ML14293A002
| ML14293A002 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 10/21/2014 |
| From: | David Pelton Plant Licensing Branch III |
| To: | Mccartney E Point Beach |
| Beltz T | |
| References | |
| TAC MF3163, TAC MF3164 | |
| Download: ML14293A002 (10) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Mr. Eric McCartney Site Vice President NextEra Energy Point Beach, LLC Point Beach Nuclear Plant 6610 Nuclear Road Two Rivers, WI 54241-9516 October 21, 2014
SUBJECT:
POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2-RELIEF FROM THE REQUIREMENTS OF THE AMERICAN SOCIETY OF MECHANICAL ENGINEERS (ASME) BOILER AND PRESSURE VESSEL CODE (CODE),
SECTION XI, FOR THE FIFTH 10-YEAR INSERVICE INSPECTION INTERVAL (TAC NOS. MF3163 AND MF3164)
Dear Mr. McCartney:
By letter dated November 27, 2013, NextEra Energy Point Beach, LLC (NextEra), submitted to the U.S. Nuclear Regulatory Commission (NRC) a request for relief proposing an alternative to the weld examination requirements specified in the 1998 Edition of the American Society of Mechanical Engineers Boiler and Pressure Code (ASME Code),Section XI, at the Point Beach Nuclear Plant (Point Beach), Units 1 and 2.
Specifically, pursuant to Title 10 of the Code of Federal Regulations (1 0 CFR), Part 50, Section 50.55a(a)(3)(ii), NextEra proposed an alternative to the weld examination requirements specified in the 1998 Edition of the ASME Code,Section XI, Table IWB-2500-1, explicitly required for use by the condition in 10 CFR 50.55a(b)(2)(xxi) for Category B-D, Item B3.120, components. The requirements pertain to volumetric examination and the alternative enhanced visual examination of the nozzle inner radius section of the integrally cast pressurizer surge line nozzle at Point Beach, Units 1 and 2. The basis for the alternative is that complying with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
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(a)(3)(ii). Therefore, the staff determined that complying with the specified requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Furthermore, the NRC staff determines that authorizing the use of the alternative provides reasonable assurance of structural integrity of the subject components.
If you have any questions, please contact Terry Beltz at (301) 415-3049, or via e-mail at Terry.Beltz@nrc.gov.
Docket Nos. 50-266 and 50-301
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 REGARDING RELIEF REQUEST RR-5 FOR THE FIFTH 10-YEAR INSERVICE INSPECTION INTERVAL NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT. UNITS 1 AND 2 DOCKET NOS. 50-266 AND 50-301 (TAC NOS. MF3163 AND MF3164)
1.0 INTRODUCTION
By letter dated November 27, 2103 (Agencywide Document Access and Management System (ADAMS) Accession No. ML13331A910), NextEra Energy Point Beach, LLC (NextEra, the licensee) requested relief from certain requirements for the fifth 1 0-year inservice inspection (lSI) interval of the lSI Program for the Point Beach Nuclear Plant (Point Beach), Units 1 and 2, currently scheduled to end on July 30, 2022.
Specifically, pursuant to Title 10 of the Code of Federal Regulations (1 0 CFR), Part 50, Section 50.55a(a)(3)(ii), the licensee submitted Relief Request (RR)-5 proposing an alternative to the weld examination requirements specified in the 1998 Edition of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code),Section XI, Table IWB-2500-1, explicitly required for use by the condition in 10 CFR 50.55a(b)(2)(xxi) for Category B-D, Item B3.120, components. The requirements pertain to volumetric examination and the alternative enhanced visual examination of the nozzle inner radius (NIR) section of the integrally cast pressurizer surge line nozzle at Point Beach, Units 1 and 2. The basis for the alternative is that complying with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
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, "Rules for lnservice Inspection of Nuclear Power Plant Components," to the extent practical within the Enclosure
' limitations of component design, geometry, and materials of construction. The regulations require that repair and replacement activities 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 limitations and modifications listed therein.
10 CFR 50.55a(a)(3}, states, in part, that alternatives to the requirements of paragraph (g) may be used, when authorized by the NRC, if the licensee demonstrates that (i) the proposed alternatives would provide 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.
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 relief, and that the regulatory authority exists to authorize the proposed alternative pursuant to 10 CFR 50.55a(a)(3).
