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Influence of Nanoconfinement and Elevated Temperatures on Geocolloid-Facilitated Transport of Energy-related Contaminants

The main objective of this project is to obtain a better understanding of the transport behavior and interactions of geocolloids in the presence of energy- related contaminants under bulk and nanoconfined conditions. In that, we first established solid protocols to synthesize and fabricate numerous types of geocolloids with a. various sizes ranging from 40 – 800 nm and b. different hydrophile-lipophile balance (HLB) ratios in the range of 25:75 to 75:25. Direct force measurements with geocolloids having different degrees of surface coverages (i.e. HLB to realistically mimic adsorption of energy related contaminants) were conducted using the Surface Forces Apparatus (SFA) over a distance regime starting from 8μm all the way down to molecular contact. Repulsive forces were observed on approach starting from > 3μm, followed by an exponential increase of which magnitude appears to be larger than a decay length obtained from Derjaguin–Landau–Verwey–Overbeek (DLVO) theory in pure water. When the geocolloids were confined in salted water, the magnitude of onset of repulsion was varied as a function of salinity in solution, which can significantly alter a purely repulsive screened electrostatic (coulombic) interaction arising from, among geocolloids as well as between geocolloids and geosurfaces. The viscosity and flow characteristics of geocolloidal suspensions at different degrees of confinement were also investigated where we identified highly discontinuous rheological behaviors below a critical nanoconfinement level. We anticipate that the knowledge gained through this study will enable the scientists and researchers to better assess transport and fate behaviors of geocolloidal dispersions that can carry energy-related contaminants under realistically emulated geosystem.

03 NATURAL GAS↗

Microstructure of In-Core Molten Salt Corrosion Hastelloy N{sup R} and 316 Stainless Steel

Since 2011, many materials have been discussed for potential application in Fluoride salt-cooled High-temperature nuclear Reactors (FHRs), which are emerging as leading reactor concept for the next generation of nuclear reactors thanks to the benefits of a high degree of passive safety, high thermal efficiency, and low spent fuel volume. Among these discussed materials, Hastelloy N{sup R} and 316L stainless steel have been selected as the most likely candidate alloys for FHRs. To provide direct experience in integrating a real system using molten FLiBe salt for reactor operation, these two candidate alloys were experimentally tested in Li-7 enriched Li{sub 2}BeF{sub 4} (FLiBe) molten salt in the core of MIT Research Reactor (MITR) at 700 deg. C for 1000 hours. After in-core corrosion testing, some basic post-irradiation examinations (PIE) including weight change per unit area and surface profile as well as surface morphology were conducted in the MIT Nuclear Reactor Laboratory (NRL) and compared with out-of-reactor corrosion samples. The deep understanding of these alloys in high-temperature molten salt with strong neutron irradiation requires further microstructural analysis such as electron microscopy and energy dispersive spectroscopy on the surface and cross-section of the in-core corrosion tested samples. The challenge in the microstructural analysis is due to the high radioactivity from the gamma emitting isotopes formed in alloys during neutron irradiation. Additionally, the facilities that are authorized for radioactive materials are very limited on campus. In this study, the microstructure of the in-core corrosion tested Hastelloy N{sup R} and 316L stainless steel specimen was analyzed by using the facilities at the Center for Advanced Energy Studies (CAES) at Idaho National Laboratory (INL) through the Nuclear Science User Facilities (NSUF) Rapid Turnaround Experiment (RTE) program.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DECOVALEX-2023 Task F Specification (Rev. 9)

This report is the revised (Revision 9) Task F specification for DECOVALEX-2023. Task F is a comparison of the models and methods used in deep geologic repository performance assessment. The task proposes to develop a reference case for a mined repository in a fractured crystalline host rock (Task F1) and a reference case for a mined repository in a salt formation (Task F2). Teams may choose to participate in the comparison for either or both reference cases. For each reference case, a common set of conceptual models and parameters describing features, events, and processes that impact performance will be given, and teams will be responsible for determining how best to implement and couple the models. The comparison will be conducted in stages, beginning with a comparison of key outputs of individual process models, followed by a comparison of a single deterministic simulation of the full reference case, and moving on to uncertainty propagation and uncertainty and sensitivity analysis. This report provides background information, a summary of the proposed reference cases, and a staged plan for the analysis.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Solvent Hold Tank Sample Results for MCU-19-2-3-4 (January 2019), MCU-19-83-84-85 (February 2019), MCU-19-208-209-210 (March 2019) and MCU-19-138-139-140 (March special trim): Quarterly Report

