Conference Agenda

In order to enhance the safety of nuclear power plants, it is required to establish emergency cooling methods for reactor accidents. In PWR, the using In Vessel Retention (IVR) method is considered to prevent melt-through in meltdown. In the IVR method, the cavity surrounding the reactor pressure vessel is filled with water in the IVR method to enable cooling by boiling heat transfer. The maximum cooling capacity of boiling heat transfer is determined by the critical heat flux (CHF), and improving CHF is crucial to implement the IVR. In this study, we found that forming a porous copper structure on the boiling surface improved the CHF to approximately 5 MW/m², which is about four times higher than that of an uncoated plain surface. The boiling experiments were conducted using a heat transfer surface with a diameter of 10 mm and porous copper structures with thicknesses from 0.5 mm to 3.4 mm under saturated temperature conditions at atmospheric pressure. It was observed that CHF increased as the thickness of the porous structure increased up to 2 mm, but decreased when the thickness reached 3.4 mm. In this presentation, we will discuss the factors contributing to the CHF improvement.

In the face of the escalating energy crisis, the advancement of nuclear energy assumes paramount significance. Fuel assemblies, crucial elements of reactor cores, play a pivotal role in this domain. During the long-term operation of reactors, deformation of the pressure vessel may occur, leading to displacement between the fuel rods and the pressure vessel, resulting in eccentricity. Studying the subcooled boiling and critical heat flux (CHF) phenomena within annular pipes and exploring the effects of eccentricity are crucial for clarifying the two-phase boiling processes in fuel assemblies and enhancing reactor safety. In the reactor core, usually the heating curves of rod bundles are non-uniform heating. Comparing to the uniform heating methods, the non-uniform heating mothods bring more uncertainness in CHF locations and values. Using CFD analysis software, in this paper the RPI boiling model combined with the Eulerian-Eulerian two-fluid model are employed to analyze the subcooled boiling and CHF heat transfer characteristics of annular pipes under non-unifrom and uniform heating. On this basis, the mechanisms underlying the effects of different eccentricities on annular pipes with non-uniform and uniform heating are explored. The findings of this paper provide valuable insights for a deeper understanding of heat transfer phenomena within fuel assemblies and offer guidance for their practical application.

During the operation of nuclear reactors, corrosion particles deposited on fuel cladding alter its surface characteristics, thereby influencing the critical heat flux (CHF). The modification of surface properties, such as roughness, wettability, and porosity, plays a significant role in determining the heat transfer efficiency and safety margins of the reactor. However, the absence of detailed structural parameters for the deposition layer has hindered a comprehensive theoretical analysis of the mechanism by which the deposition layer affects CHF. To address this gap, the present study conducted systematic pool boiling deposition experiments under varying deposition times and heat fluxes. The experimental results demonstrated that the average CHF of smooth rods was 1240 kW/m², with a deviation of less than 10% from model predictions, thereby confirming the measurement accuracy and stability of the experimental apparatus. This validation establishes a robust and reliable experimental platform for subsequent research on CHF behavior under deposition conditions. Notably, the average CHF of rods with deposition reached 1479 kW/m², representing a 19.2% enhancement compared to smooth rods. This significant improvement suggests that, in terms of CHF enhancement for fuel rods, corrosion particle deposition exerts a positive influence. These findings contribute to a deeper understanding of the complex interactions between surface properties and CHF, offering critical insights into the role of deposition layers in enhancing thermal performance.

Enhancing our understanding of the link between the near-wall void-fraction profile and nucleation characteristics at the boiling surface in subcooled-flow conditions is crucial for improving subcooled-flow and DNB models. This interaction is being investigated at a flow-boiling facility at MIT, which features a deionized water loop capable of operating at pressures up to 10 bars and mass fluxes up to 2000 kg/m²/s. The test section includes a rectangular flow channel (3 cm x 1 cm).

Subcooled nucleate boiling was generated using a custom-made heater, consisting of a sapphire substrate coated with a thin layer of chromium, housed in a heating cartridge mounted on one of the test section walls. An optical probe, translated with sub-millimeter accuracy to and from the heated surface, measured the void-fraction profile near the surface. Infrared measurements through the back of a sapphire substrate allowed for the quantification of wall temperature, heat flux, and relevant nucleation properties at several heat fluxes.

The collected measurements are the first step towards forming a comprehensive picture to elucidate the mechanisms linking the void fraction distribution in the vicinity of the boiling surface with the boiling process and the boiling crisis. Insights gained from this study may inform the development and validation of next-generation models for flow boiling simulations.

To prevent reactor vessel failure, the in-vessel corium retention through the external reactor vessel cooling (IVR-ERVC) has been adoptedas the severe accident mitigation strategy in nuclear reactors. In existing large nuclear reactor types, IVR-ERVC is conducted in a natural flow condition, which forms between the insulator and the reactor cavity. However, in Small Modular Reactors (SMRs) currently under development in the Republic of Korea, there is no insulator due to the integral design, and thus IVR-ERVC is conducted in a pool condition. Additionally, since the reactor lower head outer wall of SMRs is made of stainless steel, it is essential to study the Critical Heat Flux (CHF) phenomena occurring on stainless steel surfaces to ensure the safety and effectiveness of the IVR-ERVC process in these reactors. In this study, focusing on the integral SMR design, the experiment measured CHF under various conditions, including the effects of heater surface inclination and material properties for stainless steel. Experiments using an SS304 heater in pool boiling conditions are conducted to develop a modified CHF correlation, reflecting the specific characteristics of integral SMR, challenging existing models and contributing to safer nuclear power technology.

The critical heat flux (CHF) is a key thermal limit in water-cooled nuclear reactors and accurate and reliable modelling of boiling and CHF remains an unresolved challenge in nuclear thermal hydraulics. In large majority, CHF is still estimated using empirical models derived from expensive, full-scale experiments. Due to the empirical nature of the models, significant engineering margins are applied, restricting reactors to operate at a power level that is below their theoretical potential.

In the last few decades, computational fluid dynamics (CFD) models based on the multifluid Eulerian-Eulerian method and the heat flux partitioning approach have shown promise in reducing conservative design margins through more accurate predictions. However, the large number of modelling closures required (e.g., nucleation site density, bubble departure diameter) and the overfitting of the numerous constants on limited datasets has so far prevented developing a universally accepted, best-possible model and deliver the anticipated improvements.

In the last few years, advancements in measuring techniques have made possible detailed, small-scale measurements that enable the validation of boiling models at a level of detail that was not possible before. In this work, we have developed and implemented in MATLAB a heat flux partitioning framework and, leveraging these new data, assessed the most frequently used and recent heat flux partitioning models in pool and flow boiling conditions. Strengths and weaknesses of each model, and some physical inconsistencies are identified. Impact of uncertainty in closure models is quantified, improvements implemented and validated and areas for future development suggested.