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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 307 records · Page 17
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Identification of $^{3}$He–$^{3}$H clusters in the $^{6}$Li+$^{89}$Y experiment using particle-$\gamma$ coincidence measurement

The 6 Li+ 89 Y experiment was performed to explore the reaction mechanism induced by a weakly bound nucleus 6 Li and its cluster configuration. Here, the particle-$\gamma$ coincidence method was used to identify the different reaction channels. The $\gamma$-rays coincident with 3 He/ 3 H indicate that the 3 H/ 3 He stripping reaction plays a significant role in the formation of Zr/Nb isotopes. The obtained results support the existence of a 3 He- 3 H cluster in 6 Li. Direct and sequential transfer reactions are adequately discussed, and the FRESCO code is used to perform precise finite-range cyclic redundancy check calculations. In the microscopic calculation, direct cluster transfer is more predominant than sequential transfer in 3 H transfer. However, the direct cluster transfer is of comparable magnitude to the sequential transfer in the 3 He transfer.

CRC calculations↗
DOE OSTI2022

High Flux Isotope Reactor Low Enriched Uranium U-10Mo Fuel Design Parameters

Activities to convert the HFIR from HEU to LEU are ongoing as part of the US Department of Energy (DOE) National Nuclear Security Administration (NNSA) nuclear nonproliferation mission. Design activities to study the conversion of HFIR from HEU to LEU fuel explored different fuel design features and shapes with a uranium-molybdenum (U-10Mo) monolithic alloy fuel. This high-density alloy contains 90 wt % uranium and 10 wt % molybdenum and has a uranium density of 15.318gU/cm 3 . The goal of these studies is to generate several candidate HFIR LEU fuel designs of varying fuel fabrication complexity that meet the current HEU performance metrics and safety requirements. Recent advancements in modeling and simulation tools and design methods enabled a thorough analysis of the available design space with U-10Mo fuel. A surrogate model used this analysis as training data to quickly determine the performance of a design given specific design parameters. An optimization module used this surrogate model to quickly search this multidimensional search space given specific desired performance characteristics. This approach was made possible by the large available design space with U-10Mo fuel. Shift, a Monte Carlo tool optimized for high-performance computing (HPC) architectures, was used for faster calculation and better data management for reactor physics simulations. Once most of these design studies were complete, a new suite called the Python HFIR Analysis and Measurement Engine (PHAME) was developed to connect all fuel design analysis steps, making design studies more efficient and reproducible. The post-processing capabilities of these new tools are leveraged for the information provided herein. Leveraging these tools, several candidate fuel designs were selected with varying levels of feature complexity and reactor performance. This report provides design feature details for four selected HFIR LEU U-10Mo fuel designs and their corresponding performance and safety metrics. Nominal best-estimate design parameters and irradiation conditions, including fission rate densities, power densities, heat fluxes, and cumulative fission densities, are provided. Simulations show that the high uranium density of U-10Mo fuel provides a large potential design space that enables various LEU designs to meet HEU core performance metrics and safety requirements with a power increase from 85 MW (HEU) to 95 MW or 100 MW (LEU).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

The Baghdad Atlas: A relational database of inelastic neutron-scattering (n,n ' γ) data

A relational database has been developed based on the original (n,n'γ) work carried out by A. M. Demidov et al., at the Nuclear Research Institute in Baghdad, Iraq (Demidov et al., 1978) for 105 independent measurements comprising 76 elemental samples of natural composition and 29 isotopically-enriched samples. The information from this Atlas includes: γ-ray energies and relative intensities; nuclide and level data corresponding to the residual nucleus and meta data associated with the target sample that allows for the extraction of the flux-weighted (n,n'γ) cross sections for a given transition relative to a defined value. The optimized angular-distribution-corrected fast-neutron flux-weighted partial γ-ray cross section for the production of the 846.8-keV 21+→0gs+γ-ray transition in 56Fe, determined to be $\langle$σγ$\rangle$=143(29) mb, is used for this purpose. However, different values for the adopted cross section can be readily implemented to accommodate user preference based on revised determinations of this quantity. The Atlas (n,n'γ) data has been compiled into a series of CSV-style ASCII data sets and a suite of Python scripts have been developed to build and install the database locally. The database can then be accessed directly through the SQLite engine, or using alternative methods such as the Jupyter Notebook Python-browser interface. Several examples exploiting different interaction methodologies are distributed with the complete software package.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

