Nejdet Erkan

Three-component velocity measurement in microscale flows using time-resolved PIV

The measurement of a three-component (3C) velocity field in microfluidic devices with conventional techniques and conventional micro-PIV (particle image velocimetry) is still difficult due to limited optical access. Since Santiago et al (1998 Exp. Fluids 25 316–9), micro-PIV flow velocity measurements have remained mainly limited to 2C velocity vector field realizations. In this study, the third component of the velocity, i.e. out-of-plane velocity extraction from two-dimensional time-resolved (TR) micro-PIV images, is proposed. The method is based on PIV and performs cross-correlation (CC) peak height tracking inside the small ensembles of the TR-PIV flow images. This concept was verified basically by an experiment performed on a microscale fluid flow inside a 100 µm diameter inclined micro tube. Despite the inevitable background noise which affects the measurement negatively, the extracted steady-state depthwise velocity profile was in agreement with the analytical result.

Measurement of Two Overlapped Velocity Vector Fields in Microfluidic Devices Using Time-Resolved PIV

*1 Department of Quantum Engineering and Systems Science, The University of Tokyo, 7-3-1 Hongo, Bunkyo-ku, Tokyo 113-8656, Japan. E-mail: erkan@vis.k.u-tokyo.ac.jp *2 Department of Human Engineered and Environmental Studies, The University of Tokyo, 5-1-5 Kashiwa, Chiba 277-8563, Japan. *3 Institute of Industrial Science, Underwater Technology Research Center, The University of Tokyo, 4-6-1-FW601 Komaba, Meguro-ku, Tokyo 153-8505, Japan.

Two-phase flow degradation on Fukushima-Daiichi Unit 2 RCIC turbine performance

After the Fukushima accident, several investigation reports, including experiments and simulations have been done for each of the affected units to completely understand the accident progression and use their results to improve the knowledge of severe accident management and the severe codes performance. In Unit 2, the major uncertainties are related with the reactor core isolation cooling (RCIC) system performance during the accident progression especially focused in the RCIC turbine, which is assumed to work in two-phase flow. The main objective of this study is to analyze the RCIC turbine performance under two-phase flow scenarios under the assumption that the power produced by the turbine is lower than expected due to the liquid phase in the flow. A degradation coefficient quantifying the turbine power reduction is developed as a function of the flow quality by using the sonic speed reduction at critical flow conditions principle obtained by applying the non-homogeneous equilibrium model (NHEM). The degradation coefficient was applied to RELAP/ScdapSIM severe accident code showing a drastic reduction of the turbine-generated power during two-phase flow and obtaining a RCIC system behavior closer to the Tokyo electric power company (TEPCO) investigation report conclusions.

Reactor core isolation cooling system analysis of the Fukushima Daiichi Unit 2 accident with RELAP/ScdapSIM

ABSTRACT The reactor core isolation cooling (RCIC) system is an auxiliary system of a boiling water reactor (BWR) that provides makeup water in the case of a severe accident. During the Fukushima accident, the extended operation of the RCIC had a large influence on the accident progression and delayed the core meltdown by almost 70 h. During the Fukushima accident, the water level in the reactor pressure vessel (RPV) was assumed to rise enough to flood the main steam line (MSL), which caused the water to move through the RCIC steam turbine and reduce the overall system water injection capability. A RELAP/ScdapSIM analysis was carried out by using RCIC nodalization to reproduce the Fukushima accident and evaluate the impact of the RCIC system on the accident progression. A coefficient based on the critical flow model was included in the RELAP/ScdapSIM source code to reproduce the degradation suffered by the turbine due to the presence of water. Although highly simplified, the analysis demonstrated the RCIC systems feedback capability, which allows the RCIC to control the plant conditions for a long period of time without any human interaction.

Numerical simulation of droplet deposition onto a liquid film by VOF---MPS hybrid method

Droplet entrainment and deposition are a couple of significant mechanisms for the heat transfer in annular two-phase flows existing in some heat exchange systems. The basic physics include the peeling of droplets from the liquid film due to high friction with the gas phase and the collision of droplets with the liquid film or deposition into the liquid film. Droplet deposition particularly plays a crucial role in the course of film dryout events which might have a vital importance for particular systems. In this study, a new numerical method (named as VOF–MPS hybrid method) based on the moving particle semi-implicit (MPS) method was developed to analyze the droplet deposition onto a stagnant thin liquid film. That proposed method combines the volume of fluid (VOF) solver of the open-source CFD code OpenFOAM with the MPS method. VOF–MPS technique introduces the surface tension force calculation of VOF model into the MPS method. MPS method formerly employed the continuum surface force (CSF) approach based on the particles. Three-dimensional (3D) VOF–MPS simulation with the novel surface tension modeling addition to the MPS-based modeling provides a smoother liquid–gas interface on the crown formed after the impact. Droplet deposition experiments were also carried out for the validation and comparison of two models. VOF–MPS method could predict the crown parameters such as crown thickness relatively better than the MPS method employing the CSF. However, the instabilities formed at the tip of the crown, observed in the experiments, could not be resolved with both methods.Graphical Abstract

