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2019 Transactions of the American Nuclear Society, ANS 2019

作者： Wang, Donghui Zeng, Qingna Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China 328# 1st section Chuangshun Dadao Avenue Chengdu610213 China

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Evaluation and Analysis of RT Suspected Defects on reactor Pressure Boundary Weld

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IOP Conference Series: Earth and Environmental science 2021年第3期632卷

作者： Bo Yang Qiang Deng Cong Chen FengShou Zhang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institution of China Sichuan ChengDu China

The control rod driving mechanism (CRDM) is one of the important equipments in the nuclear reactor, and its sealing shell is part of the pressure boundary of the nuclear reactor. The sealing shell is a pressure vessel made of two different material (austenitic stainless steel and nickel base alloy) cylinders welded by dissimilar metal welds. Product production process, in this dissimilar metal weld ray detection found suspected defects show. In this paper, the mechanism and process of the suspected defects are expounded through experimental verification, and it is proved that the suspected defects do not affect the quality of the weld.

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Limit load estimation for finite plates with semi-elliptical defects under tensile loading

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

作者： K K Shi X L Fu H Xie X J Shao J Tian Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu Sichuan China

The ASME code allows the designer to circumvent many of the problems arising in design by elastic analysis by performing plastic or limit analysis of the component. Unlike elastic analysis, these analyses take account of the redistribution of stress upon yield. In limit analysis, the calculation of the limit load for finite plate with different semi-elliptical defects is an important stage for mechanical safety evaluation. The elastic compensation technique proposed by Mackenzie can be used to provide the limit load of a structure with a finite element method and aims at producing a suitable stress filed for the lower bound theory. In the present study, it is assumed that the finite plate exists three kinds of semi-elliptical cracks with low aspect ratio a/c (a/c=0.1, 0.5 and 1.0), respectively. The improved elastic compensation technique realized by finite element method is applied in the limit load estimation. At last, the results show that the improved elastic compensation technique can be used to obtain the limit value for the finite plate with different semi-elliptical defects under tensile loading. Moreover, the improved elastic compensation technique can be easily implemented in any finite element method and enables the limit load for any complex structure to be obtained swiftly.

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Comparisons of Subgrid-Scale Models for OpenFoam Large- Eddy Simulation

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Journal of Physics: Conference Series 2021年第4期1802卷

作者： Zhipeng Feng Huanhuan Qi Xuan Huang Shuai Liu Jian Liu Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu Sichuan China

OpenFOAM is a free, open-source software package that can be used for the solutions of computational fluid dynamics and simulation of various fluid flow processes. Nevertheless, OpenFOAM still lacks default settings and a large number of different numerical schemes and turbulent models should be validated. In this paper, the unsteady flow around a cylinder (Re=3900) is calculated by the large eddy simulation of OpenFOAM. The predictions include the drag and lift coefficient, the pressure distribution around the cylinder, the velocity distribution and Reynolds stress distribution in the wake region, as well as the prediction of the recirculation length and separation angle. Thanks to several simulations, these five subgrid-scale (SGS) models are compared and studied: The Smagorinsky SGS model, wall adaptive local eddy visibility SGS model, dynamic SGS model with Lagrangian averaging, dynamic one equation eddy visibility model, one equation eddy visibility model. The numerical results are verified with the published experimental data.

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Research on Optimization of Key Parts of Nuclear reactor Based on Finite Element Simulation

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IOP Conference Series: Earth and Environmental science 2021年第3期632卷

作者： Bo Yang Tianda Yu Qiang Deng Cong Chen Fengshou Zhang Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institution of China Sichuan ChengDu 610041 China

In view of the higher technical requirements of the third generation PWR for the control rod drive mechanism (CRDM), China's third generation nuclear reactor CRDM hook parts are designed to multi-tooth hook with wear-resistant surfacing. The surfacing welding process is simulated and optimized based on finite element simulation, and reasonable welding process and parameters are determined by simulation results in this *** to the optimized welding process, the parts are trial-produced, and the parts of the hook are subjected to metallographic inspection, hardness test and thermal life test. The results show that the optimized hook parts have high hardness and wear resistance, and meet the requirements of the third generation PWR nuclear power plant.

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Numerical analyses and experimental verification on multiphase flow measurements using a gamma-ray densitometer with blind-tee mixing 27

Numerical analyses and experimental verification on multipha...

