The Experts below are selected from a list of 249 Experts worldwide ranked by ideXlab platform
Wenjun Kuang - One of the best experts on this subject based on the ideXlab platform.
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a high resolution characterization of the initiation of stress corrosion crack in alloy 690 in simulated Pressurized Water Reactor primary Water
Corrosion Science, 2020Co-Authors: Wenjun Kuang, Gary S WasAbstract:Abstract Stress corrosion crack initiation of alloy 690 in simulated Pressurized Water Reactor primary Water was studied through high-resolution characterization of grain boundaries at different stages of the initiation process. It was found that a compact layer of Cr2O3 initially forms over a migrated grain boundary, driven by the diffusion of Cr. After the surface Cr2O3 is breached by straining, oxygen diffusion results in formation of a mixture of NiO and Cr2O3 along the grain boundary. The crack nucleates along either the previous grain boundary or the heavily oxidized migration zone when the boundary strength falls below the local stress.
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the oxidation of alloy 690 in simulated Pressurized Water Reactor primary Water
Corrosion Science, 2017Co-Authors: Wenjun Kuang, Miao Song, Peng WangAbstract:Abstract The oxide film formed on alloy 690 in simulated Pressurized Water Reactor primary Water is decorated with NiFe 2 O 4 particles which remain epitaxial with matrix. The penetrative inner oxides on both surface and crack wall are composed of Cr-rich (Fe,Cr,Ni) 3 O 4 and Cr 2 O 3 which were formed by solid state reactions of substrate with inwards diffusing oxygen and have orientation relationships with the substrate. Cr 2 O 3 forms first along the widely spaced planes of substrate. Compact Cr 2 O 3 layer cannot develop as there is no long-range outward diffusion of Cr in substrate with low defect density at low temperature.
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microstructural study on the stress corrosion cracking of alloy 690 in simulated Pressurized Water Reactor primary environment
18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors 2017, 2017Co-Authors: Wenjun Kuang, Miao Song, Chad M Parish, Gary S WasAbstract:This study was aimed at investigating the intergranular attack near a stress corrosion crack (SCC) of alloy 690 in simulated Pressurized Water Reactor (PWR) primary Water environment. Solution annealed alloy 690 was evaluated for its SCC initiation susceptibility in 360 °C hydrogenated pure Water using slow strain rate tensile technique. After the test, a grain boundary showing SCC initiation was sampled with Focused Ion Beam (FIB) milling. The microstructure and elemental distribution near the crack tip were studied using transmission electron microscopy (TEM) and scanning transmission electron microscopy (STEM). The results show that intergranular oxidation occurs ahead of the crack tip and is preceded by diffusion induced grain boundary migration. The oxides at the crack tip are mainly composed of NiO and Cr2O3 which maintain rigid orientations with the neighboring grains. The adjacent migration zone is free of oxidization as a compact layer of Cr2O3 dominates at the oxide/substrate interfaces and the very tip region.
Gary S Was - One of the best experts on this subject based on the ideXlab platform.
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a high resolution characterization of the initiation of stress corrosion crack in alloy 690 in simulated Pressurized Water Reactor primary Water
Corrosion Science, 2020Co-Authors: Wenjun Kuang, Gary S WasAbstract:Abstract Stress corrosion crack initiation of alloy 690 in simulated Pressurized Water Reactor primary Water was studied through high-resolution characterization of grain boundaries at different stages of the initiation process. It was found that a compact layer of Cr2O3 initially forms over a migrated grain boundary, driven by the diffusion of Cr. After the surface Cr2O3 is breached by straining, oxygen diffusion results in formation of a mixture of NiO and Cr2O3 along the grain boundary. The crack nucleates along either the previous grain boundary or the heavily oxidized migration zone when the boundary strength falls below the local stress.
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microstructural study on the stress corrosion cracking of alloy 690 in simulated Pressurized Water Reactor primary environment
18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors 2017, 2017Co-Authors: Wenjun Kuang, Miao Song, Chad M Parish, Gary S WasAbstract:This study was aimed at investigating the intergranular attack near a stress corrosion crack (SCC) of alloy 690 in simulated Pressurized Water Reactor (PWR) primary Water environment. Solution annealed alloy 690 was evaluated for its SCC initiation susceptibility in 360 °C hydrogenated pure Water using slow strain rate tensile technique. After the test, a grain boundary showing SCC initiation was sampled with Focused Ion Beam (FIB) milling. The microstructure and elemental distribution near the crack tip were studied using transmission electron microscopy (TEM) and scanning transmission electron microscopy (STEM). The results show that intergranular oxidation occurs ahead of the crack tip and is preceded by diffusion induced grain boundary migration. The oxides at the crack tip are mainly composed of NiO and Cr2O3 which maintain rigid orientations with the neighboring grains. The adjacent migration zone is free of oxidization as a compact layer of Cr2O3 dominates at the oxide/substrate interfaces and the very tip region.
Ahmet Durmayaz - One of the best experts on this subject based on the ideXlab platform.
