bluuebunny/crossref_metadata_2025_split huggingface dataset:
Research on Electromagnetic Interference Protection of Nuclear Power Unit Equipment Under the Application of New Wireless Technology in Smart Nuclear Power
Authors: Chen Bing , Zhao Guidan , Zhong Zhiping , Zheng Yong , Yu Charlie 
Published: November 2022

Abstract
Intelligence is the trend of industrial development in today’s world, which will inevitably affect the nuclear power field, thus promoting nuclear power industry from digital to intelligent forward.5G and Wi-Fi6, as a new generation of wireless communication technologies, have important advantages in terms of technical performance and security due to the characteristics of dense personnel, wide equipment distribution, complex facility environment and special materials used in nuclear power scenarios. Considering the high requirement of electromagnetic compatibility in nuclear power environment, it is urgent to further study the electromagnetic interference and protection requirements of 5G, Wi-Fi6 and other new technologies on nuclear power unit equipment in nuclear power environment.
In this paper, combined with relevant national document standards and industry standards, we will sort out the requirements of Immunity about Radio Frequency Radiation (RFR) for the important equipment in nuclear power plants, and use signal generators to simulate new radio signals such as 5G and Wi-Fi6 to test the immunity of nuclear power instruments and meters. Combined the results of test, we will study the Electro Magnetic Compatibility (EMC) between new wireless technologies and nuclear power plant equipment from different factors, such as Power, bandwidth, modulation, distance, direction. And then, we will summarize the protection requirements about nuclear power equipment.
Research and Application of Automatic Modeling Technology for Mechanical Analysis of Nuclear Power Pipelines Based on PepS
Authors: Gong Xiao Cheng , Jin Xian Jing , Zhang Yu 
Published: November 2022

Abstract
Mechanical analysis of nuclear power pipeline is an important part of nuclear power design. At present, the widely used nuclear-grade pipeline mechanical analysis software in the world is PepS, which almost contains the commonly used standards and specifications in the nuclear power industry. PepS software is used to carry out the mechanical analysis of nuclear-grade pipeline. The key step is building a calculation input pipeline model that accounts for more than 70% of the entire pipe mechanical analysis workload. At the same time, in order to meet the strict process layout requirements of nuclear power, a pipeline usually needs to undergo multiple mechanical analysis and demonstration. There are thousands of pipelines in nuclear power projects, and its mechanical analysis needs to invest a lot of manpower cost. This research deeply analyzes the procedures and principles of PepS calculation input model, the topological structure and geometric characteristics of the 3D model of the pipeline based on the three-dimensional layout design results, and realizes the automatic modeling technology of mechanical calculation input model of nuclear power pipeline based on PepS software through digital and intelligent information technology. The implementation of this technology completely replaces the manual modeling, eliminates the human error of manual modeling, fundamentally liberates the productivity, obtains remarkable economic benefits, and helps the nuclear power design company go hand in hand with multiple projects to realize batch quantitative design and production.
Fault Diagnosis of CRDM Based on EEMD-Wavelet Threshold Denoising and WOA-KELM
Authors: Jingdong Luo , Ling Chen , Liming Zhang 
Published: November 2022

Abstract
CRDM (Control Rod Drive Mechanism) is the key equipment of nuclear reactors. Accurate fault diagnosis can improve the reliability of CRDM equipment and effectively and quickly eliminate related faults, thus ensuring the safety of reactors. However, due to its complex operating environment, the fault characteristics of CRDM are easily masked by complex and changeable noises, resulting in difficulty in fault identification. To solve this problem, this paper proposes a fault diagnosis method for KELM based on ensemble empirical Mode decomposition (EEMD), wavelet threshold denoising and whale optimization (WOA). Firstly, an EEMD-Wavelet threshold signal denoising method is proposed according to correlation criterion and kurtosis criterion. Then, the time domain feature, frequency domain feature, EEMD modal component energy and sample entropy feature of noise reduction signal were extracted to construct a multi-scale feature set that can accurately reflect the fault state. Finally, a classification method based on WOA-KELM is proposed to diagnose the fault feature set accurately. The proposed method is validated with CRDM vibration signal data set, and compared with common fault diagnosis methods, the results show that the fault diagnosis method based on EEMD-Wavelet threshold denoising and WOA-KELM has higher fault classification accuracy.
A MIN-TPN-Based Remote Disaster Recovery and Backup Scheme of Power Industrial Control System in Nuclear Power Plants
Authors: Gang Cui , Yang Xin , Li Hui , Wang Han 
Published: November 2022

