蒸汽发生器传热管氦检漏定量定位技术分析及验证

蒸汽发生器传热管氦检漏是核电厂蒸汽发生器重要检查项目。本文针对蒸汽发生器传热管氦检漏漏点定量定位分析提出了一套算法理论,并通过试验平台进行试验,验证了算法的准确性,为漏点分析提供了可靠的理论依据。     

核电厂蒸汽发生器传热管氦气检漏系统研究

本研究介绍了某核电厂蒸汽发生器传热管在役氦气检漏系统的原理及系统组成,并模拟了某核电厂蒸汽发生器在役大修期间传热管检漏试验。试验结果表明,最佳参数可设置为:蒸汽发生器二次侧氦气浓度份额为30%;抽气速率为 20 L/min;蒸汽发生器二次侧压力为0.6 MPa;系统漏点定位误差在0.5 m以内。本文研究的蒸汽发生器传热管在役氦气检漏系统可为国内核电厂安全、稳定地运行提供可靠的技术保障。      

立式和卧式蒸汽发生器传热管在役检查技术

美国立式蒸汽发生器传热管在役检查技术与经验。在役检查要求美国核电站技术规格书(USNRC 1981)规定了压水堆核电站蒸汽发生器传热管的在役检查要求(取样规模和频率)，核电站在首次在役检查中接受检查的蒸汽发生器传热管的数量取决于该电站中蒸汽发生器的数目和是否对这些管子进行过役前检查．在随后的检查中每次检查一台待查的蒸汽发生器，进行轮流安排．     

大亚湾核电站蒸汽发生器传热管的涡流检查

介绍了大亚湾核电站蒸汽恨生器传热管的涡流检查设备与技术,１号机组蒸汽发生器首闪在役检查的关键路径,技术改进及其检查结果。首次在役检查结果表明;３台蒸汽发生器没有发现传热管破损,保证了核电站的安全运行。     

核电站蒸汽发生器的更换

老一代核电站的蒸汽发生器相继发生传热管腐蚀破损，传热管泄漏率不断增高，计划外停堆的几率增加，维修费用也随之增加。当传热管的缺陷超过堵管限制时，一般要采用堵管的措施。随着堵管数量的增加，核电站难以维持其额定功率，这样就必须考虑更换蒸汽发生器。更换蒸汽发生器是保证核电站安全与经济效益的最好策略。在欧洲，广泛采取更换蒸汽发生器以增加传热面积提高热功率，热功率提高可以增加发电量，以便尽早收回更换所需费用。     

秦山核电站蒸汽发生器的氦质谱检漏技术

介绍了核电站蒸汽发生器管子-管板焊缝的氦质谱检漏及总漏率测试方法的研究及为提高检测灵敏度所采取的措施。获得了单管检漏的系统最小可检灵敏度为10~(-8).L/s;总漏率测试系统的灵敏度为10~(-7)Pa.L/s的好结果。为核电站蒸汽发生器建造质量的鉴测提供了可靠的技术保证,也为大容器压力设备的检漏提供了好的方法。     

秦山核电站蒸汽发生器的氦质谱检漏技术

介绍了核电站蒸汽发生器管子-管板焊缝的氦质谱检漏及总漏率测试方法的研究及为提高检测灵敏度所采取的措施。获得了单管检漏的系统最小可检灵敏度为10~(-3)PaL/s;总漏率测试系统的灵敏度为10~(-7)PaL/s的好结果。为核电站蒸汽发生器建造质量的鉴测提供了可靠的技术保证,也为大容器压力设备的检漏提供了好的方法。     

核电站蒸汽发生器降质预防和在役检查

介绍了法国核电站蒸汽发生器在运行初期所发生的传热管降质现象,重点论述了大亚湾核电站１ 号机组蒸汽发生器对降质预防所采取的措施和在役检查,包括二回路水化学监控、泄漏率监测、传热管涡流检验、二次侧的机械清洗、清洁度检查和外来物取出等。实践证明,采取了上述降质预防措施和在役检查,对核电站的安全运行起到了重要作用     

蒸汽发生器传热管焊接堵管用小型化激光头研制

核电站蒸汽发生器传热管破损将导致放射性冷却剂外泄，因此需进行堵管作业。激光焊接精度高，可用于蒸汽发生器的焊接堵管。然而，受限于蒸汽发生器的空间结构，现有的激光焊接头往往难以满足实际要求，因此需设计定制化的小型激光头。首先，分析和计算了小型化激光头的光路；其次，设计出小型激光头的机械结构，并完成其在蒸汽发生器内的运动仿真和干涉模拟；最后，对所设计的激光头装置进行了模拟传热管焊接验证。结果表明，所设计的小型化激光头具有良好的焊接性能和可操控性能，可用于蒸汽发生器的焊接堵管作业。     

