反应堆安全壳热电偶贯穿件

反应堆安全壳热电偶贯穿件，属于核工程过程变量检测技术领域。为了提供一种结构简单、制造工艺成熟、成本低、性能可靠、气密封和承压性能好的热电偶贯穿件，本发明公开了一种反应堆安全壳热电偶贯穿件。它包括预埋管、泄漏检测装置、连接法兰、贯穿件本体及通过热电偶电缆连接的热电偶贯穿组件。热电偶贯穿组件包括通过密封连接材料和机械承插口连接成一体的热电偶接插件、铠装热电偶和密封管，     

专利简讯

反应堆安全壳热电偶贯穿件 [公开日]2004.04.21 [分类号]G21C13/036 [公开号]1490822 [申请号]03157054.2 [申请日]2003.09.12 [申请人]清华大学 [文献]反应堆安全壳热电偶贯穿件,属于核工程过程变量检测技术领域。为了提供一种结构简单、制造工艺成熟、成本低、性能可靠、气密封和承压性能好的热电偶贯穿件,本发明公开了一种反应堆安全壳热电偶贯穿件。它包括预埋管、泄漏检测装置、连接法兰、贯穿件本体及通过热电偶电缆连接的热电偶贯穿组件。热电偶贯穿组件     

核电站反应堆热电偶机械密封件研究

压水反应堆堆内温度测量装置是监测反应堆热功率输出及堆内功率分布情况、保证反应堆安全运行的手段之一。该装置以热电偶作为测量元件,用不锈钢作保护外套。由于热电偶有一定的使用寿命,须定期更换,为此研制了一种适用于核电厂反应堆堆内温度测量的热电偶密封结构。这种结构既能在高温高压下密封牢靠,没有泄漏和渗漏,又能在带放射性环境下方便地逐根更换热电偶。它是一种结构简单实用的新型密封结构。     

核电站反应堆热电偶机械密封件研究

压水反应堆堆内温度测量装置是监测反应堆热功率输出及堆内功率分布情况、保证反应堆安全运行的手段之一。该装置以热电偶作为测量元件,用不锈钢作保护外套。由于热电偶有一定的使用寿命,须定期更换,为此研制了一种适用于核电厂反应堆堆内温度测量的热电偶密封结构。这种结构既能在高温高压下密封牢靠,没有泄漏和渗漏,又能在带放射性环境下方便地逐根更换热电偶。它是一种结构简单实用的新型密封结构。     

从核废物中回收贵金属

【美国《核学会汇刊))1982年第43卷第90页报道】三个贵金属——钯、铑、钉以及锝都可以从反应堆的裂变产物中回收。表l列出1973—2000年美国累积需要量和从商用反应堆的裂变产物废物中可以回收的量。1吨从商用反应堆卸出的辐照燃料含有的这些贵金属约值20,000美元。从裂变产物中回收这些贵金属可以满足美国对铂     

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。     

10MW高温气冷实验堆反应堆压力容器热电偶贯穿件

介绍了 10 MW高温气冷实验堆反应堆压力容器热电偶贯穿件。热电偶贯穿件用于堆芯部件温度的测量,由贯穿筒体、铠装热电偶组件和焊接保护管组成。采用过渡管结构、激光焊和钨极氩弧焊方法,实现两种壁厚相差甚大的铠装热电偶套管和贯穿筒体的焊接,有效地解决了高温高压下氦气密封的困难。铠装热电偶组件直接贯穿反应堆容器,在容器外采用卡套密封,避免了高温容器内信号的转接。经 ANSYS程序分析计算,热电偶贯穿件的结构设计满足应力强度和抗震要求。经氦检漏试验,热电偶贯穿件泄漏率小于 1× 10－ 7 Pa· m3/s。该热电偶贯穿件现已在反应堆上安装完毕。     

反应堆堆芯液位监测装置

本发明涉及一种液位监测装置，具体涉及一种反应堆堆芯液位监测装置。该装置包括液位探测器和二次仪表，液位探测器包括电加热器、焊接在电加热器上的热电偶和密封件，液位探测器通过与其相连的插接件、电加热器电缆和信号电缆与二次仪表相连，液位探测器的电加热器外壁上开有不同高度的纵向凹槽，热电偶嵌焊在电加热器的凹槽内，电加热器外部套有薄壁套管，电加热器、热电偶和薄壁套管紧密配合，液位探测器的外部为焊接在密封体上的多孔保护套管。本装置结构简单、性能稳定可靠、信号响应时间短且寿命长。     

热电偶在反应堆工程应用中几个值得注意的问题

本文介绍了反应堆上使用热电偶测量堆芯温度中发现的问题——流量交混不充分、安装不合理、γ发热等引起的热电偶指示值不准以及两种难于判断的热电偶故障。对使用热电偶测量堆芯温度提出了几点建议。     

裂变热电偶的研制

介绍了裂变热电偶研制中一些问题的考虑，以及裂变热电偶的初步性能测试。裂变热电偶中裂变珠的直径为１ｍｍ左右，探测热中子的灵敏度约为３×１０＾９ｎ／ｃｍ＾２·℃，时间响应好于１μｓ。裂变热电偶最明显的应用是描绘堆内和堆周围的中子通量，可作为反应堆诊断学和反应堆控制方面的新工具。     

In-core coolant flow monitoring of pressurized water reactors using temperature and neutron noise

