bluuebunny/crossref_metadata_2025_split huggingface dataset:
Research on Simulation of Emergency Evacuation Outside Nuclear Power Plants Based on Genetic Algorithm and Variable Neighborhood Search Algorithm
Authors: Yan Jin , Zhang Jiemin , Cao Qun , Wang Mengxi , Long Liang , Xue Na 
Published: November 2024

Abstract
This paper focuses on the evacuation process of the public during off-site emergency, and studies a simulation method for off-site emergency evacuation based on genetic algorithm and variable neighborhood search algorithm. And this model is used to calculate and analyze the off-site emergency evacuation plan for a nuclear power plant, based on information such as population, vehicles, assembly points, settle places, transportation networks, and radiation dose fields around the nuclear power plant. The optimization objectives are evacuation time, collective dose, or individual maximum effective dose. The consumption time of the evacuation process and the dose exposure received by personnel during the evacuation process are provided. And compare the results with traditional evacuation simulation algorithm. The results indicate that emergency evacuation plans should not only consider the evacuation time, but also comprehensively consider the dose received by personnel during the evacuation process. In addition, this paper analyzed the impact of adding private cars on the evacuation process, and the results showed that when adding more than 50% of private cars to the number of buses, the impact on evacuation time, collective dose, and individual maximum effective dose significantly increased. In addition, when the evacuation time is taken as the optimization objective, the evacuation time of the optimized evacuation plan through genetic algorithm combined with variable neighborhood search algorithm is shorter than that of the traditional simulated evacuation plan. This algorithm can generate a more optimized evacuation plan.
Development of a Nonintrusive ROM for Parametric Study in the RPV Lower Plenum
Authors: Shang He , Peng Minjun , Xia Genglei , Zhang Shiqi , Zhu Shijie 
Published: November 2024

Abstract
System thermal hydraulic codes are extensively used to assess the response of large-scale systems due to their computational efficiency in simulation. Nevertheless, they cannot simulate inherently multi-dimensional physical phenomena because of the one-dimensional approach adopted. A common example is the multi-dimensional flow in the reactor pressure vessel lower plenum. One solution is to employ computational fluid dynamics codes to simulate the lower plenum and obtain the required multi-dimensional flow. However, computational fluid dynamics computation is typically time-consuming, introducing a new challenge. In this paper, a nonintrusive reduced order model for steady state calculation of two-dimensional lower plenum has been developed using proper orthogonal decomposition and sparse grids methods to overcome this challenge. The input of reduced order model is the coolant mass flow rate at the inlet of the lower plenum, and the output is the velocity field distribution in the lower plenum. The range of the inlet flow rate is from 500 kg/s to 600 kg/s, and all data in snapshots are generated through repetitive simulation under different sampling points using OpenFOAM. The performance of the constructed reduced order model is validated on a test set including 16 new sampling points. The results indicate that the established reduced order model can rapidly provide velocity field distributions in the lower plenum for any inlet flow rate ranging from 500 kg/s to 600 kg/s, and the errors are all acceptable to meet simulation requirement.
Design and Calculation of Nuclear Heating System Utilizing a Molten Salt Heat Storage Circuit for Waste Heat Absorption and Peak Regulation
Authors: Chong Haofang , Xue Ruojun , Wang Sen , Cao Yaowu , Jing Furu 
Published: November 2024

