• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.2_spc
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• pp.291-303
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• 2020

The Kori-Unit 1 nuclear power plant, which is scheduled to be decommissioned after permanent shutdown, is expected to generate large amounts of various types of radioactive waste during the decommissioning process. Among these, nuclear reactors and internal structures have high levels of radioactivity and the dismantled structure must have the proper size and weight on the primary side. During decommissioning, it is important to prepare an appropriate and efficient disposal method through analysis of the disposal status and the legal restrictions on wastes generated from the reactors and internal structures. Nuclear reactors and internal structures generate radioactive wastes of various levels, such as medium, very low, and clearance. A radiation evaluation indicates that wastes in the clearance level are generated in the reactor head and upper head insulation. In this study, a clearance waste safety evaluation was conducted using the RESRAD-RECYCLE code, which is a safety evaluation code, based on the activation evaluation results for the clearance level wastes. The clearance scenario for the target radioactive waste was selected and the maximum individual and collective exposure doses at the time of clearance were calculated to determine whether the clearance criteria limit prescribed by the Nuclear Safety Act was satisfied. The evaluation results indicated that the doses were significantly low, and the clearance criteria were satisfied. Based on the safety assessment results, an appropriate metal recycle and disposal method were suggested for clearance, which are the subject of the deregulation of internal structures of nuclear power plant.

• Journal of Radiation Protection and Research
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• v.28 no.3
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• pp.199-206
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• 2003

Because the MIRD phantom, the representative mathematical phantom was developed for the calculation of internal radiation dose, and simulated by the simplified mathematical equations for rapid computation, the appropriateness of application to external dose calculation and the closeness to real human body should be justified. This study was intended to modify the MIRD phantom according to the comparison of the organ absorbed doses in the two phantoms exposed to monoenergetic broad parallel photon beams of the energy between 0.05 MeV and 10 MeV. The organ absorbed doses of the MIRD phantom and the Zubal yokel phantom were calculated for AP and PA geometries by MCNP4C, general-purpose Monte Carlo code. The MIRD phantom received higher doses than the Zubal phantom for both AP and PA geometries. Effective dose in PA geometry for 0.05 MeV photon beams showed the difference up to 50%. Anatomical axial views of the two phantoms revealed the thinner trunk thickness of the MIRD phantom than that of the Zubal phantom. To find out the optimal thickness of trunk, the difference of effective doses for 0.5 MeV photon beams for various trunk thickness of the MIRD phantom from 20 cm to 36 cm were compared. The optimal thunk thickness, 24 cm and 28 cm for AP and PA geometries, respectively, showed the minimum difference of effective doses between the two phantoms. The trunk model of the MIRD phantom was modified and the organ doses were recalculated using the modified MIRD phantom. The differences of effective dose for AP and PA geometries reduced to 7.3% and the overestimation of organ doses decreased, too. Because MIRD-type phantoms are easier to be adopted in Monte Carlo calculations and to standardize, the modifications of the MIRD phantom allow us to hold the advantage of MIRD-type phantoms over a voxel phantom and alleviate the anatomical difference and consequent disagreement in dose calculation.

• The Korean Journal of Nuclear Medicine Technology
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• v.16 no.2
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• pp.126-130
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• 2012

