核动力工程

Research and Application of Transient Satistical Method for Nuclear Power Plant

Bai Xiaoming, Cao Guochang, Cao Hongsheng, Yu Xinyang, Xiong Furui, Jiang He

2024, 45(S1): 1-5. doi: 10.13832/j.jnpe.2024.S1.0001

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Abstract:
The number of transient occurrences of nuclear power plants is closely related to the fatigue life of equipment, so transient statistics are of great significance for improving the level of intelligent operation and maintenance of nuclear power plants and the renewal of operation licenses. At present, transient statistical methods at home and abroad have the disadvantages of large training data and poor generalization ability, and are rarely used in engineering. In this paper, a transient classification method based on equivalent distance measurement is established according to the design transient variation law, and the automation of transient classification, segmentation and counting process is realized. The current transient classification method is verified by the operation data of nuclear power plants, and the results show that the current method can effectively realize the statistical work of various operating transients. The application of transient statistical method plays an important role in the improvement of the intelligence level of nuclear power plants and the continuation of operation permits.

Research on Architecture Design Technology of Digital Experimental System Based on Systems Engineering

Zeng Xiaokang, Huang Yanping, Zhang Liqin, Yuan Dewen, Xu Jianjun, Sun Yuxiang

2024, 45(S1): 6-12. doi: 10.13832/j.jnpe.2024.S1.0006

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In order to establish a digital integrated R&D and delivery environment based on system engineering (SE), in this paper, the top-level architecture of the digital integrated R&D platform for nuclear power engineering experiments (i.e. the digital experimental system) is optimized, and the detailed architecture of functional composition and operational logic is designed. On this basis, the digital experimental system of nuclear power engineering is developed, and the experimental project of passive residual heat removal system on secondary side is used as an example for application verification. The verification results show that the established architecture of digital experimental system for nuclear power engineering can meet the requirements of digital integrated R&D and delivery of nuclear power engineering experiments, such as experimental task planning and decomposition, quality process control, tool software packaging and integration, collaborative design and simulation, knowledge accumulation and accompanying, process data tracing, and structured and lightweight delivery of experimental results, and can realize the integration of tasks, tools, knowledge, data and results, effectively improving the efficiency and quality of experimental research and development.

Research and Implementation of Anomaly Detection and Fault Diagnosis System for Chemical Control in Nuclear Power Plants

Chen Bo, Cao Zhongcai, Yao Xiangying, Guo Tianyu

2024, 45(S1): 13-18. doi: 10.13832/j.jnpe.2024.S1.0013

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At present, nuclear power plants mainly rely on chemical engineers for the anomaly detection and troubleshooting of chemical control, which is limited by the personal experience and skill level of chemical engineers, resulting in the problems of poor timeliness and inaccuracy in anomaly detection and troubleshooting. To address this issue, a set of anomaly detection and fault diagnosis system for chemical control in nuclear power plant is designed and implemented. During the research process, we focused on exploring key technologies such as anomaly detection rule databse, fault diagnosis database, and anomaly detection and fault diagnosis. Through the practical application in a nuclear power plant, it is proved that the system can effectively and automatically detect abnormalities in chemical monitoring data and diagnose abnormal faults, thereby improving the chemical control level of nuclear power plants and providing guarantees for the safe operation of nuclear power plants.

Research on Key Technologies of Cross Regional Collaborative Big Data in Nuclear Power Industry

Cheng Minmin, Jing Yinggang, Xu Kui, Qi Kelin, Wu Jize, Ren Zengpeng, Li Min

2024, 45(S1): 19-25. doi: 10.13832/j.jnpe.2024.S1.0019

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Due to the numerous production links, complex organizational structures and processes in nuclear power, a unified industrial big data platform at the nuclear power group level should be established to eliminate data silos, integrate massive data resources, and improve data utilization efficiency. In this study, a multi-source heterogeneous data access framework adapted to the nuclear power industry protocol is adopted to aggregate the cross-regional nuclear power industry data. According to the characteristics of nuclear power industry data, the coding and storage standards of nuclear power industry data are defined, and the efficient storage of massive data is realized. Based on metadata management, data extraction, data processing, data storage, data scheduling, quality audit, data service and other links are coordinated, and an end-to-end and closed-loop data governance and control mechanism is studied. This method was used to govern nuclear power data and open up assets, achieving full lifecycle management of nuclear power data. Represented by China Nuclear Power Data Center and six major nuclear power bases, the construction and engineering practice of big data platforms have formed a data sharing and collaboration channel between the group's data center and the edge side; We have gathered and processed data on China's nuclear power production and management, established data standards, and achieved effective integration and processing of data. Therefore, the nuclear power industry big data platform and data processing methods established in this study can provide support for the digitalization of nuclear power, have broad application prospects, and provide technical references for building a nuclear power data ecosystem.

Fluid-structure Interaction Dynamic Modeling of Tube Bundle Based on Data-driven

Feng Zhipeng, Cai Fengchun, Zhang Yixiong, Jiang Xiaozhou, Liu Shuai

2024, 45(S1): 26-31. doi: 10.13832/j.jnpe.2024.S1.0026

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Abstract:
The tube bundle is the core component in the pressurized water reactor steam generator, which is prone to flow-induced vibration. In the mechanism of flow-induced vibration, vortex shedding and fluidelastic instability are typical strong fluid-structure interaction problems, which cannot be solved by decoupling the structural field and flow field. In order to better carry out the mechanical design of tube bundle in steam generator, the present work completed the fluid-structure interaction dynamic modeling of two flow-induced vibration mechanisms, vortex shedding and fluidelastic instability, using data-driven method. The prediction results were verified with existing experimental data, and they were in good agreement with the literature results. This method combines the advantages of the theoretical models and the consideration of actual structural dynamics characteristics in CFD calculations. It not only avoids the need for massive computing resources of three-dimensional fluid-structure interaction simulation in complex tube bundle, but also reduces the dependence of traditional theoretical models on experimental data, which is conducive to promoting application in engineering.

