Analysis of a pressurized power reactor using thorium mixed fuel under regular operation

Detalhes bibliográficos
Autor(a) principal: GOMES, DANIEL de S.
Data de Publicação: 2020
Outros Autores: STEFANI, GIOVANNI L. de, OLIVEIRA, FABIO B.V. de, INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE
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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spelling 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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