Karytha M.S. Corrêa , Natália G.P.L. Oliveira , Claubia Pereira , Carlos E. Velasquez
{"title":"基于ARC反应堆的混合聚变裂变系统中插入的后处理燃料分析","authors":"Karytha M.S. Corrêa , Natália G.P.L. Oliveira , Claubia Pereira , Carlos E. Velasquez","doi":"10.1016/j.anucene.2025.111291","DOIUrl":null,"url":null,"abstract":"<div><div>This study proposes evaluating hybrid fusion-fission systems based on the ARC reactor, which uses reprocessed fuels in the fission transmutation layer. The analysis compared two reprocessed fuels: one based on dioxide ((TRU,Th)O<sub>2</sub>) and the other based on nitride ((TRU,Th)N). The analysis focused on the neutronic parameters and the burnup of these fuels. The MCNP particle transport code was used for the simulations, while the burnup process of the fuels in the (TRU,Th)O<sub>2</sub> and (TRU,Th)N systems was conducted using the MONTEBURNS code, which links MCNP and the ORIGEN2.1 depletion code. The results indicate that the (TRU,Th)N system may exhibit superior efficiency in the transmutation of minor actinides during fuel burnup compared to the dioxide-based system.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"216 ","pages":"Article 111291"},"PeriodicalIF":2.3000,"publicationDate":"2025-06-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":"0","resultStr":"{\"title\":\"Analysis of reprocessed fuels inserted into a hybrid fusion-fission system based on the ARC reactor\",\"authors\":\"Karytha M.S. Corrêa , Natália G.P.L. Oliveira , Claubia Pereira , Carlos E. Velasquez\",\"doi\":\"10.1016/j.anucene.2025.111291\",\"DOIUrl\":null,\"url\":null,\"abstract\":\"<div><div>This study proposes evaluating hybrid fusion-fission systems based on the ARC reactor, which uses reprocessed fuels in the fission transmutation layer. The analysis compared two reprocessed fuels: one based on dioxide ((TRU,Th)O<sub>2</sub>) and the other based on nitride ((TRU,Th)N). The analysis focused on the neutronic parameters and the burnup of these fuels. The MCNP particle transport code was used for the simulations, while the burnup process of the fuels in the (TRU,Th)O<sub>2</sub> and (TRU,Th)N systems was conducted using the MONTEBURNS code, which links MCNP and the ORIGEN2.1 depletion code. The results indicate that the (TRU,Th)N system may exhibit superior efficiency in the transmutation of minor actinides during fuel burnup compared to the dioxide-based system.</div></div>\",\"PeriodicalId\":8006,\"journal\":{\"name\":\"Annals of Nuclear Energy\",\"volume\":\"216 \",\"pages\":\"Article 111291\"},\"PeriodicalIF\":2.3000,\"publicationDate\":\"2025-06-15\",\"publicationTypes\":\"Journal Article\",\"fieldsOfStudy\":null,\"isOpenAccess\":false,\"openAccessPdf\":\"\",\"citationCount\":\"0\",\"resultStr\":null,\"platform\":\"Semanticscholar\",\"paperid\":null,\"PeriodicalName\":\"Annals of Nuclear Energy\",\"FirstCategoryId\":\"5\",\"ListUrlMain\":\"https://www.sciencedirect.com/science/article/pii/S0306454925001082\",\"RegionNum\":3,\"RegionCategory\":\"工程技术\",\"ArticlePicture\":[],\"TitleCN\":null,\"AbstractTextCN\":null,\"PMCID\":null,\"EPubDate\":\"2025/2/26 0:00:00\",\"PubModel\":\"Epub\",\"JCR\":\"Q1\",\"JCRName\":\"NUCLEAR SCIENCE & TECHNOLOGY\",\"Score\":null,\"Total\":0}","platform":"Semanticscholar","paperid":null,"PeriodicalName":"Annals of Nuclear Energy","FirstCategoryId":"5","ListUrlMain":"https://www.sciencedirect.com/science/article/pii/S0306454925001082","RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":null,"EPubDate":"2025/2/26 0:00:00","PubModel":"Epub","JCR":"Q1","JCRName":"NUCLEAR SCIENCE & TECHNOLOGY","Score":null,"Total":0}
Analysis of reprocessed fuels inserted into a hybrid fusion-fission system based on the ARC reactor
This study proposes evaluating hybrid fusion-fission systems based on the ARC reactor, which uses reprocessed fuels in the fission transmutation layer. The analysis compared two reprocessed fuels: one based on dioxide ((TRU,Th)O2) and the other based on nitride ((TRU,Th)N). The analysis focused on the neutronic parameters and the burnup of these fuels. The MCNP particle transport code was used for the simulations, while the burnup process of the fuels in the (TRU,Th)O2 and (TRU,Th)N systems was conducted using the MONTEBURNS code, which links MCNP and the ORIGEN2.1 depletion code. The results indicate that the (TRU,Th)N system may exhibit superior efficiency in the transmutation of minor actinides during fuel burnup compared to the dioxide-based system.
期刊介绍：
Annals of Nuclear Energy provides an international medium for the communication of original research, ideas and developments in all areas of the field of nuclear energy science and technology. Its scope embraces nuclear fuel reserves, fuel cycles and cost, materials, processing, system and component technology (fission only), design and optimization, direct conversion of nuclear energy sources, environmental control, reactor physics, heat transfer and fluid dynamics, structural analysis, fuel management, future developments, nuclear fuel and safety, nuclear aerosol, neutron physics, computer technology (both software and hardware), risk assessment, radioactive waste disposal and reactor thermal hydraulics. Papers submitted to Annals need to demonstrate a clear link to nuclear power generation/nuclear engineering. Papers which deal with pure nuclear physics, pure health physics, imaging, or attenuation and shielding properties of concretes and various geological materials are not within the scope of the journal. Also, papers that deal with policy or economics are not within the scope of the journal.
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