bluuebunny/crossref_metadata_2025_split huggingface dataset:
The Characterization of Radiation Response for MAPS
Authors: Qin Zhiwei , Xu Shoulong , Dong Hanfeng 
Published: November 2022

Abstract
This paper studies the response of γ-ray ionizing radiation under the irradiation of single active Pixel Sensor in the irradiation chamber and the Calibration Chamber, Designs Radiation Experiments, and analyzes and discusses the characteristic values in different statistical regions of pixels in the bright and dark environment, the linear relationship and dispersion degree between the statistical eigenvalue and the dose rate under different statistical algorithms are studied. The experimental results show that in the dark environment, the background gray values of any region in the image are in the range of 10–30, and the eigenvalue have a good linear response. Under the bright environment, the global region of the image will be confused with the neighboring region because of the different background gray values of the color region in the Byer array, which will affect the characterization of the dose rate. In low dose rate, the signal of radiation response is too small, dark current and background dark noise produce interference signal, so the signal-to-noise ratio is too low, and 300 frames of image data are needed to achieve the detection accuracy. In high dose rate, the response signal is strong, and 2 consecutive frames of image data are needed to achieve the detection accuracy. The characteristic value of threshold pixel at low dose rate is good, and the characteristic value of threshold Pixel and total gray value at high dose rate are good, which meet the requirement of wide range detection. In the radiation field experiment, we can choose the corresponding algorithm to detect the dose rate according to the field environment. This study provides theoretical basis and data support for improving gamma-ray dose rate detection technology based on MAPS image and realizing radiation field dose rate characterization. In order to use the static environment to do the radiation experiment, we can design an algorithm to cut the edge of the frame image, and use the computer image processing technology to choose the area which is advantageous to the dose rate Characterization, To realize radiation Detection in dynamic environment.
Effects of Stoichiometry and High-Temperature Annealing on Zirconium Carbide Coating Layer in TRISO Particles
Authors: Cheng Xinyu , Liu Rongzheng , Liu Bing , Yang Xu , Liu Malin , Chang Jiaxing , Shao Youlin 
Published: November 2022

Abstract
Very-high-temperature gas-cooled reactors (VHTR) are being developed to provide higher thermal efficiency and high-temperature process heat. Zirconium carbide (ZrC) has been proposed as a potential coating material for TRistructural-ISOtropic (TRISO) coated fuel particles because of its excellent resistance to fission products corrosion, good thermal stability and higher mechanical strength under elevated temperatures. The integrity and performance of the ZrC coating of the TRISO particles are very important as it provides the main barrier for fission product release. Therefore, the microstructure and property evolution of ZrC coating deserve to be investigated. Fluidized-bed chemical vapor deposition (FB-CVD) has been conducted to fabricate the ZrC coating in a ZrCl4−C3H6-Ar-H2 system. The stoichiometry of ZrC was changed by controlling the feeding rate of ZrCl4 and the flow rate of C3H6. The ZrC coatings were annealed from 1700 °C to 2200 °C to study the possible changes in microstructures and temperature-dependent performances. The effect of stoichiometries on ZrC coating was studied by X-ray diffraction (XRD), scanning electron microscopy (SEM), Raman spectroscopy (Raman), and nanoindenter. Results showed that free carbon prevents grain growth under high-temperature annealing, and it reacts with ZrC1-x at higher temperatures to form pure phase ZrC. In addition, the microstructure evolution mechanism of ZrC at high temperatures was proposed.
Research and Treatment of Periodic Fluctuation of Steam Feed Pump Speed Regarding 900MW Nuclear Power Plant
Authors: Li Heng , Liu Xiaoyu , Wang Xufeng 
Published: November 2022

