bluuebunny/crossref_metadata_2025_split huggingface dataset:
Experimental Research on the Retention Characteristics of Hydrophilic and Hydrophobic Aerosols in Jet Regime
Authors: Ma Qianchao , Zhou Yanmin , Gu Haifeng , Sun Zhongning , Yang Jinghao 
Published: November 2022

Abstract
Aerosol is one of the main forms of fission products in nuclear reactor accidents, which are derived from direct leakage of solid fission products and homogeneous and heterogeneous nucleation of gaseous fission products. Pool scrubbing may reduce the leakage of radioactive aerosol, and it is vital for the prediction of radioactivity to accurately grasp the retention efficiency. In this paper, experimental research on aerosol retention characteristics in jet regime was carried out. Titanium dioxide and cesium iodide were used to simulate radioactive aerosols with different hydrophilicity, and the retention efficiency under different inlet pressure and aerosol size was obtained. The results show that the retention efficiency of aerosol increases significantly while the inlet pressure increases from 43 kPa to 110 kPa. While the inlet pressure is higher than 110 kPa, the retention efficiency still increases with the increase of inlet pressure, although the increasing trend is not obvious. The most easily permeable range of cesium iodide aerosol is smaller than that of titania aerosol, and the retention efficiency of cesium iodide aerosol is always higher than that of titania aerosol, which may be caused by hydrophilicity. However, with the increase of stokes number, the difference of retention efficiency between the two aerosols gradually decreases. The difference between different aerosols was analyzed in this paper and it was found that the influence of hydrophilicity was negatively correlated with stokes number. The results of this research can be used to assist the validation of pool scrubbing model.
Reactivity Introduction Accident and Control Strategy Analysis of Small Modular Supercritical CO2 Brayton-Cycle Reactor System
Authors: Ming Yang , Jin Yang , Qiao Yanlong , Zhao Fulong , Tian Ruifeng 
Published: November 2022

Abstract
In this paper, a dynamic model of a 5MW small modular supercritical CO2 Brayton-cycle reactor system is established based on Modelica language, and the control rod ejection accident simulation with different reactivity is carried out. This paper takes a 5MW small modular supercritical CO2 Brayton cycle reactor system as the research object. Based on the modular modeling idea, the dynamic model of the system including reactor, turbine, compressor, regenerator, cooler, and other main equipment is built, and the reactor power control system is based on the PID control principle is designed. The results show that the system has certain inherent safety, and the designed control system can coordinate the main equipment under accident conditions, maintain the important parameters such as reactor core fuel temperature, coolant temperature, and reactor pressure within the limit, and control the system flow, permeability and compressor rotor speed within a reasonable range. The work of this paper provides a certain reference for the selection of control strategy and safety system design of the new supercritical CO2 reactor system.
Molecular Dynamics Study on Thermal Conductivity of Unirradiated and Irradiated Symmetrical Tilt Grain Boundary 3C-SiC
Authors: Cai Ziqi , Zhang Qingmin , Shao Zhuang , Li Yuanming 
Published: November 2022

Abstract
The cubic silicon carbide (3C-SiC) has been considered as a candidate structural material for several types of advanced nuclear reactors. The effects of cascade collision on thermal conductivity in symmetrical tilt grain boundary (GB) were studied by Molecular dynamics (MD) simulations. The thermal conductivity of 3C-SiC at Σ5(210)[001] GB was calculated using non-equilibrium molecular dynamics (NEMD) methods. A relatively small simulation unit was used to analyze the effect of different energies of incident PKA (primary knock-on atoms) on the thermal conductivity of 3C-SiC and to compare the results with perfect structure GB system. Finally, the vibrational density of states (VDOS) of atoms in the GB region was calculated to analyze the phonon mismatch at the interface. Calculations show that cascade collisions generated by energetic atoms will result in a decrease in thermal conductivity of the Σ5(210) GB system, but the effect varies in different regions, with a sharp decrease in thermal conductivity and an increase in thermal resistance for the intracrystalline region, while the magnitude of change in either thermal resistance or thermal conductivity is not significant in the GB region. Irradiated model shows a higher GB energy compared to the unirradiated model. For all irradiated models, lattice defects have a significant effect on the thermal conductivity of the GB system, depending on the spatial structure of the GBs. the results of the VDOS analysis suggest that an increase in the degree of atomic lattice mismatch near the interface is responsible for a further increase in the thermal resistance of the irradiated GB system.
Development of the Digital Human-Computer Interaction Platform of the M2LFR-1000 Reactor Based on QT and ICoCo
Authors: Liu Tianyi , Ding Tao , Shen Yanyi , Chen Hongli 
Published: November 2022

