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In this work, the syntheses of bidentate and tripodal phosphine ligands containing aryl chlorides were achieved in 4 and 7 steps respectively, starting from diethyl malonate.
In response to acidic pH, the widely expressed proton-activated chloride (PAC) channel opens and conducts anions across cellular membranes. By doing so, PAC plays an important role in both cellular physiology (endosome acidification) and diseases associated with tissue acidosis (acid-induced cell death). Despite the available structural information, how proton binding in the extracellular domain (ECD) leads to PAC channel opening remains largely unknown. Here, through comprehensive mutagenesis and electrophysiological studies, we identified several critical titratable residues, including two histidine residues (H130 and H131) and an aspartic acid residue (D269) at the distal end of the ECD, together with the previously characterized H98 at the transmembrane domain–ECD interface, as potential pH sensors for human PAC. Mutations of these residues resulted in significant changes in pH sensitivity. Some combined mutants also exhibited large basal PAC channel activities at neutral pH. By combining molecular dynamics simulations with structural and functional analysis, we further found that the β12 strand at the intersubunit interface and the associated “joint region” connecting the upper and lower ECDs allosterically regulate the proton-dependent PAC activation. Our studies suggest a distinct pH-sensing and gating mechanism of this new family of ion channels sensitive to acidic environment.
This study presents a method that can be used to gain information relevant to determining the corrosion risk for spent nuclear fuel (SNF) canisters during extended dry storage. Currently, it is known that stainless steel canisters are susceptible to chloride-induced stress corrosion cracking (CISCC). However, the rate of CISCC degradation and the likelihood that it could lead to a through-wall crack is unknown. This study uses well-developed computational fluid dynamics and particle-tracking tools and applies them to SNF storage to determine the rate of deposition on canisters. The deposition rate is determined for a vertical canister system and a horizontal canister system, at various decay heat rates with a uniform particle size distribution, ranging from 0.25 to 25 µm, used as an input. In all cases, most of the dust entering the overpack passed through without depositing. Most of what was retained in the overpack was deposited on overpack surfaces (e.g., inlet and outlet vents); only a small fraction was deposited on the canister itself. These results are provided for generalized canister systems with a generalized input; as such, this paper is intended to demonstrate the technique. This study is a part of an ongoing effort funded by the U.S. Department of Energy, Nuclear Energy Office of Spent Fuel Waste Science and Technology, which is tasked with doing research relevant to developing a sound technical basis for ensuring the safe extended storage and subsequent transport of SNF. This work is being presented to demonstrate a potentially useful technique for SNF canister vendors, utilities, regulators, and stakeholders to utilize and further develop for their own designs and site-specific studies.
There is considerable interest in molten chloride salts for several applications including thermal storage and next-generation molten salt reactors (MSRs). Several studies have concluded that Cl salts are highly corrosive to structural materials. Using Oak Ridge National Laboratory’s established methodology, Cl salt compatibility was assessed for candidate Ni-based alloys 230, 600 and 740 H at 600°–800°C in static welded capsules and in flowing thermal convection loop (TCL) salt experiments. Simply drying commercial Cl salt at 550°–650°C and adding~0.05 wt.%Mg was able to produce small specimen mass changes and limited surface attack after 100–1000 h exposures. Intentional additions of H 2 O, NiCl 2 and undried salt were used to better understand the role of impurities and achieve the>50 µm levels of attack reported in other studies. Further, characterisation of Cr depletion and oxide formation in conjunction with pre- and post-test salt chemistry measurements are being used to understand salt compatibility of structural materials.
The objective of the chloride based volatility project (CBV) was to demonstrate the ability to separate uranium from used fuel to enable process improvements resulting in 10x reduction of waste volume, while maintaining safeguards standards and global backend costs at $\$$1/MW-hr. Current industrial practices perform separations of used fuel using solvent media in the form of aqueous and molten salt processes, resulting in contaminated process waste. The CBV approach utilized solid state chemistry with no solvent media and was successful in chlorinating uranium and fission product oxides that sublimed into the gas phase and were collected in targeted condensation zones based on temperature gradients of chlorinated products. Recovery of better than 95% of initial uranium in the form of UCl4 was demonstrated when simulated used nuclear fuel was used. Laser induced breakdown spectroscopy, LIBS, and ultraviolet-visible spectroscopy, UV-Vis, were combined into a high temperature flow cell design and utilized as process monitoring techniques to observed chlorinated products leave the reaction vessel in real time.
