• Journal of Ocean Engineering and Technology
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• v.35 no.2
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• pp.161-171
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• 2021

For the fatigue design of offshore structures, it is essential to understand and use the S-N curves specified in various industry standards and codes. This study compared the characteristics of the S-N curves for five major codes. The codes reviewed in this paper were DNV Classification Rules (DNV GL, 2016), ABS Classification Rules (ABS, 2003), British Standards (BSI, 2015), International Welding Association Standards (IIW, 2008), and European Standards (BSI, 2005). Types of stress, such as nominal stress, hot-spot stress, and effective notch stress, were analyzed according to the code. The basic shape of the S-N curve for each code was analyzed. A review of the survival probability of the basic design S-N curve for each code was performed. Finally, the impact on the conservatism of the design was analyzed by comparing the S-N curves of three grades D, E, and F by the five codes. The results presented in this paper are considered to be a good guideline for the fatigue design of offshore structures because the S-N curves of the five most-used codes were analyzed in depth.

• Journal of Welding and Joining
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• v.32 no.1
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• pp.71-78
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• 2014

Although the fatigue design curve of ASME Code has enough margin with respect to alternating stress and cycles, the welding residual stress(WRS) should be included in fatigue analysis. In this paper, WRS distribution in a nozzle with dissimilar metal weldment was obtained by finite element analysis and was added in fatigue analysis. The fatigue analysis was performed by following the ASME Code including thermal and stress analysis applying with postulated 30 transient conditions. The calculated results of a cumulative fatigue usage factors(CUF) were compared for the case of the models with or without WRS effects. The results showed that the CUF at weldment and heat affected zone was affected by the WRS.

• Journal of the Korean Society of Safety
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• v.16 no.4
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• pp.65-73
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• 2001

This paper is a study on the fatigue analysis using BS5400 Part10(1980), code of practice for fatigue for a double link type level luffing jib crane which has been operated since the crane was built in 1985 at a domestic pier. South Korea. In addition, on the basis of the design lift due to fatigue analysis and the number of cracks detected from the nondestructive test the structural reliability and the residual life of the crane is predicted by evaluating the expected development of detectable fatigue oracle during the next five to ten years using Paris's Law for predicting fatigue crack growth and Gaussian probability density function to be reasonable for stress ranges below the mean values determined by laboratory tests. The statistical data used for the analysis of the structural reliability and life expectancy is given in the above referred code.

• Nuclear Engineering and Technology
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• v.52 no.12
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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.

• Nuclear Engineering and Technology
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• v.53 no.7
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• pp.2407-2417
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• 2021

A structural integrity evaluation program (STEP) was developed for the high temperature reactor design evaluation according to the ASME Boiler and Pressure Vessel Code (ASME B&PV), Section III, Rules for Construction of Nuclear Facility Components, Division 5, High Temperature Reactors, Subsection HB. The program computerized HBB-3200 (the design by analysis procedures for primary stress intensities in high temperature services) and Appendix T (HBB-T) (the evaluation procedures for strain, creep and fatigue in high temperature services). For evaluation, the material properties and isochronous curves presented in Section II, Part D and HBB-T were computerized for the candidate materials for high temperature reactors. The program computerized the evaluation procedures and the constants for the weldment. The program can generate stress/temperature time histories of various loads and superimpose them for creep damage evaluation. The program increases the efficiency of high temperature reactor design and eliminates human errors due to hand calculations. Comparisons that verified the evaluation results that used the STEP and the direct calculations that used the Excel confirmed that the STEP can perform complex evaluations in an efficient and reliable way. In particular, fatigue and creep damage assessment results are provided to validate the operating conditions with multiple types of cycles.

