Изучено строение матрицы с имитаторами отходов пирохимической переработки облученного ядерного топлива. Состав имитаторов определен с учетом свойств оксидного высокоактивного остатка, полученного в ОАО “ГНЦ НИИАР” при переработке облученного нитридного и металлического топлива. Муратаит выбран в качестве матрицы благодаря возможности его структуры изоморфно включать различные элементы отходов. Он доминирует в образце и содержит основную часть актинидов и РЗЭ. Другие фазы матрицы - титанаты (кричтонит, цирконолит, перовскит, рутил) и щелочно-земельные молибдаты. Последние растворимы в воде, поэтому в течение 300 лет, необходимых для распада Sr - 90, контакт матрицы с водой должен быть исключен за счет использования коррозионно - устойчивых контейнеров и водонепроницаемого глинистого буфера. Альтернативный способ изоляции таких отходов - синтез матрицы в инертной среде для стабилизации молибдена (IV) и его вхождения в титанатные фазы. Это позволит предотвратить образование растворимых молибдатов щелочных земель и щелочных металлов.

Изучены емкостные свойства в отношении имитаторов РЗЭ-актинидной фракции радиоактивных отходов двух типов промышленных сорбентов «термоксид-5» на основе оксида титана и «термоксид-3» на основе оксида циркония. Установлено, что титанатный сорбент «термоксид-5» обладает более высоким и равномерным распределением имитаторов отходов в гранулах по сравнению с цирконатным сорбентом. Степень его насыщения (емкость) имитаторами РЗЭ-актинидной фракции радиоактивных отходов составляет 22-24 мас.%, что в 3-5 раз выше, чем у цирконатного сорбента марки «термоксид-3». Индукционным плавлением в холодном тигле сорбента «термоксид-5» с имитаторами РАО изготовлены кристаллические минералоподобные матрицы, состоящие из рутила и редкоземельного титаната (Ln-титанат). Имитаторы РЗЭ-актинидной фракции входят только в состав Ln-титаната. Задачи дальнейших исследований изучение растворимости матриц в воде, а также их радиационной устойчивости и физико-механических свойств.

Охарактеризованы условия изоляции отходов, содержащих технеций. Это достигается его включением в металлические сплавы и размещением в подземных хранилищах с восстановительной обстановкой. Рассмотрены способы изготовления таких матриц и особенности строения образцов, полученных самораспространяющимся высокотемпературным синтезом.

Parameters of the process of vitrification of simulator of the slurry of high-level Na-, Fe, and Alcontaining waste (made in the plant in Savannah River, USA) in the presence of lithium borosilicate frits by the method of induction melting in a cold crucible are determined. With the help of X-ray diffraction analysis, optical and electron microscopy, and IR spectroscopy, phase composition of the glassy material obtained is investigated. All the samples cut from different parts of the ingot are composed of U-containing glassy phase and the crystals of spinel. A tendency to enrichment of the lower zone of the ingot with heavier ions of transition and heavy metals as well as its depletion with oxides of lighter elements (Na, Cs, Ca, Al, Si) is observed.

Изучено строение матрицы с имитаторами отходов пирохимической переработки облученного ядерного топлива. Состав имитаторов определен с учетом свойств оксидного высокоактивного остатка, полученного в ОАО “ГНЦ НИИАР” при переработке облученного нитридного и металлического топлива. Муратаит выбран в качестве матрицы благодаря возможности его структуры изоморфно включать различные элементы отходов. Он доминирует в образце и содержит основную часть актинидов и РЗЭ. Другие фазы матрицы - титанаты (кричтонит, цирконолит, перовскит, рутил) и щелочно-земельные молибдаты. Последние растворимы в воде, поэтому в течение 300 лет, необходимых для распада Sr - 90, контакт матрицы с водой должен быть исключен за счет использования коррозионно - устойчивых контейнеров и водонепроницаемого глинистого буфера. Альтернативный способ изоляции таких отходов - синтез матрицы в инертной среде для стабилизации молибдена (IV) и его вхождения в титанатные фазы. Это позволит предотвратить образование растворимых молибдатов щелочных земель и щелочных металлов.

