ML20063H743

Continuous Ae Crack Monitoring of a Dissimilar Metal Weldment at Limerick Unit 1

ML20063H743
Person / Time
Site:	Limerick
Issue date:	12/31/1993
From:	Dawson J, Friesel M, Hutton P Battelle Memorial Institute, PACIFIC NORTHWEST NATION
To:	NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
References
CON-FIN-B-2913, CON-FIN-L-1100 NUREG-CR-5963, PNL-8844, NUDOCS 9402220168
Download: ML20063H743 (100)

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NUREG/CR-5963 PNL-8844.

R5,GS Continuous AE Crack Monitoring l of A Dissimilar Metal Weldment

. at Limerick Unit 1 k

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Manuscript Completed: October 1993 l Date Published: December 1993 f~

Prepared by P. H. Hutton, M. A. Friesel, J. F. Dawson l

Pacific Northwest Laboratory l Richland, WA 99352 a

i Prepared for Division of Engineering Omce of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 NRC FINS B2913 and L1100

1

'; Abstract J

Acoustic emission (AE) technology for continuous Evaluation of the flaw indication showed that it could surveillance of a reactor component (s) to detect crack remain in place during the subsequent fuel cycle with-initiation and/or crack growth has been developed at out compromising safety. The existence of this flaw Pacific Northwest Laboratory (PNL), operated by indication offered a long sought opportunity to validate Battelle MemorialInstitute, under support from the AE surveillance to detect and evaluate crack growth U. S. Nuclear Regulatory Commission, Office of Nucle- during reactor operation. Through the cooperation and ar Regulatory Research (U. S. NRC-RES). The tech- support of ?ECO and the U. S. NRC-RES, AE instru-k, nology was validated off-reactor in several major tests, mentation was installed by PNL and PECO under but it had not been validated by monitoring crack PECO M< d. No. 043-002 to monitor the flaw indication growth on an operating reactor system. A flaw indica- during two complete fuel cycles. This report discusses tion was ientified during normal inservice inspection of the results obtained from the AE monitoring over the l piping at Philadelphia Electric Company (PECO) Lim- period May 1989 to March 1992 (two fuel cycles).

erick Unit i reactor during the 1989 refueling outage.

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3 iii NUREG/CR-5%3 I

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i Contents Abst ract . . . . . . . . . . . . . . . . . . . . . . . . . . .. ..........................................m 1

l L Executive Summary . . . . . . . . . . . . . . . . . . ... ...... ........ ..... . . . . . . . . . .......... ix I Previous Reports in Series . . . . . . . . . . . . . . . . . . . ............................................,xi Acknowledgments . . . . . . . . . . . . . . . . . . . . . . . . . ............................................ xiii 1.0 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . .. . .. . . . . . . . . . . . . . . . . . . . . . . . 1.1 2.0 AE System Installation and Calibration . . . . . . . .. .... .... ...... .... ... .. ......... . ... 2.1 2.1 Preparation . . . . . . . . . . . . . . . . . . . . . . ....... . . . . . . . . . . . . .. . . . .. . . . . . . . . . . . . . . . . . 2.1 l 2.2 AE System 1cstallation . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . .. . .. . . . . ... . . . . . . . 2.2

! 23 System Qualification . . . . . . . . . . . . . ...........................................23 2.4 Pr oble m s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 2.5 D ata Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.5 2.6 S um m ary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.5 l

3.0 Int erim D ata Res ults . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1 3.1 AE System Performance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1 l

3.2 AE Data Analysis . . . . . . . . . . . . . . . .... ...... .. ........... . . . . . . . . . ... . ... .. 3.1 33 Flaw Indication from ISI Results and Crack-Arrest-Verification-Specimen Indications . . . . . . . . . . . . 3.2 3.4 Correlation between AE, CAVS, and UT Results . . . . . . . . . . . . . . . . .................... . 33

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4.0 Final Monitoring Period . . . . . ............ .......... . . . . . . . .. . . . . . . . 4.1 4.1 AE Data Analysis . . . . . . . . . . . . . .... . .......... ....... . . . . . . . . . . . . . . . . . . 4.1 4.2 ISI and Crack. Arrest-Verification-Specimen Results . . . . ................ . . . . . .. . . . . . . . 4.1 43 Comparison of AE, ISI, and CAVS Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.2 4.4 AE System Performance . . . . . . . . ....... . . .. . . . . . . . . . . . . . . .. .. . . . . . .. . . . . . . . 4.2 5.0 Observations for Future Monitoring . . . . . . . .. .. ................. . . . . . . . . . . . . .. . . . . . . . 5.1 5.1 Reference Monitor Point . . . . . . . . . . . ....... ...................... . . . . . . . . . . . . . . 5.1 1 5.2 Containment Penetration .

. . . . . . ... ..... . .......... . . . . . . . . . . . . . . . . . .. . 5.1 l 53 Monitor Instrument Location . . . .. ... . .. ., ... . . . . . . . . . . . . . . . . . . . . . . 5.1 5.4 Calibration of the AE Source Location . .... , ... . ... . . . . . . .. . . . . . . . . . . . . . 5.1 5.5 AE Signal Identification . . . . . . . .... ..... ..... .............. . . . . . . . . . . . 5.1 5.6 Digitized AE Information . . . . . . . . . ............. ....... . . . .. . . . .. . . . 5.1 6.0 Summary and Conclusions . . . . . . . . .. . .................. ... . . . . . . . ... . . . . . . . 6.1 7.0 References . . . . . . . . . . . . . . . . . . . . . . . ...... .................... . . . . . . ...... .. .. 8.1 i

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Figures 2.1. Construction of the Waveguide AE Sensors . .. .... .................... .. ....... 2.6 2.2. Sensitivity Test of Waveguide AE Sensors on Calibration Block . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.7 p.

23. Location of AE Sensors on Pipe Weld NRR-1RD-1A N2H Limerick Unit 1. . . . . . . . . . . . . . . . . . . . 2.8 2.4. Waveguide AE Sensors Mounted Near Weld VRR 1RD-1A N2H - Limerick Unit 1. . . . . . . . . . . . . . . 2.9 2.5. Waveguide AE Sensor Being Pressure Coupled to the Pipe Surface - Limerick Unit 1 . . . . . . . . . . . . 2.10 2.6. Electronic Pulser Mounted Near AE Sensor No. 3 - Limerick Unit 1. . . . . . . . . . . . . . . . . . . . . . . . . 2.11

2.7. Waveguide AE Sensors Emerging from Weld Region - Limerick Unit 1. . . . . . . . . . . . . . . . . . . . . . . 2.12 2.8. Waveguide AE Sensor Head . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.13

, 2.9. AE System Preamplifiers Mounted in Junction Box in Reactor Enclosure, Elevation 283' - Limerick Unit 1.......................................................................2.14 1 2.10. AE Monitor Instrument Mounted in a Cabinet in Reactor Enclosure, Elevation 283' - Limerick Unit 1...........................................................................115 1

, 2.11. Test of Signal Source Location by AE System Using Signal Input from 0.5-mm Pencil Lead Breaks -

Limerick Unit 1. . . ...........................................................2.16 2.12. Sensitivity Test of Waveguide AE Sensors Installed on Reactor Pipe - Limerick Unit 1. . . . . . . . . . . . 2.17 2.13. Response of AE Sensors to Pulser Signal on 6/9/89 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.18

! 2.14. Lead Break Data Analyzed with Three-Sensor Algorithm . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2.19 f 3.1. AE Source Location Elements Superimposed on Pipe Cross Section with Flaw Profile - N2H Weld at j Lim e rick Unit 1 R eact or . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.4 4

3.2. AE Rate - Crack Growth Rate Relationship for Cyclic Fatigue Crack Growth (Hutton 1985) . . . . . . . . 3.5

. 33. Crack Growth Over the Period 5/89 to 9/90 as Indicated by AE - N2H Weld at Limerick Unit 1 Reactor.......................................................................3.6 i

3.4. Crack Growth Values Derived from AE Data over the Period 8/89 to 9/90 - Shown on Pipe Crack Section N2H Weld at Limerick Unit 1 Reactor . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7 3.5. UT Detection Threshold (20% of Wall) Superimposed on Crack Growth Prediction from AE Data -

N2H Weld at Limerick Unit 1 Reactor . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , 3.8 4.1. Crack Growth Values 2 0.1* Estimated from AE Data over the Period 5/89 to 9/90 N2H Weld at Lim e rick Unit 1 R eact or . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.4 4.2. Crack Growth Values 2 0.1" Estimated from AE Data over the Period 12/90 to 12/91 - N2H Weld at Lim erick Unit 1 Reactor . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -. 4.5 NUREG/CR-5%3 vi

L Figures  ;

4.3. Crack Growth '!alues 2 0.1" Estimated from AE Data over the Full AE Monitoring Period 5/89 to 12/91 - N2H Weld at Limerick Unit 1 Reactor . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . 4.6 - ,

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4.4. AE Source location Elements Superimposed on Pipe Cross Section with Flaw Profilm - N2H Weld at Limerick Unit 1 React or . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .' . . . . . . . . . . . . . 4.7 4.5. Plaw Indication Plot - N2H Weld at Limerick Unit 1 Reactor . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.8 4.6. OverallIndication Plot Sheet ........................................................'4.9 4.7. 60' RL Indication 2 - 34.89" to 44.9" .................................................4.10 4.8. 60* RL Indication 3 - 34.89" to 44.9" . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.11 4.9. Indication Plot Sheet for 0" to 7.5" . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.12 .

4.10. 45 RL Indication 1 - 7.85" to 11.776" . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.13 4.11. 60 RL Indication 1 5.13" to 5.921" . . . . . . . . . . . . . . . . . . . . . ............................4.14.

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i Tables 2.1. Comparison of Watts Bar and Limerick AE Installations . .. ... . ....... .... .............. 2.4 3.1. AE Event Count by Source location Element - 5/12/89 10/6/89 - N2H Weld at Limerick Unit 1 R e a ctor . . . . . . . . . . . . . . . . . . ....................................................3.9 3.2. AE Event Count by Source Location Element - 10/6/89 - 6/4/90 - N2H Weld at Limerick Unit 1 Reacor................................................,.....................3.11 1 i

33. AE Event Count by Source location Element - 6/4/90 - 9/11/90 - N2H Weld at Limerick Unit 1 )

Reactor . . . . . . . . . . . .......................................................... 3.13 3.4. Fihered AE Event Count by Source Location Element - 5/12/89 - 10/6/89 - N2H Weld at Limerick Unit 1 Reactor . . . .............................................................3.15 3.5. Filtered AE Event Count by Source Location Element - 10/6/89 - 6/4/90 N2H - Weld at Limerick Unit 1 Reactor . . ......... ............................ .......................3.17 3.6. Filtered AE Event Count by Source Location Element - 6/4/90 - 9/11/90 - N2H Weld at Limerick Unit 1 R e actor . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.19 3.7. Crack Growth Predicted from AE Data 5/12/89 to 10/6/89 - N2H Weld at Limerick Unit 1 R e a ct or . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.21 3.8. Crack Growth Predicted from AE Data - 10/6/89 to 6/4/90 - N2H Weld at Limerick Unit 1 Reactor . 3.23 3.9. Crack Growth Predicted from AE Data - 6/4/90 to 9/11/90 - N2H Weld at Limerick Unit 1 Reactor . 3.25 3.10.

Indicated Crack Growth Rate at the N2H Weld Over Period 5/89 to 9/90 . . . . . . . . . . . . . . . . . . . . . 3.26 4.1. Filtered AE Event Count by Source Location Element - 12/13/90 to 4/27/91 - N2H Weld at Limerick Unit 1 Reactor . .............. .........................................4.15 4.2. Filtered AE Event Count by Source Location Element - 4/28/91 to 8/9/91 - N2H Weld at Limerick Unit 1 Reactor .... .... ...... ... ............... ..........................4.17

43. Filtered AE Event Count by Source Location Element - 8/9/91 to 12/18/91 - N2H Weld at Limerick Unit 1 Reactor . . . . . .. .............. ......................................... 4.19 4.4.

Crack Growth Predicted from AE Data - 12/13/90 to 4/27/91 - N2H Weld at Limerick Unit 1 Reactor . . ... .......................................................... . . .. . 4.21 4.5.

Crack Growth Predicted from AE Data - 4/28/91 to 8/8/91 - N2H Weld at Limerick Unit 1 Reactor . 4.23 4.6.

