• Journal of Biomedical Engineering Research
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• v.18 no.3
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• pp.267-272
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• 1997

he scintillation detector having $BaF^2$ crystal with 3.6cm dia${\times}$2.0 cm thick was provided. The energy and timing characteristics were measured and compared with NaI(Tl) scintillation detectors, which widely used in unclear medicine. In order to measure the energy spectrum, the radioactive sources used were $^{22}Na,\;^{54}Mn,\;^{57}Co,\;^{137}Cs$ and the source to detector distance was 7cm. For the timing characteristic, NaI(Tl)(1" ${\times}$ 1")-$BaF^2$ and NaI(Tl)(3" ${\times}$ 3")-$BaF^2$ timing coincidence systems were prepared and the used source was $^{22}Na$ emitting 511keV annihilation photons. For the 511keV gamma-ray emitted from $^{22}Na$, It was revealed that the timing response of the $BaF^2$ detector was faster than NaI(Tl)(1" ${\times}$ 1") and NaI(Tl)(3" ${\times}$ 3") detector used in this experimental investigation. The energy characteristics of the $BaF^2$ detector had a good values for about 500keV energy range.

• Journal of the Korean Society of Radiology
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• v.7 no.5
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• pp.333-338
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• 2013

Generally, it's known fact that intensity of radioactivity satisfies inverse-square law. However, the law was dissatisfied with practical experiment because of limited shape of scintillation detector. Especially, in the case of near distance between the surface of detector and the radioactive source, the difference grows larger. In the present study, reason of this difference was confirmed by experiment with $2^{{\prime}{\prime}}{\times}2^{{\prime}{\prime}}{\phi}$ NaI(Tl) scintillation detector and $^{60}Co$(1.174 MeV, 1.333 MeV)and $^{137}Cs$(0.662 MeV) gamma ray sources. From the experiment, the correction coefficient was obtained with gamma ray detection efficiency and geometrical volume. In the result of the present study, the efficiency difference of the detector was corrected with the coefficient. In the present result, we obtained that the inverse-square law experiment have to consider the efficiency and geometrical value of the detector.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.3
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• pp.321-328
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• 2019

The detection performances of the NaI(Tl), $LaBr_3$(Ce) and $CeBr_3$ scintillation detectors, which can be used to rapidly evaluate the major artificial radionuclides deposited on the soil surface in a nuclear accident or radiological emergency, were compared. Detection performance was assessed by calculating the minimum detectable activity (MDA). The detection efficiency of each detector for artificial radionuclides was semi-empirically determined using mathematical modelling and point-like sources having certified radioactivity. The background gamma-ray energy spectrum for MDA evaluation was obtained from relatively wide and flat grassland, and the MDA values of each detector for the major artificial radionuclides that could be released in nuclear accidents were calculated. As a result, the relative MDA values of each detector regarding surface deposition distribution at normal environmental radiation level were evaluated as high in the order of the NaI(Tl), $LaBr_3$(Ce), and $CeBr_3$ detectors. These results were compared based on each detector's intrinsic and measurement environment background, detection efficiency, and energy resolution for the gamma-ray energy region of the radionuclide of interest.

• Journal of Radiation Protection and Research
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• v.13 no.2
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• pp.29-40
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• 1988

A spectrum-exposure rate conversion operator G(E) for a portable HPGe detector used for field environmental radiation survey was theoretically developed on the basis of a space distribution function of gamma flux emitted from a disk source and an areal efficiency of the detector. The radiation exposure rates measured using this G(E) and the portable HPGe. detector connected to a portable multichannel analyzer were compared with those measured by a 3' ${\phi}\;{\times}$3' NaI(Tl) scintillation detector with the reported G(E) and a pressurized ionization chamber. A comparison of the three results showed that the result obtained using the HPGe detector was lower than those determined using the NaI(Tl) detector and ionization chamber by 17% to 29%, The difference obtained is close to that reported in literature. The method developed here can be easily applicable to obtain a G(E) factor suitable to any detector for detecting the exposure rate of environmental gamma radiation, since the spectrum-exposure rate conversion operator can be calculated by a hand calculator.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.11 no.3
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• pp.245-251
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• 2013

