Ammonia is a potential fuel for producing and storing hydrogen, but its usage is constrained by the high cost of the noble metal catalysts to decompose NH3. Utilizing non-precious catalysts to decompose ammonia increases its potential for hydrogen production. In this study, carborundum (SiC)-supported cobalt catalysts were prepared by impregnating Co3O4 nanoparticles (NPs) on SiC support. The catalysts were characterized by high-resolution transmission electron microscope, X-ray photoelectron spectroscopy, temperature programmed reduction, etc. The results show that the large specific surface area of SiC can introduce highly distributed Co3O4 NPs onto the surface. The amount of Co in the catalysts has a significant effect on the catalyst structure, particle size and catalytic performances. Due to the interaction of cobalt species with SiC, the 25Co/SiC catalyst provided the optimal ammonia conversion of 73.2% with a space velocity of 30,000 mL gcat −1 h− 1 at 550 °C, corresponding to the hydrogen production rate of 24.6 mmol H2 gcat −1 min− 1. This research presents an opportunity to develop highly active and cost-effective catalysts for hydrogen production via NH3 decomposition.

In this study, a bipolar visible light responsive photocatalytic fuel cell (PFC) was constructed by loading a Z-scheme g-C3N4/ carbon black/BiOBr and a Ti3C2/ MoS2 Schottky heterojunction on the carbon brush to prepare the photoanode and photocathode, respectively. It greatly improved the electron transfer and achieved efficient degradation of organic pollutants such as antibiotics and dyes simultaneously in two chambers of the PFC system. The Z-scheme g-C3N4/carbon black/BiOBr formed by adding highly conductive carbon black to g-C3N4/BiOBr not only effectively separates the photogenerated carriers, but also simultaneously retains the high reduction of the conduction band of g-C3N4 and the high oxidation of the valence band of BiOBr, improving the photocatalytic performance. The exceptional performance of Ti3C2/ MoS2 Schottky heterojunction originated from the superior electrical conductivity of Ti3C2 MXene, which facilitated the separation of photogenerated electron–hole pairs. Meanwhile, the synergistic effect of the two photoelectrodes further improved the photocatalytic performance of the PFC system, with degradation rates of 90.9% and 99.9% for 50 mg L− 1 tetracycline hydrochloride (TCH) and 50 mg L− 1 rhodamine-B (RhB), respectively, within 180 min. In addition, it was found that the PFC also exhibited excellent pollutant degradation rates under dark conditions (79.7%, TCH and 97.9%, RhB). This novel pollutant degradation system is expected to provide a new idea for efficient degradation of multiple pollutant simultaneously even in the dark.

Bacterial metabolisms influence the behavior of uranium (U) in deep geological repository (DGR) system because bacteria are ubiquitous in the natural environment. Nevertheless, most studies for the U(VI) bioreduction have focused on a few model bacterium, such as Shewanella putrefaciens, Desulfovibrio desulfuricans, and Geobacter sulfurreducens. In this study, the potential of aqueous U(VI) ((U(VI)aq) reduction by indigenous bacteria was examined under anaerobic conditions with addition of 20 mM sodium acetate for 24 weeks. Three different indigenous bacterial communities obtained from granitic groundwater at depths of 44–60 m (S1), 92–116 m (S2), and 234–244 m (S3) were applied for U(VI)aq reduction experiments. The S2 groundwater contained the highest U concentration of 885.4 μg/L among three groundwater samples, where U mainly existed in the form of Ca2UO2(CO3)3(aq). The S2 groundwater amended 20 mM of sodium acetate was used for the U(VI)aq bioreduction experiment. Variations in the U(VI)aq concentration and redox potential were monitored for 24 weeks to compare U(VI)aq removal efficiency in response to indigenous bacteria. The U(VI)aq removal efficiencies varied among three indigenous bacteria: 57.8% (S3), 43.1% (S2), and 37.7% (S1). The presence of the thermodynamically stable uranyl carbonate complex resulted in the incomplete U(VI)aq removal. Significant shifts in indigenous bacterial communities were observed through highthroughput 16S rRNA gene sequencing analysis. Two SRB species, Thermodesulfovibrio yellowstonii and Desulfatirhabdium butyrativorans, were dominant in the S3 sample after the anaerobic reaction, which enhanced the bioreduction of U(VI)aq. The precipitates produced by bacterial activity were determined to be U(IV)-silicate nanoparticles by a transmission electron microscope (TEM)-energy dispersive spectroscope (EDS) analysis. These results demonstrated that considerable U immobilization is possible by stimulating the activity of indigenous bacteria in the DGR environment.

