Biofilms are complexly structured communities of microorganisms composed of surface-attached microorganisms, where their effects on the host have been controversial. In this study, we investigated the potential biofilm-forming capacity of Lacticaseibacillus rhamnosus LRH020 (DSM25568) by detecting genes known to promote biofilm formation. It was shown that the aggregation substance gene (asa 1) was presented in the LRH020 strain. Therefore, we investigated the phenotypic activities of the gene asa1 via two methods: biofilm formation and autoaggregation activity. It was shown that the strain LRH020 had significantly less ability to form biofilm compared to the positive control strain Enterococcus faecalis ATCC 19433. Furthermore, LRH020 exhibited biofilm-forming activity comparable to Lacticaseibacillus rhamnosus GG (LGG), widely used probiotics. The auto-aggregation activity of LRH020 was also within the safe range similar to that of LGG. In conclusion, this study shows that both biofilmforming and auto-aggregation activities of the LRH020 are comparable to one of the most studied probiotics strains, LGG.

Probiotics are live microorganisms that confer health benefits onto the host when administered at adequate doses. Most widely used probiotics, such as lactobacilli and bifidobacteria, are known to be elements of healthy gut microflora and hence are not considered a threat to the host. However, probiotics may pose a risk in certain populations with compromised immune systems or defects in gut barrier functions. Herein, we evaluated the safety of Bifidobacterium breve BB077, according to the safety evaluation guidelines for probiotics produced by the National Institute of Food and Drug Safety Evaluation (NIFDS). The results show that B. breve BB077 is both non-hemolytic and non-cytolytic. In contrast, B. breve BB077 exhibited higher streptomycin and tetracycline resistance than the suggested NIFDS standard cut-off values. Hence, a genetic analysis of the streptomycin and tetracycline resistance genes was performed to determine the origin of antimicrobial resistance. Streptomycin and tetracycline resistance was shown have arisen from chromosomal mutations and considered intrinsic to the taxonomic group. In conclusion, the B. breve BB077 strain might be safe for human consumption.

In Korea, 483,102 assemblies of spent fuel have been discharged and stored in sites, as of 2019. However, total capacity for site storage is 529,748 assemblies, and more than 90% is already saturated. Wolsong site, the most saturated site, started to construct more dry storage to extend the capacity in 2020. Spent fuel and high-level waste (HLW) is a big concern in Korean nuclear industry. Then, master plan for management of spent fuel is once announced by Ministry of Trade, Industry and Energy (MOTIE) in 2016 and reviewed by civil committee in 2019. The core contents of the plan are establishing schedule for construction of HLW management facility in one area, and construction of temporary dry storage in each site, if unavoidable. For HLW management facility, there are three following schedules: siting of Underground Research Laboratory (URL) and Interim Storage by 2020, operation of facilities initiated by 2030, and operation of final disposal facility initiated by 2050. Final repository will be designed as deep geological repository. The concept of deep geological disposal is that spent nuclear fuel is placed in disposal containers that can withstand corrosion and pressure in long-term, permanently isolated from the human sphere of life, and dumped in deep geological media, such as crystalline rocks and clay layer, at a depth of 300 to 1,000 meters underground. The safety assessment of waste disposal sites focuses on determining whether the disposal sites meet the safety requirements of national regulatory authority. This safety assessment evaluates the potential radiation dose of radionuclides from the disposal site to humans or the environment. In this case, the calculation is performed assuming that all engineering barriers of the disposal site have collapsed in a long-term period. Then radionuclides are released from the waste, and migrated in groundwater. The dose resulting from the release and migration of radionuclides on the concentration of nuclides in groundwater. In general, metallic nuclides may exist in water in various ionic states, but some form colloids. This colloid allows more nuclides to exist in water than in solubility. Therefore, more doses may occur than we know generally predict. To determine the impact of colloids, we performed the safety assessment of the Yucca Mountain repository as an example.

