Plutonium Solubility in High-Level Waste Alkali Borosilicate Glass

Marra, J.C.; Crawford, Charles L.; Fox, Kevin M.; Bibler, N.E.

doi:10.2172/1001774

Cited by 4 publications

(3 citation statements)

References 29 publications

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“…La atoms uniformly distributed in the glass structure without any clustering surrounded by about 6 nonbridging oxygen (O) atoms mixed with sodium (Na) and calcium (Ca) atoms to the impairment of the BO 4 groups. Additionally, because of their facile ionization behavior compared with those of lanthanides, various actinides (U, Pu, Np, Am, and Th) can be easily immobilized by conditioning borosilicate glasses. − …”

Section: Types Of Waste Formsmentioning

confidence: 99%

Inorganic Waste Forms for Efficient Immobilization of Radionuclides

et al. 2021

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The immobilization of long-lived radionuclide wastes generated from nuclear power plants (NPPs) during NPP operation, fuel reprocessing, and decommissioning of nuclear reactors is of great environmental concern. Designing suitable matrices with high durability, stability, and resistivity to various physical and chemical conditions (temperature, pressure, radiation, acidity/alkalinity, etc.) are required for the safe disposal and effective immobilization of radioactive wastes. In this review, we highlight and discuss the key developments in the design and applications of major inorganic waste forms for radioactive waste immobilization. Specifically, we review (i) glasses (borosilicates, phosphates, and sintered glasses), (ii) cements (Portland cement, cast stone/saltstone, hydroceramics, and geopolymer), (iii) synrocs (hollandite, zirconolite, and perovskite), and (iv) ceramics (titanates, sodalite, apatite, monazites, and goethite/magnetite). Additionally, we present future challenges that should be addressed during the design and selection of suitable waste forms for the immobilization of radionuclides. We believe that this paper can deepen our understanding of various waste forms and their roles in radioactive waste immobilization.

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Section: Types Of Waste Formsmentioning

confidence: 99%

Inorganic Waste Forms for Efficient Immobilization of Radionuclides

et al. 2021

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“…Separation of Pd from BSG-melt containing simulated high-level radioactive waste (Pd, Mo, Cs, Sr, Zr, Nd, Ni, Fe) using Sn as the solvent metal was studied, on lab-scale, by Hrudananda [77]. The behavior of Pu in the melter was evaluated employing paper works and corresponding analyses of DWPF melter pour samples by Bibler [78]. Rose et al produced a simulated (inactive) borosilicate high-level waste (HLW) glass being on a full-scale vitrification line with composition simulating vitrified oxide fuel (UO 2 ) reprocessing waste [79].…”

Section: Vello Processmentioning

confidence: 99%

Borosilikat Camlar

Karasu¹,

Demirel²,

Aydin³

et al. 2020

El-Cezeri Fen Ve Mühendislik Dergisi

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Thanks to their distinguished properties borosilicate glasses (BSGs) are preferred in many application fields: laboratory glassware, microwave glass cookware, high-quality beverage glassware, aquarium heaters, flashlights, spotlights (high-powered lighting products), high-intensity discharge lamps, telescopes, biomaterials, immobilization and disposal of radioactive wastes, microscope lenses and slides, art construction, aviation (airplanes, UAVs, missiles), etc. This work aims to present general information about BSGs.

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“…The distribution of up to 11.4 wt% plutonium oxide to glass has been demonstrated for lanthanum borosilicate glasses specially formulated to accommodate high plutonium concentrations (Vienna et al 1996). A PuO 2 solubility of ~1% to 7% is more typical for defense HLW glass melted at 1150° C (e.g., Marra et al (2010) and Muller et al (2002)). The molten plutonium metal also would be unlikely to survive exposure to molten silica as the Ellingham diagram shows plutonium metal capable of reducing SiO 2 and many other metal oxides, including Fe 2 O 3 , to their elemental forms.…”

Section: Re4 -Reduction Of Iron Due To Excess Sugar Additionmentioning

confidence: 99%

Effects Influencing Plutonium-Absorber Interactions and Distributions in Routine and Upset Waste Treatment Plant Operations

Delegard

Sinkov

Fiskum

2015

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Permanganate Effects -The solubilities of (hydr)oxide compounds of aluminum, cadmium, chromium, iron, and uranium (but not manganese and nickel), and of plutonium itself, increase with alkaline concentration and thus help preserve the plutonium/absorber ratio in solution. However, the ratio may be altered in permanganate oxidative dissolution of discrete plutonium phases such as PuO 2 ,PuO 2 •xH 2 O, or other low-solubility tetravalent plutonium because of oxidation to more soluble pentavalent or hexavalent plutonium. The ratio also may be altered because of oxidative leaching of Cr(III) phases while compounds of other absorber elements (e.g., aluminum, iron, nickel) are largely redox-indifferent. Plutonium dissolution from genuine washed sludge with permanganate is significantly enhanced by an increase in NaOH concentration, increasing by an average factor of 70 when permanganate oxidative leaching is undertaken at 3 M NaOH as compared with 0.1 or 0.25 M NaOH. The increase is likely due to plutonium being more readily oxidized to the more soluble hexavalent state. Dissolved plutonium species in alkaline solution, irrespective of oxidation state, are anionic so they are not expected to sorb onto the cation-sorbing resin used for 137 Cs removal. Indeed, such sorption is observed to be negligible. Excess permanganate does not appear to have an appreciable effect on the distribution of plutonium to solution based on oxidative leaching of REDOX Process sludge simulants and genuine Cr(III)-bearing Hanford tank sludges. In lab testing, intentional over-dosage of permanganate did not lead to enhanced plutonium leaching at lower alkalinity (0.09 to 0.25 M NaOH), while at higher alkalinity (3 M NaOH), the plutonium concentration increase to ~0.00036 g Pu/L was relatively low and could be mitigated by adding Cr(III) nitrate to the permanganate/manganate-bearing slurry to eliminate excess oxidant, form MnO 2 , and remove ~95% of the solubilized plutonium by coprecipitation. Separate testing showed that plutonium dissolution in the presence of excess permanganate at 2 to 4 M NaOH also can be mitigated by addition of hydrogen peroxide, removing ~75% of the solubilized plutonium. In treatment with Sr/Mn(VII) for 90 Sr and transuranic element removal from Envelope C supernates, both coprecipitated plutonium that remains undissolved and the residual soluble plutonium, should be protected with 100 to 1000 times higher neutron absorber concentrations based on AN-102 supernate studies. Although experimental data on 239 Pu decontamination from AN-107 supernates are lacking, it is likely that plutonium decontamination by Sr/Mn(VII) treatment was similar to that of AN-102 supernates based on solution composition and total alpha analyses. Decontamination of tank SY-101 from dissolved plutonium by permanganate treatment also is found. Overall, treatment of Envelope C wastes by Sr/Mn(VII) should improve criticality safety by carrying plutonium into the solid phase in the presence of co-precipitated iron and especially manganese.Cerium an...

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Plutonium Solubility in High-Level Waste Alkali Borosilicate Glass

Cited by 4 publications

References 29 publications

Inorganic Waste Forms for Efficient Immobilization of Radionuclides

Inorganic Waste Forms for Efficient Immobilization of Radionuclides

Borosilikat Camlar

Effects Influencing Plutonium-Absorber Interactions and Distributions in Routine and Upset Waste Treatment Plant Operations

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