Disposal Of High-level Radioactive Waste Research Articles

Anaerobic corrosion of carbon steel in compacted bentonite exposed to natural Opalinus clay porewater: Bentonite alteration study

Abstract. Carbon steel is a potential canister material for the disposal of high-level radioactive waste in deep geological repositories in clays and clay rocks. Bentonite is considered as a potential backfill material for those multi-barrier systems. To predict the long-term performance and for safety assessment the knowledge of canister corrosion behavior is important. The corrosion products formed and mineralogically altered bentonite at the canister/bentonite interface can potentially provide an additional barrier against radionuclide migration. In-situ corrosion experiments were performed at the Mont Terri underground research laboratory. Coupons of carbon steel were embedded in Volclay MX-80 bentonite with controlled densities, installed in a borehole under simulated repository and anaerobic conditions and exposed to natural Opalinus clay porewater for a period up to 5.5 years (Smart et al., 2017). In the present study, the bentonite layer at the canister/bentonite interface was characterized by complementary microscopic and spectroscopic techniques (XPS, SEM-EDX, µXANES) under anoxic conditions. The interface revealed reddish-brown staining up to 2 mm depth into the bentonite in the zone adjacent to the steel in all three obtained samples. The XPS analysis revealed formation of sulfides at the interface consisting of iron and other trace metals present in the steel. The SEM-EDX analyses of the interface (embedded cross-cut with steel removed) showed different degrees of calcium enrichment in the bentonite adjacent to the metal for various samples. The µXRF analysis performed on the bentonite at the interface showed a scarce or distinct calcium-enriched rim up to 100 µm into the bentonite and iron-enriched rim depending on the sample (one sample in Fig. 1), while µXANES analysis revealed formation of iron silicate compounds in the reacted reddish-brown zone. The iron appears to displace calcium from the interlayer sites in montmorillonite. The calcium then precipitates at the interface as calcite. The extent of this process seems to be strongly related to the bentonite density. The steel coupon was removed prior to embedding, with the location marked as resin in Fig. 1. A line scan from the edge towards the bulk bentonite did not indicate any systematic gradient in the Fe2+/3+ ratio. The formation of mixed Fe2+/3+ silicate compounds appears to be heterogeneous. This work contributes to an increasing understanding of steel corrosion mechanisms in clay, which can improve the robustness of canister lifetime predictions.

Evaluation of the corrosion behavior of modern spent nuclear fuels under repository conditions

