In the decommissioning site of Korean Research Reactor 1&2 (KRR-1&2), according to Low and Intermediate-level Radioactive Waste Disposal Acceptance Criteria of the Korea Radioactive Waste Agency (WAC-SIL-2022-1), characteristics of radioactive waste was conducted on approximately 550 drums of concrete and soil waste for a year starting from 2021. Among them, 50 drums of concrete waste transported and disposed to Gyeongju LILW disposal facility at the end of 2022. For the remaining approximately 500 drums of concrete and soil waste stored on-site, they were reclassified into two categories: permanent disposal grade and clearance grade. This classification was based on calculating the sum of fractions (SOF) per drum for each radionuclides. The plan is to dispose of around 200 drums in the permanent disposal grade and about 300 drums in the clearance grade by the end of 2023. Since concrete and soil decommissioning wastes are generated in large quantities over a short period with similar origins, they were grouped within five drums as suggested by the acceptance criteria. Mixed samples were collected from each group and used for radionuclide analysis. When utilizing mixed samples, three distinct samples are collected and analyzed for each group. The maximum value among these three radionuclide analysis results is then uniformly applied as the radionuclide concentration value for all drums within that group. Radioactive nuclides contained in similar types of radioactive waste with similar origins can be expected to have some statistical distribution. However, There has been no verification as to whether the maximum value among the three mixed samples exists within the statistical distribution or if it deviates from this distribution to represent a different value. In this study, we confirmed characteristics of radionuclide concentration distribution by examining and comparing radionuclide concentration distributions for radioactive wastes drum grouped for nuclear characteristic among 50 concrete wastes drum disposed in year 2022 and 500 concretes & soils drum scheduled for disposal (clearance or permanent disposal) in year 2023. In particular, when comparing tritium to other nuclides, it was observed that the standard deviation for the distribution of maximum values was approximately 318 times larger.

Domestic commercial low- and intermediate-level radioactive waste storage containers are manufactured using 1.2 mm thick cold-rolled steel sheets, and the outer surface is coated with a thin layer of primer of 10~36 μm. However, the outer surface of the primer of the container may be damaged due to physical friction, such as acceleration, resonance, and vibration during transportation. As a result, exposed steel surfaces undergo accelerated corrosion, reducing the overall durability of the container. The integrity of storage containers is directly related to the safety of workers. Therefore, the development of storage containers with enhanced durability is necessary. This paper provides an analysis of mechanical properties related to the durability of WC (tungsten carbide)-based coating materials for developing low- and intermediate-level radioactive waste storage containers. Three different WC-based coating specimens with varied composition ratios were prepared using HVOF (high-velocity oxy-fuel) technique. These different specimens (namely WC-85, WC-73, and WC-66) were uniformly deposited on cold-rolled steel surfaces ensuring a constant thickness of 250 μm. In this work, the mechanical properties of the three different WCbased coaitng materials evaluated from the viewpoints of microstructure, hardness, adheision force between substrate and coating material, and wear resistance. The cross-sectional SEM-EDS (Scanning Electron Microscope-Energy Dispersive X-ray Spectroscopy) images revealed that elements W (tungsten), C (carbon), Ni (nickel), and Cr (chromium) were uniformly distributed within the each coating layers which was approximately 250 μm thick. The average hardness values of HWC-85 and HWC-73 were found to be 1,091 Hv (Vickers Hardness) and 1,083 Hv, respectively, while the HWC-66 exhibited relatively lower hardness value of 883 Hv. This indicates that a higher WC content results in increased hardness. Adhesion force between and substrates and coating materials exceeded 60 MPa for all specimens, however, there were no significant differences observed based on the tungsten carbide content. Furthermore, a taber-type abrasion tester was used for conducting abrasion resistance tests under specific conditions including an H-18 load weight at 1,000 g with rotational speed set at 60 RPM. The abrasion resistance tests were performed under ambient temperatures (RT: 23±2°C) as well as relative humidity levels (RH: 50±10%). Currently, the ongoing abrasion resistance tests will include some results in this study.

