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2024 Vol. 45, No. S2

Column of Digital Reactor
Overall Architecture and Key Technologies of Digital Reactor
Zeng Wei, Wang Conglin, Liu Chengmin, Liu Jia, Li Songwei, Gong Zhaohu, Wang Jie, Huang Qingyu, Fang Haoyu
2024, 45(S2): 1-13. doi: 10.13832/j.jnpe.2024.S2.0001
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A digital reactor represents a comprehensive integrated platform that harnesses the power of high-precision numerical simulations, interdisciplinary data models, big data, and artificial intelligence to implement the full life-cycle management of nuclear reactors within a digital realm. To address the demands of multi-level business activities of nuclear reactors, leveraging advancements in digital technology, a digital reactor platform architecture characterized by a unified platform and two fundamental cores has been developed. Furthermore, the design of the technical architecture has been implemented through comprehensive integration and application, numerical computation cores, and big data frameworks, thereby empowering the full life-cycle management of nuclear reactors. The key technologies, problems and challenges in the application of digital reactor are analyzed and studied, and the subsequent development trend is analyzed and prospected.
Research on the Architecture Design and Modularization Technology of Digital Reactor Platform
Hao Jiangtao, Li Songwei, Zeng Wei, Liu Chengmin, Fang Haoyu, Liu Jia, Yuan Peng, Chen Di
2024, 45(S2): 14-19. doi: 10.13832/j.jnpe.2024.S2.0014
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Based on the architecture description method and modularization description method, combined with the characteristics of nuclear reactor research and development and the theoretical method of system engineering based on models, the architecture design and key technology study of modularization for the digital reactor platform are carried out. Through the analysis of the models, algorithms, data, and processes carried by the digital reactor platform, as well as the complex heterogeneous software and hardware environment centered on the supercomputer, the business architecture, application architecture, data architecture, and technical architecture design of the digital reactor platform are proposed to guide the research and development of the digital reactor platform. Through the functional verification of the digital reactor platform, it is shown that under the guidance of the architecture of the digital reactor platform, the platform has reconfigurability, reuseability, scalability, and robustness. The developed functions can meet the engineering requirements, and its architecture design practice has an important guiding significance for the subsequent design of complex system platforms.
Research on the Overall Design Method of Nuclear Reactor Based on MBSE
Wang Shuai, Zhang Xiaohua, Li Haiying, Yu Ping
2024, 45(S2): 20-27. doi: 10.13832/j.jnpe.2024.S2.0020
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Nuclear reactor engineering is a typical multidisciplinary and interdisciplinary complex system engineering with complex technology and high reliability requirements. Model based system engineering (MBSE) is recognized as one of the most likely design techniques to meet the needs of complex systems engineering in the digital context. This paper focuses on the internal requirements of R&D and design of nuclear reactor and coolant system, and discusses the application and role of model-based system engineering design method in the R&D of PWR. By discussing the design method of MBSE, the advantages of MBSE are compared and explored. The MBSE modeling of the PWR nuclear reactor and coolant system is carried out. The functional behavior analysis of the system is carried out by using the positive design concept, and the system functional decomposition and overall architecture design are completed, which can guide the tradeoff design and scheme optimization of the functional architecture of nuclear reactor and coolant system.
Global Optimization Analysis of Nuclear Power Plant Parameters Based on Modelica
Hao Chengming, Liu Lizhi, Han Yifu, Xia Junbao, Yu Qiao, Liu Chengmin, Wang Jie
2024, 45(S2): 28-34. doi: 10.13832/j.jnpe.2024.S2.0028
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To solve the problem of low efficiency caused by the need to determine the configuration mode in advance when developing design optimization codes for nuclear power plant, this study explores the implementation method of the global optimization analysis of nuclear power plant parameters based on Modelica. The standardized model construction of the primary circuit system and equipment of the nuclear power plant is carried out, the hierarchical framework from the bottom evaluation model to the top system model is established, and the free configuration and visual design modeling of the primary circuit system and equipment are realized. With coupling sensitivity analysis tool and multi-objective optimization algorithm, sensitivity analysis and multi-objective optimization of typical design parameters of the primary circuit system are realized. Taking weight, volume, natural circulation capacity, safety (the minimum departure from nucleate boiling ratio, MDNBR), and thermal efficiency of the plant as optimization objectives, a multi-objective optimization case study was carried out for the nuclear power plant studied. The research results show that the global parameter optimization tool of nuclear power plant based on Modelica can realize the global multi-objective optimization design of nuclear power plant. The hierarchical modeling method used by the tool is standardized, efficient, and flexible, and is a technical approach suitable for solving the overall design optimization problem of a model-based nuclear power plant system.
Development and Validation of Multi-physics Coupling Framework for Reactors
Pan Junjie, Qiang Shenglong, Gong Zhaohu, Tang Qifen, Li Zhigang, Zeng Wei, Wang Yuan, Wang Jie, Fan Jiakun
2024, 45(S2): 35-41. doi: 10.13832/j.jnpe.2024.S2.0035
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Developing coupling codes based on a coupling framework and using the framework for coupling data transmission and process control can improve the development efficiency and reduce difficulty. This paper introduces the overall architecture, core modules, and use of the coupling framework named MORE. Through verification and application, it has been shown that the coupling framework MORE can quickly establish the coupling process, achieve high-precision mapping and transmission of coupling data between different codes, greatly improve the development efficiency of coupling codes, and ensure that the flexibility of modeling and grid division of various professional codes for coupling is not affected, thus realizing the autonomy of the coupling framework.
Validation of HPR1000 Core Modeling and Startup Test with Three-dimensional Characteristic Neutronics Calculation Code SHARK
Wang Bo, Zhao Wenbo, Zhang Hongbo, Zhao Chen, Chen Zhang, Liu Kun, Zhang Lerui, Gong Zhaohu, Zeng Wei, Li Qing
2024, 45(S2): 42-48. doi: 10.13832/j.jnpe.2024.S2.0042
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To validate the accuracy and applicability of the three-dimensional characteristic neutronics calculation code SHARK for large PWRs, the startup test of HPR1000 is selected. The HPR1000 is a third-generation nuclear pressurized water reactor with independent intellectual property rights in China. The validation contents include critical effective multiplication coefficient keff, control rod integral value and assembly power distribution. The results show that the critical effective multiplication coefficient keff, assembly power distribution and control rod integral value are in good agreement with the measured values. Therefore, the code SHARK can be applied to the physical calculation of digital reactors with good calculation accuracy.
