法以自编软件形式对大亚湾核电站模型进行模拟,并在开发过程中采用UML统一建模语言进行分析设计,实现了对该大亚湾核电站控制与保护系统的动态仿真.瞬态工况测试结果显示该仿真能较好地模拟反应堆一、二回路的控制与保护功能.本仿真程序具有通用、便携以及廉价等优点.     

RRA入口隔离阀控制逻辑改进分析

针对大亚湾核电站原始设计中反应堆余热排出系统(RRA)的入口隔离阀控制逻辑的设计不满足单一故障的情况,提出增加2台压力变送器的改进方案。通过定量化计算,评价改进方案对RRA及机组堆芯损坏的影响。     

Conceptual design of multi-purpose heat reactor “Nuclear Heat Generator”

Small heat reactors can apply to on site demand such as district heat and air conditioning, industrial process heat, greenhouse, and seawater desalination in urban and rural areas. The purpose of this paper is to design conceptually a multi-purpose reactor named “Nuclear Heat Generator (NHG)” which could be installed in energy consuming area. The reactor of 1MWt output is designed without any needs for fuel exchange and decommissioning on site. This cassette typed reactor vessel with sealing is transported to specified fuel fabrication shop every 3 to 4 years in order to exchange used fuels. Steam generators are involved in the self-pressurized integrated reactor with natural circulation. Generated steam pressure from heating reactor is 0.88 MPa (saturated) which is so less than that of current water reactors. Under low steam pressure it is considerably easy to make design of containment vessel and safety device. For economic competition overcoming scale demerit it will be necessary for the cassette type reactor to optimize its system design for the multi-production effect as well as modular construction and recycling system.     

The HTR module pressure vessel unit, design criteria and safety philosophy

A modular-helium-cooled high temperature reactor system for the cogeneration of electricity and process heat has been developed by Siemens—Interatom.Design, manufacture and operation of the pressure vessel unit will conform to German nuclear codes and standards for LWR's, some deviations or peculiarities for their application to HTR's are unavoidable. These are for instance:

• - The main steam nozzle, through which the steam line at 530°C penetrates the steam generator pressure vessel with a nominal design temperature of 350°C.

• - The pressure test concept in which the preservice pressure test will be performed in complete accordance with the codes and standards at 1.3 times the design pressure of 70 bar using water. Afterwards, the presence of graphite structures, ceramic insulation and, of course, the pebble bed core has to be considered. Pneumatic pressure tests are performed at 1.1 times design pressure accompanied by more detailed ultrasonic examinations.

• - The position of operational material irradiation surveillance specimens has to be chosen carefully. Design postulates concerning the incrase of ΔRT_NDT will pe confirmed in a separate program.

In general, the requirements of the assured safety concept, aimed to rule out catastrophic failure of the pressure vessel unit during lifetime are fulfilled.     

球床模块式高温气冷堆失冷事故特性研究

利用高温气冷堆专用系统分析软件THERMIX程序,对球床模块式高温气冷堆(HTR-PM)失冷失压和失冷不失压事故的动态特性进行了研究,分析了堆芯功率、燃料最高温度及堆舱水冷壁余热载出功率等关键参数的变化过程,并对影响余热排出功率和燃料最高温度的不确定性进行了评价.研究结果表明,在失冷事故下,堆芯余热可通过热传导、辐射和自然对流等非能动方式传至最终热阱大气,燃料元件和压力容器等重要部件的最高温度均在设计限值内.这为HTR-PM保持模块式高温气冷堆固有安全性不变的同时实现单堆250 MW的功率方案奠定了基础,也为后续高温气冷堆电站示范工程进一步的深入设计研究提供了依据.     

Probabilistic assessment of the fracture mechanics behaviour of an HTR-module primary circuit pressure boundary

In this paper the fracture mechanical behaviour of the primary circuit pressure boundary of a planned HTR-module reactor for electricity and steam generation under normal operation is assessed probabilistically. First and second order reliability methods (FORM-SORM) are used to calculate failure probabilities. They also allow a simplified analysis of the leak-before-break (LBB) behaviour. No LBB was probabilistically identified for the reactor pressure vessel. However, failure of the pressure vessel unit in normal operation probably originates from the connecting pressure vessel or the steam generator pressure vessel. They show LBB in probabilistic terms.     