Various radioactive metal wastes are generated during operation and decommissioning of nuclear facilities. Radioactive metal wastes with complex geometries or volumetric contamination can be difficult to decontaminate and disposal costs may increase. To solve these problems, the radioactive metal wastes can be treated by melting method. In this study, we designed a melting furnace system of air induction melting type, which is widely utilized due to its advantages of good thermal efficiency, uniform heating and guaranteed safety for radioactive material. By utilizing the melting furnace system, volatile radionuclides existed in the base material can be captured in the form of gas or dust by the filter. The radionuclides whose chemical properties can easily form metal oxides present as slag. For this reason, the specific radioactivity of the base material can be reduced. Radionuclides that are difficult to transport to slag and dust are uniformly distributed in the base material. A dedicated power supply and a transformer were necessary to be included in the melting furnace system since the induction furnace uses high-frequency currents. In addition, a hood is placed on top of the furnace to capture fumes generated during melting, and additional hoods were installed around the furnace to remove airborne dust. In particular, a dust collection unit consisting of a cyclone and a HEPA filter were constructed to effectively collect dust containing radionuclides. During the melting process, the slag is removed and accumulated separately, and the ingot production system was designed to produce the ingot using molten metal. The furnace was constructed for tilting the molten metal by moving the furnace using hydraulic system. The water cooling system and cooling tower were prepared to cool off the equipment with high temperature during melting is cooled off. The above process was specified in the operating procedure developed for this melting furnace system, and the operator shall operate and inspect according to the prescribed procedures. The radioactivity concentration in the sample taken in the step of tilting shall be analyzed whether they meet clearance level for self-disposal determined and publicly announced by the Commission. We can conduct self-disposal for the product of melting furnace system confirmed by the Commission as having the radioactivity concentration by nuclide not exceeding the value determined by the Commission.

The radioactive waste generated within radiation-controlled areas is classified and processed according to relevant laws and regulations based on contamination levels. In cases where such radioactive waste complies with the legally defined clearance concentration or dose criteria, it is disposed of as non-radioactive waste by means of incineration, reclamation, recycling, etc. Within radiation controlled areas, various consumables are periodically replaced to ensure the proper operation of the area. It is necessary to have appropriate disposal methods for these consumables. In particular, waste items such as fire extinguishers, fluorescent lamps, batteries, and pressure vessels (hereinafter referred to as “Special Waste Type”), which may contain hazardous substances within their internal components and contents, should be considered for appropriate disposal methods that comply with nuclear safety and environmental laws. In the present case, the specified special waste type do not come into direct contact with radiation sources, and they have impermeable surfaces, which significantly reduces the risk of external contamination infiltrating the interior. However, the current method of clearance is not suitable for these items (Typically, nuclear energy-related business operators are required to classify clearance target waste based on internal and external components and demonstrate compliance with the criteria. Nevertheless, for special waste type, it is difficult to separate and measure internal and external components within the radiation-controlled area). In this case, the Clearance Procedure for special waste type applied to Korea Atomic Energy Research Institute was introduced. Additionally, we have extracted considerations for future domestic clearance of the type.

To evaluate the inventory of radionuclides for the disposal of waste generated from nuclear power plants, indirect assessment methods such as the scaling factor method or average radioactivity concentration method can be applied. A scaling factor represents the average concentration ratio between key radionuclides and difficult-to-measure (DTM) radionuclides, while the average radioactivity concentration refers to the average concentration of DTM radionuclides, regardless of the concentration of key radionuclides or within specific ranges of key radionuclide concentrations. These indirect assessment methods can be statistically derived through the analysis of representative drums. This study will address how to apply these scaling factors and average radioactivity concentrations. Firstly, the concentration of gamma-emitting radionuclides will be analyzed using a drum radionuclide analyzer, and the concentration of DTM radionuclides will be determined by applying scaling factors specific to each DTM radionuclide. In the case of using the average radioactivity concentration method, the average concentration of DTM radionuclides will be applied independently of the concentration of gamma-emitting radionuclides. It is crucial to perform radioactive decay correction based on the date of generation or disposal when applying scaling factors or average radioactivity concentration. Additionally, for repackaged 320 L drums, determining which drum among the two 200 L drums inside should serve as the reference is of utmost importance

