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23rd Pacific Basin Nuclear Conference, PBNC 2022

作者： Shuai, Liu Xuan, Huang Zhipeng, Feng Zhongxiu, Zeng Huanhuan, Qi Xiaozhou, Jiang Harbin Engineering University Harbin China Science and Technology on Reactor System Design Technology Laboratory Sichuan Chengdu China

ISBN: (纸本)9789811987793

The reactor system has a large scale of components. And the system dynamic analysis model for loss of coolant accidents (LOCA) contains many non-linear factors. The transient calculation analysis takes a long time and the convergence is difficult to guarantee. In this paper, two different analysis models (complete three-loop model and single-loop model) of the first loop system of reactor are established for comparative analysis. The analysis shows that the calculation error of the main position load in the two different model is within 10%. The decoupling influence of the wave tube and the reactor inlet and outlet on the system dynamic characteristics is within an acceptable range. When the single loop model is used, it is necessary to decouple the reactor pressure vessel and pressurizer. And the simplified boundary is adapted to simulate its effect on the whole loop. The single-loop dynamic analysis model based on the decoupling criterion can greatly reduce the calculation scale and meet the calculation accuracy, which can be used to calculate the LOCA dynamic characteristics of the reactor system quickly. The model can also be used for seismic analysis of reactor system. And using this model for reactor system dynamic analysis can greatly improve the calculation efficiency. Please copy you summary here. © 2023, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Loss of coolant accidents

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Effect of He/dpa ratio on bubble characteristics in Fe9Cr1.5W0.4Si F/M steel during irradiation and annealing

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Journal of Materials science & technology 2024年第30期197卷 17-24页

作者： Dewang Cui Kefei Pei Ziqi Cao Yipeng Li Yifan Ding Yuanming Li Shichao Liu Guang Ran College of Energy Xiamen UniversityXiamen361102China Science and Technology on Reactor System Design Technology Laboratory Chengdu610213China College of Chemistry and Chemical Engineering Xiamen UniversityXiamen361005China Fujian Provincial Nuclear Energy Engineering Technology Research Center Xiamen361102China Nuclear Power Institute of China Chengdu610213China

The He production rate(i.e.,He/dpa)in nuclear reactors strongly affects the degradation of material *** is an important but not yet fully understood ***,the effect of He/dpa on bubble char-acteristics in Fe9Cr1.5W0.4Si ferrite-martensitic(F/M)steel was in situ studied during 400 keV Fe^(+)and 30 keV He^(+)dual-beam irradiation at 723 K with three ratios of 100,500,and 2500 appm He/dpa and subsequent stepwise annealing using transmission electron microscopy(TEM).He/dpa strongly affected the bubble *** irradiation,the higher the He/dpa,the smaller the size of irradiated bubbles,but the higher their ***,He/dpa didn’t affect the final saturation size of irradiated bubbles for all three cases,which was-2.2 *** annealing,high He/dpa caused large,immobile,dense polyhedral bubbles with a wider bubble size distribution,while low He/dpa caused small,low-mobility,and relatively low-density spherical *** was found that the higher the He/dpa ratio,the greater the swelling during irradiation and annealing,and annealing further enhanced the ***,the tunnel structure was first found in body-centered cubic(BCC)F/M steel during in-situ *** current work provides valuable and potential insights for further understanding the He/dpa effects in materials serving in different nuclear reactors.

关键词： He effect Ferritic-martensitic steel He bubble Ion irradiation Annealing

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Study on Multi-Motor Synchronization Control for Nuclear Island Main Equipment Transporter Based on Improved Virtual Line-Shafting Structure 11th

Study on Multi-Motor Synchronization Control for Nuclear Isl...

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11th Frontier Academic Forum of Electrical Engineering, FAFEE 2024

作者： Sheng, Beining Song, Beibei An, Yanbo Du, Hua Cui, Shumei School of Electrical Engineering and Automation Harbin Institute of Technology Harbin150001 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

