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At least 415 records · Page 23
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Probabilistic Analysis of Long-Term Degradation of Microwave Cavity Flow Sensor

We are investigating a microwave resonant cavity transducer for flow sensing in the vessel of a high temperature fluid advanced reactor (AR), such as a molten salt cooled reactor (MSCR) or a sodium fast reactor (SFR). This transducer is a hollow metallic cylindrical cavity, with the flat wall of the cylinder flexible enough to undergo microscopic deflection due to dynamic fluid pressure. Membrane deflection leads to a shift in the resonant frequency, which can be detected with a spectrum analyzer. We have performed a proof-of-concept experiment of flow sensing with the transducer in liquid sodium at 340°C in impinging liquid jet geometry. The transducer remained in liquid sodium for 70 days. After removal, no structural damage was observed, and the expected transducer response was verified in a water test. Because long-term (multi-year) experimental tests of transducer resilience to harsh environment are not practical, we have developed a probabilistic model of creep to estimate transducer resilience to the harsh environment. The probabilistic model considers diffusion creep under the condition of high temperature and low stress, where the stress and temperature are allowed to be random variables with Gaussian distributions. Using the probabilistic model, we estimate inelastic membrane deflections due to creep for several temperature ranges. We conclude that for temperatures less than 650°C, creep has negligible long-term effect on the transducer performance. Since a yellowish residue was observed on the transducer surface after 70 days of immersion in liquid sodium, we have investigated possible evidence of corrosion. Chromium depletion is a typical indicator of the corrosion process in stainless steel. Scraping off a residue from the transducer and performing scanning electron microscopy (SEM) with energy dispersive analysis (EDS) did not find any chromium in the residue. Approximately 60% of the residue consisted of copper, which can be attributed to contamination of sodium due to powder residue from machining of copper and brass components of the transducer.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

MOOSE Framework Meshing Enhancements to Support Reactor Analysis

MOOSE-based physics codes require an input finite element mesh on which the physics solution is calculated, reported, and transferred to other physics codes. The use of difficult-touse, external licensed software is often required to generate high quality meshes for reactor geometries. High-fidelity geometry modeling also requires elaborate tracking of groups of elements for material property assignment and output reporting which can be considerably complex for the user to identify and maintain. Under the U.S. Department of Energy Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, several meshingrelated enhancements have been developed for the MOOSE framework to address user challenges in creating finite element meshes for advanced reactor geometries. MOOSE mesh generators have been developed to mesh hexagonal geometries (pins, ducted assemblies, and cores) commonly found in liquid-metal cooled fast reactor concepts. The mesh generator used for hexagonal pin cells is generic for regular polygons and therefore may also be used for Cartesian pin cells. Hexagonal pin cells can be stitched into ducted assemblies, and assemblies can be stitched together into a core. The user may specify region ids, region names, and other preferences on the mesh. This control is useful for later material mapping in the MOOSE-based physics codes input. A capability was also developed for meshing rotating control drums including determination of material volume fractions in each mesh element as a function of time. Control drum meshes may be stitched to other hexagonal assemblies to create a core configuration. Additional mesh generators were developed that wrap around the hexagonal meshing capabilities and utilize “extra element integer” ID values on each element. In regular Cartesian or hexagonal assemblies or cores, the bookkeeping of element groups for both material assignment and output reporting can now be automated through assignment of pin, assembly, core, axial and depletion id values stored as extra element integers. The extra element tags on the mesh greatly speed the reactor analyst’s efforts to map materials to meshes, track depletion zones, and parse output such as axial pin power distributions. At the highest level, pin, assembly, and core mesh generators (with this reactor terminology) have also been developed to easily generate regular Cartesian and hexagonal cores, including axial extrusion. These reactor geometry builders call upon the previously mentioned capabilities to produce analysis-ready 3D meshes including material assignments. Open source mesh triangulation capabilities were also investigated for integration into the MOOSE framework to address the need for meshing the core periphery region which extends from the irregular outer assembly border to a cylindrical boundary. Options are limited due to licensing constraints, and the recommendation is pursue building a native MOOSE Delaunay triangulator routine with full functionality. Finally, a series of verification problems were performed with NEAMS physics tools. All developed capabilities will be available in the new open-source “Reactor” module of the MOOSE framework, which is accessible to any MOOSE-based NEAMS physics tool.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1991

