• Journal of Radiation Protection and Research
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• v.31 no.4
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• pp.181-186
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• 2006

A radiophotoluminescent glass rod detector (GRD) system has recently become commercially available. We investigate the dosimetric properties of the GRD regarding the reproducibility of signal, dose linearity and energy dependence. The reproducibility of five measurements for 50 GRDs is presented by an average of one standard deviation of each GRD and it is ${\pm}1.2%$. It is found to be linear in response to doses of $^{60}Co$ beam in the range 0.5 to 50 Gy with a coefficient of linearity of 0.9998. The energy dependence of the GRD is determined by comparing the dose obtained using cylindrical chamber to that by using the GRD. The GRD response for each beam is normalized to the response for a $^{60}Co$ beam. The responses for 6 and 15 MV x-ray beams are within ${\pm}1.5%$ (1SD). The energy response of GRD for high-energy photon is almost the same as the energy dependence of LiF:Mg:Ti (TLD-100)and shows little energy dependence unlike p-type silicon diode detector. The GRDs have advantages over other detectors such diode detector, and TLD: linearity, reproducibility and energy dependency. It has been verified to be an effective device for small field dosimetry for stereotactic radiosurgery.

• Journal of Radiation Protection and Research
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• v.28 no.3
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• pp.225-231
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• 2003

Up-front irradiation technique as 3-dimensional conformation, or intensity modulation has kept large proportion of brain tumors from being complicated with acute radiation reactions in the normal tissue during or shortly after radiotherapy. For years, we've cannot help but counting on 2-D vertex beam technique to reduce acute reactions in the brain tumor patients because we're not equipped with 3-dimensional planning system. We analyzed its advantages and limitations in the clinical application. From 1998 to 2001, vertex or oblique vertex beams were applied to 35 patients with primary brain tumor and 25 among them were eligible for this analysis. Vertex(V) plans were optimized on the reconstructed coronal planes. As the control, we took the bilateral opposed techniques(BL) otherwise being applied. We compared the volumes included in 105% to 50% isodose lines of each plan. We also measured the radiation dose at various extracranial sites with TLD. With vertex techniques, we reduced the irradiated volumes of contralateral hemisphere and prevented middle ear effusion at contralateral side. But the low dose volume increased outside 100%; the ratio of V to BL in irradiated volume included in 100%, 80%, 50% was 0.55+/-0.10, 0.61+/-0.10, and 1.22+/-0.21, respectively. The hot area within 100% isodose line almost disappeared with vertex plan; the ratio of V to BL in irradiated volume included in 103%, 105%, 108% was 0.14+/-0.14, 0.05./-0.17, 0.00, respectively. The dose distribution within 100% isodose line became more homogeneous; the ratio of volume included in 103% and 105% to 100% was 0.62+/-0.14 and 0.26+/-0.16 in BL whereas was 0.16+/-0.16 and 0.02+/-0.04 in V. With the vertex techniques, extracranial dose increased up to $1{\sim}3%$ of maximum dose in the head and neck region except submandibular area where dose ranged 1 to 21%. From this data, vertex beam technique was quite effective in reduction of unnecessary irradiation to the contralateral hemispheres, integral dose, obtaining dose homogeneity in the clinical target. But it was associated with volume increment of low dose area in the brain and irradiation toward the head and neck region otherwise being not irradiated at all. Thus, this 2-D vertex technique can be a useful quasi-conformal method before getting 3-D apparatus.

• The Journal of Korean Society for Radiation Therapy
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• v.24 no.2
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• pp.107-114
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• 2012

Purpose: Unlike the existing linear accelerator with photon, proton therapy produces a number of second radiation due to the kinds of nuclide including neutron that is produced from the interaction with matter, and more attention must be paid on the exposure level of radiation workers for this reason. Therefore, thermoluminescence dosimeter (TLD) that is being widely used to measure radiation was utilized to analyze the exposure level of the radiation workers and propose a basic data about the radiation exposure level during the proton therapy. Materials and Methods: The subjects were radiation workers who worked at the proton therapy center of National Cancer Center and TLD Badge was used to compare the measured data of exposure level. In order to check the dispersion of exposure dose on body parts from the second radiation coming out surrounding the beam line of proton, TLD (width and length: 3 mm each) was attached to on the body spots (lateral canthi, neck, nipples, umbilicus, back, wrists) and retained them for 8 working hours, and the average data was obtained after measuring them for 80 hours. Moreover, in order to look into the dispersion of spatial exposure in the treatment room, TLD was attached on the snout, PPS (Patient Positioning System), Pendant, block closet, DIPS (Digital Image Positioning System), Console, doors and measured its exposure dose level during the working hours per day. Results: As a result of measuring exposure level of TLD Badge of radiation workers, quarterly average was 0.174 mSv, yearly average was 0.543 mSv, and after measuring the exposure level of body spots, it showed that the highest exposed body spot was neck and the lowest exposed body spot was back (the middle point of a line connecting both scapula superior angles). Investigation into the spatial exposure according to the workers' movement revealed that the exposure level was highest near the snout and as the distance becomes distant, it went lower. Conclusion: Even a small amount of exposure will eventually increase cumulative dose and exposure dose on a specific body part can bring health risks if one works in a same location for a long period. Therefore, radiation workers must thoroughly manage exposure dose and try their best to minimize it according to ALARA (As Low As Reasonably Achievable) as the International Commission on Radiological Protection (ICRP) recommends.

