The Experts below are selected from a list of 7581 Experts worldwide ranked by ideXlab platform
G H Su - One of the best experts on this subject based on the ideXlab platform.
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experimental simulation of liquid entrainment in ads 4 depressurization line in ap1000
Progress in Nuclear Energy, 2016Co-Authors: Yan Xiang, Yingwei Wu, W X Tian, P Zhang, G H SuAbstract:Abstract The fourth stage Automatic Depressurization System is an important passive safety feature in Westinghouse AP1000 which enables controlled depressurization of Reactor Coolant System in small break LOCA. However, large amount of Coolant can be carried to the containment via the ADS-4 branch entrainment and the upper plenum entrainment during the depressurization process, which poses great threats to core uncovering and melting. The automatic Depressurization and Entrainment TEst Loop (ADETEL) modeled after AP1000 with a scaling ratio of 1:5.6 was constructed to investigate the entrainment and depressurization behavior after the actuation of ADS-4 valves. The entrainment and depressurization features were investigated under different initial pressure, mixture liquid level in the pressure vessel and heating power. The entrainment deposition effect of the Reactor internals was also investigated. The test data reveals that large amount of water are entrained through the ADS-4 branch line within a short period of time. The liquid entrainment rate and the reduced rate of the mixture liquid level in the pressure vessel increase with the initial System pressure. It is notable that the core uncovery was experienced when the initial pressure was set to 0.5 MPa in current experimental conditions. The Reactor internals have little effect on the entrained mass and the mixture liquid levels in the pressure vessel.
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experimental investigation of liquid entrainment in ads 4 depressurization line with steam water
Experimental Thermal and Fluid Science, 2015Co-Authors: Yan Xiang, W X Tian, P Zhang, J Zhang, G H SuAbstract:Abstract The fourth stage Automatic Depressurization System (ADS-4) is an important part in Westinghouse AP600/AP100 which enables controlled depressurization of Reactor Coolant System during small break LOCA. However, the Coolant may be entrained into the containment through the ADS-4 branch line simultaneously, which poses great threats to Reactor core uncovering and melting. To investigate the entrainment phenomenon of ADS-4 in AP1000, ADS-4 Depressurization and Entrainment TEst Loop (ADETEL) with a scaling ratio of 1:5.6 was constructed and visualization experiments were conducted with steam–water. The experiment were conducted at atmospheric pressure with the maximum steam flow rate of 700 kg/h. The entrainment processes were recorded by a high speed camera and analyzed in detail. Onset of entrainment and entrainment rate data were obtained and compared with existing test data and correlations. Empirical correlation of ADS-4 branch entrainment onset has been developed. A clear difference exists between ADETEL entrainment onset data and available correlations, which may caused by the differences in fluids physical properties, horizontal pipe flow conditions, determination of entrainment onset and d / D ratios. The entrainment onset is more likely to be reached in small d / D ratios conditions due to the stronger Bernoulli effect. Besides, AP1000 has a higher branch quality than AP600 at the same liquid level in the hot leg.
Xuewu Cao - One of the best experts on this subject based on the ideXlab platform.
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assessment of passive residual heat removal System cooling capacity
Progress in Nuclear Energy, 2014Co-Authors: Jie Zou, Lili Tong, Xuewu CaoAbstract:Abstract Advanced passive PWR relies on passive safety Systems to provide core cooling capacity and deal with design basis accidents and beyond design basis accidents. However, the passive safety System is lack of practical operating experience and their performance is heavily influenced by other Systems. The cooling capacity of passive residual heat removal System (PRHR), which is designed to remove decay heat when normal heat removal approach is not available, requires specific assessment during different accidents. In this study, a detail model of advanced passive PWR, including Reactor Coolant System (RCS), simplified secondary side and Engineered Safety Features (ESF), has been built using mechanism accident analysis code. The plant transient has been simulated, and cooling capacity of PRHR been analyzed during loss of normal feedwater and main feedwater line rupture. Conservative assumptions were made specially based on different accident scenarios and one of the two fail-open valves arranged in parallel at the PRHR heat exchanger (HX) outlet line was assumed not open, as the worst single failure. The progress of the two accident sequence is calculated and the thermalhydraulic behavior of RCS is investigated and the main transient parameters are obtained, including primary side pressure, steam generator pressure, pressurizer water level. The cooling power and System response are calculated. The results show that PRHR, with CMT injection, can remove the decay heat from RCS to IRWST, keeping the pressures of RCS and steam generators remaining below 110 percent of the design values and the pressurizer overfilling is prevented. Sensitivity study has been performed to study the System resistance effects on the capacity of PRHR, which shows that increase in System resistance coefficient reduces the cooling capacity of PRHR.
