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Browsing by Author "Gregg, DJ"

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Candidate glass–ceramic wasteforms for the immobilisation of Cs-loaded IONSIV® wastes: a scoping study
(Springer Nature, 2024-03-28) Bahmanrokh, G; Whitelock, E; Dayal, P; Farzana, R; Koshy, P; Gregg, DJ
In some cases, nuclear wastes can be treated with ion exchange materials to remove specific radionuclides from solution via cationic exchange. A promising inorganic ion exchange material, crystalline silicotitanate (CST) or IONSIV®, has been previously employed to remove Cs-137 from contaminated aqueous systems with high specificity. Once the radioactive Cs-137 has been incorporated within the IONSIV® structure, the ion exchange material itself becomes radioactive waste and requires immobilisation within a nuclear wasteform. The current scoping study investigated design and development of advanced glass–ceramic wasteforms for the immobilisation of Cs-loaded IONSIV®. Two well-established Cs-bearing ceramic phases, hollandite, and pollucite, were considered as the ceramic component of the novel glass–ceramic design. Hollandite appeared to react with the borosilicate glass-component to form celsian and rutile. The pollucite system produced a phase assemblage of pollucite, rutile, srilankite, and glass, as targeted, and is therefore considered a promising wasteform design for Cs-loaded IONSIV® material. © 2024 Springer Nature.
Cation antisite disorder in uranium-doped gadolinium zirconate pyrochlores
(Elsevier, 2014-09-01) Gregg, DJ; Zhang, ZM; Thorogood, GJ; Kennedy, BJ; Kimpton, JA; Griffiths, GJ; Guagliardo, PR; Lumpkin, GR; Vance, ER
The incorporation of uranium into gadolinium zirconate (Gd2Zr2O7) is investigated by synchrotron X-ray powder diffraction and X-ray absorption near-edge structure (XANES) spectroscopy. The results suggest that the uranium cation is largely located on the pyrochlore B-site instead of the targeted A-site. Cation disorder in Gd2Zr2O7 and U-doped Gd2Zr2O7 is investigated by positron annihilation lifetime spectroscopy (PALS) which demonstrates the absence of cation vacancies in these systems. This work provides direct evidence for cation antisite (A- and B-site mixing) disorder in U-doped and off-stoichiometric Gd2Zr2O7 pyrochlore. © 2014, Elsevier B.V.
Ceramic conversion and densification of zirconium phosphonate sorbent materials
(Elsevier, 2018-04-01) Veliscek-Carolan, J; Thorogood, GJ; Gregg, DJ; Tansu, M; Hanley, TL
The simple conversion of zirconium phosphonate sorbent materials, with known affinity for lanthanide elements, to durable ceramic waste forms via thermal treatment has been demonstrated. The use of zirconium phosphonate enables both removal of targeted elements from spent nuclear fuel and immobilisation into leach resistant solid products to be achieved using a single material. Thermal conversion was performed on the zirconium phosphonate both before and after loading with europium, which acted as a surrogate for the chemically similar minor actinides. Without europium loaded, the zirconium phosphonate sorbent formed predominantly KZr2(PO4)3 upon heating, independent of the processing conditions used. A maximum relative density of 87% was achieved with cold isostatic pressing (200 MPa) and sintering at 1200 °C for 12 h. When the zirconium phosphonate sorbent was loaded with europium, the phase composition formed upon thermal treatment was more complex. Specifically, mixtures of ZrP2O7, Eu0.33Zr2(PO4)3, EuPO4 and Zr2O(PO4)2 were formed, with phase compositions depending on the temperatures and pressures used. The simplest phase composition for the europium loaded material was achieved via uniaxial pressing (120 MPa) and sintering at 1300 °C for 1 h, although the ceramic pellet produced under these conditions had a relative density of only 53%. The loaded europium deported primarily to a EuPO4 phase, which is known to be highly stable and leach resistant. As such, these zirconium phosphonate materials have potential utility for treatment of nuclear wastes. © 2019 Elsevier B.V.
