Radioactive contamination distribution in nuclear facilities is typically measured and analyzed using radiation sensors. Since generally used detection sensors have relatively high efficiency, it is difficult to apply them to a high radiation field. Therefore, shielding/collimators and small size detectors are typically used. Nevertheless, problems of pulse accumulation and dead time still remain. This can cause measurement errors and distort the energy spectrum. In this study, this problem was confirmed through experiments, and signal pile-up and dead time correction studies were performed. A detection system combining a GAGG sensor and SiPM with a size of 10 mm × 10 mm × 10 mm was used, and GAGG radiation characteristics were evaluated for each radiation dose (0.001~57 mSv/h). As a result, efficiency increased as the dose increased, but the energy spectrum tended to shift to the left. At a radiation dose intensity of 400 Ci (14.8 TBq), a collimator was additionally installed, but efficiency decreased and the spectrum was distorted. It was analyzed that signal loss occurred when more than 1 million particles were incident on the detector. In this high-radioactivity area, quantitative analysis is likely to be difficult due to spectral distortion, and this needs to be supplemented through a correction algorithm. In recent research cases, the development of correction algorithms using MCNP and AI is being actively carried out around the world, and more than 98% of the signals have been corrected and the spectrum has been restored. Nevertheless, the artificial intelligence (AI) results were based on only 2-3 overlapping pulse data and did not consider the effect of noise, so they did not solve realistic problems. Additional research is needed. In the future, we plan to conduct signal correction research using ≈10×10 mm small size detectors (GAGG, CZT etc.). Also, the performance evaluation of the measurement/analysis system is intended to be performed in an environment similar to the high radiation field of an actual nuclear facility.

To evaluate the inventory of radionuclides for the disposal of waste generated from nuclear power plants, indirect assessment methods such as the scaling factor method or average radioactivity concentration method can be applied. A scaling factor represents the average concentration ratio between key radionuclides and difficult-to-measure (DTM) radionuclides, while the average radioactivity concentration refers to the average concentration of DTM radionuclides, regardless of the concentration of key radionuclides or within specific ranges of key radionuclide concentrations. These indirect assessment methods can be statistically derived through the analysis of representative drums. This study will address how to apply these scaling factors and average radioactivity concentrations. Firstly, the concentration of gamma-emitting radionuclides will be analyzed using a drum radionuclide analyzer, and the concentration of DTM radionuclides will be determined by applying scaling factors specific to each DTM radionuclide. In the case of using the average radioactivity concentration method, the average concentration of DTM radionuclides will be applied independently of the concentration of gamma-emitting radionuclides. It is crucial to perform radioactive decay correction based on the date of generation or disposal when applying scaling factors or average radioactivity concentration. Additionally, for repackaged 320 L drums, determining which drum among the two 200 L drums inside should serve as the reference is of utmost importance

The primary objective of this study is to evaluate a systematic design’s effectivity in remediating actual uranium-contaminated soil. The emphasis was placed on practical and engineering aspects, particularly in assessing the capabilities of a zero liquid discharge system in treating wastewater derived from soil washing. The research method involved a purification procedure for both the uranium-contaminated soil and its accompanying wastewater. Notably, the experimental outcomes demonstrated successful uranium separation from the contaminated soil. The treated soil could be self-disposed of, as its uranium concentration fell below 1.0 Bq·g−1, a level endorsed by the International Atomic Energy Agency for radionuclide clearance. The zero liquid discharge system’s significance lay in its distillation process, which not only facilitated the reuse of water from the separated filtrate but also allowed for the self-disposal of high-purity Na2SO4 within the residues of the distilled filtrate. Through a comparative economic analysis involving direct disposal and the application of a remediation process for uranium-contaminated soil, the comprehensive zero liquid discharge system emerged as a practical and viable choice. The successful demonstration of the design and practicality of the proposed zero liquid discharge system for treating wastewater originating from real uranium-contaminated soil is poised to have a lasting impact.

