Cobalt ferrite (CoFe2O4) nanoparticles were successfully synthesized by polymer matrix templated synthesis. Synthetic ion-exchange resins were used as hosts for the coprecipitation of cobalt ferrite where the nanopores of the resin acted as the constrained environment. The weight fraction of the particles within the resin was roughly 16%. X-ray diffraction analysis indicated that cobalt ferrite crystallites were about 1–7 nm in diameter. Magnetic properties measured using an alternating gradient magnetometer showed the particles (prepared using initial salt solution ratio of Fe3+:Co2+ of 1.5:1.0) were superparamagnetic with magnetization M (of the particle-resin system) in the range of 3.0 to 4.4 emu/g at 10 kOe of applied field. The least upper bound of the magnetic size was about 3 nm in diameter. Ratios of Fe3+ to Co2+ within the matrix of the resin before and after precipitation were investigated by x-ray fluorescence spectrometry and were found to be sensitive to the initial salt solution mixture. The ratio Fe3+:Co2+ of 1.5:1.0 was found to be a better mixture in terms of magnetization value. Spherical shape cobalt ferrite particles 2–8 nm in diameter were observed using transmission electron microscopy. The close agreement between the physical and crystallite size indicated that the particles are largely monocrystals.

Two alkali-tin-silicate (ATS) glasses have been prepared at Argonne NationalLaboratory (ANL) as part of our ongoing research in radioactive waste glassdevelopment. These glasses dissolved 5% and approximately 7% Pu. Earlycorrosion test results indicate that Pu-bearing ATS glass is extremelydurable. The initial goal in this project concerned equally both thesolubility of Pu and the durability of the ATS glasses; however, our primaryemphasis has changed recently to maximizing the loading of Pu in the glass.ATS-based glasses, using Th(VI) and Ce(III) as surrogates for Pu(IV), arenow being investigated to increase the solubility of Pu withoutsubstantially sacrificing the durability of the current ATS formulations.The solution data from various corrosion tests on the original Pu-containingATS glasses are also presented.

The release fractions of the five elements in the ε-phase (99Tc, 97Mo, Ru, Rh, and Pd) as well as that of 238U arereported for the reaction of two oxide fuels (ATM-103 and ATM-106) inunsaturated tests under oxidizing conditions. The 99Tc releasefractions provide a lower limit for the magnitude of the spent fuelreaction. The 99Tc release fractions indicate that a surfacereaction might be the rate controlling mechanism for fuel reaction underunsaturated conditions and the oxidant is possibly H2O2, a product of alpha radiolysis of water.

Several alternatives for disposal of surplus plutonium are being considered. One method is incorporating Pu into glass and in this paper we discuss the development and corrosion behavior of an alkali-tin-silicate glass and update results in testing Pu doped Defense Waste Processing Facility (DWPF) reference glasses. The alkali-tin-silicate glass was engineered to accommodate a high Pu loading and to be durable under conditions likely to accelerate glass reaction. The glass dissolves about 7 wt% Pu together with the neutron absorber Gd, and under test conditions expected to accelerate the glass reaction with water, is resistant to corrosion. The Pu and the Gd are released from the glass at nearly the same rate in static corrosion tests in water, and are not segregated into surface alteration phases when the glass is reacted in water vapor. Similar results for the behavior of Pu and Gd are found for the DWPF reference glasses, although the long-term rate of reaction for the reference glasses is more rapid than for the alkalitin-silicate glass.

Tests designed to be similar to the unsaturated and oxidizing conditions expected in the candidate repository at Yucca Mountain are in progress with spent fuel at 90°C. The similarities and the differences in release behavior for 137Cs during the first 2.6 years and the actinides during the first 1.6 years of testing are presented for tests done with (1) water dripped on the fuel at a rate of 0.075 and 0.75 mL every 3.5 days and (2) in a saturated water vapor environment.

Preliminary results for the composition of the leachate from unsaturated tests at 90°C with spent fuel for two successive periods of ~60 days each with pretreated J-13 groundwater are reported. The pH of the leachate solutions ranged from 4 to 7. The americium concentration was 104 to 105 greater than that reported for saturated spent fuel tests in which the leachate pH was 8. The major fraction of material in the leachate was present as colloids containing both americium and curium. The presence of actinides in a form not currently directly included in repository radionuclide transport models provides information that can be used in spent fuel reaction modeling, the performance assessment of the repository and the design of the engineering barrier system.