bluuebunny/crossref_metadata_2025_split huggingface dataset:
Vertical-Downward Two-Phase Flow Regime Identification by Probabilistic Neural Network (PNN) and Nonlinear Support Vector Machine (SVM)
Authors: Zhong Wenyi , Qiao Shouxu , Hao Sijia , Li Xupeng , Tan Sichao 
Published: October 2021

Abstract
The present study proposes a new feature extraction method based on non-stationary conductivity probe signals. Two types of discriminative network models, i.e., the probabilistic neural network (PNN) and nonlinear support vector machine (SVM), are established for flow regime identification using small sample sets. The eigenvectors are composed of 16 feature quantities obtained by wavelet packet decomposition (WPD) and 8 feature quantities in the time-domain derived from the reconstructed low-frequency signals. The 8 features include maximum, minimum, standard deviation, arithmetic mean, kurtosis, peak factor, impulse factor and margin factor. The signals are normalized based on features rather than samples before flow regime identification. In the current study, WPD results show that the conductivity probe signals in two-phase flow are mostly in low frequency. The identification accuracy of the nonlinear SVM is 90.47%, which is better than 83.33% by the PNN method. This study verifies the superiority of nonlinear SVM in solving small samples and nonlinear flow regime classification problems. However, the accuracy of flow regime identification near flow regime transitional boundaries still remains questionable and needs further improvement.
Application of Combining 3D Model and Survey on Site to Simulate Dome Lifting
Authors: He Wei-ting , Zhou Yuan-xia , Yang Jie 
Published: October 2021

Abstract
Dome lifting is one of the most important milestones for nuclear power project. The lifting operation must be properly planned and carried out in safety manner. The simulation and analysis are usually done before lifting activities to ensure the smooth lifting implementation. By using of 3D model and survey data on site to assist dome lifting, the risks related to lifting activities can be discovered and the risk elimination scheme can be formulated at early stage. Combining with Power Plant Design Management System (PDMS), design inspection software NAVISWORKS and survey data on site, the accuracy of simulation and verification can be increased, the deviations and non-conformities can be easily found. Through the implementation of this technical scheme, it helps to ensure the smooth and safe implementation of the milestone of dome lifting for nuclear power project. This paper gives an example for combining 3D model and survey on site to simulate dome lifting, methods and suggestions are given for guidance for other projects.
Experimental Study on the Flow Characteristics of Rod Bundle Under Rolling Motion
Authors: Li Xin , Qi Peiyao , Tan Sichao , Qi Chao , Qiao Shouxu 
Published: October 2021

Abstract
The floating reactor system will be rolling, heaving and other movements affected by waves, wind, etc. These motions will introduce additional inertial force field into the rod bundle, thus affecting the flow characteristics in the rod bundle channel. In order to study the influence of rolling motion on the flow characteristics of rod bundle, a visualization research of the flow field of rod bundle channel with a pitch-to-diameter ratio of 1.326 was carried out under rolling motion. The results show that under a small driving force, the rolling motion has a significant effect on the flow field in the rod bundle, affecting the velocity distribution in different sub-channels, and there is transverse mixing between adjacent sub-channels. With the increase of driving force, the influence of rolling motion is gradually weakened. The flow field distribution under rolling motion is significantly different from that under pulsating flow. The experimental results show that the influence of rolling motion on the middle sub-channel of the rod bundle channel is small, and the influence on the edge sub-channel is large. The velocity field of the subchannels on both sides of the edge fluctuates periodically, and the wave phase is inverse. This study shows that the flow field variation caused by rolling motion is different from that caused by pulsating flow, and the flow field of fuel assembly in floating reactor system under ocean condition is further studied.
Investigation of Applicability of Subchannel Analysis Code ASFRE on Thermal Hydraulics Analysis in Fuel Assembly With Inner Duct Structure in Sodium Cooled Fast Reactor
Authors: Kikuchi Norihiro , Imai Yasutomo , Yoshikawa Ryuji , Doda Norihiro , Tanaka Masaaki 
Published: October 2021

