High Waste Loading Research Articles

Devitrified and Phase Separated Material Found in Simulated High Level Nuclear Waste Glasses Containing Ca and Zn Additions

AbstractIn the UK, blended high level nuclear waste (HLW) streams from the Magnox and THORP reprocessing plants are currently vitrified using a lithium sodium borosilicate base glass frit. Laboratory and full size non-radioactive simulations (produced on the Vitrification Test Rig at Sellafield [1]) of these compositions have shown that these glasses need to be melted at circa 1050°C to obtain a reasonable viscosity for pouring. Also, at high waste loadings an alkali molybdate phase (termed “yellow phase”) can form in these glasses [e.g. 2, 3]. Vitrification flowsheets are set to avoid yellow phase formation as this phase is highly corrosive to the inconel melter in the molten state and is partially water soluble at ambient temperature and so may challenge product quality.Ca and Zn additions to the base glass frit have been found to reduce viscosity and allow melt homogeneity and pouring at lower temperatures. It was also theorised that Ca additions could increase the solubility of Mo and thus reduce the likelihood of yellow phase formation. The composition of the phase separated material in as-cast and heat treated specimens of Ca and Zn HLW glasses produced at both laboratory and full scale is examined in this work

Widening the Envelope of UK HLW Vitrification – Experimental Studies with High Waste Loading Formulations Containing Platinoids

AbstractSellafield Ltd operates 3 vitrification lines to convert highly active concentrate liquor arising as a waste product of reprocessing operations into glass for safe interim storage in the Vitrified Product Store (VPS) prior to long term disposal. Highly Active Liquor (HAL) is stored in Highly Active Storage Tanks and transferred to WVP in batches to the liquid stock tank. It is metered in a semi-continuous batch operation to a calciner (rotating tube furnace) where it is converted into an oxide powder (calcine). Glass frit is fed at the lower end of the calciner where it discharges into an Inconel melter vessel controlled at approximately 1100 C. The glass and calcine are melted together and then poured into a container as a batch operation. After two pours the container is allowed to cool, a lid is then fitted to the container, which after further cooling is welded to the container. This container is then cleaned and transferred to the VPS. Platinoid species containing ruthenium, rhodium and palladium present in the HAL form insoluble oxide phases in the glass product. The platinoid concentration in the glass will increase with increasing waste oxide loading to an extent that settling of the platinoids in the glass may occur, leading to heel enrichment, poor melter performance and difficulty in draining the melter. The viscosity will also increase, which may require higher melter temperatures to mix and pour the molten glass and could result in enhanced corrosion of the melter. Inactive laboratory scale experiments with different glass frit formulations have been performed to determine whether product quality could be maintained with higher platinoid concentrations. Operational envelopes with existing formulations were expanded to observe laboratory trial performance and determine any changes to resulting glass qualities. Also, glasses with high waste incorporations have been produced to test process capability and to ascertain any potential phase separation or devitrification issues that could affect either the process or product performance. Physical properties of the different glass formulations were performed to measure changes in viscosity, density and the rates of settling to examine the amount of phase separation that can occur. The results have shown that ruthenium, palladium or rhodium were insoluble in the melt and were not evenly distributed throughout the glass but clustered together. These results will be used as a basis for further development work. This paper presents some findings of these experiments.

Anaerobic digestion of municipal solid wastes containing variable proportions of waste types

In many parts of the world there are significant seasonal variations in the production of the main organic wastes, food and green wastes. These waste types display significant differences in their biodegradation rates. This study investigated the options for ensuring process stability during the start up and operation of thermophilic high-solids anaerobic digestion of feedstock composed of varying proportions of food and green wastes. The results show that high seed sludge to feedstock ratio (or low waste loading rate) is necessary for ensuring process pH stability without chemical addition. It was also found that the proportion of green wastes in the feedstock can be used to regulate process pH, particularly when operating at high waste loading rates (or low seed sludge to feedstock ratios). The need for chemical pH correction during start-up and digestion operation decreased with increase in green wastes content of the feedstock. Food wastes were found to be more readily biodegradable leading to higher solids reduction while green wastes brought about pH stability and higher digestate solid content. Combining both waste types in various proportions brought about feedstock with varying buffering capacity and digestion performance. Thus, careful selection of feedstock composition can minimise the need for chemical pH regulation as well as reducing the cost for digestate dewatering for final disposal.

