SBK-L-09196, License Amendment Request to Revise Technical Specification (TS) Sections 6.7.6.k, Steam Generator (SG) Program and TS 6.8.1.7, Steam Generator Tube Inspection Report for One-Time Alternate Repair Criteria

License Amendment Request to Revise Technical Specification (TS) Sections 6.7.6.k, Steam Generator (SG) Program and TS 6.8.1.7, Steam Generator Tube Inspection Report for One-Time Alternate Repair Criteria

ML092720883
Person / Time
Site:	Seabrook
Issue date:	09/18/2009
From:	St.Pierre G NextEra Energy Seabrook
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
SBK-L-09196, TAC ME1386
Download: ML092720883 (20)

Text

NExTera ENERGY S--

SEABROOK September 18, 2009 SBK-L-09196 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 License Amendment Request to Revise Technical Specification (TS) Sections 6.7.6.k, "Steam Generator (SG) Program" and TS 6.8.1.7, "Steam Generator Tube Inspection Report" for One-Time Alternate Repair Criteria

References:

1. NextEra Energy Seabrook letter SBK-L-09118, License Amendment Request 09-03, Revision to Technical Specification 6.7.6.k, "Steam Generator (SG) Program," for Permanent Alternate Repair Criteria; May 28, 2009

2. NRC letter Seabrook Station, Unit No. 1 - Request for Additional Information Regarding Steam Generator Program (TAC NO. ME1386), August 13, 2009

3. NRC Letter Seabrook Station, Unit No. 1 - Second Request for Additional Information (RAI) Regarding Steam Generator Program (TAC NO. ME1386), September 1, 2009

4. NextEra Energy Seabrook letter SBK-L-09168, Response to Request for Additional Information Regarding Permanent H* Alternate Repair Criteria for Steam Generator Inspections, September 16, 2009 On May 28, 2009, NextEra Energy Seabrook, LLC (NextEra) submitted a license amendment request [Reference 1] to revise Seabrook Station Technical Specification (TS) 6.7.6.k, "Steam Generator (SG) Program," and TS 6.8.1.7, "Steam Generator Tube Inspection Report." The proposed changes would revise the inspection scope and repair requirements of TS 6.7.6.k, "Steam Generator (SG) Program", and the reporting requirements of TS 6.8.1.7, "Steam Generator Tube Inspection Report." The proposed changes would establish a permanent alternate repair criterion to exclude portions of the tube below the top of the steam generator tube sheet from periodic steam generator tube inspections. Westinghouse WCAP-17071-P, Revision 0, NextEra Energy Seabrook, LLC, P.O. Box 300, Lafayette Road, Seabrook, NH 03874 o^

U. S. Nuclear Regulatory Commission SBK-L-09196 / Page 2 "H*: Alternate Repair Criteria for the Tubesheet Expansion Region in Steam Generators with Hydraulically Expanded Tubes (Model F)," was submitted as Attachment 4 and provided the basis for the proposed change.

On August 13, 2009, NextEra received a request for additional information (RAI) letter

[Reference 2], which contained 24 questions. On September 1, 2009 NextEra Energy Seabrook received a second request for additional information [Reference 3], which revised questions 4, 21, 24, and added one additional question, number 25. On September 16, 2009 NextEra submitted Reference 4, which provided responses to questions 1 through 24 of the August 13, 2009 letter and revised questions 4, 21, and 24 and new question 25 of the September 1, 2009 letter.

On September 2, 2009, in a teleconference between the NRC staff and industry personnel, the NRC staff indicated that concerns with eccentricity of the tube sheet tube bore in normal and accident conditions (RAI question 4 of the August 13, 2009 letter and revised question 4 of the September 1, 2009 letter) have not been completely resolved to the satisfaction of the staff. The staff further indicated that there was insufficient time to resolve these issues to support approval of the permanent amendment request to support the fall 2009 refueling outage. As a result, NextEra proposes to revise the requested changes to the TS in Reference 1 to be one-time changes to TS 6.7.6.k and TS 6.8.1.7 during refueling outage 13 and the subsequent operating cycles until the next scheduled inspection of the steam generator tubing. NextEra requests that the staff provide the specific questions remaining to be resolved and that the review of the amendment request for permanent alternate repair criteria continue.

