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

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STUDY ON AUTOMATIC DEPRESSURIZATON system OF SMR

STUDY ON AUTOMATIC DEPRESSURIZATON SYSTEM OF SMR

引用

2020 International Topical Meeting on Advances in Thermal Hydraulics, ATH 2020

作者： Yin, Shasha Sun, Yan Li, Yi Yang, Hongfa Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Sichuan Chengdu China

ISBN: (纸本)9780894487774

Small modular reactor (SMR) is highly cared and researched by people all over the world. There are many differences during reactor operating between the small modular reactor and the traditional large-scale pressurized water reactor, because of the differences in structure and design issues for both reactors. This means that more and better research especially on SMR will be needed. The continued improvements in designing conception of SMR, the research of SMR get to be more in depth and detail. Passive safety system is commonly accepted in the present study, and developed into an indispensable system for the advance nuclear reactor. However, automatic depressurization system (ADS) as an important part of a passive safety system also is a unique system compared with conventional PWR. So ADS system is worthy of comprehensive and in-depth study. In this paper, the severe accident analysis model was established based on the design characteristics of the small modular nuclear. Then, the severe accident process, the thermalhydraulic phenomena and the system safety were researched. Then, based on the results of the severe accident process for SMR, analyze the depressurization capability of ADS system during a station blackout initiated severe accident in SMR, and analysis the effect on accident process. Finally, this paper proposed the optimization and improvement of the automatic depressurization system (ADS). The results of this study are the necessary foundations for the development of small modular reactor severe accident prevention and mitigation measures. © ATH 2020 - International Topical Meeting on Advances in Thermal *** rights reserved.

关键词： Pressurized water reactors

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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

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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RESEARCH OF MODEL FOR LOCA NONLINEAR DYNAMIC ANALYSIS OF reactor COOLANT system 30

RESEARCH OF MODEL FOR LOCA NONLINEAR DYNAMIC ANALYSIS OF REA...

引用

30th International Congress on Sound and Vibration, ICSV 2024

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)9789090390581

The Loss of Coolant Accident (LOCA) is one of the most serious accidents in nuclear power plants. According to the requirements of the United States Nuclear Regulatory Commission (USNRC) SPR, the dynamic response analysis of LOCA transients occurring in the reactor coolant system (RCS) must be performed to verify the design of equipment, supports, anti-whip devices and civil structures[1]. At present, the LOCA dynamic response of reactor coolant systems is mainly analysed using a single loop model, and there are few relevant literatures on LOCA analysis with the complete loop model. The influence on the destructive loop due to the hydraulic load of the unbroken loop cannot be considered when using a single loop model for LOCA dynamic response analysis, which may lead to unreasonable analysis results. In this paper, the two models for nonlinear LOCA dynamic analysis of reactor coolant system are established: one is single loop, and the other is three-loop. The nonlinear time-history dynamic analysis of LOCA is analysed by ANSYS software, and the analysis results from the two models are compared. The results show that the loads from the three-loop LOCA dynamic analysis model are consistent with those of the single loop LOCA analysis. By comparing the results calculated by single loop LOCA analysis model with different boundary conditions at inlet nozzles, it is found that the expansion displacements due to the internal pressure at the inlet nozzles of the reactor pressure vessel cannot be ignored during the LOCA dynamic response analysis. © 2024 Proceedings of the International Congress on Sound and Vibration. All rights reserved.

关键词： Nuclear power plants

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Research on Switch Signal Transmission system Based on Optical Fiber 1

引用

6th International Symposium on Software Reliability, Industrial Safety, Cyber Security and Physical Protection of Nuclear Power Plant, ISNPP 2021

作者： Xiang, Mei-Qiong Zhu, Jia-Liang He, Zheng-Xi Deng, Zhi-Guang Dong, Chen-Long Li, Zhuo-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)9789811911811

ISBN: (纸本)9789811911804

Nuclear power plants need to transmit a large number of switch signals, while traditional electrical transmission of switch signals requires a large number of cables, which makes it difficult to realize self-inspection of transmission path faults and is vulnerable to electromagnetic interference. In view of the needs of a large number of switch signal transmission and the shortcomings of traditional electrical methods, this paper developed two sets of switch signal transmission systems based on optical fiber, which realized the simultaneous transmission of 16 channels and 2 channels of switch signals on one optical fiber, and had the function of transmission path fault self-checking. It has the advantages of wide frequency response range, small transmission delay, high reliability and strong environmental tolerance, and has broad application prospects. © 2022, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Optical fibers

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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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The sensitivity analysis of the core lower head molten pool model based on variance decomposition 27

The sensitivity analysis of the core lower head molten pool ...

