Medium-level Waste Research Articles

A contribution towards a standard leach test for immobilized low and medium level waste

The evaluation of the safety, acceptability and environmental impact of an immobilized waste requires the identification and characterization of its relevant properties. One of the more important properties is its chemical stability in aqueous environments. This has led to the development of a wide variety of so-called “leach tests”. There is now internationally a strong body of opinion that there should be a standardized approach to this particular test. This paper discusses the categories of application of test results as they apply to the management of immobilized low and medium level waste products. It then describes a procedure for carrying put a leach test that has been adopted as an interim national standard within the United Kingdom (UK). It concludes that there is now a sufficient concensus of expertise to allow the establishment of an international standard.

Bituminization of radioactive wastes: Safety studies

The bitumen embedding of low- and medium-level radioactive wastes has reached the industrial stage. Bituminization plants are operating satisfactorily in France and throughout the world. However, bituminization presents certain risks due to the nature of bitumens themselves. These risks must be taken into account. This paper specifies those risks and presents the results of leachability tests performed in accordance with the International Atomic Energy Agency proposed standard on coated concentrates, on chemical coprecipitation sludges with and without addition of heavy solvents such as TBP and TLA, and on ion exchange resins. The results show that the leaching rates in demineralized water are on the order of 10/sup -6/ to 10/sup -8/ for cesium 137, 10/sup -5/ to 10/sup -7/ for strontium-90, 10/sup -5/ to 10/sup -6/ for ruthemium-106, 10/sup -5/ to 10/sup -6/ for cobalt-60, 10/sup -5/ to 10/sup -6/ for total gamma activity, and 10/sup -5/ to 10/sup -8/ for the plutonium-238, the plutonium-239, and the americium 241 alpha emitters. Even if this conditioning is not perfect, the storage of low- and medium-level radioactive wastes is made possible with adequate safety and at a reasonable cost.

Nuclear Waste Management in France

On October 6, and 7 1982, after a two-day Parliamentary Debate, the continuation of the French nuclear power program was approved. Installed nuclear capacities, should reach 56000 elMN by 1990, i.e. 28% of energy supplies.The nuclear waste volumes entailed by such a nuclear electricity program, with the accompanying fuel cycle operations and the other nuclear waste producing activities, are the following: Low and Medium Level waste volumes should reach 800,000 cubic meters by the year 2000 - slightly less than was planned for two years ago, because of reduced power consumption, and therefore capacities forecasts. Alpha-bearing and vitrified Wastes should reach respectively 45,000 cubic meters and 3,000 cubic meters by the year 2000.These waste quantities now make it urgent to set up a wholly comprehensive industrial scale program for their treatment, conditioning and final disposal, together with the relevant research and development activities.

Chemical Processing Facilities for Partitioning Actinides from Nuclear Fuel Cycle Waste Mixtures

This paper discusses the U.S. Department of Energy Actinide Partitioning Program, which led to the conceptual design of highly integrated chemical processing facilities to achieve the required levels of actinide element partitioning from all fuel cycle wastes. These facilities utilize both conventional tributyl phosphate extraction cycles and several new extractants such as a carbamylmethylene phosphonate (bidentate) and 2-ethylhexyl alcohol. The consequences for a light water reactor fuel cycle that result from operating these partitioning facilities include increased volumes of low- and medium-level wastes and significantly increased capital investment requirements and operating costs. 18 refs.

The effect of hot pressing additives on the leachability of hot pressed sodium hydrous titanium oxide

Sodium hydrous titanium oxide is an ion exchange resin which can be used for immobilizing medium level waste (MLW) liquors. When hot pressed, it undergoes conversion to a ceramic. Three low melting point materials (borax, bismuth trioxide, and a mixture of PbO/CuO) were added to the (Na)HTiO and the effect that each of these had on aiding densification was assessed. Hot pressing temperature, applied pressure, and percentage addition of hot pressing aid were varied. Percentage open porosity, flexural strength, and leachability were measured. Leachability was determined by weight loss after three day exposure in water at 100 °C in a Soxhlet apparatus. There was a linear relationship between the percentage open porosity and the logarithm of the leach rate for a constant percentage addition of each additive. The development of microstructure is explained by microscopic examination of the hot pressed samples before and after leaching. The leach rate of samples with almost zero open porosity hot pressed without additives or with 10 wt.% PbO/CuO are about 1 × 10 −9 kg m −2 s −1, which is better than typical Soxhlet leach rates at 100 °C for high level waste (HLW) glass formulations.

