The nuclear licensee must ensure that the nuclear or radiological emergency preparedness and response organization is explicitly defined and staffed with adequate numbers of competent and assessed personnel for their roles. This paper describes the responsibilities of medical and support personnel for the medical action of casualties in the event of a radiological emergency at the KAERI. Currently, there is one medical personnel (nurse) in KAERI, and a total of eight medical support personnel are designated for medical response in the event of a radiological emergency. These medical support personnel are designated as one or two of the on-site response personnel at each nuclear facility, operating as a dedicated team of A, B (4 people each). In the event of a radiological emergency, not all medical support personnel are mobilized, but members of the dedicated medical team, which includes the medical support personnel of the nuclear facility where the accident has occurred, are summoned. Medical and support personnel will first gather in the onsite operational support center (OSC)/technical support center (TSC) to prepare and stand by for the medical response to injured when a radiological emergency is declared. They should take radiation protective measures, such as wearing radiation protective clothing and dosimeters, before entering the onsite of a radiological emergency, because injuries sustained during a radiological emergency may be associated with radioactive contamination. In the event of an injury, direct medical treatment such as checking the patient’s vitals, first aid, and decontamination will be carried out by medical personnel, while support personnel are mainly responsible for contacting the transfer hospital, reporting the patient’s condition, accompanying the ambulance, filling out the emergency medical treatment record, and supporting medical personnel. In order to respond appropriately to the occurrence of injuries, we regularly conduct emergency medical supplies education and medical training for medical support personnel to strengthen their capabilities.

In the nuclear fuel cycle (NFC) facilities, the failure of Heating Ventilation and Air Conditioning (HVAC) system starts with minor component failures and can escalate to affecting the entire system, ultimately resulting in radiological consequences to workers. In the field of air-conditioning and refrigerating engineering, the fault detection and diagnosis (FDD) of HVAC systems have been studied since faults occurring in improper routine operations and poor preventive maintenance of HVAC systems result in excessive energy consumption. This paper aims to provide a systematic review of existing FDD methods for HVAC systems therefore explore its potential application in nuclear field. For this goal, typical faults and FDD methods are investigated. The commonly occurring faults of HVAC are identified through various literature including publications from International Energy Agency (IEA) and American Society of Heating, Refrigerating and Air-Conditioning Engineers (ASHRAE). However, most literature does not explicitly addresses anomalies related to pressure, even though in nuclear facilities, abnormal pressure condition need to be carefully managed, particularly for maintaining radiological contamination differently within each zone. To build simulation model for FDD, the whole-building energy system modeling is needed because HVAC systems are major contributors to the whole building’s energy and thermal comfort, keeping the desired environment for occupants and other purposes. The whole-building energy modeling can be grouped into three categories: physics-based modeling (i.e., white-box models), hybrid modeling (i.e., grey-box models), and data-driven modeling (i.e., black-box models). To create a white-box FDD model, specialized tools such as EnergyPlus for modeling can be used. The EnergyPlus is open source program developed by US-DOE, and features heat balance calculation, enabling the dynamic simulation in transient state by heat balance calculation. The physics based modeling has the advantage of explaining clear cause-and-effect relationships between inputs and outputs based on heat and mass transfer equations, while creating accurate models requires time and effort. Creating a black-box FDD model requires a sufficient quantity and diverse types of operational data for machine learning. Since operation data for HVAC systems in existing nuclear cycle facilities are not fully available, so efforts to establish a monitoring system enabling the collection, storage, and management of sensor data indicating the status of HVAC systems and buildings should be prioritized. Once operational data are available, well-known machine learning methods such as linear regression, support vector machines, random forests, artificial neural networks, and recurrent neural networks (RNNs) can be used to classify and diagnose failures. The challenge with black-box models is the lack of access to failure data from operating facilities. To address this, one can consider developing black-box models using reference failure data provided by IEA or ASHRAE. Given the unavailability of operation data from the operating NFC facilities, there is a need for a short to medium-term plan for the development of a physics-based FDD model. Additionally, the development of a monitoring system to gather useful operation data is essential, which could serve both as a means to validate the physics-based model and as a potential foundation for building data-driven model in the long term.

