In the field of 3H decontamination technology, the number of patent applications worldwide has been steadily increasing since 2012 after the Fukushima nuclear accident. In particular, Japan has a relatively large number of intellectual property rights in the field of 3H processing technology, and it seems to have entered a mature stage in which the growth rate of patent applications is slightly reduced. In Japan, tritium is being decontaminated through the Semi-Pilot-class complex process (ROSATOM, Russia) using vacuum distillation and hydrogen isotope exchange reaction, and the Combined Electrolysis Catalytic Exchange (CECE, Kurion, U.S.) process. However, it is not enough to handle the increasing number of HTOs every year, so the decision to release them to the sea has been made. Another commercial technology in foreign countries is the vapor phase catalyst exchange process (VPCE) in operation at the Darlington Nuclear Power Plant in Canada. This process is a case of applying tritium exchange technology using a catalyst in a high-temperature vapor state. The only commercially available tritium removal technology in Korea is the Wolseong Nuclear Power Plant’s Removal Facility (TRF). However, TRF is a process for removing HTO from D2O of pure water, so it is suitable only for heavy water with high tritium concentration, and is not suitable for seawater caused by Fukushima nuclear power plant’s serious accident, and surface water and groundwater contaminated by environmental outflow of tritium. Until now, such as low-temperature decompression distillation method, water-hydrogen isotope exchange method, gas hydrate method, acid and alkali treatment method, adsorption method using inorganic adsorbent (zeolite, activated carbon), separator method using electrolysis, ion exchange adsorption method using ion exchange resin, etc. have been studied as leading technologies for tritium decontamination. However, any single technology alone has problems such as energy efficiency and processing capacity in processing tritium, and needs to be supplemented. Therefore, in this study, four core technologies with potential for development were selected to select the elemental technology field of pilot facilities for treating tritium, and specialized research teams from four universities are conducting technology development. It was verified that, although each process has different operating conditions, tritium removal performance is up to 60% in the multi-stage zeolite membrane process, 30% in the metal oxide & electrochemical treatment process, 43% in the process using hydrophilic inorganic adsorbent, and 8% in the process using functional ion exchange resin. After that, in order to fuse with the pretreatment process technology for treating various water quality tritium contaminated water conducted in previous studies, the hybrid composite process was designed by reflecting the characteristics of each technology. The first goal is to create a Pilot hybrid tritium removal facility with 70% tritium removal efficiency and a flow rate of 10 L/hr, and eventually develop a 100 L/hr flow tritium removal system with 80% tritium removal efficiency through performance improvement and scale-up. It is also considering technology for the postprocessing process in the future.

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.

Recently, Japan’s government has announced Tokyo Electric Power Company’s plan to discharge contaminated water stored from the tanks of the Fukushima Daiichi nuclear power plant site into the sea. The contaminated water is treated by advanced liquid processing system (ALPS) to remove 62 radionuclide containing cesium, strontium, iodine and etc. using co-precipitation (or precipitation) and adsorption for other nuclides (except for tritium and carbon-14). The total amount of the contaminated water generated by ALPS facility is 1,311,736 m3 (as of August 18, 2022). The amount of contaminated water is estimated same as Tokyo dome volume. Under the sea discharge plan, the contaminated water will be diluted in seawater more than 100 times, and tritium concentration lowered 1/7 of the drinking water standard set by the World Health Organization (10,000 Bq/liters). The diluted water will then move through an undersea tunnel and be discharged about 1 kilometer off the coast.

