超高温气冷堆是国际公认的第四代先进核能系统六种堆型之一。其原型堆——高温气冷堆，具有固有安全特性和高的热转换效率。堆芯出口温度的氦气可用于提供工艺热、制氢等，有广阔的发展应用空间。辐射安全受到广泛关注。核电站中的放射性物质从根本上来说，均产生于堆芯。对反应堆一回路冷却剂放射性水平进行监测，进而掌握堆芯燃料元件状况，才能准确的开展放射性源项分析和科学的评价反应堆辐射安全。10MW高温气冷堆（HTR-10）是中国的第一座气冷堆，也是目前世界上唯一在运行的球床式高温堆。以此为基础，由清华大学核研院设计的高温气冷堆示范电站(HTR-PM)已在中国山东石岛湾建造。HTR-10和HTR-PM均采用氦气作为冷却剂，石墨作为反射层，含有TRISO包覆颗粒的石墨球作为燃料元件。本论文针对高温堆堆芯燃料元件破损与一回路放射性水平进行研究，提出一回路冷却剂放射性水平的监测方案。以HTR-10上一回路冷却剂放射性测量结果为基础，对实验谱图进行了标定、校正和分析，研究惰性气体裂变产物的产生来源和释放行为，确定能够有效指示堆芯燃料元件状态的典型裂变核素。结果显示Kr、Xe的活度浓度在反应堆3MW运行时展现了良好的稳定性和一致性，并讨论影响一回路冷却剂放射性水平的关键因素（包覆颗粒破损率、自由铀污染份额等）。由于HTR-PM还在安装和调试过程中，根据HTR-10的研究结论，以HTR-PM理论计算源项为基础，提出HTR-PM一回路冷却剂放射性水平监测方案，采用在线式总?连续监测和实验室取样核素分析相结合的方法。前者采用NaI(Tl)探测器，连续监测一回路冷却剂总的放射性水平，在反应堆出现异常工况时及时警报；后者采用HPGe探测器，准确测量一回路冷却剂各核素的活度浓度，用于确定堆芯燃料元件的破损状态。本论文采用蒙特卡罗方法模拟计算在线式总?监测仪的探测效率，开展灵敏度分析，研究影响探测效率的关键因素及典型核素活度浓度变化对测量结果的影响。结果表明，位于氦净化系统入口前的在线式总?监测仪，单位时间的计数比较高，可及时报警并有效反映一回路冷却剂放射性水平的变化；而位于氦净化系统出口的在线式总?监测仪，结合氦净化系统入口前的监测结果，可有效指示低温吸附器的吸附效率，并对氦净化系统的功能恶化发出早期警报。本论文的研究结论和设计方案已应用于国家科技重大专项HTR-PM，并为高温气冷堆燃料元件破损研究和源项分析工作提供重要参考。

The very high temperature gas-cooled reactor (VHTR) has been identified as one of the six Generation IV concept reactors internationally. As its prototype, the high temperature gas-cooled reactor (HTGR), possesses inherent safety characteristics and high heat transfer efficiency. The helium at outlet temperature from the core can be used for process heat supply, hydrogen production, etc., which brings a broad range of HTGR’s application in future. The radiation safety has received wide attention. Fundamentally, all the radioactive substances in a nuclear power plant come from the reactor core. To perform the radioactive source term analysis precisely and evaluate the reactor radiation safety scientifically, it is essential to monitor the radioactivity in the primary coolant, thus control the status of the fuel elements in the reactor core.The 10 MW high temperature gas-cooled reactor (HTR-10) is the first gas-cooled test reactor in China, and also the only pebble bed reactor currently operating in the world. On the basis of HTR-10, the Institute of Nuclear and New Energy Technology (INET) at Tsinghua University has designed the high temperature gas-cooled pebble-bed modular reactor (HTR-PM), which is under construction in Shidao Bay, Shandong Province, China. Both HTR-10 and HTR-PM adopt helium as primary coolant, graphite as reflectors, and graphite spheres containing tristructural-isotropic (TRISO) coated particles as fuel elements. In this thesis, the fuel particle failure is studied, the radioactive level in the primary coolant in HTR-10 and HTR-PM is estimated, and the scheme for the radioactivity monitoring of primary coolant is proposed. Based on the experimental measurement of radioactive level of primary coolant in HTR-10, the calibration, correction, and analysis of the experimental spectra are implemented. The generation sources and release behavior of noble gas fission products are studied. The variation of the activity concentration of typical nuclides along with the operational time of the reactor is revealed. The types of nuclides which can be indicators for the status of fuel elements in the core effectively are determined. From current research, the low activity concentrations of Kr and Xe exhibited well performance of stability and consistency when the HTR-10 was operated at 3 MW thermal power. In addition, the key factors which can influence the radioactivity level of the primary coolant are discussed, including the defect fraction of coated particles in fuel elements and fraction of free uranium contamination. Since the HTR-PM is still in equipment installation and commissioning phase, on the basis of the research in HTR-10 and the theoretical calculations on the source terms in the primary circuit, a radiation monitoring scheme on the radioactivity level of primary coolant in HTR-PM is proposed, which adopts the combination of on-line gross γ continuous monitoring and primary helium sampling for nuclide analysis in the radiochemistry. The former uses the NaI(Tl) detector to continuously monitor the total radioactivity level of primary coolant in real time, which can initiate the alarm in time when abnormal operational affairs occur; while the latter employs the HPGe detector to measure the activity concentrations of each nuclides in primary coolant accurately, which can determine the breakage status of the fuel elements in the core. The Monte Carlo method is adopted to simulate the detection efficiencies of two kinds of on-line gross γ monitors in HTR-PM, and carry out sensitivity analysis to illustrate the key factors which can affect the detection efficiency and the impacts of the activity concentrations variation of typical nuclides on the measurement results. Research result indicates that the on-line gross γ monitor upstream of the entrance of the helium purification system (HPS), has relatively high count rates, which can give an alarm timely and reflect the variation of radioactivity of primary coolant effectively. However, the on-line gross γ monitor downstream of the exit of the HPS can indicate the adsorption efficiency of the low temperature adsorber efficaciously in combination with the measurement results of the entrance of the HPS, and send an alarm when the purification function of the HPS degenerates in early stage. The research results and design scheme obtained in this thesis have been applied to HTR-PM which belongs to the National S & T Major Projects, and can provide important reference for the study on the failure of fuel elements and source term analysis in HTGRs.