Wasteform is the first barrier to prevent radionuclide release from repositories into the biosphere. Since leaching rates of nuclides in wasteform significantly impact on safety assessment of the repository, clarifying the leaching behavior is critical for accurate safety assessment. However, the current waste acceptance criteria (WAC) of the Gyeongju repository only evaluates leachability indexes for Cs, Sr, and Co, which are tracers for nuclear power plant waste streams. Furthermore, ANS 16.1, the current leaching test method used in WAC, applies deionized water (DI) as leachant. However, the interactions between wasteform and groundwater environment in the repository may not be reflected. Therefore, it is necessary to review the current leaching test method and nuclides that may require the extra evaluation of leachability beyond the Cs, Sr, and Co. Tc and I are key nuclides contributing to high radioactive dose in safety assessment due to their high mobility and low retardation factor. The groundwater conditions within the repository, such as pH and Eh significantly affect the chemical form of Tc and I. For example, Tc in H2O system tends to form hydroxide precipitates in neutral pH condition and TcO4 - in strong alkaline environments according to the Pourbaix diagram. In case of I, it generally exists in the form of I-, while it exists as IO3 - as Eh increases. Although the current leaching test at the Gyeongju repository applies DI as a leachant, the actual repository is expected to have a highly alkaline environment with a substantial amount of various ions in the groundwater. Consequently, the leaching behavior in the ANS 16.1 test and the actual disposal condition is different. Thus, it is necessary to analyze the leaching behavior of Tc and I with reflecting the actual disposal environment. In this study, the leaching behavior of Tc and I is investigated by following ANS 16.1 leaching test method. The solidified waste specimens containing 10 mmol of Re and I were manufactured with cement, which is widely used as a solidification material. Re was applied instead of Tc, which has similar chemical behavior to Tc, and NH4ReO4 and NaI were used as surrogates for Re and I. As a leachant, deionized water and cement-saturated groundwater were prepared and the concentration of nuclides in the leachant is analyzed by ICP-OES. As the result of this study, experimental data can be applied to improve the WAC and disposal concentration standards in the future.

Dry active wastes (DAWs) are combustible waste generated during the operation and decommissioning of nuclear facilities, and are known to be generated in the amount of approximately 10,000 to 40,000 drums (based on 200 L) per unit. It consists of various types of protective clothing, paper, and plastic bags, and is stored in radioactive waste storage facilities. Therefore, reducing the volume of DAWs is an important issue in order to reduce storage costs and utilize the limited space of waste storage facilities. Heat treatment such as incineration can dramatically reduce the volume of waste, but as the waste is thermally decomposed, CO2, a global warming gas, is generated and there is a risk of emissions of harmful gases including radionuclides. Therefore, a heat treatment process that minimizes the generation of CO2 and harmful gases is necessary. One of the alternatives to incineration is to carbonize DAWs, dispose of carbonized materials below the release standard as non-radioactive waste, and selectively separate and stabilize inorganic components, including radionuclides, from carbonized DAWs. In this study, 13 types of DAWs generated from nuclear power plants were selected and their thermal decomposition characteristics were investigated to design a heat treatment process that replaces incineration. As a result of TGA analysis, the temperature at which thermal decomposition of each waste begins is 260-300°C for cotton, 320-330°C for paper, 315-420°C for synthetic fiber, 350°C for latex gloves. The mass of most samples decreased to less than 1 % of the initial weight after heat treatment, and dust suit and latex gloves had residues of 13.83% and 13.71% of the initial mass, respectively. The metal components of the residue produced after heat treatment of the sample were analyzed by EDS. According to the EDS results, cotton contains Ca and Al, paper contains Ca, Al and Si, synthetic fiber contains Ca, Cu and Ti, latex gloves contain Ca and Mg. Additionally, ICP analysis was performed to quantify the inorganic components. These results are expected to be applicable to the processing of DAW generated at nuclear facilities in the future.

