bluuebunny/crossref_metadata_2025_split huggingface dataset:
Fission Gas Release Grain Boundary Network Percolation Mechanistic Studies in Oxide Fuels Based on COMSOL Multiphysics Framework
Authors: Guo Jingyu , Zhou Wenzhong 
Published: November 2022

Abstract
Fission gas release was modeled by COMSOL Multiphysics in oxide fuels, whose theory is based on the effect of grain boundary network percolation. The model is contributed to the conventional Booth model. In this model, the fuel pellet microstructure adopts 2D axisymmetric geometry. The effect of different bubble growth and coalescence rates on the independent grain boundaries are investigated, such as bubble contact angle, resolution rate, and radial position. The modeled physical phenomena are similar to the existing experiments and can be observed in the fission gas release process. The fission gas release is sensitive to the variations of these parameters. Therefore, the parameters are essential to the fission gas release on the microscopic or macroscopic scales. The long-range percolation on the networked grain boundaries is also considered in this work, but it is not considered in the Booth model. The gas resolution on the grain boundaries is also taken into account. At last, this model’s results will be compared with the outcomes of the Booth model as well as the other models.
Research on Risk Characterization Method of Marine Biological Disaster Affecting Water Intake Safety of Nuclear Power Plant and Application of Design Protection
Authors: Fan Yu , Jie Bai , Botao Xu , Xingbing Lv , Yong Li , Xueqing Zhang 
Published: November 2022

Abstract
Marine biological outbreaks in coastal waters of coastal nuclear power plants have caused abnormal water intake systems of nuclear power plants for many times, and the risk Valuation method of marine biological outbreaks to nuclear power safety is still in the research and exploration stage. For nuclear power plants, the outbreak of marine organisms near the water intake is a natural event that affects the safety of nuclear power. According to the basic framework of “four-step method” on risk Valuation of National Academy of Sciences of the United States, aiming at the disaster risk caused by marine organisms that potentially affects the reliability of cold sources of nuclear power plants, the risk characterization method is studied, and the design protection measures are put forward.
Key Impact Factors Analysis of Fuel Rod End Plug Welding Based on QFD
Authors: Li Yinghong , Lv Zhipeng , Li Changzheng , Lin Yuyu , Lin Shaofang , Cai Zhiyi , Deng Yongjun , Nie Lihong 
Published: November 2022

Abstract
Manufacturing defects of fuel rod end plug during welding affect fuel reliability. Failure of ring welding or sealing welding of fuel rod end plug during operation will cause fission gas to enter primary coolant of reactor. In order to determine the sensitivity of the influence of relevant characteristic parameters on the performance of fuel rods during end plug welding, based on the Quality Function Deployment method (QFD) and taking tungsten-inert-gas arc welding (TIG) as an example, the characteristic parameters of fuel rod end plug welding are analyzed. Combined with the experimental results, a correlation matrix between welding parameters and welding performance is constructed by LIKERT scale method with focus on the main welding performance criteria such as welding joint size, color, depth of fusion, metallography (bloating, porosity, inclusion, grain boundary separation, etc.), oxidation resistance, etc. And after rating its correlation degree and importance degree, the relevant House of Quality of end plug welding is constructed, and the key impact parameters are identified. Based on the above analysis, some feasible optimization and improvement proposals are put forward for the quality supervision and welding process of fuel element manufacturing factory.
Explainable Neural Network Algorithm for Rapid Radionuclide Identification Under Low Count Gamma-Ray Spectrum Data
Authors: Wang Yu , Li Sufen , Huo Yonggang , Yang Jianqing , Zhang Quanhu 
Published: November 2022

