bluuebunny/crossref_metadata_2025_split huggingface dataset:
Experimental Study on Bubble and Aerosol Behavior During Pool Scrubbing
Authors: Yoshida Kohei , Fujiwara Kota , Kaneko Akiko , Abe Yutaka 
Published: October 2021

Abstract
Understanding radioactive aerosol behavior is important during severe accidents in nuclear power plants. In particular, pool scrubbing at a suppression pool is a key factor for estimating the leakage of radioactive aerosol into the environment. Therefore, we have to elucidate the mechanism of the aerosol transfer from the gas phase to the liquid phase during bubbly flow. Currently, models to evaluate the amount of aerosol removal by pool scrubbing have been incorporated into integrated severe accident analysis codes such as MELCOR. However, the models for bubbles and aerosols behavior have not been validated enough by experiment. Therefore, the measurement of bubbles and aerosols behavior is required for validation and reconstruction pool scrubbing model. However, the quantitative measurement of bubbly flow is very difficult by complicated bubble shape and overlapping of bubbles in visualization. The wire-mesh sensor (WMS) is one of the most effective measuring methods for complicated bubbly flow. The reconstruction method of bubbly flow by WMS measurement is suggested by Prasser et al. This method can measure the bubble surface, meanwhile, individual bubble velocity and volume are not one-to-one correspondence because bubble volume is calculated by using radial average velocity. Kanai et al. proposed a method for acquisition bubble three-dimensional velocity method by bubble tracking. In this method, bubble velocity and volume are one-to-one correspondence. Thus, we suggested an algorithm of bubble surface measurement by applying a bubble tracking method. The bubble three-dimensional velocity and surface area is obtained by this algorithm. As a result, bubble rising velocity in bubbly flow is higher than bubble terminal velocity according to causing liquid phase velocity. In addition, the gas-liquid total interfacial area is overestimated in MELCOR predict by overestimating bubble breakup. This result suggests that the overestimating of DF at downstream is caused by overestimating bubble breakup in MELCOR model.
Microscopic Heat Transfer Characteristics During Cooling of High Temperature Surface by a Falling Liquid Film
Authors: Umehara Yutaro , Okawa Tomio 
Published: October 2021

Abstract
When the dry region on the high-temperature wall is cooled down by liquid, the dry region is transformed into wet region as the temperature shifts in a short time. Vigorous boiling occurs during the temperature transition. Especially, it is called the rewetting phenomenon that liquid contacts dry region on the high-temperature wall in the first time. In nuclear plants, it is essential for nuclear power plant safety to understand this phenomenon. Under severe accidents, a liquid film formed by an emergency core cooling system (ECCS) removes the decay heat of fuel rods. To determine the start time of ECCS, the time required for the liquid film to cover the fuel rods needs to be clarified. Working time is calculated by using rewetting velocity which means the boundary movement speed between liquid and dry region. Previous research reported rewetting velocity was calculated by heat conduction equation using the boundary condition (= heat transfer coefficient (HTC) distribution). However, proposed HTC distributions were not defined directly from experimental data but defined to match experimental results. This is because the temperature was measured by thermocouples which were not enough to obtain temperature distribution. In this study, two kinds of experiments were done to define HTC distribution from experimental results and to elucidate the rewetting phenomenon. First, as liquid film fell along high temperature wall, the transition of temperature distribution of the wall was measured by an infrared camera. Copper plate was used to heat transfer surface. From the results of temperature distribution, HTC distribution was calculated. And then, four critical parameters that consisted of HTC distribution were investigated against experimental parameters.
Second, a similar experiment was conducted using silicon wafer which is transparent for infrared rays as the heat transfer surface to observe the rewetting phenomenon, especially the precursory cooling region. The precursory cooling region is considered as one of the important regions to calculate exact rewetting velocity. An infrared camera and high speed camera was used in this experiment. An infrared camera took temperature distribution from the backside which is the opposite side of the liquid film. A high-speed camera took liquid film condition from the front side. Since both cameras were synchronized, this experimental apparatus was able to observe both HTC distribution and liquid film condition at the same time. As a result of observation, the precursory cooling region is the area where drying and wetting are repeated by boiling.
Outline of Guideline for Seismic Response Analysis Method Using 3D Finite Element Model of Reactor Building
Authors: Choi Byunghyun , Nishida Akemi , Shiomi Tadahiko , Kawata Manabu , Li Yinsheng 
Published: October 2021

