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At least 469 records · Page 26
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

IER 517: Molybdenum Critical Experiment Design [Slides]

In nuclear fuels, U-Mo metallic fuels are used in new space reactor designs and new research reactor fuels. MITR, MURR, and NBSR reactors plan on converting from different HEU fuels to a U-Mo HALEU fuel. In spend nuclear fuel, 95 Mo is one of the 15 main absorbing fission products in irradiated LWR fuel. This makes it important for criticality safety studies in transportation and reprocessing. In structural materials. Molybdenum is found commonly in alloys that make up the structural materials of nuclear reactors such as type 316 SS. Molybdenum helps to improve high temperature performance and corrosion resistance.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Solvent Recovery and Management at the Savannah River Site H-Canyon Facility

NIOWAVE, Inc. is a domestic supplier of medical and industrial isotopes from uranium and radium. The Savannah River National laboratory (SRNL) is currently providing support to NIOWAVE, which plans to deploy a superconducting electron linear accelerator (LINAC) to fission uranium for Mo-99 production without the need for a nuclear reactor or HEU. The uranium from the Mo-99 production targets will be purified using a modified PUREX (Plutonium Uranium Reduction Extraction) solvent extraction process to recover the uranium in the product stream. The uranium will then be precipitated as an oxalate which is calcined to U 3 O 8 to fabricate pellets for new Mo-99 targets. In previous support provided to NIOWAVE, the SRNL demonstrated a solvent washing process to remove degradation products from the tributyl phosphate (TBP) solvent used in the modified PUREX process under development for uranium recovery. To supplement this technology demonstration, NIOWAVE requested the SRNL to provide summary information on the solvent recovery and management activities which are used at the Savannah River Site (SRS) H-Canyon facility. An existing reference document for the reprocessing of irradiated HEU fuels at the SRS was used as the primary reference for the solvent management activities; although, other reference documents were used to provide supplementary information. The information provided includes a brief summary of the solvent degradation issues which have been observed in the H-Canyon solvent extraction cycles and resulting process safety concerns. The solvent recovery processes for the three cycles of solvent extraction used in the H-Canyon were subsequently described including the process equipment which consists of the continuous and batch solvent washers, pumps, and tanks. A final section is provided on the monitoring and analysis of solvent quality based on the previous work performed at the SRNL for NIOWAVE and past research and development activities performed to support the solvent extraction processes in both the SRS F-Canyon and H-Canyon facilities.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2022

Highly Recyclable Thermosets for Lightweight Composites: CRADA 517 [Abstract only]

Lightweight composites are of increasing need to enable reduced fuel usage, reduced carbon emissions and increased range in air and ground transportation. The superior weight savings of carbon fiber thermoset composites are impeded by material and production costs and the limited options available for composite repair and recycle. The materials recovery and recycling sector is currently overwhelmed by an abundance of low-value materials and limited market opportunities for downcycled content, rendering landfilling a more attractive alternative. Vitrimers represent a new paradigm for repairable and recyclable thermosets, but this relatively new technology will require development, including government investment, to reduce risks for its adoption by aerospace and automotive OEMs. The overall goal of the project is to create recyclable carbon fiber composites using vitrimer resins that are more energy efficient to produce and have improved properties over baseline technology. The targeted composites will retain their tensile strength after multiple recycling and reprocessing steps. In addition, monomer will be recoverable through depolymerization and re-usable carbon fibers will be retrieved.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Discrete-event Simulation Process Model for the Pyrochemical Processing of Plutonium at Los Alamos National Laboratory

