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At least 1,027 records · Page 57
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

A sensitivity analysis to predict the neutronics behavior of samples irradiated in the VTR rabbit system

We report a low-order neutronics model is developed to carry out hundreds of simulations efficiently and investigate the neutronics behavior of samples being irradiated in a test reactor setting under different geometrical constraints. The low-order model allowed for simulations that yield the expected neutronics behavior of any irradiated sample in any environment and allows for the calculation of highly accurate spatially averaged statistics and idealized spatial distributions in the neutron flux. Several benchmarks are performed to evaluate the performance and limitations of the low-order model revealing many important findings. The low-order model predicted the LHGR in the EBR-II driver fuel to within 2.34% by only simulating the fuel rod by itself, which served as a validation for the model. Sensitivity studies investigated 3% enriched UO 2 and U-10Zr being irradiated in the Versatile Test Reactor rabbit system. The analyses investigated a range of combinations of 15 radii and 5 heights for each sample in the rabbit system. Similar data sets are also provided for irradiations in the Advanced Test Reactor’s B-10 irradiation position, which is a thermal neutron spectrum environment. Generalized fits and fit coefficients are obtained for sample heating, reaction rate densities, and local multiplication rate characteristics, allowing the predictions of the neutronics behavior of the samples based on their geometrical constraints. The analyses and fits laid the groundwork for developing a user-end Multiphysics analysis framework to assist and accelerate irradiation experiment design and optimization.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

MURR LEU structural and thermal hydraulics analyses: Part II – Impacts of irradiation thermo-mechanical behavior on thermal hydraulics safety analyses

A series of structural analyses have been performed to support the conversion of the University of Missouri Research Reactor (MURR) from the use of highly enriched uranium (HEU; ≥20 wt% U-235) to low-enriched uranium (LEU; <20 wt% U-235) fuel. The irradiation thermo-mechanical analysis evaluated the effects of fuel swelling, irradiation creep, thermal expansion, as well as thermal resistance from the oxide layer growth for the MURR LEU element in prototypic thermal and irradiation conditions as presented in Part I of this article. Overall, this irradiation thermo-mechanical analysis predicts smaller gap thickness reductions in previously limiting regions, and larger reductions in the middle of the outermost end channels where power density is not typically a maximum. Due to substantial differences between the channel gap reductions assumed for the previous safety analyses and those predicted by the irradiation thermo-mechanical analysis, a need to evaluate their impact on the thermal hydraulics safety analyses arose. This article presents the results from the steady-state safety analysis for normal operation as well as the two most limiting accident scenarios. The calculation models were revised in order to account for the spatial and temporal variation of the channel gap thicknesses. The results show that sufficient safety margins are still maintained for normal operation as well as during the postulated accident transients. This work provides a methodology of incorporating the irradiation thermo-mechanical behavior of plate-type fuel into the thermal hydraulics safety analyses.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Development of mesopores in superfine grain graphite neutron-irradiated at high fluence

Microstructural changes induced by neutron irradiation of superfine grain graphite G347A (Tokai Carbon, Japan) were examined by nitrogen adsorption at 77 K and by three microscopy techniques (SEM, TEM and FIB-SEM tomography). The specimens were irradiated at doses of up to 30 dpa, covering stages before and after the turnaround fluence at three temperatures (300, 450, 750 °C) of their irradiation envelope. The initial graphite densification at low fluences did not produce any detectable effect in the pore size range (<350 nm) measured by gas adsorption. However, graphite irradiated at high fluences, after turnaround, showed severe structural changes. At all three temperatures and high irradiation fluences, gas adsorption revealed significant increase of the volume of narrow mesopores (<5–20 nm) and up to five times increase of BET surface area, both in linear relationship with the relative volume expansion. Analysis of microscopy images showed multiplication of fine macropores (>50 nm) at high irradiation fluences and more structural changes on multiple scales, from nanometers to microns. As a result, this work demonstrates the unique ability of gas adsorption techniques to analyze open pores with sizes between sub-nanometer and sub-micron in bulk nuclear graphite, with supporting microscopy results.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Effects of neutron irradiation on hydrogen isotope permeation on W-Re

