Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “Pebble”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 271 records · Page 15
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Examining Graphite Degradation in Molten Salt Environments: A Chemical, Physical, and Material Analysis

Molten-salt reactors (MSRs) are Generation IV nuclear reactors that use liquid salt as a coolant and/or fuel. In several MSR designs, graphite serves as a moderator and/or reflector. However, due to limited experimental data and operational experience, our understanding of graphite behavior in molten salt environments remains incomplete. This report aims to identify the degradation mechanisms of nuclear graphite in MSRs, detail the mechanisms of each factor, and provide an initial assessment of their impact on the structural integrity of graphite components. This assessment is based on an extensive literature review and insights from subject matter experts. Furthermore, given the limited data, a modeling strategy using existing Grizzly software is proposed for a more thorough analysis where appropriate. Additionally, it presents mitigation strategies where applicable. The report covers physical degradation mechanisms such as infiltration, erosion, and abrasion, as well as chemical degradation mechanisms including fluorination, intercalation, corrosion, and oxidation. Molten salt can infiltrate the porous structure of graphite, leading to several detrimental effects. Entrapment of fissile products within the graphite pores can cause radiation damage and could pose challenges in the handling and disposal of contaminated components. The differential thermal expansion between the infiltrated salt and graphite, along with internal stress from pressurized molten salt and volumetric heating, can compromise the structural integrity of graphite. To mitigate these effects, employing ultra-fine graphite grades and applying sealants and coatings are effective strategies. A computational model based on coupled solid mechanics and heat transfer phenomena could be used to predict the internal stresses using Grizzly software. In pebble-bed MSRs, graphite fuel pebbles can cause abrasion against reactor components due to friction and wear. The severity of wear is influenced by various factors such as temperature, environment, and the presence of lubricants. Tribological studies reveal that higher temperatures and molten salt environments, such as FLiBe, significantly reduce wear rates compared to dry conditions. Additionally, the chemical composition of the salt can further optimize graphite's tribological performance. Long-term wear effects can be modeled by incorporating surface defects into the geometry and predict stresses under thermal and radiation effects using Grizzly software. Chemical degradation of graphite in a molten salt environment can occur through fluorination and intercalation. Fluorination can occur via replacement of hydrogen or oxygen atoms, or at the active sites, but does not cause structural degradation. Intercalation, on the other hand, can lead to exfoliation, where layers of graphite separate and peel away, damaging the graphite. Protective coatings can enhance graphite's resistance to intercalation. Graphite generally exhibits good chemical stability in molten salt environments, though it can corrode under specific conditions, particularly in the presence of impurities or oxidants. Studies have shown that protective coatings, such as plasma-sprayed partially stabilized zirconia (PSZ), can effectively prevent such degradation. Corrosion behavior varies significantly with different graphite grades and coating applications, underscoring the need for detailed studies on uncoated and coated graphite to understand and mitigate corrosion mechanisms in MSRs. Research indicates that the presence of oxidants and impurities can accelerate graphite degradation in molten salts, making it essential to explore acceptable impurity limits. Oxidation is another critical degradation mechanism, leading to weight loss and structural damage due to the formation of CO and CO 2 from the reaction of carbon atoms with oxygen. This process creates new porosity and compromises graphite's integrity. While extensive research on graphite oxidation has been conducted for gas-cooled reactors, studies specific to MSRs are limited. Findings from the coal industry suggest that molten alkali metal salts can significantly accelerate graphite oxidation, a hypothesis worth exploring for fluoride salts in MSRs. Understanding oxidation behavior in MSRs is vital for developing protective measures. The analysis of post-irradiated graphite from the MSRE experiment demonstrated exceptional chemical compatibility with molten fluoride salt, suggesting that the extent of chemical attack on graphite largely depends on the salt's infiltration capability. Therefore, the use of ultra-fine grade graphite could help mitigate chemical degradation effects. Existing oxidation modeling capabilities in Grizzly, which use reaction-diffusion equations to model graphite-air interactions, could be adapted to simulate the chemical degradation effects of graphite in molten salt environments.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Identify and Assess Technical Challenges in Safeguards Measurements of Spent Advanced Reactor Fuels

