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基于统一几何建模的高保真物理-热工耦合方法研究及其在SPERT实验堆堆芯的计算应用
引用本文:张旻婉,刘宙宇,王博,曹璐,赵晨,曹良志.基于统一几何建模的高保真物理-热工耦合方法研究及其在SPERT实验堆堆芯的计算应用[J].核动力工程,2021,42(5):42-50.
作者姓名:张旻婉  刘宙宇  王博  曹璐  赵晨  曹良志
作者单位:西安交通大学,西安,710049;中国核动力研究设计院核反应堆系统设计技术重点实验室,成都,610213
基金项目:中国核工业集团有限公司领创科研项目
摘    要:针对各类小型动力堆或实验堆开展物理-热工耦合模拟计算时,由于非规则几何结构的存在而带来物理-热工网格映射关系复杂且不可统一预置的问题,基于数值反应堆高保真物理计算程序NECP-X开展了基于统一几何建模的物理-热工耦合方法研究,基于中子学模型建立物理-热工耦合的映射关系,并结合NECP-X程序中的瞬态计算方法实现了特殊功率偏移实验(SPERT)实验堆堆芯的直接瞬态计算;计算了SPERT实验堆稳态算例并与蒙特卡罗程序的结果进行对比,在此基础上,对SPERT实验堆进行了瞬态计算分析并与实验值进行对比。结果表明,NECP-X程序中子学计算的特征值和棒功率分布计算结果具有较高的精度;基于统一几何建模的网格映射方法可以方便快捷地实现复杂几何压水堆的物理-热工耦合计算;与实验值相比,瞬态计算的总功率、反应性随时间的变化曲线具有较高的精度,并且可提供精细的功率及温度分布。 

关 键 词:复杂几何  物理-热工耦合  SPERT实验堆
收稿时间:2020-08-25

Study on High-Fidelity Thermal-Neutronic Coupling Method Based on the Unified Geometry Modeling and its Application in Experimental Reactor Core Calculation for SPERT
Affiliation:1.Xi’an Jiaotong University, Xi’an, 710049, China2.Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China, Chengdu, 610213, China
Abstract:To solve the problem that the thermal-neutronic grid mapping relation is complicated and cannot be preset in a centralized manner due to the existence of irregular geometry in performing the thermal-neutronic coupled simulation calculation for various small power reactors and experimental reactors, this paper studies the thermal-neutronic coupling method based on the unified geometric modeling, using the high-fidelity numerical code for reactor neutronics calculation, NECP-X. This study establishes the mapping relation for the thermal-neutronic coupling on the basis of the neutronics model, and enables the direct transient calculation of the experimental reactor core for the special power excursion reactor test (SPERT) via combination with the transient calculation method in NECP-X. Then, this study calculates the steady-state case for the experimental reactors of SPERT, and compares the calculation results with the results gained from the Monte Carlo code. On this basis, this study conducts transient calculation and analysis for these experimental reactors and compares the corresponding results with the experimental results. The final results show that the eigenvalues from the neutronics calculation by the NECP-X and the rod power distribution calculation results are of high accuracy; that the grid mapping method based on the unified geometric modeling allows a easy and fast thermal-neutronic coupled calculation of the PWRs of complex geometry; and that compared with the experimental values, the curve of change in the total power and reactivity gained from transient calculation with time is more accurate and can provide refined distributions of power and temperature. 
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