• Nuclear Engineering and Technology
• /
• v.55 no.5
• /
• pp.1718-1733
• /
• 2023

The virtual Frisch-grid method for room-temperature radiation detectors has been widely used because of its simplicity and high performance. Recently, side electrodes were separately attached to each surface of the detectors instead of covering the entire detector surface with a single electrode. The side-electrode structure enables the measurement of the three-dimensional (3D) gamma-ray interaction in the detector. The positional information of the interaction can then be utilized to precisely calibrate the response of the detector for gamma-ray spectroscopy and imaging. In this study, we developed a 3D position-sensitive 5 × 5 × 12 mm³ cadmium-zinc-telluride (CZT) detector and applied a flattening method to correct detector responses. Collimated gamma-rays incident on the surface of the detector were scanned to evaluate the positional accuracy of the detection system. Positional distributions of the radiation interactions with the detector were imaged for quantitative and qualitative evaluation. The energy spectra of various radioisotopes were measured and improved by the detector response calibration according to the calculated positional information. The energy spectra ranged from 59.5 keV (emitted by ²⁴¹Am) to 1332 keV (emitted by ⁶⁰Co). The best energy resolution was 1.06% at 662 keV when the CZT detector was voxelized to 20 × 20 × 10.

• Nuclear Engineering and Technology
• /
• v.50 no.3
• /
• pp.439-445
• /
• 2018

Experimental measurements of the response of $BF_3$ detector to a 3 Ci Am-Be neutron source for three different reflecting materials, i.e., aluminum, wood, and Perspex of varying thicknesses have been carried out. The varying contribution of wall effect to the response due to change in active volume of the detector has also been determined experimentally. Then, a Monte Carlo code has been developed for the calculation of the neutron response function of the $BF_3$ detector using source biasing and importance sampling. This code simulates the $BF_3$ detector response exposed to the neutron field in a three-dimensional source, detector, and reflecting medium configurations. The results of simulation have been compared with the corresponding experimental measurements and are found to be in good agreement. The experimental neutron albedo measurements for various values of Perspex thickness show saturating behavior, and results agree very well with the data obtained by Monte Carlo simulation.

• The Korean Journal of Nuclear Medicine Technology
• /
• v.27 no.2
• /
• pp.129-132
• /
• 2023

Purpose: Positron emisson tomography (PET) is a crucial medical imaging scanner for the detection of cancer lesions. In order to maintain the improved image quality, it is crucial to apply detectors of superior performance. Therefore, the purpose of this study was to compare PET image quality using Monte Carlo simulation based on the detector materials of BGO, LSO, and LuAP. Materials and Methods: The Geant4 Application for Tomographic Emission (GATE) was used to design the PET detector. Scintillations with BGO, LSO and LuAP were modelled, with a size of 3.95 × 5.3 mm2 (width × height) and 25.0 mm (thickness). The PET detector consisted of 34 blocks per ring and a total of 4 rings. A line source of 1 MBq was modelled and acquired with a radius of 1 mm and length of 20 mm for 20 seconds. The acquired image was reconstructed maximum likelihood expectation maximization with 2 iteration and 10 subsets. The count comparison was carried out. Results and Discussion: The highest true, random, and scatter counts were obtained from the BGO scintillation detector compared to LSO and LuAP. Conclusion: The BGO scintillation detector material indicated excellent performance in terms of detection of gamma rays from emitted PET phantom.

• Nuclear Engineering and Technology
• /
• v.54 no.11
• /
• pp.4215-4219
• /
• 2022

The Cd_0.95Mn_0.05Te_0.98Se_0.02 (CMTS) ingot was grown by the vertical Bridgman technique at low pressure. All wafers showed high resistivity, which suggests potential as a room-temperature semiconductor detector. The resistivity of the CMTS planar detector was 1.47 × 10¹⁰ Ω·cm and mobility lifetime product of electrons was 1.29 × 10^-3 cm²/V. The spectroscopic property with Am-241 and Co-57 was evaluated. The energy resolution about 59.5 keV gamma-ray of Am-241 was 11% and the photo-peak of 122 keV gamma-ray from Co-57 was clearly distinguished. The result shows the first detector-grade CMTS in the world and proves CMTS's potential as a radiation detector operating at room temperature.

