bluuebunny/crossref_metadata_2025_split huggingface dataset:
Simulation Research on Safe Load-Reducing Operation Margin of Nuclear Power Unit With Clogging of the CFI Drum Net
Authors: Liu Jinchao , Lin Yaozu , Chen Bijun , Lyu Xue , Li Jiayi , Zhang Heng 
Published: November 2022

Abstract
As the final heat sink of the nuclear power plant, the cool source is responsible for the safety function of the waste heat export of the reactor. The explosive invasion of marine organisms and other wastes have a great impact on the cold source system of nuclear power plants, which will lead to transient events, load reduction and shutdown of nuclear power unit, and even affect the availability of the cold source system in severe cases. which are huge threats and challenges for the safe and stable operation of unit. thereby reducing the power generation efficiency, safety, economy, and sustainable development of nuclear power plants. Clogging of the CFI drum net is one of the important factors causing cold source system accidents. this paper puts forward the solution idea of maintaining the water level of the forebay through the variable frequency pump, so as to ensure that the CRF pump keeps running state under the condition of cold source blockage, so as to maintain the function of the cold source system. and based on the full range simulator of nuclear power units, consider CFI drum network differential pressure high alarm failure combined with a monitoring device for the blockage degree of the drum net, the scheme is simulated and verified, and the load that the nuclear power unit can operate safely under different blocking degrees of drum network is given. the research results of this paper provide a solution for alleviating the cold source problem, and the solution can also provide a reference for the safety improvement of the cold source of actual unit and the design of new reactor.
3-D Radiation Field Reconstruction for a Facility With Multiple Radioactive Sources
Authors: Zhu Shangzhen , Dong Xinwen , Zhuang Shuhan , Fang Sheng , Cao Jianzhu , Li Wenqian 
Published: November 2022

Abstract
The radiation field measurement and surveying play important roles in optimizing and planning operation work in the radioactive area. In this work, a 3-D radiation field reconstruction method is applied to reconstruct a 3-D radiation field of a facility with two radioactive drums and a shielding, not limited to the existing methods under two-dimensional grids. A numerical scheme of the method is introduced. Monte Carlo simulation of the gamma radiation field was performed to be an original field so that the reconstruction results can be verified. Random and regular sampling ways are taken into consideration while the sampling rates are kept at 4.12%. Quantitative evaluation of the results evaluates MSE (mean relative error) and MRE (mean squared error), MRE is kept at less than 4%, which shows a good reconstruction accuracy of the method in almost all 3D space. The work in this paper has good reference value for applications such as radiation field detection, inversion, and reconstruction, operations under radioactive environments like nuclear waste decommissioning, and nuclear power radiation modeling.
Simulation of Neutron and Proton Displacement Damage in Ultra-Wide Bandgap Semiconductor Ga2O3
Authors: Shao Zhuang , Cai Ziqi , Zhang Qingmin 
Published: November 2022

Abstract
Recently, gallium oxide (Ga2O3) as an ultra-wide bandgap oxide semiconductor has aroused enormous attention in research and development due to its prospects for future power electronic, optoelectronic, and radiation detection applications. Ga2O3-based devices could be subject to fluxes of protons or neutrons if used in aerospace or radiation-hard nuclear systems, which leads to internal defects in Ga2O3 crystals and the degradation of device performance. Therefore, the radiation effects of Ga2O3 under irradiation have become a research focus, and it is of great significance to study the defect behavior and performance influence of Ga2O3 after irradiation. In this paper, the number of displacement defects produced by 1∼20 MeV neutrons in Ga2O3 were calculated using Geant4 simulations, and the factors that impact the displacement damage were studied. The results show that the displacement defects generated by neutrons do not increase monotonously with neutron energy but are closely related to the reaction cross-section and the generation of (PKA). We simulated and calculated the radiation damage by 10∼100keV protons in Ga2O3 using SRIM. It is found that ionization damage is much greater than displacement damage; the number of vacancies generated by proton radiation in Ga2O3 increases with the energy and incident angle of the incident proton. The irradiation resistance of Ga2O3 is between silicon and diamond semiconductor materials.
Neutronics Analysis of Fusion Blanket Based on the Spherical Harmonic Function and Finite Element Method
Authors: Li Kang , Cao Liangzhi , Miao Jianxin , Zhang Haoyu , Dai Tao 
Published: November 2022

