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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 361 records · Page 20
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Full-Scale Retrieval and Transfer System Mockup for Pneumatic Conveyance of High-Level Calcine Waste - 19563

The Idaho National Laboratory reprocessed spent nuclear fuel from 1953 to 1994. The reprocessing used solvent extraction to separate uranium, resulting in a high-level radioactive liquid waste. A calcination process was used to convert the liquid waste to a high-level radioactive granular solid, referred to as calcine. The calcine is stored in stainless-steel bins contained within six concrete vaults called the Calcined Solids Storage Facilities (CSSFs). The Calcine Disposition Project is tasked with demonstrating the ability to retrieve the calcine from the bins of CSSF 1 and transferring it to the bins of CSSF 6. CSSF 1 was constructed to be a final calcine repository, with no forethought of retrieval. It represents the greatest technical hurdle to calcine extraction of all the CSSFs. The challenges inherent with accessing the calcine, safely extracting it, and securely transporting it are copious and complex. To overcome those challenges in a timely and objective manner, a full-scale mockup will be constructed. The mockup includes a full-size model of a concentric group of bins from CSSF 1 and a storage vessel representing CSSF 6. Mockup equipment will include an air compressor, filtration, blower/vacuum, and retrieval devices. In addition to defining process parameters, the mockup testing scope will demonstrate the amount of residual calcine that can be expected to be in the bins following extraction. This will be key to preparing for closure of CSSF 1 following removal of the calcine. Once testing is complete, the mockup will constitute a model of the calcine extraction and transport system and will be prepared to fulfill a second advantageous objective: the operators that will carry out the calcine transfer from CSSF 1 to CSSF 6 will be trained using the mockup. The control system will be identical, the major pieces of equipment will be representative, and the procedures will be analogous; hence, high-fidelity training can take place at the mockup. Once radiological operations commence at CSSF 1, the mock-up's commission will still not be completed. If any problems arise during calcine transport, the mockup will be ready to serve as a simulator to more effectively reproduce dilemmas and deduce solutions. The mock-up's mission spans the duration of the project. Its usefulness begins with helping to identify process criteria. Operations personnel will benefit greatly from its use as a training tool. And finally, the immediate availability of a simulator to reproduce problems encountered during radiological extraction and transfer will be invaluable. The knowledge gained during this process will also be of continued benefit as future work is done to remove various types of calcine from the remaining CSSFs. The methodology of safe and secure pneumatic transfer of radioactive material may be of benefit to others throughout the U.S. Department of Energy complex. (authors)

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

A Dose Assessment Model for Radioactive Waste Exposed by Gully Erosion at West Valley - 20513

The Western New York Nuclear Service Center (WNYNSC), located approximately 48 km south of Buffalo, New York, is the site of a former nuclear fuel reprocessing and radioactive waste disposal facility. Spent nuclear fuel was processed there from 1966 to 1972, leaving behind radioactive and chemical wastes in two disposal areas and a waste tank farm. Site operations also resulted in releases of radioactivity to site soils, groundwater, and to surface waters draining the site. The New York State Energy Research and Development Authority (NYSERDA) and the U.S. Department of Energy (DOE) are collaborating in a process of decision making for decommissioning those facilities remaining at the WNYNSC following the completion of Phase 1 decommissioning. Neptune and Company, Inc. (Neptune) was contracted to develop a probabilistic performance assessment (PPA) computer model to assist the agencies in this process. The PPA Model includes a contaminant transport component focusing on the movement of contaminants within and among environmental media including groundwater and surface water transport, contaminant translocation by plants and animals, diffusion, and erosion. The model also includes evaluation of potential exposure and health effects for a Resident Farmer exposure scenario, where the Resident Farmer represents a critical group, described as that group of individuals reasonably expected to receive the greatest exposure to residual radioactivity for any applicable set of circumstances. The West Valley Site is located in the glaciated Allegheny Plateau region of western New York State. The waste reprocessing and disposal areas were constructed on a relatively fat area of plateau dissected by drainages of Buttermilk Creek, including Erdman Brook, Franks Creek, and Quarry Creek. An important aspect of the contaminant transport component of the PPA model is consideration of erosive processes such as slumping of the stream slopes and the advance of gullies from these streams. These erosion processes remove material from the plateau, growing the size of the creek valleys and making them wider and deeper. Of particular interest for the impact of erosion is radioactive waste contained in the Nuclear Regulatory Commission (NRC)-Licensed Disposal Area (NDA), the New York State-Licensed Disposal Area (SDA), and residual radiological inventory in the underground storage tanks at the Waste Tank Farm (WTF). The PPA Model is organized around geographically-defined facilities which were constructed upon the plateau, including the NDA, SDA, and WTF. Ongoing stream erosion processes will potentially transfer radioactive waste and residual inventory from these facilities to the ground surface on adjacent hillslope areas where erosion has breached the facility. Hence, it is important to evaluate the consequences of potential exposures to a Resident Farmer on the hillslopes below a breached facility. Two interrelated aspects of the dose assessment model related to hillslope exposure are discussed: 1) representation of the physical processes related to transport of radionuclides from facilities onto the hillslopes, and from the hillslopes into adjoining creeks where contaminated material migrates downstream with surface water and sediment, and; 2) adaptation of the activities associated with the Resident Farmer scenario to assess potential exposures to contamination in the hillslope areas. This discussion will cover the conceptual basis of the hillslope exposure and transport models, and also implementation in the PPA computer model. (authors)

