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In this paper we present a novel experimental technique aimed at the simultaneous measurement of velocity and scalar fluctuations with high spatial and time resolution. Measurements have been performed using wire-mesh sensors in combination with Stereo Particle Image Velocimetry (SPIV) to achieve synchronized measurements of both velocity and scalar fluctuations, and therefore of the turbulent fluxes u j ′ φ ′ ‾ . The proof-of-concept of this measurement technique was obtained on a jet flow experimental facility HiRJet (High Resolution Jet) developed and built in the Experimental and Computational Multiphase Flow (ECMF) laboratory at the University of Michigan. Acquired experimental data indicates a high temporal resolution of proposed technique and capabilities of achieving a high level of synchronization which is confirmed by cross-correlation between wire-mesh and SPIV signals. [ABSTRACT FROM AUTHOR]

In the present paper we report about the first attempt to demonstrate and assess the ability of state-of-the-art high-fidelity computational tools to reproduce the complex patterns of CRUD deposits found on the surface of operating Pressurized Water Reactors (PWRs) fuel rods. A fuel assembly of the Seabrook Unit 1 PWR was selected as the test problem. During Seabrook Cycle 5, CRUD induced power shift (CIPS) and CRUD induced localized corrosion (CILC) failures were observed. Measurements of the clad oxide thickness on both failed and non-failed rods are available, together with visual observations and the results from CRUD scrapes of peripheral rods. Blind simulations were performed using the Computational Fluid Dynamics (CFD) code STAR-CCM+ coupled to an advanced chemistry code, MAMBA, developed at Los Alamos National Laboratory. The blind simulations were then compared to plant data, which were released after completion of the simulations. [ABSTRACT FROM AUTHOR]

Abstract: A computational fluid dynamic (CFD) model for the pressure vessel of the evolutionary pressurized reactor (EPR™) was developed and validated. The aim of this model is the simulation of transients where three-dimensional effects play a strong role, such as boron dilution and main steam line break (MSLB) scenarios. First, a full solid (CAD) model has been built, that includes all details of the reactor pressure vessel (RPV) and the internals which are important for fluid dynamic analyses. The solid model has then been used as basis for the generation of the computational mesh necessary to carry out CFD simulations. Both a hexahedral and a polyhedral mesh have been created. The CFD model has been validated against experimental results of the JULIETTE facility, a 1:5 scaled mock-up of the EPR™ reactor RPV built by AREVA and equipped with advanced instrumentation. The performances of the hexahedral and the polyhedral meshes are investigated in relation to the agreement with experimental data, convergence and CPU requirements. In addition, the effect of the cold-leg swirls on the velocity field inside the RPV is investigated. These swirls mimic the effects of the main coolant recirculation pumps on the flow field at the entrance of the RPV. It is shown that the CFD model is able to capture the shift of the maximum velocity in the downcomer annulus observed in the experimental results. Good qualitative as well as quantitative agreement with the experimental data is achieved. [Copyright &y& Elsevier]

