Thermal-hydraulic analysis of FAIDUS design for severe accident mitigation strategy in a medium size sodium-cooled fast reactor

被引:3
|
作者
Panigrahi, Prasant Kumar [1 ]
Velusamy, K. [1 ]
机构
[1] Homi Bhabha Natl Inst, Nucl Syst Anal Grp, Indira Gandhi Ctr Atom Res, Kalpakkam 603102, Tamil Nadu, India
关键词
FAIDUS; Melt relocation; Flow blockage; Multi-phase model; Sodium-cooled fast reactor; Accident mitigation strategy; TOTAL INSTANTANEOUS BLOCKAGE; SEVERE RECRITICALITY; FUEL-DISCHARGE; ELIMINATION; FLOW; PROPAGATION; BEHAVIOR; TESTS; MODEL;
D O I
10.1016/j.nucengdes.2021.111222
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
The impediment of core disruption is realized through the provision of diverse and redundant safety features in the design complemented by the use of inherent passive mechanisms that respond to the accident condition by ensuring convenient dispersal of molten core material from the core in the early stage of the accident. In light of this, a promising design strategy Fuel Assembly Inner DUct Structure (FAIDUS) core concept is proposed to enhance the axial discharge capability of molten fuel from core for the exclusion of core wide melt pool formation. The sequence of event during the course of relocation in the FAIDUS is analysed employing a transient integrated thermal-hydraulic model. The effectual performance of various FAIDUS design options in mitigating the accident is evaluated. Based on this, the preferred design choice conforming the laid down criteria is selected in the present study. The simulation results suggest that FAIDUS design concept delivers the best alternative way for extenuating accident progress through a comprehensive melt relocation strategy of greater reliability in an intrinsic way.
引用
收藏
页数:15
相关论文
共 50 条
  • [41] Progress on the plant design concept of Sodium-cooled Fast Reactor
    Hishida, Masahiko
    Kubo, Shigenobu
    Konomura, Mamoru
    Toda, Mikio
    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, 2007, 44 (03) : 303 - 308
  • [42] MULTIPHYSICS ANALYSIS SYSTEM FOR TUBE FAILURE ACCIDENT IN STEAM GENERATOR OF SODIUM-COOLED FAST REACTOR
    Uchibori, Akihiro
    Kikuchi, Shin
    Kurihara, Akikazu
    Hamada, Hirotsugu
    Ohshima, Hiroyuki
    PROCEEDINGS OF THE 21ST INTERNATIONAL CONFERENCE ON NUCLEAR ENGINEERING - 2013, VOL 3, 2014,
  • [43] Research on Shutdown Protection of Sodium-cooled Fast Reactor for Reactivity Accident Introduction
    Xu W.
    Duan T.
    Feng W.
    Fu H.
    Fu, Hao (fuhao@chinansc.cn), 1600, Atomic Energy Press (54): : 1433 - 1440
  • [44] Theoretical Research on Fragmentation Characteristics of Molten Corium Jet during Severe Accident in Sodium-cooled Fast Reactor
    Ge K.
    Zhang Y.
    Tian W.
    Su G.
    Qiu S.
    Yuanzineng Kexue Jishu/Atomic Energy Science and Technology, 2023, 57 (02): : 251 - 263
  • [45] Reactor design strategy to support spectral variability within a sodium-cooled fast spectrum materials testing reactor
    Scherr, Jonathan
    Tsvetkov, Pavel
    ANNALS OF NUCLEAR ENERGY, 2018, 113 : 15 - 24
  • [46] Development of relocation model for debris particles in core disruptive accident analysis of sodium-cooled fast reactor
    Teng, Chunming
    Zhang, Bin
    Shan, Jianqiang
    Zhang, Xisi
    Cao, Yonggang
    NUCLEAR ENGINEERING AND DESIGN, 2020, 368
  • [47] A Porous Media Model for Thermal-hydraulic Analysis of Wire-wrapped Fuel Assembly in Sodium Cooled Fast Reactor
    Wang X.
    Zhang D.
    Wang T.
    Qiu S.
    Su G.
    Hedongli Gongcheng/Nuclear Power Engineering, 2024, 45 (02): : 147 - 153
  • [48] Early detection of an Instantaneous Total Blockage accident in the core of a Sodium-cooled Fast Reactor
    Martinez-Martinez, Sinuhe
    Messai, Nadhir
    Nuzillard, Danielle
    Jeannot, Jean-Philippe
    2013 7TH IEEE GCC CONFERENCE AND EXHIBITION (GCC), 2013, : 450 - 455
  • [49] Prediction of sodium leakage droplet size distribution in sodium-cooled fast reactor loop
    Xu, Zhen
    Tong, Lili
    Cao, Xuewu
    PROGRESS IN NUCLEAR ENERGY, 2024, 168
  • [50] Research on Segment Design of Control Rod in Sodium-cooled Fast Reactor
    Xu L.
    Hu Y.
    Zhang J.
    1879, Atomic Energy Press (54): : 1879 - 1884