Abstract
Nuclear power plants are important industrial production facilities whose safety issues can have disastrous consequences for the social environment. With the deep fusion of industrialization and information, the nuclear power control system is gradually transformed into automatic digital systems. Such development improves the production efficiency and economic benefits, but introduce the security problem from traditional IT systems into industrial control systems. However, the existing power industrial control system of nuclear power plant mainly adopts firewall, anti-virus software, network gate and other passive security strategies, which lack the safety of active. Besides, only part of the data can be backed up locally and the remote backup is completed manually by operators regularly. The security of industrial control systems cannot be guaranteed and real-time remote backup cannot be realized. In this paper, we analyze and summarize the characteristics of industrial control systems used in nuclear power plants. Then we propose a Truly Private Network (TPN) for nuclear power plants based on the Multi-identifier network (MIN), which integrates blockchain, cryptography, and other security mechanisms to provide a trusted environment. Lastly, we build a network remote real-time backup scheme based on Mimic distributed storage technology providing data tamper-resistance and traceability for nuclear power plants.
The Research on Welding Procedure and Performance of Zr-3 Thick Plate
Authors: Gao Yu , Liang Xiaoru , Wang Yuxin , Guo Lifeng 
Published: November 2022

Abstract
Due to the excellent corrosion resistance and small absorption interface of thermal neutrons, Zr has been widely used in nuclear industry. As an important manufacturing process, the welding of Zr is usually performed by manual tungsten inert-gas arc welding (TIG). Due to the low deposition efficiency of tungsten inert-gas arc welding (TIG), it is more suitable for thin plates. If this method is used for thick plates, it will increase the cost of time, manpower and material resources. Therefore, for thick plates, an advanced and efficient welding method is required to improve the welding quality and construction efficiency. In this paper, a combination of plasma arc welding (PAW) for root bead and tungsten inert-gas arc welding (TIG) for filling and cap beads was successfully developed for 20mm Zr-3 plate, with specifically designed clamp and gas shielding devices based on the welding characteristics of Zr. Different post-weld heat treatment was performed for the weld pieces. Moreover, the mechanical properties and corrosion resistance of welded joints were evaluated. The results demonstrates that welded joints which meet the requirements specified in code and standard and engineering requirements can be obtained by developed welding process and welding system; the influence of different heat treatment on mechanical properties and corrosion resistance of welded joints was not obvious.
Analysis Method and Application of Maintenance Rules for Main Steam System in Qinshan no.1 Nuclear Power Plant
Authors: Jin Yan Jun 
Published: November 2022

Abstract
Nuclear power plants have traditionally adopted regular preventive and corrective maintenance methods, but extensive international operation experience shows that inadequate maintenance, excessive maintenance and improper maintenance will adversely affect nuclear power plant structures, systems and equipment, and thus affect the safety level of nuclear power plant operation. According to China’s National Nuclear Safety Administration of the improved technology policy (try out) “the effectiveness of the nuclear power plant maintenance requirements, with the first nuclear power qinshan owner steam system as an example, the implementation based on the structures, systems and equipment effectiveness analysis method and risk management of the maintenance rules, can be more reasonable and effective to improve the reliability of maintenance.
The method mainly includes: determining the system boundary and the scope of structures, systems and equipment involved in the maintenance rules, screening the system maintenance rules function; List the equipment and failure modes that affect the function; The failure probability and function importance of each function group are calculated. The reliability index and availability index of the system are determined and verified by historical data. Supervise each function group of the system according to the key indicators, evaluate the actual operation of these structures, systems and equipment regularly, determine whether to meet the established performance indicators, optimize and adjust the maintenance rules according to the evaluation results, so as to avoid insufficient and excessive maintenance.
Development and Application of High Temperature and Radiation Protection Equipment for Spent Fuel Dry Storage System
Authors: Cheng Cheng , Yuanyuan Wei , Yongjie Pan , Yongjun Ge , Zhiguo Wei , Zhihui Zhou , Zengfen Li 
Published: November 2022

Abstract
Based on the safety analysis, the risk of high temperature and radiation are found in the local operation during the automatic welding and vacuum drying of typical spent fuel dry storage system. To enhance the operational safety, a high temperature and radiation protection equipment has been developed through the structural design and numerical analysis and has been applied in demonstration project in China. It is proved by the local monitoring data that in the case of loading spent fuels with a total heat rate of 26kW, with the help of the protection equipment, the highest temperature of the operational area is 60% lower than the case without the equipment in the same environmental conditions. The dose rate in the area can also be 60% lower. The equipment is proved to be of significant help to ensure the safety of operators in high temperature and radiation environments, during both normal operation and equipment failure conditions.
Analysis and Improvement of Flow Field Calculation in Thermal Hydraulic Program for High Temperature Gas-Cooled Reactor
Authors: Haoming Dou , Minggang Lang , Han Zhang , Ding She 
Published: November 2022