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Influence of cooling Water temperature on the efficiency of a Pressurized-Water Reactor nuclear-power plant
International Journal of Energy Research, 2006Co-Authors: Ahmet Durmayaz, Oguz Salim SogutAbstract:In this study, the influence of the cooling Water temperature on the thermal efficiency of a conceptual Pressurized-Water Reactor nuclear-power plant is studied through an energy analysis based on the first law of thermodynamics to gain some new insights into the plant performance. The change in the cooling Water temperature can be experienced due to the seasonal changes in climatic conditions at plant site. It can also come into the question of design processes for the plant site selection. In the analysis, it is considered that the condenser vacuum varies with the temperature of cooling Water extracted from environment into the condenser. The main findings of the paper is that the impact of 1°C increase in temperature of the coolant extracted from environment is predicted to yield a decrease of ∼0.45 and ∼0.12% in the power output and the thermal efficiency of the Pressurized-Water Reactor nuclear-power plant considered, respectively. Copyright © 2005 John Wiley & Sons, Ltd.
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exergy analysis of a Pressurized Water Reactor nuclear power plant
Applied Energy, 2001Co-Authors: Ahmet Durmayaz, Hasbi YavuzAbstract:An exergy analysis based on the second law of thermodynamics is performed to evaluate the plant and subsystem irreversibility of a nuclear power plant (NPP) with a Pressurized-Water Reactor (PWR). The construction of such a system having a maximum Reactor core thermal power of 4250 MW is proposed in Turkey and China. This study concentrates on the questions of where and how much of the available work is lost in such a plant. The evaluated exergy destruction of this plant indicates that the Reactor pressure vessel including PWR is the most inefficient equipment in the whole NPP, while the turbines take the second place.
Hasbi Yavuz - One of the best experts on this subject based on the ideXlab platform.
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exergy analysis of a Pressurized Water Reactor nuclear power plant
Applied Energy, 2001Co-Authors: Ahmet Durmayaz, Hasbi YavuzAbstract:An exergy analysis based on the second law of thermodynamics is performed to evaluate the plant and subsystem irreversibility of a nuclear power plant (NPP) with a Pressurized-Water Reactor (PWR). The construction of such a system having a maximum Reactor core thermal power of 4250 MW is proposed in Turkey and China. This study concentrates on the questions of where and how much of the available work is lost in such a plant. The evaluated exergy destruction of this plant indicates that the Reactor pressure vessel including PWR is the most inefficient equipment in the whole NPP, while the turbines take the second place.
Salah Ud-din Khan - One of the best experts on this subject based on the ideXlab platform.
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neutronics and thermal hydraulic coupling analysis of integrated Pressurized Water Reactor
International Journal of Energy Research, 2013Co-Authors: Salah Ud-din Khan, Minjun Peng, Shahab Uddin KhanAbstract:SUMMARY In this paper, three-dimensional (3D) power distribution of newly designed small nuclear Reactor core has been achieved by using neutron kinetic/thermal hydraulic (NK/TH) coupling. This is Pressurized Water Reactor-based small nuclear Reactor in which plate type fuel element has been used and the core of the Reactor has hexagonal type geometry. This paper depicts the design of the Reactor core by using coupling approach of neutronics(Neutron Kinetic) and thermal hydraulic studies. For this purpose, neutronic analysis has been obtained by using lattice physics code, i.e. HELIOS and neutron kinetic code, i.e. REMARK. HELIOS code gives the cross-section data which is being used as input to the REMARK code. At the same time, THEATRe code was used for the thermal hydraulic analysis of the Reactor core. In the coupling process, some data (fuel temperature, moderator temperature, void fraction, etc.) from THEATRe code has been used in conjunction with HELIOS and REMARK codes. After finalizing the NK/TH coupling, 3D evaluation of the power distribution of the Reactor core has been achieved and is included in the paper. The purpose of this paper is to evaluate the design and get the normal operational behavior of the Reactor core by NK/TH coupling approach. Copyright © 2012 John Wiley & Sons, Ltd.
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Simulation of Loss of Coolant Accident in an Integral Pressurized Water Reactor (IPWR)
Advanced Materials Research, 2011Co-Authors: Salah Ud-din Khan, Minjun Peng, Muhammad ZubairAbstract:In this paper research has been carried out on the Loss of Coolant Accident (LOCA) in an Integral Pressurized Water Reactor(IPWR) by using thermal hydraulic system code Relap5/Mod3.4.The designing of Integrated Pressurized Water Reactor (IPWR) incorporates the safety and reliability of the Reactor to withstand under accidental vulnerabilities. In this study, the Reactor under consideration is Uranium Zirconium Hydride Nuclear Power Reactor INSURE-100 with the power output of 100MW.In the current research, the Reactor has been described in detail according to the requirement for the simulation of LOCA using Relap5 code with the possibility of occurrence of the time sequence of events. The graphs obtained shows good agreement for the safe operation of IPWR under LOCA.