Recently, the nuclear decommissioning and environmental restoration industries has significantly attracted as a new industry field due to the decision to decommission the KORI#1 and WOLSONG #1 nuclear power plant. In order to dispose of the decommissioning radioactive wastes generated during nuclear decommissioning, proper analysis is required, and disposal decisions are determined based on the analysis results. When dismantling a nuclear power plant, a few thousand of tons decommissioning waste are produced, so these require analysis for proper disposal. Therefore, a radionuclide facility for decommissioning waste analysis is essential for the disposal of the large quantities of decommissioning waste generated during nuclear power plant decommissioning. Korea Research Institute of Decommissioning (KRID) was established radionuclide analysis facilities to address above issues and support nuclear power plant decommissioning projects. The plan is to perform classification by type and radionuclide for all waste produced during nuclear power plant decommissioning and to support the disposal of radioactive wastes. In addition, we plan to establish validation methods for samples where verification methods are not established, in order to conduct efficient analysis and management. In this presentation, we will introduce the radionuclide facility currently under construction at KRID and present the space design, equipment layout, and utilization plans.

In general, radioactive waste with high radioactivity is made into a solid form with performance such as leaching restriction, shape retention, and structural stability so that radioactive waste does not affect humans and the environment as much as possible. This should be applied equally to radioactive waste, whether homogeneous or heterogeneous. The requirements are stipulated in the “Low and Intermediate Level Radioactive Waste Delivery Regulations” notice of the Korea Nuclear Safety and Security Commission. On the other hand, the waste acceptance criteria for domestic disposal facilities require immobilization of heterogeneous waste when the activity concentration is above a certain level, but do not provide specific immobilization performance requirements. In this study, the immobilization requirements applied to heterogeneous radioactive waste in various overseas countries operating low and intermediate-level radioactive waste disposal facilities were studied. First, the IAEA’s safety standards for radioactive waste immobilization, domestic regulations, and disposal facility waste acceptance criteria were reviewed. Countries operating surface disposal facilities such as the United States, France, Spain, and Japan and countries operating underground disposal facilities such as Sweden and Finland were divided to review the current status of immobilization application to heterogeneous waste in overseas countries. When reviewing overseas cases, each country’s disposal methods, types of disposal waste, and waste treatment criteria were also reviewed. It was found that the immobilization requirements for heterogeneous radioactive waste vary depending on the disposal method and the type of barrier used to ensure disposal safety in each country. The common point is to surround heterogeneous radioactive waste within a concrete lining of a certain thickness, and to apply the thickness, compressive strength, and diffusion coefficient of the concrete lining as immobilization performance requirements. Through this study, the immobilization performance requirements for heterogeneous radioactive waste in various overseas countries that stably operate low- and intermediate-level radioactive waste were confirmed, which is expected to contribute to specifying the performance requirements for immobilization of heterogeneous radioactive waste in domestic disposal facilities.

Every engineering decision in radioactive waste management should be based on both technical and economic considerations. Especially, the management of low-level radioactive waste (LLW) is more critical on economic concerns, due to its long-term and continuous nature, which emphasizes the importance of economic analysis. In this study, economic factors for LLW management were discussed with appropriate engineering applications. Two major factors that should be taken into account when assessing economic expectations are the accuracy of the results and its proper balancing with ALARA philosophy (As Low As Reasonably Achievable). The accuracy of the results depends on the correct application of alternatives within a realistic framework of waste processing. This is because the LLW management process involves variables such as component type, physical dimensions, and the monetary value at the processing date. Two commonly used alternatives are the simplified lump sum present worth and levelized annual cost methods, which are based on annual and capital costs. However, these discussions on alternatives not only pertain to the time series value of operational costs but also to future technical advancements, which are crucial for engineers. As new research results on LLW treatment emerge, proper consideration and adoption should be given to technical cost management. As safety is the core value of the entire nuclear industry, the ALARA philosophy should also be considered in the cost management of LLW. The typical cost of exposure in man-rem has ranged from $1,000 to $20,000 over the past decades. However, with increasing concerns about health and international political threats, the cost of man-rem should be subject to stricter criteria, even the balancing of costs and safety concerns is much controverse issue. Throughout the study, the importance of incorporating proper engineering insights into the assessment of technical value for the financial management of LLW was discussed. However, it’s essential to remember that financial management should not be solely assessed based on the size of expenses but rather by evaluating the current financial status, the value of money at the time, and anticipated future costs, considering the specific context and timeframe.

