scholarly journals Ceramic Waste Form for Residues from Molten Salt Oxidation of Mixed Wastes

1995 ◽  
Vol 412 ◽  
Author(s):  
Richard A. Van Konynenburg ◽  
Robert W. Hopper ◽  
Joseph A. Rard ◽  
Frederick J. Ryerson ◽  
Douglas L. Phinney ◽  
...  
Keyword(s):  
Molten Salt ◽  
Chemical Elements ◽  
Chemical Durability ◽  
Waste Form ◽  
Treatment Standards ◽  
Oxidation Treatment ◽  
Ceramic Waste ◽  
Low Level ◽  
Mixed Wastes ◽  
Universal Treatment

AbstractA ceramic waste form based on Synroc-D is under development for the incorporation of the mineral residues from molten salt oxidation treatment of mixed low-level wastes. Samples containing as many as 32 chemical elements have been fabricated, characterized, and leach-tested. Universal Treatment Standards have been satisfied for all regulated elements except two (lead and vanadium). Efforts are underway to further improve chemical durability.

scholarly journals Incorporation of radionuclides from the electrometallurgical treatment of spent fuel into a ceramic waste form

1999 ◽  
Vol 556 ◽  
Author(s):  
C. Pereira ◽  
M. C. Hash ◽  
M. A. Lewis ◽  
M. K. Richmann ◽  
J. Basco
Keyword(s):  
Ion Exchange ◽  
Molten Salt ◽  
Nuclear Fuel ◽  
Argonne National Laboratory ◽  
Fission Products ◽  
Experimental Results ◽  
Waste Form ◽  
Chloride Salt ◽  
Zeolite A ◽  
Ceramic Waste

AbstractAn electrometallurgical process is being developed at Argonne National Laboratory to treat spent metallic nuclear fuel. In this process, the spent nuclear fuel is electrorefined in a molten salt to separate uranium from the other constituents of the fuel. The treatment process generates a contaminated chloride salt that is incorporated into a ceramic waste form. The ceramic waste form, a composite of sodalite and glass, contains the fission products (rare earths, alkalis, alkaline earth metals, and halides) and transuranic radionuclides that accumulated in the electrorefiner salt. These radionuclides are incorporated into zeolite A, which can fully accommodate the salt in its crystal structure. The radionuclides are incorporated into the zeolite by hightemperature blending or by ion exchange. In the blending process the salt and zeolite are simply tumbled together at >450°C (723 K), but in the ion exchange process, which yields a product more highly concentrated in fission products, the molten salt is passed through a bed of the zeolite. In either case, the salt-loaded zeolite A is mixed with glass frit and hot isostatically pressed to produce a monolithic leach resistant waste form.Zeolite is converted to sodalite during hot pressing. This paper presents experimental results on the experimental results on the fission product uptake of the zeolite as a function of time and salt composition.

scholarly journals Preparation and characterization of multiphase ceramic designer waste forms

2021 ◽  
Vol 11 (1) ◽  
Author(s):  
Braeden M. Clark ◽  
Priyatham Tumurugoti ◽  
Shanmugavelayutham K. Sundaram ◽  
Jake W. Amoroso ◽  
James C. Marra
Keyword(s):  
Fixed Ratio ◽  
Chemical Durability ◽  
Waste Form ◽  
Ceramic Waste ◽  
Waste Forms ◽  
Plasma Sintering ◽  
Long Term Performance ◽  
Term Performance ◽  
Elemental Release

AbstractThe long-term performance, or resistance to elemental release, is the defining characteristic of a nuclear waste form. In the case of multiphase ceramic waste forms, correlating the long-term performance of multiphase ceramic waste forms in the environment to accelerated chemical durability testing in the laboratory is non-trivial owing to their complex microstructures. The fabrication method, which in turn affects the microstructure, is further compounding when comparing multiphase ceramic waste forms. In this work, we propose a “designer waste form” prepared via spark plasma sintering to limit interaction between phases and grain growth during consolidation, leading to monolithic high-density waste forms, which can be used as reference materials for comparing the chemical durability of multiphase waste forms. Designer waste forms containing varying amounts of hollandite in the presence of zirconolite and pyrochlore in a fixed ratio were synthesized. The product consistency test (PCT) and vapor hydration test (VHT) were used to assess the leaching behavior. Samples were unaffected by the VHT after 1500 h. As measured by the PCT, the fractional Cs release decreased as the amount of hollandite increased. Elemental release from the zirconolite and pyrochlore phases did not appear to significantly contribute to the elemental release from the hollandite phase in the designer waste forms.