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

A Degradation Model and Testing Protocol for Metallic Waste Forms

Metallic waste forms developed to immobilize high-level radioactive wastes generated in advanced (closed) fuel cycles are being evaluated as part of the US DOE Fuel Cycle Research and Development program. The electrometallurgical reprocessing of used metallic fuels to recover U and Pu generates two waste streams: salt waste comprised of electro-refiner salt contaminated with oxidized fission products and metallic waste comprised of activated cladding and metallic fission products that were not oxidized under the processing conditions. Metallic waste forms can be made by directly melting the cladding and fuel wastes with small amounts of trim metals to form a multiphase alloy material in which radionuclides and other waste constituents are immobilized within chemically durable intermetallic and iron-based solid solution phases. Some components of the waste streams serve as important constituents of phases comprising the waste form, such as Mo and Zr, which form Fe{sub 2}Mo and Fe{sub 2}Zr intermetallic phases. Other waste constituents are accommodated as substitutional contaminants, such as Tc within the Fe{sub 2}Mo intermetallic and U/TRU within the Fe{sub 2}Zr intermetallic. Sufficient amounts of these host phases must be available to accommodate all radionuclides in the waste stream. The compositional flexibility of the host phases to incorporate waste components is being utilized to formulate metal waste forms that will accommodate variations in waste streams compositions that will occur over long-term fuel reprocessing operations. Small amounts of trim metals can be added to tailor the phase assembly to match the waste stream composition and facilitate waste form processing. For example, additional Zr may be added during waste form production to ensure all U/TRU is sequestered in a durable Fe{sub 2}Zr host phase and trim Cr, Mo, and Ni can be added to supplement the amounts provided by the steel cladding to improve the durability of a metallic waste form. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Multicomponent adsorption of radioactive iodine and krypton using ETS-10 supported carbon nano-polyhedrons

Over 13% of the world's energy demand is supplied by nuclear power plants in 31 countries. This demand is growing; by 2030 the IAEA predicts a 100% increase in the demand for nuclear energy. The increase in nuclear energy production imposes a growing strain on the uranium fuel supply. At current consumption rates the world's supply of uranium is projected to last ∼100 years. Following the IAEA's prediction for increased nuclear energy, uranium supplies will dwindle even more quickly. However, this consequence can be limited as ∼95% of uranium is left in used nuclear fuel (UNF). Reprocessing of UNF will utilize this remaining uranium and extend the lifetime of the world's uranium supply. One of the concerns with nuclear reprocessing is that it frees radionuclides created by the fission of uranium to be released into the atmosphere through off-gas. The major constituents of off-gas include: {sup 85}Kr, {sup 129}I, {sup 14}C, {sup 3}H, and various Xe isotopes. {sup 85}Kr and {sup 129}I present the largest threat due to their harmful environmental effects and abundance. They are the most practical contaminants to be removed to meet requirements for radionuclide release. Removal of these contaminants has been a challenge since the 1950's, and several methods including cryogenic distillation, fluorocarbon absorption, wet scrubbing, and adsorption have been employed. Currently, adsorption on porous solid sorbents shows the most promising results. Until now, several sorbents such as silver impregnated mordenite, silver Engelhard titanosilicate 10, chalcogen based aerogels, and metal organic framework, MOFs have been investigated for this purpose. Despite these advances, development of economical and highly active sorbents for complete removal and immobilization of volatile radioactive gases remains a key challenge in realizing advanced fuel cycles. This work is focused on the multicomponent of {sup 85}Kr and {sup 129}I from a simulated off-gas with an Engelhard titanosilicate 10 (ETS-10) supported carbon nano-polyhedron sorbent. Adsorption studies were performed to determine iodine sorption capacity as a function of temperature and iodine inlet concentration. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Overview of the Salt at Wall Thermal Exchanges (SWATH) Experiment

