Feasibility Study on Strain-Based Seismic Design Criteria for Nuclear Components

Koo, Gyeong-Hoi; Kim, Jong Sung; Kim, Yun Jae

doi:10.3390/en13174435

Cited by 7 publications

(15 citation statements)

References 5 publications

(5 reference statements)

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“…In this study, the seismic test mockup is designed to be able to simulate the actual nuclear component installed in the in-structure building. This will be consistent with the purpose of validating the strain-based seismic design method developed in previous studies [9,13]. To do this, the metal component nozzle, which is known as one of the seismically fragile components in nuclear steam supply systems (NSSS), was selected as the reference for the test mockup.…”

Section: Configuration and Dimensionsmentioning

confidence: 79%

“…Recently, as one of the approaches against large earthquakes, studies on the new seismic design acceptance criteria for large earthquakes have been actively conducted for the application of nuclear metal components [6][7][8][9]. The main concept of this design criteria provides strain-based design limits to protect the pressure-retaining nuclear metal components against an inelastic strain-induced failure mode.…”

Section: Introductionmentioning

confidence: 99%

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Shaking Table Tests to Validate Inelastic Seismic Analysis Method Applicable to Nuclear Metal Components

Koo

Ahn²,

Hwang³

et al. 2021

Applied Sciences

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The main purpose of this study is to perform shaking table tests to validate the inelastic seismic analysis method applicable to pressure-retaining metal components in nuclear power plants (NPPs). To do this, the test mockup was designed and fabricated to be able to describe the hot leg surge line nozzle with a piping system, which is known to be one of the seismically fragile components in nuclear steam supply systems (NSSS). The used input motions are the displacement time histories corresponding to the design floor response spectrum at an elevation of 136 ft in the in-structure building in NPPs. Two earthquake levels are used in this study. One is the design-basis safe shutdown earthquake level (SSE, PGA = 0.3 g) and the other is the beyond-design-basis earthquake level (BDBE, PGA = 0.6 g), which is linearly scaled from the SSE level. To measure the inelastic strain responses, five strain gauges were attached at the expected critical locations in the target nozzle, and three accelerometers were installed at the shaking table and piping system to measure the dynamic responses. From the results of the shaking table tests, it was found that the plastic strain response at the target nozzle and the acceleration response at the piping system were not amplified by as much as two times the input earthquake level because the plastic behavior in the piping system significantly contributed to energy dissipation during the seismic events. To simulate the test results, elastoplastic seismic analyses with the well-known Chaboche kinematic hardening model and the Voce isotropic hardening model for Type 316 stainless steel were carried out, and the results of the principal strain and the acceleration responses were compared with the test results. From the comparison, it was found that the inelastic seismic analysis method can give very reasonable results when the earthquake level is large enough to invoke plastic behavior in nuclear metal components.

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Section: Configuration and Dimensionsmentioning

confidence: 79%

Section: Introductionmentioning

confidence: 99%

Shaking Table Tests to Validate Inelastic Seismic Analysis Method Applicable to Nuclear Metal Components

Koo

Ahn²,

Hwang³

et al. 2021

Applied Sciences

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show abstract

“…In this paper, two types of failure modes are considered to prevent the seism induced failure of a pressure-retaining safety-related nuclear component. The first o the ductile fracture failure mode, and the other is the fatigue-induced failure mode these two failure modes, the strain-based acceptance criteria, which is specified for D Service Limits for nuclear component, are used as follows [17];…”

Section: Definitions Of Failure Modesmentioning

confidence: 99%

“…In this paper, the general procedures of seismic FA with the probabilistic approach are investigated with the previously developed strain-based seismic design criteria [17] for the nuclear metal components retaining high-pressure boundary, which inevitably require an inelastic seismic time history analysis. The procedures for performing fragility calculations of the high confidence of low probability of failure (HCLPF) capacity are based on the recommendations in the EPRI technical reports [15,18].…”

Section: Introductionmentioning

confidence: 99%

Study on Inelastic Strain-Based Seismic Fragility Analysis for Nuclear Metal Components

2021

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The main purpose of this study is to investigate the feasibility of the seismic fragility analysis (FA) with the strain-based failure modes for the nuclear metal components retaining pressure boundary. Through this study, it is expected that we can find analytical ways to enhance the high confidence of low probability of failure (HCLPF) capacity potentially contained in the conservative seismic design criteria required for the nuclear metal components. Another goal is to investigate the feasibility of the seismic FA to be used as an alternative seismic design rule for beyond-design-basis earthquakes. To do this, the general procedures of the seismic FA using the inelastic seismic analysis for the nuclear metal components are investigated. Their procedures are described in detail by the exampled calculations for the surge line nozzles connecting hot leg piping and the pressurizer, known as one of the seismic fragile components in NSSS (Nuclear Steam Supply System). To define the seismic failure modes for the seismic FA, the seismic strain-based design criteria, with two seismic acceptance criteria against the ductile fracture failure mode and fatigue-induced failure mode, are used in order to reduce the conservatism contained in the conventional stress-based seismic design criteria. In the exampled calculation of the inelastic seismic strain response beyond an elastic regime, precise inelastic seismic analyses with Chaboche’s kinematic and Voce isotropic hardening material models are used. From the results of the seismic FA by the probabilistic approach for the exampled target component, it is confirmed that the approach of the strain-based seismic FA can extract the maximum seismic capacity of the nuclear metal components with more accurate inelastic seismic analysis minimizing the number of variables for the components.

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“…In South Korea, studies have been conducted to improve the seismic performance of nuclear power plant piping systems and pressure vessels. Strain-based criteria have been developed for cases in which plastic deformation occurs in the major components of nuclear power plant piping systems under large seismic load conditions that exceed the design basis earthquake [17]. Additionally, studies have been conducted to assess the seismic fragility of nuclear power plant piping systems.…”

Section: Introductionmentioning

confidence: 99%

Quantitative Limit State Assessment of a 3-Inch Carbon Steel Pipe Tee in a Nuclear Power Plant Using a Damage Index

et al. 2020

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Seismic motions are likely to cause large displacements in nuclear power plants because the main mode of their piping systems is dominated by the low-frequency region. Additionally, large relative displacement may occur in the piping systems because their supports are installed in several places, and each support is subjected to different seismic motions. Therefore, to assess the seismic performance of a piping system, the relative displacement repeated by seismic motions must be considered. In this study, in-plane cyclic loading tests were conducted under various constant amplitudes using test specimens composed of SCH 40 3-inch pipes and a tee in the piping system of a nuclear power plant. Additionally, an attempt was made to quantitatively express the failure criteria using a damage index based on the dissipated energy that used the force–displacement and moment–deformation angle relationships. The failure mode was defined as the leakage caused by a through-wall crack, and the failure criteria were compared and analyzed using the damage index of Park and Ang and that of Banon. Additionally, the method of defining the yield point required to calculate the damage index was examined. It was confirmed that the failure criteria of the SCH 40 3-inch carbon steel pipe tee can be effectively expressed using the damage index.

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Feasibility Study on Strain-Based Seismic Design Criteria for Nuclear Components

Cited by 7 publications

References 5 publications

Shaking Table Tests to Validate Inelastic Seismic Analysis Method Applicable to Nuclear Metal Components

Shaking Table Tests to Validate Inelastic Seismic Analysis Method Applicable to Nuclear Metal Components

Study on Inelastic Strain-Based Seismic Fragility Analysis for Nuclear Metal Components

Quantitative Limit State Assessment of a 3-Inch Carbon Steel Pipe Tee in a Nuclear Power Plant Using a Damage Index

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