The design life of the radioactive waste carrier, the CHEONG JEONG NURI, is in the year 2034, when the decommissioning of Kori Unit 1 is expected. As only IP-2 type transport containers (7.5- tons, 1.6 m (W) × 3.4 m (L) × 1.2 m (H)) can be loaded onto the CHEONG-JEONG-NURI, the radioactive decommissioning waste (RDW) transport containers neither of 35-tons maximum weight nor ISO type can be accommodated. Accordingly, either a new vessel (NV) to replace the CHEONGJEONG- NURI or a change in the loading dock design of the CHEONG-JEONG-NURI is required. In this study, the necessity of building a NV capable of accommodating the issued containers above is analyzed focusing, (1) the estimated building and operating costs of the NV, and (2) the economic feasibility of the NV ‘s RDW transportation scenarios. Among bulk carriers, the CHEONG-JEONG-NURI was designed as handy-size ship type. It is operated reflecting various design requirements to satisfy the domestic/international legal requirements. To estimate the cost of the NV, the same vessel type and design criteria of the CHEONG-JEONGNURI were considered. The shipping price information of the Korea Ocean Business Corporation, as of August 2022, the building cost of bulk carrier Handysize (building NV type) is about USD 30 million. Considering domestic/overseas variables, such as future labor costs, international inflation, interest rate hike, etc., the building costs are expected to continuously rise. Furthermore, vessel operation costs of crew labor, vessel, fuel, and insurance are incurred separately. Due to the increase in oil price, and wages of special positions, such as general seafarers and radiation safety managers, the NV’s operating cost is expected to be about KRW 3.8 billion every year, which is about KRW 1.1 billion higher than that of the CHEONG-JEONG-NURI. The expected total cost of building and operating the NV is about KRW 65 billion. Assuming the repayment period of the NV building cost is the same as that of the CHEONG-JEONG-NURI building cost reimbursement agency and analyzing the economic feasibility of the transport scenario of the NV built by adding up about KRW 3.8 billion of the operating cost, cost about KRW 880 million per voyage of the NV built is expected, which being KRW 620 million more than the current cost (KRW 260 million) per trip of the CHEONG-JEONG-NURI. Therefore, transporting the RDW to the disposal facility through sustainable use of the CHEONGJEONG- NURI (considering design life extension and design change) is evaluated as more appropriate than building NV.

Treatment methods such as interim storage and immobilization are being considered to dispose of intermediate level waste (ILW), but some wastes that have been treated in the past may require repackaging. Re-packaging means to cover repackaging of waste that has already been packaged in a waste container and re-packaging is required for the following reasons: loss of shielding or containment, damage to external handling features, package out-of-specification, insufficient records and external policy. The re-packaging includes various methods such as non-intrusive treatment, overpacking of waste package, external treatment of waste container, repair waste container, injection of stabiliser, disassemble waste package, high temperature process, and dissolve waste package. The purpose of this paper is to evaluate the re-packaging possibility for various wastes by identifying the main repackaging methods among the above various re-packaging methods. 1) Disposal outside of the waste container is a viable technique for most packages, as coating with a portable spray gun for low dose rate packages or remotely using a robotic arm for high dose rate packages. 2) Waste container repair is divided into welding repair and patching of waste container according to the degree of damage. Weld repair and patching are important techniques that can be used to add additional shielding, repair damaged areas, and improve the integrity of lifting gears that may not be compliant. 3) In general, disassembly of waste packages has been applied to loose drummed waste. Packages and waste forms are physically disassembled, reduced in size, and placed in different new packages. For practical solution, grouted waste is repackaged by cutting using proprietary equipment such as diamond saws, wire saws, core drilling and rupture techniques. 4) High-temperature process involves cutting the waste package and placing the pieces in a hot bath of inorganic liquid or molten metal, and the process is applicable to all waste types. However, treatment of all gases produced, compliance with waste types and acceptance criteria. Finally, dissolving waste packages, which is generally considered impractical due to the variety of chemicals and radionuclides present in ILW, is a process that is easier to perform on raw ILW than conditioned waste. An example of waste being re-packaged is when old drummed waste is recovered from an old storage facility and the waste needs to be repackaged into a form that meets modern standards for interim storage and disposal.

