Compacted bentonite buffer blocks placed in the engineered barrier system for high-level nuclear waste disposal can undergo swelling, intrusion into rock fractures, and erosion with saturation. Bentonite erosion and intrusion can lead to bentonite mass loss via groundwater flow and can ultimately compromise the overall integrity of the disposal system. To ensure the long-term safety of deep geological disposal, it is essential to assess the hydro-mechanical interactions between the bentonite buffer and surrounding rock. In this study, the impact of bentonite erosion and intrusion on the mechanical properties of the jointed rock mass were assessed via elastic wave propagation measurements using the quasi-static resonant column test. Granite rock discs obtained from the Korea Underground Research Tunnel and Gyeongju bentonite were used to simulate jointed rock specimens with different bentonite intrusion conditions. Different degrees of bentonite intrusion were simulated by mixing bentonite and water to create bentonite paste and gel. The longitudinal and shear wave velocities under different normal stress levels were used to quantify the effects of bentonite intrusion on the mechanical characteristics of the rock joint. Complementary numerical analysis using the three-dimensional distinct element code (3DEC) was conducted to provide improved understanding of wave propagation within bentonite gouge-filled rock mass.

Some of the metal waste generated from KEPCO NF is being disposed of in the form of ingots. An ingot is a metal that is melted once and then poured into a mold to harden, and it is characterized by a uniform distribution of radioactive material. When measuring the uranium radioactivity in metal ingot with HPGe detector, 185.7 keV of U-235 is used typically because most gamma rays emitted at U-235 are distributed in low-energy regions below 200 keV. To analyze radioactivity concentration of U-235 with HPGe detector more accurately, self-attenuation due to geometrical differences between the calibration source and the sample must be corrected. In this study, the MCNP code was used to simulate the HPGe gamma spectroscopy system, and various processes were performed to prove the correlation with the actual values. First an metal ingottype standard source was manufactured for efficiency calibration, and the GEB coefficient was derived using Origin program. And through the comparison of actual measurements and simulations, the thickness of the detector’s dead layers were defined in all directions of Ge crystal. Additionally instead of making an metal ingot-type standard source every time, we analyzed the measurement tendency between commercially available HPGe calibration source (Marinelli beaker type) and the sample (metal ingot type), and derived the correction factor for geometry differences. Lastly the correction factor was taken into consideration when obtaining the uranium radioactivity concentration in the metal ingot with HPGe gamma spectroscopy. In conclusion, the U-235 radioactivity in metal ingot was underestimated about 25% of content due to the self-attenuation. Therefore it is reasonable to reflect this correction factor in the calculation of U-235 radioactivity concentration.

Several tests should be performed to estimate the structural and chemical stability of the radioactive waste. Among the tests in Gyeongju LILW repository, the leaching test which follows ANS 16.1 standard test method should be conducted for Cs, Sr, and Co radionuclides and must satisfy leachability index larger than 6 which applies deionized water as a leachant. However, the expected leachant inside the silo is groundwater that contains various ions and a high pH condition is predicted due to the concrete structures inside the silo. According to the chemical environment of the leachant, the chemical form of the radionuclides varies from precipitate to ion. Cobalt precipitates when the leachant has high pH environment which is similar condition to the cement-saturated leachant. Unlike the cobalt, cesium is preferred to exist as ion in the high pH condition. Therefore, the significant effect of the chemical environment of the leachant on the leachability of the radionuclides should be considered for the waste acceptance criteria of the radioactive waste repository. From the ‘NRC, Technical position on the waste form, rev1’, the leaching test method should follow the ANS 16.1 methods by using deionized water as leachant, however, a new leachant showing more aggressive leachability can be applied instead of deionized water. In the other hand, ASTM C1308 leaching test method recommends applying actual groundwater of the repository as a leachant. FT-04-020, the leaching test method of France, suggests the ion composition of the groundwater including the pH value. Therefore, the adequacy of using deionized water as leachant for the leaching test method of Cs, Sr, and Co should be re-examined. In this study, the leaching behavior of Cs, Sr, and Co under the several leachant types is estimated. The cement solidified specimen containing single Cs, Sr, and Co were manufactured. The leaching test following ANS 16.1 was performed by applying deionized water, simulated groundater, and cement-saturated groundwater. As a result, a leachability index difference according to the leachant type was discussed. The result of this study is expected to be a background data that helps understanding the actual leaching behavior of the Cs, Sr, and Co in the Gyeongju LILW repository.

