The domestic representative nuclear fuel cycle facilities are post-irradiation examination facility (PIEF) and Irradiated Examination Facility (IMEF) at KAERI. They have regularly operated since 1991 and 1993, respectively. Due to the long period of use, the facilities are ageing, and maintenance costs are increasing every year. The maintenance methods have mainly been breakdown maintenance (BM) and partially preventive maintenance (PM). They involve replacing components that have problems through periodic inspections by on-site inspectors. However, these methods are not only uncertain in terms of replacement cycles due to worker’s deviation on the inspection results, but also make it difficult to respond accidents developed through failures on the critical equipment that confines radioactive material. Therefore, an advanced operation and maintenance studied in 2022 through all of nuclear facilities operated at KAERI. Advancement strategy in four categories (safety, sustainability, performance, innovativeness) was analyzed and their priorities according to a facility environment were determined so a roadmap for advanced operation and maintenance could be developed. The safety and sustainability are higher importance than the performance and innovativeness because facilities at KAERI has an emphasis on research and development rather than industrial production. Thus, strategy for advancement has focused even more on strengthening the safety and sustainability. To enhance safety, it has been identified that immediate improvement of aged structures, systems, and components (SSCs) through large-scale replacement is necessary, while consideration of implementing an ageing management program (AMP) in the medium to long term is also required. Facility sustainability requires strengthening operation expertise through training, education, and cultivation of specialized personnel for each system, and addressing outstanding regulatory issues such as approval of radiation environment report on the nuclear fuel processing facilities and improvement work according to fire hazard analysis. One of the safety enhancement methods, AMP, is a new maintenance approach that has not been previously applied, so it had to be thoroughly examined. In this study, an analysis was conducted on the procedure and method for introducing an AMP. An AMP for nuclear fuel cycle facilities was developed by analyzing the AMP applied to the BR2 research reactor in Belgium and modifying it for application to nuclear fuel cycle facilities. The ageing management for BR2 has the objective to maintain safety, availability and cost efficiency and three-step process. The first step is the classification of SSCs into four classes to apply graded approach. Secondly, ageing risk is assessed to identify critical failure modes, their frequency and precursors. Final step involves defining measures to reduce the ageing risk to an acceptable level in order to integrate the physical and economic aspects of ageing into a strategy for inspection, repair, and replacement. Similar approach was applied to the nuclear fuel cycle facility. Firstly, the SSCs of nuclear fuel cycle facilities have been classified according to their safety and quality classifications, as well as whether they are part of the confinement boundary. The SSCs involved in the confinement boundary were given more weight in the classification process, even if they are not classified as safety-class. A risk index for ageing was introduced to determine which prevention and mitigation measure should be chosen. By multiplying the health index and the impact index, the ageing risk matrix provides a numerical score that represents guidance on the prevention and mitigation of ageing effect. The health index is determined by combining the likelihood of failure and engineering evaluation of the current condition of SSCs, whereas the impact index is calculated by taking into account the severity of consequences and the duration of downtime resulting from a failure. This ageing management has to be thoroughly reviewed and modified to suit each facility before being applied to nuclear fuel cycle facilities.

This study monitored temperature using electronic sensors and developed a prediction model for compost maturity. The experiment used swine manure in a mechanical composting facility equipped with a screw-type agitator, and the composting process was conducted for 60 d during the summer season in South Korea. Four electronic temperature sensors were installed on the inner wall between the compost piles on Days 7, 14, 21, and 28 for daily temperature monitoring. Compost samples were collected daily for 60 d, and compost maturity was analyzed using the Solvita method. Multiple comparisons, correlations, and modeling were performed using the stat package in R software. The average compost pile temperatures was 39.1±3.9, 36.4±4.3, 31.3±4.5, and 35.4±8.1 on days 7, 14, 21, and 28, respectively, after composting. The average compost maturity according to the composting date was 3.61±0.60, 4.13±0.59, 4.26±0.47, and 4.32 ±0.56 on days 7, 14, 21, and 28, respectively. A significant negative correlation was observed between the compost composting periods (seven, 14, 21, and 28 d) and the temperature of all compost piles (p<0.05), where the correlation coefficients were -0.329, -0.382, -0.507, and -0.634, respectively. A significant positive correlation was observed between the compost composting periods (seven, 14, 21, and 28 d) and the maturity of the compost (p<0.05), where the correlation coefficients were 0.410, 0.550, 0.727, and 0.840, respectively. The model for predicting the maturation of the 14 d average compost pile according to the compost composting period and the average temperature for 14 d was y=0.026 x d – 0.021 x mt.x_14 d (mean temperature for 14 d) + 4.336 (R2=0.7612, p<0.001). This study can be considered a basic reference for predicting compost maturity by the proposed model using electronic temperature sensors.