Gyeongju radioactive waste repository has been operated to dispose low and intermediate level radioactive waste in Korea since 2016. Currently, only deep geological disposal facility (1st) is in operation, surface disposal facility (2nd) is scheduled to operate from 2024. As a result, the annual amount of radioactive waste that can be disposed of at deep geological disposal facilities and surface disposal facilities is almost determined. According to this result, it was possible to derive the total annual disposal amount to dispose of all radioactive waste at the Gyeongju repository after landfill disposal facility (3rd) construction. To evaluate it, a predictive model has been designed and radioactive waste generation, storage, and disposal data were input. The predictive model is based on system dynamics, which is useful to analyze the correlation between input variables. As a result of analysis, radioactive waste generation amount and maximum annual radioactive waste disposal were predicted to reach 741,615 drum and 17,030 drum per year respectively. From these results, it seems that the expansion of radioactive waste acceptance system or temporary storage is necessary.

The treatment of radioactive waste by melting has been mainly discussed with low-level waste (LLW). Considering that a large amount of waste in RV or RVI is intermediate-level waste (ILW), however, it is necessary to examine the possibility of treatment by melting of ILW. Different from LLW, melting of ILW with a high content of long-lived nuclides would lead to no free releasee, but has advantages in volume reduction, homogenization, and delay of release. In this paper, the possibility of melting as an alternative technology for the treatment of ILW in the future is reviewed by analyzing the benefits generated by melting ILW in the following aspects: 1) Similar to melting techniques of LLW, them of ILW are mostly based on well-known techniques, but it is necessary to review the feasibility of performing operations such as removal of solidified melt using remote equipment in abnormal situations such as loss of electricity. 2) It is necessary to specify radiation limits for the melting operation unless the ILW melting operation technique can guarantee that the risk of abnormal occurrence is very low. The main quantified radiation parameter is the ingot dose rate, which of 10 mSv/h is considered more reasonable. 3) Although the treatment of ILW by melting leads to a reduction in volume, the main characteristics of the waste still remain, and no waste can be disposed of for free release. Thus, the main potential benefits are improved long-term safety and reduced waste volume. 4) Reducing the surface-to-volume ratio of the molten material could reduce the amount of corrosive material per unit time and, consequently, increase long-term safety. Its effect on long-term safety is difficult to quantify precisely as it depends on several factors, such as the geometry of the original component or whether radionuclides were distributed on the surface of the original component or the induced radioactivity. 5) The volume reduction of ILW is estimated to be reduced by about 1/4 compared to the generated amount when assuming a disposal volume reduction factor of 3 and considering the dose reduction due to radioactive decay after long-term storage, however, due to the lack of knowledge about non-hazardous facility alternatives, it is difficult to evaluate cost-benefit. This is heavily influenced by both the final volume reduction and the potential to reduce the complexity of the repository’s technical barriers.

This study introduces the licensing process carried out by the regulatory body for construction and operation of the 2nd phase low level radioactive waste disposal facility in Gyeongju. Also, this study presents the experience and lessons learned from this regulatory review for preparing the license review for the next 3rd phase landfill disposal facility. Korea Radioactive Waste Agency (KORAD) submitted a license application to Nuclear Safety and Security commission (NSSC) on December 24, 2015 to obtain permit for construction and operation of the national engineered shallow land disposal facility at Wolsong, Gyeongju. NSSC and Korea Institute of Nuclear Safety (KINS) started the regulatory review process with an initial docket review of the KORAD application including Safety Analysis Report, Radiological Environmental Report and Safety Administration Rules. After reflecting the results of the docket review, the safety review of revised 10 application documents began on November 29, 2016. Total 856 queries and requests for additional information were elicited by thorough technical review until November 16, 2021. As the Gyeongju and Pohang earthquakes occurred in September 2016 and November 2017, respectively, the seismic design of the disposal facility for vault and underground gallery was enhanced from 0.2 g to 0.3 g and the site safety evaluation including groundwater characteristics was re-investigated due to earthquake-induced fault. Also, post-closure safety assessments related to normal/abnormal/human intrusion scenarios were re-performed for reflecting the results of site and design characteristics. Finally, NSSC decided to grant a license of the 2nd phase low level radioactive waste disposal facility under the Nuclear Safety Laws in July 2022. This study introduces important issues and major improvements in terms of safety during the review process and presents the lessons learned from the experience of regulatory review process.

