bluuebunny/crossref_metadata_2025_split huggingface dataset:
Design, Development and Testing of Large Diameter High Temperature Alkaline Metal Heat Pipe
Authors: Zhang Jiarui , Wang Chenglong , Tian Zhixing , Guo Kailun , Zhang Dalin , Tian Wenxi , Qiu Suizheng , Su Guanghui 
Published: November 2024

Abstract
Heat pipe is the most critical heat transfer element in the silent heat pipe reactor, which utilizes the phase change of the work material to carry out heat transfer work through gravity, centripetal force and capillary force, and works stably under high temperature conditions with high heat transfer efficiency. Large-diameter high-temperature heat pipe relies on a larger cross-sectional area of the vapor chamber and a larger amount of liquid filling, so that the heat transfer limit of the high-temperature heat pipe is greatly improved, increasing the heat transfer capacity of the heat pipe under the same length. In this paper, the design optimization and development test of large diameter high temperature heat pipe is carried out, the heat pipe shell material is stainless steel 310S, the length is 2 m, the diameter is 89 mm, and the wall thickness is 5 mm. The heat transfer medium is sodium and potassium. The filling of the heat pipe is carried out in the glove box using the hot state filling method to ensure the purity of the work material. The manufactured heat pipe is placed vertically for performance testing, utilizing electromagnetic induction to heat the evaporator section, and the condenser section is cooled under air-cooling conditions. It is found that the wick has a greater impact on the heat transfer performance of the heat pipe. Heat pipe #1 is a thermal siphon without the wick, the maximum heating power is 10 kW, due to the intermittent boiling of the large liquid pool at the bottom of the test process led to the existence of a violent vibration. The selection of the wick also has a great impact on the heat transfer performance of the heat pipe. Heat pipe #2 and heat pipe #3 are capillary core heat pipes with different mesh numbers of stainless steel wire mesh as the wick. The maximum electromagnetic power of heat pipe #2 is 12.3 kW, the cooling power is 3.72 kW, the maximum temperature of heat pipe is 903 °C, and there is local high temperature in the evaporator section. The maximum electromagnetic power of the heat pipe #3 is 17 kW, the cooling power is 6.11 kW, the maximum temperature is 826.1 °C, the heat transfer power is higher, and the overall temperature of the heat pipe is flattened, and no overheating phenomenon occurs. Compared with the small diameter heat pipes, the temperature fluctuation during the startup and steady state of the large diameter heat pipes is greater, and the temperature fluctuation in the condenser section reaches ± 10 °C, and the entrainment phenomenon is more obvious. However, relying on a larger cross-sectional area of the vapor chamber and a larger amount of liquid filling, the heat transfer power of large diameter heat pipe is higher.
Coalescence-Induced Irregular Particle-Droplet Removal
Authors: Li Yanzhi , Du Jiayu , Min Qi , Sun Libin 
Published: November 2024

Abstract
In this work, we experimentally investigated the coalescence-induced irregular particle-droplet removal phenomenon. We chose graphite and iron particles, common contaminant dust in HTR-PM, with large density and irregular (non-spherical) shapes at the sub-millimeter scale. It is found that the coalescence mechanism achieves excellent performance in droplet jumping enhancement and particle removal. The maximum transport height of graphite particles was 0.55mm, and the maximum transport height of iron particles was 0.40mm. It is revealed that since the particles always sunk at the bottom of the droplet, their maximum transport height was related to the droplet departure velocity. Besides, irregular particles enhanced the droplet departure velocity. The droplet departure velocity with all graphite particles and part of iron particles was higher than that without particles. The maximum dimensionless departure velocity is 0.35 (graphite) and 0.29 (iron), 40% and 16% higher than droplet jumping with no particles. Notably, it is illustrated that the mechanism of droplet jumping enhancement with particles was similar to that with macrostructures, which reduced the viscous dissipation and improved the velocity consistency of droplets by shortening the duration of the merge stage and prolonging the duration of the contact stage. This study provided an efficient method for irregular particle removal and opened up a new possible research direction for dust removal and aerosol removal in nuclear industry applications.
High-Fidelity Depletion-Pebble-Flow Coupling Simulation of Pebble Bed Reactor HTR-PM
Authors: Li Ruihan , Liang Jingang , She Ding , Zhang Weijian 
Published: November 2024