Research on Operation Optimization of AP1000 Boric Acid Makeup Subsystem

Han Jie, Dong Shubiao, Zhang Hongsheng, Zhang Xian, Wang Ying

2024, 45(S1): 32-38. doi: 10.13832/j.jnpe.2024.S1.0032

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Abstract:
Compared with the traditional second-generation unit, the boric acid makeup subsystem of AP1000 unit has advanced design concept and simple system structure. However, in the actual operation and maintenance process of the unit, boron deviation alarms frequently occur, which affects the safe operation of the unit. It is necessary to further optimize the system's operational performance. This study aims to identify the key factors that influence the performance of the subsystem by constructing a comprehensive database of unit operations. Subsequently, simulation models are established using Flowmaster and RinSim software, and the impact mechanisms of various factors are analyzed through simulation experiments. The results show that the relative pressure of demineralized water and boric acid at the inlet of the three-way valve is the key factor to determine the effectiveness of the boron concentration control. Additionally, the performance of pressure relief valve, boron tank liquid level, and measurement instrument deviation are also important influencing factors, while the effects of regulating valve and makeup water pump are relatively small. Based on the analysis results, this study further explores operational optimization strategies for the system and provides optimization schemes from both short-term and long-term perspectives. The analysis results and optimization schemes of this study can provide reference for optimizing unit operation regulations or technical transformation.

Research on Fast Prediction Method of Fuel Rod Steady-state Temperature Distribution Based on PINN

Liu Zhenhai, Zhang Tao, Qi Feipeng, Zhang Kun, Li Yuanming, Zhou Yi, Li Wenjie

2024, 45(S1): 39-44. doi: 10.13832/j.jnpe.2024.S1.0039

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Abstract:
A fast prediction method of fuel rod steady-state temperature distribution base on Physical Informed Neural Network (PINN) is established in this research. The burnup, linear power, boundary temperature and space position are taken as characteristic parameters to solve the parametric solid heat conduction equations using PINN. Based on this method, rapid prediction models for the steady-state temperature distribution of fuel pellet and cladding were constructed. The calculation results show that the calculation speed of fast prediction models are about 1000 times faster than that of commercial finite element method software, and they also have high accuracy. The maximum relative deviation of the steady-state temperature prediction of fuel pellets and cladding is about 0.318% and 0.013% respectively compared with the validation set. The established PINN model can quickly and accurately predict the steady-state temperature distribution of fuel rods.

Modeling and Simulation of Heat Pipe Radiator Based on Modelica

Qi Lin, Li Yangliu, Wang Xuesong, Yin Hao, Wang Shuguang

2024, 45(S1): 45-51. doi: 10.13832/j.jnpe.2024.S1.0045

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In order to verify the design of heat pipe type radiation radiators, and evaluate whether the performance of the radiator can meet the requirments of engineering application indicators, this paper establishes a radiator collector ring model, a heat pipe thermal resistance network model, a pseudo wick thermal conduction model, and a fin and radiation unit model. A radiator simulation code using the Modelica language was developed based on the MWORKS platform, and simulation were conducted on different diversion schemes of a two-stage heat pipe type radiator. The research results indicate that the radiator simulation code can perform reasonable caculations on different design schemes. Therefore, the models and modeling methods used in this paper can be used for radiator design and optimization analysis.

Research on the Application of SPDM System in the Simulation and Verification Platform of Nuclear Power Digital Design

Qi Wei, Hu Xupeng, Lu Xingyan, Zhang Dazhi, Qu Ming

2024, 45(S1): 52-57. doi: 10.13832/j.jnpe.2024.S1.0052

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Abstract:
As the digital design paradigm of nuclear power evolves, the complexity of collaborative design tasks across departments, disciplines, and domains escalates. Aiming at the common problems existing in the simulation verification process of digital nuclear power design, this paper studies the enterprise-level simulation process and data management (SPDM) technology, and develops a simulation verification platform for digital nuclear power design. The system achieves software integration and data control in various domains, including design, modeling, simulation, and testing software. This development has laid the foundation for the digital transformation of design verification in nuclear power scenarios.

Study on Automatic Plant Startup and Shutdown Operation Schemme for Gas-cooled Micro Reactor

Si Tianqi, Du Yu, Yi Ke, Zhang Ganghe, Su Yihui

2024, 45(S1): 58-62. doi: 10.13832/j.jnpe.2024.S1.0058

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Abstract:
Gas-cooled micro reactor (referred to as "micro-reactor"), as a micro-nuclear power source, can provide reliable power supply for the micro-grid in remote areas. Because of the characteristics of small output power and remote construction location, and considering the economic impact, it is necessary to reduce the personnel intervention as much as possible to improve the automatic operation capability of the gas-cooled micro reactor. Therefore, it needs to vigorously carry out research on the automatic operation of the gas-cooled micro reactor. The application of automatic startup and shutdown control system (APS) is a key technology to realize the automatic operation. This paper introduces the design method for the automatic startup and shutdown operation of gas-cooled micro reactor. Based on the system design and characteristics of micro reactor, firstly, the startup and shutdown processes are analyzed, then APS design and sequential control logic model construction are carried out, and finally the APS design scheme of the micro reactor is simulated and verified. The verification results show that for the implementation of APS logic in startup and shutdown processes, the change curves of important parameters are in line with the expected results of logic. Therefore, the design method of automatic startup and shutdown of micro reactor proposed in this paper is effective and feasible.

