Analysis of Fission Products on the AGR-1 Capsule Components

Demkowicz, Paul A.; Harp, Jason; Winston, Philip; Ploger, Scott

doi:10.2172/1097137

Cited by 7 publications

(16 citation statements)

References 11 publications

(13 reference statements)

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“…Fission products in the deconsolidation and leach solutions from Compact 5-3-1 expressed as decay-corrected activity (Bq) and compact fraction (based on calculated inventories [Sterbentz 2013] Uranium inventory in all of the DLBL solutions was very low (Table 4), indicating that no kernels were exposed during any of the leaching steps. This is consistent with data from the AGR-1 irradiation, indicating that there were no particles with failed TRISO layers in Capsule 5 (Collin 2015), as well as the observation that there were no elevated levels of cesium found adjacent to Compact 5-3-1 in the graphite holder from Capsule 5 (Demkowicz et al 2013) that might indicate particles with failed SiC.…”

Section: Deconsolidation-leach-burn-leachsupporting

confidence: 89%

AGR-1 Compact 5-3-1 Post-Irradiation Examination Results

Demkowicz

Harp

Winston

et al. 2016

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Destructive post-irradiation examination was performed on Advanced Gas Reactor (AGR)-1 fuel Compact 5-3-1, which was irradiated to a final compact-average burnup of 16.93% fissions per initial metal atom and a time-average, volume-average temperature of 1040°C. Analysis of this compact focused on characterizing the extent of fission product release from the particles, examining particles to determine the condition of the kernels and coating layers, and experimentally verifying the fuel burnup. Work included deconsolidation of the compact and leach-burn-leach analysis, visual inspection and gamma counting of individual particles, measurement of isotopic inventories in the fuel by mass spectrometry and subsequent burnup calculations, analyzing the fission product inventory in the kernels and coating fragments separately in a small number of particles, metallurgical preparation of selected particles, and examination of particle cross-sections with optical microscopy. Deconsolidation-leach-burn-leach (DLBL) analysis revealed no particles with failed tristructural isotropic or failed silicon carbide (SiC) layers (as indicated by a very low uranium inventory in all of the leach solutions). The DLBL data also provided information on the inventory of fission products retained in the compact outside of the SiC layers. The total fraction of the predicted Ag-110m inventory in the DLBL solutions was 9.07 × 10-3 , indicating significant release from the particles and retention in the compact matrix or outer pyrolytic carbon during irradiation. Cesium inventory was very low; the fraction of the compact inventory in the DLBL solutions was 2.45 × 10-6 and 6.46 × 10-6 for Cs-134 and Cs-137, respectively. Cerium and europium were found at levels greater than the equivalent inventory in one particle, and strontium levels were almost as high, indicating the release of small quantities through intact SiC. The data also revealed that the Pd-105 inventory in the compact outside of SiC was approximately 0.5% of the predicted compact inventory, indicating release through intact coatings and retention in the matrix or outer pyrolytic carbon at nearly the same level as silver. Jeffrey Berg (technical direction and data analysis of deconsolidation-leach-burn-leach results), Dr. Jeffrey Giglio and Daniel Cummings (inductively coupled plasma-mass spectroscopy experiments and data analysis), Brian Storms (particle gamma counting and Sr-90 analysis), and Brian Frickey and Cad Christensen (preparation of ceramographic mounts).

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Section: Deconsolidation-leach-burn-leachsupporting

confidence: 89%

AGR-1 Compact 5-3-1 Post-Irradiation Examination Results

Demkowicz

Harp

Winston

et al. 2016

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“…It retains gases and most metallic fission products below 1600 C. It has long been known, however, that silver exhibits relatively high release from intact SiC-coated TRISO fuel particles if they are maintained at sufficiently high temperatures for sufficiently long times [15][16][17][18][19][20][21][22][23]. The release of radioactive 110m Ag in particular is of considerable concern since this strong ray emitter (half-life of 249.8 days) can plate out on the various helium-wetted surfaces in the primary circuit of direct-cycle HTGRs and thus significantly complicate plant operations and maintenance [24][25].…”

Section: Triso-coated Particle Fuelmentioning

confidence: 99%

The measurement of silver diffusivity in zirconium carbide to study the release behavior of 110mAg in the ZrC TRISO-coated nuclear fuel particle

Yang

Allen

2016

Journal of Nuclear Materials

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The tri-structural isotropic (TRISO) coated particle fuel has been developed and used for high temperature gas-cooled reactors (HTGRs). It provides a unique robustness of the first barrier for the fission products. The TRISO fuel particle has typically consisted of a UO 2 or UCO kernel, surrounded by successive layers of porous carbon, dense inner pyrocarbon, silicon carbide, and dense outer pyrocarbon. During operation, however, the SiC layer has been known to release radioactive silver 110m Ag which makes maintenance more difficult and thus costly.Zirconium carbide has been considered as a promising alternative to the SiC fission product barrier. ZrC exhibits high temperature stability and possibly possesses superior Pd resistance, while the retention properties especially for silver have not been adequately studied.To help elucidate the diffusive behavior of silver in the ZrC coating of the TRISO-coated particle, a new diffusion experimental technique, called the encapsulating source method, has been developed by constructing a constant source diffusion couple between ZrC and Ag gas originated from Zr-Ag solid solution. SEM, WDS, EBSD Scanning electron microscopy (SEM), wavelength-dispersive X-ray spectroscopy (WDS), electron backscatter diffraction (EBSD) and optical methods were used to analyze the diffusion couple annealed at 1500 C. The resultant diffusion coefficient of Ag in single-crystalline ZrC 0.84 at 1500 C was experimentally determined to be about 2.8 (± 1.2) × 10 -17 m 2 /s.

