Once decommissioning begins, it is expected that large amount of radioactive wastes will be produced in a short period of time. The expected amount of radioactive wastes from Kori unit 1 NPP are approximately 80,000 drums (base on 200 L). By minimizing the amount of radioactive wastes generated through decontamination and reduction, KHNP has set the final target for the amount of radioactive wastes to be delivered to the disposal site at approximately 14,500 drums. Here, plasma torch melting technology is an essential technology for radioactive wastes treatment during nuclear power plants decommissioning and operation, because of its large volume reduction effects and the diversity of disposable wastes. KEPCO KPS was able to secure experience in operating Plasma Torch Melter (PTM) by conducting a research service for ‘development of plasma torch melting system advancement technology’ at KHNP-CRI. This study will compare kilo and Mega-Watt class PTM, largely categorized into facility configurations, operating parameters, and waste treatment. Based on this study, it would be desirable to operate PTM with approximate capacity according to the frequency and amount of waste production, and suggest volume for a kilo and Mega-watt class plasma torch in the melting furnace respectively. This plays to its strengths for both a kilo and Mega-watt class PTM.

Dry active wastes (DAWs) are combustible waste generated during the operation and decommissioning of nuclear facilities, and are known to be generated in the amount of approximately 10,000 to 40,000 drums (based on 200 L) per unit. It consists of various types of protective clothing, paper, and plastic bags, and is stored in radioactive waste storage facilities. Therefore, reducing the volume of DAWs is an important issue in order to reduce storage costs and utilize the limited space of waste storage facilities. Heat treatment such as incineration can dramatically reduce the volume of waste, but as the waste is thermally decomposed, CO2, a global warming gas, is generated and there is a risk of emissions of harmful gases including radionuclides. Therefore, a heat treatment process that minimizes the generation of CO2 and harmful gases is necessary. One of the alternatives to incineration is to carbonize DAWs, dispose of carbonized materials below the release standard as non-radioactive waste, and selectively separate and stabilize inorganic components, including radionuclides, from carbonized DAWs. In this study, 13 types of DAWs generated from nuclear power plants were selected and their thermal decomposition characteristics were investigated to design a heat treatment process that replaces incineration. As a result of TGA analysis, the temperature at which thermal decomposition of each waste begins is 260-300°C for cotton, 320-330°C for paper, 315-420°C for synthetic fiber, 350°C for latex gloves. The mass of most samples decreased to less than 1 % of the initial weight after heat treatment, and dust suit and latex gloves had residues of 13.83% and 13.71% of the initial mass, respectively. The metal components of the residue produced after heat treatment of the sample were analyzed by EDS. According to the EDS results, cotton contains Ca and Al, paper contains Ca, Al and Si, synthetic fiber contains Ca, Cu and Ti, latex gloves contain Ca and Mg. Additionally, ICP analysis was performed to quantify the inorganic components. These results are expected to be applicable to the processing of DAW generated at nuclear facilities in the future.

In this research, the dose rate was measured using a backpack-type scan survey device at 4 sites in sites around Nuclear Power Plants (Kori, Wolsong, Hanbit, Hanul), and the radioactivity ratio for each nuclide was evaluated using an high-purity germanium (HPGe) detector. Kori, Wolsong and Hanul power plants were measured within 2 km of the power plant, and Hanbit power plants were measured about 6.7 km from the power plant. As a result of measuring the dose rate with a backpacktype scan survey device, the average dose rate was the lowest in the measurement site 1 at 0.090 μSv/h, and the highest in the measurement site 4 at 0.145 μSv/h. All measurement points showed the domestic environmental dose rate level. The data obtained by the scan survey was visualized using the classed post and gridding functions of the surfer program. As a result of measurement with the HPGe detector, 137Cs was not detected, and only natural nuclides were detected. Among the detected natural nuclides, the radioactivity ratio was the highest for 40K with an average of 94.56%, and the lowest for 214Pb with an average of 0.26%. The results of this research can be used as basic data for radiation environment surveys around nuclear power plants. Further studies are needed to evaluate the radiation impacts by region and environment through periodic measurements.

