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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 667 records · Page 37
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Origin and effect of film sub-stoichiometry on ultraviolet, ns-laser damage resistance of hafnia single layers

Understanding the origin of laser damage-prone precursors in high index materials such as hafnia holds the key to the development of laser damage-resistant multilayer dielectric coated optics for high power and energy laser systems. In this study, we investigate the source of sub-stoichiometry, a potent laser damage precursor, in hafnia films produced by an ion beam sputtering (IBS) deposition method and the effect of such defects on the film performance upon ns ultraviolet (UV) laser (8 ns, 355 nm) exposure. Chemical analysis of data obtained via Rutherford backscattering spectroscopy (RBS) suggests that hafnia films deposited at two different planetary locations from the same deposition run exhibit anisotropic and location-dependent stoichiometries. While the oxygen-to-hafnium ratio is at the stoichiometric value of 2 for the hafnia film at the edge location, the ratio is significantly deviated and is 1.7 for that deposited at the planetary center. The sub-stoichiometric hafnia films display a much lower 1-on-1 damage onset at 1.6 ± 0.2 J/cm 2 compared to 2.3 ± 0.2 J/cm 2 in a stoichiometric film. The low damage performance films also have an over three times higher damage density at fluences above initiation. Coupled with Monte Carlo simulations, we reveal that sub-stoichiometry is primarily attributed to preferential removal of oxygen during film deposition by the bombardment of energetic reflected argon neutrals. The resulting oxygen deficiencies create the sub-bandgap states which facilitate the strong laser energy coupling and reduce the resistance to laser-induced damage in the hafnia single layer films.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Radiative pumping of exciton-polaritons in 2D hybrid perovskites

In addition to their attractive technological applications in photovoltaics and light emitters, the perovskite family of semiconductors has recently emerged as an excellent excitonic material for fundamental studies. Specifically, the 2D hybrid organic-inorganic perovskite (HOIP) offers the added advantage of room temperature investigations owing to their large exciton binding energy. In this work, we strongly couple excitons in 2D HOIP crystals to planar microcavity photons sustaining exciton-polaritons under ambient conditions resulting in a Rabi splitting of 290 meV. Dark excitons directly pump the polariton branch along its dispersion in resonance with the Stokes shifted emission state (radiative pumping), creating a high density of polaritons at higher in-plane momentum ( k || ). We further probe the nonlinear polariton dispersion dynamics at varying input laser fluence, which indicates efficient polariton-polariton scattering and decay to k || = 0 from higher k || . The observation of Stokes shift-assisted energy exchange of dark states with lower polaritons coupled with evidence of efficient polariton-polariton scattering makes 2D HOIPs an attractive platform to study exciton-polariton many-body physics and Bose-Einstein like condensation (BEC) at room temperature.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Impact of high-temperature annealing on hafnia-silica composite coatings deposited via ion beam sputtering for high-peak power 1064 nm lasers

The maximum power handling fluence of high-peak and average power laser systems is often limited by the laser damage of the coatings on optical components. Furthermore, these multilayer dielectric coatings are limited in their maximum power handling due to laser-damage-prone defects in the lower optical bandgap, higher optical index material. Some of these defects can be mitigated by thermal annealing to high temperatures, which can greatly reduce the linear absorbative precursors. Typically, hafnia and silica are the materials of choice for high-peak and average power laser systems in the ultraviolet through infrared spectral range; however, hafnia crystalizes readily when annealed at high temperatures. In this study, we prepare composite HfO 2 -SiO 2 coatings by co-sputtering hafnia and silica in an ion beam sputtering system and compare them to pure hafnia-based coatings. We demonstrate that crystallinity in hafnia can be completely suppressed when it is mixed with silica, such as the composite coatings in this study. High reflectors were fabricated and annealed, demonstrating that the multilayer dielectric stacks can survive high-temperature annealing and exhibit an excellent linear absorption of 0.2 +/- 1 ppm at 1064 nm. Short- and long-pulse laser damage was explored, demonstrating the complex relationship between linear absorption and the non-linear absorption which drives pulsed laser damage. These results provide an excellent route to the creation of very low linear absorption optical coatings, which also utilize low scattering materials that are best suited for high-peak and average power applications.

