Iron phosphate glass waste forms for vitrifying Hanford AZ102 low activity waste (LAW), part I: Glass formation model
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Materials Science↗
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Zinc-in-titania waste form for immobilizing lanthanide fission products from electrochemical reprocessing
This paper provides microstructural, microchemical, and bulk X-ray diffraction information for a zinc-in-titania waste form for immobilizing rare-earth oxide (REOx) fission products recovered through precipitation from electrochemical salt wastes. This waste form can be used to immobilize REOx at loadings as high as 40 mass% in a chemically durable matrix consisting of glass-bonded minerals including Zn2TiO4 as well as RE-containing phases of REOx, RE2Ti2O7, and REPO4 (monazite). Chemical durability data is also provided, and comparisons are drawn between this waste form and one made using a similar process with half the REOx loading.
rare earth, rare earth waste form, electrochemical↗
Effect of reduced dehalogenation on the performance of Y zeolite-based sintered waste forms
Not Available
Gardner, Levi (ORCID:0000000273277620)↗
Synergy in Materials: Leveraging Phosphosilicate Waste Forms for Electrochemical Salt Waste
Here, waste forms containing glassy and crystalline phosphate and silicate phases were produced to immobilize salt waste simulants from pyroprocessing and characterized by using Raman spectroscopy, Mössbauer spectroscopy, X-ray diffraction, scanning electron microscopy, heat capacity, and chemical durability measurements. In this work, a phosphosilicate waste form is presented to leverage the benefits of both borosilicate glasses and iron phosphate glasses. To improve waste loading, prior to immobilization, salt simulants were successfully dechlorinated using ammonium dihydrogen phosphate, mixed with a borosilicate frit (5–30 wt %) and Fe 2 O 3 , and vitrified. Additions of 2.5–15 wt % borosilicate glass (NBS3) improved normalized release rates for Cs relative to iron-phosphates without NBS3, resulting in chemical durabilities similar to high-level waste borosilicate glass reference materials. The release rates of the alkalis (i.e., Li, Na, K, Cs) were the lowest with the addition of 5 wt % NBS3. Although Sr was not specifically targeted in this study, evidence exists that it preferentially partitioned with Si to form an amorphous droplet phase within the iron phosphate glass matrix.
37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
Synthesis and characterization of sintered H–Y zeolite-derived waste forms for dehalogenated electrorefiner salt
Not Available
Gardner, Levi D.↗
Molten Salt Reactors and Electrochemical Reprocessing: Synthesis and Chemical Durability of Potential Waste Forms for Metal and Salt Waste Streams
The molten salt reactor (MSR) is one of the leading advanced nuclear reactor candidates to replace current nuclear reactor technologies in the U.S. Besides having more economical and reliable designs, MSRs have are amenable to a closed fuel cycle, in which electrochemical reprocessing can be performed to recycle the used nuclear fuel. This review intends to provide information about potential waste forms for metal and salt waste streams from these salt-based nuclear processes. Metal waste streams arise from reactor components and structural materials. Salt waste streams are generated during reactor operations as fission products build up in salt-fueled systems. Waste forms that have the highest waste loading and/or have shown the most commercial promise are discussed with an emphasis on the current state of efforts to understand the synthesis and chemical durability of metal and ceramic waste forms.
molten salt reactors, electrochemical reprocessing↗
Semi-Dynamic Leach Testing of Densified Silicon-based Iodine Waste Forms
Iodine waste forms (IWF) require a conceptual corrosion release model (CCRM) to provide iodine (I) release rates for performance modeling nuclear waste disposal repositories. To develop a CCRM, an understanding of the corrosion mechanisms of the IWF are required along with data from consistent test methods to parameterize the model. The present study has advanced both areas by providing and assessing the corrosion resistance of IWF types based on their processing history in minor variations of semi-dynamic leach tests. The test included a series of semi-dynamic leach tests using monolithic IWFs in deionized water (leachant). Several experiments were conducted under alternate test conditions with changes to temperature, leachant replacement, leachant pH, leachant volume, masking, and surface finish to elucidate if varying these conditions impacted IWF corrosion behavior. Tests were conducted on two classes of IWFs: (1) I-bearing silver-mordenite (AgZ) materials processed by hot isostatic pressing (HIP) at different temperatures, pressures, sizes, and times; and (2) I-bearing silver-functionalized silica aerogels (SFA) processed by either HIP or spark plasma sintering (SPS). The corrosion susceptibility of AgZ samples was influenced by HIP temperature and pressure. The SPS SFAs retained I far better than HIP SFAs. Additional findings in this study include: (1) The iodine dissolution rate decreased with decreasing temperature, (2) a common ion effect may occur and slow dissolution of the host phase if the leachant is not regularly replaced, (3) pH controls the dissolution rate, and (4) the iodine dissolution rate slows with extended test time (up to 224 days). Based on this work, these parameters should thus be represented when developing a CCRM.
