bluuebunny/crossref_metadata_2025_split huggingface dataset:
Gadolinium Borosilicate Glass-Bonded Gd-Silicate Apatite: A Glass-Ceramic Nuclear Waste Form for Actinides
Authors: Zhao Donggao , Li Liyu , Davis L.L. , Weber W.J. , Ewing R.C. 
Published: May 2011

ABSTRACTA Gd-rich crystalline phase precipitated in a sodium gadolinium alumino-borosilicate glass during synthesis. The glass has a chemical composition of 45.4-31.1 wt% Gd2O3, 28.8-34.0 wt% SiO2, 10.8-14.0 wt% Na2O, 4.3-5.9 wt% Al2O3, and 10.8-14.9 wt% B2O3. Backscattered electron images revealed that the crystals are hexagonal, elongated, acicular, prismatic, skeletal or dendritic, tens of μm in size, some reaching 200 μm in length. Electron microprobe analysis confirmed that the crystals are chemically homogeneous and have a formula of NaGd9(SiO4)6O2 with minor B substitution for Si. The X-ray diffraction pattern of this phase is similar to that of lithium gadolinium silicate apatite. Thus, this hexagonal phase is a rare earth silicate with the apatite structure. We suggest that this Gd-silicate apatite in a Gd-borosilicate glass is a potential glass-ceramic nuclear waste form for actinide disposition. Am, Cm and other actinides can easily occupy the Gd-sites. The potential advantages of this glass-ceramic waste form include: 1) both the glass and apatite can be used to immobilize actinides, 2) silicate apatite is thermodynamically more stable than the glass, 3) borosilicate glass-bonded Gd-silicate apatite is easily fabricated, and 4) the Gd is an effective neutron absorber.
Apparent Solubility Limit of Nuclear Glass
Authors: Vernaz Étienne , Gin Stéphane 
Published: May 2011

ABSTRACTMost nuclear glass alteration models are based on a first-order kinetic law of the general type r = r0 (1–Csi/C*) or r = r0 (1–Q/K). It is generally assumed that the establishment of quasi steady-state concentrations in solution corresponds to an intrinsic glass solubility limit that may be expressed either in terms of the dissolved silicon concentration C* or as an ion activity product at saturation K. Experimental research over the last five years in France has shown that the quasi steady-state concentrations observed in solution do not correspond to a thermodynamic solubility limit with respect to the glass. This assertion is supported by a large body of experimental evidence that is discussed in this paper. The signification of the C* parameter is therefore reconsidered, and its implications on long-term behavior modeling are discussed.
Structural Effect of Pu Substitutions on the Zr-Site in Zirconolite
Authors: Begg B.D. , Day R.A. , Brownscombe A. 
Published: May 2011

ABSTRACTAs the level of Pu4+ substituted on the Zr-site in CaZr1-xPuxTi2O7 zirconolite increased, from x=0.1 to 0.6, a series of structural transitions occurred from zirconolite-2M to zirconolite-4M and subsequently from zirconolite-4M to pyrochlore. The solid-solution limit for Pu4+ substituted on the Zr-site in zirconolite-2M was ~ 0.15 formula units. Zirconolite-4M was only stable over a narrow compositional range, centered about CaZr0.59Pu0.41Ti2O7, whilst the pyrochlore structure was stabilized with CaZr0.4Pu0.6Ti2O7 stoichiometry. The stability of the zirconolite polytypes is therefore sensitive to the average effective ionic size of the ions occupying the seven-coordinated Zr-site. The reduction in Pu from Pu4+ to Pu3+ destabilized the zirconolite-4M, producing a mixture of perovskite and possibly zirconolite-3T. The CaZr0.4Pu0.6Ti2O7 pyrochlore was also predominantly transformed to perovskite as a result of this reduction of Pu.
Durability of Zirconolite in Hydrothermal Fluids: Implications for Nuclear Waste Disposal
Authors: Gieré R. , Malmström J. , Reusser E. , Lumpkin G.R. , Düggelin M. , Mathys D. , Guggenheim R. , Günther D. 
Published: May 2011