Harthcock, Colin [Lawrence Livermore National Labo↗
DOE OSTI2021

Ptychographic wavefront characterization for single-particle imaging at x-ray lasers

A well-characterized wavefront is important for many x-ray free-electron laser (XFEL) experiments, especially for single-particle imaging (SPI), where individual biomolecules randomly sample a nanometer region of highly focused femtosecond pulses. We demonstrate high-resolution multiple-plane wavefront imaging of an ensemble of XFEL pulses, focused by Kirkpatrick–Baez mirrors, based on mixed-state ptychography, an approach letting us infer and reduce experimental sources of instability. From the recovered wavefront profiles, we show that while local photon fluence correction is crucial and possible for SPI, a small diversity of phase tilts likely has no impact. Our detailed characterization will aid interpretation of data from past and future SPI experiments and provides a basis for further improvements to experimental design and reconstruction algorithms.

47 OTHER INSTRUMENTATION↗
DOE OSTI2024

Anomalous Purcell decay of strongly driven inhomogeneous emitters coupled to a cavity

We perform resonant fluorescence lifetime measurements on a nanocavity-coupled erbium ensemble as a function of cavity-laser detuning and pump power. Our measurements reveal an anomalous three-fold suppression of the ensemble Purcell factor at zero cavity detuning and high pump fluence. We capture qualitative aspects of this decay rate suppression using a Tavis–Cummings model of non-interacting spins coupled to a common cavity.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Zinc oxide (ZnO) hybrid metasurfaces exhibiting broadly tunable topological properties

Extreme light confinement observed in periodic photonic structures, such as the vortex singularities in momentum (k) space, has been associated with their topological nature. Consequently, by exploiting and tuning their topological properties, optical metasurfaces have been demonstrated as an attractive platform for active photonics. However, given the fact that most active media under external excitations can only provide limited refractive index change, the potential advancements offered by the topological character of active metasurfaces have remained mostly unexplored. Zinc oxide (ZnO), which has recently exhibited optically-induced extraordinarily large permittivity modulations at visible and near-infrared frequencies, is an excellent active material for dynamic metasurfaces exhibiting strong tuning. This work demonstrates that a hybrid metasurface consisting of an array of ZnO nanodisks on a silver backplane displays broadly tunable topological properties. In particular, by performing k-space scattering simulations using measured pump-fluence-dependent material properties of ZnO, we study in detail the light reflection from the hybrid metasurface. Our results validate that the large k-space topology tuning of the metasurface can result in enormously strong polarization manipulation of near-infrared light in the vicinity of the topological features. The observed polarization switching effect is highly sensitive to the polarization and wavelength of an incident wave, owing to the symmetry and dispersion characteristics of the proposed system. Our study indicates that leveraging a combination of the extraordinary material properties and the k-space topology, hybrid metasurfaces based on ZnO may open new avenues for creating all-optical switchable metadevices.

36 MATERIALS SCIENCE↗
DOE OSTI2022

4D evolution of Cr 23 C 6 precipitates in neutron-irradiated and annealed HT-UPS steel observed via synchrotron micro-computed tomography

High-temperature-ultrafine precipitate strengthened (HT-UPS) steel is a potential structural material for advanced nuclear reactors; however, its irradiation response is not well understood. This research provides insight into irradiation-induced effects, such as precipitate evolution mechanisms and four-dimensional morphological evolution, in HT-UPS steel using synchrotron micro-computed tomography. Identical specimens were characterized pre-irradiation and post-irradiation following neutron exposure up to 0.3 displacements per atom at 600 °C. Irradiation effects were also differentiated from the annealing response of precipitates. Following neutron irradiation, the average Cr 23 C 6 precipitate size reduced, affected by the synergy of nucleation and growth, ballistic dissolution, and inverse coarsening, which was observed at fluences an order of magnitude lower than previously observed. Annealing at 600 °C for 32 h increased the average Cr 23 C 6 precipitate size and decreased the phase fraction, attributed to precipitate coarsening. The precipitate morphology evolution and resultant mechanisms can be utilized to parameterize and validate microstructural models simulating radiation damage or annealing.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Irradiation-induced damage in concrete-forming aggregates: revisiting literature data through micromechanics

