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At least 1,135 records · Page 63
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

CNL Technologies for Heavy Water Detritiation - 19437

Canadian Nuclear Laboratories (CNL) is responsible for a large amount of heavy water owned by the Government of Canada. A significant portion of this heavy water has been downgraded and is chemically and/or radiologically contaminated (besides tritium). Storage, monitoring and disposal of this water incurs significant costs. Upgrading following chemical purification (including removal of radionuclides other than tritium) of this water is feasible with commercial processes. However, removal of the tritium to the level that would allow the heavy water to be available for non-nuclear applications requires very specialized processes. CNL has developed and demonstrated the Combined Electrolysis and Catalytic Exchange (CECE) detritiation technology to reduce the tritium to the level at which the heavy water can be classified as non-radioactive. The CECE process consists of a Liquid Phase Catalytic Exchange (LPCE) bed, where the transfer of hydrogen isotopes between water and gaseous hydrogen occurs, and an electrolysis cell (E-cell) that produces hydrogen and separates the isotopes of hydrogen, adding to the enrichment achieved in the LPCE catalyst bed. The application of the CECE technology to the tritium-contaminated heavy water would restore its full market value. As part of a project supported by CNL Decommissioning and Waste Management, a business case is being developed to build and operate a facility that will purify and detritiate heavy water to render it commercially attractive. The main goal of the current study is to evaluate the net benefits of detritiation and product sale versus the cost of permanent storage or disposal. Such a facility would not only eliminate the need to store the heavy water as a radioactive waste material and provide revenue, but would also serve as a demonstration of CNL's technology for similar applications elsewhere. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Water and Ammonia Desorption from 5A Zeolite

The objective of this project is to demonstrate the adsorption and desorption characteristics of water and ammonia on 5A zeolite during regeneration conditions. As discussed in literature, zeolite material is a crystalline hydrated aluminosilicate material. Zeolites have high cation exchange ability and are commonly used as adsorbents in separation and purification processes. The property of uniformly sized pores allows zeolites to selectively separate gases. Because they have varying ion selectivity, there is competitive adsorption in a multicomponent mixture. 5A zeolite has a uniform pore diameter of 5 Angstroms. Water and ammonia are miscible components with similar molecular weights and size. This makes distinguishing the two compounds difficult using mass spectrometry. In a previous study, it was shown that water tends to desorb from 5A zeolite from 40 deg. C to 353 deg. C with peaks at 177 deg. C and 244 deg. C, while ammonia tends to desorb from 300 deg. C to 580 deg. C with a peak at 436 deg. C as determined using a Micromeritics AutoChem 2920 Instrument. The one drawback of using a Micromeritics Instrument is that only a few milligrams of sample material can be used. Bench scale tests can better demonstrate desorption characteristics of water and ammonia during regeneration from 5A zeolite material. Water in argon carrier gas and ammonia in hydrogen carrier gas will be tested separately, then a mixture of the two will be tested. Initial bake-out conditions: hydrogen carrier gas, slowly raise temp to approximately 450 deg. C until desired moisture content <100 ppm is achieved. Water loading: argon carrier gas through bubbler until breach is observed using the residual gas analyzer (RGA). Ammonia loading: hydrogen carrier gas through system until breach is observed using RGA. Regeneration conditions: hydrogen carrier gas, raise temp to approximately 300 deg. C, then slowly to 450 deg. C. Observe signals on RGA. Weigh samples between tests and integrate RGA signals. Accomplishments: Developed work control documentation and completed safety related requirements; Flow sheet and test system development; Initial bake-out of first sample. Testing will determine if regenerating zeolite material can be safely accomplished by high-temperature bake-out to remove ammonia. If unsuccessful, ammonia would only be removed at the end of life.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

DEM and MELT In-Can Vitrification Process for Fukushima Daiichi Water Treatment Secondary Waste - 20034

