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13 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

AGC 4 Graphite Specimen Postirradiation Characterization Plan

This characterization plan describes the thermal, physical, and mechanical measurement techniques that will be used to characterize graphite samples being tested in the fourth Advanced Graphite Creep experiment (AGC-4). Instruments, fixtures, and methods are currently in place for both pre and postirradiation material property measurements of bulk density, thermal diffusivity, coefficient of thermal expansion, elastic modulus, and electrical resistivity. Postirradiation testing procedures used to characterize the samples are described and discussed in the plan. Where they exist, American Society for Testing and Materials (ASTM) International testing standards will apply to the tests. Any departure from ASTM International testing standards or the approved laboratory procedures are documented within this characterization plan. Deviations that occur during testing will be documented in data reports.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Postirradiation characterization of palladium as an additive for fuel cladding chemical interaction mitigation in metallic fuel

This work describes the microstructural and elemental characterization of irradiated metallic fuels containing palladium as an additive. The use of additives has been proposed to control Fuel-Cladding Chemical Interaction (FCCI) and thus to promote higher fuel utilization (i.e., higher burnup). In this work, Pd has been investigated as a potential additive to metallic fuel to bind lanthanides, impeding their migration and attack on the cladding. The influence of Pd on the microstructure, chemistry and performance of metallic fuel has been characterized via scanning electron microscopy for two metallic fuel designs—namely, annular and solid fuel. Pd was observed to play an important role in the chemistry of the fuel. Indeed, the addition of Pd leads to the formation of new phases. Pd was detected to combine not only with the lanthanides, as intended, but also with Zr, a main element of the fuel matrix. While Pd proved to be effective in preventing lanthanide migration and their attack on the cladding, the Pd-Zr compound may potentially lead to other unexpected fuel-performance issues, such as the formation of low-melting point phases and increased unalloyed U available for FCCI interaction with Fe in the cladding. Even the increase of Zr to 13wt%. did not completely mitigate this adverse phenomenon generated by the Pd-Zr interaction. Furthermore, the efficacy of using this additive needs further investigation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Testing fast reactor fuels in a thermal reactor: Comparison of transmutation metallic fuel alloys behavior by scanning electron microscopy

To optimize nuclear waste repository performance, the destruction of minor actinide elements, particularly Np and Am, in a neutron fast spectrum reactor is possible by incorporating these elements into nuclear fuel. Evaluating the performance of minor actinide containing fuel is of paramount importance to enabling this technology. However, such a task is challenging without an available domestic fast spectrum test reactor. A comparison of fuel performance tested in an available domestic thermal reactor at the Idaho National Laboratory, the Advanced Test Reactor, and in a fast spectrum reactor in France (Phénix) is presented here in this study. This study evaluates the capability of using a cadmium shrouded test position to mimic the power profile along the fuel radius present in fast spectrum reactors so that thermally driven phenomenon (e.g., constituent redistribution) can be evaluated in a thermal reactor and determined to be prototypical of a fast reactor. Thus, optical microscopy and scanning electron microscopy has been performed on irradiated 35U-29Pu-4Am-2Np-30Zr fuel samples (where the number preceding the element is the weight percent concentration) from the two mentioned reactors that present similar irradiation temperatures and power conditions. The results indicate that fuel performance phenomena are reproducible in the two irradiation conditions. The redistribution of Zr occurred in the same manner for the two samples. Similar partitioning of U-Pu-Zr phases was observed, and the behavior of Am was similar in the analyzed specimens. Finally, the overall microstructure evolution seems not to be affected by minor actinides addition compared to expected behavior of conventional U-19Pu-10Zr ternary metal fuels for both specimens. Slight differences in fuel cladding chemical interaction were, however, observed. This difference is likely driven by difference in cladding composition rather than irradiation conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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↗
DOE OSTI2023