Concerns with colloids, dispersed 1~1,000 nm particles, in the LILW repository are being raised due to their potential to enhance radionuclide release. Due to their large surface areas, radionuclides may sorb onto mobile colloids, and drift along with the colloidal transport, instead of being sorbed onto immobile surfaces. To prevent adverse implications on the safety of the repository, the colloidal impact must be evaluated. In this paper, colloid analysis done by SKB is studied, and factors to be considered for the safety assessment of colloids are analyzed. First, the colloid generation mechanism should be analyzed. In a cementitious repository, due to a highly alkaline environment, colloid formation from wastes may be promoted by the decomposition of organic materials, dissolution of inorganic materials, and corrosion of metals. Radiolysis is excluded when radionuclide inventory is moderate, as in the case of SKB. Second, colloid stability should be evaluated to determine whether colloids remain in dispersion. Stable colloids acquire electric charges, allowing particles to continuously repel one another to prevent coagulation. Thus, stability depends on the pH and ionic condition of the surroundings, and colloid composition. For instance, under a highly alkaline cementitious environment, colloids tend to be negatively charged, repelling each other, but Ca2+ ion from cement, acting as a coagulant, makes colloid unstable, promoting sedimentation. As in the case of SKB, the colloidal impact is assumed negligible in the silo, BMA, and BTF due to their extensive cement contents, but for BLA, with relatively less cement source, the colloidal impact is a potential concern. Third, colloid mobility should be assessed to appraise radionuclide release via colloid transport. The mobility depends on the density and size of colloids, and flow velocity to commence motion. As a part of the assessment, the filtration effect should also be included, which depends on pore size and structure. As in the case of SKB, due to static hydraulic conditions and engineering barriers, acting as efficient filters, colloidal transport is expected to be unlikely. In the domestic underground repository, the highly alkaline environment would lead to colloid formation, but due to high Ca2+ concentration and low flow velocity, colloids would achieve low stability and mobility, thus colloidal impact would be a minor concern. In the future, with further detailed analysis of each factor, waste composition, and disposal condition, reliable data for safety evaluation could be generated to be used as fundamental data for planning waste acceptance criteria.

A plasma torch is a kind of equipment that utilizes an electric arc to dissociate a gas and transfer an electric energy to the gas to generate very high temperature flame. KHNP-CRI has been developed the Plasma Torch Melter (PTM) to reduce radioactive waste disposal volumes and drop the radiation level of wastes. As you guess, there is required condition for proper start-up operation like current, voltage, plasma gas flow, cooling water flow, temperature in melter and so on. Thus, the optimum start-up operation condition of plasma torch will be estimated experimentally in this paper.

90Sr is considered a hazardous radionuclide due to its long half-life of 28.8 years, radiotoxicity, and potential to bioaccumulate in various organisms. In the environment, strontium typically exists as divalent cation Sr2+ or in different complexes, depending on the environmental physical and chemical factors. Despite its mobility, Sr2+ transport remains affected by adsorption from solid phases, such as soil and sediments. This research aimed to investigate the efficiency of a magnetic flocculant (MNP/IF) in separating suspended soil and Sr2+ from a soil suspension. MNP/IF was prepared via the electrostatic interaction between magnetite particles and an inorganic flocculant (IF) composed of CaCO3 and Na2SO4. Analysis of the physical properties of MNP/IF confirmed that MNP/IF was successfully imparted with magnetism and had excellent adsorption capacity for Sr2+. The optimal MNP/IF dosage for the sedimentation of suspended soil was determined to be 0.3 g/g (mass ratio of flocculant to soil). The lower the pH, the more favorable the flocculation-sedimentation process of the suspended soil by MNP/IF, since Ca2+ and Mg2+, which are the most common strong flocculators, were further eluted from IF under acidic conditions. Besides, MNP/IF exhibited outstanding removal performance for Sr2+, with maximum adsorption capacities of 163.6 mg/g observed during the flocculation-sedimentation reaction of suspended soil. The adsorption of Sr2+ exhibited consistency with the Langmuir model and followed pseudo-second-order kinetics. These findings suggest that MNP/IF can be used for the simultaneous removal of suspended soil particles and Sr2+ from a radioactive soil suspension.

