Improvements on Fatigue Analysis Methods for the Design of Nuclear Components Subjected to the French RCC-M Code

Grandemange, J.M.; Héliot, J.; Vagner, J.; Morel, Anne-Françoise; Faidy, C.

doi:10.1016/b978-1-4832-8430-9.50013-7

Cited by 7 publications

(3 citation statements)

References 7 publications

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“…For the stepped pipe FE model, Ke R was generally found to be 10-30% conservative for Sn > 3Sm, showing a decreasing trend which aligned closely with the RCC-M thermal-plastic correction factor, Ke ther . This is to be expected since both approaches share the same technical basis outlined in WRC-361 [5]. For the PWR nozzle FE models, Ke R showed a similar trend in conservatism to the stepped pipe, albeit with slightly higher scatter depending on the relative proportions of Sn,tb and Sn,tm at the assessment location.…”

Section: Ranganath's Methodsmentioning

confidence: 59%

“…A recent proposal by Ranganath et al [4] is currently under consideration by the ASME Working Groups on Design Methodology (WGDM) and Fatigue Strength (WGFS) for inclusion as an ASME Code Case (Case 17-225). The approach is based partially on the methodology outlined in Welding Research Council (WRC) Bulletin 361 [5], and proposes a weighted average correction factor, Ke R , defined by Eq. ( 5):…”

Section: Ranganath's Methodsmentioning

confidence: 99%

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Critical Review of ASME III Plasticity Correction Factors for Fatigue Design-By-Analysis of Nuclear Power Plant Components

Clarkson

Bell

Mackenzie

2021

Journal of Pressure Vessel Technology

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Despite significant technological progress in recent years, elastic-plastic fatigue analysis of pressure-retaining components remains a time-consuming venture. Accordingly, nuclear pressure vessel design codes such as ASME Section III provide simplified elastic-plastic analysis procedures as a practical alternative. This approach can be excessively conservative under certain conditions due to the bounding nature of the applied plasticity correction factor, Ke. Whilst this overconservatism was tolerable in the past, recent technical challenges arising due to consideration of environmentally-assisted fatigue (EAF) and design for long-term operation have posed difficulty in achieving acceptable fatigue usage based on extant Code assessment procedures for certain components. The incorporation of more accurate Ke factors has since been identified as a nuclear industry priority. This paper presents a critical review of Ke factors within ASME Section III, with particular attention given to a recently proposed approach by Ranganath, which has recently been approved for publication as an ASME Section III Code Case. Correction factors adopted within other nuclear and non-nuclear codes and standards (C&S) were also considered. The code-based Ke factors were compared with Ke factors obtained directly from various elastic-plastic finite element (FE) models of representative plant components. The results revealed a considerable difference in conservatism between the code-based methods. Based on the elastic-plastic FE results, an alternative improved plasticity correction method was proposed. The need for a harmonised approach to determining Ke based on elastic-plastic FE analysis is discussed as a desirable industry objective.

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Section: Ranganath's Methodsmentioning

confidence: 59%

Section: Ranganath's Methodsmentioning

confidence: 99%

Critical Review of ASME III Plasticity Correction Factors for Fatigue Design-By-Analysis of Nuclear Power Plant Components

Clarkson

Bell

Mackenzie

2021

Journal of Pressure Vessel Technology

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“…According to the edition of RCCM 2002, The value of S alt takes the following value [4][5][6] : S alt =(K emech S pmech+ K etherm S ptherm )/2…”

Section: Fatigue Analysis Approach and Ke Modifiymentioning

confidence: 99%

Design Optimization and Fatigue Analysis for Seal Joint of Control Rod Drive Mechanism

Shao

Dujuan²

2018

MATEC Web Conf.

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Two kinds of seal joint thickness of control rod drive mechanism seal house were analyzed on stress and fatigue by using finite element method, and it provides analytical basis for equipment optimization, design manufacturing and problem handing. In the fatigue computing, all the transients were grouped into several groups, so as to separate the transients not at the same stage obviously. Meanwhile, the transients can be respectively combined in every group to reduce the redundant conservatism of computation. The elastoplastic strain correction factor was modified by analyzing thermal and mechanical load separately referring the rules of RCC-M 2002.By comparing the stress of two thickness joints, the results are found that the fatigue results of the thinner thickness are safer, but this is contrary to the rule of other results. The fatigue usage factor can be reduced by using modified elastoplastic strain correction factor when the amplify of the primary and secondary stress is large produced by both thermal load and mechanical load.

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Cumulative damage assessment of pressurised piping branch junctions under in-plane run pipe simulated seismic bending

Yahiaoui

Moffat

Moreton

1995

International Journal of Pressure Vessels and Piping

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Improvements on Fatigue Analysis Methods for the Design of Nuclear Components Subjected to the French RCC-M Code

Cited by 7 publications

References 7 publications

Critical Review of ASME III Plasticity Correction Factors for Fatigue Design-By-Analysis of Nuclear Power Plant Components

Critical Review of ASME III Plasticity Correction Factors for Fatigue Design-By-Analysis of Nuclear Power Plant Components

Design Optimization and Fatigue Analysis for Seal Joint of Control Rod Drive Mechanism

Cumulative damage assessment of pressurised piping branch junctions under in-plane run pipe simulated seismic bending

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