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Actinide host phases as radioactive waste forms
Abstract
An effective strategy for dealing with high-level waste is to partition the short-lived fission product elements from the long-lived actinides, creating separate waste streams. Once there are two waste streams, the properties and durability of the waste form can be designed to a level appropriate to the toxicity and time required for isolation from the environment. With such a strategy the fission product elements may be incorporated into a borosilicate glass and the actinides into more durable crystalline ceramics. Although special glass compositions may be developed for actinide incorporation, their long-term durability is less easily assured, particularly on the time scales required for actinide immobilization and confinement. The final selection of any waste form should depend on its ability to incorporate the radionuclides of interest, its chemical durability, response to a radiation-field, and physical properties as well as the time required for isolation to protect the environment.
... One widely accepted method is the deep geological disposal for the final disposal of radioactive waste. Glass, glass-ceramic, and ceramic waste forms have been developed for the host matrices for HLW immobilization before deep geological disposal [1][2][3][4][5][6]. Glass waste forms have been used on an industrial scale in many countries because of their wide range loading of radionuclides and appropriate chemical durability. ...
... Glass waste forms have been used on an industrial scale in many countries because of their wide range loading of radionuclides and appropriate chemical durability. However, the very low solubility of actinides in glass is one of the drawbacks of the immobilization of modern HLW [1][2][3]. Crystalline ceramic waste forms have been developed as potential candidates owing to their high solubility of actinides, high chemical durability, and excellent radiation tolerance [1][2][3][4][5][6]. Glass-ceramic waste forms have the possibility of combining the advantageous properties of both ceramic and glass waste forms, which require more studies [1][2][3]. ...
... However, the very low solubility of actinides in glass is one of the drawbacks of the immobilization of modern HLW [1][2][3]. Crystalline ceramic waste forms have been developed as potential candidates owing to their high solubility of actinides, high chemical durability, and excellent radiation tolerance [1][2][3][4][5][6]. Glass-ceramic waste forms have the possibility of combining the advantageous properties of both ceramic and glass waste forms, which require more studies [1][2][3]. ...
Fluorite-derived A2B2O7 and ABC2O7 structures, i.e. pyrochlore and zirconolite, are promising ceramic matrices for high-level radioactive waste immobilization, especially for minor actinides. Though substitutions of low valent cations were reported by many studies, fewer researches were carried out on the incorporation of high valency elements on these fluorite-derived matrices. In this study, a series of CaZr1-xNdxTi2-xNbxO7 (x = 0.05–1.0) ceramics was investigated to unravel the effect of Nb⁵⁺ - one of the main impurities in natural zirconolite, on the evolutions of crystal structure, substitution mechanism and phase transformation. X-ray photoelectron spectroscopy was used to confirm the existence of Nb⁵⁺ in the samples. Fluorite-derived structures (zirconolite-2M, 3T, 4 M and pyrochlore) with increasing Nb⁵⁺ contents simultaneously appeared as single-phase or multi-phase assemblages in the matrices, revealed by the combination of synchrotron X-ray diffraction and scanning electron microscopy-energy dispersive x-ray spectroscopy. Almost single-phase zirconolite-2M and pyrochlore were observed only in the ranges of 0.05 ≤ x ≤ 0.1 and 0.5 ≤ x ≤ 0.9, respectively. Rietveld structural analysis of the zirconolite-2M phase showed that Nd³⁺ ions preferably occupied the Zr sites while Nb⁵⁺ ions substituted the Ti sites. This result is different from the mechanisms reported in the literature when lower valency (≤+3) was as charge compensator or there was no charge compensator in the REE doped zirconolite (REE = rare earth element). In the pyrochlore structure, the Ca²⁺, Nd³⁺ and Zr⁴⁺ ions occupied the A sublattice while Ti⁴⁺ and Nb⁵⁺ ions filled in the B sublattice, and there is charge compensation between A and B sublattices. The results in this study showed a different charge compensation mechanism of the fluorite-derived structures when higher valency elements incorporate into the lattice and will be very helpful for high-level radioactive waste immobilization.
... Murataite, more exactly polytypes of the modular murataite/ pyrochlore series, whose structures are built from pyrochlore (twofold fluorite unit cell) and murataite (three-fold fluorite unit cell) blocks (modules) [1], is considered to be one of the perspective host for immobilization of complex rare earth-and actinide-bearing wastes [2,3]. Naturally-occurred murataite is a rare cubic structure mineral with space group F-43 m, lattice parameter a ¼ 14.9 Å, Z ¼ 4, and formula [8]R 6 [6]М1 12 [5]М2 4 [4]ТХ 43 , where R ¼ Y, Na, Са, Мn; М1 ¼ Ti, Nb, Na; М2 ¼ Zn, Fe, Ti, Na; T ¼ Zn; X ¼ O, F, OH [4]. ...
... The IV T sites are filled with small-sized cations (Zn 2þ , Si 4þ ) in the natural mineral and normally empty in the synthetic phases. Currently the general synthetic murataite formula is written as VIII A 3 - [3]. hus the phases of the murataite-pyrochlore series are capable to incorporate three-(Am, Cm) and four-charged (U, Pu, Np) actinide cations, fission products (Zr and rare earth elements) and corrosion products, formed at cladding dissolution (Zr, Fe, Mn) as well. ...
... Previously [3] we produced murataite-based ceramics with baseline composition (wt%): 5 Al 2 O 3 , 10 CaO, 55 TiO 2 , 10 MnO, 5 Fe 2 O 3 , 5 ZrO 2 , 10 An/REs. The goal of this work is comparative study of the phase composition and elemental partitioning among coexisting phases in the modified murataite-containing ceramics with different REs and An at slightly reduced TiO 2 and increased ZrO 2 contents. ...
Murataite-based ceramics containing 10 wt% rare earth (REs = La, Ce, Nd, Ho) or actinide (An = Th, U) oxides were produced in a resistive furnace at a temperature of 1500 °C. The samples were composed of major murataite-type phases and minor perovskite, crichtonite, and zirconolite/pyrochlore. Light (Ce-group) REs are concentrated in the perovskite whereas Nd and Ho are mainly accommodated in the murataite. U and Th enter predominantly murataite. Ce and U regardless from their oxidation state in the batch (Ce2O3/CeO2, UO2/UO3) exist in the murataite as Ce(III) or U(IV) with some contribution due to U(V) for the uranium-doped sample.
... The The possibility for incorporation of An and RE elements in the structure of the murataite-type phases has been demonstrated and a number of actinide waste forms has been produced and examined. Such actinide wastes as An and An/RE fractions of high level waste (HLW), Np, minor actinides, excess weapons plutonium, actinide-bearing ashes and slags were converted into high chemically durable and radiation resistant ceramic waste forms [1][2][3][4][5][6][7][8][9][10][11][12][13][14]. In our previous works we have determined the baseline chemical composition of the An/RE bearing ceramic with total content of the murataite polytypes of as many as 90 vol.%. ...
... Concentrations of Mn, Fe, and Al oxides vary in the reverse order as it is characteristic of the ceramics with different An and RE oxides [2,3,6]. ...
... The reflections with maxima at d = 2.847…2.858 Å are due to 5C polytype [2,3,6,7,13]. The reflections at 2.829…2.837 ...
The samples of U-bearing ceramics with a composition (wt%) 5 Al2O3, 10 CaO, 55 TiO2, 10 MnO, 5 Fe2O3, 5 ZrO2, 10 UO2 were produced by melting of oxide mixtures in a Pt ampoule, glassy carbon crucibles in a resistive furnace, and cold crucible inductive melting (CCIM). In total 85-95 vol% of the samples produced were composed of murataite polytypes with 5- (5C), 8- (8C), and 3- (3C) fold elementary fluorite unit cell forming respectively core, medium part and rim of the murataite grains. Crichtonite, rutile, and Mn/Fe titanate (pyrophanite/ilmenite) were found to be minor phases. The latter was primarily present in the ceramics produced by melting in glassy carbon crucibles. Cold crucible melted ceramics contained the traces of vitreous phase due to melt contamination with a cold crucible putty material. Average U leach rates from the ceramics measured by SPFT procedure were found to be 10⁻³ g m⁻² d⁻¹ and lower.
... Since the 1970s, there have been numerous investigations of alternative waste forms. Considerable efforts have been focused on crystalline ceramics (Sinclair and Ringwood, 1981;Ringwood, 1982;Kesson and Ringwood, 1983;Weber et al., 1997Weber et al., , 2009Burakov et al., 1999;Yudintsev et al., 2002Yudintsev et al., , 2007Ewing et al., 2004;Rusakov et al., 2005;Urusov et al., 2005;Utsunomiya et al., 2005;Lumpkin, 2006;Omel'yanenko et al., 2007), which may serve as more durable and thus safer waste forms with high loadings of radionuclides. Various tailored ceramic forms containing mineral phases such as garnet, perovskite, monazite, pyrochlore, zircon, zirconolite, and defect fluorite, (Lumpkin, 2006;Yudintsev et al., 2007;Navrotsky et al., 2013) have been studied in terms of their chemical durability, waste loading, radiation tolerance and other related properties. ...
... Considerable efforts have been focused on crystalline ceramics (Sinclair and Ringwood, 1981;Ringwood, 1982;Kesson and Ringwood, 1983;Weber et al., 1997Weber et al., , 2009Burakov et al., 1999;Yudintsev et al., 2002Yudintsev et al., , 2007Ewing et al., 2004;Rusakov et al., 2005;Urusov et al., 2005;Utsunomiya et al., 2005;Lumpkin, 2006;Omel'yanenko et al., 2007), which may serve as more durable and thus safer waste forms with high loadings of radionuclides. Various tailored ceramic forms containing mineral phases such as garnet, perovskite, monazite, pyrochlore, zircon, zirconolite, and defect fluorite, (Lumpkin, 2006;Yudintsev et al., 2007;Navrotsky et al., 2013) have been studied in terms of their chemical durability, waste loading, radiation tolerance and other related properties. ...
... Garnet, a relatively new host for accommodation of actinides, was proposed in a number of recent studies (Burakov and Strykanova, 1998;Burakov et al., 2000;Zamoryanskaya and Burakov, 2000;Yudintsev et al., 2002Yudintsev et al., , 2007Utsunomiya et al., 2002aUtsunomiya et al., ,b, 2005Rusakov et al., 2005;Livshits, 2008;Galuskina et al., 2010;Laverov et al., 2003Laverov et al., , 2010Stefanovsky et al., 2010;Zhang et al., 2010;Rak et al., 2013b;Guo et al., 2015a) due to its large loading of actinides and good chemical flexibility. The garnet structure has been shown to incorporate actinides in various oxidation states (Burakov and Strykanova, 1998;Burakov et al., 2000;Zamoryanskaya and Burakov, 2000;Yudintsev et al., 2002;Rusakov et al., 2005;Galuskina et al., 2010;Laverov et al., 2010;Stefanovsky et al., 2010;Rak et al., 2013b;Guo et al., 2014a). ...
Use of crystalline garnet as a waste form phase appears to be advantageous for accommodating actinides from nuclear waste. Previous studies show that large amounts of uranium (U) and its analogues such as cerium (Ce) and thorium (Th) can be incorporated into the garnet structure. In this study, we synthesized U loaded garnet phases, Ca3UxZr2-xFe3O12 (x = 0.5 - 0.7), along with the endmember phase, Ca3(Zr2)SiFe3+2O12, for comparison. The oxidation states of U were determined by X-ray photoelectron and absorption spectroscopies, revealing the presence of mixed pentavalent and hexavalent uranium in the phases with x = 0.6 and 0.7. The oxidation states and coordination environments of Fe were measured using transmission 57Fe-Mössbauer spectroscopy, which shows that all iron is tetrahedrally coordinated Fe3+. U substitution had a significant effect on local environments, the extent of U substitution within this range had a minimal effect on the structure, and unlike in the x = 0 sample, Fe exists in two different environments in the substituted garnets. The enthalpies of formation of garnet phases from constituent oxides and elements were first time determined by high temperature oxide melt solution calorimetry. The results indicate that these substituted garnets are thermodynamically stable under reducing conditions. Our structural and thermodynamic analysis further provides explanation for the formation of natural uranium garnet, elbrusite-(Zr), and supports the potential use of Ca3UxZr2-xFe3O12 as viable waste form phases for U and other actinides.
... The structure of synthetic murataite-3C intended for long-term immobilization of high-level radioactive waste has been solved using crystals prepared by melting in an electric furnace at 1500 C. The material is cubic, F 4 43m, a ¼ 14:676(15) Å , V ¼ 3161:31(57) Å 3 . The structure is based upon a three-dimensional framework consisting of a-Keggin [Al [4] Ti 12 ...
... The structure of synthetic murataite-3C intended for long-term immobilization of high-level radioactive waste has been solved using crystals prepared by melting in an electric furnace at 1500 C. The material is cubic, F 4 43m, a ¼ 14:676(15) Å , V ¼ 3161:31(57) Å 3 . The structure is based upon a three-dimensional framework consisting of a-Keggin [Al [4] Ti 12 [6] O 40 ] clusters linked by sharing the O5 atoms. The Keggin-cluster-framework interpenetrates with the metal-oxide substructure that can be considered as a derivative of the fluorite structure. ...
... The Keggin-cluster-framework interpenetrates with the metal-oxide substructure that can be considered as a derivative of the fluorite structure. The crystal chemical formula of synthetic murataite-3C derived from the obtained structure model can be written as [8] Ca 6 [8] Ca 4 [6] Ti 12 [5] Ti 4 [4] AlO 42 . Its comparison with the natural murataite shows that the synthetic material has a noticeably less number of vacancies in the cation substructure and contains five instead of four symmetrically independent cation positions. ...