Fukushima Daiichi Nuclear Power Station was severely damaged by a tsunami induced by the earthquake on March 11, 2011. A large amount of water treatment secondary waste has been generated from the purification work of contaminated water that consists of the water flowing into the reactor building. The Mid and Long-Term Road map specifies that solid waste at Fukushima Daiichi Nuclear Power Station should be stored safely and reasonably according to its characteristics. In order to meet this goal, a versatile treatment solution is needed, and the In-Can vitrification seems to be a suitable solution. The DEM and MELT In-Can vitrification process is being developed and qualified in France to treat waste arising from decommissioning and dismantling operations. This qualification is carried out through the consortium of the DEM and MELT project coordinated by CEA which gathers Orano, ECM technologies and Andra. The DEM and MELT In-Can vitrification process is a robust, simple and versatile in situ vitrification process. It is designed for high or intermediate level waste and is compact enough to be implemented in an existing facility or close to the waste to be treated. It is developed to treat liquid and solid waste, to produce a small amount of secondary waste and to minimize investment and operating costs. The DEM and MELT In-Can vitrification process is also developed with a modular design in order to be adapted to nuclear operators needs and requirements. The In-Can melting process has already been developed and fully qualified by the CEA for Alpha liquid waste and can be rated TRL 7 for this scope. The In-Can vitrification technology is considered as a qualified solution that could be applied for the treatment of waste resulting from the D and D work in Fukushima Daiichi Nuclear Power Station. The feasibility evaluation of the DEM and MELT In-Can vitrification process for Fukushima Daiichi water treatment secondary waste is a project led by Orano, CEA and ANADEC. This R and D project is performed by the funds from the Ministry of Economy, Trade and Industry as The Subsidy Program 'Project of Decommissioning and Contaminated Water Management'. This paper presents the first results of this feasibility evaluation, from the scenarios and glass formulation definition to the industrial implementation feasibility. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Analysis of Using a Hybrid ZIRCEX Process to Generate High-Assay Low-Enriched Uranium (HALEU) - 20158

Idaho National Lab (INL) is investigating the feasibility of providing High-Assay Low Enriched Uranium (HALEU) by recovering Highly Enriched Uranium (HEU) from spent nuclear fuel (SNF) and downblending the HEU with natural, low-enriched, or depleted uranium (NU, LEU, or DU). That investigation identified that significant quantities of HALEU could be produced with a Hybrid ZIRCEX process to further advanced reactor research while reducing INL's environmental liability. Many new advanced reactor designs are being developed with improved safety, efficiency, and economics. Most require nuclear fuel with U-235 enrichment between 5% and 20%, which is defined as HALEU. Although technically feasible, there is no current domestic capability to make HALEU in the US. While it is anticipated that industry will provide HALEU through commercial enrichment once advanced reactors mature, until a market is established, an interim source of HALEU is needed to enable research and demonstration. An interim source of HALEU could be provided by recovering the HEU in some DOE-managed SNF and downblending it to between 5% and 20% using the Hybrid ZIRCEX Process. ZIRCEX is a dry head-end process to remove cladding (zirconium or aluminum) from SNF. Drying of the SNF is needed prior to introduction to ZIRCEX. After cladding removal, uranium and fission products in the bed material are oxidized and elutriated. In a Hybrid ZIRCEX process, the HEU process stream is sent to a very compact, modular solvent extraction process for uranium purification. The uranyl nitrate from solvent extraction is denitrated and calcined to produce a HALEU product ready for fuel fabrication. The long-lived fission products from solvent extraction are immobilized in glass using a small in-can melt. By removing the cladding before dissolution, the volume of glass waste is reduced by a factor of up to 300 times (when compared to not removing the cladding from the fuel). Other radioactive wastes from the process could be disposed of as Class A or Class B low-level waste. In addition to producing HALEU for advanced reactors, consumption of SNF would reduce DoE's environmental liability. Incorporating low-enriched spent fuel (versus exclusive use of LEU, DU, or NU) into the downblending scheme could reduce the SNF inventory even further. Several downblending and enrichment scenarios using INL fuels were investigated in this study. The results of this analysis show that the Hybrid ZIRCEX process could make a significant contribution to needed HALEU feedstock and reduce the environmental liability of managing SNF at INL. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Contaminant Migration Modeling to Support the In Situ Decommissioning of Hardened Nuclear Facilities - 20215

