Seabrook Station - License Amendment Request 12-02, Request … · 2012. 7. 20. · TS 6.8.1.7 i....

NExTeraMi_ ENERGY ,,._ _SEABROOK

April 10, 2012

10 CFR 50.90

SBK-L-12072Docket No. 50-443

U. S. Nuclear Regulatory CommissionATTN: Document Control DeskWashington, DC 20555-0001

Seabrook Station

License Amendment Request 12-02

Request for Permanent Application of Steam Generator Tube Alternate Repair Criteria, H*

In accordance with the provisions of Section 50.90 of Title 10 of the Code of FederalRegulations (10 CFR), NextEra Energy Seabrook, LLC (NextEra) is submitting LicenseAmendment Request (LAR) 12-02 to revise the Seabrook Station Technical Specifications (TS).This LAR proposes a change to TS 6.7.6.k, Steam Generator (SG) Program, to exclude a portionof the tubes below the top of the SG tube sheet from periodic tube inspections and plugging. Theproposed change also establishes permanent reporting requirements in TS 6.8.1.7, SteamGenerator Tube Inspection Report, that were previously implemented on a temporary basis. Asimilar amendment was approved for Catawba Unit 2 on March 12, 2012.

Attachment 1 to this request provides NextEra's evaluation of the proposed changes, andAttachment 2 includes a markup of the TS showing the proposed changes. Attachments 3 and 4contain proprietary and non-proprietary versions, respectively, of Westinghouse document LTR-SGMMP- 11-28 Rev. 1, "Response to USNRC Request for Additional Information Regarding theLicense Amendment Requests for Permanent Application of the Alternate Repair Criterion, H*,to the Model D5 and Model F SGs." An errata letter that corrects two errors in LTR-SGMMP-11-28 Rev. 1 (Attachment 4) is included in Attachment 5. Attachment 6 contains aWestinghouse Application for Withholding Proprietary Information from Public Disclosure,CAW-12-3405; accompanying affidavit, proprietary information notice, and copyright notice.Attachment 7 provides a response to a request for additional information for questions specific toSeabrook Station model F SGs.

As Attachment 3 contains information proprietary to Westinghouse Electric Company LLC, it issupported by an affidavit signed by Westinghouse, the owner of the information. The affidavit

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

United States Nuclear Regulatory CommissionSBK-L-12072 / Page 2

sets forth the basis on which the information may be withheld from public disclosure by theCommission and addresses with specificity the considerations listed in paragraph (b)(4) of10 CFR 2.390. Accordingly, it is requested that the information that is proprietary toWestinghouse be withheld from public disclosure in accordance with 10 CFR 2.390.Correspondence with respect to the copyright or proprietary aspects of the document inAttachment 3 or the supporting Westinghouse affidavit should reference CAW-12-3405 andshould be addressed to J. A. Gresham, Manager, Regulatory Compliance, Westinghouse ElectricCompany, Suite 428, 1000 Westinghouse Drive, Cranberry Township, PA 16066.

As discussed in the enclosed evaluation, the proposed change does not involve a significanthazards consideration pursuant to 10 CFR 50.92, and there are no significant environmentalimpacts associated with the change.

The Station Operation Review Committee has reviewed this LAR. A copy of this LAR has beenforwarded to the New Hampshire State Liaison Officer pursuant to 10 CFR 50.91 (b).

Attachment 8 identifies commitments made as a result of this change.

NextEra requests NRC review and approval of LAR 12-02 with issuance of a license amendmentby September 17, 2012 to support refueling outage 15. The approved amendment will beimplemented prior to entering Mode 4 following the outage.

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

Paul FreemanSite Vice President

Enclosure

cc: NRC Region I AdministratorJ. G. Lamb, NRC Project ManagerW. J. Raymond, NRC Senior Resident Inspector

United States Nuclear Regulatory CommissionSBK-L-12072 / Page 3

Mr. Christopher M. Pope, Director Homeland Security and Emergency ManagementNew Hampshire Department of SafetyDivision of Homeland Security and Emergency ManagementBureau of Emergency Management33 Hazen DriveConcord, NH 03305

Mr. John Giarrusso, Jr., Nuclear Preparedness ManagerThe Commonwealth of MassachusettsEmergency Management Agency400 Worcester RoadFramingham, MA 01702-5399

ENERGYSEABROOK

AFFIDAVIT

SEABROOK STATION UNIT 1

Facility OeaigLicense NPF-86D)ocket No. 50-443

License Amendment Request 12-02

r Permanent Application of Ste-am Generator Tube Alternate Repair Criteri

The following information is enclosed in support of this License Amendment Request:

1. NextEra's evaluation of the proposed changes2. Markup of the TS3. LTR-SGMMP-1 1-28 Rev. 1 P-Attachment, "Response to USNRC Request for Additional

Information Regarding the License Amendment Requests for Permanent Application of theAlternate Repair Criterion, H*, to the Model D5 and Model F SGs" (Proprietary)

4. LTR-SGMMP- 11-28 Rev. 1 NP-Attachment, "Response to USNRC Request for AdditionalInformation Regarding the License Amendment Requests for Permanent Application of theAlternate Repair Criterion, H*, to the Model D5 and Model F SGs" (Non-Proprietary)

5. LTR-SGMMP- 11-28 Attachment Errata6. Application for Withholding Proprietary Information from Public Disclosure7. Response to Request for Additional Information Questions Specific to Seabrook Station Model F

Steam Generators8. List of Regulatory Commitments

I, Paul Freeman, Site Vice President of NextEra Energy Seabrook, LLC hereby affirm thatthe information and statements contained within this license amendment request are basedon facts and circumstances which are true and accurate to the best of my knowledge andbelief.

