The structure of lead-technetium pyrochlore has been refined in space group F d 3 ¯ m with a = 10.36584(2) Å using a combination of synchrotron X-ray and neutron powder diffraction data and confirmed via Electron Diffraction. The oxide is found to be oxygen deficient with a stoichiometry of Pb2Tc2O7-d. Displacive disorder of the Pb cations is evident from the refinements, as has been observed in Bi2Tc2O7-d. X-ray absorption spectroscopic measurements at the Tc K-edge demonstrate the valence of the Tc is greater than 4.0 as anticipated from the refined oxygen stoichiometry. Raman spectroscopy confirms the presence of disorder leading us to conclude that this pyrochlore is the first example of a valence V technetium oxide.

L.P. Hatch, “Ultimate Disposal of Radioactive Wastes.” American Scientist, vol. 41, pp. 410-421 (1953). R.E. Isaacson and L.E. Brownell, “Ultimate Storage of Radioactive Wastes in Terrestrial Environments.” Manage ment of Radioactive Wastes from Fuel Reprocessing, OECD Proceedings, Paris, pp. 953-986 (1972). G.J. McCarthy and M.T. Davidson, “Ceramic Nuclear Waste Forms: I, Crystal Chemistry and Phase Formation.” Bull. Am. Ceram. Soc., vol. 54, pp. 782–786 (1975). R.W. Lynch, R.G. Dosch, B.T. Kenna, J.K. Johnstone, and E.J. Nowak, “The Sandia Solidification Process-A Broad Range Aqueous Waste Solidification Method.” Proceedings of the IAEA Symposium on the Management of Radioactive Waste, IAEA-SM-207/75, Vienna, Austria, pp. 361-373 (Mar. 1976). R. Roy, “Rational Molecular Engineering of Ceramic Mate rials,” J. Am. Ceram. Soc., vol. 60, pp. 3580-35963 (1977). A.E. Ringwood, “Safe Disposal of High Level Nuclear Reactor Wastes: A New Strategy,” Australian National Uni versity Press, Canberra, Australia and Norwalk, CT (1978). A.E. Ringwood, S.E. Kesson, N.G. Ware, W. Hibberson, and A. Major, “Immobilisation of High Level Nuclear Reactor Wastes in Synroc,” Nature, vol. 278, pp. 219–223 (1979). A.G. Solomah, T.M. Hare, and H. Palmour III, “Demon stration of the Feasibility of Subsolidus Sintering of Rad waste-Containing Synroc-B Composition,” Nucl. Technol., vol. 49, pp. 183–185 (1980). J. Campbell, C. Hoenig, F. Bazan, F. Ryerson, M. Guinan, R. Van Konynenburg, and R. Rozsa, “Properties of Synroc-D Nuclear Waste Form: A State-of-the-Art Review.” UCRL-53240, pp. 1-23, Lawrence Livermore National Laboratory, Livermore, CA (Jan. 1982). S.E. Kesson and A.E. Ringwood, “Safe Disposal of Spent Nuclear Fuel,” Radioactive Waste Management and the Nuclear Fuel Cycle, vol. 4(2), pp. 159–174 (1983).

We present evidence for possibly the highest magnetic ordering temperature in any compound without 3d transition elements. Neutron powder diffraction measurements, at both time-of-flight and constant wavelength sources, were performed on two independently prepared SrTcO3 powders. SrTcO3 adopts a distorted perovskite structure with G-type antiferromagnetic ordering and has a moment of 1.87(4)μB per Tc cation at room temperature with an extraordinarily high Néel point close to 750 °C. Electronic structure calculations reveal extensive mixing between the technetium 4d states and oxygen states proximal to the Fermi level. This hybridization leads to a close relationship between magnetic ordering temperature and moment formation in SrTcO3.

Solid solubility limits of U, Pu, and the neutron absorbers Hf and Gd have been measured for zircon (ZrSiO4), monazite (CePO4), titanite (CaTiSiO5), perovskite (CaTiO3), apatite (Ca10(PO4)6O), in almost all cases where these limits were not known beforehand. The method used was to oversaturate the host phase with the dopant, using a nominated substitutional scheme, and then establish the dopant content of the host phase by microanalysis/scanning electron microscopy. Tetravalent U has limited solid solubilities in titanite, perovskite and apatite. X-ray absorption near-edge and diffuse reflectance spectroscopies were used to show that U was tetravalent in U-doped perovskite prepared in both argon and hydrogen-nitrogen atmospheres, with different charge compensating schemes. Tetravalent Pu has solubilities of 0.13 and 0.02 formula units (f.u.) in perovskite and titanite respectively. Trivalent Pu has a solubility of 0.05 f.u. in titanite. Pu3+ dominates tetravalent Pu in monazite fired in air at 1400°C. At least 0.5 and < 0.1 f.u. of Hf are soluble in titanite and monazite respectively.Hf solubility in apatite is estimated as < 0.1 f.u. Approximately 0.3 and < 0.1 f.u. of Gd are soluble in titanite and zircon respectively

The technetium perovskite CaTcO(3) has been synthesized. Combining synchrotron X-ray and neutron diffraction, we found that CaTcO(3) is an antiferromagnetic with a surprisingly high Neel temperature of ∼800 K. The transition to the magnetic state does not involve a structural change, but there is obvious magnetostriction. Electronic structure calculations confirm the experimental results.

