ML053070118

Letter, Second 10-Year Inservice Inspection Interval Steam Generator C Hot Leg Nozzle Welds Flaw Evaluation

ML053070118
Person / Time
Site:	Catawba
Issue date:	01/03/2006
From:	Marinos E Plant Licensing Branch III-2
To:	Jamil D Duke Energy Corp
Saba F
References
TAC MC4993
Download: ML053070118 (9)

Text

January 3, 2006 Mr. Dhiaa Jamil Vice President Catawba Nuclear Station Duke Energy Corporation 4800 Concord Road York, SC 29745

SUBJECT:

CATAWBA NUCLEAR STATION, UNIT 2, SECOND 10-YEAR INSERVICE INSPECTION INTERVAL STEAM GENERATOR C HOT LEG NOZZLE WELDS FLAW EVALUATION (TAC NO. MC4993)

Dear Mr. Jamil:

By letters dated October 19 and December 2, 2004, and September 22, 2005, Duke Energy Corporation (Duke, the licensee), submitted an evaluation of a flaw indication in the reactor coolant hot leg to steam generator (SG) inlet nozzle connection for Catawba Nuclear Station (Catawba), Unit 2. The licensee discovered the indication by radiographic examination on October 7, 2004, during the units 13th refueling outage. Duke intended to demonstrate through a flaw evaluation using WCAP-15658-P, Revision 1, Flaw Evaluation Handbook for Catawba Unit 2 Steam Generator Primary Nozzle Weld Regions, that the unit can be operated without repair of the subject SG nozzle connection for an additional 30 years until the end of the license.

The Nuclear Regulatory Commission (NRC) staff has completed its review and found that the flaw evaluation meets the rules in Section XI of the American Society of Mechanical Engineers, Boiler and Pressure Vessel Code (ASME Code). With a conservatively assumed flaw depth, the detected circumferential crack in the SG inlet nozzle connection will not reach the allowable flaw depth reported in WCAP-15658-P, Revision 1 after 30 years of crack growth under the limiting loading condition (normal and upset). Hence, the NRC staff concludes that Catawba Unit 2 can be operated without repair of the subject nozzle connection for 30 years until the end of the license. As mentioned in the submittal, successive inspections at the flaw location will be conducted during the next three inspection periods in accordance with IWB-2420, Successive Inspections, in Section XI of the ASME Code.

D. Jamil The enclosed Safety Evaluation contains the NRC staff's evaluation and conclusions.

Sincerely,

/RA/

Evangelos C. Marinos, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-414

Enclosure:

As stated cc w/encl: See next page

D. Jamil The enclosed Safety Evaluation contains the NRC staff's evaluation and conclusions.

Sincerely,

/RA/

Evangelos C. Marinos, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-414

Enclosure:

As stated cc w/encl: See next page DISTRIBUTION:

Public LPL2-1 R/F RidsNrrDorlLplc(EMarinos)

RidsNrrPMJStang RidsNrrLAMOBrien RidsOgcRp RidsAcrsAcnwMailCenter RidsNrrDpr RidsRgn2MailCenter(MErnstes)

SSheng MMitchell BWetzel, EDO, RII ADAMS Accession No: ML053070118 NRR-106 OFFICE NRR/LPL2-1/PM NRR/LPL2-1/LA NRR/EMCB/SC NRR/LPL2-1/BC NAME JStang: ckg MOBrien MMitchell by EMarinos Memo dated DATE 12/29/05 12/29/05 10/11/05 01/03/06 OFFICIAL AGENCY RECORD

Catawba Nuclear Station, Units 1 & 2 Page 1 of 2 cc: w/encl.

Mr. Lee Keller, Manager North Carolina Electric Membership Corp.

Regulatory Compliance P.O. Box 27306 Duke Energy Corporation Raleigh, North Carolina 27611 4800 Concord Road York, South Carolina 29745 Senior Resident Inspector U.S. Nuclear Regulatory Commission Ms. Lisa F. Vaughn 4830 Concord Road Duke Energy Corporation York, South Carolina 29745 526 South Church Street P. O. Box 1006 Mr. Henry Porter, Assistant Director Mail Code = EC07H Division of Waste Management Charlotte, North Carolina 28201-1006 Bureau of Land and Waste Management Dept. of Health and Environmental Control North Carolina Municipal Power 2600 Bull Street Agency Number 1 Columbia, South Carolina 29201-1708 1427 Meadowwood Boulevard P.O. Box 29513 Mr. R.L. Gill, Jr., Manager Raleigh, North Carolina 27626 Nuclear Regulatory Issues and Industry Affairs County Manager of York County Duke Energy Corporation York County Courthouse 526 South Church Street York, South Carolina 29745 Mail Stop EC05P Charlotte, North Carolina 28202 Piedmont Municipal Power Agency 121 Village Drive Saluda River Electric Greer, South Carolina 29651 P.O. Box 929 Laurens, South Carolina 29360 Ms. Karen E. Long Assistant Attorney General Mr. Peter R. Harden, IV, Vice President North Carolina Department of Justice Customer Relations and Sales P.O. Box 629 Westinghouse Electric Company Raleigh, North Carolina 27602 6000 Fairview Road 12th Floor NCEM REP Program Manager Charlotte, North Carolina 28210 4713 Mail Service Center Raleigh, North Carolina 27699-4713 Mr. T. Richard Puryear Owners Group (NCEMC)

Duke Energy Corporation 4800 Concord Road York, South Carolina 29745

Catawba Nuclear Station, Units 1 & 2 Page 2 of 2 cc: w/encl.

