Given the limited terrestrial reserves of uranium (about 4.6 million tons), exploring alternative resources is essential to ensure the long-term supply and sustainability of nuclear energy. Uranium extraction from seawater (UES) is a potential solution to this issue since the amount of uranium dissolved in seawater (about 4.5 billion tons) is approximately 1000 times that of terrestrial reserves. However, the ultra-low concentration of uranium in seawater (about 3.3 ppb) makes it a challenging task to make UES economically feasible. This paper provides an overview of the current status of UES technology, which has evolved over the past seven decades. Starting from inorganic adsorbents such as hydrous titanium oxide in the 1960s, amidoxime-based fibrous adsorbents gained the most attention until the early 2010s due to their ease of deployment in actual seawater conditions and high affinity for uranium. Nowadays, research on organic adsorbents with microstructures is prevailing due to their ability to easily control surface area and compositions. In addition, this study identifies the key issues that need to be addressed to make UES technology economically viable.

The configuration management system was implemented on the basis of a document management system that secured stable understanding, scalability, document security, and convenience in small modular reactor. To reduce the cost and risk of errors, configuration management is implemented to maintain a balance between design requirements, physical configuration, facility configuration information. In the initial stages, configuration change review procedures was developed with the main purpose of change management such as classification system management, configuration control committee management, configuration change review preparation, configuration control committee operation, followup measures, current status and tracking management. The preparation of the configuration change review consisted of preparation, distribution approval, designation of reviewers, review, collection of review opinions, and preparation of resolution results. In the operation of the configuration control committee, it was conducted by designating review members, reviewing members, collecting operation, and approval them. The next step is to supplement and develop the requirements of IEEE Std 828-2012, such as configuration management planning, configuration management control, configuration identification, configuration change control, configuration status monitoring, configuration audit, interface management, and release management. Through this, issue raising, action management, and baseline management will be implemented.

According to the “Law on protection and response measures for nuclear facilities and radiation”, Nuclear Power Plant (NPP) licensees should conduct periodic exercises based on hypothetical cyberattack scenarios, and there is a need to select significant and probable ones in a systematic manner. Since cyber-attacks are carried out intentionally, it is difficult to statistically specify the sequences, and it is not easy to systematically establish exercise scenarios because existing engineering safety facilities can be forcibly disabled. To deal with the above situation, this paper suggests a procedure using the Probabilistic Safety Assessment (PSA) model to develop a cybersecurity exercise scenario. The process for creating cyber security exercise scenarios consists of (i) selecting cyber-attack-causing initiating events, (ii) identifying digital systems, (iii) assigning cyber-attack vectors to a digital system, (iv) determining and adding type for operator’s response, (v) modifying a baseline PSA model, and (vi) extracting top-ranked minimal cut sets, and (vii) selecting a representative scenario. This procedure is described in detail through a case study, an expected cyber-attack scenario General Transient-Anticipated Transient Without Scram (GTRN-ATWS). It refers to an accident scenario for ATWS induced by GTRN. Since ATWS is targeted for cyber training in some NPPs, and GTRN is one of the most common accidents occurring in NPPs, GTRN-ATWS was chosen as an example. As for the cyber-attack vector, portable media and mobile devices were selected as examples based on expert judgment. In this paper, only brief examples of GTRN-ATWS events have been presented, but future studies will be conducted on an analysis of all initiating events in which cyber-attacks can occur.

