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This paper uses the finite element analysis software COMSOL to build the models of short string and long string composite insulators and the typical defects of composite insulators are simulated. For the short string insulator, this paper designs the air gap of different positions with different dimensions, simulates the electric field of air gap. This paper also analyses the whole and local moisture of insulator mandrel, designs the different moisture degrees for the insulator mandrel. Meanwhile, the axial electric field at the mandrel - sheath interface is simulated by combining the moisture defects with the air gap defects.

As a promising mitigation strategy, In-vessel-retention (IVR) has been an important research focus in realms of nuclear safety analysis. For IVR analysis, it’s of importance to study the natural convection behaviors of the corium pool formed in the lower plenum of the reactor vessel. Because of the different densities between corium compositions, typically the corium pool is believed to evolve to two-layer configurations, with a heavier oxide layer at bottom and a metal layer on top. To know better the convection behaviors of the layered corium pools, research was performed both experimentally and numerically. The test section for the experiment is a two-dimensional semi-ellipsoidal slice, which measures a span of 1.2m, a total pool depth of 0.75m and a thickness of 0.2m. Two immiscible fluid, i.e. fluorinert fluid FC-40 and water, were chosen to simulate the oxide layer and the metal layer respectively. Based on the experimental settings, numerical simulations were performed, and the results were compared with the experimental data. Further, discussions were made on the focusing effect, the relationship between the temperature and the velocity field, and also the coupling modes in the two-layer system. These results offer more details to the knowledge of layered corium pools and can be helpful for IVR analysis.

The Large Eddy Simulation (LES) methods were employed to analyze the turbulent heat transfer characteristics in detail for volumetrically heated corium pools. The main components and property parameters were calculated for the one-layer and two-layer configurations in the HPR1000 reactor. The numerical simulations were performed based on the WMLES turbulence model, phase change model, as well as the VOF model to describe thermal behaviors in the corium pool such as natural convection flow, pool temperature field and heat flux distribution. Then based on the Kelvin-Helmholtz (KH) instability theory, the calculated velocity vector difference between the two-layer flows at the interface was lower than the KH instability critical value. The high Rayleigh number natural convection in the bottom oxide pool was still not able to entrain the top metal layer. The molten salt (20 mol% NaNO3-80 mol% KNO3) and the Lead-Bismuth eutectic (44.5 wt% Pb-55.5 wt% Bi) were chosen as the appropriate simulant materials for the two-layer corium pool. The simulation results presented similar thermal characteristics compared with those from the prototypical reactor two-layer case, which could provide guidance for further research.

In-vessel retention (IVR) has been proposed as an effective mitigation strategy during severe accidents in pressurized water reactors (PWRs). To help understand better the application of IVR, the large-scale COPRA facility was used for experimental research and the eutectic salt, i.e. the binary mixture of 50 mol% NaNO3 – 50 mol% KNO3 was selected as the simulant material for corium. Besides, numerical studies were done to capture the detailed flow field information. The experiment studied the effects of a sudden change in boundary condition and the influences of power changes on the transient behaviors and the steady-state characteristics of melt pool. Experimental results show that a sudden introduction of sidewall cooling and the melt pouring are the most threatening factors to the reactor vessel safety, but which can be guaranteed as long as the vessel is kept undamaged during melt pouring phases. Also, power changing won’t lead to rapid variation of sideward heat flux, so the correlations for steady state heat transfer calculation are basically qualified to evaluate the reactor safety in analysis of long time cooling. Besides, the power changing only exerts obvious impacts on the crust growth on the curved sidewall inner surface with polar angle below 30°, contrary to which the sideward heat flux is clearly influenced by different power densities only above the angle of 30°. Finally, the numerical simulation results show that vortexes could form near the region where the crust formed, and lead to a more homogeneous temperature distribution; generally, the convection happens between the lower part of melt pool and cooling boundary, with flow velocity at a magnitude of 0.01 m/s.

Single-phase liquid sodium as a kind of ideal coolant is widely applied in the fields of nuclear reactor engineering, aerospace, solar energy and industrial waste heat utilization, due to the good heat transfer and flow characteristics. Therefore, numerous studies have been carried out experimentally and theoretically to investigate the heat and mass transfer performance by using liquid sodium in past decades, and most of them are concentrated on the experimental results, mechanisms and models of heat transfer and flow of liquid sodium. However, no comprehensive review article concerns on the thermal hydraulic characteristics of single-phase for liquid sodium. Thus, in this study, an analytical overview on thermal hydraulic characteristics of single-phase for liquid sodium is presented, using widely scattered available information from existing literature. It can be taken into account as a quick reference guide to have an overview of the thermal hydraulic characteristics of single-phase for liquid sodium in different channel geometries (circular tube, annular tube and rod bundles etc.) under different conditions.

With the development of sodium-cooled reactor (SFR), the research on the flow and heat transfer characteristics of liquid sodium becomes increasingly important and necessary. Through decades of experimental and theoretical research under different conditions, certain achievements have been obtained. To further carry out relevant studies, a review work on this topic is of necessity. As pointed out by scholars, sodium boiling might occur during hypothetical accidents, which might cause great damage to the SFRs. For safety analysis, the boiling of liquid sodium has already become a focus in this research area. Based on an extensive review on the published literature, this paper concerns two major topics, i.e. the boiling heat transfer and two-phase flow characteristics of liquid sodium. The former topic focuses on the characteristics of sodium boiling in pools, round tubes and annuli, including the correlations for the boiling heat transfer coefficient, the incipient boiling wall superheat and critical heat flux of dryout type. Also, the correlations for the single-phase flow in bundle channels are presented here due to the fact that relevant research on the sodium boiling in rod bundles is relatively insufficient and the relevant correlations are rarely found. For the part of two-phase flow characteristics, the correlations for the two-phase pressure drop multiplier are presented and analyzed. Finally, conclusions and prospects in this research field are summarized and proposed.