Division of Radiation Protection NC Dept. of Environment, Health, and Natural Resources 3825 Barrett Drive Raleigh, North Carolina 27609-7721 Mr. Henry Barron Group Vice President, Nuclear Generation and Chief Nuclear Officer P.O. Box 1006-EC07H Charlotte, NC 28201-1006 Diane Curran Harmon, Curran, Spielbergy &

Eisenberg, LLP 1726 M Street, NW Suite 600 Washington, DC 20036

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SECOND 10-YEAR INSERVICE INSPECTION INTERVAL STEAM GENERATOR C HOT LEG NOZZLE WELDS FLAW EVALUATION CATAWBA NUCLEAR STATION, UNIT 2 DUKE ENERGY CORPORATION DOCKET NO. 50-414

1.0 INTRODUCTION

By letters dated October 19 and December 2, 2004, and September 22, 2005 (Agencywide Documents Access Management System Accession Nos. ML043010387, ML043490612, and ML052770358), Duke Energy Corporation (Duke, the licensee), submitted an evaluation of a flaw indication in the reactor coolant hot leg to steam generator (SG) inlet nozzle connection for Catawba Nuclear Station (Catawba), Unit 2. The licensee discovered the indication by radiographic examination on October 7, 2004, during the units 13th refueling outage. Duke intended to demonstrate through a flaw evaluation using Westinghouse Commercial Atomic Power (WCAP)-15658-P, Revision 1, Flaw Evaluation Handbook for Catawba Unit 2 Steam Generator Primary Nozzle Weld Regions, that the unit can be operated without repair of the subject SG nozzle connection for additional 30 years until the end of the license.

2.0 REGULATORY EVALUATION

The inservice inspection of the American Society of Mechanical Engineers, Boiler and Pressure Vessel Code (ASME Code), Class 1, 2 and 3 components shall be performed in accordance with Section XI of the ASME Code and applicable editions and addenda as required by Title 10 of the Code of Federal Regulations (10 CFR), Section 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(I).

Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1, 2, and 3 components (including supports) shall meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in Section XI of the ASME Code to the extent practical within the limitations of design, geometry, and materials of construction of the components. When flaws are detected by volumetric examinations, acceptance of them by supplemental examination, repairs, replacement, or analytical evaluation shall be in accordance with IWB-3130, Inservice Volumetric and Surface Examinations. In this application, the licensee applied IWB-3600, Analytical Evaluation of Flaws, specified in IWB-3132.3, Acceptance by Analytical Evaluation, to demonstrate that the unit can be operated for Enclosure

30 years until the end of the license without repair of the reactor coolant hot leg to SG inlet nozzle connection.

3.0 TECHNICAL EVALUATION

A typical flaw evaluation for detected flaws includes five elements: (1) flaw sizing; (2) the applied stress intensity factor (Kapplied) calculation and the associated crack growth evaluation; (3) a driving force evaluation for the final flaw size using linear elastic fracture mechanics (LEFM), elastic plastic fracture mechanics (EPFM), or limit load analysis according to the projected failure mode; (4) a failure resistance evaluation considering embrittlement due to various environmental conditions and the projected failure mode; and (5) a stability evaluation using Nuclear Regulatory Commission (NRC)-approved acceptance criteria including appropriate structural factors. The five elements of the licensees flaw evaluation are evaluated in the following sections. Since the detected flaw is circumferential and located in the nozzle interface between the stainless steel safe end (including buttering) and the stainless steel field weld, the NRC staff will evaluate the part of the licensees methodology applicable to circumferential embedded flaws in stainless steel materials only.

3.1 Flaw Sizing The licensee based its flaw sizing on its 2004 radiographic test (RT) results, which showed that the indication was located in the reactor coolant loop hot leg to SG inlet nozzle connection at approximately 1.01 inches from the pipe outer surface. Duke obtained this flaw location by taking radiographs with the source placed at two different axial locations relative to the nozzle connection. The indication was characterized as linear and circumferential, and was approximately 1 inch long. Additional information in the licensees September 22, 2005, response to the NRC staffs Request for Additional Information (RAI) confirmed that the flaw was located in the interface between the nozzle safe-end/butter and the nozzle field weld. Duke also reported that the flaw evaluation assumed that the detected flaw was planar instead of linear.

