In order to ensure the long-term safety of a deep geological repository, the performance assessment of the Engineered Barrier System (EBS) considering a thermal process should be performed. The maximum temperature at the side wall of a disposal canister for the technical design requirement should not exceed 100°C. In this study, the thermal modelling was conducted to analyze the effects of the thermal process from a disposal canister to the surrounding near-field host rock using the PFLOTRAN code. The mesh was generated using the LaGriT code and the material properties were assigned by applying the FracMan code. Initial conditions were set as the average geothermal gradient (25.7°C/km) and an average surface temperature (14.7°C) in Korea. The highest temperature was observed at the middle of the canister side wall. The temperature of the buffer was lower than that of the canister, and the temperature increase of the deposition tunnel and the host rock was insignificant due to the lower effect of the heat source. The result of the thermal evolution of the EBS represented the highest thermal effects in the vicinity of the canister. In addition, the thermal effects were largely decreased after 10 years of the entire simulation period. It demonstrated that the model took 3 years to heat up the buffer around the canister. The temperature at the canister side wall increased until 3 years and then decreased after that time. This is because that the radioactive decay heat from the heat source was emitted enough to raise the overall temperature of the EBS by 3 years. However, the decay heat rate of the canister decreased exponentially with the disposal time and then its decay heat was not emitted enough after 3 years. In conclusion, the peak temperature results of the EBS were lower than 70°C to meet the technical design requirement.

The Barcelona Basic Model (BBM) is an elasto-plastic model used to describe the coupled thermo-hydro-mechanical behaviors of unsaturated soil. BBM is frequently adopted to model the unique swelling behavior of bentonite, which is generally considered as the buffer material between the host rock and the canister containing high-level radioactive waste in deep geological repositories, under the changing thermal, hydraulic, mechanical and chemical conditions during the lifetime of repository. Therefore, a variety of the continuum-based numerical methods tried to add the BBM for modelling the multi barrier systems of geological repository and succeeded to describe the elasto-plastic deformation of bentonite. However, to demonstrate the entire barrier systems the host rock should be modelled simultaneously with the buffer materials, and the continuum-based methods may be limited in their ability to reflect the fracture networks in the host rock which could be the major flow channels of groundwater. This research applies BBM in 3DEC, a three-dimensional block-based discrete element method, and validates the model by comparing the change of specific volume and mean effective stress during three numerical test cases. Discontinuum-based numerical methods with BBM can be extended to describe the coupled thermo-hydro-mechanical processes of multi-barrier systems in geological repositories, with a focus on the interaction between the host rock and bentonite.

The natural barrier system surrounding the geological repository for high-level radioactive waste plays a crucial role in preventing or delaying the leakage of radionuclides. Therefore, the natural barrier should ensure low permeability to prevent groundwater flow into the engineered barrier system throughout the repository’s lifetime. Crystalline rock, considered as the host rock for the geological repository in Korea, exhibits low intact rock permeability, but the crystalline rock often contains the multiple discontinuities due to its high brittleness that can allow the unexpected fluid flow. Therefore, the long-term hydraulic behavior of the discontinuity should be characterized while considering additional thermal, mechanical, and chemical effects. In comparison to thermal, hydraulic, and mechanical processes, the chemical processes on the discontinuities progress relatively slowly, resulting in limited researches to include these chemical processes. This research introduces mechanisms the involving coupled thermal-hydraulic-mechanicalchemical processes focusing on the rough fracture surfaces and asperities. The chemically-induced changes in mechanical and hydraulic properties are described based on pressure solution and precipitation concepts. A comprehensive review of laboratory tests, field tests, and numerical simulations is conducted related to the chemically-induced coupled processes in fractured rock. Laboratory tests, in particular, concentrate on microscopic changes in fracture asperities induced by pressure solution to analyze chemically-induced aperture changes. The TOUGHREACT, an integral finite difference method program for thermal-hydraulic-chemical simulations, is generally employed to model the chemical response of pressure solution and precipitation on fracture surfaces. The TOUGHREACT includes a module to describe effective porosity and permeability changes based on the modified cubic law, so the real-time change of the fracture permeability can be reflected during the flow simulation. Considering the coupled thermal-hydraulic-mechanicalchemical processes of discontinuity, it becomes evident that the chemical processes under repository conditions (long-term, high temperature, and high pressure) can disturb the hydraulic performance of the natural barrier, so further research is required to characterize the chemically-induced coupled processes for assessing the long-term performance of the natural barrier system.

