Materials Interactions Test Method to Measure Radionuclide Release from Waste Forms Under Repository-Relevant Conditions

1984 ◽  
Vol 44 ◽  
Author(s):  
Richard G. Strickert ◽  
Robert L. Erikson ◽  
John W. Shade
Keyword(s):  
Experimental Testing ◽  
Reliable Data ◽  
Test Method ◽  
Materials Characterization ◽  
Waste Form ◽  
Test Material ◽  
Borosilicate Glasses ◽  
Waste Forms ◽  
Test Conditions ◽  
Radionuclide Release

AbstractAt the request of the Basalt Waste Isolation Project, the Materials Characterization Center has collected and developed a set of procedures for a waste form compliance test method (MCC-14.4). The purpose of the test is to measure the steady-state concentrations of specified radionuclides in solutions contacting a waste form material. The test method uses a crushed waste form and basalt material suspended in a synthetic basalt groundwater and agitated for up to three months at 150°C under anoxic conditions. Elemental and radioisotopic analyses are made on filtered and unfiltered aliquots of the solution. Replicate experiments are performed and simultaneous tests are conducted with an approved test material (ATM) to help ensure precise and reliable data for the actual waste form material. Various features of the test method, equipment, and test conditions are reviewed. Experimental testing using actinide-doped borosilicate glasses are also discussed.

scholarly journals Investigating the Durability of Iodine Waste Forms in Dilute Conditions

Materials ◽  
2019 ◽  
Vol 12 (5) ◽  
pp. 686 ◽  
Author(s):  
R. Asmussen ◽  
Joseph Ryan ◽  
Josef Matyas ◽  
Jarrod Crum ◽  
Joelle Reiser ◽  
...  
Keyword(s):  
Test Method ◽  
Waste Form ◽  
Corrosion Mechanism ◽  
Secondary Phases ◽  
Dissolution Rates ◽  
Used Nuclear Fuel ◽  
Waste Forms ◽  
The Matrix ◽  
Alkaline Conditions ◽  
Spark Plasma

To prevent the release of radioiodine during the reprocessing of used nuclear fuel or in the management of other wastes, many technologies have been developed for iodine capture. The capture is only part of the challenge as a durable waste form is required to ensure safe disposal of the radioiodine. This work presents the first durability studies in dilute conditions of two AgI-containing waste forms: hot-isostatically pressed silver mordenite (AgZ) and spark plasma sintered silver-functionalized silica aerogel (SFA) iodine waste forms (IWF). Using the single-pass flow-through (SPFT) test method, the dissolution rates respective to Si, Al, Ag and I were measured for variants of the IWFs. By combining solution and solid analysis information on the corrosion mechanism neutral-to-alkaline conditions was elucidated. The AgZ samples were observed to have corrosion preferentially occur at secondary phases with higher Al and alkali content. These phases contained a lower proportion of I compared with the matrix. The SFA samples experienced a higher extent of corrosion at Si-rich particles, but an increased addition of Si to the waste led to an improvement in corrosion resistance. The dissolution rates for the IWF types are of similar magnitude to other Si-based waste form materials measured using SPFT.

Synroc tailored waste forms for actinide immobilization

2017 ◽  
Vol 105 (11) ◽  
Author(s):  
Daniel J. Gregg ◽  
Eric R. Vance
Keyword(s):  
Glass Ceramic ◽  
Waste Treatment ◽  
Waste Form ◽  
Borosilicate Glasses ◽  
Process Technology ◽  
Treatment Technology ◽  
Ceramic Waste ◽  
Waste Forms ◽  
Glass Matrices ◽  
Key Aspects

AbstractSince the end of the 1970s, Synroc at the Australian Nuclear Science and Technology Organisation (ANSTO) has evolved from a focus on titanate ceramics directed at PUREX waste to a platform waste treatment technology to fabricate tailored glass–ceramic and ceramic waste forms for different types of actinide, high- and intermediate level wastes. The particular emphasis for Synroc is on wastes which are problematic for glass matrices or existing vitrification process technologies. In particular, nuclear wastes containing actinides, notably plutonium, pose a unique set of requirements for a waste form, which Synroc ceramic and glass-ceramic waste forms can be tailored to meet. Key aspects to waste form design include maximising the waste loading, producing a chemically durable product, maintaining flexibility to accommodate waste variations, a proliferation resistance to prevent theft and diversion, and appropriate process technology to produce waste forms that meet requirements for actinide waste streams. Synroc waste forms incorporate the actinides within mineral phases, producing products which are much more durable in water than baseline borosilicate glasses. Further, Synroc waste forms can incorporate neutron absorbers and

Application of the TSPA Glass Degradation Model to Non-Conforming Waste Forms

2004 ◽  
Vol 824 ◽  
Author(s):  
W. L. Ebert
Keyword(s):  
Waste Form ◽  
Total System ◽  
Borosilicate Glasses ◽  
Degradation Mechanisms ◽  
Release Rates ◽  
Degradation Model ◽  
Waste Forms ◽  
Glass Model ◽  
High Level ◽  
Fractional Release

AbstractAn approach is presented for determining if the models used to calculate the release of radionuclides from defense high-level radioactive waste (HLW) glass for total system performance assessment (TSPA) calculations can be used to account for the release of radionuclides from waste forms other than standard borosilicate glasses. The fractional release rates of radionuclides due to waste form degradation, the available surface area, and the radionuclide inventory in an alternative waste form can be compared with the corresponding models used in TSPA for HLW glasses to determine if those models adequately represent the waste form. This approach is demonstrated for the ceramic and metallic waste forms developed for electrometallurgically treated spent sodium-bonded nuclear fuel. Depending on the waste form, comparisons made with aspects of the HLW glass model may be based on similarities in degradation mechanisms or purely empirical.

