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

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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...

引用

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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NUMERICAL INVESTIGATION ON THE THERMAL-HYDRAULIC AND FIV CHARACTERISTICS IN LBE-COOLED HELICAL CRUCIFORM FUEL AND WIRE-WRAPPED FUEL 31

NUMERICAL INVESTIGATION ON THE THERMAL-HYDRAULIC AND FIV CHA...

引用

2024 31st International Conference on Nuclear Engineering, ICONE 2024

作者： Qi, Zhang Haoyu, Wang Cao, Junxian Yuanming, Li Chenxi, Li Zhenhai, Liu Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780791888261

The lead bismuth eutectic (LBE) cooled fast reactor in one of the General 4 advanced reactors, which plays significant role in the nuclear fuel cycle. In the recent years, the wire-wrapped fuel with four wires surrounds the cylindrical fuel is developed and promoted, which may be accept by the LBE-cooled fast reactor. The helical cruciform fuel of the Lightbridge corporation has similar geometrical features with the wire-wrapped fuel. The cross-section of helical cruciform fuel consists of the convex blades and the concave elbows, and the cross-section is rotated around the axis. The blade and elbow of helical cruciform fuel function as the wires while they are part of the nuclear fuel. In the present work, the comparison on the thermal-hydraulic and flow-induced vibration (FIV) characteristics of the helical cruciform fuel and wire-wrapped fuel is carried out by numerical simulation. The fluid-structure interaction is solved by the coupling between the fluid and solid domains, and the transient turbulence fluctuation is solved by large eddy simulation (LES). The results show that the convective heat transfer is enhanced by the spiral structures even though the hot spot phenomenon is caused simultaneously. The hot spot temperature of the helical cruciform fuel is larger than the wire-wrapped fuel duo to the non-uniform distributed heat flux. As the coolant in the flow channel is blocked by the blades and wires, the directional transverse flow is generated and thus the transverse mixing is enhanced. The one-way mixing with opposite direction occurs in the two sides of the single rod. There are similarities in the FIV characteristics for the spiral shaped fuels. The maximum vibration displacement and dominant vibration frequency of the helical cruciform fuel and wire-wrapped fuel are almost the same. © 2024 by ASME.

关键词： Fluid structure interaction

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Comparative study on thermal stress analysis and fatigue curve in stress and fatigue calculation of nuclear equipment 28

Comparative study on thermal stress analysis and fatigue cur...

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2021 28th International Conference on Nuclear Engineering, ICONE 2021

作者： Shao, Xuejiao Xie, Hai Yixiong, Y. Zhang, Liping Fu, Xiaolong Mi, Xue Li, Hui Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780791885246

In the stress and fatigue evaluation of primary nuclear equipment, thermal stress calculation and fatigue curve have a great impact on the calculation results. There is a problem of stability in the step integral when calculating the thermal stress. The transient heat conduction changes rapidly near the initial temperature time point. It is critical to choose the right sub steps. In the nuclear code, the design fatigue curve in the air environment is given by applying a factor of 2 on stress or a factor of 20 on cycles. The factors are used to consider various influences such as data scatter, structure size, surface roughness and load. The factors of fatigue curves in ASME BPVC 2007 edition + 2009 addendum and French AFCEN's RCC-M 2017 edition RPP 2 have changed. Due to the changes of the original test data and modificative factors, the new fatigue curve and the original fatigue curve are crossed. In code case N-761 of ASME, fatigue curves considering the influence of primary coolant environment are given under different strain rates. In the calculation of thermal transient using commercial ANSYS software, the setting of time step has a great influence on the thermal stress. This paper compares and analyzes the influence of different calculation methods on the thermal stress at different positions of the model. The effect of different fatigue curves on fatigue life is also discussed considering different air design curve and different strain rate fatigue curve in water. The results show that the fatigue results obtained by using the modified ASME or RCC-M air fatigue curve are smaller than those before the change when the usage factor is large to a certain extent. After considering the environmental fatigue curve, the fatigue results increase significantly. The accuracy of thermal stress calculation can be significantly improved by adding time points and refining time steps near the time points after the transient sharply changes, otherwise the influence of thermal

关键词： Thermal stress

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Research on the effect of grid clamping failure on flow elastic stability and vortex shedding of fuel rod 26

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2018 26th International Conference on Nuclear Engineering, ICONE 2018

作者： Huan-Huan, Qi Zhi-Peng, Feng Nai-Bin, Jiang Qian, Huang Xuan, Huang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780791851456

Flow elastic stability and vortex shedding were two important mechanisms of the flow induced vibration analysis. Due to the influence of manufacturing process, transportation and irradiation, the clamping action of grid on fuel rods may be invalid. Taking one fuel assemblies as an example, the effects of clamping failure on the natural frequencies, mode shapes, flow elastic stability and vortex shedding were studied. The results show that the effect of the rigid convex support failure on the natural frequency was directly related to the mode shape. The effect of the grid rigid convex failure near the nodes with larger amplitude on the natural frequency was obvious. The velocity of flow at the top and bottom of the fuel rods were larger and the size was comparable, this induced that the rigid convex failure of the top and bottom grids had a significant effect on the ratio of flow elastic stability and the vortex shedding. Copyright © 2018 ASME.

