이 연구는 2022 개정 교육과정에서 신설된 고등학교 ｢금융과 경제생활｣의 개발 절차, 기본적인 개발 방향과 주안점 등을 다루고 있다. 개발 과정에서 연구진 내부 회의는 물론 각론 조정팀, 현장 교사, 전문가의 검토 과정, 폭넓은 여론 수렴 등의 과정을 거쳤다. 주요 개발 방향은 학생들의 삶과 연계한 교육과정 구성, 성취기준 및 학습량 적정화, 안정된 금융 생활에 필요한 역량 강화로 잡았고, 개발의 주안점은 사회적 변화를 반영한 맞춤형 금융교 육, 범교과 학습 주제 및 국가·사회적 요구 사항 반영, 융합 선택 과목의 취지를 반영한 통합적 접근 및 중복성 배제였다.

Molten salt reactors have several advantages over conventional light water reactors. These include producing less nuclear waste, operating at higher power efficiency and inherent safety due to the low operating pressure. NaCl-MgCl2 eutectic salt is one of the candidates for the molten salt reactor coolant. However, because the salt is very hygroscopic, structural material corrosion occurs resulting in the high cost to maintain. To mitigate corrosion there have been many studies for the dehydration of the salt, especially focusing on the magnesium chloride. The reason is that the moisture adsorbed to the magnesium chloride undergoes hydrolysis over 200 degrees Celsius and decomposes to MgOHCl while the moisture associated with the NaCl is easily liberated during the heating procedure without chemical reaction. As the operating temperature of the molten salt is between 500 and 700 degrees Celsius, the MgOHCl is believed as the main cause for the structural corrosion. In this research, thermal dehydration of the salt with elemental Mg, for the NaCl-MgCl2 eutectic, was studied based on the previous dehydration methods and considering scalable and easy to handle. The MgOHCl was removed both through the thermal decomposition and the reduction by Mg metal. After the removal of MgOHCl, based on the difference between the freezing points and the density, the salt cooled down very slowly to ensure the separation between the purified salt and the disposals such as MgO and remaining Mg metals. The efficiency of the dehydration method was determined by the concentration of the MgOHCl. The concentration was determined by cyclic voltammetry and the result was compared with undehydrated salt and salt dehydrated thermally without the addition of Mg metal. To qualify and quantify the MgOHCl content through the cyclic voltammetry, it was necessary to observe the signal by adding MgOHCl to each sample. Based on the thermogravimetric analysis result of MgCl2· 6H2O, MgOHCl powder was formed through heating the MgCl2·6H2O.

For licensees who face the decommissioning project for the first time, even if they can utilize their experience in operation, they should be well prepared and assessed for the risks of dismantling activities reflecting the characteristics of decommissioning. This can be included in the risk management of the decommissioning project, but what we want to discuss in this study is the evaluation of the industrial risk of the actual work before the dismantling work is carried out. We would like to focus more on the review of dismantling activities subject to industrial risk assessment and a series of processes for risk assessment. The dismantling work plan will need to obtain approval from the supervisory department before work on the Systems, Structures, and Components (SSCs) can be carried out. At this time, risk assessment may be included among many safety-related required documents, which are divided into radiological and non-radiological risks. The target activities at Level 1 level can include preparation for dismantling and maintenance of facilities, dismantling big components, removing the contamination of concrete structures, managing radioactive waste, etc. In addition, it can be composed of preparation work, removal of connections, lifting/installation, cutting, radiation/radioactivity measurement, and withdrawal as detailed work stages of each item’s activities. For domestic nuclear decommissioning projects, two major performance organizations, licensees and contractors, must be considered. Regarding risk assessment, the licensee will have a supervisory department controlling decommissioning activities and an HSE department at the site, and a process will need to be established in consideration of the contractor’s work organization. Therefore, activities in the risk assessment process may be established. In this study, risk assessment was reviewed as safety-related matters to be considered when carrying out the dismantling work. Safety-related risk assessment is a necessary procedure for performing practical dismantling activities, and this should be considered well in advance. Therefore, work activities and criteria were established for risk assessment, and the performance process was assumed to apply them. In terms of the performance organization and the responsibilities and roles of the processes to be performed by each organization were constructed, and this can be referred to in the process of preparing for the decommissioning project.