Abstract
The fusion neutronics simulation has been a great challenge to the numerical calculation of the neutron-transport equation. The key is how to deal with the features of fusion devices, such as large-scale, complex geometric models, large vacuum regions, etc. NECP-FISH, a code developed by Nuclear Engineering Computational Physics (NECP) laboratory of Xi'an Jiaotong University, is used to address this challenge. NECP-FISH adopts a deterministic numerical method instead of the Monte Carlo method because the deterministic numerical method is of higher computational efficiency and costs less computational time. To deal with large vacuum region, large-scale and complex geometric model, the first order neutron-transport equation is solved, the spherical harmonics function and the finite element method are applied to the expansion of angle and space. NECP-FISH has been validated by benchmark problems such as strong absorption problem, internal void problem, and Kobayashi series of problems. What’s more, NECP-FISH builds the user interface based on the platform SALOME so that users can visually build the necessary models for problems. NECP-FISH has been applied to the neutronics calculation of the breeder unit of Helium Cooling Ceramic Breeder (HCCB) and the blanket of CFETR. The numerical results demonstrate that the NECP-FISH code can efficiently solve the neutron transport problem of the fusion reactor.
Research on Technical Scheme of Rapid Power Reduction System Based on Gray Control Rods
Authors: Li Wei , Shi Jianfeng , Qin Yulong , Wang Lihua 
Published: November 2022

Abstract
The paper designed a small rapid power reduction system (SRPRS) based on gray control rods for pressurized nuclear power plants, and technical scheme of SRPRS was proposed in the paper. Different to current RPRS for turbine trip, SRPRS was designed for feed water pump trip. When one FWP happened, partial of gray banks dropped into the reactor core to reduce the core power, and reactor trip due to power mismatch between primary and secondary sides would be avoided. The paper researched the basic principles of SRPRS, and demonstrated the feasibility of using SRPRS in pressurized nuclear power plants with gray control rods. The small rapid power reduction system designed in the paper includes three stages: activation, recovery and normal operation. Each stage can automatically complete control actions with a few human intervention. The small rapid power reduction system makes full use of the gray control rod characteristics: low reactivity worth and small disturbance to core power distribution, so it can achieve rapid reduction of core power while maintaining a flat power distribution. Meanwhile, since the control rod insertion limit set in the technical specification does not limit the insertion of gray control rods, the operation of small rapid power reduction system will not exceed the requirements of the technical specification for normal operation. The paper designed a “slope factor” in small rapid power reduction system. By calculating the slope factor in advance and judging the rods to be dropped in real time, the small rapid power reduction system can perform its function at any initial gray control rod position.
Determination of Control Rod Position by Global Optimization Methods
Authors: Qian Zhang , Liu Zhihong , Zhao Jing 
Published: November 2022

Abstract
The method is developed to find the implicit position of the control rod using the flux value shown by the neutron detectors. When using the 3D core physics analysis code CPACT, the detector response of a set of predicted rod positions can be calculated. On a double duel problem, functions can be built to describe the difference between the prediction and the target. Under the frame of the inverse problem, combinatorial optimization methods are operated to renew the prediction of rod positions by iteration to a result with convergence. By such a method, the results are guaranteed to achieve the self-consistent with the physical and thermal field while reaching stable convergence when the forward calculation software has been fully verified to simulate the real case. In the meanwhile, the appropriate computational cost is needed and can be adjusted with time tolerance. So, we have compared several global search algorithms (Genetic Algorithm, Particle Swarm Optimization, and Characteristic Statistic Algorithm) with their performance. We can also get the conclusion that hidden features and intermediate calculation results are important to improve the search efficiency. However, the responses of the detectors to some remote rods are not obvious, which are not easy to determine. All the methods were tested on the numerical calculation examples of a pressurized water reactor case.
Study on Ultrasonic Inspection Technology of Fastening Bolts of Reactor Internals
Authors: Tang Jianbang , Wu Yukun , Wang Weiqiang , Ma Guanbing , Zeng Chenming , Xu Junlong 
Published: November 2022

