본 연구는 개가 나타내는 공격 행동 원인을 살펴보고, 이로부터 피해를 예방하는 대처 방안을 살펴보는데 목적을 가지고 있다. 이를 위해 2010년부터 2023년까지 국내·외 연구 및 출판 도서 24편을 바탕으로 개의 공격 행동의 원인이 되는 내적·외적·심리적 요인과 대처방안을 제시하기 위해 문헌고찰을 수행하였다. 연구 결과 첫째, 개의 영양상태와 심 리적 안정을 포함한 내인적 요인이 공격 행동에 영향을 미치는 요인임이 나타났다. 둘째, 견주의 책임감 있는 양육태도와 질병 및 스트레스로부터의 보호, 적절한 훈련과 사회화 과정이 필요함을 확인하였다. 셋째, 개에 대한 이해와 적절한 접근에 대한 인식 통해 공격 행동으로 인한 사고를 예방할 수 있음을 확인하였다. 나아가 개의 이해와 견주의 역할, 지역사회와 시스템 차원의 접근을 통해 근본적인 대책을 수립할 수 있다. 본 연구는 개의 공격 행동의 원인과 대처방안에 대한 후속 연구의 기초자료를 제공하고, 실효성 있는 정 책의 방향 제안에 근거를 제공하였다는 점에서 의의를 가진다.

In today’s world, carbon-based materials research is much wider wherein, it requires a lot of processing techniques to manufacture or synthesize. Moreover, the processing methods through which the carbon-based materials are derived from synthetic sources are of high cost. Processing of such hierarchical porous carbon materials (PCMs) was slightly complex and only very few methods render carbon nano-materials (CNMs) with high specific surface area. Once it is processed, which paves a path to versatile applications. CNMs derived from biological sources are widespread and their application spectrum is also very wide. This review focuses on biomass-derived CNMs from various plant sources for its versatile applications. The major thrust areas of energy storage include batteries, super-capacitors, and fuel cells which are described in this article. Meanwhile, the challenges faced during the processing of biomass-derived CNMs and their future prospects are also discussed comprehensively.

Nanofillers, by virtue of their minute size, when incorporated inside a matrix, have the capability to enhance the physical parameters of the complex matrix. Graphene, the wonder material of the twenty-first century, has established itself in the field of nanofillers. However, it still has yet to find its way into the mining industry. This review paper focuses on a novel way of attaining sustainability in mining methodology using graphene as a nanofiller. The implementations can be subdivided into three categories based on their impact—economic, environment, and safety. To achieve economic welfare in mine methodology; Graphene is used to enhance the productivity of machinery. Electric-Heavy Earth Moving Machinery using LiFePO4/Graphene hybrid cathode battery is not only an ideal replacement to fossil-powered vehicles considering the contribution of environmental strain but also a more-efficient model than Electric-Heavy Earth Moving Machinery using conventional Lithium Ion Phosphate Battery batteries. Heavy Earth Moving Machinery having tires of Styrene-Butadiene Rubber/Graphene composite would have better efficacy and longer life cycle than the conventional ones. Graphene derivative Magnetic Graphene Oxide is used to achieve environmental welfare by its implication as an additive in the effluent treatment plant for its capability of removing heavy metal ions and negative-strain bacteria from the mine water. To improve the safety standards of the mine workers, graphene and its derivatives have environmental implications to constitute a safer surrounding concerning precarious situations due to the unpredictable behavior of geomaterials. Graphene can assist in constituting a more economical and reliable slope model as incorporating graphene induces restructuration and improvement in strength parameters. This enables a miner to extract more minerals in tranquility from the resources as there is an increase in compaction and shear strength. A combination of a graphene sheet and auxetic graphene foam can be placed over the blast holes to not only restrict the trajectory of the fly rocks but also attenuate some part of the explosive energy. The objective of this coagulation is to upgrade the traditional practice by replacing the conventional products, and the effect is observed in the form of achieving sustainability in the mine.

