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At least 289 records · Page 16
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

An Investigation on the Sustainability of a Sodium-cooled Breed-and-Burn Fast Reactor (B and BR)

Fast reactor can effectively handle the used nuclear fuel (UNF) by transmuting long-living transuranic (TRUs) into relatively short-living fission products. It can also be used to produce fissile from fertile nuclides as fuels for another reactor. The breed-and-burn fast reactor (B and BR) is a unique concept of fast reactor, which can breed the fissile fuels and use the bred fuels in situ. Thanks to this characteristic, the fuel utilization in a B and BR can be extremely high and even the recycled fuel of a B and BR can be used as a fuel to spawn the next generation B and BR. The UNF from the first B and BR should be reprocessed or reconditioned so that it can be reused in the second B and BR. A fuel reprocessing technology named melt refining was used in EBR-II from 1964 to 1969 and it was also applied in our previous study which showed that second B and BR can be started using the recycled UNF of first B and BR. However, about 95% of rare earths are removed during the melt refining and the associated processes are not very simple. Moreover, the actinide loss can also be significant. To enhance the proliferation-resistance and economy of the reprocessing, a super-simplified melt and treatment (SSMT) process without removing rare earth nuclides is suggested and the feasibility study to start the second B and BR using SSMT is carried out in this paper. The neutronics simulations were done using the Monte Carlo code McCARD in conjunction with ENDF/B-VII cross-section library. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Used Nuclear Fuel Management in Asia

Used fuel discharged from nuclear reactors are presently stored in wet pools or in dry cask at-reactor and/or away-from- reactor sites. Only a few countries reprocess their civilian used fuel, notably France, Russia, and UK. No country has yet operated a geologic repository to dispose of civilian used fuel or high-level radioactive wastes (HLW). As a result, the management and disposal of used fuel and HLW are viewed as the Achilles' heel of nuclear power. Figure 1 shows the used fuel inventory in the United States (US) and East Asian nuclear power programs. The inventory is estimated based on that in 2010 and the projected growth of nuclear power in these programs to 2020 and 2030. Figure 1 indicates that in 2010, the US had about 63,900 tons of used fuel from light-water reactors (LWR), more than the statutory limit of what was supposed to go into the now-abandoned Yucca Mountain Repository, and accounting for more than 33% of global used LWR fuel total. Currently, about 2,000 tons of used fuel are generated in the existing US LWRs annually. Russia had 25,100 tons of used fuel in 2010. Based on a modest growth of nuclear power, its used fuel inventory is expected to increase to about 31,800 and 38,000 tons by 2020 and 2030, respectively. The nuclear energy programs in East Asia (China, Japan, ROK, and Taiwan) account for 19% of global used LWR fuel inventory in 2010. China's used fuel inventory in 2010 was relatively small at 1,900 tons, but is expected to grow rapidly to {approx}12,400 and {approx}31,000 tons in 2020 and 2030, respectively, based on its expanding nuclear power program (from the current installed capacity of 21.8 GW to {approx}60 and {approx}200 GW in 2020 and 2030, respectively). Besides PWRs, China also operates 2 Candu reactors which each produces roughly 8 times more used fuel in tons than a PWR of the same power. The used Candu fuel inventory would increase from 950 tons in 2010 to 2,850 and 4,750 tons in 2020 and 2030, respectively. Used fuel in Japan, ROK, and Taiwan are US-obligated. Japan's future trends of used fuel inventory are uncertain due to the unknown numbers and timing of the currently-shutdown nuclear power plants that would be allowed for restart, and the uncertain future of Japan's reprocessing at Rokkasho. Under two scenario cases (a no-nuke/no-reprocessing and a 80%-nuke with reprocessing), the Japanese used fuel inventory could be capped from 2020 to 2030 to {approx}21,500 tons for the first case, and to {approx}22,500 tons for the second. The ROK's used PWR fuel inventory would increase from 4,900 tons in 2010 to about 8,800 and 14,000 tons in 2020 and 2030, respectively. Its used Candu fuel inventory would also increase from 5,900 tons in 2010 to about 9,700 and 13,500 tons in 2020 and 2030, respectively. The Taiwan used fuel inventory should be kept at current level as nuclear power is being phase out. Its growth should be modest from 3,000 tons in 2010 to about 4,300 and 5,000 tons in 2020 and 2030, respectively even if the 4. nuclear power station would start and the operation of existing LWRs is extended. Used fuel inventory in other East Asian countries (DPRK, Vietnam, etc.) should be small, if any before 2030.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Dynamic Diketoenamine Crosslinking Unlocks Vinyl Polymer Vitrimers From β ‐Triketone Chemistry