Nuclear power plants use ion exchange resins to purify liquid radioactive waste generated while operating nuclear power plants. In the case of PHWR, ion exchange resins are used in heavy water and dehydration systems, liquid waste treatment systems, and heavy water washing systems, and the used ion exchange resins are stored in waste resin storage tanks. The C-14 radioactivity concentration in the waste resin currently stored at the Wolseong Nuclear Power Plant is 4.6×106 Bq/g, exceeding the low-level limit, and if all is disposed of, it is 1.48×1015 Bq, exceeding the total limit of 3.04×1014 Bq of C-14 in the first stage disposal facility. Therefore, disposal is not possible at domestic low/medium-level disposal facilities. In addition, since the heavy water reactor waste resin mixture is stored at a ratio of about 20% activated carbon and zeolite mixture and about 80% waste resin, mixture extraction and separation technology and C-14 desorption and adsorption technology are required. Accordingly, research and development has been conducted domestically on methods to treat heavy water waste resin, but the waste resin mixture separation method is complex and inefficient, and there are limitations in applying it to the field due to the scale of the equipment being large compared to the field work space. Therefore, we would like to introduce a resin treatment technology that complements the problems of previous research. Previously, the waste resin mixture was extracted from the upper manhole and inspection hole of the storage tank, but in order to improve limitations such as worker safety, cost, and increased work time, the SRHS, which was planned at the time of nuclear power plant design, is utilized. In addition, by capturing high-purity 14CO2 in a liquid state in a high-pressure container, it ensures safety for long-term storage and is easy to handle when necessary, maximizing management efficiency. In addition, the modularization of the waste resin separation and withdrawal process from the storage tank, C-14 desorption and monitoring process, high-concentration 14CO2 capture and storage process, and 14CO2 adsorption process enables separation of each process, making it applicable to narrow work spaces. When this technology is used to treat waste resin mixtures in PHWR, it is expected to demonstrate its value as customized, high-efficiency equipment that can secure field applicability and safety and reflect the diverse needs of consumers according to changes in the working environment.

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.

Advanced countries in the field of nuclear research and technology are currently examining the feasibility of deep geological disposal as the most appropriate method for the permanent management of high-level radioactive waste, with no intention of future retrieval. Deep geological disposal involves the placement of such waste deep underground within a stable geological formation, ensuring its permanent isolation from the human environment. To guarantee the enduring isolation and retardation of radionuclides with half-lives spanning tens of thousands to millions of years from the broader ecosystem, it is imperative to comprehend the long-term evolution of deep disposal systems, especially the role of natural barriers. These natural barriers, typically consisting of bedrock, encase the repository and undergo long-term evolutions due to tectonic movements and climate variations. For the effective disposal of high-level radioactive waste, a thorough assessment of the site’s long-term geological stability is essential. This necessitates a comprehensive understanding of its tectonic evolution and development characteristics, including susceptibility to seismic and magmatic events like earthquakes and intrusions. Furthermore, a detailed analysis of alterations in the hydrogeological and geochemical environment resulting from tectonic movements over extended time frames is required to assess the potential for the migration of radionuclides. In this paper, we have examined international evaluation methodologies employed to elucidate the predictive long-term evolution of natural barriers within disposal systems. We have extracted relevant methods from international case studies and applied a preliminary scenario illustrating the long-term evolution of the geological environment at the KURT (KAERI Underground Research Tunnel) site. Nevertheless, unlike international instances, the scarcity of quantitative data limits the depth of our interpretation. To present a dependable scenario in the future, it is imperative to develop predictive technologies aimed at comprehensively studying the geological evolution processes in the Korean peninsula, particularly within the context of radioactive waste disposal.

