In-situ electrochemical study of Zr1nb alloy corrosion in high temperature Li +  containing water

Krausová, Aneta; Macák, Jan; Sajdl, Petr; Novotný, Radek; Renčiuková, Veronika; Vrtílková, Věra

doi:10.1016/j.jnucmat.2015.10.005

Cited by 24 publications

(9 citation statements)

References 23 publications

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“…Therefore, the oxide films on uncoated and hot-steam-treated ZIRLO were found to exhibit solely n-type semiconductive behavior. This result was consistent with observations of pure zirconium 22 and of Zr-Nb 23 and Zr-Sn alloys 10 – 14 , 24 . The results of the doping density analysis are presented in Table 2 , where N A and N D denote the acceptor and donor densities, respectively.…”

Section: Resultssupporting

confidence: 93%

“…SIMS is notable among chemical-analytical methods in that it combines high sensitivity and distinguishing capabilities with the capacity to analyze all elements and isotopes in the periodic table 23 . We used SIMS to determine the changes in the C depth profile of a 300 nm NCD-coated ZIRLO tube before and after 4 days of exposure to 400 °C hot steam.…”

Section: Resultsmentioning

confidence: 99%

“…It does not influence the slope of the linear part of the Mott-Schottky plot, from which the donor density is determined 23 .…”

Section: Experimental Methodsmentioning

confidence: 99%

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Nanocrystalline diamond protects Zr cladding surface against oxygen and hydrogen uptake: Nuclear fuel durability enhancement

Škarohlíd

Ashcheulov

Škoda

et al. 2017

Sci Rep

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In this work, we demonstrate and describe an effective method of protecting zirconium fuel cladding against oxygen and hydrogen uptake at both accident and working temperatures in water-cooled nuclear reactor environments. Zr alloy samples were coated with nanocrystalline diamond (NCD) layers of different thicknesses, grown in a microwave plasma chemical vapor deposition apparatus. In addition to showing that such an NCD layer prevents the Zr alloy from directly interacting with water, we show that carbon released from the NCD film enters the underlying Zr material and changes its properties, such that uptake of oxygen and hydrogen is significantly decreased. After 100–170 days of exposure to hot water at 360 °C, the oxidation of the NCD-coated Zr plates was typically decreased by 40%. Protective NCD layers may prolong the lifetime of nuclear cladding and consequently enhance nuclear fuel burnup. NCD may also serve as a passive element for nuclear safety. NCD-coated ZIRLO claddings have been selected as a candidate for Accident Tolerant Fuel in commercially operated reactors in 2020.

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Section: Resultssupporting

confidence: 93%

Section: Resultsmentioning

confidence: 99%

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Nanocrystalline diamond protects Zr cladding surface against oxygen and hydrogen uptake: Nuclear fuel durability enhancement

Škarohlíd

Ashcheulov

Škoda

et al. 2017

Sci Rep

Self Cite

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“…The notation [defect] is used to represent concentration of the defect species per unit formula. Figure 1 [34]. We also examined a range of μ Li which shows that the compensating mechanism does not change within the stability range.…”

Section: LImentioning

confidence: 94%

Electro-chemo-mechanical effects of lithium incorporation in zirconium oxide

Yang¹,

Youssef²,

Yildiz³

2018

Phys. Rev. Materials

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Understanding the response of functional oxides to extrinsic ion insertion is important for technological applications including electrochemical energy storage and conversion, corrosion, and electronic materials in neuromorphic computing devices. Decoupling the complicated chemical and mechanical effects of ion insertion is difficult experimentally. In this work, we assessed the effect of lithium incorporation in zirconium oxide as a model system, by performing first-principles based calculations. The chemical effect of lithium is to change the equilibria of charged defects. Lithium exists in ZrO 2 as a positively charged interstitial defect, and raises the concentration of free electrons, negatively charged oxygen interstitials, and zirconium vacancies. As a result, oxygen diffusion becomes faster by five orders of magnitude, and the total electronic conduction increases by up to five orders of magnitude in the low oxygen partial pressure regime. In the context of Zr metal oxidation, this effect accelerates oxide growth kinetics. In the context of electronic materials, it has implications for resistance modulations via ion incorporation. The mechanical effect of lithium is in changing the volume and equilibrium phase of the oxide. Lithium interstitials together with zirconium vacancies shrink the volume of the oxide matrix, release the compressive stress that is needed for stabilizing the tetragonal phase ZrO 2 at low temperature, and promote tetragonal-to-monoclinic phase transformation. By identifying these factors, we are able to mechanistically interpret experimental results in the literature for zirconium alloy corrosion in different alkali-metal hydroxide solutions. These results provide a mechanistic and quantitative understanding of lithium-accelerated corrosion of zirconium alloy, as well as, and more broadly, show the importance of considering coupled electro-chemo-mechanical effects of cation insertion in functional oxides.

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“…1) je patrná rychlá oxidace v prvním týdnu experimentu, poté následuje zpomalení růstu hodnot impedance. Takové chování je v souladu s předchozími experimenty, které byly s touto slitinou provedeny [13] i s obecnou představou o parabolickém/kubickém růstu oxidické vrstvy v pre-transientním stavu [14]. V impedančním spektru jsou patrné celkem tři časo-vé konstanty.…”

Section: Výsledky a Diskuzeunclassified

Chromium Coating as a Surface Protection of Zirconium Alloys

Krausová

Tůma

Novák

et al. 2017

Koroze a Ochrana Materialu

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Increasing of nuclear safety, higher demands for longer nuclear fuel campaign and higher levels of burnout are based on improving the properties of zirconium alloys. Protective coatings appears to be a promising way to reduce their chemical reactivity and increase resistance to hydriding. In this work, a thin chromium coating applied to a standard Zr1Nb zirconium alloy was studied using in-situ impedance spectroscopy. Exposure was carried out at a temperature of 280 ° C and a pressure of 8 MPa in a simulated WWER primary coolant environment. The results show that the chromium-coated Zr1Nb alloy is under these conditions oxidized significantly slower than the non-coated alloy.

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In-situ electrochemical study of Zr1nb alloy corrosion in high temperature Li + containing water

Cited by 24 publications

References 23 publications

Nanocrystalline diamond protects Zr cladding surface against oxygen and hydrogen uptake: Nuclear fuel durability enhancement

Nanocrystalline diamond protects Zr cladding surface against oxygen and hydrogen uptake: Nuclear fuel durability enhancement

Electro-chemo-mechanical effects of lithium incorporation in zirconium oxide

Chromium Coating as a Surface Protection of Zirconium Alloys

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