• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.11 no.4
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• pp.333-349
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• 2013

As the capacity of spent nuclear fuel storage pool at reactor sites becomes saturated in ten years, long term dry storage strategy has been recently discussed as an alternative option in Korea. In this study, we reviewed safety-criteria-related research results on spent nuclear fuel performance and integrity under long-term dry storage and proposed the direction and the scope of future domestic research and development. Creep and hydride effect in relation to the embrittlement are known to be the major degradation mechanisms of the spent fuels during the long term dry storage. However, recent research results showed that hydride reorientation and hydride embrittlement are one of the most critical factors to the spent fuel integrity. Accordingly safety criteria of US and Japan for the storage system are basically founded on those mechanisms. However, in Korea, not only in-pile but out-of-pile experimental data have not been generated to understand fuel cladding degradation and to determine the criteria to ensure the safety. In addition, the transient behavior of the spent fuel during transportation also needs to be thoroughly examined. Therefore, various experimental research and development will be required to establish our own safety criteria for future long-term dry storage of domestic spent fuels.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.4
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• pp.411-422
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• 2016

Recently, because the wet pool storage facilities of NPPs in Korea has become saturated, there has been much active R&D on an interim dry storage system using a transportation and storage cask. Generally, the shielding evaluation for the design of a spent fuel transportation and storage cask is performed by the design basis fuel, which selects the most conservative fuel among the fuels to be loaded into the cask. However, the loading of actual spent fuel into the transportation metal cask is not limited to the design basis fuel used in the shielding evaluation; the loading feasibility of actual spent fuel is determined by the shielding evaluation that considers the characteristics of the initial enrichment, the maximum burnup and the minimum cooling period. This study describes a shielding analysis method for determining the minimum cooling period of spent fuel that meets the domestic transportation standard of the dual purpose metal cask. In particular, the spent fuel of 3.0~4.5wt% initial enrichment, which has a large amount of release, was evaluated by segmented shielding calculations for efficient improvement of the results. The shielding evaluation revealed that about 81% of generated spent fuel from the domestic nuclear power plants until 2008 could be transported by the dual purpose metal cask. The results of this study will be helpful in establishing a technical basis for developing operating procedures for transportation of the dual purpose metal cask.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.4
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• pp.455-472
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• 2018

It is important to study how to manage dry storage casks of spent nuclear fuels (SNF), because wet storage spaces for SNF will shortly be at full capacity in the Republic of Korea. The US has operated a dry storage cask system for several decades, and has carried out significant studies into how to successfully manage dry storage cask for SNF. This type of expertise and experience is currently lacking in the Republic of Korea. The degradation of dry casks is an important issue that must be considered. In particular, chloride-induced stress corrosion cracking (CISCC) is known to lead to the release of radioisotopes from canisters. The U.S. Department of Energy, U.S. Nuclear Regulatory Commission, and the Electric Power Research Institute have undertaken research into the CISCC mechanism. In addition, Sandia National Laboratories (SNL) has extensively researched CISCC and how to manage it in dry storage canisters. In this review paper, the probabilistic model proposed by the SNL is analyzed and, based on this model, US-based CISCC research is reviewed in detail. This paper will inform the management of dry cask storage of SNF from light water reactors in austenite stainless steel canisters in the Republic of Korea.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.2
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• pp.167-181
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• 2019

Currently, the development of evaluation technology for vibration and shock load characteristics and spent nuclear fuel structural integrity under normal conditions of transport is being conducted in the Republic of Korea. This is the first such research conducted in the Republic of Korea and, thus, previous international studies need to be investigated and will be referred to in the ongoing project. Before 2000, several studies related to measurement of vibration and shock loads on spent nuclear fuel were conducted in the US. US national research institutes conducted uniaxial fuel assembly shaker tests, concrete block tests, and multi-axis fuel assembly tests between 2009 and 2016. In 2017, multi-modal transportation tests including road, sea, and rail transport were also performed by research institutes from the US, Spain and the Republic of Korea. Therefore, test preparation procedures, acceleration and strain measurement results, and finite-element and multi-body dynamics analysis were investigated. Based on the measured strain data, the preliminary conclusion was obtained that the measured strain was too small to cause damage to spent nuclear fuel rods. However, this conclusion is a preliminary conclusion that only reviews part of the results; a detailed review is being conducted in the US. The investigation of international studies on spent nuclear fuel structural integrity tests considering vibration and shock loads under normal conditions of transport in the US will be useful data for the project being conducted in the Republic of Korea.

