Journal Article

IMPULSE-2015-00030

http://join2-wiki.gsi.de/foswiki/pub/Main/Artwork/join2_logo100x88.png Annealing tests of in-pile irradiated oxide coated U–Mo/Al–Si dispersed nuclear fuel

Zweifel, T. (Corresponding Author)MLZ* ; Valot, C. ; Pontillon, Y. ; Lamontagne, J. ; Vermersch, A. ; Barrallier, L. ; Blay, T. ; Petry, W.MLZ* ; Palancher, H.

2014
Elsevier Science Amsterdam [u.a.]

Please use a persistent id in citations: doi:10.1016/j.jnucmat.2014.05.052

Abstract: U–Mo/Al based nuclear fuels have been worldwide considered as a promising high density fuel for the conversion of high flux research reactors from highly enriched uranium to lower enrichment. In this paper, we present the annealing test up to 1800 °C of in-pile irradiated U–Mo/Al–Si fuel plate samples. More than 70% of the fission gases (FGs) are released during two major FG release peaks around 500 °C and 670 °C. Additional characterisations of the samples by XRD, EPMA and SEM suggest that up to 500 °C FGs are released from IDL/matrix interfaces. The second peak at 670 °C representing the main release of FGs originates from the interaction between U–Mo and matrix in the vicinity of the cladding.

Keyword(s): Chemical Reactions and Advanced Materials (1st) ; Key Technologies (1st) ; Fundamental Science (1st) ; Materials Science (2nd)

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Contributing Institute(s):

1. RPHYSICS

Experiment(s):

1. HEU-MEU: Brennstoffentwicklung

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Database coverage:
Current Contents - Engineering, Computing and Technology ; Current Contents - Physical, Chemical and Earth Sciences ; IF < 5 ; JCR ; Nationallizenz ; SCOPUS ; Science Citation Index ; Science Citation Index Expanded ; Thomson Reuters Master Journal List ; Web of Science Core Collection

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