В статье рассматриваются теоретические основы построения эффективной системы наблюдения в Арктике.Основой для написания данной статьи явились положительные результаты практического применения сегментов системы наблюдения в океанологии, в том числе в арктическом регионе.Анализируются четыре подсистемы наблюдения: космическая, радиолокационная, гидроакустическая, а также автоматическая система идентификации судов.Показано, что только комплексное использование всех подсистем наблюдения позволит обеспечить эффективный контроль над обстановкой и информационную поддержку принятия решений

Multidisciplinary investigations carried out in the Cape Zhelaniya area and on the Severny ice dome of Severny Island in the Novaya Zemlya Archipelago during cruise 63 of the R/V Akademik Mstislav Keldysh in September 2015 included a study of the environmental radiation level. The landscape‒geochemical and radiation‒glaciological data show that the Severny ice dome serves as a secondary source of radionuclides on the surface of the ice sheet; this source originated from past nuclear weapons testing in the atmosphere over the Severnaya Zemlya test site. Some samples taken from the periglacial zone near the edge of the Severny ice dome yielded specific activity levels of radioactive cesium of 450–650 Bq/kg. The study of ice cores obtained by shallow (up to 5.4 m) drilling of three boreholes revealed no significant activity values. At the same time, glaciological investigations made it possible to obtain the first data on the previously unexamined glacier, which indicate that the radioactively contaminated layer is located at a depth of 15‒20 m at the boundary of the glacier alimentation zone. No similar investigations had been conducted earlier either by Russian or international scientific teams.

For comprehensive study of heavy mineral placers the modern methods areneeded for integrated analysis. The generalization of methodological approaches and techniques has done for two kinds of modeling, widely used in the study of placers. The digital structural-lithological (DSL) modeling [1] is based on the principles of authoring partition for geological objects using a wide range of lithological, geochemical, metallogenic properties and other features. DSL-model is built as the result of computer processing of the target database, prepared on the basis of these principles. Methodology of geological-dynamic modeling of heavy mineral placers [2] is based on the study of the dynamic mechanisms of the main stages of placer formation processes - from the mobilization in the load sources of placer-forming material to deposition in the terminal sedimentary basin. Technology of hybrid expert systems [3] is the best for development of expert systems based on digital models for target prediction and evaluation of placer objects, because it permits the using of different knowledge about the object of study in various forms, allowing their automatically applying during expert system operation. The experimental sample model shave shown a high level oftargetfunction.

The first Russian underground repository for high- and intermediate-level radioactive waste (HLW and ILW, respectively) will be built in the crystalline Archean granite-gneisses at Yeniseisky (Krasnoyarsk region, Siberia). The geological and hydrogeological characteristics of this site are similar to those found in Forsmark (Sweden) and Olkiluoto (Finland). However, the Russian disposal strategy is different. HLW will be disposed in the form of an aluminophos phate glass and ILW (with long-lived radionuclides) will be cemented. Preliminary research on all aspects of repository design (stability of waste forms, waste packages, and bentonite buffer; evaluation of the geologic barrier; and simulation of radionuclide transport by groundwater) will be performed at an on-site underground research laboratory.

The report is devoted to methodology of prognostic-search model construction of ore placer deposit on the basis of its multifactor multitag models under study by logic-informational methods. The logic-informational model is represented by feature set, informative for solution of prognostic problem. All informative tags receive a quantitative estimation of its importance. Digital modeling allows to visualize the informative tags of geological object in graphic form.

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 m 3 liquid LILW with a salt concentration of ∼300 kg/m 3 leaves behind only 0.2 m 3 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.

The tectonic evolution of the Arctic Region in the Mesozoic and Cenozoic is considered with allowance for the Paleozoic stage of evolution of the ancient Arctida continent. A new geodynamic model of the evolution of the Arctic is based on the idea of the development of upper mantle convection beneath the continent caused by subduction of the Pacific lithosphere under the Eurasian and North American lithospheric plates. The structure of the Amerasia and Eurasia basins of the Arctic is shown to have formed progressively due to destruction of the ancient Arctida continent, a retained fragment of which comprises the structural units of the central segment of the Arctic Ocean, including the Lomonosov Ridge, the Alpha-Mendeleev Rise, and the Podvodnikov and Makarov basins. The proposed model is considered to be a scientific substantiation of the updated Russian territorial claim to the UN Commission on the determination of the Limits of the Continental Shelf in the Arctic Region.