We conducted multi-elements determination of reference material certified by the Inorganic Ventures, IV-26, using iCAP 7400 ICP-OES of Thermo Fisher Scientific. And we statistically evaluated analysis results by introducing the in-house proficiency evaluation method implemented at the Ministry of Food and Drug Safety. Ca, Co, Fe, Mg, Ni, and V were selected as target elements, and extended uncertainty was estimated at a confidence level of about 95% and coverage factor k = 2. Five parameters incurred at manufacturing process (standard solution, calibration curve, repeated measurement and dilution factor of the test sample) were considered when determining the uncertainty. En-score can be calculated using the formula En=(x-X)/(Ulab 2+Uref 2)1/2 described in KS Q ISO 13528, where x, Ulab, X, and Uref are the test results, the uncertainty of the result, and the certified value and the uncertainty of the value. And if the absolute value |En| is less than 1, it can be evaluated as a satisfied value. As a result of ICP-OES analysis, each concentration of the elements to be measured was almost similar to the certified concentration of the reference material, and the uncertainty was slightly different. Also since evaluation on multi-elements determination had an En-score within 1, it was confirmed that the analysis results satisfied En evaluation.

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.

n this research, the dose rate was measured using backpack-type scan survey device at 4 sites on Jeju Island, and the radioactivity ratio for each nuclide was evaluated using an high-purity germanium (HPGe) detector. As a result of measuring the dose rate with a backpack-type scan survey device, the average dose rate was the lowest in the measurement site 3 at 0.049 Sv/h, and the highest in the measurement site 1 at 0.066 Sv/h. The average dose rate of the 4 sites on Jeju Island was 0.056 Sv/h, and the dose rate on Jeju Island was lower than that of other regions. The data acquired by scan survey were interpreted using classed post and gridding function of surfer program. The radioactivity ratio of each nuclide in the gamma spectrum measured by the HPGe detector was the highest for K-40 with an average of 87.62%, and the lowest for Pb-214 with an average of 0.63%. In the case of the Jeju Island site, Cs-137 was detected, and the average radioactivity ratio of Cs-137 was 3.27%, which was the background level. The results of this research can be used as basic data on the radioactivity ratio for each nuclide and dose rate at the Jeju Island site. Further studies on the assessment of dose rates and radioactivity ratios in other regions are needed.

Cement is widely used as representative industrial material. In Korea, about 50 million tons of cement are consumed every year. In the manufacture of cement, raw materials containing NORM such as fly ash and bauxite are used. Therefore, the workers can be subjected to radiation exposure. The major exposure pathway in NORM industries is internal exposure due to inhalation of aerosol. Internal radiation dose due to aerosol inhalation varies depending on physicochemical properties of the aerosol. Therefore, the objective of this study was to investigate aerosol properties influencing inhalation dose in cement industries. In this study, aerosol properties were measured for two cement manufacturers. A particulate size distribution and concentration at various processing areas in cement manufacturing industries in Korea were analyzed using a cascade impactor. The mass density of raw materials and byproducts were measured using pycnometer. Shape of particulates was analyzed using SEM. The radioactivity concentration of Ra-226, Ra-228 for U/Th decay series was measured using HPGe. Particulate concentration by size was distributed log-normally with maximum at particle size about 7.2 μm in manufacturer A and 5.2 μm in manufacturer B. The mass density of fly ash and cement were 2.3±0.06, 3.2±0.02 g/cm3 respectively in manufacturer A. In manufacturer B, the mass density of bauxite and cement were 3.4±0.02, 2.9±0.01 g/cm3 respectively. The shape of particulates appeared as spherical shape in manufacturer A and B regardless of sampling area. Thus, a shape factor of unity could be assumed. The radioactivity concentrations of Ra-226, Ra-228 were 82±9, 82±8 Bq/kg for fly ash, and 25±4, 23±3 Bq/kg for cement in manufacturer A. In manufacturer B, the radioactivity concentrations of Ra-226, Ra-228 were 344±34, 391±32 Bq/kg for bauxite, and 122±13, 145±12 Bq/kg for cement. The radioactivity concentrations of Ra-226, Ra-228 in cement were less than raw materials such as fly ash and bauxite. It is because the dilution of the radioactivity concentration occurred during mixing with other raw materials in cement production process. This study results will be used as database for accurate dose assessment due to airborne particulate inhalation by workers in cement industries.

