• Radioisotope journal
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• v.20 no.2
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• pp.28-30
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• 2005

• Proceedings of the Korean Vacuum Society Conference
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• 2011.02a
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• pp.235-235
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• 2011

진단용 또는 의료용 동위원소들은 안정한 표적물질에 높은 에너지의 양성자가 조사 될 때 핵반응에 의해서 생성된다. 양성자를 충분한 에너지로 가속하기 위해서 이용되는 사이클로트론의 주요 부분은 (1) 진공시스템, (2) 자석시스템, (3) RF 시스템, (4) 외부 이온원, (5) 수직 축 방향빔의 수평방향 전환 시스템, (6) 빔 인출 장치, 그리고 빔전송과 표적장치로 구성된다. 인출된 빔은 표적까지 손실 없이 전송 될 수 있도록 빔 라인에 설치된 광학적 요소에 의해 집속되어 전송된다. 방사성동위원소의 생산량은 양성자 빔의 특성과 표적 물질의 종류에 따라 결정된다. 즉, 표적 물질에 조사하는 입자의 종류, 적절한 핵반응 선택, 최소량의 불순핵종과 원하는 방사핵종의 최대수율을 얻을 수 있는 최적 에너지 범위결정, 표적 물질의 냉각능력과 입자전류의 세기 등을 고려 하여야 한다. 동위원소 생산에 있어서 예측되는 수율은 입자전류와 비례하며, 에너지에 대한 핵반응 단면적 즉, 여기함수를 적분하여 아래와 같이 얻을 수 있다. 주 생성핵종의 생산 효율을 최대로 높이고 불순 핵종의 생성량을 최소로 감소시키기 위해서는 정확한 여기 함수 자료를 바탕으로 최적 입자를 결정하여야 한다. 또한 이론적인 생산 수율은 입자 전류에 정비례하지만, 입자 전류가 클경우 생산수율은 이론적인 수율보다 적다. 입자빔의 불균일성, 표적의 방사선 피폭에 의한 손상, 높은 입자전류에 의해 발생하는 열로 인하여 생성 핵종이 증발하여 생산 수율이 감소된다. 본 발표에서 방사핵종 C-11과 Tc-99m을 개발하기 위한 최적 조건에 관한 연구결과를 보고하고자 한다.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 2004.11a
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• pp.311-316
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• 2004

• Proceedings of the Korean Vacuum Society Conference
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• 2012.02a
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• pp.173-173
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• 2012

사이클로트론은 암진단에 사용되는 방사성동위원소를 생산하기 위한 중요한 입자 가속장치이다. 현재 핵의학 의료진단에 필요한 방사성동위원소를 제공하기 위해 세계적으로 사이클로트론의 활용도가 점점 증가하고 있다. 한국원자력의학원에 설치된 MC50 (양성자 최대 가속에너지 50 MeV, 60 uA)과 C30 (양성자 최대 에너지 30 MeV, 250 uA) 사이클로트론은 생명의학, 반도체 검출기, 핵자료 데이터, 방사성동위원소 개발 등 다양한 분야의 연구를 지원하고 있다. MC50 사이클로트론은 수소 입자를 포함하여 중양자, 알파 입자를 가속할 수 있으며 중성자 빔을 인출 할 수 있다. 수소 음이온 또는 양이온을 가속 할 수 있으며 표적에는 고에너지의 양이온이 조사되며, 핵반응을 통해 방사성동위원소가 생성된다. 양성자 빔을 이용하여 암세포를 사멸 시키는 치료법, 돌연변이로 새로운 종의 개발 등 다양한 응용성이 있다. 하전입자를 가속하는 사이클로트론의 주요 구성요소는 (1) 진공시스템, (2) 전자석 시스템, (3) 고주파 시스템, (4) 이온원 (5) 빔 인출장치 (6) 빔전환 장치 (수직에서 수평 방향으로 전환), (7) 빔 집속 및 진단 장치 등 이다. 본 발표에서는 85년부터 운영한 MC50 사이클로트론과 02년부터 가동된 사이클로트론의 운영 현황 및 다양한 응용분야와 향후 RI 빔 인출을 위한 계획을 소개하고자 한다.

• 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 radiological science and technology
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• v.40 no.1
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• pp.49-55
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• 2017

In the case of nuclear medicine practitioners in medical institutions, a wide range of exposure dose to individual workers can be found, depending on the type of source, the amount of radioactivity, and the use of shielding devices in handling radioactive isotopes. In this regard, this study evaluated the organ dose on practitioners as well as the dose reduction effect of the L-block shielding device in handling the diagnostic radiation source through the simulation based on the Monte Carlo method. As a result, the distribution of organ dose was found to be higher as the position of the radiation source was closer to the handling position of a practitioner, and the effective dose distribution was different according to the ICRP tissue weight. Furthermore, the dose reduction effect according to the L-block thickness tended to decrease, which showed the exponential distribution, as the shielding thickness increased. The dose reduction effect according to each radiation source showed a low shielding effect in proportion to the emitted gamma ray energy level.