Переработка облученного ядерного топлива (ОЯТ) с целью извлечения делящихся элементов – непременное условие развития ядерной энергетики в долгосрочном плане. Препятствие в решении этой задачи – образование большого количества жидких отходов, из которых более 99% объема составляют низко- и среднеактивные отходы (НСАО). Объем высокоактивных отходов (ВАО), характеризующихся высоким тепловыделением, не превышает долей процента. Растворимость стекол при повышенных температурах делает их малопригодными для иммобилизации ВАО, изоляцию которых способны обеспечить только минералоподобные матрицы. Стекла служат идеальной матрицей для отходов низкого и среднего уровней радиоактивности (НСАО), которые характеризуются низким тепловыделением. Растворимость боросиликатных стекол при низких температурах столь мала, что даже стекла с относительно невысокой устойчивостью способны обеспечить безопасность подземных хранилищ НСАО без дополнительных инженерных барьеров. Оптимальный способ изоляции жидких НСАО – их концентрирование, иммобилизация в боросиликатных стеклах и захоронение в приповерхностных геологических хранилищах. При остекловывании 1 м3 жидких НСАО с концентрацией солей около 300 кг/м3 получается всего 0.2 м3 отходов, что в 4–6 раз меньше, чем при битумировании и в 10 раз, чем при цементировании. Экологические и экономические преимущества остекловывания НСАО определяются: 1) низкой растворимостью остеклованных НСАО в природных водах; 2) высокой степенью уменьшения объема НСАО; 3) возможностью размещения остеклованных отходов без дополнительных инженерных барьеров в приповерхностных условиях в разнообразной геологической среде; 4) хорошими прочностными характеристиками стекол, облегчающими их хранение и транспортировку; 5) а также надежными гарантиями безопасности хранилищ в долгосрочном плане. При выборе состава стекломатриц для НСАО главное внимание должно уделяться факторам, обеспечивающим высокую технологическую и экономическую эффективность остекловывания. Результаты изучения остеклованных НСАО Курской (реакторы РБМК) и Калининской (реакторы ВВЭР) АЭС после их 14-ти летнего нахождения в приповерхностном хранилище полигона НПО “Радон” подтверждают возможность обеспечения безопасности хранилищ за счет изоляционных свойств боросиликатной матрицы. Наиболее эффективным способом остекловывания является индукционное плавление в “холодном тигле” (ИПХТ). При содержании в отходах химического элемента выше его растворимости в стеклах при остекловывании образуется кристаллическая фаза и матрицей для таких отходов становится стеклокерамика. Приведена характеристика остеклованных отходов с высоким содержанием Fe; Na и Al; Na, Fe и Al; Na и B. Состав стеклофритты и ее количественное соотношение с отходами подбираются в зависимости от состава отходов. Решение этой задачи для отходов конкретного типа требует проведения серьезных лабораторных исследований.

Добыча урана для производства ядерного топлива неизбежно ведет к истощению его природных запасов и росту рыночных цен. Обеспечить АЭС энергетическим сырьем можно за счет переработки отработанного ядерного топлива (ОЯТ), в котором сохраняется более 90% энергетического ресурса. Основные трудности в решении этой задачи связаны с образованием при переработке ОЯТ огромного количества жидких отходов, которые надо концентрировать, отверждать и захоранивать. Эти отходы принято разделять на низко (НАО)-, средне (САО)- и высокоактивные (ВАО), на которые приходится 95, 4.4 и 0.6% суммарного объема отходов соответственно. Несмотря на небольшой объем, радиоактивность ВАО примерно соответствует суммарной радиоактивности НАО и САО. Главная опасность ВАО связана с их очень высокой радиоактивностью, наличием долгоживущих радионуклидов и тепловыделением, а также необходимостью изоляции на практически неограниченный срок. Основные трудности обращения с отходами низкого и среднего уровней радиоактивности (НСАО) обусловлены их колоссальными объемами. Дан анализ существующих способов изоляции НСАО, на основе которого обосновывается вывод о том, что их концентрирование, остекловывание и захоронение в близповерхностных хранилищах является необходимым условием широкомасштабной переработки ОЯТ реакторов ВВЭР-1000. Значительное снижение объема НСАО при остекловывании и очень низкая скорость растворения при низких температурах делает боросиликатное стекло идеальной консервирующей матрицей для иммобилизации НСАО. Вместе с тем очень высокая скорость коррозии стекломатриц при повышенной температуре заставляет усомниться в эффективности их использования для иммобилизации тепловыделяющих ВАО. Повышенные затраты на остекловывание НСАО по сравнению с цементацией и битумированием могут быть компенсированы за счет уменьшения затрат на создание дополнительных инженерных барьеров, значительного уменьшения объема НСАО, возможности размещения близповерхностных хранилищ в разнообразной геологической среде, использования в качестве матриц наиболее дешевых в производстве боросиликатных стекол.

