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All sources NASA NTRS DOE OSTI

NASA NTRSDate not supplied

Updates in Developing a Prototype Science Pipeline and Full-Volume, Global Hyperspectral Synthetic Data Sets for NASA’s Earth System Observatory’s Upcoming Surface, Biology and Geology Mission

The Surface Biology and Geology (SBG) mission recently passed mission confirmation review and has entered phase A – design and development. SBG will acquire high resolution solar-reflected spectroscopy and thermal infrared observations at a data rate of ~2.5 TB/day and generate products at ~40 TB/day. Given that the per-day volume is greater than NASA’s total extant airborne hyperspectral data collection, collecting, processing, disseminating, and exploiting the SBG data present new challenges. To meet these challenges, we have developed a prototype science pipeline and a full-volume global hyperspectral synthetic data set to help prepare for SBG’s flight (see poster GC42D-0730). Our science pipeline is based on the science processing technology developed for NASA’s Kepler and TESS planet-hunting missions. The pipeline infrastructure, Ziggy, provides a scalable architecture for robust, repeatable, and replicable science and application products that can be run on a range of systems from a laptop to the cloud or a supercomputer. Ziggy is compliant with NASA Procedural Requirement (NPR) 7150.2C, is at a technical readiness level (TRL) of 7 and has been released to github.com/nasa/ziggy. We integrated Ziggy with EO-1/Hyperion workflows to build a prototype pipeline and ingested the 17-year mission archive that provides globally sampled visible through shortwave infrared spectra that are representative of SBG data types and volumes. We fully implemented the first stage and processed the entire 55 TB Hyperion data set from the raw data (Level 0) to top-of-the-atmosphere radiance (Level 1R). We are currently evaluating the ISOFIT atmospheric correction module to convert the L1R data to surface reflectance (Level 2) before reprocessing the full data set to L2. Crosschecks are being performed with RadCalNet as well as with coincident observations by AVIRIS. We are also investigating modern methods for georectifying the Hyperion scenes. Finally, we describe an analysis of the cost to conduct forward processing and reprocessing campaigns for SBG on HECC with dedicated compute and storage resources using the resurrected Hyperion pipeline as a proxy for full-volume SBG data. The analysis demonstrates that SBG L0 data can be processed to L2 on HECC with full reprocessing campaigns every two years for ~$2.6M over a 7-year lifespan. Moreover, 69% of the system capacity would be available for other activities, possibly enabling future open-source science activities, including algorithm development, L3+ processing, .etc.

ESD↗

NASA NTRSDate not supplied

Considerations For Waste-to-Base Future Research Paths

Developing innovative technologies to reprocess low value waste into high value base products will enable a new “Waste-to-Base” (WtB) capability leading to innovative approaches to space mission design. Systems engineers and mission architects will work together under a new paradigm in logistics planning. Equipment, components, packaging, storage, and structures will be understood not only by functional requirements, but also in terms of useful lifetimes and materials of construction. The advent of a new technical discipline - reprocess engineering - will grow from the intersection between chemical, mechanical, and systems engineering, mission design, and logistics planning. A crowdsourcing approach was employed to gather low Technology Readiness Level (TRL) concepts into workable solutions for issues in trash management, CO 2 processing, foam reprocessing, and fecal processing. The results of the crowdsourcing activities are summarized along with a discussion for future technology development work.

Logistics reduction↗

NASA NTRSDate not supplied

Considerations For Waste-to-Base Future Research Paths

Developing innovative technologies to reprocess low value waste into high value base products will enable a new “Waste-to-Base” (WtB) capability leading to innovative approaches to space mission design. Systems engineers and mission architects will work together under a new paradigm in logistics planning. Equipment, components, packaging, storage, and structures will be understood not only by functional requirements, but also in terms of useful lifetimes and materials of construction. The advent of a new technical discipline - reprocess engineering - will grow from the intersection between chemical, mechanical, and systems engineering, mission design, and logistics planning. A crowdsourcing approach was employed to gather low Technology Readiness Level (TRL) concepts into workable solutions for issues in trash management, CO2 processing, foam reprocessing, and fecal processing. The results of the crowdsourcing activities are summarized along with a discussion for future technology development work.