ISBN: (纸本)9789819788231

The existing drive method of the main equipment transporter of the nuclear island makes the transporter have frequent jamming problems when the transporter going straight and turning, which affects the safety during the transporter of the main equipment. Therefore, it is necessary to design an on-board multi-motor synchronous system with differential synchronization function. Although the traditional virtual line-shafting structure has the ability of same-speed synchronization and differential synchronization, there is a large synchronization error. For this reason, this paper proposes a multi-motor synchronous control method based on the improved virtual line-shafting structure, which can effectively solve the inherent synchronous delay problem of the virtual line-shafting. First, analyze and research the combination method of virtual line-shafting structure and deviation coupling structure, and designs the electronic differential to be able to adapt to the requirements of same-speed and differential synchronization. Then the influence of virtual line-shafting parameters on the system is analyzed using the root locus method, and a simplified parameter model of the virtual line-shafting is established, on the basis of which an improved virtual line-shafting control method with low synchronous delay is proposed. The results show that this method can significantly reduce the synchronization error compared with the traditional virtual line-shafting. © Beijing Paike Culture Commu. Co., Ltd. 2025.

关键词： Root loci

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REVIEW OF RISK-INFORMED SAFETY design CONCEPT, METHOD AND APPLICATION 30

REVIEW OF RISK-INFORMED SAFETY DESIGN CONCEPT, METHOD AND AP...

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30th International Conference on Nuclear Engineering, ICONE 2023

作者： Gong, Lanxin Zhong, Mingjun Peng, Changhong School of Nuclear Science and Technology University of Science and Technology of China Hefei China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9784888982566

Since the NRC issued the PSA policy in 1995, PSA containing risk information has been widely used in supervision and design. Subsequently, the concept of "risk-informed" has gained more and more recognition, and its application has expanded from the initial equipment maintenance to the current design. By analyzing the regulations, standards and other guidance documents of various countries and institutions on safety design, we reviewed the history of risk-informed safety design concepts, methods and their applications, especially the progress of LMP projects in the United States in recent years. We put forward prospects for the development and application of risk-informed concepts and methods in new reactors in the future. © 2023 by JSME.

关键词： design

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FEASIBILITY AND SAFETY ANALYSIS OF SOLID-STATE reactor FOR LUNAR SURFACE EXPLORATION 31

FEASIBILITY AND SAFETY ANALYSIS OF SOLID-STATE REACTOR FOR L...

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2024 31st International Conference on Nuclear Engineering, ICONE 2024

作者： Yugao, Ma Jiahao, Lu Yue, Liu Suyi, Zhang Muhao, Zhang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China College of Nuclear Technology and Automation Engineering Chengdu University of Technology Chengdu610059 China

ISBN: (纸本)9780791888247

The Fission Surface Power system is a feasible solution for future lunar surface exploration energy supply, with the solid-state reactor being considered a primary candidate. This paper presents a modeling and analysis of a solid-state reactor based on publicly available information. Using the reactor Monte Carlo program RMC and the commercial finite element analysis software ANSYS, the reactor is modeled and analyzed. The power distribution of the active zone and reactivity feedback coefficients of the entire core are obtained. The result indicates a negative reactivity feedback coefficient, confirming this reactor design's inherent safety. Furthermore, a two-dimensional reactor model is constructed to obtain the temperature distribution during full-power operation, and non-passive residual heat removal analysis and radiative heat loss failure analysis are performed. At the failure condition of all heat pipes, the maximum core temperature is 742 K, which is below the reactor safety temperature limit, confirming the residual heat removal capability of the solid-state reactor in the loss of heat sink. Additionally, the self-developed transient reactor system analysis program is used for thermal safety analysis. The thermal parameters and heat transfer limits of high-temperature alkali metal heat pipes within the reactor are obtained, with the ratio of heat transfer limit to the heat load of 2.05 demonstrating sufficient safety margins of the heat pipe operation in the reactor. The results manifest that the core is satisfied with the existing criteria for anticipated operating occurrences of solid-state reactors, verifying the excellent safety performance of this reactor design. Based on the analysis and calculations in this study, it is concluded that this reactor design exhibits superior safety performance, which is attributed to the neutron physics design concept of the intermediate neutron spectrum and the thermal-hydraulic design concept of non-passive residual h

关键词： Lunar surface analysis

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Investigation on influence of gamma irradiation on piezoelectric and resonance characteristics of surface acoustic wave resonators 5

Investigation on influence of gamma irradiation on piezoelec...