The disposal of spent nuclear fuel (SNF) poses a significant challenge due to its high radioactivity and heat generation. However, SNF contains reusable materials, such as uranium and trans-uranium, which can be recovered through aqueous reprocessing or pyrochemical processes. Prior to these processes, voloxidation is necessary to increase reaction kinetics by separating fuels from cladding and reducing the particle size. In the voloxidation, uranium dioxide (UO2) from SNF is heated in the presence of oxygen and oxidized to triuranium octoxide (U3O8), resulting a release of gaseous fission products (FPs), including technetium-99 (Tc-99), which poses a risk to human health and the environment due to its high mobility and long half-life of 2.1×105. To date, various methods have been developed to capture Tc in aqueous solutions. However, a means to capture the gaseous form of Tc (Tc2O7) is essential in the voloxidation. Due to the radioactive properties of technetium isotopes, rhenium is often used as a substitute in laboratory settings. The chemical properties of rhenium and technetium, such as their electronic configurations, oxidation states, and atomic radii, are similar and these similarities indicates that the adsorption mechanism for rhenium can be analogous to that for technetium. In the previous study, a disk-type adsorbent based on CaO developed was effective in capturing Re. However, this study lacked sufficient data on the chemical properties and capture performance of the adsorbent. Furthermore, the fabrication of disk-type adsorbents is time-consuming and requires multiple steps, making it impractical for mass production. This study introduces a simple and practical method for preparing CaO-based pellets, which can be used as an adsorbent to capture Re. The results provide a better understanding of the adsorption behavior of CaO-based pellets and their potential for capturing Tc-99. To the best of our knowledge, this is the first study to apply a CaO-based pellet to capture Re and investigate its potential for capturing Tc-99.

The radiolytic decomposition of oxalic acid was investigated using gamma irradiation for decontamination of nuclear power systems. The study used high-purity analytical grade chemicals, with initial concentrations of oxalic acid prepared at 1, 2, 5, and 10 mM, and the initial pH was adjusted to 2-3 at each test condition. Gamma irradiation was performed using a high-level Co-60 source, and absorbed doses were 5, 10, 20, 30, and 50 kGy. The results showed that the efficiency of gamma irradiation decreased with longer gamma exposure time, and the G-value increased with the initial concentration of oxalic acid. Interestingly, the G-value decreased with accumulated radiation dose, but the removal increased. The dose constant ranged from 0.1695 to 0.0536 kGy-1 at different initial concentrations, and the G-value was inversely proportional to the dose constant. The study concluded that oxalic acid was successfully degraded by gamma irradiation, and 92% removal was obtained at the initial concentration of 10 mM. The mineralization of oxalic acid at higher concentrations was more difficult due to the great number of generated intermediates.

Transport packages have been developed to transport the decommissioning waste from the nuclear power plant. The packages are classified with Type IP-2 package. The IAEA requirements for Type IP-2 packages include that a free drop test should be performed for normal conditions of transport. In this study, drop tests of the packages were performed to prove the structural integrity and to verify the reliability of the analysis results by comparing the test and analysis results. Half-scale models were used for the drop tests and drop position was considered as 0.3 m oblique drop on packages weighing more than 15 tons. The strain and impact acceleration data were obtained to verify the reliability of the analysis results. Before and after the drop tests, radiation shielding tests were performed to confirm that the dose rate increase was within 20% at the external surface of the package. Also, measurement of bolt torque, and visual inspection were performed to confirm the loss or dispersion of the radioactive contents. After each drop test, slight deformations occurred in some packages. However, there was no loss of pretension in the lid bolts and the shielding thickness was not reduced for metal shields. In the package with concrete shield, the surface dose rate did not increase and there was no cracks or damage to the concrete. Therefore, the transport packages met the legal requirements (no more than a 20% increase of radiation level and no loss or dispersion of radioactive contents). Safety verifications were performed using the measured strain and acceleration data from the test, and the appropriate conservatism for the analysis results and the validity of the analysis model were confirmed. Therefore, it was found that the structural integrity of the packages was maintained under the drop test conditions. The results of this study were used as design data of the transport packages, and the packages will be used in the NPP decommissioning project in the future.

