James C. Cunnane

Chemically bonded phosphate ceramics for low‐level mixed‐waste stabilization

Abstract Novel chemically bonded phosphate ceramics are being developed and fabricated for low‐temperature stabilization and solidification of mixed‐waste streams that are not amenable to conventional high‐temperature stabilization processes because volatiles, such as heavy‐metal chlorides and fluorides, and/or pyrophorics are present in the wastes. Phosphates of Mg, Mg‐Na, and Zr are being developed as candidate matrix materials. In this paper, we present the fabrication procedures for phosphate waste forms with surrogate compositions of three typical mixed‐waste streams, namely ash, cement sludges, and salts. This study was focused, but not limited to, magnesium phosphate‐ash wastestream final waste form. The performance of the final waste forms, such as compression strength, leachability of the contaminants, and durability in aqueous environments were conducted. In addition, parametric studies have been conducted to establish the optimal ash waste loading in the magnesium phosphate binder system. Based...

High-Level Nuclear-Waste Borosilicate Glass: A Compendium of Characteristics

With the imminent startup, in the United States, of facilities for vitrification of high-level nuclear waste, a document has been prepared that compiles the scientific basis for understanding the alteration of the waste glass products under the range of service conditions to which they may be exposed during storage, transportation, and eventual geologic disposal. A summary of selected parts of the content of this document is provided. Waste glass alterations in a geologic repository may include corrosion of the glass network due to groundwater and/or water vapor contact. Experimental testing results are described and interpreted in terms of the underlying chemical reactions and physical processes involved. The status of mechanistic modeling, which can be used for long-term predictions, is described and the remaining uncertainties associated with long-term simulations are summarized.

Aqueous Dissolution of Urania-Thoria Nuclear Fuel

Abstract The aqueous dissolution of irradiated and unirradiated uranium-thorium dioxide, (U,Th)O2, fuel pellets in Yucca Mountain well water has been investigated. Whole and crushed pellets were reacted at 25 and 90°C for periods of up to 195 days. The fuel dissolution was measured by analyzing the concentrations of soluble uranium, thorium, and important fission products (137Cs, 99Tc, 237Np, 239Pu, 240Pu, and 241Am) in the well water. The surface-area-normalized fractional uranium release rates for unirradiated crushed uranium dioxide (UO2) pellets were 10 to 40 times higher than the values for (U,Th)O2 fuel. Similarly, the dissolution rates of irradiated (U,Th)O2 pellets with compositions ranging from 2.0 to 5.2% UO2 were at least two orders of magnitude lower than reported literature values for pure UO2. These results demonstrate an advantage of (U,Th)O2 over UO2 in terms of matrix dissolution in groundwater and suggest that (U,Th)O2 fuel is a more stable long-term waste form than conventional UO2 fuel.

Re-Evaluating Neptunium in Uranyl Phases Derived from Corroded Spent Fuel

Abstract Interest in mechanisms that may control radioelement release from corroded commercial spent nuclear fuel (CSNF) has been heightened by the selection of the Yucca Mountain site in Nevada as the repository for high-level nuclear waste in the United States. Neptunium is an important radionuclide in repository models owing to its relatively long half-life and its high aqueous mobility as neptunyl [Np(V)O2+]. The possibility of neptunium sequestration into uranyl alteration phases produced by corroding CSNF would suggest a process for lowering neptunium concentration and subsequent migration from a geologic repository. However, there remains little experimental evidence that uranyl compounds will, in fact, serve as long-term host phases for the retention of neptunium under conditions expected in a deep geologic repository. To directly explore this possibility, we examined specimens of uranyl alteration phases derived from humid-air–corroded CSNF by X-ray absorption spectroscopy to better determine neptunium uptake in these phases. Although neptunium fluorescence was readily observed from as-received CSNF, it was not observed from the uranyl alteration rind. We establish upper limits for neptunium incorporation into CSNF alteration phases that are significantly below previously reported concentrations obtained by using electron energy loss spectroscopy (EELS). We attribute the discrepancy to a plural-scattering event that creates a spurious EELS peak at the neptunium-MV energy.