Logistics reduction↗

NASA NTRS2023

TESS Data Release Notes: Sector 59, DR84

TESS Sector 59 observations include physical orbits 125 and 126 of the spacecraft around the Earth. Data collection was paused for a total of 0.63 days to download data. Data were downloaded approximately every seven days, once near apogee of each orbit and once near perigee. Note that the Sector 59 data were briefly made available on 2023 January 18–19 before an issue was identified that required reprocessing. Data have since been reprocessed. The reprocessing has a minimal impact on the final light curves, but resulted in a different set of TCEs than initially released. Users should re-download the data if retrieved before January 19th. In total, there are 25.79 days of science data collected in Sector 59.

TESS↗

DOE OSTI2020

Sensitivity and Uncertainty Quantification of Transition Scenario Simulations

This report documents the first collective attempt at developing and applying capabilities to quantify uncertainties, assess parametric sensitivities, and optimize multiple parameters and metrics in fuel cycle simulations generated by the SA&I Campaign. To do this, external codes that were designed to perform sensitivity analysis and uncertainty quantification (SA&UQ) needed to be coupled to the SA&I Campaign’s nuclear fuel cycle simulators (NFCS). In FY20, two approaches were pursued: 1) coupling Cyclus to an ORNL-internal code called MOT (Metaheuristic Optimization Tool) and 2) coupling DYMOND to the opensource SA&UQ tool kit Dakota. The primary objective of having these NFCS/SA&UQ coupled capabilities is to better inform DOE-NE and other stakeholders on the results generated from the NFCS. For a given set of fuel cycle strategies, policies, and technology assumptions that make up a fuel cycle scenario, these NFCS have traditionally been used by the SA&I Campaign to provide quantitative answers in terms of year-by-year mass flows, infrastructure requirements, costs, etc. With these newly developed coupled capabilities, the SA&I Campaign can now efficiently simulate hundreds or thousands of these scenarios, sample large ranges of parameters and assumptions, and use the unique features of the SA&UQ tools to process the data. This enables providing answers with known and propagated uncertainties, determining the sensitivity of important metrics to different parameters and assumptions, quantifying how much fuel cycle and technology parameters impact each other, and producing optimized fuel cycle strategies for single and multiple variables. To demonstrate these new capabilities, the Cyclus/MOT was used to model several scenarios ranging from simple fleet retirements to transitions to advanced reactors. Specifically, for a transition scenario from LWRs to SFRs and advanced LWRs, uncertainty quantification, sensitivity analysis, and optimization studies were applied to cases involving single and multiple parameter (input) and single and multiple metric (output) variations. In addition, a similar transition scenario was modeled to demonstrate how to optimize the reprocessing capacity parameter to minimize two performance metrics while taking into account uncertainties from two other parameters. Lastly, a depletion module based on SCALE/ORIGEN was added in Cyclus to simulate the third scenario that was designed to quantify the impact of the modeling assumption that all LWR used nuclear fuel have the same burnup. The newly developed DYMOND/Dakota capability was also applied to a transition scenario from the existing fleet to small modular reactors and fast reactors. This particular scenario involves not only explicit isotopic depletion via ORIGEN-2, but also includes multirecycling and utilizing the criticality search feature to determine the fresh fuel composition of recycled fuel, a feature unique to the DYMOND NFCS. A large database of simulations were run with 4 main parameters that were sampled: start date of reprocessing, reprocessing capacity, energy demand growth rate, and advanced reactor share of the fleet. The 4 main metrics were uranium consumption, enrichment requirements, waste generation, and levelized cost of electricity using data from the Cost Basis Report. The demonstrated SA&UQ results include those that inform on how to choose parameters to avoid “failed” scenarios, Sobol’ indices that inform on the importance of various parameters individually and synergistically, and Analysis of Variance (ANOVA) studies that decompose parameter ranges into groups and informs on whether variations are statistically significant.