Crack Growth Predicted from AE Data - 8/9/91 to 12/18/91 Plus Totals for 5/12/89 to 12/18/91 -

N2H Weld at Lim erick Unit 1 Reactor . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.25 NUREG/CR-5%3 viii

2 Executive Summary 4

This program, with the objective of validating the appli- an operating reactor plant. The system was installed, cation of acoustic emission (AE) monitoring to reactor calibrated. and maintained without causing any pertur-components to detect crack initiation and growth during bations in te reactor schedule. AE and ultrasonic ISI reactor operation, was started in March 1989 at Phila- results gene. ally agreed relative to growth of the flaw delphia Electric Co. (PECO) Limerick Unit 1 Generat- indication rbserved in the N2H weld. Both showed ing Station in Pennsylvania. At that time, PECO made limiid giowth during the first monitoring period (fuel the decision to apply AE technology developed by Pa- cycle) and no growth during the second period (fuel cific Northwest Laboratory (PNL), operated by Battelle cycle). The relationship identified for relating AE to MemorialInstitute, to monitor a flaw indication detect- crack growth produced rational results which were ed in a recirculation nozzle-to-safe end weld during similar in magnitude to that indicated by the ISI.

routine inservice ultrasonic inspection. The technology was developed over several years under a major pro- AE indicated limited flaw growth in an area adjacent to i gram supported by the U. S. Nuclear Regulatory Com- the ISI flaw indication which was not confirmed by ISI.

mission, OfIice of Nuclear Regulatory Research Inspection using an advanced technique which has (NRC-RES). The first monitoring period (fuel cycle) at greater sensitivity such as Synthetic Aperture Focusing Limerick was supported by funding from PECO with Technique for ultrasonic testing may provide a resolu-instrumentation supplied by NRC-RES from the PNL tion of the validity of the indication.

research program. The second monitoring period (fuel cycle) was funded jointly by PECO and NRC-RES. Several chervations were noted in the course of the Installation and operation of the AE system has been a effort which will contribute to enhanced effectiveness of joint PECO-PNL cffort with PNL performing the data future AE monitoring of reactor components. Future analysis function. The guidelines of American Society applications of AE monitoring will also benefit from the of Mechanical Engineers (ASME) Code Case N-471, transfer of technology to a commercial company (ies)

, " Acoustic Emission for Successive Inspections, Section that is proceeding outside this program. It is important l XI, Div.1" were followed in applying the AE system. for the effective use of the technology that it be avail-able on a commercial basis. '

I

! The AE monitoring of the N2H weld demonstrated that continuous AE monitoring can be effectively applied to I RES IM Budget No. LH00; RES

Contact:

A Muscare ix NUREG/CR-5%3 i

Previous Reports in Series By intention, this list does not include routine monthly Hutton, P. H. and R. J. Kurtz, Pacific Northwest Labo-and quarterly reports. ratory, " Acoustic Emission for On-line Monitoring:

Results of Intermediate Vessel Test Monitoring and Reactor Hot Functional Testing," Vol. 4, p.1-19, in Hutton, P. H. and E. B. Schwenk, Pacific Northwest Er.oceedings of the Eleventh Water Reactor Safety Laboratory, " Program to Develop Acoustic Emission- Research Information Meeting, USNRC Conference Flaw Relationship for Inservice Monitoring of Nuclear Proceeding NUREG/CP-0048, Vol. 4 (1984).

Pressure Vessels, Progress Report No.1, July 1,1976-February 1,1977," USNRC Report NUREG-0250-1 Hutton, P. H., J. F. Dawson, M. A. Frisel, J. C. Harris, (1977). and R. A. Pappas, Pacific Northwest Laboratory,

" Acoustic Emission Monitoring of Hot Functional Test-Hutton, P. H. and E. B. Schwenk, Pacific Northwest ing," USNRC Report NUREG/CR-3693 (1984).

Laboratory, " Program to Develop Acoustic Emission-Flaw Relationship for Inservice Monitoring of Nuclear Hutton, P. H. and R. J. Kurtz, Pacific Northwest Labo-Pressure Vessels, Progress Report No.1, July 1,1976- ratory, " Acoustic Emission / Flaw Relationship for In-February 1,1977," USNRC Report NUREG-0250-2 serdce Monitoring of Nuclear Pressure Vessels, Quar.

(1977). terly Report, October 1983-March 1984," USNRC Re-port NUREG/CR-3825, Vols.1 and 2 (1984).

Hutton, P. H., R. J. Kurtz, E. B. Schwenk, and C. Pay-loff, Pacific Northwest Laboratory, " Program to Devel- Hutton, P. H., R. J. Kurtz, M. A. Friesel, R. A. Pappas, op Acoustic Emission-Flaw Relationship for Inservice J. R. Skorpik, J. F. Dawson, Pacific Northwest Labora-Monitoring of Nuclear Pressure Vessels, Annual Re- tory, " Summary of Detection, Location, and Character-port, July 1,1976-October 1,1977," USNRC Report ization Capabilities of AE for Continuous Monitoring NUREG/CR-0123 (1978). of Cracks in Reactors," Vol. 4, p. 362-380 in Proceed-ines of the Twelfth Water Reactor Safety Research Hutton, P. H., R. J. Kurtz, E. B. Schwenk, and C. Pav- Information Meetine. USNRC Conference Procceding loff, Pacific Northwest Laboratory, " Program to Devel- NUREG/CP-0058, Vol. 4 (1985),

op Acoustic Emission-Flaw Relationship for Inservice Monitoring of Nuclear Pressure Vessels, Progress Re- Hutton, P. H. and R. J. Kurtz, Pacific Northwest Labo-port, October 1,1977-January 1,1978," USNRC Report ratory, " Acoustic Emission / Flaw Relationship for In- I NUREG/CR-0124 (1978). service Monitoring of Nuclear Pressure Vessels, Quar-terly Report, April-September 1984," USNRC Report Doctor, P. G., T. P. Harrington, and P. H. Hutton, NUREG/CR-3825, Vols. 3 and 4 (1985).

Pacific Northwest Laboratory, " Pattern Recognition Methods for Acoustic Emission Analysis," USNRC Hutton, P. H., R. J. Kurtz, R. A. Pappas, J. F. Dawson, Report NUREG/CR-0910 (1979). L. S. Dake, and J. R. Skorpik, Pacific Northwest Labo-ratory, " Acoustic Emission Results Obtained From 3^

Hutton, P. H., E. B. Schwenk, and R. J. Kurtz, Pacific Testing the ZB-1 Intermediate Scale Pressure Vessel, Northwest Laboratory," Estimate of Feasibility to De- USNRC Report NUREG/CR-3915 (1985).

velop Acoustic Emission Flaw Relationships for Inser-vice Monitoring of Nuclear Pressure Vessels," USNRC Hutton, P. H. and R. J. Kurtz, Pacific Northwest Labo-Report NUREG/CR 0800 (1979). ratory, " Acoustic Emission / Flaw Relationship for In-service Monitoring of Nuclear Pressure Vessels, Prog-Hutton, P. H., T. T. Taylor, J. F. Dawson, R. A. ress Report, October 1984-March 1985, USNRC Report Pappas, and R. J. Kurtz, Pacific Northwest Laboratory, NUREG/CR-4300, Vol.1 (1985).

" Acoustic Emission Monitoring of ASME Section III Hydrostatic Test," USNRC Report NUREG/CR-2880 (1982).

xi NUREG/CR-5%3 I

Previous Reports Hutton, P. H., R. J. Kurtz, and M. A. Friesel, Pacific Hutton, P. H., M. A. Friesel, and R. J. Kurtz, Pacific Northwest Laboratory, " Progress for On-Line Acoustic Northwest laboratory, " Progress for On-Line Acoustic Emission Monitoring of Cracks in Reactor Systems," Emission Monitoring of Cracks in Reactor Systems,"

Vol. 2, p. 553-564 in Proceedings of the Thirteenth Vol. 2, p. 43-56, in Proceedines of the Fourteenth Wa-Water Reactor Safety Research Information Meetine. ter Reactor Safety Research Information Meeting, USNRC Conference Proceeding NUREG/CP-0072, USNRC Conference Proceeding NUREG/CP-0082, Vol. 2 (1986). Vol. 2 (1987).

Hutton, P. H. and R. J. Kurtz, Pacific Northwest Labo- Hutton, P. H., M. A. Friesel, J. F. Dawson, and J. C.

ratory, " Acoustic Emission / Flaw Relationship for In- Harris, Pacific Northwest Laboratory, " Acoustic Emis-service Monitoring of Nuclear Pressure Vessels, Prog- sion System Calibration at Watts Bar Unit 1 Nuclear ress Report, April-September 1985," USNRC Report Reactor," USNRC Report NUREG/CR-5144 (1988).

NUREG/CR-4300, Vol. 2 (1986).

Hutton, P. H., M. A. Friesel, and R. J. Kurtz, Pacific Hutton, P. H. and R. J. Kurtz, Pacific Northwest Labo. Northwest Laboratory, "On-line Acoustic Emission ratory, " Acoustic Emission / Flaw Relationship for In- Monitoring for Crack Growth in LWRs," Vol. 2, p. 333-service Monitoring of Nuclear Pressure Vessels, Prog- 342,in Praeaadiana of the Fifteenth Water Reactor ress Report, October 1985 March 1986, USNRC Report Kafety Ba=aa ch Tafar=attaa Maa'ia- USNRC Confer-NUREG/CR 4300, Vol. 3, No.1 (1986). ence Proceeding NUREG/CP-0091, Vol. 2 (1988).

Hutton, P. H., Pacific Northwest Laboratory, "Acountie Hutton, P. H., R. J. Kurtz, M. A. Priesel, J. R. Skorpik, Emission / Flaw Relationship for In-service Monitoring and J. F. Dawson, Acoustic Emission / Flaw Relation-of Nuclear Pressure Vessels, Progress Report, April ships for Inservice Monitorine of LWRs. USNRC Re-1986-September 1986," USNRC Report NUREG/CR- port NUREG/CR-5645 (1991),

4300, Vol. 3, No. 2 (198).

1 l

l l

NUREG/CR-5963 xii

Acknowledgements The authors wish to recognize the progressive attitude from the outstanding cooperation on the part of PECO on the part of PECO in permitting the test application staff in the planning and execution of the work Ron of AE monitoring to a real problem on an operating DeGregorio, Ken Collier, John Turner, Jose Vega, reactor and the continued outstanding support by the Anthony Wesstrom, Laura Carlson, Tony Gryscavage, U. S. NRC-RES under the program direction of Dr. J. Jim O'Brien, Jack McElroy, and Mike Lind have been Muscara in the development and application of continu- prominent among the many who have been very sup-aus AE monitoring. This program benefitted portive of the work.

1 xiii NUREG/CR-5963

1.0 Introduction In the course of normalinservice inspection at Limerick e Validate continuous AE monitoring on a nuclear Unit I reactor during the 1989 refueling outage, it was power reactor system.

found that the recirculation nozzle-to-safe end weld VRR-1RD-1A N2H showed a flaw indication. The The first two steps had been completed and the weld

, indication appears to be an intergranular stress monitoring at Limerick provided an opportunity to corrosion crack (IGSCC) crack 7 inches long with a address the third step.

. nominal depth of 0.25 inches and a maximum depth of 0.4 inches. It is located between 31.8 and 38.8 inches The AE monitoring at Limerick Unit 1 uniquely (257' to 314*) circumference measuring clockw'ise from provided AE data from suspected IGSCC in a reactor top-dead-center looking with the flow. The flaw component which could be compared to the amount of indication was evaluated for continued service of the crack growth measured at the end of the monitoring 4

piping using fracture mechanics methods, and a crack- cycle by ultrasonic methods. There was no previous

. arrest-verification specimen was installed for assurance calibration of AE data produced by ISGC cracking as it that the indicated flaw would not grow undetected. In progresses slowly in a reactor component. Previous addition, Philadelphia Electric Company (PECO) made AE/IGSCC data obtained during technology the decision to apply acoustic emission (AE) monitoring development activities proved that IGSCC could be on a continuous monitoring basis to validate a more detected by AE and gave a generalindication of the direct method of determining if a flaw indication was overall pattern of AE vs. IGSCC (Electric Power -

growing during reactor operation. The purpose of this Research Institute 1980). This information was derived effort was to validate technology for continuous AE from accelerated tests (two weeks to a month to grow monitoring of reactor components which had been through the wall of a 4-in. Schedule 80 pipe). It is developed by Pacific Northwest Laboratory (PNL) impracticalin an R&D program to conduct a test truly under the support of the U. S. Nuclear Regulatory simulating reactor conditions where the time for Commission (NRC), Office of Research. The significant IGSCC growth would normally be measured ,

development program followed a planned sequence of: in years.

e Develop initial AE/ flaw relationships by This report discusses the program results in three parts: l laboratory testing of crack growth specimens, e AE system installation and calibration e Evaluate and refine the technology developed in the laboratory by AE monitoring pipe rupture e Interim data resuhs tests, heavy section steel technology vessel tests, and fatigue testing of a thick wall vessel under e Final data results and overall summary. I simulated reactor conditions. l l

l 1.1 NUREG/CR-5963

1 2.0 AE System Installation and Calibration i

l 2.1 Preparation Working from the top down, the logarithmic scale is in dB below reference (10 dB/ division). The analyzer 1 I

A planning meeting was held at the Limerick Plant on reference value is 1 milliwatt when a 500 input is used.