Abstract
In the design study of an advanced sodium-cooled fast reactor (Advanced-SFR) in Japan Atomic Energy Agency (JAEA), the use of a specific fuel assembly (FA) with an inner duct structure called FAIDUS has been investigated to enhance safety of Advanced-SFR against a core disruptive accident. Due to the asymmetric layout of fuel rods by the inner duct, it is necessary to estimate the temperature distribution to confirm feasibility of FAIDUS. In JAEA, an in-house subchannel analysis code named ASFRE has been developed as a FA design tool. For the typical FAs, the numerical results of ASFRE were validated by comparisons with experimental data. As for the FAIDUS, confirmation of validity of the numerical results by ASFRE was not enough because the reference data on the thermal hydraulics in FAIDUS have not been obtained by the mock-up experiment, yet. In this paper, therefore, the code-to-code comparisons with numerical results of ASFRE and those of an in-house CFD code named SPIRAL was conducted to make further discussion on applicability of ASFRE to the thermal-hydraulics analysis of FAIDUS. The applicability of ASFRE was indicated through the confirmation of the consistency of mechanism of the specific temperature and velocity distributions appearing around the inner duct between the numerical results by ASFRE and those by SPIRAL.
Heat Transfer Performance for Helium Gas Flowing in a Minichannel With Different Inner Diameters
Authors: Xu Feng , Liu Qiusheng , Shibahara Makoto 
Published: October 2021

Abstract
The high heat load on the first wall of the helium cooled blanket is removed by tube flow of helium gas. Heat transfer augmentation is considered to be acquired by downsizing of channels. Therefore, this paper experimentally studied the influence of inner diameter on the heat transfer performance of helium gas flowing in a minichannel. The helium gas flowed in the small platinum tubes with the inner diameters of 0.8 mm and 1.8 mm, respectively. The heat generation rate of the tube was controlled by a heat input subsystem and raised with an exponential equation. The surface temperature and heat flux of the tubes were obtained under a wide range of e-folding time at different flow velocities. The heat transfer coefficients of different inner diameter tubes were compared at the same conditions. The heat transfer performance of the 0.8 mm-diameter tube was compared with a classical correlation. The experimental results showed that the heat transfer performance in the minichannel is better than a conventional large-diameter tube. The heat transfer coefficients of the 0.8 mm-diameter tube were higher than those of the 1.8 mm-diameter tube. The heat transfer process was enhanced with reducing the inner diameter of the minichannel. The heat transfer process was divided into two parts including transient and quasi-steady-state regions.
Study on Laminar Turbulent Transition in Square Arrayed Rod Bundles
Authors: Dutra Carolina S. B. , Merzari Elia 
Published: October 2021

Abstract
The study of coolant flow behavior in rod bundles is of relevance to the design of nuclear reactors. Although laminar and turbulent flows have been researched extensively, there are still gaps in understanding the process of laminar-turbulent transition. Such a process may involve the formation of a gap vortex street as the consequence of a related linear instability.
In the present work, a parametric study was performed to analyze the spatially developing turbulence in a simplified geometry setting. The geometry includes two square arrayed rod bundle subchannels with periodic boundary conditions in the cross-section. The pitch-to-diameter ratios range from 1.05 to 1.20, and the length of the domain was selected to be 100 diameters. No-slip condition at the wall, and inlet-outlet configuration were employed. Then, to investigate the stability of the flow, the Reynolds number was varied from 250 to 3000. The simulations were carried out using the spectral-element code Nek5000, with a Direct Numerical Simulation (DNS) approach. Data were analyzed to examine this Spatio-temporal developing instability. In particular, we evaluate the location of onset and spatial growth of the instability.
Phase-Field Model for Recrystallization of Impurities in Sodium Coolant
Authors: Kawaguchi Munemichi 
Published: October 2021

