In this study, we examined the antagonistic effects of sprout-borne lactic acid bacteria (LAB) on Salmonella enterica serovar Enteritidis. This antagonism is promoted as a means of controlling contamination during sprout production and provides additional LAB for consumers. We isolated a total of 24 LAB isolates in nine species and five genera from seven popular vegetable sprouts: alfalfa (Medicago sativa), clover (Trifolium pratense), broccoli (Brassica oleracea ssp. italica), vitamin (B. rapa ssp. narinosa), red radish (Raphanus sativus), red kohlrabi (B. oleracea var. gongylodes), and Kimchi cabbage (B. campestris var. pekinensis). Based on 16S rRNA gene sequences, the LAB species were identified as Enterococcus casseliflavus, E. faecium, E. gallinarum, E. mundtii, Lactococcus taiwanensis, Leuconostoc mesenteroides, Pediococcus pentosaceus, and Weissella cibaria, and W. confusa. A total of 16 LAB isolates in seven species including E. faecium, E. gallinarum, E. mundtii, L. taiwanensis, L. mesenteroides, P. pentosaceus, and W. cibaria showed antagonistic activity toward S. enterica. The growth inhibition of sprout LAB on S. enterica was confirmed by co-culture. Unexpectedly, sprout LAB failed to suppress the growth of S. enterica in alfalfa sprouts, whereas all LAB strains stimulate S. enterica growth even if it is not significant in some strains. The findings of this study indicate that S. enterica-antagonistic LAB are detrimental to food hygiene and will contribute to further LAB research and improved vegetable sprout production.

This study presents a rapid and sequential radiochemical separation method for Pu and Am isotopes in radioactive waste samples from the nuclear power plant with anion exchange resin and TRU resin. After radionuclides were leached from the radioactive waste samples with concentrated HCl and HNO3, the sample was allowed to evaporate to dryness after filtering the leaching solution with 0.45 micron filter. The Pu isotopes were separated in HNO3 medium with anion exchange resin. For leaching solution passed through anion exchange column, the Am isotopes were separated with TRU resin. The purified Pu and Am isotopes were measured by alpha spectrometer, respectively, after micro-precipitation of neodymium. The sequential radiochemical separation of Pu and Am isotopes in radioactive waste samples using anion exchange resin and TRU resin was validated with ICP-MS system.

Heavy water (deuterium oxide, D2O) is water in which hydrogen atoms (1H, H), one of the constituent elements of water molecules, have been replaced with deuterium (2H, D), a heavier isotope. Heavy water is used in a variety of industries, including semiconductors, nuclear magnetic resonance, infrared spectroscopy, neutron deceleration, neutrino detection, metabolic rate studies, neutron capture therapy, and the production of radioactive materials such as plutonium and tritium. In particular, heavy water is used as a neutron moderator or coolant in nuclear reactors and as a fuel for nuclear fusion energy, methods for measuring heavy water are becoming increasingly important. There are methods with density, mass spectrometry, and infrared (IR) spectroscopy. In this study, Fourier transform infrared spectroscopy (FT-IR) was used, which is commonly used in IR spectroscopy because of its relatively high analytical sensitivity, low operating costs, and easy online analysis. Heavy water was identified in the range of 2,300-2,600 cm-1 wavenumber (O-D) and the range of 1,200-1,300 cm-1 wavenumber (D-O-D), which are known to be the range with strong infrared absorption. As a result, the linearity of infrared absorbance for each heavy water concentration was confirmed within the relative expansion uncertainty (k=2).

In this study, we introduce the validation of the analysis guidelines through preliminary experiments of the draft analysis guidelines before analyzing waste materials (non-combustible). This validation data was applied the accuracy and efficiency of the separation and analysis for the waste such as steel generated from NPP. Steel (non-flammable) was leached the mixed acid and the leaching solution was separated by using the separation guidelines. Steel was corroded with radioactive RM (Co-60, Cs-137) and mixed acid. After drying, the corroded steel was measured the initial radioactivity by a HPGe detector (10,000 seconds). The sample was inserted in a beaker and leached with mixed acid (10 M HNO3 + 4 M HCl) for 2 hours. In this solution, it added 2 ml of H2O2 to increase the leaching effect. The ultrasonic device was adjusted so that the temperature does not exceed 60°C. After elution, the surface of the sample was washed with pure water. The weight of the sample was measured accurately, and recorded the weight loss rate after infiltration. The leaching sample was measured radioactivity by a HPGe detector (10,000 seconds). It was calculated the recovery rate based on the difference in total radioactivity before and after leaching. Before the test, radioactive RM (Co-60, Cs-137) was radioactive deposited by corrosion, but Cs- 137 was not detected in the initial gamma measurement and only Co-60 nuclides were deposited. The recovery rate test results were confirmed to be about 100%.

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.

