DN300钠阀冷冻密封腔室的分析与设计

钠阀的冷冻密封腔室是一个重要的部件,直接关系到钠阀是否会发生泄漏。本文以大口径钠阀的国产化项目为依托,对冷冻密封腔室进行分析与设计。冷冻密封腔室是一个狭小的空间,内部有氩气、液态钠、凝固钠。本文利用FLUENT的冷凝模型结合氩气的基本气体方程,计算冷冻密封腔室内部钠的冷凝情况,确定凝固钠的位置;在此基础上,利用FLUENT计算得到的温度和文献中钠剪切力与温度的关系式,分析钠固封塞的受力情况。最终确定冷冻密封腔室加工尺寸为1.5 mm,钠固封塞高度为197 mm。     

EAST速调管快速保护系统的研制

介绍了EAST速调管快速保护系统的研制.EAST速调管快速保护系统的功能是对磁场电流,主回路电流,管体电流和打火保护信号进行监测.当这些信号超过设定的阈值,快速保护系统立即给后面的控制系统发出过阈报警信号,从而快速保护速调管.     

Design of Controller for New EAST Fast Control Power Supply

The effectiveness of the magnetic confinement of plasma can be improved by elongat- ing the plasma cross-section in tokamak devices. But elongated plasma has vertical displacement instability, so a feedback control system is needed to restrain the plasma＇s vertical displacement. A fast control power supply is needed to excite the active feedback coils, which produces a magnetic field to control the plasma＇s displacement. With the development of EAST, the fast control power supply needs to keep on enhancing the fast response and output current. The structure of a new power supply is introduced in this paper. The method of multiple inverters paralleled with the current sharing reactor is presented to meet the need for large current and fast control. According to the design demands of the EAST fast control power supply, the adjuster of the current close loop is applied to the inverter, which can advance its ability to restrain the loop current in low frequency and DC output. The result of the experiment confirms the validity of the proposed scheme and control strategy.     

Development of a Fast Valve for Disruption Mitigation and its Preliminary Application to EAST and HT-7

In large tokamaks, disruption of high current plasma would damage plasma facing component surfaces (PFCs) or other inner components due to high heat load, electromagnetic force load and runaway electrons. It would also influence the subsequent plasma discharge due to production of impurities during disruptions. So the avoidance and mitigation of disruptions is essential for the next generation of tokamaks, such as ITER. Massive gas injection (MGI) is a promising method of disruption mitigation. A new fast valve has been developed successfully on EAST. The valve can be opened in 0.5 ms, and the duration of open state is largely dependent on the gas pressure and capacitor voltage. The throughput of the valve can be adjusted from 0 mbar·L to 700 mbar·L by changing the capacitor voltage and gas pressure. The response time and throughput of the fast valve can meet the requirement of disruption mitigation on EAST. In the last round campaign of EAST and HT-7 in 2010, the fast valve has operated successfully. He and Ar was used for the disruption mitigation on HT-7. By injecting the proper amount of gas, the current quench rate could be slowed down, and the impurities radiation would be greatly improved. In elongated plasmas of EAST discharges, the experimental data is opposite to that which is expected.     

TCB混合堆Li自冷包层MHD流动设计

从磁流体动力学ＭＨＤ压降的物理原理出发，对ＴＣＢ商用混合堆Ｌｉ自冷包层的ＭＨＤ流动方式进行了改进，提出了第一壁环向流动（平行环向磁场流动），核燃料增殖区径向流动的ＭＨＤ流动的设计，以解决混合堆为改善堆的经济性而采取提高包层核燃料富集度的途径所速来的热工，ＭＨＤ压降和安全问题。分析和数值计算结果表明，第一壁环向流动设计可以满足包层核燃料富集度从０．５％增加到１％，相应的热功率从４５００ＭＷ增加到...     

