Commenced in January 2007
Frequency: Monthly
Edition: International
Paper Count: 30458
Evaluation of As-Cast U-Mo Alloys Processed in Graphite Crucible Coated with Boron Nitride

Authors: Tercio Pedrosa, Kleiner Marques Marra

Abstract:

This paper reports the production of uranium-molybdenum alloys, which have been considered promising fuel for test and research nuclear reactors. U-Mo alloys were produced in three molybdenum contents: 5 wt.%, 7 wt.%, and 10 wt.%, using an electric vacuum induction furnace. A boron nitride-coated graphite crucible was employed in the production of the alloys and, after melting, the material was immediately poured into a boron nitride-coated graphite mold. The incorporation of carbon was observed, but it happened in a lower intensity than in the case of the non-coated crucible/mold. It is observed that the carbon incorporation increased and alloys density decreased with Mo addition. It was also noticed that the increase in the carbon or molybdenum content did not seem to change the as-cast structure in terms of granulation. The three alloys presented body-centered cubic crystal structure (g phase), after solidification, besides a seeming negative microsegregation of molybdenum, from the center to the periphery of the grains. There were signs of macrosegregation, from the base to the top of the ingots.

Keywords: Solidification, uranium-molybdenum alloys, incorporation of carbon, macrosegregation and microsegregation

Digital Object Identifier (DOI): doi.org/10.5281/zenodo.3669279

Procedia APA BibTeX Chicago EndNote Harvard JSON MLA RIS XML ISO 690 PDF Downloads 83

References:


[1] Y. S. Kim; G. L. Hofman. Fission product induced swelling of U-Mo alloy fuel. Journal of Nuclear Materials. 419. p. 291-301 (2011).
[2] M. Ugajim et al. Irradiation behavior of high uranium-density alloys in the plate fuels. Journal of Nuclear Materials. 254. p. 78-83 (1988).
[3] J. L. Snelgrove et al. Development of very-high-density low-enriched-uranium fuels. Nuclear Engineering and Design. v.178. n. 1. p.119-126 (1997).
[4] S.Negoy et al.. Microstructural study of gamma phase stability in U-9wt% Mo alloy. Journal of Nuclear Materials. 422. p. 77-85 (2012).
[5] M. K. Meyer et al. Low-temperature irradiation behavior of uranium-molybdenum alloy dispersion fuel. Journal of Nuclear Materials. 304. p. 221-236 (2002).
[6] S. Rudolph. Boron nitrite release coatings. High-performance ceramic performance. Interceram. V. 42. N. 5. p. 302-305 (1993).
[7] D. Adorno et al. Thermal cycling effect in U-10Mo/Zry-4 monolithic nuclear fuel. Journal of Nuclear Materials. 473. p. 136-142 (2016).
[8] T. S. Pedrosa et al. Phase transitions during artificial ageing of segregated as-cast U-Mo alloys. Journal of Nuclear Materials. 457. p. 100-117 (2015).
[9] M. Ugajin; J. Asbe; M. Kurihara. Phase behavior and thermodynamics of U-Mo-C system. Journal of Nuclear Science and Technology. V. 12. N. 9. p. 560-566 (1975).
[10] V. P. Sinha et al. Development. preparation and characterization of uranium molybdenum alloys for dispersion fuel application. Journal of Alloys and Compounds. 473. p. 238-244 (2009).
[11] B. R. T. Frost. The carbides of uranium. Journal of Nuclear Materials. N. 4. p. 265-300 (1963).
[12] A. Devaraj at al. Theoretical model for volume fraction of UC. 235U enrichment and effective density of final U-10Mo alloy. Pacific Northwest National Laboratory. Washington. USA. 9p (2016).