ML120790647
| ML120790647 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 03/29/2012 |
| From: | Jacob Zimmerman Plant Licensing Branch III |
| To: | Pacilio J Exelon Nuclear |
| Mozafari B, NRR/ADRO/DORL, 415-2020 | |
| References | |
| I3R-09, I3R-20, TAC ME6071, TAC ME6072, TAC ME6073, TAC ME6074 | |
| Download: ML120790647 (4) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 March 29, 2012 Mr. Michael J. Pacilio President and Chief Nuclear Officer Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555
SUBJECT:
BRAIDWOOD STATION, UNITS 1 AND 2 AND BYRON STATION, UNIT NOS. 1 AND 2 - RELIEF REQUESTS 13R-09 AND 13R-20 REGARDING ALTERNATIVE REQUIREMENTS FOR REPAIR OF REACTOR VESSEL HEAD PENETRATIONS (TAC NOS. ME6071, ME6072, ME6073 and ME6074)
Dear Mr. Pacilio:
By letter dated June 14, 2011 (Agencywide Documents Access and Management System Accession No. ML111650286), Exelon Generation Company, LLC (EGC, the licensee),
submitted for U.S. Nuclear Regulatory Commission (NRC) review and approval relief requests (RR) 13R-09 and 13R-20, regarding alternative requirements for repair of reactor vessel head penetrations (VHPs) for Braidwood Station, Units 1 and 2 and Byron Station, Unit Nos. 1 and 2.
The NRC staff has reviewed the subject requests and finds that the licensee has provided sufficient technical basis to support that the proposed alternative repair provides an acceptable level of quality and safety. As described in the enclosed safety evaluation, we conclude that the licensee has adequately addressed all of the regulatory requirements set forth in Title 10 of the Code of Federal Regulations, Section 50.55a(a)(3)(i). Therefore, the NRC authorizes the licensee's proposed alternatives, IR3-09 and IR3-20, for reactor VHPs at Braidwood and Byron U1 and U2, for the remainder of the third ten-year lSI interval, which is scheduled to conclude on July 15, 2016 for Byron Unit Nos. 1 and 2, July 28, 2018 for Braidwood Unit 1 and October 16, 2018, for Braidwood Unit 2.
All other American Society of Mechanical Engineers Code,Section XI, requirements for which relief was not specifically requested and approved remain applicable, including a third-party review by the Authorized Nuclear Inservice Inspector. The NRC staff's safety evaluation is enclosed.
M. Pacilio
- 2 Please contact Brenda Mozafari at (301) 415-2020 if you have any questions on this action.
Sincerely,
~zb Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. STN 50-456, STN 50-457, STN 50-454 and STN 50-455
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 RELIEF REQUEST 13R-09 AND 13R-20 EXELON GENERATION COMPANY. LLC BRAIDWOOD STATION, UNITS 1 AND 2 BYRON STATIONS, UNIT NOS. 1 AND 2 DOCKET NOS. STN 50-454, STN 50-455. STN 50-456 AND STN 50-457
1.0 INTRODUCTION
By letter dated June 14, 2011 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML111650286), Exelon Generation Company, LLC (EGC, the licensee), submitted for U.S. Nuclear Regulatory Commission (NRC, the Commission) review and approval relief requests (RR) 13R-09 and 13R-20, regarding alternative requirements for repair of reactor vessel head penetrations (VHP) at Braidwood Station, Units 1 and 2 and Byron Station, Unit Nos. 1 and 2 (Braidwood and Byron).
