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dc.contributor.authorBrooks, Adam
dc.contributor.otherQueen's University (Kingston, Ont.). Theses (Queen's University (Kingston, Ont.))en
dc.date.accessioned2016-11-04T17:11:10Z
dc.date.available2016-11-04T17:11:10Z
dc.identifier.urihttp://hdl.handle.net/1974/15235
dc.description.abstractAs nuclear energy systems become more advanced, the materials encompassing them need to perform at higher temperatures for longer periods of time. In this Master’s thesis we experiment with an oxide dispersion strengthened (ODS) austenitic steel that has been recently developed. ODS materials have a small concentration of nano oxide particles dispersed in their matrix, and typically have higher strength and better extreme temperature creep resistance characteristics than ordinary steels. However, no ODS materials have ever been installed in a commercial power reactor to date. Being a newer research material, there are many unanswered phenomena that need to be addressed regarding the performance under irradiation. Furthermore, due to the ODS material traditionally needing to follow a powder metallurgy fabrication route, there are many processing parameters that need to be optimized before achieving a nuclear grade material specification. In this Master’s thesis we explore the development of a novel ODS processing technology conducted in Beijing, China, to produce solutionized bulk ODS samples with ~97% theoretical density. This is done using relatively low temperatures and ultra high pressure (UHP) equipment, to compact the mechanically alloyed (MA) steel powder into bulk samples without any thermal phase change influence or oxide precipitation. By having solutionized bulk ODS samples, transmission electron microscopy (TEM) observation of nano oxide precipitation within the steel material can be studied by applying post heat treatments. These types of samples will be very useful to the science and engineering community, to answer questions regarding material powder compacting, oxide synthesis, and performance. Subsequent analysis performed at Queen’s University included X-ray diffraction (XRD) and inductively coupled plasma optical emission spectrometry (ICP-OES). Additional TEM in-situ 1MeV Kr2+ irradiation experiments coupled with energy dispersive X-ray (EDX) techniques, were also performed on large (200nm+) non-stoichiometric oxides embedded within the austenite steel grains, in an attempt to quantify the elemental compositional changes during high temperature (520oC) heavy ion irradiation.en_US
dc.language.isoenen_US
dc.relation.ispartofseriesCanadian thesesen
dc.rightsQueen's University's Thesis/Dissertation Non-Exclusive License for Deposit to QSpace and Library and Archives Canadaen
dc.rightsProQuest PhD and Master's Theses International Dissemination Agreementen
dc.rightsIntellectual Property Guidelines at Queen's Universityen
dc.rightsCopying and Preserving Your Thesisen
dc.rightsThis publication is made available by the authority of the copyright owner solely for the purpose of private study and research and may not be copied or reproduced except as permitted by the copyright laws without written authority from the copyright owner.en
dc.subjectengineeringen_US
dc.subjectmaterialsen_US
dc.subjectnuclearen_US
dc.titleFABRICATION, CHARACTERIZATION, AND IRRADIATION OF AN AUSTENITIC OXIDE DISPERSION STRENGTHENED STEEL SUITED FOR NEXT GENERATION NUCLEAR APPLICATIONSen_US
dc.typethesisen_US
dc.description.degreeMaster of Applied Scienceen_US
dc.contributor.supervisorYao, Zhongwenen
dc.contributor.departmentMechanical and Materials Engineeringen


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