bluuebunny/crossref_metadata_2025_split huggingface dataset:
The association between perioperative pupillary parameters and postoperative acute pain: A pilot cross‐sectional study
Authors: Fang Pan‐pan , Wu Jun‐qi , Tang Li‐li , Gao Shan , Liu Xue‐Sheng 
Published: October 2021

AbstractObjectiveWe aim to explore the capacity of perioperative pupillary variables to predict acute pain in the post‐anesthesia care unit (PACU).MethodsPatients scheduled to undergo thoracic or abdominal surgery under general anesthesia between April 2021 and June 2021 were enrolled. We measured the pupil diameter, pupillary light reflex (PLR), and pupillary reflex dilatation 5 min before anesthesia induction (T1), 5 min after intubation (T2), at the end of anesthesia (T3), immediately before extubation (T4), and 5 min after extubation (T5). We assessed the early postoperative pain intensity in the PACU using Numeric Rating Scales (NRS) at recovery, 5 min after recovery, and 10 min after recovery. Logistic regression models were used to evaluate the association between perioperative pupillary variables and postoperative pain intensity.ResultsFifty‐one patients were enrolled, 50 of whom were included in the final analysis. A total of 13 patients (26%) needed remedial analgesia in the PACU. Pupil parameters at T1, T2, T3, and T5 were not associated with NRS in the PACU. Multiple logistic regression models and receiver operating characteristic (ROC) curves indicated that only latency of PLR at T4 can predict postoperative acute pain. The ROC analysis showed that the cutoff value for latency of PLR at T4 was 0.29 s to discriminate between no pain and pain, and the area under the curve was 0.778 (95% CI 0.634–0.922, p = 0.002) with sensitivity 50.0% and specificity 91.7%.ConclusionThe latency of PLR immediately before extubation may be a useful predictor for postoperative acute pain in the PACU.
Genome‐wide association study reveals that the cupin domain protein OsCDP3.10 regulates seed vigour in rice
Authors: Peng Liling , Sun Shan , Yang Bin , Zhao Jia , Li Wenjun , Huang Zhibo , Li Ziyin , He Yongqi , Wang Zhoufei 
Published: October 2021

SummarySeed vigour is an imperative trait for the direct seeding of rice. In this study, we examined the genetic regulation of seedling percentage at the early germination using a genome‐wide association study in rice. One major quantitative trait loci qSP3 for seedling percentage was identified, and the candidate gene was validated as qSP3, encoding a cupin domain protein OsCDP3.10 for the synthesis of 52 kDa globulin. Disruption of this gene in Oscdp3.10 mutants reduced the seed vigour, including the germination potential and seedling percentage, at the early germination in rice. The lacking accumulation of 52 kDa globulin was observed in the mature grains of the Oscdp3.10 mutants. The significantly lower amino acid contents were observed in the mature grains and the early germinating seeds of the Oscdp3.10 mutants compared with those of wild‐type. Rice OsCDP3.10 regulated seed vigour mainly via modulating the amino acids e.g. Met, Glu, His, and Tyr that contribute to hydrogen peroxide (H2O2) accumulation in the germinating seeds. These results provide important insights into the application of seed priming with the amino acids and the selection of OsCDP3.10 to improve seed vigour in rice.
Intelligent Model for Predicting Downhole Vibrations Using Surface Drilling Data During Horizontal Drilling
Authors: Saadeldin Ramy , Gamal Hany , Elkatatny Salaheldin , Abdulraheem Abdulazeez 
Published: October 2021

