bluuebunny/crossref_metadata_2025_split huggingface dataset:
Study on Laser-Controlled Surface Preparation and Condensation Heat Transfer Characteristics of Copper-Based Micro-Nano Structure Surface
Authors: Wang Zhikai , Chen Feng , Chen Jing , Huang Zhiyong , Yan He , Bo Hanliang , Dong Yujie 
Published: November 2024

Abstract
The micro-nano structure surface has significant influence on the condensation heat transfer. Studying and designing micro-nano structure parameters help to better explore methods for improving surface condensation heat transfer performance. In this paper, we firstly processed hill shaped microstructure with customizable structural parameters on the copper surface using femtosecond laser technology. Subsequently, copper oxide nanowire structures were prepared on copper surfaces with hill shaped microstructures utilizing thermal oxidation method by tube furnace. We have been conducted a series of condensation heat transfer experiments on the copper-based micro-nano structure surface. Further the condensation heat transfer coefficient on the surface of micro-nano structures were analyzed. Additionally, the dynamic behavior of condensation droplets was also investigated. The surface processing results show that the surface structure parameters can be well customized using femtosecond laser technology by adjusting the laser processing parameters and designing patterns. The experimental results of condensation heat transfer indicate that the surface heat transfer coefficient of the copper-based micro-nano structure has been significantly improved compared with smooth copper surface. It has been found that there is droplet jumping phenomenon at low subcooling, while flooding phenomenon occurs at high subcooling.
Thermo-Neutronics Coupled Simulation of a Heat Pipe Reactor Based on OpenMC/COMSOL
Authors: Nie Jingyu , Li Binqian , Wu Yingwei , Zhang Jing , Zhang Guoliang , Ren Qisen , He Yanan , Su Guanghui 
Published: November 2024

Abstract
As an advanced small nuclear reactor, the heat pipe reactor possesses several advantages, including high energy density, long operational lifetime, compact size, and strong adaptability to various environments, making it an optimal choice for specialized energy needs in future applications, such as deep-sea and deep-space domains. In this study, we developed a code system using OpenMC/COMSOL for neutron and thermodynamic simulations. The continuous-energy Monte Carlo code, OpenMC, was employed to generate homogenized cross-section databases, offering significant modeling flexibility compared to traditional deterministic lattice transport codes. The generated multi-group cross-sections from OpenMC were utilized in COMSOL for the coupled neutron and thermodynamic simulations of the entire core. To validate the OpenMC/COMSOL code system, benchmark problems for pressurized water reactors were computed, and the results of the “two-step” scheme for neutron physics were compared with full-core Monte Carlo neutron results. Furthermore, to investigate the applicability of the thermal-hydraulic coupling in the heat pipe reactor neutron physics model, typical heat pipe reactor assemblymodels were established and verified under various energy group numbers and homogenized regions. The results demonstrated good agreement for multiplication factors and power distributions. The research results indicate that the utilization of the cross-section library generated by OpenMC enables the capability for steady-state analysis and core design. The results obtained from COMSOL exhibit good overall agreement with respect to multiplication factors and power distribution.
Ultrasonic Doppler Velocimetry (UDV) for Cold Shock Transients in a Scaled Liquid Metal Cooled Reactor Plenum
Authors: Sieh Broderick M. , Bindra Hitesh 
Published: November 2024