Abstract
The utilization of nuclear energy heating systems is prevalent in the heating sector. However, the peak regulating capacity of nuclear heating systems is typically inadequate. To improve it, a nuclear energy heating system integrated with molten salt energy storage is devised. Molten salt energy storage systems are widely adopted in thermal power units and photovoltaic power generation to regulate peak load and store excess energy. By integrating the molten salt energy storage system into the nuclear energy heating system, surplus heat from the reactor can be temporarily stored during periods of low demand in the central heating network and released when there is high demand, enhancing the self-regulating peak capacity and operational flexibility of the system. Furthermore, coupling with the molten salt energy storage system allows for partial replacement of certain functions of the original residual heat removal system by absorbing shutdown waste heat at the initial stage of reactor shutdowns and providing it to either the central heating network or other users. In this paper, we establish a self-programmed one-dimensional calculation model for evaluating whether using molten salt to absorb initial waste heat from reactor shutdowns in a nuclear heating system is feasible and if this absorbed heat can be applied to supplementing central heating networks; additionally, we conduct preliminary calculations on utilizing molten salt for peak regulation purposes within the central heating networks, to preliminary verify the usability of the design scheme.
Effect of Corrosion Behavior on the Heat Transfer Performance of Fuel Cladding in Lead-Cooled Fast Reactors
Authors: Feng Wenpei , Yang Shangdong , Yu Hongxing , Lu Xingqing , Zhang Xue , Gu Zhixing 
Published: November 2024

Abstract
In lead-cooled fast reactors, preventing corrosion of materials like stainless steel by liquid metal is crucial. The protective oxide film on stainless steel surfaces, formed to mitigate corrosion in lead/lead-bismuth systems, can, paradoxically, degrade heat transfer in cladding during high-oxygen concentration or normal operation. This study investigates the impact of corrosion on heat transfer, focusing on steady-state conditions. By analyzing temperature changes with and without corrosion, the study aims to understand the balance between oxide film formation and heat conduction. Additionally, it assesses corrosion effects under accident conditions to enhance safety in lead-cooled fast reactors. The findings contribute valuable insights for designing and operating cooling systems, ensuring the safety and reliability of nuclear energy systems.
A Nonlocal Meso-Scale Damage Model for Nuclear Graphite
Authors: Ma Yan , Qian Yue , Hao Yuchen , Lin Musen , Li Yue , Wang Jinhua , Wang Haitao 
Published: November 2024

Abstract
Due to its excellent nuclear physical properties and high-temperature resistance, graphite is widely used in many areas of the nuclear industry. In high-temperature gas-cooled reactors (HTGRs), graphite is used not only in the core structural components as the neutron moderator and reflector, but also as the matrix material for the fuel elements. During the operation of HTGRs, the core structure and fuel element are subject to various static and dynamic loading. Therefore, it is particularly important to study the reliability evaluation of nuclear graphite for the safe operation of reactors. As a typical brittle material, the quantity, size and shape of internal defects of nuclear graphite have strong randomness, which leads to a large discrete strength of nuclear graphite. It is more reasonable to use a probabilistic method to evaluate the strength of nuclear graphite. At present, the dominant approach to the probabilistic evaluation of nuclear graphite is based on finite element methods, such as the Burchell model. However, compared with the meshless method, the finite element method make it difficult to handle the stress singularity brought by the displacement discontinuity, the computation complexity brought by the crack bifurcation and the non-locality of failure. As an emerging meshless method, Peridynamics has been adopted by most scholars. In this paper, we develop a uniaxial tensile fracture model of nuclear graphite based on the ordinary state-based Peridynamic theory. After that, a random number and size of initial defects are added to the nuclear graphite according to the pore-size distribution in the graphite microstructurally parameters. In order to verify the reliability of the model, this paper compares the prediction results of the model with the experimental data on two typical graphites using the two-parameter Weibull function. Besides, this paper compares the model in this paper with the Burchell model, and the results show that the spherical defect shape used in this paper is more suitable for predicting isostatic graphite.
Bubble Features on SiC Surfaces in Flow Boiling
Authors: Bing Tan , Ren Haoliang , Sun Rulei , Cheng Songbai 
Published: November 2024