The influence of dynamics on two-dimensional model results - Simulations of C-14 and stratospheric aircraft NO(x) injections

The influence of dynamics on model results is investigated via a 2D photochemical model. Three representations of atmospheric transport are tested in order to simulate total ozone and C-14 amounts after nuclear tests in the early 1960s. Three scenarios of NO(x) injections from a proposed fleet of stratospheric aircraft and their effects on ozone are simulated. The three dynamical formulations used are Dynamics A, a base dynamics used in previous work with this model; Dynamics B, a strong circulation dynamics discussed by Jackman et al. (1989); and Dynamics C, the dynamics used by Shia et al. (1989). The advective component of the stratosphere to the troposphere mass exchange rate is largest for Dynamics B (5.8 x 10 exp 17 kg/yr) and smallest for Dynamics C (1.4 x 10 exp 17 kg/yr), with the advective strat/trop exchange rate for Dynamics A (2.4 x 10 exp 17 kg/yr) being these two extremes. Ozone depletion from NO(x) injections of stratospheric aircraft showed a strong sensitivity to dynamics.

Jackman, Charles H.↗
DOE OSTI2020

Overview of CMOS Sensors for Future Tracking Detectors

Depleted Complementary Metal-Oxide-Semiconductor (CMOS) sensors are emerging as one of the main candidate technologies for future tracking detectors in high luminosity colliders. Their capability of integrating the sensing diode into the CMOS wafer hosting the front-end electronics allows for reduced noise and higher signal sensitivity, due to the direct collection of the sensor signal by the readout electronics. They are suitable for high radiation environments due to the possibility of applying high depletion voltage and the availability of relatively high resistivity substrates. The use of a CMOS commercial fabrication process leads to their cost reduction and allows faster construction of large area detectors. In this contribution, a general perspective of the state of the art of CMOS detectors for High Energy Physics experiments is given. The main developments carried out with regard to these devices in the framework of the CERN RD50 collaboration are summarized.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2020

Dynamic resource allocation drives growth under nitrogen starvation in eukaryotes

Cells can sense changes in their extracellular environment and subsequently adapt their biomass composition. Nutrient abundance defines the capability of the cell to produce biomass components. Under nutrient-limited conditions, resource allocation dramatically shifts to carbon-rich molecules. Here, we used dynamic biomass composition data to predict changes in growth and reaction flux distributions using the available genome-scale metabolic models of five eukaryotic organisms (three heterotrophs and two phototrophs). We identified temporal profiles of metabolic fluxes that indicate long-term trends in pathway and organelle function in response to nitrogen depletion. Surprisingly, our calculations of model sensitivity and biosynthetic cost showed that free energy of biomass metabolites is the main driver of biosynthetic cost and not molecular weight, thus explaining the high costs of arginine and histidine. We demonstrated how metabolic models can accurately predict the complexity of interwoven mechanisms in response to stress over the course of growth.