Abstract
Liquid metal reactors experiencing low convective flux face a handful of issues that may deteriorate reactor safety. These long-standing issues, such as thermal stratification or thermal striping, can negatively impact the thermal behavior, mechanical systems, and neutronics of the nuclear reactor. To overcome these issues, safety analysis codes can be established. However, high-quality, validation grade experimental data are required for the development of such computational models, yet, such high-quality data are limited for low Prandtl number fluids such as liquid metals. To provide validation grade data for liquid metal, researchers at Purdue University have constructed the Gallium Thermal-hydraulic Experiment (GaTE).
The GaTE is an experimental facility designed to study liquid metal cooled reactor transient conditions within a 1/20th scale model of the Advanced Burner Test Reactor (ABTR) outlet plenum. Gallium serves as a surrogate for sodium in the facility as the metal also exhibits a low Prandtl number in liquid form. Additionally, gallium is not cost prohibitive, nontoxic, and non-reactive with air and water, unlike sodium. To produce the desired transient conditions within the scaled plenum, the GaTE is capable of controlling the plenum inlet jet velocity and pool temperature as well as the system circulation. Control of the flow throughout the system is permissible through use of a moving magnet pump. An electromagnetic flow meter (EMFM) is employed to provide flow rate feedback data. The temperature of the gallium entering the plenum is regulated via a heat exchanger utilizing hot oil on the secondary side. Similarly, the fluid exiting the plenum passes through a heat exchanger using cool water on the secondary to reduce the temperature of the liquid metal. Thermocouples monitor the temperature of the gallium throughout the loop. A fiber optic distributed temperature sensing system is positioned within the plenum to provide temperature measurements of the liquid metal pool.
Previously, the GaTE has been used to examine thermal stratification in a scaled liquid metal cooled reactor plenum based on flux Richardson number (Rif). These tests utilized the distributed temperature sensing system to characterize thermal stratification and the fluctuations associated with the stratification thermocline based on the proximity to the critical Rif. To relate the flux parameters to global parameters, flow velocity data within the plenum must be collected in addition to the temperature data. Ultrasonic Doppler velocimetry (UDV) is employed to obtain the desired data under cold shock transients in the scaled liquid metal cooled reactor plenum.
Load-Following Control of Nuclear Reactor Using Variable-Order Fuzzy Fractional PID Controller Based on the Fractional-Space Neutron Point Kinetics Model
Authors: Sun Hetao , Mengsitu Hailemichael Guadie , Luan Xiuchun , Zhou Jie , Wang Junling , Yang Zhida 
Published: November 2024

Abstract
With the increase in the proportion of nuclear power, the load-following capability of NPPs has increasingly become an important factor affecting the development of a system security economy. In this paper, a variable-order fuzzy fractional PID controller is applied for load-following control of a nuclear reactor based on the fractional-space neutron point kinetics (F-SNPK) model, which can show the dynamics of the nuclear reactor in a better way than the classical neutron point kinetics (CNPK) model. Three-group of delayed neutrons and reactivity feedback effects (temperature and poison feedback effects) are considered. The simulation results are compared with PID and PIλDμ controller and it is shown that the variable-order fuzzy fractional PID controller exhibits a better satisfaction and capacity during the load-following operations compared to the PID and PIλDμ controllers.
Quantifying Software Safety in Nuclear Power Plants: A Framework for Requirements Phase Assessment
Authors: Li Boyuan , Li Duo , Li Jianghai , Guo Chao , Xiong Huasheng , Zhou Shuqiao , Huang Xiaojin 
Published: November 2024

Abstract
With the increasing dependence on digital instrumentation and control (I&C) systems in nuclear power plants, software has become a significant determinant of system safety assurance. To expand the use of digital technology in the nuclear industry, systematic methods are required for quantifying the safety of software-based I&C systems in safety critical applications. A software safety quantification model limited to the requirements phase is built in this paper based on the causal mechanisms that challenge safety. A preliminary mathematical method was developed to assess the number of requirements faults, their sub-types, their hazard rates and a control gate that allows determination of when to move beyond the requirements phase. A case study is conducted on a function of a reactor protection system to verify the validity of the quantification model. The framework allows us to assess software safety quantitatively.
Dynamic Entropy Investigation of Feed Water Heater in Nuclear Power Plant During Start-Up Process
Authors: Zhang Zhijiang , Tian Zhaofei , Ma Xiaoyu 
Published: November 2024

