• Nuclear Engineering and Technology
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• v.46 no.6
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• pp.825-836
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• 2014

Predicting radioactive fission product (FP) behaviors in the reactor coolant system and the containment of a nuclear power plant (NPP) is one of the major concerns in the field of reactor safety, since the amount of radioactive FP released into the environment during the postulated accident sequences is one of the major regulatory issues. Radioactive FPs circulating in the primary coolant loop and released into the containment are basically in the form of gas or aerosol. In this study, a multi-component and multi-sectional analysis module for aerosol fission products has been developed based on the MAEROS model [1,2], and the aerosol transport model has been developed and verified against an analytic solution. The deposition of aerosol FPs to the surrounding structural surfaces is modeled with recent research achievements. The developed aerosol analysis model has been successfully validated against the STORM SR-11 experimental data [3], which is International Standard Problem No. 40. Future studies include the development of the resuspension, growth, and chemical reaction models of aerosol fission products.

• 유체기계공업학회:학술대회논문집
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• 1999.12a
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• pp.245-251
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• 1999

The structural analysis of a reactor coolant pump(RCP) of a nuclear power plant is very important for the safety assessment of the plant. Accurate boundary conditions for the heat transfer coefficient are required for reliable thermal stress analysis of the pump casing, especially in transient operations of the pump since the coolant properties are largely dependent on operational conditions. In the present study, a 3D mixed flow type coolant pump was modeled from the RCP drawings and analyzed in the steady state and number of transient flow conditions by using a commercial code STAR-CD. From the result of the computation, it is seem that the average heat transfer coefficients for the cases considered are found to be the suggested values of the manufacturer, Westinghouse Energy System. The unevenness in local heat transfer coefficients, however, is found to be considerable so that the use of average heat transfer coefficients in all boundaries might not give reliable thermal stresses.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.14 no.2
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• pp.16-23
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• 2018

In order to investigate the ways to enhance the seismic performance of APR1400 seismic fragile equipment by direct design changes, four equipment such as Reactor Vessel Support, Integrated Head Assembly, Remote Shutdown Console, and Pressurizer are reviewed using information of the main dimensions, seismic stress evaluation results, design FRS, etc. in this paper. In addition to the direct reinforcement of equipments, the feasibility of seismic isolation for the safety related cabinet is also investigated and the actual adaption plan of a commercial spring-damper system is briefly reviewed.

• Nuclear Engineering and Technology
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• v.53 no.1
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• pp.322-336
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• 2021

This paper presents a numerical prediction of the transient hydraulic loads acting on the tubes and external supports of a pressurized water reactor (PWR) steam generator (SG) during blowdown following a sudden feedwater line break (FWLB). A simplified SG model was used to easily demonstrate the prediction. The blowdown discharge flow was treated as a flashing flow to realistically simulate the transient flow fields inside the SG and the connected broken feedwater pipe. The effects of the SG initial pressure or the broken feedwater pipe length on the intensities or magnitudes of transient hydraulic loads were investigated. Then predictions of the decompression pressure wave-induced impulsive pressure differential loads on SG tubes and the transient blowdown loads on SG external supports were demonstrated and the general aspects of transient responses of such transient hydraulic loads to the FWLB were discussed.

• The Journal of the Acoustical Society of Korea
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• v.16 no.5
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• pp.83-91
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• 1997

A reactor vibration monitoring has been performed using neutron noise obtained from excore detectors for the safety operation, Traditionally, the spectral estimator based on Fourier analysis has been widely used in the noise analysis of the reactor system. If the bias is too severe, the resolution would not be adequate for a given application. One major motivation for the current interests in the parametric approach to spectral estimation is the apparent higher resolution achievable with these modern techniques. In considering an unbias, a consistency, an efficency, and a minimum lower bound of the statictic estimation, an AR model is appropriate for noise spectral estimation with sharp peaks but not deep valley. In order to select an appropriate model order, the lag value of autocorrleaton function is applied. Burg method to trace the vibration mode of RPV internal is the most sucuessful.