Covalent adaptable networks (CANs) offer a compelling strategy to unite the mechanical robustness of thermosets with the reprocessability of thermoplastics, yet achieving simultaneous durability, processability, and true recyclability remains challenging. We introduce diketoenamine (DKE) vitrimers derived from β-triketone methacrylate monomers and demonstrate how rational monomer design dictates network processability, viscoelasticity, and recyclability. By systematically varying the spacer length between the β-triketone (TK) moiety and the polymer backbone, we identify a key structure–property relationship that dictates vitrimer behavior. Networks bearing TK pendants minimally displaced from the backbone suppress creep but exhibit limited stress relaxation, whereas extended spacers yield lower glass transition temperatures, higher effective crosslink densities, and efficient stress dissipation, enabling optical transparency and reprocessability. Extending this platform to ultra-high molecular-weight prepolymers introduces physical entanglements as secondary crosslinks, further enhancing dimensional stability without compromising processability. Both mechanical and chemical recycling validate the closed-loop circularity of these materials. Furthermore, these results establish TK methacrylates as a versatile platform for designing high-performance vitrimers that integrate durability, reprocessability, and true closed-loop recyclability.

closed-loop recyclability↗
DOE OSTI2023

Pertechnetate/perrhenate–capped Zr/Hf–Dihydroxide Dimers: Elucidating Zr–TcO 4 Co–Mobility in the Nuclear Fuel Cycle

Spent nuclear fuel contains heavy element fission products that must be separated for effective reprocessing for a safe and sustainable nuclear fuel cycle. 93 Zr and 99 Tc are high-yield fission products that co-transport in liquid-liquid extraction processes. Here we seek atomic-level information of this co-extraction process, as well as fundamental knowledge about Zr IV (and Hf IV ) aqueous speciation in the presence of topology-directing ligands such as pertechnetate (TcO 4 – ) and non-radioactive surrogate perrhenate (ReO 4 – ). In this context, we show that the flat tetrameric oxyhydroxyl-cluster [M IV 4 (OH) 8 (H 2 O) 16 ] 8+ (and related polymers) is dissociated by perrhenate/pertechnetate to yield isostructural dimers, M 2 (OH) 2 (XO 4 – ) 6 (H 2 O) 6 • 3H 2 O (M=Zr/Hf IV ; X=Re/Tc VII ), elucidated by single-crystal X-ray diffraction. We used these model compounds to understand the pervasive 93 Zr- 99 Tc coextraction with further speciation studies in water, nitric acid, and tetrabutylphosphate (TBP) -kerosene; where the latter two media are relevant to nuclear fuel reprocessing. SAXS (small angle X-ray scattering), compositional evaluation, and where experimentally feasible, ESI-MS (electrospray ionization mass spectrometry) showed that perrhenate/pertechnetate influence Zr/Hf IV -speciation in water. Here, in Zr-XO 4 solvent extraction studies to simulate fuel reprocessing, we provide evidence that TcO 4 – enhances extraction of Zr IV , and compositional analysis of the extracted metal-complexes (Zr-ReO 4 study) is consistent with the crystallized Zr IV 2 (OH) 2 (Re VII O 4 – ) 6 (H 2 O) 6 •dimer.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2024

