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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

A compact dication source for Ba 2+ tagging and heavy metal ion sensor development

Here, we present a tunable metal ion beam that delivers controllable ion currents in the picoamp range for testing of dry-phase ion sensors. Ion beams are formed by sequential atomic evaporation and single or multiple electron impact ionization, followed by acceleration into a sensing region. Controllability of the ionic charge state is achieved through tuning of electrode potentials that influence the retention time in the ionization region. Barium, lead, and cadmium samples have been used to test the system, with ion currents identified and quantified using a quadrupole mass analyzer. Realization of a clean Ba 2+ ion beam within a bench-top system represents an important technical advance toward the development and characterization of barium tagging systems for neutrinoless double beta decay searches in xenon gas. This system also provides a testbed for investigation of novel ion sensing methodologies for environmental assay applications, with dication beams of Pb 2+ and Cd 2+ also demonstrated for this purpose.

47 OTHER INSTRUMENTATION↗
DOE OSTI2018

Heating Tests of AGR-2 Compact 6-4-1 Loose Kernels and AGR-3/4 Compacts 3-2 and 10-4

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program was established to perform research and development on tristructural isotropic (TRISO)-coated particle fuel to support deployment of a high-temperature gas-cooled reactor (HTGR). The work continues as part of the Advanced Reactor Technologies (ART) TRISO Fuel Program. The overarching goal of the program is to provide a baseline fuel qualification data set to support licensing and operation of an HTGR. To achieve these goals, the program includes the elements of fuel fabrication, irradiation, post-irradiation examination (PIE) and safety/heating testing, fuel performance modeling, and fission product transport (INL 2016). Several fuel irradiation experiments have been performed at the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL), and additional irradiations are planned. These experiments are intended to provide data on fuel performance under irradiation, support fuel fabrication process development, qualify fuel for operating and accident conditions, provide irradiated fuel for accident testing, and support development of fuel performance and fission product transport models.

42 ENGINEERING↗
DOE OSTI2012

AGR-1 Irradiated Compact 6-1-1 PIE Report: Evaluation of As-Irradiated Fuel Performance Using Leach Burn Leach, IMGA, Materialography, and X-ray Tomography

The AGR-1 irradiation test train consisted of six independently monitored capsules. All capsules contained twelve compacts in three stacks of four compacts each [Pope 2010]. Compact 6-1-1 was located in capsule 6, at the top of the test train. The compact was in stack 1 (one of the two stacks closest to the core) in position 1 (at the bottom of the 4 compact stack). Compact 6-1-1 was identified in the fabrication data package as LEU01-46T-Z32 [Hunn et al. 2006]. Details of the compact fabrication and measured properties can be found in that data package. The compact lot was made using TRISO-coated particles designated as AGR-1 Baseline fuel. These particles contained 19.7% enriched mixed uranium oxideuranium carbide (UCO) kernels with a O/U atomic ratio of ~1.36 and a C/U atomic ratio of ~0.33 [Pope 2010]. Details of the coated particle fabrication and measured properties can be found in the data package for the AGR-1 Baseline particles [Hunn and Lowden 2006]. The compact was irradiated at the INL Advanced Test Reactor (ATR) to an average calculated burn-up of 15.1% fissions per initial metal atom (FIMA) and an average calculated fast fluence of 3.0 x 10 25 n/m 2 [Sterbentz 2011] at a calculated volume averaged and time averaged temperature of approximately 1111°C [Hawkes 2012]. Capsules were disassembled and preliminary compact and graphite holder inspection was performed [Demkowicz et al. 2011]. Compact 6-1-1 was shipped to ORNL and post-irradiation examination (PIE) was performed according to examination plan PLN-3624 [Demkowicz 2010-2]. All the examination plan objectives listed below were completed except item 6, which was deleted due to limited funding. 1. Determine the inventory of fission products retained in the compact matrix, but not contained by the SiC and determine the number of exposed kernels or particles with defective SiC layers using the deconsolidation-leach-burn-leach method. 2. Gamma count all particles to quantify 137Cs inventory for comparison with calculated values and gamma spectrometric measurements made on the whole compact. Also, attempt to quantify the inventory of any other fission products for which gamma emissions are observed. 3. Characterize each particle based on the 137 Cs/ 144 Ce ratio to identify particles with low Cs retention that may be the result of a failed SiC coating. 4. Gamma count a subset of particles to determine the 110m Ag inventory for comparison with calculated values and gamma spectrometric measurements made on the whole compact. Also, attempt to quantify the inventory of other fission products observed. 5. Microscopically inspect selected particles to determine the condition of the kernels and coatings. This could include kernel migration, coating microstructures and defects, fission product attack of the SiC layer, and fission product migration in the coating layers. 6. Perform high temperature fission product release tests on a subset of particles in order to calibrate the collection efficiency of the Core Conduction Cooldown Test Facility (CCCTF) deposition cup for selected fission products. Gamma count particles before and after the heating test to quantify the released fission product fractions. 7. Archive remaining particles for later use. Compact 6-1-1 LBL began with compact deconsolidation in the Building 3525 hot cell, followed by two nitric acid leaches of the particles and matrix residue using a Soxhlet extraction apparatus. The matrix residue was then burned and the ash was leached twice with boiling nitric acid. The deconsolidated particles underwent an IMGA analysis to determine gamma-emitting isotope inventories for individual particles; the most conspicuous elements with reduced inventories, indicating fission product release, were silver and europium. During the IMGA analysis, particles with low silver were identified and a subset of these was set aside for materialography and x-ray. After the completion of the IMGA study, about 10% of the particles (421) were set aside as an archive and the remainder burned back and leached twice in the Soxhlet apparatus. Figure 1 shows a flow chart detailing the PIE of Compact 6-1-1.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

