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At least 127 records · Page 7
All sourcesNASA NTRSDOE OSTI
DOE OSTI2014

Characterization of the neutron irradiation system for use in the Low-Dose-Rate Irradiation Facility at Sandia National Laboratories.

The objective of this work was to characterize the neutron irradiation system consisting of americium-241 beryllium (241AmBe) neutron sources placed in a polyethylene shielding for use at Sandia National Laboratories (SNL) Low Dose Rate Irradiation Facility (LDRIF). With a total activity of 0.3 TBq (9 Ci), the source consisted of three recycled 241AmBe sources of different activities that had been combined into a single source. The source in its polyethylene shielding will be used in neutron irradiation testing of components. The characterization of the source-shielding system was necessary to evaluate the radiation environment for future experiments. Characterization of the source was also necessary because the documentation for the three component sources and their relative alignment within the Special Form Capsule (SFC) was inadequate. The system consisting of the source and shielding was modeled using Monte Carlo N-Particle transport code (MCNP). The model was validated by benchmarking it against measurements using multiple techniques. To characterize the radiation fields over the full spatial geometry of the irradiation system, it was necessary to use a number of instruments of varying sensitivities. First, the computed photon radiography assisted in determining orientation of the component sources. With the capsule properly oriented inside the shielding, the neutron spectra were measured using a variety of techniques. A N-probe Microspec and a neutron Bubble Dosimeter Spectrometer (BDS) set were used to characterize the neutron spectra/field in several locations. In the third technique, neutron foil activation was used to ascertain the neutron spectra. A high purity germanium (HPGe) detector was used to characterize the photon spectrum. The experimentally measured spectra and the MCNP results compared well. Once the MCNP model was validated to an adequate level of confidence, parametric analyses was performed on the model to optimize for potential experimental configurations and neutron spectra for component irradiation. The final product of this work is a MCNP model validated by measurements, an overall understanding of neutron irradiation system including photon/neutron transport and effective dose rates throughout the system, and possible experimental configurations for future irradiation of components.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2026

Summary of the Initial Post-Irradiation Characterization of HFIR-Irradiated Low-N and High-N HT-9 Steel

Reference cladding systems for sodium fast reactors are based on the historical steel, HT-9. HT-9 is a Fe12Cr ferritic/martensitic steel with additions of Mo, W, V, and other minor elements and demonstrates low irradiation swelling and adequate mechanical properties. Extensive irradiation literature exists on the use of HT-9 as cladding for metal fuel, primarily irradiation on the U-Zr/HT-9 system from the Experimental Breeder Reactor-II (EBR-II) and Fast Flux Test Facility (FFTF) sodium fast reactor, and as a structural material from experiments in the FFTF. The large amount of historical data makes the U-Zr/HT-9 system the reference fuel specification for many nuclear reactor vendors that seek to license modern sodium-cooled fast reactors in the United States. However, it is yet unclear how variations in impurity content within HT-9 fundamentally affect irradiation performance at various irradiation temperatures. Recent work suggests that impurity content may noticeably alter the production of helium through nuclear transmutation. For these reasons, High-Flux Isotope Reactor (HFIR) irradiation of HT-9 steels with known variations in the impurity content is particularly timely to generate data to enable more accurate refinement of the chemical specification for nuclear-grade HT-9 material. This report summarizes the initial transmission electron microscopy characterization of HFIR-irradiated HT-9 steels following mechanical property measurements by the Advanced Fuels Campaign (AFC). This report includes qualitative results of the cavity, dislocation loop and cluster/precipitate microstructures as well as radiation-induced segregation. Quantitative results are being shared with partner institutions and will be included in more detail in a future report in FY2026.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1995

Comparison of the Changes in the Visible and Infrared Irradiance Observed by the SunPhotometers on EURECA to the UARS Total Solar and UV Irradiances