AbstractThe title compound is prepared by solid state reaction of NH4TcO4 and Ca(NO3)2 (flowing Ar, 700 °C, 1 h).

Sessions C603system.The crystal structure was determined by single crystal X-ray diffraction (Xcalibur diffractometr, CCD detector, Мо Кα radiation).The structure belongs to TbFe 2 structure type (space group R-3m, a = 5.5900(8) Å, c = 13.684(3)Å), which is deformed variant of MgCu 2 structure type.

The structure of the perovskite SrTcO(3) has been investigated using both synchrotron X-ray and neutron powder diffraction. At room temperature SrTcO(3) is orthorhombic as a consequence of cooperative tilting of the corner sharing TcO(6) octahedra. The tilts are sequentially removed as the sample is heated with the oxide displaying the sequence of structres Pnma→Imma→I4/mcm→Pm ̅3m. Neutron powder diffraction data collected in the temperature range 4-1023 K indicate that SrTcO(3) has G-type antiferromagnetic structure, in which each Tc moment is antiparallel to its six nearest neighbours, below ∼1000 K. The magnetic structure is collinear antiferromagnetic with the technetium moments parallel to c-axis and can be described by the propagation vector k = [0,0,0] and the basis vector (0,0,A(z)). The same magnetic structure is observed in each of the four crystal structures.

X‐ray photoelectron spectroscopy was used to detect an intergranular impurity film in two (BaxCsy)(MnzTi8−z)O16 hollandite samples sintered in argon and air, respectively. This impurity film is enriched in Cs and Ba, as a result of segregation during high‐temperature processing. The Cs ions in the grain‐boundary layer have a different chemical bonding state from that in the hollandite phase, as evidenced by a large chemical shift of the Cs 3d and Cs 4d levels. The chemical environment around the Ba ions in the intergranular film is also likely to be different from that in the hollandite. The presence of this grain‐boundary film is responsible for the relatively high initial loss of Cs and Ba during the aqueous dissolution testing.

ANSTO has developed a combination of tailored nuclear waste form chemistries coupled with the use of flexible hot-isostatic pressing processing technology to enable the successful incorporation of problematic nuclear wastes into dense, durable monoliths. This combined package also enables the design of waste forms with waste loadings well in excess of those achievable via baseline melting routes using borosilicate glass, as hot-isostatic pressing is not constrained by factors such as glass viscosity, crystallisation and electrical conductivity. In this paper we will discuss some of our experiences with problematic wastes, namely plutonium wastes, sludges and HLW such as the Idaho calcines.

Hot isostatic pressing (HIP) is a technology with wide applicability in consolidating calcined intermediate-level and high-level nuclear waste, especially with wastes that are not able to be readily processed by vitrification at reasonable waste loadings. The essential process steps during the HIP cycle will be outlined. We have demonstrated the effective consolidation via HIP technology of a wide variety of tailored glass-ceramic and ceramic waste forms, notably simulated ICPP waste calcines, I sorbed upon zeolite beads, Pu-bearing wastes, inactive Cs/Sr/Rb/Ba mixtures, simulated waste pyroprocessing salts from spent nuclear fuel recycling, Tc, U-rich isotope production waste, and simulated K-basin (Hanford, WA, USA) and Magnox sludges (UK). Can-ceramic interactions have been carefully studied. The principal advantages of the HIP technology include: negligible offgas during the high temperature consolidation step, relatively small footprint, and high waste loadings. As a batch process, the wasteform chemistry can be readily adjusted on a given process line, to deliver wastes into different end states (e.g. direct HIP versus chemically tailored). This flexibility allows the treatment of multiple waste streams on the one process line.

Analysis of multiphase (Ca,U) titanate pyrochlore samples produced under argon or in air, using laboratory X‐ray powder diffraction (XRD) and electron microscopy revealed discrete pyrochlore phase compositions, together with either perovskite or rutile impurities. Investigations of these samples using diffuse reflectance and X‐ray photoelectron spectroscopies revealed mixed U 4+ /U 5+ oxidation states for argon‐annealed samples and U 5+ /U 6+ oxidation states for air‐annealed samples. Single‐phase (Ca 1.25 U 4+ 0.25 U 5+ 0.50 )Ti 2 O 7 (argon) and (Ca 1.40 U 5+ 0.60 )(Ti 1.90 U 6+ 0.10 )O 7 (air) compositions were synthesized and characterized using a combination of synchrotron X‐ray and neutron powder diffraction. Severe intensity reduction of hkl odd XRD peaks from the sample produced in air was found to be the result of U 6+ sharing the Ti site. The refined crystal structures for both single‐phase samples were found to be consistent with the compositions and mixed oxidation states observed by the above spectroscopic measurements. Although the normalized uranium leach rate of the argon‐annealed sample was found to be approximately twice that of the air‐annealed sample, in 7‐day Product Consistency Test the durability of both compounds with respect to leaching by water was found to be excellent. Ca releases were in the range of 0.01–0.03 g/L and U releases were <3 × 10 −5 g/L.

