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.

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.

Interim Waste Forms for High-Level Radioactive Wastes: Processing and Properties

1981 ◽  
Vol 6 ◽  
Author(s):  
G. Bandyopadhyay
Keyword(s):  
Sodium Silicate ◽  
Water Solubility ◽  
Mechanical Stability ◽  
Solid Material ◽  
High Water ◽  
Radioactive Wastes ◽  
Waste Form ◽  
Waste Forms ◽  
High Level ◽  
Fused Salt

ABSTRACTSeveral simulated interim waste forms have been investigated in the laboratory to study their suitability for application in handling and transportation of high-level radioactive wastes to terminal processing sites. In the fused-salt/sludge option, the neutralized supernatant liquid and the precipitated sludge are treated simultaneously to form fused-salt cakes. Silicate-based options, in which sodium silicate or sodium silicate and Ca(OH)2 act as binders for the sludge, require prior separation of the sludge and the soluble radioactive constituents from the supernatant before the waste form can be prepared. The results from tests on simulated fused-salt waste forms indicated that the process simplicity of this option is partially offset by the high water solubility and hygroscopicity of the product, which would necessitate special precautions during transportation and storage. The most promising silicate-based option is the ambienttemperature silicate sludge process, in which the sludge is mixed with sodium silicate [and sometimes with Ca(OH)2] and subsequently exposed to a contrelled-humidity environment at room temperature to form a chemical bond. Solid material containing 75 wt % synthetic calcined sludge, prepared by this process, has sufficient physical, chemical, and mechanical stability for use as an interim waste form.

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 Incorporation of High-Level Wastes in Synroc: Results from Recent Process Engineering Studies at Lawrence Livermore National Laboratory

1981 ◽  
Vol 11 ◽  
Author(s):  
J. H. Campbell ◽  
C. L. Hoenig ◽  
F. J. Ackerman ◽  
P. E. Peters ◽  
J. Z. Grens
Keyword(s):  
Borosilicate Glass ◽  
Selection Process ◽  
Lawrence Livermore National Laboratory ◽  
National Laboratory ◽  
Process Engineering ◽  
Waste Form ◽  
Performance Properties ◽  
Engineering Studies ◽  
Waste Forms ◽  
High Level

In October 1981 SYNROC-D was selected as the reference alternate waste form to borosilicate glass for immobilization of defense wastes. A total of eight candidate waste forms competed in this selection process and the decision of which alternate waste form to choose was based primarily on performance properties.

The Strategy of the Belgian Nuclear Research Centre in the Area of High-Level Waste Form Compatibility Research

2007 ◽  
Author(s):  
Karel Lemmens ◽  
Christelle Cachoir ◽  
Elie Valcke ◽  
Karine Ferrand ◽  
Marc Aertsens ◽  
...  
Keyword(s):  
Nuclear Research ◽  
Waste Form ◽  
Research Centre ◽  
High Level Waste ◽  
Spent Fuel ◽  
National Agency ◽  
Waste Forms ◽  
In Situ Experiments ◽  
High Level

The Belgian Nuclear Research Centre (SCK•CEN) has a long-standing expertise in research concerning the compatibility of waste forms with the final disposal environment. For high level waste, most attention goes to two waste forms that are relevant for Belgium, namely (1) vitrified waste from the reprocessing of spent fuel, and (2) spent fuel as such, referring to the direct disposal scenario. The expertise lies especially in the study of the chemical interactions between the waste forms and the disposal environment. This is done by laboratory experiments, supported by modeling. The experiments vary from traditional leach tests, to more specific tests for the determination of particular parameters, and highly realistic experiments. This results in a description of the phenomena that are expected upon disposal of the waste forms, and in quantitative data that allow a conservative long-term prediction of the in situ life time of the waste form. The predictions are validated by in situ experiments in the underground research laboratory HADES. The final objective of these studies, is to estimate the contribution of the waste form to the overall safety of the disposal system, as part of the Safety and Feasibility Case, planned by the national agency ONDRAF/NIRAS. The recent change of the Belgian disposal concept from an engineered barrier system based on the use of bentonite clay to a system based on a concrete buffer has caused a reorientation of the research programme. The expertise in the area of clay-waste interaction will however be maintained, to develop experimental methodologies in collaboration with other countries, and as a potential support to the decision making in those countries where a clay based near field is still the reference. The paper explains the current R&D approach, and highlights some recent experimental set-ups available at SCK•CEN for this purpose, with some illustrating results.

