Correlation Between Fabrication Parameters, Microstructure, and Texture in Zircaloy Tubing

Moulin, Ludovic; Reschke, Stefan; Tenckhoff, E.

doi:10.1520/stp34473s

Cited by 9 publications

(4 citation statements)

References 13 publications

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“…The compound influence of texture and microstructure has been recognized for a long time. Case studies on zirconium, titanium as well as dispersion hardened aluminium alloys are known [Moulin (1984), Campagnac (1986) and Jensen (1985)]. Steel however, has received in this context a different kind of attention.…”

Section: Introductionmentioning

confidence: 99%

The Compound Influence of Textureand Microstructure on theMechanical Properties of Low‐C SteelWires

Gangli

Szpunar

Sugondo

1990

Texture, Stress, and Microstructure

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A series of experiments were made determining textural, microstructural, and mechanical properties in cold drawn, and spheroidization heat treated low-C steel wires (AISI-1018 and 1033 grades). It was found that texture exerted a significant influence on the mechanical properties, while microstructure had a comparable influence. Mechanical properties are represented by yield strength (YS), ultimate compressive strength (UCS) and by homogeneous strain energy (EHOM), defined by the integral of stress up to uniform elongation. Textural properties are represented by the Taylor-factor, M, the R-value, and by the maximum of the orientation distribution function (ODFMAX). Micro-structural properties are treated with the help of the aspect ratio parameter (I[AR), where AR is the grain aspect ratio (length to ellipsoidal width), the grain size parameter (1VD), and the mean free path between second phase spheroidized cementites X/N. For cold drawn steel wires, homogeneous strain energy (EHOM) is well correlated to (1//AR) and (ODFMAX). Yield strength, on the other hand, appears to be chiefly influenced by the aspect ratio parameter, thus here ODFMAX exerts less influence. The yield strength (YS) of annealed, spheroidization treated low-C wires are equally influenced by the grain size parameter (1/V'D), the mean distance between spherulites (X/N) and by ODFMAX.The textures of the cold drawn wires could be well described by the (110) fibre parallel to wire axis, and by the (111) fibre normal to wire axis. The annealed wires, while also featuring these two fibres, displayed a distinct {111}(110) single orientation.

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Section: Introductionmentioning

confidence: 99%

The Compound Influence of Textureand Microstructure on theMechanical Properties of Low‐C SteelWires

Gangli

Szpunar

Sugondo

1990

Texture, Stress, and Microstructure

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“…For Zircaloy cladding, a thermal treatment (usually <600°C) is the final processing step; it usually results in full recrystallization (manifested as equiaxed grains) or partial retention of the cold work (manifested as elongated grains along the rolling direction) from the final size-reducing treatment [18][19][20]. Fig.…”

Section: Electron Backscatter Diffractionmentioning

confidence: 99%

Fast, quantitative, and nondestructive evaluation of hydrided LWR fuel cladding by small angle incoherent neutron scattering of hydrogen

Yan¹,

Qian²,

Littrell³

et al. 2015

Journal of Nuclear Materials

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a b s t r a c tA nondestructive neutron scattering method to precisely measure the uptake of hydrogen and the distribution of hydride precipitates in light water reactor (LWR) fuel cladding was developed. Zircaloy-4 cladding used in commercial LWRs was used to produce hydrided specimens. The hydriding apparatus consists of a closed stainless-steel vessel that contains Zr alloy specimens and hydrogen gas. Following hydrogen charging, the hydrogen content of the hydrided specimens was measured using the vacuum hot extraction method, by which the samples with desired hydrogen concentrations were selected for the neutron study. Optical microscopy shows that our hydriding procedure results in uniform distribution of circumferential hydrides across the wall thickness. Small angle neutron incoherent scattering was performed in the High Flux Isotope Reactor at Oak Ridge National Laboratory. Our study demonstrates that the hydrogen in commercial Zircaloy-4 cladding can be measured very accurately in minutes by this nondestructive method over a wide range of hydrogen concentrations from a very small amount (%20 ppm) to over 1000 ppm. The hydrogen distribution in a tube sample was obtained by scaling the neutron scattering rate with a factor determined by a calibration process using standard, destructive direct chemical analysis methods on the specimens. This scale factor can be used in future tests with unknown hydrogen concentrations, thus providing a nondestructive method for determining absolute hydrogen concentrations.