Molten Salt Reactors have attracted increased attention in recent years because of the design and safety possibilities offered by the use of a liquid fuel. The European project SAMOFAR1 (2015-2019) is currently coordinating the research efforts on the Molten Fast Salt Reactor (MSFR) concept in Europe. The MSFR is a fast-spectrum breeder reactor with a large negative power coefficient that can be operated in a Thorium fuel cycle. The fast spectrum allows the reduction of the reprocessing requirements and a better reactor breeding ratio. Other advantages of the MSFR design are the possibility for actinide burning and extending fuel resources, on-line fuel loading and reprocessing and the use of novel passive safety systems such as the fuel salt draining system. The reference MSFR concept is a 3000 MW(th) reactor with three different circuits: the fuel circuit, the intermediate circuit and the power conversion system. The main components of the fuel circuit are the fuel salt which serves as fuel and coolant, the core cavity, the inlet and outlet pipes, the gas injection system (not shown in the figure), the salt-bubble separators, the fuel heat exchangers and the pumps. A lithium fluoride salt has been selected as the fuel matrix. The initial composition (non-irradiated) of the MSFR fuel salt is a mixture of a lithium fluoride, thorium fluoride salts and actinides fluoride (LiF-ThF{sub 4}-{sup 233}UF{sub 4} or LiF-ThF{sub 4}- {sup enr}UF{sub 4}-(Pu-MA)F{sub 3}), with the proportion of LiF fixed at about Development of accurate numerical models for the MSFR can only be accomplished by the use of a multiphysics approach which allows taking into account the coupling between relevant neutronics, thermal hydraulics and thermomechanics phenomena. As it is discussed in the next section, such a reactor model requires further progress on the knowledge of molten salt thermal hydraulics, in particular regarding the accuracy of the fuel salt Computational Fluid Dynamics (CFD) models. The Salt at Wall: Thermal Exchanges (SWATH) experiment is being designed to contribute to this particular point. SWATH is one of the experimental activities of the European H2020 SAMOFAR project and includes both experimental and numerical developments. The main objectives of SWATH are to: (i) Improve molten salt numerical models used for design and safety studies and (ii) Demonstrate the working principles of the cold plug device. The cold plug device is a key safety component of the salt draining system of the MSFR. The cold plug is intended to passively melt during accidental conditions thus allowing the flow of the nuclear fuel from the core cavity to the fuel salt draining tanks by gravity forces. The fuel salt draining tanks are designed to ensure that adequate margins are obtained regarding the fuel cooling capabilities and the reactivity control. The next sections provide an overview of the current challenges in molten salt thermal hydraulics modeling and the SWATH experiment design. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Separation of Actinides from Lanthanides in Molten Fluorides by the Modulated-Current Electrolysis