Abstract. In Germany it is planned to directly dispose spent nuclear fuel (SNF) from nuclear power plants together with other high-level radioactive wastes (HLW) from former SNF reprocessing (e.g., vitrified waste), in a deep geological repository for heat-generating wastes – the siting process for this repository was started in 2017 and is ongoing. Based on several decades of research, development, and demonstration (RD&D) it is generally accepted at the technical and scientific level that direct disposal of HLW and SNF in deep mined geological repositories is the safest and most sustainable option (CEC, 2011; IAEA, 2004). The current efforts to improve the performance and accident tolerance of fuels in nuclear power generation resulted in an increased utilization of a variety of new types of light-water reactor (LWR) fuels such as fuels doped with Cr, Al, and Si. This doping leads to a significant change of the microstructure of the fuel matrix. The corrosion behavior of these types of fuels under conditions relevant to deep geological disposal has hardly been studied so far; however, this is of crucial importance as the development of a robust safety case for deep geological disposal of SNF requires a solid understanding of its dissolution behavior over very long time scales (up to 1 million years). To fill this knowledge gap, additional systematic studies on modern doped UO2 fuels were needed. Corrosion experiments with SNF cannot entirely unravel all of the various concurring effects of the dissolution mechanism due to the chemical and structural complexity of SNF and its high beta and gamma radiation field during the first 1000 years; moreover, technical restrictions only allow a very limited number of experiments. Therefore, within the EU-DisCo project (https://www.disco-h2020.eu, last access: 11 October 2021), a very ambitious programme of corrosion studies on irradiated Cr and Al/Cr doped fuels was carried out, which was complemented by systematic single-effect dissolution studies (e.g., with respect to doping level, grain size and thermodynamic aspects) performed on carefully prepared and characterized, simplified UO2-based model materials. Here, we present recent results on the dissolution behavior of tailor-made UO2 model materials in accelerated static batch experiments using H2O2 as simulant for radiolytic oxidants, present in long-term disposal scenarios for SNF in failed container conditions due to the alpha irradiation of water. In these dissolution experiments pure UO2 reference pellets exhibiting different densities and grain sizes, as well as Cr-doped UO2 pellets with various Cr-doping levels, produced using different doping methods having different grain sizes, were used. In addition, Nd-doped and industrially produced Cr- and Cr/Nd-doped UO2 pellets were used to determine the influence of these parameters on the dissolution rates. The dissolution experiments were performed under strictly controlled conditions with respect to exclusion of oxygen, temperature control, and exclusion of light. This bottom-up approach was followed to understand how the addition of Cr-oxide into the fuel matrix affects SNF dissolution behavior under repository relevant conditions. The results of the dissolution experiments performed with real SNF and the model materials obtained by the DisCo partners build the basis for numerical simulations on the dissolution behavior of modern SNF. First results of the data evaluation indicate that the addition of dopants and the consequential modification of the fuel matrix does not lead to a significant change of the dissolution behavior of these fuels under repository relevant conditions compared to standard SNF (i.e. dissolution rates agree within an order of magnitude).

Dissolution Behavior of Simulated Spent Nuclear Fuel in LiCl-KCl-UCl3 Molten Salt

Considering the necessity of the development of methods to reduce the burden of storage and disposal of high-level radioactive waste, in this study, we propose a nuclide separation technique using molten salt immersion. The dissolution behavior of simulated spent nuclear fuels (SSFs) immersed in a LiCl-KCl-UCl3 (LKU) molten salt at 500°C was investigated using a combination of thermodynamic and experimental studies. Surrogates of transuranic elements (TRUs), that is, rare earth elements (REs), in the SSFs were dissolved into the molten salt without any damage to the UO2 structure of the SSFs. The results suggest that the LKU salt technique can be used to separate REs and, potentially, TRUs from actual spent nuclear fuels (SNFs). It is thought that this technique is advantageous over the conventional TRU recovery techniques because the majority of the SNFs (i.e., UO2) remained stable, thus reducing the process burden. Several SNF treatment process options using this technique were suggested. This study will serve as a guide for future studies on the management of high-level waste discharged from nuclear reactors.

Coupled Thermo–Hydro–Mechanical Characteristics of Compacted GMZ Bentonite: Parameters Identification and Validation

AbstractCompacted Gaomiaozi (GMZ) bentonite has been considered as the first choice of buffer materials in the deep geological disposal of high-level radioactive wastes (HLW) in China. During the l...

Residual lateral stress effect on the swelling pressure measurements of compacted bentonite/claystone mixture

The design of sealing/backfill elements in high level radioactive waste disposal repository often depends on the laboratory results in terms of swelling pressure and hydraulic conductivity. In this study, a series of swelling pressure test were performed on compacted specimens of MX80 bentonite and Callovo-Oxfordian (COx) claystone mixture with different initial stress state (with and without residual stress relaxation), aiming at investigating the effects of compaction induced residual lateral stress on the swelling pressure. Two methods named “Trimming” and “Transfer” were applied for the residual stress relaxation; whereas specimens were directly compacted in the hydration ring to simulate the case without residual stress relaxation. A higher swelling pressure was obtained on the specimens directly compacted in the testing ring compared to their counterparts with residual stress relaxation. Regarding to the relaxation methods, “Transfer” is more practical for specimen preparation in the laboratory, as it allows the relaxation of the stress without changes in specimens' diameter and mass.

