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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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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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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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Multiple branches coupling property and design of PRHRs

Multiple branches coupling property and design of PRHRs

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AISTech 2018 Iron and Steel technology Conference and Exposition

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

ISBN: (纸本)9781935117728

关键词： Heat transfer

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Research on the response of core barrel under random turbulence excitation 25

Research on the response of core barrel under random turbule...

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

作者： Huang, Xuan Wu, Wan-Jun Qi, Huan-Huan Feng, Zhi-Peng Liu, Shuai Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu Sichuan China

ISBN: (纸本)9781510868458

The flow-induced vibration of reactor structures is directly related to nuclear safety and has long been closely watched by the nuclear engineering community. A large number of theoretical, experimental and numerical analysis research work are carried out at home and abroad. In one hand, the test means cost a lot of money and time;on the other hand due to the restriction of test conditions, we only can get the measurement results about part of the structure under specific conditions. Therefore, from the design point of view, the theoretical or numerical analysis methods should be developed to predict structural flow induced vibration response in order to compensate for the lack of test methods. For the purpose of completing the flow-induced vibration analysis of the components in the reactor, it is necessary to analyse the vibration characteristics of the components in the reactor. According to the fluid excitation situation, this paper based on the test study two kinds of core barrel calculation method of forced vibration response under turbulent excitation. The advantages and disadvantages of these two methods are described from the aspects of calculation model, calculation input, analysis method and calculation result. © 25th International Congress on Sound and Vibration 2018, ICSV 2018: Hiroshima Calling. All rights reserved.

关键词： Vibration analysis

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The implicit runge kutta method for solving point burnup equations

The implicit runge kutta method for solving point burnup equ...

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2018 Transactions of the American Nuclear Society, ANS 2018

作者： Cai, Yun Peng, Xingjie Li, Qing Qin, Dong Jiang, Zhumin Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China

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Study on melting behavior and related influential factors of shroud and basket of acp1000 during a severe accident

Study on melting behavior and related influential factors of...

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

作者： Hou, Li-Qiang Zhu, Da-Huan Wu, Qing Deng, Jian Wu, Xiao-Li Science and Technology on Reactor System Design Technology Nuclear Power Institute of China Chengdu of Sichuan Prov 610213 China

ISBN: (纸本)9780894487552

The focusing effect of the metal molten pool plays an important role in elevating the validity of the In-Vessel Retention (IVR) during a severe accident while the melting of the shroud and basket can contribute to the formation of the metal molten pool. Therefore, the study on the melting behavior of the shroud and basket can offer technical support for the validity analysis of IVR. The method of computational fluid dynamics (CFD) has been used to establish a two-dimensional calculation model of the 1/8 core of ACP1000 to study the melting behavior of the shroud and basket during the severe accident of large LOCA. The core has been divided into ten nodes in the axial direction and the radiation model and the solidification and melting model have been considered during the calculation. The results show that the shroud at six out of ten nodes and the basket at five out of ten nodes have melted totally before the first significant migration of the core and that the basket always starts to melt after the shroud has melted totally at the same node. © 2018 American Nuclear Society. All rights reserved.

关键词： Melting

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Optimization design of the reactor coolant pump flywheel comprised of multi-ring packing heavy metal alloy

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Journal of Physics: Conference Series 2021年第1期1777卷

作者： L Jiang C W Wu X J Shao Y Zhang M D Yu National Key Laboratory of Science and Technology on Reactor System Design Technology Nuclear Power Institute of China Chengdu 610041 China State Key Laboratory of Structural Analysis for Industrial Equipment Dalian University of Technology Dalian 116024 China

To improve energy storage performance of the multi-ring RCP flywheel comprised of inner hub, tungsten alloy ring and outer retainer, optimization design process for the radial thicknesses of components and interference fit assembly is demonstrated. Radial thicknesses of these components and magnitude of interference fit are set as design parameters to construct parametric analysis model. design of experiments (DOE) technique is adopted to generate design matrix for sensitive analysis. Subsequently, radial basis function (RBF) approximation is employed to search for the optimal design considering structural integrity requirement. Finally, a proper range of the magnitude of interference fit is obtained which can not only prevent neither slippage nor separation between the inner hub and tungsten alloy ring but also ensure the safety of structural strength. In addition, it is found that to ensure a high moment of inertia and energy density, a reasonable large radial thickness of tungsten alloy ring and a reasonable small radial thickness of outer retainer are required.

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Strength-based concurrent design for structural topology and building orientation of additively manufactured lattice structure

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Chinese Journal of Aeronautics 2024年

作者： Chenyang LI Yamin LI Shangqin YUAN Shaoying LI Jihong ZHU Weihong ZHANG State IJR Center of Aerospace Design and Additive Manufacturing School of Mechanical Engineering Northwestern Polytechnical University Xi’an 710072 China MIIT Lab of Metal Additive Manufacturing and Innovative Design Northwestern Polytechnical University Xi’an 710072 China Research & Development Institute of Northwestern Polytechnical University in Shenzhen Sanhang Science & Technology Building Shenzhen 518063 China Unmanned System Research Institute Northwestern Polytechnical University Xi’an 710072 China Modern Control Technology Research Institute Xi’an 710065China

A novel additive manufacturing (AM)-driven concurrent design strategy based on the beam characterization model considering strength constraints is proposed. The lattice topology, radius size, building orientation (BO), and structural yield strength can be simultaneously adjusted by integrating the overall process-structure-performance relationship of the AM process into the optimization. Specifically, the transverse isotropic material model is adopted to describe the material properties induced by the layer-by-layer manner of additive manufacturing. To bolster lattice strength performance, the stress constraints and ratio constraints of lattice struts are employed. The Tsai-Wu yield criterion is implemented to characterize the lattice strut’s strength, while the P-norm method streamlines the handling of multiple constraints, minimizing computational overhead. Moreover, the gradient-based optimization model is established, where both the individual struts diameters and BO can be designed, and the buckling-prone spatial struts are strategically eliminated to improve the lattice strength further. Furthermore, several typical structures are optimized to verify the effectiveness of the proposed method. The optimized results are quite encouraging since the heterogeneous lattice structures with optimized BO obtained by the strength-based concurrent method show a remarkably improved performance compared to traditional designs.

关键词： Strength-based design AM-driven design Concurrent design Building orientation Additive manufacturing

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