The advanced pressurized water reactor with a tight hexagonal lattice has its potential to improve uranium utilization. With a smaller pitch-to-diameter ratio(p/d) ranging from 1.06 to 1.20, a major problem is the eme...
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The advanced pressurized water reactor with a tight hexagonal lattice has its potential to improve uranium utilization. With a smaller pitch-to-diameter ratio(p/d) ranging from 1.06 to 1.20, a major problem is the emergency core cooling in a loss-of-coolant accident. The NEPTUN-III and FLORESTAN flooding experiments at PSI and KfK have been performed during the verification period 1985-1990, and the results shows that the reflooding behavior of LWHCR is quite different from that of LWR with higher peak clad temperatures and smaller quench rates. The existing system codes like RELAP5, CATHARE, TRAC may not be able to be used in the calculations of the reflooding phase of LWHCR. Though some efforts have been put into the validation of the improvement of codes, the results are not so good. In this study, the characteristics of the reflooding behavior of tight lattice are pointed out, and the deficiencies of the reflooding models of RELAP5/MOD3 are also fount out after a simulation of the NEPTUN-III reflooding test. The precursory cooling in tight lattice is not effective especially at low reflooding rates (<4cm/s), the droplet enhancement is not significant compared to LWR, and the vapor Reynolds number is much smaller. Finally, a few modifications to RELAP5/MOD3 are made. Compared with the original code, the modified one can better predict the reflooding phase of tight lattice especially the peak clad temperatures at low reflooding rates. More work should be done to deal with the heat transfer characteristics in the transition region of reflooding in tight lattice..
plate fuel assembly is widely used as the research reactor core geometry due to the narrow rectangular channel benefit, but the research on the post-CHF thermal-hydraulic characteristics in the narrow rectangular chan...
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plate fuel assembly is widely used as the research reactor core geometry due to the narrow rectangular channel benefit, but the research on the post-CHF thermal-hydraulic characteristics in the narrow rectangular channel is few. Once the loss of coolant accident (LOCA) happens the post-CHF thermal-hydraulic characteristics is important for the accident mitigation. This paper studies the characteristics of the inverted annular flow regime during the narrow channel through the comparison between narrow and general channel inside tube heat reflooding experiment, give the assumptions of the post-CHF reflooding flow regime. Then through the comparison of narrow annular channel experimental wall temperature with the RELAP5 simulation wall temperature for this experiment process. The far downstream of the quench front heat flux is obviously lower than the calculated value, so the wall temperature is higher than the calculated wall temperature. The quality is assumed lower during the far downstream of the quench front in the narrow channel experiment. Because the liquid is consumed during the close downstream of the quench front. Through the comparison of the wall temperature between the calculated by code and experimental value, the liquid core in the inverted annular flow regime which may appears during close downstream of the quench front shortly will breakup earlier due to the more severe surrounding steam disturbing, bringing on the quickly consumption of the liquid, then the quality during far downstream of the quench front is higher than general channel.
Direct numerical simulation is a powerful research tool, which provides a level of information and accuracy that cannot be equaled with other approaches. In this paper, the development of a CFD code for direct numeric...
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Direct numerical simulation is a powerful research tool, which provides a level of information and accuracy that cannot be equaled with other approaches. In this paper, the development of a CFD code for direct numerical simulation is described, which is named dnsPisoFoam. In order to validate the code, direct numerical simulation of channel flow with dnsPisoFoam was performed. The present results are compared with existing simulation results and shows good agreement.
Introduce the basic concept of best rational approximation and constructing procedure of the best rational approximation function in Laplace domain. By an illustrative example the design method of FOC based on best ra...
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Introduce the basic concept of best rational approximation and constructing procedure of the best rational approximation function in Laplace domain. By an illustrative example the design method of FOC based on best rational approximation is discussed. The performance comparison of transfer functions between the FOC obtained by best rational approximation method and conventional PID shows that it is effective to apply the method in design of FOC, which can guarantee maximum absolute error of amplitude frequency characteristic of approximation function in a given error tolerance.
This paper describes a calculation model for pre-stressed bolted joints used to connect turntable bearing to the flanges. A traditional pre-stressed bolted joint calculation model is described and its insufficiencies ...
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Based on BPR theory, the new product development and manufacturing process status of a discrete manufacturing enterprise was analysed with finding out its existing problems, then a new reasonable process was designed ...
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In this study, a precise micro press forming system was developed and applied to fabricate micro parts/units. The parts can be formed in several processes and assembled in a unit in the progressive die automatically. ...
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作者：
LI Feng-yu李凤宇LU Gu-bing陆古兵CHEN Zhi陈智College of Naval Architecture&Power
Naval University of Engineering Wuhan 430033 China 海军工程大学 船舶与动力学院湖北 武汉 430033 Science and Technology on Reactor System Design Technology LaboratoryNuclear Power Institute of China Chendu 610041 China 中国核动力研究设计院核反应堆系统设计技术重点实验室四川成都 610041
Without considering the echo signal power changes caused by target distance changes during the accumulated time the traditional radar equation is hardly applied. We take account of the effects of velocity, signal-to-n...
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In this paper, nonlinear dynamic response under loss of coolant accident (LOCA) transient in the nuclear reactor coolant system (RCS) is investigated with ANSYS program. Many nonlinear factors, such as different struc...
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In this paper, nonlinear dynamic response under loss of coolant accident (LOCA) transient in the nuclear reactor coolant system (RCS) is investigated with ANSYS program. Many nonlinear factors, such as different structural stiffness in tension and compression, gap and plastic strain caused by main pipes' break etc. are considered. The Secondary development of ANSYS is performed to form the customized module for implementing effective parameterized and modular modeling and LOCA nonlinear analysis. Comparison of the results calculated by ANSYS and program-specific shows that the results are generally consistent, but there are some differences locally. According to the experience of transient dynamic analysis, and there are so many non-linear factors in the reactor coolant system, the difference is acceptable. The efficiency of this engineering analysis is improved remarkably with the convenience of input and modeling, viewable layout and automatic creating of reports, when using the customized module based on ANSYS.
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