Medium-level Waste Research Articles

Immobilization of sulfate-containing wastes in borosilicate glass using fluoride additives

The choice of borosilicate glasses as a matrix for immobilizing radioactive wastes is largely justified by the properties of the final vitrified product to be stored for a long time: High chemical, thermal, and radiation resistance. The structure of borosilicate glasses makes it possible to use them as a matrix for high and medium level wastes of diverse chemical composition. However, individual components of the wastes - sulfate, chloride, and molybdate-containing components - have a limited solubility in borosilicate glasses. This results in phase separation (stable liquation) in the vitrification process and decreases the quality of the vitrified product.

Water-resistance of irradiated glass-composite materials for disposal of medium-level wastes

For safe long-term storage radioactive wastes are embedded in a chemically inert (with respect to the environment) matrix (cement, glass, ceramic, and so on). The most promising matrices are borosilicate glasses. They have high radiation and chemical resistance, they are mechanically strong, and they hold well many elements of wastes. During the founding of the glass, however, sulfates form a separate water-soluble phase. One way to neutralize sulfate-containing wastes is to disperse the sulfate phase in the glass melt. The resulting product consists of a two-phase material, in whose glass matrix inclusions of the sulfate phase are distributed.

Handling and Transport of Low Level Waste in Indonesia

At present, radioactive wastes in Indonesia are generated mainly from nuclear research with a small amount from industrial applications. The wastes, mostly containing short-life radionuclides, consist of a large quantity of low, and a small quantity of medium level wastes. Some research centres have their own waste treatment facility, but do not have an adequate long-term storage for their packaged wastes. Other research centres, like those at Serpong Nuclear Research Centre (SNRC), centralise their waste treatment and storage at the Radioactive Waste Management Centre (RWMC). The waste transport from waste generators to the RWMC are carried out using truck trailers. The practices of low level waste handling and transport are described. Principles covering waste handling before, during, and after, waste treatment are implemented by considering the health and safety of the transport workers and environmental safety. Waste collection has been arranged in such a way as to facilitate further waste processing. The description covers the low level wastes in the form of inorganic, organic liquid waste, spent resin, compactable and burnable solid wastes, and embedded wastes.

Deeds and misdeeds of cement composites in waste management

A waste management option considers the cement matrix as a reliably engineered barrier against the release of radionuclides (medium and low level wastes) and particular toxic and hazardous industrial wastes. The cement matrix must have a high resistance to accidental events not only as far as the material itself is concerned, but also for the materials conditioned in cement. The integrity of the cement composites will be affected in conditions in which water can accidentally enter into contact with the cement drums which are put in repositories under the earth' surface or in shallow land. But the physical characteristics of the matrix will play a role in the penetration of water in the pores of the material. Laboratory of ions confined in a cement matrix in order to assess the leaching mechanism ans to increase the retention of soluble species (cesium). In this situation, the amount of dissolved gases in water must not be neglected with a view to the effects of pressure or the reactions with the soil which affect the decomposition of the matrix and the leachability of blended compounds. The release of wastes from the cement matrix, characterized by the diffusion coefficient, is one of the most important issues which determine whether or not a prepration procedure assures the good management option scheme. A diffusion model should take into account that, eventually, the dissolution process of the matrix is slow in comparison with the diffusion of the elements considered. Leach tests have put into evidence that the kinetics of the dissolution of cement constituents seems to be affected by the leach test method in static or dynamic water conditions (modified Soxhlet procedure or water renewal). Both physical properties and chemical resistance, which are linked together, must ensure that the radioactive and toxic/hazardous industrial waste are not transported to man. The ability of a cement matrix to retain elements under leaching conditions depends on its chemical state. Several elements (Cs, Cr, Cu, Fe, Mn, Mo, Ru, Zr, Ni, actinides), contained in a Medium Level Wastes(MLW) stream from a reprocessing plant, exhibit different valence states which determine the stabilization in the matrix or its release. Nevertheless, the chemical properties of the cement matrix (high pH) contribute to the retention of some elements, but cesium. A matrix preparation procedure must also consider the final physical properties which influence the stability of blended compounds. The objective of physical tests is to provide a parameter evaluation on a comparative scale in order to characterize the product obtained in different stabilization conditions.

