Highly radioactive waste is solidified to restrict leaching, retain its shape, and maintain its structural stability to prevent it from affecting humans and the environment as much as possible. This operation should be performed consistently regardless of whether the waste is homogeneous or heterogeneous. However, currently, there are no specific performance requirements for heterogeneous waste in Korea. This study reviewed domestic research results and the status of overseas applications, and proposed immobilization requirements for heterogeneous waste to be applied in Korea. IAEA safety standards, domestic laws, and waste acceptance criteria were reviewed. The status of heterogeneous waste immobilization in countries such as the United States, France, and Spain was reviewed. Most countries treat heterogeneous waste by encasing it in concrete, and impose immobilization requirements on this concrete. Based on these data, safety standards for the thickness, compressive strength, and diffusion limit of this concrete material were proposed as immobilization requirements for heterogeneous waste disposal in Korea. Quantitative values for the above requirements need to be derived through quantitative assessments based on the characteristics of domestic heterogeneous waste and disposal facilities.

In this research, a detailed analysis of the decay heat contributions of both actinides and non-actinides (fission fragments) from spent nuclear fuel (SNF) was made after 50 GWd·tHM−1 burnup of fresh uranium fuel with 4.5% enrichment lasted for 1,350 days. The calculations were made for a long storage period of 300 years divided into four sections 1, 10, 100, and 300 years so that we could study the decay heat and physical disposal ratios of radioactive waste in medium- and long-term storage periods. Fresh fuel burnup calculations were made using the code MCNP, while isotopic content and then decay heat were calculated using the built-in stiff equation solver in the MATLAB code. It is noted that only around 12 isotopes contribute more than 90% of the decay heat at all times. It is also noted that the contribution of actinides persists and is the dominant ether despite decreasing decay heat, while the effect of fission products decreases at a very rapid rate after about 40 years of storage.

The initial development plans for the six reactor designs, soon after the release of Generation IV International Forum (GIF) TRM in 2002, were characterized by high ambition [1]. Specifically, the sodium-cooled fast reactor (SFR) and very-high temperature reactor (VHTR) gained significant attention and were expected to reach the validation stage by the 2020s, with commercial viability projected for the 2030s. However, these projections have been unrealized because of various factors. The development of reactor designs by the GIF was supposed to be influenced by events such as the 2008 global financial crisis, 2011 Fukushima accident [2, 3], discovery of extensive shale oil reserves in the United States, and overly ambitious technological targets. Consequently, the momentum for VHTR development reduced significantly. In this context, the aims of this study were to compare and analyze the development progress of the six Gen IV reactor designs over the past 20 years, based on the GIF roadmaps published in 2002 and 2014. The primary focus was to examine the prospects for the reactor designs in relation to spent nuclear fuel burning in conjunction with small modular reactor (SMR), including molten salt reactor (MSR), which is expected to have spent nuclear fuel management potential.

The aim of this study is to ensure the structural integrity of a canister to be used in a dry storage system currently being developed in Korea. Based on burnup and cooling periods, the canister is designed with 24 bundles of spent nuclear fuel stored inside it. It is a cylindrical structure with a height of 4,890 mm, an internal diameter of 1,708 mm, and an inner length of 4,590 mm. The canister lid is fixed with multiple seals and welds to maintain its confinement boundary to prevent the leakage of radioactive waste. The canister is evaluated under different loads that may be generated under normal, off-normal, and accident conditions, and combinations of these loads are compared against the allowable stress thresholds to assess its structural integrity in accordance with NUREG-2215. The evaluation result shows that the stress intensities applied on the canister under normal, off-normal, and accident conditions are below the allowable stress thresholds, thus confirming its structural integrity.

The objectives of this paper are: (1) to conduct the thermal analyses of the disposal cell using COMSOL Multiphysics; (2) to determine whether the design of the disposal cell satisfies the thermal design requirement; and (3) to evaluate the effect of design modifications on the temperature of the disposal cell. Specifically, the analysis incorporated a heterogeneous model of 236 fuel rod heat sources of spent nuclear fuel (SNF) to improve the reality of the modeling. In the reference case, the design, featuring 8 m between deposition holes and 30 m between deposition tunnels for 40 years of the SNF cooling time, did not meet the design requirement. For the first modified case, the designs with 9 m and 10 m between the deposition holes for the cooling time of 40 years and five spacings for 50 and 60 years were found to meet the requirement. For the second modified case, the designs with 35 m and 40 m between the deposition tunnels for 40 years, 25 m to 40 m for 50 years and five spacings for 60 years also met the requirement. This study contributes to the advancement of the thermal analysis technique of a disposal cell.

