• Proceedings of the Korean Society of Medical Physics Conference
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• 2002.09a
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• pp.328-330
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• 2002

Large balloon angio catheter is used for Percutaneous Transluminal Angioplsty(TPA) of the iliac, femoral and renal arteries as well as after Transjugular Intrahepatic portosystemic shunt(TIPS). The use of angioplasty balloon filled with liquid form of radioisotope reduces the rate of restenosis after PTA. The purpose of this study was to evaluate the absorbed dose to the target vessels from various sized large balloon filled with liquid form of Ho-166-DTPA. Four balloons of balloon dilatation catheters evaluated were 5, 6, 8 and 10 mm in diameter. GafChromic film was used for the estimation of the absorbed dose near the surface of the balloon catheters. Absorbed dose rates are plotted in units of Gy/min/GBq/ml as a function of radial distance in mm from the surface of balloon. The absorbed dose rate was 1.1, 1.6, 2.2 and 2.3 Gy/min/GBq/ml at a balloon surface, 0.3, 0.4, 0.5 and 0.6 Gy/min/GBq/ml at 1 mm depth for various balloon diameter 5, 6, 8 and 10 mm in diameter respectively. The study was conducted to estimate the absorbed doses to the vessels from various sized large balloons filled with liquid form of Ho-166-DTPA for clinical trial of radiation therapy after the PTA. The absorbed dose distribution of Ho-166 appeared to be nearly ideal for vascular irradiation since beta range is very short avoiding unnecessary radiation to surrounding normal tissues.

• Radiation Oncology Journal
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• v.13 no.3
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• pp.285-289
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• 1995

Purpose : We compared the calcualted percent depth dose curves of 6 MeV electron beam to that of measured to evaluate the usefulness of Monte-carlo simulation method in radiation physics. Materials and Methods : The radiation dose values of 6 MeV electron beam using EGS4 code with one million histories in water were compared values that were measured from the depth dose curve of electron beam irradiated by medical accelerator ML6M. The central axis dose values were calculated according to the changing field size. such as $5{\times}5,\;10{\times}10,\;15{\times}15,\;20{\times}20cm^2$. Results : The value calculated showed a very similar shape to depth dose curve. The calculated and measured value of $D_max$ at $10{\times}10cm^2$ cone is 15mm and 14mm respectively. The calculated value of the surface radiation dose rate is $65.52\%$ and measured one is $76.94\%$. The surface radiation dose rate has varied from $64.43\%$ to $66.99\%$. The calculated values of $D_max$ are in the range between 15mm and 18mm. The calculated value was fitted well with measured value around the $D_max$ area, excluding build up range and below the $90\%$ depth dose area. Conclusion : This result suggested that the calculation of dose value can be replace the direct measurement of the dose for radiation therapy. Also, EGS4 may be a very convenient program to assess the effect of radiation dose using by personal computers.

• Journal of radiological science and technology
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• v.43 no.4
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• pp.259-264
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• 2020

Radiation causes radiation hazards in the human body. In Korea, a case of radiation necrosis occurred in 2014. In this study, the scatter and shielding efficiency according to lead shielding were classified into epidermis and dermis for 0.511 MeV used in nuclear medicine. In this study, experiments were conducted using the slab phantom that represents calibration and the dose of human trunk. Experimental results showed that the shielding rate of 0.25 mmPb was 180% in the epidermis and 96% in the dermis. Shielding at 0.5mmPb showed shielding rates of 158%in the epidermis and 82% in the dermis. As a result of measuring the absorbed dose by subdividing the thickness of the dermis into 0.5 mm intervals, when the shielding was carried out at 0.25 mmPb, the dose appeared to be about 120% at 0.5 mm of the dermis surface, and the dose was decreased at the subsequent depth. Shielding at 0.5 mmPb, the dose appeared to be about 101% at the surface 0.5 mm, and the dose was measured to decrease at the subsequent depth. This result suggests that when lead aprons are actually used, the scattering rays would be sufficiently removed due to the spaces generated by the clothes and air, Therefore, the scattered ray generated from lead will not reach the human body. The ICRU defines the epidermis (0.07), in which the radiation-induced damage of the skin occurs, as the dose equivalent. If the radiation dose of the dermis is considered in addition, it will be helpful for the evaluation of the prognosis for radiation hazard of the skin.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.19 no.1
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• pp.133-140
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• 2021