• Journal of Radiation Protection and Research
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• v.17 no.2
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• pp.61-70
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• 1992

Photon, neutron and beta radiation fields were measured at PWR plants which are the representative types of nuclear power plant operated in Korea. The photon energy spectra were measured at locations in the auxiliary building during operation period and in the containment vessel(C/V) during shutdown period using a portable gamma spectrometer with a HPGe detector. The distribution of average energy was found to range from 440 to 780 keV in the C/V and from 280 keV to 760 keV in the auxiliary building, respectively. The average neutron energy measured at the five locations around the operation deck in the C/V in operation using a BMSS (Bonner Multi-Sphere Spectrometer) ranged from 20 keV to 210 keV. A computer code, BUNKI was used to unfold the spectrum. The beta energy spectra in the C/V and in the auxiliary building in annual outage were determined using 14 smear samples taken from the highly contaminated areas. The analysis showed that the representative corrosion product, $^{60}Co$ made main contribution to the beta energy field.

• Journal of Radiation Protection and Research
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• v.45 no.2
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• pp.76-80
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• 2020

Background: To improve the measurement accuracy of liquid-scintillation counting for activity standardization, it is necessary to significantly reduce the background caused by thermal noise or after-pulses. We have therefore improved a movable 3 photomultiplier (3PM)-γ coincidence-counting method using the logical sum of three double coincidences for β events. Materials and Methods: We designed a new data-acquisition system in which β events are obtained by counting the logical sum of three double coincidences. The change in β-detection efficiency can be derived by moving three photomultiplier tubes sequentially from the liquid-scintillation vial. The validity of the method was investigated by activity measurement of ¹³⁴Cs calibrated at the Korea Research Institute of Standards and Science (KRISS) with 4π(PC)β-γ(NaI(Tl)) coincidence counting using a proportional counter (PC) for the β detector. Results and Discussion: Measurements were taken over 14 counting intervals for each liquidscintillation sample by displacing three photomultiplier tubes up to 45 mm from the sample. The dead time in each β- and γ-counting channel was adjusted to be a non-extending type of 20 ㎲. The background ranged about 1.2-3.3 s^-1, such that the contributions of thermal noise or after-pulses were negligible. As the β-detection unit was moved away from the sample, the β-detection efficiencies varied between 0.54 and 0.81. The result obtained by the method at the reference date was 396.3 ± 1.7 kBq/g. This is consistent with the KRISS-certified value of 396.0 ± 2.0 kBq/g within the uncertainty range. Conclusion: The movable 3PM-γ method developed in the present work not only succeeded in reducing background counts to negligible levels but enabled β-detection efficiency to be varied by a geometrical method to apply the efficiency extrapolation method. Compared with our earlier work shown in the study of Hwang et al. [2], the measurement accuracy has much improved. Consequently, the method developed in this study is an improved method suitable for activity standardization of β-γ emitters.

• Korean Journal of Medicinal Crop Science
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• v.19 no.4
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• pp.287-299
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• 2011

Pharmacological effects of Panax ginseng have been demonstrated in cardiovascular system, endocrine secretion and immune system, together with antitumor, anti-stress and anti-oxidant activities. Modern scientific data show protective effect of ginseng against bone marrow cell death, increased survival rate of experimental animals, recovery of hematopoietic injury, immunopotentiation, reduction of damaged intestinal epithelial cells, inhibition of mutagenesis and effective protection against testicular damages, caused by radiation exposure. And also, ginseng acts in indirect fashion to protect radical processes by inhibition of initiation of free radical processes and thus reduces the radiation damages. The research has made much progress, but still insufficient to fully uncover the action mechanism of ginseng components on the molecule level. This review provides the usefulness of natural product, showing no toxic effects, as an radioprotective agent. Furthermore, the further clinical trials on radioprotection of ginseng need to be highly done to clarify its scientific application. The effective components of ginseng has been known as ginsenosides. Considering that each of these ginsenosides has pharmacological effect, it seems likely that non-saponin components might have radioprotective effects superior to those of ginsenosides, suggesting its active ingredients to be non-saponin series. These results also show that the combined effects of saponin and non-saponin components play an important role in the radioprotective effects of ginseng.

