When the parent radionuclide decays, the progeny radionuclide is produced. Accordingly, the dose contribution of the progeny radionuclide should be considered when assessing dose. For this purpose, European Commission (EC) and International Atomic Energy Agency (IAEA) provide weighting factors for dose coefficient. However, these weighting factors have a limitation that does not reflect the latest nuclide data. Therefore, in this study, we analyzed the EC and IAEA methodology for derivation of weighting factor and used the latest nuclide data from ICRP 107 to derive weighting factors for dose coefficient. Weighting factor calculation is carried out through 1) selection of nuclide, 2) setting of evaluation period, and 3) derivation based on ICRP 107 radionuclide data. Firstly, in order to derive the weighting factor, we need to select the radionuclides whose dose contribution should be considered. If the half-life of progeny radionuclides sufficiently short compared to the parent radionuclide to achieve radioactive equilibrium, or if the dose coefficient is greater of similar to that of the parent radionuclide and cannot be ignored, the dose contribution of the progeny radionuclide should be considered. In order not to underestimate the dose contribution of progeny radionuclides, the weighting factors for the progeny nuclides are taken as the maximum activity ratio that the respective progeny radionuclides will reach during a time span of 100 years. Finally, the weighting factor can be derived by considering the radioactivity ratio and branch fraction. In order to calculate the weighting factor, decay data such as the half-life of the radionuclide, decay chain, and branch fraction are required. In this study, radionuclide data from ICRP 107 was used. As a result of the evaluation, for most radionuclides, the weighting factors were derived similarly to the existing EC and IAEA weighting factors. However, for some nuclides, the weighting factors were significantly different from EC and IAEA. This is judged to be a difference in the half-life and branch fraction of the radionuclide. For example, in the case of 95Zr, the weighting factor for 95mNb showed a 35.8% difference between this study and previous study. For ICRP 38, when 95Zr decays, the branch fraction for 95mNb is 6.98×10-3. In contrast, for ICRP 107, the branch fraction is 1.08×10-2, a difference of 54.7%. Therefore, the weighting factor for the dose coefficient based on ICRP 107 data may differ from existing studies depending on the half-life and decay information of the nuclide. This suggests the need for a weighting factor based on the latest nuclide data. The results of this study can be used as a basis for the consideration of dose contributions for progeny radionuclides in various dose assessments.

In Korea, most temporary storage facilities for spent nuclear fuel are nearing saturation. As an alternative to this, the 2nd basic plan for high-level radioactive waste management specified the operation plan of dry interim storage facility. Meanwhile, the NSSC No. 2021-19 stipulates that it is necessary to evaluate the possibility and potential effect of accident before operating interim storage facility. Therefore, this study analyzed the categories of accident scenarios that may occur in dry storage facility as part of prior research on this. We investigated the case of categorization of dry storage facility accident scenarios of IAEA, NRC, KAREI, and KINS. The IAEA presented accident scenarios that could occur in on-site dry storage facility operated with silo and cask method. NRC has classified accident scenarios in dry storage facility and estimated the probability of accidents for each. KAERI and KINS selected major accident scenarios and analyzed the processes for each, in preparation for the introduction of dry storage facility in Korea in the future. Overall, a total of 10 accident scenarios were considered, and the scenarios considered by each institution were different. Among 10 scenarios, cask drop and aircraft collision were included in the categorization of most institutions. The results of this study can be used as basic data for cataloging accidents subject to safety evaluation when introducing dry interim storage facility in Korea in the future.