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

Radioactive waste generated during nuclear power plant decommissioning is classified as radioactive waste before the concentration is identified, but more than 90% of the amount generated is at a level that can be by clearance. However, due to a problem in the analysis procedure, the analysis is not carried out at the place of on-site but is transported to an external institution to identify concentration, which implies a problem of human error because 100% manual. As a way to solve this problem, research is underway to develop a mobile radioactive waste nuclide analysis facility. The mobile radionuclide analysis facility consists of a preparation room, a sample storage room, a measurement room, a pretreatment room, and a waste storage room, and is connected to an external ventilation facility. In addition, since the automation module is built-in from the sample pre-threatening step to the separation step, safety can be improved and rapid analysis can be performed by being located in the decommissioning site. As an initial study for the introduction of a mobile nuclide analysis facility, Visiplan, a preliminary external exposure evaluation code, was used to derive the analysis workload by a single process and evaluate the exposure dose of workers. Based on this, as a follow-up study, the amount of analysis work according to the continuous process and the exposure dose of workers were evaluated. As a result of the evaluation, the Regulatory dose limit was satisfied, and in future studies, internal and external exposure doses were evaluated in consideration of the route of movement, and it is intended to be used as basic data in the field introduction process.

The domestic Nuclear Power Plant (NPP) decommissioning project is expected to be carried out sequentially, starting with Kori Unit 1. As a license holder, in order to smoothly operate a new decommissioning project, a process in terms of project management must be well established. Therefore, this study will discuss what factors should be considered in establishing the process of decommissioning NPPs. Various standards have been proposed as project management tools on how to express the business process in writing and in what aspects to describe it. Representatively, PMBOK, ISO 21500, and PRICE 2 may be considered. It will be necessary to consider IAEA safety standards in the nuclear decommissioning project. GSR part 6 and part 2 can be considered as two major requirements. GSR part 6 presents a total of 15 requirements, including decommissioning plans, general safety requirements until execution and termination. GSR part 2 presents basic principles for securing the safety of nuclear facilities, and there are a total of 14 requirements. Domestic regulatory guidelines should be considered, and there will be largely laws and regulations related to the decommissioning of nuclear facilities, guidelines for regulatory agencies, and guidelines and regulations related to HSE. The Nuclear Safety Act, Enforcement Decree, Enforcement Rules, and NSSC should be considered in the applicable law for nuclear facilities. Since the construction and operation process has been established for domestic decommissioning project, there will be parts where existing procedures must be applied in terms of life cycle management of facilities and the same performance entity. As a management areas classification in the construction and operation stage, it seems that a classification similar to Level 1 and Level 2 should be applied to the decommissioning project. This study analyzed the factors to be considered in the management system in preparing for the first decommissioning project in Korea. Since it is project management, it is necessary to establish a system by referring to international standards, and it is suggested that domestic regulatory reflection, existing business procedures, and domestic business conditions should be considered.