March 4,1989, at which P. IL Hutton, PNL, reviewed The corresponding voltage values can be determined by:

the status of AE technology for continuous monitoring i of reactor components to detect crack growth. The 0.244 decision was made at that meeting to proceed with V a

s " (log (dBm/20)] x 0.707 installation of AE equipment to monitor the flaw indi-cation in weld VRR 1RD-1A N2H at Limerick Unit 1.

This decision was conditional on the basis that all work inside the drywell associated with the installation must Thus, on a voltage scale, the reference corresponds to i 317 millivolts peak with the analyzer on the 0 dB scale.

be completed by March 24,1989. This requirement was met; however, further access to the drywell result-ing from unforeseeable delay in reactor startup was Figure 2.2 shows that the peak response of the sensors ultimately utilized for additional system checkout. occurs at 400 to 425 kHz and that the sensitivity to the helium gas excitation is 0.08,0.18,0.18, and 0.11 milli-By prior agreement from the NRC, Office of Research, V0II8 W f r sensors #1, #2, #3, and #4 respectively at Materials Engineering Branch, equipment developed at the sensor output taking into account the 40 dB exter-PNL under the NRC AE research program was utilized nal gam used to produce the traces. The frequency in the installation at Limerick. This consisted of wave- range of the peak response and the overall response guide AE sensors and mounting clamps (4), preamplifi- Profile are very important to the resistance of the AE crs (4), and an eight-channel AE data acquisi, system to mterference from noise such as coolant flow tion / recording instrument. All signal cabling was pro, noise. Sensors with a respcase profile similar to those vided and installed by PECO. Preparation and installa- seen in Figure 2.2 have been used successfully in the tion of the AE system at Limerick followed the require- Presence of flow noise on a PWR. The sensor sensitivi-ments of the American Society of Mechanical Engi. ty was considered acceptable even though the wave-neers (ASME) Code Case titled " Acoustic Emission for guide sensors were 9 feet long.

Successive Inspections,Section XI, Div.1."

The performance of the monitor system minus the As the first step in preparation, existing three-foot'long sensors was checked using an Acoustic Emission Simu-waveguide AE sensors were modified to provide nine- lator, Model No. AES-1, from Acoustic Emission Asso-foot-long waveguides by welding on additional lengths ciates, Laguna Niguel, California 92677. This is a spe-of 0.125-in. diameter Type 304 stainless steel wire. cial-purpose signal generator which facilitates imposing Construction of the waveguide AE sensors is shown in a time delay between similar signals out of two different Figure 2.1. The AE system was then assembled in the channels to qualify delta time determination by an AE laboratory at PNL, and the sensors were mounted on a monitor instrument. This, in turn, influences the signal 4-in. x 12-in. x 12 in. steel calibration block one at a s urce location capability of the instrument. For input time to calibrate the sensor response characteristics and delays of 100,150, and 200 seconds, the measurement sensitivity. The sensors were pressure coupled to the accuracy was within 4% For inputs of 25 and 50 psec-surface of the block in a manner similar to ths used to nds, the measurement accuracy was within 10%

mount them on a reactor. A broad-band acoustic signal These input time delays represent the range of values was then generated by impinging a helium gas jet from expected to be significant in this test. The accuracy a 30-psi source through a #18 hypodermic needle onto with the larger time delays is quite good, but the accu-the surface of the block with the needle held 1/8 inch racy with the smaller time delays is marginal because above the surface of the block and 1-1/2 inches from the inherent measurement error is a larger portion of the sensor waveguide tip. The resulting spectral re- the total time delay.

sponses shown in Figure 2.2 are a logarithmic measure.

The analyzer reference point is the top of the grid The signal identification module was evaluated in accor-which corresponds to 0 dB on the logarithmic scale. dance with the method described in the ASME Code 2.1 NUREG/CR-5%3

2.0 AE System Installation and Calibration Case, " Acoustic Emission for Successive Inspections, In Figure 2.3, the specific location of the AE sensorsSection XI. Div.1." This calls for installation of a and the flaw indication are shown. Location of the flew representative waveguide AE sensor on a calibration indication is taken from Limerick Unit 1 ISI report block, connecting the sensor to the monitoring system, #89 003. Figure 2.4 shows the .waveguide sensors

] and exciting the sensor ten times by each of the follow- mounted on the nozzle and safe-end. A force of at

ing three methods

least 30 pounds is applied to the waveguide with the pressure bolt which is threaded into the mounting

1. Fracture a 0.3-mm,2H pencil lead against the bracket (Figure 2.5). Since the waveguide is tapered surface of the block in accordance with ASTM from 0.125-in, diameter to 0.050-in. diameter at the tip, E976. the 30 pounds force results in about 15,000 pounds per square inch interface paessure between the waveguide

2. Strike the surface of the block with a 0.25-in. di- tip and the surface of the pipe. Experimental evalua-ameter steel ball dropped from a uniform height. tion of the interface pressure vs. acoustic coupling efficiency in the 300 to 500 kHz frequency range has

. 3. Inject a multi-cycle (five cycles minimum) burst shown that 15,000 pounds per square inch is effective.

Some improvement can be achieved with higher pres-signal into the block with a transducer and a wave-form generator, sure, but it becomes a trade-off between the difficulty of achieving higher interface pressures vs. the limited l The requirement in the Code Case is that the signal gain to be obtained. An Inconel spring is included in identification module flag at least 8 out of 10 lead frac- the line-of-force to help maintain a uniform pressure in ture signals as crack growth AE signals and at least 8 the presence of temperature changes. A high-dut of 10 of the other types of signals as not being temperature electronic pulser is mounted in the vicinity crack growth AE signals. The results obtained in this of the AE sensors (Figure 2.6) to provide a means of evaluation were: qualitatively testing AE sensor sensitivity during reactor operation.

O Pencil lead break - 10 out of 10 identified as crack growth signals As shown in Figure 2.7, after the waveguides exit the opening in the shielding wall, the outer end is support-O Ball drop - 6 out of 10 identified as not being ed in a location away from the opening. This removes

, crack growth signals the critical sensor head (Figure 2.8), which contains the piezoelectric crystal and a 20 dB gain amplifier, from o Pulser - 10 out of 10 identified as not being crack the area of high temperature and possible neutron growth signals. beams through the shielding annulus. It is important that the sensor head be maintained at no more than 200*F and away from neutron beams.

2.2 AE System Installation

, PECO personnel installed RG58 coaxial cables from The initir,1 emphasis was focused on installing the AE the AE sensor heads to a containment boundary pene-sensors on the pipe, installing signal cable from the tration where the wires were connected through a Can-sensors to a containment penetration where the leads n n plug to two-wire pairs which were installed in the could be connected to leads through the containment Penetration plug. The RG58 cable picked up again at wall, and calibration of the installed sensors. All of this the outer end of the penetration plug and went to the work required entry to the drywell. Installation was Prmnplifiers (Figure 2.9) and to the AE monitor in-started March 22,1989 and proceeded as a joint effort strument (Figure 2.10). The AE monitor instrument is by PECO and PNL personnel. The installed AE sys- installed in the same cabinet with the crack-arrest-verifi-tem is shown in Figures 2.3 through 2.10. cation specimen instrumentation. A set of differential amplifiers has been added between the preamp 11fiers and the AE monitor instrument to help overcome a NUREG/CR 5963 2.2

1 1

2.0 AE System Installation and Calibration i

noise problem; this is discussed further under Section tends to transmit to the sensors better with thinner 2.4, Problems, material.

1 l

1 2.3 System Qualification 2A Problems After the AE system is installed, there are two qualifi. The AE monitor system setup included linear preampli- l cation steps to be performed. One evaluates the signal fiers with a linear main amplifier in the AE monitor l detection and source location function of the total sys- instrument. A total of 90 dB electronic gain was in-tem and the other evaluates the response sensitivity of cluded (20 and 40 dB in the preamplifiers and 30 dB in ,

the installed waveguide AE sensors. the main amplifier). Under these conditions, the sys- I tem performance looked very acceptable initially; i.e.,

The signal source location results obtained from crack. the background noise was about 2 volts p ad ibe

, ing 0.5-mm pencilleads on the pipe surface to simulate installed sensor sensitivity checks were satisfactory.

l an AE signal are shown in Figure 2.11. This is an Subsequently, however, a spike transient noise signal accepted method of simulating an AE signal, and it is appeared on all sensors which was in excess of 10 easily used in a field circumstance. Figure 2.11 is a voltsg out of the monitor instrument. This condition roll-out of the pipe section and shows the AE sensors precludes effective AE detection. The detection thresh-(black squares) plus the signal source determination old of the AE instrument cannot be raised high enough (letters). The axial spacing of the sensors is 3-1/2 to ignore the noise signals because it would also be far inches, and they are located at 90* intervals circumfer- too high to detect AE signals. The repetition rate of entially around the pipe (ref. Figure 23). Pencilleads the noise signal was such that it was filling up a tape were broken within 1 inch to 2 inches of each sensor in cartridge (2.1 million signals) record in a few days and turn. This plot shows that the circumferential location in a few instances the data rate became high enough to  ;

accuracy is very good. The axial accuracy is not as cause the monitor instrument to saturate and stop good but still within the acceptabic limits of two wall processing data. This noise problem was unexpected thicknesses (wall thickness at this point is 13 inches). because a similar AE system had been installed at The very narrow axial spacing of the AE sensors con. Watts Bar Unit 1 reactor and operated during hot tributes to the location scatter in the axial. direction; i.e., functional testing of the plant without any noise prob-a given error in signal time-of-arrival determination has lem. In spite of the very diligent efforts on the part of greater significance with the narrow spacing. PECO personnel and the PNL specialist, the specific source of the noise signal could not be identified. The The helium gas jet technique similar to that used during magnitude of the noise transient reduced substantially

" Preparation" (see Section 2.1) to measure AE sensor during reactor startup and has remained low enough to sensitivity on the calibration block was again used to allow effective AE monitoring. The AE monitor instru-evaluate sensitivity of the sensors installed on the reac- ment has been modified to eliminate recording any tor. Figure 2.12 gives the results. The output was signals that appear simultaneously on all sensors (noise taken directly from the sensor which omits two filter transients appear on all sensors simultaneously while stages in the midamplifier; hence, the response curve AE signals from cracking will show a difference in time appears rather poor relative to the similar data taken of arrival at the various sensors depending on the loca-earlier on the calibration block in the laboratory (Fig- tion of the cracking). The fact remains, however, that i

ure 2.2) where the 40 dB gain midamplifier was used. the AE system is vulnerable to noise signal pickup.

