• Journal of Radiation Protection and Research
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• v.12 no.2
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• pp.19-27
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• 1987

In order to determine the $^{56}Mn\;{\gamma}$-detection efficiency of a $MnSO_4$ bath system, it is essential to do the absolute activity measurement of $^{56}Mn$ solution. For the fabrication of $^{56}Mn$ samples, a 13.718 mg of $^{56}Mn$ metal flake with 99.99% purity was irradiated for 12 minutes at the thermal neutron field of about $10^{13}n/cm^2s$ of flux density. The neutron activated $^{56}Mn$ metal sample was dissolved in 50 ml of 0.1 N-HCl solution. The $^{56}Mn$ samples were fabricated by using the dissolved stock solution and the activity of each of them was measured by the $4{\pi}{\beta}-{\gamma}$ coincidence counting technique. The obtained result was 408.070 kBq/mg with total uncertainty of 0.366% at reference date, 0 h on October 15, 1987.

• Journal of Radiation Protection and Research
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• v.32 no.4
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• pp.178-183
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• 2007

Neutrons are high LET (linear energy transfer) radiation and cause more damage to the target cells than x-rays or gamma rays. The damage from neutrons is generally considered fatal to a cell and neutrons have a greater tendency to cause cell death through direct interaction on DNA. We performed experiments to measure growth delay ratio and oxygen enhancement ratio (OER) in mouse model system. We inoculated EMT-6 cells to the right hind leg of BALB-c mouse and X-rays and neutron beams were given when the average volume of tumors reached $200-300mm^3$. We irradiated 0, 11, 15.4 Gy of X-ray and 0, 5, 7 Gy of fast neutron beam at normoxic and hypoxic condition. The volume of tumors was measured 3 times per week. In x-ray experiment, growth delay ratio was 1.34 with 11 Gy and 1.33 with 15.4 Gy in normoxic condition compared to in hypoxic condition, respectively. In neutron experiment, growth delay ratio was 0.94 with 5 Gy and 0.98 with 7 Gy, respectively. The OER of neutron beam was 0.97. The neutron beam was more effective than X-ray in the control of hypoxic tumors.

• The Korean Journal of Nuclear Medicine Technology
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• v.16 no.1
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• pp.80-85
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• 2012

Purpose: The whole body bone scan is an examination that visualizing physiological change of bones and using bone-congenial radiopharmaceutical. The patients are intravenous injected radiopharmaceutical which labeled with radioactive isotope ($^{99m}Tc$) emitting 140 keV gammarays and scanned after injection. The 3 principles of radiation protection from external exposureare time, distance and shielding. On the 3 principles of radiation protection basis, radiopharmaceutical might just as well be injected rapidly for reducing radiation because it might be the unopened radiation source. However the radiopharmaceuticals are injected into patient directly and there is a limitation of distance control. This study confirmed the change of radiation exposure as change of distance from radiopharmaceutical and observed the change of radiation exposure afte rsetting a shelter for help to control radio-technician's exposure. Materials & methods: For calculate the average of injection time, the trained injector measured the injection time for 50 times and calculated the average (2 minutes). We made a source as filled the 99mTc-HDP 925 MBq 0.2 mL in a 1 mL syringe and measured the radiation exposure from 50 cm,100 cm,150 cm and 200 cm by using Geiger-Mueller counter (FH-40, Thermo Scientific, USA). Then we settled a lead shielding (lead equivalent 6 mm) from the source 25 cm distance and measured the radiation exposure from 50 cm distance. For verify the reproducibility, the measurement was done among 20 times. The correlation between before and after shielding was verified by using SPSS (ver. 18) as paired t-test. Results: The radiation doses according to distance during 2 minutes from the source without shielding were $1.986{\pm}0.052{\mu}$ Sv in 50 cm, $0.515{\pm}0.022{\mu}$ Sv in 100 cm, $0.251{\pm}0.012{\mu}$ Sv in 150 cm, $0.148{\pm}0.006{\mu}$ Sv in 200 cm. After setting the shielding, the radiation dose was $0.035{\pm}0.003{\mu}$ Sv. Therefore, there was a statistical significant difference between the radiation doses with shielding and without shielding ($p$<0.001). Conclusion: Because the great importance of whole body bone scan in the nuclear medicine, we should make an effort to reduce radiation exposure during radiopharmaceutical injections by referring the principles of radiation protection from external exposure. However there is a limitation of distance for direct injection and time for patients having attenuated tubules. We confirmed the reduction of radiation exposure by increasing distance. In case of setting shield from source 25 cm away, we confirmed reducing of radiation exposure. Therefore it would be better for reducing of radiation exposure to using shield during radiopharmaceutical injection.

• Journal of the Korean Society of Manufacturing Technology Engineers
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• v.6 no.4
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• pp.9-18
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• 1997

It is proposed a new approach to the development of the automatic vision system to examine and repair the steam generator tubes at remote distance. In nuclear power plants, workers are reluctant of works in steam generator because of the high radiation environment and limited working space. It is strongly recommended that the examination and maintenance works be done by an automatic system for the protection of the operator from to radiation exposure. Digital signal processors are used in implementing real time recognition and examination of steam generator tubes in the proposed vision system. Performance of proposed digital vision system is illustrated by simulation and experiment for similar steam generator model.