Decommissioning of nuclear power plants generates a large amount of radioactive waste in a short period. Moreover, Radioactive waste has various forms including a large volumes of metal, concrete, and solid waste. The disposal of decommissioning waste using 200 L drums is inefficient in terms of economics, work efficiency, and radiation safety. Therefore, The Korea Radioactive Waste Agency is developing large containers for the packaging, transportation, and disposal of decommissioning waste. Assessing disposability considering the characteristics of the radioactive waste and facility, convenience of operation, and safety of workers is necessary. In this study, the exposure dose rate of workers during the disposal of new containers was evaluated using Monte Carlo N-Particle Transport code. Six normal and four abnormal scenarios were derived for the assessment of the dose rate in a near surface disposal facility operation. The results showed that the calculated dose rates in all normal scenarios were lower than the direct exposure dose limitation of workers in the safety analysis report. In abnormal scenarios, the work hours with dose rates below 20 mSv·y−1 were calculated. The results of this study will be useful in establishing the optimal radiation work conditions.

본 연구는 재가노인복지시설 중 데이케어센터 종사자인 요양보호사의 직무스트레스가 직무만족에 미치 는 영향을 살펴보고, 이들 간의 관계에서 데이케어센터 요양보호사의 자아탄력성이 조절효과를 갖는지 검 증하였다. 데이케어센터 요양보호사의 다양한 업무요인에 따른 직무스트레스를 확인하고, 요양보호사의 자아탄력성이 직무만족에 미치는 영향을 규명함으로써 데이케어센터 요양보호사의 직무만족을 위한 직무 환경 개선방안을 도출 하였다. 본 연구는 요양보호사의 직무스트레스와 직무만족도의 관련성 분석을 위해 자아탄력성을 조절변수로 하 였다. 자료 수집을 위해 서울, 충북, 경남의 데이케어센터에 근무하는 요양보호사들을 대상으로 총 471부를 최종분석 자료로 사용하였다. 자료 분석을 위해 IBM SPSS 및 AMOS 25.0 프로그램을 활용하였다. 본 연 구 결과 첫째, 직무스트레스는 직무만족에 영향을 미치는 것으로 확인했다. 둘째, 자아탄력성은 직무만족 에 영향을 미치는 것으로 확인했다. 셋째, 직무스트레스의 하위요인인 행정적 지원에 스트레스가 높을수 록, 업무관련 스트레스가 높을수록, 보호자 관계에 스트레스가 높을수록 자아탄력성이 낮아짐을 확인했다.

Climate change due to global warming causes a rise in atmospheric air temperature to rise and extreme shift in precipitation patterns. Carbon dioxide (CO2) is widely known as one of the major cause of global warming and accounts for about 72% of total greenhouse gas emissions. Agriculture is more vulnerable to climate change than other industries. Many studies have been conducted to investigate how agroecosystems, both natural and controlled, will respond to the rising level of CO2. Studies on the responses of crops and agricultural environments to climate change are crucial in predicting changes in agro-ecosystems. Research facilities for various types of CO2 treatment have been developed. The representative research facilities are SAR (Soil-Plant-Atmosphere-Research), OTC (Open Top Chamber), FACE (Free Air CO2 Enrichment System), and TGC & CTGC (Temperature Gradient Chamber & CO2-Temperature Gradient Chamber). Therefore, this study reviewed research data and their application in agriculture.