2.0 REGULATORY EVALUATION
The inservice inspection (lSI) of the American Society of Mechanical Engineers (ASME) Code Class 1,2 and 3 components is to be performed in accordance with the ASME Boiler & Pressure Vessel Code (Code),Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," and applicable editions and addenda as required by Title 10, Code of Federal Regulations, Part 50, (10 CFR 50.55a(g)), except where specific written relief has been granted by the Commission. Pursuant to 1 0 CFR 50.55a(g)(4), throughout the service life of a pressurized water-cooled nuclear power facility, components which are classified ASME Code Class 1, 2 and 3 must meet the requirements, except the design and access provisions and preservice examination requirements, set forth in the ASME Code,Section XI, to the extent practical within the limitations of design, geometry and materials of construction of the components. Further these regulations require that inservice examination of components and system pressure tests conducted during the first ten-year interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference in paragraph (b) of 10 CFR 50.55a on the date 12 months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein. For Braidwood, the ASME Code of Record for the third ten-year lSI interval, which is currently scheduled to end on July 28, 2018 and October 16, 2018, respectively, is the 2001 Edition through the 2003 Addenda. For Byron, the ASME Code of Record for the third ten-year lSI interval, which is currently scheduled to end on July 15, 2016, is the 2001 Edition through the 2003 Addenda.
Enclosure
- 2 Alternatives to requirements may be authorized or relief granted by the NRC pursuant to 10 CFR 50.55a(a)(3)(i), 10 CFR 50.55a(a){3)(ii), or 10 CFR 50.55a(g)(6)(i). In proposing alternatives or requests for relief, the licensee must demonstrate that: (1) the proposed alternatives would provide an acceptable level of quality and safety; (2) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety; or (3) conformance is impractical for the facility.
By letter dated June 14, 2011, the licensee proposed alternative 13R-09 and 13R-20 in accordance with 10 CFR 50.55a(a)(3)(i) for the repair of reactor pressure VHP nozzles and associated J-groove welds at Braidwood and Byron for the plant's third ten-year lSI interval.
3.0 TECHNICAL EVALUATION
3.1.1 Component Identification Braidwood, VHP Numbers 1 through 78 and Byron, VHP Numbers 1 through 78.
3.1.2 Code Requirements for Which Relief is Requested ASME Code,Section XI, 2001 Edition through 2003 Addenda, 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, ASME Code,Section XI, 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 reactor vessel is the ASME Code,Section III, 1971 Edition through summer 1973 Addenda. The licensee requests relief from the ASME Code,Section III, subparagraphs NB-4131, NB-2538, NB-2539.1, and NB-2539.4 which pertain to the removal of base material defects prior to repair by welding, and NB-4451, NB-4452, and NB-4453.1 which pertain to the removal of weld material defects prior to repair by welding.
3.1.3 Licensee's Proposed Alternative As an alternative, VHPs would be repaired in accordance with Westinghouse Commercial Atomic Power (WCAP)-15987, Revision 2-A, "Technical Basis for the Embedded Flaw Process for Repair of Reactor Vessel Head Penetrations," December 2003 (ADAMS Accession Number ML040290246) as described in Section 5.1 of the licensee's submittal.
3.1.4 Licensee's Duration of Relief Request The licensee requested relief for the remainder of the third ten-year lSI interval at Braidwood U1
& U2, which is currently scheduled to end on July 28,2018, and October 16, 2018, respectively, and Byron U1 & U2, which is currently scheduled to end on July 15, 2016.
3.1.5 Licensee's Basis for Relief The licensee states that the embedded flaw repair technique is considered a permanent repair.
- 3 The licensee believes that as long as a primary water stress corrosion cracking (PWSCC) flaw remains isolated from the primary coolant environment, it cannot propagate. Further, the licensee reasons that since an Alloy 52 or 52M weldment is considered resistant to PWSCC, a new PWSCC flaw cannot initiate and grow through the Alloy 52 or 52M overlay to reconnect the PW environment with the embedded flaw. Structural integrity of the affected J-groove weld and nozzle will be maintained by the remaining unflawed portion of the weld overlay.
The licensee stated that the residual stresses produced by the embedded '1:law technique have been measured and found to be relatively low, indicating that no new flaws will initiate and grow in the area adjacent to the repair weld. Therefore, fatigue-driven crack growth is not a mechanism for further crack growth after the embedded flaw repair process is implemented.
