• 전기의세계
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• 제16권1호
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• pp.19-28
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• 1967

Indigenous energy resources available in Korea are limited to the three major resources such as Korean anthracite, hydraulic potential and wood and straws. As reported in various reports concerning energy problem in Korea, unfortunately these three major resources are not only poor in quality but also limited in quantity. The amount of energy to be imported, which will be increased at a considerably high rate by years due to the shortage in the supply by domertic sources against the demand, is studied in the view-pint of sound and logical energy dependence upon the external sources. What would occur, if the imported energy would be exclusively limited to an energy source only, has an enough reason to be paid a significant consideration. As a result, the feasibility is discarded in favour of nuclear power plants after an extensive prospect for electric power development plan covering more than coming thirty years, i.e., up to the year of 2,000 A.D. In briefing, this paper indicates that a measures to accomodate as large amount of nuclear power plants as possible in the electric power system is not only inevitable for a sound solution of the severe energy problem with which Korea is to be confronted but also leads to the national benefit.

• Nuclear Engineering and Technology
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• 제40권6호
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• pp.517-526
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• 2008

The widespread concern regarding the management of spent fuel that mainly contributes to nuclear waste has led to the development of the sodium-cooled fast reactor (SFR) as one of the most promising future types of reactors at both national and international levels. Various reactor deployment scenarios with SFR introductions with different conversion ratios in the existing PWR-dominant nuclear fleet have been assessed to optimize the SFR deployment strategy to replace PWRs with the view toward a reduction in the level of spent fuel as well as efficient uranium utilization through its reuse in a closed fuel cycle. An efficient reactor deployment strategy with the SFR introduction starting in 2040 has been drawn based on an SFR deployment strategy in which burners are deployed prior to breakeven reactors to reduce the amount of PWR spent fuel substantially at the early deployment stage. The PWR spent fuel disposal is reduced in this way by 98% and the cumulative uranium demand for PWRs to 2100 is projected to be 445 ktU, implying a uranium savings of 115 ktU. The SFR mix ratio in the nuclear fleet near the year 2100 is estimated to be approximately 35-40%. PWRs will remain as a main power reactor type until 2100 and SFRs will support waste minimization and fuel utilization.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(1)
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• pp.571-576
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• 1996

Currently most nuclear power plants(NPPs) are adopted the mid-loop operation to minimize the overhaul period and save the operating cost. For mid-loop operation it is essential to install nozzle dam between RCS pipe and steam generator(SG). Because SG remains more highly contaminated with radioactive material than any other parts of the NPPs, the repairmen are very reluctant to carry out installing nozzle dam inside the SG. Until now, unfortunately, it appears that no practically applicable device was developed to provide the longstanding demand. Also the accidents have been reported by licenser event report during this operation mode due to loss of residual heat removal(RHR). The purpose of this paper is to conduct remotely blocking and disintegration of nozzle of a SG which has the highest radiation exposure during the maintenance in NPPs. The remote nozzle blocking device of a SG includes three bladders, hubs, air controller provisions to supply and contact air pressure into the bladders. This remote nozzle block device will give the larger operation margin to prevent the loss of RHR and minimize the radiation exposure dose to the repairman and shorten the overhaul periods.

• Nuclear Engineering and Technology
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• 제49권5호
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• pp.905-913
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• 2017

Advanced reactor designs often feature longer operating cycles between refueling and new concepts of operation beyond traditional baseload electricity production. Owing to this increased complexity, traditional proportional-integral control may not be sufficient across all potential operating regimes. The prototypical advanced reactor (PAR) design features two independent reactor modules, each connected to a single dedicated steam generator that feeds a common balance of plant for electricity generation and process heat applications. In the current research, the PAR is expected to operate in a load-following manner to produce electricity to meet grid demand over a 24-hour period. Over the operational lifetime of the PAR system, primary and intermediate sodium pumps are expected to degrade in performance. The independent operation of the two reactor modules in the PAR may allow the system to continue operating under degraded pump performance by shifting the power production between reactor modules in order to meet overall load demands. This paper proposes a Takagi-Sugeno (T-S) fuzzy logic-based power distribution system. Two T-S fuzzy power distribution controllers have been designed and tested. Simulation shows that the devised T-S fuzzy controllers provide improved performance over traditional controls during daily load-following operation under different levels of pump degradation.

