High dose ion irradiation response of large strain extrusion machined ultrafine grain HT-9 steel
Not Available
36 MATERIALS SCIENCE↗
SEARCH · Engineering Papers
Results for “material irradiation”
Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.
Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.
The effects of neutron and ionizing irradiation on the aqueous corrosion of SiC
Chemical vapor deposited (CVD) SiC and nuclear-grade SiC/SiC composite materials were exposed to light water reactor (LWR) conditions at the MIT Nuclear Reactor (MITR) here. Three sets of samples were exposed within flowing PWR-chemistry-coolant under either 1) neutron and ionizing irradiation with resulting radiolysis products, 2) radiolysis products without neutron flux, 3) loop coolant. Post-irradiation examination (PIE) exhibited an increase in corrosion from radiolysis products (particularly H 2 O 2 ), and potentially a further increase from radiation damage. For CVD SiC, regardless of exposure location, the in-situ observed corrosion rate was less than 1 µm per year, potentially within acceptable limits for LWR application.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Metal Halide Perovskites Demonstrate Radiation Hardness and Defect Healing in Vacuum
Herein, we subject formamidinium lead iodide films to oxygen-containing gases (flowing O 2 or free diffusion of lab atmosphere), inert gases (flowing He, Ar, or N 2 ), and vacuum. Our films are irradiated by Cu Ka X-rays and held at 75 °C while X-ray diffraction is recorded. Under all gas conditions, we observe a reproducible 1.1 ± 0.5 A 3 perovskite lattice contraction from an initial unit cell volume of 256.5 ± 0.8 A 3 concurrent with continuous perovskite loss and lead iodide growth. Oxygen-containing gases increase the reaction rates without materially altering perovskite structural changes. Under the same temperature and irradiation conditions in vacuo, a self-healing reaction is observed, exhibited by a reproducible (0.9 ± 0.3 A 3 ) lattice expansion and stabilization of the perovskite. Furthermore, interactions between the perovskite, defects, and minority phases are simulated by generalized gradient approximation Perdew-Burke-Ernzerhof (GGA-PBE) density functional theory. Lattice contraction indicates an increase in the concentration of Schottky defects-pairs of formamidinium and iodine vacancies. Under irradiation in every atmospheric condition, a solid solution of Schottky defects with a concentration of several percent diffuses and precipitates forming lead iodide and consuming the defects. In the presence of ionized gases, this framework is modified to include the continual loss of formamidinium and iodine ions from the perovskite forming Schottky defects.
14 SOLAR ENERGY↗
Oxidative degradation in polyethylene: the microscopic properties responsible for macroscopic weathering
Radiation-initiated crosslinking of polyethylene has proven to be an effective method of reducing creep and wear. This has allowed polyethylene to find widespread use in engineering and medical sectors where extreme conditions would otherwise cause it to degrade. However, despite their resilience, irradiated polyethylene products are still vulnerable to oxidative degradation in some long-term applications. A good deal of literature exists to discuss the effects that irradiation and oxidative degradation have on polyethylene’s mechanical properties, but not much literature exists to describe the microscopic properties that make irradiated polyethylene susceptible to this process. This review presents a micro-scale look at polyethylene’s morphology and chemistry, with the intention of describing which characteristics make the material more or less favorable for oxidation. By describing the factors that make irradiated polyethylene vulnerable to oxidation in the first place, the hope is to provide insight on how to mitigate oxidation in future applications.
