Ductile-to-brittle transition of oxidised Zircaloy-4 and E110 claddings

Hózer, Zoltán; Csaba, Gyori; Matus, L.; Horváth, Márta

doi:10.1016/j.jnucmat.2007.07.002

Cited by 58 publications

(28 citation statements)

References 11 publications

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“…The hydrogen uptake of the remaining zirconium alloy during isothermal steam oxidation was studied experimentally in [5][6][7][8][9]. As [7,8] have shown hydrogen concentration in the remaining zirconium alloy is strongly influenced by the stability of the oxide layer formed during steam oxidation.…”

Section: Introductionmentioning

confidence: 99%

“…As [7,8] have shown hydrogen concentration in the remaining zirconium alloy is strongly influenced by the stability of the oxide layer formed during steam oxidation. At certain temperatures depending on the alloy composition the oxide growth starts with an undercooled tetragonal structure.…”

Section: Introductionmentioning

confidence: 99%

“…In [7] it was reported that the hydrogen concentration is as higher as earlier the breakaway starts. In [8] the hydrogen uptake of the classical cladding material Zry-4 and the E110 used in Russian type VVER reactors is compared. Whereas hydrogen is detected in Zry-4 only at 1273 K up to 30 at.% hydrogen was found in E110 at all temperatures investigated between 1173 and 1473 K.…”

Section: Introductionmentioning

confidence: 99%

See 2 more Smart Citations

Influence of oxide layer morphology on hydrogen concentration in tin and niobium containing zirconium alloys after high temperature steam oxidation

Große

Lehmann

Steinbrück

et al. 2009

Journal of Nuclear Materials

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Section: Introductionmentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

See 1 more Smart Citation

Influence of oxide layer morphology on hydrogen concentration in tin and niobium containing zirconium alloys after high temperature steam oxidation

Große

Lehmann

Steinbrück

et al. 2009

Journal of Nuclear Materials

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“…The breakaway oxidation in steam (without nitrogen) occurs at temperatures up to 1300K [10] and after time ranged between 2400 and 5500 s [6]. In other research on the same alloy, the oxide layer was indeed compact at 1200°C, even if the oxide scale formed at 1000°C indicated breakaway phenomena [11]. Oxides grown on pure zirconium were less permeable that those on multiphase alloys because of a greater number of grain boundaries, interfaces and diffusion pathways [12].…”

Section: Oxidationmentioning

confidence: 96%

Hydrogen degradation of pre-oxidized zirconium alloys

Szoka

Gajowiec

Zieliński

et al. 2017

Advances in Materials Science

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The presence of the oxide layers on Zr alloys may retard or enhance the hydrogen entry and material degradation, depending on the layer features. This research has been aimed to determine the effects of preoxidation of the Zircaloy-2 alloy at a different temperature on hydrogen degradation. The specimens were oxidised in laboratory air at 350ºC, 700ºC, and 900ºC. After, some samples were tensed at 10-5 strain rate and simultaneously charged with hydrogen under constant direct voltage in 1 N sulfuric acid at room temperature. Other specimens were charged without any tension, then annealed at 400ºC for 4 h and finally tensed at above strain rate. The SEM examinations were performed on the cross-sections and fracture faces of specimens. The obtained results demonstrate the effects of the oxide layer on the cathodic current and hydrogen entry, mechanical properties and the appearance of hydrides and fracture behaviour.

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“…4,5) The tests have recently been performed with prehydrided nonirradiated cladding and high-burnup fuel cladding (70 GWd/t, <700 ppm) in order to confirm the safety of the high-burnup fuel and revise the criteria if necessary. [6][7][8][9][10][11] As a result, hydrogen produced the main burnup effect on embrittlement and the zero-ductility was seen at the oxidations much lower than the safety limit when the cladding absorbed hundreds of ppm of hydrogen, which is inconsistent with the results from the LOCA-simulated experiments.…”

Section: Introductionmentioning

confidence: 99%

Ring Compression Ductility of High-Burnup Fuel Cladding after Exposure to Simulated LOCA Conditions

Nagase

Chuto

Fuketa

2011

Journal of Nuclear Science and Technology

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Ring compression tests were conducted with specimens sampled from high-burnup PWR and BWR fuel cladding segments that had been ruptured, oxidized at 1,405-1,484 K, and quenched in LOCAsimulated experiments. The calculated oxidation amount for cladding thickness ranged from 11.3 to 25.3% and the hydrogen concentration ranged from 100 to 870 ppm in the ring specimens. The plastic strain to failure and maximum load measured in the ring compression tests decrease with increasing oxidation and hydrogen. Embrittlement of the cladding is seen when the hydrogen concentration is above 300-400 ppm. Although the examined fuel cladding segments did not fracture in the LOCA-simulated experiments, most of the specimens sampled from the segments exhibited a brittle nature in the ring compression tests. This obvious discrepancy between the fracture/no-fracture criterion and the embrittlement criterion is likely caused by a difference in the loading conditions in the two tests. The advantages and drawbacks of the test methodologies should be well considered on application of the test results to the regulatory judgment.

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Ductile-to-brittle transition of oxidised Zircaloy-4 and E110 claddings

Cited by 58 publications

References 11 publications

Influence of oxide layer morphology on hydrogen concentration in tin and niobium containing zirconium alloys after high temperature steam oxidation

Influence of oxide layer morphology on hydrogen concentration in tin and niobium containing zirconium alloys after high temperature steam oxidation

Hydrogen degradation of pre-oxidized zirconium alloys

Ring Compression Ductility of High-Burnup Fuel Cladding after Exposure to Simulated LOCA Conditions

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