• 한국지진공학회논문집
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• 제22권2호
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• pp.45-54
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• 2018

The seismically isolated nuclear power plants shall be designed for design basis earthquake (DBE) and considered to ensure safety against beyond design basis earthquake (BDBE). In order to limit the excessive displacement of the seismic isolation system of the seismically isolated structure, the moat is installed at a certain distance from the upper mat supporting the superstructure. This certain distance is called clearance to stop (CS) and is calculated from the 90th percentile displacement of seismic isolation system subjected to BDBE. For design purposes, the CS can be obtained simply by multiplying the median displacement of the seismic isolation system against DBE by scale factor with a value of 3. The DBE and BDBE used in this study were generated by using 30 sets of artificial earthquakes corresponding to the nuclear standard design spectrum. In addition, latin hyper cube sampling was applied to generate 30 sets of artificial earthquakes corresponding to maximum - minimum spectra. For the DBE, the median displacement and the 99th percentile displacement of the seismic isolation system were calculated. For the BDBE, the suitability of the scale factor was assessed after calculating the 90th percentile displacement of the seismic isolation system.

• 한국지진공학회논문집
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• 제23권6호
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• pp.289-299
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• 2019

The tendency to use a probabilistic design method rather than a deterministic design method for the design of nuclear power plants (NPPs) will increase because their safety should be considered and strictly controlled in relation to various causes of damage. The distance between a seismically isolated NPP structure and a moat wall is called the clearance to stop. The clearance to stop is obtained from the 90th percentile displacement response of a seismically isolated NPP subject to a beyond design basis earthquake (BDBE) in the probabilistic design method. The purpose of this study is to analyze the effects of heating and buckling effects on the 90th percentile displacement response of a lead-rubber bearing (LRB) subject to a BDBE. The analysis results show that considering the heating and buckling effects to estimate the clearance to stop is conservative in the evaluation of the 90th percentile displacement response. If these two effects are not taken into account in the calculation of the clearance to stop, the underestimation of the clearance to stop causes unexpected damage because of an increase in the collision probability between the moat wall and the seismically isolated NPP.

• 한국지진공학회:학술대회논문집
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• 한국지진공학회 2003년도 추계 학술발표회논문집
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• pp.58-65
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• 2003

The seismic response analysis of a freestanding spent fuel storage cask model are performed for an artificial time history acceleration generated by the basis on the US NRC RG1.60 response acceleration spectrum. This paper focuses on the structural stability by seismic loads to check the overturing possibility of storage cask and the slipping displacement on bed. Parametric analyses of a simplified cask model are performed to take into account the variations in seismic load magnitude and cask/bed interface friction. The analyses results show that the storage cask has a large marginal integrity in the response acceleration and slipping distance for both design seismic and beyond design seismic loads.

• Nuclear Engineering and Technology
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• 제44권3호
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• pp.225-236
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• 2012

The Fukushima-Daiichi accident shook the world, as a well-known plant design, the General Electric BWR Mark I, was heavily damaged in the tsunami, which followed the Great Japanese Earthquake of 11 March 2011. Plant safety functions were lost and, as both AC and DC failed, manoeuvrability of the plants at the site virtually came to a full stop. The traditional system of Emergency Operating Procedures (EOPs) and Severe Accident Management Guidelines (SAMG) failed to protect core and containment, and severe core damage resulted, followed by devastating hydrogen explosions and, finally, considerable radioactive releases. The root cause may not only have been that the design against tsunamis was incorrect, but that the defence against accidents in most power plants is based on traditional assumptions, such as Large Break LOCA as the limiting event, whereas there is no engineered design against severe accidents in most plants. Accidents beyond the licensed design basis have hardly been considered in the various designs, and if they were included, they often were not classified for their safety role, as most system safety classifications considered only design basis accidents. It is, hence, time to again consider the Design Basis Accident, and ask ourselves whether the time has not come to consider engineered safety functions to mitigate core damage accidents. Associated is a proper classification of those systems that do the job. Also associated are safety criteria, which so far are only related to 'public health and safety'; in reality, nuclear accidents cause few casualties, but create immense economical and societal effects-for which there are no criteria to be met. Severe accidents create an environment far surpassing the imagination of those who developed EOPs and SAMG, most of which was developed after Three Mile Island - an accident where all was still in place, except the insight in the event was lost. It requires fundamental changes in our present safety approach and safety thinking and, hence, also in our EOPs and SAMG, in order to prevent future 'Fukushimas'.

• 한국압력기기공학회 논문집
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• 제14권2호
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• pp.32-40
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• 2018

Since 2000, the frequency of earthquakes beyond the 5.0 magnitude quake has been increasing in the Korean peninsula. For instance, the 5.0-magnitude earthquake in Baekryong-do in 2003 has occurred, and recent earthquake with Gyeongju(2016) and Pohang(2017) measured respectively magnitude of 5.2 and 5.8 on the Richter scale. As results, the public concern and anxiety about earthquakes are increasing, and therefore it is necessarily required for social infrastructure to reinforce seismic design and energy production facilities directly related to the national economy and security. This study represents the analysis of seismic performance evaluation methodology such as Seismic Margin Assessment (SMA), Seismic Probabilistic Risk Assessment (SPRA), High Confidence Low Probability Failure (HCLPF) in nuclear power plants in order to develop optimal seismic performance improvement. Current methodologies to evaluate nuclear power plants are also addressed. Through review of the nuclear structure evaluation past and current trend, it contributes to be the basis for the improvement of evaluation techniques on the next generation of nuclear power plants.