Development of System Analysis Code for Sodium-cooled Fast Reactor and its Verification on SHRT-45R Benchmark Problem

Sun Lin, Wu Zongyun, Zhang Zhenyu, Xue Fangyuan, Wang Xuesong, Liu Tiancai

2024, 45(S1): 63-67. doi: 10.13832/j.jnpe.2024.S1.0063

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Abstract:
The system analysis software for sodium-cooled fast reactors based on best estimation methods is crucial for both reactor system design and safety review. In this study, the Fast Reactor Transient Analysis Code（FRTAC） of sodium-cooled fast reactor system suitable for general reactor type is developed. Besides traditional input card, a graphical user interface is added. Through internal testing and third-party testing, FRTAC simulation function can cover the normal operation, transient operation, design basis conditions and some design extended conditions. In order to verify the accuracy of the software, the International Atomic Energy Agency SHRT-45R benchmark problems were used for modeling and analysis. The results prove that the errors of key parameters such as core coolant temperature and flow rate between calculation results and experimental values were less than 10%. The FRTAC calculation results of the software are accurate and can be used for accident analysis of sodium-cooled fast reactor.

Research and Application of Intelligent Maintenance Assistance System

Wan Shu, Cai Wanrui, You Bing, Liu Peibang

2024, 45(S1): 68-71. doi: 10.13832/j.jnpe.2024.S1.0068

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The management of instrument and control technology, such as risk analysis, safety measures analysis, and DCS file management, is a difficulty and focus in the field of instrument and control work in nuclear power plants. In order to reduce unexpected defects and shutdown risks of units caused by inadequate preparation for instrument control work, based on the study of DCS engineering documents and design documents, a set of intelligent maintenance assistance system is developed. This system interfaces with high-frequency usage scenarios of on-site personnel, supplemented by computer database information modeling technology, achieving intelligent analysis of DCS configuration, DCS system status monitoring, automatic risk warning, and one-stop retrieval functions. It can effectively improve the level of information management in the field of nuclear power instrument and control debugging and maintenance, and improve personnel work efficiency.

Research on CANDU Channel Power Prediction Based on Stacking Ensemble Learning

Wang Deying, Hu Wei, Wu Tong, Zhu Kerun, Zhang Liang, Yang Meng, Du Min, Zhang Ran

2024, 45(S1): 72-77. doi: 10.13832/j.jnpe.2024.S1.0072

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The accuracy of CANDU reactor channel power prediction directly affects the quality of the refueling plan, which in turn affects the economy and safety of reactor operation. To improve the prediction effect, it is advocated to introduce artificial intelligence algorithms to mine the potential variable relationship from historical operation data. After data cleaning and feature selection, a feature termed "Refueling Impact Index" is designed. With XGBoost, random forest, support vector regression, and BP neural network as primary learners and support vector regression as secondary learners, an ensemble learning model based on Stacking is constructed. Through comparative analysis, the Stacking ensemble learning model has achieved a "secondary improvement" in prediction effect on the basis of a single learning model. Moreover, the Stacking ensemble learning model has significantly better effect than the traditional RFSP methods in terms of average prediction deviation rate, maximum prediction deviation rate, and prediction deviation rate variance. This enables physical engineers to obtain more accurate power feedback in the process of formulating refueling plans, scientifically select refueling channels, and thereby improve economic benefits while ensuring reactor safety.

Development and Validation of Non-Inertial Coordinate System Motion Model Based on System-level 3D Thermal Hydraulic Code

Ye Qian, Tan Chao, Xiong Yan, Li Fei, Shan Fuchang

2024, 45(S1): 78-84. doi: 10.13832/j.jnpe.2024.S1.0078

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To realize the simulation for three-dimensional thermal hydraulic characteristics of offshore nuclear power plants and provide technical support for operation training, accident diagnosis and safety analysis, this study, based on the PANTHER code, has developed the 1D/3D additional force model, the 3D motion coordinate calculation model, the 3D non-inertial coordinate system motion calculation model and the IO integrated interaction module, which are integrated into the RINSIM simulation platform. Real-time interaction of parameters and on-line switching of motion conditions are realized, and the comparative test verification is completed based on the two-loop single-phase natural circulation test device. The verification results show that the calculation results of each motion condition meet the physical laws, and the calculation error with the experimental value is below 5%, which proves the reliability of the code calculation results under the condition of marine motion. Therefore, the 3D thermal-hydraulic system analysis code developed in this study can be used to simulate the 3D thermal-hydraulic characteristics of offshore nuclear power plants.

Optimization of Nuclear Power Accident Diagnosis Procedures Based on SAC Reinforcement Learning

Zhang Dazhi, Wang Zhihui, Zhou Huabing, Fu Yongjie, Xi Jiaxuan

2024, 45(S1): 85-90. doi: 10.13832/j.jnpe.2024.S1.0085

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This paper proposes an optimization method for nuclear accident diagnosis procedures based on the Soft Actor-Critic (SAC) reinforcement learning model. Using a decision tree model as the foundation to optimize the judgment strategy of accident detection procedures, which significantly improves the performance of accident detection while maintaining the interpretability of the decision model. The model employs SAC as the reinforcement learning algorithm, which defines the state as a combination of current operating data and historical data, sets the actions as the adjustment of the decision threshold of diagnostic procedures, and reflects the accuracy of diagnosis through the returns. With the help of SAC algorithm, the system constantly adjusts the threshold to optimize the strategy to obtain the best diagnosis effect. In a simulated Main Steam Line Break (MSLB) accident scenario, the model can better adapt to and comprehend complex high-dimensional data, find the optimal control strategy under specific performance indicators, and the accuracy is steadily approaching 1. The proposed method significantly reduces the false positive rate, and it not only detects nuclear power accidents more accurately, but also shows excellent results in reducing false alarms, thus improving the safety of nuclear power operation.

