A radiation shielding resin with thermal stability and high radiation shielding effect has been developed for the neutron shielding resin filled in the shielding shell of dry storage/transport cask for spent nuclear fuel. Among the most commercially available neutron shielding resins, epoxy and aluminum hydroxide boron carbide are used. But in case of the resin, hydrogen content enhances the neutron shielding effect through optimization of aluminum hydroxide, zinc borate, boron carbide, and flame retardant. We developed a radiation shielding material that can increase the boron content and have thermal stability. Flame retardancy was evaluated for thermal stability, and neutron shielding evaluation was conducted in a research reactor to prove the shielding effect. As a result of the UL94 vertical burning test, a grade of V-0 was received. Therefore, it was confirmed that it had flame retardancy. According to an experiment to measure the shielding rate of the resin against neutron rays using NRF (Neutron Radiography Facility), a shielding rate of 91.54% was confirmed for the existing resin composition and a shielding rate of 96.30% for the developed resin composition. A 40 M SANS (40 M Small Angle Neutron Scattering Instrument) neutron shielding rate test was performed. Assuming aging conditions (6 hours, 180 degrees), the shielding rate was analyzed after heating. As a result of the experiment, the developed products with 99.8740% and 99.9644% showed the same or higher performance.

Poor mechanical properties and bacterial infection are the main problems faced by dental restorative resins in clinical use. In this study, graphene quantum dots (GQDs) grafted with imidazole groups and mesoporous silica (MSN) are co-filled in a dental resin to impart excellent antimicrobial activity and mechanical properties to the dental resin. The higher specific surface area of GQDs and MSN results in an increased contact area with the resin matrix, which enhances the strength of the dental composite resin. The introduction of GQDs significantly improves the antimicrobial activity of the resin. The inhibition efficiency of the composite resin against Streptococcus mutans reached 99.9% with the addition of GQDs at only 0.2 wt.%. When MSN and GQDs are co-filled, MSN interferes with the release of GQDs, thus reducing the antimicrobial activity of the dental resin but improving the cyto-compatibility. By reasonably adjusting the amount of GQDs and MSN, the dental composite resin can exhibit excellent antimicrobial properties, mechanical properties and cyto-compatibility at the same time.

We established pretreatment method of solidified cement ion-exchange resin samples generated before 2003 in nuclear power plants for measurement of non-volatile radionuclide activity. A microwave digestion system (MDS) with mixed acid (HCl-HNO3-HF-H2O2) was used to dissolve cement and to desorb non-volatile elements such as Ce, Co, Cs, Fe, Nb, Ni, Re, Sr and U from mixed ion-exchange resin. The content of Ce, Co, Fe, Nb, Ni, Re, Sr, U and Cs after pretreatment of cement plus mixed ion-exchange resin was measured by ICP-AES and ICP-MS, respectively. As iron and strontium are also present in cement, their content after dissolving a certain amount of cement was measured by ICP-AES. All elements except Nb were quantitatively recovered. Especially since the Nb recovery was low at 72.0±2.5%, the MDS following addition of the mixed acid to the resin was operated once more for desorbing Nb from it. Finally the recovery of Nb was over 95%. This sample pretreatment method will be applied to solidified cement ion-exchange resin samples generated in nuclear power plants for assessment of radionuclide inventory.

