The objective of this experiment was to investigate the effect of drip irrigation volume on tomatoes (Solanum lycopersicum L.) grown in a greenhouse using perlite medium. Plants were treated by three different irrigation treatment I0, I25, and I50 (where irrigation volume of I25 and I50 was 25% and 50% higher than I0, having limited or no leaching). Growth characteristics of plants, yield and water use efficiency were measured. The result showed that plant height, leaf length and leaf width were lowest in the I0 treated plants. However, these parameters were not statistically significant differences between the plants that were grown in the I25 and I50 treatment. Soluble solids content, acidity and dry matter of 111th, 132nd, and 143rd days harvested tomato were higher in the plants irrigated with lowest volume (I0) than the higher volume (I25 or I50). In addition, water content was lower in the 111th and 132nd days of harvested tomatoes from the I0 treatment. The number of big-size tomatoes (>180 g) was significantly higher in the I25 irrigated plants. There was no significant difference in the total number of harvested fruits among the treatments. The average fruit weight and total yield of harvested tomatoes were lowest in the I0 treated plants. The water consumption of tomato was not significantly different amongst the treatments but water use efficiency was lowest in the I0 treatment. Principal component analysis revealed that total soluble solid and acidity of tomato showed a positive correlation between each other. These results suggest that I25 was the optimum irrigation treatment for tomato based on its measured growth characteristics, yield and water use efficiency.

Republic of Korea (ROK) is operating the Integrated Environmental Radiation Monitoring Network (IERNet) in preparation for a radioactive emergency based on Article 105 of the Nuclear Safety Act (Monitoring of Nationwide Radioactive Environment). 215 radiation monitoring posts are monitoring a wide area, but their location is fixed, so they can’t cover areas where the post is not equipped around the Nuclear Power Plants (NPPs). For this, a mobile radiation monitoring system was developed using a drone or vehicle. However, there are disadvantages: it is performed only at a specific cycle, and an additional workforce is required. In this study, a radiation monitoring system using public transportation was developed to solve the above problems. Considering the range of dose rates from environmental radiation to high radiation doses in accidents, the detector was designed by combining NaI (TI) (in the low-dose area) and GM detector (in the high-dose area). Field test was conducted by installed on a city bus operated by Yeonggwang-gun to confirm the performance of the radiation monitoring system. As a result of the field test, it was confirmed that data is transmitted from the module to the server program in both directions. Based on this study, it will be possible to improve the radiation monitoring capability near nuclear facilities.

Boric acid-containing B-10 is used in a nuclear reactor as a coolant and absorbs thermal neutrons generated during nuclear fission in the primary circuit. Boron-containing coolant water waste is generated from maintenance, floor drain, decontamination, and reactor letdown flows. There are two options for aqueous solution waste of boric acid. One is recycling and discharge through filtration, ion exchange, and reverse osmosis. The other is immobilization after evaporation and crystallization processes. The dry powder of boric acid waste liquid can be immobilized by cement, polymer, etc. Before the mid-1990s, concentrated boric acid waste was solidified with a cement matrix. To overcome the disadvantage of low waste loading of cement waste form, a method of solidifying with paraffin was adopted. However, paraffin solids were insufficient to be disposed of as final waste. Paraffin is a kind of soft solidified material and has low compressive strength and poor leaching resistance. As a result, it was decided as an unsuitable form for disposal. In KOREA, paraffin waste form was adopted for boric acid waste treatment in the 1990s. A large amount of paraffin waste forms about 20,000 drums (200 l drum) were generated to treat boric acid waste and were stored in nuclear power sites without disposal. In this study, we want to obtain high-purity boric acid waste by oxidizing and decomposing solid paraffin waste form through a boric acid catalytic reaction. In this reaction, paraffin is separated in the form of various by-products, which can then be treated through a liquid waste treatment device or an exhaust gas treatment device. The proper temperature for sample decomposition during the catalytic reaction was set through TGA analysis. Compositions of by-products and residues generated at each stage of the reaction could be analyzed to determine the state during the reaction. Finally, the boric acid waste powder was perfectly separated from paraffin waste form with disposable products through this pyrolysis process.