Characterisation of hot isostatically pressed (HIPed) hollandite wasteform-canister interaction zone
(Elsevier, 2024-02) Mann, J; Farzana, R; Aughterson, RD; Dayal, P; Sorrell, CC; Koshy, P; Gregg, DJ
A potential hollandite wasteform for immobilising waste containing Cs, Ba, Sr, and Rb, projected from a solvent extraction process that separates Cs/Sr from spent nuclear fuel, was fabricated via hot isostatic pressing (HIPing) within a stainless-steel (SS) canister at 1250 °C / 30 MPa / 2 h. Before HIPing, 2 wt.% Ti metal was added to the precursor, as a redox control additive. Detailed elemental profiling and microstructural analysis at the interaction zone between the wasteform and the SS HIP canister were thoroughly investigated with transmission electron microscopy (TEM) using a lamella extracted by focused ion beam (FIB) milling, scanning electron microscopy (SEM) and X-ray diffraction (XRD). The interaction zone towards the wasteform was ∼20–30 µm in distance and in this region, a hollandite composition with varying chemistry was observed relative to the bulk wasteform. Moreover, the regular Cr-oxide layer, often observed previously for HIPed Synroc-type materials, was not present due to the achievement of reducing condition by adding Ti-metal as redox additive and simultaneous diffusion of canister material towards the ceramic. Predominant Cr-diffusion was observed with incorporation in the hollandite phase along with minor Fe, Mn and Co from the SS canister. This study provides a detailed understanding of the HIP canister – wasteform interaction zone for a hollandite-rich wasteform design for the first time. Importantly, no deleterious phases were formed that may otherwise reduce the performance of the wasteform. This study further demonstrates the flexibility of HIPing as a consolidation process for the treatment of radioactive wastes. © 2023 Elsevier B.V.
Crystal chemistry and structures of uranium-doped gadolinium zirconates
(Elsevier, 2013-07-01) Gregg, DJ; Zhang, YJ; Zhang, ZM; Karatchevtseva, I; Blackford, MG; Triani, G; Lumpkin, GR
A series of uranium-containing gadolinium zirconate samples have been fabricated at 1450 °C in oxidizing, inert and reducing atmospheres. X-ray diffraction, Raman spectroscopy and transmission electron microscopy have been utilized to confirm adoption of pyrochlore or defect fluorite structures. X-ray diffraction allowed determination of the bulk averaged structure while Raman spectroscopy and transmission electron microscopy were used to determine ordering at the microdomain scale. Diffuse reflectance, X-ray absorption near edge structure and X-ray photoelectron spectroscopies indicated a predominantly U6+ oxidation state for all the air-sintered samples, even when Ca2+ or A-site vacancies were incorporated to charge balance for U4+, a mixed U5+/U6+ oxidation state was found for samples sintered in argon, while a mixed U4+/U5+ oxidation state occurred for sintering under N2–3.5%H2. This demonstrates a degree of uranium oxidation state control through sintering conditions, and the potential of using gadolinium zirconates as host materials for uranium in nuclear waste applications.© 2013, Elsevier B.V.
Crystal structure and phase transitions in the uranium perovskite, Ba2SrUO6
(Elsevier, 2013-02-01) Reynolds, EM; Kennedy, BJ; Thorogood, GJ; Gregg, DJ; Kimpton, JA
The structure of one of the oxides proposed to be present in the grey phase of irradiated mixed oxide fuel, the double perovskite Ba2SrUO6 has been investigated from room temperature to 1300 K using synchrotron X-ray powder diffraction methods. The divalent strontium and hexavalent uranium are found to be fully ordered in the double-perovskite arrangement of alternating octahedra sharing corner oxygen atoms. At room temperature Ba2SrUO6 adopts a monoclinic structure in space group P21/n. Heating to above 900 K induces a first order transition to a rhombohedral structure, and further heating to above 1200 K results in a continuous transition to a cubic structure. The sequence of structures is associated with the progressive loss of cooperative tilting of the corner sharing SrO6 and UO6 octahedra. © 2012, Elsevier B.V.
Current advances on titanate glass-ceramic composite materials as waste forms for actinide immobilization: a technical review
(Elsevier, 2022-05) Zhang, YJ; Kong, L; Ionescu, M; Gregg, DJ
As the emerging versatile waste forms for immobilizing actinide-rich radioactive wastes, glass-ceramic composite materials based on some durable ceramic phases are being developed. They have apparent advantages over the conventional borosilicate glasses and multi- or single- phase ceramics as they essentially combine the chemical and processing flexibilities of glasses to accommodate processing impurities and excellent chemical durability of ceramic phases to host actinides. More recently, some new advances have been made on scientific and technological aspects including new glass-ceramic systems; improved understanding of ceramic phase evolution in glass; actinide validation studies and simplified processing techniques. This review is intended to cover the current advances on the development of glass-ceramic composite waste forms focusing on titanate ceramic phases (zirconolite, pyrochlore and brannerite) for immobilizing various actinide-rich radioactive wastes arising from the nuclear fuel cycle. © 2021 Published by Elsevier Ltd.