Abstract
The high-fidelity depletion and pebble-flow coupling simulation of pebble bed reactor is realized based on Monte Carlo (MC) and Discrete Element Method (DEM). An HTR-PM full scale model with 420,000 fuel pebbles is built using the open-source MC code OpenMC, and the pebble trails are obtained with the open-source DEM code LAMMPS. The OpenMC code is modified so that the model geometry can change from one depletion step to another, and thus can be coupled with pebble-flow calculation. To obtain the high-fidelity results, each fuel pebble is specified to be a single burnup-zone. The pebble-wise power distribution and nuclide composition results are obtained and agree with the physical expectations. The model is depleted for 39 steps with each step representing 2 days. The time-consumption is 2.5 hours per step with a neutron history of 50,000,000 and a parallel scale of 40 MPI processes and 28 OpenMP threads per process. The memory-consumption is 40 GB per process.
Research on Long Term Trend Prediction of Nuclear Power Plants Based on Integrated Framework
Authors: Tan Canyi , Liang Biao , Wang Bo , Li Jiangkuan , Han Rui , Tan Sichao , Tian Ruifeng 
Published: November 2024

Abstract
In the current field of nuclear energy, the development of the third generations nuclear power plants (NPPs) are the mainstream, and its safety and economy are the important direction of nuclear energy development. Therefore, the normal, stable and safe operation of the nuclear reactor system is particularly important. In order to improve the safety and stability of the nuclear power system, this paper proposes a life prediction model of a NPP based on a comprehensive framework. In this study, an integrated framework based on BP, LSTM, Informer and other models is built, and the model is verified by the running data of Qinshan simulator. The results show that, compared with the traditional BP and LSTM, the long-term trend prediction model of NPP based on integrated framework has lower MAE and MSE values, indicating that the long-term trend prediction model of NPP based on integrated framework has lower error, which provides strong support for the safe and stable operation of nuclear power system.
Reconstruction of 3D Nuclear Graphite Based on Generative Adversarial Neural Networks With Gradient Penalty
Authors: Peng Lei , Liu Zhiyong , Wang Pengyu , Liu Hao , Zhang Huang , Yin Huaqiang , He Xuedong , Ma Tao 
Published: November 2024

Abstract
Nuclear graphite is an important structural material in reactor, and it is important to study the dehumidification method to improve the economy of reactor. In order to accurately obtain various physical property parameters of nuclear graphite from the pore scale, constructing a 3D digital graphite that can accurately characterizes the pore structure of micropores and nanopores is crucial. Due to the high experimental cost or a long reconstruction time of tradigital reconstruction methods, we use the generative adversarial neural networks with gradient penalty (WGAN-GP) to reconstruct 3D digital graphite. Taking binary image of graphite as training sample for model training, the generator model that can generate 3D graphite images is obtained, and then multiple graphite images are obtained. The generated images and original images were compared and analyzed from two-point statistics, pore distribution, porosity and specific surface area. The results show the reconstructed graphites are consistent with the original graphite, which proves the reliability of this method.
Numerical Simulation of Condensation-Induced Water Hammer With Dynamic Valve Control Using a Phase Change Model
Authors: Zhao Danfeng , Zhang Zihao , Chen Xiaoxia , Tu Ruiyang , Zhao Jiaming , Xiong Feng , Chen Zhengyu , Guo Wentao , Wang Shengfei 
Published: November 2024

