scholarly journals Leach Resistance Properties and Release Processes for Salt-Occluded Zeolite A

1992 ◽  
Vol 294 ◽  
Author(s):  
M. A. Lewis ◽  
D. F. Fischer ◽  
J. J. Laidler
Keyword(s):  
Alkali Metal ◽  
Alkali Metal Cation ◽  
Fission Products ◽  
Waste Form ◽  
Zeolite A ◽  
Spent Fuel ◽  
Release Process ◽  
Initial Release ◽  
Pyrometallurgical Processing ◽  
Release Processes

ABSTRACTThe pyrometallurgical processing of spent fuel from the Integral Fast Reactor (IFR) results in a waste of LiCI-KCI-NaCI salt containing approximately 10 wt% fission products, primarily CsCI and SrCI2. For disposal, this waste must be immobilized in a form that it is leach resistant. A salt-occluded zeolite has been identified as a potential waste form for the salt. Its leach resistance properties were investigated using powdered samples. The results were that strontium was not released and cesium had a low release, 0.056 g/m2 for the 56 day leach test. The initial release (within 7 days) of alkali metal cations was rapid and subsequent releases were much smaller. The releases of aluminum and silicon were 0.036 and 0.028 g/m2, respectively, and were constant. Neither alkali metal cation hydrolysis nor exchange between cations in the leachate and those in the zeolite was significant. Only sodium release followed t0.5 kinetics. Selected dissolution of the occluded salt was the primary release process. These results confirm that salt-occluded zeolite has promise as the waste form for IFR pyroprocess salt.

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 Microstructure and Leaching Characteristics of a Technetium Containing Metal Waste Form

1999 ◽  
Vol 556 ◽  
Author(s):  
S. G. Johnson ◽  
D. D. Keiser ◽  
M. Noy ◽  
T. O'Holleran ◽  
S. M. Frank
Keyword(s):  
Stainless Steel ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
Fission Products ◽  
Point Of View ◽  
Waste Form ◽  
Nominal Composition ◽  
Spent Fuel ◽  
Waste Forms ◽  
Metal Waste

AbstractArgonne National Laboratory is developing an electrometallurgical treatment for spent fuel from the experimental breeder reactor II. A product of this treatment process is a metal waste form that incorporates the stainless steel cladding hulls, zirconium from the fuel and the fission products that are noble to the process, i.e., Tc, Ru, Pd, Rh, Ag. The nominal composition of this waste form is stainless steel/15 wt% zirconium/ 1–4 wt% noble metal fission products. The behavior of technetium is of particular importance from a disposal point of view for this waste form due to its long half life, 2.14E5 years, and its mobility in groundwater. To address these concerns a limited number of spiked metal waste forms were produced containing Tc. These surrogate waste forms were then studied using scanning electron microscopy (SEM) and selected leaching tests.

scholarly journals Evaluation of Standard Durability Tests Towards the Qualification Process for the Glass-Zeolite Ceramic Waste Form

1996 ◽  
Vol 465 ◽  
Author(s):  
L. J. Simpson ◽  
D. J. Wronkiewicz
Keyword(s):  
Ionic Strength ◽  
Ion Exchange ◽  
Structural Integrity ◽  
Spent Nuclear Fuel ◽  
Fission Products ◽  
High Ionic Strength ◽  
Waste Form ◽  
Zeolite A ◽  
Acceptance Criteria ◽  
Ceramic Waste

ABSTRACTGlass-bonded zeolite is being developed as a potential ceramic waste form for the disposition of radionuclides associated with the U.S. Department of Energy's (DOE's) spent nuclear fuel conditioning activities. The utility of several standard durability tests [e.g., Materials Characterization Center Test #1 (MCC-1), Product Consistency Test-B (PCT-B), and Vapor Hydration Test (VHT)] was evaluated as a first step in developing methods and criteria that can be applied towards the process of qualifying this material for acceptance into the DOE Civilian Radioactive Waste Management System. The effects of pH, leachant composition, and sample surface-area-to-leachant-volume ratios on the durability test results are discussed, in an attempt to investigate the release mechanisms and other physical and chemical parameters that are important for the acceptance criteria, including the establishment of appropriate test methodologies required for product consistency measurements.Results from PCT-Bs conducted with 4 μm diameter salt-loaded zeolite powder indicate that a good correlation exists between release rate and ionic size and/or charge for the release behavior of the simulated fission products in deionized water (DRV), EJ-13 groundwater, and brine solutions. Simulated divalent and trivalent fission products [Sr, Ba, and rare earth (RE) ions] were preferentially retained in the zeolite (relative to the singly ionized cations) after tests with the salt-loaded zeolite in DIW. In general, the preferential cation release order for salt-loaded zeolite A in DrW is Li > Na ≥ K > Cs > Al > Si > RE > Sr > Ba. Results from PCT-Bs with the salt-loaded zeolite A immersed in high-ionic-strength brines at 90°C indicate a significant increase, relative to DIW tests, in the release rates of the Sr, Ba, and RE ions despite a decrease in the release of the Si and Al ions that make up the framework matrix of the zeolite. An increase in the Mg and Ca concentrations in the reacted zeolites suggests that an ion exchange process may be responsible for this increase.Vapor hydration and MCC-1 tests were performed with ceramic waste form monoliths of glass-bonded zeolite. The VHTs (temperatures at 120,150, and 200°C) provided useful information about the effect of glass composition on corrosion rates and alteration phase formation, and about the overall toughness and structural integrity of the ceramic waste form. The MCC-1 test was investigated as an alternative to the PCT for acceptance criteria measurements. The MCC-1 results indicate that corrosion testing with both DIW and high-ionic-strength leachants (that specifically affect the ion exchange behavior of the fission products) are required to fully assess the durability of the ceramic waste form. These preliminary results establish the utility of the MCC-1 test for providing possible acceptance criteria measurements, including elemental release comparisons between the environmental assessment benchmark and the ceramic waste form.

