Nuclear power plants in Korea stores approximately 3,800 drums of paraffin solidification products. Due to the lack of homogeneity, these solidification products are not allowed to be disposed of. There is therefore a need for the separation of paraffin from the solidification products. This work developed an equipment for a selective separation of paraffin from the solidification product using the vacuum evaporation and condensational recovery method in a closed system. The equipment mainly consists of a vacuum evaporator and a condensational deposition recovery chamber. Nonisothermal vacuum TGAs, kinetic analyses and kinetic predictions were conducted to set appropriate operation conditions. Its basic operability under the established conditions was first confirmed using pure paraffin solid. Simulated paraffin solidification product fixing dried boric acid waste including nonradioactive Co and Cs were then fabricated and tested for the capability of selective separation of paraffin from the simulated waste. Paraffin was selectively separated without entertainment of Co and Cs. It was confirmed that the developed equipment could separate and recover paraffin in the form of nonradioactive waste.

During PIV (Physical Inventory Verification), the IAEA has been inspecting the CANDU-Type spent fuels using an optical fiber-based scintillation detector. KINAC has developed a new verification instrument to deal with problems of the existing one such as low sensitivity, heavy and large dimension, and inconvenience-in-use. Our previous studies focused on how to develop the new instrument and had not included its performance tests. Field tests were carried out recently at Wolsung unit 4 to evaluate performance of the existing and new instruments. The objective of this paper is to discuss background noise produced in the optical fiber signal cable itself. The verification equipment for the CANDU-type Heavy Water Reactor spent fuels uses a scintillation detector to bond a scintillation material to the end of an optical signal cable. At this time, the radiation signal obtained by a data acquisition system is the signal generated from the scintillator (p-terphenyl organic scintillator) and the optical signal cable ; The signal produced in the optical cable itself is background noise to degrade the spent fuel verification equipment. To characterize the background radiation noise, the spent fuel bundles at Wolsung Unit 4 were measured using the optical fiber cable without the radiation scintillator. This signal is generated by reaction of the optical cable and the radiation emitted from the spent fuel. From experimental results, it was observed that the background noise signal of the optical cable increased as the optical cable went down in the downward direction, because the cable length irradiated by the radiation increased with the optical cable area in the spent fuel storage pool. Difference in the background noise signal was dependent on the location of the vertical direction and the signal of the new optical cable was up to about 5 times higher than that of the existing cable. While, the new cable has the cross-section area about 3.2 times larger than the old cable. Our past studies showed that total signal amplitude – sum of signals generated from the scintillator and optical fiber - of the new verification instrument was at least about 15 times greater than that of the existing one. Considering the total signal and background noise signal, from this measured results, it was confirmed that the scintillator characteristics – in particular, light output and decay time – has a dominant impact on the signal sensitivity of the newly developed instrument. More details will be discussed at the conference.

A lot of CANDU Spent Fuels (CSFs) have been stored in spent nuclear fuel pools and dry storage facilities. In accordance with the enhanced nuclear regulations, the initial characteristics of CSF should be inspected to ensure the integrity of CSF and the reliable operation of storage system before loading it into a cask for long-term dry storage. For the inspections, an initial characteristics measurement equipment was designed, which is used for Pool-Side Examination (PSE) in the spent fuel pool of the pressurized heavy water reactor nuclear power plant. Measurements using the equipment consist of non-contact inspections and contact inspections. The non-contact inspections do not affect CSF integrity, whereas the integrity of CSF can be reduced during the contact inspections under abnormal operating conditions because the probe of equipment may apply specific loads to the CSF. Therefore, the structural integrity evaluations of equipment and CSF are performed using Finite Element (FE) analyses for four combinations based on two abnormal conditions and two probe positions. The used abnormal conditions are the pressing load condition and the scratching load condition, and two probe positions are the center and bottom of the fuel rod in the longitudinal direction, respectively. In this evaluation, the bottoms of the fuel rod or CSF are defined as the regions facing the bottom surface of equipment. The analysis of the pressing load condition is performed by pressing the probe of the equipment in radial direction of the CSF fuel rod. That of the scratching load condition is carried out by applying a specific radial load to the CSF fuel rod using the probe and then applying the load to the surface of the fuel rod while moving axially along the surface. All combinations are analyzed considering geometric, boundary and material non-linearity under the dynamic load, which is dependent on the equipment operating velocity. The stresses of CSF and equipment components were obtained from these analyses. The maximum stress of each component was generated at the combination on the scratching load condition for the bottom position among the four combinations. The obtained maximum stresses are lower than the yield stress for each component material. Also, the CSF is not overturned due to the support plate of the equipment in all analyses. Therefore, the structural integrity and safety of the equipment and the CSF are maintained under abnormal operating conditions during the inspection using the initial characteristic measurement equipment.