Abstract
Drillstring vibration is a major concern during drilling wellbore, and it can be split into three types: axial, torsional, and lateral. Many problems associate with the high drillstring vibrations as tear and wear in downhole tools, inefficient drilling performance, loss of mechanical energy, and hole wash-out. The high cost for the downhole measurement of the drillstring vibrations encourages machine learning applications toward downhole vibration prediction during drilling. Consequently, the objective of this paper is to develop an artificial neural network (ANN) model for predicting the drillstring vibration while drilling a horizontal section. The ANN model uses the surface drilling parameters as model inputs to predict the three types of drillstring vibrations. These surface drilling parameters are flowrate, mud pumping pressure, surface rotating speed, top drive torque, weight on bit, and rate of penetration. The study utilized a data set of 13,927 measurements from a horizontal well that was used to train the ANN model. In addition, a different data set (9284 measurements) was employed to validate the developed ANN model. Correlation coefficient (R) and average absolute percentage error (AAPE) are statistical metrics that are used to evaluate the model accuracy based on the difference between the actual and predicted values for the axial, torsional, and lateral vibrations. The results of the optimized parameters for the developed model showed a high correlation coefficient between the predicted and the actual drillstring vibrations that showed R higher than 0.95 and AAPE below 3.5% for all phases of model training, testing, and validation. The developed model proposed a model-based equation for real-time estimation for the downhole vibrations.
Manufacturability Estimation on Burnable Poison Mixed Fuel For Improving Criticality Safety of HTGR Fuel Fabrication
Authors: Hasegawa Toshinari , Fukaya Yuji , Ueta Shohei , Goto Minoru 
Published: October 2021

Abstract
Manufacturability of the BP-mixed uranium dioxide (UO2) kernel for burnable poison (BP) credit concept was estimated to ensure criticality safety in 14 wt% enrichment fuel fabrication facility for commercial high temperature gas-cooled reactor (HTGR). The case of the coating process has not been examined, so this study is different from manufacturability estimation on Tri-ISOtropic (TRISO) coated fuel. As a BP, boron, gadolinium, erbium, and hafnium were investigated. It is essential that mixing BP does not affect the shape of the fuel kernel, and that BP does not melt or vaporize during the heat treatment in the reduction process. We determined to add BP as a nitrate by considering the existing kernel fabrication process. The results of thermodynamic equilibrium analysis showed boric acid would melt and vaporize during the heat treatment processes. On the other hand, it was found that gadolinium, erbium, and hafnium nitrate would change to solid oxides that do not melt even at 2000 °C. These results indicate the existing external sol-gel method could be applied for the fabrication of the UO2 fuel kernel mixed with gadolinium, erbium, or hafnium.
Flow Induced Vibration and Fretting Wear Characteristics of Fuel Rods
Authors: Feng Zhipeng , Qi Huanhuan , Huang Xuan , Chen Guo , Liu Shuai , Zhang Yixiong 
Published: October 2021

Abstract
Fuel assembly is located in the reactor core, which is the key component of nuclear power plant reactor. Its structural integrity will directly affect the safety and reliability of nuclear reactor operation. Fretting wear between fuel rod and grid supports caused by flow induced vibration is currently one of the main causes of fuel rod failure, which is also the most concerned phenomenon in fuel assembly design. Based on the random vibration theory and considering the interaction between fuel rod and support grid, a theoretical model and evaluation criteria for response prediction are proposed for the three possible flow induced vibration mechanisms in Pressurized Water Reactor (PWR) fuel assemblies, namely, turbulent excitation, fluidelastic instability and vortex shedding. At the same time, based on the Archard’s equation, the fretting wear analysis model of fuel rods is established, which is suitable for industrial application. The effects of transverse velocity, axial velocity and fluid density on the response of flow induced vibration and fretting wear of fuel rods are studied. The results show that the amplitude/wear depth varies with the transverse velocity in a quadratic way, exponentially with the axial velocity, and linearly with the fluid density. The established analysis process and evaluation method can provide technical support for the design and optimization of fuel assembly.
Core Thermal-Hydraulic Analysis During Dipped-Type Direct Heat Exchanger Operation in Natural Circulation Conditions
Authors: Hamase Erina , Doda Norihiro , Ono Ayako , Tanaka Masaaki , Miyake Yasuhiro , Imai Yasutomo 
Published: October 2021

