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Browsing by Author "Dayal, P"

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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.
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.
Effect of double ion implantation and irradiation by Ar and He ions on nano-indentation hardness of metallic alloys
(Elsevier, 2013-07-01) Dayal, P; Bhattacharyya, D; Mook, WM; Fu, EG; Wang, YQ; Carr, DG; Anderogluc, O; Mara, NA; Misra, A; Harrison, RP; Edwards, L
In this study, the authors have investigated the combined effect of a double layer of implantation on four different metallic alloys, ODS steel MA957, Zircaloy-4, Ti–6Al–4V titanium alloy and stainless steel 316, by ions of two different species – He and Ar – on the hardening of the surface as measured by nano-indentation. The data was collected for a large number of indentations using the Continuous Stiffness Method or “CSM” mode, applying the indents on the implanted surface. Careful analysis of the data in the present investigations show that the relative hardening due to individual implantation layers can be used to obtain an estimate of the relative hardening effect of a combination of two separate implanted layers of two different species. This combined hardness was found to lie between the square root of the sum of the squares of individual hardening effects, (ΔHA2 + ΔHB2)0.5 as the lower limit and the sum of the individual hardening effects, (ΔHA + ΔHB) as the upper limit, within errors, for all depths measured.© 2013, Elsevier B.V.
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.
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.
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 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.
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.
Ion beam irradiation effects in strontium zirconium phosphate with NZP-structure type
(Elsevier Science BV, 2014-03-01) Gregg, DJ; Karatchevtseva, I; Thorogood, GJ; Davis, J; Bell, BDC; Jackson, M; Dayal, P; Ionescu, M; Triani, G; Short, KT; Lumpkin, GR; Vance, ER
Ceramics with the sodium zirconium phosphate or NZP type structure have potential as nuclear waste form and inert matrix materials. For both applications the material will be subjected to self-radiation damage from alpha-decay of the incorporated actinides. In this study, ion-beam irradiation using Au- and He-ions has been used to simulate the consequences of a-decay and the effects of irradiation on the structural and macroscopic properties (density and hardness) have been investigated. Irradiation by Au-ions resulted in a significant volume contraction of similar to 7%, a reduction in hardness of similar to 30% and a loss in long-range order at fluences above 10(14) Au-ions/cm(2). In contrast, little effect on the material properties was noted for samples irradiated with He-ions up to a fluence of 10(17) ions/cm(2). Thermal annealing was investigated for the highest fluence Au-ion irradiated sample and significant decomposition was observed. © 2014, Elsevier Ltd.
Mechanistic impacts of long-term gamma irradiation on physicochemical, structural, and mechanical stabilities of radiation-responsive geopolymer pastes
(Elsevier, 2021-04-05) Yeoh, MLY; Ukritnukun, S; Rawal, A; Davies, JB; Kang, BJ; Burrough, K; Aly, Z; Dayal, P; Vance, ER; Gregg, DJ; Koshy, P; Sorrell, CC
The mechanistic effects of long-term γ irradiation on the mineralogical, microstructural, structural, physical, and chemical properties of 40 wt% blast furnace slag + 60 wt% fly ash geopolymer pastes have been examined. Ambient curing for 28 days during normal equilibration was followed by exposure to 60Co irradiation (1574, 4822, 10,214 kGy). The material characteristics are controlled largely through the competing mechanisms of beneficial equilibration at initial lower dosages, which enhances gelation and crosslinking, and detrimental equilibration at subsequent higher dosages, which causes structural and microstructural destabilisation. Irradiation for 2 months (1574 kGy) increases the compressive strength ~45% (~57 to ~83 MPa) through conversion of less-crosslinked (Q0/Q1/Q1′) to more-crosslinked (Q2/Q3/Q4) silicate species. The transition between these regimes occurs after ~5 months of irradiation (~4000 kGy). Beyond this, the rates of beneficial equilibration and detrimental equilibration equalise upon completion of normal geopolymerisation. Additional geopolymerisation from γ irradiation is controlled by the rate-limiting release of Si4+ from the unreacted aluminosilicates and silicates and their rapid incorporation in the geopolymer network. The aqueous leaching of the geopolymer pastes is not affected significantly by γ irradiation. These data reveal the potential for these materials as intermediate-level wasteforms that can outperform Portland cement-based materials. © 2020 Elsevier B.V.
