ML14084A549

Braidwood Station, Units 1 and 2 - Relief from the Requirements of the ASME Code (TAC MF2768 and MF2769)

ML14084A549
Person / Time
Site:	Braidwood
Issue date:	04/28/2014
From:	Tate T L Plant Licensing Branch III
To:	Pacilio M J Exelon Generation Co
References
TAC MF2768, TAC MF2769
Download: ML14084A549 (12)

Text

Mr. Michael J. Pacilio Senior Vice President UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 April 28, 2014 Exelon Generation Company, LLC President and Chief Nuclear Officer (CNO) Exelon Generation Company, LLC 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

BRAIDWOOD STATION, UNITS 1 AND 2-RELIEF FROM THE REQUIREMENTS OF THE ASME CODE (TAC NOS. MF2768 AND MF2769)

Dear Mr. Pacilio:

By letter dated September 19, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 13263A372), with a supplement dated January 24, 2014 (ADAMS Accession No. ML 14024A588), Exelon Generation Company, LLC (EGC) requested relief from IWA-4000 of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) requirements at Braidwood Station, Units 1 and 2. Specifically, pursuant to Title 10 of the Code of Federal Regulations (1 0 CFR) 50.55a(a)(3)(ii), the licensee requested to use an alternative to repair degraded canopy seal welds in lieu of defect removal requirements in the ASME Code,Section XI, IWA-4422, 1, on the basis that complying with the specified requirement would result in hardship or unusual difficulty.

The NRC staff has reviewed the subject request and concludes, as set forth in the enclosed safety evaluation, that the proposed alternative provides reasonable assurance of leak tightness of the subject component and that complying with the specified requirement would result in hardship or unusual difficulty without a compensating increase m the level of quality and safety. All other ASME Code,Section XI, requirements for which relief was not specifically requested and approved in the subject requests remain applicable, including third-party review by the Authorized Nuclear lnservice Inspector.

The NRC staff has determined that EGC has adequately addressed all of the regulatory requirements set forth in 10 CFR) 50.55a(a)(3)(ii).

J. Pacilio If you have any questions, please contact the Senior Project Manager, Joel S. Wiebe, at 301-415-6606 or via e-mail at Joei.Wiebe@nrc.gov.

Docket Nos. 50-456 and 50-457

Enclosure:

Safety Evaluation cc w/encl: Distribution via ListServ Sincerely, '

Travis L. Tate, Chief Plant Licensing 111-2 and Planning and Analysis Branch Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST 13R-11 REGARDING CONTROL ROD DRIVE MECHANISM CANOPY SEAL WELD

1.0 INTRODUCTION

EXELON GENERATION COMPANY. LLC BRAIDWOOD STATION. UNITS 1 AND 2 DOCKET NOS. 50-456 AND 50-457 By letter dated September 19, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 13263A372), with a supplement dated January 24, 2014 (ADAMS Accession No. ML 14024A588), Exelon Generation Company, LLC (the licensee) requested relief from IWA-4000 of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) for the contingency repair of degraded canopy seal welds associated with the control rod drive mechanism (CRDM) nozzle penetrations at Braidwood Station, Units 1 and 2 Specifically, pursuant to Title 10 of the Code of Federal Regulations (1 0 CFR) 50.55a(a)(3)(ii), the licensee requested to use the proposed alternative in relief request (RR) 13R-11 to repair degraded canopy seal welds in lieu of defect removal requirements in the ASME Code,Section XI, IWA-4422.1, on the basis that complying with the specified requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

2.0 REGULATORY EVALUATION

In the September 19, 2013, letter, the licensee submitted the proposed request pursuant to 10 CFR 50.55a(a)(3)(i).

However, as part of the response to the U.S. Nuclear Regulatory Commission (NRC) staff's request for additional information, the licensee requested NRC staff review the RR pursuant to 10 CFR 50.55a(a)(3)(ii).

The NRC staff finds that it is more appropriate to submit the proposed relief request under Paragraph (3)(ii) than (3)(i) because the basis for the proposed alternative is in accordance with the requirements of Paragraph (3)(ii) rather than that of (3)(i). Pursuant to 10 CFR 50.55a(g)(4), lnservice Inspection (lSI) Requirements, ASME Code Class 1, 2, and 3, components (including supports) shall meet the requirements.

except the design and access provisions, and the pre-serv1ce examination requirements set forth in the ASME Code,Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components." to the Enclosure extent practical within the limitations of design, geometry, and materials of construction ofthe components.