Numerical Verification of SP3 Based Computational Physics Code for Rod-type PWRs
Liu Kun, Zhao Wenbo, Gong Zhaohu, Chen Zhang, Chai Xiaoming, Zhang Bin, Fang Chao, Zeng Wei
2024, 45(S2): 49-54. doi: 10.13832/j.jnpe.2024.S2.0049
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In order to verify the new R&D code, a new generation of Pin-by-pin solution package KYLIN V2.0-CORCA-SPn is used to verify and analyze the measured data of the first cycle of Fuqing Unit 5. The results show that the maximum integral rod worth difference is 7.17%, which occurs in N2 rod bank, while the others are lower than 5%. The maximum critical reactivity difference in the cycle is −0.559%. The deviation between fuel pin power and RMC code is within 8%. The numerical model proposed in this paper has high discrete accuracy and good computational stability, and the relevant verification work can provide technical support for the core code design of a new generation of pressurized water reactors, thus meeting the research and development design requirements of new complex reactors.
Extended Development and Application of CAD-based Transport in RMC
Shen Pengfei, Wang Kan, Liu Zhaoyuan, Liang Jingang, Liu Shichang
2024, 45(S2): 55-62. doi: 10.13832/j.jnpe.2024.S2.0055
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To enhance the capability of CAD-based transport in RMC, and provide a new calculation method for the CAD-based geometry transport modeling in Monte Carlo (MC) for new numerical reactors, in this study, firstly, an open-source meshing kernel is employed to process CAD models, realizing an autonomous CAD-based transport method throughout the entire process in the Monte Carlo code RMC. Subsequently, a hybrid geometric transport framework combining CAD and CSG geometry is developed, which improves the efficiency of geometric processing and Monte Carlo transport. The correctness of the autonomous CAD geometric processing method and the hybrid geometric transport method is verified through numerical calculations of fuel spheres, VERA-3A fuel assembly and other models. Compared with the existing CAD geometric modeling transportation methods, the hybrid geometric modeling transportation method has significantly improved the geometric processing time and the efficiency of Monte Carlo calculation. Therefore, the enhanced CAD-based transport method in RMC developed in this study can be used for neutron transport analysis calculations of new numerical reactors.
Development and Application of Autonomous Computational Fluid Dynamics Code WINGS-CFD for Nuclear Reactors
Deng Jian, Wei Zonglan, Zeng Wei, Li Songwei, Qiu Zhifang, Liu Luguo
2024, 45(S2): 63-69. doi: 10.13832/j.jnpe.2024.S2.0063
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To meet the requirements of high-precision numerical simulation of reactor flow and heat transfer, this article introduces WINGS-CFD (Workbench of Intelligent Nuclear reactor desiGn and Simulation-Computational Fluid Dynamics), an autonomous Computational Fluid Dynamics (CFD) code for nuclear reactors developed by Nuclear Power Institute of China, which is designed based on the object-oriented and hierarchical architecture principles with a high degree of extension. This article comprehensively outlines the overall design concepts of WINGS-CFD, covering aspects such as theoretical models, numerical discretization methods, and code architecture. Numerical calculations for typical reactor scenarios involving flow and heat transfer conditions are conducted using WINGS-CFD. The results show that the accuracy of WINGS-CFD calculation results is equivalent to that of commercial CFD code. WINGS-CFD has excellent parallel performance, which can support large-scale numerical simulation of billions of grids and coupled simulation of neutron transport and flow and heat transfer, and provides an autonomous numerical technique for refined multi-physical field analysis of reactor system.
Investigation of Pivotal Models for Subcooled Boiling Based on the Eulerian-Eulerian Framework
Luo Hanwen, Wang Hongbin, Xiong Jinbiao
2024, 45(S2): 70-76. doi: 10.13832/j.jnpe.2024.S2.0070
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Wall boiling, interphase interaction and bubble coalescence/breakup models involved in the Eulerian-Eulerian two-fluid model framework within Computational Fluid Dynamics (CFD) are investigated. In the framework, the five-component wall boiling model is employed to account for the on-wall bubble behavior, including nucleation, growth, detachment, sliding and lift-off. The inhomogeneous multi-size group (iMUSIG) model is utilized to consider the bubble size distribution and its effects on interphase interaction. The Prince-Blanch model and Luo-Svendsen model are used to model the bubble coalescence/breakup. The model performance is assessed based on the DEBORA benchmark, which indicates that the models can accurately predict the distribution of void fraction and bubble size under the condition of high subcooling in normal operation of PWR. However, the void fraction in the pipe center in the case of low subcooling and high void fraction cannot be predicted. It is also found that the five-component model may underestimate the evaporation at the heated wall by comparing the calculated bubble velocity and liquid temperature with the measured ones.
Study on Non-classical PCMI Behavior Based on FUPAC3D
Liu Zhenhai, Zhou Yi, Qi Feipeng, Xin Yong, Li Wenjie, Gong Zhaohu, Zeng Wei, Zhang Tao
2024, 45(S2): 77-83. doi: 10.13832/j.jnpe.2024.S2.0077
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For the fuel rods of pressurized water reactor, strong mechanical interaction between the fuel pellet and cladding (PCMI) occurs under Class Ⅱ transient. When the pellet exhibits missing surface defects, the interaction are further intensified. Traditional 1.5D fuel performance analysis codes cannot analyze this non-classical PCMI phenomenon. In this paper, the developed 3D fuel performance analysis code FUPAC3D is used for simulation. Taking the common pellet end missing surface defect in manufacturing as an example, the impact of defect on the heat transfer and mechanical behavior of the fuel rod during transient is analyzed. The results show that the temperature distribution around the cladding near the pellet defect is uneven, with increased temperature at the edge of the defect and decreased temperature in the center. The cladding near the defect is indented inwards, causing an increase in the hoop stress of the cladding near the center of the defect. Compared to the corresponding "bamboo joint" part (the pellet end) of a complete pellet, it increases by about 14.3%, and compared to the middle plane of the pellet, it increases by about 62.9%.