2021년 우리나라 성범죄 발생은 6,321건, 범죄률 13.5%로 교정시설에서는 성범죄 예방을 위해 다양한 프로그램을 운영하고 있지만, 여전히 성범죄의 재범률은 증가하 고 있다. 그래서 본 연구에서는 변증법적 행동 치료이론을 기반으로 성범죄자의 재범 예방 프로그램을 개발하고자 한다. 프로그램의 개발은 선행연구와 요구조사를 바탕으 로 하였다. 프로그램은 변증법적 행동치료(DBT)의 4가지 핵심기술인 마음 챙김, 정서 조절훈련, 고통 감내, 대인 관계 증진 기술로 구성하였다. 연구설계는 비동등성 대조 군 사전-사후 설계(Nonequivalent control group pretest-posttest design)로 혼 합연구 방법(Mixed Methods Design)으로 프로그램의 효과를 검증하였다. 연구대상 자는 G*Power 3.1 program의 표본 크기에 맞게 G시 교정시설에 수감 된 남성 성범 죄자 중에서 총 28명 선정하고, 무작위 할당 표집으로 실험집단 14명과 대조집단 14 명으로 배정하였다. 하지만 이감과 중도 연구 참여 거부로 인한 탈락자를 뺀 실험집단 13명과 대조 집단 12명의 자료를 최종분석하였다. 양적 분석은 SPSS 27.0 프로그램 독립표본 검증(Independent t-test)으로 동질성을 확보를 확인하고, 집단 간 변화를 이원 혼합설계 반복측정 변량분석(Repeated measures ANOVA)으로 살펴보았다. 질적 자료는 Braun과 Clark(2006)가 제시한 주제 분석방법으로 분석하였다. 연구 결 과는 다음과 같다. 첫째, 변증법적 행동치료 프로그램에 참여한 실험집단이 대조집단 보다 대인 간의 공감 반응과 성인 애착에서 통계적으로 유의미하였으며, 변화의 효과 는 추후검사에서도 통계적으로 유의미하게 유지되었다. 둘째, 프로그램 참여 경험을 분석한 결과 36개의 하위 주제, 몇10개의 상위 주제가 나타났다. 이를 분석영역인 충 동성, 공격성, 자기통제력에 재배열하였다. 충동성에서는 이성에의 의지 폭력 행동의 자발적 제어, 예측 능력의 강화, 분노의 원인 탐색 주제가 돌출되었다. 공격성에서는 파괴 본능을 건강한 에너지로 전환, 타인에 대한 적대적 감정 없애기, 낮은 자리에 서 기라는 주제가 출연했다. 자기통제력에서는 유혹을 이길 수 있는 힘의 배양, 현재의 만족보다는 미래의 성공, 규범적인 행동 목표설정이 나타났다. 연구자들은 연구 결과에 근거하여 논의하였고, 교정기관 내에서 성범죄자들의 사회 재적응을 지지할 수 있는 실천적 접근을 제안함으로써 재범을 예방하고자 하였다. 본 연구는 기존의 교정시설 에서 시도하지 않은 변증법적 행동 치료이론을 접목한 프로그램의 개발하였기에 성범 죄자 프로그램에 새로운 패러다임을 시도하였다는 점에서 선행연구들과 차이가 있다.