引用

27th International Conference on Nuclear Engineering: Nuclear Power Saves the World!, ICONE 2019

作者： Li, Zhigang Liu, Wei Ming, Pingzhou Pan, Junjie An, Ping Liu, Dong Lu, Wei Yu, Hongxing Nuclear Power Institute of China Science and Technology on Reactor System Design Technology Laboratory ChengDu610041 China

ISBN: (纸本)9784888983051

The lower head molten pool model is an important model for the In Vessel Retention (IVR) effectiveness evaluation. It has the characteristics of complex relationship, many input parameters and large uncertainty. The traditional local sensitivity analysis method has the disadvantages of large amount of calculation and low efficiency when performing sensitivity analysis of complex models. Based on variance decomposition of the global sensitivity analysis method, the sensitivity analysis library(SALib) and IVR analysis code(CISER) developed by Nuclear Power Institute of China(NPIC) are adopt to analyze the sensitivity of the input parameters for four key parameters, such as the lower head wall heat flux ratio. In the paper, the sensitivity coefficient and influence tread of input parameters on key results are obtained, which provides a reference for the optimization of the lower head molten pool model and severe accident mitigation strategies. Copyright © 2018-2019 by JSME

关键词： Sensitivity analysis

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Verification Method and Basic Guarantee of RPS Availability in Nuclear Power Plant 5th

Verification Method and Basic Guarantee of RPS Availability ...

引用

5th International Symposium on Software Reliability, Industrial Safety, Cyber Security and Physical Protection of Nuclear Power Plant, ISNPP 2020

作者： He, Xian-Jian Liu, Ming-Ming Zhang, Qing Luo, Xiao-Jun Wen, Jing Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

ISBN: (纸本)9789811634550

The third generation nuclear power reactor protection system (RPS) based on computer technology, that is, safety DCS system, is designed with large-scale integrated circuits and real-time embedded software, which has the characteristics of complex structure, large system scale and high verification difficulty. In order to ensure the safe operation and economic benefits of nuclear power plants, RPS needs not only reliability and stability, but also high availability. In order to explore the feasibility of verification of DCS availability index, based on Hualong One safety DCS engineering prototype, a set of availability testing methods and basic guarantee conditions suitable for DCS suppliers are summarized for discussion and exchange. © 2021, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Nuclear power plants

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An Ergonomic Analysis of Main Control Room Console in Nuclear Power Plant Based on Jack 5th

An Ergonomic Analysis of Main Control Room Console in Nuclea...

引用

5th International Symposium on Software Reliability, Industrial Safety, Cyber Security and Physical Protection of Nuclear Power Plant, ISNPP 2020

作者： Li, Yu-Tong Liu, Ming-Ming Huang, Qing-Huai Gao, Nan Liu, Guan-Ron Zhou, Mao Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

ISBN: (纸本)9789811634550

In the process of human factor design and preliminary verification of the control room console in Nuclear Power Plant, traditional 2D human model design and analysis method is commonly used. However the results turn to be inaccurate and time-consuming. With the help of human simulation tool Jack, quick and precise analysis can be achieved. This paper puts forward the human factor engineering analysis of main control room console in nuclear power plant based on Jack including the visibility, accessibility, and collision detection. By comparing this method with traditional 2D design, conclusion can be drawn that Ergonomic Analysis of main control room console Based on Jack is more comprehensive and accurate, thus largely increasing the design efficiency and quality. © 2021, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Nuclear power plants

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CFD ANALYSIS ON THE SMALL MODULE reactor (ACP100) PRS RETURN LINE THERMAL STRATIFICATION PHENOMENON 29

CFD ANALYSIS ON THE SMALL MODULE REACTOR (ACP100) PRS RETURN...

引用

2022 29th International Conference on Nuclear Engineering, ICONE 2022

作者： Peiying, Li Cai, Zhiyun Zhang, Yulong Liu, Hang Liu, Minghao Ye, Zhu Tao, Shuchang Liu, Xianghong Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Sichuan Chengdu China

ISBN: (纸本)9780791886434

The return line of the passive residual system(PRS) of ACP100 is connected to the direct injection nozzle of the pressure vessel. It is a non-isolated dead pipe containing coolant, and there is no insulation layer. The temperature difference between the two sides of the pipe is large, which is prone to thermal stratification. This paper uses FloEFD software to carry out CFD simulation analysis and research on the thermal stratification phenomenon of the pipeline in normal operation and accident conditions. The scope of the return line analysis in this paper begins with the isolation valve at the outlet and ends with the direct injection nozzle on the pressure vessel. The finite volume method is used to solve the model, and the standard k-Ε two-equation model is used for turbulence simulation. The pipeline system adopts a mesh refinement strategy according to the flow channel and wall thickness in the pipeline to obtain a reasonable mesh distribution. The results show that during normal operation, there is obvious thermal stratification at the exit of the PRS isolation valves, and there is a certain vortex in the tube, which can promote the heat exchange between the reactor pressure vessel and the PRS return pipeline;When PRS put into operation, a large temperature gradient is generated at the interface of the annular cavity of the pressure vessel, and the influence of thermal stress and cyclic thermal shock needs to be considered, and an obvious vortex will be formed nearby to promote the rapid heat exchange of cold and hot water in the nearby area. The result of this paper can provide input for PRS return line thermal stress and thermal cycling impact analysis. Copyright © 2022 by ASME and a non-US government agency.

关键词： Vortex flow

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