The US NRC developed a program called NRCDose3 to evaluates the environmental impact of radiation around nuclear facilities. The NRCDose3 code is a software suite that integrates the functionality of three individual LADTAP II, GASPAR II, and XOQDOQ Fortran codes that were developed by the NRC in the 1980’s and have been in use by the nuclear industry and the NRC staff for assessments of liquid effluent and gaseous effluent, and meteorological transport and dispersion, respectively. Through the integrated program, it is possible to conduct safety assessment and environmental impact assessment from liquid and gaseous effluent when operating permits are granted. In addition to a more user-friendly graphic user interface (GUI) for inputting data, significant changes have been made to the data management and operation to support expanded capabilities. The basic calculation methods of the LADTAP II, GASPAR II, and XOQDOQ have not been changed with this update to the NRCDose3 code. Several features have been added. The previous program used only ICRP-2 dose conversion factor, but the new program can additionally use dose conversion factor of ICRP-30 and ICRP-72. In the previous program, 4 age groups (infant, child, teen, and adult) were evaluated during dose evaluation, but when ICRP-72 was selected, 6 age groups (infant, 1-year, 5-year, 10-year, 15-year, and adult) could be evaluated. In addition, when selecting ICRP-72, many user-modifiable parameters such as food intake and exposure time were added. It will be referred to E-DOSE60, a program currently under development.

After the Fukushima nuclear power plant accident in 2011, interest in technology for evaluating residents’ exposure to effluents generated from nuclear power plants at the time of the accident has increased. KHNP has developed the S-REDAP program and is using it to evaluate radiation dose and recommend resident protection measures in the event of a nuclear power plant emergency. Its main functions are source term evaluation, atmospheric diffusion evaluation, radiation dose evaluation, etc. Based on these evaluations, resident protection measures are evaluated. In Japan, evaluation is conducted through a program called SPEEDI-MP (System for Prediction of Environmental Emergency Dose Information Multi-model Package) created by JAEA (Japan Atomic Energy Agency). Similar to S-REDAP, the program also evaluates effluents emitted from nuclear facilities through source term evaluation and atmospheric diffusion factor evaluation. In JAEA, through a program using SPEEDI-MP, the source term evaluation was performed in collaboration with NSC (Nuclear Safety Commission) in the event of the Fukushima nuclear plant accident, and dose evaluation in Japan was performed 2 months as an atmospheric diffusion factor using meteorological data for 2 days. Through comparative analysis of evaluation data from Japan, improvements to the current program be derived.

In 2022 and 2023, the Korea Institute of Nuclear Safety (KINS), a regulatory body, revised the regulatory guidelines for off-site dose evaluation to residents, marine characteristics surveys around nuclear facilities, and environmental radiation surveys and evaluation around nuclear facilities. In addition, the NRC, a US regulatory body, has revised regulatory guide 1.21 (MEASURING, EVALUATING, AND REPORTING RADIOACTIVE MATERIAL IN LIQUID AND GASEOUS EFFLUENTS AND SOLID WASTE) to change environmental programs for nuclear facilities. The domestic regulatory guidelines were revised and added to reflect the experience of site dose evaluation for multiple units during the operation license review of nuclear facilities, the resident exposure dose age group was modified to conform to ICRP-72, and the environmental monitoring plan was clarified. In the case of the US, the recommended guidelines for updating the long-term average atmospheric diffusion factor and deposition factor, the clarification of the I-131 environmental monitoring guidelines for drinking water, and the clarification of the procedures described in the technical guidelines when changing environmental programs have been revised and added. Through such regulatory trend review, it is necessary to preemptively respond to changes in the regulatory environment in the future.

