多功能小型铅铋堆氧分析探头设计研究

作为第四代核能系统国际论坛技术路线反应堆之一,小型铅铋冷却剂块堆可为城市供热、紧急情况下移动供电、工业区热源等提供一种可靠经济的选择,因此近年来发展迅速。这种小型反应堆的冷却剂为铅铋合金,高温液态铅铋合金具有熔点低、沸点高、导热性能好、化学活性低等优良的物理与化学性能,是良好的快中子反应堆冷却剂材料,具有较高的安全性。然而液态铅铋对结构材料的腐蚀问题是制约铅铋冷却反应堆发展的关键因素之一,其中氧含量的控制是解决材料腐蚀的核心。为此设计一种适合于测量高温液态铅铋合金氧含量的探头,以便为小型铅铋块堆发展提供解决氧控的方案。     

铋对铅钙合金腐蚀行为的影响

通过恒电流阳极失重法,扫描电镜观察和腐蚀电阻测定,研究了秘对Pb—Ca—Sn—Al合金耐蚀性能的影响。结果表明,在0.0025w％～0.0648w％范围内,随着铋含量增加,合金耐蚀性能提高,腐蚀更为均匀,腐蚀层与基体金属之间断裂减少。     

铋含量对铅铋合金析氧行为的影响

制备了铋含量为0~7.33%(质量百分数)的铅铋合金。用线性电位扫描、交流阻抗法以及气体收集实验研究了铅铋合金在硫酸溶液中的析氧行为,确定了相关的电化学动力学参数。研究结果表明,当铅铋合金中铋含量小于0.100%时,铋的存在对析氧反应几乎无影响,而当铋含量大于0.83%时,铋加快氧气的析出,且随铋含量增加,析氧过电位降低,电荷交换反应电阻减小,析氧量增加。     

液态铅铋合金热物性程序开发研究

液态铅铋合金是加速器驱动次临界系统(ADS)重要的散裂靶材料和冷却剂候选材料,其热力学物理性质是ADS研发过程中必须解决的基础问题。以Fortran语言为基础编制的程序NCEPU-LBETP,实现了对液态铅铋合金熔沸点、密度、比热容、动力粘度和热导率的计算,且计算步序大大简化,精度得到提高。并对液态铅铋合金热物性程序进行了校核,其结果和国内外相关计算结果符合良好,这表明物性程序是成功的。     

铅铋合金在H2SO4溶液中析氧反应研究

采用线性电位扫描法、气体收集实验以及电化学阻抗法研究了纯铅电极和铅铋[w(Bi)=0.05%]合金电极在硫酸溶液中的析氧反应.阳极极化曲线测量表明,相同电位下,纯铅电极上的析氧电流稍小于铅铋合金电极上的析氧电流;氧气收集实验.表明,相同电位下,纯铅电极上收集到的氧气的体积稍小于铅铋合金电极;交流阻抗实验表明,在两种电极上的析氧反应均受电荷传递控制,且相同电位下,纯铅上的析氧反应电阻略大于铅铋合金电极上的反应电阻.铅铋合金中0.05%铋的存在对析氧反应有催化作用,但影响不大.因而,适量铋的存在并不会增大电池的水损失.     

超临界二氧化碳与液态铅铋合金耦合换热模型与换热特性研究

铅冷快堆和超临界二氧化碳（S-CO2）布雷顿循环因其高的热效率、紧凑式设计被认为是 最具前景的发电系统之一。液态铅铋合金（Pb/Bi）和S-CO2在中间换热器耦合换热,然 而2种流体的湍流输运特性与耦合传热行为相较于常规流体差异巨大,常规湍流模型无法准确预测其耦合换热性能。为获得两者间耦合传热的准确预测模型、掌握耦合传热规律,首先针对管内液态Pb/Bi和S-CO2的湍流普朗特数（Prt）模型的适用性进行了比较分析,发现冷、热侧分别选择Tang、Cheng和Tak提出的Prt模型可获得准确结果;对两者的耦合换热模型进行了深入分析与校验,模拟结果与实验数据吻合良好;全面探讨了雷诺数、工质温度对2种特殊流体耦合传热能力的影响,发现套管式换热器热阻主要存在于S-CO2侧,提升S-CO2侧参数可以迅速提高传热性能,且当S-CO2工质温度处于拟临界区时换热器的换热能力将大幅增强。     