Molecular dynamics study on creep behavior of montmorillonite

Clay-rich rocks are widely used in high temperature and stress environment in a very long term like disposal of high-level radioactive waste, exploring its creep deformation mechanism is essential. Clay minerals, as the essential components of clayey rocks, have a crucial influence on the creep behavior. However, it is of great challenge to experimentally investigate the mechanical property of a single crystal mineral due to its small size. In recent years, molecular dynamics (MD) simulations are widely used to address a series of scientific issues including creep mechanism from atomic level. In this study, we selected montmorillonite (MMT), a representative clay mineral in clayey rocks, as the principal object to explore the creep behavior in different temperatures and stresses. The temperature varies from 300 K to 1200 K while the stress is applied in the range of 14GPa to 32GPa. Meanwhile, the anisotropic creep behavior along the two orthogonal crystal orientations is studied as well. It is found the atomic numerical results show a good agreement with macroscopic data that the deformation increases with the temperature raise and applied stress, the creep curve contains two different phases under low temperature and stress while three phases are observed under high values of temperature and stress. Besides, the creep deformation exhibits an obvious anisotropic trend, due to oriented atomic structures. This study provides a better understanding of the creep mechanism of single crystal of MMT and an atomic insight on macroscopic time-dependent deformation of clayey rocks.

Study on Rock Mass Classification Methods Used in the Geological Disposal of High-level Radioactive Waste

The conventional methods of rock quality classification mainly focus on the constructability of rock mass, while the long-term safety requirements of the high-level radioactive waste (HLW) disposal are not considered. Therefore, two rock suitability classification methods for geological disposal projects, the Rock Suitability Criterion (RSC) and the QHLW methods were systematically compared by evaluating the rock suitability of the ONKALO site for geological disposal in Finland. The suitability of the ONKALO site was analyzed by evaluating the permeability, geochemical condition, and integrity of rock mass, according to the RSC method. Meanwhile, the QHLW method was used to evaluate the rock mass suitability by analyzing the fault zone influence, hydrogeochemical characteristics, temperature influence, permeability, strength-stress ratio, and integrity of rock mass. Then, the advantages and disadvantages of the two methods were discussed. We found that the results obtained with RSC are consistent with those with QHLW. While RSC gives more detailed chemical properties of groundwater and mechanical characteristics of rock mass, QHLW performs better in evaluating the permeability of rock mass. Besides, RSC is a qualitative or semi-qualitative method with some site-specific criteria, while the QHLW is a quantitative method with higher universality. Based on the above results and conclusions, a new version of QHLW at the tunnel scale was finally proposed by incorporating the advantages of RSC.

Measurement of thermal expansion coefficient of rock minerals using XRD and its implications to thermal damage mechanism

The measurement of thermal expansion coefficient (TEC) of rocks is of vital importance to various geotechnical engineering applications, such as high-level radioactive waste disposal, enhanced geothermal energy exploitation, ultra-deep borehole stability, liquid nitrogen storage, etc. However, previous researchers mainly focused on the thermal expansion coefficient of various rocks, instead of the constituent minerals. In this manuscript, we measured the both temperature-dependent volume and linear thermal expansion coefficients of two rock-forming minerals (quartz and albite) using both high-temperature X-ray diffraction (XRD) as well as low-temperature XRD. It is found that there exist strong anisotropic linear thermal expansion coefficients of the two minerals, which can provide strong reference for the relevant engineering applications and numerical simulations.

An optimized potential field geological modeling method for a high-level radioactive-waste geological disposal repository

A three-dimensional geological model that incorporates spatial information efficiently and intuitively is an important basis for the planning, construction, and utilization of high-level radioactive waste (HLW) disposal geological repository sites. However, traditional modeling methods cannot describe extremely complex geological contexts accurately due to the sparsity of effective data. This paper proposed a cokriging gradient interpolation method based on integrated potential fields that combines relationship and orientation data from outcrop descriptions, contour lines, boreholes, and stratigraphic contacts. A multi-stratigraphic sequence was constructed via a curve/surface tracking algorithm, and optimal geological interfaces were screened using universal cokriging equations. The geological modeling for a HLW disposal repository shows that under different data completeness conditions, this method can reveal the spatial distribution of geological structures (including fault networks and intrusive bodies), achieve better approximations for complex geological environments, and provide strong support for HLW repository roadway planning.

