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Research Papers

Overview of Boiling Water Reactor Steam Dryer Alternating Stress Assessment Procedures

[+] Author and Article Information
Stephen A. Hambric

Fellow ASME
ARL/Penn State,
P.O. Box 30,
State College, PA 16804
e-mail: sah19@arl.psu.edu

Samir Ziada

Fellow ASME
Department of Mechanical Engineering,
McMaster University,
Hamilton, ON L8S 4L7, Canada
e-mail: ziadas@mcmaster.ca

Richard J. Morante

Brookhaven National Laboratory,
98 Rochester Street,
Upton, NY 11967
e-mail: morante@bnl.gov

Manuscript received June 9, 2017; final manuscript received September 1, 2017; published online March 5, 2018. Assoc. Editor: Jovica R. Riznic.The United States Government retains, and by accepting the article for publication, the publisher acknowledges that the United States Government retains, a nonexclusive, paid-up, irrevocable, worldwide license to publish or reproduce the published form of this work, or allow others to do so, for United States Government purposes.

ASME J of Nuclear Rad Sci 4(2), 021002 (Mar 05, 2018) (8 pages) Paper No: NERS-17-1059; doi: 10.1115/1.4037898 History: Received June 09, 2017; Revised September 01, 2017

The United States Nuclear Regulatory Commission (USNRC) has approved several extended power uprates (EPU) for Boiling Water Reactors (BWRs). In some of the BWRs, operating at the higher EPU power levels and flow rates led to high-cycle fatigue damage of Steam Dryers, including the generation of loose parts. Since those failures occurred, all BWR owners proposing EPUs have been required by the USNRC to ensure that the steam dryers would not experience high-cycle fatigue cracking. This paper provides an overview of BWR steam dryer design; the fatigue failures that occurred at the Quad Cities (QC) nuclear power plants and their root causes; a brief history of BWR EPUs; and a discussion of steam dryer modifications/replacements, alternating stress mechanisms on steam dryers, and structural integrity evaluation methods (static and alternating stress).

