铅铋冷却快堆堵流事故下堵块参数对流动传热的影响

尧俊; 张熙司; 胡文军; 柴翔; 杨燕华

doi:10.11889/j.0253-3219.2018.hjs.41.020604

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铅铋冷却快堆堵流事故下堵块参数对流动传热的影响

核能科学与工程 | 更新时间：2021-05-19

- 铅铋冷却快堆堵流事故下堵块参数对流动传热的影响
- Effect of block parameter on fluid flow and heat transfer in LBE-cooled fast reactor under blockage accident
- 核技术 2018年41卷第2期文章编号：020604
- 作者机构：
  
  1.上海交通大学上海 200240
  2.中国原子能科学研究院北京 102413
- 作者简介：
  
  [ "尧俊, 男, 1994年出生, 2015年毕业于中国科学技术大学, 现为硕士研究生, 研究领域为热工水力学", "YAO Jun, male, born in 1994, graduated from University of Science and Technology of China in 2015, master student, focusing on thermal hydraulics" ]
- 基金信息：
- DOI：10.11889/j.0253-3219.2018.hjs.41.020604
  中图分类号： TL364.4
- 收稿日期：2017-11-01，
  
  修回日期：2017-12-20，
  
  纸质出版日期：2018-02-10
- 稿件说明：
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尧俊, 张熙司, 胡文军, 等. 铅铋冷却快堆堵流事故下堵块参数对流动传热的影响[J]. 核技术, 2018,41(2):020604

Jun YAO, Xisi ZHANG, Wenjun HU, et al. Effect of block parameter on fluid flow and heat transfer in LBE-cooled fast reactor under blockage accident[J]. Nuclear techniques, 2018, 41(2): 020604
尧俊, 张熙司, 胡文军, 等. 铅铋冷却快堆堵流事故下堵块参数对流动传热的影响[J]. 核技术, 2018,41(2):020604 DOI： 10.11889/j.0253-3219.2018.hjs.41.020604.

Jun YAO, Xisi ZHANG, Wenjun HU, et al. Effect of block parameter on fluid flow and heat transfer in LBE-cooled fast reactor under blockage accident[J]. Nuclear techniques, 2018, 41(2): 020604 DOI： 10.11889/j.0253-3219.2018.hjs.41.020604.

摘要

铅铋冷却快堆是第四代核能系统之一，其具有许多运行与安全性优势。但铅铋冷却快堆在运行过程中，堆芯结构材料会受到铅铋合金冷却剂的腐蚀作用，腐蚀产物在堆内堆积可能会引发堵流事故，从而导致包壳传热恶化，并影响冷却剂的流动传热效果。通过对铅铋冷却快堆单盒燃料组件建模，使用商用计算流体力学软件STAR-CCM+对不同堵块参数下的5个堵流事故工况开展了计算分析。通过对事故后包壳内壁面温度、子通道中心温度的轴向发展和堵块周围流场的轴向速度分布进行对比分析，获得了各种堵块参数对堵流事故后传热恶化、流场性质的不同影响规律。

Abstract

Background

Lead-cooled fast reactor (LFR) is one of six candidates for Generation Ⅳ nuclear systems. Although lead-bismuth eutectic (LBE) has excellent neutron properties

thermal-hydraulic performance and inherent safety features

the structure and cladding materials could be eroded by LBE coolant. The erosion products could accumulate in fuel assembly and grow as a blockage influencing the heat transfer of cladding and coolant.

Purpose

This study aims to investigate the block parameter effect on thermal-hydraulic performance of single fuel assembly in lead-bismuth subcritical reactor under various blockage accidents.

Methods

Based on actual geometry and material properties

modeling of single fuel assembly was established. Five blockage accidents were numerically simulated by the commercial computational fluid dynamics (CFD) software STAR-CCM+ with different block parameters.

Results

Temperature variations along the axial direction of cladding inner face and subchannel center were obtained and compared. Axial velocity distributions around block were analyzed. Influence of block parameters for the coolant backflow and the property of flow fields were achieved.

Conclusion

The maximum temperature rise unanimously occurs inside block. Different block parameters diversely influence flow and heat transfer.

关键词

Keywords

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