In reality, the response sensitivity of the installed sen- The Watts Bar and Limerick installations are compared sors was significantly better as indicated by the voltage in Table 2.1. The most prominent difference in signal values from the plots (0.25,0.71,0.4, and 0.56 milli- lead shielding is in the containment penetration voltsg orf sensors #1, #2, #3, and #4 respectively), arrangement where the Limerick installation has only This could be due to better coupling of the installed two wires while at Watts Bar there are three wires sensors and the thinner pipe wall (13 inches) compared which are twisted to provi<le shielding. Another differ-to the calibration block (4 inches). The helium gas jet ence that could be significant is the signal cabling inside 23 NUREG/CR-5%3

2.0 AE System Installation and Calibration Table 2.1. Comparison of Watts Bar and Limerick AE Installations Item '  ? Watts Bar  : Linierick>

Sensor 3' long two-piece waveguide 9'long one-piece waveguide Sensor tuning 375 kHz and 500 kHz 425 kHz Sensor-to-cable comiect BNC BNC Cable inside containment RG141 in conduit RG58 in cable tray Cable-to-penetration inside BNC Cannon plug connection Penetration Three-wire twisted with 2' of Two wire straight with 3' of wire at inside end and 6' of wire at each end of plug wire at outside end of plug Cable-to-penetration outside BNC Cannon plug connection Cable outside containment RG58 RG58 Preamplifiers 20 dB at sensor and 40 dB 20 dB at sensor and 40 dB outside penetration outside penetration of containment, which is a high-temperature coaxial A second problem arose in early June 1989 when it was cable contained in metal conduit at Watts Bar com. found that Sensor #2 was no longer responding. This pared to standard RG58 cable in a tray at Limerick. was discovered in a pulser check on June 9 when Sen-This emphasizes the need for extreme care in maintain- sor #2 did not show any responses (Figure 2.13).

ing good shielding in such a high-gain electronic system. There are two possible causes of this - cither the collar on the end of the waveguide, which is the bearing sur-Another step was taken to guard against the return of face for the spring holding the waveguide against the the noise transient. This consisted of including a differ- pipe surface, slipped or the integral preamplifier had ential amplifier between the midamplifier and the AE failed. In the numerous applications of this type of monitor instrument. A differential amplifier (unity sensor, including monitoring molten vitrified waste gain) is intended to cancel out signals which appear experiments, this is the first sensor failure we have simultaneously on different signallines. This is reeded experienced. The source location program has been because if the noise transient returns, it can stilllock up modified to give source location using three sensors. In the monitor system buffer memory if the repetition rate order to do this, it was necessary to split the Channel 1 is sufficiently high. Two versions of differential amplifi- output and run it into both Chanrel 1 and 2 of the ers were tried with very limited benefit due to ground- monitor instrument because the logic is arranged to ing problems and difficulty with passing the sensor work only with a four- or two-sensor array; and if power supply which is carried on the signal lead. These Channel 2 was left open, it would never recognize valid problems were remedied, and the amplifier was data from the three-channel array. Figure 2.14 shows installed in the AE system. It reduced background the results of testing the three-sensor program on re-noise by about a factor of two, corded data from the pencillead break input during the -

checkout of the installed system. These results can be NUREG/CR-5963 2.4

2.0 AE System Installation and Calibration compared with Figure 2.11. The three-sensor program monitor instrument. PNL analyzed the data for source results are very good, location of recorded signals, number of signals from a given location, and signal classification. This informa-As of mid-September 1989, the AE system setup con- tion was then compiled in a report to be submitted to sisted of: PECO at least once per month, e Original linear preamplifiers (40 dB gain) plus a Originally, the AE monitoring period was planned for unity gain differential amplifier one fuel cycle (=18 months); however, at the end of the first fuel cycle, the decision was made to continue for a e Sensor #2 was not functional second fuel cycle under joint funding by PECO and NRC-RES.

e Sensor #1 output was split to feed both Channel 1 and 2 e Detection threshold was set at 2.5 volts, down from the previous 3.7 volts Tb: installed AE system was functional and capable of e detecting and locating AE signals. It was installed and An AE signalidentification module was installed checked within a very stringent time window. Coolant m the monitor mstrument flow noise did not cause a problem of increased back-ground noise level. This was a point of concern at the outset because we had not previously monitored a i 2.5 Data Anal.'As swr during operation. The remaining point of con- 1 ccen was the fact'that the AE system was subject to PECO personnel managed the normal operation of the picking up electrical / electronic noise signals due to a 1 AE system. The plans for data analysis called for weak point in the shielding (felt to be primarily at the PECO personnel to remove the data tape cartridges at containment penetration). Day-to-day operation of the  ;

least once per month and send them to PNL for analy- AE system and replacement of the data storage car-sis. Instructions have been provided on loading and tridges for transfer of data to PNL was handled by removing the tape cartridges associated with the AE PECO personnel.

l 2.5 NUREG/CR-5963

I 2.0 AE System Installation and CaLhation e  : BNC Connector T Stainless Steel Housing k 21/2 in. Ing. X 1 1/2 in. WdiX 1 1/4 in. Dp.

20 dB Gain Differential-Preamplifier Tuning Inductor (Variable l

pJ with Freq. Requirements)

Isolation Disk I Al2Or0.010 - PZT Crystal (Chamfered) in.thk. Hysol Adhesive EA934 l l Approx. 0.02 in. thk.

I I I "

Isolation Plate (Delrin) 1 Il l l . .

% Stainless Steel Plate I Weld Nyttite Isolation Bushing (Typ. 4 Places) ,

10-24 Machine Screw b- (Type. 4 Places) b T

. Stainless Steel Type 304-L Waveguide,0.125" Diamater I- : Tip 0.050" Diameter Figure 2.1. Construction of the Waveguide AE Sensors NUREG/CR-5%3 2.6 m m

1 2.0 AE System Installation and Calibration 0 (317 mV Ref.)

OdBm 500,000 KHz 10 10 dB/30 KHz 100 KHz 20 Sensor #1 301

'40 50 60 70  ! 7 Electro ic b p' Background N

0dBm 500 KHz 10 10 dB/30 KHz 100 KHz 20 Sensor #2 4

40 A

8 50

$ 60

$ 7o [ Electroni[

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$ 0dBm 500 KHz g 10 10 dB/30 KHz 100 KHz '

S 20 g Sensor #3 fg-o / 'Y '

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60 70 Electron y Background

-,e O dNm ' 500 KHz 10 10 dB/30 KHz 100 KHz 20  :

Sensor #4

' ^

30 40 50 "

60 ,v v %s 70  ? Electronic 80 M Background N

0 0.2 0.4 O.6 0.8 1'.0 Frequency- MHz Figure 2.2. Sensitivity Test of Waveguide AE Sensors on Calibration Block 2.7 NUREG/CR-5963

2.0 AE System Installation and Calibration 19 0*

AE Sensor No.

I 314*

Max. Depgp of Raw N y

307*

28S' N N

Raw s Indication  % j 270 "

'> - 90' 3, Pulser Locafon 1

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a 3.1 AE System Performance . Tables 3.1 through 3.3 give the AE event count byc source location element for 'each of the data tapes i  :)

The AE system withstood the reactor environment weH analyzed during the first monitoring period.1This'is ;

over the 16-1/2 month long first monitoring period. basic data _- no special fikers were applied in analyzing .- 1 Two functional problems of significance developed in . the information. As we began to ' analyze the data for a s 1 the course of the monitoring period. In June 1989, it correlation between the AE data and the crack growth - 4 was discovered that Sensor #2 (Figure 2.3 shows the as indicated by the ISI iresults, it became evident that i location of the AE sensors on the safe end-nozzle cir, the basic data still contained a significant number of 1. i cumference) was not functioning properly. This did not noiseyi nduced counts in spite of steps taken during data i

, q result in any significant loss 'of data because the AE acquasation o t' ' alleviate the tw-d noise problem ,

instrument was moddied to operate with three instead .wi& the wiginal AE system instaHadonc Med - .

of four sensors. Originally, the instrument was set to . examination of the basic data showed that there were ; j operate with four sensors which required that a given . many accepted signals with two identical deka-dme '

AE event be detected by all four sensors before the values such that they could.not represent real AE sig c event would be recorded. The event detection require- nals fan &c weld area. A very restrictive filter which ment was modified to require only three sensors after Precludes accepting data with two simultaneous delta-;

Sensor #2 failed. Loss of the fourth sensor has some time values and which limits deha-time values to"those : 1 effect on the AE source location by a moderate reduc- rational fw &c size d me pipe (weld) and the AL _.

tion in the longitudinal location resolution, but it does . sensa locations was applied to the basic data recorded '  ;

not adversely affect the circumferentiallocation accura, a the digital tapes with the results shown in Tables 3.4'-  !

cy. This was tested by analyzing system calibration ' through 3.6. The use of this filter could eliminate a . .;

signalinformation recorded during the AE system maH amount of legitimate data from crack growth?  !

installation using the three-sensor arrangement. because there is a smaH source area fan whicEAE ,

signals could reach two of the sensors at the same time.' O Tlie second problem resulted in loss of AE data over : It was concluded that the possible loss of a small: q the period 11/2/89 to 1/5/90. This was caused by whatl aniountlof real AE data was justified for the benefit of? q was ultimately determined to be a failure of the central the fiker. This is'a legitimate method to use in thel processor unit board in the AE monitor instrument. . context of the N2H weld monitoring and also'in the ;

once this was replaced, the system performed satisfac. context of the ASME Code Case N 471 for AE monit _. ,

torily for the remainder of the reactor fuel cycle. toring of reactor components because both concern AE q monitoring of a limited and well-defined volume of thel j structure? This filter was not used originally in the, hope d achieving a mme unimsal muk' which would '

3.2 AE Data AnalIsis  : be applicable to a case where the volume of material to i '

AE data [AE event count, difference-in-time-of-arrival be monitored was'not defined or limited.' This should D still be attainable with a normally quiet AE monitoring '

of each event signal at the various AE sensors (delta- ~ '

anangement. i ti ne), sequence of signal arrival at the sensors, time for each event signal to reach maximum amplitude, peak amplitude of each event signal, and clock time informa- Examination of the basic AE data'and the revised AE '

data showed a significant feature. The AE count in -

tion] was recorded in digital form on magnetic tape ~

source location element 29 for the' period 2/2/90 to .  ;

cartridges. The cartridges were replaced on approxi-2/26/90 is de same in bo& sets d datm i.e., &c deha-mately monthly intervals by PECO personnel and the time filter did not affect this particular data at all.-

used cartridge was sent to PNL for analysis. The data Since location 29 coincides'wie a region d crack on the tape cartridge was played back to a computer-growth indicated by the ISI (see Figure 3.1), it appears -

azed analysis program to compile AE event count by.

. that this is all valid AE data and ' suitable for determin-source location. Source location format is in terms of -

ing an initial correlation between AE and crack growth :

signals originating within each of 36 equal segments

. in the N2H weld. The AE/ Fatigue Crack Growth ;

(source location elements) around a cross-section of the relationship shown in Figure 3.2 which was developed ;

pipe. -

3.1 _ 1 NUREG/CR-5963[ x

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3.0 Interim Data Results j from experimental fatigue crack growth data was tested using the location 29 data and the crack growth indi- -

3.3 Flaw Isdication from ISI:Results -

i. ccted by the October 1990 ISI results. It was found :and Crack-Arrest-Vesification-i that the less conservative curve of the two; i.e., dN/dt SPecim, en. Indications'

! = 1.6 x 10'4(da/dt)l#, fit quite well.2 So the equa- . .. ,

j tion for da/dt gives da/dt = 1(n (dN/dt)05 where . Information on the profile of the flaw indication isl da/dt is crack growth rate in -inches /second 'and ' 'taken from inservice inspection '(ISI) results ah8*6d

, dN/dt is AE count /second/deg ee.; Count /second/ during the refueling outages ' and made available by~

degree is used in this case instes d of just count /second . PECO. h pro 61e of the flaw indication is_shown in~

l to normalize the count over the 10 degree location . 1 Figure 3.1. ' The ISI information has been translated to j segment. This recognizes the fact that the AE count  : a pictorial form on a pipe cross section with the inten-' t

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accumulated is distributed in some fashion across the tion of making it more easily relatable to the AE infor-l 10 degrees. Taking the 22 AE counts from Location ;. mation.L While Figure 3.1 is approximately to scale, it is t j

29,22 counts + 10 degrees = 2.2 counts / degree. The not intended to be an' accurate scale presentation. The i  ;

j data accumulation period was 24 days'or 2,073,600 . light cross-hatched area describes the flaw indication as c i seconds. 2.2 counts /degrec + 2.073,600 seconds = 1.06 4

measured using ultrasonic' methods in February 1989.3 ; ]j l x 10 AE counts /second/ degree. Solving now for Enlargement of the flaw indication' as measured by) .

i da/dt, 8 ultrasonics in October 1990 is shown in Figure 3.1 by L

' the dark cross-hatched area. The AE ~ source location i .;

j 4 da/dt

= 103 (1.06 x 10 )033 element pattern ir ovtr!ayed on the flaw indicationi .

= 0.07 -in./sec. . profile to provide a connection between the AE sources l

= 7.0 x 10-8 in./sec. and the flaw'iridication.1To facilitate conversion ben  !

