More than 50 tons of weapon-related plutonium (WPu) is expected to become excess andavailable for disposition by the year 2003 in Russia and a similar amount in the United States.Per two reports from the Committee on International Security and Arms Control of the NationalAcademy of Sciences (1994 1 and 1995 2), the hazard of theft and incorporation into nuclearweapons impels us to guard this Pu carefully and as soon as possible to transform it into a formless accessible for use in nuclear weapons. To this end, CISAC adopted the “Spent FuelStandard” for the disposed WPu, which, if met, renders the WPu no greater hazard per kg thanthe much larger amount of reactor Pu in the form of spent fuel, since CISAC finds that separatedRPu can be used for nuclear weapons with little additional difficulty beyond that posed byseparated WPu. Many disposition routes can be eliminated on the basis of cost or other metric.The two principal survivors (of about equal cost and difficulty for the United States) are thepartial burning of WPu as MOX and the direct vitrification of WPu (as oxide) with high-level‘defense wastes’. Both these approaches should be pursued urgently, with experiments toqualify the processes, until one is selected on the basis of hard evidence. Either approach wouldcost about $1 B, within a factor two, to dispose of 50 tons of excess WPu. The CISAC analysiswill be presented, with comments on utility of RPu in weapons, on the DOE PlutoniumDisposition Study, and on ‘explosive criticality’ in the repository.

Plutonium incineration in a uranium-free fuel by a once-through burning cycle in LWR’s followed by geological disposal of the rock-like material as a high level waste is discussed here. For burning plutonium of various origins, zirconium oxide is a promising candidate as inert matrix because it is stabilised by rare earth oxides (Er, Ho, Eu … Y) in a single phase solid solution with a stable cubic structure. In this material, selected rare earth isotopes can also act as burnable poisons. The spent fuel may be licensed as waste material on the basis of the inventory, the stability of the material and the behaviour of natural analogue material (e.g. baddeleyite). A fuel composed of 90-80% ZrO2, 7–14% PuO2 and 3–6% Er2O3 (At%), with potential addition of Y2O3 (as additional stabiliser), is suggested for experimental study. Such a fuel employed in LWR's could generate power effectively while transmuting about 95% of the 239Pu

Zircon (ZrSiO4) is proposed as a waste form for excess weapons-grade plutonium. Zircon is an extremely durable ceramic that is often found as an accessory mineral in Precambrian terranes with ages up to 4 billion years. The chemical durability of zircon in groundwater far exceeds that of other waste forms, as modeled leach rates may be as low as 10-11 g/m2d. At least 10 wt% Pu can substitute for Zr in zircon. Self-radiation damage from alpha decay leads to a crystalline-to-amorphous transformation that is modeled as a function of time and temperature for deep borehole conditions. Based on the results of this assessment, zircon could meet all necessary durability and criticality criteria required for a Pu waste form. The types of data used in this analysis are generally not available for other crystalline ceramics or glasses.

A new method of synthesis for actinide-doped zircon is presented based on studies of zircons formed by crystallization from the reactor core melt generated in the course of the accident at the Chernobyl Nuclear Power Plant. These zircons have compositions in the range (Zr0.94, U0.06)SiO4 to (Zr0.9, U0.1)SiO4. Hot-pressing of oxides was studied to make Zr-based waste forms. The results demonstrate the efficacy of using metallic zirconium in synthesizing high-actinide zircons. In the event of deviations from zircon's ideal stoichiometry, ZrO2 forms, which is also an effective host phase for actinide elements. Waste streams high in zirconium and actinides could be converted into Zr-based waste forms. The adaptation and modification of the mixed-oxide reactor fuel (MOX) production process is proposed as a process for the production of (Zr,Pu)SiO4, a durable waste form for excess weapons plutonium.