Abstract
In order to enhance the safety of sodium-cooled fast reactors, a direct reactor auxiliary cooling system (DRACS) under natural circulation conditions with a dipped-type direct heat exchanger (D-DHX) in an upper plenum of the reactor vessel has been investigated. During the DRACS operation, the complicated thermal-hydraulic phenomena that cold coolant from D-DHX flowed into the fuel subassemblies and narrow gaps between them, which is well-known as inter-wrapper flow (IWF) was observed. Therefore, a multi-dimensional thermal-hydraulic analysis model in the reactor vessel for computational fluid dynamics (CFD) code (RV-CFD model) has been developed to evaluate the core cooling performance under natural circulation conditions. For the design study, the RV-CFD model is demanded to simulate with reasonable calculation costs while maintaining accuracy. In this paper, the application of the subchannel analysis method by CFD code for fuel subassemblies (subchannel CFD model) to the RV-CFD model was attempted. In the subchannel CFD model, the porous media approach was used to consider local geometry in the fuel subassembly, and the effective heat conductivity coefficients in diffusion term of the energy equation were set to fit the actual radial thermal diffusion between subchannels. Analysis results were compared to the experimental data obtained in the sodium experimental apparatus PLANDTL-1 and the calculated sodium temperature in the core had good agreement with the experimental result. It was confirmed that the RV-CFD model with subchannel CFD model was applicable to the core thermal-hydraulic analysis during the DRACS with the D-DHX operation under natural circulation conditions.
The Development of the NPP Nuclear Emergency Drilling Assistant System
Authors: Yanfang Chen , Xueyan Hou , Fei Chao , Longze Li , Chuan He , Wen Yang 
Published: October 2021

Abstract
The study carries out the rapid simulation calculation by using the integrated severe accident analysis program MAAP, performs the 3D modeling of nuclear power plant with 3DMax software, and implements the effect rendering with Unity3d. A set of nuclear emergency drilling assistant system has been developed to meet the requirements of nuclear emergency drilling scripting and the decision-making support for emergency drilling in emergency headquarters. It realizes the super real-time simulation calculation of nuclear accident, generation and storage of nuclear emergency drilling scenario database, generation of nuclear emergency drilling script information, and real-time scenario reconstruction. The application of the system can effectively reduce the difficulty and workload of nuclear emergency drilling scripting, and improve the efficiency and credibility of the scripting. Additionally, it can vividly present the main phenomena and processes of nuclear accident, and provide the real-time dynamic data under the whole range and all working conditions for nuclear emergency drilling.
Key Element Analysis and Suggestion for Strengthen the Quality Management of Nuclear Power Plant Fasteners
Authors: Lu Yan , Ling Ligong , Xu Yu , Gao Chen 
Published: October 2021

Abstract
This article uses the letters and notices issued by the China National Nuclear Safety Administration on fastener quality incidents as the main line to review the process of fastener quality incidents from clues, investigation to handling. Based on the problems encountered in the process of handling this incident, the improvement direction and improvement measures for strengthening the quality management of fasteners were analyzed from the elements of quality assurance, design, procurement, manufacturing, installation, non-conformance management, inspection and testing, and documents. The design of nuclear power plants should clarify the classification and marking of fasteners, and the design of fasteners should be as universal and standardized as possible, and specifications and models should be reduced. Procurement should focus on supplier evaluation, limit the lowest bidding price, prevent cutting corners, etc.; manufacturing should set up witness points for key links, inspect the quality assurance system while witnessing; pay attention to the factory inspection of fasteners, and it is recommended to increase the re-inspection of fasteners before storage or use. Fasteners managed by equipment category are not recommended to be classified as bulk materials, but they still need to be re-inspected. Fastener re-inspection standards should take into account the requirements of different reactor types and different specification systems. Re-inspection should not only focus on mechanical properties, but also pay attention to the re-inspection of fastener surface quality. The re-inspection unit shall have a certain degree of credibility or be bound by nuclear safety regulations. The idea of establishing a professional nuclear industry fastener supply and testing platform was proposed.
The Development Status of Decommissioning Technology of Nuclear Facilities: An Insight From Patents
Authors: Zhang Yading , Mo Dan , Su Ran , Zhong Haoliang 
Published: October 2021

