ML053070118
| 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, 2006Mr. Dhiaa JamilVice President Catawba Nuclear Station Duke Energy Corporation 4800 Concord Road York, SC 29745
SUBJECT:
CATAWBA NUCLEAR STATION, UNIT 2, SECOND 10-YEAR INSERVICEINSPECTION 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 EnergyCorporation (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 unit's 13th refueling outage. Duke intended to demonstratethrough 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 operatedwithout 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 theflaw 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 allowableflaw 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 CatawbaUnit 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 beconducted during the next three inspection periods in accordance with IWB-2420, "Successive Inspections," in Section XI of the ASME Code.
D. Jamil-2-The enclosed Safety Evaluation contains the NRC staff's evaluation and conclusions.Sincerely,/RA/Evangelos C. Marinos, ChiefPlant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor RegulationDocket No. 50-414
Enclosure:
As stated cc w/encl: See next page D. Jamil-2-The enclosed Safety Evaluation contains the NRC staff's evaluation and conclusions.Sincerely,/RA/Evangelos C. Marinos, ChiefPlant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor RegulationDocket No. 50-414
Enclosure:
As stated cc w/encl: See next pageDISTRIBUTION:
Public LPL2-1 R/F RidsNrrDorlLplc(EMarinos)
RidsNrrPMJStang RidsNrrLAMOBrien RidsOgcRp RidsAcrsAcnwMailCenter RidsNrrDpr RidsRgn2MailCenter(MErnstes)
SSheng MMitchell BWetzel, EDO, RIIADAMS Accession No: ML053070118NRR-106OFFICENRR/LPL2-1/PMNRR/LPL2-1/LANRR/EMCB/SCNRR/LPL2-1/BC NAMEJStang: ckgMO'BrienMMitchell byMemo datedEMarinosDATE12/29/0512/29/05 10/11/0501/03/06OFFICIAL AGENCY RECORD Catawba Nuclear Station, Units 1 & 2 Page 1 of 2cc: w/encl.
Mr. Lee Keller, ManagerRegulatory Compliance Duke Energy Corporation 4800 Concord Road York, South Carolina 29745Ms. Lisa F. VaughnDuke Energy Corporation 526 South Church Street P. O. Box 1006 Mail Code = EC07H Charlotte, North Carolina 28201-1006North Carolina Municipal Power Agency Number 1 1427 Meadowwood Boulevard P.O. Box 29513 Raleigh, North Carolina 27626County Manager of York CountyYork County Courthouse York, South Carolina 29745Piedmont Municipal Power Agency 121 Village DriveGreer, South Carolina 29651Ms. Karen E. LongAssistant Attorney General North Carolina Department of Justice
P.O. Box 629 Raleigh, North Carolina 27602NCEM REP Program Manager4713 Mail Service Center Raleigh, North Carolina 27699-4713North Carolina Electric Membership Corp.P.O. Box 27306 Raleigh, North Carolina 27611Senior Resident InspectorU.S. Nuclear Regulatory Commission 4830 Concord Road York, South Carolina 29745Mr. Henry Porter, Assistant DirectorDivision of Waste Management Bureau of Land and Waste Management Dept. of Health and Environmental Control 2600 Bull Street Columbia, South Carolina 29201-1708Mr. R.L. Gill, Jr., Manager Nuclear Regulatory Issues and Industry Affairs Duke Energy Corporation 526 South Church Street Mail Stop EC05P Charlotte, North Carolina 28202Saluda River Electric P.O. Box 929 Laurens, South Carolina 29360Mr. Peter R. Harden, IV, Vice PresidentCustomer Relations and Sales Westinghouse Electric Company 6000 Fairview Road 12th Floor Charlotte, North Carolina 28210Mr. T. Richard PuryearOwners Group (NCEMC)
Duke Energy Corporation 4800 Concord Road York, South Carolina 29745 Catawba Nuclear Station, Units 1 & 2 Page 2 of 2cc: w/encl.