ABSTRACTSynthetic zirconolite (CaZrTi2O7) doped with rare earth elements (REEs) and Hf has been subjected to corrosion tests in a closed system at elevated temperature and pressure in fluids with various compositions. Together with previous studies, the results indicate only a very weak corrosion below 250°C at a pressure of 50 MPa. Above that temperature and up to 500°C zirconolite suffers from relatively rapid corrosion and, depending on the fluid composition, it may be covered by various secondary phases. Above 500°C in Na-rich fluids, zirconolite is replaced by perovskiteand calzirtite (nominally CaTiO3 and Ca2r5Ti2O16, respectively), but theREEs and Hf (acting as actinide analogues and/or neutron absorbers) are almost quantitatively incorporated into the secondary phases. The breakdown of zirconolite and its replacement by other phases in the laboratory tests are comparable to reactions observed in natural systems. Additional experiments with U-doped zirconolite revealed differences in the behavior of the used actinide analogues (Nd, Ce, Gd)and U during corrosion. The results of this study, together with observations on natural samples, strongly support the use of zirconolite-based ceramic waste forms for actinide-rich wastes.
Effects of Cesium, Iodine and Strontium Ion Implantation on the Microstructure of Cubic Zirconia
Authors: Wang L.M. , Zhu S. , Wang S.X. , Ewing R.C. 
Published: May 2011

ABSTRACTCesium, iodine and strontium ions have been implanted into yttria-stabilized cubic zirconia (YSZ) to determine the effects of fission product incorporation in YSZ that is considered as an inert nuclear fuel matrix. The ion implantation was conducted at room temperature to 1 × 1021ions/m2 for each ion with ion energies ranging from 70 to 400 keV. The peak displacement damage level and the peak ion concentration in YSZ reached 205-330 dpa and 11-26 at%, respectively. The microstructure of the implanted YSZ was studied by in situ and cross-sectional transmission electron microscopy. In the iodine and strontium implanted samples, a damaged layer with a high density of defect clusters was observed, while in the cesium implanted specimen, the damaged layer is amorphous. Nanocrystalline precipitates were observed in the strontium implanted specimen after annealing at 1000°C. The results are discussedin terms of the ionic size, mobility and the solubility of the implanted species in YSZ.
The Influence of Chemical Composition of Liquid Radioactive Waste on the Evaporation Process
Authors: Poluektov P.P. , Soukhanov L.P. , Zhicharev M.I. 
Published: May 2011

ABSTRACTA method is suggested to assess the tolerable salt content of the evaporator bottoms from the data on solubility in salt systems taken as simplified models of liquid radioactive waste (LRW) arising from nuclear power plants (NPP) with boiling reactors. It has been demonstrated that the degree of evaporation may be substantially increased by implementing the process in nitric acid. Equations have been derived that allow the calculation of the minimum needed acidity of the solution to allow maximum evaporation.
Synthesis and Study of 239Pu-Doped Ceramics Based on Zircon, (Zr,Pu)Sio4, and Hafnon, (Hf,Pu)SiO4
Authors: Burakov B.E. , Anderson E.B. , Zamoryanskay M.V. , Yagovkina M.A. , Strykanova E.E. , Nikolaeva E.V. 
Published: May 2011

ABSTRACTZircon, ZrSiO4, as well as its Hf-analogue hafnon, HfSiO4, have been proposed for use as durable Pu host phases for the immobilization of weapons grade Pu and other actinides. Four samples of Pu-doped ceramics based on the zircon and hafnon structures were synthesized through sintering in air using precursors containing 5-6 and 10 wt% 239Pu. Synthesized ceramic samples were studied by X-ray powder diffraction (XRD), scanning electron microscopy (SEM), microprobe method, MCC-1 leach test at 25 and 90°C. Inclusions of separated a PuO2phase in the matrix of zircon-based ceramic and presumably, (Pu,Hf)O2 phase in the hafnon-based ceramic were observed for samples obtained from precursors doped with 10 wt% Pu. No separated Pu-phases in significant amounts were identified in the matrices of both ceramics obtained from the precursors doped with 5-6 wt% Pu. It was found that normalized Pu mass losses (without correction on ceramic porosity) for samples doped with 10 wt% Pu which contain separated inclusions of PuO2 or (Pu,Hf)O2 after 14/28 days were approximately (in g/m2) - for zircon: 0.2/0.2 - at 90°C and 0.03/0.04 - at 25°C and for hafnon: 0.02/0.04 - at 90°C and 0.01/0.01 - at 25°C. The losses of Pu from samples doped with 5-6 wt% are 1-2 order of magnitude less. It was suggested that optimal amount of Pu which could be incorporated by zircon and hafnon lattices does not exceed 7 wt%. An important additional conclusion is that Pu- doped ceramic based on zircon or hafnon can be successfully fabricated excluding hot pressing method.
Durabilities of Pyrochlore-Rich Titanate Ceramics Designed for Immobilization of Surplus Plutonium
Authors: Zhang Y. , Hart K.P. , Blackford M.G. , Thomas B.S. , Aly Z. , Lumpkin G.R. , Stewart M.W. , Mcglinn P.J. , Brownscombe A. 
Published: May 2011