The compacted bentonite buffer is a key component of the engineered barrier system in deep geological repositories for high-level radioactive waste disposal. Groundwater infiltration into the deep geological repository leads to the saturation of the bentonite buffer. Bentonite saturation results in bentonite swelling, gelation and intrusion into the nearby rock discontinuities within the excavation damaged zone of the adjacent rock mass. Groundwater flow can result in the erosion and transport of bentonite colloids, resulting in bentonite mass loss which can negatively impact the long-term integrity and safety of the overall engineered barrier system. The hydro -mechanicalchemical interactions between the buffer, surrounding host rock and groundwater influence the erosion characteristics of the bentonite buffer. Hence, assessing the critical hydro-mechanicalchemical factors that negatively affect bentonite erosion is crucial for the safety design of the deep geological repository. In this study, the effects of initial bentonite density, aperture, discontinuity angle and groundwater chemistry on the erosion characteristics of Bentonil WRK are investigated via bentonite extrusion and artificial fracture experiments. Both experiments examine bentonite swelling and intrusion into simulated rock discontinuities; cylindrical holes for bentonite extrusion experiments and plane surfaces for artificial fracture experiments. Compacted bentonite blocks and bentonite pellets are manufactured using a compaction press and granulation compactor respectively and installed in the transparent extrusion cells and artificial fracture cells. The reference test condition is set to be 1.6 g/cm3 dry density and saturation using distilled water. After distilled water or solution injection, the axial and radial expansion of the bentonite specimens into the simulated rock discontinuities are monitored for one month under free swelling conditions with no groundwater flow. Subsequent flow tests are conducted using the artificial fracture cell to determine the critical flow rate for bentonite erosion. The intrusion and erosion characteristics are modelled using a modified hydro-mechanicalchemical coupled dynamic bentonite diffusion model and a fluid-based hydro-mechanical penetration model.

The presence of technological voids in deep geological repositories for high-level radioactive nuclear waste can have negative effects on the hydro-mechanical properties of the engineered barrier system when groundwater infiltrates from the surrounding rock. This study conducted hydration tests along with image acquisition and X-ray CT analysis on compacted Korean bentonite samples, which simulated technological voids filling to investigate the behavior of fracturing (piping erosion) and cracking deterioration. We utilized a dual syringe pump to inject water into a cell consisting of a bentonite block and technological voids at a consistent flow rate. The results showed that water inflow to fill technological voids led to partial hydration and self-sealing, followed by the formation of an erosional piping channel along the wetting front. After the piping channel generated, the cyclic filling-piping stage is characterized by the repetitive accumulation and drop of water pressure, accompanied by the opening and closing of piping channels. The stoppage of water inflow leads to the formation of macro- and micro cracks in bentonite due to moisture migration caused by high suction pressure. These cracks create preferential flow paths that promote longterm groundwater infiltration. The experimental test and analysis are currently ongoing. Further experiments will be conducted to investigate the effects of different dry density in bentonite, flow rate, and chemical composition of injected water.

