• Journal of Radiation Protection and Research
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• v.38 no.1
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• pp.44-51
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• 2013

ICRP recommended that cosmic ray exposure to the pilot and cabin crew would be considered as an occupational exposure due to their relatively high exposure. Since 2012 with the Act No. 10908 (Natural radiation management), the guideline of cosmic ray exposure to the pilot was established in Korea. The applicability of the solid-state nuclear track detector for personal dose assessment of pilot and cabin crew was evaluated. Dose linearity and angle dependence of dosimeters to the neutron were evaluated by $^{252}Cf$ neutron emitting source. The track density has a good agreement with the dose ($r^2$=0.99) and highly dependent on the degree of an angular of the dosimeter to the neutron source. In addition, the dosimeters (SSNTD) were exposed to cosmic ray in an aircraft during its cruising for more than two months in collaboration with Airline Pilots Association of Korea. Although the correlation between the track density from aircraft cruising altitude and expected neutron dose is low, however RSNS dosimeter could be used for personal neutron dosimeter. For application of RSNS as a personal dosimeter for pilot and cabin crew, additional studies are required.

• Nuclear Engineering and Technology
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• v.4 no.4
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• pp.340-344
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• 1972

Isomeric ratios were measured for the capture of thermal neutron by $^{79}$ Br, $^{80}$ Se, $^{103}$ Rh, $^{115}$ In and $^{133}$ Cs as well as those of epi-cadmium neutron by $^{79}$ Br, $^{80}$ Se and $^{l33}$Cs. The measurements were performed by analysing decay curves obtained by ${\gamma}$-ray spectrometry after irradiation. The counting efficiency curve was determined by using the calibrated standard sources with overall uncertainties of about 1%. Isomeric ratios, given in $\sigma$ high spin/($\sigma$ high spin + $\sigma$ low spin), of $^{80, 80m}$Br, $^{81,81m}$Se, $^{014, 104m}$Rh, $^{116,116m}$In and $^{134, 134m}$Cs produced by thermal neutron activation were found to be 0.21$\pm$0.01, 0.14$\pm$0.02, 0.12$\pm$0.02, 0.69$\pm$0.07 and 0.058$\pm$0.004, respectively, Those values of $^{80, 80m}$Br, $^{81,81m}$Se, and $^{134, 134m}$Cs Produced by epi-cadmium neutron were found to be 0.19$\pm$0.02, 0.29$\pm$0.02 and 0.074$\pm$0.011, respectively. The experimental values obtained were compared with the theoretical values deduced from the statistical model. There were the general agreements between the theory and the experiment.t.

• Journal of Radiation Protection and Research
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• v.37 no.3
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• pp.149-158
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• 2012

In this paper, we measured the variations of radon concentrations in groundwater using low-level Liquid Scintillation Counter (LSC), an instrument for analyzing the alpha and beta radionuclides at its 10 sites around the Kumjung-Gu, north-western of Busan. Optimization of Pulse Shape Analyzer (PSA) to determinate the highest value of figure of merit (FM) was decided using Quantulus 1200 LSC with radium-226 source, the optimal PSA level was shown in the range of 100 to 110. The results show that the Minimum Detectable Activity (MDA) of radon concentrations is 0.61 $Bq{\cdot}L^{-1}$ for 20 minutes in PSA level. We find that the average radon concentration in groundwater is high in granitic rock area and low in volcanic rock area. (Biotite granite : 191.39 $Bq{\cdot}L^{-1}$, Micro graphic granite : 141.88 $Bq{\cdot}L^{-1}$, Adamellite : 92.94 $Bq{\cdot}L^{-1}$, Andesite (volcanic) : 35.35 $Bq{\cdot}L^{-1}$). No significant seasonal variation pattern is observed from the long-term variation analysis from 10 selected sites. We have not seen the significant correlation of radon concentration to groundwater temperature, atmospheric temperature, atmospheric pressure and rainfall. The concentration variation is probably caused by more complex factors and processes.

• Journal of radiological science and technology
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• v.40 no.4
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• pp.605-611
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• 2017

The structure of the actual detector was computed using the code of the PENELOPE. Using the standard mixed sources (450, 1,000 ml), compare the effectiveness of each energy according to various densities and height of the PENELOPE computer simulation, and calculate the effectiveness of the various environmental specimens and apply them to various environmental specimens to determine the lower limit. The values obtained by the obtained value were obtained by applying the obtained efficiency to the actual environmental specimens and obtaining the lower limit values. The density correction factor is 1.155 g of the density correction factor of $0.4g/cm^3$ (59.54keV), 1.153 (661 keV), $1.06g/cm^3$ 1.064 (1,836.04keV), 1.03, and 1.033. It was confirmed that the radioactivity concentration of environmental samples decreased as the amount of specimen was measured increases, and the MDA value decreased as time measured increases.

