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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 811 records · Page 45
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Specimen Size and Geometry Effects on the Master Curve Fracture Toughness Measurements of EUROFER97 and F82H Steels

EUROFER97 and F82H are two leading reduced-activation ferritic-martensitic (RAFM) steels for fusion blanket applications. Exposure to the harsh environment of fusion reactors can result in severe degradation of materials fracture toughness (FT). Thus, the post-irradiation evaluation of FT is critical to understanding the material behavior. Due to the space constraint of irradiation facilities, the development of small specimen test techniques (SSTT) is necessary to evaluate the performance of irradiated materials. In this study, we evaluated the specimen size and geometry effects on the ductile-to-brittle transition FT of EUROFER97 batch-3 and F82H-BA12 steels. The specimen thicknesses ranged from 1.65 to 12.7 mm and the geometries included 1.65 mm bend bar, 4 mm mini-compact tension (miniCT), and 0.5T compact tension (CT) specimens. Fracture toughness testing and evaluations were performed using the Master Curve method in the ASTM E1921-19 standard. After size correction to 1T size using the Master Curve method, no specimen size effect was observed between the 4 mm miniCT and 0.5T CT specimens for the Master Curve reference temperature T0Q, while the bend bars yielded a higher T0Q. A strong effect of fatigue precrack front straightness on T0Q for 0.5T CT specimens was observed. The minimum number of specimens needed for each specimen geometry has been determined.

Chen, Xiang↗
NASA NTRS1975

Wavefront-error evaluation by mathematical analysis of experimental Foucault-test data

The diffraction theory of the Foucault test provides an integral formula expressing the complex amplitude and irradiance distribution in the Foucault pattern of a test mirror (lens) as a function of wavefront error. Recent literature presents methods of inverting this formula to express wavefront error in terms of irradiance in the Foucault pattern. The present paper describes a study in which the inversion formulation was applied to photometric Foucault-test measurements on a nearly diffraction-limited mirror to determine wavefront errors for direct comparison with ones determined from scatter-plate interferometer measurements. The results affirm the practicability of the Foucault test for quantitative wavefront analysis of very small errors, and they reveal the fallacy of the prevalent belief that the test is limited to qualitative use only. Implications of the results with regard to optical testing and the potential use of the Foucault test for wavefront analysis in orbital space telescopes are discussed.

Wilson, R. G.↗
DOE OSTI2024

BISON Simulated and Experimental Fission Product Release Comparisons from Reradiated AGR-3/4 Compacts During High Temperature Heating Tests

The fuel performance modeling code BISON was used to predict the release of fission products iodine-131 (131I), xenon-133 (133Xe), and krypton-85 (85Kr) from four re-irradiated AGR-3/4 fuel compacts containing tristructural isotropic (TRISO) coated particles during high-temperature isothermal heating tests. The AGR-3/4 fuel compacts were irradiated in the Advanced Test Reactor (ATR) as part of the third and fourth series of planned experiments to support the Advanced Gas Reactor (AGR) Program. They were subsequently stored and re-irradiated in the Neutron Radiography (NRAD) reactor for approximately five days and then stored for another five to eight days before being subjected to isothermal heating tests in the Fuel Accident Condition Simulation (FACS furnace) for 200 to 300 hours at temperatures between 1000°C and 1600°C to evaluate fission product release at elevated temperatures. New nuclide-specific fission product source term models for the three nuclides of interest were developed using the reactor multiphysics code Griffin and implemented into BISON to support this work. The new source term models were incorporated into coupled compact- and particle-scale BISON simulations, which predict spatially- and temporally-resolved radionuclide generation, radioactive decay, transport, and release throughout the entire irradiation history, including the initial ATR irradiation, NRAD re-irradiations, FACS heating tests, and intermediate periods spent in storage. The experimentally measured fission product release from the heating tests were compared to modeling release predictions calculated by BISON to evaluate how well the code compares to experimental results. Overall, the experimental measured and BISON predicted comparative release results varied but generally agreed to within 5 particle equivalents. Comparative release results identified general observations to take into consideration to help refine future models and reduce uncertainties associated with both the measurement results and predictive results. This includes developing new uranium oxycarbide (UCO) specific kernel diffusivities for the three isotopes examined to more accurately reflect the material properties of the fuel form. Deriving new diffusivities will aid in producing a more informed BISON model

