In nuclear power plants, insulation is used to protect equipment and block heat. Insulation materials include asbestos, glass fiber, calcium silicate, etc. Various types and materials are used. This study aims to ensure volume reduction and disposal safety by applying plasma torch melting technology to insulation generated at operating and dismantling nuclear power plants. After the evaluation of characteristics by securing thermal insulation materials or similar materials in use at the operational and dismantling nuclear power plant. It is planned to perform pyrolysis and melting tests using the MW plasma torch melting facility owned by KHNP CRI Before the plasma test, check the thermal decomposition and melting characteristics (fluidity, etc.) of the insulation in a 1,600°C high-temperature furnace. The insulation is stored in a 200 L drum and injected into a plasma facility, and the drum and the insulation are to be pyrolyzed and melted by the high temperature inside the plasma torch melting furnace. Through this test, thermal decomposition and melting of the insulation, solidification/ stabilization method, maximum throughput, and exhaust characteristics are confirmed at a high temperature (1,600°C) of the plasma torch. Through this study, it is expected that the stable treatment and disposal of insulation generated from operating and dismantling nuclear power plants will be possible.

Currently, KHNP has 24 operating nuclear power plant units with a toal combined capacity of about 23 GWe and two units are under construction. However, permanent stop of Kori unit 1 nuclear power plant was decided in 2017. Accordingly, interest in how to dispose of waste stored inside a permanently stopped nuclear power plant and waste generated as decommissioning process is increasing. KHNP CRI is conducting research on the advancement of plasma torch melting facilities for waste treatment generated during the plant decommissioning and operation period. Plasma torch melting facility is composed of various equipment such as a melting furnace (Melting chamber, Pyrolsis chamber), a torch, an exhaust system facility, a waste supply device, and other equipment. In demonstration test, concrete waste was put in a 200 L drum to check whether it can be pyrolyzed using a plasma torch melting facility. Reproducibility for waste treatment in the form of a 200 L drum and discharge of molten slag could be confirmed, the amount of concrete waste in 200 L Drum that could be treated according to power of plasma torch was confirmed. This demonstration test confirmed the field applicability and stability of plasma torch melting facility, and improved expectations for long-term operation.

Currently, the Gyeongju disposal facility is planned to be operated as a complex disposal facility with three types: cave disposal, surface disposal, and landfill disposal. Approximate method and arrangement have been decided up to the 1st, 2nd, and 3rd stage disposal facilities, but the optimal method for the arrangement of the entire complex disposal facility has not been established. When establishing the subsequent disposal facility arrangement plan, the generation prospect and disposal capacity setting plan for each level of radioactive waste was established, and the disposal capacity of the subsequent disposal facility, the disposal facility method, etc. were reviewed and reflected. Among the items for deriving an efficient management plan, KEPCO E&C is going to first derive a site arrangement plan for each disposal facility of 800,000 drums of radioactive waste, and has drawn up a plan for each scenario through collaboration with other organizations. When establishing the subsequent disposal facility arrangement plan, the generation prospect and disposal capacity setting plan for each level of radioactive waste was established, and the disposal capacity of the subsequent disposal facility, the disposal facility method, etc. were reviewed and reflected.