The radiation-induced volumetric expansion (RIVE) of aggregate-forming minerals causes damage in concrete exposed to high levels of fast neutrons fluence ( \(>\,\sim \,10^{19}\, \hbox {n}\,\hbox {cm}^{-2}\) at kinetics energy above 0.1 MeV). Historical post-irradiation RIVE and Young’s modulus data obtained in test reactors were revisited using a polycrystalline homogenization model (self-consistent scheme) accounting for the aggregates’ minerals content and the formation of voids/cracks during irradiation. It was found that the formation of extra voids/cracks can contribute to the aggregate expansion more substantially than the cumulated expansions of aggregate-forming minerals. Furthermore, the rate of void creation appears to be inversely correlated to the silica content, although aggregates of higher silicate contents exhibit higher RIVEs. The loss of Young’s modulus decreases exponentially with the aggregates’ expansion. At a comparable RIVE level, the relative loss of modulus is more significant in aggregates of higher silica content.

36 MATERIALS SCIENCE↗
DOE OSTI2022

FFT-based model for irradiated aggregate microstructures in concrete

The concrete biological shield of light water reactors is exposed to neutron and gamma irradiation throughout its lifetime, which results in the long-term degradation of the concrete’s mechanical properties. Under neutron irradiation, the concrete’s aggregates are subjected to radiation-induced volumetric expansion (RIVE), which strongly depends on the mineral content of the aggregate and exhibits the largest expansion in silicate-bearing minerals. In this work, the authors used the fast Fourier transform-based code Microstructure-Oriented Scientific Analysis of Irradiated Concrete (MOSAIC) in 2D to model the expansion of five different aggregates provided by the Japan Concrete Aging Management Program (JCAMP). Comparable rock specimens were irradiated at the JEEP-II test reactor. The model uses realistic aggregate microstructure reconstruction based on high-resolution characterization images. The model accounts for anisotropic RIVE, thermal expansion, and the associated initiation and propagation of damage. The RIVE models are calibrated based on expansion data in the literature. The authors assume that damage occurs exclusively at interfaces between the particles that compose an aggregate and that these interfaces also exhibit swelling. Using a micromechanical model, the evolution of Young’s modulus with RIVE is calculated for each aggregate and compared with Russian irradiation data. The modeled linear expansion agrees well with the experimentally measured expansion. Furthermore, the model also predicts that anisotropic RIVE and thermal expansion result in an earlier onset of damage with neutron fluence than in the isotropic case.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Dose Coefficient Calculation for Use in Dosimetry Assessment of a Fission-Based Weapon

In the event of a fission-based weapon or improvised nuclear device (IND) detonation, dose coefficients can be harnessed to provide dose assessments for defense, emergency preparedness, and consequence management, as well as to prospectively inform the assessment of radiation biomarkers and development of medical prophylaxis countermeasures for defense and homeland security stakeholders and decision-makers. Although dose coefficients have previously been calculated for this group, they would apply specifically to the studied population, the 1945 Japanese cohort, after which their anthropomorphic computational phantoms were modeled. For this reason, applications to other populations may be limited, and instead, an assessment of a more standardized population is desired. We employed a series of computational human phantoms representing international reference individuals: UF/NCI voxel phantom series containing newborn, 1-, 5-, 10-, 15-, and 35-year-old males and females. Irradiation of the phantoms was simulated using the Monte Carlo N-Particle transport code to determine organ dose coefficients under four idealized irradiation geometries at three distances from the detonation hypocenter at Hiroshima and Nagasaki using DS02 free-in-air prompt neutron and photon fluence spectra. Through these simulations, age-specific dose coefficients were determined for individual organs. Various articulated PIMAL stylized phantoms were simulated as well to estimate the effect of body posture on dose coefficients and determine the effect of posture on dosimetric estimation and reconstruction. Results additionally demonstrate that 137 Cs and the Watt fission spectra are not ideal general surrogate sources for fission weapons, which may be considered for experimental testing of medical countermeasures. Supplementary data provided tabulates the compilation of organ dose-rate coefficients in this study.