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“…Dilute zirconium alloys, with minor alloying additions, meet most of the considerations made in selection of nuclear structural materials (Dietz 1994;Lemaignan & Motta 1994;Banerjee et al 2002;Williams 1970;Ibrahim & Cheadle 1985;Coleman et al 1996;Krishnan & Asundi 1981;Cox 1990;Karlsen & Vitanza 1994;Harding 1993;Bai et al 1991;Moulin et al 1984). Zirconium alloys like zircaloys (Williams 1970;Ibrahim & Cheadle 1985;Coleman et al 1996;Krishnan & Asundi 1981), Zr-Nb alloy (Lemaignan & Motta 1994;Williams 1970;Ibrahim & Cheadle 1985;Coleman et al 1996), Zr-1.2Cr alloy, Ozhennite-0.5 (Williams 1970), Zirlo (Moulin et al 1984) and Excel (Ibrahim & Cheadle 1985) are being used or developed for use in aqueous, organic and gaseous environment. Zirconium alloys, which have been extensively used in water cooled reactors are zircaloys and Zr-1-2.5 wt.% Nb alloys.…”

Section: Development Of Zirconium Alloysmentioning

confidence: 99%

“…Zirconium alloys, which have been extensively used in water cooled reactors are zircaloys and Zr-1-2.5 wt.% Nb alloys. The zircaloys (Krishnan & Asundi 1981;Cox 1990;Karlsen & Vitanza 1994;Harding 1993;Bai et al 1991;Moulin et al 1984), a series of alloy with tin as primary alloying additions, are widely used as cladding material in both pressurized water and boiling water reactors, and as cladding material, pressure tube and calandria tube material in PHWRs (Dietz 1994;Lemaignan & Motta 1994). Typical compositions of zircaloys are given in table 2 (Krishnan & Asundi 1981).…”

Section: Development Of Zirconium Alloysmentioning

confidence: 99%

Material development for India’s nuclear power programme

Suri

2013

Sadhana

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The area of materials research has registered a phenomenal growth in the recent years, assiduously accepting and assimilating ideas, concepts and analytical as well as experimental methodologies and techniques form almost all scientific disciplines, thereby demonstrating its remarkably multidisciplinary and interdisciplinary character. The focus of the materials programme of this centre is to provide materials, processes and processing solutions to the emerging needs of evolving indigenous nuclear energy systems by proactive research and development on a continuing basis. The initial stage of our activities was formulated around three stage Indian nuclear power programme. In stage I, material issues related to in-core materials with emphasis on development of fabrication routes of zirconium alloys for structural application were addressed. Subsequently the thrust areas were development and characterization of mixed oxide fuel, advanced zirconium alloys, structural steels, superalloys, neutron absorber materials based on boron carbides and borides, and shape memory alloys. The research was useful for in-service performance evaluation, safety assessment, residual life estimation and life extension of nuclear reactors built during stage I i.e., PHWRs and BWRs. It also included developments which would permit rapid expansion of nuclear power initially through fast breeder reactor based on mixed oxide fuel and later based on metallic fuels. For the 3rd stage, multilayer coatings, graphite coolant tube, BeO, refractory metals and alloys, heat-treated zirconium alloys are being developed for CHTR, ADSS and AHWR. The materials being developed for fusion programme are low Z and high Z material for plasma facing application, Cu-alloys for heat sink, austenitic steels, RAFMS and ODS for structurals and NbTi, Nb 3 Sn and Nb 3 Al superconductors, lithium titanate, lithium silicate breeders, and Pb-Bi coolant. A brief overview of the materials research activities currently being pursued at Bhabha Atomic Research Centre is presented in this article.

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Correlation Between Fabrication Parameters, Microstructure, and Texture in Zircaloy Tubing

Cited by 9 publications

References 13 publications

The Compound Influence of Textureand Microstructure on theMechanical Properties of Low‐C SteelWires

The Compound Influence of Textureand Microstructure on theMechanical Properties of Low‐C SteelWires

Fast, quantitative, and nondestructive evaluation of hydrided LWR fuel cladding by small angle incoherent neutron scattering of hydrogen

Material development for India’s nuclear power programme

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