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Zirconolite glass-ceramics are being developed as potential wasteforms for the disposition of Pu wastes in the UK. Previous studies utilised a variety of surrogates whilst this work uses both cold-press and sinter and hot isostatic press methods to validate the wasteform with PuO2. A cold press and sinter sample was fabricated as part of a validation study for plutonium incorporation in hot isostatically pressed (HIPed) wasteforms. The results confirmed the cold-press and sinter, achieved successful waste incorporation and a microstructure and phase assemblage that was in agreement with those expected of a HIPed equivalent. A HIP sample was fabricated of the same composition and characterised by SEM and XRD. Results were in agreement with the sintered sample and achieved complete waste incorporation into the glass-ceramic wasteform. These samples have demonstrated successful incorporation of PuO2 into glass-ceramic HIPed wasteforms proposed for processing Pu-based waste-streams in the UK.
Since the year 2000, Synroc has evolved from the titanate full-ceramic waste forms developed in the late 1970s to a hot isostatic pressing (HIP) technology platform that can be applied to produce glass, glass–ceramic, and ceramic waste forms and where there are distinct advantages over vitrification in terms of, for example, waste loading and suppressing volatile losses. This paper describes recent progress on waste form development for intermediate-level wastes from 99Mo production at ANSTO, spent nuclear fuel, fluoride pyroprocessing wastes and 129I. The microstructures and aqueous dissolution results are presented where applicable. This paper provides perspective on Synroc waste forms and recent process technology development in the nuclear waste management industry.
There is growing interest in reducing the use of ordinary Portland cement (OPC) owing to its high energy consumption and CO2 emissions. An environmentally-friendly alternative is the use of geopolymers, which can potentially reduce direct CO2 emissions through the appropriate choice of raw materials, mix design, and curing regimes. In this regard geopolymer mortars are also realistic candidates for the replacement of OPC mortars in nuclear waste immobilisation applications as they provide a more durable incorporation matrix as well as suppressing the formation of radiolytic hydrogen. The advantages of geopolymers over OPC for nuclear waste immobilisation include i) lower water content as alkaline activator is the main component that drives geopolymerisation, ii) higher thermal stability (<600°-800°C) compared to OPC concrete (<300°C), iii) higher compressive strength (50-80 MPa), and iv) lower leachability of radioactive ions when the mix design and curing temperature are appropriately balanced. UNSW and ANSTO have embarked on a long-term research program to investigate the possibility of using geopolymers for the immobilisation of Intermediate Level Liquid Waste (ILLW), the focus of which will be around the influence of gamma-irradiation on the durability.
Synroc has evolved over the last 40 years from the titanate full-ceramics developed in the late 1970s to a technology platform that can be applied to produce glass, glass–ceramic, and ceramic waste forms and where there are distinct advantages in terms of waste loading and suppressing volatile losses.
A first of a kind Synroc plant for immobilizing intermediate level waste arising from Mo-99 production is currently in detailed engineering at ANSTO.
Since the year 2000, Synroc has evolved from the titanate full-ceramics developed in the late 1970s to a technology platform that can be applied to produce glass, glass–ceramic, and ceramic waste forms and where there are distinct advantages in terms of waste loading and suppressing volatile losses. Furthermore recent efforts have focused strongly on waste form development for plutonium-bearing wastes in the UK, for different options for the immobilization of Idaho calcines and most recently developing an engineered waste form for the intermediate level wastes arising from 99Mo production, for the Australian Nuclear Science and Technology Organisation (ANSTO). A variety of other studies are currently in progress, including engineered waste forms for spent fuel and investigating the proliferation risks for titanate-based waste forms containing highly enriched uranium or plutonium. This paper also attempts to give some perspective on Synroc waste forms and process technology development in the nuclear waste management industry.
Various transmission routes contribute to spread of carbapenem-resistant Klebsiella pneumoniae (CRKP) in hospitalized patients. Patients with readmissions during which CRKP is again isolated (“CRKP readmission”) potentially contribute to transmission of CRKP.
To evaluate CRKP readmissions in the Consortium on Resistance against Carbapenems in K. pneumoniae (CRaCKLe).
Cohort study from December 24, 2011, through July 1, 2013.
Multicenter consortium of acute care hospitals in the Great Lakes region.
All patients who were discharged alive during the study period were included. Each patient was included only once at the time of the first CRKP-positive culture.
All readmissions within 90 days of discharge from the index hospitalization during which CRKP was again found were analyzed. Risk factors for CRKP readmission were evaluated in multivariable models.
Fifty-six (20%) of 287 patients who were discharged alive had a CRKP readmission. History of malignancy was associated with CRKP readmission (adjusted odds ratio [adjusted OR], 3.00 [95% CI, 1.32–6.65], P<.01). During the index hospitalization, 160 patients (56%) received antibiotic treatment against CRKP; the choice of regimen was associated with CRKP readmission (P=.02). Receipt of tigecycline-based therapy (adjusted OR, 5.13 [95% CI, 1.72–17.44], using aminoglycoside-based therapy as a reference in those treated with anti-CRKP antibiotics) was associated with CRKP readmission.
