The paper presents the assessment of RELAP5/SCDAP code capabilities to simulate the thermal-hydraulic behavior of liquid metal coolants. The preliminary part of the study dealt with a bibliographic review of the heat transfer correlations available for liquid metals, in particular for lead-bismuth eutectic. The most appropriate correlation, according to the thermal-hydraulic condition of the Chinese ADS design has been implemented and tested in the code. The experiment facilities used for the assessment study are the South Korean HELIOS facility, the KYLIN-II facility and the TALL facility. The first one is a T-H loop which is down scaled by a factor of 5000 of the reference PEACER-300 concept reactor, and has been recently used for an international benchmark organized by the OECD/NEA. As the first phase of the benchmark is concluded, the data is available in the open literature. The second facility is constructed and operated in the Institute of Nuclear Safety Energy of the Chinese Academy of Science in Hefei (CAS). The third facility has been constructed and operated at KTH Royal Institute of Technology of Stockholm. The full height facility was designed and operated to investigate the heat transfer performance of different heat exchangers and the thermal-hydraulic characteristics of natural and forced circulation flow under steady and transient conditions. A consistent and systematic approach for the nodalization development and assessment procedures that respond to the IAEA guidelines is discussed and thoroughly applied. The present paper discusses the results of the assessment study of the RELAP5/SCDAP capability when working with liquid metal fluid. The procedures and the database developed constitute the base in our institute for further study in case more experimental data will be available.
Assessment Study of RELAP5/SCDAP Capability to Reproduce Liquid Metal Fluid Thermal Hydraulic Behaviour
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Fiori, F, & Zhou, Z. "Assessment Study of RELAP5/SCDAP Capability to Reproduce Liquid Metal Fluid Thermal Hydraulic Behaviour." Proceedings of the 2014 22nd International Conference on Nuclear Engineering. Volume 2B: Thermal Hydraulics. Prague, Czech Republic. July 7–11, 2014. V02BT09A010. ASME. https://doi.org/10.1115/ICONE22-30612
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