Please wait a minute...
 首页  期刊介绍 期刊订阅 联系我们 横山亮次奖 百年刊庆
 
最新录用  |  预出版  |  当期目录  |  过刊浏览  |  阅读排行  |  下载排行  |  引用排行  |  横山亮次奖  |  百年刊庆
清华大学学报(自然科学版)  2021, Vol. 61 Issue (11): 1301-1307    DOI: 10.16511/j.cnki.qhdxxb.2020.22.036
  核能与新能源工程 本期目录 | 过刊浏览 | 高级检索 |
HTR-10超高温运行堆芯温度场分析
孙世妍, 张佑杰, 郑艳华, 夏冰
清华大学 核能与新能源技术研究院, 先进核能技术协同创新中心, 先进反应堆工程与安全教育部重点实验室, 北京 100084
Core temperature distributions in HTR-10 operating at very high temperatures
SUN Shiyan, ZHANG Youjie, ZHENG Yanhua, XIA Bing
Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Collaborative Innovation Center of Advanced Nuclear Energy Technology, Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084, China
全文: PDF(4237 KB)   HTML
输出: BibTeX | EndNote (RIS)      
摘要 10 MW高温气冷实验堆(HTR-10)在额定工况下满功率稳态运行时,燃料的温度裕度较大,存在将冷却剂出口温度在700℃的基础上进一步提升的潜力,对实现球床式高温气冷堆超高温运行具有重要意义。该文根据堆芯发热功率和冷却剂载出热量之间的平衡关系,为HTR-10设计了多个稳态超高温运行工况,运用改进的热工模型分析了初装堆芯下各工况的堆芯温度场,给出其分布特征,并讨论了燃料分布不均匀性对堆芯最高温度的影响。结果表明:当堆芯出口冷却剂温度达到1 000℃,且假设最高温度区燃料球和石墨球分布的不均匀程度达到极限时,堆芯最高温度仍未达到燃料温度限值。
服务
把本文推荐给朋友
加入引用管理器
E-mail Alert
RSS
作者相关文章
孙世妍
张佑杰
郑艳华
夏冰
关键词 球床式高温气冷堆10 MW高温气冷实验堆(HTR-10)超高温运行热工水力分析堆芯温度场    
Abstract:The 10 MW high temperature gas-cooled reactor test module (HTR-10) has a relatively large fuel temperature margin when operating at full power and steady state for the rated condition; thus, the coolant outlet temperature can be further increased above 700℃, which is useful for pebble-bed very high temperature gas-cooled reactors (HTGR) to operate with high coolant temperature at the core outlet. Several steady-state very high temperature operating conditions were designed for HTR-10 in this study based on a heat balance between the core heating power and the heat carried by the coolant. An improved thermal hydraulic model was used to analyze the core temperature distribution at each condition for the initial core to determine the temperature distribution characteristics, and the maximum core temperature for the nonuniform fuel distribution was discussed. For the coolant outlet temperature of 1 000℃, the predicted maximum core temperature is still below the fuel temperature limit, even when the most serious nonuniform distribution of the fuel and graphite balls occurs.
Key wordspebble-bed high temperature gas-cooled reactor (HTGR)    10 MW high temperature gas-cooled reactor test module (HTR-10)    operating at very high temperatures    thermal hydraulic analysis    core temperature distribution
收稿日期: 2020-09-11      出版日期: 2021-10-19
基金资助:国家高技术研究发展计划项目(2014AA052701)
通讯作者: 张佑杰,教授,E-mail:zhangyj@tsinghua.edu.cn     E-mail: zhangyj@tsinghua.edu.cn
引用本文:   
孙世妍, 张佑杰, 郑艳华, 夏冰. HTR-10超高温运行堆芯温度场分析[J]. 清华大学学报(自然科学版), 2021, 61(11): 1301-1307.
SUN Shiyan, ZHANG Youjie, ZHENG Yanhua, XIA Bing. Core temperature distributions in HTR-10 operating at very high temperatures. Journal of Tsinghua University(Science and Technology), 2021, 61(11): 1301-1307.
链接本文:  
http://jst.tsinghuajournals.com/CN/10.16511/j.cnki.qhdxxb.2020.22.036  或          http://jst.tsinghuajournals.com/CN/Y2021/V61/I11/1301
  