Natural uranium-contaminated soil in Korea Atomic Energy Research Institute (KAERI) was generated by decommissioning of the natural uranium conversion facility in 2010. Some of the contaminated soil was expected to be clearance level, however the disposal cost burden is increasing because it is not classified in advance. In this study, pre-classification method is presented according to the ratio of naturally occurring radioactive material (NORM) and contaminated uranium in the soil. To verify the validity of the method, the verification of the uranium radioactivity concentration estimation method through γ-ray analysis results corrected by self-absorption using MCNP6.2, and the validity of the pre-classification method according to the net peak area ratio were evaluated. Estimating concentration for 238U and 235U with γ-ray analysis using HPGe (GC3018) and MCNP6.2 was verified by 􀟙-spectrometry. The analysis results of different methods were within the deviation range. Clearance screening factors (CSFs) were derived through MCNP6.2, and net peak area ratio were calculated at 295.21 keV, 351.92 keV(214Pb), 609.31 keV, 1120.28 keV, 1764.49 keV(214Bi) of to the 92.59 keV. CSFs for contaminated soil and natural soil were compared with U/Pb ratio. CSFs and radioactivity concentrations were measured, and the deviation from the 60 minute measurement results was compared in natural soil. Pre-classification is possible using by CSFs measured for more than 5 minutes to the average concentration of 214Pb or 214Bi in contaminated soil. In this study, the pre-classification method of clearance determination in contaminated soil was evaluated, and it was relatively accurate in a shorter measurement time than the method using the concentrations. This method is expected to be used as a simple pre-classification method through additional research.

Metals such as stainless steel and alloy 600 are used as structures and materials in nuclear power plants due to their excellent mechanical properties and heat resistance. And recently thermal and mechanical cutting technologies are being actively researched and developed for dismantling NPP. Among them, the mechanical cutting method has the advantage of less secondary waste generation such as fume and fine dust, but according to the wider the cutting range, the reaction force and the cutting device size are increased. In this paper, plasma assisted milling has been proposed to reduce the reaction force and device size, and the plasma efficiency was measured for SUS 316L. The plasma torch was operated at the level of 3 to 4 kW so that it was heated only without cutting. And the feedrate was set at 150 to 250 mm/min. The test confirmed that the plasma efficiency was 35% about SUS 316L, and it is expected that the numerical analysis using these test results can be used as basic data for plasma assisted milling.

Various cutting technologies such as thermal and mechanical are being researched and developed to dismantle shutdown nuclear power plants. Each technology has the following advantages and disadvantages. The thermal cutting method has low reaction force and fast cutting speed, but secondary waste such as fume, dross, and fine dust is generated. The mechanical cutting method has the advantage of low generation of secondary waste such as fume, dross, and fine dust, but has the disadvantage of increasing the size of the device due to its large reaction force. In this study, the performance of plasma milling robot cutting technology for nuclear power plant materials was evaluated. First, before applying plasma auxiliary milling to the robot, tests were conducted on SUS 316 L and Alloy 600 to secure processing conditions such as plasma torch output and transfer speed. The test have shown that the mechanical strength was decreased of each material at the output power of the plasma torch of 4.4 and 8.4 kW, the transfer speed of 200 and 100 mm/min. Based on the test results, a plasma milling was attached to the robot and tested, and it was confirmed that even a small robot with a load of 140 kg can cut without any major problems.