KeproVt : underwater robotic system for visual inspection of nuclear reactor internals

An underwater robotic system for visual inspection of reactor vessel internals has been developed. The Korea Electric Power Robot for Visual Test (KeproVt) consists of an underwater robot, a vision processor based measuring unit, a master control station and a servo control station. The vision processor based measuring unit employs a first-of-a-kind engineering technology in nuclear robotics. The vision processor makes use of a camera located at the top of the water level referenced to the reactor center line to get an image of the robot, and computes the location and orientation of the robot. The robot guided by the control station with the measuring unit can be controlled to have any motion at any position in the reactor vessel with ±1 cm positioning and ±2° heading accuracies with enough precision to inspect reactor internals. A simple and fast installation process is emphasized in the developed system. The installation process consists of hooking a vision camera on the guide rail of the refueling machine and putting a small robot (14.5 kg in weight) in the reactor cavity pool. The easy installation and automatic operation meet the demand of shortening the reactor outage and reducing the number of inspection personnel. The developed robotic system was successfully deployed at the Yonggwang Nuclear Unit 1 for the visual inspection of reactor internals.     

A quantitative methodology for reactor vessel pressurized thermal shock decision making

The recent operating experience of the Pressurized Water Reactor (PWR) Industry has focused increasing attention on the issue of reactor vessel pressurized thermal shock (PTS). Previous reactor vessel integrity concerns have led to changes in vessel and plant system design and to operating procedures, and increased attention to the PTS issue is causing consideration of further modifications. Events such as excess feedwater, loss of normal feedwater, and steam generator tube rupture have led to significant primary system cooldowns. Each of these cooldown transients occurred concurrently with a relatively high primary system pressure. Consideration of these and other postulated cooldown events has drawn attention to the impact of operator action and control system effects on reactor vessel PTS.A methodology, which couples event sequence analysis with probabilistic fracture mechanics analyses, was developed to identify those events that are of primary concern for reactor vessel integrity. Operating experience is utilized to aid in defining the appropriate event sequences and event frequencies of occurrence for the evaluation.Once the specific event sequences of concern are identified, detailed thermal-hydraulic and structural evaluations can be performed to determine the conditions required to minimize the extension of postulated flaws or enhance flaw arrest in the reactor vessel. This paper addresses key aspects of the thermal-hydraulic and fracture mechanics analyses of the reactor vessel. The effects of incomplete mixing of safety injection flow in the primary cold leg and vessel downcomer and the application of warm prestressing are emphasized. The results of these analyses are being used to define further modifications in vessel and plant system design and to operating procedures.Previous design considerations that have evolved as a result of reactor vessel integrity evaluations are mentioned. These include the development of realistic design analysis tools and selection of plant system modifications. Modifications that are being developed or are under consideration are also mentioned. These include vessel fluence reductions, additional modifications to operating procedures, increased use of probabilistic event sequence and fracture mechanics analysis methods, enhanced material fracture toughness, and reductions in the severity or frequency of occurrence of dominant reactor vessel PTS transients.     

堆芯段筒体快速断裂计算

本工作研究反应堆压力容器在60年寿期末是否会出现快速断裂.文章采用断裂力学分析方法计算寿期末堆芯段筒体的应力强度因子,其计算结果满足RCC-M规范的要求,即在寿期末堆芯段筒体不会发生快速断裂.     

Integrated design approach of the pebble BeD modular reactor using models

The pebble bed modular reactor (PBMR) is the first pebble bed reactor that will be utilised in a high temperature direct Brayton cycle configuration. This implies that there are a number of unique features in the PBMR that extend from the German experience base. One of the challenges in the design of the PBMR is developing an understanding of the expected behaviour of the reactor through analyses and simulations and managing the integrated design process between the designers, the physicists and the analysts.This integrated design process is managed through model-based development work. Three-dimensional CAD models are constructed of the components and parts in the reactor. From the CAD models, CFD models, neutronic models, shielding models, FEM models and other thermodynamic models are derived. These models range from very simple models to extremely detailed and complex models. The models are used in legacy software as well as commercial off-the-shelf software. The different models are also used in code-to-code comparisons to verify the results.This paper will briefly discuss the different models and the interaction between the models, and how the models are used in the iterative design process that is used in the development of the reactor at PBMR.     

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。     

大亚湾核电站2号机组反应堆压力容器老化现状的初步分析

为确保核电站设备在整个寿期内设计安全裕度要求能够得到满足,必须对设备老化进行有效的管理。对影响反应堆压力容器(RPV)的老化机理进行了初步分析,并结合大亚湾核电站的实际情况对2号机组RPV的目前状态进行了分析评估：