Leaching Scale Effect for Cement-Waste Forms

1988 ◽  
Vol 127 ◽  
Author(s):  
J. C. Nomine ◽  
A. Billon ◽  
G. Courtois
Keyword(s):  
Scale Effect ◽  
Practical Method ◽  
Full Scale ◽  
Time Limit ◽  
Waste Form ◽  
Leaching Tests ◽  
Local Policy ◽  
Waste Package ◽  
Waste Forms

The confinement ability of a waste package is one of the major safety characteristics to consider in shallow land burial. In order to determine if the confinement is acceptable, in accordance with local policy, one way is to proceed to leaching tests. The practical method, for sake of simplicity, cost and time limit, is to carry out the leaching tests on laboratory samples which are easier to prepare than full-scale blocks, but the representativity of which needs to be treated with caution; it is in this context, that one of the aspect of our work concerns what is known as the “scale effect”.This study has been conducted using blocks the volumes of which are respectively of 200, 20, 2 and 0, 2 1, and made with the same cement-waste form (13 Cs) system.

Immobilization of rocky flats graphite fines residues

1999 ◽  
Vol 556 ◽  
Author(s):  
T. S. Rudisill ◽  
J. C. Marra ◽  
D. K. Peeler
Keyword(s):  
Scale Up ◽  
Glass Frit ◽  
Heating Time ◽  
Full Scale ◽  
Waste Form ◽  
Small Scale ◽  
Rocky Flats ◽  
Waste Forms ◽  
Average Residue ◽  
The Impact

AbstractThe Savannah River Technology Center (SRTC) is developing an immobilization process for graphite fines residues generated during nuclear materials production activities at the Rocky Flats Environmental Technology Site (Rocky Flats). The continued storage of this material has been identified as an item of concern. The residue was generated during the cleaning of graphite casting molds and potentially contains reactive plutonium metal. The average residue composition is 73 wt% graphite, 15 wt% calcium fluoride (CaF2), and 12 wt% plutonium oxide (PuO2 ). Approximately 950 kg of this material are currently stored at Rocky Flats.The strategy of the immobilization process is to microencapsulate the residue by mixing with a sodium borosilicate (NBS) glass frit and heating at nominally 700°C. The resulting waste form would be sent to the Waste Isolation Pilot Plant (WIPP) for disposal. Since the PuO2 concentration in the residue averages 12 wt%, the immobilization process was required to meet the intent of safeguards termination criteria by limiting plutonium recoverability based on a test developed by Rocky Flats. The test required a plutonium recovery of less than 4 g/kg of waste form when a sample was leached using a nitric acid/CaF2 dissolution flowsheet.Immobilization experiments were performed using simulated graphite fines with cerium oxide (CeO2) as a surrogate for PuO2 and with actual graphite fines residues. Small-scale surrogate experiments demonstrated that a 4:1 frit to residue ratio was adequate to prevent recovery of greater than 4 g/kg of cerium from simulated waste forms. Additional experiments investigated the impact of varying concentrations of CaF2 and the temperature/heating time cycle on the cerium recovery. Optimal processing conditions developed during these experiments were subsequently demonstrated at full-scale with surrogate materials and on a smaller scale using actual graphite fines.In general, the recovery of cerium from the full-scale waste forms was higher than for smaller scale experiments. The presence of CaF2 also caused a dramatic increase in cerium recovery not seen in the small-scale experiments. However, the results from experiments with actual graphite fines were encouraging. A 4:1 frit to residue ratio, a temperature of 700°C, and a 2 hr heating time produced waste forms with plutonium recoveries of 4±1 g/kg. With an increase in the frit to residue ratio, waste forms fabricated at this scale should meet the Rocky Flats product specification. The scale-up of the waste form fabrication process to nominally 3 kg is expected to require a 5:1 to 6:1 frit to residue ratio and maintaining the waste form centerline temperature at 700°C for 2 hr.

Host Phases for Actinide Elements in the Metallic Waste Form

2002 ◽  
Vol 757 ◽  
Author(s):  
D. E. Janney
Keyword(s):  
Stainless Steel ◽  
Electron Microscope ◽  
Electron Diffraction ◽  
Transmission Electron Microscope ◽  
Waste Form ◽  
Dissolution Behavior ◽  
Release Rates ◽  
Waste Forms ◽  
Transmission Electron ◽  
Simulated Waste