Среди продуктов деления, образующихся в атомных реакторах, 99 является наиболее экологически опасным. Это связано с большим периодом полураспада (213 000 лет), высоким содержанием в облученном ядерном топливе (0.8–1.0 кг в тонне ОЯТ), низкой сорбционной способностью и высокой подвижностью в аэробных условиях. Основная масса 99 ( 200 т) находится в составе ОЯТ, при переработке которого этот радиоизотоп выделяется в виде отдельной фракции или вместе с актинидами. Более 60 т 99 к настоящему времени накоплено в высококонцентрированном виде. Актуальность решения задачи по изоляции 99 от окружающей среды становится все более очевидной. Необходимым звеном в решении этой задачи должна стать иммобилизация технеция в высокоустойчивой труднорастворимой матрице. В качестве матриц, способных удерживать в своем составе наряду с технецием также и актиниды, авторами предложены керамики, состоящие из титанатов со структурой пирохлора, перовскита и рутила. Высокая их химическая устойчивость была ранее подтверждена экспериментально. Трудности получения таких матриц связаны с летучестью Тс и необходимостью его перевода в Tc(IV). Для преодоления этого предлагается самораспространяющийся высокотемпературный синтез – СВС, характерными особенностями которого являются восстановительные условия и быстротечность протекания реакций. В состав шихты для синтеза матриц входят: восстановители – порошки металлов с большим сродством к кислороду (Ti, Zr), окислители – оксиды (MoO3, Fe2O3, CuO), а также добавки (TiO2, ZrO2, Y2O3, СаО и др.), которые вместе с другими элементами формируют целевые фазы. При синтезе матриц вместо Тс в качестве имитатора использован близкий к нему по геохимическим свойствам Мо. Методом СВС синтезированы образцы молибденсодержащих матриц, охарактеризованы их фазовый состав и закономерности распределения молибдена. Показана возможность включения в них до 40 мас. % Мо, который находится в виде металла и как примесь в структуре титанатных фаз. Для совместной иммобилизации актинидов, редких земель и 99 оптимальными свойствами обладают титанатно-цирконатные матрицы на основе пирохлора.

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

The paper deals with some approaches to long-term solutions to managing radioactive wastes (RW) and spent nuclear fuel (SNF) volumes that have been accumulated by enterprises of the Russian Federation (RF) Minatom and RF Ministry of Defence. Short descriptions of the RW and SNF management practices are presented. The paper considers the principle of trend management of high-level waste (HLW), including the separation of fractions containing short-, intermediate-, and long-lived radionuclides; their conversion to stable matrices; and separate disposal of the matrices taking into account safety precautions for the repositories located at various depths and in different geological conditions. The paper also discusses some directions for future work in the fields of novel matrix material development and geophysical-geological research for site selection and characterisation.

Phase relations in the system: CaO-TiO 2 -SiO 2 -(Na 2 O,Al 2 O 3 ,Gd 2 O 3 ,UO 2 ) were studied. This system is of interest due to the formation of sphene, perovskite,. and other phases potentially suitable for immobilization of high level waste (HLW) elements. Along with sphene, other phases found in the samples were ru-tile, chevkinite, anorthite, crystobalite, and pyrochlore-structured phases. Sphene is able to incorporate up to 21.5 wt.% Gd 2 O 3 and 9.3 wt.% UO 2 or, in formula units: 0.25 Gd 3+ and 0.07 U 4+