Korea Atomic Energy Research Institute (KAERI) has been operating the Post Irradiation Examination Facility (PIEF). The facility has many PIE equipment and one of them is a hydrogen analyzer for measuring hydrogen contents in Zr cladding of spent fuel. The cladding tube of fuel is oxidized in the core environment of high temperature and pressure and absorbs some of the hydrogen generated during the oxidation. The hydrogen content increases with the increase of burn-up, and causes hydriding of the material, which degrades the mechanical properties. Therefore, hydrogen content analysis of the cladding tube is required for the performance and integrity evaluation of spent fuel. In PIEF, the hydrogen analyzer extracts hydrogen gas from Zr cladding by the hot extraction method. The hydrogen gas flows with inert gas and oxidizes to H2O through a CuO reagent. Finally, an IR detector measures the hydrogen amount from the absorbed IR intensity at a specific wavelength. Because the equipment is in the glove box and has some consumable parts, the maintenance work was performed as a radiation work.

Nuclear fuel assemblies are exposed to high temperature and high pressure environments underwater for long periods of time in a reactor, leading to deterioration of the assembly structure. These assembly consists of fuel rods, grids, a top nozzle, a bottom nozzle and guide tubes. In particular, the integrity of the guide tube made of Zircaloy-4 is a very important part in handling the assembly. In the Post Irradiation Examination Facility (PIEF), there are 14×14 Westinghouse STD assemblies that have lost their handleability due to the top nozzle being removed for damaged fuel rod test. To handle these assemblies, it is reasonable to use cut guide tubes whenever possible. Therefore, it is necessary to determine the irradiation embrittlement state of the guide tube before designing or manufacturing parts that can connect the top nozzle and the guide tubes. Therefore, in this paper, the location for installing the top nozzle-guide tube connection parts was selected in the height range of 3,460 to 3,713 mm, and guide tube specimens were made within that range. Offset strain was derived from the load-displacement curve obtained through compression testing to confirm whether the ductility of guide tubes was maintained. As a result, there was no significant difference in strength and ductility of the guide tube within the above length range. In addition, it was confirmed that the ductility was maintained enough to install the top nozzle-guide tube connection parts. Therefore, it is judged that there will be no problem even if the top nozzle-guide tube connection parts are installed in the guide tube to restore the handleability of the assemblies.

For efficient design and manufacture of PWR spent fuel burnup detector, data simulated with various condition of spent fuel in the NPP storage pool is required. In this paper, to derive performance requirements of spent fuel burnup detector for neutron flux and dose rates were evaluated at various distances from CE16 and WH17 types of fuel, representatively. The evaluation was performed by the following steps. First, the specifications of the spent fuel, such as enrichment, burnup, cooling time, and fuel type, were analyzed to find the conditions that emit maximum radioactivity. Second, gamma and neutron source terms of spent fuel were analyzed. The gamma source terms by actinides and fission products and neutron source terms by spontaneous and (α, n) reactions were calculated by SCALE6 ORIGAMI module. Third, simulation input data and model were applied to the evaluation. The material composition and dose conversion factor were referred as PNNL-15870 and ICRP-74 data, respectively and dose rates were displayed with the MCNP output data. It was assumed that there was only one fuel modeled by MCNP 6.2 code in pool. The evaluation positions for each distance were selected as 5 cm, 10 cm, 25 cm, 50 cm, and 1 m apart from the side of fuel, respectively. Fourth, neutron flux and dose rates were evaluated at distance from each fuel type by MCNP 6.2 code. For WH 17 types with a 50 GWd/MTU burnup from 5 cm distance close to fuel, the maximum neutron flux, gamma dose rates and neutron dose rates are evaluated as 1.01×105 neutrons/sec, 1.41×105 mSv/hr and 1.61×101 mSv/hr, respectively. The flux and dose rate of WH type were evaluated to be larger than those of CE type by difference in number of fuel rods. The relative error for result was less than 3~7% on average secured the reliability. It is expected that the simulated data in this paper could contribute to accumulate the basic data required to derive performance requirements of spent fuel burnup detector.