Low-and intermediate level waste (LILW) should be solidified and satisfy the waste acceptance criteria (WAC) to be disposed of in the LILW repository. The LILW should be uniformly solidified and should maintain its structural stability under the expected condition according to the WAC. Compressive strength of cement solidified waste should satisfy at least 3.44 MPa to be disposed of in the repository. In addition, its compressive strength should satisfy at least 3.44 MPa after the irradiation, immersion and leaching test. The compressive strength test and dimension of test specimen differ according to countries. However, measured compressive strength of solidified waste is affected by geometry of specimen and test condition. Diameter, ratio between diameter and height, and porosity are one of factors that affect to the compressive strength of cement solidified waste. Generally, specimen with larger diameter shows higher value of measured compressive strength. The ratio of height and diameter shows similar tendency to the diameter while larger porosity generally lowers the compressive strength. In other hands, higher compressive strength is expected when the loading rate is higher during the compressive strength test. U.S. is applying loading rate from ASTM C39 (0.25±0.05 MPa) for the compressive strength test while Korea is applying loading rate from KS F 2405 (0.6 MPa·s−1). France applies loading rate following FT-02-010 (0.5 MPa·s−1) for cement solidified waste. As the measured compressive strength increases when the loading rate increases, the effect of loading rate to the compressive strength of cement solidified waste should be assessed by quantification and consider its effect on the sight of regulation. In this study, the effect of geometric parameters of specimen and test condition to the compressive strength are checked by manufacturing specimen by solidifying mock sludge waste with cement. To prevent increasing amount of secondary waste, effects of ratio of height and diameter and porosity to the compressive strength are checked while diameter value is fixed. For loading rate, loading rate from ASTM C39 and KS F 2405 were compared. Existence of significant variance of measured compressive strengths of cement solidified waste are check by performing statistical analysis. Finally, by analyzing the relationship between test condition and measured compressive strength, the test method that measures the compressive strength conservatively is aimed to be derived.

Kori and Wolsong unit 1 were permanently shutdown in 2017 and 2019, respectively. During the decommissioning of a nuclear power plant, various types and levels of decommissioning waste will be generated sporadically in many areas in a relatively short period of time, so safe management of decommissioning waste is expected to emerge as a very important issue in the future. Since Korea has no experience in decommissioning nuclear power plants, radionuclides added by abnormal routes or errors in data can be identified through the list of expected nuclides and radioactivity data during decommissioning by analyzing cases of overseas nuclear power plants decommissioning. Therefore, it is expected that safety information of nuclear power plants in the United States (i.e. all information related to safety, such as radioactive waste characteristics and accident or decommissioning information at nuclear power plants) can be utilized when decommissioning Korea nuclear power plants. Therefore, in this study, the characteristics of solid radioactive waste were analyzed by collecting solid radioactive waste data during operation and after permanent shutdown of nine PWR nuclear power plants in the United States, and the correlation between the characteristics data of solid radioactive waste was analyzed. However, in the case of Korea, only data from the United States were analyzed because there was no data for each radionuclide that were disclosed when disposing of radioactive waste in LILW repository and there was no nuclear power plant that had been decommissioned. Correlation analysis of solid radioactive waste was performed by linking radioactivity of radionuclides, volume of waste, and total radioactivity data based on decommissioning work and accident data after permanent shutdown or during operation. The correlation analysis of total radioactivity, volume, and radioactivity of each nuclide of solid radioactive waste during operation and after permanent shutdown was performed using XLSTAT, an Excel add-in software, for carrying out Mann-Kendall Test and estimating Sen’s slope. Trends during operation and after permanent shutdown were compared and the effects of specific events or tasks were analyzed. This study is expected to be utilized as basic data related to safety management of decommissioning Korea nuclear power plants in future.