The structure of synthetic murataite-3C intended for long-term immobilization of high-level radioactive waste has been solved using crystals prepared by melting in an electric furnace at 1500 °C. The material is cubic, F-43m, a = 14.676(15) Å, V = 3161.31(57) Å3. The structure is based upon a three-dimensional framework consisting of α-Keggin [Al[4]Ti12[6]O40] clusters linked by sharing the O5 atoms. The Keggin-cluster-framework interpenetrates with the metal-oxide substructure that can be considered as a derivative of the fluorite structure. The crystal chemical formula of synthetic murataite-3C derived from the obtained structure model can be written as [8]Ca6[8]Ca4[6]Ti12[5]Ti4[4]AlO42. Its comparison with the natural murataite shows that the synthetic material has a noticeably less number of vacancies in the cation substructure and contains five instead of four symmetrically independent cation positions. The presence of the additional site essentially increases the capacity of synthetic murataite with respect to large heavy cations such as actinides, rare earth and alkaline earth metals in comparison with the material of natural origin.
... Radiation-tolerant materials such as silicates are at the forefront of current applied studies in the field of power engineering, heat transfer, material science. In recent decades, there have been numerous experimental and theoretical studies concerning the radiation damage mechanisms of complex oxides (see for example an inconclusive list of review works 1,2,3,4,5,6,7 ). One of the promising directions to study the materials radiation response is based on an investigation of (U, Th)-bearing minerals as natural analogues of actinide waste forms. ...
... For the latter, in the majority of amorphous and glass-like materials, the relationship C( )∼ is established as independent of the temperature at ∼ 58 . In accordance with the model of the BP as a VDOS spectrum ( ) 2 within an accuracy to C( ) coefficient, it can be concluded that the distribution of vibrational localized states in the SLZ3 sample is approximately 2 times wider than in SLZ2 (see Fig. 4 and 5 b). Note that a sharper high-energy edge of the BP (as in the SLZ2 sample, see Fig. 4 ) is usually observed in amorphous materials of Si and Ge types, while a more broad edge (as in the SLZ3 sample, see Fig. 4 ) is characteristic of glass-like materials (for example, SiO 2 , GeO 2 ) 25 . ...
The structure of the amorphous fraction and the tensile-compressive stresses in amorphous-crystalline radiation-damaged zircon ZrSiO:U,Th depending on radiation dose and temperature (8–350K) are investigated according to Raman spectroscopy of Boson peak for the first time. The Boson peak at 60-70 cm−1 associated with localized phonon states in the amorphous fraction (f𝑎) is recorded at low temperatures (T<100K) for samples with f𝑎< 30% and over the entire temperature range 8–350K for f𝑎> 70%. The wider localized states distribution in the latter case is considered as a sign of the amorphous phase structure evolution with an increase in radiation dose. The estimates of an atomic correlation radius based on the Ioffe-Regel criterion are similar to those in glasses, R𝑐 2.0–2.3 nm. The monotonic increase in R𝑐 value during heating of zircon with f𝑎> 70% is governed by thermal expansion of the percolating amorphous fraction. The non-monotonic variations of the R𝑐 value in zircon with f𝑎< 30% is determined by the stresses in the amorphous fraction due to the mismatch in thermal expansion coefficient (CTE) and elastic moduli of the amorphous and crystalline phases depending on temperature; a change in the sign of the crystalline fraction CTE at 30 K is assumed. The Boson peak disappearance at 100 K in zircon with fa < 30% during heating conforms to with the violation of the phonon localization as a consequence of amorphous fraction contraction and partial ordering. The data obtained are important for predicting the thermal and mechanical properties of heterogeneous radiation-damaged materials and nanocomposites
... and Aldecreases) in a row 3C < 8C < 5C < 7C <2C making murataite-based ceramics the promising waste forms for the immobilization of actinide and lanthanide wastes primarily with complex composition containing minor corrosion products and process chemicals [1][2][3]. ...
... The core is composed of the polytype with maximum fraction of the pyrochlore modules (normally 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 34 35 36 37 38 39 40 41 42 43 44 45 46 47 48 49 50 51 52 53 54 55 56 57 58 59 60 61 62 63 64 65 5C, sometimes, 7C) and the highest An, Ln, and Zr contents, the middle zone is formed by mainly 8C polytype with lower An/Ln/Zr and higher Ti/Al/Fe contents, and the rim composed of the 3C polytype is nearly An/Ln/Zr free but strongly enriched with Ti, Al, and Fe. Strong depletion of the rim with actinides and lanthanides provides for their low leaching and thus excellent chemical durability [1][2][3]. ...
Samples of thorium-bearing ceramic with a target composition (wt%) 5 Al2O3, 10 CaO, 55 TiO2, 10 MnO, 5 Fe2O3, 5 ZrO2, 10 ThO2 were produced by melting in glassy carbon crucibles in a resistive furnace and by cold crucible inductive melting (CCIM) at a vibration power of 10kW and operation frequency of 5.28MHz. All the samples contained 85-95vol% murataite polytypes with 5- (5C), 8- (8C), and 3-fold (3C) elementary fluorite unit cell composing core, intermediate zone and rim of the grains, respectively, and minor crichtonite, perovskite, pyrochlore, rutile, etc. A feature of the ceramics obtained by melting in glassy carbon crucibles is formation of Fe (II) titanate whereas the inductive-melted ceramics contained traces of vitreous phase due to melt contamination with a cold crucible putty material. Melting rate in the cold crucible of up to 350kg/(m² × h) has been achieved. The ceramics obtained have excellent chemical durability.
... Some of these compounds are promising as matrices for actinide bear ing HLW [2,13,16,18,20]. Characteristics of pyrochlore and garnet that make these materials appropriate to immobilize actinide bearing wastes were discussed in [2,21]. Structures of the pyrochlore type are formed by phases of the composition А 2 В 2 О 7 when the ratios of the cation sizes (r [8] A/r [6] B) range from 1.48 to 1.80. ...
... Matrices of this type can be produced by self propa gating high temperature synthesis (SHS) [45,46]. The SHS starting mixture for pyrochlore synthesis consisted of (wt %): 21 on X ray diffraction patterns because of the low concen trations of these phases. Phase 1 is REE aluminate, and phase 2 is REE, Zr, and Mo oxide ( Table 6). ...
... Some of these compounds are promising as matrices for actinide bear ing HLW [2,13,16,18,20]. Characteristics of pyrochlore and garnet that make these materials appropriate to immobilize actinide bearing wastes were discussed in [2,21]. Structures of the pyrochlore type are formed by phases of the composition А 2 В 2 О 7 when the ratios of the cation sizes (r [8] A/r [6] B) range from 1.48 to 1.80. ...
... Matrices of this type can be produced by self propa gating high temperature synthesis (SHS) [45,46]. The SHS starting mixture for pyrochlore synthesis consisted of (wt %): 21 on X ray diffraction patterns because of the low concen trations of these phases. Phase 1 is REE aluminate, and phase 2 is REE, Zr, and Mo oxide ( Table 6). ...
Complex oxides of the pyrochlore (space groups Fd3m, [8]A2
[6]B2O7) and garnet (Ia3d, [8]A3
[6]B2
[4]T3O12) structures (“A” = Ca2+, Ln3+/4+, An3+/4+; “B” = (Ti, Sn, Hf, and Zr)4+ in pyrochlore, and Al3+, Ga3+, and Fe3+ in garnet alone; “T” = (Al3+, Ga3+, and Fe3+) are promising matrices for actinide-bearing wastes. In order to identify optimal compositions of these phases, their isomorphic
capacity with respect to REE, actinides, and other components of wastes was examined. The long-term behavior of the matrix
at a repository was predicted based on data obtained on the behavior of pyrochlores and garnets under ion irradiation and
244Cm decay and on the determined leaching rates of REE from the matrices because of their interaction with aqueous solutions,
including that after amorphization. In order to propose efficient synthesis techniques, samples prepared with the use of various
methods were studied. The possibility of incorporating long-lived decay products of 99Tc into the crystalline matrices was analyzed.
... [16][17][18][19] The pyrochlore family boasts a high number of compositions that have been successfully synthesized and studied for chemical durability, waste loading, radiation tolerance, and other related properties. [20][21][22][23][24][25][26][27] Having compositions of A 2 B 2 O 7 , the ordered pyrochlore (space group, s.g. Fd3m) is a derivative of the fluorite structure (s.g. ...
Critical particle size can be determined with known surface energy. The surface enthalpy of yttrium titanate pyrochlores was determined to be 4.07 ± 0.32 J m ⁻² by calorimetry, and the lower limit of critical particle size for this is around 5.0 nm.
... Скорость выщелачивания кюрия из этой же керамики после шлифования и удаления обедненного этими фазами слоя вновь увеличилась в 2-4 раза. Таким образом, коррозионная устойчивость спеченной муратаит-содержащей керамики, полученной методами холодного прессования/спекания и плавления с последующей кристаллизацией, близка к стойкости пирохлоровых и цирконолитовых матриц [29]. ...
В работе представлены результаты исследований по включению радиоактивных отходов пирохимической переработки ОЯТ и кальцинатов от дезактивации защитного оборудования боксов и камер в керамику на основе муратаита. Изучен фазовый и химический состав полученной керамики с отходными формами технологических осадков-концентраторов продуктов деления в расплавах хлоридов. Радиационная устойчивость муратаитовых матриц изучалась путем их импрегнирования изотопом кюрия-244 (1,8 масс.%). В керамиках, полученных сплавлением при 1325 и 1350 ºС, рентгеновская аморфизация муратаитовых фаз достигается при дозах 2,46•10 18 и 2,53•10 18 α-распад/г (~0,19 смещ./ат.), а для спеченной керамики (температура спекания 1250 ºС) 2,73•10 18 α-распад/г (0,21 смещ./ат.). Структура муратаита была восстановлена после отжига при 1250 ºС в течение 5 часов на воздухе. Исходные и аморфизированные образцы имеют очень низкий уровень выщелачивания кюрия и макрокомпонентов. Продемонстрировано получение устойчивых керамик, содержащих оксидные осадки от переработки ОЯТ пирохимическим способом и кальцинаты после упаривания дезактивационных растворов
... There are also 132 some oxygen atoms that do not bridge phosphate tetrahedra and 133 these are known as non-bridging oxygen atoms. It may be noted 134 that the double bonded oxygen cannot bridge between neighbor- 135 ing tetrahedra. Therefore, any phosphate tetrahedron (PO 4 ) can 136 have one, two or at most three bridging oxygen atoms. ...
Vitrification is currently the most effective process for immobilization of nuclear waste. However, ubiquitous borosilicate glass is not suitable for immobilization of nuclear waste from advanced reactors such as Fast Breeder Reactors (FBR) because solubility of many compounds/elements existing in the spent fuel in borosilicate glasses is quite poor. In order to possess a viable immobilization strategy for wastes arising from advanced reactors, alternatives to borosilicate glasses such as phosphate glasses, glass-ceramics and crystalline waste forms are being investigated. This review aims to provide an overview of nuclear waste immobilization employing phosphate-based glasses, glass-ceramics and crystalline ceramic hosts, focusing on structure and properties that make these new matrices suitable for the challenging task of waste immobilization.
... A number of host phases were proposed for this goal such as titanates, zirconates, mixed titanozirconates with fluorite-derived structure (cubic zirconia, pyrochlore, murataite, zirconolite), phosphates and silicophosphates with monazite, apatite and kosnarite structures. They all possess high isomorphic capacity with respect to actinides and their rare earth (RE) analogs and good isolation characteristics (see, for example, [1][2][3]). Another titanate phase encountered in nuclear waste ceramics is a brannerite structure phase which was also suggested as a potential actinide form [4,5]. Brannerite with nominal formula AВ 2 O 6 –(U,Th,Ca,RE)(Ti,Fe) 2 O 6 has a monoclinic lattice symmetry (space group C2/m, Z = 2). ...
Branneritecontaining ceramics were produced by cold pressing and sintering (CPS) and cold crucible inductive melting (CCIM) methods and examined by X-ray diffraction and absorption and by scanning electron microscopy. Brannerite content in the ceramics ranged between ∼20 and 90 vol.%. Uranium is mainly partitioned between brannerite and minor mixed U/RE oxide but since brannerite is a dominant phase, it takes up to 90% of total U. Uranium in the ceramics is present as U(IV) and U(V). In the low-brannerite ceramics U occurs as U(IV) whereas in the ceramics with brannerite as major phase U(V) dominates over U(IV). Ce in the brannerite ceramics is mainly trivalent. The first coordination shell of U in ceramics produced by CPS is split into two sub-shells with U-O distances of 1.7–1.9 Å and ∼2.1 Å while in the melted ceramics this interatomic distance is 2.1–2.2 Å. The next three atoms (Ti) are positioned at a distances of 3.1–3.2 Å.
... However, the spent nuclear fuel (SNF) contains fission products, corrosion products, process contaminants, fuel components, and transmutation products. 1 These wastes would be a severe threat to ecological environment because they contain high radioactive and long-lived actinides, such as 238 Pu, 239 Pu, 237 Np, 243 Am and so on. 2 Therefore, safe and effective disposal of high level waste (HLW) is crucial to the public acceptance and sustainable development of nuclear energy. ...