Zirconolite-based titanate ceramics containing U plus Th or Pu have been prepared. The final consolidation to produce a dense monolithic waste form was carried out using hot isostatic pressing (HIPing) of the calcined materials within a stainless steel can. The ceramics were characterised and tested for their overall feasibility to immobilise impure Pu or separated actinide-rich radioactive wastes. As designed, tetravalent U and Pu are mainly incorporated in a durable zirconolite phase, together with Gd or Hf added as neutron absorbers. The interaction of the waste form with the HIP can was also examined. No changes in the U valences or the U/Pu-bearing phase distributions were observed at the waste form–HIP can interface.

Hot isostatically pressed tailored hollandite waste forms were used to demonstrate the immobilization of Cs and Sr are separable from spent nuclear fuel, as well as Ba and Rb. Four hollandite formulations were investigated, two samples with ∼12 wt% waste loading (on an oxide basis) and two with ∼18 wt% waste loading. Two of the samples were Al‐substituted and the other two contained Mg. The hollandite in the Al‐substituted samples contained all the waste cations, as designed, but this was not the case in the Mg‐substituted samples. The hollandite in the Mg‐substituted samples did not contain all the waste cations, with ∼50% of the Sr forming SrTiO3 as a secondary phase. This resulted in waste forms that were not as durable, with respect to Cs, as their Al counterparts. The formation of SrTiO3 had little effect on the Sr release rates and was not detrimental to the Mg‐substituted hollandite waste form. For the Al‐substituted samples, the MCC‐1 normalized release rates were <0.06 g·(m2·day)−1 at 0–28 days for all elements, while the Cs release rates remained at 2.0 g·(m2·day)−1 at 0–28 days for the Mg‐substituted samples.

Hot isostatically pressed (HIPed) glass-ceramics for the immobilization of uranium-rich intermediate-level wastes and Hanford K-basin sludges were designed. These were based on pyrochlore-structured Ca(1-x)U(1+y)Ti2O7 in glass, together with minor crystalline phases. Detailed microstructural, diffraction and spectroscopic characterization of selected glass-ceramic samples has been performed, and chemical durability is adequate, as measured by both MCC-1 and PCT-B leach tests.

Uranium-rich liquid wastes arising from UO2 targets which have been neutron-irradiated to generate medical radioisotopes such as 99mTc require immobilisation. A pyrochlore-rich hot isostatically pressed titanate ceramic can accommodate at least 40wt% of such waste expressed on an oxide basis. In this paper, the baseline waste form composition (containing 40wt% UO2) was adjusted in two ways: (a) varying the UO2 loading with constant precursor oxide materials, (b) varying the precursor composition with constant waste loading of UO2. This resulted in the samples having a similar phase assemblage but the amounts of each phase varied. The oxidation states of U in selected samples were determined using diffuse reflection spectroscopy (DRS) and electron energy loss spectroscopy (EELS). Leaching studies showed that there was no significant difference in the normalised elemental release rates and the normalised release rates are comparable with those from synroc-C. This demonstrates that waste forms based on titanate ceramics are robust and flexible for the immobilisation of U-rich waste streams from radioisotope processing.

The influence of CeO2 additions on the electrochemical behaviour of the MnO2 cathode in a Zn–MnO2 battery using lithium hydroxide (LiOH) as an electrolyte is investigated using microscopy and spectroscopic techniques. The results showed that such additions greatly improve the discharge capacity of the battery (from 155 to 190mAhg−1) but only from the second discharge cycle onwards. Capacity fade with subsequent cycling is also greatly reduced. With an aim to understand the role of CeO2 on the discharge–charge characteristics of MnO2 and its mechanism, we have used a range of microscopy, spectroscopy and diffraction-based techniques to study the process. The CeO2 is not modified by multiple discharged and charged cycles. The CeO2 may enhance the discharge–charge performance of the battery by raising the oxygen evolution potential during charging but does not take part directly in the redox reaction.

This paper illustrates the benefits of hot isostatically pressed (HIPed) tailored ceramic waste forms for the immobilisation of Cs and Sr separated from spent nuclear fuel. Experimental data on microstructure and aqueous durability are presented for Cs- and Sr-bearing hollanditerich tailored ceramics prepared with 12-18 wt% waste (on an oxide basis). MCC-1 type leach testing, on the sample containing 12 wt% waste at 90°C for 28 days revealed extremely low normalised 7-28-day Cs and Sr release rates of 0.003 and 0.004 g/m2day respectively.