• Nuclear Engineering and Technology
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• 제43권5호
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• pp.413-420
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• 2011

In Korea, although the concept of dry storage system for PWR spent fuels first emerged in the early 1990s, wet storage inside nuclear reactor buildings remains the dominant storage paradigm. Furthermore, as the amount of discharged fuel from nuclear power plants increases, nuclear power plants are confronted with the problem of meeting storage capacity demand. Various measures have been taken to resolve this problem. Dry storage systems along with transportation of spent fuel either on-site or off-site are regarded as the most feasible measure. In order to develop dry storage and transportation system safety analyses, development of design techniques, full scale performance tests, and research on key material degradation should be conducted. This paper deals with two topics, structural analysis methodology to assess cumulative damage to transportation packages and the effects of an aircraft engine crash on a dual purpose cask. These newly emerging issues are selected from among the many technical issues related to the development of transportation and storage systems of spent fuels. In the design process, appropriate analytical methods, procedures, and tools are used in conjunction with a suitably selected test procedure and assumptions such as jet engine simulation for postulated design events and a beyond design basis accident.

• 분석과학
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• 제16권2호
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• pp.110-116
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• 2003

사용 후 핵연료의 화학특성 연구를 위하여 요오드의 분리와 정량에 관한 연구를 수행하였다. 사용 후 핵연료를 용해시키는 과정에서 핵연료 중에 CsI로 존재하는 요오드가 $I_2$로 산화되어 휘발되지 않도록 질산과 염산의 혼합산 (80:20 mol%)을 이용하여 비휘발성 ${IO_3}^-$­로 안정화시켰다. 2.5 M $HNO_3$ 매질에서 $NH_2OH{\cdot}HCl$을 이용하여 $I_2$로 환원시킨 후 사염화탄소로 추출하여 우라늄과 핵분열생성물로부터 분리, 회수하였다. 0.1 M $NaHSO_3$을 사용하여 요오드를 역추출하였으며 수용액층으로 회수된 요오드를 이온 크로마토그래피로 정량하였다. 방사성 물질 분석에 적합한 이온 크로마토그래피/차폐 시스템을 구성하였으며 42,000~44,000 MWd/MtU 의 연소도를 갖는 사용후핵연료를 대상으로 요오드를 분석한 결과 Origin 2 연소도 전산코드에 의한 계산결과인 $324.5{\sim}343.6{\mu}g/g$와는 -8.3~-0.5%의 편차를 나타내었다.

• 방사성폐기물학회지
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• 제18권spc호
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• pp.21-36
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• 2020

With respect to spent nuclear fuels, disposal containers and bentonite buffer blocks in deep geological disposal systems are the primary engineered barrier elements that are required to isolate radioactive toxicity for a long period of time and delay the leakage of radio nuclides such that they do not affect human and natural environments. Therefore, the thermal stability of the bentonite buffer and structural integrity of the disposal container are essential factors for maintaining the safety of a deep geological disposal system. The most important requirement in the design of such a system involves ensuring that the temperature of the buffer does not exceed 100℃ because of the decay heat emitted from high-level wastes loaded in the disposal container. In addition, the disposal containers should maintain structural integrity under loads, such as hydraulic pressure, at an underground depth of 500 m and swelling pressure of the bentonite buffer. In this study, we analyzed the thermal stability and structural integrity in a deep geological disposal environment of the improved deep geological disposal systems for domestic light-water and heavy-water reactor types of spent nuclear fuels, which were considered to be subject to direct disposal. The results of the thermal stability and structural integrity assessments indicated that the improved disposal systems for each type of spent nuclear fuel satisfied the temperature limit requirement (< 100℃) of the disposal system, and the disposal containers were observed to maintain their integrity with a safety ratio of 2.0 or higher in the environment of deep disposal.