Polycrystalline samples of (Gd1−xCex)2Zr2O7+x (0 ≤ x ≤ 0.7) were synthesized by solid-state reaction using NaF as a flux at 1000 °C to simulate Pu-immobilization in the Gd2Zr2O7 matrix. Phase transformation and variation of the thermal expansion coefficients (TECs) and thermal conductivities (TCs) of the samples with temperature and composition were investigated. Powder X-ray diffraction (XRD) patterns show that pure Gd2Zr2O7 has a weakly ordered pyrochlore structure, whereas Ce-containing samples (i.e., the Pu-simulated solidified samples) exhibit a defect fluorite structure even if x is as low as 0.1. When x reaches 0.7, the XRD peaks of these samples widen. In the Ce 3d X-ray photoelectron spectrum (XPS) of (Gd1−xCex)2Zr2O7+x there are six peaks located at binding energies of 881.7, 888.1, 897.8, 900.4, 907.1, and 916.1 eV, which are almost the same as the peaks of CeO2. The Ce 3d XPS reveals that the Ce species in (Gd1−xCex)2Zr2O7+x are tetravalent. The TECs of (Gd1−xCex)2Zr2O7+x (0 ≤ x ≤ 0.7) generally increase with increasing temperature. At the same temperature, the TECs and TCs exhibit the same variation trend with the composition of the simulated solidified forms: they decrease from x = 0 to 0.1 and then linearly increase from x = 0.1 to 0.7.
... This reveals their strong resistance against ionising irradiation and their good chemical durability ( Lumpkin, 2006;Schlenz et al, 2013;Weber et al., 2009;Meldrum et al., 1997). Therefore, in the past, several investigations have been made to characterise monazite-type phases as a potential waste matrix for the conditioning of minor actinides and plutonium ( Boatner et al., 1988;Tabuteau et al., 1988;Ewing and Wang, 2002;Dacheux et al., 2004;Ewing, 2007;Yudintsev et al., 2007;Oelkers and Montel, 2008;Clavier et al., 2011;Boatner, 2002;Bregiroux et al., 2007;de Kerdaniel et al., 2007). These radionuclides have long half-lives and are responsible for the high radiotoxicity of spent nuclear fuel. ...
Raman and infrared spectroscopic investigations were performed on synthetic lanthanide orthophosphates (LnPO(4)) within this study. Seven monoclinic monazite-type phosphates (Ln = La-Gd) were synthesised via precipitation route in aqueous solution at room temperature. Linear correlation between Raman band positions and the effective cationic radii of the Ln(3+) was observed. New infrared spectroscopic data confirmed the expected steady increase of the LnPO4 wavenumbers regarding the Ln atomic number.
... 8 Uranium phosphate species are also among the most numerous, widespread, abundant, and insoluble actinidebearing materials and, therefore, have been studied as potential actinide host phases. 9 Therefore, the synthesis and crystal chemistry of uranium phosphates have been subjects of intense interest because the structures and behaviors of these compounds provide a basis from which to predict the longterm behavior of actinide-bearing phosphates. Few tetravalent uranium silicates are known; by contrast, a large number of uranium(IV) phosphates have been reported. ...
The isotypic compounds (III) and (V) are characterized by single crystal XRD, UV/VIS spectroscopy, and XPS.
... The stability and low solubility of uranyl phosphates have prompted research into immobilization of actinides and several phosphates such as monazite, Th-phosphate-diphosphate and kosnarite as potential actinide host phases. 3 Arsenates do not appear to have been considered as practical phases for actinide waste forms probably because of the toxicity of arsenic compounds. A thorough knowledge of the structures and behavior of uranium phosphates and arsenates may form a basis from which to predict the long-term behavior of actinide phosphate phases. ...
Five new uranyl arsenates, Na14[(UO2)5(AsO4)8]·2H2O (1), K6[(UO2)5O5(AsO4)2] (2a), K4[(UO2)3O2(AsO4)2] (2b), Rb4[(UO2)3O2(AsO4)2] (3), and Cs6[(UO2)5O2(AsO4)4] (4), were synthesized by high-temperature, high-pressure hydrothermal reactions at about 560 °C and 1440 bar and were characterized by single-crystal X-ray diffraction, thermogravimetric analysis, and photoluminescence spectroscopy. Crystal data for compound 1: triclinic, P1̅, a = 7.0005(3) Å, b = 12.1324(4) Å, c = 13.7428(5) Å, α = 64.175(2)°, β = 89.092(2)°, γ = 85.548(2)°, V = 1047.26(7) Å(3), Z = 1, R1 = 0.0185; compound 2a: monoclinic, P21/c, a = 6.8615(3) Å, b = 24.702(1) Å, c = 7.1269(3) Å, β = 98.749(2)°, V = 1193.89(9) Å(3), Z = 2, R1 = 0.0225; compound 2b: monoclinic, P21/c, a = 6.7852(3) Å, b = 17.3640(8) Å, c = 7.1151(3) Å, β = 98.801(3)°, V = 828.42(6) Å(3), Z = 2, R1 = 0.0269; compound 3: monoclinic, P21/m, a = 6.9783(3) Å, b = 17.4513(8) Å, c = 7.0867(3) Å, β = 90.808(3)°, V = 862.94(7) Å(3), Z = 2, R1 = 0.0269; compound 4: triclinic, P1̅, a = 7.7628(3) Å, b = 9.3324(4) Å, c = 11.9336(4) Å, α = 75.611(2)°, β = 73.136(2)°, γ = 86.329(2)°, V = 801.37(5) Å(3), Z = 1, R1 = 0.0336. The five compounds have layer structures consisting of uranyl square, pentagonal, and hexagonal bipyramids as well as AsO4 tetrahedra. Compound 1 contains chains of discrete uranyl square and pentagonal bipyramids, 2a contains three-polyhedron-wide ribbons of edge- and corner-sharing uranyl square and pentagonal bipyramids, 2b and 3 contain dimers of edge-shairing pentagonal bipyramids that share edges with hexagonal bipyramids to form chains, and 4 contains one-polyhedron-wide zigzag chains of edge-sharing uranyl polyhedra. The double sheet structure of 1 is new, but the chain topology has been observed in an organically templated uranyl sulfate. Compound 2b is a new geometrical isomer of the phosphuranylite group. The sheet anion topologies of 2a and 4 can be obtained by splitting the β-U3O8-type sheet into complex chains and connecting the chains by arsenates.
... 8 Uranium phosphate species are also among the most numerous, widespread, abundant, and insoluble actinidebearing materials and, therefore, have been studied as potential actinide host phases. 9 Therefore, the synthesis and crystal chemistry of uranium phosphates have been subjects of intense interest because the structures and behaviors of these compounds provide a basis from which to predict the longterm behavior of actinide-bearing phosphates. Few tetravalent uranium silicates are known; by contrast, a large number of uranium(IV) phosphates have been reported. ...
A uranium(IV) phosphate, Na10U2P6O24, was synthesized under hydrothermal conditions at 570 °C and 160 MPa and structurally characterized by single-crystal X-ray diffraction. The valence state of uranium was established by UV-vis and U 4f X-ray photoelectron spectroscopy. The powder sample has a second-harmonic-generation signal, confirming the absence of a center of symmetry in the structure. The structure contains UO8 snub-disphenoidal polyhedra that are linked to monophosphate tetrahedra by vertex and edge sharing such that a three-dimensional framework with intersecting 12-sided circular and rectangular channels is formed. All 10 sodium sites are situated inside the channels and are fully occupied. This is the first uranium(IV) phosphate synthesized under high-temperature, high-pressure hydrothermal conditions. The isotypic cerium(IV) phosphate, Na10Ce2P6O24, was also synthesized under the same hydrothermal conditions. It is the first structurally characterized Ce(IV) phosphate with a P/Ce ratio of 3. Crystal data of Na10U2P6O24: orthorhombic, P212121 (No. 19), a = 6.9289(3) Å, b = 16.1850(7) Å, c = 18.7285(7) Å, V = 2100.3(2) Å(3), Z = 4, R1 = 0.0304, and wR2 = 0.0522. Crystal data of Na10Ce2P6O24: orthorhombic, P212121 (No. 19), a = 6.9375(14) Å, b = 16.215(3) Å, c = 18.765(4) Å, V = 2111.0(7) Å(3), Z = 4, R1 = 0.0202, and wR2 = 0.0529.
... Hence, this garnet is an ideal model substance to investigate problems concerning potential use of garnet matrices for immobilization of high-level radioactive waste. Up to now, absence of natural garnets with significant U and Th contents forced researchers to use synthetic materials to study the substitution capacity, as well as behavior and stability of garnet matrices as potential materials for immobilization of actinides ( Burakov et al. 2000;Yudintsev et al. 2002Yudintsev et al. , 2007Yudintsev 2003;Livshits and Yudintsev 2008;laverov et al. 2010). Utsunomiya with co-authors ( Utsunomiya et al. 2002aUtsunomiya et al. , 2002bUtsunomiya et al. , 2005) investigated the radiation susceptibility of garnet matrices both natural silicate garnets and synthetic garnets of different composition irradiated with 1 MeV Kr 2+ ions. ...
Elbrusite-(Zr) Ca3(U6+Zr)(Fe3+2Fe2+)O12, a new uranian garnet (Ia 3 d, a ≈ 12.55 Å, V ≈ 1977 Å3, Z = 8), within the complex solid solution elbrusite-kimzeyite-toturite Ca3(U,Zr,Sn,Ti,Sb,Sc,Nb...)2(Fe,Al,Si,Ti)3O12 was discovered in spurrite zones in skarn xenoliths of the Upper Chegem caldera. The empirical formula of holotype elbrusite-(Zr)
with 25.14 wt% UO3 is (Ca3.040Th0.018Y0.001) ∑3.059(U6+0.658Zr1.040Sn0.230 Hf0.009Mg0.004) ∑1.941 (Fe3+1.575Fe2+0.559Al0.539 Ti4+0.199Si0.099Sn0.025V5+0.004) ∑3O12. Associated minerals are spurrite, rondorfite, wadalite, kimzeyite, perovskite, lakargiite, ellestadite-(OH), hillebrandite,
afwillite, hydrocalumite, ettringite group minerals, and hydrogrossular. Elbrusite-(Zr) forms grains up to 10–15 μm in size
with dominant {110} and minor {211} forms. It often occurs as zones and spots within Fe3+-dominant kimzeyite crystals up to 20–30 μm in size. The mineral is dark-brown to black with a brown streak. The density calculated
on the basis of the empirical formula is 4.801 g/cm3 The following broad bands are observed in the Raman spectra of elbrusite-(Zr): 730, 478, 273, 222, and 135 cm−1. Elbrusite-(Zr) is radioactive and nearly completely metamict. The calculated cumulative dose (α-decay events/mg) of the
studied garnets varies from 2.50 × 1014 [is equivalent to 0.04 displacement per atom (dpa)] for uranian kimzeyite (3.36 wt% UO3), up to 2.05 × 1015 (0.40 dpa) for elbrusite-(Zr) with 27.09 wt% UO3.
The magmatic diagenetic environment was simulated by high-temperature melting and natural cooling. A series of glass-ceramics with different Nd 2 O 3 contents were prepared by using complex component granite (aluminosilicate material). The phase evolution of the matrix at different temperatures was studied by X-ray diffraction (XRD). The structure of glass-ceramics was analyzed by infrared spectroscopy (IR) and scanning electron microscopy (SEM). The mechanical properties of glass-ceramics were also evaluated. The results showed that the glass transition of pure matrix begins at 1200 °C, and the sample with the highest degree of vitrification is obtained at 1500 °C. The addition of Nd 2 O 3 promoted the melting of Fe 3 O 4 crystal, resulting in the complete amorphous matrix when the Nd 2 O 3 amount is in the range of 20–26 wt .%. With the further increase of Nd 2 O 3 content, Nd-bearing feldspar first appeared. No raw material Nd 2 O 3 was found, indicating that the formation of Nd-bearing feldspar may increase the carrying capacity of the material. The Gaussian fitting results showed that the glass-ceramic samples with Nd 2 O 3 content of 29 wt .% are mainly composed of Q ² and Q ³ structural units. In the EDS result, part of neodymium was clustered with small bright spots, while the spots were uniformly distributed on the sample surface as a whole. Meanwhile, the addition of Nd 2 O 3 increased the mechanical properties of the samples (3.20 g/cm ³ , 8.33 GPa for the sample with 29 wt .% of Nd 2 O 3 ). The results provide a strategy for the treatment of solid waste with radioactive residual actinides.
A series of zirconolite ceramics with stoichiometry Ca1-xCexZrTi2-2xAl2xO7 (x = 0 – 0.35), considered as a host phase for the immobilisation of separated plutonium, were prepared from a mixture of oxide precursors by sintering in air at 1450°C. Ce was utilised as a structural surrogate for Pu, with Al added to provide charge compensation. XRD and electron diffraction analyses indicated crystallisation of the zirconolite-2M polytype for all compositions, accompanied by various secondary phases contingent on the doping level, consistent with microstructure observation. The relative yield of zirconolite phases remained above 94 wt.% for 0.05 < x < 0.20. It was determined that Ce was partially reduced to the Ce3+ oxidation state and Al occupied mainly octahedral Ti sites. The incorporation rate of CeO2 was calculated to be 9.27 wt.% in Ca0.80Ce0.20ZrTi1.60Al0.40O7 with a comparatively high yield of 94.7 wt.%, which is representative of a PuO2 incorporation rate of 14.86 wt.%.