corrosion↗
Dehalogenation of electrochemical processing salt simulants with ammonium phosphates and immobilization of salt cations in an iron phosphate glass waste form
This study demonstrates the conversion of simulated pyroprocessing salt wastes [i.e., ER(SF) and ERV2] into chemically durable iron phosphate glass waste forms through reactions with ammonium hydrogen phosphate precursors [i.e., NH4H2PO4 and (NH4)2HPO4] and added Fe2O3. The reaction of the salt simulants with the phosphate precursors resulted in dehalogenation of the salt through the production of NH4Cl and NH4I (along with water), of which the NH4Cl can be used to produce UCl3 oxidant for use in the electrorefiner during pyroprocessing. These iron phosphate glasses contain a notably higher waste loading than the baseline glass-bonded sodalite ceramic waste form, which requires about 70 vol% sodalite to accommodate Cl from the salt, because they are dehalogenated prior to vitrification. The high waste loadings coupled with higher bulk densities (~3×103 kg/m3) than glass-bonded sodalite (~2.3×103 kg/m3) mean that these glasses have a higher storage volume, i.e., the volume required to immobilize a given mass of waste. The iron phosphate glass waste forms were characterized using a variety of techniques including physical property measurements and chemical durability testing. Following initial testing, a proof-of-concept demonstration was conducted to show that the NH4Cl generated from this process could be used to produce UCl3 using depleted uranium dendrites as a feedstock.
phosphate glass, pyroprocessing, waste forms, salt↗
Compositional control of radionuclide retention in hollandite-based ceramic waste forms for Cs-immobilization
Hollandite materials, as a class of crystalline nuclear waste forms, are promising candidates for the immobilization of radioactive elements, such as Cs, Ba, as well as a variety of lanthanide and transition-metal fission products. In this study, three Ga-doped titanate hollandite-type phases, Ba 1.33 Ga 2.67 Ti 5.33 O 16 , Ba 0.667 Cs 0.667 Ga 2 Ti 6 O 16 , and Cs 1.33 Ga 1.33 Ti 6.67 O 16 , were synthesized using a solid-state reaction route. All synthesized phases adopted a single phase tetragonal structure, as determined by powder X-ray diffraction (XRD), and elemental analysis confirmed the measured stoichiometries were close to targeted compositions. Extended X-ray absorption fine structure spectroscopy (EXAFS) was used to determine the local structural features for the framework of octahedrally coordinated cations. EXAFS data indicated that Cs 1.33 Ga 1.33 Ti 6.67 O 16 possessed the most disordered local structure centered around the Ga dopant. The enthalpies of formation of all three hollandite phases measured using high-temperature oxide melt solution calorimetry were found to be negative, indicating enthalpies of formation of these hollandites from oxides are thermodynamically stable with respect to their constituent oxides. Furthermore, the formation enthalpies were more negative and hence more favorable with increased Cs content. Finally, aqueous leaching tests revealed that high Cs content hollandite phases exhibited greater Cs retention as compared to low Cs content hollandite. While preliminary in nature, this work draws attention to the links between the capacity for radionuclide retention, atomistic level structural features and bulk thermodynamic properties of materials.
nuclear (including radiation effects), materials a↗
Silica-based waste form for immobilization of iodine from reprocessing plant off-gas streams
Not Available
Matyáš, Josef↗
A U-bearing composite waste form for electrochemical processing wastes
Metallic/ceramic composite waste forms are being developed to immobilize combined metallic and oxide waste streams generated during electrochemical recycling of used nuclear fuel. Composites were made for corrosion testing by reacting HT9 steel to represent fuel cladding, Zr and Mo to simulate metallic fuel waste, and a mixture of ZrO2, Nd2O3, and UO2 to represent oxide wastes. More than half of the added UO2 was reduced to metal and formed Fe-Zr-U intermetallics and most of the remaining UO2 and all of the Nd2O3 reacted to form zirconates. Fe-Cr-Mo intermetallics were also formed. Microstructure characterization of the intermetallic and ceramic phases that were generated and tests conducted to evaluate their corrosion behaviors indicate composite waste forms can accommodate both metallic and oxidized waste streams in durable host phases. (c) 2018 Elsevier B.V. All rights reserved.