The permanent H* submittal is based on maintaining structural and leakage integrity in the event of an accident.

From a structural perspective, tube burst cannot occur within the thickness of the tubesheet, and the 13.1 inch value of H* ensures that tube pull out from the tube sheet will not occur during normal operation or under accident conditions. Even in the event that all tubes in the steam generator have a 360 degree sever at 13.1 inches, structural integrity of the steam generator tube bundle will be maintained. This assumption bounds the current status of the Seabrook Station steam generators with significant margin.

Seabrook Station has inspected portions of the steam generator tubing that contain overexpansions (OXP) and bulges (BLG) down to 17 inches below the top of the tube sheet The inspection of 50% of the OXP/BLG population during refueling outage 11 in the fall of 2006 found no degradation. In addition, inspections of the expansion transition at the top of the tube sheet found no degradation. This inspection program is in accordance with Seabrook Station technical specifications and industry guidance.

The tube ends have not been inspected at Seabrook Station. However, the tube end roll joint was created using a hydraulic (urethane) expansion rather than a hard roll. The hydraulic expansions are less susceptible to the initiation of cracking.

U. S. Nuclear Regulatory Commission SBK-L-09196 / Page 3 Based on inspections, no flaws exist in the tube sheet region of the Seabrook steam generators.

In addition, no indications of a 360 degree sever have been detected in any steam generator at Seabrook Station. Consequently, the level of degradation in the Seabrook Station steam generators is very limited compared to the assumption of "all tubes severed" that was utilized in the development of the permanent H*. Consequently, structural integrity will be assured for the operating period between inspections allowed by TS 6.7.6.k, "Steam Generator (SG) Program".

From a leakage perspective, projections of accident induced steam generator tube leakage are based on leakage rate factors applied to leakage detected during normal operation. The multiplication factor used for Seabrook Station bounds the expected increased leakage in the event of an accident at Seabrook Station. The projected accident induced leakage remains the same for both the one-time and permanent H* amendments. No primary-to-secondary steam generator tube leakage has been reported during the current operating cycle at Seabrook Station.

(The reporting threshold is one gallon per day.)

Significant margin exists between the current state of the Seabrook Station steam generators and the conservative assumptions used as the basis for the permanent H*. Structural and leakage integrity will continue to be assured for the operating period between inspections allowed by TS 6.7.6.k, "Steam Generator (SG) Program" for one-time H*.

The requested changes do not expand the scope of the application as originally noticed, and do not impact the conclusions of the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register (74 FR 35891).

NextEra requests approval of the proposed license amendment by October 1, 2009 to support the fall refueling outage. The proposed changes would be implemented within 30 days of issuance of the amendment.

Attachments 1 and 2 contain marked up and typed TS that show the proposed changes. The NRC commitments contained in this letter are included in Attachment 3. Should you have any questions regarding this letter, please contact Mr. Michael O'Keefe, Licensing Manager, at (603) 773-7745.

Sincerely, NextEra Energy Seabrook, LLC Gene St. Pierre Vice President North

U. S. Nuclear Regulatory Commission SBK-L-09196 / Page 4 Attachments:

1. Markup of Proposed Technical Specifications

2. Typed Pages of Proposed Technical Specifications

3. List of Regulatory Commitments cc:

S. J. Collins, NRC Region I Administrator D. L. Egan, NRC Project Manager R. B. Ennis, NRC Project Manager W. J. Raymond, NRC Resident Inspector 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

NEý xTera ENERGY74 SEABROOK The following information is enclosed in support of this License Amendment Request:

1. Markup of Proposed Technical Specifications

2. Typed Pages of Proposed Technical Specifications

3. List of Regulatory Commitments I, Gene St. Pierre, Vice President North of NextEra Energy Seabrook, LLC hereby affirm that the information and statements contained within this License Amendment Request 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' day ofA 4e-,

,2 009 Gene St. Pierre Vice President North Markup of the Proposed Technical Specifications

Mark-up of the Technical Specifications (TS)

Refer to the attached markup of the TS showing the proposed changes. The attached markups reflect the currently issued version of the TS and Facility Operating License.