Experimental Study on Ultrasonic Velocity, Suction Curve, and Gas Permeability of Damaged Granite

The excavation damaged zone (EDZ) of a high-level radioactive waste (HLW) disposal repository is a key factor related to long-term safety. The essence of the damage evolution of rocks is the development of microcracks, which change the ultrasonic velocity, suction curve, and gas permeability of rocks. In this work, some granite samples with different damage degrees were generated by uniaxial compression, which is similar to the stress condition of the surrounding rocks of HLW disposal tunnels. The ultrasonic velocity was simultaneously tested during the compression process. The suction curve was then obtained by drying the damaged samples under different levels of relative humidity. The gas permeability was finally measured at different levels of water saturation. Test results indicated that the unstable development of microcracks led to a decrease in Young’s modulus and enlarged the horizontal deformability of granite. The development of microcracks did not change the longitudinal wave velocity, which tended to remain constant when the ratio of the axial stress to the crack damage stress was greater than one. However, the shear wave velocity of damaged granite was significantly decreased. Such effect indicated that the microcracks generated at the stable and unstable development stages mainly extended along the axial direction rather than the radial direction. The suction curve of damaged granite was obtained and was found to comply with van Genuchten’s model and the dual-porosity model. The results of the permeability tests revealed that almost no new gas pathways were generated at the stable development stage of the microcracks. Meanwhile, the unstable development of microcracks led to a great increase in gas permeability. Moreover, the gas permeability of the intact and damaged granite tended to be constant when the water saturation level was less than 0.5. These results will benefit the performance assessment of the EDZs in the surrounding rocks of HLW disposal repositories.

Preliminary analysis on excavation stability of Beishan Underground Research Laboratory

According to construction requirements of the first underground research laboratory (URL) for high-level radioactive waste (HLW) disposal in China, preliminary analysis on the excavation stability was carried out. Based on in-situ stress conditions and preliminary design of the URL, excavation induced stresses were revealed by numerical simulation. Then, through coordinate conversion, the excavation induced stresses were converted into local coordinate system perpendicular to the tunnel. Finally, an analytical method that compares rock strength with stress conditions is used to evaluate the excavation stability of surrounding rock in the local coordinate system. Results of this study show that: due to the high strength of Beishan granite and low level of the regional in-situ stress, the Beishan URL is highly likely to remain stable after excavation, the surrounding rock can hardly be destroyed by both tensile and compressive stress. Relatively unstable parts of the URL are the bottom of the shaft, rotation parts of the -560m-platform and some corners. These parts should be paid attention to in the future excavation process.

Investigations of the mechanical behavior and failure mechanism of fractured rock samples

As excavation is often very deep with large unloading forces in the construction of a high-level radioactive waste (HLW) disposal repository, microcracks in the surrounding rock are unavoidable. In this work, a series of uniaxial compression tests were conducted on intact samples and samples with fractures at different inclination angles (30°, 45°, and 60°), to investigate the mechanical behavior and failure mechanism of fractured rock samples. According to the experimental results, a more inclined angle of the fractures would dramatically reduce the compressive strength of the sample. After the compression tests, the granite samples with different fracture inclinations have different failure modes and distribution characteristics of cracks on sample outside surfaces and the premade fracture planes. Higher dip angles of the premade fractures have less surface density of cracks on the sample outside surface after compression. The instability failure of granite samples with 60° fracture inclination is entirely due to shear deformation between the premade fractures. Based on this single failure mode, the compressive strength of a granite sample with 60° fracture inclination has a positive correlation with the difference in fractal dimensions induced by compression. The greater the difference in fractal dimensions, the higher the compressive strength.

