Atomic Scale Degradation of Zirconium Alloys for Nuclear Applications
Doctoral thesis, 2015

Due to their low thermal neutron capture cross-section, zirconium alloys are widely used in the nuclear industry as fuel cladding and for structural components. The lifetime of the fuel assemblies in the reactors is primarily dictated by the ability of the cladding to withstand oxidation and hydrogen pick-up from the coolant water and radiation damage induced by the neutron flux. In order to study the hydrogenation and irradiation damage of zirconium on the atomic scale, atom probe tomography (APT) is utilized. This technique offers some unique virtues for nanometer scale materials analysis, such as equal sensitivity to all elements and near-atomic resolution. However, as APT has rarely been used for hydrogen studies previously, methods for accurate qualitative and quantitative analysis need to be developed. In this thesis, methods to control adsorption of hydrogen onto the APT specimen are explored. Techniques for hydrogen measurement are further developed using deuterium, whereby it is shown that hydrogen enters the alloy through grain boundaries in the oxide scale. A model for mitigation of hydrogen pick-up is proposed, in which oxide grain boundaries are decorated with transition metal alloying elements such as Fe and Ni, which affects the probability of reducing ingressing protons to form inert H2 gas. Zr alloys incur irradiation damage when subjected to the neutron flux in the reactor core, dissolving secondary phase particles and generating dislocation loops that embrittle the material. By studying the microstructure on the atomic scale before and after prolonged in-rector exposure, it is shown how the alloying elements in Zr interact with the irradiation-induced lattice defects.

Irradiation damage

Hydrogen microanalysis

Hydrogen pick-up

Zirconium alloys

Atom probe tomography


PJ Lecture Hall
Opponent: Dr. Ron Adamson


Gustav Sundell

Chalmers, Applied Physics, Materials Microstructure

Enrichment of Fe and Ni at metal and oxide grain boundaries in corroded Zircaloy-2

Corrosion Science,; Vol. 65(2012)p. 10-12

Other text in scientific journal

Hydrogen analysis in APT: Methods to control adsorption and dissociation of H2

Ultramicroscopy,; Vol. 132(2013)p. 285-289

Journal article

Redistribution of alloying elements in Zircaloy-2 after in-reactor exposure

Journal of Nuclear Materials,; Vol. 454(2014)p. 178-185

Journal article

Direct observation of hydrogen and deuterium in oxide grain boundaries in corroded Zirconium alloys

Corrosion Science,; Vol. 90(2015)p. 1-4

Other text in scientific journal

Tin clustering and precipitation in the oxide during autoclave corrosion of Zircaloy-2

Journal of Nuclear Materials,; Vol. 456(2015)p. 409-414

Journal article

Oxidation Mechanism in Zircaloy- 2—The Effect of SPP Size Distribution

ASTM Special Technical Publication: 17th International Symposium on Zirconium in the Nuclear Industry; Hyderabad, Andhra Pradesh; India; 3 February 2013 through 7 February 2013,; Vol. 1543(2015)p. 373-403

Paper in proceeding

Toward a Comprehensive Mechanistic Understanding of Hydrogen Uptake in Zirconium Alloys by Combining Atom Probe Analysis With Electronic Structure Calculations

ASTM 17th International Symposium on Zirconium in the Nuclear Industry, Hyderabad, Andhra Pradesh, India, 3-7 February 2013,; Vol. STP 1543(2015)p. 515-539

Paper in proceeding

Subject Categories

Physical Chemistry

Physical Sciences

Atom and Molecular Physics and Optics

Materials Chemistry

Condensed Matter Physics



Doktorsavhandlingar vid Chalmers tekniska högskola. Ny serie

PJ Lecture Hall

Opponent: Dr. Ron Adamson

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