钍基液态燃料熔盐实验堆氙毒分析

郭晨; 冀锐敏; 陈金根; 刘桂民

doi:10.11889/j.0253-3219.2017.hjs.40.040602

您当前的位置：

首页 >

文章列表页 >

钍基液态燃料熔盐实验堆氙毒分析

核能科学与工程 | 更新时间：2021-05-19

- 钍基液态燃料熔盐实验堆氙毒分析
- Xenon analysis of thorium molten salt experiment reactor-liquid fuel
- 核技术 2017年40卷第4期文章编号：040602
- 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
  3.中国科学院先进核能创新研究院上海 201800
- 作者简介：
  
  [ "郭晨, 男, 1989年出生, 2012年毕业于大连大学, 现为博士研究生, 研究领域为核能科学与工程", "GUO Chen, male, born in 1989, graduated from Dalian University in 2012, doctoral student, focusing on nuclear science and engineering" ]
  刘桂民, E-mail:liuguimin@sinap.ac.cn LIU Guimin, E-mail:liuguimin@sinap.ac.cn
- 基金信息：
  
  Strategic Priority Research Program of Chinese Academy of Sciences(XDA02010000);Thorium Uranium Fuel Cycle Characteristics and Key Problem Research Project(QYZDY-SSW-JSC016);National Natural Science Foundation of China(91326201)
- DOI：10.11889/j.0253-3219.2017.hjs.40.040602
  中图分类号： TL32
- 收稿日期：2016-12-26，
  
  修回日期：2017-02-25，
  
  纸质出版日期：2017-04-10
- 稿件说明：
移动端阅览
郭晨, 冀锐敏, 陈金根, 等. 钍基液态燃料熔盐实验堆氙毒分析[J]. 核技术, 2017,40(4):040602

Chen GUO, Ruimin JI, Jingen CHEN, et al. Xenon analysis of thorium molten salt experiment reactor-liquid fuel[J]. Nuclear techniques, 2017, 40(4): 040602
郭晨, 冀锐敏, 陈金根, 等. 钍基液态燃料熔盐实验堆氙毒分析[J]. 核技术, 2017,40(4):040602 DOI： 10.11889/j.0253-3219.2017.hjs.40.040602.

Chen GUO, Ruimin JI, Jingen CHEN, et al. Xenon analysis of thorium molten salt experiment reactor-liquid fuel[J]. Nuclear techniques, 2017, 40(4): 040602 DOI： 10.11889/j.0253-3219.2017.hjs.40.040602.

摘要

熔盐堆作为第四代核能系统堆型之一，液态燃料形态的特点使其可以实现在线处理和在线添料。为了提高中子经济性可以利用在线处理的氦鼓泡法，将氦气通入反应堆一回路，去除堆芯内的裂变气体（如Xe、Kr）。基于钍基熔盐液态堆（Thorium Molten Salt Reactor-Liquid Fuel1，TMSR-LF1）概念设计，结合熔盐实验堆（Molten Salt Reactor Experiment，MSRE）氙毒模型，分析了鼓泡法去除氙毒中

135

Xe扩散规律和去除效率对氙毒的影响，并给出了对应的初始有效增殖因子的变化规律。分析结果表明，虽然存在

135

Xe会大量向石墨扩散的可能性，但是鼓泡法仍然可以有效去除TMSR-LF1堆芯内的

135

Xe，减小堆芯毒性，提高反应性。

Abstract

Background

Thorium molten salt reactor-liquid fuel (TMSR-LF) is a thermal molten salt reactor (MSR) with on-line reprocessing to extract waste and poison elements. Helium bubbling technique is considered to extract waste gas i.e.

from the reactor core in order to improve neutron economics.

Purpose

This study aims to implement the MSRE

135

Xe poison model

and apply it on TMSR-LF1 of Shanghai Institute of Applied Physics (SINAP) to estimate xenon poison and effective multiplication factor.

Methods

135

Xe poison model with helium bubbling technique was built based on the molten salt reactor experiment (MSRE) research conducted at the Oak Ridge National Laboratory (ORNL) in the 1960s. Compared with present simulation which ignores the

135

Xe diffusion behaviors in the fluid salt and graphite

this model provides a better description for a MSR. The nuclear reaction from Scale6.1 and diffusion process on the MATLAB platform were combined for

135

Xe analysis in TMSR-LF1.

Results

The result shows that helium bubbling technique can remove

135

Xe and reduce xenon poison effectively. Most of

135

Xe will stay in graphite.

