SBK-L-11240, Additional Information NextEra Energy Seabrock Licnese Renewal Application Aging Managment Programs

Additional Information NextEra Energy Seabrock Licnese Renewal Application Aging Managment Programs

ML11354A235
Person / Time
Site:	Seabrook
Issue date:	12/15/2011
From:	Freeman P NextEra Energy Seabrook
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
SBK-L-11240
Download: ML11354A235 (13)

Text

NExTer'a' EN ERG'7yz SEABROK December 15, 2011 SBK-L-1 1240 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 Additional Information NextEra Energy Seabrook License Renewal Application Aging Management Programs

References:

1. NextEra Energy Seabrook, LLC letter SBK-L-10077, "Seabrook Station Application for Renewed Operating License," May 25, 2010. (Accession Number ML101590099)

2. NRC Letter "Request for Additional Information Related to the Review of the Seabrook Station License Renewal Application (TAC NO. ME4028) - Aging 'Management Programs" December 14, 2010 (Accession Number ML103260554)

3. NextEra Energy Seabrook, LLC letter SBK-L- 11002, Response to Request for Additional Information NextEra Energy Seabrook License Renewal Application Aging Management Programs - Set 4, January 13, 2011 (Accession Number ML110140809)

4. Summary Of Telephone Conference Call Held On November 22, 2011, Between The U.S.'

Nuclear Regulatory Commission And Nextera Energy Seabrook, LLC, Concerning The Response To The Request For Additional Information Pertaining To The Seabrook Station, License Renewal Application (TAC No. ME4028). (Accession Number ML11327A072)

.In Reference 1, NextEra Energy Seabrook, LLC (NextEra) submitted an application for a renewed facility operating license for Seabrook Station Unit 1 in accordance with the Code of Federal Regulations, Title 10, Parts 50, 51, and 54.

NextEra Energy Seabrook, LLC, P.O. Box 300, Lafayette Road, Seabrook, NH 03874

United States Nuclear Regulatory Commission SBK-L-1 1240 / Page 2 In Reference 2, the NRC requested additional information in order to complete its review of the License Renewal Application (LRA). In Reference 3, NextEra provided a response to RAIs related to the Metal Fatigue Aging Management Program. During staff review of the LRA an additional question regarding action limits associated with the personnel*airlock and equipment hatch wasraised (Reference 4). Enclosure 1 containsNextEra's revised response to the previous request for additional information. For clarity the revised response shows deleted, text highlighted by strikethroughs and inserted text highlighted by bold italics.

There are no new or revised regulatory commitments contained in this letter.

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.

Sincerely, NextEra Energy Seabrook, LLC.

Paul 0. Freeman Site Vice President

Enclosures:

- Revised Response to NextEra letter SBK-L- 11002, Request for Additional Information Seabrook Station License Renewal Application Aging Management Programs

United States Nuclear Regulatory Commission SBK-L-11240 / Page 3 cc:

W.M. Dean, NRC Region I Administrator G. E. Miller, NRC Project Manager, Project Directorate 1-2 W. J. Raymond, NRC Resident Inspector R. A. Plasse Jr., 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- 11240 / Page 4 N~x~era-

ENERGY, SEABROOK I, Paul 0. Freeman, Site Vice President 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

/f dayof CeA'bv, /- ,2011 Paul 0. Freeman Site Vice President Notary Pub c. *. *':,L, -

Enclosure 1 to SBK-L-11240 Revised Response to NextEra letter SBK-L-11002 Request for Additional Information Seabrook Station License Renewal Application Aging Management Programs

United States Nuclear Regulatory Commission Page 2 of 9 SBK-L- 11240 / Enclosure 1 Issue In a conference call on November 22, 2011 the staff inquired how NextEra will track design limits related to plant startups and shutdowns as listed in LRA Section 4.6.2 related to the Equipment Hatch and Personnel Air Lock. As previously noted in NextEra's response to RAI B.2.3.1-3 and RAI B.2.3.1-4 (Reference 3) the design limit tracked by FatiguePro is 200 Plant Heatups and Cooldowns with an 80% trigger level for further evaluation. This action limit would exceed the 120 cycle design limit for the Personnel Airlock and Equipment'Hatch as specified in LRA section 4.6.2.

-NextEra Energy Seabrook Response NextEra has revised LRA Table 4.3.1-2 previously submitted in response to RAI B2.3.1-3 to include the specific plant startup and shutdown design limit of 120 cycles for the Personnel Airlock and Equipment Hatch. Cycle counting for these specific components will initiate appropriate evaluations through the corrective action program if the 80% action limit is reached.