A trend summary of three sets of Solvent Hold Tank (SHT) monthly samples (MCU-19-2-3-4, MCU-19-83-84-85, MCU-19-208-209-210); and a set from a special trim (MCU-19-138-139-140) are reported. Most of the conclusions are based on the March SHT sample (MCU-18-208-209-210). Analyses of the March SHT sample (MCU-19-208-209-210) indicated that the Modifier (Cs-7SB) and the Extractant (MaxCalix) concentrations were below their nominal recommended concentrations (169,000 mg/L and 46,900 mg/L, respectively) by 3% and 6%, respectively. The Suppressor (N,N’,N”–tris(3,7-dimethyloctyl)guanidine or TiDG) concentration has increased since the January 2019 measurement to 1141 mg/L, which is above the minimum recommended concentration (479 mg/L). These new concentrations are consistent with the concentrations observed when MCU was fully operational in 2016. The Semi-Volatile Organic Analysis (SVOA) did not detect any impurities. However, the Fourier Transform Hydrogen Nuclear Magnetic Resonance (FT-HNMR) analysis detected presence of phthalates at a few ppm. The impurity concentration was highest in the January sample and consistently detected in the February and March samples (but always registered as a few ppm or less). Another impurity observed in the samples was mercury. Based on the March SHT sample, up to 25 ± 5 micrograms of mercury per mL of solvent was detected (as the average of the Direct Mercury Analysis (DMA) and X-Ray Fluorescence (XRF) methods). The higher mercury concentration in the solvent (as determined in the last three-monthly samples) is possibly due to the higher mercury concentration in Salt Batches 9 and 10. The gamma concentration (~2.55E5 dpm/mL) measured in the March 2019 SHT samples was consistent with previous values observed when MCU was operating continuously (for example, between December 2016 and January 2017). Statistical analysis of the different analytical methods detected a bias (a statistical difference at 5% level) between the FT-HNMR and HPLC methods when measuring the MaxCalix and Modifier concentration; the FT-HNMR gave a higher concentration possibly due to peak overlapping in the measured data. A bias was also detected when measuring total mercury by the DMA method versus the XRF method with the XRF method reported higher mercury concentrations. The laboratory will continue to monitor the quality of the solvent for any new impurities or degradation of the solvent components.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DECOVALEX-2023: Task F Specification (Revision 10)

This report is the revised (Revision 10) Task F specification for DECOVALEX-2023. Task F is a comparison of the models and methods used in deep geologic repository performance assessment. The task proposes to develop a reference case for a mined repository in a fractured crystalline host rock (Task F1) and a reference case for a mined repository in a salt formation (Task F2). Teams may choose to participate in the comparison for either or both reference cases. For each reference case, a common set of conceptual models and parameters describing features, events, and processes that impact performance will be given, and teams will be responsible for determining how best to implement and couple the models. The comparison will be conducted in stages, beginning with a comparison of key outputs of individual process models, followed by a comparison of a single deterministic simulation of the full reference case, and moving on to uncertainty propagation and uncertainty and sensitivity analysis. This report provides background information, a summary of the proposed reference cases, and a staged plan for the analysis.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Integrated Wash-Aid, Treatment, and Emergency Reuse System (IWATERS) for Strontium Contaminations - 17390

