• Transactions of the Korean Society of Automotive Engineers
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• v.10 no.5
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• pp.15-22
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• 2002

Considering the high temperature durability, the most important issue is to accurately predict the maximum operating temperature of the shell, mat and substrate. This temperature prediction then defines the material selections far the mat, shell and cones, and allows an assessment to be made as to the necessity of heat shielding. In this papers, The commercial code FLUENT was utilized to simulate automotive oval type catalytic converters, with the objective of predicting thermal behavior under steady-state, high-load conditions. Specialized computational models are used to account for effects of heat and mass transfer in the monolith, conjugate heat transfer in the various converter materials, and radiation heat transfer.

• Nuclear Engineering and Technology
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• v.54 no.4
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• pp.1431-1438
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• 2022

In this study, natural radioactivity concentrations and dosimetric values of fly ash samples were evaluated for the landfill area of the coal-fired power plant (CFPP) complex at Binh Thuan, Vietnam. The average activity concentrations of ²³⁸U, ²²⁶Ra, ²³²Th and ⁴⁰K were 93, 77, 92 and 938 Bq kg^-1, respectively. The average results for radon dose, indoor external, internal, and total effective dose equivalent (TEDE) were 5.27, 1.22, 0.16, and 6.65 mSv y^-1, respectively. The average emanation fraction for fly ash were 0.028. The excess lifetime cancer risks (ELCR) were recorded as 20.30×10^-3, 4.26×10^-3, 0.62×10^-3, and 25.61×10^-3 for radon, indoor, outdoor exposures, and total ELCR, respectively. The results indicated that the cover of shielding materials above the landfill area significantly decreased the gamma radiation from the ash and slag in the ascending order: Zeolite < PVC < Soil < Concrete. Total dose of all radionuclides in the landfill site reached its peak at 19.8 years. The obtained data are useful for evaluation of radiation safety when fly ash is used for building material as well as the radiation risk and the overload of the landfill area from operation of these plants for population and workers.

• The Korean Journal of Nuclear Medicine Technology
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• v.20 no.2
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• pp.27-31
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• 2016

Purpose $^{99m}Tc-DMSA$ renal scan is a test for the comparison of the function by imaging the parenchyma of the kidneys by the cortex of a kidney and by computing the intake ratio of radiation by the left and right kidney. Since the distance between the kidneys and the bladder is not far given the bodily structure of an infant, the bladder is included in the examination domain. Research was carried out with the presumption that counts of bladder would impart an influence on the kidneys at the time of this renal scan. In consideration of the special feature that only a trace amount of a RI is injected in a pediatric examination, research on the method of injection was also carried out concurrently. Materials and Methods With 34 infants aged between 1 month to 12 months for whom a $^{99m}Tc-DMSA$ renal scan was implemented on the subjects, a Post IMAGE was acquired in accordance with the test time after having injected the same quantity of DMSA of 0.5mCi. Then, after having acquired an additional image by shielding the bladder by using a circular lead plate for comparison purposes, a comparison was made by illustrating the percentile of (Lt. Kidney counts + Rt. Kidney counts)/ Total counts, by drawing the same sized ROI (length of 55.2mm X width of 70.0mm). In addition, in the format of a 3-way stopcock, a Heparin cap and direct injection into the patient were performed in accordance with RI injection methods. The differences in the count changes in accordance with each of the methods were compared by injecting an additional 2cc of saline into the 3-way stopcock and Heparin cap. Results The image prior to shielding of the bladder displayed a kidney intake rate with a deviation of $70.9{\pm}3.18%$ while the image after the shielding of the bladder displayed a kidney intake rate with a deviation of $79.4{\pm}5.19%$, thereby showing approximately 6.5~8.5% of difference. In terms of the injection method, the method that used the 3-way form, a deviation of $68.9{\pm}2.80%$ prior to the shielding and a deviation of $78.1{\pm}5.14%$ after the shielding were displayed. In the method of using a Heparin cap, a deviation of $71.3{\pm}5.14%$ prior to the shielding and a deviation of $79.8{\pm}3.26%$ after the shielding were displayed. Lastly, in the method of direct injection into the patient, a deviation of $75.1{\pm}4.30%$ prior to the shielding and a deviation of $82.1{\pm}2.35%$ after the shielding were displayed, thereby illustrating differences in the kidney intake rates in the order of direct injection, a Heparin cap and the 3-way methods. Conclusion Since a substantially minute quantity of radiopharmaceuticals is injected for infants in comparison to adults, the cases of having shielded the bladder by removing radiation of the bladder displayed kidney intake rates that are improved from those of the cases of not having shielded the bladder. Although there are difficulties in securing blood vessels, it is deemed that the method of direct injection would be more helpful in acquisition of better images since it displays improved kidney intake rate in comparison to other methods.