• Progress in Medical Physics
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• v.25 no.1
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• pp.8-14
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• 2014

The amounts of radioactive wastes to be disposed in the medical institute have been increased due to development of radiation diagnosis and therapy rapidly. They are produced mostly by the very short lived radioisotopes such as $^{18}F$ used in PET/CT, $^{99m}Tc$, $^{123}I$, $^{125}I$ and $^{201}Tl$, etc. IAEA proposed a criteria for the clearance level of waste which depends on the individual ($10{\mu}Sv/y$) and collective dose (1 man-Sv/y), and concentration of each nuclide (IAEA Safety Series No 111-P-1.1, 1992 and IAEA RS-G-1.7, 2004). Radioactive wastes of $^{18}F$, $^{99m}Tc$, $^{123}I$, $^{125}I$ and $^{201}TI$ in the several types of container like Marinelli beaker, vial and plastic, were collected to measure the concentration of the waste of each nuclide in accordance with IAEA criteria. The measurement method and procedure of determining specific activity of the wastes using gamma emitters like MCA, gamma counter and beta emitters were developed. For the efficiency calibration of the detectors, CRM (certified reference material) which has the same dimension and shape was provided by Korea Research Institute of Standards and Science (KRISS). Correction factor of the radioactivity decay was calculated based on the measurement results, and the consideration of mutual relation with theoretical equation. The result of this study will be proposed as ISO standard.

• The Korean Journal of Nuclear Medicine Technology
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• v.21 no.2
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• pp.13-27
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• 2017

Purpose In the procedure of domestic medical radioactive self-disposal, there are many requests of supplementation and difficulties on the screening process. In this regard, presentation of basic guideline will improve the work processing efficiency of medical institution radioactive waste. From 2015 to 2016, We reviewed and compared a supplementary requests of domestic fifteen medical institution radioactive self-disposal Plan & Procedure manual. In connection with this, we derive the details of the radioactive waste document based on the relative regulation of nuclear safety Act. The representative supplementary requests of Korea Institute of Nuclear Safety are disposal method of non-flammability radioactive waste, storage method of scheduled self-disposal waste, the legitimacy of self-disposal and pre-treatment of self-disposal, reference radioactivity of disused filter and output of storage period, attachment the evidential matter of measurement efficiency when using a gamma counter. Through establishing a medical radioactive waste guideline, we can clearly suggest a classification standard of radioactive nuclide and the type of occurrence. As a result, we can confirm the reduction of examination processing period while preparing a self-disposal document and there is no spending expenses for business agency. Also, the storage efficiency of facility will better and reduce the economic expenses. On the basis of this guideline, we will expect a contribution to the improvement of work efficiency for officials who has a working-level difficulty of radioactive waste self-disposal.

• Progress in Medical Physics
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• v.15 no.4
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• pp.210-214
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• 2004

Since production of radioactive isotope for using PET, a lot of neutrons were produced. The produced neutrons were mainly shielded by concrete facility. Secondary photons are generated and emitted from the concrete shielding wall of the PET cyclotron since the proton-generated neutrons are thermalized and absorbed in the concrete wall and emit secondary radiations, i.e., photons. This study calculated neutron dose and photon dose at outside of the accelerator facility using MCNPX code. As results of the calculation, total dose were calculated less than limited dose by law.

• Journal of the Korea Convergence Society
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• v.9 no.7
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• pp.87-93
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• 2018

In this study, $^{99m}Tc$, $^{123}I$, $^{201}Tl$, $^{18}F$, and $^{131}I$, which are widely used in nuclear medicine, were transmitted through a bismuth shield. We investigated the shielding rates according to the type of radioisotope and the distance of measurement. For the experiment, 6 sheets of lead equivalent 0.25 mm Pb of bismuth shielding material were stacked one by one up to 1.50 mm as the thickness increased. The distance was 30 cm, 50 cm, and 100 cm, and the transmission dose was measured. As a result, the shielding rates was measured as the thickness increased, and the measured value decreased as the distance increased. The shielding rate of $^{123}I$ and $^{201}Tl$ was higher than $^{99m}Tc$, $^{18}F$ and $^{131}I$ showed lower shielding effect when there is a shielding material than when there is no shielding material due to high energy and ${\beta}$ rays. Based on the results of experiments, it would be helpful to reduce the exposure of nuclear medicine workers and to manage the exposure if bismuth shields are used depending on the type of radioisotope.