Study on Simulation Technology of Cracking and Blistering of Particle Dispersed Plate Fuel Element
Liu Hongquan, Xiang Fengrui, Wu Yingwei, He Yanan, Zhang Jing, Su Guanghui
2024, 45(S2): 84-92. doi: 10.13832/j.jnpe.2024.S2.0084
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Cracking and blistering may occur when the particle dispersed plate fuel element runs in the reactor, which affects the safety of fuel and nuclear reactor. In this paper, numerical simulation research on cracking and blistering of particle dispersed plate fuel element is carried out. To realize the cracking analysis and blistering simulation from the fine-scale fuel particles to the macro-scale fuel elements, this paper first establishes a set of multi-dimensional multi-scale coupled simulation methods. With the multi-scale coupling scheme from fine-scale fuel particles to the macro-scale fuel element, the fuel element cracking risk region determination is realized. Based on the multi-dimensional coupling system that includes a 3D fuel element, a 1D coolant, and a 2D fuel slice, the extended finite element method (XFEM) is applied to realize the 2D cracking and blistering simulation of the fuel element. To accurately judge the occurrence of blistering, a fission gas internal pressure model is established in this paper, which is combined with the 2D extended finite element. Finally, the crack extension simulation of macroscopic fuel elements under thermal shock conditions is realized, and the temperature at which blistering occurs is predicted to be about 790 K when the fuel burnup is 120 MW·d·kg−1(U).
Calculation and Analysis of Multiscale Coupling of Dispersion Plate-type Fuel
Xiang Fengrui, He Yanan, Wu Yingwei, Qiu Suizheng, Su Guanghui
2024, 45(S2): 93-101. doi: 10.13832/j.jnpe.2024.S2.0093
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The multi-scale characteristics of two-phase dispersion materials in dispersion-type plate fuel pose challenges to its performance research. To accurately analyze and evaluate the in-pile performance of U3Si2/Al dispersion-type plate fuel, based on the existing equivalent physical properties and behavior model of U3Si2/Al dispersion fuel meat, the creep attenuation coefficient model was established to accurately simulate the creep behavior of two-phase composites, particularly focusing on the complex creep characteristics of the equivalent fuel meat. Additionally, a three-dimensional macroscopic fuel plate model and a one-dimensional sphere model were constructed, and a multi-scale coupling method was proposed to couple these models, enabling simultaneous simulation at both scales. Based on the multi-scale multi-physics coupling analysis tool, the multi-physical field analysis of U3Si2/Al dispersion-type fuel was conducted, and the impact of fuel particle size and volume fraction was evaluated. The results inidcate that increasing the particle size and volume fraction only slightly increases the central temperature of the fuel; however, when the volume fraction increases from 30% to 40%, the fuel particle stress increases by 11.6%.
Study on Irradiation Temperature Impact on Irradiation Embrittlement Behavior of RPV Material
Dong Yuanyuan, Luo Ying, Du Hua, Hu Tian, Wang Xiaotong
2024, 45(S2): 102-109. doi: 10.13832/j.jnpe.2024.S2.0102
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Reactor Pressure Vessel (RPV) suffers strong neutron radiation, which induces significant irradiation damage for RPV with the dose accumulation, and irradiation temperature is one of the key factors affecting its irradiation damage. However, at present, the research on the mechanism of irradiation temperature is insufficient. Aiming at the above problems, the existing prediction model analysis, in-situ ion simulated irradiation test and multi-scale simulation calculation are carried out. The temperature range of commonly used radiation embrittlement prediction formula is 275~310°C, which isn't applicable for low irradiation temperature condition. In-situ ion irradiation tests at different temperatures are conducted. The results show that the sizes of irradiated dislocation loops increase and the densities decrease with the increase of irradiation temperature. The results of multi-scale simulation show that the irradiation temperature has no obvious influence on the generation process of irradiation defects, but has obvious influence on the evolution and stabilization of irradiation defects. The lower the irradiation temperature, the more serious the embrittlement of materials is. The study reveals the mechanism and law of the influence of irradiation temperature on the radiation embrittlement behavior of RPV material.
Research on Rapid Calculation and Reconstruction Technology of Radiation Field Based on Bayesian Inference Method
Tang Songqian, Yu Hong, Lyu Huanwen, Wen Xingjian, Miao Jianxin, Chen Xin
2024, 45(S2): 110-114. doi: 10.13832/j.jnpe.2024.S2.0110
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In order to meet the requirements of rapid calculation of radiation field in nuclear facility decommissioning, nuclear power plant maintenance and other related field work, a rapid calculation code CSGPK based on point kernel integral method is developed in this study, which is suitable for complex geometry and source description. In order to improve the calculation accuracy of point kernel integral method, the radiation field is reconstructed by Bayesian inference method, and the developed code is verified by flat plate model and maze model. The verification results show that the calculation accuracy of CSGPK of point kernel method is equivalent to that of similar international calculation software. By using Bayesian inference method to reconstruct the radiation field, the calculation accuracy of the point kernel integration method can be greatly improved, and the application scope of CSGPK can be improved.
Study on Shielding and Weight Reduction Characteristics of Shadow Shield with Internally Tangent Structure in Small Space Reactor
Xie Lin, Liao Haoyang, Zhao Fulong, Wang Xianbo, Li Yufeng, Wang Xu, Tan Sichao, Tian Ruifeng
2024, 45(S2): 115-121. doi: 10.13832/j.jnpe.2024.S2.0115
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The mass of shadow shield of nuclear powered spacecraft is positively correlated with the power required by the mission. Aiming at the situation that the mass required for radiation shielding of high-power spacecraft is too large but the mass limit is too much, the shadow shield structure optimization analysis is carried out, and the structure optimization scheme and the comprehensive performance analysis method of the shield are put forward. Monte Carlo method is used to calculate the neutron transport in the inner and rear area of the shadow shield with internally tangent structure, and the effects of different internally tangent schemes on shielding effect, weight reduction effect and comprehensive performance of the combination of the two are analyzed. The results show that under the scheme of shielding effect weight factor of 0.4 and weight reduction effect weight factor of 0.6, the scheme using the tail internally tangent circle ring with the diameter less than 4 cm has improved the comprehensive shielding performance compared with the original scheme. Compared with B4C material, LiH material has better overall shielding performance and can achieve shielding weight reduction without greatly affecting shielding capability. It provides ideas and analysis methods for further research on shielding weight reduction methods for high-power nuclear spacecraft.