The regulations require that inservice examination of components and system pressure tests conducted during the first 1 0-year inspection interval, and subsequent 1 0-year inspection intervals, comply with the requirements in the latest edition and addenda of Section XI of the 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 conditions listed therein. Section 50.55a(a)(3) of 10 CFR states, in part, that alternatives to the requirements of Paragraph (g) of 10 CFR 50.55a may be authorized by the NRC if the licensee demonstrates that: (i) the proposed alternative provides 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. 3.0 TECHNICAL EVALUATION

3.1 Licensee's

Alternative The affected components are ASME Code Class 1 canopy seal welds associated with the CRDM. The lSI program for the third 1 0-year lSI interval is based on the ASME Code Section XI, 2001 Edition through the 2003 Addenda. The licensee designed and fabricated the CRDM assemblies to the ASME Code, Section Ill, 1974 Edition through Summer 1974 Addenda. The ASME Code,Section XI, IWA-4000, requires that repairs be performed in accordance with the owner's original construction code of the component or system, or later editions and addenda of the code. The canopy seal weld repair would require the following activities in accordance with the ASMECode:

a. Excavation of the rejectable indications in the weld, b. A surface examination of the excavated areas of the weld, c. Rewelding and restoration to the original configuration and materials, and d. Final surface examination.

The licensee stated that due to the nature of the flaw, the excavation of the leaking portion of the weld would necessitate a cavity that extends completely through wall. A surface examination using liquid penetrant testing (PT) of this cavity is required in accordance with ASME Code,Section XI, to verify the removal of the rejectable flaw or to verify that the flaw is removed or reduced to an acceptable size. The PT examination would deposit the penetrant materials onto the inner surfaces of the original seal weld. The licensee noted that the penetrant material would not be readily removed prior to rewelding due to the inaccessibility of the inside surface. The remaining penetrant material in the original seal weld would introduce contaminants to the new weld metal and reduce the quality of the repair weld. The licensee stated that the configuration of the canopy assembly would prevent the establishment and maintenance of an adequate back-purge during the welding process and would further reduce the quality of the repair weld. The licensee stated that the high radiological dose associated with a CRDM canopy seal weld repair in accordance with the ASME Code would be contrary to the as low as reasonably achievable (ALARA) radiological controls program. In order to reduce the exposure to personnel involved in the welding process, most of the repair activities would be performed remotely using robotic equipment to the extent practical.

However, the required excavation and PT examinations would necessitate hands on access to the canopy seal weld. The licensee estimated the total dose for these activities is in excess of 0.600 person-Rem per one weld repair. The ASME Code,Section XI, IWA-4200, requires that the repair material conform to the original design specification or the ASME Code, Section Ill. As stated in its request dated September 19, 2013, the licensee will use applicable portions of ASME Code Case N-504-4, "Alternative Rules for Repair of Class 1, 2, and 3, Austenitic Stainless Steel Piping,Section XI, Division 1 ,"for repair by weld overlay to provide a new leakage barrier. In lieu of performance of PT examinations of CRDM seal weld repairs or replacement, the licensee further stated it will perform a visual examination after the welding is completed.

The licensee stated that it will use nickel-based Alloy 52/52M weld material rather than the austenitic stainless steel that is required by ASME Code Case N-504-4 because of its resistance to stress corrosion cracking (SCC). The licensee noted that industry experience with failure analyses performed on leaking canopy seal welds removed from service at other plants has attributed the majority of the cases to transgranular SCC. Industry experience shows the size of the opening where the leakage occurs has been extremely small, normally a few thousandths of an inch. The crack orientations vary, but often radiate outward such that a pinhole appears on the surface as opposed to a long crack. The SCC results from exposure of a susceptible material to residual stress, which is often concentrated by weld discontinuities, and to a corrosive environment, such as water trapped in the cavity behind the seal weld that is mixed with the air initially in the cavity, resulting in higher oxygen content than is in the bulk primary coolant. The licensee stated that it will not remove the CRDM canopy seal weld flaws, but will perform an analysis of the repaired weldment before entering MODE 4, using Paragraph (g) of ASME Code Case N-504-4 to assure that the remaining flaw will not propagate unacceptably.

