DEVELOPMENT OF A THERMAL-HYDRAULIC CODE FOR CHINESE HELIUM-COOLED CERAMIC BREEDER TBM COOLING SYSTEM

被引:0
|
作者
Wang, Jie [1 ]
Su, Guanghui [1 ]
Tian, Wenxi [1 ]
Qiu, Suizheng [1 ]
机构
[1] Xi An Jiao Tong Univ, Sch Nucl Sci & Technol, Xian 710049, Peoples R China
关键词
DESIGN; LOOP;
D O I
暂无
中图分类号
TH [机械、仪表工业];
学科分类号
0802 ;
摘要
Helium was chosen as the coolant for divertor cooling loop, Korea helium cooled solid breeder TBM, European helium cooled pebble bed TBM and Chinese helium cooled ceramic breeder TBM. The thermal-hydraulic analysis for the divertor cooling loop and the TBM cooling systems were carried out by RELAP5 and MELOCR codes, which were developed for transient simulation of light water reactor coolant system during postulated accidents. In this study, a thermal-hydraulic and safety analysis code (TSACO) for helium cooling system was developed and calculation of the Chinese HCCB TBM cooling system was presented. Heat transfer and flow friction models for helium were added in the code. First, the code was verified by comparing with the RELAP5 code with the same initial and boundary conditions. The first wall temperature, pressure drop and inlet/outlet temperatures were studied and a good agreement was obtained, then ex-vessel loss of coolant accident for Chinese HCCB-TBM cooling system was investigated using TSACO. The results show that the TBM can be cooled efficiently and the TCWS pressure stays within the design limits with a large margin.
引用
收藏
页数:9
相关论文
共 50 条
  • [31] Development of subchannel thermal-hydraulic analysis code for dual cooled annular fuel
    Saffari, A. H.
    Esmaili, H.
    PROGRESS IN NUCLEAR ENERGY, 2022, 150
  • [32] DEVELOPMENT OF THERMAL-HYDRAULIC AND SAFETY ANALYSIS CODE FOR A HEAT PIPE COOLED REACTOR
    Jiao, Guanghui
    Xia, Genglei
    Zhou, Tao
    Wang, Jianjun
    PROCEEDINGS OF 2024 31ST INTERNATIONAL CONFERENCE ON NUCLEAR ENGINEERING, VOL 4, ICONE31 2024, 2024,
  • [33] Thermal-hydraulic calculation and analysis on helium cooled ceramic breeder pebble bed assembly for in-pile irradiation & in-situ tritium extraction
    Guo, Chun-Qiu
    Xie, Jia-Chun
    Liu, Xing-Min
    Yuanzineng Kexue Jishu/Atomic Energy Science and Technology, 2013, 47 (06): : 940 - 946
  • [34] Development of subchannel thermal-hydraulic analysis code for dual cooled annular fuel
    Saffari, A. H.
    Esmaili, H.
    PROGRESS IN NUCLEAR ENERGY, 2022, 150
  • [35] Research and development of a transient thermal-hydraulic code for system safety analysis of sodium cooled fast reactor
    Yu, Yang
    Liu, Dong
    Song, Xiaoming
    Li, Zhongchun
    Guo, Fengchen
    Wang, Yafeng
    Pang, Bo
    ANNALS OF NUCLEAR ENERGY, 2021, 152
  • [36] Development of a thermal-hydraulic analysis code for the Pebble Bed Water-cooled Reactor
    Cai, Xiaoyu
    Qiu, Suizheng
    Tian, Wenxi
    Su, Guanghui
    NUCLEAR ENGINEERING AND DESIGN, 2011, 241 (12) : 4978 - 4988
  • [37] Numerical research on the neutronic/thermal-hydraulic/mechanical coupling characteristics of the optimized helium cooled solid breeder blanket for CFETR
    Cui, Shijie
    Zhang, Dalin
    Cheng, Jie
    Tian, Wenxi
    Su, G. H.
    FUSION ENGINEERING AND DESIGN, 2017, 114 : 141 - 156
  • [38] NUMERICAL ANALYSIS ON THE THERMAL-HYDRAULIC CHARACTERISTICS FOR THE REACTOR MAIN VESSEL COOLING SYSTEM OF CHINESE SODIUM COOLED FAST REACTOR
    Song, Ping
    Feng, Tangtao
    Zhang, Dalin
    Chen, Lie
    Li, Shaodan
    Lin, Yuansheng
    Qiu, Suizheng
    PROCEEDINGS OF 2021 28TH INTERNATIONAL CONFERENCE ON NUCLEAR ENGINEERING (ICONE28), VOL 2, 2021,
  • [39] Neutronics influence of poloidal nonuniform neutron wall loading on helium-cooled ceramic breeder blanket for CFETR
    Lian, Qiang
    Tian, Wenxi
    Qiu, Suizheng
    Su, G. H.
    FUSION ENGINEERING AND DESIGN, 2020, 161
  • [40] Development of a thermal-hydraulic analysis code for CARR
    Tian, WX
    Qiu, SZ
    Guo, Y
    Su, GH
    Jia, DN
    ANNALS OF NUCLEAR ENERGY, 2005, 32 (03) : 261 - 279