Analysis of a pressurized power reactor using thorium mixed fuel under regular operation
Autor(a) principal: | |
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Data de Publicação: | 2020 |
Outros Autores: | , , |
Tipo de documento: | Artigo de conferência |
Título da fonte: | Repositório Institucional do IPEN |
Texto Completo: | http://repositorio.ipen.br/handle/123456789/30708 |
Resumo: | This work discusses a parametric study applied to nuclear power generation based on a mixed fuel formed by the composition of thorium-uranium oxide (Th-U)O2. Also, approached in this study the physical neutrons models of a fuel system composed of ThO2 75 wt% and UO2 25 wt%, with 19.5% enrichment of U-235. The thermodynamic features of the thorium-uranium fuel system compared with the properties of uranium dioxide. Thorium-based fuel operating extended fuel cycles reach of over 80 GWd/MTU in a pressurized water reactor (PWR). Homogenous distribution of thorium-based fuel, used on the reactor core, could reduce Pu-239, once U-233 production capacity dependent on Th-232 replacing U-238 in the fuel matrix. The mixed oxide fuel has a lower buildup of Pu-239, causing the linear heat rate distribution slope to flatten and lowering fuel porosity. The release of gaseous fission products models for (Th-U)O2 could have different diffusion coefficients when compared to uranium oxide models. Besides, resulting in lower thermal gradients than UO2 and a reduction in fuel swelling. This parametric study reviews the aspects of radioactive decay chains of uranium and thorium. It founded the simulation using approved nuclear codes, such as SERPENT for neutron physics calculations and the FRAPCON code, which defines the licensing process. The results show that thoria based fuel has a higher performance than UO2 fuel in regular operation and can improve safety margins. |
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2020-01-15T16:55:54Z2020-01-15T16:55:54ZOctober 21-25, 2019http://repositorio.ipen.br/handle/123456789/30708This work discusses a parametric study applied to nuclear power generation based on a mixed fuel formed by the composition of thorium-uranium oxide (Th-U)O2. Also, approached in this study the physical neutrons models of a fuel system composed of ThO2 75 wt% and UO2 25 wt%, with 19.5% enrichment of U-235. The thermodynamic features of the thorium-uranium fuel system compared with the properties of uranium dioxide. Thorium-based fuel operating extended fuel cycles reach of over 80 GWd/MTU in a pressurized water reactor (PWR). Homogenous distribution of thorium-based fuel, used on the reactor core, could reduce Pu-239, once U-233 production capacity dependent on Th-232 replacing U-238 in the fuel matrix. The mixed oxide fuel has a lower buildup of Pu-239, causing the linear heat rate distribution slope to flatten and lowering fuel porosity. The release of gaseous fission products models for (Th-U)O2 could have different diffusion coefficients when compared to uranium oxide models. Besides, resulting in lower thermal gradients than UO2 and a reduction in fuel swelling. This parametric study reviews the aspects of radioactive decay chains of uranium and thorium. It founded the simulation using approved nuclear codes, such as SERPENT for neutron physics calculations and the FRAPCON code, which defines the licensing process. The results show that thoria based fuel has a higher performance than UO2 fuel in regular operation and can improve safety margins.Submitted by Celia Satomi Uehara (celia.u-topservice@ipen.br) on 2020-01-15T16:55:54Z No. of bitstreams: 1 26360.pdf: 801818 bytes, checksum: dfbca60104ad3052f658cfe46d77d98b (MD5)Made available in DSpace on 2020-01-15T16:55:54Z (GMT). No. of bitstreams: 1 26360.pdf: 801818 bytes, checksum: dfbca60104ad3052f658cfe46d77d98b (MD5)4996-5009Associa????o Brasileira de Energia Nuclearcomparative evaluationsenthalpyf codesmixed oxide fuelsperformancepwr type reactorss codesthermal conductivitythoriumuranium oxidesAnalysis of a pressurized power reactor using thorium mixed fuel under regular operationinfo:eu-repo/semantics/publishedVersioninfo:eu-repo/semantics/conferenceObjectINACIRio de JaneiroSantos, SP76707606931600600600GOMES, DANIEL de S.STEFANI, GIOVANNI L. deOLIVEIRA, FABIO B.V. deINTERNATIONAL NUCLEAR ATLANTIC CONFERENCEinfo:eu-repo/semantics/openAccessreponame:Repositório Institucional do IPENinstname:Instituto de Pesquisas Energéticas e Nucleares (IPEN)instacron:IPEN263602019OLIVEIRA, FABIO B.V. deSTEFANI, GIOVANNI L. deGOMES, DANIEL de S.20-01Proceedings93176067670OLIVEIRA, FABIO B.V. de:931:420:NSTEFANI, GIOVANNI L. de:7606:310:NGOMES, DANIEL de S.:7670:420:SORIGINAL26360.pdf26360.pdfapplication/pdf801818http://repositorio.ipen.br/bitstream/123456789/30708/1/26360.pdfdfbca60104ad3052f658cfe46d77d98bMD51LICENSElicense.txtlicense.txttext/plain; charset=utf-81748http://repositorio.ipen.br/bitstream/123456789/30708/2/license.txt8a4605be74aa9ea9d79846c1fba20a33MD52123456789/307082020-04-12 20:38:39.468oai:repositorio.ipen.br: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Repositório InstitucionalPUBhttp://repositorio.ipen.br/oai/requestbibl@ipen.bropendoar:45102020-04-12T20:38:39Repositório Institucional do IPEN - Instituto de Pesquisas Energéticas e Nucleares (IPEN)false |
