Quantifying the source term: A mechanistic modelling approach for material and radioactivity transport

dc.contributor.advisorPalazhchenko, Olga
dc.contributor.authorBaker, Fiona Christine
dc.date.accessioned2024-10-03T14:15:18Z
dc.date.available2024-10-03T14:15:18Z
dc.date.issued2024-08
dc.description.abstractDuring nuclear reactor operation, dissolved and particulate corrosion products release from component walls followed by transport to the reactor core, where they deposit on fuel channels, undergoing irradiation for a certain residence time under the neutron flux. These corrosion products are released from the core and deposit throughout the primary heat transport system (PHTS). The Corrosion and Radioactivity Transport Analysis (CARTA) code is a predictive tool that uses a mechanistic approach to model surface and bulk activity within the PHTS of a Canada Deuterium Uranium (CANDU)-6 reactor. Two sources of radionuclide transport are modelled: a dissolved cobalt input term due to mechanical wear of the refuelling machine, and an out-of-core input due to changes from precipitation to dissolution in the steam generators, where previously deposited radionuclides are released from legacy deposits. These additions strengthen the code's adaptability to station events when simulating the kinetics of material release and activation processes.
dc.description.copyright©Fiona Christine Baker, 2024
dc.format.extentxiv, 100
dc.format.mediumelectronic
dc.identifier.urihttps://unbscholar.lib.unb.ca/handle/1882/38138
dc.language.isoen
dc.publisherUniversity of New Brunswick
dc.rightshttp://purl.org/coar/access_right/c_abf2
dc.subject.disciplineChemical Engineering
dc.titleQuantifying the source term: A mechanistic modelling approach for material and radioactivity transport
dc.typemaster thesis
oaire.license.conditionother
thesis.degree.disciplineChemical Engineering
thesis.degree.grantorUniversity of New Brunswick
thesis.degree.levelmasters
thesis.degree.nameM.Sc.E.

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