Air Ingress and DLOFC Studies in Scaled HTGR Geometry using Additively Manufactured TCR Fuel Elements

Broderick Sieh, Hitesh Bindra, Prashant Jain, Christian Petrie, Nathan See

Research output: Chapter in Book/Report/Conference proceedingConference contributionpeer-review

Abstract

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.

Original languageEnglish
Title of host publicationProceedings of the 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023
PublisherAmerican Nuclear Society
Pages3416-3427
Number of pages12
ISBN (Electronic)9780894487934
DOIs
StatePublished - 2023
Event20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023 - Washington, United States
Duration: Aug 20 2023Aug 25 2023

Publication series

NameProceedings of the 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023

Conference

Conference20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023
Country/TerritoryUnited States
CityWashington
Period08/20/2308/25/23

Keywords

  • DLOFC
  • TCR

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