According to the licensee, the small residual stresses produced by the embedded flaw will act constantly, and, therefore, will have no impact on the fatigue effects in this region. Since the residual stress would be additive to the maximum and minimum stress, the stress range will not change, and the already negligible fatigue usage factor for the region will not change.
In letter dated May 3,2007 (ADAMS Accession No. ML071310117), the licensee submitted WCAP-16401, "Technical Basis for Repair Options for Reactor Vessel Head Penetration Nozzles and Attachment Welds: Byron and Braidwood Units 1 and 2," Revision 0, which provides the plant-specific analysis performed for Braidwood and Byron using the same methodology as WCAP-15987-P, Revision 2-A. WCAP-16401 provides the means to evaluate a broad range of postulated repair scenarios to the reactor vessel head penetrations and J-groove welds relative to ASME Code requirements for allowable size and service life.
3.2 NRC STAFF'S EVALUATION The licensee requested authorization of this revised alternative 13R-09 and 13R-20 under 10 CFR 50.55a(a)(3)(i) which states:
The proposed alternatives would provide an acceptable level of quality and safety.
The purpose of the licensee's proposed repair is to address PWSCC, which typically initiates in susceptible materials, such as alloy 600 material and alloy 82/182 weld materials, in areas of tensile stress and certain environmental conditions, such as higher temperatures and corrosive environments. The reactor VHPs and their associated J-groove attachment welds at Braidwood and Byron meet these conditions to be susceptible to PWSCC. The proposed repair technique isolates the susceptible material using a seal weld of alloy 52M weld material, which is less susceptible to PWSCC.
The licensee's basis for the design, implementation, and inspection of the repairs for VHPs is Westinghouse WCAP-15987, Revision 2-A. In a letter dated July 3,2003, from H. N. Berkow (NRC) to H. A. Sepp (Westinghouse Electric Company), (ADAMS Accession No. ML031840237) the NRC staff provided a safety evaluation, in which the NRC staff found WCAP-15987-2 to be acceptable for referencing in licensing applications as an alternative to the 2001 Edition, 2003 Addenda of Section XI of the ASME Code, with the following conditions:
- 1.
Licensees must follow the NRC flaw evaluation guidelines provided in the
-4 R. J. Barrett (NRC) letter to A. Marion (Nuclear Energy Institute), "Flaw Evaluation Guidelines," April 11, 2003. (ADAMS Accession No. ML030980322)
- 2.
The crack growth rate referenced in WCAP-15987-P, Revision 2 is not applicable to Alloy 600 or Alloy 690 weld material, i.e., Alloy 52, 82, 152, and 182 filler material.
- 3.
The nondestructive examination (NDE) requirements listed in the table below must be implemented for examinations of repairs made using the embedded flaw process.
Repair Location Flaw Orientation Repair Weld Repair NDE lSI NDE of the Repair Note 2 VHP Nozzle ID (inside diameter)
Axial Seal UT (ultrasonic testing) and Surface UT or Surface VHP Nozzle ID Circumferential Note 1 Note 1 Note 1 VHP Nozzle OD (outside diameter) above j-groove weld Axial or Circumferential Note 1 Note 1 Note 1 VHP Nozzle OD below j-groove weld Axial or Circumferential Seal UT or Surface UT or Surface j-groove weld Axial Seal UT and Surface, Note 3 UT and Surface, Note 3 j-groove weld Circumferential I Seal UT and Surface, Note 3 UT and Surface, Note 3 Notes: 1.
Repairs must be reviewed and approved separately by the NRC.
- 2.
Inspection consistent with the NRC Order EA-03-009 dated February 11, 2003, and any subsequent changes (ADAMS Accession No. ML030380470).
- 3.
Inspect with personnel and procedures qualified with UT performance-based criteria. Examine the accessible portion of the repaired region. The UT coverage plus surface coverage must equal 100 percent.
