Development of microstructure and irradiation hardening of Zircaloy during low dose neutron irradiation at nominally 358°C

Cockeram, B.V.; Smith, Richard W.; Leonard, Keith J.; Byun, Thak Sang; Snead, Lance Lewis

doi:10.1016/j.jnucmat.2011.07.006

Cited by 34 publications

(63 citation statements)

References 29 publications

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“…For example, in FCC copper, stacking fault tetrahedra, voids, and SIA loops are experimentally observed, [2][3][4][5] whereas in BCC a-iron, spherical voids, and SIA loops are observed, [5][6][7] and in HCP zirconium, only void and SIA loops are seen. [8] Defect morphology and spatial correlation are also dependent on other factors such as temperature, displacements per atom (DPA), and DPA rate.…”

Section: Introductionmentioning

confidence: 99%

Analysis of Obstacle Hardening Models Using Dislocation Dynamics: Application to Irradiation-Induced Defects

Sobie

Bertin

Capolungo

2015

Metall Mater Trans A

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Irradiation hardening in a-iron represents a critical factor in nuclear reactor design and lifetime prediction. The dispersed barrier hardening, Friedel Kroupa Hirsch (FKH), and Bacon Kocks Scattergood (BKS) models have been proposed to predict hardening caused by dislocation obstacles in metals, but the limits of their applicability have never been investigated for varying defect types, sizes, and densities. In this work, dislocation dynamics calculations of irradiationinduced obstacle hardening in the athermal case were compared to these models for voids, selfinterstitial atom (SIA) loops, and a combination of the two types. The BKS model was found to accurately predict hardening due to voids, whereas the FKH model was superior for SIA loops. For both loops and voids, the hardening from a normal distribution of defects was compared to that from the mean size, and was shown to have no statistically significant dependence on the distribution. A mean size approach was also shown to be valid for an asymmetric distribution of voids. A non-linear superposition principle was shown to predict the hardening from the simultaneous presence of voids and SIA loops.

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Section: Introductionmentioning

confidence: 99%

Analysis of Obstacle Hardening Models Using Dislocation Dynamics: Application to Irradiation-Induced Defects

Sobie

Bertin

Capolungo

2015

Metall Mater Trans A

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“…1. Cockeram et al observed that microstructure changes, strength intensification and uniform elongation reductions of zircaloy are a function of irradiation temperature and fluences [50]. Different from other metals and alloys, zirconium and its alloys exhibit only little void formation when they are exposed to neutron irradiation [34].…”

Section: Zirconiummentioning

confidence: 99%

Physical Ageing of The Research Reactor Core Structural Materials Due To Neutron Irradiation Exposure: A Review

Purba

2017

Jurnal Pengembangan Energi Nuklir

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PHYSICAL AGEING OF THE RESEARCH REACTOR CORE STRUCTURAL MATERIALS DUE TO NEUTRON IRRADIATION EXPOSURE: A REVIEW.A research reactor (RR) is a nuclear reactor that has function to generate and utilize neutron flux and radiation ionization for research purposes and industrial applications. More than 60% of current operating RRs have been operated for 30 years or more. As the time passes, the functional capabilities of structures, systems and components (SSCs) of those RRs deteriorate by physical ageing, which can be caused by neutron irradiation exposure such as irradiation induced dislocation and microstructural changes. To extend the lifetime and/or to avoid unplanned outages, ageing on the safety related SSCs of RRs need to be properly managed. An ageing management is a strategy to engineer, operate, maintenance, and control SSC degradation within acceptable limits. The purpose of this study is to review physical ageing of the core structural materials of the RRs caused by neutron irradiation exposure. In order to achieve this objective, a wide range of literatures are reviewed. Comprehensive discussions on irradiation behaviors are limited only on reactor vessel and core support structure materials made from zirconium and beryllium as well as their alloys, which are widely used in RRs. It is found that the stability of the mechanical properties of zirconium and beryllium as well as their alloys was mostly affected by the neutron fluences and temperatures. ABSTRAK PENUAAN MATERIAL STRUKTUR TERAS REAKTOR RISET SECARA FISIK YANG DISEBABKAN OLEH PAPARAN RADIASI NEUTRON: SEBUAH KAJIAN.Reaktor riset adalah reaktor nuklir yang difungsikan untuk pembangkitan dan pemanfaatan fluks neutron dan ionisasi radiasi untuk tujuan penelitian dan aplikasi industri. Lebih dari 60% reaktor riset yang ada saat ini telah beroperasi selama 30 tahun atau lebih. Seiring dengan berjalannya waktu, kemampuan struktur, sistem dan komponen (SSK) tentunya mengalami penurunan yang dikenal dengan penuaan fisik. Penuaan fisik ini dapat disebabkan oleh karena adanya paparan radiasi neutron seperti irradiasi yang menyebabkan terjadinya dislokasi dan perubahan mikrostruktur material. Untuk memperpanjang umur reaktor dan untuk mengurangi terjadinya pemadaman yang tidak diinginkan maka perlu dilakukan manajemen penuaan SSK, khususnya yang terkait dengan keselamatan operasi. Tujuan dari penelitian ini adalah untuk mengkaji penuaan fisik yang terjadi pada material struktur teras reaktor riset yang disebabkan oleh paparan iradiasi neutron. Untuk mencapai tujuan ini maka dilakukan kajian pada berbagai macam pustaka yang terkait. Pembahasan yang komprehensif tentang sifat-sifat iradiasi dibatasi hanya pada bejana reaktor dan material struktur pendukung teras yang terbuat dari zirconium dan beryllium dan senyawanya yang umum digunakan oleh reaktor riset. Dari kajian yang telah dilakukan, ditemukan bahwa stabilitas sifat-sifat mekanik zirconium dan beryllium serta senyawanya sangat dipengaruhi oleh fluent dan temperatur neutron. Kata kunci: Reaktor riset, p...

