• Radiation Oncology Journal
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• v.7 no.2
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• pp.305-311
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• 1989

Fletcher-Suit colpostat has an internal structure to reduce dose to bladder and rectum. Some programs were developed to calculate dose at any point in water in three dimension around the colpostat containing Cs-137 tube, to find the shielding effect to dose by the internal structure, and to draw isodose curves and iso-shielding effect curves. Computer was an IBM compatible AT with EGA card and language was MS-Basic V6.0, Material, shape and geometry of the strucure, tube and colpostat were considered in algorithm for calculation of dose. Dose rates per unit mg. Ra. eq. in water calculated by a program were stored in auxiliary memory devices and retrieved in another programs. Isodose curves on medial side shrinked. Dose distribution was not symmetric about a transverse axis bisecting the colpostat. Reduction of dose was more excessive on top side than on bottom. Iso-shielding effect curve showed that the shielding effect was higher on top side than on bottom, and that there was shielding effect over almost all area of medial side. Such results were related to both shifted position of tube in the colpostat and asymmetric distribution of active source in the tube. Maximum of shielding effect was $49\%$ on top side and $44\%$ on bottom side. The direction of iso-shielding effect curve was generally radial from the center of active source. In treatment planning using Fletcher-Suit colpostat, the internal structure should be considered to find precise doses to bladder and rectum, etc.

• Progress in Medical Physics
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• v.13 no.1
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• pp.44-50
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• 2002

CsI(T1) single crystal was grown in a Bridgman growing apparatus, which has the diameter of 11 mm and the mole ratio of 0.001 mol%. Radiation sensors were made with CsITl)crystal and two photodiodes, and measured spectroscopic characteristics and linearity for gamma-ray and X-ray. The energy resolution of CsI(T1) radiation sensor has been measured with $^{22}$ Na, $^{137}$ Cs and $^{60}$ Co gamma standard sources. Also output linearity of CsITl) sensor was measured for diagnostic radiation region. The energy resolutions of CsI(T1) radiation sensor for 0.511MeV gamma-ray from Na-22 source, 0.662MeV from Cs-137 source, and 1.17MeV and 1.332MeV from Co-60 source were 13.2%, 8.3%, 6.7%, and 5.1% respectively. Also the output linearity up to 80mAs current for 60kVp, 80kvp, 100kVp, 120kVp tube voltages has been studied.

• Progress in Medical Physics
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• v.17 no.2
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• pp.89-95
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• 2006

The purpose of this study was to evaluate the radiation doses during CT transmission scan by changing tube voltage and tube current, and to estimate the radiation dose during our clinical whole body $^{137}Cs$ transmission scan and high quality CT scan. Radiation doses were evaluated for Philips GEMINI 16 slices PET/CT system. Radiation dose was measured with standard CTDI head and body phantoms in a variety of CT tube voltage and tube current. A pencil ionization chamber with an active length of 100 mm and electrometer were used for radiation dose measurement. The measurement is carried out at the free-in-air, at the center, and at the periphery. The averaged absorbed dose was calculated by the weighted CTDI ($CTDI_w=1/3CTDI_{100,c}+2/3CTDI_{100,p}$) and then equivalent dose were calculated with $CTDI_w$. Specific organ dose was measured with our clinical whole body $^{137}Cs$ transmission scan and high quality CT scan using Alderson phantom and TLDs. The TLDs used for measurements were selected for an accuracy of ${\pm}5%$ and calibrated in 10 MeV X-ray radiation field. The organ or tissue was selected by the recommendations of ICRP 60. The radiation dose during CT scan is affected by the tube voltage and the tube current. The effective dose for $^{137}Cs$ transmission scan and high qualify CT scan are 0.14 mSv and 29.49 mSv, respectively. Radiation dose during transmission scan in the PET/CT system can measure using CTDI phantom with ionization chamber and anthropomorphic phantom with TLDs. further study need to be peformed to find optimal PET/CT acquisition protocols for reducing the patient exposure with same image qualify.

