Fabrication of radioactive and non-radioactive titanate and zirconate ceramics for immobilization of
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Reiser, Joelle T.↗
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Preliminary IHLW Formulation Algorithm Description
This report documents the initial algorithm that could be used by the Waste Treatment and Immobilization Plant (WTP) in batching high-level waste (HLW) and glass-forming chemicals (GFCs) in the HLW melter feed preparation vessel (MFPV) (HFP-VSL-00001 and -00005). Not all Hanford tank waste can be accommodated by the models developed for this report and significant expansion of the model boundaries could be achievable to reduce the WTP mission life and total canister production count. The immobilized HLW (IHLW) must meet a series of constraints to be acceptable for disposal in the Monitored Geologic Repository, which are contained in the Specification 1 of the Contract (DOE 2000), the Waste Acceptance Product Specifications (WAPS, DOE 1996), and the Waste Acceptance System Requirements Document (WASRD, DOE 2007). The IHLW Waste Form Compliance Plan (WCP, 24590-HLW-PL-RT-07-0001, Rev 3) specifies that the formulation algorithm will be developed and used to comply with the constraints associated with glass composition and properties. This report is not an engineering calculation, does not provide design input, and is not an engineering study. Algorithm inputs include the chemical analyses of the blended HLW in the HLW blend vessel (HBV) (HLP-VSL-00028, the volume and composition of the MFPV heel, the volume and composition of the MFPV after waste addition, the volume and composition of MFPV batch after GFC addition, the compositions of individual GFCs, and the mass of glass in each canister. In addition to these inputs, uncertainties in the HLW composition and processing parameters are included in the algorithm. Using the above inputs, the algorithm calculates the following outputs: 1) the volume of HLW to be transferred from the HBV to the MFPV, 2) the mass of each GFC for addition to the MFPV, 3) the composition of the glass that will be produced along with uncertainties, and 4) the predicted properties, with associated uncertainties, of the resulting IHLW. The algorithm uses the property-composition models to calculate properties with associated uncertainties and compares them with various constraints to ensure that a processable feed is formulated and a compliant IHLW is produced. The GFC additions are determined using an optimization approach to provide high confidence that the HLW glass will meet all product quality requirements and key processing constraints. For most HLW batches there are many possible glass compositions that meet all constraints. In these cases, the glass composition is optimized for a series of target component concentrations and target property values. The algorithm also incorporates process measurement and product quality uncertainties, based on the work of Piepel et al. (2005). Estimates of the various process and measurement uncertainties that affect glass compositions and predicted glass properties have been previously reported (Piepel et al. 2005, 2006) and the impacts of these estimated uncertainties on the IHLW composition envelope that meets product quality and processing-related properties with sufficient confidence were evaluated. The details of work performed to date to develop this initial GFC addition and batching algorithm are summarized in Sections 4 and 5. An example data set is used to illustrate the calculations of the algorithm summarized in Section 6. Finally, in Section 7, there is a statement of the required work to achieve a final operational IHLW formulation control algorithm. This report is not an engineering calculation, does not provide design input, and is not an engineering study.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Molten Flux Synthesis of Plutonium (IV) Silicates K 2 PuSi 2 O 7 and Rb 2 PuSi 6 O 15
Few plutonium containing silicates have been reported in the literature and only one other tetravalent plutonium silicate structure has been reported using the flux growth method. With the goal of understanding actinide incorporation into silicates, we describe herein two new plutonium (IV) silicates, Rb 2 PuSi 6 O 15 and K 2 PuSi 2 O 7 , synthesized using a mixed alkali chloride/alkali fluoride flux growth. Both structures crystallize in the C2/c space group with lattice parameters a=24.2980(10) Å, b=7.1466(3) Å, c=17.2025(8) Å and $\blacksquare$96.6560(10)° for Rb 2 PuSi 6 O 15 , and a=9.9478(2) Å, b=5.59820 (10) Å, c=13.1848(2) Å, and $\blacksquare$105.7400(10)° for K 2 PuSi 2 O 7 . Comparisons to the synthesis of other reported alkali actinide silicates (Ce, U, and Th) are made. Raman spectra and DFT calculations of formation enthalpies and vibrational modes provide confirmation of structure refinements and enhanced product characterization for comparison with other materials.
