bluuebunny/crossref_metadata_2025_split huggingface dataset:
Experimental Study on Measurement of Boiling Entrainment Rate From Falling Liquid Film With Gas Sheared Flow
Authors: Permana Raka Firman Baskara , Tabuchi Junpei , Okawa Tomio , Narushima Yuki , Furuichi Hajime , Katono Kenichi 
Published: November 2022

Abstract
Study on the droplet entrainment in annular two-phase flow has attracted many researchers in connection with the liquid film dryout in the annular two-phase flow. However, the knowledge on the entrainment due to nucleate boiling is still insufficient to predict the droplet entrainment rate accurately. In this study, measurements were made of the droplet entrainment rate with falling liquid films on the locally heated test section. The visualization using a high-speed camera to clarify the accuracy of the measurement results also carried out in this experiment. For the first step in this study, the measurement of the droplet entrainment rate was performed after getting a stable or steady state conditions. Then, we investigated the influence of the separator position on the reproducibility of the droplet entrainment rate to get the droplet entrainment rate – separator position map. As a result, the reproducibility of the droplet entrainment rate increases with decreasing separation distance from the liquid film. In additions, we also investigated the influence of the heat flux and liquid film flow rate on the reproducibility of the droplet entrainment rate. As a result, the reproducibility of the droplet entrainment rate increases with increasing the heat flux and liquid film flow rate. In the future, we expect to investigate the influence of gas flow rate, and degree of subcooling on the reproducibility of the droplet entrainment rate and some correlation models are developed.
Optimization of MOX Fuel Assembly Based on Genetic Algorithm for Discussion on Plutonium Consumption
Authors: Yamamoto Yuma , Kitada Takanori , Takeda Satoshi , Takeishi Taichi 
Published: November 2022

Abstract
Japan has been promoting its Pu utilization program based on the principle of not having excess Pu from the viewpoint of non-proliferation of nuclear weapons. On the other hand, by the end of 2020, Japan had 46.1 tons of separated plutonium, and the Rokkasho reprocessing plant, which will extract separated plutonium from spent nuclear fuel, is scheduled to start operation. Currently, there is a lack of ways for Pu consumption, and the practical way of Pu consumption is the use of MOX fuel in light water reactors. However, the conventional nuclear design of MOX fuel assemblies of light water reactors is mainly determined by focusing on the local peaking factor, and the Pu consumption has not been considered much. In this study, we focused on the Pu consumption in the MOX fuel assembly and obtained the fuel pin arrangement that maximizes the Pu consumption by changing the fuel pin arrangement and Pu composition using a genetic algorithm. Focusing only on the local peaking factor, the optimization was performed to obtain the fuel pin arrangement that minimizes the local peaking factor. As a result, it was found that fuel pins with high enrichment were placed in the fuel assembly where Pu consumption was maximized, except for the fuel pins adjacent to the uranium fuel where the thermal neutron influx has a large effect. In the arrangement that minimized the local peaking factor, the local peaking factor is suppressed to 1.1. From these results, it was confirmed that the optimization can increase the amount of Pu consumption.
Molten Salt Reactor Experiment Simulation Using ThorMOC
Authors: Zhang Ao , Dai Ming , Cheng Maosong , Chen Jingen 
Published: November 2022

Abstract
High fidelity modeling and simulation of nuclear reactors with direct whole core transport calculation has become an important state of art for development of reactor computational tools with the rapidly improvement of computing power. In this work, a quasi-2D delayed neutron precursor (DNP) drift model has been implemented in a GPU-based whole core transport code ThorMOC to achieve high fidelity modeling and simulation of liquid-fueled molten salt reactors (MSRs). A rasterized coarse mesh finite difference (RCMFD) method is applied in ThorMOC to improve the convergence rate, and to handle complex MSR models with arbitrary boundaries by combining with long characteristics method. OpenMC is used to generate a multi-group cross section library and to provide reference results for stationary state calculations. The results including eigenvalue and power distributions from ThorMOC agree well with those from OpenMC for stationary state calculations of the Molten Salt Reactor Experiment (MSRE) model. The reactivity loss due to DNP drift is computed and is in a good agreement with the measured value in experiment.
Experience Feedback Report on Hydrogen Meter Discontinuity of Steam Generator Overflow Tube
Authors: Liu-Huan Zhang , Xiang Wu , Dong-Fang Ouyang 
Published: November 2022