45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗
DOE OSTI2024

Toward a long-lifetime polarized photoelectron gun for the Ce+BAF positron source

The addition of spin-polarized, continuous-wave (c.w.) positron beams to the 12 GeV Continuous Electron Beam Accelerator Facility (CEBAF) would provide a significant capability to the experimental nuclear physics program at Jefferson Lab. Based on bremsstrahlung and pair-production in a high-Z target, the positron source requires a 120 MeV spin-polarized c.w. electron beam of several milliamperes. While the beam dynamics of the high-current electron beam are tenable, sustaining this current for weeks of user operations requires an unprecedented charge lifetime from a high-polarization GaAs-based photocathode. A promising approach to exceed the kilocoulomb charge lifetime barrier is reducing the ion back-bombardment fluence at the photocathode. By increasing the laser size and managing the emittance growth with an adequate cathode/anode design, significantly enhanced charge lifetime may be achieved. Based upon a new simulation model that qualitatively explains the lifetime data previously measured at different spot sizes, we describe the practical implications on the parameter space available for a kilocoulomb-lifetime polarized photogun design.

Bruker, M.↗
DOE OSTI2020

Deciphering the PMI Surface Chemistry of Lithium-Based PFCs and its Effects on High Performance Plasmas in NSTX

The primary aims or objectives of the proposed work here focus on the dynamic measurement (Illinois PI Allain) and atomistic-based multi-scale computational atomistic simulations (UTK PI: Wirth) of: 1) lithium coatings with re-deposited hydrogenated carbon surfaces and high-Z substrates, 2) lithium coatings on boronized ATJ graphite surfaces and high-Z substrates and 3) lithium coatings on ATJ graphite and high-Z substrates with variation in temperature. Three main primary tasks are proposed to address each of the primary objectives listed above: 1) dynamic in-situ irradiation with D+, He+ species of lithium coatings with variation in interface substrate morphology (e.g. smooth vs rough, fuzz vs nanostructured); 2) validation and coupling with multi-scale computational simulations connecting the irradiated surface to the plasma edge; 3) extrapolation to long-pulse conditions (e.g. function of flux vs fluence studies). In primary objective #3, temperature-based effects will be studied to evaluate three primary mechanisms: a) D uptake and recycling, b) erosion, and c) surface melting and evaporation as a function of substrate morphology and chemistry.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Sample Management Tools for the Irradiation Test Area (ITA)

The ITA offers onsite radiation testing, and will be available starting Fall 2020. It is projected to have testing slots about 40x/year and a capacity to reach fluences of 1016 protons/cm2 in four hours. A 400 MeV proton beam is directed to samples in a concrete shielding cave, with variations in place available to the user’s preferences. The ITA regulates all its experiment data through ITA’s Data Manager (IDM).

