Irradiation Growth

Griffiths, M.

doi:10.1016/b978-0-12-803581-8.11646-7

Cited by 6 publications

(5 citation statements)

References 73 publications

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“…By taking the slope of the curve from a plot of strain as a function of dose and dividing it by the stress, one obtains the creep compliance ( B 0 ) at each stage of the creep evolution provided it is assumed (reasonably) that the creep strain would be zero at zero stress. This is only true provided the material is not swelling and exhibits no irradiation growth [ 37 ]. Taking the Lewthwaite and Mosedale data [ 36 ] and plotting the compliance against dose shows that the secondary (“steady-state”) creep compliance ( B 0 ) decreases as the dose increases and reaches a true steady state (constant) condition after about 5–10 dpa for the various austenitic stainless steels [ 8 ].…”

Section: Microstructure Effects On Irradiation Creepmentioning

confidence: 99%

“…The main difference between Zr-alloys and other austenitic engineering alloys is that they exhibit irradiation growth because of the intrinsic crystallographic anisotropy, Zr-alloys being hexagonal close-packed (HCP) compared with face-centred-cubic (FCC) austenitic alloys. Irradiation growth is not unique to HCP metals, as it can also be exhibited when the microstructure (dislocations in particular) is anisotropic because of fabrication [ 37 ]. The intrinsic radiation response (irradiation growth) of Zr alloys itself evolves as the microstructure evolves.…”

Section: Microstructure Effects On Irradiation Creepmentioning

confidence: 99%

“…The intrinsic radiation response (irradiation growth) of Zr alloys itself evolves as the microstructure evolves. In this case, the appearance of vacancy basal plane dislocation loops in the microstructure (either discrete loops or from climb on c+a screw dislocations) leads to shrinkage along the c-axis, irrespective of the stress state [ 10 , 37 ], and is manifested as a non-linear strain when irradiation growth is plotted against the irradiation dose [ 10 , 37 ]. The growth itself is manifested as a non-zero intercept at zero stress in plots of irradiation creep rate against stress [ 10 ].…”

Section: Microstructure Effects On Irradiation Creepmentioning

confidence: 99%

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Microstructural Effects on Irradiation Creep of Reactor Core Materials

Griffiths

2023

Materials

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The processes that control irradiation creep are dependent on the temperature and the rate of production of freely migrating point defects, affecting both the microstructure and the mechanisms of mass transport. Because of the experimental difficulties in studying irradiation creep, many different hypothetical models have been developed that either favour a dislocation slip or a mass transport mechanism. Irradiation creep mechanisms and models that are dependent on the microstructure, which are either fully or partially mechanistic in nature, are described and discussed in terms of their ability to account for the in-reactor creep behaviour of various nuclear reactor core materials. A rate theory model for creep of Zr-2.5Nb pressure tubing in CANDU reactors incorporating the as-fabricated microstructure has been developed that gives good agreement with measurements for tubes manufactured by different fabrication routes having very different microstructures. One can therefore conclude that for Zr-alloys at temperatures < 300 °C and stresses < 150 MPa, diffusional mass transport is the dominant creep mechanism. The most important microstructural parameter controlling irradiation creep for these conditions is the grain structure. Austenitic alloys follow similar microstructural dependencies as Zr-alloys, but up to higher temperature and stress ranges. The exception is that dislocation slip is dominant in austenitic alloys at temperatures < 100 °C because there are few barriers to dislocation slip at these low temperatures, which is linked to the enhanced recombination of irradiation-induced point defects.

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Section: Microstructure Effects On Irradiation Creepmentioning

confidence: 99%

Section: Microstructure Effects On Irradiation Creepmentioning

confidence: 99%

Section: Microstructure Effects On Irradiation Creepmentioning

confidence: 99%

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Microstructural Effects on Irradiation Creep of Reactor Core Materials

Griffiths

2023

Materials

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“…Irradiation creep is a specific deformation phenomenon that occurs under an applied stress and is activated by the fast neutron flux 4 . Irradiation induced growth is a special deformation phenomenon because it occurs under fast neutron flux without any applied stress 5,6 . Several authors 7,8 have proposed empirical macroscopic models to account for the in-reactor deformation of zirconium alloys.…”

Section: Introductionmentioning

confidence: 99%

Polycrystalline Simulations of In-Reactor Deformation of Zircaloy-4 Cladding Tubes during Nominal Operating Conditions

Gicquel,

Onimus,

Brenner

et al. 2023

Zirconium in the Nuclear Industry: 20th International Symposium

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Fuel cladding tubes made of zirconium alloys, are subjected in reactor to a complex loading history under nominal operating conditions. Furthermore, they exhibit a complex deformation behavior resulting from irradiation-induced growth, irradiation creep and thermal creep. For design and safety requirements, empirical models are usually used. To have robust physically based mechanical simulations, a self-consistent polycrystalline model has been developed. This model takes into account the various phenomena occurring at the grain scale, such as irradiation-induced growth and irradiation creep. Moreover, this model takes into account the crystallographic texture of the material and the mechanical interactions between grains, depending on their orientation. Furthermore, this model is able to handle complex mechanical loading. This model is first shown to reproduce well an experimental database of in-reactor deformation of zirconium alloys. Thanks to the polycrystalline nature of this model, the effect of grain shape and creep mechanisms at the grain scale on the simulated data have been studied in detail. Next, this polycrystalline model has been introduced into a 1D finite element method code, allowing the computation of stress and strain gradients through a thin cladding tube during a complex mechanical loading. This approach opens the way to physically based mechanical calculations at the component scale.

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“…For instance, in the large dose limit, high densities of defects and dislocations accumulate in the cladding, causing embrittlement. Another important degradation mode is the so-called 'irradiation-induced growth' (IIG) that arises from the anisotropy of the hexagonal close packed (hcp) crystal structure of zirconium (𝛼-Zr) and zirconium alloys (Griffiths, 2020;Onimus et al, 2022). This crystal structure is stable up to and beyond the reactor core's operating temperature range of 280 • C to 350 • C, exhibiting an hcp-bcc instability only at significantly higher temperatures above 860 • C (Willaime and Massobrio, 1989).…”

Section: Introductionmentioning

confidence: 99%