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tween the AE source location in' degrees and the ISIj <  ;

. Considering the full time period for data accumulation, . location nomenclature in inches around the pipe OD,- 1

. this equates to 7.0 x 104in./sec. x 2,073,600 sec. or . 8.09* equals 1 inch on the' OD ~ surface of the pipe.' One -

j 0.145 inches of growth for the 24-day period. M UT .  ; must also keep in mind that although.the AE source j ct the end of the fuel cycle indicated 0.17 inches of ' . location can be very accurate, the accepted accuracy'. y

, crack growth for the same region. Applying this equa- range is 12 wall' thicknesses'of the component being tion to the revised AE data in Tables 3.4,3.5, and 3.6 monitored. In this case,' the location could vary by 2 x !  ;

j produced crack growth indications as given in Tables ' 13 inches or 2.6 inches which translates to 20*. The  ;

. 3.7,3.8, and 3.9. The results are summarized.in Figure accuracy tolerance is due to variations in the AE signal ?

l 33. It is worth noting that most of the crack growth - propagation velocity to the various~ sensors'~and the indications derived from AE are below the nominal . difficulty of precisely defining the position in time of ~

j detection threshold for ultrasonic inspection, the'AE signal wave front. E l

A crack-arrest-verification-specimen (CAVS) assembly '

previously used at Peach Bottom 2 was installed at -

~

t Limerick Unit'1 to provide a basis for estimating. j i growth of the flaw indication in Weld N2H during ..  !

i reactor operation. Basically, the CAVS system exposes  ;

i a fracture mechames specunen of appropriate material  !

2

%is equation is based on crack growth /AE data sets generated i

with non-irradiated pressure vessel and piping stects (A-533B and j i

A106). 'Ihe data sets p3uced a conservative relationship curve of 3

$ 0eneral Electric Nuclear Energy. ' February 1989. Ihnerick Gener*

de/dt = 190 gdN/dt)0 and a median relationship curve of de/dt =

2 103(dN/dt)0 3which is less conservative. The median curve pro- atina station Unit 11nservice T-* ** -t No. 89 003_

duced a crack growth esthnete that compared quite well with IITISI 4 .;

} results from inspection of the weld; hence,it was used in au the data General Electric Nuclear Energy. September 21,1990.- LimeliGk

].. analysis for this program. Generatine Station Unit 1 Inservice Renort No. D.30014. -

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3.0 Interim Data Results to the reactor coolant water. The specimen is loaded It is evident from the numbers in Table 3.10 that the externally with the intent of simulating the stress pres- UT inspection results and the AE results are in reason-ent around the flaw indication and a potential drop ably good agreement - the UT being about 30% higher.

method is used to measure resulting crack growth. The CAVS system prediction is an order-of-magnitude Results from the CAVS system surveillance indicated lower. The crack growth indidated by the AE monitor-crack growth rate of 0.026 inches / year.5 ng is summarized in Figure 3.4. This indicates some crack growth in areas of the weld where it was not identified by UT inspection during the ISI, but this does 3.4 Correlation between AE, CAVS, not necessarily refute their validity when examined in and UT Results light of the detection threshold for UT (nominally 20%

of wall thickness) as illustrated in Fy;ure 3.5. At the same time, it must be conceded that the AE indications Results from the three surveillance methods are consid-of crack growth less than 0.1 inches may be question-cred on a common basis in Table 3.10.

able on the basis that the values are derived from an l AE event count of three to five. A frustrating dilemma The value for the UT method was derived from the in this comparison is the fact that the only positive maximum indicated crack growth of 0.47-in. deep mea-comparator is destructive examination of the weld when sured at 39-in pipe circumference during the 1990 ISI by a simple ratio of 12 months / year + 15 months total it is removed from service, which makes it very impor-tant that provision for this be considered in any future operating period x 0.47' = 038"/ year. The value for plans to replace the section of pipe.

the AE method was similarly derived from the maxi-mum indicated crack growth of 035-in. deep in Loca-tion Element 2 (see Table 3.9).

5 Ranganath, S. and D. A. Ilate. October 1990. Structurat Evalue-tion of the Indication in the Limerick 1 Recire. Inlet Nozzle N2II Safe End Weld. General Electric Nuclear Energy.

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-= 1.6 x 10-* -

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[ dt -\ dt l

~

' ' 'I ' ' 'l ' ' '"I ' ' ' ' ' ' "

0.01 0.1 1 10 100 1000 CRACK GROWTH RATE, MICRO-IN./SEC Figurt 3.2. AE Rate - Crack Growth Rate Relationship for Cyclic Fatigue Crack Growth (Hutton 1985) 3.5 NUREG/CR-5%3 ,

l 8

l

i l

i 3.0 Interim Data Results CRACK GROWTH - PERIOD 1 E TOTAL l 0.3 -

l l

l

$ 0.2 -

c o

E 0.1 -

i i

e 0.0 m, l

l 1 23456 78 9 101112131415161718192021222324252627282930313233343536 l DEGREES X 10 t

i i

i Figun 33. Crack Growth Over the Period 5/89 to 9/90 as Indicated by AE - N2H Weld at l

l Ilmerick Unit 1 Reactor t

1 e

i i

i l

! NUREG/CR-5963 3.6 i

i l

i 3.0 Interim Data Results- 3 0*

-i l l

J 36 1 2

! 35 ,

2 34 3 l kg. 33 .

4

/ i i N 5 32

.t 31 6 l i

30 4+

f 7 3 8 l l

i l 28 0.25" 90'

g.

10 4 27 l

1. 26 -11

i. @6 12 j 25 24 ~13 23 '14 ,

22 15 21 16 l

20 17 I 39 18 180' T

j 1.E Crack Size per ISI-2/89

2. M Crack Growth during 5/12/89 - 9/11/90 period per AE j 3. Numbered Sectors identify AE Source Locaton Elements Each 10' Wide I

.i i

' A9103152.3 Mgurt 3A. Crack Growth Values Derived home'AE Data over the Peded 8/89 to 9/90 -

Shown on Pipe Crack Section - N2H Weld at Ihnedck Unit 1 Reactor i

e 3.7 NUREG/CR-5963 '

3.0 Interim Data Results O'

36 1-35 .2 '

34- 3- ,

33 ' ' ' 4-

- '2' '-

'/

32 ,

/ '5-31 \ '6, 30 N6 y '7-c 29: -

8 ,

28 '

0.25" -

270'  :. , 90' 27 l l .!10 - -!

26 - 11 '

3  :

25 >

-12 I

24 >

'13- t 23 14' ,

g N #

.15 21 . 16' J 20- 17 19 18' .!

l 180' i

1. M Crack Size perISI 2/89 . .

2. M Crack Growth during 5/12/89 9/11/90 period per AE

3. Numbered Sectors identify AE Source Location Elements Each 10'Wde.

4. b~a i UT Detection Threshold @ 20% of Wall = 0.26* - ,

noiostse.: ;

Figure 33. UT Detection nreshold (20% of Wall) Superhaposed on Crack Growth P:wdic-tion kom AE Data - N2H Weld at Danerick Unit 1 Reactor 5 >

I NUREG/CR-5963 3E .t P

a 3.0 Interim Data Results Table 3.1. AE Event Count by Source Location Element - 5/12/89 - 10/6/89 -

N211 Weld at Limerick Unit 1 Reactor lIhfERICiC AE DATAllst MONITOM

5/12h i]5/25-l? }6/2(N.;l /6M 3 /104'[

t

!5/k {$/15-$ f9/15h

"5/24i ..'6/20 ) 6/29z

7/107 8/84 :L8/154

9/151 n10/6i 1 2 5 13 1 2 5 1.

, 3 o

4 5 1 6 43 7 M 1 i

8 1 9 2 1

10 11 1 1

12 1 '

, 13 2 14 1 M 2 4 4 27 3 16 4 51 88 159 201 10 23 12 l 17 4 18 25 19 44 2 18 20 23 20 41 1 4 19 8 29 31 37 159 3 1 l 20 11 44 35 .39 259 ,180 12 21 6 30 28 24 176 2 22 4 10 23 17 188 1 23 6 13 21 21 180 3 3.9 NUREG/CR 5%3

3.0 Interim Data Results .

_ . . . . , , . ,.m , .

M, 6 T

., s, , IJhERICK AE DATF,.. . ~., ., . ..1st MONITORl;,ad. ;R wf^ ypd6,,

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26 2- 15 .

27 28 29 30 31 -

32 33 i1 34 35 s 36 r

NUREG/CR-5963 3.10 f

1 i

j 3.0 Interim Data Results i

7 i

Table 3.2. AE Event Count by Source Location Elenient - 10/6/89 - 6/4/90 -

N2H Weld at Limerick Unit 1 Reactor

.i 4 .m

. . . . . .,3....- . . :. :..c.c, u. s.,-

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DATA 2 35 28 472' 35 . 39 -5 4

3 4

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5 6 5 6 39 2 '7 2

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! 16 8 41 35 861. 66 - 24 s 17 1 3 18 1 j 19 1 1 2 20 12 25 22 854 62 15 1 21 1 2 22 1 i

3 3.11 NUREG/CR-5963 i

5 I

i

x a

a

!, . 3.0' Interim Data Results i

4 5- _ .

l.

f%

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~ 22 - 15 . ' -3297 15 10 ,-

26 1- L12 27 52 8' 28 -2'  : 1- 17 ' '1L 1 N

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33 .2 - 5~  : 94  : 11, 4 34 19 21 - 2 548..

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36 1 ~2- '1 : 1; 4

L r

I9 NUREG/CR-5%3 3.12

! T we,,-- n -+ +- A w, ww r -

3.0 Interim Data Resuhs Table 33. AE Event Count by Source Location Elemect - 6/4/90'- 9/11/90 -

N211 Weld at Limerick Unit 1 Reactor LIMERICK AE DATA $1st'MONITURj

.-a;: l6/42

16/2I-) [8/k . [8/2M/

- 6/25 - 4:8/2' 4 i8/20i ?9/111 1 3 2 9 106 25 21-3 3 4 2 4 I 1 6 5 2 2 1 6 6 10 3 2 7 8 24 10- 14 8 5 9 1 1 10 1 1 11 1 1 12 3 1 13 2 14 1 2 15 5 5 1 2-16 16 24 12 6 17 18 1 1 19 1 4

20 3 35 10 2 21 1 22 23 2 2 3.13 NUREG/CR-5%3

1 l

3.0 Interim Data Results ,

l i

) LIMERICK AE DATA - 1st MONITOR .

6/4.. 6/25- .8/2- 8/20.-

l 6/25. 8/2 8/20 9/11 24 2 2 1 l

l 25 13 23 7 14 j 26  ;

i 27 1 1 i 28 1 4

29 2 1 1 30 1

31 1 32 33 28 6 1 34 134 74 20 6 35 6 7 2 36 5 2 1 NUREG/CR-5%3 3.14

3.0 Interim Data Results Table 3.4. Filtered AE Event Count by Source Location Element 5/12/89 - 10/6/89 N2il Weld at Limerick Unit 1 Reactor

': LIMERICK - 1st Period REV.'1j

-5/12-- 6/25-:3 i 6/29-1 17/100 '8'/8-L

~

8/15 ;. .; N/15 ;-

5/24l

6/29l-- ' 7/20 :-

~ 8/8 ~- 18/15. 9/15 210/61 i 1 3

2 3

4 5

6 7

8 9

I 10

! 11 12 13 14 15 16 17 18 5 19 4 20 3 21 22 3.15 NUREG/CR-5963

3.0 Interim Data Results

LIMERICK - Ist Period REV.1 5/12-- 5/25- 6/20-1 -6/29.~ -7/10-

8/8-' -8/15-L 9/15-5/24 6/20- '6/29- 7/10 8/8J !8/15- 9/15- ;10/6-23 24 25

26 i 27 28 29 30 31 32  ;

33 M

35 36 NUREG/CR-59f3 3.16

3.0 Interim Data Restits Table 3.5. Filtered AE Event Count by Source Location Element 10/6/89 - 6/4/90 N211 Weld at Limerick Unit 1 Reactor

.: LIMERICK [1st Period. REY.i1)-

10/6-L- [11/2s ?1/5 I  : 1/19-- 22/2 2/26fl d3/29-ll f4/27-T::;

11/2: 41/5l: f1/19 a i-2/2 :2/26? "3/291 54/27J i 6/4:::

1 NO DATA 2 5 14 3 3 5 4 5 5

6 4 7

8 9

10 11 21' 12 13 13 3 14 5 3 15 3 18 16 4 23 17 18 19 20 18 21 3 22 3.17 NUREG/CR-5%3

1 3.0 Interim Data Results I

d i LIMERICK - 1st Period REV.1 10/6-' 11/2-~ 1/5- - : 1/19- '2/2- 2/26. 3/29.- 4/27.-