Chelating agents, such as ethylenediaminetetraacetic acid (EDTA), diethylenetriaminepentaacetic acid (DTPA), and nitrilotriacetic acid (NTA) are widely used in industry and agriculture as water softeners, detergents, and metal chelating agents. In wastewater treatment plants, a significant amount of chelating agents can be discharged into natural waters because they are difficult to degrade. Since those compounds affect the mobility of radionuclides or heavy metals in decontamination operations at nuclear facilities and radioactive waste disposal, quantification of the amount of ligand is very important for safe nuclear waste management. To predict the behavior of the main complexation in sample matrices of radioactive wastes, it is essential to evaluate the distribution of the metal-chelating species and their stabilities in order to develop analytical techniques for quantifying chelating agents. We have investigated to collect information on the pH speciation of metal chelation and the stability constants of metal complexes depending on three chelating agents (EDTA, DTPA, and NTA). For example, Zhang’s group recently reported that the initial coordination pH of Cu(II) and EDTA4− is delayed with the addition of Fe(III), and the pH range for the stable existence of [Cu(EDTA)]2− is narrowed compared to when it is alone in the sample matrix. The addition of Fe(III) clearly impacts the chemical states of the Cu(II)-EDTA solution. Additionally, Eivazihollagh’s group demonstrated differences in the speciation and stability of Cu(II) species between Cu(II) and three chelating ligands (EDTA, DTPA, and NTA). This study will be greatly helpful in identifying the sample matrix for binding major chelating agents and metals as well as developing chemically sample pretreatment and separation methods based on the sample matrix. Finally, these advancements will enable reliable quantitative analysis of chelating agents in decommissioning radioactive wastes.

The operation of nuclear facilities involves the potential for on-site contamination of soil, primarily resulting from pipe leaks and other operational incidents. Globally, decommissioning process for commercial nuclear power plants have revealed huge-amounts of soil waste contaminated with Cs-137, Sr-90, Co-60, and H-3. For example, Connecticut Yankee in the United States produced approximately 52,800 ton of contaminated soil waste, constituting 10% of the total waste generated during its decommissioning. Environmental remediation costs associated with nuclear decommissioning in the US averaged $60 million per unit, representing a significant 10% of the whole decommissioning expenses. Consequently, this study undertook a preliminary investigation to identify important factors for establishing a site remediation strategy based on radionuclide- and site-specific media- characteristics, focusing the efficiency enhancement for the environmental remediation. The factors considered for this investigation were categorized into physical/environmental, socioeconomic, technical, and management aspects. Physical/environmental factors contained the site characteristics, contamination levels, and environmental sensitivity, while socio-economic factors included the social concerns and economic costs. Technical and management factors included subcategories such as technical considerations, policy aspects, and management factors. Especially, technical factors were further subdivided to consider the site reuse potential, secondary waste generation by site remediation, remediation efficiency, and remediation time. Additionally, our study focused the key factors that facilitate the systematic planning for the site remediation, considering the distribution coefficient (Kd) and hydrogeological characteristics associated with each radionuclide in specific site conditions. Therefore, key factors in this study focus the geochemical characteristics of site media including the particle size distribution, chemical composition, organic and inorganic constituents, and soil moisture content. Moreover, the adsorption properties of site media were examined concerning the distribution coefficient (Kd) of radionuclides and their migration characteristics. Furthermore, this study supported the development of a conceptual framework, containing the remediation strategies that incorporate the mobility of radionuclides, according to the site-specific media. This conceptual framework would necessitate the spatial analysis techniques involving the whole contamination surveys and radionuclide mobility modeling data. By integrating these key factors, the study provides the selection and simulation of optimal remediation methods, ultimately offering the estimated amounts of radioactive waste and its disposal costs. Therefore, these key factors offer foundational insights for designing the site remediation strategies according the sitespecific information such as the distribution coefficient (Kd) and hydrogeological characteristics.

In nuclear power plant (NPP) decommissioning, ventilation and purification of the building atmosphere are important to create a working environment, ensure worker safety, and prevent the release of gaseous radioactive materials into the environment. The heating, ventilation, and air conditioning (HVAC) system of each building is maintained, modified, or newly installed. In this study, based on APR1400, operation strategies were presented in case of ventilation abnormalities in the reactor containment building (RCB), where highly radioactive particles and high dust are most frequently generated during NPP decommissioning. For research, it was assumed that the entire RCB atmospheric ventilation during decommissioning would use the RCB purge system of the existing NPP and perform continuous ventilation. Additionally, it is assumed that areas where high radiation particles and high dust occur locally, such as reactor containers or internal segments, are sealed with tents and purified using a HEFA filter of a temporary portable HVAC, and a exhaust flow path is connected to the discharge duct of the existing RCB purge system. The possibility of abnormal occurrence was largely divided into two cases. First, when large amounts of uncontrolled pollutants are released into the atmosphere inside the RCB, discharge to the environment is stopped manually or automatically by a modified engineered safety function activation signal (ESFAS). Afterwards, the RCB purge system should be operated in recirculation mode to sufficiently purify the RCB atmosphere with a HEPA filter. Second, when the first train of the low volume purge system is not running due to a failure, standby train should be operated. If both low volume purge trains fail, a high volume purge system is used. Intermittent purge operation is preferred due to large capacity during high volume purge operation. In cases where it is not possible to operate all purge systems due to common issues such as power supply, atmospheric sampling is performed to determine whether to proceed with the work inside RCB.