The licensee states that their proposed alternative will use the methodology of the NRC approved WCAP-15987 A as described by Section 5.1 of the submittal, with some modification. NRC staff reviewed Section 5.1 to ensure the licensee's proposed actions would meet the requirements of WCAP-15987 A and any modifications would be acceptable under 10 CFR 50.55a(a)(3)(i). As part of this review the NRC staff identified the following technical changes between the requirements of WCAP-15987 A and Section 5.1 of the licensee's submittal:
A. The Alloy 600 tube material with a flaw will be repaired with two Alloy 52 isolation weld layers rather than three layers required in WCAP-15987 A.
- 5 B. A circumferential flaw on the nozzle or tube inside diameter can be repaired using the seal weld technique without additional submission of the repair method for approval by the NRC.
C. Prior to the application of the Alloy 52 or 52M seal weld repair on the RPV clad surface, the stainless steel head cladding will have three beads of 309L stainless steel installed 360 degrees around the interface of the clad and the J-groove weld metal as a buffer layer. The J-groove weld will be covered with three layers of Alloy 52/52M deposited 360 degrees around the nozzle over and extend to the stainless steel buffer layer.
O. In accordance with Notes 2 and 3 of the NRC acceptance for WCAP-15987 P-Revision 2-A, the NOE of the repair will be performed in accordance with ASME Code Case N-729-1, as conditioned by 10 CFR 0.55a(g)(6)(ii}(O).
The NRC staff verified that the changes in methodology identified in 13R-09 and 13R-20 from the previously approved WCAP-15987-2 would still meet the methodology approved by the NRC of an effective embedded flaw repair and provide an acceptable level of quality and safety.
The NRC staff reviewed the licensee's proposal in paragraph A, above, to allow a reduction in the maximum three layers of the seal weld over the Alloy 600 nozzle material to only 2 layers.
The licensee's basis is that the flaw would be isolated from the primary coolant environment necessary for continued PWSCC growth with a less stress being introduced in the base metal with the proposed repair. The NRC staff finds that operational experience has shown that two layers of Alloy 52 material have been sufficient to address dilution layer effects of the high chromium content of the Alloy 52 material, which is the principle reason for the material's resistance to PWSCC. The NRC staff performed a conservative deterministic assessment of hypothetical flaw growth through a similar Alloy 52 inlay material, and showed that flaw growth through the Alloy 52 seal weld would take significantly longer than the reinspection frequency for this repaired penetration nozzle in accordance with 10 CFR 50.55a(g)(6)(ii)(O). In addition, the concern of increased residual stresses in the Alloy 600 material that might still be exposed to primary coolant is a cause for concern for future flaw initiation. Therefore, since the repair can be effective using two weld layers, a smaller seal weld that generates less weld residual stresses in the base metal, would be more preferable than using 3 weld layers. Therefore, the NRC staff finds this change to be acceptable.
The NRC staff reviewed the licensee's proposal in paragraph B, above, which alleviates the licensee from the need to submit a repair plan for each circumferential flaw identified that initiates from the inside diameter of the tube or nozzle surface. The licensee detailed a generic repair plan in accordance with section 5.1.2.b of the licensee's submittal. The licensee's proposed alternative repair would be to partially excavate the flaw to reduce it to an acceptable size, examine it by UT or surface examination, inlay with Alloy 52 or 52M, and examine by UT and surface examination. This is in compliance with embedded flaw technique. Further, operational experience has shown this repair technique is effective in arresting growth of PWSCC flaws. Hence, the NRC staff finds it is an effective generic repair plan to address circumferential flaws that initiate on the inside tube or nozzle surface. Therefore, the NRC staff finds this change to be acceptable.
