bluuebunny/crossref_metadata_2025_split huggingface dataset:
Prevention and Treatment of Network Storm in Control Station of Digital I&C System in Nuclear Power Plants
Authors: Dong Wang , Qiang Leng 
Published: November 2022

Abstract
In the design of non-safety I&C system of nuclear power plants, different safety classification and power supply sequence design need to meet the characteristics of system network redundancy and self-healing. Therefore, engineers often use ring network configuration in system network design. When the ring network design or network wiring is wrong, it is easy to cause a network storm, resulting in the abnormal operation of the main control unit of the first-floor control network control station, and the mis-operation of the control unit on the field equipment, resulting in nuclear power safety accidents. This requires that the main control unit of the control station could resist the attack of network storm. This paper provides a method to prevent network storm based on link layer. The abnormal packets are filtered in the network protocol layer of the control unit, and the network storm threshold detection algorithm is used to monitor the network storm attack in time, and the network loopback protection control unit is set. Experiments show that this method can effectively resist the attack of network storm. When the main control unit is attacked by network storm, it can normally complete the task of logic unit, ensure that the field instrument and control equipment is in safe working mode, and the system can automatically return to normal state after the network storm stops. At present, this technology has been realized and applied in the digital I&C system of nuclear power plants, which has good adaptability and broad development space.
Cause Analysis of Oil Leakage of the Skeleton Oil Seal for a Diesel Engine Pre-Lubricated Oil Pump
Authors: Lai Binsheng , Chen Zhi , Li Jianming , Zhou Xin 
Published: November 2022

Abstract
The shaft seal of the pre-lubricated oil pump for the emergency diesel engine in a nuclear power plant is a skeleton oil seal. After about one year of operation, oil will leak from the shaft seal and enter the drive motor, affecting the operation reliability of the emergency diesel engine. By building a numerical simulation platform, the thermal-structure coupling analysis of the skeleton oil seal were carried out and the contact pressure, deformation, Von Misses stress contour and temperature contour of the skeleton oil seal under actual working conditions were studied. Through the establishment of a sealing test device, sealing tests for the concentric smooth shaft, the concentric shaft with a groove on the sealing contact surface, and the eccentricity shaft with a groove on the contact surface were carried out under different shaft surface roughness and hardness. And the influence on the sealing performance of the oil seal was analyzed. The morphology of the sealing lip and the change of material functional groups before and after the trial were analyzed. The study results show that the accuracy and surface quality of the shaft where the oil seal is installed will affect the sealing performance of the oil seal; the main reason for the failure of the skeleton oil seal of the pre-lubricating oil pump is the long-term wear of the transmission shaft and oil seal, which forms grooves on the shaft; and the formation of the grooves will lead to increased leakage. Therefore, improving the wear resistance of the drive shaft is a key. The research results have high reference value for the exploration of the sealing principle of the same type of skeleton oil seal and the analysis of the reasons for the failure of the seal.
Analysis of Hydrogenation Properties of Nuclear Fuel Zirconium Alloy Cladding
Authors: Qin Lin 
Published: November 2022

Abstract
During the service of the reactor, the zirconium alloy will inevitably interact with hydrogen to form zirconium hydride, which will adversely affect its mechanical properties. In this paper, the first-principles software VASP is used to analyze the phase transition of zirconium hydride, and based on the results, the subprogram of MAAP4 integrated analysis for severe accidents is optimized to further analyze the interaction between zirconium alloy cladding and hydrogen. Verification. The results show that the diffusion equation has large calculation potential and high calculation accuracy for the effect of zirconium hydride with obvious phase transition, and can realize the hydrogen performance analysis of the cladding material in the reactor. At the same time, VASP shows that δ – ZrH1.5 is the main component of cladding material hydrogenation It is also the main reason for brittle transition.
Decommissioning Considerations for Layout Design of Nuclear Buildings
Authors: He Wei-ting , Jiang Wei-feng 
Published: November 2022

Abstract
Decommissioning for nuclear power plant is complex project, which is needed to be taken into account during the early phase of design stage. Many researches and practices have already done for nuclear decommissioning activities. The decommissioning chapter is defined in preliminary safety analysis report for new build nuclear project, which needs to describe the design and management activities. According to the preliminary research for decommissioning requirement and scheme, many activities are direct related to layout design of nuclear buildings. In this paper, the decommissioning factors for layout design of nuclear buildings are summarized, which can be used as reference and supply some practices for other project.
Heat Transfer Characteristics of Different Horizontal Wires in Pools of Liquid and Supercritical Carbon Dioxide
Authors: Liu Minyun , Xi Dapeng , Fei Junjie , Huang Yanping 
Published: November 2022