11/2 1/5 1/19 2/2' 2/26- 3/29 4/27. 6/4.

i 23 24 17 3 3 4 l 26  ;

27 9 1 j 28 19 29 22 30 3 J l

, 31 I

32 1

4 33 5 34 8 d

35 36 i

a l

l

)

NUREG/CR-5%3 3.18

3.0 Interim Data Results Table 3.6, Filtered AE Event Count by Source Location Element 6/4/90 - 9/11/90 N211 Weld at Limerick Unit 1 Reactor

~ LIMERICK - 1st Period'REVl 1i.- -

I v

. 6/44 ~

6/25::

78/2-J-  : 8/20- )

6/25- '8/2~  :

L8/20: .9/11i-1 2 4 14 3 ,

3 8 4 7 5 3 5 6 3 5 7 3 5 l 8 7 l

9 3 1 10 11 12 3 13 14 15 16 17 3 18 19

% 5 21 22 3.19 . NUREG/CR-5%3

p 3.0 Interim Data Results LIMERICK'- 1st Period REV.~ 1 -

6/4-- 6/25- 8/2-; :8/20-6/25- .8/2- '8/20- 9/11 23 4 24 25 3 26 27 28 29 30 31 32 4 1

33 3 4 1 34 5 35 7 12 36 5 9 i

4 i

i NUREG/CR-5%3 3.20

3.0 Interim Data Results Table 3.7. Crack Growth Predicted from AE Data - 5/12/89 to 10/6/89 N211 Weld at Limerick Unit 1 Reactor (inches) 1

~

. CRACK GRbWTIbl'Ist Period '

$/12-- [5/24-j 16/20.t.

f.6/29-;;L l.;:7/105 (8/8'-( [5/15 ; $9/150 P5/24^- i6/20' f.6/29l l7/10i i8/8:!  ::8/15L i9/151 10/6?

1 0.06 2

3

, 4 4

5 6

7 8

l 9 10 11 12 13 14 15 16 17 s

18 0.07 j .

l 19 0.06 20 0.05 21 22 3 21 NUREG/CR 5%3

3.0 Interim Data Results CRACK GROWTH - 1st Period 5/12- 5/24- 6/20- 6/29- 7/10 8/8- 8/15- 9/15-5/24 6/20 6/29 7/10 8/8 8/15 9/15 10/6 :-

23 24 25 26 27 28 29 30 ,

l 31 1 32 l 33 M

35 36

= :r -

A i

i NUREG/CR-5%3 3.22

. 3.0 Interim Data Results Table 3.8. Crack Growth Predicted from AE Data - 10/6/89 to 6/4/90 -

N211 Weld at Limerick Unit 1 Reactor (inches) l 5 CRACK GROWTH - 1st Period :

10/6 : .il )2- . 2/2.'

- .-1/5- I1/19h .

I:2/261 3/29-'-

14/27f:

.11/2 - -1/5' 1/19 2/2- 2/26i 3/29 3:4/27? 6/4-:

l 1 NO DATA 2 0.05 0.10 0.06 3 0.05 4 0.05 i

5 6 0.05

\

8 i 9

t 10 11 0.14

{

12 0.11 13 0.05 14 0.05 0.05 15 0.03 0.13 16 0.05 0.15 17 18 19 20 0.13 21 0.03 22 3.23 NUREG/CR-5%3 i

3.0 Interim Data Results CRACK GROWTH - 1st Period

'11/2:: .'2/2'

~

10/6-. :1/5- 1/19-: -

2/26. :3/29__ 4/27 11/2 1/5  ; 1/19 - :2/2 2/26 -3/29: 4/27i 6/4 23 24 0.13 25 0.03 0.20 26 27 0.09 28 0.13 29 0.14 30 0.05

, 31 32 33 0.07 34 0.09 35 36 1

4 J

1 NUREG/CR-5%3 3.24

3.0 Interim Data Results Table 3.9. Crack Growth Predicted from AE Data - 6/4/90 to 9/11/90 -

N211 Weld at Limerick Unit 1 Reactor l

(inches) i

.: CRACK OROWTH91st Period 5

, ze-j6/4$ f $/253 [8/2-::: i8/20-):  :. Totall L6/25i i:8/2i i:.8/20 ? '9/11f 1 06 2 02 0.12 035 3 02 09 0.16 4 09 0.14 5 02 07 09 6 02 07 0.14 7 02 07 09 8 03 03 9 05 05 10 11 0.14 12 02 0.13 13 05 14 0.10 15 0.16 16 0.20 17 05 05 18 07 19 06 20 07 0.25 21 03 22 3.25 NUREG/CR-5963 l

3.0 Interim Data Results CRACK GROWTH - 1st Period -

'8/2.

6/4-1 6/25- :8/20: Toth!L 6/25' 8/2- 8/20' 9/11.

23 0.07 0.07 24 0.13 25 0.05 0.28 26 27 0,09 28 0.13 29 0,14 30 0.05 1

31 32 33 0.02 0.06 0.15 34 0.02 0.11 35 0.03 0.11 0.14 36 0.02 0.10 0.12 Table 3.10. Indicated Crack Growth Rate at the N2il Weld Over Period 5/89 to 9/90 Method '- l Crack Rate per Year - Maxt -

UT 038 in.

CAVS 0.026 in.

AE 0.28 in.

• Based on approximately 15 months of reactor operation. 'Ihe values for LTT and AE are derived from the area of maximum crack growth indication.

n.w NUREG/CR-5%3 3.26

r 4.0 Final Monitoring Period 4.1 AE Data Analysis dark cross-hatched area. The latest ISI results8 showed only slight changes in the flaw indication profile from The analysis approach used for the final monitoring that dJmed in previous inspections as shown in Figure period was the same as described in Section 3.2, first 4.5. There were no other flaw indications identified in paragraph for the first monitoring period, the latest ISI of the N2H weld. Information from the latest ISI is shown in Figures 4.'6 through 4.11. Infor-Filtered AE event count by source location element for m tion in Figures 4.6,4.7, and 4.8 cover the region of the first monitoring period (5/12/89 to 9/11/90) is the flaw indication (31.8" to 40.4" by conventional ISI recorded in Tables 3.4 through 3.6. Similar information I cation terms). Figures 4.9,4.10, and 4.11 cover the for the second monitoring period (12/13/90 to region where AE indications clustered (0* to 75) but 12/18/91) is shown in Tables 4.1 through 43. All of where the ISI did not show an identifiable flaw the data in these six tables were filtered as described in mdication.

Section 3.2 to climinate noise-induced counts. During the first monitoring period (Tables 3.4 through 3.6), the The crack-arrest-verification-specimen (CAVS) assem-AE monitor instrument was functional 61 weeks out of bly contmued m place during the second monitormg the total monitoring period of 70 weeks for an operat. Period to provide a basis for estimating possible genvth ing efficiency of 87%. During the second monitoring of the flaw indication in Weld N2H during reactor period (Tables 4.1 through 43), the AE monitor instru. Peration. During the second AE monitoring period, ment was functional 17 weeks out of the total monitor- the CAVS system indicated a crack growth rate of 0.036 ing period of 66 weeks for an operating efficiency of inches for the full period.9 This equates to a crack 26%. The operating efficiency of the AE monitor growth rate of about 0.028 inches / year. Thermal cy-system is discussed in detail in Section 4.4 - AE System Cling of the specimen resulting from cycling of the Performance of this report. Reactor Water Cleanup System was considered to be a significant factor contributing to the indicated crack Tables 4.4 through 4.6 summarize the estimated crack growth.

l growth in the various source location elements for the l second monitoring periods plus totals for each period l and for both periods together. The crack growth esti- 4.3 Comparison of AE, ISI, and CAVS l mates from AE data are summarized in Figures 4.1, Results 4.2, and 43.

l The AE results and the ultrasonic ISI results compare favorably in some respects and disagree in others.

4.2 ISI and Crt.ck-Arrest-Verification- During the first AE monitoring period, the ISI results Specimen Results showed some growth in the original flaw indication but no other identifiable indications. The ISI results for the location and profile of the flaw indication as deter, second AE monitoring period showed effectively no mined by ultrasonic ISI are shown in Figure 4.4 which growth in the flaw indication and also no other flaw is a translation from the ISI reports for examinations indications. The AE results summarized in Figures 4.1, performed 2/89 and 10/90. The original flaw indication 4.2, and 43 - estimated crack growth 2 0.1 inches -

(light cross-hatched area) in Figure 4.4 represents the show partial agreement between the AE and ISI infor results of ultrasonic ISI of the N2H weld in February 1989.6 Subsequent growth in the flaw indication over 7 the ensuing fuel cycle as detected by ultrasonic ISI of General Electric Nuclear Energy. September 1990. IJmensk 7 * " *" " '" #*'"**"" **""" "

the N2H weld in October 1990 is illustrated by the 8

General Electric Nuclear Energy. March 1992. timerick Gener.

6 General Electric Nuclear Energy. February 1989. I.imerick Gener-atine Station Unit 1 Inservice Insocetion Report No. 89 003. 9 Cottier, K. B. May 1992. Informal written communication.

4.1 NUREG/CR-5%3 1

3 4.0 Final Monitoring Period mation. AE for the first monitoring period show activi- 4.4 AE System Performance.

ty in the region of the flaw indication (source location elements 26 through 33) as did the ISI. AE for the Coolant flow noise was never a problem. The primary second monitoring period showed no activity in the operational problem encountered in the AE monitoring region of the flaw indication which agrees with the ISI. process was transient acoustic noise pickup at the shielding wall penetration which was an integral part of The two smveillance methods depart from each other, the signal transmission path from the AE sensors to the however, on the region of the concentration of AE monitor instrument. The approach to overcoming this -

indications (source location element 34 through top- . problem involved filtering of the data as it was played dead-center to element 6 in Figure 43). ISI showed back from the tapes to eliminate all signals that did not intermittent root geometry and acoustic interface indi- fit a criteria of signal-time-of-arrival that defm' ed the cations throughout the length of the weld but nothing in source volume as the weld. This sounds like a rather the area of the AE concentration that was indicative of obvious restriction to place on the data but normally in a flaw. Much of the crack growth indicated by AE in a system without abnormal noise, it is preferred to not the location element 34 to 6 region is borderline for. limit the data this way so that general volumetric moni- i detection by conventional ultrasonic (UT) methods if toring is achieved. . It was a valid solution in this case I one accepts the nominal detection threshold of 20% of since the area of interest was focused on the weld. l wall thickness (0.26" in this case) for inspection through the weld. The area of source location element 2, how-The operating efficiency of the AE monitor system was ever, should have been detected by ultrasonic inspec- quite good during the first monitoring period (87%) but tron. It is obvious there are two possible answers: 1)- during the second monitoring period, it fell to 26%

the AE indications are valid and the UT is not picking There are two significant factors which contributed to the indication up - this is not a simple weld to inspect, this situation. ' In order to stay within the program or,2) the AE indications are in error. The only irrefut- budget, it was necessary to limit the number of times able way of answering the question is through destruc- that a PNL instrument specialist was sent to Limerick tive inspection of the weld if and when it is replaced. It to address operational problems.' The alternate was to does not appear that this would be justified any tune in utilize some of the very capable instrument specialists -

the near future. One other option that may resolve the at Limerick and communicate by telephone. The Lim-question is to inspect the weld using an advanced ultra-crick staff were handicapped by being unfamiliar with sonic technique which has improved sensitivity and the design of the instrument and also, they had a full accuracy such as the Synthetic Aperture Focusmg Tech-work load in other areas. The bottom line is that this ,

nique (SAFT) method of flaw imaging during the next I approach worked but it was costly in elapsed time.

outage. The resolution of the SAFT approach may be sufficiently better than normal UT inspection to provide The second and most serious factor was the fact that an answer.

the AE monitor instrument was not adequately cooled.

This did not begin to become evident until the second -

The nominal crack growth indications frna the CAVS monitoring period. Circuit boards that were sent to system are about the same for both monitoring periods PNL for testing during one of the inoperative periods (0.026 to 0.028 inches / year). It appears that CAVS . showed heat tinting. Although temperature measure-does not agree with either UT or AE. This is not ments in the immediate vicinity of the circuit boards intended as a general commentary on the CAVS tech-were not obtained, it is estimated that the temperature nique, but simply a statement of fact as it applies to this had to have been in the Sicinity of 200*F to have caused CASC-some of the component problems experienced.