• Nuclear Engineering and Technology
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• v.55 no.11
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• pp.4295-4306
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• 2023

The vibration of fuel rods in axial flow is a universally recognized issue within both engineering and academic communities due to its significant importance in ensuring structural safety. This paper aims to thoroughly investigate the stability and nonlinear vibration of a fuel rod subjected to axial flow in a newly designed high temperature gas cooled reactor. Considering the possible presence of thermal expansion and large deformation in practical scenarios, the thermal effect and geometric nonlinearity are modeled using the von Karman equation. By applying Hamilton's principle, we derive the comprehensive governing equation for this fluid-structure interaction system, which incorporates the quadratic nonlinear stiffness. To establish a connection between the fluid and structure aspects, we utilize the Galerkin method to solve the perturbation potential function, while employing mode expansion techniques associated with the structural analysis. Following convergence and validation analyses, we examine the stability of the structure under various conditions in detail, and also investigate the bifurcation behavior concerning the buckling amplitude and flow velocity. The findings from this research enhance the understanding of the underlying physics governing fuel rod behavior in axial flow under severe yet practical conditions, while providing valuable guidance for reactor design.

• Proceedings of the KIEE Conference
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• 2005.05a
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• pp.203-205
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• 2005

A reactor protection system (RPS) plays the roles of generating the reactor trip signal and the engineered safety features (ESF) actuation signal when the monitored plant processes reach predefined limits. A Korean project group is developing a new digitalized RPS and the Cabinet Operator Module (COM) of the RPS which is used for the RPS integrity testing and monitoring by an equipment operator. A flat panel display (FPD) with a touch screen capability is provided as a main user interface for the RPS operation. To support the RPS COM user interface design, actually the FPD screen design, we developed a user interface style guide because the system designer could not properly deal with the many general human factors design guidelines. To develop the user interface style guide, various design guideline gatherings, a walk-though with a video recorder, guideline selection with respect to user interface design elements, determination of the properties of the design elements, discussion with the system designers, and a conversion of the properties into a screen design were carried out. This paper describes the process in detail and the findings in the course of the style guide development.

• Journal of Korean Society on Water Environment
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• v.26 no.1
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• pp.140-147
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• 2010

In existing design method for aeration tank water temperature was considered as governing variable for applying safety factor. This study tried a few new approach of aeration tank design using SOUR at various temperature conditions. Specific substrate utilization rate (U) and specific oxygen uptake rate (SOUR) both were analyzed at various temperature and SRT. The laboratory scale reactor was operated on various temperature ($10^{\circ}C$, $20^{\circ}C$, $25^{\circ}C$) and SRT (5day, 10day, 20day, 30day). In this study, SOUR tended to increase with the temperature increased. On the other hand, SOUR tended to decrease when SRT increased from 5 days to 30 days. Empirical equations were obtained SOUR=a/SRT+b and $SOUR=(a/m){\cdot}U+(b-a(n/m))$ from the relationship between SRT, U and SOUR. Empirical equations shows the possibility as a new design method for the aeration basin.

• Nuclear Engineering and Technology
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• v.48 no.1
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• pp.236-245
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• 2016

In a reactor coolant system of a nuclear power plant (NPP), an overpressure protection system keeps pressure in the loop within 110% of design pressure. However if the system does not work properly, pressure in the loop could elevate hugely in a short time. It would be seriously disastrous if a weak point in the pressure boundary component bursts and releases radioactive material within the containment; and it may lead to a leak outside the containment. In this study, a gross deformation that leads to a burst of pressure boundary components was investigated. Major components in the primary pressure boundary that is structurally important were selected based on structural mechanics, then, they were used to study the burst pressure of components by finite element method (FEM) analysis and by number of closed forms of theoretical relations. The burst pressure was also used as a metric of design optimization. It revealed which component was the weakest and which component had the highest margin to bursting failure. This information is valuable in severe accident progression prediction. The burst pressures of APR-1400, AP1000 and VVER-1000 reactor coolant systems were evaluated and compared to give relative margins of safety.

• The Transactions of the Korean Institute of Electrical Engineers D
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• v.52 no.8
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• pp.452-461
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• 2003

The reliability and accuracy of the information on control rod position are very important to the reactor safety and the design of the core protection system. A survey on the RSPT(Reed Switch Position Transmitter) type control rod position indicator system and its actual implementation in the existing nuclear power plants in Korea was performed first. The control rod position indicator having the high performance for SMART was developed on the basis of RSPT technology identified through the survey. The arrangement of permanent magnet and reed switches is the most important procedure in the design of control rod position indicator. The hysteresis of reed switches is one of the important factors in a repeat accuracy of control rod position indicator as well. This paper investigates efficiency of the magnetic flux concentrator and the hysteresis using FEM and verified differences in physicals characteristics by comparing the results of FEM and those of the experiment. As a result, it is shown that the characteristics of prototype control rod position indicator have a good agreement with the results of FEM.