Design Review of Sodium Cooled Fast Reactor Secondary Loop

正Sodium cooled fast reactor secondary loop is also called the intermediate circuit.It is design for insulating the radioactive sodium in the primary loop from water vapor loop,at the same time assuring the thermal energy transfer from the primary loop to the non-radioactive sodium circulation system settings for the water vapor     

He冷却LLi氚增殖包层的热工水力初步设计

完成了托卡马克工程试验混合堆ＴＥＴＢ－Ⅲ Ｈｅ冷液态金属Ｌｉ（ＬＬｉ）氚增殖包层的初步热工水力设计，探讨了包层中载氚的两种可能方式，同时，用程序完成了对第一壁和包层的温度场计算及热工水力设计参数的初步优化。分析结果表明，尽管Ｈｅ气的导热性和密度都比液态金属冷却剂低得多，但仍有可能使堆芯在２．０ＭＰａ的低压下运行，并且包层的热工水力设计参数满足设计要求。     

Analytical Analysis of the Electromagnetic Response for the Vacuum Vessel of EAST

During plasma disruptions and vertical displacement events (VDEs), time-varying eddy currents are induced on Vacuum Vessel (VV) and Plasma Facing Components (PFCs) of Tokamak. In this paper, we calculate eddy currents induced forces on VV and inner limiter (one of the PFC) during plasma disruption in Experimental Advanced Superconducting Tokamak (EAST). Various plasma transients—VDEs and major disruptions (MDs) are considered. And the study includes assessments of integral values (net vertical, hoop forces), time evolution, peaks of the force distribution. It is shown that the distribution pattern of the eddy currents for different scenarios differ greatly, therefore the resulting EMFs and torques cause different mechanical response.     

EAST离子回旋系统液态相移器的理论设计

对EAST离子回旋系统液态相移器的原理进行了理论分析和计算,结果表明:利用液态相移器可产生相位差,从而可用离子回旋波进行电流驱动。另外,液态相移器也可与单支节液态调配器结合使用,构成传输线阻抗匹配系统。     

快堆MOX元件运输容器的缓冲器研究

快堆MOX元件运输容器的缓冲器是决定其放射性包容边界安全的重要部件。某型号MOX元件运输容器的缓冲器材料首次选择泡沫铝,通过自由下落试验的标准姿态进行吸能原理分析,设计出了适用于缓冲器材料的型号、结构及关键参数。对选定的材料进行了拉伸、压缩、剪切3种准静态和动态力学性能试验,获得了用于数值模拟计算的材料本构关系参数,并对模型参数进行了测试,用弯曲试验进行了验证。有限元分析和试验结果对比显示：运输容器缓冲器材料的本构关系具有适用性,可用于快堆某型号MOX元件运输容器的自由下落分析计算。     

钠冷快堆二回路系统热工水力瞬态分析程序SELTAC的开发与应用

针对钠冷快堆二回路系统的具体结构和运行特点,对中间热交换器、直流蒸汽发生器、钠缓冲罐以及泵、管道等设备和部件建立模型,采用FORTRAN语言自主编制了二回路系统热工水力瞬态分析程序SELTAC。利用中国实验快堆的停堆试验数据对所编制程序进行了初步验证。结果表明,程序计算值与试验值趋势一致,最大相对偏差不超过4.34%,吻合程度较好。将验证后的程序与一回路系统程序耦合,分析了某600 MW钠冷快堆在主热传输系统保持排热能力时的紧急停堆工况,得到了二回路系统的瞬态特性,为大型商用快堆电站的设计提供了参考。     

三废处理设施中阀门故障原因分析及对策

简要介绍某三废处理设施的阀门类型和故障特点。检修分析认为,阀门故障多数由于机械连接处的蠕变、介质中存在破坏性杂质造成,故障周期为1² a。文章给出了检修方法、检修周期和检修流程的建议。     

中国实验快堆一回路热工水力稳态计算程序开发

针对中国实验快堆(CEFR)的具体结构和稳态运行特点,利用Fortran语言开发了CEFR一回路热工水力稳态计算程序。重点开发了有关钠的多种物性的子程序、适应不同工况的钠的流动与换热计算子程序,并对关系式进行了对比分析,最后建立了稳态计算模型并开发了程序。在此基础上,对CEFR的一回路系统在满功率下的稳态热工水力特性进行了计算分析,所获得的结果同设计参数吻合,证明了所开发的子程序及稳态程序的正确性。     

Study of Liquid Phase Shifter for ICRF on EAST

A method of current drive with Ion Cyclotron Range of Frequency （ICRF） on Experimental Advanced Superconducting Tokomak （EAST） is described. A variety of liquid silicon oil heights in the phase shifter will bring the phase difference to the current drive. It is found that the current drive can be achieved by using the phase shifter. The liquid phase shifter is one of the impedance matching systems too.     

