Toshiaki Ohe

Geochemical properties of bentonite pore water in high-level-waste repository condition

The chemically favorable nature of bentonite pore water is clarified by the PHREEQE geochemical simulation code. Bentonite is viewed as a candidate buffer material for a high-level-waste repository, and bentonite`s pore water chemistry is expected to result in a reduced Eh and weak alkaline pH region. Pyrite (Fe{sub 2}S), initially contained in bentonite, alters to magnetite (Fe{sub 3}O{sub 4}), and this redox couple reaction controls the oxidation reduction potential. A mild alkaline pH condition is produced mainly by an ion exchange reaction between the sodium in bentonite and the protons in the solution. A geochemical simulation of the ion exchange reactions and the pyrite-magnetite alteration suggests that a favorable chemical condition would exist during the waste glass dissolution and indicates that the Ph and the Eh values are {minus}7.5 to {minus}9.4 and {minus}450 to {minus}320 mV, respectively, when the granitic groundwater intrudes into the compacted bentonite in the repository.

LWR high burn-up operation and MOX introduction; fuel cycle performance From the viewpoint of waste management

From the viewpoint of waste management, a quantitative evaluation of LWR nuclear fuel cycle system performance was carried out, considering both higher burn-up operation of UO2 fuel coupled with the introduction of MOX fuel. A major parameter to quantify this performance is the number of high-level waste (HLW) glass units generated per GWd (gigawatt-day based on reactor thermal power generation before electrical conversion). This parameter was evaluated for each system up to a maximum burn-up of 70GWd/THM (gigawatt-day per ton of heavy metal) assuming current conventional reprocessing and vitrification conditions where the waste loading of glass is restricted by the heat generation rate, the MoO3 content, or the noble metal content. The results showed that higher burn-up operation has no significant influence on the number of glass units generated per GWd for UO2 fuel, though the number of glass units per THM increases linearly with burn-up and is restricted by the heat generation rate. On the other hand, the introduction of MOX fuel causes the number of glass units per GWd to double owing to the increase in the heat generation rate. An extended cooling period of the spent fuel prior to reprocessing effectively reduces the heat generation rate for UO2 fuel, while a separation of minor actinides (Np, Am, and Cm) from the high-level waste provides additional reduction for MOX fuel. However, neither of these leads to a substantial reduction in the number of glass units, since the MoO3 content or the noble metal content restricts the number of glass units rather than the heat generation rate. These results suggest that both the MoO3 content and the noble metal content provide the key to reducing the amount of waste glass that is generated, leading to an overall improvement in fuel cycle system performance.

Modeling of Neptunium(V) Sorption Behavior onto Iron-Containing Minerals

Sorption behaviors of neptunium (V) on naturally-occurring magnetite (Fe 3 O 4 ) and goethite (α-FeOOH) in 0.1M NaN0 3 electrolyte solution under aerobic conditions were interpreted using the surface complexation model (SCM). The surface properties of these materials were experimentally investigated by C0 2 -free potentiometric titration, and SCM parameters for the constant capacitance model, such as protonation/deprotonation constants of the surface hydroxyl group, were determined. The number of negatively charged sorption sites of goethite rapidly increased with the increase of the bulk solution pH compared with that of magnetite and this tendency was similar to the pH dependence of neptunium sorption. This implies that the neptunyl cation, NpO 2 + , plays a dominant role in possible sorption reactions. Assuming that the dominant surface complex is XO-NpO 2 , modeling by means of SCM was carried out, and the results were found to agree with experimental data.

Surface complexation modeling for description of actinide sorption at the buffer materials/water interface

Abstract As mechanistic modeling becomes more feasible, surface complexation modeling was applied to actinide sorption reactions with the edge-surface hydroxyl groups of montmorillinite contained in bentonite, which is a potential buffer material for radwaste disposal. The determined surface properties of Japanese bentonite samples suggested possible interactions between positively charged surface sites and anionic Np(V) species in alkaline pH region under high inorganic carbon concentration conditions. The interfacial capacitance density affected determination of the bentonite surface parameters and the Np(V) surface complexation constants. The valence of the aqueous sorbing species was clarified to be useful in determining surface complexation constants of actinide sorption.

Adsorption and diffusion of strontium in simulated rock fractures quantified via ion beam analysis

Abstract An experimental technique has been developed and applied to the problem of determining effective diffusion coefficients and partition coefficients of Sr in low permeability geological materials. This technique, the micro-reactor simulated channel method (MRSC), allows rapid determination of contaminant transport parameters with resulting values comparable to those determined by more traditional methods and also creates product surfaces that are amenable for direct chemical analysis. An attempt to further constrain mass flux was completed by detailed ion beam analysis of polished tuff surfaces (tuff is a polycrystalline polyminerallic aggregate dominated by silicate phases) that had been reacted with Sr solutions at concentrations of 10-5, 10-3 and 10-1 mol 1-1. Ion beam analysis was carried out using beams of both protons (using particle induced X-ray emission and elastic backscattering spectrometry or EBS) and alpha-particles (using Rutherford backscattering spectrometry). The ion beam analyses showed that increased solution concentrations resulted in increased surface concentrations and that in the highest concentration experiment, Sr penetrated to at least 4 μm below the primary interface. The Sr surface concentrations determined by EBS were 0.06 (±0.05), 0.87 (±0.30) and 2.40 (±1.0) atomic weight % in the experiments with starting solution concentrations of 10-5, 10-3, and 10-1 mol 1-1, respectively.

