TY - GEN
T1 - Air Ingress and DLOFC Studies in Scaled HTGR Geometry using Additively Manufactured TCR Fuel Elements
AU - Sieh, Broderick
AU - Bindra, Hitesh
AU - Jain, Prashant
AU - Petrie, Christian
AU - See, Nathan
N1 - Publisher Copyright:
© 2023 Proceedings of the 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023. All rights reserved.
PY - 2023
Y1 - 2023
N2 - Generation IV high-temperature gas-cooled reactors (HTGR) are designed to passively dissipate decay heat from the core during certain shut-down scenarios. One such scenario, known as depressurized loss of forced cooling (DLOFC), occurs after a break in the inlet/outlet header, allowing helium to leak from the reactor. The gas leak may be followed by air ingress oxidizing the core components, leading to inefficient passive decay heat removal. The Transformational Challenge Reactor (TCR) has similar features to those of an HTGR, but the primary difference is the use of an additively manufactured fuel element arrangement. The more compact, additively manufactured ceramic fuel elements can be conveniently produced with optimally configured channels that suppress the air ingress progress and improve thermofluidic performance. Computational fluid dynamics (CFD) models are used to evaluate and optimize TCR fuel designs, but experimental data are required for the validation of the simulations prior to their application in design optimization studies. DLOFC and air ingress are experimentally studied in a scaled HTGR flow test setup. The experiments use TCR additively manufactured fuel elements embedded with flow and temperature sensors. With the test elements aligned within a lowercase h-shaped tube flow geometry, further data can be gathered to improve the understanding of DLOFC. The additively manufactured parts are compared before and after several thermal transient and air ingress cycles to determine whether any macroscopic changes occur structurally. Furthermore, the data collected will be used to validate design models and improve additively manufactured reactor components.
AB - Generation IV high-temperature gas-cooled reactors (HTGR) are designed to passively dissipate decay heat from the core during certain shut-down scenarios. One such scenario, known as depressurized loss of forced cooling (DLOFC), occurs after a break in the inlet/outlet header, allowing helium to leak from the reactor. The gas leak may be followed by air ingress oxidizing the core components, leading to inefficient passive decay heat removal. The Transformational Challenge Reactor (TCR) has similar features to those of an HTGR, but the primary difference is the use of an additively manufactured fuel element arrangement. The more compact, additively manufactured ceramic fuel elements can be conveniently produced with optimally configured channels that suppress the air ingress progress and improve thermofluidic performance. Computational fluid dynamics (CFD) models are used to evaluate and optimize TCR fuel designs, but experimental data are required for the validation of the simulations prior to their application in design optimization studies. DLOFC and air ingress are experimentally studied in a scaled HTGR flow test setup. The experiments use TCR additively manufactured fuel elements embedded with flow and temperature sensors. With the test elements aligned within a lowercase h-shaped tube flow geometry, further data can be gathered to improve the understanding of DLOFC. The additively manufactured parts are compared before and after several thermal transient and air ingress cycles to determine whether any macroscopic changes occur structurally. Furthermore, the data collected will be used to validate design models and improve additively manufactured reactor components.
KW - DLOFC
KW - TCR
UR - http://www.scopus.com/inward/record.url?scp=85202980338&partnerID=8YFLogxK
U2 - 10.13182/NURETH20-40256
DO - 10.13182/NURETH20-40256
M3 - Conference contribution
AN - SCOPUS:85202980338
T3 - Proceedings of the 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023
SP - 3416
EP - 3427
BT - Proceedings of the 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023
PB - American Nuclear Society
T2 - 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023
Y2 - 20 August 2023 through 25 August 2023
ER -