The pyrochemical metal production operations that occur in the Plutonium Facility at Los Alamos National Laboratory perform plutonium purification with the aim to provide plutonium metal for a variety of defense- and non-defense missions within the National Nuclear Security Administration. The demands and constraints associated with the pyrochemical processing of plutonium are complex, making decision analyses challenging for program managers who require plutonium production for their mission applications. The construction of a discrete-event simulation process model is proposed to measure and report the process capacity, material throughput, equipment requirements, and dose accumulation for operators of the pyrochemical metal production operations. The process model, constructed in the ExtendSim™ software, will represent the cause-and-effect relationships between the pyrochemical processing environment and the process constraints, including criticality limitations, material control and accountability measures, chemical analysis requirements, and equipment availability. An accurate representation of the pyrochemical metal production process capacity through simulation modeling will be helpful to program managers in their efforts to forecast plutonium availability for mission applications. Furthermore, the proposed process model will be vital for future analyses that will measure the interactions between the pyrochemical metal production operations and the aqueous reprocessing operations and their ability to minimize transuranic waste disposal.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Fabrication and Irradiation Behavior of Thulium Sesquioxide

The results of reactor irradiation of thulium sesquioxide are discussed, along with chemical and physical properties and wafer fabrication. Sintered densities up to 90% of theoretical were obtained in target wafers using oxide reprocessed to obtain readily sinterable powder. Irradiation to about 1.8 w/g did not change the microstructure of the wafers although some wafers broke, probably because of mechanical stresses in target and disassembly.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Capture of Organic Iodides from Vessel Off-Gas Streams

In nuclear waste reprocessing, radioactive iodine is released in both organic and inorganic forms into off-gas streams. Due to its properties of long half-life and potential accumulation in human bodies, radioactive iodine is required to be removed from off-gas streams by the Environmental Protection Agency (EPA) and Nuclear Regulatory Commission (NRC). In this study, the removal of organic iodides using two prototype silver (Ag) adsorbents, i.e., reduced Ag functionalized silica aerogel (Ag 0 -aerogel) and reduced Ag exchanged mordenite (Ag 0 Z), is investigated through a combination of experimental and modeling studies.

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

Requirements Description of the REBUS Software for the Versatile Test Reactor

This report presents the modeling and simulation capabilities of Argonne National Laboratory’s fuel cycle analysis code, the REactor BUrnup System (REBUS), that will be relied on for the VTR project. These capabilities will then be used to establish the set of REBUS verification tasks necessary to verify REBUS for usage on the VTR project. A similar path was followed for the DIF3D software where its’ requirements and verification tasks were established. REBUS has been maintained by Argonne since the early 1960s to support its reactor design mission. That software transitioned from the original REBUS to REBUS-2 in the mid 1970s and to REBUS-3 in the mid 1980s. Since then REBUS has gone through many revisions to the current REBUS-11. Note that this version numbering is consistent with the progression of DIF3D, the base flux solver that REBUS is built upon. The name REBUS refers to a pictorial based puzzle as the original developers were inspired by having to track thousands of unique fuel assemblies as they are inserted into the reactor, depleted, shuffled, discharged, and reprocessed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

A Review of Advanced Test Reactor Fuel and Assessment of Its Compatibility with the ZIRCEX Chlorination Process

Advanced Test Reactor (ATR) fuel has been identified as a resource for high-assay low-enriched uranium (HALEU) production. A survey was performed on the published literature describing ATR fuel. The geometry of the fuel is complex; different parts of the fuel compact experience differing neutron flux and burnup. The literature is sparse, and access is controlled. Therefore, fundamental studies of fuel reprocessing must use a model fuel that represents the main chemical and structural features. Advanced chlorination, or chlorination with sulfur-chlorine bearing reagents is being investigated as way to separate the fuel from metal matrix alloys. A UAl x alloy will be fabricated with x = 3, 4, and 5. The potential chlorination of individual UAl x intermetallics will be assessed in the advanced chlorination process of Al-8001 and Al-6061 as well as a representative mixture. Initial studies will track the alloying elements of the Al, which are Si, Fe, Cu, Mn, Mg, Cr, Zn, and Ti, in addition to the U itself. Further studies will include fission product simulants. Because advanced chlorination solvents include sulfur, the chemistry of sulfur with major and minor constituents will also be investigated. The experimental work accompanied by neutronic calculations will allow the assessment of the feasibility of advanced chlorination to separate aluminum from uranium. If bench-scale testing appears promising, then small-scale tests in shielded facilities with irradiated cladding, lightly irradiated fuel, and spent nuclear fuel are recommended to track the complete inventory of fissile actinides, fission product impurities, and reagent solids and liquids.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Malleable Thermosets (Vitrimers) from CO 2 and Plants

