Thermophysical and Mechanical Analyses of UO2-36.4vol % BeO Fuel Pellets with Zircaloy, SiC, and FeCrAl Claddings
Author:
Publisher
MDPI AG
Subject
General Materials Science,Metals and Alloys
Link
http://www.mdpi.com/2075-4701/8/1/65/pdf
Reference24 articles.
1. Enhanced thermal conductivity of uranium dioxide–silicon carbide composite fuel pellets prepared by Spark Plasma Sintering (SPS)
2. Thermal Conductivity of UO2-BeO Pellet
3. Accident tolerant fuels for LWRs: A perspective
4. Recession of silicon carbide in steam under nuclear plant loca conditions up to 1400 °C;Markham,2012
5. Fabrication and measurement of hoop strength of SiC triplex tube for nuclear fuel cladding applications
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1. BISON analysis of FeCrAl and Zircaloy cladding deformation during simulated BWR cyclic dryout conditions;Journal of Nuclear Materials;2023-04
2. Modeling of heat transfer in a fuel pellet based on uranium dioxide and ceramics (beryllium oxide);Kompleksnoe Ispolʹzovanie Mineralʹnogo syrʹâ/Complex Use of Mineral Resources/Mineraldik Shikisattardy Keshendi Paidalanu;2021-09-12
3. BeO Utilization in Reactors for the Improvement of Extreme Reactor Environments - A Review;Frontiers in Energy Research;2021-05-10
4. Impact of fission gas bubbles on thermal conductivity of UO2 fuels with high thermal conductivity additives;Journal of Nuclear Materials;2021-04
5. Recent studies on potential accident-tolerant fuel-cladding systems in light water reactors;Nuclear Science and Techniques;2020-03
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