KHNP-CRI has developed small-capacity and Mega-Watt Class PTM (Plasma Torch Melter) for the purpose of reducing the volume of radioactive waste and immobilizing or solidifying radioactive materials. About 1 MW PTM is a treatment technology that operates a plasma torch and puts drumshaped waste into a melter and radioactive waste in the form of slag is discharged into a waste container. The small-capacity PTM is a treatment technology that operates a plasma torch and puts small amounts of radioactive waste by directly putting it into the melter through a waste input machine. Mega-Watt Class PTM was able to inject radioactive waste in drums, so it was disposed of without backloging. On the other hand, The small-capacity PTM put radioactive waste without a package, and the waste input was blocked. If even small-capacity PTM put radioactive waste in the form of small packages such as drums, it is expected that various types of radioactive waste can be processed for a long time. Packaging also reduces the risk of radioactive contamination.

Thermal treatment, such as combustion, is the most effective way to solve the spatial problem of radioactive waste disposal. Existing incineration technology has the problem of discharging harmful pollutants (CO2 and dioxin, etc.) into the environment. Therefore, it was evaluate the validity of the thermal treatment process that can reduce the volume of dry active waste (DAW) in an eco-friendly. In addition, the stability of the alternative incineration process under development was evaluated by evaluating the emission of harmful pollutants to the environment during the thermal treatment process. We selected 14 samples identical to those discarded by each nuclear power plant (Kori, Saeul, Wolsong, Hanbit, Hanul). And EA (Elemental Analysis) analysis was performed on each sample. As a result, excluded samples containing wastes containing POPs (Persistent Organic Pollutants) such as PCBs (Polychlorinated Biphenyls), which could generate harmful pollutants during thermal treatment, and halogenated organic wastes such as PVC (Polyvinyl Chloride). In addition, the thermal treatment conditions for the four DAWs were derived by Thermogravimetric Analysis/Differential Thermal Analysis (TG/DTA) analysis. At this time, Py-GC/MS analysis was performed at the temperature at which each waste causes thermal decomposition (cotton is 437°C, paper is 562°C, latex glove is 430°C, plastic bag is 485°C). As a result of analyzing the exhaust gas produced during thermal decomposition, about 77.0% of the cotton was Benzoic acid series, the paper was 41.1% Glucopyranose series, and 15.8% hydroxy acetaldehyde. Latex glove was identified to be 45.9% and 19.2% for Limonene and 2-methyl-1, 3-Butadiene, and for plastic bags, Octacosanol and 2-octyl-1-Dodecanol were 38.8% and 15.2%. In addition, it was confirmed that dioxin and harmful heavy metals, which are discussed as environmental risks, were not detected in all samples.

The characterization of nuclear materials is crucial for global nuclear safeguards efforts, as these materials can potentially be used for illicit purposes. In this study, we evaluated the applicability and performance of the In-Situ Object Counting System (ISOCS) equipment for the characterization and quantification of uranium, including uranium pellets and radioactive wastes. Our methodology involved using ISOCS to measure samples with different enrichments and total amounts of uranium, and to analyze the results in order to evaluate the ISOCS’s effectiveness in accurately characterizing the various uranium samples. To this end, we compared the ISOCS results with those of the Multi-Group Analysis for Uranium (MGAU) system, which is currently used in the field of international safeguards. The results of this study showed that the ISOCS was sensitive enough to analyze small amounts of uranium pellet, with %differences ranging from -0.7% to 19%. However, when analyzing shielded nuclear materials like in concrete waste, the uncertainty was relatively high, with %differences ranging from 11% to 67%. On the other hand, the MGAU system was unable to analyze uranium for the same spectrum, indicating the superiority of the ISOCS in terms of usability. The ISOCS instrument was also found to be effective in analyzing uranium in various types of samples without the need of standard sources. Overall, the findings of this study have important implications for the development of more effective safeguards strategies for the characterization of nuclear materials. The ISOCS instrument could be a reliable tool for analyzing nuclear materials, contributing to global safeguards efforts to reduce the risk of nuclear proliferation.