Analytical electron microscopy study of colloids from nuclear waste glass reaction

An Analytical Electron Microscopy study of colloidal particles formed during reaction of wste glass has been performed. The effect of waste glass test parameters on colloid formation is examined. Characterization of phases present in the leachate of these tests has shown that layers spalled from the glass and precipitated phases are both sources of colloids in the leachate. Elements, such as uranium, have been found to concentrate within colloidal particles in the leachate.

High-Level Waste Glass Compendium; What it Tells us Concerning the Durability of Borosilicate Waste Glass

Facilities for vitrification of high-level nuclear waste in the United States are scheduled for startup in the next few years. It is, therefore, appropriate to examine the current scientific basis for understanding the corrosion of high-level waste borosilicate glass for the range of service conditions to which the glass products from these facilities may be exposed. To this end, a document has been prepared which compiles worldwide information on borosilicate waste glass corrosion. Based on the content of this document, the acceptability of canistered waste glass for geological disposal is addressed. Waste glass corrosion in a geologic repository may be due to groundwater and/or water vapor contact. The important processes that determine the glass corrosion kinetics under these conditions are discussed based on experimental evidence from laboratory testing. Testing data together with understanding of the long-term corrosion kinetics are used to estimate radionuclide release rates. These rates are discussed in terms of regulatory performance standards.

Microscopic Examination of a Corrosion Front in Spent Nuclear Fuel

Spent uranium oxide nuclear fuel hosts a variety of trace chemical constituents, many of which must be sequestered from the biosphere during fuel storage and disposal. In this paper we present synchrotron xray absorption spectroscopy and microscopy findings that illuminate the resultant local chemistry of neptunium and plutonium within spent uranium oxide nuclear fuel before and after corrosive alteration in an air-saturated aqueous environment. We find the plutonium and neptunium in unaltered spent fuel to have a +4 oxidation state and an environment consistent with solid-solution in the U02 matrix. During corrosion in an air-saturated aqueous environment, the uranium matrix is converted to uranyl u ( v I ) o ~ ~ + mineral assemblage that is depleted in plutonium and neptunium relative to the parent fuel. At the corrosion front interface between intact fuel and the uranyl-mineral corrosion layer, we find evidence of a thin (20 micrometer) layer that is enriched in plutonium and neptunium within a predominantly U4+ environment. Available data for the standard reduction potentials for N ~ ~ ~ + / N ~ ~ + and U O ~ ~ / U ~ + couples indicate that Np(IV) may not be effectively oxidized to Np(V) at the corrosion potentials of uranium dioxide spent nuclear fuel in air-saturated aqueous solutions.

The Behavior or Light Water Reactor Fuel after the Cladding is Breached under Unsaturated Test Conditions

Experiments were conducted to examine the corrosion behavior of fuel and cladding under test conditions selected to determine how fuel with breached cladding may behave under unsaturated repository conditions. The results discussed here were obtained from two test samples, each consisting of ~3.5-in. segments of ATM103, a moderate-burnup (~30 GWd/MtU) PWR fuel, exposed to humid air at 175°C. Visual examination of the samples after 540 days revealed that each had developed an axial crack that passed through a drilled hole in the cladding and ran the full length of the sample. Destructive examination of the fuel and cladding showed that the cladding had experienced extensive fuel-side corrosion. Metallographic examination of the cladding shows the hydride distribution near the drilled hole, the fracture surfaces, and the fuel-side corrosion products. Electron microscopy (SEM, TEM) and electron diffraction analyses were used to characterize the corrosion products. The results indicate that the fuel-side corrosion of the cladding and the specific volume increase associated with the formation of the corrosion products caused the hoop stresses that resulted in the observed axial splitting. The implications of these results for the expected evolution of the spent fuel cladding after the cladding is initially breached in an unsaturated repository are discussed.

Estimating model parameter values for total system performance assessment

The intrinsic dissolution rates of nine borosilicate waste glasses were extracted from the results of MCC-1 tests conducted for durations long enough that the solution pH reached a nearly constant value but short enough that the buildup of dissolved species did not affect the dissolution rate. The effects of the pH and temperature on the measured rates were deconvoluted to determine the sensitivity of the rate to the glass composition. The intrinsic dissolution rates were similar for all of these glasses and were not correlated with the glass composition. The mean and standard deviation of the intrinsic dissolution rates of these glasses are log k{sub 0}/[g/(m{sup 2}{center_dot}d)] = 8.2 {+-} 0.2.