The radiological characterization of SSCs (Structure, Systems and Components) plays one of the most important role for the decommissioning of KORI Unit-1 during the preparation periods. Generally, a regulatory body and laws relating to the decommissioning focus on the separation and appropriate disposal or storage of radiological waste including ILW (intermediate level waste), LLW (low level waste), VLLW (very low level waste) and CW (clearance waste), aligned with their contamination characteristics. The result of the preliminary radiological characterization of KORI Unit-1 indicated that, apart from neutron activated the RV (reactor vessel), RVI (reactor vessel internals), and BS (biological shielding concrete), the majorities of contamination were sorted to be less than LLW. Radiological contamination can be evaluated into two methods. Due to the difficulties of directly measuring contamination on the interior surfaces of the pipe, called CRUD, the assessment was implemented by modeling method, that is measuring contamination on the exterior surfaces of the pipes and calculating relative factors such as thickness and size. This indirect method may be affected by the surrounding radiation distribution, and only a few gamma nuclides can be measured. Therefore, it has limitation in terms of providing detailed nuclide information. Especially, α and β nuclides can only be estimated roughly by scaling factors, comparing their relative ratios with the existing gamma results. To overcome the limitation of indirect measurement, a destructive sampling method has been employed to assess the contamination of the systems and component. Samples are physically taken some parts of the systems or components and subsequently analyzed in the laboratory to evaluate detailed nuclides and total contamination. For the characterization of KORI Unit-1, we conducted the radiation measurement on the exterior surfaces of components using portable instruments (Eberline E-600 SPA3, Thermo G20-10, Thermo G10, Thermo FH40TG) at BR (boron recycle system) and SP (containment spray system) in primary system. Based on these results, the ProUCL program was employed to determine the destructive sample collection quantities based on statistical approach. The total of 5 and 8 destructive sample quantities were decided by program and successfully collected from the BR and SP systems, respectively. Samples were moved to laboratory and analyzed for the detail nuclide characteristics. The outcomes of this study are expected to serve as valuable information for estimating the types and quantities of radiological waste generated by decommissioning of KORI Unit-1.

Korea Atomic Energy Research Institute (KAERI) has been operating the Post Irradiation Examination Facility (PIEF) for fuel examinations. The facility has pools and hot cells for handling and examining fuel assemblies and rods. Among the hot cells, the second cell is for measuring rod internal pressure (RIP) and then cutting the rod to make samples for destructive tests. Currently, the cutting machine is broken, so it has to be replaced. Because the existing cutting machine consists of many parts and its size was quite a bit large to handle and treat for the radioactive waste disposal, the disassembly work has been performed to make it smaller using manipulators. The drawings of the cutting machine were reviewed and the disassembly tools were developed considering workability when the work performed at the hot cell using the manipulators. The large parts such as motor, mirror and cable, etc., were able to be disassembled and the machine size became so smaller that it could be easily handled for the disposal.

As unit 1 of Kori was permanently shut down in June 2017, domestic nuclear industry has entered the path of decommissioning. The most important thing in decommissioning is cost reduction. And volume reduction of radioactive waste is especially important. According to the IAEA report, more than 4,000 tons of metallic waste is generated during the decommissioning of a 1,000 MWe reactor and most of these wastes are LLW or VLLW. To reduce amount of metallic waste dramatically, we should choose efficient decontamination method. In this study, we conducted dry ice and bead blasting decontamination. We prepared Inconel-600 and STS-304 specimen with dimensions of 30 mm × 30 mm × 5 mm. Loose and fixed contamination was applied on the surface of specimen using SIMCON method. Bead and dry-ice blasting was conducted by spraying alumina and dry ice pellet at the same pressure and distance for the same time. The removal of loose contamination was observed using microscope. It was found that contaminants are significantly removed using both dry ice blasting and bead blasting. However, some abrasive material remained on the surface of specimen. The removal of fixed contamination was verified by weight comparison before and after experiment and cobalt concentration comparison before and after experiment using X-ray Fluorescence Spectroscope (XRF). At least 90% of the cobalt was removed, but some abrasive particle was also remained on the surface of specimen. In this study, it is confirmed that the effectiveness of manufacturing a large-scale abrasive decontamination facility, and it is expected that this technology can be used to effectively reduce the amount of metallic waste generated during decommissioning.

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.