• Proceedings of the Korean Institute of Building Construction Conference
• /
• 2012.11a
• /
• pp.229-231
• /
• 2012

The supply chain in plant construction project can be defined as a network combined between elements which is engineering, procurement, construction, start-up. For successfully leading construction project, close relation and mutual cooperation should be preceded and Supply Chain Management(SCM) system is needed to improve the relation. In particular, because safety is a top priority for the companies work in the construction of Nuclear Power Plant(NPP), it is necessary to control all parties realated with NPP construction systematically. In this situation, it is important to purchase high-quality equipment in timely manner by developing Integrated cooperative systems. Therefore, this study aimed to propose developing solutions of SCM such as JIT and Envelop Design system, which can reduces difficulties of decision making between the parties involved in the project and will increase the efficiency of the engineering process in NPP.

• Nuclear Engineering and Technology
• /
• v.52 no.1
• /
• pp.192-205
• /
• 2020

Seismic design practices and seismic response analyses of civil structures and nuclear power plants (NPPs) have conventionally used the peak ground acceleration (PGA) or spectral acceleration (S_a) as an intensity measure (IM) of an earthquake. However, there are many other earthquake IMs that were proposed by various researchers. The aim of this study is to investigate the correlation between seismic responses of NPP components and 23 earthquake IMs and identify the best IMs for correlating with damage of NPP structures. Particularly, low- and high-frequency ground motion records are separately accounted in correlation analyses. An advanced power reactor NPP in Korea, APR1400, is selected for numerical analyses where containment and auxiliary buildings are modeled using SAP2000. Floor displacements and accelerations are monitored for the non- and base-isolated NPP structures while shear deformations of the base isolator are additionally monitored for the base-isolated NPP. A series of Pearson's correlation coefficients are calculated to recognize the correlation between each of the 23 earthquake IMs and responses of NPP structures. The numerical results demonstrate that there is a significant difference in the correlation between earthquake IMs and seismic responses of non-isolated NPP structures considering low- and high-frequency ground motion groups. Meanwhile, a trivial discrepancy of the correlation is observed in the case of the base-isolated NPP subjected to the two groups of ground motions. Moreover, a selection of PGA or S_a for seismic response analyses of NPP structures in the high-frequency seismic regions may not be the best option. Additionally, a set of fragility curves are thereafter developed for the base-isolated NPP based on the shear deformation of lead rubber bearing (LRB) with respect to the strongly correlated IMs. The results reveal that the probability of damage to the structure is higher for low-frequency earthquakes compared with that of high-frequency ground motions.

• Nuclear Engineering and Technology
• /
• v.55 no.10
• /
• pp.3787-3800
• /
• 2023

Reactivity Initiated Accident (RIA) scenarios require special attention using advanced simulation techniques due to their complexity and importance for nuclear power plant (NPP) safety. While the conservative approach has traditionally been used for safety analysis, it may lead to unrealistic results which calls for the use of best estimate plus uncertainty (BEPU) approach, especially with the current advances in computational power which makes the BEPU analysis feasible. In this work an Uncontrolled Control Element Assembly (CEA) Withdrawal at Full Power accident scenario is analyzed using the BEPU approach by loosely coupling the thermal hydraulics best-estimate system code (RELAP5/SCDAPSIM/MOD3.4) to the statistical analysis software (DAKOTA) using a Python interface. Results from the BEPU analysis indicate that a realistic treatment of the accident scenario yields a larger safety margin and is therefore encouraged for accident analysis as it may enable more economic and flexible operation.

• Proceedings of the Korean Institute of Building Construction Conference
• /
• 2017.11a
• /
• pp.137-138
• /
• 2017

Generally, a lot of reinforcements are used in nuclear power plant concrete structures in order to improve the structural safety, but it may cause several potential problems due to the overcrowded reinforcement, such as the degradation of concrete quality, the construction delay and the increase of construction cost. In order to resolve these problems, structural test researches and code change studies on using high-strength reinforcement (Gr.80) in unclear power plant structures are under way, and there is good progress in code change of ASM BPVC.III.2 and ACI 349. This purpose of this study is to review the code change status ASM BPVC.III.2, ACI 349 under way to use the high-strength reinforcement in nuclear power plant structures. Also I will introduce the design optimization of NPP structures with high-strength reinforcements in order to maximize the effect and minimize the problem when using the high-strength reinforcements in NPP structures.

• Proceedings of the KIEE Conference
• /
• 2001.11c
• /
• pp.204-208
• /
• 2001

In this paper, we induce the linear transfer function of Downcomer water level of NPP(Nuclear Power Plant) Steam Generator using Genetic Algorithm and Levenberg-Marquardt Algorithm. The characteristic of NPP S/G mechanism is so high-non-linear that it is hard to achieve mathematical expression. So we use non-mathematical Algorithms to get the model function of NPP S/G water level. S/G level controller would be designed with this transfer function as the plant.