Solar irradiance in the near-UV (335 nm), visible (500 nm) and infrared (778 nm) spectral bands has been measured by the SunPhotometers developed at the World Radiation Center, Davos, Switzerland on board the European Retrievable Carrier between August 1992 and May 1993. Study of the variations in the visible and infrared irradiance is important for both solar and atmospheric physics. The purpose of this paper is to examine the temporal variations observed in the visible and infrared spectral bands after eliminating the trend in the data mainly related to instrument degradation. The effect of active regions in these spectral irradiances is clearly resolved. Variations in the visible and infrared irradiances are compared to total solar irradiance observed by the SOVA2 radiometer on the EURECA platform and by the ACRIMII radiometer on UARS as well as to UV observations of the UARS and NOAA9 satellites. The space-borne spectral irradiance observations are compared to the photometric sunspot deficit and CaII K irradiance measured at the San Fernando Observatory, California State University at Northridge in order to study the effect of active regions in detail.

EURECA UARS↗
DOE OSTI2022

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↗
DOE OSTI2016

Metallography and fuel cladding chemical interaction in fast flux test facility irradiated metallic U-10Zr MFF-3 and MFF-5 fuel pins

Abstract The Mechanistic Fuel Failure (MFF) series of metal fuel irradiations conducted in the Fast Flux Test Facility (FFTF) provides an important potential comparison between data generated in the Experimental Breeder Reactor (EBR-II) and that expected in a larger-scale fast reactor. The irradiations were the beginning tests to qualify U-10wt%Zr as a driver fuel for FFTF. The FFTF core, with a 91.4 cm tall fuel column and a chopped cosine neutron flux profile, operated with a peak cladding temperature at the top of the fuel column, but developed peak burnup at the centerline of the core. This places the peak fuel temperature midway between the core center and the top of fuel, lower in the fuel column than in previous EBR-II experiments that had a 32-cm height core. The MFF-3 and MFF-5 qualification assemblies operated in FFTF to >10 at% burnup, and performed very well with no cladding breaches. The MFF-3 assembly operated to 13.8 at% burnup with a peak inner cladding temperature of 643°C, and the MFF-5 assembly operated to 10.1 at% burnup with a peak inner cladding temperature of 651°C. Because of the very high operating temperatures for both the fuel and the cladding, data from the MFF assemblies are most comparable to the data obtained from the EBR-II X447 experiment, which experienced two pin breaches. The X447 breaches were strongly influenced by a large amount of fuel/cladding chemical interaction (FCCI). The MFF pins benefitted from different axial locations of high burnup and peak cladding temperature, which helped to reduce interdiffusion between rare earth fission products and stainless steel cladding. Post-irradiation examination evidence illustrates this advantage. Comparing other performance data of the long MFF pins to prior EBR-II test data, the MFF fuel inside the cladding grew less axially, and the gas release data did not reveal a definitive difference.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Establishing a process of irradiating small animal brain using a CyberKnife and a microCT scanner

