ML13346A040
| ML13346A040 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 12/18/2013 |
| From: | Robert Carlson Plant Licensing Branch III |
| To: | Meyer L Point Beach |
| Beltz T | |
| References | |
| TAC MF1146 | |
| Download: ML13346A040 (7) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Mr. Larry Meyer Site Vice President NextEra Energy Point Beach, LLC Point Beach Nuclear Plant 6610 Nuclear Road Two Rivers, WI 54241 December 18, 2013
SUBJECT:
POINT BEACH NUCLEAR PLANT, UNIT 2-RELIEF FROM THE REQUIREMENTS OF THE AMERICAN SOCIETY OF MECHANICAL ENGINEERS (ASME) BOILER AND PRESSURE VESSEL CODE (CODE),
SECTION XI, FOR THE FOURTH 10-YEAR INSERVICE INSPECTION INTERVAL (TAC NOS. MF1146)
Dear Mr. Meyer:
By letter dated March 19, 2013, as supplemented by letter dated August 9, 2013, NextEra Energy Point Beach, LLC (NextEra, the licensee), submitted to the U.S. Nuclear Regulatory Commission (NRC) a Request for Relief (RR-4L2) from certain inspection coverage requirements of the American Society of Mechanical Engineers (ASME) Boiler arid Pressure Vessel Code (Code),Section XI, for steam generator nozzle inner radius examinations at the Point Beach Nuclear Plant (PBNP), Unit 2.
Specifically, pursuant to Title 10 of the Code of Federal Regulations (1 0 yFR), Section 50.55a(g)(6)(i), the licensee requested relief for in-service testing items on the basis that the code requirement is impractical. Relief is being requested because the geometry of the Unit 2 steam generator main steam outlet nozzle design does not result in a true inner radius with the bored flow restrictor holes, such that no meaningful examination could be performed.
The NRC staff has reviewed the request* and concludes, as set forth in the enclosed safety evaluation, that NextEra has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(g)(6)(i) based on the impracticality to comply with the ASME Code examination coverage requirements for the subject welds listed in RR-4L2. The staff has determined that to meet the Code requirements, design modifications would be necessary to provide access for examination, and that imposition of the ASME Code requirements would result in undue burden.
The NRC staff concludes that visual examinations conducted during system pressure tests will continue to provide reasonable assurance of structural integrity of the steam generator main steam outlet nozzles.
If you have any questions regarding this matter, please contact Terry Beltz at (301) 415-3049, or via e-mail at Terrv.Beltz@nrc.gov.
Docket No. 50-301
Enclosure:
Safety Evaluation cc w/encl: Distribution via ListServ Sincerely, Robert D. Carlson, Chief Plant Licensing Branch 111-1 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 REGARDING RELIEF REQUEST RR-4L2 FOR THE FOURTH 10-YEAR INSERVICE INSPECTION INTERVAL
1.0 INTRODUCTION
NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNIT 2 DOCKET NO. 50-301 TAC NO. MF1146 By letter dated March 19, 2013 (Agencywide Document Access and Management System (ADAMS) Accession Number ML13079A141 ), as supplemented by letter dated August 9, 2013 (ADAMS Accession No. ML13221A525), Next Era Energy Point Beach, LLC (NextEra, the licensee), requested relief from inspection coverage requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code),Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components."
Specifically, pursuant to Title 10 of the Code of Federal Regulations ( 10 CFR}, Part 50, Section 50.55a(g)(6}{i}, the licensee submitted Relief Request (RR)-4L2 requesting relief for in-service testing items on the basis that the code requirement is impractical. The request is applicable to steam generator nozzle inner radius examinations at the Point Beach Nuclear Plant (PBNP),
Unit 2.
2.0 ' REGULATORY EVALUATION Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1, 2, and 3 components (including supports) shall meet the requirements, except the design and access provisions and the 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. The regulations require that inservice examination of components and system pressure tests conducted during the first 1 0-year interval and subsequent intervals comply with the requirements in the latest edition and addenda* of Section XI of the ASME Code, which was incorporated by reference in 10 CFR 50.55a(b) twelve months prior to the start of the 120-month interval, subject to the conditions listed therein.
Enclosure 10 CFR 50.55a(g)(5)(iii), states, in part, that licensees may determine tha,t conformance with certain code requirements is impractical, and shall notify the Commission and submit information in support of the determination.
10 CFR-50.55a(g)(6)(i), states that the Commission will evaluate determinations under paragraph (g)(5) of this section that code requirements are impractical. The Commission may grant such relief and may impose such alternative requirements as it determines is authorized by law and will not endanger life or property or the common defense and security and is otherwise in the public interest giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility.
The licensee has requested relief from the ASME Code,Section XI requirements pursuant to 10 CFR 50.55a(g)(6)(i). Based on the above, and subject to the following technical evaluation, the U.S. Nuclear Regulatory Commission (NRC) staff finds that regulatory authority exists for the licensee to request and the Commission to grant the relief requested by the licensee.
