ML13101A333
| ML13101A333 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 04/19/2013 |
| From: | Robert Pascarelli Plant Licensing Branch II |
| To: | Gatlin T South Carolina Electric & Gas Co |
| Brown E NRR/DORL/LPL 2-1 | |
| References | |
| RC-12-0165, TAC ME9851 | |
| Download: ML13101A333 (8) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 April 19, 2013 Mr. Thomas D. Gatlin Vice President, Nuclear Operations South Carolina Electric & Gas Company Virgil C. Summer Nuclear Station Post Office Box 88 Jenkinsville, SC 29065
SUBJECT:
VIRGIL C. SUMMER NUCLEAR STATION, UNIT 1, ALTERNATIVE RR-III-09-ALTERNATIVE WELD REPAIR FOR REACTOR VESSEL HEAD PENETRATION (TAC NO. ME9851)(RC-12-0165)
Dear Mr. Gatlin:
By letter dated October 30,2012, as supplemented in letters dated November 5, 14, and 16, 2012 (Agencywide Document Access and Management System (ADAMS) Accession Nos. ML12307A220, ML12319A255, ML12319A256, ML12324A167 and ML12324A168, ML12325A056), South Carolina Electric &Gas Company (the licensee) submitted proposed alternative Relief Request (RR)-11I-09. The licensee requested relief from Paragraph IWA-4420, "Defect Removal Requirements," of Section XI to the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code). This paragraph requires that defects be removed or mitigated. The licensee proposed to use the provisions of Westinghouse Topical Report WCAP-15987-P-A, Revision 2, for performing an embedded flaw repair of the indications found in control rod drive mechanism (CRDM) penetration nozzle tubes 19, 31, 37, and 52 at Virgil C. Summer Nuclear Station, Unit 1 (VCNS1). The licensee has submitted the proposed alternative on the basis that the proposed alternative provides an acceptable level of quality and safety, in accordance with Section 50.55a(a)(3)(i) to Title 10 of the Code of Federal Regulations (10 CFR).
On November 16, 2012 (ADAMS Accession No. ML12325A432), the NRC staff granted verbal authorization, to allow the licensee to perform overlay repairs to reactor vessel head penetrations. Based on our review of your submittals, we have concluded that the alternative proposed in RR-III-09 provides an acceptable level of quality and safety and, therefore, it is authorized pursuant to 10 CFR 50.55a(a)(3)(i).
T. Gatlin
- 2 These reliefs are authorized for the remainder of the third 1 O-year lSI interval at VCNS 1, which began December 31, 2003, and ends December 31, 2013.
Sincerely, IRAJ Robert J. Pascarelli, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-395
Enclosure:
Safety Evaluation cc w/encl: Distribution via Listserv
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REQUEST FOR RELIEF (RR-III-09)
ALTERNATIVE WELD REPAIR FOR REACTOR PRESSURE VESSEL HEAD PENETRATION SOUTH CAROLINA ELECTRIC &GAS COMPANY SOUTH CAROLINA PUBLIC SERVICE AUTHORITY VIRGIL C. SUMMER NUCLEAR STATION, UNIT 1 DOCKET NO. 50-395
1.0 INTRODUCTION
By letter dated October 30,2012, as supplemented in letters dated November 5, 14, and 16, 2012 (Agencywide Document Access and Management System (ADAMS) Accession Nos. ML12307A220, ML12319A255, ML12319A256, ML12324A167 and ML12324A168, ML12325A056), South Carolina Electric & Gas Company (the licensee) submitted proposed alternative Relief Request (RR)-11I-09. The licensee requested relief from Paragraph IWA-4420, "Defect Removal Requirements," of Section XI to the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code). This paragraph requires that defects be removed or mitigated. The licensee proposed to use the provisions of Westinghouse Topical Report WCAP-15987-P-A, Revision 2, for performing an embedded flaw repair of the indications found in control rod drive mechanism (CRDM) penetration nozzle tubes 19, 31, 37, and 52 at Virgil C. Summer Nuclear Station, Unit 1 (VCNS1). The licensee has submitted the proposed alternative on the basis that the proposed alternative provides an acceptable level of quality and safety, in accordance with Section 50.55a(a)(3)(i) to Title 10 ofthe Code of Federal Regulations (10 CFR).