Effect of Ti‐metal addition on hot‐isostatically pressed (HIPed) Synroc‐C
(Wiley, 2023-11) Farzana, R; Dayal, P; Peristyy, A; Sutton, P; Aly, Z; Aughterson, RD; Nguyen, TH; Yeoh, M; Koshy, P; Gregg, DJ
Synroc, a candidate nuclear wasteform and Synroc technology, a waste treatment solution utilizing hot‐isostatic pressing (HIPing) have significant potential for the immobilisation of challenging nuclear wastes from both current and innovative reactors and fuel cycles. Hot isostatic press (HIP) consolidation is undertaken within sealed metal HIP canisters, where metal buffers (e.g., Ti, Fe and Ni) can be incorporated to control the redox environment within the canister. This study, for the first time, reports the effect of varying Ti‐metal addition (0, 2, 4, and 8 wt.%) on phase formation, microstructural characteristics, and wasteform performance for HIP consolidated Synroc‐C containing 20 wt.% simulated PUREX type (PW‐4b) high level waste. Quantitative X‐ray diffraction analysis, scanning electron microscopy‐energy dispersive X‐ray spectroscopy (EDS) and transmission electron microscopy‐EDS analyses were undertaken for analytical investigations. The chemical durability of the samples was assessed using ASTM C1220‐21 standard test. Hot‐isostatically pressed (HIPed) samples with 0 and 8 wt.% Ti added for redox control produced unfavourable phase formation. However, the HIPed samples with Ti additions of 2 and 4 wt.% as a redox buffer showed the desired phase formation of Synroc‐C without any significant change to the partitioning of waste elements among the phases along with compatible durability results, when compared to previous literature for hot uniaxial pressing (HUPed) or sintered materials. © 2023 Commonwealth of Australia and The Authors. Journal of the American Ceramic Society published by Wiley Periodicals LLC on behalf of American Ceramic Society. - Open Access CC-BY-NC.
Formation of (Cr,Al)UO4 from doped UO2 and its influence on partition of soluble fission products
(Elsevier, 2013-11-01) Cooper, MWD; Gregg, DJ; Zhang, YJ; Thorogood, GJ; Lumpkin, GR; Grimes, RW; Middleburgh, SC
CrUO4 and (Cr, Al)UO4 have been fabricated by a sol–gel method, studied using diffraction techniques and modelled using empirical pair potentials. Cr2O3 was predicted to preferentially form CrUO4 over entering solution into hyper-stoichiometric UO2+x by atomic scale simulation. Further, it was predicted that the formation of CrUO4 can proceed by removing excess oxygen from the UO2 lattice. Attempts to synthesise AlUO4 failed, instead forming U3O8 and Al2O3. X-ray diffraction confirmed the structure of CrUO4 and identifies the existence of a (Cr, Al)UO4 phase for the first time (with a maximum Al to Cr mole ratio of 1:3). Simulation was subsequently used to predict the partition energies for the removal of fission products or fuel additives from hyper-stoichiometric UO2+x and their incorporation into the secondary phase. The partition energies are consistent only with smaller cations (e.g. Zr4+, Mo4+ and Fe3+) residing in CrUO4, while all divalent cations are predicted to remain in UO2+x. Additions of Al had little effect on partition behaviour. The reduction of UO2+x due to the formation of CrUO4 has important implications for the solution limits of other fission products as many species are less soluble in UO2 than UO2+x. © 2013, Elsevier B.V.
