Search NASA⌕ Search

Engineering topics

McNamara, Bruce K.

Publications and source records attributed to McNamara, Bruce K..

At least 19 records

Improved production and purification of 240Am by deuteron-induced activation of 240Pu target

An optimized method for production of 240Am via the 240Pu(d,2n)240Am reaction is reported. The optimized method produced 7.94 × 108 ± 7% atoms 240Am/mg 240Pu/µA-h. The yield and purity of the products from the optimized method are evaluated in context of the intended application to produce material to measure the cross-section for the 240Am(n,f) reaction. The presented method is also compared with other production methods available in the literature. Keywords—240Am production, deuteron-induced reactions, cross-section, target preparation, post-irradiation purification Abbreviations GEA—Gamma Energy Analysis

Morrison, Erin C.↗

Gaining insight into lithium-ion battery degradation by a calorimetric approach

With the growing demand for lithium-ion batteries (LIBs) in transportation and renewable energy sectors, ensuring reliability and mitigating degradation are crucial challenges for their widespread deployment. Here this study investigates the relationship between thermal characteristics and degradation/reliability by developing an in-situ heat measurement methodology for LIBs operating under grid services. Using adiabatic mode-based calorimetry, we examine the thermal behavior of two commercial cells with different cathode chemistries (Ni-rich layered oxide and olivine) before and after two years of grid service (peak shaving) and demonstrate a strong correlation between thermal characteristics and electrochemical performance and its degradation. By further analyzing the heat sources separation (differentiating between irreversible and reversible heat), we disclose the underlying degradation mechanisms of LIBs. These findings emphasize the critical role of thermal characteristics in determining LIB reliability and deterioration, while the obtained heat data can aid in the development and calibration of LIB state-of-health models for grid applications.

25 ENERGY STORAGE↗

Chlorine Isotope Separations using Thermal Diffusion

As of the close of 2023 a thermal diffusion isotope separations (TDIS) apparatus was constructed, successful shakedown testing was achieved and enrichments of isotopic concentrations relative to natural abundance 35/37 Cl were collected. Further, the model for these enrichment experiments drove the timing for sampling and other critical, extrapolated functions that are discussed herein. Our decision to move forward in 2024 with the installation of a larger set of separation tubes and associated equipment was based on the successful development of the predictive model. The Chlorine Isotopes Project Team at PNNL is prepared to claim that the installation and hence its separative power is restricted only by the spatial limitations of the laboratory, and this at present appears to be the limiting feature to first pass high enriched 37 Cl.

07 ISOTOPE AND RADIATION SOURCES↗

Expanding the Transuranic Metal–Organic Framework Portfolio: The Optical Properties of Americium(III) MOF-76

Reported here is the synthesis, crystal structure, and solid-state characterization of a new americium containing metal–organic framework (MOF), [Am(C 9 H 3 O 6 )(H 2 O)], MOF-76(Am). This material is constructed from Am 3+ metal centers and 1,3,5-tricarboxylic acid (BTC) ligands, forming a porous three-dimensional framework that is isostructural with several known trivalent lanthanide (Ln) analogs (e.g., Ce, Nd, and Sm–Lu). The Am 3+ ions have seven coordinates and assume a distorted, capped trigonal prismatic geometry with C1 symmetry. The Am 3+ –O bonds were studied via infrared spectroscopy and compared to several MOF-76(Ln) analogs, where Ln = Nd 3+ , Eu 3+ , Tb 3+ , and Ho 3+ . The results show that the strength of the ligand carboxylate stretching and bending modes increase with Nd 3+ < Eu 3+ < Am 3+ < Tb 3+ < Ho 3+ , suggesting the metal–oxygen bonds are predominantly ionic. Optical absorbance spectroscopy measurements reveal strong f–f transitions; some exhibit pronounced crystal field splitting. The photoluminescence spectrum contains weak Am 3+ -based emission that is achieved through direct and indirect metal center excitation. The weak emissive behavior is somewhat surprising given that ligand-to-metal resonance energy transfer is efficient in the isoelectronic Eu 3+ (4f 6 ) and related Tb 3+ (4f 8 ) analogs. The optical properties were explored further within a series of heterometallic MOF-76(Tb 1–x Am x ) (x = 0.8, 0.2, and 0.1) samples, and the results reveal enhanced Am 3+ photoluminescence.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Irradiation Effect on Noble Metal Particles in Water Using in situ Liquid Cell STEM Observation

