Solid Radioactive Waste Research Articles

Modeling Cs-137 Migration in Groundwater due to Low Probability Event of Leaching from Near Surface Disposal Facility for Low Level Solid Radioactive Waste

In this paper, we present impact of Cesium-137 (Cs-137) disposal in a Near Surface Disposal Facility (NSDF) with respect to groundwater migration and its consumption for drinking. We employ the Multiple Area Source Model for simulating the fate of radionuclide in groundwater to evaluate its potential contamination due to extremely low probability event of leaching from the NSDF. The assessment conservatively assumes direct interaction between disposal trenches and the underlying aquifer neglecting the barrier provided by the unsaturated zone. Temporal profiles of concentration of Cs-137 at different down flow distances in groundwater are obtained over a period extending several years post-disposal. The results indicate that even under such a worst-case conservative scenario and for an annual disposal rate of 3 TBq/y for a period of 50y, the maximum Cs-137 concentration in groundwater remains significantly below the safe drinking water threshold of 10 Bq/L recommended by the World Health Organization. It is found that amount of Cs-137 generally disposed in NSDF is significantly lower than the allowable disposal limit corresponding to concentration defined for safe drinking water limit.

Deep eutectic solvent microemulsions with abundant hydrogen ions in supercritical CO2 for decontamination of radioactive solid waste: Overcoming roadblocks to hydrogen ions deficiency

Deep eutectic solvents (DESs) dispersed in supercritical CO2 (SC-CO2) as microemulsions have shown promise for the removal of radioactive contaminants from solid surfaces. However, the current technology needs urgent improvement in terms of decontamination performance. This study reported a novel strategy to enhance the decontamination ability of DESs-in-CO2 systems by adding hydrogen ion (H+) supplemental reagents. The type and acidity of the added H+ reagents were found to be critical factors in dictating the decontamination performance. For instance, fluorinated carboxylic acids with higher acidity and higher CO2 affinity were more competitive for decontamination compared to other non-fluorinated carboxylic acids. Additionally, when nitric acid was used as the H+ supplemental reagent for microemulsions in SC-CO2, outstanding decontamination results were achieved. It could remove 99 % of UO3 contaminants on the specimen surfaces. The formation of microemulsions was demonstrated by molecular dynamics (MD) simulations. Meanwhile, low-cost nitric acid was more promising for scale-up applications. Remarkably, the versatility of the decontamination technology with the addition of nitric acid was demonstrated by its ability to completely decontaminate multiple radioactive wastes made of various materials, as well as different simulated radionuclides. Decontamination efficiency reached more than 99.0 %. More importantly, the DESs-in-CO2 systems with nitric acid also reduced the radioactivity of real radioactive metal waste to the clearance level with the decontamination efficiency reaching 99.0 %. Overall, this study provided valuable insights into the design of advanced decontamination technologies based on microemulsions of DESs in SC-CO2 and opened new avenues for the sustainable management of radioactive waste.

Modern Features of Radioactive Substance Content Dynamics Simulation in Multi-walled Stores

The world’s sustainable development of nuclear power engineering during last decades resulted in large number of radioactive wastes. Currently, each state having nuclear power engineering systems in possession, utilizes the concept of geological disposal as a long-term strategy of radioactive wastes treatment that requires continuous advancements in methods and means due to wastes and standards of radiation monitoring stringency. The article considers the features of radioactive substance content dynamics simulation for single- and multi-walled solid radioactive waste stores, being the product of uranium production, represented by one-dimensional equations and the second and first order sets of equations. The potentials of the existing models, that take into account complex interactions between radionuclides and environment along with possible sources of pollution and mechanisms of radioactive substance transfer, were reviewed. The data availability of temporary radionuclide content evolution inside the store to assess possible effect on environment and people based on the results of mathematical modelling of the considered numerical expressions was emphasized. It was shown, that current models allow analysing and predicting transfer of radionuclide solutions and other substances due to finding solution to contemporary scientific engineering and environmental problems. It was noted, that nowadays there is no model providing forecast of radionuclide content spatial and temporal variations within various store layers to study the accumulated radioactive wastes revealing the mechanisms of their formation, to analyse the features of their distribution, to determine the areas of their active concentration and to define the radionuclide migration activity during storage.

