摘要
钠冷快堆(Sodium-cooled Fast Reactor,SFR)作为第四代核能系统的代表性堆型,凭借其燃料增殖潜力、次锕系核素嬗变能力及固有安全特性成为国际研究热点。然而,钠冷快堆设计与运行过程中普遍存在核数据随机性、模型简化误差、边界条件扰动等多源不确定性,上述不确定性与反应堆内复杂的中子学行为、钠热工水力特性及事故工况下的多物理场效应耦合传播,可能导致钠冷快堆系统响应偏离或异常波动,甚至引发事故风险。本文主要从核反应堆物理计算、热工水力分析和事故安全分析三个方向概述了钠冷快堆系统不确定性分析的研究进展,总结了国际范围内的研究成果,分析了遇到的技术挑战和未来研究的发展趋势,并探讨了我国在这些领域的研究现状及未来的研究方向。通过分析与总结,希望为未来钠冷快堆系统的安全性评估与设计优化提供一定参考。
As a representative reactor type of Generation IV nuclear energy systems,the sodium-cooled fast reactor(SFR)has become an international research hotspot due to its fuel breeding potential,transmutation capability of minor actinides,and inherent safety characteristics.However,its design and operation are universally subject to multi-source uncertainties,including nuclear data randomness,model simplification errors,and boundary condition perturbations.These uncertainties,when coupled with complex neutronics behavior within the reactor,sodium thermal-hydraulic characteristics,and multiphysics effects under accident conditions,may lead to deviations or abnormal fluctuations in the system response of SFR,and even trigger accident risks.This paper reviews research progress in SFR system uncertainty analysis from three aspects of reactor physics calculations,thermal-hydraulic analysis,and accident safety assessment.Worldwide research achievements,technical challenges and future development trends are summarized,and current research status and future directions of uncertainty analysis for SFR in China are analytically discussed.Through systematic analysis and synthesis,safety evaluation and design optimization of future SFR systems are perspectively provided.
作者
张哲
陈毅
金德升
吴天淏
赵汉宏
姜潮
ZHANG Zhe;CHEN Yi;JIN Desheng;WU Tianhao;ZHAO Hanhong;JIANG Chao(College of Mechanical and Vehicle Engineering,Hunan University,Changsha 410082,China;China Nuclear Power Technology Research Institute Co.,Ltd.,Shenzhen 518000,China)
出处
《核技术》
北大核心
2025年第7期87-97,共11页
Nuclear Techniques
基金
国家自然科学基金(No.52375242)
湖南省科技创新计划(No.2024RC3112)
湖南省自然科学基金优秀青年基金(No.2023JJ20011)资助。
关键词
钠冷快堆系统
不确定性分析
核反应堆物理计算
热工水力分析
事故安全分析
Sodium-cooled fast reactor system
Uncertainty analysis
Reactor physics calculations
Thermalhydraulic analysis
Accident safety assessment
