This study aims to classify R&D activities related to the nuclear fuel cycle using the deep learning methodology. First, R&D data of the Republic of Korea were collected from the National Science & Technology Information Service (NTIS) for the years 2021, 2022, and 2023. We use keywords such as ‘nuclear,’ ‘uranium,’ ‘plutonium,’ and ‘thorium’ to find nuclear-related R&D projects in the NTIS database. Among the numerous R&D projects found through keyword searches, overlapping and medical-related R&D projects were excluded. Finally, 495 R&D projects conducted in 2021, 430 R&D projects conducted in 2022, and 296 R&D projects conducted in 2023 were obtained for analysis. After that, Safeguards experts determine whether the R&D projects are subject to declaration under the AP. The values of the content validity index (CVI) and content validity ratio (CVR) were used to verify whether the experts’ judgments were valid. The 1,218 collected and labeled data were then divided 8:2 into training and test datasets to see if deep learning could be applied to classify nuclear fuel cycle-related R&D activities. We use the Python and TensorFlow packages, including RNN, GRU, and CNN methods. First, the collected text information was preprocessed to remove punctuation marks and then tokenized to make it suitable for deep learning. After 20 epochs of training to classify the nuclear fuel cycle-related R&D activities, the RNN model achieved 97.30% accuracy and a 5.85% error rate on the validation dataset. The GRU model achieved 96.53% accuracy and a 9.06% error rate on the validation dataset. In comparison, the CNN model achieved 94.61% accuracy and a 2.57% error rate on the validation dataset. When applying the test dataset to each model, the RNN model had a test accuracy of 83.20%, the GRU test accuracy of 82.79%, and the CNN model had a test accuracy of 85.66% for the same dataset. This study applied deep learning models to labeled data judged by various experts, and the CNN model showed the best results. In the future, this study will continue to develop an optimum deep learning model that can classify nuclear fuel cycle-related R&D activities to achieve the purpose of safeguards measures from open-source data such as papers and articles.

A seal is one of the primary means of safeguards along with surveillance. The International Atomic Energy Agency (IAEA) uses various types of seals to verify the diversion of nuclear materials and is developing new seals according to the development of technology. Independent of the IAEA, ROK uses national safeguards seals for state-level regulation. A national safeguards inspector binds the nuclear material storage by combining a seal with a metal wire and checks the serial number of the RFID chip inserted in the seal with a reader. The Wolsong spent fuel dry storage facility has 14 modules, each with 24 seals, and thus a maximum of 336 national seals will be installed. Although dependent on the sealing method, it takes about 5 minutes to verify one seal. As such, a considerable workforce is required for verification, and both the IAEA and the ROK are currently conducting random inspections. In addition, there are cases where verification is impossible because old seals are damaged due to harsh environments and long exchange periods. Therefore, in this study, we analyzed cases in areas where sealing technology has been developed to improve the problems of the existing national safeguards seals. And we proposed a method for improving national seals by finding requirements of seals considering spent fuel dry storage facility characteristics. In international logistics, sealing is essential in product transport verification, terrorism prevention, and tariff imposition. Accordingly, the field of container sealing has been extensively developed, and the International Organization for Standardization (ISO) has regulated the mechanical requirements of the seal as ISO 17712 and the electronic requirements as ISO 18185. Mechanical seals include metal and plastic seals and metal seals include bolt seals, ball seals, and cable seals. In addition, there are various electronic seals, such as radio frequency identification (RFID), near field communication (NFC), infrared (IR). Recently, there has been a trend to use active seals that have a built-in battery and can implement various additional functions. Among the various seals, the main requirements for selecting seals suitable for dry storage facilities are as follows. First, use of a sealing tube longer than 10m should be possible. Second, it should have corrosion resistance so that it can be used for more than five years in the coastal area. Third, it must be a passive seal without a power supply. Fourth, it should not be overly costly. Finally, the seal verification time should be short. As a seal that satisfies these requirements, an electronic seal with application of the passive RFID method to the mechanical form of a metal cable seal is suitable. Since it is not an active seal, it is difficult to determine the time of breakage. Therefore, designing the seal such that the RFID is also damaged when the metal seal is broken will be helpful for verification. In this study, the requirements for national safeguards seals in dry storage facilities were defined, and measures to improve the existing national seals were studied. Field applicability will be evaluated through future sealing device design and demonstration tests.