Dry storage canisters of used nuclear fuels are fabricated using SUS 304L stainless steel. Chloride-induced stress corrosion cracking (CISCC) is one of the major failure modes of dry storage canisters. The cracked canisters can be repaired by friction stir welding (FSW), a low-heat input ‘solid-phase’ welding process. It is important to evaluate the ClSCC resistance of the friction stir welded material. Stress corrosion cracking (SCC) studies were carried out on mill-annealed base materials and friction stir welded 304L stainless U-bend specimens in 3.5% NaCl + 5 N H2SO4 solution at room temperature and boiling MgCl2 solution at 155 °C. The engineering stress on the outer fiber of the FSW U-bend specimen was ~60% higher than that of the base metal (BM). In spite of the higher stress level of the FSW, both materials (FSW and BM) showed almost similar SCC failure times in the two different test solutions. The SCC occurred in the thermo-mechanically affected zone (TMAZ) of the FSW specimens in the 3.5% NaCl + 5 N H2SO4 solution at room temperature, while the stirred zone (SZ) was relatively crack-free. The failure occurred at the stirred zone when tested in the boiling MgCl2 solution. Hydrogen reduction was the cathodic reaction in the boiling MgCl2 solution, which promoted hydrogen-assisted cracking of the heavily deformed stirred zone. The emergence of the slip step followed by passive film rupture and dissolution of the slip step could be the SCC events in the 3.5% NaCl + 5 N H2SO4 solution at room temperature. However, the slip step height was not sufficient to cause passivity breakdown in the fine-grained SZ. Therefore, the SCC occurred in the partially recrystallized softer TMAZ. Overall, the friction-stirred 304L showed higher tolerance to ClSCC than the 304L base metal.
MCRE will be the first fast spectrum, molten chloride reactor in operation. Small critical experiment to provide data for future reactors. - Collaboration between Southern Company Services, TerraPower, and Idaho National Laboratory (INL), as well as other partners. - Expected to be built and operated at the LOTUS testbed (ZPPR). - Authorized by the U.S. Department of Energy. - Technology development of salt synthesis.
This work investigates the design criteria considering thermal, structural, and mate-rial selection of a ternary chloride (MgCl2-NaCl-KCl) molten salt pump that is capable of running at temperatures above 720 °C. In this work, preliminary testing of salt-wetted hydrodynamic bearings was done to investigate different material pairings including colmonoy grades, stellite grades, and a novel NiWC-based alloy (i.e. HybrimetTM NiWC3b) used as a High Velocity Ox-ygen Fuel (HVOF) coating. The coating has shown to withstand the corrosive nature of the salt, and is additionally used to protect a 316L stainless steel pump reservoir. The pump wetted components are made of Inconel 625, and it features a carbon ring seal developed by Flows-erve flooded with a N2 gas flow on the overhead to keep salt vapors from interacting with the external ambient air. This research showcases a promising design for use in the CSP Gen 3 liquid-pathway and sets the ground for further development and use of the HybrimetTM NiWC3b as a competitive alternative to available superalloys. Further work will further characterize the pump by performance & cycle testing in molten salt at higher temperatures.
Short communication on the completion and testing of a new high-temperature cell holder to support high dose rate electron beam irradiations of select sulfur chlorides (S2Cl2 and SOCl2) for advanced low temperature chlorination of zirconium-based used nuclear fuel cladding.
The Savannah River Site (SRS) produced tritium, plutonium, and special nuclear materials for national defense, medicine, and the space programs. As part of operations, the F-Area Seepage Basins operated until 1988 for the disposition of deionized acidic wastewater from the F Separations Facility. The wastewater contained dilute nitric acid and low concentrations of non-radioactive metals, and radionuclides, with the major isotopes being Cs-137, Sr-90, U-235, U-238, Pu-239, Tc-99, I-129, and tritium. The seepage basins were closed in 1988 and backfilled and capped by 1991. The groundwater emanating from beneath the closed and capped seepage basins is acidic and contains elevated levels of both chemical and radiological contaminants. Releases from the groundwater plumes sourced from the F-Area Seepage Basins have impacted the water quality of Fourmile Branch, which is a small tributary to the Savannah River, a regional water source. A large pump-and-treat system was constructed in 1997 and operated until 2003 in an attempt to capture the releases to Fourmile Branch. The system in F Area and a similar system in H Area were expensive (∼$1.3 M/month) to operate and produced large quantities of radioactive waste, with concentrations of I- 129 too high to be disposed of at the SRS. In 2004, SRS replaced pump-and-treat with a funnel and gate system that along with operation of a base injection system at the gates reduces the flux of contaminants to the wetlands adjacent to Fourmile Branch. The alkaline solution injected into the aquifer neutralizes the acidic plume and immobilizes many of the cationic constituents. However, base injection is not effective in managing the release of iodine-129, an anionic contaminant. To address iodine-129 SRS and the Savannah River National Laboratory developed an in situ technology that uses ultra-fine ground silver chloride (AgCl) as an injectable capture medium for the sequestration of iodine-129. The AgCl amendment has a very small particle size and is designed to be injected into the contaminated aquifer to capture iodine-129. Dissolved iodine-129 forms a stable and highly insoluble solid (silver iodide) upon contact with AgCl. Laboratory studies, a field scale pilot test (2009), and three deployments (2011, 2015 and 2019) of AgCl have been successfully performed at the F-Area Seepage Basins. (authors)
