Conference Agenda

This study explores critical heat flux (CHF) in flow boiling on a downward-facing surface with non-uniform heat conductance properties. While prior investigations have examined non-uniform heat conductance surface, this research uniquely focuses on the downward-facing configuration, addressing an area that has been less explored. Additionally, the study expands the scope of surface modifications compared to earlier work, enabling a broader analysis of CHF behavior and non-unform heat conductance surface.

A significant advancement in this research is the introduction of a new CHF enhancement concept designed to prevent CHF initiation due to local hot spots. This concept is experimentally validated by measuring temperatures at both upstream and downstream regions of the heated surface, ensuring that local temperature variations and CHF dynamics are thoroughly understood. The methodology incorporates low thermal conductivity tape for surface modifications, chosen for its versatility and ease in creating various thermal conductance conditions.

Experimental results reveal a substantial CHF enhancement of up to 20%, highlighting the effectiveness of the proposed approach. These findings provide valuable insights into boiling heat transfer improvement strategies, particularly for downward-facing applications, and demonstrate the practicality of mitigating CHF triggers through innovative surface designs and precise thermal management.

Furthermore, the concept designed to prevent CHF initiation due to local hot spots can be applied to enhance the in-vessel retention capacity of a pressurized vessel. Since heat flux is non-uniform across different regions of the vessel, this approach could improve the overall CHF behavior, thus enhancing the thermal management and safety of pressurized vessels.

In this study, the effect of oxidation on the critical heat flux (CHF) of a vertical copper surface with four 16mm x 3mm x 2mm horizontal grooves and an bare copper surface will be investigated. During the experimental process, the copper surface will continuously oxidize with the repetition of experiments. For the bare copper surface, the oxidation rate is faster, as evidenced by its rapid darkening and loss of metallic luster, and the corresponding CHF value gradually increases with the surface oxidation. For the grooved copper surface, the oxidation rate of the surface is similar to that of the bare copper surface, but the oxidation rate within the grooves is much slower than that of the bare surface. The CHF value will remain stable within a certain range for a period of time before rapidly increasing to another level and continuing to remain stable. Overall, compared to the bare surface, the grooved copper surface has an enhancing effect on CHF, which is due to the change in macrostructure causing changes in bubble behavior, while oxidation also enhances the CHF of the copper surface, which is due to the change in the heating surface, thereby indirectly changing the bubble behavior.

Critical heat flux (CHF) prediction is essential for high heat systems like nuclear reactor and two-phase flow systems for enhancing dependability, safety and efficiency. CHF imposes design and operational restrictions due to safety concerns. However, there is work to be done to develop a reliable and effective CHF model. Machine learning techniques can find patterns and correlations in big datasets but lacks in explaining physical laws underlying CHF. Traditional ML operates as black-box which may result in physically unrealistic predictions, when applied to unforeseen circumstances and show instability. To address this, Physics-informed machine learning (PIML) integrates physical principles into PIML framework. While conventional ML uses only data, PIML integrates data-driven learning with knowledge from physical model. The goal of this study is to see the impacts of different physical models on reducing black-box nature of PIML and improving its interpretability for CHF prediction. In this study, four physical models for calculating CHF (Zuber, Katto-Ohno, Biasi and 2006 lookup table) were used as physical part of PIML and coupled with different pure MLs. A big amount of experimental data was used for the training and validation purpose of pure MLs and PIMLs. A thorough investigation has been carried out to assess (1) the predictive power of PIMLs and (2) compare the physical behaviors of mass flux, pressure, diameter, ratio of length-to-diameter and inlet subcooling on CHF from the PIMLs and pure MLs. This work shows the necessity of selecting a suitable physical model for approaching a robust and dependable PIML model.

Curved tubes such as U-bend, helical coil tubes etc., are commonly seen in the design of steam generators, heat exchangers in power plants due to their compactness and better heat transfer characteristics. Prediction of Critical heat flux (CHF) in the curved tubes is necessary in the system design for improved performance and safe operation. The occurrence of CHF is dictated by the geometric and operating conditions such as channel diameter, mass flux, subcooling, operating pressure etc. In the present study, a dimensionless number (De) determines the heat transfer and fluid flow characteristics in the curved geometries. The effect of De on the occurrence of the CHF in the curved tubes is given less attention in the literature. To this end, the simulations are performed using the two fluid model framework coupled with a wall heat flux partition (WHFP) model. The range of mass fluxes varies from 1500-3500 kgm^-2s^-1 and the degree of subcooling varies from 10-30 K . It was observed that the secondary flows created in the bent tube due to the centrifugal acceleration causes the fluid to undergo greater turbulence and flow separation in the bent region as De increases that enhances the heat transfer characteristics. A higher De causes the fluid velocity to increase which in turn causes the wall temperatures to drop. This further lowers the vapour volume fraction thus delaying the occurrence of CHF ensuring the safe operation.

Critical heat flux (CHF) is a primary power limiting criterion for water-cooled nuclear reactors, which must operate below the CHF conditions to allow sufficient thermal margin during normal operation. Accurate prediction of CHF is important for reactor design and safety analysis to determine safety margins under normal operating conditions, evaluate the maximum sheath temperatures of fuel bundles under anticipated operational occurrences, and predict the consequences under design basis accidents. To date, the most accurate CHF prediction method covering the widest range of flow conditions is the CHF look-up table. The latest version of the CHF look-up table was published in 2006. It was developed based on the world's largest CHF database as of 2005. Since then, a large number of CHF experimental studies have become available, allowing for the existing CHF prediction methods to be further improved. An upgraded CHF look-up table was derived from the expanded CHF databank containing 172 data sets with 42667 data points. The upgraded CHF look-up table provides good predictions for the range of flow conditions covering the normal operation, anticipated operational occurrences, and anticipated accident scenarios of water-cooled nuclear reactors. The upgraded CHF look-up table is expected to aid in improving subchannel and system thermalhydraulics analyses of reactor safety margins and consequences of postulated loss of coolant accidents for water-cooled reactors. The upgraded CHF look-up table also improves the accuracy of predictions under conditions related to fusion reactor divertor applications, and conditions corresponding to supercritical water reactor applications.

In the period of an intensifying energy crisis, the development of nuclear energy is of great importance. Fuel assemblies are critical components of reactor cores. As an innovative fuel type, Helical Cruciform Fuel (HCF) offers a larger heat transfer area per unit volume compared to traditional round fuel rods. Its unique helical structure enhances fluid flow and heat transfer capabilities. Additionally, the periodic contact formed by the helical structure provides self-supporting functionality, eliminating the need for position supporting and simplifying the structure of reactor core. These advantages make this novel fuel rod a focal area of research in Small Modular Reactor (SMR). In the course of extended operation of reactors, the pressure vessel may undergo deformation, causing displacement between the fuel rods and the pressure vessel and resulting in eccentricity. The heating curves of nuclear rod bundles typically exhibit non-uniform heating patterns in the reactor core. Unlike uniform heating methods, non-uniform heating introduces greater uncertainty in the location and values of critical heat flux (CHF). In this paper the RPI boiling model combined with the Eulerian-Eulerian two-fluid model are used to investigate the subcooled boiling and critical heat flux (CHF) heat transfer characteristics of HCF under non-uniform and uniform heating. Besides the influence mechanisms of different eccentricities on HCF with non-uniform and uniform heating power curves are explored. The findings of this paper will provide valuable insights for further research and practical applications of HCF.