Radiological characterization, one of the key factors for any successful decommissioning project for a nuclear facility, is defined as a systematic identification of the types, quantities, forms, and locations of radioactive contamination within a facility. This characterization is an essential early step in the development of a decommissioning plan, in particular during transition period after permanent shutdown of the facility, and also to be used for classification of decommissioned radioactive wastes so that their disposal criteria can be met. Therefore, the characterization should be well planned and performed. In the transition period, the characterization information developed during the operational phase is usually reexamined with respect to the applied assumptions, the actual status of the facility after shutdown, the accuracy of the required measurements and changes in its radiological properties to support the development of the final decommissioning plan. Based on some national (Korean, USA’s and Japanese) laws including the related regulations, and some related documents published by OECD/NEA, IAEA, and ASTM, key elements of radiological characterization, which should be developed in the transition period, could be proposed as the followings. The key elements might be an operational history including facility operation history and contamination by events and/or accidents, radiological inventory of the facility and site area, characterization survey including in-situ survey and/or sampling and analyses, radiological mapping (which is able to identify radiological contamination levels of SSCs, and the facility area and, if contamination may be suspected, the surroundings) with tabulating, residual radioactivity (or derived concentration guideline levels) of selected major radionuclides for remediation of the site, (retainable and retrievable) recording, and quality control and quality assurance. In review process of the operational history, interviews of current or former long-tenured knowledgeable employees of the facility should be conducted to identify conditions that may have been missing from the records.

The type of accidents associated with the operation of a melting facility for radioactive metal waste is assumed to only marginally differ from those associated with similar activities in the conventional metal casting industry or the current waste melting facility. However, the radiological consequences from a mishap or a technical failure differ widely. Three critical and at the same time possible accidents were identified: (1) activity release due to vapor explosion, (2) activity release due to ladle breakthrough, (3) consequences of failure in the hot-cell or furnace chamber not possible to remedy using remote equipment.

The remote dismantling system proposed in this paper is a system that performs the actual dismantling process using the process and program predefined in the digital manufacturing system. The key to the successful applying this remote dismantling system is how to overcome the problem of the difference between the digital mockup and the actual dismantling site. In the case of nuclear facility decommissioning, compensation between the virtual world and the real world is difficult due to harsh environments such as unsophisticated dismantling sites, radiation, and underwater, while offline programming can be proposed as a solution for other industries due to its sophisticated and controllable environment. In this paper, the problem caused by the difference in the digital mockup is overcome through three steps of acquisition of 3D point cloud in radiation and underwater environment, refraction correction, and 3D registration. The 3D point cloud is acquired with a 3D scanner originally developed in our laboratory to achieve 1 kGy of radiation resistance and water resistance. Refraction correction processes the 3D point cloud acquired underwater so that the processed 3D point cloud represents the actual position of the scanned object. 3D registration creates a transformation matrix that can transform a digital mockup of the virtual world into the actual location of a scanned object at the dismantling site. The proposed remote dismantling system is verified through various cutting experiments. In the experiments, the cutting test object has a shape similar to the reactor upper internals and is made of the same material as the reactor upper internals. The 105 successful experiments demonstrate that the proposed remote dismantling system successfully solved the key problem presented in this paper.

The radwaste facility management team is preparing for clearance of 4 MCAs in The Radwaste Form Test Facility (RFTF). The targeted waste was used for clearance level radioactive waste sample analysis and has been used for this purpose since the early 2000s. Due to the characteristics of clearance level radioactive waste, the concentration of radioactivity is very low and MCA is used with Marinelli beakers the possibility of contamination is low. Moreover, the radiation detector should not be contaminated with radioactive materials, it should be less than the clearance level. These detectors were considered surface contamination materials. To detect the contaminated spot of each material, we scanned the whole surface of a material with a gamma survey meter. After that, we took a sample from 1×1 m2 and a total of 30 samples from each MCA. The wiped filter paper was analyzed with alpha, beta low background counting systems. The results of the analysis of the smear sample of total alpha and beta nuclide radioactivity were less than MDA (α: 2.88×10−5 Bq·cm−2, β: 3.07×10−5 Bq·cm−2). The major nuclide in this facility is Co-60 and Cs-137 therefore we analyzed gamma nuclide activity with HPGe. The maximum specific activity was Co-60: 2.31×10−5 Bq·cm−2, Cs-137: 1.96×10−6 Bq·cm−2. If it is satisfied with the clearance criteria, detectors will be reused at the Radioactive Waste Treatment Facility (RWTF) room # 7251 uncontrolled area.

