Behavior and failure of uniformly hydrided Zircaloy-4 fuel claddings between 25°C and 480°C under various stress states, including RIA loading conditions

Saux, Matthieu Le; Besson, Jacques; Carassou, S.; Poussard, C.; Averty, X.

doi:10.1016/j.engfailanal.2009.07.001

Cited by 75 publications

(51 citation statements)

References 56 publications

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“…The results are by some earlier reports [27,28]. The effects of interstitial hydrogen and hydride phases are opposite, but the distinct loss of plasticity observed in another studies [25,26,29,31] might appear at high hydrogen content and a great number of hydrides. The separation of these effects is not easy at low hydrogen content.…”

Section: Discussionsupporting

confidence: 67%

“…As the hydrogen content increased, an increase in the strength, a decrease in the ductility and the maximum stress were observed with increasing hydrogen content for Zircaloy-4 [31]. The ductility at room temperature decreased with increasing hydrogen content [26]. The room temperature strength of Zr-Sn-Fe-Nb tubes increased slightly, and the ductility decreased with the formation of the hydrides [32].…”

Section: Hydrogen Degradationmentioning

confidence: 85%

“…The Zr alloys are either strengthened or softened by hydrogen, depending on temperature and hydrogen content [26]. Hydrides as reported in [25] always degraded the tensile strength, ductility and fracture toughness.…”

Section: Hydrogen Degradationmentioning

confidence: 99%

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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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Section: Discussionsupporting

confidence: 67%

Section: Hydrogen Degradationmentioning

confidence: 85%

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Hydrogen degradation of pre-oxidized zirconium alloys

Szoka

Gajowiec

Zieliński

et al. 2017

Advances in Materials Science

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“…Sugiyama et al [3] results indicated that 60% and 85% of the cladding section was at a temperature lower than 350°C and 480°C respectively at cladding fracture. In laboratory tests that aimed at studying RIA conditions should therefore take into account the temperature level as it affects the material mechanical and thermal behavior [7] and the fracture process [10][11][12], the quantity of dissolved hydrides [13] and the extent of the irradiation recovery [14]. The temperature gradient in the cladding wall is likely one of the thermomechanical loading conditions that is most difficult to reproduce experimentally.…”

Section: Temperaturementioning

confidence: 99%

Thermomechanical loading applied on the cladding tube during the pellet cladding mechanical interaction phase of a rapid reactivity initiated accident

Ménibus¹,

Sercombe²,

Auzoux³

et al. 2014

Journal of Nuclear Materials

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International audienceCalculations of the CABRI REP-Na5 pulse were performed with the ALCYONE code in order to determine the evolution of the thermomechanical loading applied on the cladding tube during the Pellet-Cladding Mechanical Interaction (PCMI) phase of a rapid Reactivity Initiated Accident (RIA) initiated at 280 °C that lasted 8.8 ms. The evolution of the following parameters are reported: the cladding temperature, heating rate, strain rate and loading biaxiality. The impact of these parameters on the cladding mechanical behavior and fracture are then briefly reviewed

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“…The fuel pellets expand thermally and may cause fast straining of the surrounding Zircaloy clad tube through pellet-clad mechanical interaction. The fast loading imposed by the mechanical expansion of fuel pellet may cause rapid propagation of pre-existing cracks (Leclercq et al 2008;Chung and Kassner 1998;Saux et al 2010). These pre-existing defects may have been initiated during the service life due to mechanical and chemical interaction of fuel pellets with the clad.…”

Section: Introductionmentioning

confidence: 99%

Fracture behavior of thin-walled Zircaloy fuel clad tubes of Indian pressurized heavy water reactor

Sanyal

Samal

2012

Int J Fract

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For structural integrity analysis of thin walled tubular components such as fuel claddings used in nuclear reactors, knowledge of valid fracture mechanics parameters are required. As axiallycracked thin tubes possess a non-standard geometry from the point of view of specimen fabrication, there is no direct straight forward way to derive their fracture mechanics parameters with known functions as can be done for standard test methods. In the present work, the J-R curve for an axially cracked fuel pin from Indian 220 MWe pressurized heavy water reactor is derived after experimental and analytical derivation of the required parameters and the results are compared with earlier published data for similar material. List of symbolsa Crack length a 0 Crack length at the start of the test a bi Blunting corrected crack size at the ith data point of the load-normalization method b 0 Initial unbroken ligament length of the PLT specimen A Cracked area of the PLT specimen fracture surface A uc Area of the remaining ligament or un-cracked area A el Elastic part of area under the load-displacement curve A pl Plastic part of area under the load-displacement curve C Compliance of the PLT specimens E Young's modulus f (a/W ) Geometric shape function for evaluation of stress intensity factor as a function of a/W for a finite PLT specimen F L Limit load of the PLT specimens J J -integral J 0.2 J-integral corresponding to 0.2 mm crack growth J el Elastic part of J-integral J pl

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Behavior and failure of uniformly hydrided Zircaloy-4 fuel claddings between 25°C and 480°C under various stress states, including RIA loading conditions

Cited by 75 publications

References 56 publications

Hydrogen degradation of pre-oxidized zirconium alloys

Hydrogen degradation of pre-oxidized zirconium alloys

Thermomechanical loading applied on the cladding tube during the pellet cladding mechanical interaction phase of a rapid reactivity initiated accident

Fracture behavior of thin-walled Zircaloy fuel clad tubes of Indian pressurized heavy water reactor

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