On November 16,2012 (ADAMS Accession No. ML12325A432), the NRC staff granted verbal authorization, to allow the licensee to perform overlay repairs to reactor vessel head penetrations.
2.0 REGULATORY EVALUATION
Section 50.55a(g)(4), In service Inspection Requirements of 10 CFR requires that ASME Code Class 1, 2, and 3 components (including supports) shall meet the requirements, except the design and access provisions and the pre-service examination requirements, set forth in the ASME Code,Section XI, "Rules for In service Inspection of Nuclear Power Plant Components,"
to the extent practical within the limitations of design, geometry, and materials of construction of
-2 the components. The regulations require that inservice examination of components and system pressure tests conducted during the first 1 O-year inspection interval and subsequent 10-year inspection intervals comply with the requirements in the latest edition and addenda of Section XI of the 10 CFR 50.55a(a)(3) of 10 CFR 50 states, in part, that alternatives to the requirements of ASME Code incorporated by reference in 10 CFR 50.55a(b) 12 months prior to the start of the 120-month inspection interval, subject to the limitations and modifications listed therein.
Section 50.55a(g) of 10 CFR may be used, when authorized by the NRC, if (i) the proposed alternatives would provide an acceptable level of quality and safety or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
Based on analysis of the regulatory requirements, the staff finds that the NRC has the regulatory authority to authorize the proposed alternative on the basis that it provides an acceptable level of quality and safety. Accordingly, the staff has reviewed the licensee's proposed alternative pursuant to 10 CFR 50.55a(a)(3)(i).
The licensee cited another facilities' verbal authorization as precedent for this action. The NRC staff did not significantly consider the prior authorization in the determination of acceptability of this request.
3.0 TECHNICAL EVALUATION
3.1 Licensee's Request for Alternative 3.1.1 Components for which Relief is Being Requested Vessel Head Penetrations (VHP) Numbers 19, 31, 37 and 52, ASME Code Case N-729-1, Table No.1, Item Number B4.20 3.1.2 ASME Code Requirements The code of record for the VCNS1 third 10-year Inservice Inspection (lSI) interval that is scheduled to end on December 31,2013, is the 1998 Edition through the 2000 Addenda of the ASME Code,Section XI. Subparagraph IWA-4000 contains requirements for the removal of defects from and welded repairs performed on ASME Code components. For the removal or mitigation of defects by welding, IWA-4411 requires that repairs and installation of replacement items shall be performed in accordance with the Owner's Design Specification and the original Construction Code of the component or system.
The original construction code of the VCSNS1, reactor vessel is the ASME Code,Section III, 1971 Edition with no addenda. Subparagraphs NB-4131, NB-2538, NB-2539.1, and NB-2539.4 pertain to the removal of base material defects prior to repair by welding, and NB-4451, NB-4452, and NB-4453.1 pertain to the removal of weld material defects prior to repair by welding.
- 3 3.1.3 Licensee's Proposed Alternative The licensee proposes to repair the subject VHPs in accordance with WCAP-15987, as described and modified in Section 5.1 of the licensee's submittal. The cited topical report has been previously reviewed and generically approved for use in a safety evaluation by the NRC staff.
3.1.4 Licensee's Basis for Requesting Relief The licensee has conducted examinations of the reactor vessel head penetrations (VHPs) in accordance with ASME Code Case N-729-1, as required by 10 CFR 50.55a(g)(6)(ii)(D).
Unacceptable flaws were found in four VHP nozzles, and these needed to be repaired prior to returning the vessel head to service. The licensee proposes to use the embedded flaw process as an alternative to the defect removal requirements of ASME Code,Section XI and Section III.
The embedded flaw repair technique is considered a permanent repair since as long as a primary water stress corrosion cracking (PWSCC) flaw remains isolated from the primary water environment, it cannot propagate. Since an Alloy 52 weldment is considered highly resistant to PWSCC, a new PWSCC flaw should not initiate and grow through the Alloy 52 seal weld to reconnect the primary water environment with the embedded flaw. The structural integrity of the affected J-Groove weld and/or nozzle will be maintained by the remaining unflawed portion of the weld and/or the VHP.