Improving the Recyclability of Polymer Composites With Cellulose Nanofibrils

Cellulose nanofibers (CNFs) have been widely studied for their reinforcing potential in high-performance composites. While there are numerous publications on CNF-reinforced composites in a variety of polymer matrices, few have considered the recyclability of such thermoplastic composites and whether the incorporation of CNFs deteriorates or improves their performance upon reprocessing. In this study, two thermoplastic resins, poly(lactic acid) (PLA), and glycol-modified polyethylene terephthalate (PETg), were prepared with CNF reinforcement and thermomechanically recycled to investigate the effect of CNF inclusion on the composite properties after reprocessing as well as their effect on the composites’ number of useful life cycles. Additionally, changes in mechanical, thermal, rheological, molecular, and microstructural properties of the composites and/or base resins were monitored as a function of cycle numbers. As is typical, the polymers’ molecular weight and mechanical performance deteriorated with continued processing. However, the addition of spray dried CNF was found to better maintain the mechanical performance of both polymers throughout multiple recycling steps as compared to neat samples. For example, the tensile strength of PETg with 20 wt% CNF after 6 processing cycles was found to exceed that of virgin neat PETg, and higher loadings of CNF were found to preserve a higher yield strength during multiple rounds of reprocessing compared to PETg composites with lower CNF loadings. Ultimately this study indicates that the addition of CNF to some thermoplastic materials can increase both their sustainability by offsetting the use of high-embodied energy resins and their circularity by enabling performance retention over more use cycles.

42 ENGINEERING↗
DOE OSTI2017

Elemental source attribution signatures for calcium ammonium nitrate (CAN) fertilizers used in homemade explosives

Calcium ammonium nitrate (CAN) is a widely available fertilizer composed of ammonium nitrate mixed with some form of calcium carbonate such as limestone or dolomite. CAN is also frequently used to make homemade explosives. The potential of using elemental profiling and chemometrics to match both pristine and reprocessed CAN fertilizers to their factories for use in future forensic investigations was examined. Inductively coupled plasma-mass spectrometry (ICP-MS) analysis was performed on 64 elements in 125 samples from 11 CAN stocks from 6 different CAN factories. Fisher ratio, degree-of-class-separation, and partial least squares discriminant analysis (PLSDA) were used to develop a model using the concentrations of Na, V, Mn, Cu, Ga, Sr, Ba and U to classify a validation set of CAN samples into 5 factory groups; one group was two factories from the same fertilizer company. In terms of the pristine CAN samples, i.e., unadulterated prills, 64% of the test samples were matched to their correct factory group with zero false positives. The same PLSDA model was used to correctly match 100% of the CAN samples that were reprocessed by crushing and mixing the CAN with powdered sugar. In the case of crushed CAN samples mixed with aluminum powder, correct matches were made for zero to 100% of the samples depending on the factory the CAN originated. Remarkably, for one factory, 100% of the ammonium nitrate samples that were extracted from CAN using tap or bottled water were matched to the correct CAN factory group. Lastly, the water-insoluble (calcium carbonate) portions of CAN provided a greater degree of discrimination between factories than the water-soluble portions of CAN. In summary, this work illustrates that sourcing unadulterated CAN fertilizer can potentially be done with high frequency and high confidence using elemental profiling and chemometrics while the sourcing of reprocessed CAN is dependent on how much an adulterant alters the recovered elemental profile of CAN.

chemical forensics↗
DOE OSTI2016

Metal–organic framework with optimally selective xenon adsorption and separation

Nuclear energy is considered among the most viable alternatives to our current fossil fuel based energy economy.1 The mass-deployment of nuclear energy as an emissions-free source requires the reprocessing of used nuclear fuel to mitigate the waste.2 One of the major concerns with reprocessing used nuclear fuel is the release of volatile radionuclides such as Xe and Kr. The most mature process for removing these radionuclides is energy- and capital-intensive cryogenic distillation. Alternatively, porous materials such as metal-organic frameworks (MOFs) have demonstrated the ability to selectively adsorb Xe and Kr at ambient conditions.3-8 High-throughput computational screening of large databases of porous materials has identified a calcium-based nanoporous MOF, SBMOF-1, as the most selective for Xe over Kr.9,10 Here, we affirm this prediction and report that SBMOF-1 exhibits by far the highest Xe adsorption capacity and a remarkable Xe/Kr selectivity under relevant nuclear reprocessing conditions. The exceptional selectivity of SBMOF-1 is attributed to its pore size tailored to Xe and its dense wall of atoms that constructs a binding site with a high affinity for Xe, as evident by single crystal X-ray diffraction and molecular simulation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Application of sensitivity analysis in DYMOND/Dakota to fuel cycle transition scenarios