The decommissioning of hardened nuclear facilities provides a unique challenge in balancing current worker dose with exposure to future receptors and media. In situ decommissioning (ISD) is a cost-effective and safe option for the closure of such facilities but requires a detailed understanding of the threat that radioactive and hazardous constituents remaining in the facilities pose to groundwater and surrounding surface water. The Savannah River Site (SRS) extensively employs contaminant migration models within the decommissioning process to develop site-specific removal, grouting, and monitoring strategies in support of safe and effective final end states of nuclear facilities. This work discusses the use of contaminant migration modeling in the ISD process at SRS, including specific examples from the closure of P-Reactor and R-Reactor, Building 235-F, F-Canyon and F/H Laboratory Complex. Migration models developed during the closure of P-Reactor and R-Reactor were divided into four source areas: the reactor vessel, the process area, the disassembly basin, and the purification wing. Each source area was assigned a specific inventory and migration pathway and was then subject to varying hypothetical removal, capping, and grouting schemes. Modeling indicated that groundwater Maximum Concentration Limits (MCLs) might be exceeded if no action was taken for eleven and ten constituents of concern (COCs) at P-Reactor and R-Reactor, respectively. These exceedances could occur in as few as 200 to 500 years. Alternatively, the migration modeling demonstrated that the selected ISD actions reduced contaminant mobility which allowed for significant radioactive decay and resulted in fewer predicted exceedances of groundwater MCLs (five COCs for P-Reactor; eight COCs for R-Reactor). In response to the modeling results for P-Reactor and R-Reactor, effectiveness monitoring programs were developed to target contaminants, specific to each reactor, that may migrate to groundwater. Contaminant migration modeling also revealed that roof collapse was a large factor in the release of COCs to the environment, giving rise to roof improvements, and an inspection and vegetation control program to ensure roof stability over time. Contaminant migration modeling is also aiding in the closure planning for hardened facilities in F Area. Building 235-F housed the Actinide Billet Line, which produced Np-237 billets for irradiation in SRS reactors, and the Plutonium Fuel Form (PuFF) facility that produced Pu-238 heat sources for the space program. As a result of these missions, areas within Building 235-F contain considerable residual amounts of both Pu-238 and Np-237. Contaminant migration modeling of Building 235-F was originally performed in 2012 to identify the feasibility of ISD and the amount of radioactive material removal required to prevent the exceedance of groundwater MCLs. The original model indicated that a 60% reduction in the PuFF facility Pu-238 inventory could keep groundwater concentrations below standards, while Np-237 did not pose a threat to groundwater. However, updates to the model with an emphasis on source impact pathways revealed that, due to the orientation of the source areas relative to groundwater flow, no amount of reasonable removal of Pu-238 would keep groundwater concentrations below MCLs and that Np-237 could be a large contributor to localized MCL exceedances. With this insight, the refined 2019 model is being used to assist in the development of grouting plans specific to each facility source area, where bentonite may be utilized to slow the migration of Pu-238 and its daughter products from the PuFF facility and a reducing grout may decrease Np-237 transport by ensuring the nuclide remains in the less mobile +IV oxidation state. The beginning phases of contaminant migration modeling are underway for F-Canyon and associated facilities using lessons learned from the Reactors and Building 235-F. A contaminant migration pathway, similar to the pathway used for the reactor vessels, is being developed for the hot and warm canyons. The F-Canyon inventories are being spatially refined to identify specific and localized source zones, comparable to Building 235-F, that may impact groundwater. Contaminant migration modeling provides vital input within the ISD process, from facility investigation to post closure effectiveness monitoring, making it a valuable tool in the development of safe and effective end states for hardened nuclear facilities at SRS. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

A Higher Fidelity Cost Analysis of Wind and Uranium from Seawater Acquisition symbiotic Infrastructure

