ML18347B419

Relief Request I4R-03, Relief from ASME Requirements for the Fourth 10-Year Inservice Inspection Interval Related to Degraded Canopy Seal Welds Associated with Control Rod Drive Mechanism

ML18347B419
Person / Time
Site:	Braidwood
Issue date:	01/17/2019
From:	David Wrona Plant Licensing Branch III
To:	Bryan Hanson Exelon Generation Co
Wiebe J
References
EPID L-2018-LLR-0033
Download: ML18347B419 (9)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O.C. 20555-0001 Janu1ry 17, 2019 Mr. Bryan C. Hanson Senior Vice President Exelon Generation Company, LLC President and Chief Nuclear Officer (CNO)

Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

BRAIDWOOD STATION, UNITS 1 AND 2 - RELIEF FROM THE REQUIREMENTS OF THE AMERICAN SOCIETY OF MECHANICAL ENGINEERS BOILER AND PRESSURE VESSEL CODE (EPID L-2018-LLR-0033)

Dear Mr. Hanson:

By letter dated March 19, 2018 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML18078A185), as supplemented by letters dated May 3, 2018 and July 31, 2018 (ADAMS Accession Nos. ML18124A154 and ML18212A352, respectively),

Exelon Generation Company, LLC (Exelon, the licensee), submitted relief request (RR) 14R-03 for the fourth 10-year inservice inspection (ISi) interval at Braidwood Station (Braidwood),

Units 1 and 2. The licensee requested relief from Article 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) at Braidwood, Units 1 and 2.

Specifically, pursuant to Title 1O of the Code of Federal Regulations (1 O CFR) 50.55a(z)(2), the licensee requested to use the alternative 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.

The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the subject request and concludes, as set forth in the enclosed safety evaluation, that Exelon has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(2). Therefore, the staff authorizes the use of Relief Request 14R-03 for the fourth 10-year ISi intervals at Braidwood, Units 1 and 2, which started on August 29, 2018, and on November 5, 2018, respectively. By its letter dated March 19, 2018, the licensee indicated that the fourth 10-year ISi interval for Unit 2 was scheduled to start on October 17, 2018, but also indicated that the start was subject to the allowable changes for inspection intervals in paragraph IWA-2430. By letter dated October 11, 2018 (ADAMS Accession No. ML18284A445), the licensee indicated that the fourth 10-year inspection interval for Unit 2 was scheduled to start on November 5, 2018.

All other requirements of ASME Code,Section XI, for which relief was not specifically requested and authorized by the NRC staff remain applicable, including the third-party review by the Authorized Nuclear lnservice Inspector.

B. Hanson If you have any questions, please contact the Project Manager, Joel Wiebe, at 301-415-6606 or via e-mail at Joel.Wiebe@nrc.gov.

Sincerely,

~ 2 Qi/________

David J. Wro a, Chief Plant Licensing Branch Ill Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-456 and 50-457

Enclosure:

Safety Evaluation cc: Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF REQUEST 14R-03 REGARDING REPAIR OF DEGRADED CANOPY SEAL WELDS ASSOCIATED WITH THE CONTROL ROD DRIVE MECHANISM EXELON GENERATION COMPANY, LLC BRAIDWOOD STATION, UNITS 1 AND 2 DOCKET NOS. 50-456 AND 50-457

1.0 INTRODUCTION

By letter dated March 19, 2018 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML18078A185), as supplemented by letters dated May 3, 2018, and July 31, 2018(ADAMSAccession Nos. ML18124A154and ML18212A352, respectively),

Exelon Generation Company, LLC (the licensee), requested relief from Article 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 (CROM) at Braidwood Station, Units 1 and 2.

The licensee submitted the proposed alternative, Request Number 14R-03, Revision 1, pursuant to Title 10 of the Code of Federal Regulations ( 10 CFR) Part 50, Section 50.55a(z)(2) on the basis that the ASME Code requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality or safety.

2.0 REGULATORY EVALUATION

The U.S. Nuclear Regulatory Commission (NRC) staff considered the following regulatory requirements and guidance in its evaluation.