• Progress in Medical Physics
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• v.33 no.4
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• pp.172-179
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• 2022

Recently, tetrahedral phantoms have been newly adopted as international standard mesh-type reference computational phantoms (MRCPs) by the International Commission on Radiological Protection, and a program has been developed to convert them to computational tomography images and DICOM-RT structure files for application of radiotherapy. Through this program, the use of the tetrahedral standard phantom has become available in clinical practice, but utilization has been difficult due to various library dependencies requiring a lot of time and effort for installation. To overcome this limitation, in this study a newly developed TET2DICOM-GUI, a TET2DICOM program based on a graphical user interface (GUI), was programmed using only the MATLAB language so that it can be used without additional library installation and configuration. The program runs in the same order as TET2DICOM and has been optimized to run on a personal computer in a GUI environment. A tetrahedron-based male international standard human phantom, MRCP-AM, was used to evaluate TET2DICOM-GUI. Conversion into a DICOM-RT dataset applicable in clinical practice in about one hour with a personal computer as a basis was confirmed. Also, the generated DICOM-RT dataset was confirmed to be effectively implemented in the radiotherapy planning system. The program developed in this study is expected to replace actual patient data in future studies.

• Journal of Radiation Protection and Research
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• v.48 no.3
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• pp.131-143
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• 2023

Background: This study examined the detection limit of thyroid screening monitoring conducted at the time of the Fukushima Daiichi Nuclear Power Plant (FDNPP) accident in 2011 using a Monte Carlo simulation. Materials and Methods: We calculated the detection limit of a NaI(Tl) survey meter to measure ¹³¹I accumulation in the thyroid gland of children. Mathematical phantoms of 1- and 5-year-old children were developed in the simulation of the Particle and Heavy Ion Transport code System code. Contamination of the body surface with eight radionuclides found after the FDNPP accident was assumed to have been deposited on the neck and shoulder area. Results and Discussion: The detection limit was calculated as a function of ambient dose rate. In the case of 40 Bq/cm² contamination on the body surface of the neck, the present simulations showed that residual thyroid radioactivity corresponding to thyroid dose of 100 mSv can be detected within 21 days after intake at the ambient dose rate of 0.2 µSv/hr and within 11 days in the case of 2.0 µSv/hr. When a time constant of 10 seconds was used at the dose rate of 0.2 µSv/hr, the estimated survey meter output error was 5%. Evaluation of the effect of individual differences in the location of the thyroid gland confirmed that the measured value would decrease by approximately 6% for a height difference of ±1 cm and increase by approximately 65% for a depth of 1 cm. Conclusion: In the event of a nuclear disaster, simple measurements carried out using a NaI(Tl) scintillation survey meter remain effective for assessing ¹³¹I intake. However, it should be noted that the presence of short-half-life radioactive materials on the body surface affects the detection limit.

• Journal of Radiation Protection and Research
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• v.37 no.2
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• pp.63-69
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• 2012

The second phase of the national program for fusion energy development in Korea starts from 2012 for design and construction of the fusion DEMO reactor. Radiological assessment for the fusion reactor is one of the key tasks to assure its licensability and the starting point of the assessment is determination of the source terms. As the first effort, the activities of the coolant due to activated corrosion product (ACP) were estimated. Data and experiences from fission reactors were used, in part, in the calculations of the ACP concentrations because of lack of operating experience for fusion reactors. The MCNPX code was used to determine neutron spectra and intensities at the coolant locations and the FISPACT code was used to estimate the ACP activities in the coolant of the fusion DEMO reactor. The calculated specific activities of the most nuclides in the fusion DEMO reactor coolant were 2-15 times lower than those in the PWR coolant, but the specific activities of $^{57}Co$ and $^{57}Ni$ were expected to be much higher than in the PWR coolant. The preliminary results of this study can be used to figure out the approximate radiological conditions and to establish a tentative set of radiological design criteria for the systems carrying coolant in the design phase of the fusion DEMO reactor.

• Journal of Radiation Protection and Research
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• v.22 no.3
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• pp.183-193
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• 1997

The amount and kinds of radionuclide contained in waste volume should be known to prepare for occupational exposure management, perform safety assessment and finally to license a repository. Although the volume of filters and resins are small, activities of them comprise most of the radioactivity that made during power generation. This study aims at developing a method of estimating the radionuclide accumulation at the filters and resins of coolant systems. In this study, accumulated amount of radionuclides is estimated by a computer program which makes use of instantaneous decontamination factor, DF, instead of average DF. A FORTRAN program was developed for the estimation. Data from in-plant source-term measurements at Rancho-Seco nuclear power plant in the United States are employed for verification of the estimating method. And experimental data are employed, too. The instantaneous-DF-method showed smaller error than the average-DF-method. Accumulated amount of radionuclides can be calculated with only the DF and the radionuclide concentration, which are measured periodically according to the operating guide. However, especially, when the operating condition of nuclear power plant changes rapidly, the measuring term of DF and radionuclide should be shortened to ensure the accurate estimation.