熔盐堆物理热工耦合程序开发及验证分析

魏泉; 郭威; 王海玲; 陈金根; 蔡翔舟

doi:10.11889/j.0253-3219.2017.hjs.40.100605

您当前的位置：

首页 >

文章列表页 >

熔盐堆物理热工耦合程序开发及验证分析

核能科学与工程 | 更新时间：2021-05-19

- 熔盐堆物理热工耦合程序开发及验证分析
- Develop and verify coupling program of the neutron physics and thermal hydraulic for MSR
- 核技术 2017年40卷第10期文章编号：100605
- 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：
  
  [ "魏泉, 男, 1986年出生, 2009年毕业于长春理工大学, 博士研究生, 研究方向为反应堆物理", "WEI Quan, male, born in 1986, graduated from Changchun University of Science and Technology in 2009, doctoral student, focusing on the physics of reactor" ]
  郭威, E-mail:guowei@sinap.ac.cn GUO Wei, E-mail:guowei@sinap.ac.cn
- 基金信息：
  
  中国科学院战略性先导科技专项(XDA02010000);若干关键问题研究项目(QYZDY-SSW-JSC016);国家自然科学基金(91326201)
- DOI：10.11889/j.0253-3219.2017.hjs.40.100605
  中图分类号： TL364
- 收稿日期：2017-04-10，
  
  修回日期：2017-05-12，
  
  纸质出版日期：2017-10
- 稿件说明：
移动端阅览
魏泉, 郭威, 王海玲, 等. 熔盐堆物理热工耦合程序开发及验证分析[J]. 核技术, 2017,40(10):100605

Quan WEI, Wei GUO, Hailing WANG, et al. Develop and verify coupling program of the neutron physics and thermal hydraulic for MSR[J]. Nuclear techniques, 2017, 40(10): 100605
魏泉, 郭威, 王海玲, 等. 熔盐堆物理热工耦合程序开发及验证分析[J]. 核技术, 2017,40(10):100605 DOI： 10.11889/j.0253-3219.2017.hjs.40.100605.

Quan WEI, Wei GUO, Hailing WANG, et al. Develop and verify coupling program of the neutron physics and thermal hydraulic for MSR[J]. Nuclear techniques, 2017, 40(10): 100605 DOI： 10.11889/j.0253-3219.2017.hjs.40.100605.

摘要

熔盐堆（Molten Salt Reactor，MSR）是第四代反应堆6种堆型中唯一的液态燃料反应堆，与固态燃料-液体冷却剂反应堆相比，原理上有较大不同。在熔盐堆中，流动的熔盐既是燃料又是冷却剂与慢化剂，中子物理学与热工水力学相互耦合；由于熔盐的流动性，缓发中子先驱核会随燃料流至堆芯外衰变，造成缓发中子的丢失，导致堆芯反应性降低。正是由于熔盐堆的这些新特性，造成熔盐堆内缓发中子先驱核、温度等参数变化与固态燃料反应堆有所不同，需要研究熔盐堆在各种工况下的相关物理参数变化。本文主要工作是考虑缓发中子先驱核的流动性对熔盐堆的影响，研究适用于熔盐堆的二维圆柱几何时空中子动力学程序及与之耦合的热工水力学程序；利用该程序对熔盐堆中子物理学和热工水力学进行耦合计算，验证熔盐堆相关实验数据；并且计算了熔盐堆无保护启停泵及堆芯入口温度过冷过热工况，用于分析熔盐堆的安全特性。计算结果表明，程序能够对熔盐反应堆实验（Molten Salt Reactor Experiment，MSRE）的相关实验数据进行较好的模拟计算，并且验证了熔盐堆的固有安全性。

Abstract

Background

As one of the six candidates for the Generation IV reactor types

molten salt reactor (MSR) is characterized by its use of the fluid-fuel. Compared with solid-fuel reactors

there are some differences in physics for liquid fuel reactor. The fluid-fuel not only serves as the fuel

but also serves as both the coolant and the moderator

hence a must for the neutron physics coupling with the thermal-hydraulic. Due to fuel flow in MSR

the delayed neutron precursors (DNP) partly flow out and decay outside of the reactor core

resulting in reactivity losses in the core.

Purpose

This study aims to develop and verify coupling program of neutron physics and thermal-hydraulic for MSR.

Methods

Distribution of DNP and temperature is different from solid-fuel reactor

it is necessary to study the change of relevant physical parameters in MSR under different operating conditions. Considering the flow of DNP impacting on MSR

a new space-time neutron dynamics program was developed to couple with the thermal-hydraulics program on the basis of two-dimensional RZ cylindrical geometry. The unprotected pump driven transient and inlet fuel temperature overcooling/overheating transient was simulated to analyze the intrinsic safety of MSR.

Results

The related experimental data in molten salt reactor experiment (MSRE) were verified by coupling calculation of neutron dynamics and thermal hydraulics

and the results showed that simulated data was in good agreement with the experiment results.

Conclusion

The program is available in simulating related experimental data in MSRE

and MSR has intrinsic safety.

关键词

Keywords

references

Doe U. A technology roadmap for generation IV nuclear energy systems[C]. Proceedings of the Nuclear Energy Research Advisory Committee and the Generation IV International Forum, 2002:239-241.

H Macpherson . The molten salt reactor adventure . Nuclear Science and Engineering , 1985 . 90 ( 4 ): 374 - 380 . DOI: 10.13182/nse90-374 http://doi.org/10.13182/nse90-374 .

Haubenreich P. Molten-salt reactor experiments[R]. Oak Ridge National Laboratory, 1970:4396.

