Universal Soldering of Lithium and Sodium Alloys on Various Substrates for Batteries
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Chemistry↗
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Stable all-solid-state sodium-sulfur batteries for low-temperature operation enabled by sodium alloy anode and confined sulfur cathode
All-solid-state sodium-sulfur (Na-S) batteries are promising for stationary energy storage devices because of their low operating temperatures (less than 100 °C), improved safety, and low-cost fabrication. Using Na alloy instead of Na metal as an anode in Na-S batteries can prevent dendrite growth and improve interfacial stability between the anode and solid electrolytes to achieve long-cycling stability. A high-sulfur content cathode possessing high sulfur utilization is also important to enable an energy-dense Na-S battery. In this work, we studied Na-Sb and Na- Sn alloy anodes and demonstrated the superiority of Na 3 Sb alloy undergoing a stable Na alloying/dealloying process at 0.04 mA cm -2 for over 500 hours. Combining the optimized Na3Sb alloy anode with sulfur-carbon composites prepared by the vapor deposition approach, the full cell shows a high sulfur specific capacity and improved rate performance. Moreover, the all-solid-state Na alloy-S battery can deliver a high initial discharge specific capacity of 1377 mAh g -1 and maintain good capacity retention of 70 % after 180 cycles at 60 °C. Further, post-cycle characterizations show that both the anode and cathode perform a reversible discharge/charge process after the 1st cycle, and the cathode undergoes significantly rearranged distributions of carbon and solid-state electrolytes after 180 cycles due to severe volume change induced by repeated sodiation/desodiation process.
25 ENERGY STORAGE↗
A review of inherent safety characteristics of metal alloy sodium-cooled fast reactor fuel against postulated accidents
The thermal, mechanical, and neutronic performance of the metal alloy fast reactor fuel design complements the safety advantages of the liquid metal cooling and the pool-type primary system. Together, these features provide large safety margins in both normal operating modes and for a wide range of postulated accidents. In particular, they maximize the measures of safety associated with inherent reactor response to unprotected, double-fault accidents, and to minimize risk to the public and plant investment. High thermal conductivity and high gap conductance play the most significant role in safety advantages of the metallic fuel, resulting in a flatter radial temperature profile within the pin and much lower normal operation and transient temperatures in comparison to oxide fuel. Despite the big difference in melting point, both oxide and metal fuels have a relatively similar margin to melting during postulated accidents. When the metal fuel cladding fails, it typically occurs below the coolant boiling point and the damaged fuel pins remain cool-able. Metal fuel is compatible with sodium coolant, eliminating the potential of energetic fuel coolant reactions and flow blockages. All these, and the low retained heat leading to a longer grace period for operator action, are significant contributing factors to the inherently benign response of metallic fuel to postulated accidents. This paper summarizes the past analytical and experimental results obtained in past sodium-cooled fast reactor safety programs in the United States, and presents an overview of fuel safety performance as observed in laboratory and in-pile tests.
inherent safety↗
Heat-Exchange in One-Phase Liquid-Metal Flows: Databank (Temperature Distribution in Circular Pipes)
In order to develop accurate methods of calculating heat-exchange in forced channel flows, it is necessary to have knowledge about the temperature distribution over the cross section in differ layers of the flow. For ordinary liquids (gas, water, where the Prandtl number Pr ~ 1), such measurements are difficult to perform because of the small thickness of the boundary layer. The problem simplifies for liquid metals, where Pr << 1. This article contains the format of the data as well as an analysis of the primary data from 37 home and foreign works (mercury, sodium-potassium alloy, sodium, lithium, lead-bismuth alloy). The data are presented in electronic form.
42 ENGINEERING↗
Effect of temperature and pressure on cavitation damage to a cobalt base alloy in sodium
Effect of temperature and pressure on cavitation damage to cobalt base alloy in sodium
Johnston, J. R.↗
Effect of temperature and pressure on cavitation damage to a cobalt base alloy in sodium.
