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DOE OSTI2013

Single phase melt processed powellite (Ba,Ca)MoO4 for the immobilization of Mo-rich nuclear waste

Cs is one of the more problematic fission product radionuclides to immobilize due to its high volatility at elevated temperatures, ability to form water soluble compounds, and its mobility in many host materials. The hollandite structure is a promising crystalline host for Cs immobilization and has been traditionally fabricated by solid state sintering methods. This study presents the structure and performance of Ba{sub 1.0}Cs{sub 0.3}A{sub 2.3}Ti{sub 5.7}O{sub 16}; A = Cr, Fe, Al hollandite fabricated by melt processing. Melt processing is considered advantageous given that melters are currently in use for High Level Waste (HLW) vitrification in several countries. This work details the impact of Cr additions that were demonstrated to i) promote the formation of a Cs containing hollandite phase and ii) maintain the stability of the hollandite phase in reducing conditions anticipated for multiphase waste form processing.

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DOE OSTI2015

Maintaining department of energy (DOE) waste disposition pathways in today's uncertain budgetary environment - 15129

For many years the Department of Energy (DOE) and its prime contractors have worked collaboratively with commercial waste transportation, treatment, and disposal providers to develop and implement outlets for the efficient management and disposition of radioactive wastes, particularly 'no-path-forward' or 'difficult-to-treat' wastes. Both the DOE and commercial providers have made substantial investments to develop capabilities and infrastructure to address these wastes and both have benefited from this collaboration. For example, DOE realized significant cost savings and avoidances when a commercial facility became available that allowed closure of the TSCA incinerator in Oak Ridge or when commercial capabilities were expanded that precluded the need to build a new Hanford M-91 waste treatment facility, estimated at $540 million for design, permitting, and construction. In addition, expanded capabilities developed at commercial facilities ensured a known, if not guaranteed, source for managing waste and improving efficiencies for wastes which previously had no viable disposition outlet or that required extended schedules and cost for completion. Such wastes include sodium- and mercury-bearing wastes, classified shapes, or many forms of suspect transuranic (TRU) wastes. (authors)

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DOE OSTI2020

Key Technical Issues for Greater-Than-Class-C (GTCC) Waste Disposal - 20164

Greater-than-Class C (GTCC) waste is low-level waste (LLW) that exceeds the Class C concentrations tabulated in Title 10, Code of Federal Regulations (CFR), Part 61. Disposal of GTCC waste in the near surface (i.e. upper 30 m of the earth's surface) is prohibited in the United States. Though GTCC waste disposal is generally prohibited, the Commission can approve disposal according to 10 CFR 61.55(a)(2)(iv) on a case-by-case basis. The waste classification tables, developed in the early 1980's, were based on model projections of dose to an inadvertent intruder in agricultural, construction, and discovery exposure scenarios. The assumptions and parameters used were documented in a series of public reports. Though different disposal facility designs and site conditions were considered, the waste classification tables were based on a disposal facility design located in a specific environment. The tables were based on shallow (i.e., top few meters) trench facility designs and did not consider deeper facilities. To determine the suitability of GTCC waste disposal in the near-surface, site-specific analyses must account for differences between GTCC waste and Class A, B, and C LLW (hereafter, traditional LLW). GTCC waste can have concentrations of radionuclides that are much higher than traditional LLW. Because of these higher concentrations, processes that are typically not significant for traditional LLW may be significant with respect to disposal of GTCC waste. These processes include, but are not limited to, heat generation, criticality, and radiolysis. The form of the waste as well as the barriers to release of the waste (e.g. waste package) could be substantially different than they are for traditional LLW. These barriers need to be considered when assessing the impacts of accidents during receipt and placement and in evaluating long-term performance. Some GTCC radioactivity is either embedded in stainless steel or contained in stainless steel barriers. Stainless steel can have very low corrosion rates under a variety of environmental conditions. Finally, and possibly most importantly, GTCC waste would likely need to be disposed deeper than traditional LLW to reduce the probability of disturbance. The waste classification table values of traditional LLW are based on an inadvertent intruder excavating into the waste and bringing some of the material to the land surface. If waste is deeper than approximately 5 m, the excavation scenario becomes very unlikely. Therefore, other intruder scenarios, such as drilling exposure scenarios, need to be evaluated. This paper summarizes key technical issues for the disposal of GTCC waste. Our previous study shows that certain GTCC waste may be suitable for near surface disposal whereas others may not. This study may support ongoing technical analyses assessing potential disposal of GTCC in a near surface disposal facility. (authors)