This study evaluates the radioactivity of concrete waste that occurs due to large amounts of decommissioned nuclear wastes and then determines the surface dose rate when the waste is packaged in a disposal container. The radiation assessment was conducted under the presumption that impurities included in the bio-shielded concrete contain the highest amount of radioactivity among all the concrete wastes. Neutron flux was applied using the simplified model approach in a sample containing the most Co and Eu impurities, and a maximum of 9.8×104 Bq·g-1 60Co and 2.63×105 Bq·g-1 152Eu was determined. Subsequently, the surface dose rate of the container was measured assuming that the bio-shield concrete waste would be packaged in a newly developed disposal container. Results showed that most of the concrete wastes with a depth of 20 cm or higher from the concrete surface was found to have less than 1.8 mSv·hr-1 in the surface dose of the new-type disposal container. Hence, when bio-shielded concrete wastes, having the highest radioactivity, is disposed in the new disposal container, it satisfies the limit of the surface dose rate (i.e., 2 mSv·hr-1) as per global standards.

• Journal of Radiation Protection and Research
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• v.36 no.4
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• pp.174-182
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• 2011

With the reactor operation conditions - 4.3 wt% $^{235}U$ initial enrichment, burn-up 55,000 MWd/MTU, average power 34 MW/MTU for three periods burned time for 539.2 days per period and cooling time for 100 hours after shut down, to set up the condition to determine the minimum height (depth) of spent fuel storage pool to shut off the radiation out of the spent fuel storage pool and to store spent fuels safely, the dose rate on the specific position directed to the surface of spent fuel storage pool induced by the neutron and gamma-ray from spent fuels are evaluated. The length of spent fuel is 381 cm, and as the result of evaluation on each position from the top of spent fuel to the surface of spent fuel storage pool, it is difficult for neutrons from spent fuels to pass through the water layer of maximum 219 cm (600 cm from the floor of spent fuel storage pool) and 419 cm (800 cm from the floor of spent fuel storage pool) for gamma-ray. Therefore, neutron and gamma-ray from spent fuels can pass through below 419 cm (800 cm from the floor) water layer directed to the surface of spent fuel storage pool.

• Journal of Environmental Science International
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• v.11 no.12
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• pp.1195-1204
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• 2002

We calculated dose rate using radiative transfer equations to consider radiative processes distinctly. The dose rate at Pohang(36°02'N, 129°23'E) was calculated using measured ozone and meteorological data and two-stream approximations(quadrature, Eddington, delta Eddington, PIFM(practical improved flux method), discrete ordinate, delta discrete ordinate) are used in solving equation. The purpose of this study is to determine the most compatible radiative transfer approximation for simulating the radiative and photochemical processes of atmosphere through comparision between calculated and measured values. Dose rate of the biologically effective irradiance in the region 0.28-0.32 U m showed the highest value when quadrature and Eddington was used and lower value on condition that delta scaling was applied. Correlation coefficient between dose rate at surface using radiation transfer equation and measured UV-B at Pohang was 0.78, 0.79 and 0.81 when delta Eddington, PIFM and delta discrete ordinate were used. Also, in case of above approximations were used, MBE(Mean Bias Error) was within -0.3MED/30min and RMBE(Relative Mean Bias Error) was below 10% between 1200 LST and 1400 LST Approximations which are compatible in estimating radiative process are delta Eddington, PIFM and delta discrete ordinate. Especially, in case that radiative process is considered more detail, delta discrete ordinate increased the number of stream is proper.

• Journal of Radiation Protection and Research
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• v.41 no.4
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• pp.378-383
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• 2016

Background: Radiation dose rates in PRIDE facility is evaluated quantitatively for assessing radiation safety of workers because of large amounts of depleted uranium being handled in PRIDE facility. Even if direct radiation from depleted uranium is very low and will not expose a worker to significant amounts of external radiation. Materials and Methods: ORIGEN-ARP code was used for calculating the neutron and gamma source term being generated from depleted uranium (DU), and the MCNP5 code was used for calculating the neutron and gamma fluxes and dose rates. Results and Discussion: The neutron and gamma fluxes and dose rates due to DU on spherical surface of 30 cm radius were calculated with the variation of DU mass and density. In this calculation, an imaginary case in which DU density is zero was added to check the self-shielding effect of DU. In this case, the DU sphere was modeled as a point. In case of DU mixed with molten salt of 50-250 g, the neutron and gamma fluxes were calculated respectively. It was found that the molten salt contents in DU had little effect on the neutron and the gamma fluxes. The neutron and the gamma fluxes, under the respective conditions of 1 and 5 kg mass of DU, and 5 and $19.1g{\cdot}cm^{-3}$ density of DU, were calculated with the molten salt (LiCl+KCl) of 50 g fixed, and compared with the source term. As the results, similar tendency was found in neutron and gamma fluxes with the variation of DU mass and density when compared with source spectra, except their magnitudes. Conclusion: In the case of the DU mass over 5 kg, the dose rate was shown to be higher than the environmental dose rate. From these results, it is concluded that if a worker would do an experiment with DU having over 5 kg of mass, the worker should be careful in order not to be exposed to the radiation.