Due to the saturation of spent fuel pool of nuclear power plant in Korea, temporary storage for spent fuel will be installed, and spent fuel will be stored and managed in dry cask for a considerable period of time. Since spent nuclear fuel must withstand continuous decay heat, radiation and high internal pressure of the fuel rod in the cask, behavior of spent nuclear fuel is needed to be reviewed. Spent nuclear fuel used in Pressurized Water Reactor (PWR) in Korea is stored in a wet storage currently, but it is going to store a temporary dry-storage facility on Kori site. Therefore, it is very important and meaningful to evaluate the behavior of nuclear fuel with realistic modeling. Also, domestic PWR nuclear fuel has various burn-up. In the past, the burn-up of nuclear fuel in light water reactors was low, but in order to increase power generation efficiency, the concentration of uranium was increased and the number of new fuel was increased. Therefore, a large amount of nuclear fuel with burn-up of 45,000 MWD/MTU or higher, generally called high burn-up, is also stored in the spent fuel pool (SFP). Therefore, it is necessary to evaluate by dividing three different burn-up such as, low, medium, and high burn-up. Thus, this study will review the behavior of nuclear fuel at different burn-up during the temporary storage period with FALCON (EPRI), computational code and analyze the factors affecting the integrity of nuclear fuel, including when the temporary storage is extended its additional lifetime. And this evaluation will contribute developing the spent fuel management plan in Korea.

Regulatory agencies require burn-up verification to ensure that dry storage casks using burn-up credit are not loaded with fuel with a reactivity greater than the allowable standard. Accordingly, in preparation for dry storage of SF, the reliability of the burnup was verified and action plans for fuel with confirmed errors were reviewed. Reliability verification was performed by comparing the actual burnup calculated with combustion calculation code (TOTE, ISOTIN) used in NPP and the design burnup calculated with the nuclear design code (ANC). As a result of comparing the differences between actual burnup and design burnup for 7,414 assemblies of SF generated from CE-type NPPs, the average deviation was confirmed to be 0.79% and 220 MWD/MTU. In the CE-type NPPs, no fuel showing large deviations was identified, and it was confirmed that reliability was secured. As a result of comparing the differences in 11,082 assemblies of SF generated from WH-type NPPs, the differences were not large, averaging 1.16% or 422 MWD/MTU. However, fuels showing significant differences were identified, and cause analysis was performed for those fuels. The cause analysis used a method of comparing the burnup of symmetrically loaded fuels in the reactor. For fuels that were not symmetrically loaded, a method was used to compare them with fuels with similar combustion histories. As a result of the review, it was confirmed that the fuel was under- or over-burned compared to symmetrically loaded fuel. For fuels for which clear errors have been identified, we are considering replacing them with the design burnup, and for fuels whose causes cannot be confirmed, we are considering ways to maintain the actual burnup.

A lot of CANDU Spent Fuels (CSFs) have been stored in spent nuclear fuel pools and dry storage facilities. In accordance with the enhanced nuclear regulations, the initial characteristics of CSF should be inspected to ensure the integrity of CSF and the reliable operation of storage system before loading it into a cask for long-term dry storage. For the inspections, an initial characteristics measurement equipment was designed, which is used for Pool-Side Examination (PSE) in the spent fuel pool of the pressurized heavy water reactor nuclear power plant. Measurements using the equipment consist of non-contact inspections and contact inspections. The non-contact inspections do not affect CSF integrity, whereas the integrity of CSF can be reduced during the contact inspections under abnormal operating conditions because the probe of equipment may apply specific loads to the CSF. Therefore, the structural integrity evaluations of equipment and CSF are performed using Finite Element (FE) analyses for four combinations based on two abnormal conditions and two probe positions. The used abnormal conditions are the pressing load condition and the scratching load condition, and two probe positions are the center and bottom of the fuel rod in the longitudinal direction, respectively. In this evaluation, the bottoms of the fuel rod or CSF are defined as the regions facing the bottom surface of equipment. The analysis of the pressing load condition is performed by pressing the probe of the equipment in radial direction of the CSF fuel rod. That of the scratching load condition is carried out by applying a specific radial load to the CSF fuel rod using the probe and then applying the load to the surface of the fuel rod while moving axially along the surface. All combinations are analyzed considering geometric, boundary and material non-linearity under the dynamic load, which is dependent on the equipment operating velocity. The stresses of CSF and equipment components were obtained from these analyses. The maximum stress of each component was generated at the combination on the scratching load condition for the bottom position among the four combinations. The obtained maximum stresses are lower than the yield stress for each component material. Also, the CSF is not overturned due to the support plate of the equipment in all analyses. Therefore, the structural integrity and safety of the equipment and the CSF are maintained under abnormal operating conditions during the inspection using the initial characteristic measurement equipment.