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DOE OSTI2016

Defense Waste Processing Facility (DWPF) Durability-Composition Models and the Applicability of the Associated Reduction of Constraints (ROC) Criteria for High TiO 2 Containing Glasses

Radioactive high-level waste (HLW) at the Savannah River Site (SRS) has successfully been vitrified into borosilicate glass in the DWPF since 1996. Vitrification requires stringent product/process (P/P) constraints since the glass cannot be reworked once it has been poured into ten foot tall by two foot diameter canisters. A unique “feed forward” statistical process control (SPC) was developed for this control rather than relying on statistical quality control (SQC). In SPC, the feed composition to the DWPF melter is controlled prior to vitrification. In SQC, the glass product would be sampled after it is vitrified. Individual glass property-composition models form the basis for the “feed forward” SPC. The models transform constraints on the melt and glass properties into constraints on the feed composition going to the melter in order to determine, at the 95% confidence level, that the feed will be processable and that the durability of the resulting waste form will be acceptable to a geologic repository. The DWPF SPC system is known as the Product Composition Control System (PCCS). One of the process models within PCCS is known as the Thermodynamic Hydration Energy Reaction MOdel (THERMO™). The DWPF will soon be receiving increased concentrations of TiO 2 -, Na 2 O-, and Cs 2 O-enriched wastes from the Salt Waste Processing Facility (SWPF). The SWPF has been built to pretreat the high-curie fraction of the salt waste to be removed from the HLW tanks in the F- and H-Area Tank Farms at the SRS. In order to validate the existing TiO 2 term in THERMO™ beyond 2.0 wt% in the DWPF, new durability data were developed over the target range of 2.00 to 6.00 wt% TiO 2 and evaluated against the 1995 durability model. The durability was measured by the 7-day Product Consistency Test. This study documents the adequacy of the existing THERMO™ terms. It is recommended that the modified THERMO™ durability models and the modified property acceptable region limits for the durability constraints be incorporated in the next revision of the technical bases for PCCS and then implemented into PCCS. It is also recommended that an reduction of constraints of 4 wt% Al 2 O 3 be implemented with no restrictions on the amount of alkali in the glass for TiO 2 values ≥2 wt%. The ultimate limit on the amount of TiO 2 that can be accommodated from SWPF will be determined by the three PCCS models, the waste composition of a given sludge batch, the waste loading of the sludge batch, and the frit used for vitrification.

