bluuebunny/crossref_metadata_2025_split huggingface dataset:
Geometric Uncertainty Analysis of Fuel Assembly Based on Sub-Channel Model
Authors: Kong Deyan , Gao Zhibo , Wu Di , Cheng Jie , Wang Jianjun 
Published: November 2024

Abstract
There are inevitably geometric uncertainties during the machining and installation of fuel rods, such as the pitch and diameter of each fuel rod, which can lead to changes in the thermal and hydraulic characteristics of the reactor. In this study, a method is proposed to simulate geometric uncertainties as realistically as possible for subchannel models. In this method, each rod is disturbed individually, including the diameter of the fuel rod, the offset distance and the offset Angle. At the same time, a 3 × 3 rod bundle channel model of pressurized water reactor (PWR) is established based on the subchannel model, which contains 8 fuel rods and 1 guide tube. Latin hypercube sampling (LHS) method was used to extract samples, and different perturbations were made to 9 rods at the same time, and then the influence of geometric uncertainty was refined. Coolant temperature, and fuel rod center temperature were selected as figures of merit (FOMs) because these parameters are the most concerned in the thermal hydraulic analysis of the reactor. The LHS method was used to generate 5000 groups of parameters, their relative uncertainties were measured, and the upper and lower limits of tolerance interval and the range of tolerance interval were measured by Wilks’ method. The results show that the uncertainty of output parameters is different at different locations. In general, the tolerance range of maximum coolant temperature and maximum fuel rod center temperature is about 0.7 K and 2.2 K.
Performance Analysis and Optimization Scheme Evaluation of Control Rod Drop Collision in Gas-Cooled Microreactor
Authors: Dong Hao , Qi Anzhou , Jiang Xiaochuan 
Published: November 2024

Abstract
In the case of rod drop accident in high temperature gas-cooled reactor, the control rod is completely inserted into the core in a state of approximate free fall by its own gravity, thus realizing reactor shutdown. In this process, the control rod will generate great kinetic energy with the falling process, and it will impact with the lower reflector of the reactor, resulting in the destruction of graphite and control rod structure in the reflector, and it is impossible for graphite components to be repaired within the service life, and the control rod is not easy to be replaced, so it is urgent to study and optimize the collision between the lower end of the control rod and the graphite reflector in the rod falling accident.
At present, there are many kinds of cushioning devices in engineering to alleviate the impact. It is common to fill foam-like materials or honeycomb-like materials between the two interfaces where the impact occurs, so as to absorb the impact load and realize the protection of components. However, due to the particularity of high temperature and long life of nuclear reactors, the above-mentioned solution using filling materials cannot be adopted. In addition, there is a scheme of using crush elements. At present, some scholars have studied the theoretical model and numerical simulation of crush shell. In the field of reactors, the scheme of buffering together with fluid resistance in pressurized water reactors and arranging crush elements on the bottom reflection layer of the core in high temperature gas-cooled reactors is widely used. However, the two schemes are not suitable for the core design of high-temperature gas-cooled reactor with high temperature, and are not easy to take out and replace, so their adaptability is limited.
In this paper, ANSYS/LS-DYNA finite element simulation program is used. Firstly, the optimization effects of the innovative spring-type and cylindrical thin-wall crushing buffer structures are compared, and then the influence of wall thickness on collision and thin-wall deformation is studied with emphasis on the scheme of cylindrical thin-wall crushing buffer mechanism. It is found that both the cylindrical thin-walled tube and the spring have obvious optimization effects on the collision, but the rebound of the cylindrical thin-walled tube scheme is smaller after the collision. The wall thickness of cylindrical thin-walled cylinder has a great influence on the maximum stress and graphite layer stress of the rod, but has a relatively small influence on the rebound of the rod, and there is an optimal optimization scheme.
Further Investigation of Neutron Noise Under the Condition of Fuel Failure Based on Frequency Domain Finite Element Method
Authors: Yuan Baoxin , Lv Huiyi , Zeng Herong , Zheng Jie , Chen Zihan , Huang Huan , Zhang Songbao , Qian Dazhi 
Published: November 2024

