• 대한방사선방어학회:학술대회논문집
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• 2009.04a
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• pp.254-255
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• 2009

국내 경수로 원전의 경우, 원전의 효율적, 경제적 운영차원에서 장주기 운전으로 패턴을 바뀌면서 핵연료봉 표면상에 크러드(crud)의 침적량은 점점 증가하는 경향을 나타내고 있다. 이러한 경향은 원자로의 출력 제어와 직결되면서 이에 대한 문제 해결을 위한 대표성이 있는 시료의 채취와 재현성이 있는 부식 생성물의 측정이 요구되어져 왔다. 원자로 계통 내에서 부식생성물의 농도변화에 대한 평가, 특히 입자농도가 증가되어지면 축방향 출력편차(Axial Offset Anomaly, AOA)가 발생될 수 있는 위험에 노출되거나, 핵연료 교체를 위해 발전소 정지시(shut down) 부식생성물의 방출이 급격히 증가되는 것으로 나타났다. 특히 입자성을 띤 물질은 존재의 특성상 이들 물질에 대한 대표시료의 채취가 어려울 뿐 아니라 grab 채취로 인해, 분석결과에 대한 재현성이 낮으며 계통 선량율의 제어와 작업자 피폭관리에 많은 어려움이 뒤따르고 있어 선진 원전 운영국에서는 앞 다투어 대표시료를 채취 할 수 있는 capillary sampling 법이나 integrated sampling법을 적용해 오고 있다. 본 논문에서는 국내 경수로 원전에서 일반적으로 사용하고 있는 grab sampling 법에 대한 문제점 파악과 해외 원전에서 사용 중인 capillary sampling 법의 국내 적용 가능성에 대해 살펴보았다.

• Resources Recycling
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• v.10 no.1
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• pp.56-65
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• 2001

Chemical decontamination techniques have been employed to reuse the high cost check valves contaminated with radioactivity and to reduce the radiation exposure during the inspection and maintenance work of safety injection system containing check valves. After chemical decontamination, an ultrasonic treatment was conducted to remove the fine solid particles retained in the crevices of check valves. The decontamination process conditions and the amount of chemical reagents were determined from the results of a pre-test, using the (list arm holder. The decontamination factors (DF), estimated from the activity in the solution, ranged from 14.5 to 18.5 corresponding to the activity removal of 93-95ft. The corrosion test data indicated that the general corrosion rate during a chemical decontamination-ultrasonic treatment process are low for type 304 S tainless steel, Inconel -600 and Stellite-6 materials $ (2.1\times10^{-2}$ $6.0\times10^{-2}$ and$ 1.7\times10^{-2}$ mil, respectively).

• The Journal of the Korea Contents Association
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• v.13 no.12
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• pp.860-868
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• 2013

The purpose of this study is to ensure safety by measuring External radiation dose ratio (ERDR) by traits of patients in many ways after administering radiopharmaceutical($^{18}F$-FDG) for PET Torso scan, and to decrease ERDR of those to RI technologist, caretakers, and those who frequently exposed to radiation by arousing attention to radiation dose. Radiopharmaceutical was administered to 80 patients who conducted PET Torso from January to June, 2013. Radiation dose emitted from the patients was measured according to body shape(BMI), water hydration, height, amount of radiation administration. From the moment immediately after the radiopharmaceutical was administered, ERDR was measured by personal traits of patients. The radiation dose increased in proportion to the administered amount of the radiopharmaceutical, and there was no significant difference depending on the body shape of the patients. When water was supplied and the height was normal, the radiation dose was lower compared with the cases where water was not supplied and height was not normal. There is a need for making efforts to minimize the working time through sufficient education and mock training before those who RI technologist with sources of radiation for complying the radiation safety management rule. And they should minimize the ERDR by wearing a protective gear.