Abstract
The decommissioning technology of nuclear facilities has been one of the most important research directions of nuclear industry for several decades, and further enhancements are being carried out in the world. The global published patents about the decommissioning technology of nuclear facilities are investigated in this work. The technology forecasting and the development strategy of the decommissioning technology of nuclear facilities was provided from the view of patents, based on the nations, the technology field, the trend in the number of patents applications. This work will contribute to understanding the advanced developments of the decommissioning technology of nuclear facilities by a new perspective. Through quantitative statistics and content analysis, the understanding of relevant industries and technology development trend of personnel in the industry is promoted from the perspective of open patent data, a complex of technology, law and economy. Through the technology and economy behind the patent data to ensure its reliability and authenticity, this paper puts forward suggestions and prospects for the practitioners in the field of nuclear facilities decommissioning from the aspects of how to combine the research and analysis of patent situation with the development of the industry.
Aerosol Removal by a Heat Exchanger of Passive Containment Cooling System
Authors: Liang Yangyang , Lu Junjing , Zhang Tianqi , Han Xu , Yuan Yidan 
Published: October 2021

Abstract
The aerosol removal mechanism and capacity related to passive containment cooling system (PCS) for PWR reactors during the severe accident is an important and interesting topic among researchers, which can enhance the natural removal of suspended aerosols, reducing the radioactivity that may leak into the environment. In this study, the heat transfer and aerosol deposition models of a single heat tube of PCS heat exchanger were established. The four natural removal mechanisms (gravity deposition, Brownian diffusion, diffusiophoresis, and thermophoresis) on the tube surface were focused on. The results show that the diffusiophoresis and thermophoresis deposition velocities decease along the flow direction of cooling water inside the heat tube. The maximum total removal coefficient locates at the position that directly above the tube inlet. Aerosol particles with large diameter cannot deposit on the bottom of the heat tube under the certain thermal-hydraulic condition. With the continuing operation of PCS, the total removal coefficient for a PCS heat exchanger decreases rapidly at the beginning and then decreases slowly, while the diffusiophoresis effect is always the dominant for the natural removal.
A Study of Heat Transfer and Flow Characteristics Under Non-Uniform Thermal Boundary Condition
Authors: Wang Qiang , Xu Yuting , Wang He 
Published: October 2021

Abstract
In the industry, most of the heating equipment of the thermal system could not achieve complete constant heating in theory. It is more complicated in nuclear power system. Even if the effect of burn up life is not considered, the distribution of power still presents a special function distribution. However, in order to simplify the problem, the heat release unit is often replaced by uniform heat flux. In the process of system characteristics research, the effect of local heat transfer usually will not have a great impact on the overall performance of the device. But from the perspective of safety design, once the local heat pipe is burned out, it will directly restrict and determine the design limit of the whole device. Therefore, it is of great engineering significance to study the characteristics of unconventional thermal boundary conditions.
In order to reveal the principle of heat transfer character under specific boundary conditions, a series of investigations are conducted. Based on the conventional test section with uniform heat flux, steady-state and transient conduction differential equations are derived theoretically and extended to the experimental channel with axial cosine heating. The theoretical deduction results indicate that the amplitude and phase difference of wall temperature are influenced by Fourier number when flowrate oscillates. Meanwhile, numerical simulation is conducted on heating tube with different thermal boundary conditions. The wall temperature field of axial nonuniform heating channel is studied under steady and unsteady flow. It includes the position of peak wall temperature, excess temperature, and the non-linear distribution of fluid temperature along the channel. The results indicate that a great difference exist between uniform and non-uniform heating flux. The simulation of heat and flow characteristics in reactor core is not suggested to use conventional test channel with uniform heat flux. The related investigations could lend some experience to nuclear safety design.
The Safety Analysis of the Design of the Reactor Coolant Pump Heat Shield in Qinshan Nuclear Power Plant
Authors: Wang Zilong , Liu Deyi , Zhao Ming , Wang Li , Xu Jie 
Published: October 2021