Liquid scintillation cocktail is liquid waste, which consists of an organic solvent, scintillator, surfactant, and radionuclide. Large volumes of liquid scintillation waste are generated each year, and both the organic compound and radionuclide content can negatively affect on the health and the environment. Therefore, the liquid scintillation waste should be treated in an appropriate way. In this study, to facilitate the treatment of liquid scintillation waste, the sulfate-radical advanced oxidation process (SR-AOP) was performed for the mineralization of liquid scintillator waste. In SR-AOP, highly reactive sulfate radicals, which react more selectively and efficiently with organic compounds, are produced in situ by cleaving the peroxide bond in the persulfate molecule. For the experiment, 100 times diluted ULTIMA GOLD-LLT (initial TOC=699,800 ppm) was used as a liquid scintillation waste. The TOC removal efficiency of liquid scintillation waste by the OXONE (potassium peroxymonosulfate, PMS, 2KHSO5+KHSO4+K2SO4) and sodium persulfate (PS) with varying dosages (4–12 mM) was tested, and the effects of Co2+ and Cu2+ catalysts were compared at a range of pHs (3, 7, and 9). The experimental results demonstrated that 91% TOC removal of ULTIMA GOLD-LLT could be achieved for SR-AOP at initial pH=9, Co2+=1.2 mM (catalyst), PMS=4.8 mM (oxidant) for 60 min reaction. Compared to traditional Fenton AOP which is effective only at low pH, PMS based SR-AOP with Co2+ catalyst is effective at wide range of pHs and less dependent on the treatment efficiency of the operational pH. Therefore, it can be useful for the mineralization of liquid scintillation waste which is difficult to treat with a general treatment method due to the mixture of various organic compounds.

According to the Atomic Energy Act of Korea, radioactive waste can be cleared when it meets the criteria, less than 10 uSv·y−1 for individual dose and 1 person · Sv·y−1 for collective dose. Consequently, it is necessary to evaluate radiation dose to get permission for regulatory clearance from the regulatory body of Korea. Several computational programs can be used for dose calculation depending on disposal methods such as landfill, incineration, and recycling. As for incineration, the effects of radionuclide emitted during combusting radwaste have to be considered to figure out exposure dose. In this study, GASPAR code is described to assess exposure dose from effluents released to the atmosphere during incinerating combustible radioactive wastes for regulatory clearance. GASPAR is the code programmed by Radiation Safety Information Computational Center at Oak Ridge National Laboratory for computing annual dose due to radioactive effluents released from a nuclear power plant to the atmosphere during routine operation. The calculating methods of the code are based on the mathematical model of U.S. NRC regulatory guide 1.109, about beta and gamma radiation from noble gas in semi-infinite plume, radioiodine, and particulates. GASPAR evaluates both individual dose and population dose. The considering pathways are composed of external exposure by plume and ground deposition of effluents, and internal exposure as a result of inhalation and food ingestion. Since the calculation model of GASPAR requires various variables about the radionuclide and disposal site, the accuracy of the results is decided by inputted values. The program contains the default values to parameters such as the humidity, fraction of deposition, and storage time of foods. However, to get permission, it is important to use the appropriate data representing the condition of the combustion scenario as substitutes for the default since the values are localized to the country where the code was developed. Therefore, dose assessment by GASPAR code can be applied for regulatory clearance by incineration, when reliable values depending on the disposal plan inputted.

Radioactive waste disposal facility in Korea, radioactive waste packaged in 200 L drums is placed in a concrete disposal container and disposed of at an underground silo type (cave) disposal facility. At this time, the disposal container cover is seated on the top of the disposal container, and if the disposal container and the cover are not completely combined, the container cover is raised up from the top of the disposal container, so safety problems may occur when stacking the disposal container. Therefore, various methods exist to secure a margin for the pure height inside the disposal container. The disposal container cover only covers the upper surface of the container to shield radiation, and structural performance is not required. Therefore, the method of processing the cover, such as a method of making the cover of the disposal container thin, is the easiest method to apply. In this study, a method to reduce the thickness of the cover of a concrete disposal container was devised, and structural performance under usability conditions such as lifting and seating was analyzed. In addition, the disposal container cover has a reinforced concrete form in which dissimilar materials (concrete and steel) are combined, an integrated analysis was performed to secure the reliability of the analysis results for this, and the analysis results were described. It was found that the proposed disposal container cover structure can improve usability by reducing the stress concentration phenomenon.