3.0 TECHNICAL EVALUATION
3.1 Licensee's Relief Request Component Affected ASME Code Class:
Examination Category:
Item No.:
lSI Component ID:
==
Description:==
1 B-D ASME Code,Section XI, Item B3.120 (1998 Edition)
PZR-SurgeNoz-IRS (Unit 1 and Unit 2)
Full Penetration Welded Nozzles in Vessels Applicable ASME Code Requirement The fifth 1 0-year lSI program interval commenced at Point Beach, Units 1 and 2, on August 1, 2012. The current ASME Code of record for Point Beach, Units 1 and 2, is the 2007 Edition of the ASME Code with the 2008 addenda. The 2007 Edition with the 2008 Addenda of the ASME Code,Section XI does not require an examination of the pressurizer surge line NIR. However, the regulation in 10 CFR 50.55a(b)(2)(xxi)(A) mandates use of the 1998 Edition of Section XI for the examination requirements of full penetration welded nozzles in vessels.
Category B-D, Item B3.120, of Table IWB-2500-1 in the 1998 Edition of the ASME Code,Section XI, requires a volumetric examination of the NIR section of the pressurizer surge nozzle; however, 10 CFR 50.55a(b)(2)(xxi)(A) allows an enhanced visual (VT-1) on the inside surface in lieu of a volumetric requirement that is performed from the outside surface.
Basis for Relief At Point Beach, the Unit 1 and 2 pressurizer surge line nozzles are integrally cast into the bottom pressurizer lower head. The subject nozzle at each unit is located under the pressurizer skirt and is surrounded by 78 heater penetrations. The multi-layered, stainless steel mirror insulation and cables for the pressurizer heaters obstruct access to the subject nozzle.
Interference from the heater penetrations and heater cables, as well as the location of the nozzle under the pressurizer skirt, restricts access to the nozzle. This limits the examiner's ability to manipulate the search unit to examine the NIR. Additionally, the integrally cast nozzle contain limitations such as an irregular outside diameter (OD) profile, a rough surface condition and an attenuating grain structure. The irregular surface condition causes the beam angle to vary from point to point around the nozzle. The attenuating grain structure results in a low signal-to-noise ratio at the NIR. Limited access to the nozzle, as well as the limitations caused by the material conditions, area dose rates and the complicated nature of the examination technique would make evaluation of any indications very difficult. NextEra has performed a total of six previous "best effort" examinations between both Point Beach units with no recordable indications.
Access to the pressurizer surge line nozzle is obstructed by insulation and the cables for the pressurizer heaters. Removal of the insulation and cables would be difficult as well as labor and time intensive. The radiation exposure of the workers and technicians needed to perform the tasks is a real and relevant concern. It is almost certain that some, and possibly all, heater cables would have to be disconnected so the cables can be pulled back to allow access for removing insulation and performing the exam. It is also likely that some cable or heater damage would occur during removal. If it is assumed that all 78 heater cables have to be disconnected and pulled back, the high-end dose estimate is 12.8 man-rem. While actions would be taken to provide protection against radiation exposure, the large dose rate gradients in the under-pressurizer area present an unusual challenge. Temporary shielding is considered impractical in this situation because placement of the shielding material would obstruct, and potentially preclude, accessibility to the examination surface. The breakdown of the 12.8 man-rem dose estimate is provided in Table 3-1.
Table 3-1, Man-REM Dose Estimate for Examination of the Pressurizer Surge Line NIR TASK HOURS PEOPLE MAXIMUM DOSE TOTAL DOSE RATE (Rihr)
(Rem)
Scaffold Erection 4
4 0.2 3.2 Insulation Removal 6
3 0.2 3.6 Examination Area Preparation 1
2 0.2 0.4 Ultrasonic Examination 1
2 0.2 0.4 Insulation Installation 6
3 0.2 3.6 Scaffold Removal 2
4 0.2 1.6 TOTAL 20 12.8 Other potential personnel safety concerns with this examination include the increased risk for an unplanned exposure event and the possibility of personnel contamination with personnel wedged between the surge line and the exposed portion of the pressurizer heaters. Other likely issues are actual accessibility after removal of the various forms of interference and the likelihood of difficulties in replacing the insulation to its original configuration. Furthermore, the amount of examination coverage would be dependent on the overall accessibility obtained.
This volumetric inspection effort, and the significant potential risk associated with it are not commensurate with the limited benefit that may be obtained from the inspection.
In addition, there are several uncertainties regarding an alternative examination of the inside surface of the pressurizer surge line area by use of a remote visual tool. Such an examination requires that a baroscope be fed through the manway and down through openings in the heater support baffles. There is also a perforated basket diffuser covering the surge nozzle opening on the inside of the pressurizer, which results in additional difficulty in performing such an exam.
