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第二届材料科学应用与能源材料国际研讨会

作者： Jinghua Yang Dengfeng Lu Zhiguang Deng Xiang Guan Qing Shi Sichuan Electric Power Design & Consulting Co. Ltd National Key Labortory of Science and Technology on Reactor System Design Technology

In order to quickly calculate the flow characteristics in the simulated two-dimensional Rayleigh–Bénard convection model,a reduced-order extrapolation difference model for solving two-dimensional Rayleigh–Bénard is established based on the Proper Orthogonal Decomposition(POD) ***,the classical difference format is used to calculate the flow field on the initial time span,the sample data is to construct the POD basis,and the truncation error generated in the process of constructing the POD base is ***,the two-dimensional reduced-order extrapolation Rayleigh–Bénard model is constructed by combining the SVD and POD *** numerical examples,the classical difference model and the reduced-order extrapolation model are numerically simulated under the given initial boundary *** reliability and effective of the reduced-order method are verified by comparing the flow field temperature cloud maps in the two calculation stages.

关键词： POD A Reduced-order Extrapolation Model Based on Proper Orthogonal Decomposition Technique for Rayleigh nard convection Model

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The application of Deep Reinforcement Learning in Coordinated Control of Nuclear reactors

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Journal of Physics: Conference Series 2021年第1期2113卷

作者： Jing Li Yanyang Liu Xianguo Qing Kai Xiao Ying Zhang Pengcheng Yang Yue (Marco) Yang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu 610213 China Department of Engineering The University of Sydney Sydney 2006 Australia

The nuclear reactor control system plays a crucial role in the operation of nuclear power plants. The coordinated control of power control and steam generator level control has become one of the most important control problems in these systems. In this paper, we propose a mathematical model of the coordinated control system, and then transform it into a reinforcement learning model and develop a deep reinforcement learning control algorithm so-called DDPG algorithm to solve the problem. Through simulation experiments, our proposed algorithm has shown an extremely remarkable control performance.

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A method for predicting the configuration of corium pool in RPV lower plenum

A method for predicting the configuration of corium pool in ...

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2018 International Congress on Advances in Nuclear Power Plants, ICAPP 2018

作者： Liu, Lili Yu, Hongxing Chen, Liang Jian, Deng Chunrui, Deng Dan, Zhang Xiaoli, Wu Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9780894487552

The configuration of a corium pool in the lower plenum of reactor vessel is very important for validity of maintaining reactor vessel integrity under IVRERVC (In-Vessel corium Retention through External reactor Vessel Cooling) condition. A method is designed to predict the configuration of corium pool for an Advanced China PWR, call ACP1000 after entire melt dropping into the lower plenum. It takes into account both thermo-chemical reaction of oxides and metals in corium and the influence of different paths of corium relocated from core region into the lower plenum. The reasons why a three-layer pool has not been observed in the MASCA and COSMOS tests are given by this method. The method is applied to investigate the corium configuration of ACP1000 after a hypothetical station blackout (SBO) accident. It is shown that the stratification of the corium pool is related to the relocation paths of the corium. In the case of downward relocation, a two-layer melt pool with a metal layer on the top is formed. For the sideward relocation, the configuration shows a stratified pool consisting of a dense metal layer on the bottom, an oxide layer in the middle and a light metal layer on the top. © 2018 American Nuclear Society. All rights reserved.

关键词： Pressurized water reactors

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Effect of clamping failure on flow induced vibration and fretting wear of fuel rods

Effect of clamping failure on flow induced vibration and fre...

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ASME 2018 Pressure Vessels and Piping Conference, PVP 2018

作者： Huan-Huan, Qi Zhi-Peng, Feng Fu-Rui, Xiong 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)9780791851654

Fuel rods are subjected to both axial and lateral flow in the reactor core. In this study, we present a study on the flow induced vibration (FIV) and fretting wear of fuel rod with failed clamping at grids. First, according to the flow distribution around a type of pressurized water reactor (PWR) fuel rod, the power spectral density (PSD) is obtained to characterize the turbulence excitation. Next, by combining the correlation PSD test parameters, the mean square value of the vibration displacement of each rod mode is found, and then the wear depth of dimple position is calculated based on the ARCHARD wear formula. The clamping of fuel rod at various grids may fail due to inaccurate manufacturing, fuel transportation and in-core irradiation. The absence of clamping force would significantly influence the rod mode and thereby changes its FIV responses. Simulation results show that the failure of the leaf spring has negligible effect on the rod natural frequency whereas the dimple failure near the location with larger FIV amplitude has a much significant effect. The lateral flow velocities at the inlet and outlet of the core are larger. For the fully clamped fuel rod, the response amplitude of turbulent excitation at the bottom and top of the fuel rod is larger. This is even more obvious with a failed dimple at these locations. Comparatively, the effect of dimple support failure in the middle is less influential. The influence of dimple support failure over the rod wear depth depicts basically the same trend as on the maximum FIV amplitude. In addition to the FIV amplitude, wear depth is also related to rod natural frequency. By examining the multiplication of amplitude and frequency at the top and bottom grid, we found that the dimple failure has greater impact at the top grid. Copyright © 2018 ASME.