Recently, the Molten Salt Reactor (MSR) is taking a renewed attraction as one of the reactor systems highlighted by the Generation IV International Forum. It is a non-classical (liquid fuel) system which can be operated as a breeder within the {sup 232}Th-{sup 233}U fuel cycle with the very interesting breeding factor. The necessary precondition for its proper functionality is the online reprocessing within the MSR's fuel cycle. A lot of effort is given to the development of the on-line pyrochemical reprocessing technology. Among the processes considered to be suitable for this process, the electrochemical separations are highlighted. Within this context, it is necessary to study electrochemical behavior of both fissile material and fission products in the molten salt systems and to develop the processes of their extraction from the system at various stages of the whole cycle. In this contribution, the study of the electrochemical separation of U from Gd in the LiF-CaF{sub 2} molten system by the modulated current electrolysis is presented. The study of MSR-related electrochemistry will continue as a part of the new Czech MSR R and D program. The knowledge of the electrochemical behavior of actinides and lanthanides in molten fluorides is necessary to the development of any electrochemical separation process. In our previous work, electrochemical characteristics of uranium, thorium, as well as that of several lanthanides, were measured in LiF-BeF{sub 2} (the basic melt considered for the MSR concept) and Be-free melts (LiF-NaFKF, LiF-NaF, LiF-CaF{sub 2}), and added to available literature data. If the fluoride carrier melt is electrochemically stable enough, it is possible to deposit the element of interest in its basic, metallic state (e.g. uranium in all types of fluoride melts tested) by the electrolysis. In the case of insufficient electrochemical stability of the fluoride melt in which the actinides and lanthanides are dissolved for the experiments, it is possible to use a reactive working electrode made of a material which creates alloys with the element to be deposited. An alloying reaction can shift the deposition potential in a favorable direction and therefore achieve the separation of an element from the system. The use of reactive electrode can also improve the adhesion of the deposit. When two or more electroactive species are present in the system at the same time, the efficiency of their separation by an electrolysis depends on how different the deposition potentials of these two species are. If we input certain modulation to the electrolytic current, we can improve overall result of an electrolytic experiment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Effect of Concentration, Temperature, and Interelectrode Gap on the Voltage Drop in an Electrochemical System of GdCl{sub 3}-LiCl-KCl

One of the most common conversations for those in the nuclear field and the public when discussing nuclear power is the issue of nuclear waste. When used nuclear fuel (UNF) is removed from a reactor, it contains fission products, some of which are long lived isotopes that will continue to irradiate for hundreds of years. One of several solutions to this issue is reprocessing, a process where minor actinides and other fission products are separated from still fissile and fertile materials. In the United States, there is no reprocessing of UNF at an industrial scale; however, one of the methods that universities and national labs are currently researching is known as pyro-processing. Pyro-processing utilizes electrochemistry in a molten salt environment to treat UNF and remove uranium and plutonium. The heart of the pyro-processing system is an electrorefining process, where most of the U in the UNF is selectively collected on a solid cathode in molten lithium-potassium chloride (LiCl-KCl) at high temperatures (723 K - 773 K). To model the behavior of materials in the electro-refiner, small bench-scale electrochemical experiments are being performed at many research universities and national labs to better understand material behavior. Within the electrochemical cell, there is a small resistance to current between the cathode and anode. This results in a voltage drop, which can affect the results of the experiments. This voltage drop varies based upon concentration of the salt, the temperature, the distance between the anode and cathode, and the scan rate. For this study, cyclic voltammetry will be performed on GdCl{sub 3}-LiCl-KCl at various temperatures (723 K - 798 K), various sample concentrations (1, 5, and 10 wt%), various scan rates (50, 100, 200, and 300 mV/s), and two different electrode spacing configurations. The resulting cyclic voltammograms will be compared for both electrode configurations to see the impact of interelectrode gap on the voltage drop.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