Abstract
The study of bubble dynamics is crucial for understanding the boiling heat transfer process, predicting heat transfer efficiency, and preventing film boiling issues. The research aimed at experimental exploration of bubble characteristics on silicon carbide (SiC) surfaces during flow boiling at subcooling levels of 4K to 12K and fluid flow rates between 0.15–0.3 m3/h, all conducted under atmospheric pressure. The investigative method harnessed a U-net artificial neural network to meticulously identify attributes of the bubbles, focusing on properties such as orientation, perimeter, effective diameter, roundness, and aspect ratio distribution. Key parameters for the working fluid were set within the specified ranges for subcooling and flow rates, ensuring the experimental conditions replicated flow boiling at atmospheric pressure. Utilizing U-net as a tool allowed for precise recognition of bubble features on SiC surfaces, which are the primary subject of this study. The outcome of the experiments, facilitated by the neural network, successfully mapped the distribution of various characteristics of bubbles on SiC surfaces under flow boiling conditions. This provided valuable experimental data and insights that aid in the further understanding of SiC surface behaviors in flow boiling processes.
Research on Sealing Performance and Electrical Performance of Electrical Penetration Assembly in Nuclear Reactor Containment
Authors: Wang Yanlu , Yu Xinli , Cao Xiaxin , Liu Yu 
Published: November 2024

Abstract
Electrical penetration assembly (EPA) are electrical conduits mounted on the containment that serve to maintain the integrity of the containment pressure boundary and provide a path for cables to traverse the containment. Currently, low-voltage electrical penetration assembly (LV-EPA) is mainly made of organic materials to realize the sealing function, and it is not clear whether the performance of organic sealing materials will be degraded in the event of a nuclear reactor under extended design conditions (DEC-B). In this paper, a representative original-size LV-EPA is selected as the research object, and the equipment life test and the test for measuring the insulation and sealing performance of the LV-EPA under severe accidents are carried out by constructing a test setup that simulates severe accident conditions. The results show that: under the condition of 200°C and 0.6MPa, with the increase of test time, the electrical performance of LV-EPA remains good, but the sealing performance will be degraded step by step; after 72h simulated DEC-B test, the sealing performance and electrical performance of the equipment satisfy the standard requirements.
Study on Application of the Single-Curvature Polyhedron Hydro-Bulging Technology in Metallic Reactor Containment
Authors: Dong Jianling , Huang Xingcheng , Chen Ke , Ning Bohao 
Published: November 2024

Abstract
The integral RPV configuration and layout nuclear reactors eliminate the loop piping and the externally located steam generators, pumps and pressurizer with their individual reactor pressure vessels. Hence, the footprint of the containment can be greatly reduced. This size reduction makes them possible to use small size and high pressure vessels as their containments. Compared with cylindrical pressure vessels, spherical vessels have the advantages of higher pressure bearing capacity and material saving. So for integral RPV configuration reactors, using spherical vessels as their containments and reactor building as a barrier to protect the containment from attack from aircraft is a promising configuration and layout plan.
The traditional spherical vessel technology adopts presses and dies to form the steel sheet blanks into double-curvature orange petals, then to assemble and weld the petals to obtain a spherical vessel, the process is complicated, long production period and higher cost.
The single-curvature polyhedron hydro-bulging technology to manufacture spherical vessel is a new technology for manufacture spherical vessels, which has been used to manufacture spherical vessels in many fields. But up to now, there is no case that the technology is used in the nuclear industry. This paper tries to explore the application of the single-curvature polyhedron hydro-bulging technology for manufacturing spherical vessel in manufacturing steel spherical containment vessel of the nuclear power plant.
Overview of the Application of Thermal Management in Industry and the Application Prospect in the DCS of the Nuclear Power Plant
Authors: Xu Youyou , Deng Jian , Qiu Zhifang , Zhu Dahuan , Wang Dongwei , Han Wenbin 
Published: November 2024

