源项分析在高温气冷堆核电站的设计、运行和放射性废物管理等方面具有重要意义。球床模块式高温气冷堆（HTR-PM）采用从德国引进的KORIGEN、PANAMA、FRESCO等源项分析程序，这些程序在模型、计算方法、数据库、适用性等方面存在缺陷，结果具有较大的保守性。为进一步提高高温气冷堆源项分析的精度，本文对高温气冷堆内放射性核素的产生、释放和迁移过程进行了深入研究，研发了高温气冷堆一体化源项分析程序系统，并开展了HTR-PM源项分析应用研究。首先，本论文研究了堆芯源项关键基础算法和精细数据库加工方法。研究并实现了线性子链方法和几种矩阵指数方法，对算法性能进行了比较分析。提出了求解非齐次燃耗方程的通用方法。开展了精细燃耗数据库加工方法研究，制作了能谱相关的中子反应截面数据，并对衰变数据、裂变产额、可利用能和光子发射数据进行了更新和完善。在此基础上，研制了堆芯源项计算程序NUIT。通过与基准题、实验值及其它程序结果对比，验证了该程序的正确性。其次，研究了TRISO燃料性能分析和放射性核素释放分析计算方法。开展了燃料元件和包覆颗粒内细致的温度分布计算，利用Redlich-Kwong气体状态方程计算缓冲层内的气体内压；采用G.Miller级数展开方法求解应力应变方程，得到包覆层内详细的应力分布；采用改进的压力壳失效模型计算包覆颗粒失效率。考虑放射性核素的扩散、反冲、吸附等释放机制，采用有限差分方法求解扩散方程，得到裂变产物浓度分布和释放率等。研发了放射性释放分析程序FRAT。进一步，对放射性核素在一回路内的迁移行为进行了研究。研究了放射性核素的吸附和迁移计算模型。考虑粉尘的团聚、沉积、重悬浮等动力学行为，对粉尘载带放射性核素迁移的行为进行了研究。开展了放射性迁移分析程序FIST的研发与验证。最后，开展了HTR-PM精细化源项分析应用研究。计算得到了平衡堆芯放射性核素总量、包覆颗粒失效率、放射性核素释放率等参数，实现了一回路放射性核素分布的精细模拟。本课题的研究成果对提高高温气冷堆放射性源项分析的自主化和精细化水平具有重要意义。
Radioactive source term analysis is important in the design, operation and radioactive waste management of HTGR. Source term codes introduced from Germany, including KORIGEN, PANAMA, FRESCO, are utilized in the design of HTR-PM. These codes have significant deficiencies in model, algorithm, data library and code function applicability. Thus, the calculation results are conservative to some degree. In order to improve the precision of source term analysis, this thesis studies the generation, release and migration of radionuclides in HTGR. An integration source term analysis code system is developed, and the applied research on HTR-PM source term analysis is conducted.First of all, in this thesis, the critical and fundamental algorithm for the source term analysis, as well as the processing method for elaborate nuclear data library are studied. Both Transmutation Trajectory Analysis (TTA) and several matrix exponential methods are studied and implemented. The algorithm performances are compared. The neutron cross section data library is relevant to the specific neutron spectrum. The decay data, fission yield, recoverable energy and photon emission data are updated and improved. Based on the above, we develop a nuclide inventory code, NUIT. The validation results obtained by the comparison with benchmark, experimental data and other code data show a good accuracy of burnup algorithm and nuclear data library.The performance of the TRISO coated fuel and the radionuclides release behavior are analyzed. First, the temperature profile in the fuel and particle are determined, respectively. The inner gas pressure in the buffer layer is computed by the Redlich-Kwong actual gas state equation. G.Miller series expansion method is adopted to solve the stress-strain equation, thus, the detailed stress distribution is obtained. We calculate the failure probability by utilizing the improved pressure vessel failure model. Then, diffusion, recoil, sorption and other release mechanisms are considered. The diffusion equation is solved by using the finite difference method, through which we get the concentration distribution and release rate. Finally, the radionuclides release analysis code, FRAT, is developed and validated.Furthermore, migration behavior of the radionuclide in the primary loop is studied. We firstly exam sorption and migration model of the radionuclide. Then, coagulation, deposition, resuspension of graphite dust, as well as the transport behavior of the radionuclide attached to the dust particle, are studied. At last, the development and validation of FIST code is performed.Finally, elaborate source term analysis of HTR-PM is performed. Thus, we obtain the nuclide inventory of the equilibrium core, the failure probability of TRISO coated particle and the radionuclide release rate, respectively. An elaborate simulation of radionuclide migration in the primary loop is performed as well. This work helps to improve the localization rate and the refinement of the source term analysis of HTGR.