ABSTRACTThe 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 TiO2 (brookite and probably anatase) have been identified on the surface of the leached sample and the principal secondary phase is an unknown polymorph of TiO2 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% PuO2 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.
Phase and Chemical Stability of Murataite Containing Uranium, Plutonium and Rare Earths
Authors: Yudintsev S.V. , Stefanovsky S.V. , Nikonov B.S. , Omelianenko B.I. 
Published: May 2011

ABSTRACTMurataite –a complex actinide (An)- and rare earth element (REE)-bearing oxide with a cubic fluorite-related lattice – is one of the promising host-phases for immobilization of Pu-containing waste. The murataite phase composition corresponds to the empirical formula:A4B2C7O22-x, where the A-sites are occupied by Ca, Mn, REE, and An (U); B sites – by Mn, Ti,Zr, and An (U); and C sites – by Ti, Al, and Fe. The total amount of the actinides (U) and REE (Ce, Gd) in the murataite may exceed 20 wt%. In contrast to the other prospective hosts for actinide waste immobilization (cubic zirconia and pyrochlore), murataite accommodates higher amounts of corrosion products(Al, Fe) along with the actinides. The authors compared murataite-based ceramics having similar compositions and produced by melting in a high-temperature resistance furnace or via inductive melting in a cold crucible. Eight samples of the murataite-based ceramics were produced and investigated in detail. Murataite was found to be the major phase in four of the samples – with a basic composition, in wt%, of: 5.0 Al2O3, 10.0 CaO, 55.0 TiO2, 10.0 MnO, 5.0 Fe2O3, 5.0 ZrO2, and 10.0 UO2. These samples were produced by melting in a resistive furnace and in the cold crucible and included Gd-bearing samples and one Pu-bearing sample. The extra phases were other titanates: (from more to less typical) rutile, pyrochlore, zirconolite, crichtonite, pseudobrookite, and perovskite (in the Pu-doped samples only). Three varieties of the murataite, with 3-, 5-, and 8-fold fluorite-type lattices, were observed. Addition of uranium and rare earth oxides stabilizes pyrochlore as the major phase, whereas addition of zirconia yields zirconolite. Plutonium stabilizes the perovskite-type phase, probably due to the formation of Pu3+. The maximum waste oxide content in the murataite for the elements studied was found to be 10% ZrO2, 12% CeO2, 13% Gd2O3, and 14% UO2. Waste element partitioning among the murataite and all the other phases with similar fluorite-related structure (pyrochlore and zirconolite) was analyzed. The uranium leach rate for the sample with maximum murataite content was measured using a procedure similar to MCC-3. This leach rate was close to10-5 g/(m2*day) in a½-day test and decreased by more than one order of magnitude in 28-day tests. Investigation of the stability of the murataite structure after irradiation byheavy ions is in progress.
The Effect of Hydrogen Peroxide Concentration on the Oxidative Dissolution of Unirradiated Uranium Dioxide
Authors: Pablo J. De , Casas I. , Clarens F. , Aamrani F. El , Rovira M. 
Published: May 2011

ABSTRACTThe dissolution rate of unirradiated uranium dioxide was studied in batch experiments as a function of hydrogen peroxide concentration (from 10−5 to 10−3 mol dm−3). Unirradiated UO2(s) was used in order to differentiate surface chemical processes from radiolytic effects. Dissolution rates were determined from both uranium release and hydrogen peroxide consumption. Results showed that H2O2consumption rate was higher than UO2 dissolution rate. This observation may indicate that the overall UO2 oxidative dissolution process would be controlled by the dissolution of the oxidized solid surface. The calculated hydrogen peroxide reaction order was 1 in the H2O2 concentration range from 10−5 to 10−4 mol dm−3, while at higher concentrations no clear dependence was observed.
Congruent and Incongruent Radionuclide Release During Matrix Dissolution of Partly Oxidized High Burnup Spent Fuel
Authors: Loida A. , Grambow B. , Geckeis H. 
Published: May 2011

