The oxide films formed on spent fuel claddings are regarded as a potential source of the instantaneous release of radionuclides, such as 14C, after waste disposal. We investigated the instant release fraction using the irradiated oxide exfoliated from a Zircaloy-2 water rod, whose bundle burnup was 53.0 GWd/MTU. We performed a rapid leaching test in a dilute NaOH solution (pH of 12.5) for 10 min in an ultrasonic bath to ensure the release of radionuclides. The activity ratios of the leached amount to the total amount for 14C, 60Co, and 125Sb were extremely low at approximately 10-4 to 10-3, among which the maximum value was 2.65 × 10-3 for 125Sb. These ratios were higher than that predicted from the thermodynamic solubility of ZrO2, i.e., less than 10-6. However, given the low ratios, it is too conservative to regard the inventory of all radionuclides in the Zircaloy oxide as instantaneous release. A small part of the released 14C was found as volatile species.

Development of an iodine immobilization technique that can fix radioactive iodine in waste form for a long period and constrain its leaching into pore water is necessary in order to secure the long-term safety of geological disposal of transuranic (TRU) waste.

Lead borate glass vitrified at a low temperature is regarded as one of the promising immobilization materials of Iodine-129 which will be removed from spent AgI filters generated from reprocessing plants and may have a significant effect on long term safety of geological disposal.

Leaching experiments in bentonite-equilibrium water have been conducted to understand the lead borate glass dissolution behaviors in possible geological disposal conditions. Boron dissolved with the highest rate in all types of the solutions and was regarded as an index element to represent the glass dissolution rate. On the other hand, lead dissolved with a far slower rate. The chemical species and possible precipitating minerals of lead were examined by a geochemical calculation code for typical underground water. Altered glass surfaces were investigated by SEM, TEM and XRD. XRD analysis showed that the main constituent phase of the altered layer was hydrocerussite, Pb3(CO3)2(OH)2, that was predicted by the geochemical simulation as well.

Alumina matrix solidification is a hot isostatic pressing (HIP) technique used to immobilize radioactive iodine (129I) in the form of silver iodide. In the present study, an alumina matrix solidification sample with a porosity of 12.9% was obtained by performing HIP at 175 MPa and 1200°C for 3 hours on a simulated spent silver-sorbent saturated with stable iodine. Material Characterization Centre-1 (MCC-1) leaching tests for the simulated waste form were performed using hydrosulfide (HS-) as a reductant at concentrations ranging from 3 × 10-7 M to 3 × 10-3 M and at pH values ranging from 8.0 to 12.5. Leached iodine concentrations were below the detection limit for ICP-MS measurements at HS- concentrations of 3 × 10-7 M and 3 × 10-5 M. This result was due to the stability of AgI. At an HS- concentration of 3 × 10-3 M, iodine leaching rapidly increased within 10 days. The maximum iodine concentration in the solution was 4.33 × 10-3 M, which corresponds to 85% dissolution of the initial iodine. This value was measured after 552 days under an HS- concentration of 3 × 10-3 M at pH 11. An analysis of specimen cross-sections suggested the following reaction: 2AgI + HS- = Ag2S + 2I- + H+. The pH affected matrix aluminum dissolution but did not significantly affect the iodine leaching behavior. Furthermore, the normalized mass loss of iodine was larger than that of aluminum by a factor greater than 104, which is due to the large porosity and the dissolution of interior AgI of the solid.

Nuclear reprocessing plants in Japan produce radioactive iodine-bearing materials such as spent silver adsorbents. Japanese disposal plans classify radioactive waste containing a given quantity of iodine-129 as Transuranic Waste Group 1 for spent silver adsorbent or as Group 3 for bitumen-solidified waste, and stipulate that such waste must be disposed of by burial deep underground. Given the long half-life of iodine-129 of 15.7 million years, it is difficult to prevent release of iodine-129 from the waste into the surrounding environment in the long term. Moreover, because ionic iodine is soluble and not readily adsorbed, its migration is not significantly retarded by engineered or natural barriers. The release of iodine-129 from nuclear waste therefore must be restricted to permit reliable safety assessment; this technique is called “controlled release”. It is desirable that the release period for iodine be longer than 100,000 years. To this end, several techniques for immobilization of iodine have been developed; three leading techniques are the use of synthetic rock (alumina matrix solidification), BPI (BiPbO2I) glass, and high-performance cement. Iodine is fixed as AgI in the grain boundary of corundum or quartz through hot isostatic pressing in synthetic rock, as BPI in boron/lead-based glass, or as cement minerals such as ettringite in high-performance alumina cement. These techniques are assessed by three models: the corrosion model, the leaching model, and the solubility-equilibrium model. This paper describes the current status of these three techniques.

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.

In order to analyze the C-14 inventory and leaching rate for safety evaluation of transuranic waste disposal, it is necessary to establish an analytical method that can measure C-14 with sufficient precision [1]. Oxidative decomposition of organic compounds containing C-14 is carried out to absorb carbon dioxide (CO2) in an alkaline solution, which is mixed with a liquid scintillation cocktail, and the amount C-14 is quantified by measuring a beta ray spectrum with a liquid scintillation counter. It has been difficult to completely decompose carbon compounds in a sample, even to CO2, by using conventional oxidizing agents. In the work described here, we improved the method of oxidative decomposition used to completely decompose carbon compounds using peroxydisulfuric acid (K2S2O8). When C-14 in the form of CO2 was absorbed in a sodium hydroxide (NaOH) aqueous solution, only 80% of the actually used quantity was detected. Total organic carbon measurements showed that the entire quantity of CO2 was absorbed by NaOH. When NaOH aqueous solution was used, it was found that only the analytical value was 80%. The entire quantity of the actually used carbon could be measured by absorbing the CO2 in Carbo-Sorb®. An anion form and a neutral molecule exist in the organic compound released from activated metals. In order to identify organic compounds efficiently, fractionation into an anion and a neutral molecule and separation by high performance liquid chromatography (HPLC) are necessary. Here, we propose the combined use of an ion exchange resin and HPLC as an improved technique for identification of the chemical species.

