Preliminary feasibility analysis of Heat Pipe Cooled Bimodal Space Nuclear Reactor

• Published in: Progress in nuclear energy (New series) 2021-08, Vol.138, p.103817, Article 103817
• Main Authors: Tian, Xiaoyan, Li, Huaqi, Jiang, Duoyu, Zhu, Lei, Chen, Sen, Kang, Xiaoya
• Format: Article
• Language: English
• Subjects: Accidents; Computational fluid dynamics; Conduction heating; Conductive heat transfer; Energy distribution; Experimental nuclear reactors; Feasibility studies; Fluid flow; Fuels; Heat pipes; Heat transfer; HP-BSNR; Hydraulics; Multiplication; Neutron flux; Neutronics analysis; Nuclear energy; Nuclear fuel elements; Nuclear fuels; Nuclear power plants; Nuclear reactors; Nuclear safety; Pipes; Power mode; Propellant transfer; Propulsion mode; Space missions; Space vehicles; Thermal-hydraulics analysis
• ISSN: 0149-1970; 1878-4224
• DOI: 10.1016/j.pnucene.2021.103817

Bimodal Space Nuclear Reactors (BSNRs) are designed to operate in two different kinds of modes, which are respectively high-power propulsion mode for movement in outer space and low-power power mode for supplying power to space vehicle-mounted systems. As BSNRs can take full use of nuclear energy generated by reactors, they are of huge potential to be applied in future space missions and thus attract much attention of researchers around the world. In this paper, a new conceptual design of Heat Pipe Cooled Bimodal Space Nuclear Reactor (HP-BSNR) was proposed. In order to evaluate the feasibility and safety of the proposed conceptual design, preliminary neutronics and thermal-hydraulics analysis for HP-BSNR were performed. For neutronics analysis, MCNP4B (Monte Carlo Natural Particle Transport Code) has been adopted to calculate the effective multiplication factor keff and neutron flux distribution. For thermal-hydraulics analysis, mathematical models such as propellant thermodynamics, propellant-fuel element heat transfer, fuel heat conduction, energy distribution and so on have been established. Then, a thermal-hydraulics analysis code named TTHA_HPBSNR (Transient Thermal-hydraulics Analysis for Heat Pipe Cooled Bimodal Space Nuclear Reactor) was developed and validated by program ELM together with experimental data from the published literature. With the self-developed code, steady state and typical transient accidents of HP-BSNR including reactivity insertion accident and partial loss of flow accident were simulated. The preliminary analysis results indicate that the proposed conceptual design of Bimodal Space Nuclear Reactor with Heat Pipe (HP-BSNR) is feasible in terms of neutronics and thermal-hydraulics analysis. •Conceptual design of Heat Pipe Cooled Bimodal Space Nuclear Reactor was proposed.•Neutron physical analysis for HP-BSNR was performed with MCNP4B.•Hydrogen properties models at high temperature was established.•A thermal-hydraulics analysis code named TTHA_HPBSNR was developed and validated.•Typical transient accidents were simulated by self-developed code TTHA_HPBSNR.