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preliminary analysis of effect of the intentional depressurization on fission product behavior during tmlb severe accident
Volume 2: Structural Integrity; Safety and Security; Advanced Applications of Nuclear Technology; Balance of Plant for Nuclear Applications, 2009Co-Authors: Gaofeng Huang, Lili Tong, Xuewu CaoAbstract:It has been identified that the pressure in the Reactor Coolant System (RCS) remains high in some severe accident sequences at the time of Reactor vessel failure, with the risk of causing direct containment heating (DCH). Intentional depressurization is an effective accident management strategy to prevent DCH or to mitigate its effects. Fission product behavior is affected by intentional depressurization, especially for inert gas and volatile fission product. Because the pressurizer power-operated relief valves (PORVs) are latched open, fission product will transport into the containment directly. This may cause larger radiological consequences in containment before Reactor vessel failure. Four cases are selected, including the TMLB’ base case and opening one, two and three pressurizer PORVs. The results show that inert gas transports into containment more quickly when opening one and two PORVs, but more slowly when opening three PORVs; more volatile fission product deposit in containment and less in Reactor Coolant System (RCS) for intentional depressurization cases. When opening one PORV, the phenomena of revaporization is strong in the RCS.Copyright © 2009 by ASME
Yan Xiang - One of the best experts on this subject based on the ideXlab platform.
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experimental simulation of liquid entrainment in ads 4 depressurization line in ap1000
Progress in Nuclear Energy, 2016Co-Authors: Yan Xiang, Yingwei Wu, W X Tian, P Zhang, G H SuAbstract:Abstract The fourth stage Automatic Depressurization System is an important passive safety feature in Westinghouse AP1000 which enables controlled depressurization of Reactor Coolant System in small break LOCA. However, large amount of Coolant can be carried to the containment via the ADS-4 branch entrainment and the upper plenum entrainment during the depressurization process, which poses great threats to core uncovering and melting. The automatic Depressurization and Entrainment TEst Loop (ADETEL) modeled after AP1000 with a scaling ratio of 1:5.6 was constructed to investigate the entrainment and depressurization behavior after the actuation of ADS-4 valves. The entrainment and depressurization features were investigated under different initial pressure, mixture liquid level in the pressure vessel and heating power. The entrainment deposition effect of the Reactor internals was also investigated. The test data reveals that large amount of water are entrained through the ADS-4 branch line within a short period of time. The liquid entrainment rate and the reduced rate of the mixture liquid level in the pressure vessel increase with the initial System pressure. It is notable that the core uncovery was experienced when the initial pressure was set to 0.5 MPa in current experimental conditions. The Reactor internals have little effect on the entrained mass and the mixture liquid levels in the pressure vessel.
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experimental investigation of liquid entrainment in ads 4 depressurization line with steam water
Experimental Thermal and Fluid Science, 2015Co-Authors: Yan Xiang, W X Tian, P Zhang, J Zhang, G H SuAbstract:Abstract The fourth stage Automatic Depressurization System (ADS-4) is an important part in Westinghouse AP600/AP100 which enables controlled depressurization of Reactor Coolant System during small break LOCA. However, the Coolant may be entrained into the containment through the ADS-4 branch line simultaneously, which poses great threats to Reactor core uncovering and melting. To investigate the entrainment phenomenon of ADS-4 in AP1000, ADS-4 Depressurization and Entrainment TEst Loop (ADETEL) with a scaling ratio of 1:5.6 was constructed and visualization experiments were conducted with steam–water. The experiment were conducted at atmospheric pressure with the maximum steam flow rate of 700 kg/h. The entrainment processes were recorded by a high speed camera and analyzed in detail. Onset of entrainment and entrainment rate data were obtained and compared with existing test data and correlations. Empirical correlation of ADS-4 branch entrainment onset has been developed. A clear difference exists between ADETEL entrainment onset data and available correlations, which may caused by the differences in fluids physical properties, horizontal pipe flow conditions, determination of entrainment onset and d / D ratios. The entrainment onset is more likely to be reached in small d / D ratios conditions due to the stronger Bernoulli effect. Besides, AP1000 has a higher branch quality than AP600 at the same liquid level in the hot leg.
Cao Xuewu - One of the best experts on this subject based on the ideXlab platform.
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Safety analysis of increase in heat removal from Reactor Coolant System with inadvertent operation of passive residual heat removal at no-load conditions
Published by Elsevier B.V., 2015Co-Authors: Ge Shao, Cao XuewuAbstract:AbstractThe advanced passive pressurized water Reactor (PWR) is being constructed in China and the passive residual heat removal (PRHR) System was designed to remove the decay heat. During accident scenarios with increase of heat removal from the primary Coolant System, the actuation of the PRHR will enhance the cooldown of the primary Coolant System. There is a risk of power excursion during the cooldown of the primary Coolant System. Therefore, it is necessary to analyze the thermal hydraulic behavior of the Reactor Coolant System (RCS) at this condition. The advanced passive PWR model, including major components in the RCS, is built by SCDAP/RELAP5 code. The thermal hydraulic behavior of the core is studied for two typical accident sequences with PRHR actuation to investigate the core cooling capability with conservative assumptions, a main steam line break (MSLB) event and inadvertent opening of a steam generator (SG) safety valve event. The results show that the core is ultimately shut down by the boric acid solution delivered by Core Makeup Tank (CMT) injections. The effects of CMT boric acid concentration and the activation delay time on accident consequences are analyzed for MSLB, which shows that there is no consequential damage to the fuel or Reactor Coolant System in the selected conditions
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effect of depressurization on hydrogen generation during severe accident in pwr nuclear power plant
Nuclear Instruments & Methods in Physics Research Section B-beam Interactions With Materials and Atoms, 2011Co-Authors: Cao XuewuAbstract:The effect of depressurization on hydrogen generation during a typical high pressure severe accident sequence in a 1 000 MWe pressurized water Reactor(PWR) nuclear power plant was analyzed.Analyses results indicate that the hydrogen generation rate is obviously increased by the Reactor Coolant System depressurization of opening one,two or three power operated relief valves(PORVs) at three core damage states.The first is peak core temperature from 1 500 K to 2 100 K.The second is peak core temperature from 2 500 K to 2 800 K.The third is from formation of molten pool supported by crust to slumping of molten materials into Reactor pressure vessel lower head.The more PORVs are opened the more increment of hydrogen generation rate.