목적: 본 연구는 하타요가 프로그램이 매우 민감한 사람(Highly Sensitive Person, 이하 HSP)의 스트레스, 불안, 우울 및 자아존중감에 미치는 영향을 규명하고자 하였다. 방법: 연구참여자는 성인 31명(남 8명, 여 23명, 평균연 령 24.29±4.33세)으로, 하타요가 프로그램을 적용한 HSP(n=7), 비HSP(n=9) 실험집단, HSP(n=8), 비HSP(n=7) 통 제집단의 하타요가 프로그램 전·후 효과를 검증하였다. 이를 위해 Highly Sensitive Person Scale, 한국판 스트레 스 자각 척도(PSS), 한국판 우울-불안-스트레스 척도(DASS-21), Rosenberg 자아존중감 척도를 사용하였다. 연구자 료는 기술통계, 사전동질성 검정을 위한 분산분석, 반복측정 분산분석을 활용하여 분석하였으며, 실험집단의 참여 자 중 HSP(n=7), 비HSP(n=6) 참가자와 반구조적 면담을 수행하였다. 결과: 사전·사후 측정시기(집단내 효과)에 따라 DASS우울과 자아존중감의 유의한 차이를 보였으며, HSP의 DASS우울 감소와 자아존중감 향상 폭이 가장 컸다. 측정시기×집단(사전/사후×HSP/비HSP)에 따른 상호작용 효과는 스트레스 자각, DASS스트레스, DASS우울 에서 유의했으며, HSP실험집단의 사후 DASS스트레스 및 우울 감소와 자아존중감 향상이 큰 것으로 나타났다. 측정시기×실험여부(사전/사후×실험/통제)에 따른 상호작용 역시 DASS스트레스, DASS우울에서 유의했는데, 실험 집단은 사후 DASS스트레스 및 우울 감소와 자아존중감의 향상을 보였다. 실험집단만을 대상으로 하였을 때, 측 정시기(집단내 효과)에 따른 스트레스 자각, DASS스트레스·우울·불안, 자아존중감에서 유의한 차이가 나타났다. 측정시기×집단에 따른 상호작용 효과는 스트레스 자각과 DASS우울에서 유의하게 나타나, HSP의 실험 후 스트 레스와 우울감소가 더욱 큰 폭으로 나타난 것을 알 수 있었다. 결론: 본 연구결과 하타요가 프로그램은 HSP여부 와 관계없이 스트레스, 우울 감소 및 자아존중감의 향상에 효과가 있으며, 하타요가 프로그램의 효과는 HSP의 스트레스와 우울 감소에 보다 긍정적이다. 이는 정신건강 문제에 쉽게 노출될 수 있는 매우 민감한 사람(HSP)에 게 하타요가가 긍정적이고 효과적인 정신건강관리 수단이 될 수 있음을 시사한다.

As part of the safety case development for generic disposal sites in Korea, it is necessary to develop generic assessment models using various geosphere–biosphere interfaces (GBIs) and potentially exposed groups (PEGs) that reflect the natural environmental characteristics and the lifestyles of people in Korea. In this study, a unique modeling strategy was developed to systematically construct and select Korean generic biosphere assessment models. The strategy includes three process steps (combination, screening, and experts’ scoring) for the biosphere system conditions. First, various conditions, such as climate, topography, GBIs, and PEGs, were combined in the biosphere system. Second, the combined calculation cases were configured into interrelation matrices to screen out some calculation cases that were highly unlikely or less significant in terms of the exposure dose. Finally, the selected calculation cases were prioritized based on expert judgment by scoring the knowledge, probability, and importance. The results of this study can be implemented in the development of biosphere assessment models for Korean generic sites. It is believed that this systematic methodology for selecting the candidate calculation cases can contribute to increasing the confidence of future site-specific biosphere assessment models.

The nuclear criticality analyses considering burnup credit were performed for a spent nuclear fuel (SNF) disposal cell consisting of bentonite buffer and two different types of SNF disposal canister: the KBS-3 canister and small standardized transportation, aging and disposal (STAD) canister. Firstly, the KBS-3 & STAD canister containing four SNFs of the initial enrichment of 4.0wt% 235U and discharge burnup of 45,000 MWD/MTU were modelled. The keff values for the cooling times of 40, 50, and 60 years of SNFs were calculated to be 0.79108, 0.78803, and 0.78484 & 0.76149, 0.75683, and 0.75444, respectively. Secondly, the KBS-3 & STAD canister with four SNFs of 4.5wt% and 55,000 MWD/MTU were modelled. The keff values for the cooling times of 40, 50, and 60 years were 0.78067, 0.77581, and 0.77335 & 0.75024, 0.74647, and 0.74420, respectively. Therefore, all cases met the performance criterion with respect to the keff value, 0.95. The STAD canister had the lower keff values than KBS-3. The neutron absorber plates in the STAD canister significantly affected the reduction in keff values although the distance among the SNFs in the STAD canister was considerably shorter than that in the KBS-3 canister.

