• Journal of the Korean GEO-environmental Society
• /
• v.12 no.9
• /
• pp.35-45
• /
• 2011

This paper is to analyze the behaviors of tunnel support by TDR(Time Domain Reflectometry) sensor using electrical pulse. To analysis the behaviors of tunnel support, Copper tape as sensing materials was studied for on-site installation. Copper tape to the top of the glass tape, foam tape, and shielding the lower part was used electromagnetic shield sheet. For a high sensitivity to load and fill out the measurement noise emissions has been developed for the production of materials. This sensing material through the tunnel model tests for the change by surcharge load in TDR data were analyzed. Varing stiffness and support of conditions were determined the change of TDR data through PVC pipe tunnel section model tests. By comparing TDR data and finite element analysis, the behaviors of the tunnel support materials were analyzed qualitatively.

• Journal of radiological science and technology
• /
• v.45 no.1
• /
• pp.57-67
• /
• 2022

CZT detectors, which are compound semiconductors that have been widely used recently for gamma-ray detection purposes, are difficult to detect neutrons because direct interaction with them does not occur unlike gamma-rays. In this paper, a method of detecting and determining energy levels (fast neutrons and thermal neutrons) of neutrons, in addition of identifying energy and nuclide of gamma-rays, and evaluating gamma dose rates using a CZT semiconductor detector is described. Neutrons may be detected by a secondary photoelectric effect or compton scattering process with a characteristic gamma-ray of 558.6 keV generated by a capture reaction (¹¹³Cd + ¹n → ¹¹⁴Cd + 𝛾) with cadmium (Cd) in the CZT detector. However, in the case of fast neutrons, the probability of capture reaction with cadmium (Cd) is very low, so it must be moderated to thermal neutrons using a moderator and the material and thickness of moderator should be determined in consideration of the portability and detection efficiency of the equipment. Conversely, in the case of thermal neutrons, the detection efficiency decreases due to shielding effect of moderator itself, so additional CZT detector that do not contain moderator must be configured. The CZT detector that does not contain moderator can be used to evaluate energy, nuclide, and gamma dose-rate for gamma-rays. The technology proposed in this paper provides a method for detecting both neutrons and gamma-rays using a CZT detector.

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

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

• Journal of radiological science and technology
• /
• v.38 no.3
• /
• pp.205-211
• /
• 2015

We evaluated the effectiveness of TL (Time Limit) method by comparing with NTL (Non-time limit) method when it is used for examinations for abdomen Anterior Posterior (AP) in this paper. The evaluation was conducted based on the comparison of dose, and of signal to noise ratio (SNR) and contrast to ratio (CNR) on both methods. The experiments were conducted with XGEO GC 80 (Samsung, Korea), Unfors ThinX RAD (Unfors, Sweden) and Rando Phantom (Alderson research laboratories, USA) and shielding material with the size of $5.5{\times}9{\times}0.1cm^3$. It was set to activate only two upper ionization chambers in automatic exposure control(AEC) mode and the tube-voltage was set to 80kVp. When the exposure time was limited, it is limited to 51 msec. The images both by NTL AEC method and TL AEC method were acquired when with and without attachment of shielding material on the upper ionization chambers. The images were evaluated by SNR and CNR which are the image evaluation methods using 'Image J'. The NTL AEC method showed increases in dose as much as 130.7% at maximum and 80% at minimum than other methods. The TL AEC method showed decreases in mAs and exposure dose than the NTL AEC method as much as 43.8% and 44.4% respectively. There were no significant differences in SNR or CNR for the experiments (($p{\geq}0.05$). Therefore, it is suggested that the TLAEC mode is more effective when examining patients who have high BMI index or a patient with a metallic substance in the body after surgery.

• Proceedings of the Korean Powder Metallurgy Institute Conference
• /
• 2002.07a
• /
• pp.25-37
• /
• 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 the Korean Wood Science and Technology
• /
• v.42 no.6
• /
• pp.714-722
• /
• 2014

In the previous study, a variety of wood-based panels was thermally decomposed to manufacture carbonized boards that had been proved to be high abilities of insect and fungi repellence, corrosion and fire resistant, electronic shielding, and formaldehyde adsorption as well as sound absorption performance. Based on the previous study, carbonized medium density fiberboard (c-MDF) was chosen to improve sound absorption performance by holing and sanding process. Three different types of holes (cross shape, square shape, and line) with three different sanding thickness (1, 2, and 3 mm) were applied on c-MDF and then determined sound absorption coefficient (SAC). The control c-MDF without holes had 14% of SAC, however, those c-MDFs with holes had 16.01% (square shape), 15.68% (cross shape), and 14.25% (line) of SAC. Therefore, making holes on the c-MDF did not significantly affect on the SAC. As the degree of sanding increased, the SAC of c-MDF increased approximately 65% on sanding treated c-MDFs (21.5, 21.83, and 19.37%, respectively) compared to the control c-MDF (13%). Based on these results, composite sound absorbing panel was developed with c-MDF and MDF (11 mm). The noise reduction coefficient of composite sound absorbing panel was 0.45 which was high enough to certify as sound absorbing material.