37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
A GPU Accelerated Mixed‐Precision Finite Difference Informed Random Walker (FDiRW) Solver for Strongly Inhomogeneous Diffusion Problems
In nature, many complex multi‐physics coupling problems exhibit significant diffusivity inhomogeneity, where one process occurs several orders of magnitude faster than others temporally. Simulating rapid diffusion alongside slower processes demands intensive computational resources due to the necessity for small time steps. To address these computational challenges, we have developed an efficient numerical solver named Finite Difference informed Random Walker (FDiRW). In this study, we propose a GPU‐accelerated, mixed‐precision configuration for the FDiRW solver to maximize efficiency through GPU multi‐threaded parallel computation and lower precision computation. Numerical evaluation results reveal that the proposed GPU‐accelerated mixed‐precision FDiRW solver can achieve a 117× speedup over the CPU baseline, while an additional 1.75× speedup is achieved by employing lower precision GPU computation. Notably, for large model sizes, the GPU‐accelerated mixed‐precision FDiRW solver demonstrates strong scaling with the number of nodes used in simulation. When simulating radionuclide absorption processes by porous wasteform particles with a medium‐sized model of 192 × 192 × 192, this approach reduces the total computational time to 10 min, enabling the simulation of larger systems with strongly inhomogeneous diffusivity.
97 MATHEMATICS AND COMPUTING↗
Chapter 10 - Thermal Energy Harvesting Using Thermomagnetic Effect
Low-grade thermal energy, either in the form of waste heat or nature heat, is available abundantly around us. Since Carnot efficiency reduces with decrease in operating temperature difference, there are very limited practical options for low-grade thermal energy conversion. Thermomagnetic effect is the effect of heat on magnetic properties of a material. This phenomenon can be used to generate electricity using either direct thermal to electrical energy conversion or indirect thermal to mechanical energy conversion. This chapter provides a broad overview of various aspects of thermomagnetic energy conversion. We first explain the working principle behind the thermomagnetic energy harvesting. We then formulate mathematical equations needed to calculate the work output and energy conversion efficiency of a thermomagnetic cycle. Based on this formulation, we provide a list of working materials that can be employed in the thermomagnetic harvesters. At last, we describe various thermomagnetic harvesters proposed in the literature.
ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATION↗
Chemical durability of multicomponent lanthanide zirconate solid solutions
Multicomponent zirconate solid solutions A 2 Zr 2 O 7 (A= (Sm,Yb), (Sm,Yb,Gd), (Sm,Yb,Gd,Er), and (Sm,Yb,Gd,Er,Dy)) were synthesized through spark plasma sintering and their chemical durability was investigated via semi-dynamic leaching tests in pH = 1 nitric acid at 90°C. Incongruent dissolution occurred for all samples with the preferential elemental release of zirconium in all A 2 Zr 2 O 7 except for high entropy (Sm,Yb,Gd,Er,Dy)Zr 2 O 7 which displays a congruent release of the A-site elements and Zr. A strong positive correlation was found between the lanthanide elemental release rates and the A-site cation size disorder, whereas a negative correlation was observed between the elemental release rates and mixing entropy.
36 MATERIALS SCIENCE↗
Thermodynamics of CeSiO 4 : Implications for Actinide Orthosilicates
Zircon (ZrSiO 4 : I4 1 /amd) can accommodate actinides, such as thorium, uranium, and plutonium. The zircon structure has been determined for several of the end member compositions of other actinides, such as plutonium and neptunium. However, the thermodynamic properties of these actinide zircon structure-types are largely unknown due to the difficulties in synthesizing these materials and handling transuranium actinides. Thus, we have completed a thermodynamic study of cerium orthosilicate, stetindite (CeSiO 4 ), a surrogate of PuSiO 4 . For the first time, the standard enthalpy of formation of CeSiO 4 was obtained by high temperature oxide melt solution calorimetry to be -1971.9 ± 3.6 kJ/mol. Stetindite is energetically metastable with respect to CeO 2 and SiO 2 by 27.5 ± 3.1 kJ/mol. The metastability explains the rarity of the natural occurrence of stetindite and the difficulty of its synthesis. Applying the obtained enthalpy of formation of CeSiO 4 from this work, along with those previously reported for USiO 4 and ThSiO 4 , we developed an empirical energetic relation for actinide orthosilicates. Here, the predicted enthalpies of formation of AnSiO 4 are then made with a discussion of future strategies to efficiently immobilize Pu or minor actinides in the zircon structure.