Advanced reactor (AR) designs use various nuclear fuel types that can be significantly different than conventional light-water reactor (LWR) fuels, including differences in sizes, compositions, and chemical forms (e.g., oxide, carbide, metal). Nearly all the proposed AR fuels use high-assay low-enriched uranium (HALEU), which will have higher enrichments (5–20 wt% 235 U) than LWR fuels (currently limited to <5 wt% 235 U). In advance of the wide use of these new fuel types around the world, international safeguards organizations such as the International Atomic Energy Agency (IAEA]) are working with some of the AR vendors to formulate safeguards approaches for these AR fuel cycles. As part of the overall safeguards approach, it is important to identify the potential technical challenges in performing safeguards verification measurements of these AR fuels (both fresh and spent fuels) in advance of the widespread adoption of these new fuel types, because new safeguards technologies can take several years to develop, test, and approve for use. This report documents work performed in fiscal year 2024 based on modeling and simulation to assess the performance of the existing safeguards measurement technologies for irradiated or spent AR fuel elements or items. This work is a continuation of the work performed in fiscal year 2023 that focused on fresh AR fuels. Spent AR fuels have a distinct difference from their LWR counterparts: unlike the spent LWR fuels typically stored in a water-filled pool, some spent AR fuels—such as tristructural-isotropic (TRISO)-based fuels—will most likely be stored in air-filled hot cells. Because most safeguards measurements on spent fuel performed to date have been conducted under water, the air-filled hot cell environment could present unique challenges to safeguards measurements. Fork detector (FDET) and Cerenkov viewing device (CVD) systems have been the two primary instruments used by the IAEA for several decades to measure spent LWR fuel assemblies stored in pools for safeguards verification purposes. Because the lower refractive index of air causes Cerenkov light to be of lower intensity in air than in water, existing CVDs are likely unable to perform safeguards verification measurements for spent fuel stored in an air-filled hot cell, as is the case for the TRISO-based spent fuel elements (e.g., pebbles, graphite fuel blocks). Unlike FDET measurements, CVD measurements do not require fuel be moved, so they are a simpler and faster to take than FDET measurements. The inability to perform CVD measurements on the TRISO-based AR fuel types presents a major technical challenge in the effort to use existing technology to perform safeguards measurements on spent AR fuels. This study was mainly conducted through the modeling and simulation of an FDET or an FDET-like system on five spent AR fuel types, including one metallic fuel type and four TRISO-based fuel types in both pebble and graphite block forms in their respective storage configurations and environments. Because the various AR fuel types have significantly different dimensions, FDET systems must be adapted to accommodate them. Partial defect tests were also simulated in this study to assess the FDET’s ability to detect potential fuel diversions. The FDET measures the fuel’s total passive neutron and gamma emissions. The simulated FDET results from spent AR fuel items are compared against results from a typical spent pressurized water reactor (PWR) assembly. High-purity germanium (HPGe) gamma detector measurements were also simulated for the spent AR fuel types and the PWR assembly because the signature photopeaks have been used in LWR safeguards verifications, although HPGe is usually not used to detect diversions because of the fuel’s self-attenuation effects on those photopeaks. The results indicate that these detectors have significant challenges in performing safeguards measurements of the spent AR fuel items, including incompatibilities between AR fuel items and existing FDETs, lower neutron count rates, lower sensitivities to fuel diversions in certain AR fuel items, and significantly higher interference from a neighboring fuel item when the measurement is performed in air. These results suggest that an alternative technology or significant and timely technology development is needed to perform adequate safeguards measurements of some of these AR fuel items.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

High-Temperature Gas-Cooled Reactors Multiphysics Simulation Demonstration and Code Validation