36 MATERIALS SCIENCE↗
Effects of irradiation damage on the hardness and elastic properties of quaternary and high entropy transition metal diborides
Multi-principal component transition metal (TM) diborides represent a class of high-entropy ceramics (HECs) that have received considerable interest in recent years owing to their promising properties for extreme environment applications that include thermal/ environmental barriers, hypersonic vehicles, turbine engines, and next-generation nuclear reactors. While the addition of chemical disorder through the random distribution of TM elements on the cation sublattice has offered opportunities to tailor elastic stiffness and hardness, the effects of irradiation-induced structural damage on the physical properties of these complex materials have remained largely unexplored. To this end, changes in the hardness and elastic moduli of a high-entropy TM diboride (Hf 0.2 Nb 0.2 Ta 0.2 Ti 0.2 Zr 0.2 )B 2 and three of its quaternary subsets following irradiation with 10 MeV gold (Au) ions to fluences of up to 6 × 10 15 Au cm -2 are investigated at the micrometer and sub-micrometer length-scales via the dispersion of laser-generated surface acoustic waves (SAW) and nanoindentation, respectively. The nanoindentation measurements show that the TM diborides exhibit an initial increase in hardness following irradiation with energetic Au ions, with a subsequent decrease in hardness following further irradiation. One quaternary composition, (Hf 1/3 Ta 1/3 Ti 1/3 )B 2 , exhibits a notable exception to the trend and continues to exhibit an increase in hardness with ion irradiation fluence. Although differences in the absolute values of the effective elastic moduli obtained from the measured SAW dispersion and nanoindentation are observed (and attributed to microstructural variations at the measurement length-scale), both techniques yield similar trends in the form of an initial reduction and subsequent saturation in the elastic modulus with increasing ion irradiation fluence. The quaternary TM diboride (Hf 1/3 Ta 1/3 Ti 1/3 )B 2 again exhibits a departure from this trend. The high-entropy TM diboride (Hf 0.2 Nb 0.2 Ta 0.2 Ti 0.2 Zr 0.2 )B 2 exhibits the greatest recovery in hardness and modulus when irradiated to high ion fluences following initial changes at low fluence, indicating superior resistance to radiation-induced damage over its quaternary counterparts. Opportunities for designing HECs with superior hardness and modulus for enhanced radiation resistance (compared to their single constituent counterparts) by tailoring chemical disorder and bond character in the lattice are discussed.
36 MATERIALS SCIENCE↗
Evaluation of BISON metallic fuel performance modeling against experimental measurements within FIPD and IMIS databases
Simulations were conducted using the BISON fuel performance code on an automated process to read initial and operating conditions from two databases—the Fuels Irradiation and Physics Database (FIPD) and Integral Fast Reactor Materials Information System (IMIS) database. These databases contain metallic fuel data from the Experimental Breeder Reactor-II (EBR-II) and the Fast Flux Test Facility (FFTF). The work demonstrates use of an integrated framework to access EBR-II fuel pin data for evaluating fuel performance models contained within BISON to predict fuel performance of next-generation metallic fuel systems. Between IMIS and FIPD, there is enough information to conduct 1,977 unique EBR-II metallic fuel pin histories from 29 different experiments, and 338 pins from FFTF MFF-3 and MFF-5 with varying levels of details between the two databases. Each of these fuel performance histories includes a high-resolution power history, flux history, coolant channel flow rates, and coolant channel temperatures, and new model developments in BISON since the initial demonstration of this integrated framework. Fission gas release (FGR), cumulative damage fraction, fuel axial swelling, FCCI wastage thickness, cladding profilometry, and burnup were all simulated in BISON and compared to post-irradiation examination (PIE) results to evaluate BISON fuel performance modeling. Implementation of new fuel performance models into a generic BISON input file coupled with IMIS and FIPD yielded results with a better representation of physics than the initial evaluation of the integrated framework. Cladding profilometry, FGR, and fuel axial swelling were found to be in good agreement with PIE measurements for most of the pins simulated. The chosen mechanical contact solver was found to significantly impact the axial fuel swelling and cladding strain predictions when used in conjunction with the U-Pu-Zr hot-pressing model since it bound the fuel to prevent further swelling and increased hydrostatic stresses. This work suggests that fuel performance modeling in BISON under steady-state conditions represents the PIE data well and should be reassessed when new PIE data become available in IMIS and FIPD databases and when improved physical models to better capture fuel performance are added to BISON.
