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Curi urn-244 Initial Amount in Sample A, , 0.030863 C i " Concentration in leachate, show units. **8,=corrected concentration x VL x factor to conven to same units as A,.

Spent light-water-reactor (LWR) fuel with an average burnup of 28,000 MWd/ MTU was leach-tested at 25°C using a modified version of the International Atomic Energy Agency (IAEA) procedure. Leach rates were determined from tests conducted in five different solutions: deionized water, sodium chloride (NaCl), sodium bicarbonate (NaHC0 3), calcium chloride (CaC1 2) and Waste Isolation Pilot Plant (WIPP) "B" brine solutions. Elemental leach rates are reported based on the release of 90Sr + 90y, 106Ru , 137Cs , 144Ce , 154 Eu , 239+240pu , 244' Cm and total uran i urn. After 467 days of cumulative leaching, the elemental leach rates are highest in deionized water. The elemental leach rates in the different solutions generally decreased from deionized water to the O.03~ NaCl solution to the WIPP "B" brine solution to the 0.03~ NaHC0 3 solution and was a factor of 20 lower in 0.015~ CaC1 2 solution than in deionized water. The leach rates of spent fuel and borosilicate waste-glass were also compared. In sodium bicarbonate solution, the leach rates of the two waste forms were nearly equal, but the glass was increasingly more resistant than spent fuel in calcium chloride solution, followed by sodium chloride solution, WIPP "B" brine solution and deionized water. In deionized water the glass, based on the elemental release of plutonium and curium, was 50 to 400 times more leach resistant than spent fuel.

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DESCRIPTION OF WORK REPORTED This report presents the results of a test run that produced glass from simulated radioactive wastes. The simulated wastes approximated wastes that would result from the radionuclide removal process. The wastes were vitrified in an electric melter, and the glass product was cast in metal canisters. These glass castings were subsequently evaluated to determine chemical composition, density, leachability, and surface area both for the "as-cast" condition and after a 7.6-m (25-f) drop onto a massive pad. The objective of the week-long run was to demonstrate the technical feasibility of vitrifying Hanford wastes in a joule-heated ceramic melter and of producing a large prototypic casting in metal containers. APPROACH USED The technical feasibility of incorporating Hanford wastes into borosilicate glass was explored by converting 8180 kg of simulated waste and glass formers into glass in a joule-heated ceramic melter. The waste feed was made by mixing powdered chemicals into a chemical blend or batch(a) whose composition was typical of that for a Hanford waste depleted of sodium salts. Glass produced was collected in four canisters that had 0.41-, 0.61-, or 0.91-m (16-, 24-, or 36-in.) diameters. The glass in the canisters was subjected to various tests to determine its characteristics. CONCLUSIONS Experience gained in the week-long vitrification test and characterization of the glass produced in the run support the following conclusions: (a) "Batch" is the common word used in the glass industry to describe a chemical blend of oxides, salts and/or minerals. Batch should not be confused with an operational mode-i.e., batch vs. continuous. iii Summary of Porosity Characteristics of Simulated Waste Glass Specimens, Test Series RHO-2H-3• • 45 Relative Peak Heights from X-Ray Fluorescence Analyzer Results for Samples from Canister N o. 3 .-47 X-Ray Fluorescence Analysis of Samples from Canister No.1.