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CsAlSi/sub 5/O/sub 12/: a possible host for /sup 137/Cs immobilization

Description: CsAlSi/sub 5/O/sub 12/ exhibits more acid resistance than pollucite (CsAlSi/sub 2/O/sub 6/). At pH values of 1.02 and 1.40, the extraction of Cs from CsAlSi/sub 5/O/sub 12/ at 25/sup 0/C was approximately proportional to the square root of leach time. The Cs extraction at 25/sup 0/C varied as (H/sup +/)/sup 0/ /sup 36/ over the pH range of 1 to 6. Also, the Cs extraction in various brines at 300/sup 0/C/30 MPa was comparable with that for pollucite. CsAlSi/sub 5/O/sub 12/ can be crystallized at about 1000/sup 0/C from calcines if a small amount of CaO is present, but in the absence of such sintering acids, crystallization temperatures of about 1400/sup 0/C are necessary. Compatibility data were also obtained with respect to several other phases with which CsAlSi/sub 5/O/sub 12/ might be expected to coexist in tailored ceramics designed for high-level defense waste.
Date: March 31, 1982
Creator: Adl, T. & Vance, E.R.
Partner: UNT Libraries Government Documents Department

Comparison of sodium zirconium phosphate-structured HLW forms and synroc for high-level nuclear waste immobilization

Description: The incorporation of (a) Cs/Sr as simulated heat-generating isotopes contained in Purex reprocessing waste, (b) simulated actinides, and (c) simulated Purex waste in sodium zirconium phosphate (NZP) has been studied. The samples were prepared by sintering, by hot pressing and by hot isostatic pressing in metal bellows containers. The short-term chemical durability of the phosphate-based material containing Purex waste was within an order of magnitude of that for Synroc-C, as measured by 7-day MCC-1 tests at 90{degrees}C. The dissolution behavior showed evidence of re-precipitation phenomena, even after times as short as 28 days. Potential for improvement of NZP-based ceramics for HLW management is discussed. 19 refs., 4 figs., 3 tabs.
Date: December 31, 1996
Creator: Zyryanov, V.N. & Vance, E.R.
Partner: UNT Libraries Government Documents Department

Comparison of ceramic waste forms produced by hot uniaxial pressing and by cold pressing and sintering

Description: Synroc C waste form specimens prepared using the Australian-developed technology are uniaxially pressed in stainless steel bellows at 1200{degrees}C and 20MPa. This produces a material with high chemical and physical durability and with the radioactivity enclosed inside both the waste form and the bellows. An alternative method of producing the ceramic product is to use cold pressing of pellets followed by reactive sintering to provide densification and mineralization. Depending on the scale of waste form preparation required and on the activity level and nature of the waste streams, the cold press and sinter method may have advantages. To evaluate the effects of production method on waste form characteristics, especially resistance to dissolution or leaching of waste elements, we have prepared two simulated waste samples for evaluation. Both samples were prepared from liquid precursor materials (alkoxides, nitrates, and colloidal silica) and then doped with waste elements. The precursor material in each case corresponded to a basic phase assemblage of 60% zirconolite, 15% nepheline, 10% spinel, 10% perovskite, and 5% rutile. One sample was doped with 25% by weight of U; the other with 10% by weight each of U and Gd. Each sample was calcined at 750{degrees}C for 1 hr. in a 3.5% H{sub 2} in N{sub 2} atmosphere. Then one portion of each sample was hot pressed at temperatures ranging from 1120 to 1250{degrees}C and 20MPa pressure in steel bellows. A separate portion of each sample was formed into pellets, cold pressed, and sintered in various atmospheres at 1200{degrees}C to produce final products about 2/3 cm in diameter. Samples were then examined to determine density of the product, grain sizes of the phases, phase assemblage, and the location of the U and Gd in the final phases. Density data indicate that sintering gives good results provided that the samples are ...
Date: September 1, 1994
Creator: Oversby, V.M. & Vance, E.R.
Partner: UNT Libraries Government Documents Department

Radiation and transmutation effects relevant to solid nuclear waste forms

Description: Radiation effects in insulating solids are discussed in a general way as an introduction to the quite sparse published work on radiation effects in candidate nuclear waste forms other than glasses. Likely effects of transmutation in crystals and the chemical mitigation strategy are discussed. It seems probable that radiation effects in solidified HLW will not be serious if the actinides can be wholly incorporated in such radiation-resistant phases as monazite or uraninite.
Date: March 15, 1981
Creator: Vance, E.R.; Roy, R. & Pillay, K.K.S.
Partner: UNT Libraries Government Documents Department

Comparison of sodium zirconium phosphate and Synroc matrices for immobilization of high-level waste

Description: The aims of the present work were to investigate possible compatibility between sodium zirconium phosphate (NZP) and Synroc titanate phases, to prepare NZP-based waste forms by hot-pressing rather than sintering, and to investigate the incorporation in NZP of (a) Cs/Sr as simulated heat-generating nuclides; (b) simulated actinides; and (c) simulated Purex waste. The NZP samples were prepared by methods similar to those used for Synroc. The precursor NZP phase was formed from tetrabutyl zirconate Zr(OC{sub 4}H{sub 9}){sub 4}, sodium nitrate, and 85% orthophosphoric acid. Simulated waste nitrate solutions were then mixed with the liquid precursor. After stir drying of the precursor, calcination was carried out at 700{degree}C to remove nitrates and organics.
Date: December 31, 1996
Creator: Zyryanov, V. N. & Vance, E. R.
Partner: UNT Libraries Government Documents Department