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

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

To efficiently manage the waste packages like drums, it is meaningful to utilize the data of placement and emplacement of disposed waste in geological storage. For the transparent and real-time management of disposal data, both technical as well as administrative factors must be included. To this end, MIRAE-EN Co., Ltd. has developed a radioactive waste tracking and management system (m-trekⓇ v1.0) through radioactive waste management life cycle which is supported by KETEP. Enhancing the functional features of m-trekⓇ, IoT-based drum location measuring and data of those drums, such as position, radionuclides, activity, and dose etc., are to be collected and monitored through data modeling and visualization, which might be utilized in emplacing the loaded drums according to specifically certain criteria of internal and external data of disposal site. Position measuring using Beacon utilizes Received Signal Strength Indicator (RSSI) to locate the correct position in 3D area. Since RSSI is affected by the surrounding environment, it is required to corrective optimization. In addition, error and deviation of previously applied Gaussian filter method, was corrected and improved through AI learning model. According to this location information and those data, the prototype in future provides the visualization of drum position and its relevant data for administrative purpose such as monitoring, emplacement and other management policy.

Lubricant oil waste contaminated with radioactive materials generated at nuclear facilities can be disposed of as industrial waste in accordance with self-disposal standards if only radioactive materials are removed. Lubricant oil used in nuclear facilities consists of oil of 75-85% and additives of 15-25%, and lubricant oil waste contains heavy metals, carbon, glycol, etc. In addition, lubricant oil waste from nuclear facilities contains metallic gamma-ray emission radionuclides including Co-60, Cs-137 and volatile beta-ray emission radionuclides such as C-14 and H-3, which are not present in lubricant oil waste from general industries and these radionuclides must be eliminated according to the Atomic Energy Act. In general industries, the wet treatment technologies such as acid-white soil treatment, ion purification, thin film distillation, high temperature pyrolysis, etc. are used as the refining technology of lubricant oil waste, but it is difficult to apply these technologies to nuclear industrial sites due to restrictions related with controlling the generation of secondary radioactive waste in sludge condition containing radionuclides of metal components, and limiting the concentration of volatile radioactive elements contained in refined oil to be below the legal threshold. In view of these characteristics, the refinement system capable of efficiently refining and treating lubricant oil waste contaminated with radioactive materials generated in nuclear facilities has been developed. The treatment process of this R&D system is as follows. First, the moisture in the radioactive lubricant oil waste pretreated through the preprocessing system is removed by the heated evaporating system, and the beta-emission radionuclides of H-3 and C-14 can be easily removed in this process. Second, the heated lubricant oil waste by the heated evaporating system is cooled through the heat exchanging system. Third, the particulate matters with gamma-ray emission radionuclides are removed through the electrostatic ionizing system. Forth, the lubricant oil waste is stored in the storage tank and the purified lubricant oil waste is discharged to the outside after sampling and checking from the upper, middle and lower positions of the lubricant oil waste stored in the storage tank. Using this R&D system, it is expected that the amount of radioactive waste can be reduced by efficiently refining and treating lubricant oil waste in the form of organic compounds contaminated with radioactive materials generated in nuclear facilities.