Research on a novel γ-ray spectrum analysis method for low- and intermediate-level radioactive solid waste in nuclear power plants

Accurate nuclide identification in γ-spectrum analysis of low- and intermediate-level radioactive waste with high-purity germanium detectors necessitates initial forced fitting with a nuclide library, yet inaccuracies in library data may lead to misidentification and missing nuclides. To this end, background clipping strategies were hereby analyzed, and a novel deconvolution spectrum analysis method was proposed, which utilized continuous wavelet transform for peak searching and Gaussian first-order derivative quadratic convolution for calculating peak width. Furthermore, to effectively realize the nuclide identification and peak area calculation, a response filter function model was established through the peak shape calibration. By eliminating the need for nuclide library parameter settings prior to overlapping peak separation, the issue of inaccurate matching arising from reliance on the precision of the nuclide library was addressed. Moreover, spectrum analysis experiments were carried out on standard point sources and 200 L drums, and the results were compared and analyzed using GammaVision. Compared to the GammaVision results set by the accurate nuclide library, the area error of strong peaks decreased from 27.5 % to 4.82 %, while that of weak peaks witnessed a decline from 49.98 % to 27.5 %. Finally, the accuracy of the proposed method was verified using the Pakistan Nuclear Library.

Advanced strategies for conversing and fixing sulfur during waste resins oxidation through Ca/Zn oxides enhanced Na2CO3-K2CO3 system

The effective reduction of sulfur pollutant emissions during the treatment of waste resins has always posed a challenge. This study put forward a solution to using the metallic oxide enhanced Na2CO3-K2CO3 system for the treatment of waste resins (CERs). Na2CO3-K2CO3 was used as a medium for heat transfer and desulfurization, while CaO or ZnO functioned as catalysts in the sulfur conversion process. The thermodynamic analysis substantiated the reliability of enhanced sulfur fixation by CaO and ZnO. H2S, SO2 and COS primarily generated during the destruction stage of -SO3H (280–500°C) and residues (∼700°C), which were largely absorbed through a vapor-liquid-solid reaction to generate CaCO3, ZnS and K2SO4. CaO improved the conversion of sulfur to sulfate and effectively mitigated the excessive emissions of CO2. SO2 generated by the oxidation of ZnS at about 800 °C could be further absorbed by molten Na2CO3-K2CO3 to form K2SO4. The proposed method exhibited in-situ desulfurization capability and effectively inhibited the sulfur migration to gaseous products. At 800°C, the sulfur retention rate, the formation rate of SO42- and the oxidation efficiency of CERs reached 91.39 %, 80.61 % and 99.99 %, respectively. The existence of CaO and ZnO facilitated the inorganic transformation of organic sulfur, and the efficient conversion of organic sulfur into inorganic compounds was achieved. This approach provides a new opportunity to improve the treatment efficiency of radioactive solid waste containing organic sulfur.

Application and mechanism of carbonate material in the treatment of heavy metal pollution: a review.

Among the many heavy metal pollution treatment agents, carbonate materials show strong flexibility and versatility by virtue of their high adsorption capacity for heavy metals and the characteristics of multiple and simple modification methods. It shows good potential for development. This review summarizes the application of carbonate materials in the treatment of heavy metal pollution according to the research of other scholars. It mainly relates to the application of surface-modified, activated, and nano-sized carbonate materials in the treatment of heavy metal pollution in water. Natural carbonate minerals and composite carbonate minerals solidify and stabilize heavy metals in soil. Solidification of heavy metals in hazardous waste solids is by MICP. There are four aspects of calcium carbonate oligomers curing heavy metals in fly ash from waste incineration. The mechanism of treating heavy metals by carbonate in different media was discussed. However, in the complex environment where multiple types of pollutants coexist, questions on how to maintain the efficient processing capacity of carbonate materials and how to use MICP to integrate heavy metal fixation and seepage prevention in solid waste base under complex and changeable natural environment deserve our further consideration. In addition, the use of carbonate materials for the purification of trace radioactive wastewater and the safe treatment of trace radioactive solid waste are also worthy of further exploration.