The decommissioning of nuclear facilities produces various types of radiologically contaminated waste. In addition, dismantlement activities, including cutting, packing, and clean-up at the facility site, result in secondary radioactive waste such as filters, resin, plastic, and clothing. Determining of the radionuclide content of this waste is an important step for the determination of a suitable management strategy including classification and disposal. In this work, we radiochemically characterized the radionuclide activities of filters used during the decommissioning of Korea Research Reactors (KRRs) 1 and 2. The results indicate that the filter samples contained mainly 3H (500–3,600 Bq·g−1), 14C (7.5–29 Bq·g−1), 55Fe (1.1– 7.1 Bq·g−1), 59Ni (0.60–1.0 Bq·g−1), 60Co (0.74–70 Bq·g−1), 63Ni (0.60–94 Bq·g−1), 90Sr (0.25–5.0 Bq·g−1), 137Cs (0.64–8.7 Bq·g−1), and 152Eu (0.19–2.9) Bq·g−1. In addition, the gross alpha radioactivity of the samples was measured to be between 0.32–1.1 Bq·g−1. The radionuclide concentrations were below the concentration limit stated in the low- and intermediatelevel waste acceptance criteria of the Nuclear Safety and Security Commission, and used for the disposal of the KRRs waste drums to a repository site.

Anderson-type polyoxometalate (POM) with general formula of [Hy(XO6)M6O18]n- (y=0-6, n=2-8, M=addenda atom, X=heteroatom) represents one of the basic topological structures among POM-type family. Anderson-type POMs have a planar arrangement and two terminal oxygen atoms attached to each addenda metal atom unlike other type. Thus, the Anderson-type POMs have high reactivity and various coordination modes. The various multifunctional organic-inorganic hybrid materials can be synthesized using the Anderson-type POMs as an inorganic building block. Another important feature of the Anderson-type POMs is the incorporation of the heteroatoms with various sizes and oxidation states, which can lead to tune chemical properties. No Anderson-type POMs with early transition metal ions in the heteroatom site have been reported previously. Recently, we reported the synthesis of titanium-containing Anderson-type POM, Na2K6Ti0.92W6.08O24∙12H2O (Ti-POM), which consists of pure inorganic framework built from a central Ti atom surrounded by six WO6 inorganic scaffold. Herein, in-depth studies were conducted to find optimal synthesis conditions such as composition and pH. The success of synthesis was confirmed with Powder X-ray Diffraction that the Ti-POM has a rhombohedral structure with space group of R-3m (No. 166) when the TiOSO4·xH2SO4∙yH2O/ Na2WO4∙2H2O molar ratio is in the range of 0.07 to 0.33. But outside of this range, other unwanted phases coexist. In a basic condition (pH=12), a single-phase Ti-POM with good crystallinity can be obtained, while a Keggin-type POM, NaxK10-x(H2W12O42), was formed through the decomposition of Ti-POM at pH lower than 7.

Attempts to use the molten salt system in various aspects such as MSR or energy storage systems are increasing. However, there are limitations in the molten salt-assisted technique due to the harsh corrosiveness of the molten salt, and a more detailed study on salt-induced corrosion is needed to solve this problem. In this study, corrosion behaviors of 80Ni-20Cr alloy in various salt environments such as eutectic NaCl-MgCl2 with NiCl2, CrCl2, and EuCl3 additives were investigated. Meanwhile, the corrosion acceleration effects of 80Ni-20Cr specimens were analyzed for various ceramic materials such as SiC, Al2O3, SiO2, graphite, and BN, and metallic materials such as Ni-based alloy, Fe-based alloy, and pure metals in a molten salt environment. The experiments were conducted at 973 K for up to 28 days, and after the experiment, the microstructural change of the specimen was analyzed through SEM-EDS, and salt condition was analyzed by ICP-OES.