Radioactive spent resin and concentrate waste powder generated from the primary system of nuclear power plants (NPPs) should be treated and disposed of in the form of solidified products or high integrity container (HIC) packages. We are preparing for the application of polymer concrete high integrity containers (PC-HICs) that has been approved for disposal and field application after going through the disposal suitability review of the repository operator and the license review process of the regulatory body. A reliable assessment of nuclide inventory in waste drum is required for the disposal of the radioactive waste drums, and the representative samples should be collected for both the indirect (non-destructive assessment based on the scaling factor, average radioactivity concentration, etc.) and direct (destructive analysis) evaluation of the difficult-to-measure (DTM) nuclides. It is important to secure the representativeness of samples for reliable and accurate evaluation of radionuclide inventory and approval of methodologies for highly radioactive waste such as spent resin and concentrate waste poser, and in order to secure the radiation safety of the sampling workers and representativeness of the samples, a remote sampling method is required with excellent convenience and safety and sufficient representativeness of the sample. The simple sampling device used in the past to collect samples for the scaling factor does not have a remote control function, so high-radiation samples must be collected within a very short time and it is difficult to obtain sufficiently representative samples due to structural characteristics that cannot collect the entire sample in the axial direction of the package. Therefore we developed concept designs for a remote sampling device that can satisfy both sample representativeness, operator convenience and safety.

In KAERI, Waste storage facility in the radiation management area has stored a large amount of wood waste. The amount of waste is approximately 27,000 kg, it accounts for 17% of the total waste in waste storage facility. Proper disposal of wood waste improves the fire resistance performance, secure storage space and reduce disposal costs. In order to self-disposal of wood waste, it is necessary to satisfy the self-disposal standards stipulated by the domestic Atomic Energy Act and the treatment standards of the Waste Management Act. The main factors of standards are surface contaminant, radionuclide activity and radiation dose effects. To confirm the contamination of wood waste, direct indirect measurement methods and gamma nuclide analysis were performed. To evaluate radiation dose, various computational programs were used. The results of the analysis were satisfied with domestic regulations on the classification and self-disposal of radioactive wastes. Based on this results, KAERI submitted the report on wood waste self-disposal plan to obtain approval. After final approval, wood waste is to be incinerated and incineration ash is to be buried in the designated place. The objective of this study is to provide total procedure of wood waste self-disposal and effective representative sampling method.

Waste containers for packaging, transportation and disposal of NPP (Nuclear Power Plant) decommissioning wastes are being developed. In this study, drop tests were conducted to prove the safety of containers for packaging of the wastes and to verify the reliability of the analysis results by comparing the test and analysis results. The drop height of the waste containers was considered to be 30 mm, which is the maximum lifting speed of a 50 tons crane in the waste treatment facility converted to the drop height. Drop orientation of the containers was considered for bottom-end on drop. The impact acceleration and strain data were obtained to verify the reliability of the analysis results. Before and after the drop tests, measurement of the dose rate and the radiographic testing for concrete wall, and measurement of the wall thickness of steel plate were conducted to evaluate the radiation shielding integrity. Also, measurement of bolt torque, and visual inspection were conducted to evaluate the loss or dispersion of radioactive contents. After the drop tests, the radiation dose rate on the container surface did not increase by more than 20%, and there was no crack in the concrete. In addition, the thickness of the steel plate did not change within the measurement error. Therefore, the radiation shielding integrity of the container was maintained. After the drop tests, the lid bolts were not damaged and there was no loss of pretension in the lid bolts. In addition, there was no loss or dispersion of the contents as a result of visual inspection. In order to prove the reliability of the drop analysis results, safety verifications were performed using the drop test results, and the appropriate conservatism for the analysis results and the validity of the analysis model were confirmed. Therefore, the structural integrity of the waste containers was maintained under the drop test conditions.