The Derived Concentration Guideline Level (DCGL) using RESRAD code is generally obtained for the reuse of the site and remaining buildings of the decommissioning of nuclear facilities. At this time, the evaluation first considers wide DCGL assuming homogenous contamination for the entire target site. The DCGL derived through this will be compared with the actual contamination measured at the Final Status Survey (FSS) stage to determine whether the site is compliance with criteria. Guidelines for Survey units are presented in MARSSIM and suggested in Class 1 through 3. Therefore, DCGL for the survey unit of a certain smaller area is established by applying a correction factor from wide DCGL, which is define as an Area Factor (AF). Therefore, this study reviewed the AF applied in overseas cases, reviewed the necessary factors for derivation, and compared them by applying factors to the preliminary experimental target area for domestic nuclear installations. The AF is the ratio of the dose from the base-case contaminated area to the dose from a smaller contaminated area with the same radioactive concentration. To this end, an unrestricted resident farmer scenario was applied as the site reuse scenario, which deals with all exposure pathways considered in the RESRAD. The potential exposure pathways considered in resident farmer scenarios are largely divided into external and internal exposures, which are based on NUREG/CR-5512. In addition, in order to calculate the AF, a change in the contaminated area occurs, and accordingly, a variable that varies according to the area, i.e., length parallel to aquifer flow (LCZPAQ), the contaminated fraction of plant food ingested (FPLANT), the contaminated fraction of meat and milk (FMEAT and FMILK), is accompanied. As the contamination area decreases, these variables decrease, and the criteria for reduction were reflected through overseas cases. In this study, three nuclides (C-14, Co-60, and Cs-137) were assumed as representative nuclides, and the area of the contaminated site was selected as 50,000 m2 and reduced at a certain rate. As a result, each nuclide showed different characteristics, but in general, AF increases as the area decreases. Compared to the area of this study, AF values were calculated to be smaller than those of overseas cases, but it was confirmed that the area of the values showed similar patterns. In addition, in the case of C-14, the slope of AF increased rapidly as the area decreased, while Co-60 and Cs-137 showed similar slopes.

The decommissioning of the Nuclear Power Plant (NPP) is a long-term project of more than 15 years and will be carried out as a project, which will require project management skills accordingly. The risk of decommissioning project is a combination of many factors such as the decommissioning plan, the matters licensed by the regulatory agency, the design and implementation of dismantling, the dismantling plan and organization, and stakeholders. There will be some difficulties in risk management because key assumptions about many factors and the contents of major risks should be well considered. Risk management typically performs a series of processes ranging from identification and analysis to evaluation. In order to analyze and evaluate risks here, identification of potential risks is the first step, and in order to reasonably select potential risks, various factors mentioned should be considered. Therefore, the purpose of this study is to identify possible risks that should be considered for the decommissioning project in various aspects. The risk of the decommissioning project can be defined using the hazard keyword, and the risk family presented in the IAEA safety series can also be referred. It would be better to approach the radiological or non-radiological risks that may occur in the dismantling work with the hazard keyword, and if the characteristics of the decommissioning project are reflected, it would be a good idea to approach it on a risk family basis. There are 10 top risks in the risk family, 25 risks at the level 2 and 61 risks at the level 3 are presented. It may be complex to consider these hazards and risks recommended as risk families at the same time, so using the results of safety evaluation as input data for risk identification can be a reasonable approach. Therefore, this study intended to derive the possible risks of the decommissioning project based on the risk family structure. At this point, the reflection of the safety assessment results was intended to be materialized by considering the hazards checklist. As a result, this study defined and example of 38 possible risks for the decommissioning project, considering the 10 top risk family and lower level risk categories. This result is not finalized, and it will be necessary to further strengthened through expert workshops or HAZOP in the future.

At Nuclear Power Plant (NPP), aging management is performed as part of the Periodic Safety Review (PSR) in accordance with the Nuclear Safety Act. The purpose of the aging management program (AMP) is to manage the integrity of structures, systems and components (SSCs) in NPPs over time and use. Through this, aging deterioration is mitigated to increase equipment life and secure long-term operation safety. Fuel Oil Chemistry is one of the AMPs. Through this program, aging management is performed for storage tanks, piping and other metal components that contact with diesel fuel oil. The program is focused on managing loss of material due to general, pitting, crevice, and microbiologically-influenced corrosion (MIC) and fouling that leads to corrosion of the diesel fuel tank internal surfaces. The fuel oil aging management method currently applied to NPPs in Korea measures the concentration of water and particulate contamination in the oil, analyzed the trend, and periodically cleans and inspect the inside of tanks. Among them, in monitoring MIC, a direct analysis and monitoring of the amount of microorganisms may be more effective. In this study, a method for improving the MIC monitoring system for diesel fuel oil systems was reviewed by reviewing reference documents including NUREG 1801 and examining the methods actually applied in US NPPs.