Cardiovascular disease remains as one of the most common causes of high morbidity and mortality worldwide, despite remarkable medical advances in recent decades. Non-invasive techniques play a preeminent role in prevention of cardiovascular disease by diagnosing it at an early stage and guiding optimal patient management. Nuclear imaging is one of the most powerful means available for noninvasive diagnosis and management of poorly perfused myocardial region resulting from the cardiovascular disease. Several radionuclides are available for monitoring blood flow to cardiac tissue. The most validated radionuclides for these measurements are 13N, 15O, 99mTc, 201Tl and 82Rb. Each of 13N, 15O and 201Tl require the presence of an on-site cyclotron, whereas, 82Rb and 99mTc require only a generator. Rubidium (Rb) is an alkali metal ion that acts biologically like potassium and accumulates in cardiac muscle tissue. Rb has a rapid blood clearance profile which allows the use of 82Rb. It also has an ultra-short physical half-life of 75 sec for non-invasive evaluation of regional cardiac blood flow. There are several advantages of 82Rb over other radionuclides. Having a short half-life significantly reduces the radiation dose to the patient. In addition, 82Rb is a positron emitter, which gives the full advantages of PET such as image quantification with superior sensitivity. Several reports have shown superior diagnostic performances of 82Rb-PET over conventional 99mTc-SPECT. 82Rb can be produced from a generator system by the decay of its 25.6-day half-life parent 82Sr. However, the 82Sr parent is difficult to prepare. In routine generator production, certain purity is required to meet the specification of the product. Since there has been no the use of 82Rb radionuclide for research or medical purpose in Korea, we have plans to produce 82Sr with certain purity and develop a 82Sr/82Rb generator system. These studies can also be applied to remove radioactive Sr from radioactive waste waters. Because ion exchange resin, used for purification of 82Sr from impurities, is also utilized to trap radioactive Sr2+ ions from radioactive waste water. After Fukushima Daiichi nuclear accident, interest in the treatment of radioactive waste water has surged. As one of main fission products of nuclear reactor, 90Sr has been regarded as a hazardous radionuclide with half-life of about 29 years. Therefore, the investigation on ion exchange resin is important for removal of 90Sr from radioactive waste water. Here, we optimized 82Sr purification method using ion exchange resin to establish the most suitable procedure.

This study presents a rapid and quantitative radiochemical separation method for Nb isotopes in radioactive waste samples from the nuclear power plant with anion exchange resin after Fe coprecipitation. After radionuclides were leached from the radioactive waste samples with concentrated HCl and HNO3, the Nb isotopes were coprecipitated with Fe after filtering the leaching solution with 0.45 micron HA filter, while the Sr, Tc and Ni isotopes were in the solution. The Nb isotopes were separated in HCl medium with anion exchange resin. The purified Nb isotopes were measured using a low level liquid scintillation counter after installing quenching curve with standard Nb-94 isotopes. The separation method for Nb isotopes investigated in this study was applied to neutron dosimeter samples from the nuclear power plant after validating the Nb activity concentration with gamma spectrometry system.

HANARO, a multi-purpose research reactor, uses a reflector as heavy water to obtain high neutron flux. Therefore, two ion exchangers were installed to manage the heavy water quality of the reflector system. The operator of HANARO manages it according to the limit value (Conductivity: less than 0.5 mS/m, pH: 5.5~6.5), and if the limit value is not satisfied, the resin must be replaced. The reflector system is in the enclosed structure and it is designed to delay the release of tritium to the outside. Tritium is produced by a nuclear reaction between neutrons and deuterium. Tritium is inhaled into the body in the form of water or vapor, which is likely to cause internal exposure problem. In addition, since tritium spreads to other regions, thorough management is required. Therefore, HANARO measures and manages tritium in Rx and RCI using the bubbler collection method. In this paper, the change in the behavior of tritium due to the replacement of the reflector ion exchanger resin was analyzed. Due to the increase in conductivity of the reflector, the ion exchanger resin was replaced on March 3, 2022. Therefore, the concentration of tritium was measured to be about 5 times higher than usual. It did not exceed the emission limit, and the concentration values of tritium is stably managed by constant monitoring and analysis.