As previously stated in RAI B.2.3.1-4, an action limit of 80% will be used by the Metal Fatigue of Reactor Coolant Pressure Boundary Program for all limits tracked in FatiguePro. This action limit will provide sufficient margin and time to allow for appropriate corrective actions as defined in the Metal Fatigue of Reactor Coolant Pressure Boundary Program to be implemented prior to reaching the design limit.

NextEra has reviewed the LRA and did not identify additional non conservative design limits utilizedin TLAA analysis.

Revised NextEra Enerzy Seabrook Response to RAI B.2.3.1-3: LRA Table 4.3.1-2

[ See Following Pages for Revised Table 4.3.1-2]

United States Nuclear Regulatory Commission Page 3 of 9 SBK-L-1 1002 / Enclosure I Heatup and Cooldown Plant Cooldown @ < 100 OF/hr 200 at 100F per hour Plant (RCS) Cooldown Y Y UFSAR Section 3.8.2.3 PlantStartup and Shutdown 120 EquipmentHatch and N Y PersonnelAirlock Pressurizer Heatup 200 . Not Specified N Y Pressurizer cooldown Pressurizer Cooldown 200 Presurper 200TF per hour N Y Unit Loading @ 5% full 13,200 Unit Loading and power/mi Unloading at 5 Percent Unit Unloading @ 5% full 3 of Full Power per power/min 13,200 Minute Step Load Increase of 10% of full 2,000 Step Load Increase and y power Decrease of 10 Percent Step Load Decrease of 10% of full power por2,000 of Full Powery Large step load decrease with 2 Large Step Load S Yr steam dump 200 Decrease with Steam Dump Larease Dces Load

United States Nuclear Regulatory Commission Page 4 of 9 SBK-L-1 1002 / Enclosure 1 initial - i.- x 105 Steady-State Steady state fluctuations (7)

Fluctuations N N Random - 3.0 x 105 Feedwater Cycling at Hot 2,000 Feedwater Cycling at Feedwater Cycling Shutdown Hot Shutdown Loop out of service (4)

Loop out of service Normal loop shutdown 80 *Normal loop shutdown Normal loop startup 70 Normal loop startup Feedwater Heaters out Feedwater Heaters out of service of service One heater out of service One heater out of 120 N Y One bank of heaters out of service 120 service One bank of heaters Out of service Unit Loading and y Y Unloading Between 0

• and 15 Percent of Full Power Y Y Boron Concentration N N Equalization Refueling Refueling Y " ______Y_"

United States Nuclear Regulatory Commission Page 5 of 9 SBK-L- 11002 / Enclosure I Keduceci temperature return to Reduced Temperature Y 2,000 Y Dower Return to Power Reactor Coolant Pumps Reactor Coolant Pumps 3,000 (3) (RCP) Startup and Y Y startup/shutdown Shutdown Letdown Flow Step Decrease and Letdown Flow Step N Y 2,000 Not Specified Decrease and Return Return (6)

Upset Trasi~ents: ____________

Loss of load without immediate Loss of Load (Without 80 Immediate Turbine Loss of Turbine Load Y .Y turbine trip Trip)

Loss of all offsite power (blackout with natural circulation in the 40 Loss of Power Loss of Offsite Power Y Y RCS)

Partial loss of flow (loss of one 80 Partial Loss of Flow. Partial Loss of RCS Flow Y Y pump)

United States Nuclear Regulatory Commission Page 6 of 9 SBK-L-1 1002 / Enclosure 1 Reactor Trip from Full Power: -Reactor Trip from Full."

Reactor trip from full power: Reactor trip with no Power - with no Without cooldown inadvertent cooldown Inadvertent Cooldown Y Y 230 With cooldown, without Reactor trip with -Reactor Trip from Full Y Y 160 safety injection cooldown but no Power - with Cooldown With cooldown and safety safety injection and no SI -Reactor Trip Y Y injection 10 Reactor trip with from Full Power - with cooldown actuating Cooldown and SI (HHSI) safety injection Inadvertent reactor coolant Inadvertent Reactor Inadvertent RCS 20 Coolant System D ua depressurization Depressurization Depressurization Inadvertent Pressurizer Auxiliary Spray Y Y Actuation (5 Inadvertent startup of inactive loop 10 Inadvertent Startup of y y t san Inactive Loop Control rod drop 80 Control Rod Drop Y Y Inadvertent ECCS actuation 60 Inadvertent Safety Inadvertent Safety y y Injection Actuation Injection (SI) Actuation Operating Basis Earthquake (5 Operating Basis Operating Basis 50 N Y earthquakes of 10 cycles each) Earthquake Earthquake (OBE) 50 N Y

