• 한국분말재료학회지
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• 제15권2호
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• pp.87-94
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• 2008

In order to investigate a nitriding process of spent oxide fuel and the subsequent change in thermal properties after nitriding, simulated spent fuel powder was converted into a nitride pellet with simulated fission product elements through a carbothermic reduction process. Nitriding rate of simulated spent fuel was decreased with increasing of the amount of fission products. Contents of Ba and Sr in simulated spent fuel were decreased after the carbothermic reduction process. The thermal conductivity of the nitride pellet was decreased by an addition of fission product element but was higher than that of the oxide fuel containing fission product elements.

• Nuclear Engineering and Technology
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• 제35권4호
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• pp.309-317
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• 2003

The mass balance of the unit processes of the Advanced spent fuel Conditioning Process was calculated to obtain basic information. Based on this mass balance, the changes in decay heat and radioactivity of the spent fuel due to the metallization in the high temperature molten salt system were estimated. The decay heat and the radioactivity were calculated by using the ORIGEN2 computer code, and the result showed that the decay heat and the radioactivity of the metallized spent fuel ingot were $24.27\%\;and\;24.24\%$, respectively, compared to those of oxide spent fuel.

• Nuclear Engineering and Technology
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• 제45권3호
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• pp.377-384
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• 2013

During the last three decades, South Korean nuclear power plants have discharged about 5,950 tons of spent fuel and the maximum burn-up reached 55 GWd/MTU in 2002. This study was performed to support the development of Korean dry spent fuel storage alternatives. First, we chose V5H-$17{\times}17$ and KSFA-$16{\times}16$ as representative domestic spent fuels, considering current accumulation and the future generation of the spent fuels. Examination reveals that their average burn-ups have already increased from 33 to 51 GWd/MTU and from 34.8 to 48.5 GWd/MTU, respectively. Evaluation of the fuel characteristics shows that at the average burn-up of 42 GWd/MTU, the oxide thickness, hydrogen content, and hoop stress ranged from $30{\sim}60{\mu}m$, 250 ~ 500 ppm, and 50 ~ 75 MPa, respectively. But when burn-up exceeds 55 GWd/MTU, those characteristics can increase up to 100 ${\mu}m$, 800 ppm, and 120 MPa, respectively, depending on the power history. These results demonstrate that most Korean spent nuclear fuels are expected to remain within safe bounds during long-term dry storage, however, the excessive hoop stress and hydrogen concentration may trigger the degradation of the spent fuel integrity early during the long-term dry storage in the case of high burn-up spent fuels exceeding 45 GWd/MTU.

• 방사성폐기물학회지
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• 제8권1호
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• pp.77-84
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• 2010

상용원자로에서 발생하는 산화물 사용후핵연료의 부피감용과 재활용을 위하여 산화물을 금속으로 환원시키는 공정에 대한 연구가 수행되어 왔다. 다양한 환원법 중에서, 한국원자력연구원은 LiCl-$Li_2O$ 용융염을 반응매질로 사용하는 전해환원공정을 현재 개발 중이다. 파이로 공정의 전단부에 해당하는 전해환원 공정은 PWR 산화물 연료 주기를 소듐냉각 고속로의 금속연료 주기에 연결시켜 준다. 이 논문은 금속전환 공정을 개발/개선하고, 용량 증대를 수행한 한국원자력연구원의 노력을 요약한다.

• Nuclear Engineering and Technology
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• 제43권4호
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• pp.317-328
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• 2011

Pyroprocessing technology was developed in the beginning for metal fuel treatment in the US in the 1960s. The conventional aqueous process, such as PUREX, is not appropriate for treating metal fuel. Pyroprocessing technology has advantages over the aqueous process: less proliferation risk, treatment of spent fuel with relatively high heat and radioactivity, compact equipment, etc. The addition of an oxide reduction process to the pyroprocessing metal fuel treatment enables handling of oxide spent fuel, which draws a potential option for the management of spent fuel from the PWR. In this context, KAERI has been developing pyroprocessing technology to handle the oxide spent fuel since the 1990s. This paper describes the current status of pyroprocessing technology development at KAERI from the head-end process to the waste treatment. A unit process with various scales has been tested to produce the design data associated with the scale up. A performance test of unit processes integration will be conducted at the PRIDE facility, which will be constructed by early 2012. The PRIDE facility incorporates the unit processes all together in a cell with an Ar environment. The purpose of PRIDE is to test the processes for unit process performance, operability by remote equipment, the integrity of the unit processes, process monitoring, Ar environment system operation, and safeguards related activities. The test of PRIDE will be promising for further pyroprocessing technology development.