Around the world, Nuclear Power Plants (NPPs) have been operated since the 1950s and are used as a major power source. In Korea, Kori unit 1 stared commercial operation for the first time in 1978, and as of 2023, 25 units of NPPs are in operation. NPPs produce electricity for about 40 to 60 years after receiving an operating license, and after securing safety through a safety evaluation, the operating period is extended. NPPs that operate for a long time are systematically evaluated for safety at regular intervals through Periodic Safety Review (PSR) recommended by the IAEA. In Korea, PSR has been introduced and performed since 2000. This study reviewed the process of the PSR by comparing with the international PSR procedure. The PSR process is established through the IAEA SSG-25 document and proceeds in the order of establishment of basis document - individual factor evaluation - global assessment - integrated improvement plan. In Korea, PSR is carried out in a similar process, but there are some differences from the IAEA’s procedure. The safety factor review is conducted under the agreement of basis document between the licensee and the regulatory body, but the prior agreement procedure with the regulatory body is not reflected in Korea. As a result, if the licensee and the regulatory body have different opinions on the current licensing basis and the modern safety standards after the evaluation is performed, a difference may occur in the review results and safety enhancement items, which may lead to inefficient PSR progress. PSR is conducted for the continuous safe operation and management of NPPs, and it is important to refer to overseas standards and cases. Although procedures, guidelines, and regulatory requirements are in place in Korea, continuous review and improvement are required. It is necessary to improve procedures such as basis document and global assessment in order to more efficiently carry out PSR evaluation by regulatory agency and licensee’s safety enhancement actions of domestic NPPs

As an initial part of Kori-1 & Wolsung-1 Unit decommissioning planning, a characterization plan is developed to define the nature, extent and location of contaminants, determine sampling locations and protocols, determine quality assurance objectives for characterization, and define documentation requirements. The actual characterization of a facility is an iterative process that involves initial sampling according to the characterization plan, field management (such as labeling, packaging, storing, and transport) of the samples, laboratory analysis, conformance to the data quality objectives (DQOs), and then identifying any additional sampling required, refining the DQOs, and modifying the characterization plan accordingly. The final product of the facility characterization is a document that describes the type, amount, and location of contaminants that will require consideration and removal during the decommissioning operations sufficient to prepare a decommissioning plan. In this study, implementing a characterization plan, developed in accordance with this standard, will result in obtaining or deriving the above information.

Kori Nuclear Power Plant Unit 1, which began operating in 1978, is Korea’s oldest commercial nuclear reactor. The reactor was permanently shut down in June 2017, and now the decommissioning process has begun. The decommissioning process will generate a significant amount of waste that requires appropriate management to minimize the impact on the environment and human health. And the waste routing, i.e. the activities and logistics for managing the material generated, is a key point in a decommissioning project. It determines the routes from the material inventory to the envisaged material end states. In this study, we review on several factors for the selection of the waste routes in a decommissioning project. In terms of sustainability, the ‘waste hierarchy’ should be applied to routing materials from nuclear facilities. According to the waste hierarchy, the preferred end state is reuse or recycling of the waste as material or, more preferably, the avoidance of waste generation. In addition, treatments (such as decontamination and thermal treatment) that can reduce the volumes requiring disposal as radioactive waste should be considered. Another important parameter is the need to secure availability and capacity of waste routes. Short-term bottlenecks or any delay in the removal of the waste from the site often has an impact on other site activities. If possible, at least two alternative waste routes should be identified for the main categories of waste and kept available throughout the decommissioning project. All routes should be direct to the material end state if possible, but it is more important that waste is removed from the site so that other site operations are not impeded.

To develop technology for extracting energy resources from seawater, we first investigated the research experiences of domestic experts. The survey items included the types of adsorbents that can adsorb dissolved resources in seawater, the subjects of experiments, and the scope of research. We divided the types of adsorbents into organic and inorganic categories and compared their adsorption performance. We also examined how adsorption experiments were conducted using simulated solutions and confirmed whether there were any experiences of conducting experiments in actual seawater. A total of 14 domestic research papers on extracting dissolved resources from seawater were reviewed, excluding topics such as removing dissolved resources from seawater and seawater desalination. This review provides an understanding of domestic research trends and will be helpful in setting directions for future research and development.

