• Nuclear Engineering and Technology
• /
• v.54 no.7
• /
• pp.2453-2466
• /
• 2022

Nuclear power plant operators in the main control room are exposed to stressful conditions in emergency situations as immediate and appropriate mitigations are required. While emergency operating procedures (EOPs) provide operators with the appropriate tasks and diagnostic guidelines, EOPs have static properties that make it difficult to reflect the dynamic changes of the plant. Due to this static nature, operator workloads increase because unrelated information must be screened out and numerous displays must be checked to obtain the plant status. Generally, excessive workloads should be reduced because they can lead to human errors that may adversely affect nuclear power plant safety. This paper presents a framework for an operator support system that can substitute the initial responses of the EOPs, or in other words the immediate actions and diagnostic procedures, in the early stages of an emergency. The system assists operators in emergency operations as follows: performing the monitoring tasks in parallel, identifying current risk and latent risk causality, diagnosing the accident, and displaying all information intuitively with a master logic diagram. The risk causalities are analyzed with a functional modeling methodology called multilevel flow modeling. This system is expected to reduce workloads and the time for performing initial emergency response procedures.

• Korean Journal of Digital Imaging in Medicine
• /
• v.10 no.1
• /
• pp.5-6
• /
• 2008

In this study, using an experiment, in certified an error between each system of SUV (standardized uptake value) that is one of the main analyses to diagnose a tumor in PET/CT, so, it would assure reliability and help to diagnose any lesions accurately. That is, a detailed analysis progressed. as all images reconstructed every setting time, then, clinical reliability between each system was expressed numerically at MBq/mL and SUV. Therefore, this study is considered that flexibility of follow-up using diverse system was insured, and it helps to offer accurate and beneficial information for diagnosis of various fields.

• Nuclear Engineering and Technology
• /
• v.32 no.2
• /
• pp.157-168
• /
• 2000

The nuclear steam generator level control system is designed by robust control methods. The feedwater controller is designed by three methods of the H$\infty$, the mixed weight sensitivity and the structured singular value. Then the controller located on the feedback loop of the level control system is designed. For the system performance, the controller of simple PID whose coefficients vary with the power is selected. The simulations show that the system has a good performance with proper stability margins.

• Nuclear Engineering and Technology
• /
• v.47 no.3
• /
• pp.293-305
• /
• 2015

In this paper, we suggest the conceptual design of a water-cooled reactor system for a low-pressure inherent heat sink nuclear desalination plant (LIND) that applies the safety-related design concepts of high temperature gas-cooled reactors to a water-cooled reactor for inherent and passive safety features. Through a scoping analysis, we found that the current LIND design satisfied several essential thermal-hydraulic and neutronic design requirements. In a thermal-hydraulic analysis using an analytical method based on the Wooton-Epstein correlation, we checked the possibility of safely removing decay heat through the steel containment even if all the active safety systems failed. In a neutronic analysis using the Monte Carlo N-particle transport code, we estimated a cycle length of approximately 6 years under 200 $MW_{th}$ and 4.5% enrichment. The very long cycle length and simple safety features minimize the burdens from the operation, maintenance, and spent-fuel management, with a positive impact on the economic feasibility. Finally, because a nuclear reactor should not be directly coupled to a desalination system to prevent the leakage of radioactive material into the desalinated water, three types of intermediate systems were studied: a steam producing system, a hot water system, and an organic Rankine cycle system.

• Nuclear Engineering and Technology
• /
• v.50 no.8
• /
• pp.1314-1323
• /
• 2018

For nuclear power plants (NPPs) to have long lifetimes, ageing is a major issue. Currently, ageing management for NPP systems is based on correlations built from generic experimental data. However, each system has its own characteristics, operational history, and environment. To account for this, it is possible to resort to prognostics that predicts the future state and time to failure (TTF) of the target system by updating the generic correlation with specific information of the target system. In this paper, we present an application of particle filtering for the prediction of degradation in steam generator tubes. With a case study, we also show how the prediction results vary depending on the uncertainty of the measurement data.

