• Proceedings of the Korean Nuclear Society Conference
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• 2001.10a
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• pp.65-65
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.441.1-441
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• 2002

• Nuclear Engineering and Technology
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• v.47 no.7
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• pp.875-883
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• 2015

The use of subcritical aqueous homogenous reactors driven by accelerators presents an attractive alternative for producing $^{99}Mo$. In this method, the medical isotope production system itself is used to extract $^{99}Mo$ or other radioisotopes so that there is no need to irradiate common targets. In addition, it can operate at much lower power compared to a traditional reactor to produce the same amount of $^{99}Mo$ by irradiating targets. In this study, the neutronic performance and $^{99}Mo$, $^{89}Sr$, and $^{131}I$ production capacity of a subcritical aqueous homogenous reactor fueled with low-enriched uranyl nitrate was evaluated using the MCNPX code. A proton accelerator with a maximum 30-MeV accelerating power was used to run the subcritical core. The computational results indicate a good potential for the modeled system to produce the radioisotopes under completely safe conditions because of the high negative reactivity coefficients of the modeled core. The results show that application of an optimized beam window material can increase the fission power of the aqueous nitrate fuel up to 80%. This accelerator-based procedure using low enriched uranium nitrate fuel to produce radioisotopes presents a potentially competitive alternative in comparison with the reactor-based or other accelerator-based methods. This system produces ~1,500 Ci/wk (~325 6-day Ci) of $^{99}Mo$ at the end of a cycle.

• Journal of the Korean Society of Safety
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• v.12 no.3
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• pp.97-105
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• 1997

Mo-99(Molybdenum) is the only source of Tc-99m(Technetium) which is most frequently used in nuclear medical diagnostics and the demand is on the increase recently. Separation and refining of Mo-99 was investigated by adsorption and desorption on alumina. At pH=0.63, adsorption isotherm of Mo was fitted by Redlich ＆ Peterson equation using the adsorption experimental data. It was found that the pore diffusion model ($D_p=1.4{\times}10^{-6}cm^2/s, K_f/=0.4 cm/s$) agreed well with batch adsorption experimental data. RTDs(Residence Time Distributions ) were measured and axial dispersion coefficients were obtained in the fixed bed absorber according to the changes of the flow rate using 0.05% -NaCl. From the adsorption experimental data, it was shown that the behavior of breakthroughs depended on flow rate. Mo recovery yield was increased as adsorption flow rate was increased and desorption flow rate was decreased.

• The Korean Journal of Nuclear Medicine Technology
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• v.14 no.2
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• pp.83-86
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• 2010

Purpose: The Molybdenum which is the raw material of $^{99}Mo$-$^{99m}Tc$ generator (generator) is produced from the nuclear reactor. However, output has dwindled as the two nuclear reactors supplying the bulk of radioactive material-one in Chalk River, Ontario and the other in Petten, the Netherlands-have been closed for repairs or maintenance. This resulted in the enhancement of its price. Therefore we have tried to seek the new method which could run generator to increase activity of $^{99m}Tc$ in this study. Materials and Methods: The $^{99m}Tc$ activity obtained from 5 times elution for 5 days from Monday to Friday using two generators was compared with 10 times elution. Appearance test, pH test, LAL test, sterility test, chemical impurity(Al) test, radio chemical purity test, ratio of $^{99}Mo$/$^{99m}Tc$ activity test have been done to check the stability of $^{99m}Tc$ eluting from generator respectively. Results: The $^{99m}Tc$ activity obtained from 5 times elution for 5 days was 168.2 GBq (4545 mCi) and 10 times was 230.5 GBq (6230 mCi). All quality control tests were within normal limit. Conclusion: We got to know that 2 times elution a day obtained more $^{99m}Tc$ activity than one time elution in this study.