Hospitalized patients with CRKP—specifically those with a history of malignancy—are at high risk of readmission with recurrent CRKP infection or colonization. Treatment during the index hospitalization with a tigecycline-based regimen increases this risk.
Infect. Control Hosp. Epidemiol. 2016;37(3):281–288
A series of uranium-containing gadolinium zirconate samples have been fabricated at 1723 K in air. X-ray diffraction and Raman spectroscopy have confirmed pyrochlore or defect fluorite structures, while diffuse reflectance, X-ray absorption near edge structure and X-ray photoelectron spectroscopies indicate a predominantly U6+ oxidation state, even when Ca2+ was added to charge balance for U4+. The results demonstrate the potential of gadolinium zirconates as host materials for actinides.
Research has been carried out to optimize the consolidation stage for the immobilization of pyrochemical wastes with a sodium aluminophosphate glass. The alternative techniques of hot pressing and hot isostatic pressing of the calcined wastes with the glass have been investigated. This has been performed on simulant waste material and the products investigated by scanning electron microscopy and X‑ray diffraction. The consolidation techniques were compared to each other and to the original process for suitability as a waste-form.
Calcium and barium zirconium phosphates were prepared by hot isostatic pressing and their thermophysical properties investigated for potential use as actinide hosts for inert matrix fuels (IMF) in light water reactors. The materials are thermally stable up to at least 1600°C in air, however they degrade above around 1400°C in an inert atmosphere. The heat capacity and thermal conductivity were measured from room temperature up to 1200°C. The thermal conductivity coefficient for both CZP and BZP at 1000°C is 1.0 W m-1 K-1, a relatively low thermal conductivity that requires NZP-type materials to be dispersed in a composite cercer or cermet IMF.
With the increasing demand for the development of nuclear power comes the responsibility to address the issue of waste, including the technical challenges of immobilizing high-level nuclear wastes in stable solid forms for interim storage or disposition in geologic repositories. The immobilization of high-level nuclear wastes has been an active area of research and development for over 50 years. Borosilicate glasses and complex ceramic composites have been developed to meet many technical challenges and current needs, although regulatory issues, which vary widely from country to country, have yet to be resolved. Cooperative international programs to develop advanced proliferation-resistant nuclear technologies to close the nuclear fuel cycle and increase the efficiency of nuclear energy production might create new separation waste streams that could demand new concepts and materials for nuclear waste immobilization. This article reviews the current state-of-the-art understanding regarding the materials science of glasses and ceramics for the immobilization of highlevel nuclear waste and excess nuclear materials and discusses approaches to address new waste streams.
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.
A synroc-D ceramic consisting mostly of spinel, hollandite, pyrochlore-structured CaUTi2O7, UO2, and Ti-rich regions shows promise for immobilisation of a HLW containing mainly Al and U, together with fission products. Ceramics with virtually zero porosities and waste loadings of 50-60 wt% on an oxide basis were prepared by cold crucible melting (CCM) at ∼1500°C, and also by subsolidus hot isostatic pressing (HIP) at 1100°C to prevent volatile losses. PCT leaching test values for Cs were < 13 g/L, with all other normalised elemental extractions being well below 1 g/L.
We have characterized thermally annealed perovskite (CaTiO3) surfaces, both before and after aqueous dissolution testing, using scanning electron microscopy, cross-sectional transmission electron microscopy, x-ray photoelectron spectroscopy, and atomic force microscopy. It was shown that mechanical damage caused by polishing was essentially removed at the CaTiO3 surface by subsequent annealing; such annealed samples were used to study the intrinsic dissolution behavior of perovskite in deionized water at RT, 90 °C, and 150 °C. Our results indicate that, although mechanical damage caused higher Ca release initially, it did not affect the long-term Ca dissolution rate. However, the removal of surface damage by annealing did lead to the subsequent spatial ordering of the alteration product, which was identified as anatase (TiO2) by both x-ray and electron diffraction, on CaTiO3 surfaces after dissolution testing at150 °C. The effect of Ca2+ in the leachant on the dissolution reaction of perovskite at 150 °C was also investigated, and the results suggest that under repository conditions, the release of Ca from perovskite is likely to be significantly slower if Ca2+ is present in ground water.
We have studied the aqueous durability of pyrochlore-structured yttrium-titanate (Y2Ti2O7) and Nd/Al-bearing zirconolite [(Ca0.8Nd0.2)Zr(Ti1.8Al0.2)O7] in both neutral and acidic solutions, with and without the presence of 0.001 M of NaF. Scanning Electron Microscopy (SEM), X-ray Diffraction (XRD), X-ray Photoelectron Spectroscopy (XPS) and Atomic Force Microscopy (AFM) were used to characterize the composition, structure and morphology of the pyrochlore (Y2Ti2O7) and zirconolite surfaces, both before and after static dissolution testing at 90 and 150°C for four weeks. The leachates were also analyzed using Inductively Coupled Plasma Mass Spectrometry (ICP-MS) to estimate the individual elemental releases. The results show that the presence of F- ions only has a significant effect in acidic media on the dissolution behavior of pyrochlore and zirconolite. This detrimental effect is more pronounced for pyrochlore than zirconolite; the Y2Ti2O7 surface was replaced completely by alteration products after dissolution testing at 150°C for 4 weeks in acidic media with 0.001 M fluoride ions.