The CTBTO is the Comprehensive Test Ban Treaty Organization to ban all forms of nuclear testing (underwater, air, and underground) worldwide and was adopted at the UN’s 50th annual general meeting in September 1996. As of September 2023, 187 out of 196 countries signed and 178 ratified. The Republic of Korea signed it in 1996 and ratified it in 1999. Several major Annex 2 countries still need to ratify it, and certain countries have not even signed it, so it has not come entry into force. The CTBTO has three verification systems for nuclear tests and consists of the International Monitoring System (IMS), the International Data Center (IDC), and On-Site Inspections (OSI). IMS consists of seismic, hydroacoustic, infrasound, and radionuclide monitoring. The measured data are delivered to IDC, analyzed by CTBTO headquarters, distributed raw data, and analyzed forms to member states. The final means of verification is in the field of OSI and will be operated when CTBT takes effect. Based on the IMS data, inspectors will be dispatched to the Inspected State Party (ISP) to check for nuclear tests. KINAC is attending the Working Group B, OSI technology development verification along with KINS and KIGAM. Since OSI is a means for final verification, integrated capabilities such as seismic and data interpretation and nuclides detection are required. CTBTO continues its efforts to foster integrated talent and modernize OSI equipment. Types of equipment include measurement, flight simulation equipment, and geographic information monitoring systems etcetera. KINAC is also developing equipment to detect contaminated areas using drones and probes. Development equipment is the nuclides detection and measurement of contaminated areas, and it is the equipment that prepares a control center and drops probes into suspected contamination areas to find a location of the radiation source. The probe can be used to track the location where the dose is most substantial through Bayesian estimation and source measurement.

Non-structural elements, such as equipment, are typically affixed to a building’s floor or ceiling and move in tandem with the structure during an earthquake. Seismic forces acting upon non-structural elements traverse the ground and the building’s structure. Considering this seismic load transmission mechanism, it becomes imperative to account for the interactions between soil, structure, and equipment, establishing seismic design procedures accordingly. In this study, a Soil-Structure-Equipment Interaction (SSEI) model is developed. Through seismic response analysis using this model, how the presence or absence of SSEI impacts equipment behavior is examined. Neglecting the SSEI aspect when assessing equipment responses results in an overly conservative evaluation of its seismic response. This emphasizes the necessity of proposing an analytical model and design methodology that adequately incorporate the interaction effect. Doing so enables the calculation of rational seismic forces and facilitates the seismic design of non-structural elements.