Organic complexing agents may affect the mobility of radionuclides at low- and intermediate-level radioactive waste repositories. Especially, isosaccharinic acid (ISA) is the main cellulose degradation product under high pH conditions in cement pore water. ISA can combine with radionuclides and form stable complexes that adversely influence adsorption in the concrete phase, resulting in radionuclides to leach to the near- and far-fields of repositories. This study focuses on investigating the sorption of ISA onto engineered barriers such as concrete, thereby studying adsorption isotherms of ISA on concrete and comparing various isotherm models with the experimental data. The adsorption experiment was conducted in three background solutions, groundwater (adjusted to pH 13 using NaOH), State 1 (artificial cement pore water, pH 13.3), and State 2 (artificial cement pore water, pH 12.5), in a batch system at a temperature of 20°C. Concrete was characterized using BET, Zeta-potential analyzer, XRD, XRF, and SEM-EDS. ISA concentrations were detected using HPLC. The experimental data were best fitted to one-site Langmuir isotherm; On the other hand, either two-site isotherm or Freundlich isotherm couldn’t give reasonable fitting to the experimental data. The observed ISA sorption behavior on concrete is crucial for the disposal of radioactive waste because it can significantly lower the concentration of ISA in the pore water. Although one-site Langmuir isotherm might effectively represent the sorption behavior of ISA on concrete, the underlying mechanism is still unknown, and further investigation should be done in the near future.

According to the ‘Regulations on the Delivery of Low and Medium Level Radioactive Waste’, Notification No. 2021-26 of the Nuclear Safety and Security Commission, a history of radioactive waste and a total amount of radioactivity in a drum are mandatory. At this time, the inventory of radionuclides that make up more than 95% of the total radioactivity contained in the waste drum should be identified, including the radioactivity of H-3, C-14, Fe-55, Co-58, Co-60, Ni-59, Ni-63, Sr- 90, Nb-94, Tc-99, I-129, Cs-137, Ce-144, and total alpha. Among nuclides to be identified, gamma-emitting nuclides are usually analyzed with a gamma ray spectrometer such as HPGe. When a specific gamma-ray is measured with a detector, several types of peaks generated by recombination or scattering of electrons are simultaneously detected in addition to the corresponding gamma-ray in gamma-ray spectroscopy. Among them, the full energy peak efficiency (FEPE) with the total gamma energy is used for equipment calibration. However, this total energy peak efficiency may not be accurately measured due to the coincidence summing effect. There are two types of coincidence summing: Random and True. The random coincidence summing occurs when two or more gamma particles emitted from multiple nuclides are simultaneously absorbed within the dead time of the detector, and this effect becomes stronger as the counting rate increases. The true coincidence summing is caused by simultaneous absorption of gamma particles emitted by two or more consecutive energy levels transitioning from single nuclide within the dead time of the detector. This effect is independent of the counting rate but affected by the geometry and absolute efficiency of the detector. The FEPE decreases and the peak count of region where the energy of gamma particles is combined increases when the coincidence summing occurs. At the Radioactive Waste Chemical Analysis Center, KAERI, samples with a dead time of 5% or more are diluted and re-measured in order to reduce the random coincidence summing when evaluating the gamma nuclide inventory of radioactive waste. In addition, a certain distance is placed between the sample and the detector during measurement to reduce the true coincidence summing. In this study, we evaluate the coincidence summing effect in our apparatus for the measurement of radioactive waste samples.

Encapsulation using cement as a solidification method for disposal of radioactive waste is most commonly used in most advanced countries in the nuclear technology to date due to its advantages such as low material cost and accumulated technology. However, in case of cement solidification, since moisture or hydroxyl group in cement is decomposed by radioactivity, it may happen that gas is generated, structural stability is weakened, and leachability is increased due to low chemical durability. Therefore, the various new solidification methods are being developed to replace it. As one of these alternative technologies, for dispersible metal compounds generated through the incineration replacement process, the study on engineering element technology for powder metallurgy is under way, which overcomes the interference problem between mechanical elements and media that may occur during the process such as the homogeneous mixing process of the target powder substance and additives used in the powder metallurgy concept-based sintering process for the solidification of the final glass composite material (GCM), the process of creating a compressed molded body using a specific mold, the process of final sintering treatment. The solidification process of dispersible radioactive waste can be largely divided into pre-treatment stage, molding stage, and sintering stage, and the characteristics of the final radioactive waste solidification material can vary depending on the solidification treatment characteristics of each stage. In relation with these characteristics, the matters to be considered when designing device for each stage to solidify dispersible radioactive waste (property of super-mixing device for homogenized powder formation, structural geometry and pressure condition of molding device for production of compressed molded body, temperature and operation characteristics of sintering device for final glass composite material (GCM), etc.) are drawn out. It is expected that the solidification device design reflecting these considerations will meet all disposal conditions of radioactive waste material, such as compressive strength and leaching characteristics of solidified radioactive waste material, and create a uniformized solidification of radioactive waste material.