A large amount of small and medium-sized metal waste is generated during the decommissioning of nuclear power plants (NPPs). Metal waste is mostly contaminated with low-level radioactive, so it needs decontamination for self-disposal and recycling. A large amount of Organic Decontamination Liquid Waste during decontamination will be generated. The generated organic liquid waste is low in concentration, so the decomposition efficiency is low in the decomposition process. A conditioning process is necessary to concentrate at a high concentration. For effective treatment for Organic Decontamination Liquid Waste, the composition of organic liquid waste and conditioning process were analyzed. Organic acids, metal ions, radioactive nuclides, surfactants, etc. are present in the Organic Decontamination Liquid Waste, and suspended solids are sometimes generated by various reactions. According to previous studies, the concentration of organic acids including surfactants obtained results from several tens of ppm to a maximum of 1,000 ppm, so the maximum value of 1,000 ppm was assumed. For the composition and total amount of metal ions, the average value (52.7wt% Fe, 16.3wt% Ni, 15.1wt% Cr, 15.9wt% Mn) of the distribution of metal species removed by the actual decontamination process is applied, and the total amount is 1,000 ppm was assumed. As for the radionuclides, only 60Co and 137Cs, which are expected to be mainly present, were considered, and 60Co was assumed to be 2,000 Bq/g and 137Cs to be 360 Bq/g by referring to the literature. The amounts of suspended solids were assumed to be 500 ppm by referring to the characteristics of the liquid waste generated in the decontamination process of the NPPs. Based on the estimated value, a reaction formula was established and a simulated Organic Decontamination Liquid Waste was prepared. As a result of measurement using an analysis device, the composition of the estimated and simulated Organic Decontamination Liquid Waste had similar values. The conditioning and treatment process largely consists of pretreatment, conditioning, decomposition processes. Organic Decontamination Liquid Waste goes through a pretreatment process to remove impurities with large particles. In the conditioning process, treated water that has passed through the UF/RO membrane system is discharged into the environment. At this time, Concentrated water goes through a decomposition process for processing the Organic Decontamination Liquid Waste, and is discharged to the environment through a secondary RO membrane system. The conditioning process is the low-concentration Organic Decontamination Liquid Waste in the UF membrane system is forming a micelles in an RO membrane system, concentrating it to a high concentration and then go through a recirculation process in the UF membrane system. An experiment was conducted to confirm whether the concentration of surfactants occurred during the conditioning process. As a result of the experiment confirmed that the highly concentrated surfactant formed micelles and was filtered out in the UF membrane system.

The spent filters used to purify radioactive materials and remove impurities from primary systems at nuclear power plants (NPPs) have been stored for long periods in filter storage rooms at NPPs due to concerns about the unproven safety of the treatment method, absence of disposal facilities, and risk of high radiation exposure. In the storage room at Kori Unit 1, there are approximately 227 spent filters of 9 different types. The radiation dose rates of filters range from 0.01 to 500 mSv/hr. Recently, a comprehensive plan has been established for the treatment and disposal of radioactive waste that has not yet been treated to facilitate decommissioning of NPPs. As a follow-up measure, compression and packaging optimization processes are being developed to treat the spent filters. KHNP plans to dispose of the spent filters after compressing, packaging, and immobilizing them. However, the spent filters are currently stored without being sorted by type or radiation intensity. If the removal and packing of the filters are done randomly without a plan for the order of withdrawal and subsequent processes, issues may arise such as a decrease in drum loading efficiency and exceeding the dose limit of the package. In this study, the number of drums needed to pack the spent filters was calculated, considering the filter size, weight, quantity, dose rate, shielding thickness of drum, and loadable quantity in a shielding drum (SD). Then, the spent filters that can be loaded on each drum were classified into one group. In addition, the withdrawal order for each group was set so that the filter withdrawal, compression, and packaging processes could be performed efficiently. The spent filter groups are as follows: (1) compression/12 cm SD (17 groups), (2) compression/16 cm SD (6 groups), (3) non-compression/ intermediate storage container (17 groups, additional radiation attenuation required due to high dose rate), and (4) unclassified (5 groups, determined after measurement due to lack of filter information). The withdrawal order of the groups was determined based on several factors, including visual identification of the filter, ease of distribution after withdrawal, work convenience, and safety. Due to the decay of radioactivity over time, the current dose rate of the spent filters is expected to be much lower than at the time of waste generation. Therefore, in the future, sample filters will be taken from the storage room to measure their radioactivity and radiation dose rate. Based on these measurements, a database of radiological characteristics for the 227 filters will be created and used to revise the filter grouping.