Abstract
The hydrogen meter sampling branch of 1st ring steam generator overflow pipe of CEFR secondary loop from 2016 to March 2020 has no access to sodium flow. The unavailability affects the way judges the small leakage in the upper part of the steam generator. After investigation, in March 2020, the sampling branch restored sodium flow, and the hydrogen meter returned to normal. In this paper, the cause analysis, troubleshooting process and solutions of hydrogen meter sampling branch are introduced, and the operation and maintenance experience of diffused hydrogen meter system is summarized, providing reference for the design, optimization, maintenance and personnel training of diffusion hydrogen meter in China experimental fast reactor (CEFR) and demonstration fast reactor.
Mechanical Performance Assessment of Independently-Developed Zirconium Alloy Under RIA Condition
Authors: Wang Jie , Wang Yang , Wei Jun , Deng Yongjun 
Published: November 2022

Abstract
Reactivity-initiated accident (RIA) is postulated design-basis accidents (DBAs) in light-water reactor (LWR). Moreover, Pellet-cladding mechanical interaction (PCMI) can cause a failure of the cladding in the early transient. In recent RIA regulatory guide released by NRC which points out that PCMI failure threshold depends on total hydrogen content in cladding during RIA. In order to evaluate the performance of independently-developed zirconium alloy under RIA conditions, the mechanical behavior and the fracture of 2 different cladding tubes (CZ, and SR (Stress Relief) Zr-4) with different hydrogen contents are investigated under thermal-mechanical loading conditions representative of PCMI during RIAs. Ring tensile tests are performed at room temperature, 350 °C on 2 different materials containing various hydrogen concentrations up to 1000 wt. ppm. Test results indicate that the ductility of the material decreases with increasing hydrogen content at room temperature due to damage nucleation by hydride cracking, the ductility and strength results of SR Zr-4 have a good agreement with reference paper, confirming the rationality of experimental method applied and reliability of test facilities. According to the results of Zr-4 and CZ, a conclusion can be made is that the ductility of independently-developed zirconium alloy (CZ) is better than Zr-4, which can provide the technical support when licensing.
Comprehensive Comparison and Analysis for Fast Reactors
Authors: Yan Zheng , Zhao Weiguang , Cao Xiaxin , Yan Changqi 
Published: November 2022

Abstract
Fast reactor is the main reactor type of the fourth generation reactors. It has the advantages of good fuel breeding characteristics, strong transmutation ability, high thermal efficiency and good safety. The research of fast reactor is the focus of current advanced reactor. According to the characteristics of fast reactors, the differences between fast reactors and pressurized water reactors in the fuel breeding, transmutation and safety are analyzed in this paper. On this basis, the characteristics of sodium-cooled fast reactors, lead and lead-bismuth eutectic cooled fast reactors and gas-cooled fast reactors are analyzed and their feature differences are compared. According to the basic characteristics of reactor physics, thermal-hydraulics and safety, the characteristics of these fast reactors are analyzed and compared, their respective advantages are given, the technical problems and difficulties are pointed out. The advantages and suitable application fields of these fast reactors are illustrated, the development opportunities and technical challenges are discussed. Some analysis results can provide reference for the research, design and application of advanced nuclear reactors.
Design Verification of Minimum Flow Valve of the Main Feed Water Pump Based on the Full-Range Simulation Platform of CPR 1000
Authors: Zhang Zhenhua , Zhang Bo , Zheng Chaoying 
Published: November 2022

