• Journal of Radiation Protection and Research
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• v.47 no.4
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• pp.214-225
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• 2022

Background: The conventional cerium-doped Gd₂Al₂Ga₃O₁₂ (GAGG(Ce)) scintillator-based gamma-ray imager has a bulky detector, which can lead to incorrect positioning of the gammaray source if the shielding against background radiation is not appropriately designed. In addition, portability is important in complex environments such as inside nuclear power plants, yet existing gamma-ray imager based on a tungsten mask tends to be weighty and therefore difficult to handle. Motivated by the need to develop a system that is not sensitive to background radiation and is portable, we changed the material of the scintillator and the coded aperture. Materials and Methods: The existing GAGG(Ce) was replaced with Bi₄Ge₃O₁₂ (BGO), a scintillator with high gamma-ray detection efficiency but low energy resolution, and replaced the tungsten (W) used in the existing coded aperture with lead (Pb). Each BGO scintillator is pixelated with 144 elements (12 × 12), and each pixel has an area of 4 mm × 4 mm and the scintillator thickness ranges from 5 to 20 mm (5, 10, and 20 mm). A coded aperture consisting of Pb with a thickness of 20 mm was applied to the BGO scintillators of all thicknesses. Results and Discussion: Spectroscopic characterization, imaging performance, and image quality evaluation revealed the 10 mm-thick BGO scintillators enabled the portable gamma-ray imager to deliver optimal performance. Although its performance is slightly inferior to that of existing GAGG(Ce)-based gamma-ray imager, the results confirmed that the manufacturing cost and the system's overall weight can be reduced. Conclusion: Despite the spectral characteristics, imaging system performance, and image quality is slightly lower than that of GAGG(Ce), the results show that BGO scintillators are preferable for gamma-ray imaging systems in terms of cost and ease of deployment, and the proposed design is well worth applying to systems intended for use in areas that do not require high precision.

• Nuclear Engineering and Technology
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• v.56 no.4
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• pp.1204-1212
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• 2024

To accurately calculate the heating distribution of the fast reactor, a neutron-photon library in MATXS format named Knight-B7.1-1968n × 94γ was processed based on the ENDF/B-VII.1 library for ultrafine groups. The neutron cross-section processing code MGGC2.0 was used to generate few-group neutron cross sections in ISOTXS format. Additionally, the self-developed photon cross-section processing code NGAMMA was utilized to generate photon libraries for neutron-photon coupled heating calculations, including photo-atom cross sections for the ISOTXS format, prompt photon production cross sections, and kinetic energy release in materials (KERMA) factors for neutrons and photons, and the self-shielding effect from the capture and fission cross sections of neutron to photon have been taken into account when the photon source generated by neutron is calculated. The interface code GSORCAL was developed to generate the photon source distribution and interface with the DIF3D code to calculate the neutron-photon coupling heating distribution of the fast reactor core. The neutron-photon coupled heating calculation route was verified using the ZPPR-9 benchmark and the RBEC-M benchmark, and the results of the coupled heating calculations were analyzed in comparison with those obtained from the Monte Carlo code MCNP. The calculations show that the library was accurately processed, and the results of the fast reactor neutron-photon coupled heating calculations agree well with those obtained from MCNP.

• Journal of the Korean Society of Radiology
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• v.14 no.5
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• pp.503-510
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• 2020

Concrete is one of the most widely used materials as the shielding structures of a nuclear facilities. It is also the most generated radioactive waste in quantity while dismantling facilities. Since the concrete captures neutrons and generates various radionuclides, radiation measurement and analysis of the sample was fulfilled prior to dismantle facilities. An HPGe detector is used in general for the radiation measurement, and effective correction factors such as geometrical correction factor, self-absorption correction, and absolute detector efficiency have to be applied to the measured data to decide exact radioactivity of the sample. Correction factors are obtained by measuring data using a standard source with the same geometry and chemical states as the sample under the same measurement conditions. However, it is very difficult to prepare standard concrete sources because concrete is limited in pretreatment due to various constituent materials and high density. In addition, the concrete sample obtained by core drill is a volumetric source, which requires geometric correction for sample diameter and self absorption correction for sample density. Therefore in recent years, many researchers are working on the calculation of effective correction factors using Monte carlo simulation instead of measuring them using a standard source. In this study we calculated, using Geant4, one of the Monte carlo codes, the correction factors for the various diameter and density of the concrete core sample at the gamma ray energy emitted from the nuclides ¹⁵²Eu and ⁶⁰Co, which are the most generated in radioactive concrete.