The disposal of spent nuclear fuel (SNF) in a deep geological repository (DGR) is a widely accepted strategy for the long-term sequestration of radiotoxic SNF. Ensuring the safety of a DGR requires the prediction of various reactions and migration behaviors of radionuclides (RNs) present in SNF within its geochemical surroundings. Understanding the dissolution behaviors of mineral phases harboring these RNs is crucial, as the levels of RNs in groundwater are basically linked to the solubility of these solid phases. Accurate measurements of solubility demand the use of welldefined solid materials characterized by chemical compositions and structures. Herein, we attempted the synthesis of sklodowskite, a magnesium-uranyl (U(VI))-silicate, employing a twostep hydrothermal synthetic approach documented previously. Subsequently, we subjected this synthesized sklodowskite to various analytical techniques, including powder X-ray diffraction (pXRD), scanning electron microscopy/energy dispersive X-ray spectrometry (SEM/EDX), and vibrational spectroscopies (FTIR and Raman). Based on our findings, we confidently identify the obtained mineral phase as sklodowskite (Mg[UO2SiO3OH]2·5H2O). This identification is primarily based on the similarity between its pXRD pattern and the reference XRD pattern of sklodowskite. Furthermore, the measured infrared and Raman spectra show the vibrational modes of UO2 2+ and SiO4 4- ions, particularly within the 700~1,100 cm-1 region, which support that the synthetic mineral has a characteristic layered uranyl-silicate structure of crystalline sklodowskite. Finally, we utilized synthetic minerals to estimate its solubility up to about three months in a model groundwater, where the dissolved species composition is analogous to that of granitic groundwater from the KAERI Underground Research Tunnel. In this presentation, we will present in detail the results of spectroscopic characterizations and the methodology employed to assess the solubility of the U(VI)-silicate solid phase.

The thermal evaluations for the conceptual design of the deep geological repository considering the improved modeling of the spent fuel decay heat were conducted using COMSOL Multiphysics computational program. The maximum temperature at the surface of a disposal canister for the technical design requirement should not exceed 100°C. However, the peak temperature at the canister surface should not exceed 95°C considering the safety margin of 5°C due to several uncertainties. All thermal evaluations were based on the time-dependent simulation from the emplacement time of the canister to 100,000 years later. In particular, the heat source condition was set to the decay heat rate and axial decay heat profile of the PLUS7 fuel with 4.0wt% U-235 and 45 GWD/MTU. The thermal properties of the granitic rock in South Korea were applied to the host rock region. For the reference design case, the cooling time of the SNF was set to 40 years, the distance between the deposition holes 8 meters and that between the deposition tunnels 30 meters. However, the peak temperature at the canister surface at 10 years was 95.979°C greater than 95°C. This design did not meet the thermal safety requirement and needed to be modified. For the first modified case, when the distance between the deposition tunnels was set to 30 meters, three cooling time cases of 40, 50 and 60 years and five distances of 6, 7, 8, 9 and 10 meters between the deposition holes were considered. The design with the distances of 9 and 10 meters between the deposition holes for the cooling time of 40 years and all five distances for 50 and 60 years were less than 95°C. For the second modified case, when the distance between the deposition holes was set to 8 meters, three cooling time cases of 40, 50 and 60 years and five distances of 20, 25, 30, 35 and 40 meters between the deposition tunnels were considered. The design with the distances of 35 and 40 meters between the deposition tunnels for the cooling time of 40 years, the distances of 25, 30, 35 and 40 meters for 50 years and all five distances for 60 years were less than 95°C. As a result, the peak temperature at the canister surface decreased as the cooling time and the distance between the deposition holes and the tunnels increased.

Spent nuclear fuel management is a high-priority issue in South Korea, and addressing it is crucial for the country’s long-term energy sustainability. The KORAD (Korea Radioactive Waste Agency) is leading a comprehensive, long-term project to develop a safe and effective deep geological repository for spent nuclear fuel disposal. Within this framework, we have three primary objectives in this work. First, we conducted statistical analysis to assess the inventory of spent nuclear fuel in South Korea as of 2021. We also projected future generation rates of spent nuclear fuels to identify what we refer to as reference spent nuclear fuels. These reference spent nuclear fuels will be used as the design basis spent fuels for evaluating the safety of the repository. Specifically, we identified four types of design basis reference spent nuclear fuels: high and low burnup from PLUS7 (with a 16×16 array) and ACE7 (with a 17×17 array) assemblies. Second, we analyzed radioactive nuclides’ inventory, activities, and decay heats, extending up to a million years after reactor discharge for these reference spent nuclear fuels. This analysis was performed using SCALE/TRITON to generate the burnup libraries and SCALE/ORIGEN for source term evaluation. Third, to assess the safety resulted from potential radioactive nuclides’ release from the disposal canister in future work, we selected safety-related radionuclides based on the ALI (Annual Limit of Intake) specified in Annex 3 of the 2019-10 notification by the NSSC (Nuclear Safety and Security Commission). Conservative assumptions were made regarding annual water intake by humans, canister design lifetime, and aquifer flow rates. A safety margin of 10-3 of the ALI was applied. We selected 56 radionuclides that exceed the intake limits and have half-lives longer than one year as the safety-related radionuclides. However, it is crucial to note that our selection criteria focused on ALI and half-lives. It did not include other essential factors such as solubility limits, distribution coefficients, and leakage processes. So, some of these nuclides can be removed in a specific analysis area depending on their properties.

