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15th IEEE International Conference on Mechatronics and Automation, ICMA 2018

作者： Ni, Fenglei Li, Yongqiang Zhou, Yunhu Liu, Hong Tang, Yuan State Key Laboratory of Robotics and System Harbin Institute of Technology Harbin Heilongjiang Province China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

ISBN: (纸本)9781538660720

Self-reconfigurable modular robots have shown great potential in carrying out multiple tasks in different conditions. A connection of both flexibility and reliability is critical for realizing the docking and transformation of the modular robots. We propose a new design of connectors based on SMA which has the advantages of high autonomy, small size and high speed. The couples of connectors consist of active and passive connector. In the active connector, a two-way rotary SMA actuator is used as the actuation mechanism. The double-layer cam and claws are used as an executive mechanism. The passive connector can be securely docked in five different facets for 3D reconfigurations. The multisensory system and hardware architecture provide the foundation for control strategies. The design and the prototype of the connector are described in this paper. Finally, experimental results to evaluate the performance of docking are presented. © 2018 IEEE.

关键词： Modular robots

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Study on optimization design for CSR1000 core

Study on optimization design for CSR1000 core

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作者： Wang, Lianjie Yang, Ping Lu, Di Zhao, Wenbo Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China No. 328 Section 1 Changshun Avenue Shuangliu County Chengdu610213 China

An optimization design of China supercritical water-cooled reactor (SCWR) with the rated electric power of 1000 MWe (CSR1000) conceptual core is proposed. Steady-state performance of the proposed core is then studied with the SCWR core steady-state analysis code system SNTA. These key parameters such as burnup performance, reactivity control capability, power distribution, maximum fuel cladding temperature, and maximum linear power density are analyzed. The relative coolant flow rate of the second flow path, which is suited with assembly power, is also presented. The study shows that the refueling cycle of CSR1000 core can be extended effectively under the optimization design. Copyright © 2018 by ASME.

关键词： Water cooled reactors

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RMC/CTF multiphysics solutions to VERA benchmark problem 9

RMC/CTF multiphysics solutions to VERA benchmark problem 9

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2018 Transactions of the American Nuclear Society, ANS 2018

作者： Ma, Yugao Liu, Shichang Luo, Zhen Wang, Kan Yu, Ganglin Huang, Shanfang Department of Engineering Physics Tsinghua University Beijing100084 China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610213 China

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Methods and Verification of 3D direct reactor calculation code KYCORE

Methods and Verification of 3D direct reactor calculation co...

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

作者： Tang, Xiao Li, Qing Chai, Xiaoming Tu, Xiaolan Peng, Xingjie Wang, Kan Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China Department of Engineering Physics Tsinghua University Beijing 100084 China

ISBN: (纸本)9781713808510

KYCORE is a 3D direct reactor calculation code capable of performing reactor hetero geneous calculation without cross section homogenization and group condense. It is developed in Nuclear Power Institute of China, involving radial characteristics method (MOC) and axial discrete ordinate method (SN) coupled 3D neutron transport and subgroup method with fixed source problem for resonance treatment and depletion, aiming at providing mini core design and commercial reactor benchmark solution. This article briefly introduces optimized MOC/SN coupling, adjusted coarse mesh finite difference (CMFD) and resonance treatment. The accuracy of the code is verified with 3D VERA Problem 3 benchmark and proves to be credible. © 2018 International Conference on Physics of reactors, PHYSOR 2018: reactor Physics Paving the Way Towards More Efficient systems. All rights reserved.

关键词： Codes (symbols)

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Lattice Boltzmann method for simulation of time-dependent neutral particle transport

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Nuclear science and Techniques 2017年第3期28卷 74-84页

作者： Ya-Hui Wang Li-Ming Yan Bang-Yang Xia Yu Ma School of Energy Science and Engineering Harbin Institute of Technology Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China

In this paper, a novel model is proposed to investigate the neutron transport in scattering and absorbing medium. This solution to the linear Boltzmann equation is expanded from the idea of lattice Boltzmann method(LBM) with the collision and streaming process. The theoretical derivation of lattice Boltzmann model for transient neutron transport problem is proposed for the first *** fully implicit backward difference scheme is used to ensure the numerical stability, and relaxation time and equilibrium particle distribution function are obtained. To validate the new lattice Boltzmann model, the LBM formulation is tested for a homogenous media with different sources, and both transient and steady-state LBM results get a good agreement with the benchmark solutions.

关键词： Transient neutron transport Lattice Boltzmann method Linear Boltzmann equation

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Development of a sensitivity analysis code for 3-D C5G7 benchmarks based on the pertubation theary

Development of a sensitivity analysis code for 3-D C5G7 benc...

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

作者： Wu, Qu Peng, Xingjie Yu, Yingrui Li, Qing Wang, Kan Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China Department of Engineering Physics Tsinghua University Beijing100084 China

ISBN: (纸本)9781713808510

Sensitivity and Uncertainty (S&U) analysis plays a major role in reactor physics. Uncertainty in reactor engineering comes from model error, numerical error and input parameter error. Nuclear data are regarded as crucial input parameters and proven to be a major uncertainty source. Therefore, effective evaluation of nuclear data can promote the accuracy and fidelity of nuclear system simulation. In deterministic codes, however, most studies about S&U analysis are based on the traditional two-step method, i.e. assembly calculation and core simulation. With the development of advanced 2-D/l-D transport codes, the corresponding S&U analysis model is required. A new S&U analysis code, namely KYADJ, achieved the sensitivity analysis of eigenvalue with respect to material macroscopic cross sections based on classical perturbation theory. Forward and adjoint flux were solved by the coupling of 2-D Method of Characteristics (MOC) and 1-D discrete ordinates method (Sn). Besides, the Direct Numerical Perturbation (DNP) method in KYADJ and the Iterated Fission Probability (UP) method in reactor Monte Carlo code (RMC) were adopted to calculate sensitivity coefficients for comparison. A 3x3 pin-cell benchmark and 3-D OECD/NEA C5G7 benchmarks were applied to verify the ability of sensitivity analysis in KYADJ. Results showed that a good agreement among three approaches was achieved. KYADJ can accurately calculate eigenvalue sensitivity coefficients with respect to input nuclear data. © 2018 International Conference on Physics of reactors, PHYSOR 2018: reactor Physics Paving the Way Towards More Efficient systems. All rights reserved.