Study on Internal Flooding in Nuclear Power Plants Based on 3D Simulation Software CNIFA

Zhao Ziwei, Wang Chenhui, Huang Xiaoyun

2024, 45(S1): 91-95. doi: 10.13832/j.jnpe.2024.S1.0091

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In order to carry out safety assessment for internal flooding of nuclear power plants, the 3D simulation software CNIFA with independent intellectual property rights is developed, which is specially used for the analysis of internal flooding in nuclear power plants. In this paper, the 3D simulation software is used to analyze the internal flooding of the small reactor (ACP100) building, and suggestions for design improvement are put forward. Firstly, the physical model of scaled-down building is built, and the flooding spread in the room of nuclear power plant is simulated by experiments. Compared with the simulation results of CNIFA software, the results show that the software simulation has high accuracy. At the same time, the internal flooding of ACP100 reactor building is analyzed. Through software simulation analysis, it is known that the flooding height of the lowest floor room of the reactor building reaches 1.367 m, and relevant drainage measures should be considered. The 3D simulation analysis of internal flooding by CNIFA software has strong adaptability and good visualization, which can dynamically simulate the internal flooding scene of the nuclear power plant and show the whole process of flooding spreading, so as to formulate the flooding protection measures in a better way.

Study on Thermal Conductivity of Accident Tolerant Fuels using Laser-based Thermoreflectance Technology

Wang Yuzhou, Zhang Qiang, Ma Xianfeng, Zhu Fei, Liao Jingjing

2024, 45(S1): 96-102. doi: 10.13832/j.jnpe.2024.S1.0096

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In order to improve the thermal conductivity of new accident tolerant nuclear fuel and improve the detection efficiency and accuracy of thermal conductivity testing methods, this paper introduces the laser-based thermoreflectance technology with high spatial resolution and high testing frequency, expounds the basic principle, experimental equipment and testing method of this technology, and emphatically introduces the space-domain thermoreflectance technology developed for nuclear fuel research. Taking ion irradiated samples and coated nuclear fuel coatings as examples, the application scenarios of space-domain thermoreflectance technology in the field of in-situ testing are introduced. Aiming at the inhomogeneous damage distribution of ion irradiated samples, a multi-layer heat transfer model was developed to characterize the thermal conductivity of materials more accurately, and the quantitative attenuation law of thermal conductivity of UO₂ fuel with ion implantation dose was obtained by using this method. The heat transport properties of coatings in microencapsulated nuclear fuel particles at high temperature were accurately characterized, and the influence of defects on the thermal conductivity of pyrolytic carbon coatings was revealed by combining microstructure research. The thermoreflectance technology provides a powerful tool for investigating the impact of irradiation damage and structural defects on the thermal transport in nuclear fuel, and provides a reference for further improving fuel performance and developing high-fidelity simulation codes.

Experimental Investigation for the Fretting Wear of Cr-coated Zircaloy Cladding

Yang Siyuan, Yuan Bo, Wen Qinglong, Wen Shuang, Zhang Ruiqian, Yang Hongyan

2024, 45(S1): 103-109. doi: 10.13832/j.jnpe.2024.S1.0103

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To study the mechanism and microscopic mechanism of fretting wear of Cr-coated zircaloy cladding under the condition of turbulent excitation of PWR, this paper mainly takes Cr-coated zircaloy cladding as the research object, carries out experimental research on fretting wear under multi-parameter coupling, and expounds the influence laws of parameters such as frequency, load, displacement and cycle times on fretting wear. The maximum wear depth and wear volume of 19 sets of fretting wear experiments were obtained, of which the maximum wear depth was 12.052 µm and the maximum wear volume was 3.301×10⁻³ mm³. The results show that the main influencing parameters of fretting wear include fretting amplitude, normal load, cycle times and material hardness, while fretting frequency has little influence on wear volume and maximum wear depth. The least square method is used to fit the experimental results to obtain the wear volume calculation formula, and the deviation between the experimental value and the calculated value of the formula is within ±50%. This study provides data support for the evaluation of wear resistance of Cr-coated zircaloy cladding.

Development and Application of a Mechanical Model for Multilayer Anisotropic Cladding

Zhang Ruixiao, He Yanan, Wu Yingwei, Tian Wenxi, Qiu Suizheng, Su Guanghui

2024, 45(S1): 110-116. doi: 10.13832/j.jnpe.2024.S1.0110

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The multilayer SiC composites cladding composed of monolithic SiC and SiC fiber/matrix composite (SiC_f/SiC) is a popular choice for accident-tolerant fuel cladding. However, typical fuel performance analysis codes currently lack the modeling capability for the anisotropic mechanical behavior of SiC_f/SiC materials. In order to enhance the precision of mechanical calculations for composite SiC cladding in fuel performance analysis, a mechanical model was developed for multilayer anisotropic materials and integrated into the fuel performance analysis code FRAPCON4.0. Validation of the model's accuracy was conducted using a multilayer SiC cladding thermal-mechanical coupling case. The mechanical calculation capability for SiC composite cladding with orthotropic mechanical properties and multiaxial pseudoplastic behavior was achieved, and the performance of duplex-layer SiC cladding fuel elements under normal operation condition was analyzed. The developed mechanical model in this study is adaptable to arbitrary multilayer cylindrical structure fuel elements and has the capability to analyze orthotropic mechanical parameters and behaviors, rendering it applicable to various types of novel fuel element analyses.