Non-ideal thermodynamics of solid solutions can greatly impact materials degradation behavior. We have investigated an actinide silicate solid solution system (USiO 4 –ThSiO 4 ), demonstrating that thermodynamic non-ideality follows a distinctive, atomic-scale disordering process, which is usually considered as a random distribution. Neutron total scattering implemented by pair distribution function analysis confirmed a random distribution model for U and Th in first three coordination shells; however, a machine-learning algorithm suggested heterogeneous U and Th clusters at nanoscale (~2 nm). The local disorder and nanosized heterogeneous is an example of the non-ideality of mixing that has an electronic origin. Partial covalency from the U/Th 5 f –O 2 p hybridization promotes electron transfer during mixing and leads to local polyhedral distortions. The electronic origin accounts for the strong non-ideality in thermodynamic parameters that extends the stability field of the actinide silicates in nature and under typical nuclear waste repository conditions.
Low-temperature mineral-like magnesium potassium phosphate matrix of various chemical compositions have been synthesized. Their phase composition, structure and distribution of radioactive waste components have been studied. The mechanical, thermal, radiation and hydrolytic stability of the synthesized materials have been investigated according to the valid regulatory requirements. It has been demonstrated that the magnesium potassium phosphate matrix is promising materials for industrial solidification of the liquid waste, including highly saline actinide-containing waste with complex chemical composition.
La-Ce-Gd titanate-zirconate pyrochlore-based ceramics was synthesized at the lab-scale inductive melting unit with a 56 mm inner diameter cold crucible. Batch feeding rate and melting ratio were 3 kg/h and 3 kW h/kg respectively. The ceramics is composed of 85–90 vol% zoned grains of the pyrochlore structure phase, the sample has excellent chemical durability in hot water and may be considered as a promising matrix for actinide – rare earth fraction of high level waste.
Incorporation of waste from spent nuclear fuel pyrochemical reprocessing and calcines from the decontamination of glove box and hot cell equipment into murataite-based ceramics was studied. The phase and chemical compositions of the ceramics containing precipitates simulating fission products in molten chlorides were examined. The radiation stability of murataite matrices was studied by incorporating ²⁴⁴Cm isotope (1.8 wt%). In the ceramics produced by melting at 1325 and 1350 °C the murataite phases was rendered to be X-ray amorphous at doses of 2.46 × 10¹⁸ and 2.53 × 10¹⁸ α-decay/g (0.21 dpa), while for the sample sintered at 1250 °C the amorphization dose was found to be 2.73 × 10¹⁸ α-decay/g (0.21 dpa). The murataite structure was recovered after the annealing at 1250 °C for 5 h in air. Both the pristine and amorphized samples had very low leachability of Cm and major elements. Production of highly durable murataite-based ceramics containing waste surrogate after spent nuclear fuel (SNF) pyrochemical reprocessing and calcine after the evaporation of decontamination solutions were demonstrated.
The behavior of rare earth titanate phases of monoclinic and orthorhombic symmetry under ion irradiation was studied. Three samples contain a set of rare earth elements (REE), and in one sample the high-level waste (HLW) is simulated by neodymium. The samples were prepared by sintering at 1400°С or cold crucible induction melting followed by melt crystallization. The samples were irradiated with 1-MeV Kr²⁺ ions at temperatures in the range 298–1023 K on a tandem installation consisting of an ion accelerator and a transmission electron microscope. The critical amorphization doses for the samples at 298 K were (1.5–2.5) × 10¹⁴ ion cm–2. These values correspond to those for monoclinic titanates Nd2Ti2O7 and La2Ti2O7, are close to those for REE–Ti pyrochlores and brannerite, and are lower by an order of magnitude than those for zirconolites at the same irradiation mode. The critical amorphization temperature of the phases studied (900 K) is comparable to that of Nd2Ti2O7 (920 K), but higher than that of La2Ti2O7 (840 K). Close radiation resistance of the REE2Ti2O7 and REE4Ti9O24 phases in which the rare earth elements are represented by a mixture of elements similar in composition to their set in spent nuclear fuel and liquid HLW from its reprocessing or solely by Nd confirms the assumption that Nd is a suitable surrogate for the whole rare earth–actinide fraction.
It has been found that leaching of AmIII from the surface after the sorption is characterized by slow kinetics; over 10 days are required to reach a dynamic equilibrium in the system. The leaching process is accompanied by the formation of Am-containing phosphate phase on the sorbent surface. Annealing the sample at 400 °C for 5 h after sorption results in incorporation of the sorbed cation into the sorbent structure, which allows subsequent leaching to be reduced considerably.
Ca10-2xSmxNax(PO4)6F2 (x = 0.8–1.2) solid solutions were successfully synthesized by the solid state reaction method using samarium (Sm) as the surrogate for trivalent minor actinide neptunium (Np). The influences of calcining temperature, holding time and Sm doping content on the phase composition and microstructure of Ca10-2xSmxNax(PO4)6F2 were investigated. The results indicated that the optimized calcining temperature and holding time for preparing Ca10-2xSmxNax(PO4)6F2 were 1000 °C and 2 h. Ca10-2xSmxNax(PO4)6F2 were confirmed to be the discontinuous solid solutions and solid solubility limit of Sm in the Ca10-2xSmxNax(PO4)6F2 was 1.2 formula units. The rod-like grains of Ca10-2xSmxNax(PO4)6F2 had typical hexagonal characteristics with a diameter of 3–5 μm and a length of 10–20 μm. No significant changes on the microstructures of Ca10-2xSmxNax(PO4)6F2 were observed with the increase of Sm doping content. The Ca, Sm, Na, P, O and F elements were nearly distributed uniformly in the Ca8Sm1Na1(PO4)6F2 solid solution.
Glass-ceramics, with a specific crystalline phase assembly, can combine the advantages of glass and ceramic and avoid their disadvantages. In this study, both cubic-zirconia and zirconolite-based glass-ceramics were obtained by the crystallization of SiO2–CaO–Al2O3–TiO2–ZrO2–Nd2O3–Na2O glass. Results show that all samples underwent a phase transformation from cubic-zirconia to zirconolite when crystallized at 900, 950 and 1000 °C. The size of the cubic-zirconia crystal could be controlled by temperature and dwelling time. Both cubic-zirconia and zirconolite crystals/particles show dendrite shapes, but with different dendrite branching. The dendrite cubic-zirconia showed highly oriented growth. SEM images show that the branches of the cubic-zirconia crystal had a snowflake-like appearance while those in zirconolite were composed of many individual crystals. Rietveld quantitative analysis revealed that the maximum amount of zirconolite was ~19 wt.%. A two-stage crystallization method was used to obtain different microstructures of zirconolite-based glass-ceramic. The amount of zirconolite remained approximately 19 wt.%, but the individual crystals were smaller and more homogeneously dispersed in the dendrite structure than those obtained from one-stage crystallization. This process-control feature can result in different sizes and morphologies of cubic-zirconia and zirconolite crystals to facilitate the design of glass-ceramic waste forms for nuclear wastes.
This paper considers various matrices that are able to incorporate components of radioactive wastes (RAW) of different origin. It is noted that attempts to develop the single phase crystalline matrix to immobilize all RAW components failed. The only single phase matrix brought to the industrial application is glass, which is able to accumulate practically all RAW components but in limited concentrations. Prospects are related with some types of ceramics for immobilization of narrow fractions of RAW or individual radionu-clides (for instance, minor actinides), as well as some types of low-temperature matrices (iron-phosphate, magnesium–potassium–phosphate, and geopolymers). Approaches to choosing the technology of waste form synthesis are considered. Perspectives of application of both high-temperature (cold-crucible induction melting, self-propagating high-temperature synthesis) methods and modified cementation technologies are demonstrated. It is noted that the final isolation of RAW from the biosphere suggests their disposal in underground repositories. The most difficult technical problem is the disposal of RAW containing long-lived radio-nuclides. It is shown that the quantitative assessment of repository safety with allowance for their characteristics and all possible processes and phenomena is required to substantiate the safe disposal of long-lived radionuclides.
High-temperature oxide melt solution calorimetric measurements were completed to determine the enthalpies of formation of the uranothorite, (USiO4)x - (ThSiO4)1-x, solid solution. Phase - pure samples with x = 0, 0.11, 0.21, 0.35, 0.71, and 0.84 were prepared, purified, and characterized by powder X-ray diffraction, electron probe microanalysis, thermogravimetric analysis and differential scanning calorimetry coupled with in situ mass spectrometry, and high temperature oxide melt solution calorimetry. This work confirms the energetic metastability of coffinite, USiO4, and of U-rich intermediate silicate phases with respect to a mixture of binary oxides. However, variations in unit cell parameters and negative excess volumes of mixing, coupled with strongly exothermic enthalpies of mixing in the solid solution, suggest short-range cation ordering that can stabilize intermediate compositions, especially near x = 0.5.
Presented here are two isostructural uranyl coordination polymers [UO2(EDO)(H2O)]·H2O (1) and [UO2(BDO)(H2O)]·2H2O (2) (EDO2−=ethylene-1,2-dioxamate; BDO2−=butylene-1,2-dioxamate) with identical stepwise zigzag chain structure and distinct interchain hydrogen bonding interaction, prepared from hydrothermal reaction of DEEDO or DEBDO (DEEDO=diethyl ethylene-1,2-dioxamate; DEBDO=diethyl butylene-1,2-dioxamate) with uranyl ions. The monomeric uranyl-based fluorescence emissions of compounds 1 and 2 are red-shifted by about 6 and 5 nm respectively, compared to that of uranyl nitrate hexahydrate. Compound 1 has stronger emission than compound 2, but both their emissions exhibit triple-exponential decay. The photophysics of uranyl oxalate trihydrate was also investigated for comparison. The selective crystallization of compound 1 in alkaline solution was applied to the sequestration of uranyl ions, showing a kinetic preference.
Uranium is one of the most important nuclear materials. It is the core of the nuclear fuel cycle and the element to which much attention has been paid since it is inevitable that uranium is released in the process of uranium mining, enrichment, oxide fuel manufacturing and spent nuclear fuel reprocessing. Inorganic acid and carboxylic acid group of the humus can be coordinated to uranyl thus change its migration and sorption behavior in the environment. By analyzing the synthesis and crystal structure of uranyl complexes we can provide essential structure data for sorption modeling and sorption mechanism in the molecular level. The structural characters of uranyl primary building units and the complexes that may form with common inorganic and organic acids in the aquatic solutions have been reviewed. A proposal followed by the summary and analysis of the work carried out in the groups in the world of uranyl complexes based on environmentally concerned acids and some other oxo containing ligands is presented.
One of the most important problems related to the use of nuclear energy and of advanced nuclear fuel cycle is the immobilization of minor actinides as well as of rare earth fractions and corrosion products present in high level radioactive waste (HLW) (Lutze and Ewing 1988; Ewing 1999, 2005; Stefanovsky et al. 2004; Yudintsev et al. 2007a; Weber et al. 2009; Livshits and Yudintsev 2008). The most enduring proposal was made by Ringwood et al. (1979), who invented Synroc, a polyphase waste form consisting of titanates. © 2012 Springer-Verlag Berlin Heidelberg. All rights are reserved.
The new trends in the development of nuclear power industry are summarized with the special focus on their effects on the environment.
Discussing the difference of Al−Nd co-doped zirconolite derived from glass matrix and powder sintering. Using split-atom model to construct the crystal structure of zirconolite crystal with heavy stacking faults. Revealing that the heavy stacking faults may lead to substantial differences in the Al-Nd co-doped zirconolite structures prepared by these two fabrication routes. Firstly using TEM-EDX to determine the chemical compositions (Ca0.83Nd0.25Zr0.85Ti1.95Al0.11O7) of the zirconolite nano-crystals derived from glass matrix.
Sample of zirconate ceramic with a composition corresponding to formula Gd1.7
241Am0.3Zr2O7 was synthesized by heat-treatment of mechanically activated and compacted in pellet oxide mixture at 1500 °C for 30 min. The d values on XRD pattern of the sample soon after synthesis (D = 7.9×1015 α-decays/g or 0.001 dpa) demonstrated fluorite structure with the most intensive peak with d
111 =3.042 Å (a = 5.269 Å) and very weak diffuse reflections due to d-pyrochlore. At a dose of 7.9×1017 α-decays/g or 0.11 dpa the reflections were broadened by approximately 20% and their relative intensity slightly reduced. At higher doses all the weak superstructure reflections disappeared and the growth in intensity and narrowing of the main reflection occurred. Lattice parameter a increased with the dose and reached 5.343 Å (d
111 = 3.085 Å) at a dose of 4.6×1018 α-decays/g or 0.42 dpa. At a dose of 5.5×1018 α-decays/g or 0.78 dpa positions of reflections were shifted to lower d-spaces (d
111 value reduced to 3.071 Å) and the half-width of the major reflection was 67% of initial. For the 241Am-doped Gd-zirconate the structure recovery rate exceeds disordering rate and no amorphization occurred at doses higher than ∼0.2-0.3 dpa.