Primary Results of Lithium Coating for the Improvement of Plasma Performance in EAST

First lithium coating associated with ion cyclotron range of frequency (ICRF) plasma was performed successfully in EAST. Results in reduction of both residual impurity and deuterium in the vacuum vessel were obtained. Particularly the partial pressure of deuterium after the lithium coating was reduced by about a factor of 5. Impurity radiation in the plasma was reduced and electron temperature increased by about 50%. Moreover, reproducible plasma discharges with high parameters, such as higher plasma current and density, could be easily obtained. These results showed that plasma performance was improved. Even though only 2 g of lithium were injected, the effective lifetime of the Li film was raised up to 40 shots.     

直动电磁阀参数设计与分析

控制棒水压驱动技术是清华大学核能与新能源技术研究院具有自主知识产权的一项新型发明专利技术,组合阀属于该项技术的关键部件。组合阀由3个直动电磁阀组成,电磁阀的性能直接影响组合阀的性能,从而影响控制棒水压驱动技术的运行性能。本工作运用二次回归正交试验拟合方法和复合形优化方法,对直动电磁阀的参数进行了设计和分析,并进行了部分实验验证。分析结果表明：输入直动电磁阀的电流数值对电磁力大小影响最大;通过优化方法可以得出直动电磁阀的优化设计参数和最大电磁力。     

Design and Analysis of the Thermal Shield of EAST Tokamak

EAST （Experimental Advanced Superconducting Tokamak） is a tokamak with su- perconducting toroidal and poloidal magnets operated at 4.5 K. In order to reduce the thermal load applied on the surfaces of all cryogenically cooled components and keep the heat load of the cryogenic system at a minimum, a continuous radiation shield system located between the magnet system and warm components is adopted. The main loads to which the thermal shield system is subjected are gravity, seismic, electromagnetic and thermal gradients. This study employed NASTRAN and ANSYS finite element codes to analyze the stress under a spectrum of loading conditions and combinations, providing a theoretical basis for an optimization design of the structure.     

Design,Analysis and R＆D of the EAST In-Vessel Components

In-vessel components are important parts of the EAST superconducting tokamak. They include the plasma facing components, passive plates, cryo-pumps, in-vessel coils, etc. The structural design, analysis and related R＆D have been completed. The divertor is designed in an up-down symmetric configuration to accommodate both double null and single null plasma operation. Passive plates are used for plasma movement control. In-vessel coils are used for the active control of plasma vertical movements. Each cryo-pump can provide an approximately 45 m^3/s pumping rate at a pressure of 10^-1 Pa for particle exhaust. Analysis shows that, when a plasma current of 1 MA disrupts in 3 ms, the EM loads caused by the eddy current and the halo current in a vertical displacement event （VDE） will not generate an unacceptable stress on the divertor structure. The bolted divertor thermal structure with an active cooling system can sustain a load of 2 MW/m^2 up to a 60 s operation if the plasma facing surface temperature is limited to 1500 ℃. Thermal testing and structural optimization testing were conducted to demonstrate the analysis results.     

核电厂一回路系统模拟实验设计与分析

研究核电站特定运行工况下,一回路系统传热流动的规律。应用Ishii模化方法模拟压水堆核电厂的一回路系统,设计出主泵与关联系统耦合实验回路的主要热工参数。同时,应用机理性程序对设计的实验回路进行分析。结果表明,基于Ishii模化方法设计的实验回路主要参数合理可行;模型可以研究反应堆原型事故运行瞬态工况下,一回路各系统间传热流动相互影响规律。