• Publications of The Korean Astronomical Society
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• v.27 no.3
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• pp.81-86
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• 2012

The International Space Station (ISS) orbits the Earth within the inner radiation belt, where high-energy protons are produced by collisions of cosmic rays to the upper atmosphere. About 6 astronauts stay in the ISS for a long period, and it should be important to monitor and assess the radiation environment in the ISS. The tissue equivalent proportional counter (TEPC) is an instrument to measure the impact of radiation on the human tissue. KASI is developing a TEPC as a candidate payload of the ISS. Before the detailed design of the TEPC, we performed simulations to test whether our conceptual design of the TEPC will work propertly in the ISS and to predict its performance. The simulations estimated that the TEPC will measure the dose equivalent of about 1:1 mSv during a day in the ISS, which is consistent with previous measurements.

• Tribology and Lubricants
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• v.16 no.1
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• pp.15-21
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• 2000

Radiation of Nd-YAG laser changes and refines the surface microstructure of steels, which gives rise to enhancement of hardness and resulting wear resistance. In the present work, the effect of processing parameters during the surface modification with laser on the wear behavior of the SM45C steel was studied by means of wear testing. The counter material was alumina ceramics. The microstructure observation revealed the dependence of molten depth and width on the defocusing distance. The laser modification of steel surface give rise to improved wear resistance in the testing speed range of either <0.2 m/s or >0.9 m/s Material transfer from steel was wear observated the surface of counter material when testing speed was lower than 0.7 m/s.

• Proceedings of the Korean Society of Medical Physics Conference
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• 2002.09a
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• pp.225-227
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• 2002

In order to improve the accuracy of charged-particle radiation therapy, the beam energy was measured precisely using a TOF-system, and the range using a counter telescope system. A Si detector and a Ge detector were used to estimate the range straggling as a $\Delta$E and an E detector, respectively, because they have good energy resolution and the output pulse heights don't depend on the atomic number of detected particles. The results were compared with the theoretical values by a calculation code.

• Journal of the Korean Institute of Telematics and Electronics
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• v.8 no.1
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• pp.1-8
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• 1971

A survey meter which is used a G-M counter sensitive to beta and gamma radiation is studied. This device is completely transistorized, operated with battery, and can be read directly the 3 full-scale meter range: 2.5, 25 and 250 MR/HR respectively. The collector-coupled monostabel multivibrator consisting of a counting-rate meter circuit, and the astable blocking oscillator consisting of a dc-de converter for power supply are analyzed and derived the design dquations. To improve the resolving time of the G-M counter the device is designed to be triggered by low pulse in the order of 0.5v.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.31 no.5
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• pp.491-498
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• 2007

Flame characteristics of a methane-air premixed flame stabilized in a heat-regenerative small combustor were investigated experimentally. A small combustor having two counter-current shallow channels and a combustion space at one side was developed. In which the channel-gap was less the ordinary quenching distance of a stoichiometric methane-air premixed flame. Two design parameters of channel gap and thickness of the middle wall, which is located between two channels for unburned and burned gases, were varied. Flame stabilization conditions and characteristic flame behaviors were experimentally examined. Conclusively, Blowout conditions were governed mostly by the scale of the combustion space, and flashback conditions into the channel are dominated by the channel gap. Surface temperatures of the combustor were between 100 to 500$^{\circ}C$. Additionally, two distinctive flame stabilization modes of radiation and well-stirred?reaction were observed and their applicability was discussed.

• Journal of Radiation Protection and Research
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• v.34 no.1
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• pp.9-14
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• 2009

High density polyethylene (HDPE) is degraded due to a radiation-induced oxidation when it is used as a neutron moderator in a neutron counter for a nuclear material accounting of spent fuels. The HDPE exposed to the gamma-ray emitted from the fission products in a spent nuclear fuel results in a radiation-induced degradation which changes its original molecular structure to others. So a neutron moderating power variation of HDPE, irradiated by a gamma radiation, was investigated in this work. Five HDPE moderator structures were exposed to the gamma radiation emitted from a $^{60}Co$ source to a level of $10^5-10^9$ rad to compare their post-irradiation properties. As a result of the neutron measurement test with 5 irradiated HDPE structures and a neutron measuring system, it was confirmed that the neutron moderating power for the $10^5$ rad irradiated HDPE moderator revealed the largest decrease when the un-irradiated pure one was used as a reference. It implies that a neutron moderating power variation of HDPE is not directly proportional to the integrated gamma dose rate. To clarify the cause of these changes, some techniques such as a FTIR, an element analysis and a densitometry were employed. As a result of these analyses, it was confirmed that the molecular structure of the gamma irradiated HDPEs had partially changed to others, and the contents of hydrogen and oxygen had varied during the process of a radiation-induced degradation. The mechanism of these changes cannot be explained in detail at present, and thus need further study.