Verification of Sodium-cooled Fast Reactor SUPERFACT-1 SF4/SF16 Fuel Rod Experiment using LoongCALF Code

Peng Xinhang, Zhang Tian, Shao Shihao, Liu Zhouyu

2024, 45(S1): 117-122. doi: 10.13832/j.jnpe.2024.S1.0117

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Due to the high power density and deep burn-up of sodium-cooled fast reactor, its fuel has the characteristics such as high temperature, high fission gas release rate, large deformation, and the formation of central voids during operation. Therefore, the sodium-cooled fast reactor poses new challenges to the development of fuel performance codes. The LoongCALF code is a fast reactor fuel performance analysis code based on the finite element method and JFNK method. To verify the applicability of the LoongCALF code in the analysis of sodium-cooled fast reactor fuel performance, this work uses the LoongCALF code to simulate the SF4/SF16 fuel rods in the SUPERFACT-1 irradiation experiment, and compares the simulation results with those of fast reactor fuel performance codes such as TRANSURNUS, GERMINAL and MACROS in public literature. The research results show that the cladding temperature, fuel rod internal pressure, and pellet temperature calculated by the LoongCALF code are in good agreement with the literature results, and the axial central void diameter is in good agreement with the experimental results, which can meet the needs of sodium-cooled fast reactor simulation. Therefore, the LoongCALF code can be used for the simulation work of sodium-cooled fast reactors, but the related models of fission gas release and gap width need to be further improved.

Effect of pH on High-temperature Electrochemical Corrosion Behavior of AlCrNbSiTi High Entropy Alloy Coatings on Zr-Sn-Nb Alloy

Wang Yuhui, Liu Chao, Hu Yong, Peng Dequan

2024, 45(S1): 123-129. doi: 10.13832/j.jnpe.2024.S1.0123

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In order to improve the corrosion resistance of Zr-Sn-Nb alloy cladding material in PWR, the AlCrNbSiTi high entropy alloy coatings were prepared on Zr-Sn-Nb alloy substrates by arc ion plating, and the open-circuit potential and potentiodynamic polarization electrochemistry were measured in three different pH water environments. The morphology, chemical composition and structure of the surface oxide film were analyzed by microscopic characterization, and the influence of pH change on the high-temperature electrochemical behavior of AlCrNbSiTi high entropy alloy coating was studied. The results show that with the increase of pH (7.4-8.5), the open circuit potential of the coating decreases, the corrosion current density of polarization curve increases and the corrosion degree of the coating surface intensifies. Therefore, the corrosion resistance of AlCrNbSiTi high entropy alloy coating decreases with the increase of pH(7.4~8.5).

Research Progress and Development Trend of Accident Tolerant Fuel UN Pellets

Chen Xiangyang, Ding Yang, Ding Jie, Li Cong, Zhang Xintao

2024, 45(S1): 130-137. doi: 10.13832/j.jnpe.2024.S1.0130

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UN pellet has high uranium density and high thermal conductivity, and is a kind of potential accident tolerant fuel pellet. In this paper, the research progress of UN pellets is summarized from five aspects: preparation process and physical properties, environmental compatibility, irradiation properties, pellet-cladding interaction, economy and safety. The research results show that the advantages of using UN pellets in PWR outweigh the disadvantages, and it is generally beneficial to promote the safety of the reactor under accident conditions. It has the remarkable characteristics of reducing the operating temperature of pellets and reducing the release of energy storage in accidents. The main problems to be solved are poor water corrosion resistance and high cost of ¹⁵N enrichment. The possible solutions to improve water corrosion resistance and oxidation resistance include doping or adding antioxidant components, and the high cost problem needs to reduce the cost of ¹⁵N enrichment. This review comprehensively summarizes the overall research progress and development trend of UN pellets, and provides reference for understanding its feasibility and existing problems as accident tolerant fuel pellets.

Safety Analysis on Accident-tolerant Fuel during LBLOCA Based on LOCUST Code

Xiong Yiran, Ma Zehua, Liang Ren, Lin Zhikang, Ju Zhongyun, Peng Zhenxun

2024, 45(S1): 138-144. doi: 10.13832/j.jnpe.2024.S1.0138

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Accident tolerant fuel (ATF) aims to improve the safety characteristics of nuclear fuel under normal operating conditions and accidents. In order to evaluate the safety performance of ATF in large break loss-of-coolant accident (LBLOCA) of large commercial pressurized water reactor, based on LOCUST code, this study analyzes and describes the main thermal hydraulic phenomena and key influencing parameters of HPR1000 using using UO₂-Cr coated zirconium alloy cladding fuel at different stages of LBLOCA. The results indicate that compared to the traditional UO₂-Zr fuel, UO₂-Cr coated zirconium alloy cladding fuel can reduce the peak cladding temperature (PCT) and the thickness of cladding oxide film under LBLOCA, improve the safety margin in accidents, and have better accident-tolerance characteristics.