The treatment of waste generated during operation as a part of preparation for decommissioning is coming to the fore as a pending issue. Non-fuel waste stored in the spent fuel pool (SFP) of PWRs in Korea includes Dummy fuel, damaged fuel rod storage container, reactor vessel specimen cask, spent in-core instrumentation, spent control element assemblies, spent neutron source assemblies, burnable poison rods, etc. In order to treat such waste, it is necessary to classify radioactive waste level and analyze kinds of nuclide in accordance with legal requirements. In order to solve the problem, the items that KHNP-CRI is trying to conduct like followings. First, KHNP-CRI will identify the current status of non-fuel waste stored in the SFP of all domestic nuclear power plants. In order to consider the treatment of non-fuel waste, it is essential to know what kind of items and how many items are stored in the SFP. Second, to identify the dimension and characteristics of non-fuel waste stored in the SFP would be conducted. The configuration of non-fuel waste is important information to handle them. Third, the way to handle non-fuel waste would be deduced including analysis of their dimension, whether the equipment should be developed to handle each kind of non-fuel waste or not, how to transport them. In order to classify radioactive waste level and analyze the nuclide for the non-fuel waste, handling tools and the cask to transport them into the facility which nuclide analysis is able to be performed would be required. Fourth, the nuclide analysis technology would be identified. Also, domestic holding technology would be identified and which technology should be developed to classify the radioactive waste level for the non-fuel waste would be deduced. This preliminary study will provide KHNP-CRI with the insight for the nuclide analysis technology and future work which is following action for the non-fuel waste. Based on the result of above preliminary study, the feasibility of the research for the treatment of non-fuel waste would be evaluated and research plan would be established. In conclusion, the treatment of non-fuel waste stored in the spent fuel pool of domestic PWR should be considered to prepare the decommissioning. KHNP-CRI will identify the quantity, the dimension and kinds of non-fuel waste in the SFP of domestic PWR. Also, the various nuclide analysis technology would be identified and the technology which should be developed would be defined through this preliminary study.

The rise of nuclear power plants to meet escalating global energy needs has made environmental pollution including the contamination of uranium due to improper disposal of radioactive wastewater during uranium milling and mining processes. Adsorption, a water purification method known for its fast kinetics, high selectivity, and ease of use, has emerged as a popular choice for the treatment of radioactive wastewater. In response to the critical need for the purification of radioactive wastewater contaminated with U(VI), this review provides a comprehensive summary of the various types of materials, synthetic methods, and adsorption mechanisms used for the purification process. The materials are categorized into four main groups: organic, inorganic, composite/nanomaterials, and framework materials. To enhance the adsorption capacity for U(VI), researchers have explored physical and chemical modifications as well as the development of organic-inorganic hybrids. The improved adsorption performance resulting from these modifications is mostly attributed to electrostatic interaction, surface complexation, and ion exchange mechanisms. However, despite the present understanding of the processes involved, further research is still needed to fully determine the optimal approach for purifying contaminated radioactive wastewater.

After melting glass at a high temperature of about 1,100 degrees in the Cold Crucible Induction Melter (CCIM) of the vitrification facility, radioactive waste is fed into the CCIM to vitrify radioactive waste. Accordingly, since the metal sector of the CCIM contacts the high-temperature molten glass, cooling water is supplied to continuously cool the metal sector. The cooling system is divided into primary and secondary cooling water systems. The primary cooling water flows inside the metal sector of the CCIM to maintain the metal sector within normal temperature, thereby forming a glass layer between the metal sector and the high-temperature melting glass. The secondary cooling system is a system that cools the primary cooling water that cools the metal sector, and removes heat generated from the primary cooling system. In addition, it is designed to stably supply cooling water to the secondary cooling water system through an emergency cooling water system so that cooling water can be stably supplied to the secondary cooling water system in the event of secondary cooling water loss. Therefore, it is designed to maintain the facility stably in the event of loss of cooling water for the CCIM of the vitrification facility.

A vitrification facility control area is formed to control and monitor the vitrification facility process, and the control system is designed to manage the vitrification facility more safely and effectively. The control system is largely composed of a process control system and an off-gas monitoring system. The process control system is operated so that operation variables can be maintained in a normal state even in normal and transient conditions, and is designed so that the vitrification facility can be stably maintained in the event of an abnormality in the facility. The process control system consists of Programmable Logic Controller (PLC) and Local Control Panel (LCP), which controls and monitors each unit device. In addition, operation variables are provided to the operator so that the operator can manage operation variables during process control in a centralized manner for the operation of the vitrification facility. The off-gas monitoring system is operated to monitor whether the off-gas discharged to the environment is stably maintained within the standard level, and the off-gas is monitored through an independent monitoring system.