Background: Back stabilization exercises have been applied to decrease low back pain. Pilates exercise can help stabilize the low back and pelvic alignment, and uses equipment such as cadillac, reformer, chair, and barrel. Objectives: This study was conducted to confirm the effects of pilates exercise using cadillac equipment and self-exercise using back pain educational materials on pain, pelvic inclination, and single leg stance ability in adult women with back pain. Design: Randomized controlled trial. Methods: Twenty-five participants (female 25) with unspecified low back pain with lumbar hyperextension and anterior pelvic tilt were recruited. 25 adult women with back pain were randomly divided into a pilates group and a selfexercise group. Pilates exercise intervention using a cadillac was administered to the pilates group, and posture and lifestyle education materials for back pain were provided to the self-exercise group. All of the interventions were conducted 3 times a week for 4 weeks, 50 minutes per session, and the results are as follows. Results: Data analysis in each group revealed statistical differences between self-exercise and divided into a Pilates intervention. A visual analog scale (VAS)/oswestry disability index (ODI) decreased pain in both groups, Independent t-test decreased pain in both groups, but there was no significant difference. Data analysis divided into a pilates group revealed statistical differences between a VAS/ODI and pelvic tilt angle and static balance. It has been shown to increase scores or relieve pain. Conclusion: Through this study, it was confirmed that the pilates exercise using cadillac can be effectively used to reduce back pain, improve pelvic alignment, and improve single leg stance ability in adult women with low back pain.

In the era of the 4th Industrial Revolution, Logistic 4.0 using data-based technologies such as IoT, Bigdata, and AI is a keystone to logistics intelligence. In particular, the AI technology such as prognostics and health management for the maintenance of logistics facilities is being in the spotlight. In order to ensure the reliability of the facilities, Time-Based Maintenance (TBM) can be performed in every certain period of time, but this causes excessive maintenance costs and has limitations in preventing sudden failures and accidents. On the other hand, the predictive maintenance using AI fault diagnosis model can do not only overcome the limitation of TBM by automatically detecting abnormalities in logistics facilities, but also offer more advantages by predicting future failures and allowing proactive measures to ensure stable and reliable system management. In order to train and predict with AI machine learning model, data needs to be collected, processed, and analyzed. In this study, we have develop a system that utilizes an AI detection model that can detect abnormalities of logistics rotational equipment and diagnose their fault types. In the discussion, we will explain the entire experimental processes : experimental design, data collection procedure, signal processing methods, feature analysis methods, and the model development.

Various types of tanks are used in nuclear power plants, and sludge composed of various organic substances and inorganic oxides contaminated with radioactive materials may be present at the bottom of a tank of a radioactive waste treatment device. In addition, glassy and fixative oxide contamination layers are accumulated on the inner wall of the tank depending on the tank material, usage and degree of oxidation. Such contaminated sludge is the main cause of radiation exposure to workers when dismantling nuclear power plant tanks. In addition, the waste filters generated by filtration of contaminated sludge is treated as secondary radioactive waste, and this radioactive waste not only occupies a lot of disposal space, but also the disposal cost is continuously increasing. Therefore, it is necessary to develop a technology that does not generate waste filters as much as possible. To solve this problem, NILEPLANT Co., Ltd. registered a patent named “Filtering apparatus” based on previous research and manufactured a rotary filtration membrane device through detailed design. The rotary filtration membrane device is composed of three or more multiple rotary filtration membranes, and can remove fine particles in wastewater as well as sludge accumulated inside a radioactive contamination tank. In addition, considering the site characteristics of special conditions such as nuclear power plants, it was designed to show excellent performance in removing fine particles while minimizing the area where the device is installed. The rotary filtration membrane device is designed and manufactured as a double cylinder structure that combines a hydro cyclone filter type body and an inner partition wall, and is equipped with a filter cloth-based rotary cylinder filter to process sludge through the filter cloth in addition to inertial. In addition, the patented principle enables self-backwashing without stopping the filtration process, extending the life of the filter and minimizing waste filters. The filtration performance, self-backwashing function, and sludge behavior of the rotary filtration membrane device manufactured based on the detailed design were evaluated through experiments, and improvements to obtain more effective filtration performance were derived. Accordingly, it is expected that the more improved rotary filtration membrane device can be effectively used to remove sludge generated during the dismantling of nuclear power plants in the future.

