Text

Relief from the Requirements of the ASME Code(Cac Nos. MF4809, MF4810, MF4811, and MF4812)
	ML16007A185
Person / Time
Site:	Byron, Braidwood
Issue date:	01/21/2016
From:	Justin Poole; Plant Licensing Branch III
To:	Bryan Hanson; Exelon Generation Co, Exelon Nuclear
	Wiebe J, NRR/DORL 415-6606
References
	CAC MF4809, CAC MF4810, CAC MF4811, CAC MF4812
	Download: ML16007A185 (8)
	v • d • e

Mr. Bryan C. Hanson Senior Vice President UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 January 21,2016 Exelon Generation Company, LLC President and Chief Nuclear Officer (CNO)

Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

BYRON STATION, UNIT NOS. 1AND2, AND BRAIDWOOD STATION, UNITS 1 AND 2 - RELIEF FROM THE REQUIREMENTS OF THE ASME CODE (CAC NOS. MF4809, MF4810, MF4811, AND MF4812)

Dear Mr. Hanson:

By letter dated September 8, 2014 (Agencywide Documents and Management System (ADAMS) Accession No. ML14251A536), as supplemented by letters dated May 29, 2015 (ML15152A254), and November 5, 2015(ML15309A227).

Exelon Generation Company, LLC (the licensee) requested relief from the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) penetrant testing examinations on repair weld surfaces for reactor vessel head penetrations at Braidwood, Units 1 and 2, and Byron, Unit Nos. 1 and 2.

Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(a)(3)(i)

(retitled paragraph 50.55a(z)(1) (79 FR 65776, dated November 5, 2014)), the licensee requested to use the proposed alternative on the basis that the alternative examination provides an acceptable 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 the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(z)(1 ).

Therefore, the NRC staff authorizes the proposed alternative for the remainder of the third ISi interval at Byron, Unit Nos. 1 and 2, and Braidwood, Units 1 and 2.

Docket Nos. STN 50-456, STN 50-457, STN 50-454 and STN 50-455

Enclosure:

Safety Evaluation cc w/encl: Distribution via Listserv Sincerely, -;

.~-*1

~---~~--

Justin C. Poole, Acting Branch Chief Plant Licensing Branch 111-2 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 NOS. 13R-09 AND 13R-20, REVISION 2 REGARDING REACTOR VESSEL HEAD PENETRATIONS

1.0 INTRODUCTION

EXELON GENERATION COMPANY, LLC BYRON STATION, UNIT NOS. 1AND2, AND BRAIDWOOD STATION, UNITS 1 AND 2 DOCKET NOS. STN 50-454, STN 50-455, STN 50-456 AND STN 50-457 By letter dated September 6, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML14251A536), and as supplemented by letters dated May 29, 2015 (ADAMS Accession No. ML15152A254), and November 5, 2015 (ADAMS Accession No. ML15309A227), Exelon Generation Company, LLC (the licensee) requested relief from the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) penetrant testing examinations on repair weld surfaces for reactor vessel head penetrations at Braidwood, Units 1 and 2, and Byron, Unit Nos. 1 and 2.

Specifically, pursuant to Title 1 O of the Code of Federal Regulations (10 CFR) 50.55a(z)(1 ), the licensee requested to use the proposed alternative on the basis that the alternative examination provides an acceptable level of quality and safety.

2.0 REGULATORY EVALUATION

The licensee proposes to use alternatives to the requirements of ASME Code,Section XI, IWA-5220.

10 CFR 50.55a(g)(4) states, in part, that ASME Code class 1, 2, and 3 components (including supports) must meet the requirements, except the design and assess provisions and the pre-servicexamination requirements, set forth in the ASME Code,Section XI, "Rules for In-service Inspection (ISi) of Nuclear Power Plant Components," to the extent practical within the limitations of design, geometry, and materials of construction of the components.

Enclosure 10 CFR 50.55a(z), states, in part, that alternatives to the ASME Code requirements may be authorized by the U. S. Nuclear Regulatory Commission (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.