43 PARTICLE ACCELERATORS↗
DOE OSTI2020

Destructive PIE and Safety Testing of Six AGR-2 UO 2 Capsule 3 Compacts

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program’s second irradiation experiment (AGR-2) was irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) from June 2010 to October 2013 (Collin 2014). The fuel compacts in this experiment held either tristructural isotropic (TRISO)-coated spherical kernels of uranium oxide (UO2) or TRISO-coated kernels containing both uranium carbide and uranium oxide phases (UCO). There were six separately monitored and controlled capsules in the AGR-2 test train. Capsule 3 held twelve compacts containing UO2-TRISO particles fabricated by BWX Technologies Nuclear Operations Group. The AGR-2 TRISO particles were fabricated in a pilot-scale fluidized-bed chemical vapor deposition (FB-CVD) furnace with a coating chamber inner diameter of 150 mm (Phillips, Barnes, and Hunn 2010), which was a change from the first irradiation experiment (AGR-1) particles that had been coated in a lab-scale FB-CVD coating system with a chamber inner diameter of 50 mm (Lowden 2006). The TRISO particles were overcoated with resinated graphite flake at Oak Ridge National Laboratory (ORNL), and the overcoated particles were pressed into one-inch-long, half-inch-diameter cylinders (Hunn, Montgomery, and Pappano 2010). Each cylindrical compact held an average of 1,543 TRISO particles with 9.6% enriched UO2 kernels that had a nominal diameter of 500 μm (Hunn, Savage, and Silva 2012). Capsule 3 compacts were irradiated to average calculated burnups of 9.01–10.69% fissions per initial metal atom (FIMA), and the average calculated fluences of fast neutrons with energies E n > 0.18 MeV were 3.05–3.53×10 25 n/m 2 (Sterbentz 2014). The calculated time-average, volume-average Capsule 3 compact temperatures were 996–1,062°C. However, Capsule 3 compact temperatures varied several hundred degrees across each compact, and the timeaverage minimum (TA min ) and time-average maximum (TA max ) temperatures were between 889–999°C and 1,072–1,105°C, respectively (Hawkes 2014). After irradiation, the AGR-2 test train was transferred from ATR to the INL Materials and Fuels Complex for inspection and disassembly (Ploger, Demkowicz, and Harp 2015). The initial inspection included dimensional metrology of the compacts and graphite fuel holders. Like all the AGR-2 compacts, the compacts in Capsule 3 shrank slightly during irradiation, as expected, with an average length reduction of 1.07–1.24% and an average diameter reduction of 0.13–0.41%. Post-irradiation examination (PIE) of the capsule components was completed to measure select fission products ( 90 Sr, 110m Ag, 134 Cs, 137 Cs, 144 Ce, and 154 Eu) outside the compacts (Stempien and Demkowicz 2020). This involved gamma counting of the graphite and graphoil spacers at the top and bottom of each capsule, acid leaching for radiochemical analysis of fission products on the metallic capsule components, and burn-leach analysis of the graphite holders. The total amount of 110mAg measured on the Capsule 3 components was 13% of the calculated capsule inventory. This was significantly lower than the amount of 110m Ag measured on the three UCO capsule components, which ranged from 32–70%. The lower 110m Ag release in Capsule 3 was likely due to lower peak temperatures compared with the UCO fuel capsules (Hawkes 2014). Measured inventories of the other select fission products were also lower in Capsule 3.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Active Irradiation Testing of Temperature Sensing Capability of Clad Sapphire Optical Fibers with Type 2 Bragg Gratings using Optical Backscatter Reflectometry

This work investigated the temperature-sensing capabilities of Type II fiber Bragg gratings (FBGs) inscribed in sapphire optical fiber in the Ohio State University Research Reactor (OSURR). A Luna Innovations 4600 optical backscatter reflectometer (OBR) was used to interrogate FBGs in three sapphire optical fibers with an internal cladding that was developed by Ohio State. The sapphire fibers were fusion-spliced to silica lead fibers. In addition to the sapphire fibers, two silica fibers with the same Type II FBGs were included in the experiment as a reference. The fiber-based sensors were irradiated in the Central Irradiation Facility (CIF) of the OSURR for a total of 40 hours over 5 days at a reactor power of 450 kW, resulting in a total neutron fluence of approximately 2.5 × 10 18 n/cm 2 and a gamma dose of approximately 3.48 Grad. Two K-type thermocouples were included within the rig: one at the core centerline and the other 12 inches above the top of the core. The temperature was not actively controlled, but the thermocouples were used to provide a reference to which the fiber-based temperature measurements could be compared.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Transition Core Planning and Safety Analyses in Support of LEU Fuel Conversion of the University of Missouri Research Reactor (MURR)