The design philosophy of a zirconolite-rich Synroc variant, hot-pressed at 1200°C/20 MPa, for immobilizing excess weapons Pu is summarized. Phase analysis of Synroc variants containing different (nominated) proportions of zirconolite, Pu and Gd is presented. The presence of hollandite allows the incorporation of radioactive Cs to reduce diversion risks. The influence of chloride impurities on Synroc properties was studied to bear on the usage of encapsulated 137CSCI for this purpose. The maximum incorporation of the neutron poisons Gd and Hf in solid solution in zirconolite from XRD/SEM is 0.5–0.7 and 1.0 formula units respectively, with the value for Gd depending on the charge compensation scheme employed. The use of deep boreholes for disposal would promote the self-annealing of radiation damage from α-decays. The production of zirconolite-rich Synroc ceramics via fired PuO2 or a liquid Pu/microsphere precursor is also discussed. The choice of processing atmosphere for a zirconolite-rich ceramic containing Pu is oxidizing to neutral for Pu-only and Pu + Cs immobilization options, and reducing for a Pu + fission product option

A titanate Synroc ceramic for the immobilization of Pu-bearing waste was designed to consist of 70 wt% zirconolite (CaZrTi2O7) + 15 wt% nepheline (NaAlSiO4) + 15 wt% rutile (TiO2). It contained 10 wt% of Pu plus 6 wt% of Gd as a neutron poison. The material was made by our standard sol-gel route, using a mixture of alkoxides and nitrates, followed by stirdrying and calcination. It was fabricated by hot-pressing at 1150–1250°C/20 MPa for 2 hours in a collapsible metal bellows. Though zirconolite was the majority phase, ∼20 wt% of perovskite also formed. Some of the Na, intended for nepheline, partitioned into the titanate phases. 84-day differential total leach rates of Pu were in the order of 10−5 g/m2/d at 90 and 200°C. Companion ceramics using molar substitution of Ce for Pu confirmed the idea that Ce is a good simulant of Pu from a solid state chemical view, but that there are limitations in terms of leach rate parallels.

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.

This paper presents results of an investigation of the microstructures and durabilities of glasses for immobilization of excess Pu, Am, and Cm at the Savannah River Site (SRS). The glasses investigated had compositions based on commercial lanthanide glasses developed in the 1930's. All the glasses were prepared remotely in shielded cells and the analyses performed using instruments in radio hoods or benches. Durabilities were measured using the ASTM C-1285 standard leach test (PCT). Results for three glasses are presented. Two glasses contained 15 and 7 wt.% Pu, respectively. The third glass contained Am and Cm. The 15 wt% Pu glass was not completely amorphous and contained crystals of undissolved PuO2 and as well as dissolved PuO2. The 7 wt% Pu glass was completely amorphous. The Am/Cm glass contained <1 wt% actinides and was homogenous. The PCT tests for the Pu glasses indicated that B and Ba were leached congruently. Sm and Pu were leached at slower rates. In some cases release rates for specific Ba, Sm, and Pu isotopes were measured by analyzing the leachates by mass spectroscopy. For the Am-Cm glass release rates for B and Ba were equal indicating congruent dissolution. All three glasses were 25 to 50 times more durable than the borosilicate glasses developed at SRS for immobilization of the fission product HLW resulting from reprocessing operations at SRS.

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.

Simulated high-bumup nuclear fuel (SIMFUEL) has been leached in synthetic groundwater under oxic conditions. SIMFUEL pellets were ground and sieved to two particle sizes (50–100 and 100–315 μm). An extensive solid characterization of the fragments was carried out by various techniques. Elemental analysis has also been performed prior to the leaching tests.

The release of U and the minor components (Mo, Ba and Sr) was monitored during the long term dissolution experiments (350 days). These minor components exhibit a trend similar to uranium, high release at the beginning followed by a plateau. The M/U calculated ratios show different behavior although after a period of time, depending on the particle sizes, constant ratios were observed.

SIMHUEL powder was used in order to simulate the physical effect of bum-up on the fuel structure. This fact seems to play an important role on the uranium release. A comparison with the results given in the literature for SIMFUEL pellet leaching tests shows good agreement with the values reported.