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effect analysis of the intentional depressurization on fission product behavior during tmlb severe accident
Nuclear Science and Techniques, 2009Co-Authors: Huang Gaofeng, L I Jingxi, Tong Lili, Cao XuewuAbstract:It has been found that the pressure in the Reactor Coolant System (RCS) remains high in some severe accident sequences at the time of Reactor vessel failure, with the risk of causing direct containment heating (DCH). Intentional depressurization is an effective accident management strategy to prevent DCH or to mitigate its consequences. Fission product behavior is affected by intentional depressurization, especially for inert gas and volatile fission product. Because the pressurizer power-operated relief valves (PORVs) are latched open, fission product will transport into the containment directly. This may cause larger radiological consequences in containment before Reactor vessel failure. Four cases are selected, including the TMLB’ base case and the opening one, two and three pressurizer PORVs. The results show that inert gas transports into containment more quickly when opening one and two PORVs, but more slowly when opening three PORVs; more volatile fission product deposit in containment and less in Reactor Coolant System (RCS) for intentional depressurization cases. When opening one PORV, the phenomenon of revaporization is strong in the RCS.
W X Tian - One of the best experts on this subject based on the ideXlab platform.
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experimental simulation of liquid entrainment in ads 4 depressurization line in ap1000
Progress in Nuclear Energy, 2016Co-Authors: Yan Xiang, Yingwei Wu, W X Tian, P Zhang, G H SuAbstract:Abstract The fourth stage Automatic Depressurization System is an important passive safety feature in Westinghouse AP1000 which enables controlled depressurization of Reactor Coolant System in small break LOCA. However, large amount of Coolant can be carried to the containment via the ADS-4 branch entrainment and the upper plenum entrainment during the depressurization process, which poses great threats to core uncovering and melting. The automatic Depressurization and Entrainment TEst Loop (ADETEL) modeled after AP1000 with a scaling ratio of 1:5.6 was constructed to investigate the entrainment and depressurization behavior after the actuation of ADS-4 valves. The entrainment and depressurization features were investigated under different initial pressure, mixture liquid level in the pressure vessel and heating power. The entrainment deposition effect of the Reactor internals was also investigated. The test data reveals that large amount of water are entrained through the ADS-4 branch line within a short period of time. The liquid entrainment rate and the reduced rate of the mixture liquid level in the pressure vessel increase with the initial System pressure. It is notable that the core uncovery was experienced when the initial pressure was set to 0.5 MPa in current experimental conditions. The Reactor internals have little effect on the entrained mass and the mixture liquid levels in the pressure vessel.
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experimental investigation of liquid entrainment in ads 4 depressurization line with steam water
Experimental Thermal and Fluid Science, 2015Co-Authors: Yan Xiang, W X Tian, P Zhang, J Zhang, G H SuAbstract:Abstract The fourth stage Automatic Depressurization System (ADS-4) is an important part in Westinghouse AP600/AP100 which enables controlled depressurization of Reactor Coolant System during small break LOCA. However, the Coolant may be entrained into the containment through the ADS-4 branch line simultaneously, which poses great threats to Reactor core uncovering and melting. To investigate the entrainment phenomenon of ADS-4 in AP1000, ADS-4 Depressurization and Entrainment TEst Loop (ADETEL) with a scaling ratio of 1:5.6 was constructed and visualization experiments were conducted with steam–water. The experiment were conducted at atmospheric pressure with the maximum steam flow rate of 700 kg/h. The entrainment processes were recorded by a high speed camera and analyzed in detail. Onset of entrainment and entrainment rate data were obtained and compared with existing test data and correlations. Empirical correlation of ADS-4 branch entrainment onset has been developed. A clear difference exists between ADETEL entrainment onset data and available correlations, which may caused by the differences in fluids physical properties, horizontal pipe flow conditions, determination of entrainment onset and d / D ratios. The entrainment onset is more likely to be reached in small d / D ratios conditions due to the stronger Bernoulli effect. Besides, AP1000 has a higher branch quality than AP600 at the same liquid level in the hot leg.
