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.

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.

Interpreting Uranyl Mineral Diffraction Patterns

1998 ◽  
Vol 4 (S2) ◽  
pp. 560-561
Author(s):  
Edgar C. Buck
Keyword(s):  
Corrosion Behavior ◽  
Nuclear Waste ◽  
Equatorial Plane ◽  
Waste Form ◽  
Spent Fuel ◽  
Secondary Phases ◽  
Form Complex ◽  
Waste Forms ◽  
Diffraction Patterns ◽  
Nuclear Waste Forms

Secondary phases that form during the corrosion of nuclear waste forms may influence both the rate of waste form dissolution and the release of radionuclides [1]. The identification of these phases is critical in developing models for the corrosion behavior of nuclear waste forms. In particular, the secondary uranyl (VI) minerals that form during waste form alteration may control uranium solubility and release of radionuclides incorporated into these phases [2].The U6+ cation in uranyl minerals is almost always present as a linear (UO2)2+ ion [3]. This uranyl (Ur) ion is coordinated by four, five, or six anions (ϕ) in the equatorial plane resulting in the formation of square (Urϕ4), pentagonal (Urϕ5), and hexagonal (Urϕ6) bipyramids, respectively [3]. These bipyramid polyhedra may polymerize to form complex infinite sheet structures. The linking of Urϕ5 is observed in a number of uranyl minerals formed during waste glass and spent fuel corrosion [2,4], such as weeksite [Na,K(UO2)2(Si205)3*4H2O] and β-uranophane [Ca[(UO2)(SiO3OH)]2*5H2O].

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.

Fabrication and Chemical Durability of Ceramic Technetium-Based Pyrochlores and Perovskites as Potential Waste Forms

2014 ◽  
Vol 94 ◽  
pp. 85-92 ◽  
Author(s):  
Thomas Hartmann ◽  
Izua J. Alaniz-Ortez
Keyword(s):  
Borosilicate Glass ◽  
Standard Test ◽  
Test Method ◽  
Layered Perovskite ◽  
Synthesis Conditions ◽  
Ceramic Waste ◽  
Used Nuclear Fuel ◽  
Waste Forms ◽  
Good Crystallinity ◽  
Hydrodynamic Data

Technetium-99 is a key radioisotope from a nuclear waste perspective because of its long half-life (t1/2= 2.13x105 years) and its abundance in used nuclear fuel. As such, it is targeted in separation strategies such as UREX+, for isolation and encapsulation in solid waste forms for final disposal in a nuclear repository. We report here results regarding the incorporation of Tc-99 into ternary oxides of different structure types: pyrochlore (Nd2Tc2O7), perovskite (SrTcO3), and layered perovskite (Sr2TcO4). The objective was to determine synthesis conditions of these ceramic waste forms to immobilize Tc-99 as Tc (IV) and to harvest crystallographic, thermophysical and hydrodynamic data. The fabricated ceramic technetates exhibit good crystallinity and lattice parameters and atomic coordinates could be refined with high accuracies. Thermophysical properties of the oxides, such as the critical temperature (Tc) for superconductivity, were analyzed using AC magnetic susceptibility measurements. In our efforts to compare hydrodynamic properties of the ceramic waste forms with those to Tc-bearing borosilicate glass, we applied the standard test method ASTM C1220-10. Hereby matrix corrosion and Tc-leaching of Tc-99 containing LAWE4 type borosilicate glass and ceramic specimens were compared by static leaching experiments.

scholarly journals Influence of Alumina Nanofibers Sintered by the Spark Plasma Method on Nickel Mechanical Properties

Metals ◽  
2021 ◽  
Vol 11 (4) ◽  
pp. 548 ◽  
Author(s):  
Leonid Agureev ◽  
Valeriy Kostikov ◽  
Zhanna Eremeeva ◽  
Svetlana Savushkina ◽  
Boris Ivanov ◽  
...  
Keyword(s):  
Mechanical Properties ◽  
Tensile Strength ◽  
Elastic Modulus ◽  
Strength Properties ◽  
Alumina Nanoparticles ◽  
Metal Powders ◽  
Wide Range ◽  
The Matrix ◽  
Spark Plasma ◽  
Average Size

The article presents the study of alumina nanoparticles’ (nanofibers) concentration effect on the strength properties of pure nickel. The samples were obtained by spark plasma sintering of previously mechanically activated metal powders. The dependence of the grain size and the relative density of compacts on the number of nanofibers was investigated. It was found that with an increase in the concentration of nanofibers, the average size of the matrix particles decreased. The effects of the nanoparticle concentration (0.01–0.1 wt.%) on the elastic modulus and tensile strength were determined for materials at 25 °C, 400 °C, and 750 °C. It was shown that with an increase in the concentration of nanofibers, a 10–40% increase in the elastic modulus and ultimate tensile strength occurred. A comparison of the mechanical properties of nickel in a wide range of temperatures, obtained in this work with materials made by various technologies, is carried out. A description of nanofibers’ mechanisms of influence on the structure and mechanical properties of nickel is given. The possible impact of impurity phases on the properties of nickel is estimated. The tendency of changes in the mechanical properties of nickel, depending on the concentration of nanofibers, is shown.

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.