The Integrated Wash-Aid, Treatment, and Emergency Reuse System (IWATERS) is being developed to mitigate the release of radioactivity to the urban environment by disseminating, collecting, and processing decontamination wash water. Previously, IWATERS was designed for the treatment of surfaces contaminated with radioactive cesium. Experiments were performed to investigate the expansion of IWATERS for strontium contamination. Bench-scale tests were performed to investigate various counter-ions and chelators to promote ion exchange of strontium from building surfaces. Desorption results for Cs, Sr, and Eu from concrete aggregate and concrete, brick, and asphalt coupons using low- and high-pressure flow with Ca{sup 2+} and Ba{sup 2+} rich wash water are reported along with those from non-porous surfaces such as glass and vinyl siding. Several sorbents were tested for their ability to specifically separate Sr{sup 2+} from these high ionic strength wash waters. Alkali metal and alkaline earth salts were found to be effective counter-ions for decontamination of Sr{sup 2+}, which may facilitate the use of common sorbents, even if their K{sub d} values are <100 mL/g. In practice, low K{sub d} values require the use of large quantities of sorbent. Computer simulations were run using the GoldSim Contaminant Fate Module software package to optimize the efficiency of IWATERS for strontium and account for the technical limitations of the sorbents. This modeling allows greater flexibility to design the wash water recycling system for various end-user specifications including options from rain barrel designs for treating runoff from roofs and facades to larger options like drainage basins for larger-scale wash down activities. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Pilot Plant Test of Removal of Co-60 and Cs-137 from the Evaporator Concentrates of Forsmark unit 3 - 17158

A pilot plant test has been conducted to treat evaporator concentrates at Forsmark BWR NPP unit 3 in Sweden using Fortum's selective ion exchanger technology. The concentrate contains an array of radionuclides of which C-60 and Cs-137 are radiologically the most important ones. Concentrate also contained a high salt content ∼ 20 g/l and high organic carbon loading ∼ 1 g/l The pilot system consisted of pre-treatment (mechanical filters, activated carbon, UV oxidation) and Cs and Co selective ion exchangers. CsTreat ion exchanger was very effective in removing Cs from the evaporator concentrate: {sup 137}Cs activity was lowered from the initial value 280 MBq/m{sup 3} to below detection limit 550 Bq/m{sup 3}. Removing Co was more complicated: results suggest that Co is bound to organic chelating compounds, thus fairly compete oxidation of organic carbon was required in order to achieve good ion exchange results. Finally, after total organic carbon (TOC) was treated to below 50 ppm, {sup 60}Co activity was lowered from the initial value of 57 MBq/m{sup 3} to 2.7 kBq/m{sup 3}. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Solvent Hold Tank Sample Results for MCU-19-366- 367-368 (April 2019), MCU-19-469-470-471 (May 2019), MCU-19-524-525-526 (June 2019): Quarterly Report

A trend summary of three sets of Solvent Hold Tank (SHT) monthly samples MCU-19-366-367-368 (April), MCU-19-469-470-471 (May), and MCU-19-524-525-526 (June) are reported. Most of the conclusions are based on the June SHT sample (MCU-19-524-525-526). Analyses of the June SHT sample (MCU-19-524- 525-526) indicated that the Modifier (Cs-7SB) and the Extractant (MaxCalix) concentrations were below their nominal recommended concentrations (169,000 mg/L and 46,900 mg/L respectively) by 4% and 2% respectively but consistent with previously observed levels during normal runs. The Suppressor (N,N’,N”– tris(3,7-dimethyloctyl)guanidine or TiDG) concentration has decreased since the March 2019 measurement to 793 mg/L, which is above the minimum recommended concentration (479 mg/L). The Semi-Volatile Organic Analysis (SVOA) did not detect any impurities. However, the Fourier Transform Hydrogen Nuclear Magnetic Resonance (FT-HNMR) analysis detected presence of phthalates at a few ppm. The impurity concentration was highest in the April sample and consistently detected in the May and June samples (but always registered as a few ppm or less). Another impurity observed in the samples was mercury. Based on the June SHT sample, up to 25 ± 5 micrograms of mercury per mL of solvent was detected (as the average of the Direct Mercury Analysis (DMA) and X-Ray Fluorescence (XRF) methods). The higher mercury concentration in the solvent (as determined in the last three-monthly samples) is possibly due to the higher mercury concentration in Salt Batches 9 and 10. The gamma concentration (~2.90E4 dpm/mL) measured in the June 2019 SHT samples was consistent with previous values observed when the Modular Caustic-Side Solvent Extraction Unit (MCU) was operating continuously (for example, between December 2016 and January 2017). Statistical analysis of the different analytical methods detected a bias between the FT-HNMR and the High Performance Liquid Chromatography (HPLC) methods when measuring the MaxCalix and Modifier concentrations; the FT-HNMR gave a higher concentration possibly due to peak overlapping in the measured data. A bias was detected when measuring total mercury by the DMA method versus the XRF method with the XRF method reported higher mercury concentrations. The laboratory will continue to monitor the quality of the solvent for any new impurities or degradation of the solvent components.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Dry Deactivation of Sodium Metal in a Molten LiCl-KCl-CsCl-NaCl System