Purpose: Establish and validate a process of accurately irradiating small animals using the CyberKnife G4 System (version 8.5) with treatment plans designed to irradiate a hemisphere of a mouse brain based on microCT scanner images. Methods: These experiments consisted of four parts: (1) building a mouse phantom for intensity modulated radiotherapy (IMRT) quality assurance (QA), (2) proving usability of a microCT for treatment planning, (3) fabricating a small animal positioning system for use with the CyberKnife's image guided radiotherapy (IGRT) system, and (4)in vivo verification of targeting accuracy. A set of solid water mouse phantoms was designed and fabricated, with radiochromic films (RCF) positioned in selected planes to measure delivered doses. After down-sampling for treatment planning compatibility, a CT image set of a phantom was imported into the CyberKnife treatment planning system—MultiPlan (ver. 3.5.2). A 0.5 cm diameter sphere was contoured within the phantom to represent a hemispherical section of a mouse brain. A nude mouse was scanned in an alpha cradle using a microCT scanner (cone-beam, 157 × 149 pixels slices, 0.2 mm longitudinal slice thickness). Based on the results of our positional accuracy study, a planning treatment volume (PTV) was created. A stereotactic body mold of the mouse was “printed” using a 3D printer laying UV curable acrylic plastic. Printer instructions were based on exported contours of the mouse's skin. Positional reproducibility in the mold was checked by measuring ten CT scans. To verify accurate dose delivery in vivo, six mice were irradiated in the mold with a 4 mm target contour and a 2 mm PTV margin to 3 Gy and sacrificed within 20 min to avoid DNA repair. The brain was sliced and stained for analysis. Results: For the IMRT QA using a set of phantoms, the planned dose (6 Gy to the calculation point) was compared to the delivered dose measured via film and analyzed using Gamma analysis (3% and 3 mm). A passing rate of 99% was measured in areas of above 40% of the prescription dose. The final inverse treatment plan was comprised of 43 beams ranging from 5 to 12.5 mm in diameter (2.5 mm size increments are available up to 15 mm in diameter collimation). Using the Xsight Spine Tracking module, the CyberKnife system could not reliably identify and track the tiny mouse spine; however, the CyberKnife system could identify and track the fiducial markers on the 3D mold.In vivo positional accuracy analysis using the 3D mold generated a mean error of 1.41 mm ± 0.73 mm when fiducial markers were used for position tracking. Analysis of the dissected brain confirmed the ability to target the correct brain volume. Conclusions: With the use of a stereotactic body mold with fiducial markers, microCT imaging, and resolution down-sampling, the CyberKnife system can successfully perform small-animal radiotherapy studies.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2014

Irradiation performance of U-Mo monolithic fuel

High-performance research reactors require fuel that operates at high specific power to high fission density, but at relatively low temperatures. Research reactor fuels are designed for efficient heat rejection, and are composed of assemblies of thin-plates clad in aluminum alloy. The development of low-enriched fuels to replace high-enriched fuels for these reactors requires a substantially increased uranium density in the fuel to offset the decrease in enrichment. Very few fuel phases have been identified that have the required combination of very-high uranium density and stable fuel behavior at high burnup. U-Mo alloys represent the best known tradeoff in these properties. Testing of aluminum matrix U-Mo aluminum matrix dispersion fuel revealed a pattern of breakaway swelling behavior at intermediate burnup, related to the formation of a molybdenum stabilized high aluminum intermetallic phase that forms during irradiation. In the case of monolithic fuel, this issue was addressed by eliminating, as much as possible, the interfacial area between U-Mo and aluminum. Based on scoping irradiation test data, a fuel plate system composed of solid U-10Mo fuel meat, a zirconium diffusion barrier, and Al6061 cladding was selected for development. Developmental testing of this fuel system indicates that it meets core criteria for fuel qualification, including stable and predictable swelling behavior, mechanical integrity to high burnup, and geometric stability. In addition, the fuel exhibits robust behavior during power-cooling mismatch events under irradiation at high power.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Current status of post-irradiation examination for the ATF-1 irradiation

A series of irradiations, referred to as the ATF series, is currently underway to evaluate the feasibility of different Accident Tolerant Fuels for commercial light water reactors. For initial screening, concepts consisting of an uranium bearing phase and cladding were irradiated in stainless steel capsules. These tests provide screening data among concepts to ensure viability of a fuel form going forward and can provide some initial data on fuel performance particularly information about fuel-cladding interaction and fuel behavior at operating temperatures and fission densities. Screening concepts are under irradiation or planned for irradiation at the Advanced Test Reactor of Idaho National Laboratory for a number of industrial partners (AREVA, GE, Westinghouse) and national laboratories (Los Alamos National Laboratory, Oak Ridge National Laboratory). Baseline PIE will be performed on all irradiated rodlets and will consist of visual examinations, neutron radiography, gamma spectrometry, dimensional inspections, fission gas analysis, optical microscopy, and chemical analysis. Additional evaluations may also be performed. The first capsules have arrived at the hot cell for initial post-irradiation examination. Results from initial examinations are presented along with example results from equivalent exams on similar fuel. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Post Irradiation Examination of SiC Tube Subjected to Simultaneous Irradiation and Radial High Heat Flux

