In Korea, the construction of dry storage facilities for spent nuclear fuel is being promoted through the 2nd basic plan for high-level radioactive waste management. When operating dry storage facilities, exposure dose assessment for workers should be performed, and for this, exposure scenarios based on work procedures should be derived prior. However, the dry storage method has not yet been sufficiently established in Korea, so the work procedure has not been established. Therefore, research is needed to apply it domestically based on the analysis of spent nuclear fuel management methods in major overseas leading countries. In this study, the procedure for receiving and storing spent nuclear fuel in a concrete overpack-based storage facility was analyzed. Among the various spent nuclear fuel management systems, the metal overpack-based HI-STAR 100 system and the concrete overpackbased HI-STORM 100 system are quite common methods in the United States. Therefore, in this study, work procedures were analyzed based on each final safety analysis report. First, the HI-STAR 100 overpack enters the facility and is placed in the transfer area. Remove the impact limiter of the overpack and install the alignment device on the top of the overpack. Place the HI-TRAC, an on-site transfer device, on top of the alignment unit and remove the lids of the two devices to insert the canister into the HI-TRAC. When the canister transfer is complete, reseat the lid to seal it, and disconnect the HI-TRAC from the HI-STAR 100. Raise the canister-loaded HI-TRAC over the alignment device on the top of the HI-STORM 100 overpack and remove the lids of the two devices that are in contact. Insert the canister into the HI-STORM 100 and reseat the lid. The HI-STORM 100 loaded with spent nuclear fuel is transferred to the designated storage area. In this study, the procedure for receiving and storing spent nuclear fuel in a concrete overpack-based storage facility was analyzed. The main procedure was the transfer of canisters between overpacks, and it was confirmed that HI-TRAC was used in the work procedure. The results of this study can be used as basic data for evaluating the exposure dose of operating workers for the construction of dry storage facilities in Korea.

In the wake of the Fukushima NPP accident, research on the safety evaluation of spent fuel storage facilities for natural disasters such as earthquakes and tsunamis has been continuously conducted, but research on the protection integrity of spent fuel storage facilities is insufficient in terms of physical protection. In this study, accident scenarios that may occur structurally and thermally for spent fuel storage facilities were investigated and safety assessment cases for such scenarios were analyzed. Major domestic and international institutions and research institutes such as IAEA, NEA, and NRC provide 13 accident scenario types for Spent Fuel Pool, including loss-of-coolant accidents, aircraft collisions, fires, earthquakes. And 10 accident scenario types for Dry Storage Cask System, including transportation cask drop accidents, aircraft collisions, earthquakes. In the case of Spent Fuel Pool, the impact of the cooling function loss accident scenario was mainly evaluated through empirical experiments, and simulations were performed on the dropping of spent nuclear fuel assembly using simulation codes such as ABAQUS. For Dry Storage Cask System, accident scenarios involving structural behavior, such as degradation and fracture, and experimental and structural accident analyses were performed for storage cask drop and aircraft collision accidents. To evaluate the safety of storage container drop accidents, an empirical test on the container was conducted and the simulation was conducted using the limited element analysis software. Among the accident scenarios for spent fuel storage facilities, aircraft and missile collisions, fires, and explosions are representative accidents that can be caused by malicious external threats. In terms of physical protection, it is necessary to analyze various accident scenarios that may occur due to malicious external threats. Additionally, through the analysis of design basis threats and the protection level of nuclear facilities, it is necessary to derive the probability of aircraft and missile collision and the threat success probability of fire and explosion, and to perform protection integrity evaluation studies, such as for the walls and structures, for spent fuel storage facilities considering safety evaluation methods when a terrorist attack occurs with the derived probability.

Radiation dose rates for spent fuel storage casks and storage facilities of them are typically calculated using Monte Carlo calculation codes. In particular, Monte Carlo computer code has the advantage of being able to analyze radiation transport very similar to the actual situation and accurately simulate complex structures. However, to evaluate the radiation dose rate for models such as ISFSI (Independent Spent Fuel Storage Installation) with a lot of spent fuel storage casks using Monte Carlo computational techniques has a disadvantage that it takes considerable computational time. This is because the radiation dose rate from the cask located at the outermost part of the storage facility to hundreds of meters must be calculated. In addition, if a building is considered in addition to many storage casks, more analysis time is required. Therefore, it is necessary to improve the efficiency of the computational techniques in order to evaluate the radiation dose rate for the ISFSI using Monte Carlo computational codes. The radiation dose rate evaluation of storage facilities using evaluation techniques for improving calculation efficiency is performed in the following steps. (1) simplified change in detailed analysis model for single storage cask, (2) create source term for the outermost side and top surface of the storage cask, (3) full modeling for storage facilities using casks with surface sources, (4) evaluation of radiation dose rate by distance corresponding to the dose rate limit. Using this calculation method, the dose rate according to the distance was evaluated by assuming that the concrete storage cask (KORAD21C) and the horizontal storage module (NUHOMS-HSM) were stored in the storage facility. As a result of calculation, the distance to boundary of the radiation control area and restricted area of the storage facility is respectively 75 m / 530 m (KORAD21C case), and 20 m / 350 m (NUHOMS-HSM case).

The IAEA states that in the event of sabotage, nuclear material and equipment in quantities that can cause high radiological consequences (HRC), as well as the minimum systems and devices necessary to prevent HRC, must be located within one or more vital areas. Accordingly, in Article 2 of the ACT ON PHYSICAL PROTECTION AND RADIOLOGICAL EMERGENCY, the definition of the vital area is specified, and a nuclear facility operator submits a draft to the Nuclear Safety and Security Commission to establish vital areas and must obtain approval from Nuclear Safety and Security Commission. Since the spent fuel pool and new fuel storage area are areas where nuclear material is used and stored, they can be candidates for vital areas as direct targets of sabotage. The spent fuel pool is a wet spent fuel storage facility currently operated by most power plants in Korea to cool and store spent nuclear fuel. Considering the HRC against sabotage, it is necessary to review whether sepnt fuel pool needs to establish a vital area. In addition, depending on the status of plant operation during the spent fuel management cycle, the operation status of safety systems to mitigate accidents and power system change, so vital areas in fuel handling building (including spent fuel pool) also need to be adjusted flexibly. This study compares the results of the review on whether the essential consideration factors are reflected in the identification of essential safety systems and devices to minimize HRC caused by sabotage in the spent fuel storage system with the procedure for identifying the vital area in nuclear power plants. It was reviewed from the following viewpoints: Necessity to identify necessary devices to minimize the radiation effects against sabotage on the spent fuel pool, Review of necessary elements when identifying vital areas to minimize the radiation effects of spent fuel pool against sabotage, Necessity to adjust vital areas according to the spent fuel management cycle. The main assumptions used in the analysis of the vital area of the power plant need to be equally reflected when identifying vital areas in spent fuel pool. And, the results of this study are for the purpose of minimizing the radiological consequences against sabotage on the spent fuel storage system including the spent fuel pool and used to establish regulatory standards in the spent fuel storage stage.