Vitrification of sulfate- and chloride-containing radioactive waste in an electric furnace

In the present study are reported which were obtained in the experimental treatment of RAW (medium-level radioactive waste) from the Moscow Purification Station. The batch was added to the furnace in the form of a paste of 20-25% moisture content. The temperature of the glass was 1150°C. The glass-forming additives were datolite concentrate, Lyuberetsk quartz sand, and kaolinite

Investigations on the Treatment of Low- and Medium-Level Radioactive Liquid Wastes

Sorption properties of six different coagulants with respect to the retention of cobalt, caesium, europium and uranium have been studied. The precipitating agents are ferrous sulphate, ferric chloride, aluminium sulphate, tri-sodium phosphate, ferric sulphate and zinc sulphate. The parameters investigated are pH, concentration of the precipitating agents, volume of coagulants vs time and the nature of radionuclides. The data showed that the order of hydroxide volume formed is the following: zinc sulphate > ferric chloride > ferric sulphate > ferrous sulphate, aluminium sulphate, trisodium phosphate. Hexacyanocobaltferrate was used as an adsorbant for caesium from medium-level liquid waste.

Problems of improving engineering procedures and processing flow diagrams for low- and medium-level liquid wastes

The process flow diagrams presently used for lowand medium-level liquid waste disposal produce large volumes of secondary waste that must be solidified and later buried. The cost of the expensive processes of solidification and burial must be reduced by perfecting precipitation, ion-exchange, evaporation, and other methods of treating liquid radioactive waste. The technical and economic indices of the operation of treatment facilities could be substantially improved if certain engineering procedures and process flow diagrams were employed: elimination of coagulation, breakdown of organic matter using ozone, preliminary desalination of waste by electrodialysis, and extraction of valuable components during waste concentration, and also if the technologies of continuous ion exchange and ultrafiltration were implemented.

Treatment of low- and medium-level residues and wastes from reprocessing

Treatment of low- and medium-level residues and wastes from reprocessing

Measurements, calculations and stability prognoses at the SFR undersea repository for low- and medium-level nuclear waste

Measurements, calculations and stability prognoses at the SFR undersea repository for low- and medium-level nuclear waste

Storage of Nuclear Waste in Sweden

Abstract Sweden has been active in developing a comprehensive program for the disposal of nuclear waste. The author describes the programme for radioactive waste management, which comprises: a sea-transportation system; a central repository for final disposal of low- and medium-level radioactive waste (SFR); a central facility for interim storage of spent fuel (CLAB); and a repository for final disposal of spent fuel and long-life waste. The research and development work in progress in an abandoned iron-ore mine (Stripa Project) is described.

Leachability of technetium from concrete

Technetium-bearing medium-level waste is in the Federal Republic of Germany supposed to be conditioned in concrete and stored in a salt mine repository. An accidental ingress of water into the repository leads to the formation of a brine solution which may corrode the barrier system components and leach radionuclides from the concrete. The long half-life of 99Tc may then permit for a slow migration of this radionuclide into the biosphere. The leachability of technetium was investigated with samples prepared from portland, aluminous and blast furnace cement with or without the additives bentonite and Microsilica. 99Tc was used as NH 4TcO 4 resp. TcO 2. A quinary brine solution and distilled water were applied as leachants. 137Cs and 36Cl were also used for comparative measurements. With aluminous cement/bentonite only unfavourable results were obtained: 20% of the technetium had been leached into brine after 60 days. Later, but in water already after 20 days with 40% of the technetium leached, the slope of the leaching curve turned negative, which was possibly due to the formation of new mineral phases with Tc as a constituent. Bentonite-free aluminous cement behaved somewhat better than the mixture. Portland cement with or without bentonite showed an overall medium leachability. The best results were obtained with blast furnace cement. Leachability generally remains below 0.1%; bentonite is of no significance. Here Tc is possibly rapidly fixed as a species of low solubility, whose solubility product is being reached within short time. Leachability of TcO 2 is higher than expected, it amounts to about 10% as compared with pertechnetate, which is probably due to a transformation of TcO 2 into a more soluble species. Cs and Cl are leached by a factor of >100 faster than Tc from blast furnace cement, supporting the assumption of a mineralogical fixation of Tc and a diffusion-controlled migration of Cs- and Cl-ions.