PIE on Five Irradiated AGR-1 Compacts

Post-irradiation examination (PIE) is in progress on coated particle fuel compacts from the first Advanced Gas Reactor irradiation experiment (AGR-1). PIE has been completed on five compacts at the Oak Ridge National Laboratory (ORNL). Three compacts (Compacts 6-1-1, 4-4-2, and 5-2-3) were examined in the as-irradiated condition and two compacts (Compacts 6-4-3 and 3-3-2) were subjected to safety testing at 1600°C, followed by post-safety testing examination similar to the PIE performed on the as-irradiated compacts. All compacts were subjected to a standard set of analyses that included the following: (1) detection of exposed fission products by Deconsolidation-Leach-Burn-Leach (DLBL), (2) measurement of gamma-emitting fission product inventory using the Irradiated Microsphere Gamma Analyzer (IMGA), and (3) microstructural examination by x-ray tomography and materialography. The equipment and methods used are reported in detail in ORNL/TM-2012/233, AGR-1 Irradiated Compact 6-1-1 PIE Report, and have been summarized in the next section of this report. This report also provides a brief summary of the results of the completed PIE. Preliminary trends are discussed in this report, but final conclusions will require the support of additional data and further analysis. A separate report is being generated for each AGR-1 compact after PIE is complete that describes in detail the examinations performed and fully documents the data obtained. A final AGR-1 PIE summary report will compile all available data and provide a more comprehensive analysis of the AGR-1 fuel performance. In general, it was observed that the tristructural isotropic (TRISO) coated particles remained intact during irradiation and subsequent safety testing to 1600°C. No significant gaseous 85 Kr release was detected during the irradiation or high temperature safety testing, which would be indicative of simultaneous failure of all three outer layers of the TRISO coating. Only three particles with defective SiC layers were detected in the five compacts included in this summary. These particles were detected because they released a significant fraction of their cesium inventory, which was generally well retained. Two particles from Compact 5-2-3 released cesium during irradiation and one particle from Compact 3-3-2 released cesium during safety testing. However, it is evident that some fission products were released through intact SiC during irradiation. Most of these fission products were trapped in the compact matrix, where they could be analyzed by PIE. In terms of fractional release, the main fission products detected outside intact SiC were silver, palladium, and europium. The main impact of irradiation on the microstructure of the TRISO-coating layers was densification and shrinkage of the buffer layer, which often led to buffer fracture and subsequent enhanced swelling of the kernel. Some inner pyrocarbon (IPyC) layers were cracked by the radiation-induced changes in the buffer layer, and on rare occasion these cracks propagated into the SiC. In the two particles that released cesium during irradiation, these SiC cracks penetrated through the entire layer. The particle from Compact 3-3-2 that released cesium was not directly related to irradiation damage. Its SiC layer was determined to be porous as a result of a fabrication defect that occurs when a particle is over-fluidized in the coater, causing it to pick up carbon soot on the surface of the IPyC prior to SiC deposition. This porous SiC layer remained sufficiently intact during irradiation to retain cesium, but finally cracked during high temperature testing. Analysis of fission products remaining in the coated particles showed that the retention of silver varied dramatically from particle-to-particle within each compact. Examination with the IMGA determined that some particles contained a level of 110m Ag equivalent to that calculated to have been generated over the three-year irradiation period, while other particles contained too little to be detected by the gamma analysis. The mechanism for silver release is still being investigated. Attempts to correlate the varied release behavior to obvious radiation-induced changes in the coating layers have not been successful. Scanning electron microscopy (SEM) has provided some evidence that particles with higher silver release also show greater penetration of palladium silicide into the SiC layer, and this should be studied further.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