Study of Natural Minerals of U-Ryrochlore-Type Structure as Analogues of Plutonium Ceramic Waste-form

2002 ◽  
Vol 713 ◽  
Author(s):  
Roman V. Bogdanov ◽  
Yuri F. Batrakov ◽  
Elena V. Puchkova ◽  
Andrey S. Sergeev ◽  
Boris E. Burakov
Keyword(s):  
Chemical Shifts ◽  
Pyrochlore Structure ◽  
Type Structure ◽  
Waste Form ◽  
X Ray ◽  
Ceramic Waste ◽  
Natural Minerals ◽  
The Us ◽  
Pyrochlore Type

ABSTRACTAt present, crystalline ceramic based on titanate pyrochlore, (Ca,Gd,Hf,Pu,U)2Ti2O7, is considered as the US candidate waste form for the immobilization of weapons grade plutonium. Naturally occuring U-bearing minerals with pyrochlore-type structure: hatchettolite, betafite, and ellsworthite, were studied in orders to understand long-term radiation damage effects in Pu ceramic waste forms. Chemical shifts (δ) of U(Lδ1)– and U(Lβ1) – X-ray emission lines were measured by X-ray spectrometry. Calculations were performed on the basis of a two-dimensional δLá1- and δLδ1- correlation diagram. It was shown that 100% of uranium in hatchettolite and, probably, 95-100% of uranium in betafite are in the form of (UO2)2+. formal calculation shows that in ellsworthite only 20% of uranium is in the form of U4+ and 80% of the rest is in the forms of U5+ and U6+. The conversion of the initial U4+ ion originally occurring in the pyrochlore structure of natural minerals to (UO2)2+ due to metamict decay causes a significant increase in uranium mobility.

A Description of the Ceramic Waste Form Production Process from the Demonstration Phase of the Electrometallurgical Treatment of EBR-II Spent Fuel

2001 ◽  
Vol 134 (3) ◽  
pp. 263-277 ◽  
Author(s):  
Michael F. Simpson ◽  
K. Michael Goff ◽  
Stephen G. Johnson ◽  
Kenneth J. Bateman ◽  
Terry J. Battisti ◽  
...  
Keyword(s):  
Production Process ◽  
Waste Form ◽  
Spent Fuel ◽  
Ceramic Waste

Aging Studies of Pu-238 and -239 Containing Calcium Phosphate Ceramic Waste-forms

2013 ◽  
Vol 1518 ◽  
pp. 73-78 ◽  
Author(s):  
Shirley K. Fong ◽  
Brian L. Metcalfe ◽  
Randall D. Scheele ◽  
Denis M. Strachan
Keyword(s):  
Calcium Phosphate ◽  
Accelerated Aging ◽  
Waste Form ◽  
Calcium Phosphate Ceramic ◽  
Ceramic Waste ◽  
Waste Forms ◽  
Aging Studies ◽  
Flow Through ◽  
Pyrochemical Reprocessing

ABSTRACTA calcium phosphate ceramic waste-form has been developed at AWE for the immobilisation of chloride containing wastes arising from the pyrochemical reprocessing of plutonium. In order to determine the long term durability of the waste-form, aging trials have been carried out at PNNL. Ceramics were prepared using Pu-239 and -238, these were characterised by PXRD at regular intervals and Single Pass Flow Through (SPFT) tests after approximately 5 yrs.While XRD indicated some loss of crystallinity in the Pu-238 samples after exposure to 2.8 x 1018 α decays, SPFT tests indicated that accelerated aging had not had a detrimental effect on the durability of Pu-238 samples compared to Pu-239 waste-forms.

Progress on Ongoing Waste form HIP projects at ANSTO

MRS Advances ◽  
2018 ◽  
Vol 3 (20) ◽  
pp. 1059-1064 ◽  
Author(s):  
Eric R. Vance ◽  
Dorji T. Chavara ◽  
Daniel J. Gregg
Keyword(s):  
Spent Nuclear Fuel ◽  
Technology Development ◽  
Hot Isostatic Pressing ◽  
Waste Form ◽  
Process Technology ◽  
Ceramic Waste ◽  
Waste Forms ◽  
Produce Glass ◽  
Form Development ◽  
Year 2000

Abstract:Since the year 2000, Synroc has evolved from the titanate full-ceramic waste forms developed in the late 1970s to a hot isostatic pressing (HIP) technology platform that can be applied to produce glass, glass–ceramic, and ceramic waste forms and where there are distinct advantages over vitrification in terms of, for example, waste loading and suppressing volatile losses. This paper describes recent progress on waste form development for intermediate-level wastes from 99Mo production at ANSTO, spent nuclear fuel, fluoride pyroprocessing wastes and 129I. The microstructures and aqueous dissolution results are presented where applicable. This paper provides perspective on Synroc waste forms and recent process technology development in the nuclear waste management industry.

Preliminary Studies of the Disposition of Cesium in a Glass-Bonded Sodalite Waste form

2002 ◽  
Vol 713 ◽  
Author(s):  
Marsha J. Lambregts ◽  
Steven M. Frank
Keyword(s):  
National Laboratory ◽  
Argonne National Laboratory ◽  
Fission Products ◽  
Waste Form ◽  
Resonance Spectroscopy ◽  
Ceramic Waste ◽  
Long Term Storage ◽  
Geological Repository ◽  
Term Storage

ABSTRACTArgonne National Laboratory has developed an electrometallurgical treatment for DOE spent metallic nuclear fuel. Fission products are immobilized in a durable glass bonded sodalite ceramic waste form (CWF) suitable for long term storage in a geological repository. Cesium is estimated to be in the waste form at approximately 0.1 wt.%. The exact disposition of cesium was uncertain and it was believed to be uniformly distributed throughout the waste form. A correlation of X-ray diffractometry (XRD), electron microscopy (EM), and nuclear magnetic resonance spectroscopy (NMR) performed on surrogate ceramic waste forms with high cesium loadings found a high cesium content in the glass phase and in several non-sodalite aluminosilicate phases. Cesium was not detected in the sodalite phase.

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