Purpose : Methode an effective block was investigated to deal with volatile radioactive gas, short lived radioactive waste generated as a result of the routinely produced radiopharmaceuticals FDG (2-deoxy-2-[$^{18}F$]fluoro-D-glucose) and compound with $^{11}C$. Materials and Methods : All components of the radiation stack monitoring and data management system for continuous radioactive gas detection in the air extract system purchase from fixed noble gas monitor of Berthold company. TEDLAR gas sampling bags purchase from the Dongbanghitech company. TEDLAR gas sampling bags (volume: 10 L) connected via paraflex or PTFE tubing and Teflon 3 way stopcock. When installing TEDLAR gas sampling bags in Hot cell on the inside and not radioactive gas concentrations were compared. According to whether the Hot cell inside a activated carbon filter installed, compare the difference in concentration of the radioactive gas $^{18}F$. Comparison of radiation emission concentration difference of module a FASTlab and TRACElab. Results : Activated carbon filter are installed in the Hot cell, a measure of the concentration of radioactive gas was 8 $Bq/m^3$. Without activated carbone filter in the hot cell was 300 $Bq/m^3$. Tedlar bag prior to installation of the radioactive gases a measure of the concentration was 3,500 $Bq/m^3$, $^{11}C$ synthesis of the measured concentration was 27,000 $Bq/m^3$. After installed a Tedlar bag and a measure concentration of the radioactive gases was 300 $Bq/m^3$ and $^{11}C$ synthesis was 1,000$Bq/m^3$. Conclusion : $^{11}C$ radioactive gas that was ejected out of the Hot cell, with the use of a Tedlar gas sampling bag stored inside. A compound of 11C is not absorbed onto activated carbon filter. But can block the release out by storing in a Tedlar gas sampling bag. We was able to reduce the radiation exposure of the worker by efficient radiation protection.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.18 no.3
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• pp.657-665
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• 2017

The aim of this study was to develop and distribute a dedicated program that can easily calculate the effective dose of a patient undergoing nuclear medicine examinations, and assist in the study of dose of nuclear medicine examinations and information disclosure. The program produced a database of the effective dose per unit activity administered (mSv/MBq) of the radiopharmaceuticals listed in ICRP 80, 106 Report and the fourth addendum, was designed through Microsoft Visual Basic (In Excel) to take the effect of 5 different (Area, Clark, Solomon(=Fried), Webster, Young) of pediatric dose calculation methods and 7 different body surface area calculation methods. The program calculates the effective dose (mSv) when the age, radionuclide, substance, and amount injected in the human body is inputted. In pediatric cases, when the age is entered, the pediatric method is activated and the pediatric method to be applied can be selected. When the BSA (Body Surface Area) formula is selected in the pediatric calculation method, a selection window for selecting the body surface area calculation method is activated. When the adult dose is input, the infant dose and the effective dose (mSv) are calculated automatically. The patient effective dose calculation program of the nuclear medicine examinations produced in this study is meaningful as a tool for calculating the internal exposure dose of the human body that is most likely to be obtained in nuclear medicine examinations, even though it is not the actual measurement dose. In the future, to increase the utilization of the program, it will be produced as an application that can be used in mobile devices, so that the public can access it easily.

• Proceedings of the Korean Geotechical Society Conference
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• 2006.03a
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• pp.42-50
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• 2006