The licensee noted that the canopy seal weld is not a structural weld, nor a pressure-retaining weld, but provides a seal to prevent reactor coolant leakage if the mechanical joint leaks. The licensee stated that the weld buildup is considered a repair in accordance with IWA-411 0, which requires application of the original code of construction or design specification because the weld is performed on an appurtenance to a pressure retaining component.

The licensee stated it will use a gas tungsten arc welding (GTAW) process controlled remotely for the repair. Should the need arise, the licensee stated it may use a manual GTAW repair. The licensee stated that it will visually examine the repaired weld using methods and personnel qualified to the standards of ASME VT-1 visual examination requirements.

The licensee stated that a post-maintenance VT-2 visual examination will be performed at normal operating temperature and pressure during the system leakage test in lieu of the hydrostatic test specified in Paragraph (h) of ASME Code Case N-504-4. The licensee stated that the high radiological dose associated with a CRDM canopy seal weld repair in accordance with the ASME Code would be contrary to the as low as reasonably achievable (ALARA) radiological controls program. In order to reduce the exposure to personnel involved in the welding process, most of the repair activities would be performed remotely using robotic equipment to the extent practical.

However, the required excavation and PT examinations would necessitate hands on access to the canopy seal weld. The licensee estimated the total dose for these activities is in excess of 0.600 person-Rem per one weld repair. The ASME Code,Section XI, IWA-4200, requires that the repair material conform to the original design specification or the ASME Code, Section Ill. As stated in its request dated September 19, 2013, the licensee will use applicable portions of ASME Code Case N-504-4, "Alternative Rules for Repair of Class 1, 2, and 3, Austenitic Stainless Steel Piping,Section XI, Division 1 ,"for repair by weld overlay to provide a new leakage barrier. In lieu of performance of PT examinations of CRDM seal weld repairs or replacement, the licensee further stated it will perform a visual examination after the welding is completed.

The licensee stated that it will use nickel-based Alloy 52/52M weld material rather than the austenitic stainless steel that is required by ASME Code Case N-504-4 because of its resistance to stress corrosion cracking (SCC). The licensee noted that industry experience with failure analyses performed on leaking canopy seal welds removed from service at other plants has attributed the majority of the cases to transgranular SCC. Industry experience shows the size of the opening where the leakage occurs has been extremely small, normally a few thousandths of an inch. The crack orientations vary, but often radiate outward such that a pinhole appears on the surface as opposed to a long crack. The sec results from exposure of a susceptible material to residual stress, which is often concentrated by weld discontinuities, and to a corrosive environment, such as water trapped in the cavity behind the seal weld that is mixed with the air initially in the cavity, resulting in higher oxygen content than is in the bulk primary coolant. The licensee stated that it will not remove the CRDM canopy seal weld flaws, but will perform an analysis of the repaired weldment before entering MODE 4, using Paragraph (g) of ASME Code Case N-504-4 to assure that the remaining flaw will not propagate unacceptably.

The licensee noted that the canopy seal weld is not a structural weld, nor a pressure-retaining weld, but provides a seal to prevent reactor coolant leakage if the mechanical joint leaks. The licensee stated that the weld buildup is considered a repair in accordance with IWA-411 0, which requires application of the original code of construction or design specification because the weld is performed on an appurtenance to a pressure retaining component.

The licensee stated it will use a gas tungsten arc welding (GTAW) process controlled remotely for the repair. Should the need arise, the licensee stated it may use a manual GTAW repair. The licensee stated that it will visually examine the repaired weld using methods and personnel qualified to the standards of ASME VT-1 visual examination requirements.

The licensee stated that a post-maintenance VT-2 visual examination will be performed at normal operating temperature and pressure during the system leakage test in lieu of the hydrostatic test specified in Paragraph (h) of ASME Code Case N-504-4. The licensee requested the alternative to be approved for the remainder of the Third 10-Year Inspection Interval for Braidwood Station, Units 1 and 2, which is currently scheduled to end on July 28, 2018, for Unit 1, and October 16, 2018, for Unit 2. 3.2 NRC Staff Evaluation The NRC staff evaluated the proposed alternative to demonstrate that: ( 1) the proposed seal weld is intended to provide leak tightness and is not relied upon for structural integrity, (2) the seal weld is designed to ensure leak tightness, (3) the weld material is appropriate to ensure leak tightness of the weld, (4) the welding requirements satisfy the appropriate ASME Code, (5) the proposed nondestructive examinations ensure leak tightness, (6) the flaw growth calculation ensures leak tightness of the weld for a sufficient amount of time, (7) the overall design; operation, and inspection activities at the plant demonstrate defense-in-depth even if a leak should occur, and (8) hardship justification is sufficient.