06/01639 Commissioning and operation of high level radioactive waste vitrification and storage facilities: the Indian experience: Raj, K. International Journal of Nuclear Energy Science and Technology, 2005, 1, (2–3), 148–163

A simulated high level waste (HLW) containing 4 mass% chrome oxide, whose overall composition is representative of the high chrome oxide wastes at Hanford WA USA, was easily vitrified in a phosphate glass at temperatures ranging from 1150 °C, for waste loadings of 55 mass%, to 1250 °C for waste loadings of 75 mass%. Even at these high waste loadings, these wasteforms had an excellent chemical durability. The best chemical durability was achieved when the O/(Si + P) atomic ratio was between 3.5 and 3.8. These wasteforms were also resistant to crystallization although trace amounts of crystalline Cr2O3 were present in wasteforms containing more than 70 mass% HLW. It is concluded that up to 45 mass% of the total HLW at Hanford, especially that containing as high as 4.5 mass% chrome oxide, could be directly vitrified into an iron phosphate glass, that meets all of the current chemical durability requirements by simply adding 25–35 mass% P2O5 to the waste and melting the mixture at 1150–1250 °C for a few (<6) hours.

Vitrification of high chrome oxide nuclear waste in iron phosphate glasses

A simulated high level waste (HLW) containing 4 mass% chrome oxide, whose overall composition is representative of the high chrome oxide wastes at Hanford WA USA, was easily vitrified in a phosphate glass at temperatures ranging from 1150 °C, for waste loadings of 55 mass%, to 1250 °C for waste loadings of 75 mass%. Even at these high waste loadings, these wasteforms had an excellent chemical durability. The best chemical durability was achieved when the O/(Si + P) atomic ratio was between 3.5 and 3.8. These wasteforms were also resistant to crystallization although trace amounts of crystalline Cr 2O 3 were present in wasteforms containing more than 70 mass% HLW. It is concluded that up to 45 mass% of the total HLW at Hanford, especially that containing as high as 4.5 mass% chrome oxide, could be directly vitrified into an iron phosphate glass, that meets all of the current chemical durability requirements by simply adding 25–35 mass% P 2O 5 to the waste and melting the mixture at 1150–1250 °C for a few (<6) hours.

Modeling solid waste decomposition

The hydrolysis rate coefficients of sorted municipal waste were evaluated from the biochemical methane potential tests using non-linear regression. A distributed mathematical model of anaerobic digestion of rich (food) and lean (non-food) solid wastes with greatly different rates of polymer hydrolysis/acidogenesis was developed to describe the balance between the rates of hydrolysis/acidogenesis and methanogenesis. The model was calibrated using previously published experimental data [Biores. Technol. 52 (1995) 245] obtained upon various initial food waste loadings. Simulations of one- and two-stage digestion systems were carried out. The results showed that initial spatial separation of food waste and inoculum enhances methane production and waste degradation in a one-stage solid-bed digester at high waste loading. A negative effect of vigorously mixing at high waste loading reported in some papers was discussed. It was hypothesized that the initiation methanogenic centers developing in time and expanding in space under minimal mixing conditions might be a key factor for efficient anaerobic conversion of solid waste into methane.

Geochemical behaviour of host phases for actinides and fission products in crystalline ceramic nuclear waste forms