Measurements, calculations and stability prognoses at the SFR undersea repository for low- and medium-level nuclear waste

Abstract A central repository for low- and medium-level radioactive waste from the Swedish nuclear power plants has been constructed at Forsmark, Sweden. The most active waste will be disposed of in a rock cavern, designed as a silo 69 m high and 30 m in diameter. Because the rock stability requirements are stringent, pre-investigations were very extensive. The initial rock stresses were measured at several points. Based on these investigations, an analysis of the behavior of the silo was performed using the Finite-Element Method. During construction, the behavior of the rock mass around the silo was monitored with extensometers. Very small deformations of a few millimeters were measured; these were comparable with the calculated results.

Water Corrosion and Release Mechanism of Cement Matrix Incorporating Simulated Medium-Level Waste

Previous works on cement leached in water and containing radioactive wastes like cesium and strontium agree to attribute the release in the liquid phase to a diffusion mechanism. The kinetic release can be represented by an empirical relationship in which the dependence of the leached fraction Fr = C/C/sub o/ versus time t assumes the form Fr = Bt/sup n/ and the factor n = 0.5 is typical of a diffusion process. On the other hand, the results of the authors' studies on cement leached in static water demonstrate that the release of calcium, considered to be representative of matrix degradation, follows a time dependence of t/sup 0.25/. A model is suggested for which the release of calcium depends on superposition of two processes: a diffusion through a reaction layer of calcium silicate hydrate around the cement particles during the hydration step and a diffusion of elements from the bulk of cement toward the external surface of the specimen. Based on this schematic diffusion mechanism, some suggestions are advanced to improve the physical characteristics and to increase the retention of the radioactive waste encapsulated in the cement matrix.

Water leachability of cement forms for medium-level waste immobilization

An evaluation of the time dependence and mechanism of the corrosion of cement containing chemical compounds simulating a medium-level waste (MLW) stream has been carried out by measuring in the leachate the amount of Ca, Al, and Si as structural elements and SO 4 −2 and NO 3 −1 as representative of MLW compounds. The samples, measuring 2-cm diam, have been prepared at a water-to-cement ratio of 0.35 and a solid-to-cement ratio of 0.20. The tests in static distilled water at 50, 70, and 90°C show good cement resistance to water action. A slowing-down of a diffusion mechanism through a layer of silica gel or back-reaction to form calcium silicate seems to influence the overall concentration of calcium in the liquid phase. A rate of cement dissolution of about 5.07 10 −16 m sec −1 (1.6 × 10 −6 cm yr −1) has been calculated. In a more realistic condition of flowing water around the cement drums in an accident event occuring in a repository, the cement resistance is found to depend on the water flow rate conditions. Dissolution of the matrix at a rate of about 9.5 × 10 −11 and 5.7 × 10 −10 m sec −1 (0.3 and 1.8 cm yr −1) at 50 and 70°C, respectively, has been calculated under extreme dynamic conditions.

Powder technological vitrification of medium-level waste by in-can hot pressing

Powder technology is used to immobilize medium level nuclear waste (MLW) in sintered glass. By in-can hot pressing (temperature 923 K; pressure 0.4 MPa; heating rate ≈ 100 K/h; cooling rate 5 K/h) glass products containing simulated MLW (20 wt%) have been produced in stackable steel cans (⩽ 300 mm diameter). High densities (≈ 93% TD), bulk integrities and homogeneities for the waste element distributions are realized. The advantages of powder technology as, for example: (i) no segregation due to solid state vitrification, (ii) low evaporation losses and no compatibility problems due to low densification temperatures (sinter temperature ≈ 0.6 softening temperature of glass), (iii) production in easily arrangable and interchangeable stacking units at modest pressures adequate especially for the use in hot cells under remote handling conditions, all together hold for the process which therefore is appropriated for high level waste (HLW), medium level waste and HLW/MLW common vitrification as well.

