Tc-99 and Cs-137 Volatility from the DWPF Production Melter During Vitrification of the First Macrobatch of HLW Sludge at the Savannah River Site

Bibler, N.E.; Fellinger, T.L.; Marra, S.L.; O’Driscoll, Rosalind; Ray, J.W.; Boyce, W. T.

doi:10.1557/proc-608-697

Cited by 27 publications

(29 citation statements)

References 4 publications

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“…[21] During the processing of SB1A (Macrobatch 1) in the DWPF, sufficient information was obtained to determine the fraction of Tc-99 that was lost to the DWPF off gas system and not immobilized in the glass. [22] This determination was made by comparing the amount of Tc-99 in the feed to the DWPF process to the amount measured in three glass samples from the pour stream of the melter during processing. These results have been published in an earlier report.…”

Section: Possible Losses Of Tc-99 In Dwpf Processingmentioning

confidence: 99%

“…These results have been published in an earlier report. [22] They will be presented here in order to complete the picture of the behavior of Tc-99 in the sludge waste processing. Prior to discussing results of that report, it is helpful to summarize the DWPF process as background information.…”

Section: Possible Losses Of Tc-99 In Dwpf Processingmentioning

confidence: 99%

“…For Canister 409 the measured concentration was greater than the predicted but it was shown that this is due to the analytical uncertainty of the data. [22] However, evidence for the presence of Tc-99 in the off gas deposits was obtained by analyzing three radioactive deposits from the off gas system of Melter One at the DWPF. [24] Melter One was used in the DWPF for eight years during the processing of SB1A, SB1B, and part of SB2 before Melter One failed.…”

Section: Possible Losses Of Tc-99 In Dwpf Processingmentioning

confidence: 99%

“…Also in a small scale melter operated remotely at SRTC, Tc-99 losses were determined to be much higher than in the DWPF melter (31% compared to the 4%). [22] Even though this melter was slurry fed, the feed rate was too slow to allow the formation of a cold cap and thus the Tc-99 losses were higher. Another possibility for the lower volatility of the Tc-99 is that the Tc-99 could have been reduced to the +4 state by the formic acid added in one of the DWPF pretreatment steps.…”

Section: Possible Losses Of Tc-99 In Dwpf Processingmentioning

confidence: 99%

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Technetium-99 Behavior in Savannah River Site High Level Waste Sludges During Waste Processing

Bibler¹,

Fellinger²,

Hobbs³

2004

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Section: Possible Losses Of Tc-99 In Dwpf Processingmentioning

confidence: 99%

Section: Possible Losses Of Tc-99 In Dwpf Processingmentioning

confidence: 99%

Section: Possible Losses Of Tc-99 In Dwpf Processingmentioning

confidence: 99%

Section: Possible Losses Of Tc-99 In Dwpf Processingmentioning

confidence: 99%

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Technetium-99 Behavior in Savannah River Site High Level Waste Sludges During Waste Processing

Bibler¹,

Fellinger²,

Hobbs³

2004

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“…For example, the RSM-1 and -2 data for Cs show a total DF of 3-5 and 9-49, respectively, and the Clemson data show a DF of 2, but data from the DWPF melter show a Cs DF of about 50 (Ref. 32).…”

Section: Dbe-24 Partition Factors For Feed Species In Meltermentioning

confidence: 99%

Review of FY 2001 Development Work for Vitrification of Sodium Bearing Waste

Taylor¹,

Barnes²

2002

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Treatment of sodium-bearing waste (SBW) at the Idaho Nuclear Technology and Engineering Center (INTEC) within the Idaho National Engineering and Environmental Laboratory is mandated by the Settlement Agreement between the Department of Energy and the State of Idaho. This report discusses significant findings from vitrification technology development during 2001 and their impacts on the design basis for SBW vitrification. iv SUMMARYWaste currently stored in tanks at the Idaho Nuclear Technology and Engineering Center (INTEC), located at the Idaho National Engineering and Environmental Laboratory (INEEL), must be processed into waste forms suitable for permanent disposal as part of a Settlement Agreement between the DOE Idaho Operations Office (DOE-ID), the State of Idaho, and the Department of the Navy. In late FY 2000, a roadmap was prepared outlining the technology development required in order to treat this waste, called "sodium-bearing waste" (SBW), using three different technologies. At the direction of DOE-ID, development of vitrification, one of these three technologies, was vigorously pursued in FY-2001. Results from these development activities are contained in twenty-two documents and summarized in tabular form in this report.During FY 2001, a baseline flowsheet for SBW vitrification was defined and a mass balance generated. To manage the assumptions used to generate the process mass balance, the process functional requirements, and all other data that will ultimately be part of the design basis for the SBW vitrification process, a database was created. Much of the knowledge gained in FY 2001 was incorporated into the database and the mass balance, and was published in September, 2001 (Ref. 28). However, results from experiments that were performed late in the year or work that extended into early FY 2002, were not incorporated into the database. Thus a thorough review of the development results from FY 2001 was performed in order to capture in a single source both the progress made in filling data gaps and the impact of these results on the database and hence the basis for the SBW vitrification flowsheet.Significant results were obtained in the areas of feed characterization, feed pretreatment, glass formulation, melter operation, offgas characterization, offgas treatment and secondary waste disposal. Feed characterization data which suggested that the waste may likely contain a higher sulfate content than previously believed together with development of a flowsheet that recycled much of the sulfur in the offgas resulted in development of a glass formulation that tolerates higher sulfur concentration inf the feed, capturing a higher percentage of the feed sulfate in the glass. A maximum waste loading in the glass that forms no salt-layer was found, and the effects of different reductant additives, reductant concentrations, and other process parameters on the formation and growth of a salt layer on the melt were determined. Simulant formulations were developed for both waste from tank WM-180 and...

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Reduction of Pertechnetate by Chemical and Photochemical Approaches and Incorporation of Tc(IV) into Titanium Dioxide

et al. 2022

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Technetium-99 is a prevalent fission product from nuclear waste. The long half-life (211,000 yr) and environmental mobility of pertechnetate (TcO 4 À ) render Tc particularly challenging to isolate and stabilize. Here we present two approaches for development of potential wasteforms using titanium dioxide, TiO 2 . Approach 1 is a low temperature chemical synthesis of TiO 2 doped with Tc(IV) from TcO 4 À intended to mimic the Tc waste stream from the UREX family of separations and removes 98.5 % of the Tc, mainly present as edge-shared Tc(IV) pairs. Approach 2 utilizes TiO 2 to photocatalytically reduce TcO 4 À to Tc(IV) stabilized on the surface of or within the TiO 2 lattice. The %Tc removed from solution and adsorbed to TiO 2 is pH dependent, with the maximum Tc(IV) adsorbed at pH 3-4 as either TcO 2 or edge-sharing Tc(IV) octahedra. The Tc(IV)-TiO 2 composites materials formed by both approaches are suitable for consolidation into a dense wasteform by Hot Isostatic Pressing (HIPing).

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Tc-99 and Cs-137 Volatility from the DWPF Production Melter During Vitrification of the First Macrobatch of HLW Sludge at the Savannah River Site

Cited by 27 publications

References 4 publications

Technetium-99 Behavior in Savannah River Site High Level Waste Sludges During Waste Processing

Technetium-99 Behavior in Savannah River Site High Level Waste Sludges During Waste Processing

Review of FY 2001 Development Work for Vitrification of Sodium Bearing Waste

Reduction of Pertechnetate by Chemical and Photochemical Approaches and Incorporation of Tc(IV) into Titanium Dioxide

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