When decommissioning and operating nuclear power plants, a lot of radioactive waste in concentrated waste powder, slurry, sludge, and powder is generated. The radioactive waste, non-conformity for disposal, cannot be treated or disposed of, but is currently being stored instead. To dispose of the waste, the waste can be solidified by mixing with an appropriate solidification agent. However, when the solidification agent and powder particles are mixed as in the conventional method, the final volume of the waste form to be disposed of increases. In order to solve this problem, in this study, volume reduction was achieved, compared to the existing powder, by applying the roll compaction technology to mold the radioactive waste into compressed pellets. Soil, concrete, concentrate waste, and contaminated soil powder were used as test materials, and pellets were prepared under different operating conditions. Subsequently, a compressive strength test was performed to confirm the integrity and optimal process conditions of the manufactured pellets. However, in order to perform the compressive strength test, the upper and lower surfaces of the pellets must be horizontal, but the pellet has the shape of two tetrahedrons joined together. Hence, test specimens for measuring compressive strength were prepared by making a surface treatment jig. The compressive strength test showed a high strength of 5.20~28.20 MPa. The process conditions showing high compressive strengths were selected as the optimal process conditions. Finally, the volume reduction ratios were calculated by measuring the weight, density and volume of the manufactured pellets. The degrees of volume reduction of the manufactured pellets compared to the existing powder were checked. When the roll gap was 0 mm, the average reduction ratios of the test materials were 3.7 for the soil, 4.0 for the concrete, 4.6 for the concentrate waste, and 3.8 for the contaminated soil. When roll gap was 1 mm, the ratios were 2.7 for the soil, 2.9 for the concrete, 3.4 for the concentrate waste, and 2.8 for the contaminated soil. Therefore, from a conservative point of view (Roll gap = 1 mm), when powdered waste is formed into pellets, it means that the volume is reduced by 1/2.7 for soil, 1/2.9 for concrete, 1/3.4 for concentrated waste, and 1/2.8 for contaminated soil.

Radioactive waste must be stored for at least 300 years and must bear astronomical costs. In addition, unexpected potential risk factors are also a considerable burden. In the case of low-level radioactive waste, combustible and liquid low-contamination radioactive waste can be treated relatively easily through high-temperature plasma which the volume can be reduced by 1/250 and the weight by 1/30. It is possible to permanently dispose of the ash leached after plasma treatment in a more stable manner compared to the conventional methods. Types of low-level combustible radioactive waste, including paper, vinyl, clothing, filters, and resins, account for more than 30% of the total waste volume. Furthermore, high-temperature plasma treatment of low-level radioactive waste from petrochemical plants and medical institutions have many advantages, namely astronomical cost savings, securing free space in existing storage facilities, and improving the image of nuclear energy. Korea is preparing to decommission the Kori No. 1 nuclear power plant, and small and mediumsized enterprises and related organizations are conducting various studies to incinerate radioactive waste. In foreign countries, Britain began incineration technology in the 1970s, and Plasma Energy Group, LLC, headquartered in Florida, USA, physically changed the molecular structure of the material by combining plasma chambers and plasma arcs and obtained a patent application in 1992. Germany was approved for operation in 2002, and Switzerland completed a trial run of a plasma technologybased facility in 2004. Important radionuclides in terms of radioactive gas waste treatment include inert gases, radioiodine, and radioactive suspended particles. Gas waste is compressed in a compressor through a surge tank in the gas waste treatment system and filters at each stage. after that, the shortlife nuclide is naturally collapsed for 30 to 60 days in the storage or activated carbon adsorbent in the attenuation tank and released through HEPA filters. The radioactive concentration at discharge is monitored and managed using continuous monitoring equipment, and the oxygen concentration is managed in the gas waste treatment system to prevent explosion risk. The problem of radioactive waste disposal is not only a problem for people living in the present era, but also a big social issue that brings a burden to future generations While interest in plasma treatment is increasing from the decommissioning of the Kori Unit 1. in Korea, it is showed that there is a lack of systematic management and research especially in the radioactive volatile gases fields, that’s why I propose some ideas as follows. First, the government and related institutions should invest to the continuous radioactive monitoring system to produce and distribute continuous radioactive monitoring facilities with an affordable price. Second, it is recommended that radioactive waste incineration would be connected to the GRS system of the plant’s gas radwaste treatment system, and radioactive volatile materials should be monitored through continuous monitoring system. Third, radioactive volatile materials generated according to the temperatures and times during plasma incineration treatment are different. Therefore, prior classification of each expected radioactive volatile substance must be performed before incineration.