Division of Radiation ProtectionNC Dept. of Environment, Health, and Natural Resources 3825 Barrett Drive Raleigh, North Carolina 27609-7721Mr. Henry BarronGroup Vice President, Nuclear Generation and Chief Nuclear Officer P.O. Box 1006-EC07H Charlotte, NC 28201-1006Diane CurranHarmon, Curran, Spielbergy &
Eisenberg, LLP 1726 M Street, NW Suite 600 Washington, DC 20036 EnclosureSAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATIONSECOND 10-YEAR INSERVICE INSPECTION INTERVALSTEAM GENERATOR C HOT LEG NOZZLE WELDS FLAW EVALUATIONCATAWBA NUCLEAR STATION, UNIT 2DUKE ENERGY CORPORATIONDOCKET NO. 50-41
41.0 INTRODUCTION
By letters dated October 19 and December 2, 2004, and September 22, 2005 (AgencywideDocuments 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 unit's 13th refueling outage. Dukeintended 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 thesubject 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 PressureVessel Code (ASME Code), Class 1, 2 and 3 components shall be performed in accordancewith 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 writtenrelief 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 (includingsupports) 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 30 years until the end of the license without repair of the reactor coolant hot leg to SG inletnozzle connection.
3.0 TECHNICAL EVALUATION
A typical flaw evaluation for detected flaws includes five elements: (1) flaw sizing; (2) theapplied stress intensity factor (K applied) 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 evaluationusing Nuclear Regulatory Commission (NRC)-approved acceptance criteria includingappropriate structural factors. The five elements of the licensee's flaw evaluation are evaluated in the following sections. Since the detected flaw is circumferential and located in the nozzleinterface between the stainless steel safe end (including buttering) and the stainless steel field weld, the NRC staff will evaluate the part of the licensee's methodology applicable tocircumferential embedded flaws in stainless steel materials only.
3.1 Flaw SizingThe licensee based it's flaw sizing on its 2004 radiographic test (RT) results, which showed thatthe 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 licensee's September 22, 2005, response to the NRC staff's Request for Additional Information (RAI) confirmed that the flawwas 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 oflinear.The NRC staff's review confirmed that the licensee's flaw characterization was in accordancewith 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 tothe 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 licenseehas established qualitatively that the indication is at worst a linear flaw with narrow depth andwidth and the licensee's 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. Duke's fatigue crack growth calculation employed the fatigue crack growth rate (CGR) for austenitic steels in air environment from Appendix C of Section XIof the ASME Code, of which the key parameter K applied was calculated considering thermal anddeadweight 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 licensee's
K applied calculations were based on a paper by Shah and Kobayashi (1971), which is applicableto cases with embedded flaws not too close to the nozzle inner-diameter or outer-diameter surface. Duke's approach in calculating the CGR as described above is consistent with industry practiceand is consistent with Section XI of the ASME Code, and is, therefore, acceptable to the NRCstaff. The licensee's approach in calculating the K applied deviates from the non-mandatoryAppendix A approach in Section XI of the ASME Code. However, the NRC staff f ound noshortcoming in the licensee's K applied methodology. Since endorsing a methodology for ageneral application would require much more effort, the NRC staff accepted the embedded flawmethodology 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) toperform the flaw stability analysis. The allowable flaw size, adjusted for growth over 10, 20, or30 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 materialsare very ductile. Instead of calculating the final crack size by adding crack growth to the detected flaw size, the WCAP's 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 flawsize to determine acceptability from the appropriate chart. This measure is conservativebecause the WCAP's 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 sizethe failure resistance. ASME Code,Section XI rules will be violated when the growing cracksize exceeds the allowable flaw size. 3.4 Evaluation of the Detected Flaw Using WCAP-15658-P, Revision 1 Duke's September 22, 2005, response to the NRC staff's RAI evaluated two assumedembedded crack configurations using WCAP-15658-P, Revision 1, Figure A-3.7. The licensee's evaluation results showed that the Catawba SG nozzle connection with the assumed1-inch deep limiting embedded flaw can be operated for 30 years.The NRC staff has reviewed the licensee's bounding flaw evaluation using WCAP-15658-P, Revision 1 and agreed with the licensee's conclusion that the unit can be operated withoutrepair of the subject connection for an additional 30 years till the end of the license.
4.0 CONCLUSION
SThe NRC staff has completed the review of the submittal and found that the licensee'sbounding 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 beoperated 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 beconducted 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