• Journal of the Computational Structural Engineering Institute of Korea
• /
• v.15 no.1
• /
• pp.59-68
• /
• 2002

The cylindrical Upper Guide Structure assembly of the reactor intervals wish the Core Support Barrel and the Inner Barrel Assembly is subjected to flow induced loads horizontally which include random pressure fluctuation due to turbulent flow and pump pulsation pressures. The purpose of this papers is to perform random vibration and harmonic response analyses fort flow induced loads. The dynamic response characteristics due to random turbulence and pump pulsation loads were evaluated using the lumped mass beam model. Especially the model considered the annulus effects due to water gaps existing between cylindrical structures such as the Upper Guide Structure Barrel, the Core Support Barrel, and the Inner Barrel Assembly. The effect of the Inner Barrel Assembly inside the Upper Guide Structure assembly was studied. The peak dynamic responses lot each loading condition due to the addition of IBA were affected by the natural frequencies of the structures. Therefore the peak dynamic responses of the structures should be conservatively obtained from evaluation of dynamic analysis for various loading conditions.

• Proceedings of the KSME Conference
• /
• 2007.05a
• /
• pp.300-304
• /
• 2007

The reactor head structure assembly(RHSA) is the structure located on the reactor assembly. The purpose of the assembly is providing interface location for cables, preventing pipe whips, prohibiting instruments from becoming missiles, and restraining CEDMs' horizontal motion. On the RHSA, reactor disconnect panels(RDP) are installed. The installation location of RDP is to be decided to minimize the geometric interface with other components. Since the neighborhood of RHSA is crowded due to many connectors and cables, it is necessary to find the good design of RHSA to make an intricate situation attenuated and the required function maintained. The geometric shape and overall configuration of RHSA are determined by axiomatic design approach. The FRs of RHSA are specified and the corresponding DPs are found to satisfy FRs in sequence. The finite element analysis is carried out based on the result of the axiomatic design to evaluate the structural integrity.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
• /
• 2012.10a
• /
• pp.306-311
• /
• 2012

A reactor vessel internals comprehensive vibration assessment program (RVI CVAP) of an advanced power reactor 1400 (APR1400) is being performed as a non-prototype category-2 type of reactor based on the US Nuclear Regulatory Commission Regulatory Guide (NRC RG) 1.20. The aim of this paper is to present the results of structural response analysis and measuring locations of a upper guide structure (UGS) assembly of the APR1400 reactor. The analysis results of the UGS assembly results show that meet the specified integrity levels of the design acceptance criteria. Also, the measuring locations are set by the analysis results of the UGS assembly and selection criteria of measuring locations prior to this study. These analysis results and measuring locations will be used as fundamental materials to design a measurement system for the APR1400 RVI CVAP.

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

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
• /
• 2012.04a
• /
• pp.907-910
• /
• 2012

Since the inner barrel assembly of the Advanced Power Reactor 1400 reactor vessel is a new design feature introduced instead of CEA(control element assembly) shroud assembly, the inner barrel assembly can be a significant object of structural integrity assessment. This paper covers the structural responses of top plate, which is a component of the inner barrel assembly, against the deterministic hydraulic load induced by pump pulsation and the random hydraulic load induced by turbulence of coolant. The top plate responds to the deterministic hydraulic load more than to the random hydraulic load and shows enough structural integrity. The results of this paper will be important basis for the selection of instruments and measurement location.

• Transactions of the Korean Society for Noise and Vibration Engineering
• /
• v.23 no.1
• /
• pp.49-55
• /
• 2013

A reactor vessel internals comprehensive vibration assessment program(RVI CVAP) of an advanced power reactor 1400(APR1400) is being performed as a non-prototype category-2 type of reactor based on the US nuclear regulatory commission regulatory guide(NRC RG) 1.20. The aim of this paper is to present the results of structural response analysis and measuring locations of a upper guide structure(UGS) assembly of the APR1400 reactor. The analysis results of the UGS assembly show that the specified integrity levels meet the design acceptance criteria. Also, the measuring locations are determined by the analysis results of the UGS assembly and selection criteria of previous study. These analysis results and measuring locations will be used as a guide to design a measurement system for the APR1400 RVI CVAP.