• Analytical Science and Technology
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• v.33 no.6
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• pp.274-284
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• 2020

This study described the analytical method for simultaneous determination of 89Sr and 90Sr in liquid sample using automated separation system. Radiostrontium in 0.5 kg of liquid sample was concentrated as SrCO3 to reduce the volume of sample, and purified from the sample using Sr-resin 2 mL (BV, Bed volume). The behavior of Sr and interferences such as Ba, Ca and Y were estimated with various flow rate ranging from 1 to 4 mL min-1. The detailed procedure for the purification of Sr on Sr-resin was presented. The purified radiostronitum was measured in Cerenkov mode and then measured in Scintillation mode by mixing scintillation cocktail. The measured value in both modes were used to calculate the activity of 89Sr and 90Sr. The performance tests were carried out the lab-control-sample having various activity ratio of between 89Sr and 90Sr. The recovery of Sr was ranged from 68 to 94 %. The relative bias of 89Sr activity was ranged from -5 to 20 %, and it was ranged from -10 to 10 % for 90Sr.

• Journal of IKEEE
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• v.26 no.3
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• pp.515-519
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• 2022

In this paper, we propose the development of a high-sensitivity entry-level nuclide analysis module. The proposed measurement sensor module consists of an electronic driving circuit for nuclide analysis resolution, prototype production with nuclide analysis function, and GUI development applied to prototypes. The electronic part driving circuit for nuclide analysis resolution is divided into nuclide analysis resolution process by the electronic part driving circuit block diagram, MCU circuit design used for radiation measurement, and PC program design for Spectrum acquisition. Prototyping with nuclide analysis function is made by adding a 128×128 pixel OLED display, three buttons for operation, a Li-ion battery, and a USB-C type port for charging the battery. The GUI development department applied to the prototype develops the screen composition such as the current time, elapsed measurement time, total count, and nuclide Spectrum. To evaluate the performance of the proposed measurement sensor module, an expert witness test was conducted. As a result of the test, it was confirmed that the calculated result by applying the resolution formula to the Spectrum (FWHM@662keV) obtained using the Cs-137 standard source in the nuclide analysis device had a resolution of 17.77%. Therefore, it was confirmed that the nuclide analysis resolution method proposed in this paper produces improved performance while being cheaper than the existing commercial nuclide analysis module.

• Journal of the Korean Society of Radiology
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• v.16 no.7
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• pp.953-960
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• 2022

The measurement of branching ratio of ¹⁶⁷Yb radioactive isotopes from gamma-ray spectrum of ¹⁶⁹Tm(p,3n)¹⁶⁷Yb reaction were performed by using a 100-MeV proton linear accelerator of the Korea Multi-purpose Accelerator Complex (KOMAC). The ¹⁶⁷Yb isotope has a half-life of 17.5 minutes and decays to ¹⁶⁹Tm. The gamma rays generated from the ¹⁶⁷Yb isotope were measured using an HPGe detector gamma ray spectroscopy system. The energy calibration of the detector and the efficiency measurement of the detector were determined using a standard source. The gamma rays of known main energy (62.9, 106.2, 113.3, 143.5 and 176.3 keV) were measured. On the other hand, information about the intensity of the generated gamma rays is very inaccurate. Therefore, in this study, the decay strength of the main gamma rays was accurately measured. Overall, it was different from the previously known results, and in particular, it was found that the intensity of the main decay gamma ray, such as the 113.3 and 106.2 keV gamma ray, was overestimated, and it was found that the gamma ray, such as 62.9, 116.7 and 143.5 keV was underestimated. The present results are considered to be important information in the fields of nuclear fusion, astrophysics and nuclear physics in the future.