Abstract
Reactor internals are the core components of the reactor pressure vessels in Nuclear Power Plant (NPP), which directly compress, support and position the nuclear fuel assemblies. They are exposed to high temperature, high pressure and high radiation during a long-term service. The fastening bolt may break and fall off due to irradiation stress corrosion crack (IASCC), which directly affects the safe operation of the reactor. Along with the in service time increase, it has gradually become the focus of the in-service supervision of NPP, and it is urgent to perform Nondestructive Examination (NDE) for its structural integrity. In this study, according to the structural characteristics and inspection difficulties of typical bolts, a special ultrasonic combined probe and ultrasonic process are designed, by using simulation technology. The design principles are summarized. In addition, the automatic inspection system of underwater submarine is introduced. The results of the process verification show that currently technology can effectively detect with a minimum flaw sizes of 17.5% cross-sectional area (CSA), which meets the in-service inspection (ISI) requirements of NPP.
Regulation without Transparency Impedes Compliance
Authors: Miranda Samuel 
Published: November 2022

Abstract
In November 2020, the Office of the Inspector General (OIG) of the Nuclear Regulatory Commission (NRC) issued its annual audit plan, which includes an audit concerning transparency (i.e., the public trust and confidence) in the NRC’s regulatory activities. [3] The audit will examine the NRC’s practice of allowing “drop-in” visits, by senior executives of licensees, especially during times when the NRC staff could be evaluating and deciding on regulatory matters that directly affect the interests of those licensees.
In 2017, the OIG conducted a related audit that also focused upon the public’s trust and confidence in the NRC. [4] That audit examined the NRC’s procedure that governed its evaluation of 10 CFR §2.206 enforcement petitions. The OIG found that the NRC staff had not issued a single enforcement order, as the result of 38 enforcement petitions that it had received in the prior three fiscal years, ending in 2016. The OIG concluded that, the lack of such actions could adversely affect the public’s perspective on the effectiveness of the agency’s 10 CFR 2.206 petition process. Both audits are discussed herein, via examples that illustrate the NRC’s implementation of its policy of transparency, in theory and practice.
Numerical Study of the Effect of δ-Ferrite on Crack-Tip Mechanical Field
Authors: Wu Yule , Zhang Lefu 
Published: November 2022

Abstract
A dual-phase microstructure with 5–12% δ-ferrite is needed to prevent hot cracking of stainless steel weld metal. However, the δ-ferrite also makes it susceptible to long-term thermal aging embrittlement. Recent studies revealed that the stress corrosion cracking (SCC) susceptibility might either be reduced or increased by the δ-ferrite, depending on the embrittlement degree of the ferrite during the operation. One possible explanation was that the harder δ-ferrite inhibited the crack growth by changing the cracking direction and creating a highly branched crack path. To further reveal the mechanism of δ-ferrite effect on SCC behavior of those materials in high temperature water, a finite element investigation for the distribution of crack-tip stress and plastic strain field with and without δ-ferrite under a constant load was conducted. The result shows that the crack tip stress and plastic strain decrease when the crack encounters the δ-ferrite, indicating a lower cracking susceptibility, and the enhancement of δ-ferrite hardness can further reduce the crack-tip plastic strain. the severe stress and strain states occur along the ferrite edge, this indicates that the cracks are more tend to initiate and develop along the interface of δ-ferrite and austenite matrix.
Discussion of Nuclear and Radiation Safety Regulator Training in China
Authors: Li Zhaotong , Xing Dan , Chen Fangqiang , Jia Wei , Zhang Yang , Chen Shuai 
Published: November 2022

