聚变堆包层模块第一壁不同冷却剂传热性能研究
Research on Heat Transfer Performance of Different Coolants for the First Wall of the Test Blanket Module of Nuclear Fusion Reactor
张豪磊 1周涛 1薛春辉 1刘鹏1
作者信息
- 1. 东南大学 能源与环境学院核科学与技术系,江苏南京 211189;核热工安全与标准化研究所,江苏南京 211189;大型发电装备安全运行与智能测控国家工程研究中心,江苏南京 211189
- 折叠
摘要
[目的]核聚变作为一种清洁、高效的能源,是实现全球可持续发展的未来希望.针对中国氦冷固态增殖剂包层模块的第一壁,基于核热工安全与标准化研究团队提出的 4根冷却剂道的设计方案,计算了氦气、氩气、氮气作为冷却剂的温度场.[方法]选择B.S.Petukhov公式,计算聚变堆包层模块第一壁不同冷却剂的传热性能.[结果]研究表明:氦气、氩气、氮气作为冷却剂,致使Be板和RAFM钢中温度场的变化趋势是相似的;温度场出现的最大温度均小于许用温度,符合温度的安全要求;氮气作为冷却剂可以实现的安全裕度是最大的,氩气次之,氦气实现的安全裕度相较偏小.[结论]可由此对聚变堆实验包层第一壁的冷却剂选择提供更多的优化可能,对聚变堆实验包层第一壁的安全增加更多的裕度.
Abstract
[Introduction]As a clean and efficient energy source,nuclear fusion promises to achieve global sustainable development in the future.For the first wall of Chinese helium-cooled solid breeder test blanket module,the temperature fields of helium,argon and nitrogen as coolants were calculated based on the design scheme of four coolant channels proposed by the Nuclear Thermal Security and Standardization Research Team.[Method]To calculate the heat transfer of different coolants in the first wall of the cladding module of fusion reactor,the B.S.Putukhov formula is used to calculate the heat transfer.[Result]Research indicates that the temperature field variations in the Be panel and RAFM steel are similar for helium,argon,and nitrogen as coolants.The maximum temperatures in the temperature fields are below the allowable temperature,meeting the safety requirements for temperatures.Nitrogen as a coolant offers the largest safety margin,followed by argon,with helium presenting a relatively smaller safety margin.[Conclusion]This study provides more optimization options for selecting the coolant for the first wall of the test blanket module of nuclear fusion reactor,thereby enhancing the safety margin of the first wall of the test blanket module of nuclear fusion reactor.
关键词
核聚变/第一壁/氦气/氩气/氮气/冷却剂Key words
nuclear fusion/the first wall/helium/argon/nitrogen/coolant引用本文复制引用
基金项目
江苏省双创人才经费项目(110300297)
科技部重点研发计划(2020YFB1901700)
中核科技工业核动力技术创新中心项目(6903003092)
出版年
2024
国家科技期刊平台
NSTL
万方数据