This study presents a comprehensive benchmarking and verification effort of several thermal-hydraulic and multiphysics capabilities for high-temperature gas-cooled reactor applications. The first part of this effort focuses on the running-in verification of Griffin’s multiphysics capabilities, specifically for simulating the evolution of pebble-bed reactor cores from startup to equilibrium. Since Fiscal Year 2024, improvements and enhancements have been implemented in Griffin, including simplifying the process to specify streamlines and developing the online cross-section generation capability. In the absence of validation data, code-to-code comparisons are conducted with kugelpy, showing good agreement for integral quantities like k-eff predictions and predictions for maximum power density. However, accuracy issues are noted for more detailed quantities like the spatial distribution of fission rate densities which will require further work to address. The second part of this report presents an improved System Analysis Module (SAM) core channel model where the effects of cross flow are considered during the pressurized loss of forced cooling transient, resulting in an improved agreement of the predicted pebble temperature with respect to the predictions from the SAM 2D porous media model. Additionally, the wall channeling effect due to variable porosity at the near wall region of the core is also investigated. Furthermore, to demonstrate Griffin’s online cross-section generation capability, a Multiphysics simulation is performed by coupling Griffin to the SAM core channel model. In the third part of the report, as a part of the Organisation for Economic Co-operation and Development/Nuclear Energy Agency (OECD/NEA) thermal-hydraulic code validation benchmark activity for a high-temperature gas-cooled reactor, the High Temperature Test Facility (HTTF) is investigated first using the NekRS computational fluid dynamics (CFD) code to study the flow mixing phenomenon in the lower plenum of the facility. Then, code-to-code and code-to-data comparisons are performed for Test PG27, which is a pressurized conduction cooldown (PCC) test, using five different codes by six organizations from five countries. The different simulations show good agreements in terms of the general trend but there are differences in some results such as the peak temperatures of different regions and heat removal rate.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Gamma and Neutron Measurement and Modeling of Irradiated TRISO Fuel

Given the unique characteristics of the PBR fuel cycle, both gamma and neutron measurements are expected to play important roles in performing and maintaining nuclear material control and accounting for spent pebbles to safeguard the fuel cycle. Given the lack of irradiated pebbles in the US, a variety of irradiated TRISO fuel samples with wide ranges of burnups and cooling times available at ORNL were used in this work. A large number of gamma and neutron measurements have been performed on these samples to collect data to test the various detectors and to benchmark the computer models to simulate the depletion and decay of the fuel and the measurements themselves. Two neutron detectors, including a custom-made detector and the Very High-Performance Neutron Multiplicity Counting, were used to measure the neutrons emitted by these TRISO samples. Three gamma spectrometry detectors, including an HPGe and the M400 CZT detector, were used to measure gamma-ray emissions from these samples. The M400 was recently adopted by the IAEA for fresh uranium measurements, but it was tested for spent fuel measurements prior to this project. Detailed MCNP models were developed to simulate these neutron and gamma measurements. Some GADRAS models were also developed to cross check the MCNP models for the gamma measurements. It was found challenging to perform neutron measurements in the hot cell due to the high background counts. Close agreements were observed between the simulated and measured neutron count rates in both detectors’ measurements of californium calibration sources. Both the HPGe and M400 detectors were able to measure the 604 and 662 keV peaks from these samples, which are the two most important peaks used to infer fuel burnup. Although the M400 detector did not have nearly good energy resolution and did not detect some of the minor peaks as the HPGe detector, it was found to be capable of handling significantly higher dose rates than HPGe. Given the complexities in the TRISO samples (e.g., different samples sizes) and uncertainties in the alignments between the detector and the TRISO fuel inside the containers, large scatters were found between the peak area rates and the samples’ burnups. However, the 604/662 peak ratios were found to trend well with the samples’ burnups among most samples in both measured and simulated results.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Cross Section Generation Capability in Griffin

The Griffin code is a Multiphysics Object-Oriented Simulation Environment (MOOSE) based reactor multiphysics analysis application jointly developed by Idaho National Laboratory and Argonne National Laboratory. The code includes a variety of steady-state solvers for fixed-source, k-eigenvalue, adjoint, and subcritical multiplication, as well as transient solvers for point-kinetics, improved quasi-static, and spatial dynamics. The code reads multigroup cross sections in the ISOXML format generated from external deterministic or Monte Carlo cross section generation codes. The implementation of the cross section generation capability in Griffin was initiated last year by plugging in the cross section application programming interface (CSAPI) and reviewing the methodologies for treating particulate fuels. The focus this year was on improving the CSAPI integration and implementing advanced self-shielding methods for applications to advanced reactor problems with TRISO fuels. First, the process for cross section library generation was updated to accurately and rigorously produce isotopic cross section data. Second, the on-the-fly slowing down method for the resonance treatment was implemented in CSAPI to improve the accuracy of effective multigroup cross sections in the resonance energy range. Among various on-the-fly slowing down methods, the equivalent Dancoff factor cell method was employed. Third, the iterative local spatial self-shielding method was implemented under the calculation framework of the equivalent Dancoff factor cell method to accurately deal with the double heterogeneity effect of particulate fuel. The updated CSAPI with the advanced self-shielding methods, together with the cross section libraries generated based on the improved process, were tested for the very high temperature reactor (VHTR), high temperature test reactor (HTTR), and Empire benchmark problems with various resonance self-shielding conditions, indicating that the updated CSAPI in Griffin is able to produce multigroup cross sections accurately and efficiently. We also show that the methodology works well for pebble bed fuel from HTR-10, but the capability still needs to be fully integrated into CSAPI. In the future, further benchmark tests will be performed for various thermal reactor core problems, including particulate fuel-based pebble bed reactors.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