During geologic disposal of spent nuclear fuel (SNF) in an engineered nuclear waste repository, once all other barriers have degraded, oxidizing may occur at the solid-water interface owing to a self-generated radiolytic field. The repository design includes large quantities of iron (Fe), that is anticipated to corrode under an anoxic environment, and generate hydrogen (H 2 ) gas. This H 2 gas is thought to be able to suppress the dissolution of SNF through a catalytic reaction with noble metal particles (NMP) that are pre-existing in the SNF. This interaction leads to the decomposition of the major oxidant, hydrogen peroxide (H 2 O 2 ). In conclusion, these processes are described in the Fuel Matrix Degradation (FMD) model that is being used to predict SNF degradation rates. The NMP, therefore, plays an important role within the FMD model.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

A laboratory-scale process for producing dilithium beryllium tetrafluoride (FLiBe) with dissolved uranium tetrafluoride

Flibe Energy, Incorporated (FEI)'s conceptual Lithium Fluoride Thorium Reactor (LFTR) incorporates a chemical processing facility aimed at recovering uranium and other valuable volatile radionuclides while managing harmful radionuclides from the used fuel. The fuel utilized in this reactor is a combination of dilithium beryllium tetrafluoride (Li 2 BeF 4 or FLiBe) and uranium tetrafluoride (UF 4 ), (FLiBe/U). FEI's plan involves extracting the uranium and other valuable volatile fluoride-forming radionuclides using nitrogen trifluoride (NF 3 ). To facilitate laboratory-scale testing of uranium extraction using NF 3 and address the toxicity and physical hazards associated with beryllium and beryllium fluoride (BeF 2 ), we used a two-step process to prepare the simulated fuel salt. The first step entailed thermally decomposing ammonium beryllium tetrafluoride [(NH 4 ) 2 BeF 4 ] (ABeF) through a nominal 3-step process, combined with appropriate amounts of lithium fluoride (LiF) and UF 4 , resulting in the formation of beryllium fluoride (BeF 2 ). In the second step, the mixture was repeatedly melted and frozen at the melting point of FLiBe to prepare the eutectic FLiBe with dissolved UF 4 . Although the concept appears straightforward, the production of FLiBe/U involved various challenges. These challenges included transporting the gaseous decomposition products of ABeF, hydrogen fluoride (HF) and ammonia (NH 3 ), while preventing the formation of ammonium fluoride (NH 4 F). Additionally, it was necessary to control the reaction between the higher-than-anticipated water content in the commercial ABeF with NH 3 , HF, and the condensed NH 4 F, protect UF 4 from forming an unknown black compound, select suitable structural materials to mitigate fluoride corrosion, address the risks associated with beryllium toxicity through equipment design and operational protocols, and monitor process conditions. This article provides an account of the thermal decomposition chemistry observed in the commercial ABeF, describes the FLiBe/U production apparatus, describes the experiences and process refinements developed to prepare FLiBe/U, and presents our characterizations of prepared FLiBe/U.

Ammonium beryllium fluoride thermal decomposition↗

Advancing radioactive material research method: the development of a novel in situ particle-attached microfluidic electrochemical cell

This study aims to develop a microgram-scale microfluidic electrochemical cell (E-cell) for investigating the redox behavior of uranium oxide (UO 2 ). The traditional bulk electrochemical methods may require shielded facilities to investigate the hazardous materials, e.g., spent nuclear fuel, due to high radiation levels. Microfluidic E-cells offer advantages such as reduced radiation exposure, control over fluid flow rates, and high-throughput capabilities. Methods: The design of the E-cell considers electrode morphology, adhesion to a thin membrane, electrode configuration, and vacuum compatibility. Three techniques, including FIB-SEM lift-out, Au coating, and polyvinylidene fluoride (PVDF) binder, are explored for fabricating and attaching microgram quantities of UO 2 as working electrodes. The PVDF binder method proves to be the most effective, enabling the creation of a vacuum-compatible microfluidic E-cell. Results and discussion: The PVDF binder method demonstrates successful electrochemical responses and allows for real-time monitoring of UO 2 electrode behavior at the microscale. It offers chemical imaging capabilities using in situ SEM/EDS analysis. The technique provides consistent redox outcomes similar to bulk electrochemical analysis. Conclusion: The development of a microgram-scale microfluidic electrochemical cell using the PVDF binder technique enables the investigation of UO 2 redox behavior. It offers a low-risk approach with reduced radiation exposure and high-throughput capabilities. The technique provides real-time monitoring and chemical imaging capabilities, making it valuable for studying spent nuclear fuel systems and material characterization.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Measuring the Electrochemical Corrosion of 233U-doped UO2 in a Microfluidic Device

