蒸汽发生器柔性泥渣冲洗枪研制

蒸汽发生器(SG)二次侧管板表面堆积的泥渣会随运行时间的增加而变硬,使传热管受到腐蚀和挤压,导致传热管破裂。为了直接对硬泥渣堆积层进行冲洗,研制了一种SG柔性泥渣冲洗枪。本文阐述了SG柔性泥渣冲洗枪的结构、原理、主要技术指标及技术难点。应用结果表明,研制的SG柔性泥渣冲洗枪满足SG柔性冲洗的要求,冲洗效果显著。     

卧式蒸汽发生器排污穴室水力冲洗设备的研发和应用

为清理田湾核电站1、2号机组蒸汽发生器（SG）二次侧排污穴室内积聚的腐蚀产物、降低其对壳体母材和焊缝造成腐蚀的风险，研制了排污穴室水力冲洗设备并且在机组大修期间进行了应用。通过水力冲洗工作，清理了SG排污穴室内积聚十余年时间的沉积物、检查确认排污穴室内壁和焊缝无异常，提高了蒸汽发生器运行的安全性、可靠性。      

生物X射线小角散射实验站控制和数据采集系统

生物X射线小角散射光束线站(Biological Small Angle X-Ray Scattering,BioSAXS)是国家蛋白质科学研究上海设施五线六站之一,运动控制和数据采集系统是BioSAXS实验站建设的重要组成部分。介绍了BioSAXS实验站运动控制和数据采集系统,设计实现了基于EPICS(Experimental Physics and Industrial Control System)的运动控制系统,开发了EPICS下的Pilatus探测器的数据采集软件。通过对运动控制、探测器数据采集和光强检测等控制操作界面进行集成,形成了统一的用户界面,并在BioSAXS实验站的调试和运行中得到成功应用。该控制软件界面友好,操作简单,其功能和性能通过了实验验证,满足了BioSAXS线站对运动控制和数据采集系统的需求。     

核动力蒸汽发生器水位控制方法分析

核动力蒸汽发生器(SG)是一个高度复杂的非线性时变系统.SG在瞬态、启动和低功率运行工况下的"收缩"与"膨胀"现象引起的逆动力学效应使SG的水位控制变得复杂.文章分析了SG水位控制方法的特点,重点分析了SG水位模糊控制方法与神经网络控制方法.指出了传统的PI(D)水位控制方法存在的问题,就SG水位控制发展趋势提出了看法.     

核电机组大修后主蒸汽压力的变化分析

核电厂的蒸汽发生器(SG)出口蒸汽压力(PSG)也常称为主蒸汽压力,是一个重要的运行参数.在对该参数的监测过程中发现,几次机组大修后PSG都比大修前有所降低,但随着机组的运行,PSG又逐渐恢复到正常的水平.本文针对该现象,首先对这种规律性的变化进行了总结,然后做了一些初步的分析,认为对SG的水力冲洗是造成PSG变化的主要原因,并给出一些改善和提高主蒸汽压力的方法.     

秦山核电厂蒸汽发生器的老化管理

依据蒸汽发生器(SG)老化管理的PDCA (PLAN、DO、CHECK、ACT的首个字母的缩写,可以简称为戴明循环)循环,阐述了秦山核电厂SG老化管理体系的建立、SG的运行控制(主要是水化学控制)、检查、检测和评估以及SG的维护措施.通过这些措施的实施,对SG的老化降质进行了有效的管理,确保秦山核电厂运行16 a后,SG仍处于一个良好的运行状态.     

CPR1000蒸汽发生器排污结构设计优化

对中国改进型百万千瓦级压水堆(CPR1000)蒸汽发生器(SG)排污结构进行优化。通过取消排污管及阻挡块,改为在管板上直接开排污孔,提高管廊区域的可达性,便于管板二次侧上表面的检查和泥渣冲洗。应用SG热工水力分析专用软件GENEPI,对比分析优化前后的热工水力特性。结果表明:与原设计方案相比,优化后SG热工水力性能满足设计要求,虽然管板二次侧上表面流场分布发生变化,导致发生泥渣沉积的传热管数量增加,但结构优化后有利于泥渣冲洗,提高冲洗效果。分析结果从理论上证明了优化的可行性。     

蒸汽发生器液位控制系统手自动切换的前馈补偿研究

当前蒸汽发生器（SG）液位控制系统手自动切换信号复制回路的设计中,液位控制器运算基准为切换时的汽水失配信号,主给水流量调节阀由手动模式切到自动模式后导致SG液位控制系统失去快速调节给水流量的前馈作用。针对该问题,结合阳江核电厂4号机组SG液位高高跳堆事件,提出了针对手自动切换操作方式和系统设计的2种优化方案。针对操作方式的优化,在主给水流量调节阀投自动前,手动平衡汽水流量;针对系统设计的优化,增加汽水失配判断环节和前馈自动补偿环节。通过SG液位扰动试验证明,所提出的优化方案能有效提高手自动切换后控制系统的调节速度、减小超调量,对核电机组安全运行水平提升有重要贡献。      

