TY - JOUR
T1 - A review of recent advances in HTGR CFD and thermal fluid analysis
AU - Huning, Alexander J.
AU - Chandrasekaran, Sriram
AU - Garimella, Srinivas
N1 - Publisher Copyright:
© 2020
PY - 2021/3
Y1 - 2021/3
N2 - 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.
AB - 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.
KW - Core heat transfer
KW - High temperature gas reactor
KW - Plenum flow
KW - Safety analysis
KW - Thermal fluid behavior
UR - http://www.scopus.com/inward/record.url?scp=85099001054&partnerID=8YFLogxK
U2 - 10.1016/j.nucengdes.2020.111013
DO - 10.1016/j.nucengdes.2020.111013
M3 - Review article
AN - SCOPUS:85099001054
SN - 0029-5493
VL - 373
JO - Nuclear Engineering and Design
JF - Nuclear Engineering and Design
M1 - 111013
ER -