• Journal of Korea Society of Digital Industry and Information Management
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• v.18 no.2
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• pp.39-48
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• 2022

The application of digital technology to instrumentation and control systems in nuclear power plants has overcome many shortcomings of analog technology, but the threat of cybersecurity has increased. Along with other systems, the reactor protection system also uses digital-based equipment, so responding to cybersecurity threats is essential. We generally determine cybersecurity threats according to the role and function of the system. However, since the instrumentation and control system has various systems linked to each other, it is essential to analyze cybersecurity threats together between the connected systems. In this paper, we analyze the cybersecurity threat of the reactor protection system with the associated facilities. To this end, we quantitatively identified the risk of the reactor protection system by considering safety functions, a communication type, the use of analog or digital-based equipment of the associated systems, and the software vulnerability of the configuration module of the reactor protection system.

• Journal of the Korean Society of Safety
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• v.15 no.3
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• pp.66-70
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• 2000

In order to cope with environmental problems caused by harmful gases emitted from various industrial sources, a new technology which employs discharge plasma formed in ordinary atmospheric pressure has been intensively investigated in many industrialized nations. Although a plenty of useful outcomes and suggestions have been made public by scientists in this field, few commercial products which effectively decompose pollutant gases have appeared as yet. This is partly because that the energy efficiency of a most effective plasma reactor has not reached a satisfactory level in comparison with those of devices using conventional technologies. In an attempt to solve the problem mentioned above, we noticed to combine heterogeneous electrical discharges. This concepts is based on that each plasma reactor has its specific spatial region in which chemical reaction are active and by electrically affected with another reactor of different type, the activated region would increase - which may lead to cutting down the energy consumption. To prove this concept experimentally, two different discharge equipments, a plane ceramic-based surface discharge electrode and a corona electrode with tungsten needle may, are selected and combined to fabricate a hybrid plasma reactor. The results are summarized as follows; (1) Ozone concentration generated in the plasma region drastically increases when the positive corona discharge is added to the surface discharge. The rate of increase of ozone depends on the frequency of the surface discharge. The negative corona, however, does not contribute to the improvement of the ozone generation. (2) NO(nitrogen monoxide) decomposition rate also improves by simultaneously applying the surface and the positive corona discharges. The effect of the corona superposition is more evident when the level of the surface discharge is moderate. (3) By adjusting the corona level, the net energy efficiency during NO decomposition improves in comparison with the simple surface discharge reactor.

• Nuclear Engineering and Technology
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• v.54 no.7
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• pp.2395-2407
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• 2022

The sodium-cooled fast reactor is among the innovative nuclear technologies selected in the framework of the development of Generation IV concepts, allowing the irradiation of uranium-plutonium mixed oxide fuels (MOX). A fundamental step for the safety assessment of MOX-fuelled pins for fast reactor applications is the evaluation, by means of fuel performance codes, of the integral thermal-mechanical behaviour under irradiation, involving the fission gas behaviour and release in the fuel-cladding gap. This work is dedicated to the performance analysis of an inner-core fuel pin representative of the ASTRID sodium-cooled concept design, selected as case study for the benchmark between the GERMINAL and TRANSURANUS fuel performance codes. The focus is on fission gas-related mechanisms and integral outcomes as predicted by means of the SCIANTIX module (allowing the physics-based treatment of inert gas behaviour and release) coupled to both fuel performance codes. The benchmark activity involves the application of both GERMINAL and TRANSURANUS in their "pre-INSPYRE" versions, i.e., adopting the state-of-the-art recommended correlations available in the codes, compared with the "post-INSPYRE" code results, obtained by implementing novel models for MOX fuel properties and phenomena (SCIANTIX included) developed in the framework of the INSPYRE H2020 Project. The SCIANTIX modelling includes the consideration of burst releases of the fission gas stored at the grain boundaries occurring during power transients of shutdown and start-up, whose effect on a fast reactor fuel concept is analysed. A clear need to further extend and validate the SCIANTIX module for application to fast reactor MOX emerges from this work; nevertheless, the GERMINAL-TRANSURANUS benchmark on the ASTRID case study highlights the achieved code capabilities for fast reactor conditions and paves the way towards the proper application of fuel performance codes to safety evaluations on Generation IV reactor concepts.

• Nuclear Engineering and Technology
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• v.56 no.10
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• pp.4018-4030
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• 2024

The in-vessel retention through external reactor vessel cooling (IVR-ERVC) strategy is a key management strategy for early termination of a nuclear severe accident that can threaten the integrity of the reactor vessel. To simulate the physical phenomena of the molten corium, the smoothed particle hydrodynamic (SPH) method is utilized in this study. The SPH method is a Lagrangian computational fluid dynamic (CFD) method that can simulate multi-fluid stratification, turbulence, natural circulation, radiative heat transfer, thermal ablation, and crust formation. To address the external vessel cooling, it is coupled with a conventional 1-D nuclear system analysis method. The 1-D system analysis code can calculate the two-phase natural circulation of cooling water and the convective heat transfer on the external reactor vessel wall. These two simulation codes exchange the temperature and heat flux of the reactor vessel outer wall. This study numerically simulated the IVR-ERVC strategy for a Korean high-power reactor and compared it with the traditional lumped parameter method (LPM). Unlike LPM, this study provides localized detailed data about the thermal hydraulic behavior of molten corium and visualization of phenomena in the IVR-ERVC strategy. This enhances our understanding of the phenomena in IVR-ERVC strategy and introduces new perspectives.

