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DOE OSTI2023

On electron beam-induced degradation of vinyl ester thermosets

We have demonstrated that electron beam radiolysis induces scissions of the C-O-C bonds along the backbone of the chains of unsaturated polyester thermosets of different compositions based on dicyclopentadiene, isophthalic acid, epoxy vinyl ester, and terephthalic acid. The radiolysis is imminent irrespective of the degree of crosslinking in the thermosets both in neat resins and in the presence of solvents. Electron Paramagnetic Resonance (EPR) results show the formation of the alkoxyl radicals and C-centered radicals as the primary intermediate products of the C-O-C scissions. While the alkoxyl radicals of these resins exhibit very good stability even six months after the irradiation, the C-centered radicals decay very rapidly via their reactions with oxygen that is available either as adsorbed in the resins, dissolved in solvents or from the environment. The radiolytically produced •OH radicals in the unsaturated-ester aqueous solutions play a major role in inducing scissions on the backbone of the polymer chains. The solvated electrons (e s ) from organic solvents, such as dimethyl sulfoxide and isopropyl alcohol, also induce direct scission of the C-O-C bonds, giving rise to the formation of alkoxyl radicals and C-centered radicals. However, a considerable fraction of es is scavenged by the dissolved O 2 to produce O 2 - . Despite the radiation-induced scissions, irradiation of the resins at a dose level of 1000 kGy results in an increase of the glass transition temperature, Tg. This is due to the simultaneous radiation-induced polymerization of the vinyl monomers, toluene, and styrene that are present in the resins. The increase in Tg is observed in all the resins except for the terephthalic polyester resin in aqueous solution, in which the absence of these monomers results in Tg decreasing sharply with irradiation. This work demonstrates that ionizing radiation triggers continuous free radical-chain reactions that lead to the formation of recyclable oligomers.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2020

ECAR-3873 Acceptance of Nonconforming Fuel for the AGR-5/6/7 Irradiation Experiments

Low-enriched fuel compacts fabricated for the AGR-5/6/7 irradiation experiments did not conform to all fuel specifications. Reasons why the fuel is acceptable for use in the AGR-5/6/7 irradiation are discussed. Revision 1 alters the characterization data for the dispersed uranium and exposed kernel defect fractions. Additional characterization data was obtained for fuel compacts analyzed by Oak Ridge National Laboratory. This enabled statistical outliers to be rejected and to confirm that results obtained from BWX Technologies Nuclear Operations Group were statistically independent of the Oak Ridge results and should be excluded from the analyses. Revision 2 alters the characterization data for the TRISO lot, J52R-16-98005, used in the fabrication of the AGR-5/6/7 fuel. Data from the BWXT J52R-16-98005 Certification package is used to calculate the reported values.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

