• Proceedings of the Korea Information Processing Society Conference
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• 2004.05a
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• pp.85-88
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• 2004

해체 활동 시작 단계에서부터 최종처리까지 발생되는 일련의 모든 자료를 체계적이고 과학적으로 관리할 수 있는 해체 DB structure를 구축하기 위하여 해체 정보 전략 계획을 수립하였고, 해체 DB 정보 영역을 분류하여 세부항목을 도출하였으며, 시스템 개발 도구 및 운영환경을 설정하였다. 자료흐름도(DFD)와 개체 관계도(ERD)를 이용하여 해체 자료 구조를 최적화하였고, Prototype 과정을 거쳐 해체 자료가 체계적으로 저장 관리 될 수 있도록 프로그램을 개발 하였다. 현재(2001년6월부터 2003년12월)까지 연구로 해체활동을 통해 발생한 해체 현장 자료를 이용하여 해체 DB 시스템을 시험한 결과 무작위로 데이터를 추출하여 집계한 결과와 잘 일치하고 있음을 확인하였다.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.1
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• pp.19-27
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• 2007

Combined method of extraction chromatography and liquid scintillation analysis was established for determinating $^{55}Fe\;and\;^{63}Ni$ radioactivity in solid samples. Activated concrete wastes generated from dismantling KRR-2 were analysed. The sequential separation including precipitation and extraction chromatography resulted in the above 90% chemical recoveries of Fe. Above 62% recoveries of Ni were obtained by this procedure exception to 43.6 and 46.5% recoveries. The seperation and counting procedure was also confirmed with spiked samples of known quantity. The measured and spiked quantity were agreed with the 3.7% and 0.7% variations in the $^{55}Fe\;and\;^{63}Ni$ experiments, respectively. The radioactivities of $^{55}Fe$ in the dismantled concretes are shown from below MDA to maximum 362 Bq/g. The radioactivities of $^{63}Ni$ in all concrete samples are below MDA. The $^{63}Ni$ doesn't exist in dismantled concretes from KRR-2. The radioactivity of $^{55}Fe$ is decreased rapidly as the sampling depth is increased from the concrete surface.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.660-664
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• 2003

The exposure dose form recycling on a large amount of the steel scrap from the KRR-1&2 decommissioning activities was evaluated, and also the clearance level was derived. The maximum individual dose and collective dose were evaluated by modifying internal dose conversion factor which was based on the concept of effective dose in ICRP 60, applied to the RESRAD-RECYCLE ver 3.06 computing code, IAEA Safety Series III-P-1.1 and NUREG-1640 as the assessment tool. The result of assessment for individual dose and collective dose is 23.9 ${\mu}Sv$ per year and 0.11 man$\cdot$Sv per year respectively. The clearance levels were ultimately determined by extracting the most conservative value form the results of the generic assessment and specific assessment methodologies. The result of clearance level for radionuclides($Co^60$, $Cs^137$) is less than $1.67{\times}10^{-1}$ Bq/g to comply with the clearance criterion(maximum individual dose : 10 $\muSv$ per year, collective dose : 1 man$\cdot$Sv per year) provided for Korea Atomic Energy Act and relevant regulations.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 2002.05a
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• pp.43-48
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• 2002

ISP (Information Strategy Planning), which is the first step of the whole database development has been studied to manage effectively information and data related to the decommissioning activities of the Korea Research Reactor 1＆2(KRR-1＆2). In order to establish the scope of the decommissioning DB, user requirement and the importance of the decommissioning information were analyzed and set up the conceptual design of the decommissioning DB and established the classification system related to decommissioning activities. It has been extracted the principle information such as working information facilities information, radioactive waste treatment information and radiological surveying and analysis during the interviewing with an experts. And also, It has been made the code system. These results will be used as the basic data to design the prototyping for the decommissioning DB.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.707-719
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• 2003

The D&D work requires worker's safety and high reliability of operation because it has been processing in high radioactive environment. Therefore, it is necessary to select the dismantling items and applicable dismantling technologies and analyze the scenarios for selected items. In this paper, the main dismantling items were selected by the consideration of several factors, their 3D CAD models were constructed as well. The applicable dismantling technologies for each dismantling items were selected and their dismantling scenarios were setup. Finally the 3D graphic simulations for the shielding concrete, the RSR, and the core are performed.

