bluuebunny/crossref_metadata_2025_split huggingface dataset:
Preparation of SiC Composite Coating on the Surface of Graphite Spheres in VHTR
Authors: Yang Hui , Zhao Hongsheng , Zhang Kaihong , Cheng Xing , Li Ziqiang , Wang Yue , Liu Bing 
Published: November 2024

Abstract
Due to various characteristics such as high thermal efficiency, inherent safety, passive safety, and hydrogen production for industrial use, the high temperature gas-cooled reactor (HTR) is considered as one of the most promising technologies in the future. Fuel elements are the core part of a reactor. In the existing high-temperature gas cooled reactor HTR, spherical fuel elements are involved, with graphite accounting for more than 90% of the mass and volume of the fuel element. However, graphite is prone to oxidation in oxidizing atmospheres at a higher temperature than 400°C. Regarding the design of future HTR, accident conditions must be taken into consideration. Under the accident conditions of inlet and outlet water, the graphite in the spherical fuel element is easy to be oxidized. This causes damage to the fuel elements and increases the release of fission products, which threatens the safety of HTR.
In this study, two-step pack cementation is performed to prepare SiC composite coatings on the surface of graphite spheres with a diameter of 40 mm, which improves the oxidation resistance of graphite spheres. According to research, the thickness of coating is about 150 μm after two-step pack cementation, with SiC and Si as the main phase of the coating. However, it is necessary to improve the density of SiC composite coatings needs. After 7 h of oxidation in static air at 1500°C, the mass loss reaches 22.22%. To a certain extent, the compactness of the coating can be improved by the slurry impregnation and Si sol gel method.
Research on Optimization Method of Personnel Evacuation Path Based on Evacuation Time and Radiation Risk
Authors: Zhang Jiemin , Yan Jin , Gao Jiaxuan , Cao Qun , Wu Nan , Xue Na 
Published: November 2024

Abstract
In a radiation scenario, the non-emergency personnel in the site are the closest to the accident site, and the group most likely to be exposed to radiation. The evacuation preparation and evacuation plan for personnel in the nuclear power plant should be fully studied to ensure the evacuation efficiency under accident conditions and reduce the risk of irradiation. For personnel at different initial positions, the evacuation time and possible radiation dose during the evacuation process should be comprehensively considered, and the evacuation path should be reasonably planned to reduce the radiation risk of personnel during the evacuation process. This article uses an ant colony algorithm to optimize evacuation time and radiation risk to calculate and analyze the evacuation path of personnel. The results indicate that a comprehensive consideration of evacuation time and radiation risk in evacuation path planning can achieve the goal of balancing evacuation time and radiation dose, and can provide technical support for decision-making in nuclear emergency protection actions.
Transient Identification of HTR-PM Abnormal Feedwater Flow Based on Graph Neural Networks
Authors: Zhang Wenji , Zhang Tianhao , Li Jitao , Li Duo , Guo Chao , Huang Xiaojin 
Published: November 2024

Abstract
The safety requirements for nuclear power plants are very strict, as an accident may lead to radioactive leakage, causing unbearable losses. Therefore, safety related research in nuclear power plants is very important. High Temperature Reactor-Pebblebed Modules (HTR-PM) are a fourth generation advanced reactor with inherent safety features. HTR-PM adopts modular reactors, using multiple reactors to drive one turbine for power generation. To achieve a multi-module design, HTR-PM has a fluid network for steam and feedwater. When the feedwater flow rate of a modular reactor changes, the physical quantities in the fluid network are also affected, resulting in a very complex transient of the HTR-PM at this time.
In recent years, many deep learning methods have been applied to the fault diagnosis of nuclear power plants. These methods cannot consider the physical structural characteristics of nuclear power plants when extracting data features. Therefore, graph neural networks can effectively solve this problem. The graph neural network constructs a topology graph based on the fluid network structure of HTR-PM and combines it with the topology graph when the features propagate with the network layer, fully considering the physical structural characteristics of the nuclear power plant. After experiments, the network structure of LSTM-GCN has achieved excellent results in transient identification of water supply anomalies.
Research on Accident Prediction of Nuclear Power Plants Based on Deep Learning
Authors: Lv Wei , Li Tong , Wang Bo , Tan Sichao , Li Jiangkuan , Tian Ruifeng 
Published: November 2024

