In pressurized water reactor(PWR),fretting wear is one of the main causes of fuel assembly failure.Moreover,the operation condition of cladding is complex and harsh.A unique fretting damage test equipment was develope...In pressurized water reactor(PWR),fretting wear is one of the main causes of fuel assembly failure.Moreover,the operation condition of cladding is complex and harsh.A unique fretting damage test equipment was developed and tested to simulate the fretting damage evolution process of cladding in the PWR environment.It can simulate the fretting wear experiment of PWR under different temperatures(maximum temperature is 350℃),displacement amplitude,vibration frequency,and normal force.The fretting wear behavior of Zr-4 alloy under different temperature environments was tested.In addition,the evolution of wear scar morphology,profile,and wear volume was studied using an optical microscope(OM),scanning electron microscopy(SEM),and a 3D white light interferometer.Results show that higher water temperature evidently decreased the cladding wear volume,the wear mechanism of Zr-4 cladding changed from abrasive wear to adhesive wear and the formation of an oxide layer on the wear scar reduced the wear volume and maximum wear depth.展开更多
The subject of this study is the oxidation of fuel rod cladding made of material Zr1Nb(0.1% O) in steam at temperatures in the range of 660℃ to 1200℃ with a surface in the initial state (after manufacturing - grindi...The subject of this study is the oxidation of fuel rod cladding made of material Zr1Nb(0.1% O) in steam at temperatures in the range of 660℃ to 1200℃ with a surface in the initial state (after manufacturing - grinding) and after additional chemical etching. The changes in the microstructure of tubes due to the interaction with steam were investigated. A comparison was made between the oxidation rate of this material (weight gain) and the data on the oxidation of other alloys for nuclear power plants. The oxidation rate of Zr1Nb(0.1% O) is close to the oxidation rate of other zirconium alloys. It is shown that after chemical treatment of the surface of the samples there is a more even growth of oxide films, and they have a smaller thickness for the same time of exposure than after mechanical grinding. Surface treatment before oxidation also affects the change of microstructure of samples when heated to high temperatures.展开更多
With the aim of improving corrosion resistance of rod cladding for in-service and accident conditions,six new zirconium alloys(named N1-N6)have been designed.The contents of Sn and Nb were optimized for better behavio...With the aim of improving corrosion resistance of rod cladding for in-service and accident conditions,six new zirconium alloys(named N1-N6)have been designed.The contents of Sn and Nb were optimized for better behavior at high-temperature pressurized water,and Fe,Cr,V,Cu or Mo elements were added to the alloys to adjust the corrosion behavioi\The current work focused on the rapid corrosion behavior in 500℃/10.3 MPa steam for up to 1960 h,aiming to test the corrosion resistance at high temperature.The structure of matrix and properties of second-phase particles(SPPs)were characterized to find the main differences among these alloys.All the six alloys exhibited better corrosion resistance than N36,and NI was shown to have the best performance.A careful analysis of the corrosion kinetics curves revealed that Cr was beneficial for severe condition.Elements Fe,Cr,V,Cu or Mo aggregated into SPPs with diiferent concentrations and structures.This was demonstrated to be the main reason for different corrosion resistance.Due to good processing control,all alloys had a uniform structure and a uniform distribution of SPPs.As for N4,N6 and N36,the existing of large-size SPPs(450 nm)might be a contributing factor of the relatively poor corrosion resistance.展开更多
Creep behavior of the Zr-1.5Nb-0.4Sn-0.1Fe-0.1Cu alloy sheet is investigated from 300℃ to 400℃ in the stress range from 50 MPa to 180 MPa along the rolling direction. The measured strain rates range from 8.8 × ...Creep behavior of the Zr-1.5Nb-0.4Sn-0.1Fe-0.1Cu alloy sheet is investigated from 300℃ to 400℃ in the stress range from 50 MPa to 180 MPa along the rolling direction. The measured strain rates range from 8.8 × 10^-10 s^-1 to 4.7 × 10^-7 s^-1. The activation energies are estimated to assess the creep deformation mechanisms in this alloy. The strain rate is slightly different at low stress, while it shows a distinct difference at high stresses. Stress exponents of this alloy increase with increasing applied stress at all testing temperatures. It is concluded that the creep deformation of the Zr-1.5Nb-0.4Sn-0.1Fe-0. 1 Cu alloy is controlled by the diffusion creep at low stress region and by the climbing of dislocations at high stress region.展开更多
基金Supported by National Key R&D Program of China(Grant No.2022YFB3401901)Key Program of National Natural Science Foundation of China(Grant No.U2067221)+2 种基金Sichuan Provincial Science and Technology Planning Project(Grant Nos.2022JDJQ0019 and 2022ZYD0029)Funds for China Postdoctoral Science Foundation(Grant No.2022M713008)Sichuan Provincial Innovative Talent Funding Project for Postdoctoral Fellows(Grant No.BX202225).
