Actinide chemistry in the nuclear fuel cycle begins with the extraction of uranium from minerals. The common mineral phases are oxides, carbon­ates, phosphates, silicates and vanadates. Uranium recovery is typically

accomplished by leaching extracted ore with sulfuric acid (sometimes with the addition of an oxidizing agent), nitric acid, or sodium carbonate after crushing and grinding of the ore. For uranium from basic minerals, precipi­tation of Na2U2O7 is used for purification. From acid leaching processes, solvent extraction with organophosphorus or trialkylamine extraction agents selectively isolates U(VI) from the matrix. The recovered uranium is then stripped with acidic, neutral or basic solutions (depending on the extraction process employed) to allow further purification/processing and recycle of the extractant solution. Uranium can also be recovered from phosphate minerals using a somewhat more complex solvent extraction process. [9] Among the byproducts of environmental concern are mill tailing residues from uranium mining, containing most of the radioactive decay daughter activity, but also residual uranium.

The recovered uranium is then subjected to further processing to convert it to U3O8. This uranium is oxidized and converted to the fluoride with F2, providing UF6, to allow isotope enrichment. The enriched uranium stream is then sent to fuel fabrication while the residual UF6 comprising isotopi­cally depleted uranium (>99.7% 238U) is stored for later treatment and disposal. Substantial amounts of such depleted uranium reserves are found in storage at various locations.

To prepare fuel, the enriched UF6 is hydrolyzed,

UF6 (g) + 2 H2O ^ UO2F2 + 4 HF (aq) precipitated with ammonia,

2 UO2F2 (aq) + 6 NH4(OH) (aq) ^ (HN^UO (s) + NH4F (aq) + 3 H2O

and the ammonium diuranate product reduced with H2 and converted to UO2. The “green” UO2 is pressed and sintered at 1700°C in a dry H2 atmo­sphere to give a material with a small oxygen excess (UO2+x). The UO2+x ceramic material is machined to the proper size and shape then “canned” in zirconium (or aluminum) fuel pellets which are used to prepare a fuel assembly. The enriched fuel is then used to create power through sustained fission chain reactions in a reactor, producing the mixture of fission and activation products described above.

In the single pass fuel cycle, the used fuel becomes the waste form upon its discharge from the reactor, ultimately destined for disposal in a geologic repository. As long as the used fuel remains dry and it is not altered through outside intrusion, this material can be expected to remain intact, particu­larly if the surroundings remain reducing. However, it cannot be ensured that water will never contact the waste package. Over the millennia that used fuel will remain radiotoxic, alterations in climate are expected to occur, including (with high probability) periods of wet conditions and of glaciation. When this occurs, alterations and potentially oxidation of the

used fuel should be expected to occur, which ultimately leads to the poten­tial for mobilization of the radioactive materials from the fuel and migra­tion to the biosphere.

The effects of radiation damage to the solid fuel contributes an increased friability of the predominantly UO2 matrix, resulting ultimately in the con­version of the monolithic ceramic UO2 to a more readily mobilized powder. Interfacial reactions in the thermal and radiolytic environment will ulti­mately lead to deterioration of the fuel cladding and eventually to contact between the water and the fuel matrix. At this point, leaching of the fuel components by water can result in the mobilization of radioactive materials from the waste package. The distance traveled by the components from the waste package will be governed by the chemistry of the isotopes, redox conditions, strength of the interactions between the solute components and surrounding mineral surfaces, water flow rate, temperature and radiation field strength.

As noted above, those species with the greatest potential for true solubil­ity have the highest probability for significant migration from the repository. Technetium and iodine are of primary concern, followed by Np (in oxidizing conditions, but not in reducing environments) and Cs. Over time, alteration of the mineral phase or of the fuel matrix can either enhance or retard migration potential. In the case of direct disposal of fuel, the matrix is UO2. The dissolution of spent fuel releases radioactive material to the water column. Whether waste migrates any substantial distance from the reposi­tory will be a complex function of the chemistry of the radionuclides and the nature of the surroundings.

Though such assumptions must be used with caution, the release of radio­active materials from the repository will be generally predicted based on thermodynamic models. Thermodynamic models are considered to be rep­resentative, as favorable thermodynamics are a necessary condition for there to be any migration. The validity of such an assumption will depend significantly on the completeness of the model and the ambient conditions. The complexity of the system, with many solid and fluid phase reactions to be taken into account, makes accurate predictions a formidable challenge. In addition, the constant alteration provided by radiolysis, thermal gradi­ents, changes in ground water flow rates and alteration phases ultimately complicate the development of the models and could reduce the validity of thermodynamic model predictions. Colloid transport phenomena are a par­ticular challenge to making accurate predictions of migration potential using thermodynamic modeling due to the substantial variety of colloids that can be created.

A similar range of phenomena and isotopes will be present in high level wastes from reprocessing as practiced today, except for the relative absence of uranium and plutonium isotopes. In a closed loop fuel cycle, uranium is

routed to low or intermediate level waste, plutonium is recycled to the preparation of mixed-oxide fuel (MOX). The matrix of high level waste glass from reprocessing is in most cases borosilicate glass, hence the leach­ing characteristics and mineral alteration products are expected to be dif­ferent from fuel stored in cladding. For high level reprocessing wastes, the outer container is a welded stainless steel can (as opposed to zircalloy fuel cladding in direct disposal of fuel). In any disposal scenario, the tendency for radionuclides to migrate from the point of emplacement is dependent on the ambient conditions including temperature, water flow rate and redox conditions plus the contribution of engineered backfill materials. For actinide ions in general, it is expected that they should remain largely in the repository environment if the conditions remain reducing and intru­sions are limited. The primary mechanism for long distance displacement that can be envisioned for most metallic radioactive species is probably colloid transport; ionic species of low charge (Cs+, Rb+, TcO4-, I-, Br-) would be expected to manifest significantly greater mobility and to transport as simple ions.