Activated carbon (AC) is used for filtering organic and radioactive particles, in liquid and ventilation systems, respectively. Spent ACs (SACs) are stored till decaying to clearance level before disposal, but some SACs are found to contain C-14, a radioactive isotopes 5,730 years halflife, at a concentration greater than clearance level concentration, 1 Bq/g. However, without waste acceptance criteria (WAC) regarding SACs, SACs are not delivered for disposal at current situation. Therefore, this paper aims to perform a preliminary disposal safety examination to provide fundamental data to establish WAC regarding SACs SACs are inorganic ash composed mostly of carbon (~88%) with few other elements (S, H, O, etc.). Some of these SACs produced from NPPs are found to contain C-14 at concentration up to very-low level waste (VLLW) criteria, and few up to low-level waste (LLW) criteria. As SACs are in form of bead or pellets, dispersion may become a concern, thus requiring conditioning to be indispersible, and considering VLL soils can be disposed by packaging into soft-bags, VLL SACs can also be disposed in the same way, provided SACs are dried to meet free water requirement. But, further analysis is required to evaluate radioactive inventory before disposal. Disposability of SACs is examined based on domestic WAC’s requirement on physical and chemical characteristics. Firstly, particulate regulation would be satisfied, as commonly used ACs in filters are in size greater than 0.3 mm, which is greater than regulated particle size of 0.2 mm and below. Secondly, chelating content regulation would be satisfied, as SACs do not contain chelating chemicals. Also, cellulose, which is known to produce chelating agent (ISA), would be degraded and removed as ACs are produced by pyrolysis at 1,000°C, while thermal degradation of cellulose occurs around 350~600°C. Thirdly, ignitability regulation would be satisfied because as per 40 CFR 261.21, ignitable material is defined with ignition point below 60°C, but SACs has ignition point above 350°C. Lastly, gas generation regulation would be satisfied, as SACs being inorganic, they would be targeted for biological degradation, which is one of the main mechanism of gas generation. Therefore, SACs would be suitable to be disposed at domestic repositories, provided they are securely packaged. Further analysis would be required before disposal to determine detailed radioactive inventories and chemical contents, which also would be used to produce fundamental data to establish WAC.

At the end of 2022 there were 439 nuclear power reactors in operating around the world, including 25 nuclear power reactors of Korea. Domestic nuclear power plants (NPPs) continuously produce low and intermediate-level radioactive waste (LILW) and spent nuclear fuel (SNF). As amount of radioactive waste is increasing and interim storage facilities meet limitation of their capacity, radioactive waste need to be transported. Consequently, the demand for radioactive waste transportation is increasing and the importance of Radiation Risk Assessment Codes (RRACs) for radioactive waste transportation is also on the rise. Considering the domestic situation where all NPPs are located on seaside, the radioactive waste transportation by ship is inevitable and the its risk assessment is very important for safety. Although various researches on radioactive waste transportation risk assessment is being actively conducted, research on domestic radioactive waste maritime transportation is insufficient. In this study, MARINRAD and KM-RAD were used to review on the radioactive waste transportation risk assessment. The result of reviewing shows that MARINRAD used SNF as transporting radioactive materials and ‘SAND87-7067 (1987)’ as nuclide database, whereas KMRAD used LILW and ‘IAEA Technical Report Series-422 (2004)’. To complement these limitations, we present an modernized integrated database by updating data and covering the radioactive materials from LILW to SNF. These results are expected to contribute to the development of RRACs for domestic radioactive waste maritime transportation.