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“…The total quantity of Ag-110m found outside of the SiC layer was 3.3% of the compact inventory, equivalent to the inventory in 134 Compact 4-1-1 particles, clearly indicating release through intact coatings. Note that 12.5% of the total Capsule 4 inventory of Ag-110m was measured outside of the compacts on the various capsule components (Demkowicz et al 2013;Harp, Demkowicz, and Ploger 2012).…”

Section: Ag-110mmentioning

confidence: 99%

“…One possible explanation is that the predicted Ag-110m inventory for Compact 4-1-1 is significantly lower than it should be, and therefore the ratios in Figure 10 are shifted too high and the distribution in reality reflects significant Ag-110m release from many of the particles. However, available AGR-1 PIE data indicate that the predicted Ag-110m compact inventory is reasonably accurate (a total Ag-110m mass balance for Capsule 4, including Ag-110m remaining in the compacts after irradiation as well as found in all of the capsule components, accounts for 93% of the predicted inventory [Demkowicz et al 2013]). Gamma counting of the intact compacts suggests that this compact released very little of its Ag-110m inventory, a while the DLBL data (Table 3) indicate that approximately 3.3% of the total compact Ag-110m inventory was retained in the compact outside of the SiC layers.…”

mentioning

confidence: 98%

AGR-1 Compact 4-1-1 Post-Irradiation Examination Results

Demkowicz

Harp

Winston

et al. 2016

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Destructive post-irradiation examination was performed on AGR-1 fuel Compact 4-1-1, which was irradiated to a final compact-average burnup of 19.4% FIMA (fissions per initial metal atom) and a time-average, volume-average temperature of 1072°C. The analysis of this compact focused on characterizing the extent of fission product release from the particles and examining particles to determine the condition of the kernels and coating layers. The work included deconsolidation of the compact and leach-burn-leach analysis, visual inspection and gamma counting of individual particles, metallurgical preparation of selected particles, and examination of particle cross sections with optical microscopy, electron microscopy, and elemental analysis. Deconsolidation-leach-burn-leach (DLBL) analysis revealed no particles with failed tristructural isotropic (TRISO) or failed SiC layers (as indicated by very low uranium inventory in all of the leach solutions). The total fractions of the predicted compact inventories of fission products Ce-144, Cs-134, Cs-137, and Sr-90 that were present in the compact outside of the SiC layers were <2×10 −6 , based on DLBL data. The Ag-110m fraction in the compact outside the SiC layers was 3.3×10 −2 , indicating appreciable release of silver through the intact coatings and subsequent retention in the outer pyrolytic carbon (OPyC) layers or matrix. The Eu-154 fraction was 2.4×10 −4 , which is equivalent to the inventory in one average particle, and indicates a small but measurable level of release from the intact coatings. Gamma counting of 61 individual particles indicated no particles with anomalously low fission product retention. The average ratio of measured inventory to calculated inventory was close to a value of 1.0 for several fission product isotopes (Ce-144, Cs-134, and Cs-137), indicating good retention and reasonably good agreement with the predicted inventories. Measured-to-calculated (M/C) activity ratios for fission products Eu-154, Eu-155, Ru-106, Sb-125, and Zr-95 were significantly less than 1.0. However, as no significant release of these fission products from compacts was noted during previous analysis of the AGR-1 capsule components, the low M/C ratios are most likely an indication of a bias in the inventories predicted by physics simulations of the AGR-1 experiment. The distribution of Ag-110m M/C ratios was centered on a value of 1.02 and was fairly broad (standard deviation of 0.18, with values as high as 1.42 and as low as 0.68). Based on all data gathered to date, it is believed that silver retention in the particles was on average relatively high, but that the broad distribution in values among the particles represents significant variation in the inventory of Ag-110m generated in the particles. Ceramographic analysis of particle cross-sections revealed many of the characteristic microstructures often observed in irradiated AGR-1 particles from other fuel compacts. However, there was no obvious correlation between the inner particle morphology and the measured silver ...

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Analysis of Fission Products on the AGR-1 Capsule Components

Cited by 7 publications

References 11 publications

AGR-1 Compact 5-3-1 Post-Irradiation Examination Results

AGR-1 Compact 5-3-1 Post-Irradiation Examination Results

The measurement of silver diffusivity in zirconium carbide to study the release behavior of 110mAg in the ZrC TRISO-coated nuclear fuel particle

AGR-1 Compact 4-1-1 Post-Irradiation Examination Results

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