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.

Water electrolysis is an efficient method to enrich heavy hydrogen isotopes (tritium and deuterium) in the aqueous phase. Although an alkaline water electrolyzer has been commercialized for mass production of hydrogen, such a method requires additional purification steps to remove electrolytes from the final concentrates. On the other hand, proton exchange membrane water electrolysis (PEMWE) does not require additional electrolyte treatment steps, and PEMWE is operated at higher current density compared to the alkaline water electrolysis. In this study, we investigated deuterium and tritium separation from light water by PEMWE. Separation behaviors at the anode and cathode were analyzed, and H/D and H/T separation factors were compared.

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

The 2-round Delphi survey and Focus Group Interview (FGI) survey method, in this study, are sequentially applied for the level analysis of the high-level radioactive waste (HLW) management technologies, that are classified into transport/storage, site evaluation, and disposal categories. The 2- round Delphi survey was conducted on domestic 56 experts in the HLW field in Korea, and survey answers were managed with questionnaires distributed by e-mail. In the FGI survey, domestic 24 experts from management field were formed into three groups to conduct in-depth interviews. Past research achievements including journal papers, intellectual properties and the expert opinions presented at expert hearing on HLW technology were used as reference materials. As a result of the survey, in this study, the average domestic technology level compared to the leading countries was 83.1% in transport area, 79.6% in storage area, 62.2% in site evaluation area, and 57.4% in disposal area, respectively. When compared to the former level analysis results in 2017, technology level of transport-storage area increased by 8.6%, and the site evaluation-disposal technology area decreased by 7.27%. The highest factor that increase the level of technology in the transport-storage field was due to the increased R&D program resulting on journal papers, intellectual properties. In addition, the decrease factor in the level of technology in the site evaluation-disposal field was mainly due to relatively low R&D program when compared to the leading countries. Suggested method for the level survey can be used to find out the basic data of the lower tech technologies, to estimate the efficient research budgets and to prepare the R&D human resources. With this regards, R&D roadmap can be matured with suggested prediction method for the domestic technology level on HLW.

The analysis of uranium migration is crucial for the accurate safety assessment of high-level radioactive waste (HLW) repository. Previous studies showed that the migration of the uranium can be affected by various physical and chemical processes, such as groundwater flow, heat transfer, sorption/ desorption and, precipitation/dissolution. Therefore, a coupled Thermal-Hydrological-Chemical (THC) model is required to accurately simulate the uranium migration near the HLW repository. In this study, COMSOL-PHREEQC coupled model was used to simulate the uranium migration. In the model, groundwater flow, heat transfer, and non-reactive solute transport were calculated by COMSOL, and geo-chemical reaction was calculated by PHREEQC. Sorption was primarily considered as geo-chemical reaction in the model, using the concept of two-site protolysis nonelctrostatic surface complexation and cation exchange (2 SP NE SC/CE). A modified operator splitting method was used to couple the results of COMSOL and PHREEQC. Three benchmarks were done to assess the accuracy of the model: 1) 1D transport and cation exchange model, 2) cesium transport in the column experiment done by Steefel et al. (2002), and 3) the batch sorption experiment done by Fernandes et al. (2012), and Bradbury and Baeyens (2009). Three benchmark results showed reliable matching with results from the previous studies. After the validation, uranium 1D transport simulation on arbitrary porewater condition was conducted. From the results, the evolution of the uranium front with sequentially saturating sites was observed. Due to the limitation of operator splitting method, time step effect was observed, which caused the uranium to sorbed at further sites then it should. For further study, 3 main tasks were proposed. First, precipitation/ dissolution will be added to the reaction part. Second, multiphase flow will be considered instead of single phase Darcy flow. Last, the effect of redox potential will be considered.

Although the effect of elevated carbon dioxide (CO2) on Phalaenopsis plant flowering, biomass, and photosynthesis has received intensive study, whether elevated CO2 affects plant requirements and sensitivity to potassium sulfate (SOP) during the reproductive growth stage remains unclear. To evaluate the combined effect of CO2 and SOP provision on crassulacean acid metabolism orchids, we cultivated Phalaenopsis Queen Beer ‘Mantefon’ under ambient and elevated CO2 treatments (≈ 400 or ≈ 720 μmol×mol-1, respectively) and four levels of SOP supply for 20 weeks after treatments (WAT): potassium and sulfate levels by 10.41 and 1.96 mmol·L-1 (SOP1), 5.98 and 0.90 mmol·L-1 (SOP2), 12.80 and 1.96 mmol·L-1 (SOP3), and 14.83 and 3.16 mmol·L-1 (SOP4), respectively. The number of floral buds and flowers decreased in the plants grown under elevated CO2 than in those grown under ambient CO2, regardless of the SOP level; however, the reduced production of floral buds and flowers did not affect the dry mass of shoot, root, and spike at 20 WAT. There were significant interactive effects of CO2 and SOP on root biomass accumulation and net CO2 uptake. The stimulation of biomass partitioning on the root, as a sink source, observed due to the uptake of elevated CO2 was improved under increased SOP supply. Under ambient CO2, the leaf critical SOP level was SOP1 for root and spike biomass accumulation. Plants grown under elevated CO2 were more sensitive to SOP treatments, with higher essential leaf levels of SOP.