Adherence to Section XI of the ASME Code is mandated by 10 CFR 50.55a(g)(4), which states, in part, that ASME Code Class 1, 2, and 3 components (including supports) will meet the requirements, except the design and access provisions and the pre-service 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.

The regulation in 10 CFR 50.55a(z) states, in part, that alternatives to the requirements of paragraph (g) of 10 CFR 50.55a may be used, when authorized by the NRC, if the licensee demonstrates that ( 1) the proposed alternative provides an acceptable level of quality and safety, or (2) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

Enclosure

Based on the above, and subject to the following technical evaluation, the NRC staff finds that regulatory authority exists for the licensee to request and the NRC to authorize the alternative requested by the licensee.

3.0 TECHNICAL EVALUATION

3.1 Licensee's Alternative 3.1.1 ASME Code Components Affected The affected components are ASME Code Class 1 canopy seal welds associated with the CRDMs.

3.1.2 Applicable Code Editions and Addenda The ISi program for the fourth 10-year ISi interval is based on the ASME Code,Section XI, 2013 Edition. The applicable code of construction is the ASME Code, Section Ill, 1974 Edition through Summer 1974 Addenda.

3.1.3 Applicable Code Requirement The ASME Code,Section XI, Article 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.

3.1.4 Reason for Request The licensee stated that at present there are no known degraded canopy seal welds at Braidwood Station, Units 1 and 2. The proposed alternative is associated with a contingency repair of leaking canopy seal welds.

In order to repair a leaking CRDM canopy seal weld in accordance with the code of construction (ASME Code, Section Ill), a canopy seal weld repair would require excavation of indications followed by a surface examination to ensure that the rejectable indications have been removed.

After re-welding and restoration to the original configuration and materials, a surface examination is required. The licensee stated the excavation of the leaking portion of a seal weld would necessitate a cavity that extends completely through wall. A liquid penetrant testing (PT) surface examination of the excavated area, as required by code, would result in the deposit of penetrant materials onto the inner surfaces of the original seal weld which could not be removed prior to re-welding. The remaining penetrant material would introduce contaminants into the new weld metal and reduce the quality of the repair weld. 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. Although most of the repair activities associated with an ASME Code compliant repair would be performed remotely using robotic equipment, the ASME Code required excavation and PT examinations would necessitate hands-on access to the canopy seal weld. The licensee estimated the total dose for these activities would be in excess of 0.600 person-Rem per one weld repair.

The licensee requested the proposed alternative, 14R-03, Revision 1, on the basis that compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

3.1.5 Proposed Alternative and Basis for Use In lieu of repairing leaking CRDM canopy seal welds in accordance with ASME Code,Section XI, Article IWA-4000, the licensee proposed to leave flaws in place and apply a weld overlay over the existing seal weld followed by a VT-1 visual examination. A post-maintenance VT-2 visual examination would be performed at normal operating temperature during the system leakage test. The licensee stated that it will perform an analysis of the repaired weldment, prior to Mode 4, to assure that the remaining flaw will not propagate unexpectedly. The design of the weld overlay will follow the applicable portions of ASME Code,Section XI, Nonmandatory Appendix Q, "Weld Overlay Repair of Classes 1, 2 and 3 Austenitic Stainless Steel Piping."

The licensee stated that the proposed alternative uses the gas tungsten arc welding (GTAW) process controlled remotely; however, a manual GTAW repair may be utilized. All base materials associated with the canopy seal welds are stainless steel. The proposed weld overlay weld metal is Alloy 52/52M in lieu of the canopy seal weld filler material which is stainless steel.

The weld overlay weld material specified in Appendix Q is also stainless steel.

Subarticle IWA-4200 requires that the repair material conform to the original design specification or ASME Code, Section Ill.

The licensee stated that components associated with the canopy seal welds were assembled using threaded connections and seal welded using consumable inserts. The threaded joints between the cap and rod travel housing, rod travel housing and latch housing, and latch housing and head adaptor provide the primary pressure boundary and structural support for the CRDMs.