• Radiation Oncology Journal
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• v.11 no.2
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• pp.449-454
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• 1993

The skin sparing effect associated with high energy x-ray or gamma ray beams may be reduce or lost under certain conditions of treatment. Current trends in using large fields. Shield carrying trays, compensating filters, and isocentric methods of treatment have posed problems of increased skin dose which sometimes become a limiting factor in giving adquate tumor doses. We used the shallow ion chamber to measure the phantom surface dose and the physical treatment variables for Co-60 gamma ray, 4MV and 10 MV x-ray beam. The dependence of percent surface dose on field sizes, atomic number of the shielding tray materials and its distance from the surface for 4, 10MV x-rays and Co-60 gamma ray is qualitatively similar. The use of 2 mm thick tin filter is recommended for situations where a low atomic number tray is introduced into the beam at distances less than 15 cm from the surface and with the large field sized for 4 MV x-ray beam. In case of Co-60 gamma ray, the lead glass tray is suitable for enhancement of skin sparing. Also, the filter distance should be as large as possible to achieve substantial skin sparing.

• Nuclear Engineering and Technology
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• v.52 no.10
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• pp.2204-2220
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• 2020

An open-pool type research reactor is designed and operated considering the accessibility around the pool top area to enhance the reactor utilization. The reactor structure assembly is placed at the bottom of the pool and filled with water as a primary coolant for the core cooling and radiation shielding. Most radioactive materials are generated from the fuel assemblies in the reactor core and circulated with the primary coolant. If the primary coolant goes up to the pool surface, the radiation level increases around the working area near the top of the pool. Hence, the hot water layer is designed and formed at the upper part of the pool to suppress the rising of the primary coolant to the pool surface. The temperature gradient is established from the hot water layer to the primary coolant. As this temperature gradient suppresses the circulation of the primary coolant at the upper region of the pool, the radioactive primary coolant rising up directly to the pool surface is minimized. Water mixing between these layers is reduced because the hot water layer is formed above the primary coolant with a higher temperature. The radiation level above the pool surface area is maintained as low as reasonably achievable since the radioactive materials in the primary coolant are trapped under the hot water layer. The key to maintaining the stable hot water layer and keeping the radiation level low on the pool surface is to have a stable flow of the primary coolant. In the research reactor with a downward core flow, the primary coolant is dumped into the reactor pool and goes to the reactor core through the flow guide structure. Flow fields of the primary coolant at the lower region of the reactor pool are largely affected by the dumped primary coolant. Simple, circular, and duct type discharge headers are designed to control the flow fields and make the primary coolant flow stable in the reactor pool. In this research, flow fields of the primary coolant and hot water layer are numerically simulated in the reactor pool. The heat transfer rate, temperature, and velocity fields are taken into consideration to determine the formation of the stable hot water layer and primary coolant flow. The bulk Richardson number is used to evaluate the stability of the flow field. A duct type discharge header is finally chosen to dump the primary coolant into the reactor pool. The bulk Richardson number should be higher than 2.7 and the temperature of the hot water layer should be 1 ℃ higher than the temperature of the primary coolant to maintain the stability of the stratified thermal layer.

• Journal of the Microelectronics and Packaging Society
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• v.28 no.1
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• pp.39-46
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• 2021

The quantitative measurement of interfacial adhesion energy (Gc) of multilayer thin films for Cu interconnects was investigated using a double cantilever beam (DCB) and 4-point bending (4-PB) test. In the case of a sample with Ta diffusion barrier applied, all Gc values measured by the DCB and 4-PB tests were higher than 5 J/㎡, which is the minimum criterion for Cu/low-k integration without delamination. However, in the case of the Ta/Cu sample, measured Gc value of the DCB test was lower than 5 J/㎡. All Gc values measured by the 4-PB test were higher than those of the DCB test. Measured Gc values increase with increasing phase angle, that is, 4-PB test higher than DCB test due to increasing plastic energy dissipation and roughness-related shielding effects, which matches well interfacial fracture mechanics theory. As a result of the 4-PB test, Ta/Cu and Cu/Ta interfaces measured Gc values were higher than 5 J/㎡, suggesting that Ta is considered to be applicable as a diffusion barrier and a capping layer for Cu interconnects. The 4-PB test method is recommended for quantitative adhesion energy measurement of the Cu interconnect interface because the thermal stress due to the difference in coefficient of thermal expansion and the delamination due to chemical mechanical polishing have a large effect of the mixing mode including shear stress.