The baroscope would need to be positioned through the support plates, and then threaded through a perforation in the basket diffuser, if possible, to the pressurizer surge line area. This examination would also be partially obscured by the thermal sleeve, which extends beyond the inside radius area into the volume of the pressurizer. These obstructions would need to be overcome several times to achieve the required examination coverage. Furthermore, the resulting examination would only be of the cladding that covers the inside radius of the nozzle, which is considered to be only marginally beneficial in determining the structural integrity of the nozzle. Performing the visual inspection requires opening the reactor coolant system (RCS) and establishing access and foreign material exclusion controls. The baroscope itself has the potential to become lodged inside the perforated basket diffuser or behind a pressurizer heater support plate. This visual inspection effort, and the significant potential risk associated with it are not commensurate with the limited benefit that may be obtained from the inspection.
The licensee is an active member of the Electric Power Research Institute modification/rework package and has access to research results, and is therefore aware of industry trends of failure or indications in this area. A search on the Institute of Nuclear Power Operation website of operating experience involving degradation at the inside radius section of the surge line nozzle in a Westinghouse designed pressurizer as well as a query among industry experts identified no known service induced indications for pressurizer NIR sections at any Pressurized Water Reactor (PWR) plants.
Any ultrasonic examination on this nozzle could only be described as "best effort." The benefit gained would not be commensurate with the difficulty, potential to damage system components, and estimated 12.8 man-rem personnel exposure to perform the volumetric examination. An alternative examination employing a remote visual technology has very little if any reasonable probability of success and poses additional threats to the system considering foreign material controls and the potential for the equipment to get trapped during examination. As such, permission to use the proposed alternative is requested pursuant to 10 CFR 50.55a(a)(3)(ii) since compliance with the specified requirements would result in hardship or unusual difficulties without a compensating increase in the level of quality and safety.
Licensee's Proposed Alternative Examination A visual (VT-2) examination is performed at the end of each refueling outage during the system leakage test as required by Table IW8-2500-1, Examination Category 8-P, Item 8 15.1 0, of the 2007 Edition through 2008 Addenda of Section XI.
The pressurizer cast surge inner corner region is subject to a VT -2 visual examination as part of the system leakage test on the pressurizer vessel conducted each refueling outage as specified in Table IW8-2500-1, Examination Category 8-P of the 2007 Edition through 2008 Addenda of ASME Section XI.
It is NextEra's position that, based on the acceptable results of VT -2 visual examinations performed during the Class 1 system leakage test, there is a reasonable assurance of continued structural integrity of the subject component and an acceptable level of quality and safety is maintained without performing the required inner radius section examinations required by the 1998 Edition of the ASME Code,Section XI, Table IW8-2500-1, Category 8-D, Item 83.120.
3.2
NRC Staff Evaluation
The licensee is proposing to perform VT-2 examinations of Pressurizer Surge Line Nozzle-to-Vessel Weld as part of the normally scheduled ASME Code, Class 1 system leakage test each refueling outage in lieu of the 10 CFR 50.55a(b)(2)(xxi)(A) requirements. 10 CFR 50.55a(b)(2)(xxi)(A) requires the use of inspection requirements in the 1998 Edition of ASME Code,Section XI be applied for Category 8-D, Item 83.120, the pressurizer NIR section. This regulation also permits the use of enhanced VT-1 visual examination of the interior surface of the NIR section in lieu of the volumetric examination from the outside surface as required by the 1998 Edition of the ASME Code.
The requirements for examinations of inner nozzle radii were developed in the ASME Code in reaction to the discovery of thermal fatigue cracks in the inner-radius section of boiling water reactor feedwater nozzles. These thermal fatigue cracks were the result of internal water temperature fluctuations in the feedwater system. The NRC staff is unaware of any operating experience involving degradation (i.e., indications) in pressurizer NIR sections or for any reactor or steam generator NIR sections at pressurized water reactor plants.
The NRC staff agrees with the licensee that the ultrasonic examination limitations due to the design of integrally cast pressurizer surge nozzles (obstructions to search unit manipulation and coverage, irregular outer diameter profile, rough surface coupling condition and attenuating grain structure) would limit the volumetric examination coverage to below the ASME Code requirement of "essentially 100 percent" and present a significant detection challenge considering the signal-to-noise ratio expected. Therefore, the examination would only provide a "best effort" result, and limit the safety related benefit of the examination.
The NRC staff finds that in order for the licensee to volumetrically examine the pressurizer surge line nozzle-to-vessel weld and pressurizer NIR section, it would have to remove insulation and heater cables, thus exposing licensee personnel to an estimated dose of 12.8 man-rem and potentially resulting in personnel contamination from newly exposed surfaces. The NRC staff also finds that use of temporary shielding to mitigate exposure would further obstruct the examination surface. In addition, disconnecting the heater cables could also cause damage to both the cables and heaters. These issues pose a hardship on the licensee.