Based on the above, and subject to the following technical evaluation, the NRC staff finds that the regulations permit the licensee to request the use of an alternative and that the NRC has the regulatory authority to authorize the alternative proposed by the licensee.

3.0 TECHNICAL EVALUATION

3.1

Applicable Code Edition and Addenda

The ISi and Repair/Replacement Programs: ASME Code,Section XI, 2001 Edition through 2003 Addenda. Examinations of the vessel head penetrations (VHPs) are performed in accordance with 10 CFR 50.55a(g)(6)(ii)(D), which specifies the use of Code Case N-729-1, with conditions.

Code of Construction [Reactor Pressure Vessel (RPV)]: ASME Section Ill, 1971 Edition through Summer 1973Addenda.

3.2 Components for Which Relief is Requested Braidwood and Byron Stations, Units 1 and 2, reactor VHPs and J-groove welds.

3.3 Duration of Relief The duration of the proposed alternative is for the remainder of the Byron, Unit Nos. 1 and 2, third ISi interval currently scheduled to end in July 15, 2016.

The duration of the proposed alternative is for the remainder of the Braidwood, Units 1 and 2, third ISi interval currently scheduled to end in July 28, 2018, and October 16, 2018, respectively.

3.4 Proposed Alternative Licensee proposes a reinspection frequency of every other cycle when the surface examination results of the embedded flaw repair (EFR) are verified to be acceptable for two consecutive cycles after the original installation or repair of the EFR.

3. 5 Basis for Use In 2011, the licensee proposed an alternative repair technique using weld overlays on the reactor vessel head penetration housing and J-groove welds, using a Westinghouse EFR. In 2012 (ADAMS Accession No. ML120790647), the NRC provided its authorization to implement relief requests 13R-09 and 13R-20, Revision 1, as a repair method for degradation identified in reactor VHPs.

The NRC approved reactor VHPs repair weld examinations methods including surface examinations (i.e., dye penetrant test (PT)) in accordance with NB-2545 or NB-2546. The licensee has reviewed the technical basis and previous examination results for requiring PT examinations of applied weld material each outage and determined that it is appropriate to relax the required inservice PT examination frequency. The license determined that personnel radiation exposure associated with examinations would be reduced. The November 5, 2015, submittal proposes relaxation of PT examination frequency for installed repairs that have demonstrated continued satisfactory PT examination results. The proposed changes are to Section 5.1.2(g), Note (4), and to Section 5.1.3 of 13R-09 and 13R-20, Revision 1.

Westinghouse Report L TR-PSDRTAM-14-005, Revision 3 (Reference 13; provided in of the submittal) provides the technical bases for reducing surface examination requirements for J-groove weld repairs. This technical justification includes a detailed review of PT examination history, review of potential causes of PT indications in EFRs, and the use of crack resistant alloys in the EFR. The EFR is a robust design that is resistant to primary water stress-corrosion cracking (PWSCC). EFR installation, examination, and operational history indicate that the EFR performs acceptably.

Examination and removed sample history indicate that the flaws identified shortly after installation of EFR weld material were due to embedded weld discontinuities and not due to service induced degradation. With inspection of the EFR every other cycle of operation, the nozzles are adequately monitored for degradation by ultrasonic test (UT) examination methods similar to the nozzles without EFR repairs. The licensee projects that the reduction of the PT examination of nozzles would result in a dose savings of approximately 0.4 to 0.7 roentgen equivalent man (rem) per nozzle examination. The licensee states that the proposed changes to the inservice examination requirements assure that the EFR repaired nozzles are adequately monitored through a combination of volumetric and surface examinations throughout the life of the installation, thus ensuring the EFR repaired nozzles will continue to perform their required function.

4.0 NRC STAFF EVALUATION The EFR is a technique where at least two layers of Alloy 52 weld metal are deposited to isolate existing flaws and susceptible material from the primary water environment. Alloy 52 weld metal is highly resistant to PWSCC. A new PWSCC crack is highly unlikely to initiate and grow through the Alloy 52 overlay.