Molten salt reactors (MSRs) have garnered increasing attention recently with several demonstration efforts on the way. A key challenge to the licensing basis for these reactors is the lack of experimental data on fueled salts. This is expected to be crucial to the safety evaluation and licensing basis of reactors of this type deployed in the future. While capability for irradiating molten salts has been reestablished in the recent decade, no enriched fuel irradiation capability has been developed and tested as of yet. A new experiment vehicle under development at Idaho National Laboratory (INL) is presented here. The Molten-salt Research Temperature-controlled Irradiation (MRTI) experiment was developed to host enriched uranium bearing salt samples to be irradiated at a test reactor within the lab complex. One of the key scientific objectives is to provide irradiated salt samples for post-irradiation examination (PIE) to study the impact of fission product generation and neutron/gamma radioactivity on the salt solution and salt-facing wall material. This paper provides a detailed overview of the mechanical design of the experiment, followed by an overview of the fabrication and assembly of an initial prototype vehicle (with non-fuel-bearing salt). A summary of the key analyses conducted as a part of the performance and safety evaluation is then provided. Lastly, an overview of the test conducted in prototypic out-of-pile (non-neutron) environment are shown. These evaluations provide the foundation for a planned irradiation of an enriched uranium-bearing chloride salt sample in the near term. The upcoming irradiation will contain 13 cm3 of UCl3-NaCl salt (93% enrichment) generating around 20 W/cm3 of fission energy during irradiation and a temperature range that can be contained between bounds of 525-900°C.
Presentation highlight the INL Center for Radiation Chemistry Research's contributions the Material Recovery & Waste Form Development Campaign on sulfur chloride processing technologies.
Over the previous 5 years, Idaho National Laboratory (INL) has been working with Southern Co. and TerraPower on the Advanced Reactor Demonstration Program (ARDP) funded Molten Chloride Reactor Experiment (MCRE) project. As part of this effort, INL has developed a fuel synthesis process to produce the NaCl-UCl3 fuel salt that MCRE will need for operation. Dr. Phillips will present on process development and scale-up testing and results, as well as provide an update on the current status of fuel production for MCRE. To date, the fuel synthesis process has been demonstrated at full scale using depleted uranium, and the equipment necessary for production of the fuel for MCRE has been installed in the Fuel Manufacturing Facility (FMF) at the INL’s Materials and Fuels Complex (MFC). The NaCl-UCl3 produced to-date has been shown to be of 99.99% purity or greater, and be within tolerance for all relevant parameters. Overall process efficiency in terms of uranium utilization has been demonstrated to be above 90%. The synthesis operation has also been shortened to allow for completion of the reaction within a single 10 hour working shift. Consequently, the process developed is expected to be capable of meeting all project objectives for efficiency, purity, scale, and scheduling. Production of NaCl-UCl3 fuel salt for MCRE is projected to begin during the Summer of 2025.
The United States Department of Energy (DOE) is committed to the advancement of nuclear reactor technology through initiatives such as the Advanced Reactor Development Program (ARDP), to diversify the United States energy portfolio towards more sustainable energy options. The ARDP includes demonstration by industry partners of molten salt fast reactors (MSRs). Construction of molten salt reactor technology requires qualified nuclear structural materials. Unfortunately, there are no current materials that meet current qualification requirements dictated by the Nuclear Regulatory Commission for construction of MSRs. Adapting current structural material qualifications requires expansion of our current knowledgebase on corrosion performance. In this investigation, we assess microstructural changes in a Ni-based superalloy after exposure to a chloride-containing salt system through a correlated multi-modal approach combining several advanced characterization techniques. Namely we will highlight the impact of grain boundary phenomena at the onset of corrosion attack, including the role of intergranular crack propagation and development of internal corrosion products. The findings from this investigation will further expand our assessment of the corrosion performance of structural materials being investigated for construction of MSR components.