Hydrogen and deuterium permeation behavior for neutron damaged W and W-3 %Re was evaluated using a Plasma Driven Permeation device. For neutron damaged W-3 %Re at 773 K and 1473 K, precipitation of Re and Os was observed by atom probe tomography. Neutron irradiation introduced defect damage throughout the sample and enhanced H and D diffusion toward downstream side. The existence of irradiation damage by neutron irradiation was found to increase hydrogen isotope permeation. Moreover, the existence of Re has suppressed the formation of irradiation defects for hydrogen isotope trapping sites. Further, the recovery of irradiation defects by increasing the irradiation temperature led to the structural changes due to the diffusion of Re and Os, which also affected the hydrogen isotope permeation behavior.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Helium effects and bubbles formation in irradiated Ti 3 SiC 2

Ti 3 SiC 2 is a potential structural material for nuclear reactor applications. However, He irradiation effects in this material are not well understood, especially at high temperatures. Here, we compare the effects of He irradiation in Ti 3 SiC 2 at room temperature (RT) and at 750 °C. Irradiation at 750 °C was found to lead to extremely elongated He bubbles that are concentrated in the nano-laminate layers of Ti 3 SiC 2 , whereas the overall crystal structure of the material remained intact. In contrast, at RT, the layered structure was significantly damaged and highly disordered after irradiation. Our study reveals that at elevated temperatures, the unique structure of Ti 3 SiC 2 can accommodate large amounts of He atoms in the nano-laminate layer, without compromising the structural stability of the material. The structure and the mechanical tests results show that the irradiation induced swelling and hardening at 750 °C are much smaller than those at RT. Overall, these results indicate that Ti 3 SiC 2 has an excellent resistance to accumulation of radiation-induced He impurities and that it has a considerable tolerance to irradiation-induced degradation of mechanical properties at high temperatures.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Evidence of lithium mobility under neutron irradiation

Understanding the evolution of intermetallic materials in high-radiation environments is of great importance for fusion science and national security. Tritium ( 3 H) and lithium ( 6 Li, 7 Li) transport within neutron irradiated claddings coated with aluminide (FeAl3) has been investigated using state-of-the-art multimodal imaging. Specifically, scanning electron microscopy – focused ion beam (SEM-FIB) was used to prepare irradiated coating lift-out samples for follow-on microanalysis. Here, scanning transmission electron microscopy (STEM) was used to acquire atomic-scale information on the carbonaceous structure and elemental mapping. Atomic force microscopy (AFM) was used to determine lift-out dimensions nondestructively. Time-of-flight secondary ion mass spectrometry (ToF-SIMS) spectral and depth profiling show unexpected lithium isotopic distributions in the irradiated cladding, raising an important evidence of possible light isotope mobility within the cladding under elevated temperature and irradiation. ToF-SIMS three-dimensional chemical mappings of mid- and bottom-cladding coatings show light isotopic (e.g., 3 H, 6 Li, 7 Li) distributions in the irradiated coating and give new insights into the fundamental mechanism related to transport mechanisms within the cladding. Multimodal imaging is power to link microstructure, chemistry, and nanoscale defects that impact reliability of these materials. Furthermore, chemical mapping offers observations of the microstructural evolution due to irradiation and provides insights into unexpected material transport under extreme conditions.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Structural damage response of lanthanum and yttrium aluminate crystals to nuclear collisions and electronic excitation: Threshold assessment of irradiation damage