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Microstructure, stored energy, and stability of H/He-filled nanocavities in low temperature irradiated Inconel 718

The microstructure, trapped transmutation gases, stored energy, and mechanical behavior of samples from an irradiated Inconel 718 proton beam window were characterized using transmission electron microcopy, thermal desorption spectrometry (TDS), differential scanning calorimetry (DSC), and tensile testing. In the as-irradiated condition the microstructure contained a high number density of 1–3 nm gas-filled nanocavities. Emissions of trapped gases, H and He, during TDS correlated with peaks of the energy release curves from DSC examinations, which suggest these gases were likely stored in highly stable defect traps. The stored energy from radiation damage saturated at doses of a few dpa and did not increase with increasing radiation dose, but the amount of stored H and He increased with increasing dose. Effects of post-irradiation annealing were studied as well. After exposure to 700 °C, the nanocavities grew only slightly to 2–4 nm in diameter, but after exposure to 900 °C, the cavities grew to 10–20 nm in diameter and electron energy-loss spectroscopy showed these cavities contained a core of He surrounded by a shell of H. Further, this study demonstrated that the irradiation defect structures containing H and He were remarkably stable during irradiation and after exposure up to 700 °C. The effect of the irradiation temperature, defect mobility, and interaction of H, He, and irradiation defects on mechanical behavior provides insight into the processes responsible for the unusual recovery in ductility with increasing radiation dose observed in Inconel 718 after high energy proton and spallation neutron irradiation.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Measurement and Reporting Guidelines for Solar Mirror Aging Tests Using Xenon Arc Lamp Exposure (XALE)

This technical report provides a guideline for accelerated aging tests on solar reflectors. Solar reflectors must have a high durability, with goal lifespans reaching 30 years. To test and develop state-of-the-art solar reflectors, relevant durability tests must be conducted. Existing guidelines for materials testing include isothermal testing, humidity testing, corrosion, etc., but many tests do not specify to include solar irradiance or address its relevance to solar reflector durability. The guideline focuses on test design and reporting guidelines for using Xenon Arc Lamp Exposure (XALE). The XALE environmental chambers can subject samples to a combination of relevant environmental factors including concentrated solar irradiance, heat, humidity, and weathering cycles. The guideline is meant to provide information on developing relevant test conditions, exposure times, reflectance characterization procedures, and data reporting.

14 SOLAR ENERGY↗
DOE OSTI2023

Local Power Impact Experiment Design for a New Fuel Type for use in the Advanced Test Reactor

The Advanced Test Reactor (ATR), and complimentary zero-power ATR Critical (ATRC) reactor, located at Idaho National Labs (INL), are undergoing conversion from Highly Enriched Uranium (HEU) to Low Enriched Uranium (LEU). Both have a variety of testing locations that can receive large variations in flux due to its unique serpentine design, consisting of five lobes (see Figure 1). Initial criticality and power distribution throughout the core are controlled by core-external outer shim control cylinders (OSCCs). Distinct test loops allow for testing at specific temperatures, pressures, and irradiation conditions. The ATR is one of the key nuclear engineering research and testing facilities within the DOE National Laboratory Complex, and the ATRC supports its operation [1]. Currently, the Office of Material Management and Minimization (M3) within the National Nuclear Security Administration of the DOE is working to convert the remaining research reactors, including the ATR, from 93% HEU fuel to 19.75% LEU fuel (LEU) to support non-proliferation [2]. Extensive materials testing at INL and internationally has demonstrated that a high-density uranium molybdenum (U 10Mo) alloy can meet the performance requirements of the remaining high powered research reactors. However, there are many technical challenges to address before the conversion to LEU can be successful, including the accurate characterization of the reactor core physics with LEU fuel. Reactor physics safety evaluations currently use Monte Carlo for the 21st Century (MC21), a continuous-energy Monte Carlo radiation transport code [3]. Existing MC21 models of the ATR and ATRC cores have a validation basis for use in neutronics analyses with HEU fuel. The models are used to support safety analyses that include comparisons to the safety requirements for the reactors. However, the use of the LOWE element in the ATR and ATRC is not currently covered by the current model validation basis. To deploy the new fuel type, extensive computational reactor physics support is necessary to support the use of LOWE in the ATR and ATRC. Therefore, LOWE requires a rigorous validation basis, aligned with that of HEU fuel, that takes advantage of the existing software tools and processes currently used for the ATR and ATRC. The experiment to validate of the MC21 models for determining power, the Power Impact Validation Experiment, will consist of two flux runs in the ATRC, one with fully HEU loading and one with a single LOWE element. Both flux runs will be instrumented with 20 sets of azimuthal fission wires and 3 sets of axial fission wires, as shown in Figure 4. Standard flux run methodology will be used [4]. Power Impact Validation Experiment data will be compared against MC21 calculated data, both for absolute fission rate accuracy and to determine the relative change in fission rates between the two runs. The results of the Power Impact Validation Experiment and subsequent evaluations will provide the validation basis for MC21 for use with LOWE elements. Key features of the Power Impact Validation Experiment include: (1) Two flux runs to allow for LOWE perturbed measurements to be compared to already validated measurements taken from a full core of HEU fuel, (2) Optimization of instrumentation to balance analytical needs with practical considerations (e.g., limited time window to count beta particles from fission products), and (3) Standard ATRC core loading, including both driver positions and flux traps, to minimize cost while remaining representative of typical ATR core loading.