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.

Residual Stress Measurements in an Autofrettage Tube Using Hole Drilling Method

2012 ◽  
Vol 134 (5) ◽  
Author(s):  
Hamid Jahed ◽  
Mohammad Reza Faritus ◽  
Zeinab Jahed
Keyword(s):  
Residual Stress ◽  
Residual Stresses ◽  
Hoop Stress ◽  
Test Method ◽  
Test Material ◽  
Hole Drilling ◽  
Hole Drilling Method ◽  
Ring Sample ◽  
Drilling Method ◽  
American Society

Relieved strains due to drilling hole in a ring sample cut from an autofrettage cylinder are measured. Measured strains are then transformed to residual stresses using calibration constants and mathematical relations of elasticity based on ASTM standard recommendations (American Society for Testing and Materials, ASTM E 837-08, 2008, “Standard Test Method for Determining Residual Stresses by the Hole-Drilling Strain-Gage Method,” American Society for Testing and Materials). The hydraulic autofrettage is pressurizing a closed-end long cylinder beyond its elastic limits and subsequently removing the pressure. In contrast to three-dimensional stress state in the autofrettage tube, the stress measurement in hole drilling method is performed on a traction free surface formed from cutting the ring sample. The process of cutting the ring sample from a long autofrettaged tube is simulated using finite element method (FEM) and the redistribution of the residual stress due to the cut is discussed. Hence, transformation of the hole drilling measurements on the ring slice to the autofrettage residual stresses is revealed. The residual stresses are also predicted by variable material properties (VMP) method (Jahed, H., and Dubey, R. N., 1997, “An Axisymmetric Method of Elastic-Plastic Analysis Capable of Predicting Residual Stress Field,” Trans. ASME J. Pressure Vessel Technol., 119, pp. 264–273) using real loading and unloading behavior of the test material. Prediction results for residual hoop stress agree very well with the measurements. However, radial stress predictions are less than measured values particularly in the middle of the ring. To remove the discrepancy in radial residual stresses, the measured residual hoop stress that shows a self-balanced distribution was taken as the basis for calculating residual radial stresses using field equations of elasticity. The obtained residual stresses were improved a lot and were in good agreement with the VMP solution.

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 Activity of Ceftaroline against Aerobic Gram-Positive and Gram-Negative Pathogens: Effect of Test Method Variability

2011 ◽  
Vol 2011 ◽  
pp. 1-5
Author(s):  
Diane M. Citron ◽  
Yumi A. Warren ◽  
Kerin L. Tyrrell ◽  
Ellie J. C. Goldstein
Keyword(s):  
Bactericidal Activity ◽  
Test Method ◽  
In Vitro Test ◽  
Gram Negative ◽  
Methicillin Resistant ◽  
E Coli ◽  
Minimum Inhibitory Concentrations ◽  
Test Conditions ◽  
Vitro Test

Ceftaroline is a new cephalosporin with bactericidal activity against methicillin-resistant S. aureus (MRSA) as well as gram-negative pathogens. Variations of in vitro test conditions were found to affect ceftaroline activity, with 5% NaCl inhibiting growth and/or reducing the minimum inhibitory concentrations (MICs) for E. coli, K. pneumoniae, M. catarrhalis, H. influenzae, and streptococci, while an inoculum of 106 CFU/mL raised MICs of some E. coli, K. pneumoniae, and M. catarrhalis strains.

Validation of a Nuclear Waste Repository Performance Assessment Model:Comparison of Theory With Experiment

1987 ◽  
Vol 112 ◽  
Author(s):  
R. W. Geldart ◽  
B. P. Mcgrail ◽  
K. C. Rhoads ◽  
M. J. Apted
Keyword(s):  
Mass Transfer ◽  
Dissolution Kinetics ◽  
Waste Form ◽  
Nuclear Waste Repository ◽  
Transfer Model ◽  
Mass Transfer Model ◽  
Buffer Solution ◽  
Simulated Waste ◽  
Limited Mass ◽  
Radionuclide Release

AbstractLaboratory tests were conducted to validate a radionuclide mass transfer model. During the experiments, cesium and uranium releases from a simulated waste form embedded in an isotropic medium of quartz sand were measured. A 0.01 M NaH2PO4 buffer solution flowed past the waste form. Downstream concentrations obtained from computer simulations were compared with experimentally measured concentrations of cesium and uranium. Uranium release was found to be controlled by solubility-limited mass transfer, while cesium release was controlled by waste form dissolution kinetics. It was also found that the effects of dissolution on groundwater chemistry must be coupled with solubility-limited mass transfer models to defensibly predict radionuclide release rates under realistic repository conditions.

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.

Sign in / Sign up

Export Citation Format

Share Document