Fiber-reinforced composite materials are increasingly used to replace heavy metal components in transportation applications for lightweighting purposes. Polymer matrix materials used in automotive parts, such as epoxies, nylons, and polypropylenes, are almost all produced from fossil fuels. Higher performance thermoset composites are generally not repairable and lower-cost thermoplastic composites do not have enough performance to replace steel. Vitrimers are a new class of polymer materials with long shelf life, low waste in production, ease of processing, and repairability that have the potential to break through the high-cost thermoset/low-performance thermoplastic barrier to vehicle light-weighting. Vitrimers also exhibit easier chemical recyclability than conventional thermosets with reversible chemical crosslinking that enables reprocessing of scrap from production and end of life components. Vitrimers are a new class of engineered plastics that are weldable, repairable and recyclable like thermoplastics but have high mechanical properties like thermosets. Bio-based vitrimers have only recently been reported in the academic and patent literature. These materials represent an unusual opportunity to meet both VTO light-weighting targets for low-density, high performance materials and BETO sustainability targets for value-added high-volume applications of bio-fuel byproducts. Also, PNNL researchers have previously demonstrated a process to convert CO 2 into potential vitrimer precursors. Washington State University (WSU) success in demonstration of bio-based vitrimers and recent steps toward commercialization of petroleum-based vitrimers support the opportunity for PNNL success in development of sustainable vitrimer materials for recyclable, high-performance fiber-reinforced composites for the transportation sectors.

36 MATERIALS SCIENCE↗
DOE OSTI2023

On-line Measurement of Hydrogen Gas using Raman Spectroscopy for Process Gas Systems

Advanced reprocessing schemes such as the Zircex process are important for continued development of the nuclear fuel cycle. The hydrochlorination reaction of the Zircex process flow sheet for metallic fuels has a known off-gas stream of H 2 (g) and HCl. The flow sheet can be simplified through HCl recycling, which requires detection and quantification of any residual H 2 (g) for safety and purity. In this work, commercially available Raman spectroscopy was applied for low-level, on-line detection of H 2 (g) in process gas streams, and parameters were adjusted to optimize the H 2 (g) Raman signal. Increasing the number of scans and exposure time in the spectrometer increased the signal-to-noise ratio of the H 2 (g) Raman signal, while the gas flow rate was optimized at 2.3 L/min. The limit of detection for H 2 (g) was estimated to be 2,100 ppm H 2 (g) in an N 2 (g) background. This scoping study for the application of Raman spectroscopy for on-line measurements of H 2 (g) shows that H 2 (g) can be detected at low levels (approximately 5% of the lower explosive limit (LEL)) in flowing gas streams.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Technology Development Roadmap for Volatile Radionuclide Capture and Immobilization

An “Off-Gas and Waste Forms Strategy Workshop” was held in Washington, D.C. on January 31, and February 1, 2023, to review the current state of the art (baseline) technologies for capturing and immobilizing these volatile radionuclides in waste forms (WFs). The discussions in that workshop were used to develop a roadmap for future research and development to mature technologies that are presently not ready for use in a commercial reprocessing facility and develop and demonstrate technologies that provide more safety, more simplicity, or lower costs compared to the current baseline.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Entrainment and aerosol effects on marine boundary-layer clouds: An investigation using ACE-ENA data from HOLODEC, G1, Pico and ACTOS (Final Report on Project Activities)