On the Use of Ionic Liquids for the Recycling of the Contaminated Material

Dealing with spent nuclear fuel (SNF) and other radioactive waste management is one of the main issues of the nuclear industry. Today, there is an increased demand for lower environmental affection and higher safety assurances within all industries dealing with radioactivity. The resulting increasing costs for dealing with radioactive waste, altogether with the pressure of public opinion, are strong motivation factors for the development of recycling schemes, softening the effects on the environment, and bringing economic advantages for the nuclear industry. This problem is even multiplied as a significant wave of decommissioning of nuclear facilities can be expected in the next decades and it makes the problem even more urgent. The Organisation for Economic Co-operation and Development/Nuclear Energy Agency estimates that 400 nuclear power plants will have to be decommissioned between 2014 and 2050 and it will generate more than 5 million tons of radioactive scrap metal. If the decommissioning of other nuclear installations is taken into account, almost 30 million tons of scrap metal contaminated by radionuclides will need to be handled?1 According to an International Atomic Energy Agency report?2 during the nuclear facility decommissioning process, only 3 % of the material is considered contaminated, but costs for its disposal can make over 60 % of overall costs. In both of mentioned areas - SNF reprocessing and material decontamination - possible benefits of ionic liquids (ILs) are studied worldwide because of their interesting physical-chemical properties and chemical behavior. ILs (described as room-temperature molten salts in older literature) are commonly described as salts with organic cations associated with organic or inorganic anions that are liquid below 373 K and fully consisting of ions and offer advantages of nonaqueous processes like a safer environment because of substituting volatile and flammable organic solvents or less serious criticality issues. At the same time, ILs beat high-temperature molten salt in comparing energy costs. Also some other of ILs' unique properties are definitely attractive for areas of SNF reprocessing and solid waste decontamination. Among other major advantages of ionic liquids, negligible vapor pressure, non-flammability, chemical and mechanical stability, tunability as well as good electric conductivity or wide electrochemical window are usually being mentioned. In this contribution, the experimental verification of some principles needed for the development of the decontamination of material contaminated with uranium and europium are presented as model cases.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Major Considerations on the Nuclear Material Accounting of PB-HTR Commercial Plants

The pebble bed high temperature reactor (PB-HTR) is a promising reactor type with inherent safety features in the next generation nuclear systems Based on the experience of studying and operating test PB-HTR (the HTR-10) for decades, China is constructing the first commercial nuclear power plant of modular PB-HTR in the world, i.e. the HTR-PM, which is expected to commission by the end of 2017. Since the PB-HTRs are quite different from the conventional PWRs in many aspects, including the large amount of fuel pebbles, the continuous refueling, the low heavy metal density in fuels, the safeguards of PB-HTR is important to investigate. Unfortunately, the research on this topic is insufficient by far, especially for the upcoming commercial PB-HTR plants. The PB-HTR has excellent features on proliferation resistance The graphite fuel pebbles with diameter of 6 cm are used as fuel elements, embedded with more than ten thousand TRISO coated fuel particles. This kind of fuel design provides more than one order of magnitude lower heavy metal concentration in fuel than other fuels. For the typical uranium loading of 7 grams in the HTR-PM, more than 10,000 fresh fuel pebbles (about 1.2 tons) are required to provide a significant quantity of indirect-use uranium with enrichment less than 20%. Moreover, the reprocessing of TRISO particle fuels is difficult because of the rigid SiC layer in the coating. Also, the possibility of utilizing materials from PB-HTR's spent fuels to product dirty bombs is quite low just due to the same reason as reprocessing. On the other hand, the large amount of fuel pebbles and the continuous on-line fueling introduce new challenges in nuclear safeguards and material accounting of PB-HTRs. This summary discusses the major considerations on the nuclear material accounting of PB-HTRs.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Preprocessing Analysis of the Multi Isotope Process Monitor

The Multi-Isotope Process (MIP) Monitor is a nondestructive monitoring methodology to detect diversion and tampering of nuclear material at a reprocessing facility. The system measures the gamma-emitting isotopes in different stages of the reprocessing cycle through passive gamma spectroscopy. The indicators of isotopic distributions are measured and evaluated in near-real time by comparing measured spectra to expected, normal process spectra through multivariate analysis. The MIP monitor framework is compatible with medium resolution gamma detectors that can be easily deployed in an existing facility. In FY15, the MIP monitor was deployed at the H-Canyon Facility at Savannah River National Laboratory (SRNL) to develop and evaluate analysis algorithms for future in situ deployment. This paper discusses recent results from analysis of the data collected at SRNL. New results examine the preprocessing step of spectral alignment that is associated with the MIP monitor.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Feasibility of a Breed-and-Burn Molten Salt Reactor

This work addresses the feasibility and characteristics of the breed and burn (B and B) cycle for Molten Salt Reactors. In the general B and B cycle, natural Uranium or Thorium is used as fuel for a reactor which breeds a sufficient amount to maintain criticality at equilibrium without reprocessing. Interest in applying this scheme to an MSR has developed recently For the MSR B and B scheme considered here, fresh salt containing fertile material is continually fed into the core while salt with the composition of the homogeneous fuel is expelled without reprocessing. Through an analysis of the equilibrium conditions and startup requirements, the feasibility of such a cycle operating with various salts is assessed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