Abstract
A 900MW nuclear power unit has been in commercial use for 20 years. The steam-driven feed water pump uses a regulating system constructed by an analog board to control the speed of the pump. Since the analog board has been discontinued, and its reliability is reduced. The power plant has carried out digital transformation on the control part of the speed control system. Shortly after the renovation, the speed of the steam feed pump continued to fluctuate. The phenomenon accompanying the rotational speed fluctuation is a periodic fluctuation similar to a sine wave. The range of fluctuations varies from time to time, and is interspersed with non-sinusoidal fluctuations. When one of the feed pumps is placed in the manual control state, the speed fluctuation of the other pump will increase. After analysis, with the aging of the mechanical parts of the feed pump, the follow-up ability of the pump continues to weaken. Under the combined action of the four loops of speed, W/S differ., SG level and feed water flow, continuous fluctuations occur in the four parameters.
Aiming at the fluctuation problem, this paper relies on the real-time monitoring data of nuclear power plants and Matlab/Simulink to establish a nonlinear model and “Accompanying system” that reflect the actual working conditions on site, and conduct in-depth research on the mechanism of the problem by using control theory, system identification and “Accompanying system”, and propose solutions to resolve fluctuations. This method provides theoretical guidance for reducing or eliminating rotational speed fluctuations, thereby avoiding potential safety hazards caused by multi-parameter fluctuations in the operation of nuclear power units.
Study on Risk of Nuclear Spent Fuel Marine Transportation
Authors: Yiqing Liu , Wei Wang , Xiaodong Zhao , Zhong Gui , Jiangtao Chen 
Published: November 2022

Abstract
The safety of hazardous materials marine transportation is an important aspect of marine transportation safety. As the seventh category of hazardous materials, spent fuel has strong radioactivity and high decay heat, which makes the marine transportation of spent fuel concerned by the national competent department. Risk analysis is based on the identification of risk factors, through qualitative or quantitative analysis methods, to analyze the possible consequences of these risks and the possibility of these consequences, so as to determine the corresponding risk level and formulate targeted risk control measures. The safety of marine transportation depends not only on the safety of the cargo itself, but also on the navigation safety of the ship. On the basis of introducing the steps and methods of risk analysis, this paper selects various factors affecting the safety of spent fuel marine transportation from the aspects of cargo itself and ship navigation safety, and establishes a scientific and reasonable spent fuel marine transportation safety evaluation index system; The risk factors of maritime transportation are evaluated by risk evaluation methods such as expert scoring and risk coordinate diagram, some measures and suggestions are put forward according to the evaluation results.
Dynamic Modeling and Reactor Power Control of a Natural Circulation Pool-Type Lead-Cooled Fast Reactor
Authors: Feng Yun , Sun Pei-Wei , Wei Xinyu 
Published: November 2022

Abstract
The primary circuit of a natural circulation pool-type lead-cooled fast reactor adopts full natural circulation cooling technology, the reactor core power regulation process has the characteristics of long regulation time. When developing the reactor power control system, it is necessary to fully understand the dynamic characteristics of the reactor, so it is necessary to establish the nonlinear model of the natural circulation lead-cooled fast reactor. Based on the requirements of reactor dynamic characteristics research and power control system development, a nonlinear model of primary circuit is established on MATLAB/Simulink platform. The dynamic characteristics of the reactor are analyzed by introducing the reactivity disturbance of the reactor core and the flow and temperature disturbance of the secondary loop coolant. The transient response trend of the system is reasonable and conforms to the physical law, which can be used to analyze the dynamic characteristics of the system and the design of the control system. The core power control system adopts load following mode, and the load power measured value is used as the set value of the core thermal power for power regulation. When introducing the load power step change, the core thermal power change trend is consistent with the load power, and the control effect is satisfactory. The research results of this paper can provide reference for the design of power control system of natural circulation fast reactor.
Safety Analysis About Combined Heat and Power Generation of Nuclear Power Plant
Authors: Liu Shiyao 
Published: November 2022

Abstract
Nuclear energy is one of the most efficient, clean and stable energy sources. The technology of combined heat and power generation, which is widely used in the conventional thermal power plants all over the world, is also applicable to the nuclear power plants to make full use of the nuclear energy. But at present, the mode of combined heat and power generation is rarely used in the nuclear power plants in China, concerning about the nuclear safety problem. This study aims at the optimum use of nuclear energy to raise the potential margin of nuclear power plants, under the premise of ensuring nuclear safety, and analyzes the influence of the combined heat and power generation (such as the external supply of steam generated by nuclear power plants) of a pressurized water reactor on nuclear safety, including the influence on the normal operation, design basis conditions, design extension conditions and transient analysis. It shows that combined heat and power generation of nuclear power plants has no negative impact on nuclear safety. In the meanwhile, the mode of combined heat and power generation can improve the economy of nuclear power plants and has positive significance for the efficient use of nuclear energy.
First-Principles Study on Adsorption Reaction of Oxygen Molecules on Fe (110) Crystal Surface
Authors: Xiaochuan Zeng , Xuejun Li , Cuizhu He , Qiaodan Hu 
Published: November 2022

