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Anti-interference Study of Wireless Sensor Networks in Nuclear Power Plants 1

7th International Symposium on Digital Instrumentation and Control technology for Nuclear Power Plant, SICPNPP 2023

作者： Deng, Zhi-guang Wu, Qian Lv, Xin Xiang, Mei-qiong Xu, Tao Qin, Yue Sun, Yong-sheng Peng, Ye-shun Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

ISBN: (数字)9789819934553

ISBN: (纸本)9789819934546

The compatibility and anti-interference capability of wireless sensor networks within nuclear power plants in complex electromagnetic environments are the most essential elements to ensure reliable operation and effective combat of equipment within nuclear power plants in harsh environments. Interference signals generated by electrical, rotating and communication equipment and sensors attached to nuclear power plants, as well as malicious enemy interference sources, cause their electromagnetic environment to become increasingly complex. For the complex interference environment of nuclear power plant compartment, this paper firstly adopts supervised and unsupervised learning for interference cognition, and after obtaining the features of wireless interference sequences, anti-interference decision based on deep reinforcement learning and array antenna multi-domain anti-interference based on intelligent decision are implemented. From the intelligent interference simulation based on the physical transmission system, it can be seen that the proposed method can effectively identify and resist typical interference signals, which can bring reference to the anti-interference of nuclear power plants. © 2023, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Nuclear power plants

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An Unsupervised Learning-Based Framework for Effective Representation Extraction of reactor Accidents 23rd

An Unsupervised Learning-Based Framework for Effective Repre...

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23rd Pacific Basin Nuclear Conference, PBNC 2022

作者： Li, Chengyuan Li, Meifu Qiu, Zhifang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Sichuan Chengdu China

ISBN: (纸本)9789811987793

Compared to knowledge-based diagnostic systems, data-based methods tend to perform better in terms of speed and accuracy in diagnosing reactor accidents, and have significant advantages in terms of scalability of models. With the increasing use of high-precision system analysis programs in nuclear engineering, the number of high-fidelity computational data for accident simulation is exploding. Therefore, an algorithm that can achieve both automatic extraction of low-dimensional features from the data and guarantee the validity of the features is needed to improve the performance and confidence of the accident diagnosis system. This study proposes an autoencoder-based autonomous learning framework, namely Padded Auto-Encoder (PAE), which is able to automatically encode accident monitoring data that has been noise-added and with partially missing data into low-dimensional feature vectors via a Vision Transformer-based encoder, and to decode the feature vectors into noise-free and complete reconstructed monitoring data. Thus, the encoder part of the framework is able to automatically infer valid representations from partially missing and noisy monitoring data that reflect the complete and noise-free original data, and the representation vectors can be used for downstream tasks for accident diagnosis or else. In this paper, LOCA of HPR1000 was used as the study object, and the PAE was trained by an unsupervised method using cases with different break locations and sizes as the dataset. The encoder part of the pre-trained PAE was subsequently used as the feature extractor for the monitoring data, and several basic statistical learning algorithms for predicting the break locations and sizes. The results of the study show that the pre-trained diagnostic model with two stages has a better performance in break location and size diagnostic capability with an improvement of 41.62% and 80.86% in the metrics respectively, compared to the diagnostic model with end-to-end model st

关键词： Location

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Research on Thermal-Hydraulic Model and Characteristics of Lead Cooled Traveling Wave reactor 23rd

Research on Thermal-Hydraulic Model and Characteristics of L...

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23rd Pacific Basin Nuclear Conference, PBNC 2022

作者： Zhang, Yaoxiang Zhang, Xue Yu, Hongxing Du, Sijia Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

ISBN: (纸本)9789819910229

Traveling wave reactor (TWR) is an advanced nuclear power system, which can keep the total amount of fissionable nuclides constant during its lifetime through the transformation of fissionable nuclides in the reactor, so as to maintain the stability of the effective proliferation factor of the reactor core. As a coolant, lead has low neutron moderation ability and capture cross section. At the same time, it also has excellent thermal characteristics. Lead cooled traveling wave reactor is a reactor with high inherent safety and good application prospects. The research on reactor core mainly focuses on the physical calculation of the reactor, and the analysis of its thermal hydraulic model and characteristics is relatively scarce, which limits the design and safety analysis of the advanced reactor. Therefore, it is necessary to investigate the thermal hydraulic model of lead cooled traveling wave reactor. In this paper, a thermal hydraulic model suitable for the core of lead cooled traveling wave reactor is proposed. Considering the axial particularity of heat release from the core of traveling wave reactor, the model considers the axial thermal conductivity of fuel rods ignored by other thermal hydraulic models. Due to the strong thermal conductivity of core coolant, the thermal model also considers the heat transfer of fuel assemblies and the flow distribution of core flow. The research shows that the thermal hydraulic model is suitable for the analysis of thermal hydraulic characteristics of lead cooled traveling wave reactor, the thermal hydraulic characteristics are studied by using the thermal hydraulic model. © 2023, The Author(s).

