• 한국의학물리학회지:의학물리
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• 제27권3호
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• pp.156-161
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• 2016

When air discharged from a radioisotope production facility is contaminated with radiation, the public may be exposed to radiation. The objective of this study is to manage such radiation exposure. We measured the airborne radioactivity concentration at a 30 MeV cyclotron radioisotope production facility to assess whether the exhaust gas was contaminated. Additionally, we investigted the radioactive contamination of the air filter for efficient air purification and radiation safety control. To measure the airborne radiation concentration, specimens were collected weekly for 4 h after the beginning of the radioisotope production. Regarding the air purifier, five specimens were collected at different positions of each filter-pre-filter, high-efficiency particulate air filter, and charcoal filter-installed in the cyclotron production room. The concentrations of F-18, I-123, I-131, and Tl-201 generated in the radioiodine production room were $13.5Bq/m^3$, $27.0Bq/m^3$, $0.10Bq/m^3$, and $11.5Bq/m^3$, respectively; the concentrations of F-18, I-123, and I-131 produced in the radioisotope production room were $0.05Bq/m^3$, $16.1Bq/m^3$, and $0.45Bq/m^3$, correspondingly; and those of F-18, I-123, I-131, and Tl-201 generated in the accelerator room were $2.07Bq/m^3$, $53.0Bq/m^3$, $0.37Bq/m^3$, and $0.15Bq/m^3$, respectively. The maximum radiation concentration of I-123 generated in the radioiodine production room was 1,820 Bq/g, which can be disposed after 2 days. The maximum radiation concentration of Tl-202 generated in the radioisotope production room was 205 Bq/g, and this isotope must be stored for 53 days. The I-123 generated in the radioiodine production room had a maximum concentration of 1,530 Bq/g and must be stored for 2 days. The maximum radiation concentration of Na-22 generated in the radioisotope production room was 0.18 Bq/g and this isotope must be disposed after 827 days. To manage the exhaust, the efficiency of air purification must be enhanced by selecting an air purifier with a long life and determining the appropriate replacement time by examining the differential pressure through systematic measurements of the airborne radiation contamination level.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2018년도 춘계학술논문요약집
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• pp.184-185
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• 2018

• 대한방사성의약품학회지
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• 제1권1호
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• pp.23-30
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• 2015

Radioisotopes emitting low-range highly ionizing radiation such as ${\beta}$-particles are of increasing significance in internal radiotherapy. Among the ${\beta}$-particle emitting radioisotopes, $^{67}Cu$ is an attractive radioisotope for various nuclear medicine applications due to its medium energy ${\beta}$-particle, gamma emissions, and 61.83-hour half-life, which can also be used with $^{64}Cu$ for PET imaging. The production and application of the ${\beta}$-emitting radioisotope $^{67}Cu$ for therapeutic radiopharmaceutical are outlined, and different production routes are discussed. A survey of copper chelators used for antibody labeling is provided. It has been produced via proton, alpha, neutron, and gamma irradiations followed by solvent extraction, ion exchange, electrodeposition. Clinical studies using $^{67}Cu$-labelled antibodies in lymphoma, colon carcinoma and bladder cancer patients are reviewed. Widespread use of this isotope for clinical studies and preliminary treatments has been limited by unreliable supplies, cost, and difficulty in obtaining therapeutic quantities.

• Nuclear Engineering and Technology
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• 제52권7호
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• pp.1517-1523
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• 2020