Research on Optimizing Method of Vibration Control Based on Improved Vibration Isolation Mass Structure
Zhang Rui, Liu Shuai, Wang Bihao, Bai Xiaoming, Liu Chengmin, Liu Zhenyu, Gong Zhaohu
2024, 45(S2): 122-126. doi: 10.13832/j.jnpe.2024.S2.0122
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In order to further reduce the vibration level of the pipeline system and reduce the cost of pipeline vibration control, combined with the characteristics of traditional vibration control methods, this paper proposes an optimization method of pipe vibration control based on vibration isolation mass. The dynamic vibration absorption principle is used to control the vibration transmission of the pipeline system. By analyzing the contribution of three-direction participating masses to the vibration suppression of the system, a limited vibration isolation mass structure is designed, and an optimization strategy for vibration control of pipeline system is established, which provides a means for global vibration control of pipeline system. The analysis results show that this method is reasonable and feasible, and can be used in the optimization research of pipeline vibration control, which has certain engineering significance.
Research on Modeling and Simulation of Nuclear Instrumentation System Based on System Code and Simulink
Luo Tingfang, Huang Ke, Liu Yi, Bao Chao, He Zhengxi, Huang Youjun, Xiao Kai, Liu Yaolong, Sun Qi, Gao Zhiyu
2024, 45(S2): 127-132. doi: 10.13832/j.jnpe.2024.S2.0127
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The measurement process of nuclear instrumentation system has statistical fluctuation characteristics, its measurement signal is constantly changing. In the current instrument and control system simulation analysis, the nuclear instrumentation system model has no statistical fluctuation characteristics, so it is difficult to effectively assess the coherence and the signal fluctuation level of nuclear instrumentation system designs. Therefore, this paper studies the principle and composition of nuclear instrumentation system, designs the simulation model of the basic components of nuclear instrumentation system, construts a elaborate model of the nuclear instrumentation system. Through the simulation analysis of the reactor start-up process, it is shown that the signal generated by the simulation of the nuclear instrumentation system is in good agreement with the reactor start-up process, the three ranges are connected to each other by more than one order of magnitude, and the whole cycle signal is stable. The modeling method of nuclear instrumentation system proposed in this paper is able to simulate the statistical fluctuation charateristics of nuclear instrumentation system, and can be used to reflect the characteristics of the nuclear instrumentation system more comprehensively in the simulation of the instrumentation and control system, and obtain more realistic simulation results.
Design and Application of a Collaborative Design Platform for Reactor Numerical Calculation
He Tengjiao, Han Fei, Li Songwei, Zhang Shuang, Wu Bin, Cheng Xianyan, Yuan Peng
2024, 45(S2): 133-136. doi: 10.13832/j.jnpe.2024.S2.0133
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To effectively enhance the efficiency and quality of collaborative design and execution of multi-disciplinary numerical calculation of reactor physics, thermal engineering, shielding and fuel analysis, a high-scalability and high-stability collaborative design platform for multi-disciplinary numerical calculation of reactors has been established. On the basis of cluster management and service-oriented encapsulation of complex heterogeneous high-performance computing resources, such as supercomputers and workstations, this platform achieves full lifecycle management of calculation codes in multiple disciplines, from deployment, workflow design, debugging and operation tracking to the outputs. It provides a unified, standardized, customized, on-demand reusable workflow design, job scheduling and data management mechanism, effectively supporting multi-projects and multi-disciplinary numerical calculations of nuclear reactors.
Architecture Design and Implementation of the Nuclear Reactor Big Data System (Ruilong System)
Zeng Hui, Yang Hui, Liu Chengmin, Wang Yuanmei, Zhang Siyuan, Yang Kunlin, Lin Yuanfeng, He Haotian
2024, 45(S2): 137-143. doi: 10.13832/j.jnpe.2024.S2.0137
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This paper introduces the data architecture design of the nuclear reactor big data system. By identifying business objects, a flexible and stable data business model has been constructed, forming a complete, standardized and accurate data system that logically integrates multi-source and heterogeneous data. In cases where label information is missing, a high-performance label annotation tool has been developed using semi-supervised learning methods and applied to unstructured data and time-series data. For different types of data missing, methods based on mechanisms, statistics, and machine learning are used to repair and filter anomalies in the data. Through critical domain analysis and architectural design, a software view characterized by clear layers, strong scalability and high flexibility has been established. This effectively avoids design iterations and ensures development quality. To further enhance the maturity of nuclear reactor big data technology, the next phase will focus on improving data governance, data value mining and intelligent operation and maintenance capabilities, ensuring the long-term safe and reliable operation of nuclear reactor facilities.
A Nuclear Reactor Accident Diagnosis Technology Integrating Expert Knowledge and Machine Learning Algorithms
Huang Tao, Zhu Dahuan, Zeng Wei, Fang Weiyang, Xiong Qingwen, Zhang Zhuo, Huang Qingyu
2024, 45(S2): 144-149. doi: 10.13832/j.jnpe.2024.S2.0144
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The nuclear reactor accident diagnosis is the most important step in the accident handling process, and the diagnosis result directly determines the accident handling strategy. In this paper, a nuclear reactor accident diagnosis method is proposed, which combines expert knowledge and machine learning algorithms. Based on the existing mature expert knowledge, this method introduces machine learning diagnosis algorithms to realize the superposition of advantages and the complementarity of disadvantages of the two methods. In terms of expert knowledge diagnosis, the symptom-oriented accident diagnosis method is adopted to form a symptom-oriented expert knowledge base and an accident diagnosis function module based on expert knowledge; In terms of accident diagnosis based on machine learning algorithm, Extreme Gradient Boosting (XGBoost), Linear Support Vector Machines (SVM), Deep Feedforward Networks (DFN), and Long Short-Term Memory (LSTM) are used to establish the accident diagnosis model, and the voting mechanism algorithm is used to fuse all kinds of algorithms to form the machine learning intelligent diagnosis module. On this basis, this paper puts forward a diagnosis model based on expert knowledge, supplemented by machine learning intelligent diagnosis, and the verification is carried out using the Steam Generator Tube Rupture (SGTR) accident of HPR1000. The results prove the rationality of the method.