Abstract
This article focuses on the phenomenon of complex water hammer that occurred during the testing period of the steam pipeline in a nuclear power plant. Computational fluid dynamics (CFD) techniques are utilized to simulate the transient behavior of condensation-induced water hammer (CIWH). An phase change model is implemented to accurately capture the phase transition process and its impact on the system dynamics. A dynamic valve control function is implemented to investigate the impact of valves on CIWH, allowing for a comprehensive analysis of their influence. The simulation results are compared with the actual operating conditions in the steam pipeline for validation. Overall, the numerical simulation approach established in this article effectively reproduced the process of complex water hammer, laying the foundation for subsequent model optimization and triggering mechanism research.
Experimental Study of Rectangular Channel Heat Transfer With Buoyancy-Aided Flow Under Asymmetrically Heating
Authors: Ji Yongan , Guo Zehua , Ding Ming , Sun Zhongning 
Published: November 2024

Abstract
Many nuclear power facilities use natural air circulation as part of their non-energetic safety mechanisms to remove heat. Both forced and spontaneous convection are frequently involved in the flow during this cooling process in vertical channels. The flow is referred to as buoyancy-aided flow when the direction of forced convection and natural convection coincide. The mixed convection heat transfer coefficient is lower than the forced convection heat transfer coefficient, and this type of flow will be in a specific buoyancy range. The supercritical fluid and water as the working medium, four-sided test mode heating, asymmetric heating, and air as the work material of the experimental research are the main topics of current research on buoyancy-aided flow. Thus, the investigation of mixed convection under buoyancy-aided conditions is presented in this study, along with a comparison of the experimental findings with the current correlations and an analysis of the causes of the differences. Lastly, a modified Buoyancy number with an average deviation of +20% is offered as a new correlation. This will serve as a point of reference for the relevant mixed convection heat transfer calculation.
Methodology for Acquiring Reliability Data in Novel Nuclear Energy Systems Using a Non-Homogeneous Poisson Process
Authors: Li Zefeng , Chen Pu , Liu Tao 
Published: November 2024

Abstract
In nuclear power plant maintenance, Reliability-Centered Maintenance (RCM) increasingly focuses on preventing equipment failure. This is particularly important in novel reactor models where mature data for maintenance decisions is limited. Bayesian Networks (BNs) are employed to fill this gap, modeling variables and their dependencies to assess component impact on system failure rates. Innovative approaches like Bayesian updating with non-homogeneous Poisson processes (NHPP) for small samples can modeling failures more accurately. This article utilizes the Markov Chain Monte Carlo (MCMC) method for Bayesian updating of component reliability data and employs BNs to analyze the reliability parameters of first of a kind equipment. These methodologies enhance the safety and efficiency of nuclear power plants by ensuring maintenance decisions are informed by existing and comprehensive reliability data.
Key Technologies Research for the Implementation of In-Service Inspection Strategies Optimizing for the Reactor Coolant System of In-Service Nuclear Power Plants
Authors: Shen Yunhai , Deng Feng , Duan Yongqiang , Yu Xiaoquan , Ye Shuixiang 
Published: November 2024

Abstract
The in-service inspection code for the M310 nuclear power plant in China is RSE-M. According to the standard, a periodic hydrostatic test of reactor coolant system is required every ten years, which reduces the economic efficiency of the plant. Based on current research results, the integrity of the pressure boundary of the reactor coolant system can be verified through system leakage testing after refueling outage combined with appropriate in-service inspection strategies. This paper is based on previous research results and focuses on the application research of periodic hydrostatic test and in-service inspection strategy optimization for a nuclear power plant. Key technical research during the application implementation process has been carried out and completed, including the applicability analysis of ASME standards and related regulations, the analysis of pressure boundary differences between system leakage test and hydrostatic test, and the mechanical analysis of nuclear safety class 1 pipelines, and analysis of the effect of related system equipment. Finally, the application and implementation of the optimization of the in-service inspection strategy for the reactor coolant system of the M310 nuclear power plant was completed.
Direct Numerical Simulation Study on Turbulent Flow and Heat Transfer Behavior of Helium-Xenon Gas Mixture in a Round Pipe
Authors: Zhang Ruini , Deng Jian , Wei Zonglan , Lu Qi , Han Wenbin 
Published: November 2024