铅铋合金在H2SO4溶液中的析氢反应

采用线性电位扫描法、气体收集实验以及电化学阻抗法研究了纯铅电极和铅铋(0.05%Bi)合金电极在硫酸溶液中的析氢反应。阴极极化曲线测量表明,相同电位下,纯铅电极上的析氢电流大于铅铋合金电极上的析氢电流;氢气收集实验表明,相同电位下,纯铅电极上收集到的氢气的体积大于铅铋合金电极;交流阻抗实验表明,在两种电极上的析氢反应均受电荷传递控制,相同电位下,纯铅上的析氢反应电阻小于铅铋合金电极上的反应电阻。铅铋合金中铋的存在对析氢反应起抑制作用,且电位越负,铋的抑制作用越明显。     

用于水下航行器的铅铋堆S-CO_(2)循环发电系统热力学及动态特性分析EI北大核心CSCD

将铅铋堆与超临界二氧化碳(supercriticalcarbon dioxidecycle,S-CO_(2))循环相结合的发电系统可为水下航行器稳定、高效供能。然而,目前缺乏水下航行器与铅铋堆S-CO_(2)循环发电系统的匹配设计研究,其热力学分析仍不完善。因此,该文分别构建其热力学模型和动态模型,通过模拟仿真,开展发电系统与4种动力循环形式的匹配研究,发现简单回热循环与铅铋合金的温度匹配性较好,系统热效率在压缩机和透平入口压力分别为7.6和25MPa时达到26.81%,满足系统设计要求。进一步探讨循环关键运行参数对系统热力学性能的影响规律,结果表明,增大循环最高压力和最高温度有利于系统效率的提升。基于此,进行系统关键换热部件的设计,揭示发电系统在温度阶跃扰动条件下的动态响应特性。可知,当堆芯铅铋温度从410℃降至390℃时,循环效率在大约8s内下降1.69%。该文可为应用于水下航行器的铅铋堆超临界二氧化碳循环发电系统设计提供参考。     

非对称紧凑式换热器内超临界二氧化碳与液态铅铋合金耦合换热特性研究

超临界二氧化碳布雷顿循环与铅冷快堆的结合被认为是最为理想的动力循环之一,系统通过中间换热器传递热量,其性能影响着整个发电系统的高效与安全运行。由于超临界二氧化碳和液态铅铋合金（LBE）物理性质和热输运性质差异显著,对称式结构无法匹配两侧工质的换热要求,构建了1种非对称式紧凑式耦合换热器,采用数值模拟方法研究了超临界二氧化碳与液态铅铋合金耦合换热特性。结果表明：提升冷侧流体入口速度会显著增强换热;增加热侧LBE入口速度时,总换热系数先降低后增加;提升换热器冷热流体入口温度,换热器的换热系数先减小后增大,存在最优值;在拟临界区内,强浮力作用会大幅提升冷侧换热,而加速效应则抑制换热。     

铅铋合金上生长的阳极Pb(Ⅱ)膜的性质

应用交流伏安法、线性电位扫描法和电化学阻抗频谱法分别研究了铅和铅铋合金在4.5mol·L-1H2SO4溶液中以0.9V(vs.Hg/Hg2SO4)生长的阳极膜的性质,测得了阳极膜的阻抗实数部分随电位的变化曲线和阳极膜的生长速率。结果表明:铅和铅铋合金在0.9V下生长的阳极膜的主要组成为Pb(Ⅱ)(PbO PbO·PbSO4)和PbSO4。Pb和Pb 1.0%(w)Bi合金电极上Pb(Ⅱ)电量随膜生长的增长率分别为81.2mC·cm-2·h-1和61.8mC·cm-2·h-1,膜中PbO的电阻随膜生长时间的增长率分别为26.2Ω·cm2·h-1和19.2Ω·cm2·h-1。由此可见,在铅合金中添加铋,不仅可以降低阳极腐蚀层Pb(Ⅱ)氧化物的电阻,亦可较显著地抑制阳极Pb(Ⅱ)膜的生长,从而改善板栅/腐蚀层界面电阻和抑制铅合金的阳极腐蚀。     

