Thermal treatment of uranium oxide irradiated in pressurized water reactor: Swelling and release of fission gases-Reference-Cited by-同舟云学术

Thermal treatment of uranium oxide irradiated in pressurized water reactor: Swelling and release of fission gases

Published:1998-06 Issue:2-3 Volume:255 Page:85-91
ISSN:0022-3115
Container-title:Journal of Nuclear Materials
language:en
Short-container-title:Journal of Nuclear Materials

Author:

Zacharie I,Lansiart S,Combette P,Trotabas M,Coster M,Groos M

Publisher

Elsevier BV

Subject

Nuclear Energy and Engineering,General Materials Science,Nuclear and High Energy Physics

Reference10 articles.

1. I. Zacharie, Doctoral thesis, Ecole Centrale Paris, 1997.

2. Th. Bredel, Production des gaz de fission Xe et Kr dans les combustibles UO2, Rapport interne CEA (codes Apollo et Pépin), 1993.

3. C. Baker, J.C. Killeen, Fission gas release during post irradiation annealing of UO2, BNES Conf. on Materials for Nuclear Core Applications, Paper 24, 1987.

Cited by 46 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. Effect of grain size on irradiation-induced bubble evolution in high burn-up UO2: A phase-field study;Journal of Nuclear Materials;2024-12

2. Impact of grain boundary and surface diffusion on predicted fission gas bubble behavior and release in UO2 fuel;Journal of Nuclear Materials;2024-06

3. Evolution of extended defects in UO2 during high temperature annealing;Journal of Nuclear Materials;2024-05

4. Phase-field simulation study on fission gas release behavior of large grain UO2 fuel;Acta Physica Sinica;2024

5. Extended Development of a Fission Gas Release Behavior Model Inside Spherical Fuel Grains for LWR Reactors;Applied Sciences;2023-09-18