At the time of submittal, the Facility Operating License was revised through Amendment No. 122.

Listed below are the license amendment requests that are awaiting NRC approval and may impact the currently issued version of the Facility Operating License affected by this LAR.

"LAR.,:

Title NextEiaEfig DateEeg Seabr o letter Submitted None The following TS pages are included in the attached markup:

Techtiical.-.

Page

'Specification Title TS 6.7.6.k Steam Generator (SG) Program 6-13 6-14 TS 6.8.1.7 Steam Generator Tube Inspection Report 6-21

INSERT 1 For refueling outage 13 and the subsequent inspection cycle, tubes with service-induced flaws located greater than 13.1 inches below the top of the tubesheet do not require plugging. Tubes with service-induced flaws located in the portion of the tube from the top of the tubesheet to 13.1 inches below the top of the tubesheet shall be plugged upon detection.

INSERT 2 For refueling outage 13 and the subsequent inspection cycle, the portion of the tube below 13.1 inches from the top of the tubesheet is excluded from this requirement.

INSERT 3 If crack indications are found in portions of the SG tube not excluded above, INSERT 4

i.

For refueling outage 13 and the subsequent inspection cycle, the primary to secondary leakage rate observed in each SG (if it is not practical to assign the leakage to an individual SG, the entire primary to secondary leakage should be conservatively assumed to be from one SG) during the cycle preceding the inspection which is the subject of the report,

j.

For refueling outage 13 and the subsequent inspection cycle, the calculated accident induced leakage rate from the portion of the tubes below 13.1 inches from the top of the tubesheet for the most limiting accident in the most limiting SG. In addition, if the calculated accident induced leakage rate from the most limiting accident is less than 2.50 times the maximum operational primary to secondary leakage rate, the report should describe how it was determined, and

k. For refueling outage 13, the results of monitoring for tube axial displacement (slippage). If slippage is discovered, the implications of the discovery and corrective action shall be provided.

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS 6.7.6 (Continued)

j.

Technical Specification (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.

a.

Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

b.

Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:

1.

A change in the TS incorporated in the license or

2.

A change to the updated FSAR (UFSAR) or Bases that requires NRC approval pursuant to 10 CFR 50.59.

c.

The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the UFSAR.

d.

Proposed changes that meet the criteria of Specification 6.7.6j.b above shall be reviewed and approved by the NRC prior to implementation.

Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

k. Steam Generator (SG) Pro-gram A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following provisions:

a.

Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged to confirm that the performance criteria are being met.

SEABROOK - UNIT 1 6-11 Amendment No. 34, 55, 67, 88, 104, 115

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS 6.7.6 (Continued)

b.

Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.

1.

Structural integrity performance criterion: All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down and all anticipated transients included in the design specification) and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensingbasis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

2.

Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. Leakage is not to exceed 1 gpm total or 500 gpd through any one SG.

3.

The operational LEAKAGE performance criterion is specified in LCO 3.4.6.2, "Reactor Coolant System Operational Leakage."

c.

Provisions for SG tube repair criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.

SEABROOK - UNIT 1 6-12 Amendment No. 34, 39, 104, 109, 115

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS 6.7.6 (Continued)

The following alternate tube repair criteria may be applied as an alternative to the 40% depth based criteria:

1.

Durina*.refuelin l..etage 11 and the subsequent operating cycles.*ntil the S6.*-

• eXt ch einspe i

, flaws; 0fifled in the3edrtion of th be below 1 e i n rhh

• pr from the

-of

• theh 1g tubeshee, 66not re ur luggi

,ng.

During eling outa e 11 and thesd sequent/,o rating cycl ntil t!)

ne chedulegdin.spection, all tubes with flawsidentified in.te portion"of the tube within the region from the top of the hotieg tubesheet to 17 inches

d.

Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair crteia._/.fD-u~pg-efueling outage 11 and the subsequen-t oeatn ut

• ex*t-sW5uld inspectiq0.*,,,thre'4portion of tý,tbeblw RJch's fromTlJ~top

.of'-

( -T*'-,

theiot leg tubes.eteis excluded fro nspection wherrthe alternate tube re air (EEý

• criteria in TS 6.,qr.6.k.c are impleme~pnteýd.J-The t"*"'r u'be-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. An assessment of degradation shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

I/

00, I

0-

1.

