Comparison of the Leachability of Three TRU Cement Waste Forms

1982 ◽  
Vol 15 ◽  
Author(s):  
Wayne Ross ◽  
Joseph Westsik ◽  
Frank Roberts ◽  
Clark Harvey
Keyword(s):  
Elevated Temperatures ◽  
Waste Form ◽  
Cold Pressing ◽  
Waste Forms ◽  
Leach Test

ABSTRACTCement waste forms prepared by three processes, casting, cold pressing, and FUETAP (Formed Under Elevated Temperatures and Pressure) have been compared for their leachability by using the MCC-1 leach test. The results indicate that releases of plutonium are not controlled by the waste form matrix and that there is no significant overall advantage to any of the three cement processes from a leachability viewpoint.

Standardized Waste Form Test Methods

1984 ◽  
Vol 44 ◽  
Author(s):  
Steven C. Slate
Keyword(s):  
Radiation Stability ◽  
Test Methods ◽  
Test Method ◽  
Waste Form ◽  
Current Status ◽  
High Level Waste ◽  
Gas Generation ◽  
Waste Forms ◽  
Leach Test ◽  
Compliance Testing

AbstractThe Materials Characterization Center (MCC) is developing standard tests to characterize nuclear waste forms. Development of the first thirteen tests was originally initiated to provide data to compare different high-level waste (HLW) forms and to characterize their basic performance. The current status of the first thirteen MCC tests and some sample test results are presented: The radiation stability tests (MCC-6 and 12) and the tensile-strength test (MCC-11) are approved; the static leach tests (MCC-1, 2, and 3) are being reviewed for full approval; the thermal stability (MCC-7) and microstructure evaluation (MCC-13) methods are being considered for the first time; and the flowing leach test methods (MCC-4 and 5), the gas generation methods (MCC-8 and 9), and the brittle fracture method (MCC-10) are indefinitely delayed. Sample static leach test data on the ARM-I approved reference material are presented.Established tests and proposed new tests will be used to meet new testing needs. For waste form production, tests on stability and composition measurement are needed to provide data to ensure waste form quality. In transportaion, data are needed to evaluate the effects of accidents on canisterized waste forms. The new MCC-15 accident test method and some data are presented. Compliance testing needs required by the recent draft repository waste acceptance specifications are described. These specifications will control waste form contents, processing, and performance.

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.

MCC-1: A Standard Leach Test for Nuclear Waste Forms

1982 ◽  
Vol 56 (2) ◽  
pp. 306-312 ◽  
Author(s):  
D.M. Strachan ◽  
R. P. Turcotte ◽  
B. O. Barnes
Keyword(s):  
Nuclear Waste ◽  
Waste Forms ◽  
Leach Test ◽  
Nuclear Waste Forms

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.

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.

scholarly journals The Effects of Temperature on the Leaching Behavior of Cement Waste Forms - The Cement/Sodium Sulfate System

1989 ◽  
Vol 176 ◽  
Author(s):  
Mark Fuhrmann ◽  
Richard Pietrzak ◽  
John Heiser ◽  
Eena-Mai Franz ◽  
Peter Colombo
Keyword(s):  
Sodium Sulfate ◽  
Elevated Temperatures ◽  
Effective Diffusion ◽  
Finite Cylinder ◽  
Kcal Mole ◽  
Experimental Conditions ◽  
Waste Forms ◽  
Diffusion Controlled ◽  
Increasing Temperature ◽  
Relationship Of

ABSTRACTThe leaching mechanisms of simulated low-level radioactive waste forms are being determined as support for development of an accelerated leach test. Two approaches are being used: (1) comparison of leaching data with results of a model that describes diffusion from a finite cylinder, and (2) observation of the leaching process at temperatures between 20°C and 65°C. To provide results that can be used for modeling, leaching at elevated temperatures must change neither the leaching mechanism nor the structural controls of leaching such as the porosity. Releases of 137Cs, 85Sr, calcium, sodium and potassium from portland cement containing sodium sulfate, as a simulated evaporator sludge, have been determined under a variety of experimental conditions. Data from the leach tests were compared to model results for diffusion from the finite cylinder. While most leaching appears to be diffusion controlled, notable exceptions occur. For all samples, activation energies ranging between 6 and 11 Kcal/mole have been calculated from the relationship of the effective diffusion coefficient to increasing temperature, close to the expected value of 5 Kcal/mole for diffusion.

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.

scholarly journals Comparison of Ceramic Waste forms Produced by hot Uniaxial Pressing and by Cold Pressing and Sintering

1994 ◽  
Vol 353 ◽  
Author(s):  
V. M. Oversby ◽  
E.R. Vance
Keyword(s):  
Production Method ◽  
Activity Level ◽  
Waste Form ◽  
Phase Assemblage ◽  
Waste Streams ◽  
Cold Pressing ◽  
N2 Atmosphere ◽  
Precursor Material ◽  
Simulated Waste ◽  
Cold Pressed

AbstractSynroc C waste form specimens prepared using the Australian-developed technology are uniaxially pressed in stainless steel bellows at 1200°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°C for 1 hr. in a 3.5% H2 in N2 atmosphere. Then one portion of each sample was hot pressed at temperatures ranging from 1120 to 1250°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°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 held at 200°C for long enough to allow trapped gases to escape.

Comparison of 200 Liter and 40 Milliliter Leach Tests

1982 ◽  
Vol 15 ◽  
Author(s):  
J. H. Westsik ◽  
C. O. Harvey ◽  
F. P. Roberts ◽  
W. A. Ross ◽  
R. E. Thornhill
Keyword(s):  
Large Scale ◽  
Deionized Water ◽  
Waste Form ◽  
Small Scale ◽  
Matrix Elements ◽  
Elemental Concentrations ◽  
The Past ◽  
Leach Test ◽  
The Matrix ◽  
Scale Test

ABSTRACTDuring the past year we have conducted a modified MCC-1 leach test on a 145 kg block of a cast cement waste form. The leach vessel was a 200 liter Teflon®-lined drum and contained 97.5 liters of deionized water. The results of this large-scale leach test were compared with the results of standard MCC-1 tests (40 ml) on smaller samples of the same waste form. The ratio of leachate volumes between the large and small scale tests was 2500 and the ratio of sample masses was 150,000. The cast cement samples for both tests contained plutonium-doped incinerator ash.The leachates from these tests were analyzed for both plutonium and the matrix elements. Evaluation of plutonium plateout in the large-scale test indicated that the majority of the plutonium leached from the samples deposits onto vessel walls and little (<3 × 10−12M) remains in solution. Comparison of elemental concentrations in the leachates indicates some differences up to 5X in the concentration in the large- and small-scale tests. The differences are attributed to differences in the solubilities of Ca, Si, and Fe at pH ˜11.5 and at pH ˜12.5. The higher pH observed for the large-scale test is a result of the larger quantities of sodium in the large block of cement.

Sign in / Sign up

Export Citation Format

Share Document