37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
High-Temperature Thermodynamics of Cerium Silicates, A-Ce 2 Si 2 O 7 , and Ce 4.67 (SiO 4 ) 3 O
Lanthanide disilicates and oxyapatites have potential roles in high temperature applications as thermal (TBC) and environmental barrier coatings (EBC), or possible alteration phases in geological nuclear waste repositories. However, those Ce 3+ -bearing silicates have only been limitedly studied. In this work, we performed detailed structural and thermodynamic investigations on A-Ce 2 Si 2 O 7 (tetragonal, P4 1 ) and Ce 4.67 (SiO 4 ) 3 O (hexagonal, P6 3 /m). The high temperature structural behaviors and coefficients of thermal expansion were determined by in situ high temperature synchrotron X-ray diffraction (HT-XRD) implemented with Rietveld analysis and thermogravimetric analysis coupled with differential scanning calorimetry (TGA-DSC). A-Ce 2 Si 2 O 7 was found to be stable in N 2 and air up to ~1483 K with anisotropic thermal expansion along the a and c axes (α a = 12.3 × 10 -6 K -1 and α c = 12.4 × 10 -6 K -1 ). Ce 4.67 (SiO 4 ) 3 O had a slow partial oxidation between 533 K and 873 K to a new nonstoichiometric phase Ce 3+ 1.67- xCe4 + x Ce 3+ 3 (SiO 4 ) 3 O 1+0.5x , followed by a thermal decomposition to CeO 2 and SiO 2 at ~1000 K in air. By using high temperature oxide melt solution calorimetry at 973 K with lead borate as the solvent, the standard enthalpy of formation was determined for A-Ce 2 Si 2 O 7 (-3825.1 ± 6.0 kJ/mol) and Ce 4.67 (SiO 4 ) 3 O (-7391.3 ± 9.5 kJ/mol). Finally, these thermodynamic parameters were compared with those of CeO 2 , CeSiO 4 , and other silicate oxyapatites for examining their chemical stability in high temperature environments relevant for aeronautical applications, mineral formation, and nuclear fuel cycle.
36 MATERIALS SCIENCE↗
Dechlorination Apparatus for Treating Chloride Salt Wastes: System Evaluation and Scale-Up
Not Available
37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
Influence of Metal Ion Complexation on the Radiolytic Longevity of Butyramide Extractants under Direct Dissolution Process Conditions
The direct dissolution of voloxidized used nuclear fuel (UNF) into an organic solution–comprised of diluent and specialized extractants–poses a promising alternative to the traditional liquid–liquid solvent extraction approach to reprocessing UNF. However, moving to direct dissolution removes the presence of a concentrated nitric acid aqueous phase, which has been shown to significantly influence the radiolytic longevity of extractants in conventional extraction flowsheets. Given the limited knowledge of radiation effects under direct dissolution conditions, here we present a time-resolved and dose-accumulation study on the impact of direct dissolution conditions on the radiolytic longevity of two candidate butyramide extractants, N,N-di(2-ethylhexyl) butyramide (DEHBA) and N,N-di(2-ethylhexyl)isobutyramide (DEHiBA), in pre-equilibrated n-dodecane solvent in the presence and absence of process-relevant metal ions, specifically, uranium and rhenium. Loss G(DEHBA) and G(DEHiBA) values were found to be comparable to each other, with an average of 0.37 ± 0.02 μmol J –1 , and to previous data from the γ irradiation of DEHBA and DEHiBA under conventional solvent extraction conditions. Rhenium, and by extension technetium, extraction had a modest decrease (~10%) in the overall radiolytic stability of DEHiBA only, despite >2× observed increases in chemical kinetic reactivity of the corresponding complexes with the n-dodecane radical cation. Uranium loading, on the other hand, significantly improved the lifetime of both ligands (>30%) under γ irradiation, with a greater stabilization observed for DEHBA over DEHiBA. The observed radioprotective effect afforded by uranium loading is fortuitous for the longevity of direct dissolution solvent.
38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
Neutron Total Scattering Analysis of Materials for the Nuclear Fuel Cycle
Here, this paper reviews recent experimental efforts at the University of Tennessee and Oak Ridge National Laboratory to comprehensively characterize the structural details of materials relevant for the nuclear fuel cycle by employing advanced neutron scattering techniques. For the study of nuclear ceramics, neutron scattering offers distinct advantages over traditional laboratory or synchrotron X-ray diffraction, including enhanced sensitivity to elements with a low atomic mass, such as oxygen, nitrogen, and carbon. The key to these efforts is the recent advancement in the neutron scattering infrastructure at the high-flux diffractometers at the Spallation Neutron Source. The high neutron flux at these instruments enables neutron total scattering, a nondestructive bulk technique that simultaneously captures both short-range structural effects through pair distribution function analysis and long-range order through diffraction pattern analysis. This approach is particularly important for a comprehensive description of defective, disordered, or amorphous nuclear materials. The case studies presented here include analyses of the local defect structure in hyperstoichiometric uranium oxides and short-range order of ion-irradiated ceramics. This advanced analytical methodology will improve our understanding of the behavior of materials in extreme environments and contribute to the development of more resilient nuclear materials.