Proton exchange membrane fuel cells (PEMFCs) are an auspicious energy conversion technology with the potential to address rising energy demands while reducing greenhouse gas emissions. The stack’s performance, durability, and economy scale are greatly influenced by the materials used for the PEMFC, viz., the membrane electrocatalyst assembly (MEA) and bipolar flow plates (BPPs). Despite extensive study, carbon-based materials have outstanding physicochemical, electrical, and structural attributes crucial to stack performance, making them an excellent choice for PEMFC manufacturers. Carbon materials substantially impact the cost, performance, and durability of PEMFCs since they are prevalently sought for and widely employed in the construction of BPPs and gas diffusion layers (GDLs)) and in electrocatalysts as a support material. Consequently, it is essential to assemble a review that centers on utilizing such material potential, focusing on its research development, applications, problems, and future possibilities. The prime focus of this assessment is to offer a clear understanding of the potential roles of carbon and its allotropes in PEMFC applications. Consequently, this article comprehensively evaluates the applicability, functionality, recent advancements, and ambiguous concerns associated with carbonbased materials in PEMFCs.

The domestic shipbuilding industry is building high-value-added ships such as LNG and LPG, and the demand for natural gas, a clean energy source, is continuously increasing. Climate change, such as global warming, is occurring due to rising oil prices and excessive use of fossil fuels. To protect their homes from the changing environment, 121 countries announced intensive climate target policies to reduce carbon emissions to 0% by 2050. In this study, modeling and design were performed using SUS410 and SUS304L about the operating part of the Pilot valve based on the physical properties of the aluminum alloy used in the Pilot valve, a component of the gas pressure Regulating valve for LNG ships. Numerical We want to develop the optimal Pilot valve by comparing and analyzing the results using ANSYS, an analysis simulation program.

This study examined the efficacy of various chlorinating agents in partitioning light water reactor spent fuel, with the aim of optimizing the chlorination process. Through thermodynamic equilibrium calculations, we assessed the outcomes of employing MgCl2, NH4Cl, and Cl2 as chlorinating agents. A comparison was drawn between using a single agent and a sequential approach involving all three agents (MgCl2, NH4Cl, and Cl2). Following heat treatment, the utilization of MgCl2 as the sole chlorinating agent resulted in a moderate separation. Specifically, this method yielded a solid separation with 96.9% mass retention, 31.7% radioactivity, and 44.2% decay heat, relative to the initial spent fuel. In contrast, the sequential application of the chlorinating agents following heat treatment led to a final solid separation characterized by 93.1% mass retention, 5.1% radioactivity, and 15.4% decay heat, relative to the original spent fuel. The findings underscore the potential effectiveness of a sequential chlorination strategy for partitioning spent fuel. This approach holds promise as a standalone technique or as a complementary process alongside other partitioning processes such as pyroprocessing. Overall, our findings contribute to the advancement of spent fuel management strategies.

The thermal integrity of spent nuclear fuels has to be maintained during their long-term dry storage. The detailed temperature distributions of spent fuel assemblies are essential for evaluating the integrity of their dry storage systems. In this study, a subchannel analysis model was developed for a canister of a single fuel assembly using the COBRA-SFS code. The thermal parameters affecting the peak cladding temperature (PCT) of the spent fuel assembly were identified, and sensitivity analyses were performed based on these parameters. The subchannel analysis results indicated the presence of a recirculation flow, based on natural convection, between the fuel assembly and downcomer region. The sensitivity analysis of the thermal parameters indicated that the PCT was affected by the emissivity of the fuel cladding and basket, convective heat transfer coefficient, and thermal conductivity of the fluid. However, the effects of the wall friction factor of the canister, form loss coefficient of the grid spacers, and thermal conductivities of the solid materials, on the PCT were predominantly ignored.