• Nuclear Engineering and Technology
• /
• v.49 no.7
• /
• pp.1489-1494
• /
• 2017

The use of a room-temperature gamma spectrometer is an issue in environmental radiation monitoring. To monitor radionuclides released around a nuclear power plant, suitable instruments giving fast and reliable information are required. High-pressure xenon (HPXe) chambers have range of resolution and efficiency equivalent to those of other medium resolution detectors such as those using NaI(Tl), CdZnTe, and $LaBr_3:Ce$. An HPXe chamber could be a cost-effective alternative, assuming temperature stability and reliability. The CEA LIST actively studied and developed HPXe-based technology applied for environmental monitoring. Xenon purification and conditioning was performed. The design of a 4-L HPXe detector was performed to minimize the detector capacitance and the required power supply. Simulations were done with the MCNPX2.7 particle transport code to estimate the intrinsic efficiency of the HPXe detector. A behavioral study dealing with ballistic deficits and electronic noise will be utilized to provide perspective for further analysis.

• Nuclear Engineering and Technology
• /
• v.36 no.1
• /
• pp.47-52
• /
• 2004

Recently, a new sintered pellet-type LiF:Mg,Cu,Na,Si TL detector which has a high sensitivity and good reusability, named KLT-300(KAERI LiF:Mg,Cu,Na,Si TL detector), was developed by the variation of the dopants concentrations and the parameters of the preparation procedure at KAERI (Korea Atomic Energy Research Institute). In this study, the thermoluminescent characteristics of the newly developed TL detectors were investigated. The sensitivity of the TL detector was compared with that of the TLD-100 by light integration. The dose linearity of the detector was tested from $10^{-6}$ Gy up to 30 Gy. The dose response was very linear up to 10 Gy and a sublinear response was observed at higher doses. The energy response of the detector was studied for photon energies from 20 keV to 662 keV. The result shows that a maximum response of 1.004 at 53 keV and a minimum response of 0.825 at 20 keV were observed. The reproducibility study for the TL detector was also carried out. The coefficients of variation for each detector separately did not exceed 0.016, and for all the 10 detectors collectively was 0.0054. Lower limit of detection for the detector was investigated at 70 nGy by the Harshaw 4500 TLD Reader and the residual signal of the TL detector was found to be $0.57\%$.

• Nuclear Engineering and Technology
• /
• v.52 no.5
• /
• pp.1022-1028
• /
• 2020

One of the possible options for spent-fuel management in Korea is pyroprocessing, which is a process for electrochemical recycling of spent nuclear fuel. Nuclear material accountancy is considered to be a safeguards measure of fundamental importance, for the purposes of which, the amount of nuclear material in the input and output materials should be measured as accurately as possible by means of chemical analysis and/or non-destructive assay. In the present study, a neutron measurement system based on the fast-neutron energy multiplication (FNEM) and passive neutron albedo reactivity (PNAR) techniques was designed for nuclear material accountancy of a spent-fuel assembly (i.e., the input accountancy of a pyroprocessing facility). Various parameters including inter-detector distance, source-to-detector distance, neutron-reflector material, the structure of a cadmium sleeve around the close detectors, and an air cavity in the moderator were investigated by MCNP6 Monte Carlo simulations in order to maximize its performance. Then, the detector responses with the optimized geometry were estimated for the fresh-fuel assemblies with different ²³⁵U enrichments and a spent-fuel assembly. It was found that the measurement technique investigated here has the potential to measure changes in neutron multiplication and, in turn, amount of fissile material.

• Nuclear Engineering and Technology
• /
• v.54 no.2
• /
• pp.601-607
• /
• 2022

In recent years, monitoring of spent nuclear fuel inside dry cask storage has become an important area of national security. Muon tomography is a useful method for monitoring spent nuclear fuel because it uses high energy muons that penetrate deep into the target material and provides a 3-D structure of the inner materials. We designed a muon tomography system consisting of four 2-D position sensitive detector and characterized and optimized the system parameters. Each detector, measuring 200 × 200 cm², consists of a plastic scintillator, wavelength shifting (WLS) fibers and, SiPMs. The reconstructed image is obtained by extracting the intersection of the incoming and outgoing muon tracks using a Point-of-Closest-Approach (PoCA) algorithm. The Geant4 simulation was used to evaluate the performance of the muon tomography system and to optimize the design parameters including the pixel size of the muon detector, the field of view (FOV), and the distance between detectors. Based on the optimized design parameters, the spent fuel assemblies were modeled and the line profile was analyzed to conduct a feasibility study. Line profile analysis confirmed that muon tomography system can monitor nuclear spent fuel in dry storage container.