Feng, B.↗

DOE OSTI2021

Gamma Radiolysis of Biphasic Acetohydroxamic Acid (AHA) Solvent Systems

Acetohydroxamic acid (AHA) has been proposed as a substitute for uranium(IV) and hydrazine as a plutonium complexant and neptunium reductant for simplified, single-cycle used nuclear fuel reprocessing flowsheets. However, the chemical behavior of AHA in an intense multi-component radiation field is poorly understood, especially under representative biphasic reprocessing solvent system conditions. In response to this critical knowledge gap, this study has investigated the gamma radiolytic integrity of AHA in aqueous nitric solutions in contact with an organic phase, comprising current and future reprocessing ligands (TBP, DEHBA, and DEHiBA) dissolved in n-dodecane diluent. Our data show negligible effect of the organic phase on the radiolytic behavior of AHA compared to complementary single-phase experiments.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2022

Recent Advances in Radiation-Induced Actinide Redox Chemistry

The actinide series boasts many unique physical and chemical features worthy of both fundamental and applied study. However, the chemical influence of their inherent radiation field is often overlooked, especially as we begin to explore the late actinides in more detail than ever possible before. From the perspective of used nuclear fuel reprocessing, the absorption of ionizing radiation induces the formation of a variety of transient and steady-state excited states, radicals, ions, and molecular degradation products, many of which are highly redox active and can lead to significant changes in a reprocessing solvent system’s physical and chemical properties. For example, radiolysis of the actinides can drive steady-state redox distributions and the formation of non-traditional oxidation states which can complicate their separation and recovery from fission products. This scenario is further exacerbated when complexation is taken into account. Consequently, a molecular-level understanding of radiation effects on the actinides over multiple time, distance, and material domains is essential for supporting innovation in used nuclear fuel reprocessing technologies. Attaining this knowledge necessitates a firm grasp of actinide radiation chemistry to develop predictive, mechanistic, multiscale models to support engineering efforts. Presented here are several recent studies that highlight recent advances in actinide radiation chemistry, in particular, the effect of actinide complexation on ligand reactivity towards radiation-induced transients.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2018

Irradiated FA Structural Materials Conditioning by Induction-Slag Remelting in a Cold Crucible

A method of managing one type of radwaste forming during BREST-300 reactor operation – the structural materials of FA – is described. In the module developed for reprocessing, the fissile materials losses together with all forms of waste must not exceed 0.1% of the initial amount, in connection with which the maximum admissible content of facade materials and structural materials sent into long-term storage must not exceed 0.001% by weight. Only the plutonium content may reach 0.015% by weight, which makes it necessary to perform additional purification and return fissile materials into the nuclear fuel cycle. A method of reprocessing the structural materials of irradiated fuel elements is proposed on the basis of induction-slag remelting in a cold crucible, which makes it possible to conduct purification to the desired index value. The operating regime of the melting unit and the qualitative composition of the employed flux that make it possible to reach the required level of purification by removal of fissile materials are described.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2019

Whole core analysis of the single-fluid double-zone thorium molten salt reactor (SD-TMSR)

The SD-TMSR (2,250 MW th ) is a Single-fluid Double-zone Thorium-based Molten Salt Reactor. The active core of the SD-TMSR is divided into the inner zone (486 fuel tubes) and the outer zone (522 fuel tubes) to improve the Th-U3 breeding performance. This work adopted the SERPENT-2 Monte Carlo code to analyze the whole core model of the SD-TMSR. Built-in SERPENT-2 capabilities simulated online reprocessing and refueling and calculated the multiplication factor and Breeding Ratio (BR). We found that the molten salt Temperature Coefficient of Reactivity ( α T ) was negative for initial and equilibrium states. This study investigated the variation of the multiplication factor, BR, and build-up of important nuclides in the core as a function of burnup. Under online reprocessing and refueling, we studied the variation of the reactivity during 60 years of reactor operation. Additionally, the neutron spectrum shift during the reactor operation was calculated. Finally, these simulations determined the appropriate 232 Th and 233 U feed rates for maintaining criticality and enabled analysis of the overall SD-TMSR fuel cycle performance.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2023

