{"id":{"repo_id":"tamu","oai_identifier":"oai:oaktrust.library.tamu.edu:1969.1/1595129"},"canonical_url":"https://search.dev.ndltd.org/etd/tamu/oai:oaktrust.library.tamu.edu:1969.1/1595129","repository":{"repo_id":"tamu","name":"Texas A&M University","base_url":"https://oaktrust.library.tamu.edu/server/oai/request"},"display":{"title":"Modeling and Simulations of Natural Convection Technologies for Advanced Reactor Applications","abstract":"As advanced reactors are looking to rely on natural convection technologies, modeling and simulations of these systems is crucial to design, analyses, and optimization. System codes have been the primary tool for design and safety analysis of reactor technologies. While legacy system codes such RELAP and TRACE have been well validated for existing reactor technologies, more work is needed to assess code capabilities in modeling advanced reactor systems which have significant phenomenological differences compared to traditional light-water reactors. Computational fluid dynamics (CFD) and system code models of the experimental Reactor Cavity Cooling System (RCCS) facility at Texas A&M were developed and validated against two sets of experimental data to assess code performance and modelling of a typical passive heat removal system proposed for advanced reactor designs. Results showed that the CFD and system code models were able to predict both the global and local characteristics of the RCCS facility for Low and High Reynolds experimental data sets. The Low Reynolds case showed significant stratification and recirculation in the riser panel which lead to additional thermal stresses in the riser panel. System code models of a NuScale-like integral-pressurized water reactor (iPWR) were also developed to perform a code-to-code benchmark study against the NuScale simulator at Texas A&M University. Two models were developed using the TRACE code which included a traditional 1D and 3D modelling approach of the primary vessel where steady-state and a postulated station-blackout (SBO) scenario were simulated. Steady-state results of the TRACE models showed good agreement to data from the NuScale simulator, and overall trends of the primary and secondary side characteristics over the SBO transient were captured.","abstract_html":"As advanced reactors are looking to rely on natural convection technologies, modeling and simulations of these systems is crucial to design, analyses, and optimization. System codes have been the primary tool for design and safety analysis of reactor technologies. While legacy system codes such RELAP and TRACE have been well validated for existing reactor technologies, more work is needed to assess code capabilities in modeling advanced reactor systems which have significant phenomenological differences compared to traditional light-water reactors. Computational fluid dynamics (CFD) and system code models of the experimental Reactor Cavity Cooling System (RCCS) facility at Texas A&M were developed and validated against two sets of experimental data to assess code performance and modelling of a typical passive heat removal system proposed for advanced reactor designs. Results showed that the CFD and system code models were able to predict both the global and local characteristics of the RCCS facility for Low and High Reynolds experimental data sets. The Low Reynolds case showed significant stratification and recirculation in the riser panel which lead to additional thermal stresses in the riser panel. System code models of a NuScale-like integral-pressurized water reactor (iPWR) were also developed to perform a code-to-code benchmark study against the NuScale simulator at Texas A&M University. Two models were developed using the TRACE code which included a traditional 1D and 3D modelling approach of the primary vessel where steady-state and a postulated station-blackout (SBO) scenario were simulated. Steady-state results of the TRACE models showed good agreement to data from the NuScale simulator, and overall trends of the primary and secondary side characteristics over the SBO transient were captured.","abstract_has_math":false,"creators":["Gorman, Michael Spencer 1995-"],"institution":"Texas A&M University","degree_name":"Master of Science","degree_level":"Masters","degree_discipline":"Nuclear Engineering","degree_department":null,"school":null,"contributors":[],"advisors":["Hassan, Yassin"],"committee_chairs":[],"committee_members":["Peddicord, Kenneth","Ugaz, Victor"],"year":2025,"date_issued":"2025-05","date_published":"2025-05","updated_at":"2026-08-21T16:48:40Z","subjects":["Engineering, Nuclear"],"languages":["English"],"rights":[],"rights_urls":[],"identifier_entries":[]},"links":{"outbound_url":"https://hdl.handle.net/1969.1/1595129","outbound_label":"Handle","outbound_source":"dc:identifier.uri"},"source_record":{"url":"https://oaktrust.library.tamu.edu/server/oai/request?verb=GetRecord&metadataPrefix=dim&identifier=oai%3Aoaktrust.library.tamu.edu%3A1969.1%2F1595129","prefix":"dim"},"metadata_groups":[{"id":"people","label":"People","entries":[{"key":"dc:contributor.advisor","label":"Advisor","values":["Hassan, Yassin"]},{"key":"dc:contributor.committeemember","label":"Committee Member","values":["Peddicord, Kenneth","Ugaz, Victor"]},{"key":"dc:creator","label":"Author","values":["Gorman, Michael Spencer 1995-"]}]},{"id":"academic_context","label":"Academic Context","entries":[{"key":"dc:date.accessioned","label":"Dc Date Accessioned","values":["2025-09-12T21:30:47Z"]},{"key":"dc:date.issued","label":"Date","values":["2025-05"]},{"key":"dc:type","label":"Dc