Computations of Temperature Profiles in the Medium/Low-level Waste Cement and in the Rock Salt Surrounding the Disposal Cavern

Prior to disposal of cemented medium level waste and low level waste in large caverns 1000 m below ground it is important to know what temperatures will be caused by heat generation in the waste product. During the disposal period the heat produced by hydration of the cement used as the binder as well as by decaying radionuclides in the waste may give rise to inadmissibly high temperatures in the waste product resulting in a deterioration of the produc:t qualities. To avoid such undesirable processes, a maximum permissible product temperature of 90°C was specified for the investigations, i.e. a temperature that is significantly below the boiling point of the water contained in the cemented product.

Stability and Construction of Caverns for The Disposal Of Cemented Low-And Medium-Level Wastes

ABSTRACTThe geomechanical and mining-technological aspects of the construction of salt caverns as disposal chambers have been investigated during project phase 2, completed by mid 1981. With a view towards the stability analysis of such a cavern, FEM-estimates have been carried out and evaluated. From these it can be derived that- a rotational ellipsoid would be the most suitable shape- its dimensions should be 82 m (vertical axis) and 42 m (horizontal axis)- the distance (safety pillar) between the neighbouring caverns should be 170 m (vertical) and 180 m (horizontal).For practical engineering purposes the rotational ellipsoid can be modified into a cylinder with conic bottom and top. The numerical model simulated the short term as well as the long term characteristics of the surrounding salt rocks. The short term characteristics were assessed by an elastic approach, the long term characteristics by a rheological model. The input parameters have been determined by means of laboratory tests on ASSE rock salt.In a second step the characteristics of partially and completely filled caverns were simulated. It was shown clearly that deformation of the salt rock comes to a halt when counteracted by the filling.Based upon the results of the stability analysis, investigations were made to find out a suitable mining technique for the construction of the cavern. Solution mining and conventional development by means of drilling and blasting have been studied alternatively. Since both methods have their advantages and disadvantages a decision in favour of the one or the other cannot be made until the actual site has been defined.

Control of Pollution from the Generation of Nuclear Power

Since the early days of the Manhattan District, there has been concern over potential pollution from the manufacture of atomic weapons and the generation of nuclear power. With time it has become obvious that the nuclear power plants themselves are not important sources of radioactive wastes, although they can be responsible for thermal pollution, as are conventional fossil fuel burning plants. The serious problems are associated with the relatively few chemical plants in which the partially spent nuclear fuels must be reprocessed to save their "unburned" fuel. As the result of an unprecedented program of research and development carried out by the Atomic Energy Commission, methods have been devised and tested for turning the high-level liquid radioactive wastes into inert solids which will be buried in salt mines, and for disposing of the medium-level wastes by incorporation in asphalt or by hydraulic fracturing. The very low-level wastes, decontaminated to drinking water standards, can be safely discharged to the environment, or preferably, disposed of into deep wells.

欧米における放射性廃水の処理

During the world-round trip from June to September in 1959, the author observed the recent tendency of the treatments of radioactive liquid wastes in Europe and U.S.A. The conclusions are as follows:(1) Low level wastes (<10-4μc/ml) is mainly treated by the chemical coagulation, quite economically.(2) The evaporation method is applied to medium and high level wastes and the methods of coagulation and ion-exchange are used as auxiliary treatments.(3) It is urgent to settle the best procedure to treat high level wastes.(4) We must be careful to adapt these foreign technics directly for our treatment, because we have different and complicated conditions, i.e. narrow territory, dense population, many calamities and special sensibilities of people hating radioactivity. We must acquire more budget for these sorts of plants and equipments in order to assure our coming progress in atomic age.