Experimental study on gas permeability of granite close to natural joint

The metal corrosion, organic degradation and waste radiolysis in the geological disposal repository of high-level radioactive waste (HLW) will lead to gas generation and form additional pressure, which will drive the nuclide migration and thereby influence the long-term safety of geological disposal. It is crucial to study the gas permeability of the host rock for the geological disposal of HLW. Therefore, the gas permeability of some granite samples close to a natural joint in the Beishan pre-selected area for China’s geological disposal project was studied in this paper. Firstly, the effective gas permeability under different confining pressures, different gas pressures and different directions was measured by the quasi-steady method. Secondly, the absolute gas permeability and slip factor were obtained according to Klinkenberg’s law. Finally, the effect of confining pressure, direction and distance to the joint surface on gas permeability were analyzed. The gas seepage in granite is found to conform to Klinkenberg’s law. The absolute gas permeability is gradually decreased with the increase of the confining pressure. The gas permeability of granite is isotropic and homogeneous because the intrinsic permeability measured in different directions and at different distances from the joint is found to be similar. The above results can provide references for the performance assessment of the surrounding rock of the HLW disposal repository.

Cooperation between microporous frameworks and micron-sized channel in crystals for excellent chromate removal

The vitrification process represents an effective method for high-level radioactive waste disposal, but CrO42- can form crystalline spinels within the glass to cause the secondary leak of radioactive waste. Porous materials are widely used for trapping CrO42-, but the unsatisfied adsorbing capacity and poor selectivity of reported materials make CrO42- enrichment to be a challenging issue. Herein, a rarely hollow single-crystalline MOF (1) with ordered macro-micropores of 13 μm and 8.6 Å was obtained, exhibiting high N2 adsorption of 339.2 cm3/g and large BET surface area of 1150 m2/g. Cooperation effect between microporous frameworks and micron-sized channels in crystals endows 1 with excellent chromate removal capacity of 609 mg/g, which is a record value known to date among the anion-exchange materials for CrO42- removal. The kinetic analysis revealed that such adsorption was a pseudo-second-order adsorption process. Framework 1 possessed high pH stability and irradiation resistance, can repeatedly absorb/desorb CrO42- to some degree. More importantly, sample 1 can maintain its high CrO42- adsorption capacity after 400 kGy β irradiation, offering the promising application for CrO42- disposal in high-level radioactive waste.

Stability Analysis of Host Claystone under T-M Coupling of HLW Disposal Repository in China: Experiments and Numerical Simulation

This paper focuses on the stability of the high-level radioactive waste (HLW) in the proposed clay rock in Tamusu area of China. The in-situ stress as well as the variational characteristics of ambient temperature caused by nuclide decay during HLW storage should be noticeable. A series of thermal property tests and thermo-mechanical coupled strength (T-M) tests of rock samples in the target formation are carried out. Then the stability of surrounding rock of an HLW under the combination of heat release from HLW and in-situ stress is simulated and analyzed by numerical method. Thermal properties of Tamusu clay rock samples are obtained by testing their thermal conductivity. In order to obtain the characteristics and the failure modes of rock samples at different temperatures, the T-M coupling experiments in the temperature range of 100°C are conducted. Numerical model for simulating the state of operation of the nuclear waste tank buried in the tunnel within 100 years is constructed. A thermal boundary by the heat release equation of HLW and the real in-situ stress level in Tamusu area are considered in the model. While, the variation law of surrounding rock’s temperature, stress, and deformation corresponding to the embedding time is obtained from the numerical calculation. Finally, the stability of the deep geological repository is comprehensively evaluated. The results show that the temperature has a significant impact on the T-M coupling characteristics of Tamusu clay rock, and the proposed repository numerical model has no large deformation and failure problems in 100 years. However, the temperature of the surrounding rock of the repository may exceed the safety standard value during the operation period.