Abstract
The flow field calculation plays a crucial role in thermal-hydraulic calculation of High Temperature Gas-cooled Reactor (HTGR). While using the thermal-hydraulic program for HTGR to calculate the low-flow condition, the calculation of the helium pressure field distribution in the reactor core is difficult to converge, which may lead to the non-physical pressure oscillation. The analysis and improvement of flow field calculation to alleviate this non-physical oscillation is of great significance for HTGR thermal-hydraulic design. In this work, the factors affecting the flow field calculation are analyzed. The approximation assumptions of horizontal cavity modeling and the using of Gauss-Seidel (GS) partition point iteration method leading to the flow field calculation is sensitive to round-off error. For the reason of GS partition point iterative method suffers from the slow convergence rate, and may even non-convergence in some conditions, a global direct method is developed in this paper to solve the whole flow field directly to pursue strong stability and high convergence rate. Compared with original method, the numerical results show that the newly developed algorithm achieves faster convergence rate and stricter convergence results.
Research and Application on Drawing Method of Optimizing for the Intelligent Process System Flow Chart of the Nuclear Power
Authors: Chen Haiyan , Tian Licheng , Zhou Zhigang , Yang Jiaojiao 
Published: November 2022

Abstract
The intelligent process system flow chart of the nuclear power is based on the symbol library. Each component and its attributes on the system flow chart are stored in a digital form for reference to subsequent projects. The content of this paper is the automatic generation of Definitive Data Assembly and the automatic association of the drawing interface of the intelligent process system flow chart by using Diagrams. It is an optimization of the intelligent process system flow chart for the application of Diagrams by using information technology and computer language. The Definitive Data Assembly automatic generation function realizes generating Definitive Data Assembly statistical table at a fixed position of the drawing surface, and improves the neat specification of the drawing information; The single drawing Definitive Data Assembly statistical table is automatically generated, and the batch drawing Definitive Data Assembly statistical table can be automatically generated, one-click operation, which greatly improves the work efficiency; at the same time, solves the problem of disordered management of Definitive Data Assembly. Optimized the application of Diagrams to draw the interface merging function of the intelligent process system flow chart, which greatly reduced the number of files in the project file XXXMAC-&gt;Schematic path, reduced the selection obstacles and improved the correct rate; Developed interface-connected menus, one-click operation, saving time for drafters, automatically deleting MAC files, reducing the workload of project administrators, and releasing project storage space, improving work efficiency. The research results of this paper are applied to draw the process system flow chart of many specialities such as nuclear island system, radioactive waste management, hydraulic technology, HVAC, chemical water, conventional island machinery, etc.
Analysis of EBR-II Loss-of-Flow Transients With 3D Model of CATHARE Code
Authors: Chen Xiaolin , Li Simon , Cheng Songbai 
Published: November 2022

Abstract
As one of the GEN-IV nuclear reactor systems proposed by GEN-IV International Forum, the sodium-cooled fast reactor features the energy generation with fast spectrum and a closed fuel cycle for fuel breeding and actinide management. In order to validate system codes for reactor system analyses and to improve code performance, benchmark activities of experimental transients of fast reactors were performed by worldwide institutions. The Shut-down Heat Removal Tests in Experimental Breeder Reactor II were proposed in the context of the International Atomic Energy Agency. In this work, loss-of-flow tests performed in Experimental Breeder Reactor II are analyzed using system thermal-hydraulic code CATHARE. Evolutions of important reactor parameters such as core power, coolant flow rate, temperatures in core and in the intermediate heat exchanger are predicted and compared against experimental data. The sodium pool is further modeled by using a 3D model in CATHARE-3 and its verification is also conducted in the benchmark. Qualitative agreement is obtained for the reactor parameters predicted by CATHARE. The inherent passive safety characteristics of Experimental Breeder Reactor II in unprotected loss-of-flow transient is demonstrated, and the stratification effects in sodium pool are able to be identified with the 3D model. The results obtained in this work are expectable to provide some valuable insights for future sodium-cooled fast reactor transient accident analyses by coupling system thermal-hydraulic codes and CFD tools.
Optimization the Protection Configuration of the RAM Excitation System of the Nuclear Power Plant
Authors: Lin Wang , Yan-jun Zhao , Xing-Zhen Zhang 
Published: November 2022

