2 MW液态钍基熔盐实验堆主屏蔽温度场分析

何杰; 夏晓彬; 蔡军; 潘登; 彭玉; 黄建平; 张国庆

doi:10.11889/j.0253-3219.2016.hjs.39.040601

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2 MW液态钍基熔盐实验堆主屏蔽温度场分析

更新时间：2021-05-25

- 2 MW液态钍基熔盐实验堆主屏蔽温度场分析
- Temperature field analysis for the main shielding of the 2-MW thorium-based molten salt experimental reactor
- 核技术 2016年39卷第4期文章编号：040601
- 作者机构：
  
  1.中国科学院上海应用物理研究所嘉定园区上海 201800;
  2.中国科学院大学北京 100049
- 作者简介：
  
  何杰,男,1989年出生,2013年毕业于重庆大学,现为硕士研究生,研究领域核安全
  夏晓彬,E-mail:xiaxiaobin@sinap.ac.cn
- 基金信息：
  
  中国科学院战略性先导科技专项项目(No.XDA02050200)资助()
- DOI：10.11889/j.0253-3219.2016.hjs.39.040601
  中图分类号： TL364
- 收稿日期：2016-01-07，
  
  录用日期：2016-02-24，
  
  纸质出版日期：2016-04-25
- 稿件说明：
移动端阅览
何杰, 夏晓彬, 蔡军, 等. 2 MW液态钍基熔盐实验堆主屏蔽温度场分析[J]. 核技术, 2016,39(4):040601

Jie HE, Xiaobin XIA, Jun CAI, et al. Temperature field analysis for the main shielding of the 2-MW thorium-based molten salt experimental reactor[J]. Nuclear techniques, 2016, 39(4): 040601
何杰, 夏晓彬, 蔡军, 等. 2 MW液态钍基熔盐实验堆主屏蔽温度场分析[J]. 核技术, 2016,39(4):040601 DOI： 10.11889/j.0253-3219.2016.hjs.39.040601.

Jie HE, Xiaobin XIA, Jun CAI, et al. Temperature field analysis for the main shielding of the 2-MW thorium-based molten salt experimental reactor[J]. Nuclear techniques, 2016, 39(4): 040601 DOI： 10.11889/j.0253-3219.2016.hjs.39.040601.

摘要

反应堆主屏蔽是核反应堆的重要组成部分

用来有效降低反应堆运行时屏蔽体外的辐射剂量水平

以满足反应堆部件材料对辐射限制的要求。温度是影响反应堆主屏蔽性能的重要因素。针对2MWth液态熔盐堆(2-MW liquid-fueled molten salt experimental reactor

TMSR-LF1)

采用MCNP软件获得功率分布后

利用Fluent软件对主屏蔽进行温度场计算。计算过程中利用Python语言编写了程序(MCNP to Fluent

MTF)来实现将MCNP(Monte Carlo N Particle Transport Code)计算结果转换为功率密度的空间分布

以用户自定义函数(User-Defined Function

UDF)形式导入到Fluent

解决了MCNP计算结果不能直接导入到Fluent的问题

并分别计算了TMSR-LF1熔盐堆不同环境温度下的主屏蔽温度场分布情况。结果表明

在环境温度为5℃、18℃、25℃、30℃、35℃、40℃情况下

TMSR-LF1熔盐堆主屏蔽普通混凝土墙温度均低于要求限值

达到设计要求。

Abstract

Background:

Molten salt reactor is a fourth generation advanced reactor. The concrete wall is the key part of this high-temperature reactor shielding

so temperature field analysis is important.

Purpose:

This study attempts to calculate the temperature field of the TMSR-LF1(2-MW liquid-fueled molten salt experimental reactor) shielding

and judge if it meets the design requirements.

Methods:

In accordance with the problem that MCNP(Monte Carlo N Particle Tran

sport Code) results cannot be directly imported into Fluent

a program which converts MCNP results to the spatial distribution of power density

and imports the spatial distribution of power density into the Fluent in the form of User-Defined Function(UDF) was developed by using Python programming language to realize the coupling of the two. According to TMSR-LF1 design parameters

a one-eight physical and thermal model of the whole reactor is established

using code MCNP and Fluent. Reactor radiation shielding thermal analysis adopts the assumptions that the different environment temperatures are 5℃

18℃

25℃

30℃

35℃ and 40℃

respectively.

Results:

The maximal values of temperature and temperature gradient in the radiation shielding concrete wall are 67.42℃ and 78.40℃·m

-1

which are lower than limit values.

Conclusion:

The radiation shielding concrete wall can meet the design requirements.

关键词

Keywords

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