The University of Missouri Research Reactor (MURR®) is one of six U.S. High Performance Research Reactors (USHPRR), including one critical facility, that is working with the National Nuclear Security Administration (NNSA) Office of Material Management and Minimization (M3) Reactor Conversion Program to convert from highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel. The M3 Reactor Conversion USHPRR Project objectives include the development of LEU fuel element designs that will ensure safe reactor operations and to maintain the existing experimental performance of each facility. The work is being conducted through many inter-related activities being completed by four Project Pillars: Fuel Qualification (FQ), Fuel Fabrication (FF), Reactor Conversion (RC), and Cross Cutting (CC). A new type of LEU fuel based on an alloy of uranium-10 wt% molybdenum (U-10Mo) is expected to allow the conversion of those USHPRR, like MURR, requiring higher density fuels. The very-high-density LEU U-10Mo monolithic fuel is currently undergoing irradiation testing and post-irradiation examination under a planned and documented fuel qualification effort. The FQ Pillar will document fuel property and fuel performance data and qualify the fuel for use in these reactors. The FF Pillar is fabricating fuel for ongoing and future irradiation tests, as well as conducting fabrication demonstrations to validate or update preliminary fabrication assumptions. The FF Pillar is also working to develop and install commercial manufacturing capacity with the U-10Mo monolithic fuel to produce prototypic fuel. Working with the RC Pillar at Argonne, MURR has progressed through a preliminary fuel element design using preliminary data for the proposed monolithic alloy of U-10Mo. Analyses were completed in previous work that found for typical equilibrium operations with the preliminary LEU fuel element design, in conjunction with a power uprate to 12 MW and appropriate changes to the MURR Limiting safety system settings (LSSS), MURR will have adequate margins to safety for steady-state operations and postulated transient accidents and will have experimental performance in key locations that meets or exceeds current operations with HEU fuel. The purpose of this work is to develop a sequence of transition cycles that will enable MURR to transition from operation with the reactor core loaded with fresh LEU fuel elements only to typical equilibrium operations with mixed-burnup cores following conversion while meeting operational requirements on safety and experimental performance. It is expected that the use of fresh LEU fuel at conversion and subsequent low burnup of the LEU fuel elements that will initially be available for use following conversion will result in critical control blade positions that will substantially change the axial power distribution in the core and the neutron flux available in key experimental locations relative to equilibrium LEU operations. Given the constraints of MURR safety margins, operational practices, and production and research, a novel method has been developed to identify a transition sequence that minimizes the time MURR operates atypically compared to the current prototypic cycles using HEU fuel. The proposed transition sequence moves quickly to the same sort of equilibrium cycles for the LEU fuel that have already been evaluated in documented preliminary safety analyses. Although shifting the neutron flux peak to the lower half of the core during initial cycles with LEU at 12 MW reduces the experiment performance in some key locations relative to current HEU operations at 10 MW, all LEU cores provide an average performance that meets or exceeds that of HEU. An LEU cycle is reached that meets or exceeds the level of experimental performance predicted for current HEU and equilibrium LEU operations in more than 450 key locations identified by a reactor specialist at MURR by the 23rd cycle following conversion and that afterwards will enable MURR to consistently meet its experimental performance requirements. The proposed transition sequence only requires the fabrication of 34 fresh LEU elements in the first year of operation and does not exceed the anticipated availability of fresh elements that can be produced by the fuel fabricator. By the third year after conversion, 22 fresh LEU elements will be required each year, which is the same as expected for equilibrium LEU operations and the same as current operations with HEU fuel. The proposed transition sequence thus combines a relatively short time period before equilibrium burnup is achieved, a temporary increase of fuel elements needed annually relative to typical operations that are within the production capabilities of the fuel fabricator, and demonstrates comparable experimental performance of the LEU cores relative to current HEU operations. Further measures may be taken to reduce any initial experimental performance penalty even further, where possible, by repositioning certain experiments to leverage the increased performance in the lower axial experimental positions in the initial cycles following conversion or leaving the experiments in the irradiation facilities longer in order to achieve the required neutron fluence. This analysis may require refinement depending on the experimental facilities in use at the time of conversion. Nonetheless, the results presented here, including the experimental performance, core burnup, and critical control blade positions throughout the transition cycles, show that the proposed transition cycle fuel management patterns are consistent with what is expected and desired for MURR operation with LEU U-10Mo fuel. Detailed core power distributions from the neutronics models were also used to evaluate safety margins during steady-state operations for the selected transition cycles and the equilibrium LEU core. It is shown that there are adequate safety margins for both steady-state operations and postulated accident scenarios. For the steady-state operations with the preliminary LEU fuel element design the analysis predicts at least 2.49 MW margin to the onset of flow instability at the LSSS power of 15 MW. Considering the LSSS power is 125% of full license power, the margin to OFI is sufficient. In addition, the critical heat flux ratio at LSSS power is well above the requirement of CHFR > 2.0 from NUREG-1537 for all considered cases. For postulated transient accidents, the minimum margin to the fuel temperature safety limit is at least 109 °C. In summary, the proposed sequence of core loadings for MURR operations following conversion to LEU fuel and a power uprate to 12 MW provides sufficient safety margins for both steady-state operations and postulated transient accidents during a proposed sequence of transition cycles to equilibrium operations. Analysis has shown that there are some local experimental performance penalties during the initial cycles. Although there are local shifts in the experimental performance, on average all LEU cores at 12 MW have equal or higher performance than HEU at 10 MW. Temporary adjustments are being planned that will produce suitable experimental performance during these cycles. The results indicate that for the equilibrium LEU core the experimental performance exceeds that of current HEU operations in all key locations while also demonstrating sufficient safety margins.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Intermediate-Term Thermal Aging Effect Evaluation for Grade 92 and 316L at The LWR Relevant Temperature