The vastness of the worlds oceans makes the low concentration of naturally existing uranium, 3.3 ppb, appealing as a recoverable natural resource. Although ample uranium is available currently through conventional means, uranium from seawater is a widely studied topic due to the long term security it can assure the nuclear fuel cycle. It is important to mention that seawater uranium is not meant to act as a direct competitor to terrestrial uranium, but instead create a cost ceiling that provides economic stability and potential savings. Additionally, seawater uranium could circumvent some of the environmental impacts associated with recovery of any land based resource. The system currently studied by a nation-wide consortium of national lab and university partners involves the passive recovery of uranium using polymer based adsorbents. High Density Polyethylene fibers undergo a radiation induced graft co-polymerization process involving amidoxime, to attract uranium, and a polar co-monomer to increase hydrophilicity. After further chemical conditioning adsorbent fibers are braided into 60 meter strands for marine deployment. Braids are moored to the ocean floor for the duration of their soaking campaign. After sufficient seawater exposure they are winched up so the adsorbed uranium may be eluted off the braids. This deployment and elution process is repeated multiple times before the adsorbent's ultimate disposal, where its lifetime is dictated by the degradation it suffers with each re-cycle. Previous economic analyses have identified the adsorbent production and mooring as the most expensive components of the recovery process. Therefore, a later publication by Picard et al. proposed an alternative deployment method in an effort to reduce seawater uranium production cost. The Wind and Uranium from Seawater Acquisition symbiotic Infrastructure (WUSABI) couples uranium recovery with off shore wind to reduce mooring capital cost. The uranium harvesting structure attached to the underwater base of the wind turbine supports elution tanks and an adsorbent pulley system. This allows for a nearly autonomous mooring and elution procedure as adsorbent is in constant motion, reaching the elution tanks at the end of its soaking campaign. Although the original publication did include a short economic section, further analysis was carried out to provide a higher fidelity production cost estimate for uranium recovered by this system. Sufficient detail was provided regarding the capital and operating costs of the novel uranium harvesting units. The adsorbent production and uranium elution and purification steps however were estimated by simpler methods, for example applying a previously reported cost for producing a unit mass of adsorbent. This methodology, although satisfactory for the zeroth-order approximation desired in the original publication, overlooks many of the complexities and feedbacks embedded in the full cost model, which are necessary to provide a more accurate estimate. Perhaps more importantly, given the constant updates regarding adsorbent synthesis and performance this value quickly become outdated. Syncing an improved estimate for the WUSABI harvesting units with the existing economic analysis for adsorbent production and elution, which evolves in tandem with the technology, provides a more robust production cost estimate that will remain current over time. In addition to the updates and higher level of detail provided in this independent analysis, consistency in methodology allows for accurate comparison of uranium production costs across recovery methods. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2017

Engineering-Scale Development/Demonstration of Zirconium Recovery from Used Duel Cladding

Recognizing that the zirconium (Zr) in used nuclear fuel (UNF) cladding is the second largest mass in UNF assemblies, process development studies for the recovery and purification of Zr were initiated and have been in progress at Oak Ridge National Laboratory (ORNL) for the last several years as a major element of waste minimization for advanced nuclear fuel recycle. Together with uranium recovery and recycle, Zr recovery would divert >90% (wt.) of the UNF assemblies from emplacement in a geologic repository by converting the Zr to low-level waste or reuse/recycle.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Computational Studies of Gas Phase Reactions of Zirconium Tetrachloride (ZrCl{sub 4}) and Metal Chloride Impurities

Zirconium-based cladding materials that surround used nuclear fuel (UNF) rods represent a significant fraction of the total amount of high level waste that must be disposed of in long term geological repositories. The decontamination, recovery and reuse of these materials would significantly reduce the high level waste associated with light water reactors and could provide a new source of hafnium free zirconium for the fabrication of new nuclear components. One part in the strategy to recycle zirconium from used cladding materials involves developing a computational modeling (DFT) protocol to understand the reactions between ZrCl{sub 4} and metal chloride impurities that may occur in the gas phase during sublimation. Several of the metal chloride species present are known to form Lewis acid/base adducts in the gas phase and development of purification protocols require investigation into the properties of these ZrCl{sub 4} contaminates.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Cost Analysis of Wind and Uranium from Seawater Acquisition symbiotic Infrastructure using Shell Enclosures