Abstract
When the flow of the main feed water pump of CPR1000 unit is low, that is, when the unit is operating at low load, in order to protect the main feed water pump from damage, two low-flow protection recirculation pipelines are installed at the pressure stage pump outlet, and the two lines are installed with two minimum flow regulating valves. These two valves generally have two designs: proportional type and on-off type. The paper simulates the response of the minimum flow valve of main feed pump to water level regulation of steam generator at low load based on the full-range simulation platform of CPR1000 nuclear power unit. By analyzing the response curve of the water level of the steam generator, the advantages and disadvantages of the different valve designs are verified separately. In the proportional design, the valve position of the minimum flow valve is easy to oscillate with the fluctuation of the flow signal, which will affect the stability of the steam generator feed water control system and then affect the operation of the water level regulation system of the steam generator. At the same time, the proportional minimum flow control valve has the control dead zone. Under the transient conditions with frequent disturbance, the response speed of the minimum flow valve does not match the response of the water level adjustment of the steam generator, resulting in the unstable water level of the steam generator. In the on-off design, the stability of water level regulation will have certain advantages. After comparing the influence of the two valve designs on the water level regulation of the steam generator and analyzing the limitation of the proportional design and the importance of minimum flow valve, two improvements schemes are given for the design of the minimum continuous regulating flow valve.
Seismic Isolation Scheme Design of 200MW Nuclear Heating Reactor
Authors: Hou Nannan , Dong Jianling 
Published: November 2022

Abstract
Air pollution and the energy problem have become concerned and urgently needed to be solved all over the world in recent years. Coal-fired boilers in industrial parks are an important source of air pollutants. Low temperature heating reactor built in industrial zone is an ideal substitute for coal-fired boiler because of its excellent inherent safety and cleanliness. The new type nuclear heating reactor named NHR200-II was proposed by Institute of Nuclear and New Energy Technology of Tsinghua University in 2015[1]. Compared with the previous generation NHR200, the thermal parameters of NHR200-II are improved which can not only be used for regional heating but also realize the multi-purpose like power supply, steam supply and seawater desalination. And the economic benefits are also greatly improved. The design of NHR200-II is based on the design, construction and operation experience of NHR5 which has maintained running safely up to now and NHR200. The main technical features include three circuit design, integrated layout of primary circuit, full power natural circulation, and automatic self-pressurization.
NHR200-II uses the traditional anti-seismic technology at present. However, the requirements of seismic performance of reactors have generally increased after Fukushima disaster. This paper tries to apply seismic isolation in NHR200-II to meet the improved requirements of seismic performance and realize the standardized design by which the magnitude of design ground motion is not restricted in the future. Base isolation, which means setting isolation devices like elastomeric bearings or spherical sliding bearings between the foundation and the superstructure, has been widely applied in normal buildings and bridges, but still needs particular design for nuclear reactors.
This paper describes NHR200-II seismic isolation system, which consists of the foundation, raft, isolation layer and the whole nuclear island as superstructure. In addition, the paper also tells the design process of the system, including designing of the raft, selecting of isolators and their layout design based on the environment and mechanical property, etc. And the particular calculation method of the performance parameters and acceptance criteria are also shown in this paper. Finally, the overall design method of seismic isolation is given as a conclusion.
Automation of High-Fidelity Modeling of PWR Fuel Rods and Verification
Authors: Chunyu Zhang , Yang Wang , Yongjun Deng , Zifan Song , Yong Lu 
Published: November 2022