INL Intern Poster Session

Characterization of off-gas constituents from nuclear fuel reprocessing can be measured using a residual gas analyzer. Characterization of this off-gas is important in a closed fuel cycle, where gasses such as Nox, Kr, I, and C-14 need to be removed from reprocessing plants to meet regulatory requirements. Both Kr and Xe have low chemical reactivity, making capture and separation difficult. Solid sorbents can be used to capture these gases as an alternative to cryogenic distillation, which is energy intensive and hazardous. AgZ-PAN and HZ-PAN are two solid sorbents that have demonstrated selective capture of Xe and Kr. A residual gas analyzer can be used to obtain breakthrough curves that demonstrate how well this adsorption and separation occur. The breakthrough curve is dependent on packing density, column size, and flow rate. Data obtained from the RGA was verified with a gas chromatographer.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Conductive Liquid Metal Vitrimer Composites for Reconfigurable and Recyclable Flexible Electronics

Liquid metal (LM) elastomer composites exhibit excellent functionality for stretchable electronics and wearables, but limited recycling and reuse pathways constrain their sustainable use. Here, to address these challenges amid growing concerns over electronic waste, a conductive LM–vitrimer composite is presented that enables recyclable and reconfigurable electronics. This soft and stretchable composite features uniformly distributed LM inclusions that enhance thermal conductivity by 6.53× and enable the formation of conductive traces with electrical self-healing, while the vitrimer provides structural restoration. The dynamic covalent bonds of the vitrimer matrix are leveraged for both reprocessing the composite and chemically recovering 94% of the LM. This liquid-state filler slightly reduces the vitrimer's stiffness to 2.63 MPa (≈20% lower), while maintaining its high stretchability (>135% strain) and thermal stability. It is further examined how ultrasonicated LM inclusions interact with the vitrimer matrix and demonstrate the composite's self-healing and recyclability through two distinct approaches: 1) thermomechanical reprocessing, which restores fragmented composites under heat and compression for circuit reconfiguration; and 2) chemical recycling, which recovers the embedded LM for reuse in fabricating new composites and redesigned circuitry. With the integration of recyclability and diverse functional capabilities, LM–vitrimer composites emerge as a promising material platform for sustainable, flexible electronics.

Han, Youngshang [Univ. of Washington, Seattle, WA ↗
DOE OSTI2024

Supersoft Norbornene–Based Thermoplastic Elastomers with High Strength and Upper Service Temperature

With over 6 million tons produced annually, thermoplastic elastomers (TPEs) have become ubiquitous in modern society, due to their unique combination of elasticity, toughness, and reprocessability. Nevertheless, industrial TPEs display a tradeoff between softness and strength, along with low upper service temperatures, typically ≤100 °C. This limits their utility, such as in bio-interfacial applications where supersoft deformation is required in tandem with strength, in addition to applications that require thermal stability (e.g., encapsulation of electronics, seals/joints for aeronautics, protective clothing for firefighting, and biomedical devices that can be subjected to steam sterilization). Thus, combining softness, strength, and high thermal resistance into a single versatile TPE has remained an unmet opportunity. Through de novo design and synthesis of novel norbornene-based ABA triblock copolymers, this gap is filled. Ring-opening metathesis polymerization is employed to prepare TPEs with an unprecedented combination of properties, including skin-like moduli (<100 kPa), strength competitive with commercial TPEs (>5 MPa), and upper service temperatures akin to high-performance plastics (≈260 °C). Here, the materials are elastic, tough, reprocessable, and shelf stable (≥2 months) without incorporation of plasticizer. Structure–property relationships identified herein inform development of next-generation TPEs that are both biologically soft yet thermomechanically durable.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Hybrid Ultra‐Microporous Materials for Selective Xenon Adsorption and Separation

Abstract The demand for Xe/Kr separation continues to grow due to the industrial significance of high‐purity Xe gas. Current separation processes rely on energy intensive cryogenic distillation. Therefore, less energy intensive alternatives, such as physisorptive separation, using porous materials, are required. Herein we show that an underexplored class of porous materials called hybrid ultra‐microporous materials (HUMs) affords new benchmark selectivity for Xe separation from Xe/Kr mixtures. The isostructural materials, CROFOUR‐1‐Ni and CROFOUR‐2‐Ni, are coordination networks that have coordinatively saturated metal centers and two distinct types of micropores, one of which is lined by CrO 4 2− (CROFOUR) anions and the other is decorated by the functionalized organic linker. These nets offer unprecedented selectivity towards Xe. Modelling indicates that the selectivity of these nets is tailored by synergy between the pore size and the strong electrostatics afforded by the CrO 4 2− anions.

