• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.899-905
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• 1998

The analysis priority makings the recommendation to reduce the total core damage frequency (CDF) of Wolsong nuclear Power Plant nits 2/3/4 was Performed in this paper. In order to derive the recommendation, the sensitivity analysis of CDF on which major contributors effect m performed based on the accident quantification results during Level 1 Probabilistic safety assessment (PSA). Priorities were ranked in tile way that compares the CDF reduction rate with efforts required to implement those recommendations using risk matrix

• Nuclear Engineering and Technology
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• v.36 no.6
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• pp.485-496
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• 2004

Typical probabilistic safety assessment (PSA) results were used to estimate the performance indicator (PI) thresholds of unplanned reactor scram (URS) and safety system unavailability (SSU) for Korean nuclear power plants (NPPs). The changes in core damage frequency (${\Delta}$CDFs) of $10^{-6}/yr$, $10^{-5}/yr$, and $10^{-4}/yr$ were adopted as the risk criteria in setting up the PI thresholds. The PI thresholds for the URS were estimated using information pertaining to the initiating event frequencies, the CDF, and the CDF contribution of each initiating event. The PI thresholds of the SSU were estimated using information on the unavailability, the Fussell-Vesely importance, and the CDF.

• Nuclear Engineering and Technology
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• v.40 no.4
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• pp.285-298
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• 2008

This study evaluates the performance of a coil spring-viscous damper system for the vibration and seismic isolation of an Emergency Diesel Generator (EDG) by measuring its operational vibration and seismic responses. The vibration performance of a coil spring-viscous damper system was evaluated by the vibration measurements for an identical EDG set with different base systems - one with an anchor bolt system and the other with a coil spring-viscous damper system. The seismic performance of the coil spring-viscous damper system was evaluated by seismic tests with a scaled model of a base-isolated EDG on a shaking table. The effects of EDG base isolation on the fragility curve and core damage frequency in a nuclear power plant were also investigated through a case study.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.31 no.8
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• pp.870-879
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• 2007

In this study, various damage modes in bending unimorph piezoelectric composite actuators with a thin sandwiched PZT plate during bending fracture tests have been evaluated by monitoring acoustic emission (AE) signals in terms of waveform and peak frequency as well as AE parameters. Three kinds of actuator specimens consisting of woven fabric fiber skin layers and a PZT ceramic core layer are loaded with a roller and an AE activity from the specimen is monitored during the entire loading using an AE transducer mounted on the specimen. AE characteristics from a monolithic PZT ceramic with a thickness of $250{\mu}m$ are examined first in order to distinguish different AE signals from various possible damage modes in piezoelectric composite actuators. Post-failure observations and stress analyses in the respective layers of the specimens are conducted to identify particular features in the acoustic emission signal that correspond to specific types of damage modes. As a result, the signal classification based on waveform and peak frequency analyses successfully describes the failure process of the bending piezoelectric composite actuator exhibiting diverse failure mechanisms. Furthermore, it is elucidated that when the PZT ceramic embedded actuators are loaded mechanical bending loads, the failure process of actuator specimens with different lay-up configurations is almost same irrespective of their lay-up configurations.

• Steel and Composite Structures
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• v.38 no.5
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• pp.477-496
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• 2021