Chemical Recycling of Polybutadiene Rubber with Tailored Depolymerization Enabled by Microencapsulated Metathesis Catalysts

The effective management of plastic waste streams to prevent plastic land and water pollution is a growing problem that is also one of the most important challenges in polymer science today. Polymer materials that are stable over their lifetime and can also be cheaply recycled or repurposed as desired could more easily be diverted from waste streams. However, this is difficult for most commodity plastics. It is especially difficult to conceive this with intractable, cross-linked polymers such as rubbers. In this work, we explore the utility of microencapsulated Grubbs’ catalysts for the in-situ depolymerization and reprocessing of polybutadiene (PB) rubber. Second-generation Hoveyda-Grubbs catalyst (HG2) contained within glassy thermoplastic microspheres can be dispersed in PB rubber below the microsphere’s glass transition temperature (T g ) without adverse depolymerization, evidenced by rubber with and without these microspheres obtaining similar shear storage moduli of ≈16 and ≈28 kPa, respectively. The thermoplastic’s T g can be used to tune the depolymerization temperature, via release of HG2 into the rubber matrix. For example, using poly(lactic acid) (PLA) vs polysulfone results in an 85 and 162 °C depolymerization temperature, respectively. Liquefaction of rubber to a mixture of small molecules and oligomers is demonstrated using a 0.01 mol % catalyst loading using PLA as the encapsulant. Furthermore, at that same catalyst loading, depolymerization occurs to a greater extent in comparison to two ex-situ approaches, including a conventional solvent-assisted method, where it occurs at roughly twice the extent at each given catalyst loading. In addition, depolymerization of the microsphere-loaded rubbers was demonstrated for samples stored under nitrogen for 23 days. Lastly, we show that the depolymerized products can be reprocessed back into solid rubber with a shear storage modulus of ≈32 kPa. Thus, we envision that this approach could be used to recycle and reuse cross-linked rubbers at the end of their product lifetime.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2022

Review of recent developments in iodine wasteform production

Radioiodine capture and immobilization is not only important to consider during the operation of reactors (i.e., I-131), during nuclear accidents (i.e., I-131 and I-129) or nuclear fuel reprocessing (i.e., I-131 and I-129), but also during disposal of nuclear wastes (i.e., I-129). Most disposal plans for I-129-containing waste forms (including spent nuclear fuel) propose to store them in underground repositories. Here, iodine can be highly mobile and, given its radiotoxicity, needs to be carefully managed to minimize long-term environmental impacts arising from disposal. Typically, any process that has been used to capture iodine from reprocessing or in a reactor is not suitable for direct disposal, rather conversion into a wasteform for disposal is required. The objectives of these materials are to use either chemical immobilization or physical encapsulation to reduce the leaching of iodine by groundwaters. Some of the more recent ideas have been to design capture materials that better align with disposal concepts, making the industrial processing requirements easier. Research on iodine capture materials and wasteforms has been extensive. This review will act as both an update on the state of the research since the last time it was comprehensively summarized, and an evaluation of the industrial techniques required to create the proposed iodine wasteforms in terms of resulting material chemistry and applicability.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2019

Radionuclide Inventory Calculations for the Calcined Solid Storage Facility Performance Assessment - 19292

