The examination of reacted layers that form as glass reacts can provide insight into reaction mechanisms that control long-term glass reaction, and into processes by which radionuclides are released to solution. Examples are given for the natural glasses obsidian and tektite, and the nuclear waste glasses 131 and 165, where the layers form both in-situ and precipitated from solution, where they are both amorphous and crystal 1ine, where selective incorporation of actinide elements Into stable phases may offer a barrier to release, and where sloughing of the layers from the glass acts as an alternative radionuclide release process to solubility-limited control.Several reaction processes occur simultaneously as glass reacts, and the projection of glass performance to long time periods requires the identification of those processes that dominate the long-term reaction.

2 prototype alloys: RAW-1(Tc) and RAW-2(UTc) suitable for a wide range of waste stream compositions are being evaluated to support development of a waste form degradation model that can be used to calculate radionuclide source terms for a range of waste form compositions and disposal environments. Tests and analyses to support formulation of waste forms and development of the degradation model include detailed characterizations of the constituent phases using SEM/EDS and TEM, electrochemical tests to quantify the oxidation behavior and kinetics of the individual and coupled phases under a wide range of environmental conditions, and corrosion tests to measure the gross release kinetics of radionuclides under aggressive test conditions.

A metallic waste form (MWF) will be used to immobilize cladding hulls from spent sodium-bonded nuclear fuel rods. Laboratory tests were conducted to determine if the HLW glass degradation rate model used in Total System Performance Assessment (TSPA) calculations can be used as an upper bound for the degradation rate of the MWF. Static tests were conducted with monolithic MWF samples in pH buffer solutions to measure the degradation rates at 50, 70, and 90°C. The measured degradation rates of the MWF were compared to the rates calculated with the HLW glass degradation model. The degradation rates were calculated as the product of the specific dissolution rates and the anticipated exposed surface areas of glass and MWF in a breeched canister. This comparison showed that the rate calculated using the HLW glass degradation model bounds the rates measured for the MWF at all pH values and temperatures. These results support the potential use of HLW glass as a surrogate for MWF degradation for the purpose of performance assessment.

A multi-phase iron-based metallic waste form is being developed to immobilize the metallic and Tc-bearing waste streams that are generated during the reprocessing of used nuclear fuels with either aqueous or electrochemical methods. A metallic waste form provides for the efficient processing and immobilization of metallic wastes and other components that can be readily reduced to metals prior to or during immobilization. These waste streams can be processed with added iron (or stainless steels) at about 1600 degrees C to incorporate transition metal fission products and other waste components into durable iron solid solution and intermetallic phases. Work is in progress to (1) formulate and produce an alloy composition to immobilize the anticipated range of waste compositions within a small number of phases, (2) identify processing conditions for producing waste forms with high waste loadings and consistent chemical, physical, and radiological properties, and (3) develop a mechanistically-based corrosion and radionuclide release model for calculating long-term waste form performance under the range of possible disposal conditions. The experimental and modeling approaches are presented with some representative results.

This report describes the experimental work performed at Argonne National Laboratory (ANL) during fiscal year 2004 (FY 04) under the Bechtel SAIC Company, LLC (BSC) Memorandum Purchase Order (MPO), contract number B004210CM3X. Important results related to the technical bases, uncertainties, validation, and conservatism in current source term models are summarized below. An examination of specimens of commercial spent nuclear fuel (CSNF) that had been subjected to corrosion testing for up to 10 years under hydrologically unsaturated conditions was undertaken to elucidate radionuclide release pathways and mechanisms.

