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HTR-10燃料元件装卸系统的集散控制 总被引:1,自引:1,他引:0
依据10MW高温气冷实验堆(HTR-10)燃料元件装卸系统的控制要求,对燃料元件装卸系统的集散控制进行了研究,设计了集散控制系统,包括硬件设计,软件设计,步进电动机精确定位闭环控制和无位置传感器转子位置检测的设计,并进行了模拟实验,实验表明,系统的性能稳定可靠,达到了控制要求。 相似文献
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10MW高温气冷堆燃料元件装卸系统的控制系统设计 总被引:3,自引:0,他引:3
10MW高温气冷实验堆(HTR-10)是一座球床型反应堆,燃料元件的装卸和循环不需要停堆,由燃料元件装卸系统自动实现。为保证HTR-10的正常运行,燃料元件装卸系统必须安全可靠运行。为此,控制系统根据HTR-10燃料装卸系统热实验装置控制系统的设计和运行经验,采用了欧姆龙(OMRON)C200H可编程控制器(PLC)作为核心部件。本文介绍了控制系统的设计方案、控制过程和PLC控制的特点以及用其实现 相似文献
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研究了10MW高温气冷堆(HTR-10)中提升器的结构,工作原理及驱动系统的特点,在提升器的结构设计中,根据提升的性能要求,选用五相混合式步进电动机作为提升器的驱动电机,并采用了闭环控制线路,提高了提升器的定位精度和运动平稳性。通过提升器及其控制系统,已向堆芯发送了近2万个燃料元件和石墨球,使HTR-10达到临界。 相似文献
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介绍了一种基于虚拟仪器技术研铜开发的高温堆燃料元件探测系统,该系统有效地解决了10MW高温气冷堆燃料元件装卸过程中现场复杂过球信号的计数判断问题,很大程度地消除了外界干扰,保证了计数的准确性,方便地实现了整个系统的智能管理。 相似文献
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1 0MW高温气冷实验堆 (HTR 1 0 )的事故分析表明 ,在设计基准事故和严重事故条件下 ,HTR 1 0的堆芯燃料元件的最高温度和反应堆冷却剂系统的压力都低于规定的安全限值 ,燃料元件和冷却剂系统压力边界都能保持其完整性 ,不会造成裂变产物大量向外释放。根据事故分析结果并参照国外高温气冷堆安全运行的管理实践经验 ,针对HTR 1 0所提出的一系列事故对策有效地保证了HTR 1 0在较高的安全水平上进行设计、建造、运行及管理等 ,能够确保HTR 1 0、人员、社会以及环境的安全 相似文献
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《核电子学与探测技术》2018,(5)
针对高温气冷堆(HTR)燃料元件和换料方式不同于传统压水堆的特性,HTR采用燃料元件计数与核素含量计算相结合的衡算方式,详细介绍了核素含量的计算方法,并对该算法进行系统测试和MUF不平衡差验证,证明了该衡算算法的正确性和可行性,实现了对HTR核材料平衡的监督与控制。 相似文献
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The thermal hydraulic calculations of the 10 MW high temperature gas-cooled-test module (HTR-10) are among the most important indications to judge the reactor performance under design conditions. The power distribution, the temperature distribution and the flow distribution of the HTR-10 are calculated for initial and equilibrium core in this paper. The temperature distribution includes the temperature parameters of fuel elements, the helium coolant and the main components in the reactor. In the temperature calculation of fuel elements, several uncertain factors are considered carefully, including non-uniform burnup, power distribution deviation, manufacture deviation of fuel elements, graphite balls mixed with fuel balls in the core, calculation deviation of heat transfer and so on. In the flow distribution calculation, the conservative pebble bed core flow value is selected. The results show that the maximum fuel temperature is much lower than the limitation and the flow distribution can meet the cooling requirement in the reactor core. 相似文献
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Design and manufacture of the fuel element for the 10 MW high temperature gas-cooled reactor 总被引:1,自引:0,他引:1
Chunhe Tang Yaping Tang Junguo Zhu Yanwen Zou Jihong Li Xiaojun Ni 《Nuclear Engineering and Design》2002,218(1-3)
The Chinese 10 MW high temperature gas-cooled reactor (HTR-10) attained its first criticality on December 21, 2000. The fabrication of the first fuel for the HTR-10 started in February 2000 at the Institute of Nuclear Energy Technology (INET), Tsinghua University. Up to September 2000, a total of 11 721 spherical fuel elements were successfully produced. The average free uranium fraction of the first fuel-determined by the burn-leach method-was 5.0×10−5. So far, the release rate R/B of the fission gas, measured in the irradiation test, shows that not a single particle in three irradiated spherical fuel elements failed as the results of the irradiation test carried out in Russia. This paper describes the design parameter, the fabrication technology and the performance data of the HTR-10 first fuel, and the production and quality control experiences obtained from the manufacture of the first fuel for the HTR-10. 相似文献
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Hongsheng Zhao Tongxiang Liang Jie Zhang Jun He Yanwen Zou Chunhe Tang 《Nuclear Engineering and Design》2006,236(5-6):643-2004
