‘탄소중립’에 대한 관심이 높아짐에 따라 CO2를 농업적으 로 이용하기 위한 시도가 증가하고 있다. 본 실험은 발전소에 서 부산물로 배출되는 CO2를 포집하여 액화·정제 후 시설 엽 채류의 생육 및 생산성 증대를 위한 시비용 CO2로의 활용 가 능성을 평가하기 위해 수행되었다. 경상남도 하동지역의 부추, 취나물, 미나리 농장에 드라이아이스가 공급되었고 각 농 장의 온실 중 하나의 온실은 대조군, 하나의 온실은 CO2 처리 구로 사용되었다. CO2의 시비는 자체 제작한 장치를 사용하 여 드라이아이스에서 승화된 가스를 온실에 공급했다. 부추 온실은 대조군과 CO2 처리에서 온실 내 CO2 농도의 차이가 없었고 두 온실 모두 높은 CO2 농도를 보였다. 반면에 취나물 과 미나리 온실에서는 CO2 시비 처리에서 높은 CO2 농도가 측정되었다. 취나물 및 미나리의 생육은 CO2 시비 처리구에 서 유의성 있게 증가하였으며 수확량도 각각 36%와 25%로 증가하였다. 경제성 분석 결과, 취나물 농가에서는 소득률이 증가하였지만, 부추와 미나리 농가는 감소하는 것으로 나타 났다. 따라서 화력발전소에서 부산물로 발생한 드라이아이스 의 이용은 시설 엽채류의 생산성을 높일 수 있었다

본 연구는 같은 일적산광량(DLI) 조건에서 각각의 광강도 와 광주기 조합이 십자화과 식물 3종의 생육 및 글루코시놀레 이트(GSLs) 함량에 미치는 영향을 살펴보고자 수행되었다. 갓(Brassica juncea (L.) Czern.), 적겨자(Brassica juncea L.), 케일(Brassica oleracea L. var. acephala (DC.) Alef.) 종자를 암면 배지에 파종하여 3주간 육묘하였다. 육묘 후 DLI 를 10mol·m-2·d-1로 설정하여 10h-280, 14h-200, 18h-155, 22h-127μmol·m-2·s-1 처리 조합으로 3주간 처리하였다. 14h- 200μmol·m-2·s-1 처리한 결과, 적갓과 케일은 지상부 생체중/ 건물중, 엽수, 엽면적이 증가하였지만, 유의적 차이는 관찰되 지 않았다. 적갓의 GSLs 함량에서 14h-200μmol·m-2·s-1 처리 가 10h-280, 18h-155, 22h-127μmol·m-2·s-1 처리보다 각각 유의적으로 139.95, 135.87, 154.03% 증가하였으며, 케일의 경우 14h-200μmol·m-2·s-1 처리가 다른 처리구보다 각각 132.96, 122.06, 131.78% 유의하게 증가하였다. 적겨자는 18h-155μmol·m-2·s-1 처리에서 지상부 생체중/건물중, GSLs 함량이 증가하였으며, 14h-200μmol·m-2·s-1 처리에서 엽수가 10h-280, 18h-155, 22h-127μmol·m-2·s-1 처리보다 각각 15.62, 12.12, 32.14% 유의적으로 증가하였다. 유사한 십자 화과 작물이라도 종에 따라 DLI 반응이 다르기 때문에 최적의 DLI 조건 구명과 최소의 전력 소비량 및 최대 효율을 내기 위 한 광질 연구를 병행하여 더 세밀한 결과를 도출해야 할 것으 로 사료된다.

본 논문에서는 등가 단자유도를 이용하여 구조부재의 정적변위를 고려하는 해석기법을 제시하였다. 기존의 단자유도 비선형 동적 해석 알고리듬을 구조부재의 초기정적변위의 영향을 고려할 수 있도록 개선하였다. 가정된 폭발하중 지속시간과 부재의 고유주기 비 에 따라 정적변위가 최대응답에 미치는 영향의 차이와 폭발하중의 방향과 초기변위의 방향에 따른 차이를 확인하였다. 이에 따라 기 존의 응답 차트를 정적변위를 고려할 수 있도록 폭발하중의 형태에 따라 각각 제시하였다. 설계 예제를 정적변위가 고려된 응답 차트 에 적용하여 부재의 최대 변위를 비교 및 분석하였다. 본 연구의 결과를 통해 초기 정적변위를 고려한 구조부재의 최대응답을 쉽게 산 정할 수 있으며 본 연구에서 제시한 응답 차트는 플랜트 또는 군사시설물의 내폭 설계에 활용될 수 있다.