Abstract A number of polyphase or single-phase ceramic waste forms have been considered as options for the disposal of nuclear waste in geological repositories. Of critical concern in the scientific evaluation of these materials is their performance in natural systems over long periods of time (e.g., 10 3 to 10 6 years). This paper gives an overview of the aqueous durability of the major titanate host phases for actinides (e.g., Th, U, Np, Pu, Cm) and important fission products (e.g., Sr and Cs) in alternative crystalline ceramic waste forms. These host phases are compared with reference to some basic acceptance criteria, including the long-term behaviour determined from studies of natural samples. The available data indicate that zirconolite and pyrochlore are excellent candidate host phases for actinides. These structures exhibit excellent aqueous durability, crystal chemical flexibility, high waste loadings, and well-known processing conditions. Although both pyrochlore and zirconolite become amorphous due to alpha-decay processes, the total volume swelling is only 5–6% and there is no significant effect of radiation damage on aqueous durability. Hollandite also appears to be an excellent candidate host phase for radioactive Cs isotopes. Brannerite and perovskite, on the other hand, are more prone to alteration in aqueous fluids and have a lower degree of chemical flexibility. With the exception of hollandite, many of the properties of these potential host phases have been confirmed through studies of natural samples.

A distributed model of solid waste anaerobic digestion: sensitivity analysis

A distributed model of anaerobic digestion of solid waste was developed to describe the balance between the rates of polymer hydrolysis and methanogenesis during the anaerobic conversion of rich and lean wastes in batch and continuous-flow reactors. Waste, volatile fatty acids (VFAs), methanogenic biomass and sodium concentrations are the model variables. Diffusion and advection of VFAs inhibiting both polymer hydrolysis and methanogenesis were considered. A sensitivity analysis by changing the key model parameter values was carried out. The model simulations showed that the effective distance between the areas of hydrolysis/acidogenesis and methanogenesis is very important. An initial spatial separation of rich waste and inoculum enhances the methane production and waste degradation at high waste loading if relatively low VFA diffusion into the methanogenic area is taking place. When both hydrolysis and methanogenesis are strongly inhibited by high levels of VFA, fluctuations in biomass concentration are thought to be responsible for initiating the expansion of methanogenic area over the reactor space.

Iron – Phosphate Glass (IPG) Waste Forms Produced using Induction Melter with Cold Crucible

ABSTRACTA simulated sodium bearing waste (SBW), which represented a type of high sodium and sulfate waste, was successfully vitrified in iron phosphate glasses (IPG), at a maximum waste loading of 40 wt%, using a cold crucible induction melter (CCIM). Scanning electron microscopy (SEM) and X-ray diffraction (XRD) showed that all of the IPG waste forms did not contain sulfate salt segregation or crystalline phases. The calculated composition and the average analytical composition obtained by Electron Probe Microanalysis (EPMA) were in good agreement. The major elements were uniformly distributed throughout the samples. The chemical durability of the IPG waste forms containing 40wt% SBW was evaluated by the product consistency test (PCT) and met current DOE requirements. IPG waste forms were melted at a relatively low temperature and for short times compared to borosilicate glasses. These advantages, combined with those of a significantly higher waste loading and the feasibility for CCIM melting, offer a considerable savings in time, energy, and cost for vitrifying this high sodium and sulfate waste.

Failure analysis of a pilot scale melter

Failure of the pilot-scale test melter resulted from severe overheating of the Inconel 690 (690) jacketed molybdenum electrode. Extreme temperatures were required to melt the glass during this campaign because the feed material contained a very high waste loading. Metallurgical evaluation revealed the presence of an alloy containing nickel and molybdenum in several ingots found on the bottom of the melter and on a drip which had solidified on the electrode sheath. This indicates that a major portion of the electrode assembly was exposed to a temperature of at least 1317°C, the nickel/molybdenum eutectic temperature. Small regions on the end of the 690 sheath showed evidence of melting, indicating that this localized region exceeded 1345 °C, the melting point of 690. In addition to nickel, antimony was found on the grain boundaries of the molybdenum electrode. This also contributed to the failure of the electrode. The source of the antimony was not identified but is believed to have originated from the feed material. Metallurgical evaluation also revealed that nickel had attacked the grain boundaries of the molybdenum/tungsten drain valve. This component did not fail in service; however, intergranular attack led to degradation of the mechanical properties, resulting in the fracture of the drain valve tip during disassembly. Antimony was not observed on this component.