Powder Technological Vitrification of High Level Nuclear Waste by In-Can Hot Pressing Including a Sol-Gel Route

AbstractPowder technology is used to immobilize high level nuclear waste (HLW) in sintered borosilicate glass. By uniaxial in-can hot pressing(temperature 950 K; pressure 1 MPa; heating rate ∼100 K/h; cooling rate 5 K/h), glass products containing simulated HLW(15 wt.%) have been produced in stackable steel cans (≤ 200 mm diameter). High densities, bulk integrities and homogeneities for the waste element distribution are realized. The advantages of powder technology are for example: (i) no segregation due to solid state vitrification, (ii) low evaporation losses and no compatibility problems due to low densification temperatures (sinter temperature ∼0.6 softening temperature of glass), (iii) production in easily arrangeable and interchangeable stacking units at modest pressures adequate especially for the use in hot cells under remote handling conditions, (iv) choosing units with alternatively HLWo r MLWin prescribed sequences to control the heat production of the package. As demonstrated now, the process is appropriate for high level waste (HLW), medium level waste (MLW) and mixed (HLW/ MLW vitrification, and is insensitive to waste modifications.As the next step the effect of glass modifications was studied. In order to improve the glass leaching resistance and maintain low viscosity, silicon dioxide was doped with titanium dioxide. The mixture (referring to eutectic composition) was produced by a sol gel route and combined finally with 15 w/o HLW-oxides. Hot pressing of the obtained gels at 1273 K and 20 MPa provided highly homogeneous products with high densities.

Nuclear Waste Management Policy in France

The object of the nuclear waste management policy in France has always been to protect the worker and the public from unacceptable risks. The means and the structures developed to reach this objective, however, have evolved with time. One fact has come out ever more clearly over the years: Nuclear waste problems cannot be considered in a piecemeal fashion. The French nuclear waste management structure and policy aim at just this global approach. Responsibilities have been distributed between the main partners: the waste producers and conditioners, the research teams, the safety authorities, and the long-term waste manager, National Radioactive Waste Management Agency. The main technical options adopted for waste forms are embedding in hydraulic binders, bitumen, or thermosetting resins for low-level waste (LLW) and medium-level waste (MLW), and vitrification for high-level, liquid wastes. One shallow land disposal site for LLW and MLW has been in operation since 1969, the Centre of La Manche. Alpha-bearing and high-level waste will be disposed of by deep geological storage, possibly in granite formations. Further RandD aims mainly at improving present-day practices, developing more durable, long-term, alpha-bearing waste for all solid waste forms and going into all aspects of deep geological disposal characterization.

Possibilities and conditions for vitrification of medium-level wastes

Of the materials for which manufacturing processes are currently being developed, vitreous preparations have the best qualities (maximum reduction in volume, lowest rate of leaching of radionuclides, absence of requirements for fire safety in transportation and burial, etc.). In this connection, it is interesting to evaluate the possibility and conditions for vitrification of medium-level wastes. The investigatory and experimental work was performed for a setup using an electrically heated furnace, such as is being developed in the USSR for high-level wastes as well [I].

In-Can Hot Pressing of Borosilicate Glass for the Immobilization of Highand Medium-Level Wastes

ABSTRACTThe paper presents a method for the immobilization of different waste streams based in the uniaxial in-can hot pressing of a glass frit together with the waste oxides.The process reported herein offers engineering simplicity in combination with many technological advantages such as significant decrease in temperature, use of small pressure, improved homogeneity of the products, production in easily arrangeable and interchangeable units. Furthermore, the proposed method has also proved to be suitable for treating intermediatelevel waste streams containing Zry-4 cladding fragments.

Authors

Authors