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#REDIRECT [[SBK-L-12146, NextEra Energy Seabrook License Renewal Application, Comments on NRC Safety Evaluation Report with Open Items and Editorial Corrections to Supplement #25]]
| number = ML12208A110
| issue date = 07/20/2012
| title = NextEra Energy Seabrook License Renewal Application, Comments on NRC Safety Evaluation Report with Open Items and Editorial Corrections to Supplement #25
| author name = Walsh K T
| author affiliation = NextEra Energy Seabrook, LLC
| addressee name =
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000443
| license number =
| contact person =
| case reference number = SBK-L-12146
| document type = Letter
| page count = 13
| project =
| stage = Supplement
}}
 
=Text=
{{#Wiki_filter:NEXTera ENERGY_ _July 20, 2012 SBK-L-12146 Docket No. 50-443 U.S. Nuclear Regulatory Commission Attention:
Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852 Seabrook Station NextEra Energy Seabrook License Renewal Application Comments on NRC Safety Evaluation Report with Open Items and Editorial Corrections to Supplement  
#25
 
==References:==
: 1. NextEra Energy Seabrook, LLC letter SBK-L-10077, "Seabrook Station Application for Renewed Operating License," May 25, 2010. (Accession Number ML 101590099)
: 2. NRC Letter, "Safety Evaluation Report with Open Items Related to the License Renewal of Seabrook Station," dated June 8, 2012 (Accession Number MLI 2053A1 92)3. NextEra Energy Seabrook, LLC letter SBK-L-12123, "NextEra Energy Seabrook License Renewal Application Supplement
#25," June 19, 2012. (Accession Number ML12178A405)
In Reference 1, NextEra Energy Seabrook, LLC (NextEra) submitted an application for a renewed facility operating license for Seabrook Station Unit I in accordance with the Code of Federal Regulations, Title 10, Parts 50, 51, and 54.Reference 2 contains the NRC "Safety Evaluation Report with Open Items Related to the License Renewal of Seabrook Station." Enclosure I provides NextEra's comments on this report.Reference 3 contains Supplement
#25, LRA changes associated with LR-ISG-201 1-01, "Aging Management of Stainless Steel Structures and Components in Treated Borated Water." Enclosure 2 contains editorial corrections to the LRA changes contained in Supplement
#25 to the LRA.There are no new or revised regulatory commitments contained in this letter.NextEra Energy Seabrook, LLC, P.O. Box 300, Lafayette Road, Seabrook, NH 03874 United States Nuclear Regulatory Commission SBK-L-12146/Page 2 If there are any questions or additional information is needed, please contact Mr. Richard R.Cliche, License Renewal Project Manager, at (603) 773-7003.If you have any questions regarding this correspondence, please contact Mr. Michael O'Keefe, Licensing Manager, at (603) 773-7745.Sincere y, Nex Energy Seabrook, LLC.Kevin T. Walsh Site Vice President
 
==Enclosures:==
 
Enclosure 1- NextEra Comments on NRC Safety Evaluation Report with Open Items Enclosure 2- Editorial Corrections to Information Contained in SBK-L- 12123 cc: W.M. Dean, NRC Region I Administrator J. G. Lamb, NRC Project Manager, Project Directorate 1-2 W. J. Raymond, NRC Resident Inspector A.D. Cunanan, NRC Project Manager, License Renewal M. Wentzel, NRC Project Manager, License Renewal Mr. Christopher M. Pope Director Homeland Security and Emergency Management New Hampshire Department of Safety Division of Homeland Security and Emergency Management Bureau of Emergency Management 33 Hazen Drive Concord, NH 03305 John Giarrusso, Jr., Nuclear Preparedness Manager The Commonwealth of Massachusetts Emergency Management Agency 400 Worcester Road Framingham, MA 01702-5399 United States Nuclear Regulatory Commission SBK-L- 12146/ Page 3 NExTera ENERY I, Thomas A. Vehec, Plant General Manager of NextEra Energy Seabrook, LLC hereby affirm that the information and statements contained within are based on facts and circumstances which are true and accurate to the best of my knowledge and belief.Sworn and Subscribed Before me this'ZO day of July, 2012 Thomas A. Vehec I2' ,7 , /i Plant General Manager Notary Enclosure 1 to SBK-L-12146 NextEra Comments on NRC Safety Evaluation Report with Open Items United States Nuclear Regulatory Commission Page 2 of 8 SBK-L-12146
/ Enclosure 1 Comments Regarding Seabrook Station License Renewal Safety Evaluation with Open Items Section/Page Comment Suggested Resolution (Changes shown in bold and strikethrough.)
iii Editorial The unit is a 4-loop pressurized-water reactor (PWR) design. Westinghouse Genera4 E.eetf.ie
.Company supplied the nuclear steam supply system. Wetinghettse United Engineers and Constructors constructed the plant.TABLE OF Editorial Add T of C page numbers (e.g. viii) and make sectional page number corrections.