Nuclear spent fuel (SNF) disposal in deep geological repositories is considered as one of sound options for the long-term and safe sequestration of radiotoxic SNF and the sustainable use of nuclear energy. The chemical behaviors of various radionuclides originated from SNF should be well understood to evaluate the migrational behaviors of radionuclides and their reactions and interactions with various geochemical components. Formation of secondary minerals, colloids, other insoluble precipitates is of interest since the concentrations of radionuclides in groundwaters can be limited by the solubility of those solid phases. Particularly when evaluating their solubility, the use of well-defined solid materials in terms of chemical composition and molecular structure is crucial to obtain reliable measurement results. In this study, a synthetic calcium uranyl silicate (Ca-U(VI)-silicate, or uranophane) was prepared and characterized by using various analytical methods including powder X-ray diffraction (pXRD), scanning electron microscopy/energy dispersive X-ray spectrometry (SEM/EDX), and vibrational (FTIR and Raman) spectroscopies. Uranyl silicate minerals are significant to the disposal of nuclear wastes. Our simulation demonstrates that uranophane (Ca[UO2SiO3OH]2·5H2O), one having a U:Si ratio of 1:1, can be a mineral species limiting U(VI) solubility under groundwater conditions in Korea. For the preparation of Ca-U(VI)-silicate, we applied a two-step hydrothermal synthetic procedure reported in literature with modification. Briefly, we conclude that the obtained mineral phase is the ‘α-uranophane’; our characterization results show that the structural and spectroscopic properties of the synthetic Ca-U(VI)-silicate agree well with those of α-uranophane. For instance, the pXRD patterns obtained from the solid show nearly identical diffraction peak positions with those from the reference XRD pattern. From IR and Raman spectroscopy it is noticed that the stretching modes of UO2 2+ and SiO4 4- ions result in strong absorption bands in a region of 700 ~ 1,100 cm-1. Elemental compositions of the synthetic solids were also estimated by using EDX analysis, which results in a Ca:U:Si ratio close to 1:2:2 on average. However, we found that it is difficult to obtain good crystallinity of uranophane, which can be observable by using SEM and its image analysis. We believe that this work serves as a model study to provide synthetic routes of radionuclide-related mineral phases and applicable solid phase characterization methods. In the presentation, the potential use of the U(VI)-silicate solid phase for the upcoming groundwater solubility measurements will be discussed. Keywords: Hexavalent Uranium, Silicate

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.

This study presents distribution of naturally occurring radioactive materials in groundwater in Jeju island. Radon (222Rn) and potassium (40K) concentrations were performed by using RAD H2O of RAD7 and 940 Professional IC Vario, respectively. In addition, the activities of uranium and thorium nuclides were analyzed by ICP-MS. All of the groundwater samples were collected from 29 sites from August to October 2022. The radon concentrations in groundwater were in the range of 0 to 60 Bq L-1, and there was no groundwater exceeding the range of 148 Bq L-1 proposed by the US EPA. The distribution of uranium in groundwater varied from 0 to 0.6 μg L-1 and did not exceed 30 μg L-1, thresholds indicated by the US EPA.

Dose-rate monitoring instruments are indispensable to protect workers from the potential risk of radiation exposure, and are commonly calibrated in terms of the ambient dose equivalent (H*(10)), an operational quantity that is widely used for area monitoring. Plastic scintillation detectors are ideal equipment for dosimetry because of their advantages of low cost and tissue equivalence. However, these detectors are rarely used owing to the characteristics caused by low-atomic-number elements, such as low interaction coefficients and poor gamma-ray spectroscopy. In this study, we calculated the G(E) function to utilize a plastic scintillation detector in spectroscopic dosimetry applications. Numerous spectra with arbitrary energies of gamma rays and their H*(10) were calculated using Monte Carlo simulations and were used to obtain the G(E) function. We acquired three different types of G(E) functions using the least-square and first-order methods. The performances of the G(E) functions were compared with one another, including the conventional total counting method. The performance was evaluated using 133Ba, 137Cs, 152Eu, and 60Co radioisotopes in terms of the mean absolute percentage error between the predicted and true H*(10) values. In addition, we confirmed that the dose-rate prediction errors were within acceptable uncertainty ranges and that the energy responses to 137Cs of the G(E) function satisfied the criteria recommended by the International Commission.

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.

This study evaluated the synthesis of optimal materials for high efficiency adsorption and removal characteristics of Cs-137 for radioactive contaminated water, and considered thermal treatment methods to stabilize the spent adsorbent generated after treatment. We synthesized a composite adsorbent with a combination of impregnating metal ferrocyanide that improves the selectivity of Cs adsorption with zeolite capable of removing Cs as a support. The Cs removal efficiency of the composite adsorbent was evaluated, and the stability change of Cs according to the high-temperature sintering was evaluated as a stabilization method of the spent adsorbent. The metal ferrocyanide content of the adsorbent was in the range of 11.8~36.0%. The adsorption experiments were performed using a simulated liquid waste to have a total Cs concentration of 1 mg/L while containing a trace amount of Cs-137, and then gamma radioactivity was analyzed. In order to evaluate the stabilization of the spent adsorbent, heat treatment was performed in the range of 500~1,100°C, and the volatilization rate of Cs during heat treatment and the leaching rate of Cs after heat treatment were compared. In the adsorption experiment, the Cs removal efficiency was higher than 99%, regardless of the amount of metal ferrocyanide in the composite adsorbent. In the sintering experiment on the spent adsorbent, it was confirmed that there was no volatilization of Cs up to 850°C, and then the volatilization rate increased as the heating temperature increased. On the other hand, the leaching rate of Cs in the sintered adsorbent tends to significantly decrease as the heating temperature increases, so that Cs can be stabilized in the sintered body. In addition, as the content of metal ferrocyanide increases, the volatilization rate of Cs rapidly increases, indicating that the unstable metal ferrocyanide in the adsorbent may adversely affect the removal of Cs as well as the thermal treatment stability.

