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Treatment of waste salt arising from the pyrochemical treatment of used nuclear fuel using precipitation methods

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Abstract

There is currently significant interest in the pyrochemical processing of nuclear fuel for advanced fuel cycles. These high-temperature techniques generally use molten halides as electrolytes resulting in a fission product loaded salt waste stream. Due to its high aqueous solubility and low incorporation in standard borosilicate nuclear waste glasses, an effective treatment of the salt is key to demonstrating the economic feasibility and sustainability of pyro-processing. The preferred strategy for salt waste treatment is ‘clean-up’ where the fission products are separated from the salt (which can then be re-used in the main pyro-process) and immobilized in bespoke wasteforms. This allows high incorporations, minimizing final waste volumes and removing the water-soluble salt burden from the disposal facility. This paper describes some preliminary trials on the use of lithium and potassium phosphate and carbonate reagents as candidates for the removal of fission products from molten salt via a precipitation process.

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Notes

  1. A crystalline mixed lanthanide orthophosphate (LnPO4) mineral historically proposed for the immobilization of specific nuclear waste streams.

  2. The amount of precipitant was calculated to achieve a PO43− / CO32− to RE3+ and/or AE2+ mole ratio of 1.0.

  3. De-ionised water was used to remove the cooled salt in the carbonate experiments to avoid the precipitate also dissolving at pH ~ 1.

  4. LKE is deliquescent and requires handling under specialist inert atmosphere facilities.

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Acknowledgments

This research was funded under the £46m Advanced Fuel Cycle Programme as part of the Department for Business, Energy and Industrial Strategy’s (BEIS) £505m Energy Innovation Programme.

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Authors and Affiliations

  1. National Nuclear Laboratory, Sellafield, Seascale, CA20 1PG, Cumbria, UK

    M. T. Harrison & D. McKendrick

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  1. M. T. Harrison
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  2. D. McKendrick
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Correspondence to M. T. Harrison.

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Harrison, M.T., McKendrick, D. Treatment of waste salt arising from the pyrochemical treatment of used nuclear fuel using precipitation methods. MRS Advances 7, 117–121 (2022). https://doi.org/10.1557/s43580-022-00244-z

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