关键词： Sensitivity analysis

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Numerical simulation research on the performance of SCWR fuel rod

Numerical simulation research on the performance of SCWR fue...

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作者： Tang, Changbing Xing, Shuo Pang, Hua Chen, Ping Zhou, Yi Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China No. 328 First Section of Changshun Avenue Shuangliu District Sichuan610041 China

Because of the high temperature and high pressure characteristics of supercritical water-cooled reactor (SCWR), the thermal hydraulic performance of SCWR is greatly different from pressurized water reactor (PWR), which makes the current PWR fuel rod performance analysis codes are no longer applicable to SCWR. In this research, the irradiation swelling, irradiation densification, thermal expansion, thermal creep, plastic deformation, irradiation creep and irradiation hardening of UO2 pellet, and stainless steel cladding were considered;the gas conductance and radiant conductance of gap heat transfer were considered, the forced convective heat transfer on the outer surface of cladding was considered. Meanwhile, the irradiation effects and the thermal effects on the materials parameters such as thermal conductivity, specific heat, and young’s modulus were also considered in this research. With the help of ABAQUS software, the related user-defined subroutines were developed, and the irradiation effects and thermal effects of SCWR fuel were introduced into the numerical simulation, and then completed the analysis of SCWR fuel rods’ performance under steady power conditions. Some reference suggestions for the design and development of SCWR fuel could be provided by the establishment of this numerical simulation method. Copyright © 2018 by ASME.

关键词： Pressurized water reactors

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Calculation of adjoint-weighted reactor kinetics parameters in OpenMC

Calculation of adjoint-weighted reactor kinetics parameters ...

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

作者： Peng, Xingjie Forget, Benoit Smith, Kord Science and Technology on Reactor System Design Technology Laboratory ChengduSichuan610000 China Nuclear Science and Engineering Department Massachusetts Institute of Technology 77 Massachusetts Ave. CambridgeMA02139 United States

ISBN: (纸本)9781713808510

Knowledge of adjoint-weighted kinetics parameters plays an importance role in analyzing the reactor safety and in describing the transient behavior of nuclear reactors during normal operation or accidents. In this work, the capability of calculating adjoint-weighted kinetic parameters, including the effective neutron generation time, the effective delayed neutron fraction, and the ratio of these two quantities was implemented in OpenMC code based on two different methods of interpreting the physical meaning of the adjoint weighting, namely, the iterated fission probability (IFP) method and the Contribution-Linked eigenvalue sensitivity/Uncertainty estimation via Tracklength importance CHaracterization (CLUTCH) method. The results of the implementation were benchmarked by comparison to MCNP5 calculations, Serpent2 calculations and experimental data in 10 critical configurations. © 2018 International Conference on Physics of reactors, PHYSOR 2018: reactor Physics Paving the Way Towards More Efficient systems. All rights reserved.

关键词： Eigenvalues and eigenfunctions

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Performance comparison of different approximation schemes in matrix MOC

Performance comparison of different approximation schemes in...

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

作者： Yong, Zheng Peng, Minjun Zhou, Yueshan Qiang, Shenglong Ping, An Wei, Lu Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu610041 China Fundamental Science on Nuclear Safety and Simulation Technology Laboratory Harbin Engineering University Harbin150001 China

ISBN: (纸本)9781713808510

The method of characteristics (MOC) with matrix form has theoretically more favorable features than its standard application, since the time-consuming track sweeping operations are replaced with explicit solution of matrix equation. In traditional MOC, there are three optional discretization strategies: the step characteristic method, diamond difference scheme and linear source approximation. However, their performances in matrix MOC have not be reported so far. The current paper develops their matrix forms and solves the resulted linear systems using preconditioned generalized minimal residual (PGM-RES) method. Additionally, their performances with respect of computation accuracy and efficiency are compared with each other, by simulating a homogeneous hexagonal cell problem and the 2D C5G7 benchmark. The results show that their performances in matrix form are different from that of traditional form. With the same angular quadrature and region partition scheme, the step characteristic method is more efficient significantly than the diamond difference scheme and linear source approximation. Whether the diamond difference scheme and linear source approximation can improve the calculation accuracy or not depends on whether the the spatial mesh is fine enough for every scheme. For the coarser spatial partition strategy, diamond difference scheme and linear source approximation can significantly improve the accuracy of local pin power. © 2018 International Conference on Physics of reactors, PHYSOR 2018: reactor Physics Paving the Way Towards More Efficient systems. All rights reserved.

关键词： Diamonds

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A computer code for thermal-hydraulic analysis of steam generator with axial preheater/economizer

A computer code for thermal-hydraulic analysis of steam gene...

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AISTech 2018 Iron and Steel technology Conference and Exposition

作者： Wu, Ge He, Gening Wang, Haisong Nuclear Power Institute of China Science and Technology on Reactor System Design Technology Laboratory NO.328 First Sec. of Changshun Ave. Shuangliu District Chengdu China

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