Studying the Synroc, doped with a simulated HLW, we have found, along with conventional Synroc phases (zirconolite, perovskite, hollandite), an extra phase with a stoichiornetry (Ca, Mn, U, TR) 4 (U, TR, Zr, Ti) 2 (AI, Ti) 7 O 22 . XRD and TEM study has shown this phase is related to a very rare mineral murataite. In the present work a ceramic based on murataite is studied. The ceramic samples in the system: Ca-Mn-Ti-Zr-U-Ce-AI-Fe-O were produced and examined in details using XRD, SEM/EDS, TEM, and optical microscopy. Total amount of actinide (U) and rare earth (Ce, Gd) elements in the murataite exceeds 20 wt%. Isomorphic substitution schemes in the structures of synthetic and natural murataites are discussed. High isomorphic capacity of the murataite structure towards actinides and REEs, flexibility of its composition, feasibility of synthesis by melting, including a cold crucible melting, and very high chemical durability under hydrothermal conditions make the murataite-based ceramics very promising for actinides and excess weapon Pu fixation.

Pyrochlore is a complex oxide with the nominal formula A 2 B 2 X 6 Y, where A and B are cations in VIII and VI-fold co-ordination, X and Y are anions. Its structure is derived from the cubic fluorite structure. In natural pyrochlores A = Na, Mg, K, Ca, Mn, Fe, Sr, Sb, Cs, Ba, REEs, Pb, Bi, Th, and U; B = Nb, Ta, Ti, Zr, Sn, W, Fe, and Al; X = O; Y = O, OH, or F. Synthetic pyrochlores have been repeatedly described as matrices designed for actinide-bearing waste immobilization. In synthetic pyrochlores site “A” is mainly occupied by Ca, U, An, and REEs; B = Ti and Zr; X and Y = O. In this work we have studied pyrochlores in crystalline titanate-based waste forms. The ceramics were fabricated in the system: Ca-Mn-U-REE-Zr-Ti-Al-O by cold pressing and sintering, melting in a high-temperature furnace, and inductive melting in a cold crucible. All specimens were studied by XRD, SEM/EDS and TEM methods. The amount of pyrochlore in the samples varied from 10 to 70%. Other phases in these ceramics were brannerite, perovskite, zirconolite, murataite, hibonite, loverengite, pseudobrookite, and rutile. Compositions of the pyrochlores correspond to stoichiometry: A 2 B 2 O 7-x , 0.1<x<0.4, where A = Ca, Mn, REEs, U, Zr; B = Ti, Zr, Al, Mn. The positions and intensities of the peaks of pyrochlores from various ceramics were: d 222 =2.89-2.93A, 1=100; d 400 =2.51, 1=10-25; d 440 =1.779-1.809, 1=20-60; d 622 =1.512-1.540, 1=20-35; d 444 =1.451-1.477, I=10-15; d 662 =1.158-1.173, I=10-15. These data allowed the determination of the unit-cell dimensions of the pyrochlores as 1.00-1.02 nm. Results obtained from TEM research agree well with these values. Distribution of U and REEs among all phases of the ceramics was characterized. The main substitutions which have influenced the pyrochlore compositions are discussed.

Since the early of the 1990s the method of inductive melting in a cold crucible (IMCC) has been applied at SIA “Radon” for production of various wasteforms, including glasses and Synroc-type ceramics. Sphene-based glass-ceramics composed of glass and crystalline phases were considered as appropriate wasteform for High Level Waste immobilisation. Investigation of two glass-ceramic specimens prepared with the IMCC has been performed using optical microscopy, XRD, SEM/EDS, and TEM methods. The samples produced consist of vitreous and crystalline phases. The vitreous phase consists of two varieties of glass formed by the immiscibility of the initial melt onto two separate liquids. One of the glasses is observed as spherical microinclusions in the matrix glass. The glass of the microspheres are differed from the matrix glass composition by higher contents of Ca, Ti, Ce, Sr, Zr (or Cr), while the matrix glass contains higher amounts of Si, Al, and alkalies. The crystalline phases with sphene- and perrierite-like structures have been also occurred. Their total quantity reaches up to 50 vol.%. The synthetic perrierite has similar unit-cell parameters with its natural mineral analogs with the only exception in two-fold value of c dimension. Zr, Ce, and Sr are incorporated into synthetic sphene and perrierite, while Cs is hosted by the glass phases.