• Nuclear Engineering and Technology
• /
• v.48 no.6
• /
• pp.1321-1329
• /
• 2016

To get an insight into the operating characteristics of the passive residual heat removal system of molten salt reactors, a two-phase natural circulation test facility was constructed. The system consists of a boiling loop absorbing the heat from the drain tank, a condensing loop consuming the heat, and a steam drum. A steady-state experiment was carried out, in which the thimble temperature ranged from $450^{\circ}C$ to $700^{\circ}C$ and the system pressure was controlled at levels below 150 kPa. When reaching a steady state, the system was operated under saturated conditions. Some important parameters, including heat power, system resistance, and water level in the steam drum and water tank were investigated. The experimental results showed that the natural circulation system is feasible in removing the decay heat, even though some fluctuations may occur in the operation. The uneven temperature distribution in the water tank may be inevitable because convection occurs on the outside of the condensing tube besides boiling with decreasing the decay power. The instabilities in the natural circulation loop are sensitive to heat flux and system resistance rather than the water level in the steam drum and water tank. RELAP5 code shows reasonable results compared with experimental data.

• Journal of the Korean Society of Safety
• /
• v.34 no.5
• /
• pp.159-166
• /
• 2019

The seismic PSA is to probabilistically estimate the potential damage that a large earthquake will cause to a nuclear power plant. It integrates the probabilistic seismic hazard analysis, seismic fragility analysis, and system analysis and is utilized to identify seismic vulnerability and improve seismic capacity of nuclear power plants. Recently, the seismic risk of domestic multi-unit nuclear power plant sites has been evaluated after the Great East Japan Earthquake and Gyeongju Earthquake in Korea. However, while the currently available methods for system analysis can derive basic required results of seismic PSA, they do not provide the detailed results required for the efficient improvement of seismic capacity. Therefore, for in-depth seismic risk evaluation, improved system analysis method for seismic PSA has become necessary. This study develops a system analysis method that is not only suitable for multi-unit seismic PSA but also provides risk information for the seismic capacity improvements. It will also contribute to the enhancement of the safety of nuclear power plants by identifying the seismic vulnerability using the detailed results of seismic PSA. In addition, this system analysis method can be applied to other external event PSAs, such as fire PSA and tsunami PSA, which require similar analysis.

• Journal of the Korean Society of Safety
• /
• v.25 no.2
• /
• pp.24-28
• /
• 2010

The nuclear power Plant onsite AC electrical power sources are required to supply power to the engineering safety facility buses if the offsite power source is lost. Typically, Diesel Generators are used as the onsite power source. The 125 VAC buses are part of the onsite Class 1E AC and DC electrical power distribution system. The DC power distribution system ensure the availability of DC electrical power for system required to shutdown the reactor and maintain it in a safety condition after an anticipated operational occurrence or a postulated Design Base Accident. Recently, onsite DC power supply system trip occurs the loss of system function. To obtain the performance such as reliability and availability, we analyzed the cause of battery charger trip and described the improvement of DC power supply system reliability. Finally, we provide reliability performance criteria of charger in order to ensure the probabilistic goals for the safety of the nuclear power plants.

• Nuclear Engineering and Technology
• /
• v.37 no.1
• /
• pp.109-117
• /
• 2005

Limiting conditions for operations (LCOs) are evaluated dynamically using the tool of system dynamics. The LCOs de-fine the allowed outage times (AOTs) and the actions to be taken if the repair cannot be completed within the AOT. System dynamics has been developed to analyze the dynamic reliability of a complicated system. System dynamics using Vensim software have been applied to LCOs assessment for an example system, the auxiliary feed water system of a reference nuclear power plant. Analysis results of both full power operation and shutdown operation have been compared for a measure of core damage frequency. The framework developed in this study has been shown to be very flexible in that it can be applied to assess LCOs quantitatively under any operational context of the TS in FSAR.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1998.05a
• /
• pp.315-320
• /
• 1998

In this work, an automatic software verification method for Nuclear Power Plant (NPP) protection system is developed. This method utilizes Colored Petri net (CPN) for modeling and Prototype Verification system (PVS) for mathematical verification. In order to help flow-through from modeling by CPN to mathematical proof by PVS, a translator has been developed in this work. The combined method has been applied to a protection system function of Wolsong NPP SDS2(Steam Generator Low Level Trip)and found to be promising for further research and applications.