The liquid radioactive waste system of nuclear power plants treats radioactive contaminated wastes generated during the Anticipated Operational Occurrence (AOO) and normal operation using filters, ion exchange resins, centrifuges, etc. When the contaminated waste liquid is transferred to an ion exchanger filled with cation exchange resin and anion exchange resin, nuclides such as Co and Cs are removed and purified. The lifespan and replacement time of the ion exchange resin are determined by performing a performance test on the sample collected from the rear end of the ion exchanger, and waste ion exchange resin is periodically generated in nuclear power plants. In the general industry, most waste resins at the end of their lifespan are incinerated in accordance with related laws, but waste resins generated from nuclear power plants are disposed of by clearance or stored in a HIC (High Integrity Container). Plasma torch melting technology can reduce the volume of waste by using high-temperature heat (about 1,600 degrees) generated from the torch due to an electric arc phenomenon such as lightning, and secure stability suitable for disposal. Plasma torch melting technology will be used to check thermal decomposition, melting, exhaust gas characteristics, and volume reduction at high temperatures, and to ensure disposal safety. Through this research, it is expected that the stable treatment and disposal of waste resins generated from nuclear power plants will be possible.

This study described a way of developing a resin for deviceizing quantum dots. Furthermore, the following conclusions were obtained by developing light curable syrup and UV curable syrup. First, The viscosity of the mixed resin decreased as the content of the diluent increased, and the value was bewteen 4,310 and 1,473cps. Second, haze was measured by using NDH 5000, and all of the synthesized syrups were obtained a haze value of 1% or less with a transmittance of 95% or more in the visible light region. Third, the viscosity of the mixed resin decreased as the temperature increased, and at this point, the viscosity showed a value of 4,219 to 1,128cps. Lastly, As a result of measuring the viscosity of the resin before and after mixing the quantum dot nanoparticles, it appears to be little change.

Abstract In the present study, the effect of nickel nitrate addition as a catalytic precursor for the in situ formation of Ni nanoparticles during the heating process has been investigated on the modification of microstructure and graphitization of amorphous carbon resulting from pyrolysis of phenolic resin. For this purpose, the prepared resin samples were cured in carbon substrate with and without additives at temperatures of 800, 1000, and 1250 °C. XRD, FESEM, and TEM studies were performed to investigate the phase and microstructural changes in the samples during the heating process. In addition to phase and microstructural studies, thermodynamic calculations of the reactions performed for the in situ formation of nickel nanoparticles and their effective factors during the curing process were performed. The results indicated that nickel nitrate is transformed to nickel nanoparticles of different sizes during the reduction process in a reduced atmosphere. The in situ formation of nickel nanoparticles and its catalytic effect led to the graphitization of carbon resulting from the pyrolysis of phenolic resin at a temperature of 800 °C and above. By increasing temperature, the morphology of the formed graphite changed and hollow carbon nanotubes, carbon cells, and onion skin carbon were formed in the microstructure. It was also observed that by increasing the temperature and the amount of additive, carbon nanotubes and their size are increased. A noteworthy point from thermodynamic calculations during the formation of nickel nanoparticles was that the nickel nanoparticles themselves acted as accelerators of nickel oxide reduction reactions and the formation of nickel nanoparticles. This increases the amount of amorphous carbon graphitization resulting from the pyrolysis of phenolic resin which leads to the formation of more carbon nanotubes at higher temperatures.

The dose was evaluated for the workers transporting the spent resin drums from a spent resin mixture treatment facility. The treatment technology of spent resin mixture waste based on microwave was developed to compensate for the shortcoming of the existing one. The mechanism of the facility for the treatment is divided into separation, desorption, condensation and adsorption process. The treated spent resin that has passed through the microwave reactor flows into the spent resin storage tank. As the treatment time elapses, if spent resin accumulates in the spent resin storage tank, it is moved to the drum of the volume of 200 L. The drum must be moved by the worker, in which case radiation exposure to the drum transport worker occurs. It requires the dose evaluation for drum transport workers in terms of radiation safety. Dose evaluation was performed in consideration of the change in the composition ratio and weight of the spent resin mixture, where the working time for transportation was considered from 10 to 120 minutes in 10-minute increment. In the case of 100 kg of the spent resin mixture, the dose range was derived as 4.62×10−3 – 5.90×10−2 mSv for the 100 kg of spent resin, 4.72×10−3– 5.58×10−2 mSv for the 80 kg of spent resin and 20 kg of zeolite and activated carbon, and 5.38×10−3 – 6.32×10−2 mSv for the 60 kg of spent resin and 40 kg of zeolite and activated carbon. In the case of 150 kg of the spent resin mixture, the dose range was derived as 6.83×10−3 – 8.20×10−2 mSv for the 150 kg of spent resin, 7.13×10−3 – 8.22×10−2 mSv for the 120 kg of spent resin and 30 kg of zeolite and activated carbon, and 8.28×10−3 – 8.86×10−2 mSv for the 90 kg of spent resin and 60 kg of zeolite and activated carbon. The estimated maximum doses for each weight (100 kg and 150 kg of mixture) were confirmed to be 3.16×10−1% and 4.43×10−1% of the annual average dose limit of 20 mSv for radiation workers.