At current consumption rates, global conventional reserves of uranium, estimated to be 7.6 million tonnes, could be depleted in a little over a century. As these reserves decrease, extraction of uranium is expected to shift to lower quality sites, resulting in higher extraction costs and greater environmental impact. Fortunately, the ocean contains approximately 4.5 billion tonnes of uranium, as uranyl ions at low concentrations of about 3 ppb. Uranium from seawater is a widely studied topic due to the long term security it can assure the nuclear fuel cycle. Additionally, harvesting uranium from seawater could reduce some of the environmental impacts associated with the recovery of land-based uranium. Seawater uranium is meant to create a cost ceiling that provides economic stability and potential savings, not to act as a direct competitor to terrestrial uranium. A recent review of uranium recovery technologies by found uranium adsorption by chelating polymers to be the most promising in terms of adsorbent capacity, environmental footprint, and cost. In this technology, chelating polymers are deployed in seawater and remain submerged until the amount of captured uranium approaches the adsorption capacity. Then, an elution bath is used to strip metal ions, such as uranium, off the adsorbent polymer. A polymer may be immersed in a number of elution baths before it is regenerated by an alkali wash to free its functional groups, allowing for the reuse of the polymer. The output from the elution process undergoes purification and precipitation typical for mined uranium to produce yellowcake. Offshore systems for the extraction of uranium from seawater have been developed since the early 2000's. The system currently studied by a nationwide consortium of national laboratory and university partners utilizes continuous adsorbent polyethylene fibers that are braided around a porous polypropylene float which can be made into long lengths. However, because this deployment scheme requires the adsorbent be brought to a mother-ship for the elution process and redeployed afterward, it has significant practical and economic deployment challenges. Detailed economic analysis have identified the adsorbent production and mooring as the most expensive components of the recovery process. Based on these results, proposed an alternative deployment method that is attached to an offshore wind turbine and continuously takes the adsorbent, that is fabricated into a belt, from the ocean through an elution process and before submerging it back into the sea. However, it has been found that adsorbents with high tensile strength and durability often have very low uranium adsorption properties. Hence, the Wind and Uranium from Seawater Acquisition symbiotic Infrastructure (WUSABI) previously studied, which requires the adsorbent into a belt held in tension, could face difficulties in an ocean environment. Instead, the design, further referred to in this paper as WUSABI-Koosh, utilized a two-part system to decouple the mechanical and chemical needs of an adsorbent for seawater harvesting of uranium. In these designs, the uranium adsorbent material with high adsorbent capacity is enclosed in a hard permeable outer shell with sufficient mechanical strength and durability for use in an offshore environment and chemical resilience against elution treatments. The integration of the design of a uranium harvesting system into an offshore wind turbine tower is pursued because the development of offshore wind or uranium harvesting by themselves bears a high capital cost for the structures, but if the mooring function can be shared, the overall cost for each will be lower. Furthermore, initial design analysis and prototyping of these symbiotic systems has proven they are technically feasible. This paper presents preliminary cost analysis for comparison to uranium recovery via alternative deployment schemes.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2017

