TY - GEN
T1 - Modeling the Prismatic HTGR Core in SAM by Representative Channels
AU - Zhang, Taiyang
AU - Hua, Thanh
AU - Ooi, Zhiee Jhia
AU - Zou, Ling
AU - Brooks, Caleb S.
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 - 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.
AB - 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.
KW - high-temperature gas reactor
KW - prismatic core
KW - SAM
KW - system analysis
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U2 - 10.13182/NURETH20-40030
DO - 10.13182/NURETH20-40030
M3 - Conference contribution
AN - SCOPUS:85202961173
T3 - Proceedings of the 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023
SP - 3084
EP - 3097
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 -