Microstructural analysis of MTR fuel plates damaged by a coolant flow blockage

Leenaers, A.; Joppen, F.; Berghe, S. Van den

doi:10.1016/j.jnucmat.2009.08.014

Cited by 8 publications

(3 citation statements)

References 14 publications

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“…The TEM examination of the fuel shows that for both samples, the U(Mo) kernels are still crystalline at a rather high burnup of approximately 32% 235 U. A similar observation is made on atomized U(Mo) fuel but it differs from currently used research reactor fuel such as UAl x or U 3 Si 2 fuel, which turn almost instantly amorphous under low temperature irradiation [28,29]. In the fuel kernels, fission gas nanobubbles were observed in out-of-focus bright field images.…”

Section: Fission Gas Behaviorsupporting

confidence: 55%

Irradiation behavior of ground U(Mo) fuel with and without Si added to the matrix

Leenaers

Berghe

Renterghem

et al. 2011

Journal of Nuclear Materials

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Section: Fission Gas Behaviorsupporting

confidence: 55%

Irradiation behavior of ground U(Mo) fuel with and without Si added to the matrix

Leenaers

Berghe

Renterghem

et al. 2011

Journal of Nuclear Materials

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“…The most detailed study has been performed on an UAl x /Al fuel plate which had undergone a coolant flow blockage incident [43]. If blister tests on in-pile irradiated U 3 Si 2 plates have been reported in literature [39], very few associated destructive examinations are provided.…”

Section: Microstructural Evolutions In Temperaturesmentioning

confidence: 99%

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

Zweifel¹,

Valot²,

Pontillon³

et al. 2014

Journal of Nuclear Materials

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is an open access repository that collects the work of Arts et Métiers ParisTech researchers and makes it freely available over the web where possible. a b s t r a c t 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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“…However, much like the UAl x fuel does not show fission gas bubble formation up to relatively high burnup under normal operating conditions [21], bubbles are not observed inside the interaction phase in the case of U(Mo) either. To be precise, the voids considered responsible for the fuel plate pillowing are formed between the interaction layer and the matrix and can be considered to result from a transport of gas from the U(Mo) kernel through the interaction layer.…”

Section: Dispersion Fuelsmentioning

confidence: 88%

From High to Low Enriched Uranium Fuel in Research Reactors

Berghe

Leenaers

Koonen

et al. 2010

Advances in Science and Technology

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Since the 1970's, global efforts have been going on to replace the high-enriched (>90% 235U), low-density UAlx research reactor fuel with high-density, low enriched (<20% 235U) replacements. This search is driven by the attempt to reduce the civil use of high-enriched material because of proliferation risks and terrorist threats. American initiatives, such as the Global Threat Reduction Initiative (GTRI) and the Reduced Enrichment for Research and Test Reactors (RERTR) program have triggered the development of reliable low-enriched fuel types for these reactors, which can replace the high enriched ones without loss of performance. Most success has presently been obtained with U3Si2 dispersion fuel, which is currently used in many research reactors in the world. However, efforts to search for a replacement with even higher density, which will also allow the conversion of some high flux research reactors that currently cannot change to U3Si2 (eg. BR2 in Belgium), have continued and are for the moment mainly directed towards the U(Mo) alloy fuel (7-10 w% Mo). This paper provides an overview of the past efforts and presents the current status of the U(Mo) development.

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Microstructural analysis of MTR fuel plates damaged by a coolant flow blockage

Cited by 8 publications

References 14 publications

Irradiation behavior of ground U(Mo) fuel with and without Si added to the matrix

Irradiation behavior of ground U(Mo) fuel with and without Si added to the matrix

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

From High to Low Enriched Uranium Fuel in Research Reactors

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