Criticality Scoping Calculations for a Multi-Function Dissolver Insert

As one of its missions, the Savannah River Site seeks to be a leader in the safe and productive disposition of special nuclear materials. One of the premier programs is the disposition of used nuclear fuel, primarily from research reactors, through dissolution, purification, and down blend into low enriched uranium for commercial power reactors. Prior to use, this fuel contained highly enriched uranium (HEU). In an effort to advance this expertise, this study has been performed to demonstrate the feasibility to disposition a number of special geometry HEU fuels. These fuels come from a number of different domestic and foreign research reactors which had geometries considerably different from round or square bundles of plates or rods. As such, they cannot be dispositioned in current H-Canyon dissolver inserts so a new, multi-function dissolver insert may be considered. While this is not a current mission, it is believed that this study's scoping calculations demonstrate the feasibility of the site taking such a mission. At the H-Canyon facility, a dissolver is a large stainless steel pot, loaded with a nitric acid solution into which uranium-alloy fuel is dissolved. The dissolver uses a device called an insert to feed the fuels to the acid solution in a controlled manner. Most fuel forms fit into one of the existing inserts which include one with ten {approx}5-inch diameter circular wells, one with four {approx}4.5 inch by {approx}15 inch slab wells, and one specifically designed for the Oak Ridge National Laboratory High Flux Isotope Reactor (HFIR). This study considered five bounding fuel candidates for scoping the potential new insert. The fuels include a generic annular plate fuel (APF), the Tower Shield Reactor (TSR) fuel elements, the Reactor A Haute Flux (RHF) core, Knolls Atomic Power Laboratory (KAPL) reactor fuel, and SLOWPOKE cores. These are expected to bound a range of proposed fuel candidates. Scoping studies were made to establish geometric restrictions for the new dissolver insert and to determine if these fuels could safely be dissolved. Homogenous models were used for the scoping study due to time constraints. Computational modeling was performed using KENOVI in the SCALE 6.1 code package.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Chapter 3: Nanostructured Iron Oxide Hybrid Composites as Heterogeneous Fenton-Like Catalyst for Remediation of Persistent Organic Pollutants

In water purification research and technologies, heterogeneous Fenton catalysts are evolving as an excellent material recently. A summary of present research on heterogeneous Fenton/Fenton-like processes is essential for developing novel and high-efficiency organic wastewater treatment methods. In this chapter, we discussed the basics of Fenton reactions taking the example of iron oxides and their different composites, such as bimetallic composites, composites with carbon compounds, and Fe metal-organic frameworks (MOFs). The important operating parameters (solution pH, H 2 O 2 dosage, amount of catalyst used, and reaction temperature) in heterogeneous Fenton-like reactions are discussed. Strategies to increase the rate and activities of heterogeneous Fenton catalysts by using energy sources are presented. The mechanistic aspects of heterogeneous Fenton processes and reusability of catalysts along with their application in real wastewater plants are summarized and discussed. Finally, we concluded with the scope and future prospects of heterogeneous Fenton catalysis research to motivate the readers to construct novel and efficient heterogeneous Fenton/Fenton-like systems for industrial applications.

carbon composites↗
DOE OSTI2018

A Highly Crystalline Fused‐Ring n‐Type Small Molecule for Non‐Fullerene Acceptor Based Organic Solar Cells and Field‐Effect Transistors

Abstract N‐type organic small molecules (SMs) are attracting attention in the organic electronics field, due to their easy purification procedures with high yield. However, only a few reports show SMs that perform well in both organic field‐effect transistors (OFETs) and organic solar cells (OSCs). Here, the synthesis and characterization of an n‐type small molecule with an indacenodithieno[3,2‐b]thiophene (IDTT) core unit and linear alkylated side chain (C 16 ) (IDTTIC) are reported. Compared to the state‐of‐the‐art n‐type molecule IDTIC, IDTTIC exhibits smaller optical bandgap and higher absorption coefficient, which is due to the enhanced intramolecular effect. After mixing with the polymer donor PBDB‐T, IDTIC‐based solar cells deliver a power conversion efficiency of only 5.67%. In stark contrast, the OSC performance of IDTTIC improves significantly to 11.2%. It is found that the superior photovoltaic properties of PBDB‐T:IDTTIC blends are mainly due to reduced trap‐assisted recombination and enhanced molecular packing coherence length and higher domain purity when compared to IDTIC. Moreover, a significantly higher electron mobility of 0.50 cm 2 V −1 s −1 for IDTTIC in OFET devices than for IDTIC (0.15 cm 2 V −1 s −1 ) is obtained. These superior performances in OSCs and OFETs demonstrate that SMs with extended π‐conjugation of the backbone possess a great potential for application in organic electronic devices.