Research on Fuel Rod Damage Diagnosis Method Based on Big Data
Jing Futing, Lyu Huanwen, Tang Songqian, Wei Jianglin, Li Lan, Xia Mingming
2024, 45(S2): 150-155. doi: 10.13832/j.jnpe.2024.S2.0150
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Fuel rod damage diagnosis (FRDD) in nuclear power plants is a critical issue of concern for nuclear power plant operators and nuclear safety regulators. The application of big data and the nearest neighbor algorithm to fuel rod damage diagnosis has led to the development of a PWR (Pressurized Water Reactor) fuel rod damage diagnosis software. The software has been validated using operational cases of fuel rod damage in nuclear power plants and theoretical examples. The validation results are as follows: ① In terms of category analysis of fuel rod breach size, 80% of the analysis results are consistent with the theoretical examples; ② In terms of the analysis of the damaged fuel rod number, the maximum deviation from the theoretical examples is one rod. The FRDD methodology for PWRs, based on big data and the nearest neighbor algorithm, provides diagnostic results that are closer to the actual damage scenario. This allows for the timely detection of fuel rod damage and changes in the damage state, providing a reliable basis for operational decision-making and radiation protection after fuel rod damage. This approach can enhance the economic efficiency of nuclear power plant operations while ensuring safety.
Research on Sensor Fault Diagnosis of Nuclear Power Plant Based on Improved CWT-CNN
Deng Zhiguang, Li Zhengxi, He Liang, Wu Qian, Zhu Jialiang, Zhu Biwei, Xu Tao, Wang Hailin
2024, 45(S2): 156-162. doi: 10.13832/j.jnpe.2024.S2.0156
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The consequences of sensor faults in nuclear power plants are serious, and the inherent complexity of primary circuit system and equipment in nuclear power plants brings difficulties to sensor fault diagnosis based on accurate mathematical models. In this paper, an intelligent sensor fault diagnosis method for nuclear power plant based on deep learning algorithm and time-frequency analysis is proposed, which transforms the signal recognition problem into image recognition problem. Firstly, the continuous wavelet transform (CWT) is used to process the time series signals of seven common health states of typical sensors in nuclear power plants to generate a time-frequency diagram that captures the characteristics of fault signals. Then, the convolutional neural network (CNN) model improved by channel attention mechanism (CA) is trained with pre-processed and labeled data sets, and the subtle image features of the time-frequency diagram are extracted. Based on these features, sensor faults are identified and isolated. This method does not need to model and design thresholds, and it has strong robustness and an accuracy rate of more than 97%. By comparing the diagnostic effects of typical deep learning networks such as long short-term memory network (LSTM) and one-dimensional convolutional neural network (1D-CNN), the effectiveness and superiority of the improved CWT-CNN are verified.
Research on Vibration Signal Classification Method of the Key Equipment of Reactor Based on Open-set Recognition
Pang Tianfeng, Li Shujian, Yang Taibo, Luo Neng
2024, 45(S2): 163-167. doi: 10.13832/j.jnpe.2024.S2.0163
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In the actual engineering scenario of health monitoring of key equipment in reactors, the equipment status continues to deteriorate over time, and the types of monitoring data gradually increase. Traditional data-driven algorithms may experience accuracy degradation or even failure. In order to overcome the above problems, this paper proposes an open set signal classification method. Firstly, a variational coding classifier network is used to classify known classes (KCs) and learn the distribution of feature space to extract hidden features; Then the hidden features are fitted to a Weibull distribution, and whether the sample belongs to unknown classes (UCs) is determined based on Extreme Value Theory (EVT); Finally, a simulated open set experiment is conducted using a multi-class labeled vibration signal dataset collected during the actual operation of the reactor. The experimental results show that by selecting appropriate discrimination thresholds, effective recognition of KCs and UCs can be achieved. The method proposed in this article provides a feasible solution for data classification scenarios where equipment gradually transitions from normal state to unknown faults in practical engineering scenarios.
Column of Nuclear Engineering Mechanics
Numerical Simulation of RPV with Complex Mechanical Behaviors under Core-melting Accident
Li Hui, Zhang Yixiong, Bai Xiaoming, Shao Xuejiao, Fu Xiaolong, Yu Mingda
2024, 45(S2): 168-173. doi: 10.13832/j.jnpe.2024.S2.0168
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After the severe loss of coolant accident (LOCA), the lower head of reactor pressure vessel (RPV) will be ablated by the core melt. Therefore, it is of great significance to simulate and analyze the ablation and complex mechanical behavior of the pressure vessel wall under the condition of core melt for the design, accident prevention and mitigation of RPV. This paper firstly presents the peridynamic formulation of coupling thermo-mechanical problem with crack propagation. Then, based on the peridyanmic framework, a simple and efficient moving boundary model is proposed. The abated status of the material points is directly characterized by introducing a scalar field, which makes it unnecessary to constantly update the calculation domain in the calculation process, thus improving the calculation efficiency. Finally, the present method is used to simulate the dynamic ablation of the core melt on the pressure vessel wall and the crack propagation of the pressure vessel under the action of internal pressure. The calculation results show that there are both elastic deformation and plastic deformation in RPV under the core melt accident, and there are also complex mechanical behaviors such as damage and fracture.
Experimental Study on Single-phase Flow-induced Vibration Mechanism in Helical Coiled Tubes
Wang Ningyuan, Chen Deqi, Liu Hanzhou, Bu Shanshan
2024, 45(S2): 174-179. doi: 10.13832/j.jnpe.2024.S2.0174
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To investigate the evolution of single-phase flow-induced vibration response in helical coiled tubes and elucidate the vibration mechanisms in both in-plane and out-of-plane directions, this study conducts experiments of single-phase flow-induced vibration in helical coiled tubes using laser Doppler measurement technology. By performing flow-induced vibration experiments at different inlet flow velocities and integrating modal analysis results, we examined the influence of flow velocity on coil vibration response and explored the characteristics of vibration response in both in-plane and out-of-plane directions. The experimental results indicate that the in-plane vibration characteristics of the helical coil are influenced by the static force and mass, while the out-of-plane vibration characteristics are related to the system mass. The RMS of out-of-plane vibration displacement decreases with the increase of height. By analyzing the flow state within the tube, the study reveals the impact mechanism of secondary flow on the vibration response of the helical coil. These findings can provide reference for the study of flow-induced vibration in helical coils, and can provide support for the subsequent study of two-phase flow-induced vibration in the tube.