CONTENTS 2.1.4.7.2/2-30 LRA The staff noted that, after the scoping of electrical and I&C components was performed, the in-scope electrical components were categorized into electrical component types.Component types include similar electrical and I&C components with common characteristics.
Component-level intended functions of the component types were identified-such as cable, connections, fuse holders, terminal blocks, high voltage transm.issi.n ,onduetor, connections and insulators, metal enclosed bus, switchyard bus, and connections.
2.3.2.3.1/2-45 LRA
* PID-1-CBS-LR20233
* PID-1-RC-LR20841
* PID-I-RH-LR20663" PIID-l-VSL-LR20776" PID-1-CS-LR20722" PID-l-RC-LR20844
* PID-1-CS-LR20725
-PID 1 SI LR20446 PID 1 RIH LR20662* PID-1-SI-LR20446-PID-1-SI-LR20449" PID-1-RH-LR20661-PID-1-RH-LR20662" PID-l-WLD-LR20221" PID-1-SI-LR20450 2.3.3/2-48 Editorial The applicant deseiibed stated components such as instrumentation and components found on vendor drawings were already included within the scope of license renewal but not depicted on the LRA drawings.2.3.3.6.2
/2-54 LRA The containment enclosure cooling units maintain the first six areas (charging pump areas, SI pump areas, RHR equipment areas, containment spray pump and heat exchanger equipment areas, mechanical penetration area, containment enclosure ventilation equipment area, and H2 analyzer room and electrical room areas) at, or below the safety-related equipment's maximum design operating temperatures during normal operation and following a LOCA, loss of offsite power, high and moderate pipe breaks, SSE, and tornados.
United States Nuclear Regulatory Commission SBK-L-12146
/ Enclosure 1 Page 3 of 8 Section/Page Comment Suggested Resolution (Changes shown in bold and strikethrough.)
The H2 analyzer and electrical room supply fans maintain area 7 at or below the safety-related equipment's maximum design operating temperatures during normal operation and following a LOCA, loss of offsite power, high and moderate pipe breaks, and an SSE.2.3.3.17.1/
2-69 Editorial The fuel oil system provides fuel to the two diesel driven fire pumps-1l-F4P-FP-P-20A and I--FP-P-FP-P-20B.
There are two fuel tanks, each dedicated to a diesel driven fire pump.2.3.3.19.1/
2-70 Editorial In addition to providing heating to buildings not within the license renewal boundary, such as the administration building, turbine building and waste process building, the HW system provides the functions described below.2.3.3.19.2
/2-72 SBK-L-11015 In its response dated February 3, 2011, the applicant clarified that the 1/2 in. vent lines are within the scope of license renewal for 10 CFR 54.4(a)(2).
The applicant also described the 1/2 in. vent lines as 1/2 in. carbon steel piping with threaded plugs at the end.Based on its review, the staff finds the applicant's response to RAI 2.3.3.19-01 acceptable because the staff confirmed that the 1/2 in. vent-lines are already included within the scope of license renewal under 10 CFR 54.4(a)(2).
Therefore, the staff's concern described in RAI 2.3.3.19-01 is resolved.2.3.3.26.1
/2-78 Editorial The floor drains in this system are located outside any area with a potential for contamination.
The plant floor drain system is located in those areas where automatic sprinkler and spray systems are installed.
2.3.3.31.1
/2-85 Editorial The reactor makeup water system consists of one reactor makeup water storage tank, two redundant, full capacity reactor makeup water pumps, and associated piping, valves, instrumentation, and controls.2.3.3.34.2 2-87 Editorial In RAI 2.3.3.34-01, dated January 5, 2011, the staff noted on LRA drawing, PID-I-DR-LR20633, that the applicant depicts 6 in. lines that continue onto LRA drawing PID-1-SD-LR20402.