Abstract
Based on the bidirectional long short-term memory network (BILSTM) and the long short-term memory network (LSTM), this paper uses the rolling update mechanism (RU) to obtain the time series training set data and test set data of the original data, and predicts the water level and pressure changes of the Pressurizer after the LOCA accident in nuclear power plants, so as to provide a certain reference for the prediction of LOCA accidents in nuclear power plants. Specifically, the following conclusions are obtained: Compared with the LSTM model, the BILSTM model has better prediction results and less prediction loss. Although the average error of the prediction results of the two models is 10−3 orders of magnitude, the average error of BILSTM is reduced by more than 20% compared with LSTM.
Effect of Non-Condensation Gas on Heat Transfer of Direct Contact Condensation
Authors: Li Shu , Huang Wenxing , Zhang Dandi , Tong Lili 
Published: November 2024

Abstract
The small modular pressurized water reactors adopt a safety relief system to provide overpressure protection for the reactor coolant system. During this process, the mixed gas is discharged into the suppression water tank, causing direct contact condensation. The fraction of the non-condensable gas has a significant impact on the heat transfer. A visual experimental device was established to conduct the experiments of direct contact condensation within the mixed gas mass flux in the range of 177∼425 kg/m2s and the non-condensable gas mass fraction in the range of 0∼20% by using type I multi-holes sparger. The jet plume shapes for the pure steam condensation and mixed gas condensation are investigated in the condensation oscillation zone and jet stable condensation zone. The heat transfer area of pure steam and mixed gas condensation is obtained based on the jet plume shape at the interface. An empirical correlation of the heat transfer coefficient of direct contact condensation with a function of a dimensionless steam mass flux and a driving potential for the condensation process and steam mass fraction is developed and verified with different experimental data. The results show that the empirical correlation of the heat transfer coefficient applies to pure steam and mixed gas, and the error of the model is within ±50%.
Verification and Analysis of Thermal-Hydraulics System Program for a Power Plant Model
Authors: Du Lixin , Zhou Peng , Zhang Hao , Yang Yanhua 
Published: November 2024

Abstract
The thermal-hydraulic system analysis program is mainly used for the calculation and analysis of transient and accident processes in nuclear power plants, and it is one of the important tools for the design safety analysis of pressurized water reactor nuclear power plants. This article introduces a new software package that includes a thermal hydraulic system analysis program, Severe accident Calculation and Evaluation software for Nuclear Energy System (SCENES). SCENES is a thermal-hydraulic system analysis program developed using the staggered grid method based on a two-fluid and six-equations model. To verify the reliability of the thermal hydraulic system analysis program in this software package for calculating transient and accident process conditions in a nuclear power plant, this study models the main hydraulic systems and heat transfer components in the primary and secondary circuits of a nuclear power plant, and simulate and analyze its normal operating and accident conditions separately. The results were also compared and analyzed with the RELAP5 calculation results, which were in good agreement. The results indicate that the SCENES system analysis program is reasonable and reliable in calculating the main thermal hydraulic characteristic parameters of nuclear power plants during steady-state operation and certain accident transient processes. Meanwhile, this study lays the foundation for further program development in the future.
Research on Automatic Cable Layout Method for Nuclear Power Plant Based on A* Algorithm
Authors: Du Hailong , Cheng Jie , Wang Jianjun , Gao Puzhen , Deng Xuchen , Tang Kai , Wang Lei , Su Jincheng 
Published: November 2024