Abstract
The RAM system is the reactor control rod drive mechanism power supply system, which is the only power supply designed for the control rod drive mechanism. The RAM (control rod Power System) system of the Nuclear Power Plant only has two synchronous generators (RAM001/002AP) operating in parallel. On November 14, 2019,when one unit has an excitation system failure, due to the unreasonable setting of the protection value of the two units, the over-current protection of the unfailed unit precedes the loss of field protection action of the field lost unit, resulting the two generators in the RAM system tripping one after the other, therefore causing CRDM losing all power supply, resulting in the control rod dropping off and finally causing the automatic shutdown of the reactor, bringing serious consequences to the reactor and power grid. Therefore, the setting of synchronous generator protection in the RAM system is very important for the safe and stable operation of nuclear power plant. This report carries out the digital modeling and simulation of the RAM system of the Nuclear Power Plants, simulates various faults of RAM system synchronous generator RAM001/002AP and provides the important basis for verifying the settings of over-current protection and loss of excitation protection.
Pilot Application of RISMC Methodology on Specific SBLOCA Sequences of Passive NPP
Authors: Du Yun , Zhang Qinfang 
Published: November 2022

Abstract
In this paper, a typical passive nuclear power plant small break LOCA is taken as the research object. Based on the theory of Risk Informed Safety Margin Characterization methodology, the development process of the accident is studied. The specific sequence is selected to be studied. The important systems are discretely branched (such as Automatic Depressurization System), and the important uncertainty parameters are sampled (such as ADS valve and IRWST valve resistance coefficient). The event tree is re-established, and the success criterion of the head is modified. The fault tree is reestablished. After that the uncertainty parameters are sampled for specific sequences and the accident simulation of the whole plant is carried out for each condition. Thus, the frequency of each sequence and the conditional failure probability under the specific condition are obtained, and the core damage frequency (CDF) value based on RISMC method is finally calculated.
This analysis is mainly aimed at the configuration and sensitivity of ADS. The selected sequence 45,53,64 are generated from the risk significant core damage sequence S in the original model. Through the analysis and calculation of RISMC method, it is concluded that sequence 45 is ok sequence, and the assumption in the original model is too conservative. OK cases with epidemic proportions existed in sequence 53; sequence 64 is consistent with the original hypothesis, which is a CD sequence. Therefore, the calculated value of CDF is reduced to a certain extent.
This paper verifies the feasibility of RISMC method in application on passive NPPs and also proves that the method can remove unnecessary conservatism. more realistic assessment of the risk helps us quantify the safety margin of nuclear power plants better.
Numerical Study of the Effect of Different Tank Shapes on Liquid Sloshing Characteristics
Authors: Zhu Yuxuan , Lu Daogang , Li Donghao , Liu Yu 
Published: November 2022

Abstract
In nuclear power engineering, there are many tanks with irregular shapes (e.g., the cooling water tank of the containment, etc.), whose shapes are quite different from the traditional rectangular tanks. Under seismic excitation, the sloshing of the liquid in the tank produces the fluid-structure interaction (FSI) with the structure. It is crucial to study the sloshing characteristics of the water tank during the design of the reactor structure. There is a difference between the sloshing characteristics of irregularly shaped tanks and rectangular tanks, but there are few studies on this difference and the significance of the difference needs to be further investigated. To explore the aforementioned issues, based on the fluid-structure interaction theory and Housner’s equivalent mechanical model, the finite element models of the irregularly shaped water tank and the regular water tank (rectangular) are established by ANSYS software. The effect of the tank shape on the sloshing frequency and added mass of water in the tank is studied by modal analysis and calculation of the added mass. Further, the significance of the effect of tank shape on the natural and dynamic characteristics of tanks is investigated by comparing the variability of the sloshing frequency and the added mass. The results can guide the fluid-structure interaction problem in nuclear power engineering and provide a reference for the structural design of similar irregular water tanks, thus improving the safety and economy of reactor design.
Research of the Standards Compilation of Welding Material for Carbon Steel in Pressurized Water Reactor Nuclear Power Plants
Authors: Lin Bai , Xiaoru Liang 
Published: November 2022

Abstract
Carbon steel welding is a major manufacturing method for nuclear power construction. For the moment, there are two sets of standards of welding material for carbon steel which used in pressurized water reactor nuclear power plants of China. The coexistence of standards of two different technical routes brings unnecessary cost and waste to pressurized water reactor nuclear power plant manufacturing. With the rapid development of pressurized water reactor nuclear power plants in China, the research of developing a set of unified standards for carbon steel electrodes and rods for gas shielded arc welding and carbon steel covered electrodes is necessary. At present, two main standards will be revised and published. Through the interpretation of “Welding material for pressurized water reactor nuclear power plants — Part 1 Carbon steel covered electrodes” and “Welding material for pressurized water reactor nuclear power plants — Part 6 Carbon steel electrodes and rods for gas shielded arc welding” and the introduction to their compilation thoughts and technical points, the article guides engineers to understand the compilation thoughts of these two standards quickly. The standards specify classification, technical requirements, test methods and testing of carbon steel electrodes and rods for carbon steel in pressurized water reactor nuclear power plants.