Iodine filters expended after nuclear fuel reprocessing contain radioactive iodine (I-129), almost all of which exists as silver iodide (AgI). The synthetic rock technique is a solidification treatment technique using hot isostatic press (HIP), in which the alumina adsorbent base material is synthesized to form a dense solidified material (synthetic rock), and I-129 is physically confined in the form of AgI in the alumina matrix. Thus, it is necessary to understand the matrix dissolution behavior to evaluate the iodine release behavior.Experiments involving the dissolution of the matrix were carried out under various temperatures (35–70 °C) and pH values (10–12.5) that reflect the disposal conditions. The results of the experiments showed that the dissolution rate of Al visibly increases with temperature and pH. The dissolution rate constant was calculated from the initial data assuming the dissolution of the matrix as a primary reaction. The logarithmic rate constant showed a good linear correlation with the pH and the reciprocal of temperature. The 27Al-NMR analysis of the solutions of the dissolved matrix showed that the major chemical species present in the solutions was Al(OH)4-. This indicated that the dissolution of the matrix can be described by the following equation: Al2O3 + 2OH- + 3H2O → 2Al(OH)4-. Subsequently, the empirical equation of the rate of dissolution of the matrix as a function of the temperature and pH was derived. It will be used to evaluate the iodine release behavior from the synthetic rock.

Corrosion behavior is a key issue for the waste disposal of irradiated metals, such as hulls and endpieces, and is considered to be a leaching source of radionuclides including C-14. However, little information about Zircaloy corrosion in anticorrosive conditions has been provided.In the present study, long-term corrosion tests of Zircaloy-4 and Zircaloy-2 were performed in assumed disposal conditions (dilute NaOH solution, pH 12.5, 303 K) by using the gas flow system for 1500 days. The corrosion rate, which was determined by measuring gaseous hydrogen and the hydrogen absorbed in Zircaloy, decreased with immersion time and was lower than the value of 2×10−2 μm/y used in performance assessment (1500-day values: 5.84×10−3 and 5.66×10−3 μm/y for Zircaloy-4, 1000-day values: 8.81×10−3 μm/y for Zircaloy-2). The difference in corrosion behavior between Zircaloy 4 and Zircaloy-2 was negligible. The average values of the hydrogen absorption ratios for Zircaloy-4 and Zircaloy-2 during corrosion were 91% and 94%, respectively.The hydrogen generation kinetics of both gas evolution and absorption into metal can be shown by a parabolic curve. This result indicates that the diffusion process controls the Zircaloy corrosion in the early corrosion stage of the present study, and that the thickness of the oxide film in this stage is limited to approximately 25 nm and may therefore be in the form of dense tetragonal zirconia.

Corrosion tests of Zircaloy-4 were performed in a dilute NaOH solution (pH =12.5) at 303 K for 90 days using the gas flow system (oxygen; < 1 ppb) and a batch method (oxygen; < 0.1 ppm). The corrosion rate was determined by measuring gaseous hydrogen and the hydrogen absorbed into Zircaloy-4 assuming the following reaction:where x represents the Zircaloy-4 hydrogen absorption ratio. The initial hydrogen content in the Zircaloy-4 specimen was controlled to be below 10 ppm. The corrosion rate decreased with time (90-day values: 2.46×10-3 and 2.37×10-3 μm/y for the gas flow method and 6.72×10-2 μm/y for the batch test). The Zircaloy-4 hydrogen absorption ratio during corrosion was over 90%. The large amount of hydrogen absorbed in Zircaloy-4 will play an important role in the long-term safety for the disposal of irradiated Zircaloy materials.

The carbon-14 generated in Zircaloy (Zry) hull waste is considered an important radionuclide in the TRU waste geological disposal concept in Japan. Given that the metal Zry is highly corrosion-resistant in the anaerobic and low-temperature conditions of the repository, and that the C-14 release rate is assumed to be controlled by the corrosion rate, a variety of corrosion and leaching tests have been performed. However, since the Zry corrosion rate is extremely slow, it is not possible to predict long-term corrosion behavior through low-temperature corrosion tests conducted in a reasonable time period. A vast amount of testing has been conducted in the higher-temperature range of 523 to 633 K, and corrosion correlations have been obtained from these tests. Corrosion correlations have been used to predict the corrosion rate of Zry in a tuff repository. Long-term Zry autoclave corrosion data have been analyzed to develop new corrosion correlations. Extrapolating these correlations to a lower temperature range requires verification that the mechanisms do not change over the range of testing and extrapolation. Factors that influence corrosion rates under geological disposal conditions, such as material and environmental factors, should also be examined. Corrosion correlations, factors influencing corrosion rates, the results of corrosion and leaching tests, and a preliminary evaluation are discussed.

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