Recently, the status of North Korea’s denuclearization has become an international issue, and there are also indications of potential nuclear proliferation among neighboring countries. So, the need for establishment of nuclear activity verification technology and strategy is growing. In terms of ensuring verification completeness, sample collection-based analysis is essential. The concepts of Chain of Custody (CoC) and Continuity of Knowledge (CoK) can be defined in the process of sample extraction as follows: CoC is interpreted as the ‘system for managing the flow of information subjected by the examinee’, and CoK is interpreted as the ‘Continuity of information collection through CoC subjected by the inspector’. In the case of sample collection process in unreported areas for nuclear activity verification, there are additional risks such as worker exposure/kidnapping or sample theft/tampering. Therefore, the introduction of additional devices might be required to maintain CoC and CoK in the unreported area. In this study, an Environmental Geometrical Data Transfer (EGDT) was developed to ensure the safety of workers and the CoC/CoK of the samples during the collection process. This device was designed for achieving both mobility and rechargeability. It is categorized into two modes based on its intended users: sample mode and worker mode. Through the sensors, which is positioned in the rear part of device, such as radiation, gyroscope, light, temperature, humidity and proximity sensors, it can be easily achievable various environmental information in real-time. Additionally, GPS information can also be received, allowing for responsiveness to various hazardous scenarios. Moreover, the OLED display positioned on the front gives us for checking device information such as the current status of the device such as the battery level, the connectivity of wifi, and etc. Finally, an alarm function was integrated to enable rapid awareness during emergency situations. These functions can be updated and modified through Arduino-based firmware, and both the device and the information collected through it can be remotely controlled via custom software. Based on the presented design conditions, a prototype was developed and field assessments were conducted, yielding results within an acceptable margin of error for various scenarios. Through the application of the EGDT developed in this study to the sample collection process for nuclear activity verification purposes, it is expected to achieve a stable maintenance of CoC/CoK through more accurate information transmission and reception.

Emerging technologies are innovative technologies currently under development or in the early stages of introduction. These technologies have the potential to impact a wide range of industries and sectors significantly and may, therefore, be subject to export controls. The list of emerging technologies subject to export controls varies from country to country and constantly changes as new technologies are developed. For example, the U.S., EU, and South Korea have responded to these changes by adding software and technologies related to artificial intelligence and machine learning to their export control lists. Nevertheless, export control of emerging technologies still presents challenges and limitations. The rapid pace of technological advancement makes it difficult for export control regulations to keep up. For export control purposes, international cooperation on information sharing and control methods is necessary for most countries to control similar items. Several new technologies in the nuclear field may be subject to export controls. These technologies include advanced reactors, nuclear fuel cycle technologies, and nuclear waste management technologies. Small modular reactors (SMRs) and fourth-generation reactors are being developed as advanced technologies, and new technologies are being developed to improve the nuclear fuel cycle. There is also active development of technologies for space applications utilizing nuclear reactors, such as the Nuclear Thermal Propulsion System and the Nuclear Electric Propulsion System. As these technologies may include new systems and items not in existing export control, they may pose a proliferation risk or may include software design know-how for advanced materials, it is necessary to consider whether and how they should be subject to export control to prevent nuclear proliferation. Overall, export controls are an essential issue in the emerging technology and nuclear energy sectors. Countries are moving toward strengthening regulations and international cooperation to overcome these challenges and ensure safe technology transfer, and South Korea should actively participate and lead this trend.