The radwaste repository consists of a multi-barrier, including natural and engineered barriers. The repository’s long-term safety is ensured by using the isolation and delay functions of the multi-barrier. Among them, natural barriers are difficult to artificially improve and have a long time scale. Therefore, in order to evaluate its performance, site characteristics should be investigated for a sufficient period using various analytical methods. Natural barriers are classified into lithological and structural characteristics and investigated. Structural factors such as fractures, faults, and joints are very important in a natural barrier because they can serve as a flow path for groundwater in performance evaluation. Considering the condition that the radioactive waste repository should be located in the deep part, the drill core is an important subject that can identify deep geological properties that could not be confirmed near the surface. However, in many previous studies, a unified method has not been used to define the boundaries of structural factors. Therefore, it is necessary to derive a method suitable for site characteristics by applying and comparing the boundary definition criteria of various structural factors to boreholes. This study utilized the 1,000 m deep AH-3 and DB-2 boreholes and the 500 m deep AH-1 and YS- 1 boreholes drilled around the KURT (KAERI Underground Research Tunnel) site. Methods applied to define the brittle structure boundary include comparing background levels of fracture and fracture density, excluding sections outside the zone of influence of deformation, and confining the zone to areas of concentrated deformation. All of these methods are analyzed along scanlines from the brittle structure. Deriving a site-specific method will contribute to reducing the uncertainties that may arise when analyzing the long-term evolution of brittle structures within natural barriers.

The Colloid Formation and Migration (CFM) international joint research initiative continues as a part of the GTS’s Radionuclide Retardation Programme, which has been in progress since 1984. This project focuses on examining the formation of colloids from a bentonite-engineered barrier system and exploring how these colloids impact the migration of radionuclides in fractured host rock when subjected to advective flow. Phase 1 of the project was launched in 2004 and concluded in early 2008, focusing on preliminary studies related to in-situ boundary conditions, predicting models, and supplementary lab works. Following that, Phase 2 spanned from 2008 to 2013 and aimed at fortifying the field setup by adding three new monitoring boreholes and suitable instrumentation in both the boreholes and tunnel. This phase also tested the system’s resilience while mapping the flow domain. Phase 3 kicked off in January 2014 and extended until December 2018. During this period, the Long-term In-situ Test (LIT) was introduced in May 2014, featuring a set of compacted bentonite rings laced with radionuclide tracers. These were placed in a borehole to serve as a colloid and radionuclide source. CFM Phase 4 initiative commenced in January 2019, marking the successful deployment of the i-BET (In-situ Bentonite Erosion Test). This project component involves placing approximately 50 kg of compacted bentonite in a natural water-conducting shear zone. Korea Atomic Energy Research Institute (KAERI) joined CFM in 2008 to examine the behavior of colloid generation and migration with radionuclides in the Underground Research Laboratory. The fourth phase of the CFM project was also scheduled to include a post-mortem evaluation of the LIT and additional tracer experiments in the well-mapped MI shear zone. This study aims to provide an interim update on the ongoing i-BET, a key component of Phase 4 of the CFM project. We will also discuss the current status of the post-mortem analysis for the LIT experiment. In addition, we will outline plans for the forthcoming Phase VI of the project. These plans will continue to advance our understanding of radionuclide migration and the influence of bentonite-based disposal systems.

The effectiveness of a crystalline natural barrier in providing sealing capabilities is based on the behavior of numerous fractures and their intersections within the rock mass. It is important to evaluate the evolving characteristics of fractured rock, as the hydro-mechanical coupled processes occurring through these fractures play a dominant role. KAERI is actively developing a true tri-axial compression test system and concurrently conducting hydro-mechanical experiments using replicated fractured rock samples. This research is focused on a comprehensive examination of coupled processes within fractures, with a particular emphasis on the development of true tri-axial testing equipment. The designed test system has the capability to account for three-dimensional stress conditions, including vertical and both maximum and minimum horizontal principal stresses, realizing the disposal conditions at specific underground depths. Notably, the KAERI-designed test system employs the mixed true tri-axial concept, also known as the Mogi-type, which allows for fluid flow into fractures under tri-axial compression conditions. This system utilizes a hydraulic chamber to maintain constant stress in one direction through the application of oil pressure, while the other two directional stresses are applied using rigid platens with varying magnitudes. Once these mechanical stress conditions are established, control over fluid flow is achieved through the rigid platens in contact with the specimen section. This pioneering approach effectively replicates in-situ mechanical conditions while concurrently observing the internal fluid flow patterns within fractures, thereby enhancing our capacity to study these coupled phenomena. As future research, numerical modeling efforts will be proceeding with experimental data-driven approaches to simulate the coupled behavior within the fractures. In these numerical studies, two distinct fracture geometry domains will be generated, one employing simplified rough-walled fractures and the other utilizing mismatched rough-walled fractures. These investigations mark the preliminary steps in the process of selecting and validating an appropriate numerical model for understanding the hydro-mechanical evolution within fractures.

