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检索条件"机构=Science and Technology on Reactor system Design Technology Laboratory"

共 1276 条记录，以下是81-90 订阅

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A Novel Model for Dynamic Process Study of reactor Control Rod Drive Mechanism 26

A Novel Model for Dynamic Process Study of Reactor Control R...

引用

26th International Conference on Electrical Machines and systems, ICEMS 2023

作者： Yang, Yun Xu, Qiwei Zhang, Xuefeng Yu, Tianda Chen, Xinan Zhao, Yizhou Chongqing University National Key Laboratory of Power Transmission Equipment Technology Chongqing China Nuclear Power Institute of China Science and Technology on Reactor System Design Technology Laboratory Chengdu China

ISBN: (纸本)9798350317589

The magnetic jack type control rod drive mechanism (CRDM) is a critical electromagnetic actuator utilized in reactors, with its movable components subject to the combined effects of electromagnetic force, liquid resistance, spring force, and other factors, resulting in highly nonlinear and uncertain motion characteristics. This article establishes a novel model for the dynamic process analysis of the CRDM. The dynamic electromagnetic force and inductance of the CRDM are analyzed based on the equivalent magnetic network (EMN) model, taking into account mesh reconstruction and magnetic saturation. Through kinematic analysis and parameter coupling analysis, electromagnetic and kinematic parameters during the CRDM lifting process are obtained. Finally, finite element analysis (FEA) and experimentation are conducted to validate the accuracy of the proposed model. © 2023 IEEE.

关键词： Kinematics

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Development and Verification of a New Depletion, Activation and Radiation Source Term Calculation Code 23rd

Development and Verification of a New Depletion, Activation ...

引用

23rd Pacific Basin Nuclear Conference, PBNC 2022

作者： Wen, Xingjian Zhai, Zian Tang, Songqian Tian, Chao Liu, Zhouyu Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Sichuan Chengdu China Xi’an Jiaotong University Xi’anShaanxi China

ISBN: (纸本)9789819910229

Existing depletion and source term calculation codes lack flexible interfaces, which is difficult to meet the actual engineering design needs. Based on the Chebyshev rational approximation method, a new depletion, activation and source term calculation code TIST is developed and verified. Full-fidelity depletion library based on ENDF-VII.0 and EAF2010 contains 3837 isotopes is generated. Flexible interfaces are built-in to support cross sections from different sources for depletion and activation calculations. The EAF2010 library and the cross sections from the high-fidelity code NECP-X are adopted by TIST for the verification in this work. Based on the nuclide inventory of the depletion and activation calculation, the radiation source term calculation capability of TIST is developed, including the radioactivity, the decay heat, the neutron source and the photon source. The comparison results with NECP-X and FISPACT demonstrate good accuracy of TIST. © 2023, The Author(s).

关键词： Chemical activation

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STUDY ON HYDROGEN RISK VENTING MITIGATION MEASURE OF SMALL STEEL CONTAINMENT DURING SEVERE ACCIDENT 29

STUDY ON HYDROGEN RISK VENTING MITIGATION MEASURE OF SMALL S...

引用

2022 29th International Conference on Nuclear Engineering, ICONE 2022

作者： Zou, Zhiqiang Zhang, Hang Zhang, Ming Peng, Huanhuan Xu, Youyou Deng, Jian Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780791886465

The hydrogen risk in small steel containment is one of the most important safety problems of small modular reactor (SMR), effective measures need to be taken to mitigate the risk of high concentration hydrogen in a small free volume during severe accidents. In the early stage of a severe accident caused by a loss of coolant accident, the pressure in the containment may rise rapidly, probably triggering the immediate venting of the containment, which can protect the integrity of the containment in the stage of pressure rise in the short term. In the later stage of a severe accident, the generation of hydrogen from reactor core may cause hydrogen combustion or even explosion, which may cause the possibility of overpressure in the containment. In this paper, Computational Fluid Dynamics (CFD) GASFLOW-MPI code is used to analyze the effect of containment venting measures on hydrogen risk in small steel containment of SMR during severe accident initiated by direct vessel injection line break loss-of-coolant. The analysis results show that the venting action can effectively depressurize the containment, the larger the area of the venting pipe is, the more obvious the depressurization effect is. However, hydrogen cannot be effectively discharged in the middle and later stage of the accident, so the risk of hydrogen cannot be completely eliminated, and the possibility of flame acceleration in containment cannot be ruled out for the reason that the negative pressure generated by water vapor condensation will suck in the air outside the containment and form a combustible gas cloud again. Copyright © 2022 by ASME.