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“…Two zirconium-base alloys used in nuclear applications are Zircaloy-2 and Zircaloy-4 [1][2][3][4][5][6][7][8][9][10][11][12][13][14][15]. Both alloys contain nominally 1.5% Sn in solid-solution and consist primarily of a hexagonal alpha Zr-phase.…”

Section: Introductionmentioning

confidence: 99%

“…Zircaloy-2 contains low alloying levels of Ni, Fe, and Cr additions that are tied up as Laves phase precipitates (Zr(Fe, Cr) 2 ) or as Zintl phase precipitates (Zr 2 (Fe, Ni)) [1][2][3][4][5][6][7][8][9][10][11][12][13][14][15][16]. Zircaloy-4 contains only low alloying levels of Fe and Cr and contains only Laves phase precipitates [1][2][3][4][5][6][7][8][9][10][11][12][13][14][15][16]. The neutron irradiation of Zircaloy results in the formation of dislocation loops that are barriers to dislocation motion that result in irradiation hardening .…”

Section: Introductionmentioning

confidence: 99%

“…Irradiation hardening is primarily the result of the high number density of small hai loops (4-30 nm diameter) that are formed on the prism planes (f1 0 1 0g) with a Burgers vector b hai ¼ 1 3 h1 1 2 0i [1-3,6-10, [13][14][15][16][24][25][26][27][28][29][30][31][32][33]. These hai loops are nucleated after a neutron fluence of 0.3-1.1 Â 10 24 n/m 2 (E > 1 MeV) and then rapidly saturate to a relatively constant size/number density at a fluence of about 1-5 Â 10 25 n/m 2 , depending on the irradiation temperature [1][2][3][4][14][15][16]. The hai loops are both interstitial or vacancy in nature, and the fraction of loop type is a strong function of irradiation temperature, with the majority being an interstitial type for irradiations at a temperature of 300°C or lower, about 50% interstitial for irradiation at 350°C, and about 30% interstitial for irradiation at 400°C [6,14,30,32].…”

Section: Introductionmentioning

confidence: 99%

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The recovery of irradiation damage for Zircaloy-2 and Zircaloy-4 following low dose neutron irradiation at nominally 358 °C

Cockeram

Leonard

Byun

et al. 2015

Journal of Nuclear Materials

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a b s t r a c tThe recovery of irradiation damage in wrought Zircaloy-2 and Zircaloy-4 was determined following a series of post-irradiation anneals at temperatures ranging from 343°C to 510°C and for time periods ranging from 1-h to 500 h. The materials had been irradiated at nominally 358°C in the High Flux Isotope Reactor (HFIR) at neutron fluences of nominally 3 Â 10 25 n/m 2 (E > 1 MeV). Irradiation at nominally 358°C resulted in a coarser distribution of hai loops that result in a 25-45% lower irradiation hardening than reported in the literature for irradiations at 260-326°C. The irradiation hardening and recovery were determined using tensile testing at room-temperature. Post-irradiation annealing at 343-427°C was shown to result in an increase in irradiation hardening to values even higher than for the as-irradiated material in the first 1-10 h of annealing. This Radiation Anneal Hardening (RAH) was followed by a relatively slow recovery of the irradiation damage. Much faster recovery with no RAH was observed for post-irradiation annealing at temperatures of 454-510°C. Irradiation at 358°C was shown to result in different recovery kinetics than observed in the literature for irradiation at 260-326°C. While the general trend described above is true for the four materials tested (alpha-annealed and beta-treated Zircaloy-2 and Zircaloy-4), notable and yet unexplained differences in RAH and in recovery are observed between the materials that might be a result of differing solute effects. Examinations of microstructure using Transmission Electron Microscopy were used to investigate the RAH and recovery mechanisms. Agreement between the measured and calculated irradiation hardening using a generalized Orowan hardening model to account for the observed loop structure was not as close for the post irradiation annealed condition as for the as-irradiated condition, which can likely be attributed to unaccounted for changes in the configuration of the hai loops to dislocation lines, segregation of solutes to dislocation loops, and the potential for the formation of fine clusters of point defects or solutes during annealing.

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Development of microstructure and irradiation hardening of Zircaloy during low dose neutron irradiation at nominally 358°C

Cited by 34 publications

References 29 publications

Analysis of Obstacle Hardening Models Using Dislocation Dynamics: Application to Irradiation-Induced Defects

Analysis of Obstacle Hardening Models Using Dislocation Dynamics: Application to Irradiation-Induced Defects

Physical Ageing of The Research Reactor Core Structural Materials Due To Neutron Irradiation Exposure: A Review

The recovery of irradiation damage for Zircaloy-2 and Zircaloy-4 following low dose neutron irradiation at nominally 358 °C

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