In this study, radioactivity of Cs-134, Cs-137, and Eu-154, which are gamma-emitting nuclides among fission products of spent fuel, was analyzed as a tool to measure the burnup of spent fuel nondestructively. This nuclide has a unique gamma-ray energy, allowing the amount of the isotope to be estimated based on the intensity of the gamma-ray at a specific energy. The SCALE 6.2 ORIGAMI (ORIGen AsseMbly Isotopics) module and the latest ORIGEN-arp library were used for computational analysis. The spent fuel samples were selected as WH14×14 with an enrichment of 1.5~5.0wt%, a burnup of 10~60 GWD/MTU, and a cooling time of 0~40 years. The analysis results were benchmarked using SFCOMPO experimental data provided by OECD/ NEA, including isotope inventory and uncertainty measured by destructive radiochemical analysis, fuel assembly design data required for benchmark model development, reactor design information, and operating history information. 16 similar spent fuels were selected from SFCOMPO data and the calculation results of Cs-134, Cs-137, and Eu-154 were compared. The average error of the Cs-134 radioactivity calculation result was 2.81%, and the maximum error was 6.70%. The average errors of Cs-137 and Eu-154 were 2.42% and 4.95%, respectively, and the maximum errors were 5.40% and 14.91%, respectively.

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).