Abstract
The design of cable pathways is indispensable components in the layout design of nuclear power plants. In nuclear power plants, cables exhibit characteristics of being numerous in quantity, diverse in types, and having multiple installation standards. At present, the layout design of cable pathways in a typical nuclear power plant has always been carried out relying heavily on engineering experience. Depending solely on manual deployment not only results in low efficiency but may also not yield the most optimal solutions. To mitigate the human factors on the design and expedite the generation of economically superior solutions, we propose an automatic layout method.
In this method, spatial discretization of nuclear power plant civil structures and equipment is achieved using a grid-based approach, and cable path searches are conducted within the grid map using the A* algorithm. Considering the drawbacks of the traditional A* algorithm’s slow pathfinding in large plants and complex environments, we present an improved A* algorithm tailored for cable layout in nuclear power plants. Firstly, according to the characteristic of prioritizing cable installation along the walls, delineate feature lines along the walls throughout the entire plant area. The feature lines intersect with each other, forming a network that connects various rooms. The A* algorithm prioritizes pathfinding along the feature lines during the routing process, significantly reducing the number of traversed nodes. Finally, the cost function of the A* algorithm has been redesigned to better reflect the actual layout costs. In the traditional A* algorithm, the cost function is designed to assess the distance cost from the start node to the goal node. For the cable layout problem, this cost not only includes the actual length of cable routing but also account for other factors such as bend costs, wall penetration costs, etc.
The aforementioned methods have been implemented in the C# programming language and have undergone testing. The results confirm the feasibility of the proposed method.
Typical Fault Diagnosis Model of Nuclear Power Plant Combined With Knowledge Driven and Data Driven
Authors: Wang Xin , Peng Minjun , Wang Hang , Li Zikang 
Published: November 2024

Abstract
In order to prevent potential radioactive leaks and maintain personnel safety, fault diagnosis in nuclear power plants is crucial. Knowledge-driven and data-driven approaches are two predominant directions in nuclear power plant fault diagnosis. Knowledge-driven approach relies on prior professional knowledge, rules and model for state monitoring and diagnosis, offering strong logic and interpretability. However, it often yields poor results in complex systems. On the other hand, data-driven approach utilizes black-box inference to establish the mapping relationship between complex system characteristics and fault modes. However, this approach lacks interpretability in its operation mechanism, and it is deemed unacceptable for high-security industries such as nuclear power. To address these challenges, this paper introduces the concept of graph neural network and proposes an innovative idea of integrating knowledge-driven and data-driven approaches for fault diagnosis in nuclear power systems. The proposed method embeds the fault propagation path into the data-driven inference process by constructing a graph based on feature nodes and fault type nodes. Furthermore, we analyze how different methods of sampling features and aggregating data influence the diagnostic results during the process of inferring fault paths using graphs with corresponding fault labels. We implement and verify our scheme using a comprehensive simulation simulator for nuclear power devices. The results demonstrate that our method effectively carries out fault diagnosis by combining both interpretable graph-based inference with powerful nonlinear fitting capabilities offered by data-driven approaches.
Influence of Thorium Utilization on the Safety of a Small Modular Molten Salt Reactor
Authors: Wu Jianhui , Zou Chunyan , Yu Chenggang , Zhao Hongkai , Cui Yong , Bao Haotian , Cai Xiangzhou , Chen Jingen 
Published: November 2024

