The terminal solid solubility of hydrogen in irradiated Zircaloy-2 and microscopic modeling of hydride behavior

Une, Katsumi; Ishimoto, Shinji; Etoh, Yoshinori; Ito, Kouichi; Ogata, Kohichi; Baba, Tetsuya; Kamimura, Kazuo; Kobayashi, Yasuo

doi:10.1016/j.jnucmat.2009.01.017

Cited by 49 publications

(7 citation statements)

References 18 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…This ::: that ::::::: reduces :: θ ::: to :: an :::::::::: apparently :::::: smaller :::::: volume ::::::: fraction :: of :::: alloy : θ apparent = 1 − φ/ f clamp :::::::::::::::::: (19) :::…”

Section: Hydride Precipitation and Dissolutionmentioning

confidence: 99%

“…where T I is the metal-oxide interface temperature C 1 is the rate constant for pre-transition oxidation Q 1 is the activation energy for pre-transition oxidation C 2 is the rate constant for post-transition oxidation Q 2 is the activation energy for post-transition oxidation R is the universal gas constant δ trans is the transition oxide thickness For normal operating temperatures below 673 K, constants from the EPRI/KWU/C-E model are used [15,16]: [17], (b) [18], (c) [19], and (d) [20]. Note that Sawatzky assumed that TSSp ≡ TSSd.…”

Section: A C C E P T E D Accepted Manuscriptmentioning

confidence: 99%

See 1 more Smart Citation

Multidimensional simulations of hydrides during fuel rod lifecycle

Stafford

2015

Journal of Nuclear Materials

View full text Add to dashboard Cite

Section: Hydride Precipitation and Dissolutionmentioning

confidence: 99%

Section: A C C E P T E D Accepted Manuscriptmentioning

confidence: 99%

Multidimensional simulations of hydrides during fuel rod lifecycle

Stafford

2015

Journal of Nuclear Materials

View full text Add to dashboard Cite

“…Puls' [46] theoretical prediction is that the TSSD would be slightly shifted and the TSSP largely shifted to higher content. On the other hand, the effect of neutron irradiation on the hydrogen solubility limit in zirconium alloys is still a matter of controversy since Une et al [99] did not observed an irradiation effect on the solubility limit on irradiated Zircaloy-2 at 5.4−12.9×10 25 n/m 2 .…”

Section: What Does Limit the Blister Growth?mentioning

confidence: 99%

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

Ménibus

Auzoux

Dieye

et al. 2014

Journal of Nuclear Materials

View full text Add to dashboard Cite

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.

show abstract

“…Figure 16. Solubility of hydrogen in Zr and its alloys [144,221,[234][235][236][237][238][239][240]. The bold line represents the average for thirteen studies on pure Zr and Zr-based alloys, reported in Ref.…”

Section: Why Barriers Are Needed For Fusion Reactorsmentioning

confidence: 99%

Tritium Barriers and Tritium Diffusion in Fusion Reactors

Causey

Karnesky

Marchi

2012

Comprehensive Nuclear Materials

132

View full text Add to dashboard Cite

Tritium and deuterium induce radiation in reactor materials and some radioactive tritium gas may be released. Most of the materials used in fusion reactors are metals that have relatively high permeabilities for tritium. The fusion community has been working on barriers to minimize tritium release. Unfortunately, most barrier materials work very well during laboratory experiments, but fail to meet requirements when placed in radiation environments.This chapter presents tritium permeation characteristics of various materials used in fusion reactors, including plasma-facing, structural, and barrier materials. The necessary parameters for tritium release calculations for various regions of a fusion reactor are given.

show abstract

The terminal solid solubility of hydrogen in irradiated Zircaloy-2 and microscopic modeling of hydride behavior

Cited by 49 publications

References 18 publications

Multidimensional simulations of hydrides during fuel rod lifecycle

Multidimensional simulations of hydrides during fuel rod lifecycle

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

Tritium Barriers and Tritium Diffusion in Fusion Reactors

Contact Info

Product

Resources

About