• The Journal of the Convergence on Culture Technology
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• v.10 no.3
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• pp.819-825
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• 2024

The reason for recording dose data when using a diagnostic radiation source is to record and manage the dose to healthcare personnel and patients. The purpose of this study was to verify the difference in radiation dose when using diagnostic radiation generating devices and to inform users' awareness of dose reduction through measurement and analysis of dose in situations with and without shielding. The dose analysis of each equipment for two Korean C-arms and two German C-arms showed that the Korean FPD type C-arm had the highest dose value, followed by the German I.I type C-arm, German FPD type C-arm, Korean, and I.I type C-arm. The results of the dose analysis with and without shielding showed that the dose to the human phantom in a normal atmosphere increased by about 2 times due to scattered radiation, but the dose to the human phantom was reduced by about 5 times by wearing a shield (0.5mm/lead apron). More important than the management of radiation dose is the study of how to reduce exposure when using radiation, and since the radiation dose output from different equipment is different, it is necessary to provide dose information with and without shielding.

• Proceedings of the Korean Powder Metallurgy Institute Conference
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• 2002.07a
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• pp.25-37
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• 2002

The most important industrial application of gamma radiation in characterizing green compacts is the determination of the density. Examples are given where this method is applied in manufacturing technical components in powder metallurgy. The requirements imposed by modern quality management systems and operation by the workforce in industrial production are described. The accuracy of measurement achieved with this method is demonstrated and a comparison is given with other test methods to measure the density. The advantages and limitations of gamma ray densitometry are outlined. The gamma ray densitometer measures the attenuation of gamma radiation penetrating the test parts (Fig. 1). As the capability of compacts to absorb this type of radiation depends on their density, the attenuation of gamma radiation can serve as a measure of the density. The volume of the part being tested is defined by the size of the aperture screeniing out the radiation. It is a channel with the cross section of the aperture whose length is the height of the test part. The intensity of the radiation identified by the detector is the quantity used to determine the material density. Gamma ray densitometry can equally be performed on green compacts as well as on sintered components. Neither special preparation of test parts nor skilled personnel is required to perform the measurement; neither liquids nor other harmful substances are involved. When parts are exhibiting local density variations, which is normally the case in powder compaction, sectional densities can be determined in different parts of the sample without cutting it into pieces. The test is non-destructive, i.e. the parts can still be used after the measurement and do not have to be scrapped. The measurement is controlled by a special PC based software. All results are available for further processing by in-house quality documentation and supervision of measurements. Tool setting for multi-level components can be much improved by using this test method. When a densitometer is installed on the press shop floor, it can be operated by the tool setter himself. Then he can return to the press and immediately implement the corrections. Transfer of sample parts to the lab for density testing can be eliminated and results for the correction of tool settings are more readily available. This helps to reduce the time required for tool setting and clearly improves the productivity of powder presses. The range of materials where this method can be successfully applied covers almost the entire periodic system of the elements. It reaches from the light elements such as graphite via light metals (AI, Mg, Li, Ti) and their alloys, ceramics ($AI_20_3$, SiC, Si_3N_4, $Zr0_2$, ...), magnetic materials (hard and soft ferrites, AlNiCo, Nd-Fe-B, ...), metals including iron and alloy steels, Cu, Ni and Co based alloys to refractory and heavy metals (W, Mo, ...) as well as hardmetals. The gamma radiation required for the measurement is generated by radioactive sources which are produced by nuclear technology. These nuclear materials are safely encapsulated in stainless steel capsules so that no radioactive material can escape from the protective shielding container. The gamma ray densitometer is subject to the strict regulations for the use of radioactive materials. The radiation shield is so effective that there is no elevation of the natural radiation level outside the instrument. Personal dosimetry by the operating personnel is not required. Even in case of malfunction, loss of power and incorrect operation, the escape of gamma radiation from the instrument is positively prevented.

