草金鱼C型起动流场的PIV测量

鱼类的C型快速起动常常用于逃脱危险的过程中,是鱼类最重要的机动能力之一。本文利用二维粒子图像测速(Particle Image Velocity,PIV)技术对草金鱼C型起动过程中的鱼体周围流场进行了测量和分析。结果表明,在第三阶段为减速静止的情况下,第二阶段的加速过程很短,而第一阶段的加速更为明显,第一阶段与第二阶段的时长比为1:5.25;草金鱼的C型起动过程主要依赖附加质量效应产生推力。C型起动的尾迹流场具有逆卡门涡街的特征,较大的身体弯曲度使得鱼在舒展摆尾的过程中连续产生了方向相同的两个涡,而且涡在脱离鱼体后由于受到鱼后续行为和高速射流区的影响,其强度经历了先增大再降低的过程。     

枢轨接触面形貌对电枢起动特性的影响

在脉冲大电流直线驱动装置中,电枢和轨道的接触状态会改变电枢起动特性,而电枢起动过程将直接影响整个发射系统的效率和寿命,因此有必要对电枢起始阶段的滑动接触状态进行研究。搭建发射实验平台,通过高速相机观察电枢起动状态,并结合有限元软件ANSYS,对电枢的预紧力、初始接触状态以及电磁压力、电流密度进行仿真分析,研究电枢表面形貌对电枢起动的影响。结论表明:开槽电枢增加了电枢本身的柔顺性,使得预紧力增大,同时由于电流趋肤效应使得电流密度分布更加均匀,从而电磁压力增大,电枢起动变慢,接触电阻变小。实验和仿真结果对于改善电枢起动过程的接触状态,减轻烧蚀具有重要意义。     

用相似理论对单轴燃气轮机起动工况的分析

燃气轮机起动是机组可靠性的主要指标之一。本文用相似理论分析了大气参数变化和起动设备更换时对起动工况的控制方法。其结论与单轴机组长期运行的数据基本一致。 一、单轴燃气轮机的相似工况和理想起动过程 燃气轮机起动过程中的燃料控制对起动的成败有决定性的作用。很多机组随着压气机、透平气动性能变差和起动设备的出力下降使起动问题日趋尖锐。尤其是在气温较高的夏季中午,往往是起动时间延长,产生“热挂”,甚至起动失败。单轴燃气轮机运行工况     

低风速条件下沙粒起动的PIV研究

本文在对风沙流进行气固两相流分析的基础上实现了低风速条件下的沙粒起动过程,借助PⅣ测量系统对其进行了观测并对影响因素进行了分析,获得了沙粒的浓度和速度分布。实验结果表明沙粒浓度沿距离沙面的高度呈负指数分布特征。流体起动和冲击起动下的输沙率随摩阻速度和雷诺数的增加而单调增加,而起动效能比随之减小并趋向于常数0.207。测量得到的输沙率修正了低速范围内的Bagnold公式和河村龙马公式。     

多孔介质干燥过程传热传质研究

一、过程分析及数学模型的建立 多孔介质的干燥过程一般可以分为三个阶段:预热阶段、恒速率干燥阶段和降速率干燥阶段.降速率干燥阶段通常又分为第一、第二降速率干燥阶段. 预热阶段一般很短,对过程影响不大,本文不考虑它的存在,只研究后面几个阶段. 根据本文的研究范围,对问题作如下简化: (1)被干燥材料存宏观结构上是统计均匀的,且在干燥过程中不发生体积变化.     

某型航空涡轴发动机起动过程控制研究

针对涡轴发动机起动过程复杂的特点,提出了“跟踪-微分器”的二阶动态结构来对其加速过程进行控制;首先,在对涡轴发动机起动过程分析的基础上,给出了其起动控制逻辑;其次,通过对起动过程主燃油供油系统规律分析,给出了起动过程的供油计划,并采用“跟踪-微分器”的二阶动态结构来求取转速加速度;最后,对所研究的涡轴发动机起动过程控制规律进行了仿真验证;仿真结果表明,所设计的过程控制方法满足起动时限的要求,且不超温超转,达到了系统设计要求。     

CCD扩展板

Tektronix公司的TK1/DEV CCD扩展板使CCD阵列具有通用性,可起动达到3072×3072个象素的单输出CCD,另外,还易于调整不同CCD的尺寸、外部计时和信号处理。扩展板起动全帧和帧转换器件,并可选择S/H、X-Y-Z视频或放大的CCD信号输出。用于直接成象时,它只需要一个电源和X-Y-Z监视器。     

含湿毛细多孔介质干燥过程相变传热传质分析

分析了含湿毛细多孔介质干燥过程的主要机理,建立了以液相饱和度、温度和气体压力为参数的一维数学模型, 采用全隐式有限差分方法对该模型进行了数值计算。计算结果表明,干燥过程可分为两个阶段:不稳定阶段和稳定阶段。 在不稳定阶段,模拟参数变化剧烈,而在稳定阶段,模拟参数变化平稳。     