Song, Xin↗
DOE OSTI2021

Inorganic Halide Perovskitoid TlPbI 3 for Ionizing Radiation Detection

Room temperature semiconductor detector (RTSD) materials for γ-ray and X-ray radiation are in great demand for the nonproliferation of nuclear materials as well as for biomedical imaging applications. Halide perovskites have attracted great attention as emerging and promising RTSD materials. In this contribution, the material synthesis, purification, crystal growth, crystal structure, photoluminescence properties, ionizing radiation detection performance, and electronic structure of the inorganic halide perovskitoid compound TlPbI 3 are reported on. This compound crystallizes in the ABX 3 non-perovskite crystal structure with a high density of d = 6.488 g·cm –3 , has a wide bandgap of 2.25 eV, and melts congruently at a low temperature of 360 °C without phase transitions, which allows for facile growth of high quality crystals with few thermally-activated defects. High-quality TlPbI 3 single crystals of centimeter-size are grown using the vertical Bridgman method using purified raw materials. A high electrical resistivity of ≈10 12 Ω·cm is readily obtainable, and detectors made of TlPbI 3 single crystals are highly photoresponsive to Ag K α X-rays (22.4 keV), and detects 122 keV γ-rays from 57 Co radiation source. The electron mobility-lifetime product µ e τ e was estimated at 1.8 × 10 –5 cm 2 ·V –1 . A high relative static dielectric constant of 35.0 indicates strong capability in screening carrier scattering and charged defects in TlPbI 3 .

36 MATERIALS SCIENCE↗
DOE OSTI2022

Stabilization of the Polar Structure and Giant Second-Order Nonlinear Response of Single Crystal γ-NaAs 0.95 Sb 0.05 Se 2

The dearth of suitable materials significantly restricts the practical development of infrared (IR) laser systems with highly efficient and broadband tuning. Recently, γ-NaAsSe 2 is reported, and it exhibits a large nonlinear second-harmonic generation (SHG) coefficient of 590 pm V —1 at 2 µm. However, the crystal growth of γ-NaAsSe 2 is challenging because it undergoes a phase transition to centrosymmetric δ-NaAsSe 2 . Herein, the stabilization of non-centrosymmetric γ-NaAsSe 2 by doping the As site with Sb, which results in γ-NaAs 0.95 Sb 0.05 Se 2 is reported. The congruent melting behavior is confirmed by differential thermal analysis with a melting temperature of 450 °C and crystallization temperature of 415 °C. Single crystals with dimensions of 3 mm × 2 mm are successfully obtained via zone refining and the Bridgman method. The purification of the material plays a significant role in crystal growth and results in a bandgap of 1.78 eV and thermal conductivity of 0.79 Wm —1 K —1 . The single-crystal SHG coefficient of γ-NaAs 0.95 Sb 0.05 Se 2 exhibits an enormous value of |d 11 | = 648 ± 74 pm V —1 , which is comparable to that of γ-NaAsSe 2 and ≈20× larger than that of AgGaSe 2 . Importantly, the bandgap of γ-NaAs 0.95 Sb 0.05 Se 2 (1.78 eV) is similar to that of AgGaSe 2 , thus rendering it highly attractive as a high-performing nonlinear optical material.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Nanoporous Atomically Thin Graphene Membranes for Desalting and Dialysis Applications

Dialysis is a ubiquitous separation process in biochemical processing and biological research. State-of-the-art dialysis membranes comprise a relatively thick polymer layer with tortuous pores, and suffer from low rates of diffusion leading to extremely long process times (often several days) and poor selectivity, especially in the 0–1000 Da molecular weight cut-off range. Here, the fabrication of large-area (cm 2 ) nanoporous atomically thin membranes (NATMs) is reported, by transferring graphene synthesized using scalable chemical vapor deposition (CVD) to polycarbonate track-etched supports. In this study, after sealing defects introduced during transfer/handling by interfacial polymerization, a facile oxygen-plasma etch is used to create size-selective pores (≤1 nm) in the CVD graphene. Size-selective separation and desalting of small model molecules (≈200–1355 Da) and proteins (≈14 000 Da) are demonstrated, with ≈1–2 orders of magnitude increase in permeance compared to state-of-the-art commercial membranes. Rapid diffusion and size-selectivity in NATMs offers transformative opportunities in purification of drugs, removal of residual reactants, biochemical analytics, medical diagnostics, therapeutics, and nano-bio separations.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Polymers with Side Chain Porosity for Ultrapermeable and Plasticization Resistant Materials for Gas Separations