Research on synthetic zircons fabricated in system ZrO 2 -UO 2 -SiO 2 at T = 400 - 750° C, P H20 = 100-200 MPa, and Ni-NiO oxygen buffer was made in order to estimate the limits of uranium incorporation into the phase's structure. It was shown that up to 27 wt.% of uranium may locally enter into the synthetic zircons. SEM/EDS study of the samples has showed that the element is uniformly distributed within the zircon crystals. Incorporation of U into zircon increases its unit-cell dimensions up to α = 6.70 A and c = 6.00 A relative to 6.60 and 5.98 A for a pure phase without impurities. After thermal high temperature treatment of the samples under 1400° C, in air, for 6 hours inhomogeneous distribution of uranium within the zircon grains was revealed. Very small uranium rich spots probably corresponding to a newly formed phase were observed in the zircon crystals. Contents of the element throughout the grains decreased to about 10 wt.%. It is probably caused by destruction of zircon-coffinite solid solution formed at hydrothermal conditions in the course of its re-heating. Study of interaction of natural zircons with alkaline carbonate solution at 100° C shows that U release from the phase can reach up to 6 % of total uranium in the samples even under relatively short term interaction. Loss of uranium rises with the element content (or amorphisation degree) and run duration, as well as with grain size reduction. Research on uranium distribution in natural zircons picked up from altered and weathered rocks is characterised by strong depletion of U at grain margins relative to the centre of the crystals. These results indicate that zircon may lose some uranium and, probably, the other actinides with time due to its lattice destruction under radioactive decay followed by hot water attack.

Preparation and characterization of inductively-melted Synroc containing 20 wt% simulated plant “Mayak” reprocessing waste were performed. The sample bulk composition was as follows, (in wt.%): 55.4 TiO2; 15.8 ZrO2; 7.5 CaO; 7.4 BaO; 4.3 Al2O3 2.0 MnO; 1.8 SiO2; 0.7 Na2O; 1.9 K2O, 0.5 Ce2O3; 1.0 UO2; 0.9 NiO; 0.6 Cr2O3, and 0.2 FeO. The sample was produced by melting in air at 1550–1600 °C under barometric pressure. It is composed of a few crystalline phases and a minor glass phase. Most of the phases (hollandite, zirconolite, perovskite and rutile) are similar to the analogous phases found in the other Synroc formulations. An additional phase with average composition, wt.%: 59.8 TiO2; 15.6 CaO; 7.0 UO2; 5.6 ZrO2; 4.7 MnO; 4.1 Ce2O3, and 1.8 Al2O3 was found. Some elements (Ba, Si, Ni, K, Na, Fe) were present in the phase in negligible quantities. Its formula (Ca2.65U0.3Ce0.2)(Ti7.3Mn0.6Zr0.4Al0.3)O20.0 is rather close to a rare mineral uhligite - Ca3(Ti,Zr,Al)9O20. Another possible counterpart of the phase is murataite-like mineral previously described in tailored ceramic designed for Savannah River Plant wastes fixation. This phase as well as zirconolite are the major host for U in the sample Preliminary data on the material leachability in water at 350 °C and 50 MPa have been obtained Uranium contents in the solution were about 1 ppb and close to the uranium dioxide solubility in deionized water under the same P-T conditions.

Three Synroc-C samples, containing simulated high level waste were studied. One was produced by the conventional hot-pressing method at ANSTO, Australia, and the others were obtained using cold crucible technology at Radon, Russia. One of the melted samples was prepared using the Australian sol-gel precursor and the second one was obtained From an oxide-nitrate mixture. It was established that the specimens have closely similar mineral compositions, with major hollandite, perovskite, zirconolite, and rutile. Small amounts of hibonite were also found. Unlike the hot-pressed Synroc containing metallic alloy particles, melted Synrocs contain molybdates. An investigation of mineral compositions and elemental distribution in the samples was carried out. Features of hot-pressed and melted ceramics were compared. Unit cell parameters of the Synroc phases were determined and preliminary results on durability of the melted Synroc are presented.