Conclusions

Xenon poison model considering

135

Xe diffusion and flow effect is reasonable.

关键词

Keywords

references

T Abram , S Ion . Generation-Ⅳ nuclear power: a review of the state of the science . Energy Policy , 2008 . 36 4323 - 4330 . DOI: 10.1016/j.enpol.2008.09.059 http://doi.org/10.1016/j.enpol.2008.09.059 .

E S Bettis , W B Cottrell . The aircraft reactor experiment operation . Nuclear Science and Engineering , 1957 . 2 ( 6 ): 841 - 853 .

Haubenreich P N, Engel J R. Experience with the molten salt reactor experiment[R]. Oak Ridge National Laboratory, 1969.

Journee D. Helium bubbling in a molten salt fast reactor[D]. Delft: Delft University of Technology, 2014.

Scott D, Eatherly W P. Graphite and xenon behavior and their influence on molten salt reactor design[R]. Oak Ridge National Laboratory, 1969.

E Merle-Lucotte , L Mathieu , D Heuer , . Influence of the processing and salt composition on the thorium molten salt reactor . Nuclear Technology , 2008 . 163 ( 3 ): 358 - 365 . https://www.researchgate.net/profile/Ludovic_Mathieu/publication/266479452_Influence_of_the_Processing_and_Salt_Composition_on_the_Thorium_Molten_Salt_Reactor/links/54b3c2930cf2318f0f9585f6.pdf .

L Mathieu , D Heuer . The thorium molten salt reactor: moving on from the MSBR . Progress in Nuclear Energy , 2006 . 48 ( 7 ): 664 - 679 . DOI: 10.1016/j.pnucene.2006.07.005 http://doi.org/10.1016/j.pnucene.2006.07.005 .

Heuer D, Merle-Lucotte E, Allbert M, et al . Simulation tools and new developments of the molten salt fast reactor[C]. Proceedings of the European Nuclear Conference ENC2010, Barcelona, Spain, 2010. DOI: 10.1051/rgn/20106095. http://dx.doi.org/10.1051/rgn/20106095.

M Aufiero , A Cammi , C Fiorina , . An extended version of the SERPENT-2 code to investigate fuel burn-up and core material evolution of the molten salt fast reactor . Journal of Nuclear Materials , 2013 . 441 ( 1-3 ): 473 - 486 . DOI: 10.1016/j.jnucmat.2013.06.026 http://doi.org/10.1016/j.jnucmat.2013.06.026 .

Kell R J, Houtzeel A. Development of a model for computing Xe-135 migration in the MSRE[R]. ORNL-4069, Oak Ridge National Laboratory, 1967.

Engel J R, Steffy R C. Xenon behavior in the molten salt reactor experiment[R]. ORNL-TM-3464, Oak Ridge National Laboratory, 1971.

SCALE: a comprehensive modeling and simulation suite for nuclear safety analysis and design[R]. Oak Ridge National Laboratory, 2006.

B R Betzler , J J Powers , A Worrall . Molten salt reactor neutronics and fuel cycle modeling and simulation with SCALE . Annals of Nuclear Energy , 2017 . 101 489 - 503 . DOI: 10.1016/j.anucene.2016.11.040 http://doi.org/10.1016/j.anucene.2016.11.040 .

陈杰 . MATLAB宝典 , : 北京电子工业出版社 , 2007 . http://www.cnki.com.cn/Article/CJFDTOTAL-SYQY201603027.htm .

Jie CHEN . MATLAB treasure , : Beijing Electronic Industry Press , 2007 . http://www.cnki.com.cn/Article/CJFDTOTAL-SYQY201603027.htm .

谢仲生 , 邓力 . 中子输运理论数值计算方法 , : 西安西北工业大学出版社 , 2005 . http://www.cnki.com.cn/Article/CJFDTOTAL-SYQY201603027.htm .

Zhongsheng XIE , Li DENG . Numerical calculation method for the neutron transport theory , : Xi'an Northwestern Polytechnical University Press , 2005 . http://www.cnki.com.cn/Article/CJFDTOTAL-SYQY201603027.htm .

浏览量

下载量

CSCD

文章被引用时，请邮件提醒。

提交

工具集

关联资源

熔盐堆临界与⁷Li丰度变化的核安全分析

150 MWt熔盐堆（SM-MSR）反应性引入事故后果的不确定性及参数敏感性分析

双面冷却熔盐堆组件热工分析程序开发与验证

熔盐堆流体动力学模型降阶方法适用性分析

熔盐堆核燃料盐贮存的核临界安全分析