• Proceedings of the Korean Institute of Information and Commucation Sciences Conference
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• 2014.05a
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• pp.707-709
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• 2014

The radiation sources leaked from large-scale radiation leak accident like the Fukushima nuclear power plant accident or nuclear explosions can cause to the very large damage for us. So that the damage can be minimized, we have being developed a detector that can providing information about the location of the source to remove dangerous substances quickly than the conventional single detector. In this paper, we designed and implemented the radiation shield and the collimator for the development of the stereo radiation detector to detect contamination things using MCNP Simulation. And we analysed the test results of the radiation shield and collimator using the radiation source. The results of this paper will be used as the basis for improving the efficiency of the stereo radiation detector being studied currently.

• Journal of radiological science and technology
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• v.40 no.3
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• pp.407-413
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• 2017

Medical radiation offers significant benefits in diagnosing and treating patients, but it also generates unnecessary radiation exposure to those nearby. Accordingly, the objective of the present study was to analyze spatial dose rate according to types of radiation source term in multi-bed hospital rooms occupied by patients and general public. MCNPX was used for geometric simulation of multi-bed hospital rooms and radiation source terms, while the radiation source terms were established as whole body bone scan patients and imaging using a portable X-ray generator. The results of simulation on whole body bone scan patients showed $3.46{\mu}Sv/hr$ to another patient position, while experimental results on imaging using a portable X-ray generator showed $1.47{\times}10^{-8}{\mu}Sv/irradiation$ to another patient position in chest imaging and $2.97{\times}10^{-8}{\mu}Sv/irradiation$ to another patient position in abdomen imaging. Multi-bed hospital room, unnecessary radiation generated in the surrounding patients, while legal regulations and systematic measures are needed for radiation exposure in multi-bed hospital rooms that are currently lacking in Korea.

• Analytical Science and Technology
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• v.5 no.1
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• pp.57-61
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• 1992

For the neutron spectrum measurement, $^6Li$ neutron spectrometer system is installed. The characteristic of the $^6Li$ detector are investigated using a $^{137}Cs$ and $^{207}Bi$ point source, and the neutron capture peaks and the pulse height spectrum using an $^{214}Am-Be$ neutron source are measured. Furthermore, the pulse height spectrum for the irradiation time variation from the (214)^Am-Be neutron source, and for the distance variation between detector and source, and the threshold variation of discriminator are measured.

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.3073-3084
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• 2022

The propagation of radiation source uncertainties in spent nuclear fuel (SNF) cask shielding calculations is presented in this paper. The uncertainty propagation employs the depletion and source term outputs of the deterministic code STREAM as input to the transport simulation of the Monte Carlo (MC) codes MCS and MCNP6. The uncertainties of dose rate coming from two sources: nuclear data and modeling parameters, are quantified. The nuclear data uncertainties are obtained from the stochastic sampling of the cross-section covariance and perturbed fission product yields. Uncertainties induced by perturbed modeling parameters consider the design parameters and operating conditions. Uncertainties coming from the two sources result in perturbed depleted nuclide inventories and radiation source terms which are then propagated to the dose rate on the cask surface. The uncertainty analysis results show that the neutron and secondary photon dose have uncertainties which are dominated by the cross section and modeling parameters, while the fission yields have relatively insignificant effect. Besides, the primary photon dose is mostly influenced by the fission yield and modeling parameters, while the cross-section data have a relatively negligible effect. Moreover, the neutron, secondary photon, and primary photon dose can have uncertainties up to about 13%, 14%, and 6%, respectively.

• Journal of Radiation Protection and Research
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• v.28 no.3
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• pp.183-187
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• 2003

Collimator for $M{\ddot{o}}ssbauer$ source was manufactured for compton scattering experiment. Exposure dose rate was calculated and measured using GM counter for radiation evaluation. These results were well agreed to each other and used for collimator design. SUS303 was used for collimator material because exposure dose rate at 10 cm is about 2 mR/h. The radiation emited from the 35 mm, 65 mm hole was measured using gamma camera which have 4' diameter. 2-D radiation image was acquired and analyzed. The radiation size at Gamma Camera was 8.0 mm and 5.8 mm respectively.

• Journal of Sensor Science and Technology
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• v.14 no.4
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• pp.244-249
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• 2005

In this study, radiation sensor tips are fabricated for remote sensing of X or gamma ray with inorganic scintillators and plastic optical fiber. The visible range of light from the inorganic scintillator that is generated by radiation source is guided by the plastic optical fiber and is measured by optical detector and power-meter. Two kinds of sensor tips are designed and fabricated such as film type and powder type. Many kinds of inorganic scintillators are used to fabricate both sensor tips, and the different wavelength of emitting lights from them are measured to determine the optimal inorganic scintillator which has maximum light output. As a radiation source X-ray generator and Ir-192 are selected to test a performance of sensor tip. It is expected that the fiber-optic radiation sensor is widely used in nuclear industry and medical applications due to its special characteristics such as good flexibility, easy in processing, long lengths and no interference to electro magnetic field.