Analysis of double-layered buffer in high-level waste repository

Buffer is the main component of the engineering barrier system (EBS) in high-level radioactive waste (HLW) disposal repository. Reliable EBS performance assessment requires the delicate numerical modeling of the buffer layer. This study presents the sensitivity analysis of thermal and hydraulic parameters in the repository by using COMSOL Multiphysics, and proposes a new buffer barrier with double layers consisting of buffer-A with higher heat conductivity closing to the HLW canister and buffer-B with lower hydraulic conductivity closing to surrounding rock. From results of the sensitivity analysis, it is concluded that the thermal conductivity and hydraulic conductivity are high sensitivity parameters, and specific heat capacity, water density and water compressibility are low sensitivity parameters. The increase of the thermal conductivity in buffer-A results in effectively reducing the peak temperature in the buffer. Therefore, the double-layer buffer is advantageous and effective for improving the repository safety.

The Effect of Porosity Change in Bentonite Caused by Decay Heat on Radionuclide Transport through Buffer Material

Bentonite is used as a buffer material in most high-level radioactive waste (HLW) repository designs. Smectite clay is the main mineral component of bentonite and plays a key role in controlling the buffer’s physical and chemical behaviors. Moreover, the long-term functions of buffer clay could be lost through smectite dehydration under the prevailing temperature stemming from the heat of waste decay. Therefore, the influence of waste decay temperatures on bentonite performance needs to be studied. However, seldom addressed is the influence of the thermo-hydro-chemical (T-H-C) processes on buffer material degradation in the engineered barrier system (EBS) of HLW disposal repositories as related to smectite clay dehydration. Therefore, we adopted the chemical kinetic model of smectite dehydration to calculate the amount of water expelled from smectite clay minerals caused by the higher temperatures of waste decay heat. We determined that the temperature peak of about 91.3 °C occurred at the junction of the canister and buffer material in the sixth year. After approximately 20,000 years, the thermal caused by the release of the canister had dispersed and the temperature had reduced close to the geothermal background level. The modified porosity of bentonite due to the temperature evolution in the buffer zone between 0 and 0.01 m near the canister was 0.321 (1–2 years), 0.435 (3–10 years), and 0.321 (11–20,000 years). In the buffer zone of 0.01–0.35 m, the porosity was 0.321 (1–20,000 years). In the simulation results of near-field radionuclide transport, we determined that the concentration of radionuclides released from the buffer material for the porosity of 0.321 was higher than that for the unmodified porosity of 0.435. It occurs after 1, 1671, 63, and 172 years for the I-129, Ni-59, Sr-90, and Cs137 radionuclides, respectively. The porosity correction model proposed herein can afford a more conservative concentration and approach to the real release concentration of radionuclides, which can be used for the safety assessment of the repository. Smectite clay could cause volume shrinkage because of the interlayer water loss in smectite and cause bentonite buffer compression. Investigation of the expansion pressure of smectite and the confining stress of the surrounding host rock can further elucidate the compression and volume expansion of bentonite. Within 10,000 years, the proportion of smectite transformed to illite is less than 0.05%. The decay heat temperature in the buffer material should be lower than 100 °C, which is a very important EBS design condition for radioactive waste disposal. The results of this study may be used in advanced research on the evolution of bentonite degradation for both performance assessments and safety analyses of final HLW disposal.

A review on the impact of thermal history on compacted bentonite in the context of nuclear waste management