• Nuclear Engineering and Technology
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• 제52권12호
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• pp.2918-2927
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• 2020

This paper proposes a simplified elastic-plastic analysis procedure using the penalty factors presented in the Code Case N-779 for strain-based fatigue assessment of nuclear safety class 1 components under severe seismic loads such as safety shutdown earthquake and beyond design-basis earthquake. First, a simplified elastic-plastic analysis procedure for strain-based fatigue assessment of nuclear safety class 1 components under the severe seismic loads was proposed based on the analysis result for the simplified elastic-plastic analysis procedure in the Code Case N-779 and the stress categories corresponding to normal operation and seismic loads. Second, total strain amplitude was calculated directly by performing finite element cyclic elastic-plastic seismic analysis for a hot leg nozzle in pressurizer surge line subject to combined loading including deadweight, pressure, seismic inertia load, and seismic anchor motion, as well as was derived indirectly by applying the proposed analysis procedure to the finite element elastic stress analysis result for each load. Third, strain-based fatigue assessment was implemented by applying the strain-based fatigue acceptance criteria in the ASME B&PV Code, Sec. III, Subsec. NB, Article NB-3200 and by using the total strain amplitude values calculated. Last, the total strain amplitude and the fatigue assessment result corresponding to the simplified elastic-plastic analysis were compared with those using the finite element elastic-plastic seismic analysis results. As a result of the comparison, it was identified that the proposed analysis procedure can derive reasonable and conservative results.

• Earthquakes and Structures
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• 제12권6호
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• pp.653-663
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• 2017

A seismic margin assessment evaluates how much margin exists for the system under beyond design basis earthquake events. Specifically, the seismic margin for the entire system is evaluated by utilizing a systems analysis based on the sub-system and component seismic fragility data. Each seismic fragility curve is obtained by using empirical, experimental, and/or numerical simulation data. The systems analysis is generally performed by employing a fault tree analysis. However, the current practice has clear limitations in that it cannot deal with the uncertainties of basic components and accommodate the newly observed data. Therefore, in this paper, we present a Bayesian-based seismic margin assessment that is conducted using seismic fragility data and fault tree analysis including Bayesian inference. This proposed approach is first applied to the pooltype nuclear research reactor system for the quantitative evaluation of the seismic margin. The results show that the applied approach can allow updating by considering the newly available data/information at any level of the fault tree, and can identify critical scenarios modified due to new information. Also, given the seismic hazard information, this approach is further extended to the real-time risk evaluation. Thus, the proposed approach can finally be expected to solve the fundamental restrictions of the current method.

• 한국압력기기공학회 논문집
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• 제14권2호
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• pp.59-68
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• 2018

This study compares true stress-true strain curves obtained by tensile tests of various piping materials with bi-linear stress-strain approximation suggested in the JSME Code Case(CC) Draft, a guideline for piping seismic inelastic response analysis. Based on the comparisons, the reliability of the bi-linear approximation is evaluated. It is found that bi-linear stress-strain curve of TP316 stainless steel is in good agreement with its true stress-true strain curve. However, Bi-linear stress-strain curves of TP304 stainless steel and carbon steels determined by the approximation cannot appropriately estimate their stress-strain behavior. Accordingly new bi-linear approximations for carbon steels and low-alloy steels are proposed. The proposed bi-linear approximations for carbon and low-alloy steels, which include the temperature effect on strength and hardening of material, estimate their stress-strain behavior reasonably well.

• 한국지진공학회논문집
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• 제27권1호
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• pp.25-35
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• 2023

Considering the non-linear behavior of structure and soil when evaluating a nuclear power plant's seismic safety under a beyond-design basis earthquake is essential. In order to obtain the nonlinear response of a nuclear power plant structure, a time-domain SSI analysis method that considers the nonlinearity of soil and structure and the nonlinear Soil-Structure Interaction (SSI) effect is necessary. The Boundary Reaction Method (BRM) is a time-domain SSI analysis method. The BRM can be applied effectively with a Perfectly Matched Layer (PML), which is an effective energy absorbing boundary condition. The BRM has a characteristic that the magnitude of the response in far-field soil increases as the boundary interface of the effective seismic load moves outward. In addition, the PML has poor absorption performance of low-frequency waves. For this reason, the accuracy of the low-frequency response may be degraded when analyzing the combination of the BRM and the PML. In this study, the accuracy of the analysis response was improved by adjusting the PML input parameters to improve this problem. The accuracy of the response was evaluated by using the analysis response using KIESSI-3D, a frequency domain SSI analysis program, as a reference solution. As a result of the analysis applying the optimal PML parameter, the average error rate of the acceleration response spectrum for 9 degrees of freedom of the structure was 3.40%, which was highly similar to the reference result. In addition, time-domain nonlinear SSI analysis was performed with the soil's nonlinearity to show this study's applicability. As a result of nonlinear SSI analysis, plastic deformation was concentrated in the soil around the foundation. The analysis results found that the analysis method combining BRM and PML can be effectively applied to the seismic response analysis of nuclear power plant structures.