Abstract
Definition of a high-fidelity finite element model of a fuel rod is complex and error-prone. In order to make the models affordable for the purpose of design and analysis, this study developed an automated procedure for generation of the finite element models of fuel rods within the general-purpose finite element modeling environment, i.e., ABAQUS/CAE, by using the parametric modeling technique with embedding the best modeling strategy that balances both efficiency and accuracy. The generated models are then verified against typical experimental measurements. Based on the results, several physics models, such as the thermal conductivity model and the fission gas release model, are evaluated. It was shown the automated modeling program can build a wide range of rods within several minutes. Important parameters such as the centerline temperature, the burnup distribution, the elongation of the fuel active zone and the diameter change of the cladding can be predicted with high accuracy under real power histories. The fission gas release can be qualitatively analyzed as well. A combination of the automated modeling program and the fuel performance analysis program therefore allows rapid high-fidelity modeling of real fuel rods and quantitative evaluation of their performance, which provides an important tool for assisting the design of advanced fuels.
Integral Simulation on the Thermo-Mechanical Behavior of the UMo/Al Dispersion Fuel Plate Using Finite Element Method
Authors: Yang Guangliang , Feng Wenpei , Ding Tao , Chen Hongli 
Published: November 2022

Abstract
Fuel plate performance under irradiation is an important topic for the research reactor. The fuel plate is expected to deform after a long-time operation, which is a threat to the reactor safety. In this research, irradiation behaviors like interaction layer growth, swelling, creep and plasticity of the UMo/Al dispersion fuel plate are carefully reviewed from literatures. Taking these behaviors into consideration, the longtime three-dimensional stress update algorithms for the fuel meat and cladding are developed respectively based on the large-deformation incremental constitutive relation. Since local properties in the fuel meat change with burnup, equivalent material properties models are used to take account of this characteristic. These models and algorithms are implemented into Abaqus user subroutines to simulate the thermo-mechanical performance of the fuel plate under the typical operating condition in research reactors. In order to illustrate the macroscopic deformation of the fuel plate and reduce the computational cost, 1/4 symmetrical geometry of the fuel plate is employed in the simulation. The evaluation and distribution of temperature, stress and deformation of the fuel plate are obtained and analyzed. Results show that the growth of interaction layer affects the temperature profiles heavily. The stress concentration mainly locates at the contact interface between fuel meat and cladding, especially at the side of the interface.
Modeling and Analysis of Tritium Fuel Cycle for Magnetic Confinement Fusion Reactor
Authors: Xu Jinggang , Zeng Qin , Wang Sheng , Wang Huaihuai , Zhou Junjie , Xiong Yangbin , Yang Jingchen 
Published: November 2022

Abstract
The realization of tritium self-sufficiency has become an important step to the study of tritium cycle for fusion reactor. To realize tritium self-sufficiency of tritium cycle system of fusion reactor, a typical tritium cycle model of magnetically confined fusion reactor is established on the basis of CFETR in this paper. Tritium cycle model and tritium self-sufficiency are also studied. Based on the mean residence time method, the mass balance equations of the model are developed. Considering the tritium retention of the feeding system and plasma, the specific scheme of the model for calculating the minimum initial startup tritium inventory and the minimum tritium breeding ratio required for tritium self-sufficiency is given. The tritium inventory and the tritium treatment capacity requirements of each subsystem are obtained. In addition, the tritium inventory is also obtained through Simulink modeling and analysis. This is also a method to solve the problem, which can be used as a reference for the modeling and analysis of fusion reactor.
Evaluation of Effect of Depressurization Timing on the Thermal Load To Primary Containment Vessel During Severe Accident at BWR
Authors: Matsumoto Taisei , Yamaji Akifumi , Fujiwara Daisuke 
Published: November 2022

Abstract
According to accident of a Boiling Water Reactor (BWR), the operators generally do not depressurize the Reactor Pressure Vessel (RPV) until low-pressure core injection systems are recovered. However, when the core water level decreases below a certain limit, the RPV is depressurized to prevent the RPV failure at high pressure, which may lead to failure of the Primary Containment Vessel (PCV) by direct heating of the ejected melt. In such a case, if recovery of the core cooling cannot be expected, earlier depressurization may be preferable from the viewpoint of reducing thermal load on PCV, because it may effectively reduce the amount of fission product (FP) aerosol deposition on the RPV dome and reduce the risk of PCV top head heating via convective and radiative heat transfers from the RPV dome. However, significance of such risk is not evident as the steam dryer and separator may hinder direct aerosol deposition on the RPV dome and the RPV dome is thermally insulated. Moreover, the narrow space between the RPV dome and the PCV top head may hinder convective heat transfer. In this study, sensitivity analyses have been carried out using MELCOR-2.2 to support such understanding.
Experimental and Numerical Research on Droplet Drag Coefficient and Deformation Coefficient in Nuclear Power Plants
Authors: Li Ru , Tian Ruifeng , Chen Bowen , Wang Bo 
Published: November 2022