New thermal decomposition pathway for TATB

Abstract Understanding the thermal decomposition behavior of TATB (1,3,5-triamino-2,4,6-trinitrobenzene) is a major focus in energetic materials research because of safety issues. Previous research and modelling efforts have suggested benzo-monofurazan condensation producing H 2 O is the initiating decomposition step. However, early evolving CO 2 (m/z 44) along with H 2 O (m/z 18) evolution have been observed by mass spectrometric monitoring of head-space gases in both constant heating rate and isothermal decomposition studies. The source of the CO 2 has not been explained, until now. With the recent successful synthesis of 13 C 6 -TATB ( 13 C incorporated into the benzene ring), the same experiments have been used to show the source of the CO 2 is the early breakdown of the TATB ring, not adventitious C from impurities and/or adsorbed CO 2 . A shift in mass m/z 44 (CO 2 ) to m/z 45 is observed throughout the decomposition process indicating the isotopically labeled 13 C ring breakdown occurs at the onset of thermal decomposition along with furazan formation. Partially labeled (N 18 O 2 ) 3 -TATB confirms at least some of the oxygen comes from the nitro-groups. This finding has a significant bearing on decomposition computational models for prediction of energy release and deflagration to detonation transitions, with respect to conditions which currently do not recognize this oxidation step.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Structural, Criticality, and Radiation Dose Calculations to Support SNF Loading into a DOE Standard Canister

The DOE Standard Canister Demonstration Project includes the development of an internal support structure (ISS) for the 4.6-m-long and 45.7-cm-diameter canister. This study presents structural evaluations, criticality safety assessments, and dose rate calculations conducted within the scope of the ISS design process to support smooth canister loading operations and to ensure safe storage, transportation, and disposal of Peach Bottom 1 Core II (PB2) and Fort St. Vrain (FSV) SNF, currently stored at INL’s CPP-603 facility. The ISS includes a 316L stainless-steel basket that holds 12 PB2 SNF rods. The PB2 basket rests on top of six individual, 78.7-cm-long, 316L stainless-steel columns that are equally spaced and welded to the inner canister wall at the lower end of the shell. These columns represent the FSV basket and can hold one FSV SNF element. An A92014 T6 aluminum spacer disc is bolted to the bottom plate of the PB2 basket to vertically restrain the FSV SNF element after the PB2 basket is placed inside the canister above the FSV basket. The structural evaluations of the ISS followed applicable ASME BPVC.III.3 guidelines and included finite element (FE) analyses of the PB2 basket structure; analyses of welds and bolds; buckling analyses of selected components, and acceptability assessments of the expected basket deformations under loading operations. The criticality safety assessments used the Monte Carlo N-Particle (MCNP) software architecture Version 6.2, including ENDF/B-V continuous-energy cross-section libraries, considering intact SNF in a single storage overpack or two multistorage overpack configurations and intact or failed SNF configured for disposal. The dose rate computations are based on source terms taken from the DOE Spent Fuel Database. The isotopic composition was decay corrected for the year 2022 using the ORIGEN module in the SCALE suite. A 19-group photon spectrum and a 27-group neutron-source spectrum were generated and used in MCNP to calculate estimated dose-equivalent rates, both on DOE Standard Canister contact and at a radial distance of 1 m from the canister surface. The results of this study indicate a structurally sound system that can uphold its criticality safety functions throughout its intended operational phases. Furthermore, this study provides confidence that sufficient radiological protection is technically achievable.