In order to start decommissioning domestic nuclear facilities, the Final Decommissioning Plan (FDP) must be prepared and approved by the regulatory agency. The contents of domestic FDP consist of 12 chapters, and there is the decommissioning feasibility design that should be described in Chapter 5 as contents to be considered from the construction stage of nuclear facilities. The design of decommissioning feasibility for nuclear facilities seems to be largely divided into three items. In summary, there ae minimization of contaminations to facilities and the environment, easy of dismantling, and minimization of the radioactive waste generation. In addition, the design characteristics to which the ALARA principle is applied in terms of optimizing the exposure dose of workers and residents may also correspond to the decommissioning feasibility design. The design characteristics for decommissioning feasibility during the period leading up to the design, operation, and decommissioning of nuclear facilities can be listed as the main points as follows. Minimization of facility contamination will include contents related to the leakage of systems and components, minimization of effluents to the environment will involve gaseous and liquid effluents from systems and components to the environment, easy of dismantling will involves history and inspection records during operation, and minimization of radioactive waste generation can be the contents related to the radioactive waste management plans. The design characteristics of facilities and equipment to meet the ALARA principles can be listed as follows. It means taking into account the benefits and costs of the design improvement plan, and the elimination of unnecessary radiation exposure can be maintained at the exposure dose ALARA, which is in line with the decommissioning feasibility design. Among the requirements of licensing documents for decommissioning domestic nuclear facilities is the decommissioning feasibility design. This item relates to the design characteristics for decommissioning considered in the construction stage of the facility and should present the effectiveness of measures for them until operation and decommissioning. In this study, the regulatory requirements presented in the construction and operation stage and the contents presented in the U.S. case were reviewed, and it is hoped that it will be used as reference for the preparation of FDP.

The type of radioactive waste that may occur in the process of nuclear power plant dismantling can be classified into solid, liquid, gas, and mixed waste. The amount of these wastes must be defined in the Final Decommissioning Plan for approval of the licensing. Also, in the case of liquid radioactive waste, it is necessary to calculate the generation amount in order to treat radioactive waste at a Radioactive Waste Treatment Facility (RWTF) or on-site. In this regard, there is no Code and Standard for the amount of liquid radioactive waste generated during NPP are dismantled, but ANSI/NS-55.6 describes the amount of liquid radioactive waste generated from a light water reactor type NPP. This code is applied to nuclear power-related facilities such as domestic NPP and radioactive waste disposal facility. Therefore, this review intends to suggest an application plan for domestic NPP decommissioning through codes for liquid radioactive waste expected to generate during nuclear power plant decommissioning.

The type of radioactive waste that may occur in the process of NPP dismantling can be classified into solid, liquid, gas, and mixed waste. Most of the radioactive waste generated during the dismantling of a NPP is metal solid waste, but liquid radioactive waste is also a very important factor in terms of radiation environmental impact assessment. In the case of liquid radioactive waste, it is necessary to calculate the generation amount in order to design liquid radioactive waste processing system of Radioactive Waste Treatment Facility (RWTF). Depending on the amount of liquid radioactive waste generated, the type of liquid radioactive waste processing system included in the RWTF is different. In addition, in order to apply to the domestic RWTF, it is important to secure the site area occupied by the each system, the liquid radioactive waste treatment capacity of the system, and how to secure circulating water used for dilution and discharge of liquid radioactive waste. Therefore, this review aims to suggest an optimal method for the treatment system for liquid radioactive waste included in RWTF of Wolseong.