- 6 The NRC staff reviewed the licensee's proposal in paragraph C above that the stainless steel head cladding will have three beads of stainless steel (SS) 309L buffer layer installed 360 degrees around the interface of the clad and the J-groove weld metal. The licensee's application of this layer is to prevent contamination of the Alloy 52 weld wire which is susceptible to produce fabrication defects in welding. The NRC staff finds the proposed alternative to be appropriate as the Alloy 52 weld is less susceptible to PWSCC than the Alloy 182 weld at the existing nozzle penetration and will fully cover the Alloy 182 weld material. In addition the SS 309L layer, applied only to the periphery of the J-groove weld, will allow for a better-quality seal weld. Therefore, the NRC staff finds this change to be acceptable.
The NRC staff reviewed the licensee's proposal in paragraphs D above, for alternatives for NDE examination requirements of the seal weld and future lSI requirements. During the time period in which WCAP-15987-2 was approved by the NRC staff, the regulatory requirements for upper head inspection were dictated under NRC Order EA-03-009. In September 2008, by rule, the NRC established 10 CFR 50.55a(g)(6)(ii)(D) which 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 nozzles under the current regulatory guidelines of ASME Code Case N-729-1! "Alternative Examination Requirements for PWR Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds. Section X1. Division 1", satisfy the previous NRC conditions on the NDE required for implementation of an embedded flaw repair under WCAP-15987-2. Therefore, the NRC staff finds these changes to be acceptable.
Therefore, the NRC staff finds that the changes in the license's proposed alternative from the NRC approved WCAP-15987-2. 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-15987-2 with a plant specific technical basis for the use of the embedded flaw repair, the licensee previously submitted WCAP-16401. The NRC staff finds WCAP-16401 provides a basis for any remaining ligaments of the flaws identified by the licensee in VHP nozzle base material to be safely encapsulated for 20 years of operation. The NRC staff finds WCAP-16401 provides a basis for any remaining ligaments of the flaws identified by the licensee in VHP nozzle J-groove weld material to be safely encapsulated for 10 years of operation.
The NRC staff has previously approved this proposed alternative repair method as adequate to allow weld repairs of embedded axial and circumferential flaws in the outside diameter surface of VHP nozzles at or below the J-groove weld and similar flaws in the J-groove weld itself at D.C. Cook Unit 1, Beaver Valley Generating Station. Units 1 and 2, San Onofre Power Station, Units 2 and 3, and Byron.
In accordance with the previous NRC conditions imposed on the use of WCAP-15987-2, and plant-specific technical basis for the embedded flaw repair. the NRC staff confirms that the licensee has followed the NRC flaw evaluation guidelines and will implement the appropriate NDE for the repairs to VHP nozzles and their associated J-groove welds at Braidwood and Byron. In accordance with the July 3, 2003, NRC safety evaluation, the embedded flaw repair process is considered to be an alternative to Code requirements that provides an acceptable level of quality and safety. as required by 10 CFR 50.55a(a}(3)(i).
- 7
4.0 CONCLUSION
S As set forth above, the NRC staff concludes that the licensee has provided sufficient technical basis to support that the proposed alternative repair provides an acceptable level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(a)(3)(i). Therefore, the NRC staff authorizes the licensee's proposed alternatives, IR3-09 and IR3-20, for reactor VHPs at Braidwood and Byron, for the remainder of the third ten-year lSI interval which is scheduled to conclude on July 15, 2016 for Byron U1 &U2, July 28,2018 for Braidwood U1 and October 16, 2018 for Braidwood U2.
All other ASME Code,Section XI requirements for which relief was not specifically requested and approved in this relief request remain applicable, including third party review by the Authorized Nuclear Inservice Inspector.
Principal Contributor: J. Collins. NRR Dated: March 29, 2012
ML120790647 NRR-028
- SE memo date OFFICE LPL3-2/PM LPL3-21LA DCI/CPNB/BC*
LPL3-2/BC LPL3-21PM NAME BMozafari KGoldstein TLupold (JTsao for)
JZimmerman BMozafari DATE 3/28112 3/28112 3119/12 3/28/12 3/29/12