On LRA drawing PID-1-DR-LR20633, the applicant depicts 6 in. piping that enters the continuation flag marked "B" as being included within scope of license renewal under 10 CFR 54.4(a)(2).
The applicant depicts the other 6 in. piping that enters the continuation flag, marked "C," as being excluded from scope of license renewal.However, the applicant depicts on the continuation LRA drawing, PID-1-SD-LR 20402, the 6 in. piping for "B" as being excluded from scope of license renewal, while the other 6 in. piping for "C" is shown as being included within scope of license renewal. The applicant was asked to clarify the scoping classifications of both 6 in. piping sections on both LRA drawings.2.3.3.37.1
/2-90 Editorial The four service water pumps take suction from a common bay in the service water pumphouse.
Seawater flow is supplied to the service water pumphouse from the Atlantic Ocean due to the static head of the ocean above the elevation of the service water pump suctions pumflps.
United States Nuclear Regulatory Commission SBK-L-12146
/ Enclosure 1 Page 4 of 8 Section/Page Comment Suggested Resolution (Changes shown in bold and strikethrough.)
2.3.3.45.2
/2-101 Editorial In RAI 2.32.3.45-02, dated January 5, 2011, the staff noted on LRA drawing PID-1-WLD-LR20218, at location H-5, that the applicant depicts a relief valve tailpipe connected to relief valve V83 as not being within scope of license renewal.Based on its review, the staff finds the applicant's response to RAI 2.32.3.45-02 acceptable because the applicant revised its position regarding the relief valve tailpipe to include it within the scope of license renewal under 10 CFR 54.4(a)(2) for spatial interaction.
Therefore, the staff's concern described in RAI 2.32.3.45-02 is resolved.T3.0-1/3-9, -10 SBK-L-10204 ASME Code Section XI, Consistent with Subsection IWE Program enhancements LRA Inaccessible Power Medium Voltage-Cables Not Subject to 10 CFR 50.49 EQ Requirements Program Editorial 345 kV sulfur hexafluoride (SF6) A.244 2.2.1 Bus Program B.2.2.1 3.0.3.1.2
/3-15 Editorial The applicant further stated that during the April 2008 refueling outage, it replaced the faulty floating lid seal, and since this time, the specific conductivity has remained below the 0.1 (8)/em micro siemens (,US) limit.3.0.3.1.13
/3-41 SBK-L-10192 The applicant also stated that this program applies to sensitive instrumentation cable and connection circuits with low-level signals in the in-scope portions of the in-eere neutron flux monitoring cable in the nuclear instrumentation system.3.0.3.2/3-49 LRA
* Inaccessible Mediwm- Veltage Power Cables Not Subject to 10 CFR 50.49 EQ Requirements Program 3.0.3.2.3
/3-62 SBK-L-12084 With respect to corrective actions taken in response to this event, the applicant stated that a design change will be developed to replace the PVC-lined piping in 2012 with a corrosion-resistant, unlined material, and that the associated service water piping will be periodically inspected to verify adequate pipe wall thickness, until replaced.3.0.3.2.18
/3-133 SBK-L-10204 From 2000-2004, testing and inspection continued in the spent fuel pool, the spent fuel transfer canal, and the cask handling area to identify the source of the leak. In 2004 2-2N, the applicant installed a nonmetallic liner in an attempt to stop the leakage.3.0.3.2.21
/3-155 Editorial In addition, in order to support the 60-year TLAAs associated with metal fatigue of the RCS pressure boundary, the applicant analyzed the projected CUF, incorporating United States Nuclear Regulatory Commission SBK-L-12146
/ Enclosure 1 Page 5 of 8 Section/Page Comment Suggested Resolution (Changes shown in bold and strikethrough.)
environmental effects for seven locations specified in NUREG/CR-6260, and found that the CUFs when including environmental effects for the surge line hot-leg nozzle and the charging nozzle will exceed 1.0 for 60 years of service.3.0.3.3.1
/3-161
* test the pressure-retaining capability of the piping system by an elevated pressure test which can detect current or near tern leaks T3.0.3.3.1-1/3-162 Editorial See page 7 3.0.3.3.2
/3-168 SBK-L-1 1173 (Annual Update)
* The first path includes the SF6 bus from 345 kV Power Circuit Breakers 11 and 12 4-6 to the generator step-up transformer and to the unit auxiliary transformers via the isolated phase bus.3.0.3.3.3
/3-173 LRA The applicant stated that the standard coupons are placed in the spent fuel pool for monitoring of the aging effects and control coupons were supplied, in addition to standard coupons, to benchmark coupon initial conditions, to monitor possible uncontrolled changes in Boral material that are unrelated to the spent fuel pool conditions, and to demonstrate comparisons between different examination techniques and service contractors.