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

Autotrophic and mixotrophic metabolism of an anammox bacterium revealed by in vivo 13 C and 2 H metabolic network mapping

Anaerobic ammonium-oxidizing (anammox) bacteria mediate a key step in the biogeochemical nitrogen cycle and have been applied worldwide for the energy-efficient removal of nitrogen from wastewater. However, outside their core energy metabolism, little is known about the metabolic networks driving anammox bacterial anabolism and use of different carbon and energy substrates beyond genome-based predictions. Here, we experimentally resolved the central carbon metabolism of the anammox bacterium Candidatus ‘Kuenenia stuttgartiensis’ using time-series 13 C and 2 H isotope tracing, metabolomics, and isotopically nonstationary metabolic flux analysis. Our findings confirm predicted metabolic pathways used for CO 2 fixation, central metabolism, and amino acid biosynthesis in K. stuttgartiensis , and reveal several instances where genomic predictions are not supported by in vivo metabolic fluxes. This includes the use of the oxidative branch of an incomplete tricarboxylic acid cycle for alpha-ketoglutarate biosynthesis, despite the genome not having an annotated citrate synthase. We also demonstrate that K. stuttgartiensis is able to directly assimilate extracellular formate via the Wood–Ljungdahl pathway instead of oxidizing it completely to CO 2 followed by reassimilation. In contrast, our data suggest that K. stuttgartiensis is not capable of using acetate as a carbon or energy source in situ and that acetate oxidation occurred via the metabolic activity of a low-abundance microorganism in the bioreactor’s side population. Together, these findings provide a foundation for understanding the carbon metabolism of anammox bacteria at a systems-level and will inform future studies aimed at elucidating factors governing their function and niche differentiation in natural and engineered ecosystems.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI1968

Nuclear Physics Made Very, Very Easy

The fundamental approach to nuclear physics was prepared to introduce basic reactor principles to various groups of non-nuclear technical personnel associated with NERVA Test Operations. NERVA Test Operations functions as the field test group for the Nuclear Rocket Engine Program. Nuclear Engine for Rocket Vehicle Application (NERVA) program is the combined efforts of Aerojet-General Corporation as prime contractor, and Westinghouse Astronuclear Laboratory as the major subcontractor, for the assembly and testing of nuclear rocket engines. Development of the NERVA Program is under the direction of the Space Nuclear Propulsion Office, a joint agency of the U. S. Atomic Energy Commission and the National Aeronautics and Space Administration. This report is being reprinted for use in the U. S. Atomic Energy Commission and National Aeronautics and Space Administration educational and technology utilization programs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Qualitative and quantitative analysis of neutron irradiation effects in SiC/SiC composites using X-ray computed tomography

Silicon carbide fiber-reinforced silicon carbide matrix (SiC/SiC) composites are candidate materials for cladding of light water reactor (LWR) fuels. Loss of fission product gas retention due to the formation of microcrack networks is considered a potential failure mechanism for SiC/SiC-cladded fuels. In this work, a variety of SiC/SiC composite tubes were irradiated with and without an LWR-relevant radial heat flux in the High Flux Isotope Reactor, followed by detailed characterization with X-ray computed tomography (XCT). This first set of XCT data for neutron-irradiated samples confirmed that the internal stresses arising from a combination of temperature gradients and irradiation-induced swelling act as the primary driver for cracking. Consequently, while the observed cracking patterns varied depending on the tube architectures, the sharp edges of relatively large pores were found to be the common stress concentrator. These findings are useful to help improve the design and manufacturing of SiC/SiC fuel claddings for reduced failure probability.

