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30th International Conference on Nuclear Engineering, ICONE 2023

作者： Pu, Xingwu Chen, Deqi Liu, Hanzhou Wang, Ningyuan Zi, Binbin Deng, Jian Key Laboratory of Low-grade Energy Utilization Technologies and Systems Chongqing University Ministry of Education Chongqing400044 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China

ISBN: (纸本)9784888982566

In order to study the characteristics of single rod flow-induced vibration in axial flow, the vibration response of single rod flow-induced vibration in axial flow with different constraints at both ends and different flow velocity was measured with a Laser Doppler Vibrometer. Based on spectral analysis and statistical analysis, the transient vibration response was analyzed in depth. The results show that the vibration amplitude increases with the increase of the flow velocity, and the two ends of the constraints have a significant influence on the flow-induced vibration response. The root means square amplitude value of the simply-supported single rod is 35% larger than that of the fixed one at 3 m/s. Spectrum analysis shows that with the increase of flow velocity, the dominant frequency decreases from 17.5 Hz at 1.5 m/s to 13.8 Hz at 3.0 m/s under simply-supported constraints, while the dominant frequency basically fluctuates around 13.0 Hz under fixed support constraints. The vibration frequency of simply-supported constraints are distributed between 6 Hz and 28 Hz, while those of fixed constraints are mainly distributed between 12 Hz and 16 Hz. The statistical results show that in the X direction of the simply-supported constraints, the geometric center point of the rod is shifted, which indicates that the rod is buckling. © 2023 by JSME.

关键词： Nuclear reactors

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Investigation on the Influence of Entrained Air to Internal Flow of Centrifugal Pumps under Cavitation Condition

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International Journal of Fluid Machinery and systems 2022年第1期15卷 56-63页

作者： Fanjie, Deng Jianping, Yuan Song, Yang Biaobiao, Wang Qiaorui, Si Research Center of Fluid Machinery Engineering and Technology Jiangsu University Jiangsu Zhenjiang212013 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Sichuan Chengdu610000 China

The introduction of air is often used in engineering to delay the occurrence of cavitation. A steady simulation on a single-stage and single-suction centrifugal pump was used to study the influence of the air on the internal flow under cavitation condition at 1% inlet void fraction. Steady simulation based on SST k-ω turbulence model and Rayleigh-Plesset cavitation model included in Ansys CFX 17.0 were processed to obtain vapor volume fraction, gas volume fraction and the turbulent energy of the impeller under the condition of natural cavitation and cavitation with entrained air. A homogeneous two-phase flow model was adopted to describe the air-water mixed flow. The experimental cavitation performance curve provides boundary condition support for the numerical simulation. In conclusion, it shows that the head of the pump keeps steady after 1% air entrained like natural cavitation. Then it goes worse when the cavitation number decreases but with lower NPSHr which means cavitation performance is improved by 1% air entrained. It indicated that the entrained air changes the distribution of the vapor in the passage of the impeller. However, the descent of the pressure over a certain value will also do harm to the internal flow of the model pump whether it is ventilati ng or not. The internal flow of the pump will be seriously damaged and even be blocked completely due to more and more vapor generated causing a dramatic decline of the head. © 2022, Turbomachinery Society of Japan. All rights reserved.

关键词： Cavitation

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RESEARCH ON LOCA DYNAMIC ANALYSIS MODEL OF reactor system 29

RESEARCH ON LOCA DYNAMIC ANALYSIS MODEL OF REACTOR SYSTEM

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2022 29th International Conference on Nuclear Engineering, ICONE 2022

作者： Liu, Shuai Harbin Engineering University Harbin China Science and Technology on Reactor System Design Technology Laboratory Chengdu China

ISBN: (纸本)9780791886472

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 reactor loop system 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 surge 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. Copyright © 2022 by ASME.

关键词： Loss of coolant accidents

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design of Code Generator for Safety Level I&C Software of NPPs 6th

Design of Code Generator for Safety Level I&C Software of NP...

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

作者： Wei, Jiang Lin, Lan Bin, Yang Hou, Rong-Bin Science and Technology On Reactor System Design Technology Laboratory Nuclear Power Institute of the China Sichuan Chengdu China

ISBN: (纸本)9789811911804

The function block diagram is used to design control algorithm of nuclear safety-level I&C (Instrumentation & Control) system. Then it is converted into C code by code generator. And the C code is downloaded into embedded software platform of NASPIC after compiled by a common C compiler. Therefore, the credibility of the code generator is directly related to the safety of nuclear power plant operation. At present, the methods which are used to guarantee the credibility of code generators mainly include test and V&V. However, the formal method describes the software system from a mathematical point of view and verifies its correctness of translation processes, which can improve the credibility of the code generator to the greatest extent. Based on the formal method and compilation technology, this paper introduces the design and implement of code generator for safety I&C software. A initial test has been performed to check the requirement of nuclear I&C software development. The test results show that the code generator can translate Lustre program into C program. © 2022, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Nuclear power plants

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NUMERICAL SIMULATION OF THERMAL CHARACTERISTICS OF BOILING TWO-PHASE FLOW IN A SPIRAL TUBE 30

NUMERICAL SIMULATION OF THERMAL CHARACTERISTICS OF BOILING T...