Study on Internal Pressure Burst, Creep and Fatigue Properties of ODS-FeCrAl Alloy Tube

Liu Yang, Lu Zhiwei, Ge Hongen, Wu Lixiang, Xue Jiaxiang, Liao Yehong

2024, 45(S1): 145-151. doi: 10.13832/j.jnpe.2024.S1.0145

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The microstructure, internal pressure burst, creep and fatigue properties of oxide dispersion strengthened (ODS)-FeCrAl alloy tube were studied by transmission electron microscopy (TEM), internal pressure burst, creep and fatigue test machine. The results show that a large number of nano-second phase particles are dispersed in the matrix of ODS-FeCrAl alloy tube, with an average diameter of about 8.76 nm and a volume density of 6.8×10²² m^–3. The burst strength of ODS-FeCrAl alloy tube is 1158 MPa at room temperature; The burst strength of the ODS-FeCrAl alloy decrease gradually with the increase of temperature; At 1000℃, the ODS-FeCrAl alloy tube does not lose its pressure-bearing ability, and its burst strength is 81 MPa. The ODS-FeCrAl alloy tube shows excellent internal pressure creep resistance (the creep deformation is 0.9%) under the test condition of 350℃/30 MPa. When the peak fatigue loading pressure is lower than 30 MPa at 350°C, there is no fatigue failure of ODS-FeCrAl alloy tube after 1000000 cycles of loading. The internal pressure burst, creep, and fatigue properties of ODS FeCrAl alloy tube are significantly excellent.

Numerical Study on Mechanical Characteristics of Storage Canister Module for Spent Fuel Dry Storage System

Yuan Bo, Chen Kang, Wen Qinglong, Xu Shijia, Cheng Cheng, Nie Zhaoyu, Xu Xiao

2024, 45(S1): 152-158. doi: 10.13832/j.jnpe.2024.S1.0152

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Spent fuel dry storage is an important method for handling spent fuel, however, the high temperature generated by spent fuel may cause great thermal stress, thereby leading to permanent deformation and damage. Therefore, it is of great significance to study the mechanical properties of storage canister module in spent fuel storage system. In this paper, the storage canister module in the spent fuel dry storage system is taken as the research object, and a 1/2 scaled model of the storage canister module is established. Based on the calculated temperature distribution of the storage canister module, the numerical study of the mechanical characteristics of the storage canister module under normal storage conditions is carried out, which provides data support for the scale test of the spent fuel dry storage system. The results show that: ①The stress in grid is high in center and low around, and there is a large shear stress in the top and bottom parts, while the overall stress of the aluminum support block is small, and there is a large stress at the junction of the canister shell and the top cover plate; ②Under the lowest ambient temperature condition, the maximum stress of the grid, the aluminum support block and the storage canister is 253.71 MPa, 89.99 MPa and 55.35 MPa, respectively. The stress of each component does not exceed the limits.

Effect of Cold-rolling Deformation on Microstructure and Mechanical Properties of Fe-11Cr-5Al-2Mo Alloy

Wang Xinmin, Wang Yurong, Wu Yu, Pan Qianfu, Yao Lifu, Xu Qi

2024, 45(S1): 159-166. doi: 10.13832/j.jnpe.2024.S1.0159

Abstract(59) HTML (25) PDF(10)

Abstract:
The purpose of the study is to understand the cold deformation rule of Fe-11Cr-5Al-2Mo alloy, so as to provide data and theoretical basis for the cladding tube manufacturing. The rolling experiments of Fe-11Cr-5Al-2Mo alloy plates with different cold-rolled deformation were carried out. The cold-rolled plates were annealed at 800℃ for 1 h, and the microstructure evolution of the cold-rolled plates and annealed plates was characterized. The strength and plasticity of Fe-11Cr-5Al-2Mo alloy were evaluated by the tensile test at room temperature. The deformed microstructure of Fe-11Cr-5Al-2Mo alloy plate is elongated with the increase of cold rolling deformation, and the strength of the plate increases with the increase of cold rolling deformation. After annealing at 800℃ for 1h, the plate with 50% cold rolling deformation showed the smallest and uniform recrystallization microstructure, and had a good ductile-plastic fit. In this paper, the relationship between cold rolling deformation, microstructure and properties of the alloy is expounded, which provides a reference for the design of cold rolling deformation and the study of annealing process of Fe-11Cr-5Al-2Mo alloy tube.

Study on Bubble Behavior Mechanism of Saturated Pool Boiling on SiC Cladding Material Surface under Atmospheric Pressure

Jin Desheng, Yan Yalun, Cheng Yanhua, Fu Xuefeng, Peng Zhenxun, Liao Yehong, Mao Yulong

2024, 45(S1): 167-174. doi: 10.13832/j.jnpe.2024.S1.0167

Abstract(43) HTML (15) PDF(14)

Abstract:
SiC material is a kind of accident tolerant fuel (ATF) with high temperature resistance and good oxidation resistance. Its heat transfer and critical heat flux (CHF) are important indexes to evaluate the material performance, and the study of bubble behavior mechanism is helpful to evaluate its heat transfer performance. In this paper, the bubble behavior mechanism of SiC cladding material was studied by using the visualization experimental device of pool boiling at atmospheric pressure, and the bubble growth and detachment process in different sections of pool boiling curve were analyzed. According to the observed images of pool boiling bubbles on the surface of SiC cladding, the whole pool boiling heat transfer process can be divided into four sections: natural convection section, isolated bubble nucleate boiling section, slug bubble nucleate boiling section and film bubble nucleate boiling section. In the nucleate boiling zone of isolated bubbles, the growth time of bubbles on the surface of SiC cladding is short and the detachment frequency of bubbles is high. In the nucleate boiling zone of slug bubbles, a large number of bubbles are produced on the surface of SiC, and the interaction between bubbles is intense, resulting in strong heat transfer on the surface of SiC. The relationship between contact angle and detachment diameter is established, which can provide important support for the establishment of subsequent heat transfer model.

