scholarly journals Investigation of the insulating properties of the Cherkasy deposit clays for the creation of underlying screens of radioactive waste at the ‘Vector’ site

2021 ◽  
pp. 71-81
Author(s):  
Borys Shabalin ◽  
Olena Lavrynenko ◽  
Kostiantyn Yaroshenko
Keyword(s):  
Radioactive Waste ◽  
Waste Disposal ◽  
Alkaline Earth ◽  
Bentonite Clay ◽  
Barrier Properties ◽  
Radioactive Waste Disposal ◽  
Near Surface ◽  
Safe Disposal ◽  
Bentonite Clays ◽  
Main Component

The lack of scientifically substantiated requirements, comprehensively developed and approved in a prescribed manner, for the usage of clays as a barrier material poses risks for the safe disposal of radioactive waste in facilities at the ‘Vector’ site for the period of their operation and closure. The bentonite clay from Ukraine’s largest Cherkasy deposit of bentonite and palygorskite clays is considered the most durable as the main component of the insulating (underlying) screens of radioactive waste disposal facilities. The main properties and compositional features of the Cherkasy natural bentonite clay (Dashukovskaya site, layer II) and its variety such as alkaline earth bentonite (activated soda bentonite), which provide isolation of radioactive waste in disposal, are considered. It is shown that the Cherkasy field has good waterproofing and barrier properties, including a high sorption capacity with respect to 90Sr and 137Cs, which is one of the main characteristics that ensure the safe disposal of radioactive waste. The alkaline earth bentonite absorbs 90Sr and 137Cs more efficiently than natural bentonite does. However, 90Sr is sorbed in larger quantities than 137Cs on both types of bentonite. With increasing time of interaction with an aqueous solution, both types demonstrate a redistribution of the mobile (exchangeable) and immobile (non-replaceable) forms of radionuclides. The contribution of the stationary form that does not participate in migration processes also increases. A comprehensive analysis of the bentonite clays of the Cherkasy deposit was carried out, taking into account the significance of recoverable reserves and the potential for improving the technical and economic parameters of clays. Thus, the Cherkasy bentonite clays can be recommended as an additional anti-migration engineering barrier for ground/near-surface facilities for the disposal of radioactive waste. When choosing the type of bentonite clay for use as a barrier in a radioactive waste disposal facility, one could take into account the data published in the article, but the question of applying the bentonite clays of the Cherkasy deposit to ensure the safe disposal of radioactive waste remains to be  further studied.

scholarly journals Peculiarities of 137Cs Sorption/Desorption by Bentonite Clays of Cherkasy Deposit from Groundwater Model Solutions of Radioactive Waste Disposal Facilities at the “Vector” Production Complex

2021 ◽  
Vol 21 (2) ◽  
pp. 78-87
Author(s):  
B. H. Shabalin ◽  
◽  
К. К. Yaroshenko ◽  
S. P. Buhera
Keyword(s):  
Radioactive Waste ◽  
Waste Disposal ◽  
Water Soluble ◽  
Radioactive Waste Disposal ◽  
Groundwater Model ◽  
Static Mode ◽  
Near Surface ◽  
Model Solutions ◽  
Production Complex ◽  
Bentonite Clays

The main feature of bentonite clays is their high sorption capacity with respect to various radionuclides. The study of sorption kinetics of 137Cs was performed in the static mode by natural and industrial soda modified (PBA-20) samples of bentonite clays of Cherkasy deposit of bentonite and paligorskite clays from groundwater model solutions of radioactive waste disposal facilities of “Vector” production complex under various pH and solution mineralisation. The desorption of occluded samples was studied in distilled water and acetateammonium buffer solution. The value of the degree of sorption (S) for 137Cs on the modified samples exceeds 90%, for natural bentonite this indicator is lower (about 83–85%). On both types of bentonite with increasing time of their contact with aqueous solution and pH, there is a redistribution of water-soluble, ion-exchange and fixed forms of radionuclide and the share of the latter, that is not participating in migration processes increases, indicating the ability of bentonites to immobilize effectively for a long time. It is shown that Na-modified bentonite has higher proportion of sorption in fixed form compared to natural one and its application increases the probability of irreversible fixation of migrating radionuclides under non-optimal conditions of sorption (high pH (>11) of water after prolonged contact with cement-concrete components of engineering barriers) and thus increases the environmental safety of the storage facility. It is shown that bentonite clays of the Cherkasy deposit can serve as an effective material for creating anti-migration barriers of I and II stages of surface/near-surface storage facilities for radioactive waste disposal at the “Vector” production complex. At the same time, the issue of practical application of bentonite clays of Cherkasy deposit for accurate predictions of securing radioactive waste disposal of Chornobyl origin requires further study of sorption-desorption properties of bentonite clay with respect to other fission products and actinides

