• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.18 no.2
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• pp.54-60
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• 2022

ASME BPVC provides stress evaluation rules for Class 2 and 3 nuclear piping. However, such rules are difficult to be applied to reinforced wall-thinned pipes during service. To resolve this issue, a new method for stress evaluation of reinforced wall-thinned pipes is proposed in this work, based on the equivalent stiffness concept. By converting a reinforced wall-thinned pipe to an equivalent straight pipe having the same stiffness, stress evaluation can be proceeded using the current ASME BPVC rules. The proposed method is applied to pipes with 4 different normal pipe size and the effects of reinforcement and wall-thinning dimensions on evaluated stresses are discussed.

• Proceedings of the Korea Water Resources Association Conference
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• 2022.05a
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• pp.191-191
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• 2022

우리나라 강우는 대부분 여름에 편중되어 수자원 이용의 불확실성이 크기 때문에 댐과 같은 수공구조물을 활용하여 대응하고 있다. 기존 수도권의 용수공급은 한강수계 다목적댐들을 활용하여 충족시켰으나 기후변화에 따른 이상가뭄, 용수 수요 증가 등에 의해 미래 용수부족이 전망된다. 수도권의 안정적인 용수공급을 위해 환경부와 한국수력원자력(주)는 2020년 4월에 우리나라 최초로 기 건설된 발전용댐을 다목적화하는 『한강수계 발전용댐 다목적 활용 협약』을 체결하였다. 협약에 따라 화천댐은 시범운영기간 동안 기본계획 공급량 22.2 m³/sec (연간 7억 톤)를 상시 공급하고 있다. 따라서, 본 연구에서는 팔당댐에 기여하는 화천댐의 용수공급량을 산정하여 용수공급 효과를 검증하였다. 화천댐 용수공급 기여도를 정량적으로 제시하기 위해 HEC-ResSim을 활용하여 시범운영기간 운영 결과와 유사한 한강수계 댐 운영 모형을 구축하였다. 또한, 시범운영 기간 화천댐과 다목적댐들의 팔당댐 용수공급 기여도를 정량적으로 제시하는 방안을 구축하였다. 시범운영이 수행됨에 따라 발전용댐의 다목적 활용 효과를 정량적으로 평가하는 방안 구축으로발전용댐의 용수공급 효과를 검증할 수 있을 것으로 기대된다.

• Corrosion Science and Technology
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• v.18 no.1
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• pp.33-38
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• 2019

Alloy 600/182 with excellent mechanical/chemical properties have been utilized for nuclear power plants. Although both alloys are known to have superior corrosion resistance, stress corrosion cracking failure has been an issue in primary water environment of nuclear power plants. Therefore, primary water stress corrosion crack (PWSCC) growth rate tests were conducted to investigate crack growth properties of Alloy 600/182. To investigate PWSCC growth rate, test facilities including water chemistry loop, autoclave, and loading system were constructed. In PWSCC crack growth rate tests, half compact-tension specimens were manufactured. These specimens were then placed inside of the autoclave connected to the loop to provide primary water environment. Tested conditions were set at temperature of $360^{\circ}C$ and pressure of 20MPa. Real time crack growth rates of specimens inside the autoclave were measured by Direct Current potential drop (DCPD) method. To confirm inter-granular (IG) crack as a characteristic of PWSCC, fracture surfaces of tested specimens were observed by SEM. Finally, crack growth rate was derived in a specific stress intensity factor (K) range and similarity with overseas database was identified.

• Proceedings of the Korean Institute of Information and Commucation Sciences Conference
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• 2019.05a
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• pp.464-466
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• 2019

Controllers used in nuclear power plants require high reliability. A controller including a Field Programmable Gate Array (FPGA) and a Complex Programmable Logic Device (referred to hereinafter as FPGA) has been applied to many Nuclear Power Plants (NPP) in the past, including the APR1400 (Advanced Power Reactor 1400), a Korean digital nuclear power plant. Initially, the FPGA was considered as a general IC (Integrated Circuit) and verified only by device verification and performance testing. In the 1990s, research on FPGA verification began, and until the FPGA became a chip, it was regarded as software and the software Verification and Validation (V&V) using IEEE 1012-2004 was implemented. Currently, IEC 62566, which is a European standard, has been applied for a lot of verification. This method has been evaluated as the most sensible method to date. This is because the method of verifying the characteristics of SoC (System on Chip), which has been a problem in the existing verification method, is sufficiently applied. However, IEC 62566 is a European standard that has not yet been adopted in the United States and maintains the application of IEEE 1012 for FPGA. IEEE 1012-2004 or IEC 62566 is a technical standard. In practice, various methods are applied to meet technical standards. In this paper, we describe the procedure and important points of verification method of Nuclear Safety Class FPGA applying SoC verification method.

