Liutao Chen

Influence of intermediate heat treatment on the creep properties of CZ1 zirconium alloys: Creep mechanisms and mechanistic modeling

Weitong Lin, Wenya Li, Qingyang Lv, Liutao Chen et autres

Oxidation Behaviour of CZ Alloys Under High Temperature Steam

Minli Chen, Changyuan Gao, Guoliang Zhang, Lin Shi et autres

Abstract CZ Alloys, including CZ1 and CZ2, are new advanced zirconium alloys developed by China General Nuclear Power Group. This paper presents the oxidation behaviour of CZ Alloys and Zircaloy-4 in flowing steam. A SETSYS Evolution TGA (ThermoGravimetric Analyzer) is used for …

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Study of Iodine-Induced Stress Corrosion Cracking of CZ2 Alloy

Lin Shi, Changyuan Gao, Guoliang Zhang, Guocheng Sun et autres

Abstract This paper is dealing with the iodine-induced stress corrosion cracking behavior of CZ2 zirconium alloy which is developed by China General Nuclear Power Group. The alloy in this study was fabricated with four different final annealing temperature in the range of …

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Study on Out-Pile Performance of CZ Zirconium Alloy Guide Tube

Changyuan Gao, Lin Shi, Guoliang Zhang, Yang Xu et autres

Abstract Based on the composition of CZ1 alloy and CZ2 alloy, guide tubes were fabricated. The performance of CZ1 and CZ2 alloy guide tube was studied. Through microscopic observation, it can be seen that the CZ alloy guide tube has a completely …

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Microstructual and Micro-Chemical Analysis of Zr Cladding Alloys on Corrosion and Creep Properties

Q.Y. Lv, Libin Zhang, Minli Chen, Changyuan Gao et autres

Abstract This paper conducts in-depth research on zirconium alloys by combining results of out-of-pile corrosion and creep performance with microstructual and micro-chemical analysis using state-of-the-art (S)TEM, EDX and EBSD. The mechanism of different heat treatment processes on the macroscopic properties of zirconium …

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Study on Water Chemistry and Temperature Adaptability of CZ Zirconium Alloy

Hansen Chen, Changyuan Gao, Liutao Chen, Xu Wang et autres

Abstract The corrosion behavior of zirconium alloys is closely related to the water chemistry and operating temperature of the primary circuit. The increase of operating temperature, Li concentration and pH value will have a significant negative impact on the corrosion behavior of …

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Study of Pre-Oxidization Law and Fretting Wear Resistance of CZ2 Alloy Cladding

Guocheng Sun, Lin Shi, Xu Wang, Liutao Chen

Abstract In the core of pressurized water nuclear reactor, coolant flow-induced vibration of Grid to rod fretting (GTRF) is the dominant factor leading to fuel rod damage. pre-oxidization treatment of zirconium cladding forming a ceramic layer on its surface is the main …

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Study on Fatigue Properties of Hydrided CZ2 Alloy

Jieming Huang, Changyuan Gao, Liutao Chen, Xu Wang et autres

Abstract CZ2 alloy is a new Zr-Nb zirconium alloy independently developed by China Nuclear Power Research and Technology Institute (CNPRI), which has independent intellectual property rights. Zirconium will react with coolant to absorb hydrogen in reactor service, which will degrade the performance …

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Effect of Final Annealing Temperature on Axial Creep Property of CZ Alloys

Lin Shi, Liutao Chen, Yang Xu, Changyuan Gao et autres

Two kinds of new cladding material for pressurized water reactor were developed by China Nuclear Power Research and Technology Institute (CNPRI) which named CZ1 and CZ2. The cladding tubes of CZ alloys used in the study were fabricated by different final annealing …

Investigation on Breakaway Oxidation Behavior of CZ Alloys

Changyuan Gao, Liutao Chen, Lin Shi, Yang Xu et autres

Breakaway oxidation is a potential fuel failure mechanism during a loss-of-coolant accident (LOCA), especially small-break LOCA. Two kinds of new advanced zirconium alloy, CZ1 and CZ2, were developed by China Nuclear Power Research and Technology Institute of China General Nuclear Power Group. …

Annealing Temperature Effect on Creep Property of CZ Alloy for PWR Modern Fuel Claddings

Yang Xu, Jun Tan, Liutao Chen, Dungu Wen et autres

Zirconium alloys remain as the main cladding materials in most water reactors. CZ zirconium alloy is recently developed for advanced PWR fuel assembly (STEP) designed by China General Nuclear Power Group (CGN). There are two kinds of zirconium alloys, designated as CZ1 …

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Effects of Annealing on the Corrosion Property of CZ1 Alloy

Liutao Chen, Tan Jun, Changyuan Gao, Dungu Wen et autres

CZ alloys were developed by China General Nuclear Power Group (CGN) for pressurized water reactor. To improve the corrosion resistance, the impacts of annealing on corrosion of CZ1 alloy were studied. The fuel cladding tubes of CZ1 alloy was fabricated by different …