A metallic waste form (MWF) will be used to immobilize contaminated cladding hulls recovered after electrometallurgical treatment of spent sodium-bonded nuclear fuel from the Experimental Breeder Reactor-II (EBR-II). Tests were conducted to determine if the high-level waste (HLW) glass degradation model developed for total system performance assessment (TSPA) calculations for the Yucca Mountain repository system can be used to represent the degradation of disposed MWF. Static tests were conducted at 50, 70, and 90 C with monolithic samples of MWF in pH buffer solutions spiked with NaCl at a MWF surface-to-solution volume ratio of about 200 m{sup -1}. Test specimens were prepared from a surrogate MWF ingot containing about 10 mass% U. Solutions were exchanged after 14, 28, and 70 days. The cumulative amount of U released into solution through 70 days was used to calculate the MWF degradation rate for each test condition. The rate was independent of temperature. The rate was highest in acidic solutions, lowest in neutral solutions, and intermediate in alkaline solutions. The uranium release rate from a breached canister, which is the product of the MWF degradation rate and the surface area of two MWF ingots in a canister, was compared with the release rate calculated with the HLW glass degradation model for a glass log at the same temperature and pH values. The uranium release rates measured for MWF are less than the degradation rates calculated for HLW glass (compared on a mass per time basis).

An approach is presented for determining if the models used to calculate the release of radionuclides from defense high-level radioactive waste (HLW) glass for total system performance assessment (TSPA) calculations can be used to account for the release of radionuclides from waste forms other than standard borosilicate glasses. The fractional release rates of radionuclides due to waste form degradation, the available surface area, and the radionuclide inventory in an alternative waste form can be compared with the corresponding models used in TSPA for HLW glasses to determine if those models adequately represent the waste form. This approach is demonstrated for the ceramic and metallic waste forms developed for electrometallurgically treated spent sodium-bonded nuclear fuel. Depending on the waste form, comparisons made with aspects of the HLW glass model may be based on similarities in degradation mechanisms or purely empirical.

A rate expression and model parameter values that provide an upper bound for the dissolution rates of high-level waste glasses were determined for use in performance assessment calculations to evaluate the suitability of the Yucca Mountain site for use as a high-level radioactive waste repository. The effects of temperature and solution pH on the glass dissolution rate were modeled explicitly, whereas the effects of glass composition, solute feedback, and alteration phase formation were bounded. The range and distribution of model parameter values are being redefined to provide realistic glass dissolution rates for total system performance assessment calculations for the Yucca Mountain license application. The results of MCC-1 tests, product consistency tests, vapor hydration tests, and unsaturated (drip) tests are being used to determine model parameter values. This paper describes the model and how test results are used to develop model parameter values.

dFive glasses are being subjected to long-term product consistency tests (PCTs) and vapor hydration tests (VHTs) to (1) study the effect of glass composition on corrosion rates, (2) compare the relative responses in short-term and long-term tests, and (3) provide a data base to support performance assessment calculations for the Hanford low-activity waste disposal system. These glasses were selected for detailed study from a suite of 56 glasses that were formulated to span the composition range of possible low-activity waste glasses for Hanford tank wastes and are being subjected to short-term PCTs and VHTs. The results of PCTs conducted through 140 days are discussed in this paper.

A ceramic waste form (CWF) has been developed to immobilize radioactive electrorefiner salt from the electrometallurgical treatment of spent metallic fuel from the experimental breeder reactor EBR-II. Tests are being carried out to qualify CWF for disposal in a high-level waste repository. There are two major phases in U/Pu-loaded CWF, sodalite (Na8Al6Si6O24Cl2) and glass; and two important minor phases, halite (NaCl) and oxide (U,Pu)O-2. The (U,Pu)O-2 phase is present as colloid-sized particles in the glass, near sodalite-glass phase boundaries. Tests have been conducted to measure the release behavior of the (U,Pu)O-2 particles as the CWF corrodes. The releases of matrix and radioactive elements from CWF samples into water were determined from tests in which crushed material was reacted with water at 90 or 120 degreesC for durations from 7 to 365 days. Colloids in the test solutions were characterized by sequential filtration, followed by analysis of filtrates for cations. Plutonium is released partially as (U,Pu)O-2 colloids similar to the particles in the uncorroded CWT.

