Due to the Fukushima nuclear accident, a large amount of radioactive material was released into the atmosphere, and consequently, it spread over a wide area and was deposited into the soil. As a result of this, a wide area of radioactive contamination site was created. Due to the contaminated site, the need for research on various exploration platforms for efficient situation management and field response is being emphasized. Backpack-type radiation survey & monitoring equipment is useful for creating a contamination maps containing information such as Dose Rate, Radionuclide, Activity Concentration accompanied by spatial analysis when performing a Scan Survey that moves with a backpack on a wide area site. contamination maps are based on accurate radiological characteristic information. However, there is a problem in that the accuracy of the evaluation results is lowered due to changes in environment conditions or the variability of the dose rate and counting rate during scan survey. This problem should be solved by applying the influence of each variable to the underlying data. However, prior to this, it is most important to prepare the base underlying first. And this can be obtained through evaluation of detection performance through static survey. Therefore, in this study, the change in detection efficiency for the measurement height and radius of the backpack-type radiation survey & monitoring equipment based on the 3"×3" NaI(Ti) detector was evaluated. First, the height of the source and Backpack-type radiation survey & monitoring equipment was evaluated from 0 cm to 1 m, which is the height of the soil and detector when an adult male wears a backpack. The experiment was conducted using the 137Cs (383 kBq) point source, which is a nuclide mainly detected at the contaminated site. The measurement time was measured five times per one minute, considering that it was backpack-type equipment and a future scan survey. In addition, in order to evaluate the detection radius, the measurement was performed by changing the measurement distance up to 5 m at intervals of 50 cm. As a result of evaluating the detection performance of the backpack-type radiation surveys & monitoring equipment, it was confirmed that increasing the detection height and radius reduces the count rate in the form of an exponential function. In addition, it was confirmed that the detected radius varies depending on the height. Based on these results, we plan to conduct additional research to understand the scan survey and its sensitivity to various factors. Through this, the company plans to develop various models for exploring the site by improving the accuracy of backpack-type radiation surveys & monitoring equipment.

Attention has been paid to the source term released after Chernobyl and Three Mile Island (TMI), which were the representative accidents of nuclear power plants, and has been studied several times in order to predict and evaluate radiation source term, which can be released in the event of a virtual accident. In particular, the impact of the accident was assessed on the basis of Deterministic Safety Analysis (DSA) and after the WASH-1400, the technology of the Probabilistic Safety Assessment (PSA) was introduced, supplementing safety by taking into account the existence of uncertainty. After the Fukushima accident, a SOARCA report was published to evaluate the specific classification of each type of accident, the realistic progress of the accident, and the leakage of radioactive materials. In this paper, the evaluation methodology and results of the source term of severe accident before and after the Fukushima accident were compared, and the evaluation methods applied to domestic nuclear power plants were compared. Prior to the Fukushima accident, the behavior of the accident and source term were evaluated for Loss of Coolant Accident (LOCA), which led to design based accidents, Total Loss of Feed Water (TLOFW) followed by Station Blackout (SBO) the results were compared to Chernobyl and TMI based on the resulting data to evaluate safety and reliability. After the Fukushima accident, the Interfacing System Loss of Coolant Accident (ISLOCA) and the Steam Generator Tube Rupture (SGTR), which is classified as containment’s bypass accident, were included for predictive assessment. This is due to the analysis that the risk of cancer and early mortality are affected. MACST facilities and strategies were added to domestic nuclear power plants, and accidents with a high core damage frequency were mainly interpreted. In addition, source term was evaluated with the addition of a Basement Melt-Through (BMT) accident that had not previously been considered as a focus. As a result of the comparison of source term evaluation, accidents can be caused by a number of unidentified problems, and Korea’s experience on Level 2 and 3 has not been accumulated, making it difficult to predict the results of source term evaluation or lack of reliability.