Thermal treatment, such as combustion, is the most effective way to solve the spatial problem of radioactive waste disposal. Existing incineration technology has the problem of discharging harmful pollutants (CO2 and dioxin, etc.) into the environment. Therefore, it was evaluate the validity of the thermal treatment process that can reduce the volume of dry active waste (DAW) in an eco-friendly. In addition, the stability of the alternative incineration process under development was evaluated by evaluating the emission of harmful pollutants to the environment during the thermal treatment process. We selected 14 samples identical to those discarded by each nuclear power plant (Kori, Saeul, Wolsong, Hanbit, Hanul). And EA (Elemental Analysis) analysis was performed on each sample. As a result, excluded samples containing wastes containing POPs (Persistent Organic Pollutants) such as PCBs (Polychlorinated Biphenyls), which could generate harmful pollutants during thermal treatment, and halogenated organic wastes such as PVC (Polyvinyl Chloride). In addition, the thermal treatment conditions for the four DAWs were derived by Thermogravimetric Analysis/Differential Thermal Analysis (TG/DTA) analysis. At this time, Py-GC/MS analysis was performed at the temperature at which each waste causes thermal decomposition (cotton is 437°C, paper is 562°C, latex glove is 430°C, plastic bag is 485°C). As a result of analyzing the exhaust gas produced during thermal decomposition, about 77.0% of the cotton was Benzoic acid series, the paper was 41.1% Glucopyranose series, and 15.8% hydroxy acetaldehyde. Latex glove was identified to be 45.9% and 19.2% for Limonene and 2-methyl-1, 3-Butadiene, and for plastic bags, Octacosanol and 2-octyl-1-Dodecanol were 38.8% and 15.2%. In addition, it was confirmed that dioxin and harmful heavy metals, which are discussed as environmental risks, were not detected in all samples.

The acceptance criteria for low and intermediate level radioactive waste disposal facilities in Korea to regulate that homogeneous waste, such as concentrated waste and spent resin, should be solidified. In addition, solidification requirements such as compressive strength and leaching test must be satisfied for the solidified radioactive waste solidified sample. It is necessary to develop technologies such as the development of a solidification process for radioactive waste to be solidified and the characteristics of a solidification support. Radioactive waste solidification methods include cement solidification, geopolymer solidification, and vitrification. In general, low-temperature solidification methods such as cement solidification and geopolymer solidification have the advantage of being inexpensive and having simple process equipment. As a high-temperature solidification method, there is typically a vitrification. Glass solidification is generally widely used as a stabilization method for liquid high-level waste, and when applied to low- and intermediate-level radioactive waste, the volume reduction effect due to melting of combustible waste can be obtained. In this study, the advantages and disadvantages of the solidification process technology for radioactive waste and the criteria for accepting the solidified material from domestic and foreign disposal facilities were analyzed.

The rise of nuclear power plants to meet escalating global energy needs has made environmental pollution including the contamination of uranium due to improper disposal of radioactive wastewater during uranium milling and mining processes. Adsorption, a water purification method known for its fast kinetics, high selectivity, and ease of use, has emerged as a popular choice for the treatment of radioactive wastewater. In response to the critical need for the purification of radioactive wastewater contaminated with U(VI), this review provides a comprehensive summary of the various types of materials, synthetic methods, and adsorption mechanisms used for the purification process. The materials are categorized into four main groups: organic, inorganic, composite/nanomaterials, and framework materials. To enhance the adsorption capacity for U(VI), researchers have explored physical and chemical modifications as well as the development of organic-inorganic hybrids. The improved adsorption performance resulting from these modifications is mostly attributed to electrostatic interaction, surface complexation, and ion exchange mechanisms. However, despite the present understanding of the processes involved, further research is still needed to fully determine the optimal approach for purifying contaminated radioactive wastewater.

The most important thing in development of a process-based TSPA (Total System Performance Assessment) tool for large-scale disposal systems (like APro) is to use efficient numerical analysis methods for the large-scale problems. When analyzing the borehole in which the most diverse physical phenomena occur in connection with each other, the finest mesh in the system is applied to increase the analysis accuracy. Since thousands of such boreholes would be placed in the future disposal system, the numerical analysis for the system becomes significantly slower, or even impossible due to the memory problem in cases. In this study, we propose a tractable approach, so called global-local iterative analysis method, to solve the large-scale process-based TSPA problem numerically. The global-local iterative analysis method goes through the following process: 1) By applying a coarse mesh to the borehole area the size of the problem of global domain (entire disposal system) is reduced and the numerical analysis is performed for the global domain. 2) Solutions in previous step are used as a boundary condition of the problem of local domain (a unit space containing one borehole and little part of rock), the fine mesh is applied to the borehole area, and the numerical analysis is performed for each local domain. 3) Solutions in previous step are used as boundary conditions of boreholes in the problem of global domain and the numerical analysis is performed for the global domain. 4) steps 2) and 3) are repeated. The solution derived by the global-local iterative analysis method is expected to be closer to the solution derived by the numerical analysis of the global problem applying the fine mesh to boreholes. In addition, since local problems become independent problems the parallel computing can be introduced to increase calculation efficiency. This study analyzes the numerical error of the globallocal iterative analysis method and evaluates the number of iterations in which the solution satisfies the convergence criteria. And increasing computational efficiency from the parallel computing using HPC system is also analyzed.

