To non-destructively determine the burnup of a spent nuclear fuel assembly, it is essential to analyze the nuclear isotopes present in the assembly and detect the neutrons and gamma rays emitted from these isotopes. Specifically, gamma-ray measurement methods can utilize a single radiation measurement value of 137Cs or measure based on the energy peak ratio of Cs isotopes such as 134Cs/137Cs and 154Eu/137Cs. In this study, we validated the extent to which the results of gamma-ray measurements using cadmium zinc telluride (CZT) sensors based on 137Cs could be accurately simulated by implementing identical conditions on MCNP. To simulate measurement scenarios using a lead collimator, we propose equations that represent radiation behavior that reaches the detector by assuming “Direct hit” and “Penetration with attenuation” situations. The results obtained from MCNP confirmed an increase in measurement efficiency by 0.47 times when using the CZT detector, demonstrating the efficacy of the measurement system.

In this study, we conduct a multi-frequency analysis of the gamma-ray bright blazar 1308+326 from February 2013 to March 2020, using the Korean VLBI Network at 22 and 43 GHz and gamma-ray data from the Fermi Large Area Telescope (LAT). Our findings reveal spectral variations around the 2014 gamma-ray flare, aligning with the shock-in-jet model. A strong correlation is observed between gamma-ray and 43 GHz emissions, with a 27-day lag in the VLBI core light curve, indicating a 50-day delay from the beginning of a specific radio flare to the gamma-ray peak. This radio flare correlates with a new jet component, suggesting the 2014 gamma-ray flare resulted from its interaction with a stationary component. Our analysis indicates the 2014 gamma-ray flare originated 40–63 parsecs from the central engine, with seed photons for the gamma-ray emission unlikely from the broad-line region.

In this study, we examined the effects of gamma irradiation dosage on the mycelial growth of Auricularia auriculajudae and performed analyses of fruiting body yield, growth characteristics, taste, fragrance, and mineral composition. Assessments of mycelial growth in response to gamma irradiation at different intensities revealed an enhancement in the growth of fungi exposed to irradiation at 200 Gy. Fruiting body yield was also highest at 200 Gy, followed by 800 Gy and the control group. On the basis of these observations, we subsequently applied gamma ray doses of 200 and 800 Gy to examine the effects of irradiation on fungal quality characteristics. In terms of the taste of fruiting bodies, we detected no significant differences among the control, 200 Gy, and 800 Gy groups. Contrastingly, with respect to fragrance, we found that fungi treated with 200 Gy were characterized by a pattern that differed from those of the control and other treatment groups. Furthermore, whereas we detected no significant difference among treatments with respect total dietary fiber content, calcium content was found to be higher in the treatment groups compared with the control group, with the highest content being measured in fungi exposed to 800 Gy irradiation. Copper content was confirmed to be higher in the control group, whereas there were no significant differences between the fungi irradiated with 200 and 800 Gy. Contrastingly, the highest levels of zinc were detected in response to 200 Gy irradiation, followed by 800 Gy. Collectively, our findings thus indicate that gamma irradiation can contribute to promoting increases in the fruiting body yield and mineral contents of mushrooms.

The inorganic scintillator used in gamma spectroscopy must have good efficiency in converting the kinetic energy of charged particles into light as well as high light output and high light detection efficiency. Accordingly, various studies have been conducted to enhance the net-efficiency. One way to improve the light yield has been studied by coating scintillators with various nanoparticles, so that the scintillation light can undergo resonance on surface between scintillators and nanoparticles resulting in higher light yield. In this study, an inorganic scintillator coated with CsPbBr3 perovskite nanocrystals using dip coating technique was proposed to improve scintillation light yield. The experiment was carried out by measuring scintillation light output, as the result of interaction between inorganic scintillator coated with CsPbBr3 perovskite nanocrystals and gamma-ray emitted from Cs-137 gamma source. The experimental results show that the channel corresponding to 662 keV full energy peak in the Cs-137 spectrum shifted to the right by 14.37%. Further study will be conducted to investigate the detailed relationships between the scintillation light yield and the characteristics of coated perovskite nanoparticles, such as diameter of nanoparticles, coated area ratio and width of coated region.

