• Nuclear Engineering and Technology
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• v.53 no.6
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• pp.1955-1963
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• 2021

As the IAEA has applied integrated safeguards and a state level approach to member states, the importance of national inspection has increased. However, the requirements for national inspection for some member states are different from the IAEA safeguards. In particular, the national inspection for the ROK requires on-site U concentration analysis due to a domestic notification. This research proposes an on-site U concentration analysis (OUCA) method for UO₂ pellets using gamma spectroscopy to satisfy the domestic notification requirement. The OUCA method calculates the U concentration of UO₂ pellets using the measured net X-ray counts and declared ²³⁵U enrichment. This research demonstrates the feasibility of the OUCA method using both MCNP simulation and experiment. It simulated and measured the net X-ray counts of different UO₂ pellets with different U concentrations and ²³⁵U enrichments. The simulated and measured net X-ray counts were fitted to polynomials as a function of U concentration and ²³⁵U enrichment. The goodness-of-fit results of both simulation and experiment demonstrated the feasibility of the OUCA method.

• Journal of the Korea institute for structural maintenance and inspection
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• v.23 no.7
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• pp.172-180
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• 2019

In this study, the microstructure and gamma-ray shielding efficiency of CRT glass mortar specimen were evaluated with replacement ratio and material properties. The results show that as the replacement ratio of CRT waste glass increases, the volume of pores with diameters below 50 nm and above 400 nm is increased. Also, the half-value layer of CRT glass mortar decreased with the increasing of linear attenuation coefficient. In addition, compressive and flexural strength were reduced when CRT waste glass was replaced as the fine aggregate, but the mechanical performance of CRT mortar specimen could be obtained by substitution of the mineral admixture.

• Nuclear Engineering and Technology
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• v.51 no.3
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• pp.800-806
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• 2019

This paper presents a laboratory experiment on data acquisition technique that applied to the gamma radiation scanning coupled with computed tomography (CT) technique for inspection of broken nozzle inside the vertical vessel. The acquisition technique was developed to inspect a large diameter vessel when suspicious problem location is not easily accessed. This technique allows the installation of gamma radiation source (Cesium 137, Cs-137), and detectors (Sodium Iodine. NaI(Tl)) from the accessible location to the required location and performs the scanning by designed pattern. To demonstrate the designed technique, top opened tank which installed with six cut steel pipes diameter of 76.2 mm (3") at a certain position was selected. They were assumed to be a gas riser pipes inside the vessel. Three studied cases were performed, (a) projection of well installed six pipes, (b) projection of one out of six broken pipe and (c) one of nozzle was assumed to be failure and fell down until one out of six pipes was broken and obstructed by nozzle. Results clearly indicated the capability of developed technique to distinguish between normal situation case and abnormal situation cases.

• Journal of the Korean Society for Nondestructive Testing
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• v.19 no.5
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• pp.356-362
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• 1999

Based on the assumption that film density increases exponentially with exposure in the industrial radiographic film. an equation representing the characteristic curves of industrial radiographic films and a new density-thickness relation are suggested. The accuracy and reliability of the suggested relation has been tested using radiographs of a carbon steel step wedge with known thickness variation by polychromatic X-ray and ${\gamma}$-ray ($Ir^{192}$). The experimental results were well agreed to the proposed relation in the range of film densities from 1.0 to 3.5 and it was more accurate than the conventional relation. It is also found that ${\gamma}$-ray is more effective in this purpose than polychromatic X-ray, which results in variation of effective linear absorption coefficient due to beam hardening effect as thickness increases. Therefore using the equation and experimentally determined parameters the quantitative evaluation of thickness variation is possible and it can be used to evaluate the depth of local corrosion of pressure vessels in plants.

• 연구논문집
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• s.31
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• pp.135-147
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• 2001

The wall-thickness of insulated pipelines can be easily evaluated by measuring the gamma-ray transmission intensity because this intensity is inversely proportional to the thickness of insulated pipeline. The main purpose of this study is to develop the nondestructive and filmless on-line inspection system of corrosion by measuring the wall thickness of insulated pipeline. The inspection system is constructed with radioisotope, 64 channel photo diode array detector, crawler system and data taking and operating software. The traditional off-line radiographic method carried out by exposing film cassettes can be replaced by this cost-effective on-line digital imaging method and the application will be greatly expected especially in the chemical and petrochemical industries.

• Journal of Radiation Protection and Research
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• v.48 no.4
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• pp.167-174
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• 2023

Background: A nondestructive test is commonly used to inspect the surface defects and internal structure of an object without any physical damage. X-rays generated from an electron accelerator or a tube are one of the methods used for nondestructive testing. The high penetration of X-rays through materials with low atomic numbers makes it difficult to discriminate between these materials using X-ray imaging. The interaction characteristics of neutrons with materials can supplement the limitations of X-ray imaging in material discrimination. Materials and Methods: The radiation image acquisition process for air-cargo security inspection equipment using X-rays and neutrons was simulated using a GEometry ANd Tracking (Geant4) simulation toolkit. Radiation images of phantoms composed of 13 materials were obtained, and the R-value, representing the attenuation ratio of neutrons and gamma rays in a material, was calculated from these images. Results and Discussion: The R-values were calculated from the simulated X-ray and neutron images for each phantom and compared with those obtained in the experiments. The R-values obtained from the experiments were higher than those obtained from the simulations. The difference can be due to the following two causes. The first reason is that there are various facilities or equipment in the experimental environment that scatter neutrons, unlike the simulation. The other is the difference in the neutron signal processing. In the simulation, the neutron signal is the sum of the number of neutrons entering the detector. However, in the experiment, the neutron signal was obtained by superimposing the intensities of the neutron signals. Neutron detectors also detect gamma rays, and the neutron signal cannot be clearly distinguished in the process of separating the two types of radiation. Despite these differences, the two results showed similar trends and the viability of using simulation-based radiation images, particularly in the field of security screening. With further research, the simulation-based radiation images can replace ones from experiments and be used in the related fields. Conclusion: The Korea Atomic Energy Research Institute has developed air-cargo security inspection equipment using neutrons and X-rays. Using this equipment, radiation images and R-values for various materials were obtained. The equipment was reconstructed, and the R-values were obtained for 13 materials using the Geant4 simulation toolkit. The R-values calculated by experiment and simulation show similar trends. Therefore, we confirmed the feasibility of using the simulation-based radiation image.