综合传动装置起动扭矩测试试验研究

低温环境起动性能是表征车辆动力性能的一项重要指标。对液力机械传动车辆而言,由于综合传动装置自身技术特点,改善车辆起动性能除应当考虑发动机自身起动阻力外,还应当考虑综合传动装置的起动阻力。在系统分析综合传动装置技术特点和功率流向的基础上,总结了发动机起动过程的主要特征,提出采用稳态试验与动态试验相结合的台架试验方法进行综合传动装置起动扭矩测试。以某高速履带车辆为研究对象,进行了综合传动装置起动扭矩测试。试验结果表明,传动油温的变化和各类油泵的功率消耗是决定综合传动装置起动扭矩大小的关键因素,为改进车辆总体设计和提高车辆起动性能提供了依据。     

基于矿物质定量热分析的灰熔融反应动力学

利用矿物质定量热分析理论对三种典型煤样灰熔融过程进行了详细分析,借助化学反应动力学相关概念,计算了灰样熔融反应动力学参数,结果表明:灰样熔融过程呈明显的三阶段反应过程,即初始熔融阶段、矿物共熔阶段以及熔融末期;熔融初始阶段反应活化能均大于后面两个阶段,矿物共熔过程活化能较初始阶段明显下降;在理论上证明了灰熔融逐渐加剧的过程。     

Plasma-facing materials for fusion devices

In a future D/T fusion reactor the walls of the vessel containing the magnetically confined hot plasma have to stand simultaneously very high power, particle and neutron loads. In today’s high temperature plasma experiments at the areas of the highest load, i.e. the divertor and the limiters, W, Mo and Carbon (CFC) are used and Be, W, Mo, Inconel and stainless steel are at the other wall areas. These materials are also envisaged for future bigger fusion experiments, such as ITER [1–3]. The resistance of these materials to the different expected higher loads in a fusion reactor is only partly known and more investigations are needed with respect to find better materials and/or a modification of the divertor.     

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研制出了用于计算氚投料量在FEB聚变堆各个子系统中的分布及其随时间变化的数值模拟程序包SWITRIM。通过近5年的使用，表明其运行良好、计算结果可靠。用SWITRIM数值模拟研究了聚变堆起动过程中的“氚坑深度和氚坑时间”新现象。简单介绍了SWITRIM程序包的组成和用户使用说明以及最新的运用等。     

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In part one published in the last issue, the tritium retention and the total tritium inventory in PFC materials of FEB-E fusion reactor had been calculated. The tritium well depth, tritium well time during the FEB-E fusion reactor start-up and initial operation phase had been obtained. In this part, how to improve tritium recovery efficiency in the ITER TBM solid breeder blanket with using purge gas has been discussed. Some new innovative schemes for reducing tritium retention and improving tritium recovery efficiency are proposed. Such as, sponge mechanism based on deuterium saturated PFC materials; deuterium and beryllium co-deposition layer created on first wall surface; SPB scheme for enhancing tritium recovery efficiency of purge gas in ceramic breeder blanket based on the electrical polarization rotations catalyzing isotope exchange rate enhancement resulted from applied low frequency electric-field, of Li4SiO4 grain and purge gas molecular particles and so on, are explored.     

聚变堆氚系统设计中的一些重要问题研究(Ⅱ)

在文献[1]中,计算了FEB-E 聚变堆PFC 材料内的氚滞留量、堆系统总的氚投料量、启动运行开始阶段的氚坑深度和氚坑时间大小。这里将讨论在ITER 的TBM 氚增殖包层内固体氚增殖剂中的氚如何高效率地被载氚气体带出并且以高效率地提取回收。本部分将进行创新的探索性研究并且提出某些减少氚滞留量和改善氚提取回收效率的新方案,例如：基于氘饱和的海绵效应;第一壁表面建立氘和铍的伴同沉积层;基于在低频外电场作用下载氚气分子和硅酸锂颗粒电极化旋转催化同位素交换速率的增强载氚气提取氚效率“SPB 方法”。     

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聚变堆第一壁表面和PFC材料内的氚滞留量、堆系统总的氚投料量多高？在启动和运行的开始阶段的氚坑深度,氚坑时间的大小是多少？在TBM氚增殖包层内固体氚增殖剂中的氚能否高效率地被载氚气体带出来并且以高效率地提取回收？能否找到某些新机制解决这些问题是决定实现ITER的预期目标和最终实现聚变能的实际运用成败的关键问题。本文第(Ⅰ)部分回答前面两个问题,在下期第(Ⅱ)部分将进行创新的探索性研究并且提出某些减少氚滞留量和改善氚提取回收效率的新方案,例如：基于氘饱和的海绵效应;第一壁表面建立氘和铍的伴同沉积层;基于在低频外电场作用下载氚气分子和硅酸锂颗粒电极化旋转催化同位素交换速率增强提高载氚气提取氚效率“SPB方法”等等。     

