Hui CHENG

• PhD
• Associate Porfessor at Harbin Engineering University

As compared with the other Generation-IV nuclear reactor systems, Sodium-cooled Fast Reactor (SFR) has become the “first seed” in the world due to its great breeder characteristics and the richest construction and operation experience. In this chapter, the development history and current situation, typical configurations, and safety characteristics...

Motivated to understand the pressure-buildup characteristics in a circumstance of a water droplet immerged inside a heavy liquid metal pool, which is a key phenomenon during a Steam Generator Tube Rupture accident of Lead-cooled Fast Reactor, many experiments have been conducted by injecting water lumps into a molten lead pool at Sun Yat-sen Univer...

Micro and Small Modular Reactor (MSMR) is an emerging energy technology that meets the requirements of market demand, safety, efficiency, and sustainability. This paper summarizes the advantages, application scenarios, and advanced technologies to support MSMR. Now that the energy market is more flexible and the requirements are more complex, while...

A R T I C L E I N F O Keywords: Critical heat flux Nucleate boiling Bubble behavior Percolative nature Nucleation site density A B S T R A C T Recent findings showed that the boiling crisis might be caused by the percolative nature of bubble interactions. Based on the new funding, previous work has shown the possibility of predicting critical heat...

In this study, melt-jets with elliptical cross sections are employed with different melt temperature, water temperature, penetration velocity and water depth to explore the influence of the jet cross-sectional shape on melt-jet fragmentation behavior. When free fall distance is larger than 40 cm, the jet cross-sectional shape has no obvious influen...

Three-dimensional numerical study on the release and migration of bubbles in liquid metal pool is carried out for severe accidents in advanced reactors. The fission gas bubble migration will affect the release and distribution of radioactive materials in the molten pool, and then affect the evaluation of the source term in the reactor. The simulati...

To study the interaction between molten non-eutectic alloys and subcooled water during severe nuclear accidents, an experimental investigation was carried out by injecting molten lead–bismuth non-eutectic alloy (LBNE, 70% Pb-30% Bi) into the water in a free-fall style using the visualized thermo-hydraulic characteristics in melt coolant interaction...

The lead-cooled fast reactor (LFR) is one of the most promising fast neutron reactors using molten lead or the lead–bismuth eutectic (LBE) alloy as a coolant. Under postulated severe accidents, the fuel rod of LFR may be damaged, which would cause the release of fission gas, and the migration of fission gas bubbles in the reactor molten pool will a...

During a core disruptive accident of Sodium-cooled Fast Reactors (SFR), discharged corium may contact with coolant in form of jet, which can be an imminent threat to the integrity of reactor vessel if the jet does not sufficiently break up and cooldown. It is significant to ascertain the morphologies, penetration depth and breakup mechanism of disc...

In a core meltdown accident in light water reactors, molten corium may drop into the lower plenum of the pressure vessel and interact with water, which is called fuel–coolant interaction (FCI). The behavior of the corium jet breakup in water during FCIs is important for the in-vessel retention strategy and has been extensively studied. While in pre...

Investigations on the molten-pool sloshing behavior are of essential value for improving nuclear safety evaluation of Core Disruptive Accidents (CDA) that would be possibly encountered for Sodium-cooled Fast Reactors (SFR). This paper is aimed at synthesizing the knowledge from our recent studies on molten-pool sloshing behavior with solid particle...

In the present work, to investigate the thermo-hydraulic characteristics of molten lead jet fragmentation in water, visualized fragmentation experiments are carried out by releasing superheated molten lead into subcooled water at different experimental conditions using the VTMCI (Visualized Thermo-hydraulic characteristics in Melt-Coolant Interacti...

Studies about using nanofluids to enhance the Critical Heat Flux (CHF) of In-vessel Retention (IVR) strategy in the third-generation reactor have been conducted extensively and show a significant CHF enhancement effect. However, low carbon steel SA508 used in the reactor vessel is easy to oxidize and the oxidation can lead to changes in the surface...