Profiling hot isostatically pressed canister–wasteform interaction for Pu-bearing zirconolite-rich wasteforms
(John Wiley & Sons, Inc, 2022-04-02) Dayal, P; Farzana, R; Zhang, YJ; Lumpkin, GR; Holmes, R; Triani, G; Gregg, DJ
Zirconolite-rich full ceramic wasteforms designed to immobilize Pu-bearing wastes were produced via hot isostatic pressing (HIP) using stainless steel (SS) and nickel (Ni) HIP canisters. A detailed profiling of the elemental compositions of the major and minor phases over the canister–wasteform interaction zone was performed using scanning electron microscopy combined with energy-dispersive X-ray spectroscopy (SEM-EDS) characterization. Bulk sample analyses from regions near the center of the HIP canister were also conducted for both samples using X-ray diffraction and SEM-EDS. The sample with the Ni HIP canister showed almost no interaction zone with only minor diffusion of Ni from the inner wall of the canister into the near-surface region of the wasteform. The sample with the SS HIP canister showed ∼100–120 μm of interaction zone dominated by high-temperature Cr diffusion from canister materials to the wasteform with the Cr predominantly incorporated into the durable zirconolite phase. We also examined, for the first time, changes to the HIP canister wall thickness caused by HIPing and demonstrated that no canister wall thinning occurred. Instead, in the areas examined, the canister wall thickness was observed to increase (up to ∼20%) due to the compression occurring during the HIP cycle. Further, only sparse formation of (Cr, Mn)-rich oxide particles were noted within the HIP canister inner wall area immediately adjacent to the ceramic material, with no evidence for reverse diffusion of ceramic materials. Though the HIP canister–wasteform interaction extends to ∼120 μm when using an SS HIP canister for the system investigated, this translates to <<1 vol.% for an industrial scale HIPed wasteform. Importantly, the HIP canister–wasteform interactions did not produce any obviously less durable phases in the wasteform or had any detrimental impact on the HIP canister properties. © 2022 Commonwealth of Australia. Journal of the American Ceramic Society published by Wiley Periodicals LLC on behalf of American Ceramic Society.
Pyrochlore glass-ceramics for the immobilization of molybdenum-99 production wastes: demonstrating scalability and flexibility to waste stream variance
(Elsevier, 2021-11) Farzana, R; Zhang, YJ; Dayal, P; Aly, Z; Holmes, R; Triani, G; Vance, ER; Gregg, DJ
Pyrochlore glass ceramics have been fabricated via in-situ crystallization under reducing conditions by both sintering and hot isostatic pressing (HIPing) as candidate wasteforms for the acidic waste biproduct of Mo-99 radiopharmaceutical production. The tailored wasteform demonstrates flexibility in the wasteform design to receive the required waste variability, it also suitably passes high-level waste performance requirement, and successfully scales to 1 kg scale with 45 wt.% waste loading. U-rich pyrochlore as the major phase was confirmed by X-ray diffraction, scanning electron microscopy and energy dispersive X-ray spectroscopy, with residual glass and minor secondary phases. The presence of both U4+ and U5+ valences in the wasteforms was revealed by diffuse reflectance spectroscopy. Addition of glass content had little influence on the pyrochlore composition but facilitated minor perovskite formation. The up-scaled dense, HIPed sample showed elemental releases of < 2 g/L for all elements in durability experiments. © 2021 Elsevier Ltd
Radiation effects on microstructure and hardness of a titanium aluminide alloy irradiated by helium ions at room and elevated temperatures
(Elsevier B.V., 2015-04) Wei, T; Zhu, HL; Ionescu, M; Dayal, P; Davis, J; Carr, DG; Harrison, RP; Edwards, L
A 45XD TiAl alloy possessing a lamellar microstructure was irradiated using 5MeV helium ions to a fluence of 5×1021ionm−2 (5000appm) with a dose of about 1dpa (displacements per atom). A uniform helium ion stopping damage region about 17μm deep from the target surface was achieved by applying an energy degrading wheel. Radiation damage defects including helium-vacancy clusters and small helium bubbles were found in the microstructure of the samples irradiated at room temperature. With increasing irradiation temperature to 300°C and 500°C helium bubbles were clearly observed in both the α2 and γ phases of the irradiated microstructure. By means of nanoindentation significant irradiation hardening was measured. For the samples irradiated at room temperature the hardness increased from 5.6GPa to 8.5GPa and the irradiation-hardening effect reduced to approximately 8.0GPa for the samples irradiated at 300°C and 500°C. © 2015 Elsevier B.V.