The garnet structure has been proposed as a potential crystalline nuclear waste form for accommodation of actinide elements, especially uranium (U). In this study, yttrium iron garnet (YIG) as a model garnet host was studied for the incorporation of U analogs, cerium (Ce) and thorium (Th), incorporated by a charge-coupled substitution with calcium (Ca) for yttrium (Y) in YIG, namely, 2Y 3+ = Ca 2+ + M 4+ , where M 4+ = Ce 4+ or Th 4+. Single-phase garnets Y 3−x Ca 0.5x M 0.5x Fe 5 O 12 (x = 0.1−0.7) were synthesized by the citrate−nitrate combustion method. Ce was confirmed to be tetravalent by X-ray absorption spectroscopy and X-ray photo-electron spectroscopy. X-ray diffraction and 57 Fe−Mö ssbauer spec-troscopy indicated that M 4+ and Ca 2+ cations are restricted to the c site, and the local environments of both the tetrahedral and the octahedral Fe 3+ are systematically affected by the extent of substitution. The charge-coupled substitution has advantages in incorporating Ce/Th and in stabilizing the substituted phases compared to a single substitution strategy. Enthalpies of formation of garnets were obtained by high temperature oxide melt solution calorimetry, and the enthalpies of substitution of Ce and Th were determined. The thermodynamic analysis demonstrates that the substituted garnets are entropically rather than energetically stabilized. This suggests that such garnets may form and persist in repositories at high temperature but might decompose near room temperature.
Cold crucible inductive melting is a promising method for production of high-temperature materials. The method is based on direct heating of conductive materials by high-frequency (105-107 Hz) electromagnetic field from an external source. Application of the CCIM to production of vitreous borosilicate and alumino/iron phosphate and ceramic waste forms such as Synroc and its varieties and pyrochlore, murataite and garnet-based ceramics has been successfully demonstrated. Currently a full-scale low level waste vitrification plant based on a 418 mm inner diameter cold crucibles energized from a 1.76MHz/160 kW generators is under operation at SIA Radon. This plant was used for demonstration of feasibility of cold crucible vitrification of Savannah River Site high-iron and high iron/aluminum high level wastes. Numerous ceramic and glass ceramic materials containing high level and actinide waste surrogates such as actinide and actinide/rare earth fractions of high level waste have been successfully produced in the Radon lab- and bench-scale cold crucible based units operated at 5.28 and 1.76 MHz. Large-scale cold crucibles may be applied for vitrification of liquid and solid low and high level wastes whereas small-scale cold crucible may be efficiently used for immobilization of actinide-bearing waste generated from advanced nuclear fuel cycle reprocessing.
Zirconolite (CaZrTi2O7) is one of the components of Synroc materials, which are regarded throughout the world nuclear as the second generation of high-level nuclear waste forms. The zirconolite phase was synthesized by a sol-gel method, with one variant of the method making use of ascorbic acid as a strong complexing agent. Into the structure of the zirconolite was incorporated 10 mol% Sr. Undoped and doped samples were examined by thermal analyses and X-ray diffraction. Addition of ascorbic acid to the sols lowered the firing temperature and promoted formation of the zirconolite phase.
The thermodynamic stability of Th-doped yttrium iron garnet (Y 3 Fe 5 O 12 , YIG) as a possible actinide-bearing material has been investigated using calorimetric measurements and first-principles electronic-structure calculations. Yttrium iron garnet with thorium substitution ranging from 0.04 to 0.07 atoms per formula unit (Y 3Àx Th x Fe 5 O 12 , x ¼ 0.04–0.07) was synthesized using a citrate–nitrate combustion method. High-temperature oxide melt solution calorimetry was used to determine their enthalpy of formation. The thermodynamic analysis demonstrates that, although the substitution enthalpy is slightly endothermic, an entropic driving force for the substitution of Th for Y leads to a near-zero change in the Gibbs free energy. First-principles calculations within the density functional theory (DFT) indicate that the main limiting factors for Th incorporation into the YIG structure are the narrow stability domain of the host YIG and the formation of ThO 2 as a secondary phase. Nevertheless, the defect formation energy calculations suggest that by carefully tuning the atomic and electronic chemical potentials, Th can be incorporated into YIG. The thermodynamic results, as a whole, support the possible use of garnet phases as nuclear waste forms; however, this will require careful consideration of the repository conditions.
The garnet structure is a promising nuclear waste form because it can accommodate various actinide elements. Yttrium iron garnet, Y3Fe5O12 (YIG), is a model composition for such substitutions. Since cerium (Ce) can be considered an analogue of actinide elements such as thorium (Th), plutonium (Pu), and uranium (U), studying the local structure and thermodynamic stability of Ce-substituted YIG (Ce:YIG) can provide insights into the structural and energetic aspects of large ion substitution in garnets. Single phases of YIG with Ce substitution up to 20 mol % (Y3–xCexFe5O12 with 0 ≤ x ≤ 0.2) were synthesized through a citrate–nitrate combustion method. The oxidation state of Ce was examined by X-ray absorption near edge structure spectroscopy (XANES); the oxidation state and site occupancy of iron (Fe) as a function of Ce loading also was monitored by 57Fe–Mössbauer spectroscopy. These measurements establish that Ce is predominantly in the trivalent state at low substitution levels, while a mixture of trivalent and tetravalent states is observed at higher concentrations. Fe was predominately trivalent and exists in multiple environments. High temperature oxide melt solution calorimetry was used to determine the enthalpy of formation of these Ce-substituted YIGs. The thermodynamic analysis demonstrated that, although there is an entropic driving force for the substitution of Ce for Y, the substitution reaction is enthalpically unfavorable. The experimental results are complemented by electronic structure calculations performed within the framework of density functional theory (DFT) with Hubbard-U corrections, which reproduce the observed increase in the tendency for tetravalent Ce to be present with a higher loading of Ce. The DFT+U results suggest that the energetics underlying the formation of tetravalent Ce involve a competition between an unfavorable energy to oxidize Ce and reduce Fe and a favorable contribution due to strain-energy reduction. The structural and thermodynamic findings suggest a strategy to design thermodynamically favorable substitutions of actinides in the garnet system.
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.
One of the major issues related to the expanded use of nuclear power and the development of advanced nuclear fuel cycles is the fate of plutonium and “minor” actinides. In addition, substantial quantities of plutonium and highly enriched uranium from dismantled nuclear weapons now require disposition. There are two basic strategies for the disposition of the actinides: (1) to “burn” or transmute the actinides using nuclear reactors or accelerators; (2) to “sequester” the actinides in chemically durable, radiation-resistant materials that are suitable for geologic disposal. This paper deals with actinide-bearing materials that support the latter approach. During the past two decades, a considerable amount of research and development has been done in an effort to develop matrices for the immobilization of plutonium and the “minor actinides”, Np, Am and Cm. A variety of waste form materials – oxides, silicates and phosphates – have been developed that have a high capacity for the incorporation of actinides, are chemically durable and, in some cases, resistant to the radiation-induced transformation to the aperiodic state. These waste forms can be selected depending on the composition of the waste stream that contains the actinides, the desired materials' properties of the waste form, and the geochemical and hydrologic conditions of the specific repository. The present state-of-knowledge for these materials is such that now one can design materials for very specific conditions, such as the thermal history and accumulated radiation dose, in a repository.
Lanthanide-Borosilicate (LaBS) glass capable to dissolve up to ∼10 wt.% PuO2 is designed for immobilization of plutonium-bearing wastes. The sample of LaBS glass with a target chemical composition (wt %): 9.0 Al2O3, 11.8 B2O3, 12.2 Gd2O3, 6.3 HfO2, 17.2 La2O3, 13.6 Nd2O3, 9.5 PuO2, 18.1 SiO2, 2.3 SrO was prepared from PuO2 powder and mechanically activated mixture of chemicals at 1500 °C. The obtained product was visually homogeneous. Xraydiffraction of the as-prepared glass showed that it mostly consists of a vitreous phase withsmall amounts of crystalline PuO2 (or PuO2-HfO2 solid solution with minor HfO2), britholite andan oxide with a fluorite structure and a composition close to GdHfO1.75. The crystalline fractionincreased after storage for ∼1 year. Magnitude of the FT of EXAFS spectrum at Pu LIII edgeshows that the peak due to first coordination sphere is much more intense than that of the secondshell. This indicates that though some Pu entered crystalline phases (mainly distorted PuO2), itsmajor fraction remained in the vitreous phase. Fit of the spectrum (R-factor = 0.02) gives thefollowing distances: R (Pu-O1)1 = 1.98 (σ = 0.04), R (Pu-O1)2 = 2.18 (σ = 0.04), R (Pu-O1)3= 2.35, R (Pu-Pu) = 3.72 (σ = 0.04), R (Pu-O2) = 4.4 (σ = 0.06). Oxygen environment ofthe Pu4+ ions in the vitreous phase resembles axially squeezed tetragonal pyramid with acoordination number ∼5. The distances ∼2.35 and ∼4.4 correspond to the pairs Pu-O in thefirst and the second shells in the crystalline PuO2. The distance ∼3.72 corresponds to the Pu-Pu and/or Pu-M (M = Ln, Hf) distances. EXAFS spectra of Hf show that Hf is present mostly in thevitreous phase with major neighbors at 2.17 and ∼3.2 A.
Various compounds with fluorite (cubic zirconia) and fluorite-derived (pyrochlore, zirconolite) structures are considered as promising actinide host phases at immobilization of actinide-bearing nuclear wastes. Recently some new cubic compounds — stannate and stannate-zirconate pyrochlores, murataite and related phases, and actinide-bearing garnet structure compounds were proposed as perspective matrices for complex actinide wastes. Zirconate pyrochlore (ideally Gd2Zr2O7) has excellent radiation resistance and high chemical durability but requires high temperatures (at least 1500 °C) to be produced by hot-pressing from sol-gel derived precursor. Partial Sn4+ substitution for Zr4+ reduces production temperature and the compounds REE2ZrSnO7 may be hot-pressed or cold pressed and sintered at ~1400 °C. Pyrochlore, A2B2O7−x (two-fold elementary fluorite unit cell), and murataite, A3B6C2O20−y (three-fold fluorite unit cell), are end-members of the polysomatic series consisting of the phases whose structures are built from alternating pyrochlore and murataite blocks (nano-sized modules) with seven- (2C/3C/2C), five- (2C/3C), eight- (3C/2C/3C) and three-fold (3C — murataite) fluorite unit cells. Actinide content in this series reduces in the row: 2C (pyrochlore) > 7C > 5C > 8C > 3C (murataite). Due to congruent melting murataite-based ceramics may be produced by melting and the firstly segregated phase at melt crystallization is that with the highest fraction of the pyrochlore modules in its structure. The melts containing up to 10 wt. % AnO2 (An = Th, U, Np, Pu) or REE/An fraction of HLW form at crystallization zoned grains composed sequentially of the 5C → 8C → 3C phases with the highest actinide concentration in the core and the lowest — in the rim of the grains. Radiation resistance of the "murataite" is comparable to titanate pyrochlores. One more promising actinide hosts are ferrites with garnet structure. The matrices containing sometime complex fluorite structure oxide as an extra phase have leach and radiation resistance similar to the other well-known actinide waste forms.
Over the past few decades, many studies of actinides in glasses and ceramics have been conducted that have contributed substantially
to the increased understanding of actinide incorporation in solids and radiation effects due to actinide decay. These studies
have included fundamental research on actinides in solids and applied research and development related to the immobilization
of the high level wastes (HLW) from commercial nuclear power plants and processing of nuclear weapons materials, environmental
restoration in the nuclear weapons complex, and the immobilization of weapons-grade plutonium as a result of disarmament activities.
Thus, the immobilization of actinides has become a pressing issue for the twenty-first century (Ewing, 1999), and plutonium
immobilization, in particular, has received considerable attention in the USA (Muller et al., 2002; Muller and Weber, 2001).
The investigation of actinides and
Recycling plutonium is dangerous and costly. Britain should take the
lead on direct disposal, say Frank von Hippel, Rodney Ewing, Richard
Garwin and Allison Macfarlane.
Comprehensive analysis of 'murataites' compositions and structures derived using optical and electron microscopy, including high resolution, X-ray diffraction, Mossbauer techniques and their comparison with that of pyrochlore allow concluding the similarity of model and experimentally observed phase compositions. It is also concluded that 'murataites' represent one more group of subtraction derivatives in fluorite-like structure family and can be regarded as modular members of polysomatic series, in which ultimate members are pyrochlore and murataite Mu-3.
The elemental and ionic quantitative compositions and oxidation states of elements in the ceramics CaCe0.9Ti2O6.8 (I) and CaCeTi2O7 (II) - matrices for long-term storage of long-lived radionuclides constituting high-level radioactive wastes - are studied by X-ray photoelectron spectroscopy. Compound I, obtained at a pressure of 300 MPa and T = 1400°C in air, has two types of surface cerium ions, 63 at. % Ce3+ and 37 at. % Ce4+; compound II, fabricated at a pressure of 300 MPa and T = 1300°C in an oxygen atmosphere, also has two types of surface ions, 36 at. % Ce3+ and 64 at. % Ce4+, which agrees satisfactorily with X-ray powder diffraction and scanning electron microscopy data. On exposure to environmental conditions, calcium and/or cerium carbonates can form on the surface of the ceramics under consideration, which entails their failure.
The use of confinement matrices is a key part of safe management of high-level radioactive wastes (HLWs) derived in the nuclear fuel cycle. The matrices should immobilize radioisotopes after HLW deposition in the geological environment with possible groundwater filtration. The glasses currently used for this purpose on an industrial scale are not capable of incorporating sufficient amounts of plutonium and have low stability to chemical corrosion. This paper summarizes the results of structural analysis of crystalline phases that could be used for immobilization of actinide wastes of various compositions. It was suggested that pyrochlore-type phases can be used for incorporation of the actinide-zirconium-rare-earth element fraction of HLWs, while ferrites with a garnet structure could be used for immobilization of wastes of complex composition with high contents of corrosion products (Fe, Al, Ga). Ceramics of such composition were synthesized and analyzed for concentrations of actinides (Th, U), rare-earth elements (Gd, Ce), and Zr. It is necessary to study the stability of these phases to radiation and chemical corrosion to select suitable matrix materials.