• Nuclear Engineering and Technology
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• 제53권7호
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• pp.2396-2406
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• 2021

The Pacific Earthquake Engineering Research (PEER) Center has been developing a performance-based earthquake engineering (PBEE) methodology, which is based on explicit determination of performance, e.g., monetary losses, in a probabilistic manner where uncertainties in earthquake ground motion, structural response, damage estimation, and losses are explicitly considered. To carry out the PEER PBEE procedure for a component of the nuclear power plant (NPP) such as the cable tray system, hazard curve and spectra were defined for two hazard levels of the ground motions, namely, operation basis earthquake, and safe shutdown earthquake. Accordingly, two sets of spectral compatible ground motions were selected for dynamic analysis of the cable tray system. In general, the PBEE analysis of the cable tray in NPP was introduced where the resulting floor motions from the time history analysis (THA) of the NPP structure should be used as the input motion to the cable tray. However, for simplicity, a finite element model of the cable tray was developed for THA under the effect of the selected ground motions. Based on the structural analysis results, fragility curves were generated in terms of specific engineering demand parameters. Loss analysis was performed considering monetary losses corresponding to the predefined damage states. Then, overall losses were evaluated for different damage groups using the PEER PBEE methodology.

• Nuclear Engineering and Technology
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• 제54권10호
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• pp.3878-3887
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• 2022

For the future energy-mix policy for carbon neutrality, demand for the capability of load-follow operation has emerged in nuclear power plants in order to accommodate the intermittency of renewable energy. The short-term decay heat analysis is also required to evaluate the decay heat level varied by the power level change during the load-follow operation, which is a very important parameter in terms of short-term decay heat removal during a grace time. In this study, the short-term decay heat level for 10 days after the shutdown was evaluated for both seasonal and daily load-follow cases. Additionally, the nuclide-wise contribution to the accumulated decay heat for 10 days was analyzed for further understanding of the short-term decay heat behavior. The result showed that in the seasonal case, the decay heat level was mainly determined by the power level right before the shutdown and the amount of each nuclide was varied with the power variation due to the long variation interval of 90 days. Whereas, in the daily case, the decay heat level was strongly impacted by the average power level during operation and meaningful mass variations for those nuclides were not observed due to the short variation interval of 0.5 days.

• 방사선산업학회지
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• 제9권3호
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• pp.153-159
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• 2015

Reliability of printed circuit board (PCB), which is based on high integrated circuit technology, is having been important because of development of electric and self-driving car. In order to answer these demand, automated X-ray inspection (AXI) is best solution for PCB non-destructive test. PCB is consist of plastic, copper, and, lead, which have low to high Z-number materials. By using dual energy X-ray imaging, these materials can be inspected accurately and efficiently. Dual energy X-ray imaging, that have the advantage of separating materials, however, need some solution such as energy separation method and enhancing efficiency because PCB has materials that has wide range of Z-number. In this work, we found out several things by analysis of X-ray energy spectrum. Separating between lead and combination of plastic and copper is only possible with energy range not dose. On the other hand, separating between plastic and copper is only with dose not energy range. Moreover the copper filter of high energy part of dual X-ray imaging and 50 kVp of low energy part of dual X-ray imaging is best for efficiency.