Paaren, Kyle M.↗
Development of Innovative Non-Concentrated Alloy (NCA) Fuel Cladding for Advanced Nuclear Reactors
As the nuclear energy sector advances toward next-generation reactors, the need for high-performance fuel cladding materials has become increasingly urgent. Traditional alloys like zirconium and stainless steel are reaching their performance limits under higher temperatures, more corrosive coolants, and extended irradiation. This report presents the development of a new class of fuel cladding materials based on Non-Concentrated Alloys (NCAs) ? multi-element systems designed to deliver enhanced mechanical strength, corrosion resistance, and radiation tolerance. Through a combination of computational modeling (Computer Coupling of Phase Diagrams and Thermochemistry: CALPHAD), simulation-guided alloy selection, and experimental fabrication via arc melting and spark plasma sintering (SPS), three strategic alloy design paths were explored: (1) FeCrAl-based NCAs, (2) refractory-lean neutron-efficient alloys, and (3) equimolar high-entropy compositions. Microstructural analysis confirmed the formation of stable body-centered cubic BCC_A2 phases, while mechanical testing demonstrated hardness values significantly exceeding those of conventional cladding materials. The results highlight the tunability of NCA systems and their potential for balancing strength and ductility ? a critical consideration for in-reactor performance. Looking forward, future work will focus on thermomechanical optimization, CALPHAD refinement, and benchmarking against industry standards to enable scalable deployment. This work not only advances the science of nuclear materials but also supports broader goals in nuclear safety, performance, and nuclear energy innovation.
36 - MATERIALS SCIENCE↗
Simultaneous isotopic analysis of fission product Sr, Mo, and Ru in spent nuclear fuel particles by resonance ionization mass spectrometry
Abstract Fission product Sr, Mo, and Ru isotopes in six 10-μm particles of spent fuel from a pressurized water reactor were analyzed by resonance ionization mass spectrometry (RIMS) and evaluated for utility in nuclear material characterization. Previous measurements on these same samples showed widely varying U, Pu, and Am isotopic compositions owing to the samples’ differing irradiation environments within the reactor. This is also seen in Mo and Ru isotopes, which have the added complication of exsolution from the UO 2 fuel matrix. This variability is a hindrance to interpreting data from a collection of particles with incomplete provenance since it is not always possible to assign particles to the same batch of fuel based on isotopic analyses alone. In contrast, the measured 90 Sr/ 88 Sr ratios were indistinguishable across all samples. Strontium isotopic analysis can therefore be used to connect samples with otherwise disparate isotopic compositions, allowing them to be grouped appropriately for interpretation. Strontium isotopic analysis also provides a robust chronometer for determining the time since fuel irradiation. Because of the very high sensitivity of RIMS, only a small fraction of material in each of the 10 μm samples was consumed, leaving the vast majority still available for other analyses.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Migration energy barriers and diffusion anisotropy of point defects on tungsten surfaces
We report tungsten is one of the most promising candidates for plasma-facing materials in future fusion devices, owing to its high performance under extreme irradiation conditions. However, irradiation-induced surface morphology varies significantly, depending on the irradiation type, fluence, and flux. Therefore, it is critical to examine the dynamics of point defects on tungsten (W) surfaces to understand how irradiation affects surface morphology. In this study, we employ the Self-Evolving Atomistic Kinetic Monte Carlo (SEAKMC) method to search for potential migration paths of point defects on W (1 0 0), (1 1 0) and (1 1 1) surfaces. The first-principles calculations are then used to accurately determine the migration energy barriers. The obtained paths and barriers are incorporated into a kinetic Monte Carlo (KMC) model to determine trajectories and diffusivities, which are described by diffusion tensors to demonstrate their anisotropic features. Multiple diffusion mechanisms have been identified on different surfaces, with various anisotropy factors. Particularly, point defects on the W (1 1 0) surface have the highest diffusivities, with anisotropy factors independent of temperatures. In comparison, the anisotropy of the W (1 0 0) surface decreases as temperature increases, while the W (1 1 1) surface is isotropic for point defects. This study provides insights into defect transport properties on different surfaces, which are essential for understanding the early stages of irradiation-induced microstructural evolutions and surface morphology of tungsten.
36 MATERIALS SCIENCE↗
The effect of phosphorus on precipitation in irradiated reactor pressure vessel (RPV) steels
Embrittlement of light water reactor pressure vessels (RPV) steels by fast neutron irradiation may limit extended nuclear plant life. Embrittlement, which is manifested as increases in various indexes of a ductile to brittle transition temperatures (ΔT), is primarily due to hardening by nanoscale precipitates containing Cu, Ni, Mn, and Si, which form under irradiation. In addition to these elements, P has also been found to play a role in embrittlement. While only slightly enriched in the precipitates, hardening and embrittlement increase with trace P concentrations in low-Cu steels. Here, we characterize the individual and synergistic irradiation precipitation and hardening mechanisms in a series of RPV steels containing no to low-Cu and with systematic variations in Ni and P. In this study, the steels were irradiated to a fluence of ~ 1.38 x 10 20 n/cm 2 at ~ 292 °C in the UCSB ATR-2 experiment. In Cu-free medium and high-Ni RPV steels, atom probe tomography shows that P and Ni promotes precipitation of P-Mn-Si-Ni and Mn-Si-Ni precipitates, respectively. The precipitate microstructure correlates with the observed irradiation hardening.