Concrete radioactive waste is divided into surface-contaminated concrete and activated concrete, and although the generation rate varies depending on the operating conditions of the nuclear power plant, it is reported that the amount of surface-contaminated concrete generated is greater. It is reported in the ‘US-NRC Inventory Report’ that 99% of radionuclides in surface-contaminated concrete are distributed within 1 mm of the surface. Since concrete radioactive waste accounts for a large amount of generation after metal radioactive waste, it is necessary to reduce the amount of radioactive waste disposal by applying appropriate treatment techniques to surface-contaminated concrete. In this study, a similar contamination environment work space with the size of 5.4 (W) × 3.6 (L) × 2.5 (H) [m] in which concrete specimens can be fixed on the wall and floor was established. And an integrated decontamination equipment was verified the automation performance for ‘location accuracy’, ‘radioactive contamination level measurement’ and ‘concrete surface laser scabbling’. It was confirmed that the average was 8.3 [mm] in the evaluation of the ‘location accuracy’ for the remote control and movement of the integrated decontamination equipment. For performance verification of ‘radioactive contamination level measurement’ and ‘laser scabbling’, it were used that size of 30×30×8 [cm] ordinary concrete specimens and concrete radioactively contaminated with Co-60 below the regulatory exemption concentration. ‘Radioactive contamination level measurement’ is measured as much as the set range, divied and display the measured values in different colors on the map of the control program. Ordinary concrete specimens are 0.066~0.089 μ Sv/hr, and contaminated concrete specimens are 0.107~0.121 μ Sv/hr, and it was confirmed that they are expressed in different colors on the map. For ‘laser scabbling’, the performance according to the laser scabbling speed and reproducibility with ordinary concrete specimens was verified. As a result, a weight change of up to 1.48 kg was confirmed. Contaminated concrete specimens were subjected to a direct method using a surface contamination detector and an indirect method using a smear paper to measure surface contamination before and after scabbling, and the debris generated after scabbling was analyzed using HPGe.

The purpose of this study is to provide technical issues in upgrade and modification of fuel handling equipment at operating nuclear power plants. The improvement for safety function and performance enhancement of fuel handling equipment has been going on for 20 years since the early 2000’s. This improvement is recently focused on the replacement of components through the performance analysis and the operation and maintenance plan based on replacement cycle of its component. Additionally, it is required to secure spare parts so that it can be operated at all times with compatibility and standardization to other domestic nuclear power plants. The fuel handling equipment is consisted of refueling machine, upender and carriage of fuel transfer system, spent fuel handling machine, new fuel elevator and various tools, and the equipment are linked in systematic interlocks. Fuel handling is a critical task during a nuclear power plant refueling outage. Even minor component defects may stop operation of the whole system and have a significant impact on the overall system process. To achieve this goal, major components that are expected to be replaced for reliable operation are summarized as follows; 1) motor assembly with AC servomotors and driver for bridge, trolley and hoist of refueling machine and spent fuel handling machine, 2) winch motor and drive for upender and carriage of fuel transfer system, 3) operator control console with a HMI PC base PLC (Programmable Logic Controller) control system, 4) positioning and load weighing sensors such as an encoder and a load cell with its support for periodic calibration and maintenance, 5) main power drapped style festoon cable assembly for bridge of refueling machine, 6) pneumatic control assembly for gripper operation of refueling machine, 7) active components (e. g., air motor, hydraulic cylinder and limit switch) to be removable and reinstallable without requiring the water level to be lowered. It is advisable to utilize such various information as it can help to improve reliability of fuel handling as a critical path in upgrade and modification of fuel handling equipment at operating nuclear power plants.