• Nuclear Engineering and Technology
• /
• v.53 no.9
• /
• pp.2888-2898
• /
• 2021

Vital area identification (VAI) is an essential procedure for the design of physical protection systems (PPSs) for nuclear power plants (NPPs). The purpose of PPS design is to protect vital areas. VAI has been improved continuously to overcome the shortcomings of previous VAI generations. In first-generation VAI, a sabotage fault tree was developed directly without reusing probabilistic safety assessment (PSA) results or information. In second-generation VAI, VAI model was constructed from all PSA event trees and fault trees. While in third-generation VAI, it was developed from the simplified PSA event trees and fault trees. While VAIs have been performed for NPPs in full-power operations, VAI for NPPs in low-power and shutdown (LPSD) operations has not been studied and performed, even though NPPs in LPSD operations are very vulnerable to sabotage due to the very crowded nature of NPP maintenance. This study is the first to research and apply VAI to LPSD operation of NPP. Here, the third-generation VAI method for full-power operation of NPP was adapted to the VAI of LPSD operation. In this study, LPSD VAI for a few plant operational states (POSs) was performed. Furthermore, the operation strategy of vital areas for both full-power and LPSD operations was discussed. The LPSD VAI method discussed in this paper can be easily applied to all POSs. The method and insights in this study can be important for future LPSD VAI that reflects various LPSD operational states. Regulatory bodies and electric utilities can take advantage of this LPSD VAI method.

• Journal of Radiation Protection and Research
• /
• v.47 no.4
• /
• pp.226-236
• /
• 2022

Background: An effective communication strategy for reducing conflicts in South Korea has been designed through the analysis of public perception and communication variables on nuclear power under the conditions of rapidly changing nuclear power policies. Materials and Methods: This study conducted both qualitative research through group discussions based on social psychology and quantitative research through surveys. Results and Discussion: Nuclear power plant (NPP) area residents in favor of nuclear power indicated higher levels of communication, safety perception, and contribution than those against it. NPP area residents trusted the civilian expert groups (18.3%) and local government (17.3%) the most, while metropolitan city residents trusted the Nuclear Safety and Security Commission and the Korea Institute of Nuclear Safety (20.7%) the most. In determining nuclear power policy, both the NPP area residents (18.1%) and metropolitan city residents (17.1%) prioritized safety, health, and the environment. While metropolitan city residents thought that energy security and economic growth (16.4%) were important, NPP area residents thought the current issue of spent fuel rods (14.1%) to be important. Conclusion: It is necessary for the nuclear power industry to have and actively implement communication and conflict resolution strategies based on the patterns obtained in the study results.

• Journal of the Korean Society of Systems Engineering
• /
• v.18 no.2
• /
• pp.58-74
• /
• 2022

Deterministic accident analysis plays a central role in the nuclear power plant (NPP) safety evaluation and licensing process. Traditionally the conservative approach opted for the point kinetics model, expressing the reactor core parameters in the form of reactivity and power tables. However, with the current advances in computational power, high fidelity multi-physics simulations using real-time code coupling, can provide more detailed core behavior and hence more realistic plant's response. This is particularly relevant for transients where the core is undergoing reactivity anomalies and uneven power distributions with strong feedback mechanisms, such as reactivity initiated accidents (RIAs). This work addresses a RIA, specifically a control element assembly (CEA) withdrawal at power, using the multi-physics analysis tool RELAP5/MOD 3.4/3DKIN. The thermal-hydraulics (TH) code, RELAP5, is internally coupled with the nodal kinetics (NK) code, 3DKIN, and both codes exchange relevant data to model the nuclear power plant (NPP) response as the CEA is withdrawn from the core. The coupled model is more representative of the complex interactions between the thermal-hydraulics and neutronics; therefore the results obtained using a multi-physics simulation provide a larger safety margin and hence more operational flexibility compared to those of the point kinetics model reported in the safety analysis report for APR1400. The systems engineering approach is used to guide the development of the work ensuring a systematic and more efficient execution.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1996.05a
• /
• pp.340-345
• /
• 1996

Because of limited resources of time and information processing capability during abnormal situation, diagnosis is difficult tasks in nuclear power plant (NPP) operators. Moreover since minimizing of adverse consequences according to process abnormalities is vital for the safety of NPP, introducing of diagnosis support systems have particularly emphasized. However, considerable works to develop effective diagnostic support system are not sufficiently fulfilled because of the complexity of NPP is one of the major problems. To cope with this complexity, a lot of model-based diagnosis support systems have considered and implemented worldwide. In this paper, as a prior step to development of model-based diagnosis support systems, primary side of pressurized water reactor is functionally modeled by multilevel flow modeling (MFM) concept. MFM is suitable for complex system modeling and for diagnosis of abnormalities. Furthermore, knowledge-based diagnosis process, of NPP operator could be supported because this diagnosis strategy can represent operator's one.

• Nuclear Engineering and Technology
• /
• v.51 no.6
• /
• pp.1669-1680
• /
• 2019

Seismic safety is considered to be one of the key design objectives of AP1000 nuclear power plant (NPP) in strong earthquakes. Dynamic behavior, damage development and aggravation effect are studied in this study for the three main components of AP1000 NPP, namely reinforced concrete shield building (RCSB), steel vessel containment (SVC) and reinforced concrete auxiliary building (RCAB). Characteristics including nonlinear concrete tension and compressive constitutions with plastic damage are employed to establish the numerical model, which is further validated by existing studies. The author investigates three earthquakes and eight input levels with the maximum magnitude of 2.4 g and the results show that the concrete material of both RCSB and RCAB have suffered serious damage in intense earthquakes. Considering RCAB in the whole NPP, significant damage aggravation effect can be detected, which is mainly concentrated at the upper intersection between RCSB and RCAB. SVC and reinforcing bar demonstrate excellent seismic performance with no obvious damage.