• Proceedings of the Korean Radioactive Waste Society Conference
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• 2016.10a
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• pp.103-104
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• 2016

• Nuclear Engineering and Technology
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• v.22 no.2
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• pp.83-94
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• 1990

The radioactivity in the spent fuel storage pool is calculated to ensure to maintain its concentration below the permissible limit, when the storage capacity of Uljin nuclear power plant unit 1&2 is extended from 9/3 to 32/3 core using consolidated fuels in maximum density rack (MDR). For this evalulation, two models to calculate the spent fuel pool activities on the continuous and intermittent operating its purification system are developed and these results compared, The results of above two cases show that the current water purification system can not guarantee the radioactivity concentration below the design limit, 5$\times$10$^{-4}$ $\mu$Ci/ml, for the extention to 32/3 core. Therefore, it has been concluded that a modification of the current purification system is necessary to extend the spent fuel storage capacity with the above method. The alternative way suggested in this study is to increase the number of cation bed demineralizers.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.4
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• pp.231-236
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• 2011

When the Fukushima nuclear power plant accident occurred in March of 2011, a hydrogen explosion in the reactor building at the 4th unit of Fukushima plants led to a big surprise because the full core of the unit 4 reactor had been moved and stored underwater at the spent nuclear fuel storage pool for periodic maintenance. It was because the possible criticality in the fuel storage pool by coolant loss may yield more severe situation than the similar accident happened inside the reactor vessel. Fortunately, it was assured to be evitable to an anxious situation by a look of water filled in the storage pool later. In the paper, the safety state of the spent fuel storage pool and rack structures of the domestic nuclear plants would be roughly reviewed and compared with the Fukushima plant case by engineering viewpoint of potential severe accidents.

• Journal of the Korean Society for Nondestructive Testing
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• v.19 no.2
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• pp.93-99
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• 1999

SCAI(spent CANDU fuel bundle serial number identifier) was developed to read serial numbers of spent fuel bundles in the spent fuel storage. For the purpose of effectively identifying the serial number of fuel bundle. SCAI was composed of underwater camera & light part. guiding & supporting part and control & monitor part. So it is easy to assemble and disassemble, and operate. It was tested to read serial numbers of spent fuel bundles loaded in basket during the recent spent fuel transfer campaign at Wolsong Unit 1. And it was also applied to read serial numbers of spent fuel bundles discharging from the initial core at Wolsong Unit 3 by slight change of camera and light. Inspectors could easily operate SCAI after several practices in the storage pond, which was a user friendly. And SCAI provided clear and immediate picture for identification of serial numbers of spent fuel bundles. It was interally evaluated that SCAI greatly contributed to cut inspection efforts for national and international safeguards at Wolsong power plant.

• Journal of Radiation Protection and Research
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• v.9 no.2
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• pp.90-96
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• 1984

Estimated dose rates in spent fuel pool storage with the extended fuel cycle core management were reviewed and compared with design limit after calculation with the aid of DLC-23/CASK(22 n, 18 g) nuclear data and ANISN code. Radioactivity and gamma spectrum within spent fuel assemblies were calculated with ORIGEN code by extended fuel cycle model. In the calculation of dose rate, the fuel pool geometry was assumed to be infinite slab. Also, composition materials and radiation source within assemblies which are being stored in pool storage were assumed to be uniformly distributed throughout all the assemblies. As a result of culculation of dose rate from stored assemblies and waterborne radionuclides in pool water, the calculated dose rates appear to be lower than design basis limit under normal condition as well as abnormal condition.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.344-344
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• 2004

한국원자력연구소에서는 사용후핵연료의 체적을 감소시켜 저장 안전성 및 경제성을 확보키 위한 사용후핵연료 차세대관리 공정(ACP)을 개발하고 있다. 이 기술의 개발을 위해서는 사용후핵연료를 사용한 실증시험이 필수적이며, 이를 위한 ${\alpha}-{\gamma}$ type의 hot cell 시설 및 부속시설이 필요하다. 연구소는 별도의 실증시설에 요구되는 고 비용을 줄이기 위해 현재 연구소가 보유하고 있는 조사재시험시설(IMEF)의 지하에 위치한 예비 hot cell을 활용키로 하고 차세대관리 종합공정의 특성 및 용도에 맞는 시설의 수정/보완 업무를 수행해오고 있다.(중략)