• Nuclear Engineering and Technology
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• 제31권6호
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• pp.83-100
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• 1999

As disposal packaging concepts of spent fuels generated from the domestic NPP, two types, one is to package PWR and CANDU spent fuels in different containers and the other is to package them together, were proposed. The configuration of the containers and the layout of underground repository, such as the container spacing and the deposition tunnel spacing, were developed. The layout of underground repository satisfies the thermal constraint of the bentonite buffer surrounding disposal container, which should be lower than $100^{\circ}C$ in order to keep the physical and chemical properties of bentonite From the spent fuel packaging concepts and container emplacement methods, seven options were developed. With a typical pair-wise comparison methods, AHP, the most promising disposal concept was selected based on the technology Point of view.

• 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.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2003년도 가을 학술논문집
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• pp.609-613
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• 2003

사용후핵연료 및 방사성물질을 저장하기 위한 저장시스템은 사용후핵연료를 저장하는 동안 안전성 문제를 야기하지 않도록 격납을 설계하고 평가하여야 하며, 격납 평가는 ANSI Nl4.5 또는 ISO 12807에서 규정하고 있는 절차에 따른 허용 누설률을 계산하여 평가할 수 있다. 따라서, ISO 12807에서 규정한 평가방법에 따라 PWR 사용후핵연료 24 다발을 저장하였을 경우와 금속전환체 24다발을 저장하였을 경우에 대한 허용 누설률을 평가하였다. OWR 사용후핵연료 24다발을 저장하였을 경우 허용 누설률은 $1.38{\times}10_{-10}m_3/s$로, 금속전환체 24다발을 저장하였을 경우 $4.46{\times}10_{-10}m_3/s$로 평가되었다. 따라서, 사용후핵연료를 저장하였을 경우보다 금속전환체를 저장하였을 경우 격납 조건이 수월해 짐을 알 수 있었다.

• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2007년도 추계학술대회 논문집
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• pp.421-422
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• 2007

• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2008년도 춘계학술대회 논문집
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• pp.785-786
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• 2008

• 방사성폐기물학회지
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• 제17권4호
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• pp.405-418
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• 2019

국내 원자력발전소에서 발생하는 사용후핵연료의 제원 및 방출시점 등 특성과 현재의 고준위 방사성폐기물 기본계획에 근거한 처분시나리오를 도출하여 기존 심층 처분시스템을 바탕으로 처분효율과 경제성을 향상시킨 개선된 처분시스템을 제안하였다. 이를 위하여 국내 원자력발전소에서 발생하는 사용후핵연료의 길이에 따라 2종류의 처분용기 개념을 도출하고, 사용후핵연료 발생 년도와 현재의 기본계획에 근거한 처분 시나리오 설정에 따른 처분시점에서의 냉각기간을 고려하여 처분용기내 수용 가능한 붕괴열 량을 결정하였다. 그리고 2종류의 처분용기에 대한 처분시스템과 결정된 붕괴열을 바탕으로 열적 안정성 분석을 통하여 제안된 처분시스템의 설계요건에 대한 적합성 여부를 확인하고, 처분효율을 평가하였다. 개선된 처분시스템은 기존 처분시스템에 비하여 처분면적은 약 20% 감소되고 처분밀도는 약 20% 향상됨을 확인하였고, 처분용기와 완충재 재료도 상당량 절감됨을 확인하였다. 본 연구의 결과는 향후 사용후핵연료 관리정책 수립 및 실제 사업을 위한 처분시스템 설계를 위한 자료로 활용될 수 있다.

• 분석과학
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• 제17권5호
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• pp.438-442
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• 2004

To determine the contents of $^{99}Tc$ in the spent PWR (pressurized water reactor) nuclear fuels by ICP-MS (inductively coupled plasma-mass spectrometry), a technetium separation method using an extraction chromatographic resin (TEVA Spec resin) has been established. $^{99}Tc$ was separated from a spent PWR nuclear fuel solution by this separation procedure and its concentration was determined by ICP-MS. The result agrees well with the value calculated by the program ORIGEN 2 and also the value measured by AG MP-1 resin/ICP-MS method described in our previous paper. It can be concluded that the present separation procedure is superior to the AG MP-1 resin procedure with respect to the time required for technetium separation as well as the efficiency of decontamination from other radioactive nuclides.