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References

NRC, 2017, “ Boiling Water Reactor (BWR) Systems,” U.S. Nuclear Regulatory Commission, Rockville, MD, accessed Sept. 23, 2017, www.nrc.gov/reading-rm/basic-ref/students/for-educators/03.pdf
Hambric, S. , Mulcahy, T. , Shah, V. , Scarbrough, T. , and Wu, J. , 2006, “ Acoustic Loading on BWR Steam Dryers Caused by Valve Singing,” 9th NRC/ASME Symposium on Valves, Pumps, and Inservice Testing, Washington, DC, July 17–19, Paper No. NUREG/CP-0152. https://www.nrc.gov/docs/ML0727/ML072700042.pdf
NRC, 2002, “ Failure of Steam Dryer Cover Plate After a Recent Power Uprate,”U.S. Nuclear Regulatory Commission, Washington, DC, NRC Information Notice 2002-26. https://www.nrc.gov/docs/ML0225/ML022530291.pdf
NRC, 2003, “ Additional Failure of Steam Dryer After a Recent Power Uprate,” U.S. Nuclear Regulatory Commission, Washington, DC, NRC Information Notice 2002-26, Suppl. 1.
NRC, 2004, “ Additional Flow-Induced Vibration Failures After a Recent Power Uprate,” U.S. Nuclear Regulatory Commission, Washington, DC, NRC Information Notice 2002 - 26, Suppl. 2.
GE Nuclear Energy, 2006, “ BWR Steam Dryer Integrity,” GE SIL No. 644, Rev. 2, GE Nuclear Energy, ADAMS Accession No. ML082530175. https://www.nrc.gov/docs/ML0500/ML050040048.pdf
Ziada, S. , and Lafon, P. , 2014, “ Flow-Excited Acoustic Resonance Excitation Mechanism, Design Guidelines, and Counter Measures,” ASME Appl. Mech. Rev., 66(1), p. 010802.
Baldwin, R. M. , and Simmons, H. R. , 1986, “ Flow-Induced Vibration in Safety Relief Valves,” ASME J. Pressure Vessel Technol., 108(3), pp. 267–272. [CrossRef]
Ruggles, A. , Moore, E. , Shehane, M. , Zhang, B. , and Sparger, J. , 2009, “ Side Branch Interaction With Main Line Standing Waves and Related Component Load Definition,” ASME Paper No. IMECE2009-12214.
DeBoo, G. , Ramsden, K. , Gesior, R. , and Strub, B. , 2007, “ Identification of Quad Cities Main Steam Line Acoustic Sources and Vibration Reduction,” ASME Paper No. PVP2007-26658.
ASME, 2013, “  Boiler and Pressure Vessel Code Section III—Rules for Construction of Nuclear Facility Components, Subsection NG, Design Fatigue Curves,” American Society of Mechanical Engineers, New York http://bsb.co.in/ASME_BPVC_2013/OUT/BPVC-III-1A_2013.pdf.
ASME, 2013, “  Boiler and Pressure Vessel Code Section III—Rules for Construction of Nuclear Facility Components, Appendix I, Design Fatigue Curves,” American Society of Mechanical Engineers, New York http://bsb.co.in/ASME_BPVC_2013/OUT/BPVC-III-1A_2013.pdf.
Ohtsuka, M. , Fujimoto, K. , Takahashi, S. , Hirokawa, F. , and Tsubaki, M. , 2006, “ Study on Acoustic Resonance and Its Damping of BWR Steam Dome,” International Congress on Advances in Nuclear Power Plants (ICAPP), Reno, NV, June 4--8, Paper No. 6186 https://inis.iaea.org/search/search.aspx?orig_q=RN:39042819.
Szasz, G. , Fujikawa, K. , and DeBoo, G. , 2006, “ Assessment of Steam Line Dynamic Pressures Using External Strain Gage Measurements,” ASME Paper No. PVP2006-ICPVT-11-93206.
Continuum Dynamics, 2008, “  Acoustic and Low Frequency Hydrodynamic Loads at CLTP Power Level on Browns Ferry Nuclear Unit 1 Steam Dryer to 250 Hz, Rev. 1,” Continuum Dynamics, Inc., Ewing, NJ, Report No. 08-04NP, Document No. ML081750084. https://www.nrc.gov/docs/ML0817/ML081750084.pdf
Forsyth, D. , and Longoni, G. , 2010, Solving the Steam Dryer Degradation Problem, Westinghouse Engineering Energy Magazine.
Seybert, A. F. , and Ross, D. F. , 1977, “ Experimental Determination of Acoustic of Acoustic Properties Using a Two Microphone Random Excitation Technique,” J. Acoust. Soc. Am., 61(5), pp. 1362–1370. [CrossRef]
Seybert, A. F. , 1988, “ Two-Sensor Methods for the Measurement of Sound Intensity and Acoustic Properties in Ducts,” J. Acoust. Soc. Am., 83(6), pp. 2233–2239. [CrossRef]
Abom, M. , and Boden, H. , 1988, “ Error Analysis of Two-Microphone Measurements in Ducts With Flow,” J. Acoust. Soc. Am., 83(6), pp. 2429–2438. [CrossRef]
Karplus, H. B. , 1961, “ Propagation of Pressure Waves in a Mixture of Water and Steam,” Armour Research Foundation of Illinois Institute of Technology, Chicago, IL, United States Atomic Energy Commission Contract No. AT (11-1) 528, Report No. D132A13.
Petr, V. , 2004, “ Wave Propagation in Wet Steam,” Proc. Inst. Mech. Eng., Part C, 218(8), pp. 871–882. [CrossRef]
ASME, 2013, “ Boiler and Pressure Vessel Code Section III—Rules for Construction of Nuclear Facility Components, Appendix N, Dynamic Analysis Methods,” American Society of Mechanical Engineers, New York http://bsb.co.in/ASME_BPVC_2013/OUT/BPVC-III-1A_2013.pdf.
NRC, “ Comprehensive Vibration Assessment Program for Reactor Internals During Preoperational and Startup Testing.” U.S. Nuclear Regulatory Commission, Washington, DC, Regulatory Guide 1.20 https://www.federalregister.gov/documents/2017/02/13/2017-02864/comprehensive-vibration-assessment-program-for-reactor-internals-during-preoperational-and-startup.
NRC, 2013, “ Programs for Monitoring Boiling-Water Reactor Steam Dryer Integrity,” U.S. Nuclear Regulatory Commission, Washington, DC, NRC Information Notice 2013 - 10, ADAMS Accession No. ML13003A049. https://www.nrc.gov/docs/ML1300/ML13003A049.pdf

Figures

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Fig. 4

Instrumented QC2 replacement steam dryer with pressure transducers on the hood (top) and skirt (bottom). From source material previously published in Ref. [2].

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Fig. 5

Example of steam dryer surface pressure spectra, left−hood; right−skirt. From source material previously published in Ref. [2].

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Fig. 3

Cross-sectional views of steam dryer hood regions (from Ref. [6]). The RPV walls and MSL inlets are to the left of the figure.

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Fig. 2

Schematic and photos of cracks and subsequent loose parts in original QC dryers (from Ref. [6])

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Fig. 1

Top—Cutaway of BWR RPV and steam dryer (left), and curved hood dryer (right); Bottom—schematic of typical original BWR steam dryer: assembly (left), single panel (right). From Ref. [1,2].

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Fig. 6

Typical flow rates within an RPV (top) and low-order RPV steam volume acoustic resonances (bottom); analyses of a BWR nuclear power plant using FLUENT compressible flow computational fluid dynamics analysis. From source material previously published in Ref. [2].

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Fig. 7

MSL strain gage arrays (left) and typical individual and summed “hoop” strain, from QC2. From source material previously published in Ref. [2].

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