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Crud deposition behavior on zirconium alloy fuel cladding in high-temperature pressurized water environments
C Xue, Y Mao, Z Zhang, J Tan, X Wu, EH Han… - Journal of Nuclear …, 2022 - Elsevier
Crud deposition behavior on zirconium alloy fuel cladding in high-temperature pressurized
oxygenated water environments was investigated. Porous crud with typical boiling chimney …
oxygenated water environments was investigated. Porous crud with typical boiling chimney …
Effects of Zn injection on corrosion behavior and crud deposition of FeCrAl fuel cladding under subcooled nuclear boiling condition in high-temperature pressurized …
C Xue, Z Zhang, J Tan, X Wu, EH Han, W Ke - Corrosion Science, 2023 - Elsevier
The effects of Zn injection on corrosion behavior and crud deposition of FeCrAl fuel cladding
in high-temperature pressurized water were investigated under subcooled nuclear boiling …
in high-temperature pressurized water were investigated under subcooled nuclear boiling …
The transformation of fretting corrosion mechanism of zirconium alloy tube mating with 304 stainless steel in high temperature high pressure water
W Zhuang, P Lai, H Lu, Z Han, J Lu, L Zhang… - Journal of Nuclear …, 2023 - Elsevier
Studying the fretting wear mechanism of zirconium alloy in high temperature high pressure
water is of great importance for understanding the failure behavior of fuel assembly. In this …
water is of great importance for understanding the failure behavior of fuel assembly. In this …
Characterization and corrosion behavior of plasma electrolytic oxidation coating on zirconium alloy in superheated steam condition
The plasma electrolytic oxidation (PEO) coating on Zirconium (Zr) alloy was employed and
immersed under 400° C/10.3 MPa steam conditions. The surface morphology, phase …
immersed under 400° C/10.3 MPa steam conditions. The surface morphology, phase …
Systematic investigations on the oxidation mechanism of Cr coated Zr-4 alloy under different high-temperature steam conditions
Y Wang, Y Wang, S Wang, J Geng, C Zhang… - Ceramics …, 2024 - Elsevier
Chromium (Cr) coating is a potential candidate material for improving the oxidation
resistance and extending rail service life of zirconium (Zr) alloys used in nuclear fuel …
resistance and extending rail service life of zirconium (Zr) alloys used in nuclear fuel …
Electrochemical and modelling study of ZrNbO alloys aged under high temperature and high pressure PWR simulated conditions
This work proposes a kinetic study of Zirconium alloy when working under controlled
temperature and pressure conditions to simulate pressurized water reactor (PWR) …
temperature and pressure conditions to simulate pressurized water reactor (PWR) …
Long-term oxidation of zirconium alloy in simulated nuclear reactor primary coolant—Experiments and modeling
Oxidation of Zr-1% Nb fuel cladding alloy in simulated primary coolant of a pressurized
water nuclear reactor is followed by in-situ electrochemical impedance spectroscopy. In …
water nuclear reactor is followed by in-situ electrochemical impedance spectroscopy. In …
Analyses of electrochemical behavior of plasma electrolytic oxidation film on Zirlo alloy in lithium borate buffer solution at 25–300° C
K Wei, X Wang, C Xu, J Du, W Xue, G Cheng - Surface and Coatings …, 2022 - Elsevier
The electrochemical behavior of plasma electrolytic oxidation (PEO) film on Zirlo alloy in 25–
300° C lithium borate buffer solution was evaluated by in-situ electrochemical impedance …
300° C lithium borate buffer solution was evaluated by in-situ electrochemical impedance …
Effect of Plasma Electrolytic Oxidation Coating on Corrosion Behavior of Zirconium Alloy in Superheated Steam Condition
Plasma electrolytic oxidation (PEO) coating on Zr alloy was prepared, and the corrosion
behavior was investigated under 400° C/10.3 MPa steam condition. The surface …
behavior was investigated under 400° C/10.3 MPa steam condition. The surface …
Nickel Ferrite as a Model Corrosion Product and its Deposition on Current and Future Nuclear Fuel Cladding Materials
BS Nagothi - 2024 - scholarsarchive.library.albany.edu
The primary corrosion products in Light Water Reactors (LWRs) are nickel ferrites (nominal
stoichiometry NiFe 2 O 4) having the spinel crystal structure. These products, commonly …
stoichiometry NiFe 2 O 4) having the spinel crystal structure. These products, commonly …