A transfer cask serves as the container for transporting and handling canisters loaded with spent nuclear fuels from light water reactors. This study focuses on a cylindrical transfer cask, standing at 5,300 mm with an external diameter of 2,170 mm, featuring impact limiters on the top and bottom sides. The base of the cask body has an openable/closable lid for loading canisters with storage modules. The transfer cask houses a canister containing spent nuclear fuels from lightweight reactors, serving as the confinement boundary while the cask itself lacks the confinement structure. The objective of this study was to conduct a structural analysis evaluation of the transfer cask, currently under development in Korea, ensuring its safety. This evaluation encompasses analyses of loads under normal, off-normal, and accident conditions, adhering to NUREG-2215. Structural integrity was assessed by comparing combined results for each load against stress limits. The results confirm that the transfer cask meets stress limits across normal, off-normal, and accident conditions, establishing its structural safety.

With South Korea increasingly focusing on nuclear energy, the management of spent nuclear fuel has attracted considerable attention in South Korea. This study established a novel procedure for selecting safety-relevant radionuclides for long-term safety assessments of a deep geological repository in South Korea. Statistical evaluations were performed to identify the design basis reference spent nuclear fuels and evaluate the source term for up to one million years. Safety-relevant radionuclides were determined based on the half-life criteria, the projected activities for the design basis reference spent nuclear fuel, and the annual limit of ingestion set by the Nuclear Safety and Security Commission Notification No. 2019-10 without considering their chemical and hydrogeological properties. The proposed process was used to select 56 radionuclides, comprising 27 fission and activation products and 29 actinide nuclides. This study explains first the determination of the design basis reference spent nuclear fuels, followed by a comprehensive discussion on the selection criteria and methodology for safety-relevant radionuclides.

Within the air purification system of a nuclear power plant, specific radioactive isotopes are extracted from gases through adsorption onto activated carbon. To properly dispose of used activated carbon, it is essential to determine the concentration of radioactive nuclides within it. This study discusses the application of the pyrolysis method for analyzing the concentrations of 3H and 14C in spent activated carbon. The pyrolysis was conducted using Raddec’s Pyrolyser, with adjustments made to parameters such as temperature profiles, airflow rates, sample quantities, and trapping solution volumes. The evaluation method for the pyrolysis of activated carbon to analyze 3H and 14C involved adding 3H and 14C sources to the activated carbon before use and subsequently assessing the recovery rates of the added sources in comparison to the analysis results.

Currently, non-volatile nuclides such as 94Nb, 99Tc, 90Sr, 55Fe, and 59/63Ni are used a sequential separation. In this study, we developed a separation for 99Tc and 90Sr by a carbonate precipitation. Sodium Carbonate (Na2CO3) was inserted in the aqueous sample from a Dry Active Waste (DAW) and a carbonate precipitation was produced. The precipitate is composed of di- or tri-valent element such as Co, Sr, Fe, Ni and the supernatant is composed of mono-valent element (Cs) and anion materials (ReO4 -, TcO4 -). In DAW, it was confirmed that the recovery of 90Sr (precipitate) and 99Tc (supernatant) were > 90%, respectively. The precipitate and supernatant separated by using a Sr-resin and an anion-exchange resin, respectively. The separated samples were measured by a Liquide Scintillation Counter (LSC, 90Sr) and Induced-Coupled Plasma-Mass Spectroscopy (ICPMS, 99Tc).

Bis (2-ethylhexyl)phosphoric acid (HDEHP) is a renowned extractant, favored for its affinity to selectively remove uranium via its P=O groups. We previously synthesized HDEHP-functionalized mesoporous silica microspheres for solid-phase uranium adsorption. Herein, we investigated the kinetic and isothermal behavior of uranyl ion adsorption in mesoporous silica microspheres functionalized with phosphate groups. Adsorption experiments were conducted by equilibrating 20 mg of silica samples with 50 mL of uranium solutions, with concentrations ranging from 10 to 100 mgU L−1 for isotherms and 100 mgU L−1 for kinetics. Three distinct samples were prepared with varying HDEHP to TEOS molar ratios (x = 0.16 and 0.24) and underwent hydrothermal treatment at different temperatures, resulting in distinct textural properties. Contact times spanned from 1 to 120 hours. For x = 0.16 samples, it took around 50 and 11 hours to reach equilibrium for the hydrothermally treated samples at 343 K and 373 K, respectively. Adsorbed quantities were similar (99 and 101 mg g-1, respectively), indicating consistent functional group content. This suggests that the key factor influencing uranium adsorption kinetics is pore size of the silica. The sample treated at 373 K, with a larger pore size (22.7 nm) compared to 343 K (11.5 nm), experienced less steric hindrance, allowing uranium species to diffuse more easily through the mesopores. The data confirmed the excellent fit of pseudo-second-order kinetic model (R2 > 0.999) and closely matched the experimental value, suggesting that chemisorption governs the rate-controlling step. To gain further insights into uranium adsorption behavior, we conducted an adsorption isotherm analysis at various initial concentrations under a constant pH of 4. Both the Langmuir and Freundlich isotherm models were applied, with the Langmuir model providing a superior fit. The relatively high R2 value indicated its effectiveness in describing the adsorption process, suggesting homogenous sorbate adsorption on an energetically uniform adsorbent surface via a monolayer adsorption and constant adsorption site density, without any interaction between adsorbates on adjacent sites. Remarkably, differences in surface area did not significantly impact uranium removal efficiency. This observation strongly suggests that the adsorption capacity is primarily governed by the loading amount of HDEHP and the inner-sphere complexation with the phosphoryl group (O=P). Our silica composite exhibited an impressive adsorption capacity of 133 mg g-1, surpassing the results reported in the majority of other silica literature.