To obtain confidence in the safety of disposal facilities for radioactive waste, it is essential to quantitatively evaluate the performance of the waste disposal facilities by using safety assessment models. Thus, safety assessment models require uncertainty management as a key part of the confidencebuilding process. In application to the numerical modelling, the global sensitivity analysis is widely employed for dealing with parametric and conceptual uncertainties. In particular, the parametric uncertainty can be effectively reduced by minimizing the uncertainty of critical parameters in the safety assessment model. In this paper, the numerical model of each step disposal facility (Silo, Near surface, and Trench type) at Wolsong Low and Immediate Level Waste (LILW) Disposal Center is designed by using a two-dimensional finite element code (COMSOL Multiphysics). In order to determine the critical parameters for non-adsorbed nuclides such as H-3, C-14, Tc-99, we introduced the variance-based sensitivity analysis methodology of the global sensitivity analysis. In the case of Silo type, the density of waste is highly sensitive to the total leakage quantity of all nuclides. Additionally, the initial nuclide concentration of H-3 was identified as another important parameter of H-3. On the other hands, the mass transport coefficient showed a high contribution in C-14 and Tc-99. In other types of disposal facilities, the leaking properties of H-3 are significantly affected by the amount of infiltration water. However, C-14 and Tc-99 were found to be more sensitive to the density of waste.

Safety for the radioactive waste disposed of in the disposal facility should be secured through safety assessment in consideration of the various situations. In this study, the influence and correlation of EDTA and ISA, which are the factors that can impede the safety of the disposal facility, were analyzed using the PHREEQC computational code. Thermodynamic database (TDB) of Andra, specific ion interaction theory (SIT) model as ionic strength correction model, radionuclides (Ni, Am, Pu) were adopted to perform the calculation on the distribution of chemical species by pH. According to the results, EDTA dominated the system and the effect of ISA is relatively small for the distribution of the chemical species of divalent and trivalent cations in neutral and weak base conditions. In the case of the tetravalent cations, the effect of ISA increased compared to the previous case especially in the strong base conditions. In conclusion, EDTA has a more significant effect on the system than ISA under the environment of the domestic disposal facility. Furthermore, when EDTA and ISA are present simultaneously in the system, the effects of two materials are inversely proportional and this characteristic should be considered during the safety assessment.

Near-surface disposal facility is more susceptible to intrusion than underground repository, resulting in more possible pathways for contaminant release. Alike human intrusion, animals (e.g. Ants, Moles, etc.) could intrude into the disposal site to excavate burrows, which could cause direct release of contaminants to biosphere. In this paper, animal intrusion is demonstrated using GoldSim’s commercial contaminant transport module and impact on the integrity of the near-surface disposal facility is evaluated in terms of fractional release rate of the contaminants. In this study, the near-surface disposal facility is modelled with a single concrete vault to contain radionuclide according to LLW concentration limit stated in NSSC notice No.2020-6. The release of contaminants is modelled to occur directly after the institutional control period, and the contaminants are mostly transported from the concrete vault to cover layers via diffusion. To produce mathematical model of the release of the contaminants due to animal intrusion, firstly, the fraction of burrow volume for each cover layer is calculated separately for each animal species, based on their maximum possible intrusion depth. In this study, fractions of burrow volume for ants and moles are calculated based on their maximum possible intrusion depths, where for ants is 2–3 m, and for moles is 0.1–0.135 m. Then, assuming that the contaminants are distributed homogeneously throughout each cover layers by diffusion, fraction of contaminants transported into the uppermost layer via excavation of the burrow is calculated for each layer based on burrow volume, and fraction of contaminants removed from the uppermost layer to the layers below via collapse of the burrow is also calculated based on the burrow volume. Lastly, the net transportation of contaminants into and out of the burrow via excavation and collapse, respectively, is calculated and demonstrated using direct transfer rate function of the GoldSim. Based on the simulated result, the maximum mass flux is too minor to cause a meaningful impact on the safety. The peak mass flux of the most sensitive radionuclide, I-129, is witnessed at around year 1,470, with a flux value of 5.36×10−6 g·yr−1. This minor release of the contaminants could be due to cover layers being much thicker than the maximum possible intrusion depth of the animals, preventing the animal intrusion into the deeper layers of higher radionuclide concentration. In future, this study can be used to provide a guidance and fundamental data for scenario development and safety evaluation of the near-surface disposal facility.