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.18 no.2
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• pp.43-49
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• 2022

The purpose of this study is to evaluate the radiation effect on damage when the external shield of the spent nuclear fuel transport cask is damaged due to impact as the cause of an unexpected accident. The neutron and gamma-ray intensities and spectra are calculated using the ORIGEN-Arp module in the SCALE 6.2.4 code package⁽¹⁾ and then using MCNP6.2⁽²⁾ code calculate the dose rate. In order to evaluate the radiation dose according to the size of damage caused by external impact, various sized holes of 0.3~13.7% are assumed in the outer shield of the cask to evaluate the sensitivity to the dose. In the case of radiation source leakage, damage to the nuclear fuel assembly is assumed to be up to 6% based on overseas test cases. When only the outer shield is damaged, the maximum surface dose is calculated as 3.12E+03 mSv/hr. However, if the radiation source is leaked due to damage to the nuclear fuel assembly, it becomes 7.00E+05 mSv/hr which is about 200 times greater than the former case.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.11 no.1
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• pp.31-39
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• 2013

Nuclear fuel can be damaged under unexpected circumstances in a nuclear reactor. Fuel rod failure can be occurred due to debris fretting or excessive hydriding or PCI (Pellet-to-clad Interaction) etc. It is important to identify the causes of such failed fuel rods for the safe operation of nuclear power plants. If a fuel rod failure occurs during the operation of a nuclear power plant, the coolant water is contaminated by leaked fission products, and in some case the power level of the plant may be lowered or the operation stopped. In addition, all spent fuels must be transferred to a dry storage. But failed fuel can not be transferred to a dry storage. Therefore, the purpose of this study is to develop a system which is capable of inspecting whether the spent fuel in the storage pool is failed or not. The sipping technology is to analyze the leakage of fission products in state of gas and liquid. The failed fuel inspection system with gamma analyzer has successfully demonstrated that the system is enough to find the failed fuel at Wolsong plant.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05b
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• pp.350-355
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• 1997

가압중수로(PHWR)형 원자력발전소의 저장수조에 보관중인 사용후핵연료를 대상으로 하는 핵물질 보장조치(safeguards) 이행에 필요한 핵연료다발 수중검증장치를 개발하였다. 본 장치는 CdTe 감마선검출기, 차폐체 및 시준기등으로 구성된 검출부와 이를 지지 및 구동하기 위한 구동부로 구성되어 있다. 검출부에 대하여 감마선 표준선원 및 사용후핵연료 시료를 사용하여 성능시험을 수행한 결과 현장검증시의 요건을 만족하였고, 구동부의 경우 건식조(dry pit)에서 수행한 예비실험 결과 검증목적에 적합하였다. 따라서, PHWR형 원자력발전소인 월성 1 호기의 수중저장조에 있는 사용후CANDU핵연료에 대한 현장성능시험을 현재 준비중에 있으며 그 결과를 바탕으로 하여 국가사찰시에 본 장치를 사용할 예정이며, 향후 IAEA의 공인을 획득하여 IAEA 사찰용 장비로도 활용할 계획이다.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.4
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• pp.237-245
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• 2011

As necessity of safety re-evaluation for spent fuel storage facility has emphasized after the Fukushima accident, accuracy improvement of heat load evaluation has become more important to acquire reliable thermal-hydraulic evaluation results. As groundwork, parametric and sensitivity analyses of various storage conditions for Kori Unit 4 spent fuel storage pool and spent fuel depletion parameters such as axial burnup effect, operation history, and specific heat are conducted using ORIGEN2 code. According to heat load evaluation and parametric sensitivity analyses, decay heat of last discharged fuel comprises maximum 80.42% of total heat load of storage facility and there is a negative correlation between effect of depletion parameters and cooling period. It is determined that specific heat is most influential parameter and operation history is secondly influential parameter. And decay heat of just discharged fuel is varied from 0.34 to 1.66 times of average value and decay heat of 1 year cooled fuel is varied from 0.55 to 1.37 times of average value in accordance with change of specific power. Namely depletion parameters can cause large variation in decay heat calculation of short-term cooled fuel. Therefore application of real operation data instead of user selection value is needed to improve evaluation accuracy. It is expected that these results could be used to improve accuracy of heat load assessment and evaluate uncertainty of calculated heat load.