AGR-5/6/7 Fuel Fabrication Report

The U.S. Department of Energy Office of Nuclear Energy (DOE NE) and the Idaho National Laboratory (INL) Advanced Reactor Technologies (ART) Advanced Gas Reactor (AGR) Fuel Development and Qualification program (referred to as AGR Fuel program hereafter) are pursuing qualification of tristructural isotropic (TRISO) coated particle fuel for use in high temperature gas cooled reactors (HTGRs). The AGR Fuel program was established to provide a fuel qualification data set in support of the licensing and operation of an HTGR. BWX Technologies Nuclear Operations Group (BWXT-NOG) was subcontracted to fabricate the fuel for the AGR program. Several investments and innovations were realized in preparation to fabricate fuel for the AGR-5/6/7 irradiation experiments that brought fuel fabrication fully out of the laboratory and into engineering-scale operations. These included: • Increased the kernel fabrication line capacity and uniformity • Upgraded ancillary support equipment and processes for the tristructural isotropic (TRISO) coating furnace • Demonstrated efficient production of the matrix precursor powder by dry jet milling of co mingled components • Demonstrated an engineering-scale method for quick and efficient overcoating TRISO particles with the matrix precursor • Demonstrated an automated, multi cavity compacting system with a volumetric feed system • Demonstrated a combined-cycle thermal treatment furnace These changes increased production rates of some of these processes by an order of magnitude or more while eliminating the use of flammable solvents, multiple grinding and sorting operations, and the weighing out of individual die charges. Three fuel kernel lots were fabricated for production of the fuel for AGR-5/6/7. The initial lot (J52R-16-39316) was certified to fuel specifications but was not used because the kernels had a high fraction of internal fissures that caused an unacceptable fraction of the kernels to fragment when charged to the coating furnace where the TRISO coating would be deposited. Fragmented kernels increased the dispersed uranium in the particles and produced a worrisome fraction of dimpled particles with an elevated probability of in-pile failure. After some efforts to identify the cause of the fissure formation, two additional lots were produced with much lower fissure fractions, J52R-16-69317 and 69318. The latter kernel lot was a backup to the first and was not needed. Multiple kernel batches were composited to form each of the lots so as to simulate a commercial-scale operation where kernel batches would also be composited. Multiple TRISO coating runs were performed and the product characterized so that several could be composited into a TRISO lot. TRISO lot J52R-16-98005 conformed to all fuel specifications except the mean outer pyrocarbon (OPyC) layer thickness was thinner than specified. Furthermore, it was determined that the TRISO lot had a dispersed uranium fraction (DUF) that might result in the compacts not meeting the DUF specification. A review of the role of the OPyC layer and consequences of the DUF by the Technical Coordination Team and INL concluded that the fuel was acceptable for use in the AGR-5/6/7 irradiation experiment. The TRISO particles were overcoated with the matrix precursor that had been produced in a jet-milling operation. The overcoating was performed in equipment originally designed to coat pharmaceuticals. The overcoating process performed well; producing highly spherical and uniform overcoats requiring little upgrading and no recycle or rework. TRISO particles were overcoated with the matrix precursor to achieve nominal volumetric packing fractions (PFs) of TRISO particles of 25% and 40% for the irradiation experiments. The 40% PF compacts occupy the first and fifth test capsule in the test train while the inner three capsules are loaded with 25% PF compacts. The resinated graphite matrix precursor powder was a derivative of the German A3-27 matrix formulation, which differs from previous AGR irradiation campaigns that used an A3-3 formulation. Jet milling of the matrix powder precursor produced a finer mean graphite particle size than the milling operations used for the A3-3 matrix powder precursor. Changes made in the matrix formula and equipment yielded compacts with significantly higher matrix density than was attained in previous AGR irradiation campaigns. The changes in the matrix formulation and the means of milling the powders also complicated resolution of the three fuel compact defect fractions, DUF, exposed kernel fraction (EKF), and the silicon carbide defect fraction (SDF). Characterization data from BWXT-NOG had some anomalous results, so samples of the fuel compacts and overcoated TRISO particles were also analyzed by Oak Ridge National Laboratory (ORNL) to ensure that the defect fractions were accurately characterized.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

The Complexity and Challenges of Irradiation-Induced Phase Instability Phenomenon in Nuclear Power Plant Components

High energy irradiation by neutrons or ions of metastable austenitic 300-series steels appears to initiate formation of bcc-phases with different morphology and elemental composition compared to that of retained ferrite. Extensive formation of specific Fe-rich ferrite was observed on the grain boundaries of annealed Russian steel 12Cr18Ni10Ti, an analog of AISI 321, after irradiation in the center-core region of sodium-cooled BN-350 fast reactor (SFR) doses up to 57 dpa at ~1×10-6dpa/sec. The ferrite fraction determined from scanning electron microscopy (SEM) closely matched the magnetic fraction determined using a ferritscope confirmed by X-ray diffraction analysis and selected area diffraction patterns. Chemical composition of the secondary phases was determined by energy dispersive X-ray analysis using transmission electron microscope, showing Fe-rich ferrite domains. Fe-ion irradiation (2.3 MeV) was used to study the accumulation of ferrite and martensite phases at higher dpa rates. The very high etchability of these ferrite particles in alcohol-based solutions may signal a very high etching in hot water, leading to accelerated intergranular cracking, especially upon long-term exposure during extended plant lifetimes.

