• 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.17 no.4
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• pp.375-387
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• 2019

In this study, the radiation dose rates for the design basis fuel of 360 assemblies CANDU spent nuclear fuel transportation cask were evaluated, by measuring radiation source terms for the design basis fuel of a pressurized heavy water reactor. Additionally, radiological safety evaluation was carried out and the validity of the results was determined by radiological technical standards. To select the design basis fuel, which was the radiation source term for the spent fuel transportation cask, the design basis fuels from two spent fuel storage facilities were stored in a spent fuel transportation cask operating in Wolsung NPP. The design basis fuel for each transportation and storage system was based on the burnup of spent fuel, minimum cooling period, and time of transportation to the intermediate storage facility. A burnup of 7,800 MWD/MTU and a minimum cooling period of 6 years were set as the design basis fuel. The radiation source terms of the design basis fuel were evaluated using the ORIGEN-ARP computer module of SCALE computer code. The radiation shielding of the cask was evaluated using the MCNP6 computer code. In addition, the evaluation of the radiation dose rate outside the transport cask required by the technical standard was classified into normal and accident conditions. Thus, the maximum radiation dose rates calculated at the surface of the cask and at a point 2 m from the surface of the cask under normal transportation conditions were respectively 0.330 mSv·h^-1 and 0.065 mSv·h^-1. The maximum radiation dose rate 1 m from the surface of the cask under accident conditions was calculated as 0.321 mSv·h^-1. Thus, it was confirmed that the spent fuel cask of the large capacity heavy water reactor had secured the radiation safety.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.4
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• pp.491-505
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• 2018

As an alternative to deep geological disposal technology, which is considered as a reference concept, the domestic applicability of deep borehole disposal technology for high level radioactive waste, including spent fuel, has been preliminarily evaluated. Usually, the environment of deep borehole disposal, at a depth of 3 to 5 km, has more stable geological and geo-hydrological conditions. For this purpose, the characteristics of rock distribution in the domestic area were analyzed and drilling and investigation technologies for deep boreholes with large diameter were evaluated. Based on the results of these analyses, design criteria and requirements for the deep borehole disposal system were reviewed, and preliminary reference concept for a deep borehole disposal system, including disposal container and sealing system meeting the criteria and requirements, was developed. Subsequently, various performance assessments, including thermal stability analysis of the system and simulation of the disposal process, were performed in a 3D graphic disposal environment. With these analysis results, the preliminary evaluation of the domestic applicability of the deep borehole disposal system was performed from various points of view. In summary, due to disposal depth and simplicity, the deep borehole disposal system should bring many safety and economic benefits. However, to reduce uncertainty and to obtain the assent of the regulatory authority, an in-situ demonstration of this technology should be carried out. The current results can be used as input to establish a national high-level radioactive waste management policy. In addition, they may be provided as basic information necessary for stakeholders interested in deep borehole disposal technology.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.7 no.3
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• pp.167-174
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• 2009

A comprehensive analysis has been conducted on the purity of the uranium product generated from a pyroprocessing of EBR-II spent fuel. The analysis results were compared to the low-level waste criteria for both ROK and USA under a collaborative program between INL and KAERI. It is found that the US LLW definition does not include the activity from any U isotopes, but the Korean one does. The analysis results show that Pu-239 is the only alpha emitting isotope other than U isotopes that exceed the limit in the EBR-II U product. Pu contamination of the product seems to be drastically reduced in a preliminary test of the modified cathode process, and the further development of the proposed technology may be possible to meet the US LLW criteria.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.12 no.4
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• pp.299-314
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• 2014

A lifting device is used to deal with transport cask for the transportation of spent fuels from nuclear power plants. This study performed theoretical analysis and numerical simulation to evaluate the structural integrity of the lifting device based on Nuclear Safety and Security Commission(NSSC) Notice No.2013-27 and US 10CFR Part 71 ${\S}71.45$. The results of theoretical analysis showed that the maximum stresses of all components were below the allowable values. This result confirmed that the lifting device was structurally safe during operation. The results of finite element analysis also showed that it was evaluated to satisfy the design criteria bothyielding and ultimate condition. All components have been shown to ensure the structural safety due to sufficient safety margins. In other words, the safety factor was 3 or more for the yielding condition and was 5 or more for the ultimate condition.

• Journal of Radiation Protection and Research
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• v.21 no.4
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• pp.313-327
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• 1996

As an at-reactor(AR) storage method o( spent fuel, there are horizontal concrete module type, metal storage cask type, concrete storage cask type, dual purpose (transportation and storage) cask type and multi-purpose (transportation, storage and disposal) cask type. All other types except multi-purpose one have been already used for AR dry storage of spent fuels after obtaining operation license in various foreign countries. Also the development of multi-purpose type has been continued for operation license. In America, Japan, Germany, Canada, Spain, Switzerland, and Czech Republic, etc., AR dry storage facilities are under operation or on propulsion, and spent fuels are transported to interim storage facility or reprocessing plant after dry storage at reactor temporarily. At Wolsung site, in case of Korea, concrete silo type has already been introduced, and it is believed to be inevitable to store spent fuels at reactor temporarily, considering the reality that storage capacity of spent fuel is approaching to the limit in some nuclear power plants. In this report, the system characteristics, design requirements, technical standards and status of AR storage system, which is suitable for domestic site such as Kori, have been studied. In most cases, the licensed period of storage cask is limited up to 20 years and the integrity of material and maintenance of leaktightness are required during the whole service life.