• Nuclear Engineering and Technology
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• v.33 no.6
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• pp.566-584
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• 2001

In this paper, the progressive deformation and the creep-fatigue damage for the conceptually designed reactor internals of KALIMER(Korea Advanced Liquid MEtal Reactor) are carried out by using the elastic analysis method in the RCC-MR code for normal operating conditions including the thermal load, seismic load (OBE) and dead weight. The maximum operating temperature of this reactor is 53$0^{\circ}C$ and the total service lifetime is 30 years. Thus, the time- dependent creep and stress-rupture effects become quite important in the structural design. The effects of the thermal induced membrane stress on the creep-fatigue damage are investigated with the risk of the elastic follow-up. To calculate the thermal stress, detailed thermal analyses considering conduction, convection and radiation heat transfer mechanisms are carried out with the ANSYS program. Using the results of the elastic analysis, the progressive deformation and creep-fatigue damages are calculated step by step using the RCC-MR in detail. This paper ill be a very useful guide for an actual application of the high temperature structural design of the nuclear power plant accounting for the time-dependent creep and stress-rupture effects.

• Journal of Welding and Joining
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• v.35 no.1
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• pp.61-67
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• 2017

Recent studies on the fatigue assessment of high strength steel weld based on the fracture mechanics have frequently raised the problems related to the conservatism in the fatigue crack growth rate specified in the relevant design code. The purpose of this study is to evaluate the effect of the fatigue crack growth parameter on the fatigue life for the low carbon steel weld. In order to do it, the fatigue tests with the constant stress ratio were performed to evaluate the fatigue crack growth rate in the butt weld of SM490. And the fatigue crack growth parameters of the weld were evaluated in accordance with ASTM E647. From the comparative fatigue assessment results, it was found that the fatigue crack growth rate specified in the relevant design code was too conservative to estimate the residual fatigue life of welded structure. So, in order to get the more reliable results, it was recommended that the fatigue life estimation based on the fracture mechanics be performed with the fatigue crack growth parameter specified by test.

• Wind and Structures
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• v.3 no.1
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• pp.59-67
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• 2000

The cruciform shaped mast over 47 storey, Telecom Corporate Building in Melbourne, Australia rises to a height of approximately 25 m above the roof level. As the members are subjected to very high fluctuating loads under wind, the design was mainly governed by wind-induced fatigue. A detailed fatigue analysis was carried out according to the requirements of the Australian Steel Structures Code, AS4100. The wind-induced fatigue analysis procedure is described in the paper. The fatigue design of this mast is used as an example to illustrate some potential problems of relevant specifications in AS4100 and to outline some of the more important parameters in the fatigue analysis.

• Proceedings of the Korea Concrete Institute Conference
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• 2006.05a
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• pp.394-397
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• 2006

For checking fatigue safety and endurance of precast concrete deck and loop joint system in the steel plate girder composite bridge, the test composite bridge model was made for the fatigue experiment by the wheel load machine. The fatigue tests of 1,000,000 cycles were implemented according to wheel load condition of DB24 rear axle of Korea Highway Design Code. From the test results, the loop joint system for the precast deck has a sufficient flexural capacity. Although a little lower longitudinal continuity capacity is evaluated than general sound cast-in-place RC bridge deck, there is no problem about the safety. The overall fatigue level of safety defined by the code is satisfied.

• Journal of the Korean Society of Safety
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• v.33 no.5
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• pp.164-169
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• 2018

A detailed fatigue evaluation procedure was developed to mitigate the excessive conservativeness of the conventional environmental fatigue evaluation method for the pressurizer spray line elbow of domestic new nuclear power plants. The pressurizer spray line is made of austenitic stainless steel, which is relatively sensitive to the environmentally assisted fatigue, and has a low degree of design margin in terms of environmentally assisted fatigue due to the thermal stratification phenomenon on the pipe cross section as a whole or locally. In this study, to meet the environmental fatigue design requirements of the pressurizer spray line elbow, the new environmental fatigue evaluation has been performed, which used the ASME Code NB-3200-based detailed fatigue analysis and the environmental fatigue correction factor instead of the existing NB-3600 evaluation method. As a result, the design requirements for environmentally assisted fatigue were met in all parts of the pressurizer spray line elbow including the fatigue weakened zones by thermal stratification.