A model for fission gas release and gaseous swelling of the uranium dioxide fuel coupled with the FALCON code-Reference-Cited by-同舟云学术

A model for fission gas release and gaseous swelling of the uranium dioxide fuel coupled with the FALCON code

Published:2011-08 Issue:8 Volume:241 Page:2983-3007
ISSN:0029-5493
Container-title:Nuclear Engineering and Design
language:en
Short-container-title:Nuclear Engineering and Design

Author:

Khvostov G.,Mikityuk K.,Zimmermann M.A.

Funder

Eidgenössisches Nuklearsicherheitsinspektorat

Publisher

Elsevier BV

Subject

Mechanical Engineering,Waste Management and Disposal,Safety, Risk, Reliability and Quality,General Materials Science,Nuclear Energy and Engineering,Nuclear and High Energy Physics

Reference38 articles.

1. Isothermal grain growth kinetics in sintered UO2 pellets;Ainscough;J. Nucl. Mater.,1974

2. Development of the fission gas behaviour model in the START-3 code and its experimental support;Bibilashvili,2000

3. Interpretation of high-burnup fuel annealing tests;Blair;J. Nucl. Sci. Technol.,2008

4. Stochastic problems in physics and astronomy;Chandrasekhar;Rev. Mod. Phys.,1943

5. A study of xenon aggregates in uranium dioxide using X-ray absorption spectroscopy;Garcia;J. Nucl. Mater.,2006

Cited by 47 articles. 订阅此论文施引文献订阅此论文施引文献，注册后可以免费订阅5篇论文的施引文献，订阅后可以查看论文全部施引文献

1. In-Situ TEM Study of Microstructural Evolution in Proton Irradiated Single Crystal UO <sub>2</sub> under High-temperature Annealing;SSRN Electronic Journal;2024

2. Equation of state for helium-xenon gas mixture studied by molecular dynamics simulations;Journal of Nuclear Materials;2023-12

3. The SCIANTIX code for fission gas behaviour: Status, upgrades, separate-effect validation, and future developments;Journal of Nuclear Materials;2023-12

4. Analysis of thermal fuel behaviour under steady-state irradiation using selected cases of the IAEA CRP FUMEX;Journal of Nuclear Materials;2023-10

5. Calibration and Validation of thermal fuel behaviour models based on the first case of the first IAEA CRP FUMEX;Journal of Nuclear Materials;2023-10