Life extension of the existing nuclear reactors imposes accumulated damages, such as higher fluences and longer period of corrosion, to structural materials, which would result in significant challenges to the traditional reactor materials such as type 304 and 316 stainless steels. Advanced alloys with superior radiation resistance will increase safety margins, design flexibility, and economics for not only the life extension of the existing fleet but also new builds with advanced reactor designs. The Electric Power Research Institute (EPRI) teamed up with Department of Energy (DOE) on the Advanced Radiation Resistant Materials (ARRM) program, aiming to develop and test degradation resistant alloys from current commercial alloy specifications by 2021 to a new advanced alloy with superior degradation resistance in light water reactor (LWR)-relevant environments by 2024. Based on a comprehensive microstructure and property screening performed in Phase-1 of the ARRM program, a total of five alloys, together with 316L and X-750 as references, were down-selected for Phase-2 neutron irradiation studies. Because thermal aging could exert a synergistic effect on neutron irradiation due to the low neutron damage rate on the order of 10 –7 displacements per atom per second (dpa/s), Grade 92 (one of the five down-selected alloys) and two heats of 316L were selected in this task to study the effect of aging at 350°C for 12.5–12.7 kh on microstructure and mechanical properties. The aging time is approximately corresponding to 5 dpa neutron irradiation, which can be served as a reference for the 5-dpa-irradiated samples of the alloys to help understand the independent neutron damage influence on microstructure and mechanical properties. Optical microscopy and scanning electron microscopy were used for microstructural characterization. Hardness, tensile, Charpy impact toughness, and fracture toughness in the ductile regime of the aged samples were examined. The aging of Grade 92 led to the formation of many Laves phase in sizes of ~100–200 nm and resulted in some reduction in hardness and yield/ultimate tensile strength with some increases in uniform and total plastic elongations, which may have helped the enhancement of Charpy impact toughness, e.g., ~4 J increase in upper-shelf energy and 20.5°C reduction in ductile-brittle transition temperature compared with the unaged condition. The fracture toughness of the aged Grade 92 showed decent toughness of ~303 MPa√m (K Jq ) with ~87 tearing modulus at 22°C, which reduced to ~241 MPa√m (~20% reduction) with ~75 tearing modulus at 300°C. The two heats of 316L are differentiated by their amounts of δ-ferrite, i.e., ~1 vol% in heat T1103 and ~4 vol% in heat N5B8. The two heats of 316L were subjected to 15% cold work (CW) as this condition is often used in nuclear reactors to help trapping radiation-induced defects and thus postpone the steady state swelling stage. The aging did not result in any noticeable microstructure changes, except for possible segregations at grain boundaries that need to be further investigated. The aging resulted in slight reduction in hardness of 316L-T1103 and slight increase in hardness of 316L-N5B8, which is consistent with their tensile testing results. The 15%CW might have introduced some inhomogeneity, resulting in large standard deviations in hardness, which were reduced after the aging. The aged 316L followed the same trend as the unaged condition for the strength and elongation results from the tensile testing. Minimum elongations appeared within ~300–600°C, above which the elongations seem to increase. Unlike the negligible or minor changes in microstructure, hardness and tensile results of the two heats of 316L, their fracture toughness showed noticeable difference. The aged 316L-T1103 showed good toughness of ~350 MPa√m with ~80 tearing modulus at 22°C, which reduced to ~271 MPa√m (~23% reduction) with ~83 tearing modulus at 300°C. In contrast, the aged 316L-N5B8 showed lower toughness of ~244 MPa√m with ~53 tearing modulus at 22°C, which reduced to ~180 MPa√m (~26% reduction) with ~31 tearing modulus at 300°C. The preliminary results indicate that the presence of high volume of δ-ferrite would noticeable impair fracture toughness although it may not influence hardness and tensile properties.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Wireless Instrumented RB Experiment Preliminary Design and Analysis