• The Journal of the Korea institute of electronic communication sciences
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• v.7 no.2
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• pp.279-285
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• 2012

Most national critical key infrastructure, such like electricity, nuclear power plant, and petroleum is run on SCADA (Supervisory Control And Data Acquisition) system as the closed network type. These systems have treated the open protocols like TCP/IP, and the commercial operating system, which due to gradually increasing dependence on IT(Information Technology) is a trend. Recently, concerns have been raised about the possibility of these facilities being attacked by cyber terrorists, hacking, or viruses. In this paper, the method to minimize threats and vulnerabilities is proposed, with the virtual honeynet system architecture and the attack detection algorithm, which can detect the unknown attack patterns of Zero-Day Attack are reviewed.

• Nuclear Engineering and Technology
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• v.27 no.5
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• pp.710-720
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• 1995

The propagation of pump-induced pressure pulsation in a reactor is important because of the potential for vibration and resultant damage of reactor internals. A hydrodynamic model has been developed to obtain the pressure fluctuation due to the operation of pumps in the annulus(between the core support barrel and reactor vessel of a pressurized water reactor) including the coolant inlet pipe. The mathematical analysis is formulated in accordance with the linearized Navier-Stokes equation by assuming a compressible, inviscid flow. Two regions are considered separately and by coupling the solutions of the inlet pipe and the annulus, the inlet nozzle pressure(pressure at pipe and annulus interface) is to be calculated without assumptions. The geometric parameter effect on the pump-induced pressure pulsation is evaluated. Comparison of predicted and measured inlet nozzle pressure values for each forcing frequency shows good order of magnitude agreement.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.06a
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• pp.457-463
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• 2005

KHNP(Korea Hydro and Nuclear Power Ltd., Co.) is developing a HIC transport package which is satisfying domestic and IAEA regulations and NETEC(Nuclear Environment Technology Institute) is conducting a conceptual design. In this study, the shielding thickness was calculated using the data from radionuclide assay program which is currently using in nuclear sites and Micro Shield code. Considering the structural safety, carbon steel was chosen as shielding material and the shielding thickness was calculated for 500 R/hr and 100 R/hr at HIC surface, respectively. Through the shielding analysis, it was evaluated that the regulation limit is satisfied when the shielding thickness is 22 cm for 500 R/hr and 17 cm for 100/hr.

• Journal of Energy Engineering
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• v.28 no.3
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• pp.26-33
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• 2019

A comparison of CL 13% and 17% IBLOCA counterpart tests(CPTs) between the ATLAS and LSTF facilities was carried out and the behavior of peak cladding temperatures(PCTs) and related thermal hydraulic phenomena were investigated and discussed. There appeared quite a big difference in PCT behavior between the two CPTs and a further comparison of reactor coolant system design between the two facilities was performed. As a result, there was a difference in fuel alignment plate (FAP) design, e.g., one FAP in ATLAS, a combination of upper core plate and upper end box in LSTF, respectively. The FAP design mainly affects the reflux condensate behavior in IBLOCA tests and any difference in FAP design can be a possible reason for different PCT behavior between the two facilities. It should be a further study to find the reason of different PCT behvior between the two facilites.

• Journal of Korea Water Resources Association
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• v.55 no.7
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• pp.515-529
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• 2022

Identifying the available water resources amount is an essential process in establishing a sustainable water resources management plan. Dam facility is a major infrastructure storing and supplying water during the dry season, and the water supply yield of the dam varies depending on dam inflow conditions or operation rule. In South Korea, water supply yield of dam is calculated by reservoir simulation based on observed historical dam inflow data. However, the water supply capacity of a dam can be underestimated or overestimated depending on the existence of historical drought events during the simulation period. In this study, probabilistic inflow data was generated and used to estimate the appropriate range of the water supply yield of hydropower dams. That is, a method for estimating the probabilistic dam inflow that fluctuates according to climatic and socio-economic conditions and the range of water supply yield for hydropower dams was presented, and applied to hydropower dams located in the Han river in South Korea. It is expected that the understanding water supply yield of the hydropower dams will become more important to respond to climate change in the future, and this study will contribute to national water resources management planning by providing potential range of water supply yield of hydropower dams.

• KSCE Journal of Civil and Environmental Engineering Research
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• v.44 no.5
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• pp.649-661
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• 2024

Recently, the need for water supply from hydropower dams has been increasing due to drought. In order to supply water through hydropower dam, the inflow into hydropower dam should be evaluated first. Some of the hydropower dams in the Han River basin are located downstream of multipurpose dam, so it is important to analyze its own inflow of the hydropower dam. The purpose of this study is to evaluate the contribution of inflow by water source to the hydropower dam located in the Han River basin. Water use-related data provided by various domestic institutions were investigated and collected, and a location-based water supply and demand network was constructed. Unlike the existing domestic water balance analysis method, the simulation was conducted in consideration of the amount of transmission loss. The applicability of the analysis method was confirmed through the results of the fitness evaluation (NSE 0.95~0.99 and correlation coefficient 0.98~0.99) comparing the simulated flow with the observed flow at the representative point. Based on the water intake method of the facility and the release method of the remaining multipurpose dam water, a water balance analysis was performed assuming four cases, and the contribution of inflow by water sources into each hydropower dam was evaluated and presented. The research results are expected to provide various information for evaluating the water supply capacity of hydropower dams in the future.