As nuclear power plants are operated in Korea, low and intermediate-level radioactive wastes and spent nuclear fuels are continuously generated. Due to the increase in the amount of radioactive waste generated, the demand for transportation of radioactive wastes in Korea is increasing. This can have radiological effect for public and worker, risk assessment for radioactive waste transportation should be preceded. Especially, if the radionuclides release in the ocean because of ship sinking accident, it can cause internal exposure by ingestion of aquatic foods. Thus, it is necessary to analyze process of internal exposure due to ingestion. The object of this study is to analyze internal exposure by ingestion of aquatic foods. In this study, we analyzed the process and the evaluation methodology of internal exposure caused by aquatic foods ingestion in MARINRAD, a risk assessment code for marine transport sinking accidents developed by the Sandia National Laboratory (SNL). To calculate the ingestion internal exposure dose, the ingestion concentrations of radionuclides caused by the food chain are calculated first. For this purpose, MARINRAD divide the food chain into three stages; prey, primary predator, and secondary predator. Marine species in each food chain are not specific but general to accommodate a wide variety of global consumer groups. The ingestion concentrations of radionuclides are expressed as an ingestion concentration factors. In the case of prey, the ingestion concentration factors apply the value derived from biological experiments. The predator's ingestion concentration factors are calculated by considering factors such as fraction of nuclide absorbed in gut, ingestion rate, etc. When calculating the ingestion internal exposure dose, the previously calculated ingestion concentration factor, consumption of aquatic food, and dose conversion factor for ingestion are considered. MARINRAD assume that humans consume all marine species presented in the food chain. Marine species consumption is assumed approximate and conservative values for generality. In the internal exposure evaluation by aquatic foods ingestion in this study, the ingestion concetration factor considering the food chain, the fraction of nuclide absorbed in predator’s gut, ingestion rate of predator, etc. were considered as influencing factors. In order to evaluate the risk of maritime transportation reflecting domestic characteristics, factors such as domestic food chains and ingestion rate should be considered. The result of this study can be used as basis for risk assessment for maritime transportation in Korea.

After Fukushima nuclear power plant accident in 2011, Concerns about accident of spent fuel pool increase. In Korea, the time of saturation of spent fuel pool is coming, but regulatory measures and safety evaluation are insufficient when occurring spent fuel pool accident. Thus, it is necessary to review of spent fuel pool accident in foreign countries to establish regulatory measures and safety evaluation of spent fuel pool accident suitable for domestic spent fuel pool. Therefore, we reviewed spent fuel pool accident that occurred at Fukushima Unit 4, SONGS Unit 2 and PAKS. In Japan, spent fuel pool accident occurred at Fukushima NPP in 2011. Tsunami was cause of the accident. Station Black Out occurred at Fukushima NPP and Emergency Diesel Generator lost their functions due to Tsunami. As a result, Loss of cooling happened in spent fuel pool at Fukushima NPP. For Unit 4, wall of spent fuel pool in Unit 4 was damaged due to hydrogen explosive, so loss of coolant in spent fuel pool of Unit 4 occurred. After the accident, the temperature of spent fuel pool increases to 75°C, but there was no damage to the spent fuel. In USA, spent fuel pool accident occurred at SONGS Unit 2 in 2013. The debris of nearby ocean is cause of the accident. The debris entered the system through a damaged Salt Water Cooling pump suction strainer. The debris obstructed flow through the Component Cooling Water heat exchanger and operation of Salt Water Cooling. The maximum spent fuel pool temperature during this event was 25.6°C. It was a value that satisfied the technical specifications of the SONGS NPP. In Ukraine, spent fuel pool accident occurred at PAKS in 2003. Unintentionally opened valve of cleaning tank is cause of the accident. Loss of coolant occurred in spent fuel pool of PAKS. Due to loss of coolant, spent fuels were exposed to the vapor state atmosphere, and oxidation occurred in the cladding tube of the spent fuel that rose to 1,400°C. In this study, Review of spent fuel pool accident in major foreign countries was conducted as basic studies for establishing regulatory measures and safety evaluation of spent fuel pool in Korea. Causes of each accident were different by structure of spent fuel pools. Result of this study will be contributed to establish safety measures of spent fuel pool accident suitable for domestic spent fuel pool facility.

Currently, low and intermediate-level radioactive wastes and spent nuclear fuels are continuously generated in Korea. For the disposal of the radioactive wastes, the transport demand is expected to increase. Prior to transportation, it is necessary to evaluate the radiation risk of transportation to confirm that is not high. In Korea, there is no transportation risk assessment code that reflects domestic characteristics. Therefore, foreign assessment codes are used. In this study, before developing the overland transportation risk assessment code that reflects domestic characteristics, we analyzed the radiation risk assessment methodology in transportation accident codes developed in other countries. RADTRAN and RISKIND codes were selected as representative overland transportation risk assessment codes. For the two codes we analyzed accident scenarios, exposure pathways, and atmospheric diffusion. In RADTRAN, the user classifies accident severity for possible accident scenarios, and the user inputs the probability for each accident severity. On the other hand, in the case of RISKIND, the accident scenarios are classified and the probabilities are determined according to the NRC modal study (LLNL, 1987) in consideration of the cask impact velocity, cask impact angle, and fire temperature. In the case of RISKIND, the accident scenarios are applied only to transportation of spent nuclear fuel, and cannot be defined for low and intermediate-level radioactive waste. However, in the case of RADTRAN, since the severity and probability of accidents are defined by user, it can be applied to low and intermediate-level radioactive wastes. As the exposure pathways considered in transportation accident, both RADTRAN and RISKIND consider external exposure (cloudshine and groundshine), and internal exposure (inhalation, resuspension inhalation and ingestion). In the case of RADTRAN, additionally, external exposure due to loss of shielding (LOS) is considered. Atmospheric diffusion calculation is essential to determine the extent to which radioactive materials are diffused. In both RADTRAN and RISKIND, atmospheric diffusion calculations are based on Gaussian diffusion model. Users must input Pasquill stability class, release height, heat release, wind speed, temperature and mixing height, etc. Additionally, RADTRAN can input weather information relatively simply by inputting only the Pasquill stability class fraction and selecting the US average weather option. This study results will be used as a basis for developing radioactive waste overland transportation risk assessment code that reflects domestic characteristics.