This study was performed to assess the cosmic-ray effect caused by altitude in the aerial gammaray measurement. For the gamma-ray measurement experiment by altitude, the aerial survey system composed of four 4×4×16 inches large volume NaI (Tl) detectors was used. The aerial survey system was installed in a rotor-craft to stably keep its flight altitude and position. In addition, in order to avoid to time-dependent shielding effects with the amount of fuel, a rotor-craft of which the fuel tank is not located beneath the cabin floor was selected. In this study, the ROI (Region Of Interest) was set to the 3~6 MeV range to assess the cosmic-ray contribution to the gamma-ray spectrum that could ignore the contribution of the dominant natural radionuclides. The gamma-ray spectra measured inside and outside of the rotor-craft on the ground were compared to evaluate the shielding effects of the aircraft body. As a result, the count rate of the 40K photo peak was decreased by about 10% when measuring the inside compared to the outside. On the other hand, the total count rate of the 3~6 MeV region was decreased by about 0.7% under the same condition. Therefore, the aircraft body effect was insignificant in 3~6 MeV region considering the relative uncertainty of 0.04~0.78% (1σ). In addition, the count rate in the 3~6 MeV range according to altitude was evaluated to assess the cosmic-ray effect. In order to evaluate the change in the ROI count rate according to the altitude, the gamma-ray spectrum was measured in the range of 300~2,000 m above the sea to avoid the effect of terrestrial radiation. As a result, the relationship between altitude and count rate in the 3~6 MeV range showed a high correlation with the R2 value of 0.99, when the approximate equation was derived in the form of a quadratic polynomial. Also, the count rate of 3~6 MeV at 50~500 m above the ground was estimated using the correlation equation, and this value was compared with the measured count rate. As a result of comparing the average value of estimated count rate and measured count rate, the relative difference is less than 2%. Considering the relative uncertainty of 0.78~4.11% (1σ), it was possible to evaluate the count rate of the 3~6 MeV region relatively accurately. The results of this study could be used for further study on background dose corrections in aerial survey.

The IAEA recommended considerations for exemption regulations of consumer products containing greater amounts of radioactive isotopes than the amounts specified for generic exemption. One of the major considerations is the expected exposure dose should be less than 10 μSv/y and 1 mSv/y for general cases and low probability cases, respectively, in all predictable scenarios. Under this recommendation, many countries evaluated the radiation dose for exposure scenarios of various products in consideration of the national circumstances and, then, established their own specific exemption regulation. In Republic of Korea, the “Regulation on substances excluded from radioactive isotopes” was legislated to specify consumer products excluded from regulation. However, as the usage status and product specifications has changed over time, it is necessary to periodically verify the validity of the regulation criteria in the view of exemption justification. In this study, we developed the use and disposal scenarios in consideration of the domestic use of thorium-containing gas mantle and evaluated radiation dose of each scenario accordingly. The gas mantles are used as a wick for gas lanterns and the maximum activity of natural thorium contained among the currently available gas mantles is 12.5 kBq. Radioactive isotopes in the decay chain of natural thorium can be divided into three groups according to their physical characteristics, and exposure routes suitable for each group were considered in dose calculation. Currently, most gas mantles are installed in camping lanterns. Therefore, we developed use scenarios related to camping. The average number of camping trips and time spent at the campground were set by the data from Korea Tourism Organization. Tent sizes and vehicle specifications were determined by referring to surveys and products in Korea. The used gas mantle is disposed of in a garbage bag for general waste and transported to landfill or incinerator. We determined the amount of gas mantle discarded in landfill and incinerator by the data from Korea Environment Corporation. The exposure time and amount handled by an individual were determined by considering the number of waste collection vehicles, landfills, and incinerators. Although we assumed the maximum activity of the gas mantle for conservative evaluation, the calculated radiation doses for the use and disposal scenarios were below the general requirement (i.e., 10 μSv/y) in all scenarios.

