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HL-2M, a new generation of tokamak, has been built continuously in 2020 in Southwestern Institute of Physics, China National Nuclear Corporation, and the first plasma discharge was successfully achieved in 12th, Dec. This tokamak experiment device has high design parameters, and its maximum operating current reaches 3 MA. As a satellite device of ITER, it aims to explore and solve the severe physical and engineering problems. Cryogenic system is an auxiliary system of HL-2M, which mainly meets the stable operation of cryopumps and superconducting magnets. The refrigeration capacity under the condition of Liquid Nitrogen Precooling will up to 500W@4.5K, the liquefaction rate is 160L/h. In this report, we present the main characteristic parameters of the refrigeration system. At present, the system has completed the field assembly, and is expected to start commissioning in June 2021.

The heat transfer characteristics of post-CHF (critical heat flux) is very important and complicated, which has aroused wide attention in the topic of nuclear reactor safety. In this paper, a phase change model of liquid droplet during mist flow is proposed to consider the heat transfer of mist flow. Couple with this phase change model, three dimensional CFD simulation has been carried out to study the flow and heat transfer characteristics of mist flow in a wire-wrapped rod bundle. The discussion and analysis focus on the turbulent flow behavior and the heat transfer performance of post-CHF mist flow in the rod bundle. The effect of vapor properties such as density, heat capacity, thermal conductivity and viscosity on heat transfer are considered in the CFD simulation, and those properties are dependent on temperature. According to the CFD simulation result, the pressure distribution, temperature distribution, velocity distribution and phase distribution in the wire-wrapped rod bundle are analyzed in detail.

The 500 W@4.5 K helium refrigeration system operated in HL-2M Tokamak is the first domestically made cryogenic system independently designed and manufactured in China. It is expected to cool down HL-2M cryousers for the first time around June 2021. In the requirement document of helium refrigeration system, the mean time between failures (MTTF) should be at least 8000 hours, and the inherent availability of the subsystem should be 60%. This paper presents the RAMI (Reliability, Availability, Maintainability and Inspectability) analysis of the failure modes of the HL-2M helium refrigeration system as well as the corresponding effects and criticality to guide the design of the components and the preparation for operation and maintenance. The description of the risk mitigation actions regarding the design, testing, maintenance and operation in order to reach the expected targets of enhancing the reliability and availability of the refrigeration system.

The effectiveness of in-vessel retention (IVR) by external reactor vessel cooling (ERVC) strongly depends on the critical heat flux (CHF). As long as the local CHF does not exceed the local heat flux, the lower head of the pressure vessel can be cooled sufficiently to prevent from failure. In this paper, a CFD simulation is carried out to investigate the CHF of ERVC. This simulation is performed by a CFD code fluent couple with a boiling model by UDF (User-Defined Function). The experimental CHF of ERVC obtained by State Nuclear Power Technology Research and Development Center (SNPTRD) is used to validate this CFD simulation, and it is found that the simulation result agrees well with the experimental result. Based on the CFD simulation, detailed analysis focusing on the pressure distribution, velocity distribution, void fraction distribution, heating wall temperature distribution are proposed in this paper.

In pressurized water reactors (PWRs), the spacer grids of the fuel assembly has significant impact on the thermal-hydraulic performance of the fuel assembly. Particularly, the spacer grids with the mixing vanes can dramatically enhance the secondary flow and have significant effect on the void distribution in the fuel assembly. In this paper, the CFD study has been carried out to analyze the effects of the spacer grid with the steel contacts, dimples and mixing vanes on the boiling two-phase flow characteristics, such as the two-phase flow field, the void distribution, and so on. Considered the influence of the boiling phase change on two-phase flow, a boiling model was proposed and applied in the CFD simulation by using the UDF (User Defined Function) method. Furthermore, in order to analyze the effects of the spacer grid with mixing vanes, the adiabatic (without boiling) two-phase flow has also been investigated as comparison with the boiling two-phase flow in the fuel assembly with spacer grids. The CFD simulation on two-phase flow in the fuel assembly with the proposed boiling model can predict the characteristics of two-phase flow better.

The turbulent flow behavior and heat transfer performance of the rod bundle with spacer grid have significant effect on the thermal–hydraulic characteristics in Pressurized Water Reactor (PWR). The complex structure of the spacer grid includes dimple, mixing vane and spring. In this paper, different turbulence models are coupled with the Computational Fluid Dynamics (CFD) analysis for the complex flow in rod bundle, and the numerical results are compared with the experimental results with temperature distribution at outlet. This paper focuses on thermal–hydraulic behavior downstream of the spacer grid caused by mixing vane, dimple and spring. Two types of rod bundles are investigated, including with spacer grid and without spacer grid. According to the numerical study, the strong secondary flow induced by the spacer grid gradually decay along the rod bundle after the mixing vane, and this secondary flow has significant effect on the heat transfer. In addition, a normalized Nusselt number (Nu) and swirl factor are defined to investigate the thermal–hydraulic performance of the rod bundle.

In order to investigate single bubble evolution, a boiling phase change model in subcooled flow boiling is proposed in this paper, and VOF model combined with phase change model is adopted to simulate the single bubble growth and movement. The effects of flow velocity, liquid subcooling, wall superheat, and vapor-liquid contact angle are considered in this model. The predicted bubble growth curve agrees well with the experimental result. Based on the analysis of bubble shape evolution and temperature field, it is found that the average bubble growth rate, flow velocity, and dynamic contact angle have significant effect on the bubble shape evolution during the bubble growth and movement while the temperature gradient in superheated liquid does not change with bubble growing. The character of dynamic contact angle during bubble growth and movement is also obtained in different working condition.