  
  
  
  
  
  
  
  
  
  
[1] 王捷. 高温气冷堆技术背景和发展潜力的初步研究[J]. 核科学与工程, 2002, 22(4):325-330.WANG J. Preliminary study on technical base and future potential of high temperature gas-cooled reactor[J]. Chinese Journal of Nuclear Science and Engineering, 2002, 22(4):325-330. (in Chinese)
[2] SCHULTEN R. The AVR nuclear power plant:A milestone in high-temperature reactor development[J]. Nuclear Engineering and Design, 1985, 90(4):388-390.
[3] BAUST E, RAUTENBERG J, WOHLER J. Results and experience from the commissioning of the THTR 300[J]. Atomkernenergie-Kerntechnik, 1985, 47:141-144.
[4] WU Z X, LIN D C, ZHONG D X. The design features of the HTR-10[J]. Nuclear Engineering and Design, 2002, 218(1-3):25-32.
[5] US DOE Nuclear Energy Research Advisory Committee, Generation IV International Forum. A technology roadmap for generation IV nuclear energy systems[R]. Washington DC, USA:US DOE Nuclear Energy Research Advisory Committee, Generation IV International Forum, 2002.
[6] 陈福冰. 利用HTR-10试验数据对安全分析程序THERMIX的验证[D]. 北京:清华大学, 2009.CHEN F B. Validation of the safety analysis code THERMIX by using the HTR-10 experimental data[D]. Beijing:Tsinghua University, 2009. (in Chinese)
[7] 徐小琳, 曲荣红, 边晖. 10 MW高温气冷实验堆首次装料和趋近临界[J]. 高技术通讯, 2001(3):104-106.XU X L, QU R H, BIAN H. The first loading and critical experiment for 10 MW high temperature gas-cooled reactor[J]. High Technology Letters, 2001(3):104-106. (in Chinese)
[8] 刘俊杰, 王敏稚, 张征明, 等. 10 MW高温气冷实验堆的堆体结构特点[J]. 核动力工程, 2001(1):53-56.LIU J J, WANG M Z, ZHANG Z M, et al. Features of reactor structure design for 10 MW high temperature gas-cooled reactor[J]. Nuclear Power Engineering, 2001(1):53-56. (in Chinese)
[9] GAO Z Y, SHI L. Thermal hydraulic calculation of the HTR-10 for the initial and equilibrium core[J]. Nuclear Engineering and Design, 2002, 218(1):51-64.
[10] CLEVELAND J C, GREENE S R. Application of THERMIX-KONVEK code to accident analyses of modular pebble bed high temperature reactors (HTRs)[R]. Oak Ridge, USA:Oak Ridge National Laboratory, 1986.
[11] SUN S Y, ZHANG Y J, ZHENG Y H. Research on influence of different simulation methods of bypass flow in thermal hydraulic analysis on temperature distribution in HTR-10[J]. Science and Technology of Nuclear Installations, 2020, 2020:4754589.
[12] KADAK A C, BAZANT M Z. Pebble flow experiments for pebble bed reactors[C]//Proceedings of 2nd International Topical Meeting on High Temperature Reactor Technology. Beijing, China, 2004:H05.
[1] 郝琛, 李富, 郭炯. 球床式高温气冷堆球流混流的模拟[J]. 清华大学学报(自然科学版), 2014, 54(5): 624-628.
Viewed
Full text


Abstract

Cited

  Shared   
  Discussed   
版权所有 © 《清华大学学报(自然科学版)》编辑部
本系统由北京玛格泰克科技发展有限公司设计开发 技术支持:support@magtech.com.cn