The ability to perform sensitivity analysis has been enabled for the nuclear fuel cycle simulator DYMOND through its coupling with the design and analysis toolkit Dakota. To test and demonstrate these new capabilities, a transition scenario and multi-parameter study were devised. The transition scenario represents a partial transition from the US nuclear fleet to a closed fuel cycle with small modular LWRs and fast reactors fueled by reprocessed used nuclear fuel. Four uncertain parameters in this transition were studied – start date of reprocessing, total reprocessing capacity, the nuclear energy demand growth, and the rate at which the fast reactors are deployed – with respect to their impact on four response metrics. The responses – total natural uranium consumed, maximum annual enrichment capacity required, total disposed mass, and total cost of the nuclear fuel cycle – were chosen based on measures known to be of interest in transition scenarios and to be significantly impacted by the varying parameters. Furthermore, analysis of this study was performed both from the direct sampling and through surrogate models developed in Dakota to calculate the global sensitivity measures Sobol’ indices. This example application of this new capability showed that the most consequential parameter to most metrics was the share of new build capacity that is fast reactors. However, for the cost metric, the scaling factor of the energy demand growth was significant and had synergistic behavior with the fast reactor new build share.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Extraction of Neptunium, Plutonium, Americium, Zirconium, and Technetium by Di-(2-Ethylhexyl)- Iso -Butyramide (DEH i BA) at High Metal Loadings

Increased focus on carbon neutral energy has generated a resurgence of interest in nuclear power, and in particular advanced reactors which are likely to utilize high assay low enriched uranium (HALEU). This in turn could increase the economic attractiveness of recovering still partially enriched uranium from used nuclear fuel. Concomitant to development of advanced reactors, advanced reprocessing schemes should be developed which address the disadvantages to well established reprocessing schemes. The present study focuses on using di-(2-ethylhexyl)-iso-butyramide (DEHiBA) under high metal loading conditions for the reprocessing of used nuclear fuel. The elements examined in the study include the dominant transuranic actinides (Np, Pu, Am) as well as the often-problematic Tc and Zr. Further, by increasing the concentration of the extractant from the more commonly reported 1.0 M – 1.5 M, the extraction of hexavalent actinides is substantially increased, while maintaining effective rejection of tri, tetra, and pentavalent actinides, particularly in the presence of high loadings of uranium. In conclusion, the extraction of Zr by 1.5 M DEHiBA is noted to be negligible by comparison to tributyl phosphate (TBP), however the coextraction of Tc with U is observed to be nominally twice the quantity that is extracted by TBP indicating a need for effective Tc management.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Isotope anomalies in the FE-group elements in meteorites and connections to nucleosynthesis in AGB stars

We study the effects of neutron captures in AGB stars on “Fe-group” elements, with an emphasis on Cr, Fe, and Ni. These elements show anomalies in {sup 54}Cr, {sup 58}Fe, and {sup 64}Ni in solar system materials, which are commonly attributed to supernovae (SNe). However, as large fractions of the interstellar medium (ISM) were reprocessed in AGB stars, these elements were reprocessed, too. We calculate the effects of such reprocessing on Cr, Fe, and Ni through 1.5 M{sub ⊙} and 3 M{sub ⊙} AGB models, adopting solar and 1/3 solar metallicities. All cases produce excesses of {sup 54}Cr, {sup 58}Fe, and {sup 64}Ni, while the other isotopes are little altered; hence, the observations may be explained by AGB processing. The results are robust and not dependent on the detailed initial isotopic composition. Consequences for other “Fe group” elements are then explored. They include {sup 50}Ti excesses and some production of {sup 46,47,49}Ti. In many circumstellar condensates, Ti quantitatively reflects these effects of AGB neutron captures. Scatter in the data results from small variations (granularity) in the isotopic composition of the local ISM. For Si, the main effects are instead due to variations in the local ISM from different SN sources. The problem of Ca is discussed, particularly with regard to {sup 48}Ca. The measured data are usually represented assuming terrestrial values for {sup 42}Ca/{sup 44}Ca. Materials processed in AGB stars or sources with variable initial {sup 42}Ca/{sup 44}Ca ratios can give apparent {sup 48}Ca excesses/deficiencies, attributed to SNe. The broader issue of galactic chemical evolution is also discussed in view of the isotopic granularity in the ISM.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2024