We have developed a microfluidic, ‘particle-attached microfluidic electrochemical’ device, or PAMEC, that uses microgram quantities of radioactive materials, such as uranium dioxide (UO2), rather than bulk specimens to both observe electrochemical changes and measure spectroscopic properties. The PAMEC device reduces the need for running experiments in a shielded facility and allows for in-operando, real-time investigations of electrochemical properties and morphological changes of micrograms of particles. This novel design offers a promising platform to investigate spent nuclear fuel (SNF) and other nuclear materials under various controlled condition, and to provide insights to nuclear waste management, and fuel performance

Electrochemistry↗

In-Situ Liquid Cell Transmission Electron Microscopy of Nanoparticles from Spent Nuclear Fuel

Noble Metal Particles (NMP) that are present in SNF play an important role in the Fuel Matrix Degradation (FMD) Model that is being considered for predicting the release rate of radionuclides from a waste package. Previous characterization of chemically separated NMP from SNF showed that these were well-crystallized nanoparticles containing Mo, Tc, Ru, Rh, and Pd with some evidence of actinides. These same chemically separated particles were re-characterized following 10 years of storage as the separated material in water. Scanning Transmission Electron Microscopy (STEM) analysis revealed that significant alteration had occurred in this material during storage, resulting in the almost complete degradation of the NMP. We hypothesized that radiolytic damaged had occurred in the material. To support these concepts, we synthesized a mixed Ru-Pd nanoparticle and subjected it to irradiation in the electron microscope in an in-situ liquid cell. We demonstrated that intense radiation can lead to the decomposition of these metallic particles.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Plutonium Hybrid Materials: A Platform to Explore Assembly and Metal–Ligand Bonding

In this work, we report the synthesis of five new hybrid materials containing the [PuCl 6 ] 2- anion and charge balancing, non-covalent interaction donating 4-X-pyridinium (X = H, Cl, Br, I) cations. Single crystals of the title compounds were grown and harvested from acidic, chloride-rich, aqueous media and their structures were determined via X-ray diffraction. Compounds 1-4, (4XPyH) 2 [PuCl 6 ] and 5, (4IPyH) 4 [PuCl 6 ] · 2Cl, exhibit two distinct sheet-like structure types. Structurally relevant non-covalent interactions were tabulated from crystallographic data and verified computationally using electrostatic surface potential maps and the quantum theory of atoms in molecules (QTAIM) approach. The strength of the hydrogen and halogen bonds was quantified using Kohn-Sham density functional theory and a hierarchy of acceptor-donor pairings established. In turn, the PuIV-Cl bonds were studied using the QTAIM and natural localized molecular orbital (NLMO) approaches to delineate the underlying bond mechanism and hybrid atomic orbital contributions therein. Energy decomposition (ED) and natural ED analyses were also explored to probe the bond mechanism and, more broadly, explore the efficacy of these techniques in studying these anionic systems. The results of the PuIV-Cl bond analyses were compared across composition via analogous treatments of previously reported [PuO 2 Cl 4 ] 2- and [PuCl 3 (H 2 O) 5 ] molecular units. In summary, our study indicates that the Pu-Cl bonds are predominately ionic, yet exhibit small varying degrees of covalent character that increase from [PuCl 3 (H 2 O) 5 ], [PuO 2 Cl 4 ] 2- , to [PuCl 6 ] 2- , while the participation of the Pu based s/d and f orbitals concurrently decrease and increase, respectively.