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DOE OSTI2012

Form and Aging of Plutonium in Savannah River Site Waste Tank 18

Process modeling; injection molding; long-fiber thermoplastic; fiber orientation; fiber lengt This report provides a summary of the effects of aging on and the expected forms of plutonium in Tank 18 waste residues. The findings are based on available information on the operational history of Tank 18, reported analytical results for samples taken from Tank 18, and the available scientific literature for plutonium under alkaline conditions. These findings should apply in general to residues in other waste tanks. However, the operational history of other waste tanks should be evaluated for specific conditions and unique operations (e.g., acid cleaning with oxalic acid) that could alter the form of plutonium in heel residues. Based on the operational history of other tanks, characterization of samples from the heel residues in those tanks would be appropriate to confirm the form of plutonium. During the operational period and continuing with the residual heel removal periods, Pu(IV) is the dominant oxidation state of the plutonium. Small fractions of Pu(V) and Pu(VI) could be present as the result of the presence of water and the result of reactions with oxygen in air and products from the radiolysis of water. However, the presence of Pu(V) would be transitory as it is not stable at the dilute alkaline conditions that currently exists in Tank 18. Most of the plutonium that enters Savannah River Site (SRS) high-level waste (HLW) tanks is freshly precipitated as amorphous plutonium hydroxide, Pu(OH) 4(am) or hydrous plutonium oxide, PuO 2(am,hyd) and coprecipitated within a mixture of hydrous metal oxide phases containing metals such as iron, aluminum, manganese and uranium. The coprecipitated plutonium would include Pu 4+ that has been substituted for other metal ions in crystal lattice sites, Pu 4+ occluded within hydrous metal oxide particles and Pu 4+ adsorbed onto the surface of hydrous metal oxide particles. The adsorbed plutonium could include both inner sphere coordination and outer sphere coordination of the plutonium. PuO 2(am,hyd) is also likely to be present in deposits and scales that have formed on the steel surfaces of the tank. Over the operational period and after closure of Tank 18, Ostwald ripening has and will continue to transform PuO 2(am,hyd) to a more crystalline form of plutonium dioxide, PuO 2(c) . After bulk waste removal and heel retrieval operations, the free hydroxide concentration decreased and the carbonate concentration in the free liquid and solids increased. Consequently, a portion of the PuO 2(am,hyd) has likely been converted to a hydroxy-carbonate complex such as Pu(OH) 2 (CO 3 ) (s) . or PuO(CO 3 ) · xH 2 O (am) . Like PuO 2(am,hyd) , Ostwald ripening of Pu(OH) 2 (CO 3 ) (s) or PuO(CO 3 ) · xH 2 O (am) would be expected to occur to produce a more crystalline form of the plutonium carbonate complex. Due to the high alkalinity and low carbonate concentration in the grout formulation, it is expected that upon interaction with the grout, the plutonium carbonate complexes will transform back into plutonium hydroxide. Although crystalline plutonium dioxide is the more stable thermodynamic state of Pu(IV), the low temperature and high water content of the waste during the operating and heel removal periods in Tank 18 have limited the transformation of the plutonium into crystalline plutonium dioxide. During the tank closure period of thousands of years, transformation of the plutonium into a more crystalline plutonium dioxide form would be expected. However, the continuing presence of water, reaction with water radiolysis products, and low temperatures will limit the transformation, and will likely maintain an amorphous Pu(OH) 4 or PuO 2(am,hyd) form on the surface of any crystalline plutonium dioxide produced after tank closure. X-ray Absorption Spectroscopic (XAS) measurements of Tank 18 residues are recommended to confirm coordination environments of the plutonium. If the presence of PuO(CO 3 ) (am,hyd) is confirmed by XAS, it is recommended that experiments be conducted to determine if plutonium carbonates are transformed back into PuO 2(am,hyd) upon contact with grout.

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DOE OSTI2024

Electrochemically Enhanced Carbonate Precipitation into Building Materials: A Scalable Carbon Sequestration Strategy

Decarbonization goals across hard-to-abate industries have prompted an urgent need for advanced carbon capture and storage technologies. Sequestering CO2 into carbonate minerals is a scalable method of carbon management with the ability to produce value-added carbon negative materials from waste streams for the construction industry. Waste streams rich in Ca and Mg such as nickel mine tailings, iron/steel slag, and reverse osmosis brines can store 7.6 Mt CO2/year as minerals. Additionally, CO2 mineralization in acid-neutralization processes currently present in industrial waste treatment can eliminate associated CO2 emissions of lime processing by improving process circularity. The carbonate minerals formed from these waste sources are valuable as components of carbon-negative concrete, which have the potential to sequester 1.8 billion Mt of CO2/year. Electrochemical means of CO2 mineralization improves the kinetics of the thermodynamically favorable mineralization process, lessening or eliminating the high energy requirements of traditional methods. Here, we investigate benchtop scale electrochemical CO2 mineralization of alkaline mining waste, highlighting the effects of key constituents in mining waste on the mineralization process.

carbon capture↗

DOE OSTI2025

Strontium Speciation in Relevant Tank Waste Components Examined by Electrospray Ionization Mass Spectrometry