Abstract
The adsorption reaction between oxygen (O2) molecule and ferrum (Fe) (110) crystal surface in the oxidation process of Fe surface was studied by using the first-principles method. The differential charge density analysis of the adsorption sites of oxygen molecule on Fe (110) crystal surface, the calculation of adsorption energy at different sites and the analysis of electronic density of states showed that the stable adsorption position of oxygen molecule was parallel to Fe (110) crystal surface, and the oxygen atom tended to adsorb at the triangular gap of Fe atoms. The electronic structure of the adsorption system showed that the 2p electron orbital of oxygen atom plays a major role in the adsorption, and only O-Fe electron interaction exists when oxygen molecule is adsorbed in the parallel orientation, which makes the whole Fe (110) crystal surface lose electrons, increase the system potential and the risk of electrochemical corrosion. The research conclusions can provide theoretical support for the further insight in the oxidation corrosion mechanism of nuclear metal surface.
Numerical Investigation on Heat Transfer Features of Gas-Cooled Open Lattice Reactor in Normal Operations
Authors: Fang Junlin , Sun Qian , Ji Yu , Ye Zishen , Sun Jun , Sui Zhe 
Published: November 2022

Abstract
The gas-cooled reactor combined with Brayton cycle energy conversion system has been investigated with the application potential to meet the requirement of large power nuclear reactor systems in deep space explorations. Conceptual core design configured in the open lattice has attracted extensive attention due to the compact structure, and its heat transfer features need further investigation, including conduction, convection, and radiation. Many studies on thermal-hydraulic characteristics of similar structure reactors such as Pressurized Water Reactor (PWR) and Sodium Cooled Fast Reactor (SFR) have been carried out. However, different operating conditions bring the distinct heat transfer features. Referring to Prometheus program, this paper analyses the heat transfer features of gas-cooled open lattice reactor by numerical simulations. The computational fluid dynamic tool (ANSYS Fluent) is mainly used to investigate the effect of circumferential rod heat conduction, thermal radiation among fuel rods, and gas flow distribution. The results show that contributions of conduction, convection and radiation to the total reactor core heat transfer should be taken into account. The maximum temperature on the walls can be lower by increasing the rod thermal conductivity, enhancing the thermal radiation among fuel rods, and increasing the gas flow of inner core. This study will help to develop the system code for the thermal-hydraulic design and safety analysis of gas-cooled open lattice reactors.
Study on the Policy and Legislative System of Spent Nuclear Fuel Management in Different Developed Countries
Authors: Shi Lei , Chen Chao , Liu Hongjun , Su Jiqiang , Zhang Honglin , Liu Qun , Li Yanrui , Hu Jian 
Published: November 2022

Abstract
Spent nuclear fuel is an inevitable product from the development of nuclear energy. Almost all of the fuel content is radioactive, and long systematic process are required for the safety management, which has always been an important global issue. In order to make sure that spent nuclear fuel should be safely managed in different countries developing nuclear power, IAEA is establishing a sharing system of spent nuclear fuel management by concluding joint conventions and issuing safety standards. For different countries, the United States, France and Russia with nuclear power have all established a complete policy and legislative system for spent fuel management. In the US, policy decision of open-cycle has been made, and no commercial reprocessing is being conducted. In France and Russia, closed-cycle strategy is implemented with industrial-scale reprocessing plant in operation. At present, China has become the country with the largest scale of nuclear power under construction in the world. There will be a large number of spent nuclear fuel requiring properly and safely managed. The lessons-learning of how developed countries managing spent nuclear fuel arising is important for China. The authors suggest that it is necessary to combine the top-level design to the legal practice, so that there are laws to respect during all steps of spent fuel management, and responsibilities of all parties are clear.
Die Design and Process Study of Zr-4 Alloy Profile Extrusion
Authors: Qi Peng , Gao Bo , Zhang Zhihao , Yin Jianming 
Published: November 2022

Abstract
According to the application requirements of Zr-4 profile with outer square and inner circle for small nuclear reactors, the extrusion die structure and extrusion process parameters are optimized through numerical simulation and calculation. Based on the study of alloy friction boundary conditions and heat exchange conditions, the influences of die inlet angle and working belt structure size on metal flow behavior were clarified. Combined with the analysis of related factors such as flow stress and temperature rise variation law in extrusion process, determine the structure size of the die and the process parameter range of the extrusion, use the on-site 1250T horizontal extruder to carry out the pilot scale experiment of profile extrusion, verify the results of the preliminary analysis and calculation, determine the optimal process technology plan, and trial-produce The required outer square and inner circle zirconium alloy profiles that every sizes and performances meet the design requirements. The cross-sectional size of this profile is 36mm × 36mm, the inner hole diameter is 20mm, and the length is 1500∼2500(mm). It is the first zirconium alloy profile produced by the extrusion process in China. This process can ensure the grains are broken under large deformation, and obtain profiles with good performance, uniform wall thickness and longer length. At the same time, it can also reduce material loss, improve production efficiency.
Study on Forging Technology of Integrated Topcover Forging for AP1000 Reactor Pressure Vessel
Authors: Li Zhaoyin , Dong Kai , Tao Zhiyong 
Published: November 2022