Investigation on Corrosion Model of Cr-coated Zirconium Alloy Cladding

Shen Yong, Zeng Xiehu, Duan Zhengang, Wen Qinglong, Yuan Bo, He Liang, Gao Shixin

2024, 45(S1): 175-180. doi: 10.13832/j.jnpe.2024.S1.0175

Abstract(91) HTML (17) PDF(22)

Abstract:
As one of the candidate materials for accident tolerant fuel (ATF) cladding, Cr coating can significantly improve the corrosion resistance and oxidation resistance of zirconium alloy cladding, which is expected to prolong the service life. In order to evaluate the corrosion and oxidation behavior of Cr-coated zirconium alloy cladding, a corrosion model of Cr-coated zirconium alloy cladding under normal operating conditions of PWR was established in this paper, and the model was verified based on the experimental data in the literature. Based on this model, the effects of heat flux and mass flow rate on the corrosion of Cr-coated zirconium alloy cladding were analyzed. The results show that the corrosion thickness increases with the increase of heat flux. In addition, the increase of coolant mass flow rate leads to the decrease of cladding wall temperature, which eventually leads to the decrease of cladding corrosion thickness.

Experimental Study on Steam Reforming Engineering of Radioactive Waste Oil

Gao Ruixi, Lin Li, Liang Yi, Li Zhenchen, Zhang Hangzhou

2024, 45(S1): 181-185. doi: 10.13832/j.jnpe.2024.S1.0181

Abstract(39) HTML (11) PDF(8)

Abstract:
In order to achieve the objectives of minimizing radioactive waste and safer disposing of waste, the steam reforming technology for radioactive organic waste was selected in this study, and the cold test was carried out on the successfully developed steam reforming engineering prototype with clean engine oil and simulated waste oil. The cold test mainly verifies the volume reduction effect of the engineering prototype and measures the tail gas. The test results show that the capacity reduction multiple of the engineering prototype is greater than 7, and the waste treatment capacity is 5 kg/h. The tail gas discharged by the tail gas treatment system meets the first-class emission requirements of GB 18484—2020 .

Effect of Molybdenum Electrode Heating on Microstructure and Properties of Radioactive Borosilicate Glass

Li Xiaobo, Yuan Qingqing, Li Pingchuan, Cheng Changming, Tang Deli, Ye Xinnan, Wang Wei

2024, 45(S1): 186-191. doi: 10.13832/j.jnpe.2024.S1.0186

Abstract(55) HTML (25) PDF(6)

Abstract:
Borosilicate glass is widely used in the solidification of radioactive waste. In the plasma high temperature melting system, the bottom of the borosilicate glass melting pool containing radionuclides and the glass body far from the heat source have low temperature, which leads to poor flow uniformity of the molten pool and greatly hinders the process of waste discharge. In this paper, the effect of molybdenum electrode on the structure and properties of radioactive borosilicate glass is investigated by introducing molybdenum electrode to assist heating the molten pool. The experimental results show that sodium molybdate and molybdenum oxide can be avoided when the mass fraction of Na₂O in the glass formula is 10%. After the molybdenum electrode was put into use, the initial high resistance characteristics of the molten glass between the electrodes made it release a lot of Joule heat, and the temperature distribution uniformity and fluidity of the molten pool were improved, thus effectively promoting the discharging process. The weight, density and compressive strength of the discharged glass increased by about 8.2%, 14.6% and 13.6% respectively. This paper can provide some solutions for developing a high temperature plasma melting system with large waste capacity and long running time.

Design Research and Verification of Pneumatic Driven Long-distance Transportation of Spent Resin

Liu Wenlei, Zou Qinghua, Li Zhenchen, Lin Li, Luo Feng, Chen Li, Zhang Jianbing

2024, 45(S1): 192-196. doi: 10.13832/j.jnpe.2024.S1.0192

Abstract(50) HTML (17) PDF(3)

Abstract:
In order to solve the engineering problem of local radioactive hot spots caused by spent resin deposition in the process of radioactive spent resin long-distance pipeline transportation, a set of 1300 m spent resin ultra-long-distance pipeline system was constructed on the basis of theoretical calculation of pneumatic driven flow, comprehensive stress analysis of pipeline and engineering design combined with field practice. The pipeline adopts the design idea of overall smooth inner wall, pipeline self-compensation for the thermal expansion compensation, large bending radius square compensator and space Z-shaped compensation structure, etc. The valves and fittings are customized to match the inner diameter of the pipeline. After installation, the hydraulic test, simulated transportation test and engineering verification are carried out. The verification results show that the spent resin transportation pipeline has solved the engineering problem of local hot spots caused by spent resin deposition during long-distance pipeline transportation. The system has high inherent safety and simple operation, which significantly shortens the operation time, reduces the generation of secondary radioactive waste, reduces the labor intensity and radiation dose of personnel, and has good engineering application effect.

Development of Docking Device for Discharge Pipe of Vitrification Melter

Zhou Qiang, Wu Shuaizhen

2024, 45(S1): 197-202. doi: 10.13832/j.jnpe.2024.S1.0197

Abstract(67) HTML (18) PDF(7)

Abstract:
Joule heated ceramic melter is the core equipment in the high level liquid waste vitrification process, and the vitrified waste in the melter is poured into the canister by means of bottom discharge. In order to repair the melter discharge pipe, the damaged discharge pipe was inspected, a special device was developed to cut, grind and clean the upper section of the discharge pipe, and a docking device was designed for easy installation, fixing and disassembly. Various sealing structures designed were tested by experimental methods, and the optimal sealing scheme was selected. Through the test, the reliability of the overall structure and thermal compensation function of the docking device was verified. The installation and use of the docking device show that it is simple in structure, convenient to install and reliable in sealing. The function of the discharge pipe is successfully restored, and the replacement of high-level vitrified waste in the melter is successfully completed, which creates favorable conditions for ensuring the good state of the melter and welding repair of the discharge pipe, and accumulates experience for similar work in radioactive high-temperature environment.