Long-term climate and surface environment changes can influence the geological subsurface environment evolution. In this context, a fluid flow pathway developing and connection possibility can be increased between the near-surface zone and deep depth underground. Thus, it is necessary to identify and prepare for the overall fluid flow at the entire geological system to minimize uncertainty on the spent nuclear fuel (SNF) disposal safety. The fluid flow outside the subsurface environment is initially penetrated through the surface and then the unsaturated area. Thus, the previously proved reports, POSIVA in Finland, suggested that sequential research about the fluid infiltration experiment (INEX) and the investigation is necessary. Characterizing the unsaturated zone can help predict changes and ensure the safety of SNFs according to geological long-term evolution. For example, the INEX test was conducted at the upper part of ONKALO, about 50 to 100 m depth, to understand the geochemical evolution of the groundwater through the unsaturated zone, to evaluate the main flow of groundwater that can approach the SNF disposal reservoir, and to estimate the decreasing progress of the buffering capacity along the pathway through the deep geological disposal. In the present study, a preliminary test was performed in the UNsaturated-zone In-situ Test (UNIT) facility near the KAERI underground research tunnel to design and establish a methodology for infiltration experiments consistent with the regional characteristics. The results represented the methodological application is possible for characterizing unsaturated-zone to perform infiltration experiments. The scale of the experiment will be expanded sequentially, and continuous research will be conducted for the next application.

Nuclear power is responsible for a large portion of electricity generation worldwide, and various studies are underway, including the design of permanent deep geological disposal facilities to safely isolate spent nuclear fuel generated as a result. However, through the gradual development of drilling technology, various disposal option concepts are being studied in addition to deep geological disposal, which is considered the safest in the world. So other efforts are also being made to reduce the disposal area and achieve economic feasibility, which requires procedures to appropriately match the waste forms generated from separation process of spent nuclear fuel with disposal option systems according to their characteristics. And safety issue of individual disposal options is performed through comparison of nuclide transport. This study briefly introduces the pre-disposal nuclide management process and waste forms, and also introduces the characteristics of potential disposal options other than deep geological disposal. And environmental conditions and possible pathways for nuclide migration are reviewed to establish transport scenarios for each disposal option. As such, under this comprehensive understanding, this study finally seeks to explore various management methods for high-level radioactive waste to reduce the environmental burden.

Graphene oxide (GO) and ultrafine slag (UFS) have been applied to reinforce cement mortar cubes (CMC) in this research. The consequences of GO and UFS on the mechanical attributes of the CMC were explored through experimental investigations. Established on the results, at the 28 days of hydration, the CMC compressive and flexural strength with 0.03% of GO and 10% UFS were 89.8 N/mm2 and 9.1 N/mm2, respectively. Furthermore, the structural changes of CMC with GO and UFS were qualitatively analysed with instrumental techniques such as scanning electron microscope (SEM), X-ray fluorescence (XRF), thermogravimetric analysis (TGA), Fourier transform infrared spectroscopy (FT-IR), FT Raman spectroscopy, atomic force microscopy (AFM), and 27Al, 29Si-Nuclear magnetic resonance spectroscopy (NMR). SEM results reported that GO and UFS formed an aggregated nanostructure that improved the microstructural properties of the CMC. TGA analysis revealed the quantum of calcium hydrate and bound water accomplished by supplementing GO bound to the UFS aggregates. FT-IR analysis of the CMC samples confirmed the ‘O-’comprising functional groups of GO which expedited the formation of complexes between calcium carbonate ( CaCO3) and UFS. 0.03% GO was the optimum dosage that enhanced the compressive and flexural attributes when combined with 10% UFS in CMC.

Recently a biosafety level-3(BL-3) mobile laboratory has been set up for the virus scanning and vaccine development because of the COVID-19 pandemic. The study on air flow inlet and outlet location and its flow direction with ventilation in the mobile laboratory needs to prevent spread of COVID-19 virus because the COVID-19 virus is primarily transmitted to people through respiratory droplets and aerosol coming out as their coughing. This study is conducted on the air flow pattern optimization in BL-3 mobile laboratory with various design specifications of position of air supply & exhaust port and particle source. Air flow patterns of ceiling supply-exhaust and ceiling supply-bottom side exhaust with particle source were determined to compare the impact of the infection prevention. CFD simulation was used to analyze for two air flow patterns and particle source position. Numerical results showed that air flow pattern of air conditioning system with ceiling supply-exhaust in a row is more effective than that of ceiling supply-bottom side exhaust air flow pattern in terms of infection prevention in biosafety mobile laboratory.