본 연구는 국내 수상스포츠 시설의 서비스스케이프가 고객만족 및 참여행동에 어떠한 영향을 미치는지를 규명하여 운영방안의 기초자료를 제공하고자 했다. 따라서 수상스포츠 시설을 경험한 사람들 중 20대 이상 참여자를 표본으로 선정했다. 조사는 2022년 4월 25일부터 7월 15일까지 비대면으로 실시했 으며, 총 243부의 데이터를 최종 분석에 이용했다. 자료처리는 SPSS(ver. 21.0) 프로그램을 활용해 빈도분 석, 탐색적 요인분석, 문항내적일관성, 상관분석, 단순 및 다중회귀 분석을 실시했다. 본 연구의 주요 결과, 첫째, 수상스포츠 시설 서비스스케이프가 편리성, 공감성, 시설환경 순으로 가치에 영향을 미치는 것으로 나타났다. 둘쨰, 수상스포츠 시설 고객만족이 참여행동에 영향을 미치는 것으로 나타났다. 셋째, 수상스포 츠 시설 서비스스케이프가 시설환경, 서비스, 편의성, 공감성 순으로 참여행동에 영향을 미치는 것으로 나 타났다.

The Korea Atomic Energy Research Institute operates the Nuclear Cycle Experimental Research Facility which has radiation controlled area in the laboratory with the aim of realizing pyroprocessing technology. In this Facility, depleted Uranium feed material and a depleted Uranium mixed with some surrogate material are used for performing experiments. Therefore the facility is using uranium, users should be careful of radiation. This paper will explain the radiation protection of the Nuclear Cycle Experimental Research facility and will also explain how much alpha radiation comes out from the facility. The RMS (Radiation Monitoring System) detector is made by CANBERRA and the model name is ICAM. ICAM RMS is the detector which can detect Alpha Radiation by absorbing the air in the facility. The RMS detector is installed in the HVAC room on the third floor to check the air contamination through the chimney. The RMS is connected to the air ventilation line for detecting Alpha radiation in the whole facility. Experiment was performed for two weeks to check the radiation level and the air ventilation fan continued to operate 24 hours a day. the results are below the required value which is 0.1 Bq/m3, indicating that the facility is safe in terms of radiation safety management.

For deep geological repository of the spent nuclear fuel, the fuel assemblies loaded in the storage cask are transferred to the disposal cask and the operation is performed in the fuel handling hot cell at the fuel re-packaging facility. As the fuel handling hot cell shielding is accomplished by the concrete wall and the viewing glass window, the required shielding thickness was evaluated for both materials. The ordinary concrete is applied to hot cell wall and two kinds of glasses, i.e., single layer of lead glass and double layer of lead glass and borosilicate glass, are considered for the viewing glass window. A bare spent PWR fuel assembly exposed to the environment in the hot cell was considered as the neutron and gamma radiation sources. The neutron and gamma transport calculations were performed using the MAVRIC program of the SCALE code system for the dose rate evaluation. The dose limit of 10 μSv/h is applied as the target dose to establish the required shielding thickness. The concrete wall of 94 cm thickness reduces the total dose rate to 6.9 μSv/h, which is the sum of neutron dose and gamma dose. Penetrating the concrete wall, both of the neutron dose and the gamma dose decrease constantly with shield thickness and the gamma dose is always dominant through whole penetrating distance. Single layer lead glass of 74 cm thickness reduces total dose rate to 6.2 μSv/h. Applying double layer shield glass combined of lead glass and borosilicate glass, the total dose rate reduces to 3.6 μSv/h at same shield thickness of 74 cm. Through the shield glass, gamma dose decreases rapidly and neutron dose decreases slowly compared with those for concrete wall. In result, neuron dose becomes dominant on the window glass shielding. The more efficient dose reduction of double layer glass is achieved by the borosilicate glass’s superior neutron shielding power. Thus, the use of double layer glass of lead glass and borosilicate glass is recommended for the viewing glass of the fuel handling hot cell. Finally, it is concluded that about 1 m thick concrete wall and 75 cm thick viewing glass window are sufficient for the radiation shielding of the hot cell at the spent fuel repackaging facility.