Vergnes J, Lecarpentier D, Garzenne C, et al. The AMSTER concept[C]. Proceedings of the proceedings of the 6th OECD/NEA Information Exchange Meeting on Actinide and Fission Product Partitioning and Transmutation, Madrid, Spain, 2000.

Sekimoto H. Innovative nuclear reactor development for future[R]. Inter-COE, 2006:1.

江绵恒 , 徐洪杰 , 戴志敏 . 未来先进核裂变能——TMSR核能系统 . 中国科学院院刊 , 2012 . ( 3 ): 366 - 374 . DOI: 10.3969/j.issn.1000-3045.2012.03.016 http://doi.org/10.3969/j.issn.1000-3045.2012.03.016 .

Mianheng JIANG , Hongjie XU , Zhimin DAI . Advanced fission energy program-TMSR nuclear energy system . Bulletin of Chinese Academy of Sciences , 2012 . ( 3 ): 366 - 374 . DOI: 10.3969/j.issn.1000-3045.2012.03.016 http://doi.org/10.3969/j.issn.1000-3045.2012.03.016 .

G Lapenta , F Mattioda , P Ravetto . Point kinetic model for fluid fuel systems . Annals of Nuclear Energy , 2001 . 28 ( 17 ): 1759 - 1772 . DOI: 10.1016/s0306-4549(01)00012-3 http://doi.org/10.1016/s0306-4549(01)00012-3 .

J Kophazi , M Szieberth , S Feher , . MCNP based calculation of reactivity loss in circulating fuel reactors . Nuclear Mathematical and Computational Sciences , 2003 .

J Křepel , U Grundmann , U Rohde , . DYN1D-MSR dynamics code for molten salt reactors . Annals of Nuclear Energy , 2005 . 32 ( 17 ): 1799 - 1824 . DOI: 10.1016/j.anucene.2005.07.007 http://doi.org/10.1016/j.anucene.2005.07.007 .

J Křepel , U Rohde , U Grundmann , . DYN3D-MSR spatial dynamics code for molten salt reactors . Annals of Nuclear Energy , 2007 . 34 ( 6 ): 449 - 462 . DOI: 10.1016/j.anucene.2006.12.011 http://doi.org/10.1016/j.anucene.2006.12.011 .

D L Zhang , S Z Qiu , G H Su . Development of a safety analysis code for molten salt reactors . Nuclear Engineering and Design , 2009 . 239 ( 12 ): 2778 - 2785 . DOI: 10.1016/j.nucengdes.2009.08.020 http://doi.org/10.1016/j.nucengdes.2009.08.020 .

D L Zhang , S Z Qiu , G H Su , . Analysis on the neutron kinetics for a molten salt reactor . Progress in Nuclear Energy , 2009 . 51 ( 4 ): 624 - 636 . DOI: 10.1016/j.pnucene.2008.11.008 http://doi.org/10.1016/j.pnucene.2008.11.008 .

庄坤 , 曹良志 , 郑友琦 , . 熔盐堆动力学软件开发 . 核动力工程 , 2014 . S2 183 - 185 . http://www.cnki.com.cn/Article/CJFDTOTAL-HDLG2014S2050.htm .

Kun ZHUANG , Liangzhi CAO , Youqi ZHENG , . Development of neutron kinetic code for molten salt reactors . Nuclear Power Engineering , 2014 . S2 183 - 185 . http://www.cnki.com.cn/Article/CJFDTOTAL-HDLG2014S2050.htm .

胡田亮 , 吴宏春 , 曹良志 , . 熔盐堆稳态物理-热工耦合计算研究 . 原子能科学技术 , 2016 . 50 ( 10 ): 1823 - 1827 . DOI: 10.7538/yzk.2016.50.10.1823 http://doi.org/10.7538/yzk.2016.50.10.1823 .

Tianliang HU , Hongchun WU , Liangzhi CAO , . Neutronics and thermal-hydraulics coupling analysis of molten salt reactor at steady state . Atomic Energy Science and Technology , 2016 . 50 ( 10 ): 1823 - 1827 . DOI: 10.7538/yzk.2016.50.10.1823 http://doi.org/10.7538/yzk.2016.50.10.1823 .

Křepel J. Dynamics of molten salt reactor[D]. Prague:Czech Technical University, 2006.

Marleau G, Ebert A, Roy R. A user guide for DRAGON[R]. Version 4. Institut de génie nucléaire, Département de génie mécanique, École Polytechnique de Montréal, 2011.

N Z Cho . Fundamentals and recent developments of reactor physics methods . Nuclear Engineering and Technology , 2005 . 37 ( 1 ): 25 - 78 . http://www.dbpia.co.kr/Journal/ArticleDetail/NODE02239232 .

李磊 , 张志俭 . 并联通道瞬态流量分配方法研究 . 核动力工程 , 2010 . 31 ( 5 ): 97 - 101 . http://www.cnki.com.cn/Article/CJFDTOTAL-HDLG201005022.htm .

Lei LI , Zhijian ZHANG . Research on transient flux distribution in parallel channels . Nuclear Power Engineering , 2010 . 31 ( 5 ): 97 - 101 . http://www.cnki.com.cn/Article/CJFDTOTAL-HDLG201005022.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

熔盐堆散体石墨组件堆芯结构变化对核反应性的影响

反应堆瞬时短周期与反应性引入速率约束问题研究

液态燃料熔盐堆三维动力学程序开发及验证

后处理效率对熔盐堆燃料循环性能的影响

RELAP5缓发中子先驱核输运模型扩展及验证