Currently, Japan is undertaking a nationwide project to measure and map radioactive contamination around Fukushima, as part of the efforts to restore normalcy following the nuclear accident. The Japan Atomic Energy Agency (JAEA) manages the Fukushima Environmental Safety Center, located approximately 20 km north of the Fukushima Daiichi nuclear power plant in Minamisōma City, Fukushima Prefecture. In collaboration with the JAEA, this study involved conducting comparison experiments and analyses with radiation detectors in high radiation environments, a challenging task in Korean environments. Environmental radiation surveys were conducted using three types of detectors: CZT, NaI(Tl), and LaBr3(Ce), across two contaminated areas. Dose rate values were converted using dose rate conversion factors for each detector type, and dose rate maps were subsequently created and compared. The detectors yielded similar results, demonstrating their feasibility and reliability in high radiation environments. The findings of this study are expected to be a crucial reference for enhancing the verification and supplementation of procedures and methods in future radiation measurements and mobile surveys in high-radiation environments, using these three types of radiation instruments.

In-situ gamma spectrometer with mobile equipment can be used for rapid determination of radioactivity in the environment within a very short interval. 2”×2” NaI(Tl) scintillator are used to build a mobile radiation measurement system (called as Monitoring of Ambient Radiations of KAERI for Backpack, MARK-B3) with a signal processing unit, and GPS and interface units to a PC for wireless controlling system. Development of the survey system is to measure ambient gamma-ray spectrometry for estimating ground radioactivity and radiation dose in the environment. The ambient dose rate is estimated using G-factor method. For determination of G-factor, we conducted MCNP simulations in assumptions of various incident photons into the detector system. And the scintillator was exposed to Cs-137 source in the range of 1- 300 mGy/hr. Calculated dose rates for different simulation results were compared to the irradiated dose rate to derive correction factor of G-factor. To evaluate performance of the MARK-B3, in-situ gamma spectrometry was conducted in Jeju island.

Airborne surveys are an essential analysis method for rapid response and contamination identification in the early event of a radiation emergency. On the other hand, airborne surveys are far from the ground, so it is necessary to convert the dose rate at a height of 1 m above the ground. In order to improve the accuracy of the analysis value, a lot of analysis of the measurement data is required. In our previous research, we developed MARK-A1, an instrument for rapid radiation aerial survey in high radiation environment, and MARK-M1, a multipurpose instrument for radiation detection. In this study, a large unmanned aerial vehicle (UAV) was used to conduct airborne surveys using three types of detectors in the Jeju Island environment. The NaI(Tl) detector uses one 3-inch scintillator (38 mm φ × 38 mm H), and the LaBr3 detector uses two 2-inch scintillators (25 mm φ × 25 mm H). The CZT detector uses a detector with dimensions of (15 mm × 15 mm × 7.5 mm). The UAV has a payload of 15 kg (J10, JCH systems Inc. Seoul, Korea). Three different detectors were operated at a constant height of 20 m, 30 m, and 50 m. The flight experiments were performed in the west near Jeju City. Dose rate conversion factors were used to convert the dose rate from the measured spectra, and hovering flights were conducted between 1 and 50 meters to derive altitude correction factors. In this paper, the data measured with each detector in the same area were compared and the differences were derived.

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.

“Radiation measurement situation virtual reality program” is a three-dimensional modeling program. It is a program that virtually implements the measurement situation by applying visualization techniques such as animation effect charts, and effectively delivers gamma radiation energy spectrum data. This program was developed to respond to various measurement situations by visualizing major analysis objects such as radiation detectors, radioactive waste drums, radioactive building structures, and the ground surface as a 3D model. User-friendliness was secured by supporting various control functions such as distance, size, and angle while checking the three-dimensionally produced detector and analysis target in a virtual space. By using high-resolution photos obtained through 360-degree shooting, a virtual space was implemented to approximate the actual situation, such as the ground surface measurement. In addition, data communication safety was secured so that a large number of users could use it through a local area network in consideration of the actual operating environment.

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.