With the rapid growth of nuclear power in China, a large number of dry wastes, which mainly include the high efficiency particulate air filters (glass fiber), cotton, polyethylene, and absorbent paper with low-level radioactivity and high volume, will be produced during the operation and maintenance of the nuclear power plants. Thermal plasma treatment is a world acceptable technology to incinerate and immobilize radioactive wastes, owing to the high volume reduction factor and the excellent chemical durability of the vitrified waste form. China has developed thermal plasma technology for the treatment of dry wastes from nuclear power plants for more than 15 years and the pilot plant has been constructed. This work will concentrate on the formulation of waste glass fiber to adapt to the vitrification process. A three-component (glass fiber-CaO-Na2O) constrained-region mixture experiment was designed and their viscosity data was mainly studied. The quadratic Scheffé model was used to plot the component effect on melting temperature. The retentions of simulated nuclides, such as Co, Sr, and Cs in the glasses were analyzed. In addition, the glass fiber as a glass matrix to immobilize residual ashes from the thermal plasma gasification of cotton, polyethylene, and absorbent paper was investigated as well.

The soils contaminated with radionuclides such as Cs-137 and Sr-90 should be solidified using a binder matrix, because radioactively contaminated soils pose environmental concerns and human health problems. Ordinary Portland cement has been widely used to solidify various radioactive wastes due to its low cost and simple process. In this study, simulant soil waste was solidified using cement waste form. The soils were collected around ‘Kori Nuclear Power Plant Unit 1’ and they were contaminated with the prepared simulant liquid waste containing Fe, Cr, Cs, Ni, Co, and Mn. The water-to-dry ingredients (W/D) ratio of cement waste form was 0.40. The cement paste was poured into a cubic mold (5×5×5 cm) and then cured for 28 days at room temperature. The 28-day compressive strength, water immersion, and EPA1311-toxicity characteristic leaching procedure (TCLP) tests were performed to evaluate the structural stability of cement waste form. The compressive strength was not proportional to soil waste loading, and the lowest compressive strength (4±0.1 MPa) was achieved in cement waste form containing 50wt% soil waste. After the water immersion test for 90 days, the compressive strength of cement waste form with 50wt% soil waste increased to 7.5±0.6 MPa, meeting the waste form acceptance criteria in the repository. It is believed that long-term water immersion test contributed to the additional curing and hydration reaction, resulting in the enhanced compressive strength. As a result of the TCLP test, the released amount of As, Ba, Cd, Cr, Pb, Se, Co, Cs, and Sr was less than the domestic and international standards. These results imply that cement waste form can be a promising candidate for the solidification of radioactive soil wastes.

As the importance of radioactive waste management has emerged, quality assurance management of radioactive waste has been legally mandated and the Korea Radioactive Waste Agency (KORAD) established the “Waste Acceptance Criteria for the 1st Phase Disposal Facility of the Wolsong Lowand Intermediate-Level Waste Disposal Center (WAC)”, the detailed guideline for radioactive waste acceptance. Accordingly, the Korea Atomic Energy Research Institute (KAERI) introduced a radioactive waste quality assurance management system and developed detailed procedures for performing the waste packaging and characterization methods suggested in the WAC. In this study, we reviewed the radioactive waste characterization method established by the KAERI to meet the WAC presented by the KORAD. In the WAC, the characterization items for the disposal of radioactive waste were divided into six major categories (general requirements, solidification and immobilization requirements, radiological, physical, chemical, and biological requirements), and each subcategories are shown in detail under the major classification. In order to satisfy the characterization criteria for each detailed item, KAERI divided the procedure into a characterization item performed during the packaging process of radioactive waste, a separate test item, and a characterization item performed after the packaging was completed. Based on the KAERI’s radioactive waste packaging procedure, the procedure for characterization of the above items is summarized as follows. First, during the radioactive waste packaging process, the characterization corresponding to the general requirements (waste type) is performed, such as checking the classification status of the contents and checking whether there are substances unsuitable for disposal, etc. Also, characterization corresponding to the physical requirements is performed by checking the void fraction in waste package and visual confirmation of particulate matter, substances containg free water, ect. In addition, chemical and biological requirements can be characterized by visually confirming that no hazardous chemicals (explosive, flammable, gaseous substances, perishables, infectious substances, etc.) are included during the packaging process, and by taking pictures at each packaging steps. Items for characterization using separate test samples include radiological, physical, and chemical requirements. The detailed items include identification of radionuclide and radioactivity concentration, particulate matter identification test, free water and chelate content measurement tests, etc. Characterization items performing after the packaging is completed include general requirements such as measuring the weight and height of packages and radiological requirements such as measurements of surface dose rate and contamination, etc. All of the above procedures are proceduralized and managed in the radioactive waste quality assurance procedure, and a report including the characterization results is prepared and submitted when requesting acceptance of radioactive waste. The characterization of KAERI’s radioactive waste has been systematically established and progressed under the quality assurance system. In the future, we plan to supplement various items that require further improvement, and through this, we can expect to improve the reliability of radioactive waste management and activate the final disposal of KAERI’s radioactive waste.