Gd2Zr2O7 and Nd2Zr2O7 pyrochlore prepared by aqueous chemical synthesis
(Elsevier, 2013-12-01) Kong, L; Karatchevtseva, I; Gregg, DJ; Blackford, MG; Holmes, R; Triani, G
Pyrochlore structured Gd2Zr2O7 and Nd2Zr2O7 are produced via complex precipitation processing. A suite of characterization techniques, including FTIR, Raman, X-ray and electron diffraction, TEM, SEM as well as nitrogen sorption are employed to investigate the structural and grain size evolution of the synthesized and calcined powder. Results show that Gd2Zr2O7 with the pyrochlore structure are produced after calcination at 1400 °C for 12 h while Nd2Zr2O7 has already formed the pyrochlore structure at 1200 °C. This method allows the formation of dense materials at relatively low temperature, with bulk densities over 92% of the theoretical values achieved after sintering at 1400 °C for 50 h. This unique aqueous synthetic method provides a simple pathway to produce pyrochlore lanthanide zirconate without using either organic solvent and/or mechanical milling procedures, making the synthesis protocol an attractive potential scale-up production of highly refractory ceramics. © 2013, Elsevier Ltd.
Glass-bonded ceramic waste forms for immobilization of radioiodine from caustic scrubber wastes
(Elsevier, 2024-01-26) Lere-Adams, AJ; Wilkins, MCD; Bollinger, D; Stariha, S; Farzana, R; Dayal, P; Gregg, DJ; Chong, S; Riley, BJ; Heiden, ZM; McCloy, JS
Glass-bonded sodalite composite waste forms have been developed for the immobilization of liquid radioactive wastes resulting from off-gas treatment during aqueous reprocessing of used nuclear fuel, with a particular focus on 129I. The proposed composite waste form is comprised of aluminosilicate ceramic phases containing volatile radionuclides bonded with a glassy matrix. In this work, a suite of ten candidate low-temperature glass binders (ZnO-Bi2O3-based glasses and a Na2O-B2O3-SiO2 glass) were examined. Six glasses were mixed with caustic scrubber waste simulant previously converted into a sodalite-rich material (to provide glass fractions of 10 and 20 wt.%), uniaxially pressed into pellets, and sintered at 350 °C or 550 °C for 8 h in air. Iodine retention after heat treatment was assessed by neutron activation analysis, showing retention of 67–100 % of expected iodine. The aqueous durabilities of the resulting materials were then determined, following the ASTM C1308 standard test, showing iodine releases of 1 to 23 g m−2 after 4 d. The cumulative iodine release for the best performing system (a zinc-bismuth-borate glass binder) was <1 g m−2, and its iodine retention from processing was 67 %. The iodine releases compared favorably with other waste forms. In parallel, this best-performing composition was also consolidated via hot isostatic pressing (HIP) in a stainless-steel canister at 550 °C for 2 h under 100 MPa pressure. The HIPed sample was produced at the ∼20 g scale and showed improved densification and minimal reaction with the canister. © 2024 Elsevier B.V.
He and Au ion radiation damage in sodalite, Na4Al3Si3O12Cl
(Elsevier, 2014-10) Vance, ER; Gregg, DJ; Davis, J; Ionescu, M
Sodalite, a candidate ceramic for the immobilisation of pyroprocessing nuclear waste, showed no observable lattice dilatation in grazing incidence X-ray diffraction when irradiated with up to 1017 5 MeV He ions/cm2. However micro-Raman scattering showed considerable spectral broadening characteristic of radiation damage near the end of the ∼22 μm He range. Partial amorphism plus nepheline formation was observed in grazing incidence X-ray diffraction when sodalite was irradiated by 1016 12 MeV Au ions/cm2. Nepheline appeared less susceptible to 12 MeV Au ion damage than sodalite, with ∼25% less amorphous fraction at 1016 ions/cm2. © 2014, Elsevier B.V.