Neutron scattering↗
Nanoscale TiO 2 coating improves water stability of Cs 2 SnCl 6
To improve the stability of Cs 2 SnCl 6 under aqueous/moisture environments, we applied a concept of artificial passivation by depositing a protective TiO 2 coating of 10 nm on the surface of Cs 2 SnCl 6 . Static leaching experiments results indicate that the initial release rates of Cs + and Cl - are decreased by 20–30 times with TiO 2 coating, suggesting its possibility to improve the short-term water/environmental stability of Cs 2 SnCl 6 . An amorphous-to-crystalline phase transition in TiO 2 film was observed, possibly resulting in degradation of Cs 2 SnCl 6 . However, the crystalline TiO 2 film still remains after 21 days water exposure and can still act as an effective passivation layer to reduce the release rates of Cs + and Cl - by as much as about 17 and 7 times, respectively, relative to static leaching without artificial coatings. Therefore, the water/environmental stability of metal halide perovskite Cs 2 SnCl 6 , which is a highly soluble molecular salt, can be enhanced by the nanoscale TiO 2 coating as an artificial passivation film.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Updated Medical Waste Management Plans and Emergency Action Plans for the Lawrence Livermore National Laboratory Main Site
Attached are the updated Medical Waste Management Plans and Emergency Action Plans for Lawrence Livermore National Laboratory’s (LLNL) Biosafety Level 2 (BSL-2) facilities and Biosafety Level 3 (BSL-3) facility. A redline version of each document has also been provided. Minor changes such as facility personnel, contact information, and titles for those that support onsite Biosafety Labs at LLNL have been made to the document. It should be noted that the most recent versions of the Medical Waste Large Quantity Generator Registration Application Form and Medical Waste Management Plan were incorporated and sourced from the Alameda County Department of Environmental Health website.
54 ENVIRONMENTAL SCIENCES↗
CFA-17-12798: Nanostructured Ceramic Membranes for Enhanced Tritium Management
This project focuses on the novel use of interfacial hydrated layers in ionic conducting ceramic materials for tritiated water (HTO) adsorption and recovery. Low-temperature water adsorption properties of nanoscale structured ceramics combined with hydrogen isotope exchange in their hydrated layers could provide a low-cost and transformative avenue to address tritium management challenges.
08 HYDROGEN↗
Effects of Borate and Organics on U(VI) Solubility in WIPP Brine
The solubility of uranium (VI) in Waste Isolation Pilot Plant (WIPP)-relevant brine was determined to support ongoing WIPP recertification activities. This research was performed by the Los Alamos National Laboratory Carlsbad Operations (LANL-CO) Actinide Chemistry and Repository Science Program (ACRSP). The WIPP Actinide Source Term Program (ASTP) did not develop a model for the solubility of actinides in the VI oxidation state. The solubility of UO 2 2+ , in the absence of WIPP specific data, is presently set to be equal to a conservatively high 1 mM within the WIPP Performance Assessment (PA) for all expected WIPP conditions (SOTERM, 2019) as selected at the recommendation of the Environment Protection Agency (EPA) (EPA, 2005). According to the current WIPP chemistry model assumptions and conditions, the expected pC H+ is about 9.5 and controlled by MgO buffering CO 3 2- . The goal of this study is to perform screening experiments that account for the contributions of organics and borate on uranium solubility. In this report, the solubility of U(VI) was determined at pC H+ 9 WIPP brine in the absence or presence of borate and organics at under-saturation approach. Experiments were equilibrated for about 135 days. Organic compounds present in WIPP waste can form strong complexes with actinides and can affect the oxidation states of actinides. The organic compounds addressed in WIPP performance assessment include EDTA (Ethylenediaminetetraacetic Acid), oxalate, citrate, and acetate (SOTERM, 2019). These data quantify the effects of WIPP-relevant concentrations of borate and organics effects on the solubility of U(VI) to challenge the predictions of the WIPP actinide model and inform decisions and recommendations made in the upcoming recertification of the WIPP (CRA-2024). The experiments performed were done according to the U.S. Department of Energy (DOE) approved Test Plan entitled “Experimental Strategy to Challenge Actinide Solubility Predictions” and designated LCO-ACP-26. All data reported were obtained under the LANL-CO Quality Assurance Program, which is compliant with the DOE Carlsbad Field Office, Quality Assurance Program Document (CBFO/QAPD) (QAPD, 2017).