3.2
NRC Staff Evaluation
PWSCC of nickel-based pressure boundary materials is a safety concern. Operational experience has shown that PWSCC can occur as the result of the combination of susceptible material, corrosive environment, and tensile stresses, resulting in leakage and the potential for loss of structural integrity. The subject RVH penetrations meet these conditions thus may be susceptible to PWSCC. Examination per ASME Code Case N-729-1, as required by 10 CFR 50.55a(g)(6)(ii)(D), is intended to ensure the structural integrity and leak tightness of RVH penetrations.
The licensee's ultrasonic (UT) examination of the VHP nozzle tubes, in accordance with ASME Code Case N-729-1, found unacceptable indications in 4 nozzle tubes. The penetrations need to be repaired prior to returning the vessel head to service. In order to perform the design, implementation, and inspection of the VHP repairs, the licensee proposes to use the provisions of WCAP-15987. The NRC staff has reviewed the technical basis of WCAP-15987 and has generically accepted it, subject to the specified limitations and conditions, for referencing in licensing applications as an alternative to Section XI of the ASME Code (Reference 5).
The licensee states that the proposed alternative, as described by Section 5.1 of the submittal, will use the methodology of the NRC approved WCAP-15987 with modifications. The NRC staff reviewed Section 5.1 of the proposed alternative to ensure the licensee's proposed actions would meet the requirements ofWCAP-15987, and that any modifications would be acceptable.
As part of this review the NRC staff identified the following technical changes:
- 1. Cracks in the Alloy 600 penetration nozzle tube material will be embedded with two weld layers of Alloy 52 rather than the three layers specified in WCAP-15987;
-4
- 2. A single layer consisting of at least 3 beads of stainless steel 309L will be installed on the reactor vessel head clad surface 360 degrees around at a distance of approximately 0.5 inches from the toe of the J-groove weld prior to deposition of the first Alloy 52 layer;
- and,
- 3. Nondestructive examination of the repair will be performed in accordance with ASME Code Case N-729-1, as required by 10 CFR 50.55a(g)(6)(ii)(D).
The NRC staff reviewed the licensee's proposal to allow a reduction in the maximum three layers of the seal weld over the Alloy 600 nozzle material to only 2 layers. The NRC staff finds that operational experience has shown that two layers of Alloy 52 material deposited on Alloy 600 material are adequate to maintain the high chromium content of the Alloy 52 material, the principle reason for the material's resistance to PWSCC, even after weld dilution effects are considered. The NRC staff finds that the embedded flaw will be isolated from the primary coolant environment necessary for continued PWSCC growth and a lower residual stress will be introduced in the base metal with the proposed repair. Therefore, the NRC staff finds that two layers of Alloy 52 are acceptable.
The licensee indicated that for the application of 309L weld metal on the RPV cladding, that the 309L is used to mitigate the risk of solidification cracking or hot cracking in the region where the Alloy 52 weld beads intersect the RPV cladding, insulating the outermost Alloy 52 weld beads from possible contaminants in the cladding. The NRC staff finds the 309L layer, applied only to the periphery, will allow for a quality seal weld by mitigating the potential for solidification cracking and hot cracking. The NRC staff notes that this technique has been previously accepted for another licensee. Therefore, the NRC staff finds that the application of the 309L weld metal is acceptable.
The NRC staff reviewed the licensee's proposed alternative for NDE examination requirements of the seal weld and future lSI requirements. During the time period in which WCAP-15987 was approved by the NRC staff, the regulatory requirements for upper head inspection were found under NRC Order EA-03-009. In September 2008, by rule, the NRC established 10 CFR 50.55a(g)(6)(ii)(D) that defines the current regulatory requirements for upper head inspections and rescinded NRC Order EA-03-009. The NRC staff finds that the licensee's proposed alternative inspections for the upper head penetration nozzle conform to the current regulatory requirements and satisfy the previous NRC limitations on the NDE required for implementation of an embedded flaw repair under WCAP-15987, thus the NRC staff finds them acceptable.