The R&D of spherical fuel elements for the 10 MW high temperature gas-cooled reactor (HTR-10) started in 1986 in China. A process known as cold quasi-isostatic molding was used for manufacturing spherical fuel elements, and about 20,540 spherical fuel elements were produced in 2000 and 2001. Fabrication technology and graphite matrix materials were investigated and optimized. Cold properties of the spherical fuel elements met the design specifications. The mean free uranium fraction of 44 batches was 4.57 × 10−5. In-pile irradiation test results showed that irradiation did not lead to apparent change in linear dimensional, geometrical density, porosity and strength of matrix graphite samples. No cracks and blisters were observed in spherical fuel elements. This indicated that matrix graphite and spherical fuel elements of HTR-10 met the requirement of design specifications. 相似文献
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In the 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) fuel elements move through the core driven by gravity. To reach their design burn-up the fuel elements are re-shuttled five times. This transportation outside the core is mainly achieved pneumatically. Although, adopting the international experience at design and operation of similar systems some key components were improved so that the fuel handling system (FHS) becomes simpler and more reliable. The improved components were tested in full-scale testing facilities. The debugging test and the first loading operation for the FHS indicate that the FHS meets the demands of the HTR-10. In this paper, the functions, design parameters, technological processes, main components and design characteristics of the FHS are described in detail. The flow schemes, design parameters of the full-scale testing facilities and the experimental results are briefly introduced. 相似文献
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H. Werner 《Nuclear Engineering and Design》1997,170(1-3):147-164
Irradiation experiments with HTR fuel elements in the AVR reactor are described. It will be shown how fuel elements are inserted into the fuelling process so selectively that, after a single core passage, it is possible to uniquely identify them by means of the burn-up measuring system so as to enable a selective discharge. The method of chemical digestion of the fuel elements is described and the results of mass spectrometric analyses are presented. Computer analyses with the HTR-2000 reactor simulation program are described. The experimental results will be compared with the results from computer simulation. 相似文献
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在球床式高温气冷堆中,对排出堆芯的乏燃料球的探测和数量统计是燃料监测的重要内容。按国际原子能机构针对球床式高温气冷堆核安保的要求,对于燃料装卸系统管道内的燃料监测应开发一种独立于现有涡流检测原理的新监测方案。本文提出了一种基于γ测量原理的新探测方案,设计了探测器构型,对其在堆稳态运行时的探测功能进行验证。结合球床式高温气冷堆HTR-10的燃料球放射性核素数据,及对不同球速不同燃耗的燃料球经过探测区域过程的蒙特卡罗模拟分析,验证了此方案对单个燃料球鉴别和计数的可靠性,同时证明了该方案对于燃料球球流探测的可行性,为今后该探测方案的完善和实际装置的制作提供了设计基础。 相似文献
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Chunhe Tang Ziqiang Li Yanwen Zou Xiaoming Fu 《Nuclear Engineering and Design》2008,238(11):2886-2892
In order to evaluate if the fuel elements and their components (matrix material and coated fuel particles) can meet the design requirements of 10 MW high temperature gas-cooled reactor (HTR-10), an irradiation testing with four spherical fuel elements, 60 matrix material specimens of 5 mm × 5 mm × 40 mm and 13,500 coated fuel particles was performed in Russian IVV-2M reactor from July 2000 to February 2003. The irradiation temperature was 1000 °C. The fast neutron fluence of matrix material specimens reached 1.3 × 1021 cm−2 (E > 0.1 MeV). Post-irradiation examination contained the visual inspection, dimension measurement and determining the density, porosity, specific electrical resistance and bending strength. The irradiation results are given in this paper, and show that the matrix material for spherical HTR-10 fuel elements made from the domestic raw materials and fabricated by the quasi-isostatic room-temperature moulding process is suitable as a structural material for spherical HTR fuel elements. 相似文献