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。     

核电站蒸汽发生器二次侧清洗技术

核电站运行中,二回路工质中的杂质会浓缩并沉积在蒸汽发生器(SG)二次侧。SG杂质的沉积会对其热交换效率产生影响。目前,国内第一批核电站接近退役和延寿阶段,保证SG具有良好的清洁度对SG的延寿起关键作用。同时,国内某些核电站也出现SG污垢系数增大影响核反应堆满功率运行的情况。为此,参照国内外SG清洗工艺对机械清洗、化学清洗和鼓泡清洗进行定性分析,致力于比较得出一种适合我国SG特点的清洗工艺。     

Predicting the accumulated number of plugged tubes in a steam generator using statistical methodologies

A steam generator (SG) plays a significant role not only with respect to the primary-to-secondary heat transfer but also as a fission product barrier to prevent the release of radionuclides. Tube plugging is an efficient way to avoid releasing radionuclides when SG tubes are severely degraded. However, this remedial action may cause the decrease of SG heat transfer capability, especially in transient or accident conditions. It is therefore crucial for the plant staff to understand the trend of plugged tubes for the SG operation and maintenance. Statistical methodologies are proposed in this paper to predict this trend. The accumulated numbers of SG plugged tubes versus the operation time are predicted using the Weibull and log–normal distributions, which correspond well with the plant measured data from a selected pressurized water reactor (PWR). With the help of these predictions, the accumulated number of SG plugged tubes can be reasonably extrapolated to the 40-year operation lifetime (or even longer than 40 years) of a PWR. This information can assist the plant policymakers to determine whether or when a SG must be replaced.     

Sensitivity analysis of the MASLWR helical coil steam generator using TRACE

Accurate simulation of transient system behavior of a nuclear power plant is the goal of systems code calculations, and the evaluation of a code's calculation accuracy is accomplished by assessment and validation against appropriate system data. These system data may be developed either from a running system prototype or from a scaled model test facility, and characterize the thermal hydraulic phenomena during both steady state and transient conditions. The identification and characterization of the relevant thermal hydraulic phenomena, and the assessment and validation of thermal hydraulic systems codes, has been the objective of multiple international research programs. The validation and assessment of the best estimate thermal hydraulic system code TRACE against the Multi-Application Small Light-Water Reactor (MASLWR) Natural Circulation (NC), helical coil Steam Generator (SG), Nuclear Steam Supply System (NSSS) design is a novel effort, and is the topic of the present paper. Specifically, the current work relates to the assessment and validation process of TRACE code against the NC database developed in the OSU-MASLWR test facility. This facility was constructed at Oregon State University under a U.S. Department of Energy grant in order to examine the NC phenomena of importance to the MASLWR reactor design, which includes an integrated helical coil SG. Test series have been conducted at this facility in order to assess the behavior of the MASLWR concept in both normal and transient operation and to assess the passive safety systems under transient conditions. In particular the test OSU-MASLWR-002 investigated the primary system flow rates and secondary side steam superheat, used to control the facility, for a variety of core power levels and Feed Water (FW) flow rates. This paper illustrates a preliminary analysis, performed by TRACE code, aiming at the evaluation of the code capability in predicting NC phenomena and heat exchange from primary to secondary side by helical SG in superheated condition and to evaluate the fidelity of various methods to model the OSU-MASLWR SG in TRACE. The analyses of the calculated data show that the phenomena of interest of the OSU-MASLWR-002 test are predicted by the code and that one of the reasons of the instability of the superheat condition of the fluid at the outlet of the SG is the equivalent SG model used to simulate the different group of helical coils. The SNAP animation model capability is used to show a direct visualization of selected calculated data.     

Thermal hydraulic behaviors during loss of RHR system at mid-loop operation of Chinese 300 MWe PWR NPP

Thermal hydraulic behavior of nuclear power plant (NPP) is analyzed by using mechanistic computer code for loss of residual heat removal (RHR) system during mid-loop operation of Chinese 300 MWe two-loop pressurized water reactor is presented. In the absence of recovery of RHR or other accident management measures, the reactor core will be uncovered for a long term resulting in core heat-up, degradation and relocation to the lower plenum. The effectiveness of available mitigate measures, such as safety injection system, gravity feed from refueling water storage tank (RWST) and steam generator (SG) reflux-condensation, are investigated. Coolant injection is highly effective in halting the accident progression and make the core recovered. The cooling capability of SG reflux-condensation has a relationship with different availabilities of steam generators and decay heat power. 6 days after shutdown, 2SG operation can keep the water level at mid-line of hot leg. 12 days after shutdown, both 2SG operation and 1SG operation can keep the water level at mid-line of hot leg. The analyses also indicate that the cooling mechanism of safety injection system is more effective than gravity feed from RWST and SG reflux-condensation. Through confirming the success criteria of SG reflux-condensation, time windows can be devided. Then, event trees for loss of RHR system under mid-loop operation are built with considering the analysis results and abnormal procedure.     

