{"id":{"repo_id":"uoit","oai_identifier":"oai:ontariotechu.scholaris.ca:10155/87"},"canonical_url":"https://search.dev.ndltd.org/etd/uoit/oai:ontariotechu.scholaris.ca:10155/87","repository":{"repo_id":"uoit","name":"Ontario Institute of Technology","base_url":"https://ontariotechu.scholaris.ca/server/oai/request"},"display":{"title":"Transport-theory-equivalent diffusion coefficients for node-homogenized neutron diffusion problems in CANDU lattices","abstract":"Calculation of the neutron flux in a nuclear reactor core is ideally performed by solving the neutron transport equation for a detailed-geometry model using several tens of energy groups. However, performing such detailed calculations for an entire core is prohibitively expensive from a computational perspective. Full-core neutronic calculations for CANDU reactors are therefore performed customarily using two-energy-group diffusion theory (no angular dependence) for a node-homogenized reactor model. The work presented here is concerned with reducing the loss in accuracy entailed when going from Transport to Diffusion. To this end a new method of calculating the diffusion coefficient was developed, based on equating the neutron balance equation expressed by the transport equation with the neutron balance equation expressed by the diffusion equation. The technique is tested on a simple twelve-node model and is shown to produce transport-like accuracy without the associated computational effort.","abstract_html":"Calculation of the neutron flux in a nuclear reactor core is ideally performed by solving the neutron transport equation for a detailed-geometry model using several tens of energy groups. However, performing such detailed calculations for an entire core is prohibitively expensive from a computational perspective. Full-core neutronic calculations for CANDU reactors are therefore performed customarily using two-energy-group diffusion theory (no angular dependence) for a node-homogenized reactor model. The work presented here is concerned with reducing the loss in accuracy entailed when going from Transport to Diffusion. To this end a new method of calculating the diffusion coefficient was developed, based on equating the neutron balance equation expressed by the transport equation with the neutron balance equation expressed by the diffusion equation. The technique is tested on a simple twelve-node model and is shown to produce transport-like accuracy without the associated computational effort.","abstract_has_math":false,"creators":["Patel, Amin"],"institution":"University of Ontario Institute of Technology","degree_name":"Master of Science (MSc)","degree_level":null,"degree_discipline":"Modelling and Computational Science","degree_department":null,"school":null,"contributors":[],"advisors":["Nichita, Eleodor M."],"committee_chairs":[],"committee_members":[],"year":2010,"date_issued":"2010-04-01","date_published":"2010-04-01","updated_at":"2026-07-24T05:35:30Z","subjects":["Applied reactor physics","Transport theory","Diffusion theory","CANDU","Nuclear reactor"],"languages":["en"],"rights":[],"rights_urls":[],"identifier_entries":[]},"links":{"outbound_url":"https://hdl.handle.net/10155/87","outbound_label":"Handle","outbound_source":"dc:identifier.uri"},"metadata_groups":[{"id":"people","label":"People","entries":[{"key":"dc:contributor.advisor","label":"Advisor","values":["Nichita, Eleodor M."]},{"key":"dc:creator","label":"Author","values":["Patel, Amin"]}]},{"id":"academic_context","label":"Academic Context","entries":[{"key":"dc:date.accessioned","label":"Dc Date Accessioned","values":["2010-06-04T16:20:34Z","2022-03-29T17:06:13Z"]},{"key":"dc:date.available","label":"Dc Date Available","values":["2010-06-04T16:20:34Z","2022-03-29T17:06:13Z"]},{"key":"dc:date.issued","label":"Date","values":["2010-04-01"]},{"key":"dc:type","label":"Dc Type","values":["Thesis"]},{"key":"thesis:degree_discipline","label":"Discipline","values":["Modelling and Computational Science"]},{"key":"thesis:degree_name","label":"Degree Name","values":["Master of Science (MSc)"]},{"key":"thesis:institution_name","label":"Thesis Institution Name","values":["University of Ontario Institute of Technology"]}]},{"id":"subjects_keywords","label":"Subjects and Keywords","entries":[{"key":"dc:subject","label":"Dc Subject","values":["Applied reactor physics","Transport theory","Diffusion theory","CANDU","Nuclear reactor"]}]},{"id":"language_rights","label":"Language and Rights","entries":[{"key":"dc:language.iso","label":"Language (ISO)","values":["en"]}]},{"id":"identifiers","label":"Identifiers","entries":[{"key":"dc:identifier.uri","label":"Identifier URI","values":["https://hdl.handle.net/10155/87"]}]},{"id":"additional","label":"Additional Metadata","entries":[{"key":"dc:description.abstract","label":"Abstract","values":["Calculation of the neutron flux in a nuclear reactor core is ideally performed by solving the neutron transport equation for a detailed-geometry model using several tens of energy groups. However, performing such detailed calculations for an entire core is prohibitively expensive from a computational perspective. Full-core neutronic calculations for CANDU reactors are therefore performed customarily using two-energy-group diffusion theory (no angular dependence) for a node-homogenized reactor model. The work presented here is concerned with reducing the loss in accuracy entailed when going from Transport to Diffusion. To this end a new method of calculating the diffusion coefficient was developed, based on equating the neutron balance equation expressed by the transport equation with the neutron balance equation expressed by the diffusion equation. The technique is tested on a simple twelve-node model and is shown to produce transport-like accuracy without the associated computational effort."]},{"key":"dc:title","label":"Title","values":["Transport-theory-equivalent diffusion coefficients for node-homogenized neutron diffusion problems in CANDU lattices"]}]}],"canonical_facts":{"dc:contributor.advisor":["Nichita, Eleodor M."],"dc:creator":["Patel, Amin"],"dc:date.accessioned":["2010-06-04T16:20:34Z","2022-03-29T17:06:13Z"],"dc:date.available":["2010-06-04T16:20:34Z","2022-03-29T17:06:13Z"],"dc:date.issued":["2010-04-01"],"dc:description.abstract":["Calculation of the neutron flux in a nuclear reactor core is ideally performed by solving the neutron transport equation for a detailed-geometry model using several tens of energy groups. However, performing such detailed calculations for an entire core is prohibitively expensive from a computational perspective. Full-core neutronic calculations for CANDU reactors are therefore performed customarily using two-energy-group diffusion theory (no angular dependence) for a node-homogenized reactor model. The work presented here is concerned with reducing the loss in accuracy entailed when going from Transport to Diffusion. To this end a new method of calculating the diffusion coefficient was developed, based on equating the neutron balance equation expressed by the transport equation with the neutron balance equation expressed by the diffusion equation. The technique is tested on a simple twelve-node model and is shown to produce transport-like accuracy without the associated computational effort."],"dc:identifier.uri":["https://hdl.handle.net/10155/87"],"dc:language.iso":["en"],"dc:subject":["Applied reactor physics","Transport theory","Diffusion theory","CANDU","Nuclear reactor"],"dc:title":["Transport-theory-equivalent diffusion coefficients for node-homogenized neutron diffusion problems in CANDU lattices"],"dc:type":["Thesis"],"thesis:degree_discipline":["Modelling and Computational Science"],"thesis:degree_name":["Master of Science (MSc)"],"thesis:institution_name":["University of Ontario Institute of Technology"]},"updated_at":"2026-07-24T05:35:30Z"}