Study on Impact Environment of Submarine under Combined Effect of Shock Wave and Bubble
Yang Bo, Li Yilei, He Jian, Yao Di, Fan Jinpeng
2024, 45(S2): 180-188. doi: 10.13832/j.jnpe.2024.S2.0180
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In order to investigate the impact environment of submarine and its equipment under underwater explosion and increase the vitality of the submarine, the dynamic response of the model under underwater explosion load is calculated by acoustic-structural coupling method. The propagation of shock wave in water is discussed, and the response spectrum of cabin platform under the combined action of shock wave and bubble pulsation is calculated. The platform response of each cabin under different diving depth, explosive impact factor and explosive attack angle is compared, and the influence of various parameters on the impact environment of equipment is analyzed by combining the acceleration curve and response spectrum curve of cabin plate. The simulation results show that the bubble load will aggravate the platform response, especially in the low frequency band. Therefore, through the study of the impact environment of submarine equipment under different conditions, this paper puts forward reasonable suggestions for the installation position and vitality improvement of the equipment.
Column of Digitalization in Nuclear Power
Research on Automatic Control of SGTR Post-accident Safety Injection Process Based on Intelligent Algorithm
Du Ming, Bian Shujie, Niu Yuguang, Yuan Jinxiao, Chen Rigang
2024, 45(S2): 189-196. doi: 10.13832/j.jnpe.2024.S2.0189
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Following a steam generator tube rupture (SGTR) accident at a nuclear power plant, the manual adjustment control method is commonly employed, and the adjustment of the safety injection process is one of the difficulties faced by operators. For this process, the main adjustment mechanism and characteristics of the object are analyzed firstly in this paper. Combined with the experience of the operator, the automatic control structure of the safety injection control process is summarized, and the corresponding parameter setting method is designed. The intelligent elements such as adaptive PID, intelligent feedforward and reinforcement learning are integrated, so that the intelligent control strategy of the safety injection control process is innovatively proposed. In order to complete the strategy validation on the M310 full-range simulator developed by a company, this paper first develops the intelligent computing engine, then completed the control strategy configuration on the intelligent computing engine, then completes the control strategy configuration on the intelligent computing engine, and completes the real-time communication of the control signals through the MySQL. Through the simulation and test verification of different breach accidents in the simulator, the intelligent control strategies proposed in this paper are all able to realize the automatic adjustment of the injection control process, the actual cooling rate deviation is 5.89% of the cooling rate set value, the matching effect of the subcooling degree and the regulator level adjustment is good, and the performance is higher than the average of the manipulator's manual execution of the task.
Research on Intelligent Control Method of Operating Temperature of Reactor Thermal System Based on Deep Reinforcement Learning
Liu Yongchao, Tan Sichao, Li Tong, Cheng Jiahao, Wang Bo, Gao Puzhen, Tian Ruifeng
2024, 45(S2): 197-205. doi: 10.13832/j.jnpe.2024.S2.0197
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Traditional proportional-integral-differential (PID) control method is difficult to achieve good and stable control effect. In this paper, an intelligent control method of operating temperature of reactor thermal system based on deep reinforcement learning is proposed. The steps are as follows: RELAP5 model of reactor thermal system is built and extended interactively, so that it can support deep reinforcement learning technology. Secondly, based on the Soft Actor-Critic (SAC) algorithm and coupled with the multivariable Long Short-Term Memory (LSTM) neural network, the time series characteristics of the control history information are effectively extracted. Then, the control model driven by optimization goal can collect data samples by itself, and complete the optimization of control strategy through self-learning mechanism. According to the multivariable state characteristics and time series characteristics, the end-to-end control of operating temperature is realized. Compared with the simulation experiment of PID controller, the proposed method has excellent load tracking ability and disturbance suppression ability, and has good environmental adaptability and control robustness.
Research on the Conversion Method of Critical Calculation Model for Fuel Storage Rack
Ma Jiancong, Wang Yiqi, Liu Shichang, Yu Miao, Shao Zeng
2024, 45(S2): 206-213. doi: 10.13832/j.jnpe.2024.S2.0206
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To improve the modeling efficiency and accuracy of RMC (Reactor MonteCarlo Code) in critical calculations for spent fuel storage racks, and to promote the application of autonomous Monte Carlo code in critical analysis of spent fuel storage racks, a conversion code for the input files of the Monte Carlo code MONK and RMC has been developed. For the description cards of geometry, hole, material, control and source of MONK code, the corresponding reading code was developed based on python language. The code could read the card calculation input file of MONK 3D Monte Carlo code and generate the input file of RMC 3D Monte Carlo code. After comparison, the generated RMC input file model is consistent with MONK input file model in terms of geometry, material, source terms and computational control conditions. The effective multiplication factors before and after conversion are calculated respectively, and the effective multiplication factors are the same within the allowable error range. The results show that the code meets the requirements for the conversion of critical calculation code input cards for spent fuel storage racks, which can improve the efficiency and accuracy of modeling spent fuel storage racks, and also verifies the correctness of applying RMC to critical calculations for spent fuel storage racks.