Temperature and pressure effect on cobalt base alloy cavitation in liquid sodium, using vibratory apparatus and loss rate contour diagram
Johnston, J. R.↗
Handling and Treatment of Sodium-Potassium Alloy (NaK) after the Loss of Key Personnel - 19002
In September 2010, during cleanup and removal efforts associated with a glovebox containing a sodium-potassium alloy (NaK) cooling system, a NaK reaction event occurred. Near the end of the evolution, which was at a Materials and Fuels Complex (MFC) research facility at the Idaho National Laboratory, a 2 to 3-ounce sample of NaK was being transferred from a metal can to a polyethylene bottle so it could be sent to the disposal outlet for verification analysis. During the transfer, a reaction of the NaK with ambient air was observed in both the metal can and sample container. The potential for a reaction was identified during the planning of this activity and mitigating steps were included in the procedures. In response, the two containers were flooded with argon, closed, and observed overnight. No additional reaction was noted. The sample container and metal can were packaged into a 10-gallon drum and transferred to a permitted storage location pending disposal. While the other containers generated from the deactivation and decommissioning activity were sent off-site for disposal, the 10-gal drum remained due to concerns by a staff specialist that the material could contain potassium superoxide (KO{sub 2}) as a result of the reaction event. The intent at that time was to treat the NaK in a Resource Conservation and Recovery Act (RCRA)-permitted treatment facility at MFC, instead of sending it off-site for disposal. The container was labeled, inspected, and managed as a hazardous waste. Due to the complexities associated with safely handling and treating NaK and KO{sub 2}, the container remained in storage. Treatment and disposal of the NaK and KO{sub 2} container was revisited in 2016, when researchers and chemists were consulted regarding the potential for superoxide formation and the best means to safely store and treat the container. The facility came to realize that due to the retirements of key personnel, it needed to provide current employees operational experience with the handling and treatment of NaK. An outside company was identified and contracted to provide training and consulting services. Several Battelle Energy Alliance, Inc. (BEA) personnel then attended an alkali metals safety training class. The training course provided participants with classroom and hands-on instruction associated with: - Hazards associated with handling and using sodium and NaK; - Proper personal protection equipment, spill control techniques, and first aid; - Common design approaches for buildings, systems, and equipment used to contain sodium and NaK; - What to expect in the event of a water reaction and how to respond; - How to properly prepare for a fire and contain/extinguish small metal fires; - A hands-on demonstration of NaK reaction and controls. Following the training, a procedure was developed to open the container, and inspect and treat the NaK. In June 2017, facility operations personnel successfully completed the inspection and treatment of the NaK drum, eliminating the legacy liability. (authors)
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Handling and Treatment of Sodium-Potassium Alloy (NaK) after the Loss of Key Personnel – 19002
ABSTRACT In September 2010, during cleanup and removal efforts associated with a glovebox containing a sodium-potassium alloy (NaK) cooling system, a NaK reaction event occurred. Near the end of the evolution, which was at a Materials and Fuels Complex (MFC) research facility at the Idaho National Laboratory, a 2 to 3-ounce sample of NaK was being transferred from a metal can to a polyethylene bottle so it could be sent to the disposal outlet for verification analysis. During the transfer, a reaction of the NaK with ambient air was observed in both the metal can and sample container. The potential for a reaction was identified during the planning of this activity and mitigating steps were included in the procedures. In response, the two containers were flooded with argon, closed, and observed overnight. No additional reaction was noted. The sample container and metal can were packaged into a 10-gallon drum and transferred to a permitted storage location pending disposal. While the other containers generated from the deactivation and decommissioning activity were sent off-site for disposal, the 10-gal drum remained due to concerns by a staff specialist that the material could contain potassium superoxide (KO2) as a result of the reaction event. The intent at that time was to treat the NaK in a Resource Conservation and Recovery Act (RCRA)-permitted treatment facility at MFC, instead of sending it off-site for disposal. The container was labeled, inspected, and managed as a hazardous waste. Due to the complexities associated with safely handling and treating NaK and KO2, the container remained in storage. Treatment and disposal of the NaK and KO2 container was revisited in 2016, when researchers and chemists were consulted regarding the potential for superoxide formation and the best means to safely store and treat the container. The facility came to realize that due to the retirements of key personnel, it needed to provide current employees operational experience with the handling and treatment of NaK. An outside company was identified and contracted to provide training and consulting services. Several Battelle Energy Alliance, Inc. (BEA) personnel then attended an alkali metals safety training class. The training course provided participants with classroom and hands–on instruction associated with: • Hazards associated with handling and using sodium and NaK • Proper personal protection equipment, spill control techniques, and first aid • Common design approaches for buildings, systems, and equipment used to contain sodium and NaK • What to expect in the event of a water reaction and how to respond • How to properly prepare for a fire and contain/extinguish small metal fires • A hands-on demonstration of NaK reaction and controls. Following the training, a procedure was developed to open the container, and inspect and treat the NaK. In June 2017, facility operations personnel successfully completed the inspection and treatment of the NaK drum, eliminating the legacy liability.
