Current status of materials development of nuclear fuel cladding tubes for light water reactors

Duan, Zhengang; Yang, Huilong; Satoh, Yuhki; Murakami, Kenta; Kano, Shinzo; Zhao, Zhenyu; Shen, Jingjie; Abe, Hiroaki

doi:10.1016/j.nucengdes.2017.02.031

Cited by 269 publications

(92 citation statements)

References 108 publications

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“…Due to the exothermic zirconium-steam oxidation reaction, a large quantity of heat is released and hydrogen gas is produced increasing the risks of hydrogen detonation and radioactive fission products release [3][4][5]. Therefore, the development of advanced accident-tolerant fuels (ATFs) with larger safety margins became one primary focus after the Fukushima accident in 2011 [6,7].…”

Section: Introductionmentioning

confidence: 99%

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Influence of composition and heating schedules on compatibility of FeCrAl alloys with high-temperature steam

Tang

Jianu

Steinbrueck

et al. 2018

Journal of Nuclear Materials

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FeCrAl alloys are proposed and being intensively investigated as alternative accident tolerant fuel (ATF) cladding for nuclear fission application. Herein, the influence of major alloy elements (Cr and Al), reactive element effect and heating schedules on the oxidation behavior of FeCrAl alloys in steam up to 1500°C was examined. In case of transient ramp tests, catastrophic oxidation, i.e. rapid and complete consumption of the alloy, occurred during temperature ramp up to above 1200°C for specific alloys. The maximum compatible temperature of FeCrAl alloys in steam increases with raising Cr and Al content, decreasing heating rates during ramp period and doping of yttrium. Isothermal oxidation resulted in catastrophic oxidation at 1400°C for all examined alloys. However, formation of a protective alumina scale at 1500°C was ascertained despite partial melting. The occurrence of catastrophic oxidation seems to be controlled by dynamic competitive mechanisms between mass transfer of Al from the substrate and transport of oxidizing gas through the scale both toward the metal/oxide scale interface.

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Section: Introductionmentioning

confidence: 99%

“…Alternative ATF cladding concepts, including coated Zr-based cladding, hybrid ceramic/metal cladding, or advanced ceramic and metallic cladding, owning excellent high-temperature oxidation resistance are being investigated aiming to enhance the accident tolerance [2,6,[8][9][10][11].…”

Section: Introductionmentioning

confidence: 99%

Influence of composition and heating schedules on compatibility of FeCrAl alloys with high-temperature steam

Tang

Jianu

Steinbrueck

et al. 2018

Journal of Nuclear Materials

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“…Higher fuel consumption and power up-rating for reactor operation require the development of the advanced zirconium alloys exhibiting better corrosion resistance and lower hydrogen uptake [12,13]. An alternative way to enhance corrosion resistance and reduce hydrogenation of zirconium alloys is coating deposition or surface modification [14]. Currently, there are various deposition technologies and coatings providing better resistance of zirconium alloys during steam corrosion and accident conditions, such as Cr [15,16], TiN, and TiN/TiAlN [17][18][19], micro arc oxidation coatings [20], CrN and AlCrN [21], and others.…”

Section: Introductionmentioning

confidence: 99%

Hydrogen Interaction with Deep Surface Modified Zr-1Nb Alloy by High Intensity Ti Ion Implantation

Kashkarov

Ryabchikov

Kurochkin

et al. 2018

Metals

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A deep surface modified TiZr layer was fabricated by high-intensity low-energy titanium ion implantation into zirconium alloy Zr-1Nb. Gas-phase hydrogenation was performed to evaluate protective properties of the modified layer against hydrogen permeation into Zr-1Nb alloy. The effects of ion implantation and hydrogen on microstructure, phase composition and elemental distribution of TiZr layer were analyzed by scanning electron microscopy, X-ray diffraction, and glow-discharge optical emission spectroscopy, respectively. It was revealed that TiZr layer (~10 μm thickness) is represented by α′ + α(TiZr) lamellar microstructure with gradient distribution of Ti through the layer depth. It was shown that the formation of TiZr layer provides significant reduction of hydrogen uptake by zirconium alloy at 400 and 500 °C. Hydrogenation of the modified layer leads to refinement of lamellar plates and formation of more homogenous microstructure. Hydrogen desorption from Ti-implanted Zr-1Nb alloy was analyzed by thermal desorption spectroscopy. Hydrogen interaction with the surface modified TiZr layer, as well as its resistance properties, are discussed.