42 ENGINEERING↗
DOE OSTI2026

Ceramic Composite Experimental Testing Status

Over recent years, ceramic matrix materials such as SiC–SiC and C–C have been gaining interest for use in fusion reactors, light water reactors (LWRs), and high-temperature reactors (HTRs). These materials are good candidates to operate in very high temperature and moderate to high radiation environments. The evaluation of composite materials, in general, is challenging because of variations in precursor materials, variations in the fabrication process across fabricators, and the wide range of potential fiber architectures, to name a few. However, the need to evaluate neutron-irradiated properties adds another layer of complexity, which includes cost, timeline, and specimen size limitations (often associated with irradiation testing). A qualification methodology for the use of ceramic composites is provided in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section III-5-HHB. The methodology is supported by ASTM International (ASTM) guides, which provide a pathway to accomplish this effort. Part of the qualification strategy is for the designer to collect material property data on environmental conditions representative of its design envelope. These data include irradiation effects. This report presents an experimental study and test campaign developed to partially address this gap by providing initial mechanical and physical property data required for design. A variety of different materials using different manufacturing techniques are considered as part of this campaign. The test plan suggests performing a screening or partial irradiation study to assist the designer during the material selection process. The designer can then perform a more comprehensive qualification study if the material performance is promising. This work focuses on the status of the specimen preparations (machining of samples), the current test methods and failure analysis as well as the preparation of irradiation vehicles for the irradiation campaign. The irradiation will be performed at Oak Ridge National Laboratory (ORNL) in the High Flux Isotope Reactor (HFIR) and at Idaho National Laboratory (INL) in the Advanced Test Reactor (ATR).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS2013

Neutron Characterization for Additive Manufacturing

Oak Ridge National Laboratory (ORNL) is leveraging decades of experience in neutron characterization of advanced materials together with resources such as the Spallation Neutron Source (SNS) and the High Flux Isotope Reactor (HFIR) shown in Fig. 1 to solve challenging problems in additive manufacturing (AM). Additive manufacturing, or three-dimensional (3-D) printing, is a rapidly maturing technology wherein components are built by selectively adding feedstock material at locations specified by a computer model. The majority of these technologies use thermally driven phase change mechanisms to convert the feedstock into functioning material. As the molten material cools and solidifies, the component is subjected to significant thermal gradients, generating significant internal stresses throughout the part (Fig. 2). As layers are added, inherent residual stresses cause warping and distortions that lead to geometrical differences between the final part and the original computer generated design. This effect also limits geometries that can be fabricated using AM, such as thin-walled, high-aspect- ratio, and overhanging structures. Distortion may be minimized by intelligent toolpath planning or strategic placement of support structures, but these approaches are not well understood and often "Edisonian" in nature. Residual stresses can also impact component performance during operation. For example, in a thermally cycled environment such as a high-pressure turbine engine, residual stresses can cause components to distort unpredictably. Different thermal treatments on as-fabricated AM components have been used to minimize residual stress, but components still retain a nonhomogeneous stress state and/or demonstrate a relaxation-derived geometric distortion. Industry, federal laboratory, and university collaboration is needed to address these challenges and enable the U.S. to compete in the global market. Work is currently being conducted on AM technologies at the ORNL Manufacturing Demonstration Facility (MDF) sponsored by the DOE's Advanced Manufacturing Office. The MDF is focusing on R&D of both metal and polymer AM pertaining to in-situ process monitoring and closed-loop controls; implementation of advanced materials in AM technologies; and demonstration, characterization, and optimization of next-generation technologies. ORNL is working directly with industry partners to leverage world-leading facilities in fields such as high performance computing, advanced materials characterization, and neutron sciences to solve fundamental challenges in advanced manufacturing. Specifically, MDF is leveraging two of the world's most advanced neutron facilities, the HFIR and SNS, to characterize additive manufactured components.