• The Korean Journal of Nuclear Medicine Technology
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• v.14 no.2
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• pp.177-180
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• 2010

Purpose: As PET test came to be covered by the pay system of medical insurance (July 1, 2006) and the needs for it becoming increased for laboratory purpose, it became necessary to purchase expensive medical equipments to solve those problems. However, as most of equipments that are operated by cyclotron are very expensive as to amount from tens of millions up to hundreds of millions of won, it is difficult to purchase those equipments from the point of medical organizations. It may be possible to self manufacture those equipments with least costs if their parts functions that meets the operators demands. The Nuclear Medicine department of National Cancer Center (NCC) is trying to manufacture and use equipments that can be made with least costs, including introducing 2 medical equipments that can improves the operator's works. Materials and Methods: Example 1: Self production of radioisotope($^{18}F$) divider was fabricated. The NCC's Nuclear Medicine department acquired one acrylic panel, seven 3-way valve, tubing etc. that can be found in the market to make the main body of divider in cooperation with biomedical engineering, and placed them inside hot cell, and installed switching box outside of hot cell to make it possible to control them from outside. This main body of divider were placed in radioisotope transfer line that are manufactured in the cyclotron. Example 2: Self production of $^{18}F$-FDG automated divider was fabricated. The NCC's Nuclear Medicine department used cavro pump syringe that consists the main body of divider in cooperation with biomedical engineering, biomedical engineering developed programs that divides a certain amount. $^{18}F$-FDG automated divider is placed inside hot cell, and cable chords were used in the equipment, and then it was connected to PC outside hot cell to make it possible to control the $^{18}F$-FDG automated divider. Results: From the NCC's Nuclear Medicine department tests that were carried out from March, 2007 until now, we found out that radioisotope can be sent to radiopharmaceuticals composite module we want, and from the tests that are carried out at NCC's Nuclear Medicine department using $^{18}F$-FDG automated divider since August, 2009 it was possible to distribute radiopharmaceuticals into vial intended. Conclusion: Through the two examples above, we found out that costs can be reduced by self manufacturing expensive equipments from NCC's cyclotron room with least costs. Also, it decreased radiation exposure dose on workers, and set up problem solving processes in cooperation with lots of parties related.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.12 no.2
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• pp.171-178
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• 2014

After the expiration of operating license of a plant, all infrastructures within the plant must be safely dismantled to the point that it no longer requires measures for radiation protection. Despite the fact that Kori 1 and Wolsong 1 are close to the expiration of their operating license, sufficient technologies for radiological characterization, decontamination and dismantling is still under development. The purpose of this study is to develop one of methods for radiological inventory assessment on measuring object by using direct measure of large component with In-Situ measurement technique. Radiological inventory was assessed by analyzing nuclide using portable gamma spectroscopy without dismantling reactor head, and the result of direct measurement was supplemented by performing indirect measurement. Radiochemical analysis were performed on surface contamination samples as well. During the study, radiological inventory of reactor vessel calculated expanding the result. Based on the result and the radioactivity variation of each radionuclides time frame for decommissioning can be decided. Thus, it is expected that during the decommissioning of plants, the result of this study will contribute to the reduction of radiation exposure to workers.

• Journal of IKEEE
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• v.19 no.3
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• pp.282-287
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• 2015

In this study, an optical fiber detector was constructed by using a Ce:GAGG scintillator, optical fiber, and photomultiplier. The single crystal size of the scintillator was set to $3{\times}3{\times}20mm^3$ after simulating the counting efficiency of gamma rays in the scintillator by using the MCNPX code. The constructed detector used the standard gamma ray sources $^{137}Cs$ and $^{133}Ba$ to measure radiation and analyze the spectral characteristics of gamma rays. The resulting trend curve showed excellent linearity with an R-squared value of 0.99741, and the detector characteristics were found to vary 2% or less with distance based on comparison with the MCNPX value. Furthermore, the spectroscopic analysis of the gamma ray energy from the single-ray and mixed-ray sources showed that $^{137}Cs$ had its peak energy at 662 keV, and $^{133}Ba$ had at 356 keV. It seems that if the fiber-optics detector is used, working hours and exposure of worker can be reduced.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.4
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• pp.343-356
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• 2016