Residual stresses distribution measured by neutron diffraction in fabricated square high strength steel tubes
(Trans Tech Publications, 2014-02-01) Mashiri, FR; Paradowska, AM; Uy, B; Tao, Z; Khan, M; Dayal, P
Engineers are increasingly encouraged to consider sustainability in the design and construction of new civil engineering infrastructure. Sustainability can be achieved through the use of high strength materials thereby reducing quantity of materials required in construction where possible. Knowledge of residual stresses in fabricated columns is important in identifying whether the fabricated columns can be classified as heavily welded (HW) or lightly welded (LW). The determination of residual stresses can be used to determine the local buckling of stub columns. Residual stress magnitudes are also essential in the numerical modelling of buckling behaviour of columns. This paper outlines the challenges in measurement of residual stresses using neutron diffraction in fabricated high strength steel square tubes. The residual stress line scans and maps were measured using the Kowari Strain Scanner located at the Australian Nuclear and Science Organisation (ANSTO) in Australia. © 2014, Trans Tech Publications.
Sodium zirconium phosphate‐based glass‐ceramics as potential wasteforms for the immobilization of nuclear wastes
(Wiley, 2021-10-25) Scales, N; Dayal, P; Aughterson, RD; Zhang, YJ; Gregg, DJ
A comprehensive study on the development of sodium zirconium phosphate (NZP)‐based glass‐ceramic composites as potential wasteforms for the immobilization of nuclear wastes is reported. Two complementary waste treatment routes, the ex situ and in situ crystallisation of NZP with a sodium aluminoborosilicate glass, were investigated with various processing conditions including sintering temperature, cooling rate and NZP to glass ratios. While the ex situ route with mixing of pre‐made NZP and glass is a robust and reliable means of producing the glass‐ceramic composites, the in situ crystallisation of NZP from an amorphous NZP precursor is a more realistic processing route. The formation of ZrO2 as a minor phase was observed especially for high NZP to glass ratios due to the solubility difference between Zr and P oxides in glass. The addition of extra phosphate can overcome this and yield glass‐ceramic composites with appropriate NZP stoichiometry. Overall, the NZP glass‐ceramic system is versatile offering multiple processing options for nuclear waste management. © 2024 The American Ceramic Society.
Structural and phase evolution in U3Si2 during steam corrosion
(Elsevier, 2022-08-01) Liu, J; Burr, PA; White, JT; Peterson, VK; Dayal, P; Baldwin, C; Wakeham, D; Gregg, DJ; Sooby, ES; Obbard, EG
U3Si2 nuclear fuel is corroded in deuterated steam with in situ neutron diffraction. Density functional theory is coupled with rigorous thermodynamic description of the hydride including gas/solid entropy contributions. H absorbs in the 2b interstitial site of U3Si2Hx and moves to 8j for x ≥ 0.5. Hydriding forces lattice expansion and change in a/c ratio linked to site preference. Rietveld refinement tracks the corrosion reactions at 350–500 °C and preference for the 8j site. Above 375 °C, formation of UO2, U3Si5 and USi3 take place in the grain boundaries and bulk. Hydriding occurs in bulk and precedes other reactions. © 2022 Published by Elsevier Ltd.
Synroc technology: perspectives and current status (review)
(John Wiley & Sons, Inc., 2020-06-22) Gregg, DJ; Farzana, R; Dayal, P; Holmes, R; Triani, G
Dr Eric (Lou) Vance spent 32 years at the Australian Nuclear Science and Technology Organisation (ANSTO), where he was dedicated to the development of Synroc technology, a waste treatment solution for intractable nuclear wastes. The original form of Synroc, a multiphase ceramic wasteform based on stable and leach resistant titanate minerals, was invented by Australian scientists in the late 1970s. This formulation was directed toward the immobilization of PUREX wastes from the reprocessing of nuclear fuels. Synroc at ANSTO under the scientific leadership of Dr Vance since evolved beyond these original titanate ceramics into a waste treatment technology platform. This platform can be applied to produce glass, glass‐ceramic and ceramic wasteforms and offers distinct advantages in terms of waste loading and suppressing volatile losses. The platform therefore provides an opportunity to treat those waste streams that are problematic for glass matrices alone or existing vitrification process technologies. Such wastes include, for example, actinide‐bearing wastes, those that contain large proportions of refractory elements, those with significant fission product or corrosive volatile emissions and those wastes resulting from radiopharmaceutical production. The implementation of the latter will see the industrialization of Synroc technology via a first‐of‐a‐kind Synroc Waste Treatment Facility that is currently under construction at ANSTO. This paper will review Synroc technology, particularly noting the substantial and essential contributions from the late Dr Vance. The review will also provide some perspective on the development of the technology for nuclear waste immobilization and describe the significant recent advancements at ANSTO. © 1999-2020 John Wiley & Sons, Inc.

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