A review of recent advances in HTGR CFD and thermal fluid analysis

被引:27
作者
Huning, Alexander J. [1 ]
Chandrasekaran, Sriram [2 ]
Garimella, Srinivas [2 ]
机构
[1] Oak Ridge Natl Lab, Oak Ridge, TN 37830 USA
[2] Georgia Inst Technol, Sustainable Thermal Syst Lab, GWW Sch Mech Engn, Atlanta, GA 30332 USA
关键词
High temperature gas reactor; Safety analysis; Thermal fluid behavior; Core heat transfer; Plenum flow;
D O I
10.1016/j.nucengdes.2020.111013
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 High Temperature Gas-cooled Reactor (HTGR) is an advanced reactor design being pursued by several different domestic and international organizations due to its high outlet temperature and inherent safety features. This paper spotlights some of the recent advances in experimental thermal fluid behavior and safety studies for the HTGR designs. Core heat transfer, plenum flow, and transient event sequence phenomena, or potential accident phenomena, are principally discussed here. Most of these advances arise from the increasing application of computational fluid dynamics (CFD) to fluid behavior in the reactor vessel under normal and transient conditions. With advanced modeling, some novel design improvements could reduce or eliminate potentially undesirable phenomena such as 'hot streaking' and vessel heat up in excess of their design limits. For air ingress accident purposes, CFD simulations are necessary to predict time scales and gas concentration fractions in the vessel. These modeling advances, however, suggest the need for additional experimental validation. Still somewhat lacking, however, are analyses that tie recent vessel and reactor cavity experimental flow results to expected HTGR operation, which is necessary to validate Loss of Forced Circulation (LOFC) type events. Modeling and simulation of these events have the potential to illustrate the hallmark safety feature of the HTGR, which is indefinite or near-indefinite safe-shutdown without any operator intervention or electrical power. Future experiments should then estimate or measure core heat transfer effects to show that fuel design limits are met over the entire length of the accident. The corresponding results could validate the various industry thermal fluid and systems analysis tools for HTGRs.
引用
收藏
页数:23
相关论文
共 123 条
  • [91] Siefken L. J., 2001, Report No. INEL-96/0422
  • [92] Staudenmeier J., 2007, 205550001 DC US NUCL
  • [93] Analysis of dust and fission products in a pebble bed NGNP
    Stempniewicz, M. M.
    Winters, L.
    Caspersson, S. A.
    [J]. NUCLEAR ENGINEERING AND DESIGN, 2012, 251 : 433 - 442
  • [94] Sterbentz J.W., 2003, INEELEXT0300870 INL, DOI [10.2172/910732, DOI 10.2172/910732]
  • [95] A review of HTGR graphite dust transport research
    Sun, Qi
    Peng, Wei
    Yu, Suyuan
    Wang, Kaiyuan
    [J]. NUCLEAR ENGINEERING AND DESIGN, 2020, 360
  • [96] Study of the deposition of graphite dust in the inlet passageway of intermediate heat exchanger in VHTR
    Sun, Qi
    Hai, Xiao
    Wang, Kaiyuan
    Peng, Wei
    [J]. EXPERIMENTAL AND COMPUTATIONAL MULTIPHASE FLOW, 2019, 1 (01) : 29 - 37
  • [97] A numerical study of particle deposition in HTGR steam generators
    Sun, Qi
    Chen, Tao
    Peng, Wei
    Wang, Jie
    Yu, Suyuan
    [J]. NUCLEAR ENGINEERING AND DESIGN, 2018, 332 : 70 - 78
  • [98] T.A.C. Technologies, 2000, NEV SOFTW PACK QUICK
  • [99] Numerical investigation of a heat transfer within the prismatic fuel assembly of a very high temperature reactor
    Tak, Nam-Il
    Kim, Min-Hwan
    Lee, Won Jae
    [J]. ANNALS OF NUCLEAR ENERGY, 2008, 35 (10) : 1892 - 1899
  • [100] DEVELOPMENT OF A CORE THERMO-FLUID ANALYSIS CODE FOR PRISMATIC GAS COOLED REACTORS
    Tak, Nam-Il
    Lee, Sung Nam
    Kim, Min-Hwan
    Lim, Hong Sik
    Noh, Jae Man
    [J]. NUCLEAR ENGINEERING AND TECHNOLOGY, 2014, 46 (05) : 641 - 654