Radioactive waste generated in large quantities from NPP decommissioning has various physicochemical and radiological characteristics, and therefore treatment technologies suitable for those characteristics should be developed. Radioactively contaminated concrete waste is one of major decommissioning wastes. The disposal cost of radioactive concrete waste is considerable portion for the total budget of NPP decommissioning. In this study, we developed an integrated technology with thermomechanical and chemical methods for volume reduction of concrete waste and stabilization of secondary waste. The unit devices for the treatment process were also studied at bench-scale tests. The volume of radioactive concrete waste was effectively reduced by separating clean aggregate from the concrete. The separated aggregate satisfied the clearance criteria in the test using radionuclides. The treatment of secondary waste from the chemical separation step was optimally designed, and the stabilization method was found for the waste form to meet the final disposal criteria in the repository site. The final volume reduction rates of 56.4~75.4% were possible according to the application scenario of our processes under simulated conditions. The commercial-scale system designs for the thermomechanical and chemical processes were completed. Also, it was found that the disposal cost for the contaminated concrete waste at domestic NPP could be reduced by more than 20 billion won per each unit. Therefore, it is expected that the application of this technology will improve the utilization of the radioactive waste disposal space and significantly reduce the waste disposal cost.

Decommissioning waste is generated at all stages during the decommissioning of nuclear facilities, and various types of radioactive waste are generated in large quantities within a short period. Concrete is a major building material for nuclear facilities. It is mixed with aggregate, sand, and cement with water by the relevant mixing ratio and dried for a certain period. Currently, the proposed treatment method for volume reduction of radioactive concrete waste was involved thermomechanical and chemical treatment sequentially. The aggregate as non-radioactive materials is separated from cement components as contaminated sources of radionuclides. However, to commercialize the process established in the laboratory, it is necessary to evaluate the scale-up potential by using the unit equipment. In this study, bench-scale testing was performed to evaluate the scale-up properties of the thermomechanical and chemical treatment process, which consisted of three stages (1: Thermomechanical treatment, 2: Chemical treatment, 3: Wastewater treatment). In the first stage, lab, bench, and pilot scale thermomechanical tests were performed to evaluate the treated coarse aggregate and fines. In the second stage, the fine particles generated by the thermomechanical treatment process, were chemically treated using dissolution equipment, after then the removal efficiency and residual of cement in the small aggregate was compared with laboratory results. The final stage, the secondary wastewater containing contaminant nuclides was treated, and the contaminant nuclides could be removed by chemical precipitation method in the scale-up reactors. Furthermore, an additional study was required on the solid-liquid separation, which connected each part of the equipment. It was conducted to optimize the separation method for the characteristics of the particles to be separated and the purpose of separation. Therefore, it is expected that the basic engineering data for commercialization was collected by this study.