Abstract
Real-time and automatic radioisotope identification using gamma spectrum is an important issue in the field of nuclear safety. It is widely used in vehicle mounted radioisotope monitoring, Marine radioisotope monitoring and nuclear decommissioning verification scenarios. At present, the focus of radionuclide identification is fast and stable recognition under low count conditions. In this paper, a radionuclide recognition method based explainable artificial neural network is proposed, and a synthetic gamma spectrum data set is created. The data set contains gamma-ray spectra of 12 different types of radionuclides, which were obtained by Geant4 Monte Carlo simulation software and gaussian broadening of the detector. Data Augmentation was achieved by simulating gamma spectra at different measuring times, different measuring distances and different ambient temperatures. The training results of the neural network optimized by hyperparameter show that it has a high accuracy on the test set with shorter measurement time, longer measurement distance and larger energy spectrum drift range, which provides a method for rapid identification of nuclides in the case of low count. Using t-SNE dimension reduction technology, the twelve dimensions data output by the neural network is reduced to two dimensions for feature visualization, which vividly explains and verifies the recognition results of neural network.
Research of Dissimilar Steel Welding for HTR-PM SG Heat Exchange Tube
Authors: Zheng Mingtao , Wang Li 
Published: November 2022

Abstract
The modular High-Temperature Gas-Cooled Reactor nuclear power plant has inherent safety, simple system, high power generation efficiency and wide use. Incoloy800H and 2.25Cr-1Mo heat exchange tubes are used for High-Temperature Gas-Cooled Reactor (HTR-PM) Steam Generator. The steam generator uses the Incoloy800H and 2.25Cr-1Mo heat exchange tubes to transfer the high temperature helium medium to the water medium. The dissimilar steel welds are designed between Incoloy800H and 2.25Cr-1Mo heat exchange tubes. The heat exchange tubes of HTR-PM Steam Generator with small OD, large thickness, and design structure complicated. The dissimilar steel welds between two kinds of heat exchange tubes with limited welding operation space and welding seam quality requirements high. During the HTR-PM Steam Generator operation process, Incoloy800H heat exchange tube, 2.25Cr-1Mo heat exchange tube and the dissimilar steel welds between Incoloy800H and 2.25Cr-1Mo heat exchange tube are service in high temperature environment for a long time, until the steam generator fails. The surface forming quality, conventional properties and high temperature stress-rupture of dissimilar steels weld is required. The quality of dissimilar steel weld of Incoloy800H and 2.25Cr-1Mo heat exchange tubes is directly related to the safe operation of steam generator. Through systematic research including tube-to-tube joint type, the welding material, welding parameters, heat treatment process etc. The results show that using appropriate GTAW process with Nickel base alloy consumable inserts, welded joints can meet all design requirements such as strict inner and outside surface forming quality, conventional performance and the strength of stress-rupture test in HTR-PM Steam Generator service environment. This process has been used in the product manufacturing of demonstration projects, and provides good demonstration experience for similar application scenarios in the industry.
Rapid Assessment Method for Corrosion Flaws of Seawater Cooling Pipes in Nuclear Power Plants
Authors: Han Zhichao , Chang Haijun , Shen Mengling , Rong Zhishen , Li Zhengpeng 
Published: November 2022

Abstract
The corrosion of the fluid to the pipes could damage the pipe structure, and then affect the safe operation of the whole pipeline system in the nuclear power plant. In order to assess the influence of the corrosion flaws on pipe integrity, it is necessary to perform the stress analysis and calculate the remaining strength of the pipes containing flaws. In this paper, the corrosion flaws in the seawater cooling pipes are considered as the volumetric type flaws. The remaining strength of the pipes containing the flaws has been calculated, the minimum required wall thickness of the pipes has been obtained, and the flaws have been assessed according to SY/T 6477 Standard. The finite element method has been used to analyze the pipes containing corrosion flaws. Stress distribution of the pipes has been obtained under the pressure and nozzle loading. Based on the RCC-M Standard, the strength assessment has been conducted. Results showed that the stress calculated from SY/T 6477 Standard is close to that from the finite element method, and it can be considered that the assessment method summarized from SY/T 6477 Standard is suitable for the current pipes containing flaws under the nozzle loadings. Therefore, this method can be used to assess the corrosion flaws of the seawater cooling pipes rapidly, and provide advices for the monitoring and repair of the pipes containing flaws in nuclear power plants.
Study on the Method for the Safety-Related Instrument Calibration Surveillance Interval Extension of Nuclear Power Unit With Analog I&C System Based on Setpoint and Uncertainty Analysis
Authors: Zhang Yong , Zhang Zhao , Wang Zi-Chun , Zhou Yong-Hui , Tang Bo-Wen , Su Min-Hua , Luo Wen-Bo 
Published: November 2022

