田湾核电站换料期间中子通量密度监测方式优化研究  

Optimization Research of Neutron Flux Density Monitoring Method during Refueling for Tianwan NPP

作　　者：吕牛[1] 李东朋 夏兆东[1] 王杨 LYU Niu;LI Dongpeng;XIA Zhaodong;WANG Yang(Reactor Engineering Technology Research Institute,China Institute of Atomic Energy,Beijing 102413,China;Jiangsu Nuclear Power Corporation,Lianyungang 222042,China)

机构地区：[1]中国原子能科学研究院反应堆工程技术研究所,北京102413 [2]江苏核电有限公司,江苏连云港222042

出　　处：《原子能科学技术》2023年第1期140-146,共7页Atomic Energy Science and Technology

摘　　要：针对田湾核电站换料期间中子通量密度监测方式的不足,研究了用源量程探测器取代换料监测量程探测器完成换料期间中子通量密度监测的优化方案。基于全进全出和堆芯倒料两种堆芯换料方式,应用蒙特卡罗程序MCNP模拟计算源量程探测器的响应,计算结果表明,当靠近源量程探测器位置的乏燃料组件达到一定燃耗深度时,源量程探测器中子计数率达到0.5 s^(-1),满足换料过程中子计数率监测要求,与测量值符合得很好。应用源量程探测器替代换料监测量程探测器监测换料期间中子通量密度是可行的。The ex-core nuclear instrument equipment consists of the source range detector,the start-up and work range detector and the refueling monitoring range detector in Tianwan nuclear power plant(NPP).The neutron flux density of the reactor is monitored by the refueling monitoring range detector,which is a temporary device,installed into the measurement tubes before refueling and dismounted after refueling manually.It will increase radiation risk to worker,prolong overhaul period and have an impact on the economic benefits due to installing and dismounting the refueling monitoring range detector.To cope with the deficiency of neutron flux density monitoring method during refueling in Tianwan NPP,an optimization scheme was investigated to replace the refueling monitoring range detector with the source range detector to complete the neutron flux density monitoring during refueling.How to calculate source range detector neutron counting rate exactly was critical to this scheme.Firstly,the neutron emission rate versus burnup of spent fuel assemblies with initial enrichment of 2.4%,3.6%,4.0%and 4.9%were calculated using the ORIGENS program,mainly considering the spontaneous fission neutron and(α,n)neutron.Secondly,based on the actual core structure,the axial distribution of the neutron emission rate and the neutron emission spectrum of the spent fuel assembly calculated by the ORIGENS program,a Monte Carlo program MCNP was used to simulate the calculation of the neutron flux density where the source range detector was located.Finally,the neutron counting rates of the source range detector were calculated with considering about the deviation coefficient K and the conservative factor B and compared with the experimental data.In this paper,the calculation of the neutron counting rates of the source range detector were performed according to the two core refueling methods of all-in and all-out and core feeding.The calculation results show that:1)The neutron counting rate of the 6th channel source range detector is affected mo

关 键 词：换料期间 探测器响应 中子通量密度监测 

分 类 号：TL327[核科学技术—核技术及应用]