M. W. A. Stewart

Leaching Behavior of Zirconolite-Rich Synroc Used to Immobilize High-Fired Plutonium Oxide

This study reports on the use of zirconolite-rich Synroc to demonstrate the safe immobilisation of ‘high-fired’ Pu0 2 . The zirconolite-rich Synroc used in this study was prepared by adding 13 wt% Pu with equimolar amounts of Gd and Hf, relative to Pu, as neutron absorbers. The incorporation of the Pu and neutron absorbers has been studied microstructurally as well as by longer-term leach testing. This work has shown that the sintered ceramic can immobilise 13 wt% of Pu with almost complete incorporation of the Pu (≃ 98%) into the zirconolite phase. Durability studies have shown that under a wide range of leaching conditions there is no major separation of the Pu and neutron absorbers, with the majority of these elements either remaining in the matrix or leaching at low ( −4 g m −2 d −1 ) and comparable rates from the waste form.

Immobilization of sodium and potassium in Synroc

Substitution of K for Na in certain nuclear fuel reprocessing cycles may allow an increase of waste loading in Synroc, because K can be incorporated in the barium hollandite phase more easily than Na. The use of rare-earth additions to stabilize Na in the perovskite phase may also have merit.

Durabilities of Pyrochlore-Rich Titanate Ceramics Designed for Immobilization of Surplus Plutonium

The chemical durabilities of two Pu-doped pyrochlore samples were studied by Single-Pass-Flow-Through (SPFT) tests at 70°C. The dissolution of pyrochlore is incongruent with preferential releases of Ca and Gd over Ti, close to stoichiometric releases of U and Ti, and lower releases of Hf and Pu than Ti. Altered pyrochlore and polymorphs of TiO 2 (brookite and probably anatase) have been identified on the surface of the leached sample and the principal secondary phase is an unknown polymorph of TiO 2 containing Hf and varying amounts of Gd and Pu. These surface alteration phases are consistent with reported studies of natural samples. The releases of U, Gd, Ca and Ti into solution follow linear kinetics, whereas the releases of Pu and Hf exhibit non-linear behavior. The presence of ∼5% PuO 2 and trace amounts of glass does not appear to have an effect on the overall durability of the material. Further, the low Pu release rate and the similar kinetics for Pu and Hf releases limit the possibility of nuclear criticality under repository conditions. Overall, this study provides useful information on the lower bounds of durabilities of the materials.

The replacement of titanium by zirconium in ceramics for plutonium immobilization

Zirconates and titanates, based on the nominal baseline composition developed for the Plutonium Immobilization Project, have been prepared with and without process impurities. The titanates form pyrochlore as the major phase and the zirconates form a defect-fluorite. Very little, if any, of each impurity is accommodated in the defect-fluorite with powellite, kimzeyite, a spinel and a silicate glass appearing as extra phases in this ceramic. In the titanate ceramics the pyrochlore incorporates more impurities, with the remainder forming zirconolite and a small amount of silicate glass. At extreme levels of impurities, traces of magnetoplumbite, perovskite and loveringite were found. The defect-fluorite zirconate phase is more radiation damage resistant than the titanate pyrochlore, though the secondary phases in the zirconate will reduce the radiation damage resistance of zirconate monoliths. To produce a dense product the oxide-route zirconate required sintering temperatures of about 1550 C, 200 C higher than that required for the titanate. Silicate impurities reduce the sintering temperatures appreciably.

Advantages Hot Isostatically Pressed Ceramic and Glass-Ceramic Waste Forms Bring to the Immobilization of Challenging Intermediate- and High-Level Nuclear Wastes

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.

Immobilisation of Higher Activity Wastes from Nuclear Reactor Production of 99Mo

A variety of intermediate- and low-level liquid and solid wastes are produced from reactor production of 99Mo using UAl alloy or UO2 targets and in principle can be collectively or individually converted into waste forms. At ANSTO, we have legacy acidic uranyl-nitrate-rich intermediate level waste (ILW) from the latter, and an alkaline liquid ILW, a U-rich filter cake, plus a shorter lived liquid stream that rapidly decays to low-level waste (LLW) standards, from the former. The options considered consist of cementitious products, glasses, glass-ceramics, or ceramics produced by vitrification or hot isostatic pressing for intermediate-level wastes. This paper discusses the progress in waste form development and processing to treat ANSTO’s ILW streams arising from 99Mo. The various waste forms and the reason for the process option chosen will be reviewed. We also address the concerns over adapting our chosen process for use in a hot-cell environment.