To transport radioactive waste generated during the decommissioning of Kori Unit 1, transport containers of various sizes are being developed. Since these radioactive decommissioning waste transport containers are larger than the specifications of the existing IP-2 type transport containers, which are for operational radioactive waste, design of the CHEONG-JEONG-NURI needs to be changed when transporting them to disposal facility using the CHEONG-JEONG-NURI, which carries operational radioactive waste. In this study, design changes of the CHEONG-JEONG-NURI, cargo hold modification plan for efficient loading of radioactive decommissioning waste transport containers and radioactive decommissioning waste container loading arrangement (plan) were evaluated during the design life period (year 2034). First, as only the IP-2 type transport container with a weight of 7.5 tons and size of 1.6 m (W) × 3.4 m (L) × 1.2m (H) can be loaded in the cargo hold, if only the decommissioning radioactive waste containers are to be loaded and transported, cargo hold needs to be reinforced. Second, when both the radioactive decommissioning waste transport container of the same size as the current operating radioactive waste transport container, and the radioactive decommissioning waste transport container of the same size as the ISO-type transport container are to be loaded in the cargo hold of the CHEONG-JEONG-NURI and transported, the overall design changes (cargo hold size and load reinforcement) are required. Third, since the safe working load of the CHEONG-JEONG-NURI crane is 12.5-tons, it shall be replaced with a ship crane of 35-tons or more to handle the decommissioning radioactive waste container smoothly, or a gantry crane used in general port facilities shall be installed. When replacing with a ship crane of 35-tons or more, ship buoyancy, ship stability, and ship structural safety shall be considered. The possibility of moving in all 4 directions for smooth operation, and the possibility of lifting the transport container to a position higher than the height of the CHEONG -JEONG-NURI shall be considered. Loading and transporting all decommissioning radioactive waste containers, which are the same size as IP-2 and ISO-type transport containers, in the cargo hold of the CHEONG-JEONG-NURI is uneconomical due to the need for overall design changes (cargo size and load reinforcement, etc.). Also, delay in delivery of the operation wastes is expected due to a long-term design change period. Therefore, it is considered reasonable to load and transport only the decommissioning radioactive waste transport container, which is the same size as the IP-2 transport container, in the cargo hold.

Reactor pressure vessels and steam generators generated in the process of dismantling nuclear power plants or replaced steam have various shape and occupy a considerable amount of the disposal site when disposed of in original shape. For the development of domestic technologies related to the disposal of large wastes, it is necessary to secure technologies for reducing large radioactive metal wastes, including technologies such as decontamination, cutting, melting, and residual radioactivity evaluation. Cases of disposal of steam generators in Europe and the United States were investigated. Except for u-tubes, steam generators are less contaminated or easily decontaminated, so it is possible to reduce the volume of waste subject to final disposal by exempting a significant amount through appropriate treatment. Korea Hydro & Nuclear Power Co. is currently temporarily storing 24 steam generators at 41.6 billion won. This paper presents a method to exempt more parts of the steam generator and reduce the volume of waste by properly combining mechanical cutting thermal cutting and melting to dispose of the steam generator. Currently the decontamination and dismantling industries of nuclear facilities are gradually expanding around the world. Therefore, it is necessary to localize the treatment technology for metal waste generated during maintenance and dismantling. The result of this study can be used to establish waste reduction and disposal method for dismantling steam generators.

In preparation for the decommissioning of Kori unit 1 of the nuclear power plant (NPP), new containers of package, transportation, and disposal are being developed that reflect the type, generation amount, and radiological characteristics of decommissioning waste. The containers under development have internal volumes of 1 m3 ~ 14 m3 and loading weights of 1 ton ~ 35 tons, which are larger in size and have a higher loadable weight compared to the 200 L drum and IP-2 type transport container currently being used for packaging and transporting waste. So, there is a limit to handling new containers using existing transport systems (cranes, spreaders, forklifts, transport vehicles, etc.). Therefore, in this study, the status of handling equipment in NPP and disposal facilities was reviewed, the flow from the generation to disposal of decommissioning waste was analyzed, and the possibility of handling new container or the necessity of introducing new systems were derived. Except for some high-dose/high-radioactive wastes among decommissioning wastes, all wastes are finally disposed of through decommissioning area, temporary storage facility, waste treatment facility, waste storage facility, and receipt and storage building. The decommissioning area, temporary storage facility, and waste treatment facility are newly established areas for the decommissioning and should be equipped with a spreader/crane with a lifting weight of 15 tons, 35 tons, and 40 tons in consideration of the weight of the package to be handled in the zone. The waste storage facility has a 7.5 tons crane, so it can handle only some of the lower weight of the 5 to 35 tons package that is expected to be handled. Therefore, additional installation of spreaders/cranes, each with a lifting capacity of 15 tons and 40 tons, is required. The maximum loading weight of forklifts and transport vehicles operating at NPP, and disposal facilities is 10 tons and 12.6 tons, respectively. To transport the package, the facility must additionally install 15 tons and 40 tons forklifts, and 40 tons transport vehicles. Since the lifting weight of the crane installed on the transport vessel is also low at 12.5 tons, it is necessary to change the design of the existing or replace it with 40 tons to handle high-weight package. The results of this study will be used as basic data for the establishment of transport systems in the relevant area and facility, and design requirements for each equipment will be derived through additional research.