36 MATERIALS SCIENCE↗
Fast neutron irradiation capability in existing thermal test reactors
In today’s nuclear industry, momentum towards the design, licensing, and construction of advanced nuclear demonstration plants, including fast reactors, is at a remarkably high level. However, there are currently no dedicated fast spectrum irradiation test facilities in the United States to support the development of fast spectrum technologies. As a result, a unique situation is developing where most of these plants will likely be designed by leveraging historic nuclear material technologies, but where the further optimization and advancement is impeded by the lack of fast neutron irradiation test facilities. While these circumstances present a challenge, there are some near-term opportunities that, if seized, can still help develop advanced fast reactor materials to a meaningful level of readiness to support future commercial fast reactors. Here, in this paper, we assess the feasibility of using thermal neutron filtering materials in existing experiment positions in the Advanced Test Reactor (ATR) at Idaho National Laboratory and the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory to simulate fast reactor test environments for nonfuel test specimens. Items investigated include the incident neutron flux (both fast and thermal), the total neutron fluence and cumulative atom displacements, helium production rate due to thermal neutron capture in nickel, and the potential impact that the thermal neutron filter material has on the cycle length of a given reactor. It is concluded that while HFIR provides the highest fast flux of all the options investigated, it is limited in the amount of thermal neutron filtering material that can be introduced into an experiment position without significantly affecting the operation of the reactor. Irradiation in Outboard-A positions in the ATR was found to be the most realistic near-term experiment avenue due to having ample space for several capsules in a moderately fast flux.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Cryogenic tracer irradiation facility at the university of Texas at Austin
We report on a cryogenic gas irradiation facility in the 1.1 MW TRIGA reactor at The University of Texas at Austin. The system was designed to produce radioactive xenon and argon for environmental studies, and it can be applied to produce other gaseous radiotracers. The system design includes modeling accident scenarios to ensure operation does not risk damage to the reactor and minimizes risk of release of radioactive material. In conclusion, the SCALE code was used to model a 30-day irradiation of one liter of Xe-126 and the predicted activity is 0.9 Ci of Xe-127 at irradiation end.
Ar-37↗
Thermal Contributions to the Degradation of Ground-Laboratory- and Space-Irradiated Teflon Investigated
The Hubble Space Telescope (HST) is covered with two primary types of thermal control materials, radiators and multilayer insulation blankets, which passively control temperatures during orbit. Both of these thermal control materials utilize back-surface metalized Teflon FEP (DuPont, fluorinated ethylene propylene) as the exterior (spacefacing) layer because of its excellent optical properties (low solar absorptance and high thermal emittance). The aluminized-FEP (Al-FEP) outermost layer of the multilayer insulation blankets on the HST has become embrittled while in space, resulting in severe on-orbit cracking (see the photographs). During the second servicing mission, an extremely embrittled piece of Al-FEP was retrieved that had curled, exposing the backsurface aluminum to space (see the photograph on the right). Because the aluminum surface has a lower thermal emittance than the FEP, this curled piece reached 200 C during orbit, 150 C higher than the nominal temperature extreme. To better understand the effect of temperature on the rate of degradation, and on the mechanism of degradation, of this insulation material in the low-Earth-orbit environment, researchers at the NASA Glenn Research Center conducted experiments to determine the effect of heating on the degradation of FEP that has been irradiated in a ground laboratory facility or in space on the HST. For this study, Teflon FEP retrieved from the HST during the third servicing mission after 9.7 years of space exposure was provided to Glenn by the NASA Goddard Space Flight Center.
deGroh, Kim K.↗
Irradiation response of innovatively engineered metastable TRIP high entropy alloy
For this work, properties and radiation responses of a metastable high alloy (HEA) exhibiting the transformation induced plasticity (TRIP) effect were studied. Innovative engineering used to manufacture this HEA has shown superior mechanical and corrosion properties in 3.5% NaCl than most advanced stainless steels. The microstructural evolution and corresponding mechanical response after irradiation have been evaluated using detailed transmission electron microscopy, and nanoindentation. The study shows a change in metastability of the alloy with irradiation via a recovery mechanism, where the irradiation-induced transformation is reversed by the temperature-induced transformation, thereby introducing the concept of self-healing, made possible due to the TRIP behavior of HEA.