The canopy seal weld is for secondary pressure retention only and does not provide any structural support. The licensee provided the general configuration of the canopy seal welds along with their general location on the CRDM assembly in Figures 14R-03-1 and 14R-03-2 in its May 3, 2018 letter. The approximate diameter of the canopy seal welds ranges from 3.75" (for the upper canopy seal weld) to 6.45" (for the lower canopy seal weld).

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 stress-corrosion cracking (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 selected nickel-based Alloy 52/52M, in lieu of the original stainless steel seal weld filler material, because of its resistance to SCC. In addition, the licensee stated that the suitability of the replacement material will be evaluated for each application and determined to be compatible with the existing component and will provide a leakage barrier for the remainder of the intended life of the CRDM.

3.1.6 Duration of Proposed Alternative The licensee requested the alternative to be approved for the remainder of the fourth 10-year ISi interval for Braidwood Station, Units 1 and 2, which is currently scheduled to end on July 28, 2028, for Unit 1, and October 16, 2028, for Unit 2.

3.2 NRC Staff Evaluation In 2013, the licensee requested an alternative, Relief Request 13R-11, to repair CROM canopy seal welds using a weld overlay for Braidwood Units 1 and 2 for the third 10-year ISi interval (ADAMS Accession Nos. ML13263A372 and ML14024A588). The alternative was approved by the NRC on April 28, 2014 (ADAMS Accession Nos. ML14084A549). The current fourth interval request is similar to the previous interval request. There are currently no known degraded canopy seal welds at Braidwood Station, Units 1 and 2. The licensee's request is intended as a contingency in the event that leaking canopy seals are detected during the fourth interval.

There is a history of degradation resulting in leaking of CROM canopy seal welds in several domestic and foreign nuclear plants.

Article IWA-4000 of the ASME Code,Section XI, 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 applicable code of construction is the ASME Code, Section Ill, 1974 Edition through Summer 1974 Addenda. The NRC staff notes that in order to repair a leaking canopy seal weld in accordance with ASME Code, Section Ill, a canopy seal weld repair would require excavation of indications followed by a surface examination to ensure that the rejectable indications have been removed. A purge of the cavity area during welding would be required. After welding and restoration to the original configuration using the original welding filler material, a surface examination would be required to be performed.

The NRC staff recognizes that excavation of the flaw would create a sizable-opening in the weld from the exterior surface of the canopy seal weld to the cavity between the threaded connection and the back side of the canopy seal weld (open-root weld joint). Performing a code-required surface examination prior to welding would introduce contaminants into this cavity area that could not be removed prior to welding. In addition, boric acid and other contaminants in the cavity resulting from the leak would not be able to be cleaned out. Contaminants in the cavity would be drawn into the weld on the first weld pass and could result in welding defects. The contaminating material deposited in the first weld pass could be drawn into subsequent weld passes resulting in through-weld flaws. Due to the configuration of the canopy seals, it would not be possible to establish an adequate purge prior to welding. The lack of an adequate purge would result in severe oxidation of the cavity side of the first weld pass (root pass). Although the welding would mostly be performed using remotely operated GTAW equipment to reduce dose, excavation of the defect area, a surface examination of the excavated area and the final weld surface examination would result in a considerable amount of radiological dose. The licensee estimated the radiological dose to be 0.600 person-Rem per one weld repair.

In lieu of ASME Code-required weld repair, the licensee proposed to perform the repair of leaking CROM canopy seal welds by applying an Alloy 52 weld overlay using the applicable design provisions of ASME Code,Section XI, Appendix Q. The licensee's NRG-approved third interval alternative, Relief Request 13R-11, to perform CROM canopy seal weld overlays, was based on applicable portions of ASME Code Case N-504-4. Code Case N-504-4 has been incorporated into ASME Code,Section XI, Appendix Q. The NRC staff notes that several U.S.

and foreign nuclear plants have successfully applied weld overlays to leaking CROM canopy seal welds.