Abstract
AP1000 topcover forging, as a key component in the Reactor Pressure Vessel (RPV), is located at the top of the RPV and is an important reactor primary component within the pressure boundary of the primary circuit, and its integrity is of great significance to the safe operation of nuclear power equipment. The traditional topcover part is welded by three parts: the upper head, the topcover flange and the stacking tube socket. When the components are in service, the in-service inspection workload of the equipment is large, and the safety risk is high under complex working conditions. Therefore, the study of a large forging integrated top cover with a stacking tube socket is carried out to realize the upper head, the topcover flange and the stacking tube as integrated topcover. The integral forging of the integrated topcover is an urgent problem to be solved.
This paper mainly studies the development of forging process for topcover forging of AP1000 RPV with SA508 III material. Based on the shape characteristics and forging rheological characteristics of the topcover forgings, the topcover tube seat was identified as the key point of forming, so the concave round cake preform was designed, and the final forging was formed by spinning the preform. The preform forming simulation and the final spinning forming simulation were carried out by DEFORM software respectively. The simulation results show that the metal filling is good, the forging allowance is sufficient, but the fracture factor of the bottom head is high. In order to more comprehensively verify the feasibility of the forming scheme, Trial production of 1:20 small scale simulation parts was carried out, and the forming results showed that the small scale parts of the integrated topcover were well filled after spinning. Through the manufacture of theoretical and practical simulation parts, the forging process scheme of spinning preforms through two sets of dies is verified, which effectively solves the problem that the size of the top cover forgings is difficult to guarantee. It was manufactured according to the designed forging process using the 300-ton series of steel ingots, and the final product size and non-destructive testing met the requirements, and the qualified product was delivered.
Common Cause Failure Analysis and Countermeasure of Digital Instrument and Control System in Nuclear Power Plant
Authors: Wang Ping , Hu Lisheng 
Published: November 2022

Abstract
The design philosophy of defence in depth for nuclear power plant establishes in consideration of technical, artificial and organizational management failures. Defence in depth principle is a safety design idea that runs through all stages of production activities in the whole life of nuclear power plants. Each of defence lines must be reliable and independent of each other as much as possible. According to the system function and operation requirements of nuclear power plants, this paper analyses latest requirements of defence in depth and diversity line of nuclear power plants, based on the research and analysis of laws and regulations, standard and good practices at domestic and abroad. According to the requirements of independence, common cause fault and diversity, the independence and diversity of defence line in depth of instrument and control systems are analyzed, the weak points of common cause fault of defence line in depth of instrument and control system are identified, and the improved analysis method and improved realization steps are put forward. And then each of its lines of defence provided diversity for common cause failure. This paper discusses the types of common cause faults and typical countermeasures of defence in depth and diversity line for digital instrument and control system. This paper summarizes the common cause failure analysis methods, processes and countermeasures for digital instrument and control systems in nuclear power plant.
The Numerical Method Research for Simulating the SMR Steady-State Primary System Flow and Head Loss
Authors: Ma Zhongying , Li Ge , Wang Ting , Li Jinggang 
Published: November 2022

Abstract
In order to establish and predict the flow and pressure distribution of the reactor coolant system of SMR under steady-state operating conditions, so as to provide an important design input for the thermal-hydraulic design of the reactor. This study is based on the flow characteristics of the SMR with multiple flow paths and multiple nodes and hydraulics property, and two numerical methods are established for the hydraulic analysis of the circulation system. For the flow that can be simplified into the flow path and the node, the finite element analysis method is combined with the quasi-linear processing of the flow and the head loss to realize the flow distribution calculation of the circulation system; For the upper head bypass flow with a certain pressure distribution, the method of bisection is used to achieve calculation of such type flow. The research results show that the numerical method established can be used for the flow and pressure distribution of the reactor coolant system of SMR, and the numerical calculation results are in line with expectations.