The permanent shutdown of Wolseong 1, PHWR (Pressurized Heavy Water Reactor) was decided. Accordingly, there is need for C-14 treatment technology to spent resin generated by PHWR in classified Medium Level Radioactive Waste by C-14 specific activity. However, spent resin by PHWR is mixed and stored with activated carbon and zeolite (mixture), not a single storage, and separation from the mixture must be carried out in advance for C-14 treatment in the spent resin. This study developed a C-14 treatment facility that combined with the technology of separating spent resin from spent resin mixture by PHWR NPP and the technology of C-14 treatment for disposal. The C-14 treatment facility consists of spent resin separation (Part 1) and treatment of separated spent resin. (Part 2) Part 1 is applied with a process of separating the mixed and stored spent resin from the spent resin mixture by applying a drum screen method. In the case of Part 2, spent resin treatment process for desorbing and collecting C-14 nuclides in the separated spent resin using microwave reactor was applied. Except for the adsorbent used to collect C-14 detached in the process of separating and treating spent resin, no additional material is introduced into the facility, and thus secondary waste is significantly reduced. In addition, pollution prevention banks at the bottom of the facility and a sealed automated circulation system were applied to prevent unexpected leakage and diffusion of radioactive materials and ensure stability of workers. Currently, the C-14 treatment facility has been verified for spent resin separation and spent resin treatment using simulated spent resin mixture, and the facility will be demonstrated and verified for field applicability. According to derived results, it is believed that it will be possible to apply the C-14 treatment facility when decommissioning of PHWR.

In nuclear power plants and nuclear facilities, radioactive waste containing hazardous substances (Mixed waste) is continuously generated due to research such as radiochemical study and nuclide analysis. In addition, radioactive waste including heavy metals and asbestos is generated during the dismantling process of nuclear power plants. Mixed wastes have both radiation hazards and chemical hazards, and there’s a possibility of synergistic effects generation. However, in most countries except the United States, there are no regulatory standards for the chemical hazards of mixed waste. The regulations applicable to mixed waste in Korea include the Nuclear Safety Act and the Waste Management Act. The Nuclear Safety Act prohibits the acceptance of hazardous radioactive waste in disposal facilities, but there is no definition or characteristic identification procedure for “hazardous.” The Waste Management Act also does not state the regulation for radioactive waste. In the Gyeongju disposal facility in Korea, the leachate in the disposal facility is expected to be a groundwater saturated with concrete and is expected to irradiated by radioactive waste. On the other hands, most of the non-radioactive waste landfill facilities are built on the surface, and the leachate is expected to be rainwater that reacts with the soil. Due to the differences in leaching environments, there’s a potential to overestimate or underestimate the leaching properties of hazardous substances if the standard leaching test is applied. To show for this, a leaching test simulating disposal facility’s environment were applied to sample waste containing heavy metals. The leaching solution was groundwater collected from the area near the Gyeongju disposal facility, which is then saturated with concrete and adjusted to pH 12.5. In addition, gamma-ray irradiation was conducted during the leaching test to observe changes in the leaching behavior of heavy metals in the actual radioactive waste disposal environment. As a result, lead showed significantly increased leaching compared to the standard test method, and cadmium was not detected in all experimental conditions except heavy irradiation. This study suggested that regulations on the hazardous of mixed waste should be settled, which should be established in sufficient consideration of the types and characteristics of substances contained in the waste.