ABSTRACTArgonne National Laboratory has developed an electrometallurgical process for conditioning spent sodium-bonded metallic reactor fuel prior to disposal. A waste stream from this process consists of stainless steel cladding hulls that contain undissolved metal fission products such as Tc, Ru, Rh, Pd, and Ag; a small amount of undissolved actinides (U, Np, Pu) also remains with the hulls. These wastes will be immobilized in a waste form whose baseline composition is stainless steel alloyed with 15 wt% Zr (SS-15Zr). Scanning electron microscope (SEM) observations of simulated metal waste forms (SS-15Zr with up to 11 wt% actinides) show eutectic intergrowths of Fe-Zr-Cr-Ni intermetallic phases with steels. The actinide elements are almost entirely in the intermetallics, where they occur in concentrations ranging from 1–20 at%. Neutron- and electron-diffraction studies of the simulated waste forms show materials with structures similar to those of Fe2Zr and Fe23Zr6.Dissolution experiments on simulated waste forms show that normalized release rates of U, Np, and Pu differ from each other and from release rates of other elements in the sample, and that release rates for U exceed those for any other element (including Fe). This paper uses transmission electron microscope (TEM) observations and results from energy-dispersive X-ray spectroscopy (EDX) and selected-area electron-diffraction (SAED) to characterize relationships between structural and chemical data and understand possible reasons for the observed dissolution behavior.Transmission electron microscope observations of simulated waste form samples with compositions SS-15Zr-2Np, SS-15Zr-5U, SS-15Zr-11U-0.6Rh-0.3Tc-0.2Pd, and SS-15Zr-10Pu suggest that the major actinide-bearing phase in all of the samples has a structure similar to that of the C15 (cubic, MgCu2-type) polymorph of Fe2Zr, and that materials with this structure exhibit significant variability in chemical compositions. Material whose structure is similar to that of the C36 (dihexagonal, MgNi2-type) polymorph of Fe2Zr was also observed, and it exhibits less chemical variability than that displayed by material with the C15 structure. The TEM data also demonstrate a range of actinide concentrations in materials with the Fe23Zr6 (cubic, Mn23Th6-type) structure.Microstructures similar to those produced during experimental deformation of Fe-10 at% Zr alloys were observed in intermetallic materials in all of the simulated waste form samples. Stacking faults and associated dislocations are common in samples with U, but rarely observed in those with Np and Pu, while twins occurred in all samples. The observed differences in dissolution behavior between samples with different actinides may be related to increased defect-assisted dissolution in samples with U.

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.

scholarly journals A study into the scale effect on the strength properties of polymer composite materials

2019 ◽  
Vol 5 (2) ◽  
pp. 103-108
Author(s):  
Valentina V. Kiryushina ◽  
Yuliya Yu. Kovaleva ◽  
Petr A. Stepanov ◽  
Pavel V. Kovalenko
Keyword(s):  
Composite Materials ◽  
Polymer Composite ◽  
Scale Effect ◽  
Mechanical Characteristics ◽  
Full Scale ◽  
Sample Thickness ◽  
Polymer Composite Materials ◽  
Test Machine ◽  
Test Results ◽  
Laboratory Sample

Polymer composite materials (PCM) are used extensively and are viewed as candidates for application in various industries, including nuclear power. Despite a variety of methods and procedures employed to investigate the mechanical characteristics of PCMs, the use of the laboratory sample mechanical test results to design and model large-sized structures is not always fully correct and reasonable. In particular, one of the problems is concerned with taking into account the scale parameter effects on the PCM strength and elastic characteristics immediately in the product. The purpose of the study is to investigate the scale effects on the mechanical characteristics of glass reinforced plastics using phenolformaldehyde and silicon-organic binders and a fabric quartz filler. Samples of four different standard sizes under GOST 25604-82 and GOST 4648-2014 were tested for three-point bending using an LFM-100 test machine to estimate the scale effect. The thicknesses of the model samples were chosen with regard for the wall thicknesses of full-scale products under development or manufactured commercially and the test machine features, and varied in the limits of 1.6 to 7.5 mm. The tests showed that strength decreased as the sample thickness was increased to 3 mm and more both at room and elevated (200 to 500 °C) temperatures, which can be described by an exponential function based on the Weibull statistical model. The values of the Weibull modulus that characterizes the extent of the scale effect on the strength of the tested materials were 4.6 to 6.7. The average bend strength in the sample thickness range of 3 mm and less does not vary notably or tends to increase slightly as the thickness is increased. This fact makes it possible to conclude that estimation of allowable stresses in a thin-wall product requires the use of test results for samples with a thickness that is equal to the product wall thickness since standard samples may yield overestimated allowable stress values and lead, accordingly, to incorrect calculations of the strength factor. The results obtained shall be taken into account when defining the allowable levels of operation for full-scale products and structures of polymer composites based on the laboratory sample strength data as well as when estimating their robustness as a characteristic of the product’s fail-safe operation.

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.

Materials Interactions Relating to Long-Term Geologic Disposal of Nuclear Waste Glass

1986 ◽  
Vol 84 ◽  
Author(s):  
Ned E. Bibler ◽  
Carol M. Jantzen
Keyword(s):  
Nuclear Waste ◽  
Review Paper ◽  
Waste Glass ◽  
Waste Form ◽  
Glass Waste ◽  
Waste Package ◽  
Geologic Disposal ◽  
Nuclear Waste Glass ◽  
Backfill Materials

AbstractIn the geologic disposal of nuclear waste glass, the glass will eventually interact with groundwater in the repository system. Interactions can also occur between the glass and other waste package materials that are present. These include the steel canister that holds the glass, the metal overpack over the canister, backfill materials that may be used, and the repository host rock. This review paper systematizes the additional interactions that materials in the waste package will impose on the borosilicate glass waste form-groundwater interactions. The repository geologies reviewed are tuff, salt, basalt, and granite. The interactions emphasized are those appropriate to conditions expected after repository closure, e.g. oxic vs. anoxic conditions. Whenever possible, the effect of radiation from the waste form on the interactions is examined. The interactions are evaluated based on their effect on the release and speciation of various elements including radionuclides from the glass. It is noted when further tests of repository interactions are needed before long-term predictions can be made.