Abstract
In the seismic safety assessment and design of building structures in nuclear facilities, lumped mass models are conventionally used. However, they cannot possess the required high-accuracy evaluation of nuclear facilities, such as the local response at the equipment location in a reactor building. From this point of view, a seismic response analysis method using a three-dimensional finite element (3D FE) model is indispensable. Furthermore, because analysis results obtained using 3D FE models vary to a large degree depending on the experience and knowledge of analysts, the quality of analysis results should be ensured by developing a standard analysis method. In the Japan Atomic Energy Agency, we have developed a guideline for seismic response analysis methods that adopt 3D FE models of reactor buildings, and we plan to publish it as JAEA report. The guideline consists of a main body, commentary, and several appendixes; it also includes procedures, recommendations, points of attention, and a technical basis for conducting seismic response analysis using 3D FE models of reactor buildings. In this paper, the outline of the guideline and analysis examples based on the guideline are presented.
Research on Eccentricity Performance of Capacitance Rod Position Measurement Sensor for Measuring Non-Metallic Rod
Authors: Li Yanlin , Qin Benke , Bo Hanliang 
Published: October 2021

Abstract
Rod position measurement sensor is the key equipment of control rod hydraulic drive line, the measurement accuracy of the sensor is directly related to the safety and reliability of the reactor. Capacitance rod position measurement sensor is generally composed of excitation and detection electrodes, ceramic tube, outer shielding layer and measuring rod. For the double helix capacitance rod position measurement sensor with non-metallic measuring rod, the static characteristic experiment of the sensor was completed, the finite element method was used to establish the calculation model and the model was verified by the experimental results. By the calculation model, the influence of electrode angle, pitch, thickness and relative permittivity of ceramic tube on the eccentricity error of the sensor was systematically analyzed, the optimal combination of design parameters were obtained by orthogonal test method. The results show that the calculation results are in good agreement with the experimental results, the eccentricity error can be reduced by reducing the electrode angle, increasing the electrode pitch, increasing the thickness and reducing the relative permittivity of the ceramic tube. The design scheme of the sensor obtained by the orthogonal test can greatly reduce the eccentricity error and realize the requirement of rod position measurement without losing step. The research results provide guidance for the design and optimization of capacitance rod position measurement sensor.
A New Concept for Irradiation Experiments in Fast-Reactor Environment: CFD Simulation of the LBE Loop in HYST
Authors: Kong Ran , Kim Seungjin , Wahlen Robert , Grimm Terry 
Published: October 2021

Abstract
The current work investigates the thermal-hydraulics of the fast core of a HYbrid fast/thermal core configuration Subcritical Testbed (HYST), a new concept for irradiation experiments in fast-reactor-like environment. The Reynolds-Averaged Navier Stokes (RANS) simulations using ANSYS Fluent are performed. A benchmark study is first performed to validate the simulation of liquid LBE flow in a wire-wrapped rod bundle geometry using the experimental data in Pacio et al. (2016), which provides the most detailed data for rod surface temperature, LBE temperature in the subchannel, and pressure drop along the test section. Two turbulence models (Standard k-ε model with Standard Wall Function and Realizable k-ε model with Enhanced Wall Treatment), two approaches to simplify the wire-rod contact and different models for turbulent Prandtl number (Prt) are tested. It is found that the selected models can predict the data well and are therefore used for simulations of the LBE loop in HYST. A mesh independence study is performed in simulating the LBE loop in HYST with three different meshes being generated (number of cells: 14.8 million, 25.8 million, and 45.4 million). It is found that the mesh independency is reached using the mesh with 25.8 million cells. The effect of non-uniform power distribution in the radial direction of the core is investigated. It is found that the temperature profiles change significantly when uniform power distribution changes to non-uniform. An experimental facility is being constructed and newly collected data will be used to evaluate the RANS simulations. Once validated, the current model will be valuable for the design, licensing, and operation of the HYST prototype.
Study on Heat Transfer Coefficient of Supercritical Water Based on Factorial Analysis
Authors: Xu Peng , Zhou Tao , Chen Ning , Chen Juan , Fu Zhongguan 
Published: October 2021

