CFETR集成设计平台的核热耦合模块开发

徐兵兵; 徐坤; 叶民友; 毛世峰; 雷明准; 彭学兵

doi:10.11889/j.0253-3219.2018.hjs.41.010601

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CFETR集成设计平台的核热耦合模块开发

核能科学与工程 | 更新时间：2021-05-19

- CFETR集成设计平台的核热耦合模块开发
  封面论文
- Neutronics-thermal hydraulics coupling module for CFETR integration design platform
- 核技术 2018年41卷第1期文章编号：010601
- 作者机构：
  
  1.中国科学技术大学核科学技术学院合肥 230027
  2.中国科学院等离子体物理研究所合肥 230031
- 作者简介：
  
  [ "徐兵兵, 男, 1990年出生, 2012年毕业于武汉大学, 现为硕士研究生, 研究领域为核科学与核技术", "XU Bingbing, male, born in 1990, graduated from Wuhan University in 2012, master student, focusing on nuclear science and technology" ]
  叶民友, E-mail:yemy@ustc.edu.cn YE Minyou, E-mail:yemy@ustc.edu.cn
- 基金信息：
  
  国家磁约束核聚变能发展研究专项(2014GB110000)
- DOI：10.11889/j.0253-3219.2018.hjs.41.010601
  中图分类号： TL99
- 收稿日期：2017-02-21，
  
  修回日期：2017-05-16，
  
  纸质出版日期：2018-01-10
- 稿件说明：
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徐兵兵, 徐坤, 叶民友, 等. CFETR集成设计平台的核热耦合模块开发[J]. 核技术, 2018,41(1):010601

Bingbing XU, Kun XU, Minyou YE, et al. Neutronics-thermal hydraulics coupling module for CFETR integration design platform[J]. Nuclear techniques, 2018, 41(1): 010601
徐兵兵, 徐坤, 叶民友, 等. CFETR集成设计平台的核热耦合模块开发[J]. 核技术, 2018,41(1):010601 DOI： 10.11889/j.0253-3219.2018.hjs.41.010601.

Bingbing XU, Kun XU, Minyou YE, et al. Neutronics-thermal hydraulics coupling module for CFETR integration design platform[J]. Nuclear techniques, 2018, 41(1): 010601 DOI： 10.11889/j.0253-3219.2018.hjs.41.010601.

摘要

正在研发的中国聚变工程实验堆（China Fusion Engineering Test Reactor，CFETR）集成设计平台包括物理设计平台和工程设计平台，工程设计平台采用模块化方式，包括磁体、真空室、偏滤器、中子学等模块。模块设计中涉及用计算流体动力学（Computational Fluid Dynamics，CFD）软件来对堆内部件开展热工水力分析，CFD的热源项包括中子学计算产生的核热，而中子学分析软件（如Monte Carlo N particle transport code，MCNP）核热输出结果文件存在CFD软件无法直接读取等问题。因此，基于网格-网格插值和点-点插值法，开发了中子学与CFD核热耦合模块，提供两种途径实现高精度的三维核热耦合。使用CFX软件，以CFETR的一种氦冷陶瓷包层（Helium Cooled Ceramic Breeder blanket，HCCB）中增殖单元模块为对象，进行了热工水力分析，计算结果表明了核热耦合模块功能的可靠性。

Abstract

Background

The China fusion engineering test reactor (CFETR) under development

mainly consists of a physical design platform and an engineering design platform

and the latter includes various modules. In the design of the modules

computational fluid dynamics (CFD) software is used for thermal-hydraulic analysis of the fusion in-vessel components of which the nuclear heating is a necessary source term generated from neutronics calculations of Monte Carlo N particle transport code (MCNP).

Purpose

This study aims to transfer the 3D nuclear heating profiles from MCNP to CFD software and develop neutronics-thermal hydraulics coupling module for CFETR integration design platform.

Methods

The 3D nuclear heating profiles were processed by two different interpolation methods: element-to-element mapping method and point-to-point mapping method

and the 1D nuclear heating profiles were also processed for comparison. Thermal-hydraulic analysis based on breeder unit of the helium cooled ceramic breeder blanket (HCCB) for CFETR was performed on ANSYS CFX for verification. The total interpolated nuclear heating

heating distributions and temperature distributions were compared in detail according to the CFX simulation results.

Results

Comparison results showed that all schemes satisfied the conservation of nuclear heating with appropriate mesh. Element-to-element mapping method and point-to-point mapping method differs in the heating peak value due to the different interpolation method.

Conclusion

The neutronics-thermal hydraulics coupling module proposed in this paper possesses high reliability for CFETR integration design platform.

关键词

Keywords

references

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