One AE sensor failed in the course of each monitoring period. It was possible to adjust algorithms to permit effective monitoring to continue with three sensors.

l The sensors are being held at Limerick for the present l NUREG/CR-5963 4.2

1 m

4.0 Final Monitoring Period because the radiation level from them is too high for The AE monitor system was in continuous service w'.th unconditional release, but the level has been steadily limited attention for 11/4 years during de Cr t moni-dropping to where they are now barely above the limit. toring period and continued on into the second moni-When the sensors are released, they will be carefully toring period before significant problems appeared.

examined and tested to determine what caused the one The amount of down time could have been greatly sensor to fail and the condition of the remaining sen- reduced if qualified technical personnel could have been sors. available at all tim s The problems were significantly aggravated by instrument overheating. All factors con-

! sidered, the AE monitor system performed well.

d 1

1 1

4.3 NUREG/CR-5%3 i

.)

1 4.0 Final Monitoring Period 0 10*

l

/

35 2 84 3 0 g "

0.12 0.33 33 4 0.0 16 1.3 in.

32 op O's 5 0

31 ' 'O .g r

30 o.7-s, 29 y ,a e'

28 E- '

9 270*

90*

27 @'

d -

10 v

26 oT 11-0 25 e o' 12 0

4 24 o g 13 ro n.

23 p 14-22 io.@ 15 21 . 0.2s ja 20 jg 3, 17 -

180*

1. Crack growth in inches per AE - May 12,1989 :

to September 11,1990 (Period 1).

2. Numbered sectors identify AE source location elements -each 10 wide. N i2

• 2 Figure 4.1. Crack Growth Values k 0.1' Estimated koma AE Data over the Period 5/89 to 9/90 - N2H Weld at LAmerick Unit 1 Reactor NUREG/CR-5%3 4.4

l' i- 4.0 Final Monitoring Period i

0* 1 o*

! 35 2 34 8

! 0.16 0,77 33 0.15 '0.16

4 l g,g6 O. 77 1.3in.

j 32 0 ,

7o o

31 4 6 4

o'

- y0 30 -7

28 9 '

270* 90'.

-i 27 .10 l 26' 11 25 12 24  ::13 l 23 14-22 15 21 16- i 8 17 19 18 180*

1. Crack growth in inches per AE - September 13,1990 to February 18,1992 (period 2).

2. Numbered sectors identify AE source location elements -each 10' Wide, anim.3 ,

Figure 4.2.

~

Crack Growth Values k 0.1' Estimated firoan AE' Data over the Pedod 12/90 to 12/91 - N2H Weld at - -

Umedek Unit 1 Reactor e

4.5 NOREG/CR'5963 -

e

. *-

• 4 < , - - . - . V w , 4 's.e-

4 1

. 4.0 Final Monitoring Period 1 0* 10=

I 35 2

34 0.28 0,18 0 Se 3

, g,29 #

1 33 oS1 4ep 4 1.3 in.

EV 32 o$ O 5

g i 31 oo 4

s-30 41 o. 7-

.gs .

29 2 '

8 i

t 28E 9 o

90*

270*

27 "ci -

0.25 10 l

l 26 .

o' 11-i O $~

! 25

• o C' 12

• /g 24 g 13 4 o'O o.

• 23 gor '14 o.

l 22 D -15 OS I 21

'#8 16 20 18; 17.

19 I

1 i

! 180*

1. Crack growth in inches per AE-total both periods.

t'

2. $49 Flaw size per ISI - 2/89 . R9211002.2 .,

3. Numbered sectors identify AE source

^

location elements - each 10* wide.

Figure 43. Crack Growth Values k 0.1" Estimated from AE Data over the Full AE Monitodag Period'5/89 to '

12/91 - N2H Weld at IJmerick Unit 1 Reactor NUREG/CR-5%3 4.6

4.0 Final Monitoring Period i

i

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3. Numbered Sectors identify AE Source Location Elements - Each 10'Wde M103152.4 Figure 4.4. AE Source Location Elements Superimposed on Pipe Cross Sectice with Maw Prettle - N2H Weld at Limerick Unit 1 Reactor 4.7 NUREG/CR 5963 c

4.0 Final Monitoring Period

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D ", 3' O GENuclear EnergyINDICATION PLOT SHEET paoseer Ho An

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Figure 4.6. Overall Indication Plot Sheet 1

._ ____ _ _ = _ _ - _ _ _

4.0 Final Monitoring Period M@M bit 1 @-I% M .l R.D -l A d 2.d hERf_.r D - do5 l

'O nJh*.05 10 a, e Of il sa-260 {j  !

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i i

l l

l NUREG/CR-5%3 4,10 l

L_-_______________...._....___----__.--.-

4.0 Final Monitoring Period U HEF4C*'.- Oder .'C L M -t 9 2. Yl22. IR.O.1 A MZN

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{ - --

REPORT NO. 'A SITE:LI/hC2.IC /C UNir 1 A//jJ o

O GENuclear EnergyINDICATION PLOT SHEET eRosecT no; Un R L f

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, Figure 4.9. Indication Plot Sheet for 0" to 7.5" .

i

4.0 Final Monitoring Period l

l 4 lOJ n2h45r 10: A, 45ft. 88-270: l@  !

90.

80.  ;

s.

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+

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f i

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4.13 NUREG/CR-5963

i i

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! 4.0 hal Monitoring Period l

i j

v85C I Bj n2n6 erie. 9,60st 88-260: ]@ i :t l 90.

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l l

l I

l NUREG/CR-5%3 4,14

. . _ _ . . _ . ~ . , - . . , , _ , . _ , , _ _ . . , . _ , , , , . . . , , ,,,..,,,,,.,,,__y..__,.~_,__.,-. ,._7.,m , . . . - , . ,

l l

4.0 Final Monitoring Period Table 4.1. Filtered AE Event Count by Source Location Element -

12/13/90 to 4/27/91 - N211 Weld  ;

at Limerick Unit 1 Reactor  !

1 1

iLIMERICK 2nd Period 2

12/13- -

~ 1/24:' L4/111~

l4/12.' i4/17-1 i.4/20-)  :: 4/23-1 3/260. l

-: 1/23' 4/10: :4/12L i4/16. 4/19 ':- 4/22 : i4/25 :4/271 I 1 3 No 3 1 1 2 1 2 7 1 5 1 1 3 10 1 1 4 6 1 1 1

5 10 3 2 1 i

6 13 7 1 1 I 8 9

l 10 11 12 13 14 1 15 16 17 18 19 20 21 1 22 3 23 3 4.15 NUREG/CR-5963

4.0 Final Monitoring Period

~ LIMERICK - 2nd Per!.d

.12/13.: [1/24-- '4/11.- '4/12..

4/17 _-- i4/20  : 4/23 :. ~4/26-

1/23. '.4/10 :. E;4/12. '- 4/16 -'

'4/19l '4/22 i:4/25 T4/27 24 1 25 2 26 1 27 1 28 1 29 1 30 31 32 33 3 34 2 4 3 35 1 5 3

% 0 1 0 0 1 0 3-NUREG/CR-5%3 4.16

4.0 Final Monitoring Period Table 4.2. Filtered AE Event Count by Source Location Element -

4/28/91 to 8/9/91 - N211 Weld at Limerick Unit 1 Reactor llMERICK - 20d Periodi;'

4/28-5/3 J 5/3_-5/7 y '5/8-5/12 i 5/13-I ~5/19.. .g7/22( .17/25-8/5; L 8/6-8/9i

~

~

i'5/184  ;;7/211 47/24::

1 2 1 1 No 1 No Data Data 2 3 3 4 3 2 4 1 5 1 2 6 5 3 7 1 8 -

9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 4.17 NUREG/CR-5%3

4.0 Final Monitori:ig Period

' LIMERICK 2nd Period -

4/28-5/3 - 5/3-5/7 5/8-5/12 5/13- - 5/19-- 7/22-- 7/25-8/5.' : 8/6-8/9 -

5/18 7/21 .7/24-24 25 2 26 27 28 29 30 1 31 32 1 33 3 M 3 4 3 4 35 3 4 2

% 4 2 2 3 1 l

NUREG/CR-5%3 4.18

4.0 Final Monitoring Period ;

i Table 43. Filtered AE Event Count by Source Location Element -

8/9/91 to 12/18/91 - N2H Weld at Limerick Unit 1 Reactor t

. .,- .m.m... , , i

' ^ '

' ILIMERICK52nd PeridM - 4-  ; , - >

~

, ,s, .

I j - - . f' 9

' ' G '^,. ..f(

c12/18:-l

, < s ,  % N'

• 1 No 4 Data 3 1 4 2 .

5 1 -l 6 1 l 7 < y

.1 8 ,

i 9

10 11 12 ,

13 14 15 16 17 18 19 1 20 21 22 23' 2

~

4.19 ' NUREG/CR-5%3 ;

i 4.0 Final Monitoring Period l

LIMERICK - 2nd Period -

8/9-12/9

~

12/10- j 12/18'-

24 25 26 27 j 28 29 i

~

I' 30 31 32 2 33 i

34 1 35 5 36 1 NUREG/CR-5%3 4.20

4.0 Final Monitoring Period Table 4.4. Crack Growth Predicted from AE Data - 12/13/90 to 4/27/91 -

N211 Weld at Lirr.erick Unit 1 Reactor Onches)

-CRACK GROWTH - 2nd Period

12/13 1/24'  :-4/11-s* -4/12- -- 4/17-: "4/20-; '4/23-;1 l;4/26 :::

.1/23

.4/101  ::4/12c )4/16L- T4/19:-  :;4/22 i- L4/25.. 4/27';

1 0.03 No 0.01 0.01 0.01 0,01' O.01 2 0.04 0.01 0.02 0.01 0.01 3 0.05 0.01 0.01 4 0.04 0.01 5 0.05 0.01 0.01 6 0.06 I

7 0.01 0.01 8

9  ;

10 11 12 13 14 0.01 15 16 17 18 19 20 21 0.01 22 0.01 23 0.01 4.21 NUREG/CR-5963

i 4.0 Final Monitoring Period l

l CRACK GROWTH 2nd Period 12/13- 1/24 4/11- 4/12. 4/17- 4/20I /./23- -- ' 4/26 ..

1/23: 4/10 4/12 4/16 4/19 4/22 4/25 4/27' 1

24 0.01 25 0.01 26 0.01 27 0.01 28 0.01 29 0.01 31

_ l 32 33 0.03 34 0.01 0.02 0.01 35 0.01 0.02 0.01 26 0.00 0.01 0.00 0.01 0.01 0.00 0.01 NUREG/CR-5%3 4.22

i e

4.0 Final Monitoring Period - j P

Table 43. Crack Growth Predicted from AE Data - 4/28/91 to 8/8/91 - >

N2H Weld at Limerick Unit 1 Reactor - 'i (inches) i ir -

jf M ^

g .., ' , , _ . . _

s ,

S iMCK GROWTH M2Ed Phlsil . M C'  :

h7/22s" E4/28-5/3 ' j;5/3-5/7.-J f$/S$12? .J5/13-j; tis /19@.. . (7/25-8/5l- (8/6 8/83  :!

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1 0.02 0.01 0.01 No ' No .

Data -

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3 0.02 i 4 0.01 5 0.01 0.02 ,

6 ' O.03

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9 10 11 <

I 12 13 i

14 I l

16 17 I 18 19

.i 21 22 23 g ..

5 4.23 NUREG/CR-5963 -

.l 4

l 4.0 Final Monitoring Period CRACK GROWTH - 2nd Period -

4/28-5/3 5/3-5/7 5/8 5/12 5/13- 5/19- 7/22- 7/25-8/5- 8/6-8/8 5/18 7/21- 7/24 24 25 0.02 26 27 28 29 30 0.01 31 32 0.01 33 0.02 34 0.02 0.03 0.02 0.03 35 0.02 6.03 0.02 36 0.02 0.02 0.02 0.03 0.01 0.00 l

NUREG/CR-5%3 4.24

' - - ~ ~ ^~ ' '

4.0 Final Monitoring Periad Table 4.6. Crack Growth Predicted from AE Data - 8/9/91 to 12/18/91 Plus Totals for 5/12/89 to 12/18/91 -

N2H Weld at Limerick Unit 1 Reactor (inches)

ICRA'CK GRO'WTH - 2Ad P$rfoE ASD'

SUMMARY

!, Total #27 < dotAlY1[ t iToNBothl!