The licensee noted that the industry operating experience has shown that canopy seal degradation has been characterized as pinhole leaks (i.e., rounded indications) with no significant flaws extending to the outside weld surface. Currently, canopy seal leaks have been limited to welds at the lower and middle canopy seals. The licensee stated that there were no instances of catastrophic failures associated with canopy seal leakage in any domestic or foreign plants or any instances of applied weld overlays failing after being placed in service. Industry experience has attributed degradation of the existing seal welds to transgranular stress corrosion cracking.

The canopy seal weld and associated overlay weld is separate from the structural pressure retaining threaded segments of the CRDM. The threaded segments are not designed to be leak tight and, therefore, canopy seal welds are used as part of the plant design to minimize leakage. The proposed overlay simply replaces a leaking section of the canopy seal weld and, because it does not affect the threaded CRDM joint, does not change the probability of a CRDM failure. The NRC staff verified that the canopy seal weld does not provide structural and pressure boundary support of the CRDM housing based on the design requirements of ASME Section Ill. The pressure boundary of the CRDM housing is supported by threaded joints between the cap and rod travel housing, rod travel housing and latch housing, and latch housing and head adapter as the licensee specified.

At present, the licensee does not have a weld design because it has not identified any degraded canopy seal weld. The licensee stated that the final weld design will be based on the applicable portions of ASME Code Case N-504-4. The NRC staff finds that adherence to the requirements of ASME Code Case N-504-4 ensures an adequate overlay thickness to support leak tight integrity of the repaired seal weld. The NRC conditionally approved ASME Code Case N-504-4 in Regulatory Guide (RG) 1.147, Revision 16. None of the conditions identified by the NRC staff in RG 1.147 apply to the proposed repair because the subject weld does not support the structural integrity of the canopy seal. As stated above, the licensee will use nickel-based Alloy 52/52M weld metal to deposit on the degraded seal weld. The NRC staff determined that based on laboratory tests and operating experience, Alloy 52/52M weld metal has shown sufficient resistance to primary SCC. This weld metal has been used in many weld overlay applications in pressurized water reactor plants. Therefore, the NRC staff finds that Alloy 52/52M will provide reasonable assurance of the leak tight integrity of the repaired seal weld. The licensee explained that weld overlay repairs would be performed using an appropriate welding procedure specification along with welders or welding operators qualified in accordance with the ASME Code,Section IX, "Qualification Standard for Welding and Brazing Procedures, Welders, Brazers, and Welding and Brazing Operators." In addition, the licensee stated that it will follow a repair plan consistent with ASME Code,Section XI, IWA-4150, "Repair/ Replacement Program and Plan." The licensee noted that because the existing base aterials are relatively thin, special consideration to travel speed and heat input would be critical parameters to be addressed during the weld qualification process. Based on a review of industry experience, the licensee stated that most of the difficulties attributed to Alloy 52/52M overlay repairs have been associated with overlays onto existing Alloy 82/182 dissimilar metal welds on nozzle-to-safe end configurations or embedded flaw repairs on pressurized water reactor upper head penetrations (i.e., overhead welding).

All base materials associated with the canopy seal welds are stainless steel. The licensee noted that overhead welding is not applicable for the proposed repair and, therefore, is not a concern as the welds at the lower and upper canopy seals are oriented horizontally as shown in the 2G position in Figure QW-461.4(b), "Groove Welds in Pipe-Test Position," of the ASME Code,Section IX. The weld at the middle canopy seal is also oriented horizontally as shown in the 1 G position in Figure QW-461.3(a) of the ASME Code,Section IX. The NRC staff notes that the overhead welding tends to cause fabrication defects as compared to the horizontal welding. The licensee stated that it will evaluate specific weld repairs and determine whether mock-ups are necessary to ensure a sound weld overlay is successfully applied. Based on the above, the NRC staff finds that the licensee will follow the ASME Code,Section IX, for the welding,Section XI, IWA-4150, for the repair activities, and applicable sections of Code Case N-504-4 for the weld design. The licensee will consider appropriate travel speed and heat input to minimize fabrication defects in the repaired weld. Therefore, the NRC staff finds that the welding for the proposed repair is acceptable.

