• Journal of radiological science and technology
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• v.31 no.2
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• pp.177-182
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• 2008

Digital imaging for general rediography has many advantages over the film/screen systems, including a wider dynamic range and the ability to manipulate the images produced. The wider range means that acceptable images may by acquired at a range of dose levels, and therefore repeat exposures can be reduced. Digital imaging can result in the over use of radiation, however, because there is a tendency can be reduced. Digital imaging can result in the over use of radiation, however, because there is a tendency for images to be acquired at too high a dose. We investigated the actual exposure dose conditions on general radiography and a questionnaire survey was conducted with radiotechnologiest at medical institutions using digital radiology system. As a results, the dose of exposure was not controlled with patient's figure and dose optimization but was controlled by worker's convenience and image quality. Radio-technologiests often set up the exposure dose regardless of patient figure and body part to be examined. Many organizations, such as the International Commission on Radiological Protection, recommend to keep the dose as low as possible. In addition, they strongly recommend to keep the optimal but minimal dosage by proper training programs and constant quality control, including frequent patient dose evaluations and education.

• Fire Science and Engineering
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• v.31 no.2
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• pp.61-73
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• 2017

The Fukushima Nuclear Disaster in 2011 and California Power Failure in 2001 are examples of the importance of the power plant safety management that caused huge national loss with a power-related mass casualty incident. In a situation where humans cannot live without electricity, efforts to strengthen the systematic firefighting safety management in power plants that produce electricity with large amounts of hazardous materials as fuel, such as nuclear energy, coal and gas, are essential to protect life and prevent property loss and stable economic growth from fire explosion accident or radiation leak due to the negligence of safety management and natural disasters such as earthquakes, which has recently become an issue. This study examined the operating situation of firefighting safety management in power plants with firefighting officials employed by five power generation companies including Korea Southern Power Co., Ltd. and Korea Hydro & Nuclear Power Co. Ltd., which are in charge of the domestic power supply. As a result, for the systematic firefighting safety management of power plants, improvement plans were drawn, including the development of an effective business manual and a comprehensive management system, the substantiality of firefighting safety education, and the strengthening of seismic designs to prepare for earthquakes.

• Journal of the Korean Society of Radiology
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• v.13 no.2
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• pp.175-183
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• 2019

Cerebral nervous system intervention has been reported frequently due to radiation exposure such as blistering of the skin, hair loss, and erythema due to prolonged procedures. By applying ergonomically manufactured Bismuth (atomic number 83; Bi) shield to endovascular treatment of cerebral aneurysms, we aimed to minimize radiation exposure of scalp and lens from medical radiation exposure. The measurement site was the posterior part of the head, bilateral temporal part, bilateral quadriceps part, nose part, and the measuring part was attached to the optically stimulated Luminescence dosimeter (OSLD) Before and after the use, the entrance surface dose was compared and analyzed. The average entrance surface dose of group A (unshield) was 92.44 mGy, and group B was measured at 67.55 mGy. The average decrease in Group B was 26.92% compared to Group A. The entrance surface dose mean of the occipital region was measured at 146.08 mGy B group at 103.23 mGy and decreased by an average of 29.32% in group B compared to group A. The average entrance surface dose of the bilateral temporal part was measured in group A at 101.90 mGy group B at 72.69 mGy and decreased by an average of 28.67% in group B compared to group A. The average entrance surface dose for bilateral quadriceps part was measured at 27.51 mGy group B at 21.39 mGy and averaged 22.26% less in group B than group A. It is believed that the use of bismuth shields will be an alternative to reducing radiation disturbance due to temporary hair loss and other stochastic effects that may occur after the endovascular treatment of cerebral aneurysms procedure.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.1
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• pp.45-63
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• 2017

The oldest commercial reactor in South Korea, Kori-1 Nuclear Power Plant (NPP), will be shut down in 2017. Proper treatment for decommissioning wastes is one of the key factors to decommission a plant successfully. Particularly important is the recycling of clearance level or very low level radioactively contaminated metallic wastes, which contributes to waste minimization and the reduction of disposal volume. The aim of this study is to introduce a conceptual design of a recycle system and to evaluate the doses incurred through defined work flows. The various architecture diagrams were organized to define operational procedures and tasks. Potential exposure scenarios were selected in accordance with the recycle system, and the doses were evaluated with the RESRAD-RECYCLE computer code. By using this tool, the important scenarios and radionuclides as well as impacts of radionuclide characteristics and partitioning factors are analyzed. Moreover, dose analysis can be used to provide information on the necessary decontamination, radiation protection process, and allowable concentration limits for exposure scenarios.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.17 no.7
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• pp.616-622
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• 2016