Innovative deep decontamination of radioactive solid surfaces using microemulsions of deep eutectic solvents in supercritical carbon dioxide

Radioactive solid wastes resulting from the application of nuclear energy pose a risk to the environment, yet traditional decontamination methods generate many secondary wastes. Thus, sustainability-driven innovations are receiving increasing attention to achieve eco-friendly decontamination methods characterized by minimizing secondary waste generation. Here, we proposed a sustainable and pioneering approach to decontaminate radioactive solid wastes using microemulsions of deep eutectic solvents (DESs) in supercritical carbon dioxide (SC-CO2). We also determined the optimal ratios of the additives as follows: prepared DESs with a molar ratio of choline chloride (ChCl) to p-toluenesulfonic acid monohydrate (ptsa) of 1:2, and nonyl phenol polyoxyethylene ether (NP-10) to ChCl:ptsa DESs in a molar ratio of 0.8. Remarkably, after optimization of the decontamination process in terms of pressure, temperature and cleaning time, the radioactivity on the sample surface can be reduced to a level that is reusable in the nuclear field. Detailed analyses provided insights into the intermolecular interactions within the DESs. And the decontamination mechanisms involved the spontaneous evolution of microemulsions with ChCl:ptsa as the core to facilitate the solvation and removal of radioactive contaminants. Additionally, repeated cleaning was proved to be effective for ultra-deep decontamination, with the radioactivity on the sample surface reaching a clearance level and a decontamination efficiency of 99.8%. The DESs-in-CO2 systems successfully decontaminated various materials and simulated radionuclides, even real radioactive metal wastes from the nuclear industry. It was notable that the DESs-in-CO2 systems achieved decontamination efficiencies of more than 95.0% for radioactive contaminants on these solid surfaces. Overall, this innovative approach provided a green alternative to conventional radioactive solids decontamination, meeting the need for sustainable development of nuclear energy.

URANIUM BEHAVIOR DURING THE FORMATION AND SUBSEQUENT STORAGE OF SOLID RADIOACTIVE WASTES FROM CONVERSION PRODUCTION

A complex of physicochemical studies of solid radioactive wastes (RW) of the near-surface storage facility from conversion uranium production was carried out. It has been established that the composition of the wastes is represented mainly by calcium compounds: gypsum, calcite, and calcium fluoride. Uranium is distributed throughout the entire volume of wastes. At the same time, uranium is unevenly concentrated along the depth of the sludge. The content of uranium in the wastes is 0.005-0.65 wt. %. Uranium was predominantly adsorbed on the surface of sludge by main components during the formation of the wastes solid phase. Uranium was partially precipitated in the form of calcium-containing weeksite Ca2(UO2)2(Si2O5)3·10H2O, uranophane-alpha Ca(UO2)2(SiO3OH)2·5H2O, and urankalcarite Ca(UO2)3CO3(OH)6·3H2O. The isotopic composition of uranium contained in the storage wastes corresponds, within the error, to the normal isotopic composition of natural uranium. Near-surface storage is constantly exposed to natural factors. Infiltration of atmospheric condensation leads to leaching of some components from solid RW. At the same time, uranium goes into solution in the form of [(UO2)2CO3(OH)3]-, [UO2(CO3)2]2-, [UO2(CO3)3]4- and Ca2UO2(CO3)3. The subsequent migration of uranium with infiltrating water contributes to its distribution within the entire sludge field, including the reclamation soil layer and the near-surface pond. In some areas of the storage facility, the uranium content in the reclamation soil layer reaches 0.043 wt.%. The concentration of uranium in the pond was 17.97 mg L-1. The chemical composition of the pond water was characterized by a high content of chloride-ions (1.29 g L-1), calcium (630 mg L-1), sulfate-ions (224 mg L-1), hydrocarbonates (122 mg L-1). The pond is connected to groundwater, which increases the risk of pollutants escaping beyond the storage facility. For citation: Skripchenko S.Yu., Titova S.М., Nalivaiko K.A., Semenishchev V.S. Uranium behavior during the formation and subsequent storage of solid radioactive wastes from conversion production. ChemChemTech [Izv. Vyssh. Uchebn. Zaved. Khim. Khim. Tekhnol.]. 2024. V. 67. N 5. P. 77-86. DOI: 10.6060/ivkkt.20246705.6925.