This study focuses on the development of coatings designed for storage containers used in the management of radioactive waste. The primary objective is to enhance the shielding performance of these containers against either gamma or neutron radiation. Shielding against these types of radiation is essential to ensure the safety of personnel and the environment. In this study, tungsten and boron cabide coating specimens were manufactured using the HVOF (High-Velocity Oxy Fuel) technuqe. These coatings act as an additional layer of protection for the storage containers, effectively absorbing and attenuating gamma and neutron radiation. The fabricated tungsten and boron carbide coating specimens were evaluated using two different testing methods. The first experiment evaluates the effectiveness of a radiation shielding coating on cold-rolled steel surfaces, achieved by applying a mixture of WC (Tungsten Carbide) powders. WC-based coating specimens, featuring different ratios, were prepared and preliminarily assessed for their radiation shielding capabilities. In the gamma-ray shielding test, Cs-137 was utilized as the radiation source. The coating thickness remained constant at 250 μm. Based on the test results, the attenuation ratio and shielding rate for each coated specimen were calculated. It was observed that the gammaray shielding rate exhibited relatively higher shielding performance as the WC content increased. This observation aligns with our findings from the gamma-ray shielding test and underscores the potential benefits of increasing the tungsten content in the coating. In the second experiment, a neutron shielding material was created by applying a 100 μm-thick layer of B4C (Boron Carbide) onto 316SS. The thermal neutron (AmBe) shielding test results demonstrated an approximate shielding rate of 27%. The thermal neutron shielding rate was confirmed to exceed 99.9% in the 1.5 cm thick SiC+B4C bulk plate. This indicates a significant reduction in required volume. This study establishes that these coatings enhance the gamma-ray and neutron shielding effectiveness of storage containers designed for managing radioactive waste. In the future, we plan to conduct a comparative evaluation of the radiation shielding properties to optimize the coating conditions and ensure optimal shielding effectiveness.

Low- and intermediate level waste (LILW) repository in Gyeongju, Korea is in operation and the radioactive waste should satisfy the waste acceptance criteria (WAC) of the repository. Among the WAC of the Gyeongju LILW repository, the leachability index criterion is considered to be the criterion that is directly related to the isolation of the radionuclides from biosphere. Cesium, strontium, and cobalt should satisfy the leachability index larger than six by following the ANS 16.1 leaching test method. Several research were performed for the leachability index of Cs, Sr, and Co by following the ANS 16.1 leaching test method. However, the test condition of the previous research is expected to be different to the condition of the actual waste. Due to the radioactivity of the radionuclide such as Cs, Sr and Co, most of the research applied the surrogate of those radionuclides. The concentration of those nuclides was generally measured by the inductively coupled plasma (ICP) equipment, however, high concentration compared to the disposal limit of those nuclides due to the detection limit of the ICP was applied. From the Freundlich and Langmuir adsorption isotherms, the adsorption of the nuclides differs according to the concentration of the nuclides. As the leachability index of the nuclides is affected by the adsorption of the nuclides on the binding material, the effect of nuclide concentration is expected to be not ignorable. Therefore, the leachability index difference according to the nuclide concentration should be compared to avoid over- or underestimation of the leachability index. In this study, the difference in the leachability index according to the concentration of nuclides is aimed to be checked. Cs, Sr, and Co, which should satisfy the leachability index criterion in the WAC of the Gyeongju repository, were selected as target nuclides. Three concentrations were selected to compare the leachability index: 0.1 mol, 0.001 mol and below the regulatory exemption concentration. The concentration of non-radioactive nuclides in the leachant was measured by ICPOES and ICP-MS while the concentration of radionuclides was measured by HPGe. The result of this study can be applied as background data enhancing the WAC or disposal concentration limit of the radionuclides in Gyeongju LILW repository.

The decommissioning of nuclear power plants will generate a lot of low and intermediate-level radioactive waste (LILW), and preliminary radioactive evaluation for these wastes should be carried out before decommissioning work. Mainly, Concrete, Carbon Steel, Stainless Steel-304 (SS304) and Inconel are used in many parts of nuclear power plants and considered as main resource of nuclear wastes. Depending on the material location, the number of neutrons irradiated to material varies, which can range from self-disposal waste to LILW. In this paper, activation analysis was performed to compare the radiation dose according to the presence or absence of impurity elements present in SS304. For the calculation, SS304 composition and impurity elements were used as described in the report of NUREG-3474. This report lists 41 impurity elements for SS304 and other materials. Calculation code is used ORIGEN-S module in SCALE 6.1 code. Neutron flux is used as arbitrary value that around 1E+11 level and irradiation time is set as 30 year with 10-year cooling time. In the ORIGEN-S calculation, 1g of SS304 is used for easy calculation of specific activity. The ORIGEN-S calculation results are as follows. All impurity elements contained case calculated 9.32E+07 Bq activity. In the absence of all impurity elements case and most cases shows that total Becquerel value after 10-year cooling time around 9.11E+07 Bq, and Co impurity case had larger result. The calculation was performed again by increasing the amount of impurity substances by 100 times to perform the sensitivity evaluation more reliably. Representatively, Li, N, Co, and Ba impurity elements cases were calculated to have a particularly high Becquerel. Especially Co impurity element case, a total Becquerel of 3.03E+08 was calculated. Accordingly, evaluation of impurities mixed in SS304 must be considered, and in particular, the inclusion rate for Co must be considered.