In nuclear power plants and nuclear facilities, radioactive waste containing hazardous substances (Mixed waste) is continuously generated due to research such as radiochemical study and nuclide analysis. In addition, radioactive waste including heavy metals and asbestos is generated during the dismantling process of nuclear power plants. Mixed wastes have both radiation hazards and chemical hazards, and there’s a possibility of synergistic effects generation. However, in most countries except the United States, there are no regulatory standards for the chemical hazards of mixed waste. The regulations applicable to mixed waste in Korea include the Nuclear Safety Act and the Waste Management Act. The Nuclear Safety Act prohibits the acceptance of hazardous radioactive waste in disposal facilities, but there is no definition or characteristic identification procedure for “hazardous.” The Waste Management Act also does not state the regulation for radioactive waste. In the Gyeongju disposal facility in Korea, the leachate in the disposal facility is expected to be a groundwater saturated with concrete and is expected to irradiated by radioactive waste. On the other hands, most of the non-radioactive waste landfill facilities are built on the surface, and the leachate is expected to be rainwater that reacts with the soil. Due to the differences in leaching environments, there’s a potential to overestimate or underestimate the leaching properties of hazardous substances if the standard leaching test is applied. To show for this, a leaching test simulating disposal facility’s environment were applied to sample waste containing heavy metals. The leaching solution was groundwater collected from the area near the Gyeongju disposal facility, which is then saturated with concrete and adjusted to pH 12.5. In addition, gamma-ray irradiation was conducted during the leaching test to observe changes in the leaching behavior of heavy metals in the actual radioactive waste disposal environment. As a result, lead showed significantly increased leaching compared to the standard test method, and cadmium was not detected in all experimental conditions except heavy irradiation. This study suggested that regulations on the hazardous of mixed waste should be settled, which should be established in sufficient consideration of the types and characteristics of substances contained in the waste.

The mixing powder of vitrification material and metallic oxide sludge was solidified by hot isostatic press method and was tested to check whether the solidified waste disposal acceptance criteria were met or not. From various contaminated tank in nuclear power plants, and other nuclear energy facilities, radioactive sludge based on metallic oxide can be generated. The most of tank consist of stainless steel can be oxidated by the long-term exposure on oxygen and moisture, and then can be made sludge layer based on metallic oxide on the inner wall of contaminated tank. Radioactive sludge waste should be solidified and disposed. Melting and hardening is the most basic method for solidification. The melting points of metallic oxide of stainless steel as Fe3O4, NiO, Cr2O3 are 1597, 1955, 2435, respectively. Those are very high temperature. To melt these metallic oxides, a furnace capable of raising the temperature to a very high temperature is required, which requires a lot of thermal energy, which may lead to an increase in disposal cost. Therefore, it is necessary to lower the melting point and solidify non-melted metallic oxide powder by adding vitrifying material powder as Na2O, SiO2, B2O3. The more vitrification material is added, the easier it is to solidify the sludge based on metallic powder at a low temperature, but there is a problem in that the total waste volume increases due to the addition of vitrification material. In this study, the mixing ratio and temperature conditions that can fix the sludge while adding a minimum amount of vitrification material will be confirmed. Mixing ratio conditions of the vitrification material and sludge powder are 10:90, 15:85, 20:80, 25:75. To fix the metallic oxide sludge by melting only the vitrification material without completely melting the metallic oxide, compression by external pressure is required. Therefore, the HIP (Hot Isostatic Pressing) method was used to solidify the metallic oxide sludge by simultaneously heating and pressurizing it. Because the softening points of most of vitrification based on Na2O, SiO2, B2O3 are ranged from 800 to 1000, temperature conditions are 800, 900, 1000. Since the compressive strength for disposing of the solidified materials was 3.4 MPa, the maximum pressure condition was set to 5000 psi (about 34 MPa), which is 10 times 3.4 MPa. And optimal mixing ratio, temperature, pressure conditions that meet the solidified waste disposal acceptance criteria will be found.

