• Nuclear Engineering and Technology
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• v.40 no.3
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• pp.191-198
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• 2008

The influence of fission products' contents on the DUPIC fuel powder and pellet properties was experimentally evaluated using SIMFUEL as a surrogate for actual spent PWR fuel due to the high radioactivity of spent fuel. Pure $UO_2$ and SIMFUEL pellets with fission products equivalent to a burn-up of 35,000 MWd/tU and 60,000 MWd/tU were used as impurities in this study. The specific surface area of the powder milled after the OREOX treatment increased and resulted in sintered pellets with a theoretical density (TD) higher than 95%, regardless of the impurity contents. However, the grain size of the sintered pellets decreased with the increasing impurity contents. As a result of the dissolved oxides in $UO_2$ from the impurity groups, the specific surface area of the OREOX powder increased with an increase of the impurities. The grain size of the sintered pellets was significantly decreased by the metallic and oxide precipitates.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.4
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• pp.343-353
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• 2017

Processing and equipment were tailored for engineering scale fabrication of $UO_2$ porous pellets, a feed material for the electrolytic reduction process in the PRIDE (PyRoprocessing Integrated DEmonstration) facility at KAERI (Korea Atomic Energy Research Institute). The starting materials, $UO_2$ powder and pre-milled surrogate oxide powders, were proportioned to simulate the chemical composition of spent fuel (so-called Simfuel). The Simfuel powders were homogenized by mixing, compacted into a pellet shape, and finally heat treated using a tumbling mixer, rotary press, and sintering furnace. After sintering at $1450^{\circ}C$ for 24 h in $4%\;H_2-Ar$, the average bulk density of the $UO_2$ Simfuel pellets was $6.89g{\cdot}cm^{-3}$, which meets the standard of the following electrolytic reduction process. In addition, the results of a microstructural analysis demonstrated that the sintered Simfuel $UO_2$ porous pellets accurately simulate the properties of spent fuel in terms of the formation of second phases. These results provide essential information for the massive fabrication of $UO_2$ porous pellets for engineering scale pyroprocessing research.

• Analytical Science and Technology
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• v.13 no.6
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• pp.781-788
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• 2000

Tellurium in simulated nuclear spent fuels (SIMFUEL) has been determined by hydride generation-inductively coupled plasma atomic emission spectrometry (HG-ICP-AES). Parameters such as concentrations of HCl and $NaBH_4$, flow rate of HCl and $NaBH_4$ were optimized and then the effects of U, Mo, Pd, Rh and Ru on the Te intensity were investigated. A thiourea as a masking agent was used to eliminate or minimize such interferences specially caused by palladium. Tellurium was measured by HG-ICP-AES and ICP-MS after separation of tellurium from SIMFUEL with cation exchange chromatography. The relative deviation between direct measurement and separation method was less than 6% based on the data by ICP-MS.

• Analytical Science and Technology
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• v.13 no.4
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• pp.446-452
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• 2000

The simulated nuclear spent fuel (SIMFUEL) containing the platinum group elements which will not be dissolved in a nitric acid was completely dissolved with a acid digestion bomb. The metallic elements separated in the SIMFUEL were measured by inductively coupled plasma atomic emission spectrometry (ICP-AES). Because the peaks of metallic elements were spectrally interfered by uranium spectrum, uranium and metallic elements were separated by cation exchange resin for Mo, Pd, Rh and Ru and by anion exchange resin for Ba, Ce, La, Nd, Rh, Sr, Y and Zr, respectively. The recovery of Mo, Pd, Rh and Ru after separation by cation exchange chromatography found to be 99-103% and anion exchange separation showed 96.5-107% of recovery except Y with the simulated solution whose concentration was similar to the spent nuclear fuel. The relative standard deviation of this method showed 1.3-6.7% in the SIMFUEL whose concentrations of metallic elements were between several $10^2-10^3$ppm.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.553-558
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• 1996

가압 경수로형 핵연료에 대한 장기 저장거동을 연구하기 위하여 모의 사용후핵연료(SIMFUEL) 및 조사 핵연료에 대한 산화시험을 공기중에 수행하였다. 연소도가 15,33 및 50 GWD/MTU로 모의한 핵연료를 300-375$^{\circ}C$ 구간에서 산화 시험한 결과, 모의 사용후핵연료는 미조사 $UO_2$시편과 같이 S-형 곡선의 무게증가 특성을 보여 주었으며, 미조사 $UO_2$시편에 비해 산화가 느리게 일어났으며, 모의 사용후핵연료는 연소도가 높을수록 산화속도가 느리다. 고리 2호기에서 2주기 연소한 우라늄 및 가돌리니아 핵연료를 275$^{\circ}C$에서 산화 시험한 결과, 조사 $UO_2$는 연소도가 증가할수록 산화가 느리게 일어나며, 우라늄 핵연료는 가돌리니아 핵연료에 비해 산화가 빨리 일어난다.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2007.05a
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• pp.147-148
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• 2007

• Proceedings of the Korean Ceranic Society Conference
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• 2003.04a
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• pp.92.1-92
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• 2003

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.302.1-302
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• 2002

• Proceedings of the Korean Nuclear Society Conference
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• 2003.10a
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• pp.342.1-342.1
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• 2003

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2010.05a
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• pp.265-266
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• 2010