ABSTRACTWith respect to the assessment of the long-term behavior of the waste form spent fuel it is of high importance to study in particular the UO2 matrix dissolution behavior and the associated release/retention of radionuclides in contact with groundwater. During long term fuel storage, fuel oxidation may not be avoided. Main issue of this work is to identify the impact on the corrosion of partly oxidized fuel of environmental conditions such as (1) the nature of solution contacting the matrix, the (2) presence/absence of CO2, (3) fixed pH values within a range between pH 7- pH 11, and (4) the presence/absence of corroding container material (Fe-powder). Dissolution tests with powdered oxidized spent fuel in various granite waters, and NaCl-brine resulted in matrix dissolution rates in the same order of magnitude for all investigated media (ca.5×10−4/d). The presence of CO2 and fixed pH values (pH 5 – 11) was without a distinct effect. The independence of the dissolution rate of the oxidized fuel matrix upon the nature of solution, pCO2, fixed pH values (5-11) can probably be explained by a masking effect of radiolysis. In presence of Fe powder the matrix dissolution rate was found to be slowed down by a factor of ca. 20, associated with strong retention effects of radionuclides.
On Metal Aggregates in Spent Fuel, Synthesis and Leaching of Mo-Ru-Pd-Rh Alloy
Authors: Cui Daqing , Eriksen Trygve , Eklund Ulla-Britt 
Published: May 2011

ABSTRACTThe results of leaching tests on the synthesised 46%Mo-29%Ru-21%Pd-4%Rh alloy show selective release Mo at both reducing and air saturated conditions. The leach rate increases with rising redox potential (Eh) of leaching solution and temperature, and slightly decreases with leaching time. The electrochemical study indicates that the alloy is more stable than carbon steel against corrosion but less stable than copper and stainless steel. Based on the leach rates of Mo, Ru, Rh, Pd and oxidation potentials of these four metals and Tc0, the leach rates of Tc from alloy aggregates, 40%Mo, 30%Ru, 10%Tc, 15% Pd, 5%Rh in spent fuel at reducing and air saturated conditions are estimated, to be 4.10−9 g.cm−2.day−1 and 4. 10−7 g.cm−2.day−1, respectively.
Release of Radiotoxic Elements from High Burn-Up UO2 and MOX Fuel in a Repository
Authors: Glatz Jean-Paul , Carbol Paul , Cobos-Sabaté Joaquin , Gouder Thomas , Miserque Frédéric , Gimenez Javier , Wegen Detlef 
Published: May 2011

ABSTRACTIn a spent fuel repository the processes that govern the release of radionuclides are dissolution and transport in a possible groundwater flow. The cladding will be the last barrier before the water comes into contact with the fuel, namely with the outer rim of the pellet. Here the heterogeneity of the material due to the irradiation process is responsible for a complex release process. Fission products and minor actinides inventories are considerably higher at the pellet periphery as a result of increased epithermal neutron capture and of migration in the case of the volatile fission products.The present paper gives a review of experimental activities at the Institute for Transuranium Elements (ITU). Both single effects studies and integral tests are carried out to study the behavior of spent fuel under storage conditions.Leaching of irradiated UO2 (up to 50 GWd/tU) and MOX (up to 25 GWd/tU) fuel rods with preset cladding defects at 100°C under anoxic or reducing conditions should simulate the realistic case of groundwater coming into contact with a spent nuclear fuel repository. For all main radionuclides the release process can be described considering a two-step dissolution mechanism that includes the initial dissolution of an oxidized layer present on the fuel surface followed by a long-term oxidative matrix dissolution. By means of α-doped (238Pu) UO2 it could be demonstrated, that radiolysis has a significant influence on this dissolution. Especially high initial release rates were found for the volatile cesium and iodine for the reasons mentioned above.Besides the conventional leaching experiments electrochemical techniques are used to investigate for instance the complex corrosion behavior of the heterogeneous MOX fuel materials or the influence of α-radiolysis on spent fuel dissolution.In the integral tests mentioned above with large S/V values, reprecipitation of U is likely to happen. Therefore special dynamic test are conducted where this reprecipitation is prohibited and true U solubility can be determined.Thin layer of UO2 and (U,Pu)O2 doped with various fission products and minor actinides are prepared to study the influence of these elements on the matrix dissolution. When Cs is for instance co-deposited, the U oxidation state changes from U4+ to U6+ for the same O2 pressure possibly indicating a stable Cs uranate. This could be an indirect proof of the existence of such a species in irradiated fuel (e.g. at the grain boundaries).
Iodine Immobilization by Sodalite Waste Form
Authors: Nakazawa T. , Kato H. , Okada K. , Ueta S. , Mihara M. 
Published: May 2011