D a, Merezhko↗

DOE OSTI2021

Post Examination of Candidate NpO 2 Targets for 238 Pu Production

Four targets composed of sintered NpO 2 pellets enclosed in Zircaloy cladding were irradiated in the High Flux Isotope Reactor for 2 to 4 cycles to determine if an oxide pellet and Zircaloy cladding would be usable candidates for 238 Pu production. The pellets were sintered to a high fraction of theoretical density (TD) and melt wires were placed in small holes that had been drilled in the center of selected pellets to determine if the pellet centers exceeded desirable operating temperatures. After irradiation, the targets were punctured to determine the internal target pressure and the fission gas release. None of the targets had excessive pressure, and fission gas release was in the range of 4 to 38%. The targets were then cut into segments for radiochemical analysis, and metallography (MET) mounts were created to examine the pellet microstructure and to search for the melt wires to determine if they had melted. Three of the four targets resulted in successful MET mounts; the two-cycle pellets were too friable to be handled. No signs of pellet melting were noted, and unfortunately, none of the melt wires could be recovered. The results indicate that this pellet/clad design offers a practical option for 238 Pu production.

36 MATERIALS SCIENCE↗

DOE OSTI2021

New approach to precisely measure γ-ray intensities for long-lived fission products, with results for the decay of 95 Zr

For many fission products, the rays emitted following decay provide an easily-detectable signature that can be used to identify their quantities and distributions in a sample. As a result, γ-ray spectroscopy is often exploited to study fission-product yields, provided sufficiently accurate information on the γ-ray intensity is available. However, in many cases, the uncertainties in the existing nuclear data are large enough that they compromise the precision achievable for modern experiments and applications. To address this need, we have developed a new experimental method that is well suited to precisely measure absolute γ-ray intensities in the β decay of long-lived fission products. The approach involves the production of a radiopure sample by implantation of a mass-separated ion beam from the CAlifornium Rare Isotope Breeder Upgrade (CARIBU) facility on a thin carbon foil. The emitted β-decay radiation is detected with a 4π gas proportional counter and a meticulously efficiency-calibrated high-purity germanium (HPGe) detector. As a first measurement to demonstrate the approach, we studied the absolute γ-ray intensities of the strongest transitions following the β decay of 95 Zr and its decay-daughter 95 Nb, and determined them to fractional precisions of better than 1–2%. In addition, with a larger sample of activity produced through neutron irradiation of an isotopically-enriched Zr foil, we performed a high-precision measurement of the relative γ-ray intensities following the decay of 95 Zr with just the HPGe detector. Here, the sample-production method at CARIBU and the coincidence detection approach demonstrated here can be applied to study fission products with half-lives longer than a day, which includes isotopes important not only for nuclear-energy and national-security applications, but also for medical-isotope research and environmental monitoring.

4π↗

DOE OSTI2020

Plutonium Dedicated Surface Probe Microscopy Capability

An atomic force microscope (AFM) equipped with gas dosing capabilities and a scanning tunneling microscope (STM) housed within a vacuum chamber with Pu sputtering capabilities and the potential for rapid heating via electron bombardment have been devised for investigations of Pu surface aging (such as oxidation, corrosion, and self-irradiation damage), reactivity, and rapid heating. These instruments are capable of probing Pu surface areas ranging from a fractions of nm2 (STM) to hundreds of μm2 (AFM) to obtain three dimensional imaging of surface microstructure and morphology, as well as mapping of the local electronic structure (i.e., the occupied and unoccupied electronic states), surface chemical, and mechanical properties. A detailed description of our surface probe microscopy (SPM) capability, specifically designed for handling of Pu coupons is given in the references. In FY20, four tasks have been undertaken: Resumed maintenance of Pu AFM and modification of STM capabilities including calibration, annual measurement of Pu surface, and implementation of additional instrumentation for rapid heating effort; Analyzed AFM FY19 gas dosing experiment data of 7 at% gallium-stabilized δ-phase Pu (δ-Pu) coupon; Determined the direction of future gas dosing experiments to investigate reversibility of surface features; and, Published a paper titled "Effects of ion sputtering on plutonium surfaces" in the Journal of Nuclear Materials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Isotope Production at the Jefferson Lab LERF Facility (Final Report)

