The vibration responses and bubble behaviors are simultaneously measured when subcooled pool boiling occurs on a vertical heated surface. The vibration velocity of the heated surface is collected with a non-contact laser system while the bubble behaviors are captured by a high-speed camera. Experiments with different heat flux and subcooling are carried out, and the vibration responses correlated with the bubble behaviors are analyzed. The results show that with the increase of heat flux, more nucleation sites on the heated surface are activated, and bubble interactions occur. Four typical modes are extracted from the experimental data with isolated bubbles from a single site, continuous bubbles from a single site, bubbles from two sites and bubbles from multiple sites. The main excitation mechanism for all the cases is the pressure fluctuations caused by the rapid growth of the bubble at the initial stage, and other bubble behaviors such as departure, collapse and bubble interactions show insignificant influence. Moreover, the vibration responses in time and frequency domain are strongly related to the temperature distribution before bubble initiation and the position of the nucleation site for isolated boiling bubbles. The generation behaviors of the bubbles also affect the vibration responses especially in frequency domain.

The stratified molten pool, consisting of an upper metal layer and a lower oxide layer, may form in the lower head of the reactor pressure vessels (RPV) during severe accidents of a pressurized water reactor (PWR) with in-vessel melt retention (IVR) strategy. It is therefore of paramount importance to quantify the thermal load distribution of the stratified structure on the RPV lower head. Motivated by such interest, the test facility SAMPO (StrAtified Molten Pool) is developed at CNPE to investigate the convection and heat transfer in stratified pools in a slice of semispherical vessel with the inner diameter in 1100 mm. In the present study two naturally stratified material pairs, i.e. water/thermal oil and nitrate salt (50 %NaNO3-50 %KNO3)/thermal oil, are used as the simulant materials of two types of stratified molten pool denoted by W-O and S-O, respectively. The upper layer has a free surface and the lower layer has an isothermal boundary due to the intensive cooling of water flowing through the external channel. The test results indicate that the profiles of central temperature and heat flux in the lower layers of molten pools are similar to those of a homogenous molten pool with a radiation upper boundary condition. The focusing effect is observed only in the S-O pool of nitrate salt/thermal oil, and the peak heat flux is strongly affected by heating power of the lower layer, thickness of the upper layer and the interlayer crust next the sidewall.

In the context of severe nuclear accidents, the migration of corium into the reactor pressure vessel (RPV) poses significant hazards, prompting the proposal of the in-vessel melt retention (IVR) strategy, particularly the external reactor vessel cooling (ERVC) approach. Evaluating the accuracy of turbulence models within the melt pool is crucial for assessing the feasibility of IVR. However, previous studies have yet to reach a consensus about the most suitable model due to the lack of data comparison. We aim to conduct a comprehensive comparative analysis of turbulence models to evaluate their performance across a range of Rayleigh numbers, particularly under conditions relevant to IVR scenarios. Therefore, this study employs six commonly used turbulence models in computational fluid dynamics (CFD) software, ANSYS Fluent, to simulate three natural convection experiments (Kulacki-Goldstein, BALI, and LIVE-3D). The results demonstrate that the choice of turbulence model significantly impacts the accuracy of temperature and heat flux predictions within the melt pool. Although the relative temperature deviation is less than 0.1% in all the simulations of the Kulacki-Goldstein experiment, the differences among turbulence models become increasingly pronounced with rising Rayleigh numbers. Among the models tested, wall-modeled large eddy simulation (WMLES) proved the most reliable for complex geometries and high Rayleigh numbers, while the realizable k-epsilon and generalized k-omega (GEKO) models also showed consistent performance. However, the Reynolds stress model (RSM)-baseline (BSL) and detached eddy simulation (DES) models exhibited notable limitations, particularly in scenarios involving solidification and melting. These findings provide valuable guidance for selecting appropriate turbulence models in IVR-related natural convection simulations and highlight the need for further refinement to improve model accuracy across varying melt pool conditions.

The MELCOR code as a severe accident simulation tool does not have the capability to capture the quench process of a debris bed which may form in the wet cavity during a severe accident of light water reactors (LWRs). Although the coupled MELCOR/COCOMO simulation could overcome the limitation (Chen et al., 2022), the calculation time was explosively escalated due to mechanistic modeling of debris bed thermal-hydraulics in COCOMO. To suppress the computational cost, a surrogate model (SM) was developed in our previous study (Wang et al., 2023), and its coupling with MELCOR could realize a quick estimation of the quench process of one-dimensional debris beds. The present study is an extension of the previous work, aiming at the development of a new surrogate model for the quench process of two-dimensional conical debris beds. The new surrogate model (SM) was based on artificial neural networks (ANNs) and trained by the database from COCOMO calculations of various conical debris beds quenched in the reactor cavity of a Nordic boiling water reactor (BWR). The MELCOR was then coupled with the new SM to simulate a postulated station blackout (SBO) scenario in the BWR. The results show that the coupled MELCOR/SM simulation could provide similar ex-vessel debris bed quench period and containment pressure/temperature trends as the coupled MELCOR/COCOMO. Compared with the MELCOR standalone calculation, the coupled calculations predicted earlier points of time for water pool saturation and containment venting, since the heat transfer from conical debris bed to water pool is faster in the coupled simulations.

During postulated severe accidents of a light water reactor, a debris bed may form in the lower head of the reactor pressure vessel due to Fuel-Coolant Interaction (FCI), and re-melt into a molten pool if the debris bed is uncoolable. The debris bed remelting is therefore an important process in a severe accident scenario. To predict the dynamic process of debris bed remelting, a computer program is developed in the present study using lumped-parameter models. The melt in the lower head is split into different zones of molten metal, molten oxide and solid debris particles submerged in molten pools. Correlations are employed to calculate the heat transfer within each zone and between zones. The developed lumped-parameter code is employed to calculate the COREM experiments. The comparison of the simulation results with the experimental shows a reasonable agreement for melting processes of single-material and two-material debris beds. The code is also used to investigate some factors which may affect debris bed remelting, such as internal heating power, volume ratio of components, and thermophysical properties.