• Nuclear industry
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• v.7 no.12 s.58
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• pp.71-77
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• 1987

사용후핵연료를 재이용하는 일련의 흐름을 핵연료사이클이라 부르지만, 재이용$\cdot$재처리에 대한 생각, 대응방안에는 각국의 원자력정책의 차이가 있어서 각기 다르다. 일본에서는 청삼현$\cdot$육개소촌에서 핵연료사이클 3개시설(재처리, 우라늄농축, 저준위폐기물저장)의 건설이 진행되고 있다. 다음은 일본, 미국, 영국, 프랑스, 서독 등 5개국의 재처리와 방사성폐기물 처리$\cdot$처분에 대한 대응방안을 소개한 것이다.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.381-387
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• 2003

Hull arising from the spent PWR fuel elements is classified as a high-level radioactive waste. This report describes the radio-chemical characteristics of the hull -from PWR spent fuel of 32, 000MWd/tU burn-up and 15 years cooling, discharged from Gori Unit I cycled 4 -7-by examination and literature survey. On the basis of the results, a method of degradation to middle and low-level radio active waste was proposed by dry process such as laser or plasma technique with removing the nuclides deposited on the surface of the hull.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.596-600
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• 2003

The Advanced spent fuel Conditioning Process(ACP) is under development for the effective management of spent fuel which had been generated in nuclear plants. The ACP needs a hot cell where most operations will be peformed. To give priority to the environment safety, radiation doses evaluation for the radioactive nuclides were preliminarily peformed in both normal operation and accident case. The evaluation result shows a safe margin for regulation limits and SAR limit of IMEF where this facility will be constructed.

• Journal of Energy Engineering
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• v.13 no.4
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• pp.259-266
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• 2004

Specific heat mechanism of oxide fuel is contributed by lattice vibration, dilatation, conduction electron and defect and excess specific heat. Model of oxide fuel for specific heat consists of specific heat at constant pressure term, dilatation specific heat term and defect specific heat term. In this study experimental and published data on the specific heats of oxide nuclear fuels have been reviewed and analyzed to recommend the best fitting model. The oxide fuels considered in this paper were UO$_2$, mixed (U, Pu) oxides and spent fuel. The specific heat data of spent fuel has been replaced by that of simulated fuel.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.1
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• pp.51-58
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• 2006

Since 2002, more than 400 PWR spent nuclear fuel assemblies have been transported and stored on-site using transport casks in order to secure the storage capacity of PWR spent nuclear fuel of Kori nuclear power plant. The complete on-site transport system, which includes KN-12 transport casks, the related equipment and transport vehicles, had been developed and provided. KN-12 transport casks were designed, fabricated and licensed in accordance with Korean and IAEA's transport regulations, and the related equipment was also provided in accordance with the related regulations. The on-site transport and storage operation using two KN-12 casks and the related equipment has been conducted, and the strict Quality Control and Radiation Safety Management through the whole process has been carried out so as to achieve the required safety and reliability of the on-site transport of spent nuclear fuel.

• Analytical Science and Technology
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• v.16 no.2
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• pp.110-116
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• 2003

A study has been done on the separation of iodide from spent pressurized water reactor (PWR) fuels and its quantitative determination using ion chromatography. Spent PWR fuels were dissolved with mixed acid of nitric and hydrochloric acids (80 : 20 molL%) which can oxidize iodide to iodate to prevent it from be vaporized. After reducing ${IO_3}^-$ ­to $I_2$ in 2.5 M $HNO_3$ with $NH_2OH{\cdot}HCl$, Iodine was selectively separated from actinides and all other fission products with carbontetrachloride and back-extracted with 0.1 M $NaHSO_3$. Recovered iodide was determined using the ion chromatograph of which the column was installed in a glove box for the analysis of radioactive materials. In practice, spent PWR fuel with 42,000~44,000 MWd/MtU was analyzed and its quantity was compared to that calculated by burnup code, ORIGEN2. The agreement was achieved with a deviation of -8.3~-0.5% from the ORIGEN 2 data, $324.5{\sim}343.6{\mu}g/g$.