10.2172/918137

The Defense Waste Processing Facility (DWPF) is about to process High Level Waste (HLW) Sludge Batch 4 (SB4). This sludge batch is high in alumina and nepheline can crystallize readily depending on the glass composition. Large concentrations of crystallized nepheline can have an adverse effect on HLW glass durability. Several studies have been performed to study the potential for nepheline formation in SB4. The Phase 3 Nepheline Formation study of SB4 glasses examined sixteen different glasses made with four different frits. Melt rate experiments were performed by the Process Science and Engineering Section (PS&E) of the Savannah River National Laboratory (SRNL) using the four frits from the Phase 3 work, plus additional high B2O3/high Fe2O3 frits. Preliminary results from these tests showed the potential for significant improvements in melt rate for SB4 glasses using a higher B2O3-containing frit, particularly Frit 503. The main objective of this study was to investigate the durability of SB4 glasses produced with a high B2O3 frit likely to be recommended for SB4 processing. In addition, a range of waste loadings (WLs) was selected to continue to assess the effectiveness of a nepheline discriminator in predicting concentrations of nepheline crystallization that would be sufficient to influencemore » the durability response of the glass. Five glasses were selected for this study, covering a WL range of 30 to 50 wt% in 5 wt% increments. The Frit 503 glasses were batched and melted. Specimens of each glass were heat-treated to simulate cooling along the centerline of a DWPF-type canister (ccc) to gauge the effects of thermal history on product performance. Visual observations on both quenched and ccc glasses were documented. A representative sample from each glass was submitted to the SRNL Process Science Analytical Laboratory (PSAL) for chemical analysis to confirm that the as-fabricated glasses corresponded to the defined target compositions. The Product Consistency Test (PCT, ASTM C1285) was performed in triplicate on each Frit 503 quenched and ccc glass to assess chemical durability. The experimental test matrix also included the Environmental Assessment (EA) glass and the Approved Reference Material (ARM-1) glass. Representative samples of all the ccc glasses were examined for homogeneity visually and by X-ray diffraction (XRD) analysis. Chemical composition measurements indicated that the experimental glasses were close to their target compositions. PCT results showed that all of the Fit 503 quenched glasses had an acceptable durability compared to the EA benchmark glass. The durability of one of the ccc glasses, NEPHB-04, was statistically greater than its quenched counterpart. However, this was shown to be of little practical significance, as the durability of the NEPHB-04 ccc glass was acceptable when compared to the durability of the EA benchmark glass. Visual observations and PCT results indicated that all of the Frit 503 quenched glasses were free of any crystallization that impacts durability. For the ccc glasses, XRD results indicated that the lower WL glasses (30 to 40 wt%) were amorphous, which was consistent with visual observations and PCT responses. The higher WL glasses (45 and 50 wt%) were shown by XRD to contain spinel (trevorite, NiFe2O4). It is possible that some of the other high WL glasses also contained some nepheline, but that the amount of nepheline crystallization was below the detection limit (0.5 vol%) associated with XRD. The results indicate that Frit 503 is a good candidate for SB4 processing, based on chemical durability of homogeneous and devitrified glasses over a WL range of 30 - 50%. It should be noted that the higher WL glasses would not be fit for processing in DWPF as they exceed other process related criteria (such as liquidus temperature). However, this is only one of many factors influencing the frit selection. Melt rate and the final SB4 composition are also important factors in frit selection. Additional melt rate studies are currently underway, and the final composition projection for SB4 is expected shortly.« less

Vitrification of radioactive waste by reaction sintering under pressure

Silicate nuclear waste glasses were synthesized by reaction sintering of powdered precursors under pressure. The glass samples contained a glass matrix phase with embedded zirconia (baddeleyite) particles. A waste composition with 38 wt% of ZrO 2 was prepared with a waste loading of 30–50 wt% at 800°C and 28 MPa, by hot isostatic pressing. The glass former was commercial amorphous silica powder to which simulated waste was added as calcined oxides. Phase compositions and microstructure of the sintered glass samples were characterized using scanning and analytical electron microscopy. The results show that extensive sintering took place and that a continuous glass phase was formed, particularly at higher waste loading. Waste components such as Na 2O, CaO, MnO 2, La 2O 3, Fe 2O 3, Cr 2O 3, and P 2O 5 dissolved completely in the glass phase. ZrO 2 was also dissolved but recrystallized from the glass as aggregates of baddeleyite crystallites surrounding the original silica particles. MCC-1 type chemical durability tests showed that the glasses are durable with dissolution rates similar to or lower than that of the highly durable French R7T7 borosilicate glass. This glass contains 13 wt% high-level radioactive waste from light water reactor fuel reprocessing and has a melting temperature of 1150°C. The long-term chemical durability of our sintered glasses is expected to be as high as that of rhyolitic glasses, based on hydration energies of 3.7 and 3.3 kJ/mole, respectively. Rhyolitic glasses show little alteration over geological periods of time with a typical corrosion rate of 1 μm/1000 yr.