In a postulated Steam Generator Tube Rupture (SGTR) accident of pool-type Lead-cooled Fast Reactor (LFR), highly-pressurized water from the secondary circuit will be discharged into the primary vessel containing low-pressure molten lead-based alloy (commonly pure lead or low-melting-point Lead-Bismuth Eutectic (LBE)). One of the dangers of such acc...

In this paper, the melt jet breakup behavior are numerically studied in 3D with nonorthogonal central-moment MRT color-gradient lattice Boltzmann method, which could significantly enhance the numerical stability and accuracy when applied to flows with very high Reynolds number. Firstly, the methodology to simulate immiscible two-phase flow is valid...

The loss of coolant accident (LOCA), as one of the design basis accidents (DBAs), is a hypothetical accident that is usually considered in the design of nuclear power plant. LOCA is caused by small/large breaks in the reactor primary coolant and pressure boundary and may result in a loss of reactor coolant at a rate in excess of the reactor makeup...

In this chapter the applications of advanced multiphase lattice Boltzmann methods (LBMs) to enhance the understanding of the molten fuel–coolant interaction (FCI) during a severe accident are presented. Since the FCI is very complex due to the strong thermodynamic and hydrodynamic coupling, which means that it is difficult to understand all the mec...

In this study, to enhance the understanding of the thermo-hydraulic characteristics during the direct contact of molten lead-bismuth eutectic (LBE) and water, a series of visualized fragmentation experiments were conducted by releasing molten LBE into a subcooled water pool using the VTMCI (Visualized Thermo-hydraulic characteristics in Melt-Coolan...

Studies on the sloshing motion of a molten fuel pool are important for the improved assessment of core disruptive accidents that might occur in sodium-cooled fast reactors. In this paper, the progress of recent studies on sloshing motion in a liquid pool with solid particles performed at the Sun Yat-Sen University are summarized and discussed. In t...

In this paper, the color-gradient lattice Boltzmann model (Saito et al., Phys. Re. E 98, 013305, 2018) is used to study 3D hydrodynamic corium jet breakup and fragmentation in sodium. Firstly, an in-house GPU-accelerated color-gradient lattice Boltzmann solver is introduced in detail. Then, the solver is validated by comparing the simulation result...

Diameter effect on heat transfer deterioration (HTD) for supercritical water upward flow in circular tubes and annular channels at low mass flux and high heat flux was studied numerically based on validated turbulence model. When the same boundary conditions were applied, i.e., mass flux, heat flux, and inlet temperature, it was observed that for c...

Heat transfer deterioration is numerically studied for supercritical fluids flowing upward in circular tubes at high heat fluxes and low mass fluxes. The simulations are conducted with Shear Stress Transport (SST) k-ω turbulent model in commercial software Fluent 15.0. Both water and CO2 are simulated and the results are consistent well with the ex...

During a severe accident in nuclear power plant, core damage may occur due to decay heat and molten fuel can pour into and interact with water resulting in steam explosion. The energetics of steam explosion strongly depends on the initial premixing stage during which the molten fuel undergoes a coarse fragmentation process, which determines the sur...

With the advantages of the thermophysical property of supercritical carbon dioxide (SCO2), SCO2 has been proposed for being used as the coolant of the secondary system in a nuclear reactor to promote a higher thermal efficiency. However, heat transfer deterioration (HTD) in supercritical fluid became a potential operational problem for the supercri...

Supercritical water reactor (SCWR) is one of the most promising nuclear reactor system among generation IV reactors thanks to its high thermal efficiency and simplicity. One of the main features of supercritical water is the strong variation of thermal-physical properties in the vicinity of the pseudo-critical temperature, which makes it very hard...

In AP1000 reactor system, two canned motor pumps are directly attached to the cold side of the steam generator. In order to investigate the effect of the velocity distortion generated by the steam generator on the performance of the two pumps, CFD method is used to simulate the model pump with three different suction flows, one comes from the strai...

Background: Canned nuclear coolant pump is welded to the steam generator and main pipe line. The pressure fluctuation generated in the pump may affect both the steam generator and main pipe line. Purpose: The research was conducted for the purpose of studying the characteristics of pressure fluctuation at inlet and outlet of the nuclear coolant pum...