Abstract
Usually, some fuel elements are allowed to be damaged under normal operating conditions of the reactor. At present, the design index of fuel element damage rate of pressurized water reactor at home and abroad is 0.25%. In view of fuel element damage accidents, there are two measurement methods for operating units: fuel accident monitoring and primary loop dose monitoring. Fuel accident monitoring depends on the measurement of delayed neutron activity flowing out of the core to determine whether there is fuel damage, and primary loop dose monitoring depends on measuring the activity of fission products and the concentration of inert gas flowing into the primary coolant to determine whether there is fuel damage. It takes a long time for the two measurement methods to discharge the measured fluid from the core outlet to the measuring device, and the limited damage event is also perplexed by the lower limit of dosimeter measurement after being widely diluted by the total flow of the first loop. For the plate fuel core, the axial flow of the delayed neutron precursor after the fuel element is damaged will break the inherent balance of the normal core and cause the transient change of neutron noise. Based on this idea, on the basis of the work of the author’s team reported by ICONE30, this work assumes three different degrees of plate fuel assembly damage to the assembly position of the core, corresponding to three axial flow scenarios of delayed neutron precursor. Based on the frequency domain finite element method, the neutron dynamics calculations are carried out for the above scenarios, and the fast group and thermal group neutron noise responses of neutron detectors at four different positions in the core are obtained.
Safety Function Based HPR1000 Accident Mitigation Strategy Development for SGTR
Authors: Yang Benlin , Wang Jiansheng , Zhang Yating , Xiong Zhenpeng , Zhao Guanghui , Zheng Junming , Gong Yihong 
Published: November 2024

Abstract
Based on the safety function design of HPR1000, a systematic and structurized process has been provided to show how SGTR accident mitigation strategies are developed. The evolution of SGTR accident progress and variation of main physical parameters are used for the analysis of HPR1000 safety functions which are needed to be performed in the accident, and all safety features performing the corresponding safety functions are then identified for accident mitigation to reach a controlled state or safe sate. All safety features represent the system configuration of HPR1000 safety design, and are the key elements for the development of HPR1000 accident mitigation strategy. According to the operation mode of HPR1000, considering accident progression and logic of the actuation of safety features, the SGTR accident mitigation strategy can be made comprehensively and systematically.
Evaluation of Expansion Reactivity Based on Reactivity Coefficient for Sodium-Cooled Fast Reactor
Authors: Takeda Satoshi , Kitada Takanori , Hoashi Eiji , Okita Takafumi 
Published: November 2024

Abstract
In this paper, formulas for evaluating expansion reactivity based on reactivity coefficients in sodium-cooled fast reactors are derived, and the approximation errors used in the derivation are also evaluated. This research provides formulas for both Z-axis expansion reactivity and radial expansion reactivity. These reactivities are separated into the effect of change in geometric shape and that of change in the cross-section. Since the major approximation in the present derivation is the separation of effects due to geometric shape changes and cross-sectional area changes, the approximation error due to this separation was evaluated using results of neutron transport calculations for fast reactor. This evaluation was performed for two geometries with different ratios of the reactor core’s height to diameter. The results show that the approximation error is less than 20 pcm. Therefore, it can be expected that expansion reactivity can be accurately evaluated using reactivity coefficients.
Hydrogen Absorption Behavior of the R-SUS304ULC/Ta/Zr Dissimilar Metal Joint Under NaOH Immersion
Authors: Ishijima Yasuhiro , Ueno Fumiyoshi , Abe Hitoshi , Igarashi Takahiro 
Published: November 2024