We produced an activated carbon using sodium-lignosulfonate, in which we investigated how the sodium salt in lignin served as the activating agent during heat treatment. Our process resulted in a product with a high specific surface area of 1324 m2/ g at 800 °C and microporous structure. During the activation process, we observed the consumption of carbon due to the dehydration reaction of NaOH and the reduction of Na2CO3 to metallic Na, which created pores through oxidation/ reduction reactions. The intercalation of metallic Na between the lattices at high temperatures formed additional pores and increased the specific surface area. Our proposed mechanism holds promise for enhancing the control of the microstructure and porosity of activated carbons through the thermal treatment of biomass.

The development of food packaging materials with mechanical and antimicrobial properties is still a major challenge. N, P-doped carbons (NPCs) were synthesized. Poly(butylene adipate-co-terephthalate) (PBAT), which has an adverse effect on the environment and affects petroleum resources, has been commonly used for applications as food packaging. The development of PBAT composites reinforced with NPCs and studies on their structure and antimicrobial properties are presented in this study. The composite materials in the PBAT/NPCs were processed by solution casting. The plasticizing properties of NPCs enhanced the mechanical strength of composites produced of PBAT and NPCs. The thermal properties of PBAT composites were enhanced with addition of NPCs, according to thermogravimetric analysis (TGA). After reinforcement, PBAT/NPCs composites became more hydrophobic, according to contact angle measurements. In studies against S. aureus and E. coli food-borne pathogenic bacteria, the obtained composites show noticeably improved antimicrobial activity. The composite materials, according to the results of PBAT and NPCs may be a good choice for packing for food that prevents microorganisms.

증산은 적정 관수 관리에 중요한 역할을 하므로 수분 스트레스에 취약한 토마토와 같은 작물의 관개 수요에 대한 지식이 필요하다. 관수량을 결정하는 한 가지 방법은 증산량을 측정하는 것인데, 이는 환경이나 생육 수준의 영향을 받는다. 본 연구는 분단위 데이터를 통해 수학적 모델과 딥러닝 모델을 활용하여 토마토의 증발량을 추정하 고 적합한 모델을 찾는 것을 목표로 한다. 라이시미터 데이터는 1분 간격으로 배지무게 변화를 측정함으로써 증산 량을 직접 측정했다. 피어슨 상관관계는 관찰된 환경 변수가 작물 증산과 유의미한 상관관계가 있음을 보여주었다. 온실온도와 태양복사는 증산량과 양의 상관관계를 보인 반면, 상대습도는 음의 상관관계를 보였다. 다중 선형 회귀 (MLR), 다항 회귀 모델, 인공 신경망(ANN), Long short-term memory(LSTM), Gated Recurrent Unit(GRU) 모델을 구 축하고 정확도를 비교했다. 모든 모델은 테스트 데이터 세트에서 0.770－0.948 범위의 R2 값과 0.495mm/min－ 1.038mm/min의 RMSE로 증산을 잠재적으로 추정하였다. 딥러닝 모델은 수학적 모델보다 성능이 뛰어났다. GRU 는 0.948의 R2 및 0.495mm/min의 RMSE로 테스트 데이터에서 최고의 성능을 보여주었다. LSTM과 ANN은 R2 값이 각각 0.946과 0.944, RMSE가 각각 0.504m/min과 0.511로 그 뒤를 이었다. GRU 모델은 단기 예측에서 우수한 성능 을 보였고 LSTM은 장기 예측에서 우수한 성능을 보였지만 대규모 데이터 셋을 사용한 추가 검증이 필요하다. FAO56 Penman-Monteith(PM) 방정식과 비교하여 PM은 MLR 및 다항식 모델 2차 및 3차보다 RMSE가 0.598mm/min으로 낮지만 분단위 증산의 변동성을 포착하는 데 있어 모든 모델 중에서 가장 성능이 낮다. 따라서 본 연구 결과는 온실 내 토마토 증산을 단기적으로 추정하기 위해 GRU 및 LSTM 모델을 권장한다.