AGR-3/4 TRISO Fuel Compact Ceramography

The combined third and fourth irradiation in the Advanced Gas Reactor (AGR) program (AGR-3/4) contained tristructural isotropic (TRISO)-coated particle fuel and designed-to-fail (DTF) fuel particles. The DTF particles were only coated with highly anisotropic pyrocarbon (PyC) so they would purposely fail during the AGR-3/4 irradiation and provide a source of fission products for measurement. To observe the post-irradiation morphology of these DTF particles and the TRISO-coated particles, three AGR-3/4 compacts were mounted in epoxy, sectioned above their centerlines, and ground/polished. Three rounds of grinding/polishing and optical microscopy were performed so that the particles could be observed at multiple planes. Each compact contained approximately 1,872 TRISO particles and exactly 20 DTF particles. The three AGR-3/4 compacts examined covered a wide temperature range and featured both the hottest average irradiation temperature (1375°C) and the coldest/lowest burnup (872°C and 5.5% fissions per initial metal atom) of any compacts to undergo post-irradiation examination (PIE) in the AGR program to date. A total of 29 DTF particles were located and observed via microscopy across the three compacts. All the DTF particles observed via microscopy had completely failed PyC coatings. Some different DTF kernel and PyC morphologies were observed that have been attributed to the differences in irradiation temperature and/or burnup. At low and medium irradiation temperatures (roughly 850 to 1050°C), it appears irradiation-induced dimensional changes in the DTF PyC caused it to fracture and fold in on itself, and the kernel deformed to accommodate this PyC deformation. The extent of DTF kernel deformation to accommodate DTF PyC buckling tended to be greater for the medium-temperature fuel compared to the cold fuel. In the high-irradiation-temperature compact (1375°C), the DTF PyC layer appears to have completely reacted chemically with the kernel material. The only material surrounding the DTF particles in this hot compact resembles the compact matrix material. Observations of the TRISO-coated driver fuel particles and fuel compact graphitic matrix were also made. The TRISO particle kernels in the medium-temperature compact (1047°C), and many in the high-temperature compact (1375°C), showed morphologies consistent with what was seen in AGR-1 and AGR-2. In other TRISO particles from the high-temperature compact, a spatial gradient in the kernel appearance was evident. Here the cool side of the kernel (away from the center of the compact) had a much darker appearance (like that of the buffer and pyrocarbon). This is the first observation of this kind of kernel spatial gradient in AGR fuel. It is believed that the high irradiation temperature of that compact (an average of 1375°C) activates or accelerates an unidentified chemical reaction, and the temperature gradient within the fuel causes a sharp spatial gradient. The low-temperature/low-burnup compact also showed unique kernel morphologies where the TRISO-coated particles had clearly distinguishable oxide and carbide phases at the center of the kernel, an oxide rind surrounding this, and remnants of a carbide skin surrounding the oxide rind. These are features observed in the as-fabricated fuel that are no longer present in higher burnup fuel. Occasional gaps between the outer pyrolytic carbon (OPyC) and graphitic matrix material were observed in all compacts. Gaps were most often found in the small, matrix-filled spaces between adjacent particles. Finally, buffer morphologies in AGR-3/4 TRISO particles were like those observed in AGR-1 and AGR-2, and the buffer fracture frequencies in AGR-3/4 and AGR-2 TRISO particles were plotted versus fast neutron fluence and irradiation temperature. In AGR-3/4 the buffer fracture rate was highest (23%) for an irradiation temperature of 1047°C and a fast fluence of 5.18E25 n/m 2 . Increasing the irradiation temperature to 1375°C reduced the buffer fracture rate to 14%. Reducing the irradiation temperature to 872°C and the fast fluence by a factor of 3 reduced the buffer fracture rate to 7%. The temperature and fluence dependencies observed for AGR-3/4 buffer fracture were consistent with those observed in AGR-2. Lower fluences and/or higher temperatures significantly reduced the buffer fracture frequencies. Lower buffer fracture frequencies were also observed at lower temperatures as long as the fluence was also significantly reduced. This is believed to be due to lower fluence resulting in less irradiation-induced dimensional change (shrinkage) of the buffer, and a higher temperature promoting enhanced creep relaxation of stresses within the buffer, leading to less buffer fracture.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