Level measurement of liquid radwaste is essential for inventory management of treatment system. Among various methods, level measurement based on differential pressure has many advantages. First, it is possible to measure the liquid level of the system regardless of liquid type. Second, as the instrument doesn’t need to be installed near the tank, there is no need to contact the tank when managing it. Therefore, workers’ radiation dose from the system can be decreased. Finally, although it depends on the accuracy, the price of the instrument is relatively low. With these advantages, in general, liquid radwaste level in a tank is measured using differential pressure in the treatment system. Not only the advantages described above, there are some disadvantages. As the liquid in the system is waste, it is not pure but has some suspended materials. These materials can be accumulated in tanks and pipes where the liquids move to come into direct contact with pneumatic pipes that are essential in differential pressure instruments. As a result, in case of a treatment using heat source, the accumulated materials may become sludge causing interference in pneumatic pipes. And this can change the pressure which also affects the level measured. In conclusion, in case of liquid storage tanks in which the situation cannot be checked, the proficiency of an operator becomes important.

In KHNP CRI, the PTMs (plasma torch melting system) was developed as a treatment technology of a wide variety of radioactive wastes generated by nuclear power plants. The facility is made of melting zone, thermal decomposition zone, melt discharge zone, waste feeding device, MMI, and offgas treatment system. In this study, demonstration test was conducted using NaOH solution as liquid waste to evaluation the applicability of the PTM system. For demonstration test of NaOH solution treatment, the plasma melting zone is sufficiently pre-heated by the plasma torch for 5 hours. The temperature inside the plasma melting zone is about 1,600°C. The NaOH solution as simulant was put into the thermal decomposition zone by the spray feeding device with the throughput of maximum 30 liter/hour. During the test, the power of plasma torch is about 100 kW on the transferred mode. The 160 liters of liquid waste was treated for 500 minutes. After the demonstration test, the final product in the form of salt was remained in the melting zone, and the disposal of the final product are still under consideration.

Plasma torch melting has been considered as a promising treatment technology for radioactive waste generated by nuclear power plants. The IAEA reported in 2006, the plasma melting technology could be treated regardless of the type of radioactive wastes such as combustible, non-combustible and liquid. Also, the technology has the advantage of being an eco-friendly technology. It emits less harmful gases such as NOx, SOx, HCl and CO because it does not use fossil fuels. In KHNP CRI, the plasma torch melting system was developed as the new radioactive waste treatment technology. In this study, to evaluate the long-term integrity of the new facility, a demonstration test with concrete as a simulant was carried out for about 3 days. For the 3 days, the evaluation was conducted in view of abnormal shutdown, soundness of waste feeding device, electrode consumption, and so on.

Plasma melting technology is a high-temperature flame of about 1,600°C or higher generated using electrical arc phenomena such as lightning, and radioactive waste generated during the operation and dismantling of nuclear power plants is not classified according to physical characteristics. It is a technology that can meet waste disposal requirements through treatment and reduction. Plasma torch melting technology was used for volume reduction and stable treatment of HVAC filters generated from nuclear power plants HVAC (Heating Ventilation and Air Conditioning). filter was treated by placing 1 to 3 EA in a drum and injecting it into a plasma melting furnace at 1,500°C, and the facility was operated without abnormal stop. A total of 132.5 kg of filter was treated, and the high-temperature melt was normally discharged four times. It was confirmed that the plasma torch melting facility operates stably at 500 LPM of nitrogen and 370-450 A of current during filter treatment. Through this study, the possibility of plasma treatment of filters generated at nuclear power plants has been confirmed, and it is expected that stable disposal will be possible in the future.

Japan’s government has announced plan to release the contaminated water stored from the tanks of the Fukushima Daiichi nuclear power plant site into the sea in June. The contaminated water is treated by SARRY (Cesium removal facility) and ALPS (advanced liquid processing system) to remove 62 radionuclide containing Cesium, Strontium, Iodine, and so on using filtration, precipitation (or coprecipitation) and adsorption for other nuclides (except for H-3 and C-14). The total amount of the contaminated water stored at tanks is 1,328,508 m3 (as of March 23, 2023). Currently, three ALPS systems which are existing ALPS, improved ALPS, high performance ALPS have been operated to meet the regulatory standard for release to the sea. According to the release plan, they have announced that 30 nuclides and H-3 concentration of the contaminated water will be measured and assessed before/after the discharge of the contaminated water into the sea. Before the release, the contaminated water is re-treated by reverse osmosis membrane facility and additional ALPS. And then, the water will be diluted with seawater more than 100 times. The diluted water will then move through an undersea tunnel and be released about 1 kilometer off the coast.