We present the analysis results of the simultaneous multifrequency observations of the blazar 4C +28.07. The observations were conducted by the Interferometric Monitoring of Gamma-ray Bright Active Galactic Nuclei (iMOGABA) program, which is a key science program of the Korean Very Long Baseline Interferometry (VLBI) Network (KVN). Observations of the iMOGABA program for 4C +28.07 were conducted from 16 January 2013 (MJD 56308) to 13 March 2020 (MJD 58921). We also used γ-ray data from the Fermi Large Array Telescope (Fermi-LAT) Light Curve Repository, covering the energy range from 100 MeV to 100 GeV. We divided the iMOGABA data and the Fermi-LAT data into five periods from 0 to 4, according to the prosody of the 22 GHz data and the presence or absence of the data. In order to investigate the characteristics of each period, the light curves were plotted and compared. However, a peak that formed a hill was observed earlier than the period of a strong γ-ray flare at 43–86 GHz in period 3 (MJD 57400–58100). Therefore, we assumed that the minimum total CLEANed flux density for each frequency was quiescent flux (Sq) in which the core of 4C +28.07 emitted the minimum, with the variable flux (Svar) obtained by subtracting Sq from the values of the total CLEANed flux density. We then compared the variability of the spectral indices (α) between adjacent frequencies through a spectral analysis. Most notably, α22–43 showed optically thick spectra in the absence of a strong γ-ray flare, and when the flare appeared, α22–43 became optically thinner. In order to find out the characteristics of the magnetic field in the variable region, the magnetic field strength in the synchrotron self-absorption (BSSA) and the equipartition magnetic field strength (Beq) were obtained. We found that BSSA is largely consistent with Beq within the uncertainty, implying that the SSA region in the source is not significantly deviated from the equipartition condition in the γ-ray quiescent periods.

Radioactive waste can be classified according to the concentration level for radionuclides, and the disposal method is different through the level. Gamma analysis is inevitably performed to determine the concentration of radioactive waste, and when a large amount of radioactive waste is generated, such as decommissioning nuclear facilities, it takes a lot of time to analyze samples. The performance of a lot of analysis can cause human errors and workload. In general, gamma analysis is performed using by HPGe detector. Recently, for convenience of analysis, commercial automatic sample changers applicable to the HPGe detectors were developed. The automatic sample changers generate individual analysis reports for each sample. In this study, gamma analysis procedure was improved using the application of the automatic sample changer and the automated data parsing using by Python. The application of automatic sample changers and data parsing technique can solve the problems. The human errors were reduced to 0% compared to the previous method by improving the gamma analysis procedure, and working time were also dramatically reduced. This automation of analysis procedure will contribute to reducing the burden of analysis work and reducing human errors through various improvements.

Prevention of radiation hazards to workers and the environment in the event of decommissioning nuclear power plants is a top priority. To this end, it is essential to continuously perform radiation characterization before and during decommissioning. In operating nuclear power plants, various detectors are used depending on the purpose of measurement. Portable detectors used in power plants have excellent portability, but there is a limit to the use of a single measuring device alone to quantify radioactive contamination, nuclide analysis, and ensure representation of measurement results. In foreign countries, gamma-ray visualization detectors are being actively used for operating and decommissioning nuclear power plants. KHNP is also conducting research on the development of gamma-ray visualization detectors for multipurpose field measurement at decommissioning nuclear power plants. It aims to develop detectors capable of visualizing radioactive contamination, analyzing nuclides, estimating radioactivity, and estimating dose rates. To this end, we are developing related software according to the development process by purchasing sensors from H3D, which account for more than 75% of the US gamma-ray visualization detector market. In addition, field tests are planned in the order of Wolsong Unit 1 and Kori Unit 1 with Research reactor in Gongneung-dong in accordance with the progress of development. The detector will be optimized by analyzing the test results according to various gamma radiation field environments. The development detector will be used for various measurement purposes for Kori unit 1 and Wolsong