We conducted multi-elements determination of reference material certified by the Inorganic Ventures, IV-26, using iCAP 7400 ICP-OES of Thermo Fisher Scientific. And we statistically evaluated analysis results by introducing the in-house proficiency evaluation method implemented at the Ministry of Food and Drug Safety. Ca, Co, Fe, Mg, Ni, and V were selected as target elements, and extended uncertainty was estimated at a confidence level of about 95% and coverage factor k = 2. Five parameters incurred at manufacturing process (standard solution, calibration curve, repeated measurement and dilution factor of the test sample) were considered when determining the uncertainty. En-score can be calculated using the formula En=(x-X)/(Ulab 2+Uref 2)1/2 described in KS Q ISO 13528, where x, Ulab, X, and Uref are the test results, the uncertainty of the result, and the certified value and the uncertainty of the value. And if the absolute value |En| is less than 1, it can be evaluated as a satisfied value. As a result of ICP-OES analysis, each concentration of the elements to be measured was almost similar to the certified concentration of the reference material, and the uncertainty was slightly different. Also since evaluation on multi-elements determination had an En-score within 1, it was confirmed that the analysis results satisfied En evaluation.

n this research, the dose rate was measured using backpack-type scan survey device at 4 sites on Jeju Island, and the radioactivity ratio for each nuclide was evaluated using an high-purity germanium (HPGe) detector. As a result of measuring the dose rate with a backpack-type scan survey device, the average dose rate was the lowest in the measurement site 3 at 0.049 Sv/h, and the highest in the measurement site 1 at 0.066 Sv/h. The average dose rate of the 4 sites on Jeju Island was 0.056 Sv/h, and the dose rate on Jeju Island was lower than that of other regions. The data acquired by scan survey were interpreted using classed post and gridding function of surfer program. The radioactivity ratio of each nuclide in the gamma spectrum measured by the HPGe detector was the highest for K-40 with an average of 87.62%, and the lowest for Pb-214 with an average of 0.63%. In the case of the Jeju Island site, Cs-137 was detected, and the average radioactivity ratio of Cs-137 was 3.27%, which was the background level. The results of this research can be used as basic data on the radioactivity ratio for each nuclide and dose rate at the Jeju Island site. Further studies on the assessment of dose rates and radioactivity ratios in other regions are needed.

KHNP-CRI has developed Mega-Watt Class PTM (Plasma Torch Melter) for the purpose of reducing the volume of radioactive waste and immobilizing or solidifying radioactive materials. About 1 MW PTM is a treatment technology that operates a plasma torch and puts drum-shaped waste into a melter and radioactive waste in the form of slag is discharged into a waste container. Since only the overflowing slag is discharged from the melter, the discharge is intermittent. Therefore, solidification occurs in the process of discharging the melt. It is difficult to accumulate evenly in the waste container, and there is also an empty space. Solid radioactive waste must be disposed of to meet the acceptance criteria for radioactive waste. Plasma-treated solid waste raised concerns about the requirements. The waste solidification output in a slag container gave us some concerns for the waste package’s solidification and encapsulation requirements. The plasma-treated solid waste process to meet the acceptance criteria will be cost and need time consuming. Thus, a induction heating will be introduced to meet solidification requirements and test criteria of the solidification waste for the waste package disposal.

To ensure the peaceful use of nuclear energy, nuclear safeguards are applied in member states of the International Atomic Energy Agency (IAEA) under the Non-Proliferation Treaty. The two major considerations in implementing nuclear safeguards are effectiveness and efficiency. In terms of efficiency, the IAEA has a great interest in using containment and surveillance (C/S) technology to maintain continuity of knowledge. A representative means of C/S technology is a sealing system to detect tampering. The existing sealing systems used by the IAEA are of limited functionality in realtime verification purposes. To address this limitation, the present study develops a real-time verification sealing system. First, we analyzed the design requirements of a sealing system proposed by various institutions including the IAEA, the U.S. Nuclear Regulatory Commission, and a number of national laboratories and companies. Then, we identified the appropriate design requirements of this system for real-time verification. Finally, the prototype system was developed and tested based on the identified design requirements. The validation tests of the prototype system were performed for anticipated environmental conditions, radiation resistance, and safeguards functionality. Additionally, we are developing user-friendly verification software. The software validation is planned to perform for functionality, performance efficiency, and security. The next step is to develop a commercialized realtime verification sealing system based on the results of validation tests. Using this commercialized system, we plan to evaluate the performance in various actual use cases. Such a system is expected to significantly enhance the efficiency of nuclear safeguards.

