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Author

Bang, In Cheol
Nuclear Thermal-Hydraulics & Reactor Safety Lab
Research Interests
  • Nuclear Thermal-Hydraulics

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Feasibility study of fuel cladding performance for application in ultra-long cycle fast reactor

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Title
Feasibility study of fuel cladding performance for application in ultra-long cycle fast reactor
Author
Jung, Ju AngKim, Seung HyunShin, Sang HunBang, In CheolKim, Ji Hyun
Keywords
FISSION-GAS RELEASE; IRRADIATION CREEP; FUSION APPLICATIONS; FERRITIC ALLOY; SIC COMPOSITES; METALLIC FUEL; DEFORMATION; BEHAVIOR; STEELS; MODEL
Issue Date
201309
Publisher
ELSEVIER SCIENCE BV
Citation
JOURNAL OF NUCLEAR MATERIALS, v.440, no.1-3, pp.596 - 605
Abstract
As a part of the research and development activities for long-life core sodium-cooled fast reactors, the cladding performance of the ultra-long cycle fast reactor (UCFR) is evaluated with two design power levels (1000 MWe and 100 MWe) and cladding peak temperatures (873 K and 923 K). The key design concept of the UCFR is that it is non-refueling during its 30-60 years of operation. This concept may require a maximum peak cladding temperature of 923 K and a cladding radiation damage of over 200 dpa (displacements per atom). Therefore, for the design of the UCFR, deformation due to thermal creep, irradiation creep, and swelling must be taken into consideration through quantitative evaluations. As candidate cladding materials for use in UCFRs, ferritic-martensitic (FM) steels, oxide dispersion strengthened (ODS) steels, and SiC-based composite materials are studied using deformation behavior modeling for a feasibility evaluation. The results of this study indicate that SiC is a potential UCFR cladding material, with the exception of irradiation creep due to high neutron fluence stemming from its long operating time of about 30-60 years.
URI
http://scholarworks.unist.ac.kr/handle/201301/3262
DOI
http://dx.doi.org/10.1016/j.jnucmat.2013.04.062
ISSN
0022-3115
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