HTR-10燃料球和石墨球反应性当量测量实验

10MW高温气冷实验堆（HTR－10）是球床型反应堆，燃料元件的装卸和循环可不停堆，为了保证其安全运行，必须准确地知道燃料元件的装卸和循环过程可能引起的反应性变化，尤其是向堆芯装料时引入的反应性变化，HTR－10首次临界后对燃料球和石墨球的反应性价值进行了测量，本文介绍了燃料球和石墨球反应性当量刻度的方法，给出了初始装载截芯和空气气氛下的燃料球和石墨球反应性当量实验测量结果。     

HTR-10燃料元件装卸系统的集散控制

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。     

HTR-10燃料元件的气体输送

为了保证高温气冷实验堆球形燃料元件可靠地输送，采用了传递管输送方法，本文介绍了10MW高温气冷堆（HTR－10）燃料元件气体输送系统的关键设备，管路设计及输送气体流量计算，通过初装料的运行，证明该系统运行良好。     

HTR-10的调试管理

10MW高温气冷实验堆（HTR－10）属于研究堆类型，但又具有小型核动力堆的运行模式；HTR－10的调试队伍由设计者、运行者和合同单位有关人员组成。针对HTR－10自身的特点和调试队伍人员组合的特点，确定了调试管理模式。本文重点介绍了HTR－10调试管理中的调试组织的规范化和试验活动的程序化，并给出了试验活动程序流程图。     

10MW高温气冷堆燃料元件装卸系统的控制系统设计

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     

10 MW高温气冷堆的关键设备--提升器

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。     

10MW高温气冷堆球形燃料元件虚拟仪器式探测系统的研制

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。     

10 MW高温气冷实验堆事故分析的结果与对策

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     

HTR-10装料前的调试

10MW高温气冷实验堆（HTR－10）调试工作分为三个阶段进行，设计的调试项目共100个，本文简要介绍了HTR－10的系统划分及调试项目的设计原则，并重点介绍了第一阶段装料前冷态调试试验的内容及结果以及对调试不符合荐的处理办法，通过调试，全面验证了系统设计的合理性以及设备的可靠性和运行相容性，并获得了满意的结果。     

高温气冷堆核材料衡算方法研究

高温气冷堆（HTR）采用球形包覆颗粒燃料元件，采用不停堆换料运行方式。因此，其运行方式、燃料元件的形式、换料方式等与压水堆核电站差别较大。HTR的特点决定了其核材料的监管方式既不同于传统压水堆，也不同于散料核设施，不易采用传统压水堆的件料管理模式和散料核设施的散料管理模式进行核材料衡算管理。为此，本文针对HTR核材料管理，提出一种适于HTR核材料衡算及其不明损失量（MUF）评价的方法。该方法根据HTR的燃料元件、运行方式和换料方式的特点，综合考虑件料和散料衡算两种模式，通过对HTR核材料衡算平衡区合理划分、关键测量点设置和实物盘存方式选取等的研究，最终选取件料+散料的衡算模式进行核材料衡算管理和评估，为HTR核材料监管提供技术基础。目前，该方法已应用于我国HTR的核材料管理，取得了预期的效果。     

10MW高温气冷堆乏燃料元件的贮存

10MW高温气冷堆（HTR-10）在设计寿命内共卸出约9万个乏燃料元件，其放射性裂变产物的活度高达1.0×10¹⁶Bq，必须妥善处置。HTR-10乏燃料元件卸在密封和屏蔽的乏燃料罐内，每罐可容纳2000个乏燃料元件。这些罐暂存在反应堆建筑物最底层的乏燃料暂存库内，在库内采取通风冷却。若干年后，通过转运小车运至反应堆大厅竖井下方，再用大厅吊车从竖井吊至地面，最后用卡车运至最终贮存库。     

Thermal hydraulic calculation of the HTR-10 for the initial and equilibrium core

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.     

Design and manufacture of the fuel element for the 10 MW high temperature gas-cooled reactor

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.     

Manufacture and characteristics of spherical fuel elements for the HTR-10

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.     

Design and full scale test of the fuel handling system

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.     

HTR-10一回路流量变化试验的模拟

10 MW高温气冷实验堆(HTR-10)是我国第一座高温气冷堆。一回路流量变化试验是HTR-10的三个动态特性试验之一,该试验不仅证明了反应堆的功率自调节性能,也为系统分析程序的验证提供了实测数据。基于实际的试验工况,利用THERMIX程序对一回路流量变化试验进行了模拟,分析了反应堆主要参数的变化。关于反应堆功率,计算结果与试验结果符合得很好,证明程序能够满意地再现HTR-10在该试验中的动态特性。试验过程中,燃料元件中心最高温度始终低于1 230℃的温度限值。     

基于γ测量的球床式高温气冷堆内乏燃料球探测方案的模拟分析

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。     

Build-up of plutonium isotopes in HTR fuel elements a comparison between computed prediction and chemical analysis

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.     

Irradiation testing of matrix material for spherical HTR-10 fuel elements

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.