수소는 다양한 신재생에너지 중 환경친화적인 에너지로 각광받고 있지만 농업에 적용된 사례는 드물다. 본 연구는 수소 연료전지 삼중 열병합 시스템을 온실에 적용하여 에너지를 절 약하고 온실가스를 줄이고자 한다. 이 시스템은 배출된 열을 회수하면서 수소로부터 난방, 냉각 및 전기를 생산할 수 있다. 수소 연료 전지 삼중 열 병합 시스템을 온실에 적용하기 위해 서는 온실의 냉난방 부하 분석이 필요하다. 이를 위해서는 온 실의 형태, 냉난방 시스템, 작물 등을 고려해야 한다. 따라서 본 연구에서는 건물 에너지 시뮬레이션(BES)을 활용하여 냉 난방 부하를 추정하고자 한다. 전주지역의 토마토를 재배하 는 반밀폐형 온실을 대상으로 2012년부터 2021년까지의 기 상데이터를 수집하여 분석했다. 온실 설계도를 참고하여 피 복재와 골조를 모델화하여 작물 에너지와 토양 에너지 교환을 실시했다. 건물 에너지 시뮬레이션의 유효성을 검증하기 위 해 작물의 유무에 의한 분석, 정적 에너지 및 동적 에너지 분석 을 실시했다. 또한 월별 최대 냉난방 부하 분석에 의해 평균 최 대 난방 용량 449,578kJ·h-1, 냉방 용량 431,187kJ·h-1이 산정 되었다.

To ensure radiological safety margin in the transport and storage of spent nuclear fuel, it is crucial to perform source term and shielding analyses in advance from the perspective of conservation. When performing source term analysis on UO2 fuel, which is mostly used in commercial nuclear power plants, uranium and oxygen are basically considered to be the initial materials of the new fuel. However, the presence of impurities in the fuel and structural materials of the fuel assembly may influence the source term and shielding analyses. The impurities could be radioactive materials or the stable materials that are activated by irradiation during reactor power operation. As measuring the impurity concentration levels in the fuel and structural materials can be challenging, publicly available information on impurity concentration levels is used as a reference in this evaluation. To assess the effect of impurities, the results of the source term and shielding analyses were compared depending on whether the assumed impurity concentration is considered. For the shielding analysis, generic cask design data developed by KEPCO-E&C was utilized.

바이오디젤은 중립연료로써 친환경 연료로 알려져 있으며, 육상에서는 일정 비율을 의무 혼합하는 정책을 시행하고 있다. 본 연구에서는 바이오디젤의 선박 연료로써의 사용 가능성을 검증하기 위해 선박용 경유와 바이오디젤의 혼합비율 0 %, 5 %, 10 %, 20 %에 대해 성분 분석, 금속 부식성 실험, 저장 안정성 실험을 수행하였다. 성분 분석은 ISO 8217:2017 기준에 따라 밀도, 동점도, 인화점 등 총 8가지를 평가하였으며, 180일 동안 상온과 가혹 조건(60 ℃)에서 금속 부식성 실험과 저장 안정성 실험을 통해 바이오디젤 신뢰성을 검증 하였다. 연구 결과, 성분 분석은 바이오디젤 모든 혼합비율에서 ISO 8217:2017 기준을 만족하였으며, 바이오디젤 비율에 따라 동점도, 밀 도, 산값은 혼합비율이 높아질수록 높게 나타났으며, 황분은 혼합비율이 높아질수록 낮게 나타났다. 금속 부식성은 탄소강, 철, 알루미늄, 니켈의 경우 부식이 거의 발생하지 않았으나, 구리의 경우 60 ℃ 환경 바이오디젤 20 % 혼합에서 산소가 풍부한 바이오디젤의 영향으로 부식이 발생하였다. 저장 안정성은 모든 바이오디젤 혼합비율을 180일 동안 상온과 가혹 조건에서 저장한 결과, 변색, 슬러지 발생, 연료 분리가 육안으로 확인되지 않았다.

In order to investigate the optimum condition of the autofrettage process for the diesel engine fuel injection pipe, different values of autofrettage pressure, pressure rising time, pressure holding time, and repetition of autofrettage process were applied. Autofrettage was performed by applying the hydrostatic internal pressures of 604MPa, 535MPa, 500MPa on the fuel injection pipe, corresponding to theoretical 50%, 30%, and 20% overstrain levels, respectively. The autofrettage residual stresses in the injection pipe were experimentally determined by using X-ray diffractometer. As the overstrain level increased, the magnitude of compressive residual stress at the bore increased. It was found that the rising time to reach the autofrettage pressure, holding time at the autofrettage pressure, and repeated application of the autofrettage pressure on the pipe had no significant influence on the residual stress distributions.