Currently, the Gyeongju disposal facility is planned to be operated as a complex disposal facility with three types: cave disposal, surface disposal, and landfill disposal. Approximate method and arrangement have been decided up to the 1st, 2nd, and 3rd stage disposal facilities, but the optimal method for the arrangement of the entire complex disposal facility has not been established. When establishing the subsequent disposal facility arrangement plan, the generation prospect and disposal capacity setting plan for each level of radioactive waste was established, and the disposal capacity of the subsequent disposal facility, the disposal facility method, etc. were reviewed and reflected. Among the items for deriving an efficient management plan, KEPCO E&C is going to first derive a site arrangement plan for each disposal facility of 800,000 drums of radioactive waste, and has drawn up a plan for each scenario through collaboration with other organizations. When establishing the subsequent disposal facility arrangement plan, the generation prospect and disposal capacity setting plan for each level of radioactive waste was established, and the disposal capacity of the subsequent disposal facility, the disposal facility method, etc. were reviewed and reflected.

Currently, KHNP has 24 operating nuclear power plant units with a toal combined capacity of about 23 GWe and two units are under construction. However, permanent stop of Kori unit 1 nuclear power plant was decided in 2017. Accordingly, interest in how to dispose of waste stored inside a permanently stopped nuclear power plant and waste generated as decommissioning process is increasing. KHNP CRI is conducting research on the advancement of plasma torch melting facilities for waste treatment generated during the plant decommissioning and operation period. Plasma torch melting facility is composed of various equipment such as a melting furnace (Melting chamber, Pyrolsis chamber), a torch, an exhaust system facility, a waste supply device, and other equipment. In demonstration test, concrete waste was put in a 200 L drum to check whether it can be pyrolyzed using a plasma torch melting facility. Reproducibility for waste treatment in the form of a 200 L drum and discharge of molten slag could be confirmed, the amount of concrete waste in 200 L Drum that could be treated according to power of plasma torch was confirmed. This demonstration test confirmed the field applicability and stability of plasma torch melting facility, and improved expectations for long-term operation.

Recently, concern regarding disposal of cellulosic material is growing as cellulose is known to produce complexing agent, isosaccharinic acid (ISA), upon degradation. ISA could enhance mobility of some radionuclides, thus increasing the amount of radionuclide released into the environment. Evaluation on the possible impact of the cellulose degradation would be an important aspect in safety evaluation. In this paper, the maximum safe disposal amount cellulose is evaluated considering the disposal environment of silos of 1st phase disposal facility. The key factor governing the impact of cellulose degradation is pH of disposal environment, as cellulose is known to degrade partially at pH above 12.5, and completely at pH above 13. Thus, disposal environment should be analyzed as to determine the extent of degradation. As silos are constructed with large amount of cement, porewater within concrete walls would be of very high pH. However, for high pH porewater to be released into the pores of crushed rock, which is filling up the silos, lower pH groundwater (commonly pH 7) should flow into the silos through the concrete walls. This causes dilution of the high pH concrete porewater, resulting in a lower pH as the silos are filled, reaching to expected pH of 11.8–12.3, which is below cellulose degradation condition. Thus, cellulose degradation is not expected, but to quantitatively evaluate safe disposal amount of cellulose, partial degradation is assumed. Upon literature review, the most conservative ISA concentration, enhancing radionuclide mobility, is determined to be 1.0×10−4 M and to reach this concentration, cellulose mass equivalent to 6wt% of cement of the repository, is required to be degraded. However, this ratio is derived based on complete degradation of cellulose into ISA, so for partial degradation, degradation ratio and yield ratio of ISA should be considered. Commonly, cellulosic material (e.g. cotton, paper, etc.) has degree of polymerization (DP) between 1,000–2,000, and with this DP, degradation ratio is estimated to be about 10%. Furthermore, yield ratio of ISA is known to be 80%. Considering all these aspects, about 1.79×107 kg of cellulose could be disposed, which if converted into number of drums, considering cellulose content of dry active waste, more than 100,000 drums (200 L) could be disposed with negligible impact on safety. Based on the result, negligible impact of cellulose degradation is expected for safety of 1st phase disposal facility. In future, this study could be used as fundamental data for revising waste acceptance criteria.