Distribution and solubility of radionuclides and neutron absorbers in waste forms for disposition of plutonium ash and scraps, excess plutonium, and miscellaneous spent nuclear fuels. 1998 annual progress report

Description: 'The objective of this research is to gain a fundamental understanding of the distributions and the solubility limits for actinides Pu and U and rare earth neutron absorbers such as Gd and Hf in waste forms. This will be accomplished by systematically studying the local structural environments of these constituents in representative waste forms such as glass, ceramics, and vitreous ceramics. Basic knowledge of these issues will provide a technical and scientific basis that can be used by the US Department of Energy (DOE), Environment Management (EM) Program in developing, evaluating, and selecting waste forms for the safe disposal of Pu, spent nuclear fuel, and other transuranic wastes. The work presented here is a summary of the research activity from November 1997 to May 1998. The elucidation of the correlations between the local structural environments of actinides and rare earth neutron absorbers in waste forms as functions of waste form compositions, and waste form processing conditions will also advance basic material science. The work presented here is a summary of the research activity from November 1997 to May 1998. Currently being studied is the effect of the Pu oxidation state on its solubility in borosilicate-based glasses. When glasses are melted in ambient atmosphere, Pu(IV) has been shown to be the dominant oxidation state as determined by ultraviolet-visible-near infrared spectroscopy (UV-VIS-NIR) and x-ray absorption fine structure (XAFS) techniques. However, no literature data are available for glasses containing Pu predominantly as Pu(III) nor the solubility for Pu(III) in the glass. The results of the study demonstrate that in borosilicate glass, Pu(III) is significantly more soluble than Pu(IV). Using x-ray diffraction analysis the solubility of Pu(III) as oxide was determined to be at least 25 mass% in the reduced glass, while it was no greater than 10 mass% in the same glass under ...
Date: June 1, 1998
Creator: Fen, X.; Vance, E. R. & Shuh, D. K.
Partner: UNT Libraries Government Documents Department

Ceramic phases for immobilization of /sup 129/I. [Sodalite and boracite]

Description: Materials for ultimate disposal of /sup 129/I have been studied. At present, iodide-sodalite, though not ideal, appears to be the best material for /sup 129/I immobilization from the aspects of ease of preparation, thermal stability, cost of materials, and leach resistance. Good consolidation of the material was achieved by sintering in air at 1000 to 1200/sup 0/C, but the iodine content was significantly below stoichiometric expectations. Hot aqueous media preferentially removed iodine, apparently by OH/sup -/ substitution in near-neutral solutions, and I reversible reaction Cl/sup -/ exchange occurred in brine. Alternation of the sodalite also took place. Soxhlet leach rates were about 5 x 10/sup -4/ g/cm/sup 2/-day by total weight loss, but physical weathering contributed significantly to this value. Moderate doses of radiation had no observable deleterious structural effects. Iodoboracites seemingly cannot be prepared by ceramic or nonhydrothermal wet chemical techniques. Fe-iodoboracite has inferior thermal stability to iodide-sodalite and was completely altered to hematite after treatment at 200/sup 0/C in deionized water. Silver zeolites retained some iodine in the form of crystalline ..cap alpha..-AgI at temperatures up to 1300/sup 0/C even though heating above approx. 700/sup 0/C altered the alumino-silicate framework. However, some of the iodine appeared to be present as soluble iodine, even in heated materials. Treatment at 200/sup 0/C in deionized water or 2M NaCl significantly decreased the crystallinity of the aluminosilicate framework and the ..cap alpha..-AgI reflections in the x-ray patterns were enhanced. Mild ..gamma.. irradiations (approx. 50 MR) affected the x-ray diffraction patterns of some of the zeolites. Various lead oxyhalides had very poor thermal stability.
Date: July 31, 1981
Creator: Vance, E.R.; Agrawal, D.K.; Scheetz, B.E.; Pepin, J.G.; Atkinson, S.D. & White, W.B.
Partner: UNT Libraries Government Documents Department


Description: Current ANSTO scientific research on wasteform development for mainly high-level radioactive waste is directed towards practical applications. Titanate wasteform products we have developed or are developing are aimed at immobilization of: (a) tank wastes and sludges; (b) U-rich wastes from radioisotope production from reactor irradiation of UO2 targets; (c) Al-rich wastes arising from reprocessing of Al-clad fuels; (d) 99Tc; (e) high- Mo wastes arising from reprocessing of U-Mo fuels and (f) partitioned Cs-rich wastes. Other wasteforms include encapsulated zeolites or silica/alumina beads for immobilization of 129I. Wasteform production techniques cover hot isostatic and uniaxial pressing, sintering, and cold-crucible melting. In addition, building on previous work on speciation and leach resistance of Cs in cementitious products, we are studying geopolymers. Although we have a strong focus on candidate wasteforms for actual wastes, we have a considerable program directed at basic understanding of the wasteforms in regard to crystal chemistry, their dissolution behavior in aqueous media, radiation damage effects and processing techniques.
Date: February 27, 2003
Creator: Vance, E.R.; Perera, D.S.; Stewart, M.W.A.; Begg, B.D.; Carter, M.L.; Day, R.A. et al.
Partner: UNT Libraries Government Documents Department