A recyclable self-healing composite with advanced sensing property

Polymer-based composites frequently encounter damage, often lurking beneath the surface and proving challenges to their early detection and repair. While material-based sensors show promise for encoding self-sensing properties within these composites, their in situ healing and reprocessability remain significant challenges. Therefore, the overarching goal of this study is the creation of a reprocessable polymeric composite encoded with self-healing attributes and the ability to autonomously sense damage. At the core of this innovation are vitrimers, a polymeric material characterized by a covalently adaptive dynamic network responsive to external factors such as heat. They combine thermoset-like resilience with thermoplastic-like flowability on demand under external stimuli. We nanoengineer a polyester-based vitrimeric polymer by incorporating piezoresistive carbon nanotubes (CNTs) as reinforcing elements that not only enhance its mechanical strength but also create a percolation network within the composite, thereby enabling piezoresistive self-sensing properties, all the while preserving the intrinsic self-healing capabilities offered by the vitrimeric matrix. The fabrication process of the composite involves a solvent-free in situ polymerization method that combines epoxy and anhydride-containing monomers with ~ 0.1 wt.% of CNTs. Once it was established that the introduction of CNTs into the polymeric matrix did not compromise the mechanical properties of the composite, their strain-sensing properties were characterized by applying cyclic loading while measuring their electrical resistance. Strikingly, CNT-enhanced vitrimer composite consistently retains its mechanical and sensing properties through repeated cycles of reshaping and reprocessing, underscoring its potential as a robust distributed strain sensor. This polyester-based vitrimeric composite is also easily recyclable without harsh chemical treatments. Preliminary findings from this study conclusively demonstrate that the bulk composite boasts both self-sensing capabilities and in situ detect healing properties, charting a promising course towards the development of a mechanically resilient multifunctional composite that seamlessly integrates selfsensing and healing capabilities.

Rohewal, Sargun Singh↗
DOE OSTI2019

Non-Aqueous Electrochemical Fluorination of Used Nuclear Fuel as an Advanced Separation Process

Development of efficient and environmentally benign methods to reprocess used nuclear fuel (UNF) will enable technologies for a nuclear renaissance. Electrochemical and fluorination methods for reprocessing UNF have been proposed, but combinations of electrochemical and fluorination methods have not been investigated. Electrochemical fluorination may reduce waste volumes compared to the main large-scale aqueous methods for processing used nuclear fuel. A non-aqueous electrochemical fluorination reprocessing method has been developed and demonstrated that enables gas phase uranium recovery while allowing for potential control of the reaction using a single process. Thermodynamic modeling showed a galvanic reaction between UNF and a fluorinating agent, such as NF 3 , in a molten fluoride electrolyte was possible and could selectively fluorinate U to UF 6 . To verify the model results, a reactor system for electrochemical fluorination was constructed along with vessels for product collection and analysis. Initial trials of the electrochemical reaction were performed and characterized by several electrochemical methods. The electrochemical oxidation of U using NF 3 as an oxidizing agent was demonstrated and UF 6 as a reaction product was detected.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Utah FORGE: Seismic Reflection Data

This is 2D and 3D seismic reflection data from Utah FORGE reprocessed during Phase 2c. The readme file containing an explanation of the data including data formats, software that can be used, processing, and projection and datum used. The Reprocessing document gives the rationale for reprocessing and shows examples of the improvements that were obtained. For all 3D and 2D data the following datasets were created and output in SEG-Y format: - Unmigrated Time - Prestack Time Migration (PSTM), Unenhanced (UnEnh) and Enhanced (Enh) - Prestack Depth Migration (PSDM), Unenhanced (UnEnh) and Enhanced (Enh) - Velocity Model used for Migration

