Economical radioactive soil treatment technology is essential to safely and efficiently treat of high-concentration radioactive areas and contaminated sites during operation of nuclear power plants at home and abroad. This study is to determine the performance of BERAD (Beautiful Environmental construction’s RAdioactive soil Decontamination system) before applying magnetic nanoparticles and adsorbents developed by the KAERI (Korea Atomic Energy Research Institute) which will be used in the national funded project to a large-capacity radioactive soil decontamination system. BERAD uses Soil Washing Process by US EPA (402-R-007-004 (2007)) and can decontaminate 0.5 tons of radioactive soil per hour through water washing and/or chemical washing with particle size separation. When contaminated soil is input to BERAD, the soil is selected and washed, and after going through a rinse stage and particle size separation stage, it discharges decontaminated soil separated by sludge of less than 0.075 mm. In this experiment, the concentrations of four general isotopes (A, B, C, and D which are important radioisotopes when soil is contaminated by them.) were analyzed by using ICP-MS to compare before and after decontamination by BERAD. Since BERAD is the commercial-scale pilot system that decontaminates relatively large amount of soil, so it is difficult to test using radioactive isotopes. So important general elements such as A, B, C, and D in soil were analyzed. In the study, BERAD decontaminated soil by using water washing. And the particle size of soil was divided into a total of six particle size sections with five sieves: 4 mm, 2 mm, 0.850 mm, 0.212 mm, and 0.075 mm. Concentrations of A, B, C, and D in the soil particles larger than 4 mm are almost the lowest regardless of before and after decontamination by BERAD. For soil particles less than 4 mm, the concentrations of C and D decreased constantly after BERAD decontamination. On the other hand, the decontamination efficiency of A and B decreased as the soil particle became smaller, but the concentrations of A and B increased for the soil particle below 0.075 mm. As a result, decontamination efficiency of one cycle using BERAD for all nuclides in soil particles between 4 mm and 0.075 mm is about 45% to 65 %.

The process of carbonization followed by a high-temperature halogenation removal of radionuclides is a promising approach to convert low-radioactivity spent ion-exchange (IE) resins into freereleasable non-radioactive waste. The first step of this process is to convert spent ion-exchange resins into the carbon granules that are stable under high-temperature and corrosive-gas flowing conditions. This study investigated the kinetics of carbonization of cation exchange resin (CER) and the changes in structures during the course of carbonization to 1,273 K. Both of model-free and modelfitted kinetic analysis of mixed reactions occurring during the course of carbonization were first conducted based on the non-isothermal TGAs and TGA-FTIR analysis of CER to 1,272 K. The structural changes during the course of carbonization were investigated using the high-resolution FTIR and C-13 NMR of CER samples pyrolyzed to the peak temperature of each reaction steps established by the kinetic analysis. Four individual reaction steps were identified during the course of carbonization to 1,273 K. The first and the third steps were identified as the dehydration and the dissociation of the functional group of —SO3-H+ into SO2 and H2O, respectively. The second and the fourth steps were identified as the cleavage of styrene divinyl benzene copolymer and carbonization of pyrolysis product after the cleavage, respectively. The temperature and time positions of the peaks in the DTG plot are nearly identical to those of the peaks of the Gram Schmidt intensity of FTIR. The structural changes in carbonization identified by high-resolution FTIR and DTG are in agreement with those by C-13 NMR. The results of a detailed examination of the structural changes according to NMR and FTIR were in agreement with the pyrolysis gas evolution characteristics as examined by TGA-FTIR.

KORAD (Korea Radioactive Waste Agency, http://www.korad.or.kr) has stored slightly contaminated ascon (asphalt coated concrete mixture) that was introduced to Gyeongju repository about a decade ago waiting for a final disposal. It is believed to be mainly contaminated by radioisotope 137Cs due to impurities introduced from the outside during the ascon manufacturing process. We studied characteristics of the radioactive waste to see whether this material would be proper enough to be disposed in Gyeongju LILW repository or be other ways to reduce the disposal volume including self-disposal before its final disposal otherwise. KORAD looked into the properness of characteristics of ascon in terms of WAC (Waste Acceptance Criteria) documented by KORAD that includes general chemical and physical properties of asphalt, density, size of grains, content of organic material and possibility of existence of chelate materials that qualitatively limited to be disposed by the criteria. And other associated characteristics such as gas generation and bio degradation were also investigated. Based on the data obtained from the study, we proposed various plausible solutions in associated with operational and disposal safety and economic view points. This study will be used for KORAD’s decision on how to control and safely dispose the spent ascon within a reasonable time period. And also those experiences may be applied for other LILW issues that require treatment or conditioning of radioactive wastes in the future.