United States Nuclear Regulatory Commission Page 7 of 9 SBK-L-1 1002 / Enclosure 1 txcessive t eeawater Excessive Feedwater

-Excessive feedwater flow 30 Y Y Flow Flow RCS Cold RCS Cold Overpressurization 10 Overpressurization N Y Charging and Letdown Flow Charging and Letdown 60 Not Specified Flow Shutoff and Return N Y Shutoff and Return (6)

Charging Flow Shutoff with Charging Flow Shutoff N Y 20 Not Specified with Delayed Return Delayed Return (6)

Charging Flow Shutoff with 20 Not Specified Charging Flow Shutoff N Y Prompt Return (6) with Prompt Return Letdown Flow Shutoff with Letdown Flow Shutoff 20 Not Specified, with Delayed Return N Y Delayed Return (6)

Letdown Flow Shutoff with Letdown Flow Shutoff 200 Not Specified with Prompt Return N Y Prompt Return (6)

Em er]gency Transients: _____________ __________________-__ ____________________

Small LOCA Small Loss-of-Coolant Aciet5__________

Accident Small Steam Line Small steam break 5Break Complete loss of flow 5 Complete Loss of Flow Faulted *ransients: _____________ _____________________ _______________________

Reactor Coolant Pipe Main reactor coolant pipe break Break (Large Loss-of-(LOCA) Coolant Accident)

United States Nuclear Regulatory Commission Page 8 of 9 SBK-L- 11002 / Enclosure I Large Steam Line N N Large steam line break I Break Feedwater line break 1 Feedwater Line Break N N Reactor Coolant Pump locked Reactor Coolant Pump N N rotor Locked Rotor Control rod ejection 1 Control Rod Ejection N N Included under Reactor Steam Generator tube rupture Trip with Steam Generator Tube N N cooldown and Rupture safety injection Safe Shutdown Safe Shutdown Earthquake 1 Earthquake N N

ýTest TrAnsients:

Primary Side Primary Side RCSyy Primary side hydrostatic test 10 Prmr SiePiaySd Hydrostatic test Hydrostatic TestC Secondary Side y y Secondary side hydrostatic test 10 Hydrostatic Test Turbine roll test 20 Turbine Roll Test Turbine Roll Test Y Y Primary side leak test 200 Primary Side Leakage Primary Side RCS y y Test Leakage Test

.Secondary Seco r Side y y Secondary side leak test 80 Leakage Test Tube leak test 800 Tube Leakage Test Y Y

United States Nuclear Regulatory Commission Page 9 of 9 SBK-L-1 1002 / Enclosure 1

1. For the design transient of Unit Loading and Unit Unloading @ 5% full power/min., the Reactor Vessel, Steam Generators and Pressurizers are designed for 13,200 cycles, where the Class 1 piping is designed for 18,300 cycles. The most limiting value of these major components is used as a monitoring limit in the Metal Fatigue of Reactor Coolant Pressure Boundary Program (B.2.3.1).

2. For the design transients of Unit load and unload between 0% to 15% of full power, the Reactor Vessel, Steam Generators and Class 1 piping are designed for '500 cycles, where the Pressurizer is designed for 1,510 cycles. The most limiting value of these major components is used as a monitoring limit in the Metal Fatigue of Reactor Coolant Pressure Boundary Program (B.2.3.1).

3. For the design transient of Reactor Coolant Pump startup/shutdown, the limit specified in the UFSAR is 3800 cycles. The Pressurizer is designed for 4,000 cycles, where Steam Generators are designed for 3,000 cycles. The Steam Generators has have the mosf limiting value (3,000 cycles) of these components is and are used as a monitoring limit in the Metal Fatigue of Reactor Coolant Pressure Boundary Program (B.2.3.1).

4. Categorization of the Loop out of Service transient is taken from UFSAR' Section 3.9(N). 1.1 .a.7.

5. Inadvertent Pressurizer Auxiliary Spray Actuation transient specified in the PBD is one of the five subevents included in the Inadvertent Reactor Coolant System Depressurization event.

6. Transients identified as auxiliary transients in Westinghouse Systems Standard.

7. The Steady-State Fluctuation event does not contribute to the computed fatigue usage for any analyzed component and is not specifically counted.

8. The Boron Concentration Equalization event is a load-following event, and is not specifically counted.