Inorganic and organic ion exchange materials were generally applied to liquid processes in nuclear reactor. In the case of heavy-water reactor (HWR), zeolite, active carbon, anion resin, and cation resin were used to treat liquid processes such as reactor primary coolant cleanup and liquid radioactive waste management system. Then, used ion exchangers were stored at storage tanks. Various kinds of nuclides were adsorbed in ion exchange materials. Especially, C-14, long half-life nuclide, was highly concentrated in anion resin, and waste resin was treated as intermediated level radioactive waste (ILW). Thermal and non-thermal methods such as pyrolysis, incineration, catalytic extraction, acid digestion, and wet oxidation have been studied for treating spent resin. However, destructive methods are not suitable due to massive off gas waste containing radioactive species. To solve this problem, various kinds of processes were developed such as acid stripping, PLO process, activity stripping, thermal treatment, and etc. In this study, microwave method is suggested to treat HWR waste resin. C-14 nuclide was selectively removed from waste resin without decomposition of main structure in waste resin. Radioactive waste resin generated from Wolsung HWR unit 1 and unit 2 was treated using microwave method and 95% of C-14 was successfully removed from the radioactive waste resin.

The feasibility study of synthesizing graphene quantum dots from spent resin, which is used in nuclear power plants to purify the liquid radioactive waste, was conducted. Owing to radiation safety and regulatory issues, an uncontaminated ion-exchange resin, IRN150 H/OH, prior to its use in a nuclear power plant, was used as the material of experiment on synthesis of graphene quantum dots. Since the major radionuclides in spent resin are treated by thermal decomposition, prior to conducting the experiment, carbonization of ion-exchange resin was performed. The experiment on synthesis of graphene quantum dots was conducted according to the general hydrothermal/solvothermal synthesis method as follows. The carbonized ion-exchange resin was added to a solution, which is a mixture of sulfuric acid and nitric acid in ratio of 3:1, and graphene quantum dots were synthesized at 115°C for 48 hours. After synthesizing, procedure, such as purifying, filtering, evaporating were conducted to remove residual acid from the graphene quantum dots. After freeze-drying which is the last procedure, the graphene quantum dots were obtained. The obtained graphene quantum dots were characterized using atomic force microscopy (AFM), Fourier-transform infrared (FT-IR) spectroscopy and Raman spectroscopy. The AFM image demonstrates the topographic morphology of obtained graphene quantum dots, the heights of which range from 0.4 to 3 nm, corresponding to 1–4 graphene layers, and the step height is approximately 2–2.5 nm. Using FT-IR, the functional groups in obtained graphene quantum dots were detected. The stretching vibrations of hydroxyl group at 3,420 cm−1, carboxylic acid (C=O) at 1,751 cm−1, C-OH at 1,445 cm−1, and C-O at 1,054 cm−1. The identified functional groups of obtained graphene quantum dots matched the functional groups which are present if it is a graphene quantum dot. In Raman spectrum, the D and G peaks, which are the characteristics of graphene quantum dots, were detected at wavenumbers of 1,380 cm−1 and 1,580 cm−1, respectively. Thus, it was verified that the graphene quantum dots could be successfully synthesized from the ionexchange resin.