• Nuclear industry
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• v.4 no.2 s.18
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• pp.12-19
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• 1984

• Nuclear Engineering and Technology
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• v.7 no.1
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• pp.49-52
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• 1975

• Nuclear Engineering and Technology
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• v.50 no.3
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• pp.396-410
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• 2018

To assist operators to properly assess the current situation of the plant, accurate fault diagnosis methodology should be available and used. A reliable fault diagnosis method is beneficial for the safety of nuclear power plants. The major idea proposed in this work is integrating the merits of different fault diagnosis methodologies to offset their obvious disadvantages and enhance the accuracy and credibility of on-line fault diagnosis. This methodology uses the principle component analysis-based model and multi-flow model to diagnose fault type. To ensure the accuracy of results from the multi-flow model, a mechanical simulation model is implemented to do the quantitative calculation. More significantly, mechanism simulation is implemented to provide training data with fault signatures. Furthermore, one of the distance formulas in similarity measurement-Mahalanobis distance-is applied for on-line failure degree evaluation. The performance of this methodology was evaluated by applying it to the reactor coolant system of a pressurized water reactor. The results of simulation analysis show the effectiveness and accuracy of this methodology, leading to better confidence of it being integrated as a part of the computerized operator support system to assist operators in decision-making.

• Nuclear Engineering and Technology
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• v.52 no.7
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• pp.1417-1428
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• 2020

The design of a nuclear reactor core requires basic thermal-hydraulic information concerning the heat transfer regime at which onset of nucleate boiling (ONB) will occur, the pressure drop and flow rate through the reactor core, the temperature and power distributions in the reactor core, the departure from nucleate boiling (DNB), the condition for onset of flow instability (OFI), in addition to, the critical velocity beyond which the fuel elements will collapse. These values depend on coolant velocity, fuel element geometry, inlet temperature, flow direction and water column above the top of the reactor core. Enough safety margins to ONB, DNB and OFI must-emphasized. A heat transfer package is used for calculating convection heat transfer coefficient in single phase turbulent, transition and laminar regimes. The main objective of this paper is to study the possibility of power upgrading of WWR-S research reactor from 2 to 10 MW_th. This study presents a one-dimensional mathematical model (axial direction) for steady-state thermal-hydraulic design and analysis of the upgraded WWR-S reactor in which two types of plate fuel elements are employed. FOR-CONV computer program is developed for the needs of the power upgrading of WWR-S reactor up to 10 MW_th.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.37 no.10
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• pp.961-969
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• 2013

In this study, the numerical analysis of a turbulent flow around both a staggered and an inline tube bundle was conducted using ANSYS CFX V.13, a commercial CFD software. The flow was assumed to be steady, incompressible, and isothermal. According to the CFD Best Practice Guideline, the sensitivity study for grid size, accuracy of the discretization scheme for convection term, and turbulence model was conducted, and its result was compared with the experimental data to estimate the applicability of the CFD Best Practice Guideline. It was concluded that the CFD Best Practice Guideline did not always guarantee an improvement in the prediction performance of the commercial CFD software in the field of tube bundle flow.

• Structural Engineering and Mechanics
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• v.50 no.6
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• pp.755-772
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• 2014

A simplistic approach towards evaluation of complete load deflection response of Reinforced Concrete (RC) flexural members under post fire (residual) scenario is presented in this paper. The cross-section of the RC flexural member is divided into a number of sectors. Thermal analysis is performed to determine the temperature distribution across the section, for given fire duration. Temperature-dependent stress-strain curves for concrete and steel are then utilized to perform a moment-curvature analysis. The moment-curvature relationships are obtained for beams exposed to different fire durations. These are then utilized to obtain the load-deflection plots following pushover analysis. Moreover one of the important issues of modeling the initial stiffness giving due consideration to stiffness degradation due to material degradation and thermal cracking has also been addressed in a rational manner. The approach is straightforward and can be easily programmed in spreadsheets. The presented approach has been validated against the experiments, available in literature, on RC beam subjected to different fire durations viz. 1hr, 1.5hrs and 2hrs. Complete load-deflection curves have been obtained and compared with experimentally reported counterparts. The results also show a good match with the results obtained using more complicated approaches such as those involving Finite element (FE) modeling and conducting a transient thermal stress analysis. Further evaluation of the beams during fire (at elevated temperatures) was performed and a comparison of the mechanical behavior of RC beams under post fire and during fire scenarios is made. Detailed formulations, assumptions and step by step approach are reported in the paper. Due to the simplicity and ease of implementation, this approach can be used for evaluation of global performance of fire affected structures.