With the aging of nuclear power plants (NPPs) in 37 countries around the world, 207 out of 437 NPPs have been permanently shutdown as of August 2022 according to the IAEA. In Korea, the decommissioning of NPPs is emerging as a challenge due to the permanent shutdown of Kori Unit 1 and Wolsong Unit 1. However, there are no cases of decommissioning activities for Heavy Water Reactor (HWR) such as Wolsong Unit 1 although most of the decommissioning technologies for Light Water Reactor (LWR) such as Kori Unit 1 have been developed and there are cases of overseas decommissioning activities. This study shows the development of a decommissioning waste amount/cost/process linkage program for decommissioning Pressurized Heavy Water Reactor (PHWR), i.e. CANDU NPPs. The proposed program is an integrated management program that can derive optimal processes from an economic and safety perspective when decommissioning PHWR based on 3D modeling of the structures and digital mock-up system that links the characteristic data of PHWR, equipment and construction methods. This program can be used to simulate the nuclear decommissioning activities in a virtual space in three dimensions, and to evaluate the decommissioning operation characteristics, waste amount, cost, and exposure dose to worker. In order to verify the results, our methods for calculating optimal decommissioning quantity, which are closely related to radiological impact on workers and cost reduction during decommissioning, were compared with the methods of the foreign specialized institution (NAGRA). The optimal decommissioning quantity can be calculated by classifying the radioactivity level through MCNP modeling of waste, investigating domestic disposal containers, and selecting cutting sizes, so that costs can be reduced according to the final disposal waste reduction. As the target waste to be decommissioning for comparative study with NAGRA, the calandria in PHWR was modeled using MCNP. For packaging waste container, NAGRA selected three (P2A, P3, MOSAIK), and we selected two (P2A, P3) and compared them. It is intended to develop an integrated management program to derive the optimal process for decommissioning PHWR by linking the optimal decommissioning quantity calculation methodology with the detailed studies on exposure dose to worker, decommissioning order, difficulty of work, and cost evaluation. As a result, it is considered that it can be used not only for PHWR but also for other types of NPPs decommissioning in the future to derive optimal results such as worker safety and cost reduction.

Regulations on the concentration of boron discharged from industrial facilities, including nuclear power plants, are increasingly being strengthened worldwide. Since boron exists as boric acid at pH 7 or lower, it is very difficult to remove it in the existing LRS (Liquid Radwaste System) using RO and ion exchange resin. As an alternative technology for removing boron emitted from nuclear power plants, the electrochemical boron removal technology, which has been experimentally applied at the Ringhal Power Plant in Sweden, was introduced in the last presentation. In this study, the internal structure of the electrochemical module was improved to reduce the boron concentration to 5 mg/L or less in the 50 mg/L level of boron-containing waste liquid. In addition, the applicability of the electrochemical boron removal technology was evaluated by increasing the capacity of the unit module to 1 m3/hr in consideration of the actual capacity of the monitor tank of the nuclear power plant. By applying various experimental conditions such as flow rate and pressure, the optimum boron removal conditions using electrochemical technology were confirmed, and various operating conditions necessary for actual operation were established by configuring a concentrated water recirculation system to minimize secondary waste generation. The optimal arrangement method of the 1 m3/hr unit module developed in this study was reviewed by performing mathematical modeling based on the actual capacity of monitor tank and discharge characteristics of nuclear power plant.

For the decommissioning or continuous long-term power generation of nuclear power plants, it is necessary to transfer the spent nuclear fuel from the wet storage pool to the dry storage. Spent nuclear fuel should go through the drying process, which is the first step of dry storage. The most important part in the drying process is the removal of the residual water. The spent fuel might be stored in a dry storage system for a long time. The integrity of internal components and spent fuel cladding should be maintained during the storage period. If residual water is present, problems such as aging of metal materials, oxidation of cladding, and the hydride-reorientation could occur. The presence or absence of residual water after vacuum drying is evaluated by pressure. If there is residual water in the vacuum drying process, it evaporates easily at low pressure to form water vapor pressure and the internal pressure rises. In the recent EPRI High burn up demonstration test, the gas inside the canister that satisfied the dryness criteria was extracted and analyzed. It showed that the water content was higher than the expected value. We are conducting verification studies on the pressure evaluation method, which is an indirect evaluation method of vacuum drying. The vacuum drying test was performed on small specimens at Sandia National Laboratory, and quantitative residual water evaluation was also performed. The report did not mention a detailed method for the assessment of residual water. Based on the test results of SNL, direct residual water evaluation was performed using energy balance. If the dryness criteria were satisfied, the quantitative amount of residual water was also evaluated. As a result, almost the same result as the evaluation result of SNL was derived, and it was confirmed that most of the water was removed when the dryness criteria was satisfied.