Abstract
The design of the reactor coolant pump heat shield of Qinshan Nuclear Power Plant in China National Nuclear Power Co., Ltd. is a weakness in the design of the EDF standard 900MW nuclear power plant. When the reactor coolant pump heat shield is broken, and the single failure of the equipment cooling water system valve (RRI219VN (220VN) or RRI225VN (226VN) valve) is considered, it is inevitable that the high-pressure coolant will damage the equipment cooling system (RRI), resulting in a LOCA accident of the containment bypass type. The Unit 3 and Unit 4 of Qinshan Nuclear Power Plant were improved during the construction stage, and the upstream of the heat shield was changed to 2 check valves and the downstream was 2 electric valves. The modification hasn’t been performed in Unit 1 and Unit 2 of Qinshan Nuclear Power Plant. Qinshan Nuclear Power Plant has entrusted the research institute to carry out the plan design, and the on-site verification of the plan was limited by the on-site space and could not be implemented. There are also factors such as a large amount of work to be carried out by the main control room board replacement, and insufficient time for outage implementation. This article makes a safety analysis of this technical modification using PSA method to determine the necessity of the technical modification.
Eigenvalue Analysis of Well-Posedness of Two-Fluid Single Pressure Model With Virtual Mass Force and Interfacial Pressure
Authors: Chao Fei , Yang Wen , Li Longze , Qiu Jinrong , Shan Jianqiang 
Published: October 2021

Abstract
It is well know that the two-fluid single pressure model is currently widely used to analyze reactor transient accidents such as LOCA. Current mainstream reactor safety analysis codes such as RALAP5 and CATHARE are based on this single pressure model. However this model has been proved to be ill posed in the sense that the equation system is non-hyperbolic which will lead to numerical unphysical oscillation. Currently reactor safety analysis codes use the interfacial pressure and virtual mass force to solve the ill-posed non-hyperbolicity issue. The development of China’s self-owned reactor high fidelity system analysis code has attracted much attention in recent years, and studying on the ill-posedness of the two-fluid model and the improvement of the ill-posedness is an important basis to analyze these reactor accidents. In this paper, the ill-posedness regions of the two-fluid single pressure model is first investigated using the eigenvalue analysis method based on the Cauchy problem with initial conditions. Then the effect of the virtual mass force and the interfacial pressure is studied by this eigenvalue analysis method, and three types of virtual mass term in the momentum equation are discussed. The results show that the appropriate virtual mass force and interfacial pressure can well improve the ill-posedness of the single pressure model, and the appropriate combination of them can significantly improve the ill-posedness.
Modeling and Sensitivity Analysis of the Sodium-Water Reaction Accident in Parallel Channels
Authors: Luo Gang , Sun Peiwei , Bai Xi , Cao Huasong , Wang Kai , Zhu Huanjun 
Published: October 2021

Abstract
A pressure wave propagation across the second loop of a sodium-cooled fast reactor may lead to severe damage to the pipes and the equipment due to a large leakage sodium-water reaction accident. Therefore, the pressure source and the pressure wave propagation calculation and analysis can be significant for a sodium-cooled fast reactor’s design and operation.
A mathematical model with code was built to calculate and predict both the pressure source and the pressure wave propagation after the large leak sodium-water reaction accident occurred in an SFR steam generator.
In the pressure source model, the sodium-water reaction produces hydrogen and squeezes sodium nearby. It is assumed that the sodium-water reaction carries out instantaneously, and a spherical-to-column bubble is assumed to grow up at the reaction zone. Under such assumptions, the pressure source model can obtain the hydrogen bubble’s volume and pressure in the steam generator.
In the pressure wave propagation model, flow is assumed to be one-dimensional. With the basic theory of mass conservation, momentum conservation, and the Isentropic relation of the working fluid, the numerical solution is calculated by a grid divided by the method of characteristics, considering both precision and efficiency.
However, plenty of parameters, both of the pressure source and the parallel channels, can affect the accident in different ways and scales. Considering that to practice a large leakage sodium-water reaction experiment can be somewhat difficult because of its danger, inconvenience, and high cost, it is quite essential to conduct a sensitivity analysis through the program for the purpose of facilitating the design of an SFR.
Since most sodium-cooled fast reactor designs are multimodules, a parallel channel structure with multiple pressure boundaries is modeled for sensitivity analysis. Sensitivity analysis was performed on multiple sets of important parameters, including different hydrogen bubble properties, different pressure boundaries, and different equipment types. Following the sensitivity study, the code can be used for the accident analysis in a future commercial demonstration fast reactor by indicating various parameter guidance for the design and construction.