排序方式： 共有10条查询结果，搜索用时 140 毫秒
1
1.
S. V. Stefanovsky S. V. Yudintsev S. E. Vinokurov B. F. Myasoedov 《Geochemistry International》2016,54(13):1136-1155
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 radionuclides (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 radionuclides. 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. 相似文献
2.
S. V. Yudintsev A. M. Pervukhina A. V. Mokhov V. I. Malkovsky S. V. Stefanovsky 《Doklady Earth Sciences》2017,473(2):427-432
In Russia, highly radioactive liquid wastes from recycling of spent fuel of nuclear reactors are solidified into Na–Al–P glass for underground storage. The properties of the matrix including the radionuclide fixation will change with time due to crystallization. This is supported by the results of study of the interaction between glassy matrices, products of their crystallization, and water. The concentration of Cs in a solution at the contact of a recrystallized sample increased by three orders of magnitude in comparison with an experiment with glass. This difference is nearly one order of magnitude for Sr, Ce, and Nd (simulators of actinides) and U due to their incorporation into phases with low solubility in water. Based on data on the compositional change of solutions after passing through filters of various diameters, it is concluded that Cs occurs in the dissolved state in runs with a glass and recrystallized matrix. At the same time, Sr, lanthanides, and U occur in the dissolved state and in the composition of colloids in runs with glass, and mostly in colloid particles after contact with the recrystallized sample. These results should be regarded for substantiation of safety for geological waste storage. 相似文献
3.
Laverov N. P. Yudintsev S. V. Stefanovsky S. V. Ewing R. Ch. 《Doklady Earth Sciences》2012,443(2):526-531
Doklady Earth Sciences - 相似文献
4.
N. P. Laverov V. S. Urusov S. V. Krivovichev A. S. Pakhomova S. V. Stefanovsky S. V. Yudintsev 《Geology of Ore Deposits》2011,53(4):273-294
Synthetic analogues of murataite, a very rare mineral—a complex oxide of REE, actinide, Ti, Fe, and other elements,—are of
great interest as confinement matrices of radioactive wastes. They are produced by sintering (1200–1300°C) and melting (1500–1600°C)
with subsequent crystallization of the melt. Four structural varieties of murataite distinguished by unit-cell parameters
have been established by TEM study. All these varieties are derivatives of the fluorite structure designated as murataite
3C, 5C, 7C, and 8C depending on repetition factor of a parameter of the fluorite subcell. The structural features of the synthetic murataite varieties are analyzed in this paper
based on data obtained from high-resolution electron microscopy, microdiffraction, and X-ray refinement. The hypothesis of
a modular structure of the members of polysomatic pyrochlore-murataite series has been confirmed. At the same time, the structural
modules are zero-dimensional rather than two-dimensional as had previously been suggested. The combinations of zero-dimensional
modules in 3D space create the entire structural diversity of the polysomatic series. 相似文献
5.
N. P. Laverov S. V. Yudintsev T. S. Livshits S. V. Stefanovsky A. N. Lukinykh R. C. Ewing 《Geochemistry International》2010,48(1):1-14
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. 相似文献
6.
Yudintsev S. V. Nickolsky M. S. Stefanovsky O. I. Nikonov B. S. 《Doklady Earth Sciences》2022,504(2):403-409
Doklady Earth Sciences - Light rare earth (REE) titanates, such as REE2TiO5, REE2Ti2O7, and REE4Ti9O24, are potential matrices for the REE-actinide fraction of high-level waste from the... 相似文献
7.
N. P. Laverov B. I. Omel’yanenko S. V. Yudintsev S. V. Stefanovsky 《Geology of Ore Deposits》2012,54(1):1-16
Mining of uranium for nuclear fuel production inevitably leads to the exhaustion of natural uranium resources and an increase
in market price of uranium. As an alternative, it is possible to provide nuclear power plants with reprocessed spent nuclear
fuel (SNF), which retains 90% of its energy resource. The main obstacle to this solution is related to the formation in the
course of the reprocessing of SNF of a large volume of liquid waste, and the necessity to concentrate, solidify, and dispose
of this waste. Radioactive waste is classified into three categories: low-, intermediate-, and high-level (LLW, ILW, and HLW);
95, 4.4, and 0.6% of the total waste are LLW, ILW, and HLW, respectively. Despite its small relative volume, the radioactivity
of HLW is approximately equal to the combined radioactivity of LLW + ILW (LILW). The main hazard of HLW is related to its
extremely high radioactivity, the occurrence of long-living radionuclides, heat release, and the necessity to confine HLW
for an effectively unlimited time period. The problems of handling LILW are caused by the enormous volume of such waste. The
available technology for LILW confinement is considered, and conclusion is drawn that its concentration, vitrification, and
disposal in shallow-seated repositories is a necessary condition of large-scale reprocessing of SNF derived from VVER-1000
reactors. The significantly reduced volume of the vitrified LILW and its very low dissolution rate at low temperatures makes
borosilicate glass an ideal confinement matrix for immobilization of LILW. At the same time, the high corrosion rate of the
glass matrix at elevated temperatures casts doubt on its efficient use for immobilization of heat-releasing HLW. The higher
cost of LILW vitrification compared to cementation and bitumen impregnation is compensated for by reduced expenditure for
construction of additional engineering barriers, as well as by substantial decrease in LLW and ILW volume, localization of
shallow-seated repositories in various geological media, and the use of inexpensive borosilicate glass. 相似文献
8.