Nuclear power generation is expected to be enlarged for domestic electricity supply based on the 10th Basic Plan of Long-Term Electricity Supply and Demand. However, the issues on the disposal of spent nuclear fuel or high-level radioactive waste has not been solved. KBS-3 concept of the deep geological disposal and pyroprocessing has been investigated as options for disposal and treatment way of spent nuclear fuel. In other way, the radionuclide management process with 6 scenarios are devised combining chlorination treatment and alternative disposal methods for the efficient disposal of spent nuclear fuel. Various scenarios will be considered and comprehensively optimized by evaluation on many aspects, such as waste quantity, radiotoxicity, economy and so on. Level 0 to 4 were identified with the specialized nuclide groups: Level 0 (NFBC, Hull), Level 1 (Long-lived, volatile nuclides), Level 2 (High heat emitting nuclides), Level 3 (TRU/RE), Level 4 (U). The 6 options (Op.1 to 6) were proposed with the differences between scenarios, for examples, phase types of wastes, the isolated nuclide groups, chlorination process sequences. Op.1 adopts Level 0 and 1 to separate I, Tc, Se, C, Cs nuclides which are major concerns for long-term disposal through heat treatment. The rest of spent nuclear fuel will be disposed as oxide form itself. Op.2 contains Sr separation process using chlorination by MgCl2 and precipitation by K2CO3to alleviate the burden of heat after heat treatment process. U/TRU/RE will be remained and disposed in oxide form. Op.3 is set to pyroprocessing as reference method, but residual TRU/RE chlroides after electrorefining will be recovered as precipitates by K3PO4. Op.4 introduces NH4Cl to chlorinate TRU/RE from oxides after Op.2 applied and precipitates them. TRU/RE/Sr will be simultaneously chlorinated by NH4Cl without MgCl2 in Op.5. Then, chlorinated Sr and TRU/RE groups will be separated by post-chlorination process for disposal. But, chlorinated Sr and TRU/RE are designed not to be divided in disposal steps in Op.6. In this study, the mass flow analysis of radionuclide management process scenarios with updated process variables are performed. The amount and composition of wastes by types will be addressed in detail.

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.

To address the pressing societal concern in Korea, characterized by the imminent saturation of spent nuclear fuel storage, this study was undertaken to validate the fundamental reprocessing process capable of substantially mitigating the accumulation of spent nuclear fuel. Reprocessing is divided into dry processing (pyro-processing) and wet reprocessing (PUREX). Within this context, the primary focus of this research is to elucidate the foundational principles of PUREX (Plutonium Uranium Redox Extraction). Specifically, the central objective is to elucidate the interaction between uranium (U) and plutonium (Pu) utilizing an organic phase consisting of tributyl phosphate (TBP) and dodecane. The objective was to comprehensively understand the role of HNO3 in the PUREX (Plutonium Uranium Redox Extraction) process by subjecting organic phases mixed with TBPdodecane to various HNO3 concentrations (0.1 M, 1.0 M, 5.0 M). Subsequently, the introduction of Strontium (Sr-85) and Europium (Eu-152) stock solutions was carried out to simulate the presence of fission products typically contented in the spent nuclear fuel. When the operation proceeds, the complex structure takes the following form. 􀜷􀜱􀬶 􀬶􀬾(􀜽􀝍) + 2􀜰􀜱􀬷 􀬿(􀜽􀝍) + 2􀜶􀜤􀜲(􀝋􀝎􀝃) ↔ 􀜷􀜱􀬶(􀜰􀜱􀬷)􀬶 ∙ 2􀜶􀜤􀜲(􀝋􀝎􀝃) Subsequently, separate samples were gathered from both the organic and aqueous phases for the quantification of gamma-rays and alpha particles. Alpha particle measurements were conducted utilizing the Liquid Scintillation Counter (LSC) system, while gamma-ray measurements were carried out using the High-Purity Germanium Detector (HPGe). The distribution ratio for U, Eu (Eu-152), and Sr (Sr-84) was ascertained by quantifying their activity through LSC and HPGe. Through the experiments conducted within this program, we have gained a comprehensive understanding of the selective solvent extraction of actinides. Specifically, uranium has been effectively separated from the aqueous phase into the organic phase using a combination of tributyl phosphate (TBP) and dodecane. Subsequently, samples containing U(VI), Eu(III), and Sr(II) underwent thorough analysis utilizing LSC and HPGe detectors. Our radiation measurements have firmly established that the concentration of nitric acid enhances the selective separation of uranium within the process.