The permanent shutdown of Wolseong 1, PHWR (Pressurized Heavy Water Reactor) was decided. Accordingly, there is need for C-14 treatment technology to spent resin generated by PHWR in classified Medium Level Radioactive Waste by C-14 specific activity. However, spent resin by PHWR is mixed and stored with activated carbon and zeolite (mixture), not a single storage, and separation from the mixture must be carried out in advance for C-14 treatment in the spent resin. This study developed a C-14 treatment facility that combined with the technology of separating spent resin from spent resin mixture by PHWR NPP and the technology of C-14 treatment for disposal. The C-14 treatment facility consists of spent resin separation (Part 1) and treatment of separated spent resin. (Part 2) Part 1 is applied with a process of separating the mixed and stored spent resin from the spent resin mixture by applying a drum screen method. In the case of Part 2, spent resin treatment process for desorbing and collecting C-14 nuclides in the separated spent resin using microwave reactor was applied. Except for the adsorbent used to collect C-14 detached in the process of separating and treating spent resin, no additional material is introduced into the facility, and thus secondary waste is significantly reduced. In addition, pollution prevention banks at the bottom of the facility and a sealed automated circulation system were applied to prevent unexpected leakage and diffusion of radioactive materials and ensure stability of workers. Currently, the C-14 treatment facility has been verified for spent resin separation and spent resin treatment using simulated spent resin mixture, and the facility will be demonstrated and verified for field applicability. According to derived results, it is believed that it will be possible to apply the C-14 treatment facility when decommissioning of PHWR.

In this introduction, test devices for radwaste characterization specimen was developed and utilized. In order to permanently dispose of solidified radwastes, not only radioactive characterization but also physical & chemical characterization shall be performed to assess compliance with the waste acceptance criteria. Waste acceptance criteria can be made up measurement of free standing water, compressive strength test, thermal cycling test, radiation resistance test, leaching test, immersion test. Approximately, the equipment for each test is sorted out five types. equipment for making characterization specimen, equipment for compressive strength test, equipment for thermal cycling test, equipment for radiation resistance test, equipment for Immersion test and leaching test. Equipment for making characterization specimen is operated the dry process. The equipment of two types: one (sampling device) that cores solidified radioactive waste in a drum, and the other (cutting machine) that properly cuts the coring samples. Sampling device is not used in industry, so it is specially manufactured, cutting machine is using ready-made products. In addition, devices for compressive strength test and thermal cycling test are using ready- made products. Facility for Radiation resistance test is located in Jeong-eup. For the efficient test, a table was manufactured in the columnar form like the specimen. Finally, devices for immersion test and leaching test are so transformed that contact all surfaces of the specimen with the liquid.

The feasibility study of synthesizing graphene quantum dots from spent resin, which is used in nuclear power plants to purify the liquid radioactive waste, was conducted. Owing to radiation safety and regulatory issues, an uncontaminated ion-exchange resin, IRN150 H/OH, prior to its use in a nuclear power plant, was used as the material of experiment on synthesis of graphene quantum dots. Since the major radionuclides in spent resin are treated by thermal decomposition, prior to conducting the experiment, carbonization of ion-exchange resin was performed. The experiment on synthesis of graphene quantum dots was conducted according to the general hydrothermal/solvothermal synthesis method as follows. The carbonized ion-exchange resin was added to a solution, which is a mixture of sulfuric acid and nitric acid in ratio of 3:1, and graphene quantum dots were synthesized at 115°C for 48 hours. After synthesizing, procedure, such as purifying, filtering, evaporating were conducted to remove residual acid from the graphene quantum dots. After freeze-drying which is the last procedure, the graphene quantum dots were obtained. The obtained graphene quantum dots were characterized using atomic force microscopy (AFM), Fourier-transform infrared (FT-IR) spectroscopy and Raman spectroscopy. The AFM image demonstrates the topographic morphology of obtained graphene quantum dots, the heights of which range from 0.4 to 3 nm, corresponding to 1–4 graphene layers, and the step height is approximately 2–2.5 nm. Using FT-IR, the functional groups in obtained graphene quantum dots were detected. The stretching vibrations of hydroxyl group at 3,420 cm−1, carboxylic acid (C=O) at 1,751 cm−1, C-OH at 1,445 cm−1, and C-O at 1,054 cm−1. The identified functional groups of obtained graphene quantum dots matched the functional groups which are present if it is a graphene quantum dot. In Raman spectrum, the D and G peaks, which are the characteristics of graphene quantum dots, were detected at wavenumbers of 1,380 cm−1 and 1,580 cm−1, respectively. Thus, it was verified that the graphene quantum dots could be successfully synthesized from the ionexchange resin.