The identification of chemical species formed in complex nuclear waste is crucial for the development and employment of advanced separations technologies to remediate the Hanford site by processing tank waste. The current Tank Side Cesium Removal (TSCR) process deployed at Hanford utilizes crystalline silicotitanate (CST) ion exchange (IX) media to aid in the separation of low-activity waste for proper treatment and disposal. The inorganic IX media is highly selective for Cs but has been shown to also remove Sr from caustic simulants and small-scale IX processing of Hanford tank waste.(Fiskum, Rovira et al. 2019, Fiskum, Campbell et al. 2021, Westesen, Campbell et al. 2022) Quantitative Sr removal has not been observed in all tank waste supernates tested; thus, to better understand Sr removal and effectively predict processing behavior through TSCR, it is necessary to first investigate Sr speciation in tank waste. This work utilized electrospray ionization mass spectrometry (ESI-MS) to identify ionic Sr complexes that form in the presence of NO 3 –, NO 2 –, OH–, and Cl–. Although our results show that NO 3 –, NO 2 –, and OH– are competitive for Sr 2+ binding, previous data from IX studies indicate that [SrOH] + is not the dominant species of concern in tank waste processing schemes.(Fiskum, Campbell and Trang-Le 2020) Our results show that the [Sr(NO3)]+ species and the [Sr(NO2)] + species form in considerable abundances, which may affect the ability to separate Sr using CST in nuclear waste separation processes.

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DOE OSTI2013

Networks of recyclable material waste-picker’s cooperatives: An alternative for the solid waste management in the city of Rio de Janeiro

Highlights: ► In the marketing of recyclable materials, the waste-pickers are the least wins. ► It is proposed creating a network of recycling cooperatives to achieve viability. ► The waste-pickers contribute to waste management to the city. - Abstract: The objective of this study is to discuss the role of networks formed of waste-picker cooperatives in ameliorating problems of final disposal of solid waste in the city of Rio de Janeiro, since the city’s main landfill will soon have to close because of exhausted capacity. However, it is estimated that in the city of Rio de Janeiro there are around five thousand waste-pickers working in poor conditions, with lack of physical infrastructure and training, but contributing significantly by diverting solid waste from landfills. According to the Sustainable Development Indicators (IBGE, 2010a,b) in Brazil, recycling rates hover between 45% and 55%. In the municipality of Rio de Janeiro, only 1% of the waste produced is collected selectively by the government (COMLURB, 2010), demonstrating that recycling is mainly performed by waste-pickers. Furthermore, since the recycling market is an oligopsony that requires economies of scale to negotiate directly with industries, the idea of working in networks of cooperatives meets the demands for joint marketing of recyclable materials. Thus, this work presents a method for creating and structuring a network of recycling cooperatives, with prior training for working in networks, so that the expected synergies and joint efforts can lead to concrete results. We intend to demonstrate that it is first essential to strengthen the waste-pickers’ cooperatives in terms of infrastructure, governance and training so that solid waste management can be environmentally, socially and economically sustainable in the city of Rio de Janeiro.

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NASA NTRS1980

U.S. program assessing nuclear waste disposal in space - A status report

Various concepts for the space disposal of nuclear waste are discussed, with attention given to the destinations now being considered (high earth orbit, lunar orbit, lunar surface, solar orbit, solar system escape, sun). Waste mixes are considered in the context of the 'Purex' (Plutonium and Uranium extraction) process and the potential forms for nuclear waste disposal (ORNL cermet, Boro-silicate glass, Metal matrix, Hot-pressed supercalcine) are described. Preliminary estimates of the energy required and the cost surcharge needed to support the space disposal of nuclear waste are presented (8 metric tons/year, requiring three Shuttle launches). When Purex is employed, the generated electrical energy needed to support the Shuttle launches is shown to be less than 1%, and the projected surcharge to electrical users is shown to be slightly more than two mills/kW-hour.