42 ENGINEERING↗
DOE OSTI2024

Preliminary analysis of TREAT free-field experiments using openmc

This work analyses activation calculations for dosimetry materials during a steady-state irradiation in the Transient Reactor Test (TREAT) reactor core. Hence, we developed a workflow based on the Monte Carlo code OpenMC alongside a custom depletion solver. The irradiation-induced activity as a function of time is computed, and several sensitivity studies are performed to evaluate uncertainty. This study has shown activity computations are sensitive to flux amplitude, irradiation time, atoms quantity and microscopic cross sections. Stochastic uncertainties have been propagated to evaluate the activity uncertainty for each dosimetry material. Most uncertainties are below our target of 3%, which demonstrates OpenMC as a powerful predictive and analysis tool. The precise results obtained through this newly developed computation scheme will be used in future experiments to characterize quantities of interest when operating the TREAT reactor in new configurations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1982

Engineering sciences area and module performance and failure analysis area

Photovoltaic-array/power-conditioner interface studies are updated. An experiment conducted to evaluate different operating-point strategies, such as constant voltage and pilot cells, and to determine array energy losses when the array is operated off the maximum power points is described. Initial results over a test period of three and a half weeks showed a 2% energy loss when the array is operated at a fixed voltage. Degraded-array studies conducted at NE RES that used a range of simulated common types of degraded I-V curves are reviewed. The instrumentation installed at the JPL field-test site to obtain the irradiance data was described. Experiments using an optical filter to adjust the spectral irradiance of the large-area pulsed solar simulator (LAPSS) to AM1.5 are described. Residential-array research activity is reviewed. Voltage isolation test results are described. Experiments performed on one type of module to determine the relationship between leakage current and temperature are reviewed. An encapsulated-cell testing approach is explained. The test program, data reduction methods, and initial results of long-duration module testing are described.

Ross, R. G., Jr.↗
NASA NTRS2010

Cosmogenic Samarium-150 and Calcium-41 in Norton County

Though brecciated, the Norton County (NC) aubrite contains little or no trapped noble gas and has been widely assumed to have a simple if unusually long cosmic ray exposure (CRE), 115 Ma. One goal of this ongoing study of NC has been to search for signs of pre-irradiation as proposed. One may test for multiple stages of CRE by comparing thermal neutron fluences inferred from Ca-41 (t(sub 1/2)=0.1 Ma) activities, which reflect irradiation conditions over the last approximately 0.3 Ma, with those inferred from (stable) Sm isotope abundances, which integrate over the entire CRE history. In the case of a one-stage exposure the fluences should agree. We focus on these particular comparisons because the properties of NC - its long CRE exposure, relatively large size, and low iron concentration - all promised high production rates and ease of measurement. Previously, we reported on several cosmogenic nuclides in NC. Here we present new Ca-41 data, Sm isotope measurements, and comparisons with model calculations of cosmic ray production.