In Korea, Kori Unit 1, a commercial pressurized water reactor (PWR), was permanently shut down in June 2017, and an immediate decommissioning strategy is underway. Therefore, it is essential to understand the characteristics of radioactive waste during the decommissioning process of nuclear power plants (NPP). Because radioactive waste must be handled with care, radioactive waste is treated in a hot cell facility. Hot cell facility handles radioactive waste, and worker safety is essential. In this study, it was dealt with whether or not the radiation safety regulations were satisfied when processing the core beltline metal of the dismantling waste treated at the post irradiation examination facility (PIEF) of the hot cell facility. Core beltline metal used for the pressure vessel in the reactor is carbon steel, and it is continuously irradiated by neutrons during the operation of the NPP. A radiological safety estimation of the behavior of radioactive aerosols during the cutting process within the PIEF was carried out to ensure the safety of the environment and workers. When processing the core beltline metal in PIEF, dominant six nuclides (60Co, 63Ni, 55Fe, 3H, 59Ni, 14C) of aerosol are generated. Accordingly each cutting device, amount of aerosol and value of dose is different. Using a 99.97% efficiency HEPA filter, the emission concentration of the dominant nuclides (60Co, 63Ni, 55Fe, 3H, 59Ni, 14C) in the air source term was satisfied with the emission control standard of Nuclear Safety Commission No. 2016-16. It was confirmed that the radioactivity concentration in the airborne source term inside the PIEF is in equilibrium state, when ventilation is considered. Also, the mass of aerosol and the concentration of airborne source term differed according to the thickness of the saw blade of the cutting tool, and the exposure dose of the worker was different through Monte Carlo N-Particle (MCNP). At that time, 60Co accounted for 95.4% of the exposure dose, showing that 60Co had the highest impact on workers, followed by 55Fe with 2.7%. The worker’s dose limit is satisfied in accordance with Article 2 of the Nuclear Safety Act and the dose limit of radiation-controlled area is found to be satisfied in accordance with Article 3 of the rules on technical standards for radiation safety management at this time.

Radiological characterization is important in decommissioning and dismantling of nuclear facilities, in order to assess the radioactivity concentration, classify the wastes, and secure workers’ safety. The Some components such as Reactor Pressure Vessel (RPV) in nuclear facilities has dose rate higher than Sv/hr, thus in-situ gamma spectroscopy systems suffer from a very high count rate which causes energy resolution degradation, photo-peak shift, and count loss by pile-up and dead-time. The system must be operated in a very high count rate, in order to measure spectra precisely and to quantify radionuclide contents. In order to apply in-situ measurement in high radiation dose rate environment, the sensor, front-end electronics, and data acquisition (DAQ) should be carefully selected and designed as well as precise design of collimators and radiation shield. In this paper, the components of the detector system were selected and performance was evaluated in a high count rate before design the collimator and shield. A LaBr3 coupled with a PMT having short decay time constant (16 nsec) was selected for high count rate application, and two different amplifiers (a conventional charge sensitive preamplifier with 50 usec decay time constant, and wide-band voltage amplifier) were tested. As DAQs, DT5781 (14 bit, 100 MS/s, CAEN) of Pulse Height Analysis (PHA) which is conventionally used signal processing method in the gamma spectroscopy, and DT5730 (14 bit, 500MS/s, CAEN) of Pulse Shape Discrimination (PSD) which is similar to Charge to Digital Convertor (QDC) were used. The number of photons incident to the detector was varied by changing the detector-source distance with Certificate Radiation Material (CRM), and compared to the output count rate. The count rate capability, and energy resolution with different amplifier and DAQ was evaluated. Additionally, the performance of DAQs in extremely high count rate was evaluated with signal data generated by the emulator which can simulate the detector signal waveforms fed into the DAQ based on the measured spectrum.