Rice, E. E.↗

DOE OSTI2016

Effect of B on Crystallization of Li and Na aluminosilicates - 16323

The vitrification of Hanford high-level waste (HLW) requires careful balance of glass-forming additives and nuclear waste components to maximize the waste loading but prevent adverse effects such as crystallization of aluminosilicates upon cooling in canisters. Aluminosilicate crystallization removes chemically durable glass-formers from the melt, leaving behind a residual glass that is enriched in less durable components, such as alkali and boron. In our previous work, it has been shown that the composition of the starting melt has a large influence on the phase and composition of the nucleating crystals, typically nepheline (NaAlSiO{sub 4}) when wastes are simultaneously high in Na and Al. Chemical partitioning allows for additional crystallization in the residual glass. Interestingly, in HLW glass the primary phases to form are those of sodium or lithium aluminosilicates despite the presence of boron, a common additive known to reduce crystallization and lower viscosity and melting point. In this work, we seek to better understand aluminosilicate crystallization in the presence of boron. Twelve glasses were prepared along the tie lines of NaAlSiO{sub 4}-NaBSiO{sub 4} and LiAlSiO{sub 4}-LiBSiO{sub 4}. Glasses were heat-treated following a schedule that traversed the crystallization temperatures of carnegieite (NaAlSiO{sub 4}), nepheline, and eucryptite (LiAlSiO{sub 4}) as determined by thermal analysis. Samples were analyzed using X-ray diffraction (XRD), Raman spectroscopy, and differential scanning calorimetry (DSC). It was observed that Li glasses tended to crystallize at higher boron levels than Na glasses. This behavior may result from the relationship between the structure of the glass network and the structure of the resulting crystal and the relative need for atomic rearrangement prior to crystallization in the Na bearing glass. (authors)

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DOE OSTI2013

Hanford and Savannah River Site Programmatic and Technical Integration

Abstract only. The Hanford Site and the Savannah River Site (SRS) were the primary plutonium production facilities within the U.S. nuclear weapons complex. Radioactive wastes were generated as part of these missions and are stored in similar fashion. The majority of radioactivity maintained by the two sites is located in underground carbon steel tanks in the physical form of supernatant, saltcake, or sludge. Disposition of SRS tank waste is ongoing by converting it into glass (pathway for sludge and radionuclides separated from supernatant or dissolved saltcake) or cement (pathway for the decontaminated supernatant and dissolved saltcake). Tank closure activity has also begun at SRS and will continue for the duration of mission. The Hanford tank waste inventory is roughly 2/3rds larger than SRS's by volume- but nominally half the radioactivity. The baseline disposition path includes high-level and low-activity waste vitrification with separate disposition of contact-handled transuranic tank waste. Retrieval of tank waste from aging single shell tanks (SSTs) into double-shell tanks (DSTs) is currently ongoing. As vitrification commences later this decade, Hanford will be in a similar operations mode as SRS. Site integration is increasing as the missions align. The ongoing integration is centered on key issues that impact both sites- regardless of mission timeframe. Three recent workshop exchanges have been held to improve communication with the primary intent of improving operations and technical work organization. The topics of these workshops are as follows: DST space utilization, optimization, and closure; Waste Feed Qualification; and, Cementitious Waste Forms. Key goals for these and future exchanges include aligning research and technology, preparing for joint initiatives (to maximize budgetary value for the customer), and reviewing lessons learned. Each site has played a leading role in the development of technology and operational practices that can be used to meet current challenges and to minimize the impact of future challenges. This paper provides an overview of the exchanges held, but predominately focuses on the team development and actions leading from the workshops.

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NASA NTRS1992

Glass corrosion in natural environments

Experiments carried out during the progress period are summarized. Experiments carried out involving glass samples exposed to solutions of Tris have shown the appearance of 'spikes' upon monitoring glass dissolution as a function of time. The periodic 'spikes' observed in Tris-based media were interpreted in terms of cracking due to excessive stress in the surface region of the glass. Studies of the interactions of silicate glasses with metal ions in buffered media were extended to systems containing Al. Caps buffer was used to establish the pH. The procedures used are described and the results are given. Preliminary studies were initiated as to the feasibility of adding a slowly dissolving solid compound of the additive to the glass-water system to maintain a supply of dissolved additive. It appears that several magnesium compounds have a suitable combination of solubility and affinity towards silicate glass surfaces to have a pronounced retarding effect on the extraction of uranium from the glass. These preliminary findings raise the possibility that introducing a magnesium source into geologic repositories for nuclear waste glass in the form of a sparingly soluble Mg-based backfill material may cause a substantial reduction in the extent of long-term glass corrosion. The studies described also provide mechanistic understanding of the roles of various metal solutes in the leachant. Such understanding forms the basis for developing long-term predictions of nuclear waste glass durability under repository conditions. From what is known about natural highly reduced glasses such as tektites, it is clear that iron is dissolved as ferrous iron with little or no ferric iron. The reducing conditions were high enough to cause metallic iron to exsolve out of the glass in the form of submicroscopic spherules. As the nuclear waste glass is much less reduced, a study was initiated on other natural glasses in addition to the nuclear waste glass. Extensive measurements were carried out on these glasses in order to characterize their magnetic properties. Results of these studies are described.