Solid-electrolyte-based molten-metal batteries have attracted considerable attention for grid-scale energy storage. Although ZEBRA batteries are considered one of the promising candidates, they still have the potential concern of metal particle growth and ion exchange with the β”-Al 2 O 3 electrolyte. Herein, a Li 6.4 La 3 Zr 1.4 Ta 0.6 O 12 solid-electrolyte-based molten lithium–molybdenum–iron(II) chloride battery (denoted as Li–Mo–FeCl 2 ) operated at temperature of 250 °C, comprising a mixture of Fe and LiCl cathode materials, a Li anode, a garnet-type Li-ion ceramic electrolyte, and Mo additive, is designed to overcome these obstacles. Different from conventional battery reaction mechanisms, this battery revolutionarily synchronizes the reversible Fe–Mo alloying–dealloying reactions with the delithiation–lithiation processes, meaning that the porous Mo framework derived from Fe–Mo alloy simultaneously suppresses the growth of pure Fe particles. By adopting a Li anode and a Li-ion ceramic electrolyte, the corrosion problem between the cathode and the solid electrolyte is overcome. With similar battery cost ($12 kWh –1 ), the theoretical energy density of Li–Mo–FeCl 2 battery surpasses that of a Na–FeCl 2 ZEBRA battery over 25%, to 576 Wh kg –1 and 2216 Wh L –1 , respectively. Experimental results further prove this cell has excellent cycling performance (472 mAh g LiCl –1 after 300 cycles, 50 mg active material) and strong tolerance against the overcharge–overdischarge (3–1.6 V) and freezing–thawing (25–250 °C) incidents.
All-solid-state sodium batteries (ASSBs) employing halide solid electrolytes (SEs) offer a cost-effective and energy-dense alternative to conventional liquid electrolyte systems. However, their high voltage (>4 V vs. Na/Na + ) performance remains limited by interfacial instability between the cathode active material (CAM) and the SE. We present here the electrochemical and interfacial behaviors of crystalline NaAlCl 4 and amorphous sodium–aluminum–oxychloride (NACO) SEs when combined with NaNi 0.5 Mn 0.5 O 2 cathode. While oxygen incorporation in NACO enhances ionic conductivity by nearly three orders of magnitude relative to NaAlCl 4 , it does not improve high-voltage cycling stability. Cells employing NACO exhibit accelerated capacity fade, increased cell impedance growth, and intrinsic oxygen evolution above 4.5 V vs. Na 3 Sn, as revealed by operando electrochemical mass spectrometry. In contrast, the NaAlCl 4 -based cells show no detectable gas release, underscoring their superior high-voltage stability and safety. Time-of-flight secondary-ion mass spectrometry confirms the formation of Al─O and Ni/Mn─Cl species, respectively, in the SE and CAM, indicating redox-driven anion exchange that contributes to kinetic hindrance of high-voltage phase transitions. The findings establish that while oxygen incorporation enhances ionic transport, it can compromise interfacial stability, suggesting pure chloride SEs may offer a more robust and intrinsically safer pathway for developing high-energy ASSBs.
Abstract A series of glycerol‐based compounds were investigated for their application as plasticizers for flexible poly(vinyl chloride) (PVC) blends. The effect of plasticizer chemical structure on the performance, migration behavior, and blend morphology were evaluated and compared to blends produced using the commercial plasticizer dioctyl terephthalate (DOTP). Blends containing 40 phr (parts per hundred rubber) of glycerol‐based plasticizer showed a considerable reduction (between 54 to 86°C) in glass transition temperature ( T g ) relative to neat PVC ( T g ~ 80°C). Tensile testing of samples prepared with the glycerol analogs demonstrated higher ductility (elongation at break values of up to 97%) than DOTP (elongation at break value of 75%) at identical plasticizer loadings. The surface morphologies showed excellent incorporation of the glycerol plasticizers functionalized with alkyl chains longer than four carbons in length into the PVC matrix, whereas droplet formation was observed in blends with shorter chain glycerol derivatives. Leaching behavior of the plasticized samples were evaluated into different media and showed that plasticizers comprised of branched, or longer alkyl chains produced 2‐ to 4‐fold lower migration rates compared to those with shorter alkyl chains into polar solutions.
Abstract A highly ordered carbon structure based on polyvinylidene chloride precursor has been developed to host silicon nanoparticles. The stoichiometric ratio of sacrificial H and Cl elements facilitated full utilization of the carbon content of the precursor that produced robust carbon coatings on silicon nanoparticles with excellent mechanic properties and electrochemical stabilities. The optimal sintering temperature and carbon content have been investigated. When evaluated as the anode of a lithium‐ion battery (LIB), the Si : C 1 : 2 composite sintered at 800 °C (SiC12‐800) provided excellent capacity retention of 85 % at 0.1 C after 50 cycles with an enhanced CE, which reached 99 % only after 10 cycles. The SiC12‐800 electrode maintained a specific capacity of 709.2 mAh/g after 300 cycles at 0.3 C, and delivered a high rate performance of 737.9 mAh/g and 485.6 mAh/g at 5 C and 10 C, respectively. The results indicate that polymer precursors with stoichiometric ratio of sacrificial elements have high potential for generating highly robust carbon coatings for silicon anodes in high energy density LIBs.