Comparative Study of Neural Network-based Source Term Inversion Methods for Nuclear Accidents
Peng Dingping, Li Zhonghao, Cao Bo, Miao Xuewei, You Qingyue
2024, 45(S2): 214-222. doi: 10.13832/j.jnpe.2024.S2.0214
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The nuclear accident source term is an important basis for the evaluation of the consequences of nuclear accidents. The accurate and fast acquisition of the source term is of great significance for the development of nuclear accident emergency response decision-making. This study utilizes three typical neural network algorithms, namely radial basis neural network, back propagation neural network and deep feed-forward neural network, to compare the inverse estimation of nuclear accident source terms. For the inversion of single nuclide and multi-nuclide source term, taking the release rate of radionuclides as the inversion target, the simulation experiment is conducted to obtain the monitoring data by using the radioactive atmospheric dispersion simulation code RADC developed by the group. The results of error analysis and inversion time comparison show that the nuclide release rate inversion model based on deep feedforward neural network exhibits excellent performance. For the release rate inversion of single nuclide, the relative error of the inversion is within 1.7%~3.0% and the average absolute percentage error is 2.2% under the ten-fold cross-validation. For the multi-nuclide case, the average absolute percentage error of the inversion of deep feedforward neural network is 8.05%. Its inversion estimation accuracy and stability are better than that of the BP neural network and radial basis neural network.
Study on Risk-informed Safety Evaluation and Optimization under Adaptive Sampling in Nuclear Power Plant
Li Linfeng, Xu Anqi, Dong Xiaomeng, Zhang Zhen, Yang Ming, Wen Ting, Liu Yong
2024, 45(S2): 223-230. doi: 10.13832/j.jnpe.2024.S2.0223
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In order to meet the dual requirements of safety and economy of advanced reactors, the Deterministic Safety Analysis and Probabilistic Safety Analysis are coupled in this study. Based on the Risk-informed System Analysis (RISA), a RISA optimization method under adaptive sampling strategy is proposed to realistically evaluate the safety margin of nuclear power plants. It focuses on solving the problems of massive sampling times and low calculation efficiency under high precision requirements. Support Vector Machine (SVM) and K-Nearest Neighbor (KNN) Algorithm are used to train the surrogate model, which replaces a large number of Best Estimation and Uncertainty Analysis code simulations. The adaptive sampling strategy is used to identify the limit surface and reduce the sampling range and times. Taking the Small Break Loss of Coolant Accident (SBLOCA) as an example, the test results show that, compared with the random sampling RISA results, the peak temperature of fuel cladding predicted by the surrogate model is close, and the calculation time is reduced by more than 50%. Therefore, this method can support the practical engineering application of RISA in the future, and provide realistic and accurate decision support for the risk-informed design, operation and management.
Excellent Papers from the Academic Annual Conference of Chinese Nuclear Society
Preparation and Performance Evaluation of Fuel Elements for Ultra-high Temperature Gas-cooled Reactors
Liu Malin, Cheng Xinyu, Liu Bing, Wang Taowei, Liu Zebing, Yang Xu, Liu Rongzheng, Shao Youlin
2024, 45(S2): 231-237. doi: 10.13832/j.jnpe.2024.S2.0231
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According to the research and development requirements of future ultra-high temperature gas cooled reactors, the design of relevant fuel elements, preparation of new coating layers, high-temperature testing and evaluation of TRISO particles before and after irradiation have been carried out. Specifically, the preparation methods of fluidized bed chemical vapor deposition for SiC coating layers, ZrC coating layers, NbC coating layers, and carbide composite coating layers are studied. And the SiC coating layer was used in the ultra-high temperature testing, including pre irradiation (up to 2500℃) and post irradiation (up to 1770℃). The research results indicate that single-phase SiC coating, ZrC coating, NbC coating, and carbide composite coating can be prepared using fluidized bed chemical vapor deposition method with liquid methyltrichlorosilane, solid ZrCl4, and NbCl5 as precursors. The gas phase carrier transportation and gas powder transportation methods are developed. The large-scale preparation of SiC coating has been successfully obtained. It was found that the SiC coating layer of TRISO particles could withstand the high temperature of 2200℃ for a short time in the high-temperature test before irradiation. At temperatures above 2100℃, partial phase transformation, grain growth, and micro decomposition could be found, but the overall coating layer remained relatively intact. The high temperature test at 1770℃ after irradiation showed that high temperatures would accelerate the diffusion of some fission elements such as Cs in the dense pyrolysis carbon layer of TRISO particles. And no damage was found in the SiC coating layer, indicating that the ability to block fission products continued to be maintained. Besides, molecular simulations were used to simulate various microstructures and mixed crystal SiC coating materials after high-temperature testing and irradiation. The above research results provide reference for the research and development and performance evaluation of fuel elements in ultra-high temperature gas-cooled reactors in China, and are of great significance to the future development of ultra-high temperature gas-cooled reactors.
Optimization Design of Grooved Sodium Heat Pipe Based on NSGA Algorithm
Zhang Jiansong, Mei Huaping, Li Taosheng
2024, 45(S2): 238-244. doi: 10.13832/j.jnpe.2024.S2.0238
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Alkali metal sodium heat pipe has broad application prospects in hypersonic aircraft, micro nuclear reactor and utilization of solar energy because of its high operating temperature and great heat transfer performance. In this paper, according to the requirements of a heat pipe cooling space nuclear reactor project, a trapezoidal grooved sodium heat pipe is designed. Based on the genetic algorithm of NSGA-Ⅲ, the effects of different operating temperature, steam chamber diameter and groove number on the heat transfer limit of sodium heat pipe are studied, and the optimal solution of heat transfer capacity of sodium heat pipe under the service condition of heat pipe cooling space nuclear reactor is proposed. The study found that the capillary limit initially increased with temperature, then decreases. Within the working temperature range, both the entrainment limit and the condensation limit were likely to become the heat transfer limit of the designed heat pipe. At an operating temperature of 1256 K, with a vapor chamber diameter of 12.8 mm and 20 grooves, the heat transfer limit of the designed heat pipe was 4.56 kW.The designed trapezoidal grooved sodium heat pipe meets the requirements of Mach number and effective capillary radius.