99 GENERAL AND MISCELLANEOUS↗
Syringe for injecting sodium-potassium alloy.
Syringe injection system for seeding plasma with Na-K alloy in study of MHD power generation and propulsion
SEEDING↗
Effects of Sodium Exposure on the Tensile Properties of Grade 91 steel
Two heats of G91 steel have been investigated in liquid sodium at 550- 650°C to understand their corrosion behaviour, microstructural evolution, and tensile properties. Sodium exposure experiments of subsized, sheet-type tensile specimens of G91 were performed in Argonne’s forced convection sodium materials testing loops, SMT-1 and SMT-2. Maximum exposure times of ~49,000 h at ~38,000 h, and ~20,000 h have been achieved at 550, 600, and 650°C, respectively. Thermal aging experiments of G91 steel were also conducted in parallel, and the results were compared with those of the sodium-exposed specimens to isolate the effect of thermal aging from the effect of sodium exposure. It was found that the yield stress and ultimate tensile strength of G91 were reduced by more than 50% after sodium exposures at 650°C due to decarburization. The reduction in tensile strength was accompanied by the dissolution of M 23 C 6 carbides and excessive grain growth. The effect of sodium at 550 and 600°C was insignificant for exposure times up to ~49,000 h and ~38,000 h, respectively. The effect of sodium on the microstructural stability and tensile properties of G91 is largely dependent on the carbon mass transfer in the alloy-sodium system.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
A Heterogeneous Sodium Fast Reactor Designed to Transmute Minor Actinide Actinide Waste Isotopes into Plutonium Fuel
An axial heterogeneous sodium fast reactor design is developed for converting minor actinide waste isotopes into plutonium fuel. The reactor design incorporates zirconium hydride moderating rods in an axial blanket above the active core. The blanket design traps the active core’s axial leakage for the purpose of transmuting Am-241 into Pu-238. This Pu-238 is then co-recycled with the spent driver fuel to make new driver fuel. Because Pu-238 is significantly more fissile than Am-241 in a fast neutron spectrum, the fissile worth of the initial minor actinide material is upgraded by its preconditioning via transmutation in the axial targets. Because, the Am-241 neutron capture worth is significantly stronger in a moderated epithermal spectrum than the fast spectrum, the axial targets serve as a neutron trap which recovers the axial leakage lost by the active core. The sodium fast reactor proposed by this work is designed as an overall transuranic burner. Therefore, a low transuranic conversion ratio is achieved by a degree of core flattening which increases axial leakage. Unlike a traditional “pancake” design, neutron leakage is recovered by the axial target/blanket system. This heterogeneous core design is constrained to have sodium void and Doppler reactivity worth similar to that of an equivalent homogeneous design. Because minor actinides are irradiated only once in the axial target region; elemental partitioning is not required. This fact enables the use of metal targets with electrochemical reprocessing. Therefore, the irradiation environment of both drivers and targets was constrained to ensure applicability of the established experience database for metal alloy sodium fast reactor fuels.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Efficient and stable emission of warm-white light from lead-free halide double perovskites
Lighting accounts for one-fifth of global electricity consumption. Single materials with efficient and stable white-light emission are ideal for lighting applications, but photon emission covering the entire visible spectrum is difficult to achieve using a single material. Metal halide perovskites have outstanding emission properties; however, the best-performing materials of this type contain lead and have unsatisfactory stability. Here we report a lead-free double perovskite that exhibits efficient and stable white-light emission via self-trapped excitons that originate from the Jahn–Teller distortion of the AgCl 6 octahedron in the excited state. By alloying sodium cations into Cs 2 AgInCl 6 , we break the dark transition (the inversion-symmetry-induced parity-forbidden transition) by manipulating the parity of the wavefunction of the self-trapped exciton and reduce the electronic dimensionality of the semiconductor. This leads to an increase in photoluminescence efficiency by three orders of magnitude compared to pure Cs 2 AgInCl 6 . The optimally alloyed Cs 2 (Ag 0.60 Na 0.40 )InCl 6 with 0.04 per cent bismuth doping emits warm-white light with 86 ± 5 per cent quantum efficiency and works for over 1,000 hours. We anticipate that these results will stimulate research on single-emitter-based white-light-emitting phosphors and diodes for next-generation lighting and display technologies.