37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
Canister Centerline Cooling Experiments for DPF5-336 Reference Material Made with ERV3b Salt Simulant
This report provides experimental details and results on the elemental distributions (compositions), crystalline/amorphous phase distribution, and microstructures for DPF5-336 reference materials made with ERV3b salt that were subjected to the canister centerline cooling (CCC) heat treatment process. To fabricate these materials, ammonium dihydrogen phosphate was mixed with ERV3b salt simulant and a dechlorination procedure was run in the generation-2 dechlorinator at the Pacific Northwest National Laboratory up to 600°C in an alumina crucible. After the dechlorination process, the product was removed from the crucible, lightly crushed, loaded into a new alumina crucible, Fe 2 O 3 was added, and this was vitrified at 1100°C for 1 hour and quenched. The quenched material was mostly amorphous but showed some small diffraction peaks attributed to Li 3 Fe 2 (PO 4 ) 3 . This quenched material was ground to a fine particle size in a milling chamber, and this was subjected to the CCC cooling profile in duplicate within alumina and fused quartz crucibles. The CCC experiment is designed to simulate the slowest cooling profile (the vertical centerline) of a 0.61-m diameter canister where the slower cooling profile often results in the production of a variety of different crystalline phases upon formation of which result in a residual glass of unknown composition and properties. The information gleaned from these types of experiments will lead to improved composition-property predictions for materials with similar compositions. The results of the alumina and fused quartz experiments show different microstructures but similar diffraction patterns providing evidence of similar crystalline phases being present in each sample. Both samples had monazite [i.e., (Ce,Nd)PO 4 ] and mixtures of phases containing P, Fe, and alkali metal elements. This report completes the milestone M4FT-23PN030104042 with details provided in Appendix B.
36 MATERIALS SCIENCE↗
Trash to Treasure: Waste Carbon into Construction Materials
Mineralization of industrial waste and earth-abundant CO2 into value-added products is a powerful method of utilizing domestic resources to produce value from waste. Carbonate minerals formed from Ca and Mg rich waste streams are key constituents in cement and concrete, which are in demand at the gigaton-scale every year. While thermodynamically favorable, CO2 mineralization is hindered by slow kinetics, and traditional thermochemical routes are too expensive to scale. Here, we demonstrate the advantages of an electrochemical method of waste valorization, which eliminates this limitation and improves kinetics. We highlight this method's ability to initiate crystallization under favorable local conditions, producing carbonates from mining waste at mildly acidic pH. We highlight the effects of key constituents in mining waste on the mineralization process.
29 ENERGY PLANNING, POLICY, AND ECONOMY↗
Performance Comparison of LPG and Gasoline in an Engine Configured for EGR-Loop Catalytic Reforming
In prior work, the EGR loop catalytic reforming strategy developed by ORNL has been shown to provide a relative brake engine efficiency increase of more than 6% by minimizing the thermodynamic expense of the reforming processes, and in some cases achieving thermochemical recuperation (TCR), a form of waste heat recovery where waste heat is converted to usable chemical energy. In doing so, the EGR dilution limit was extended beyond 35% under stoichiometric conditions. In this investigation, a Microlith®-based metal-supported reforming catalyst (developed by Precision Combustion, Inc. (PCI)) was used to reform the parent fuel in a thermodynamically efficient manner into products rich in H2 and CO. We were able to expand the speed and load ranges relative to previous investigations: from 1,500 to 2,500 rpm, and from 2 to 14 bar break mean effective pressure (BMEP). Experiments were conducted to determine the effects of the H/C ratio of the fuel on H2 production and on the engine efficiency in order to compare E10 gasoline (H/C = 1.95) and liquified petroleum gas (LPG), comprised primarily of propane (H/C = 2.67). Additionally, the compression ratio of the engine was increased to ascertain whether further efficiency improvements could be realized based on a reduced knock propensity of the dilute EGR mixture with the reformed fuel. Both the gasoline and propane reforming strategies provided efficiency gains up to 1.85 percentage points and further efficiency improvements with the increased compression ratio were realized. The fuel specific effects of gasoline vs. LPG, the effect of engine operating condition on reforming, and knock limits of the reformed mixture are discussed in detail.
33 ADVANCED PROPULSION SYSTEMS↗