For safe management of spent nuclear fuels, they should be delivered to repository or waste disposal site. As the amount of spent nuclear fuel transportation is expected to increase in the future due to the provision of an intermediate storage facility, the necessity to secure transportation cask is emerging. In order to secure the spent nuclear fuel transportation cask, it is necessary to analyze the regulatory processes for domestic and foreign spent nuclear fuel transportation cask. In this study, the regulatory processes for domestic and foreign spent nuclear fuel transportation cask was analyzed. In this study, the IAEA, US, and Korea spent nuclear fuel transportation cask regulatory processes were analyzed. The domestic and foreign spent nuclear fuel transportation cask regulatory processes consist of design phase, manufacturing phase, and operation phase. In the design stage, the transport requirements are designed in accordance with the safety requirements of international organizations and countries. The application to be submitted when applying for approval should include a safety analysis report, evidence proving compliance with safety requirements et al. In the manufacturing stage, it is a stage to check whether the safety requirements are satisfied before the first use after manufacturing the transportation cask. Inspections include welding inspection, leakage inspection, shielding inspection, and thermal inspection. In the operation stage, it is a stage of periodically performing inspections for continuous maintenance of the package when the transportation cask is used. The inspection items to be performed are similar to the manufacturing stage and typically include performance inspection of components and leakage inspection. In this study, domestic and foreign spent nuclear fuel transportation cask regulatory processes were analyzed. It was found that the domestic and foreign spent nuclear fuel transportation cask regulatory processes consist of the design phase, the manufacturing phase, and the operation phase. The results of this study can be used as basic data for policy decision-making for the spent nuclear fuel cask.

In Korea, research on the introduction of dry storage facility is being conducted as an alternative to saturation of temporary storage facilities for spent nuclear fuel. The introduction of dry storage facilities requires a radiological impact assessment on the workers of the facility, and for this, an appropriate exposure scenario must be derived through work procedure analysis. In this study, the procedure for storing spent nuclear fuel in dry storage facilities was analyzed based on the case of evaluating the radiological impact of workers in dry storage facilities abroad. We investigated cases of radiological impact assessment on workers at on-site dry storage facilities by PNNL, Dominion, and P. F. Weck. PNNL and Dominion analyzed the storage work procedure of the VSC (Vertical Storage Cask) method using CASTOR V/21, TN-32, respectively, and conducted a radiological impact assessment. P. F. Weck analyzed the storage work procedure of various spent nuclear fuel casks for VSC and HSM (Horizontal Storage Module), conducted a radiological impact assessment. As a result of comparing the procedure for storing spent nuclear fuel by case, it was found that the storage procedure was determined by the storage method and the cask type. In the case of VSC method, canister-type casks and basket-type casks are used, and the storage procedure are partially different according to each. Canister-type cask requires repackaging from transfer overpack to storage overpack, but basket-type cask doesn’t require that procedure. In the case of the HSM method, only the canister type cask was found to be used. However, the storage procedure was different depending on the type of HSM system. Depending on the type of HSM system, the necessity of cask for on-site transport was different. In this study, we investigated and analyzed the work procedure according to the storage method of dry storage facilities abroad. It was found that the dry storage procedure of spent nuclear fuel different according to the storage method and type of cask. The results of this study can be used as basic when deriving the exposure scenario for spent nuclear fuel dry storage workers suitable for the domestic situation.