59 BASIC BIOLOGICAL SCIENCES↗
DOE OSTI2022

Microstructural evolution and precipitation in γ-LiAlO 2 during ion irradiation

Polycrystalline γ-LiAlO 2 pellets were sequentially irradiated with 120 keV He + and 80 keV D 2 + ions to the fluences of 1 × 10 17 and 2 × 10 17 (He + + D + )/cm 2 at 573 K. Additional irradiation was performed to a fluence of 2 × 10 17 (He + + D + )/cm 2 at 773 K. The irradiated pellets were characterized using scanning transmission electron microscopy, time-of-flight secondary ion mass spectrometry, and grazing incidence x-ray diffraction. Lattice damage, amorphization, and fractures are observed with no evidence for the formation of secondary-phase precipitates in the pellets irradiated up to an ion fluence of 2 × 10 17 (He + + D + )/cm 2 at 573 K. In contrast, faceted precipitates with sizes larger than 100 nm formed in a pellet irradiated to 2 × 10 17 (He + + D + )/cm 2 at 773 K. Analyses of the diffraction and composition data suggest that the precipitates have a spinel-type structure, likely a non-stoichiometric LiAl 5 O 8 with Li depletion. This could be an intermediate phase with Li atoms at the octahedral and possibly tetrahedral sites as well. It is speculated that as the dose increases, Li loss will continue and the precipitates will approach a composition of alumina primarily in phases of α-Al 2 O 3 and amorphized Al 2 O 3 .

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2020

Pump depletion in Raman lasing in a graded index multimode fibre

We examine multimode diode laser pump depletion in a graded index multimode (MM) fibre Raman laser generating a high-quality Stokes beam. Spatial ‘hole burning’ in the transmitted pump beam is demonstrated. A balance model is constructed for interaction of the pump and Stokes waves in the MM Raman laser. Analytical formulas are derived and used in calculations demonstrating qualitative agreement with experimental data. The origin of the observed quantitative discrepancies is discussed. (paper)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

Characterization of the Aerosol Source Term in Dry Storage Canisters

As long-term dry storage of used nuclear fuel at independent spent fuel storage installations (ISFSI) trends toward the de facto back end of the US fuel cycle, it becomes appropriate to investigate potential degradation and dispersion scenarios for suitable risk mitigation purposes. Pitting and subsequent stress corrosion cracking of the canister wall is currently viewed as a potential scenario leading to a through-wall pathway for contamination to be transferred from within the storage container to the surrounding environment. While stress corrosion cracking measurements are currently underway to further characterize this scenario, a parallel effort endeavors to perform a consequence analysis of conditions in which through wall cracks are indeed formed. This effort consists of engineering scale modeling using the GOTHIC and MELCOR software packages along with experimental depletion and penetration tests.

spent fuel storage, aerosol deposition, Stress Cor↗
DOE OSTI2021

Fuel-cladding chemical interaction of a prototype annular U-10Zr fuel with Fe-12Cr ferritic/martensitic HT-9 cladding

As an alternative fuel form, the annular metallic fuel design eliminates the liquid sodium bond between the fuel and the cladding, providing back-end fuel cycle and other benefits. The fuel-cladding chemical interaction (FCCI) of annular fuel also presents new features. In this work, state-of-the-art electron microscopy and spectroscopy techniques were used to study the FCCI of a prototype annular U-10wt%Zr (U-10Zr) fuel with ferritic/martensitic HT-9 cladding irradiated to 3.3% fission per initial heavy atom. Compared with sodium-bonded solid fuels, negligible amounts of lanthanides were found in the FCCI layer in the investigated helium-bonded annular fuel. Instead, most lanthanides were retained in the newly formed UZr 2 phase in the fuel center region. The interdiffusion of iron and uranium resulted in tetragonal ( U ,Zr) 6 Fe phase (space group I4/mcm) and cubic ( U ,Zr)( Fe ,Cr) 2 phase (space group Fd 3 ¯ m). The ( U ,Zr)( Fe ,Cr) 2 phase contains a high density of voids and intergranular uranium monocarbides of NaCl-type crystal structure (space group Fm 3 ¯ m). At the interdiffusion zone and inner cladding interface, a porous lamellar structure composed of alternating Cr-rich layers and U-rich layers was observed. Next to the lamellar region, the unexpected phase transformation from body-centered cubic ferrite (α-Fe) to tetragonal binary Fe-Cr σ phase (space group P4 2 /mnm) occurred, and tetragonal Fe-Cr-U-Si phase (space group I4/mmm) was identified. Due to the diffusion of carbon into the interdiffusion zone, carbon depletion inside the HT-9 led to the disappearance of the martensite lath structure, and intergranular U-rich carbides formed as a result of the diffusion of uranium into the cladding. These detailed new findings reveal the unique features of the FCCI behavior of annular U-Zr fuels, which could be a promising alternative fuel form for high burnup fast reactor applications.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Radiochemical transport analysis of gamma spectroscopic data to support estimation of molten salt reactor off-gas inventories