Abstract
Human-computer interaction interface refers to a human-oriented, humanized and comfortable information display system. Building an interactive relationship with users in design is a bridge between users and information. The purpose of the human-computer interaction interface is to exert mutual influence between humans and machines, so that both parties understand each other’s information and keep the information exchange channel smooth. However, at present, there is a lack of visual human-computer interaction interface for lead reactor, and the results calculated by the thermal hydraulic calculation program cannot be presented directly to users. Moreover, it is very inconvenient for users to modify some parameters. Therefore, for the M2LFR-1000 lead-cooled fast reactor, this paper will introduce the development of a digital human-computer interaction platform based on the ICoCo concept and QT framework. The system code RELAP5 has been selected as the underlying thermo-hydraulic program. The platform can visually display the overall structure of the primary circuit of the M2LFR-1000, and can display parameters during calculation such as the inlet and outlet temperature of the primary circuit coolant in the core, the inlet and outlet temperature of the secondary circuit coolant in the main heat exchanger, and the mass flow of the core. The user can also adjust the parameters of the time-related components such as the inlet and outlet temperature of the secondary circuit coolant in the main heat exchanger when the calculation is being performed. In this way, the calculation of the thermal-hydraulic program can be synchronized with the display of the calculation results. At the same time, the user’s needs can be fed back to the thermal-hydraulic program in time. In addition, since QT is a cross-platform interface development software, the visual human-computer interaction interface developed by QT also has a good code compatibility, which is convenient for subsequent development and upgrade.
Experimental Investigation of Cable Arrangement on the Cable Burning Behavior in a Cone Calorimeter
Authors: Yang Ping , Huang Xianjia , Peng Lan , Xing Chaoliang , Zhao Chunyang 
Published: November 2022

Abstract
Cables are one of the main fire hazards in nuclear power plants (NPPs). Cable arrangement, especially cable spacing in an electrical cable tray has a significant effect on cable tray fire. Predicting cable fire behavior in NPPs is necessary for the nuclear safety analysis. In this work, cable fire experiments with halogen-free flame retardant cable (XLPE cable with a diameter of 8.55mm) were carried out in a cone calorimeter. Cable spacing was changed from 0 mm to triple of cable diameter, including 7 different spacing arrangements under 3 different external heat fluxes. Besides, the experimental data of two other halogen-free flame retardant cables in the literature available are used. With the increase of cable spacing, the heat release rate per unit area (HRRPUA) was significantly increased and the fire duration was considerably decreased, while the time to ignition (TTI) of the cable was basically invariable with different cable spacings. Furthermore, the effects of the cable spacing on the cable burning become constant when the cable spacing was larger than a critical cable spacing. This work provides experimental support for industrial cable fire risk assessment, and the above findings are beneficial for fire prevention in the design stage of NPPs.
Research and Application of Pre-Development Software Qualification Technology for Nuclear Safety I&C Systems
Authors: Wang Shengchao 
Published: November 2022

Abstract
With the continuous advance of the mass construction of nuclear power in China, it has become an irresistible trend to use mature commercial-off-the-shelf (COTS) to perform safety functions in the safety important instrumentation and control (I&C) systems to reduce the construction cost of nuclear power plants (NPPs). The quality of pre-developed items especially the pre-developed software (PDS) has been paid more and more attention by stakeholders. By combining with the specific requirements of international mainstream guidelines and standards related to nuclear safety I&C systems or equipment PDS qualification, a set of strong commonality, operational PDS qualification scheme has been proposed. The scheme includes qualification process, activities, tasks and methods. The main qualification activities are suitability evaluation, quality evaluation, operating experience evaluation, and additional test. According to the software requirements of implementing different safety function categories, the corresponding software integrity level qualification tasks are matched to the activities. In addition, according to the characteristics of each task, the recommended execution methods are given, which mainly include document review, traceability analysis, FMEA, code review, code static analysis, code dynamic test and so on. Finally, the application effect of the PDS qualification scheme for safety I&C systems of NPP is introduced, and the concerns in the qualification process are expounded, so as to provide reference for the safety items software qualification of NPPs.
Research on GIL Fault Diagnosis and Location Method Based on Partial Discharge and Vibration Monitoring
Authors: Liu Sen 
Published: November 2022