A comparative analysis is performed on the structural damage response and associated mechanisms in lanthanum aluminate and yttrium aluminate crystals under various irradiation conditions by a combination of experimental and theoretical approaches. Under low-energy Au+ irradiation, the damage accumulation curve shows a higher damage rate for LaAlO 3 crystals than YAlO 3 crystals. The relatively low irradiation tolerance of LaAlO 3 to the action of nuclear collisions is ascribed to the large amorphization cross-section and effective cross-section for defect-stimulated amorphization. Under swift Ar 12+ , Ni 19+ and Kr 17+ irradiation with different ion energies and velocities, the formed highly-disordered/amorphous latent tracks with different morphologies in pristine and predamaged crystals are discussed, and the corresponding electronic energy loss and lattice temperature thresholds are quantitatively determined. Compared to YAlO 3 , LaAlO 3 exhibits lower sensitivity and higher damage tolerance to the electronic energy loss process, attributing to its relatively high recrystallization efficiency during the rapid quenching process. Furthermore, the introduction of lattice defects into LaAlO 3 and YAlO 3 crystals considerably enhances the sensitivity and intensity of thermal spike response to the electronic energy loss, and the induced effective modification of track morphologies demonstrates the synergistic effect between the electronic energy loss and pre-existing defects created by nuclear collisions. In this case, even under the action of electronic energy loss below the threshold, the lattice temperature in the nuclear-collision damaged crystalline system could still meet the criterion for track production. The irradiation energy deposited to atoms and induced lattice temperature evolution discussed in this work provide a deeper insight into the complex processes involved in irradiation-induced latent track behaviors.

36 MATERIALS SCIENCE↗
DOE OSTI2015

High Dose Neutron Irradiation of Hi-Nicalon Type S Silicon Carbide Composites, Part 2. Mechanical and Physical Properties

Nuclear-grade silicon carbide (SiC) composite material was examined for mechanical and thermophysical properties following high-dose neutron irradiation in the High Flux Isotope Reactor at a temperature range of 573–1073 K. Likewise, the material was chemical vapor-infiltrated SiC-matrix composite with a two-dimensional satin weave Hi-Nicalon Type S SiC fiber reinforcement and a multilayered pyrocarbon/SiC interphase. Moderate (1073 K) to very severe (573 K) degradation in mechanical properties was found after irradiation to >70 dpa, whereas no evidence was found for progressive evolution in swelling and thermal conductivity. The swelling was found to recover upon annealing beyond the irradiation temperature, indicating the irradiation temperature, but only to a limited extent. Moreover, the observed strength degradation is attributed primarily to fiber damage for all irradiation temperatures, particularly a combination of severe fiber degradation and likely interphase damage at relatively low irradiation temperatures.

36 MATERIALS SCIENCE↗
DOE OSTI2015

Irradiation response of commercial, high-Tc superconducting tapes: Electromagnetic transport properties

Effects of low dose irradiation on the electrical transport current properties of commercially available high-temperature superconducting, coated-conductor tapes were investigated, in view of potential applications in the irradiative environment of fusion reactors. Three different tapes, each with unique as-grown flux-pinning structures, were irradiated with Au and Ni ions at energies that provide a range of damage effects, with accumulated damage levels near that expected for conductors in a fusion reactor environment. Measurements using transport current determined the pre- and post-irradiation resistivity, critical current density, and pinning force density, yielding critical temperatures, irreversibility lines, and inferred vortex creep rates. Results show that at the irradiation damage levels tested, any detriment to as-grown pre-irradiation properties is modest; indeed in one case already-superior pinning forces are enhanced, leading to higher critical currents.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Radiation tolerance of neutron-irradiated model Fe-Cr-Al alloys

