Recently, as interest in safe food and ingredient content has increased, the demand for highly concentrated health functional food that can be manufactured at home is increasing. Accordingly, in this study, we developed a galenic maker that can increase the extraction efficiency of herbal medicine(ginseng) ingredients and shorten the manufacturing time. After verifying the safety of the designed components through structural analysis and thermal flow analysis, the dimensions and shape of the galenic maker were determined. A prototype was produced based on the design process and the performance of the product was verified through extraction efficiency and ingredient analysis.

In this paper, durability verification of forged wheels for automobiles were performed using the finite element method for bending fatigue analysis and impact analysis. In addition, the durability analysis environment of forged wheels was implemented. By analyzing the stress distribution on the surface of the forged wheel, the area with a high possibility of breakage was identified and improved. The durability analysis of the initial model forged wheel was performed by bending fatigue analysis and impact analysis, The stress distribution of the forged wheel surface was analyzed through the analysis results of the initial model. and the spokes, flanges, hubs, and rear parts are less likely to be damaged were cut to reduce the weight by about 10%, and the reliability of the improved model was confirmed.

Considering the Fukushima nuclear accident and the marine discharge plan of contaminated (or treated) water, it is necessary a seafood monitoring system for radioactive material screening. Currently, radioactivity tests in seafood are conducting in Korea. Although current method using a HPGe detector can provide very low uncertainty in determining radioactivity, there is a limitation in that rapid inspection cannot be performed because of a time-consuming pretreatment process as well as long measurement time (typically 10,000 s). To overcome this limitation, we are developing an insitu inspection device, a kind of screening system, which can monitor the radioactivity in seafood in a near real-time basis. In this study, the actual seafood with a check source was measured to verify the reliability of the Monte Carlo simulation model. The detector used in the experiment was a 5-cm-thick polyvinyl toluene (PVT) plastic scintillator with a 0.5-cm-thick lead shield for background reduction. A Cs-137 check source was placed within seafood. The seafood used in the experiment was fishcake, raw oyster, and dried laver, which is the representative seafood of fish, shellfish, and seaweed. These three seafood products of the same size and shape as the manufacturing process were used to predict the performance realistically. We compared the energy spectrum of the Cs-137 check source obtained from measurements and simulations. The region of interest (ROI) around the Compton edge was set to reduce the influence of the background and electronic noise. The results showed that the measured and simulated spectrum were in good agreement.

To ensure the peaceful use of nuclear energy, nuclear safeguards are applied in member states of the International Atomic Energy Agency (IAEA) under the Non-Proliferation Treaty. The two major considerations in implementing nuclear safeguards are effectiveness and efficiency. In terms of efficiency, the IAEA has a great interest in using containment and surveillance (C/S) technology to maintain continuity of knowledge. A representative means of C/S technology is a sealing system to detect tampering. The existing sealing systems used by the IAEA are of limited functionality in realtime verification purposes. To address this limitation, the present study develops a real-time verification sealing system. First, we analyzed the design requirements of a sealing system proposed by various institutions including the IAEA, the U.S. Nuclear Regulatory Commission, and a number of national laboratories and companies. Then, we identified the appropriate design requirements of this system for real-time verification. Finally, the prototype system was developed and tested based on the identified design requirements. The validation tests of the prototype system were performed for anticipated environmental conditions, radiation resistance, and safeguards functionality. Additionally, we are developing user-friendly verification software. The software validation is planned to perform for functionality, performance efficiency, and security. The next step is to develop a commercialized realtime verification sealing system based on the results of validation tests. Using this commercialized system, we plan to evaluate the performance in various actual use cases. Such a system is expected to significantly enhance the efficiency of nuclear safeguards.

