Abstract
Critical equipment of nuclear power plants and petrochemical industries is sometimes subjected to both creep and fatigue loading simultaneously. Under combined creep–fatigue loading, the creep deformation affects the fatigue behavior of the material depending on the relative duration of stress relaxation due to creep within service life. In the present paper, evaluation of creep and fatigue damage is carried out for a process reactor using elastic analysis method of ASME-NH code. The reactor material is 2.25Cr-1Mo steel. The stress evaluations are carried out at outlet nozzle where stresses are observed to be maximum. Effect of three parameters, that is, the maximum hold temperature, the duration of hold time at highest temperature and the rate of temperature change on combined creep–fatigue damage, is studied. This work provides guidelines for performing creep–fatigue analysis for similar pressure components.
This is a preview of subscription content, log in via an institution to check access.
Access this chapter
Tax calculation will be finalised at checkout
Purchases are for personal use only
Abbreviations
- D :
-
Total creep–fatigue damage
- K :
-
Local geometric concentration factor
- K e :
-
Stress ratio factor at yield
- K t :
-
Factor for reduction in extreme fiber bending stress due to effect of creep
- K v :
-
Multiaxial plasticity and Poisson ratio adjustment factor
- P b :
-
Primary bending equivalent stress
- P L :
-
Local primary membrane equivalent stress
- P m :
-
Primary membrane equivalent stress
- (Q R )max :
-
Maximum secondary stress range
- S*, \( \bar{S} \) :
-
Stress indicators
- S alt :
-
Alternating stress intensity
- S j :
-
Initial stress
- S m :
-
Allowable stress
- S rH :
-
Hot relaxation strength
- S t :
-
Temperature and time-dependent stress intensity limit
- S y :
-
Yield strength of material
- X :
-
Primary stress parameter
- Y :
-
Secondary stress parameter
- Z :
-
Dimensionless effective creep stress parameter
- ∆Ɛ max :
-
Maximum equivalent strain range
- ∆Ɛ mod :
-
Modified maximum equivalent strain range
- ∆Ɛ c :
-
Creep strain increment
- Ɛ t :
-
Total strain range
- σ c :
-
Effective creep stress
- ∆t :
-
Duration of time interval
References
Plumbridge WJ, Dean M, Sand Miller DA (1982) The importance of failure mode in fatigue-creep interactions. In: Fatigue of engineering materials and structures, vol 5, pp 101–114
Hormozi R, Biglari F, Nikbin K (2015) Experimental and numerical creep-fatigue study of type 316 stainless steel failure under high-temperature LCF loading condition with different hold time. Eng Fract Mech 141:19–43
Jawad MH, Jetter RI (2009) “Creep-fatigue analysis” in design and analysis of ASME Boiler and pressure vessel components in the creep range. ASME Press, New York, pp 151–176
Koo GH, Yoo B (2000) Elevated temperature design of KALIMER internals accounting for creep and rupture effects. J Korean Nucl Soc, vol 32, pp 66–594
Gurumurthy K, Balaji S et al (2014) Creep-fatigue design studies for process reactor components subjected to elevated temperature service as per ASME-NH. Procedia Eng 86:327–334
Fournier B, Sauzay M et al (2008) Creep-fatigue oxidation interactions in 9Cr-1Mo martensitic steel. Part III: lifetime prediction. Int J Fatigue 30:1797–1812
AFCEN (2002) Design and construction rules for mechanical components of FBR Nuclear Islands. RCC-MR, 2002 edn. AFCEN
Oldham J, Abou-Hanna J (2011) Numerical investigation of creep-fatigue life prediction utilizing hysteresis energy as a damage parameter. Int J Pressure Vessels Piping 88:149–157
ASME Boiler and Pressure Vessel Code (2015) Section III, Subsection NH
Manson SS, Halford GR (2009) Fatigue and durability of metals at high temperatures. ASM International, Materials Park, Ohio
Guodong ZB, Yanfen Z et al (2011) Creep-fatigue interaction damage model and its application in modified 9Cr–1Mo steel. Nucl Eng Des 241:4856–4861
Fan ZC, Chen XD et al (2009) A CDM-based study of fatigue-creep interaction behavior. Int J Pressure Vessels Piping 86:628–632
Aoto K, Komine R et al (1994) Creep-fatigue evaluation of normalized and tempered modified 9Cr-1Mo. Nucl Eng Des 153:97–110
ASME Boiler and Pressure Vessel Code (2015) Section VIII, Division 1
ANSYS Workbench (2015) Release 14.5, ANSYS Inc.
ASME Boiler and Pressure Vessel Code (2015) Section II, Part D
ASME Boiler and Pressure Vessel Code (2015) Section VIII, Division 2
Koo GH, Lee JH (2006) High-temperature structural integrity evaluation method and application studies by ASME-NH for the next generation reactor design. J Mech Sci Technol 20:2061–2078
Author information
Authors and Affiliations
Corresponding author
Editor information
Editors and Affiliations
Appendix: Flowchart for Creep and Fatigue Damage Evaluation
Appendix: Flowchart for Creep and Fatigue Damage Evaluation
Rights and permissions
Copyright information
© 2018 Springer Nature Singapore Pte Ltd.
About this paper
Cite this paper
Dukare, S.R., Raykar, N.R. (2018). Creep–Fatigue Damage Evaluation of 2.25Cr-1Mo Steel in Process Reactor Using ASME-NH Code Methodology. In: Seetharamu, S., Rao, K., Khare, R. (eds) Proceedings of Fatigue, Durability and Fracture Mechanics. Lecture Notes in Mechanical Engineering. Springer, Singapore. https://doi.org/10.1007/978-981-10-6002-1_21
Download citation
DOI: https://doi.org/10.1007/978-981-10-6002-1_21
Published:
Publisher Name: Springer, Singapore
Print ISBN: 978-981-10-6001-4
Online ISBN: 978-981-10-6002-1
eBook Packages: EngineeringEngineering (R0)