Abstract
Helium-Xenon (He-Xe) gas mixture is widely adopted as the working fluid for space gas cooled nuclear reactors with CBC engines. The Prandtl number of He-Xe gas mixture in the preferred mixing-ratio range is approximately at 0.2–0.3, which is significantly lower than air or water. To better reveal the flow and heat transfer mechanism of He-Xe gas mixture, 2 direct numerical simulation (DNS) cases of He-Xe gas mixture flowing through a heated pipe were performed. One is based on the experiment case (RUN696) of Taylor (1988), and another one is set with a higher temperature range. The DNS results are in good agreement with the experiment data. The Taylor Nu heat transfer correlation was evaluated by DNS results. Further analyses of other turbulent statistics distribution, such as Reynolds stress, and turbulent heat flux were conducted. Reynolds-Averaged Navier-Stokes (RANS) simulations with 6 different turbulent models of the same case were also performed. The DNS data was used to evaluate the calculation results of different RANS turbulent models. The result shows that the three k-ε and the linear pressure-strain RSM models perform barely the same, and the SST k-ω model relatively behaves the best. It is also found that the current default turbulent Prandtl number model (Prt = 0.85) deviates largely from the calculated DNS data, especially in the near wall region.
Remote Impact Acoustic Inspection of Structures Using Water Jet Impacts
Authors: Tokuomi Saeko , Mori Kazuya , Ohshima Yasutaka 
Published: November 2024

Abstract
We have developed an advanced method for remotely detecting defects in concrete structures using pulsed water jets, offering significant improvements over traditional continuous flow impact echo testing. Unlike the conventional approach, our method utilizes intermittent, high-pressure water jets, which enable more distant and efficient inspections while minimizing water consumption. Notably, the pulsed nature of the water jets makes the testing process less susceptible to wind interference and broadens the excitation frequency range, potentially enhancing defect detection capabilities. In our experiments, we targeted concrete specimens embedded with artificial cavities to simulate common defects, employing a 4 mm diameter nozzle at 1 MPa pressure to release water pulses for 0.01 seconds. The testing distances extended up to 5 meters, demonstrating the method’s effectiveness in detecting 300 mm diameter cavities from a significant range. Defect identification was facilitated by analyzing the acoustic impact signatures through spectrogram visualization, focusing specifically on frequency bands resonating with the defects’ natural frequencies.
Research Methods and Protective Measures for Internal Flooding Analysis of Nuclear Islands
Authors: Du Guang , Zhao Danfeng , Lin Daping , Chen Xiaoxia 
Published: November 2024

Abstract
Internal flooding of the nuclear island in nuclear power plants is a common mode failure of systems and equipment, and the potential hazards must be considered during design. The protective design for such hazards should ensure the realization of various safety functions of the power plant system. Based on the analysis and research on internal flooding of the nuclear island in third-generation nuclear power units, this article follows the standards and specifications of advanced pressurized water reactors, draws on engineering design experience of nuclear power plants, and studies the analysis methods and protective measures for internal flooding of the nuclear island. Firstly, it determines the background and evaluation criteria for internal flooding protection design in nuclear power plants, compares the recommended flooding hazard evaluation methods in the standards, then analyzes in detail the steps and process of screening internal flooding sources, analyzes the spreading path of water sources and water level calculation methods, and conducts risk analysis on protective equipment within the subarea, so as to determine protective measures. The analysis results are given for the safety objectives of nuclear power plants.
Effect of Different P/D Ratios on the Thermal-Hydraulic Performance of Fuel Assemblies in a Lead-Cooled Fast Reactor
Authors: Zhang Yangguang , Lu Daogang , Cao Qiong , Wang Xiaotian 
Published: November 2024