Marine boundary-layer clouds cover large regions of the globe and are known to strongly influence radiative balances. The microphysical properties and persistence of these clouds are tightly coupled with cloud-top entrainment and aerosol properties within both the boundary layer and the overlying free troposphere. This work addressed the microphysical response to entrainment and aerosol properties in marine stratocumulus clouds, using data from the Aerosol and Cloud Experiments in the Eastern North Atlantic (ACE-ENA) project. Specifically, emphasis was placed on 1. extensive in situ measurements taken with the Atmospheric Radiation Measurement (ARM) G1 aircraft using the Holographic Detector for Clouds (Holodec) instrument; 2. airborne measurements obtained with the helicopter-borne Airborne Cloud Turbulence Observation System (ACTOS) operated by the Leibniz Institute for Tropospheric Research (TROPOS) during the first phase of ACE-ENA; and 3. above-boundary-layer, mountain-top measurements taken at the Pico Mountain Observatory (OMP) by the research group of the co-investigator and scientists from TROPOS during the first phase of ACE-ENA. The work is aligned with the topic “Warm Boundary-Layer Atmospheric Processes.” Key results include the following. Data from the Holographic Detector for Clouds (Holodec) were reprocessed, refined, and validated through a careful instrument intercomparison, resulting in a high-quality dataset available on the ARM archive for ACE-ENA. High-resolution ACTOS measurements of thermodynamics, microphysical, and turbulence properties were analyzed to explore the behavior of the entrainment velocity at cloud top. Analysis of the vertical variability of cloud droplet size distribution shape suggests that mixing is more inhomogeneous near cloud top, and more homogeneous deeper into the cloud. Analysis of aerosol measurements from OMP explored the mixing state and cloud condensation nucleus properties, as well as the implications for radiative forcing. Finally, a machine learning algorithm was developed for identifying characteristic cloud droplet size distributions, and was employed to explore microphysical regimes in stratocumulus clouds observed during ACE-ENA. The characteristic size distributions are narrow, and only when spatially averaged do they produce the broad “gamma” distributions typically assumed in models. This implies that precipitation development should account for variability and correlations in the distribution shape as well as the number concentration of cloud droplets.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2023

Technology Development Roadmap for Volatile Radionuclide Capture and Immobilization

An "Off-Gas and Waste Forms Strategy Workshop" was held in Washington, D.C. on January 31, and February 1, 2023, to review the current state of the art (baseline) technologies for capturing and immobilizing these volatile radionuclides in waste forms (WFs). The discussions in that workshop were used to develop a roadmap for future research and development to mature technologies that are presently not ready for use in a commercial reprocessing facility and develop and demonstrate technologies that provide more safety, more simplicity, or lower costs compared to the current baseline.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Scaled up Process Report – Apparatus and Model

Advanced voloxidation with NO 2 is a proposed process for used nuclear fuel head-end reprocessing scheme that converts UO 2 to higher oxides, and it also converts partitioning volatile fission products into the gas phase, thus facilitating fuel dissolution and actinide recovery. NO 2 voloxidation is being studied on small batches of UO 2 Simfuel, but the real test of process feasibility will be when it is scaled up to work with >100 g of irradiated material. This report discusses the aspects of scale-up that must be considered for NO 2 voloxidation, including development of a stirred reactor to promote agitation of the mixture during processing, online process monitoring, and automation controls. Brief details on parallel efforts are also provided in this report, including (a) demonstration of iodine release from Simfuel made by Spark Plasma Sintering, and (b) development of an order-of-magnitude scale-up to react 100 g of UO 2 Simfuel in a metal reactor.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Sludge Processing Options for early HLW Treatment at Hanford

The U.S. Department of Energy’s (DOE) Hanford Site has 177 underground storage tanks that contain wastes from past nuclear fuel reprocessing and waste-management operations. Over 20% of this waste is in the form of an insoluble sludge that will require solids concentration and washing prior to vitrification for long-term disposal. An assessment of potential flowsheet operations to support feed preparation activities prior to high level waste (HLW) vitrification has been conducted to better evaluate pretreatment processing options. Settling studies assessing the baseline approach of a settle-decant method were explored as well as a crossflow filtration system to be used alternatively for concentrating and washing HLW sludge. Significant variations in behavior of settling rates and sludge characteristics give reason to evaluate alternative pretreatment options for the HLW. Non-radioactive sludge containing iron oxide, boehmite, and gibbsite were evaluated via gravity settling and crossflow filtration to determine the behavior of these compounds in various tank waste matrices. Understanding the predictive capabilities of HLW solids settling as well as sludge concentration via crossflow filtration can help provide technical guidance during flowsheet planning.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Alternative Materials Literature Review: A Review of Alternative Materials for Xenon Capture in Comparison to AgZ-PAN