In this study, to develop and verify the Jeju-type groundwater thermal system design program, the energy consumption and system performance derived by input into the design program based on the load calculated on the demonstration site and the groundwater temperature were compared and analyzed with the actual measured values. The theoretical values ​​of energy consumption and heating and cooling COP ​​obtained through the Jeju-type groundwater thermal system design program were 11.24kW, 5.28 for heating, 13.31kW, 3.94 for cooling respectively. The measured values ​​of energy consumption and COP of the Jeju-type groundwater thermal system were 3967.2kW and 4.5, respectively. The error between the theoretical value and the measured value obtained through the design program is 0% and 2.39%, respectively. The errors that occur in the predicted values ​​and the actual values ​​are due to variables that are ignored in the system assumptions. If users consider errors and use it when designing groundwater thermal systems, they can estimate the cost of required drilling works, heat exchangers, and heat pumps and analyze economic feasibility.

본 연구는 데이터를 기반으로 한 인공지능 기계학습 기법을 활용하여 온실 내부온도 예측 시뮬레이션 모델을 개발을 수행 하였다. 온실 시스템의 내부온도 예측을 위해서 다양한 방법 이 연구됐지만, 가외 변인으로 인하여 기존 시뮬레이션 분석 방법은 낮은 정밀도의 문제점을 지니고 있다. 이러한 한계점 을 극복하기 위하여 최근 개발되고 있는 데이터 기반의 기계 학습을 활용하여 온실 내부온도 예측 모델 개발을 수행하였 다. 기계학습모델은 데이터 수집, 특성 분석, 학습을 통하여 개 발되며 매개변수와 학습방법에 따라 모델의 정확도가 크게 변 화된다. 따라서 데이터 특성에 따른 최적의 모델 도출방법이 필요하다. 모델 개발 결과 숨은층 증가에 따라 모델 정확도가 상승하였으며 최종적으로 GRU 알고리즘과 숨은층 6에서 r2 0.9848과 RMSE 0.5857℃로 최적 모델이 도출되었다. 본 연 구를 통하여 온실 외부 데이터를 활용하여 온실 내부온도 예 측 모델 개발이 가능함을 검증하였으며, 추후 다양한 온실데이 터에 적용 및 비교분석이 수행되어야 한다. 이후 한 단계 더 나아 가 기계학습모델 예측(predicted) 결과를 예보(forecasting)단 계로 개선하기 위해서 데이터 시간 길이(sequence length)에 따른 특성 분석 및 계절별 기후변화와 작물에 따른 사례별로 개발 모델을 관리하는 등의 다양한 추가 연구가 수행되어야 한다.

This study is about the production of radiation sources of simulated concrete and soil reference materials to verify the validity of the quality establishment and measurement of the detector (HPGe) of the radioactive soil and concrete waste classification system, which is being developed to quickly and accurately classify nuclear decommissioning waste. Specific activity of gamma nucleus among radioactive wastes is evaluated using gamma spectroscopy. At this time, in order to verify the validity and reliability of measuring equipment, it shall be a standardized substance of the same medium as nuclear decommissioning waste (chemical ingredients, particles, density, etc.) in order to correct the energy and efficiency of gamma nuclide analysis equipment. The CRM used for the detector’s energy correction used a 1 L Marinelli beaker standard correctional radiation source consisting of 10 radioactive isotopes. In order to correct efficiency, in accordance with the production and certification process of the Korea Standards and Research Institute, it has produced artificial simulated radioactive concrete similar to nuclear decommissioning waste (30% for cement, 60% for regulation and 10% for bentonite). The radioactive homogeneity of the simulated concrete reference materials was evaluated using dispersion analysis (ANOVA) in accordance with ISO Guide 35, while 137Cs and 60Co of concrete reference materials were able to obtain homogeneous measurements both in and between bottles. The self-absorption rate of the simulated concrete reference material was determined by the MCNP computer simulation measurement method, and the self-absorption correction coefficients of 137Cs and 60Co were assessed at 0.995 and 0.996, respectively, and the standard value for the radiation of the simulated concrete reference material was calculated on the weighted average of the measurements of 20 samples. The uncertainty about the reference value was calculated by combining measurement uncertainty (Type B evaluation), bottle to bottle standard deviation, and uncertainty within bottle by modifying the formula suggested in ISO Guide 35. The concentration of 137Cs and 60Co of reference materials was divided into high-speed measurement mode and precision measurement mode in consideration of the self-disposal standard. The reference value and uncertainty of expansion among reference materials for high-speed measurement mode were rated at 1,032.7 ± 64.0 Bq·kg−1and 1,083.7 Bq·kg−1, respectively. The standard value and expansion uncertainty for 137Cs and 60Co among reference materials for precision measurement mode were rated at 113.7 ± 10.0 Bq·kg−1 and 122.3 ± 10.3 Bq·kg−1, respectively.