In nuclear fuel development research, consideration of the back-end cycle is essential. In particular, a review of an in-reactor performance of nuclear fuel related to the various degradation phenomena that can occur during spent fuel dry storage is an important area. The important factors affecting the degradation of zirconium-based cladding during dry storage are the cladding’s hydrogen concentration and rod internal pressure after irradiation. In this study, a preliminary analysis of the in-reactor behavior of the HANA cladding, which has been developed and is currently undergoing licensing review, was performed, and based on this result, a comparative analysis between nuclear fuel with HANA cladding and current commercial fuel under storage conditions was performed. The results show that the rod internal pressure of nuclear fuel with HANA cladding is not significantly different from that of commercial cladding, and the hydrogen concentration in the cladding tends to reduce due to the increased corrosion resistance, so fuel integrity in a dry storage conditions is not expected to be a major problem. Although the lack of cladding creep data under dry storage conditions, the results from the Halden research reactor test comparing in-reactor creep behavior with Zircaloy-4 showed that there is sufficient margin for degradation due to creep during storage.

The radionuclide management process is a conditioning technology to reduce the burden of spent fuel management, and refers to a process that can separate and recover radionuclides having similar properties from spent fuels. In particular, through the radionuclide management process, high heat- emitting, high mobility, and high toxicity radionuclides, which have a significant impact on the performance of disposal system, are separated and managed. The performance of disposal system is closely related to properties (decay heat and radioactivity) of radioactive wastes from the radionuclide management process, and the properties are directly linked to the radionuclide separation ratio that determines the composition of radionuclides in waste flow. The Korea Atomic Energy Research Institute have derived process flow diagrams for six candidates for the radionuclide management process, weighing on feasibility among various process options that can be considered. In addition, the GoldSim model has been established to calculate the mass and properties of waste from each unit process of the radionuclides management process and to observe their time variations. In this study, the candidates for the radionuclide management process are evaluated based on the waste mass and properties by using the GoldSim model, and sensitivity analysis changing the separation ratio are performed. And the effect of changes in the separation ratio for highly sensitive radionuclides on waste management strategy is analyzed. In particular, the separation ratio for high heat-emitting radionuclides determines the period of long-term decay storage.

Owing to the great demand for portable and wearable chemical sensors, the development of all-solid-state potentiometric ion sensors is highly desirable considering their simplicity and stability. However, most ion sensors are challenged by the penetration of water and gas molecules into ion-selective membranes, causing unstable and undesirable sensing performances. In this study, a hydrophobic ionic liquid-modified graphene (Gr) sheet was prepared using a fluid dynamics-induced exfoliation and functionalization process. The high hydrophobicity and electrical double-layer capacitance of Gr make it a potential solid-state ion-to-electron transducer for the development of potentiometric sodium-ion ( Na+) sensors. The as-prepared Na+ sensors effectively prevented the formation of the water layer and penetration of gas species, resulting in stable and high sensing performances. The Na+ sensors showed a Nernstian sensitivity of 58.11 mV/[Na+] with a low relative standard deviation (0.46), fast response time (5.1 s), good selectivity (K < 10− 4), and good durability. Furthermore, the Na+ sensor demonstrated its feasibility in practical applications by measuring accurate and reliable ion concentrations of artificial human sweat and tear samples, comparable to a commercial ion meter.

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.