The resistance of Alloy 690 and its associated weld metals, Alloys 52 and 152, has been demonstrated by laboratory testing in which crack initiation and growth at rates of significance has been observed to be unlikely to occur in simulated pressurized-water reactor environments.

The EFR for reactor VHPs was first applied at the DC Cook plant in 1996, and since that time it has been implemented on over 45 reactor VHPs world-wide. As an additional requirement of the original repair and each subsequent repair, a liquid penetrant examination was to be performed on the EFR weld each refueling outage following the EFR. This requirement was put in place when the process was new, however, due to the dose required for this examination, the NRC has decided to revisit the frequency of the examination as requested in the licensee's submittal.

The licensee has requested relaxation of the current PT requirement of every outage to a reinspection frequency of every other cycle, after the surface examination results of the EFR are verified to be acceptable for two consecutive cycles, after the original installation or repair of the EFR. This is an extension of an additional 1.5 to 2 years between inspections. The NRC considered the following points when reviewing this request:

1. Alloy 52/Alloy 152 applied weld material is highly resistant to PWSCC, with no crack initiations in over 18 years of service history.

2. The initial and post installation PT's performed on the EFR weld established a clean baseline of the weld. No evidence of PWSCC of these repairs has been found to date.

3. Follow-up PT examinations have not revealed any service-induced cracking or structural degradation. All indications found to date in EFR after installation have been shown to be fabrication defects and not PWSCCs.

4. There is significant radiation dose associated with the penetrant examinations.

5. Continued UT examination each outage of the repaired nozzle provides reasonable assurance that any degradation of the EFR and nozzle material would be found.

The service history of the PT examinations shows that out of 7 4 PT examinations, 68 have been acceptable and six have identified rejectable PT indications. As indicated above in bullet points 2 and 3, the six penetrations with rejectable PT indications represent a relatively small portion of installed EFR welds and each of these six penetrations has subsequently been repaired and found acceptable. In each case, the EFR was restored to a fully acceptable condition by minor buffing/grinding and, where needed, limited manual repair welding. In no case was PWSCCs identified in these welds.

Based on the lack of PWSCC flaws in EFRs; the excellent service history of Alloy 690/52/152 in other applications; the continued performance of UT examinations by the licensee during each outage, and the high amount of radiological dosage involved with performing the PT exams, the NRC staff concludes that, in this instance, the performance of PT exams every outage is not necessary to ensure the structural and leak tight integrity of the subject welds.

Additionally, there is reasonable evidence that relaxing the current PT requirement of every outage to a reinspection frequency of every other cycle only after the surface examination results of the EFR are verified to be acceptable for two consecutive cycles after the original installation or repair of the EFR will continue to provide a sufficient level of quality and safety in the upper head EFR welds. As indicated in the licensee's proposed alternative, if any indications are found during subsequent UT or PT examinations that require a repair, the licensee will be required to return to the inspection schedule of performing a PT examination each refueling outage until two consecutive acceptable cycles are obtained.

Based on the above analysis, the NRC staff finds that the technical requirements of 10 CFR 50.55a(z)(1) have been met and, therefore, that the licensee's proposal provides an acceptable level of quality and safety. The NRC staff, therefore, finds no technical basis that would preclude the authorization of the alternative to the current PT examination requirement, as requested by the licensee.

5.0 CONCLUSION

S As set forth above, the NRC staff determines that the proposed alternative 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(z)(1 ).

Therefore, the NRC staff authorizes the proposed alternative for the remainder of the third ISi interval at Byron Station, Units 1 and 2 and Braidwood Units 1 and 2.

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

Reviewer: M Audrain

ML16007A185 OFFICE LPL3-2/PM LPL3-2/LA NAME JWiebe SRohrer DATE 1/21/16 1/12/16 Sincerely, IRA/