In gamma-ray spectrometry for volume samples, the self-attenuation effect should be considered in the case of differences in chemical composition and density between the efficiency calibration source for quantitative analysis of sample and the sample actually measured. In particular, the lower the gamma-ray energy, the greater the gamma-ray attenuation due to the self-attenuation effect of the sample. So, the attenuation effect of low-energy gamma-rays in the sample should be corrected to avoid over- or under-estimation of its radioactivity. One of the most important factors in correcting the self-attenuation effect of the sample is the linear attenuation coefficient for the sample, which can be directly calculated using a collimator. The larger the size of the collimator, the more advantageous it is to calculate the linear attenuation coefficient of the sample, but excessive size may limit the use of the collimator in a typical environmental laboratory due to its heavy weight. Therefore, it is necessary to optimize the collimator size and structure according to the measurement environment and purpose. This study is to optimize a collimator that can determine the effective linear attenuation coefficient of low-energy gamma-rays, and verify its applicability. The overall structure of the designed collimator was optimized for gamma-ray energy of less than 100 keV and cylindrical plastic bottle with diameter of 60 mm and a height of 40 mm. The materials of optimized collimator consisted of tungsten. Acryl and acetal were used to form the housing of the collimator, which fixes the central axis of the bottle, collimator and point-like source. In addition, using the housing, the height of the tungsten is adjusted according to the height of the sample. For applicability evaluation of the optimized collimator, IAEA reference material in solid form were used. The sample was filled in the bottle with heights of 1, 2, 3 and 4 cm respectively. Using the collimator and point-like source of 210Pb (46.5 keV), 241Am (59.5 keV), and 57Co (121.1 keV), the linear attenuation coefficient and the radioactivity for the samples were calculated. As a result, to calculate the linear attenuation coefficient using the optimized collimator, a relatively high sample height is required. However, the optimized collimator can be used to determine the linear attenuation coefficients of low-energy gamma-rays for the self-attenuation correction regardless of the sample height. It is concluded that the optimized collimator can be useful to correct the sample selfattenuation effect.

Gamma spectrometry is one of the main analysis methods used to obtain information about unknown radioactive materials. In gamma-ray energy spectrometry, even for the same gamma-ray spectrum, the analysis results may be slightly different depending on the skill of the analyst. Therefore, it is important to increase the proficiency of the analyst in order to derive accurate analysis results. This paper describes the development of the virtual spectrum simulator program for gamma spectrometry training. This simulator program consists of an instructor module and trainee module program based on an integrated server, in which the instructor transmits a virtual spectrum of arbitrarily specified measurement conditions to the students, allowing each student to submit analysis results. It can reproduce a virtual gamma-ray energy spectrum based on virtual reality and augmented reality technique and includes analysis function for the spectrum, allowing users to experience realistic measurement and analysis online. The virtual gamma-ray energy spectrum DB program manages a database including theoretical data obtained by Monte Carlo simulation and actual measured data, which are the basis for creating a virtual spectrum. The currently developed database contains data on HPGe laboratory measurement as well as in-situ measurements (ground surface, decommissioned facility wall, radiowaste drum) of portable HPGe detectors, LaBr3(Ce) detector and NaI detector. The analysis function can be applied not only to the virtual spectrum, but also to the input measured spectrum. The parameters of the peak analysis algorithm are customizable so that even low-resolution spectra can be properly analyzed. The validity of the database and analysis algorithm was verified by comparing with the results derived by the existing analysis programs. In the future, the application of various in-situ gamma spectrometers will be implemented to improve the profiling of the depth distribution of deposited nuclides through dose rate assessment, and the applicability of the completed simulator in actual in-situ gamma spectrometry will be verified.