The Fe Cr Al alloy system has the potential to form an important class of enhanced accident-tolerant cladding materials in the nuclear power industry owing to the alloy system's higher oxidation resistance in high-temperature steam environments compared with traditional zirconium-based alloys. However, radiation tolerance of Fe Cr Al alloys has not been fully established. In this study, a series of Fe Cr Al alloys with 10 18 wt % Cr and 2.9 4.9 wt % Al were neutron irradiated at 382 C to 1.8 dpa to investigate the irradiation-induced microstructural and mechanical property evolution as a function of alloy composition. Dislocation loops with Burgers vector of a/2 111 and a 100 were detected and quantified. Results indicate precipitation of Cr-rich is primarily dependent on the bulk chromium composition. Mechanical testing of sub-size-irradiated tensile specimens indicates the hardening response seen after irradiation is dependent on the bulk chromium composition. Furthermore, a structure property relationship was developed; it indicated that the change in yield strength after irradiation is caused by the formation of these radiation-induced defects and is dominated by the large number density of Cr-rich α' precipitates at sufficiently high chromium contents after irradiation.

36 MATERIALS SCIENCE↗
DOE OSTI2015

Ion irradiation induced defect evolution in Ni and Ni-based FCC equiatomic binary alloys

In order to explore the chemical effects on radiation response of alloys with multi-principal elements, defect evolution under Au ion irradiation was investigated in the elemental Ni, equiatomic NiCo and NiFe alloys. Single crystals were successfully grown in an optical floating zone furnace and their (100) surfaces were irradiated with 3 MeV Au ions at fluences ranging from 1 × 10 13 to 5 × 10 15 ions cm –2 at room temperature. The irradiation-induced defect evolution was analyzed by using ion channeling technique. Experiment shows that NiFe is more irradiation-resistant than NiCo and pure Ni at low fluences. With continuously increasing the ion fluences, damage level is eventually saturated for all materials but at different dose levels. The saturation level in pure Ni appears at relatively lower irradiation fluence than the alloys, suggesting that damage accumulation slows down in the alloys. Here, under high-fluence irradiations, pure Ni has wider damage ranges than the alloys, indicating that defects in pure Ni have high mobility.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Defect evolution in single crystalline tungsten following low temperature and low dose neutron irradiation

The tungsten plasma-facing components of fusion reactors will experience an extreme environment including high temperature, intense particle fluxes of gas atoms, high-energy neutron irradiation, and significant cyclic stress loading. Irradiation-induced defect accumulation resulting in severe thermo-mechanical property degradation is expected. For this reason, and because of the lack of relevant fusion neutron sources, the fundamentals of tungsten radiation damage must be understood through coordinated mixed-spectrum fission reactor irradiation experiments and modeling. In this study, high-purity (110) single-crystal tungsten was examined by positron annihilation spectroscopy and transmission electron microscopy following low-temperature (~90 °C) and low-dose (0.006 and 0.03 dpa) mixed-spectrum neutron irradiation and subsequent isochronal annealing at 400, 500, 650, 800, 1000, 1150, and 1300 °C. The results provide insights into microstructural and defect evolution, thus identifying the mechanisms of different annealing behavior. Following 1 h annealing, ex situ characterization of vacancy defects using positron lifetime spectroscopy and coincidence Doppler broadening was performed. The vacancy cluster size distributions indicated intense vacancy clustering at 400 °C with significant damage recovery around 1000 °C. Coincidence Doppler broadening measurements confirm the trend of the vacancy defect evolution, and the S–W plots indicate that only a single type of vacancy cluster is present. Furthermore, transmission electron microscopy observations at selected annealing conditions provide supplemental information on dislocation loop populations and visible void formation. This microstructural information is consistent with the measured irradiation-induced hardening at each annealing stage. This provides insight into tungsten hardening and embrittlement due to irradiation-induced matrix defects.