For Dry Storage of Spent Nuclear Fuel (SNF), all moisture must be removed from the dry storage canister through subjected to a drying process to ensure the long-term integrity. In NUREG-1536, the evacuation of most water contained within the canister is recommended a pressure of 0.4 kPa (3 torr) to be held in the canister for at least 30 minutes while isolated from active vacuum pumping as a measure of sufficient dryness in the canister. In the existing drying process, the determination of drying end point was determined using a dew point sensor indirectly. Various methods are being studied to quantify the moisture content remaining inside the canister. We presented a moisture quantification method using the drying process variables, like as temperature, pressure, and relative humidity operation data. During the drying process, it exists in the form of a mixed gas of water vapor and air inside the canister. At this time, if the density of water vapor in the mixed gas discharged out of the canister by the vacuum pump is known, the mass of water removed by vacuum drying can be calculated. The canister is equipped with a pressure gauge, thermometer and dew point sensor. The density of water vapor is calculated using the pressure, temperature and relative humidity of the gas obtained from these sensors. First, calculate the saturated water vapor pressure, and then calculate the humidity ratio. The humidity ratio refers to the ratio of water vapor mass to the dry air mass. After calculating the density of dry gas, multiply the density by the humidity ratio to calculate the density of water vapor (kg/m3). Multiply the water vapor density by the volume flow (m3/s) to obtain the mass value of water (kg). The calculated mass value is the mass value obtained per second since it is calculated through the flow data obtained every second, and the amount of water removed can be obtained by summing all the mass values. By comparing this value with the initial moisture content, the amount of moisture remaining inside the canister can be estimated. The validity of the calculations will be verified through an experimental test in the near future. We plan to conduct various research and development to quantify residual water, which is important to ensure the safety of the drying process for dry storage.

PURPOSES : The increasing heat wave warnings during the summer season in Korea have significant impacts on daily life and industry as a whole, especially in urban areas (such as areas with asphalt and sidewalk pavements). Heat waves directly affect urban heat island and heat dome phenomena. Various urban temperature reduction measures are being discussed to reduce urban heat islands and heat dome phenomena and to improve citizen safety against summer heat waves; suggestions include thermal packaging, rooftop greening, and expansion of vegetation areas. There is a lack of analysis on the methodology for increasing the road spraying effect during summer heat waves (e.g., there is no systematic engineering study on the effect from reducing the temperature of the road spraying during a heat wave in the city) and on the types of road pavements in the city. In addition, as the asphalt pavements of roadways and block pavements installed in sidewalks account for a considerable portion of all pavements, this study provides a more systematic and scientific approach and procedures for reducing temperatures through road spraying in the city by tracking the effects of heat waves. METHODS : In this preliminary experiment, four types of road pavement materials were selected as test specimens: asphalt test specimens (AP- 300 mm × 300 mm × 50 mm), concrete test specimens (CP-300 mm × 300 mm × 50 mm), impermeable blocks (IB-200 mm × 200 mm × 60 mm), and self-permeable blocks (PB-200 mm × 200 mm × 60 mm). As a test method to evaluate the size and duration of each spray effect package type, the surface temperature of each specimen was measured using thermal imaging cameras every 20 min after spraying at the maximum temperature point of each specimen, and the average surface temperature was analyzed based on the collected temperature data. In addition, to conduct a quantitative analysis of the effect of reducing the surface temperature of road pavements by road spraying in summer, field tests were conducted on asphalt roads and watertight blocks for sidewalks. RESULTS : As a result of the comparative analysis of the spray effect under a 36 ℃ air temperature based on a heat wave warning, the surface temperatures were, from high to low, the asphalt (68.8 ℃), concrete (59.1 ℃), impermeable block (57.3 ℃), and permeable block (58.7 ℃). The asphalt pavement had the greatest effect on the heat island and heat dome phenomena. From measuring the temperature reduction effect and sustainability of each type of road pavement, the surface temperature reduction effects were ranked in the following order: water-permeable block (Δ18.0 ℃), asphalt test piece (Δ17.5 ℃), concrete test piece (Δ12.2 ℃), and water-permeable block (over 240 min). In the report pitching block, the average road surface temperature reduction between the pore recovery and treatment was expected to continue to decrease by approximately -4.3 ℃ on the day of work and approximately -2.4 ℃ on the next day. The expected effect of the temperature reduction owing to simple spraying on the surface of the pore block was evaluated to be limited to the day. CONCLUSIONS : In the road spray effect analysis conducted on the common asphalt road, there was a slight difference in the initial temperature reduction size as the test specimen was measured, but the surface temperature difference between the non-spray section and spray section tended to be approximately Δ3°C after 140 minutes of spraying. Therefore, it was determined that the asphalt pavement temperature reduction plan through road spraying in urban areas in summer would be the most effective if it was repeated twice or more in an hour (between 13:00 and 14:00) on the day of the heat wave.