The heating load using solar hot water is lower in summer than in the other seasons. This decreased heating load leads to the overheating solar collectors and related components. To prevent overheating of the solar collectors, air cooling and shading shields were used. On the other hand, it requires additional mechanical components, and reduces the system reliability. The geothermal heat dump system to release the high temperature heat (over $150^{\circ}C$) transferred from the heat pipe solar collectors was investigated in the present study. Research on the heat dump to cool the solar collector is rare. Therefore, the present study was carried out to collect possible data of a geothermal heat dump to cool the solar collector. A helical type geothermal heat exchanger was buried at a 1.2m depth. Experimentally and numerically, the geothermal heat dump was investigated in terms of the effects of parameters, such as the quantity of solar radiation, aperture area of the collector and the mass flow rate. A pipe length of 50m on the geothermal heat exchanger was suitable with a 0.33 kg/s flow rate. The water reservoir was a possible co-operation solution linked to the geothermal heat exchanger.

• Journal of the Institute of Electronics and Information Engineers
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• v.52 no.12
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• pp.99-108
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• 2015

The embedded microcontroller which is operated by the logic gates synchronized on the clock pulse, is gradually used as main controller of mission-critical systems. Severe electrical situations such as high voltage/frequency surge may cause malfunctioning of the clock source. The tolerant system operation is required against the various external electric noise and means the robust design technique is becoming more important issue in system clock failure problems. In this paper, we propose on-chip backup clock change architecture for the automatic clock failure detection. For the this, we adopt the edge detector, noise canceller logic and glitch-free clock changer circuit. The implemented edge detector unit detects the abnormal low-frequency of the clock source and the delay chain circuit of the clock pulse by the noise canceller can cancel out the glitch clock. The externally invalid clock source by detecting the emergency status will be switched to back-up clock source by glitch-free clock changer circuit. The proposed circuits are evaluated by Verilog simulation and the fabricated IC is validated by using test equipment electrical field radiation noise

• Journal of Radiation Protection and Research
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• v.40 no.1
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• pp.1-9
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• 2015

Compton cameras overcome several limitations of conventional mechanical collimation based gamma imaging devices, such as pin-hole imaging devices, due to its electronic collimation based on coincidence logic. Especially large-scale Compton camera has wide field of view and high imaging sensitivity. Those merits suggest that a large-scale Compton camera might be applicable to monitoring nuclear materials in large facilities without necessity of portability. To that end, our research group have made an effort to design a large-scale Compton camera for safeguard application. Energy resolution or position resolution of large-area detectors vary with configuration style of the detectors. Those performances directly affect the image quality of the large-scale Compton camera. In the present study, a series of Geant4 Monte Carlo simulations were performed in order to examine the effect of those detector parameters. Performance of the designed large-scale Compton camera was also estimated for various monitoring condition with realistic modeling. The conclusion of the present study indicates that the energy resolution of the component detector is the limiting factor of imaging resolution rather than the position resolution. Also, the designed large-scale Compton camera provides the 16.3 cm image resolution in full width at half maximum (angular resolution: $9.26^{\circ}$) for the depleted uranium source considered in this study located at the 1 m from the system when the component detectors have 10% energy resolution and 7 mm position resolution.

• The Korean Journal of Nuclear Medicine Technology
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• v.16 no.2
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• pp.126-130
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• 2012