Originally named the Idaho Chemical Processing Plant when it was constructed in the early 1950's, the Idaho Nuclear Technology and Engineering Center reprocessed SNF to recover highly enriched uranium and other nuclear-related products for the U.S. DOE. Reprocessing SNF generated millions of gallons of aqueous radioactive waste. Calcination was a process developed at the INL Site to convert liquid waste into a solid, granular form from 1963 to 2000. The calcination process produced a safer product for storage while reducing the volume of stored liquid waste by an average factor of seven. Calcine is currently stored in the Calcined Solids Storage Facility (CSSF). The CSSFs are often called 'bin sets' because each CSSF (CSSFs 1 through 6) contains several stainless-steel storage bins housed in a concrete vault. The CSSF is an interim waste storage facility. In the future, calcine will be retrieved, treated (if necessary), and sent to an off-Site facility for storage or disposal. To close the CSSF, the bulk of the calcine will be removed and stored or disposed of elsewhere. The bins and vaults will then be placed in a configuration that is protective of human health and the environment. For purposes of the performance assessment, the depth of the residual calcine at closure was assumed to be 5-cm in each bin. The 5-cm of residual calcine in each bin is assumed to be mixed in grout during closure. The activities of the residual radionuclides are the inventory used for the dose assessment calculations for the CSSF performance assessment. The objective of this analysis was to calculate the radionuclide activity in the residual calcine remaining in the CSSF after closure. The purpose of this paper is to present the approach used to calculate the radionuclide inventory of the calcine currently stored in CSSF and the residual calcine in the CSSF after closure using data from the calcined waste Historical Processing Model (HPM). Limited sampling and analysis of the calcine in the CSSF bins was conducted in the past. Future similar calcine sampling and analysis efforts would be costly, generate additional mixed waste, and may not provide the desired detailed calcine composition data. However, during the later years of calcining, an alternative to calcine sampling and analysis was developed and the information was captured in the HPM. The HPM uses multiple sources to provide definitive calcine volume, mass, and composition, including: compositions and volumes of the liquid wastes that were calcined, ORIGEN2 - and ORIGEN-ARP-based radioactivity calculations, calciner and CSSF operating data, and process knowledge of INTEC fuel and waste-processing systems. HPM calcine composition was used to develop an inventory of radionuclides for use in the performance assessment evaluation for closure of CSSF. There are 43 bins that contain calcine, with six to twelve segments in each bin for a total of 337 bin segments for which the HPM calculates the average calcine composition. The HPM calculates the composition of each of the nearly 7,000 batches of material (startup bed, nonradioactive material, and radioactive waste) charged to the calciners. The calculations use the CSSF dimensions, including the CSSF number of bins, heights, outside diameters, annular air space diameter (if necessary), and volume of calcine currently in the bins. For each of the CSSFs, for a 5-cm residual, the percent of total calcine was calculated along with the residual activity for the radionuclides. Residual activity was estimated for 148 different radionuclides. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

Panel Session 32: Panel: Bringing First-of-a-Kind Treatment Facilities On-Line: Idaho IWTU Commissioning Approach (2/5)

This panel discussed the activities of bringing a first-of-a-kind treatment facility online: The Integrated Waste Treatment Unit (IWTU). The design was designed to treat 900,000 gallons of sodium-bearing liquid waste. Construction of this first-of-a-kind treatment facility was completed in 2012. A new approach was adopted in 2016 after multiple challenges were encountered during facility commissioning. This session focused on the Integrated Waste Treatment Unit (IWTU), which was constructed to convert the remaining 900,000 gallons of liquid sodium-bearing waste at the Idaho National Laboratory Site, Idaho, USA, to a granular solid. The liquid waste was generated during historic spent nuclear fuel reprocessing campaigns at the Idaho Nuclear Technology and Engineering Center and was transferred to three underground storage tanks. The waste was not first-cycle waste but was rather the liquid rinse waste from flushing waste transfer lines between reprocessing runs. IWTU will employ a steam-reforming process to treat the liquid waste. The resulting granular solid will be transferred to stainless steel canisters and then to concrete vaults, which will be stored on site until an off-site national geologic repository is available to accept the waste. Presentation proposed: Bringing First-of-a-Kind Treatment Facilities On-Line: Idaho IWTU Commissioning Approach (Joel Case, US DOE (USA), Leo Thompson, Sterling Engineering and Consulting Group, LLC (USA), Joseph Giebel, Fluor Idaho, LLC (USA), Richard Boardman, Idaho National Laboratory (USA), Kevin O'Neill, US DOE (USA))

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

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 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↗