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5th International Conference on Mechanical Engineering and Materials, ICMEM 2024

作者： Sun, Lei Peng, Bin Yang, Yuntao Sun, Zhaofeng Zhu, Jialiang Zhang, Wanli State Key Laboratory of Electronic Thin Films and Integrated Devices University of Electronic Science and Technology of China Chengdu China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (数字)9781510688971

ISBN: (纸本)9781510688964

In this work, an evaluation towards influence of high-dose gamma irradiation on surface acoustic wave (SAW) resonators, is reported. The single-port SAW resonators are fabricated using Lithium niobate (LiNbO3) and Catangasite (CTGS, Ca3TaGa3Si2O14) substrates, along respective cut directions of (0º, 38º, 0º) and (0º, 90º, 0º). Piezoelectric coefficient of substrates, and several main resonance parameters covering resonance frequency, amplitude of resonance peak in reflection coefficient, quality factor together with electromechanical coupling factor of manufactured SAW resonators are investigated, as suitable representatives of piezoelectric and resonance characteristics of SAW resonators. After exposure to gamma ray with a maximum dose of 250 kGy, substrates show insignificant variation not exceeding 6% in their piezoelectric coefficient. Meanwhile, resonance parameters of SAW resonators also alter slightly. Such excellent stability of these characteristics demonstrates credible irradiation hardness owned by both SAW resonators. Results in this study can provide supportive references for usage of LiNbO3 and CTGS-based SAW devices in harsh environments. © 2025 SPIE.

关键词： Gamma rays

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Fault Severity Classification based on Information Flow and Support Vector Machines 8

Fault Severity Classification based on Information Flow and ...

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8th International Conference on Energy system, Electricity, and Power, ESEP 2023

作者： Yang, Xiaohua Liu, Zhikun Fang, Weiyang Huang, Tao Liu, Jie Yang, Jinghua Ren, Changan Department of Compute Science University of South China Hengyang421001 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China

ISBN: (数字)9781510679801

ISBN: (纸本)9781510679795

Fault severity describes the level of impact caused by a fault. In the event of a malfunction in a nuclear power plant, a vast amount of data and alarms are generated, making it challenging for operators to identify and determine the state of the power plant in a short time. If the fault severity can be effectively identified during a malfunction, it would allow the operators to respond quickly and prevent further losses. Compared to other machine learning methods, the Support Vector Machine (SVM) has superior classification capability in scenarios with limited nuclear power samples. However, if the feature representation of the SVM training sample is inadequate, resulting in similar, indistinguishable training data, it will affect the SVM's classification of fault severity. To address this issue, using the Liang-Kleeman Information Flow (IF) can characterize information transmission from a causal perspective. By introducing IF to combine features of the training data, insufficient sample features can be replaced with the IF that describes information from a causal viewpoint, thereby enhancing the SVM's accuracy in fault severity classification. This paper takes the Main Steam Line Break accident in the primary loop of the nuclear reactor as an example and conducts simulation experiments on the PCTRAN simulation platform. The experimental results show that under 5 different fault degrees, the accuracy of fault degree evaluation of SVM model combined with IF is improved. This method can effectively improve the classification accuracy of different fault severity degrees in MSLB steam flow data with similar changes. © 2024 SPIE.

关键词： Steam piping systems

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A Transformer-based Time Series Prediction Model with Contrastive Learning for Nuclear Power system State Monitoring 15

A Transformer-based Time Series Prediction Model with Contra...

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15th IEEE Global Reliability and Prognostics and Health Management Conference, PHM-Beijing 2024

作者： Gou, Pengfei Zheng, Daiwei Tang, Lei Liu, Mengjuan School of Information and Software Engineering University of Electronic Science and Technology of China UESTC Chengdu China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China NPIC Chengdu China

ISBN: (纸本)9798350354010

Prognostics and Health Management (PHM) is currently the mainstream solution for the condition monitoring and maintenance of industrial systems. In PHM, the system condition is monitored by monitoring the operational status parameters of the system to diagnose equipment faults, predict the remaining useful life, etc. Utilizing sequential operational status parameters to complete prediction and classification tasks is usually modeled as a time series processing task. Therefore, a high-performance time series prediction model is crucial for PHM. This paper focuses on the use of time series prediction models to achieve the prediction of key state parameters of the nuclear power system. However, there are still some challenges when directly applying existing time series prediction models to industrial systems. First, obtaining operational data on the nuclear power system is relatively difficult, resulting in a small amount of training data, which causes many models to face the risk of insufficient training or overfitting. Second, most of the existing time series prediction models are composed of high- and low-frequency decoupling modules, while the operational data of nuclear power systems often have fewer high-frequency components, making it difficult for many existing SOTA models to achieve good results. In order to address these challenges, we have developed a Transformer-based time series prediction model with contrastive learning for monitoring the status of nuclear power systems. We evaluated the new model on three nuclear power simulation datasets and four power transformer datasets. The experimental results show that the new model's performance is superior to baseline models. We also proved the effectiveness of our improvement through ablation experiments. © 2024 IEEE.