dc.title.pt_BR.fl_str_mv |
Analysis of a pressurized power reactor using thorium mixed fuel under regular operation |
title |
Analysis of a pressurized power reactor using thorium mixed fuel under regular operation |
spellingShingle |
Analysis of a pressurized power reactor using thorium mixed fuel under regular operation GOMES, DANIEL de S. comparative evaluations enthalpy f codes mixed oxide fuels performance pwr type reactors s codes thermal conductivity thorium uranium oxides |
title_short |
Analysis of a pressurized power reactor using thorium mixed fuel under regular operation |
title_full |
Analysis of a pressurized power reactor using thorium mixed fuel under regular operation |
title_fullStr |
Analysis of a pressurized power reactor using thorium mixed fuel under regular operation |
title_full_unstemmed |
Analysis of a pressurized power reactor using thorium mixed fuel under regular operation |
title_sort |
Analysis of a pressurized power reactor using thorium mixed fuel under regular operation |
author |
GOMES, DANIEL de S. |
author_facet |
GOMES, DANIEL de S. STEFANI, GIOVANNI L. de OLIVEIRA, FABIO B.V. de INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE |
author_role |
author |
author2 |
STEFANI, GIOVANNI L. de OLIVEIRA, FABIO B.V. de INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE |
author2_role |
author author author |
dc.contributor.author.fl_str_mv |
GOMES, DANIEL de S. STEFANI, GIOVANNI L. de OLIVEIRA, FABIO B.V. de INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE |
dc.subject.por.fl_str_mv |
comparative evaluations enthalpy f codes mixed oxide fuels performance pwr type reactors s codes thermal conductivity thorium uranium oxides |
topic |
comparative evaluations enthalpy f codes mixed oxide fuels performance pwr type reactors s codes thermal conductivity thorium uranium oxides |
description |
This work discusses a parametric study applied to nuclear power generation based on a mixed fuel formed by the composition of thorium-uranium oxide (Th-U)O2. Also, approached in this study the physical neutrons models of a fuel system composed of ThO2 75 wt% and UO2 25 wt%, with 19.5% enrichment of U-235. The thermodynamic features of the thorium-uranium fuel system compared with the properties of uranium dioxide. Thorium-based fuel operating extended fuel cycles reach of over 80 GWd/MTU in a pressurized water reactor (PWR). Homogenous distribution of thorium-based fuel, used on the reactor core, could reduce Pu-239, once U-233 production capacity dependent on Th-232 replacing U-238 in the fuel matrix. The mixed oxide fuel has a lower buildup of Pu-239, causing the linear heat rate distribution slope to flatten and lowering fuel porosity. The release of gaseous fission products models for (Th-U)O2 could have different diffusion coefficients when compared to uranium oxide models. Besides, resulting in lower thermal gradients than UO2 and a reduction in fuel swelling. This parametric study reviews the aspects of radioactive decay chains of uranium and thorium. It founded the simulation using approved nuclear codes, such as SERPENT for neutron physics calculations and the FRAPCON code, which defines the licensing process. The results show that thoria based fuel has a higher performance than UO2 fuel in regular operation and can improve safety margins. |
publishDate |
2020 |
dc.date.evento.pt_BR.fl_str_mv |
October 21-25, 2019 |
dc.date.accessioned.fl_str_mv |
2020-01-15T16:55:54Z |
dc.date.available.fl_str_mv |
2020-01-15T16:55:54Z |
dc.type.status.fl_str_mv |
info:eu-repo/semantics/publishedVersion |
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info:eu-repo/semantics/conferenceObject |
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conferenceObject |
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publishedVersion |
dc.identifier.uri.fl_str_mv |
http://repositorio.ipen.br/handle/123456789/30708 |
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http://repositorio.ipen.br/handle/123456789/30708 |
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7670 7606 931 |
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600 600 600 |
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info:eu-repo/semantics/openAccess |
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openAccess |
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4996-5009 |
dc.coverage.pt_BR.fl_str_mv |
I |
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Associa????o Brasileira de Energia Nuclear |
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Associa????o Brasileira de Energia Nuclear |
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