Inspect 100% of the tubes in each SG dur following SG replacement.

ing the first refueling outage

2.

Inspect 100% of the tubes at sequential periods of 120, 90, and, thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SG shall operate for more than 48 effective full power months or two refueling outages (whichever is less) without being inspected.

SEABROOK-UNIT 1 6-13 Amendment No. 34, 104, 1 09,Q)

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS 6.7.6 (Continued)

3. UIt.ck ations arwfoundfr'an y,6 tuGb then the next

_inspection-for each SG fo-rth-e degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.

e.

Provisions for monitoring operational primary to secondary leakage.

I. Control Room Envelope Habitability Proqram A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Emergency Makeup Air and Filtration System (CREMAFS), CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem total effective dose equivalent (TEDE) for the duration of the accident. The program shall include the following elements:

a.

The definition of the CRE and the CRE boundary.

b.

Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance.

c.

Requirements for (i) determining the unfiltered air in-leakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.

SEABROOK - UNIT 1 6-14 Amendment No. 34, 78, 104, 115, 119

ADMINISTRATIVE CONTROLS 6.8.1.6.c The core operating limits shall be determined so that all applicable limits (e.g.,

fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as SHUTDOWN MARGIN, and transient and accident analysis limits) of the safety analysis are met. The CORE OPERATING LIMITS REPORT for each reload cycle, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, to the NRC Document Control Desk with copies to the Regional Administrator and the Resident Inspector.

STEAM GENERATOR TUBE INSPECTION REPORT 6.8.1.7 A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with Specification 6.7.6.k, Steam Generator (SG) Program. The report shall include:

a.

The scope of inspections performed on each SG,

b.

Active degradation mechanisms found,

c.

Nondestructive examination techniques utilized for each degradation mechanism,

d.

Location, orientation (if linear), and measured sizes (if available) of service induced indications,

e.

Number of tubes plugged during the inspection outage for each active degradation mechanism,

f.

Total number and percentage of tubes plugged to date,

g.

The results of condition monitoring, including the results of tube pulls and in-situ testing,

h.

The effective plugging percentage for all plugging in each SG.

SPECIAL REPORTS 6.8.2 Special reports shall be submitted to the U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, Attn: Document Control Desk, with a copy to the NRC Regional Administrator within the time period specified for each report.

6.9 (THIS SPECIFICATION NUMBER IS NOT USED)

SEABROOK - UNIT I 6-21 Amendment No. 22, 66, 88, 104, 107, 115 Typed Pages of the Proposed Technical Specifications

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS 6.7.6 (Continued)

The following alternate tube repair criteria may be applied as an alternative to the 40% depth based criteria:

For refueling outage 13 and the subsequent inspection cycle, tubes with service-induced flaws located greater than 13.1 inches below the top of the tubesheet do not require plugging. Tubes with service-induced flaws located in the portion of the tube from the top of the tubesheet to 13.1 inches below the top of the tubesheet shall be plugged upon detection

d.

Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria. For refueling outage 13 and the subsequent inspection cycle, the portion of the tube below 13.1 inches from the top of the tubsheet is" excluded from this requirement. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. An assessment of degradation shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

1.

Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement.

2.

Inspect 100% of the tubes at sequential periods of 120, 90, and, thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SG shall operate for more than 48 effective full power months or two refueling outages (whichever is less) without being inspected.

SEABROOK - UNIT 1 6-13 Amendment No. 3, 104, 109,115, XXX

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS 6.7.6 (Continued)

3.

If crack indications are found in portions of the SG tube not excluded above, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.

e.

Provisions for monitoring operational primary to secondary leakage.

I. Control Room Envelope Habitability Program A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Emergency Makeup Air and Filtration System (CREMAFS), CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem total effective dose equivalent (TEDE) for the duration of the accident. The program shall include the following elements:

a.

The definition of the CRE and the CRE boundary.

b.

Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance.

c.