Aluminum clad fuel and target elements represent approximately 10% of the DOE owned spent nuclear fuels. The uranium in a large fraction of these fuels is highly enriched and is present as uraniumaluminides which are distributed relatively uniformly within an Al-U alloy core. Emerging acceptance criteria are expected to limit the dry storage temperature for aluminum based fuels to approximately 200°C. The rock temperature near the center of a repository may exceed 200°C if the thermal loading approaches 110 kW/acre. This combination may force the placement of canisters containing aluminum based fuels near the repository periphery. The warm, moist environment anticipated at the periphery may provide aggressive conditions for corrosion of the canister and the highly enriched, aluminum based fuels. Peripheral locations may also be the most vulnerable to covert fuel removal operations. Possible consequences of mixing aluminum based fuels with other fuels in a repository are discussed in this paper.

An engineered system for dry storage of aluminum-clad foreign and domestic research reactor spent fuel owned by the United States Department of Energy is being considered to store the fuel up to a nominal period of 40 years prior to ultimate disposition. Scientifically-based criteria for environmental limits to drying and storing the fuels for this system are being developed to avoid excessive degradation in sealed and non-sealed (open to air) dry storage systems. These limits are based on consideration of degradation modes that can cause loss of net section of the cladding, embrittlement of the cladding, distortion of the fuel, or release of fuel and fission products from the fuel/clad system. Potential degradation mechanisms include corrosion mechanisms from exposure to air and/or sources of humidity, hydrogen blistering of the aluminum cladding, distortion of the fuel due to creep, and interdiffusion of the fuel and fission products with the cladding.

The aluminum-clad research reactor fuels are predominantly highly-enriched aluminumuranium alloy fuel which is clad with aluminum alloys similar to 1100, 5052, and 6061 aluminum. In the absence of corrodant species, degradation due to creep and diffusion mechanisms limit the maximum fuel storage temperature to 200°C. The results of laboratoryscale corrosion tests indicate that this fuel could be stored under air up to 200°C at low relative humidity levels (< 20%) to limit corrosion of the cladding and fuel (exposed to the storage environment through assumed pre-existing pits in the cladding). Excessive degradation of fuels with uranium metal up to 200°C can be avoided if the fuel is sufficiently dried and contained in a sealed system; open storage can be achieved if the temperature is controlled to avoid excessive corrosion even in dry air.

We have measured the transuranic content of two transuranic-doped, simulated waste glasses, using inductively coupled plasma-mass spectrometry (ICP-MS), γ-spectrometry, and α-spectrometry. Average concentrations measured by each technique were within ±10% of the as-doped concentrations. We also report the transuranic content of three fully radioactive SRL waste glasses that were determined using γ- and α-spectrometry measurements to deconvolute isobaric interferences present in the ICP-MS analyses.

In the Idaho Chemical Processing Plant (ICPP) waste streams, zirconia is often the waste load limiting species. It modifies the glass network, enhances durability, increases viscosity and induces crystallization. The limits of its dissolution in boroaluminosilicate glass, with magnesia and soda additions were experimentally determined. A ternary compositional surface is evolved to present the isothermal regimes of liquid, liquid+zircon, liquid+forsterite, and liquid phase sintered ceramic. The potential of partitioning the transuranics, transition elements and solutes in these regimes is discussed. The visible Raman spectroscopic results are presented to elucidate the dependence among glass composition, structure and chemical durability.

A TEM study on the kinetics of metastable phase separation in a lithium bearing HLW glass frit is presented. As nothing was known about the kinetics of metastable phase separation below the glass transformation temperature we had to carry out annealing experiments just above and below the transformation temperature, in order to achieve phase separation on a laboratory time scale- Particle growth in the WAKID5 HLW glass frit at 550 °C occurs by a diffusion controlled Ostwald ripening process (about 40 °C above Tg). Metastable phase separation in the lithium bearing glasses even at lower temperatures cannot be excluded from the present study. In NaBSi3O8 glass (Tg ≈ 580 °C) phase separation could be detected at 500 °C well below Tg after about 160 days.

