• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2005.11a
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• pp.267-270
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• 2005

Finite element model of pressurized water reactor internals were obtained using ANSYS software package to analyze dynamic characteristics. The pressure vessel, hold-down ring, alinement key, core support barrel(CSB), upper guide structure(UGS) and fluid gap were fully modeled using structural solid element(SOLID45) and fluid element(FLUID80) which is one of element types. Also modal analysis using the above finite element model has been performed. As a result, it was found that the fundamental beam mode natural frequency of the CSB were 8.2 Hz, the shell mode one 14.5 Hz. To verify the Finite Element Analysis(FEA), we compare the analysis result with experimental data that is obtained from the plant IVMS(internal Vibration Monitoring System). The experimental results are good agreement with the FEA model.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.37 no.9
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• pp.1127-1132
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• 2013

The failure of reactor internals may have a significant effect on the safe operation and shutdown of a reactor. Various agings related to neutron irradiation occur or can potentially occur in the reactor internals owing to high neutron irradiation levels. Austenitic stainless steel, one of the principal materials constituting the reactor internals, shows different mechanical material behaviors such as tensile/creep properties and fracture toughness with neutron irradiation levels. This variation should be considered when the structural integrity of the reactor internals against agings during the design lifetime or continued operation period is evaluated. In this study, user subroutine programs considering the variation of mechanical material behaviors with neutron irradiation levels were developed. The programs were validated by testing them for various conditions.

• Computational Structural Engineering
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• v.6 no.4
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• pp.73-82
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• 1993

Investigated in this paper is the effect of fluid-structure interaction between reactor internal components due to their immersion in a confining fluid on the dynamic responses. A non-linear mathematical model is developed for the dynamic analysis of the reactor internals, which includes lumped masses, stiffnesses and hydrodynamic couplings. The hydrodynamic mass matrix which characterizes the fluid-structure interaction is calculated. Also, the equations of motion containing hydrodynamic mass matrix are presented. The responses of the reactor internals due to seismic and pipe break excitations are obtained for the case of with- and without-hydrodynamic couplings and the different response characteristics are investigated.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 1998.04a
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• pp.582-589
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• 1998

This paper presents the analysis results of ex-core and in-core neutron noise, acceleration signals and pressure fluctuation measured at Ulchin Nuclear Unit 1 & 2 to identify and monitor the reactor internals vibration including fuel motion. A phase separation algorithm developed by authors was applied to the neutron noises to clearly identify the reactor internals vibratory motion. The beam mode frequency of the core support barrel was identified to be 8Hz and the shell mode to be 20Hz. The first frequency of the fuel assembly was also found to be 3Hz, while first two acoustic frequencies of the primary coolant system were 6 and 17.5Hz. By monitoring and analyzing these frequencies periodically, it is possible to diagnose the operating condition of reactor internals and to provide an early detection of faults for the predictive maintenance.

• Journal of the Korean Society of Systems Engineering
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• v.12 no.1
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• pp.25-40
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• 2016

The Systems Engineering (SE) approach is characterized by the application of a structured engineering methodology for the design of a complex system or component. In this study, the SE methodology is used to design a nondestructive inspection system for Bottom Mounted Instrumentation (BMI) nozzles. We developed a system that enables nondestructive inspection of BMI nozzles during regular refueling outage without removing the reactor internals. A special ultrasonic (UT) probe is introduced to scan and detect cracks within the weld region of the nozzle. A 3D model of the inspection structure system was developed along with the reactor pressure vessel (RPV) and internals which permits a virtual 3D simulation of the operation to check the design concept and effectiveness of the system and to provide a good visualization of the system. This approach allows for a virtual walk through to verify the proposed BMI nozzle inspection system.

• Nuclear Engineering and Technology
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• v.29 no.5
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• pp.417-431
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• 1997

To resolve a general argument about OBE elimination for the future nuclear power plant design, seismic responses of reactor vessel internals and fuel assembly for Ulchin nuclear power plant units 3 and 4 in Korea are investigated as an example. Dynamic analyses of the coupled internals and core are performed for the seismic excitations using the reactor vessel motions. By investigating the response relations between OBE and SSE and their response characteristics, the critical components for OBE loading are addressed. Also the fuel assembly responses are calculated using the core plate motions and their behavior is found to be insignificant for OBE elimination.

• Nuclear Engineering and Technology
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• v.25 no.1
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• pp.28-36
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• 1993

Much effort is being done to standardize the PWR-type nuclear power plant in Korea. This paper presents the development of seismic design criteria for the reactor internals as a part of the standardization program for nuclear power plant. The seismic design loads of the reactor internals are calculated using the reference input motions of reactor vessel taken from Yong-gwang Nuclear Power Plant Units 3 and 4. An overview of analysis related to the basic parameters and methodologies is presented. Also, the response of internal components for the reactor vessel motions is carefully investigated.

• Journal of KSNVE
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• v.10 no.1
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• pp.129-137
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• 2000

This paper presents the vibration characteristics of reactor internals of Ulchin-1 nuclear power plant, which are identified by using the conventional and the phase separated spectral analysis of the pressure vessel acceleration and ex-core neutron signals. These identified vibration characteristics show excellent agreement with those of Tricastin-1 nuclear power plant that is the prototype of Ulchin-1. And the trend of ex-core neutron signals has been observed during one reactor cycle. These results can be used as basic data for fault diagnosis of reactor internals.

• Nuclear Engineering and Technology
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• v.49 no.7
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• pp.1513-1523
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• 2017

Improvement of numerical analysis methods has been required to solve complicated phenomena that occur in nuclear facilities. Particularly, fluid-structure interaction (FSI) behavior should be resolved for accurate design and evaluation of complex reactor vessel internals (RVIs) submerged in coolant. In this study, the FSI effect on dynamic characteristics of RVIs in a typical 1,000 MWe nuclear power plant was investigated. Modal analyses of an integrated assembly were conducted by employing the fluid-structure (F-S) model as well as the traditional added-mass model. Subsequently, structural analyses were carried out using design response spectra combined with modal analysis data. Analysis results from the F-S model led to reductions of both frequency and Tresca stress compared to those values obtained using the added-mass model. Validation of the analysis method with the FSI model was also performed, from which the interface between the upper guide structure plate and the core shroud assembly lug was defined as the critical location of the typical RVIs, while all the relevant stress intensities satisfied the acceptance criteria.

• Journal of KSNVE
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• v.7 no.1
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• pp.33-41
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• 1997

To cover a range of possible site conditions where the Korean standard nuclear power plant may be constructed, a range of generic site conditions is selected for geologic and seismologic evaluation. To envelop other Asian countries as well as the Korean peninsula, there is an attempt to increase the seismic level to 0.3g ground motions for the safe shutdown earthquake. The dynamic analyses of the reactor vessel internals and fuel assemblies are performed for the increased motions and the effect of seismic level on the response is investigated. Also the nonlinear response characteristics are discussed by comparing the loads between operating basis earthquake and safe shutdown earthquake excitations. The design adequacy of the reactor vessel internals and fuel assemblies for the increased seismic level is addressed.