• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.820-821
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 1998.10a
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• pp.105-105
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• 1998

• Journal of the Korea institute for structural maintenance and inspection
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• v.27 no.5
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• pp.16-22
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• 2023

In this study, a salt water immersion test was performed on concrete specimens simulating the concrete mix design of the nuclear power plant, and the correlation between the amount of chloride and the XRF component according to the depth of the concrete was analyzed. The amount of chloride on the surface of the nuclear power plant concrete increased slightly with increasing immersion time in salt water, but the amount of chloride in the depth of 5.5 mm or more showed a clear tendency to increase with increasing immersion time in salt water. As a result of analyzing the correlation between the amount of chloride in concrete and the XRF component, the concrete with 20% FA substitution compared with the OPC concrete showed a very high correlation between the composition ratio of Cl ions and the evaluation result of salt damage resistance by XRF component analysis. Accordingly, it was confirmed that chlorine ion analysis and salt damage resistance performance evaluation by XRF component analysis were possible through repeated data accumulation in the nuclear power plant concrete mix with 20% fly ash replacement.

• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1890-1901
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• 2022

We have assessed the applicability of the thermal-hydraulic system analysis code, SPACE, to a small modular reactor called SMART. For the assessment, the experimental data from a scale-down integral-test facility, SMART-ITL, were used. It was conformed that the SPACE code unrealistically calculates the safety injection flow rate through the CMT and SIT during a small-break loss-of-coolant experiment. This unrealistic behavior was due to the overprediction of interfacial heat transfer at the steam-water interface in a vertically stratified flow in the tanks. In this study, a special thermal-hydraulic component model has been developed to realistically calculate the interfacial heat transfer when a strong non-equilibrium two-phase flow is formed in the CMT or SIT. Additionally, we developed a special heat structure model, which analytically calculates the heat transfer from the hot steam to the cold tank wall. The combination of two models for the tank are called the special component model. We assessed it using the SMART-ITL passive safety injection system (PSIS) test data. The results showed that the special component model well predicts the transient behaviors of the CMT and SIT.

• Nuclear Engineering and Technology
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• v.53 no.10
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• pp.3344-3351
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• 2021

The Rare Isotope Science Project was launched in 2011 in Korea toward constructing the Rare isotope Accelerator complex for ON line experiments (RAON). RAON will house several experimental systems, including the Muon Spin Rotation/Relaxation/Resonance (μSR) facility in High Energy Experimental Building B. This facility will use 600-MeV protons with a maximum current of 660 pμA and beam power of 400 kW. The key μSR features will facilitate projects related to condensed-matter and nuclear physics. Typical experiments require a few million surface muons fully spin-polarized opposite to their momentum for application to small samples. Here, we describe the design of a muon transport beam line for delivering the requisite muon numbers and the electromagnetic-component specifications in the μSR facility. We determine the beam-line configuration via beam-optics calculations and the transmission efficiency via single-particle tracking simulations. The electromagnet properties, including fringe field effects, are applied for each component in the calculations. The designed surface-muon beamline is 17.3 m long, consisting of 2 solenoids, 2 dipoles affording 70° deflection, 9 quadrupoles, and a Wien filter to eliminate contaminant positrons. The average incident-muon flux and spin rotation angle are estimated as 5.2 × 10⁶ μ⁺/s and 45°, respectively.

• Journal of Software Engineering Society
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• v.25 no.1
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• pp.11-18
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• 2012

Characteristic of decommissioning target facility investigate and understand is very important. because radioactive materials occurs in the decommissioning and dismantling, so it is difficult to use a general dismantling method. Decommissioning nuclear facilities, the characteristics of the target of research to predict the amount of decommissioning waste, decommission projects costing is largely utilized. For this purpose, we developed DEFACS(Decommissioning Facility Characterization DB System) that manage characteristic of decommissioning target facility. But nuclear facility decommissioning takes long time. so we inevitably developed system during decommissioning works, it occurs many system changes. For this reason, it is difficult to apply general development process, so we take CBD process that divide CD(Component Development) and CBSD(Component Based Software Development) for handling change of requirement. it make Component of the overall system for changes to minimize changes by strengthening the independence of components and processes due to changes in requirements were to minimize stopping of the process.