Fink, D.↗
DOE OSTI2024

Preliminary analysis of TREAT free-field experiments using OpenMC

This work analyses activation calculations for dosimetry materials during a steady-state irradiation in the Transient Reactor Test (TREAT) reactor core. Hence, we developed a workflow based on the Monte Carlo code OpenMC alongside a custom depletion solver. The irradiation-induced activity as a function of time is computed, and several sensitivity studies are performed to evaluate uncertainty. This study has shown activity computations are sensitive to flux amplitude, irradiation time, atoms quantity and microscopic cross sections. Stochastic uncertainties have been propagated to evaluate the activity uncertainty for each dosimetry material. Most uncertainties are below our target of 3%, which demonstrates OpenMC as a powerful predictive and analysis tool. The precise results obtained through this newly developed computation scheme will be used in future experiments to characterize quantities of interest when operating the TREAT reactor in new configurations.

61 - RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2024

Comparison of hardening and microstructures of ferritic/martensitic steels irradiated with fast neutrons and dual ions

Ferritic/martensitic steels T91 and HT9 were irradiated with neutrons (BOR-60 reactor) and dual ions (9 MeV Fe 3+ and 3.42 MeV energy degraded He 2+ ) from 369 to 520 °C and damage levels of 16.6 to 72 dpa to quantify the possibility of using ion irradiation to simulate neutron irradiation in terms of microstructures and mechanical properties. Nanoindentation testing was performed to obtain the bulk equivalent hardness of the dual-ion irradiated samples. For the neutron irradiated samples, both nanoindentation and Vickers hardness testing were conducted. Transmission Electron Microscopy (TEM) characterizations of the cavities, dislocation loops and precipitates were conducted to account for the strengthening contribution of each microstructure element. The good agreement between the microstructure-predicted (dispersed barrier hardening) and measured strength of the irradiated specimens demonstrated the accuracy of the strengthening model and the nanoindentation tests. Furthermore, the comparison of mechanical property and microstructure changes in ion and neutron irradiated structural materials indicated that ion irradiation replicated many neutron irradiation features. However, a single 70 °C temperature shift is insufficient to match all complex microstructures of neutron vs. ion irradiation over the irradiation temperature range of 369–520 °C.

36 MATERIALS SCIENCE↗
DOE OSTI2023

High-temperature steam oxidation study of irradiated FeCrAl defueled specimens

Post irradiation examinations (PIE) were performed on irradiated iron-chromium-aluminum (FeCrAl) specimens. These FeCrAl specimens were fabricated at the US Department of Energy's Oak Ridge National Laboratory (ORNL). The experimental setup involved subjecting FeCrAl cladding, along with UO 2 pellets, to irradiation in the Idaho National Laboratory Advanced Test Reactor (ATR). In parallel, the FeCrAl alloy tubing without UO 2 pellets was irradiated at ORNL's High Flux Isotope Reactor (HFIR). After irradiation, the ATR-irradiated rodlet was transported to an ORNL hot cell, where it was sectioned into multiple samples for the PIE and severe-accident testing. The sectioning process revealed that the fuel was not bonded to the cladding and could be easily detached from sectioned cladding slices. Microstructural analysis of the fuel cross sections demonstrated no significant interaction between the fuel and the cladding. Additionally, high-temperature steam oxidation tests on defueled cladding segments showed minimal oxygen uptake even at 1200 °C. Here, the ATR-irradiated specimens began to exhibit signs of enhanced oxidation upon reaching a temperature of 1300 °C. Furthermore, enhanced oxidation was observed on the inner surface of the ATR-irradiated FeCrAl specimen, which had been subjected to 1300 °C for a duration of 1 min. By contrast, high-temperature steam oxidation experiments indicated that the HFIR-irradiated FeCrAl cladding provided good thermal stability when exposed to 1300 °C for up to 4 h. Comparative analysis encompassing the oxidation behavior of the ATR-irradiated fueled FeCrAl, HFIR-irradiated unfueled FeCrAl, and unirradiated FeCrAl suggests that the fuel–cladding interaction, although not visible via standard microscale electron microscopy measurements, may accelerate the deterioration of FeCrAl cladding in beyond-design-basis accident scenarios.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