36 MATERIALS SCIENCE↗
Development of Solid Isotope Harvesting Methods in Preparation for FRIB (Final Report)
In a collaborative effort between Hunter College (HC), Lawrence Livermore National Laboratory (LLNL), and Michigan State University (MSU), we aimed to investigate the recovery of 88 Zr produced as the secondary beam at the National Superconducting Cyclotron Laboratory (NSCL) from solid collectors via solid isotope harvesting. We developed separation schemes to isolate Zr from sub-gram-scale collectors composed of various materials including carbon, aluminum, copper, tungsten, and gold. The solid harvesting endstation was constructed and then used as part of a dedicated beamtime at the NSCL using 88 Zr beam. After irradiation, the 88 Zr was radiochemically separated from each of the collector materials and the recovery efficiencies from each was assessed. The results of this work provide a path forward for harvesting hydrolysable metals at FRIB as a complementary approach to aqueous isotope harvesting.
07 ISOTOPE AND RADIATION SOURCES↗
Io - A surface evaporite deposit
A model is suggested for Io's surface composition involving evaporite salt deposits, rich in sodium and sulfur. According to this model, these deposits were produced as a result of the migration of salt-saturated aqueous solutions to Io's surface from a warm or hot interior followed by loss of the water to space. This model satisfies cosmochemical constraints based on Io's initial composition, current density, and thermal history. Salt-rich assemblages are easily derivable from the leaching of carbonaceous chondritic material. The chemical and optical properties of such deposits, after modification by irradiation, can be used to explain Io's overall albedo and spectral reflectance, its dark reddish poles, and the observed sodium emission as well as or better than other currently suggested materials.
Fanale, F. P.↗
Origins of radiation-induced optical attenuation in neutron-irradiated single-crystal sapphire at elevated temperatures
Sapphire (α-Al 2 O 3 ) is a candidate fiber-optic sensor material for extreme temperature environments, potentially including those of nuclear reactors. However, its optical transmission under high-dose neutron irradiation is not well understood compared with that of conventional fused silica. This study examined dimensional changes, optical transmission, and irradiation-induced defects in neutron-irradiated α-Al 2 O 3 at temperatures of 298 °C to 688 °C and doses of 3.2 to 12 dpa. Although previous studies attributed radiation-induced attenuation (RIA) at the highest irradiation temperatures to increased optical scattering from radiation-induced voids, our findings indicate that scattering from neither voids nor dislocation loops can explain the measured attenuation. Instead, absorption due to aluminum vacancy centers appears more likely based on a comparison of the spectral features of the measured optical attenuation with previous literature. Significant c-axis swelling (5.51 % ± 0.83 %) was observed in the 12 dpa, 592 °C irradiated sample, much higher than earlier measurements, suggesting temperature sensor drift of 543 °C to 1,140 °C. Void patterning was predominantly observed along the a-axis, differing from previous studies on polycrystalline samples, which showed c-axis patterning. Dislocation loops evolved into network dislocations with increasing temperature and dose; voids formed within these structures, showing no size or density changes, indicating an atypical growth mechanism.
Neutron irradiation↗
Hydrophobic Surfaces of Spacecraft Components Enhance the Aggregation of Microorganisms and May Lead to Higher Survival Rates on Mars
In order to minimize the forward contamination of Mars, spacecraft are assembled under clean-room conditions that often require several procedures to clean and sterilize components. Surface characteristics of spacecraft materials may contribute to microbial survival by protecting spores from sterilizing agents, including UV irradiation on the surface of Mars. The primary objective of this study was to evaluate the effects of surface characteristics of several spacecraft materials on the survival of Bacillus subtilis spores under simulated Martian conditions.
A C Schuerger↗