A summary of the results from the DOE advanced gas reactor (AGR) fuel development and qualification program

Petti, David A.

doi:10.2172/1369357

Cited by 14 publications

(7 citation statements)

References 17 publications

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“…The initial fuel evaluated in MiniFuel irradiations is related to the development of UN tristructural isotropic (TRISO) fuel [3]. This fuel form was developed as a high actinide density variation on TRISO fuel, which was developed under the US Department of Energy (DOE) Advanced Gas Reactor program [4][5][6][7][8][9]. This utilization of nitrides in coated particle nuclear fuel is different from historical applications of nitrides for fast reactors as driver fuel [10,11] or as an actinide destruction targets [12,13].…”

Section: Summary Of Irradiation Testmentioning

confidence: 99%

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

Harp

Morris

Petrie

et al. 2020

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Section: Summary Of Irradiation Testmentioning

confidence: 99%

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

Harp

Morris

Petrie

et al. 2020

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“…JAEA has confirmed that a conventional overcoating method by HTTR fuel technology can achieve up to around 33 volumetric percent with keeping less than HTTR fuel design limit, 8 × 10 -5 of as-fabricated SiC-defective fraction (Mizuta, 2017). As a latest overcoating technology in USA, well-qualified compaction technologies with graphite matrix is being developed over 40 volumetric percent in target and ~37 volumetric percent as the current status (Petti, 2017). On the contrary, in case of applying the high burnup TRISO-CFP for the HTTR core, packing fraction around 45 % shall be needed to satisfy the required uranium mass, approximately 13.5 grams of uranium per a fuel compact.…”

Section: Other Requirements For the High Burnup Htgr Fuelmentioning

confidence: 99%

“…UCO TRISO-CFP is one of candidates to prevent from the failure due to internal gas pressure (Petti, 2017). This is being established in worldwide, however, its fabrication technology and its irradiation properties might not be matured comparing with those of UO2 TRISO-CFP, especially in Japan.…”

Section: Other Requirements For the High Burnup Htgr Fuelmentioning

confidence: 99%

Research and development on high burnup HTGR fuels in JAEA

Ueta

Mizuta

Sasaki

et al. 2020

Mechanical Engineering Journal

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JAEA has been progressing to design HTGR fuels for not only small-type practical HTGRs but also VHTR proposed in GIF which can be utilized for various purposes with high-temperature heat at 750 to 950 °C. To increase economy of these HTGRs, JAEA has been upgrading the design method for the HTGR fuel, which can maintain their integrities at the burnup of three to four times higher than that of the conventional HTTR fuel. Design principles and specifications of various concepts of the high burnup HTGR fuels designed by JAEA are reported. As the latest results on post-irradiation examinations of the high burnup HTGR fuel progressing in a framework of international collaboration with Kazakhstan, irradiation shrinkage rate of the fuel compact as a function of fast neutron fluence was obtained at around 100 GWd/thm. Furthermore, the future R&Ds needed for the high burnup HTGR fuel are described based on these experimental results.

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“…The first irradiation, AGR-1, and its associated post-irradiation examination (PIE) and safety testing demonstrated the fuel performance of lab-scale-produced TRISO fuel [1,2]. The AGR-2 irradiation and its associated PIE was designed to demonstrate the performance of fuel compacts fabricated using TRISO particles produced on an engineering scale [3,4]. Additionally, while AGR-1 contained exclusively uranium oxide / uranium carbide (UCO) TRISO kernels, AGR-2 contained a capsule with uranium oxide (UO2) TRISO fuel [5].…”

Section: Introductionmentioning

confidence: 99%

Fission product inventory and burnup evaluation of the AGR-2 irradiation by gamma spectrometry

Harp

Demkowicz

Stempien

2018

Nuclear Engineering and Design

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Gamma spectrometry has been used to evaluate the burnup and fission product inventory of different components from the U.S. Advanced Gas Reactor (AGR) Fuel Development and Qualification Program's second TRISO-coated particle fuel irradiation test (AGR-2). TRISO fuel in this irradiation included both uranium carbide / uranium oxide (UCO) kernels and uranium oxide (UO2) kernels. Four of the 6 capsules contained fuel from the U.S. Advanced Gas Reactor program, and only those capsules will be discussed in this work. The inventories of gamma-emitting fission products from the fuel compacts, graphite compact holders, graphite spacers and test capsule shell were evaluated. These data were used to measure the fractional release of fission products such as Cs-137, Cs-134, Eu-154, Ce-144, and Ag-110m from the compacts. The fraction of Ag-110m retained in the compacts ranged from 1.8% of the predicted inventory to approximately full retention. Additionally, the activities of the radioactive cesium isotopes (Cs-134 and Cs-137) have been used to evaluate the burnup of all U.S. TRISO fuel compacts in the irradiation. The experimental burnup evaluations compare favorably with burnups predicted from physics simulations. Predicted burnups for UCO compacts range from 7.26 to 13.15 % fission per initial metal atom (FIMA) and 9.01 to 10.69 % FIMA for UO2 compacts. Measured burnup ranged from 7.3 to 13.1 % FIMA for UCO compacts and 8.5 to 10.6 % FIMA for UO2 compacts. Results from gamma emission computed tomography performed on the graphite holders that reveal the distribution of different fission products in a component will also be discussed. Gamma tomography of graphite holders was also used to locate the position of TRISO fuel particles suspected of having silicon carbide layer failures that lead to in-pile cesium release.

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A summary of the results from the DOE advanced gas reactor (AGR) fuel development and qualification program

Cited by 14 publications

References 17 publications

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

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

Research and development on high burnup HTGR fuels in JAEA

Fission product inventory and burnup evaluation of the AGR-2 irradiation by gamma spectrometry

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