Uranium Recovery from a Nuclear Fuel Waste Form

Miller, R.L.; Welch, J.M.; Flinn, J.E.

doi:10.1007/bf03402596

Cited by 4 publications

(3 citation statements)

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“…11,12 A silicate-based fused-cast ceramic waste was tested for uranium recovery by direct sulfuric-acid leaching. 11 Although sulfuric-acid/hydrogen-peroxide leaching is simpler to operate than alkali leaching, the recoveries are marginal. An eightstage, continuous-ion-exchange plant operating on a counter-current fluidizedbed concept to recover uranium from low-grade solutions (about 120 mg/L U 3 O 8 ) was investigated for the recovery of uranium.…”

Section: Enthalpymentioning

confidence: 99%

The pyro- and hydrometallurgical processing of uranium-containing waste

Reddy

2001

JOM

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Overview Waste Treatment & MinimizationAs large amounts of uranium-containing (both high and low-level) waste are generated from activities such as fuel fabrication, fuel reprocessing, and R&D, concern is growing over the safe disposal of these radioactive materials. Over the past few decades, numerous disposal options have been investigated, including pyrometallurgical high-temperature fusion/vitrification, hydrometallurgical processing, biological remediation, polymerization, clay back-filling, and zeolite adsorption. This paper reviews the techniques, thermodynamics, and kinetic aspects of processing uranium-containing waste. In particular, the removal of uranium waste by zeolite adsorption and mechanisms are discussed. In addition, the pyro-and hydrometallurgical conceptual flow sheet is proposed for the disposal of uranium-containing waste.

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Section: Enthalpymentioning

confidence: 99%

The pyro- and hydrometallurgical processing of uranium-containing waste

Reddy

2001

JOM

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“…Ammonium nitrate was tested because ammoniacal solutions have been used to leach Hg from fly ash and other matrices [12][13][14][15]. Caustic dissolution tests for the recovery of U from radioactive waste have also been performed [16,17]. Kitts and Perona tested dilute aqua regia for the treatment of Stainless Steel-Uranium Fuels [18].…”

Section: Introductionmentioning

confidence: 99%

Uranium, Cesium, and Mercury Leaching and Recovery from Cemented Radioactive Wastes in Sulfuric Acid and Iodide Media

Reynier¹,

Lastra²,

Laviolette³

et al. 2015

Minerals

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The Canadian Nuclear Laboratories (CNL) is developing a long-term management strategy for its existing inventory of solid radioactive cemented wastes, which contain uranium, mercury, fission products, and a number of minor elements. The composition of the cemented radioactive waste poses significant impediments to the extraction and recovery of uranium using conventional technology. The goal of this research was to develop an innovative method for uranium, mercury and cesium recovery from surrogate radioactive cemented waste (SRCW). Leaching using sulfuric acid and saline media significantly improves the solubilization of the key elements from the SRCW. Increasing the NaCl concentration from 0.5 to 4 M increases the mercury solubilization from 82% to 96%. The sodium chloride forms a soluble mercury complex when mercury is present as HgO or metallic mercury but not with HgS that is found in 60 °C cured SRCW. Several leaching experiments were done using a sulfuric acid solution with KI to leach SRCW cured at 60 °C and/or aged for 30 months. Solubilization yields are above 97% for Cs and 98% for U and Hg. Leaching using sulfuric acid and KI improves the solubilization of Hg by oxidation of Hg0, as well as HgS, and form a mercury tetraiodide complex. Hg and Cs were selectively removed from the leachate prior to uranium recovery. It was found that U recovery from sulfuric leachate in iodide media using the resin Lewatit TP260 is very efficient. Considering these results, a process including effluent recirculation was applied. Improvements of solubilization due to the recycling of chemical reagents were observed during effluent recirculation.

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“…This section documents the results of arc melter capability demonstration tests (Tests [8][9][10][11][12][13][14] for various simulated waste type compositions stored at the RWMC. The capability tests also evaluated the effects of material addition during arc melter operating conditions.…”

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confidence: 99%

Bench-scale arc melter for R&D in thermal treatment of mixed wastes

Kong¹,

Grandy²,

Watkins³

et al. 1993

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A small dc arc melter was designed and constructed to run bench-scale investigations on various aspects of development for high-temperature (1,500-1,800°C) proccssing of simulatcd transuranic-contaminatcd waste and soil located at the Radioactive Waste Management Complex (RWMC). Several recent system design and treatment studies have shown that high-tempcrature melting is the preferred treatment. The small arc melter is needed to establish techniques and procedures (with surrogates) prior to using a similar melter with the transuranic-contaminatcd wastes in appropriate facilities at the site. This report documents the dcsign and construction, starting and heating procedures, and tests evaluating the mclter's ability to process several waste types stored at the RWMC. lt is found that a thin graphite strip provides reliable starting with initial high current capability for partially melting the soil/waste mixture. The heating procedure includes (1) the initial high current-low voltage mode, (2) a low current-high voltage mode that " commences after some slag has tbrmed and arcing dominates ovcr the receding graphite conduction path, and (3) a predominantly Joule heating mode during which thc currcnt can bc increased within the limits to maintain relatively quiescent operation. Scvcral experiments involving the melting of simulated wastes arc discussed. Energy balancc, slag tcmpcraturc, and electrode wear mcasurements are presented. Recommendations for further refinements to enhance its processing capabilities are identified. Future studies anticipated with the arc melter include waste form processing development; dissolution, retention, volatilization, and collection for transuranic and low-levcl radionuclides, as well as high vapor prcssure metals; clcctrode material development to minimize corrosion and erosion; refractory corrosion and/or skull formation effects; crucible or melter geometry; metal oxidation; and desirable melt reduction/oxidation (redox) conditions.

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Uranium Recovery from a Nuclear Fuel Waste Form

Cited by 4 publications

References 0 publications

The pyro- and hydrometallurgical processing of uranium-containing waste

The pyro- and hydrometallurgical processing of uranium-containing waste

Uranium, Cesium, and Mercury Leaching and Recovery from Cemented Radioactive Wastes in Sulfuric Acid and Iodide Media

Bench-scale arc melter for R&D in thermal treatment of mixed wastes

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