Titanate Ceramic Matrices for Alumina-Rich Wastes

2003 ◽  
Vol 807 ◽  
Author(s):  
Bruce D. Begg ◽  
Eric R. Vance ◽  
Huijun Li ◽  
Terry McLeod ◽  
Nicholas Scales ◽  
...  
Keyword(s):  
Savannah River ◽  
Fuel Cladding ◽  
Process Flexibility ◽  
Magnesia Spinel ◽  
Savannah River Plant ◽  
Waste Loading ◽  
High Level ◽  
Sodium Incorporation ◽  
Titanate Ceramic ◽  
Ceramic Matrices

ABSTRACTIn the early 1980s a synroc variant, SYNROC-D, was developed for immobilisation of high-level defence waste stored at the Savannah River Plant, USA. A key phase in the immobilisation matrix was spinel, used to immobilise the large proportion of iron and alumina in the waste. Here we examine the feasibility of this approach for other alumina-rich wastes, not necessarily containing iron, derived from the dissolution of aluminium fuel cladding. The advantages of using a magnesia spinel, as opposed to hercynite (FeAl2O4), as the primary alumina-bearing phase are discussed in terms of an increase in waste loading and process flexibility. Two options for sodium incorporation, glass and the titanate phase freudenbergite, are considered.

Effects of Alpha, Gamma, and Alpha-Recoil Radiation on Borosilicate Glass Containing Savannah River Plant Defense High-Level Nuclear Waste

1981 ◽  
Vol 6 ◽  
Author(s):  
Ned E. Bibler
Keyword(s):  
Borosilicate Glass ◽  
Deionized Water ◽  
Savannah River ◽  
Γ Radiation ◽  
Leach Rate ◽  
Geologic Storage ◽  
Savannah River Plant ◽  
Pb Ions ◽  
High Level ◽  
Doped Glass

ABSTRACTAt the Savannah River Plant, the reference process for the immobilization of defense high-level waste (DHLW) for geologic storage is vitrification into borosilicate glass. During geologic storage for 106y, the glass would be exposed to ∼3 × 1010 rad of β radiation, ∼1010 rad of γ radiation, and 1018 particles/g glass for both α and α-recoil radiation. This paper discusses tests of the effect of these radiations on the leachability and density of the glass. No effect of the radiations was detected that reduced the effectiveness of the glass for long-term storage of DHLW even at doses corresponding to 106 years storage for the actual glass. For the tests, glass containing simulated DHLW was prepared from frit of the reference composition. Three methods were used to irradiate the glass: external irradiations with beams of ∼200 keV or Pb ions, internal irradiations with Cm–244 doped glass, and external irradiations with Co–60 γ rays. Results with both Xe and Pb ions indicate that a dose of 3 × 1013 ions/cm2 (simulating >106 years storage) does not significantly increase the leachability of the glass in deionized water. Tests with Cm–244 doped glass show no increase in leach rate in deionized water up to a dose of 1.3 × 1018 α and α-recoils/g glass. The density of the Cm–244 doped glass has decreased by 1% at a dose of 1018 particles/g glass. With γ-radiation, the density has changed by <0.05% at a dose of 8.5 × 1010 rad. Results of leach tests in deionized water and brine indicated that this very large dose of γ-radiation increased the leach rate by only 20%. Also, the leach rates are 3 to 4 times lower in brine.

Up Close: Immobilization of High-Level Nuclear Waste at the Savannah River Plant

MRS Bulletin ◽  
1987 ◽  
Vol 12 (5) ◽  
pp. 61-65 ◽  
Author(s):  
M.J. Plodinec
Keyword(s):  
Nuclear Waste ◽  
Melting Process ◽  
Melt Processing ◽  
Savannah River ◽  
Good Product ◽  
Glass Processing ◽  
Savannah River Plant ◽  
Commercial Glass ◽  
High Level Nuclear Waste ◽  
High Level

At the Savannah River Plant (SRP), construction of what will be the world's largest solidification facility for nuclear waste has been under way since 1983. Beginning in 1990, the nearly 100 million liters of liquid high-level nuclear waste now stored on the site will be made into a durable borosilicate glass in this Defense Waste Processing Facility (DWPF).In developing a slurry-fed melting process for the DWPF, we made advances in understanding both glass processing and glass durability. This article focuses on what we learned and what further advances are likely to be made.Generally speaking, the goal of any glass technologist is to make a good glass and to make it well. In the glass industry a good product is whatever people will buy. To make it well means, above all, to make the product as economically as possible. Thus, the commercial glass technologist will control the composition of the melter feed material very closely to ensure that only the components necessary for glass performance are included, and in the least expensive form possible. The commercial glass technologist may also tolerate low yields or specify several stages of post-melt processing if it is necessary to produce a product to demanding specifications.To the nuclear waste glass technologist, however, a good product is one which will be stable in geologic environments for millions of years.