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2021

Isotope Production at the Jefferson Lab LERF Facility

The project initially proceeded well, meeting all of the Milestones. However, the need to create a new irradiation area in the LERF (the initial proposal was to use the CEBAF Injector) used up a larger fraction of the budget than foreseen. A further problem arose when the hot-cell became unavailable at VCU, followed by VCU dropping out of the collaboration. This meant that we were unable to separate the 67 Cu from the gallium. We replaced this by carrying out detailed gamma spectroscopy to determine the quantity of 67 Cu produced. This work was published as presentations in November of 2021 by the two doctoral students (Geno Santistevan and Robert Bentley) and proceedings articles for American Nuclear Society’s topical meeting AccApp21 (https://www.ans.org/pubs/proceedings/issue-3206/), as well as in invited, expanded versions in the peer-reviewed journal “Nuclear Science and Engineering”. Geno and Robert have subsequently successfully leveraged this research to achieve “doctoral candidacy” status and are expected to defend their dissertations in the Fall of 2023.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2020

Comparison of aircraft measurements during GoAmazon2014/5 and ACRIDICON-CHUVA

The indirect effect of atmospheric aerosol particles on the Earth's radiation balance remains one of the most uncertain components affecting climate change throughout the industrial period. The large uncertainty is partly due to the incomplete understanding of aerosol–cloud interactions. One objective of the GoAmazon2014/5 and the ACRIDICON (Aerosol, Cloud, Precipitation, and Radiation Interactions and Dynamics of Convective Cloud Systems)-CHUVA (Cloud Processes of the Main Precipitation Systems in Brazil) projects was to understand the influence of emissions from the tropical megacity of Manaus (Brazil) on the surrounding atmospheric environment of the rainforest and to investigate its role in the life cycle of convective clouds. During one of the intensive observation periods (IOPs) in the dry season from 1 September to 10 October 2014, comprehensive measurements of trace gases and aerosol properties were carried out at several ground sites. In a coordinated way, the advanced suites of sophisticated in situ instruments were deployed aboard both the US Department of Energy Gulfstream-1 (G1) aircraft and the German High Altitude and Long-Range Research Aircraft (HALO) during three coordinated flights on 9 and 21 September and 1 October. Here, we report on the comparison of measurements collected by the two aircraft during these three flights. Such comparisons are challenging but essential for assessing the data quality from the individual platforms and quantifying their uncertainty sources. Similar instruments mounted on the G1 and HALO collected vertical profile measurements of aerosol particle number concentrations and size distribution, cloud condensation nuclei concentrations, ozone and carbon monoxide mixing ratios, cloud droplet size distributions, and downward solar irradiance. We find that the above measurements from the two aircraft agreed within the measurement uncertainties. The relative fraction of the aerosol chemical composition measured by instruments on HALO agreed with the corresponding G1 data, although the total mass loadings only have a good agreement at high altitudes. Furthermore, possible causes of the discrepancies between measurements on the G1 and HALO are examined in this paper. Based on these results, criteria for meaningful aircraft measurement comparisons are discussed.

47 OTHER INSTRUMENTATION↗

DOE OSTI2021

Initial demonstration of automated fuel performance modeling with 1977 EBR-II metallic fuel pins using BISON code with FIPD and IMIS databases