The NRC staff finds that, based on the description of the pressurizer access provided in the documents referenced within the licensee's submittal, the alternative VT-1 examination with a remote visual technology would have limited benefit and a limited probability of success. It would be very difficult for the licensee to feed a baroscope through the pressurizer access manway, down through openings in the heater support baffles and perforated basket diffuser covering the surge nozzle opening. The perforated basket diffuser covering the nozzle opening on the inside of the pressurizer would make it difficult, if not impossible, to place a probe into the subject nozzle opening. The examination would be partially obscured by the thermal sleeves and the examination would only be of the non-structural cladding covering the welds.
Additionally, the threat of the foreign material introduction, including the potential for the baroscope to become lodged in the many orifices it must pass through, poses an operational safety hazard to the pressurizer system.
The NRC staff reviewed the license renewal documents and the updated final safety analysis report for Point Beach. The staff finds that the pressurizer surge line location applicable to the NUREG/CR-6260 analysis is the hot leg surge line nozzle, which like the surge line pressurizer nozzle is susceptible to environmentally assisted fatigue (EAF). This location is also the fatigue limiting location in the Point Beach pressurizer surge line as identified by fatigue analysis based on real plant operational data; therefore, it is bounding for the pressurizer surge nozzle. The plant-specific analysis of the hot leg surge line nozzle produces a projected 60-year cumulative usage factor (CUF) of 0.00147 (Unit 1) and 0.00079 (Unit 2), based on actual plant transients and the design number of heatup and cooldown cycles. Since this location is stainless steel, the worst-case multiplier of 15.35 produces an EAF-adjusted CUFs of 0.0226 (Unit 1) and 0.0122 (Unit 2), which is acceptable.
The environmental effects on fatigue evaluation for the hot leg surge line nozzle were also evaluated using an alternate approach. The evaluation used over nine years of actual plant data from 1994 through 2003. This evaluation used a mean plus one standard deviation methodology to produce 60-year CUFs of 0.0383 (Unit 1) and 0.007132 (Unit 2). The most conservative value of 0.0383 was multiplied by the worst-case multiplier of 15.35 to yield an EAF-adjusted CUF of 0.588, which is also acceptable. The CUFs for Units 1 and 2 are considerably less than the design limit of 1.0. In addition, the NRC staff is unaware of any operating experience involving degradation in pressurizer NIR sections or for any reactor or steam generator NIR sections at PWR plants. NextEra also performed six "best effort" volumetric examinations between the two units of the subject components with no recordable indications. This provides insight into the very low potential for failure in this system component and therefore adds reasonable assurance of the component integrity and leak tightness for the duration of the Point Beach fifth 1 0-year lSI interval.
The licensee has an active Boric Acid Corrosion Control Program that identifies and monitors borated water leakage to prevent boric acid related degradation of the RCS. The Point Beach technical specification surveillance requirements regarding RCS leak rate and containment atmosphere radioactivity will further ensure the integrity of the pressurizer surge line nozzle.
These programs ensure that a small amount of leakage will be identified and corrected prior to adversely affecting the overall level of plant quality and safety. The NRC staff finds this to be an acceptable approach.
Therefore, the NRC staff concludes that based on the above, the ASME Code-required volumetric examination and/or the optional visual examination discussed in 10 CFR 50.55a(b)(2)(xxi)(A) would impose a hardship on the licensee without a compensating increase in quality and safety. The staff has determined that the licensee's requirement to perform a VT-2 visual examination each outage, maintain a Boric Acid Corrosion Control Program, monitor technical specification leak rate indications, and the low component CUF with no industry operational experience of pressurizer NIR material degradation, provides reasonable assurance of structural integrity and leak tightness.
4.0 CONCLUSION
As set forth above, the NRC staff has reviewed the licensee's submittal and determines that use of the proposed alternative presented in RR-5 provides reasonable assurance of structural integrity and leak tightness, and that complying with the specified requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
Accordingly, the 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 compliance with the ASME Code's requirements. Therefore, the NRC staff authorizes RR-5 for the Point Beach Nuclear Power Plant, Units 1 and 2, pursuant to 10 CFR 50.55a(a)(3) for the fifth 1 0-year inservice lSI interval currently scheduled to end on July 30, 2022.
All other ASME Code,Section XI requirements for which relief was not 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: October 21, 2014
ML14293A002 OFFICE LPL3-1/PM LPL3-1/LA NAME TBeltz MHenderson DATE 10/20/14 10/20/14 Sincerely, IRA/
David L. Pelton, Chief Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation RidsNrrDeEpnb Resource RidsRgn3MaiiCenter Resource RidsAcrsAcnw_MaiiCTR Resource J. Jandovitz, EDO Rill S. Cumblidge, NRR
- via email EPNB/BC*
LPL3-1/BC DAlley DPelton 10/10/14 10/21/14