• Nuclear Engineering and Technology
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• v.51 no.6
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• pp.1626-1632
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• 2019

Monitoring of environmental radioactivity is essential for ensuring the radiological safety of residents who live near nuclear power plants. Ulsan, South Korea, is surrounded by 16 nuclear power plants, the highest density in the country. In addition, the city contains facilities for conducting radiological nondestructive testing and using radioisotopes for medical purposes. It makes the confirmation of radiological safety particularly necessary. In this study, sampling points were selected based on regional characteristics, and surface water samples were pretreated and analyzed for gross beta and gamma radiation levels. In addition, the distribution of the city's gamma dose rate was determined using a mobile monitoring system and distribution visualization program. The results showed that there is no effect on the gross beta and gamma nuclides of artificial radionuclides, and the gamma dose rate of the entire region did not exceed the environmental radiation level in South Korea overall, confirming the radiological safety of the city.

• Nuclear Engineering and Technology
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• v.53 no.11
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• pp.3704-3710
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• 2021

The characteristics of the external radiation on the surface of the casks for spent nuclear fuel (SNF) storage by dry method are investigated for the case when the spatial distribution of SNF in the basket changes due to the destruction of the fuel rod claddings. The surface areas are determined, where the changes in fluxes of neutrons, produced by ²⁴⁴Cm actinide, and γ-quanta, produced by long-lived isotopes, are maximum in the result of the decrease in the height of the SNF area. Concrete (VSC-24) and metal (SC-21) casks are considered as examples. The procedure of periodic measurement of the dose rate of neutrons or γ-quanta at the specified points of the cask surface is proposed for identifying the fuel rod cladding destruction. Under normal operation, the decrease in the dose rate produced by neutrons as the function of SNF storage duration is determined by the half-life of ²⁴⁴Cm, and for γ-quanta - by the half-lives of long-lived SNF isotopes. Consequently, a stepwise change in the dose rate of neutrons or γ-quanta, detected by the measurements, as compared to the previous one, would indicate the destruction of the fuel rod claddings.

• The Journal of Korean Society for Radiation Therapy
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• v.8 no.1
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• pp.41-54
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• 1996

The secondary electrons developed by interaction between primary beam and a tray mounted for blocks in Megavoltage irradiation result in excess soft radiation dose to the surface layer. To reduce this electron contamination, electron filters have been used to be attached under a tray. Various filters with Cu and Al plates in six different thickness and Cu/Al combined plates in 3 different thickness were tested to measure the reduction rate of secondary electron contamination to the surface layer. The measurement to find optimal filter was performed on 6MV linear accelerator in $10 cm{\times}10 cm$ field size and fixed 78.5cm source to measurement points distance from surface to maximum build up point in 2mm intervals. The result was analyzed as the ratio of measured doses with using filters, to standard doses of measured open beam. The result of this study was fellowing : 1. The contaminated low energy radiation were mainly produced by blocking tray. 2. The surface absorbed dose was slowly increased by increasing irradiation field size but rapidly increased at field size above $15cm{\times}15cm$. 3. Al plate upto 2.5mm thickness used as a filter was found to be inadequate due to the failure of reduction of the surface absorbed dose below doses of the under surface upto the maximal build up. Cu 0.5mm plate and Cu 0.28mm/A1 1.5mm compound plate were found to be optimal filters. 4. By using these 2 filters, the absorbed dose to the surface were effectively reduced $5.5\%$ in field size $4cm{\times}4cm,\;11.3\%$ in field size $10cm{\times}10cm,\;22.3\%$ in field size $25cm{\times}25cm$. 5. In field size $10cm{\times}10cm$, the absorbed dose to the surface of irradiation was reduced by setting TSD 20cm at least,. but effective and enough dose reduction could be achieved by setting TSD 30cm as 2 optimal filters used. 6. More surface dose absorbed at TSD less than 7.4cm with a tray and filters together indicated that soft radiation was also developed by filters. 7. The variation of PDD by the different size of irradiation field was minimal as 2 optimal filters used. There was also not different in variation of PDD according to using any of two different filters. 8. PDD was not effected either by various TSD or by using the different filter among two.