Due to the Fukushima nuclear accident, a large amount of radioactive material was released into the atmosphere, and consequently, it spread over a wide area and was deposited into the soil. As a result of this, a wide area of radioactive contamination site was created. Due to the contaminated site, the need for research on various exploration platforms for efficient situation management and field response is being emphasized. Backpack-type radiation survey & monitoring equipment is useful for creating a contamination maps containing information such as Dose Rate, Radionuclide, Activity Concentration accompanied by spatial analysis when performing a Scan Survey that moves with a backpack on a wide area site. contamination maps are based on accurate radiological characteristic information. However, there is a problem in that the accuracy of the evaluation results is lowered due to changes in environment conditions or the variability of the dose rate and counting rate during scan survey. This problem should be solved by applying the influence of each variable to the underlying data. However, prior to this, it is most important to prepare the base underlying first. And this can be obtained through evaluation of detection performance through static survey. Therefore, in this study, the change in detection efficiency for the measurement height and radius of the backpack-type radiation survey & monitoring equipment based on the 3"×3" NaI(Ti) detector was evaluated. First, the height of the source and Backpack-type radiation survey & monitoring equipment was evaluated from 0 cm to 1 m, which is the height of the soil and detector when an adult male wears a backpack. The experiment was conducted using the 137Cs (383 kBq) point source, which is a nuclide mainly detected at the contaminated site. The measurement time was measured five times per one minute, considering that it was backpack-type equipment and a future scan survey. In addition, in order to evaluate the detection radius, the measurement was performed by changing the measurement distance up to 5 m at intervals of 50 cm. As a result of evaluating the detection performance of the backpack-type radiation surveys & monitoring equipment, it was confirmed that increasing the detection height and radius reduces the count rate in the form of an exponential function. In addition, it was confirmed that the detected radius varies depending on the height. Based on these results, we plan to conduct additional research to understand the scan survey and its sensitivity to various factors. Through this, the company plans to develop various models for exploring the site by improving the accuracy of backpack-type radiation surveys & monitoring equipment.

Attention has been paid to the source term released after Chernobyl and Three Mile Island (TMI), which were the representative accidents of nuclear power plants, and has been studied several times in order to predict and evaluate radiation source term, which can be released in the event of a virtual accident. In particular, the impact of the accident was assessed on the basis of Deterministic Safety Analysis (DSA) and after the WASH-1400, the technology of the Probabilistic Safety Assessment (PSA) was introduced, supplementing safety by taking into account the existence of uncertainty. After the Fukushima accident, a SOARCA report was published to evaluate the specific classification of each type of accident, the realistic progress of the accident, and the leakage of radioactive materials. In this paper, the evaluation methodology and results of the source term of severe accident before and after the Fukushima accident were compared, and the evaluation methods applied to domestic nuclear power plants were compared. Prior to the Fukushima accident, the behavior of the accident and source term were evaluated for Loss of Coolant Accident (LOCA), which led to design based accidents, Total Loss of Feed Water (TLOFW) followed by Station Blackout (SBO) the results were compared to Chernobyl and TMI based on the resulting data to evaluate safety and reliability. After the Fukushima accident, the Interfacing System Loss of Coolant Accident (ISLOCA) and the Steam Generator Tube Rupture (SGTR), which is classified as containment’s bypass accident, were included for predictive assessment. This is due to the analysis that the risk of cancer and early mortality are affected. MACST facilities and strategies were added to domestic nuclear power plants, and accidents with a high core damage frequency were mainly interpreted. In addition, source term was evaluated with the addition of a Basement Melt-Through (BMT) accident that had not previously been considered as a focus. As a result of the comparison of source term evaluation, accidents can be caused by a number of unidentified problems, and Korea’s experience on Level 2 and 3 has not been accumulated, making it difficult to predict the results of source term evaluation or lack of reliability.