Abstract
At present, the ultra-high voltage transmission lines in nuclear power plants mostly use Gas Insulated transmission lines (GIL). This type of transmission line transmits large natural power, and GILs are mostly laid through corridors and themselves are metal-enclosed gas -insulated structures, which are affected by the natural environment and have low reliability. high. However, due to the long length of each independent air chamber of the GIL, when it fails, it is accurately difficult to monitor the fault and locate the fault, which reduces the availability of the unit and brings about greater economic losses. In this paper, through the simulation test of the vibration signal propagation when the GIL fails, the signal attenuation law of the vibration signal passing through the insulator is tested, and the fault cell is located through the signal analysis of the vibration sensors in different directions. At the same time, through the simulation test research on the discarge faults of metal spikes, floating potential and metal particles in the GIL, after obtaining the partial discharge signal through the sensor, the corresponding partial discharge eigenvalues are extracted, and the eigenvalues are classified based on the SVM method to realize the GIL fault. Which can provide support for engineering practice.
Study on the Applicability of Flat Ceramic Membrane in Simulated Primary Coolant Water Environment of PWR Nuclear Power Plant
Authors: Guan Yuxin , Chen Jiguo , Zhang Guodong , Ren Xuqiu 
Published: November 2022

Abstract
A flat ceramic membrane has a separating function and is a novel type of material membrane. High temperature resistance, superior mechanical strength, huge permeation flow, and reusability are all advantages. It’s been employed in water treatment industries like food, medicine, and fine chemicals with great success. However, there has been no mention of application research in nuclear power plants so far. This paper evaluates the feasibility of ceramic membranes in a primary coolant water environment in terms of filtration materials and fluid compatibility. The researchers tested the chemical stability of the flat ceramic membranes and observed the variation of dissolved elements concentration with time, mass and temperature. The dissolved elements are mostly Al, Si, Mg, and Ti, according to the results. The leachable Si element is the most abundant, followed by Al and Mg, with a very low eluted Ti percentage. With increasing duration, mass, and temperature, the dissolved elements of the flat ceramic membrane in the simulated primary coolant water environment rise. With time, the rate of dissolved element release diminishes. The dissolving rate tends to a constant value until a particular concentration (about 10 mg/L in this experiment) is reached. The existing flat ceramic membrane is not suited for the primary coolant water environment of PWR nuclear power plants due to the high dissolving rate of each element.
Study on the Feasibility of Power Supply Migration in Production Preparation Area of the Nuclear Power Station
Authors: Yonggang Lin 
Published: November 2022

Abstract
Since 2016, Some nuclear power plants in the northern region has been facing severe situation due to seasonal heating; and because of seasonal hydropower generation, the generation situation of some nuclear power plants in the southern region is more serious. At present, the power supply in the production preparation area of most nuclear plants is mostly purchased electricity. There is a big difference between the price of electricity purchased from external power grid and the cost of power generation from nuclear plant. Therefore, if the production preparation area of the nuclear power plant power consumption is changed to power supply by itself (Auxiliary power), it will make a great contribution to energy saving and power saving, cost reduction and efficiency promotion. There is no study on the efficiency promotion of nuclear power generating units at home. This study mainly focuses on the project of power supply migration in production preparation area of nuclear power plant. This study is carried out from four aspects: transformer capacity, bus voltage drop of 6.6kV, capacitive current, and economy. The definition of production preparation area of nuclear power plant: the living area of the on-duty dormitory of the plant, administrative office area and other relevant areas.
Study on UC Phase Control of UC Ceramic Microspheres Prepared by Sol-Gel Method
Authors: Wen Feng , Guo Bolong , Li Mingyang , Wang Jiachen , Zhang Fan , Liu Wei , Ma Wenkai , Liu Zhihua 
Published: November 2022

Abstract
Since uranium carbide (UC) fuel has higher thermal conductivity and uranium density than UO2 fuel, good inherent safety and economy can be achieved by UC dispersion fuel. This paper focuses on the research of UC phase control during sol-gel preparation of UC ceramic microsphere. Using nano-carbon black powder as carbon source, UC ceramic microspheres were prepared by sol-gel process combined with carbothermal reaction process. Specific amount of carbon black was uniformly added into the glue solution, which was then dispersed to prepare carbon ADU gel spheres. With heat treatment of carbon ADU gel spheres subject to carbothermic reaction, ceramic microspheres containing UC phase were obtained. During the experiment, the effects of different carbon black addition ratio, thickener addition level and dispersion process on the chemical composition, sphericity and surface morphology on UC ceramic microspheres were investigated. The effects of heat treatment temperature, atmosphere and pressure on the preparation of UC microspheres were also studied during the carbothermic experiment. The process parameters for UC phase control were determined through the experiment and XRD, and the ceramic microsphere samples containing the UC phase were obtained as well. Analysis results show that crack-free UC microspheres with smooth surface can be obtained after drying when sol solution C/U is 2.5∼6.0 and thickener addition level is 1.2∼2.0 kg/kgU. By reaction treatment at high temperature, ceramic microspheres with good sphericity, compact structure and UC phase composition were obtained. A preparation method of UC ceramic microspheres was developed as a result.
Analysis of Single-Line Fast Starting Technology for Emergency Diesel Generators in Nuclear Power Plants
Authors: Li Hongzhou 
Published: November 2022