The NRC staffs review confirmed that the licensees flaw characterization was in accordance with IWA-3370, Radiographic Examination, of Section XI of the ASME Code, which states,

[a]n indication detected by radiographic examination shall be considered to be a linear flaw....

By definition, a linear flaw has only limited dimensions in the plane perpendicular to the length direction of the indication. Since cracks may only be detected if they lie in a direction parallel to the radiation beam, the linear indication in the first radiograph, if it was caused by a planar flaw, would disappear in the second radiograph when the radiation source is displaced axially.

The fact that a linear indication was shown on both radiographs with very little contrast supports that the indication is a slag inclusion. Therefore, the NRC staff determined that the licensee has established qualitatively that the indication is at worst a linear flaw with narrow depth and width and the licensees approach of performing a bounding flaw evaluation based on an embedded planar flaw configuration of one inch deep is conservative.

3.2 The Applied Stress Intensity Factor and the Crack Growth Rate The licensee determined that crack growth of the detected flaw due to stress corrosion cracking is insignificant for stainless steel materials, and, consequently, the licensee only calculated crack growth due to fatigue. Dukes fatigue crack growth calculation employed the fatigue

crack growth rate (CGR) for austenitic steels in air environment from Appendix C of Section XI of the ASME Code, of which the key parameter Kapplied was calculated considering thermal and deadweight piping loads, pressure, thermal transient loads, and residual stresses. The primary system transients and their associated number of occurrences are defined for normal, upset, emergency, and faulted conditions in WCAP-15658-P, Revision 1, Table 2-1. The licensees Kapplied calculations were based on a paper by Shah and Kobayashi (1971), which is applicable to cases with embedded flaws not too close to the nozzle inner-diameter or outer-diameter surface.

Dukes approach in calculating the CGR as described above is consistent with industry practice and is consistent with Section XI of the ASME Code, and is, therefore, acceptable to the NRC staff. The licensees approach in calculating the Kapplied deviates from the non-mandatory Appendix A approach in Section XI of the ASME Code. However, the NRC staff found no shortcoming in the licensees Kapplied methodology. Since endorsing a methodology for a general application would require much more effort, the NRC staff accepted the embedded flaw methodology by Shah and Kobayashi only in the present application.

3.3 Limit-Load Analysis (Driving force, Failure Resistance, and Flaw Stability)

The licensee used the limit-load analysis of Appendix C (Section XI of the ASME Code) to perform the flaw stability analysis. The allowable flaw size, adjusted for growth over 10, 20, or 30 years, was determined and presented in various charts for surface and embedded flaws and for different locations in the SG primary nozzle region. Examples are provided in the WCAP for using these charts.

The use of limit load analysis in this application is appropriate because stainless steel materials are very ductile. Instead of calculating the final crack size by adding crack growth to the detected flaw size, the WCAPs approach subtracts the crack growth corresponding to 10, 20, or 30 years from the allowable flaw size so that the licensee needs to use only the detected flaw size to determine acceptability from the appropriate chart. This measure is conservative because the WCAPs crack growth calculation is based on the allowable flaw size, which is larger than the growing flaw sizes at different progressing time steps. Unlike LEFM and EPFM analyses where the driving force and failure resistance are clearly defined, the driving force and failure resistance of the limit load analysis can not be separated cleanly. For limit load analysis, the NRC staff considers the growing crack size as the driving force and the allowable flaw size the failure resistance. ASME Code,Section XI rules will be violated when the growing crack size exceeds the allowable flaw size.

3.4 Evaluation of the Detected Flaw Using WCAP-15658-P, Revision 1 Dukes September 22, 2005, response to the NRC staffs RAI evaluated two assumed embedded crack configurations using WCAP-15658-P, Revision 1, Figure A-3.7. The licensees evaluation results showed that the Catawba SG nozzle connection with the assumed 1-inch deep limiting embedded flaw can be operated for 30 years.

The NRC staff has reviewed the licensees bounding flaw evaluation using WCAP-15658-P,

Revision 1 and agreed with the licensees conclusion that the unit can be operated without repair of the subject connection for an additional 30 years till the end of the license.

4.0 CONCLUSION

S The NRC staff has completed the review of the submittal and found that the licensees bounding flaw evaluation meets the rules in Section XI of the ASME Code. Since the allowable flaw size bounds the conservatively assumed flaw size that complements the RT information considering 30 years of crack growth, the NRC staff concludes that Catawba Unit 2 can be operated without repair of the SG inlet nozzle connection for 30 years until the end of the license. As mentioned in the submittal, successive inspections at the flaw location will be conducted during the next three inspection periods in accordance with IWB-2420, Successive Inspections, in Section XI of the ASME Code.

Principal Contributor: S. Sheng Date: January 3, 2006