In the early 1980s a synroc variant, SYNROC-D, was developed for immobilisation of high-level defence waste stored at the Savannah River Plant, USA. A key phase in the immobilisation matrix was spinel, used to immobilise the large proportion of iron and alumina in the waste. Here we examine the feasibility of this approach for other alumina-rich wastes, not necessarily containing iron, derived from the dissolution of aluminium fuel cladding. The advantages of using a magnesia spinel, as opposed to hercynite (FeAl2O4), as the primary alumina-bearing phase are discussed in terms of an increase in waste loading and process flexibility. Two options for sodium incorporation, glass and the titanate phase freudenbergite, are considered.
Static dissolution tests on Nd-bearing zirconolite were conducted in deionized water at 150°C for up to 6 months. Surfaces, both before and after aqueous dissolutions, were examined using X-ray Photoelectron Spectroscopy (XPS) and Scanning Electron Microscopy (SEM). Individual submicron-sized crystals wereobserved only on some zirconolite grains after hydrothermal treatment for one week, and were identified as brookite (possibly plus anatase) by Transmission Electron Microscopy (TEM). The number of secondary crystals present on the zirconolite surface was, however, reduced significantly after 6 months of durability testing. The results of electron microscopy studies are consistent with those obtained from XPS in that the average Ca and Ti concentrations of the top surface layer (< 5 nm) decreased slightly with respect to Zr, while the average Al concentration increased after durability testing.
Optical emission spectra in the 300-700 nm range were collected from zirconolite and rutile specimens irradiated with a 3 μs pulsed electron beam using a Febetron 706 variable energy pulsed electronbeam generator. The long-lived emissions (up to microseconds after the electron pulse) consist of broad (halfwidths ~ 100 nm) bands centred around ~400 nm. Over the range 0.2 MeV to 0.6 MeV, the emission intensity per unit dose versus electron beam energy data from the rutile sample showed a single stage dependence on electron beam energy, whereas the zirconolite data suggested a two stage dependence. Rutile has a threshold of 0.23 ½ 0.02 MeV, which gives an Ed value of 39 ½ 4 eV for oxygen. Zirconolite has a threshold of 0.26 ½ 0.02 MeV, which gives an Ed value of 45 ½4 eV for oxygen. These data are discussed in the context of previously measured and calculated Ed values for other oxides.
Eight perovskites of different compositions and pre-existing vacancy contents were irradiated with 1.5 MeV Kr+ ions using the HVEM-Tandem User Facility at Argonne National Laboratory. The critical dose of 1.5 MeV Kr+ ions for amorphisation (Dc) of NaNbO3, SrTiO3 and two natural perovskites varies with composition and may decrease with Na content. The D,c values of 5 members of the solid solution series from SrTiO3 to La0.67TiO3 do not exhibit a simple relationship with vacancy content: In particular, Dc(Sr0 7La0.2TiO3) is extraordinarily high (>170 × 1014 ions cm−2) compared to those of the other (Sr, La)-perovskites in the series, which have Dc values between 2.4 and 11 × 1014 ions cm−2.
Eight perovskites of different compositions and pre-existing vacancy contents were irradiated with 1.5 MeV Kr+ ions using the HVEM-Tandem User Facility at Argonne National Laboratory. The critical dose of 1.5 MeV Kr+ ions for amorphisation (Dc) of NaNbO3, SrTiO3 and two natural perovskites varies with composition and may decrease with Na content. The Dc values of 5 members of the solid solution series from SrTiO3 to La0.67TiO3 do not exhibit a simple relationship with vacancy content: In particular, Dc(Sr0.7La0.2TiO3) is extraordinarily high (>170 × 1014 ions cm−2) compared to those of the other (Sr, La)-perovskites in the series, which have Dc, values between 2.4 and 11×1014 ions cm−2.
New X-ray diffraction and scanning electron microscopy data are given for the incorporation of Np and Pu in zirconolite, at levels of tens of percent. The actinide valences and the cations they replace are deduced from the microanalysis of the zirconolite compositions, and X-ray absorption data are used to obtain more direct information on the valences of Ce and Nd, which are used as simulants of Pu and trivalent actinides respectively. Trivalent rare earths and actinides have extensive solid solubility in zirconolite, mainly but not exclusively in the Ca site. Tetravalent rare earths and actinides have considerable solid solubility in the Zr site of zirconolite, and some solubility in the Ca site, but the strong tendency of zirconolite with ions substituted in the Zr site to undergo phase separation complicates structural interpretation. In zirconolite-rich Synroc-type ceramics designed to immobilise waste actinides, the target actinide waste loading has been set at 20 wt% and early leach results indicate the durability is at least as good as that of Synroc-C.