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Title: High-temperature leaching of an actinide-bearing, simulated high-level waste glass

Abstract

The chemical durability of a simulated high-level waste glass when exposed to high-temperature geologic solutions was investigated. In this study, simulated high-level waste glass-beads (76 to 68 glass)l doped with technetium, uranium, neptunium, plutonium, curium and americium were leached in deionized water, Waste Isolation Pilot Plant salt brine B, and 0.03M sodium bicarbonate solution at 150 and 250/sup 0/C for 2, 4, 8, 16, and 32 days. The resulting solutions were analyzed for several nonradioactive glass components and for the radioactive dopants. The glass exhibited incongruent leaching behavior, i.e., the normalized releases (g-glass/m/sup 2/) based on the different elements spanned four orders of magnitude. Normalized releases based on boron, molybdenum, sodium, cesium, silicon, and technetium were the same within a factor of three. Most of the nonradioactive components of the glass were released more to the salt brine than to the other two solutions. However, silicon, boron, molybdenum, technetium, and the actinides had their lowest releases in the salt brine. Reaction-layer thickness on the glass surface and weight losses of the glass beads were also smallest in the brine solution. Actinide releases were highest in the sodium bicarbonate solution. Calcium, strontium and barium releases decreased with time and temperature; themore » releases of most other elements increased with time and temperature. Solubility appears to be limiting the release of most elements. The leachate pH is controlled by chemical species within the original leachant and by species released as the glass leached. Carbonate ion complexes with some elements including uranium, effectively increasing their release. The more soluble elements including sodium, boron, molybdenum and technetium provide an indication of the actual rate of reaction between the glass and water.« less

Authors:
; ;
Publication Date:
Research Org.:
Pacific Northwest Lab., Richland, WA (USA)
OSTI Identifier:
6494366
Report Number(s):
PNL-3172
ON: DE83008755
DOE Contract Number:
AC06-76RL01830
Resource Type:
Technical Report
Resource Relation:
Other Information: Portions are illegible in microfiche products. Original copy available until stock is exhausted
Country of Publication:
United States
Language:
English
Subject:
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE WASTES FROM NUCLEAR FACILITIES; 36 MATERIALS SCIENCE; GLASS; LEACHING; TEMPERATURE EFFECTS; AMERICIUM; BORON; BRINES; CESIUM; CURIUM; HIGH-LEVEL RADIOACTIVE WASTES; MOLYBDENUM; NEPTUNIUM; PLUTONIUM; RADIOACTIVE WASTE DISPOSAL; SALT DEPOSITS; SILICON; SIMULATION; SODIUM; SODIUM CARBONATES; TECHNETIUM; UNDERGROUND DISPOSAL; URANIUM; WATER; ACTINIDES; ALKALI METAL COMPOUNDS; ALKALI METALS; CARBON COMPOUNDS; CARBONATES; DISSOLUTION; ELEMENTS; GEOLOGIC DEPOSITS; HYDROGEN COMPOUNDS; MANAGEMENT; MATERIALS; METALS; OXYGEN COMPOUNDS; RADIOACTIVE MATERIALS; RADIOACTIVE WASTES; SEMIMETALS; SEPARATION PROCESSES; SODIUM COMPOUNDS; TRANSITION ELEMENTS; TRANSPLUTONIUM ELEMENTS; TRANSURANIUM ELEMENTS; WASTE DISPOSAL; WASTE MANAGEMENT; WASTES; 052002* - Nuclear Fuels- Waste Disposal & Storage; 360604 - Materials- Corrosion, Erosion, & Degradation

Citation Formats

Westsik, J.H. Jr., Harvey, C.O., and Kuhn, W.L. High-temperature leaching of an actinide-bearing, simulated high-level waste glass. United States: N. p., 1983. Web. doi:10.2172/6494366.
Westsik, J.H. Jr., Harvey, C.O., & Kuhn, W.L. High-temperature leaching of an actinide-bearing, simulated high-level waste glass. United States. doi:10.2172/6494366.
Westsik, J.H. Jr., Harvey, C.O., and Kuhn, W.L. 1983. "High-temperature leaching of an actinide-bearing, simulated high-level waste glass". United States. doi:10.2172/6494366. https://www.osti.gov/servlets/purl/6494366.
@article{osti_6494366,
title = {High-temperature leaching of an actinide-bearing, simulated high-level waste glass},
author = {Westsik, J.H. Jr. and Harvey, C.O. and Kuhn, W.L.},
abstractNote = {The chemical durability of a simulated high-level waste glass when exposed to high-temperature geologic solutions was investigated. In this study, simulated high-level waste glass-beads (76 to 68 glass)l doped with technetium, uranium, neptunium, plutonium, curium and americium were leached in deionized water, Waste Isolation Pilot Plant salt brine B, and 0.03M sodium bicarbonate solution at 150 and 250/sup 0/C for 2, 4, 8, 16, and 32 days. The resulting solutions were analyzed for several nonradioactive glass components and for the radioactive dopants. The glass exhibited incongruent leaching behavior, i.e., the normalized releases (g-glass/m/sup 2/) based on the different elements spanned four orders of magnitude. Normalized releases based on boron, molybdenum, sodium, cesium, silicon, and technetium were the same within a factor of three. Most of the nonradioactive components of the glass were released more to the salt brine than to the other two solutions. However, silicon, boron, molybdenum, technetium, and the actinides had their lowest releases in the salt brine. Reaction-layer thickness on the glass surface and weight losses of the glass beads were also smallest in the brine solution. Actinide releases were highest in the sodium bicarbonate solution. Calcium, strontium and barium releases decreased with time and temperature; the releases of most other elements increased with time and temperature. Solubility appears to be limiting the release of most elements. The leachate pH is controlled by chemical species within the original leachant and by species released as the glass leached. Carbonate ion complexes with some elements including uranium, effectively increasing their release. The more soluble elements including sodium, boron, molybdenum and technetium provide an indication of the actual rate of reaction between the glass and water.},
doi = {10.2172/6494366},
journal = {},
number = ,
volume = ,
place = {United States},
year = 1983,
month = 3
}