Abstract
To extend the intervals of Safety-related Instrument Calibration Periodic Test can shorten the refueling outage period and improve unit economy. Based on the general requirements for extending the Safety-related Instrument Calibration Periodic Test Intervals and in consideration of the characteristics of M310 Nuclear Power Unit with analog I&C System, an evaluation method is proposed and investigated, which consist of instrument drift analysis, protection channel analysis, channel uncertainty calculation, setpoint analysis and relevant channel tests evaluation. The research on the extension of steam generator level transmitter calibration test interval is carried out in detail. The results show that the drift resulting from the interval extension will has no effect on the protection channel setpoint, the acceptance criteria for channel consistency test and channel function test should remain unchanged. The research of this paper has a reference value for unit economy improvement and outage optimization.
Study on Methodology and Application of Seismic Level 2 Probabilistic Safety Assessment for Nuclear Power Plants
Authors: Zheng Hongjia 
Published: November 2022

Abstract
Many countries have paid more attention to seismic risk in nuclear power plants after the Fukushima nuclear accident. It indicates that evaluating the risk for seismic event is vitally important. Seismic Level 1 probabilistic safety assessment (PSA) has a mature methodology, but how to perform the study on seismic Level 2 PSA has become a crucial problem needed to be considered deeply by both regulators and operators. In this paper, the methodology of seismic Level 2 probabilistic safety assessment (PSA) is developed for nuclear power plants. Seismic Level 2 PSA is focused on the severe accident phenomenon, progression and source terms release to evaluate the containment response after core damage caused by seismic event. To demonstrate the effectiveness of the proposed approach, this study took a nuclear power plant under construction in China as an example. Seismic Level 2 PSA focuses on crucial elements including Level 1/ Level 2 PSA interface analysis, severe accident design features, severe accident phenomenon, containment event tree analysis, seismic fragility analysis, etc. A risk-informed seismic Level 2 PSA model is developed to evaluate large release frequency (LRF) caused by seismic event. The proposed methodology can be expected to provide an effective framework for performing the seismic Level 2 PSA and decision support to improve the nuclear power plant safety.
Development and Application of Nuclear Power Plant Life Cycle Management Database Platform
Authors: Guochang Cao , Zhao Liu , Fangming Lv , Zhenzhong Yang , He Jilang , Yixin Cui 
Published: November 2022

Abstract
In the long life extension operation of nuclear power plant, life cycle management technology can provide reliable support for the economic operation of nuclear power plant, and is an important support technology for the evaluation of nuclear power plant’s continuous operation. In order to promote the development of life cycle management technology in nuclear power plants, information technology is used to transform specific problems into standardized databases, relying on the Operation License Renewal Project, the software platform of life cycle management database for important equipment in nuclear power plant is developed. In addition to the storage function of the equipment life cycle management information, the database has the functions of cost calculation and economic comparison of the long-term operation plan. It includes seven modules: Original Information, Program Management, Aging Mechanism, Out-of-date Evaluation, Failure Rate, Economic Analysis and External Feedback. The database built-in nuclear power plant life cycle management typical equipment such as steam turbines, generators, main transformers, and so on, while users can be independent in life-cycle equipment maintenance to add the need to expand the object. It is considered that the acquisition and reliability of cost data, equipment reliability data, and equipment life and failure rate data are the difficulties of life cycle management of nuclear power plants at present.
Analysis of the Influence of RCCA Swelling on Rod Falling and Study on Mitigation Measures
Authors: Lv Zhipeng , Zhang Guoliang , Li Changzheng , Liu Weichao , Lin Shaofang , Nie Lihong , Li Zhijun , Zhang Anlong 
Published: November 2022