• Nuclear Engineering and Technology
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• v.45 no.7
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• pp.979-986
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• 2013

Uranium metal particle dispersion plates have been proposed as targets for Molybdenum-99 (Mo-99) production to improve the radioisotope production efficiency of conventional low enriched uranium targets. In this study, uranium powder was produced by centrifugal atomization, and miniature target plates containing uranium particles in an aluminum matrix with uranium densities up to 9 $g-U/cm^3$ were fabricated. Additional heat treatment was applied to convert the uranium particles into UAlx compounds by a chemical reaction of the uranium particles and aluminum matrix. Thus, these target plates can be treated with the same alkaline dissolution process that is used for conventional $UAl_x$ dispersion targets, while increasing the uranium density in the target plates.

• Progress in Medical Physics
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• v.19 no.1
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• pp.1-8
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• 2008

Technisium $(^{99m}Tc)$ is one of the most widely used radioactive isotopes for diagnosis in nuclear medicine. In general, technisium is produced inside the so called $^{99m}Tc$ generator which is usually made out of lead to shield relatively high energy radiation from $^{99}Mo$ and its daughter nuclide $^{99m}Tc$. In this paper, a GEANT4 simulation is carried out to test the safety of the $^{99m}Tc$ generator, taking the Daiichi product with radioactivity of 500 mCi as an example. According to the domestic regulation on radiation safety, the dose at 10 cm and 100 cm away from the surface of shielding container should not exceed 2.0 mSv/h and 0.02 mSv/h, respectively. The simulated dose turned out to be less than the limit, satisfying the domestic regulation.

• Nuclear Engineering and Technology
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• v.4 no.4
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• pp.327-330
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• 1972

$^{99m}$Tc Ferric Hydroxide Macroaggregates for Lung Perfusion Studies were prepared from home made Na $^{99m}$TcO$_4$ which was extracted by methyl ethyl ketone from low activity $^{99m}$Mo. Particle size was in between 20 and 60$\mu$. Rabbit and human body tests gave excellent results.sults.

• The Korean Journal of Nuclear Medicine
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• v.30 no.3
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• pp.361-366
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• 1996

This study was to determine the effect of $Al^{3+}$ in $^{99m}Tc$ eluate from $^{99}Mo-^{99m}Tc$ generator on labeling efficiency and biodistribution of $^{99m}Tc$-MDP. The chromatographic analysis of $^{99m}Tc$-MDP preparations containing $Al^{3+}(0-62.5{\mu}g/ml)$ showed decreased labeling efficiency $^{99m}Tc$ pertechnetate and hydrolyzed reduced $^{99m}Tc$ fraction increased with increasing concentrations of aluminum. However, the chromatography system could not discern between hydrolyzed reduced $^{99m}Tc$ and $^{99m}Tc$ labeled colloid. $^{99m}Tc$-MDP preparations containing aluminum were relatively stable. Chromatographic analysis also confirmed that no significant differences were observed in the radiochemical purity of the filtered and the unfiltered $^{99m}Tc$-MDP preparations containing aluminum by $0.22{\mu}m$ syringe filter. In biodistribution data of ICR-mice, blood and heart uptake were increasing with increasing concentrations of aluminum, because of decreasing labeling efficiency of $^{99m}Tc$-MDP and increasing of $^{99m}Tc$ pertechnetate. However, liver and bone uptake were not significantly increased. In rat images no difference were observed at $5{\mu}g/ml\;Al^{3+}$ compare with at $0{\mu}g/ml\;Al^{3+}$, but at $10{\mu}g/ml\;Al^{3+}$ lumbar uptake was increased. As a practical conclusion, a concentration below $10{\mu}g/ml\;Al^{3+}$($10{\mu}g/ml\;Al^{3+}$ is the maximum allowed in pertechnetate eluate from $^{99}Mo-^{99m}Tc$ generator by USP.) in $^{99m}Tc$-MDP radiopharmaceutical result in low labeling efficiency. Radiochemical purity 90% of $^{99m}Tc$-MDP is the minimum allowed by USP. Therefore, when soft tissue uptake is observed in $^{99m}Tc$-MDP bone scan and labeling efficiency is above 90%, we can expect that $Al^{3+}$ in pertechnetated eluate is not the cause of soft tissue uptake.

• Proceedings of the KWS Conference
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• 1997.10a
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• pp.95-99
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• 1997