• The Journal of Korean Society for Radiation Therapy
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• v.26 no.2
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• pp.329-336
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• 2014

Purpose : To evaluate the dose on a contralateral breast and the usefulness of shielding according to the distance between the contralateral breast and the side of the beam by breast size when patients who got breast implant receive radiation therapy. Materials and Methods : We equipped 200 cc, 300 cc, 400 cc, and 500 cc breast model on the human phantom (Rando-phantom), acquired CT images (philips 16channel, Netherlands) and established the radiation treatment plan, 180 cGy per day on the left breast (EclipseTM ver10.0.42, Varian Medical Systems, USA) by size. We set up each points, A, B, C, and D on the right(contralateral) breast model for measurement by size and by the distance from the beam and attached MOSFET at each points. The 6 MV, 10 MV and 15 MV X-ray were irradiated to the left(target) breast model and we measured exposure dose of contralateral breast model using MOSFET. Also, at the same condition, we acquired the dose value after shielding using only Pb 2 mm and bolus 3 mm under the Pb 2 mm together. Results : As the breast model is bigger from 200 cc to 500 cc, The surface of the contralateral breast is closer to the beam. As a result, from 200 cc to 500 cc, on 180 cGy basis, the measurement value of the scattered ray inclined by 3.22 ~ 4.17% at A point, 4.06 ~ 6.22% at B point, 0.4~0.5% at C point, and was under 0.4% at D point. As the X-ray energy is higher, from 6 MV to 15 MV, on 180 cGy basis, the measurement value of the scattered ray inclined by 4.06~5% at A point, 2.85~4.94% at B point, 0.74~1.65% at C point, and was under 0.4% at D point. As using Pb 2 mm for shield, scattered ray declined by average 9.74% at A and B point, 2.8% at C point, and is under 1% at D point. As using Pb 2 mm and bolus together for shield, scattered ray declined by average 9.76% at A and B point, 2.2% at C point, and is under 1% at D point. Conclusion : Commonly, in case of patients who got breast implant, there is a distance difference by breast size between the contralateral breast and the side of beam. As the distance is closer to the beam, the scattered ray inclined. At the same size of the breast, as the X-ray energy is higher, the exposure dose by scattered ray tends to incline. As a result, as low as possible energy wihtin the plan dose is good for reducing the exposure dose.

• The Korean Journal of Nuclear Medicine Technology
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• v.14 no.2
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• pp.159-165
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• 2010

Purpose: Positron emission tomography scan has been growing diagnostic equipment in the development of medical imaging system. Compare to $^{99m}Tc$ emitting 140 keV, Positron emission radionuclide emits 511 keV gamma rays. Because of this high energy, it needs to reduce radioactive emitting from patients for radiotechnologist. We searched the external dose rates by changing distance from patients and measure the external dose rates when we used shielder investigate change external dose rates. In this study, the external dose distribution were analyzed in order to help managing radiation protection of radiotechnologists. Materials and Methods: Ten patients were searched (mean age: $47.7{\pm}6.6$, mean height: $165.5{\pm}3.8$ cm and mean weight: $65.9{\pm}1.4$ kg). Radiation were measured on the location of head, chest, abdomen, knees and toes at the distance of 10, 50, 100, 150 and 200 cm. Then, all the procedure was given with a portable radiation shielding on the location of head, chest and abdomen at the distance of 100, 150 and 200 cm and transmittance was calculated. Results: In 10 cm, head (105.40 ${\mu}Sv/h$) was the highest and foot (15.85 ${\mu}Sv/h$) was the lowest. In 200 cm, head, chest and abdomen showed similar. On head, the measured dose rates were 9.56 ${\mu}Sv/h$, 5.23 ${\mu}Sv/h$, and 3.40 ${\mu}Sv/h$ in 100, 150 and 200 cm respectively. When using shielder, it shows 2.24 ${\mu}Sv/h$, 1.67 ${\mu}Sv/h$, and 1.27 ${\mu}Sv/h$ in 100, 150 and 200 cm on head. On chest, the measured dose rates were 8.54 ${\mu}Sv/h$, 4.90 ${\mu}Sv/h$, 3.44 ${\mu}Sv/h$ in 100, 150 and 200 cm, respectively. When using shielder, it shows 2.27 ${\mu}Sv/h$, 1.34 ${\mu}Sv/h$, and 1.13 ${\mu}Sv/h$ in 100, 150 and 200 cm on chest. On abdomen, the measured dose rates were 9.83 ${\mu}Sv/h$, 5.15 ${\mu}Sv/h$ and 3.18 ${\mu}Sv/h$ in 100, 150 and 200cm respectively. When using shielder, it shows 2.60 ${\mu}Sv/h$, 1.75 ${\mu}Sv/h$ and 1.23 ${\mu}Sv/h$ in 100, 150 and 200 cm on abdomen. Transmittance was increased as the distance was expanded. Conclusion: As the distance was further, the radiation dose were reduced. When using shielder, the dose were reduced as one-forth of without shielder. The Radio technologists are exposed of radioactivity and there were limitations on reducing the distance with Therefore, the proper shielding will be able to decrease radiation dose to the radiotechnologists.