The background of the development is to contribute to the reduction of radioactive waste, recycling of resources and effectively purifying the air in the workplace. Ultimately, it affects the reduction of internal exposure of workers. According to the standard procedure of KHNP,「Use and Management of Respiratory Protection Equipment」, the expiration date of mask filter is indicated by the manufacturer before opening. It is 1 year from the date of first combination after opening. We have developed an air purifying equipment that can recycle and reuse expired mask filter waste in nuclear power plant. In order to confirm the performance, we observed air pollution level by operation time. The location was measured at 3 locations including the decontamination product warehouse in NPP, and the size of the measurement locations were 53 m3, 150 m3, 180 m3, and 900 m3. As a result of measurement, significant air purification effect was found in 53 m3 and 150 m3. Decontamination effect of 80% was shown after 1 hour of operation, and 20% of decontamination effect was shown gently for 3 hours thereafter. On the other hand, there was no significant decontamination effect in the 180 m3 and 900 m3 spaces. Significant results were derived by statistical methods. Statistical procedure involves the collection of data leading to test of the relationship between two statistical data sets, or a data set and synthetic data drawn from an idealized model. The basic composition and product characteristics was as follows: Blower, filter fixing unit, Air purifier outlet round shape, Differential pressure gauge, inverter (200 V, 3π, 200 W). The developed product weigh is 25 kg. This is lighter than the existing product weighing 100 kg. It is judged that it is suitable for convenient use. Because the area where the major air pollution level occurs is isolated by a room in NPP. This developed product has a greater significance in that it recycles radioactive waste within the radiation management area rather than air purification efficiency.

Near-surface disposal facility is more susceptible to intrusion than underground repository, resulting in more possible pathways for contaminant release. Alike human intrusion, animals (e.g. Ants, Moles, etc.) could intrude into the disposal site to excavate burrows, which could cause direct release of contaminants to biosphere. In this paper, animal intrusion is demonstrated using GoldSim’s commercial contaminant transport module and impact on the integrity of the near-surface disposal facility is evaluated in terms of fractional release rate of the contaminants. In this study, the near-surface disposal facility is modelled with a single concrete vault to contain radionuclide according to LLW concentration limit stated in NSSC notice No.2020-6. The release of contaminants is modelled to occur directly after the institutional control period, and the contaminants are mostly transported from the concrete vault to cover layers via diffusion. To produce mathematical model of the release of the contaminants due to animal intrusion, firstly, the fraction of burrow volume for each cover layer is calculated separately for each animal species, based on their maximum possible intrusion depth. In this study, fractions of burrow volume for ants and moles are calculated based on their maximum possible intrusion depths, where for ants is 2–3 m, and for moles is 0.1–0.135 m. Then, assuming that the contaminants are distributed homogeneously throughout each cover layers by diffusion, fraction of contaminants transported into the uppermost layer via excavation of the burrow is calculated for each layer based on burrow volume, and fraction of contaminants removed from the uppermost layer to the layers below via collapse of the burrow is also calculated based on the burrow volume. Lastly, the net transportation of contaminants into and out of the burrow via excavation and collapse, respectively, is calculated and demonstrated using direct transfer rate function of the GoldSim. Based on the simulated result, the maximum mass flux is too minor to cause a meaningful impact on the safety. The peak mass flux of the most sensitive radionuclide, I-129, is witnessed at around year 1,470, with a flux value of 5.36×10−6 g·yr−1. This minor release of the contaminants could be due to cover layers being much thicker than the maximum possible intrusion depth of the animals, preventing the animal intrusion into the deeper layers of higher radionuclide concentration. In future, this study can be used to provide a guidance and fundamental data for scenario development and safety evaluation of the near-surface disposal facility.

