Lawrence Johnson

Mechanisms of Leaching and Dissolution of UO2 Fuel

An integrated experimental approach to mechanistic studies of the leaching and dissolution of irradiated UO2fuel is described. The program includes an investigation of the solubility of the UO2 matrix under thermodynamically well-defined conditions, detailed measurements of the leaching and dissolution of irradiated fuel under simulated disposal conditions, and electrochemical measurements with a novel UO2 electrode to elucidate dissolution mechanisms. Initial experiments show that the solubility of UO2 under alkaline reducing conditions is relatively insensitive to temperature changes, that the leach rates of irradiated fuel are also not strongly temperature dependent, and that surface films on the UO2 fuel may play an important role in the dissolution process. Several aspects of the UO2 matrix dissolution process are now understood, and the approach taken has indicated where future work is needed.

Oxidation of Candu UO 2 Fuel by the Alpha-Radiolysis Products of Water

The oxidation of CANDU fuel (UO/sub 2/) by the alpha-radiolysis products of water has been investigated using electrochemical and X-ray photoelectron spectroscopic experiments. Experiments with O/sub 2/ and H/sub 2/O, two of the expected products of radiolysis of water, indicate that the rate of oxidation of UO/sub 2/ by H/sub 2/O/sub 2/ is about 200 times faster than by dissolved oxygen. Oxidation by both H/sub 2/O/sub 2/ and O/sub 2/ shows pH dependence. Possible reaction paths for the oxidation of UO/sub 2/ by radiolysis products are discussed.

Quantitative Assessment of the Instant Release Fraction (IRF) for Fission Gases and Volatile Elements as a Function of Burnup and Time under Geological Disposal Conditions

The Instant Release Fraction at container failure time, IRF(t), is here considered as being the sum of (i) the initial labile fraction, corresponding to the sum of gap and grain boundary inventories of certain radionuclides on exit from the reactor, with a further possible contribution from segregation in the rim region and (ii) the time-dependent fraction of radionuclides accumulating at grain boundaries due to a self-irradiation enhanced diffusion through the grains. The initial labile fraction of radionuclides such as 14 C, 36 Cl, 79 Se, 129 I, and 135 Cs has been estimated based on leaching experiments, post-irradiation fission gas release measurements and studies of solid-state chemistry of spent fuel, along with estimates of fission product segregation in the rim zone. The contribution of the a self-irradiation enhanced diffusion has also been estimated based on a diffusion coefficient decreasing with time proportionally with the volume α-activity of the spent fuel. Its contribution to the IRF is limited for UO 2 fuels. The proposed bounding values of the IRF for fuel with a burnup of 55 GWd/t IHM for 14 C, 36 Cl, 79 Se, 129 I, and 135 Cs are 11 % at t=0 and close to 15 % at a container failure time of 10,000 y.

Mechanisms governing the release of radionuclides from spent nuclear fuel in geological repository: major outcomes of the European Project SFS

European Commission supported a wide research project entitled “Spent Fuel Stability under repository conditions” (SFS) within the 5 th FWP, the aim of which was to develop a common understanding of the radionuclides release from spent nuclear fuel in geological disposal and build a RN release model in order to assess the fuel performance. This project achieved by the end of 2004 focuses both on the Instant Release Fraction (IRF) model and the Matrix Alteration Model (MAM). A new IRF model was developed based on the anticipated performances of the various fuel microstructures (gap, rim, grains boundaries) and the potential diffusion of RN before the canister breaching. However, this model lets the choice to the end-user about the degree of conservativeness to consider. In addition, fuel alteration has been demonstrated to be linked to the production of radiolytic oxidants by water radiolysis at the fuel interface, the oxidation of the fuel interface by radiolytic oxidants and the subsequent release of uranium under the influence of aqueous ligands. A large set of experimental data was therefore acquired in order (i) to upgrade the current radiolytic kinetic scheme, (ii) to experimentally correlate the fuel alteration rate and the fuel specific alpha activity by performing experiments on alpha doped samples, (iii) to experimentally test the potential inhibitor effect of hydrogen on fuel dissolution. Based on these results, a new MAM was developed, which was also calibrated using the experiments on inactive UO 2 samples. This model was finally applied to representative granitic, salt and clayey environment to predict spent fuel long-term fuel performance.