In this work, we report test results for direct melting of non-combustible wastes by using a 100 kW class transferred type plasma torch. For this purpose, non-combustible wastes consisting of metals and sands were prepared, weighed and melted by a transferred arc in a ceramic crucible with inner diameter of 150 mm. Test results reveal that 75wt% M6 iron bolts mixed with 25wt% sands were completely melted down within 140 seconds at the plasma power level of 83.8 kW, producing melting speed of 100 kg/hr and volume reduction rate of 62.8%. In addition, for simulated wastes consisting of 77.3wt% metal chips and 22.7wt% sands, the volume reduction rate high than 88% was achieved at 50 kW plasma power. These results indicate that non-combustible wastes can be treated efficiently when directly melting them by using transferred type plasma torch.

It is important that the plasma torch used in the waste treatment field has a high output to increase throughput. In order to increase the output of the plasma torch, there is a method of increasing the current or extending the length of the plasma arc. Among these methods, high power can be easily achieved simply by increasing current, but it is difficult to ensure electrode life. Therefore, it is necessary to check the appropriate current and arc length conditions to achieve high power and stable operation. In this paper, the power performance according to the arc length, current, and operation mode was confirmed in the transfer mode plasma torch. The test conditions are the distance (arc length) between the plasma torch and the external electrode was set to 5-180 mm, and the current was set to be in the range of 90-460 A. As a result of the test, it was confirmed that the reverse polarity operation had a maximum output of 159 kW depending on the arc length and current, and the positive polarity operation had a maximum output of 138 kW. Through this result, it was confirmed that the arc length had an effect on increasing the output, and that the reverse polarity operation had a longer arc than the positive polarity operation.

In underground repository environments, various types of engineered barriers are installed to hinder the mobility of radionuclides. Cement admixtures, especially used to improve workability for concrete, are composed of fairly high organic molecules and have a dispersing effect through bonding with the C-S-H of the concrete. Previous studies have shown that complex-forming organics like EDTA, NTA, and ISA have a significant effect on the mobility of radionuclides, but the studies on the behavior and stability of combined complexes in hydrated cement are lacking. So, we selected a commonly used polycarboxylic-ester (PCE) type cement admixture and stable Co as a surrogate of Co-60 to perform desorption experiments from hydrated cement containing the admixture. Radioactive Co is known to be a common contaminant in nuclear fission and medical facilities and considered to exist as a relatively stable phase in repositories. In addition, the evaluation of cobalt can be a standard of safety issue for other radionuclides with the presence of cement admixture in repository. In this study, cement samples were prepared at water/cement ratio of 0.55 and cured for 28 days at 23-25°C and at least 80% of humidity with varying cement admixtures of 0.0, 0.1, and 2.0wt%. To evaluate the stability of cobalt in the weathered cement, a 0.001 M HCl solution was used to simulate cement weathering conditions on a hot plate at 60°C for 1 day using a solid/liquid ratio of 1:100. Degree of weathering was confirmed using XRD analysis. The adsorption experiments were performed by adding 0.0042 mmol of cobalt (CoCl2, Sigma-Aldrich, anhydrous ≥ 98.0%) to the weathered cement for 3 days using a platform shaker at 200 rpm, and the supernatant was separated using a syringe filter (<0.20 um) before ICP-MS analysis to determine the amount of Co adsorption. Cobalt desorption was tested for the Co-adsorbed cement using 0.019 mmol of calcium (Ca(NO3)2·4H2O, Sigma-Aldrich, 99%) for 3 hours to 14 days. The results showed that adsorbed cobalt with and without cement admixture was stably bound to cement, and did not increase any noticeable Co release by 2.0wt% PCE admixture. However, additional experiments using varying contents of PCE and other admixtures should be conducted to provide a standard for assessing the safety of cement admixtures in repositories.