Abstract
A dynamic simulation model has been established to study the entropy characteristics of the feed water heater start-up process, and the dynamic entropy analysis was conducted. The effects of step and continuous changes in extraction steam flow on the entropy characteristics of the start-up process were investigated. With the increase of the extraction mass flow rate, the entropy generation rate at steady state increased gradually. To study the effects of the continuous change in extraction steam flow, the entropy was characterized and analyzed with three types changes: the quick opening (QO), the linear and the equal percentage (EP). The optimal start-up method should have the EP extraction steam flow change characteristics with the smallest accumulated entropy generation (28.16 MJ/K), which was 32.01% less than that of the linear flow characteristic and 47.548% less than that of the QO flow characteristic. The proportion of entropy generation rate of each control volume under the design condition was obtained, and the proportion of steam condensing zone was the largest (46.08%). This study could provide important guiding significance for the start-up process of the feed water heater.
Investigating the Impact of Various Molten Salt Combinations on Reactor Criticality and Thermal Neutron Flux Distribution in SD-TMSR
Authors: Yilmaz Seda , Revankar Shripad , Xu Yunlin 
Published: November 2024

Abstract
Molten salt reactors represent a groundbreaking advancement in both advanced and conventional reactor technologies, leveraging fluid fuel/coolant salt for enhanced temperature control and thermal efficiency. In the pursuit of optimizing molten salt reactor efficiency, our research focuses on exploring diverse salt compositions within the structured framework of the SD-TMSR core geometry. This investigation maintains a fixed hexagonal lattice structure, distinguishing two core zones, including a crucial blanket region for thorium-based fuel breeding. The study systematically alters salt composition with various isotopes, specifically fissile (U-233) and fertile (Th-232) materials, providing critical insights into reactor criticality and thermal neutron flux distribution. The molten salts examined here include different combinations of materials listed as lithium (Li), fluoride (F), beryllium (Be), sodium (Na), zirconium (Zr), and rubidium (Rb). Nine molten salts were tested for their criticality performance, and the reactor reached criticality with only four of them. Therefore, the neutron flux distribution simulations were performed only with those salts. The primary purpose of the research is to compare the performance of the core reactivity with different molten salts and identify the most suitable salt for SD-TMSR. Through the broadening of our knowledge and understanding of the complexities in MSR design, this study emphasizes the crucial position of molten salt reactors in advanced next-generation nuclear technology, establishing them as robust and sustainable contributors to the global energy perspective.
Methodology Development for Explosion Hazard Evaluation in Hydrogen Production System Using High Temperature Gas-Cooled Reactor
Authors: Morita Keisuke , Aoki Takeshi , Shimizu Atsushi , Sato Hiroyuki 
Published: November 2024

Abstract
High temperature gas-cooled reactor (HTGR) is expected to use nuclear heat to wide range of industrial applications such as hydrogen production due to its high temperature heat supply capability and inherent safe characteristics. One of the key technologies is a safety evaluation method which can simulate an impact of explosion hazards occurred in the hydrogen production plant to safety functions in the nuclear plant because HTGR hydrogen production system contains large amount of combustible gases such as hydrogen.
A computational fluid dynamics code FLACS has been sufficiently validated for dispersion and explosion of combustible gases such as hydrogen and methane gas worldwide, however, only few attempts have been made for validation of analysis in closed area with small space. A leak of combustible gases to the piping in HTGR hydrogen production system may occur in case of abnormal condition in hydrogen production plant and therefore an explosion in the piping should be considered.
Two experiments on combustible gas explosion in piping are selected to compare with calculated results obtained by FLACS code. One is the deflagration of natural gas in a simple closed system consisting of two spherical vessels of different sizes and a thin straight piping (Experiment 1), and the other is the deflagration of mixture of hydrogen and methane gas in an opened complex piping including multiple branches (Experiment 2).
The comparison of calculated and experimental results showed good agreement for maximum pressure and temperature in Experiment 1. In Experiment 2, the maximum pressure and flame spread velocity in a section of straight piping from the ignition point to the first branch also agreed well in calculated and experimental results.
Remote Tapping Inspection Method Employing a String Shooter
Authors: Tokuomi Saeko , Mori Kazuya , Ohshima Yasutaka 
Published: November 2024

