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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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Research on the Alarm Threshold for Steam Generator Tube Leak Monitoring of Small Modular reactor 23rd

Research on the Alarm Threshold for Steam Generator Tube Lea...

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23rd Pacific Basin Nuclear Conference, PBNC 2022

作者： Xia, Ming-ming Wang, Jun-long Xu, Tao Huang, Bo-chen Zhu, Jian-ping Wu, Yi-rui Miao, Jian-xin Chen, Xin Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Sicuan Chengdu China

ISBN: (纸本)9789819910229

On the basis of fully learning from the existing principles of leak monitoring alarm thresholds setting of the steam generator tube in pressurized water reactor nuclear power plants, the principle of setting for N-16 leak monitoring alarm thresholds of steam generator tube in small modular reactors is proposed, and the N-16 monitoring alarm thresholds are given. The calculation method for total gamma count rate of steam generator tube leak monitoring alarm thresholds of the small modular reactor is established, and the total gamma count rate under the condition of the leak corresponding to the N-16 monitoring alarm thresholds is calculated to guide the leak monitoring total gamma alarm thresholds set. © 2023, The Author(s).

关键词： Nuclear power plants

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High-Order Fully Actuated System Models for Strict-Feedback Systems With Increasing Dimensions

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IEEE/CAA Journal of Automatica Sinica 2024年第12期11卷 2451-2462页

作者： Xiang Xu Guang-Ren Duan Shenzhen Key Laboratory of Control Theory and Intelligent Systems Southern University of Science and TechnologyShenzhen 518055 the School of System Design and Intelligent Manufacturing Southern University of Science and TechnologyShenzhen 518055China

This paper mainly addresses control problems of strict-feedback systems(SFSs)with increasing *** with the commonly-considered SFSs where the subsystems have the same dimension,we aim to handle more complex cases,i.e.,the subsystems in the considered SFSs are assumed to have increasing *** transforming the systems into highorder fully-actuated system(HOFAS)models,the stabilizing controllers can be directly *** first-order SFSs,secondorder and high-order SFSs are also considered.

关键词： Increasing dimensions high-order fully-actuated system nonlinear systems strict-feedback systems

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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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Research and Application of Endogenous Security Active Defense technology for Domestic Nuclear Safety-Level Gateway Equipment 7th

Research and Application of Endogenous Security Active Defen...

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

作者： Li, Yong Liu, Ming-Xing Peng, Hao Hou, Rong-Bin Ma, Quan Wang, Ru-Qiao Liang, Gen-Hua Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

ISBN: (纸本)9789819934546

Gateway is a protective equipment at the boundary of the nuclear safety-level I&C system. Aiming at the unknown threat across the boundary of information physical space, this paper studies several key technologies of active defense, and constructs the gateway deep secure architecture that can resist multilevel and multi-dimensional complex attacks. Firstly, the systematic control mechanism of the gateway is constructed, which mechanism includes access control for data flow, IP whitelist at network layer, function code whitelist at application layer, IP/MAC binding and the isolation mechanism of chip’s external ports. Secondly, based on the domestic trust root, the trusted chain from bootloader to the upper-layer application is built to guarantee the secure startup of the embedded system, and the key application processes are measured periodically and verified dynamically during the running process. Based on the above, from building equipment embedded trusted system environment to its security functions, a set of active security defense mechanisms of the gateway equipment sare formed, which enhances the security of the system and improves the information security protection n capability of the equipment. © 2023, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Network security

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Research on Application of High Temperature and High Pressure Radar Liquid Level Measurement in Nuclear Power Plant 1

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

作者： Deng, Zhi-Guang Zhu, Bi-Wei Wu, Qian Wang, Xue-Mei Xu, Si-Jie Zhu, Jia-Liang Dong, Chen-Long Sun, Yong-Sheng Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