Zirconolite-rich ceramics were produced by the cold crucible melting technique in an air atmosphere, at 1550±50 °C and 1 atm. Four samples with overall composition (in wt.%): 4.9–14.3 CaO; 19.0–41.3 ZrO2; 24.1–42.6 TiO2; 1.3–11.3 Al2O3; 6.8–30.0 Gd2O3, and 1.1–8.5 SiO2 have been studied. Total phases in the ceramics consist of major zirconolite and minor rutile, perovskite, zirconia, aluminium titanate, and glass. The Gd2O3 content in zirconolite reaches up to 31.4 wt.% corresponding to the formula: (Ca0.4,Gd0.7)Zri1.0(Ti1.4,Al0.5)O7.0. The data on the phase composition agree well with coupled Gd incorporation into the mineral structure. Ca(H) + Ti(IV) = Gd(III) + Al(III), and 2Gd(III) = Ca(II) + Zr(IV). The highest Gd contents observed in the other phases are 25.4 % for zirconia, 12.6 % in glass, 8.8 % in perovskite, and 1.4 % for rutile. The rest of the elements' distribution in the samples are analyzed.

The principal requirements for employing natural minerals as buffer and backfill material in high-level waste (HLW) repositories are high sorptive properties, low water permeability, relatively high thermal conductivity, and thermostability. The major task of the buffer is to prevent the penetration of radionuclides into groundwater. The authors of this report examined weathered basic rocks from three regions of Russia in consideration as a suitable radioactive waste barrier.

Sorptive properties of weathered dunites, gabbro-diabases and basic volcanic rocks for Sr and Cs were studied. The results show that the sorptive capacities of these rocks are equivalent to or, in some cases, superior to industrial sorptive materials. Results of a uranium distribution study by fission-track radiography suggest that material from weathered basic rocks is characterized by high sorptive properties for uranium also. One can assume that other radionuclides of the transuranic group will be intensely sor-bed by the residuum of weathered basic rocks. Low-temperature hyd-rothermal transformation leads to sealing fissures of the basic rocks with highly sorptive minerals, for example, smectite, chlorite, serpentine, talc, zeolite, hydroxides of Fe, Ti, Mn. The process results in contemporaneous decreasing hydraulic conductivity and increasing sorptive capacity of the rocks. HLW disposal at the radiochemical plant “Mayak” is expected to be produced in deep wells situated in basaltic rocks. The safety of disposal is based on high sorptive properties of the crust of weathering and protective capacities of volcanic rocks. This method is not expensive and may allow the disposal of HLW in the near future.

Sample glasses have been made using SB6 high level waste (HLW) simulant (high in both Al and Fe) with 12 different frit compositions at a constant waste loading of 36 wt.%. As follows from X-ray diffraction (XRD) and optical and scanning electron microscopy (SEM) data, all the samples are composed of primarily glass and minor concentration of spinel phases which form both isometric grains and fine cubic (∼1 μm) crystals. Infrared spectroscopy (IR) spectra of all the glasses within the range of 400-1600 cm{sup -1} consist of the bands due to stretching and bending modes in silicon-oxygen, boron-oxygen, aluminum-oxygen and iron-oxygen structural groups. Raman spectra showed that for the spectra of all the glasses within the range of 850-1200 cm{sup -1} the best fit is achieved by suggestion of overlapping of three major components with maxima at 911-936 cm{sup -1}, 988-996 cm{sup -1} and 1020-1045 cm{sup -1}. The structural network is primarily composed of metasilicate chains and rings with embedded AlO{sub 4} and FeO{sub 4} tetrahedra. Major BO{sub 4} tetrahedra and BO{sub 3} triangles form complex borate units and are present as separate constituents. (authors)