关键词： Natural frequencies

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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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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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Development of three-dimensional neutron kinetics code based on high order nodal expansion method in hexagonal-Z geometry 26

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2018 26th International Conference on Nuclear Engineering, ICONE 2018

作者： Chao, Guo Yu, Liu Hangxing, He Luguo, Liu Xiaoyu, Wang Sufang, Xin Peiying, Li Xiaoli, Wu Hongsheng, Yuan Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

ISBN: (纸本)9780791851456

To solve three-dimensional kinetics problems, a high order nodal expansion method for hexagonal-z geometry(HONEM) and a Runge-Kutta (RK) method are respectively adopted to deal with the spatial and temporal problem. In the HONEM, 1D partially-integrated flux are approximated by using four order polynomial. The two order polynomial is adopted to the approximation of partially-integrated leakages. The Runge-Kutta method is adopted as a tool for dispersing the time term of 3D kinetics equation. A flux weighting method (FWM) is used for obtaining homogenized cross sections of mix node. The three-dimensional hexagonal kinetics code has been developed based on this method and tested with two benchmark problems of VVER which are the control rod ejection without any feedback and with simple adiabatic Doppler feedback. The results calculated by this code agree well with the reference results and the code is validated. Copyright © 2018 ASME.

关键词： Runge Kutta methods

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Multiple branches coupling property and design of PRHRs

Multiple branches coupling property and design of PRHRs

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2018 Transactions of the American Nuclear Society, ANS 2018 and Embedded Topical Meeting Nuclear Fuels and Structural Materials

作者： Zhuohua, Zhang Xu, Ran Feng, Li Lin, Xian Ke, Zhou Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

Nuclear reactor applied on ship or floating platform recommend miniaturization and optimization of PRHRs. Due to much larger volume and heat transfer area in PRHRs heat exchanger, miniaturization and design of PRHRs need consider many constraints such as volume and systems configuration in containment. In normal reactor design such constraints are not so rigorous, but in reactor design for floating platform it takes significant difficulties. A multiple branches coupling system is constructed in the paper which describes the actual relationship between PRHRs and other systems or equipment, and by combining temperature distribution in heat exchanger and conservation equations of coupling system, the design parameters set satisfied design requirement can be calculated and optimization of design can be completed by programming solving method. Copyright © 2018 AMERICAN NUCLEAR SOCIETY, INCORPORATED, LA GRANGE PARK , ILLINOIS 60526Printed in USA.

关键词： Heat transfer

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Maap5 simulation of pwr high-pressure core melting severe accident due to smallbreak loca

Maap5 simulation of pwr high-pressure core melting severe ac...

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2018 International Congress on Advances in Nuclear Power Plants, ICAPP 2018

作者： Wu, Xiaoli Li, Wei Deng, Jian Deng, Chunrui Hou, Liqiang Zhang, Dan Yuan, Hongsheng Science Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

ISBN: (纸本)9780894487552

MAAP5 is capable of simulating the whole response of a PWR nuclear power plant under severe accident conditions. An analysis of the severe accident scenarios in a PWR by MAAP5 were carried out. It was assumed that the severe accident was initiated by an unusual smallbreak loss-of-coolant accident (SBLOCA) at the reactor pressure vessel (RPV) lower head and the failure of highpressure injection system (HPIS) in a PWR by using MAAP5 were carried out. Different break sizes with and without manual depressurization of the reactor coolant system (RCS) were studied. The results showed that without manually depressurization the RCS, reactor core underwent slow heatup and completely melted and eventually failed the RPV lower head at the primary system pressure of 6.87 MPa for 0.4-inch-break LOCA and 2.71 MPa for 1.0-inch-break LOCA. On the other hand, timely manually opening the pressurizer (PZR) safety valves (SVs) was an effective mitigation measure to recover the core coolabilty by cold-leg accumulator injection system (AIS) and low-pressure injection system (LPIS). Besides, reasonably manually opening the steam generators (SGs) SVs while keeping the auxiliary feedwater work also helped to depressurize the RCS and prevent the severe accident. Both of the two mitigation measures successfully prevented the core from complete melt, but the latter one is preferable to the former one provided no steam generator tube rupture took place. © 2018 American Nuclear Society. All rights reserved.

关键词： Loss of coolant accidents

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Preliminary study on few group cross section representation method based on model tree algorithm

Preliminary study on few group cross section representation ...

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2018 International Conference on Physics of reactors: reactor Physics Paving the Way Towards More Efficient Systems, PHYSOR 2018

作者： Zhao, Wenbo Ju, Haitao Yu, Yingrui Chai, Xiaoming Yu, Junchong Peng, Xingjie Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China

ISBN: (纸本)9781713808510

The tabular representation of few-group cross section is widely used for node wise two-step core simulations. However, it may take too much memory usage for pin-by-pin two-step simulations. A few-group cross section representation method for PWR has been developed to reduce memory usage in this paper. The few-group cross sections are represented as the sum of a base term and a collection of partial terms. Each of these terms is fitted by modified model tree. The case matrix of training data set is similar with that of the conventional methods such as the piece wise interpolation and the tabular representation. Two modifications have been made for the model tree algorithm: 1) burnup is the only variable to split data set, 2) linear regression method on nodes of the tree is replaced with polynomial fitting method according to priori knowledge of the relationship between cross sections and state variables. Through the modifications accuracy and generalization ability are achieved without the pruning procedure. Then the lattice calculations for test data set could be canceled. Tests on the typical PWR lattice are performed. Numerical results show that for the worst case the memory usage of this method is less than 30% of the tabular representation, when the maximum error of reactivity is less than 500 pcm. The memory usage for other cross sections is less than 10% with high accuracy. Optimization of the target tolerance criteria should be tested for fast group cross sections. Large errors in thermal group absorption and fission production cross sections lead to large errors in Kinf and it needs more study on regression model. The method combined with the tabular representation might be more suitable for the pin-by-pin two-step simulations. © 2018 International Conference on Physics of reactors, PHYSOR 2018: reactor Physics Paving the Way Towards More Efficient Systems. All rights reserved.

关键词： Errors

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