This comprehensive study delves into the intricate process of exfoliating and functionalizing boron nitride nanosheets (BNNSs) extracted from hexagonal boron nitride (h-BN), and meticulously explores their potential application within epoxy composites. The extensive research methodology encompasses a sequence of treatments involving hydrothermal and sonication processes aimed at augmenting the dispersion of BNNSs in solvents. Leveraging advanced analytical techniques such as Raman spectroscopy, X-ray diffraction, and FTIR spectroscopy, the study rigorously analyzes a spectrum of changes in the BNNS’s properties, including layer count variations, interlayer interactions, crystal structure modifications, and the introduction of functional groups. The research also rigorously evaluates the impact of integrating BNNSs, specifically glycidyl methacrylate (GMA)-functionalized BNNSs, on the thermal conductivity of epoxy composites. The conclusive findings exhibit notable enhancements in thermal properties, predominantly attributed to the enhanced dispersion of fillers and enhanced interactions within the epoxy matrix. This pioneering work illuminates the wide potential of functionalized BNNSs for significantly enhancing the thermal conductivity of epoxy composites, paving the way for advanced materials engineering and practical applications.

In NPP (nuclear power plant), boric acid is used as a neutron absorbent. So radioactive boric acid waste are generated from various waste streams such as discharge or leakage of reactor coolant water, floor drains, drainage of equipment for operation or maintenance, reactor letdown flows and etc. Depending on KHNP, 20,015 drum (200 L drum) of concentrated boric acid waste were stored in KOREA NPP until 2019. In previous study, our group suggested the waste upcycling process synthesizing B4C neutron absorber using boric acid waste and activated carbon waste to innovatively reduce radioactive wastes. Radioactive activated carbon waste was utilized in off gas treatment system of NPP to capture nuclide such as I-131, C-14 and H-3. Activated carbon waste is treated as low-level radioactive waste and pre-treatment system for removing nuclide from the activated carbon waste is needed to use B4C up-cycling process. In this study, microwave treatment system is suggested to treat the activated carbon waste. Activated carbon waste was exposed to microwave for a few minutes and temperature of the waste was dramatically increased over 400°C. Nuclide in the activated carbon waste were selectively removed from the waste without massive production of secondary off gas waste.

Radioactive liquid waste generated during the operation of domestic nuclear power plants is treated through a somewhat different liquid radwaste system (LRS) for each plant. Prior to the introduction of standard nuclear power plants, LRS used a concentrated water dry system (CWDS) to evaporate liquid waste and manage it in the form of dry powder. The boron-containing radioactive liquid waste dry powder was solidified using paraffin from 1995 to 2010, and about 3,650 drums (based on 200 L) of paraffin solidified drums are currently stored in nuclear power plants. Paraffin solidification drums do not meet the acceptance criteria for radioactive waste repositories because it is difficult to secure the homogeneity of the solidified body and there are concerns about leaching of radioactive waste due to the low melting point of paraffin. In order to solve this problem and safely permanently dispose of paraffin solidification drums, the characteristics of dry powder paraffin solidification drums containing boron-containing radioactive liquid waste must be analyzed and appropriate treatment technology utilizing the results must be introduced. This study analyzes the physical properties of paraffin, the chemical properties of boron-containing radioactive waste dry powder, and the physicochemical properties of paraffin solidification powder, and proposes an appropriate alternative technology for treating boron-containing radioactive waste dry drum. When disposing of the paraffin solidification drum with boron-containing radioactive liquid waste dry powder, the solidification body must be effectively withdrawn from the drum and the paraffin must be completely separated from the solidification body. When disposing the drum, the solidified material must be effectively extracted from the drum and the paraffin must be completely separated from the solidified material. Afterwards, the paraffin must be self-disposed, and the radioactive waste must be disposed of in accordance with acceptance criteria of repository. We looked at how each characteristic of the paraffin solidification drum with boron-containing radioactive liquid waste dry powder can be utilized in each of the above treatment processes.

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

Domestic nuclear power plants can affect the environment if multiple devices are operated on one site and even a trace amount of pollutants that may affect the environment after power generation are simultaneously discharged. Therefore, not only radioactive substances but also ionic substances such as boron should be discharged as minimally as possible. We adopted pilot CDI and SD-ELIX sytem to separating and concenrating of boron containing nulcear power plant discharge water. The boron concentration of the initial inflow water tended to decrease over time. The water quality of concentrated water also reached its peak until the initial 60 minutes, but tended to decrease in line with the decrease in the inflow water concentration. The boron removal rate was in the range of 85 to 99% with respect to the initial boron concentration of 15 to 25 mg/L. On the other hand, performance degradation due to the use of electrochemical modules is also observed, and regeneration through low ion-containing water cleaning effective. We shortened processing time by considering the optimal flow rate conditions and conductivity conditions and converting electrochemical modules into series or parallel.