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.

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.

Matrices for actinide wastes, consisting of complex murataite-type oxides, were studied. The samples were prepared by sintering of the oxide charge at 1100–1400°C or by fusion at 1450–1600°C followed by crystallization. Along with the structural analog of the natural mineral (murataite 3C), appreciable role in the samples is played by the phases (hereinafter murataites 5C, 7C, and 8C) consisting of pyrochlore and murataite 3C units. The fraction of wastes in the samples is about 10 wt %, which is close to the values for the pyrochlore matrix for Pu immobilization. In the ceramics prepared from the melt, the murataite grains have a zonal structure. They are built of murataite 3C at edges and murataite 5C (rarely 7C) in the center where the actinide content is maximal. This fact accounts for their high capability to isolate radionuclides. Amorphization of the structure only slightly affects the stability of murataite in solution. The optimal procedure for the industrial production of the matrices is their melting by induction heating in cold crucible (IMCC) and crystallization.

The methodological issues in safety assessment of repositories of spent nuclear materials (high level wastes and untreatable irradiated fuel) in crystalline rocks are discussed, and a concept of such assessment is proposed. The problem is urgent because of the need for scientific confirmation of the safety assessment procedure for such repositories in Russian territory. The concept develops the provisions of Russian normative base in the field of handling of radioactive wastes of this type, taking into account the world experience. The concept is based on theoretically sound principles to be followed in safety assessment for geological repositories. It takes into account the results of analysis of a vast literature on this issue published both in Russia and abroad. Radiological safety criteria and methods for improving the reliability of prediction calculations for determining safety indicators are characterized. The arguments that supplement radiological indicators are shown to be of importance for substantiating the safety of the anticipated repository. The regular implementation of the assessment procedure at each completed stage in the implementation of the project of terminal geological isolation of wastes is regarded as a means for quality control of the performed works.

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).

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.

Crystalline rocks are common for areas of dislocation of radiochemical enterprises in Russia and are the most probable media for construction of underground repositories of irradiated nuclear fuel and high-level radioactive waste. The repositories are supposed to be sited in seismically stable blocks with a low speed of vertical movements, where there is no volcanism and mineral resources. Under these conditions, the only mechanism of biosphere pollution is radionuclide transport by ground waters from the repository to environment. The article considers the isolation properties of crystalline rocks and conditions preventing release of radionuclides from underground repositories.

Rate of zirconolite (CaZrTi2O7) and pyrochlores (Gd2Ti2O7, CaCeTi2O7, and Gd2Zr2O7) formation by cold pressing and sintering (CPS) was studied. Batches were prepared from CaCO3, ZrO2, TiO2, CeO2, and Gd2O3 milled to size of 20–30 micrometers. Powders were compacted and sintered in air (oxygen) at 800–1600 °C for 0.5–55 hours. Samples were examined with XRD, SEM, and TEM. Phase formation rate was the fastest in Gd-Ti-O and Ca-Ce-Ti-O systems and the reactions were the slowest in Gd-Zr-O system. In zirconolite-based system reaction rate had intermediate value. For three systems at 1400 C equilibrium was reached in 3–5 hours, while in Gd-Zr-O system required about ten time longer reaction time. High formation rate was evidently one of the reasons to select the pyrochlore (Ca,Gd,U,Pu)2(Hf,Ti)2 O7 as a host for Pu in the USA. We have also studied an effect of titanium on pyrochlore formation in the Gd-Zr-Ti-O system. The batches with nominal composition Gd2Ti0.4Zr1.6O7, Gd2Ti0.2Zr1.8O7, and Gd2Ti0.1Zr1.9O7 were compacted and annealed at 1500–1600 °C for 3–98 hours. In most of them along with pyrochlore non-reacted zirconium and gadolinium oxides were found. Their content reduced with temperature and sintering duration increasing. Pyrochlore composition was varied in different parts of the samples. Variation range decreased with rise of temperature, runs duration, and titanium content in the precursors. Partial substitution of Ti for Zr increased rate of pyrochlore formation. However, even at the highest titanium content (0.4 formula units) and temperature of synthesis (1600 °C) too long sintering duration in tens of hours is required to form a single-phase pyrochlore matrix. This makes their production from oxide precursor via CPS route inefficient. Inductive melting in a cold crucible or self-sustaining high-temperature synthesis are more promising methods for fabrication of the zirconate pyrochlores.