核电厂卧式蒸汽发生器热工性能的计算和分析

文章基于卧式蒸汽发生器的工作原理及内部结构特点,建立了卧式蒸汽发生器数学物理模型,开发了针对卧式蒸汽发生器的热工水力程序。基于在役核电站卧式蒸汽发生器的设计参数,对程序进行了校核。该程序可以用来研究卧式蒸汽发生器内主要热工参数的分布情况,为卧式蒸汽发生器设计、安全分析提供指导;也可以根据在役核电站的历史运行数据对蒸汽发生器现阶段热性能进行分析评定,对蒸汽发生器一段时间内的热性能进行预测,为蒸汽发生器的运行、检修以及更换提供依据。     

10MW高温气冷堆蒸汽发生器传热管束应力分析

10MW高温气冷堆蒸汽发生器（SG）传热管是带放射性的一回路与无放射性的给水蒸汽二回路的屏障。管束的破裂将会引起二回路的水蒸汽进入一回路,从而导致堆芯压力的升高和放射性产物的外泄,因此确保传热管的完整性是十分必要的。传热管的结构采用小弯曲半径的螺旋管结构,对于这种无法进行体积性在役检查的螺旋管,利用破前漏思想确保传热管的完整性是一个重要的选择。本文利用管道有限元程序PIPESTRESS对高温气冷堆蒸汽发生器传热管的应力进行了计算,得到了传热管的最大应力和应力与材料的不利组合位置。     

Development of automatic algorithm for localizing loose parts with a steam generator

In this paper, an automatic localization algorithm for estimating the impact location by loose parts in a Steam Generator using modified triangular method is proposed and applied to the impact test data of YongGwang Nuclear Power Plant Unit 3 Steam Generator. The algorithm, at first, was developed at the Mock-up system and modified to apply for the real plant. The Steam Generator is modeled as a cylinder shape and the modified method is used to find out the impact point of a loose part on the model. The result of estimated impact point applying to the developed algorithm has below about 5% average error. If the algorithm will be installed in the existing plant or next generation plant, the safety and reliability of Nuclear Power Plant will be improved.     

Experimental evidence of imperfection influence on the buckling of thin cylindrical shell under uniform external pressure

The load carrying behaviour of cylindrical thin-walled shell structures under pressure load is strongly dependent on the nature and magnitude of the imperfections invariably caused by various manufacturing processes. The present paper examines instabilities of long homogeneous and isotropic thin elastic tubes, characterized by geometric imperfections like eccentricity or ovality, on the buckling behaviour in conditions for which, at present, a complete theoretical analysis was not found in literature. Moreover, the additional aspect of the influence of the welded joint geometry and position is investigated over a wide range of diameter to thickness ratio, extending the findings of previous works. The problem of buckling for variable load conditions is relevant in the context of NPP applications as, for instance the optimisation of an integrated and innovative LWR Steam Generator (SG) tubes, according to the updated ASME rules.To the purpose, at Pisa University a rather intense research activity is being carried out on the buckling of thin walled metal specimens in the dimensional range suitable for the above mentioned application. Therefore a test equipment (with the necessary data acquisition facility), suitable for carrying out test series on this issue, as well as numerical models implemented on the MARC FEM code, were set up. The experiments were conducted on test specimens with different materials, e.g. A-316 ASTM (with and without seam weld) and Inconel 690 TT, as well as different loading conditions (lateral and hydrostatic external pressure). A validation of numerical evaluations by comparison with test results is also performed. A good agreement has been observed between the experimental data and the elasto-plastic finite element analyses results, highlighting also the different influence of the mentioned imperfections on the buckling loads.     

基于RS-FNN的核电厂设备智能故障诊断方法的研究

将粗糙集(RS)理论与模糊神经网络(FNN)相结合,能充分发挥各自的优点.本文利用RS方法对知识的约简技术,从大量的原始数据中提取精简的规则,基于这些规则建立的FNN网络具有更好的拓扑结构,学习速度大大提高、判断准确、容错能力强,具有更高的实用价值.为了验证该方法的有效性,以核电厂设备蒸汽发生器U形管破裂等故障为例,进行了仿真实验研究.诊断结果表明,将基于RS理论的FNN智能故障诊断方法引入核电厂设备故障诊断中是可行的,并且具有简单方便、计算量小、诊断结果可靠等特点.     