This study analyzed the flow inside floating seedling equipment for Scapharca subcrenata. Due to the aging society of fishing villages, it is impossible to continuously input the labor force. Therefore, it is necessary to improve efficiency. Scapharca subcrenata has high per capita consumption. It serves as an important aquatic food resource. Scapharca subcrenata culture tends to be highly dependent on the natural environment. Production of Scapharca subcrenata is difficult to predict with low stability. In the past, manpower directly installed bamboo nets in mudflats. The seedling equipment devised in this study is a floating type and can be freely moved on the sea according to the prediction of Scapharca subcrenata generation. The flow around the floating seedling equipment was analyzed by numerical analysis. The physical phenomena of the flow around the net inside the floating seedling equipment were visualized. As a result, the space between the floating seedling equipment and the bottom net and the space between the net groups showed a lower flow rate than the inlet flow rate. It is expected that the low flow rate of the floating seedling equipment will have a positive effect on the attachment of Scapharca subcrenata.

본 연구의 목적은 무중력 운동기구를 활용한 하이플랭크 운동 동작 차이에 따른 상완, 체간 및 코어근육의 근활성도를 비교하는 것이었다. 본 연구의 대상자는 20대 남성 6명을 대상으로 진행하였고(연 령, 23.00±0.73 세; 신장, 172.95±2.05 cm; 체중, 66.83±2.75 kg; 신체질량지수, 22.33±0.72 kg/m2), 4가지(HP, HPAW, HPSB 및 HPWT)의 하이플랭크 동작을 수행하였으며, 근활성도 분석을 위하여 구체 적인 표면전극의 부착 부위는 신체 근육의 우측 삼각근, 상완삼두근, 광배근 및 외복사근으로 설정하였다. 본 실험 결과, 삼각근, 상완삼두근, 광배근 및 외복사근의 근활성도는 HPSB 동작 시 가장 높게 나타났다 (p=.000). 따라서 본 연구의 결과는 추후 무중력 운동기구를 활용한 하이플랭크 동작 시 효과적인 운동 프 로그램의 기초자료가 될 것으로 기대된다.

Thin-film shape technology is recognized for its core technology to enhance the technology of LCD, PDP, semiconductor manufacturing processes, hard disks and optical disks, and is widely used to form coated thin films of products. In addition, resistance (electron beam filament) technology for heating is used to manufacture filament for ion implants used in semiconductor manufacturing processes. By establishing an electronic beam filament production system and developing seven specifications of electronic beam filament, it is contributing to improving trade dynamics and increasing exports to Japan through localized media of theoretical imports to domestic companies. In this study, CAE analysis was performed after setting electron beam filament specification and development objectives, facilities and fabrication for electron beam filament production, electron beam filament JIG & fixture design and fabrication followed by electron beam filament prototype. Then, the automation and complete inspection equipment of the previously developed electronic beam filament manufacturing facilities was developed and researched to mass-produce them, to analyze and modify prototypes, design and manufacture automation facilities, and finally, to design and manufacture the complete inspection equipment. In this paper, design and manufacture of electronic beam filament total inspection equipment for mass production were dealted with.

In this study, we performed thermal safety design of the electric module of a heat-loaded equipment with consideration of its heat dissipation performance. Initially, we calculated the heat dissipation of natural convection to choose a cooling method. Based on this, we found that some modules required forced convection and selected an air-cooling method with an outdoor temperature of 43 degrees Celsius, which is the maximum temperature in Korea. Prior to module production, we performed thermal analysis of each module and proceeded with a design to increase the thermal conductivity of the module as a primary step, and subsequently proceeded with Heat Sink design to maximize the heat dissipation performance. After considering various constraints according to the system requirements and designing the cooling path, we experimentally and analytically secured thermal safety at the operating temperature of the equipment.

