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Sub‐channel‐based multiscale thermal‐hydraulic analysis of a medium‐power, lead‐cooled fast reactor
Author(s) -
Cao Liankai,
Zhang Xilin,
Chen Hongli
Publication year - 2020
Publication title -
international journal of energy research
Language(s) - English
Resource type - Journals
SCImago Journal Rank - 0.808
H-Index - 95
eISSN - 1099-114X
pISSN - 0363-907X
DOI - 10.1002/er.5279
Subject(s) - thermal hydraulics , nuclear engineering , coolant , transient (computer programming) , materials science , cladding (metalworking) , thermal , scram , rod , mechanics , mechanical engineering , environmental science , computer science , engineering , heat transfer , thermodynamics , physics , medicine , alternative medicine , pathology , metallurgy , operating system
Summary The lead‐cooled fast reactor (LFR) offers enhanced safety and reliability with the fine properties of liquid lead and lead alloy. To study accurately the thermal characteristics of fast reactors, the multiscale thermal‐hydraulic coupling simulation is an effective way. Multiscale coupling based on the sub‐channel code has evident advantages on the analysis of fuel assemblies. In this study, a multiscale thermal‐hydraulic analysis of a forced‐circulation, medium‐power LFR under steady‐state and transient conditions is performed with the system code ATHLET and sub‐channel code KMC‐SUBtraC which was developed based on the previous version by modifying the pressure drop correlations and adding the assembly‐level calculation. The codes are one‐way‐coupled, with good efficiency and precision. Transient verification of the sub‐channel code is conducted with the CFD code. In the steady‐state analysis of M 2 LFR‐1000, mass flow and temperature distributions of the assemblies, sub‐channels, and fuel rods in the hottest assembly are analyzed and the safety performance is investigated. In the transient analysis, two typical DECs (unprotected overpower transient and ULOF+ULOHS) are simulated and the multiscale thermal‐hydraulic characteristics are analyzed. With the negative reactivity feedback, the variations of the temperatures of the coolant and fuel rods are within the safe limits, which shows the inherent safety of the reactor. And the results indicate that the loss of primary flow could increase the risk of cladding corrosion.