钠冷快堆发展综述

作为第4代核电堆型之一的钠冷快堆,能更有效地利用铀资源并能嬗变长寿命放射性废物,因此得到了世界各国的重视。本文介绍了国际上钠冷快堆的发展规划以及钠冷快堆的安全性和运行情况。     

The EUCLID/V1 Integrated Code for Safety Assessment of Liquid Metal Cooled Fast Reactors. Part 1: Basic Models

The article describes the basic models included in the EUCLID/V1 integrated code intended for safety analysis of liquid metal (sodium, lead, and lead-bismuth) cooled fast reactors using fuel rods with a gas gap and pellet dioxide, mixed oxide or nitride uranium–plutonium fuel under normal operation, under anticipated operational occurrences and accident conditions by carrying out interconnected thermal–hydraulic, neutronics, and thermal–mechanical calculations. Information about the Russian and foreign analogs of the EUCLID/V1 integrated code is given. Modeled objects, equation systems in differential form solved in each module of the EUCLID/V1 integrated code (the thermal–hydraulic, neutronics, fuel rod analysis module, and the burnup and decay heat calculation modules), the main calculated quantities, and also the limitations on application of the code are presented. The article also gives data on the scope of functions performed by the integrated code’s thermal–hydraulic module, using which it is possible to describe both one- and twophase processes occurring in the coolant. It is shown that, owing to the availability of the fuel rod analysis module in the integrated code, it becomes possible to estimate the performance of fuel rods in different regimes of the reactor operation. It is also shown that the models implemented in the code for calculating neutron-physical processes make it possible to take into account the neutron field distribution over the fuel assembly cross section as well as other features important for the safety assessment of fast reactors.     

The Challenging Thermophysical Issues of Fast Reactors

The article presents and analyzes the results from topical thermophysical investigations aimed at substantiating the characteristics and assessing the safety of liquid metal cooled fast reactors: an advanced largecapacity fast sodium-cooled reactor and a fast lead-cooled reactor. We also outline the results from experimental investigations into hydrodynamics and heat transfer for stratified flow of coolant, into thermal hydraulics of a large-modular “sodium–water” steam generator in different modes of reactor operation, and into the fuel assembly degradation process in the course of a severe accident involving loss of sodium flowrate in a fast reactor. The basic mechanisms governing degradation of dummy fuel rod claddings are identified; the distribution of marker materials over the fuel assembly height in its ultimate state is estimated, and the phenomena of blocking the fuel assembly flow pass section and the marker materials being thrown to beyond the assembly boundaries are studied. The article demonstrates the advisability of using a combined sodium purification system built into the reactor vessel that contains cold traps as a mandatory element along with hot traps, which serve to perform accelerated removal of oxygen during the NPP operation at its nominal parameters. The results from thermal–hydraulic tests of the lead-cooled reactor core carried out on the 6B experimental setup and of the steam generator carried out on the SPRUT experimental setup installed at the SSC RF–IPPE are presented.1 It has been found from the experiments on the thermal–hydraulic model of the lead-cooled reactor’s steam generator that the steam temperatures at the outlets from both headers are identical with each other and so are the lead temperatures at the downcomer section outlet and in the main lead path. The experiments did not show pulsations of feedwater flowrate or pressure in the loops, which points to the stable nature of operation at partial load. The state of the technology of heavy metal coolants and the prospects for its future development are considered. It is shown that there is a possibility in principle to realize the required parameters of a high-temperature sodium-cooled fast reactor for producing large quantities of hydrogen, e.g., based on one of the thermochemical cycles or high-temperature electrolysis with a high electricity thermal utilization factor. The problems that must be solved during the further thermophysical investigations are analyzed.     