Noise measurements were performed at the Loss-of-Fluid-Test (LOFT) and Sequoyah-1 pressurized water reactors (PWRs) in order to investigate the possibility of inferring in-core coolant velocities from cross-power spectral density (CPSD) phases of core-exit thermocouple and in-core neutron detector signals. These noise measurements were used to investigate the effects of inlet coolant temperature, core flow, reactor power, and random heat transfer fluctuations on the noise-inferred coolant velocities. The effect on the inferred velocities of varying in-core neutron detector and core-exit thermocouple locations was also investigated. Theoretical models of temperature noise were developed, and the results were used to interpret the experimental measurements.Results of these studies indicate that the neutron detector/thermocouple phase is useful for monitoring core flow in PWRs. Our results show that the interpretation of the phase between these signals depends on the source of temperature noise, the response times and locations of the sensors, and the neutron dynamics of the reactor. At Sequoyah-1 we found that the in-core neutron detector/core-exit thermocouple phase can be used to infer in-core coolant velocities, provided that the measurements are corrected for the thermocouple response time.     

Temperature Measuring System of the In-core Components for Chinese 10 MW High Temperature Gas-cooled Reactor

The 10 MW High Temperature Gas-cooled Reactor (HTR-10) constructed at the Institute of Nuclear Energy Technology (INET), Tsinghua University in China reached its first criticality by the end of 2000. The temperature measuring system of the in-core components is described in this paper. This system consists of the thermocouple penetration assembly of the reactor pressure vessel (TPARPV), the thermocouple penetration assembly of the reactor containment (TPARC) and the distributed computer-based data acquisition and processing system (DCS). Some new techniques were developed and applied, such as the thermocouple penetration technology under the high temperature and high-pressure environment and the laser welding technique. The TPARPV is the key measurement device and is described in detail. The general behavior of the TPARPV and TPARC was confirmed under HTR-10 operating conditions. The helium leakage rate of the TPARPV is 1×10^-7Pa-m³/s while the helium leakage rate of the TPARC is less than 1×10^-2 Pa-m³/s. The insulation resistance of the sheathed thermocouple is more then 10⁹Ω. The temperature measurement error of the system is 2.3°C. The results of testing and field inspection and operation demonstrate that the design of the temperature measuring system is reasonable and reliable and that the performance of the system satisfies the design requirements of the HTR-10. These new techniques used in the temperature measuring system can be applied not only to other high temperature gas-cooled reactors but to various reactor types as well.     

裂变热电偶时间响应特性研究

裂变热电偶是一种能在强γ辐射环境下测量中子的快响应小型化探测器。本文从理论上估算了裂变热电偶的时间响应,好于1μs,并通过热阶跃响应实验验证了理论计算,     

Determination of thermocouple transfer-functions and fluid-flow velocities by temperature-noise measurements in liquid sodium

The time response of TI-, TM- and intrinsic thermocouples has been investigated in sodium by determination of frequency dependent thermocouple transfer functions and related delay times. Experimentally temperature fluctuations (temperature noise), generated in sodium by heated channels or injection of hot sodium, have been used as driving sources. The measured transfer functions and delay times have been compared to predictions from theoretical models. Good agreement was found. Intrinsic and TM-type thermocouples are best suited if fast response is required. Delay times in the range of one millisecond or upper break frequencies of 1000 Hz were determined for intrinsic thermocouples.

The determination of sodium flow velocities by the transit time correlation technique showed systematic error sources for analysis at low frequencies due to intermediate heat storage between the thermocouple positions. Statistical errors have been determined as function of thermocouple distance, frequency range and time of analysis. Minimum errors are obtained for thermocouple distances from 100 to 150 mm and in the frequency range 90 to 120 Hz. A transit time correlator, which automatically tracks the maximum of the cross-correlation function and displays the velocity, has been successfully used for signals with correlation coefficients greater than 0.3.     

高稳定性高可靠性耐辐照新型热电偶

新研制的镍铬硅/镍硅镁热电偶合金及其铠装热电偶具有高稳定性、高可靠性及良好的耐辐照性能。它的性能远优于常规的镍铬/镍铝(K 型)热电偶,某些性能也优于 Nicrosil/Nisil(N 型)热电偶。这种新型热电偶合金及其铠装热电偶可在要求高精确度的科研部门及要求高可靠性的核的或非核的工业部门中应用。     

自动化量热测氚装置研制

以量热测氚法为基础,设计了一套自动化测氚装置。该装置通过综合应用双杯对接测量、大规模热电偶组均布、标准加热电阻预热等技术,使预热时间由原来的15h减至5h以内,在300~3000mW时,测量不确定度提高至0.4%,实现了含氚部件氚含量的快速、高精度测量。     

堆芯热电偶密封结构失效机理分析

核电厂堆芯热电偶密封结构属于主系统的压力边界,其可靠性对于核电厂主系统的严密性和安全性具有重要意义.文章以秦山核电站的堆芯热电偶密封结构为例,对目前国内压水堆核电厂堆芯热电偶密封结构中应用广泛的CONOSEAL和GRAYLOC的组合密封结构进行分析,对各种密封失效事件进行分析,并提出相应的改进措施.     

Calculation of Beam Extraction Pulse Duration for Beam Angular Divergence Measurement of EAST Neutral Beam Injection

Distributed thermocouple measurement will be used on EAST neutral beam injection beam angular divergence measurement. In order to determine the pulse duration of neutral beam extraction, the plates of calorimeter are analyzed by using heat transfer science. Mathematic derivations and the results are given according to the installation location of thermocouple in the plate. Those results could point the way to beam angular divergence measurement of EAST neutral beam injection.