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

作者： Ma, Yugao Cui, Yinghuan Hao, Yisheng Huang, Shanfang Pan, Yanzhi Yu, Hongxing Department of Engineering Physics Tsinghua University Beijing China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China Xi'an Jiaotong University Xi'an China

ISBN: (纸本)9784888983051

Gas-liquid two-phase flow widely exists in nuclear industry, where the heat flux is quite high. Accurate predictions of key parameters such as void fraction and gas volume fraction are important to the insurance of reactor safety operation. However, the accuracy of most multiphase flowmeters is flow-pattern dependent. A common way to homogenize the gas-liquid mixture and to eliminate the flow pattern effects is to install a blind-tee upstream of the flow measurement section to transform the horizontal flow to vertical upward flow. It is therefore of great interest to verify the effectiveness of flow mixing caused by blind-tee via experiments or simulation. In this study, the three-dimensional transient numerical simulation on gas-liquid two-phase flow was carried out by a commercial CFD software Fluent, using VOF model as multiphase model and Realizable k-Ε model as turbulence model. The results show that the flow develops from centrifugal flow to homogenous flow after passing around the bend and recovers its intermittent nature further downstream. CFD simulation analysis shows that there is a relatively ideal mixing effect after about 7 pipe diameters downstream. Therefore, the complexity of horizontal flow and intermittent flow are avoided by means of device design and measurement position optimization, which simplifies the measurement and calculation model. A flow-pattern based flow rate model is established by introducing the homogeneous-degree evaluation factor. The formula is simple and yet accurate. The relative average error of the model for predicting the total mass flow rate is within 2.66%. Copyright © 2018-2019 by JSME

关键词： Flowmeters

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Effects of heaving condition on characteristics of system under forced circulation and natural circulation 27

Effects of heaving condition on characteristics of system un...

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

作者： Cai, Rong Mi, Zhengpeng Yue, Nina Fang, Hongyu Qiu, Suizheng Wang, Kai Qiu, Zhifang Science and Technology on Reactor System Design Technology Laboratory Chengdu China CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology Chengdu China Xi'an Jiaotong University Xi'an Shaanxi China Department of Nuclear Engineering and Management School of Engineering University of Tokyo Tokyo Japan

ISBN: (纸本)9784888983051

Under ocean conditions, the nuclear power plants have complex movements due to the effects of wind and wave. The thermal-hydraulic characteristics of the nuclear power plants under ocean conditions are influenced by its various motions. To ensure the safety of the reactor, the research on the effects of various motions on its thermal-hydraulic characteristics are necessary. The previous research has shown that heaving condition has more serious effects on the thermal-hydraulic characteristics of the nuclear power plants. In this paper, a thermal-hydraulic system code for ocean conditions named STAC was developed, and the STAC code was validated by test data. The effects of heaving conditions on the thermal-hydraulic characteristics of the system under forced circulation and natural circulation are studied with the STAC code. The simulation results show that the thermal parameters of the reactor system under heaving condition fluctuate periodically, and the fluctuation period is the same as the heaving period. The delay of the variation of circuit flow and the thermal inertia of the fuel material cause a phase difference between the fluctuations of the thermal parameters of reactor core. Heaving motion has less effect on the nuclear power system under forced circulation than that under natural circulation. As the amplitude of the acceleration increases, the fluctuation amplitudes of thermal-hydraulic parameters in the nuclear power system increase. Under heaving condition with short period, the fluctuation amplitudes of thermal-hydraulic parameters are affected by the heaving period. With the increase of the fluctuation period, the fluctuation amplitudes of thermal-hydraulic parameters increase gradually, and finally approach certain values. Copyright © 2018-2019 by JSME

关键词： Nuclear power plants

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Structural design and strength analysis of multi-ring packed flywheel with heavy metal alloy segments 2

Structural design and strength analysis of multi-ring packed...