BISON validation to in situ cladding burst test and high-burnup LOCA experiments

The process for developing and qualifying nuclear fuels for commercial nuclear application requires fundamental material development, characterization, and design; out-of-pile testing on unirradiated materials; integral fuel rod irradiations, testing, and postirradiation examinations; and transient analyses. The historical approach depends on the generation of large empirical datasets and series of integral fuel rod irradiations, and this approach ultimately takes ~20 years—or sometimes longer—to acquire data through extensive sequential testing. Thus, the qualification and eventual deployment of new fuel systems constitute a long process. However, recent technological advancements have provided researchers the opportunity to perform out-of-cell, in situ measurements to assess material performance for the duration of the experiment. One such example of this capability is the use of digital image coordination and thermal imaging to assess Zircaloy cladding performance under a simulated loss-of-coolant accident (LOCA) transient condition. In situ measurements generally provide high-fidelity strain, strain rates, and temperature surface maps. This is critical for the US nuclear industry, which is actively developing a technical basis to support extending the peak rod average burnup from 62 to ~75 GWd/tU and the deployment of accident-tolerant fuel. However, the US Nuclear Regulatory Commission (NRC) outlined in its research information letter several technical issues that the industry must address before extending burnup. One topic of specific interest is understanding the cladding balloon and rupture geometry during the LOCA heat-up phase. By leveraging these advanced in situ capabilities, this work used in situ data generated from a simulated LOCA to better understand high-temperature creep and its effect on Zircaloy balloon and rupture performance. Here, this work used the BISON fuel performance code to assess the high-temperature creep model predictions with in situ data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Characterization of a minor actinides bearing metallic fuel pin irradiated in EBR-II

The X501 transmutation experiment irradiated in Experimental Breeder Reactor II (EBR-II) was intended to evaluate the safety and performance of the addition of minor actinides (Am, Np) into the metallic nuclear fuel system. The irradiation of this experiment is also of paramount importance for comparisons between true fast spectrum irradiations (e.g. EBR-II) and shrouded irradiation capsules performed at the thermal neutron spectrum Idaho National Laboratory Advanced Test Reactor. The X501 experimental fuel pins were irradiated to a peak measured burnup of 6.2% fission per initial heavy metal atom. This paper reports and discusses Postirradiation Examination results for one of two minor actinide bearing pins from the EBR-II X-501 experiment (X501-G591) with a focus on fuel microstructure evolution. Additionally, fuel swelling, fission product distribution, cladding strain, fission gas release, fuel microstructure and fuel cladding chemical interaction were all evaluated. The fission gas release was 79.9% and the microstructure along the fuel pin presented several characteristics very similar to other EBR-II U-Pu-Zr ternary metallic fuel experience. Minor actinides seem to not dramatically affect the overall performance of this candidate transmutation fuel. Scanning Electron Microscopy was also performed and an in-depth characterization of the different regions and phases is presented. Performance data from these irradiations can be used to inform the feasibility of minor actinide transmutation in future reactor systems.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1979

Characterization of solar cells for space applications. Volume 5: Electrical characteristics of OCLI 225-micron MLAR wraparound cells as a function of intensity, temperature, and irradiation

Computed statistical averages and standard deviations with respect to the measured cells for each intensity temperature measurement condition are presented. Display averages and standard deviations of the cell characteristics in a two dimensional array format are shown: one dimension representing incoming light intensity, and another, the cell temperature. Programs for calculating the temperature coefficients of the pertinent cell electrical parameters are presented, and postirradiation data are summarized.

Anspaugh, B. E.↗
NASA NTRS2001

Response of thyroid follicular cells to gamma irradiation compared to proton irradiation. I. Initial characterization of DNA damage, micronucleus formation, apoptosis, cell survival, and cell cycle phase redistribution