• Nuclear Engineering and Technology
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• v.27 no.6
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• pp.911-917
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• 1995

로듐 자기 기전력형 중성자 계측기는 단위 중성자당 발생되는 전류 신호가 매우 커 계측성이 우수하나 연소율이 빨라 자주 교체해야 하므로 재장전 기간 연장 및 새로운 로듐계측기 구입 등의 문제점이 있다. 75% 연소에 해당하는 제5핵 연료 주기 기간 동안 영광 3, 4호기와 같은 C-E 원자로에 사용되고 있는 로듐 자기 기전력형 중성자 계측기의 연소 거동이 C-E에 의해서 연구되었다. 약 제3핵연료주기까지 분석한C-E의 초창기 연구에서는 중성자 방사화율 개념에 근거하여 약 66%연소시점까지 로듐계측기 연소특성곡선은 선형적임이 밝혀진 바 있다. ［l, 2］ 그 후 C-E의 연구에 의하면 약 75% 연소에 해당하는 제5핵연료 주기까지도 로듐계측기 연소 특성 곡선은 선형적임이 밝혀졌다. 그 결과로 C-E형 원자력발전소에서 사용되는 로듐 노내계측기의 수명을 약 60%연소에서 66%연소 시점까지 연장시킬 수 있게 되었다. ［3］이 정도의 계측기 수명 연장은 약 반년의 원자로 운전 기간에 해당되며 차기 핵연료주기에서 많은 로듐 노내계측기를 계속 사용할 수 있게 한다. 특히 영광 3, 4호기가 12개월 핵연료 주기에서 18개월 핵연료 주기로 재장전 전략을 바꿀 경우 로듐 노내계측기의 수명 이 연장되지 않으면 계측기 교체가 빈번해 질 것으로 사료되어 로듐 수명 연장과 관련된 기술 특히 C-E 및 B&W의 로듐 노내계측기 연소도 특성곡선 불확실도 평가 및 출력 측정 계통 오차 분석 기술을 소개하고자 한다. 영광3, 4호기에서 사용중인 로듐 노내계측기 수명을 현재 연소도 기준 66%내로 한정하고 있는데 C-E 및 B&W의 로듐 노내계측기연소 특성에 관한 연구 내용을 분석한 결과 노내계측기 수명을 연소도가 66%를 초과하는 계측기가 있어도 전체적으로 불확실도가 안전한계를 넘지 않으면 노내상주가 가능한 것으로 평가되었다.

• Journal of Radiation Protection and Research
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• v.40 no.2
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• pp.92-100
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• 2015

For the purpose of improving the efficiency of the radiation impact assessment of dry interim storage facilities for the spent nuclear fuel of pressurized water reactors (PWRs), radiation impact assessment was performed after the application of sensitivity assessment according to the radiation source term designation method, development of a 2-step calculation technique, and cooling time credit. The present study successively designated radiation source terms in accordance with the cask arrangement order in the shielding building, assessed sensitivity, which affects direct dose, and confirmed that the radiation dosage of the external walls of the shielding building was dominantly affected by the two columns closest to the internal walls. In addition, in the case in which shielding buildings were introduced into storage facilities, the present study established and assessed the 2-step calculation technique, which can reduce the immense computational analysis time. Consequently, results similar to those from existing calculations were derived in approximately half the analysis time. Finally, when radiation source terms were established by adding the storage period of the storage casks successively stored in the storage facilities and the cooling period of the spent nuclear fuel, the radiation dose of the external walls of the buildings was confirmed to be approximately 40% lower than the calculated values; the cooling period was established as being identical. The present study was conducted to improve the efficiency of the Monte Carlo shielding analysis method for radiation impact assessment of interim storage facilities. If reliability is improved through the assessment of more diverse cases, the results of the present study can be used for the design of storage facilities and the establishment of site boundary standards.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.1
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• pp.63-75
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• 2012

This paper gives some basic requirements and preferences of various geological environmental conditions for the final deep geological repository of spent nuclear fuel (SNF). This study also indicates how the requirements and preferences are to be considered prior to the selection of sites for a site investigation as well as the final disposal in Korea. The results of the study are based on the knowledge and experience from the IAEA and NEA/OECD as well as the advanced countries in SNF disposal project. This study discusses and suggests preliminary guideline of the disposal requirements including geological, mechanical, thermal, hydrogeological, chemical and transport properties of host rock with long term geological stabilities which influence the functions of a multi-barrier disposal system. To apply and determine whether requirements and preferences for a given parameter are satisfied at different stages during a site selection and suitability assessment of a final disposal site, the quantitative criteria in each area should be formulated with credibility through relevant research and development efforts for the deep geological environment during the site screening and selection processes as well as specific studies such as productions of safety cases and validation studies using a generic underground research laboratory (URL) in Korea.

• 제어로봇시스템학회:학술대회논문집
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• 1991.10a
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• pp.911-916
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• 1991

Spent nuclear fuel is highly radioactive and requires remote operation due to radiation exposure. Motors which have been used in the highly radioactive environments are a DC type because of their easy implementation on control system. However there are some problems such as mechanical maintenance of brush and commutator, high cost, and heavy material control. AC servo motors are applied and tested on fuzzy and conventional control algorithms. Fuzzy logic controls of AC servo give adequate control accuracy and power for spent fuel handling in radioactive environments.