N. P. Laverov

Synthetic minerals with the pyrochlore and garnet structures for immobilization of actinide-containing wastes

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.

Self-propagating high-temperature synthesis and characteristics of cermet matrices for isolation of wastes with long-lived radionuclides

The possibility of preparing by self-propagating high-temperature synthesis (SHS) metal-ceramic (cermet) matrices with simulated wastes of REE-actinide fraction and Tc was examined. The specimens consist of oxide crystalline phases, glass, and melts. In the aluminate composite, the component (Sm) simulating the REE-actinide fraction is in the garnet and glass phases, and in the titanate composite, in the pyrochlore, titanosilicate of perrierite structure, and glass phases. Rhenium (Tc simulator) is incorporated in alloy phases. To evaluate the prospects for radioactive waste immobilization by SHS, it is necessary to synthesize matrices containing actinide isotopes (Am) and Tc and to study their structure and isolation properties.

Glasses for immobilization of low- and intermediate-level radioactive waste

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.

Confinement matrices for low- and intermediate-level radioactive waste

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.

Geochemical conditions for the isolation of the long-lived radioisotope technetium-99

Conditions required for the isolation of Tc-bearing wastes include their incorporating into metallic alloys and placement in underground waste repositories under reduced conditions. The paper reports means of synthesizing these matrices and distinctive structural features of samples produced by self-propagating high-temperature synthesis (SHS).