There is a large amount of radioactive waste in waste storage in the Korea Atomic Energy Research Institute. Some of the radioactive waste was generated during the dismantling process due to Korea Research Reactor 1&2 and it accounts for 20% of the total waste. Radioactive waste must be reduced by appropriate disposal methods to secure storage space and to reduce disposal costs. Research Reactor wastes include wastes that are below the acceptable criteria for selfdisposal and non-contaminated wastes, so they can be treated as wastes subject to self-disposal through contamination analysis and reclassification. In order to deregulation radioactive waste, it is necessary to meet the self-disposal standards stipulated in the Domestic Nuclear Act and the treatment standards of the Waste Management Act. The main factors of deregulation are surface contaminant, radionuclide activity and dose assessment. To confirm the contamination of waste, surface contaminant and gamma nuclide analysis were performed. After homogenizing the waste sample, it was placed in 1 L Mariinelli beaker. When collecting waste samples, 1 kg per 200 kg of waste was collected. The concentrations of the major radionuclides Co-60, Cs-134, Cs-137, Eu-152, and Eu-154 were analyzed using HPGe detector. To evaluate radiation dose, various computational programs were used. A dose assessment was performed with the analyzed nuclide concentration. The concentrations of representative nuclides satisfied the deregulation acceptance criteria and the results of the dose assessment corresponding to self-disposal method was also satisfied. Based on this results, KAERI submitted the report on waste self-disposal plan to obtain approval. After final approval, Research Reactor waste is to be incinerated and incineration ash is to be buried in the designated place. Some metallic waste has been recycled. In this study, the suitability of deregulation for self-disposal was confirmed through the evaluation of the surface contaminant analysis, radionuclide concentration analysis and dose assessment.

The primary objective of this study is to evaluate a systematic design’s effectivity in remediating actual uranium-contaminated soil. The emphasis was placed on practical and engineering aspects, particularly in assessing the capabilities of a zero liquid discharge system in treating wastewater derived from soil washing. The research method involved a purification procedure for both the uranium-contaminated soil and its accompanying wastewater. Notably, the experimental outcomes demonstrated successful uranium separation from the contaminated soil. The treated soil could be self-disposed of, as its uranium concentration fell below 1.0 Bq·g−1, a level endorsed by the International Atomic Energy Agency for radionuclide clearance. The zero liquid discharge system’s significance lay in its distillation process, which not only facilitated the reuse of water from the separated filtrate but also allowed for the self-disposal of high-purity Na2SO4 within the residues of the distilled filtrate. Through a comparative economic analysis involving direct disposal and the application of a remediation process for uranium-contaminated soil, the comprehensive zero liquid discharge system emerged as a practical and viable choice. The successful demonstration of the design and practicality of the proposed zero liquid discharge system for treating wastewater originating from real uranium-contaminated soil is poised to have a lasting impact.