Hot isostatic pressed pyrochlore glass‐ceramics: revealing structure insides at the reaction interface
(Wiley, 2021-06-15) Zhang, YJ; Wei, T; Xu, A; Dayal, P; Gregg, DJ
As potential waste forms for immobilizing actinide‐rich radioactive wastes, Eu2Ti2O7 (Eu as a surrogate for minor actinides) pyrochlore glass‐ceramics were fabricated via hot isostatic pressing (HIPing) at 1200°C. The structure and microstructure at the reaction interface between the glass‐ceramic waste form and the stainless steel (SS) canister under HIPing conditions were carefully investigated with scanning electron microscopy (SEM), transmission electron microscopy (TEM), and synchrotron single crystal X‐ray diffraction (SC‐XRD). The interactions at the reaction interface led to the formations of a ~10‐µm‐thick Cr2O3 layer as the oxidation front of the SS and a layer of a mixed oxide phase (Eu1.25SiCr0.8Ti1.2O7.5) on the glass‐ceramic side of the reaction interface. The crystal structure of such a unique mixed oxide phase was revealed indubitably with a combination of synchrotron SC‐XRD and TEM assisted with a focused ion beam (FIB) SEM system. The improved structural understanding of the reaction interface will help to support the utilization of HIPing as a versatile hot consolidation process for the treatment of radioactive wastes. © 2024 The American Ceramic Society/
Hot isostatic pressing of ceramics, glass and glass-ceramics for immobilisation of intermediate-and high-level nuclear waste
(Australian Institute of Physics, 2018-01-31) Vance, ER; Gregg, DJ; Chavara, DT
Different classes of nuclear wastes are briefly described, together with an idea of the required qualities of the processed waste for geological disposal. The main Synroc project at ANSTO these days is immobilisation of the upcoming ~5000L/yr intermediate-level waste from the production of the 99Mo radiopharmaceutical and this has been demonstrated inactively by hot isostatic pressing (HIP) a glass at full scale, together with showing its resilience to compositional variations of 10% for all main components, minor impurities and HIP temperature and pressure. Other work deals with HIP immobilisation of zirconolite-based glass-ceramics for surplus PuO2 immobilisation, glass-ceramics and ceramics for U-rich legacy ANSTO 99Mo intermediate-level and low-level waste, immobilisation of pyroprocessing fluoride-salt waste, spent power plant fuel, refractory glasses and CuI.
Hot isostatically pressed (HIPed) fluorite glass‐ceramic wasteforms for fluoride molten salt wastes
(John Wiley & Sons, Inc., 2020-06-07) Gregg, DJ; Vance, ER; Dayal, P; Farzana, R; Aly, Z; Holmes, R; Triani, G
Molten pyroprocessing salts can be used to dissolve used nuclear fuel from a reactor allowing recovery of the actinides. Previously, ANSTO have demonstrated hot isostatically pressed (HIPed) sodalite glass‐ceramic wasteforms for eutectic (Li,K)Cl salts containing fission products, but this system cannot be used for the analogous molten alkali fluoride salts (eg, FLiNaK), which have utility in the application of the next generation of nuclear reactors. In this work, a novel glass‐ceramic composite wasteform has been prepared by HIPing, as a candidate for the immobilization of fission product‐bearing FLiNaK salts. The wasteform has been tailored to immobilize the high fluoride content of the waste within fluorite, whereas the waste alkali elements are incorporated in a durable sodium aluminoborosilicate glass, with total waste loadings of ~17‐21 wt% achieved. It was also demonstrated that the speciation of Mo‐ and Sb‐simulated fission products was altered by adding Ti metal due to a controlled redox environment. The resulting candidate wasteform has been studied by X‐ray diffraction and scanning electron microscopy, including the HIP canister‐wasteform interaction zone, and its performance assessed via leaching studies using the PCT and ASTM C1220 leaching protocols. Dr Vance very much enjoyed the challenge of wasteform design for problematic nuclear wastes, for which fission product‐bearing FLiNaK salts are a clear example. His ability to hone in on a wasteform solution with viable waste loadings that meet performance requirements was testament to his nearly 40 years experience in nuclear waste immobilization. The samples discussed in this work represent the last wasteform materials that he prepared. © 1999-2020 John Wiley & Sons, Inc.
Immobilization of iodine via copper iodide
(Elsevier, 2018-07) Vance, ER; Grant, C; Karatchevtseva, I; Aly, Z; Stopic, A; Harrison, JJ; Thorogood, GJ; Wong, HKY; Gregg, DJ
CuI is a candidate wasteform for the immobilization of the fission product 129I. CuI can be made simply by the addition of CuCl to an I− bearing solution such that exchange of Cl− with I− takes place. The CuI material can then be consolidated into a wasteform by sintering at approximately 550 °C in argon or by hot isostatically pressing at 550 °C with 100 MPa of pressure. A waste loading of greater than 60 wt.% is achievable with good water leach resistance, in keeping with the low solubility product of CuI. However, like the well known wasteform candidate AgI, CuI decomposes in water containing metallic Fe. To compensate this deficiency, the sintered CuI wasteform can be further protected by surrounding it by Sn powder and HIPing at the low temperature of 200 °C. © 2018 Elsevier B.V
In-situ neutron characterization of advanced nuclear ruels - the road to a new neutron irradiation testing capability
(The Minerals, Metals & Materials Society, 2020-02-23) Obbard, EG; Gasparrini, C; Burr, PA; Johnson, KD; Lopes, DA; Anghel, C; Middleburgh, SC; Gregg, DJ; Liss, KD; Griffiths, GJ; Scales, N; Thorogood, GJ; Lumprin, GR
Not available.