Effects of Iron and pH on Glass Dissolution Rate

2002 ◽  
Vol 757 ◽  
Author(s):  
Seung-Young Jeong ◽  
William L. Ebert
Keyword(s):  
Dissolution Rate ◽  
Ph Dependence ◽  
High Level Waste ◽  
Total System ◽  
Degradation Model ◽  
Static Tests ◽  
Dissolved Iron ◽  
Glass Dissolution ◽  
License Application ◽  
High Level

ABSTRACTShort-term static tests were conducted with a surrogate high-level waste glass to measure the effects of pH and dissolved iron on the glass dissolution rate. The tests were conducted to determine if a term to account for the effects of dissolved iron is needed in the glass degradation model developed for Total System Performance Assessment (TSPA) calculations for the Yucca Mountain disposal system license application. The glass degradation model includes terms for dependencies on temperature, pH, and chemical affinity. A series of tests was conducted at 90 °C in various pH solutions without iron and with added FeCl3, Fe2O3, Fe3O4, and FeOOH. Tests were conducted at glass surface area/solution volume (S/V) ratios about 2 and 10 m-1for between 2 and 21 days. Solution concentrations of boron were used to measure the extent of reaction and calculate the glass dissolution rates. Similar rates were measured in tests conducted with and without iron at each pH. Both the results of the tests with and without iron showed V-shaped pH dependence curves with minima at near-neutral pH values. The pH dependencies (η) are about 0.44 in basic solutions and –0.49 in acidic solutions, based on the combined results of tests with and without iron. These are within the range of values for the pH dependence in the TSPA model for site recommendation.

scholarly journals Corrosion of Glass-Bonded Sodalite and its Components as a Function of ph and Temperature

2002 ◽  
Vol 713 ◽  
Author(s):  
Seung-Young Jeong ◽  
Lester R. Morss ◽  
William L. Ebert
Keyword(s):  
Dissolution Rate ◽  
Hot Isostatic Pressing ◽  
Waste Form ◽  
High Level Waste ◽  
Model Parameters ◽  
Dissolution Rates ◽  
Degradation Model ◽  
Modified Binder ◽  
Ph Range ◽  
High Level

ABSTRACTA glass-bonded sodalite ceramic waste form (CWF) has been developed to immobilize electrorefiner salt wastes from electrometallurgical treatment of spent sodium-bonded reactor fuel for disposal. A degradation model is being developed to support qualification of the CWF for disposal in the federal high-level waste disposal system. The parameter values in the waste form degradation model were previously determined from the dissolution rates measured in MCC-1 tests conducted at 40, 70, and 90°C. The results of several series of tests that were conducted to confirm the applicability of the dissolution rate model and model parameters are presented in this paper: (1) Series of MCC-1 tests were conducted in five dilute buffer solutions in the pH range of 4.8 – 9.8 at 20°C with hot isostatic pressing (HIP) sodalite, HIP glass, and HIP CWF. The results show that the model adequately predicts the dissolution rate of these materials at 20°C. (2) Tests at 20 and 70°C with CWF made by pressureless-consolidation (PC) indicate that the model parameters extracted from the results of tests with HIP CWF can be applied to PC CWF. (3) The dissolution rates of a glass made with a composition corresponding to 80 wt. % glass and 20 wt. % sodalite were measured at 70°C to evaluate the sensitivity of the rate to the composition of binder glass in the CWF. The dissolution rates of the modified binder glass were indistinguishable from the rates of the binder glass.

Corrosion Testing of Stainless Steel-Zirconium:Metal Waste Forms

1999 ◽  
Vol 556 ◽  
Author(s):  
D. P. Abraham ◽  
L. J. Simpson ◽  
M. J. Devries ◽  
S. M. Mcdeavitt
Keyword(s):  
Stainless Steel ◽  
Spent Nuclear Fuel ◽  
Fission Products ◽  
Waste Form ◽  
Waste Forms ◽  
Immersion Corrosion ◽  
Corrosion Studies ◽  
High Level Nuclear Waste ◽  
High Level ◽  
Nuclear Waste Forms

AbstractStainless steel-zirconium (SS-Zr) alloys have been developed as waste forms for the disposal of metallic waste generated during the electrometallurgical treatment of spent nuclear fuel. The waste forms incorporate irradiated cladding hulls, components of the alloy fuel, noble metal fission products, and actinide elements. The baseline waste form is a stainless steel- 15 wt% zirconium (SS-15Zr) alloy. This article presents microstructures and some of the corrosion studies being conducted on the waste form alloys. Electrochemical corrosio n, immersion corrosion, and vapor hydration tests have been performed on various alloy compositions to evaluate corrosion behavior and resistance to selective leaching of simulated fission products. The SS-Zr waste forms immobilize and retain fission products very effectively and show potential for acceptance as high-level nuclear waste forms.

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

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