Abstract
Molten salt reactors (MSRs) are one type of reactor system that are characterized by their application of liquid fuel and thorium utilization. Up till now, the performance of thorium-uranium fuel cycle has been extensively studied for MSRs, while the influence of thorium utilization on the MSR core safety is rarely investigated. In this study, a systematic research on the impact of thorium utilization on the core safety for a small modular MSR with a thermal power of 500 WM is conducted. The influence of thorium utilization on the decay heat and radioactivity of spent fuel salt is firstly investigated by comparing the performance of cores fueled with low enriched uranium (LEU) and LEU+Th mixed fuel. It reveals that thorium utilization would raise the decay heat and radioactivity of spent fuel salt due to the yield increase of Pa-233, Sr-93, La-142, La-140, La-143 and Cs-141. Then, the influence of thorium utilization on negative temperature feedback and core delayed neutron (DN) fraction is analyzed, and the results demonstrated that thorium addition to the core will weaken the negative temperature coefficient at the beginning of fuel cycle (BOC) and decrease the core DN fraction, but the negative temperature feedback turns to be better than the core fueled with LEU at the end of fuel cycle (EOC). Finally, the influence of thorium utilization on the response of core to six typical transients, including overcooling/overheating of the inlet fuel salt, increase/decrease of the fuel salt flow rate, reactivity insertion, are studied. The results show that thorium utilization would deteriorate the core response to the transient at BOC, while improve the safety of core at EOC. Most importantly, the safety of the core varies less through the fuel cycle than that of the core fueled with LEU, which would benefit for the design of safety control system.
Transport Characteristics of Two-Phase Flow in Rod Bundle Channels With a Spacer Grid
Authors: Ren Jiaxing , Wang Ruohao , Wang Fangdong , Qiao Shouxu , Tan Sichao , Tian Ruifeng , Gao Puzhen 
Published: November 2024

Abstract
The variation of void fraction and interfacial area concentration can affect the heat transfer, mass transfer, and interfacial drags. This study investigates the air-water two-phase flow in 5 × 5 Rod Bundle Channels with a spacer grid. The four-sensor conductivity probe was used to measure the distribution of local two-phase flow parameters (void fraction, interfacial area concentration, bubble Sauter mean diameter, bubble frequency, and bubble velocity) across channels cross-sections at 36.5, 56, 65.2, 77, and 87.6 hydraulic diameters. The experimental data was used to analyze the development characteristics of two-phase flow parameters across the channel cross-section and along the flow direction, the effect of spacer grid and gas-phase superficial velocities, and the one-dimensional transport of the averaged two-phase flow parameters. The results show that the distribution of the local interfacial structure is mainly affected by the combined effects of the radial forces, such as the lift force, the wall lubrication force, and the turbulent dispersion force. The spacer grid mainly affects the distribution characteristics of interfacial parameters through turbulent vortex aggregation and grid shear fracture, with opposite effects; There is a spacer grid influence zone downstream of the spacer grid, and the distribution of interfacial parameters inside it undergoes drastic changes.
Development and Preliminary Validation of the Fine Fuel Cell Subchannel Analysis Code CUNLUN
Authors: Jiang Hongwei , Chen Guangliang , Tian Zhaofei 
Published: November 2024

Abstract
In response to the low resolution caused by the assumption of homogenization in traditional subchannel analysis methods and the high computational resource consumption of computational fluid dynamics (CFD). This article proposes a novel subchannel calculation and analysis method for pressurized water reactor cores that combines the advantages of CFD and subchannel analysis methods, and develops a thermal hydraulic calculation and analysis code CUNLUN for pressurized water reactor cores based on this method using C++language. Preliminary verification was carried out with the numerical simulation results of ANSYS FLUENT, and the calculation results were basically consistent with the experimental data and FLUENT numerical simulation results, indicating that the code can improve the accuracy of predicting local thermal and hydraulic parameters in the subchannel of the pressurized water reactor core, and can predict the flow characteristics of the coolant in the pressurized water reactor core. This code can provide effective design and research methods for thermal hydraulic analysis of pressurized water reactor fuel assemblies.
Prediction of Sensor Data Accuracy in Thermal Experimental Benches Using GRU-GCN Neural Network Model
Authors: Yang Linjun , Li Tong , Liu Yongchao , Wang Bo , Li Jiangkuan , Wen Jiming , Tan Sichao , Tian Ruifeng 
Published: November 2024

