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Preparation and characterization of UO2-based AGR SIMFuel

Published online by Cambridge University Press:  30 June 2014

Zoltan Hiezl,
David Hambley and
William E. Lee
Zoltan Hiezl
Affiliation:
Imperial College London, Centre for Nuclear Engineering, London, SW7 2AZ, UK
David Hambley
Affiliation:
Spent Fuel Management and Disposal, UK National Nuclear Laboratory (NNL), Springfields, PR4 0XJ, UK
William E. Lee
Affiliation:
Imperial College London, Centre for Nuclear Engineering, London, SW7 2AZ, UK
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Abstract

Preparation and characterization of a Simulated Spent Nuclear Fuel (SIMFuel), which replicates the chemical state and microstructure of Spent Nuclear Fuel (SNF) discharged from UK Advanced Gas-cooled Reactor (AGR) after a cooling time of 100 years is described. Thirteen stable elements were added to depleted UO2 and sintered to simulate the composition of fuel pellets after burn-ups of 25 and 43 GWd/tU and, as a reference, pure UO2 pellets were also investigated. The fission product distribution was calculated using the Fispin code provided by NNL. SIMFuel pellets exhibit a microstructure up to 92% TD. During the sintering process in H2 atmosphere Mo-Ru-Rh-Pd metallic precipitates and grey-phase ((Ba, Sr)(Zr, RE)O3 oxide precipitates) formed within the UO2 matrix. These secondary phases are present in real PWR and AGR SNF, although they are smaller in size than those examined in this study. The grain size of the produced SIMFuel is in good agreement with literature references.

Keywords

USEMsintering

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References

REFERENCES

Strausberg, S. and Murbach, E.W.. (1963). Multicycle Reprocessing and Refabrication Experiments on Sintered UO2-fissia Pellets. I&EC Process Design and Development, 2(3), 228–231.CrossRefGoogle Scholar
Lucuta, P.G., Verral, R.A., Matzke, H., and Palmer, B.J.. (1991). Microstructural Features of SIMFUEL-Simulated High-burnup UO2-based Nuclear Fuel. Journal of Nuclear Materials, 178, 48–60.CrossRefGoogle Scholar
Mignanelli, M.A. and Shaw, T.L., Information on Phase Chemistry in UK Spent AGR Fuels, NNL Report, (2009).Tech. No: 9734.Google Scholar
Shirsat, A.N., Ali, M., Kolay, S., Datta, A., and Das, D.. (2009). Transport Properties of I, Te and Xe in Thoria–Urania SIMFUEL. Journal of Nuclear Materials, 392(1), 16–21.CrossRefGoogle Scholar
Rauff-Nisthar, N., Boxall, C., Farnan, I., Hiezl, Z., Lee, W., Perkins, C., and Wilbraham, R.. (2013). Corrosion Behavior of AGR Simulated Fuels – Evolution of the Fuel Surface. Corrosion in Nuclear Energy Systems: From Cradle to Grave “ECS Transactions”, 53.Google Scholar
Grimes, R.W. and Catlow, C.R.A.. (1991). The Stability of Fission Products in Uranium Dioxide. Philosophical Transactions of the Royal Society A: Mathematical, Physical and Engineering Sciences, 335(1639), 609–634.Google Scholar
Lucuta, P.G., Matzke, H., and Verral, R.A.. (1995). Thermal Conductivity of Hyperstoichiometric SIMFUEL. Journal of Nuclear Materials, 223, 51–60.CrossRefGoogle Scholar
Cooper, M.W.D., Middleburgh, S.C., and Grimes, R.W.. (2013). Partition of Soluble Fission Products Between the Grey Phase, ZrO2 and Uranium Dioxide. Journal of Nuclear Materials, 438(1-3), 238–245.CrossRefGoogle Scholar
Morgan, S., Size, Grain and Variations, Porosity in Unfailed Spent AGR Fuel, NNL Memorandum, (2011).Google Scholar
Middleburgh, S.C., Parfitt, D.C., Grimes, R.W., Dorado, B., Bertolus, M., Blair, P.R., Hallstadius, L., and Backman, K.. (2012). Solution of Trivalent Cations into Uranium Dioxide. Journal of Nuclear Materials, 420(1-3), 258–261.CrossRefGoogle Scholar
Cooper, M.W.D., Middleburgh, S.C., and Grimes, R.W.. (2013). Swelling due to the partition of soluble fission products between the grey phase and uranium dioxide. Progress in Nuclear Energy, In Press.Google Scholar
Kleykamp, H.. (1989). Constitution and Thermodynamics of the Mo-Ru, Mo-Pd, Ru-Pd and Mo-Ru-Pd Systems. Journal of Nuclear Materials, 167, 49–63.CrossRefGoogle Scholar