The safety of a KTC-360 transport cask, a large-capacity pressurized heavy-water reactor transport cask that transports CANDU spent nuclear fuel discharged from the reactor after burning in a pressurized heavy-water reactor, must be demonstrated under the normal transport and accident conditions specified under transport cask regulations. To confirm the thermal integrity of this cask under normal transport and accident conditions, high-temperature and fire tests were performed using a one-third slice model of an actual KTC-360 cask. The results revealed that the surface temperature of the cask was 62°C, indicating that such casks must be transported separately. The highest temperature of the CANDU spent nuclear fuel was predicted to be lower than the melting temperature of Zircaloy-4, which was the sheath material used. Therefore, if normal operating conditions are applied, the thermal integrity of a KTC-360 cask can be maintained under normal transport conditions. The fire test revealed that the maximum temperatures of the structural materials, stainless steel, and carbon steel were 446°C lower than the permitted maximum temperatures, proving the thermal integrity of the cask under fire accident conditions.

The acoustic emission (AE) is proposed as a feasible method for the real-time monitoring of the structural damage evolution in concrete materials that are typically used in the storage of nuclear wastes. However, the characteristics of AE signals emitted from concrete structures subjected to various environmental conditions are poorly identified. Therefore, this study examines the AE characteristics of the concrete structures during uniaxial compression, where the storage temperature and immersion conditions of the concrete specimens varied from 15℃ to 75℃ and from completely dry to water-immersion, respectively. Compared with the dry specimens, the water-immersed specimens exhibited significantly reduced uniaxial compressive strengths by approximately 26%, total AE energy by approximately 90%, and max RA value by approximately 70%. As the treatment temperature increased, the strength and AE parameters, such as AE count, AE energy, and RA value, of the dry specimens increased; however, the temperature effect was only minimal for the immersed specimens. This study suggests that the AE technique can capture the mechanical damage evolution of concrete materials, but their AE characteristics can vary with respect to the storage conditions.

Molten salt consisting primarily of eutectic LiCl-KCl is currently being used in electrorefiners in the Fuel Conditioning Facility at Idaho National Laboratory. Options are currently being evaluated for storing this salt outside of the argon atmosphere hot cell. The hygroscopic nature of eutectic LiCl-KCl makes is susceptible to deliquescence in air followed by extreme corrosion of metallic cannisters. In this study, the effect of occluding the salt into a zeolite on water sorption/desorption was tested. Two zeolites were investigated: Na-Y and zeolite 4A. Na-Y was ineffective at occluding a high percentage of the salt at either 10 or 20wt% loading. Zeolite-4A was effective at occluding the salt with high efficiency at both loading levels. Weight gain in salt occluded zeolite-4A (SOZ) from water sorption at 20% relative humidity and 40℃ was 17wt% for 10% SOZ and 10wt% for 20% SOZ. In both cases, neither deliquescence nor corrosion occurred over a period of 31 days. After hydration, most of the water could be driven off by heating the hydrated salt occluded zeolite to 530℃. However, some HCl forms during dehydration due to salt hydrolysis. Over a wide range of temperatures (320–700℃) and ramp rates (5, 10, and 20℃ min−1), HCl formation was no more than 0.6% of the Cl− in the original salt.

Drought stress is a condition that occurs frequently in the field, it reduces of the agricultural yield of field crops. The aim of the study was to screen drought-adapted genotype of Italian rye grass. The experiments were conducted between the two Italian ryegrass (Lolium multiflorum L.) cultivars viz. Hwasan (H) and Kowinearly (KE). The plants were exposed to drought for 14 days. The results suggest that the morphological traits and biomass yield of KE significantly affected by drought stress-induced oxidative stress as the hydrogen peroxide (H2O2) level was induced, while these parameters were unchanged or less affected in H. Furthermore, the cultivar H showed better adaptation by maintaining several physiological parameter including photosystem-II (Fv/Fm), water use efficiency (WUE) and relative water content (RWC%) level in response to drought stress. These results indicate that the cultivar H shows improved drought tolerance by generic variation, improving photosynthetic efficiency and reducing oxidative stress damages under drought stress. These findings can be useful to the breeder and farmer for improving drought tolerance in Italian rye grass through breeding programs.