Performance of AGR-1 high-temperature reactor fuel during post-irradiation heating tests

Morris, Robert; Baldwin, Charles A.; Demkowicz, Paul A.; Hunn, John D.; Reber, E. L.

doi:10.1016/j.nucengdes.2016.04.031

Cited by 52 publications

(48 citation statements)

References 9 publications

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“…In perspective of the overall AGR-1 fuel performance, the particles discussed in this study represent SiC failure fractions of 1.3×10 -5 during irradiation, 9.1×10 -5 during 1600°C safety testing, 4.8×10 -4 during 1700°C safety testing, and 1.4×10 -3 during 1800°C safety testing. It should also be noted that compacts were held for 300 or more hours at the maximum safety-test temperature [Morris et al, 2014] and time was a factor in the number of particles that released cesium due to SiC degradation. Failure fractions at 1700°C and 1800°C would have been lower for shorter test periods.…”

Section: Discussionmentioning

confidence: 99%

“…Cesium release from four particles with failed SiC was observed in the PIE of the graphite holders that surrounded the compacts in the irradiation test capsules [Demkowicz et al, 2014]; three of these particles were recovered for further analysis discussed below. Three particles released cesium during 1600°C safety testing, and numerous particles released cesium during safety testing at 1700 and 1800°C [Morris et al, 2014]; many of these were also recovered for analysis.…”

Section: Introductionmentioning

confidence: 99%

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Detection and analysis of particles with failed SiC in AGR-1 fuel compacts

Hunn¹,

Baldwin²,

Gerczak³

et al. 2016

Nuclear Engineering and Design

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Section: Discussionmentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

Detection and analysis of particles with failed SiC in AGR-1 fuel compacts

Hunn¹,

Baldwin²,

Gerczak³

et al. 2016

Nuclear Engineering and Design

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“…As part of the AGR-1 campaign, out-of-pile safety testing of individual compacts was performed to understand fuel performance at accident conditions and to test the margins of fuel performance. Individual compacts were exposed to temperatures ranging 1600-1800°C for exposure times of 300 hours or more while periodically measuring the release of select fission products, notably 110m Ag, 154 Eu, and 90 Sr [39]. These fission products were selected as they are radiologically [39].…”

Section: Irradiated Triso Fuel Performancementioning

confidence: 99%

“…Individual compacts were exposed to temperatures ranging 1600-1800°C for exposure times of 300 hours or more while periodically measuring the release of select fission products, notably 110m Ag, 154 Eu, and 90 Sr [39]. These fission products were selected as they are radiologically [39].…”

Section: Irradiated Triso Fuel Performancementioning

confidence: 99%

SiC layer microstructure in AGR-1 and AGR-2 TRISO fuel particles and the influence of its variation on the effective diffusion of key fission products

Gerczak

Hunn

Lowden

et al. 2016

Journal of Nuclear Materials

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Tristructural isotropic (TRISO) coated particle fuel is a promising fuel form for advanced reactor concepts such as high temperature gas-cooled reactors (HTGR) and is being developed domestically under the US Department of Energy's Nuclear Reactor Technologies Initiative in support of Advanced Reactor Technologies. The fuel development and qualification plan includes a series of fuel irradiations to demonstrate fuel performance from the laboratory to commercial scale. The first irradiation campaign, AGR-1, included four separate TRISO fuel variants composed of multiple, laboratory-scale coater batches. The second irradiation campaign, AGR-2, included TRISO fuel particles fabricated by BWX Technologies with a larger coater representative of an industrial-scale system. The SiC layers of as-fabricated particles from the AGR-1 and AGR-2 irradiation campaigns have been investigated by electron backscatter diffraction (EBSD) to provide key information about the microstructural features relevant to fuel performance. The results of a comprehensive study of multiple particles from all constituent batches are reported. The observations indicate that there were microstructural differences between variants and among constituent batches in a single variant. Insights on the influence of microstructure on the effective diffusivity of key fission products in the SiC layer are also discussed.