Abstract
Reactor digital control system (DCS) is the brain and central nervous system of nuclear power plant. It is the most complex system of nuclear power plant, which can ensure the safety and stable operation of the nuclear power plant. The DCS in nuclear power plants can be divided into safety-related DCS and non-safety-related DCS. With the improvement of digital, intelligent, integrated and miniaturized technology of the DCS in nuclear power plant, thermal management is becoming an increasingly integral part of advanced DCS system design. In electronic devices, temperature is one of the key factors affecting the performance of the electronic components. In the field of electronic heat dissipation, there is a famous ten-degree rule: the reliability of electronic devices is closely related to the temperature. Within a certain temperature range, the temperature of electronic components rises by 10 degrees, its reliability will decline by 50%. Therefore, effective thermal management of electronic equipment is particularly important. Usually, thermal management technology mainly focus on three key elements: heat source, heat sink and the channel to removing heat. Large amount of researches about these three elements were carried out in order to meet the requirements of application scenarios and various heat indexes. On one hand, thermal management can reduce the failure probability of the DCS, on the other hand, more efficient thermal management can reduce the weight, volume and energy consumption of the DCS. Firstly, the application of thermal management technology in various industrial fields is systematically reviewed and summarized in this paper. Then, combined with the three elements of thermal management mentioned above, the application scenario of thermal management in DCS of nuclear power plant is analyzed. Systematic thermal management research on DCS of nuclear power plant aims to lay the groundwork and support for the subsequent application of advanced thermal management technology in DCS of nuclear power plant, so as to reduce the weight, size and the energy consumption, reduce the space occupied by DCS and improve the reliability of the instrument control system.
Application of Artificial Neural Networks in the Optimization of Feedwater Heater Models
Authors: Dong Shubiao , Xiaoyu Zhang , Qianping Zhang 
Published: November 2024

Abstract
The traditional lumped parameter model for feedwater heaters is insufficient to accurately predict changes in heat transfer performance caused by factors such as aging and fouling during power plant operation. This study established an operational database consisting of 68 million datasets and employed it to train an artificial neural network model. This neural network model establishes a mapping relationship between operating time, thermal hydraulic conditions, and heat transfer efficiency correction coefficients. Subsequently, the neural network model was employed to optimize the heat transfer calculation process of the lumped parameter model, and a new composite model was formed after integration. Finally, the optimization effect of the model was tested using operational data. The results show that compared with the lumped parameter model, the composite model exhibits higher and more stable computational accuracy at different operation times. This means that the composite model can, to some extent, reflect the changes in heat transfer performance of the feedwater heater during operation.
Experimental Study on Two-Phase Flow Instability in Parallel Helically Coiled Tubes Under Rolling Motion
Authors: Wang Ruohao , Qi Chao , Ren Jiaxing , Qiao Shouxu , Tan Sichao , Tian Ruifeng 
Published: November 2024

Abstract
Parallel helically coiled tubes are extensively employed in heat exchangers of floating nuclear platform owing to their compact configuration and exceptional heat transfer efficiency. The helical structure of the coiled tube leads to the occurrence of secondary flow, making the flow instability phenomena more complex. Under ocean conditions, the helically coiled tube will suffer from more serious two-phase flow instability. Thus, research on the two-phase flow instability of parallel helically coiled tubes under rolling conditions is crucial.
The present study experimentally investigates the two-phase flow instability in parallel helically coiled tubes under static and rolling conditions. Two vertically parallel helically coiled tubes which have 8mm hydraulic diameter are used in the experiments. Ledinegg flow instability and Thermal Oscillation (TO) are observed under static conditions. The single-phase oscillations, Ledinegg flow instability and TO are observed under rolling conditions. The experimental phenomenon under static and rolling conditions are compared.
The stability map of parallel helically coiled tubes is established. The Fast Fourier Transform (FFT) method are used to investigate the evolution characteristic of the flow instability under rolling conditions. The variation of amplitude spectrums indicates that impact of rolling conditions on flow instability diminishes as heating power is increased. Finally, the parameter effects of rolling angle and rolling period on two-phase flow stability are indicated.
A Summary of Historical Evolution of Radiation Protection Policy From ICRP
Authors: Huang Yufei , Zhang Zongyuan , Liu Ling , Wang Yu , Guo Jingni , Wei Liqiang , Xie Feng , Tong Jiejuan 
Published: November 2024