As-run thermal-hydraulics analysis of the EPRI-2 experiment

The EPRI-2 experiment was designed to irradiate various types of reactor pressure vessel steels at a temperature of 288°C (PLN-3934). The specimens were irradiated in a non-instrumented test train inside a pressurized water loop in the center lobe of the ATR during PALM cycle 153B. Temperature was indicated by post-irradiation examination of melt wires placed within the test train. The purpose of this analysis is to calculate specimen temperature using measured data on reactor power and as-run calculations of heating rates of the test train. The experiment contains several melt wires spanning the temperature range 239°C to 327°C. The accuracy of the model is assessed by comparing the measured and calculated in-pile tube inlet to outlet temperature difference, and comparing the calculated specimen temperature to the temperature range indicated by examination of the melt wires.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1986

Long term effects of low doses of 56Fe ions on the brain and retina of the mouse: ultrastructural and behavioral studies

Eight month old male C57BL6 mice were exposed without anesthesia to whole-body irradiation in circular holders. The mice were tested for behavioral decrements after 0.5 and 50 rads of Fe particle irradiation at 6 and 12 months post irradiation to obtain long term results. A standard maze was used and the animals were timed for completion thereof. A string test also was administered to the mice, testing their ability to grasp and move along a string to safety. The results from animals exposed to 50 rads were significantly different from [correction of fron] control results to p = < .001 in both systems of testing. The hippocampus (believed to be the location of environmental interaction in the brain) and the retina were examined for ultrastructural changes. The ultrastructural changes were similar to those we found in our Cosmos 782, 936 and in our Argon experiments. The mouse data indicate that iron particles were able to induce long term changes in the central nervous system which lead to behavioral impairment.

NASA Program Space Medicine↗
DOE OSTI2020

Characterization of mechanical properties and deformation behavior of highly irradiated 316L stainless steel from target modules at the Spallation Neutron Source using digital image correlation analysis

Tensile specimens with irradiation doses ranging from approximately 0.7 to 7.6 displacements per atom (dpa) were fabricated from 316L samples removed from Spallation Neutron Source (SNS) target vessels after service. Tensile tests were performed in conjunction with digital image correlation (DIC) characterization, which permitted detailed analysis of the strain distribution and deformation behavior of the specimen gauge sections during testing. The results show, while considerable hardening (500–700 MPa increase in yield strength) occurred in the 316L SNS target material with increasing dose, appreciable ductility (19–23% total elongation) remained in the material after irradiation to 7.6 dpa in the mixed neutron/proton environment of the SNS target region. The DIC data showed the strain distribution and deformation behavior of all irradiated specimens were nonuniform during testing. Strain localization in the form of deformation bands or “hotspots” was observed during testing, and in some specimens, the bands propagated as deformation waves along the gauge section. The “traveling wave” deformation behavior affected the amount of elongation each specimen achieved, and the scatter in the elongation data was attributed to the large variation in deformation behavior observed among different specimens. Bands of localized strain, or deformation bands, oriented at specific angles were observed in two-dimensional plots of the strain distribution in the gauge sections of some specimens. Overall, this paper describes the experimental procedures, results, and deformation behavior of tensile specimens from SNS targets after service and illustrates the value of DIC characterization during tensile testing of irradiated materials.

36 MATERIALS SCIENCE↗
NASA NTRS1966

Nondestructive test method accurately sorts mixed bolts

Neutron activation analysis method sorts copper plated steel bolts from nickel plated steel bolts. Copper and nickel plated steel bolt specimens of the same configuration are irradiated with thermal neutrons in a test reactor for a short time. After thermal neutron irradiation, the bolts are analyzed using scintillation energy readout equipment.

Dezeih, C. J.↗
NASA NTRS1973

Unrestrained swelling of uranium-nitride fuel irradiated at temperatures ranging from 1100 to 1400 K (1980 to 2520 R)

Six fuel pins were assembled, encapsulated, and irradiated in the Plum Brook Reactor. The fuel pins employed uranium mononitride (UN) in a stainless steel (type 304L) clad. The pins were irradiated for approximately 4000 hours to burnups of about 2.0 atom percent uranium. The average clad surface temperature during irradiation was about 1100 K (1980 deg R). Since stainless steel has a very low creep strength relative to that of UN at this temperature, these tests simulated unrestrained swelling of UN. The tests indicated that at 1 percent uranium atom burnup the unrestrained diametrical swelling of UN is about 0.5, 0.8, and 1.0 percent at 1223, 1264, and 1306 K (2200, deg 2273 deg, and 2350 deg R), respectively. The tests also indicated that the irradiation induced swelling of unrestrained UN fuel pellets appears to be isotropic.

Rohal, R. G.↗