The objective of this study is development of graphite-boron composite material as a replacement for metal canisters to Improve the heat dissipation and radiation shielding performance of dry spent nuclear fuel storage system and reduce the volume of waste storage system. KEARI research team plan to use the graphite matrix manufacturing technology to pelletize the graphite matrix and adjust the content of phenolic resin binder to minimize pore formation. Specifically, we plan to adjust the ratio of natural and synthetic graphite powder and use uniaxial pressing technology to manufacture black graphite matrix with extremely high radial thermal conductivity. After optimizing the thermal conductivity of the graphite matrix, we plan to mix it with selected boron compounds, shape it, and perform sintering and purification heat treatments at high temperatures to manufacture standard composite materials.

International Atomic Energy Agency defines the term “Poison” as a substance used to reduce reactivity, by virtue of its high neutron absorption cross-section, in IAEA glossary. Poison material is generally used in the reactor core, but it is also used in dry storage systems to maintain the subcriticality of spent fuel. Most neutron poison materials for dry storage systems are boron-based materials such as Al-B Carbide Cermet (e.g., Boral®), Al-B Carbide MMC (e.g., METAMIC), Borated Stainless Steel, Borated Al alloy. These materials help maintain subcriticality as a part of the basket. U.S.NRC report NUREG-2214 provides a general assessment of aging mechanisms that may impair the ability of SSCs of dry storage systems to perform their safety functions during longterm storage periods. Boron depletion is an aging mechanism of neutron poison evaluated in that report. Although that report concludes that boron depletion is not considered to be a credible aging mechanism, the report says analysis of boron depletion is needed in original design bases for providing long-term safety of DSS. Therefore, this study aimed to simulate the composition change of neutron poison material in the KORAD-21 system during cooling time considering spent fuel that can be stored. The neutron source term of spent fuel was calculated by ORIGEN-ARP. Using that source term, neutron transport calculation for counting neutrons that reach neutron poison material was carried out by MCNP®-6.2. Then, the composition change of neutron poison material by neutron-induced reaction was simulated by FISPACT-II. The boron-10 concentration change of neutron poison material was analyzed at the end. This study is expected to be the preliminary study for the aging analysis of neutron poison material about boron depletion.

In this work, we investigated a modern combined processing technique for the synthesis of lightweight superhard composites based on boron–carbon. We used traditional B4C with precipitates of free graphite and Al powder as initial materials. In the first stage, the composites were fabricated by the self-propagating high-temperature synthesis (SHS) with the subsequent hot pressing of the compound. Further, by the disintegration and attrition milling, the ultrafine-grained powder was obtained. We used HCl and HNO3 acids for the chemical leaching of the powder to remove various impure compounds. At the last stage, a solid composite was obtained by the spark plasma sintering (SPS) method under nitrogen pressure. The main feature of this approach is to implement different synthesis techniques and chemical leaching to eliminate soft phases and to obtain superhard compounds from low-cost materials. The phases were studied by X-ray diffraction and scanning electron microscopy with energy-dispersive spectroscopy. The composites compacted by the SPS method contained superhard compounds such as B13C2, B11.7C3.3, and c-BN. The fabricated composite has an ultrafine-grained microstructure. Using a Berkovich indenter, the following nanohardness results were achieved: B13C2 ~ 43 GPa, c-BN ~ 65 GPa (all in Vickers scale) along with a modulus of elasticity ranging between ~ 400 GPa and ~ 450 GPa.

Hydroxyl radical (OH radical) is the most harmful free radical amongst the Reactive Oxygen Species (ROS) responsible for numerous diseases of DNA damage like mutagenesis, carcinogenesis and ageing. Therefore, it is important to find a suitable scavenger for OH radical. In the present contribution, we aim to investigate the ability of pristine armchair-SWCNT and B/N/P-doped armchair-SWCNT to scavenge OH radicals using DFT calculations. The calculations reveal that the B/Pdoped armchair-SWCNTs can act as a better scavenger for OH radical compared to pristine armchair-SWCNT but N-doped armchair-SWCNT does not act as a better scavenger for OH radical compared to pristine armchair-SWCNT. Furthermore, the developed scavenger is examined in terms of large-scale availability, biocompatibility, conductivity, stability and reactivity. For both in vivo and in vitro studies, the work is found to useful for enhancing SWCNT as a free radical scavenger.

