Development of the Technological Process for the IGR Reactor's Highly-Enriched Irradiated Uranium-Graphite Fuel Immobilization

IF 2.8 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY

Journal of Nuclear Materials Pub Date : 2025-04-01 DOI:10.1016/j.jnucmat.2025.155801

Kuanysh K. Samarkhanov , Yuliya Yu. Baklanova , Olga S. Bukina , Viktor V. Baklanov , Yerbolat T. Koyanbayev , Ivan M. Kukushkin , Igor M. Bolshinsky , Kenneth J. Bateman

{"title":"Development of the Technological Process for the IGR Reactor's Highly-Enriched Irradiated Uranium-Graphite Fuel Immobilization","authors":"Kuanysh K. Samarkhanov , Yuliya Yu. Baklanova , Olga S. Bukina , Viktor V. Baklanov , Yerbolat T. Koyanbayev , Ivan M. Kukushkin , Igor M. Bolshinsky , Kenneth J. Bateman","doi":"10.1016/j.jnucmat.2025.155801","DOIUrl":null,"url":null,"abstract":"<div><div>The immobilization of irradiated highly enriched uranium (HEU) fuel is a critical component of nuclear waste management and non-proliferation efforts. In Kazakhstan, at National Nuclear Center of the Republic of Kazakhstan special attention is given to managing legacy HEU fuel from research reactors. One such case involves the IGR research reactor, whose first core containing irradiated HEU uranium-graphite fuel was operated from 1961 to 1966 and removed following reactor modernization. This fuel now requires a reliable and secure immobilization strategy.</div><div>This paper presents the development of a technological process for immobilizing this fuel to reduce its enrichment to below 20% in terms of <sup>235</sup>U content. The proposed method involves down-blending irradiated HEU fuel with depleted uranium, followed by encapsulation in a Portland cement matrix. Full-scale experiments were conducted to assess the uniformity of uranium distribution within the matrix.</div><div>The results confirm the effectiveness of this approach, ensuring reliable immobilization of fuel in accordance with international requirements, including IAEA standards and Kazakhstan's regulatory framework. These findings contribute to the broader effort of adapting immobilization strategies for the safe management of spent fuel from research reactors.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"610 ","pages":"Article 155801"},"PeriodicalIF":2.8000,"publicationDate":"2025-04-01","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/S0022311525001965","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}

引用次数: 0

Abstract

The immobilization of irradiated highly enriched uranium (HEU) fuel is a critical component of nuclear waste management and non-proliferation efforts. In Kazakhstan, at National Nuclear Center of the Republic of Kazakhstan special attention is given to managing legacy HEU fuel from research reactors. One such case involves the IGR research reactor, whose first core containing irradiated HEU uranium-graphite fuel was operated from 1961 to 1966 and removed following reactor modernization. This fuel now requires a reliable and secure immobilization strategy.

This paper presents the development of a technological process for immobilizing this fuel to reduce its enrichment to below 20% in terms of ²³⁵U content. The proposed method involves down-blending irradiated HEU fuel with depleted uranium, followed by encapsulation in a Portland cement matrix. Full-scale experiments were conducted to assess the uniformity of uranium distribution within the matrix.

The results confirm the effectiveness of this approach, ensuring reliable immobilization of fuel in accordance with international requirements, including IAEA standards and Kazakhstan's regulatory framework. These findings contribute to the broader effort of adapting immobilization strategies for the safe management of spent fuel from research reactors.

查看原文本刊更多论文

开发 IGR 反应堆高浓缩辐照铀-石墨燃料固定化技术工艺

本文章由计算机程序翻译，如有差异，请以英文原文为准。

求助全文

约1分钟内获得全文求助全文

来源期刊

Journal of Nuclear Materials 工程技术-材料科学：综合

CiteScore

5.70

自引率

25.80%

发文量

601

审稿时长

63 days

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.