• The purpose of this work is to create a novel experimental database to investigate dead leg flows in pressurized water reactors (PWR). • The interaction between the cold, stagnant branch line water and the high-flow hot main line generates a secondary flow known as the "dead leg vortex." • The dead leg vortex can interact with a stratified layer in the branch, leading to temperature variations on the solid surface. • Such a flow behavior is not well understood, and current screening methods do not adequately account for thermohydraulic forcing. • The aim is to develop more accurate tools by validating advanced prediction methods, with the methodology involving experiments to reproduce the phenomena and validate computational fluid dynamics (CFD) simulations. • A new pressurized experimental setup of a T-junction has been designed and constructed at Vattenfall R&D labs, with temperature, flow rates, and pressures controlled and stable. • Measurements include temperature fluctuations inside the solid pipe wall material in the isolated branch line and temperature measurements on the outer pipe surface using forward-looking infrared (FLIR) cameras. The current screening criteria for thermal fatigue used in isolated branch lines at nuclear power plants are based on simplified approaches, which, on one hand, tend to be overly conservative, and on the other hand, appear to miss important physics. This is demonstrated by the fact that a number of affected pipes were missed by the current screening methods. The purpose of the present work is to produce novel and accurate experimental data, which can be used for a benchmark of Computational Fluid Dynamics (CFD) in order to advance the prediction capabilities of thermal mixing and fatigue in dead leg T-junctions and improve the accuracy of the screening criteria. The ultimate value creation is increased safety and availability at nuclear power plants. This is of great interest and utmost importance to the nuclear reactor safety community. A new pressurized experimental setup of a T-junction has been designed and constructed for the purpose. The experimental setup, with a temperature difference of 120 °C, more closely resembles plant conditions than in earlier experiments. The boundary conditions in terms of the flowrates and temperatures are controlled and stable within their measurement uncertainty – a prerequisite for CFD-grade experiments. The novelty of the present work is measurements of the temperature fluctuations both in the near wall fluid mixing region and throughout the solid pipe wall material. It has been shown that the interaction between the swirl penetration of hot water and the stratified layer in the cold dead leg has an intermittent character and that random clusters of bursts of hot water penetrate into the stratification and the stagnant cold water below. The average time scale of this process is of the order of hundreds of seconds. The stratification between the hot and the cold water in the dead leg has also shown to have a low frequency oscillation of the order of 0.1 Hz, which wiggles in the circumferential direction. After hot water has penetrated downward and excited the instability, the temperature falls back to the initial one in a manner similar to a damped oscillation. The vertical temperature profiles have shown to be self-similar, for moderate temperatures, when scaled with the temperature difference between the hot and the cold water and with the penetration length. The penetration length increases with the velocity in the hot leg confirming earlier reported correlations. A small (less than 0.1 % of the hot main flow) cold leakage flow into the T-junction is sufficient to suppress the swirl penetration and substantially reduces the penetration length. However, the temperature fluctuations have a larger amplitude and range across the whole mixing zone. The fluid temperature fluctuations are shifted towards a more high frequency content, however, the low frequency nature of the swirl penetration is still present. The obtained novel dataset will be provided for a blind CFD benchmark exercise as part of an OECD/NEA project in 2023 and 2024. [ABSTRACT FROM AUTHOR]

• Adiabatic air–water two-phase flows in a helically coiled tube were investigated. • Void fraction measurements were performed by means of X-ray radiography. • Flow regimes were identified and characterized. • Bubbly, plug, slug, slug-annular and annular flow patterns were encountered. • Identified flow regimes are in line with published transition criteria. Helical coil heat exchangers are known for their superior heat transfer, compactness and robustness against thermal expansion compared to systems with straight pipes. Due to all these benefits, they became quite popular in the last few decades in various industries such as pharmaceutical, chemical or nuclear. Generally, they are used in a single-phase environment or with evaporation taking place on the shell side. However, in most of the steam generator designs adopted in advanced nuclear reactors, the two-phase mixture flows upward within the vertically oriented helical coils. In addition, physical two-phase phenomena taking place inside helical tubes differ significantly from those in straight pipes due to inherent centrifugal and torsion effects. To improve the computational models used to predict the thermal hydraulic performance of such steam generators, experimental characterization and identification of different two-phase flow patterns in helical coils are essential. In the present paper, high-resolution measurements of the void fraction distribution for an adiabatic upward air–water two-phase flow were conducted in a transparent helically coiled acrylic pipe by means of high-speed X-ray radiography at ambient conditions. Various void fraction based features such as probability density function estimates, profiles over the height of the pipe cross-section and temporal evolutions were used to identify and characterize flow regimes. A total of 13 measurements are presented in this paper, classified into six flow regimes, namely: bubbly, plug, wavy, slug, slug-annular and annular flow. The obtained flow regimes are in accordance with flow regime characterizations and transition criteria already reported in literature. [ABSTRACT FROM AUTHOR]