Wolsong Unit 1 is about 679 MW Pressurized Heavy Water Reactor (PHWR). Canada AECL was responsible for Nuclear Steam Supply System (NSSS) design and supply. Wolsong Unit 1 was operated from 1983 to 2019. Currently, Wolsong Unit 1 is under safety management after permanent shutdown. Wolsong unit 1, a heavy water reactor, has the following characteristics. • Unlike Light Water Reactor, vertical reactors, Heavy Water Reactor is installed horizontally. • The internal structure of the reactor is more complex than that of a light water reactor (380 pressure tubes in reactor as called Calandria) • The Calandria Vault, a large concrete structure filled with light water, is located outside of Calandria In the case of the decommissioning plan of PHWR in Canada, they have adopted a deferred decommissioning strategy that decommissioning begins after permanent shutdown and long-term safety management (30 to 40 years). So, Decommissioning of PHWR in Canada is expected to start in the 2050s. Nuclear Safety Act stipulates that if a commercial nuclear power plant is permanently suspended, the utility must submit a Final Decommissioning Plan (FDP) within 5 years. So, KHNP, the utility, is developing the FDP for Wolsong Unit 1 and have a plan to submit it to the government by the end of 2024. And then licensing review is expected to take at least two years. The key milestone for decommissioning project has a plan to start decommissioning in 2027 and complete it by 2034, but this is flexible depending of the government’s approval for decommissioning and the completion of prerequisites such as spent fuel transfer, etc. KHNP has prepared a strategy and system consisting of three areas such as R&D, Engineering and licensing document development to prepare the final decommissioning plan for Wolsong Unit 1. The promotion system for the preparation of the FDP for Wolsong Unit 1 is consisted of Engineering (HAS Characterization, Process/Work Package/Cost Estimation, Dismantling Safety Evaluation, Radiological Environmental Report, Radioactive Waste Treatment and Facility Construction), R&D (COG cooperation, KHNP R&D Results), Kori unit 1 lessons learned, etc. KHNP have the plan that the FDP Draft development by the end of 2023, reflecting engineering services results, R&D results, COG technical cooperation results and lessoned learned on Kori Unit 1. After collecting opinions from residents through a public hearing, the FDP will be submitted to the government by the end of 2024. It is expected that there will be many difficulties in the development process as it is the world’s first FDP development for the commercial Pressurized Heavy Water Reactors.

Kori Unit 1 is about 600MW Pressurized Light Water Reactor as WH type. KHNP got an approval for construction and operation of Kori unit 1 on May 31, 1972 and started commercial operation from Apr. 29, 1987. And then, it was decided to permanently suspend it on Jun. 18, 2017 after 40 years of commercial operation. The Nuclear Safety Act stipulates that if a commercial nuclear power plant is permanently suspended, the utility must submit a Final Decommissioning Plan (FDP) within 5 years. So, KHNP, the utility, developed a FDP for Kori Unit 1 and submitted it to the government in May 2021. In South Korea, the FDP is to be prepared in accordance with the relevant notices and consists of 11 major chapters such as (1) Decommissioning Plan Overview, (2) Project management, (3) Status of Site and Environmental, (4) Decommissioning Strategies and Method, (5) Ease of Decom. Design characteristic, (6) Safety Analysis, (7) Radiation Protection, (8) Decontamination and Dismantling, (9) Radioactive Waste Management, (10) Environmental Impact Analysis, (11) Fire Protection and (12, 13) Etc., References and Glossary. KHNP has prepared a strategy and system consisting of three areas such as R&D, Engineering and licensing document development to prepare the final decommissioning plan for Kori Unit 1. The promotion system for the preparation of the FDP for Kori Unit 1 is composed of Engineering (HAS Characterization, Dismantling Safety Evaluation, Radiological Environmental Report, Radioactive Waste Treatment and Facility Construction), R&D(KHNP R&D Results such as Process/Work Package /Cost Estimation, Safety Analysis, Contamination and Exposure, Guide for Detailed Characteristic, Site Safety Analysis, RV & RVI Dismantling Process, etc.), Overseas case lessons learned(Taiwan unit 1 NPP FDP and Spain Zorita NPP FDP analysis) and Development of Licensing Document. KHNP completed the initial completion of the Final Decommissioning Plan for Kori Unit 1 in March 2020 and carried out collecting residents’ opinions through public hearings. And then, KHNP supplemented the results of the residents’ opinions and applied for license to the Nuclear Safety and Security Commission in May, 2021. Now, KHNP are responding to the FDP licensing review.