Structural damage to actinide-bearing matrices upon their irradiation with Kr and Xe ions with energies of 1.0 and 1.5 MeV was studied. Actinides are incorporated into oxides with fluorite-type structures (zirconolite, pyrochlore, and murataite), uranium titanate (brannerite), silicates and ferrites with a garnet lattice, and a (Ca, REE) silicate with an apatite structure (britholite). The radiation doses for complete structural amorphization at 25°C were as follows (in units of 1014 ions/cm2): 2.9 for zirconolite, 1.8-2.4 for pyrochlore, 1.5-2 for garnet, 1.7-1.9 for murataite, 1.4 for brannerite, and 0.4 for britholite. The radiation resistance of phases expressed as the number of displacements per atom (dpa) ranges from 0.1 to 0.4 dpa. According to these data, a phase containing 10 Wt % 239Pu will be completely amorphized over 500-2000 yr. This will increase actinide leaching from the matrix by tens of times. Amorphization doses are several times higher for natural analogues because of healing of radiation damage to the mineral structure with time. Disposal of highly radioactive waste matrices in deep-well repositories with an elevated temperature of ambient rocks favors an increase in the resistance of the crystal structure to radiation and maintains the immobilizing properties of radionuclide-bearing matrices over longer periods.
The crystal structure of murataite, space group F43m, a 14.886(2) Å, has been solved by Patterson methods and refined to an R of 4.91 (wR 4.71 %) for 265 observed reflections (MoKα radiation). The ideal formula is (Y,Na)6(Zn,Fe)5Ti12O29(O,F)10F4 with Z = 4, but its simple appearance conceals extensive cation disorder within the structure. There are four distinct cation sites: the X site is [8]-coordinated and contains (Y,HREE,Na,Ca,Mn); the T site is tetrahedrally coordinated and contains (Zn,Si); the M1 site is octahedrally coordinated and contains (Ti,Nb,Na); the M2 site is [5]-coordinated by a triangular bipyramid and contains (Zn,Fe3+,Ti,Na). Three M1 octahedra share edges to form a compact M3φ13 group. Four of these M3φ13 groups link by sharing comers to form a tetrahedral cage in the center of which is the T site. The resulting Keggin-structured [M12Tφ40]n- unit may be considered as the fundamental building block of the structure. The net formed by linkage of the M1 polyhedra is topologically the same as the B net of UB12, which accounts for the similarity of X-ray properties of murataite and this compound.
The heat treatment conditions are a key factor in fabricating zirconolite ceramics and glass-ceramics following high-temperature melting. An oxide mixture melted at 1450 C and subsequently heat-treated at 1200 degreesC yielded a glass-ceramic containing crystallized zirconolite-2M. The silica-enriched residual glass represented about 60-70 vol% of the total; the actinide surrogates (Nd, Ce) were equally distributed between the residual glass and the zirconolite crystals. Zirconolite ceramics obtained after melting an oxide mixture at 1600-1700 degreesC consisted of zirconolite, perovskite and rutile. Rapid cooling rates (> 100degreesC(.)min(-1)) were obtained by pouring the melt into ingot molds; the resulting zirconolite ceramics were characterized by crystals of zirconolite-2M ranging from 1 to no more than 20 mum. Slow cooling (< 25 degreesC(.)min(-1)) produced ceramics with crystals several hundred micrometers long. Despite the microstructural differences, the chemical durability of the zirconolite ceramics was identical. The initial alteration rates r(0) were about two orders of magnitude lower than those measured for the residual aluminosilicate glass of the zirconolite glass-ceramics. Moreover, during long-term leach tests at high S/V ratios to obtain advanced degrees of reaction progress, the alteration rates of all the materials diminished by over 3 to 4 orders of magnitude below r(0).
Zirconolite-glass and sphene-glass specimens, doped with REE as simulants for trivalent actinides, were leached under two conditions. The first was a solubility test using a powdered sample in deionized water at an SA/V ratio of 200 cm(-1), to examine the long-term leaching behavior of the composite materials. The other test was carried out in the presence of moist clay, to assess the degree of surface alteration of the composites in the presence of potential geological repository materials. Both tests were carried out at 90 degreesC. The specimens leached in clay showed signs of preferential attack on the glassy matrix along zirconolite and sphene grain boundaries. EDS results showed no gross changes in composition of the constituent phases as a result of leaching. For the solubility tests, steady state conditions of elemental release were attained within 7 days of leaching, suggesting development of a surface passivation layer hindering movement of reactive species between the surface of the material and the leachant. Calcium, Si and Al releases were similar between composites. Titanium and Cc releases were also similar between composites, and were two orders of magnitude lower than those for Ca, Si and Al. The actinide simulants partitioned into the glass phase and into the crystalline component of the composite materials in approximately similar proportions. Although the surrogates were contained in the less durable glassy phase at these levels, this was not reflected in the release of Cc, for example, which was similar to that for a single-phase zirconolite.
Neodymium-doped zirconolite materials may be synthesized by two melting processes. One involves devitrification of an aluminosilicate parent glass containing titanium, zirconium and neodymium oxides, yielding a glass ceramic comprised of submicron zirconolite needles embedded in a silica-rich glass matrix. The second method consists of melting an oxide mixture with the stoichiometry of a highly Nd-enriched zirconolite, then quickly cooling the melt to produce a ceramic rich in zirconolite crystals several hundred microns long containing a large fraction of the initial neodymium.
The disposal of fission products and actinides generated by the nuclear-fuel cycle is one of the major challenges in Environ- mental Sciences of the 21st Century. Because some fission products (e.g., 99Tc, 129I, 79Se and 135Cs) and actinides (e.g., 239Pu and 237Np) are long-lived, they have a major impact on the risk assessment of geological repositories. Thus, demonstrable long-term chemical and mechanical durability are essential properties of waste forms for the immobilization and disposal of radionuclides. Mineralogical and geological studies provide excellent candidate phases for immobilization and a unique database that cannot be duplicated by a purely Materials Science approach. The "mineralogical approach" is illustrated by a discussion of zircon as a phase for the immobilization of plutonium from dismantled nuclear weapons. Other minerals, e.g., monazite, apatite, pyrochlore, zirconolite and zeolites, also are important candidates for the immobilization of actinides.
A comprehensive understanding of radiation effects in zircon, ZrSiO4, over a broad range of time scales (0.5 h to 570 million years) has been obtained by a study of natural zircon, Pu-doped zircon, and ion-beam irradiated zircon. Radiation damage in zircon results in the simultaneous accumulation of both point defects and amorphous regions. The amorphization process is consistent with models based on the multiple overlap of particle tracks, suggesting that amorphization occurs as a result of a critical defect concentration. The amorphization dose increases with temperature in two stages (below 300 K and above 473 K) and is nearly independent of the damage source (α-decay events or heavy-ion beams) at 300 K. Recrystallization of completely amorphous zircon occurs above 1300 K and is a two-step process that involves the initial formation of pseudo-cubic ZrO2.
Single-pass flow-through (SPFT) experiments were conducted on a set of non-radioactive Ti-based ceramics at 90 °C and pH =
2 to 12. The specimens contained 27.9 to 35.8 wt% CeO2 as a surrogate for UO2 and PuO2. Compositions were formulated as TiO2-saturated pyrochlore (CeP1) and pyrochlore-rich baseline (CePB1) ceramic waste forms. Pyrochlore + Hf-rutile and pyrochlore
+ perovskite + Hf-rutile constituted the major phases in the CeP1 and CePB1 ceramics, respectively. Results from dissolution
experiments between pH = 2 to 12 indicate a shallow pH-dependence with an ill-defined minimum. Element release rates determined
from experiments over a range of sample surface areas (S) and flow rates (q) indicate that dissolution rates become independent of q/S values at 10−8 to 10−7 m/s. Dissolution rates dropped sharply at lower values of q/S, indicating rates that are subject to solution saturation effects as dissolved constituents become concentrated. Forward
dissolution rates were 1.3(0.30) × 10−3 and 5.5(1.3) × 10−3 g/m2·d for CeP1 and CePB1 ceramics, respectively. Dissolution rates obtained in other laboratories compare well to the findings
of this study, once the rate data are placed in the context of solution saturation state. These results make progress toward
an evaluation of CeO2 as a surrogate for UO2 and PuO2 as well as establishing a baseline for comparison with radiation-damaged specimens.
Synthetic ferrites with garnet-type structures were studied as possible matrices for immobilization of highly radioactive wastes (HLWs). Unlike other promising matrices, such as pyrochlore, monazite, britholite, etc., natural garnets contain no radioactive elements. Thus, the properties of garnet-type matrices, e.g., their capacity to accommodate HLW components, should be studied using synthetic compounds. According to similarity in degree of oxidation and ionic radii, we chose Th4+ and Ce 4+ as imitators of tetravalent actinides (U, Np, and Pu) and Gd3+ as an imitator of trivalent actinides (Am and Cm); Gd3+, together with La3+, composes the REE component of fractionated HLWs. The compositions of the studied samples corresponded to the following formulas: [(Ca1.5-x NaxGdTh0.5) (ZrFe) (Fe3-xSix,) O12] and [(Ca1.5-xNax, (Gd, La)Ce0.5) (ZrFe) (Fe3-xSix) O12], where x = 0.25, 0.5, and 0.75. Our study was aimed at revealing effects of Na2O and SiO2 contents and synthesis temperature (in a system with ThO2) on the phase composition of samples and capacity of the target phases to incorporate REE and actinides. It was found that the increase in Na2O and SiO2 contents in the starting material led to some decrease of ThO2 and CeO2 contents in garnets, while Gd2O3 and La2O3 contents slightly increased. The temperature increase resulted an increase of SiO2 contents in garnets and consequent decrease of ThO2 Solubility in these phases by several weight percent. Our study demonstrated that it is possible to synthesize ferrite garnets with significant amounts of actinides, REE, and admixture elements. Thus, these compounds may be promising matrices for immobilization of complex actinide-bearing HLW.
Processes of phases formation in the ceramic mixtures with basic compositions (wt.%) 10 CaO, 10 MnO, 5 Al 2 O 3 , 5 Fe 2 O 3 , 55 TiO 2 , 5 ZrO 2 , 10 UO 2 (M1) and 8 CaO, 8 MnO, 4 Al 2 O 3 , 4 Fe 2 O 3 , 20 Gd 2 O 3 , 44 TiO 2 , 4 ZrO 2 , 8 UO 2 (M4) were studied using X-ray diffraction, scanning and transmission electron microscopy. The batches were milled, compacted in pellets at 200 MPa, and heat-treated in a resistive furnace at 1100 °C, 1300 °C, 1400 °C, and 1500 °C as well as melted in a cold crucible at ∼1600 °C. Reactions in the mixtures heat-treated at 1100 °C were not completed and samples contained significant amount of unreacted and intermediate (altered rutile, cubic oxide solid solution, perovskite) phases. Within the temperature range 1100–1300 °C reactions are mainly completed and ceramics sintered at 1300 °C are composed of major murataite and minor rutile (M1) or major murataite and pyrochlore and minor zirconolite and perovskite (M4). However full homogenization at 1300 °C has not been reached yet and to obtain the ceramics with uniform compositions of the phases sintering at 1400 °C or melting at 1500–1600 °C were required.
In the ceramic sample M1 two murataite varieties with five- (murataite-5C) and eight-fold (mu-rataite-8C) fluorite-type unit cells were found. The sample M4 is composed of pyrochlore, murataite-8C and zirconolite-3O. In the sample M1 murataite-5C is enriched with U and Ca and depleted with iron group elements as compared to murataite-8C. Fraction of murataite-5C concentrates about 80% of total U and about 70% of Mn+Fe (corrosion products). Waste elements partitioning among the phases in the M4 sample depends significantly on temperature of heat-treatment.
Scientific basis for nuclear waste management XXI
Breaking rock rolling cutters are cutters of breaking rock with rolling in the way of the push of the cutterhead, they are used widespread in the full face rock tunnel boring machine. On the analysis of the motional characteristics of disc cutters, cutters' wear and tear, broken rocks, it is shown that the rock-cut must exist and which would come from the side-force. In order to calculate this unbalance side-force, the mechanics model of the rolling breaking rock cutters is put out, the formulas for calculating this unbalance side-force is give out and its correctness are tested.
Perovskite-based ceramics ABO3 (A = La or Gd; B = Al or Fe) and a pseudobinary system: LaAlO3 + CaZrTi2O7 are promising matrices for immobilization of actinide fraction of HLW. The ceramic samples containing 241Am, 238Pu or 147Pm were prepared by cold pressing in pellets at 100-300 MPa and sintering at 1300-1500 °C. Leach rates of radionuclides and matrix elements (La, Gd, Al and Fe) from powders were measured using Soxhlet unit. Sintered ceramics in the series: LaAl1-xFexO3 were composed of perovskite phase. Ceramics in the compositional series xLaAlO3 + CaZrTi 2O7 (0.4 ≤ x ≤ 6) consisted of perovskite, zirconolite, and baddeleyite at x = 0.4 and x = 0.7, and perovskite and baddeleyite at higher x values. Leach rate of radionuclides and matrix elements from all the ceramics were lower than 10-8 g·cm-2 day.