Ionic Size Limits for A Ions in Brannerite (ATi 2 O 6 ) and Pyrochlore (CaATi 2 O 7 ) Titanate Structures ( A = tetravalent rare earths and actinides)

The lower limit of the size of the octahedral A 4+ ion in the ATi 2 O 6 brannerite structure is just smaller than that of Ce/Pu. Attempts to expand the A ion size beyond that of Th by (a) substituting a Ba ion plus two U 5+ ions for three A ions or (b) substituting one Ba plus one hexavalent ion for two A ions did not succeed. Ge, Sn and Zr substitutions in the Ti site of ThTi 2 O 6 do not exceed 0.2 formula unit in ceramic preparations. These and other coupled substitutions in the B site of ThTi 2 O 6 showed that the average B site size could tolerate deviations of 4+ is unusually stabilised in air atmospheres at temperatures close to the melting point of 1400°C in the A site of brannerite. Lattice parameter data on different endmember ATi 2 O 6 brannerites are given. The lower and upper size limits for the eightfold A ions in the pyrochlore structure are around 0.100 and 0.117 nm respectively. A BaUTi 2 O 7 stoichiometry did not produce a pyrochlore structure, and when fired in either argon or air yielded a mixture of BaUTiO 6 , whose structure is still uncertain, plus brannerite and rutile.

The Use of Hot-Isostatic Pressing to Process Nuclear Waste Forms

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.Copyright

Progress at ANSTO on a Synroc Plant for Intermediate-Level Waste from Reactor Production of 99Mo

Intermediate level waste from ANSTO’s expanded 99Mo production plant will consist of ~5000L/year of 6M NaOH + 1.4 NaAlO2 + fission products. Detailed engineering is being carried out on a synroc plant to immobilise this waste in a glass-ceramic, with completion scheduled for 2016. The liquid waste will be mixed with precursors and dried before being calcined in a reducing atmosphere to control fission product volatility. The calcine will be transferred to 30L metal cans which will be hot isostatically pressed at 1000°C/30MPa for 2h, then cooled to room temperature and stored preparatory to final disposal. Laboratory scale waste form material will pass 90°C PCT tests. In addition, legacy intermediate level uranyl nitrate-based liquid waste from 99Mo production at ANSTO between the 1980s and 2005 via irradiation of UO2 targets will also be immobilised by the same process to form a Synroc-type waste form. Some examples illustrating the wide applicability of hot isostatic pressing to consolidate nuclear waste forms will be given showing the advantages for particular wastes, notably high waste loadings and the absence of off-gas in the high temperature consolidation step. The immobilisation of a variety of low-level liquid and solid wastes from 99Mo production will also be discussed.

Durability of Pu-doped Titanate and Zirconate Ceramics Designed for Pu Immobilisation

Ceramics rich in pyrochlore-structured titanate and fluorite-structured zirconate phases designed for surplus Pu immobilisation, with and without process impurities, have been leach tested at 90°C in deionised water. The zirconates consist mainly of a defect-fluorite with secondary impurity-containing phase-powellite/scheelite (when sintered in Ar but not when sintered in air), a spinel or magnetoplumbite type phase, a glass forming silicate and a secondary U-rich phase (when sintered in air with added impurities). The pyrochlore-rich baseline titanate ceramic consists of pyrochlore, brannerite and Hf-rutile. When impurities are added zirconolite and a silicate glass are also present. The pyrochlore-rich titanate with 5 wt% of impurities sintered at 1300°C is highly durable. A well-densified zirconate version without impurities has comparable elemental releases to those of the titanate ceramic but a zirconate with 5 wt% of impurities sintered at 1400°C in air or Ar shows much higher U and Ca releases than the titanate ceramic. Sintering atmospheres, changing from Ar to air, can influence Pu and U release rates up to an order of magnitude. High Ga releases from zirconates with impurities show that the secondary phase containing Ga is not durable. The higher processing temperature and the apparent inability to incorporate many impurity elements suggest that zirconates are not as flexible as titanates in respect of processing conditions and aqueous durability.