Conference Agenda

Improving thermal-hydraulic performance is a major goal for many applications since fluid flow is essential to many natural and artificial systems. This study focuses on assessing thermal and fluid flow performance in a corrugated mini channel, which has a distinct separation and mixing zone arrangement that influences its thermal-hydraulic behavior. To investigate how various geometric parameters affect this channel's hydraulic performance, experiments and CFD simulations were carried out. Using water as the working fluid and volumetric flow rates ranging from 1 to 7 L/min, increasing in increments of 0 to 1 L/min, an experimental investigation was carried out. The Reynolds numbers for these flow rates ranged from 1000 to 4000. The study also explores the effect of the mixing-to-separation-zone length ratio (Lm/Ls) on hydraulic operations. A crucial metric for evaluating hydraulic performance, the friction factor, and Lm/Ls are clearly correlated in the experimental results. This experimental result had a maximum deviation of 5% from the numerical calculation. Consequently, a power law-based novel correlation with a variance of less than 5% is suggested to forecast the friction factor and heat transfer. This emphasizes how the Reynolds number and geometric parameters both affect the friction factor, a crucial hydraulic performance metric.

The accurate prediction of two-phase flow void fraction and wall superheat under pressurized conditions is crucial for understanding reactor safety margins. Existing void fraction and flow boiling heat transfer models used in numerical simulations exhibit significant uncertainty, limiting their accuracy in two-phase CFD simulations. This paper presents high-resolution vertical upward flow boiling experimental data from the PCHT test facility at the University of Michigan, using a gamma densitometer and X-ray imaging system. Experimental results are compared with established one-dimensional models to validate their applicability and identify limitations. The Saha and Zuber correlation is used to predict the thermodynamic equilibrium quality at the net vapor generation point. The slip ratio model proposed by Chisholm, Thom, Zivi, Cai, Lockhart, and Martinelli was used to estimate the void fraction and heat transfer coefficient in sub-cooled flow boiling conditions.

Hence, based on the existing correlations from previous research and experiment data from the PCHT test facility, a new one-dimensional model is proposed to better predict the void fraction and wall heat transfer coefficient under the vertical, upward, sub-cooled flow boiling conditions.

This study presents an innovative use of boiling acoustics techniques to examine the cooling system of the Isotope Production Facility (IPF) at Los Alamos National Laboratory (LANL). In the IPF target station, multiple stacked targets are arranged with a series of water channels interspersed in between, leveraging forced convection for cooling. During operation, the rastered high-energy proton beam can initiate various boiling regions from subcooled boiling to critical heat flux. Identifying these boiling characteristics is challenging because of the extreme radiation environment. For that, a prototypical facility setup with a transparent window is utilized for the visualization of boiling phenomena. In this paper, we employ a hydrophone and high-speed video camera to capture acoustic signals and images indicative of various boiling phenomena. By applying signal processing techniques such as Fast Fourier Transform (FFT) and Short-Time Fourier Transform (STFT), we aim to discern distinct boiling behaviors from the hydrophone data. The insights gained from this analysis will guide the installation of hydrophones within IPF, allowing real-time monitoring to prevent boiling crisis by adjusting operational parameters such as beam intensity. While this methodology is tailored for IPF, its implications extend to other systems where boiling dynamics are critical, particularly in the nuclear industry and research sectors. This research enhances our understanding of thermal hydraulics and heat transfer in isotope production facilities and contributes to the safety and efficiency of nuclear systems.

This study investigates the nucleation site density (NSD) on nuclear fuel cladding materials under flow boiling conditions in an annular gap geometry. Experiments were conducted at the Karlsruhe Institute of Technology (KIT) COSMOS-L facility using three cladding samples: uncoated Zircaloy-4, and two physical vapor deposition (PVD)-coated variants, CrN and Cr. The test section comprised a 9.5 mm diameter cladding heated over a 330 mm length, with data collection focused on a 25 mm segment near the outlet. Measurements were performed at 300 kPa outlet pressure, approx 500 kg/m^2/s mass flux, and an 85°C inlet temperature, with variable heat flux. High-speed videography captured bubble dynamics, and nucleation sites were identified using an in-house KIT code that tracks brightness changes in individual frames to calculate the frequency and spatial distribution of departing bubbles. To distinguish true nucleation sites from passing bubbles, noise, and non-uniform illumination, an adaptive filter based on proper orthogonal decomposition was implemented. The NSD comparison across the three samples revealed observable variations, which require further evaluation under different flow and heat flux conditions to better understand surface modification effects on boiling behavior.

An experimental setup has been designed and manufactured at the Royal Institute of Technology (KTH) to investigate the internal structure of boiling two-phase water flow under prototypical Light Water Reactor core conditions, including those relevant to PWR, BWR and SMR designs. The setup is based on the High-pressure WAter Test (HWAT) loop, designed for 25 MPa pressure, 1 kg/s water mass flow rate and 1 MW thermal power. The facility has been updated with a new test section and advanced instrumentation systems to enable measurements under steady-state and transient operations. This novel experimental setup allows for the first-time measurements of radial distributions of local two-phase flow parameters under high-pressure LWR core conditions. The resulting data is intended to enhance the fundamental understanding of boiling two-phase flow phenomena, contribute to the development of closure laws and support the validation of computational codes. The paper presents the loop design, the updated instrumentation with associated uncertainties, and data post-processing methods (including the derivation of dispersed phase length scales). Results from commissioning tests, such as heat balance tests and single-phase tests, are presented. Examples of high-pressure boiling two-phase flow measurements are presented and discussed. Fundamental behavior and associated key parameters, including radial distributions of void fraction, mixture velocity, interfacial length scales and polydispersed characteristics, are identified and quantified.