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Journal Articles

OECD/NEA benchmark on pellet-clad mechanical interaction modelling with fuel performance codes; Influence of pellet geometry and gap size

Soba, A.*; Prudil, A.*; Zhang, J.*; Dethioux, A.*; Han, Z.*; Dostal, M.*; Matocha, V.*; Marelle, V.*; Lasnel-Payan, J.*; Kulacsy, K.*; et al.

Proceedings of TopFuel 2021 (Internet), 10 Pages, 2021/10

Journal Articles

The Effect of base irradiation on failure behaviors of UO$$_{2}$$ and chromia-alumina additive fuels under simulated reactivity-initiated accidents; A Comparative analysis with FEMAXI-8

Udagawa, Yutaka; Mihara, Takeshi; Taniguchi, Yoshinori; Kakiuchi, Kazuo; Amaya, Masaki

Annals of Nuclear Energy, 139, p.107268_1 - 107268_9, 2020/05

AA2019-0372.pdf:0.81MB

 Times Cited Count:3 Percentile:29.11(Nuclear Science & Technology)

Journal Articles

Analytical study of SPERT-CDC test 859 using fuel performance codes FEMAXI-8 and RANNS

Taniguchi, Yoshinori; Udagawa, Yutaka; Amaya, Masaki

Annals of Nuclear Energy, 139, p.107188_1 - 107188_7, 2020/05

 Times Cited Count:1 Percentile:9.57(Nuclear Science & Technology)

Journal Articles

OECD/NEA benchmark on pellet-clad mechanical interaction modelling with fuel performance codes; Impact of number of radial pellet cracks and pellet-clad friction coefficient

Dost$'a$l, M.*; Rossiter, G.*; Dethioux, A.*; Zhang, J.*; Amaya, Masaki; Rozzia, D.*; Williamson, R.*; Kozlowski, T.*; Hill, I.*; Martin, J.-F.*

Proceedings of Annual Topical Meeting on Reactor Fuel Performance (TopFuel 2018) (Internet), 10 Pages, 2018/10

Journal Articles

Coupled computer code study on irradiation performance of a fast reactor mixed oxide fuel element with an emphasis on the fission product cesium behavior

Uwaba, Tomoyuki; Nemoto, Junichi*; Ishitani, Ikuo*; Ito, Masahiro*

Nuclear Engineering and Design, 331, p.186 - 193, 2018/05

 Times Cited Count:4 Percentile:34.56(Nuclear Science & Technology)

A computer code for the analysis of the overall irradiation performance of a fast reactor mixed-oxide (MOX) fuel element was coupled with a specialized code for the analysis of fission product cesium behaviors in a MOX fuel element. The coupled code system allowed for the analysis of the radial and axial Cs migrations, the generation of Cs chemical compounds and fuel swelling due to Cs-fuel-reactions in association with the thermal and mechanical behaviors of the fuel element. The coupled code analysis was applied to the irradiation performance of a fast reactor MOX fuel element attaining high burnup for discussion on the axial distribution of Cs, fuel-to-cladding mechanical interaction owing to the Cs-fuel-reactions by comparing the calculated results with post irradiation examinations.

Journal Articles

Pellet-cladding mechanical interaction of PWR fuel rod under rapid power transient

Yanagisawa, Kazuaki; Katanishi, Shoji;

Journal of Nuclear Science and Technology, 31(7), p.671 - 676, 1994/07

 Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)

no abstracts in English

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