In general, after the decommissioning of nuclear facilities, buildings on the site can be demolished or reused. The NSSC (Nuclear Safety and Security Commission) Notice No. 2021-11 suggests that when reusing the building on the decommissioning site, a safety assessment should be performed to confirm the effect of residual radioactivity. However, in Korea, there are currently no decommissioning experiences of nuclear power plants, and the experiences of building reuse safety assessment are also insufficient. Therefore, in this study, we analyzed the foreign cases of building reuse safety assessment after decommissioning of nuclear facilities. In this study, we investigated the Yankee Rowe nuclear power plant, Rancho Seco nuclear power plant, and Hematite fuel cycle facility. For each case, the source term, exposure scenario, exposure pathway, input parameter, and building DCGLs were analyzed. In the case of source term, each facility selected 9~26 radionuclides according to the characteristics of facilities. In the case of exposure scenario, building occupancy scenario which individuals occupy in reusing buildings was selected for all cases. Additionally, Rancho Seco also selected building renovation scenario for maintenance of building. All facilities selected 5 exposure pathways, 1) external exposure directly from a source, 2) external exposure by air submersion, 3) external exposure by deposited on the floor and wall, 4) internal exposure by inhalation, and 5) internal exposure by inadvertent ingestion. For the assessment, we used RESRAD-BUILD code for deriving building DCGLs. Input parameters are classified into building parameter, receptor parameter, and source parameter. Building parameter includes compartment height and area, receptor parameter includes indoor occupancy fraction, ingestion rate, and inhalation rate, and source parameter includes source thickness and density. The input parameters were differently selected according to the characteristics of each nuclear facility. Finally, they derived building DCGLs based on the selected source term, exposure scenario, exposure pathway, and input parameters. As a result, it was found that the maximum DCGL was 1.40×108 dpm/100 cm2, 1.30×107 dpm/100 cm2, and 1.41×109 dpm/100 cm2 for Yankee Rowe nuclear power plant, Rancho Seco nuclear power plant, and Hematite fuel cycle facility, respectively. In this study, we investigated the case of building reuse safety assessment after decommissioning of the Yankee Rowe nuclear power Plant, Rancho Seco nuclear power plant, and Hematite fuel cycle facility. Source terms, exposure scenarios, exposure pathways, input parameters, and building DCGLs were analyzed, and they were found to be different depending on the characteristics of the building. This study is expected to be used in the future building reuse safety assessment after decommissioning of domestic nuclear power plants. This work was

In Korea, the NUREG-0017 methodology based on realistic model for reactor coolant concentrations are used to estimate the annual radioactive effluent releases for normal operation of nuclear power plant. The realistic model to estimate the radionuclide concentrations in reactor coolant is formulated as a standard, ANSI/ANS-18.1. This standard has provided a set of the reference radionuclide concentrations and adjustment factors for estimating the radioactivity in the principal fluid systems of target plant. Since ANSI/ANS-18.1 was first published in 1976, it was revised in 1984, 1999, 2016, and most recently in 2020. Therefore, this study analyzed revision history of assessment methodology of radioactive source term of light water reactors, which is ANSI/ANS-18.1. Assessment methodology of radioactive source term given ANSI/ANS-18.1 is by using radionuclide concentrations for reactor coolant and steam generator fluid of the reference plant and adjustment factors, which is modifying radioactive source term according to differences in design parameters between reference plant and target plant. There are three type of reference plant: PWR with u-tube steam generator, PWR with once-through steam generator, and BWR. This study analyzed for PWR with u-tube steam generator. Although the standard was revised, evaluation methodology and formula of adjustment factor have been retained, but some of items have been revised. First revision item is reduction of the number of radionuclides and decrease of radioactive concentration in reactor coolant. In the 1976 version of the standard, there were 71 target radionuclides, but the target nuclides have reduced to 57 in 1984 and 56 after 1999. In the case of radioactive concentration in reactor coolant, as the version of standard was updated, the radioactive concentration of 18 nuclides in 1984, 14 nuclides in 1999, and 25 radionuclides in 2016 was decreased. Most of the radionuclides with decrease radioactivity concentration were fission product, it is resulted from improvement of nuclear fuel performance. Second revision item is change of adjustment factors. After the revision in 2016, the adjustment factors for zinc addition plants using natural or depleted zinc are changed. This study analyzed revision history of evaluation methodology of radioactive source term of light water reactors. Furthermore, result of this study will be contributed to the improvement of understanding of assessment methodology and revision history for the radioactive source term.

