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检索条件"机构=National Key Laboratory of Science and Technology on Reactor System Design Technology"

共 2032 条记录，以下是51-60 订阅

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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...

引用

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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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...

引用

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

引用

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

引用

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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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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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...

引用

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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Openfoam Simulation of Damping Controlled Fluidelastic Instability 23rd

Openfoam Simulation of Damping Controlled Fluidelastic Insta...

引用

23rd Pacific Basin Nuclear Conference, PBNC 2022

作者： Feng, Zhipeng Liu, Shuai Qi, Huanhuan Huang, Xuan Lv, Xi Science and Technology On Reactor System Design Technology Laboratory Nuclear Power Institute of China Sichuan Chengdu China

ISBN: (纸本)9789811987793

A parallel triangular tube bundle with a single flexible tube is taken as the object, and the most dangerous mechanism of flow induced vibration in the tube bundle structure, fluidelastic instability (FEI) is studied based on the open-source CFD tool OpenFOAM. Firstly, a numerical model describing the fluid-structure interaction of the tube bundle is established by combining computational fluid dynamics with the vibration equation of the tube;Then, the variation of lift coefficient, drag coefficient, lateral displacement and lift spectrum with velocity is discussed. The results show that the open-source tool OpenFOAM can be easily used to predict the fluidelastic instability of tube bundle. © 2023, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Computational fluid dynamics

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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...

引用

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...

引用

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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Research on Output Torque Optimization of Ultrasonic Motor Based on Genetic Algorithm 17th

Research on Output Torque Optimization of Ultrasonic Motor B...

引用

17th Annual Conference of China Electrotechnical Society, CES 2022

作者： Li, Xiaoyang Tang, Jiankai Yang, Bo Wu, Hao Xu, Qiwei Yang, Yun The State Key Laboratory of Power Transmission Equipment & System Security and New Technology Chongqing University Chongqing400030 China Science and Technology on Reactor System Design Technology Laboratory Chengdu610041 China

ISBN: (纸本)9789819904501

Aiming at the problem of small output torque in the design process of traveling to wave ultrasonic motor, this paper proposes a method of optimizing the output torque of traveling to wave ultrasonic motor based on multi-objective genetic algorithm. It can effectively improve the output torque of the motor. Firstly, the structure and operation mechanism of the motor is introduced, and the key factors affecting the output torque of the motor is analyzed reasonably;then the parameterized model of the stator of the ultrasonic motor is established by WORKBENCH, the sample points are randomly selected in the design variable space, and the combination of each sample point is obtained. The modal analysis and harmonic response analysis of the stator structure was carried out to establish the response surfaces model of the stator;finally, the global optimal solution was obtained based on the response surface model by using the multi-objective genetic algorithm (MOGA). The results show that the output torque of the ultra-sonic motor can be increased by 8% before and after optimization. © 2023, Beijing Paike Culture Commu. Co., Ltd.

关键词： Stators

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