This study presents a rapid and quantitative radiochemical separation method for Nb isotopes in radioactive waste samples from the nuclear power plant with anion exchange resin after Fe coprecipitation. After radionuclides were leached from the radioactive waste samples with concentrated HCl and HNO3, the Nb isotopes were coprecipitated with Fe after filtering the leaching solution with 0.45 micron HA filter, while the Sr, Tc and Ni isotopes were in the solution. The Nb isotopes were separated in HCl medium with anion exchange resin. The purified Nb isotopes were measured using a low level liquid scintillation counter after installing quenching curve with standard Nb-94 isotopes. The separation method for Nb isotopes investigated in this study was applied to neutron dosimeter samples from the nuclear power plant after validating the Nb activity concentration with gamma spectrometry system.

The decommissioning of a nuclear power plant is a project that consists of several stages, and various technologies are applied when performing various tasks at each stage. And it is essential to secure safety and economic feasibility. As the paradigm has changed due to digital transformation in various industries, digitalization is applied to the life cycle of nuclear power plant from construction, operation and decommissioning project. Element technologies are being developed for decommissioning plan establishment, process design, econtamination method, decommissioning work process, waste management, environmental monitoring and radiation dose simulation. The utilization of digital twin in the decommissioning stage is classified into three categories. ① Process Monitoring (decommissioning work procedure, work progress (plan/actual), real-time work status and etc.) ② Facility Monitoring (real-time sensing and video data monitoring, decommissioning SSCs information, work alarm and etc.) ③ Safety Monitoring (work safety, radiation exposure, fire monitoring, work risk and etc.) A system suitable for the decommissioning stage and work should be developed in consideration of the target of use, development function, and when to create data according to the purpose of the system. Simulation module according to user purpose should be provided. In addition, data-base management should be performed according to the decommissioning characteristics in consideration of the data associated with the existing operating system. The system to be developed should support the project management to comply with the domestic standards and regulations to be determined in the future. This will improve the competitiveness of domestic and foreign markets.

Under Article 17 of the Radioactive Waste Management Act and Article 12 of the Enforcement Decree of the Radioactive Waste Management Act, KHNP shall reserve the cost for the decommissioning of NPPs as provisions. To preserve the value, an additional amount considering the discount rate is to be added annually. The initial provision is decided by estimating the decommissioning cost of NPP at the time of commercial operation, calculating the future cost by applying the inflation rate to the expected start date of decommissioning, and then discounting it at a discount rate to the present value. According to the current notice, the period for applying inflation and discount rate is defined as the period of 5 years added to the design life of NPP, which is presumed to be due to the assumption that all decommissioning costs are incurred at once 5 years after the permanent shutdown of the power plant. However, assuming that the actual decommissioning period of a domestic nuclear power plant is generally planned for 15 years, it can be expected that most of the decommissioning activities will begin after the decommissioning preparation and transition period, or 5 years after permanent shutdown of the plant. Considering this, it can be said that the current period (5 years + design life) for applying inflation and discount rate is set a little conservatively. In this paper, the initial provision is calculated by appropriately distributing the decommissioning costs of overseas NPPs categorized by International Structure for Decommissioning Costing (ISDC) during the planned decommissioning period of domestic NPPs, and then adding up the decommissioning cost each year by separately applying the inflation and discount period, which was compared with the results calculated using the current method. Through this, it was confirmed that the revised method had the effect of reducing the initial provision by 2.2% to 5.7% compared to the current method depending on the gap between inflation rate and discount rate, which can be converted to about 8 years of inflation and discount period used in the current method. It is expected that this paper will be used in the future as a basic reference for developing a more accurate method for calculating the initial provision of decommissioning cost.

The decommissioning of nuclear power plant (NPP) consists of various activities, such system decontamination, take out of activated components, segmentation of the activated components, site remediation, etc. During various activities, the generation of radioactive wastes and radiation exposure to workers is expected. The systematic waste management during the activities is important to implement the decommissioning. The inefficient waste management usually bring significant delay in decommissioning process and results in increase of decommissioning cost. The radiation exposure management is also an important issue. It is generally accepted that the hot spot, generated from operation and decommissioning of NPP, is observed in many places within containment building. Although the health physicists measure the radiation in various points, the unintended hot spots are sometimes generated and observed. The effective radiation exposure management also requires the control of personnel and space during various activities. In this study, the radiation exposure and waste management experiences of Zion NPP is reviewed. The primary nuclides and radiation exposure during various activities are systematically studied to achieve the main objectives of this paper.