• International Journal of Contents
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• v.16 no.2
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• pp.102-108
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• 2020

The purpose of this study was to estimate and analyze the potential radiation dose that the future visitors and the cleaning staff will be exposed to when the KRR-1 reactor is converted into a memorial hall. The radiation doses were estimated using the RESRAD-BUILD software, where case, building, receptor, shielding, and source parameters were applied as the input data. Also, the basic data for the assessment of the radiation doses were determined in an indirect manner using the data on the waste generated during the decommissioning process of the reactor. The assessment results indicate that the potential radiation dose to the visitors and the cleaning staff will be less than 1 mSv, the annual dose limit for the general public. However, if anyone for a significant period of time is close to the reactor, the overall dose will increase. The radiation dose for the future visitors and the cleaning staff was determined to be lower than the annual dose limit for the general public. Given such a risk, systematic measures, such as periodic monitoring or limiting hours, are imperative.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.43-51
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• 2004

In order to decommission the shielding concrete of KRR(Korea Research Reactor) -1＆2, it must be exactly determined activated level and range by neutron irradiation during operation. To determine the activated level and range, it must be sampled and analyzed the core sample. But, there are difficulties in sample preparation and determination of the measurement efficiency by self-absorption. In the study, the full energy efficiency of the HPGe detector was compared with the measured value using standard source and the calculated one using Monte Carlo simulation. Also. self-absorption effects due to the density and component change of the concrete were calculated using the Monte Carlo method. Its results will be used radioactivity analysis of the real concrete core sample in the future.

• The Transactions of The Korean Institute of Electrical Engineers
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• v.62 no.10
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• pp.1465-1469
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• 2013

The resent developed high speed train called HEMU-430X (Highspeed Electric Multiple Unit - 430km/h eXperiment) recorded a 421.4 km/h in Kyoungbu high speed line in Korea. A verity of measurement systems are used to check the performance between pantograph and catenary system. An innovative measurement system is adopted to check the current of catenary wire in the track side during HEMU-430X test running. This paper presents the measurement results of catenary current in kyoungbu high-speed line and describe its utilizing technology in the experimental results of catenary current. In order to analyze field testing results, the current ratio between contact and catenary current have been analyzed by means of Carson-Pollaczek equation. And the current wave forms between catenary and contact wire are presented based on the simulation results.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.681-686
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• 2003

It has been designed GUI(graphic user interface) to consult the convenience of a input data and the flexibility that can be retrieved dismantling information relation to decommissioning DB of KRR1&2. The GUI can proceed an input materials and a search and output of a saved data in server based on a facility code and also have a function of explorer which can find the lower dismantling objects in each facilities. It has added the structure of the multimedia that could be showed a series of dismantling activities with a Mpeg and pictures into the GUI. In the future work, Decommissioning DB and User Interface are intend to contribute a functions that could be evaluate and analyze for a dismantling activities with a engineering theory.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.11 no.4
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• pp.271-280
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• 2013

Since the decommissioning of nuclear plants and facilities, large quantities of slightly contaminated concrete waste have been generated. In Korea, the decontamination and decommissioning of the KRR-1, 2 at the KAERI have been under way. And concrete waste was generated about 800 drums of 200 L. The conditioning of concrete waste is needed for final disposal. The concrete waste is conditioned as follows: mortar using coarse and fine aggregates is filled void space after concrete rubble pre-placement into 200 L drum. Thus, this research has developed an optimizing mixing ratio of concrete waste, water, and cement and has evaluated characteristics of a cement waste form to meet the requirements specified in disposal site specific waste acceptance criteria. The results obtained from compressive strength test, leaching test, thermal cycling test of cement waste forms conclude that the concrete waste, water, and cement have been suggested to have 75:15:10wt% as the optimized mixing ratio. Also, the compressive strength of cement waste form was satisfied that including fine powder up to maximum 40wt% in concrete debris wastes about 75%. As a result of scale-up test, the mixture of concrete waste, water, and cement is 75:10:15wt% meet the satisfied compressive strength because the free water increased with and increased in particle size.