36 MATERIALS SCIENCE↗
Supramolecular assembly of borate with quaternary ammonium: Crystal structure and tunable luminescent properties
A new borate [C{sub 6}H{sub 16}N][B{sub 5}O{sub 6}(OH){sub 4}] (1) is synthesized hydrothermally by the reaction of isopropyltrimethylammonium hydroxide with boric acid. It crystallizes in the triclinic space group P-1 with the parameters a=9.1578(10) Å, b=9.372(9) Å, c=9.9812(10) Å, α=66.508(2)°, β=74.751(2)°, γ=81.893(2)°. The [B{sub 5}O{sub 6}(OH){sub 4}]{sup −} anions are interlinked via hydrogen bonding forming a 3D supramolecular network containing large cavities, where reside the (CH{sub 3}){sub 3}(i-C{sub 3}H{sub 7}) N{sup +} cations. This borate shows tunable luminescent properties with temperature, heating-treatment, exciting-light, and solvents. The fluorescent intensity of 1 enhances 6-fold with decreasing the temperature from 25 K to 78 K. By treatment under different temperatures, the luminescence of 1 shifted from blue to white and the sample treated at 230 °C emits bright white light to naked eyes. The hybrid borate can disperse in different solvents, and shows a red-shifted and intense emission in polar solvents. - Graphical abstract: The new quaternary ammonium borate [C{sub 6}H{sub 17}N][B{sub 5}O{sub 6}(OH){sub 4}] contains a 3D supramolecular network formed by hydrogen bond linked [B{sub 5}O{sub 6}(OH){sub 4}]{sup −} anions and shows tunable luminescent properties with temperature, excitation light, and solvents. Highlights: ► A novel quaternary ammonium borate was synthesized. ► It possesses a supramolecular network fomed by H-bonded [B{sub 5}O{sub 6}(OH){sub 4}]{sup −} anions. ► This borate shows tunable luminescent properties with temperature, heating treatment, excitation light, and solvents.
37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
Photoluminescence and electrical characterization of unfilled tetragonal tungsten bronze Ba{sub 4}La{sub 1−x}Eu{sub x}TiNb{sub 9}O{sub 30}
Graphical abstract: PL spectra of the unfilled TTB structure BLTN: Eu{sup 3+}x samples (x = 0.00, 0.25, 0.50, 0.75, and 1.00) excited by 399 nm. The inset is a schematic diagram of the unfilled TTB structure. - Highlights: • Unfilled TTB structure BLTN: Eu{sup 3+}x ceramics have been synthesized. • Photoluminescenct properties of the BLTN: Eu{sup 3+}x ceramics have been first reported. • Bright red emission excited by NUV light has been observed at room temperature. • Obvious variations of dielectric characteristics have been confirmed. • Relaxor-like ferroelectric phase transitions have been detected. - Abstract: Unfilled tetragonal tungsten bronze (TTB) structure Ba{sub 4}LaTiNb{sub 9}O{sub 30} doped by Eu{sup 3+} (BLTN: Eu{sup 3+}x) with different x have been prepared, and their structural, photoluminescence, dielectric, and ferroelectric properties are carefully investigated in this work. Bright red emission, originating from {sup 5}D{sub 0} → {sup 7}F{sub 1} and {sup 5}D{sub 0} → {sup 7}F{sub 2} transitions of Eu{sup 3+} ions, has been observed by naked eyes at room temperature under near ultraviolet (NUV) light excitation. Optimized emission intensity is obtained when x = 1.00 for present unfilled TTB-type BLTN: Eu{sup 3+}x samples. Furthermore, with increasing x, the dielectric and ferroelectric characteristics of the unfilled TTB-type BLTN: Eu{sup 3+}x samples also display remarkable variation. When x ≥ 0.50 relaxor-like ferroelectric phase transitions are detected above room temperature, it is believed that unfilled TTB-type BLTN: Eu{sup 3+}x = 1.00 involving bright photoluminescence and enhanced ferroelectric properties may act as a potentially multifunctional optical-electro material.