Cementation of spent radioactive ion-exchange resin ashes using alkali-activated cements: Physicochemical and structural changes

Spent ion-exchange resins (IERs) constitute the major volume of solid radioactive wastes generated by the nuclear industry and Portland cement wasteforms exhibit limitations for their conditioning and storage. In this paper, an alternative approach is conducted involving the solidification of IER ashes (IERAs) through the use of “one-part” alkali-activated cements (AACs). Systems with different chemistry based on blast furnace slag (BFS) and metakaolin (MK) were formulated and the physicochemical and structural changes of the gel products after incorporating IERAS were investigated. The incorporation of the IERAs reduced the amount of C-A-S-H and (N,C)-A-S-H products while changing the pore structure and microstructure, reducing the mechanical strengths of the cement matrices; these effects were more remarkable in the high BFS systems. New insights are elucidated on the microstructural changes of the gel phases and suggest that AACs with BFS/MK ratios of 25/75 and 50/50 are promising candidates for conditioning IERA wastes.

Institutional Experience of Routine Radiation Surveillance of Delay and Decay Tanks Facility in a Department Having High-dose Iodine Therapy Unit.

Radioactive solid and liquid waste generated by patients after high-dose iodine therapy may lead to significant radiation exposure if not properly handled. This study was conducted to monitor the radiation exposure along the sewerage drainage system of the high-dose iodine therapy ward and to rule out leakage if any, that might pose a potential radiation hazard to the general public (sewerage workers) and radiation health professional. The sewerage drainage system from isolation wards has multiple gate valves to regulate sewerage flow from the high-dose iodine therapy ward into delay and decay tanks (DDT) built, especially for the purpose. Radiation surveillance was done using a Geiger-Muller counter-based survey meter at 11 different locations on a weekly basis for 12 weeks. A total of 26 patients underwent high-dose iodine ablation therapy during the study period in our department, with the highest recorded radiation exposure rate in the sewerage draining system in the 9th week of patient admission. This was at the common gate valve junction (location B) that directed sewerage waste from all four isolation rooms into the common pipeline leading to DDT. Minimal radiation exposure was recorded within Atomic Energy Regulatory Board -prescribed limits with no evidence of leakage. A routine radiation survey is an important component of overall radiation safety in the nuclear medicine department, including sewerage delay tank facilities, which helps keep the radiation exposure to acceptable levels by identifying timely leakage.

Solidification of Radioactive Wastes Using Recycled Cement Originating from Decommissioned Nuclear-Energy Facilities

Hundreds of thousands of tons of waste are generated from decommissioned nuclear- power facilities, and it has become a critical global issue to secure technology for reducing and recycling this waste. Concrete waste (CW) is estimated to comprise 60–80% of the total waste, and concrete-waste powder (CWP) includes enough inorganic substances used as effective materials for waste treatment. Accordingly, it can be used to produce recycled cement (RC). This study aimed to evaluate the performance of a solidification agent manufactured using recycled cement (SRC) for the safe packing of radioactive wastes, such as coarse aggregates of CW, waste soil, and metal wastes originating from decommissioned nuclear facilities. The experimental results indicated that the most relevant incineration temperature of CWP for RC was 700 °C. The optimum water-to-binder ratio was determined to be 0.4, and the most relevant substitution ratio of ground granulated blast furnace slag for CWP was determined to be 15%. In addition, calcium silicate hydrate is the most effective hydration product for improving the compressive strength of SRC. The maximum packing capacities of the SRC for coarse aggregates, waste soil, and metal waste, which were simulated as radioactive wastes, were determined to be 30, 5, and 7 wt%, respectively. The results of leaching tests using SRC containing radioactive wastes contaminated with Co, Cs, and Sr indicated that their leachability indices met the acceptance level for disposal. Consequently, the RC composed of CWP can be used as a solidifying agent to safely dispose of radioactive wastes, such as coarse aggregates, waste soil, and metal waste.

РАДИОЭКОЛОГИЧЕСКАЯ ОБСТАНОВКА В РАЙОНЕ ПЛОЩАДКИ УРАНОВОГО НАСЛЕДИЯ – ШАХТА «СТЕПНАЯ» (КАЛМЫКИЯ)

