Xenon analysis of thorium molten salt experiment reactor-liquid fuel

Chen GUO; Ruimin JI; Jingen CHEN; Guimin LIU

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

您当前的位置：

首页 >

文章列表页 >

Xenon analysis of thorium molten salt experiment reactor-liquid fuel

NUCLEAR ENERGY SCIENCE AND ENGINEERING | 更新时间：2021-05-19

- Xenon analysis of thorium molten salt experiment reactor-liquid fuel
- NUCLEAR TECHNIQUES Vol. 40, Issue 4, Article number: 040602(2017)
- 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：
- 基金信息：
- DOI：10.11889/j.0253-3219.2017.hjs.40.040602
  CLC： TL32
- Received：26 December 2016，
  
  Revised：25 February 2017，
  
  Published：10 April 2017
- 稿件说明：
移动端阅览
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：

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 .

Views

下载量

CSCD

Alert me when the article has been cited

提交

Tools

Publicity Resources

Nuclear safety analysis of molten salt reactor criticality with changes in ⁷Li abundance

Uncertainty and sensibility analysis of reactivity insertion transient accident of a 150 MWt molten salt reactor (SM-MSR)

Development and validation of a thermal-hydraulic analysis code for dual cooled assemblies in molten salt reactors

Applicability analysis of reduced order modeling methods for fluid dynamics in molten salt reactor

Criticality safety analysis of nuclear fuel storage of molten salt reactor

Related Author

YANG Zhen

DAI Zhimin

ZOU Yang

YAO Yongkui

WANG Kai

WANG Chaoqun

YANG Qun

HE Zhaozhong

Related Institution

Shanghai Institute of Applied Physics, Chinese Academy of Sciences

North-Western China Regional Office of Nuclear and Radiation Safety Inspection, National Nuclear Safety Administration

North-Western China Regional Office of Nuclear and Radiation Saftety Inspection, National Nuclear Safety Administration

ShanghaiTech University

Chengdu University of Technology

AI问答

Address：2019 Jialuo Road, Jiading, Shanghai, China Postal code：201800
Tel：021-39194977 Email：LHB@sinap.ac.cn
Technical support is provided by Beijing Founder electronics co., LTD 沪ICP备05005479号-1 京公网安备11010802024621
It is recommended to read the content of this site in Chrome&IE9+. Please switch to extreme mode in browser 360.
Cookies We use cookies to help provide and enhance our service and tailor content. By continuing, you agree to the use of cookies.
Total Visits：881268 Visits Today：795

⁰