36 MATERIALS SCIENCE↗
Report on the Initial Loading of SS3 Specimens from the First Commercial Heat of Alloy 709 in Sodium Materials Test Loops for Long-term Exposure
This report provides an update on the understanding of the effects of sodium exposures on tensile properties of advanced alloy A709 in support of the design and operation of structural components in sodium-cooled fast reactors (SFRs). The report is a Level 2 deliverable in FY18(M2NT-18AN050503011), under the Work Package NT-18AN05050301, “A709 Sodium Compatibility” performed by Argonne National Laboratory (ANL), as part of the Advanced Materials Development Program for Fast Reactors.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Effect of pH on Corrosion Performance of Cu-Se and Cu-Se-Te Alloys in 3.5% Sodium Chloride Solution
The influence of pH (2, 7, 12) on the corrosion behavior of Cu-Se-Te and Cu-Se alloys in a 3.5% sodium chloride solution was investigated using static immersion testing and electrochemical methods. The morphologies of the alloys after corrosion were observed by scanning electron microscopy. The distribution of elements was detected by energy-dispersive spectrometry, and the phases were identified by x-ray diffraction. The results indicate that the corrosion rate decreases significantly with an increase in pH. No compact protective film forms on the alloy surface when pH is 2; meanwhile, Cu{sub 2}(OH){sub 3}Cl and Cu{sub 2}O films form on the alloy surface when pH is 7 or 12. The corrosion rates of alloys in the NaCl solution with different pH levels vary because the cathode reaction and the protective films affect the corrosion rate synthetically. In addition, the Cu-0.5Se alloy generally exhibits satisfactory conductivity, strength, and corrosion resistance. The corrosion resistance of the Cu-0.5Se alloy is higher than that of the Cu-Se-Te alloys, particularly at low pH levels.
36 MATERIALS SCIENCE↗
Assessment of Aging Degradation Mechanisms of Alloy 709 for Sodium Fast Reactors
This project was initiated in October 2015 to investigate the creep and creep-fatigue behavior of Alloy 709 for structural components in Gen IV sodium cooled fast spectrum reactors. The goal of this project is to develop an understanding of the mechanisms responsible for the behavior of austenitic stainless steel alloy 709 under accelerated testing conditions to confidently predict long term creep and creep-fatigue behavior at the times (500,000 hours) and temperatures (550°C) of interest for fast reactor structural applications. Alloy 709 has exhibited significantly better creep resistance than Alloy 316H, which is currently specified in design of fast reactor structural components. However, the alloy has not been code qualified for elevated temperature nuclear design. The elevated temperature microstructure stability and creep and creep-fatigue behavior need to be better understood so service life can be accurately predicted. Alloy 709 was developed for fossil boiler applications, which operate at a higher temperature than the nuclear components. Thus, there is a strong need to characterize its creep and creep-fatigue behavior at these lower temperatures and corresponding longer service lifetimes. The overall objective of this project is to develop sufficient understanding of the mechanisms responsible for behavior of Alloy 709 under accelerated testing conditions to confidently predict long term creep and creep-fatigue behavior at the times (500,000 hours) and temperatures (550°C) of interest for fast reactor structural applications. The following report summarizes the findings of the creep-fatigue and creep behavior with respect to the research questions identified during the course of the project. Additionally, the training and professional development that occurred through the project and the products resulting from the project are summarized.
36 MATERIALS SCIENCE↗
Corrosion Resistance of Nickel Alloys in Molten Sodium Hydroxide
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Probst, H B↗