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• Concept of "LESS is MORE" is proposed in the design of lead-bismuth cooled reactor. • Core design of lead-bismuth cooled small modular reactor is conducted. • Neutronic and thermal-hydraulic performances of reactor core are analyzed and improved. • Preliminary core design of LBE cooled reactor could satisfy the design criteria and is reasonable. Small modular reactors (SMR) are in great demands in the near future and the lead-based fast reactor is one of the most important directions in the development of SMR. In this paper, core design and analysis of a LEad-bismuth Small MOdular Reactor (LESMOR) was developed and examined to realize the notion of "LESS is MORE". LESMOR is a 300 MWt pool-type reactor where all facilities are mounted in a reactor vessel. Advanced nitride fuel is adopted, and the spent-plutonium is used as the driven fuel and thorium is used as the fertile fuel. Sub-channel analysis was performed in the assembly design using an in-house sub-channel code SUBAS, and the 11 × 11 scheme with a pitch-to-diameter (P/D) ratio of 1.4 was adopted. The full core neutronic performance was studied and evaluated in this paper. Results show that the reactor can be maintained critical for 20 years and the maximum radial power factor is 1.28. The steady-state thermal-hydraulic performance of the core was analyzed, including the characteristics of the mass flow distribution and the hottest assembly. The peaking fuel cladding temperature is 519.2 °C, and the maximum fuel center temperature is 723.4 °C, both within the temperature limit. This paper provides valuable information in the neutronic and thermal-hydraulic design and analysis of LBE nuclear reactor. [ABSTRACT FROM AUTHOR]

• A code is developed for studying the operating thermal-hydraulic characteristics of the reactor core, which consists of core thermal-hydraulic model, neutron kinetics model and heat pipe startup model. • In order to verify the code, a series of experiments about potassium heat pipe are operated. • Steady-state and startup transient are performed by the code. • This research provides valuable experience for designing and formulating control strategies of the heat pipe cooled reactor. Heat pipe cooled reactor has been proposed for applying in the space station and underwater exploration featured with high reliability, low noise level and compact structure. A typical design of a heat pipe cooled reactor system is HP-STMCs designed by the University of New Mexico. It is conceptually designed as a fast reactor, lithium-filled heat pipes that transfer heat to potassium heat pipes radiator and thermoelectric generators. In this paper, a code is developed for studying the operating thermal–hydraulic characteristics of the reactor core, which consists of core thermal–hydraulic model, neutron kinetics model and heat pipe startup model. In order to verify the code, a series of experiments about potassium heat pipe are operated. Steady-state and startup transient are performed by the code. The highest temperatures of fuel and matrix in reactor core show at the 2nd channel, which reach 1720 K and 1545 K from the results of steady-state. As for the startup transient, the code for heat pipe cooled reactor core successfully predicts key parameters during startup transient, such as each layer material temperature of reactor core, a single heat pipe's temperature response and heat rejection power. The code has initially predicted the transient response of the heat pipe cooled reactor under power increasing rates of 800 W/s and130W/s. It is discussed that the startup strategy of power increasing rates is based on heat pipe transition temperature (916 K). This research provides valuable experience for designing and formulating control strategies of the heat pipe cooled reactor. [ABSTRACT FROM AUTHOR]

The lead-based reactor is one of the fourth-generation nuclear energy systems with excellent development prospects. It has attracted much attention because of its unique advantages in fuel transmutation and breeding, passive safety, and economic efficiency. In this study, the main features of typical lead-based fast reactors (LFRs) worldwide are reviewed and current challenges in their development are pointed out. Moreover, the research progress in four main thermal-hydraulic aspects is overviewed and discussed, including the flow and heat transfer characteristics of the coolant, thermal-hydraulic analysis of the reactor core, lead pool and reactor system. These aspects are further divided into specific issues according to the developing needs of LFR. For each issue, up-to-date research findings, existing difficulties, and future trends are presented. This study can serve as a reference for the design and thermal-hydraulic analysis of LFRs. • Research on the main thermal-hydraulic features of LFRs is discussed. • Flow and heat transfer characteristics of lead-based coolants are summarized and compared. • Up-to-date research and developments of Core/Pool/System Level thermal-hydraulics are reviewed. [ABSTRACT FROM AUTHOR]