The homogeneity of radioactive spent ion exchange resins (IERs) distribution inside waste form is one of the important characteristics for acceptance of waste forms in long-term storage because heterogenous immobilization can lead to the poor structural stability of waste form. In this study, the homogeneity of metakaolin-based geopolymer waste form containing simulant IERs was evaluated using a laser-induced breakdown spectroscopy (LIBS) and statistical approach. The cation-anion mixed IERs (IRN150) were used to prepare the simulant spent IERs contaminated by non-radioactive Cs, Fe, Cr, Mn, Ni, Co, and Sr (0.44, 8.03, 6.22, 4.21, 4.66, 0.48, and 0.90 mg/g-dried IER, respectively). The K2SiO3 solution to metakaolin ratio was kept constant at 1.2 and spent IERs loading was 5wt%. For the synthesis of homogeneous geopolymer waste form, spent IERs were mixed with K2SiO3 solution and metakaolin first, and then the fresh mixture slurry was poured into plastic molds (diameter: 2.9 cm and height: 6.0 cm). The heterogeneous geopolymer waste form was also fabricated by stacking two kinds of mixtures (8wt% IERs loading in bottom and 2wt% in top) in one mold. Geopolymers were cured for 7d (1d at room temperature and 6d at 60°C). The hardened geopolymers were cut into top, middle, and bottom parts. The LIBS spectra and intensities for Cs were obtained from the top and bottom of each part. Cs was selected for target nuclide because of its good sensitivity for measurement. Shapiro-Wilk test was performed to determine the normality of LIBS data, and it revealed that data from the homogeneous sample is normal distribution (p-value = 0.9246, if p-value is higher than 0.05, it is considered as normal distribution). However, data from the heterogeneous sample showed abnormal distribution (p-value = 7.765×10-8). The coefficient of variation (CoV) was also calculated to examine the dispersion of data. It was 31.3% and 51.8% from homogeneous and heterogeneous samples, respectively. These results suggest that LIBS analysis and statistical approaches can be used to evaluate the homogeneity of waste forms for the acceptance criterion in repositories.

Currently, KHNP-CRI has developed 100 kW plasma torch melting facility to reduce the amount of radioactive waste in nuclear power plant. Plasma torch melting technology uses electric arc phenomena like lightning to melt the target material at a high temperature of about 1,600°C. The technology is applicable to treatment for various types of waste such as combustible, non-combustible and mixed wastes. The volume reduction ratio by the technology is respectively expected to be about 1/60 of combustible wastes and about 1/5 for non-combustible wastes. It is important to discharge the melt without problems in the melting technology. In general, molten slag has properties such as high viscosity and quick solidification. Because of the properties, when discharging into slag container, the final product is accumulated like a mountain. To improve this problem, there is three suggestions; 1) rotation of the slag container, 2) vibration of the slag container, and 3) heating of the slag container.

Since high-level radioactive wastes contain long-lived nuclides and emit high energy, they should be disposed of permanently through a deep geological disposal system. In Korea, the first (2016.07) and the second (2021.12) basic plans for the management of high-level disposal systems were proposed to select sites for deep geological disposal facilities and to implement business strategies. Leading countries such as Finland, Sweden and France have developed and applied safety cases to verify the safety of deep geological disposal systems. By examining the regulatory status of foreign leading countries, we analyze the safety cases ranging from the site selection stage of the deep geological disposal system to the securing of the permanent disposal system to the investigation, analysis, evaluation, design, construction, operation, and closure. Based on this analysis, we will develop safety case elements for long-term safety of deep geological disposal systems suitable for domestic situation. To systemically analyze data based on safety cases, we have established a database of deep geological disposal system regulations in leading foreign countries. Artificial intelligence text mining and data visualization techniques are used to provide database in dashboard form rather than simple lists of data items, which is a limitation of existing methods. This allows regulatory developers to understand information more quickly and intuitively and provide a convenient interface so that anyone can easily access the analyzed data and create meaningful information. Furthermore, based on the accumulated bigdata, the artificial intelligence learns and analyzes the information in the database through deep learning, and aims to derive a more accurate safety case. Based on these technologies, this study analyzed the legal systems, regulatory standards, and cases of major international leading countries and international organizations such as the United States, Sweden, Finland, Canada, Switzerland, and the IAEA to establish a database management system. To establish a safety regulation base suitable for the domestic deep geological disposal environment, the database is provided as data to refer to and apply systematic information management on regulatory standards and regulatory cases of overseas leading countries, and it is expected that it will play a key role as a forum for understanding and discussing the level of safety of deep geological disposal system among stakeholders.

