TY - GEN
T1 - VALIDATION OF THE THERMAL-HYDRAULIC MODELS IN COBRA-TF BASED ON THE NRC/PSU RBHT BENCHMARK
AU - Jin, Yue
AU - Cheung, Fan Bill
AU - Bajorek, Stephen M.
AU - Tien, Kirk
AU - Hoxie, Chris L.
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 - In order to increase the safety margin and economic efficiency of nuclear power plants, accurate prediction of the reactor core thermal-hydraulic behavior especially during accidents is crucial. However, current thermal-hydraulic codes are still largely limited by their modeling capabilities for the post-dryout flow and heat transfer regimes. Fortunately, with the NRC/PSU RBHT experimental data recently obtained under the ARTHUR program, we are able to answer many challenging questions in post-dryout regimes and to further improve the code performance. In the current study, detailed code validation work is performed utilizing the COBRA-TF sub-channel analysis code to evaluate its liquid entrainment prediction capabilities for the two-phase flow reflood transient simulations based on the NRC/PSU RBHT benchmark. Key figures of merit (FoMs) for reflood thermal-hydraulics are evaluated including: the quench front propagation, peak cladding temperature (PCT) variations, two-phase flow mixture mass flow as well as separate phase mass flow rate. The predicted results are investigated for each of these quantities in detail to gain a better understanding of the code capabilities. A comprehensive evaluation of the models used for liquid entrainment in COBRA-TF is presented, which will be instructive and useful for future development of improved models and code upgrade.
AB - In order to increase the safety margin and economic efficiency of nuclear power plants, accurate prediction of the reactor core thermal-hydraulic behavior especially during accidents is crucial. However, current thermal-hydraulic codes are still largely limited by their modeling capabilities for the post-dryout flow and heat transfer regimes. Fortunately, with the NRC/PSU RBHT experimental data recently obtained under the ARTHUR program, we are able to answer many challenging questions in post-dryout regimes and to further improve the code performance. In the current study, detailed code validation work is performed utilizing the COBRA-TF sub-channel analysis code to evaluate its liquid entrainment prediction capabilities for the two-phase flow reflood transient simulations based on the NRC/PSU RBHT benchmark. Key figures of merit (FoMs) for reflood thermal-hydraulics are evaluated including: the quench front propagation, peak cladding temperature (PCT) variations, two-phase flow mixture mass flow as well as separate phase mass flow rate. The predicted results are investigated for each of these quantities in detail to gain a better understanding of the code capabilities. A comprehensive evaluation of the models used for liquid entrainment in COBRA-TF is presented, which will be instructive and useful for future development of improved models and code upgrade.
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U2 - 10.13182/NURETH20-40025
DO - 10.13182/NURETH20-40025
M3 - Conference contribution
AN - SCOPUS:85202947090
T3 - Proceedings of the 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023
SP - 5512
EP - 5525
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 -