The product consistency test (PCT) that is used for qualification of borosilicate high-level radioactive waste (HLW) glasses for disposal can be used for the same purpose in the qualification of the glass-bonded sodalite ceramic waste form (CWF). The CWF was developed to immobilize radioactive salt wastes generated during the electrometallurgical treatment of spent sodium-bonded nuclear fuels. An interlaboratory study was conducted to measure the precision of PCTs conducted with the CWF for comparison with the precision of PCTs conducted with HLW glasses. The six independent sets of triplicate PCT results generated in the study were used to calculate the intralaboratory and interlaboratory consistency based on the concentrations of Al, B, Na, and Si in the test solutions. The results indicate that PCTs can be conducted as precisely with the CWF as with HLW glasses. For example, the values of the reproducibility standard deviation for Al, B, Na, and Si were 1.36, 0.347, 3.40, and 2.97 mg/l for PCT with CWF. These values are within the range of values measured for borosilicate glasses, including reference HLW glasses.

This paper describes how the results of vapor hydration tests (VHTs) are used to model the corrosion of waste glasses exposed to humid air in the glass degradation model for total system performance assessment (TSPA) calculations for the proposed Yucca Mountain disposal system. Corrosion rates measured in VHTs conducted at 125, 150, 175, and 200°C are compared with the rate equation for aqueous dissolution to determine parameter values that are applicable to glass degradation in humid air. These will be used to determine the minimum for the range and distribution of parameter values in calculations for the Yucca Mountain disposal system license application (TSPA-LA). The rate equation for glass dissolution is rate = k E • 10 η • pH • exp(–E a /RT). Uncertainties in the calculated rate due to the range of waste glass compositions and water exposure conditions are taken into account by using a range of values for the rate coefficient k E . The parameter values for the pH dependence (η) and temperature dependence (E a ) and the upper limit for k E are being determined with other tests. Using the values of η and E a from the site recommendation model, the VHT results described in this paper provide a value of log k E = 5.1 as the minimum value for the rate expression. This value will change slightly if different pH-and temperature-dependencies are used for the TSPA-LA model.

Glass bonded sodalite is the ceramic waste form (CWF) being developed to immobilize salt waste generated during electrometallurgical conditioning of spent sodium-bonded fuel. CWF will be prepared using a pressureless consolidation (PC) method. The processing conditions will be selected to optimize waste loading while maintaining physical and chemical durability. The possible use of the Product Consistency Test (PCT) to measure the consistency of CWF products is being evaluated by measuring test repeatability, reproducibility, sensitivity to different compositions and processing conditions, and production consistency. The response of the PCT was the same, within 95% uncertainty limits, in replicate tests with the same product, indicating test repeatability. The response of the PCT was the same for replicate products, within 95% uncertainty limits, indicating production consistency. The response of the PCT was sensitive to PC processing conditions. Releases of Al, B, Na, and Si from the PC CAT in 7-day PCTs were lower than those from Environmental Assessment glass. These observations show that the PCT is an appropriate test for confirming that processing conditions meet specifications during CWF production as well as determining that the CWF has adequate and consistent chemical durability.

An interlaboratory study was conducted to determine the precision with which the composition and chemical durability of a borosilicate glass could be measured and to generate a data base of expected values for that glass. The study was conducted with a low-activity reference material (LRM) glass that was developed for use as a standard material for acceptance testing of immobilized low-activity waste (ILAW) products, including those to be made with Hanford tank wastes. The study provided nine independent measurements of the LRM glass composition and eight independent sets of triplicate product consistency tests (PCTs) at 40°C and 90°C. Statistical analysis of these data indicates that LRM glass is suitable for use as a composition and test standard. The results from this study can be used to evaluate the accuracy of composition analyses and PCTs conducted with LRM glass at other laboratories in conjunction with acceptance tests conducted with ILAW products.