In Korea, it is expected that the decommissioning of nuclear reactors will increase due to the license termination of reactors constructed in the 1960s to the 80s. According to the investigation of KORAD, VLLW accounts for 67.10% of decommissioning wastes and amounts to about 413,336 drums. Due to their huge amount, it is necessary to create an appropriate decommissioning waste management plan even though VLLW is disposed at the second-phase disposal facility of the Gyeongju repository. For efficient reduction in decommissioning wastes, it is required to actively use a clearance of metallic and concrete radioactive wastes. Regulations of nuclear safety and security commission notice that the radioactive waste can be reused or recycled if it meets the clearance criterion, 10 μSv·y−1 for individual dose. Therefore, it is important to develop a computational code which calculate individual doses for each scenario, and determine whether the clearance criterion is satisfied. However, in the case of metallic waste, RESRAD-RECYCLE used in dose assessment for the clearance has no longer been maintained or updated since 2005 and there is no code for recycling of concrete waste. For this reason, a dose assessment code RUCAS (Recycle-Underlying Computational dose Assessment System) has been developed by Ulsan National Institute of Science and Technology (UNIST). A point kernel method is adopted into external dose assessment model to calculate more realistic options, which are various geometries of source, and shielding effect. In the case of internal radiation, equations of internal dose from IAEA are used. This research conducts a verification of dose assessment model for recycling of metallic radioactive waste. RESRAD-RECYCLE is the comparison object and results from RESRAD-RECYCLE validation report are referenced. Targets are 14 recycling scenarios composed up to the smelting metal step of four steps, which are arising scrap metal, smelting scrap metal, and fabrication of metal product, and reusing/recycling of product. Seven isotopes, which are Ac-227, Am-241, Co-60, Cs-137, Pu-239, Sr- 90, and Zn-65, are selected for calculation. Validation results for external dose vary by isotopes, but show acceptable differences. It seems to be caused by difference in the calculation method. In the case of internal dose using same calculation formula, results are exactly matched to RESRAD-RECYCLE for all isotopes. Consequently, RUCAS can conduct functions supported by RESRAD-RECYCLE well and future work will be conducted related to domestic recycling scenarios considering public acceptance, and verification with radiation shielding codes for various geometries of source.

APro, developed by KAERI as a process-based total system performance assessment model, can simulate the radionuclide transport affected by thermal, hydraulic, mechanical and geochemical changes that may occurs in the engineering and natural barriers of a geological disposal system. APro targets a large-scale and heterogeneous 3D system that includes more than 10,000 boreholes located about 500 m underground and hundreds of fractures of different sizes distributed within an area of several km2. Simulating transport and reaction phenomena for such a system through the global implicit approach (GIA) may require considerable computational resources or be intractable in some cases. Therefore, APro adopts the sequential non-iterative approach (SNIA), one of the operator splitting (OS) methods, to separate the mass transport and reaction phenomena into independent problems. By using SNIA, the parallel computation performance in APro with multiple cores is expected to be improved. In this study, the effect of SNIA on the parallel computation performance was analyzed through a simple 1D reactive transport problem. Without SNIA, finite difference equations, discretized from the partial differential equations (PDEs) describing the reactive transport problem, have to be solved at once because all dependent variables are nonlinearly and spatially interconnected through reaction and mass transport terms. When the reaction and mass transport terms are separated through SNIA, the mass transport problem can be converted into independent linear equations for each chemical and the efficient linear system solver can be applied to each linear equation. In particular, since the reaction problem is changed to independent nonlinear equations for each node, the parallel computation performance can be greatly improved. To verify this, the 1D reactive transport problem was implemented in MATLAB, and SNIA and GIA were applied to solve the problem. As a result, there was no significant difference in results between SNIA and GIA for proper spatial and temporal discretization, which verified the accuracy of SNIA. In order to see the parallel computation performance, the calculation times for SNIA and GIA with increasing number of cores were measured and compared. As the number of cores increased, the SNIA calculation speed became faster than that of GIA, which verified that SNIA could improve parallel computation performance in APro. In the future, the effect of SNIA on the parallel computation performance will be verified for the numerical analysis of large-scale geological disposal systems.

