Fuel performance code BERKUT-U to simulate the in-pile behavior of a single oxide or nitride fuel rod for fast reactors

IF 2.8 2区工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY Journal of Nuclear Materials Pub Date : 2024-09-20 DOI:10.1016/j.jnucmat.2024.155417

A.V. Boldyrev, A.P. Dolgodvorov, I.O. Dolinskiy, V.D. Ozrin, P.V. Polovnikov, V.E. Shestak, V.I. Tarasov, A.V. Zadorozhnyi

{"title":"Fuel performance code BERKUT-U to simulate the in-pile behavior of a single oxide or nitride fuel rod for fast reactors","authors":"A.V. Boldyrev, A.P. Dolgodvorov, I.O. Dolinskiy, V.D. Ozrin, P.V. Polovnikov, V.E. Shestak, V.I. Tarasov, A.V. Zadorozhnyi","doi":"10.1016/j.jnucmat.2024.155417","DOIUrl":null,"url":null,"abstract":"<div><div>This paper describes the fuel performance code BERKUT-U, which the Nuclear Safety Institute of the Russian Academy of Sciences (IBRAE RAN) designed as part of the \"Codes of The New Generation\" subproject of the Proryv Project. The code aims to model oxide or nitride single fuel rod behavior with a gas or liquid metal sublayer under normal and accident conditions of a liquid metal-cooled fast reactor operation. The BERKUT-U code's models, incorporating the MFPR/R code, are grounded in the contemporary understanding of mechanisms governing the fundamental processes in fuel rods under irradiation, which substantially enhances the code's predictive ability in comparison with the engineering analogs. Simulations of the nitride and oxide fuel rod behavior in BN-600 and BOR-60 fast reactors have demonstrated good agreement with the post-irradiation examination data. Further validation is foreseen as the corresponding data are available.</div></div>","PeriodicalId":373,"journal":{"name":"Journal of Nuclear Materials","volume":"603 ","pages":"Article 155417"},"PeriodicalIF":2.8000,"publicationDate":"2024-09-20","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/S002231152400518X","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

This paper describes the fuel performance code BERKUT-U, which the Nuclear Safety Institute of the Russian Academy of Sciences (IBRAE RAN) designed as part of the "Codes of The New Generation" subproject of the Proryv Project. The code aims to model oxide or nitride single fuel rod behavior with a gas or liquid metal sublayer under normal and accident conditions of a liquid metal-cooled fast reactor operation. The BERKUT-U code's models, incorporating the MFPR/R code, are grounded in the contemporary understanding of mechanisms governing the fundamental processes in fuel rods under irradiation, which substantially enhances the code's predictive ability in comparison with the engineering analogs. Simulations of the nitride and oxide fuel rod behavior in BN-600 and BOR-60 fast reactors have demonstrated good agreement with the post-irradiation examination data. Further validation is foreseen as the corresponding data are available.

查看原文

微信好友朋友圈 QQ好友复制链接

本刊更多论文

燃料性能代码 BERKUT-U，用于模拟快堆中单根氧化物或氮化物燃料棒的堆内行为

本文介绍了俄罗斯科学院核安全研究所（IBRAE RAN）设计的燃料性能代码 BERKUT-U，该代码是 Proryv 项目 "新一代代码 "子项目的一部分。该代码旨在模拟在液态金属冷却快堆运行的正常和事故条件下，带有气态或液态金属下层的氧化物或氮化物单燃料棒的行为。BERKUT-U 代码的模型结合了 MFPR/R 代码，以当代对辐照下燃料棒基本过程的机制的理解为基础，与工程模拟相比，大大提高了代码的预测能力。对 BN-600 和 BOR-60 快堆中氮化物和氧化物燃料棒行为的模拟结果表明，与辐照后的检查数据非常吻合。随着相应数据的获得，预计将进行进一步验证。

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分钟内获得全文去求助

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.