Integration of Online Cross-Section Generation Capability with Depletion and Transient Solvers in Griffin

Griffin is a Multiphysics Object-Oriented Simulation Environment (MOOSE)-based reactor multiphysics analysis application jointly developed by Argonne and Idaho National Laboratories under the DOENE Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. In FY25, an online crosssection generation capability based on the Self-Shielding Application Programming Interface (SSAPI) was demonstrated for TRISO-fueled reactor problems under steady-state conditions. This fiscal year, that capability was extended to support depletion and transient multiphysics calculations, enabling high-fidelity analyses that generate self-shielded cross sections on the fly from the actual evolving composition and temperature states rather than from pre-tabulated libraries. For depletion, a two-way coupling was established in which SSAPI computes compact-averaged self-shielded cross sections that the depletion solver then uses to advance the Bateman equations, with the updated compositions returned to SSAPI at each step; the depletion module was refactored to support both library-based and SSAPI-based cross sections, and additional logic was added to track daughter isotopes and to exclude minor isotopes for efficiency. For transient analysis, the SSAPI multigroup library was extended with the kinetics data required for time-dependent calculations, the Improved Quasi-Static (IQS) scheme was coupled with SSAPI, and several supporting capabilities were implemented, including a self-shielding treatment that lets control rods and drums move within a self-shielded model, which had previously been impossible and had ruled out rod- and drum-movement transients with on-the-fly cross sections altogether, a new mixing scheme for delayed-neutron precursor decay constants, a checkpoint-based restart workflow, and performance improvements such as pointwise cross-section interpolation and the bypassing of unnecessary Dancoff factor calculations. The implemented capabilities were verified against Serpent Monte Carlo solutions. For depletion, a prismatic pin-cell problem based on a Next Generation Nuclear Plant (NGNP) Very High Temperature Reactor benchmark showed excellent agreement, with eigenvalue differences within 200 pcm over the entire burnup range (up to 140 MWD/kgU) and fission-product and actinide inventories agreeing to within 0.8% and 2.5%, respectively; a heat-pipe microreactor assembly problem with a much higher fuel loading confirmed the same behavior and quantified the bias introduced when the multigroup equivalence effect is neglected. For transient analysis, a pin-cell problem with a step reactivity insertion and temperature feedback reproduced the analytically expected asymptotic power and showed close agreement between the direct and IQS solutions, and a two-dimensional microreactor core problem with control-drum rotation exercised the new moving-drum self-shielding treatment and demonstrated successful coupling of the online crosssection generation with both the direct and IQS transient methods. The capability was further exercised on a full-core pebble-bed problem, in which Griffin was coupled with the System Analysis Module (SAM) to simulate load-following operation of the gPBR with the Doppler feedback resolved at the TRISO fuel kernel temperature. These developments in Griffin provide a convenient, high-fidelity approach to cross-section generation for advanced thermal reactors with geometrically complex and highly heterogeneous configurations, including TRISO-fueled prismatic and pebble-bed systems, and support steady-state, depletion, and transient multiphysics calculations. They also enable self-shielded cross sections to be evaluated directly at the actual coupled state of the system, thereby establishing a foundation for high-fidelity, fully coupled multiphysics analysis of advanced reactors