역들의 공간적 분포가 불균질하고 역의 크기가 큰 역암의 경우 암석 전체를 대표하는 물성치($E_m,\;c,\;\Phi$ 등) 구하기 위해서는 매우 큰 시험기기가 필요하다. 따라서 커다란 역을 포함하는 역암의 경우 직접 암석실내시험을 통한 물성치 산정은 현실적으로 거의 불가능하다. 이러한 문제를 극복하기 위하여 이 연구에서는 개별요소법을 이용하여 역암의 물성치를 산출하는 방법을 제안한다. 그 방법은다음과 같다. (1) 역암내의 역의 물성과 기질부의 물성을 각각 실내실험을 통하여 파악한 후 이들 (2) 두 물질의 거동양상을 구현할 수 있는 개별요소집합체의 개별요소간의 물성을 결정한다. (3) 역의 함량, 크기 모양 공간적 분포양상등의 역암 조직과 유사한 개별요소 수치해석시료를 만든 후, (4) 이를 수치 해석실험 (이축압축실험)에 사용한다. 이러한 수치해석실험을 통해 현재까지 만들어진 결과는 다음과 같다. 첫째, 역의 강도가 기질의 강도보다 높은 역암의 경우, 역의 양이 증가할수록 일축압축강도, 내부 마찰각, 점착력이 증가하지만 증가 양상은 선형이 아니다. 탄성계수의 경우 역의 양과 상관 없이 변화하지 않는다. 둘째, 역과 기질 사이 표면의 점착력이 약할 경우 이러한 표면에서 최초 미세 균열이 형성되기 시작하므로 이 점착력은 물성치를 산출하는 중요한 인자이다. 따라서, 향후 이에 대한 자세한 연구가 필요하다고 판단된다. 결론적으로,설계 또는 시공시 직접시험에 의한 물성치의 파악이 어려운 역암 또는 직접시험을 위해 대량의 시료를 필요로 하는 함력 미고결지층, 핵석층, 풍화암과 같은 시료의 물성치는 별도로 측정된 물성들 (예, 역과 기질)을 이용한 개별요소법을 통해 구할 수 있다.로 나타났다.TEX>, DIN/DIP비 표층수 $23.91\pm3.42$, 저층수 $23.43\pm3.38$이었으며, 전반적으로 해역별 수질기준 I등급 내지는 II등급을 유지하고 있었고, 공간적으로는 외해측으로 갈수록 외해수와 혼합 확산되어 양호한 수질을 나타내었다. 장기적인 변동특성은 세그룹으로 구분되어진다.기 실험결과 용출용매로 증류수와 해수를 이용했을 때, 제강 슬래그에서 용출되는 납, 구리, 카드뮴, 수은의 용출 경향의 차이를 확인할 수 있었고 이에 따라서, 납, 구리, 카드뮴의 용출 유해성은 낮기 때문에 해양구조물로의 제강슬래그 유효이용은 적합할 것으로 판단되었다.im80%$로 계산되었다. 열형광선량계로 측정된 방사선량은 각각 1.8, 1.2, 0.8, 1.2, 0.8 (70 cm 거리) cGy로 측정되었으며, 환자의 복부 표면에서의 서베이메터를 이용한 측정량은 10.9 mR/h였다. 차폐구조물의 사용 시 전체 치료 동안에 태아선량은 약 1 cGy 정도로 평가되었다. 결론 : AAPM Report No.50의 자료에 따르면, 임산부의 방사선 치료 시 태아의 방사선 피폭선량은 5 cGy 이하일 경우에 방사선 피폭에 따른 태아의 위험이 거의 없는 것으로 제시되고 있다. 본원에서 차폐 구조물을 설치하였을 경우에 측정된 태아선량은 약 1 cGy로 측정되었고, 고안된 차폐구조물은 태아에 도달하는 방사선량을 감소시키기에 적합한 설계임이 입증되었다. 아니라 일반종합병원에서도 CTX-M형 ESBL 생성 E. coli와 K. pneumoniae가 존재하며 확산 중임을 시사한다. 앞으로 CTX-M형 ESBL의 만연과 변종 CTX-M형 ESBL의 출연을 감시하기 위한

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.1
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• pp.103-111
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• 2020

Pulling-type cutting devices, which use a diamond wire saw, have been used generally for cutting concrete structures. In this study, a pushing-type cutting device with a collection cover was developed by overcoming the disadvantages of pulling-type devices. In this device, dry or liquid methods can be selected to cool frictional heat. Operation and leakage tests of the dust generated during the dismantling of a concrete structure were carried out, confirming the suitable operation of the fabricated cutting device; the leakage rate was approximately 1.7%. For a conservative evaluation, the internal dose of workers was estimated in dismantling the core center part of biological shield concrete with a specific activity of 99.5 Bq·g^-1. The committed effective dose per worker was 0.25 mSv. The developed cutting device contributed to reducing radioactive concrete waste and minimizing worker exposure due to its easy installation. Therefore, it can be utilized as a cutting apparatus for dismantling not only reinforced concrete structures but also radioactive biological shield concrete in nuclear power plant decommissioning efforts.