Watkins, Thomas↗
NASA NTRS2006

Overview of Mars Science Laboratory (MSL) Environmental Program

This viewgraph presentation is an overview of the Mars Science Laboratory (MSL) program. The engineering objectives of the program are to create a Mobile Science Laboratory capable of one Mars Year surface operational lifetime (670 Martian sols = 687 Earth days). It will be able to land and operation over wide range of latitudes, altitudes and seasons It must have controlled propulsive landing and demonstrate improved landing precision via guided entry The general science objectives are to perform science that will focus on Mars habitability, perform next generation analytical laboratory science investigations, perform remote sensing/contact investigations and carry a suite of environmental monitoring instruments. Specific scientific objectives of the MSL are: (1) Characterization of geological features, contributing to deciphering geological history and the processes that have modified rocks and regolith, including the role of water. (2) Determination of the mineralogy and chemical composition (including an inventory of elements such as C, H, N, O, P, S, etc. known to be building blocks for life) of surface and near-surface materials. (3) Determination of energy sources that could be used to sustain biological processes. (4) Characterization of organic compounds and potential biomarkers in representative regolith, rocks, and ices. (5) Determination the stable isotopic and noble gas composition of the present-day bulk atmosphere. (6) Identification potential bio-signatures (chemical, textural, isotopic) in rocks and regolith. (7) Characterization of the broad spectrum of surface radiation, including galactic cosmic radiation, solar proton events, and secondary neutrons. (8) Characterization of the local environment, including basic meteorology, the state and cycling of water and C02, and the near-surface distribution of hydrogen. Several views of the planned MSL and the rover are shown. The MSL environmental program is to: (1) Ensure the flight hardware design is capable of surviving all the environments throughout its mission life time, including ground, transportation, launch, cruise, entry decent and landing (EDL) and surface operation environments. (2) Verify environmental testing and analysis have adequately validated the flight hardware's ability to withstand all natural, self-induced, and mission-activity-induced environments. The planned tests to ascertain the capability of the MSL to perform as desired are reviewed.

spacecraft↗
DOE OSTI2021

High Flux Isotope Reactor Low-Enriched Uranium Low Density Silicide Fuel Design Parameters

High Flux Isotope Reactor (HFIR) highly enriched uranium (HEU) to low-enriched uranium (LEU) conversion activities are ongoing as part of the Department of Energy (DOE) National Nuclear Security Administration (NNSA)’s nuclear nonproliferation mission. Design activities studying the conversion of HFIR from HEU to LEU fuel explored different fuel design features and shapes with a low density uranium-silicide dispersion (U 3 Si 2 -Al) fuel, which has a uranium density of 4.8 gU/cm 3 . The goal of these studies is to generate several HFIR LEU fuel designs of varying fuel fabrication complexity that meet the current HEU performance metrics and safety requirements. The documented designs will serve as references for fuel fabrication and qualification activities. Recent advancements in modeling and simulation tools enable quick prototyping of fuel designs. Shift, a Monte Carlo neutron transport and depletion tool optimized for high-performance computing (HPC) architectures, is used for efficient fuel cycle and performance metrics calculations. The HFIR Steady State Heat Transfer Code (HSSHTC) is used to vet the thermal safety margin. Also, a new automation tool that connects all fuel design analysis steps, named Python HFIR Analysis and Measurement Engine (PHAME), has been developed to expedite the design study in an efficient and reproducible manner. Leveraging these tools, several candidate fuel designs were selected for varying fabrication complexity. This report provides design feature details for four selected HFIR LEU low density U 3 Si 2 -Al fuel designs and their corresponding performance and safety metrics. Nominal, best-estimate design parameters and irradiation conditions, including fission rate densities, power densities, heat fluxes, and cumulative fission densities are provided for candidate fuel designs relevant to framing irradiation experiments to support fuel qualification efforts. Simulations show that the low density U 3 Si 2 -Al, with design features to enhance safety, can meet HEU core performance metrics and safety requirements if the reactor power is increased from 85 MW (HEU) to 95 MW (LEU) and if the active fuel length is increased from 50.80 cm (HEU) to 55.88 cm (LEU).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2001