• Journal of Radiation Industry
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• v.8 no.1
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• pp.23-27
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• 2014

Industrial single photon emission computed tomography (SPECT) is an useful technique to investigate the dynamic behavior of process flow. In the present study, to evaluate the influence of a void on the performance of industrial SPECT, industrial SPECT with various radioisotope sources and gas holdups was modeled by the Monte Carlo simulation. The results are very encouraging; that is, the void little influences the performance of industrial SPECT, which means that industrial SPECT could be a suitable tool to investigate the dynamic characteristics of the flow in a water-air phase process.

• Journal of the Korean Society of Radiology
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• v.11 no.4
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• pp.183-188
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• 2017

In recent years, the automatic remote control controller of the gamma ray irradiator malfunctions, and radiation workers are continuously exposed to radiation exposure accidents. In the non-destructive testing field, much time and resources are invested in establishing a radioactive source monitoring system in order to prevent potential incidents of radiation. In this study, the gamma-ray response properties of the lead monoxide-based radiation detector were estimated through monte carlo simulation as a previous study for the development of a radioactive source location monitoring system that can be applied universally to various non-destructive testing equipment. As a result of the study, the optimized thickness of the radiation detector varies according to the gamma-ray energy emitted from the radioactive source, and the optimized thickness gradually increases with increasing energy. In conclusion, the optimized thickness of the lead monoxide-based radiation detector was $200{\mu}m$ for the Ir-192, $150{\mu}m$ for the Se-75 and $300{\mu}m$ for the Co-60. Based on these results, the appropriate thickness of lead monoxide-based radiation detector considering secondary-electron equilibrium was evaluated to be $300{\mu}m$ for general application. These results can be used as a basic data for determining the appropriate thickness required in the radiation detector when developing a radiation source location monitoring system for universal application to various non-destructive testing equipment in the future.

• Journal of Radiation Protection and Research
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• v.26 no.3
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• pp.275-279
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• 2001

In uranium enrichment plants and nuclear fuel fabrication facilities, exact measurement of fissile isotope enrichment of uranium is required for material accounting in international safeguards inspection as well as process quality control. The purpose of this study was to develop a simple measurement system which can portably be used at nuclear fuel fabrication plants especially dealing with low enriched uranium. For this purpose, a small size CZT (CdZnTe) detector was used, and the detector performance in low uranium gamma/X -rays energy range was investigated by use of various enriched uranium oxide samples. New enrichment measurement technique and analysis method for low enriched uranium oxide, so-called, 'semi-peak ratio technique' was developed. The newly developed method was considered as an alternative technique for the low enrichment and would be useful to account nuclear material in safeguarding activity at nuclear fuel fabrication facility.

• Korean Journal of Digital Imaging in Medicine
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• v.12 no.2
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• pp.113-117
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• 2010

This research attempts to qualitatively evaluate the intensity change by radiopharmaceuticals and obtain computed tomography using phantom injected with various nuclide. Cylindrical phantom is used for comparing and analysing the effect on diagnosis image during radiopharmaceuticals inspection. Inside of the phantom, water is injected and computed tomography image is scanned. During nuclear medicine invitro, frequently used radiopharmaceuticals, $^{99m}TcO_4$ 20 mCi and $^{18}F$ 14 mCi, is diluted in the water phantom and scanned in the same method. Traverse image obtained by CT scan is divided into six traverse image in the same slice of each scanned image. CT-number(HU) value of 10 measuring point is measured in 2 cm interval based on the center of the phantom. Measured HU value, based on the water phantom, is compared with the image after injecting $^{99m}TcO_4$ and $^{18}F$. Average scale of water is 2.8~1.6 HU, $^{99m}TcO_4$ is 3.0~1.6 HU and $^{18}F$ is 1.2~0 HU. Average of water is $2.3{\pm}0.17$ HU, $^{99m}TcO_4$ is $2.2{\pm}0.85$ HU and F-18 is $0.7{\pm}0.95$ HU. Based on water, reduced value of about 0.1 HU and about 0.5 HU is acquired from $^{99m}TcO_4$ and F-18. Radionuclide used in nuclear medicine inspection utilizes 100~200 KeV energy and obtains image through scintillation camera and PET-CT utilizes 511 KeV positron annihilation energy to obtain image. What we learned from this research is that gamma rays from these energies used in CT scan for diagnosis purpose or radioactive therapy plan can change the intensity of the image. The nuclear medicine inspection for reducing the effect of emitted gamma ray diagnosis image should be obtained after a period of time considering half-life which would be reduced distortion or changed in image.