S. van Til , P.R. Hania , A.V. Fedorov , E. D'Agata , D. Freis , S. Bejaoui , F. Delage , A. Gallais-During
{"title":"海洋辐照试验中含镅包层燃料的辐照性能及初步检验","authors":"S. van Til , P.R. Hania , A.V. Fedorov , E. D'Agata , D. Freis , S. Bejaoui , F. Delage , A. Gallais-During","doi":"10.1016/j.jnucmat.2023.154699","DOIUrl":null,"url":null,"abstract":"<div><p><span>Americium<span> (Am) is a strong contributor to the long-term radiotoxicity of high-level waste from nuclear fuels. Transmutation of long-lived nuclides like </span></span><sup>241</sup>Am by irradiation in nuclear reactors is therefore an option for the reduction of radiotoxicity and heat production of waste volumes to be stored in a repository. The MARINE irradiation experiment is the latest in a series of European experiments on americium transmutation (e.g. EFTTRA-T4, EFTTRA-T4bis, HELIOS, MARIOS, SPHERE) performed in the High Flux Reactor (HFR) in Petten (The Netherlands). The development and irradiation of MARINE was carried out in the framework of the collaborative research project PELGRIMM of the EURATOM 7th Framework Programme (FP7). Dismantling was completed and post-irradiation examinations (PIE) were started within the Dutch national research programme PIONEER. Destructive PIE is foreseen within the Euratom H2020 funded project PATRICIA.</p><p><span>The main objective of the MARINE experiment is to study the in-pile behaviour of uranium oxide fuel containing 13% of americium and to compare the behaviour of sphere-pac versus </span>pellet fuel<span>, in particular the role of microstructure and temperature on fission gas and helium release dynamics on fuel swelling.</span></p><p>The MARINE experiment was irradiated for 359 Full Power Days in the HFR in 2016 and 2017. This paper discusses results from irradiation, i.e. power and temperature history and transmutation rates as well as preliminary results from post irradiation examinations, assessing a.o. clad strains and helium and fission gas release and first ceramographic observations, putting a preliminary upper bound on fuel swelling.</p></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"587 ","pages":"Article 154699"},"PeriodicalIF":2.8000,"publicationDate":"2023-08-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":"0","resultStr":"{\"title\":\"Irradiation performance and first examinations of Americium bearing blanket fuel from the MARINE irradiation experiment\",\"authors\":\"S. van Til , P.R. Hania , A.V. Fedorov , E. D'Agata , D. Freis , S. Bejaoui , F. Delage , A. Gallais-During\",\"doi\":\"10.1016/j.jnucmat.2023.154699\",\"DOIUrl\":null,\"url\":null,\"abstract\":\"<div><p><span>Americium<span> (Am) is a strong contributor to the long-term radiotoxicity of high-level waste from nuclear fuels. Transmutation of long-lived nuclides like </span></span><sup>241</sup>Am by irradiation in nuclear reactors is therefore an option for the reduction of radiotoxicity and heat production of waste volumes to be stored in a repository. The MARINE irradiation experiment is the latest in a series of European experiments on americium transmutation (e.g. EFTTRA-T4, EFTTRA-T4bis, HELIOS, MARIOS, SPHERE) performed in the High Flux Reactor (HFR) in Petten (The Netherlands). The development and irradiation of MARINE was carried out in the framework of the collaborative research project PELGRIMM of the EURATOM 7th Framework Programme (FP7). Dismantling was completed and post-irradiation examinations (PIE) were started within the Dutch national research programme PIONEER. Destructive PIE is foreseen within the Euratom H2020 funded project PATRICIA.</p><p><span>The main objective of the MARINE experiment is to study the in-pile behaviour of uranium oxide fuel containing 13% of americium and to compare the behaviour of sphere-pac versus </span>pellet fuel<span>, in particular the role of microstructure and temperature on fission gas and helium release dynamics on fuel swelling.</span></p><p>The MARINE experiment was irradiated for 359 Full Power Days in the HFR in 2016 and 2017. This paper discusses results from irradiation, i.e. power and temperature history and transmutation rates as well as preliminary results from post irradiation examinations, assessing a.o. clad strains and helium and fission gas release and first ceramographic observations, putting a preliminary upper bound on fuel swelling.