重水堆核电站主蒸汽旁排阀故障分析和应对策略

根据重水堆电站的电厂控制模式及蒸汽发生器压力控制的原理和特点,结合实际事例对主蒸汽旁排阀故障开启对机组运行的影响及人员响应进行了分析和探讨,总结了处理类似事件的原则和经验.     

Experimental and Operational Verification of the HTR-10 Once-Through Steam Generator (SG)

In January 2003, the 10MW High-temperature Gas-cooled Reactor (HTR-10) reached its full power for continuous operation of seventy-two hours in the Institute of Nuclear Energy Technology, Tsinghua University. The reactor was operated smoothly at the designated parameters. The once-through steam generator (SG) is one of key equipments of the HTR-10 reactor. The SG includes 30 modular heating helical tube assemblies. Design of the SG includes hydraulics, heat transfer and stability designs. Based on the design requirement, it is necessary to ensure sufficient heat removal from the reactor in order to maintain stable operation. In order to confirm the thermal hydraulic reliability of the SG, a series of experiments had been carried out. The purpose of this paper is to introduce the design features and experimental verification of HTR-10 SG, and the research results of small bending radius helical coil-pipe used in HTR-10, for example, the heat transfer coefficient of water, superheat steam and the two phase flow in the helical tube, the heat transfer coefficient of the He flow across the helical tube, and the centrifugal force effect on the heat transfer for the helical tube. In the paper, some operational experimental data of the HTR-10 SG have been presented.     

基于SPSA的蒸汽发生器液位MPC系统性能优化方法研究

核电厂蒸汽发生器（SG）液位变化过程具有强非线性且存在“虚假水位”现象,传统SG液位控制系统多采用固定参数比例-积分-微分（PID）控制器,但传统PID控制方法不具备自优化、自适应、自学习等能力,使得控制系统性能难以达到并保持最佳。为提高机组瞬态响应能力以及核电厂的稳定性、安全性和经济性,提出了一种基于并行摄动随机逼近（SPSA）算法的模型预测控制（MPC）方法。该方法采用MPC系统代替传统PID控制系统,并利用SPSA实现液位控制系统参数的整定优化,从而实现SG液位控制系统的性能优化。通过仿真试验验证了本方法能够有效提高SG液位控制参数的整定效率以及控制系统稳定性。     

SGTR事故SG满溢分析扩展研究

采用热工水力系统程序进行核电厂蒸汽发生器传热管破裂（SGTR）事故蒸汽发生器（SG）满溢分析,验证在该事故下SG不会发生满溢;对SGTR事故进行扩展研究,考虑多种传热管破裂情况,包括单根传热管双端断裂、多根传热管双端断裂和传热管破口,并将3种情况的分析结果进行比较,给出SGTR事故最极限的工况。研究结果表明,单根传热管双端断裂工况下,SG不会发生满溢,且与其他2种工况相比满溢裕量最小,在所有分析工况中最极限。      

蒸汽发生器传热管的降质及对其完整性的评估

蒸汽发生器传热管是反应堆冷却剂压力边界的主要组成部分，这就意味着必须保持传热管的完整性。然而，运行经验表明，蒸汽发生器传热管会出现各种降质。这些降质可能会导致管子的泄漏或破裂，使反应堆冷却剂丧失，并提供了直接通向二回路和释放到环境中去的途径。本文将介绍几种已知的传热管降质，传热管完整性性能准则．并对蒸汽发生器传热管完整性进行评估。     

高温气冷堆二回路水质控制方法研究

参考压水堆二回路和火电直流炉水质标准,结合高温气冷堆二回路结构和材料特点,对高温气冷堆蒸汽发生器（SG）原设计中的进水标准和精处理出水标准进行了修正,得出比较合理的水质控制标准;研究了提高高温气冷堆二回路水-汽品质的3种方法,即机组启动冲洗方式控制、加药控制和精处理运行方式控制,对3种方法进行了详细的论述,提出了科学、完整的高温气冷堆二回路水-汽品质优化方法。      

基于模糊控制的核电蒸汽发生器水位的串级自抗扰控制

针对核电站蒸汽发生器所存在的非线性,时变性,大时滞等特点,本文提出了基于模糊控制的蒸汽发生器水位的串级自抗扰控制方案。该方案采用双闭环控制,内环采用带前馈的一阶线性自抗扰控制调节阀,并分别前馈补偿蒸汽扰动和给水扰动,外环采用二阶模糊自抗扰,设计了新型的幂次控制率。仿真结果表明,该控制方案对蒸汽发生器水位具有良好的控制效果,与串级ADRC-PID控制系统相比,不仅具有优良的鲁棒性和抗干扰能力,而且具有可行性。