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“…3.1.1. TRISO-coated fuel particles TRISO-coated fuel particle consists of a fissionable fuel kernel, which is surrounded by four coating layers [8,9]. Those four coating layers are porous pyrolytic carbon (PyC) buffer, dense inner pyrolytic carbon (IPyC), chemically vapour deposited silicon carbide (SiC), and dense outer pyrolytic carbon (OPyC), which are respectively designed from the kernel to the outside of each fuel particle as in Fig.…”

Section: Htgr Inherent Safety Featuresmentioning

confidence: 99%

The Evaluation of the High Temperature Gas Cooled Reactor Safety to Fulfill the Requirement of the Next Generation Nuclear

Purba

Waskita

Tjahyani

2020

Jurnal Pengembangan Energi Nuklir

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THE EVALUATION OF THE HIGH TEMPERATURE GAS COOLED REACTOR SAFETY TO FULFILL THE REQUIREMENT OF THE NEXT GENERATION NUCLEAR. High temperature gas cooled reactor (HTGR) has been considered to be the most promising option to meet energy demands in the future. It has also been selected as the next generation nuclear plant. The primary safety requirement of the next generation nuclear plant design is to limit radioactive material releases to practically eliminate the need for public evacuation or sheltering beyond the exclusion area boundary. The purpose of this study is to evaluate the safety design of HTGRs in order to fulfill the requirement of the next generation nuclear plant. To achieve this objective, inherent safety features, fundamental safety functions, and confinement functions realized into the design of HTGRs are comprehensively evaluated. It is found that design provisions of HTGRs can fulfill the intention of keeping radionuclides at their original sources. The layers of the coated fuel particles are very robust to retain nuclear fission products for all foreseeable reactivity events. There will be no possibility of radioactive materials to be released even though related safety systems and operator intervention are not involved in the recovery actions. This design has complied with the requirement of the next generation nuclear plant, which is to practically eliminate the need for public evacuation or sheltering beyond the exclusion area boundary. ABSTRAK EVALUASI KESELAMATAN REAKTOR BERPENDINGIN GAS TEMPERATUR TINGGI DALAM MEMENUHI PERSYARATAN PEMBANGKIT NUKLIR MASA DEPAN.Reaktor berpendingin gas temperatur tinggi diprediksi akan menjadi pilihan yang menjanjikan untuk memenuhi kebutuhan energi masa depan. Persyaratan utama keselamatan bagi pembangkit nuklir masa depan adalah pembatasan pada lepasan material radioaktif sehingga secara praktis evakuasi dan sheltering diluar exclusion area boundary dapat dihilangkan. Tujuan penelitian ini adalah mengevaluasi desain keselamatan HTGR dalam memenuhi persyaratan pembangkit nuklir masa depan. Untuk mencapai tujuan ini maka fitur keselamatan melekat, fungsi keselamatan fundamental, dan fungsi pengungkungan yang diimplementasikan pada desain HTGR akan dievaluasi. Hasil evaluasi menunjukkan bahwa desain HTGR dapat mengungkung material radioaktif tetap berada pada sumbernya. Lapisan partikel bahan bakar mampu menahan produk fisi untuk setiap kejadian reaktifitas. Pelepasan material radioaktif tidak akan pernah terjadi meskipun intervensi sistem keselamatan dan operator tidak terlibat dalam aksi pemulihan. Dengan demikian, desain HTGR ini telah memenuhi persyaratan pembangkit nuklir masa depan yaitu tidak diperlukannya evakuasi dan shelter di luar exclusion area boundary. Kata kunci: Reaktor berpendingin gas temperatur tinggi, fitur keselamatan melekat, fungsi keselamatan fundamental, fungsi pengungkungan, pembangkit nuklir masa depan Keywords:high temperature gas cooled reactor inherent safety features fundamental safety functions confinement functions next gener...

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Performance of AGR-1 high-temperature reactor fuel during post-irradiation heating tests

Cited by 52 publications

References 9 publications

Detection and analysis of particles with failed SiC in AGR-1 fuel compacts

Detection and analysis of particles with failed SiC in AGR-1 fuel compacts

SiC layer microstructure in AGR-1 and AGR-2 TRISO fuel particles and the influence of its variation on the effective diffusion of key fission products

The Evaluation of the High Temperature Gas Cooled Reactor Safety to Fulfill the Requirement of the Next Generation Nuclear

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