Abstract
Silver-Indium-Cadmium (Ag-In-Cd) is commonly used as the neutron absorption material in PWR, and its swelling under reactor irradiation condition may lead to outward deformation of control rod cladding, and finally affect RCCA dropping. In severe cases, rod dropping time exceeding its limit regulated in operation technical specification and even RCCA sticking caused by Ag-In-Cd excessive swelling may occur, which will threaten reactor safety shutdown. The influence of different swelling forms on the dropping time of RCCA is studied in this paper, and finally the absorber design is optimized based on current PWR absorber design. The results show that the swelling of control rods has little effect on T5, but mainly affects the dropping time T6 in the dashpot section. Different swelling positions have different effects on the dropping time, and the optimized design can improve this swelling effects.
Application of Detail Design Data of Nuclear Power Plant in Project Construction and Operation Under the Background of Digitization
Authors: Huang Huizhen , Tian Licheng , Yan Jiafu 
Published: November 2022

Abstract
Building Information Modeling theory has been more and more welcome all over the world, and has been introduced into nuclear power industry. The so-called ‘smart power station’ being established by many nuclear power enterprises is the embodiment of BIM. With the deepening of digital design and application, the design data has become more and more detailed and accurate, to support all design work such as data exchange between systems and software, automatic drafting, automatic material and working counting, and so on. In addition to design, how should the procurement, construction, installation and power station operation engineers use these data to improve their work efficiency and accuracy is a research topic, which has been highly followed by engineering companies at present. At the same time, the design data is more for design itself. Whether its expression form, depth and breadth can meet the requirements of downstream engineers and how to improve efficiently when they are not satisfied is an important issue that all project participants should pay attention to. This paper gives a sketch of 3D design data, with pipework as an example, to elaborate how to use design data to help other engineers to break down old-style of working mode, and gives some thinking about optimizing design data efficiently to fulfill the requirement of other project participants.
Research on Redundancy Strategy of Nuclear Instrument Control Data Service System
Authors: Zhang Yingying , Hu Yiwu , Li Hualong , Sun Jingle 
Published: November 2022

Abstract
The nuclear instrument control data service system is the data monitoring background of the monitoring layer of the nuclear instrument control system, which has higher requirements for system reliability and data security, and usually adopts the fully redundant configuration structure of dual machine hot standby and double storage. How to arbitrate the server master-salve state according to the real-time diagnosis results, is the key problem faced in the redundant design of the nuclear instrument control system. According to the characteristics of the nuclear instrumentation control system, server failure types are analyzed, and an intelligent hierarchical redundant switching strategy is proposed. The decision-making module is divided into two layers. First of all, different business failure scenarios are divided, and each fault scenario gives a prime and slave arbitration result. Secondly, according to the importance priority, the arbitration result of the first layer is sorted, and the final arbitration result is given. The proposed redundancy strategy supports arbitration under multiple concurrent faults, and can support single-machine deployment and dual-machine deployment, with good switching performance. The technology has been successfully applied in a nuclear power instrumentation control project. The system runs normally in the case of local faults such as single-machine power failure, single-machine network failure, process collapse, etc., ensuring the safe and stable operation of the unit.
Study on Synthesis of the Organophosphorus Functionalized MCM-41 And its Adsorption Property for Dy(III)
Authors: Mao Cong , Sang Hongji , Zheng Jiawei , Wu Yan 
Published: November 2022