• The Journal of Korean Society for Radiation Therapy
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• v.25 no.1
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• pp.9-14
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• 2013

Purpose: High-energy radiotherapy with 10 MV or higher develops photoneutron through photonuclear reaction. Photoneutron has higher radiation weighting factor than X-ray, thus low dose can greatly affect the human body. An accurate dosimetric calculation and consultation are needed. This study compared and analyzed the dose change of photoneutron in terms of space according to the size of photon beam energy and treatment methods. Materials and Methods: To measure the dose change of photoneutron by the size of photon beam energy, patients with the same therapy area were recruited and conventional plans with 10 MV and 15 MV were each made. To measure the difference between the two treatment methods, 10 MV conventional plan and 10 MV IMRT plan was made. A detector was placed at the point which was 100 cm away from the photon beam isocenter, which was placed in the center of $^3He$ proportional counter, and the photoneutron dose was measured. $^3He$ proportional counter was placed 50 cm longitudinally superior to and inferior to the couch with the central point as the standard to measure the dose change by position changes. A commercial program was used for dose change analysis. Results: The average integral dose by energy size was $220.27{\mu}Sv$ and $526.61{\mu}Sv$ in 10 MV and 15 MV conventional RT, respectively. The average dose increased 2.39 times in 15 MV conventional RT. The average photoneutron integral dose in conventional RT and IMRT with the same energy was $220.27{\mu}Sv$ and $308.27{\mu}Sv$ each; the dose in IMRT increased 1.40 times. The average photoneutron integral dose by measurement location resulted significantly higher in point 2 than 3 in conventional RT, 7.1% higher in 10 MV, and 3.0% higher in 15 MV. Conclusion: When high energy radiotherapy, it should consider energy selection, treatment method and patient position to reduce unnecessary dose by photoneutron. Also, the dose data of photoneutron needs to be systematized to find methods to apply computerization programs. This is considered to decrease secondary cancer probabilities and side effects due to radiation therapy and to minimize unnecessary dose for the patients.

• Journal of IKEEE
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• v.21 no.4
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• pp.408-411
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• 2017

In this paper, we propose an integrated control system that measures neutrons, gamma ray, and x-ray. The proposed system is able to monitor and control the data measured and analyzed on the remote or network, and can monitor and control the status of each part of the system remotely without remote control. The proposed system consists of a gamma ray/x-ray sensor part, a neutron sensor part, a main control embedded system part, a dedicated display device and GUI part, and a remote UI part. The gamma ray/x-ray sensor part measures gamma ray and x-ray of low level by using NaI(Tl) scintillation detector. The neutron sensor part measures neutrons using Proportional Counter Detector(low-level neutron) and Ion Chamber Type Detector(high-level neutron). The main control embedded system part detects radiation, samples it in seconds, and converts it into radiation dose for accumulated pulse and current values. The dedicated display device and the GUI part output the radiation measurement result and the converted radiation amount and radiation amount measurement value and provide the user with the control condition setting and the calibration function for the detection part. The remote UI unit collects and stores the measured values and transmits them to the remote monitoring system. In order to evaluate the performance of the proposed system, the measurement uncertainty of the neutron detector was measured to less than ${\pm}8.2%$ and the gamma ray and x-ray detector had the uncertainty of less than 7.5%. It was confirmed that the normal operation was not less than ${\pm}15$ percent of the international standard.

• Journal of Radiation Protection and Research
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• v.32 no.4
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• pp.190-193
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• 2007

The objective of this study is to recommend the radiation protection design parameters from the shielding point of view for concrete wall between the decay tank room and the primary pump room in TRIGA Mark-II Research Reactor Facility. The shield design for this concrete wall has been performed with the help of Point-kernel Shielding Code Micro-Shield 5.05 and this design was also validated based on the measured dose rate values with Radiation Survey Meter (G-M Counter) considering the ICRP-60 (1990) recommendations for occupational dose rate limit ($10{\mu}Sv/hr$). The recommended shield design parameters are: (i) thickness of 114.3 cm Ilmenite-Magnetite Concrete (IMC) or 129.54 cm Ordinary Reinforced Concrete (ORC) for concrete wall A (ii) thickness of 66.04 cm Ilmenite-Magnetite Concrete (IMC) or 78.74 cm Ordinary Reinforced Concrete (ORC) for concrete wall B and (iii) door thickness of 3.175 cm Mild Steel (MS) on the entrance of decay tank room. In shielding efficiency analysis, the use of I-M concrete in the design of this concrete wall shows that it reduced the dose rate by a factor of at least 3.52 times approximately compared to ordinary reinforced concrete.

• Journal of Radiation Protection and Research
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• v.45 no.2
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• pp.76-80
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• 2020

Background: To improve the measurement accuracy of liquid-scintillation counting for activity standardization, it is necessary to significantly reduce the background caused by thermal noise or after-pulses. We have therefore improved a movable 3 photomultiplier (3PM)-γ coincidence-counting method using the logical sum of three double coincidences for β events. Materials and Methods: We designed a new data-acquisition system in which β events are obtained by counting the logical sum of three double coincidences. The change in β-detection efficiency can be derived by moving three photomultiplier tubes sequentially from the liquid-scintillation vial. The validity of the method was investigated by activity measurement of ¹³⁴Cs calibrated at the Korea Research Institute of Standards and Science (KRISS) with 4π(PC)β-γ(NaI(Tl)) coincidence counting using a proportional counter (PC) for the β detector. Results and Discussion: Measurements were taken over 14 counting intervals for each liquidscintillation sample by displacing three photomultiplier tubes up to 45 mm from the sample. The dead time in each β- and γ-counting channel was adjusted to be a non-extending type of 20 ㎲. The background ranged about 1.2-3.3 s^-1, such that the contributions of thermal noise or after-pulses were negligible. As the β-detection unit was moved away from the sample, the β-detection efficiencies varied between 0.54 and 0.81. The result obtained by the method at the reference date was 396.3 ± 1.7 kBq/g. This is consistent with the KRISS-certified value of 396.0 ± 2.0 kBq/g within the uncertainty range. Conclusion: The movable 3PM-γ method developed in the present work not only succeeded in reducing background counts to negligible levels but enabled β-detection efficiency to be varied by a geometrical method to apply the efficiency extrapolation method. Compared with our earlier work shown in the study of Hwang et al. [2], the measurement accuracy has much improved. Consequently, the method developed in this study is an improved method suitable for activity standardization of β-γ emitters.