文摘In pressurized water reactor(PWR),fretting wear is one of the main causes of fuel assembly failure.Moreover,the operation condition of cladding is complex and harsh.A unique fretting damage test equipment was developed and tested to simulate the fretting damage evolution process of cladding in the PWR environment.It can simulate the fretting wear experiment of PWR under different temperatures(maximum temperature is 350℃),displacement amplitude,vibration frequency,and normal force.The fretting wear behavior of Zr-4 alloy under different temperature environments was tested.In addition,the evolution of wear scar morphology,profile,and wear volume was studied using an optical microscope(OM),scanning electron microscopy(SEM),and a 3D white light interferometer.Results show that higher water temperature evidently decreased the cladding wear volume,the wear mechanism of Zr-4 cladding changed from abrasive wear to adhesive wear and the formation of an oxide layer on the wear scar reduced the wear volume and maximum wear depth.
文摘The subject of this study is the oxidation of fuel rod cladding made of material Zr1Nb(0.1% O) in steam at temperatures in the range of 660℃ to 1200℃ with a surface in the initial state (after manufacturing - grinding) and after additional chemical etching. The changes in the microstructure of tubes due to the interaction with steam were investigated. A comparison was made between the oxidation rate of this material (weight gain) and the data on the oxidation of other alloys for nuclear power plants. The oxidation rate of Zr1Nb(0.1% O) is close to the oxidation rate of other zirconium alloys. It is shown that after chemical treatment of the surface of the samples there is a more even growth of oxide films, and they have a smaller thickness for the same time of exposure than after mechanical grinding. Surface treatment before oxidation also affects the change of microstructure of samples when heated to high temperatures.
基金funded by the Major Project of CNNC (China National Nuclear Corporation): Key Technology Research on CF4 Fuel Assembly and Associated Assembly (No.[2016] 298)
文摘With the aim of improving corrosion resistance of rod cladding for in-service and accident conditions,six new zirconium alloys(named N1-N6)have been designed.The contents of Sn and Nb were optimized for better behavior at high-temperature pressurized water,and Fe,Cr,V,Cu or Mo elements were added to the alloys to adjust the corrosion behavioi\The current work focused on the rapid corrosion behavior in 500℃/10.3 MPa steam for up to 1960 h,aiming to test the corrosion resistance at high temperature.The structure of matrix and properties of second-phase particles(SPPs)were characterized to find the main differences among these alloys.All the six alloys exhibited better corrosion resistance than N36,and NI was shown to have the best performance.A careful analysis of the corrosion kinetics curves revealed that Cr was beneficial for severe condition.Elements Fe,Cr,V,Cu or Mo aggregated into SPPs with diiferent concentrations and structures.This was demonstrated to be the main reason for different corrosion resistance.Due to good processing control,all alloys had a uniform structure and a uniform distribution of SPPs.As for N4,N6 and N36,the existing of large-size SPPs(450 nm)might be a contributing factor of the relatively poor corrosion resistance.
基金supported by Korea Science & Engineering Foundation and the Ministry of Science & Technology,Korean government,through its national nuclear technology program.
文摘Creep behavior of the Zr-1.5Nb-0.4Sn-0.1Fe-0.1Cu alloy sheet is investigated from 300℃ to 400℃ in the stress range from 50 MPa to 180 MPa along the rolling direction. The measured strain rates range from 8.8 × 10^-10 s^-1 to 4.7 × 10^-7 s^-1. The activation energies are estimated to assess the creep deformation mechanisms in this alloy. The strain rate is slightly different at low stress, while it shows a distinct difference at high stresses. Stress exponents of this alloy increase with increasing applied stress at all testing temperatures. It is concluded that the creep deformation of the Zr-1.5Nb-0.4Sn-0.1Fe-0. 1 Cu alloy is controlled by the diffusion creep at low stress region and by the climbing of dislocations at high stress region.