A disposal system for spent nuclear fuel divides into two parts; (1) engineered barriers including spent nuclear fuel, canister, buffer and backfill, (2) natural barriers surrounding engineered barriers. Sorption and diffusion are main retardation mechanisms for the migration of released radionuclides. We analyzed the sorption properties of radionuclides for bentonite as a buffer material and collected/ evaluated the distribution coefficients for the purpose of safety assessment for the deep geological disposal of a spent nuclear fuel. Through this, we presented recommended distribution coefficients for radionuclides required for the safety assessment. This work included the radionuclides as follows; alkali and alkaline earth metals (Cs, Sr, Ba), lanthanides (Sm), actinides (Ac, Am, Cm, Np, Pa, Pu, Th U), transition elements (Nb, Ni, Pd, Tc, Zr), and others (C, Cl, I, Rn, Se, Sn). The sorption of radionuclides affected various geochemical conditions such as pH/carbonates, redox potential, ionic strength, radionuclide concentration, kinds and amounts of minerals, and microbes. Among the evaluated radionuclides, Cs, Ni, Pd, and Ra is sensitive to the ionic strength, while Np, Pu, U, Se, and, Tc is sensitive to the redox condition. For the evaluation of distribution coefficients, the data from Sweden (SKB), Finland (Posiva), Switzerland (Nagra), and Japan (JAEA) were collected, analyzed, and the recommended distribution coefficients were suggested.

Timely detection of nuclear activity is important for the management and supervision of nuclear materials, and inspections on North Korea as a method of safety measures for this monitoring system seem to be a not far future in the rapidly changing North Korea relationship. However, a simpler and more reliable monitoring device is required since the inspection period is limited and the possibility of revisiting is unclear. The seal is a simple but easily used security device for monitoring forgery and falsification in the IAEA. IAEA presents matters related to 1) project engineering, 2) remote monitoring, and 3) seal development as the three major tasks of the Department of safeguards and Division of Technical Support. The importance of development, modernization, and application of new seal devices is emphasized, and advanced sealing and identification system development is in progress at many research institutes such as JRC, ISCN, and JAEA. Since the existing seal devices used by IAEA and KINAC can only be confirmed through on-site inspections for damage, it is difficult to respond immediately in the event of similar situations such as theft of nuclear materials and loss of continuity of knowledge. Unlike facilities that comply with the requirements for safety measures, such as domestic nuclear facilities, in the case of facilities subject to denuclearization, it is very likely that various hazardous environments will exist that make it difficult to apply safety measures. Hence, a real-time seal device has developed through prior research due to the high possibility of situations in which Continuity of Knowledge (COK) is not maintained, such as damage, malfunction, and power loss of sealing and monitoring equipment. Through previous studies, the real-time seal device was loaded with server-based operating software and improved its performance by utilizing feedback from real users (KINAC) after use. In this study, the effectiveness of the previously developed sealing system was verified through performance evaluation, and the authentication of the equipment was secured through environmental tests.

This study uses first-principles calculations to investigate the mechanical properties and effect of strain on the electronic properties of the 2D material 1H-PbX2 (X: S, Se). Firstly, the stability of the 1H Pb-dichalcogenide structures was evaluated using Born’s criteria. The obtained results show that the 1H-PbS2 material possesses the greatest ideal strength of 3.48 N/m, with 3.68 N/m for 1H-PbSe2 in biaxial strain. In addition, 1H-PbS2 and 1H-PbSe2 are direct semiconductors at equilibrium with band gaps of 2.30 eV and 1.90 eV, respectively. The band gap was investigated and remained almost unchanged under the strain εxx but altered significantly at strains εyy and εbia. At the fracture strain in the biaxial direction (19 %), the band gap of 1H-PbS2 decreases about 60 %, and that of 1H-PbSe2 decreases about 50 %. 1H-PbS2 and 1H-PbSe2 can convert from direct to indirect semiconductor under the strain εyy. Our findings reveal that the two structures have significant potential for application in nanoelectronic devices.