Abstract
Tapping inspections of structures are commonly performed through manual efforts by inspectors. However, in the case of nuclear structures, there may be situations where it is preferable to avoid close proximity to the objects under inspection. Moreover, there is a desire for the automation of these inspections. This study proposes a completely new tapping inspection method that enables inspections without the need to approach the object.
Our specific approach involves the use of a string shooter for tapping inspections. Known commonly as a toy that launches a looped string into the air using two rollers, we hypothesized that attaching a steel ball for impact to this string could enable remote tapping of the inspection object. Consequently, we attached a 4mm diameter stainless steel ball to the string of a string shooter and attempted to detect a circular void defect with a diameter of 300mm within a concrete test specimen measuring 600mm × 600mm × 175mm from a distance. When the specimen was randomly tapped from a 2-meter distance, a unique acoustic signal was generated above the void defect. By assigning colors to the impact points based on the intensity of the acoustic frequency spectrum, we were able to differentiate the area corresponding to the void defect from healthy areas by color, thus facilitating the identification of void defect areas. This method allows for tapping inspections to be conducted remotely from several centimeters to several meters away, and it is capable of tapping inspection surfaces in any direction, from horizontal to vertical. By mounting the inspection device on a drone, it also becomes possible to inspect high places.
Numerical Investigation of Convective Heat Transfer Performance of Hybrid Nanofluid Flowing Through Square Channel in PWR Fuel Rod Assembly
Authors: Oke Samuel , Gao Puzhen , Fadodun Olatomide , Jayeola Mathew A. 
Published: November 2024

Abstract
The invention of hybrid nanofluid provides a better cooling ability than the ordinary nanofluid. Its convective heat transfer performance and thermal conductivity foster a new opportunity for reducing thermal hydraulic challenges of nuclear reactor. This work investigates the effect of Reynold’s number (1.5 × 104 ≤ Re ≤ 2.8 × 104) and concentration of nanoparticle (0 ≤ VR ≤ 2%) on cladding wall temperature, average Nusselt number, pressure drop and performance evaluation criterion of copper-alumina hybrid water based nanofluid flowing through square sub channel of nuclear fuel rod assemblies in typical pressurized water reactor. The flow of turbulent hybrid nanofluid in square channel PWR nuclear fuel rod assembly was modeled using single phase and the k-ϵ turbulent models. The results are compared with existing correlations and give good agreement. The result shows that the addition of very moderate copper-alumina hybrid nanoparticles into water causes significant augmentation in heat transfer performance of about 40% than ordinary water. Also, copper-alumina hybrid nanofluid gives better heat enhancement ranging from 10% to 30 % compared to ordinary alumina nanofluid. with higher volume fraction. This implies that usage of hybrid nanofluid should be considered instead of mono nanofluid. since hybrid nanofluid has capability to increase the performance evaluation criteria (PEC) and lower the cladding wall temperature than water and mono nanofluid.
Numerical Investigation of Vapor-Liquid Direct Contact Condensation Heat Transfer Behavior Under the Spray Condition
Authors: Wang Yiwei , Wei Hongyang , Gao Puzhen , Tian Ruifeng , Tan Sichao 
Published: November 2024

Abstract
The direct contact condensation heat transfer between the liquid film formed by the nozzle and the saturated steam is a complex process, and many experiments and simulations have been done by predecessors. In this study, CFD was used to calculate the heat and mass transfer process of direct contact condensation between liquid film and saturated steam. The VOF model was used to obtain a clear two-phase interface. The fixed cone angle formed by the liquid film was used to improve the phase interface area density of contact condensation heat transfer between liquid film and steam in the heat and mass transfer model. The results are in good agreement with the previous experimental and simulation results. Comparing the numerical results obtained before and after the treatment of the interfacial area density, the variation of the temperature and thickness of the liquid film is basically the same, but there are some differences in the local heat transfer coefficient of the liquid film. It can be seen that the treatment of the interfacial area density has a great influence on the heat transfer coefficient of the liquid film.
A Novel Framework for Spatiotemporally Decoupled Source Reconstruction of Radionuclides Released Into the Atmosphere
Authors: Xu Yuhan , Dong Xinwen , Zhuang Shuhan , Hu Hao , Fang Sheng 
Published: November 2024