Currently, at the Idaho Chemical Processing Plant (ICPP) there are about 6800 m3 of liquid sodium-bearing and liquid high-level wastes (HLW), and 3800 m3 of solid calcined HLW. One of the waste processing options under consideration includes separation of the HLW into high activity and low activity (LAW) wastes, followed by immobilization. Preliminary glasses were synthesized for the sodium-bearing, alumina-bearing, and the zirconia-bearing LAW fractions after radionuclide separations. The glasses were formed by crucible melting of a mixture of reagent chemicals representative of the LAW waste streams and frit additives at 1200 °C for 5 hours, followed by overnight annealing at 550 °C and furnace cooling of the melt. These glasses were characterized for density, elastic property, viscosity, chemical durability, structural parameters, and glass phase separation. The results are compared with that of the Hanford's standard glass ARM-i, Savannah River's benchmark glass EA, and the ICPP's grout waste form prepared using the simulated non-radioactive sodium-bearing waste fraction.

Using simulated low-level radioactive waste glasses, interaction between phosphate and sulfate in the glass network was systematically studied using Raman spectroscopy. Phosphate and sulfate interaction was shown to depend on the extent of saturation of both species in the glass. At elevated concentration level of phosphate and sulfate in combination, yet below the solubility limits, phosphate and sulfate both separated from the glass. In the 7-day product consistency test (PCT), the phase separated glasses were found to have higher normalized releases of sodium, phosphorous, sulfur, boron, and silicon than those of their baseline glasses without phase separation.

Chromium, iron, and nickel are present in many waste-streams from DOE sites. This study considers a simplified lime-aluminosilicate glass system, similar to one proposed for high temperature (1250–1450°C) vitrification of soil and wastes from DOE sites, in which concentrations of Cr, Fe. and Ni are being varied and different redox states induced for selected melting conditions. The solubilities of Cr 2O3 and NiO in the simplified system are determined at 1350°C, 1400°C, and 1450°C. The enthalpies of solution calculated from the solubility data for Cr2O3 and NiO are 12.2 and 15.2 kcal/mol, respectively. The iron redox ratio, Fe2+/Fetotal in selected glasses is determined by Mössbauer Spectroscopy and related to changes in microstructure. The crystallization of eskolaite (Cr2O3) seems to be favored at intermediate states of reduction (Fe2+/Fetotal ≈0.8) in Cr-containing glass at near the solubility limit. Pseudoternary diagrams are plotted for melts in air at 1450°C in which Fe2O3 + SiO2 + X (X = NiO or Cr2O33) is held constant. Time-Temperature-Transformation (TTT) curves are plotted using data from heat treatment studies. Liquidus temperatures of base, 1.5 wt% Cr2O3 glass and 8 wt% NiO, 7 wt% Fe2O3 glass are found to be 1150°C, 1175°C, and 1175°C, respectively.

Reduction and separation of Pd were investigated in soda-lime-alumino-borosilicate glass. The dependence on raw materials, melting atmosphere, concentration of Pd and melting temperature was examined by electrochemical oxygen activity measurement, optical absorbance, X-ray diffraction and microscopic observation. Pd ions were slowly reduced during melting and metal particles were formed. The separation was enhanced in a reducing atmosphere, at high temperature, for long melting periods and with high concentration of Pd. Pd metal particles of about 10 to 30 μm in diameter were observed by SEM and X-ray diffraction.

Comparative tests on SGN and Radon simulated ILW vitrification with a cold crucible based experimental plant were performed. The batch was fed as paste with 20 wt.% moisture. An operation conditions, the ability of the cold crucible to maintain the differential pressure during the operation, process variables, product properties, off-gas and condensate compositions, the effect of melt agitation on melter capacity and cesium loss have been determined. Melt viscosity and resistivity have been measured. Material structures were studied using infra-red spectroscopy and electron microscopy. Leach rates of sodium and potassium cations as well as Cs-137 have been measured. The behavior of sulfate and chloride ions in the vitrification process as well as their solubility in SGN and “Radon” glasses have been investigated.