Recently, concern regarding disposal of cellulosic material is growing as cellulose is known to produce complexing agent, isosaccharinic acid (ISA), upon degradation. ISA could enhance mobility of some radionuclides, thus increasing the amount of radionuclide released into the environment. Thus, evaluation on the possible impact of the cellulose degradation would be an important aspect in safety evaluation. In this paper, safety assessments conducted in Sweden and UK are studied, and the factors required to be considered for appropriate safety assessment of cellulose is analyzed. SKB (Sweden) conducted safety assessment of cellulose degradation as a part of long-term safety assessment of SFR. SKB determined that ISA would impact sorption of trivalent and tetravalent radionuclides (Eu, Am, Th, Np, Pa, Pu, U, Tc, Zr and Nb) at concentration higher than 10−4–10−3 M, and impact sorption of divalent radionuclides (Ni, Co, Fe, Be and Pb) at concentration higher than 10−2 M. Then, SKB conservatively set the upper limit of ISA concentration to be 10−4 M and conducted cellulose degradation evaluation on each waste package type, considering the expected disposal environment of SFR. Based on the calculated results, some of the waste packages showed concentration of ISA to be higher than 10−4 M, so SKB conservatively developed waste acceptance criteria to prevent ISA being produced to an extent of affecting the safety of the repository. SKB conducted safety assessment only for the repositories with pH above 12.5 and excluded 1BLA from the safety assessment as the expected pH of 1BLA is around 12, which is insufficient for cellulose to degrade. However, SKB set disposal limit for 1BLA as well, to minimize potential impact in future. Serco (UK) conducted safety assessment of cellulose degradation for the conceptual repository, which is a concrete vault with cementitious backfill. Serco estimated that the pH of repository would maintain around 12.4. Serco conservatively assumed that the pH would be sufficient for cellulose degradation to occur partially, and suggested application of appropriate degradation ratio for safety assessment of cellulose degradation. To conduct appropriate safety assessment of cellulose degradation, an appropriate ISA concentration limit based on radionuclide inventory list, and an appropriate cellulose degradation ratio based on the pH of disposal environment should be determined. As for guidance, below pH 12.5, cellulose degradation is not expected, and between pH 12.5–13, partial cellulose degradation is expected. In future, this study could be used as fundamental data to evaluate safety of the repository.

Plasma melting technology has been considered as promising technology for treatment of radioactive wastes. According to the IAEA TECDOC-1527 report (2006), the technology has an advantage that it can treat regardless of waste types which is both combustible and non-combustible wastes. In particular, it is expected that a large amount of concrete, a representative non-combustible wastes, will be generated during the operation and dismantling of nuclear power plants. In order to treat the concrete waste in plasma torch melting system, various factors could be considered like the slag of electric conductivity, viscosity and melting temperature. Above all, as a critical factor, the viscosity of the melt is very important to easily discharge the melt. The viscosity of slag (SiO2-CaO-Al2O3 system) can be lowered by adding a basic oxide such as CaO, Na2O, MgO and MnO. The basic oxides are donors of oxygen ions. These oxides are called notwork breakers, because they destroy the network of SiO2 by reacting with it. In this study, the slag composition of the concrete waste was developed to apply the plasma torch melting. Also, demonstration test was performed with the developed slag composition and 100 kW plasma torch melting system.