This work introduces a novel application of radiochronometry to estimate nuclide inventories in molten salt reactor off-gas systems based on gamma spectroscopic data from the Molten Salt Reactor Experiment. By analyzing isotopic, isobaric, and isomeric activity ratios, key depletion model parameters related to species transport within the reactor system could be inferred. The findings demonstrate the potential of leveraging a limited subset of gamma spectroscopy measurements to accurately estimate nuclide inventories throughout the off-gas system. The approach can be useful in reactor design activities and support analyses relevant to operations, safety, security, and safeguards.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Demonstration of In Situ Down Blending During ZIRCEX Process

The Material Recovery Pilot Plant has successfully demonstrated in situ down-blending of HEU fuel to support INL’s commitment to HALEU production technologies. Depleted uranium pellets, along with an HEU fuel piece were placed in the hydrochlorinator and subsequently declad and oxidized. UOx product was recovered, analyzed, and demonstrated to have reduced enrichment within the targeted range of 20 – 30 % U-235. This first attempt at in situ down-blending will be followed by continued process refinement over the remaining fuel operations at MRPP.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Experimental Results for Large Particle Separation from Non-Newtonian Slurries Using Tapered Bump Arrays

Lateral displacement arrays are useful for separating particles such as blood cells and sand from carrier fluids. These arrays consist of staggered posts, which allow smaller particles to follow streamlines and larger particles to flow around the posts and migrate to one side. This migration increases the particle concentration in one direction and depletes the particle concentration in the other direction allowing particle separation to occur. Experiments were conducted to separate large particles in non-Newtonian yield stress slurries using tapered bump arrays. The non-Newtonian slurry used was composed of a bentonite kaolin blend with the inclusion of larger diameter particles. These experiments were conducted to evaluate the performance of particle separation using a tapered array of staggered posts as configured in a bump array for non-Newtonian yield stress slurries, an application that has not been explored experimentally. A theory for large particle separation was developed for slurries that exhibit Bingham or Cross rheology. The results of these experiments are described. This information could be applied in industrial settings such as separation of particles from nuclear waste slurries including those to be processed at the Hanford site, where removing large particles from waste streams is important to processing.

slurry, solids removal, deterministic lateral disp↗
DOE OSTI2020

Studies on small charge packet transport in high-resistivity fully-depleted CCDs

In this work, we will present a physical model and measurements of the transport of small charge packets in the bulk of thick high resistivity CCD before being collected by the pixel potential wells. A new technique to measure the lateral spread of the charge as a function of the ionization depth in the bulk is presented. Results from measurements on CCD currently in use for several scientific instruments are shown and validated with a new mathematical algorithm to extend the current modeling based only on the diffusion of the charge in silicon.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2025