The U.S. Nuclear Regulatory Commission (NRC) states that every environmental report prepared for the licensing stage of a Pressurized Water Reactor shall contain a statement concerning risk during the transportation of fuel and radioactive wastes to and from the reactor. Thus, the licensee should ensure that the radiological effect in accidents, as well as normal conditions in transport, do not exceed certain criteria or be small if cannot be numerically quantified. These are specified in 10 CFR Part 51 and applied in NUREG-1555 Supplement 1 Revision 1, which deals with Environmental Standard Review Plan. Corresponding regulations in Korea would be the Nuclear Safety and Security Commission Notice No. 2020-7. In Appendix 2 of the Notice, guides on the radiological environmental report for production and utilization facilities, spent nuclear fuel interim storage facilities, and radioactive waste disposal facilities. In this guide, unlike the regulations in the U.S., there are no obligations for radiological dose assessment for workers and public during the transportation. Therefore, overall regulations and their legal basis related to risk assessment during transportation conducted for the environmental report in the U.S. were analyzed in this study. On top of that, through the comparison with regulations in Korea, differences between the two systems were figured out. Finally, this study aims to find the points in terms of assessing transport risk to be revised in the current regulatory system in Korea.

A radioactive waste disposal facility needs to be developed in a way to protect present and future generations and its environment. A safety assessment is implemented for normal and abnormal scenarios and human intrusion scenarios as a part of a safety case in developing a disposal facility for the radioactive waste. The human intrusion scenarios include a well scenario which takes into account various potential exposure groups (PEGs) who use a groundwater well contaminated with radionuclides released from the disposal facility. It is observed that a pumping rate has a negative correlation with the biosphere dose conversion factor (BDCF) in the well scenario. C-14 is shown to be a key radionuclide in the well scenario, and a special model based on the carbon cycle is applied for C-14. For Tc-99, an adsorption coefficient should be adjusted to be suitable for the site. The safety assessment for the radioactive waste disposal facility is successfully carried out for the well scenario. However, it is observed that site-specific models needs to be developed and sitespecific input data need to be collected in order to avoid unnecessary conservatism.

Glass fiber (GF) insulation is a non-combustible material, light, easy to transport/store, and has excellent thermal insulation performance, so it has been widely used in the piping of nuclear power plants. However, if the GF insulation is exposed to a high-temperature environment for a long period of time, there is a possibility that it may be crushed even with a small impact due to deterioration phenomenon and take the form of small particles. In fact, GF dust was generated in some of the insulation waste generated during the maintenance process. In the previous study, the disposal safety assessment of GF waste was performed under the abnormal condition of the disposal facility to calculate the radiation exposure dose of the public residing/ residents nearby facilities, and then the disposal safety of GF waste was verified by confirming that the exposure dose was less than the limit. However, the revised guidelines for safety assessment require the addition of exposure dose assessment of workers. Therefore, in this study, accident scenarios at disposal facilities were derived and the exposure dose to the workers during the accident was evaluated. The evaluation was carried out in the following order: (1) selection of accident scenario, (2) calculation of exposure dose, (3) comparison of evaluation results with dose limits, and confirmation of satisfaction. The representative accident scenarios with the highest risk among the facility accident were selected as; (a) the fire in the treatment facility, (b) the fire in the storage facility, and (c) fire after a collision of transport vehicles. The internal and external exposure doses of the worker by radioactive plume were calculated at 10m away from the accident point. In evaluation, the dose conversion factors ICRP-72 and FGR12 were used. As a result of the calculation, the exposure dose to workers was derived as about 0.08 mSv, 0.20 mSv, and 0.10 mSv, due to fire accidents (vehicle collision, storage facilities, treatment facilities). These were 0.2%, 0.4%, and 0.2% of the limit, and the radiation risk to workers was evaluated to be very low. The results of this study will be used as basic data to prove the safety of the disposal of GF waste. The sensitivity analysis will be performed by changing the radiation source and emission rate in the future.