Using the BISON fuel performance code, simulations were conducted using an automated process to read initial and operating conditions from the Fuels Irradiation and Physics Database (FIPD) and Integral Fast Reactor materials information system (IMIS) database, which contains metallic fuel data from the Experimental Breeder Reactor-II (EBR-II). This work demonstrates use of an integrated framework to access the vast majority of EBR-II experimental fuel pin data to support rapid development of fuel performance models for next-generation metallic fuel systems. With this capability, validation for fuel qualification can be performed rapidly. Between IMIS and FIPD, there is enough information to conduct 1977 unique EBR-II metallic fuel pin histories from 24 different experiments, at varying levels of detail between the two databases. Each of these histories includes a high-resolution power history, flux history, coolant channel flow rates, and coolant channel temperatures. Fission gas release (FGR), cumulative damage fraction (CDF), fuel axial swelling, cladding profilometry, and burnup were all simulated in BISON. The results were compared to post-irradiation examination (PIE) results for the initial demonstration of automated BISON modeling. BISON simulations conducted with IMIS and FIPD were in rough agreement with PIE measurements and calculations. Cladding profilometry, FGR, and fuel axial swelling were found to be in rough agreement with PIE measurements, depending on the physics used within the BISON input files. Here, the mechanical contact solver chosen was found to significantly impact axial fuel swelling and cladding strain predictions. CDF values were assessed to see whether pin failure may have been predicted (CDF ≥ 1). This work suggests that continued development of an automated tool for BISON should focus on inclusion of the Fast Flux Test Facility (FFTF) experimental data for a larger database for metallic fuel, improved physical models to better capture fuel performance, such as fuel-cladding interactions, and a more detailed comparison with available PIE data to further the BISON model development.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

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.↗

DOE OSTI2025

Gallium oxide (Ga2O3) as a radiation detector distinguishing neutrons and gammas

Gallium oxide (Ga2O3) is a promising ultrawide bandgap semiconductor for radiation detection with the potential of integrating electronic and scintillation functions within a single crystal device. This study establishes the scintillation response of β-Ga2O3 gamma irradiation from yttrium-88 (88Y). Then, californium-252 (252Cf) is used as a spontaneous fission source of mixed neutron and gamma radiation field to measure scintillation signals. Pulse shape discrimination and constant fraction discrimination techniques were used to separate neutron and gamma interaction events. Further investigation indicates that the prompt temporal responses of β-Ga2O3 for gammas and neutrons may enable discrimination of the two by prompt pulse fitting methods, focused around the initial peak. For gamma irradiation, we observed a rise time (τr) of 2.1 ns, decay time (τd) of 9.5 ns, and a full width at half maximum (FWHM) of 6.2 ns. For neutrons, it showed a τr of 2.3 ns, a τd of 12.1 ns, 9.4 ns FWHM, and reduced peak intensity. A diamond detector exhibited a more symmetrical τr and τd for both gamma and neutron signals and therefore is less effective at discriminating between the two by this method. This draws attention to β-Ga2O3’s ability to distinguish neutron and gamma particles. These findings showcase Ga2O3’s potential as a next-generation semiconductor for applications in nuclear safety and medical imaging, where precise discrimination between neutron and gamma interactions is essential.

Valdes, D. J. (ORCID:0000000304373720)↗

DOE OSTI2023

Gas-phase Radiolysis in Plutonium Dioxide Powder

A sample of PuO 2 powder undergoes self radiolysis, causing water present in the solid to generate hydrogen, oxygen, and minor species (e.g. H 2 O 2 ). This is a matter of concern for stored PuO 2 , since the gases could react exothermically and rupture the container. In experiments at LANL, most PuO 2 /salt samples generate mostly H 2 , but a few samples generate a mixture of H 2 and O 2 together. In these special cases, the mixture of H 2 and O 2 in the pore volumes of the powder will then be subject to radiolysis, causing the reverse reaction H 2 + O 2 $\Rightarrow$ H 2 O. In order to understand the steady-state quantities of these gases, it is necessary to estimate the amount of energy from the alpha irradiation that drives this reaction. This report outlines one possible approach for this process, yielding a value for the fractional amount of radiation deposited in the gas phase of 294 ppm. This corresponds to the case of a powder with porosity of 78%. We then use a comprehensive chemical kinetics model based on elementary gas phase reactions to calculate the rate of conversion back to water. We estimate a reduction in the H 2 pressure of about 12 kPa/month, which applies only when O 2 is simultaneously present in the gas phase.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2023

Microstructure and in-service degradation of baffle former bolts – in-core components of light water reactors