An experimental study was performed with the objective of investigating and characterizing a dry technique for deactivation of sodium metal in a molten salt system. The study was performed in three parts. First, proof-of-principle testing of the technique was performed at bench scale. It involved loading and melting tens of grams of clean sodium metal atop a pool of LiCl-KCl-CsCl eutectic at ∼300°C and then adding ammonium chloride particles while mixing. The ammonium chloride reacted with sodium to form sodium chloride and nitrogen/hydrogen off-gases. The sodium chloride assimilated into the salt pool, forming a quaternary salt mixture of LiCl-KCl-CsCl-NaCl. The proof-of-principle testing repeatedly exhibited complete deactivation of the loaded sodium metal. Second, characterization of a LiCl-KCl-CsCl-NaCl system was performed using differential scanning calorimetry to produce a partial phase diagram of LiCl-KCl-CsCl eutectic versus NaCl, which identified the liquidus, solidus, and two-phase regions of the quaternary system. Third, the dry deactivation technique was demonstrated at kg-scale in an inert atmosphere radiological glovebox with sodium metal that was previously separated in the same glovebox from uranium metal in unirradiated blanket elements for the Fermi-1 nuclear reactor. The quaternary salt product at the end of the demonstration was sampled and showed complete deactivation of the sodium metal. Here, in short, this study qualified a technique to completely deactivate batches of sodium metal in the absence of air or water. While this study focused on the deactivation of sodium metal, including bond sodium from unirradiated blanket elements, the technique is applicable to other sources of sodium metal, such as sodium metal from batteries. Furthermore, the technique could also be extended to other alkali metal systems, including lithium, sodium, potassium, rubidium, cesium, and mixtures thereof.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Results from Accelerated Environmental Testing of Barcoded Identifiers for UF 6 Cylinders

This report provides the qualitative and quantitative results from the accelerated environmental testing campaign. A total of 444 samples were subjected to 10 tests that included UV, salt fog, temperature, blowing dust, high pressure and temperature water, and impact exposure. Each sample included a two-dimensional barcode and an alphanumeric string representing the cylinder identification number. When evaluating the results, primary attention was placed on the decodability of the barcodes using a barcode verifier device, since machine scanning a barcode offers the greatest efficiency and accuracy gains when reading cylinder identification numbers during safeguards activities. Contrast was considered the critical test parameter because maintaining contrast under a wide range of environmental conditions is of fundamental importance to decoding the identification number of the barcodes. Additionally, manual readability of the alphanumeric string was considered an important factor in determining whether an inspector or operator could still read the identification number of the cylinder if the barcode proved to be undecodable.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Transforming E-Waste Into Strategic Resources: Techno-Economic Analysis of Gallium and By-products Recovery from LEDs via Bioleaching

The growing demand for gallium in optoelectronics and renewable energy applications raises concerns about supply security and production sustainability. This study evaluates the techno-economic feasibility of recovering gallium and by-products (copper and nickel) from waste GaN-based LEDs via bioleaching. A process flowsheet encompassing transportation, robotic disassembly, ball milling, bioleaching, solvent extraction/electrowinning, and refining was modeled. Based on mass balance analysis, more than 53 tons of LED waste are required annually to yield 1 kg of gallium alongside substantial copper and nickel co-products. Preliminary techno-economic analysis (TEA) shows an average total cost (ATC) of 6.84 USD/kg metal when costs are allocated by mass-weighted economic value (market price) fraction, corresponding to 6.75 USD/kg for copper, 15.91 USD/kg for nickel, and 470.95 USD/kg for gallium. For gallium, direct operational costs account for more than 70% of the cost. Monte Carlo simulations further quantify cost uncertainties under market price fluctuations. This work represents the first TEA of gallium recovery from GaN-based LEDs and highlights potential pathways for future cost reduction.