Incorporation of Ba in Al and Fe pollucite
(Elsevier B.V., 2016-09-01) Vance, ER; Gregg, DJ; Griffiths, GJ; Gaugliardo, PR; Grant, C
Ba, the transmutation product of radioactive Cs, can be incorporated at levels of up to ∼0.07 formula units in Cs(1−2x)BaxAlSi2O6 aluminium pollucite formed by sol-gel methods and sintering at 1400 °C, with more Ba forming BaAl2Si2O8 phases. The effect of Ba substitution in pollucite-structured CsFeSi2O6 was also studied and no evidence of Ba substitution in the pollucite structure via cation vacancies or Fe2+ formation was obtained. The Ba entered a Fe-silicate glass structure. Charge compensation was also attempted with a Cs+ + Fe3+ ↔ Ba2+ + Ni2+ scheme but again the Ba formed a glass and NiO was evident. PCT leaching data showed CsFeSi2O6 to be very leach resistant. © 2016 Elsevier B.V.
The incorporation of Li2SO4 into barium borosilicate glass for nuclear waste immobilisation
(Elsevier, 2022-03-15) Farzana, R; Dayal, P; Karatchevtseva, I; Aly, Z; Gregg, DJ
This study has systematically investigated the effect of Li2SO4 addition (2.75 −16.5 wt%) in barium borosilicate glass, to provide a pathway to optimise the glass composition and maximise sulphate incorporation. The work also provides a mechanistic understanding as to how SO42- is incorporated within the glass structure. The highest sulphate incorporation of 2.78 wt% SO3 (from 11 wt% Li2SO4 addition) was achieved without crystallisation following melting at 1200 °C. Sulphate incorporation in glass was confirmed by XRF, ICP, EDS and Raman analysis. Addition of Li2SO4 along with sodium and barium oxides improved the sulphate incorporation by mixed alkali network depolymerisation and the larger Ba cations helped to create sufficient space within the boron-silicate network to incorporate sulphate ions into the glass. An immiscible sulphate layer rich in BaSO4 and Na2SO4 formed on top of the glass at lower temperature (800–1100 °C) and subsequent diffusion of Na, Ba oxides and sulphur from this layer increased with increasing time and temperature to form a sulphate incorporated amorphous glass. Addition of Na2O played an important role to improve sulphate incorporation in the glass, as well as formation of an immiscible layer on top of the glass however, the formation of Na2SO4 lowered the sulphur incorporation rate at high temperature compared to BaSO4. Increasing the Li2SO4 content in the glass decreased the glass transition temperature. Aqueous durability testing using the standard PCT tests indicated the glass had satisfactory aqueous durability compared to benchmark environmental assessment glass. This study provides opportunities to convert Li+ and SO42- rich nuclear wastes into appropriate glass wasteforms. © 2021 Elsevier B.V.
The incorporation of neptunium and plutonium in thorutite (ThTi2O6)
(Elsevier, 2013-12-25) Zhang, YJ; Gregg, DJ; Lumpkin, GR; Begg, BD; Jovanovic, M
The incorporation of neptunium (Np) and plutonium (Pu) into the brannerite structured lattice was studied using thorutite (ThTi2O6) as host lattice and sintering in air. The uncompensated Np and Pu doped samples and the low Y-charge compensated Np and Pu doped samples showed main phases as designed together with trace amounts of rutile. Those samples with larger amounts of Y produced yttrium pyrochlores as an additional minor phase. XRD analyses reveal anisotropic changes of the cell parameters; the a-parameter contracts while b- and c-parameters expand with mean cationic radius. This is in reasonable agreement with previous experimental data on ThTi2O6 and Ce0.975Ti2O5.95. Attempts to form Np or Pu valences >4+ by adding Y as a charge compensator were unsuccessful, suggesting that tetravalent Np and Pu ions are favoured in air-fired thorutite. © 2013, Elsevier B.V.

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