The RBE of protons has been assumed to be equivalent to that of photons. The objective of this study was to determine whether radiation-induced DNA and chromosome damage, apoptosis, cell killing and cell cycling in organized epithelial cells was influenced by radiation quality. Thyroid-stimulating hormone-dependent Fischer rat thyroid cells, established as follicles, were exposed to gamma rays or proton beams delivered acutely over a range of physical doses. Gamma-irradiated cells were able to repair DNA damage relatively rapidly so that by 1 h postirradiation they had approximately 20% fewer exposed 3' ends than their counterparts that had been irradiated with proton beams. The persistence of free ends of DNA in the samples irradiated with the proton beam implies that either more initial breaks or a quantitatively different type of damage had occurred. These results were further supported by an increased frequency of chromosomal damage as measured by the presence of micronuclei. Proton-beam irradiation induced micronuclei at a rate of 2.4% per gray, which at 12 Gy translated to 40% more micronuclei than in comparable gamma-irradiated cultures. The higher rate of micronucleus formation and the presence of larger micronuclei in proton-irradiated cells was further evidence that a qualitatively more severe class of damage had been induced than was induced by gamma rays. Differences in the type of damage produced were detected in the apoptosis assay, wherein a significant lag in the induction of apoptosis occurred after gamma irradiation that did not occur with protons. The more immediate expression of apoptotic cells in the cultures irradiated with the proton beam suggests that the damage inflicted was more severe. Alternatively, the cell cycle checkpoint mechanisms required for recovery from such damage might not have been invoked. Differences based on radiation quality were also evident in the alpha components of cell survival curves (0.05 Gy(-1) for gamma rays, 0.12 Gy(-1) for protons), which suggests that the higher level of survival of gamma-irradiated cells could be attributed to the persistence of nonlethally irradiated thyrocytes and/or the capacity to repair damage more effectively than cells exposed to equal physical doses of protons. The final assessment in this study was radiation-induced cell cycle phase redistribution. Gamma rays and protons produced a similar dose-dependent redistribution toward a predominantly G(2)-phase population. From our cumulative results, it seems likely that a majority of the proton-irradiated cells would not continue to divide. In conclusion, these findings suggest that there are quantitative and qualitative differences in the biological effects of proton beams and gamma rays. These differences could be due to structured energy deposition from the tracks of primary protons and the associated high-LET secondary particles produced in the targets. The results suggest that a simple dose-equivalent approach to dosimetry may be inadequate to compare the biological responses of cells to photons and protons.

NASA Discipline Radiation Health↗
DOE OSTI2020

Out-of-pile and postirradiated examination of lanthanide and lanthanide-palladium interactions for metallic fuel

Palladium is being investigated as a fuel additive to bind with and potentially immobilize lanthanide fission products. A primary cause of fuel-cladding chemical interaction (FCCI) is the lanthanide fission products migrating to the fuel periphery and interacting with the cladding. This interaction will lead to wastage of the cladding and eventually to a cladding breach. Palladium has previously been identified as a promising additive used to prevent or decrease FCCI by reacting with the lanthanide fission products. In the current study, an alloy cast from the four highest abundant lanthanides found in irradiated metallic fuel, Nd, Ce, Pr, and La, with and without Pd, has been characterized using neutron diffraction, scanning electron microscopy, and electron probe microanalysis. In the lanthanide-Pd intermetallic compounds, all of the constituent compounds, i.e. Nd-Pd, Ce-Pd, La-Pd and Pr-Pd are known. There is very good agreement, both structurally and compositionally, between the out-of-pile lanthanide alloy and lanthanide fission products characterized in irradiated fuels. In both cases, the lanthanide elements form a solid solution in a hexagonal crystal structure. The out-of-pile lanthanide alloy follows Vegard's Law, with the measured and calculated (weighted average of constituents) lattice parameters being within 1% for both the a and c parameters. Pd bonds with the lanthanides (Ln) forming the phases LnPd and Ln7Pd3. The results indicate the properties of lanthanide compounds in irradiated metallic fuel can be reliably simulated in out-of-pile experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Summary of the Postirradiation Examination of the First Samples from a MiniFuel Irradiation

An overview of postirradiation examination results for uranium nitride kernels and uranium nitride coated particles irradiated in the High Flux Isotope Reactor are presented. This is the first postirradiation examination of the MiniFuel irradiation vehicle that was recently developed to rapidly screen different nuclear fuel concepts. Observations on fission gas release, irradiation conditions and microstructure of the irradiated fuel show good fuel performance at the low burnup achieved in this initial irradiation. The burnup measured by mass spectrometry ranged from 5.9 to 10 MWd/kgU as was achieved after only 68 effective full power days of irradiation. Results from silicon carbide thermometry measurements further benchmarked the MiniFuel irradiation vehicle and indicated fuel was irradiated at temperatures ranging from 410-460°C. Extensive microstructural characterization on the irradiated fuel was performed and no significant irradiation induced changes were observed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Postirradiation examination from separate effects irradiation testing of uranium nitride kernels and coated particles