ISBN: (数字)9789819934553

ISBN: (纸本)9789819934546

At present, the water levels of high-temperature and high-pressure vessels such as voltage stabilizers and steam generators on ships' nuclear power plants are measured by the pressure principle. Affected by the load conditions in the vessels, the differential pressure water level measurement has problems of low accuracy and low reliability. The radar type liquid level gauge has the advantages of high precision and high reliability, and is an effective means of measuring the liquid level of high temperature and high pressure containers. In this paper, a FMCW system split guided wave radar based on the principle of frequency domain reflection (FDR) is proposed for liquid level measurement. The antenna probe part and the electronics part adopt a split structure, and the electronics part is installed in the secondary instrument cabin with better environmental conditions., the radar is easy to meet the index requirements of shock, vibration and irradiation. The high-frequency signal emitted by the radar antenna is conducted along the coaxial rod probe, and the signal will be reflected only when it encounters the surface of the material, which effectively eliminates the influence of obstacles such as the heating plate in the container, the internal reference pipe water replenishment device and the reference pipe connection. Non-blind zone measurement of liquid level to meet the liquid level measurement requirements of high temperature and high pressure vessels in ship 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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SMR FEEDWATER SYSTEM ANALYSIS 30

SMR FEEDWATER SYSTEM ANALYSIS

引用

30th International Conference on Nuclear Engineering, ICONE 2023

作者： Zhu, Ye Liang, Qiao Xianwei, Liao Zhiyun, Cai Minghao, Liu Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Sichuan Chengdu China

ISBN: (纸本)9784888982566

The once through steam generator has great heat transfer capacity and compact structure. Some small module reactors (SMR) applied once through steam generatorfor its high thermal efficiency. The thermal hydraulic characteristic of once through steam generator is different from traditional natural cycle steam generator as it doesn't contain water level. Thus the related feed water system is important for the steady operation of the power plant. Base on the thermal hydraulic characteristic of once through steam generator, SMR feed water system that contains process and control model is developed using APROS. The operating parameters are simulated under both different steady states and transient condition. The results show that both the thermal hydraulic and automatic control parameters of SMR feed water system can be simulated accurately in APROS. Reference and guidance can be provided for system optimization and safety analysis. © 2023 by JSME.

关键词： Water levels

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Response Time Analysis Method of Safety-Level DCS in Nuclear Power Plant Based on Probability Theory 1

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

作者： Wen, Jing He, Xian-Jian Chen, Zhao Wan, Zi-Yuan Wu, Zhi-Qiang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610000 China

ISBN: (数字)9789819934553

ISBN: (纸本)9789819934546

Aiming at the incompleteness and inadequacy of the response time analysis method of the current nuclear power plant safety-level digital-control system (DCS), this paper proposes an analysis method based on probability theory for the response time of nuclear power plant safety-level DCS. Firstly, according to the structure and characteristics of the safety-level DCS, the probability model of the response time of the safety-level DCS is established, and the calculation method of the probability distribution of the response time is obtained. Then, according to the obtained calculation method and the parameters of the engineering example, the rang of the response time and the probability density function are calculated, and the response time of the engineering example is simulated, and the simulated response time rang, frequency and probability density function fitting curve are obtained. Finally, the response time of the engineering example is tested by the response time automatic test device, a large amount of data with statistical significance is obtained. After the data is processed, the range, frequency and probability density fitting curve of the response time are obtained. By comparing the theoretical calculation results, simulation results and test results of engineering examples. The correctness and feasibility of the analysis method are verified. © 2023, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Nuclear power plants

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Effect of Corrosion Damage on Structural Failure Models Under Different Boundry Conditions 23rd

Effect of Corrosion Damage on Structural Failure Models Unde...

引用

23rd Pacific Basin Nuclear Conference, PBNC 2022

作者： Qian, Sheng Sun, Yu Huangfu, Yuzhao Zhang, Ke Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Sichuan Chengdu China

ISBN: (纸本)9789811987793

Nuclear energy is an important means to deal with the energy crisis and global climate change, and the long-term safe operation of nuclear power plants strongly depends on the service behavior of structural materials. Nuclear power equipment is subject to complex multiple loads in actual situations, especially under compression, and even the buckling of tiny structures may have important safety hazards. Experience has shown that corrosion is an unavoidable result of nuclear power structures, and structures damaged by corrosion will cause internal stress redistribution under compression, resulting in changes in failure modes. This paper takes the most common uniform corrosion and pitting corrosion as the research objects, and explores the effects of different corrosion forms and corrosion parameters on the buckling failure of compression members under the same corrosion rate. © 2023, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Nuclear fuels

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