Abstract
In reprocessing plants of Japan, Zr and ultra-low carbon 304 stainless steel (R-SUS304ULC) piping are connected by a joint formed by explosive bonding with Ta sandwiched between R-SUS304ULC and Zr. The joint exhibits excellent corrosion resistance during normal operation. However, Ta causes corrosion and generates hydrogen in NaOH solutions, which are used to decontaminate the equipment in reprocessing plants. This has led to concerns about hydrogen embrittlement of the joints via hydrogen absorption. The authors examined the hydrogen absorption behavior of pure Ta immersed in various NaOH solutions. However, there is a lack of studies on the hydrogen absorption behavior of such joints, and it is difficult to evaluate the amount of hydrogen absorbed by the joints under various decontamination conditions. In this study, we conducted immersion tests of the R-SUS304ULC/Ta/Zr joint in NaOH solution to investigate the effect of immersion environment on hydrogen absorption behavior. Electrochemical measurements were conducted to examine the hydrogen absorption environment. The results showed that the amount of hydrogen absorbed into Ta in the joints decreased compared with that in pure Ta, regardless of the immersion time. The galvanic current measurements of Ta connected R-SUS304ULC or Zr in an NaOH solution indicated that the hydrogen evolution reaction was separated to the R-SUS304ULC and Zr surfaces and that the hydrogen absorption of Ta was suppressed.
Research on the Re-Entrainment Mechanism in Corrugated Plate Driers in the Nuclear Field
Authors: Wang Bo , Li Ru , Li Jiangkuan , Tan Sichao , Tian Ruifeng 
Published: November 2024

Abstract
The corrugated plate dryer is an important steam-water separation equipment in the secondary loop of the mainstream third-generation nuclear power plant in the field of nuclear engineering. The important parameter indicator of the corrugated plate dryer is its separation efficiency. The Re-entrainment phenomenon is a key issue that prevents further improvement of the separation efficiency of the corrugated plate. Its connotation is the phenomenon of secondary liquid droplets being carried by the airflow, leading to a decrease in the efficiency of steam water separation. Starting from the mechanism of Re-entrainment, this article analyzes the fluctuation of the liquid film inside the corrugated plate, the occurrence process of liquid film rupture, and the location and size of secondary liquid droplets generated by liquid film rupture. Based on the establishment of a two-dimensional curve coordinate system at the corner position of the corrugated plate, the force analysis of a thin liquid film with a width of dr and a curvature radius of r in the radial and axial directions is also conducted. Define dimensionless parameters to obtain the order of each item in the liquid film control equation, and then ignore the higher order, ultimately obtaining the liquid film rupture model. Based on the above liquid film rupture model, high-precision three-dimensional calculation of the waveform plate was achieved, thereby achieving accurate calculation of the position of secondary droplet generation. The results indicate that there are obvious nonlinear characteristics in the location and size of the liquid film rupture inside the corrugated plate. The size of secondary droplets is mostly in the millimeter range. The above model provides important technical support for the elimination of Re-entrainment phenomenon and provides important reference for the design of corrugated plate dryers in nuclear power plants and nuclear power plants.
Flow Regime Visualization of Wet Steam Flow in a Horizontal Circular Pipes Having Different Diameter
Authors: Uchiyama Yuta , Morita Ryo 
Published: November 2024

Abstract
To evaluate the occurrence conditions and rate of progression of flow-induced aging degradation phenomena such as pipe wall thinning and vibration fatigue in steam piping systems, it is essential to accurately understand the steam flow conditions inside pipes. However, the flow structure (flow pattern and sound velocity) of water-steam two-phase flow under high steam quality conditions is not sufficiently clear at present, and the occurrence and progression of degradation phenomena may not be properly evaluated. In this study, the pattern of water-steam two-phase flow under high-steam-quality conditions was visually observed to clarify the flow pattern. The database of flow patterns under a wide range of conditions was constructed by expanding the data, especially in the annular dispersed flow region for systems with pipe inner diameters of about 40 mm, and by acquiring new flow visualization data for piping systems with pipe inner diameters larger than 40 mm. It was shown that the existing flow regime maps can be used to determine the flow pattern of wet steam flow through comparison with existing flow regime maps of water-air two-phase flow, thereby improving the reliability of the flow pattern determination method.
Assessment of CFD Methods in Simulating Flowfield in a Vane-Type Separator
Authors: Zeng Xiaobo , Xu Yifan , Fan Guangming , Yan Changqi 
Published: November 2024