• Nuclear Engineering and Technology
• /
• v.55 no.10
• /
• pp.3913-3923
• /
• 2023

Todays, medium energy resolution detectors are preferably used in radioisotope identification devices(RID) in nuclear and radioactive material categorization. However, there is still a need to develop or enhance « automated identifiers » for the useful RID algorithms. To decide whether any material is SNM or NORM, a key parameter is the better energy resolution of the detector. Although masking, shielding and gain shift/stabilization and other affecting parameters on site are also important for successful operations, the suitability of the RID algorithm is also a critical point to enhance the identification reliability while extracting the features from the spectral analysis. In this study, a RID algorithm based on Bayesian statistical method has been modified for medium energy resolution detectors and applied to the uranium gamma-ray spectra taken by a LaBr3:Ce detector. The present Bayesian RID algorithm covers up to 2000 keV energy range. It uses the peak centroids, the peak areas from the measured gamma-ray spectra. The extraction features are derived from the peak-based Bayesian classifiers to estimate a posterior probability for each isotope in the ANSI library. The program operations were tested under a MATLAB platform. The present peak based Bayesian RID algorithm was validated by using single isotopes(²⁴¹Am, ⁵⁷Co, ¹³⁷Cs, ⁵⁴Mn, ⁶⁰Co), and then applied to five standard nuclear materials(0.32-4.51% at.²³⁵U), as well as natural U- and Th-ores. The ID performance of the RID algorithm was quantified in terms of F-score for each isotope. The posterior probability is calculated to be 54.5-74.4% for ²³⁸U and 4.7-10.5% for ²³⁵U in EC-NRM171 uranium materials. For the case of the more complex gamma-ray spectra from CRMs, the total scoring (ST) method was preferred for its ID performance evaluation. It was shown that the present peak based Bayesian RID algorithm can be applied to identify ²³⁵U and ²³⁸U isotopes in LEU or natural U-Th samples if a medium energy resolution detector is was in the measurements.

• The Journal of the Korea Contents Association
• /
• v.14 no.8
• /
• pp.233-242
• /
• 2014

The purpose of this study is to compare the reduction of the dose radioactivity by CARE kV with that of the Bismuth shielding. First, CT was performed with transparent materials, including a Bismuth shielder which is a well-known material for decreasing the dose of radiation. Moreover, we have estimated and compared the affects of the reduction of dose on eye lens, thyroid, breast and genitals. These steps aim to compare reactions with and without the application of the Rando phantom with PLD as well as with CARE kV or not. As a result, during the Brain angio scan, the dose of CARE kV set inspection test methods showed the least dose. Depending on whether we use CARE kV, which showed the effect of dose reduction by 63%. During the Carotid angio scan, the dose was increased by 13% by how to set CARE kV+Bismuth. During the Cardiac angio scan, which showed the effect of dose reduction by 31% by how to set CARE kV+Bismuth. During the Lower extremity angio scan, the dose was measured least by how to set up the whole Bismuth. Compared with CARE kV set of test methods, which showed the effect of dose reduction by 9%.

• Journal of the Korea Concrete Institute
• /
• v.26 no.3
• /
• pp.315-321
• /
• 2014

Concrete as a construction material is widely used in nuclear vessel and plant for excellent radiation shielding. However the isolation characteristics in concrete may affect adversely in the case of fire and melt-down in nuclear vessel since temperature cooling down is very difficult from outside. This study is for development of high thermal conductive concrete, and its mechanical and thermal properties are evaluated. Magnetite aggregates with volume ratio of 42.3% (maximum) and steel powder of 1.5% are replaced with normal aggregates and thermal properties are evaluated. Thermal conductivity little increases by 30% addition of magnetite but rapidly increases afterwards. Finally thermal conductivity is magnified to 2.5 times in the case of 42.3% addition of magnetite. Steel powder has a positive effect on high thermal conduction to 106~113%. Several models for thermal conduction like ACI, DEMM, and MEM are compared with test results and they are verified to reasonably predict the thermal conductivity with increasing addition of magnetite aggregates and steel powder.