During the decommissioning of nuclear facilities, 3D digital model that precisely describes the work environment can expedite the accomplishment of the work. Thus, the workers’ exposure to radiation is minimized and the safety risk to the workers is reduced, while precluding inadvertent effects on the environment. However, it is common that the 3D model does not exist for legacy nuclear facilities as most of the initial design drawings are 2D drawings and even some of the 2D drawings are missing. Even in the case that all of the 2D drawings are intact, these initial design drawings need to be updated using asbuilt data because facilities get modified through years of operation. In those cases, 3D scanning can be a good option to quickly and accurately generate a structure’s actual 3D geometric information. 3D scanning is a technique used to capture the shape of an object in the form of point cloud. Point cloud is a collection of large number of points on the external surfaces of objects measured by 3D scanners. The conversion of point cloud to 3D digital model is a labor-intensive process as a human worker needs to recognize objects in the point cloud and convert the objects into 3D model, even though some of the conversion process can be automated by using commercial software packages. With the aim of full automation of scan-to-3D-model process, deep learning techniques that take point cloud as input and generate corresponding 3D model have been studies recently. This paper introduces an efficient scan simulation method. The simulator generates synthetic point cloud data used to train deep learning models for classifying reactor parts in robotic nuclear decommissioning system. The simulator is built by implementing a ray-casting mechanism using a python library called ‘Pycaster’. In order to improve the speed of simulation, multiprocessing is applied. This paper describes the ray casting simulation mechanism and compares the in-house scan simulator with an open source sensor simulation package called Blensor.

This study introduces the licensing process carried out by the regulatory body for construction and operation of the 2nd phase low level radioactive waste disposal facility in Gyeongju. Also, this study presents the experience and lessons learned from this regulatory review for preparing the license review for the next 3rd phase landfill disposal facility. Korea Radioactive Waste Agency (KORAD) submitted a license application to Nuclear Safety and Security commission (NSSC) on December 24, 2015 to obtain permit for construction and operation of the national engineered shallow land disposal facility at Wolsong, Gyeongju. NSSC and Korea Institute of Nuclear Safety (KINS) started the regulatory review process with an initial docket review of the KORAD application including Safety Analysis Report, Radiological Environmental Report and Safety Administration Rules. After reflecting the results of the docket review, the safety review of revised 10 application documents began on November 29, 2016. Total 856 queries and requests for additional information were elicited by thorough technical review until November 16, 2021. As the Gyeongju and Pohang earthquakes occurred in September 2016 and November 2017, respectively, the seismic design of the disposal facility for vault and underground gallery was enhanced from 0.2 g to 0.3 g and the site safety evaluation including groundwater characteristics was re-investigated due to earthquake-induced fault. Also, post-closure safety assessments related to normal/abnormal/human intrusion scenarios were re-performed for reflecting the results of site and design characteristics. Finally, NSSC decided to grant a license of the 2nd phase low level radioactive waste disposal facility under the Nuclear Safety Laws in July 2022. This study introduces important issues and major improvements in terms of safety during the review process and presents the lessons learned from the experience of regulatory review process.

Important medical radionuclides for Positron Emission Tomography (PET) are producing using cyclotrons. There are about 1,200 PET cyclotrons operated in 95 countries based upon IAEA database (2020). Besides, including PET cyclotrons, demands for particle accelerators are continuously increasing. In Korea, about 40 PET cyclotrons are in operating phases (2020). Considering design lifetime (about 30-40 years) and actual operating duration (about 20-30 years) of cyclotrons, there will be demands for decommissioning cyclotron facilities in the near future. PET cyclotron produces radionuclides by irradiating accelerated charged particles to the targets. During this phase, nuclear reactions (18O(p,n)18F etc.) produce secondary neutrons which induce neutron activation of accelerator itself as well as surrounding infrastructures (the ancillary subsystems, peripheral equipment, concrete walls etc.). Generally, experienced cyclotron personnel prefer an unshielded cyclotron because of the repair and maintenance time. In unshielded cyclotron, water cooling systems, air compressor, and other equipment and structures could be existed for operating purposes. Almost all the equipment and structures are consisted of steel, and these affect neutron distribution in vault especially thermal neutron on the concrete wall. In addition, most of them can be classified as very low level radioactive wastes by Nuclear Safety and Security notice (NSSC Notice No. 2020-6). However, few studies were estimating radioactivity concentrations (Bq/g) of surrounding structures using mathematical calculation/simulation codes, and they were not evaluating the effect of surrounding structures on neutron distribution. In this study, by using computational neutron transport code (MCNP 6.2), and source term calculation code (FISPACT- II), we evaluated effect of the interaction between surrounding structures (including surrounding equipment) and secondary neutrons. Discrepancies of activation distribution on/in concrete wall will be occur depending on thickness of structure, distance between structures and walls, and consideration of interaction between structures and neutrons. Throughout this study, we could find that the influence of those structures can affect neutron distribution in concrete walls even if, thickness of the structure was small. For estimating activation distribution in unshielded cyclotron vault more precisely, not only considering cyclotron components and geometry of target, but also, considering surrounding structures will be much more helpful.

