• Nuclear Engineering and Technology
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• v.52 no.7
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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.

• Journal of Radiopharmaceuticals and Molecular Probes
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• v.5 no.1
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• pp.3-10
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• 2019

The activities per volume of $^{188}Re$ and $^{99m}Tc$ from their generators are dependent on the specific activity of their mother nuclides $^{188}W$ and $^{99}Mo$ respectively. After a particular lapse of time, the eluted RI activity is exponentially reduced and thus cannot satisfy the needs of clinical application. The purpose of this study is to develop a $^{188}Re$ and $^{99m}Tc$ concentration device with a compact size that can extend the period of use as well as conveniently concentrate the RI. We designed the concentration module by including two-different check valves that do not required any manual on-off operations. In these concentration process, cation exchange resin embedded with Ag and anion exchange resins were used. After completing the concentrating step, the recovering yield was identified to be more than 93% for $^{188}Re$ generators and 88% for $^{99m}Tc$ generators. Moreover, all these procedures were done within 5 min.

• Nuclear Engineering and Technology
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• v.48 no.3
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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.

• The Korean Journal of Nuclear Medicine Technology
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• v.20 no.2
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• pp.36-41
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• 2016

Purpose Sentinel lymphoscintigraphy (SLS) was using only $^{99m}Tc-phytate$. If the supply is interrupted temporarily, there is no alternative radiopharmaceuticals. The aim of this study measure the particle size of radiopharmaceuticals and look for radiopharmaceuticals which can be substituted for $^{99m}Tc-phytate$. Materials and Methods The particle size of radiopharmaceuticals were analyzed by a nano-particle analyzer. This study were selected known radiopharmaceuticals to be useful particle size for SLS. We were divided into control and experimental groups using $^{99m}Tc-DPD$, $^{99m}Tc-MAG3$, $^{99m}Tc-DMSA$ with $^{99m}Tc-phytate$. For in-vivo experiment, radiopharmaceuticals were injected intradermally at both foot to perform lymphoscintigraphy. Imaging was acquired to dynamic and delayed static image and observe the inguinal lymph nodes with the naked eye. Results Particle size was measured respectively Phytate 105~255 nm (81.9%), MAG3 91~255 nm (98.7%), DPD 105~342 nm (77.3%), DMSA 164~ 342 nm (99.2%), MAA 1281~2305 nm (90.6%), DTPA 342~1106 nm (79.4%), and HDP 295~955 nm (94%). In-vivo delayed static image, inguinal lymph nodes of all experiment groups and two control groups are visible to naked eye. however, $^{99m}Tc-MAG3$ of control groups is not visible to naked eye. Conclusion We were analyzed to the particle size of the radiopharmaceuticals that are used in in-vivo. Consequently, $^{99m}Tc-DPD$, $^{99m}Tc-DMSA $are possible in an alternative radiopharmaceuticals of emergency.

• Nuclear Engineering and Technology
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• v.55 no.1
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• pp.304-309
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• 2023

In the last decade, nuclear medicine appears to be a good choice of medicine. ⁵⁸Co, ⁹⁹Mo, ⁹⁹Tc, ⁹⁹Re, ¹³³Xe and ¹⁸⁶Re are very important radionuclides for nuclear medicine. In this study, the excitation functions of ⁵⁸Ni (n, p) ⁵⁸Co, ⁹⁹Tc (n, p) ⁹⁹Mo, ⁹⁹Ru (n, p) ⁹⁹Tc, ¹³¹Xe (n, p) ¹³¹I, ¹³³Cs (n, p) ¹³³Xe and ¹⁸⁶Os (n, p) ¹⁸⁶Re nuclear reactions were calculated at neutron energies between 1 and 20 MeV using TALYS 1.95 and EMPIRE 3.2 nuclear codes. Furthermore, the cross sections were calculated with the empirical formula derived in our past study at 14-15 MeV. The obtained results were compared with the measured values in EXFOR library, and with the evaluated data of (JENDL-4.0/HE, JEFF-3.3, TENDL-2019, ENDF/B-VIII.0, IRDFF-II, JENDL/ImPACT-18). The results are in good agreement with those of the evaluated data libraries and experimental results and indicates that these radioisotopes can be produced by smaller cyclotrons.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.108-113
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• 1997

하나로(HANARO)를 이용하여 핵분열 방법으로 진단용 방사선원인 $^{99m}$ Tc의 모핵종인 Mo-99를 생산할 경우, HEU 및 LEU UO2 표적이 사용될 수 있다. 표적연료로서 HEU(93w/o $^{235}$ U)가 LEU(19.75w/o $^{235}$ U)에 비해 생성수율(Ci/gU)이 높게 나타났으며 제품의 질(quality)을 좌우하는 비방사능(Ci$^{99}$Mo/gMo)은 같게 나타났다. HEU가 같은 Mo-99의 방사능량을 얻기 위해서는 우라늄 장전량이 적어지므로 폐기물측면과 용해측면에서 이득이나 농축도를 고려하면, 큰 이득이 발생하지 않으므로 하나로에 LEU를 사용하는 것도 타당하다 할 수 있다.

• Proceedings of the Korean Society of Radiological Science
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• 2001.06a
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• pp.88.1-88.1
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• 2001

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2009.06a
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• pp.407-408
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• 2009

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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• pp.521-526
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• 1998

군분리공정의 관심대상 핵종인 잔존 U, MA인 Am, Np 및 백금족원소의 장수명핵종인 Tc을 방사성모의용액으로부터 효과적으로 분리할 수 있었다. 각 단위공정에 주입되는 용액의 농도를 기준으로 하여 볼 때 잔존 U 제거 공정에서 U, Np, 및 Tc의 회수율은 각각 99.1%. 31.9%, 및 99.5%였다. 개미산 탈질 공정에서는 탈질 용액의 총 산도를 0.4M 이하까지 감소시켰으며, 부수적으로 후속용매추출 공정 전반에 걸쳐 문제 핵종으로 작용하는 Zr, Mo, Fe 등을 각각 99.8%, 94.2%, 22.4% 침전 제거시킬 수 있었다. 마지막으로 DEHPA에 의한 AM/RE 상호분리 공정에서는 연속적인 선택적역추출에 의해 Am 및 Np 의 회수율은 각각 99.8% 및 98.2％ 정도였고, Am의 경우 7.6%읜 Fe및 81.6% 의 Mo가, Np의 경우 5.8%의 Fe이 불순물로서 여전히 존재하고 있다. 한편 RE 생성물에서는 97～99.9%의 RE가 회수 제거되었다.

• Journal of Radiation Industry
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• v.4 no.1
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• pp.79-83
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• 2010

The pollutants in subsurface soil are advected by groundwater flow and transported by the hydrodynamic dispersion. In this study, laboratory-scale experiments by using a radioisotope were conducted to evaluate the characteristics of the transport and dispersion of pollutants in the soil. The hydraulic model of the laboratory-scale was manufactured based upon its geometric similarity. Tc-99m having a short half-life was used with a tracer and it was injected instantaneously into the soil. Tc-99m milked from a $^{99}Mo/^{99m}Tc$ portable generator fabricated for medical purposes had 0.141 MeV of gamma radiation. The experiments are performed by the different conditions like the variations of groundwater velocity and the results are analyzed by the measured CPS of Tc-99m.