Abstract
This paper analyses the reasons for the failure of five consecutive single-line fast starting of emergency diesel generators in nuclear power plants. Based on the starting dynamics of diesel generators and governor control principles, and considering the starting characteristics of the respective units, the study identifies the root causes and key reasons affecting single-line fast starting among various contributing factors. The study extracted a set of methods for handing emergency diesel engine set faults. Employing a comparative analysis of test results and diagrams, the paper proposes solutions to ensuring the success of five consecutive single-line fast starting of emergency diesel generators for the first time in the world.
A Systematic Research on Prevention and Control Measures to Blockages of Cooling Water Intake System in Coastal Nuclear Power Plant
Authors: Cao Ranran , Yang Jia , Bai Wei 
Published: November 2022

Abstract
The cooling water intake system of coastal nuclear power plant involves multiple parts from the sea to the land. Moreover, the interception facilities of cooling water intake system are technically complex and face variable blocking risks. In order to improve the safety, stability and economy of nuclear power, for the first time, a systematic study on prevention and control measures to blockages of cooling water intake system in coastal nuclear power plant is carried out from the systematic perspective. The principle of active defense, depth setting, stepped filtration and permanent or temporary combination are adopted in this study. The research contents concern system design, equipment selection and technology layout, etc. Some key technologies, such as cold source blockages investigation, flow field active guidance and avoidance, cascade filtering system, pump station design, monitoring and early warning system, emergency response, etc., have been emphatically studied and optimized. As a result, a reasonable and systematic scheme of prevention and control measures to blockages of cooling water intake system is proposed. The application of this systematic scheme will enable nuclear power plant to form a strong-matrix prevention and control model containing all parts of the cooling water intake system. It can improve the active defense capability of cooling water intake system against blockages and reduce the occurrence of cooling water intake blockage incidents in nuclear power plant.
Thermal Experimental and Analysis of Vertical Spent Fuel Dry Storage Canister
Authors: Tang Qionghui , Li Mao-Yun , Cheng Cheng , Xie En-Fei 
Published: November 2022

Abstract
A full-scale prototype of vertical ventilated spent fuel dry storage canister is tested to evaluate its heat removal performance. Specifically, a temporary shelter is built to resemble the factory encompassing dry storage canisters. Electrical heating components with designed heat power are designed to represent the spent fuel assemblies inside the canister. From the test data its accordance with thermal safety related rules and guidance is proved. Furthermore, these data are compared with the simulated results, and the simulated results using CFD computer code of CFX. It is shown that the temperatures obtained from test agree well with the simulated results, and the simulated temperature of each component is no more than 10% higher than the test data, the velocity inside air-channel is in range of 1m/s, which is close to the environmental wind velocity of mid-summer of the testing site. It is also shown that conductive heat transfer vial air channel between the concrete over-pack and the steel canister covers more than 2/3 of the total heat load, indicating that conductive heat transfer plays a major role compared with heat conduction and radiation. From the full-scale prototype testing, it is shown that the vertical dry storage canister design has inherited conservatism from the aspect of thermal safety, And it is reasonable and acceptable to use CFD numerical calculation method for thermal analysis of this type of ventilated spent fuel dry storage system.
Reactor Pit Injection Isolation Design Modification and Experience Feedback of UK HPR1000 In-Vessel Retention
Authors: Zheng Hua , Wei Shuhong 
Published: November 2022

Abstract
During UK HPR1000 Generic Design Assessment (GDA), a shortfall about In-Vessel Retention (IVR) strategy is identified after a detailed timeline analysis according to UK nuclear safety requirements “Safety assessment principles for nuclear facilities” (SAP): the reactor pit injection manual action available time is less than the required time. Four possible modification schemes are proposed and analyzed. After ALARP (as low as reasonably practicable) analysis, the design of reactor pit injection isolation is modified to make the reactor pit injection manual action available time more than the required time : a remote manual permissive action in the main control room by a switch in the Emergency Control Panel replaces the local administrative isolation. Improvements for advanced nuclear power plant R&D, design and future operation are suggested, including enhanced requirement management and configuration management, especially for the design basis of functional requirements and sizing requirements; systematic identification of time-critical actions (TCA), performing analysis, preliminary validation and integrated validation of TCA during design and construction phase, implementing a TCA programme during operation phase; improving Severe Accident Management Guidelines (SAMG) implementation approval process by adding “the point of entry to the use of SAMGs is reached” to reduce the time of SAMG implementation approval process.