Appendix Q allows deposition of one or more layers of the weld overlay to seal unacceptable indications in the area to be repaired without excavation. Additional layers of weld metal provide a redundancy to prevent leakage. The NRC staff notes that a weld overlay has the backing of the existing weld and will provide a higher quality weld than an open-root weld joint, as discussed above, resulting from complete flaw removal in accordance with the construction code.

Appendix Q further requires a stress analysis of the repaired weldment to assure that the existing flaw will not propagate unacceptably for the design life of the repair. The analysis will consider potential flaw growth due to fatigue and SCC which is most likely the mechanism that initiated the flaw. The licensee will perform its analysis prior to entering Mode 4. The weld metal proposed by the licensee for its alternative repair, Alloy 52, is less susceptible to SCC than the original stainless steel seal weld filler metal. Alloy 52 has been used to perform hundreds of successful weld overlay repairs in piping welds and CROM welds. The NRC staff is not aware of any instances of SCC in Alloy 52 weld metal in U.S. or foreign nuclear plants. The seal weld itself is neither a structural weld nor a pressure-retaining weld (i.e., the weld does not support any loads and is not part of reactor coolant system pressure boundary). Based on the above, the NRC staff finds that the licensee's proposed alternative will not impact the structural integrity of the CROM. The NRC staff finds acceptable that the licensee will perform a flaw evaluation to ensure that the flaw will not propagate to unacceptable size and that the licensee will use Alloy 52 filler metal which is less susceptible to sec.

In lieu of a PT examination of the completed weld overlay, the licensee proposes to conduct a VT-1 visual examination through direct observation where possible. In locations where direct observation is not possible, the licensee will conduct VT-1 examinations using remote visual equipment. The qualification and performance of VT-1 examinations will meet ASME Code, Sections XI and V, "Nondestructive Examination." The NRC staff finds that any major surface-breaking defects in the weld overlay would most likely be identified with a Code-qualified VT-1 examination. The licensee stated that in the event that a minor canopy seal leakage develops during startup, there are no components in the near vicinity of the CROMs that would be adversely impacted. Leakage could be detected by the containment area or process radiation monitoring system; or by the containment sump monitor. Reactor coolant system leak rate surveillance conducted every shift would also identify leakage. The licensee stated that based on operating experience, there were no instances of significant degradation associated with any target surfaces as a result of identified canopy seal leakage. Based on the above, the NRC staff finds the licensee's use of a VT-1 in lieu of a PT examination acceptable.

A post-maintenance VT-2 will be performed at normal operating pressure and temperature during the system leakage test. The canopy seal weld overlays will also be subjected to a VT-2 examination at the beginning and end of each refueling outage as well during forced outages.

These VT-2 examinations provide additional assurance that should leakage occur, it will be identified in a timely manner before significant degradation can occur. The NRC staff finds acceptable that the licensee will perform required system leakage test and associated VT-2 visual examination to monitor the leak tightness of the canopy seals.

Based on the information provided in the licensee's May 3, 2018, letter, 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 CROM 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, when an acceptable alternative is available, is contrary to the philosophy of maintaining radiological dose exposure to a worker's ALARA. In addition, the canopy seal weld is not designed to provide structural integrity to the CROM 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 XI, the design requirements of ASME Code,Section XI, Appendix Q and the system leakage testing requirements of the ASME Code,Section XI.

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(z){2).

Therefore, the NRC staff authorizes the use of Request Number 14R-03, Revision 1, for the fourth 10-year ISi interval at Braidwood Station, Units 1 and 2, which is currently scheduled to end on July 28, 2028, for Unit 1, and October 16, 2028, 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: R. Davis Date of issuance: January 1 7, 201 9

ML183478419 *via email OFFICE NRR/DORL/LPL3/PM NRR/DORL/LPL3/LA NRR/DMLR/MPHB/BC NAME JWiebe SRohrer (JBurkhardt for) SRuffin*

DATE 1/2/19 12/21/18 11/21/18 OFFICE NRR/DORL/LPL3/BC NAME DWrona DATE 1/17/19