Abstract
In this paper, the stress analysis of the support structure (skirt) of a Boiling Water Reactor (BWR-5) vessel is presented. Three transient loading conditions, which take place during the life of the plant, were evaluated, in order to determine the structural integrity under fatigue loading. The concept of Cumulative Usage Fatigue (CUF) was used. The guidelines of the subsection NB and subsection NF of Section III of the ASME Code were taken into account in the evaluation in conjunction with the Finite Element Analyses (FEA). In addition, the environmental fatigue penalty factors (Fen) were also considered during the evaluation. The environmental conditions affected the integrity of the skirt. Besides, a perturbation in the stresses was observed due to the dynamic forces generated by the jet pumps, which increased the CUF in the skirt.
Access this chapter
Tax calculation will be finalised at checkout
Purchases are for personal use only
Abbreviations
- BWR:
-
Boiling Water Reactor
- CUF:
-
Cumulative Usage Factor
- FEA:
-
Finite Element Analysis
- NPP:
-
Nuclear Power Plant
- FEM:
-
Finite Element Method
- PWR:
-
Pressure Water Reactor
- PVR:
-
Pressure Vessel Reactor
- AEF:
-
Assisted Environmental Fatigue
- ppm:
-
Parts per million
- DO:
-
Dissolved Oxygen
- CS:
-
Carbon Steel
- LAS:
-
Low Alloy Steel
References
NUREG/CR-2940, Greene, S.R.: Realistic simulation of severe accidents in BWRs computer modeling requirements. ORNL/TM8517, April 1984
McManus, J.P.: Thermal stress analysis of pressure vessels with cylindrical skirt supports. Thesis and Dissertation Collection, Calhoun: The NPS Institutional Archive, Dudley Knox Library, Monterey, California, USA, 93943, 1964
Cruz, J.R.B., Mattar, M.N., de Miranda, C.A.J., Bezerra, L.M.: Aspects of design and stress classification of a PWR support structure. Structural Mechanics Division, COPES/IPEN-CNEN/SP, Sao Paulo, Brazil
Wilkie, D.: Reactor vessel support skirt concerns. Crowd Sourced Information & Analysis Without Focus On Profit, SimplyInf.Org, consulted in November 2016
Mkrtchyan, L., Schau, H., Wolf, W., Holzer, W., Wernicke, R., Trieglaff, R., Stress analyses for reactor pressure vessels by the example of a product line ’69 Boiling Water Reactor. Kerntechnik, Agosto 2011. https://doi.org/10.3139/124.110173
Hodge, S.A., Ott, L.J.: BWR reactor vessel head failure modes. Boiling Water Reactor Severe Accident Technology (BWRSAT) Program, Oak Ridge National Laboratory, Oak Ridge, Tennessee, 10 de Mayo 1989
NUREG/CR-6260 INEL-95/0045, Ware, A.G., Morton, D.K., Nitzel, M.E.: Application of NUREG/CR-5999 Interim fatigue curves to selected nuclear power plants components. Idaho Technologies Company, February 1995
EPRI 1024995, Chu, S.: Environmentally assisted fatigue screening. Process and technical basis for identifying EAF limiting locations, Final report. Palo Alto, California, August 2012
ASME Boiler and Pressure Vessel Code, Section III, Division 1—Subsection NB, Class I Components: Rules for construction of nuclear facility components. American Society of Mechanical Engineers, 2010
NUREG-1801, Revision 2: Generic Aging Lessons Learn (GALL) Report. U.S. Nuclear Regulatory Commission, December 2010
NUREG/CR-5704 (ANL-98/31): Effects of LWR coolant environments on fatigue design curves of austenitic stainless steels, 1999
NUREG/CR-6583 (ANL-97/18): Effects of LWR coolant environments on fatigue design curves of carbon and low-alloys steels, 1998
NUREG/CR-6909 (ANL-06/08): Effects of LWR coolant environments on the fatigue life of reactor materials, Final report, February 2007
Ranganath, S., Carter, B.: Simplified methods for including environmental effects in ASME code fatigue analysis for BWR2, XGEN engineering, Electric Power Research Institute, U.S., NRC Office, Washington, DC, March 2010
Poloski, J.P., Grant, G.M., Gentillon, C.D., Galyean, W.J., Roesener, W.S.: Reactor core isolation cooling system reliability, 1987–1993, NUREG/CR-5500, Idahi National Engineering Laboratory, Nuclear Risk Management Technologies Department, Lockheed Martin Idaho Technologies Company, Idaho Falls, Idaho 83415, June, 1997
Cuahquetzi, N.M., Gómez, L.H.H., López, P.R., Calderón, G.U., Fernández, J.A.B., Sosa, G.U., Cuamatzi, E.F., Ramírez, A.O., Ángeles, B.R.: “Evaluation of the structural integrity of the jet pumps of a Boiling Water Reactor under hydrodynamic loading”, Defect and diffusion forum, vol. 348, pp. 261–270, Trans Tech Publications, Switzerland (2014)
Acknowledgements
The grant for the development of the project 211704 awarded by the National Council of Science and Technology (CONACYT) is kindly acknowledged.
Author information
Authors and Affiliations
Corresponding author
Editor information
Editors and Affiliations
Rights and permissions
Copyright information
© 2019 Springer International Publishing AG, part of Springer Nature
About this chapter
Cite this chapter
Grijalba, Y.L. et al. (2019). Evaluation of the Structural Integrity of a Boiling Water Reactor Skirt Under Stationary and Transient Loading Conditions. In: Öchsner, A., Altenbach, H. (eds) Engineering Design Applications. Advanced Structured Materials, vol 92. Springer, Cham. https://doi.org/10.1007/978-3-319-79005-3_31
Download citation
DOI: https://doi.org/10.1007/978-3-319-79005-3_31
Published:
Publisher Name: Springer, Cham
Print ISBN: 978-3-319-79004-6
Online ISBN: 978-3-319-79005-3
eBook Packages: EngineeringEngineering (R0)