Creation of system of final isolation (disposal) of radioactive waste in the Russian Federation

MRS Advances ◽  
2020 ◽  
Vol 5 (5-6) ◽  
pp. 275-282 ◽  
Author(s):  
Vsevolod Igin ◽  
Victor Krasilnikov
Keyword(s):  
Radioactive Waste ◽  
Waste Management ◽  
Waste Disposal ◽  
Russian Federation ◽  
Radioactive Waste Disposal ◽  
High Level Waste ◽  
Federal State ◽  
Radioactive Waste Management ◽  
Near Surface ◽  
High Level

Abstract:The paper provides generic overview of legal and regulatory framework of radioactive waste management activities held in Russian Federation and national operator responsibilities and accomplishments. It gives a short description of waste classification scheme used and plans for radioactive waste disposal. In particular the paper provides information on the plans of the FEDERAL STATE UNITARY ENTERPRISE "National operator for radioactive waste management" to construct and operate several near-surface disposal facilities for low and intermediate level waste with total capacity up to 550 000 cubic meter. The paper also provides detailed information on the steps of high-level waste disposal program including site-selection, construction phase of the underground research laboratory (URL) near the city of Zheleznogorsk, Krasnoyarsk Region and research program after the construction of the URL. The paper also describes Russian system and state policy in the field of RW management and gives recommendations for future implementers.

How Thick Should Cover Layer be for Radioactive Waste Disposal Facility? The Case of the Yamin Plain, Israel

2017 ◽  
Vol 3 (3) ◽  
Author(s):  
Avraham Dody ◽  
Ravid Rosenzweig ◽  
Ran Calvo ◽  
Eyal Shalev
Keyword(s):  
Radioactive Waste ◽  
Waste Disposal ◽  
Erosion Rate ◽  
Soil Surface ◽  
Radioactive Waste Disposal ◽  
Rain Event ◽  
Infiltration Depth ◽  
Near Surface ◽  
Cover Layer ◽  
Disposal Facility

Two main natural processes which control the migration of radioactive contaminants to the biosphere were studied in the Yamin Plain in order to evaluate the thickness of a cover layer needed for near-surface radioactive waste disposal facility. The first is the natural erosion of the cover layer, and the second is the infiltration during rain and runoff events. The erosion rate of the soil surface was studied by optical stimulation luminescence technique. It was found that during the last 14,000 years, the erosion rate was 0.3 mm/y which are 3 m for 10,000 years. The infiltration depth assessment was based on water content measurements and numerical modeling. It shows that under the most extreme rain event having an equivalent rain of 84 mm, infiltration depth was limited to 4.5 m. Taking into account the two processes, the effective cover layer thickness over 10,000 years should be at least 7.5 m thick.

Analytical Methodology for Optimization of Waste Management Scenarios in Nuclear Installation Decommissioning Process

2009 ◽  
Author(s):  
Matej Zachar ◽  
Vladimi´r Danisˇka ◽  
Ivan Reha´k ◽  
Marek Vasˇko ◽  
Vladimi´r Necˇas
Keyword(s):  
Radioactive Waste ◽  
Waste Management ◽  
Waste Disposal ◽  
Radioactive Waste Disposal ◽  
Management Scenarios ◽  
Planning Period ◽  
Analytical Methodology ◽  
Processing Technologies ◽  
Near Surface ◽  
Nuclear Installation

The nuclear installation decommissioning process is characterized by production of large amount of various radioactive and non-radioactive waste that has to be managed, taking into account its physical, chemical, toxic and radiological properties. Waste management is considered to be one of the key issues within the frame of the decommissioning process. During the decommissioning planning period, the scenarios covering possible routes of materials release into the environment and radioactive waste disposal, should be discussed and evaluated. Unconditional and conditional release to the environment, long-term storage at the nuclear site, near surface or deep geological disposal and relevant material management techniques for achieving the final status should be taken into account in the analysed scenarios. At the level of the final decommissioning plan, it is desirable to have the waste management scenario optimized for local specific facility conditions taking into account a national decommissioning background. The analytical methodology for the evaluation of decommissioning waste management scenarios, presented in the paper, is based on the materials and radioactivity flow modelling, which starts from waste generation activities like pre-dismantling decontamination, selected methods of dismantling, waste treatment and conditioning, up to materials release or conditioned radioactive waste disposal. The necessary input data for scenarios, e.g. nuclear installation inventory database (physical and radiological data), waste processing technologies parameters or material release and waste disposal limits, have to be considered. The analytical methodology principles are implemented into the standardised decommissioning parameters calculation code OMEGA, developed in the DECOM company. In the paper the examples of the methodology implementation for the scenarios optimization are presented and discussed.

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