An overview of postirradiation examination results for uranium nitride kernels and uranium nitride coated particles irradiated in the High Flux Isotope Reactor are presented. This is the first postirradiation examination of the MiniFuel irradiation vehicle that was recently developed to rapidly accumulate burnup during separate effects irradiation testing. In general, the burnup and fuel temperatures measured postirradiation were consistent with the design calculations. The burnup measured by mass spectrometry ranged from 5.9 to 10 MWd/kgU and was achieved after only 68 effective full-power days of irradiation. The dilatometric evaluation of passive silicon carbide thermometry indicated that the fuel was irradiated at temperatures ranging from 410 to 460 °C. Because the irradiation temperatures and burnup were low, the UN kernels showed minimal fission gas release that was within the range of the expected recoil (athermal) release. While it is possible to measure fuel swelling using x-ray computed tomography, the observed swelling was too small to quantify in this case. Extensive microstructural characterization of the irradiated fuel was performed in this study, and no significant irradiation induced changes were observed.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Demonstration of Dimensional and Microstructural Postirradiation Examination Capabilities on MiniFuel Disk Specimens

This work demonstrates two new MiniFuel postirradiation examination (PIE) capabilities, laser profilometry and serial sectioning, for future use on alpha-U specimens for the upcoming Mini-99 irradiation campaign. Representative U–Mo disk specimens—3 mm in diameter and approximately 0.75 mm in thickness—were subjected to dimensional measurements using standard calipers, as well as a new Keyence CL-3000 series laser profilometry system designed to measure the surface of both sides of the disk simultaneously. The laser profilometry system captured thickness gradients across the specimens as large as 0.2 mm. Additionally, because of the limitations inherent to the use of calipers, caliper measurements consistently overestimated the initial volume of MiniFuel disk specimens compared to the laser profilometry method by up to 5%. However, the results show that optimal data collection and analysis procedures are crucial to avoiding significant artifacts during data collection that can lead to large errors (approximately 10%). Serial sectioning using a combination of ion milling and electron imaging was also demonstrated for the analysis of micron-sized control volumes. Nanometer- to micron-scale porosity was resolved on a representative U–Mo specimen before irradiation, even when the surface of the specimen appeared free of porosity. Future efforts to relate microstructural features (porosity, fission product distributions, and so on) to macroscale effects (dimensional changes, thermal conductivity degradation) require accurate characterization methods such as those demonstrated in this work. These advanced PIE capabilities will be applied to optimized alpha-U specimens for the upcoming Mini-99 neutron irradiation campaign.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Destructive Examination of a FeCrAl-UO 2 Irradiation Test

Several destructive postirradiation examinations were performed on an irradiation specimen that coupled an early iron-chromium-aluminum (FeCrAl) candidate alloy cladding with UO2 pellets. This irradiation test was designed to investigate the early life performance and compatibility of the FeCrAl-UO 2 cladding-fuel system under prototypic light water reactor neutronic conditions. Additionally, these tests were expected to provide neutron irradiated samples for severe accident testing of this fuel system. The irradiation studied in this work was part of the ATF-1 series of drop-in style irradiations performed in the Idaho National Laboratory Advanced Test Reactor. The rodlet studied in this work is one of three similar rodlets irradiated in ATF-1 that had approximately 7.6 cm of cladding machined from a wrought FeCrAl alloy and fueled with a 6.1 cm stack of UO 2 pellets. The outer diameter of the cladding was ~0.94 cm and the inner diameter was ~0.83 cm. After irradiation and non-destructive examination in Idaho this rodlet was shipped to the Oak Ridge National Laboratory hot-cells for further examination. The irradiated rodlet was sectioned into several samples for microstructural, micromechanical, and severe accident testing. During sectioning, it was noted that the fuel was not firmly bonded to the cladding and could be readily removed from small cladding slices. Microstructural characterization of fuel cross sections also revealed no significant interaction between the fuel and the cladding. Samples were also prepared for microhardness testing. To prepare for high temperature oxidation testing, the fuel was dissolved from segments of the cladding. High temperature oxidation testing of cladding segments was performed at 1200°C and 1300°C in a steam environment. Comparisons between the oxidation of this FeCrAl alloy in its neutron irradiated state, as-fabricated state, and the oxidation of Zircaloy-2 are made. The oxidation testing will be followed by ring compression testing to evaluate ductility. The microstructure of the samples after oxidation and ring compression testing will also be analyzed.

Harp, Jason↗