Assessment of the models in RELAP/SCDAPSIM with QUENCH-06 analysis

A fundamental principle of accident management in a nuclear power plant is the injection of water to cool the core. In this framework, a series of QUENCH tests have been conducted at Karlsruhe Institute of Technology (formerly Forschungszentrum Karlsruhe). The test results constitute a significant experimental database not only for further understanding of reflooding behavior, but also for code validation and improvement. The RELAP/SCDAPSIM code is a system code that is used to model reactor behavior and is widely used around the world. To date, assessment and validation have been performed with numerous experiments, including QUENCH tests. In the previous studies, the results of QUENCH simulations were referred to be sensitive to two main parameters: the electrical resistance and the thermal conductivity of the shroud insulator, which are subject to relatively large uncertainty. It is important to investigate these two parameters in detail, because this would enable identification of those SCDAP models that require further improvement. In this study, the uncertainty of the electrical resistance was reduced by modification of the code and subsequent validation with experimental data. In addition, modification of the thermal properties of the shroud insulator is suggested with consideration of the argon atmosphere in the facility. Finally, upcoming problems and questions are discussed. A rather good agreement was obtained than those of previous studies. As a result, more accurate modeling of the electrical resistance and the thermal properties of the shroud insulator was conducted and the importance of these parameters was evaluated.

Experimental Investigation Into Thermal Stratification by Direct Condensation in a Scaled Suppression Pool of Fukushima Daiichi Nuclear Power Plant

Experimental and numerical studies into thermal stratification by direct steam condensation in a torus type suppression pool were carried out to investigate the reactor core isolation cooling in the accidents of Fukushima Daiichi nuclear power plants. The suppression pool was manufactured to be a 1/22 scaled model of a Fukushima Daiichi nuclear power plant. Two different types of spargers were employed to simulate different units of the plants. In a sparger, 132 holes were uniformly drilled on the side of a pipe. However, the other sparger injected steam to the bottom. Flow rate was varied in a wide range to examine the effect on thermal stratification in the suppression pool. The experimental results showed that the sparger type influenced formation of thermal stratification. Moreover, steam flow rate strongly affected the onset time of thermal stratification, and the disappearance of the thermal stratification was affected by subcooling temperature. Computer simulation using a commercial software was conducted and the results show similar temperature profiles to the experimental results. Steam condensation was visualized in a vicinity of the spargers using high speed camera.Copyright

Fluctuation transfer velocity measurement in a boundary layer around a thin edge plate using dynamic PIV

The dynamic (time resolved) PIV (particle imaging velocimetry) measurement technique was applied to high-speed gas flow in a narrow channel with an obstacle. The boundary layer was visualized with a high-speed APX RS camera and an Nd:YLF high repetition double-pulse laser. Nitrogen gas seeded with oil particles using Laskine nozzle flows through a 10 ? 10 mm2 square channel with Reynolds numbers of 11?000 and 34?000. Although a sufficient quantity of images was difficult to capture for the Re = 34?000 flow to visualize the vortex evolution in time for the time resolved analysis of the boundary layer, large scale structures of turbulence at the edge of the thin plate are clearly visualized in the temporal domain. Fluctuation transfer velocities in the boundary layer were measured employing the whole field two-point velocity correlation. It is proposed that dynamic PIV can open a way of measuring the fluctuation transfer velocities in the whole flow target area simultaneously for high-speed turbulent flows even in small scales.

Dynamic visualization of eutectic reaction between boron carbide and stainless steel

ABSTRACT To investigate the eutectic reaction process of control-rod materials in a boiling water reactor (BWR), fundamental tests using boron carbide (B4C) powder inserted between stainless steel (SS) plates were performed and dynamically visualized. The eutectic reaction process near the contact area of the two materials and the behavior of molten material and B4C powder were visualized in real time. The temperature, reaction area, and maximum reaction-layer thickness were obtained. The average temperature range of the test was 1455–1481 K. Through dynamic visualization, some important and previously undiscovered phenomena were observed. The solid part of the SS plate and the strong surface tension of the melt retained the melt inside the specimen, preventing it from flowing out from the surface; the melt then invaded the B4C powder region during the reaction. Diffusion of the B4C powder and migration of the nonreacted B4C powder from the B4C powder region through the retained melt to the SS region were observed. This migration accelerated the local reaction growth rate. The time-resolved observation of these dynamic phenomena offers significant insights to the improvement of numerical calculation codes for severe accident analyses of BWRs, including the Fukushima Daiichi nuclear reactors.

Comparison of pool boiling CHF of a polished copper block and carbon steel block on a declined slope

ABSTRACT This study conducts a critical heat flux (CHF) experiment on a carbon steel block, and the block is positioned on slope that is declined at angles of 5° and 10°. The results of the carbon steel block experiment were then analyzed and compared with the results obtained from a copper block experiment that had been conducted previously at the same test facility. The comparison showed that several different types of phenomena had occurred, and the carbon steel block CHF at both 5° and 10° was much lower than that of the copper block. Detailed images of the heating surface of each material were acquired by a high-speed camera under different heat fluxes and analyzed. The carbon steel block surface generates more bubbles compared to the copper block under the same heat flux, which indicates that the carbon steel block should have a large number of nucleation sites. This causes a higher CHF. Finally, several existing theories on CHF mechanisms were also analyzed in an attempt to explain the difference of copper and carbon steel. It seemed that the contact angle alone was not sufficient to explain the large CHF decrease in the carbon steel block.