Studies in the Carbonate-Uranium System

Blake, Charles A.; Coleman, C.F.; Brown, K.B.; Hill, Douglas G.; Lowrie, R.S.; Schmitt, Jean-Michel

doi:10.1021/ja01604a004

Cited by 65 publications

(26 citation statements)

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“…Because of the strong affinity of uranyl ion to form carbonate complexes, the net reaction for the conditions of our experiments is more appropriately written as (Blake et al 1956):…”

Section: Uranium Mineral Solubility Measurementsmentioning

confidence: 99%

“…The equilibrium solubility of this material was found to maintain a uranium concentration in pure water of about 0.26 M at 30°C. In addition, Blake et al (1956) measured the solubility of Na 4 UO 2 (CO 3 ) 3 (s) in various solutions. In pure water, they determined the equilibrium uranium concentration to be 0.286 M at 26°C.…”

Section: Uranium Mineral Solubility Measurementsmentioning

confidence: 99%

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Hanford Tanks 241-C-203 and 241-C-204: Residual Waste Contaminant Release Model and Supporting Data

Deutsch¹,

Krupka²,

Lindberg³

et al. 2004

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This report describes the development of release models for key contaminants that are present in residual sludge remaining after closure of Hanford Tanks 241-C-203 (C-203) and 241-C-204 (C-204). The release models were developed from data generated by laboratory characterization and testing of samples from these two tanks. Key results from this work are that future releases from the tanks of the primary contaminants of concern ( 99 Tc and 238 U) can be represented by relatively simple solubility relationships between infiltrating water and solid phases containing the contaminants. In addition, it was found that high percentages of 99 Tc in the sludges (20 wt% in C-203 and 75 wt% in C-204) are not readily water leachable, and, in fact, are very recalcitrant. This is similar to 99 Tc results found in related studies of sludges from Tank AY-102 (Lindberg and Deutsch 2003;Krupka et al. 2004). These release models are being developed to support the tank closure risk assessments performed by CH2M HILL Hanford Group, Inc., for the U.S. Department of Energy.The material used for testing consisted of sludge samples collected from the tanks in September 2003. Initial (Tier 1) testing of the sludges consisted of 1) fusion analysis and acid digestion to determine the total composition of the sludges and 2) water leaching to estimate the soluble portion of the solids. Based on the results of these tests, subsequent analyses were conducted (Tier 2). These included X-ray diffraction (XRD) measurements to identify crystalline solids and scanning electron microscopy/energy dispersive spectroscopy (SEM/EDS) analysis to provide a close-up view of the morphologies of the sludge material and identify the major constituents of individual minerals and amorphous solids. Periodic replenishment and selective extraction tests were also conducted during the Tier 2 tests to further evaluate water leachability and identify solid phases limiting the release of contaminants to solution. The test data were used to develop conceptual contaminant release models based on a series of solubility controls.The 99 Tc release model developed from the laboratory testing of C-203 and C-204 sludges is based on the concentrations and solubilities of technetium-bearing solids in contact with pore water migrating through the sludges. There are two stages of technetium release to solution: The 238 U release model for C-203 and C-204 sludges is based on the concentrations and solubilities of uranium-and sodium-bearing minerals in contact with pore water migrating through the sludges. There are three stages of uranium release to solution:1. Initially, soluble čejkaite [Na 4 (UO 2 )(CO 3 ) 3 ] and a sodium nitrate solid are present in the sludges at concentrations of 0.16 g U/g-sludge and 0.22 g NaNO 3 /g-sludge for C-203 and 0.068 g U/g-sludge and 0.04 g NaNO 3 /g-sludge for C-204. The common ion effect due to Na + limits the solubility of čejkaite to 0.19 g U/L until all the NaNO 3 is dissolved. The solubility of NaNO 3 is constant at 629 g/L.2. After all the ...

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“…Because of the strong affinity of uranyl ion to form carbonate complexes, the net reaction for the conditions of our experiments is more appropriately written as (Blake et al 1956):…”

Section: Uranium Mineral Solubility Measurementsmentioning

confidence: 99%

Section: Uranium Mineral Solubility Measurementsmentioning

confidence: 99%

Hanford Tanks 241-C-203 and 241-C-204: Residual Waste Contaminant Release Model and Supporting Data

Deutsch¹,

Krupka²,

Lindberg³

et al. 2004

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show abstract

“…Scanlan (1977), Bullwinkel (1954), and Blake et al (1956) examined the uranyl carbonate complex ions in aqueous solutions that contained additional carbonate added as sodium carbonate or HCOq. Ciavatta et al (1979 and examined changes in the aqueous uranyl species associated with changes in the concentration of initial dissolved uranium and with variations in the pressure of C02 gas in equilibrium with the solution, but in the absence of added carbonate or bicarbonate ions.…”

Section: A Model For the Deposition Of Coffinite And Some Uraninitesmentioning

confidence: 99%

Thermodynamic properties of selected uranium compounds and aqueous species at 298.15 K and 1 bar and at higher temperatures; preliminary models for the origin of coffinite deposits

Hemingway¹

1982

Open-File Report

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“…The solubility of uranyl peroxo carbonato complex ions in 0.5 M Na 2 CO 3 is approximately 80 g/l [9,10]. The solubility of uranyl tris-carbonato complex ions is known to be lower than that of uranyl peroxo carbonato complex ions [26][27][28], but there is no exact data on solubility of uranyl tris-carbonato complex in a system of Na 2 U 2 O x (OH) y (H 2 O) z -UO 2 (CO 3 ) 3 4− -CO 3 2− . The solubility of uranyl tris-carbonato ions in the system of Na 2 U 2 O x (OH) y (H 2 O) z -UO 2 (CO 3 ) 3 4− -UO 2 (O 2 )(CO 3 ) 2 4− -CO 3 2− , as a concentration of uranyl tris-carbonato complex ions in the solution system where the uranium precipitated, was 9.2 g/l.…”

Section: Resultsmentioning

confidence: 99%

Evaluation of the stability of uranyl peroxo-carbonato complex ions in carbonate media at different temperatures

Lee

Chung

Lee

et al. 2012

Journal of Hazardous Materials

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Studies in the Carbonate-Uranium System

Cited by 65 publications

References 0 publications

Hanford Tanks 241-C-203 and 241-C-204: Residual Waste Contaminant Release Model and Supporting Data

Hanford Tanks 241-C-203 and 241-C-204: Residual Waste Contaminant Release Model and Supporting Data

Thermodynamic properties of selected uranium compounds and aqueous species at 298.15 K and 1 bar and at higher temperatures; preliminary models for the origin of coffinite deposits

Evaluation of the stability of uranyl peroxo-carbonato complex ions in carbonate media at different temperatures

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