36 MATERIALS SCIENCE↗

Analysis of Tank 30H (HTF-30-19-33, -34), Tank 32H (HTF-32-19-35, -36), and Tank 37H (HTF-37-19-53, -54) Samples for Support of the Evaporator Feed Qualification and Corrosion Control Programs for the 3H-Evaporator

SRNL analyzed samples from Tank 30H, Tank 32H, and Tank 37H to support the Evaporator Feed Qualification and Corrosion Control Programs for the 3H-Evaporator system. The six samples from Tanks 30H, 32H, and 37H all contain highly concentrated salt solutions with sodium concentrations ranging from 13.8 M to 16.9 M. Hydroxide is the dominant anion in all of the samples followed by nitrate and nitrite. The preponderance of the analytical determinations (density, anions, cations) on all three tanks show slightly higher concentrations of solute in the sub-surface sample than in the surface sample, indicating some stratification within the tanks. With the exception of the Tank 30H VDS, all of the samples are somewhat more concentrated compared with the compositions of the previous samples. The sum of the major cations versus the sum of the major anions show a difference of <5% for all of the samples indicating good data quality for the non-radioactive analytes in the table. The silicon concentrations measured in the VDS and surface samples from Tank 30H, Tank 32H, and Tank 37H were all below detection.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Analysis of Tank 30H (HTF-30-19-33, -34), Tank 32H (HTF-32-19-35, -36), and Tank 37H (HTF-37- 19-53, -54) Samples for Support of the Evaporator Feed Qualification and Corrosion Control Programs for the 3H-Evaporator

SRNL analyzed samples from Tank 30H, Tank 32H, and Tank 37H to support the Evaporator Feed Qualification and Corrosion Control Programs for the 3H-Evaporator system. The six samples from Tanks 30H, 32H, and 37H all contain highly concentrated salt solutions with sodium concentrations ranging from 13.8 M to 16.9 M. Hydroxide is the dominant anion in all of the samples followed by nitrate and nitrite. The preponderance of the analytical determinations (density, anions, cations) on all three tanks show slightly higher concentrations of solute in the sub-surface sample than in the surface sample, indicating some stratification within the tanks. With the exception of the Tank 30H VDS, all of the samples are somewhat more concentrated compared with the compositions of the previous samples. The sum of the major cations versus the sum of the major anions show a difference of <5% for all of the samples indicating good data quality for the non-radioactive analytes in the table. The silicon concentrations measured in the VDS and surface samples from Tank 30H, Tank 32H, and Tank 37H were all below detection.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

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↗

Assessing I 2 S-LWR economic competitiveness using systematic differential capital cost evaluation methodology