The ability to deploy new nuclear fuels for current or future reactor concepts requires carefully designed experiments to generate data to support fuel qualification. Ideally these experiments would include state of-the-art sensing to maximize the amount of in situ data that can be collected during operation. Furthermore, advanced reactor systems can take advantage of integrated in-core sensing technologies to maximize fuel utilization, reduce unnecessary conservativism in design margins, and improve operator’s understanding of limiting peaking factors. Before any novel sensing technologies can be readily adopted for nuclear applications, they must first demonstrate acceptable performance in test reactors. This report summarizes the preliminary design and analysis of the most highly instrumented irradiation experiment ever performed in the removable beryllium (RB) positions of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). The Wireless Instrumented RB Experiment 2021 (WIRE-21) will test a wide range of sensors including wireless sensors being developed by Westinghouse Electric Company (WEC) that could provide in situ measurements of peak fuel temperatures and fuel rod pressurization due to fission gas release. The ability to wirelessly transmit a signal through the fuel rod’s cladding is critical to improving fuel monitoring capabilities without requiring signal penetrations through the cladding pressure boundary, which would significantly impact fuel fabrication, handling, and operation. Other sensors that will be tested in WIRE-21 include an array of thermocouples, self-powered neutron detectors (SPNDs), and spatially distributed fiber-optic temperature sensors. More generally, WIRE-21 will establish a flexible irradiation vehicle design to allow accelerated, economical testing of advanced sensor technologies while leveraging the extremely high neutron flux that is available in HFIR. This report summarizes the mechanical design for WIRE-21, the experimental test matrix, initial neutronic and thermal design analyses, and the active monitoring and control system enhancements necessary to support testing of advanced sensor technologies. The containment for WIRE-21 is similar to previous RB irradiation vehicles but includes a few modifications, most notably the use of integrated compression seals to pass a larger number of sensor leads through the experiment’s pressure boundary. In addition to the sensor leads, inert gas lines are passed into the experiment to enable active temperature control and the ability to pneumatically actuate a bellows-driven pressure sensor. WIRE-21 is targeting temperatures (300–350°C) and neutron fluence levels (~10 22 n/cm 2 ) relevant to light water reactors (LWRs), but the flexible design of the experiment vehicle allows much higher operating temperatures (>1,100°C). Neutronic calculations determine the neutron flux conditions as well as the nuclear heating within the experiments. These results are used as inputs to detailed thermal finite element calculations, which are required to evaluate the complex, three-dimensional heat transfer that occurs within WEC’s wireless sensor enclosures. Initial results show that the temperatures of the sensors’ enclosures and the metal bellows can be operated near the temperature range of LWR coolants and cladding while simultaneously increasing the temperature of a surrogate fuel material to values in the range of 800–1200°C to simulate centerline fuel temperatures during LWR operation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