Wet solid wastes including spent ion exchange resins, evaporator concentrates and sludges require solidification to transform wastes into an acceptable solid, monolithic form for final disposal. The development of the process control program for the solidification of radioactive sludges generated at nuclear power plants has been in progress to provide reasonable assurance that the solidified product will meet the established waste acceptance criteria for solidified waste. A mobile solidification system to produce the solidified waste in the size of a 200 L drum was used, which adopts the in-line mixing method where the waste and binder are mixed and then transferred to the disposable container. To simulate radioactive sludges, non-radioactive sludges are synthesized and the specimens are prepared by using them. The qualification tests on the prepared specimens including the compressive strength test, the thermal cycling test, the irradiation test, the leach test, the immersion test, etc. have been performed to qualify recipes for a range of waste compositions. The results of the tests will be analyzed and discussed.

Radioactive waste (hereinafter referred to as mixed waste) containing hazardous substances (heavy metals, organic and inorganic waste liquids, asbestos, etc.) has been continuously generated from domestic nuclear power plants, nuclear facilities, and other industrial facilities, and heavy metals were released during the dismantlement of Kori Unit 1 and Wolseong Unit 1. Lead, cadmium, mercury, arsenic), asbestos, decontamination waste liquid (organic/inorganic waste liquid), etc. may be generated. Although hazardous waste related to the nuclear industry continues to be generated, only the regulation direction for hazardous substances is presented in the provisions related to hazardous substances in the delivery regulations for low and intermediate-level radioactive waste and the acceptance criteria for low and intermediate-level radioactive waste disposal facilities. In particular, because there is no clear definition of “hazardousness” and specific standards such as concentration and characteristics for classification of hazardous substances, as well as hazard removal procedures when the hazardousness of radioactive waste is confirmed, no hazardous substances have been delivered in Korea to date and many mixed wastes are stored at each generation facility or at the NPP. As a plan to improve delivery standards related to mixed waste is being prepared recently, it is believed that if the acceptance standards are revised accordingly, it will be possible to confirm the suitability for disposal of drums produced after the establishment of the acceptance standards in 2015. However, it is believed that securing disposal suitability for waste that was packed in 200L drums and compressed under super high pressure in the absence of specific technical standards and regulatory guidelines for the disposal of radioactive waste containing hazardous substances would still remain a difficult problem. In this report overseas acceptance standards related to hazardous waste were reviewed and a plan to secure the disposal suitability of 200 L drums compressed with of super high pressure was proposed.

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

In order to apply indirect methods (such as scaling factors) to assess the radionuclide inventory of waste generated by nuclear power plants, it is essential to first evaluate the correlation coefficient between key radionuclides and those that are difficult to measure (DTM). The benchmark for the correlation coefficient (r) applied in indirect assessments is set at 0.6, and its significance can vary based on both its value and the size of the dataset. For instance, deriving a correlation coefficient using three data points versus utilizing a dataset with a hundred data points would yield different implications. This study addresses the variance in correlation coefficients based on data selection and presents a methodology for validating the significance of these coefficients. Additionally, we will discuss how these variances may impact the results of indirect assessments, such as scaling factor evaluations.

To effectively assess the inventory of radionuclides generated from nuclear power plants using a consistent evaluation method across diverse groups, it is imperative to analyze the similarity in radioactive distribution between these groups. Various methodologies exist for evaluating this similarity, and the application of statistical approaches allows us to establish similarity at a specific confidence level while accounting for the dataset size (degrees of freedom). Initially, if the variance characteristics of the two groups are similar, a t-test for equal variances can be employed. However, if the variance characteristics differ, methods for unequal variances should be applied. This study delineates the approach for assessing the similarity in radioactive distribution based on the analytical characteristics of the two groups. Furthermore, it delves into the results obtained through two case studies to offer insights into the assessment process.