Purpose: To study the radio-ecological situation on the “uranium legacy” site of the former Stepnaya mine in the Republic of Kalmykia. Material and methods: To measure the ambient dose equivalent rate (ADER), the pedestrian gamma survey method was used using a portable spectrometric complex MKC-01A Multirad-M and dosimeter-radiometer MKC-AT6101c. The activity of gamma-emitting radionuclides in soil samples was measured using a stationary gamma spectrometer from CANBERRA. The activities of 210Po and 210Pb were measured using a radiometric installation UMF-2000 following their radiochemical separation from samples. Short-term measurements of activity concentration (AC) and equivalent equilibrium activity concentration (EEAC) of radon were carried out with an aerosol alpha radiometer for radon and thoron RAA-20P2 Poisk. Doses of radiation exposure to biological objects were estimated using dose coefficients provided by ICRP Publication 136 taking into account recommendations R52.18.820-2015. Results: Gamma ADER values at the mine site vary over the range from 0.1 to 0.36 µSv/h, and on 80 % of the area these values do not exceed the background value of 0.14 µSv/h. Along the road from the mine to Narta village the ADER values do not exceed background values with exception of the area around the dam, where in a local part of this area of about 300 m2 these values reach 0.49 µSv/h. The specific activities of natural radionuclides in the soil are below the criteria for classification as solid radioactive waste (SRW). Under the certain weather conditions, radon EEAC inside the buildings on the site reaches 13 kBq/m3, and on the territory 1-1.5 kBq/m3. Ecological risk for the terrestrial biological objects under consideration (grass, soil worm, snake and mouse-like rodents) does not exceed 0.025. Conclusions: The radiation situation at the Stepnaya mine site meets the requirements of SP LKP-91, which were in force until 2020. However, in order to transfer the facility to a local government body, reclamation work should be carried out in accordance with the Federal Law “On the Transfer of Lands or Land Plots from One Category to Another” dated December 21, 2004 No. 172-FZ and GOST R 59057— 2020 «Environmental Protection. Lands. General Requirements for Reclamation of Affected Lands». Doses of exposure to biological objects do not impact significantly on morbidity, reproduction and life expectancy of terrestrial biological objects.

Application of vacuum membrane distillation-Fenton oxidation process for deep purification of low-level radioactive organic wastewater

During the production of UO2 kernel for the high temperature gas-cooled nuclear reactor, a large amount of low-level radioactive organic wastewater (LLROW) containing ammonia, uranium and organic compounds are generated. The current treatment process is suffered from the production of radioactive solid wastes, low adsorption efficiency of organic matters and the frequent membrane fouling during RO process. In this study, vacuum membrane distillation (VMD)-Fenton oxidation process was developed for the deep purification of the LLROW. The process parameters were optimized and the system performance was evaluated. As a result, all major contaminates such as uranium, NH3-N, TN and COD could meet the GB8978-1996 (secondary standard) and GB/T31962-2015 (B standard) for direct discharge. The membrane fouling was also studied by membrane autopsy as well as the quantum calculation. It was found that weak hydrogen bonds were formed between PTFE membrane and contaminants while the membrane fouling could be recovered by 0.1 M HCl cleaning.

Study of γ-Ray Spectrum Measurement Analysis Algorithm of Radioactive Solid Waste Steel Boxes

Abstract In accordance with safety requirements for the near-surface disposal of low and medium-level radioactive solid waste, accurate identification and measurement of radionuclide types and activities within steel boxes containing such waste are necessary. Conventional sampling techniques are impractical due to nonuniform internal filling and radioactive distribution in the steel boxes used for storage. Nondestructive measurement analysis techniques offer significant advantages by providing comprehensive measurements without causing damage to the steel boxes or generating secondary waste. Therefore, nondestructive techniques are widely applicable and superior to chemical analysis methods. This study presents an algorithm for analyzing γ-ray spectrum measurements of radioactive solid waste steel boxes by combining the traditional computed tomography (CT) principle with a discrete treatment approach. The progressive split voxel measurement method is utilized to discretely handle the radioactive solid waste steel boxes during the measurement and calculation process while considering the impact of voxel interactions on detection efficiency. A detection efficiency calibration model is established using Monte Carlo (MC) simulation in conjunction with the algebraic reconstruction technique (ART). Additionally, an inversion correction of the efficiency for the physically coincident portion of the detected voxel volume is performed during the measurement process. The proposed algorithm is initially applied to the analysis platform of steel boxes that store FA-IV-type radioactive solid waste. The algorithm's performance is further investigated and validated using standard radioactive sources with known activity levels. The results indicate that the relative deviation between the reconstruction results and the nominal values of radioactive activities in the steel boxes is less than 30%. These findings align with the expected results and satisfy the measurement analysis and warehousing processing requirements for waste generated from the decommissioning of China's 101 nuclear reactor. The proposed algorithm holds great potential for widespread application in related detection work.