Abstract
In researches and developments (R&Ds) on sodium (Na) management technology, the experimental data related to a cold trap and a plugging meter have been accumulated in each country because the impurities such as sodium oxide (Na2O), sodium hydride (NaH) and/or sodium hydroxide (NaOH) in Na coolant accelerate the corrosion on the stainless-steel surface. The cold trap has stainless-steel wire mesh and wall where the saturated impurities are recrystallized to remove the impurity in the Na coolant. Similarly, the plugging meter has stainless-steel orifice where the saturated impurities are recrystallized to measure the saturation concentration (plug temperature).
The recrystallization is common physical chemistry phenomena, in which is dominated by the temperature and the concentration. To date, a phase-field model (PFM) has been developed extensively as a powerful tool to predict microstructure evolution of micro–meso scale. Especially, the PFM has some simulation merits: it can calculate movement of interface without explicit trace of the interface (calculate robustly the large deformation), and this simulation results are known to be consistent with the thermodynamics data. In this study, we developed the simulation models for complex recrystallization of one impurity in Na coolant using the PFM, and confirmed that the simulation results are reasonable for the experimental data with the operating experience. In the simulation results for the one impurity of Na2O or NaH, we concluded that the kinetics of recrystallization was determined by the solubility and/or diffusion behavior in liquid Na.
Research on Acoustic Characteristic of Steam Injection Under Small delta-T Below Saturation
Authors: Li Hui , Li Yong , Xiao Qi , Li Dongyang , Tan Sichao 
Published: October 2021

Abstract
Bubbling deaeration is widely applied in marine steam power system (MSPS), and direct contact condensation (DCC) of steam injected into water under low subcooled water is the thermodynamic foundation of this technology. Due to the small delta-T below saturation of the flow field, the steam condensation potential is small, and there is a violent steam plume evolution. The process and the growth, merging, breaking and collapse of bubbles, which produce severe pressure oscillation and vibration noise. In response to this problem, the flow pattern evolution and acoustic characteristics of underwater steam injection in the range of different subcooling degrees (10–0.2°C) and steam mass flow (18kg/h) were studied through experiments. Studies have found that the degree of subcooling has an important impact on the evolution of the steam plume and the spectral characteristics of noise. The change in the shape of the steam plume determines the amplitude and spectral characteristics of the noise sound pressure. As the ambient water temperature rises, the steam plume shape changes differently, and the phenomena such as “tail oscillation”, “congestion”, and “steam recirculation” appear in sequence. Affected by the evolution of the above-mentioned steam plume shape, the noise sound pressure amplitude changes with the subcooling. The temperature first rises and then falls, and it rises slightly when it is close to the saturation temperature. At the same time, the noise spectrum shifts from medium and low frequency to low frequency. When it is close to the saturation temperature, the spectrum tends to move to medium frequency.
Comparison of Pebble Bed Velocity Profiles Between High-Fidelity and Intermediate-Fidelity Codes
Authors: Reger David , Merzari Elia , Balestra Paolo , Schunert Sebastian , Hassan Yassin A. 
Published: October 2021

Abstract
Recent interest for the development of high-temperature gas reactors has increased the need for more advanced understanding of flow characteristics in randomly packed pebble beds. A proper understanding of these flow characteristics can provide a better idea of the cooling capabilities of the system in both normal operation and accident scenarios. In order to enhance the accuracy of computationally efficient, intermediate fidelity modeling, high-fidelity simulation may be used to generate correlative data. For this research, NekRS, a GPU-enabled spectral-element computational fluid dynamics code, was used in order to produce the high-fidelity flow data for beds of 1,568 and 45,000 pebbles. Idaho National Lab’s Pronghorn porous media code was used as the intermediate fidelity code. The results of the high-fidelity model were separated into multiple concentric regions in order to extract porosity and velocity averages in each region. The porosity values were input into the Pronghorn model and the resulting velocity profile was compared with that from NekRS. Both cases were run with a Reynolds number of 20,000 based on pebble diameter. The Pronghorn results were found to significantly overestimate the velocity in the outermost region indicating that changes in the porosity alone do not cause the difference in fluid velocity. We conclude that further work is necessary to develop a more effective drag coefficient correlation for the near-wall region and improve predictive capabilities of intermediate fidelity models.
U.S. Nuclear Power Plant Performance Assessment Using the Versatile Economic Risk Tool (VERT)
Authors: Miller Jaden C. , Ercanbrack Spencer C. , Pope Chad L. 
Published: October 2021

