A technology for thermovacuum drying of spent radioactively contaminated ion-exchange resins was developed. It was shown experimentally that thermovacuum drying yields a product suitable for long-term storage and subsequent disposal. Recommendations were made for using NZK and KMZ concrete and metal casks for holding dry, spent, low-and medium-activity ion-exchange resins without additional solidification. Because the volume of the product obtained is greatly reduced, thermovacuum drying will decrease by at least 15-fold (from 100 to 7) the number of casks required for long-term storage and subsequent disposal of spent ion-exchange resins produced in one year of operation of a single VVER-1200 power-generating unit.The development and implementation of new-generation nuclear power plant designs (AES-2006 and promising new nuclear power plants) require optimization of the approach to reprocessing and storage of liquid and solid radwastes. The main problem is to decrease the volume of solid wastes subject to intermediate storage and final disposal, including the liquid wastes obtained from reprocessing and conditioning.One form of liquid wastes produced during the operation of nuclear power plants are spent ion-exchange resins, which are not reprocessed and are stored in the form of pulps in metal casks. These wastes must be reprocessed and conditioned. The main technology under consideration for conditioning ion-exchange resins is inclusions of the resins in a solid matrix, for example, cement [1, 2]. A facility for cementing the filtering materials which have accumulated over more than 30 years of operation of power-generating units is under construction at the Leningradskaya nuclear power plant. Cementing of spent ion-exchange resins is also included in the AES-2006 designs. Spent resins as well as the bottoms of an evaporation facility are supposed to be immediately solidified. The AES-2006 design provides a facility for cementing spent ion-changers in non-retrievable NZK-150-1,5P protective casks.For all the advantages of cementing as a method of conditioning spent ion-exchangers, there are drawbacks, one of which is the several-fold larger volume of the conditioned product. In reality, 10-15% ion-exchanger is incorporated into cement [1] and the maximum attainable inclusion does not exceed 25% [2]. The casks with the cement compound must be stored for a long time on site in a solid-wastes repository at the nuclear power plant followed by disposal at regional points, which increases considerably the cost of these stages in the handling of ion-exchange resins. According to design data from the St. Petersburg State Research and Design Institute ATOMENERGOPROEKT (SPbAEP), the yearly volume of mediumlevel ion-exchange resins from one unit of the Leningradskaya NPP-2 will be 15 m 3 , the volume of low-level resins will be 10 m 3 , and the specific activity of both will be 100 and <0.37 GBq/m 3 , respectively. The radionuclide composition of the medium-level ion-exchange resins is as follows, %: 60
There are a large number of works (specifically, [1][2][3][4][5]) concerning methods for calculating the concentration of contaminating substances, propagating with ground waters, outside the pollution sources: holding ponds, disposal sites for tailings, contaminated water-bearing strata. However, there are no works, either here or abroad, on processes occurring simultaneously in the source of pollution and in its environment as a whole.The present paper fills this lacuna from both the methodological and computational standpoints. We present here a mathematical model and a numerical algorithm for solving the problem of the escape of radionuclides from containers with solidified radioactive wastes in a surface disposal site and their propagation along the site and the lower-lying layer of rock down to the surface of the water-bearing stratum located beneath the disposal site. Interstitial water, which enters the disposal site through the top insulating screen and the cover with the precipitation and snow melt, is the main agent which transports the radionuclides.Physical Model and Sehematization of the Conditions. The object of investigation is a surface disposal site for radioactive wastes in activity groups I and II. Concrete containers in the form of cylinders or rectangular parallelepipeds with solidified or solid wastes are placed here. The wastes themselves can be packed in steel barrels. The storage site consists of a square iron-concrete cavity in the surface layer of soil. The disposal site is covered with concrete slabs. An isolating screen consisting of compacted clay covered with a layer of earth is constructed above the top cover 9 A sorption layer consisting of a mixture of sand and clay cart be present beneath the disposal site. A water-bearing zone is located underneath the disposal site and separates it from a water-bearing stratum. According to the standard requirements, the thickness of this zone must be greater than 4 m.The containers in the disposal site are piled one on top of another. In some designs the space between the containers is filled with a filling material -sand or bentonite. The filling material can also be packed between the rows of containers.As a possible scenario of escape of radionuclides from the disposal site we consider a situation when the protective screen, the top cover, and the base at the bottom have fractured with time so that precipitation from the surface freely passes through them. It is assumed that the container walls have not changed and the radionuclides can escape through them to the outside only by diffusion in the moisture located between the interstitials. It can be assumed that the precipitation at the location of the disposal site is distributed uniformly throughout the year.In the model at hand, it is assumed that the concentration of radionuclides reaching the inner wall of the container and the change in this concentration in time are known.The aim of the calculations is to determine the concentration of the migrating radionuclides on the surface of t...
An important dement of the engineered-barriers system used for burying radioactive wastes is the container in which the wastes are placed in solid or solidified form. The construction of the containers and the materials from which they are made can differ substantially. When choosing and developing a new construction, container types must be compared with one another according to their capability to retain radionuclides. For underground storage or burial of wastes radionuclides escape primarily by diffusion into the interstitial moisture through the container wall. Many factors influence the escape of radionuclides. The main ones are the character and rate of leaching from the solid matrix or solid wastes, the diffusion and sorption properties of the container wails and the medium surrounding the container, and the construction of the container, specifically, the shape of the container.It is most convenient to compare different container types or container constructions, taking account of the large number of variable factors, on the basis of a mathematical model describing the escape of radionuclides. As a measure or criterion characterizing the retaining properties of a container, it is convenient to choose the quantity of radionuclides escaping from the container surface per unit time into the surrounding medium (the flux from the entire surfaceor the escape rate). This quantity, as indicated in [1], can serve as a direct characteristic of an engineered barrier, such as a container. The amount of radionuclides entering the external medium and the amount remaining inside the container can be determined by summing the amount of radionuclides escaping from the container in a definite time interval.In the present paper a numerical-analytical method of calculating approximately the rate of escape of radionuclides from containers with different shapes -sphere, cylinder, parallelepiped -is proposed and investigated. It is assumed that a container with solid or solidified waste is placed in a medium whose properties are different from those of the container walls. The proposed computational method makes it possible to find the rate of escape of radionuclides when the diffusion properties of the container walls and of the surrounding medium differ substantially. The difference in the effective diffusion coefficients can reach a factor of 105. Ordinarily, the finite-difference methods employed in such cases are unstable near the location where different media abut one another. This instability results in the appearance of oscillating solutions and makes it impossible to calculate with acceptable accuracy the rate of escape of radionuclides.To confirm that the accuracy of the calculation of the rate of escape of radionuclides is adequate, the proposed method is compared with the grid method for cubic and cylindrical containers.The proposed method is used to investigate the influence of the container shape on the escape rate of radionuclides. The retaining properties of concrete containers with the most common shapes and s...
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