In the engineered barrier system of deep geological disposal repository, complex physicochemical phenomena occur throughout the entire disposal time, consequently impacting the safety function. The bentonite buffer, a significant component of the engineered barrier system, can be geochemically altered due to the changes in host rock groundwater, temperature, and redox condition. Such changes may have direct or indirect effects on radionuclide migration in case of canister failure. Therefore, a modeling tool that accounts for coupled thermal-hydraulic-mechanical-chemical (THMC) processes is necessary for the safety assessment. To this end, the Korea Atomic Energy Research Institute (KAERI) has developed the APro, a modeling interface for conducting safety assessment of deep geological disposal repository. The APro considers coupled THMC processes that influence radionuclide migration. Here, the solute transport considering thermal and hydraulic processes are calculated using the COMSOL multi-physics, while geochemical reactions are carried out in PHREEQC. The two software are coupled using a sequential non-iterative operator splitting approach, and transport of non-water H, non-water O, and charge were additionally considered to enhance the coupling model stability. Finally, the applicability of APro to simulate long-term geochemical evolution of bentonite was demonstrated through benchmark studies to evaluate the effects of mineral precipitation/dissolution, temperature, redox, and seawater intrusion.

Korea Atomic Energy Research Institute is developing a radionuclide management processes as a conditioning technology to reduce the burden of spent fuel disposal. The radionuclide management process refers to a process managing radionuclides with similar properties by introducing various technology options that can separate and recover radionuclides from spent fuels. In particular, it is a process aimed at increasing disposal efficiency by managing high-heat, high-mobility, and high-toxic radionuclides that can greatly affect the performance of the disposal system. Since the radionuclide management process seeks to consider various technology options for each unit process, it may have several process flows rather than have a single process flow. Describing the various process flows as a single flow network model is called the superstructure model. In this study, we intend to develop a superstructure model for the radionuclide management process and use it as a model to select the optimal process flow. To find the optimal process flow, an objective function must be defined, and at the fuel cycle system level multiple objectives such as effectiveness (disposal area), safety (explosure dose), and economics (cost) can be considered. Before performing the system-level optimization, it is necessary to select candidates of process flow in consideration of waste properties and process efficiency at the process level. In this study, a sensitivity analysis is conducted to analyze changes in waste properties such as decay heat and radioactivity when the separation ratio varies due to the performance change for each unit process of the radionuclide management process. Through this analysis, it is possible to derive a performance range that can have waste properties suitable for following waste treatment, especially waste form manufacturing. It is also possible to analyze the effect of waste properties that vary according to the performance change on waste storage and management approaches.

In case of damaged spent fuels, it would require additional treatment for their transportation and storage to capture the radioactive fission products in a defined space. The canning container for the damaged spent fuels is one way to seal the radioactive fission products inside the container. In the Post Irradiation Examination Facility (PIEF) of KAERI, the Quiver container has been introduced for canning damaged spent fuels from Westinghouse Sweden. The main container body has been manufactured for particle-tightness of spent fuel. In addition, drying equipment is being prepared for gas-tightness of spent fuel. The drying equipment can remove water and fill the inert gas inside the container. Before drying inside the container, we evaluated the volatile fission products inventory because volatile fission products could be released during the drying process. Despite assuming highly conservative hypotheses for the inventory remaining in damaged fuel rods, the amount that could be released during the drying process was less and dose rate levels around the evacuation piping system were low.