关键词： Vortex shedding

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POST-HOC INTERPRETABILITY BASED PARAMETER SELECTION FOR DATA ORIENTED NUCLEAR reactor ACCIDENT DIAGNOSIS system 29

POST-HOC INTERPRETABILITY BASED PARAMETER SELECTION FOR DATA...

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2022 29th International Conference on Nuclear Engineering, ICONE 2022

作者： Li, Chengyuan Li, Meifu Qiu, Zhifang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780791886427

During applying data-oriented diagnosis systems to distinguishing the type of and evaluating the severity of nuclear power plant initial events, it is of vital importance to decide which parameters to be used as the system input. However, although several diagnosis systems have already achieved acceptable performance in diagnosis precision and speed, hardly have the researchers discussed the method of monitoring point choosing and its layout. For this reason, redundant measuring data are used to train the diagnostic model, leading to high uncertainty of the classification, extra training time consumption, and higher probability of overfitting while training. In this study, a method of choosing thermal hydraulics parameters of a nuclear power plant is proposed, using the theory of post-hoc interpretability theory in deep learning. At the start, a novel Time-sequential Residual Convolutional Neural Network (TRES-CNN) diagnosis model is introduced to identify the position and hydrodynamic diameter of breaks in LOCA, using 38 parameters manually chosen on HPR1000 empirically. Afterwards, post-hoc interpretability methods are applied to evaluate the attributions of diagnosis model’s outputs, deciding which 15 parameters to be more decisive in diagnosing LOCA details. The results show that the TRES-CNN based diagnostic model successfully predicts the position and size of breaks in LOCA via selected 15 parameters of HPR1000, with 25% of time consumption while training the model compared the process using total 38 parameters. In addition, the relative diagnostic accuracy error is within 1.5% compared with the model using parameters chosen empirically, which can be regarded as the same amount of diagnostic reliability. Copyright © 2022 by ASME.

关键词： Nuclear power plants

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Flow-induced vibration analysis of the check valve 25

Flow-induced vibration analysis of the check valve

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25th International Congress on Sound and Vibration 2018: Hiroshima Calling, ICSV 2018

作者： Shuai, Liu Peng, Feng Zhi Zhou, Jiang Xiao Xi, Lv Shou, Zhang Feng Xuan, Huang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9781510868458

As an important over-current adjustment unit in the nuclear power plant piping system, check valve often causes some vibration and noise problems. The study of the existing valves mainly focuses on the simple or fixed opening two-dimensional steady-state flow field, which can't be used to explain the vibration and noise problems. In this paper, a complex passive check valve which has the asymmetric distribution with one import and two exports is proposed. In order to investigate the coupling mechanism between the internal pressure fluctuations and the vibration of system with the automatic adjusting process, the three-dimensional fluid simulation model is constructed. In addition, the vibration characteristic of internal flow field and the irregular fluctuating of the flow field are also calculated when the spool is moving. The coupling vibration characteristics of the flow field is obtained, which provides a meaningful guidance for the vibration reduction of the valve. Copyright © (2018) by International Institute of Acoustics & Vibration. All rights reserved.

关键词： Flow fields

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Preliminary research on the thermal-mechanical coupling behavior simulation method of m3 fuel 28

Preliminary research on the thermal-mechanical coupling beha...

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2021 28th International Conference on Nuclear Engineering, ICONE 2021

作者： Changbing, Tang Yongjun, Jiao Yuanming, Li Kun, Zhang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institution of China Chengdu China

ISBN: (纸本)9780791885253

M3 fuel which microencapsulated TRISO particles in zirconium metal matrix is an integrated fuel designed to replace the traditional pellet-cladding fuel rod. The fuel design can effectively solve the problems of LWR fuel, such as water side corrosion, abrasion, pellet-cladding interaction, stress corrosion cracking and internal pressure bracing. Because of the multiple coating structure of TRISO particles and complex contacting behavior and deformation between TRISO particle and metal matrix, M3 fuel exhibits complex in-pile behavior. In order to establish the simulation method for the M3 fuel, the three-dimensional model was established by using ABAQUS software. The complex contacting behavior and deformation of coated layers and matrix for M3 fuel was considered. The results indicated that bonded strength between TRISO particle and Zr matrix has great effect on the stress distribution of SiC layer and matrix. The vulnerable area of Zr matrix located in the nearest position of TRISO particle to TRISO particle. The maximum tensile hoop stress of SiC layer was up to 150 MPa under strong bonding state, while SiC layer suffered compressive stress during all operation time for the weak bonding state. The failure probability of SiC layer was about 7 10-3 under strong bonding state which much higher than the weak ones. Thus, the integrity of SiC layer can be maintained by controlling the bonding strength between TRISO particle and Zr matrix during manufacturing process. © 2021 American Society of Mechanical Engineers (ASME). All rights reserved.