This study investigated changes of milk production in dairy cows intramuscularly injected with drugs containing dexamethasone (DXM). Three types of dexamethasone formulations (Bueunde® (DXM 0.5 mg/mL), Dexason INJ.® (DXM 1 mg/mL) and Dexolone-20 inj.® (DXM 1 mg/mL)) were intramuscularly injected into sixteen healthy dairy cows each. Bueunde® was intramuscularly injected into 8 dairy cows with 5 mg (BED-1) and 10 mg (BED-2) of DXM once a day for 3 consecutive days, respectively. Dexason INJ.® was intramuscularly administered once into dairy cows with 20 mg (DXS-1, n=8) and 40 mg (DXS-2, n=8) of DXM, respectively. Dexolone-20 inj.® was intramuscularly injected once into dairy cows with 20 mg (DXS-1, n=8) and 40 mg (DXS-2, n=8) of DXM, respectively. Milk production (MP) of BED-1 and BED-2 significantly decreased during the drug administration and up to 48-hour post-drug treatment. Compared with the MP before drug administration, the MP of DXS-1 and DXL-1 was meaningfully decreased by 36 and 24-hour post-drug administration, respectively, and that in both DXS-2 and DXL-2 significantly decreased until 48-hour post-drug treatment. In conclusion, it was confirmed that the MP temporarily decreased by 48 hours after administration of DXM to dairy cows.

RADTRAN is a code that assesses the radiation risk of radioactive material transportation. RADTRAN assumes that the package is a point source or a line source regardless of package type and corrects the external dose rate using a shape factor which depends on the critical dimension of the package. However, the external dose rate calculated using a shape factor may be different from the actual external dose rate. Therefore, it is necessary to analyze the effect of the shape factor on the external dose rate. In this study, the effect of the shape factor on the external dose rate in RADTRAN was analyzed by comparison with MCNP. This study analyzed change in external dose rate depending on the distance from the package and the critical dimension. The distance from the package was in the range of 1–800 m. The shape of the package was assumed to be cylindrical with a radius of 1 m, and the critical dimensions of the package were assumed to be 2, 4, and 8 m. Attenuation and build-up in the air were not considered to consider only the effect on the shape factor. When simulating the exposure situation using MCNP, the package was assumed to be a volume source, and flux by distance from the package was calculated using F5 tally. The dose rate at 1 m from the package was normalized to 2 mSv·hr−1. As a result of the analysis, the external dose rates of the package were higher in RADTRAN than in MCNP. For the critical dimension of 2, 4, and 8 m, when the distance from package is 1–10 m, the RADTRAN was 1.83, 4.08, and 5.27 times higher on average than MCNP, respectively. And when the distance from the package was 10–100 m and 100–800 m, RADTRAN was 1.10, 2.02, 3.01 times and 1.04, 1.92, 2.43 times higher than MCNP, respectively. It was found that the larger the distance from the package is and the smaller the critical dimension of the package is, the less conservatively RADTRAN assessed. It is because the shape of the package gets closer to the point source as the distance from the package increases, and the shape factor decreases as the critical dimension of the package decreases. The result of this study can be used as the basis for radiation risk assessment when transporting radioactive materials.

Detritiation of low-level tritiated water has become global issue after Fukushima accident. Several attempts have been made to reduce the radioactivity of Fukushima tritiated water below legal limit of nuclear plant effluents (~104 Bq·L−1). Various technologies such as water distillation, electrolysis, and catalytic exchange were tested to treat the tritiated water, however, those demand enormous expense to achieve the goal due to low process efficiency. It highlights that the performance enhancement of current technologies is necessary to improve economic feasibility. We have quantitatively evaluated the separation performance of various polymers toward low-level tritium (~105 Bq·L−1) through batch experiments. The polystyrene with grafted by 20 types of functional groups (Tris (2-aminoethyl) amine, dimethylaminomethyl, isocyanate, mercaptomethyl, aminomethyl, hydroxymethyl, triphenylphosphine, morpholine, 2-chlorotrityl amine, 4-(dimethylamino) pyridine, poly (vinyl chloride) carboxylated, poly (4-vinyl pyridine), p-toluenesulfonic acid, p-toluenesulfonyl hydrazide, piperidine, acetyl polystyrene, 2-chlorotrityl chloride, piperazine, diethylene triamine, poly (vinyl chloride)) were suspended in HTO solution (initial activity = 4.7 × 105 Bq·L−1). After the equilibration, the suspension was filtered using 3 kDa membrane filter and the activity in filtrates were quantified by LSC (HIDEX-300 SL). The results demonstrate the detritiation efficiency and separation factors of functional groups toward tritium. Carboxylic group (COOH) showed the most reactive performance as detritiation efficiency of ~4%. Compared to other functional groups, styrene sulfonyl groups including sulfonyl amide (SNH2) and sulfonyl hydrazide (SNHNH2) revealed promising performance for tritium separation as separation factor of 10.97 and 3.85, respectively. However, sulfonyl hydroxide (SOH) which is known as reactive functional group to tritium exchange showed the poor performance (detritiation efficiency: 0.68%; separation factor: 3.02). This study could suggest the promising functional groups for detritiation of low-level tritiated water which can be utilized to enhance the performance of current technologies. For example, reactive functional groups can be grafted on the surface of packing material within distillation tower resulting in the increasing detritiation efficiency.