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2nd International Conference on Numerical Modelling in Engineering, NME 2019

作者： Jiang, L. Zhang, W. Wu, C.W. Zhang, L.P. Zhang, Y.X. Liu, Z.Y. National Key Laboratory of Science and Technology on Reactor System Design Technology Nuclear Power Institute of China Chengdu610041 China State Key Laboratory of Structural Analysis for Industrial Equipment Dalian University of Technology Dalian116024 China

Flywheel, as a kinetic energy storage and conversion device, has promising applications due to its high efficiency, no pollution and long lifetime. In particular, with a three-layer structure packed with heavy metal alloy middle layer, a flywheel can achieve both a large rotational inertia and a high energy density. To guarantee structural strength safety of this type of multi-ring packed flywheel operating at high temperature, structural and thermal stress analysis for a flywheel with a tungsten alloy layer sandwiched between two stainless steel layers is carried out using finite element method (FEM). Based on plane stress assumption, flywheel stress characteristic along radius under assembly force, centrifugal force and thermal loading is demonstrated. The results show that, the rise of temperature causes inconsistent thermal expansion between neighbouring layers and results in a large thermal stress. Particularly, a high circumferential tensile stress occurs in the middle layer. It is also found that, the segments of the middle tungsten alloy layer can efficiently relieve the inconsistent thermal expansion and lower the thermal stress. Moreover, the number of segments significantly influences the efficiency of thermal stress releasing and the reserved gap between adjacent segments significantly improves the contact pressure between inner hub and tungsten alloy layer. This work may provide reference for structural design of the multi-ring packed flywheels under thermal loading. © Published under licence by IOP Publishing Ltd.

关键词： Kinetics

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Effect of finned structure on critical heat flux (CHF) in downward-face pool boiling 27

Effect of finned structure on critical heat flux (CHF) in do...

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

作者： Wang, Kai Erkan, Nejdet Cai, Rong Okamoto, Koji Department of Nuclear Engineering and Management School of Engineering University of Tokyo Tokyo Japan Science and Technology on Reactor System Design Technology Laboratory Chengdu China Nuclear Professional School School of Engineering University of Tokyo Tokyo Japan

ISBN: (纸本)9784888983051

CHF plays an important role in the heat transfer process during the external reactor vessel cooling (IVR-ERVC) which is deemed as a very effective way to release the decay heat after the core melted. In this paper, the experiment of the macro-fin structure experiment positioned on slope that is declined at angles of 5°and 10°was done. The results of the copper bar experiment were then analyzed and compared with the results obtained from a copper block experiment that had been conducted previously at the same test facility, and it was found that the CHF decreased. Finally, the hot/dry spot theory for the copper bar was brought up in an attempt to explain the difference of copper and copper bar. It seemed that using this theory can explain the decrease of CHF for copper bar. Copyright © 2018-2019 by JSME

关键词： Heat flux

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Boron concentration field measurement and mixing analysis in the annular downcomer of reactor pressure vessel 27

Boron concentration field measurement and mixing analysis in...

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

作者： Zhang, Qi Tan, Sichao Zhao, Hangbin Wang, Peng Gu, Hanyang Yang, Yunjia School of Nuclear Science and Engineering Shanghai Jiao Tong University Shanghai China Fundamental Science on Nuclear Safety and Simulation Technology Laboratory Harbin Engineer University Harbin Heilongjiang China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9784888983051

In nuclear power plant, the reactor annular downcomer acts as the upstream of the reactor core, and boric distribution inside the downcomer is focused in the ECC situation. The relevant study indicates that the inlet coolant does not flow along annular channel directly, and the current multi primary loops layout makes the fluid mixing of the downcomer far more complex. In this paper, boron transport and mixing phenomenon inside the reactor annular downcomer were analyzed by Laser Induced Fluorescence (LIF) method, a reduziertreduced RPV model was built with transparent material, the boric acid was equivalently simulated by rhodamine B. The single loop and double loops flow mixing tests were respectively performed, the qualitative and quantitative data at cross sections were acquired. For the single loop test, the injected tracer transported along the circumferential position and was constantly diluted by the ambient pure water;meanwhile, a portion of the injection entered the lower space and filled up the annular channel in the same manner. From the time evolution of concentration gradient distribution, the increasement of inlet velocity is favorable to the uniform mixing. For the double loops fluid mixing tests, the tracer solution and pure water were injected into the model simultaneously through two independent entrances. The location of the intersection boundary of the two fluids was affected by the entrance momentum ratio, and the tracer distribution fluctuated remarkably in the lower space of the downcomer. Copyright © 2018-2019 by JSME

关键词： Boric acid

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