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AbstractAbstract
[en] In the design of Qinshan II nuclear power plant, Fluid-Induced Vibration of the reactor internals has been studied with the combination method of experimental and theoretical analysis. For the theoretical analysis of core barrel, a special methodology has been adopted. This paper describes the analysis process of this method, deduces the theoretical basis, puts forward the restriction conditions for usage, and studies the applicability in Qinshan II. It is considered that this method has definite theoretical foundation and it could be used in the engineering design if the restriction conditions are satisfied. The engineering practice of Qinshan II may have some limitation in satisfying the restriction conditions simultaneously, especially for the characteristics of excitation spectrum, damping, and nonlinear effects. Farther theoretical and engineer application study should be devoted in this field
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Journal Article
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Nuclear Power Engineering; ISSN 0258-0926; ; v. 25(3); p. 198-202
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AbstractAbstract
[en] The hypothetical loss-of-coolant accident (LOCA) in the primary piping is assumed in the design of Qinshan II reactor system. This accident causes transient pressure variation of the coolant and induces dynamic hydraulic loads on reactor components. The author describes the non-linear dynamic analysis procedure of reactor system including the engineering method for nonlinear factors, the establishment of the nonlinear dynamic model and the nonlinear dynamic analysis method. The generation and the application of the LOCA dynamic response for the system are also presented
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Journal Article
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Nuclear Power Engineering; ISSN 0258-0926; ; v. 20(4); p. 342-347
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AbstractAbstract
[en] The author presents the stress analysis and evaluation of pipe layout and support adjustment process for Qinshan phase II pressurizer discharge system. Using PDL-SYSPIPE INTERFACE software, the characteristic parameters of the system are gained from 3-D CAD engineering design software PDL and outputted as the input date file format of special pipe stress analysis program SYSPIPE. Based on that, SYSPIPE program fast stress analysis function is applied in adjusting pipe layout , support layout and support types. According to RCC-M standard, the pipe stress analysis and evaluation under deadweight, internal pressure, thermal expansion, seismic, pipe rupture and discharge loads are fulfilled
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Journal Article
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Nuclear Power Engineering; ISSN 0258-0926; ; v. 21(2); p. 117-120
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Mao Qing; Zhang Jinghui; Luo Yushan; Wang Haijun; Duan Quan
Proceedings of 18th international conference on structural mechanics in reactor technology2005
Proceedings of 18th international conference on structural mechanics in reactor technology2005
AbstractAbstract
[en] This paper presents the hydraulic test for studying the orifice induced pressure fluctuation and vibration in pipeline. Based on the preliminary test results, which has been presented in paper ICONE13-50931, modification of experimental facility has been made. With more test conditions, the statistical characteristics of the fluctuating pressure and structure acceleration response have been studied. The latest test results confirm that these modifications effectively solved the previous problems. The natural frequencies and the strain response of the structure are also obtained to provide comparison to the numerical mathematical model of the pipe and the response calculation in the near future. (authors)
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International Association for Structural Mechanics in Reactor Technology (United States); Chinese Nuclear Society, Beijing (China); Chinese Socity of Theoretical and Applied Mechanics, Beijing (China); Tsinghua Univ., Beijing (China); 4896 p; ISBN 7-5022-3421-7; ; Jul 2005; p. 2463-2469; 18. international conference on structural mechanics in reactor technology; Beijing (China); 7-12 Aug 2005; 9 figs., 6 refs.
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Book
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Conference
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Mao Qing; Yang Yu; Zang Fenggang; Zeng Zhongxiu; Xiao Zhong
The 13th pacific basin nuclear conference. Abstracts2002
The 13th pacific basin nuclear conference. Abstracts2002
AbstractAbstract
No abstract available
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Source
Chinese Nuclear Society, Beijing (China); China National Nuclear Corporation, Beijing (China); China Guangdong Nuclear Power Holding Co., Ltd., Shenzhen (China); State Power Corporation of China, Beijing (China); 347 p; ISBN 7-5022-2682-6; ; 2002; p. 111; 13. pacific basin nuclear conference; Shenzhen (China); 21-25 Oct 2002
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Book
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Conference
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AbstractAbstract
[en] In this paper,based on the experimental results of power spectrum density of wall pressure fluctuations, the response of pipe vibration caused by orifice plate was calculated with random vibration analysis function of ANSYS code. The effect of correlation between wall pressure fluctuations on the response of pipe vibration was discussed. Furthermore, a simplified method was given, and the calculation result was compared with that of the detailed calculation. It was shown that the simplified method was convenient and efficient. The response of the pipe flow-induced vibration calculated with simplified method was conservative, so it can be used in the engineering evaluation. (authors)
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Source
8 figs., 3 tabs., 7 refs.