One of the major factors impeding nuclear power plant (NPP) construction is their economics. In this study, a systematic differential economics evaluation approach was developed through the use of the Code of Accounts guidelines to assess the costs of nuclear power plants. The methodology, entirely based on publicly available data, may serve as a template to evaluate direct costs for reactors of any size and design at any stage of developments. In particular, this approach was used to assess the costs of the Integral Inherently Safe Light Water Reactor (I 2 S-LWR). The I 2 S-LWR is a design concept of a large (~1000 MWe) light water reactor. One of its key design features promoting inherent safety is implementation of an integral primary circuit configuration, which however necessitates a compact design of the core and primary circuit components. Through the methodology here presented, a representative loop PWR design was taken as a reference and the differential cost was estimated for each individual account based on the design difference, or similarity. Cost scaling techniques were applied to the accounts representing systems that differ from the ones of the reference PWR. Cost estimating techniques were used to evaluate cost of innovative components that are not part of standard PWR designs. By evaluating the cost difference of the I 2 S-LWR from the standard PWR, rather than the absolute cost, the uncertainty in the estimate is reduced. A similar approach was used by ORNL to estimate the cost of a Fluoride-salt High-temperature Reactor (FHR). A traditional four-loop PWR plant with a core thermal power of 3417 MWth (1144 MWe) was selected as the reference. Costs for that plant were prepared in 1978 by the Department of Energy (DOE) Energy Economics Data Base (EEDB), averaging actual cost incurred in the construction of several nuclear power plants (NPP), itemized with a great level of detail according to the Code of Accounts. This best estimate costs are denoted PWR12-BE. For each account, the cost of equipment, site labor and site material is provided. Industry experts at Westinghouse Electric Company performed a “sanity check” of the cost items, adjusting the cost of several items to match the current market and supply chain data. The detailed cost assessment of I 2 S-LWR was performed, systematically analyzing cost for each account, and applying the differential economics approach. First, Relative importance of each account, i.e., its contribution to the total cost was established, to help focus analysis on the most significant contributors. Moreover, the accounts describing components that are different than that of the PWR12-BE were identified. The integral configuration of the reactor has important implications on the cost of the reactor plant equipment (accounts 22x). Turbine generator equipment (Accounts 23x) is not believed to be much different than that of the reference design. I 2 S-LWR structures (Accounts 21x) mainly differ from that of a standard LWR as several buildings (containment building, shield building, annex building, waste processing building, fuel storage building) are integrated into a single building (nuclear island). Yardwork has a higher cost for the I 2 S-LWR, as the NI is partially below grade. On the other hand, due to its compact Nuclear Island footprint, I 2 S-LWR facilitates (and includes in its reference version) the use of seismic isolators, which contribute to reducing the direct cost. The total capital investment cost (TCIC), on the $/kW basis, of I 2 S-LWR is 5.84% lower than that of PWR12-BE, in spite of the lower power output of I 2 S-LWR. However, for the Western US (0.7 g), benefits of the seismic isolation are more pronounced, and the I 2 S-LWR total capital investment cost is 13.02% lower than that of PWR12-BE. If the I 2 S-LWR is compared to a PWR10-BE (traditional loop design, but scaled to the same power level as I 2 S-LWR), the savings are even higher, in the range 11.12%–17.89%. The analysis indicates that I 2 S-LWR has potential to offer an economically attractive design, with TCIC lower than that of a nuclear power plant based on a traditional loop PWR design. In other words, I 2 S-LWR design offers significantly enhanced safety, while at the same time improving economics. The differential economics approach developed in this paper can be used in identifying changes in cost of key components in order to improve the economics of a nuclear reactor design. The method can also help compare the economics of advanced Generation IV reactors and innovative water-cooled SMR (Small Modular Reactors) with respect to standard LWR technologies. In summary, the differential economics approach and can be used as a tool capable of helping stakeholder decisions.

Economic↗

Exploring the Role of Neutral 4-Amino-1,2,4-triazole in the Formation of Hexanuclear f-Element Hydrolysis Products

Our recent observations of an unexpected Ce(III) hydrolysis product from the reaction of 4-amino-1,2,4-triazole (4-NH 2 -1,2,4-Triaz) with CeCl 3 ·7H 2 O, [Ce 6 (μ 3 -O) 4 (μ 3 -OH) 2 (μ 3 -Cl) 2 (Cl) 6 (μ 2 -4-NH 2 -1,2,4-Triaz) 12 ]·7H 2 O, the first high-nuclearity lanthanide complex where all Ln atoms are connected pairwise through 12 N-donor ligands or 12 neutral bridging ligands of any type, prompted us to explore the utility of this ligand in trapping additional f-element examples. Reactions of LnCl 3 ·6H 2 O (Ln = Nd, Eu, Ho) with a large excess of 4-NH 2 -1,2,4-Triaz (20 equiv) and with the addition of small amounts of water to help solubilize the metal salts led to the isolation of the unique hydrolysis products [Nd 6 (μ 3 -OH) 8 Cl 6 (μ 2 -4-NH 2 -1,2,4-Triaz) 12 ][Cl 4 ]·2H 2 O, [Eu 6 (μ 6 -Cl) 0.23 (μ 3 -O 0.77 ) 4 (μ 3 -O) 2.6 (μ 3 -Cl) 0.4 Cl 6 (μ 2 -4-NH 2 -1,2,4-Triaz) 12 ], and [Ho 6 (μ 6 -Cl) 0.21 (μ 3 -O 0.79 ) 4 (μ 3 -OH) 2 Cl 6 (μ 2 -4-NH 2 -1,2,4-Triaz) 12 ][Cl] 3.4 . Here, we also report a Ce(III) analogue prepared in glassware contaminated with Pb(OAc) 2 , namely, [Ce 6 (μ 3 -OH) 8 (BrPbBr 5 )(μ 2 -4-NH 2 -1,2,4-Triaz) 11.5 (OH 2 ) 6 ][Pb 0.84 Br 4.2 ][Br] 3.8 ·2(4-NH 2 -1,2,4-Triaz)·3.6H 2 O. The Nd(III) complex is the structurally most ordered with a clear [Nd 6 (μ 3 -OH) 8 ] cluster core, while the Eu(III) and Ho(III) compounds contain partial occupancy of a μ 6 position and thus result in an incomplete Ln 6 O 9 cluster core formation. The crystallographic results suggest that the 4-NH 2 -1,2,4-Triaz ligand brings Ln(III) ions together, followed by the formation of an Ln 6 O 8 or Ln 6 O 9 core with whatever remaining anions or ligands can be incorporated. Given the complexity of the hydrolysis products of nuclear waste, we expect to continue to find a myriad of closely related complex structures of these types for the f-elements.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