Intelligent system for transient data collection and fatigue monitoring of pressurised water reactors nuclear steam supply system

Electricité de France (EDF), the French national electricity company, is operating 54 standardised pressurised water reactors. This about 500 reactor-years experience in nuclear stations operation and maintenance area has allowed EDF to develop its own strategy for monitoring of age-related degradations of NPP systems and components relevant for plant safety and reliability. After more than fifteen years of experience in regulatory transient data collection and seven years of successful fatigue monitoring prototypes experimentation, EDF decided to design a new system called SYSFAC (acronym for SYstème de Surveillance en FAtigue de la Chaudière) devoted to transient logging and thermal fatigue monitoring of the reactor coolant pressure boundary. The system is fully automatic and directly connected to the on-site data acquisition network without any complementary instrumentation. A functional transient detection module and a mechanical transient detection module are in charge of the general transient data collection. A fatigue monitoring module is aimed towards a precise surveillance of five specific zones particularly sensible to thermal fatigue. After a first step of preliminary studies, the industrial phase of the SYSFAC project is currently going on, with hardware and software tests and implementation. The first SYSFAC system will be delivered to the pilot power plant by the beginning of 1996. The extension to all EDF’s nuclear 900 MW is planned after one more year of feedback experience.     

测温热电偶自动插装技术及插装机械手

核电站反应堆测温热电偶插装机械手是堆芯热电偶安装，检修更换的专用设备。本文结合３００ＭＷ的反应堆型，介绍了一种新型的热电偶插装技术以及插装机械手的结构和工作原理。经实例应用认为，该插装技术可作为导管引导式铠装热电偶的安装设备。     

Steam Generator control in Nuclear Power Plants by water mass inventory

Control of water mass inventory in Nuclear Steam Generators is important to insure sufficient cooling of the nuclear reactor. Since downcomer water level is measurable, and a reasonable indication of water mass inventory near steady-state, conventional feedwater control system designs attempt to maintain downcomer water level within a relatively narrow operational band. However, downcomer water level can temporarily react in a reverse manner to water mass inventory changes, commonly known as shrink and swell effects. These complications are accentuated during start-up or low power conditions. As a result, automatic or manual control of water level is difficult and can lead to high reactor trip rates.This paper introduces a new feedwater control strategy for Nuclear Steam Generators. The new method directly controls water mass inventory instead of downcomer water level, eliminating complications from shrink and swell all together. However, water mass inventory is not measurable, requiring an online estimator to provide a mass inventory signal based on measurable plant parameters. Since the thermal-hydraulic response of a Steam Generator is highly nonlinear, a linear state-observer is not feasible. In addition, difficulties in obtaining flow regime and density information within the Steam Generator make an estimator based on analytical methods impractical at this time.This work employs a water mass estimator based on feedforward neural networks. By properly choosing and training the neural network, mass signals can be obtained which are suitable for stable, closed-loop water mass inventory control. Theoretical analysis and simulation results show that water mass control can significantly improve the operation and safety of Nuclear Steam Generators.     

蒸汽发生器二次侧^16N迁移时间的计算模型

用测量蒸汽中＾１６Ｎ的放射性活度来监测蒸汽发生器传热管的泄漏，是一项新的技术，该监测系统确定泄漏率的一个重要参数是＾１６Ｎ在蒸汽发生器二次侧的迁移时间，本文认为＾１６Ｎ泄漏工质是以汽相形成随二次测工质运动，根据蒸汽发生器二次侧工质的流动特点，将迁移的时间分成四段计算，并重点提出了管束区汽相运动的速度分布计算模型，用本文模型对秦山核电厂蒸汽发生器的＾１６Ｎ迁移时间进行了计算，并与法国电力公司的计算结     

Total Neutron Cross Section in Heavy Ice at 77 and 15 K for Neutron Energies between 0.3 and 100 meV

A 10MW High Temperature Gas Cooled Reactor (HTR-10) designed by the Institute ofNuclear Energy Technology (INET) of Tsinghua University is being constructed now. The steam generator (SG) of the HTR-10 is one of the most important facilities for reactor safety. In order to investigate the thermal-hydraulic performance of the SG, a full scale HTR-10 Steam Generator Two Tube Engineering Model Test Facility (SGTM-10) was installed and tested at INET. This paper describes the SGTM-10 in detail. The test assembly of the SGTM-10 simulates practical thermal and structural parameters of the HTR-10. The SGTM-10 consisted of three separated loops, primary-helium loop, secondary-water loop, and third-cooling water loop. There are two parallel tubes arranged in the test assembly. The main experimental equipment is shown in this paper. Analysis shows that for once-through steam generator simulation experiment, the electric-heated simulation method could not match practical operating condition. The results may not reflect true phenomena. The main results of experiments, for example effects of the outlet pressure, effects of the heating power, effects of the inlet sub-cooling are described. Experiments indicated, when the heat load of the HTR-10 is more than 30% the SG will be stable.