Due to the high-Rayleigh-number natural convection and strong turbulence in the volumetrically heated corium melt pools, it is difficult to capture the heat transfer characteristics in detail. The Large Eddy Simulation (LES) methods are more and more applicable to nuclear industry and have become powerful tools to analyze the complicated turbulent and multi-phase flows. In this paper, the Wall-Modeled LES (WMLES) method was employed for the simulation of three typical corium pool heat transfer experiments: BALI, LIVE and COPRA experiments. The melt pool temperature and heat flux from numerical simulations were in good comparison with experimental data. Then numerical simulation for the transient formation of two-layer corium pool was performed with prototypical corium to obtain the heat transfer data in reactor situation. There was a distinctive gap between the temperatures in two layers. The heat flux showed a decrease at the two-layer interface and a huge jump at the metal layer side. The thin metal layer on the top at the early stage of the two-layer formation will threaten the vessel integrity.

In realms of nuclear safety analysis, In-vessel retention (IVR) has been a research focus as it is a promising mitigation strategy for the postulated severe accidents in pressurized water reactors (PWRs). However, the recent development of ocean floating reactors for civil use has greatly motivated the pertinent studies under motion conditions, which can be much different from the known investigations conducted under static condition without the additional acceleration forces induced by motions. Thus, through experimental research, this paper extensively studied the influence of rolling motions on the convection behaviors inside two-layer corium pools, which are the major concerns for IVR research. Transient behaviors of the layered corium pools were obtained, and sensitivity analyses were conducted on three critical parameters, i.e. maximum rolling angle, rolling period and the thickness of metal layer. Experimental results show that the two-layer fluids can stay basically stratified under rolling conditions but mixing phenomena can be observed in the interface nearby region when rolling motions become intense enough. Generally, rolling motions can weaken the thermal stratification inside the layered corium pool, and enhance heat removal capacity from the corium pool to the outside cooling boundary. These phenomena indicate a good chance for the efficient decay heat removal from the corium pool under rolling motions. This study is supposed to provide valuable references to assess the success probability of IVR design applied in reactors operated under ocean conditions.

In view of the safety characteristics of the offshore floating nuclear plant (OFNP), an ocean condition available analysis code based on RELAP5 MOD3.3 is developed and the influences of single heaving, single rolling and combined motion of heaving and rolling on a OFNP were analyzed under station blackout accident (SBO). The results indicate that heaving motion accelerates the drop in the reactor pressure. The flow fluctuation caused by heaving motion promoted the derivation of residual heat after SBO; Rolling motion accelerates the declination of the reactor pressures and the coolant temperature after SBO, and also causes a large flow oscillation in pipes. The results also shows that the combined motion is more severe than these two single conditions. The parameters variation under the combined motion is more similar to that under its rolling condition. The residual heat can be removed effectively under all these conditions and the security will not be threatened.

In-vessel-retention (IVR) has become an important subject of severe accident mitigation strategy. Up to now, many experimental and numerical investigations have been performed on the natural convection characteristics in melt pools with volumetric heating. But these studies are limited to the melt pools under static condition. As floating nuclear reactors become increasingly popular among both commercial and military ships, for successful application of IVR in this occasion, research should be done on the heat transfer characteristics of melt pool under moving conditions. Currently, the specially-designed facility is under construction in Xi'an China for the relevant experiment and numerical studies are performed beforehand. In this paper, a hemisphere with an inner radius of 0.5m, similar to LIVE experimental facility, is chosen to simulate the melt pool. Its flow behaviors under periodic rolling condition are simulated by means of CFD calculation. This work may cast a light on the melt pool characteristics under moving conditions and could be further evaluated by future experimental data.

For some Pressurized Water Reactors (PWR) operated on automobiles, boats or deep sea vessels, system characteristics is important for understanding their safety during severe accidents. The development of an analysis code and the transient thermal beaviors of a floating nuclear reactor under heaving motion are described in this paper. By modifying the control equations based on the mathematical models of ocean conditions, an ocean condition available system analysis code named RELAP5/GR was developed from RELAP5 MOD3.2 to simulate the transient thermal-hydraulic response of the nuclear reactor systems to the motion conditions in accidents, which is an advanced and independent node programming code. Using the code, the analysis model was established for a small 200MW offshore floating nuclear plants (OFNP). The transient thermal behaviors of the whole system were analyzed in the cases of the station blackout accident under heaving motion conditons. The analysis shows that all the results can be reasonably explained and the code development is successful at this stage.

Convection inside a layered system has been a significant research focus. Different from the known research under static condition, this study experimentally investigated the convection phenomena of internal-heated two-layer fluids systems under swinging motion conditions. The effects of swinging motions on the major convection phenomena related parameters were studied, including the system temperature fields and heat transfer capacities, together with the sensitivity analyses of maximum swinging angle and swinging period. Additionally, phenomenological research was carried out on the pattern of interface deformation between the two layers under swinging motions. From the experiment, it’s found that swinging motions can weaken the thermal stratification occurring in the internal-heated two-layer system under static condition, and intensify heat transfer from the system to the outer cooling boundary. A wave shape of interface is observed with the declining swinging period, and unstable interface deformations can be seen during swinging motions with a large maximum swinging angle but a small swinging period, with mixing or penetration observed between the two liquids. Moreover, a thicker layer of upper fluid refrains interface deformation and thus helps to stabilize two-layer system during swinging motions. This study can be helpful for the nuclear safety management strategy related to the analysis of stratified corium pools in the marine environment.