• Progress in Medical Physics
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• v.23 no.2
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• pp.114-122
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• 2012

For the determination of absorbed dose to water from a linear accelerator photon beams, it needs a exposure calibration factor $N_x$ or air kerma calibration factor $N_k$ of air ionization chamber. We used the exposure calibration factor $N_x$ to find the absorbed dose calibration factors of water in a reference source through the TG-21 and TRS-277 protocol. TG-21 used for determine the absorbed dose in accuracy, but it required complex calculations including the chamber dependent factors. The authors obtained the absorbed dose calibration factor $N_{dw}{^{Co-60}}$ for reduce the complex calculations with unknown $N_{dw}$ only with $N_x$ or $N_k$ calibration factor in a TM31010 (S/N 1055, 1057) ionization chambers. The results showed the uncertainty of calculated $N_{dw}$ of IC-15 which was known the $N_x$ and $N_{dw}$ is within -0.6% in TG-21, but 1.0% in TRS-277. and TM31010 was compared the $N_{dw}$ of SSDL to that of PSDL as shown the 0.4%, -2.8% uncertainty, respectively. The authors experimented with good agreement the calculated $N_{dw}$ is reliable for cross check the discrepancy of the calibration factor with unknown that of TM31010 and IC-15 chamber.

• Journal of the Korean Society of Radiology
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• v.14 no.1
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• pp.23-29
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• 2020

The standard uptake values (SUVs) strongly depend on positron emission tomographs (PETs) and image reconstruction methods. Various image reconstruction algorithms in GE Discovery MIDR (DMIDR) and Discovery Ste (DSte) installed at Department of Nuclear Medicine, Seoul Samsung Medical Center were applied to measure the SUVs in an anthropomorphic torso phantom. The measured SUVs in the heart, liver, and background were compared to the actual SUVs. Applied image reconstruction algorithms were VPFX-S (TOF+PSF), QCFX-S-350 (Q.Clear+TOF+PSF), QCFX-S-50, VPHD-S (OSEM+PSF) for DMIDR, and VUE Point (OSEM) and FORE-FBP for DSte. To reduce the radiation exposure to radiation technologists, only the small amount of radiation source ¹⁸F-FDG was mixed with the distilled water: 2.28 MBq in the 52.5 ml heart, 20.3 MBq in the 1,290 ml liver and 45.7 MBq for the 9,590 ml in the background region. SUV values in the heart with the algorithms of VPFX-S, QCFX-S-350, QCFX-S-50, VPHD-S, VUE Point, and FOR-FBP were 27.1, 28.0, 27.1, 26.5, 8.0, and 7.4 with the expected SUV of 5.9, and in the background 4.2, 4.1, 4.2, 4.1, 1.1, and 1.2 with the expected SUV of 0.8, respectively. Although the SUVs in each region were different for the six reconstruction algorithms in two PET/CTs, the SUV ratios between heart and background were found to be relatively consistent; 6.5, 6.8, 6.5, 6.5, 7.3, and 6.2 for the six reconstruction algorithms with the expected ratio of 7.8, respectively. Mean SNRs (Signal to Noise Ratios) in the heart were 8.3, 12.8, 8.3, 8.4, 17.2, and 16.6, respectively. In conclusion, the performance of PETs may be checked by using with the SUV ratios between two regions and a relatively small amount of radioactivity.

• Progress in Medical Physics
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• v.27 no.1
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• pp.31-36
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• 2016

The paper discusses radiation dose of dual energy CT on which copper modulation layer, is mounted in order to improve diagnostic performance of the dual energy CT. The radiation dose is estimated using MCNPX and its results are compared with that of the conventional dual energy CT system. CT X-ray spectra of 80 and 120 kVp, which are usually used for thorax, abdominal, head, and neck CT scans, were generated by the SPEC78 code and were used for the source specification 'SDEF' card for MCNPX dose modeling. The copper modulation layer was located 20 cm away from a source covering half of the X-ray window. The radiation dose was measured as changing its thickness from 0.5 to 2.0 mm at intervals of 0.5 mm. Since the MCNPX tally provides only normalized values to a single particle, the dose conversion coefficients of F6 tally for the modulation layer-based dual energy CBCT should be calculated for matching the modeling results into the actual dose. The dose conversion coefficient is $7.2*10^4cGy/output$ that is obtained from dose calibration curve between F6 tally and experimental results in which GAFCHORMIC EBT3 films were exposed by an already known source. Consequently, the dose of the modulation layer-based dual energy cone beam CT is 33~40% less than that of the single energy CT system. On the basis of the results, it is considered that scattered dose produced by the copper modulation layer is very small. It shows that the modulation layer-based dual energy CBCT system can effectively reduce radiation dose, which is the major disadvantage of established dual energy CT.