The change of surface environments (e.g., climate change, uplift/subsidence, and erosion) can undermine the long-term safety of a high-level radioactive waste repository. Therefore, understanding the water cycle between atmosphere, surface, and subsurface is essential to ensure the long-term safety of deep geological disposal and consequently to gain public acceptance for the repository. Among hydrologic components (e.g., precipitation, interception, runoff, infiltration, evapotranspiration (ET), and recharge) which constitute the water cycle, ET is more than half of the total precipitation and plays a crucial role in the water and energy transfer among the three systems. Although various methods for ET evaluation (e.g., Bowen Ratio, Eddy Covariance, Optical Scintillation, and Weighing Lysimeter methods) have been developed, many influential factors such as vegetation, climate, and moisture content make its accurate evaluation still tricky. In this work, we chose weighing lysimeter and Penman-Monteith methods for direct/indirect estimation of ET, and installed a smart field lysimeter and a micro-meteorological station around KAERI Underground Research Tunnel. Water balance in the unsaturated zone and five climatic variables (air temperature, humidity, precipitation, radiation, and wind speed/direction) were measured more than once per 10 minutes for six months from April to September, 2022. From the measurements, daily actual and potential ET values at the study site were calculated and compared. We also discussed the applicability and limitation of current methods and ET assessments at different spatial scales regarding verifying and validating the developing numerical models.

For the geological risk assessment of deep-depth underground condition by excavation work or tunneling, since rocks and geologic structure of each country is different, it is necessary to objectify or classify quantitatively deep-depth underground risk evaluation in accordance with Korean geologic characteristics. It could be summarized major factors of rock failure and underground space deformation by geological and geotechnical features as geologic structures, overburden, rock mass characteristics, groundwater, high stress and additional categories. Induced main factors that could be identified and predicted intermediate to deep-depth underground risk through literature investigation and analysis study on research trend related to the underground geological engineering. In order to assessment the risk of rock mass excavated from 100 m or more to several kilometers deep below the ground are classified into about 19 factors, and can be divided into 6 categories. Using these risk factors as basic data, weights for each factor for each category can be set, and further, the risk of excavated rock mass can be calculated.

The saturation rates of the spent fuel (SF) wet storage at the Kori Nuclear Power Plant (NPP), Hanbit, and Hanul are 83.3%, 74.2%, and 80.8% as of the fourth quarter of 2021. The storages of Kori NPP and Hanbit NPP are expected to be saturated in 2031, and Hanul is expected to be saturated in 2032. Therefore, the construction of an interim storage facility to store the SF temporarily stored in the NPP was planned, and preparations for the safe transport of the SF are required. In this paper, radiological preliminary assessment using NRC-RADTRAN in the process of sea transport of SF from the wet storage or ISFSI of the Hanbit NPP to the optional interim storage facility was performed. Since domestic SF transport vessels are not currently in operation, the specifications of the UK Pacific Grebe vessel which can carry up to 20 casks were used. The transport cask used the specifications of KORAD-21, a transport container developed in Korea. Because it can carry more SF assemblies than the existing KN-18. In addition, a land transport safety test was conducted in 2020 and a sea transport test is planned. The sea transport route was entered by referring to the transport route of domestic low and intermediate level waste. The accidents rate was calculated using statistics on maritime accidents from 2017 to 2021. The probability accidents along the transportation route were evaluated as 3.152E -10. When transporting to an interim storage facility, the SF expected to be the main transport target was selected as WH 17X17, combustion 45,000 MWD/MTU, and concentration of 4.5%. The source term was calculated and entered according to this data and the release fraction was entered with reference to the DOE report. In the case of normal transport without accident, the individual dose of the crew member and public residents were estimated to be 0.0525% and 0.000492% of the annual limit of 1 mSv/yr for the general public. Under the accident conditions, the annual individual doses of residents were 0.0011%, 0.0023%, 0.0034%, and 0.0046% of the annual limit of 1 mSv/yr when carrying 5, 10, 15, and 20 casks. Currently, the site of the interim storage facility has not been precisely determined, but a preliminary radiation assessment through sea transport resulted in a significantly lower than the limit. Combined scenario sea transport followed by land transport will be carried out in the next stage of study.