Tritium Burn-up Depth and Tritium Break-Even Time

Similarly to but quite different from the xenon poisoning effects resulting from fission-produced iodine during the restart-up process of a fission reactor, we introduce a completely new concept of the tritium burn-up depth and tritium break-even time in the fusion energy research area. To show what the least required amount of tritium storage is used to start up a fusion reactor and how long a time the fusion reactor needs to be operated for achieving the tritium break-even during the initial start-up phase due to the finite tritium breeding time that is dependent on the tritium breeder, specific structure of breeding zone, layout of coolant flow pipe, tritium recovery scheme, extraction process, the tritium retention of reactor components, unrecoverable tritium fraction in breeder, leakage to the inertial gas container, and the natural decay etc., we describe this new phenomenon and answer this problem by setting up and by solving a set of equations, which express a dynamic subsystem model of the tritium inventory evolution in a fusion experimental breeder （FEB）. It is found that the tritium burn-up depth is 317g and the tritium break-even time is approximately 240 full power days for FEB designed detail configuration and it is also found that after one-year operation, the tritium storage reaches 1.18kg that is more than the least required amount of tritium storage to start up three of FEB-like fusion reactors.     

中国聚变工程实验堆启动氚投料量与氚平衡分析

根据中国聚变工程实验堆(CFETR)堆芯设计参数及燃料系统流程模型，采用平均停留时间方法，建立燃料循环子系统的氚转移模型用来描述氚在各子系统之间的输运、滞留等过程。采用该模型，分析了不同聚变功率水平、运行因子以及燃烧率对中国聚变工程实验堆的氚平衡以及启动氚投料量的影响。     

Fuel cycle for a fusion neutron source

The concept of a tokamak-based stationary fusion neutron source (FNS) for scientific research (neutron diffraction, etc.), tests of structural materials for future fusion reactors, nuclear waste transmutation, fission reactor fuel production, and control of subcritical nuclear systems (fusion–fission hybrid reactor) is being developed in Russia. The fuel cycle system is one of the most important systems of FNS that provides circulation and reprocessing of the deuterium–tritium fuel mixture in all fusion reactor systems: the vacuum chamber, neutral injection system, cryogenic pumps, tritium purification system, separation system, storage system, and tritium-breeding blanket. The existing technologies need to be significantly upgraded since the engineering solutions adopted in the ITER project can be only partially used in the FNS (considering the capacity factor higher than 0.3, tritium flow up to 200 m³Pa/s, and temperature of reactor elements up to 650°C). The deuterium–tritium fuel cycle of the stationary FNS is considered. The TC-FNS computer code developed for estimating the tritium distribution in the systems of FNS is described. The code calculates tritium flows and inventory in tokamak systems (vacuum chamber, cryogenic pumps, neutral injection system, fuel mixture purification system, isotope separation system, tritium storage system) and takes into account tritium loss in the fuel cycle due to thermonuclear burnup and β decay. For the two facility versions considered, FNS-ST and DEMO-FNS, the amount of fuel mixture needed for uninterrupted operation of all fuel cycle systems is 0.9 and 1.4 kg, consequently, and the tritium consumption is 0.3 and 1.8 kg per year, including 35 and 55 g/yr, respectively, due to tritium decay.     

Radiation effect of neutrons produced by D–D side reactions on a D–3He fusion reactor

One of the most important characteristics in D–³He fusion reactors is neutron production via D–D side reactions. The neutrons can activate structural material, degrading them and ultimately converting them into high-level radioactive waste, while it is really costly and difficult to remove them. The neutrons from a fusion reactor could also be used to make weapons-grade nuclear material, rendering such types of fusion reactors a serious proliferation hazard. A related problem is the presence of radioactive elements such as tritium in D–³He plasma, either as fuel for or as products of the nuclear reactions; substantial quantities of radioactive elements would not only pose a general health risk, but tritium in particular would also be another proliferation hazard. The problems of neutron radiation and radioactive element production are especially interconnected because both would result from the D–D side reaction. Therefore, the presentation approach for reducing neutrons via D–D nuclear side reactions in a D–³He fusion reactor is very important. For doing this research, energy losses and neutron power fraction in D–³He fusion reactors are investigated. Calculations show neutrons produced by the D–D nuclear side reaction could be reduced by changing to a more ³He-rich fuel mixture, but then the bremsstrahlung power loss fraction would increase in the D–³He fusion reactor.     

Level densities for basic and applied nuclear physics

Nuclear level densities are a key ingredient in compound nuclear reaction calculations. They are, therefore, of importance to fields as diverse as fission and fusion reactor design and astrophysics. A review of present methods for calculating level densities is presented, which includes global approximation methods and rotational enhancement factors. New information about level density parameter systematics should be forthcoming from Hartree-Fock and shell model methods. The strengths and limitations of these techniques for calculating level densities are summarized. It is concluded that more work on both empirical and microscopic models is needed to improve our ability to calculate level densities for nuclei off the stability lines.