Abstract
The peculiar heat transfer mechanism of supercritical fluids has been studied for a long time, especially the flow and heat transfer in the tubes. Numerous experiments have demonstrated the three heat transfer modes of supercritical fluids: normal heat transfer, improved heat transfer and deteriorated heat transfer. The effects of the pressure, pipe diameter, mass flow rate, heat flux and flow direction also have been well studied in the experiments. From the perspective of theory, the pseudo-boiling theory noticed the similarities between the subcritical boiling and the supercritical fluid heat transfer, and explained the heat transfer mechanism. Another theory focuses the buoyancy and flow acceleration effect caused by the property change also made great success. However, the existing research still yields inconsistent results and more evidences are needed. In this work, the heat transfer characteristics from different horizontal wires immersed in both liquid and supercritical carbon dioxide are studied. For a wide range of bath temperatures (5 ∼ 60 °C) and pressures (5 ∼ 8 MPa), the heat transfer coefficients were measured with a quasi-steady increase and decrease of the heat generation rate. The surface temperature of the wires was derived from the measured resistance. Finally, the boiling curves of both the subcritical cases and supercritical cases were obtained. Meanwhile, the near-wall fluid fields were visualized using the high-speed camera. Different from the prior experiments, this pool-type experiment has avoided the complex turbulence flows, and provides clearer evidences of the different heat transfer modes of the supercritical fluids.
A Numerical Study of the Spray Cooling Effect on Improving the IVR Safety Margin
Authors: Yu Peng , Jia Ningxi , Guo Qiang , Liu Zhuo , Jin Yu , Yuan Yidan 
Published: November 2022

Abstract
In-vessel retention (IVR) is a widely adopted severe accident mitigation strategy in the advanced commercial PWRs. External water cooling on the outer surface of the reactor pressure vessel (RPV) is designed to efficiently remove the decay heat from the in-vessel corium, such that the RPV integrity can be kept and the radioactive corium retained inside. The effectiveness of the IVR strategy requires that the maximum heat flux along the RPV lower than the critical heat flux (the switch point from nucleate boiling to film boiling). In existing engineering design, water cooling is realized by flooding the RPV outer surface. Spray cooling on the outer vessel surface is considered to be a potential method to increase the safety margin. In this paper, we investigate the effect of spray cooling in the IVR applications with respect to the influences on the heat transfer and cooling capacity (limit). Molten pool heat transfer analysis is performed to analysis its influence on the thermal behavior of the melt pool and RPV. The cooling capability (limit) is also analyzed between the spray cooling and conventional cooling measurements. Results indicate that the spray cooling does not show significant influence on heat removal as the changes of the heat flux profile and the wall thickness (that are of interest in safety analysis) are insignificant; the main contribution to the enhancement of the safety margin by spray cooling would be from the expected significant improvement of critical heat flux (CHF).
Mixing and Stratification Phenomena Under Natural Convection in the Multi-Openings Compartment
Authors: Jiang Jinghua , Li Ying , Wang Tianlin , Tong Lili , Cao Xuewu 
Published: November 2022

Abstract
The mixing and stratification of hydrogen-vapor-air in the compartment is extensively studied under inertia jets and turbulent buoyant jets. The characteristics of buoyancy-driven flows during fission and fusion reactor accidents need to be further studied to provide a benchmark for the validation of computational procedures. In this paper, a large-scale facility of a vessel with multiple openings is established to study the mixing and stratification of the mixture driven by buoyancy under natural convection. The test is carried out with well-controlled boundary conditions, in which, a mixture of helium and vapor with a volume fraction of 1:10 is injected in the compartment filled with air. As the injection begins, the gas mixture is driven by the injection source and the buoyancy of the light gas. The gas is released through the top and bottom openings with the flow velocity of 5 m/s and 7.5 m/s, respectively. When the injection is stopped at 4600s, the density gradient induces the natural convection, air enters the vessel through the bottom opening with a relatively low velocity, and the low-density mixture is discharged through the top opening at the same time. During this period, the velocity of the mixture at the opening decreases due to the reduction of the drive force of natural convection.
Study on the Application of Electromagnetic Stepping CRDM at Tilted Environment
Authors: Lu Zhaohui , Li Zewen , Zhou Jianming , Liu Qingsong 
Published: November 2022