Thorpe, Arthur N.↗

NASA NTRS1976

Space disposal of nuclear wastes. Volume 1: Socio-political aspects

The history and interpretation of radioactive waste management in the U.S., criteria for choosing from various options for waste disposal, and the impact of nuclear power growth from 1975 to 2000 are discussed. Preconditions for the existence of high level wastes in a form suitable for space disposal are explored. The role of the NASA space shuttle program in the space disposal of nuclear wastes, and the impact on program management, resources and regulation are examined.

Laporte, T.↗

DOE OSTI2019

Structural Characterization of Ternary Salt Melts for Low Activity Waste Applications

Reactions of alkali salts (nitrates, sulfates, carbonates, halides, borates) play a key role in the low temperature feed conversion occurring at the cold cap during processing of Hanford Low Activity Waste (LAW) glass melters. An alkali salt phase can sometimes form, and preferentially incorporate radionuclides of Cs, Cl, I, and Tc. During melting of the slurry feed, some of the feed components sequentially break down with increasing temperature to form gases (i.e., nitrates → NO x , carbonates → CO 2 , and boric acid → H 2 O) or partially volatilize (halides). Sulfate, however, tends not to volatilize but has limited solubility in the final borosilicate glass waste form. To improve understanding of these low temperature processes and their composition dependencies, a scoping study was undertaken to synthesize salt systems that remain amorphous at room temperature, thus facilitating structural study. Melts of equimolar ratios of K2SO4-ZnSO4 (a known ionic glass-forming system) with added nitrates, halides, or carbonates, were melted and quenched. Some of the materials formed single phase glasses and some underwent crystallization upon quenching. Characterization of these quenched materials by thermal analysis, infrared absorption, and diffraction was performed. Addition of other anions to the sulfate base glass resulted in a distortion of the sulfate tetrahedron, as evidenced by infrared absorption. Carbonates strongly promoted crystallization, mostly of carbonate phases. Nitrates promoted crystallization of ZnO, and the nitrate volatilized with some incorporating into the glass. Here, halides tended to incorporate into the glass, but the small (F) and large (I) halogens promoted crystallization of sulfate-containing crystals, while moderate sized (Cl) halogens produced single-phase ionic glasses.

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DOE OSTI2025

Foam and Unreacted Material Reduction in Magnesium Oxysulfate for SRPPF Waste Solidification

A magnesium oxysulfate (MOS) grout formulation has been proposed to solidify the liquid effluent from the Savannah River Plutonium Processing Facility (SRPPF) Aqueous Recovery System (ARS). This formulation uses a combination of light-burnt MgO, anhydrous MgSO 4 , and dead-burnt MgO resulting in a grout with good mixability and acceptable density. The setting time for the grout is within the operational limits of the SRPPF facility and the leachate pH is ~9.4, which is within the assumed pH range of the brine from the Waste Isolation Pilot Plant (WIPP). When first tested at the 1-gallon small-scale and 55-gallon full-scale, there was a significant foam layer that raised concerns of a nonhomogeneous final form. A nonhomogeneous mixture could have variable density throughout the waste form and therefore allow for unequal shielding. To reduce the potential for unequal shielding in the final form, a series of foam reduction tests were carried out and a final formulation of 28 wt% light burnt MgO, 10 wt% MgSO 4 , and 62 wt% dead burnt MgO was used to remove the foam in under two hours while still maintaining a mix that has an acceptable density, mix time, peak temperature, and set time.