The 3.0.3.3.4
/ Editorial Page 3-179 3-179, 3-183 SRP-LR GALL Repe,'t Table 3.1-1, IDs 31 and 34,and their subordinate items, are unique in the GALL Report in that they do not recommend that aging be managed through the use of an AMP contained in the SRP-LR GALL Repe,4. Rather, they recommend specific aging management activities.
For nickel-alloy materials addressed by these AMR items, the recommended aging management activities consist of the following: " use of Inservice Inspection (IWB, IWC, and IWD) AMP" use of Water Chemistry AMP" compliance with all NRC Orders" commitment to implement applicable bulletins and generic letters" commitment to implement staff accepted industry guidelines The approach taken by the SRP-LR GALL Repert for these AMR items permits license renewal applicants to demonstrate consistency with the GALL Report by citing these aging management activities in their LRA AMR items. Alternatively, as has been done in this case, consistency with the GALL Report may be demonstrated for these items by developing an AMP and citing it for each applicable AMR item, which is consistent with SRP-LR Section A. 1.2.3 and which addresses all of the recommended aging management activities listed above. The staff's review of this AMP is, therefore, designed to verify consistency with SRP-LR Section A. 1.2.3 and to ensure that the aging management activities listed for nickel-alloy components included in SRP-LR GALL Repeo4 Table 3.1-1, IDs 31 and 34, are addressed by the AMP.Page 3-183 The staff finds this response acceptable because the applicant committed to the United States Nuclear Regulatory Commission SBK-L- 12146/ Enclosure 1 Page 6 of 8 Section/Page Comment Suggested Resolution (Changes shown in bold and strikethrough.)
implementation of applicable bulletins, generic letters, and staff-accepted industry guidelines, and is consistent with SRP-LR GALL .Repe4 Table 3.1-1, IDs 31 and 34, and their subordinate items. The staff's concern described in RAI B.2.2.3-1 is resolved.3.0.3.3.5/3-183 Editorial Summary of Technical Information in the Application.
LRA Section B.2.1.7 describes the PWR Vessel Internals Program as plant-specific.
The applicant stated that the PWR Vessel Internals Program is a new program which manages the aging effect of cracking due to irradiation-assisted stress corrosion cracking (IASCC), PWSCC, ifitefgfanfflar stress corrosion cracking (IGSCC), reduction in fracture toughness due to radiation and thermal embrittlement, void swelling, and loss of preload in RVI components.
3.4.2.1.7/3-413 Editorial The staff agrees with the applicant's observation that SRP-LR Table 3.1-1, ID 84, applies specifically only to once through steam generators (Table IV.D2).3.5.2.2.1
/3-469 Editorial The concrete and the moisture barrier where the liner becomes embedded are monitored under the ASMIE Code Section XI, Subsection IWE gWL Program.T3.6-1/3-489 LRA Inaccessible Medium Veltage Power Cables Not Subject to 10 CFR 50.49 EQ Requirements 3.6.2.1 /3-491 LRA
* Inaccessible Power Medium Veltage Cables Not Subject to 10 CFR 50.49 EQ Requirements Program SECTION 4/1-86 Editorial Change pages 1 through 86 to 4-1 through 4-86 4.2.2.2/11 Editorial The staff confirmed the applicant's projected USE values at the end of the period of extended operation for beltline and extended beltline materials by using Position 1.2 of RG 1.99, Revision 2. The staff's analysis verified that the lower shell plate B 1808-2 was the limiting material with a projected USE value of 59 ft-lbs after -56 55 EFPY.4.2.4.2/14 This letter was a draft During its review of the LAR, the staff, requested by letter dated April 25, 2012, that the applicant provide additional information related to whether the methodology used to develop the P-T limits is consistent with the requirements in 10 CFR 50, Appendix G.4.3.2.2.1
/23 Editorial The applicant dispositioned the "NRC Bulletin 88-11, Pressurizer Surge Line Thermal Stratification," related ASME Section III, Class 1, fatigue TLAAs (without environmental effects) for the PSL, PSN, and HSLN HLSN in accordance with 10 CFR 54.2 1(c)(l)(i), that the analyses remain valid for the period of extended operation.