Abstract
Electromagnetic stepping control rod drive mechanism is most widely used in onshore pressurized water reactor in vertical direction. However, due to the bias load caused by long-term tilting, the wear and failure of the sliding block kinematic pair and carrier and buffer shaft kinematic pair are caused. The researchers carried out targeted research on the existing problems, and improved the design from the structural design, material improvement and surface treatment, and verified the rationality of the improvements through the full-size prototype tests, which also proved that the electromagnetic stepping control rod drive mechanism is feasible in tilted environment.
Weighted Data-Based Fault Detection Approach for Nonlinear Nuclear Power System
Authors: Chen Zhaoxu , Zhou Hongkuan , Ke Zhiwu , Qi Xiao , Qiu Zhiqiang 
Published: November 2022

Abstract
Mechanism-based and data-based methods are widely used in fault detection of nuclear power plant. Each method has its advantages and disadvantages. In this paper, we propose weighted data-based fault detection method based on data which we called Implicit Model Approach (IMA) for nuclear power plant. By introducing weighted factor and Takagi-Sugeno modeling method, we extend the IMA-based fault detection from linear system to nonlinear system. At local operating point, linear IMA is utilized to generate sub-system residual function. Inspired by Parallel Distributed Control (PDC) method, we generate overall residual function and overall residual evaluation mechanism. Simulation analysis has verified the validity of the proposed algorithm. The significance of the weighted IMA lies in establishing a data-based and mechanism-friendly approach. This work can provide new route for fault detection of nonlinear nuclear power system.
Detection Efficiency Fitting Method Research by Γ-Ray Spectroscopy for Quantitative Nuclear Material Holdup
Authors: Li Yurong , Wang Sijia , Wang Jing , He Lixia 
Published: November 2022

Abstract
The term “holdup” refers to nuclear material that has been deposited in the processing facility’s equipment, containers, pipelines, etc. Nuclear material control and accounting require the monitoring of nuclear material holdup in process. Holdup is useful for indicating the state of facility operations, critical radiation safety status, and so on. It is more difficulty to quantitative analyzed nuclear material holdup by destructive analysis method. So non-destructive analysis (NDA) technology for in-situ and on-line measurement satisfied to the actual situation are the most recommended nominator for holdup quantitative analysis. In order to get the detection efficiency of γ-ray spectroscopy for quantitative nuclear material holdup. The intrinsic detection efficiency, geometric detection efficiency, and self-absorption attenuation correction methods are researched in this paper by combining Monte Carlo simulation and experiment. Three-dimensional and the least square fitting methods, and the distribution of the nuclear material holdup in the simulation pipeline system of nuclear facilities’ production line is simplified. Resulted to the experiment, the deviation between declared and the measured value is less than 30%. Compared with the requirements of national nuclear safeguards supervision and the needs of facility operation, the detection efficiency fitting method research by γ-ray spectroscopy for quantitative nuclear material holdup is achievable. It also can be widely used to provide relevant data in radiation dose monitoring, decommissioning source item investigation, and so on.
Dynamic Modeling and Simulation of MHTGR Plant Start-Up and Shutdown System
Authors: Li Bowen , Dong Zhe , Li Junyi 
Published: November 2022

Abstract
The start-up and shutdown system is an important auxiliary subsystem for the low power level operation of once-through steam generator (OTSG) equipped coal-fired plants or nuclear power plants (NPPs). It is difficult to generate qualified steam through OTSG at a low power level or in the start-up and shutdown process. The start-up and shutdown system can carry out steam-water separation for the working fluid flowing out of OTSG and realize the recovery of mass and energy. It is necessary to maintain the stability of pressure, water level, and flowrate of each branch’s during the start-up and shutdown process, which enables coordination between OTSGs and thermal loads. A modular high-temperature gas-cooled reactor (mHTGR) is a typical small modular reactor (SMR) equipped with the OTSG. So as to realize the automatic start-up and shutdown function of mHTGR, it is necessary to carry out the modeling, simulation, and control design of the start-up and shutdown system. Driven by the above requirements, a lumped parameter dynamic model is proposed according to the conservation law of mass, energy, and momentum. In addition, a novel modeling method of the two-phase vessel is raised in the study which can describe the ‘swell and shrink’ phenomenon in the critical equipment of start-up and shutdown systems. Additionally, the open-loop response to external disturbances is to verify the accuracy of mechanism modeling with numerical simulation.
The Research and Simulation of Beyond Design Basis Accidents in PWR Nuclear Power Plant
Authors: Cheng Guo 
Published: November 2022