The NRC staff finds that the changes in the license's proposed alternative from the NRC-approved WCAP-15987 either meet or provide additional quality for the embedded flaw repair technique and, as such, provide an acceptable level of quality and safety.
In order to support the use of WCAP-15987, the licensee has submitted a plant-specific technical basis, Westinghouse LTR-PAFM-12-137-NP, "Technical Basis for Westinghouse Embedded Flaw Repair for V. C. Summer Unit 1 Reactor Vessel Head Penetration Nozzles" as to the submittal. The submission evaluated the expected lifetime based on fatigue crack growth of flaws in the penetration nozzle, but not in the J-groove weld. In a letter dated November 5,2012, the licensee provided inspection information for the position and length of the flaws, and stated that the PT examinations showed that the flaws had not propagated to the
- 5 toe of the J-groove weld, thus a flawed J-groove weld did not need to be considered in the lifetime calculation.
The NRC staff reviewed the UT examination data provided by the licensee in Attachment 2 to the November 5, 2012, responses. The UT data showed that Indication No.1 in Penetration 37 had propagated past the toe of the J-groove weld. Since nondestructive examination techniques are not currently able to volumetrically determine the extent of cracking in the J-groove weld, the staff requested information concerning a lifetime analysis for a flaw which bounded the flaw in the penetration nozzle and a completely cracked J-groove weld. In letters dated November 14, 2012, the licensee provided Revision 2 of the site-specific flaw analysis assuming that the entire J-groove weld was cracked in addition to the UT examination indication. The results of this analysis showed that the fatigue crack growth of the flaw and cracked J-groove weld is excess of 20 years.
The NRC staff examined the results of this calculation and noted that a previous flaw analysis (ADAMS Accession No. ML112990783) had a predicted "at least 10 years of service life time" for a similar flaw size, a value significantly less than the lifetime "in excess of 20 years" calculated. In a letter dated November 16, 2012, the licensee stated that the previous calculation only considered the five most significant thermal transients out of the full set of normal/upset thermal transients and accounted for the remaining transients by adding design cycles for conservatism. The licensee provided data for the VCNS1 stress intensity factor and number of cycles for each of the three most important thermal transients. The NRC staff has examined the data provided and finds that a fatigue life in excess of 20 years is reasonable.
The NRC staff accepts the licensee's explanation for the plant-specific differences and finds that the present 20-year life time prediction calculation is acceptable, thus the NRC staff accepts the plant-specific lifetime calculation.
In summary, the NRC staff concludes that the proposed alternative for embedded flaw repair of the CRDM nozzle indications provides an acceptable level of quality and safety in accordance with 10 CFR 50.55a(a)(3)(i).
4.0 CONCLUSION
As set forth above, the NRC staff finds that the proposed alternative for embedded flaw repair of indications found on control rod drive mechanism penetration tube numbers 19, 31, 37, and 52 will provide an acceptable level of quality and safety. Accordingly, the staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(a)(3)(i) and therefore authorizes the use of the proposed alternative at VCSN1, for the remainder of the third 10-year inservice inspection interval that ends December 31, 2013.
All other ASME Code,Section XI requirements for which relief was not specifically requested and authorized in the subject proposed alternative remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
Principal Contributors: Jay Wallace, NRR Date: April 19, 2013
T. Gatlin
- 2 These reliefs are authorized for the remainder of the third 10-year lSI interval at VCNS 1, which began December 31, 2003, and ends December 31, 2013.
Sincerely, IRAJ Robert J. Pascarelli, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-395
Enclosure:
Safety Evaluation cc w/encl: Distribution via Listserv DISTRIBUTION:
PUBLIC LPL2-1 RlF RidsNrrPMSummer Resource RidsAcrs_AcnwMailCTR Resource RidsNrrLASFigueroa Resource RidsOgcRp Resource RidsNrrDorlDpr Resource RidsRgn2MailCenter Resource RidsNrrDeEpnb Resource JCollins, NRR
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RPascarelli DATE 04/16/13 04/16/13 03/15/13 04/19/13 OFFICIAL RECORD COpy