Abstract
China’s National Nuclear Safety Administration (NNSA) has always attached great importance to the training of regulators. In 2020, NNSA issued the training program for nuclear and radiation safety regulators. In this study, research was conducted on the implementation of the training program. Analyze the current status of our training systems were drawn from the aspect of implementation and management. Main problems of regulators’ systems were summarized from this study: (1) the training course needs to be further improved according to the training program. (2) The construction of curriculum materials needs to be continuously promoted. (3) personnel training experience needs to be systematically managed. (4) training knowledge management and application need to explore effective ways. Solutions to each problem are proposed to help improve the training efficiency of regulators and explore the use of training knowledge.
Research on Fatigue Analysis Method of Nuclear Equipment Under Ocean Conditions
Authors: Liu Guangdong , Zhu He , Zhang Zhui , Zhang Guoxun , Cao Guitao , Tan Lin , Qi Jing 
Published: November 2022

Abstract
Under the conditions of the marine environment (wind, wave, current), the floating platform on the ocean undergoes heeling, trimming, rolling, pitching, swaying, surging, heaving movements. The floating reactor on the floating platform within the design life, in addition to withstand the low-frequency fatigue caused by primary loop temperature and pressure, it also has to withstand the inertial force generated by the sway of the wave for the service period. The inertial force generated by the wave is a constantly changing dynamic load. Alternating stresses inside of the structure may cause fatigue damage to the structure. In order to estimate the fatigue life of floating reactor equipment under various ocean conditions, a finite element analysis model of equipment was established to carry out research for fatigue analysis methods under sway loads, meanwhile the operating temperature and pressure loads were considered. The probability of sway at each angle obeys the Rayleigh distribution, through the analysis of the status for the platform, it can estimate the number of swings at each angle. Based on Miner linear cumulative damage theory, the cumulative fatigue usage factors meet the requirements of the ASME BPVC III NB-3220 code. This analysis method provides a reference for the fatigue analysis of the floating reactor nuclear equipment.
Feasibility Analysis and Demonstration of In-Vessel-Injection in the Early Stage of Severe Accident
Authors: Xia Shaoxiong , Zhan Dekui , Wu Zijie , Zhao Xinhai , Yu Yiqun 
Published: November 2022

Abstract
The core of the nuclear reactor will melt during severe accidents, which may damage the integrity of the reactor pressure vessel and containment, and release radioactive materials to the environment. The third-generation pressurized water reactor is equipped with severe accident mitigation systems to prevent the high temperature (3000K) corium from melting through the reactor pressure vessel. The severe accident mitigation system mainly includes: primary depressurization system, reactor pit flooding system, containment combustible system, containment heat removal system, etc.. Benefit from these systems, large release frequency (LRF) is restricted to a low level. However, these systems cannot stop the process of core degradation. If severe accident happens, the core will melt and the whole reactor cannot be reused again, causing irreversible economic losses. In order to improve the economy and safety of nuclear power plants, this paper proposes in-vessel-injection (IVI) in the early stage of severe accident, and proves the effectiveness of this measure. According to the analysis, in-vessel-injection in the early stage of severe accident can prevent the large-scale melting of the core during severe accident, and the risk of hydrogen and source items will be greatly reduced.
Research on Establishing Efficient Scheduling System of Spent Fuel From Chinese Nuclear Power Plants
Authors: Su Jiqiang , Liu Qun , Zhang Honglin , Shi Lei , Liu Hongjun , Li Yanrui , Hu Jian 
Published: November 2022

Abstract
As the number of nuclear power plants put into operation in China continues to increase, more and more spent fuel is produced. Before the spent fuel leaves the spent fuel pool to the final disposal of the spent fuel, it needs to go through multiple and various forms of transportation and storage. By establishing an efficient scheduling system for spent fuel, the efficient scheduling and overall coordination of each link is crucial for the safe management of spent fuel. This paper studies the connotation of the efficient scheduling system of spent fuel. Based on the enlightenment of the establishment of the efficient scheduling system for spent fuel in major nuclear power countries, it analyzes the necessity and current situation of establishing an efficient scheduling system for spent fuel in China’s nuclear power plants, and puts forward relevant measures and suggestions.
Digital Twin-Driven Development of Online Monitoring and Data Management Systems in TMSR-SF0
Authors: Liu Wen-Qian , Han Li-Feng , Huang Li 
Published: November 2022