Silicon carbide (SiC) based composite is considered among leading candidate materials for accident tolerant fuel (ATF) cladding in light water reactors (LWRs) because of its outstanding irradiation and oxidation resistance, as well as low neutron absorption. Under normal operation conditions, the synergism between neutron irradiation and a high radial heat flux though the cladding thickness is expected to produce significant stress. This stress is attributed to a significant through-thickness temperature gradient and the resulting differential swelling strain due the highly temperature-dependent point-defect swelling of SiC. Although modeling of the in-pile stress state of SiC cladding has been conducted using finite element analysis, the analysis is challenging because of the complex stress state that depends upon swelling, irradiation creep, and thermal conductivity, all of which are both temperature and dose-dependent. These integral models have not yet been verified by experiments. In order to experimentally validate the multi-physics thermo-mechanical models of ATF SiC cladding during operation, SiC tube specimens have been irradiated under a high radial heat flux. This paper reports preliminary results from the post-irradiation examination (PIE) in on-going project supported by the U.S. Department of Energy's Nuclear Science User Facility program. The objectives of the PIE are (1) determination of the irradiation temperature distribution within the tube specimens, and (2) evaluation of the state and magnitude of the stress within the tube specimens arising from the temperature gradient. This paper mainly presents findings for objective (1)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Raman study of amorphization in nanocrystalline 3C-SiC irradiated with C+ and He+ ions

This study examines C+ and He+ ion irradiation induced amorphization processes in 3C-SiC nanograins embedded in an amorphous SiC matrix. Raman spectroscopy and Rutherford backscattering spectrometry are used for damage characterization. SiC grains with an average size of either ~6 or ~20 nm were observed to be fully amorphized to a lower dose at room temperature compared to their monocrystalline counterpart under the identical irradiation condition. In addition to damage accumulation induced amorphization in the grains, preferential amorphization at the crystalline/amorphous interfaces could play a significant role. This interface-driven amorphization proceeds at comparable rates for C+ and He+ ion irradiations. The results may have an important implication for nanocrystalline SiC to be applied in advanced nuclear reactors.

Amorphization, Ion irradiation, Silicon carbide, N↗
DOE OSTI2018

Evolution of irradiation defects in Ti 2 AlC ceramics during heavy ion irradiation

Here, the evolution of irradiation defects in Ti 2 AlC under Kr ion irradiation at room and elevated temperatures has been investigated. The dependence of defect size and density on the irradiation dose and temperature was characterized by in-situ transmission electron microscopy (TEM) irradiation experiments. At 25-800 °C, interstitial-type dislocation loops with several nanometers in diameter form on the basal plane of Ti 2 AlC mostly with a Burgers vector of 1/2[0 0 0 1], and no voids have been observed. The number density of dislocation loops approaches saturation at a level of about 2 dpa at 800 °C. Ti 2 AlC samples remain crystalline in structure without amorphization or phase transformation. However, the localized disorder of atomic positions on the basal plane of the irradiated Ti 2 AlC was revealed by high-resolution transmission electron microscopy (HRTEM) studies, which increases with irradiation dose while decreases with elevated temperatures.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Emulation of fast reactor irradiated T91 using dual ion beam irradiation

In this work, dual ion irradiations using 5 MeV defocused Fe 2+ ions and co-injected He 2+ ions were conducted on a ferritic-martensitic steel alloy, T91, in the temperature range of 406 °C–570 °C over a damage range of 14.6–35 dpa followed by characterization of the microstructure using transmission electron microscopy (TEM) and scanning transmission electron microscopy (STEM). Dislocation loops were observed to increase in diameter and decrease in density with temperature until only network dislocations were observed at the highest temperatures of 520 °C and 570 °C. Swelling exhibited the expected bell-shaped trend with temperature following the number density of cavities, peaking at 460 °C and with a bimodal size distribution except at 520 °C and 570 °C. Nickel- and silicon-rich clusters formed under dual ion irradiations near the surface at all but the highest temperatures of 520 °C and 570 °C. Very little Cr and Si segregation was observed at lath boundaries while Ni enriched at all temperatures examined. Segregation of Cr and Ni appeared to saturate by 17 dpa, while Si enriched up to 35 dpa. The dislocation and cavity microstructures of dual ion irradiated T91 and T91 irradiated in the BOR-60 fast reactor matched extremely well using a temperature shift of +60–70 °C. However, segregation to grain boundaries and formation of nickel-silicon rich clusters were minimal in the dual ion irradiated T91 and less than that in T91 irradiated in the BOR-60 fast reactor.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Irradiation-induced formation of G-phase precipitates and M 2 X carbides in self-ion irradiated HT-9