When decommissioning a nuclear power plant, the structure must be made to a disposable size. In general, the cutting process is essential when dismantling a nuclear power plant. Mainly, thermal cutting method is used to cutting metal structures. The aerosols generated during thermal cutting have a size distribution of less than 1 μm. The contaminated structures are able to generate radioactive aerosols in the decommissioning. Radioactive aerosols of 1 μm or less are deposited in the respiratory tract by workers’ breathing, causing the possibility of internal exposure. Therefore, workers must be protected from the risk of exposure to radioactive aerosols. Prior knowledge of aerosols generated during metal cutting is important to ensure worker safety. In this study, the physical and chemical properties of the aerosol were evaluated by measuring the number and mass concentrations of aerosols generated when cutting SUS304 and SA508 using the laser cutting method. High-resolution aerosol measuring equipment (HR-ELPI+, DEKATI) was used to measure the concentration of aerosols. The HR-ELPI+ is an impactor-type aerosol measuring equipment that measures the aerosol number concentration distribution in the aerodynamic diameter range of 6 nm to 10 um in real-time. And analyze the mass concentration of the aerosol according to the diameter range through the impactor. ICP-MS was used for elemental mass concentration analysis in the aerosol. Analytical elements were Fe, Cr, Ni and Mn. For the evaluation of physical and chemical properties, the MMAD of each element and CMAD were calculated in the aerosol distribution. Under the same cutting conditions, it was confirmed that the number concentration of aerosols generated from both materials had a uni-modal distribution with a peak around 0.1 um. CMAD was calculated to be 0.072 um for both SUS304 and SA508. The trend of the CMAD calculation results is the same even when the cutting conditions are changed. In the case of MMAD, it was confirmed that SUS304 had an MMAD of around 0.1 μm in size for only Fe, Cr and Mn. And SA508, Fe, Cr, Ni and Mn were all confirmed to have MMAD around 0.1 μm in size. The results of this study show that a lot of aerosols in the range of less than 1 μm, especially around 0.1 μm in size, are generated when metal is cut using laser cutting. Therefore, in order to protect the internal exposure of workers to laser metal cutting when decommissioning NPPs, it is necessary to protect from nano-sized aerosols beyond micron size.

The fuel fabrication facility has been built and is being operated by KAERI since licensing research reactor fuel fabrication in 2004. After almost 20 years of operation, outdated equipment for fabrication or inspection has been replaced by automated, digitalized ones to assure a higher quality of nuclear fuels. However, the generation of a large amount of radioactive waste is another concern for the replacement in terms of its volume and various types of it that should be categorized before disposal. The regulatory body, NSSC (Nuclear Safety and Security Commission) released a notice related to the classification of radioactive wastes, and most accessory equipment can be classified into the clearance levels, called self-disposal waste. In this study, the practice of self-disposal of metal radioactive waste is carried out to reduce its volume and downgrade its radioactivity. For metal radioactive waste, which is expected to occupy the most amount, analysis status and legal limitations were performed as follows: First, the disposal plan was established after an investigation of the use history for equipment. Second, those were classified by types of materials, and their surface radio-contamination was measured for checking self-disposable or not. After collecting data, the plan for the self-disposal was written and submitted to the Korea Institute of Nuclear Safety (KINS) for approval.

In the design of a spent-fuel (SF) storage, the consideration of burnup credit brings the benefits in safety and economic views. According to it, various SF burnup measurement systems have been developed to estimate high fidelity burnup credit, such as FORK and SMOPY. Recently, there are a few attempts to localize the SF burnup measurement system in South Korea. For the localization of SF burnup measurement systems, it is very important to build the isotope inventory data base (DB) of various kinds of SFs. In this study, we performed DeCART2D/MASTER core follow calculations and McCARD single fuel assembly (FA) burnup analyses for Hanbit unit 3 and confirmed the characteristic of the isotope inventory over burnup. Firstly, the core follow calculations for Cycles 1~7 were performed using DeCART2D/MASTER code system. The core follow calculation is very realistic and practical because it considers the design conditions from its nuclear design report (NDR). Secondly, the Monte Carlo burnup analyses for single FAs were conducted by the McCARD Monte Carlo (MC) transport code. The McCARD code can utilize continuous energy cross section library and treat complex geometric information for particle transport simulation. Accordingly, the McCARD code can provide accurate solutions for burnup analyses without approximations, but it needs huge computing resources and time burden to perform whole-core follow calculations. Therefore, we will confirm the effectiveness of the single McCARD FA burnup analyses by comparing the DeCART2D/MASTER core follow results with the McCARD solution. From the results, the use of single FA burnup analyses for the establishment of the DBs will be justified. Various FAs, that have different 235U enrichments and loading pattern of fuel rods and burnable absorbers, were considered for the burnup analyses. In addition, the results of the sensitivity analyses for power density, initial enrichment, and cooling time will be presented.