Abstract
The vane-type separator is an important device to remove the fission gas from molten fuel in Molten salt reactor. To better understanding its working mechanism and provide a reliable reference for its practical application, numerical study was carried out in this paper. Due to the centrifugal force, the transition of flow pattern happens, leading to the change in interface structure. Before bunching to the gas core, the scale of interface is less than millimeter, while for the scale of interface between gas core and continuous phase is much larger than that. Because of this multi-scale interface phenomenon, it is important to assess the CFD models to obtain a high-fidelity result. Therefore, two multiphase models were compared. One is the Euler-Euler method and another is a hybrid approach of Euler-Euler and Volume of Fraction methods. Besides, Large Eddy Simulation (LES) was used to capture the turbulence information of the flowfield. The result shows that a sharper interface between gas core and liquid is able to simulate using the hybrid method because of the introduction of the transportation equation of volume fraction.
Dynamic Response of Nuclear Fuel Transport Package Under Multiple Impact Angles
Authors: Hao Yuchen , Xu Bin , Zhang Xiaohong , Sun Wenzhao , Li Yulin , Liu Guang , Ma Yan , Li Yue , Lin Musen 
Published: November 2024

Abstract
Nuclear power is one of the most efficient way to provide the clean and green electrical energy. As the uranium in Earth is sufficient to meet global energy needs for several hundred years, it is a sustainable energy source. Meanwhile it can generate power 24 hours a day, unlike renewable energy sources that are rather dependent on weather conditions. To develop the nuclear energy better, the safety performance of nuclear energy cannot be ignored where the importance of safely transporting nuclear fuel has received increased attention. Nuclear fuel includes radioactive material that can pose a significant health threat to human and environment if not transported properly. Therefore, as nuclear transportation packages need to avoid radioactive material leakage in public domain, 9-m drop test is serious but necessary.
Although extensive research has been carried out on the most severe condition under the 9-m drop test, the main challenge of the actual test faced by many reseachers is the impact angle which is directly influenced by the weather and operating conditions. For example, the wind cannot be ignored in the actual because it can affect the impact angle which make it more difficult to predict the orientation of the nuclear fuel transport package. Therefore, this study aims to compare the dynamic responses under multiple impact angles where the typical nuclear transport package is chosen as the research objective. First the geometry model of the nuclear fuel transport package is introduced. The possible impact angles caused by the accidental factors are discussed and then used as the boundary condition in the 9-m drop analysis model. The finite element method (FEM) is used to solve the drop analysis model and then obtain the dynamic responses of nuclear fuel transport package via explicit non-linear software. Finally, the analysis and comparison based on the numerical result is carried out. It is hoped that this research will contribute to a deeper understanding of the influence of impact angle caused by the accidental factors in the actual test. Furthermore, this research approach can be applied in the design and validation of nuclear fuel transport package, which could make the package more robust and reliable.
Advanced Waste Monitoring: Field Trials of a Wireless Sensor Network for Radioactive Waste Package Integrity Assessment
Authors: Chierici Andrea , Ciolini Riccardo , Lo Frano Rosa , d’Errico Francesco 
Published: November 2024