Thermal Impact on Geological Disposal of Hull and End Piece Wastes Resulting from High-Burn-up Operation of LWR and Introduction of MOX Fuels into LWR

The thermal impacts of hull and end piece wastes from high-burn-up UO2 and MOX fuels on a conventional disposal system were investigated. The heat generation rates in the canister containing these wastes were obtained using burn-up calculations of PWR fuels. For wastes from spent MOX fuel, the heat generation rates increase to 3.2–4.5 times that from present-day burn-up spent UO2 fuel when these canisters are disposed of. The temperature distributions in the area around the disposal galleries for these wastes were obtained using two-dimensional thermal analyses byassuming a maximum 80_C temperature exposure of the cement mortar. For wastes from spent MOX fuel, the temperature of the surrounding rock remains at about 60–70°C after disposal, even after 1,000 years. In this case, the number of canisters loaded in a waste package must be decreased from four to around one. This increases the number of waste packages to contain the required number of canisters. It will be important to apply alternative approaches to increase the amount of wastes in a waste package by reducing the amounts of FPs and actinides adhering to hulls and to optimize the layout design of galleries, which may be done by significantly increasing the distance separating neighboring galleries.

Numerical Analysis of Uranium Solubility in Compacted Bentonite by Applying the Activity Correction for Strong Interaction between Liquid/Solid Interface

The solubility of radionuclide in the compacted bentonite is one of the key factors to reduce the radiological impact derived from the high-level radioactive waste disposal. Although efforts to accumulate the experimental evidence and thermodynamic data have been accomplished so far, direct measurements of the solubility-limited concentration of insoluble elements such as Np, Pu, and U in compacted bentonite have not yet been reported. The reason is mainly due to the experimental difficulties to collect the sufficient volume of pore water for concentration measurements because of the extremely low liquid to solid ratio. The effort to overcome such experimental limitation, the theoretical calculations using a geochemical code such as PHREEQE are often achieved 1) as substitution. These calculations are commonly based on the theory for the aqueous system, thus a large disagreement would arise because of the low liquid/solid ratio in the water-saturated compacted bentonite. The low liquid/solid ratio implies strong interaction between the ion and the solid surface and the resulting ionic activity becomes quite different from that in the bulk water in aqueous system. One suggestive finding for the existence of this strong interaction may be the difference of the sorption coefficients measured by the two different techniques; the batch technique confirmed in the high liquid/solid ratio with using suspensions and the column technique using the compacted bentonite with the relatively low liquid/solid ratio. Torikai et al.2) indicated the strong interaction between the water molecule and the montmorillonite surface by using the equilibrium water pressure measurements and concluded that the pore water was not a dilute electrolytic solution. The main objective of the short report is to derive the

The long-term alteration rate of Na-smectite in natural bentonite formation

Abstract Alteration of the natural bentonite in Kuroishi ore deposit located in the north Japan was estimated by the vertical element profiles of Na, Ca and Mg in the drilled core samples. The exchangeable Na depleted from the ground surface to 20 m in depth and the total loss of Na coincided with the accumulation of Ca and Mg. This suggests the Na depletion was caused by the ion exchange reactions. A simple analytical calculation using the steady state approximation indicated the average alteration rate was about 1 cm/1000 years. This value is equivalent to that by geomorphological studies.

Burning of MOX fuels in LWRs; fuel history effects on thermal properties of hull and end piece wastes and the repository performance

The thermal impacts of hull and end piece wastes from the reprocessing of MOX spent fuels burned in LWRs on repository performance were investigated. The heat generation rates in MOX spent fuels and the resulting heat generation rates in hull and end piece wastes change depending on the history of MOX fuels. This history includes the burn-up of UO2 spent fuels from which the Pu is obtained, the cooling period before reprocessing, the storage period of fresh MOX fuels before being loaded into an LWR, as well as the burn-up of the MOX fuels. The heat generation rates in hull and end piece wastes from the reprocessing of MOX spent fuels with any of those histories are significantly larger than those from UO2 spent fuels with burn-ups of 45 GWd/THM. If a temperature below 80°C is specified for cement-based materials used in waste packages after disposal, the allowable number of canisters containing compacted hull and end pieces in a package for 45 and 70 GWd-MOX needs to be limited to a value of 0.4–1.6, which is significantly lower than 4.0 for 45 GWd-UO2.

Experience With Technical Advisory Groups in the Japanese HLW Disposal Programme

Recognising the benefits to be gained by integrating national and international experience into a rapidly growing programme, NUMO already in June 2001 set up both domestic and international technical advisory committees. These groups supported NUMO during the time when many of the basic, often innovative, concepts that characterise the Japanese deep geological disposal programme were developed. Overall experience with those technical advisory committees was very positive: most of the essential programme elements have been put in place and, in some technical areas, NUMO is now up at the front along with other leading programmes. On the other hand, after a decade, NUMO recognized the necessity to revisit their roles and rethink the overall NUMO policy on technical support, advice and review. This paper provides a summary of NUMO’s experience with advisory committees and an outlook on the future use of such bodies.Copyright