With regard to lSI, the licensee stated that there is no applicable Category or Item number associated with canopy seal welds in the ASME Code,Section XI, Table IWB-2500-1, "Examination Categories," and, therefore, canopy seal welds are not subject to the periodic surface or volumetric examinations.

Also, replacement seal welds are specifically exempted from post-welding pressure testing per ASME Code,Section XI, IWA-4540, "Pressure Testing of Classes 1, 2, and 3 Items," Paragraph (b)(8). However, the licensee noted that a final surface examination is required after the weld repair completion in accordance with the Code of Construction (i.e., ASME Code, Section Ill) and the original Design Specification.

The licensee stated that the canopy seal welds are part of the Class 1 system leakage test boundary and are required to be tested for leakage during each refueling outage as specified in the ASME Code,Section XI, Table IWB-2500-1, Category B-P, "All Pressure Retaining Components," Item No. B15.1 0, "Pressure Retaining Components." As such, the licensee will perform the system leakage test per Table IWB-2500-1 in lieu of the hydrostatic test required in paragraph (h) of ASME Code Case N-504-4. Based on the pressure testing discussion, above, the NRC staff finds that the hydrostatic test is not needed for the proposed repair and that a system leakage test is sufficient to demonstrate the leak tightness of the repaired weld. The licensee stated that depending on the location of a particular canopy seal weld, the accessibility necessary to perform the required surface (PT) examination in accordance with the Code of Construction after the weld repair may not be possible.

In lieu of the surface examination after the weld repair, the licensee requested to use a VT-1 examination.

The licensee explained that conduct of the VT-1 examination would be dependent on accessibility of the inspection.

The licensee stated that if the repair was performed on a canopy seal that is located on the outer periphery of the core, it may be possible to inspect the repaired weld through direct visual observation and a mirror. If inspection access is limited, the licensee stated that a VT-1 examination would be performed using remote visual equipment such as a video probe or camera equipment that accompanies the welding equipment.

The licensee stated the remote visual equipment resolution would be demonstrated prior to, and upon completion of, the examination(s) in accordance with ASME Code, Section XIIWA-2216, "Remote Visual Examination," and ASME Code,Section V, "Nondestructive Examination," Article 9, "Visual Examination." In its September 19, 2013, letter, the licensee stated that the VT-1 visual examination will be demonstrated to resolve a 0.001-inch thick wire against the surface of the weld. The NRC staff determined that the qualification ofVT-1 examination based on 0.001-inch thick wire is not acceptable.

The NRC staff determined that the ASME Code,Section XI, Table IWA-221 0-1, "Visual Examinations," provides acceptable requirements for qualifying VT-1 examination based on character resolution.

In its letter dated January 24, 2014, the licensee stated that it will use the requirements of the ASME Code, SE;lction XI, Table IWA-2210-1, for procedure demonstrations required to support the VT-1 examinations.

The NRC staff finds that the proposed VT-1 examination is acceptable because it follows the ASME Code,Section XI, Table IWA-221 0-1. As part of the periodic system leakage test, the licensee is required to examine the canopy seal using the VT-2 visual examination method in accordance with Table IWB-2500-1.

The licensee stated that the CRDM housings are not insulated and are oriented vertically.

Accessibility to conduct a VT-2 examination of the lower canopy seal welds is relatively unobstructed once the shroud access doors are opened. Staining, due to reactor coolant system leakage, would be evident at low points in this area. According to the licensee, the intermediate and upper canopy seals cannot be observed directly and are examined to the extent practical in accordance with ASME Code,Section XI, IWA-5241, "Insulated and Noninsulated Components," Paragraph (d). In addition, the licensee verifies leakage in the beginning of each refueling outage and during forced outages in accordance with NRC Generic Letter (GL) 88-05, "Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary Components in PWR [pressurized-water reactor] Plants." The NRC staff finds that the VT-1 examination of the repaired canopy seal weld is not as robust as a surface examination in terms of detecting small surface-connecting fabrication defects. However, the licensee will perform a VT-2 examination as part of system leakage test and leakage verification per GL 88-05 during each refueling outage. The NRC staff determined these frequent visual examinations will provide reasonable assurance of the leak tightness of the repaired seal weld. As stated above, the licensee will perform an analysis of the repaired weldment before entering MODE 4, using Paragraph (g) of ASME Code Case N-504-4, to assure that the remaining flaw will not propagate unacceptably.