The energy band and the G-factor method were compared to determine the exposure rate from the measured spectrum using a NaI(Tl) scintillation detector. First, G-factors of a 3"${\Phi}X3$" NaI(Tl) detector mounted to a EFRD 3300, which means the environmental radiation monitor, in Korea Atomic Energy Research Institute (KAERI) were calculated for several directions of incident photons through the MCNP modeling, and the optimum G-factor applicable to that monitor was then determined by comparing the results both the energy band method and the G-factor method. The results for these spectrometric determinations were also compared with the dose rate from a HPIC radiation monitor around a EFRD 3300. The measured value at the EFRD 3300 based on a 3"${\Phi}X3$" NaI(Tl) detector was $7.7{\mu}R/h$ and its difference was shown about $3{\mu}R/h$, when compared with the results from a HPIC radiation moditor. Since a HPIC is known to be able to measure cosmic rays with the relatively high energy, the difference between them was caused by cosmic rays which were not detected in a 3"${\Phi}X3$" NaI(Tl) detector.

• Nuclear Engineering and Technology
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• v.51 no.4
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• pp.1091-1097
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• 2019

After the Fukushima accident in 2011, there has been increased public concern about radioactive contamination of water resources through fallout in neighboring countries. However, there is still no available initial response system that can promptly detect radionuclides. The purpose of this research is to develop the most efficient gamma spectrometer to monitor radionuclides in an aquatic environment. We chose a thallium-doped sodium iodide (NaI(Tl)) scintillator readout with a silicon photo multiplier (SiPM) due to its compactness and low operating voltage. Three types of a scintillation detector were tested. One was composed of a scintillator and a photomultiplier tube (PMT) as a reference; another system consisted of a scintillator and an array of SiPMs with a light guide; and the other was a scintillator directly coupled with an array of SiPMs. Among the SiPM-based detectors, the direct coupling system showed the best energy resolution at all energy peaks. It achieved 9.76% energy resolution for a 662 keV gamma ray. Through additional experiments and a simulation, we proved that the light guide degraded energy resolution with increasing statistical uncertainty. The results indicated that the SiPM-based scintillation detector with no light guide is the most efficient design for monitoring radionuclides in an aquatic environment.

• Nuclear Engineering and Technology
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• v.54 no.10
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• pp.3795-3802
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• 2022

NaI(Tl) scintillation materials are considered to be one of many materials that are used exclusively for γ-ray detection and spectroscopy. The gamma-ray spectrometer is not an easy-to-use device, and the accuracy of the numerical values must be carefully checked based on the rules of the calibration technique. Therefore, accurate information about the detection system and its effectiveness is of greater importance. The purpose of this study is to estimate, using an analytical-numerical formula (ANF), the purely geometric solid angle, geometric efficiency, and total efficiency of a cylindrical NaI(Tl) γ-ray detector with a side-through hole. This type of detector is ideal for scanning fuel rods and pipelines, as well as for performing radio-immunoassays. The study included the calculation of the complex solid angle, in combination with the use of various points like gamma sources, located axially and non-axially inside the through detector side hole, which can be applied in a hypothetical method for calibrating the facility. An extended γ-ray energy range, the detector, source dimensions, "source-to-detector" geometry inside the side-through hole, path lengths of γ-quanta photons crossing the facility, besides the photon average path length inside the detector medium itself, were studied and considered. This study is very important for an expanded future article where the radioactive point source can be replaced by a volume source located inside the side-trough hole of the detector, or by a radioactive pipeline passing through the well. The results provide a good and useful approach to a new generation of detectors that can be used for low-level radiation that needs to be measured efficiently.

