Development of a thermal-hydraulic analysis code for annular fuel assemblies

被引:0
|
作者
Vishnoi, A. K. [1 ]
Chandraker, D. K. [1 ]
Vijayan, P. K. [1 ]
机构
[1] Bhabha Atom Res Ctr, Bombay 400085, Maharashtra, India
关键词
D O I
暂无
中图分类号
TL [原子能技术]; O571 [原子核物理学];
学科分类号
0827 ; 082701 ;
摘要
In this work a detailed study of the annular fuel has been carried out. A thermal hydraulics code, ANUFAN (Annular Fuel Analysis), based on the bundle average method, capable of modeling both internally and externally cooled annular fuel pins is developed. Code predictions have been compared with calculations from Korea Atomic Energy Research Institute (KAERI) and MIT Heat transfer fraction difference between ANUFAN and RELAP was found about 1.7%. Analysis of a 54 - fuel rod assembly is carried out with 36 and 45 numbers of annular fuel pins keeping the same channel size and bundle power as of the solid fuel assembly. Fuel pin maximum temperature of the annular fuel is found much less than the solid fuel. MCHFR value for annular fuel is found much higher compared to that of the solid fuel of 54 - fuel rod assembly. The full paper covers the details of the computer code, the analysis carried out and the results obtained.
引用
下载
收藏
页码:12 / 17
页数:6
相关论文
共 50 条
  • [21] Development and Verification of Thermal-Hydraulic Transient Analysis Code in Plate-Type Fuel Nuclear Reactor
    Liu W.
    Zhang Y.
    Jiang X.
    Zhang C.
    Zhang D.
    Hedongli Gongcheng/Nuclear Power Engineering, 2019, 40 (05): : 18 - 22
  • [22] PRELIMINARY DEVELOPMENT ON THERMAL-HYDRAULIC ANALYSIS CODE FOR THE SPENT FUEL ROD UNDER THE CONDITION OF SPRAY COOLING
    Guo Chao
    Deng Jian
    Cai Rong
    Ma Yugao
    Liu Lili
    Zhang Yuhao
    Lv Siyu
    PROCEEDINGS OF THE 2020 INTERNATIONAL CONFERENCE ON NUCLEAR ENGINEERING (ICONE2020), VOL 3, 2020,
  • [23] Development of a thermal-hydraulic analysis code for research reactors with plate fuels
    State Key Laboratory of Multi Phase Flow in Power Engineering, Xi'an Jiaotong University, Xi'an, 710049, China
    不详
    Ann Nucl Energy, 4 (433-447):
  • [24] Development and validation of thermal-hydraulic and safety analysis code for CARR (TSACC)
    Tian, Wen-Xi
    Qiu, Sui-Zheng
    Su, Guang-Hui
    Jia, Dou-Nan
    Liu, Xing-Min
    Zhang, Jian-Wei
    Hedongli Gongcheng/Nuclear Power Engineering, 2009, 30 (01): : 40 - 44
  • [25] Uncertainty analysis of neutronic/thermal-hydraulic coupling in pressurized water reactor fuel assemblies
    Kong, Deyan
    Gao, Zhibo
    Wu, Di
    Cheng, Jie
    Wang, Jianjun
    ANNALS OF NUCLEAR ENERGY, 2024, 207
  • [26] Development of a thermal-hydraulic analysis code for research reactors with plate fuels
    Lu, Qing
    Qiu, Suizheng
    Su, G. H.
    ANNALS OF NUCLEAR ENERGY, 2009, 36 (04) : 433 - 447
  • [27] Thermal-hydraulic performance evaluation of annular fuel based on modified FROBA-ANNULAR
    Li C.
    Liao H.
    Wu Y.
    Tian W.
    Su G.H.
    Qiu S.
    International Journal of Advanced Nuclear Reactor Design and Technology, 2021, 3 : 108 - 118
  • [28] Measurements for verification and validation of thermal-hydraulic computer code used for thermal-hydraulic analysis of VVER440 typical fuel assembly
    Ezsol, Gyorgy
    Hutli, Ezddin
    Valer, Gottlasz
    Gabor, Zsiros
    MEASUREMENT, 2021, 171
  • [29] COBRA STAR GCFR, A COMPUTER CODE FOR THERMAL-HYDRAULIC ANALYSIS OF GCFR FUEL ASSEMBLY
    BAXI, CB
    BURHOP, CJ
    BENNETT, FO
    TRANSACTIONS OF THE AMERICAN NUCLEAR SOCIETY, 1978, 30 (NOV): : 543 - 545
  • [30] THERMAL-HYDRAULIC ANALYSIS OF ANNULAR FUEL-PINS IN A LOW-POWER-DENSITY PWR
    CAREY, C
    ANGHAIE, S
    DUGAN, ET
    TRANSACTIONS OF THE AMERICAN NUCLEAR SOCIETY, 1984, 47 : 482 - 483