Molecular Dynamics Simulation of Nanomechanics Behavior of SiC Layer of TRISO Particle
Yan Zefan, Tian Yu, Liu Malin, Liu Rongzheng, Liu Bing, Shao Youlin, Tang Yaping
2024, 45(S2): 245-253. doi: 10.13832/j.jnpe.2024.S2.0245
Abstract(29) HTML (13) PDF(10)
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The grain of the SiC layer of tristructural-isotropic (TRISO) particles will undergo phase transformation, fracture and abnormal growth after irradiation and high temperature test. The mechanical behavior of these SiC layers is very important for the safety study of TRISO particles. In this paper, molecular dynamics simulation is used to study the nanomechanical behavior and properties of SiC layer. Four typical SiC layer structures are constructed according to the experimental phenomena: 3C-SiC before service, 3C-SiC after irradiation test, 6H-SiC after high temperature test, and 6H/3C-SiC after high temperature & irradiation test. The nanomechanical behavior and mechanical properties of SiC layer are analyzed by load-depth curve, dislocation evolution, stress & strain, and atomic diffusion. The results show that the SiC layer after service has less interaction between dislocations during the nanoindentation loading process, which reduces the plastic deformation, resulting a decrease in Young's modulus. For the SiC layer after irradiation test and high temperature & irradiation test, the concentration degree of stress and strain directly below the indenter decreases, and the transverse distribution of stress, strain, and atomic diffusion increases, resulting in a decrease in hardness. For the SiC layer after high temperature test, the concentration degree of stress and strain directly below the indenter increases, and the vertical distribution of stress, strain, and atomic diffusion increases, resulting in an increase in hardness. The research results give a quantitative explanation for the mechanical behavior and properties of various types of SiC layers, which is helpful to understand the relationship among the microstructure, mechanical behavior and mechanical properties of SiC layers.
Analysis of Laser-TIG Hybrid Welding Performance on 15-15Ti Stainless Steel Lock Bottom Structure
Guan Huai, Xu Xiaodong, Zhang Xuewei, Zou Benhui
2024, 45(S2): 254-260. doi: 10.13832/j.jnpe.2024.S2.0254
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To investigate a kind of welding process of 15-15Ti stainless steel material for Sodium-cooled fast reactors, a method that can accommodate to the bottom lock structure while maintaining good welding performance after multiple welding cycles is explored. In this paper, laser-TIG hybrid welding process is adopted to a bottom lock structure made of domestically produced 15-15Ti stainless steel. The microstructural characteristics, microhardness, intergranular corrosion resistance, and mechanical properties of the weld seam were tested using metallurgical microscope, scanning electron microscope, hardness tester, and tensile testing machine. Comparisons were made between the weld morphology and properties after multiple welding cycles. The research shows that laser-TIG hybrid welding can realize the welding of lock bottom structure and obtain superior welding microstructure and mechanical properties. After multiple welding cycles, the weld seam solidification mode, microstructure, and mechanical properties remain unchanged, and the microstructure and mechanical properties of the weld have good consistency. This welding method of 15-15Ti material has broad engineering application potential and scenarios.
Study on the Coupled Calculation Method of Discrete-Ordinates and Point Kernel Integration
Yang Xuhui, Liu Fan, Wang Xiang, Cai Li, Marcus Seidl
2024, 45(S2): 261-267. doi: 10.13832/j.jnpe.2024.S2.0261
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To address the obvious shortcomings of most existing point kernel integration codes in shielding calculations, i.e., they produce errors of order of magnitude in the calculation process of large source and multi-layer shielding combinations. In this study, discrete-ordinates method is introduced in the point kernel integration calculation, and the characteristics of the two codes TORT and QADS are investigated. An interface code is written to realize the coupling of the two codes. The problems of poor accuracy of source discretization and lack of oblique angle decomposition technology in traditional point kernel integration codes are solved, and a simple geometric model is designed to study the feasibility of this method. The results show that the coupled discrete-ordinates and point kernel integral method greatly improves the computational accuracy in the calculation of deep penetration problems and can provide better results in the practice of radiation shielding calculation compared with the involved methods.
Implementation and Application of a Coupling Method between Molecular Dynamics and Kinetic Monte Carlo for the Evolution of Helium Bubbles in Nuclear Structural Steel
Li Liuliu, Hu Xuefei, Peng Lei
2024, 45(S2): 268-273. doi: 10.13832/j.jnpe.2024.S2.0268
Abstract(33) HTML (18) PDF(13)
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The research and development of structural materials for fourth-generation advanced reactors, such as accelerator-driven subcritical reactors and ultrahigh-temperature gas-cooled reactors, urgently requires the use of numerical simulation methods to shorten the research and development cycle and enhance the efficiency of research and development. Currently, various existing numerical simulation methods are only applicable to specific time and space scales, while the high-temperature irradiation effect of nuclear structural materials for advanced reactors involves multiple time and space scales from the irradiated microstructure evolution to the macroscopic mechanical properties. It is of great significance to develop coupling methods and codes between simulation methods at various scales, and to construct a multi-scale simulation and computation platform for the rapid research and development and service performance prediction of structural materials for advanced reactors. Based on the mutual conversion between atomic configuration and defect configuration, this paper proposes and implements a simulation method that temporally couples the microscale simulation method (molecular dynamics) with the mesoscale simulation method dynamics (Monte Carlo). By this method, the irradiation cascade process can be simulated by molecular dynamics, while the further evolution of defects is simulated by kinetic Monte Carlo, thus simulating the microstructural evolution of nuclear structural materials under the accumulation of irradiation dose. The reliability of this coupling method is demonstrated by using it to simulate the evolution of helium bubbles in the nuclear structural steel matrix material α-Fe under neutron irradiation and comparing it with experimental data.
Research on Impurity Regulation and Purification Technology of Liquid Lead-based Metal Coolant
Tang Hairong, Li Ying, Lou Ruifan, Yue Nina, Wang Suhao, Wang Sheng
2024, 45(S2): 274-278. doi: 10.13832/j.jnpe.2024.S2.0274
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Liquid lead-based metal is the mainstream candidate working fluid for the international fourth generation fast reactors and accelerator-driven subcritical systems. However, the liquid lead-based coolant has problems during the long-term operation in the non-isothermal system, such as continuous generation of impurities, large accumulation, difficulty to avoid and to deal with, which will lead to deposition and scaling, deterioration of heat transfer, and even blockage of the flow, causing significant safety risks. Therefore, the purification and regulation of impurities in lead-based liquid metal coolant is a key technology to be broken through in the design and development of lead-cooled fast reactor. This paper mainly introduces the source and occurrence form of impurities in lead-based coolant, and the research status of purification technology. The advantages and limitations of inhibition generation method, filtration capture method and reduction method are summarized. Finally, the selection and challenges of impurity purification strategies for lead-cooled reactor systems with different specifications and forms are discussed.