Park, H.↗
DOE OSTI2016

HYDROCARBON EMISSION RINGS IN PROTOPLANETARY DISKS INDUCED BY DUST EVOLUTION

We report observations of resolved C{sub 2}H emission rings within the gas-rich protoplanetary disks of TW Hya and DM Tau using the Atacama Large Millimeter Array. In each case the emission ring is found to arise at the edge of the observable disk of millimeter-sized grains (pebbles) traced by submillimeter-wave continuum emission. In addition, we detect a C{sub 3}H{sub 2} emission ring with an identical spatial distribution to C{sub 2}H in the TW Hya disk. This suggests that these are hydrocarbon rings (i.e., not limited to C{sub 2}H). Using a detailed thermo-chemical model we show that reproducing the emission from C{sub 2}H requires a strong UV field and C/O > 1 in the upper disk atmosphere and outer disk, beyond the edge of the pebble disk. This naturally arises in a disk where the ice-coated dust mass is spatially stratified due to the combined effects of coagulation, gravitational settling and drift. This stratification causes the disk surface and outer disk to have a greater permeability to UV photons. Furthermore the concentration of ices that transport key volatile carriers of oxygen and carbon in the midplane, along with photochemical erosion of CO, leads to an elemental C/O ratio that exceeds unity in the UV-dominated disk. Thus the motions of the grains, and not the gas, lead to a rich hydrocarbon chemistry in disk surface layers and in the outer disk midplane.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2021

The TW Hya Rosetta Stone Project. I. Radial and Vertical Distributions of DCN and DCO{sup +}

Molecular D/H ratios are frequently used to probe the chemical past of solar system volatiles. Yet it is unclear which parts of the solar nebula hosted an active deuterium fractionation chemistry. To address this question, we present 0."2–0."4 Atacama Large Millimeter/submillimeter Array (ALMA) observations of DCO{sup +} and DCN 2–1, 3–2, and 4–3 toward the nearby protoplanetary disk around TW Hya, taken as part of the TW Hya Rosetta Stone project, augmented with archival data. DCO{sup +} is characterized by an excitation temperature of ~40 K across the 70 au radius pebble disk, indicative of emission from a warm, elevated molecular layer. Tentatively, DCN is present at even higher temperatures. Both DCO{sup +} and DCN present substantial emission cavities in the inner disk, while in the outer disk the DCO{sup +} and DCN morphologies diverge: most DCN emission originates from a narrow ring peaking around 30 au, with some additional diffuse DCN emission present at larger radii, while DCO{sup +} is present in a broad structured ring that extends past the pebble disk. Based on a set of simple parametric disk abundance models, these emission patterns can be explained by a near-constant DCN abundance exterior to the cavity, and an increasing DCO{sup +} abundance with radius. In conclusion, the ALMA observations reveal an active deuterium fractionation chemistry in multiple disk regions around TW Hya, but not in the cold planetesimal-forming midplane and in the inner disk. More observations are needed to explore whether deuterium fractionation is actually absent in these latter regions, and if its absence is a common feature or something peculiar to the old TW Hya disk.

36 MATERIALS SCIENCE↗
NASA NTRS1975

Gravity shear waves atop the cirrus layer of intense convective storms

Recent visual satellite photographs of certain intense convective storms have revealed concentric wave patterns. A model for the generation and growth of these waves is proposed. The proposed initial generating mechanism is similar to the effect noticed when a pebble is dropped into a calm pond. The penetration of the tropopause by overshooting convection is analogous to the pebble's penetration of the water's surface. The model for wave growth involves instability due to the wind shear resulting from the cirrus outflow. This model is based on an equation for the waves' phase speed which is similar to the Helmholtz equation. It, however, does not assume an incompressible atmosphere, but rather assumes density is a logarithmic function of height. Finally, the model is evaluated on the two mid-latitude and three tropical cases. The data indicate that shearing instability may be a significant factor in the appearance of these waves.

Stobie, J. G.↗
NASA NTRS1989

Anomalous quartz from the Roter Kamm impact crater, Namibia - Evidence for post-impact hydrothermal activity?

Quartz pebbles from the Roter Kamm impact crater (the Namib Desert, SWA/Namibia) were examined for evidence of impact-induced hydrothermal activity, using results from microprobe analyses, neutron activation analyses, transmission IR spectroscopy, and X-ray diffractometry. It was found that the pebbles consisted of pure quartz, which contains three different types of fluid inclusions. These were identified as primary inclusions (5-10 microns) that record the formation conditions of the quartz, very small (less than 1 micron) secondary inclusions associated with the grain boundaries, and late inclusions of irregular size. It is concluded that the quartz and the primary inclusions may provide evidence for a postimpact phase of extensive hydrothermal activity, generated by the residual heat from the kinetic energy of the impact.