Study of Melted Synroc Doped with Simulated High-Level Waste

1996 ◽  
Vol 465 ◽  
Author(s):  
I. A. Sobolev ◽  
S. V. Stefanovsky ◽  
S. V. Ioudintsev ◽  
B. S. Nikonov ◽  
B. I. Omelianenko ◽  
...  
Keyword(s):  
Bulk Composition ◽  
Barometric Pressure ◽  
High Level Waste ◽  
Savannah River ◽  
Savannah River Plant ◽  
High Level ◽  
A Minor ◽  
Simulated High Level Waste ◽  
Additional Phase

ABSTRACTPreparation and characterization of inductively-melted Synroc containing 20 wt% simulated plant “Mayak” reprocessing waste were performed. The sample bulk composition was as follows, (in wt.%): 55.4 TiO2; 15.8 ZrO2; 7.5 CaO; 7.4 BaO; 4.3 Al2O3 2.0 MnO; 1.8 SiO2; 0.7 Na2O; 1.9 K2O, 0.5 Ce2O3; 1.0 UO2; 0.9 NiO; 0.6 Cr2O3, and 0.2 FeO. The sample was produced by melting in air at 1550–1600 °C under barometric pressure. It is composed of a few crystalline phases and a minor glass phase. Most of the phases (hollandite, zirconolite, perovskite and rutile) are similar to the analogous phases found in the other Synroc formulations. An additional phase with average composition, wt.%: 59.8 TiO2; 15.6 CaO; 7.0 UO2; 5.6 ZrO2; 4.7 MnO; 4.1 Ce2O3, and 1.8 Al2O3 was found. Some elements (Ba, Si, Ni, K, Na, Fe) were present in the phase in negligible quantities. Its formula (Ca2.65U0.3Ce0.2)(Ti7.3Mn0.6Zr0.4Al0.3)O20.0 is rather close to a rare mineral uhligite - Ca3(Ti,Zr,Al)9O20. Another possible counterpart of the phase is murataite-like mineral previously described in tailored ceramic designed for Savannah River Plant wastes fixation. This phase as well as zirconolite are the major host for U in the sample Preliminary data on the material leachability in water at 350 °C and 50 MPa have been obtained Uranium contents in the solution were about 1 ppb and close to the uranium dioxide solubility in deionized water under the same P-T conditions.

scholarly journals Vitrification of High-Level Radioactive Waste in a Small-Scale Joule-Heated Ceramic Melter

1981 ◽  
Vol 6 ◽  
Author(s):  
Gerald B. Woolsey ◽  
M. John Plodinec
Keyword(s):  
Radioactive Waste ◽  
Large Scale ◽  
Liquid Waste ◽  
Glass Frit ◽  
Small Scale ◽  
Savannah River ◽  
Savannah River Plant ◽  
Reference Process ◽  
High Level ◽  
Large Scale Tests

ABSTRACTVitrification is the reference process for the immobilization of radioactive waste from the production of defense materials at the Savannah River Plant (SRP). Since 1979, a small vitrification facility (1 Ib/hr) has been operated at the Savannah River Laboratory using actual SRP waste. In previous studies. dried waste was fed to this smaller melter. This report discusses direct feeding of actual liquid-waste slurries to the small melter. These liquidfeeding tests demonstrated that addition of premelted glass frit to the waste slurry reduces the amount of material volatilized. Results of these tests are in accord with results of large-scale tests with actual waste.

scholarly journals High-Level Radioactive Insoluble Waste Preparation for Vitrification

1984 ◽  
Vol 44 ◽  
Author(s):  
B. A. Hamm ◽  
R. E. Eibling ◽  
M. A. Ebra ◽  
T. Motyka ◽  
H. D. Martin
Keyword(s):  
Radioactive Waste ◽  
Borosilicate Glass ◽  
Savannah River ◽  
Soluble Salts ◽  
High Level Radioactive Waste ◽  
Savannah River Plant ◽  
High Level

AbstractAt the Savannah River Plant (SRP), a process has been developed for immobilizing high-level radioactive waste in a borosilicate glass. The waste is currently stored as soluble salts and insoluble solids. Insoluble waste as stored requires further processing before vitrification is possible. The processes required have been developed and demonstrated with actual waste. They include removal of aluminum in some waste, washing soluble salts out of the insoluble waste, and mercury stripping. Each of the processes and the results with actual SRP waste will be discussed. The benefits of each step will also be included.