Analysis of Tank 30H (HTF-30-18-96, -97), Tank 32H (HTF-32-18-98, -99), and Tank 37H (HTF-37-18-100, -101) Samples for Support of the Evaporator Feed Qualification and Corrosion Control Programs for the 3H-Evaporator

SRNL analyzed samples from Tank 30H, Tank 32H, and Tank 37H to support the Evaporator Feed Qualification and Corrosion Control Programs for the 3H-Evaporator system. The six samples from Tanks 30H, 32H, and 37H all contain highly concentrated salt solutions with sodium concentrations ranging from 11.8 M to 15.7 M. The samples from Tank 30H show somewhat higher concentrations of the major cations and anions in the sub-surface sample than the surface sample indicating some stratification within the tank. The two Tank 32H samples show similar concentrations of the major anions and cations between the surface and sub-surface samples. The sub-surface sample from Tank 37H shows higher concentrations for most species than the surface sample indicating some stratification within this tank as well. For all six samples, the sum of the major cations versus the sum of the major anions shows a difference of <10% providing an indication of good data quality. The VDS and surface samples from Tank 30H, Tank 32H, and Tank 37H all show silicon concentrations below 30 mg/L except for the Tank 30H VDS with a concentration of 87 mg/L.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Development of Surface Sampling Techniques for the Canister Deposition Field Demonstration (FY22 Update)

This report describes the proposed surface sampling techniques and plan for the multi-year Canister Deposition Field Demonstration (CDFD). The CDFD is primarily a dust deposition test that will use three commercial 32PTH2 NUHOMS welded stainless steel storage canisters in Advanced Horizontal Storage Modules, with planned exposure testing for up to 10 years at an operating ISFSI site. One canister will be left at ambient condition, unheated; the other two will have heaters to achieve canister surface temperatures that match, to the degree possible, spent nuclear fuel (SNF) loaded canisters with heat loads of 10 kW and 40 kW. Surface sampling campaigns for dust analysis will take place on a yearly or bi-yearly basis. The goal of the planned dust sampling and analysis is to determine important environmental parameters that impact the potential occurrence of stress corrosion cracking on SNF dry storage canisters. Specifically, measured dust deposition rates and deposited particle sizes will improve parameterization of dust deposition models employed to predict the potential occurrence and timing of stress corrosion cracks on the stainless steel SNF canisters. The size, morphology, and composition of the deposited dust and salt particles will be quantified, as well as the soluble salt load per unit area and the rate of deposition, as a function of canister surface temperature, location, time, and orientation. Previously, a preliminary sampling plan was developed, identifying possible sampling locations on the canister surfaces and sampling intervals; possible sampling methods were also described. Further development of the sampling plan has commenced through three different tasks. First, canister surface roughness, a potentially important parameter for air flow and dust deposition, was characterized at several locations on one of the test canisters. Second, corrosion testing to evaluate the potential lifetime and aging of thermocouple wires, spot welds, and attachments was initiated. Third, hand sampling protocols were developed, and initial testing was carried out. The results of those efforts are presented in this report. The information obtained from the CDFD will be critical for ongoing efforts to develop a detailed understanding of the potential for stress corrosion cracking of SNF dry storage canisters.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