• Nuclear Engineering and Technology
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• 제17권3호
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• pp.185-192
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• 1985

가압경수로에서 나오는 사용후 핵연료의 fissile 양은 CANDU형 원자로에 쓰는 천연우라늄의 농축도 보다 높다. 따라서 핵연료 활용을 다양화하고 점차 누적되고 있는 가압경수로의 사용후 핵 연료의 저장문제를 부분적으로나마 해결하기 위하여, 가압경수로의 사용후 책 연료를 CANDU 형 원자로에 사용하는 방안을 검토 하였다. 가압경수로에서 나온 사용후 핵 연료에서 가공되는 혼합핵연료(Mixed Oxide Fuel)를 CANDU형 원자로에 장전하였을 경우, WIMS/D 코드를 이용하여 핵적특성을 분석하였다. 그리고 본 분석에서는 현 CANDU형 원자로의 반응도 조절장치를 변경시키지 않고 혼합핵 연료를 CANDU형 원자로에 사용할 수있는 방안만 조사하였다.

• 한국분말재료학회지
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• 제7권3호
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• pp.123-130
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• 2000

The simulated DUPIC fuel provides a convenient way to investigate fuel properties and behaviours such as thermal conductivity, thermal expansion, fission gas release, leaching and so on. Several pellets simulating the composition and microstructure of the DUPIC fuel were fabricated from resintering powder through the OREOX process of the simulated spent fuel pellets, which were prepared from the mixture of stable forms of constituent nuclides. This study describes the powder treatment, OREOX, compaction and sintering to fabricate simulated DUPIC fuel using the simulated spent fuel. The homogeneity of additives in the powder was observed after attrition milling. The microstructure of the simulated spent fuel was in agreement with the previous studies. The densities and the grain size of simulated DUPIC fuel was pellets are higher than those of simulated spent fuel pellets. Small metallic precipitates and oxide precipitates were observed on matrix grain boundaries.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2005년도 Proceedings of The 6th korea-china joint workshop on nuclear waste management
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• pp.86-91
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• 2005

A metal product obtained from an electrolytic reduction process, possesses less volume and radioactivity than those of the unprocessed spent oxide fuels. The chemical composition of the metal product varies according to the process condition. In this work, a basic study was performed to evaluate the chemical forms of the spent oxide fuel components in an electrolytic reduction process with the operation conditions. One of the most important operation conditions is the cell potential applied for the reduction cell. It is expected that $PU_{2}O_3$ is difficult to reduce even though the cell potential is negative enough to reduce the lithium oxide when the activity of $Li_{2}O$ exceeds 0.003. The reduction of actinide oxides via the reduction of $Li_{2}O$ is assumed to have a greater reduction yield than a direct reduction of the actinide oxides.

• 방사성폐기물학회지
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• 제19권2호
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• pp.255-266
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• 2021

Solid-state mechanochemical reduction combined with subsequent melting consolidation was suggested as a technical option for the oxide reduction in pyroprocessing. Ni ingot was produced from NiO as a starting material through this technique while Li metal was used as a reducing agent. To determine the technical feasibility of this approach for pyroprocessing, which handles spent nuclear fuels, thermodynamic calculations of the phase stabilities of various metal oxides of U and other fission elements were made when several alkaline and alkali-earth metals were used as reducing agents. This technique is expected to be beneficial, not only for oxide reduction but also for other unit processes involved in pyroprocessing.

• 에너지공학
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• 제13권4호
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• pp.259-266
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• 2004

세라믹핵연료의 비열기구는 격자 진동 비열, 팽창 비열, 전도전자 및 결함비열 그리고 과잉비열로 구성된다. 비열을 표현하는 모델은 정압비열 항과 팽창비열 항 그리고 결함비열 항으로 구성된다. 본 연구에서는 세라믹 핵연료의 실험자료 또는 발표된 자료들을 종합 분석하였으며, 가장 적합한 모델을 추천하였다. $UO_2$, (U, Pu)혼합핵연료 및 사용후 핵연료의 비열 자료들이 분석되었다. 사용 후 핵연료의 경우 모의 핵연료의 비열로 대신하였다.