Koeberl, Christian↗
NASA NTRS1997

Incipient Melt Formation and Devitrification at the Wanapitei Impact Structure, Ontario, Canada

The Wanapitei impact structure is approximately 8 km in diameter and lies within Wanapitei Lake, approximately 34 km northeast of the city of Sudbury. Rocks related to the 37 Ma impact event are found only in Pleistocene glacial deposits south of the lake. Most of the target rocks are metasedimentary rocks of the Proterozoic Huronian Supergroup. An almost completely vitrified, inclusion-bearing sample investigated here represents either an impact melt or a strongly shock metamorphosed, pebbly wacke. In the second, preferred interpretation, a number of partially melted and devitrified clasts are enclosed in an equally highly shock metamorphosed arkosic wacke matrix (i.e., the sample is a shocked pebbly wacke), which records the onset of shock melting. This interpretation is based on the glass composition, mineral relicts in the glass, relict rock textures, and the similar degree of shock metamorphism and incipient melting of all sample components. Boulder matrix and clasts are largely vitrified and preserve various degrees of fluidization, vesiculation, and devitrification. Peak shock pressure of approximately 50-60 GPa and stress experienced by the sample were somewhat below those required for complete melting and development of a homogeneous melt. The rapid cooling and devitrification history of the analyzed sample is comparable to that reported recently from glasses in the suevite of the Ries impact structure in Germany and may indicate that the analyzed sample experienced an annealing temperature after deposition of somewhere between 650 C and 800 C.

Dressler, B. O.↗
NASA NTRS2013

A Mars Sample Return Sample Handling System

We present a sample handling system, a subsystem of the proposed Dragon landed Mars Sample Return (MSR) mission [1], that can return to Earth orbit a significant mass of frozen Mars samples potentially consisting of: rock cores, subsurface drilled rock and ice cuttings, pebble sized rocks, and soil scoops. The sample collection, storage, retrieval and packaging assumptions and concepts in this study are applicable for the NASA's MPPG MSR mission architecture options [2]. Our study assumes a predecessor rover mission collects samples for return to Earth to address questions on: past life, climate change, water history, age dating, understanding Mars interior evolution [3], and, human safety and in-situ resource utilization. Hence the rover will have "integrated priorities for rock sampling" [3] that cover collection of subaqueous or hydrothermal sediments, low-temperature fluidaltered rocks, unaltered igneous rocks, regolith and atmosphere samples. Samples could include: drilled rock cores, alluvial and fluvial deposits, subsurface ice and soils, clays, sulfates, salts including perchlorates, aeolian deposits, and concretions. Thus samples will have a broad range of bulk densities, and require for Earth based analysis where practical: in-situ characterization, management of degradation such as perchlorate deliquescence and volatile release, and contamination management. We propose to adopt a sample container with a set of cups each with a sample from a specific location. We considered two sample cups sizes: (1) a small cup sized for samples matching those submitted to in-situ characterization instruments, and, (2) a larger cup for 100 mm rock cores [4] and pebble sized rocks, thus providing diverse samples and optimizing the MSR sample mass payload fraction for a given payload volume. We minimize sample degradation by keeping them frozen in the MSR payload sample canister using Peltier chip cooling. The cups are sealed by interference fitted heat activated memory alloy caps [5] if the heating does not affect the sample, or by crimping caps similar to bottle capping. We prefer cap sealing surfaces be external to the cup rim to prevent sample dust inside the cups interfering with sealing, or, contamination of the sample by Teflon seal elements (if adopted). Finally the sample collection rover, or a Fetch rover, selects cups with best choice samples and loads them into a sample tray, before delivering it to the Earth Return Vehicle (ERV) in the MSR Dragon capsule as described in [1] (Fig 1). This ensures best use of the MSR payload mass allowance. A 3 meter long jointed robot arm is extended from the Dragon capsule's crew hatch, retrieves the sample tray and inserts it into the sample canister payload located on the ERV stage. The robot arm has capacity to obtain grab samples in the event of a rover failure. The sample canister has a robot arm capture casting to enable capture by crewed or robot spacecraft when it returns to Earth orbit