scholarly journals Effect of Different Glass and Zeolite-a Compositions on the Leach Resistance of Ceramic Waste Forms

1996 ◽  
Vol 465 ◽  
Author(s):  
M. A. Lewis ◽  
M. Hash ◽  
D. Glandorf
Keyword(s):  
Spent Nuclear Fuel ◽  
National Laboratory ◽  
Hot Isostatic Pressing ◽  
Argonne National Laboratory ◽  
Fission Products ◽  
Waste Form ◽  
Zeolite A ◽  
Ceramic Waste ◽  
Leach Test ◽  
Salt Phase

ABSTRACTA ceramic waste form is being developed at Argonne National Laboratory for waste generated during the electrometallurgical treatment of spent nuclear fuel. The waste is generated when fission products are removed from the electrolyte, LiCI-KCl eutectic. The ceramic waste form is a composite, fabricated by hot isostatic pressing a mixture of glass frit and zeolite occluded with fission products and salt. Past work has shown that the normalized release rate (NRR) is less than 1 g/m2d for all elements in a Material Characterization Center-Type 1 (MCC-1) leach test run for 28 days in deionized water at 90°C (363 K). This leach resistance is comparable to that of early Savannah River glasses. We are investigating how leach resistance is affected by changes in the cationic form of zeolite and in the glass composition. Composites were made with three forms of zeolite A and six glasses. We used three-day ASTM C1220–92 (formerly MCC-1) leach tests to screen samples for development purposes only. The leach test results show that the glass composites of zeolites 5A and 4A retain fission products equally well. The loss of cesium is small, varying from 0.1 to 0.5 wt%, while the loss of divalent and trivalent fission products is one or more orders of magnitude smaller. Composites of 5A retain chloride ion better in these short-term screens than 4A and 3A. The more leach resistant composites were made with durable glasses that were rich in silica and poor in alkaline earth oxides. The x-ray diffraction (XRD) results show that a salt phase was absent in the leach resistant composites of 5A and the better glasses but was present in the other composites with poorer leach performance. Thus, the data show that the absence of a salt phase in a composite's XRD pattern corresponds to improved leach resistance. The data also suggest that the interactions between the zeolite and glass depend on the composition of both.

Release of Radiotoxic Elements from High Burn-Up UO2 and MOX Fuel in a Repository

2000 ◽  
Vol 663 ◽  
Author(s):  
Jean-Paul Glatz ◽  
Paul Carbol ◽  
Joaquin Cobos-Sabaté ◽  
Thomas Gouder ◽  
Frédéric Miserque ◽  
...  
Keyword(s):  
Spent Nuclear Fuel ◽  
Electrochemical Techniques ◽  
Dynamic Test ◽  
Fission Products ◽  
Storage Conditions ◽  
Dissolution Mechanism ◽  
Minor Actinides ◽  
Spent Fuel ◽  
Release Process ◽  
Mox Fuel