TA-1/ See page 8 A-5, -9, -10 1 United States Nuclear Regulatory Commission SBK-L-12146
/ Enclosure 1 Table 3.0.3.3.1-1 Buried Piping Inspections (Page 3-162)Page 7 of 8 System Code Class or Contains Material Committed Safety Related Hazardous Inspections Materials Control Building Yes No Steel, with no 44W Air Handling cathodic protection Fire Protection No No Steel, with no 4-6--cathodic protection Diesel Generator No No Steel, with no 4164g 4/6/82 cathodic protection Plant Floor No No Steel, with no 4/6/82 Drains (not cathodic safety related) protection Condensate No No Steel, with no cathodic protection Feedwater Yes No Steel, with no 4*i8 cathodic protection Control Building Yes No Steel, with IE44TA14-Air Handling cathodic protection Instrument Air No No Steel, with 44-AN4 2  1/NA/42 cathodic protection Fire Protection No No Steel, with 44/A/4-cathodic protection Service Water Yes No Steel, with cathodic protection Condensate No No Stainless Steel -l-Diesel Generator Yes Yes Stainless Steel 4--Cooling Water'Fire Protection No No Fiberglass 1/NA/2 2 United States Nuclear Regulatory Commission SBK-L- 12146 / Enclosure 1 Table A-I. Seabrook Station License Renewal Commitments (Pages A-5, -9, -10)Page 8 of 8 Reference 29) Lubricating Oil Enhance the program to sample the A.2.1.26 Prior to the period of extended operation.
SBK-L- 11207, 11/2/2011 Analysis oil for the Switehyar-f SF6..mprcz.....
and the Reactor Coolant pump oil collection tanks.3 1) ASME Section X/ Xi, Enhance procedure to. include the A.2.1.28 Prior to the period of extended operation.
Editorial Subsection IWL definition of "Responsible Engineer".
: 50) ASME Section XI, Perform UT testing of the A.2.1.27 No lat.. than Der.emb.ber31 , 2015 and repeated-SBK-L-11154, 8/11/2011 Subsection IWE containment liner plate in the at intervals
.f no mere than  vicinity of the moisture barrier for outages loss of material.
Within the next two refueling outages, OR15 or OR16, and repeated at intervals of no more than five refueling outages 52) ASME Section XI, Implement measures to maintain A.2.1.28 By-201-3 By December 31, 2012 SBK-L- 12084, 4/26/2012 Subsection IWL the exterior surface of the Containment Structure, from elevation
-30 feet to +20 feet, in a dewatered state.60) Buried Piping and Implement the design change A.2.1.22 Prior to entering the period of extended SBK-L-12084, 4/26/2012 Tanks Inspection replacing the buried Auxiliary operation Boiler supply piping with a pipe-within-pipe configuration with leak__dieatien detection capability.
: 62) Water Chemistry Enhance the program to include a A.2.1.2 Prior to enterin the period of extended SBK-L- 12084, 4/26/2012 statement that sampling operation.
frequencies are increased when chemistry action levels are exceeded.
Enclosure 2 to SBK-L-12146 Editorial Corrections to Information Contained in SBK-L-12123 United States Nuclear Regulatory Commission SBK-L-12146
/ Enclosure 2 Page 2 of 2 a) On page 15 of 48 of in SBK-L12123/Enclosure 1, Item 58 is revised as follows: 58. In Table 3.2.2-3, on page 3.2-69, the 1st row is revised as follows: Water Treated V Chemistry A Pressure Borated Bedy Program V.D1-31 Valve Body Boundary CASS Water (E-12) 3.2.1-48>1400 F Cracking One-Time (Internal)
Inspection A Program b) On page 26 of 48 of in SBK-L12123/Enclosure 1, Item 99 is revised as follows: 99. In Table 3.3.2-3, on page 3.3-155, the 7 th row is revised as follows:}}

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