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“…The last phenomena may appear together and result in hydrogen embrittlement in forms of simultaneous hydrogen-enhanced localized plasticity and delayed hydride cracking.Some Zr alloys are used in the nuclear industry for fuel claddings [1-3], reflectors in light water reactors [4], and in spent nuclear fuel reprocessing plants [5,6]. Zirconium alloys are applied to manufacture nuclear fuel pellets due to their low thermal neutron capture cross-section, proper strength properties, and excellent corrosion resistance in the cooling medium [7]. The nuclear fuel pellets are made of the Zr-Sn Zircaloys, Zr-Nb E110, E125, and E635, Zr-Sn-Nb Zirlo, Zr-Nb M5, and X5A alloys [7-10], Zr-Nb and Zr-Nb-Fe [11], and the 702 alloys [12,13].The most important degradation mechanisms of zirconium alloys in the nuclear industry comprise high-temperature oxidation, delayed hydride cracking, electrochemical corrosion of waterside of fuel pellets, and pipelines, and creep at elevated temperatures [2,11].…”

mentioning

confidence: 99%

“…Some Zr alloys are used in the nuclear industry for fuel claddings [1-3], reflectors in light water reactors [4], and in spent nuclear fuel reprocessing plants [5,6]. Zirconium alloys are applied to manufacture nuclear fuel pellets due to their low thermal neutron capture cross-section, proper strength properties, and excellent corrosion resistance in the cooling medium [7]. The nuclear fuel pellets are made of the Zr-Sn Zircaloys, Zr-Nb E110, E125, and E635, Zr-Sn-Nb Zirlo, Zr-Nb M5, and X5A alloys [7-10], Zr-Nb and Zr-Nb-Fe [11], and the 702 alloys [12,13].…”

mentioning

confidence: 99%

Hydrogen Embrittlement and Oxide Layer Effect in the Cathodically Charged Zircaloy-2

et al. 2020

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The present paper is aimed at determining the less investigated effects of hydrogen uptake on the microstructure and the mechanical behavior of the oxidized Zircaloy-2 alloy. The specimens were oxidized and charged with hydrogen. The different oxidation temperatures and cathodic current densities were applied. The scanning electron microscopy, X-ray electron diffraction spectroscopy, hydrogen absorption assessment, tensile, and nanoindentation tests were performed. At low oxidation temperatures, an appearance of numerous hydrides and cracks, and a slight change of mechanical properties were noticed. At high-temperature oxidation, the oxide layer prevented the hydrogen deterioration of the alloy. For nonoxidized samples, charged at different current density, nanoindentation tests showed that both hardness and Young’s modulus revealed the minims at specific current value and the stepwise decrease in hardness during hydrogen desorption. The obtained results are explained by the barrier effect of the oxide layer against hydrogen uptake, softening due to the interaction of hydrogen and dislocations nucleated by indentation test, and hardening caused by the decomposition of hydrides. The last phenomena may appear together and result in hydrogen embrittlement in forms of simultaneous hydrogen-enhanced localized plasticity and delayed hydride cracking.

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Current status of materials development of nuclear fuel cladding tubes for light water reactors

Cited by 269 publications

References 108 publications

Influence of composition and heating schedules on compatibility of FeCrAl alloys with high-temperature steam

Influence of composition and heating schedules on compatibility of FeCrAl alloys with high-temperature steam

Hydrogen Interaction with Deep Surface Modified Zr-1Nb Alloy by High Intensity Ti Ion Implantation

Hydrogen Embrittlement and Oxide Layer Effect in the Cathodically Charged Zircaloy-2

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