关键词： Prediction models

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Analysis of contribution degree of vibration transmission of bolted structure 29

Analysis of contribution degree of vibration transmission of...

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2022 29th International Conference on Nuclear Engineering, ICONE 2022

作者： Liu, Shuai Li, Wanyou Huang, Xuan Xiong, Furui Jiang, Xiaozhou Deng, Dliwei Harbin Engineering University Harbin China Science and Technology on Reactor System Design Technology Laboratory Chengdu China Technology on Reactor System Design Technology Laboratory Chengdu China

ISBN: (纸本)9780791886359

In practical engineering, the vibration transmission between bolted structures supported by cantilever beams involves many factors, including the friction between plates and the bolt connection relationship. The vibration transmission effect is the embodiment of the coupling of various factors, so it is impossible to accurately and quantitatively evaluate the influence degree of various factors. In this paper, the influence factors of vibration between bolted plates in the form of cantilever beam are studied by numerical calculation method. The effects of friction coefficient, bolt preload, temperature and internal and external pressure on vibration characteristics are studied. Finally, the contribution of different influencing factors to vibration transmission is given. The preliminary research shows that the bolt connects the two plates into a whole structure, and its stiffness changes little with the friction coefficient, bolt preload, temperature and internal and external pressure. The natural frequency of each order increases with the increase of friction coefficient, bolt preload and internal and external pressure, decreases with the increase of temperature, and the overall trend changes little. At the same time, the bolt preload is verified by test, the difference of natural frequency under the same preload is compared and analyzed, the numerical calculation is compared with the test value, and the vibration contribution of preload is verified. © 2022 American Society of Mechanical Engineers (ASME). All rights reserved.

关键词： Cantilever beams

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Investigation on crack propagation of SiCf/SiC composite via PF-CZM method

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International Journal of Advanced Nuclear reactor design and technology 2024年第1期6卷 9-13页

作者： Qiu, Bowen Li, Zhuang Liang, Shuang Zhang, Cheng Wei, Chong Xiao, Zhong Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China School of Mechanical Civil Engineering and Architecture Northwestern Polytechnical University Xi'an710072 China Science and Technology on Thermo Structural Composite Materials Laboratory Northwestern Polytechnical University Xi'an710072 China

SiCf/SiC composite materials had become a research hotspot in the fields of aerospace materials and nuclear materials in recent years. This materials have high strength, radiation resistance, high temperature resistance and oxidation resistance. However, SiC is a typical brittle material prone to being damaged due to stress concentration in some cases, leading to cracks and even component failure. Therefore, this study employed an accurate and efficient numerical simulation method, phase field cohesive zone method (PFCZM), to investigate the microscopic damage behavior of SiCf/SiC composite materials. This study aimed to illustrate the influence of the thickness of pyrolytic carbon (PyC) interface layer on the failure behavior of SiCf/SiC composite materials. Firstly, this study conducted relevant verification of PFCZM. Secondly, numerical simulations of crack initiation, crack propagation, and final damage were performed by establishing a representative volume element (RVE). Observing the simulation results, it was noted that at low strain, when PyC was thicker, the SiCf/SiC composite materials exhibited slight damage to both the SiC matrix and PyC interface. Conversely, when PyC was thinner, the SiCf/SiC composite materials displayed only slight damage to PyC interface. As the PyC thickness increases, the damage to PyC interface decreases, while the damage to SiC becomes more prominent. Simultaneously, a competitive relationship in the damage behavior of SiCf/SiC composite materials can be found after the SiC matrix has cracked: When PyC is thinner, the damage primarily manifests as interlayer delamination between the SiC matrix and the SiC fiber. Conversely, when PyC is thicker, it manifests in PyC failure. Finally, this study concluded that, in this investigation, a PyC interface layer with a thickness of 0.1 μm exhibited the most effective protective effect on SiCf/SiC composite material. © 2024 Xi'an Jiaotong University

关键词： Crack propagation

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