The ultimate objective of deep geological repositories is to achieve complete segregation of hazardous radioactive waste from the biosphere. Thus, given the possibility of leaks in the distant future, it is crucial to evaluate the capability of clay minerals to fulfill their promising role as both engineered and natural barriers. Selenium-79, a long-lived fission product originating from uranium- 235, holds significant importance due to its high mobility resulting from the predominant anionic form of selenium. To investigate the retardation behaviors of Se(IV) in clay media by sorption, a series of batch sorption experiments were conducted. The batch samples consisted of Se(IV) ions dissolved in 0.1 M NaCl solutions, along with clay minerals including kaolinite, montmorillonite, and illite-smectite mixed layers. The pH of the samples was also varied, reflecting the shift in the predominant selenium species from selenious acid to selenite ion as the environment can shift from slightly acidic to alkaline conditions. This alteration in pH concurrently promotes the competition of hydroxide ions for Se(IV) sorption on the mineral surface as the pH increases and impedes the selective attachment of selenium. The acquired experimental data were fitted through Langmuir and Freundlich sorption isotherms. From the Freundlich fit data, the distribution coefficient values of Se(IV) for kaolinite, montmorillonite, and illite-smectite mixed layer were derived, which exhibited a clear decrease from 91, 110, 62 L/kg at a pH of 3.2 to 16, 6.3, 12 L/kg at a pH of 7.5, respectively. These values derived over the pH range provide quantitative guidance essential for the safety assessment of clay mineral barriers, contributing to a more informed site selection process for deep geological repositories.

In order to establish disposal plans for sludge, which is one of the untreated waste materials from domestic nuclear power plants, it is necessary to determine the radioactivity concentration of radioactive isotopes. In this study, we aim to evaluate the gross alpha radioactivity of sludge containing radioactive contaminants after pre-treatment, in order to assess the level of sludge waste and obtain analytical data for discussing disposal methods. Samples of sludge generated from nuclear power plants were pre-treated, solutionized, and prepared as analysis samples for evaluating the gross alpha radioactivity.

Typically, the bottom of the effluent treatment facility at a nuclear power plant contains sediment, which is low-contamination waste consisting of sludge, gravel, sand, and other materials from which radioactive contaminants have been removed. Among these sediments, sludge is an irregular solid form consisting of small particles that are coagulated together, with radioactive isotopes containing cobalt attached. Currently, there is a record of disposing of dry active waste from domestic nuclear power plants, and efforts are underway to gather basic data for the disposal of untreated waste such as sludge, spent filter, and spent resin. In particular, the classification and disposal methods of waste will be determined based on the radioactivity concentration. Therefore, plans are being made to determine the radioactivity concentration of radioactive isotopes and establish disposal plans for sludge samples. In this study, pre-treatment and solutionization were carried out for the analysis of radioactive isotopes in sludge sampels from nuclear power plants. The deviation of the gamma radioisotope analysis results was derived to obtain an optimal sample quantity that represents the sludge.

For the disposition of radioactive wastes generated from nuclear power plant, radioisotope inventory must be analyzed to determine an activity concentration of radionuclides. Radionuclides in low- and intermediate-low-level of radioactive wastes, however, can be easily classified to easyto- measure (ETM) and difficult-to-measure (DTM) nuclides. ETM nuclides are gamma emitting nuclides that is relatively easy to measure because they do not need to be destroyed for the preprocessing. On the other hands, DTM nuclides are alpha and beta emitting nuclides that need to be destroyed for the preprocessing and also need chemical separation. Currently, measurement methods for DTM nuclides are developed and in this paper measurement methods of Fe-55, Ni-59, Ni-63, Sr-90 and Tc-99 will be introduced.