Abstract
This paper addresses the use of a new nuclear power plant performance risk analysis tool. The new tool is called Versatile Economic Risk Tool (VERT). VERT couples Idaho National Laboratory’s SAPHIRE and RAVEN software packages. SAPHIRE is traditionally used for performing probabilistic risk assessment and RAVEN is a multi-purpose uncertainty quantification, regression analysis, probabilistic risk assessment, data analysis and model optimization software framework. Using fault tree models, degradation models, reliability data, and economic information, VERT can assess relative system performance risks as a function of time. Risk can be quantified in megawatt hours (MWh) which can be converted to dollars. To demonstrate the value of VERT, generic pressurized water reactor and boiling water reactor fault tree models were developed along with time dependent reliability data to investigate the plant systems, structures, and components that impacted performance from the year 1980 to 2020. The results confirm the overall notion that US nuclear power plant industry operational performance has been improving since 1980. More importantly, the results identify equipment that negatively or positively impact performance. Thus, using VERT, individual plant operators can target systems, structures, and components that merit greater attention from a performance perspective.
Investigation on Thermal Stability of Sintered Magnesia in Sodium for Core Catcher Application in SFRs
Authors: Shukla Prabhat Kumar , Hemanth Rao E. , M. Muthuganesh , Vetrivendan E. , Polaki S. R. , Das Sanjay Kumar , Ningshen S. , Chaurasia Pramod Kumar , Ponraju D. , Athmalingam S. , Venkatraman B. 
Published: October 2021

Abstract
Sodium cooled Fast Reactors (SFRs) are built with several engineered safety features to prevent and mitigate core melt accidents. Core catcher is a passive mitigation device, which retains core debris resulting from hypothetical core melt accident in a subcritical and coolable geometry. The core catcher of Prototype Fast Breeder Reactor (PFBR) uses Austenitic Stainless Steel (ASS) as structural material. However, in the view of enhanced safety, core catchers for future reactors need to be designed for large thermal loads resulting from whole core melting. Provision of suitable refractory material as protection layer for ASS made core catcher plate can be one of the promising options to enhance the melt retention capability. Thermal stability of protective layer during impact of molten corium or fragmented core debris at high temperature in presence of liquid sodium is very essential for its satisfactory performance. Based on preliminary studies, sintered magnesia has been chosen as candidate material and experiments were carried out to investigate its stability in liquid sodium. Accelerated sodium compatibility tests were carried out by exposing magnesia specimens in sodium for 240 h at 850 °C. Tests were also carried out to assess its stability in liquid sodium under condition envisaged during whole core relocation over core catcher. Tests were carried out by subjecting the specimens under dead weight up to the boiling point of sodium in lower plenum of a typical SFR. Sodium exposed magnesia specimens were subjected to XRD and SEM analysis to assess material degradation. The specimens were also tested for change in compressive strength due to high temperature sodium exposure. The experimental results are discussed in the paper.
Analysis of Passive Tube Condensation With Non-Condensable Gas Using Heat and Mass Analogy Model
Authors: Cotul Ugur , Revankar Shripad T. 
Published: October 2021

Abstract
In this study, we used the heat and mass analogy model to be able to predict the heat transfer properties of a condenser tube operating in passive mode. The most important advantage of analogy model comparing boundary layer model is simplicity and fast computation, that’s why it can be applied to various engineering problems for many cases. The heat and mass analogy model is based on the heat transfer balance between liquid film and gas mixture area. The main problem for the liquid film region is the heat transfer coefficient (HTC) which is affected negatively in the presence of non-condensable gas. Therefore, our main goal is to increase the HTC and condensation heat transfer rate by updating the analogy code. In the gas-vapor mixture region, heat transfer mainly occurred as latent condensation and sensible heat transfer. In order to maintain this balance between the mixture and liquid film, the interface temperature is iterated. After defining a specified tolerance value of the heat and mass analogy model codes, this iteration process was started to be used at the entrance of a condenser tube. The gas and vapor mixture is considered to be saturated at the liquid/gas interface in the heat and mass transfer analogy model. Via boundary layer study of species concentration and energy balance, the non-condensable gas effect on condensation is added into the equation. For the condensation heat transfer coefficient of turbulent vapor flow associated with laminar condensate, numerical predictions were made and they were satisfactory. The predictions were compared with the experimental data from the literature to be able to test the model. Non-condensable gas mass fraction and vapor-non-condensable mixture temperature were presented in the form of radial and axial profiles.
UHT Test Facility Updates and Oxidation Tests for Accident Tolerant Fuel Development
Authors: Byers William A. , Wang Guoqiang 
Published: October 2021