The main objective of this paper is to study vibration of sandwich open cylindrical panel with damaged core and FG face sheets based on three-dimensional theory of elasticity. The structures are made of a damaged isotropic core and two external face sheets. These skins are strengthened at the nanoscale level by randomly oriented Carbon nanotubes (CNTs) and are reinforced at the microscale stage by oriented straight fibers. These reinforcing phases are included in a polymer matrix and a three-phase approach based on the Eshelby-Mori-Tanaka scheme and on the Halpin-Tsai approach, which is developed to compute the overall mechanical properties of the composite material. Three complicated equations of motion for the panel under consideration are semi-analytically solved by using 2-D differential quadrature method. Several parametric analyses are carried out to investigate the mechanical behavior of these multi-layered structures depending on the damage features, through-the-thickness distribution and boundary conditions. It is seen that for the large amount of power-law index "P", increasing this parameter does not have significant effect on the non-dimensional natural frequency parameters of the FG sandwich curved panel. Results indicate that by increasing the value of isotropic damage parameter "D" up to the unity (fully damaged core) the frequency would tend to become zero. One can dictate the fiber variation profile through the radial direction of the sandwich panel via the amount of "P", "b" and "c" parameters. It should be noticed that with increase of volume fraction of fibers, the frequency parameter of the panels does not increase necessarily, so by considering suitable amounts of power-law index "P" and the parameters "b" and "c", one can get dynamic characteristics similar or better than the isotropic limit case for laminated FG curved panels.

• Nuclear Engineering and Technology
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• v.52 no.1
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• pp.211-221
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• 2020

One fundamental element of probabilistic safety assessment (PSA) is the initiating event (IE) analysis. Since IE frequencies can change over time, time-trend analysis is required to obtain optimized IE frequencies. Accordingly, such time-trend analyses have been employed to estimate industry-average IE frequencies for use in the PSAs of U.S. nuclear power plants (NPPs); existing PSAs of Korean NPPs, however, neglect such analysis in the estimation of IE frequencies. This article therefore provides the method for and results of estimating Korean industry-average IE frequencies using time-trend analysis. It also examines the effects of the IE frequencies obtained from this study on risk insights by applying them to recently updated internal events Level 1 PSA models (at-power and shutdown) for an OPR-1000 plant. As a result, at-power core damage frequency decreased while shutdown core damage frequency increased, with the related contributions from each IE category changing accordingly. These results imply that the incorporation of time-trend analysis leads to different IE frequencies and resulting risk insights. The IE frequency distributions presented in this study can be used in future PSA updates for Korean NPPs, and should be further updated themselves by adding more recent data.

• Journal of the Korean Society of Safety
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• v.28 no.3
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• pp.123-128
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• 2013

The Advanced Power Reactor Plus(APR+), which is a GEN III+ reactor based on the APR1400, is being developed in Korea. In order to enhance the safety of the APR+, a passive auxiliary feedwater system(PAFS) has been adopted in the APR+. The PAFS replaces the conventional active auxiliary feedwater system(AFWS) by introducing a natural driving force mechanism while maintaining the system function of cooling the primary side and removing the decay heat. As the PAFS completely replaces the conventional AFWS, it is required to verify the cooling capacity of PAFS for the core damage frequency(CDF) evaluation. For this reason, this paper discusses the cooling performance of the PAFS during transient accidents. The test case and scenarios were picked from the result of the sensitivity analysis in APR+ Probabilistic Safety Assessment(PSA). The analysis was performed by the best estimate thermal-hydraulic code, RELAP5/.MOD3.3. This study shows that the plant maintains the stable state without the core damages under the given test scenarios. The results of PSA considering this analysis' results shows that the CDF values are decreased. The analysis results can be used for more realistic and accurate performance of a PSA.

• Steel and Composite Structures
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• v.51 no.1
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• pp.73-91
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• 2024