Organic waste generated by small and medium-sized (S&M-sized) metal decontamination in NPP decommissioning. To lower the concentration of these organic substances for a level acceptable at the disposal site, the project of “Development of Treatment Process of Organic Decontamination Liquid Wastes from Decommissioning of Nuclear Power Plants” is being carried out. The conditioning and treatment process of organic liquid waste was designed. Also, the literature was investigated to make simulated organic liquid waste, and the composition of these waste was analyzed and compared. As the decontamination agent, organic acids such as EDTA, oxalic acid, citric acid are used. The sum of the concentrations of these organic materials was set to a maximum value of 1,000 ppm. The major metal ions of the decontamination liquid waste estimated are 59Fe, 51Cr, 54Mn, 63Ni, and the concentrations are respectively 527, 163, 161, 159 ppm. Additional major metal ions are 60Co, 58Co, 137Cs. 58Co is replaced by 60Co because it has the same chemical properties as 60Co. Unlike the HLW, the contamination level of S&M-sized metal in primary system was quite low, so 60Co is set to 2,000 Bq/g. Considering the contribution of fission and gamma ray dose constant, 137Cs was estimated to 360 Bq/g. Also, suspended solids of decontamination liquid waste were set at 500 ppm. Under these assumptions, the simulated organic liquid waste was made, and then organic substances and metal ions were analyzed with TOC analyzer and ICP-OES. The TOC analysis value was expected to 392 ppm in consideration of the equivalent organic quantity. the test result was 302 ppm. Some of organics appears to have been decomposed by acid. The values of metal ions (Fe3+, Cr3+, Mn2+, Ni2+) analyzed by ICP-OES are 139, 4, 152, 158 ppm, respectively. A large amount of Cr3+ and Fe3+ were expected to exist as ions, but they existed in the form of suspended solid. Mn2+ and Ni2+ came out similar to the expected values. The designed conditioning and treatment process is largely divided into pretreatment, conditioning, and decomposition processes. After collecting in the primary liquid waste storage tank, large particulate impurities and suspensions are removed through a pretreatment process. In the conditioning process, treated liquid waste passes through UF/RO membrane system, and pure water is discharged to the environment after monitoring. Concentrated water is decomposed in the electrochemical catalyst decomposition process, then this water secondarily passes through the RO membrane system and then discharged to the environment after monitoring. Through an additional experiment, the conditioning and treatment process will be verified.

In a nuclear power plant, the activated corrosion products are deposited on the reactor coolant system. The activated corrosion products must be removed to reduce the radiation exposure to workers before maintaining or decommissioning of the nuclear power plant. In order to remove the remove the activated duplex oxide layer generated on the reactor coolant system in the pressurized water reactor (PWR), the Cyclic SP (Sulfuric acid/Permanganate)-HyBRID (Hydrazine Based Reductive metal Ion Decontamination) process developed by KAERI (Korea Atomic Energy Research Institute) can be used. After applying the Cyclic SP-HyBRID process, a sulfate-rich waste powder containing the radionuclide is generated, and the radioactive powder has to be stabilized for final disposal. In the previous study, it was confirmed that the low-temperature sintering method can be applied to immobilize the sulfate-rich waste powder. Thus, immobilization of the Cyclic SP-HyBRID process waste powder was carried out by the low-temperature sintering method using a low melting point glass, and the physicochemical and radiological characteristics of a waste form were evaluated in this study. As a result, the compressive strength of the waste form increased with increasing sintering temperature and sintering time. It is considered that the result was caused by the difference in the band gap between the bismuth borate and zinc borate, which are the products during the sintering process. It was verified that the physical stability was maintained after the 107 Gy of irradiation test. In addition, it was confirmed that the radioactive metal hydroxides contained in the waste powder converted to metal oxide forms, which have the lower solubility, at the sintering temperature. Finally, the waste form was evaluated as a low-level radioactive waste from the concentration of radionuclides contained in the waste form.