The widely used medical isotope technetium-99 m (^99mTc) is a daughter of Molybdenum-99 (⁹⁹Mo), which is mainly produced using dedicated research reactors from the nuclear fission of uranium-235 (²³⁵U). ^99mTc has been used for several decades, which covers about 80% of the all the nuclear diagnostics procedures. Recently, the instability of the supply has become an important topic throughout the international radioisotope communities. The aging of major ⁹⁹Mo production reactors has also caused frequent shutdowns. It has triggered movements to establish new research reactors for ⁹⁹Mo production, as well as the development of various ⁹⁹Mo production technologies. In this context, a new research reactor project was launched in 2012 in Korea. At the same time, the development of fission-based ⁹⁹Mo production process was initiated by Korea Atomic Energy Research Institute (KAERI) in 2012 in order to be implemented by the new research reactor. The KAERI process is based on the caustic dissolution of plate-type LEU (low enriched uranium) dispersion targets, followed by the separation and purification using a series of columns. The development of proper waste treatment technologies for the gaseous, liquid, and solid radioactive wastes also took place. The first stage of this process development was completed in 2018. In this paper, the results of the hot test production of fission ⁹⁹Mo using HANARO, KAERI's 30 MW research reactor, was described.

• Nuclear Engineering and Technology
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• 제48권3호
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• pp.613-623
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• 2016

Molybdenum-99 ($^{99}Mo$) is the most important isotope because its daughter isotope, technetium-99m ($^{99m}Tc$), has been the most widely used medical radioisotope for more than 50 years, accounting for > 80% of total nuclear diagnostics worldwide. In this review, radiochemical routes for the production of $^{99}Mo$, and the aspects for selecting a suitable process strategy are discussed from the historical viewpoint of $^{99}Mo$ technology developments. Most of the industrial-scale $^{99}Mo$ processes have been based on the fission of $^{235}U$. Recently, important issues have been raised for the conversion of fission $^{99}Mo$ targets from highly enriched uranium to low enriched uranium (LEU). The development of new LEU targets with higher density was requested to compensate for the loss of $^{99}Mo$ yield, caused by a significant reduction of $^{235}U$ enrichment, from the conversion. As the dramatic increment of intermediate level liquid waste is also expected from the conversion, an effective strategy to reduce the waste generation from the fission $^{99}Mo$ production is required. The mitigation of radioxenon emission from medical radioisotope production facilities is discussed in relation with the monitoring of nuclear explosions and comprehensive nuclear test ban. Lastly, the $^{99}Mo$ production process paired with the Korea Atomic Energy Research Institute's own LEU target is proposed as one of the most suitable processes for the LEU target.

• 방사선산업학회지
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• 제17권4호
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• pp.543-549
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• 2023

Thallium-201 (²⁰¹Tl) is a medical radioisotope which emits gamma rays when it decays and used in myocardial perfusion scans in single-photon emission tomography due to its similar properties to potassium. Currently, the Korea Institute of Radiological & Medical Sciences is the only institution producing ²⁰¹Tl in Korea, and optimization of ²⁰¹Tl production research is necessary to meet supply compared to domestic demand. To this end, technical analysis of plating target production and chemical separation methods essential for ²⁰¹Tl production research is conducted. It deals with the process of generating and separating ²⁰¹Tl radioisotope and target production, It can be generated through a nuclear reaction such as ^natHg(p,xn)²⁰¹Tl, ²⁰¹Hg(p,n)²⁰¹Tl, ^natPb(p,xn)²⁰¹Bi → ²⁰¹Pb → ²⁰¹Tl, ²⁰⁵Tl(p,5n)²⁰¹Pb → ²⁰¹Tl, and considering impure nuclide generated simultaneously with the use of proton beam energy of 35 MeV or less, it is intended to be produced using the ²⁰³Tl(p,3n)²⁰¹Pb→²⁰¹Tl nuclear reaction. In particular, the chemical separation of Tl is a very important element, and the chemical separation methods that can separate it is broadly divided into four types, including solid phase extraction, liquid-liquid, electrochemical, and ion exchange membrane separation. Some chemical separations require additional separation steps, such as methods using selective adsorption. Therefore, this technical report describes four chemical separation methods and seeks to separate high-purity ²⁰¹Tl using a method without additional separation steps

• Journal of Radiation Protection and Research
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• 제38권2호
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• pp.52-59
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• 2013