关键词： Risk assessment

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SGTR ANALYSIS OFSMR ONCE-THROUGH STEAM GENERATOR 29

SGTR ANALYSIS OFSMR ONCE-THROUGH STEAM GENERATOR

引用

2022 29th International Conference on Nuclear Engineering, ICONE 2022

作者： Zhu, Ye Xiang, Cheng Wei, Liu Zhiyun, Cai Yun, Ren Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780791886410

The steam generator tube rupture (SGTR) of a nuclear power plant is a transient that is subject to numerous studies. The casing once-through steam generator has great heat-transfer capacity due to its dual-layer heat transfer structure. Base on the structure and operational characteristic of casing once-through steam generator used in small module reactor (SMR), a graphical model is designed using process module in APROS. Operating parameters of the casing once-through steam generator is simulated under steady state of different power load and SGTR condition. According to the results, the enthalpy and the temperature of secondary side outlet first increases and then decreases with the increasing power load and reaches the greatest value when the power load is 60%. Further study the impact of the position and degree of the breakage on the operating characteristics of the casing once-through steam generator when the steam generator tube rupture. The results show that the influence on the operation characteristic of the casing once-through steam generator increase when the tube rupture takes place in the position which is near the primary side inlet. And when the size of tube rupture becomes larger, the leakage mass flow from primary side to secondary side increase, thus cause the enthalpy and the temperature of secondary outlet further decrease. The simulating results fit theoretical analysis, showing that the casing once-through steam generator can be simulated accurately in APROS. Reference and guidance can be provided for the design and optimization of casing once-through steam generator as well as the safety analysis. Copyright © 2022 by ASME.

关键词： Heat transfer

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Determination of Hardware Logic Drawing Simulation Modeling Criteria and Its Application in Nuclear Safety-Class DCS 6th

Determination of Hardware Logic Drawing Simulation Modeling ...

引用

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

作者： Zhang, Xu Chen, Shi-Yong Yao, Zhang Deng, Zhi-Guang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9789811911804

In digital control system (DCS) design, independent hardware logic circuit design is usually used to deal with common causes failures of software. Due to the distinct difference between the design criteria of electrical design diagram and logic schematic diagram, the hardware logic verification should be carried out in the form of translation simulation. In this paper, the modeling criteria of DCS hardware logic simulation are summarized. The modeling criteria take into account both the standard engineering hardware logic drawing specifications and the characteristics of Simulink platform, and the DCS hardware logic electrical design information is translated to the Simulink platform for automatic redrawing according to specific criteria. A complete hardware logic simulation system was established based on the modeling criteria, and the application was verified in HPR 1000 Zhangzhou nuclear power plant project. © 2022, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Digital control systems

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DATA DRIVEN METHODS FOR BREAK SIZE AND LOCATION ESTIMATION IN LOCA BASED ON DEEP LEARNING 29

DATA DRIVEN METHODS FOR BREAK SIZE AND LOCATION ESTIMATION I...