• Proceedings of the KSME Conference
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• 2004.04a
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• pp.264-267
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• 2004

Wall thinning of carbon steel pipe components due to Flow-Accelerated Corrosion (FAC) is one of the most serious threats to the integrity of steam cycle piping systems in Nuclear Power Plants (NPP). Since the mid-1990s, secondary side piping systems in Korean NPPs have experienced wall thinning, leakages and ruptures caused by FAC. Korea Electric power Research Institute (KEPRI) and Korea Hydro & Nuclear Power Co., LTD. (KHNP) have conducted a study to develop the methodology for systematic pipe management and established the Korean Thinned Pipe Management Program (TPMP). To effectively maintain the integrity of piping system, FAC engineer should understand the criterions of the structural integrity evaluation and the safety margin assessment for the thinned pipe component. This paper describes the technical items of TPMP, and shows the example of the integrity evaluation and safety margin assessment for three thinned pipe component of a NPP.

• Nuclear Engineering and Technology
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• v.32 no.5
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• pp.485-494
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• 2000

In this work, a fuzzy neural network is used to estimate the relevant sensor signal using other sensor signals. Noise components in input signals into the fuzzy neural network are removed through the wavelet denoising technique . Principal component analysis (PCA) is used to reduce the dimension of an input space without losing a significant amount of information. A lower dimensional input space will also usually reduce the time necessary to train a fuzzy-neural network. Also, the principal component analysis makes easy the selection of the input signals into the fuzzy neural network. The fuzzy neural network parameters are optimized by two learning methods. A genetic algorithm is used to optimize the antecedent parameters of the fuzzy neural network and a least-squares algorithm is used to solve the consequent parameters. The proposed algorithm was verified through the application to the pressurizer water level and the hot-leg flowrate measurements in pressurized water reactors.

• Nuclear Engineering and Technology
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• v.39 no.2
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• pp.103-110
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• 2007

The High Temperature Gas-cooled Reactor (HTGR) possesses inherent safety features and is recognized as a representative advanced nuclear system for the future. Based on the success of the HTR-10, the long-time operation test and safety demonstration tests were carried out. The long-time operation test verifies that the operation procedure and control method are appropriate for the HTR-10 and the safety demonstration test shows that the HTR-10 possesses inherent safety features with a great margin. Meanwhile, two new projects have been recently launched to further develop HTGR technology. One is a prototype modular plant, denoted as HTR-PM, to demonstrate the commercial capability of the HTGR power plant. The HTR-PM is designed as $2{\times}250$ MWt, pebble bed core with a steam turbine generator that serves as an energy conversion system. The other is a gas turbine generator system coupled with the HTR-10, denoted as HTR-10GT, built to demonstrate the feasibility of the HTGR gas turbine technology. The gas turbine generator system is designed in a single shaft configuration supported by active magnetic bearings (AMB). The HTR-10GT project is now in the stage of engineering design and component fabrication. R&D on the helium turbocompressor, a key component, and the key technology of AMB are in progress.

• Nuclear Engineering and Technology
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• v.44 no.3
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• pp.297-310
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• 2012

Aging-related degradation of nuclear power plant components is an important aspect to consider in securing the long term safety of the plant, especially the seismic safety, since the degradation of the components affects not only their seismic capacity but their response. This can cause a change in the seismic margin of a component and the overall seismic safety of a system. To better understand the status and characteristics of degradation of components in Nuclear Power Plants (NPPs), the degradation occurrences of components in the U.S. NPPs were identified by reviewing recent publicly available information sources and the characteristics of these occurrences were evaluated and compared to observations from the past. Ten categories of components that are of high risk significance in Korean NPPs were identified, comprising anchorage, concrete, containment, exchanger, filter, piping systems, reactor pressure vessels, structural steel, tanks, and vessels. Software tools were developed to expedite the review process. Results from this review effort were compared to previous data in the literature to characterize the overall degradation trends.