Formulation and Processing of Polyphase Ceramics for High Level Nuclear Waste

1981 ◽  
Vol 6 ◽  
Author(s):  
Alan B. Harker ◽  
Peter E. D. Morgan ◽  
David R. Clarke ◽  
John F. Flintoff
Keyword(s):  
Full Range ◽  
Hot Isostatic Pressing ◽  
Fluorite Structure ◽  
Savannah River ◽  
Radioactive Elements ◽  
Fine Grain ◽  
Savannah River Plant ◽  
High Level Nuclear Waste ◽  
High Level ◽  
Tetravalent State

ABSTRACTTwo basic crystalline phase assemblages have been developed for incorporating the full range of Savannah River Plant waste compositions into polyphase ceramic forms. Both phase assemblages provide crystalline host phases, with stable mineral analogues, for all radionuclides in the waste. The first, an alumina based assemblage, immobilizes the radioactive elements in solid solutions of magnetoplumbite and uraninite with the bulk non-radioactive waste elements being present in spinel and nepheline. The second assemblage uses the titanate based “zirconolite” type fluorite structure and the alumina/iron based magnetoplumbite phases to host the radioactive nuclei with spinel and nepheline, again providing crystalline hosts for the non-radioactive elements. Both phase assemblages can be consolidated to a fine grain ceramic by hot isostatic pressing at 1040°C pressures from 20,000 to 30,000 psi. Redox control during processing, just sufficient to reduce uranium to the tetravalent state, is used.

scholarly journals Near-Surface Leaching Studies of Pb-Implanted Savannah River Waste Glass

1981 ◽  
Vol 11 ◽  
Author(s):  
G. W. Arnold ◽  
C. J. M. Northrup ◽  
N. E. Bibler
Keyword(s):  
Energy Deposition ◽  
Surface Damage ◽  
High Level Waste ◽  
Savannah River ◽  
Near Surface ◽  
Ion Tracks ◽  
Savannah River Plant ◽  
High Level ◽  
Storage Times ◽  
Long Storage

Ion implantation into simulated nuclear waste glasses is a rapid means of producing near-surface energy deposition similar to that produced by α-recoil nuclei after long storage times (typically 103 – 106 years). For example, Dran, Maurette, and Petit1 used 200 keV Pb-ion implantations in glass at a fluence of 5 × 1012/cm2 to produce surface damage. This fluence is equivalent to approximately 2 × 1018 alpha-decays/cm3 which corresponds to approximately 106 years storage for glass containing Savannah River Plant (SRP) defense high-level waste (DHLW). These authors1 found that this fluence value corresponded to a critical fluence (Φc) for enhanced etching (a factor of 20 increase as inferred by step-height changes) for several silicate glasses when etched in a NaCl solution at 100°C. This critical fluence value also corresponds very well with the fluence at which significant overlap of individual ion tracks occurs.

Mercury Reduction and Removal During High-Level Radioactive Waste Processing and Vitrification

1981 ◽  
Vol 6 ◽  
Author(s):  
Russell E. Eibling ◽  
John R. fowler
Keyword(s):  
Radioactive Waste ◽  
Formic Acid ◽  
Borosilicate Glass ◽  
Processing Temperature ◽  
Savannah River ◽  
High Level Radioactive Waste ◽  
Savannah River Plant ◽  
Reference Process ◽  
Glass Melter ◽  
High Level

ABSTRACTA reference process for immobilizing the high-level radioactive waste in borosilicate glass has been developed at the Savannah River Plant. This waste contains a substantial amount of mercury from separations processing. Because mercury will not remain in borosilicate glass at the processing temperature, mercury must be removed before vitrification or must be handled in the off-gas system. A process has been developed to remove mercury by reduction with formic acid prior to vitrification. Additional benefits of formic acid treatment include improved sludge handling and glass melter redox control.

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