It has been investigated on the management of the nuclides in KAERI. Strontium-90 is a high heatgenerating nuclide in spent nuclear fuel. It is needed to separate the salt from the salt solution for the recovery of strontium after the chlorination of the strontium oxide in molten salt. A vacuum distillation technology was used for the separation of strontium from the molten salt. It was investigated on operating conditions of reactive distillation process for the recovery of the strontium from the salt solution. At a reduced pressure, considerable amount of the carbonation agents such as K2CO3 and Li2CO3 were reduced during heating in the distiller due to the thermal decomposition. Therefore, the two step process was proposed, which is composed of a reaction step at an atmospheric pressure and a salt distillation step at a reduced pressure. In the reaction step, the condition of low temperature and high pressure is suitable to suppress the decomposition of the carbonation agent. In the salt distillation step, reduced pressure is preferable at a suitable temperature depending on the evaporation rate of the salt.

As the use of nuclear energy has been expanded, issues in a spent nuclear fuel management are raised. Several methods have been proposed and developed to manage spent fuels safely and efficiently. One method is to reduce environmental burden in disposal of spent fuels by decreasing volume of high-level waste. A nuclides management process (NMP) is one example. Through this novel process, it is able to separate highly mobile nuclides (ex. iodine, krypton), high thermal emission nuclides (ex. strontium, barium), and optionally, uranium from spent fuels. Since the NMP is a back-end fuel cycle technology, a reliable safeguards system should be employed in the facility. As international atomic energy agency (IAEA) recommends safeguards-by-design (SBD), it is desirable to investigate an appropriate safeguards approach at a step of technology development. Process monitoring (PM) is a complemental safeguards technology for traditional safeguards technologies which based on mass balance. PM traces nuclear materials indirectly but consecutively by using process parameters such as temperature, pressure, and flow of fluid. These parameters are obtainable by installing appropriate sensors. In a respect of SBD, PM is a promising approach to achieve the safeguards goal, the timely detection of diversion of a nuclear material. However, it is necessary to classify useful process parameters from all available signals which provided from PM in order to properly utilize PM. In this study, we investigated application methods of the PM approach to NMP. NMP consists of several unit processes in series. Firstly, we inspected a principle and a feature of each unit process. Based on the results, we evaluated applicability of the PM approach to each unit process according to effectiveness in enhancing safeguardability. Several unit processes were expected that their safeguards are able to be enhanced by using certain process parameters from PM.

Since the National R&D Innovation Act was enacted in 2022, it became a crucial issue how to qualify or improve R&D activities and disseminate their outcomes. Many organizations have referred to various quality management standards such as the American National Standards Institute/American Society for Quality (ANSI/ASQ) Z1.13, International Organization for Standardization (ISO) 9001, and the American Society of Mechanical Engineers Nuclear Quality Assurance-1 (ASME NQA-1), as a means to set up their own quality system. ISO is the international standard for implementing a quality management system (QMS), which provides a framework and principles for managing an organization’s QMS, with the aim of ensuring that the organization consistently provide products or services that meet regulatory requirements. ISO 9001 can cover all aspects of an organization’s operations, and it can also be expanded to include R&D areas. The introduction of ISO 9001 to R&D aims to improve R&D practices and establish a standardized process framework for conducting R&D. ANSI/ASQ Z1.13 provides quality guidelines for research and consists of 10 sections covering various aspects of research quality, emphasizing ethical conduct, clear objectives, reliable data collection, and analysis. ASME NQA-1 is one of quality assurance standards for nuclear facility applications, but it has been extended and applied to R&D activities in the nuclear fields. It just focuses on planning, procedures, documentation, competence, equipment, and material control. KINAC has conducted extensive research on verifying and regulating nuclear activities while providing support for national nonproliferation technologies and policies. In addition to the quantitative growth achieved so far, efforts are being made to establish a qualitative and integrated management system. As a first step to achieve this goal, this study reviewed international standards and methodologies for research quality and derived the key components for R&D quality management. Moreover, the appropriate outline of quality management system framework was proposed for R&D as a regulatory support process, based on the ISO 9001. The implementation of quality management standards and procedures for R&D in KINAC, which could lead to improved research practices, more reliable data collection and analysis and increased efficiency in conducting R&D activities.