</p></div>\",\"PeriodicalId\":373,\"journal\":{\"name\":\"Journal of Nuclear Materials\",\"volume\":\"587 \",\"pages\":\"Article 154699\"},\"PeriodicalIF\":2.8000,\"publicationDate\":\"2023-08-25\",\"publicationTypes\":\"Journal Article\",\"fieldsOfStudy\":null,\"isOpenAccess\":false,\"openAccessPdf\":\"\",\"citationCount\":\"0\",\"resultStr\":null,\"platform\":\"Semanticscholar\",\"paperid\":null,\"PeriodicalName\":\"Journal of Nuclear Materials\",\"FirstCategoryId\":\"5\",\"ListUrlMain\":\"https://www.sciencedirect.com/science/article/pii/S0022311523004671\",\"RegionNum\":2,\"RegionCategory\":\"工程技术\",\"ArticlePicture\":[],\"TitleCN\":null,\"AbstractTextCN\":null,\"PMCID\":null,\"EPubDate\":\"\",\"PubModel\":\"\",\"JCR\":\"Q3\",\"JCRName\":\"MATERIALS SCIENCE, MULTIDISCIPLINARY\",\"Score\":null,\"Total\":0}","platform":"Semanticscholar","paperid":null,"PeriodicalName":"Journal of Nuclear Materials","FirstCategoryId":"5","ListUrlMain":"https://www.sciencedirect.com/science/article/pii/S0022311523004671","RegionNum":2,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":null,"EPubDate":"","PubModel":"","JCR":"Q3","JCRName":"MATERIALS SCIENCE, MULTIDISCIPLINARY","Score":null,"Total":0}
Irradiation performance and first examinations of Americium bearing blanket fuel from the MARINE irradiation experiment
Americium (Am) is a strong contributor to the long-term radiotoxicity of high-level waste from nuclear fuels. Transmutation of long-lived nuclides like 241Am by irradiation in nuclear reactors is therefore an option for the reduction of radiotoxicity and heat production of waste volumes to be stored in a repository. The MARINE irradiation experiment is the latest in a series of European experiments on americium transmutation (e.g. EFTTRA-T4, EFTTRA-T4bis, HELIOS, MARIOS, SPHERE) performed in the High Flux Reactor (HFR) in Petten (The Netherlands). The development and irradiation of MARINE was carried out in the framework of the collaborative research project PELGRIMM of the EURATOM 7th Framework Programme (FP7). Dismantling was completed and post-irradiation examinations (PIE) were started within the Dutch national research programme PIONEER. Destructive PIE is foreseen within the Euratom H2020 funded project PATRICIA.
The main objective of the MARINE experiment is to study the in-pile behaviour of uranium oxide fuel containing 13% of americium and to compare the behaviour of sphere-pac versus pellet fuel, in particular the role of microstructure and temperature on fission gas and helium release dynamics on fuel swelling.
The MARINE experiment was irradiated for 359 Full Power Days in the HFR in 2016 and 2017. This paper discusses results from irradiation, i.e. power and temperature history and transmutation rates as well as preliminary results from post irradiation examinations, assessing a.o. clad strains and helium and fission gas release and first ceramographic observations, putting a preliminary upper bound on fuel swelling.
期刊介绍：
The Journal of Nuclear Materials publishes high quality papers in materials research for nuclear applications, primarily fission reactors, fusion reactors, and similar environments including radiation areas of charged particle accelerators. Both original research and critical review papers covering experimental, theoretical, and computational aspects of either fundamental or applied nature are welcome.
The breadth of the field is such that a wide range of processes and properties in the field of materials science and engineering is of interest to the readership, spanning atom-scale processes, microstructures, thermodynamics, mechanical properties, physical properties, and corrosion, for example.
Topics covered by JNM
Fission reactor materials, including fuels, cladding, core structures, pressure vessels, coolant interactions with materials, moderator and control components, fission product behavior.
Materials aspects of the entire fuel cycle.
Materials aspects of the actinides and their compounds.
Performance of nuclear waste materials; materials aspects of the immobilization of wastes.
Fusion reactor materials, including first walls, blankets, insulators and magnets.
Neutron and charged particle radiation effects in materials, including defects, transmutations, microstructures, phase changes and macroscopic properties.
Interaction of plasmas, ion beams, electron beams and electromagnetic radiation with materials relevant to nuclear systems.
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