О ВОЗМОЖНОЙ РАДИАЦИОННОЙ АВАРИИ НА АТОМНОМ ЛЕДОКОЛЕ И ОРГАНИЗАЦИИ ОБЕСПЕЧЕНИЯ БЕЗОПАСНОСТИ СПАСАТЕЛЕЙ МЧС РОССИИ ПРИ ЛИКВИДАЦИИ ЕЕ ПОСЛЕДСТВИЙ

Based on the analysis of the presence of radiation hazardous facilities in the Arctic regions of the Russian Federation, the problem of ensuring environmental safety in connection with the disposal of spent nuclear fuel of nuclear icebreakers is considered. Due to the growing number of Russian nuclear icebreakers in the Arctic, the specifics of their use, storage conditions for accumulating radioactive waste on ships serving them, there is a risk of accidents during loading and unloading operations with spent nuclear fuel. A hypothetical model of an accident with the depressurization of a container with solid radioactive waste is considered, and recommendations are given on the organization of radiation reconnaissance and ensuring the safety of personnel of EMERCOM of Russia participating in the elimination of the consequences of radiation pollution, attention is paid to the specifics of decontamination.

Evaluation of the Solidification of Radioactive Wastes Using Blast Furnace Slag as a Solidifying Agent.

The decommissioning process of nuclear power facilities renders hundreds of thousands of tons of various types of waste. Of these different waste types, the amount of concrete waste (CW) varies greatly depending on the type of facility, operating history, and regulation standards. From the previous decommissioning projects, CW was estimated to comprise 60-80 wt.% of the total weight of radioactive wastes. This represents a significant technical challenge to any decommissioning project. Furthermore, the disposal costs for the generated concrete wastes are a substantial part of the total budget for any decommissioning project. Thus, the development of technologies effective for the reduction and recycling of CW has become an urgent agenda globally. Blast furnace slag (BFS) is an industrial byproduct containing a sufficient amount (higher than 30%) of CaO and it can be used as a substitute for ordinary Portland cement (OPC). However, there have been few studies on the application of BFS for the treatment of radioactive waste from decommissioning processes. This study was conducted to evaluate the performance of the solidification agent using ground granulated BFS (SABFS) to pack radioactive wastes, such as the coarse aggregates of CW (CACW), waste soil (WS), and metal waste (MW). The analytical results indicated that the CaO content of the ground granulated BFS was 36.8% and it was confirmed that calcium silicate hydrate (CSH) could be activated as the precursor of the hydration reactions. In addition, the optimum water-to-binder ratio was determined to be 0.25 and Ca(OH)2 and CaSO4 were found to be the most effective alkaline and sulfate activators for improving the compressive strength of the SABFS. The maximum packing capacities of the SABFS were determined to be 9 and 13 wt.% for WC and WM, respectively, when the content of CW was fixed at 50 wt.%. The results of the leaching tests using SABFS containing radioactive wastes contaminated with Co, Cs, and Sr indicated that their leachability indices met the acceptance level for disposal. Consequently, the SABFS can be used as a solidifying agent for the safe disposal of radioactive waste.

Fiber-supported layered magnesium phosphate exhibits high caesium(I) capture capacity in water

Solid-supported adsorbents used for the sequestration of radiocaesium, a major hazardous nuclear fission byproduct, from water are limited by low uptake capacities, increasing the difficulty of radioactive solid waste reduction. This study dispersed potassium magnesium phosphate monohydrate (KMP), a promising Cs+ adsorbent, in polyacrylonitrile fibers (PANFs) to efficiently adsorb Cs+. The larger size of KMP compared to PANFs is conjectured to result in the exposure of some KMP particles, thereby minimizing any decline in their adsorption capacity. The KMP/PANF composite exhibited an adsorption capacity per unit mass of KMP that was nearly equal to that of the bare KMP powder, which was less than 3.52% on average. The composite adsorbent with 100 wt% KMP loading demonstrated the highest experimental capacity of 299 mg g–1 among all solid-supported adsorbents reported to date. It also exhibited relatively fast uptake kinetics (equilibration time = 1.0–4.0 h) and a high distribution coefficient (average value of 1.22 × 104 mL g–1 over pH 5.0–11). Overall, the newly developed KMP/PANF adsorbent exhibits great potential for the practical decontamination of radiocaesium-contaminated waters.