For the disposal of radioactive waste from nuclear facilities, assessing their radioactivity inventories is essential. As a result, countries with nuclear facilities are implementing assessment schemes tailored to their respective policies and available resources for radioactive waste management. This paper specifically describes the assessment scheme for radioactivity inventory applied to metal waste generated during the dismantling of the Japan Power Demonstration Reactor (JPDR), a 1.25 MW BWR. The distinctive aspect of the Japanese approach lies in the fact that, for a pair of a key nuclide and a difficult-to-measure (DTM) nuclide that lack a significant correlation in their concentrations, the mean activity concentration method was used. In this method, an arithmetic average of all measurements of the DTM nuclide from representative drums, including MDAs (Minimum Detectable Activities), was assigned to the concentration of the DTM nuclide for all drums, regardless of the concentration of its paired key nuclide. Conversely, for a specific pair of a key nuclide and a DTM nuclide with a significant correlation, the scaling factor method was applied, as is common in many other countries. This Japanese case can serve as a valuable reference for Korea, which does not have the option of using the mean activity concentration method in its assessment scheme.

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.

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.

Spent ion exchange resins have been generated during the operation of nuclear facilities. These resins include radioactive nuclides. It is needed to fabricate them into a stable form for final disposal. Cement solidification process is a useful method for the fabrication of them into a waste form for final disposal. In this study, proper conditions for the fabrication of them into a stable waste form were determined using the cement solidification process. In-drum waste forms were then produced at the conditions, where the stability of representative samples was evaluated for final disposal. The samples were satisfied to the Waste Acceptance Criteria for low and intermediate level radioactive waste disposal sites. This result can be utilized to derive optimal conditions for the fabrication of spent ion exchange resins into a final disposal form.

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.

Activated carbon (AC), extensively used across various industrial sectors, serves as a sponge for different types of gases due to its porous carbon material. These gases are attracted to the carbon substrate via van der Waals forces. In nuclear power plants, AC is commonly used to adsorb radioactive gases such as 86Kr and 134Xe, as well as radioiodine sources like 131I and 133I from gaseous effluents. Even if the adsorbed radioactive gases and radioiodine decay into non-radioactive elements, the spent AC still contains radioactive species with long half-lives, such as 3H (Tritium, T) and 14C (radiocarbon). Minimizing and separating waste that contains long-lived nuclides (e.g., 14C) are pivotal components of an efficient waste management approach. A challenging aspect of effectively managing disposed AC is to minimize long-lived radioactive substances by eliminating them. This paper explores and summarizes the technology used to remove pollutants (3H, 14C) trapped within the pores of Activated carbon through thermochemical vacuum and surface oxidation processes. By recycling and reusing spent Activated carbon, we anticipate a reduction in the volume of radioactive waste, leading to decreased disposal costs. Furthermore, this paper will contribute as a valuable reference in future studies, enhancing the understanding of vacuum thermal desorption and surface oxidation of used Activated carbon.

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.

Nuclear power plants use ion exchange resins to purify liquid radioactive waste generated while operating nuclear power plants. In the case of PHWR, ion exchange resins are used in heavy water and dehydration systems, liquid waste treatment systems, and heavy water washing systems, and the used ion exchange resins are stored in waste resin storage tanks. The C-14 radioactivity concentration in the waste resin currently stored at the Wolseong Nuclear Power Plant is 4.6×106 Bq/g, exceeding the low-level limit, and if all is disposed of, it is 1.48×1015 Bq, exceeding the total limit of 3.04×1014 Bq of C-14 in the first stage disposal facility. Therefore, disposal is not possible at domestic low/medium-level disposal facilities. In addition, since the heavy water reactor waste resin mixture is stored at a ratio of about 20% activated carbon and zeolite mixture and about 80% waste resin, mixture extraction and separation technology and C-14 desorption and adsorption technology are required. Accordingly, research and development has been conducted domestically on methods to treat heavy water waste resin, but the waste resin mixture separation method is complex and inefficient, and there are limitations in applying it to the field due to the scale of the equipment being large compared to the field work space. Therefore, we would like to introduce a resin treatment technology that complements the problems of previous research. Previously, the waste resin mixture was extracted from the upper manhole and inspection hole of the storage tank, but in order to improve limitations such as worker safety, cost, and increased work time, the SRHS, which was planned at the time of nuclear power plant design, is utilized. In addition, by capturing high-purity 14CO2 in a liquid state in a high-pressure container, it ensures safety for long-term storage and is easy to handle when necessary, maximizing management efficiency. In addition, the modularization of the waste resin separation and withdrawal process from the storage tank, C-14 desorption and monitoring process, high-concentration 14CO2 capture and storage process, and 14CO2 adsorption process enables separation of each process, making it applicable to narrow work spaces. When this technology is used to treat waste resin mixtures in PHWR, it is expected to demonstrate its value as customized, high-efficiency equipment that can secure field applicability and safety and reflect the diverse needs of consumers according to changes in the working environment.

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.