Assessment of the CTF subchannel code for modeling a large-break loss-of-coolant accident reflood transient

Journal Article · Tue Aug 20 00:00:00 EDT 2024 · Annals of Nuclear Energy

DOI:https://doi.org/10.1016/j.anucene.2024.110831· OSTI ID:2438944

Salko, Robert ^[1]; Wysocki, Aaron ^[1]; Hizoum, Belgacem ^[1]; Capps, Nathan ^[1]

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1. Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)

With increased industry interest in extending reactor operating cycles, the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program has been investigating the behavior of high-burnup fuel during design basis accidents such as the large-break loss-of-coolant accident (LBLOCA) with consideration for risk of fuel fragmentation, relocation, and dispersal (FFRD). As part of that activity, the NEAMS subchannel thermal/ hydraulics (T/H) code, CTF, is being used for modeling of LBLOCA and to determine the impact of subchannel resolution on results. Although CTF includes a wide range of models for LBLOCA conditions, the code has not been used for this application while maintained at Oak Ridge National Laboratory (ORNL) until now. Therefore, here, in this work, a preliminary assessment of several of these models was performed using openly available reflood experimental data from the Flooding Experiments in Blocked Arrays (FEBA) tests. One coarse mesh and one fine mesh model were set up in CTF for high and low flooding rate tests performed in the unblocked FEBA facility. A coarse TRACE model was set up to be as consistent as possible with the coarse CTF model to allow for code-to-code benchmarking. The assessment shows a tendency of the codes to over-predict peak cladding temperature (PCT) near the top of the bundle and to quench early. Advanced spacer grid models were shown to improve upper bundle predictions in CTF. The resolved CTF model over-predicted PCT by a larger degree in the center channels in the low-flooding rate test, and it is believed that the radiative heat transfer model, which was not used in this study, may be needed to correct this over-prediction. Finally, this work demonstrates the importance of the droplet model in determining quench time and vapor temperature and PCT prediction, which necessitates a more in-depth validation of these models in the future.

• Research Organization: Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)
• Sponsoring Organization: USDOE National Nuclear Security Administration (NNSA), Office of Naval Reactors; USDOE Office of Nuclear Energy (NE); USDOE Office of Nuclear Energy (NE), Nuclear Energy Advanced Modeling and Simulation (NEAMS)
• Grant/Contract Number: AC05-00OR22725; AC07-05ID14517
• OSTI ID: 2438944
• Alternate ID(s): OSTI ID: 2437834
• Journal Information: Annals of Nuclear Energy, Journal Name: Annals of Nuclear Energy Vol. 210; ISSN 0306-4549
• Publisher: Elsevier
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
CTF
FEBA
LOCA
ORNL