Regulations on the concentration of boron discharged from industrial facilities, including nuclear power plants, are increasingly being strengthened worldwide. Since boron exists as boric acid at pH 7 or lower, it is very difficult to remove it in the existing LRS (Liquid Radwaste System) using RO and ion exchange resin. As an alternative technology for removing boron emitted from nuclear power plants, the electrochemical boron removal technology, which has been experimentally applied at the Ringhal Power Plant in Sweden, was introduced in the last presentation. In this study, the internal structure of the electrochemical module was improved to reduce the boron concentration to 5 mg/L or less in the 50 mg/L level of boron-containing waste liquid. In addition, the applicability of the electrochemical boron removal technology was evaluated by increasing the capacity of the unit module to 1 m3/hr in consideration of the actual capacity of the monitor tank of the nuclear power plant. By applying various experimental conditions such as flow rate and pressure, the optimum boron removal conditions using electrochemical technology were confirmed, and various operating conditions necessary for actual operation were established by configuring a concentrated water recirculation system to minimize secondary waste generation. The optimal arrangement method of the 1 m3/hr unit module developed in this study was reviewed by performing mathematical modeling based on the actual capacity of monitor tank and discharge characteristics of nuclear power plant.

Water electrolysis is a representative technology for tritium enrichment in water. Proton exchange membrane (PEM) water electrolysis has received great attention to replace traditional alkaline water electrolysis which generates concentrated tritiated water containing a large amount of salts. Nafion has been widely used as a polymeric electrolyte for the PEM electrolyzer. However, its low gas barrier property causes explosion, corrosion or degradation of electrolyzer. Furthermore, the traditional polymeric electrolytes have negligible differences in conductivity between hydrogen isotopes. To enhance the tritium separation by water electrolysis, we designed a composite membrane (Nafion/ hexagonal boron nitride (hBN)). The monolayer hBN has a high proton conductivity and gas barrier property, and the hBN can enhance conductivity differences between hydrogen isotopes. We prepared Nafion/hBN composite membranes, and water electrolysis performances and proton/deuterium separation behaviors were investigated.

Extensive research is being carried out on Ni-rich Li(NixCoyMn1-x-y)O2 (NCM) due to the growing demand for electric vehicles and reduced cost. In particular, Ni-rich Li(NixCoyMn1-x-y-zAlz)O2 (NCMA) is attracting great attention as a promising candidate for the rapid development of Co-free but electrochemically more stable cathodes. Al, an inactive element in the structure, helps to improve structural stability and is also used as a doping element to improve cycle capability in Ni-rich NCM. In this study, NCMA was successfully synthesized with the desired composition by direct coprecipitation. Boron and tin were also used as dopants to improve the battery performance. Macro- and microstructures in the cathodes were examined by microscopy and X-ray diffraction. While Sn was not successfully doped into NCMA, boron could be doped into NCMA, leading to changes in its physicochemical properties. NCMA doped with boron revealed substantially improved electrochemical properties in terms of capacity retention and rate capability compared to the undoped NCMA.

In this study, three kinds of bainitic steels are fabricated by controlling the contents of vanadium and boron. High vanadium steel has a lot of carbides and nitrides, and so, during the cooling process, acicular ferrite is well formed. Carbides and nitrides develop fine grains by inhibiting grain growth. As a result, the low temperature Charpy absorbed energy of high vanadium steel is higher than that of low vanadium steel. In boron added steel, boron segregates at the prior austenite grain boundary, so that acicular ferrite formation occurs well during the cooling process. However, the granular bainite packet size of the boron added steel is larger than that of high vanadium steel because boron cannot effectively suppress grain growth. Therefore, the low temperature Charpy absorbed energy of the boron added steel is lower than that of the low vanadium steel. HAZ (heat affected zone) microstructure formation affects not only vanadium and boron but also the prior austenite grain size. In the HAZ specimen having large prior austenite grain size, acicular ferrite is formed inside the austenite, and granular bainite, bainitic ferrite, and martensite are also formed in a complex, resulting in a mixed acicular ferrite region with a high volume fraction. On the other hand, in the HAZ specimen having small prior austenite grain size, the volume fraction of the mixed acicular ferrite region is low because granular bainite and bainitic ferrite are coarse due to the large number of prior austenite grain boundaries.

The mechanical properties and microstructures of hexagonal boron nitride (h-BN)-reinforced cement composites are experimentally studied for three and seven curing days. Various sizes (5, 10, and 18 μm) and concentrations (0.1%, 0.25%, 0.5%, and 1.0%) of h-BN are dispersed by the tip ultrasonication method in water and incorporated into the cement composite. The compressive strength of the h-BN reinforced cements increases by 40.9%, when 0.5 wt% of 18 μm-sized h-BN is added. However, the compressive strength decreases when the 1.0 wt% cement composite is added, owing to the aggregation of the h-BNs in the cement composite. The microstructural characterization of the h-BN-reinforced cement composite indicates that the h-BNs act as bridges connecting the cracks, resulting in improved mechanical properties for the reinforced cement composite.