Annealing tests of in-pile irradiated oxide coated U–Mo/Al–Si dispersed nuclear fuel
dc.contributor.author
hal.structure.identifier | ZWEIFEL, T.
|
dc.contributor.author
hal.structure.identifier | VALOT, Ch.
|
dc.contributor.author
hal.structure.identifier | PONTILLON, Y.
|
dc.contributor.author
hal.structure.identifier | LAMONTAGNE, J.
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dc.contributor.author | VERMERSCH, A. |
dc.contributor.author | BLAY, T. |
dc.contributor.author | PETRY, W. |
dc.contributor.author
hal.structure.identifier | PALANCHER, H.
|
dc.contributor.author
hal.structure.identifier | BARRALLIER, Laurent
|
dc.date.accessioned | 2014 |
dc.date.available | 2014 |
dc.date.issued | 2014 |
dc.date.submitted | 2014 |
dc.identifier.issn | 0022-3115 |
dc.identifier.uri | http://hdl.handle.net/10985/8337 |
dc.description | Authors do acknowledge the MERARG team for their experimental work (CEA) and F. Charollais, J. Noirot and finally B. Kapusta for their advices and comments. This study was supported by a combined Grant (FRM0911) of the Bundesministerium für Bildung und Forschung (BMBF) and the Bayerisches Staatsministerium für Wissenschaft, Forschung und Kunst (StMWFK). |
dc.description.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. |
dc.language.iso | en |
dc.publisher | Elsevier |
dc.rights | Post-print |
dc.subject | Combustible nucléaire |
dc.subject | Microstructure |
dc.title | Annealing tests of in-pile irradiated oxide coated U–Mo/Al–Si dispersed nuclear fuel |
dc.identifier.doi | 10.1016/j.jnucmat.2014.05.052 |
dc.typdoc | Article dans une revue avec comité de lecture |
dc.localisation | Centre de Aix en Provence |
dc.subject.hal | Sciences de l'ingénieur: Matériaux |
ensam.audience | Internationale |
ensam.page | 533-547 |
ensam.journal | Journal of Nuclear Materials |
ensam.volume | 452 |
ensam.issue | 1-3 |
hal.identifier | hal-01021884 |
hal.version | 1 |
hal.status | accept |
dc.identifier.eissn | 0022-3115 |