Chen, X. (ORCID:0000000332734431)↗
Preparation, Performance and Mechanism of Tc and I Getters in Cementitious Waste Forms - 17124
Cementitious waste forms offer a low temperature, cost effective technology for the solidification of liquid wastes. At the Department of Energy Hanford Site, 99-Tc and 129-I are radionuclides contained within low activity waste (LAW) which contribute significantly to environmental impacts associated with clean-up of the site. Cast Stone, a cementitious waste form comprised of 47 wt% blast furnace slag, 45 wt% fly ash and 8 wt% ordinary Portland cement, has undergone investigation as a possible candidate technology for the solidification of LAW. In screening tests using EPA Method 1315 leach testing of various mixes of Cast Stone with simulated LAW, Tc observed effective diffusivities (Dobs) had an average of 5.3 x 10-11 cm{sup 2}/s and I Dobs had an average of 5.7 x 10-9 cm{sup 2}/s. From these measurements, an opportunity exists to improve retention of Tc and I by Cast Stone. One approach to accomplish this is through the incorporation of Tc and I getters to selectively sequester Tc and I in the waste stream and eventual waste form. Initial testing conducted at PNNL identified potassium metal sulfide (KMS-2) and Sn(II)-apatite (Sn-A) as the strongest performing Tc getters and silver exchanged zeolite (Ag-Z) as the top performing I getter. This research also demonstrated that the Tc and I getters, when added to solution simultaneously, can have deleterious interactions with one another and lower the Tc and I removal levels. However, these interactions can be overcome through sequential treatment of solution with the getters. Cast Stone samples were fabricated using a 6.5 M Na average LAW simulant spiked with Tc and I with the addition of various combinations of Tc and I getters. Through EPA Method 1315 and Method 1313 testing it was found that the KMS-2 Tc getter was most successful at lowering Tc Dobs. The I getters were not effective, as apparent interference from sulfide components of the waste form may have led to dissociation of the AgI formed by the Ag-Z getter in sequestering the I from solution. (authors)
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Evidence of Technetium and Iodine from a Sodalite-Bearing Ceramic Waste Form
Current plans for nuclear waste vitrification at the Hanford Tank Waste Treatment and Immobilization Plant (WTP) lack the capacity to treat all of the low activity waste (LAW) that is not encapsulated in the vitrified product. Several technologies are being considered to treat the excess LAW. One such technology is Fluidized Bed Steam Reforming (FBSR). The FBSR process results in a granular product composed of feldspathoid mineral phases that immobilize the major components in the LAW as well as other contaminants of concern (COCs), with Tc and I expected to be present in sodalite cages formed during the process. In order to meet compressive strength requirements at the Hanford Integrated Disposal Facility (IDF), the granular product may be encapsulated in a monolith. To demonstrate the ability of the technology to serve the mission of managing excess LAW, Single Pass Flow-Through (SPFT) tests have been performed on non-radioactive granular materials and granular materials encapsulated in a geopolymer binder produced at the engineering- and bench-scale as well as a granular product produced at the bench scale with actual Hanford tank waste. SPFT tests were conducted at 40 °C for durations up to 2 months with a flow-through solution buffered at pH 9. The forward reaction rate of the non-radioactive mineral product dissolution based on Si release for the granular product was measured to be (6.2 ± 2.1) × 10-4 g/m2d for the engineering-scale product and (13 ± 4.9) × 10-4 g/m2d for the bench-scale product. The resulting non-radioactive monoliths showed forward reaction rates based on Si release of (3.4 ± 1.1) × 10-4 g/m2d for the engineering-scale material and (4.2 ± 1.5) × 10-4 g/m2d for the bench-scale material demonstrating that encapsulation of the FBSR granular product in a monolith does not significantly alter the performance of the material. Finally, an FBSR granular product created at the bench scale using actual Hanford LAW gave similar release values and also demonstrated that Re and Tc show similar release behavior with a dilute-condition release rate of (1.1 ± 0.5) × 10-4 g/m2d for Tc and (1.6 ± 0.6) × 10-4 g/m2d for Re. We believe that this is the first demonstration to show that Re and I have similar release behavior and that similar behavior of Re and Tc also occurs. These results indicate that the FBSR mineral product may be a suitable supplementary immobilization technique at the Hanford site and they also suggest that both Tc and Re occupy similar sites in the sodalite cage structure.
fluidized bed steam reforming↗
Method of making nanostructured glass-ceramic waste forms
A waste form for and a method of rendering hazardous materials less dangerous is disclosed that includes fixing the hazardous material in nanopores of a nanoporous material, reacting the trapped hazardous material to render it less volatile/soluble, and vitrifying the nanoporous material containing the less volatile/soluble hazardous material.