Abstract
In recent years, Europe has been confronted with a series of radionuclide leak incidents, emanating from sources that remain unidentified. Such incidents are typically assumed to originate from a single, static point source, allowing the source location to be pinpointed to a specific, time-unvarying coordinate. However, the unavailability of detailed information about these leaks, including crucial data like the operating status of involved nuclear plant, renders direct estimation of release details using a top-down approach unfeasible. Consequently, a bottom-up approach, relying on environmental measurements and atmospheric dispersion modeling, becomes the only feasible choice. This approach focuses on the spatiotemporal source reconstruction of atmospheric radionuclide release, aiming to optimize the source location and release rate by minimizing the discrepancy between simulations and measurements. Owing to the interdependence of these spatiotemporal source parameters, traditional source reconstruction methods often attempt to resolve both temporal and spatial aspects simultaneously. The resulting tremendous solution space may cause substantial estimation errors in both the source location and release rate.
To tackle this issue, this paper proposes a novel framework for spatiotemporally decoupled source reconstruction. This framework splits the reconstruction into the source localization and release rate estimation stage, reducing the solution space at each stage. Initially, an artificial neural network model is employed to map a relationship between the source location and several simulation-measurement error metrics. The error metrics for each source location are obtained based on the release rates calculating by the Tikhonov regularization method. Subsequently, a likelihood function is constructed based on these error metrics, followed by assessing the posterior distribution of the source location using the Bayesian statistics. Finally, based on the estimated source location, the release rate is calculated using the Tikhonov regularization method.
The proposed framework has been first validated using a European Tracer Experiment (ETEX) whose source is previously known. The source location error was less than 100 km and the estimated release rate profile match the truth well. Furthermore, this framework was applied to reconstruct the source of the 2017 106Ru leak incident, a case where the source remains unknown. The estimated source location and release amount are consistent with public literature. The successful application of this framework to both controlled experiment and real-world radionuclide leak incident underscores its potential as a key component of the nuclear emergency response system.
Remote Measurement by Robot Arm Equipped With PAUVP and LIBS
Authors: Ji Chengzuo , Chen Yuan , Brice Christian , Kikura Hiroshige , Takahashi Hideharu , Endo Gen 
Published: November 2024

Abstract
This paper presents a prototype of a system for remotely detecting leaks and determining the shape and elements of the sediment in the primary containment vessels (PCVs) of the Fukushima Daiichi Nuclear Power Plant, which is currently undergoing decommissioning. The accident resulted in breaches of the PCVs, causing the cooling water injected to cool the fuel debris to leak, along with relatively large sediments present in the leakage field in addition to small fuel debris. Up to now, the primary focus of the investigation is the internal structure of the vessel, leakage position, and the distribution of nuclear fuel debris within the reactor.
The paper applies the Phased Array Ultrasound Velocity Profile (PAUVP) and Laser-Induced Breakdown Spectroscopy (LIBS) is employed in this study to identify the elements present in the sediment above the water surface part. Additionally, an arm robot capable of carrying both PAUVP and LIBS was developed as a result of the radioactivity.
The experiments conducted in this study confirmed that the location of the leakage and the sediment characteristics were estimated from the visualization flow field map and the elements of sediment above water were identified, enhancing the efficiency of the detection process. The results showed that the PAUVP technique could accurately capture the velocity distribution and the leakage location in the simple flow field and that the LIBS technique could effectively determine the elemental composition of the above-water part. The arm robot was able to perform the measurements remotely and safely, avoiding human radiation exposure.
This study demonstrates the feasibility and potential of using the PAUVP and LIBS techniques combined with an arm robot for investigating leak locations and sediment properties in the PCVs of the Fukushima Daiichi Nuclear Power Station. The system developed in this paper could provide valuable information for the decommissioning process and contribute to the safety and environmental protection of the nuclear power plant.