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Journal Article
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Nuclear Power Engineering; ISSN 0258-0926; ; v. 30(3); p. 22-26
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AbstractAbstract
[en] The present work is to investigate the three-dimensional non-isothermal turbulent jet into crossflow in a thick-wall T-junction pipe, which is commonly subject to the greatest thermal stress in the pressurized water reactor (PWR) cooling system. Two cases with low jet-to-crossflow velocity ratios of 0.05 and 0. 5 are computed, with a finite-volume numerical procedure utilizing κ-ε turbulent model. Temperature of the pipe is acquired by thermally coupling with the fluid. Comparison of the computations with measured data shows good qualitative agreement. Via analysis of the flow and thermal characteristics, influence of the flow structure on the temperature distribution and thermal stress of the component is studied. Major factors causing instantaneous thermal shock of the component are explored. Optimal flow rates are discussed to reduce the thermal stress
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Journal Article
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Nuclear Power Engineering; ISSN 0258-0926; ; v. 22(2); p. 127-132
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AbstractAbstract
[en] Numerical simulations were performed with the commercial computational fluid dynamics (CFD) package FLUENT 5.3 to investigate the thermal-hydraulic phenomena of thermal shock, which is caused by non-isothermal turbulent jet into crossflow in a T-junction with thermal sleeve in the pressurized water reactor (PWR) cooling systems. In allusion to the thermal sleeve configuration with vent holes and lower collar, two typical cases with jet-to-mainstream velocity ratios of 0.05 and 0.5 were computed. Experimental studies were carried out to determine the heat transfer characteristics for the main pipe and the annulus between the nozzle and the thermal sleeve. The calculations well matches the experimental data. The results indicated that the protective action of the thermal sleeve against thermal shock loading is dependent on both the sleeve geometry and the velocity ratio, obtaining improvement with appropriate lower velocity ratios. In addition, optimal flow rates and partial sizes of the thermal sleeve were discussed to reduce the thermal shock
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Journal Article
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Nuclear Power Engineering; ISSN 0258-0926; ; v. 23(3); p. 10-16
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AbstractAbstract
[en] Objective: Attempting to find a reliable method for assessing the patient's ability to tolerate carotid artery occlusion. Methods: The temporary balloon occlusion (TBO) test of carotid artery was performed for 20 patients who might have carotid artery manipulated or permanently occluded. Transfemoral artery Seldinger' s catheterization was used to introduced the temporary balloon occlusion catheters into the vessels of the concern. Neurologic testing was performed continuously by the attending neurologist. Transcranial Doppler ultrasonography (TCD) and carotid artery stump pressure (SP) were measured continuously during the TBO. The collateral circulation of Willis circle was observed with DSA. Results: Out of the 20 cases, one failed during the TBO because of CCA dissection caused by catheterization, another one failed because of a neurologic defect occurring before the balloon was inflated, the others went through the test uneventfully. Two cases finished the test before the approved schedule because neurologic defects appeared 34 min and 27 min after the vascular occlusion, respectively. These two patients were proved unable to tolerate carotid artery sacrifice. The other 16 cases passed the 45 minutes TBO. Their mean velocity of ipsilateral middle cerebral artery fell 36% ± 18%. Their SP is (53.76 ± 21.49) mmHg(30-87). Adequate collateral circulation in Willis circle was observed by DSA in all cases except the two who failed with the TBO. Conclusions: TBO is a safe and reliable method for assessing the patient's ability to tolerate carotid artery occlusion. The authors suggest it should be a routine examination prior to carotid manipulations
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Journal Article
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Journal of Interventional Radiology; ISSN 1008-794X; ; v. 11(5); p. 329-331
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[en] Simulative experiment focused on transient temperature is performed on a 5:1 model of surge line nozzle, with sealed sleeve or unsealed sleeve, of pressurizer in PWR, under inlet and outlet flow condition. The range of Re is 4500-750000. The conclusion is as follows: sealed sleeve can effectively protect surge line nozzle from thermal shock and thermal fatigue; Sealed sleeve has better protection than unsealed sleeve; at low flow rate unsealed sleeve has some protection from temperature, but it loses its protection at high flow rate
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Journal Article
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Numerical Data
Journal
Nuclear Power Engineering; ISSN 0258-0926; ; v. 23(1); p. 74-79
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