Ferritic-martensitic steels with high chromium content are a promising material group for advanced nuclear systems due to their high temperature strength and good irradiation tolerance. HT-9 is an optimized and often-studied alloy in this group, but additional studies are required on its radiation response under extreme conditions to be experienced in various types of nuclear reactors, especially with respect to phase stability under irradiation. Self-ion irradiation of HT-9 by 5 MeV Fe ions was used to simulate neutron-induced behavior reaching peak doses of 100 and 300 dpa at temperatures ranging from 450 to 550 °C. M 23 C 6 carbides that existed prior to irradiation were found to remain stable under all examined irradiation conditions. As irradiation progressed at 450 and 500 °C, however, formation of spherical-like G-phase precipitates and needle-like M 2 X carbides was observed. G-phase precipitates were found to be enriched in Ni, Si, and Mn, and show no interface segregation, whereas needle-like M 2 X carbides were rich in Cr and Mo and clearly displayed interface segregation of Ni and Si. M 2 X carbide formation is believed to be assisted by vacancies, while G-phase precipitation is thought to be assisted by interstitials. Finally, this difference in defect-mediated formation leads to a difference in distribution with depth. M 2 X carbides are distributed over shallower depths than that of G-phase precipitates, consistent with defect imbalance predictions that consider the influence of the injected interstitial effect.

atom probe tomography↗
DOE OSTI2014

Direct observation of the structural and electronic changes of Li{sub 2}MnO{sub 3} during electron irradiation

This study focuses on the effects of electron beam induced irradiation to the layered oxide Li{sub 2}MnO{sub 3}. Aberration-corrected scanning transmission electron microscopy and electron energy loss spectroscopy are used to characterize structural and electronic transitions in the material during irradiation, with a focus on changes in Mn valence and O content. This truly in situ irradiation allows for specific particle tracking, dose quantification, and real-time observation, while demonstrating many parallels to the oxide's structure evolution observed during electrochemical cycling. Furthermore, it is demonstrated that typical imaging conditions are not severe enough to induce damage to the pristine oxide.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2015

Meso-scale magnetic signatures for nuclear reactor steel irradiation embrittlement monitoring

Verifying the structural integrity of passive components in light water and advanced reactors will be necessary to ensure safe, long-term operations of the existing U.S. nuclear fleet. This objective can be achieved through nondestructive condition monitoring techniques, which can be integrated with plant operations to quantify the “state of health” of structural materials in real-time. While nondestructive methods for monitoring many classes of degradation (such as fatigue or stress corrosion cracking) are relatively advanced, this is not the case for degradation caused by irradiation. The development of nondestructive evaluation technologies for these types of degradation will require advanced materials characterization techniques and tools that enable comprehensive understanding of nuclear reactor material microstructural and behavioral changes under extreme operating environments. Irradiation-induced degradation of reactor steels causes changes in their microstructure that impacts their micro-magnetic properties. In this paper, we describe preliminary results of integrating advanced material characterization techniques with meso-scale computational models. In the future, this will help to provide an interpretive understanding of the state of degradation in structural materials. Microstructural data are presented from monocrystalline Fe and are correlated with variable-field magnetic force microscopy and micro-magnetic measurements. Ongoing research is focused on extending the measurements and models on thin films to gain insights into the structural state of irradiated materials and the resulting impact on magnetic properties. Preliminary conclusions from these correlations are presented, and next steps described.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