본 연구에서는 실리카 복합막 기반 고분자 전해질막을 5단 연료전지 스택에 적용하여 성능 평가를 수행하였다. 이를 통하여, 개별 구성 요소의 성능도 중요하지만, 전체적인 관점에서 공급되는 연료의 유량이 스택 성능에 중요한 역할을 하며, 특히 수소의 유량에 크게 의존한다는 사실이 확인하였다. 산소의 유량을 증가시켜도 성능의 변화는 미미한 반면, 수소 의 유량을 증가시키면 성능이 향상되는 것을 확인하였다. 그러나 수소의 유량 증가는 수소와 산소 유량 비율의 불균형을 초 래하여 장기적으로는 스택 성능과 내구성을 저하시키는 문제가 관찰되었다. 이러한 현상을 스택 구성 요소 및 개별 단위 셀 에서도 관찰할 수 있었으며, 따라서 스택 운전 시 각 구성 요소의 성능을 최적화하는 것 외에도 균일한 유량 제어를 위해 유 로 설계 및 운전 조건을 최적화하는 것이 중요하다는 것을 알 수 있었다. 마지막으로 실리카 복합막은 최대 출력 기준 25 W 이상의 성능을 나타내어 실제 연료전지 시스템에 적용하기에 충분한 성능을 갖춘 것으로 판단된다.

In this study, a bipolar visible light responsive photocatalytic fuel cell (PFC) was constructed by loading a Z-scheme g-C3N4/ carbon black/BiOBr and a Ti3C2/ MoS2 Schottky heterojunction on the carbon brush to prepare the photoanode and photocathode, respectively. It greatly improved the electron transfer and achieved efficient degradation of organic pollutants such as antibiotics and dyes simultaneously in two chambers of the PFC system. The Z-scheme g-C3N4/carbon black/BiOBr formed by adding highly conductive carbon black to g-C3N4/BiOBr not only effectively separates the photogenerated carriers, but also simultaneously retains the high reduction of the conduction band of g-C3N4 and the high oxidation of the valence band of BiOBr, improving the photocatalytic performance. The exceptional performance of Ti3C2/ MoS2 Schottky heterojunction originated from the superior electrical conductivity of Ti3C2 MXene, which facilitated the separation of photogenerated electron–hole pairs. Meanwhile, the synergistic effect of the two photoelectrodes further improved the photocatalytic performance of the PFC system, with degradation rates of 90.9% and 99.9% for 50 mg L− 1 tetracycline hydrochloride (TCH) and 50 mg L− 1 rhodamine-B (RhB), respectively, within 180 min. In addition, it was found that the PFC also exhibited excellent pollutant degradation rates under dark conditions (79.7%, TCH and 97.9%, RhB). This novel pollutant degradation system is expected to provide a new idea for efficient degradation of multiple pollutant simultaneously even in the dark.

Noble metal precipitates are fission products that can be found as metallic alloys in used nuclear fuel. They do not exist homogenously inside the fuel pellets, but exists in grain boundaries in the form of immiscible particles. The first drawback that comes because they exist in grain boundaries is the degradation of mechanical integrity. The particles in the grain boundaries can be considered as defect n solid solution of uranium oxide pellets, and they can change the lattice volume. Therefore, it is known that it can cause stress corrosion cracking of fuel pellets. Furthermore, there is a negative effect from the perspective of used fuel management. However, they also have a positive effect on used fuel management. Since the noble metal has galvanic reduction effect, the particles serve as an oxidation inhibitor for uranium. There are many other effects regarding to the noble metal precipitates. However, in any case, quantifying the particles is important in order to quantitatively analyze these effects from the perspective of used fuel management. SCALE/TRITON code was applied to calculate the noble metal isotopes including Mo, Tc, Ru, Rh and Pd. In order to calculate the distribution inside the pin, the multiregion cell model was selected. In particular, a cylindrical geometry was used, and the pellet was divided into several layers. In addition, coolant and cladding surrounded the pellet. Finally, the radial distribution was evaluated using the computational code, along with neutron flux map.

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.