Abstract
Heat transfer coefficient has an important influence on the flow and heat transfer of supercritical water in the core channels. The effects of different factors and their interactions on the heat transfer coefficient of the supercritical water were studied by full factorial experimental design method, such as pressure, mass flow rate, heat flux, and inlet temperature. The results show that: Within the range of the tested working conditions, effect D (inlet temperature), effect B (mass flow rate) and effect A (pressure) had a significant impact on the heat transfer coefficient, where the percentage contribution of effect D was 48.21%; effect B was 21.58%; effect A was 15.1%. The percentage contribution of other factors and their interactions on the heat transfer coefficient of the supercritical water can be ignored. At the same time, a prediction formula of heat transfer coefficient on supercritical water was fitted, and it was found that the prediction error of this formula conformed to the assumption of normality, and the prediction error was 10.5%.
Risk Factors Selection Approach for Nuclear Decommissioning Risk Assessment, Modeling, and Management
Authors: Junior Awodi Ngbede , Liu Yong-Kuo , Ayodeji Abiodun , Adibeli Justina Onyinyechukwu 
Published: October 2021

Abstract
Due to the complexity of nuclear decommissioning projects, developing a risk assessment model to manage the project risk is paramount. The current study is an attempt to identify all the possible risk factors that can arise in a nuclear decommissioning project. A literature review and the expert judgement method were used to identify and analyse all the possible risk factors. The identified factors are then further refined, and the most important risk factors are selected to be included in the final risk assessment model. This work presents the criteria developed for the simplification, refinement, and assessment of risk factors. The criteria used considered the availability of data, the clarity, and the measurability of the selected factors. Risk factors with a lower rating are eliminated, and those with a similar focus of attention are grouped, enabling the selection of a simplified final list of risk factors for the model. This work also proposed a method of assessment for the remaining factors to allow the quantification of individual risk factors within the model. From the 81 risk factors initially identified, 18 consolidated factors are considered for the development of the risk assessment model. The selection process and risk rating approach presented in this work serves as a critical foundation for the development of a robust nuclear decommissioning risk assessment and management.
Study on Eutectic Melting Behavior of Control Rod Materials in Core Disruptive Accidents of Sodium-Cooled Fast Reactors: (1) Project Overview and Progress Until 2019
Authors: Yamano Hidemasa , Takai Toshihide , Furukawa Tomohiro , Kikuchi Shin , Emura Yuki , Kamiyama Kenji , Fukuyama Hiroyuki , Higashi Hideo , Nishi Tsuyoshi , Ohta Hiromichi , Morita Koji , Nakamura Kinya 
Published: October 2021

Abstract
One of the key issues in a core disruptive accident (CDA) evaluation in sodium-cooled fast reactors is eutectic reactions between boron carbide (B4C) and stainless steel (SS) as well as its relocation. Such behaviors have never been simulated in CDA numerical analyses in the past, therefore it is necessary to develop a physical model and incorporate the model into the CDA analysis code. This study focuses on B4C-SS eutectic melting experiments, thermophysical property measurement of the eutectic melt, and physical model development for the eutectic melting reaction. The eutectic experiments involve the visualization experiments, eutectic reaction rate experiments and material analyses. The thermophysical properties are measured in a range from solid to liquid state. The physical model is developed for a CDA computer code based on the measured data of the eutectic reaction rate and the physical properties. This paper describes the project overview and progress of experimental and analytical studies conducted until 2019. Specific results in this paper are the validation of physical model describing B4C-SS eutectic reaction in the CDA analysis code, SIMMER-III, through the numerical analysis of the B4C-SS eutectic melting experiments in which a B4C block was placed in a SS pool.
Nanoindentation Test of F321 Austenitic Stainless Steel Under Fe-ion Irradiation
Authors: Liu Meidan , Nie Junfeng , Lin Pandong 
Published: October 2021