본 연구의 목적은 정적 플랭크 운동의 동작 차이에 따른 코어근육의 근활성도를 비교하는 것이 었다. 본 연구의 대상자는 “J”대학교 20대 남성 10명을 대상으로 진행하였고(연령, 23.20±0.65 세; 신장, 174.54±1.51 cm; 체중, 70.00±2.24 kg; 신체질량지수, 22.94±0.51 kg/m2), 4가지의 풀, 엘보우, 사이드 및 리버스 정적 플랭크 동작을 수행하였으며, 근전도 분석을 위한 표면전극 부착 부위는 신체 근육의 우측 복직근, 외복사근, 광배근 및 척추기립근으로 설정하였다. 본 실험 결과, 광배근과 척추기립근의 근활성도 는 리버스 플랭크 동작 시 가장 높게 나타났고(p<.001), 복직근과 외복사근의 근활성도는 엘보우 플랭크 동작 시 가장 높게 나타났다(p<.001). 따라서 본 연구의 결과는 추후 정적 플랭크 동작 시 효과적인 운동 프로그램의 자료가 될 것으로 기대된다.

북한 핵에 대한 IAEA의 사찰이 2002년 종료된 이후 북한은 2017년에 여섯 번째의 핵실험과 ICBM급인 화성-15형을 시험 발사하고 핵무력을 완성하였다고 선언하였다. 그리고 북한은 2021년 초 조선노동당 8차 대 회에서는 핵고도화 선언을 하기도 하였으며 2022년 들어 극초음속미사 일, IRBM, ICMB 등을 시험발사하며 한반도 정세를 더욱 악화시키고 있 다. 시간이 경과할수록 북한의 핵능력은 더욱 확장되고 고도화할 가능성 이 많다. 한국과 미국을 포함하는 국제사회는 북한과의 비핵화 협상을 적극 추진하고 실질적 검증을 통해 북한의 완전한 비핵화를 달성해야 한다. 본 연구의 목적은 군비통제 검증 분야의 역사적 성공 사례인 INF 조약 이행과정 분석을 통하여 북한 핵에 대한 핵심적 검증과정인 현장사찰을 촉진하는 방안을 모색하는 것이다. 북한의 완전한 비핵화를 위한 협상에 서 가장 고비가 될 수 있는 검증의 현장사찰에 관한 실행가능하고 발전 적인 방안은 다음과 같다. 첫째, 북한을 NPT에 복귀시키고 IAEA의 안전 조치협정에 가입하도록 하여 ‘IAEA 주도의 현장사찰’을 실시하여야 한 다. 둘째, 북한 핵시설 중 미신고·미공개된 핵시설에 대한 ‘선택적이고 효과적인 현장사찰’이 이루어져야 한다. 셋째, 북한의 폐쇄적인 사회적 특수성을 고려하여 ‘상호협력적 현장사찰’이 이루어져야 한다. 넷째, 다 양한 국내 및 국제기술수단을 이용하여 원격 감시, 위성 및 항공 감시, 핵관련 인원의 인터뷰, 문헌조사, 시료 채취 및 정밀분석 등을 시행하는 ‘과학기술 집약적 통합 현장사찰’이 이루어져야 한다.

The denuclearization of North Korea was unpredictable and resulted in radical changes. Despite the skepticism and disappointment surrounding denuclearization, it is important for certain verification technologies to establish what is technically possible or practically impossible, and how reliable these technical means are. This article presents the technical hurdles in nuclear verification by systematically categorizing them into issues of correctness and completeness. Moreover, it addresses the safety and security risks during the denuclearization process, including the radiological impact on humans, environmental effects, and the illegal transfer of material, information, and technologies.