Abstract
As a large-scale facility with multi-system coupling, the operation safety of nuclear power plant is of great importance to the environment and human society. Accurate monitoring of the overall operation status of nuclear power plants (NPP) is not only the key to ensure its safe operation, but also an important means to improve energy efficiency and reduce potential risks. In order to solve the problem of simultaneous prediction of multiple operating parameters of nuclear power plant, this study proposes a deep learning model combining Graph Convolution Network (GCN) and Gated Cycle Unit (GRU), which have the advantages of processing complex network data and time series data respectively. In order to verify the validity of the proposed model, a series of experiments are carried out on the thermal test-bed simulating the operation characteristics of nuclear power plants. The experimental results show that compared with the existing methods, GRU-GCN model has significantly improved the prediction accuracy. This result not only shows the potential of the deep learning model combining GCN and GRU in processing complex system data, but also provides a new idea for the safety monitoring and operation management of nuclear power plants in the future.
Research on the Natural Circulation Ability of Lead-Bismuth-Steamwater Direct Contact Multiphase Flow
Authors: Ma Xiaoyao , Cheng Daoxi , Zhai Weiming , Yu Song , Xing Mingdi , Li Ruizhi , Zhou Ping , Gao Weilong 
Published: November 2024

Abstract
Pb-Bi-cooled direct contact boiling water fast reactor (PBWFR) uses a large amount of steam generated by the direct contact heat exchange between liquid metal and water to drive the natural circulation flow of lead bismuth (Pb-Bi) alloy in the primary circuit. To explore the natural circulation flow characteristics of Pb-Bi-water-steam multiphase flow is of great significance for improving the natural circulation ability of Pb-Bi. This paper establishes a one-dimensional theoretical analysis program and numerical simulation calculation program for the PBWFR system, It’s the law of the natural circulation ability of Pb-Bi was obtained with the change of core power, ascent channel height, water injection subcooling degree and so on. At the same time, the transient process of direct contact between Pb-Bi and water at high temperature and high pressure was simulated by numerical simulation program. The key parameters such as temperature, velocity and void fraction of Pb-Bi multiphase flow in the ascending channel were obtained.
The research in this paper lays a theoretical foundation for in-depth exploration of the natural circulation mechanism of Pb-Bi, which is helpful for the optimal design of PBWFR reactors.
Numerical Analysis of Flow Field and Phase Distribution Characteristics of Sub-Cooled Boiling Flow in the PSBT 5×5 Fuel Bundle Channel
Authors: Liao Hengji , Yan Xiao , Gu Haifeng , Qin Shengjie 
Published: November 2024

Abstract
Rod-bundle coolant channel is the key component of the primary loop in pressurized water reactor. Its local thermal-hydraulic behavior is associated with safety and economic of nuclear power plants. With development of computing power and two-phase flow model, Computational Fluid Dynamics has become one of the most feasible approaches to simulate sub-cooled boiling flow phenomenon of the fuel assembly. In this paper, based on Euler-Euler two-fluid model coupled with wall heat partitioning model, flow boiling performance in the PSBT 5 × 5 full-scale fuel bundle with uniformly heated condition have been adopted to validate this CFD model. Predictions of the sub-channel averaged void fraction profiles were in good agreement with the experimental data. Then, void fraction distribution near the fuel rod and transverse velocity field in the PSBT 5 × 5 rod-bundle channels were analyzed under three axial heat flux profiles. With the increase of total heating power, void fraction on the surface of all the fuel rods increase gradually. Especially, a large number of vapor bubbles were observed to occur near the cold rods located at the certain positions. Comparative analysis of the transverse velocity field downstream from the last mixing-vane spacer grid showed that lateral flow downstream the mixing grid was the main factor that influences phase distribution in the fuel bundle channels. The swirl flow occurred in central sub-channels promoted vapor bubbles in the vicinity of the hot rod to accumulate on center of central sub-channels and induced the liquid phase to continuously wet the heated wall, which would enhance the heat transfer behavior. However, due to the limitation of flow shroud and configuration of the mixing vanes, there was no swirl flow in the center of side or corner sub-channels of the PSBT 5 × 5 rod bundle. Besides, the complex cross flow around the fuel rods in the side or corner sub-channels may guide the liquid away from cold rod surface, causing a substantial increase of near-wall void fraction as the heating power increased and ultimately heat transfer deterioration. The research rules in this paper provide an excellent support for a deep understanding of the evolution of phase distribution from the mixing grid in the 5 × 5 rod bundle.