Abstract
The International Commission on Radiological Protection (ICRP), as an unofficial organization dedicated to advancing the science of radiation protection, plays a crucial role in formulating radiation protection guidelines, providing scientific advice, and establishing principles in this field. By drawing Sankey map, combing ICRP’s past publications, and investigating history of radiation protection, this study deeply reviewed the focus of ICRP’s work in different periods, evolution and significance of specific concepts, gradual clarification and deepening of radiation protection concepts, and key directions of ICRP’s future work. By investigating the development and reasons behind the evolution of the classification of radiation situations and the underlying objectives in radiation protection over different periods, this study emphasizes that the shifts in the concept of radiation protection are rooted in the experiences gained through continuous practical applications in real world. Moreover, we provide a detailed discussion on the four major ethical values of radiation protection — beneficence and non-maleficence, prudence, justice, and dignity — highlighting their manifestations and developments during different time periods. This study delves into the relationships between the concepts and principles of radiation protection. Meanwhile, by discussing the relationship between updating of radiation protection theory and the understanding of radiation effects, re-examining the Linear No-Threshold hypothesis based on dose, balancing radiation harms with social values, and standardized management of radiation medicine, the inspiration of the continuous development of ICRP radiation protection concept was proposed.
Research on Measurement and Analysis Method of Radiation Test Section Based on Neural Network
Authors: Lee Jinlin , Zhang Yunsheng , Cheng Jie , Fan Guangming , Jin Shuai 
Published: November 2024

Abstract
In the irradiation test equipment, frequent operations such as material replacement, irradiation device surface replacement, and entry and exit of the reactor are carried out. These factors may lead to sensor failures, resulting in inaccurate test results. During the irradiation test process, the fault diagnosis is usually carried out by operators by observing the parameters during operation and combining their own knowledge. The advantage of using neural networks for fault diagnosis lies in their direct and effective statistical analysis and information extraction of massive, multi-source, and high-dimensional data. This method collects different types of data as the basis, uses the implicit effective information to characterize the operating mode and status of sensors, and thus achieves the purpose of detection and diagnosis. In this study, a neural network fault diagnosis model was established to detect and diagnose the faults of sensors. The fault diagnostic model is trained based on experimental data, and the model is synchronously trained online during the irradiation test process to continuously improve diagnostic efficiency. The fault diagnosis model adopts an unsupervised autoencoder, which takes experimental data as the target for learning. The model trained using obtained from irradiation experiments performs well in simulation tests, and can rapidly and accurately determine the data and process related experimental data when sensor faults occur, providing the location and type of sensor faults, and intuitively and quickly prompt operators with effective information.
Thermal-Hydraulic Analysis of the Upper Plenum in a Typical LFR During a Reactivity Insertion Accident
Authors: Qiu Hanrui , Zhang Ruibo , Zhang Jing , Wang Mingjun , Tian Wenxi , Su G. H. 
Published: November 2024

Abstract
MYRRHA is the Gen IV liquid metal cooled fast reactor, which is a pool-type reactor system developed and designed by the Belgian Nuclear Research Center SCK·CEN. Regarding the safety assessment of the installation, the major challenge is to analyze the complex coolant flow field characteristics in the large open regions of the reactor vessel. In this paper, a coupled sub-channel/CFD multi-scale platform is developed through implementing the sub-channel code SACOS-PB to the CFD code FLUENT based on the domain decomposition and explicit numerical scheme. The flow field characteristics in the upper plenum of reactor emergency shutdown after a reactivity insertion accident (RIA) were investigated. Based on the RIA analysis, the thermal stratification effect has been observed in the upper plenum. The analysis results well demonstrated the 3D flow and local phenomenon of the fluid in the upper plenum, which is significant to reactor design and reactor safety analysis of LFRs.