Modeling the Prismatic HTGR Core in SAM by Representative Channels

Taiyang Zhang, Thanh Hua, Zhiee Jhia Ooi, Ling Zou, Caleb S. Brooks

Research output: Chapter in Book/Report/Conference proceedingConference contribution

Abstract

High-temperature gas reactors (HTGRs) are drawing growing attention as the nuclear power industry marches towards more advanced systems.Confronting the complexity in the core geometry and multi-scale heat transfer mechanisms, system analysis of HTGRs faces the challenge of reaching reasonable fidelity and accuracy while maintaining sufficient simplicity.Previous development and validation have demonstrated that a “2-D ring model” can be implemented with acceptable performance for prismatic HTGR cores using the System Analysis Module (SAM) code.In the current study, an alternative methodology is proposed based on representative channels with better reserved local heat transfer geometry.Different from the previous ring-model approach, this methodology separately models and then combines small-and large-scale thermal conductions, by connecting representative local heat transfer structures through effective core-wise thermal resistance.The procedure is demonstrated by an example modeling practice for an integral high-temperature test facility (HTTF).Steady-state prediction is assessed against a higher-resolution benchmark from a 3D-1D coupled simulation, which shows reasonably captured global parameters and well-represented temperature fields.A postulated pressurized conduction cooldown (PCC) is also simulated and analyzed, demonstrating the model capability of transient prediction with physically captured phenomena resolved in both small and large scales.In general, this work achieves a preliminary success in proposing an effective methodology using representative channels to model prismatic HTGR cores in SAM.

Original languageEnglish (US)
Title of host publicationProceedings of the 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023
PublisherAmerican Nuclear Society
Pages3084-3097
Number of pages14
ISBN (Electronic)9780894487934
DOIs
StatePublished - 2023
Externally publishedYes
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
Period8/20/238/25/23

Keywords

  • high-temperature gas reactor
  • prismatic core
  • SAM
  • system analysis

ASJC Scopus subject areas

  • Nuclear Energy and Engineering
  • Instrumentation

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