Annealing tests of in-pile irradiated oxide coated U-Mo/Al-Si dispersed nuclear fuel

T. Zweifel; Ch Valot; Y. Pontillon; J. Lamontagne; A. Vermersch; L. Barrallier; T. Blay; W. Petry; H. Palancher

doi:10.1016/j.jnucmat.2014.05.052

Annealing tests of in-pile irradiated oxide coated U-Mo/Al-Si dispersed nuclear fuel

T. Zweifel, Ch Valot, Y. Pontillon, J. Lamontagne, A. Vermersch, L. Barrallier, T. Blay, W. Petry, H. Palancher

Department of Physics

Research output: Contribution to journal › Article › peer-review

8 Scopus citations

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.

Original language	English
Pages (from-to)	533-547
Number of pages	15
Journal	Journal of Nuclear Materials
Volume	452
Issue number	1-3
DOIs	https://doi.org/10.1016/j.jnucmat.2014.05.052
State	Published - Sep 2014

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title = "Annealing tests of in-pile irradiated oxide coated U-Mo/Al-Si dispersed nuclear fuel",

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.",

author = "T. Zweifel and Ch Valot and Y. Pontillon and J. Lamontagne and A. Vermersch and L. Barrallier and T. Blay and W. Petry and H. Palancher",

note = "Funding Information: 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 f{\"u}r Bildung und Forschung (BMBF) and the Bayerisches Staatsministerium f{\"u}r Wissenschaft, Forschung und Kunst (StMWFK).",

year = "2014",

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doi = "10.1016/j.jnucmat.2014.05.052",

language = "English",

volume = "452",

pages = "533--547",

journal = "Journal of Nuclear Materials",

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AU - Zweifel, T.

AU - Valot, Ch

AU - Pontillon, Y.

AU - Lamontagne, J.

AU - Vermersch, A.

AU - Barrallier, L.

AU - Blay, T.

AU - Petry, W.

AU - Palancher, H.

N1 - Funding Information: 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).

PY - 2014/9

Y1 - 2014/9

N2 - 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.

AB - 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.

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SP - 533

EP - 547

JO - Journal of Nuclear Materials

JF - Journal of Nuclear Materials

IS - 1-3

ER -

Annealing tests of in-pile irradiated oxide coated U-Mo/Al-Si dispersed nuclear fuel

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