• Nuclear Engineering and Technology
• /
• v.55 no.2
• /
• pp.681-689
• /
• 2023

Purpose: Muons are characterized by a strong penetrating ability and can travel through thousands of meters of rock, making them ideal to image large volumes and substances typically impenetrable to, for example, electrons and photons. The feasibility of 3D image reconstruction and material identification based on a cosmic ray muons tomography (MT) system with triangular bar plastic scintillator detectors has been verified in this paper. Our prototype shows potential application value and the authors wish to apply this prototype system to 3D imaging. In addition, an MT experiment with the same detector system is also in progress. Methods: A simulation based on GEANT4 was developed to study cosmic ray muons' physical processes and motion trails. The yield and transportation of optical photons scintillated in each triangular bar of the detector system were reproduced. An image reconstruction algorithm and correction method based on muon scattering, which differs from the conventional PoCA algorithm, has been developed based on simulation data and verified by experimental data. Results: According to the simulation result, the detector system's position resolution is below 1 ~ mm in simulation and 2 mm in the experiment. A relatively legible 3D image of lead bricks in size of 20 cm × 5 cm × 10 cm used our inversion algorithm can be presented below 1× 10⁴ effective events, which takes 16 h of acquisition time experimentally. Conclusion: The proposed method is a potential candidate to monitor the cosmic ray MT accurately. Monte Carlo simulations have been performed to discuss the application of the detector and the simulation results have indicated that the detector can be used in cosmic ray MT. The cosmic ray MT experiment is currently underway. Furthermore, the proposal also has the potential to scan the earth, buildings, and other structures of interest including for instance computerized imaging in an archaeological framework.

• The Korean Journal of Nuclear Medicine Technology
• /
• v.22 no.2
• /
• pp.105-111
• /
• 2018

Purpose Quality control of instrument takes up a large part in the Radioimmunoassays. The gamma-ray instrument, which is one of the important instruments in the laboratory, observes the condition and performance of instrument and performs quality control of the instrument by measuring the Normalization, Calibration, Background and etc. However, there are some automation instruments which can't measure the counting efficiency of gamma-ray meters, resulting in insufficient management in terms of performance evaluation of gamma-ray meters. Therefore, the purpose of this paper is to manage the quality control continuously and regularly by suggesting how to measure the counting efficiency of gamma-ray instruments. Materials and Methods In case of a comparative measurement method to a gamma-ray instrument dedicated to nuclear medical examination, the CPM and counting efficiency can be obtained after the measurement of normalization by inserting the I-125 $200{\mu}L$(CPM 50,000~500,000) into the test tube. With this CPM and counting efficiency values, it's possible to calculate the measurement of the DPM value and count the CPM from the automation instrument from the same source, and enter the DPM to calculate the counting efficiency using a comparative measurement method. Another method is to calculate the counting efficiency by estimating the half life using the radiation source information of the tracer in B test reagents of company A. Results According to the calculation formula using the DPM obtained by counting the normalization of gamma-ray meters, the detection efficiency was 75.16% for Detector 1, 76.88% for Detector 2, 77.13% for Detector 3, 75.36% for Detector 4 and 73.2% for Detector 5 respectively. Using another calculation formula estimated from the shelf life, the data of the detection efficiency from Detector 1 to Detector 5 were 74.9%, 75.1%, 76.5%, 74.9% and 73.2% respectively. Conclusion Although the accuracy of counting efficiencies of both methods are insufficient, this is considered to be useful for ongoing management of quality control if counting efficiency is managed after setting the acceptable ranges. For example, if the measurement efficiency is set to 70% or higher, the allowed %difference between measurements is within 3% and the %difference with the detector wall is set within 5%.