基于燃料棒层面核热耦合的超临界水堆堆芯研究  

Study on SCWR Core Characteristics Utilizing Pin-wise Neutronics/Thermal-Hydraulic Coupling Method

作　　者：洪谦[1] 牛刚[1] 汤春桃[1] 路璐[1] 史国宝[1] 

机构地区：[1]上海核工程研究设计院,上海200233

出　　处：《核科学与工程》2014年第2期193-200,共8页Nuclear Science and Engineering

基　　金：973计划(2007CB209800)

摘　　要：针对超临界水堆堆芯内流体物性分布非均匀性显著、核热反馈强烈的特点,建立了适用于超临界水堆运行环境的、基于燃料棒层面的精细化堆芯中子学/热工水力耦合方法,开发了子通道程序NCEDSCWR、节块扩散计算程序MRAPS、多功能程序COUPLE,结合西屋公司组件能谱计算程序PARAGON,构建了堆芯中子学/热工耦合分析程序系统SCAP。以具有121盒燃料组件的超临界水堆堆芯进行模拟分析,研究了堆芯三维功率分布和流体物性分布的特点以及反应性参数与重要同位素密度等随燃耗的变化规律。结果表明,本文提出的精细化核热耦合方法和开发的程序系统可以应用于超临界水堆堆芯的研究与分析,相关研究结果对超临界水堆堆芯设计具有一定的指导意义。According to the intense flow heterogeneity and strong neutronics/thermal-hydraulic feedback of SCWR, a pin-wise neutronics and thermal-hydraulic coupling method is developed and a core analysis code package SCAP is established. The package utilizes sub-channel program NCED-SCWR, lattice program PARAGON, nodal diffusion program MRAPS, and multi- functional code COUPLE. Furthermore, 3D core power distribution and flow distribution, reactivity parameters and isotope density are studied for a typical core with 121 fuel assemblies. The results show that both the coupling method and code package are applicable to the SCWR core analysis. The obtained conclusion will be benefit for SCWR core design.

关 键 词：超临界水堆 堆芯设计 核热耦合 

分 类 号：TL329[核科学技术—核技术及应用]