• Nuclear Engineering and Technology
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• v.50 no.3
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• pp.462-469
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• 2018

The background spectrum of a $3^{{\prime}{\prime}}{\times}3^{{\prime}{\prime}}$ NaI(Tl) well-type scintillation SILENA detector was measured without shielding, in 6 cm thick lead shielding, and with 2 mm thick electrolytic copper covering the detector inside the lead shielding. The relative remaining background of the lead shield lined with copper was found to be ideal for low-level environmental radioactive spectroscopy. The background total count rate in the (20-2160 KeV) was reduced 28.7 times by the lead and 29 times by the Cu + Pb shielding. The effective reduction of background (1.04) by the copper mainly appeared in the energy range from X-ray up to 500 KeV, while for the total energy range the ratio is 1.01 relative to the lead only. In addition, a strong relation between the full-energy peak absolute efficiency and the detector well height was found using gamma-ray isotropic radiation point sources placed inside the detector well. The full-energy peak efficiency at a midpoint of the well (at 2.5 cm) is three times greater than that on the detector surface. The energy calibrations and the resolution of any single energy line are independent of the locations of the gamma source inside or outside of the well.

• Journal of Radiation Protection and Research
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• v.40 no.2
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• pp.73-78
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• 2015

In this study, a scintillation detector was fabricated using a silicon photomultiplier (SiPM) and a Ce:GAGG scintillator single crystal, and its spectroscopic properties were compared with those of commercially available LYSO and CsI:Tl scintillators using ${\gamma}$-ray spectroscopy. The energy resolutions of the self-produced scintillation detector composed of the scintillator single crystal (volume: $3{\times}3{\times}20mm^3$) and SiPM (Photosensitive area: $3{\times}3mm^2$) for standard ${\gamma}$-ray sources, such as $^{133}Ba$, $^{22}Na$, $^{137}Cs$ and $^{60}Co$ were measured and compared. As a result, the energy resolutions of the proposed Ce:GAGG scintillation detector for g-rays, as measured using its spectroscopic properties, were found to be 13.5% for $^{133}Ba$ 0.356 MeV, 6.9% for $^{22}Na$ 0.511 MeV, 5.8% for $^{137}Cs$ 0.662 MeV and 2.3% for $^{60}Co$ 1.33 MeV.

• Journal of Radiation Protection and Research
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• v.48 no.3
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• pp.131-143
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• 2023

Background: This study examined the detection limit of thyroid screening monitoring conducted at the time of the Fukushima Daiichi Nuclear Power Plant (FDNPP) accident in 2011 using a Monte Carlo simulation. Materials and Methods: We calculated the detection limit of a NaI(Tl) survey meter to measure ¹³¹I accumulation in the thyroid gland of children. Mathematical phantoms of 1- and 5-year-old children were developed in the simulation of the Particle and Heavy Ion Transport code System code. Contamination of the body surface with eight radionuclides found after the FDNPP accident was assumed to have been deposited on the neck and shoulder area. Results and Discussion: The detection limit was calculated as a function of ambient dose rate. In the case of 40 Bq/cm² contamination on the body surface of the neck, the present simulations showed that residual thyroid radioactivity corresponding to thyroid dose of 100 mSv can be detected within 21 days after intake at the ambient dose rate of 0.2 µSv/hr and within 11 days in the case of 2.0 µSv/hr. When a time constant of 10 seconds was used at the dose rate of 0.2 µSv/hr, the estimated survey meter output error was 5%. Evaluation of the effect of individual differences in the location of the thyroid gland confirmed that the measured value would decrease by approximately 6% for a height difference of ±1 cm and increase by approximately 65% for a depth of 1 cm. Conclusion: In the event of a nuclear disaster, simple measurements carried out using a NaI(Tl) scintillation survey meter remain effective for assessing ¹³¹I intake. However, it should be noted that the presence of short-half-life radioactive materials on the body surface affects the detection limit.