During the operation of the nuclear power plant, various radioactive waste are generated. The spent resin, boron concentrates, and DAW are classified as a generic radioactive waste. They are treated and stored at radioactive waste building. In the reactor vessel, different types of radioactive waste are generated. Since the materials used in reactor core region exposed to high concentration of neutrons, they exhibit higher level of surface dose rate and specific activity. And they are usually stored in spent fuel pool with spent fuel. Various non-fuel radioactive wastes are stored in spent fuel pool, which are skeleton, control rod assembly, burnable neutron absorber, neutron source, in core detector, etc. The skeleton is composed of stainless 304 and Inconel-718. There are two types of control rod assembly, that are WH type and OPR type. The WH type control rod is composed of Ag-In-Cd composites. The OPR type control rod is composed of B4C and Inconel-625. In this paper, the characteristics and storage status of the non-fuel radioactive waste will be reported. Also, the management strategy for the various non-fuel radioactive waste will be discussed.

Most of the spent nuclear fuel generated by domestic nuclear power plants (NPPs) is temporarily stored in wet storage which is spent fuel pool (SFP) at each site. Currently, in case of Kori Unit 2, about 93.6% of spent nuclear fuel is stored in SFP. Without clear disposal policy determined for spent nuclear fuel, the storage capacity in each nuclear power plant is expected to reach saturation within 2030. Currently, the SFP stores not only spent fuel but also various non-fuel assembly (NFA). NFA apply to all device and structures except for fuel rods inserted in nuclear fuel assembly. The representative NFA is control element driving mechanism (CEDM), in-core instrument (ICI), burnable poison, and neutral resources. Although these components are irradiated in the reactor, they do not emit high-temperature heat and high radiation like nuclear fuel, so if they are classified as intermediate level waste (ILW) and low level waste (LLW) and moved outside the SFP, positive effects such as securing spent fuel storage space and delaying saturation points can be obtained. Therefore, this study analyzes the status of spent fuel and Non Fuel Assembly (NFA) storage in SFP of domestic nuclear power plants. In addition, this study predict the amount of spent fuel and NFA that occur in the future. For example, this study predicts the percentage of current and future ICIs and control rods in the SFP when stored in the spent fuel storage rack. In addition, the positive effects of moving NFA outside the SFP is analyzed. In addition, NFA withdrawn from SFP is classified as ILW & LLW according to the classification criteria, and the treatment, storage, and disposal methods of NFA will be considered. The study on the treatment, storage, and disposal methods of NFA is planned to be conducted by applying the existing KN-12 & KN-18 containers and ILW & LLW containers being developed for decommissioning waste.

The treatment of waste generated during operation as a part of preparation for decommissioning is coming to the fore as a pending issue. Non-fuel waste stored in the spent fuel pool (SFP) of PWRs in Korea includes Dummy fuel, damaged fuel rod storage container, reactor vessel specimen cask, spent in-core instrumentation, spent control element assemblies, spent neutron source assemblies, burnable poison rods, etc. In order to treat such waste, it is necessary to classify radioactive waste level and analyze kinds of nuclide in accordance with legal requirements. In order to solve the problem, the items that KHNP-CRI is trying to conduct like followings. First, KHNP-CRI will identify the current status of non-fuel waste stored in the SFP of all domestic nuclear power plants. In order to consider the treatment of non-fuel waste, it is essential to know what kind of items and how many items are stored in the SFP. Second, to identify the dimension and characteristics of non-fuel waste stored in the SFP would be conducted. The configuration of non-fuel waste is important information to handle them. Third, the way to handle non-fuel waste would be deduced including analysis of their dimension, whether the equipment should be developed to handle each kind of non-fuel waste or not, how to transport them. In order to classify radioactive waste level and analyze the nuclide for the non-fuel waste, handling tools and the cask to transport them into the facility which nuclide analysis is able to be performed would be required. Fourth, the nuclide analysis technology would be identified. Also, domestic holding technology would be identified and which technology should be developed to classify the radioactive waste level for the non-fuel waste would be deduced. This preliminary study will provide KHNP-CRI with the insight for the nuclide analysis technology and future work which is following action for the non-fuel waste. Based on the result of above preliminary study, the feasibility of the research for the treatment of non-fuel waste would be evaluated and research plan would be established. In conclusion, the treatment of non-fuel waste stored in the spent fuel pool of domestic PWR should be considered to prepare the decommissioning. KHNP-CRI will identify the quantity, the dimension and kinds of non-fuel waste in the SFP of domestic PWR. Also, the various nuclide analysis technology would be identified and the technology which should be developed would be defined through this preliminary study.