A radioactive waste disposal facility needs to be developed in a way to protect present and future generations and its environment. A safety assessment is implemented for normal and abnormal scenarios and human intrusion scenarios as a part of a safety case in developing a disposal facility for the radioactive waste. The human intrusion scenarios include a well scenario which takes into account various potential exposure groups (PEGs) who use a groundwater well contaminated with radionuclides released from the disposal facility. It is observed that a pumping rate has a negative correlation with the biosphere dose conversion factor (BDCF) in the well scenario. C-14 is shown to be a key radionuclide in the well scenario, and a special model based on the carbon cycle is applied for C-14. For Tc-99, an adsorption coefficient should be adjusted to be suitable for the site. The safety assessment for the radioactive waste disposal facility is successfully carried out for the well scenario. However, it is observed that site-specific models needs to be developed and sitespecific input data need to be collected in order to avoid unnecessary conservatism.

There are various types of level gauging method such as using float, differential pressure, hypersonic, displacement and so on. In this study, among them, the method utilizing the differential pressure was reviewed. The strengths include: the differential pressure type level gauge can measure the level without direct contact of the sensor with media. That is to say, the level can be measured even if the sensor is far away from the tank. And regardless of the size of the tank, the level can be measured if the pneumatic pipes are installed. The weaknesses include: the sensor needs intermedium to recognize the level. The intermedium utilizes a fluid, which is compressed air. It is difficult to handle that compressed air has the properties of a gas. And to make compressed air needs compressor, tank and pneumatic pipes. So if you have many tanks, you need to install exponentially the pneumatic pipes. As well, level measurement range is limited to the points where the pneumatic pipes of the tank is installed. And if a compressed air that supplies to the sensor leaks, uncertainty will increase. A compressed air is colorless and odorless, so it’s difficult to pinpoint the leak. Finally, events like cracks and clogging can cause inaccurate measurement. Rather than using only differential pressure, it is better to use another measurement method according to the situation of the facility.

Untreated waste is temporarily stored on the site of the nuclear power plant. In some nuclear power plants, saturation period of temporary storage waste is less than 10 years away. As untreated waste continues to be generated in nuclear power plants, it could also affect management of operations. Accordingly, CRI is developing the 3.5 generation plasma torch melting facility for waste treatment. The 3.5th generation plasma torch melting facility consists of melter, plasma torch, waste supply device, exhaust gas treatment facility, power supply, etc. Melter is composed of melting chamber for melting control and pyrolysis chamber for waste pretreatment, and dam-type discharge device is adopted to overflow the melt. Plasma torch is hollow type with reversed discharge, has a rating of megawatt class, and has two gas supply lines. It can be used in transfer mode, non-transfer mode and mixed mode. There are three types of device for waste supply. The first is a drum pusher for injecting 200 L drums, the second is a screw-type waste supply and hopper for injecting solid waste, and the third is a nozzle-type waste supply device for injecting liquid waste. Exhaust gas treatment facility was equipped with post combustion chamber, off-gas cooler, high-temperature filter, HEPA filter, reheater, scrubber, ID fan and etc. Power supply of plasma torch operation is designed with a capacity of 1.5 megawatt (Maximum) and consists of channels A and B. Transfer mode, non-transfer mode and mixing mode of plasma torch may be selected through the control of PLC. This paper introduces the composition and function of the 3.5th generation plasma torch melting facility of CRI. In order to solve the problems arising through the operation of the 3rd generation plasma torch melting facility, an optimization plan is applied.