The Swedish concept for a final repository for radioactive waste consists of a set of facilities, each of these is engineered to handle different kinds of radioactive waste under different timescales. The proposed facility for the long-lived radioactive waste in Sweden is the Nuclear Fuel Repository (Kärnbränsleförvaret), which is designed to contain the spent nuclear fuel from Swedens nuclear reactors for up to 100 000 years.
The facility will be built according to a multi-barrier concept, where each barrier gives isolation from, and protection for, the environment. The essentially non-soluble fuel matrix, the corrosion-resistant copper canister, the backfill of bentonite clay of low water-permeability and the stable granitic bedrock at a level of 500m below surface are all together working as a consecutive barrier system.
In a worst case scenario, a broken copper canister could bring the fuel in contact with a limited amount of groundwater that has to diffuse through the bentonite clay. In such a worst-case scenario, processes can occur that in time will affect the movement and spreading of radioactivity to the environment.
Besides the nuclear fuel, other types of radioactive waste are usually incorporated into concrete and are then to be stored in separate underground facilities, for example the currently operational SFR facility at Forsmark.
Many of these processes are solution–solid interface chemical interactions. Some will facilitate, whereas others will inhibit radionuclide transport. The repository research at Nuclear Chemistry is centered about the chemistry of solids, mainly the fuel itself, and the barrier materials of concrete and rock. Aquatic solutions in contact with these solids are studied and, more specifically, the distribution of the radioactive elements between solid and aqueous phases. Some of the research questions that we try to answer are presented below.
UO2 is the primary solid phase, and therefore the spreading of radionuclides into the environment in the event of groundwater contact with the fuel is to a large extent dependent on the solubility of the UO¬2. The solubility of UO2 is very low, but it can be affected by oxidants produced by radiolysis of water due to the radiation from the spent fuel. However, the anoxic corrosion of iron that is present in the fuel assemblies and other components will produce hydrogen, which can reduce this effect.
How fast is the dissolution of UO2 and what conditions will influence the process?
Once dissolved, the radioactive waste may precipitate in the form of secondary solid phases which in turn may undergo further changes in the form of phase transformations and possible re-dissolution.
How can these phases be characterized and what changes can they undergo?
Reactions may occur between dissolved radioactive waste and other chemicals present in the waste and the environment. Some types of radioactive waste may include organic compounds, while several inorganic compounds are present in the groundwater. When considering radioactive metals that may have dissolved from the waste, these may form secondary compounds (metal complexation) with other dissolved compounds. The complexation may substantially affect the amount substance that goes into solution.
Which complexation reactions can happen, and do they influence the transport of radioactivity to the environment?
The process of diffusion is important for transport of dissolved radioactive waste in stagnant water zones, particularly in the barrier of bentonite clay, but also in the granitic rock. Due to the small available porosity and the strong sorption behavior of the materials, the diffusion of radionuclides in these materials is 1000 times slower than their diffusion in groundwater.
How fast can the radioactive waste diffuse through the barrier materials during different groundwater conditions?
Sorption describes how dissolved radioactive waste reacts with solid surfaces and may thereby become temporarily fixed to the surface (adsorbed) or incorporated (absorbed) in them. The most important surfaces in the barrier system are those consisting of bentonite clay, granite and concrete. These materials are structurally complicated and difficult to theoretically describe in sorption models.
How strong is the sorption of radioactivity to the barrier surfaces and how can this sorption behavior be modeled?
In the case of “low-level” radioactive waste, that is, the waste beside the spent nuclear fuel, this type of waste is usually incorporated into concrete. Concrete chemistry is characterized by alkaline pH values (pH 11-13). The long term influence of alkaline pH on the waste, the organic compounds in the waste and the surrounding rock is important to have knowledge about in the safety assessments. All the processes described above, except fuel dissolution, must be re-evaluated for highly alkaline conditions.
How will concrete and the associated alkaline conditions influence long-time radioactive waste retention?
Professor Christian Ekberg
Researcher Stefan Allard
Researcher Stellan Holgersson