Abstract
Research and development on accident tolerant fuel (ATF) have been widely pursued by the nuclear industry around the world. Westinghouse is developing new accident tolerant fuel design, EnCore® Fuel. Chromium (Cr) coated zirconium alloy and silicon-carbide (SiC) fuel cladding are the excellent candidates for application in fuel cladding of light water reactors (LWRs)[1, 2]. The ATF fuel is being developed to deliver beyond-design-basis (BDB) and design-basisaltering (DB) safety margins, withstanding far more severe conditions than the current fuel. The products in this portfolio have enhanced temperature performance enabling survival during BDB and DB accidents. The products also minimize interactions of fuel and fuel rod materials with water. The Cr-coated zirconium alloy cladding exhibits sustained high temperature performance and inhibits the detrimental effects of the zirconium-steam reaction. Westinghouse’s silicon-carbide (SiC) fuel cladding offers safety benefits in severe accident conditions, particularly compared to the significant hydrogen-and-heat producing reactions that occur above 1200° C for zirconium fuel cladding. These advancements will increase safety margins and enable the transition to high burnup facilities as well.
To support the accident tolerant fuel (ATF) development, the Westinghouse Ultra High Temperature (UHT) test facility was built at the Westinghouse Churchill site in early 2016. The UHT test facility has been upgraded recently. The current UHT reactor was designed to promote better sealing of the reaction chamber and a smaller uncertainty on cladding temperature measurement. Simulating large break LOCA steam conditions, the UHT oxidation tests have been performed with SiC Optimized ZIRLO® and Cr-coated zirconium alloy cladding. In the EnCore Fuel program, one of the key tasks is to test the safety features and margin gain of the EnCore Fuel. This paper describes the designs, safety features and baseline commissioning tests for the upgraded Ultra High Temperature (UHT) test facility.
Analysis of Particle Transfer Behavior in Fuel Rod Bundles Using CFD Lagrangian Particle Tracking Method
Authors: Xu Yiban , Krammen Michael A. , Wang Guoqiang , Fisher Jesse S. , Karoutas Zeses 
Published: October 2021

Abstract
Crud has been observed on the fuel rod surfaces in a variety of fuel designs around the world, and in some limited situations fuel performance was compromised due to crud-induced power shift (CIPS) and/or crud-induced localized corrosion (CILC). It is generally believed that crud deposition depends on fuel rod surface sub-cooled nucleate boiling, coolant chemistry and the availability of particles from component corrosion or from reinserted fuel. The formation, release, and accumulation of crud on the fuel and its influence on CIPS and/or CILC is a complicated process involving multi-physics phenomena. This study uses Computational Fluid Dynamics (CFD) Lagrangian Particle Tracking (LPT) techniques in analysis of particle transfer behavior in fuel rod bundles focusing on flow swirl and turbulence impacts. It is hoped that high fidelity CFD results can provide insights into particle transfer behaviors in the bulk coolant as well as near the fuel rods, which may provide guidance for model development of lumped or integrated analysis methods.
The CFD model was built based on the best practices learned from previous single-phase analyses. The LPT options, including particle injectors, forces on particles, and solver settings, were verified by comparing the simulated results to the test data from simple geometry with various particle sizes, covering deposition mechanisms in diffusion-, turbulent- and inertial-dominated regimes. The tested model then was applied to Westinghouse fuel designs with and without Intermediate Flow Mixing (IFM) grids. Particle concentration and size distributions in the coolant around fuel rods were obtained and the effects of grid induced swirl flow on particle transfer were identified. The analysis results may be included in lumped or integrated crud formation/release analysis methods. Limitations and potential improvements of this analysis method are also discussed.