36 MATERIALS SCIENCE↗
Low Temperature Waste Form for Supplemental Immobilization of Hanford Low Activity Waste - 17485
The Hanford site has approximately 56 million gallons of radioactive waste stored in 177 underground storage tanks. The Hanford Waste Treatment and Immobilization Plant (WTP) is being constructed to treat the waste but will only have sufficient capacity to treat about one-third of the Low Activity Waste (LAW) portion within the anticipated time-frame for completing the waste treatment mission. The LAW vitrification facility will need to be supplemented with additional LAW vitrification capacity or an alternate LAW immobilization technology. The LAW immobilization technologies considered must be capable of treating approximately 65% of Hanford's LAW and will be evaluated with respect to facilities required, quantity of final waste forms and secondary wastes produced, waste form performance data, technical viability, and life cycle cost and schedule estimates. A low temperature (i.e. non-thermal) waste form and treatment process is being evaluated to provide the additional LAW immobilization capacity. Low temperature, cementitious waste forms have been evaluated previously for immobilizing Hanford tank waste but were not further developed for various reasons. Over the past few years the Hanford Tank Operations Contractor has conducted a technology development program to evaluate a low temperature waste form for immobilization of Hanford LAW including formulation development and testing with a range of simulant compositions spiked with radioactive and hazardous contaminants of concern (COCs). The objectives of this program include; developing a low temperature waste form formulation that is robust with respect to both waste form performance and processing properties over a range of LAW compositions, demonstrating waste form performance with real waste, and conducting Engineering Scale testing to advance the technical maturity of the low temperature immobilization process. The purpose of this paper is to report progress toward these objectives, in particular testing with real waste including additives to enhance technetium and iodine retention. (authors)
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Radioactive demonstration of final mineralized waste forms for Hanford waste treatment plant secondary waste (WTP-SW) by fluidized bed steam reforming (FBSR) using the bench scale reformer platform
The U.S. Department of Energy’s Office of River Protection (ORP) is responsible for the retrieval, treatment, immobilization, and disposal of Hanford’s tank waste. Currently there are approximately 56 million gallons of highly radioactive mixed wastes awaiting treatment. A key aspect of the River Protection Project (RPP) cleanup mission is to construct and operate the Waste Treatment and Immobilization Plant (WTP). The WTP will separate the tank waste into high-level and low-activity waste (LAW) fractions, both of which will subsequently be vitrified. The projected throughput capacity of the WTP LAW Vitrification Facility is insufficient to complete the RPP mission in the time frame required by the Hanford Federal Facility Agreement and Consent Order, also known as the Tri-Party Agreement (TPA), i.e. December 31, 2047. Therefore, Supplemental Treatment is required both to meet the TPA treatment requirements as well as to more cost effectively complete the tank waste treatment mission. In addition, the WTP LAW vitrification facility off-gas condensate known as WTP Secondary Waste (WTP-SW) will be generated and enriched in volatile components such as 137Cs, 129I, 99Tc, Cl, F, and SO4 that volatilize at the vitrification temperature of 1150°C in the absence of a continuous cold cap (that could minimize volatilization). The current waste disposal path for the WTP-SW is to process it through the Effluent Treatment Facility (ETF). Fluidized Bed Steam Reforming (FBSR) is being considered for immobilization of the ETF concentrate that would be generated by processing the WTP-SW. The focus of this current report is the WTP-SW.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Equilibrium Temperature Profiles within Fission Product Waste Forms
We studied waste form strategies for advanced fuel cycle schemes. Several options were considered for three waste streams with the following fission products: cesium and strontium, transition metals, and lanthanides. These three waste streams may be combined or disposed separately. The decay of several isotopes will generate heat that must be accommodated by the waste form, and this heat will affect the waste loadings. To help make an informed decision on the best option, we present computational data on the equilibrium temperature of glass waste forms containing a combination of these three streams.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