In KHNP CRI, the 100 kW PTM (plasma torch melting) system was designed for the treatment and disposal technology of various radioactive wastes including the metal, concrete, liquid waste and insulator. The facility consists of melting chamber, thermal decomposition chamber, waste feeding system and off-gas treatment system. In this study, to evaluate the applicability of the PTM system, demonstration test was conducted using the radiation hazmat suit as combustible waste. The plasma melting chamber is pre-heated by 2nd combustion device and plasma torch for 5 hours. The temperature inside the plasma melting chamber is approximately 1,600°C. The combustible waste was put into the melting chamber by the pusher feeding device with the throughput of maximum 50 kg/hour. During the test, the power of plasma torch is 60–96 kW on the transferred mod. It was evaluated in terms of long-term integrity of PTM system on operation according to the waste throughput ratio.

After the Fukushima accident in 2011, a huge amount of radioactively contaminated water is being generated by cooling the melted fuel of units 1, 2 and 3. Most of contaminated water is seawater and underwater containing not only salt elements but also nuclear fission products with radioactivity. To treat the contaminated water, Cs/Sr removal facilities such as KURION and SARRY are being operated by TEPCO. Additionally, three ALPS facilities are on operation to meet the regularity standards for discharge to the sea. However, massive secondary wastes such as Zeolite, sludge and adsorbent is being generated by these facilities for liquid water treatment. The secondary wastes containing various radionuclide with Cs and Sr is difficult to store due to highly radioactive concentration and corrosive properties. In Japan, a variety of technologies such as GeoMelt vitrification, In-Can vitrification and CCIM vitrification is considered as a promising solution. In this study, they were reviewed, and the advantage and disadvantage of each technology were evaluated as the candidate technologies for thermal treatment of sludge radwaste.

Radioactive materials emitted from nuclear accident or decommissioning cause soil contamination over wide areas. In the event of such a wide area of contaminated soil, decontamination is inevitable for residents to reside and reuse as industrial land. There are many ways to decontaminate these contaminated soils, but in urgent situations, the soil washing, which has a short process period and relatively high decontamination efficiency, is considered the most suitable. However, the soil washing process of removing fine soil and cesium by using washing liquid as water and adding a flocculating agent (J-AF) generates slurry/sludge-type secondary waste (Cs-contaminated soil + flocculating agent). Since this form of sludge contaminants cannot be disposed, solidification is needed using an appropriate solidification agent to treat wastes for disposal. Therefore, this study devised a treatment method of contaminated fine soils occurring after the soil washing process. This investigation prepared the simulated wastes of contaminated fine soils generated after the soil washing, and pelletized the samples using a roll compactor under the optimum operating conditions. The optimum conditions of the device were determined in the pre-test. Roll speed, feeding rate, and hydraulic pressure were 1.5 rpm, 25 rpm, and 28.44 MPa, respectively. The waste forms were manufactured by incorporating created pellets (H 6.5 × W 9.4 mm) using polymers as solidification agents. Used polymers were main ingredient (YD-128), hardener (G-1034), and diluent (LGE). The optimum mixing ratio was YD-128 : G-1034 = 65 : 35 phr, and LGE was added in an amount of 10wt% of the total mixture. To confirm the disposal suitability of the manufactured waste forms, characterization evaluation was carried out (compressive strength, thermal cycling, immersion, and leaching test). Characterization evaluation revealed a minimum compressive strength of 23.1 MPa, far exceeding 3.44 MPa of the disposal facility waste acceptance criteria. Compressive strength increased to the highest value of 31.90 MPa after immersion test. To examine leaching characteristics, the pH, Electrical Conductivity (EC) and leachability index (􀜮􀯜) of leachates were identified. As results, pH and EC consistently increased or remained constant with leaching time. The average of Co, Cs and Sr nuclides was 17.76, 17.38 and 14.04, respectively, exceeding the value of 6 in the waste acceptance criteria. Effective waste treatment/ disposal can be achieved without increasing volumes of sludge/slurry by enhancing the technique of this research by performing additional studies in the future.