15 GEOTHERMAL ENERGY↗
DOE OSTI2024

Hestia SW-IFL Onroad Fossil Fuel Carbon Dioxide (FFCO2) product: Road segment-level annual FFCO2 emissions across Arizona (2017-2022), version 1.1

The SW-IFL onroad fossil fuel carbon dioxide (FFCO2) emissions data product represents CO2 emissions from the combustion of fossil fuels by motor vehicles (e.g., passenger cars, trucks, buses, motorcycles) traveling on designated roadways. The emissions are represented geographically on each road segment within the state of Arizona spanning the 2017 to 2022 time period. This data product was developed as part of the Southwest Urban Corridor Integrated Field Laboratory (SW-IFL) project, which aims to provide new knowledge and tools that address urban environmental issues by integrating high-resolution observations, modeling, and civic engagement. The emissions data are provided in CSV (input data, ONR_FFCO2_AZ_county.csv) and GeoPackage form (output polyline objects - about 786,000 road segments, XXXX_AZ_v1.1.gpkg) designated by road class (interstates, arterials, collectors, local). The metadata file (Metadata_SW-IFL_Onroad_annualFFCO2_v1.1.docx) provides details about attributes and data formats. The GeoPackage emissions data are provided separately for local roads and nonlocal roads (interstates, arterials, collectors). The method file (Methods_SW-IFL_Onroad_annualFFCO2_v1.1.docx) describes the data processing flow and data sources. Update on 2024-04-17: Updates were made to both the input emission data file (.csv) and output segment-level emission file (.gpkg). There was an update in county-level emission input data (ONR_FFCO2_AZ_county.csv) and the entire road segments were reprocessed to reflect this update.Update on 2024-04-29: Update was made to one output segment-level emission file (Nonlocal_AZ_v1.0.gpkg). There was an error in the AADT values and the data were reprocessed to reflect this update.Update on 2024-10-22: Temporal coverage was extended to include 2022. VMT values were recalculated using new AADT data and the entire road segments were reprocessed to reflect these updates.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2017

Synthesis and Characterization of 5- and 6- Coordinated Alkali Pertechnetates

The local chemistry of technetium-99 ( 99 Tc) in oxide glasses is important for understanding the incorporation and long-term release of Tc from nuclear waste glasses, both those for legacy defense wastes and fuel reprocessing wastes. Tc preferably forms Tc(VII), Tc(IV), or Tc(0) in glass, depending on the level of reduction of the melt. Tc(VII) in oxide glasses is normally assumed to be isolated pertechnetate TcO 4 - anions surrounded by alkali, but can occasionally precipitate as alkali pertechnetate salts such as KTcO 4 and NaTcO 4 when Tc concentration is high. In these cases, Tc(VII) is 4-coordinated by oxygen. A reinvestigation of the chemistry of alkali-technetium-oxides formed under oxidizing conditions and at temperatures used to prepare nuclear waste glasses showed that higher coordinated alkali Tc(VII) oxide species had been reported, including those with the TcO 5 - and TcO 6 - anions. The chemistry of alkali Tc(VII) and other alkali-Tc-oxides is reviewed, along with relevant synthesis conditions. Additionally, we report attempts to make 5- and 6-coordinate pertechnetate compounds of K, Na, and Li, i.e. TcO 5 - and TcO 6 -. It was found that higher coordinated species are very sensitive to water, and easily decompose into their respective pertechnetates. It was difficult to obtain pure compounds, but mixtures of the pertechnetate and other phase(s) were frequently found, as evidenced by x-ray absorption spectroscopy (XAS), neutron diffraction (ND), and Raman spectroscopy. Low temperature electron paramagnetic resonance (EPR) measurements showed the possibility of Tc(IV) and Tc(VI) in Na 3 TcO 5 and Na 5 TcO 6 compounds. Here, it was hypothesized that the smaller counter cation would result in more stable pertechnetates. To confirm the synthesis method, LiReO 4 and Li 5 ReO 6 were prepared, and their Raman spectra match those in the literature. Subsequently, the Tc versions LiTcO 4 and Li 5 TcO 6 were synthesized and characterized by ND, Raman spectroscopy, XANES, and EXAFS. The Li 5 TcO 6 was a marginally stable compound that appears to have the same structure as that known for Li 5 ReO 6 . Implications of the experimental work on stability of alkali technetate compounds and possible role in the volatilization of Tc are discussed.