引用

2022 29th International Conference on Nuclear Engineering, ICONE 2022

作者： Yuan, Peng Deng, Jian Qiu, Zhifang Zhu, Dahuan Huang, Tao Li, Zhongchun Du, Peng Xu, Qinglan Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780791886427

Loss of coolant accident (LOCA) is one of the most important accidents in thermal hydraulic design of nuclear power plant. The traditional analysis method is difficult to realize the rapid prediction of the size and location of the break in the LOCA, while machine learning provides an idea for the rapid diagnosis of the initial cause of LOCA. In this paper HPR1000 which is independently designed by China was taken as the object, the process of LOCA under various break location and sizes is studied by using the advanced reactor system analysis code ARSAC (Advanced reactor system Analysis Code). The key thermal hydraulic parameters including temperature, pressure, flow rate and water level in the transient process of LOCA are analyzed, and a series of data sets are generated. Based on convolution neural network algorithm of deep learning, the transient thermal hydraulic parameters of HPR1000 under different working conditions are learned and the accident diagnosis model is obtained. By comparing with the test data, the trained model can quickly and accurately predict the size and location of LOCA, which can be subsequently extended to the diagnosis of other kinds of accidents in nuclear power plants and has certain application prospects. Copyright © 2022 by ASME.

关键词： Location

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COMPARISON OF NUMERICAL ANALYSIS METHODS OF COOLANT ENVIRONMENTAL FATIGUE BASED ON STRAIN RATE CONTROL 29

COMPARISON OF NUMERICAL ANALYSIS METHODS OF COOLANT ENVIRONM...

引用

2022 29th International Conference on Nuclear Engineering, ICONE 2022

作者： Shao, Xuejiao Xie, Hai Zhang, Yixiong Xiong, Furui Wang, Xinjun Shi, Kaikai Yu, Mingda Huang, Xuan Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780791886397

In this paper, the transformed strain rate of environmental fatigue correction factor (Fen), which is used to consider the influence of PWR primary coolant environment on material fatigue, is studied. EPRI's guideline and theAFCEN's probationary phase rule (RPP) No.3 of Volume VI in RCC-M 2017 edition provide two different calculation methods of transformed strain rate separately. NUREG CR-6909 gives two versions of prediction models of Fen for Low alloy steel, nickel base alloy and austenitic stainless steel which are the three main materials of the primary circuit system. In this paper, the model and transients of Sample 2 in EPRI's guideline are selected for calculation and comparison, so as to verify whether the analysis method is reasonable. Firstly, the fatigue analysis in air is carried out, and then based on the combination, the detailed calculation of strain rate is carried out to calculate environmental fatigue results. The analysis results are compared with the results of guideline, and the valuesare close. Then, the main pipeline safety injection nozzle components with small fatigue margin in PWR are selected to carry out environmental fatigue analysis. In the analysis, different methods of calculating modified strain rates are used, and different prediction models in CR-6909 are *** coefficient of Fenintegratedin RPP3 isconsidered, and the effects of various calculation methods on the results are compared. The results show that for austenitic stainless steel, the results of RCC-M RPP3 method are mostly less than those of EPRI guidelines. Compared with the prediction model of NUREG CR-6909 version 0, the prediction model version 1 greatly decreases the Fen results. Copyright © 2022 by ASME.

关键词： Strain rate

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EVALUATION OF THE PARAMETER IMPORTANCE CHANGES OVER TIME FOR NUCLEAR reactor TRANSIENTS 29

EVALUATION OF THE PARAMETER IMPORTANCE CHANGES OVER TIME FOR...

引用

2022 29th International Conference on Nuclear Engineering, ICONE 2022

作者： Xiong, Qingwen Song, Gongle Du, Peng Huang, Tao Deng, Jian Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780791886410

The uncertainty analysis and sensitivity analysis are two important parts in the safety analysis of nuclear reactor accidents, and the main purpose of the sensitivity analysis is to quantify the importance of the input on the target output. In typical sensitivity analysis for the nuclear reactor transients, specific techniques such as the regression analysis or correlation coefficient analysis are utilized to calculate the sensitivity measures of the inputs on the target output at scalar points, for instance the peak cladding temperature (PCT). However, the phenomena and parameters vary significantly over time during the transients, thus the influence of each input also changes. In order to evaluate the importance changes of the inputs on the target output over time, a time-dependent global sensitivity analysis technique was developed base on the B-spline function system of the functional data analysis, the landmark registration strategy, and an efficient moment-independent sensitivity analysis method. The developed technique was adopted in the sensitivity analysis of a large break loss of coolant accident (LBLOCA) of the nuclear reactor, and the time-dependent sensitivity measures of the inputs on the maximum cladding temperature were acquired and analyzed. Results showed that the proposed technique is capable of evaluating the importance changes of the key parameters in nuclear reactor accident transients. Copyright © 2022 by ASME.