Abstract
Nuclear technology, as a high quality, clean and reliable energy supply, is attracting broad interest from countries across the world. F321 austenitic stainless steel (F321SS) is widely utilized in key components of nuclear power plant due to its excellent corrosion resistance and high temperature mechanical properties. Irradiation can easily lead to the degradation behaviors of materials, such as irradiation hardening, irradiation embrittlement and high-temperature He embrittlement, etc. Understanding such degradation is important for predicting the evolution of material behavior under irradiation and extending the lifespan of existing nuclear reactors. Ion irradiation is most commonly used to model neutron-induced damage since the irradiation conditions (temperature, flux, spectrum, etc.) can be regulated more accurately and flexibly. In this paper, the Fe-ion irradiation experiments of F321SS at different temperatures and doses were carried out, and the nanoindentation experiments under different conditions were further conducted. Irradiation hardening is observed in all specimens and strongly depending on irradiation temperature and damage dose. The hardness after irradiating increases with doses and saturates for at least 1dpa under low temperature regimes (&lt; 300°C). However, at higher temperature (450°C and 560°C), nano-hardness reaches the peak at ∼0.5dpa and then declines. Moreover, the hardness of all specimens has a similar trend with temperature, that is, it first increases, reaches the peak, and then decreases.
Numerical Simulation of Vortex Shedding Downstream of a Thermoacoustic Engine Stack
Authors: Qiao Bowen , Yamada Shota , Tanaka Gaku 
Published: October 2021

Abstract
Thermoacoustic engines use waste heat available at a higher temperature to produce mechanical energy in the form of waves. The purpose of this study is to visualize the fluid behavior downstream of the stack of a thermoacoustic engine by applying numerical analysis and to explore the effect of vortex shedding on the temperature field. The self-sustained oscillation is reproduced by two-dimensional numerical analysis using the compressible solver of OpenFOAM, and the vorticity field and temperature field were visualized. The vortex downstream of the stack changed from symmetric to asymmetric near the peak of the flow velocity. Changes in the vortex structure by the unsteady flow caused irregular and non-uniform temperature field. It is found that the vortex significantly affects the heat transfer between the heat exchanger and the fluid.
Measurement of Liquid Film Thickness for Annular Two-Phase HFC134a Gas-Liquid Ethanol Flow in the Vertical Tube
Authors: Zhang Huacheng , Hisano Tutomo , Mori Shoji , Yoshida Hiroyuki 
Published: October 2021

Abstract
Annular gas-liquid two-phase flows, such as the flows attached to the fuel rods of boiling water reactors (BWR), are a prevalent occurrence in industrial processes. At the gas-liquid interface of such flows, disturbance waves with diverse velocity and amplitude commonly arise. Since the thin liquid film between two successive disturbance waves leads to the dryout on the heating surface and limits the performance of the BWRs, complete knowledge of the disturbance waves is of great importance for the characterized properties of disturbance waves. The properties of disturbance waves have been studied by numerous researchers through extensive experimental and analytical approaches. However, most of the experimental data and analyses available in the literature are limited to the near atmospheric condition. In consideration of the properties of liquids and gases under atmospheric pressure which are distinct from those under BWR operating conditions (7 MPa, 285 °C), we employed the HFC134a gas and liquid ethanol whose properties at relatively low pressure and temperature (0.7 MPa, 40 °C) are similar to those of steam and water under BWR operating conditions as working fluids in a tubular test section having an inside diameter 5.0mm. Meanwhile, the liquid film thickness is measured by conductance probes. In this study, we report the liquid film thickness characteristics in a two-phase HFC134a gas-liquid ethanol flow. A simple model of the height of a disturbance wave was also proposed.
Preliminary Simulations on the Atmospheric Dispersion of Radioactive Substance for the Two Sites in Tunisia
Authors: Ghannouchi Elyes , Wang Yu , Cao Jianzhu , Xie Feng , Zhang Liguo , Tong Jiejuan , Yao Rentai , Debbabi Khaled 
Published: October 2021