This paper has focused on presenting vibration analysis of trapezoidal sandwich plates with a damaged core and FG wavy CNT-reinforced face sheets. A damage model is introduced to provide an analytical description of an irreversible rheological process that causes the decay of the mechanical properties, in terms of engineering constants. An isotropic damage is considered for the core of the sandwich structure. The classical theory concerning the mechanical efficiency of a matrix embedding finite length fibers has been modified by introducing the tube-to-tube random contact, which explicitly accounts for the progressive reduction of the tubes' effective aspect ratio as the filler content increases. The First-order shear deformation theory of plate is utilized to establish governing partial differential equations and boundary conditions for the trapezoidal plate. The governing equations together with related boundary conditions are discretized using a mapping-generalized differential quadrature (GDQ) method in spatial domain. Then natural frequencies of the trapezoidal sandwich plates are obtained using GDQ method. Validity of the current study is evaluated by comparing its numerical results with those available in the literature. After demonstrating the convergence and accuracy of the method, different parametric studies for laminated trapezoidal structure including carbon nanotubes waviness (0≤w≤1), CNT aspect ratio (0≤AR≤4000), face sheet to core thickness ratio (0.1 ≤ ${\frac{h_f}{h_c}}$ ≤ 0.5), trapezoidal side angles (30° ≤ α, β ≤ 90°) and damaged parameter (0 ≤ D < 1) are carried out. It is explicated that the damaged core and weight fraction, carbon nanotubes (CNTs) waviness and CNT aspect ratio can significantly affect the vibrational behavior of the sandwich structure. Results show that by increasing the values of waviness index (w), normalized natural frequency of the structure decreases, and the straight CNT (w=0) gives the highest frequency. For an overall comprehension on vibration of laminated trapezoidal plates, some selected vibration mode shapes were graphically represented in this study.

• Journal of the Korean Society of Safety
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• v.34 no.3
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• pp.102-110
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• 2019

This paper describes the work and results of the preliminary Probabilistic Safety Assessment (PSA) for a research reactor in the design phase. This preliminary PSA was undertaken to assess the level of safety for the design of a research reactor and to evaluate whether it is probabilistically safe to operate and reliable to use. The scope of the PSA described here is a Level 1 PSA which addresses the risks associated with core damage. After reviewing the documents and its conceptual design, eight typical initiating events are selected regarding internal events during the normal operation of the reactor. Simple fault tree models for the PSA are developed instead of the detailed model at this conceptual design stage. A total of 32 core damage accident sequences for an internal event analysis were identified and quantified using the AIMS-PSA. LOCA-I has a dominant contribution to the total CDF by a single initiating event. The CDF from the internal events of a research reactor is estimated to be 7.38E-07/year. The CDF for the representative initiating events is less than 1.0E-6/year even though conservative assumptions are used in reliability data. The conceptual design of the research reactor is designed to be sufficiently safe from the viewpoint of safety.

• Journal of the Korean Society of Safety
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• v.33 no.5
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• pp.157-163
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• 2018

This paper describes design improvement to a research rector for safety enhancement using Probabilistic Safety Assessment (PSA). This PSA under reactor design was undertaken to assess the level of safety for the design of a research reactor and to evaluate whether it is probabilistically safe to operate and reliable to use. The scope of the PSA reported here is a Level 1 PSA, which addresses the risks associated with the core damage. The technical objectives of this study were to identify accident sequences leading to core damage and to derive design improvement from the dominant accident sequences through the sensitivity analysis. The AIMS-PSA and FTREX were used for the this PSA of the research reactor. The criterion for inclusion was all sequences with a point estimate frequency greater than a truncation value of 1.0E-14/yr. The final result indicates a point estimate of 6.79E-05/yr for the overall Core Damage Frequency (CDF) attributable to internal initiating events for the research reactor under design. Based on the dominant accident sequences from the PSA, the seven kinds of sensitivity analysis were performed and some design improvement items were derived. When the five methods to improve the safety were all applied to the reactor design and emergency operating procedure, its risk was reduced to about 1.21E-06/yr from 6.79E-05/yr. The contribution of LOCA and LOEP with high CDF were significantly reduced by the sensitivity analysis. The safety of the research reactor was well improved and the risk was reduced than before adapting the design improvement gotten from the sensitivity analysis. The present study indicated that the research reactor has the well-balanced safety in regard to each initiating event contribution to CDF. The PSA methodology is very effective to improve reactor safety in a conceptual design phase and especially, Risk-informed design(RID) is very nice way to find the deficiencies of research reactor under design and to improve the reactor safety by solving them.