전베타 방사능 분석법을 이용한 내부오염 스크리닝법을 검증하였고 실제 의료용 동위원소 생산시설 종사자 내부오염을 판단하는데 적용하였다. 종사자의 작업 종료 후 첫 번째로 채취된 뇨시료(spot 시료)와 24시간 동안 취합된 뇨시료(24 h 시료)를 채취하여 측정하였다. 특정 종사자의 경우를 제외하고 대부분의 측정결과는 일반인 체내 기저준위인 100 Bq $kg^{-1}$을 기준으로 22% 이내로 변동폭이 작았다. 측정결과 작업종료 후 수 시간 이내 종사자 뇨시료의 전베타 농도가 전반적으로 35% 이상 상승하는 경향이 있었다. 또한 스크리닝 결과와 작업일지를 바탕으로 작업장내부 구조상 오염을 유발하는 요인을 추정 할 수 있었으며 추가 세부 핵종별 분석법을 바탕으로 내부피폭선량을 평가해야 할 것으로 판단되었다. 한편 사업장에서 신속히 적용 가능한 내부오염평가 절차를 수립하였다.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2018년도 추계학술논문요약집
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• pp.271-272
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• 2018

• Journal of Pharmaceutical Investigation
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• 제33권2호
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• pp.145-149
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• 2003

Manufacturing facilities of the pharmaceuticals must meet certain level of the cleanness required so that foreign substances such as dust, moisture, heat, microorganism, or virus do not contaminate the product. In case of radiopharmaceuticals for medical treatment and diagnosis, not only should the operators and environment be protected from radiation but also need to be isolated from the foreign contaminant. Therefore, manufacturing facilities for radiopharmaceuticals must satisfy the design standards of both hot cell and clean room which are specified by GMP. However, standards of maintaining negative pressure for preventing spread of radioactive contaminant in isolated facilities conflict with the standards of maintaining positive pressure for keeping cleanness. To solve this problem, air pressure of hot cell was designed lower than in the adjacent area to meet standards of the radiation safety. To keep higher cleanness in certain part of the hot cell for filling, minimal relative positive pressure allows. In order to effectively maintain the cleanness that is required for production of Tc-99m generator, which takes 70% of whole demand of radiopharmaceuticals, the rooms placed in each side of production room are used as a buffer area and three lead hot cells are installed in production room. In this research, we established the appropriate engineered design concept for Tc-99m generator manufacturing facility, which satisfies both GMP cleanness standard for preventing particles, bacteria, other contaminants and the regulations of radiation safety for supervising and controlling the amount of radiation exposure and exhausted radioactivity. And the concept of multi-barrier buffer zones is introduced to apply negative air pressure for hot cell with first priority and to continue relative positive air pressure for clean room.

• Nuclear Engineering and Technology
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• 제32권4호
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• pp.361-371
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• 2000

The 123Te(p,n)123I, 124Te(p,n)124I and 124Te(p,2n)123I reactions, among the many reaction channels opened, are the major reactions under consideration from a diagnostic purpose because reaction residuals as the gamma emitters are used for most radiophamaceutical applications involving radioiodine. Based on the available experimental data, the absorption cross sections and elastic scattering angular distributions of the proton-induced nuclear reaction on Te isotopes below 60 MeV are calculated using the optical model code APMNK. The transmission coefficients of neutron, proton, deuteron, trition and alpha particles are calculated by CUNF code and are fed into the GNASH code. By adjusting level density parameters and the pair correction values of some reaction channels, as well as the composite nucleus state density constants of the pre-equilibrium model, the production cross sections and energy-angle correlated spectra of the secondary light particles, as well as production cross sections and energy distributions of heavy recoils and gamma rays are calculated by the statistical plus pre-equilibrium model code GNASH. The calculated results are analysed and compared with the experimental data taken from the EXFOR. The optimized global optical model parameters give overall agreement with the experimental data over both the entire energy range and all tellurium isotopes.