ABSTRACTLeachabilities and solubilities of the synthesized iodide sodalite and natural (chloride) sodalite were measured by leach test. The iodide sodalite was synthesized in nitrogen gas flow at 800°C for 2 hours. The crystalline structure of the product was certified by XRD analysis. The natural sodalite containing chlorine was obtained at Bancroft, Canada. The several types of solution were used to evaluate the influence of the solubility of sodalite that included the varied pH and the chemical compositions. The solubilities of chloride sodalite were calculated by the thermodynamics data. The solubility measured for the synthesized iodide sodalite was compared with that calculated. The solubility of the synthesized iodide sodalite was approximately 2 x 10-4 mol/L, which shows a good agreement with the calculated one of the natural sodalite.
Mass Transport of Multicomponents Solute in Bentonite Clay
Authors: Ichikawa Y. , Kawamura K. , Nakano M. , Kitayama K. , Saito N. , Kawamura H. , Yamamoto S. 
Published: May 2011

ABSTRACTA fundamental concept combining a molecular behavior and macro-continuum phenomenon is presented for a multicomponent solute diffusion problem in bentonite clay, which is a key component of the Engineering Barrier System (EBS) of high-level radioactive waste disposal (HLRW). Bentonite is a micro-inhomogeneous material. Properties of the saturated bentonite are characterized by the constituent clay mineral (montmorillonite) and water, namely montmorillonite hydrate. We analyze its molecular behavior by applying a molecular dynamics (MD) simulation to inquire into the physicochemical properties such as diffusivity of chemical species. For extending the microscopic characteristics of constituent materials to a macroscopic diffusion behavior of the micro-inhomogeneous material we apply a homogenization analysis (HA).
Ash and Soil Conditioning Using Exothermic Metallic Compositions
Authors: Karlina O.K. , Varlackova G.A. , Ojovan M.I. , Tivansky V.M. , Klimov V.L. , Pavlova G.Yu. , Dmitriev S.A. 
Published: May 2011

ABSTRACTA thermochemical approach to conditioning ash residue that involves the use of exothermic metallic compositions (EMC) has been previously reported. EMC intermixed with ash residue at the appropriate ratios can sustain a glass forming reaction wave that produces monolith glass-like blocks. Herein, the thermochemical conditioning process is extended for conditioning contaminated clay soils. The results are reported for a study of the conditioning process and product materials to determine the optimal processing parameters and to characterize the product materials.
The National Radioactive Waste Repository Project
Authors: Perkins C. , Harris J. , Veitch S. , Kellett J. 
Published: May 2011

ABSTRACTIn 1992, the Commonwealth Government supported by States and Territories commenced an Australia-wide search for a site for a suitable site for a national radioactive waste repository for Australia's small inventory of low level and short-lived intermediate level waste. In 1998, the central-north region of South Australia was chosen as the preferred area in which to undertake detailed site investigations as it had the largest area of potential suitability compared to relevant technical and social criteria. Public consultation has been conducted both nationally and in the region.Site investigations are currently in their final phase, and it is expected that a preferred site will be identified before the end of 2000. Environmental assessment and licensing processes will then commence, and the repository may be constructed in 2002.
SR 97: Spent Fuel Alteration/Dissolution and the Influence of Near Field Hydrogen
Authors: Spahiu Kastriot , Sellin Patrik 
Published: May 2011

ABSTRACTA discussion of the evaluation of the source term in the SR 97 safety assessment of a deep repository for spent nuclear fuel is presented. Since the majority of the radionuclides are embedded in the uranium dioxide fuel matrix, they will be released only after the alteration/dissolution of the matrix. Therefore a description of the process of alteration/dissolution of the spent fuel matrix is needed in a safety assessment.Under normal repository conditions, i.e. reducing environment and neutral to alkaline pH, uranium dioxide has a very low solubility in water. If solubility is assumed to be the limiting factor, the dissolution of the fuel matrix will proceed very slowly due to the low water exchange in the defective canister. On this basis, a solubility-limited model for the release of the radionuclides from the fuel may be formulated.The reducing conditions can be upset by the radioactivity of the spent fuel, which generates oxidizing products through water radiolysis. This causes the oxidative alteration/dissolution of the UO2(s) matrix. A model for fuel matrix conversion resulting from radiolytic oxidative dissolution is discussed, as well as parameter variations and the associated uncertainties.In a repository the spent fuel will come in contact with groundwater after the copper canister has breached. Large amounts of hydrogen are then produced through the anoxic corrosion of the cast iron insert. Recent data on spent fuel leaching in presence of repository relevant hydrogen pressures and the implications on the actual and future spent fuel dissolution modeling will also be discussed.