Since the first operation of the Gori No. 1 nuclear power plant in Korea was started to operate in 1978, currently 24 nuclear power plants have been being operated, out of which 21 plants are PWR types and the rest are CANDU types. About 30% of total electricity consumed in Korea is from all these nuclear power plants. The accumulated spent nuclear fuels (SNFs) generated from each site are temporarily being stored as wet or dry storage type at each plant site. These SNFs with their high radiotoxicity, heat generating, and long-lived radioactivity are currently the only type of high-level radioactive waste (HLW) in Korea, which urgently requires to be disposed of in deep geological repository. Studies on disposal of HLW in various kind of geological repositories have been carried out in such countries as Sweden, Finland, United States, and etc. with their own management policies in consideration of their situations. In Korea long-term R&D research program for safe management of SNF has also been conducted during last couple of decades since around 1997, during which several various type of disposal concepts for disposal of SNFs in deep geological formations have been investigated and developed. The first concept developed was KAERI Reference Disposal System (KRS) which is actually very much similar to Swedish KBS-3, a famous concept of direct disposal of SNF in stable crystalline rock at a depth of around 500 m which has been regarded as one of the most plausible method worldwide to direct disposal of SNF. The world first Finnish repository will be also this type. Since the characteristics of SNF discharged from domestic nuclear reactors have been changed and improved, and burnup has sometimes increased, a more advanced deep geological repository system has been needed, KRS-HB (KRS with High Burnup SNF) has been developed and in consideration of the dimensions of SNFs and the cooling period at the time point of the disposal time, KRS+, a rather improved disposal concept has also been subsequently developed which is especially focused on the efficient disposal area. Recently research has concentrated on rather advanced disposal technology focused on a safer and more economical repository system in recent view of the rapidly growing amount of accumulated SNF. Especially in Korea the rock mass and the footprint area for the repository extremely limited for disposal site. Some preliminary studies to achieve rather higher efficiency repository concept for disposal of SNF recently have already been emphasized. Among many possible ones for consideration of design for high-efficiency repository system, a double-layered system has been focused which is expected to maximize disposal capacity within the minimum footprint disposal area. Based on such disposal strategy a rather newly designed performance assessment methodology might be required to show long-term safety of the repository. Through the study some prerequisites for such methodological development will be roughly checked and investigated, which covers FEP identification and pathway and scenario analyses as well as preliminary conceptual modeling for the nuclide release and transport in near-field, far-field, and even biosphere in and around the conceptual repository system.

The deep geologic repository (DGR) concept is widely accepted as the most feasible option for the final disposal of spent nuclear fuels. In this concept, a series of engineered and natural barrier systems are combined to safely store spent nuclear fuel and to isolate it from the biosphere for a practically indefinite period of time. Due to the extremely long lifetime of the DGR, the performance of the DGR replies especially on the natural geologic barriers. Assessing the safety of the DGR is thus required to evaluate the impacts of a wide range of geological, hydrogeological, and physicochemical processes including rare geological events as well as present water cycles and deep groundwater flow systems. Due to the time scale and the complexity of the physicochemical processes and geologic media involved, the numerical models used for safety evaluation need to be comprehensive, robust, and efficient. This study describes the development of an accessible, transparent, and extensible integrated hydrologic models (IHM) which can be approved with confidence by the regulators as well as scientific community and thus suitable for current and future safety assessment of the DGR systems. The IHM under development can currently simulate overland flow, groundwater flow, near surface evapotranspiration in a modular manner. The IHM can also be considered as a framework as it can easily accommodate additional processes and requirements for the future as it is necessary. The IHM is capable of handling the atmospheric, land surface, and subsurface processes for simultaneously analyzing the regional groundwater driving force and deep subsurface flow, and repository scale safety features, providing an ultimate basis for seamless safety assessment in the DGR program. The applicability of the IHM to the DGR safety assessment is demonstrated using illustrative examples.