• The Korean Journal of Nuclear Medicine
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• v.33 no.1
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• pp.100-109
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• 1999

Purpose: Preamplifier and amplifier are very important parts for developing a portable counting or imaging gamma probe. They can be used for analyzing pulses containing energy and position information for the emitted radiations. The commercial Nuclear Instrument Modules (NIMs) can be used for processing these pulses. However, it may be improper to use NIMs in developing a portable gamma probe, because of its size and high price. The purpose of this study was to develop both preamplifier and amplifier and measure their performance characteristics. Materials and Methods: The preamplifier and amplifier were designed as a charge sensitive device and a capacitor resistor-resistor capacitor (CR-RC) electronic circuit, respectively, and they were mounted on a print circuit board (PCB). We acquired and analyzed energy spectra for Tc-99m and Cs-137 using both PCB and NIMs. Multichannel analyzer (Accuspec/A, Canberra Industries Inc., Meriden Connecticut, U.S.A) and scintillation detectors (EP-047(Bicron Saint-Gobain/Norton Industrial EP-047 (Ceramics Co., Ohio, U.S.A) with $2"{\times}2"$ NaI(T1) crystal and R1535 (Hamamatsu Photonics K.K., Electron Tube Center, Shizuoka-ken, Japan) with $1"{\times}1"$ NaI(T1) crystal were used for acquiring the energy spectra. Results: Using PCB, energy resolutions of EP-047 detectors for Tc-99m and Cs-137 were 12.92% and 5.01%, respectively, whereas R1535 showed 13.75% and 5.19% of energy resolution. Using the NIM devices, energy resolutions of EP-047 detector for Tc-99m and Cs-137 were measured as 14.6% and 7.58%, respectively. However, reliable energy spectrum of R1535 detector could not be acquired, since its photomultiplier tube (PMT) requires a specific type of preamplifier. Conclusion: We developed a special preamplifier and amplifier suitable for a small sized gamma probe that showed good energy resolutions independent of PMT types. The results indicate that the PCB can be used in developing both counting and imaging gamma probe.

• Journal of Sensor Science and Technology
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• v.21 no.4
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• pp.293-297
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• 2012

CsI:Gd crystal was grown by the Bridgeman method and its scintillation properties were investigated. The wavelength peak of the luminescence spectrum for the crystal excited by X-ray was 419 nm. The range of the spectrum was from 300 nm to 800 nm. The spectrum well matched to the quantum efficiency of a typical bi-alkali photo-multiplier tube(PMT). An energy resolution of 48.2 % was obtained for 662 keV ${\gamma}$-rays of $^{137}Cs$. The three decay times were obtained as a fast(557.4 ns, 42.2 %), intermediate (1.78 ${\mu}s$, 29.7 %) and slow (5.43 ${\mu}s$, 28.1 %) components, respectively.

• Radiation Oncology Journal
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• v.6 no.2
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• pp.263-268
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• 1988

This paper presents dose distributions in water around Fletcher-Suit colpostat containing $^{137}Cs$ tube, and shielding effect of Internal lead shield. Using ready packed film, author measured dose distribution in water around the colpostat containing cesium source. Nine sheets of films on one side of the colpostat are packed with acryl frames cut out so as to fill water, and irradiated in water by cesium source in the colpostat. Dose distributions on transverse plane and upper plane 0.5cm from upper surface of the colpostat were measured. Shielding effect was greater in upper medial direction than in lower medial direction. And that was the greatest around $30^{\circ}$ from the axis of the colpostat on upper side and around $50^{\circ}$on lower side. In the region 7cm from the center of the colpostat, shielding efficiency was 0.23 to 0.35 on the lower $50^{\circ}$ and 0.26 to 0.42 on the upper $30^{\circ}$, and decreased with increase of distance.