• Nuclear Engineering and Technology
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• v.53 no.8
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• pp.2646-2651
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• 2021

A rectangular-shaped PET system with an adjustable gantry (AGPET) has been developed for imaging small animals. The AGPET system employs a new depth of interaction (DOI) method using a depth dependent reflector patterns and a new digital time pickoff method based on the pulse reconstruction method. To evaluate the performance of the AGPET, timing resolution, intrinsic spatial resolution and point source images were acquired. The timing resolution and intrinsic spatial resolution were measured using two detector modules and Na-22 gamma source. The PET images were acquired in two field of view (FOV) sizes, 30 mm and 90 mm, to demonstrate the characteristic of the AGPET. As a result of in the experiment results, the timing resolution was 0.9 ns using the pulse reconstruction method based on the bi-exponential model. The intrinsic spatial resolution was an average of 1.7 mm and the spatial resolution of PET images after DOI correction was 2.08 mm and 2.25 mm at the centers of 30 mm and 90 mm FOV, respectively. The results show that the proposed AGPET system provided higher sensitivity and resolution for small animal imaging.

• Journal of Radiation Protection and Research
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• v.40 no.2
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• pp.92-100
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• 2015

For the purpose of improving the efficiency of the radiation impact assessment of dry interim storage facilities for the spent nuclear fuel of pressurized water reactors (PWRs), radiation impact assessment was performed after the application of sensitivity assessment according to the radiation source term designation method, development of a 2-step calculation technique, and cooling time credit. The present study successively designated radiation source terms in accordance with the cask arrangement order in the shielding building, assessed sensitivity, which affects direct dose, and confirmed that the radiation dosage of the external walls of the shielding building was dominantly affected by the two columns closest to the internal walls. In addition, in the case in which shielding buildings were introduced into storage facilities, the present study established and assessed the 2-step calculation technique, which can reduce the immense computational analysis time. Consequently, results similar to those from existing calculations were derived in approximately half the analysis time. Finally, when radiation source terms were established by adding the storage period of the storage casks successively stored in the storage facilities and the cooling period of the spent nuclear fuel, the radiation dose of the external walls of the buildings was confirmed to be approximately 40% lower than the calculated values; the cooling period was established as being identical. The present study was conducted to improve the efficiency of the Monte Carlo shielding analysis method for radiation impact assessment of interim storage facilities. If reliability is improved through the assessment of more diverse cases, the results of the present study can be used for the design of storage facilities and the establishment of site boundary standards.

• Nuclear Engineering and Technology
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• v.54 no.4
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• pp.1421-1430
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• 2022

The computation of the solid angle and the detector efficiency is considering to be one of the most important factors during the measuring process for the radioactivity, especially the cylindrical γ-ray NaI(Tl) detectors nowadays have applications in several fields such as industry, hazardous for health, the gamma-ray radiation detectors grow to be the main essential instruments in radiation protection sector. In the present work, a generic numerical simulation method (NSM) for calculating the efficiency of the γ-ray spectrometry setup is established. The formulas are suitable for any type of source-to-detector shape and can be valuable to determine the full-energy peak and the total efficiencies and P/T ratio of cylindrical γ-ray NaI(Tl) detector setup concerning the truncated conical radioactive source. This methodology is based on estimate the path length of γ-ray radiation inside the detector active medium, inside the source itself, and the self-attenuation correction factors, which typically use to correct the sample attenuation of the original geometry source. The calculations can be completed in general by using extra reasonable and complicate analytical and numerical techniques than the standard models; especially the effective solid angle, and the detector efficiency have to be calculated in case of the truncated conical radioactive source studied condition. Moreover, the (NSM) can be used for the straight calculations of the γ-ray detector efficiency after the computation of improvement that need in the case of γ-γ coincidence summing (CS). The (NSM) confirmation of the development created by the efficiency transfer method has been achieved by comparing the results of the measuring truncated conical radioactive source with certified nuclide activities with the γ-ray NaI(Tl) detector, and a good agreement was obtained after corrections of (CS). The methodology can be unlimited to find the theoretical efficiencies and modifications equivalent to any geometry by essential sufficiently the physical selective considered situation.

• Journal of Radiation Industry
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• v.9 no.1
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• pp.1-7
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• 2015

The trend of x-ray image sensor has been evolved from an amorphous silicon sensor to a crystal silicon sensor. A crystal silicon X-ray sensor, meaning a X-ray CIS (CMOS image sensor), is consisted of three transistors (Trs), i.e., a Reset Transistor, a Source Follower and a Select Transistor, and a photodiode. They are highly sensitive to radiation exposure. As the frequency of exposure to radiation increases, the quality of the imaging device dramatically decreases. The most well known effects of a X-ray CIS due to the radiation damage are increments in the reset voltage and dark currents. In this study, a pixel array of a X-ray CIS was made of $20{\times}20pixels$ and this pixel array was exposed to a high radiation dose. The radiation source was Co-60 and the total radiation dose was increased from 1 to 9 kGy with a step of 1 kGy. We irradiated the small pixel array to get the increments data of the reset voltage and the dark currents. Also, we simulated the radiation effects of the pixel by MCNP (Monte Carlo N-Particle) simulation. From the comparison of actual data and simulation data, the most affected location could be determined and the cause of the increments of the reset voltage and dark current could be found.