A collaborative literature review between Idaho National Laboratory (INL) and Pacific National Laboratory (PNNL) was conducted to research alternative metals/materials for xenon adsorption in gaseous effluents generated from used nuclear fuel (UNF) reprocessing. The review was conducted to research sorbents that could outperform silver mordenite (AgZ) at ambient temperature. The use of silver requires complex and expensive waste disposal processes due to it being an RCRA metal, therefore alternative materials need to be researched and evaluated. The report details two different types of sorbent material: alternative zeolite materials and non-zeolite-based materials (i.e., MOFs, COFs, etc.). This report documents a brief history of the materials, what testing has been conducted on them, and what results suggest the possibility of outperforming the silver-based materials. Sorbent performance characteristics such as durability, capacity, and operating temperatures were evaluated. The simplicity of the sorbents was also evaluated, meaning how they are acquired (synthesized vs. naturally abundant). Recommendations were made for the most promising sorbents based on available research in the subject area and these recommended sorbents will be tested in the upcoming fiscal year. The recommended sorbents that are zeolite based are CHA zeolites, ZIF-69, and Na-ETS-10. The recommended sorbents that are non-zeolite based are SBMOF-1 and CC3.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Complete Initial Evaluation of Novel Complexants for Tc Holdback for Simplified Single Cycle Separations.

Management of technetium at the back end of nuclear fuel cycle is complicated by its unique physico-chemical character, but must be addressed to due to the environmental, storage and reprocessing challenges of this fission product. In a single-cycle scenario, the recovery and purification of uranium is particularly challenging due to the tendency of a pertechnetate anion to follow uranyl cation. This report summarizes initial studies directed towards finding the solution for technetium management. Two options for controlling technetium’s chemistry in solvent extraction scenarios were studied. Molecular recognition of pertechnetate by aqueous complexants based on guanidinium moieties was compared to redox manipulation of technetium using dihydrazide reagents. Solvent extraction, potentiometric and nuclear magnetic resonance spectroscopy studies were performed. The influence of guanidinium reagents on the liquid-liquid partitioning of technetium is less pronounced, relative to dihydrazides. Strong impact exerted by dihydrazides originates from the technetium-catalyzed decomposition of such compounds which reduces technetium to a tetravalent, non-extractable state. Although this route of technetium management is very effective the destruction of dihydrazides is undesirable. Guanidinium complexants show a more tempered influence on technetium. The effect on the liquid-liquid partitioning of pertechnetate is evident and likely guided by anion recognition due to ion-pair and hydrogen bond formation. Guanidiniums are stable in presence of technetium as evidenced by nuclear magnetic resonance studies. The steady state chemistry of guanidiniums identify this class of aqueous complexants as solid candidates for structure-function pertechnetate recognition studies.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Analysis of H-Canyon Process Tanks in Preparation of Consolidation and Blending for HALEU Fuels

High-Assay Low- Enriched Uranium (HALEU) fuels are being developed to support the replacement of Highly Enriched Uranium (HEU) fuels used in U.S. High-Performance Research Reactors (USHPRR) as well as advanced nuclear power reactor designs. The projected demand for HALEU far exceeds the supply and studies are underway to assess various options to partially mitigate the potential short supply. The H Canyon facility at the Savannah River Site (SRS) Low Enriched Uranium (LEU) containing 4.95% U-235 from the reprocessing of highly enriched foreign and domestic research reactor fuel for the Tennessee Valley Authority’s (TVA) commercial power reactor market for several decades. The production of LEU at the H-Canyon facility can be readily transitioned to produce 19.75% HALEU solutions from the current separated inventory of purified HEU solutions in H-Canyon storage.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