• Nuclear Engineering and Technology
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• v.49 no.7
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• pp.1563-1570
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• 2017

In order to perform dry interim storage and transportation of the spent-fuel assemblies of the Bushehr Nuclear Power Plant, dual-purpose casks can be utilized. The effectiveness of different neutron-shield materials for the dual-purpose cask was analyzed through a set of calculations carried out using the Monte Carlo N-Particle (MCNP) code. The dose rate for the dual-purpose cask utilizing the recently developed materials of $epoxy/clay/B_4C$ and $epoxy/clay/B_4C/carbon$ fiber was less than the allowable radiation level of 2 mSv/h at any point and 0.1 mSv/h at 2 m from the external surface of the cask. By utilization of $epoxy/clay/B_4C$ instead of an ethylene glycol/water mixture, the dose rates on the side surface of the cask due to neutron sources and consequent secondary gamma rays will be reduced by 17.5% and 10%, respectively. The overall dose rate in this case will be reduced by 11%.

• Corrosion Science and Technology
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• v.20 no.4
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• pp.189-195
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• 2021

A technology for designing and licensing a dedicated radiation shielding facility needs to be developed for safe and efficient operation an R&D center. Technology development is important for smooth operation of such facilities. Causes of damage to internal structures (such as baffle former bolt (BFB) of pressurized water reactor) of a nuclear power reactor should be analyzed along with prevention and countermeasures for similar cases of other plants. It is important to develop technologies that can comprehensively analyze various characteristics of internal structures of long term operated reactors. In high-temperature, high-pressure operating environment of nuclear power plants, cases of BFB cracks caused by irradiated assisted stress corrosion cracks (IASCC) have been reported overseas. The integrity of a reactor's internal structure has emerged as an important issue. Identifying the cause of the defect is requested by the Korean regulatory agency. It is also important to secure a foundation for testing technology to demonstrate the operating environment for medium-level irradiated testing materials. The demonstration testing facility can be used for research on material utilization of the plant, which might have highest fluence on the internal structure of a reactor globally.

• 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.

• 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.

• 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.

• The Journal of Korean Institute of Electromagnetic Engineering and Science
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• v.20 no.9
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• pp.936-942
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• 2009

This paper presents the delivered power and reflection coefficient in metallic shielding enclosure with a sub-cavity, which are evaluated with the method of moments, sad describes a method for controlling the resonance characteristics of the metallic cavity by putting lossy dielectric material in the sub-cavity. In this paper we introduce carbon polystyrene-foam as lossy dielectric material and observe it's effects of reduction when the dimensions of the sub-cavity and permittivity of lossy dielectric material are changed. The results show that the reduction of the electromagnetic radiation can be achieved by controlling the amount of carbon in lossy dielectric material and the dimensions of the sub-cavity. The theoretical analysis is verified by the measured delivered power.

• Proceedings of the Korea Concrete Institute Conference
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• 2008.11a
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• pp.577-580
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• 2008

This paper was discussed about in-site application of heavyweight(or high density) concrete. Heavyweight concrete was placed with the method of conventional. Placement of conventionally mixed heavyweight concrete is subject to the same considerations of quality control as normal density concrete, except that it is far more susceptible to variations in quality due to improper handling. It is particularly subject to segregation during placement. Segregation of heavyweight concrete results not only in variation of strength but, far more importantly, in variation in density that are intolerable for work this type, since this adversely affects shielding properties. Heavyweight concrete materials and heavyweight concrete should be sampled and tested prior to and during construction to insure conformance with applicable standards and specifications.