Abstract
The core is the area of the nuclear reactor where the fission reaction occurs, therefore, it is necessary for a complete thermal-hydraulic study to offer vital research data to assure the safe and stable functioning of the nuclear reactor. Due to the large number of fuel rods in each box and the high average heat flux density on the surface of the fuel rods, the ratio of the pitch to the diameter of the fuel rods (P/D) and the width of the side channels of the fuel assembly have a large influence on the uniformity of the coolant temperature at the outlet of the fuel tank in lead-cooled fast reactors. The thermal and hydraulic calculations of the fast reactor fuel assemblies are carried out in this paper based on the parameters of the fuel assemblies of the small and micro lead-cooled fast reactor SMILE, and the feasibility of the numerical calculation method is verified. Second, three models with P/D of 1.15, 1.20, and 1.24 were created, and sensitivity analysis was performed for the width of each model’s opposite side channel. The findings demonstrate that as P/D increases, the coolant temperature uniformity at the fuel box’s outlet improves significantly. Additionally, when P/D = 1.24 and the side channel’s width is 1.95 mm, the fuel assembly’s outlet temperature is effectively flattened, increasing the thermal safety margin.
Design Calculation of Ultra-High Temperature and High Pressure Helium-Xenon Mixed Gas Radiator
Authors: Xu Yongwang , Duan Minghui , Wei Junhan , Zhao Minfu , Zhang Dongxu , Li Qingyuan 
Published: November 2024

Abstract
Helium-xenon mixed gas has higher heat transfer performance under a certain mixing ratio than pure helium. There are great limitations and uncertainties in applying existing theories and empirical models of helium-xenon mixed gases to ultra-high temperature and high pressure ranges.
Therefore, it is necessary to build a loop and conduct experiments to verify the flow heat transfer characteristics of helium-xenon mixed gas under ultra-high temperature and high pressure conditions. The working conditions requires the radiator to have good heat dissipation efficiency under different pressures, temperatures and flows. Therefore, in this loop design, the radiator is a key device. So, it is necessary to perform thermal calculations before determining the structure of an ultrahigh temperature and high pressure radiator.
Two radiator structural designs are currently proposed. Method of computational fluid dynamics (CFD) were used to study two radiators with different structures. Calculations are conducted for typical stable operating conditions. The calculation results show that under the same working conditions, the radiator designed with “fin - low temperature helium-xenon mixed gas” has higher heat dissipation efficiency and is an ideal design choice for ultra-high temperature and high pressure radiators. In addition, its performance under different operating conditions is also studied.
Evaluation and Sensitivity Study of Uniform Kernel Method in Atmospheric 41Ar Dispersion Against the Belgian Field Experiment
Authors: Sun Yeshuai , Zhang Yujie , Wang Zhaoyang , Yang Li , Li Xinpeng , Chen Yixue , Dong Xinwen , Fang Sheng 
Published: November 2024

Abstract
The Uniform kernel is an effective kernel density estimator used in the Lagrangian Particle Dispersion Model (LPDM), estimating the dispersion trajectory of radionuclides, and determining radioactive concentration to assess environmental contamination levels. This method significantly decreases computational resources compared to the traditional box counting concentration calculation in LPDM, serving as a fast and accurate tool to ensure effective nuclear emergency response in the early stages. In this paper, the effectiveness of the Uniform kernel is evaluated systematically via the measurements from eight gamma dose rate monitoring sites deployed at the Belgian field experiment by coupling the California Meteorological Model (CALMET) and the Lagrangian Particle Model (LAPMOD). Sensitivity studies are performed on three important parameters, including bandwidths multiplication factors, horizontal grid size, and particle numbers. The results demonstrate satisfactory agreement between simulations and measurements, with acceptable statistical metrics. The sensitivity analysis highlights the significant impact of bandwidths multiplication factors on the Uniform kernel, summarizing appropriate parameters for applying the Uniform kernel in small-scale dispersion for the nuclear power plant scenarios.