• The Korean Journal of Nuclear Medicine Technology
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• v.18 no.1
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• pp.127-133
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• 2014

Purpose: Low dose of PET/CT is important because of Patient's X-ray exposure. The aim of this study was to evaluate the effectiveness of low-dose PET/ CT image through the CTAC and QAC of patient study and phantom study. Materials and Methods: We used the discovery 710 PET/CT (GE). We used the NEMA IEC body phantom for evaluating the PET data corrected by ultra-low dose CT attenuation correction method and NU2-94 phantom for uniformity. After injection of 70.78 MBq and 22.2 MBq of 18 F-FDG were done to each of phantom, PET/CT scans were obtained. PET data were reconstructed by using of CTAC of which dose was for the diagnosis CT and Q. AC of which was only for attenuation correction. Quantitative analysis was performed by use of horizontal profile and vertical profile. Reference data which were corrected by CTAC were compared to PET data which was corrected by the ultra-low dose. The relative error was assessed. Patients with over weighted and normal weight also underwent a PET/CT scans according to low dose protocol and standard dose protocol. Relative error and signal to noise ratio of SUV were analyzed. Results: In the results of phantom test, phantom PET data were corrected by CTAC and Q.AC and they were compared each other. The relative error of Q.AC profile was been calculated, and it was shown in graph. In patient studies, PET data for overweight patient and normal weight patient were reconstructed by CTAC and Q.AC under routine dose and ultra-low dose. When routine dose was used, the relative error was small. When high dose was used, the result of overweight patient was effectively corrected by Q.AC. Conclusion: In phantom study, CTAC method with 80 kVp and 10 mA was resulted in bead hardening artifact. PET data corrected by ultra- low dose CTAC was not quantified, but those by the same dose were quantified properly. In patients' cases, PET data of over weighted patient could be quantified by Q.AC method. Its relative difference was not significant. Q.AC method was proper attenuation correction method when ultra-low dose was used. As a result, it is expected that Q.AC is a good method in order to reduce patient's exposure dose.

• 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).

• Journal of the Korean Society of Radiology
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• v.11 no.6
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• pp.423-428
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• 2017

In this study, the calculation of the effective spatial dose distribution of the diagnostic imaging laboratory of K university was performed by the Monte Carlo simulation. The radiation generator has a maximum tube voltage of 150 kVp and a maximum current of 700 mA. Using the results, we compared the spatial effective dose distributions of diagnostic imaging laboratory when the shielding door was closed and opened. In conclusion, it was found that the effective dose in the operating room of the diagnostic imaging laboratory does not exceed the annual dose limit (6 mSv/y) of the student (occasional visitor) even when the door is opened. However, since the effective dose when the door is open is about 16 times higher in front of the lead glass window and about 3,000 times higher in front of the doorway than the case when the door is closed, closing the shielding door at the time of the practical exercising reduces unnecessary radiation exposure by great extent.

• Journal of the Korean Society of Radiology
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• v.11 no.6
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• pp.531-538
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• 2017

Agricultural products produced in the agricultural area around the nuclear power plant are radioactive contamination, which can cause radioactive contamination to the human body. The purpose of this study was to investigate the limit of the radioactivity concentration $^{90}Sr$ for the internal exposure dose evaluation by ingesting the agricultural products collected around the nuclear power plant. The results of the gamma-isotope element analysis were freshly <0.0166-0.0336 Bq / kg for all samples and for artificial radionuclides not detected, and fresh <0.00586-0.0421 Bq / kg for Chinese cabbage, The freshness was 0.106 Bq / kg, and the freshness was 0.0114-0.0901 Bq / kg. 0.0177%, 0.0222%, 0.0376% and 0.00243%, respectively, for Chinese cabbages and large roots, which is lower than the legal standard value of $1mSv/yr{\cdot}man%$. It is considered that the formulas need to be broadly evaluated for the foods consumed by children and adults, taking into consideration the age of the food and the diet