전기 설비는 원자력 발전소와 같은 발전소에서 발전 시스템을 운영하고 시설물 전체의 안전을 유지하는데 중요한 역 할을 한다. 이러한 전기 설비들은 캐비닛 내부에 설치되어 외부의 위험으로부터 보호된다. 캐비닛 자체 구조와 내부의 전기 장 치들의 내진검증은 발전소의 안전한 운영을 위해 필수적이다. 부분 전기 캐비닛의 내진검증 연구들은 진동대 시험에 의존하여 진행되고 있다. 그러나 시험을 기반으로 한 내진검증의 경우 다양한 종류의 전기 캐비닛의 특성을 포함하기에는 한계가 있다. 이러한 문제는 상세 3D 유한요소 모델을 통한 내진 검증평가는 진동대 시험으로 인한 비용 및 시간을 절약할 수 있고 다양한 형태 및 위치의 전기설비 설치에 따른 내부의 응답 민감성을 확인해 볼 수 있다. 본 연구에서는 단문형 전기 캐비닛의 내진응 답에 대한 연구를 실시하였다. 먼저 진동대 시험으로 얻은 모달해석 결과를 바탕으로 ABAQUS software를 통해 구축된 3D 상세 유한요소 모델을 검증하였다. 그리고 캐비닛 내부의 다양한 기기 설치를 모사하기 위해 캐비닛 하부 및 상부 위치에 캐비닛 총 무게의 2%, 4%, 6%의 질량을 유한요소 모델에 추가하였다. 또한 2개의 고주파 지진과 2개의 저주파 지진을 입력 지진으로 선 정하여 캐비닛 내부 기기의 질량 변화에 따른 캐비닛 내부와 외부의 가속도 응답 변화를 분석하였다.

In Korea, most nuclear power plants were designed based on the design response spectrum of Regulatory Guide 1.60 of the NRC. However, in the case of earthquakes occurring in the country, the characteristics of seismic motions in Korea and the design response spectrum differed. The seismic motion in Korea had a higher spectral acceleration in the high-frequency range compared to the design response spectrum. The seismic capacity may be reduced when evaluating the seismic performance of the equipment with high-frequency earthquakes compared with what is evaluated by the design response spectrum for the equipment with a high natural frequency. Therefore, EPRI proposed the inelastic energy absorption factor for the equipment anchorage. In this study, the seismic performance of welding anchorage was evaluated by considering domestic seismic characteristics and EPRI's inelastic energy absorption factor. In order to reflect the characteristics of domestic earthquakes, the uniform hazard response spectrum (UHRS) of Uljin was used. Moreover, the seismic performance of the equipment was evaluated with a design response spectrum of R.G.1.60 and a uniform hazard response spectrum (UHRS) as seismic inputs. As a result, it was confirmed that the seismic performance of the weld anchorage could be increased when the inelastic energy absorption factor is used. Also, a comparative analysis was performed on the seismic capacity of the anchorage of equipment by the welding and the extended bolt.

In this study, alternative seismic force-resisting systems for plant structure supporting equipment were designed, and the seismic performance thereof was compared using nonlinear dynamic analysis. One alternative seismic force-resisting system was designed per the requirement for ordinary moment-resisting and concentrically braced frames but with a reduced base shear. The other seismic force-resisting system was designed by accommodating seismic details of intermediate and unique moment-resisting frames and special concentrically braced frames. Different plastic hinge models were applied to ordinary and ductile systems based on the validation using existing test results. The control model obtained by code-based flexible design and/or reduction of base shear did not satisfy the seismic performance objectives, but the alternative structural system did by strengthened panel zones and a reduced effective buckling length. The seismic force to equipment calculated from the nonlinear dynamic analysis was significantly lower than the equivalent static force of KDS 41 17 00. The comparison of design alternatives showed that the seismic performance required for a plant structure could be secured economically by using performance-based design and alternative seismic-force resisting systems adopting minimally modified seismic details.

The electrochemical behavior was investigated during the electrolysis of nickel oxide in LiCl-Li2O salt mixture at 650℃ by changing several components. The focus of this work is to improve anode design and shroud design to increase current densities. The tested components were ceramic anode shroud porosity, porosity size, anode geometry, anode material, and metallic porous anode shroud. The goal of these experiments was to optimize and improve the reduction process. The highest contributors to higher current densities were anode shroud porosity and anode geometry.