Space Fission Propulsion System Development Status

The world's first man-made self-sustaining fission reaction was achieved in 1942. Since then fission has been used to propel submarines, generate tremendous amounts of electricity, produce medical isotopes, and provide numerous other benefits to society. Fission systems operate independently of solar proximity or orientation, and are thus well suited for deep space or planetary surface missions. In addition, the fuel for fission systems (enriched uranium) is virtually non-radioactive. The primary safety issue with fission systems is avoiding inadvertent system start. Addressing this issue through proper system design is straight-forward. Despite the relative simplicity and tremendous potential of space fission systems, the development and utilization of these systems has proven elusive. The first use of fission technology in space occurred 3 April 1965 with the US launch of the SNAP-10A reactor. There have been no additional US uses of space fission systems. While space fission systems were used extensively by the former Soviet Union, their application was limited to earth-orbital missions. Early space fission systems must be safely and affordably utilized if we are to reap the benefits of advanced space fission systems. NASA's Marshall Space Flight Center, working with Los Alamos National Laboratory (LANL), Sandia National Laboratories, and others, has conducted preliminary research related to a Safe Affordable Fission Engine (SAFE). An unfueled core has been fabricated by LANL, and resistance heaters used to verify predicted core thermal performance by closely mimicking heat from fission. The core is designed to use only established nuclear technology and be highly testable. In FY01 an energy conversion system and thruster will be coupled to the core, resulting in an 'end-to-end' nuclear electric propulsion demonstrator being tested using resistance heaters to closely mimic heat from fission. Results of the SAFE test program will be presented. The applicability of a SAFE-powered electric propulsion system to outer planet science missions will also be discussed.

Houts, M.↗
DOE OSTI2024

Quantitative near-field water–air spray measurements at elevated pressures by neutron radiography imaging

Extensive experimental research on high-pressure spray has been conducted for decades to deepen our understanding and optimize its use in transportation, aviation, and propulsion applications; however, the near-field and in-nozzle flow characteristics are not fully understood. Dense near-field spray is among the most challenging diagnostic tasks since light is severely scattered and diffused by the liquid droplets and columns. In this work, the near-field spray and in-nozzle flow characteristics of an aeration nozzle at elevated pressures were characterized by neutron radiography imaging at the Oak Ridge National Laboratory High Flux Isotope Reactor. Neutron imaging benefits via strong penetration depths for some metals (i.e., aluminum, lead, and steel) and is sufficiently sensitive to detection of light elements, especially for hydrogen-based molecules, due to the large incoherent scattering cross section of neutrons. Both two-dimensional snapshots of the near-field spray and a three-dimensional tomographic scan of the nozzle geometry and in-nozzle water were obtained. This work provides new quantitative characterization of practical metal nozzle geometry for accurate boundary conditions, internal flow patterns inside the nozzle, and high-pressure spray flows. In conclusion, the findings may be used to improve performance and operating conditions of transportation vehicles and propulsion systems.

42 ENGINEERING↗
DOE OSTI2020

$β$-delayed-neutron studies of 135,136 Sb and 140 I performed with trapped ions

In this work, Beta-delayed-neutron $(βn)$ spectroscopy was performed using the Beta-decay Paul Trap and an array of radiation detectors. The $βn$ branching ratios and energy spectra for 135,136 Sb and 140 I were obtained by measuring the time of flight of recoil ions emerging from the trapped ion cloud. These nuclei are located at the edge of an isotopic region identified as having $βn$ branching ratios that impact the $r$-process abundance pattern around the A ≈ 130 peak. For 135,136 Sb and 140 I, $βn$ branching ratios of 14.6(13)%, 17.6(29)%, and 7.6(28)% were determined, respectively. The $βn$ energy spectra obtained for 135 Sb and 140 I are compared with results from direct neutron measurements, and the $βn$ energy spectrum for 136 Sb has been measured for the first time.