36 MATERIALS SCIENCE↗
DOE OSTI2014

SRNL CRP progress report [Development of Melt Processed Ceramics for Nuclear Waste Immobilization]

A multi-phase ceramic waste form is being developed at the Savannah River National Laboratory (SRNL) for treatment of secondary waste streams generated by reprocessing commercial spent nuclear. The envisioned waste stream contains a mixture of transition, alkali, alkaline earth, and lanthanide metals. Ceramic waste forms are tailored (engineered) to incorporate waste components as part of their crystal structure based on knowledge from naturally found minerals containing radioactive and non-radioactive species similar to the radionuclides of concern in wastes from fuel reprocessing. The ability to tailor ceramics to mimic naturally occurring crystals substantiates the long term stability of such crystals (ceramics) over geologic timescales of interest for nuclear waste immobilization [1]. A durable multiphase ceramic waste form tailored to incorporate all the waste components has the potential to broaden the available disposal options and thus minimize the storage and disposal costs associated with aqueous reprocessing.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Melt Processed Crystalline Ceramic Waste Forms for Advanced Nuclear Fuel Cycles: CRP T21027 1813: Processing technologies for high level waste, formulation of matrices and characterization of waste forms, Task 17208 (Final Report)

A multi-phase ceramic waste form is being developed at the Savannah River National Laboratory (SRNL) for treatment of secondary waste streams generated by reprocessing commercial spent nuclear. The envisioned waste stream contains a mixture of transition, alkali, alkaline earth, and lanthanide metals. Ceramic waste forms are tailored (engineered) to incorporate waste components as part of their crystal structure based on knowledge from naturally found minerals containing radioactive and non-radioactive species similar to the radionuclides of concern in wastes from fuel reprocessing. The ability to tailor ceramics to mimic naturally occurring crystals substantiates the long term stability of such crystals (ceramics) over geologic timescales of interest for nuclear waste immobilization [1]. A durable multi-phase ceramic waste form tailored to incorporate all the waste components has the potential to broaden the available disposal options and thus minimize the storage and disposal costs associated with aqueous reprocessing. This report summarizes results from three years of work on the IAEA Coordinated Research Project on “Processing technologies for high level waste, formulation of matrices and characterization of waste forms” (T21027), and specific task “Melt Processed Crystalline Ceramic Waste Forms for Advanced Nuclear Fuel Cycles” (17208).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Melt Processed Crystalline Ceramic Waste Forms for Advanced Nuclear Fuel Cycles CRP T21027 1813: Processing technologies for high level waste, formulation of matrices and characterization of waste forms, Task 17208 (Final Report)

A multi-phase ceramic waste form is being developed at the Savannah River National Laboratory (SRNL) for treatment of secondary waste streams generated by reprocessing commercial spent nuclear. The envisioned waste stream contains a mixture of transition, alkali, alkaline earth, and lanthanide metals. Ceramic waste forms are tailored (engineered) to incorporate waste components as part of their crystal structure based on knowledge from naturally found minerals containing radioactive and non-radioactive species similar to the radionuclides of concern in wastes from fuel reprocessing. The ability to tailor ceramics to mimic naturally occurring crystals substantiates the long term stability of such crystals (ceramics) over geologic timescales of interest for nuclear waste immobilization. A durable multi-phase ceramic waste form tailored to incorporate all the waste components has the potential to broaden the available disposal options and thus minimize the storage and disposal costs associated with aqueous reprocessing. This report summarizes results from three years of work on the IAEA Coordinated Research Project on “Processing technologies for high level waste, formulation of matrices and characterization of waste forms” (T21027), and specific task “Melt Processed Crystalline Ceramic Waste Forms for Advanced Nuclear Fuel Cycles” (17208).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