Abstract
Fukushima nuclear accident occurred due to earthquake and beyond-design-basis tsunami, which is beyond design basis accident (BDBA); it is necessary to carry out relevant research on pressurized water reactor (PWR) nuclear power plant (NPP). Starting from the relations between various types of operating and accident conditions, this paper summarizes the BDBAs in the accident control guideline. During simulator training exercises with BDBAs including total loss to the steam generator feedwater and total loss of on-site and off-site electrical power supplies incident conditions, the evolution of nuclear safety-related parameters is studied, and the rules and the reasons of the evaluation are summarized. All this is helpful for the operator to master the principle of the accident.
Codes Comparative Study on Mechanical Properties of Core Forgings of Reactor Pressure Vessel
Authors: Yu Hao , He Yinbiao , Wu Xinxin , Gu Shenjie , Liu Xin , Wang Hongchang 
Published: November 2022

Abstract
In this paper, the mechanical properties of core forgings in Pressurized Water Reactor (PWR )NPPs in China were studied according to different rules of minimum yield strength at room temperature stipulated in ASME codes and RCC-M codes. First, this paper investigates the origin of two different codes for core forgings, then analyses the effects of yield strength, tensile strength, tempering temperature and toughness on the mechanical properties of forgings in both of codes. And then the results showed that the acceptance value of minimum yield strength at room temperature materials does not correspond to the mechanical design and analysis value. Finally, some suggestions for compiling the c’ore area forgings mechanical properties codes suitable for China’s PWR NPPs were given. This work provided a feasible implementation path for guiding the formulation of China’s PWR NPP codes.
Research and Practice on Environmental Dose Reduction in Nuclear Power Plant Core Instrument Room
Authors: Tang Long , Gao Binghan , Yu Long , Zhao Dongdong , Luo Feihua 
Published: November 2022

Abstract
Qinshan Phase II Units 3 and 4 began to use the solvent cleaning technology for flux thimble tubes in 2017. In 2018, the environmental dose of the core instrument room increased significantly, resulting in higher radiation protection costs in the core instrument room during the overhaul and absorption dose of equipment maintenance person, etc. After the abnormality of the environmental dose of the core instrumentation room was found, the investigation into the reasons for the increase of the ambient dose in the core instrumentation room of Unit 3 found that because the room was only designed with an air inlet and no air outlet, a large amount of radioactive dust escaped from the room. The neutron flux detector was rusted for the first time and the material analysis of the radioactive dust found the activation products Fe-59 and Co-60. Aiming at this problem that has never occurred in the field of nuclear power, this paper firstly analyzes and studies the reasons for the increase of the environmental dose in the core instrument room. On this basis, the four environmental dose reduction measures adopted are described, including the improvement of the cleaning method of the flux thimble tubes, the decontamination of the equipment surface, the overall cleaning of the mechanical equipment, and all-round cleaning of room radioactive dust. From the practical results, the ambient dose in the core instrument room was reduced from nearly 1mSv/h to about 0.1mSv/h, and the room surface pollution and equipment radiation hot spots were almost eliminated.
Study on the Plutonium Holdup Measurement in Glove-Box
Authors: Cheng Yimei , Xu Xiaoming , Yin Honghe , Li Xinjun , Wang Sijia 
Published: November 2022

Abstract
Based on passive total neutron counting with the 3He detectors, an assay system that can define the plutonium holdup in glove-box directly had been established through optimization design by Monte-Carlo simulation. The subsequent debugging for the electronics of this system ensured the system staying in a good operation state. The carried out experimental measurements had defined a series of parameters for the system, including the optimal operation voltage, detection efficiency for neutrons, spatial efficiency distribution, and the relationship between sample mass and the counts. Finally, according to the on-site measurements, the RSD of the measurement system for the glove-box holdup was 0.2%, which indicates the excellent stability of this system.
Affecting Factors and Aspects of Developing Nuclear Public Acceptance Strategies
Authors: Lin Haomiao 
Published: November 2022

Abstract
Nowadays, the issue of climate change is moved from margins to the mainstream. The important role of nuclear power in the global fight against climate change has been significantly highlighted by IAEA at COP26 last year. As a low carbon base-load electricity provider, nuclear power can became progressively an important option of countries for achieving its sustainable development goals. Comparing with traditional energy supply, nuclear power projects are characterized by complex technology, long construction period, large capital demand and high sensitivity. Currently, the international nuclear power demand market is mainly concentrated in the emerging countries, most of which have a poor industrial infrastructure, “not in my backyard” phenomenon is prominent, which is seriously interrupting the government’s energy development decision. Different national conditions and development phases will also have different impacts on public perception. This paper focuses on the analysis of the common characteristics of factors affecting public acceptance in different markets, and puts forward suggestions on public communication strategies, so as to provide references for improving public acceptance.