Alpha activities can be used for categorization, transportation, and disposal of radioactive waste generated from the operation of nuclear facilities including nuclear power plants. In order to transport and dispose of such low- and intermediate-level radioactive waste (LILW) to the Wolsong LILW Disposal Center (WLDC) at Gyeongju, the gross alpha concentration of an individual drum should be determined according to the acceptance criteria. In addition, when the gross alpha concentration exceeds 10 Bq/g, the inventory of the comprising alpha emitters in the waste is to be identified. Gross alpha measurements using a proportional counter are usually straightforward, inexpensive, and high-throughput, so they are broadly used to assay the total alpha activity for environmental, health physics, and emergency-response assessments. However, several factors are thoughtfully considered to obtain a reliable approximate for the entire alpha emitters in a sample, which include the alpha particle energy of a particular radionuclide, the radionuclide that is used as a calibration standard, the uniformity of film in a planchet, time between sample collection and sample preparation, and time between sample preparation and counting. Korea Atomic Energy Research Institute (KAERI) have evaluated the inventory of radionuclides in low-level radioactive waste drums to send every year hundreds of them to the WLDC. In this presentation, we revisit the gross alpha measurement results of the drums transported to WLDC in the past few years and compare them with the concentrations of alpha emitters measured from alpha spectrometry and gamma spectrometry. This study offers an insight into the gross alpha measurement for radioactive waste regarding calibration source, self-absorption effect, composition of alpha emitters, etc.

To achieve permanent disposal of radioactive waste drums, the radionuclides analysis process is essential. A variety of waste types are generated through the operation of nuclear facilities, with dry active waste (DAW) being the most abundant. To perform radionuclides analysis, sample pretreatment technology is required to transform solid samples into solutions. In this study, we developed a dry ashing-microwave digestion method and secured the reliability of the analysis results through a validity evaluation. Additionally, we conducted a comparative analysis of the radioactivity of 94Nb nuclides with and without the chemical separation process, which reduced the minimum detectable activity (MDA) level by more than 65-fold for a certain sample.

To ensure the maintenance of the nuclear emergency response system, it is important to periodicaly conduct hazard assessments using up-to-date input variables. The results of this review are apllied to drills and exercises, enabling the inspection of emergency plan and response procedures. Therefore, this study aims to analyze off-site consequences according to the occurrence time of the Design Basis Accident (DBA) for the Hanaro Fuel Fabrication Facility (HFFF) by using the recent site-specific meteorological data and to review the appropriateness of urgent protective measures. MELCOR and SafeHanaro computer codes were used for radiation source-term estimation and environmental impact assessment, respectively. It was assumed that radioactive materials are released into environment for 2 hours due to the fire during the nuclear fuel sieving process. The following 12 scenarios for each occurrence time period was selected (0 am, 2 am, 4 am, 6 am, 8 am, 10 am, 12 pm, 2 pm, 4 pm, 6 pm, 8 pm, 10 pm) and the effective dose and thyroid dose in earlyand intermediate-phase were assessed. As a result, the most severe exposure-induced accident scenario is found to be as occurring at 0 am on July 15th, with the Most Exposed Individual (MEI) positioned 200 meters downwind from the facility. The committed effective dose for MEI is identified as to be 2.97E-02 mSv which has a significant margin against the IAEA's (Generic Intervention Level) GIL and (Generic Criteria) GC. During the passage of the radio-active plume, the estimated effective dose and thyroid dose due to inhalation were 2.97E-02 mSV (99.99%) and 5.06E-05 mSv (99.77%), respectively. External exposure appeared to be negligible. Meanwhile, the thyroid dose is noticeably below the criteria for decision-making for distribution of Potassium Iodide (KI). Accordingly, in order for local residents to participate in the exercise and drills, it is essential to develop scenarios considering simultaneous emergencies at multi-facilities and latenight accidents. In conclusion, this results will be used to improve the exercise plans for enhancing the nuclear or radiological emergency competencies of the KAERI.

The objective of this study is to investigate the safety awareness and effectiveness of the education and training for employees engaged in radiological emergency organization of the Korea Atomic Energy Research Institute (KAERI). In 2022, the questionnaire for the education satisfaction survey was revised to regulary evaluate the effect of edcation on perceptions of importance on emergency preparedness for nuclear research facilities. In line with, a standard questionnaire was created which covers 3 factors and 9 attributes, and the evaluation indicatior is based on a 5-point Likert scale. In 2023, the education on radiological emergency preparedness was conducted for 235 emergency staff. From May 24 to July 13, 2023, data was collected from a total of 235 emergency response personnels, including 28 new staffs and 207 maintenance staffs. Aa a result of response analysis, it was identified that education for radiological emergency response had a significant correlation with the promoting safety culture. It was found that senior emergency personnel with more years of experience are highly interested in radioactive disaster prevention and actively participate in and training. On the other hand, it was presented that new and less experienced groups tend to have a relatively high scored of the risk perception of nuclear research facilitites. Therefore, it is necessary to improve the practical curriculum in order to increase the participation of junior disaster prevention personnel in education and training, ensuring that they correctly recognize the risk of research facilities. This results are expected to be used to improve the quality of education and drills for radiological emergency response at KAERI.