Abstract
The composite of organophosphorus groups loaded on MCM-41(MCM-Zr-TBP) was prepared by multi-steps impregnation method to develop a novel adsorbent for radioactive lanthanides extraction from the secondary contaminated water. The synthesized hybrid material was characterized by SEM and TG. Dy(III) was taken as the representative of trivalent lanthanides. The adsorption performance of Dy(III) on MCM-Zr-TBP composite was systematically studied as the functions of solution pH, initial concentration, interaction time and aqueous temperature. The results showed that Dy(III) adsorption on MCM-Zr-TBP composite was highly dependent on aqueous pH and initial metal ion concentration. Compared with the Freundlich and pseudo-first order models, the Langmuir and pseudo-second order models presented better fitting for the adsorption data. These results indicated that MCM-Zr-TBP was found to be an effective and competent adsorbent, which could be used for the effective removal of lanthanides from the wastewater.
PWR Spent Fuel Dry Storage Loading Pattern and Safety Evaluation
Authors: Wen Li , Jinghui Gao , Gang Li , Shengdong Zhong , Jin Xu , Bingquan Zhao 
Published: November 2022

Abstract
This paper describes the loading pattern, safety evaluation and operating experiences of PWR plant off-site Spent Fuel Dry Storage System (SFDSS). According to the limits from the SFDSS criticality and heat-transfer safety analysis, ALARA principle and AFA series fuel characteristics, technical staff complete the loading pattern and safety evaluation by analyzing influence of assemblies’ initial enrichment, burn up and cooling time. This paper also shows the temperature and dose data during AFA series fuel SFDSS loading, transportation and storage, which can indicate the actual condition of SFDSS criticality, heat-transfer and radiation. In general, AFA series fuel SFDSS is in good condition and meet the expected design and safety requirements.
Research on Structural Optimization Technology of Fuel Assembly Parts Based on Topology and Optimal Design
Authors: Fan Huabin , Xu Shiyin , Jia Huili 
Published: November 2022

Abstract
The top nozzle is an important part of the fuel assembly and is located at the core coolant outlet, which has an important impact on the uniformity of the coolant flow distribution at the core outlet and the overall pressure drop of the fuel assembly[1]. The peak stress of the top nozzle under each working condition appears at the position of the adapter plate, top nozzle and the pressure drop of the top nozzle is closely related to the flow area, height and local characteristics of the adapter plate. In this paper, the overall structure of the adapter plate is optimized to reduce the pressure drop as much as possible on the premise of ensuring that the stress meets the requirements of the standard limit. The 3D modeling software NX is used to establish the geometric model of the top nozzle, and it is transferred to ANSYS to establish a finite element model for simulation optimization design. On this basis, the key dimension parameters of the adapter plate (such as the thickness of the rib, etc.) are selected as design variables, and the law of its influence on the blocking area and stress is studied, and the structure size is further optimized.
Study of Breeding Performance in a Heat Pipe Cooled Traveling Wave Reactor With Th and U Fuels
Authors: Ma Kunfeng , Hu Po 
Published: November 2022

Abstract
Heat pipe cooled traveling wave reactor (HPTWR) is a newly proposed heat pipe reactor. The HPTWR can achieve the low enrichment of loaded fuel, high power density, and long-term continuous operation for the power supply of decentralized electricity markets. Due to the excellent breeding capability of the HPTWR, the Th fuel is also added into the breeding fuel region of the reactor to achieve the Th-U fuel cycle in this work. The Monte Carlo code RMC is used to obtain the reactivity swing, propagation of axial power peak, burnup and the productions of bred fissile nuclides for the HPTWR with Th and U fuels. The study shows that the HPTWR with 14.3% 235U enrichment of ignition fuel and 20% 235U enrichment of breeding fuel can continuously operate for 41 years without refueling when the mass fraction of 232Th in heavy metals of breeding fuel region is 50% at the initial time. The propagation velocity of axial power peak and total burnup for the HPTWR with the Th and U fuels is about 0.3659 cm/years and 89.32 GWd/THM during the 41 years operation respectively. The productions of the bred 239Pu, 241Pu and 233U in the HPTWR with Th and U fuels are about 356.40 kg of 239Pu, 1.57 kg of 241Pu and 199.94 kg of 233U at the end of life respectively. The obtained results in this study demonstrate that the HPTWR can achieve the Th fuel breeding in the case of the low 235U enrichment loading (≤ 20%) and long-term continuous operation (&gt; 40 years).