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DOE OSTI2014

Application of evolved gas analysis to cold-cap reactions of melter feeds for nuclear waste vitrification

The melter feed (a mixture of nuclear waste and glass forming and modifying additives) experiences multiple gas-evolving reactions in an electrical glass-melting furnace during the vitrification of nuclear wastes. Foams from the residual gases can significantly alter the melting rate through mass and heat transfers. Here, we employed the thermogravimetry-gas chromatography-mass spectrometry (TGA-GC-MS) combination to perform quantitative evolved gas analysis (EGA) and developed a simple calibration model which correlates the overall mass loss rate with the evolution rates for individual gases. The model parameters are obtained from the least squares analysis, assuming that the gas-evolving reactions are independent. Finally, the EGA adds the ‘chemical identity’ to the reactions indicated by the ‘phenomenological’ kinetic model.

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DOE OSTI2017

33 Shafts Category of Transuranic Waste Stored Below Ground within Area G

This report compiles information to support the evaluation of alternatives and analysis of regulatory paths forward for the 33 shafts. The historical information includes a form completed by waste generators for each waste package (Reference 6) that included a waste description, estimates of Pu-239 and uranium-235 (U-235) based on an accounting technique, and calculations of mixed fission products (MFP) based on radiation measurements. A 1979 letter and questionnaire (Reference 7) provides information on waste packaging of hot cell waste and the configuration of disposal shafts as storage in the 33 Shafts was initiated. Tables of data by waste package were developed during a review of historical documents that was performed in 2005 (Reference 8). Radiological data was coupled with material-type data to estimate the initial isotopic content of each waste package and an Oak Ridge National Laboratory computer code was used to calculate 2009 decay levels. Other sources of information include a waste disposal logbook for the 33 shafts (Reference 9), reports that summarize remote-handled waste generated at the CMR facility (Reference 10) and placement of waste in the 33 shafts (Reference 11), a report on decommissioning of the LAMPRE reactor (Reference 12), interviews with an employee and manager involved in placing waste in the 33 shafts (References 13 and 14), an interview with a long-time LANL employee involved in waste operations (Reference 15), a 2002 plan for disposition of remote-handled TRU waste (Reference 16), and photographs obtained during field surveys of several shafts in 2007. The WIPP Central Characterization Project (CCP) completed an Acceptable Knowledge (AK) summary report for 16 canisters of remote-handled waste from the CMR Facility that contains information relevant to the 33 Shafts on hot-cell operations and timeline (Reference 17).

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DOE OSTI2021

Applications of Strain-Coupled Magnetoelectric Composites

This article deals with research, development and future directions of magnetoelectric composites for practical devices applications. In the past 20 years there has been a surge of research in the area of multiferroics (MF) and magnetoelectrics (ME) due to their potential to replace existing technologies based only on ferroelectric or ferromagnetic materials. Some of the magnetoelectric composites show exceptionally high potential in the area of magnetic field sensors, however, work remains before commercialization can be realized. The cross coupling among various ferroic parameters in magnetoelectric composites is several orders higher than single phase magetoelectrics, which make its favorable for low detection (nT or pT) magnetic field sensors. Advances in both layered structures or controlled three-dimensional matrix composites for applications as magnetoelectric nonvolatile memory elements are both required. Robust cross-coupling among various parameters with more than four logic states and its compatibility with complementary-symmetry metal–oxide–semiconductor (CMOS) technology are the main requirements for heterostructure magnetoelectric thin films. The major hurdles in the area of magnetoelectric nonvolatile memory elements are poor interfacial properties and weak magnetoelectric coupling for high density fast read and write processes. Another potential area is strained coupled magneto-electric composites where magnetostriction mediated dimensional change in the magnetic layer effectively modulate the change in the dimension of piezoelectric layers via piezostriction, which leads to a strong ME coupling where their coupling magnitude is sufficient for magnetic field sensors. Energy harvesters based on magnetoelectric composites are also intriguing concepts to capture various types of waste energy, in the form mechanical vibration, pressure, wind energy, hydrothermal and waste temperature.

36 MATERIALS SCIENCE↗