The Effects of Uranium on the Structure of Iron Phosphate Glasses

AbstractBecause of their high chemical durability and waste loading capacity, iron phosphate glasses are a natural candidate for a nuclear waste disposal medium. We have studied the effects of uranium on the structure of iron phosphate glasses with both neutron and high-energy x-raydiffraction. The results of neutron scattering, which is mostly sensitive to pair correlations involving light atoms i.e. O-O, Fe-O and P-O, indicate the main structural features of the base glass are largely unaffected by the addition of UO2. The nearest-neighborbour P-O, Fe-O and O- O peaks remain at the same position in real space and their intensities scale approximately with concentration. These findings are consistent with earlier results using Raman scattering and EXAFS on the Fe-K edge, where in both cases the spectra remain similar to the base glass. The results of high-energy x-ray scattering, which is sensitive to correlations involving the heavier atoms and thus complements the neutron measurements, are also consistent with the overall picture of uranium occupying interstitial sites in the relatively undisturbed base glass structure. Combining the neutron and x-ray data for a 10 mol% UO2 glass suggests the intriguing possibility of a U6+ uranyl ion configuration although further work is needed to establish the precise local structure and valence state of uranium in these glasses.

Microstructures and leach rates of glass–ceramic nuclear waste forms developed by partial vitrification in a hot isostatic press

A high level nuclear waste calcine simulant is transformed to a dense and durable glass–ceramic waste form by addition of glass and crystal forming components, and hot isostatic pressing at 1000 °C and 138 MPa. The waste forms are abundantly composed of zircon, beddeyelite, apatite, fluorite, greenockite and boroaluminosilicate glass. The crystal nucleating, glass forming and volatilizing components of the calcine are partitioned into crystalline and glass phases such that 95 wt% of the waste components, including actinide surrogates, stoichiometrically reside in the crystalline phases. This results in a high waste loading of 60–80 wt% calcine in the total glass–ceramic. The partitioning follows the natural association of elements, as a result, species like P avoid the glass phase. Instead glass accommodates the incompatible solutes like Cs. It minimizes porosity and bonds the polyphase ceramic microstructure, which resembles rhyolite or basalt volcanic rocks. Both glass and crystals contribute to high chemical durability, which is degraded when glass devitrifies with lowering of partial liquid viscosity by higher MgO additions. The devitrified phases are layered mica, dendritic nepheline and fibrous alkaline-earth borate. These phases are enriched in the mobile elements of Cs, Na and B, respectively. © 1998 Chapman & Hall

Index of authors and papers to Physica B 161–181

Cs3Bi2I9, a defect perovskite derivative, is a potential host phase to immobilize iodine and cesium with high waste loadings. In this work, two strategies were explored to form Cs3Bi2I9-silica composites and a core-shell structure in order to improve chemical durability of waste form materials meanwhile maintaining high waste loadings. Cs3Bi2I9 loadings as high as 70 wt.% were incorporated into a silica matrix to form silica-ceramic composites, and 20 wt.% Cs3Bi2I9 was encapsulated into silica to form a core–shell structure by low temperature spark plasma sintering. Chemical durability of the composite and core-shell waste forms was evaluated by semi-dynamic leaching experiments, and Cs and I were incongruently released from waste form matrices. A BiOI alteration layer formed, acting as a passivation layer to reduce the release of radionuclides. The long-term iodine release rate was low (30 mg m−2 day-1) for the 70 wt.% Cs3Bi2I9–silica composite leached in deionized water at 90 °C, which can be further reduced to 5 × 10−3 mg m−2 day−1 for the 20 wt.% core-shell structure. This work highlights a robust way to immobilize the highly mobile radionuclides with high waste loadings through encapsulation into durable matrices and a surface passivating mechanism that can greatly reduce the elemental transport from waste form materials and significantly enhance their chemical durability.