Hu, Xunxiang↗
DOE OSTI2016

Neutron-irradiation creep of silicon carbide materials beyond the initial transient

Irradiation creep beyond the transient regime was investigated for various silicon carbide (SiC) materials. Here, the materials examined included polycrystalline or monocrystalline high-purity SiC, nanopowder sintered SiC, highly crystalline and near-stoichiometric SiC fibers (including Hi-Nicalon Type S, Tyranno SA3, isotopically-controlled Sylramic and Sylramic-iBN fibers), and a Tyranno SA3 fiber–reinforced SiC matrix composite fabricated through a nano-infiltration transient eutectic phase process. Neutron irradiation experiments for bend stress relaxation tests were conducted at irradiation temperatures ranging from 430 to 1180 °C up to 30 dpa with initial bend stresses of up to ~1 GPa for the fibers and ~300 MPa for the other materials. Initial bend stress in the specimens continued to decrease from 1 to 30 dpa. Analysis revealed that (1) the stress exponent of irradiation creep above 1 dpa is approximately unity, (2) the stress normalized creep rate is ~1 × 10 –7 [dpa –1 MPa –1 ] at 430–750 °C for the range of 1–30 dpa for most polycrystalline SiC materials, and (3) the effects on irradiation creep of initial microstructures—such as grain boundary, crystal orientation, and secondary phases—increase with increasing irradiation temperature.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Microstructural characterization of an irradiated RERTR-6 U-7Mo/AA4043 alloy dispersion fuel plate specimen blister-tested to a final temperature of 500°C

The Material Management and Minimization (M3) Reactor Conversion Program, in the past called the Reduced Enrichment for Research and Test Reactor (RERTR) Program, is developing low-enriched uranium (LEU) fuels for application in research reactors. U–Mo alloy dispersion fuel is one type being developed. Blister testing has been performed on different fuel plate samples to determine the margin to failure for fuel plates irradiated to different fission densities. Microstructural characterization was performed using scanning electron microscopy and transmission electron microscopy on a sample taken from a U-7Mo/AA4043 matrix dispersion fuel plate irradiated in the RERTR-6 experiment that was blister-tested up to a final temperature of 500°C. The results indicated that two types of grain/cell boundaries were observed in the U- 7Mo fuel particles, one with a relatively low Mo content and fission gas bubbles and a second type enriched in Si, due to interdiffusion from the Si-containing matrix, with little evidence of fission gas bubbles. With respect to the behavior of the major fission gas Xe, a significant amount of the Xe was still observed within the U-7Mo fuel particle, along with microns into the AA4043 matrix. For the fuel/matrix interaction layers that form during fabrication and then grow during irradiation, they change from the as-irradiated amorphous structure to one that is crystalline after blister testing. In the AA4043 matrix, the original Si-rich precipitates, which are typically observed in as-irradiated U-Mo dispersion fuel, get consumed due to interdiffusion with the U-7Mo fuel particles during the blister test. Lastly, the fission gas bubbles that were originally around 2 nm in diameter and resided on a fission gas superlattice in the intragranular regions of as-irradiated U-7Mo fuel grew in size (up to ~20 nm diameter) during blister testing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Microstructural evolution of pure tungsten neutron irradiated with a mixed energy spectrum

Here, microstructures of single-crystal bulk tungsten (W) and polycrystalline W foil with a strong grain texture were investigated using transmission electron microscopy following neutron irradiation at ~90–800 °C to 0.03–4.6 displacements per atom (dpa) in the High Flux Isotope Reactor with a mixed energy spectrum. The dominant irradiation defects were dislocation loops and small clusters at ~90 °C. Additional voids were formed in W irradiated at above 460 °C. Voids and precipitates involving transmutation rhenium and osmium were the dominant defects at more than ~1 dpa. We found a new phenomenon of microstructural evolution in irradiated polycrystalline W: Re- and Os-rich precipitation along grain boundaries. Comparison of results between this study and previous studies using different irradiation facilities revealed that the microstructural evolution of pure W is highly dependent on the neutron energy spectrum in addition to the irradiation temperature and dose.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Postirradiation examination results of several metallic fuel alloys and forms from low burnup AFC irradiations