• 자원ㆍ환경경제연구
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• 제28권2호
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• pp.177-200
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• 2019

전력수급계획의 근간이 되는 전력수요 전망은 GDP와 기상변수 등 다양한 요인에 의해 영향을 받기 때문에 확률 프로세스로 이해할 수 있다. 이 전망치를 바탕으로 전력설비의 구성 방안이 수립되는데, 실제 의사결정 과정은 주어진 확률분포에 대한 정보가 온전하다고 가정한다는 한계를 가진다. 그러나 현실적으로는 확률분포 자체의 중첩 불확실성이 존재하기 때문에 강건한 최적계획(robust optimization)의 수립이 필요하다. 본 논문은 중첩 불확실성을 포함한 발전설비 조정의 최적의사결정을 연구한다. 구체적으로 원자력의 감축투자 관련 실물옵션 모형을 수립하고 우리나라 전력수급기본계획의 특성을 고려한 중첩 불확실성하에서 원전감축 투자를 분석한다. 분석 결과, 현재의 원전축소 정책은 전력수요 증가율이 낮다는 것을 전제로 한 정책으로서 전력수요 증가에 대응할 수 있는 정책 강건성을 갖추지는 못한다는 것을 보여준다.

• 방사선산업학회지
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• 제18권1호
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• pp.35-42
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• 2024

Demand for RW transportation is expected to increase due to the continuous generation of RW from nuclear power plants and facilities, decommissioning of plants, and saturation of spent fuel temporary storage facilities. The locational aspect of plants and radiation protection optimization for the public have led to an increasing demand for maritime transportation, necessitating to apprehend the overseas and domestic current status. Given the potential long-term radiological impact on the public in the event of a sinking accident, a pre-transportation exposure assessment is necessary. The objective of this study is to investigate the overseas and domestic RW maritime transportation current status and overseas dose assessment cases for the public in sinking accident. Selected countries, including Japan, UK, Sweden, and Korea, were examined for transport cases, Japan and the U.S were chosen for dose assessment case in sinking accidents. As a result of the maritime transportation case analysis, it was performed between nuclear power plants and reprocessing facilities, from plants to disposal or intermediate storage facilities. HLW and MOX fuel were transported using INF 3 shipments, and all transports were performed low speed of 13 kn or less. As a result of the dose assessment for the public in sinking accident, japan conducted an assessment for the sinking of spent fuel and vitrified HLW, and the U.S conducted for the sinking of spent fuel. Both countries considered external exposure through swimming and working at seashore, and internal exposure through seafood ingestion as exposure pathway. Additionally, Japan considered external exposure through working on board and fishing, and the U.S considered internal exposure through spray inhalation and desalinized water and salt ingestion. Internal exposure through seafood ingestion had the largest dose contribution. The average public exposure dose was 20 years after the sinking, 0.04 mSv yr^-1 for spent fuel and 5 years after the sinking, 0.03 mSv yr^-1 for vitrified HLW in Japan. In the U.S, it was 1.81 mSv yr^-1 5 years after the sinking of spent fuel. The results of this study will be used as fundamental data for maritime transportation of domestic RW in the future.

• Nuclear Engineering and Technology
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• 제40권4호
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• pp.255-268
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• 2008

The greater demand for electricity and the available capacity within safety margins in some operating NPPs are prompting nuclear utilities to request license modification to enable operation at a higher power level, beyond their original license provisions. Such plant modifications require an in-depth safety analysis to evaluate the possible safety impact. The analysis must consider the thermo hydraulic, radiological and structural aspects, and the plant behavior, while taking into account the capability of the structures, systems and components, and the reactor protection and safeguard systems set points. The purpose of this paper is to introduce international experiences and approaches for implementation of power uprates related to the reactor thermal power of nuclear power plants. The paper is intended to give the reader a general overview of the major processes, work products, issues, challenges, events, and experiences in the power uprates program. The process of increasing the licensed power level of a nuclear power plants is called a power uprate. One way of increasing the thermal output from a reactor is to increase the amount of fissile material in use. It is also possible to increase the core power by increasing the performance of the high power bundles. Safety margins can be maintained by either using fuels with a higher performance, or through the use of improved methods of analysis to demonstrate that the required margins are retained even at the higher power levels. The paper will review all types of power uprates, from small to large, and across various reactor types, including light and heavy water, pressurized, and boiling water reactors. Generally, however, the content of the report focuses on power uprates of the stretch and extended type. The International Atomic Energy Agency (IAEA) is developing a technical guideline on power uprates and side effects of power uprates in nuclear power plants.