Corrosion behavior is a key issue in the assessment of disposal performance for activated waste such as spent fuel assemblies (i.e., hulls and end-pieces) because corrosion is expected to initiate radionuclide (e.g., C-14) leaching from such waste. Because the anticipated corrosion rate is extremely low, understanding and modeling Zircaloy (Zry) corrosion behavior under geological disposal conditions is important in predicting very long-term corrosion. Corrosion models applicable in the higher temperature ranges of nuclear reactors have been proposed based on considerable testing in the 523−633 K temperature range.

In this study, corrosion tests were carried out to confirm the applicability of such existing models to the low temperature range of geological disposal, and to examine the influence of material, environmental, and other factors on corrosion rates under geological disposal conditions. A characterization analysis of the generated oxide film was also performed.

To confirm applicability, the corrosion rate of Zry-4 in pure water with a temperature change from 303 K to 433 K was obtained using a hydrogen measuring technique, giving a corrosion rate for 180 days of 8 × 10-3 μm/y at 303 K.

To investigate the influence of various factors, corrosion tests were carried out. The corrosion rates for Zry-2 and Zry-4 were almost same, and increased with a temperature increase from 303 K to 353 K. The influence of pH (12.5) compared with pure water was about 1.4 at 180 days at 303 K.

C-14 contained in Hull waste is one of the most important radionuclides in the safety assessment of transuranic (TRU) waste disposal. For more realistic safety assessment, it is important to clarify the release mechanism and chemical species of C-14 from Hull waste. In this research, leaching tests were conducted using an irradiated Zry cladding tube from a boiling-water reactor (BWR) to obtain leaching data and to investigate the relationship between Zry metal corrosion and C-14 release behavior. Both organic and inorganic C-14 compounds existed in the the liquid phase, and some C-14 moved to the gaseous phase. The release rate of C-14 obtained from the BWR cladding tube after two-year leaching tests was lower than the release rate from a pressurize water reactor (PWR) cladding tube. It is considered that the BWR cladding tube used in this test did not easily corrode since it used a comparatively new material. The release rate of C-14 was slightly lower as compared with the corrosion rate of unirradiated Zry. This is thought to be the result of improved corrosion resistance conferred by neutron irradiation, which encouraged the dissolution of grain boundary precipitation elements, such as Fe, Cr, and Ni, into the crystal grains. The leaching tests will be continued for 10 years.

An iodine-immobilizing cement solidification process using calcium aluminate cement with gypsum additive was developed. Powdered cement solid was repeatedly immersed in ion-exchanged water with varying liquid-to-solid ratios (L/S) in accelerated dissolution tests simulating interaction with groundwater at waste disposal sites. The measured concentrations of iodine in the water were on the order of 10−5 to 10−3 mol⋅dm−3 in the entire L/S range. These concentration levels are extremely low compared with those in the case of ordinary Portland cement. Calculations with a solution equilibrium model for the cement immersed in ion-exchanged water showed that the observed iodine release profile versus integrated L/S ratio from the immersion test was explained by a dissolution model of minerals in the cement.

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.

The development of an iodine immobilization technique that can fix radioactive iodine in waste form for a long period and constrain its leaching into pore water is necessary in order to secure the long-term safety of geological disposal of transuranic (TRU) waste. Lead borate glass vitrified at a low temperature is regarded as a promising material for immobilizing the Iodine-129 that is recovered from spent AgI filters generated by reprocessing plants in Japan and which may have a significant effect on the long-term safety of geological disposal.

Batch leaching tests were conducted to understand glass dissolution behavior in various solutions that account for geological disposal conditions. Boron dissolved at the highest rate in all types of solutions to be used as an index element for measuring the glass dissolution rate. On the other hand, lead dissolved in these solutions at a much lower rate. These results are consistent with an electron micro-probe analysis (EPMA) of the altered glass surfaces that indicated the depletion of boron and enrichment of lead near the surfaces.

The altered glass surfaces were further examined by scanning and transmission electron microscopy (SEM/TEM) and X-ray diffraction (XRD). SEM/TEM observation showed formation of a porous altered layer consisting of fine crystallites on the pristine glass and euhedral crystals on the altered layer. XRD analysis indicated that the fine crystallites and euhedral crystals are hydrocerussite, Pb3 (CO3)2(OH) 2, which was predicted by geochemical calculation as the precipitate for the experimental system.

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.

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.

Iodine-129 collected from a reprocessing plant is regarded as the dominant nuclide in terms of safety for TRU wastes disposal in Japan. AgI vitrification (AgI-Ag2O-P2O5) is a potential iodine immobilization technique, which has the advantages of less iodine volatilization (low-temperature vitrification) and high volume reduction efficiency (approx. 1/25 the original waste volume). The iodine immobilization property can be evaluated by examining the surface condition of the AgI glass immersed in water. In this study, immersion tests have been performed on AgI glass in pure water in a 3% H2-N2 atmosphere at room temperature, and the surface characteristics have been examined. The thin layer (<4.3 μm) that is formed has been found to consist of AgI, which may act as a barrier, preventing leaching of glass components. The concepts behind the iodine release model have been proposed based on diffusion and the solubility of the components at the glass surfaces.