• Computational Structural Engineering
• /
• v.7 no.3
• /
• pp.95-103
• /
• 1994

The safety related components in the nuclear power plant should be designed to withstand the seismic load. Among these components the integrity of reactor internals under earthquake load is important in stand points of safety and economics, because these are classified to Seismic Class I components. So far the modelling methods of reactor internals have been investigated by many authors. In this paper, the dynamic behaviour of reactor internals of Yong Gwang 1&2 nuclear power plants under SSE(Safe Shutdown Earthquake) load is analyzed by using of the simpled Global Beam Model. For this, as a first step, the characteristic analysis of reactor internal components are performed by using of the finite element code ANSYS. And the Global Beam Model for reactor internals which includes beam elements, nonlinear impact springs which have gaps in upper and lower positions, and hydrodynamical couplings which simulate the fluid-filled cylinders of reactor vessel and core barrel structures is established. And for the exciting external force the response spectrum which is applied to reactor support is converted to the time history input. With this excitation and the model the dynamic behaviour of reactor internals is obtained. As the results, the structural integrity of reactor internal components under seismic excitation is verified and the input for the detailed duel assembly series model could be obtained. And the simplicity and effectiveness of Global Beam Model and the economics of the explicit Runge-Kutta-Gills algorithm in impact problem of high frequency interface components are confirmed.

• Journal of Energy Engineering
• /
• v.24 no.3
• /
• pp.135-141
• /
• 2015

The core melts stratifies into lower mixture layer and upper metal layer by density in a severe accident condition. The decay heat generated from the mixture layer threatens the integrity of the reactor vessel. This study simulated the natural convection heat transfer of the mixture layer with volumetric heat source using the mass transfer system. $H_2SO_4-CuSO_4$ electroplating system was used as the mass transfer system. With the modified Rayleigh number of $3{\times}10^{14}$, the Nusselt number showed minimum at the bottom and increased along curvature to the top of the experimental apparatus.

• KSCE Journal of Civil and Environmental Engineering Research
• /
• v.13 no.3
• /
• pp.59-67
• /
• 1993

The variations of the natural frequencies and the peak response acceleration at the top of prestressed concrete reactor building due to random variability and/or model uncertainty of structural parameters are studied. The results may be used as essential input parameters in seismic probabilistic risk assessment or seismic margin assessment of the reactor building. The sensitivity test of each structural parameter is first performed to determine the most influential parameter upon the natural frequency of structure model. Then Monte Carlo simulation technique is applied to evaluate the effect of parameter variation on the natural frequencies and the peak response acceleration. The acceleration time history is obtained by direct integration scheme. As the study results, it is found that the fundamental natural frequency and the peak response acceleration at the top of the building are most strongly affected by Young's modulus among the structural parameters, in which the value of mean plus one standard deviation obtained by probabilistic approach deviates up to about (+)12% from the result of deterministic method. Considering the uncertainty of flexural rigidity, the structural responses vary in range of (-)4%~(+)14%.

• Journal of the Earthquake Engineering Society of Korea
• /
• v.2 no.1
• /
• pp.1-10
• /
• 1998

In the dynamic soil-structure interaction(SSI) analysis, numerous uncertain parameters are involved. They include the uncertainties in the definition of input motions, modeling of soil-structure interaction systems. analysis techniques, etc. This paper presents the results of parametric studies of the seismic responses of a reactor containment structure built on the viscoelastic layered soil. Among the numerous parameter, this study concentrates on the effects of definition point of the input motion, embedment of structure to the base soil, thickness of the top soil layer, and rigidity of the base soil. The substructure method using frequency independent impedances is adopted. The method is based on the mode superposition method in time domain using the composite modal damping values of th SSI system computed from the ratio of dissipated energy to the strain energy for each model. From the study results, the sensitivity of each parameter on the earthquake responses has been suggested for the practical application of the substructure method of SSI analysis.