• Journal of Radiation Protection and Research
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• v.23 no.4
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• pp.229-236
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• 1998

In order to establish a testing system for personnel dosimetry performance, the radiation fields from photons, beta particles and neutrons are required, in recent, Korea Institute of Nuclear Safety(KINS) established the reference radation fields except neutrons and tested a variety of their properties. As a result of the test, the reference beams were shown to meet satisfactorily not only the standards of the International Organization for Standardization(ISO), but also the standard levels of the developed countries which are intercomparable with the international traceability. This paper describes the reference beam of gamma radiation. The self-designed and established reference radiation fields were investigated and analyzed by ISO and other international standards. The secondary photon contribution and the beam uniformity of the gamma radiation field were measured and evaluated to fulfill those requirements suggested by the ISO-4037. The measured air kerma rate for the $^{137}$Cs and $^{60}$Co gamma fields was 0.1891 $\sim$ 23.4967 $\mu$Gy/s sand 0.5844 $\sim$ 15.9954 $\mu$Gy/s respectively. The uncertainty with 95 % confidence level of the measured air kerma rate was determined to be less than 2.5 % which is comparable to the international reference gamma radiation fields. It was found that the evaluated air kerma calibration factors of Exradin ionization chamber were in good agreement within 0.9 % and 0.03 % with those given by PTB and NIST, respectively. The gamma radiation fields installed at KINS can maintain traceability systems in Korea, Germany and United State.

• Journal of Radiation Protection and Research
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• v.23 no.3
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• pp.123-138
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• 1998

Characteristics of element responses of Panasonic UD802 personnel dosimeters in the X, ${\beta}$, ${\gamma}$, ${\gamma}/X$, ${\gamma}/{\beta}$ and ${\gamma}$/neutron mixed fields were assessed. A dose-response algorithm has been developed to decide the high probability of a radiation type and energy by using the distribution in all six ratios of the multi-element TLD. To calculate the 4-element response factors and ratios between the elements of the Panasonic TLDs in the X, $\beta$, and $\gamma$ radiation fields, Panasonic’s UD802 TLDs were irradiated with KINS’s reference irradiation facility. In the photon radiation field, this study confirms that element-3 (E3) and element-4 (E4) of the Panasonic TLDs show energy dependent both in low- and intermediate-energy range, while element-1 (E1) and element-2 (E2) show little energy dependency in the entire whole range. The algorithm, which was developed in this study, was applied to the Panasonic personnel dosimetry system with UD716AGL reader and UD802 TLDs. Performance tests of the algorithm developed was conducted according to the standards and criteria recommended in the ANSI N13.11. The sum of biases and standard deviations was less than 0.232. The values of biases and standard deviations are distributed within a triangle of a lateral value of 0.3 in the ordinate and abscissa, With the above algorithm, Panasonic TLDs satisfactorily perform optimum dose assessment even under an abnormal response of the TLD elements to the energy imparted. This algorithm can be applied to a more rigorous dose assessment by distinguishing an unexpected dose from the planned dose for the most practical purposes, and is useful in conducting an effective personnel dose control program.

• Journal of Radiation Protection and Research
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• v.25 no.4
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• pp.191-195
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• 2000

This study was undertaken to estimate the exposed dose of the medical personnel during the intracoronary radiotherapy procedure as a part of ongoing SPARE (Seoul National University Hospital Post-Angioplasty Rhenium) trial. Data of thirty-four patients among forty-two irradiated patients participating in this trial due to coronary artery stenosis were retrospectively analyzed. Intracoronary radiotherapy was delivered to the patient immediately after angioplasty ballooning. Prescribed dose was 17 Gy to media of the diseased artery and was delivered with $^{188}Re$ filled balloon catheter. Dosimetry was carried out with GM counter at eight different points. Ten centimeter and forty centimeter from the patient's heart were selected to represent maximum and whole-body exposed dose of the operator, respectively. Median delivered dose was 111.6 mCi with average treatment time of 576 seconds. Average exposed dose rate at 10 cm and 40 cm from the patient's heart were 0.43 mSv/hr and 0.30 mSv/hr, respectively. Average exposed doses per treatment were 0.07 mSv and 0.05 mSv for 10 cm and 40 cm from the patient's heart, respectively. Exposed doses measured are much lower than recommended limit of 50 mSv for radiation workers or 1 mSv for general population in ICRP-60. This study proves that current method of intracoronary radiotherapy incorporated in this trial is very safe regarding radiation protection.