Type","values":["Thesis"]},{"key":"thesis:degree_discipline","label":"Discipline","values":["Nuclear Engineering"]},{"key":"thesis:degree_level","label":"Degree Level","values":["Masters"]},{"key":"thesis:degree_name","label":"Degree Name","values":["Master of Science"]},{"key":"thesis:institution_name","label":"Thesis Institution Name","values":["Texas A&M University"]}]},{"id":"subjects_keywords","label":"Subjects and Keywords","entries":[{"key":"dc:subject","label":"Dc Subject","values":["Engineering, Nuclear"]}]},{"id":"language_rights","label":"Language and Rights","entries":[{"key":"dc:language.iso","label":"Language (ISO)","values":["English"]}]},{"id":"identifiers","label":"Identifiers","entries":[{"key":"dc:identifier.uri","label":"Identifier URI","values":["https://hdl.handle.net/1969.1/1595129"]}]},{"id":"additional","label":"Additional Metadata","entries":[{"key":"dc:description.abstract","label":"Abstract","values":["As advanced reactors are looking to rely on natural convection technologies, modeling and simulations of these systems is crucial to design, analyses, and optimization. System codes have been the primary tool for design and safety analysis of reactor technologies. While legacy system codes such RELAP and TRACE have been well validated for existing reactor technologies, more work is needed to assess code capabilities in modeling advanced reactor systems which have significant phenomenological differences compared to traditional light-water reactors. Computational fluid dynamics (CFD) and system code models of the experimental Reactor Cavity Cooling System (RCCS) facility at Texas A&M were developed and validated against two sets of experimental data to assess code performance and modelling of a typical passive heat removal system proposed for advanced reactor designs. Results showed that the CFD and system code models were able to predict both the global and local characteristics of the RCCS facility for Low and High Reynolds experimental data sets. The Low Reynolds case showed significant stratification and recirculation in the riser panel which lead to additional thermal stresses in the riser panel. System code models of a NuScale-like integral-pressurized water reactor (iPWR) were also developed to perform a code-to-code benchmark study against the NuScale simulator at Texas A&M University. Two models were developed using the TRACE code which included a traditional 1D and 3D modelling approach of the primary vessel where steady-state and a postulated station-blackout (SBO) scenario were simulated. Steady-state results of the TRACE models showed good agreement to data from the NuScale simulator, and overall trends of the primary and secondary side characteristics over the SBO transient were captured."]},{"key":"dc:format.mimetype","label":"Dc Format Mimetype","values":["application/pdf"]},{"key":"dc:title","label":"Title","values":["Modeling and Simulations of Natural Convection Technologies for Advanced Reactor Applications"]}]}],"canonical_facts":{"dc:contributor.advisor":["Hassan, Yassin"],"dc:contributor.committeemember":["Peddicord, Kenneth","Ugaz, Victor"],"dc:creator":["Gorman, Michael Spencer 1995-"],"dc:date.accessioned":["2025-09-12T21:30:47Z"],"dc:date.issued":["2025-05"],"dc:description.abstract":["As advanced reactors are looking to rely on natural convection technologies, modeling and simulations of these systems is crucial to design, analyses, and optimization. System codes have been the primary tool for design and safety analysis of reactor technologies. While legacy system codes such RELAP and TRACE have been well validated for existing reactor technologies, more work is needed to assess code capabilities in modeling advanced reactor systems which have significant phenomenological differences compared to traditional light-water reactors. Computational fluid dynamics (CFD) and system code models of the experimental Reactor Cavity Cooling System (RCCS) facility at Texas A&M were developed and validated against two sets of experimental data to assess code performance and modelling of a typical passive heat removal system proposed for advanced reactor designs. Results showed that the CFD and system code models were able to predict both the global and local characteristics of the RCCS facility for Low and High Reynolds experimental data sets. The Low Reynolds case showed significant stratification and recirculation in the riser panel which lead to additional thermal stresses in the riser panel. System code models of a NuScale-like integral-pressurized water reactor (iPWR) were also developed to perform a code-to-code benchmark study against the NuScale simulator at Texas A&M University. Two models were developed using the TRACE code which included a traditional 1D and 3D modelling approach of the primary vessel where steady-state and a postulated station-blackout (SBO) scenario were simulated. Steady-state results of the TRACE models showed good agreement to data from the NuScale simulator, and overall trends of the primary and secondary side characteristics over the SBO transient were captured."],"dc:format.mimetype":["application/pdf"],"dc:identifier.uri":["https://hdl.handle.net/1969.1/1595129"],"dc:language.iso":["English"],"dc:subject":["Engineering, Nuclear"],"dc:title":["Modeling and Simulations of Natural Convection Technologies for Advanced Reactor Applications"],"dc:type":["Thesis"],"thesis:degree_discipline":["Nuclear Engineering"],"thesis:degree_level":["Masters"],"thesis:degree_name":["Master of Science"],"thesis:institution_name":["Texas A&M University"]},"updated_at":"2026-08-21T16:48:40Z"}