Gamma-ray spectroscopy, which is an appropriate method to identify and quantify radionuclides, is widely utilized in radiological leakage monitoring of nuclear facilities, assay of radioactive wastes, and decontamination evaluation of post-processing such as decommissioning and remediation. For example, in the post-processing, it is conducted to verify the radioactivity level of the site before and after the work and decide to recycle or dispose the generated waste. For an accurate evaluation of gamma-ray emitting radionuclides, the measurement should be carried out near the region of interest on site, or a sample analysis should be performed in the laboratory. However, the region is inaccessible due to the safety-critical nature of nuclear facilities, and excessive radiation exposure to workers could be caused. In addition, in the case of subjects that may be contaminated inside such as pipe structures generated during decommissioning, surveying is usually done over the outside of them only, so the effectiveness of the result is limited. Thus, there is a need to develop a radiation measurement system that can be available in narrow space and can sense remotely with excellent performance. A liquid light guide (LLG), unlike typical optical fiber, is a light guide which has a liquid core. It has superior light transmissivity than any optical fiber and can be manufactured with a larger diameter. Additionally, it can deliver light with much greater intensity with very low attenuation along the length because there is no packing fraction and it has very high radiation resistant characteristics. Especially, thanks to the good transmissivity in UV-VIS wavelength, the LLG can well transmit the scintillation light signals from scintillators that have relatively short emission wavelengths, such as LaBr3:Ce and CeBr3. In this study, we developed a radiation sensor system based on a LLG for remote gamma-ray spectroscopy. We fabricated a radiation sensor with LaBr3:Ce scintillator and LLG, and acquired energy spectra of Cs-137 and Co-60 remotely. Furthermore, the results of gamma-ray spectroscopy using different lengths of LLG were compared with those obtained without LLG. Energy resolutions were estimated as 7.67%, 4.90%, and 4.81% at 662, 1,173, and 1,332 keV, respectively for 1 m long LLG, which shows similar values of a general NaI(Tl) scintillator. With 3 m long LLG, the energy resolutions were 7.92%, 5.48%, and 5.07% for 662, 1,173, and 1,332 keV gamma-rays, respectively.

Recently, an international issue due to the discharge of contaminated water from the Fukushima has been highlighted. Since the Fukushima nuclear power plant accident in japan, marine environmental radioactivity survey has been strengthened with increased sampling frequency and range for seawater in territorial waters. And a stationary underwater radiation monitoring system including floating equipment-based system such as oceanographic buoys, tidal stations have been deployed on-site to detect abnormal radiological events. However, stationary monitoring systems may be insufficient for the early detection of abnormal radioactivity over a wide area, since it is a passive way of waiting for radioactive materials to spread in the ocean. So, our team developed a ship-mounted seawater gammaray monitoring system that can be operated remotely and in real time. In this study, it includes a detailed description of the design, installation, monitoring method, and operation of the system.

To obtain the gamma-ray energy spectrum of artificial radionuclides which is difficult to obtain practically, virtual gamma-ray energy spectrum simulator program was developed. It can be applied for the predetermined measurement condition for which the database was developed through computational simulation and actual measurement of background radiation. For gamma spectrometry training for KHNP HPGe detectors using this program, the database for KNPG HPGe detectors was developed. First, the geometry of the detector in the simulation was adjusted to resemble the real structure by comparing the actually measured net counts rate at the main gamma peak with the value simulated by MCNP6. The Certified Reference material (CRM) of 137Cs and 60Co were used for verification. The comparison was made with respect to the situation where CRM was attached to the top and side of the detection part of the considered detector. The geometry structures of detectors were simulated by reflecting the design drawing of the products, and the simulation was performed for several thicknesses of the Ge/Li dead layer in consideration of the change in the thickness over time. As the results, the simulation geometry was tuned so that the results for 137Cs showed a difference within 10% for all detectors. At this time, in some detectors, the result for 60Co shows a 10% higher error, which is estimated to be due to the random summing. It was not considered in tuning the simulation geometry, but it was found that improvements were needed to reflect the coincidence summing when construction the virtual spectrum in the future. The determined simulation geometry was applied to generate theoretical gamma-ray energy spectra of representative artificial radionuclides. In order to create a virtual spectrum similar to the real one, the background spectrum was measured for each detector without a source, and the simulation results were calculated in the form of having the same energy channel as the background spectrum. The background spectrum and theoretical spectra of artificial radionuclides for each detector were databased so that virtual spectra could be generated under desired conditions. The virtual spectrum was generated by adding a background spectrum and a spectrum obtained by multiplying the spectrum of the desired nuclide by the concentration of the nuclide. The validity of generated virtual spectra was verified using the pre-developed gamma spectrometry program. As a results of gamma spectrometry of virtual spectra, the virtual spectra was verified by showing a difference within 20% from the radioactivity value input when generating the virtual spectra.