3.0 TECHNICAL EVALUATION
3.1 Licensee's Relief Request Component Descriptions Relief Request RR-4L2 includes steam generator main steam outlet nozzle inner radii at PBNP, Unit 2, w~ich have flow limiting devices installed in place of standard smooth transitions. The steam nozzle flow limiters consist of SA-508 Class 3A Steam Outlet Nozzle with seven venture inserts constructed from SB-564 (Alloy 690). A 0.22-inch thick weld overlay on Alloy 52 was deposited on the steam outlet nozzles. The venture inserts are welded into this overlay with Alloy 52.
Applicable ASME Code Requirement
- ASME Code Section XI Table IWC-2500-1 Category C-B, Item Number C2.22-Nozzle Inside Radius Section requires that the inner radius sections of all nozzles at terminal ends of piping runs be volumetrically examined. Note 4 of Table IWC-2500-1, Category C-B states, "in the case of multiple vessels of similar design, size, and service (such as steam generators, heat
. exchangers), the required examinations may be limited to one vessel or distributed among the vessels~"
The Code of Record for the fourth inspection interval at PBNP, Unit 2, is the ASME Code,Section XI, 1998 Edition through the 2000 Addenda. The fourth inspection interval at PBNP,
- Unit 2, *began on July 1, 2002, and en~ed on July 31, 2012.
Basis for Relief (as stated by the licensee)
The geometry of this nozzle design, with the bored flow restrictor holes, does not have a true inner radius; therefore, no meaningful examination can be performed. This design is not the same as a typical nozzle with a radius as described in [ASME Code Section XI] Figure IWC-2500-4, but instead has several "corner regions" which correspond to each bored hole. As a result, the Code requirement is not considered to be applicable to the design of the nozzle and compliance with the Code should not be required. Therefore, no alternate examinations of the main steam nozzle inner radius section are proposed.
The subject nozzles receive a visual (VT -2) examination each inspection period during the system leakage test as required by Section XI, Table IWC-2500-1, Examination Category C-H.
The required VT -2 examination performed each inspection period provides an acceptable level of quality and safety for the subject sections of the steam generators.
The licensee did not propose any alternative examinations for the subject welds.
3.2.1 NRC Staff Evaluation' The ASME Code requires.1 00-percent volumetric examination for the steam generator main steam outlet nozzle inner radius. The steam generator main steam outlet contains a flow limiter
- device within the bore of the nozzle..The flow limiter does not have the rounded inner radius or any region similar to the examination volume specified in Figure IWC-2500-4. Obtaining an inner-radius examination as per ASME Code requirements would require redesigning the steam generator nozzle, which would place a burden on the licensee.
The requirements for examinations of inner nozzle radii were developed in the ASME Code in reaction to the discovery of thermal fatigue cr9cks in the inner-radius section of feedwater nozzles. These thermal fatigue cracks were the result of internal water temperature fluctuations in the feedwater system. The main steam outlet nozzle is located at the top of the steam generator head. The steam vapor at this location has traveled through the dryers which removes excess liquid water droplets. During plant operations, this location is subjected to a relatively constant temperature and the saturated steam exiting the nozzle has a uniform temperature. The only temperature fluctuations at the nozzle inner radius are associated with reactor heat up and cool down, and these are controlled for the thermal effects on components.
Based on these factors, thermal fatigue is not expected to be an issue at this location.
The flow limiter contains nickel alloys, which can be subject to stress corrosion cracking. The flow limiter is overlaid with Alloy 52, and the venturi inserts are constructed from Alloy 690.
Additionally, these components are not in contact with primary coolant and are *normally only exposed to high-quality steam with very little water present. Based on the materials chosen and the operating environment, it is extremely unlikely that these components would be vulnerable to stress corrosion cracking.
The NRC staff concludes the ASME Code requirements are impractical for the steam generator main steam outlet nozzle inner radius examination. To meet the Code coverage requirements, design modifications would be necessary to provide access for examination. Imposition of the ASME Code requirements would result in a burden on the licensee. Based on the conditions at the nozzle inner radius, the visual examinations conducted during the system pressure tests provide reasonable assurance of structural integrity for the steam generator main steam outlet nozzles.
4.0 CONCLUSION
As set forth above, the NRC staff has reviewed the licensee's submittals and determined that granting relief pursuant to 1 0 CFR 50.55a(g)(6)(i) is authorized by law and Will not endanger life or property, or the common defense and security, and is otherwise in the public interest giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility. Furthermore, the staff concluded that the examinations performed to the extent practical provide reasonable assurance of structural integrity of the subject components. Accordingly, the NRC staff concludes that the licensee has adequately addres,sed all of the regulatory requirements set forth in 10 CFR 50.55a(g)(5). Therefore, the NRC staff grants relief for the subject examinations of the components contained in RR-4L2 for the fourth 1 0-year inspection intervals at the Point Beach Nuclear Plant, Unit 2.
All other ASME Code,Section XI requirements for which relief was riot specifically requested and approved in the subject request for relief remain applicable, including third-party review by the Authorized Nuclear lnservice Inspector.
Principal Contributor: S. Cumblidge Date: December 18 2013
ML13346A040 OFFICE LPL3-1/PM LPL3-1/LA NAME TBeltz MHenderson DATE 12/16/13 12/16/13 Sincerely,
/RAJ Robert D. Carlson, Chief Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
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LPL3-1/BC Tlupold RCarlson 12/06/13 12/18/13