The disposing method of the low-intermediate-level radioactive waste, near-surface disposal facilities are generally used. This disposal method refers to a method of constructing a concrete structure on the surface of the ground, putting radioactive waste in it, and covering it with an engineered barrier to isolate human life. Among these, engineered barriers mean covering multiple layers of heterogeneous materials such as sand, clay, and gravel. Engineering barriers have the purpose of delaying the release of radioactive materials into the natural environment as much as possible, and maintaining the isolation of radioactive waste and human life for as long as possible. In this study, the design and construction method of the facility to demonstrate the performance of the engineered barrier that isolates the surface disposal facility from nature was described. In addition, the design and construction method of monitoring technology that can monitor the safety of engineered barriers by measuring information such as moisture, temperature, and slope safety in real time was also explained.

In South Korea, the master plan for high-level radioactive waste management, announced in 2016, suggested the construction and operation of intermediate storage facilities on a permanent disposal site and specified the adoption of dry storage in consideration of the ease of operation and expansion. As of 2021, the government is again reviewing its overarching policy on the back-end fuel cycles, including intermediate storage and permanent disposal. In the case of dry storage facilities, safety evaluation is being conducted using a combination of deterministic and probabilistic approaches, similar to that of nuclear power plants. The two methods are complementary, of which Probabilistic Safety Assessment (PSA) has the advantage of being able to identify key scenarios affecting safety, but its use in storage facilities has not been highlighted so far. However, depending on the spent fuel management phases such as loading, transportation, and storage, it may be not enough to capture effective and efficient safety evaluation only deterministically, and probabilistic methods may contribute to the evaluation of long-term operation or external events such as an earthquake. There have already been cases where PSA has been performed on a part of the nuclear fuel cycle through previous studies. This paper created the safety assessment model based on open sources such as the released EPRI reports, by targeting arbitrary intermediate storage facilities. The model considered the scenarios for loading, transportation, and storage, with human error respectively. It will be able to be modified and improved to fit domestic and specific intermediate storage facilities in the future.

Since 2017, the Korea Institute of Nuclear Nonproliferation and Control (KINAC) has been implementing State System of Accounting for and Control of Nuclear Materials (SSAC) training courses for the nuclear Newcomer States. This IAEA SSAC course for Newcomer States aims to provide overall concepts and techniques, particularly on nuclear material accountancy and control systems, and address future challenges with regard to developing new nuclear power plants. Due to the restricted travels and limited in-person access to training and facilities from the COVID-19 pandemic, however, a new software was developed to substitute a technical tour on bulk handling facility (BHF) of the training course, and the course was favorably shifted to online in 2021. This newly built training software allows participants to follow each step of the technical process at a virtual bulk handing facility, and provides a video tour for such conditions where the software is found difficult to operate. Another feature of the development is a security function that prevents access of unauthorized users to the software. The achievement is expected to strengthen the SSAC of Newcomer States and ensure the practical implementation of safeguards from the initial stage of their novel nuclear power program through cooperation with IAEA. This contribution of the Republic of Korea (ROK) as one of the leading countries in the field of nuclear nonproliferation will further extend the partnership between IAEA and ROK and promote cooperation with the Newcomer States.

In this study, rainfall infiltration in vault of the second near-surface disposal facility was evaluated on the basis of various disposal scenarios. A total of four different disposal scenarios were examined based on the locations of the radioactive waste containers. A numerical model was developed using the FEFLOW software and finite element method to simulate the behavior of infiltrated water in each disposal scenario. The effects of the disposal scenarios on the infiltrated water were evaluated by estimating the flux of the infiltrated water at the vault interfaces. For 300 years, the flux of infiltrated water flowing into the vault was estimated to be 1 mm/year or less for all scenario. The overall results suggest that when the engineered barriers are intact, the flux of infiltrated water cannot generate a sufficient pressure head to penetrate the vault. In addition, it is confirmed that the disposal scenarios have insignificant effects on the infiltrated water flowing into the vault.

This paper is to study the technology of inspection and history management systems for wind power that are continuously increasing around the world. In the past, inspections and analysis of major devices in renewable energy system have been operated in an analog way that identifies problems through photography and passive method. To improve this problem, we conduct a study on VR-based inspection history management system using 3D texturing technique of drone image. The paper describes the current status and prospects of wind power, research and development of wind power blade inspection and history management systems, experiments and reviews in the field, and expected effects and future utilization of this technology. It is expected that the latest technology for inspection and management of renewable system will be secured and introduced to the site through the development research of this system to reduce maintenance costs and power generation costs.