A transmission electron microscope investigation was made of zirconolites and perovskites irradiated to amorphization with 1 MeV krypton ions using the HVEM-Tandem Facility at Argonne National Laboratory. Three specimens were examined - a prototype zirconolite CaZrTi2O7, a gadolinium doped zirconolite Ca0.75Gd0.50Zr0.75Ti2O7 and a uranium doped zirconolite Ca0.75U0.50Zr0.75Ti2O7. The critical amorphization dose Dc was determined at several temperatures between 20 K to 675 K. Dc was inversely proportional with temperature. For example, pure zirconolite requiring 10x the dose for amorphization at 475 K compared with gadolinium zirconolite. Using an Arrhenius plot, the activation energy Ea for annealing in these compounds was found to be 0.129 eV and 0.067 eV respectively. The greater ease of amorphization for the gadolinium sample is probably a reflection of this element's large cross section for interaction with heavy ions. Uranium zirconolite was very susceptible to damage and amorphized under 4 keV argon ions during the preparation of microscope specimens. In each sample, zirconolite coexisted with minor perovskite, reduced rutile (Magneli phases) and zirconia. These phases were more resistant to ion irradiation than zirconolite. Even for high gadolinium loadings, perovskite (Ca0.8Gd0.2TiO3) was 3-4 times more stable to ion irradiation than the surrounding zirconolite crystals.
Zirconolite (CaZrTi2O7) is a rare accessory mineral crystallizing under different geological conditions and in a wide range of generally SiO2-poor rock types. In this paper, we present results obtained from a statistical study of nearly 300 chemical analyses of natural zirconolite. The chemical composition of natural zirconolite often deviates significantly from its theoretical composition due to extensive substitutions involving more than 30 chemical elements which vary widely in size (from 0.40 to 1.14 Å) and charge (from 2+ to 6+). These data document a high degree of compositional flexibility in zirconolite, and imply that availability of specific elements is a key factor in determining the chemical composition of zirconolite. The composition is strongly dependent on and reflects the chemical characteristics of the geological environment: zirconolite from lunar basalts is generally poor in Nb, Ta, U and Th, but rich in Fe, Zr, Hf and rare earth elements (ΣREE2O3 up to 32 wt%); zirconolite in metasomatically altered carbonate rocks exhibits a large variation in the content of U, Th and REE, but is generally low in Nb and Ta; and zirconolite from carbonatites is poor in Al and Mg, but extremely variable with respect to its Nb, Ta, REE, U and Th contents. Distinct compositional trends are observed for individual geological localities and are attributed to chemical variations resulting from pronounced continuous and discontinuous zoning within individual crystals. The results show that the chemical composition of natural zirconolite can be simplified and satisfactorily described in terms of 9 components: Ca, Zr, Ti, Me5+ (Nb and Ta), Me3+ (mainly Al and Fe3+), Me2+ (primarily Mg and Fe2+), REE, ACT (U4+ and Th4+), and O. In metasomatically altered carbonate rocks, most of the compositional variation can be expressed by the substitutions ACT + Me2+ ??? Ca + Ti and REE + Me3+ ??? Ca + Ti; the statistical analysis further indicates that both Fe2+ and Fe3+ are often present. In carbonatites, the most important substitution is Me5+ + Me3+ ??? 2 Ti, and Fe appears to be present predominantly as Fe3+. The chemical data further suggest that Ca may be replaced by a limited amount of Mn2+ or Mg, and possibly by Zr. Although many substitutions are theoretically possible, only a few are geologically important. The results suggest that the redox conditions during crystallization exert a significant control on the substitution mechanisms and possibly on the total amounts of ACT and REE accommodated by zirconolite. The highest reported amounts of ACT (0.475 ACT per formula unit), in particular, were incorporated by a substitution that requires the presence of a charge-balancing bivalent cation (ACT + Me2+ ??? Ca + Ti) and was thus influenced by the relative abundances of Fe2+ and Fe3+. The limited number of important substitutions observed in natural samples indicates that these mechanisms produce zirconolite crystals with energetically favorable compositions and a generally high durability, implying that synthetic zirconolite designed for immobilization of high-level nuclear waste (e.g., in Synroc) should be tailored in a similar way.
Synroc, a ceramic made from a reactive mixture of Al, Ba, Ca, Ti, and Zr oxides, is proving to be a suitable and effective medium for immobilizing nuclear wastes.
Synroc-C, a titanate-based ceramic variant, was initially developed in 1978 by Ringwood et al. for immobilizing high-level nuclear waste (HLW) from nuclear power reactor fuel reprocessing. HLW is essentially a solution of radioactive fission products, actinides, and process contaminants in ~3 mol/L nitric acid. The developers of Synroc-C aimed to immobilize radioactive waste ions by incorporating them in a ceramic. They accomplished this by mixing the HLW solution (liquid waste) with a ceramic precursor, then forming the ceramic by drying, calcining, and hot-pressing the mixture in a metal container for two hours at 1200°C/20 MPa. The result, Synroc-C, is composed of hol-landite, zirconolite, perovskite, and rutile, together with a few percent of minor phases and metal alloys. The Synroc-C precursor has the following composition (wt%): Al 2 0 3 (5.4); BaO(5.6); CaO(11); TiO 2 (71.4); and ZrO 2 (6.6). Since 1984, it has been prepared by hydrolyzing a mixture of Al, Ti, and Zr alkoxides with an aqueous slurry of Ba and Ca hydroxide. The abundances of the phases, and the radionuclides contained in them in dilute solid solution, are identified in Table I.
For the HLW fractions of certain Hanford tank wastes, wasteforms consisting mainly of crystalline Synroc phases in an aluminosilicate matrix can be formed by melting at 1350-1450°C in a neutral atmosphere, with waste loadings of up to 70 wt%, depending on the waste composition. The leachability of all measured elements in 7-day PCT tests was <0.1 g/m 2d at 90°C in the best materials.
An alternative description of zirconolite, zirkelite, pyrochlore and polymignyte as modular structures is proposed. This approach not only emphasizes their crystallographic similarities but is also readily extended to include defects observed by high resolution electron microscopy. The interest in zirconolite stems from its inclusion in SYNROC. The means by which alpha -emitting radwaste is incorporated in the ceramic will influence the process of metamictization.
The aeschynite structure-type (Ce,Nd,La,Th,U,Ca)(Nb,Ti)2O6, and the rare-earth silicate apatite structure-type with the formula (Ce,La,Nd,Ca,Th)10(SiO4,PO4)6(O,F,OH)2 are important rare-earth and actinide host phases for high-level nuclear waste. Natural phases of these structure-types have calculated alpha-decay doses up to approximately 1017 α-events/mg which have accumulated over hundreds of millions of years. Transmission electron microscopy has been used to study the microstructure of α-decay damage in aeschynite and britholite. Electron diffraction analysis of natural aeschynite revealed that minerals originally crystalline gradually lost their crystallinity with increasing alpha-decay doses. Helium bubbles were found in the aeschynite which have accumulated up to approximately 2×1016 α-events/mg. These bubbles may nucleate within collision cascades during α-decay damage. Electron irradiation has an enhanced rare-gas migration and the formation of larger bubbles. High-resolution electron microscopy (HRTEM) revealed that amorphization during accumulation of α-decay damage was from alpha-recoil nuclei collision cascades, in both the aeschynite and britholite.
The possibility of production of mineral-like composites on the base of zirkonolite using the technique of self-propagating high temperature synthesis is experimentally shown. X-ray analysis of these compositions containing the nuclear wastes simulators was carried out. The thermodynamic calculation of adiabatic combustion temperatures was performed.
The valence states of 238U, 237Np, 239,Pu, and 241 Am have been studied over a period of time after initial incorporation into a lowtitanium borosilicate glass. Under the air-firing conditions used, only the following valence states were observed: U(VI), Np(IV), Pu(IV), and Am (III). Over a period of one year, no actinide valence changes were observed. Preliminary EXAFS studies on uranium-and thorium-doped glasses indicate ordering in the first coordination shell of the actinide ions, with apparent local structure resembling that existing in model crystalline compounds.
Aqueous durability has been assessed for fourteen pyrochlore- and zirconolite-rich titanate waste forms designed for the immobilisation of excess Pu. The ceramics used in this study contained about 12 wt% Pu and about 15 wt% of Hf and Gd oxides as neutron absorbers and were fabricated by cold-pressing and sintering or hot isostatic pressing.
Total release rates (i.e. unfiltered solution + vessel wall inventory) have been measured withthe MCC-1 test method at 90 °C in deionised water. For all samples, 7-day release rates of Pu are between 4 × 10−5 and about 10−3 g.m−2.d−1, reducing to between about 8 × 10−6 and 3 × 10−5 g.m−2.d−1after more than 300 days of leaching. Release rates of U from the baseline ceramic were found to decrease to values between 6 × 10−4 and 1 × 10−5 g.m−2.d−1after 200 days. Hf leach rates were generally < 5 × 10−6 g.m−2.d−l after more than 7 days. Further, the addition of several % of chemical impurities to the baseline sintered formulations was found to reduce U and Gd releases by about a factor of 10 after 200 days.
Zirconolite (CaZrTi2O7) and perovskite (CaTiO3) are key minerals in SYNROC, a ceramic material developed for the immobilization of high level nuclear reactor wastes. When these are incorporated in SYNROC, the long-lived radioactive actinide elements are preferentially partitioned into zirconolite and perovskite which are therefore subjected to the effects of alpha-recoil, resulting from the decay of these elements. These effects have been studied via X-ray and electron diffraction investigations of natural samples of zirconolite and perovskite of varying ages and varying uranium and thorium contents. The samples studied have received cumulative alpha doses ranging from 1.0 × 10¹⁸ to 1.1 × 10²⁰α/g. The upper limit corresponds to the alpha irradiation which would be received by the zirconolite in SYNROC containing 10 percent of high level waste over a period of 5 × 10⁸years. These studies show that zirconolites remain crystalline up to and beyond alpha doses of 2 × 10¹⁹α/g. This dose would have accumulated in such a SYNROC zirconolite after a million years of storage. Electron microscopy revealed that the grains were composed of small crystalline domains which possessed the defect fluorite-type structure. After a dose exceeding that which would be received by SYNROC in 100 million years, zirconolites appeared metamict when studied by X-ray diffraction. However, the electron micrographs and diffraction patterns clearly demonstrate that the mineral continues to retain a large degree of short range order and in no way resembles a glass. The density changes produced in these zirconolites by irradiation are small and range from 0 to 3% at saturation. Perovskite samples which have SYNROC ages up to 20,000 years decrease in density by 1.8 ± 0.1%. Their X-ray powder patterns are essentially unaffected. Comparative studies show that the perovskite lattice is even more resistant to the effects of alpha-recoil than the zirconolite lattice. The results demonstrate that zirconolite and perovskite are extremely resistant to the effects of nuclear radiation and will provide stable crystal structures for the containment of the radioactive waste elements during the time required for the radioactivity to decay to safe levels (typically 10⁵–10⁶years).
The thorium phosphate-diphosphate (TPD) appears as a good host matrix
for the long-term storage of radioactive waste, for its ability to
incorporate actinides (Th, U, Pu, Np) and its chemical and thermal
stability. The TPD was held at 50 MPa in hydrothermal conditions under a
thermal gradient ( T=400-320 °C) in the presence of cement and a
clay (FoCa-7, compacted benthonite), in order to evaluate the TPD
reactivity. After two months, TPD has partly reacted with the cement, to
form hydroxylapatite, Ca 5(PO 4) 3(OH)
and thorianite, ThO 2. This result when extrapolated to
natural systems could explain why TPD has no natural equivalent. To cite
this article: B. Goffé et al., C. R. Geoscience 334 (2002)
1047-1052.
Specimens of Gd 2Ti 2O 7 and CaZrTi
2O 7 were doped with 244Cm and the
effects of self-radiation damage from alpha decay were determined as a
function of cumulative dose. The macroscopic swelling of the specimens
increased exponentially with dose to limiting (saturation) values of 5.1
and 6.0% for Gd 2Ti 2O 7 and CaZrTi
2O 7, respectively. The radiation-induced
microstructure consists primarily of individual amorphous tracks from
the alpha-recoil particles which eventually overlap to produce an
amorphous state at ˜ 2.0 × 10 25 alpha decays/m
3. This radiation-induced swelling and amorphization results
in increased dissolution rate and fracture toughness, but decreased
hardness. Both materials recrystallize in a sharp recovery stage. The
stored energy release is ˜127 J/g and the activation energy for
recrystallization in CaZrTi 2O 7 is estimated to
be 5.8 eV.
The liquidus temperatures, TL, of rare earth-alumino-borosilicate glasses were measured as functions of glass composition. The TL values ranged from 1153?C to 1405?C. Three primary crystalline phases were identified in the study. The most frequently encountered was a rare earth silicate phase in which the TL values ranged from 1153?C to 1405?C. Al2O3 was encountered in glasses with an Al2O3:SiO2 mass ratio greater than 1, with TL values ranging from 1242?C to 1305?C. Alumino-silicate crystals were encountered as the primary phase in glasses with less than 43 mass% of mixed rare earth oxides (Ln2O3) with TL values between 1164?C and 1255?C. A linear relationship between total mixed rare earth oxide concentration and TL was found within all three primary phase fields. In the rare earth silicate primary phase field, normalizing the Ln2O3 with its mean ionic radius enhanced this linear relationship.