Assessment of Redox Conditions in the Near Field of Nuclear Waste Repositories: Application to the Swiss high-level and intermediate level waste disposal concept

Redox conditions were assessed for a spent fuel and high-level waste (SF/HLW) and an intermediate-level waste (ILW) repository. For both cases our analysis indicates permanently reducing conditions after a relatively short oxic period. The canister-bentonite near field in the HLW case displays a high redox buffering capacity because of expected high activity of dissolved and surface-bound Fe(II). This is contrary to the cementitious near field in the ILW case where concentrations of dissolved reduced species are low and redox reactions occur primarily via solid phase transformation processes. For the bentonite-canister near field, redox potentials of about -100 to -300 mV (SHE) are estimated, which is supported by recent kinetic data on U, Tc and Se interaction with reduced iron systems. For the cementitious near field, redox potentials of about -200 to -800 mV are estimated, which reflects the large uncertainties related to this alkaline environment.

Summary of the euratom collaborative project FIRST-nuclides and conclusions for the safety case

Abstract A consortium of 10 partners from seven European countries and the European Commission collaborated in investigating the short-term release of radionuclides from disposed spent nuclear fuel upon canister failure. The Collaborative Project FIRST-Nuclides was implemented in the scope of the 7th Euratom Framework Programme in the period from 2012 to 2014. The objectives and organization of the project are presented, as well as the experiments with highly radioactive materials under investigation. The outcome of the project summarizes the measured instant release fraction (IRF) of safety-relevant isotopes from high burnup spent UO2 nuclear fuels (SNFs). Specifically discussed are the dependencies of the IRF on the sample properties, the gap and grain boundary releases, and the behavior and IRFs of elements such as cesium, iodine, and selenium. The IRFs of nonstandard SNFs were also investigated. The summary is complemented by the presentation of the modeling approaches within the project.

The Integrated Project NF-PRO: Recent Advances in European Research on the Near-Field System

The Integrated Project NF-PRO (Sixth Framework Programme by the European Commission) investigates key-processes in the near-field of a geological repository for the disposal of high-level vitrified waste and spent fuel. The paper discusses the project scope and content and gives a summary overview of advances made by NF-PRO.Copyright

Project Opalinus Clay (Entsorgungsnachweis) - Illustrating the Safety and Robustness of the Proposed Disposal System for SF, HLW and Long-lived ILW

Nagra (the Swiss National Cooperative for the Disposal of Radioactive Waste) has completed a study to determine the suitability of Opalinus Clay as a host rock for a repository for spent fuel (SF), high-level waste from reprocessing (HLW) and long-livedintermediate-level waste (ILW). The proposed siting area is in the Zurcher Weinland region of Northern Switzerland. A repository at this site is shown to provide sufficient safety for a spectrum of assessment cases that is broad enough to cover all reasonable possibilities for the evolution of the system. Furthermore, the system is robust; i.e. remaining uncertainties do not put safety in question.

Lifetimes of Titanium and Copper Containers for the Disposal of Used Nuclear Fuel

Titanium and copper have been proposed as suitable container materials for disposal of nuclear fuel waste in plutonic rock of the Canadian Shield. Studies of the corrosion of these materials have led to the development of container failure models to predict long-term performance. Crevice corrosion and hydrogen-induced cracking of titanium have been identified as potential failure mechanisms, and these two processes have been studied in detail. Using data from these studies as well as a number of conservative assumptions, titanium container lifetimes of 1200 to 7000 a have been estimated. For copper, general corrosion has been studied in detail in bulk solution and in compacted clay-based buffer material. Results indicate that the copper corrosion rate is likely to be controlled by the rate of transport of copper species away from the container surface. An assessment of copper pitting data suggests that pitting is an extremely improbable failure mechanism. The copper container failure model predicts minimum container lifetimes of 30 000 a. The results demonstrate that long lifetime containment can be provided, should performance assessment studies indicate the need for such an option.