transuranic↗

Corrosion of U233-Doped Uranium Oxide using Microfluidics Methods

The aim of work this year has been to investigate the role of alpha (α)-radiation and hydrogen (H 2 ) on the corrosion of uranium oxide (UO 2 ) using a microfluidic device. The microfluidic device, termed the Particle-Attached Microfluidic Electrochemical Cell, (PAMEC), enables monitoring of the UO 2 electrochemical corrosion potential (E corr ) under different environments, including de-aerated conditions and in the presence of dissolved H 2 . The Si 3 N 4 window allows us to study morphological and chemical changes under an electron microscope. We have previously demonstrated that the PAMEC matches the results from bulk electrochemical tests with UO 2 [1]. We used the high specific activity uranium (U) isotope, 233 U, (t 1/2 = 160,000 years) incorporated into UO 2 , to generate a localized a-field. The objective of the experiments were to mimic the radiation environment that would be experience at the surface of aged spent nuclear fuel (SNF) during long-term geologic disposal under anoxic conditions. Wittman et al. [2] predicted that in the presence of a pure a-radiation field and under H 2 conditions, the concentration of the radiolytic oxidant H 2 O 2 would be suppressed or even eliminated. In a UO 2 corrosion experiment this would be exhibited through a lowering of the measured UO 2 corrosion potential compared to identical conditions in the absence of dissolved H 2 . We found that the predictions of Wittman and co-workers were supported and that the corrosion potential of the 233 U-doped UO 2 in solution lowered with presence of H 2 gas and increased in the absence of H 2 , when under anoxic conditions. The PAMEC experiments indicate that H 2 O 2 has been eliminated in a solution sparged with H 2 while exposed to an a-radiation field. The corrosion potential of the a-doped 233 U(10%)- 238 UO 2 in a solution sparged with Ar/H 2 matched the corrosion potential of 238 UO 2 in solution sparged with air. This clearly demonstrated that the H 2 O 2 had been eliminated and that the only oxidant present was O 2 in this system in complete agreement with the modeling results of Wittman et al. These results point to the need to improve the Fuel Matrix Degradation (FMD) Process Model training data set that is being used in the FMD surrogate model that is being developed for the repository program. The incorporation of realistic radiation chemistry will improve the scientific basis for the FMD Model.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

14 MeV Irradiation and Analysis of a 93% 239 Pu Target in Preparation for a F2019 FY22 Pu Campaign

In this work, we present the irradiation of a 93% 239 Pu with 14 MeV neutrons and subsequent analysis of the fission and activation products. The fully assembled target, a Pu metal bead encapsulated in Al, further encapsulated in welded stainless steel, was analyzed 22 times over more than 100 days using gamma emission analysis (GEA). Using the results from these analyses, R-values and fission yields for fission products were determined. To prepare for the FY22 Pu irradiation using the GODIVA critical assembly at NCERC, the irradiated Pu target was disassembled, dissolved, and separated using chemistry provided by LANL collaborators. This chemical separation was intended to remove the Pu from solution with little to no effect on the remaining elements. The chemistry was assessed to try to determine possible routes of fractionation of the sample and the fission products. The separation process used was successful for the bulk of the analyzed fission and activation products, as well as the added radiotracers. The final dissolved solution and separated fractions were analyzed by GEA looking at the remaining fission, activation and radiotracers and thermal ionization mass spectrometry analysis looking at the Pu isotopics.

14MeV↗

Removable coatings: Thermal stability and decontamination of steel surfaces from 241 Am

This study presents a comparative analysis of several strippable materials for decontamination of steel surfaces from 241 Am. The strippable coatings were chosen to have a history of application, to be commercially available, easily handled and have different composition and formulation. Carbon steel and stainless steel surfaces (coupons) were utilized as the most representative industrial materials, and the experimental series were expanded to include the rusting treatment of these surfaces as it is common for decommissioned old nuclear facilities. The experiments followed the path of pre-screening of the strippable materials (easiness of application, peeling etc.), their characterization with Fourier-transform infrared (FTIR) spectroscopy and thermogravimetric analysis (TGA), decontamination factors (DF) or removal efficiencies for different types of steel surfaces, and enhancement of the removal efficiencies of the select strippable coatings via amendment with EDTA. As a result, DF and removal efficiencies for stainless steel (both pristine and with rusting treatment) are higher than for pristine carbon steel, and decontamination factors and removal efficiencies are higher for pristine than for rusted carbon steel across all the coatings. It is also shown that EDTA amendment can improve DF and removal efficiency for a strippable coating (i.e., CC Strip), however other factors should also be considered. Thus, the coating CC Strip exhibited easiness in handling and high DF (up to 97% when EDTA-amended), but its drying period was the longest among the rest of the materials and its thermal decomposition indicates on a risk of intense release of heat in case of a fire event.