Requirements for (i) determining the unfiltered air in-leakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors,"

Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.

SEABROOK - UNIT 1 6-14 Amendment No. 34, 78, 104, 115, 119, xxx

ADMINISTRATIVE CONTROLS 6.8.1.6.c The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as SHUTDOWN MARGIN, and transient and accident analysis limits) of the safety analysis are met. The CORE OPERATING LIMITS REPORT for each reload cycle, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, to the NRC Document Control Desk with copies to the Regional Administrator and the Resident Inspector.

STEAM GENERATOR TUBE INSPECTION REPORT 6.8.1.7 A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with Specification 6.7.6.k, Steam Generator (SG) Program. The report shall include:

a.

The scope of inspections performed on each SG,

b.

Active degradation mechanisms found,

c.

Nondestructive examination techniques utilized for each degradation mechanism,

d.

Location, orientation (if linear), and measured sizes (if available) of service induced indications,

e.

Number of tubes plugged during the inspection outage for each active degradation mechanism,

f.

Total number and percentage of tubes plugged to date,

g.

The results of condition monitoring, including the results of tube pulls and in-situ testing,

h.

The effective plugging percentage for all plugging in each SG.

i.

For refueling outage 13 and the subsequent inspection cycle, the primary to secondary leakage rate observed in each SG (if it is not practical to assign the leakage to an individual SG, the entire primary to secondary leakage should be conservatively assumed to be from one SG) during the cycle preceding the inspection which is the subject of the report,

j.

For refueling outage 13 and the subsequent inspection cycle, the calculated accident induced leakage rate from the portion of the tubes below 13.1 inches from the top of the tubesheet for the most limiting accident in the most limiting SG. In addition, if the calculated accident induced leakage rate from the most limiting accident is less than 2.50 times the maximum operational primary to secondary leakage rate, the report should describe how it was determined, and SEABROOK - UNIT I 6-21 Amendment No. 22, 66, 88, 10, 107, 11-5, xxx

ADMINISTRATIVE CONTROLS 6.8.1.7 (Continued)

k.

For refueling outage 13, the results of monitoring for tube axial displacement (slippage).

If slippage is discovered, the implications of the discovery and corrective action shall be provided.

SPECIAL REPORTS 6.8.2 Special reports shall be submitted to the U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, Attn: Document Control Desk, with a copy to the NRC Regional Administrator within the time period specified for each report.

6.9 (THIS SPECIFICATION NUMBER IS NOT USED)

SEABROOK - UNIT I 6-21a Amendment No. 22, 66, 88, 104, 107, 115, xxx List of Regulatory Commitments

List of Regulatory Commitments The following table identifies those actions committed to by NextEra Energy Seabrook, LLC Seabrook in this document. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments. Please direct questions regarding these commitments to Mr. Michael O'Keefe, Licensing Manager.

Regulatory Commitment Due Date / Event

1. NextEra Energy Seabrook, LLC commits to perform a one-time verification of tube expansion locations to determine if any significant deviations exist from the top of tubesheet to the bottom of expansion transition (BET). If any significant deviations are found, the condition will be entered into the plant's corrective action program and dispositioned.

Additionally, if any significant deviations are found, those deviations will be reported to the NRC via the Steam Generator Tube Inspection Report or separate timely correspondence.

Prior to entering Mode 4 during startup following refueling outage 13 in the fall of 2009

2. NextEra Energy Seabrook, LLC commits to Each inspection of the monitor for tube slippage as part of the steam Seabrook Station Steam generator tube inspection program. Slippage Generators monitoring will occur for each inspection of the Seabrook Station Steam Generators.

3. For the Condition Monitoring assessment, the During each inspection component of operational leakage from the prior required by Technical cycle from below the H* distance will be Specification 6.7.6.k, "Steam multiplied by a factor of 2.50 and added to the Generator (SG) Program" total accident leakage from any other source and compared to the allowable accident induced leakage limit. For the Operational Assessment, the difference between the allowable accident induced leakage and the accident induced leakage from sources other than the tubesheet expansion region will be divided by 2.50 and compared to the observed operational leakage.

An administrative operational leakage limit will be established to not exceed the calculated value.