Experimental Study on Plasma High Temperature Melting Treatment Process

Yuan Qingqing, Li Xiaobo, Zhang Zhenghao, Cheng Changming, Zhang Xiaojie, Ye Xinnan, Wang Wei

2024, 45(S1): 203-207. doi: 10.13832/j.jnpe.2024.S1.0203

Abstract(104) HTML (23) PDF(7)

Abstract:
The plasma high-temperature melting engineering prototype was used to conduct high-temperature pyrolysis and melting treatment on different kinds of waste samples, mainly including simulated nuclear power plant technical waste, resin waste and waste oil adsorbent, to verify the treatment effect of this technology on different wastes. This experimental study mainly introduces the results of different treatment conditions and the operation state of each system during the experiment, and studies the overall operation performance and corresponding optimization scheme of the system under different conditions. The experimental results of plasma high-temperature melting treatment show that the system has different treatment capabilities for different types of simulated nuclear power plant low-level waste, and can meet the design requirements in engineering applications. It is suggested to strengthen the design connection before and after the main process system of plasma high-temperature melting, reduce mutual constraints, and improve the treatment capacity and effect of the system.

Overview of Research on Corrosion Properties of Additively Manufactured Products in the Nuclear Field

Zhang Yue, Lan Yang, Wang Chengyu, Yang Sha

2024, 45(S1): 208-214. doi: 10.13832/j.jnpe.2024.S1.0208

Abstract(53) HTML (17) PDF(5)

Abstract:
The corrosion resistance of additively manufactured products in the nuclear field is related to the service life and operation safety of the reactor system, so the corrosion resistance is very important. In this paper, the basic concepts of research on corrosion properties of additively manufactured products in nuclear field are summarized, and the research and development status of ex-core corrosion properties of additively manufactured products is summarized. Based on the comprehensive analysis of the types of additive manufacturing products in the nuclear field and various additive manufacturing processes such as laser powder bed fusion, directional energy deposition and laser engineered net shaping, the characteristics of the corrosion properties of additive manufacturing products are discussed. It is concluded that the corrosion resistance of additively manufactured products is different due to different manufacturing processes, material reprocessing and corrosion conditions, such as doping a small amount of hafnium, hot isostatic pressing and solution annealing, which can improve the corrosion resistance of additively manufactured products in the nuclear field. Through comprehensive demonstration and analysis, it provides ideas and methods for understanding, deepening and expanding the basic research, technical means and application of corrosion properties of additively manufactured products in the nuclear field.

Dual-sided Trench Shaped Neutron Detector Using ⁶LiF/α-Al₂O₃:C Based on Optically Stimulated Luminescence

Fan Haijun, Cui Hui, Wang Shanqiang, Wang Zungang, Zhou Hongzhao, Chen Wenzhuo, Tang Kaiyong

2024, 45(S1): 215-220. doi: 10.13832/j.jnpe.2024.S1.0215

Abstract(51) HTML (19) PDF(11)

Abstract:
Personal neutron dosimetry is of great significance to the staff of nuclear facilities such as nuclear power plants, nuclear power units, research reactors and high-energy accelerators. Optically stimulated luminescence (OSL) technique has the advantages of fast reading speed and repeated reading many times, and is an important development direction of neutron personal dose monitoring. In this paper, a dual-sided trench shaped optically stimulated luminescence neutron detector (DS-TSOSLND) using ⁶LiF/α-Al₂O₃:C is designed, and the Monte Carlo code Geant4 is used to calculate and analyze the influences of different trench width, depth and trench ratio on the detector performance, so as to explore its neutron detection mechanism. Based on the simulation results of Geant4, DS-TSOSLND was successfully developed by combining the current processing conditions of α-Al₂O₃:C crystal microstructure. The test results of ¹³⁷Cs source and heavy water moderated ²⁵²Cf neutron source show that the neutron detection threshold of the newly developed DS-TSOSLND is 10.3 μSv, and the neutron dose response is linear in the range of 0.05~20 mSv, which has broad application prospects in the field of personal neutron dosimetry.

Numerical Simulation and Optimization Analysis of Labyrinth Screw Pump in Nuclear Power Reactor Coolant Pump

Xu Xi, Pan Weilong, Cai Liang, Xie Jianghong, He Shaohua, Xu Yu, Fan Xueqing

2024, 45(S1): 221-226. doi: 10.13832/j.jnpe.2024.S1.0221

Abstract(52) HTML (22) PDF(10)

Abstract:
In order to improve the wear phenomenon caused by cavitation on the surface of the thrust bearing in a certain nuclear power plant and extend the service life of reactor coolant pump, it is considered to add a second-stage labyrinth screw pump to the lubrication circuit of the thrust bearing in the reactor coolant pump. By installing a pressure boosting device, the circuit pressure is increased and cavitation is reduced. In this paper, the three-dimensional model of the first-stage labyrinth screw pump in the reactor coolant pump is established, and the numerical calculation is carried out by using FLUENT software to analyze the pressure, velocity and temperature distribution of the flow field. The calculation results are compared with the actual operation results of the nuclear power plant to verify the accuracy of the model and the numerical calculation method. A second-stage labyrinth screw pump is designed and numerically simulated. It is found that after adding the second-stage labyrinth screw pump, the pressure rise at the inlet and outlet of the lubrication circuit is 3.30×10⁵ Pa, the head is increased by 33.67%, and the temperature rise at the inlet and outlet section is 3 K. The research results have a certain reference value for the design and application of labyrinth screw pump in engineering practice.

2024 Vol. 45, No. S1