- 8/9d2/91 ,112/10 12/181 Periddi' 1Periodi !Periodsi 1 No 0.04 0.16 0.06 0.22 '

i Data 2 0.17 035 0.52

3 0.02 0.11 0.16 ' O.27 4 0.03 0.09 0.14 0.23 t

5 0.02 0.12 0.09 0.21 6 0.02 0.13 0.14 0.27 7 0.03 ' O.09 . 0.12 8 0.03 ' O.03 ,

9 0.05 0.05 10 I l

11 0.14 0.14 12 'O.13 .0.13 1 1

13 0.05 0.05 j 14 0.01 0.10 0.11 -

15 0.16 0.16 16 0.20 0.20 I

17 0.05 0.05 1

18 0.07. 0.07 l

+

19 0.02 0.02 0.06 - 0.08 l 20 0.25- 0.25 l

21 0.01 0.03 0.04 22 0.01- 0.01 1 4

a 4.25 NUREG/CR-5%3 l

1 m

4.0 Final Monitoring Period CRACK GROWITI 2n( Period AND

SUMMARY

Total #2 -- Total #11 : Total Both :

8/9-12/9 ' '12/10-12/18 .. Period 1 Period' Periods 23 0.03 0.04 0.07 0.11 24 0.01 0.13 0.14 25 0.03 0.28 0.31 26 0.01 0.01 ,

27 0.01 0.@ 0.10 28 0.01 0.13 0.14 29 0.01 0.14 0.15 30 0.01 0.05 0.06 31 32 0.03 0.04 0.04 33 0.05 0.15 0.20 34 0.02 0.16 0.11 0.27 35 0.04 0.15 0.14 0.29 36 0.02 0.16 0.12 0.28 1

Taken from Table 3.9.

f NUREG/CR-5%3 4.26

~ . ___ ._ - _ .. . .__ ._ -_

i b

i 5.0 Observations for' Future Monitoring 1

2 l There are several lessons to be learned from this moni- 5.4 Calibration of the AE Source Loca--  :

j' toring effort to enhance the effectiveness of future AE tion

monitoring of nuclear reactor components. ,

- At a minimum, pulse sign'als such as those produced by.

l

. a pencil lead break should be input at known points

5.1 Reference Monitor Point- within the area of interest and the source location de-

' termination by the instrument recorded at the beginning j Plans for the first one or two AE monitoring efforts in .and end of a fuel cycle. Additional calibrations during -

j a given plant should include a base reference monitor the fuel cycle would be beneficialif a suitable outage.

point which would be a location on the reactor system occurs.

j in the vicinity of components of concern where cracking ('

l or other physical deterioration is highly unlikely. In the case f ee N2H weld, a location on de 12" pipe up-i stream of the elbow would have been a good reference 5.5 AE SignalIdentification location. The purpose of this reference point is to .

An u.nProved method ofimplementm.gAEs.ignal ident.i-provide data from a location that is subject to all the -

Heation is needed. The signalidentification module .

j conditions (noise, electrical transients, etc.) that the ,

used in momtoring the N2H weld did not perform as I location of concern sees except ihw growth. At a stage expected. The bas,c i concept of the three pulse signal  :

! when AE monitoring for crack giowth is still being .

Produced by an AE event when waveguide sensors are ,

I viewed with some skepticism, this would permit a more being used is valid but the method for putting this -

! positive separation of crack growth AE from other concept into practice was not satisfactory without fur-innocuous signals or noise. Such a reference point was ther testmg and refinement. It was a new design with  ;

1 origina!!v planned at Limerick but other critical work in no opportunity for prior test,mg under field conditions, ,

the imm'rdiate area precluded installation of the added and it was not practical to have'AE specialists spend. 1

,,,3o,3,

time working with the module at Limerick to see if its -)

performance could be improvedi An AE signalidentifi- i '

er is important to AE monitoring because it provide:

l 5.2 Conta.inment Penetration another defense against being mislead by noise signals.

I

! Shielding of conductors on both sides of the contain-ment penetration plug needs to receive specific consid- 5.6 : Digitized A' E Information l eration.' The ideal per.etration would be one with coax.

j ial lead wires but it is recognized that in most reactors, A modification to' current AE monitoring methodology t these are all used for uitical reactor control instrumen.

which would overcome a large part of the noise inter.

! tation. Baring coaxial conductors, twisted-shielded-pairs ference and AE signJ identification problem would be

with good shielding on both sides of the plug have been

' to digitize the informa: ion at the sensor and transmit shown to be a suitably noise free arrangement for AE

. digital information 04t to the monitor instrument. This monitoring.

approach or some vsriation thereof (such as Aghg the data at a point inside of the containment penetra- l l

ti n) has been' discussed by PNL staff and others for

5.3 Monitor Instrument Location some time. The primary problem is seen as the cost to P &c necessary &cuks that could &tand er i

The AE monitor instrument must be housed in a loca- CN mnn8 Perform reHaMy m, side of reactor tion where the temperature of the instrument boards taimnent. The approach to testing and calibrat,on i -

can be maintained below a maximum temperature of "*** '

aboct 100-125'F. This should avoid any problems re-l i sulting from heating of the components.

i i

a

! -l 7 l

?

5.1 NUREG/CR-5963 '

l t

3 J

u

1 6.0 Summary and Conclusions )

The project to apply AE monitoring to the N2H weld period. The relationship arrived at for relating AE to at Limerick Unit 1 to detect growth of an identified crack growth gives rational results which are similar in flaw indication and/or any other flaw growth in the magnitude to ultrasonic estimates. AE indicated limit-weld is considered a success. It demonstrated that an ed flaw growth in an area near the ISI flaw indication l AE monitor system can be installed and calibrated on which was not corroborated by ISI results. An absolute (

an operating reactor system within a short time period resolution of the validity of this indication can only be j as dictated by the outage schedule. It further demon- achieved by destructive examination of the weld which I strated that coolant flow noise on a BWR is not a prob- will not occur in the near future. Inspection of the tem [ previous tests at Watts Bar Unit I reactor demon- weld using an advanced technique with improved per-strated that flow noise on a PWR is not a problem formance such as the Synthetic Aperture Focusing (Hutton et at 1984)) and that an abnormal amount of Technique for ultrasonic testing may shed further light RF noise pickup can be overcome to produce useful on the question. The AE system worked well, data. The monitoring proceeded over two fuel cycles. produced reasonable results, and did not pose any Although the operating efficiency of the AE system was unwarranted problems to continuity of reactor not good during the second mcnitoring period, evidence operation.

indicates that a significant contibutor was the temperature environment which is a problem that is Several observations were noted in the course of the relatively easy to solve. effort which will contribute to enhanced effectiveness of future AE monitoring of reactor components. Future AE and ISI results generally agreed with regard to the applications of AE monitoring will benefit from transfer growth of the flaw indication. There was modest of this technology to a commercial company (ies) that is growth shown during the first monitoring period and proceeding outside this program. It is important to the essentially no growth during the second monitoring effective use of the technology that it be available on a commercial basis.

6.1 NUREG/CR-5963

l 7.0 References Electric Power Research Institute. May 1980. Devel- Hutton, P. H., et al. September 1985. Acoustic Emis-ooment of an Acoustic Emission Zone Monitor and sion Results Obtained from the ZB-1 Intermediate Recorder for BWR Pine-Cracking Detection. Report Scale Pressure Vessel. NUREG/CR-3915. Nuclear ..

No. EPRI NP-1408. Regulatory Commission, Washington, D.C.

Hutton, P. H., et al. June 1984. Acoustic Emission  !

Monitoring of Hot Functional Testing. NUREG/CR- l 3693. Nucicar Regulatory Commission, Washington, .3 D.C.

L F

)

1 7.1 NUREG/CR-5%3'

,. q--- s .-en, ,,n,,. , ,. e. , .,

NUREG/CR-5963 PN18844 RS,GS Distribution No. of No. of

_Conin Copics A. L. IIiser. Jr. B. E. Rodgers NRC/RES TVA Nudear Maintenance Mail Stop NS 217C BR5N49A Chattanooga, TN 37402-2801 J. M uscara NRC/RES R. C. Emory Mail Stop NS 217C Maintenance Engineering and Technical Section D. W. Craig MOB 1L NRC/RES Watts Bar Nudear Plant Mail Stop NS 217C PO Box 2000 Spring City, TN 37381 3 D. B. Petters, 62A4-1 Nuclear Engineering Dept. 31 Pacific Northwest Laboratory Philadelphia Electric Co.

%5 Chesterbrook Blvd. DM Boyd Wayne, PA 19087-5 % 1 JP Dawson SR Doctor (10) 2 J. McElroy MA Priesel Philadelphia Electric Co. RV Harris PO Box 650 PH Hutton (3)

Valley Porge, PA 19482 RJ Kurtz (3)

DK Lemon 3 D. D. Nunn RL Moffitt  !

l Vice President, Nudear JR Skorpik Projects JC Spanner i TVA TechnicalInformation (5)  !

LP6N20E Publication Coordination Chattanooga, TN 37402-2801 l

J. A. Scalice Plant Manager Watts Bar Nudcar Plant PO Box 2000 Spring City, TN 37381 Distr.1 NUREG/CR-5%3

U.S. NUCLEAR REGULATORY COMMISSION 1. REPOAT NUMBER NRC FORM 335 hfc% 1102.

b Yumbsq n 2m . 22" BIBUOGRAPHIC DATA SHEET NUREG/CR-5963

esee inserverions on ene r. rses g4g

2. TITLE AND SUBTITLE Continuous AE Crack Monitoring of A Dissimilar Metal

3. DATE REPORT PUBLISHED l Weldment at Limerick Unit 1 MONTH YEAR December 1993

4. FIN OR GR ANT NUMBE R FINS B2913 and L1100

6. TYPE OF REPORT

5. AUTHORIS)

P. H. Hutton, M. A. Friesel, J. F.* Dawson Technical I. PE R IOD COVER ED flectuw.e pe,mf i 5/89 - 3/92

8. F0F > AN1Z ATION - NAME AND ADDRESS tar Nac. provksr D%kon. Ortoce or nouton, U.S Nucker Moruintmy commkston, ana me, sing edoven, tr eensrector. provide e

] Pacific Northwest Laboratory

Richland, WA 99352 a

9. SPONSO R RGANIZATION - NAME AND ADDRESS rif NRC, type *seme m anove ; /(contrec,or, provide NRC Dwon, Derice or Amedon, ua Nucher AssuJa,ory Commission, Division of Engineering Office of Nuclear Regulatory Research j

U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 l i 10. SUPPLEMENTARY NOTES A I j 11. ABSTR ACT (200 words or sess>

1 Acoustic emission (AE) technology for continuous surveillance of a reactor component (s) to l

detect crack initiation and/or crack growth has been developed at Pacific Northwest Laboratory (PNL), ' operated'by Battelle Memorial Institute, under support from the U. S.

Nuclear Regulatory Commission, Office of Nuclear Regulatory Research (U. S. NRC-RES). The

] technology was validated off-reactor in several major tests, but it had not been validated

~

by monitoring crack growth on an operating reactor system. - A flaw indication was identi-2 fied during normal inservice inspection of piping at Philadelphia Electric Company (PECO) l Limerick Unit I reactor during the 1989 refueling outage. Evaluation of the flaw j indication showed that it could remain in place during the subsequent fuel cycle without compromising safety. The existence of this flaw indication offered a long sought opportunity to validate AE surveillance to detect and evaluate crack growth during reactor operation. Through the cooperation and support of PECO and the U. S. NRC-RES, AE

instrumentation was installed by BNW and PEC0 under PECO Mod. No. 043-002 to monitor the flaw indication during two complete fuel cycles. This report discusses the results'ob-tained from the AE monitoring,over the period May 1989 to March 1992 (both fuel cycles).

.oa. .,,ar.s,. ,a,,-,,, , , r. n ,<n, , orr., >2. avai'AsiuTv sTata==~T i2. x E Y woRoSiotSCR:eTOR S ru 4

Unlimited

14. $t CUAli V CLAsO F icA TIOte

! nondestructive testing, acoustic emission, AE, continuous trau r ,

inservice monitoring, flaw growth detection, reactor pressure Unclassified t ra. a-a, boundaries Unclassi fied Ib. NUMBER OF PAGES h 16. PRICE N8tC 7ORM 335 (2491

,4,y

Printed on recycled paper Federal Recycling Program i

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