Ion exchange resins are commonly employed in the treatment of liquid radioactive waste generated in nuclear power plants (NPP). The ion exchange resin used in NPP is a mixed-bed ion exchange resin known as IRN-150, which is of nuclear grade. This resin is a mixture of cation exchange resin and anion exchange resin. The cation exchange resin removes cationic radionuclides such as Cs and Co, while anion exchange resin handles anions (e.g., H14CO3 -), effectively purifying the liquid waste. Spent ion exchange resins (spent resin) containing C-14 are classified as low and intermediate level radioactive waste, and their radioactivity needs to be reduced as it exceeds the disposal limit regulated by law. Therefore, the microwave technology for the removal of C-14 from spent resin has been investigated. Previous studies have successfully developed a method for the effective removal of C-14 during the resin treatment process. However, it was observed that, in this process, functional groups in the resin were also removed, resulting in the generation of off-gases containing trimethylamine. These off-gases can dissolve in water from process, increasing its pH, which can subsequently hinder the recovery of C-14. In this study, we investigated the high-purity recovery of C-14 by adjusting the moisture content within the reactor following microwave treatment. Mock spent resins, consisting of 100 g of resin with HCO3 - ion-exchanged and 0, 25, or 50 g of deionized water, were subjected to microwave treatment for 40 or 60 minutes. Subsequently, the C-14 desorption efficiency of the mock spent resins was evaluated using an acid stripping process with H3PO4 solution. The functional group status of the mock spent resins was analyzed using 15N NMR spectroscopy. The results showed that the mock spent resins exhibited efficient C-14 recovery without significant functional group degradation. The highest C-14 desorption efficiency was achieved when 25 g of deionized water was used during microwave treatment.

Domestic waste acceptance criteria (WAC) require flowable or homogeneous wastes, such as spent resin, concentrated waste, and sludge, etc., to be solidified regardless of radiation level, to provide structural integrity to prevent collapse of repository, and prevent leaching. Therefore, verylow level (VLL) spent resin (SR) would also require to be solidified. However, such disposal would be too conservative, considering IAEA standards do not require robust containment and shielding of VLL wastes. To prevent unnecessary cost and exposure to workers, current WAC advisable to be amended, thus this paper aims to provide modified regulation based on reviewed engineering background of solidification requirement. According to NRC report, SR is classified as wet-solid waste, which is defined as a solid waste produced from liquid system, thus containing free-liquid within the waste. NRC requires liquid wastes to be solidified regardless of radiation level to prevent free liquid from being disposed, which could cause rapid release of radionuclides. Furthermore, considering class A waste does not require structural integrity, unlike class B and C wastes, dewatering would be an enough measure for solidification. This is supported by the cases of Palo Verde and Diablo Canyon nuclear power plants, whose wet-solid wastes, such as concentrated wastes and sludge, are disposed by packaging into steel boxes after dewatering or incineration. Therefore, dewatering VLL spent resin and packaging them into structural secure packaging could satisfy solidification goal. Another goal of solidification is to provide structural support, which was considered to prevent collapse of soil covers in landfills or trenches. However, providing structural support via solidification agent (ex. Cement) would be unnecessary in domestic 2nd phase repository. As the domestic 2nd phase repository is cementitious structure, which is backfilled with cement upon closure, the repository itself already has enough structural integrity to prevent collapse. Goldsim simulation was run to evaluate radiation impact by VLL SR, with and without solidification, by modelling solidified wastes with simple leaching, and unsolidified wastes with instant release. Both simulations showed negligible impact on radiation exposure, meaning that solidifying VLL SR to delay leaching would be irrational. Therefore, dewatering VLL SR and packaging it into a secure drum (ex. Steel drum) could achieve solidification goals described in NRC reports and provide enough safety to be disposed into domestic repositories. In future, the studied backgrounds in this paper should be considered to modify current WAC to achieve efficient waste management.

Nuclear power plants in Korea stores approximately 3,800 drums of paraffin solidification products. Due to the lack of homogeneity, these solidification products are not allowed to be disposed of. There is therefore a need for the separation of paraffin from the solidification products. This work developed an equipment for a selective separation of paraffin from the solidification product using the vacuum evaporation and condensational recovery method in a closed system. The equipment mainly consists of a vacuum evaporator and a condensational deposition recovery chamber. Nonisothermal vacuum TGAs, kinetic analyses and kinetic predictions were conducted to set appropriate operation conditions. Its basic operability under the established conditions was first confirmed using pure paraffin solid. Simulated paraffin solidification product fixing dried boric acid waste including nonradioactive Co and Cs were then fabricated and tested for the capability of selective separation of paraffin from the simulated waste. Paraffin was selectively separated without entertainment of Co and Cs. It was confirmed that the developed equipment could separate and recover paraffin in the form of nonradioactive waste.