关键词： Thermal conductivity

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Prospects for Next Generation Nuclear Power system 23rd

Prospects for Next Generation Nuclear Power System

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23rd Pacific Basin Nuclear Conference, PBNC 2022

作者： Zhang, Yaoxiang Yu, Hongxing Xia, Bangyang Li, Wenjie Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

ISBN: (纸本)9789811988981

systematically research has been carried out on the demand of economic development for low-carbon clean energy and the opportunities faced by the development of nuclear power. A new generation of nuclear power technical index of "Inherent safety, Economics competitiveness, Environment-friendly, and Sustainability" has been formed, which provides guidance for the innovation and development of nuclear power technology. Considering the technical potential of various advanced reactor including the fourth-generation nuclear power system, inherently safe multi-purposed lead-cooled traveling wave reactor is proposed to solve the long-standing problems that have confronted nuclear power. The advanced reactor has attracted much attention because of their unique advantages in fuel supply issue, inherent safety, and economic efficiency. In this study, the main features of the advanced reactors are reviewed and current challenges in their development are pointed out. The advanced reactor obviates the fuel supply issues, including the nuclide enrichment, while making abundant natural uranium usable nuclear fuel. Each of these novel features is technologically separable, so that these new reactor design concepts can be also applied piecewise to enhance prospects in many different utilization circumstances. © 2023, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Nuclear energy

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The Optimization design of the reactor Coolant system Based on Optimus 23rd

The Optimization Design of the Reactor Coolant System Based ...

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23rd Pacific Basin Nuclear Conference, PBNC 2022

作者： Yanli, Yuan Xianhui, Ye Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Sichuan Chengdu China

ISBN: (纸本)9789811987793

Structural dynamic analysis of reactor coolant systems is an important means of evaluating the safety of nuclear power plants under seismic loads. The seismic load distribution of the equipment can be obtained through the dynamic analysis of the reactor coolant system, and the stress of the equipment can be analyzed. In order to reduce the seismic stress of the equipment, effectively reducing the seismic load of the equipment is an important way. In this paper, aiming at the problem of excessive load at the outlet nozzle of the reactor pressure vessel, combined with the design characteristics of the reactor coolant system under the seismic, the optimal design of the reactor coolant system is carried out by the OPTIMUS [1] which is a kind of software developed by Noesis Solutions Company, by which the multidisciplinary analysis processes can be integrated for collaborative optimization design, and ANSYS [2], MATLAB and other software can be integrated to automate the simulation processes. Firstly, a nonlinear model for the dynamic analysis of the reactor coolant system is established, and then the dynamic response sensitivity analysis of the reactor coolant system is carried out, from which the design variables for subsequent optimization analysis are selected, and then the optimization analysis is processed with the force load at the outlet nozzle of the reactor pressure vessel minimized as the target function, and with the force load at the inlet nozzle of the reactor pressure vessel, the bending moment load at the inlet nozzle of the reactor pressure vessel and the bending moment load at the outlet nozzle of the reactor pressure vessel as the state variables. The results show that after the optimization, the load at the outlet nozzle of the reactor pressure vessel is effectively reduced. Relevant optimization analysis methods and analysis processes can be extended to other systems and equipment in the nuclear power plants. © 2023, The Author(s), under exclusive li

关键词： Nozzles

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Research on Ultra High Flux Research reactor 23rd

Research on Ultra High Flux Research Reactor

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23rd Pacific Basin Nuclear Conference, PBNC 2022

作者： Zhang, Xue Yu, Hongxing Xia, Bangyang Li, Wenjie Zhang, Xilin Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

ISBN: (纸本)9789819910229

The lack of fast neutron irradiation test equipment is a key practical factor that restricts the development of the new generation nuclear energy technology with fast reactor as the core in China. Scarce isotopes such as californium 242, which are urgently needed in industry and medicine, are also highly dependent on imports. In order to provide a fast reactor material irradiation environment and isotope production environment, significantly accelerate the development of fast reactor fuel and structural material technology, and improve the efficiency of rare isotope production, this paper proposes a multifunctional inherently safe ultra-high flux reactor design with fast/thermal neutron flux levels up to 1016 n/(cm2.s), namely UFR-1016. The neutron energy spectrum covers fast spectrum, superheat spectrum and thermal spectrum;several material irradiation test orifices, isotope production orifices and test loop orifices are arranged in the core, and the test loop supports many new reactor coolants such as H2O, Na, He, He-Xr, molten salt, etc., and supports the simulation of typical transient processes and accident conditions. This paper discusses the overall construction scheme of the reactor, and the current construction scheme shows that each performance index can meet the target requirements and achieve the expected functions while ensuring the safety performance of the reactor. This study can provide a basis for the development of ultra-high flux reactors and provide strong support for the cause of nuclear energy, nuclear technology, and nuclear medicine in China. © 2023, The Author(s).

关键词： Nuclear energy

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Analytical Method Research of Source Term for Steam Generator Tube Rupture Accident of Small Modular reactor 23rd

Analytical Method Research of Source Term for Steam Generato...

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23rd Pacific Basin Nuclear Conference, PBNC 2022

作者： Xia, Ming-ming Wang, Jun-long Zhu, Jian-ping Tian, Chao Liu, Jia-jia Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Sicuan Chengdu China

ISBN: (纸本)9789819910229

Based on the full reference to the existing engineering practice and safety review experience, and considering the actual design characteristics of the small modular reactor ACP100, a set of source term analysis method suitable for ACP100 of the steam generator tube rupture (SGTR) accident is proposed, and the source term analysis and consequence evaluation of ACP100 SGTR accident was carried out using this method. The analysis shows that the radiological consequences of the accident source term calculated by this method meet the acceptance criteria of small modular reactor. The analysis results of this article can provide support for the follow-up review of accident source term of ACP100. © 2023, The Author(s).

关键词： Accidents

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Study on the Influence of reactor system LOCA Modeling Mode on Dynamic Analysis 23rd

Study on the Influence of Reactor System LOCA Modeling Mode ...

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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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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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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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