ABSTRACTIn a spent fuel repository the processes that govern the release of radionuclides are dissolution and transport in a possible groundwater flow. The cladding will be the last barrier before the water comes into contact with the fuel, namely with the outer rim of the pellet. Here the heterogeneity of the material due to the irradiation process is responsible for a complex release process. Fission products and minor actinides inventories are considerably higher at the pellet periphery as a result of increased epithermal neutron capture and of migration in the case of the volatile fission products.The present paper gives a review of experimental activities at the Institute for Transuranium Elements (ITU). Both single effects studies and integral tests are carried out to study the behavior of spent fuel under storage conditions.Leaching of irradiated UO2 (up to 50 GWd/tU) and MOX (up to 25 GWd/tU) fuel rods with preset cladding defects at 100°C under anoxic or reducing conditions should simulate the realistic case of groundwater coming into contact with a spent nuclear fuel repository. For all main radionuclides the release process can be described considering a two-step dissolution mechanism that includes the initial dissolution of an oxidized layer present on the fuel surface followed by a long-term oxidative matrix dissolution. By means of α-doped (238Pu) UO2 it could be demonstrated, that radiolysis has a significant influence on this dissolution. Especially high initial release rates were found for the volatile cesium and iodine for the reasons mentioned above.Besides the conventional leaching experiments electrochemical techniques are used to investigate for instance the complex corrosion behavior of the heterogeneous MOX fuel materials or the influence of α-radiolysis on spent fuel dissolution.In the integral tests mentioned above with large S/V values, reprecipitation of U is likely to happen. Therefore special dynamic test are conducted where this reprecipitation is prohibited and true U solubility can be determined.Thin layer of UO2 and (U,Pu)O2 doped with various fission products and minor actinides are prepared to study the influence of these elements on the matrix dissolution. When Cs is for instance co-deposited, the U oxidation state changes from U4+ to U6+ for the same O2 pressure possibly indicating a stable Cs uranate. This could be an indirect proof of the existence of such a species in irradiated fuel (e.g. at the grain boundaries).

Materials Characteristics and Dissolution Behavior of Spent Nuclear Fuel

MRS Bulletin ◽  
1994 ◽  
Vol 19 (12) ◽  
pp. 24-27 ◽  
Author(s):  
L.H. Johnson ◽  
L.O. Werme
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Waste ◽  
Nuclear Power ◽  
Spent Nuclear Fuel ◽  
Fission Products ◽  
Waste Form ◽  
Dissolution Behavior ◽  
Spent Fuel ◽  
Power Stations ◽  
Geologic Disposal

The geologic disposal of spent nuclear fuel is currently under consideration in many countries. Most of this fuel is in the form of assemblies of zirconium-alloy-clad rods containing enriched (1–4% 235U) or natural (0.71% 235U) uranium oxide pellets. Approximately 135,000 Mg are presently in temporary storage facilities throughout the world in nations with commercial nuclear power stations.Safe geologic disposal of nuclear waste could be achieved using a combination of a natural barrier (the host rock of the repository) and engineered barriers, which would include a low-solubility waste form, long-lived containers, and clay- and cement-based barriers surrounding the waste containers and sealing the excavations.A requirement in evaluating the safety of disposal of nuclear waste is a knowledge of the kinetics and mechanism of dissolution of the waste form in groundwater and the solubility of the waste form constituents. In the case of spent nuclear fuel, this means developing an understanding of fuel microstructure, its impact on release of contained fission products, and the dissolution behavior of spent fuel and of UO2, the principal constituent of the fuel.

scholarly journals Densification of Salt-Occluded Zeolite a Powders to a Leach-Resistant Monolith

1993 ◽  
Vol 333 ◽  
Author(s):  
Michele A. Lewis ◽  
Donald F. Fischer ◽  
Christopher D. Murphy
Keyword(s):  
Fast Reactor ◽  
Fission Products ◽  
Glass Frit ◽  
Deionized Water ◽  
Zeolite A ◽  
Spent Fuel ◽  
Release Rates ◽  
Geologic Disposal ◽  
Radiation Resistant ◽  
Salt Waste

ABSTRACTPyrochemical processing of spent fuel from the Integral Fast Reactor (IFR) yields a salt waste of LiCI-KCI that contains approximately 6 wt% fission products, primarily as CsCI and SrCl2. Past work has shown that zeolite A will preferentially sorb cesium and strontium and will encapsulate the salt waste in a leach-resistant, radiation-resistant aluminosilicate matrix. However, a method is still needed to convert the salt-occluded zeolite powders into a monolith suitable for geologic disposal. We are thus investigating a method that forms bonded zeolite by hot pressing a mixture of glass frit and sait-occluded zeolite powders at 990 K (717°C) and 28 MPa. The leach resistance of the bonded zeolite was measured in static leach tests run for 28 days in 363 K (90°C) deionized water. Normalized release rates of all elements in the bonded zeolite were low, <1 g/m2d. Thus, the bonded zeolite may be a suitable waste form for IFR salt waste.

Conformational effect on the preferential binding of alkali metal cation with crown ether: A molecular level investigation

2009 ◽  
Vol 12 (1-3) ◽  
pp. 93-99 ◽  
Author(s):  
Sulagna De ◽  
Sk. Musharaf Ali ◽  
M.R.K. Shenoi ◽  
Sandip K. Ghosh ◽  
Dilip K. Maity
Keyword(s):  
Crown Ether ◽  
Alkali Metal ◽  
Metal Cation ◽  
Molecular Level ◽  
Alkali Metal Cation ◽  
Conformational Effect ◽  
Preferential Binding
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