Polymer membranes with ultrahigh CO 2 permeabilities and high selectivities are needed to address some of the critical separation challenges related to energy and the environment, especially in natural gas purification and postcombustion carbon capture. However, very few solution-processable, linear polymers are known today that access these types of characteristics, and all of the known structures achieve their separation performance through the design of rigid backbone chemistries that concomitantly increase chain stiffness and interchain spacing, thereby resulting in ultramicroporosity in solid-state chain-entangled films. Herein, the separation performance of a porous polymer obtained via ring-opening metathesis polymerization is reported, which possesses a flexible backbone with rigid, fluorinated side chains. This polymer exhibits ultrahigh CO 2 permeability (>21 000 Barrer) and exceptional plasticization resistance (CO 2 plasticization pressure > 51 bar). Compared to traditional polymers of intrinsic microporosity, the rate of physical aging is slower, especially for gases with small effective diameters (i.e., He, H 2 , and O 2 ). Overall, this structural design strategy, coupled with studies on fluorination, demonstrates a generalizable approach to create new polymers with flexible backbones and pore-forming side chains that have unexplored promise for small-molecule separations.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Heterologous Assembly of Pleomorphic Bacterial Microcompartment Shell Architectures Spanning the Nano- to Microscale

Many bacteria use protein-based organelles known as bacterial microcompartments (BMCs) to organize and sequester sequential enzymatic reactions. Regardless of their specialized metabolic function, all BMCs are delimited by a shell made of multiple structurally redundant, yet functionally diverse, hexameric (BMC-H), pseudohexameric/trimeric (BMC-T), or pentameric (BMC-P) shell protein paralogs. When expressed without their native cargo, shell proteins have been shown to self-assemble into 2D sheets, open-ended nanotubes, and closed shells of ≈40 nm diameter that are being developed as scaffolds and nanocontainers for applications in biotechnology. Here, by leveraging a strategy for affinity-based purification, it is demonstrated that a wide range of empty synthetic shells, many differing in end-cap structures, can be derived from a glycyl radical enzyme-associated microcompartment. The range of pleomorphic shells observed, which span ≈2 orders of magnitude in size from ≈25 nm to ≈1.8 µm, reveal the remarkable plasticity of BMC-based biomaterials. In addition, new capped nanotube and nanocone morphologies are observed that are consistent with a multicomponent geometric model in which architectural principles are shared among asymmetric carbon, viral protein, and BMC-based structures.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Porphyrin Covalent Organic Framework (POF)-Based Interface Engineering for Solar Steam Generation

Through replacement of traditional energy sources with sunlight as the heat source, solar steam generation has emerged as a promising technology for water purification and residuals management. Despite significant efforts to develop efficient photothermal materials for solar steam devices, challenges associated with scalable fabrication of high-performance materials remain. Moreover, most existing methods cannot be easily engineered to produce steam-generating devices with both arbitrary control over shape and high photothermal efficiency. Herein, a flexible porphyrin organic framework (POF)-based interface engineering method is introduced to produce high-performance solar steam generators. POFs, a recently discovered class of materials, are demonstrated to grow readily on a diverse range of porous substrates, including membranes, fabrics, sponges, and wood. Wood@POF exhibits particularly strong performance, achieving approximate to 80% light-to-steam conversion efficiency. This study demonstrates a universal, simple, and scalable interface engineering strategy for the fabrication of solar steam generators based on POF materials.

36 MATERIALS SCIENCE↗