• Journal of Radiation Protection and Research
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• v.20 no.2
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• pp.85-95
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• 1995

Radiation measuring system using wireless communication method with single channel has been diveloped and tested. In this system, radiation signals from GM tube are transformed into digital pulses in pulse processing circuit and modulated in FSK (frequency shift keying) circuit for digital communication and then wirelessly transmitted to a receiving unit. The digital pulses received are then demodulated in FSK circuit and converted into radiation dose/dose rate in the data acquisition unit to display on the screen of a personal computer. The performance of this system was evaluated by using both a pulse generator and a standard radiation source(Cs-137). In both cases, digital pulses with 5V were observed in pulse processing circuit without distortion of their shape through wireless communication system. The experimental results of radiation measurement by this system after several test-irradiation of GM detector to a standard radiation source(Cs-137), showed good agreement with irradiation dose rate within 10% difference, and proved that this system could be effectively utillized as radiation measuring instrument. It is expected that this wireless radiation measuring system developed for the first time in Korea, can be used as a radiation monitor as well as a personal dosimeter if we can further improve this system to adopt wireless multichannel communication system.

• Nuclear Engineering and Technology
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• v.52 no.2
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• pp.456-460
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• 2020

In this research, a radiation detection backpack to be used discreetly or by a wide range of users was developed using array silicon-photomultiplier (SiPM) and CsI (Tl), and its characteristics were evaluated. The R-squared value, which indicates the responsiveness of a detector based on the signal intensity, was determined to be 0.981, indicating a good linear responsivity. The energy resolutions for gamma radiation energies of Co-57 (122 keV), Ba-133 (356 keV), Cs-137 (662 keV), and Co-60 (1332 keV) were found to be 13.40, 10.50, 6.77, and 3.16%, respectively. These results confirm good energy resolution characteristics. Furthermore, in the case of mixed sources, the gamma radiation peaks were readily distinguishable, and the R-squared value for energy linearity was calculated to be 0.999, demonstrating an exceptional energy linearity. Further research based on the results of this study would enable the commercialization of lightweight SiPM-based wireless radiation detection backpacks that can be used for longer durations by replacing the photomultiplier tube, which is mainly used as the optical sensor in existing radiation detection backpacks.

• Nuclear Engineering and Technology
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• v.52 no.5
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• pp.1029-1035
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• 2020

The detector suffers from pulse pileup by overlapping of the signals when it was used in high radiation fields. The pulse pileup deteriorates the energy spectrum and causes count losses due to random co-incidences, which might not resolve within the resolving time of the detection system. In this study, it is aimed to propose a new pulse pileup correction method. The proposed method is to correct the start point of the pileup pulse. The parameters are obtained from the fitted exponential curve using the peak point of the previous pulse and the start point of the pileup pulse. The amplitude at the corrected start point of the pileup pulse can be estimated by the peak time of the pileup pulse. The system is composed of a NaI (Tl) scintillation crystal, a photomultiplier tube, and an oscilloscope. A 61 μCi ¹³⁷Cs check-source was placed at a distance of 3 cm, 5 cm, and 10 cm, respectively. The gamma energy spectra for the radioisotope of ¹³⁷Cs were obtained to verify the proposed method. As a result, the correction of the pulse pileup through the proposed method shows a remarkable improvement of FWHM at 662 keV by 29, 39, and 7%, respectively.

• Proceedings of the KIEE Conference
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• 2002.07d
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• pp.2285-2288
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• 2002

Recently, electronic personal dosimeters based upon silicon PIN photodiode or miniature GM tube were developed and have attracted a lot of attention because of the advantages of their nature such as indication of dose rate and the cumulative dose, and facilitation of record keeping. In this paper, we have developed a high-sensitivity electronic personal dosimeter with silicon PIN photodiode. The electronic personal dosimeter is constructed with silicon PIN photodiode, preamplifier, and shaping amplifier. To show the effectiveness of electronic personal dosimeter, we conducted nuclear radiation experiments using $\gamma$-ray Ba-133, Cs-137, and Co-60. The electronic personal dosimeter have a good linearity on $\gamma$-ray energy and activity.