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.

The dismantling nuclear power plant is expected to continue to change the radiation working environment compared to the operating nuclear power plant. Contamination monitors and survey meters currently in use have limitations in accurate analysis source term and dose rates for continuous changes in radiation fields at dismantling sites. Due to these limitations, the use of semiconductor detectors such as HPGe and CZT detectors with excellent energy resolution and portability is increasing. The CZT detector performs as well as the HPGe detector, but there is no proven calibration procedure yet. Therefore, in this study, the HPGe calibration method was reviewed to derive implications for the CZT detector calibration method. The operating principle of a semiconductor detector that measures gamma emission energy converts them into electrical signals is the same. Two calibrations of HPGe detectors are performed according to the standard calibration procedure for semiconductor detectors for gamma-ray measurement issued by the Korea Association of Standards & Testing Organizations. The first is an energy calibration that calculates gamma-ray peak position measurements and relational expressions using standard source term that emit gamma-rays. The channel values for energy are measured using certified reference source term to determine radionuclides by identifying channels corresponding to the measured peak energy values. The second is the measurement efficiency of measuring the coefficient calibration device, which measures gamma rays emitted from the standard source term. The detector efficiency by sample or distance is measured in consideration of the shape, size, volume, and density of the calibration device. The HPGe detector performs calibration once every six months through a verified calibration method and is being used as a source term analyzer at the power plant. The CZT detector may also establish a procedure for identifying peak positions through energy calibration and calculating radioactivity through efficiency calibration. This will be a way to expand the usability of semiconductor detectors and further monitor radiation in a more effective way.

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.

The detector response was simulated to design a fork detection system for verifying the characteristics of spent fuel. The fork detection system currently used consists of two fission chamber and an ion chamber, and it is nuclear safeguard equipment that measures the gross neutrons and gross gamma rays emitted from the spent fuel assembly to identify the characteristics of the spent fuel and verify the authenticity of the operation history. In order to improve the current fork detection system, we are developing a system that applies CZT, a room temperature semiconductor detector, and a stilbene detector, which is an organic scintillator. Depletion calculations were performed using the ORIGEN code to determine the radiological characteristics emitted from spent nuclear fuel assembly. The flux of radiation emitted from the spent nuclear fuel assembly was calculated by changing the conditions such as initial enrichment, burnup, and cooling time, which are major variables of spent fuel assembly. The calculated result is used as the source term of the particle transport code. Considering the general operating conditions of the pressurized light water reactor, the conditions were changed in the range of 3-5% for initial enrichment and 30-72 GWD/MTU for burnup, and the cooling time was given within 10 years. MCNP 6.2, a Monte Carlo simulation code, was used to simulate the detector response to radiation emitted from spent nuclear fuel assembly. According to the shape, size, and position of the CZT detector, the gamma counts incident on the detector were calculated and derived the initial design of our fork detection system.