The Development of a Ceramic Waste Form for Immobilisation of Highly Active Wastes from Radical Purex Reprocessing Operations

ABSTRACTThe development of novel, Radical Purex, reprocessing technologies, leading to fission product waste streams with high levels of inert constituents, may mean that vitrification is no longer the optimum solution for the immobilisation of highly active wastes at the back end of the nuclear fuel cycle. A ceramic phase assemblage is described which uses the inert constituents of the waste stream as a functional component of the waste form, permitting high waste loadings to be achieved and thus enabling waste minimisation considerations to be satisfied. The initial development of this phase assemblage is presented.

Optimization of Savannah River M-Area Mixed Waste for Vitrification

ABSTRACTVitrification studies of actual Savannah River M-Area mixed wastes have shown that the limiting factor for high waste loading of this waste stream is its chemical durability as defined by the toxicity characteristics leaching procedure (TCLP). As part of the optimization study of Savannah River M-Area wastes, a number of additives were examined including Na2O, Li2O, B2O3, ZrO2, and TiO2. This paper reports on the effect of varying the boron to total alkali ratio and on the effect of substitutions such as ZrO2 for waste and TiO2 for SiO2 on the chemical durability and processability of M-Area waste glasses.

An Electrodeless Melter for Vitrification of Nuclear Waste

ABSTRACTThis paper describes a new concept for a high-temperature, electrodeless melter for vitrifying radioactive wastes. Based on the principles of induction heating, it circumvents a number of difficulties associated with existing technology. The melter can operate at higher temperatures (1500–2000°C vs 1150°C), allowing for a higher quality, more durable glass which reduces the long-term leaching rate. Higher processing temperatures also enable conversion from borosilicate to high-silica glass which can accommodate 2 to 3 times as much radioactive waste, potentially halving the ultimate required long-term disposal space. Finally, with high temperatures, conversion of nuclear waste into ceramics can also be considered. This too leads to higher waste loading and the reduction of repository space. The melter is toroidal, linked by an iron core transformer that allows efficient electrical operation even at 60 Hz. One-dimensional electrical and thermal analyses are presented.

Zircon: A host-phase for the disposal of weapons plutonium

Zircon, ZrSiO4, is a well-characterized, naturally occurring phase that is extremely durable. Zircon has been synthesized with Pu-concentrations up to 10 wt. % and radiation-damage effects studied to saturation doses of nearly 0.8 displacements per atom. We propose that zircon be used as a waste form for the disposal of the more than 100 metric tons of plutonium that will result from the dismantling of nuclear weapons. There are already several demonstrated processing technologies, of which hot pressing offers the most potential. This highly durable material, even under hydrothermal conditions, with its high waste loading and smaller volume allows deep, permanent disposal of the weapons plutonium in geologic environments in which the borosilicate waste-form glass would not be stable.

Characterization of Highly Waste Loaded Glass for HLW

AbstractPossibility of highly waste loaded borosilicate glasses up to 65wt% were investigated for HLW. In the case the waste was loaded at a content higher than 45wt%, apatite crystal was identified in the glasses. Adding above 2.5wt% of M0O3, yellow phase was observed in the 45wt% waste glass. In the case the composition of actual waste was simulated, waste(25–45wt%) glasses were characterized. Chemical durability of 45wt% waste glass was equivalent to that of 25wt% waste glass under the condition of Na20 content lower than 12wt%. Rheological behavior of 45wt% waste glass was equivalent to that of 25wt% waste glass in the range of melting temperature. Electrical resistivity of 45wt% waste glass were a little lower than that of 25wt% waste glass and it depended on the noble metal content. The Time-Temperature-Transformation (T-T-T) diagram and the leach rate of heat treated glass indicated a good thermal stability of highly waste loaded glass below the glass transition temperature. Possibility of high waste loading up to 45wt% was elucidated.