N. P. Laverov B. I. Omel’yanenko S. V. Yudintsev S. V. Stefanovsky B. S. Nikonov 《Geology of Ore Deposits》2013,55(2):71-95
Reprocessing of spent nuclear fuel (SNF) for recovery of fissionable elements is a precondition of long-term development of nuclear energetics. Solution of this problem is hindered by the production of a great amount of liquid waste; 99% of its volume is low- and intermediate-level radioactive waste (LILW). The volume of high-level radioactive waste (HLW), which is characterized by high heat release, does not exceed a fraction of a percent. Solubility of glasses at an elevated temperature makes them unfit for immobilization of HLW, the insulation of which is ensured only by mineral-like matrices. At the same time, glasses are a perfect matrix for LILW, which are distinguished by low heat release. The solubility of borosilicate glass at a low temperature is so low that even a glass with relatively low resistance enables them to retain safety of under-ground LILW depositories without additional engineering barriers. The optimal technology of liquid confinement is their concentration and immobilization in borosilicate glasses, which are disposed in shallow-seated geological repositories. The vitrification of 1 m3 liquid LILW with a salt concentration of ～300 kg/m3 leaves behind only 0.2 m3 waste, that is, 4–6 times less than by bitumen impregnation and 10 times less than by cementation. Environmental and economic advantages of LILW vitrification result from (1) low solubility of the vitrified LILW in natural water; (2) significant reduction of LILW volume; (3) possibility to dispose the vitrified waste without additional engineering barriers under shallow conditions and in diverse geological media; (4) the strength of glass makes its transportation and storage possible; and finally (5) reliable longterm safety of repositories. When the composition of the glass matrix for LILW is being chosen, attention should be paid to the factors that ensure high technological and economic efficiency of vitrification. The study of vitrified LILW from the Kursk nuclear power plant with high-power channel reactors (HPCR; equivalent Russian acronym, RBMK) and the Kalinin nuclear power plant with pressurized water reactors (PWR; equivalent Russian acronym VVER) after their 14-yr storage in the shallow-seated repository at the MosNPO Radon testing ground has confirmed the safety of repositories ensured by confinement properties of borosilicate matrix. The most efficient vitrification technology is based on cold crucible induction melting. If the content of a chemical element in waste exceeds its solubility in glass, a crystalline phase is formed in the course of vitrification, so that the glass ceramics become a matrix for such waste. Vitrified waste with high Fe; Na and Al; Na, Fe, and Al; Na and B is characterized. The composition of frit and its proportion to waste depends on waste composition. This procedure requires careful laboratory testing. 相似文献
9.
N. P. Laverov S. V. Yudintsev S. V. Stefanovsky B. I. Omel’yanenko B. S. Nikonov 《Geology of Ore Deposits》2006,48(5):335-356
A new variety of matrices based on synthetic phases whose structure is close to that of murataite (a natural mineral) is proposed for immobilization of nuclear wastes. Murataite is Na, Ca, REE, Zn, and Nb titanate with a structure derived from the fluorite lattice. This very rare mineral was found in alkali pegmatites from Colorado in the United States and the Baikal region in Russia. The synthetic murataite-like phases contain manganese instead of zinc, as well as actinides and zirconium instead of sodium, calcium, and niobium. Varieties with threefold, as in the mineral, and five-, seven-, and eightfold repetition of the lattice relative to the fluorite cell have been established. Correspondingly, the structural varieties M3, M5, M7, and M8 are recognized among the synthetic murataites. A decrease in the contents of actinides, rare earth elements, and zirconium occurs in the series M7-M5-M8-M3, along with enrichment in Ti, Al, Fe, and Ga. Murataite-based ceramics are characterized by high chemical and radiation stability. The rate of U, Th, and Pu leaching with water at 90°C in static and dynamic tests is 10?6–10?5 g/m2 per day. These values are lower than the leaching rate of other actinide confinement matrices, for example, zirconolite-or pyrochlore-based. Murataite is close to other titanates in its radiation resistance. At 25°C, amorphization of its lattice is provided by a radiation dose of 2 × 1018 α decays/g, or 0.2 displacements/atom. Murataite-based matrices are synthesized within a few hours by cold compacting combined with sintering at 1300°C or by melting at 1500–1600°C and subsequent crystallization. The melting technology, including induction smelters with a cold crucible, makes it possible to produce samples with zonal murataite grains. The inner zone of such grains is composed of structural variety M5 or M7; the intermediate zone, of M8; and the outer zone, of M3. The contents of actinides, zirconium, and rare earth elements reach a maximum in the inner zone and drop to a minimum in the outer zone, while the amounts of nonradioactive elements—Ti, Al, Fe, and Ga—vary conversely. The U, Th, and Pu contents in the inner and outer zones differ by three to five times. Such a distribution precludes removal of actinides by interaction of the matrix with solution after its underground disposal. Individual actinides (Np, Pu, Am); the actinide-zirconium-rare earth fraction of high-level radioactive wastes (HLW); Am-Ga residues of weapons plutonium reprocessing with its conversion into U-Pu mixed oxide (MOX) fuel; and other sorts of HLW enriched in actinides, REE, and products of corrosion (Mn, Fe, Al, Zr) can be incorporated into a murataite-based matrix. As much as 350 kg of HLW components can be included in 1 t of such a ceramic. An actinide matrix that is composed of titanates with a pyrochlore structure is its nearest analogue. The advantage of murataite in comparison with pyrochlore consists in its universal character; i.e., a murataite-based matrix can be used for utilization of a wider range of actinide-bearing highly radioactive wastes. 相似文献
10.
1