Abstract
Global climate change, record-high oil prices and the deceasing availability of fossil fuels are forcing us to rethink the way we deal with our energy needs. In response to this growing concern, and as an indication of the seriousness of these issues, the current administration of Tunisia is executing strategic plans to increase green energy sources. Nuclear energy is one such alternative source. The evaluation of the transport in the atmosphere of radioactive substance discharged from a nuclear power plant under normal operational or accidental conditions is a requirement of design and licensing. The atmosphere is a major exposure pathway by which radioactive materials that are either routinely discharged under authorization or accidentally released from a nuclear power plant could be dispersed in the environment and transported to locations where they may reach the public. In this paper, the development plan for nuclear energy in Tunisia will be introduced, and the atmospheric dispersion model of radioactive substance for the two sites in Tunisia will be presented. The meteorological parameters for specific site will be illustrated with the evaluation of the meteorological investigation. Contamination in the air and on the ground over short and long time periods will be described in the atmospheric dispersion models. Several related issues will be discussed, including the suitability of the meteorological investigation, appropriateness of the dispersion models, etc.
Grey Correlation Study on Natural Circulation Heat Transfer Coefficient of Liquid Metal
Authors: Ning Chen , Tao Zhou , Lanyu Zhou , Tian Qi , Juan Chen , Xiang Feng 
Published: October 2021

Abstract
Liquid metals have excellent thermal, nuclear physics, and high-density power properties that allow them to efficiently take heat out of the reactor. Based on the data of Natrium, Natrium-Kalium, Plumbum and Plumbum-Bismuth simulated by CFX software, the three factors of heating power, system pressure and inlet temperature were selected, and the analysis method of grey correlation model was used to analyze the correlation degree and weight of these factors to the heat transfer coefficient of different liquid metals. Through analysis, it can be concluded that: under Natrium, Plumbum and Plumbum-Bismuth working medium, the correlation degree and weight importance of heating power to heat transfer coefficient are greater than that of system pressure, while under NatriumKalium working medium, the correlation degree and weight of system pressure and heating power to heat transfer coefficient are roughly equal; The inlet temperature has the greatest correlation to the heat transfer coefficients of the four liquid metals, and the inlet temperature contributes the largest weight to the change in the heat transfer coefficient, that is, the inlet temperature and the heat transfer coefficient are most closely related; When the resolution is 0.01, the correlation degree ofthe heat transfer coefficient can be clearly distinguished between these three factors. Through the analysis of the degree of correlation and the weight of the influencing factors of different liquid metal heat transfer coefficients, it is helpful to discover the unstable factors that cause the deterioration of liquid metal heat transfer, and it is of great significance to the choice of new reactor coolants.
A Modified Model for the Net Vapor Generation Point and Its Application on CHF Prediction in Subcooled Flow Boiling
Authors: Akand Md. Abdur Rafiq , Kitahara Kei , Matsumoto Tatsuya , Liu Wei , Morita Koji 
Published: October 2021

Abstract
Critical heat flux (CHF) is one of the most significant thermal criteria for nuclear reactor design and safety in subcooled flow boiling. The accurate prediction capabilities of the characteristic size of departure bubbles are crucial for predicting the net vapor generation point (NVG) and CHF. An experimental research facility was designed to determine the bubble departure diameter and subcooling at the net vapor generation point, not only for vertical flow boiling but also in any orientation between vertical and downward-facing horizontal. An improved force-balanced model is proposed to calculate the bubble departure diameter at the net vapor generation point to predict CHF in vertical subcooled flow boiling. The force-balance equation was solved numerically through iteration to calculate the bubble departure diameter using a novel MATLAB script. The net vapor generation point was modified using the bubble departure diameter obtained from the proposed model. The model agrees well with the experimental data. Finally, the modified departure diameter and NVG were applied to the liquid sublayer dryout model to predict the CHF on upward subcooled flow boiling. It was found that the model predicts the experimental CHF data with an average relative error of 7.17%.