The NRC staff determined that ASME Code Case N-504-4(g) requires the licensee to perform a flaw growth evaluation in accordance with the ASME Code,Section XI, IWB-3640.

The NRC staff finds that it is acceptable that the licensee will perform the flaw growth evaluation in accordance with ASME Code Case N-504-4(g) because, besides periodic visual inspections, the flaw growth evaluation will provide additional assurance of the leak tightness of the repair weld. The licensee stated that there were no instances of significant degradation associated with any surfaces impacted by identified canopy seal leakage. Should a minor leak develop in a canopy seal weld develop during reactor startup, there are no components in the near vicinity of the CRDMs that would be adversely affected.

The licensee stated that if a canopy seal leaked, the leak could be detected by the containment area or process radiation monitoring system; or by the containment sump monitor. In addition, a reactor coolant system leak rate surveillance is conducted every shift that would identify leakage. The NRC staff determined that potential leakage containing boric acid could reach the carbon steel reactor vessel head and cause corrosion of the carbon steel. However, based on operating experience, there has not been instances of significant reactor vessel head corrosion caused by the canopy seal leakage. Based on the above, discussion of leak experience and leak identification methods, the NRC staff finds that leakage from a through wall flaw at the repaired canopy seal weld, if it occurs, will not significantly affect the structural integrity of the reactor vessel head and CRDM housing. The NRC staff determined that the overall design, operation, and inspections at the plant demonstrates defense-in-depth should a leak occur. Based on the information provided in the September 13, 2013, submittal, the NRC staff determined that the canopy seal location has a relatively high radiological dose. The licensee estimates a dose in excess of 0.600 person-Rem for a single CRDM seal weld repair, if the flaw is excavated and a dye penetrant examination is performed.

The NRC staff finds that performing the ASME Code repair of a degraded canopy seal weld in a high radiological dose area with limited accessibility is contrary to the philosophy of maintaining radiological dose exposure to workers ALARA. In addition, the canopy seal weld is not designed to provide structural integrity to the CRDM housing. Therefore, the NRC staff finds that complying with the specified ASME Code requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. In summary, the NRC staff finds that the proposed alternative canopy seal weld repair will provide reasonable assurance of leak tightness of the canopy seal because the proposed repair follows the welding requirements of the ASME Code,Section IX, the examination requirements of the ASME Code,Section XI, and the design requirements of ASME Code Case N-504-4.

4.0 CONCLUSION

As set forth above, the NRC staff finds that the proposed alternative provides reasonable assurance of leak tightness of the canopy seal welds. The NRC staff finds that complying with the specified ASME Code requirements would result in hardship or unusual difficulty without a compensating increase in the 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)(ii) and is in compliance with the requirements of the ASME Code,Section XI, for which relief was not requested.

Therefore, the NRC staff authorizes relief for 13R-11 for the remainder of the third 1 0-year lSI interval at Braidwood Station, Units 1 and 2, which is currently scheduled to end on July 28, 2018, for Unit 1 and October 16, 2018, for Unit 2. All other requirements of the ASME Code,Section XI, for which relief has not been specifically requested, remain applicable, including a third party review by the Authorized Nuclear lnservice Inspector.

Principal Contributor:

John Tsao Date of issuance:

April 28, 2014 J. Pacilio If you have any questions, please contact the Senior Project Manager, Joel S. Wiebe, at 301-415-6606 or via e-mail at Joei.Wiebe@nrc.gov.

Docket Nos. 50-456 and 50-457

Enclosure:

Safety Evaluation cc w/encl: Distribution via ListServ DISTRIBUTION:

PUBLIC RidsNrrPMBraidwood Resource RidsAcrsAcnw_MaiiCTR Resource LPL3-2 R/F ADAMS Accession No. ML 14084A549 OFFICE LPL3-2/PM LPL3-2/LA NAME JWiebe SRohrer DATE 3/26/14 4/28/14 Sincerely, /RAJ Travis Tate, Chief Plant Licensing 111-2 and Planning and Analysis Branch Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation RidsNrrDorllpl3-2 Resource RidsNrrLASRohrer Resource RidsRgn3MaiiCenter Resource RidsNrrDoriDpr Resource *via e-mail DE/ENPB/BC*

LPL3-2/BC Tlupold TTate 3/11/14 4/28/14 OFFICIAL RECORD COPY