Laser Powder Directed Energy Deposition of Steels for Nuclear Applications

This comprehensive investigation examines the structure–property relationships in two nuclear alloy systems—Alloy 709 (A709) austenitic stainless steel and Grade 92 (G-92) ferritic/martensitic (F/M) steel—manufactured via directed energy deposition (DED) for sodium-cooled fast reactor applications. This study establishes the fundamental mechanisms for controlling microstructures for optimizing the mechanical performance of additively manufactured nuclear materials through systematic heat treatment optimization and multiscale characterization. As-deposited A709 steel develops a complex multiscale strengthening architecture consisting of a fine cellular solidification structure with diameter of 2-3 µm within10–50 µm grains, elevated dislocation densities from rapid thermal cycling, and grain boundary precipitates that activate concurrent Hall–Petch, dislocation, and precipitation hardening mechanisms to achieve exceptional properties [yield strength (YS): 603 MPa, ultimate tensile strength (UTS): 844 MPa, Vickers hardness: 220 HV] that achieve a 44% superior strength compared to that of the wrought material. Heat treatments produce different results. Solution annealing (SA) dissolves the cellular structure and reduces the hardness to 190 HV. Precipitation treatment (PT) keeps the cellular structure but adds carbides, allowing the hardness to reach 205 HV. The best approach combines both treatments (SA+PT) and creates uniform precipitate distributions with M 23 C 6 carbides at the grain boundaries and MX carbonitrides in the matrix, achieving a hardness of 195 HV. However, directional differences persist, with a 12%–15% strength variation between orientations due to the inherited layered microstructural architecture that survives aggressive heat treatment. While tensile testing at 550°C demonstrates 40%–50% thermal softening with dynamic strain aging, DED A709 steel still maintains a 71% higher YS than that of the wrought material. Ion irradiation studies (100–400 dpa) of DED A709 steel reveal progressive radiation damage with increasing void density and radiation-induced segregation causing nickel enrichment and chromium depletion, which will ultimately compromise mechanical properties. As-deposited G-92 exhibits exceptional strength (UTS: 1650–1700 MPa, 430 HV) through a complex microstructure containing both ferrite and martensite phases, a high geometrically necessary dislocation (GND) density (17.04×10 14 /m 2 ), and fine carbides. Heat treatments create distinct changes. Normalizing produces fresh martensite with the highest hardness (460 HV) and an increased GND density (20.23×10 14 /m 2 ). Tempering develops dual precipitation systems and reduces the hardness to 290 HV. The optimal approach uses sequential normalizing plus tempering, achieving balanced properties with the lowest hardness (250 HV) and a reduced GND density (11.01×10 14 /m 2 ). A processing-dependent anisotropy is observed: horizontal specimens achieve superior ductile behavior, while vertical specimens exhibit brittle failure. A tempering heat treatment successfully mitigates this anisotropic behavior by transforming the hard martensitic as-deposited structure into tempered martensite enabling both horizontal and vertical specimens to exhibit similar stress–strain characteristics with visible necking behavior. Remarkably, testing at 550°C reveals a reversal in the anisotropy, where as-deposited specimens achieve near isotropy with superior thermal stability (a 15%–20% strength reduction), while tempered specimens develop an orientation dependence with a 25%–30% strength reduction. Both alloy systems demonstrate that DED processing creates specimens with a superior strength through refined microstructural features, though with distinct strengthening mechanisms—austenitic through cellular structures and precipitates versus F/M through phase transformations and precipitates. Heat treatment optimization requires alloy-specific approaches, with A709 benefiting from controlled precipitation while G-92 requires careful phase transformation control. The results show that DED manufacturing can produce nuclear materials with exceptional performance, but directional effects and temperature-dependent behavior must be carefully considered for reactor component design and qualification.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

A comparison of carbon impurities in pre- and post-melt uranium Part 1: Scanning electron microscopy analysis

Although C impurities in U have been studied for decades, fundamental questions regarding their incorporation, migration, and overall transformation during metal processing still exist. In two written Parts, we compare the chemical speciation, distribution, crystallography, and morphologies (size, shape, etc.) of carbon-containing impurities in U metal both before and after melting using high resolution analytical electron microscopy (AEM). This first Part demonstrates the variability of carbide inclusions in depleted U (DU) metal with respect to their chemical phases and morphologies as observed by scanning electron microscopy (SEM). Here, a variety of inclusion types, such as the pill-shaped U monocarbide (UC), previously only hypothesized, are summarized. Additionally, delineation between hopper-shaped U carbonitride U(C,N) and dendritic UC inclusions is discussed in relationship to archived literature.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Performing k eff Validation of As-Loaded Criticality Safety Calculations Using UNF-ST&DARDS: Applicable Experiment Selection [Slides]