Parameter Sensitivity Analysis for Line-Start Permanent Magnet Motors

ABSTRACT

A comprehensive study of the parameter sensitivity on both the synchronous and asynchronous performances of permanent magnet machines is presented. Based on the 2-axis theory for permanent magnet motor, performance equations are derived In terms of normalized machine parameters. Such an analysis can be used successfully in design optimization of permanent magnet machines.

A normalized power equation for evaluating the steady-state stability limits of such type of motor is also given. Simulation results obtained are confirmed with experimental results using a number of different rotor designs. Useful conclusions have been provided, which are considered to be a good aid for designing any permanent magnet motor.     

基于参数灵敏度分析的负荷建模研究

过于复杂的负荷模型容易使辨识寻优落入局部最优,而且难以实用化。减少复杂结构负荷模型的参数是解决问题的有力手段。本文首先提出一种衡量负荷模型响应变化的量化指标,采用摄动法分析了改进综合负荷模型参数对负荷响应变化指标的灵敏度,得到灵敏度较大的参数。结合实际电网进一步分析了负荷模型对电力系统仿真计算结果的灵敏度。发现对负荷响应灵敏度较小的部分参数对电力系统仿真计算结果的灵敏度较大。然后提出基于灵敏度分析的负荷模型参数减少的方法。最后,使用某大区电网的实测数据验证了该方法的有效性。     

钠冷快堆高温钠管道应力分析与评定

本文简要介绍了RCC—MR标准的发展历程，简要说明了采用RCC—MR“RB3600管道设计标准”对高温钠管道进行应力分析与计算的步骤；作为实例，运用CASTEM2000计算程序对一个在工作寿期内承受高温热循环载荷的钠管道进行详细的应力分析与评定，以及在整个工作寿期内的总体应变和疲劳-蠕变分析。     

轨迹灵敏度分析下的TCSC参数优化设计

提出了一种基于时域仿真的可控串联电容补偿(TCSC)及其附加阻尼控制器的参数优化方案。该方案以表征系统振荡相对动能的发电机角速度的平方最小为目标函数,通过轨迹灵敏度仿真得到目标函数的梯度信息,再应用共轭梯度法优化TCSC阻尼控制器参数。数字仿真表明,在发电机振荡模式和区域间振荡模式下,可有效提高大区域系统的动态稳定性。     

同塔多回输电线路并联电抗器的参数优化分析

同塔多回输电线路间的电磁耦合使得并联电抗器与中性点小电抗的取值需重新优化。建立了同塔多回输电线路的分布参数耦合模型,得出潜供电流与弧道恢复电压的表达式,并可推广至多重故障情况;研究了影响潜供电流和恢复电压的关键因素。基于潜供电弧的动态模型,计算了电弧起始长度、中性点小电抗等对燃弧时间的影响,进而得到并联电抗器的取值准则,指出单回与多回线路中性点小电抗优化取值的差异之处。采用谐振频率分析法,综合考虑非全相运行、回路间电磁感应等引起的谐振过电压,导出了多回线路的固有谐振频率简化表达式,指出中性点小电抗的取值范围与取值方法。研究结果对超／特高压多回输电线路并联电抗器的优化设计具有指导价值。     

基于轨迹灵敏度的励磁系统参数可辨识性分析

针对在有限测点下励磁系统部分参数关联形成隐函数导致辨识结果不稳定的情况,提出一种关联参数的辨识方法。首先,利用奇异值分解选取子列集的方法将励磁系统参数划分为关联参数集和非关联参数集。其次,为降低对关联参数赋经验值消除关联性时,经验值与实际真值存在偏差对系统参数辨识结果精度的影响,选择灵敏度较小的参数作为关联参数代表赋值。为避免选取参数代表时重新引入关联参数,需进一步明确每个参数间的关联关系。为此,构造轨迹灵敏度矩阵求出判别系数,根据判别系数将相关联的系统参数划分至几个独立的隐函数参数集。仿真结果表明,该方法不仅能够有效判别励磁系统参数的可辨识性,而且可以尽可能地降低由于赋值偏差而对降阶后系统参数的辨识精度影响。