As the importance of radioactive waste management has emerged, quality assurance management of radioactive waste has been legally mandated and the Korea Radioactive Waste Agency (KORAD) established the “Waste Acceptance Criteria for the 1st Phase Disposal Facility of the Wolsong Lowand Intermediate-Level Waste Disposal Center (WAC)”, the detailed guideline for radioactive waste acceptance. Accordingly, the Korea Atomic Energy Research Institute (KAERI) introduced a radioactive waste quality assurance management system and developed detailed procedures for performing the waste packaging and characterization methods suggested in the WAC. In this study, we reviewed the radioactive waste characterization method established by the KAERI to meet the WAC presented by the KORAD. In the WAC, the characterization items for the disposal of radioactive waste were divided into six major categories (general requirements, solidification and immobilization requirements, radiological, physical, chemical, and biological requirements), and each subcategories are shown in detail under the major classification. In order to satisfy the characterization criteria for each detailed item, KAERI divided the procedure into a characterization item performed during the packaging process of radioactive waste, a separate test item, and a characterization item performed after the packaging was completed. Based on the KAERI’s radioactive waste packaging procedure, the procedure for characterization of the above items is summarized as follows. First, during the radioactive waste packaging process, the characterization corresponding to the general requirements (waste type) is performed, such as checking the classification status of the contents and checking whether there are substances unsuitable for disposal, etc. Also, characterization corresponding to the physical requirements is performed by checking the void fraction in waste package and visual confirmation of particulate matter, substances containg free water, ect. In addition, chemical and biological requirements can be characterized by visually confirming that no hazardous chemicals (explosive, flammable, gaseous substances, perishables, infectious substances, etc.) are included during the packaging process, and by taking pictures at each packaging steps. Items for characterization using separate test samples include radiological, physical, and chemical requirements. The detailed items include identification of radionuclide and radioactivity concentration, particulate matter identification test, free water and chelate content measurement tests, etc. Characterization items performing after the packaging is completed include general requirements such as measuring the weight and height of packages and radiological requirements such as measurements of surface dose rate and contamination, etc. All of the above procedures are proceduralized and managed in the radioactive waste quality assurance procedure, and a report including the characterization results is prepared and submitted when requesting acceptance of radioactive waste. The characterization of KAERI’s radioactive waste has been systematically established and progressed under the quality assurance system. In the future, we plan to supplement various items that require further improvement, and through this, we can expect to improve the reliability of radioactive waste management and activate the final disposal of KAERI’s radioactive waste.

Licensing for the application of the Polymer Concrete High Integrity Container (PC-HIC) to nuclear power plants has been completed or is in progress. Approval for the expanded application to all domestic nuclear power plants has been completed to utilize the 860 L PC-HICs for the 2nd stage surface repository, and the regulatory body is reviewing the license application to use the 510 L PCHICs for the 1st stage underground repository in the representative nuclear power plants. The 860 L PC-HICs, which have been licensed for all domestic nuclear power plants, will be used for safe storage management and disposal of low-dose dried concentrate waste and spent resin, and a total of 100 units is expected to be supplied to representative nuclear power plants that have been licensed first. The 510 L PC-HICs are planned to be used for underground disposal of high-dose spent resin and dried concentrate waste. Prior to the application of PC-HICs to nuclear power plants and disposal to the repository, it is necessary to establish realistic and reasonable requirements through close consultations between waste generator and disposal operators to ensure the suitability for disposal of PC-HIC packages and to carry out disposal delivery and acceptance work. Since the Polymer Concrete High Integrity Container (PC-HIC) has long-term integrity of more than 300 years and the barrier does not temporarily collapse, spent resin and dried concentrate waste, which are radioactive wastes to be solidified, can be disposed of much more safely in PC-HIC packages than solidified types. Acceptance criteria for the PC-HIC packages should be prepared fully reflecting the advantages of PC-HIC, and quality assurance methods for physical/chemical/radiological characterization results based on the Waste Certification Program (WCP) should be supported. In addition, infrastructure should be secured for safe transportation, handling, and storage of the PC-HIC packages. In this paper, we have tried to find a reasonable acceptance criteria, quality assurance method, and infrastructure level according to the dose and disposal conditions of PC-HIC packages.