Reliable evaluation of radioactivity inventory for the nuclear power plant components and residual materials is very important for decontamination and decommissioning. This can make it possible to define optimum dismantling approaches, to determine radioactive waste management strategies, and to estimate the project costs reasonably. To calculate radioactivity of the nuclear power plant structure, various information such as interest nuclide, cross-section, decay constant, irradiation time, neutron flux, and so on is required. Especially irradiation time and neutron flux level are very changeable due to cycle specific fuel loading pattern, the plant overhaul, cycle length. However most of the radioactivity calculations have generally been performed assuming one representative or average neutron flux during the lifetime of the nuclear power plant. This assumption may include excessive conservatism because the radioactivity level has the characteristics of saturation and decay. Therefore, considering these variables as realistically as possible could prevent overestimation. In order to perform realistic radioactivity calculation, we developed monthly relative power contribution factor applying plant-specific operation history and cycle-specific neutron flux. The factors were applied to the radioactivity calculation. The calculation results ware compared with measured values of the neutron monitors that were actually installed and withdrawn from the nuclear power plant. As a result of the comparisons, there are good agreements between the calculated values and measured values. These accurate calculation results of radioactivity could contribute to the establishment of radioactive waste dismantling strategies, the classification of radioactive waste, and the deposit of disposal costs for safe and reasonable decommissioning of the nuclear power plant.

Solid radioactive waste such as rubble, trimmed trees, contaminated soil, metal, concrete, used protective clothing, secondary waste, etc. are being generated due to the Fukushima nuclear power plant accident occurred on March 11, 2011. Solid radioactive waste inside of Fukushima NPP is estimated to be about 790,000 m3. The solid radioactive waste includes combustible rubble, trimmed trees, and used protective clothing, and is about 290,000 m3. These will be incinerated, reduced to about 20,000 m3 and stored in solid waste storage. The radioactive waste incinerator was completed in 2021. About 60,000 m3 of rubble containing metal and concrete with a surface dose rate of 1 mSv/h or higher will be stored without reduction treatment. Metal with a surface dose rate of 1 mSv/h or less are molten, and concrete undergoes a crushing process. About 60,000 m3 of contaminated soil (0.005 ~1 mSv/h) will be managed in solid waste storage without reduction treatment. The amount of secondary waste generated during the treatment of contaminated water is about 6,500 huge tanks, and additional research is being conducted on future treatment methods.

In 2017, Kori unit 1 nuclear power plant was permanently shut down at the end of its life. Currently, Historical Site Assessment (HSA) for MARSSIM characteristics evaluation is being conducted according to the NUREG-1575 procedure, this is conducted through comprehensive details such as radiological characteristics preliminary investigation and on-site interview. Thus, the decommissioning of nuclear power plant must consider safety and economic feasibility of structures and sites. For this purpose, the establishment of optimal work plan is required which simulations in various fields. This study aims to establish procedure that can form a basis for a rational decommissioning plan using the virtual nuclear power plant model. The mapping procedure for 3D platform implementation consisted of three steps. First, scan the inside and outside of the nuclear power plant for decommissioning structure analysis, 3D modeling is performed based on the data. After that, a platform is designed to directly measure the radiation dose rate and mapped the derived to the program. Finally, mapping the radiation dose rate for each point in 3D using the radiation dose rate calculation factor according to the time change the measured value created on the 3D mapping platform. When the mapping is completed, it is possible to manage the exposure dose of workers according to the ALARA principle through the charge of radiation dose rate over time because of visualization of the color difference to the radiation dose rate at each point. For addition, the exposure dose evaluation considering the movement route and economic feasibility can be considered using developed program. As the interest in safety accidents for workers increases, the importance of minimum radiation dose and optimal work plan for workers is becoming increasingly important. Through this mapping procedure, it will be possible to contribute to the establishment of reasonable process for dismantling nuclear power plant in the future.

Regulations on the concentration of boron discharged from industrial facilities, including nuclear power plants, are increasingly being strengthened worldwide. Since boron exists as boric acid at pH 7 or lower, it is very difficult to remove it in the existing LRS (Liquid Radwaste System) using RO and ion exchange resin. As an alternative technology for removing boron emitted from nuclear power plants, the electrochemical boron removal technology, which has been experimentally applied at the Ringhal Power Plant in Sweden, was introduced in the last presentation. In this study, the internal structure of the electrochemical module was improved to reduce the boron concentration to 5 mg/L or less in the 50 mg/L level of boron-containing waste liquid. In addition, the applicability of the electrochemical boron removal technology was evaluated by increasing the capacity of the unit module to 1 m3/hr in consideration of the actual capacity of the monitor tank of the nuclear power plant. By applying various experimental conditions such as flow rate and pressure, the optimum boron removal conditions using electrochemical technology were confirmed, and various operating conditions necessary for actual operation were established by configuring a concentrated water recirculation system to minimize secondary waste generation. The optimal arrangement method of the 1 m3/hr unit module developed in this study was reviewed by performing mathematical modeling based on the actual capacity of monitor tank and discharge characteristics of nuclear power plant.