Maintaining fuel sheath integrity during dry storage is important. Intact sheath acts as the primary containment barrier for both fuel pellets and fission products over the dry storage periods and during subsequent fuel handling operations. In KNF, in-house fuel performance code was developed to predict the overall behavior of a fuel rod under normal operating conditions. It includes the analysis modules to predict temperature, pellet cracking and deformation, sheath stress and strain at the mid-plane of the pellet and pellet-pellet interfaces, fission gas release and internal gas pressure. The main focus of the code is to provide information on initial conditions prior to dry storage, such as fission gas inventory and its distribution within the fuel pellet, initial volumes of storage spaces and their locations, radial profile of heat generation within the pellet, etc. To upgrade the developed code that address all the damage mechanisms, the first step was a review of the available technical information on phenomena relevant to fuel integrity. Potential degradation mechanisms that may affect sheath integrity of CANDU spent fuel during dry storage are: creep rupture under internal gas pressure, sheath oxidation in air environment, stress corrosion cracking (SCC), delayed hydride cracking (DHC), and sheath splitting due to UO2 oxidation for a defective fuel. The failure by creep rupture, SCC or DHC is in the form of small cracks or punctures. The failure by sheath oxidation or sheath splitting due to UO2 oxidation results in a gross sheath rupture. The second step was to examine the technical bases of all modules of the in-house code, identify and extend the ranges of all modules to required operating ranges. This step assessed the degradation mechanisms for the fuel integrity. The objective of this assessment is to predict the probability of sheath through-wall failure by a degradation mechanisms as a function of the sheath temperature during dry storage. Further improvements being considered include upgrades of the analysis module to achieve sufficient accuracy in key output parameters. The emphasis in the near future will be on validation of the inhouse code according to a rigorous and formal methodology. The developed models provide a platform for research and industrial applications, including the design of fuel behavior experiments and prediction of safe operating margins for CANDU spent fuel.

Currently, as the saturation capacity of wet storage pool for spent nuclear fuel (SNF) of PWR in Korea has reached approximately 75%, Dry Storage Facilities (DSF) are necessary for sustainable operation of nuclear power plants. It is necessary to develop acceptance requirements for the delivery of SNF from reactor storage site to Centralized DSF. To do this end, the mechanical integrity of SNF is directly related to its repacking, retrieving, and transporting/handling performances. And also, this integrity is a key factor associated with the criticality safety that is connected to the damaged status of SNF. According to the NUREG/CR-6835, the NRC expects that the potential for nuclear fuel failures will increase because of the increase of the fuel discharge burnup and the degradation of fuel and clad material properties. Due to such damages and/or degradation, the fuel rods in the fuel assembly may be extracted and empty for following treatments (transportation, storage, handling etc). This condition can have a detrimental effect on the criticality safety of SNF. Thus, this study investigated whether extracted and empty of damaged SNF rod affects criticality safety. In this analysis, it is assumed that up to four fuel rods are missed. As a result of the analysis, As the number of fuel rods miss up to a certain number, the value of multiplication factor value of the fuel assembly increases. In addition, since the fuel rods located at the outermost layer contained relatively less fissile material than the fuel rods located center of the lattice, and neutrons were lost by the absorption material, the effective multiplication factor value gradually decreased. Nevertheless, the criticality safety was assessed to be maintained.