36 MATERIALS SCIENCE↗

Separations of U/Pu and Np/Pu using fluoride volatility

Here the importance of uranium/plutonium and neptunium/plutonium separations at large scale or at environmental levels is such that the reactivity of nitrogen trifluoride (NF 3 ) with respect to other fluorinating reagents warranted investigation. We show that NF 3 does not convert PuO 2 or PuF 4 to PuF 6 , during separations of U/Pu and Np/Pu, even under forcing conditions. Herein, this is analytically demonstrated to trace level. This rather exclusive behavior has been reported for only one other reagent: BrF 3 . Confidence in this result now allows for rapid and selective separations of volatile UF 6 and NpF 6 from plutonium. Preliminary data concerning the separation of 233 Pa from 237 Np is also described.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Fuel Fabrication Capability Assessment in Support of Advanced Reactor Deployments

More than 30 U.S. companies are designing a variety of advanced reactor concepts, and several companies are planning to demonstrate their reactor designs in the mid-2020s to late 2030s time frame. In 2020, the U.S. Department of Energy (DOE) announced a series of awards under the Advanced Reactor Demonstration Program (ARDP) to accelerate the successful deployment of 10 of these reactors under three pathways. TerraPower and X-energy were awarded grants under the Advanced Reactor Demonstration Program to deploy their respective Natrium reactor and Xe-100 reactor designs in the next 7–10 years. These demonstrations are in addition to several parallel programs, including the U.S. Department of Defense’s (DoD’s) interest in the development of microreactors, and interest of the National Aeronautics and Space Administration in space nuclear power and propulsion. The National Reactor Innovation Center’s (NRIC’s) mission is to accelerate the demonstration and deployment of advanced reactors; NRIC is partnering with several reactor developers and harnessing the world-class capabilities of the U.S. National Laboratory system to deliver on its mission. Several of these reactor designs will require advanced fuel forms that are not commercially available today, including metal fuel, molten salt fuel, TRi-structural ISOtropic (TRISO) particle fuel, and uranium nitride fuel. Recognizing that there may be potential gaps in the laboratory-scale process development and pilot-scale first-of-a-kind (FOAK) production of these fuel forms leading to delivery of the FOAK cores, NRIC commissioned this study to look at the challenges that need to be overcome for successful deliveries, including the evaluation of existing facilities and the potential need for a new fuel fabrication facility.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Measurement of Local Magnetic Fields in Actinide Tetrafluorides

Fluorine-19 magnetic shielding tensors have been measured in a series of actinide tetrafluorides (AnF4) by solid state nuclear magnetic resonance (NMR) spectroscopy. Tetravalent actinide centers with 0 to 8 valence electrons can form tetrafluorides with the same unit cell, making these compounds an attractive choice for a systematic study of the variation in electronic structure across the 5f row of the Periodic Table. Pronounced deviations from predictions based on localized valence electron models have been detected by these experiments, which suggests that this approach may be used as a quantitative probe of electronic correlations.

Walter, Eric D.↗

Ab initio Molecular Dynamics Assessment of Thermodynamic and Transport Properties in (K,Li)Cl and (K, Na)Cl Molten Salt Mixtures

Molten salt mixtures are integral part of highly important technological applications such as nuclear reactors. However, due to inherent difficulties associated with experiment at high temperatures and the intrinsic complexity of liquid-phase multi-component systems, understanding their properties at a molecular level remains a challenge. Here, we report on an ab initio molecular dynamics investigation on structural, electronic, transport, and thermal properties of two common molten salt mixtures, (K, Li)Cl and (K,Na)Cl, at five different compositions and three temperatures. Most of the properties were found to depend on both composition and temperature. While properties, like atomic charges, show additive behaviors, other properties, such as electrical conductivity, show considerable deviations from additivity. We shall show that the mixing of the molten salt mixtures is mainly driven by entropy, and that the KCl and LiCl mix better than KCl and NaCl. Our computational results are in general consistent with available experimental data. Comparison with available theoretical data is also provided.

Nguyen, Manh Thuong↗

Potential defect structure effects in the volatility separation of sub-picogram quantities of fission and activation products from an irradiated UO2 matrix

Here, the use of fluorination as a rapid separation technique for quantification of trace-mass, short-lived isotopes is illustrated. Single crystal 238 UO 2 was exposed to a 14 MeV neutron source to produce mixed fission and activation products. Selective removal of trace volatile fluorides from non-volatile ones produced two results of interest. A decrease in background produced better detection and quantification of select analytes across the gamma spectrum. Second, at the very low burnup condition described, the volatility behaviors of common isotopes were altered relative to their well-established behaviors in higher burnup metal and metal oxide fuels.

Fission and activation product separations↗