Following Fukushima Daiichi, nuclear safety is paramount in advanced pressurized water reactors. In-Vessel Retention (IVR), notably External Reactor Vessel Cooling (ERVC), offers simplicity and cost-effectiveness. However, uncertainties in the corium thermal load and IVR processes mandate conservative design and safety margins. This study simulates the SAMPO experiment with nitrate salts and thermal oil to investigate the thermal hydraulics of a stratified melt pool. Analysis of power levels, heat transfer coefficients, and radiation heat transfer reveals key insights. Increasing input power raises the temperature and sidewall heat flux in the upper layer, leading to interlayer crust dissolution and enhanced upward heat transfer near the lower layer's melting point. Higher convection coefficients double the heat flux at the metal layer's top while reducing sidewall heat flux. Adjusting radiation emissivity of the front and back plates to 0.7 synchronously decreases heat flux from both the top and curved sidewalls, achieving an effect similar to a 50% power reduction without radiation.

To enhance the thermal safety limits, the present study experimentally investigates the performance of spray cooling on a downward-facing surface. The experiments were conducted on the KTH-SPAYCOR facility. It employs four pressure-swirl full-cone nozzles arranged in a 2 x 2 array to produce a multi-spray pattern on a thin SA302B foil with a relatively large surface area of 120 mm x 80 mm. The tests were carried out under steadystate conditions, achieved through stepwise increments in surface heat flux, until the onset of dry spots or ultimate burnout was detected. Seven tests were performed to examine the effects of nozzle-to-surface distance and the flowrate, while maintaining a fixed surface inclination angle of 30 degrees. The cooling mechanism combines direct droplet impingement and flushing by water film. The experimental results demonstrated the potential applicability of the multi-nozzle array system for cooling large surface areas, such as the reactor pressure vessel lower head. Specifically, the cooling limit was observed to enhance with increasing NTSD and flowrate, reaching a peak value of 2.33 MW/m2 at a flow rate of 13 lpm. However, a slight decline in the cooling limit was noted when the flowrate exceeded this optimal flowrate. Additionally, increasing NTSD and flowrate enhanced the uniformity of the heater temperature distribution. To demonstrate the system's performance under high heat flux conditions, the ultimate critical heat flux was measured at flow rates of 10 lpm and 13 lpm, yielding values at 1.97 MW/m2 and 2.50 MW/m2, respectively.

The thermophysical properties of corium are critical for improving the predictive accuracy of severe accident analysis codes. However, due to the high melting temperature and high volatility of corium, thermophysical property measurements are extremely challenging, resulting in a significant lack of data. This study presents a non-contact measurement facility based on the aerodynamic levitation technique, enabling the measurement of the density, surface tension, and viscosity of corium components at temperatures exceeding 3000 K. Density is measured based on the axisymmetric ellipsoid assumption of levitated drops, while the surface tension and viscosity are determined using the drop oscillation method. Experimental results for key corium components, including ZrO2 and a UO2-ZrO2 mixture, are presented, addressing data gaps in the thermophysical properties of UO2-containing materials.

To better understand the thermal behavior of a two-layer melt pool with a high Rayleigh number-a pattern observed in the RASPLAV study, which indicates a significant risk to pressure vessel integrity and the success of in-vessel retention (IVR) strategies-this paper reports experimental findings from a 2D, full-scale (1:1 ratio) prototypical stratified melt pool (SAMPO-P). A series of experimental tests were carried out with varying heating powers and top layer heights, achieving a Rayleigh number of 3.77 x 10(15), comparable to that found in light water reactors (LWR). Water was used to simulate the bottom layer, while n-octanol represented the top layer. Internal decay heat was modeled in the bottom layer using electric heating rods. After analyzing the main heat transfer parameters from the experiment, this paper derived several useful heat transfer correlations. The normalized temperature and heat flux distributions remained consistent across different power levels, and the normalized heat flux in the bottom layer aligned well with existing experimental correlations. In the bottom layer, the downward heat transfer coefficient was lower compared to other single-layer correlations, likely due to increased upward heat transfer.

During severe core meltdown accidents of a light water reactor (LWR), the core melt (molten corium) may fall into a water pool, resulting in molten fuel coolant interactions (FCI). Quantitative understanding of FCI phenomena is paramount to corium risk assessment of LWRs such as Nordic boiling water reactors which employ reactor cavity flooding as severe accident management strategy (SAMS). Melt jet breakup and droplet fragmentation play an important role in FCI, affecting debris coolability and steam explosion energetics which are considered in ex-vessel corium risk assessment. The present study is concerned with numerical simulation of melt jet breakup in a water pool using a multiphase computational fluid dynamics (MCFD) approach where a coupled Level Set and Volume of Fluid (CLSVOF) method is used to capture melt-coolant interfaces. The focus is placed on the prediction of interface instabilities and jet breakup length, and their influential factors (melt materials, jet diameter, fall height, in-pool structures, multiple jets and pitch/diameter ratio). The simulation results are compared with the data of the DEFOR-M tests carried out at KTH. There is a good agreement between simulation and experiment, in terms of jet deformation pattern and jet breakup length. It is also found that the jet breakup length is different from the values predicted by well-known correlations (e.g., Taylor's, Epstein Fauske's and Matsuo's). Based on the experimental and numerical data, a new correlation for the jet breakup length is developed in the similar formula of the Satio's correlation.