• The Journal of Korean Society for Radiation Therapy
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• v.26 no.2
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• pp.265-272
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• 2014

Purpose : To reduce the radiation dose to the thyroid that is affected to scattered radiation, the shield was used. And we evaluated the shielding effect for the thyroid during whole brain radiation therapy. Materials and Methods : To measure the dose of the thyroid, 300cGy were delivered to the phantom using a linear accelerator(Clinac iX VARIAN, USA.)in the way of the 6MV X-ray in bilateral. To measure the entrance surface dose of the thyroid, five glass dosimeters were placed in the 10th slice's surface of the phantom with a 1.5 cm interval. The average values were calculated by measured values in five times each, using bismuth shield, 0.5 mmPb shield, self-made 1.0 mmPb shield and unshield. In the same location, to measure the depth dose of the thyroid, five glass dosimeters were placed in the 10th slice by 2.5 cm depth of the phantom with a 1.5 cm interval. The average values were calculated by measured values in five times each, using bismuth shield, 0.5 mmPb shield, self-made 1.0 mmPb shield and unshield. Results : Entrance surface dose of the thyroid were respectively 44.89 mGy at the unshield, 36.03 mGy at the bismuth shield, 31.03 mGy at the 0.5 mmPb shield and 23.21 mGy at a self-made 1.0 mmPb shield. In addition, the depth dose of the thyroid were respectively 36.10 mGy at the unshield, 34.52 mGy at the bismuth shield, 32.28 mGy at the 0.5 mmPb shield and 25.50 mGy at a self-made 1.0 mmPb shield. Conclusion : The thyroid was affected by the secondary scattering dose and leakage dose outside of the radiation field during whole brain radiation therapy. When using a shield in the thyroid, the depth dose of thyroid showed 11~30% reduction effect and the surface dose of thyroid showed 20~48% reduction effect. Therefore, by using the thyroid shield, it is considered to effectively protect the thyroid and can perform the treatment.

• Radiation Oncology Journal
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• v.24 no.4
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• pp.280-284
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• 2006

$\underline{Purpose}$: To evaluate biological characteristics of neutron beam generated by MC50 cyclotron located in the Korea Institute of Radiological and Medical Sciences (KIRAMS). $\underline{Materials\;and\;Methods}$: The neutron beams generated with 15 mm Beryllium target hit by 35 MeV proton beam was used and dosimetry data was measured before in-vitro study. We irradiated 0, 1, 2, 3, 4 and 5 Gy of neutron beam to EMT-6 cell line and surviving fraction (SF) was measured. The SF curve was also examined at the same dose when applying lead shielding to avoid gamma ray component. In the X-ray experiment, SF curve was obtained after irradiation of 0, 2, 5, 10, and 15 Gy. $\underline{Results}$: The neutron beams have 84% of neutron and 16% of gamma component at the depth of 2 cm with the field size of $26{\times}26\;cm^2$, beam current $20\;{\mu}A$, and dose rate of 9.25 cGy/min. The SF curve from X-ray, when fitted to linear-quadratic (LQ) model, had 0.611 as ${\alpha}/{\beta}$ ratio (${\alpha}=0.0204,\;{\beta}=0.0334,\;R^2=0.999$, respectively). The SF curve from neutron beam had shoulders at low dose area and fitted well to LQ model with the value of $R^2$ exceeding 0.99 in all experiments. The mean value of alpha and beta were -0.315 (range, $-0.254{\sim}-0.360$) and 0.247 ($0.220{\sim}0.262$), respectively. The addition of lead shielding resulted in no straightening of SF curve and shoulders in low dose area still existed. The RBE of neutron beam was in range of $2.07{\sim}2.19$ with SF=0.1 and $2.21{\sim}2.35$ with SF=0.01, respectively. $\underline{Conclusion}$: The neutron beam from MC50 cyclotron has significant amount of gamma component and this may have contributed to form the shoulder of survival curve. The RBE of neutron beam generated by MC50 was about 2.2.