For the peaceful use of nuclear energy, the international community has devoted itself to fulfilling its obligations under the Safeguards Agreement with IAEA. In this regard, uranium in a radioactive waste drum should be analyzed and reported in terms of mass and 235U enrichment. In order to characterize radioactive wastes, gamma spectroscopy techniques can be effectively applied. In the case of high-resolution gamma spectroscopy, because an HPGe detector can provide excellent energy resolution, it can be applied to analyze a mixture having a complicated isotopic composition. However, other substances such as wood, concrete, and ash are mixed in radioactive waste with various form factors; hence, the efficiency calibration is difficult. On the other hand, In Situ Object Counting System (ISOCS) has a capability of efficiency calibration without standard materials, making it possible to analyze complex radioactive wastes. In this study, the analysis procedure with the ISOCS was optimized for quantification of radioactive waste. To this end, a standard radioactive waste drum at KEPCO NF and low-level radioactive waste drums at Korea Radioactive Waste Agency (KORAD) were measured. The performance of the ISOCS was then evaluated by Monte Carlo simulations, Multi-Group Analysis for Uranium (MGAU) code, and destructive analysis. As a result, the ISOCS showed good performance in the quantification of uranium for a drum with the homogenized simple geometry and long measurement time. It is confirmed that the ISOCS gamma spectroscopy technique could be used for control and accountancy of nuclear materials contained in a radioactive waste drum.

Background: Although recent studies using virtual reality have been actively conducted, studies on exercise therapy interventions and changes in muscle characteristics in patients with scoliosis are limited. Objectives: To investigate the effects of trunk and pelvic exercises program using virtual reality on the characteristics of paraspinal muscles in patients with scoliosis. Design: A randomized controlled trial. Methods: Thirty college students with scoliosis were randomly assigned to the trunk and pelvic exercise program using virtual reality group (VRG; n=15) and the general stretching exercise group (SG; n=15). The intervention was conducted three times a week for 8 weeks for 30 min per session. The recorded variables were muscle tone (F), muscle stiffness (S), and muscle elasticity (D). The collected data were analyzed after statistical processing. Results: In the cervical vertebrae, F showed significant differences in the amount of inter-group variation in the VRG compared that in the SG on the right side. The VRG showed significant increases in intra-group variation on both the right and left sides. S showed significant differences in the amount of inter-group variation in the VRG compared to that in the SG on the right side. With regard to the intra-group variation of F in the thoracic vertebrae, the right muscle in the VRG demonstrated significantly increased after the experiment, while D significantly increased in the left muscle in the VRG. Conclusion: Pelvic and trunk exercises program using virtual reality may be applied a better therapeutic intervention for patients with scoliosis.

This study was carried out to find out the changes in the growth characteristics and feed value of the three different whole-crop silage rice cultivars of whole-crop silage rice such as Jonong, Yeongwoo and Mogwoo to develop an efficient double cropping system. This study showed that there were significant differences biomass and feed values among cultivars but no clear difference among transplanting dates. Dry weight and height were in order of Mogwoo, Yeongwoo, Jonong (p<0.05). Dry weight and feed value of Jonong showed no significant difference after 21 days after heading (DAH), it was expected to be harvested before DAH 30 days. Yeongwoo showed a lower dry weight than Mogwoo, but heading date was earlier than Mogwoo, so one can expect a higher feed value than Mogwoo. Mogwoo had lower crude protein and total digestible nutrient than the other two cultivars but relative feed value in stem was higher than that of the other cultivars, but had higher dry weight than other cultivars so it was considered to take an advantage as a silage rice. Therefore, the results of this study suggest that the selection of whole-crop silage rice on the cropping system be made comprehensively by considering the heading characteristics of the cultivars and the feed value.