This presentation discusses the UNF-ST&DARDS which performs many analyses for as-loaded SNF canisters, including criticality safety, shielding, thermal-hydraulic, and containment. It also discusses the experiment selection based on c k assessment of similarity and that the c k value of 0.8 or greater considered applicable for validation. 11 PWR SNF canisters (MPC-32) are used in this work. 1 model represents a failed fuel assembly as fresh, per the design basis and the remaining 10 models represent all 32 assemblies with depleted fuel. In conclusion, critical experiment selection can be performed with S/U techniques for as-loaded canisters in UNF-ST&DARDS. Sufficient benchmark experiments exist to support validation and additional MOX experiments that are a good match for commercial SNF would be a benefit to provide independent data. The presentations states that S/U techniques identify different pools of experiments for different systems and that the process is amenable to automation within UNF-ST&DARDS.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

An in-line burn-up algorithm: proposal and preliminary analysis

A new burn-up algorithm is proposed for coupling neutronics and isotopic depletion. The algorithm is analogous to those proposed for enforcing xenon equilibrium and converging neutronics and thermal-hydraulics. During the corrector step of standard predictor-corrector methods, the corrector nuclide density is updated as neutronics converges. This is expected to improve stability while reducing computational expense compared to other implicit algorithms. The stability of the algorithm is investigated by Fourier analysis for a simple case whose results have been successfully compared with numerical experiments.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

SCALE 6.2 Lattice Physics Performance Assessment

The US Nuclear Regulatory Commission relies on the lattice physics analysis capabilities of the SCALE code system to perform confirmatory licensing analyses. Either SCALE lattice physics code—TRITON/NEWT or Polaris—can be used to generate cross section data used by the PARCS nodal core simulator for full-core neutronics calculations. This report presents an assessment of the accuracy of SCALE lattice physics codes for preparation of lattice physics data that are used to support simulator codes such as the NRC’s PARCS, for UO 2 -mixed oxide (MOX)/Zr fueled light water reactor (LWR) analyses. Due to the nature of lattice physics calculations, critical reactor experiment benchmarks cannot be modeled in explicit detail in a lattice physics code. However, this limitation does not mean that these measurement data are not usable for lattice physics studies. Therefore, either geometry approximations or axial buckling must be implemented to determine the critical water height. These modeling limitations have led to development of a three-phase assessment strategy. In the first phase, selected critical experiment benchmarks are modeled using the SCALE 3D continuous-energy (CE) Monte Carlo (MC) code KENO, which is the most rigorous neutron transport method available in SCALE, with no approximations in the spatial, angle, or energy treatments. Biases and statistical uncertainties in quantities of interest such as k eff and pin power distributions are determined by comparing CE KENO results to experimental data. This first phase in the accuracy assessment is to demonstrate and establish the use of CE KENO as a reference solution for the second phase. In the second phase, 14 numerical test suites are used to compare SCALE lattice physics calculations with CE KENO as a reference solution. In the third phase, SCALE lattice physics depletion calculations are performed, and the spent fuel isotopic results are compared with available radiochemical assay measurements. Isotopic measurement comparisons provide quantitative assessment of isotopic density distribution predictions with the depletion models in SCALE lattice physics codes. This report documents results for all test suites. The assessment was performed using standard production techniques unless otherwise noted. Both TRITON/NEWT and Polaris exhibited acceptable accuracy for most test cases. For the few test cases in which acceptable accuracy criteria were not met, further code and data development are planned. The computer codes used in this assessment are as follows: SCALE 6.2 rev19189 (pre-release of 6.2.1) was used for CE KENO, TRITON/NEWT and Polaris (PWR only) calculations; SCALE 6.2.2 was used for Polaris calculations to address several updates in support of support boiling water reactor (BWR) lattice geometries. For this work, there is no difference between SCALE 6.2.1 and SCALE 6.2.2 beyond the inclusion of the Polaris BWR analysis capability; The ENDF/B-VII.1 continuous energy and 252 group libraries deployed in SCALE 6.2 were generated with AMPX 6.2 that is distributed with SCALE.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