Mohamed, Mona H.↗
DOE OSTI2025

Repurposing Post-Consumer Polyethylene to Access Cross-Linked Polyethylene with Reprocessabilty, Recyclability, and Tunable Properties

Polyethylene (PE) is the most widely produced plastic but accumulation and resistance to degradation has significantly contributed to the plastic waste crisis. Upcycling has presented promising solutions to transform PE waste into value-added products. In this study, mixed post-consumer PE was successfully repurposed into reprocessable and chemically recyclable cross-linked polyethylene (XLPE). This process involved converting PE into telechelic oligomers, followed by repolymerization using a hybrid cross-linking system consisting of a dynamic cross-linker 2,4,6-triethoxy-1,3,5-triazine (TETA) and non-dynamic cross-linker tris(6-isocyanatohexyl)isocyanurate (Tri-HDI). In the resulting XLPE, TETA facilitated iterative reprocessing with minimal property degradation across cycles, whereas Tri-HDI helped preserve functional performance throughout service life. Compared to PE, XLPE exhibited enhanced mechanical properties, reduced creep deformation under application-relevant temperatures, and high temperature structural stability. Notably, copolymerizing PE oligomers with commercial macrodiols was employed to create composite XLPEs, enabling tuning material properties. After use, XLPE can be efficiently and selectively depolymerized under mild conditions, even when mixed with commercial insulator cables. This depolymerization allows for the recovery of the constituent building blocks, enabling purification and subsequent repolymerization for reuse. This approach demonstrates the potential of repurposing plastic waste into sustainable materials and fostering the development of a circular economy.

INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CHEMI↗
DOE OSTI2016

Noria: A Highly Xe‐Selective Nanoporous Organic Solid

Abstract Separation of xenon and krypton is of industrial and environmental concern; the existing technologies use cryogenic distillation. Thus, a cost‐effective, alternative technology for the separation of Xe and Kr and their capture from air is of significant importance. Herein, we report the selective Xe uptake in a crystalline porous organic oligomeric molecule, noria, and its structural analogue, PgC‐noria, under ambient conditions. The selectivity of noria towards Xe arises from its tailored pore size and small cavities, which allows a directed non‐bonding interaction of Xe atoms with a large number of carbon atoms of the noria molecular wheel in a confined space.

Patil, Rahul S.↗
DOE OSTI2018

The presence of fecal test soil protects bacteria during cleaning/disinfection

Abstract Reusable medical devices (RMDs) must be reprocessed between uses to render them safe for each use and each patient. Cleaning used devices removes organic and inorganic soil making them either safe for reuse or ready for disinfection/sterilization depending on the device. Although cleaning is an important step in a RMD's life cycle, it is not always a priority during device design. In addition, when performing cleaning validation, it is recommended that the manufacturer takes into consideration, what the most appropriate or worst case conditions are in terms of type of soil or the presence of bacteria. This study compared the ability of three different cleaning/disinfecting agents (water, alcohol, and bleach) to remove bacteria and fecal test soil from two different polymers: polypropylene and ultrahigh molecular weight polyethylene (UHMWPE) with two different roughness. There were some differences in the effects of the cleaning/disinfecting agents, the materials, and the roughness depending on the particular circumstances. However, the most consistent effect on the removal of bacteria was the presence of soil, which protected the bacteria from being removed. Conversely, the presence of bacteria played little role in the removal of soil. Although the interactions between material type and roughness, soil type, and bacteria are complicated, they should be taken into account during device design and reprocessing validation to create a device that is easy and safe to use. © 2018 Wiley Periodicals, Inc. J Biomed Mater Res Part B: Appl Biomater 107B: 1706–1710, 2019.

Nandy, Poulomi↗