Effect of Precursors on Transmittance and Microstructure of Mullite Ceramics

2009 ◽  
Vol 620-622 ◽  
pp. 429-432 ◽  
Author(s):  
Gui Min Zhang ◽  
Zheng Yi Fu ◽  
Yu Cheng Wang ◽  
Hao Wang ◽  
Wei Min Wang ◽  
...  
Keyword(s):  
Spray Drying ◽  
Ethanol Solution ◽  
Spherical Particles ◽  
Plasma Sintering ◽  
The Matrix ◽  
Conventional Drying ◽  
Spark Plasma ◽  
Mean Size ◽  
Dta Curves ◽  
Equiaxed Grains

Two different kinds of mullite precursors with composition 3Al2O3•2SiO2 (3:2) were prepared by conventional drying ethanol solution and spray-drying aqueous solution of aluminum nitrate nanohydrate and tetraethoxysilane, respectively. The results of scanning electron microscope (SEM) indicate that one powder consists of irregular particles with size of 1-10μm, the other powder is made of inhomogeneously sized hollow spherical particles with mean size of 0.5-5μm. The TG-DTA curves indicate the hollow spherical particles are unfavorable to eliminate the decomposed products. After the precursors were sintered by Spark Plasma Sintering at 1450°C for 10min, the microstructures of the former are made of fine equiaxed grains with sizes of around 0.5μm, the latter consist of elongated grains distributed in the matrix of fine grains with imhomogenous size of 0.5~10μm due to the liquid phase forming. The different microstructures lead to the former sintered body is transparent, while, the sample from spray-drying is opaque.

scholarly journals The Valance Determination of Cerium Ions in α-SiAlON by Electron Energy Loss Spectroscopy Analysis

2008 ◽  
Vol 14 (S3) ◽  
pp. 19-22 ◽  
Author(s):  
H. Yurdakul ◽  
S. Turan
Keyword(s):  
Rare Earth ◽  
Ionic Radius ◽  
High Hardness ◽  
Engineering Properties ◽  
Secondary Phases ◽  
Sintering Additive ◽  
Valence States ◽  
Domain Boundaries ◽  
Spark Plasma

SiAlON ceramics have found applications in many different areas due to their excellent engineering properties such as high hardness, fracture toughness, good thermal shock and oxidation resistance. SiAlON exist mainly in two different polymorphs: a (MxSi12-(m+n)Al(m+n)OnN16-n; M: metal and rare earth cations, x≈0,35 and n≤1,35) and β (β-Si6-zAlzOzN8-z; 0≤z≤4). In general, stable alpha and beta phases separately as well as in combination of α and β are obtained by incorporation of metal and rare earth cations as sintering additives. The metal cations such as Li, Mg, Ca, Y, and most lanthanide cations with the exception of La, Ce, Pr and Eu are able to stabilise α-SiAlON structure. Ekstrom et al. 1991 found that cerium can not occupy interstitial sites in α-SiAlON structure due to the fact that ionic radius of Ce3+ (0.103 nm) is too large, whereas ionic radius of Ce4+ (0.080 nm) is too small to stabilise α-SiAlON structure. After this work, several studies carried out to incorporate cerium cations into α-SiAlON structure. It was shown that cerium cation alone can be incorporated into α-SiAlON if the samples are either fast cooled after sintering, or when the samples are spark plasma sintered. On the other hand, cerium can also be incorporated into the α-SiAlON structure when it is used as a sintering additive together with a smaller α-SiAlON stabiliser cation such as Yb or Ca. Similar results were observed in other multi-cation doped SiAlONS that non α-SiAlON stabiliser cations like Sr2+ (0.112 nm) and La3+ (0.106 nm) are able to stabilise α-SiAlON when used together with α-SiAlON stabiliser cations such as Ca or Yb. Although it was shown that cerium existed in mixed valance state at domain boundaries in Ce-doped and spark plasma sintered α-SiAlON, there is no work on the valance determination of cerium in sintered α-SiAlON which has no domain boundaries. Therefore, in this study; it was aimed to incorporate cerium into α-SiAlON structure by combining with Yb3+ and the determination of possible cerium valence states (Ce3+/Ce4+) in both α-SiAlON grains and secondary phases.

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.

Dissolution of spent nuclear fuel fragments at high alkaline conditions under H2 overpressure

MRS Advances ◽  
2016 ◽  
Vol 1 (62) ◽  
pp. 4163-4168
Author(s):  
E. González-Robles ◽  
M. Herm ◽  
V. Montoya ◽  
N. Müller ◽  
B. Kienzler ◽  
...  
Keyword(s):  
Aqueous Solution ◽  
Partial Pressure ◽  
Nuclear Fuel ◽  
Alkaline Solution ◽  
Spent Nuclear Fuel ◽  
Dissolution Tests ◽  
The Matrix ◽  
Alkaline Conditions ◽  
Long Term Behavior

ABSTRACTThe long-term behavior of the UO2 fuel matrix under conditions of the Belgian “Supercontainer design” was investigated by dissolution tests of high burn-up spent nuclear fuel (SNF) in high alkaline solution under 40 bar of (Ar + 8%H2) atmosphere. Four fragments of SNF, obtained from a pellet previously leached during two years, were exposed to young cement water with Ca (YCWCa) under 3.2 bar H2 partial pressure in four single/independent autoclave experiments for a period of 59, 182, 252 and 341 days, respectively. After a decrease of the concentration of dissolved 238U, which is associated with a reduction of U(VI) to U(IV), the concentration of 238U in solution is constant in the experiments running for 252 and 341 days. These observations indicate an inhibition of the matrix dissolution due to the presence of H2. A slight increase in the concentration of 90Sr and 137Cs in the aqueous solution indicates that there is still dissolution of the grain boundaries. These findings are similar to those reported for spent nuclear fuel corrosion in synthetic near neutral pH solutions.

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