This report presents a microstructure evaluation of the baffle former bolt, a commercial pressurized water reactor component. Using SEM, EDS, and EBSD analysis the material was identified as AISI 316 steel with an annealed austenite microstructure. Findings include abnormal grain growth in a fraction of the grains and the absence of retained ferrite. Notable observations include an overlapping of pre- and post-irradiation deformation, including defect-free channel formation as an active deformation mechanism in the near-surface layer of BFB. Specimen surfaces, exposed to high-temperature, high-pressure water, exhibited signs of in-service corrosion degradation. EBSD and EDS analyses highlighted intergranular corrosion, possible grain boundary oxidation with depths less than 3 µm, and, unexpectedly, short cracks filled with Cr-rich oxides measuring around 5-6 µm. The presence of defect-free channels from EBSD data suggests episodes of high mechanical stresses during service. The specimens provide a potentially unique insight into the in-service degradation of NPP components, strain localization, and crack initiation processes. An intact corrosion layer reveals a complex component's in-service history. Further research is underway for a detailed analysis.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2020

Three-dimensional bubble reconstruction in high burnup UO 2

In light water reactor (LWR) UO 2 fuels, the evolution of volatile fission products is one of the critical areas of fuel behavior that is yet to be fully understood. In UO 2 irradiated to high burnups, it is well known that most released fission gases come from the central region of the fuel as opposed to the highly porous high burnup structure (HBS) on the periphery of the pellets. However, fuels with and without interconnected bubble networks at the fuel center showed high to moderate release fractions, which conceals the mechanisms responsible for the gas release in the latter scenario. In this work, focused ion beam tomography was used to investigate the three-dimensional bubble structure in an irradiated LWR UO 2 fuel pellet with high degree of fission gas retention so that the degree of bubble interconnection could be assessed. Six radial locations with different burnups and temperatures were serially sectioned and imaged to reconstruct the three-dimensional bubble structure. As expected, the highest porosity was observed at the periphery of the fuel (HBS). Here, the porosity then decreased towards the pellet center, except for the centermost location. This location had a slightly higher porosity than its adjacent mid-radial location, which was attributed to the temperature difference between the two locations. This study provides a first-time volumetric evaluation of the porosity at different radial locations on a UO 2 fuel pellet. During this investigation, no significant bubble interconnection was noted at any of the six radial locations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

In-situ observation of nano-oxide and defect evolution in 14YWT alloys

Nanostructured ferritic alloys (NFAs) are considered as candidates for structural components in advanced nuclear reactors due to their excellent radiation resistance as a result of a high density of nano-oxides (NOs) in the microstructure. Therefore, gaining an understanding on the stability of NOs under irradiation is crucial. In this study, we have investigated the evolution of defects and NOs in 14YWT NFAs under in-situ Kr ion irradiation at room temperature (RT) and 450°C up to 10 dpa. It has been found that irradiations at 450°C do not create any changes in the NOs, similar to the bulk irradiations. On the other hand, elemental mapping indicates that NOs dissolve mostly after 10 dpa irradiations at RT. Thus, while defects are both annihilated and pinned by NOs at low doses (before the dissolution of NOs), glissile loops start to escape to the foil surface at high doses (after the dissolution of NOs), justifying the significantly low fraction of (111) loops compared to the literature values. Furthermore, high resolution transmission electron microscopy analysis has shown that the NOs are mostly coherent Y 2 Ti 2 O 7 particles with pyrochlore crystal structure after both RT and 450°C irradiations, similar to those observed before irradiation.

36 MATERIALS SCIENCE↗

DOE OSTI2021

Fabrication of MiniFuel Compacts for High-Power Irradiation Testing of TRISO Fuel

MiniFuel compacts containing 20 individual tristructural isotropic (TRISO)-coated fuel particles were characterized to provide supporting preirradiation characterization data. The preirradiated characterization data supports irradiation capsule design and safety analysis, as well as downstream postirradiation examination (PIE) for a planned MiniFuel irradiation to explore high particle powers. The analysis included dimensional inspection, x-ray radiography and tomography, as-fabricated defect fraction analysis, and a matrix impurity analysis.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