Abstract
Radioactive waste materials, produced from nuclear technologies and power generation, must be safely contained and confined into suitable packaging systems that align with international global standards for disposal. One of the most important aspects related to package disposal is the possibility to monitor and control the physical and thermo-mechanical properties of the overall waste form over time. With recent technological advancements, there is now the capability to deploy flexible and distributed wireless sensor networks for long-term monitoring, diminishing reliance on intricate cabling infrastructure or mechanical scanning systems. This approach can substantially reduce inconsistencies and human errors that typically arise during the information management of packages throughout the disposal process, and integrating these technologies with decommissioning machinery, such as conveyor belts, robotic arms, and specific demolition equipment, may provide real-time radiation data during all stages of the decommissioning process. This study focused on validating and testing a novel system for identifying and continuously monitoring the integrity of stored radioactive waste packages developed at the University of Pisa. It is based on a network of wireless node devices equipped with solid-state, micro-power radiation detectors. Preliminary field trials were conducted at the Nucleco storage site in Rome, demonstrating the system’s efficiency in providing detailed insights into the activity distribution within cemented overpacks and the system’s ability to transmit data correctly even in harsh environmental conditions. Further, a joint demonstration test at ÚJV Řež, in line with the objectives of the H2020 European Project PREDIS, was conducted to emphasize the system’s ability to offer a comprehensive waste monitoring solution. The system’s novel ability to infer package deterioration and degradation based on activity data in a completely automated way sets it apart from existing methods, marking a significant advancement in destructive and non-destructive testing techniques.
Investigations on the Debris Bed Formation Behavior of the Heated Solid Particles
Authors: Zou Wenbin , Liu Zhaoran , Tong Lili , Cao Xuewu 
Published: November 2024

Abstract
During the later stages of the postulated Core Disruptive Accidents (CDAs) of a sodium-cooled fast reactor, the porous debris bed is expected to form in the lower plenum region. The geometries of the debris beds are significant for the coolability and the risk of neutronic subcriticality. In this paper, the sedimentation and formation behavior at different factors of the debris bed is experimentally investigated; the mound dimple diameter and mound dimple depth are chosen as the critical quantitative parameters, which can express the flatness of the debris bed intuitively. The results show that the particles formed a concave debris bed at normal temperature. The general tendency of developing a relatively large dimple area is observed with larger nozzle size, lower particle diameter, lower particle density, and higher water level. The prediction model of the mound dimple volume of the debris bed is established, and the error is less than 30%. When the particles are in the film boiling regime, the vapor film reduces the friction between the particle and fluid, thereby increasing the impact momentum and driving the debris bed in a flat direction.
Hydrodynamics of High-Density Particles in a Multistage Fluidized Bed With Downcomers
Authors: Zhang Yiming , Yang Ying , Wang Huan , Li Shuyue , Zhang Yongmin , Wu Xiuhua 
Published: November 2024

Abstract
Fluidized bed reactor is commonly used in uranium recycling processes of nuclear industry, particles used in these reactors usually have very high densities, which are usually several times higher than those of catalyst particles in chemical industry or fuel particles in fluidized bed boiler in power plants. Multistage fluidized bed reactor has advantages of better residence time distribution and higher gas-solids contact efficiency. However, less literature has involved the operation of multistage fluidized beds with high-density particles. In this study, we used a super-high-density powder of wolfram carbide (WC) powder, which had a true density of 15630 kg/m3 and a typical light powder of similar size, FCC particles, to investigate systematically the different hydrodynamics in a cold-model counter-current three-stage fluidized bed with downcomers. The column diameter, stage height, and downcomer diameter were 186mm, 400mm and 30mm, respectively. The ranges of superficial gas velocity and volumetric solids feeding rate were 0–0.2 m/s and 4–8 ml/s. Pressure drop, pressure fluctuation and solids flow rate at the bottom outlet were measured during stable operation status to evaluate the operating stability. The stable operating ranges were also determined in experiment. The experimental results indicated that this high-density WC powder could be fluidized and circulated continuously in this multistage fluidized bed, but lower fluidization quality was observed using WC than FCC. The minimum operating gas velocities of WC were higher than those of FCC particles, which was determined by the change of transient pressure signals. Different flow phenomena were observed at the minimum operating velocities using the two powders of different densities. The experimental results demonstrated the feasibility of operating uranium compound powders in multistage fluidized beds.