Purpose : Methode an effective block was investigated to deal with volatile radioactive gas, short lived radioactive waste generated as a result of the routinely produced radiopharmaceuticals FDG (2-deoxy-2-[$^{18}F$]fluoro-D-glucose) and compound with $^{11}C$. Materials and Methods : All components of the radiation stack monitoring and data management system for continuous radioactive gas detection in the air extract system purchase from fixed noble gas monitor of Berthold company. TEDLAR gas sampling bags purchase from the Dongbanghitech company. TEDLAR gas sampling bags (volume: 10 L) connected via paraflex or PTFE tubing and Teflon 3 way stopcock. When installing TEDLAR gas sampling bags in Hot cell on the inside and not radioactive gas concentrations were compared. According to whether the Hot cell inside a activated carbon filter installed, compare the difference in concentration of the radioactive gas $^{18}F$. Comparison of radiation emission concentration difference of module a FASTlab and TRACElab. Results : Activated carbon filter are installed in the Hot cell, a measure of the concentration of radioactive gas was 8 $Bq/m^3$. Without activated carbone filter in the hot cell was 300 $Bq/m^3$. Tedlar bag prior to installation of the radioactive gases a measure of the concentration was 3,500 $Bq/m^3$, $^{11}C$ synthesis of the measured concentration was 27,000 $Bq/m^3$. After installed a Tedlar bag and a measure concentration of the radioactive gases was 300 $Bq/m^3$ and $^{11}C$ synthesis was 1,000$Bq/m^3$. Conclusion : $^{11}C$ radioactive gas that was ejected out of the Hot cell, with the use of a Tedlar gas sampling bag stored inside. A compound of 11C is not absorbed onto activated carbon filter. But can block the release out by storing in a Tedlar gas sampling bag. We was able to reduce the radiation exposure of the worker by efficient radiation protection.

• The Journal of Korean Society for Radiation Therapy
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• v.8 no.1
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• pp.19-27
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• 1996

I. Project Title A Study of Brachytherapy for intraocular tumor II. Objective and Importance of the project The eye enucleation or external-beam radiation therapy that has been commonly used for the treatment of intraocular tumor have demerits of visual loss and in deficiency of effective tumor dose. Recently, brachytherapy using the plaques containing radioisotope-now treatment method that decrease the demerits of the above mentioned treatment methods and increase the treatment effect-is introduced and performed in the countries, Our purpose of this research is to design suitable shape of plaque for the ophthalmic brachytherapy, and to measure absorbed doses of Ir-192 ophthalmic plaque and thereby calculate the exact radiation dose of tumor and it's adjacent normal tissue. III. Scope and Contents of the project In order to brachytherapy for intraocular tumor, 1. to determine the eye model and selected suitable radioisotope 2. to design the suitable shape of plaque 3. to measure transmission factor and dose distribution for custom made plaques 4. to compare with the these data and results of computer dose calculation models IV. Results and Proposal for Applications The result were as followed. 1. Eye model was determined as a 25mm diameter sphere, Ir-192 was considered the most appropriate as radioisotope for brachytherapy, because of the size, half, energy and availability. 2. Considering the biological response with human tissue and protection of exposed dose, we made the plaques with gold, of which size were 15mm, 17mm and 20mm in diameter, and 1.5mm in thickness. 3. Transmission factor of plaques are all 0.71 with TLD and film dosimetry at the surface of plaques and 0.45, 0.49 at 1.5mm distance of surface, respectively. 4. As compared the measured data for the plaque with Ir-192 seeds to results of computer dose calculation model by Gary Luxton et al. and CAP-PLAN (Radiation Treatment Planning System), absorbed doses are within ${\pm}10\%$ and distance deviations are within 0.4mm Maximum error is $-11.3\%$ and 0.8mm, respectively. As a result of it, we can treat the intraocular tumor more effectively by using custom made gold plaque and Ir-192 seeds.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.2
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• pp.113-119
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• 2011

Safe operation and maintenance of engineered dry storage systems for spent fuel from nuclear power plants basically depends on adequately adopted design requirements. The most important design target of the system are those which provide the necessary assurances that spent fuel can be received, handled, stored and retrieved without undue risk to health and safety of workers and the public. To achieve these objectives, the design of the system incorporates features to remove spent fuel residual heat, to provide for radiation protection, and to maintain containment over the lifespan of the system as specified in the design specifications. The features also provide for all possible anticipated operational occurrences and design basis events in accordance with the design basis as guided by the designated regulations. The general performance requirements of a projected storage system are introduced in this paper. The storage system is designed to store fuel assemblies in associated with designated regulatory requirements. Small increases/decreases in maximum burnup can be adjusted with cooling time. These variations are compensated for by a corresponding small site-specific increase/decrease in the design basis-cooling period, as long as the maximum heat load and radioactivity of loaded fuel assemblies are met. Generic design basis events considered for the storage system are summarized. Shielding and radiological requirements along with mechanical and structural are derived in this study.