Abstract
There are many multiphase flow phenomena in steam generators (SG) of nuclear power plants (NPPs) and the movement of droplets affects the separation efficiency of dryers in SG. And the most important factor affecting droplet movement is drag force on droplet. Drag force coefficient (DRC) and deformation coefficient (DEC) as characterization coefficient of drag force have important research significance. In this paper, water and silicone oil are used as continuous phase and dispersed phase, respectively. Study on variation of the DRC when Re is in the range of 30–1200 is carried out. Firstly, the motion of droplets (MDs) of different sizes is visually studied by using the high-speed camera and images of MD is acquired. The droplet contour is recognized based on principle of boundary differentiation and the centroid coordinates of the droplet are determined, thus DRC and DEC are obtained. Besides, the relationship between DRC and DEC are researched. Additionally, Lattice Boltzmann Method (LBM) is used to simulate the droplet motion. The simulation results are compared with the experimental results, thus verifying the feasibility of LBM method to simulate the MD. The results show that when velocity of droplet (VD) is low, DRC is inversely proportional to VD. While VD is high, DRC is constant. When the droplet diameter (DD) is small, the final VD is proportional to DD, and when DD is large, the final VD is proportional to the square of DD. The DEC is linearly related to We. The larger the We, the smaller the DEC. Shan-Chen model of LBM is feasible for droplets with low Reynolds number (Re), while the simulation of droplets with high Re is the future prospect.
High Efficient Mineralization of Cesium in Waste Liquid by Hydrothermal Method
Authors: Gao Kai , Feng Wendong , Yan Xiaojun , Guo Xiaobin , Cui Xiaoyan , Guo Xiliang 
Published: November 2022

Abstract
During the nuclear fuel cycle, a large amount of radioactive cesium is produced, and at the same time, it is difficult to safely dispose of because of its extremely soluble in water and low boiling point. The safe disposal of radioactive cesium is a problem that needs to be solved urgently. The currently used methods to treat cesium-containing wastewater such as cement fixation, glass solidification and ion exchange, have certain drawbacks, and urgently need to be further optimized. Pollucite is an analcite molecular sieve. Many studies have shown that it is one of the potential final choices for radioactive cesium. Pollucite has an appropriate pore size, without destroying the structure of pollucite, radioactive cesium cannot diffuse out of pore of pollucite, pollucite also has a good stability and high cesium loading rate. In previous studies, pollucite needs to be synthesized under high temperature conditions above 1000°C. This method will cause the volatilization of cesium and increase the difficulty of exhaust gas treatment. How to quickly synthesize pollucite at a lower temperature has become an important research direction for the safe disposal of caesium. Different from the traditional calcination method to synthesize pollucite, this paper uses hydrothermal method to simulate the generation environment of zeolite in nature, and realizes the synthesis of pollucite at a lower temperature. Using cesium contained in simulated wastewater as synthetic raw materials, combined with commonly used silicon sources and aluminum sources as synthetic raw material, through exploring different reaction conditions such as reaction temperature, reaction time, and alkali addition, the hydrothermal synthesis of pollucite was finally successfully realized. Then the influence of different reaction conditions on synthesis and cesium removal efficiency was investigated, and its growth mechanism was also analyzed. Finally, it was realized that pollucite was successfully synthesized under 150°C hydrothermal conditions using common chemical reagents as raw materials, which further promoted pollucite application in the safe disposal of radioactive cesium.