CONTAINER STORAGE AS AN ELEMENT OF ENHANCEMENT OF ENVIRONMENTAL SAFETY AND A STATE POLICY TOOL FOR DEALING WITH TECHNOLOGICALLY ENHANCED SOURCES OF IONIZING RADIATION

In order to implement the state environmental policy and ensure environmental safety, a number of normative legal acts, laws and state standards have been implemented on the territory of Ukraine. Special attention was paid to the State Targeted Environmental Program for the Management of Radioactive Waste, the main principle of which is the implementation of the protection of the population and the natural environment in order to create safe conditions for the existence of current and future generations and ensure sustainable development. The article highlights the results of research and development works on the development and certification of a set of packaging type IR-2 for the transportation and temporary storage of low- and medium-level solid radioactive waste KTTZ-NS-1-02.00.00, as a key element of environmental protection technologies during handling with radioactive waste. The urgent task of handling radioactive waste, which is caused by the accumulation of a significant amount of sludge in the territories of the oil and gas production complex enterprises and requires temporary storage and transportation, has been solved, as a prerequisite for increasing the efficiency of ensuring the environmental safety of the mentioned objects of formation. The technical characteristics of the packaging kit are described in detail, schematic drawings are provided, and the main parameters and dimensions of the developed packaging kit type IR-2 are provided. The scheme of the main stages of the testing of this packaging set and the certificates obtained as a result of such work, which were carried out jointly with representatives of the State Inspection of Nuclear Regulation, are presented. The developed packaging set significantly exceeds the performance of the packaging set currently used in temporary storage facilities. Keywords: environmental safety, technogenically enhanced radioactive waste, protection technologies, containers, low- and medium-level waste.

An SPS-RS Technique for the Fabrication of SrMoO4 Powellite Mineral-like Ceramics for 90Sr Immobilization.

This paper reports a method for the fabrication of mineral-like SrMoO4 ceramics with a powellite structure, which is promising for the immobilization of the high-energy 90Sr radioisotope. The reported method is based on the solid-phase "in situ" interaction between SrO and MoO3 oxides initiated under spark plasma sintering (SPS) conditions. Dilatometry, XRD, SEM, and EDX methods were used to investigate the consolidation dynamics, phase formation, and structural changes in the reactive powder blend and sintered ceramics. The temperature conditions for SrMoO4 formation under SPS were determined, yielding ceramics with a relative density of 84.0-96.3%, Vickers microhardness of 157-295 HV, and compressive strength of 54-331 MPa. Ceramic samples demonstrate a low Sr leaching rate of 10-6 g/cm2·day, indicating a rather high hydrolytic stability and meeting the requirements of GOST R 50926-96 imposed on solid radioactive wastes. The results presented here show a wide range of prospects for the application of ceramic matrixes with the mineral-like composition studied here to radioactive waste processing and radioisotope manufacturing.

Progress with the regulation of radiation safety during recovery and removal of spent nuclear fuel from the site for temporary storage at Andreeva Bay on the Kola Peninsula.

In the 1960s, a shore technical base (STB) was established at Andreeva Bay on the Kola Peninsula, in northwest Russia. The STB maintained nuclear submarines and the nuclear icebreaker fleet, receiving and storing fresh and spent nuclear fuel (SNF) as well as solid and liquid radioactive waste (RW). It was subsequently re-designated as a site for temporary storage (STS) for SNF and RW. Over time, the SNF storage facilities partly lost their containment functions, leading to radioactive contamination of workshops and the site above permitted values. The technological and engineering infrastructure at the site was also significantly degraded as well as the condition of the stored SNF. At present, the STS Andreeva Bay facility is under decommissioning. This paper describes progress with the creation of safe working measures for workers involved in site remediation and SNF recovery operations, including the determination of safe shift times in high radiation areas, as part of overall optimization of safety. Results are presented for the successful application of these measures in the period 2019-2021, during which time significant SNF recovery and removal operations were completed without incident. Significant important experience has been gained to support safe removal of remaining SNF, including the most hazardous degraded fuel, as well as recovery of other higher level RW and decommissioning of the old storage buildings and structures.