TEM Analysis of Corrosion Products From a Radioactive Stainless Steel-based Alloy

2000 ◽  
Vol 6 (S2) ◽  
pp. 368-369
Author(s):  
N.L. Dietz ◽  
D.D Keiser
Keyword(s):  
Stainless Steel ◽  
Spent Nuclear Fuel ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
Fission Products ◽  
Waste Form ◽  
Metallic Sodium ◽  
Geologic Repository ◽  
Tem Analysis ◽  
Waste Forms

Argonne National Laboratory has developed an electrometallurgical treatment process for metallic spent nuclear fuel from the Experimental Breeder Reactor-II. This process stabilizes metallic sodium and separates usable uranium from fission products and transuranic elements that are contained in the fuel. The fission products and other waste constituents are placed into two waste forms: a ceramic waste form that contains the transuranic elements and active fission products such as Cs, Sr, I and the rare earth elements, and a metal alloy waste form composed primarily of stainless steel (SS), from claddings hulls and reactor hardware, and ∼15 wt.% Zr (from the U-Zr and U-Pu-Zr alloy fuels). The metal waste form (MWF) also contains noble metal fission products (Tc, Nb, Ru, Rh, Te, Ag, Pd, Mo) and minor amounts of actinides. Both waste forms are intended for eventual disposal in a geologic repository.

Ion Implantation Studies of Nuclear Waste Forms

1981 ◽  
Vol 6 ◽  
Author(s):  
Clyde J. M. Northrup ◽  
George W. Arnold ◽  
Thomas J. Headley
Keyword(s):  
Nuclear Waste ◽  
Borosilicate Glass ◽  
Alpha Decay ◽  
Lead Ions ◽  
Waste Form ◽  
Elastic Recoil ◽  
Elastic Recoil Detection ◽  
Waste Forms ◽  
Basic Studies ◽  
Nuclear Waste Forms

ABSTRACTThe first observations of physical and chemical changes induced by lead implantation damage and leaching are reported for two proposed U.S. nuclear waste forms (PNL 76–68 borosilicate glass and Sandia titanate ceramics) for commercial wastes. To simulate the effects of recoil nucleii due to alpha decay, the materials were implanted with lead ions at equivalent doses up to approximately 1 × 1019 a decays/cm3 . In the titanate waste form, the zirconolite, perovskite, hollandite, and rutile phases all exhibited a mottled appearance in the transmission electron microscope (TEM) typical of defect clusters in radiation damaged, crystalline solids. One titanate phase containing uranium was found by TEM to be amorphous after implantation at the highest dose. No enhanced leaching (deionized water, room temperature, 24 hours) of the irradiated titanate waste form, including the amorphous phase, was detected by TEM, but Rutherford backscattering (RBS) suggested a loss of cesium and calcium after 21 hours of leaching. The RBS spectra also indicated enhanced leaching from the PNL 76–68 borosilicate glass after implantation with lead ions, in general agreement with the observations of Dran, et al. [6,7] on other irradiated materials. Elastic recoil detection spectroscopy (ERD), used to profile hydrogen after leaching, showed penetration of the hydrogen to several thousand angstroms for both the implanted and unimplanted materials. These basic studies identified techniques to follow the changes that occur on implantation and leaching of complex amorphous and crystalline waste forms. These studies were not designed to produce comparisons between waste forms of gross leach rates.

The Expected Environment for Waste Packages in a Salt Repository

1983 ◽  
Vol 26 ◽  
Author(s):  
L. R. Pederson ◽  
D. E. Clark ◽  
F. N. Hodges ◽  
G. L. Mcvpy ◽  
D. Rai
Keyword(s):  
Rock Salt ◽  
Near Field ◽  
Materials Testing ◽  
Site Specific ◽  
Waste Package ◽  
Barrier Materials ◽  
Waste Forms ◽  
Temperature Irradiation ◽  
Salt Brine ◽  
Package Materials

ABSTRACTThis paper discusses results of recent efforts to define the very near-field (within approximately 2m) environmental conditions to which waste packages will be exposed in a salt repository. These conditions must be considered in the experimental design for waste package materials testing, which includes corrosion of barrier materials and leaching of waste forms. Site-specific brine compositions have been determined, and “standard” brine compositions have been selected for testing purposes. Actual brine compositions will vary depending on origin, temperature, irradiation history, and contact with irradiated rock salt. Results of irradiating rock salt, synthetic brines, rock salt/brine mixtures, and reactions of irradiated rock salt with brine solutions are reported.

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