Preparation and Characterization of a Calcium Phosphate Ceramic for the Immobilization of Chloride-containing Intermedia
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Preparation and Characterization of a Calcium Phosphate Ceramic for the Immobilization of Chloride-containing Intermediate Level Waste Brian L. Metcalfe, Ian W. Donald, Randy D. Scheele* and Denis M. Strachan* Atomic Weapons Establishment, Aldermaston, Berkshire, UK *Pacific Northwest National Laboratory, Richland, WA, USA ABSTRACT Attention has recently been given to the immobilization of special categories of radioactive wastes, some of which contain high concentrations of actinide chlorides. Although vitrification in phosphate glass has been proposed, this was rejected because of the high losses of chloride through the mobilization of volatile species. On the basis of non-radioactive and, more recently, radioactive studies, we have shown that calcium phosphate is an effective host for immobilizing the chloride constituents [1]. In this instance, the chlorine is retained as chloride, rather than evolved as a chlorine-bearing gas. The immobilized product is in the form of a free-flowing, non-hygroscopic powder, in which the chlorides are chemically combined within the synthetic mineral phases chlorapatite [Ca5(PO4)3Cl] and spodiosite [Ca2(PO4)Cl]. Data from studies on non-radioactive simulated waste consisting of a mixture of CaCl2 and SmCl3, and radioactive simulated waste composed of CaCl2 with PuCl3 or PuCl3 and AmCl3, are presented and compared. The XRD data confirm the presence of chlorapatite and spodiosite in the non-radioactive and radioactive materials. The durability of all specimens was measured with a modified MCC1 test. Normalized releases of Cl after 28 days were 1.6 x 10-3 g m-2 for the non-radioactive specimens and 7 x 10-3 g m-2 for the Pu bearing specimens. Releases of Ca after 28 days were 0.3 x 10-3 and 2.0 x 10-3 g m-2 for the non-radioactive composition and the Pu composition, respectively, whilst release of Pu from the radioactive specimens was higher for the mixed Pu/Am specimen at 1.2 x 10-5g m-2 than for the Pu only specimens. The release of Am from the mixed Pu/Am composition was exceptionally low at 2.4 x 10-7 g m-2. Overall, the release rate data suggest that the ceramics dissolve congruently, followed by precipitation of Sm, Pu and Am as less soluble phases, possibly oxides or phosphates. The differences in behaviour noted between non-radioactive and radioactive specimens are interpreted in terms of the crystal chemistry of the individual systems. INTRODUCTION Techniques being developed to treat chloride-containing wastes that result from the reprocessing of certain types of spent nuclear fuel or the pyrochemical reprocessing of plutonium metal include vitrification, utilizing iron phosphate or lead borosilicate based glasses [2-6, or reacting the waste directly with ammonium dihydrogen phosphate with subsequent heating to yield a phosphate based glass, together with volatile ammonium chloride [7-9]. These processes remove chloride in the form of volatile species, rather than immobilizing the chloride chemically into the host. They therefore require suitable off-gas treatment
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