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30th International Conference on Nuclear Engineering, ICONE 2023

作者： Hongcheng, Dong Guangchao, Yang Guo, Chen Xiaofei, Yu Simian, Qin Shanshan, Bu Deqi, Chen Key Laboratory of Low-grade Energy Utilization Technologies and Systems Chongqing University Ministry of Education Chongqing400044 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610200 China

ISBN: (纸本)9784888982566

Helical coil once-through tube steam generator (HCOTSG) has the advantages of compact structure and higher heat transfer efficiency due to its special structure. In order to study the flow boiling heat transfer characteristics in HCOTSG, this paper used the FLUENT commercial software, based on the Critical heat flux(CHF) wall boiling model, to conduct numerical simulation in the spiral tube under the given wall temperature boundary conditions, and verified its reliability by comparing the relevant experimental research. On this basis, the flow boiling heat transfer process in the spiral tube was simulated under different flow parameters, and the velocity field, temperature field and the distribution of vapor-liquid two-phases in the spiral tube were studied, It is found that under the condition of constant wall temperature boundary, the influence of different inlet temperature only exists in the single-phase water region near the inlet, and has little influence on the boiling two-phase region and single-phase steam region. The void fraction decreases at the same height in the boiling two-phase region with the increase of inlet flow velocity, which is due to the decrease of the heating time when the fluid reaches the same height. © 2023 by JSME.

关键词： Heat flux

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Constructing Basic Probability Assignment Based on the Slope Correlation Degree 5

Constructing Basic Probability Assignment Based on the Slope...

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5th IEEE International Conference on Electronics technology, ICET 2022

作者： Chen, Jie Yang, Pengcheng Pu, Xiaofei Xie, Hao Wang, Jinxin Wu, Kunren Nuclear Power Institute of China Science and Technology on Reactor System Design Technology Laboratory Chengdu610213 China

ISBN: (纸本)9781665485081

DS evidence theory can be applied to deal with multi-sensor data and get reliable and accurate decision result of system's target or object due to its good measurement and reasonable description of system's uncertainty. In practical application field, DS evidence theory is achieved by two steps, one is the acquisition of the basic probability assignment function (BPA), and the other is the synthesis of multiple evidences that is obtained from multi-sensor data. The existing relevant literatures mainly solve the combination problem of conflict evidences, while lack of systematic research on the acquisition of BPA. In this paper, we focus on the discussion and innovation of the acquisition of BPA. BPA in DS evidence theory is the basic unit to describe the uncertainty of the system and evidences, which fundamentally determines its superiority with respect to probability theory and Bayesian reasoning. Getting a reasonable and effective BPA is the premise of the application of DS evidence theory. Traditional BPA acquisition methods have the limitations like absent priori knowledge and low acquisition accuracy. Aiming to solve the technical difficulties of traditional methods, we put forward a new BPA acquisition algorithm based on the slope correlative degree. The novel algorithm proposed in this paper successfully realizes the reliable descriptions of system's uncertainty, which is conducive to product the accurate BPAs and the precise fusion results. Experimental results and analyses demonstrate that the innovative algorithm can not only produce reasonable BPAs, which can better fit the sensors' monitoring data, but also expand the gap between the support degree of different propositions in BPA, which is more beneficial to the synthesis of evidences and the decision fusion of the system. Thus, the novel BPA acquisition algorithm has great application value and research significance. © 2022 IEEE.

关键词： Uncertainty analysis

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Definition of Standard Input Data in Launch Dynamics of Multibody systems 2nd

Definition of Standard Input Data in Launch Dynamics of Mult...

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2nd International Conference of Mechanical system Dynamics, ICMSD 2023

作者： Li, Geng Zhou, Qinbo Wang, Guoping Li, Tang National Key Laboratory of Complex Multibody System Dynamics Nanjing University of Science and Technology Nanjing China Institute of Launch Dynamics Nanjing University of Science and Technology Nanjing China

ISBN: (纸本)9789819980475

At present, each program for launch dynamics of multibody systems (LDMS) has its own data structure without a unified standard, which is not beneficial to the program development and maintenance. To improve the program's versatility, it is important to study the standardization of the data structure and input data for the LDMS code. Based on the launching subsystem proposed in (Zhou et al., Def Technol 17(6), 2011–2026 [4]), we design a general data structure and standard input data (SID) for the codes of the launching subsystem dynamics in this work. Based on object-oriented programming (OOP) technology, the unified modeling language (UML) is adopted to describe the geometry, kinematics, kinetics and subjected forces of the launching subsystem. The UML class diagram is used to describe the variable properties, such as definitions, types, and default units within each data structure. At the same time, the relationship among various classes is represented by straight or dashed line segments with a specialized convention. Further, the input data structure of the launching subsystem dynamics code is designed: the SID is defined for parameter initialization, which is stored in an extensible markup language (XML) document and read by the program. This work provides a theoretical basis and important guidance for the program design of LDMS. © The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd. 2024.

关键词： XML

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Application Verification of Wireless Sensor Network in Nuclear Power Plant 1

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

作者： Deng, Zhi-Guang Wu, Qian Xiang, Mei-Qiong Dong, Chen-Long Qin, Yue Li, Zhuo-Yue Sun, Yong-Sheng Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

ISBN: (数字)9789811911811

ISBN: (纸本)9789811911804

The application of wireless sensor network in nuclear power plant can greatly reduce the workload of nuclear power plant cable wiring, and help to improve the automation and information level of nuclear power plant. However, its application also faces the constraints of strong electromagnetic interference, high temperature and other special complex nuclear environment, especially the narrow multi metal enclosed space of nuclear power plant cabin, which brings challenges to the application of wireless sensor. Based on the Lora wireless protocol, this paper studies the low-power selection of wireless sensor and MCU, sleep / wake-up low-power mechanism, dual star adaptive frequency hopping anti-interference, frequency division and time division multi node real-time communication and other aspects. On this basis, the prototype is developed, and the functional test, high and low temperature test, electromagnetic compatibility test and bench test are carried out. The feasibility of wireless sensor application in nuclear power plant is preliminarily verified, which lays a foundation for further application. © 2022, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Wireless sensor networks

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Research on Switch Signal Transmission system Based on Optical Fiber 1

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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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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...

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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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