• Journal of Radiation Protection and Research
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• v.15 no.2
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• pp.1-6
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• 1990

An attempt was made to test the possibility of a major role for the $Na^{+}-K^{+}$ adenosine triphosphatase (ATPase)system in the diuresis induced by uranyl nitrate(UN). Fisher 344 rats were intravenously injected with UN(5 mg/kg, 15 mg/kg and 30 mg/kg). Urinary excretion of $Na^{+}\;and\;K^{+}$ significantly increased in 24 h exposure on the UN and then decreased below the normal level 3 days after the treatment. $Na^{+}-K^{+}$ ATPase activity of kidney was significantly inhibited in high dosages of UN 15mg/kg and UN 30 mg/kg 3-5 days after injection. And then the recovery of the enzyme activity was observed within 5-10 days after injection, at which the regeneration of the tubular cells occurred.

• 제어로봇시스템학회:학술대회논문집
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• 1996.10b
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• pp.101-104
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• 1996

In nuclear power plants, workers are reluctant of works in steam generator because of the high radiation environment and limited working space. It is strongly recommended that the examination and maintenance works be done by an automatic system for the protection of the operator from the radiation exposure. In this paper, it is proposed a new approach to the development of the automatic vision system to examine and repair the steam generator tubes at remote distance. Digital signal processors are used in implementing real time recognition and examination of steam generator holes in the proposed vision system. Performance of proposed digital vision system is illustrated by experiment for similar steam generator model.

• Proceedings of the KSME Conference
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• 2000.04a
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• pp.542-547
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• 2000

In this paper, we propose a new approach to the development of the automatic vision system to examine and repair the steam generator tubes at remote distance. In nuclear power plants, workers are reluctant of works in steam generator because of the high radiation environment and limited working space. It is strongly recommended that the examination and maintenance works be done by an automatic system for the protection of the operator from the radiation exposure. Digital signal processors are used in implementing real time recognition and examination of steam generator tubes in the proposed vision system. Performance of proposed digital vision system is illustrated by simulation and experiment for similar steam generator model.

• Journal of the Korean Society for Precision Engineering
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• v.15 no.6
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• pp.5-14
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• 1998

In this paper we propose a new approach to the development of the automatic vision system to examine and repair the steam generator tubes at remote distance. In nuclear power plants, workers are reluctant of works in steam generator because of the high radiation environment and limited working space. It is strongly recommended that the examination and maintenance works be done by an automatic system for the protection of the operator from the radiation exposure. Digital signal processors are used in implementing real time recognition and examination of steam generator tubes in the proposed vision system. Performance of proposed digital vision system is illustrated by simulation and experiment for similar steam generator model.

• Journal of Radiation Protection and Research
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• v.45 no.1
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• pp.35-44
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• 2020

Background: Cadmium zinc telluride (CZT) is a promising material because of a high detection efficiency, good energy resolution, and operability at room temperature. However, the cost of CZT dramatically increases as its size increases. In this study, to achieve a large effective volume with relatively low cost, an array structure comprised of individual virtual Frisch-grid CZT detectors was proposed. Materials and Methods: The prototype consisted of 2 × 2 CZTs, a holder, anode and cathode printed circuit boards (PCBs), and an application-specific integrated circuit (ASIC). CZTs were used and the non-contacting shielding electrode method was applied for virtual Frisch-grid effect. An ASIC was used, and the holder and the PCBs were fabricated. In the current system, because the CZTs formed a common cathode, a total of 5 channels were assigned for data processing. Results and Discussion: An experiment using ¹³⁷Cs at room temperature was conducted for 10 minutes. Energy and timing information was acquired and the depth of interaction was calculated by the timing difference between the signals of both electrodes. Based on obtained three-dimensional position information, the energy correction was carried out, and as a result the energy spectra showed the improvements. In addition, a Compton image was reconstructed using the iterative method. Conclusion: The virtual Frisch-grid CZT detector based on the array structure was developed and the energy spectra and the Compton image were successfully acquired.

• Journal of Radiation Protection and Research
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• v.43 no.4
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• pp.154-159
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• 2018

Background: To prevent small leakage accidents, a real-time and direct detection system for small leaks with a detection limit below that of existing systems, e.g. $0.5gpm{\cdot}hr^{-1}$, is required. In this study, a small-size beta detector, which can be installed inside the reactor containment (CT) building and detect small leaks directly, was suggested and its feasibility was evaluated using MCNPX simulation. Materials and Methods: A target nuclide was selected through analysis of radiation from radionuclides in the reactor coolant system (RCS) and the spectrum was obtained via a silicon detector simulated in MCNPX. A window was designed to reduce the background signal caused by other nuclides. The sensitivity of the detector was also estimated, and its shielding designed for installation inside the reactor CT. Results and Discussion: The beta and gamma spectrum of the silicon detector showed a negligible gamma signal but it also contained an undesired peak at 0.22 MeV due to other nuclides, not the $^{16}N$ target nuclide. Window to remove the peak was derived as 0.4 mm for beryllium. The sensitivity of silicon beta detector with a beryllium window of 1.7 mm thickness was derived as $5.172{\times}10^{-6}{\mu}Ci{\cdot}cc^{-1}$. In addition, the specification of the shielding was evaluated through simulations, and the results showed that the integrity of the silicon detector can be maintained with lead shielding of 3 cm (<15 kg). This is a very small amount compared to the specifications of the lead shielding (600 kg) required for installation of $^{16}N$ gamma detector in inside reactor CT, it was determined that beta detector would have a distinct advantage in terms of miniaturization. Conclusion: The feasibility of the beta detector was evaluated for installation inside the reactor CT to detect small leaks below $0.5gpm{\cdot}hr^{-1}$. In future, the design will be optimized on specific data.