42 ENGINEERING↗
DOE OSTI2024

Specification of FIPD Fission Gas Chemistry Data

The current FIPD library contains two main sets of fission gas chemistry data. The first set is data collected during the Integral Fast Reactor (IFR) program from 1984 to 1994, using a gas mass spectrometry system located in the Analytical Laboratory (AL) at Argonne National Laboratory-West (Argonne-West). Throughout this period, numerous fission gas release and chemistry datasets were gathered from a variety of metallic fuel pins. The fission gas was sampled by the Gas Assay, Sample and Recharge (GASR) System in the Hot Fuel Examination Facility (HFEF) and transferred to the AL to perform gas composition and isotopic abundance analysis. The second set is data collected after the IFR program. The fission gas samples were also collected by the GASR system at the HFEF, but analyzed using a similar gas mass spectrometer located in Pacific Northwest National Laboratory (PNNL). Many fuel pins irradiated in Experimental Breeder Reactor II (EBR-II) and the Fast Flux Test Facility (FFTF) were measured, including the fuel pins for the MFF series of experiments, designed to qualify metal fuel for use as driver fuel in the FFTF and X496 experiment. For either set of data, fission gas was sampled with the gas sampling line in GASR using sample bottles after the capsule/element volume has been determined and the system is still full of radioactive gas. The sample bottles were then transferred to the sample packaging cylinder or an approved storage location pending transfer to the AL or prepared for shipment to another laboratory (such as PNNL) for analysis of the collected gas as directed on the GASR data form, other approved form. The receiving laboratories (AL or PNNL) required their Analytical Service Request form to be completed prior to sample transfer. Typical sample transfer processes were initiated at HFEF by the principal or process engineer. The laboratories performing the analyses (AL or PNNL) use the sample bottle numbers as well as a sample number produced by the respective laboratory. HFEF and the responsible experimenter tracked the sample using the analysis number, the gas bottle number, and the fuel pin number.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2025

RIMS analysis of isotopically tagged uranium particles with application to Intentional Forensics

Traditional nuclear forensics approaches leverage isotopic measurements for characterizing nuclear forensics signatures in fuel cycle materials. To aid in provenance assessment, isotopically perturbed transition metal taggants can be added to fuel cycle materials. We present a case study for examining natural uranium oxide powder that was tagged with isotopically perturbed Mo before and after irradiation using resonance ionization mass spectrometry. In conclusion, this method accurately and precisely measures taggant and uranium compositions, rapidly enabling the ability to clearly discriminate between tagged particles and other material with application for robust assessment of material provenance.

Forensic Anthropology↗
NASA NTRS1987

Chemical engineering design of CO oxidation catalysts

How a chemical reaction engineer would approach the challenge of designing a CO oxidation catalyst for pulsed CO2 lasers is described. CO oxidation catalysts have a long history of application, of course, so it is instructive to first consider the special requirements of the laser application and then to compare them to the characteristics of existing processes which utilize CO oxidation catalysts. All CO2 laser applications require a CO oxidation catalyst with the following characteristics: (1) active at stoichiometric ratios of O2 and CO, (2) no inhibition by CO2 or other components of the laser environment, (3) releases no particulates during vibration or thermal cycling, and (4) long lifetime with a stable activity. In all applications, low consumption of power is desirable, a characteristic especially critical in aerospace applications and, thus, catalyst activity at low temperatures is highly desirable. High power lasers with high pulse repetition rates inherently require circulation of the gas mixture and this forced circulation is available for moving gas past the catalyst. Low repetition rate lasers, however, do not inherently require gas circulation, so a catalyst that did not require such circulation would be favorable from the standpoint of minimum power consumption. Lasers designed for atmospheric penetration of their infrared radiation utilize CO2 formed from rare isotopes of oxygen and this application has the additional constraint that normal abundance oxygen isotopes in the catalyst must not exchange with rare isotopes in the gas mixture.

Herz, Richard K.↗