The decommissioning waste generated during the dismantling of a nuclear power plant has various types and radioactivity levels and is characterized by being generated in a large amount in a short time. For the safe and efficient management of decommissioning waste with these characteristics, the Korea Radioactive waste Agency (KORAD) is developing a large container for decommissioning waste. And the Waste Certification Program (WCP) requirement was developed for the development of a Waste certification program for nuclear power plant operators that can prove whether the transfer/ acceptance criteria are satisfied at the step of generation of decommissioning waste packages. The radioactive waste profile, which is a waste quality certification document required by the WCP requirements, allows the KORAD to confirm that the disposition suitability evaluation was performed for each process of decommissioning waste and radiological characteristic data were evaluated appropriately. Therefore in this study, in order to propose a draft of a radioactive waste profile for large packages of decommissioning waste, overseas cases and the draft radioactive waste profile of the WCP requirements was analyzed. In addition, it was attempted to increase the utility of the derived waste profile by clearly suggesting the treatment methods for each waste stream considering the physical and radioactive characteristics evaluation methods of large decommissioning waste packages. The proposed large decommissioning waste profile can be used in the future development of a nuclear power plant operator’s decommissioning waste certification program, as well as KORAD’s a disposal facility safety evaluation and improvement of the waste tracking management system (WTS).

Glass wool, the primary material of insulation, is composed of glass fibers and is used to insulate the temperature of steam generators and pipes in nuclear power plants. Glass fiber is widely adopted as a substitute for asbestos classified as a carcinogen. The insulations used in nuclear power plants are classified as radioactive waste and most of the insulation is Very Low-Level Waste (VLLW). It is packaged in a 200 L drum the same as a Dry Active Waste (DAW). In the case of the insulations, it is packaged in a vinyl bag and then charged into the drum for securing additional safety because of the fine particle size of the fiberglass. A safety assessment of the disposal facility should be considered to dispose of radioactive waste. As a result of analyzing overseas Waste Acceptance Criteria (WAC), there is no case that has a separate limitation for glass fiber. Also, in order to confirm that glass fibers can be treated in the same manner as DAW, research related to the diffusion of glass fibers into the environment was conducted in this paper. It was confirmed that the glass fiber was precipitated due to the low flow velocity of groundwater in the Gyeongju radioactive waste repository and did not spread to the surrounding environment due to the effect of the engineering barrier. Therefore, the glass fiber has no special issue and can be treated in the same way as a DAW. In addition, it can be disposed of in the disposal facility by securing sufficient radiological safety as VLLW.

KHNP-CRI has developed Mega-Watt Class PTM (Plasma Torch Melter) for the purpose of reducing the volume of radioactive waste and immobilizing or solidifying radioactive materials. About 1 MW PTM is a treatment technology that operates a plasma torch and puts drum-shaped waste into a melter and radioactive waste in the form of slag is discharged into a waste container. Since only the overflowing slag is discharged from the melter, the discharge is intermittent. Therefore, solidification occurs in the process of discharging the melt. It is difficult to accumulate evenly in the waste container, and there is also an empty space. Solid radioactive waste must be disposed of to meet the acceptance criteria for radioactive waste. Plasma-treated solid waste raised concerns about the requirements. The waste solidification output in a slag container gave us some concerns for the waste package’s solidification and encapsulation requirements. The plasma-treated solid waste process to meet the acceptance criteria will be cost and need time consuming. Thus, a induction heating will be introduced to meet solidification requirements and test criteria of the solidification waste for the waste package disposal.