A deep geological repository is the most sustainable way to dispose of high-level radioactive waste. It consists of a multi-barrier system with waste containing canister, sealing material, and geology that avails a facility to provide the long-term isolation of these harmful wastes. The bentonite or sand bentonite blocks serve as a sealing material in the DGR around the waste canisters to isolate them from the surrounding atmosphere. Buffer surrounding the waste canister experiences variation in the temperature for a long duration due to the continuous reduction in the radioactivity and the temperature emitted by the element with time. The thermal flux is induced on the compacted bentonite when a high-temperature waste canister comes in contact with the bentonite buffer and due to seasonal temperature change resulting from a heat source for a long time. As the temperature inside the deep geological repository varies with a long duration of time, it creates a thermal history on the bentonite buffer and changes its physicochemical and hydromechanical properties affecting its overall performance to serve as a sealant. Past reported studies had mentioned the influence of temperature on the various bentonites, but the effect of thermal history on bentonite efficiency is rarely reported. This study presents a review of existing knowledge about the high-temperature impact on bentonite buffer behavior. It has critically investigated the impact of temperature on the physicochemical, hydromechanical, and mineralogical characteristics of bentonite. Further, the analyses are made on the influence and importance of thermal history to be considered while designing the buffer material from few recently reported studies. The physicochemical and hydromechanical properties such as Atterberg’s limits or swell pressure reduced when subjected to heat. It further reduced when time also came into the picture along with the temperature, i.e., thermal history. Such research will be very helpful and will help better understand the different physicochemical and hydromechanical reactions in a bentonite buffer before its designed values are established as part of the engineering barrier systems in the deep-geological repository.

Swelling pressure of compacted MX80 bentonite/sand mixture prepared by different methods

Compacted bentonite/sand mixture have been considered as possible sealing/backfilling material in the deep geological disposal of high-level radioactive nuclear waste. The swelling pressure of the compacted bentonite/sand mixture is of significance in the design of the geological repository, which requires good consistency between data obtained by laboratory and field measurement. In this work, a series of swelling pressure tests were performed on compacted MX80 bentonite/sand specimens prepared by methods commonly adopted in laboratory (As-compacted, Transferred) and those mimicking the real block manufacturing process (Trimmed, Inserted). Results shown that with identical dry density (especially when the dry density was larger than 1.70 Mg/m3), largest swelling pressure was obtained in specimens prepared by method Inserted, followed by method As-compacted, Transferred and Trimmed. The distinct difference between the swelling pressure could be largely attributed to the effects of residual post-compaction lateral stress. More interestingly, specimens prepared by methods Trimmed and Transferred followed a similar swelling pressure-dry density relationship. From this point of view, method Transferred without causing mass loss, change of bentonite content and possible technological voids effect was recommended in lieu of method Trimmed for specimen preparation in laboratory.

MODELLING THE HORIZONTAL VELOCITY FIELD OF THE NIZHNE-KANSK MASSIF ACCORDING TO GNSS OBSERVATIONS

Within the boundaries of the Nizhne-Kansk granite-gneiss massif, which directly borders on the Atamanovskiy branch of the Yenisei Ridge, the building of an underground research laboratory for validating the safety of disposal of high-level radioactive waste began in 2019. In 2010, researchers of the Mining and Chemical Combine at Zheleznogorsk and the Geophysical Center, Russian Academy of Sciences, organized a satellite geodetic network within the boundaries of the Nizhne-Kansk massif; this network included 30 GNSS stations intended for observations of modern crustal movements.The purpose of this study is to determine vectors and simulate the field of horizontal modern crustal movements from measurements made in 2010 - 2019. The tasks included: creating a catalogue of displacement data; calculating and estimating horizontal velocities of modern crustal movements, modelling the horizontal velocity field using artificial neural networks, developing a kinematic model of the area and comparing it with geological survey data.As a result, the resulting model was found to be in good agreement with the results of structural-geological and geodynamic studies in the area. The rate of convergence between the Siberian Platform and the West Siberian Plate in the interaction zone of the southern part of the Yenisei Ridge can be estimated as 2-4 mm/year. The movements of the selected area are due to sublatitudinal compression along an azimuth of 100-110 degrees. Within the selected tectonic blocks relatively low rates of modern horizontal movements of the earth's crust were obtained, which confirms the stable geodynamic regime of the structural block hosting the underground research laboratory. Thus, the results of the work demonstrate the possibility of disposal of high-level radioactive waste in the selected structural block.