中国氦冷固态增殖剂实验包层系统氚输运初步分析
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摘要
氚输运分析是开展中国氦冷固态增殖剂实验包层系统安全分析及未来聚变堆氚自持运行的重要研究内容之一。基于氚输运理论和固态增殖剂包层系统设计,利用FDS凤麟核能团队开发的聚变系统氚分析程序TAS,构建了固态增殖剂包层系统氚输运分析系统动力学模型。该模型氚输运结果与文献报道的吻合得很好,误差小于6%,验证了模型的正确性。针对中国氦冷固态增殖剂实验包层系统氚输运问题进行了两种计算方法(稳态、脉冲模式)的初步分析,获得了氚提取系统、氦气冷却系统回路氚分压,实验包层模块冷却流道、窗口室内氚提取系统和氦气冷却系统回路材料中氚滞留量,窗口室内氚提取系统和氦气冷却系统回路氚日渗透量等数据。最终对比结果显示,脉冲模式分析方法能够实时地跟踪源项的快速变化,更符合中国氦冷固态增殖剂实验包层系统实际运行情况。窗口室内氦气冷却系统回路材料中氚滞留量占到日产氚量的31.3%,因此需要在这些氚滞留损失严重的部位考虑适当的阻氚措施。
Tritium transport analysis is an important research issue not only in the safety analysis of China Helium Cooled Ceramic Breeder Test Blanket System,but also tritium-selfavailability operation of future fusion reactors. Based on the tritium transport theory and design of ceramic breeder blanket systems,a tritium transport analysis system dynamics model of the solid breeder blanket system was built using the Tritium Analysis Program for Fusion System(TAS)developed by FDS Team. The tritium transport outcome of this model is in good agreement with that of literature and the error is less than 6%,verifying the correctness of the model. The preliminary tritium transport analysis of China Helium Cooled Ceramic Breeder Test Blanket System was performed by two calculation methods of the steady state and the pulse mode. Several data,tritium partial pressures in Tritium Extraction System and Helium Cooling System loops,tritium retentions in the cooling channel material of Test Blanket Module,Tritium Extraction System and Helium Cooling System loop material in Port Cell as well as tritium permeations from Tritium Extraction System and Helium Cooling System loop into Port Cell per operational day,were obtained. The final comparative results show that the pulse mode analysis method can track the rapid changes of source terms,more in line with the actual operating situations of China Helium Cooled Ceramic Breeder Test Blanket System. The tritium retention in Helium Cooling System loop material in Port Cell is accounted for 31.3% of tritium generation per operational day,thus suitable tritium permeation preventing measures should be taken in these areas where tritium retention losses are severe.
Tritium transport analysis is an important research issue not only in the safety analysis of China Helium Cooled Ceramic Breeder Test Blanket System,but also tritium-selfavailability operation of future fusion reactors. Based on the tritium transport theory and design of ceramic breeder blanket systems,a tritium transport analysis system dynamics model of the solid breeder blanket system was built using the Tritium Analysis Program for Fusion System(TAS)developed by FDS Team. The tritium transport outcome of this model is in good agreement with that of literature and the error is less than 6%,verifying the correctness of the model. The preliminary tritium transport analysis of China Helium Cooled Ceramic Breeder Test Blanket System was performed by two calculation methods of the steady state and the pulse mode. Several data,tritium partial pressures in Tritium Extraction System and Helium Cooling System loops,tritium retentions in the cooling channel material of Test Blanket Module,Tritium Extraction System and Helium Cooling System loop material in Port Cell as well as tritium permeations from Tritium Extraction System and Helium Cooling System loop into Port Cell per operational day,were obtained. The final comparative results show that the pulse mode analysis method can track the rapid changes of source terms,more in line with the actual operating situations of China Helium Cooled Ceramic Breeder Test Blanket System. The tritium retention in Helium Cooling System loop material in Port Cell is accounted for 31.3% of tritium generation per operational day,thus suitable tritium permeation preventing measures should be taken in these areas where tritium retention losses are severe.
引文
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[4]CHEN C A,ZHOU X,WANG Z,et al.Assessment of Tritium Release Through Permeation and Natural Leakage in ITER CN HCCB TBS Under Normal Operations[J].Fusion Science and Technology,2018,73(1):34-42.
[5]ZHANG L,CN HCCB TBM TEAM.HCCB-TBS Tritium Transport Analysis and Benchmark Exercises,ITER_D_TQ8HAM[R].France:ITER.
[6]吴宜灿,胡丽琴,龙鹏程,等.先进核能系统设计分析软件与数据库研发进展[J].核科学与工程,2010,30(1):55-64.
[7]HUANG Q.Status and improvement of CLAM for nuclear application[J].Nuclear Fusion,2017,57(8):1-9.
[8]WANG B,LIU L,XIANG X,et al.Diffusive transport parameters of deuterium through China reduced activation ferritic-martensitic steels[J].Journal of Nuclear Materials,2016,470:30-33.
[9]FORCEY K S,ROSS D K,SIMPSON J C B,et al.Hydrogen transport and solubility in 316L and1.4914 steels for fusion reactor applications[J].Journal of Nuclear Materials,1988,160(2-3):117-124.