Technical Report:

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  • Leach tests were conducted using a modified version of the IAEA procedure to study the behavior of glass waste-solution interactions. Release rates were determined for Tc, U, Np, Pu, Am, Cm, and Si in the following solutions: WIPP B salt brine, NaCl (287 g/l), NaCl (1.76 g/1), CaCl/sub 2/ (1.66 g/l), NaHCO/sub 3/ (2.52 g/l), and deionized water. The leach rates for all elements decreased an order of magnitude from their initial values during the first 20 to 30 days leaching time. The sodium bicarbonate solution produced the highest elemental release rates, while the saturated salt brine and deionized watermore » in general gave the lowest release. Technetium has the highest initial release of all elements studied. The technetium release rates, however, decreased by over four orders of magnitude in 150 days of leaching time. In the prepared glass, technetium was phase separated, concentrating on internal pore surfaces. Neptunium, in all cases except CaCl/sub 2/ solution, shows the highest actinide release rate. In general, curium and uranium have the lowest release rates. The range of actinide release rates is from 10/sup -5/ to 10/sup -8/ g/cm/sup 2//day. 25 figures, 7 tables.« less
  • This study investigated how time and temperature affect the leaching of 76-68 waste glass. Two temperature-dependent regions were identified in the results. Above about 250/sup 0/C, hydrothermal alteration and releases of B, Na, and Mo occur faster than would be expected from the results of tests at lower temperatures. At 250/sup 0/C and below, leaching appears to proceed through three steps. Characteristically, in the first step, the pH of the leachate changes with time; in the second, the pH is constant and normalized releases based on B, Mo, Na, and Si are described by the empirical equation: Release (g glass/m/supmore » 2/) = kt/sup 0.67(range 0.58-0.75)/, where k follows an Arrhenius temperature dependence (apparent activation energy = 5.3 x 10/sup 4/ J/mol /sup 0/K); and in the third step, so far observed only at 250/sup 0/C, normalized releases show a weaker time dependence than they do in the second step.« less
  • A 3.5 liter refractory-lined joule-heated glass melter was built to test the applicability of electric melting to vitrify simulated high activity waste (HAW). The HAW streams result from dissolution and separation of Idaho Chemical Processing Plant (ICPP) calcines and/or radioactive liquid waste. Pilot scale melter operations will establish selection criteria needed to evaluate the application of joule heating to immobilize ICPP high activity waste streams. The melter was fabricated with K-3 refractory walls and Inconel 690 electrodes. It is designed to be continuously operated at 1,150 C with a maximum glass output rate of 10 lbs/hr. The first set ofmore » tests were completed using surrogate HAW-sodium bearing waste (SBW). The melter operated for 57 hours and was shut down due to excessive melt temperatures resulting in low glass viscosity (< 30 Poise). Due to the high melt temperature and low viscosity the molten glass breached the melt chamber. The melter has been dismantled and examined to identify required process improvement areas and successes of the first melter run. The melter has been redesigned and is currently being fabricated for the second run, which is scheduled to begin in December 1997.« less
  • The attached vendor report was prepared for Westinghouse Hanford Company by Babcock & Wilcox as documentation of the Phase I Final Test Report, Cyclone Combustion Melter Demonstration.
  • As part of continuing Department of Energy (DOE)-sponsored studies in waste management, the Pacific Northwest Laboratory (PNL) has been conducting the High-Level Waste Immobilization Program. The purpose of this program is to develop and demonstrate technology for incorporating nuclear wastes into final waste forms. The preparation and leach testing of fully radioactive, zinc borosilicate glass, which was prepared from power reactor waste, are described. Leach testing using the International Atomic Energy Association (IAEA) procedure was performed in deionized water for a period of 1.75 years. Leach rates were determined for activation products, fission products, and actinides. These rates ranged frommore » 4 x 10/sup -5/ g of glass/cm/sup 2/-day, based on cesium, to 4 x 10/sup -9/ g of glass/cm/sup 2/-day, based on cerium. Following is the ranking of the release rates of the elements, from highest to lowest: Cs > Sr > Co > Sb > Mn > Pu > Eu > Rh > Cm > Ce. A similar leach test, using the same glass composition but with nonradioactive elements, has recently been completed. The leach rates of Cs and Sr for the nonradioactive glass were found to be in close agreement with those in this study. Slopes calculated from curves of cumulative fractions leached show that radioisotope release begins with a diffusion-type mechanism and changes gradually to a silicate lattice alteration mechanism. Changes in sampling frequency altered the apparent release mechanism when leachant changes were longer than one month. The leach rates were quite constant for samples taken from the top to the bottom of the glass melt, indicating a homogeneous product. Safety assessment studies and modeling programs use leach rates to predict the amount of radioactive material released should the waste be contacted by aqueous solutions. Further tests, focusing on geologic storage conditions and using fully radioactive wastes, are planned.« less