A thermodynamic database for zirconium alloys

Dupin, N.; Ansara, I.; Servant, C.; Toffolon, C.; Lemaignan, C.; Brachet, Jean-Christophe

doi:10.1016/s0022-3115(99)00125-7

Cited by 171 publications

(127 citation statements)

References 41 publications

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“…To reduce data noise, the cartography was filtered with an empirically chosen 5×5 Gaussian filter. The stoichiometry of δ-hydrides considered at 25 • C is not ZrH 1.66 but ZrH 1.557 , as suggested by Puls [46] and confirmed with Thermo-Calc software [61] thermodynamical calculations reported in [62]. The volumetric fraction was calculated with a zirconium density equal to 6.500 g/cm 3 [63] and a δ-hydride density equal to 5.646 g/cm 3 [64].…”

Section: Hydrogen Quantification By Erdamentioning

confidence: 67%

“…They measured the hydrogen content by Nuclear Reaction Analysis on a blister formed in a Zircaloy-2 pressure tube that had been in service for 13 years. The blister reached 2.9±0.4wt% in deuterium content close to the external surface, which is a bit less than the α+δ → δ transition point at 25 • C, which is equal to 3.4wt% (based on Thermo-Calc software calculations [61] and converting the hydrogen to deuterium limit by multiplying by two). The thermal conditions to obtain such a blister are nevertheless not representative of PWR conditions, which might result in different hydrogen distributions.…”

Section: Comparison Of Laboratory To Reactor Grown Blistersmentioning

confidence: 93%

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Formation and characterization of hydride blisters in Zircaloy-4 cladding tubes

Ménibus

Auzoux

Dieye

et al. 2014

Journal of Nuclear Materials

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This article is focused on the formation of hydride blisters in zirconium alloys an experimental and theoritical standpoint, and their characterization in terms of morphology, hydrides crystallographic phases, hardness and hydrogen concentration. An experimental setup was developed to grow hydride blisters on pre-hydrided Zircaloy-4 cladding tubes by thermo-diffusion. The thermal conditions were optimized based on thermo-diffusion calculations, that take into account the hysteresis in the hydrogen solubility limit, to obtain a high blister growth rate. Micro X-Ray Diffraction (XRD), nano-hardness and Elastic Recoil Detection Analysis (ERDA) showed that the blisters contain a hydrogen gradient, with pure δ-hydride phase close to the external surface over one third of the blister depth. thermo-diffusion * Tel.: +33 1 69 08 39 43; e-mail: arthur.hellouin-de-menibus@cea.fr 1 calculations showed these half thickness blisters should grow in only a few days in PWR conditions. Eventually, the Diffusion Equilibrium Threshold (DET) was defined as a criterion that limits the blister growth, and emphasizes that the hysteresis in the hydrogen solubility limit in zirconium must be taken into account to model hydrogen thermo-diffusion in zirconium alloys.

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Section: Hydrogen Quantification By Erdamentioning

confidence: 67%

Section: Comparison Of Laboratory To Reactor Grown Blistersmentioning

confidence: 93%

Formation and characterization of hydride blisters in Zircaloy-4 cladding tubes

Ménibus

Auzoux

Dieye

et al. 2014

Journal of Nuclear Materials

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“…The value of offset yield strength was determined for the case when hydrogenated zirconium was not subjected to external stresses in the works [22,23]. The Zr-H state diagram [24] has shown that the main hydride phase is α-zirconium during the slow cooling of hydrogenated zirconium samples. The influence of external stress was studied in [25].…”

Section: Introductionmentioning

confidence: 99%

Mechanical Property Testing of Hydrogenated Zirconium Irradiated with Electrons

Kudiiarov

Ларионов

Tyurin

2018

Metals

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Abstract:The mechanical properties of the hydrogenated zirconium alloy Zr-1Nb are studied under different conditions for hydrogen removal by an electron beam and thermal heating. The mechanical testing of zirconium samples is analyzed during hydrogenation and irradiation with a low energy electron beam. The plasticity of the samples is shown to be increased during the radiation stimulation of hydrogen removal from zirconium by even a weak electron beam. In this case, the tensile strength (ultimate strength) is practically not changed.

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“…These alloys, known as zircaloys are used in the nuclear power industry for structural parts in the core; such as pressure tubes, channels, guide tubes or fuel cladding [2] due to its low thermal neutron cross section [3]. The mineral zircon (ZrSiO 4 ) is the primary source of zirconium metal.…”

Section: Introductionmentioning

confidence: 99%

Microstructural and analytical analysis of plasma dissociated zircon

Minnaar

Neethling

Lee

et al. 2012

J. Phys.: Conf. Ser.

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Abstract. The investigation of the microstructure and distribution of impurities was carried out on plasma dissociated zircon (PDZ). The morphology of the PDZ and crystalline nature of the zirconia was determined by scanning electron microscopy (SEM) and transmission electron microscopy (TEM) while energy dispersive X-ray spectrometry (EDS) provided compositional information. The SEM and TEM results indicate that the morphology and crystalline nature of the zirconia varies as a result of the existence of a thermal gradient. The EDS results show that the majority of the impurities segregates to the silica phase of the PDZ.

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A thermodynamic database for zirconium alloys

Cited by 171 publications

References 41 publications

Formation and characterization of hydride blisters in Zircaloy-4 cladding tubes

Formation and characterization of hydride blisters in Zircaloy-4 cladding tubes

Mechanical Property Testing of Hydrogenated Zirconium Irradiated with Electrons

Microstructural and analytical analysis of plasma dissociated zircon

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