Wilson, David↗
NASA NTRS2016

Filling in the Gaps: Xenoliths in Meteorites are Samples of "Missing" Asteroid Lithologies

We know that the stones that fall to earth as meteorites are not representative of the full diversity of small solar system bodies, because of the peculiarities of the dynamical processes that send material into Earth-crossing paths [1] which result in severe selection biases. Thus, the bulk of the meteorites that fall are insufficient to understand the full range of early solar system processes. However, the situation is different for pebble- and smaller-sized objects that stream past the giant planets and asteroid belts into the inner solar system in a representative manner. Thus, micrometeorites and interplanetary dust particles have been exploited to permit study of objects that do not provide meteorites to earth. However, there is another population of materials that sample a larger range of small solar system bodies, but which have received little attention - pebble-sized foreign clasts in meteorites (also called xenoliths, dark inclusions, clasts, etc.). Unfortunately, most previous studies of these clasts have been misleading, in that these objects have simply been identified as pieces of CM or CI chondrites. In our work we have found this to be generally erroneous, and that CM and especially CI clasts are actually rather rare. We therefore test the hypothesis that these clasts sample the full range of small solar system bodies. We have located and obtained samples of clasts in 81 different meteorites, and have begun a thorough characterization of the bulk compositions, mineralogies, petrographies, and organic compositions of this unique sample set. In addition to the standard e-beam analyses, recent advances in technology now permit us to measure bulk O isotopic compositions, and major- though trace-element compositions of the sub-mm-sized discrete clasts. Detailed characterization of these clasts permit us to explore the full range of mineralogical and petrologic processes in the early solar system, including the nature of fluids in the Kuiper belt and the outer main asteroid belt, as revealed by the mineralogy of secondary phases.

Zolensky, Mike↗
NASA NTRS2016

Did Jupiter's Core Form in the Innermost Parts of the Sun's Protoplanetary Disc?

Jupiter’s core is generally assumed to have formed beyond the snow line. Here we consider an alternative scenario, that Jupiter’s core may have accumulated in the innermost parts of the protoplanetary disk. A growing body of research suggests that small particles (“pebbles”) continually drift inward through the disk. If a fraction of drifting pebbles is trapped at the inner edge of the disk a several Earth-mass core can quickly grow. Subsequently, the core may migrate outward beyond the snow line via planet-disk interactions. Of course, to reach the outer Solar System Jupiter’s core must traverse the terrestrial planet-forming region. We use N-body simulations including synthetic forces from an underlying gaseous disk to study how the outward migration of Jupiter’s core sculpts the terrestrial zone. If the outward migration is fast (τ(sub mig) ∼ 10(exp 4) years), the core simply migrates past resident planetesimals and planetary embryos. However, if its migration is slower (τ(sub mig) ∼ 10(exp 5) years) the core removes solids from the inner disk by shepherding objects in mean motion resonances. In many cases the disk interior to 0.5-1 AU is cleared of embryos and most planetesimals. By generating a mass deficit close to the Sun, the outward migration of Jupiter’s core may thus explain the absence of terrestrial planets closer than Mercury. Jupiter’s migrating core often stimulates the growth of another large (∼ Earth-mass) core – that may provide a seed for Saturn’s core – trapped in exterior resonance. The migrating core also may transport a fraction of terrestrial planetesimals, such as the putative parent bodies of iron meteorites, to the asteroid belt.

Protoplanetary disks↗
NASA NTRS2023

Can the Streaming Instability Form the First Planetesimals in Globally Turbulent Protoplanetary Disks?