In this experiment, the error according to the measurement method was analyzed for the length, angle, radius of curvature, and diameter of the measurement system analysis using a profile projector device used in the field. One-way analysis was performed on each data tested 30 times using a statistical technique. Through the experiment, it was found that an error occurred in each data when measuring the angle according to the measurement method, and the null hypothesis that no error occurred when measuring the length, inside dia and radius was established.

Abstract Recently, the circular economy aiming at elimination of waste and the continual use of resources has attracted a lot of attentions. In the circular system, the resource recovery uses the recycled wastes as the raw material to manufacture new valuable products. This work focuses on a low-cost process, in which an activated carbon (AC) adsorbent was prepared from waste cation exchange resin by calcination and HNO3 activation hydrothermal method. Surface structure and chemistry of AC were characterized by SEM, XRD, FTIR and Boehm titration. It was found that the acid treatment could increase the number of pores and the content of oxygen-containing functional groups on AC surface. In the adsorption experiment, Methylene blue (MB) was used to assess the adsorption capacity of AC. Experimental results showed that the highest efficiency of MB removal was achieved by AC with modification of 4M HNO3, showing the acidification effect on the adsorption capacity of AC. Adsorption isotherms of Langmuir and Freundlich were employed to fit the experimental data. It was shown that MB adsorption on AC is more consistent with Langmuir model that assumes a homogeneous adsorption. In the adsorption kinetic analysis, the adsorption was found to follow the pseudo-second-order model, indicating that adsorption of MB on acidified AC is dominated by chemical adsorption. The study revealed that the waste ion-exchange resin is a proper precursor of carbon adsorbent for MB dye. This low-cost method would specifically reduce the environmental cost of waste disposal.

In this study, a mixed resin containing Bis-GMA was developed to produce a light-emitting sign using quantum dots. As a result of measuring the viscosity, color coordinates change, and luminance of the mixed resin, the following conclusions were obtained. The viscosity of the mixed resin decreased as the content of the diluent increased, and viscosity values ​​ranged from 3,627 to 1,349cps showed as a result. The viscosity of the mixed resin decreased as the temperature increased, and the viscosity showed a value of 5,156 to 1,132cps. For the optical properties of InP/GaP/ZnSe/ZnS quantum dots, the absolute quantum efficiency was 91% at 522nm and 90% at 618nm when the gallium was 0.01%. The luminance of the light-emitting sign using the resin mixed with quantum dots was showed 142.6cd/m2 in white and 104.2cd/m2 in the red region.

When luting indirect restorations with dual-cure resin cement (DCRC), excess cement can be easily removed by performing tack cure of DCRC for a few seconds. The purpose of this study was to evaluate whether different tack cure times affect polymerization shrinkage (PS) of the selected DCRC. One dual-cure resin cement (G-CEM LinkAce, GC) was used for measuring PS in light-cure (LC group), self-cure (SC group), and two tack-cure modes. In the first tack-cure subgroup, tack cure was performed for 1, 2, 3, and 5 seconds, followed by light cure after 2 minutes of remnant removal time in each case (TC-LC groups). In the other tack-cure subgroup, tack cure was performed for the same lengths of time, but followed by self-cure in each case (TC-SC groups). PS was measured by a modified bonded disc method for 1,800 seconds. One-way analysis of variance followed by Duncan’s post hoc test was used to determine any statistically significant differences among the test groups (α = 0.05). When the DCRC was selfcured after tack cure, PS was significantly lower than when it was only self-cured (p < 0.05); however, tack cure time did not affect PS (p > 0.05). When the DCRC was light-cured, PS was not affected by tack cure or tack cure time (p > 0.05). Therefore, tack cure within 5 seconds did not negatively affect the final PS when the DCRC was light-cured after cement remnant removal.