The Korea Radioactive Waste Agency (KORAD) has developed a dual-purpose metal cask for the dry storage of spent nuclear fuel that has been generated by domestic light-water reactors. The metal cask was designed in compliance with international and domestic technology standards, and safety was the most important consideration in developing the design. It was designed to maintain its integrity for 50 years in terms of major safety factors. The metal cask ensures the minimization of waste generated by maintenance activities during the storage period as well as the safe management of the waste. An activation evaluation of the main body, which includes internal and external components of metal casks whose design lifetime has expired, provides quantitative data on their radioactive inventory. The radioactive inventory of the main body and the components of the metal cask were calculated by applying the MCNP5 ORIGEN-2 evaluation system and by considering each component's chemical composition, neutron flux distribution, and reaction rate, as well as the duration of neutron irradiation during the storage period. The evaluation results revealed that 10 years after the end of the cask's design life, $^{60}Co$ had greater radioactivity than other nuclides among the metal materials. In the case of the neutron shield, nuclides that emit high-energy gamma rays such as $^{28}Al$ and $^{24}Na$ had greater radioactivity immediately after the design lifetime. However, their radioactivity level became negligible after six months due to their short half-life. The surface exposure dose rates of the canister and the main body of the metal cask from which the spent nuclear fuel had been removed with expiration of the design lifetime were determined to be at very low levels, and the radiation exposure doses to which radiation workers were subjected during the decommissioning process appeared to be at insignificant levels. The evaluations of this study strongly suggest that the nuclide inventory of a spent nuclear fuel metal cask can be utilized as basic data when decommissioning of a metal cask is planned, for example, for the development of a decommissioning plan, the determination of a decommissioning method, the estimation of radiation exposure to workers engaged in decommissioning operations, the management/reuse of radioactive wastes, etc.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.2
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• pp.103-116
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• 2006

By executing corrosion experiment on Inconel 600, 690 used to material of S/G tube in domestic NPP, this paper show estimation of amount of product such as Co-58, Co-60, Cr-51, Mn-54, Fe-59 which are main exposure cause to the workers in NPP. Therefore, Making the 12 samples consisted of Inconel 600, 690, whole corrosion experiment was carried out for 60 days(each pH by 20 days). The conditions of those tests were similar or more harsh than actual conditions of domestic NPP. The Glow Discharge Spectrometer(GDS) was used for quantitative analysis of results. The results of using GDS, the Inconel 600 corrodes more than Inconel 690 at pH 7 and pH 9. However, it is observed that Inconel 690 corrodes more than Inconel 600 at pH 4. Those results is estimated that test sections had the effect of transient. The long terms of experiment is required to minimize and solve the problem.

• Journal of Radiation Protection and Research
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• v.24 no.2
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• pp.109-120
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• 1999

Residual radiation and radioactivity level in TRIGA Mark-II, III research reactors and facilities at the KAERI Seoul site, which are to be decommissioned, have been measured, analyzed and evaluated to know the current status of radiation and radioactivity level and to establish and to provide the technical requirements for the safe decommissioning of the facilities which shall be applied in minimizing the radiation exposure for workers and in preventing the release of the radioactive materials to the environment. Radiation dose rate and surface radioactivity contamination level on the experimental equipments, floors, walls of the facilities, and the surface of the activated materials within the reactor pool structure were measured and evaluated. Radioactivity and radionuclides in the pool and cooling water were also analyzed. In case of the activated reactor pool structures which are very difficult to measure the radiation and radioactivity level, a computer code Fispin was additionally used for estimation of the residual radioactivity and radionuclides. The radiation and radioactivity data obtained in this study were effectively used as basic data for decontamination and dismantling plan for safe decommissioning of TRIGA Mark-II, III facilities.