关键词： Silicon carbide

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A new turbulent Prandtl number model for forced fully-developed pipe heat transfer of liquid metals 27

A new turbulent Prandtl number model for forced fully-develo...

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

作者： Peiying, Li Hongxing, Yu Sijia, Du Yaou, Shen Mingyu, Yan Hongsheng, Yuan Yu, Liu Xiaoli, Wu Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9784888983051

With the proposal of liquid metals as coolant of the reactors, the use of accurate heat transfer correlation and reliable turbulent prandtl number（Prt）model of these low prandtl number fluids in turbulent flow is essential. The heat transfer of Liquid metals differ from traditional coolants for their good molecular heat conductivity and this property leads to a separation of the spatial extension of thermal and viscous boundary layers. A constant turbulent prandtl number is proved to be inappropriate for complex conditions for it depends on the Reynolds and molecular Prandtl number. This paper proposes a new turbulent prandtl number model Prt=C+B/Pet, which can be used for accurately predicting the heat transfer for liquid metal flows. The constants C and B of the expression are obtained from the analytic solution of renormalization group procedure. B is further corrected with the physical properties for different liquid metals. The results are in a good agreement with experimental data over the range of experimentally accessible Prandtl numbers. The Nusselt numbers calculated using the proposed model with the Lyon integral are found to be well consistent with the experimental data for forced fully-developed pipe flows with constant heat flow boundary conditions. Computational fluid dynamics(CFD) analysis is also carried out for pipe geometries. With the proposed model for turbulent prandtl number, CFD codes using RANS approach give similar results for local Nusselt numbers measured in the experiments. Copyright © 2018-2019 by JSME

关键词： Liquid metals

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STUDY ON DYNAMIC RESPONSE OF reactor CORE BARREL UNDER PUMP-INDUCED PULSATING PRESSURE 30

STUDY ON DYNAMIC RESPONSE OF REACTOR CORE BARREL UNDER PUMP-...

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30th International Congress on Sound and Vibration, ICSV 2024

作者： Ye, Xianhui Cai, Fengchun Huang, Xuan Feng, Zhipeng Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9789090390581

The flow-induced vibration (FIV) problem is widely present in nuclear power plants and has caused numerous engineering accidents. Therefore, it has been highly valued by the nuclear engineering community for a long time. The pump-induced pulsating pressure is an important excitation source of FIV. The characteristics of pump-induced pulsating pressure are periodic, it may cause resonance and have the adverse impact on the fatigue life of reactor coolant system. In order to study the vibration of reactor core barrel (CB) of newly designed HPR1000 (Hua-long Pressurized reactor) under pump-induced pulsating pressure, the finite element model of CB is established and the dynamic response characteristics are analyzed. Combined with turbulent excitation, the influence of CB under flow induced load was comprehensively evaluated .The stress analysis shows that the maximum stress of FIV at each position of CB is less than the fatigue stress limit, and the structure is safe. However, for the newly designed reactor or the replacement of the reactor coolant pump, the pump-induced vibration of the CB and reactor internal should be paid enough attention to. © 2024 Proceedings of the International Congress on Sound and Vibration. All rights reserved.

关键词： Stress analysis

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Research on the vibration reduction and sound radiation index properties of the cabin with internal equipment 26

Research on the vibration reduction and sound radiation inde...

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26th International Congress on Sound and Vibration, ICSV 2019

作者： Wenzheng, Zhang Bihao, Wang Xiaozhou, Jiang Bin, Lan Haiyang, Song Fengchun, Cai Xianhui, Ye Jiabin, Feng Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9781999181000

Suitable vibration and noise indices which provide technical direction for vibration and noise reduction, play a very important role in structure analysis and validated test. The contribution of foundation impedance is studied by analyzing the vibration reduction index of a single degree of freedom system. The finite element method (FEM) and boundary element method (BEM) are applied to simulate the vibration and noise indices of a cabin along with internal equipment. The vibration reduction and sound radiation indices of the internal isolator have been calculated to analyze the variation trend with frequency. The results indicate that the difference between vibration level difference and insert loss is very small in most frequency band. Interestingly, the power flow difference and radiated sound power are opposite in phase in middle and high frequency band, which have been proved to be appropriate elements to evaluate the vibration and noise indices of the cabin. The analysis method and procedure can be used to objectively evaluate the vibration reduction property of similar vibration isolators © Proceedings of the 26th International Congress on Sound and Vibration, ICSV 2019. All rights reserved.

关键词： Finite element method

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