关键词： Sensitivity analysis

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Numerical Simulation of Flow Boiling Heat Transfer in Helical Tubes Under Marine Conditions 23rd

Numerical Simulation of Flow Boiling Heat Transfer in Helica...

引用

23rd Pacific Basin Nuclear Conference, PBNC 2022

作者： Yuan, Leqi Cheng, Kun Bian, Haozhi Liao, Yaping Jiang, Chenxi Fundamental Science on Nuclear Safety and Simulation Technology Laboratory Harbin Engineering University Heilongjiang Harbin China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Sichuan Chengdu China

ISBN: (纸本)9789819910229

Lead-based cooled reactors in most countries and some small reactors at sea use helical tube steam generators. Compared with U-tubes, the convection heat transfer coefficient in the spiral tube is higher, the structure is more compact, and the secondary flow is generated under the action of centrifugal force and gravity, which can achieve the effect of wetting the inner wall of the tube. However, due to the importance of the steam generator in the reactor and the complexity of the flow and boiling in the helical tube, the aggregation behavior of bubbles, the distribution of the two-phase interface and the secondary flow in the tube will significantly affect the heat transfer characteristics, so the gas-liquid phase in the tube is studied. Distribution, changes in heat transfer coefficients, and fluid flow characteristics are very important. In order to study the boiling heat transfer characteristics of helical once-through steam generators under static and marine conditions to provide safe and reliable energy supply for offshore facilities such as marine floating, this study uses STAR-CCM+ software, VOF method and Rohsenow boiling model to study the heat transfer capacity and flow characteristics of flow boiling in a helical tube under swaying and tilting conditions. The gas-liquid phase distribution characteristics, secondary flow variation characteristics and convective heat transfer coefficient of the fluid under different swing functions and inclined positions are obtained by numerical calculation, and the law of physical parameters changing with the cycle is found. The research results show that the secondary flow and heat transfer capacity in the tube change with the cycle, and the change is most obvious at the tube length of 0.8m. 5% of the normal condition;when the inclination angle is 45°, the maximum increase of the convection heat transfer coefficient is 16.8%, and the maximum decrease is 6.6%. © 2023, The Author(s).

关键词： Computational fluid dynamics

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STUDY ON STEADY OPERATIONAL CHARACTERISTICS OF FLOATING NUCLEAR reactor UNDER OCEAN CONDITION 29

STUDY ON STEADY OPERATIONAL CHARACTERISTICS OF FLOATING NUCL...

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2022 29th International Conference on Nuclear Engineering, ICONE 2022

作者： Cheng, Kun Chu, Xiao Qiu, Zhifang Xian, Lin Deng, Jian Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780791886380

Different from land-based reactors, floating reactors are affected by the ocean condition. Numerical simulation is conducted to study the effects of heaving and rolling motion on typical floating nuclear reactor (FNR) under rated power operation by ARSAC code with ocean condition calculation module. The calculation results have shown that heaving motion can cause the periodic fluctuations in the flow rate of both the primary and secondary loops in FNR. The amplitude of heaving motion is the main influence factor on the flow rate fluctuation of the primary loop while the effect of heaving motion period is limited. Rolling motion can cause the out-phase flow fluctuation in the two symmetric loops of the primary system. However, the rolling motion has limited impact on the core flow rate and cannot cause dramatic changes of system parameters at the core. Within the range of motion parameters used in this paper, heaving and rolling motion cannot have a substantial impact on the safe operation of FNR under rated power operation. Copyright © 2022 by ASME.

关键词： Oceanography

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