Several different U based metallic nuclear fuel alloys and forms were irradiated as part of the Advanced Fuels Campaign (AFC) series of irradiation experiments. Results from postirradiation examinations of the AFC-3A and AFC-3B are presented here. In this irradiation, U alloyed with Mo, Zr, and Zr with Pd were investigated. Different geometric forms (solid, annular), smear densities and bonding media (sodium, helium) were also irradiated and evaluated. All fuel forms were irradiated in ferritic-martensitic HT-9 cladding. Examination of the irradiated fuel included neutron radiography, evaluation of the distribution of gamma-ray emitting radionuclides, dimensional inspection, fission gas release analysis, optical microscopy, and chemical analysis. No in-pile failure was observed with this fuel. Neutron radiography, gamma spectrometry, and dimensional inspections indicated the fuel largely performed acceptably. Optical microscopy revealed significant fuel-cladding interaction between the Mo based alloys and the cladding. The Zr based alloys generally performed better although some forms revealed behaviors that present specific fabrication challenges such as tighter fabrication tolerances that will need to be better controlled to match historical U-Zr fuel performance. Furthermore, the addition of Pd as an additive to prevent fuel cladding chemical interaction (FCCI) also appears to present some challenges to fuel performance.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

200 MeV proton irradiation of the oxide-dispersion-strengthened copper alloy (GlidCop-Al15)

In search to identify extremely stable materials to perform the collimation function of the 7-TeV proton beam halo at the Large Hadron Collider (LHC), oxide dispersion strengthened (DS) copper alloys were explored. These internally oxidized copper alloys known as GlidCop are also leading candidates for high heat flux applications in fusion reactors and as divertor and first wall structure in ITER. These considerations have led to an extensive body of research on neutron-induced changes in the microstructure and physio-mechanical properties. This study focused primarily on the damage induced by 200 MeV protons on as wrought and cold-worked GlidCop Al15 DS copper ally selected as a candidate material of the LHC beam collimation structure to damage levels up to ~10 displacements-per-atom (dpa) and irradiation temperatures up to ~600 °C. For reference, low dose spallation neutron damage at sub-zero irradiation temperature was included in the study. Proton irradiation effects on dimensional stability, mechanical behavior, electrical resistivity and X-ray diffraction-based microstructural changes are presented in the paper. GlidCop Al15 exhibited excellent thermal stability and superior to pure Cu resistance to conductivity loss. Proton irradiation-induced hardening and embrittlement at distinct irradiation temperature regimes were shown to follow similar trends to what was reported under neutron exposure to even higher dpa damage. At T irr ~600 ± 20 °C the cold-worked DS copper alloy exhibited softening accompanied by elongation reduction while at T irr ~200 ± 10 °C the as-wrought experienced hardening with elongation increase. Furthermore, high resolution X-ray diffraction revealed that even low-dose neutron irradiation has a strong influence on the size and distribution of dispersed Al 2 O 3 particles.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Post-irradiation examinations of UO 2 composites as part of the Accident Tolerant Fuels Campaign

Enhanced thermal conductivity uranium dioxide composites containing silicon carbide (UO 2 -SiC) and diamond (UO 2 -diamond) have been irradiated to low burnup. The conditions of this irradiation test and subsequent postirradiation examinations are discussed. These irradiations evaluate fuel microstructure and potential fuel cladding interaction under representative light water reactor power conditions. Both non-destructive and destructive techniques have been used to evaluate fuel integrity, fission gas release, fission product distribution, burnup, fuel swelling and cladding strain. Examination of the UO 2 -SiC pellets revealed enhanced cracking when compared to a UO 2 rodlet also irradiated under similar conditions. Additionally instability of the SiC whiskers in the uranium dioxide matrix was observed in the pellet central region, where the local temperatures exceeded 1300°C. The microstructure of the diamond-added UO 2 was severely disrupted during irradiation, resulting in local migration of cesium along the fuel stack and increased fission gas release when compared with the expected release from the Vitanza curve at corresponding values of burnup and irradiation temperature. The postirradiation examination results cast doubt on the suitability of these additives to improve UO 2 fuel performance in a way that would lead to enhanced accident tolerance.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