Rates and mechanisms of radioactive release and retention inside a waste disposal canister - in Can Processes

Sweden and Finland are planning to dispose of spent nuclear fuel in a deep underground repository constructed in granitic rock. Each country is investigating candidate sites and developing the scientific and technical basis for assessing the safety of an eventual repository. An essential part of the safety assessment involves understanding the behaviour of the spent fuel after it is placed in the geologic environment. The fuel will be sealed inside a copper canister that contains a cast iron insert. The copper functions as a corrosion resistant barrier, while the cast iron insert fills much of the internal void space, adding strength to the canister and reducing the space available for water to accumulate inside the canister after the corrosion barrier is breached. The canisters will be surrounded by compressed bentonite, which will limit the access of water and dissolved species to the canister. Oxygen that is initially present when the disposal environment is sealed will be rapidly consumed by pyrite in the bentonite, bacterial species in the rock, and reduced inorganic materials in the rock. The copper canister will prevent access of water to the iron until it is corroded through, a process that is expected to take millions of years. After water contacts the iron, anaerobic corrosion of the insert will generate hydrogen gas and introduce Fe(II) ions into the water. The long-term environment for the fuel, therefore, is a highly reducing environment. The only possible source of oxidising agents is radiolysis of the water by radiation from the fuel. In the long-term, the radioactivity in the fuel is due to isotopes that decay by alpha decay; most of the activity from beta and gamma radiation will have decayed away. Spent fuel that is available for testing contains high levels of beta and gamma activity. Even when testing is done in the presence of hydrogen or actively corroding iron, the radiolysis due to beta and gamma radiation can introduce oxidising agents into the system that will not be present under long term disposal conditions. A simulation of long-term conditions can be done using uranium dioxide that contains a short-lived isotope of uranium, but this will not include the effects of fission product and higher actinide elements on the behaviour of the spent fuel. We designed a project that had as its objective to improve the scientific understanding of the processes that control release of radioactive species from spent fuel inside a disposal canister and the chemical changes in those species that might limit release of radioactivity from the canister. If the mechanisms that control dissolution of the fuel matrix, including self-irradiation effects, can be clarified, a more realistic assessment of the long-term behaviour of spent fuel under disposal conditions can be made. By removing uncertainties concerning waste form performance, a better assessment of the individual and collective role of the engineered barriers can be made. To achieve the overall objective of the project, the following scientific and technical objectives were set. 1. Measure the actual rate of matrix dissolution of uranium dioxide under oxidising and reducing conditions. 2. Measure the effect of alpha radiolysis on the dissolution rate of uranium dioxide under oxidising and reducing conditions. 3. Measure the dissolution rate of the matrix material of spent fuel and thereby determine the additional effects of beta and gamma radiation on uranium dioxide dissolution rate under oxidising and reducing conditions. 4. Measure the ability of actively corroding iron to reduce oxidised U(VI) to U(IV) when U is present as the complex ion uranyl carbonate. 5. Measure the rate of reduction of Np(V) species in the presence of actively corroding iron. 6. Calculate the expected equilibrium and steady state concentrations of U under the conditions of the experiments used for meeting objectives 1 through 3 and compare the calculated results with those measured in the experiments. 7. Calculate from first principles the expected reaction path, and the relative reaction rate for the reduction of U(VI) to U(IV) under the conditions of experiments used to meet objective 4 and with other ligands in solution. Similar calculations will also be done for reduction of Np(V) to Np(IV) for solutions with various ligands. As the project progressed, item 6 was found to be unnecessary and item 7 was modified to the more specific task of calculating the reduction of U(VI) to U(V) by Fe(II) using ab initio methods. Each of the technical and scientific objectives was a work package within the project. (Item 7 was WP6)