The detailed structure investigations of Zr2Gd2O7 with pyrochlore- and fluorite-type structures have been carried out by X-ray single crystal method at room temperature. The bond distances between cations and oxide ions versus the site population for cations reasonably support that the fluorite phase is made up of the structure with microdomains of the pyrochlore. Sharp fundamental and diffuse superstructure reflections of the pyrochlore are interpreted assuming that anti-phase domain boundaries lie parallel to the &{211} and that domain sizes are not uniform. It should be proposed that the layers composed of the 48f site oxide ions are anti-phase boundaries. The probability density function maps and effective one-particle potentials of the pyrochlore structure reveal that the magnitudes of anharmonic thermal motions for the 48f site oxide ions are large toward the unoccupied 8b site. Compared to the s-pyrochlore phase (with sharp superstructure reflections), the 48f site oxide ions of the d-pyrochlore (with diffuse superstructure reflections) and fluorite phases have larger temperature factors and gentler potential curves. This is due to that the anti-phase domains are coherent each other and that their electron density distributions are the averages of individual domains.
Borosilicate glasses loaded with â10 wt % plutonium were found to produce plutonium-silicate alteration phases upon aqueous corrosion under a range of conditions. The phases observed were generally rich in lanthanide (Ln) elements and were related to the lanthanide orthosilicate phases of the monoclinic LnâSiOâ type. The composition of the phases was variable regarding [Ln]/[Pu] ratio, depending upon type of corrosion test and on the location within the alteration layer. The formation of these phases likely has implications for the incorporation of plutonium into silicate alteration phases during corrosion of titanate ceramics, high-level waste glasses, and spent nuclear fuel.
Over 200 chemical analyses of microlite determined by electron microprobe are reported, and they are consistent with the accepted structural formula. A/sub 2-m/BâOâY/sub 1-n/ pHâO, where principally A = Ca, Na, U, Mn; B = Ta, Nb, Ti; X = O; and Y = F, OH, O. General features of microlite crystal chemistry identified include (1) a positive correlation between A-site vacancies and the maximum Y-site vacancies, (2) a positive correlation between Na and F, and (3) a negative correlation between U and F. The latter is consistent with the interpretation that U at the A site in the pyrochlore structure is analogous to the uranyl group, UOâ/sup 2 +/, requiring O in place of F at the Y site. Microlite compositions from four of five lithologic units examined are chemically distinct in terms of linear combinations of the variables U, Fe, Ti, Bi, Ca, Ce, Pb, F, Mn, Ba, Sb, Th, Ta, and Na. The effects of alpha-recoil damage due to the decay of constituent U have been examined. Because of the wide variation in U content (0.1-10 wt% UOâ), the microlites exhibit the full range of structural periodicity from completely crystalline (20 dpa). Based on X-ray and electron diffraction analysis, the progressive structural modification of microlite with increasing alpha dose involves (1) formation of isolated defect aggregates (i.e., individual alpha-recoil tracks) up through doses of 10¹ⴠalphas/mg with no detectable effect on the materials' ability to diffract X-rays or electrons, (2) continued damage and overlap of these defect aggregates yielding coexisting regions of amorphous and crystalline domains at 10¹ⵠto 10¹ⶠalphas/mg, and (3) amorphization at doses greater than 10¹ⷠalphas/mg.
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Monazite, monoclinic CePO4, is an often used phase in geologic age dating because of its high U content and has more recently been proposed to be host phase for the immobilization of nuclear waste. A naturally occurring monazite from Petaca, New Mexico, was irradiated by 1.5 MeV Kr+ ions over temperatures ranging from 30 to 480 K. The calculated critical temperature for amorphization is 428 K. A single stage annealing process is attributed to epitaxial recrystallization due to radiation-enhanced defect mobility (activation energy = 0.064 eV). The response of monazite to irradiation is compared to that of structurally and chemically related minerals: zircon (ZrSiO4), fluorapatite [Ca5(PO4)3F], and berlinite (AlPO4), having critical temperatures of amorphization of 1101, 475, and 650 K, respectively.
Estimates are given of the total quantities of radioactivity, and of the contribution from different isotopes to this total, arising in the wastes from civil nuclear power generation; the figures are normalized to 1 GW(e)y of power production. The intensity of the heat and gamma -radiation emitted by the spent fuel has been calculated, and their decrease as the radioactivity decays. Reprocessing the spent fuel results in 95% or more of the fission products and higher actinides being concentrated in a small volume of high-level, heat-emitting waste. The total decay curve of unreprocessed spent fuel or of the separated high-level waste is dominated by the decay of some fission products for a few hundred years and then by the decay of some actinide isotopes for some tens of thousands of years. The residual activity is compared with that of the original uranium ore. Some of the long-lived activity will appear in other waste streams, particularly on the fuel cladding, and the volumes and activities of these wastes arising in this country are recorded. The long-lived activity arising from reactor decommissioning will be small compared with the annual arisings from the fuel cycle.
The solubility of synthetic NdPO4 monazite end-member has been determined experimentally from 21 to 300°C in aqueous solutions at pH = 2, and at 21°C and pH = 2 for GdPO4. Measurements were performed in batch reactors, with regular solution sampling for pH measurement, rare earths and phosphorous analysis by inductively coupled plasma mass spectrometry (ICP-MS) coupled with a desolvation system. Scanning electron microscopy (SEM) and X-ray photoelectron spectroscopy (XPS) were employed to check that no reprecipitation of secondary phases occurred and that the mineral surfaces remained those of a monazite. Coupled with speciation calculations, measured solution compositions permitted the determination of NdPO4 and GdPO4 solubility products which are in general agreement with previous experimental determination on rhabdophane at 25°C, but showing that monazite is more than two orders of magnitude less soluble than inferred on the basis of previous thermodynamic estimates. The temperature evolution from 21 to 300°C of the equilibrium constant (K) of the NdPO4 monazite end-member dissolution reaction given by:
### History and applications
The earliest description of a change in mineral property caused by radiation-damage was by Jons Jacob Berzelius in 1814. Berzelius, a Swedish physician and mineral chemist, was certainly one of the most eminent scientists of his century, discovering the elements cerium, selenium and thorium. He discovered that some U- and Th-bearing minerals glowed on moderate heating, releasing, sometimes violently, large amounts of energy. This “pyrognomic” behavior was first observed in gadolinite, (REE)2FeBe2Si2O10, as it released stored energy. Today, this same type of catastrophic energy release is a significant concern in reactor safety where rapid energy release from neutron-irradiated graphite can lead to a rise in temperature and the rupture of fuel elements.
In 1893, Brogger defined “metamikte” in an encyclopedia entry for amorphous materials. Metamict minerals were believed to have been previously crystalline, as judged by well-formed crystal faces, but had a characteristic glass-like, conchoidal fracture and were optically isotropic. Prior to the discovery of radioactivity by Becquerel in 1896, the metamict state was not recognized as a radiation-induced transformation. Hamberg (1914) was the first to suggest that metamictization is a radiation-induced, periodic-to-aperiodic transformation caused by α-particles that originate from the constituent radionuclides in the uranium and thorium decay-series. A detailed summary of the history of studies of radiation effects in minerals can be found in Ewing (1994).
In the 1950s, interest in metamict minerals was revived. Adolf Pabst rescued the term “metamict” from obscurity in his presidential address to the Mineralogical Society of America in 1951 (Pabst 1952). However, mineralogists paid limited attention to nuclear effects in minerals or the general importance of radiation effects in modifying the properties of materials. By the early 1940s, E.P. Wigner already had anticipated that the intense neutron flux in the Hanford Production reactors …
Most countries intend to dispose of their high-level radioactive wastes by converting them into a solidified wasteform which is to be buried within the earth. SYNROC is a titanate ceramic wasteform which has been designed for this purpose on the basis of geochemical principles. It comprises essentially rutile TiO2, 'hollandite' Ba(A1,Ti)Ti6016 , zirconolite CaZrTi2OT, and perovskite CaTiO a. The latter three phases have the capacity to accept the great majority of radioactive elements occur- ring in high-level wastes into their crystal lattice sites. These minerals (or their close relatives) also occur in nature, where they have demonstrated their capacity to survive for many millions of years in a wide range of geological environments. The properties of SYNROC and the crystal chemistry of its constituent minerals are reviewed in some detail and current formulations of SYNROC are summarized. A notable property of SYNROC it its extremely high resistance to leaching by groundwaters, particularly above 100 ~ In addition, it can be shown that the capacity of SYNROC minerals to immobilize high-level waste elements is not markedly impaired by high levels of radiation damage. Current investigations are focused on developing a satisfactory production technology for SYNROC and progress to- wards this objective is described. The high leach- resistance of SYNROC at elevated temperatures increases the range of geological environments in which the waste may be finally interred; in particular, SYNROC is well adapted for disposal in deep drill-holes, both in con- tinental and marine environments. The fact that SYNROC is comprised of minerals which have demon- strated long-term geological stability is significant in establishing public confidence in the ability of the nuclear industry to immobilize high-level wastes for the very long periods required.
The dissolution of thorium phosphate diphosphate (TPD) doped or not with trivalent actinides and that of associated solid solutions with tetravalent plutonium was studied from a kinetic point of view as a function of the acidity or the basicity of the leachate. From the evolution of the normalized mass losses, the dissolution rates were determined. For all the solids considered, the values were found between 1.2×10−5 and 4.4×10−9 gm−2d−1 which confirms the very good durability of TPD to aqueous corrosion. The expression of the dissolution rate was given in acidic and in basic media (10−1–10−4 M HNO3 or HClO4 and 10−1–10−4 M NaOH). The partial orders related to the proton and hydroxide ion concentrations were found to be equal to n=0.31–0.40 and to m=0.37, respectively. The associated dissolution rate constant at pH=0 and pH=14 were found to k298K,0.1M′=1.2×10−5 to 2.4×10−5 gm−2d−1 and to k298K,0.1M″, (7.8±1.9)×10−5 gm−2d−1, respectively. In these conditions, the dissolution rate value extrapolated in neutral medium was evaluated to 2.4×10−7 to 3.6×10−7 gm−2d−1 at room temperature and to 5.0×10−6 to 7.5×10−6 gm−2d−1 at 90 °C which remains very low by comparison to the other ceramics studied for the same applications.
Considering that phosphate matrices could be potential candidates for the immobilization of actinides or for the final disposal of the excess plutonium from dismantled nuclear weapons, the chemistry of thorium phosphates has been re-examined. In the ThO2-P2O5 system, the thorium phosphate-diphosphate Th-4(PO4)(4)P2O7 (TPD) can be synthesized by wet and dry chemical processes. The substitution of thorium by other tetravalent actinides like uranium or plutonium can be obtained for 0 < x < 3.0 and 0 < x < 1.63, respectively. In this work, we report the chemical conditions of synthesis of thorium-neptunium (IV) phosphate-diphosphate solid solutions Th4-xNPx(PO4)(4)P2O7 (TNPD) with 0 < x < 1.6 from a mixture of thorium and neptunium (IV) nitrates and concentrated phosphoric acid. From the variation of the cell parameters and volume, the maximum substitution of Th4+ by Np4+ in the TPD structure is evaluated to 2.08 (which corresponds to about 52 mol% of thorium replaced by neptunium (IV)). The held of existence Of solid solutions Th4-xU-xNp-xPuUxUNpxNpPuxPu(PO4)4P(2)O(7) has been calculated. These solid solutions should be synthesized for 5(xU) + 7(xNp) + 9(xPu) less than or equal to 15. In the NpO2-P2O5 system, the unit cell parameters of Np2O(PO4)(2) were refined by analogy with U2O(PO4)(2) which crystallographic data have been published recently. For Np2O(PO4)(2) the unit cell is orthorhombic with the following cell parameters: a = 7.033(2) Angstrom, b = 9.024(3) Angstrom, c = 12.587(6) Angstrom and V = 799(1) Angstrom(3). The unit cell parameter obtained for alpha-NpP2O7 (a = 8.586(1) Angstrom) is in good agreement with those already reported in literature.
In the framework of nuclear waste management aiming at the research of a storage matrix, the chemistry of thorium phosphates has been completely re-examined. In the ThO2–P2O5 system a new compound thorium phosphate–diphosphate Th4(PO4)4P2O7 has been synthesized. The replacement of Th4+ by a smaller cation like U4+ and Pu4+ in the thorium phosphate–diphosphate (TPD) lattice has been achieved. Th4−xUx(PO4)4P2O7 and Th4−xPux(PO4)4P2O7 solid solutions have been synthesized through wet and dry processes with 0
Fast neutron irradiation is being used to produce displacement damage in SYNROC B, SYNROC C, barium hollandite, zirconolite and perovskite, to simulate the actinide decay damage which will accumulate during long term disposal of SYNROC containing radioactive waste. Fast neutron and actinide decay damage is correlated by equating the calculated displacements per atom for the two cases, on the basis of 10 wt% waste addition. The macroscopic, microscopic and crystal structure effects are being measured on all five materials for ‘SYNROC ages’ up to 106 years, obtained by irradiating to a fast neutron dose of 4.8 × 1020n/cm2 (⩾ 1 MeV). Materials variables include the fabrication method and the presence or absence of simulated radwaste.All materials have exhibited a volume expansion and a corresponding decrease in density with increasing fast neutron dose. For the same dose, the volume expansions increase in the order barium hollandite ≅ hot pressed SYNROC B < perovskite ≅zirconolite ≅ SYNROC C< cold pressed and sintered SYNROC B. Microcracking occurs in SYNROC at expansions greater than 4 vol%, producing only a small increase in open porosity. For SYNROC C the neutron dose is equivalent to a SYNROC age of 2 × 105-years.