Understanding the formation of the first planetesimals remains key to deciphering the history of planet formation within our own solar system and beyond. Evidence from the vast meteorite record [e.g., 1] as well as observations [e.g., 2] strongly suggest that the first planetesimals, and perhaps giant planet core accretion occurred well within the first million years of disk evolution. Moreover, chemical and lithological mixing as well as observations of line-broadening in protoplanetary disks (PPDs) [e.g., 3-6] suggest that the solar system nebula in this epoch was weakly-to-moderately turbulent in the regions where particle growth is of the greatest interest [7-8]. It is a well-known though that global hydrodynamic turbulence complicates particle growth due to a slew of barriers that can slow or even stall particle or aggregate growth at pebble sizes (with corresponding small particle Stokes numbers St) that can lead to loss to the central star via radial drift before planetesimals can ever form [9,10], requiring that some other mechanism come into play that collects growth-frustrated pebbles into gravitationally bound multi-km bodies – objects that are “born big” [11]. The current leading candidate for such a “leap-frog” mechanism is the so-called Streaming Instability (SI), a gas-drag mediated momentum exchange resonance in which the relative velocity between the particle component and a rotating gaseous fluid can lead to high densities in the particle field [12], which has been invoked in a number of recent PPD models that include a turbulent intensity α [e.g., 13-14] as the defacto mechanism for planetesimal formation if conditions for the SI (depending on particle St and the solids-to-gas mass ratio) are satisfied. However, these previous works use conditions established from occurrence studies for the onset of SI in laminar disks [15-17] in which the only source of turbulence is that self-generated by the settling particle layer, and not externally driven global turbulence. Recent analytical theories of the SI subject to global turbulence predict much more stringent conditions for the effectiveness of the SI than the laminar case [18-19]. Thus, whether the efficient operation of the SI can be attained in realistic models of the solar nebula have yet to be established. In this work, we ask whether the conditions under which the SI can produce gravitationally bound particle overdensities can actually be met in the first million years of evolution in globally turbulent PPDs.

P R Estrada↗
DOE OSTI2022

U.S. High Temperature Materials Highlights

U.S. GIF VHTR work is continuing on graphite qualification, Alloy 617 regulatory issues beyond the Code space, Alloy 800H weldments, and ASME Codes and Standards R&D is still considering both pebble bed and prismatic and steam generator and heat exchanger U.S. DOE Advanced Reactor Demonstration Program (ARDP) Two U.S.-based teams were selected to demonstrate advanced nuclear reactors in the United States that can be operational by 2027 One of the teams is X-energy (Rockville, MD) which will demonstrate a modular gas-cooled reactor design (Xe-100) with four 80 MWe, TRISO fuel, pebble bed reactors A number of U.S.-based teams were selected to design and develop safe and affordable reactor technologies that can be licensed and deployed over the next 10 to 14 years (Risk Reduction) One of the teams is BWXT Advanced Technologies, LLC which will develop a commercially viable transportable microreactor with the design focused on using TRISO fuel particles and silicon carbide (SiC) matrix A number of U.S.-based teams were selected to assist the progression of advanced reactor designs in their earliest phases (Advanced Reactor Concepts-20) One of the teams is Massachusetts Institute of Technology which will mature the Modular Integrated Gas-Cooled High Temperature Reactor (MIGHTR) concept with a horizontal compact design from a pre-conceptual stage to a conceptual stage to support commercialization

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Study on Water-ingress Accident during the Initial Core and the Running-in Phase of HTR-PM - 18429

Water ingress accident is one of the typical accidents with potential hazard to the high temperature gas-cooled reactor (HTGR), which will cause reactivity introduction, as well as the chemical corrosion of the graphite fuel elements and reflector structure material. Besides, increase of the primary pressure may result in the opening of the safety valves, consequently leading the release of radioactive isotopes. Therefore high attention has been paid to water ingress phenomenon during the design of the HTGR to verify its inherent safety characteristics. The online continuous discharge and multi-pass of the fuel elements is another important characteristics for the pebble-bed HTGR which can effectively increase the reactor operation availability. From the initial core, a long period of time, namely the running-in phase, is needed to reach the final long-term stable equilibrium core. During this period of time, the core states are totally different, including the number of the fuel elements, the fuel enrichment, the accumulation of the fission product, the temperature feedback coefficient of the reactivity, and so on. In this paper, based on the design of 250 MW Pebble-bed Module High Temperature gas-cooled Reactor (HTR-PM) and a candidate proposal for its initial core and running-in phase, considering three typical states during the running-in phase, the design basis accident of a double-ended guillotine break of one heating tube has been analyzed with the TINTE code, and the results are compared with that of the equilibrium core. Some important parameters during the transient process, including the power, the temperature of the fuel element, the pressure of the primary coolant, as well as the graphite corrosion rate, are studied. To a certain extent, the results can verity the design of the initial core and the running-in phase, and also further prove the safety feature of the HTR. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