Abstract
With the development of a new generation of information technology, the monitoring systems in the nuclear power field are also facing the technological revolution of digital reform. Based on the example of Thorium Molten Salt Reactor-Solid Fuel (TMSR-SF0), this paper introduces the overall structure, modeling and mapping, and digital visualization expression means, then a complete development scheme of the real-time online 3D monitoring system is proposed. Firstly, the digital mapping model of the molten salt reactor was established, and the model docking and virtual scene rendering were completed in the Unity engine. Secondly, the back-end server using Experimental Physical and Industrial Control System (EPICS) and a proxy server using Node.js are established, and spatio-temporal data association is realized through Node-EPICS event driver and Socket.io. Finally, the MySql database interface and xCharts visualization framework are introduced to complete data storage and expression. Verified by practice, The 3D monitoring system is feasible in design and operation. Compared with the previous monitoring system, the data update period is increased to 20ms, and it has the whole process of data collection, network communication, graph element dynamic display, and other functions, which can promote the operation management and monitoring of nuclear power plants, and provide a reference for the digital transformation of monitoring technology in the nuclear power field.
Feasibility Study of Passive Water-Cooled Residual Heat Removal Scheme for Small Reactors
Authors: Yu Fei , Fu Yueming , Zhu Feng , Zhang Mingqian 
Published: November 2022

Abstract
With the development of nuclear energy and the need for environmental protection, some advanced small reactors have gradually entered the public’s field of vision. The passive natural circulation is an important part of the inherent safety feature for advanced small reactors. Based on the limitations of the site environment, the large heat exchange area of the air cooler, and the high cost of equipment construction for the air-cooled waste heat removal scheme of a small reactor, this paper studies the water-cooled passive residual heat removal scheme and analyzes the water-cooled water cooling by establishing a system natural circulation hydraulic calculation model. From the aspects of equipment scale, construction cost, and operation & maintenance cost, etc., the advantages and disadvantages of several schemes were analyzed.
This study shows that the water-cooled passive waste heat removal scheme can extract the core heat under accident conditions; it can reduce the equipment volume, and reduce the cost. Finally, suggestions are put forward for the follow-up improvement direction, and further research and realization of steam-water two-phase condensation heat transfer are made to improve heat exchange efficiency, reduce equipment scale, and improve the economy, which can provide a reference for the follow-up small reactor research and development.
Research on the Expected Carbon-14 Production and Discharge in Pressurized Water Reactors
Authors: Fu Pengtao 
Published: November 2022

Abstract
Carbon-14 is one of the most important radionuclides discharged to the environment from pressurized water reactors due to its long half-life and its important role in the biological chain. Carbon-14 is the largest contributor of the dose rate to the public from all radionuclides discharged to the environment during the normal operation of pressurized water reactors, and thus the production and discharge of Carbon-14 have been focused on in the industry in recent years.
Based on the generation mechanism in pressurized water reactor, one theoretical model of Carbon-14 generation has been established and the nitrogen concentration of all chemical species in the primary loops has been determined according to the measured ammonium. It predicts that the annual Carbon-14 generation in the primary loops is proportional to annual electricity output and the typical normalized Carbon-14 is 2.9E+02 GBq/GWe/yr. The theoretical model has been verified by the statistical analysis of annual Carbon-14 discharges from French PWR units. In addition, the quantity of Carbon-14 in radioactive solid waste has been estimated in these PWR units. It shows the generation of Carbon-14 in PWR cannot be effectively minimized because O-17 atoms, the predominant origin of Carbon-14, exist naturally in the primary loops during long-term operation. This approach can be applied to analyze the Carbon-14 production and discharges in operating pressurized water reactors and in the assessment of source term of the new pressurized water reactors.