Title: Thermal-hydraulic analysis of lead-bismuth cooled reactors core based on single channel model

Authors: Huaping Mei; Shuyong Liu; Chao Chen; Jiansong Zhang; Taosheng Li

Addresses: Institute of Nuclear Energy Safety Technology, Hefei Institutes of Physical Science, Chinese Academy of Sciences, Hefei, Anhui Province, China ' Institute of Nuclear Energy Safety Technology, Hefei Institutes of Physical Science, Chinese Academy of Sciences, Hefei, Anhui Province, China; University of Science and Technology of China, Hefei, Anhui, China ' Institute of Nuclear Energy Safety Technology, Hefei Institutes of Physical Science, Chinese Academy of Sciences, Hefei, Anhui Province, China; University of Science and Technology of China, Hefei, Anhui, China ' Institute of Nuclear Energy Safety Technology, Hefei Institutes of Physical Science, Chinese Academy of Sciences, Hefei, Anhui Province, China; University of Science and Technology of China, Hefei, Anhui, China ' Institute of Nuclear Energy Safety Technology, Hefei Institutes of Physical Science, Chinese Academy of Sciences, Hefei, Anhui Province, China; University of Science and Technology of China, Hefei, Anhui, China

Abstract: Lead-bismuth reactor is an important choice of micro-reactors. Based on the main design parameters of a 5 MW lead-bismuth reactor with typical fuel bundle structure, this paper discusses the interaction law of fuel element diameter, linear power density, coolant velocity, maximum cladding temperature, active zone volume and pressure drop of coolant with a developed program module which is based on a heat transfer model of fuel elements and a single coolant channel model. The result indicated that the thermal limits of fuel elements are mainly influenced by the maximum cladding temperature rather than the melting point of UO₂ pellet, and the fuel elements with small diameter can realise a higher linear power and a smaller core volume. This research can provide a reference to the miniaturisation design of a liquid metal cooled reactor.

Keywords: thermal-hydraulic analysis; lead-bismuth cooled reactor; core; single channel model.

DOI: 10.1504/IJNEST.2024.142757

International Journal of Nuclear Energy Science and Technology, 2024 Vol.17 No.2/3, pp.180 - 196

Received: 21 Jan 2024
Accepted: 02 Sep 2024
Published online: 20 Nov 2024 *

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