Sworn and Subscribedbefore me this

/0/j•day of _ 2olz,' 2,2012 g2 /Paul Freeman

Noar Site Vice President

Attachment 1

NextEra Energy Seabrook's Evaluation of the Proposed Change

Subject: Request for Permanent Application of Steam Generator Tube Alternate RepairCriteria, H*

1.0 SUMMARY DESCRIPTION

2.0 DETAILED DESCRIPTION

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria

4.2 Significant Hazards Consideration

4.3 Conclusion

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

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1.0 SUMMARY DESCRIPTION

NextEra Energy Seabrook, LLC (NextEra) proposes to revise Technical Specification(TS) 6.7.6.k, "Steam Generator (SG) Program," to permanently exclude portions of thesteam generator tubes below the top of the SG tubesheet from periodic steam generatortube inspections and repair. In addition, this amendment proposes to revise TS 6.8.1.7"Steam Generator Tube Inspection Report" to remove, reference to previous interimalternate repair criteria and provide reporting requirements specific to the permanentalternate repair criteria.

The proposed changes to the TS are based on the supporting structural analysis andleakage evaluation completed by Westinghouse Electric Company, LLC. Application ofthe supporting structural analysis and leakage evaluation results to exclude portions of thetubes from inspection and repair of tube indications constitutes a redefinition of theprimary to secondary pressure boundary. The documentation supporting theWestinghouse analysis is described in section 3.0 and provides the technical basis for thischange. Westinghouse Electric Company WCAP-7330-P, "H*: Resolution of NRCTechnical Issue Regarding Tubesheet Bore Eccentricity," Revision 1, June 2011,[Reference 8] Table 5-1, provides the 95/95 whole plant H* value of 15.21 inches forplants with Model F steam generators, which includes Seabrook.

The NRC has previously issued the following amendments revising the SG tubeinspection and repair requirements:

" Amendment Number 112 [Reference 1] to exclude degradation found in theportion of the tubes below 17 inches from the top of the hot leg tubesheetfrom the requirement to plug for refueling outage 11 and the subsequentoperating cycles.

* Amendment Number 123 [Reference 2] revised TS 6.7.6.k "Steam Generator(SG) Program," to exclude portions of the tubes within the tubesheet fromperiodic SG inspections (establish alternate repair criteria). In addition, thisamendment revised TS 6.8.1.7, "Steam Generator Tube Inspection Report," toprovide reporting requirements specific to refueling outage 13 and theinspection required by TS 6.7.6.k.d.

NextEra requests approval of this amendment application by September 17, 2012 tosupport refueling outage 15 (fall 2012), since the existing amendment expires at the endof the current operating cycle.

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2.0 DETAILED DESCRIPTION

Proposed changes to the current TS are shown below. Attachment 2 provides a markupof the TS showing the proposed changes.

TS 6.7.6. k. c

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

The following alternate tube repair criteria shall be applied as an alternative to the40% depth based criteria:

For- refueling ut.age 13 and the subsequent inspeti.n cycle,- tTubes withservice-induced flaws located greater than --34 15.21 inches below the top of thetubesheet do not require plugging. Tubes with service-induced flaws locatedin the portion of the tube from the top of the tubesheet to 4-34 15.21 inches belowthe top of the tubesheet shall be plugged upon detection.

TS 6.7.6. k. d

Provisions for SG tube inspections. Periodic SG tube inspections shall beperformed. The number and portions of the tubes inspected and methods ofinspection shall be performed with the objective of detecting flaws of any type(e.g., volumetric flaws, axial and circumferential cracks) that may be present alongthe 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 repaircriteria. F-r rIn- g outage 13 and the subsequent inspection .cyle, tThe portionof the tube below -34 15.21 inches from the top of the tubesheet is excluded from thisrequirement. The tube-to-tubesheet weld is not part of the tube. In addition tomeeting 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 SGtube integrity is maintained until the next SG inspection. An assessment ofdegradation shall be performed to determine the type and location of flaws towhich the tubes may be susceptible and, based on this assessment, to determinewhich inspection methods need to be employed and at what locations.

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TS 6.8.1.7

i. F.r .r .efuieling outage 13 and the subsequent inspe.ti.n cycle, tThe primary tosecondary leakage rate observed in each SG (if it is not practical to assign theleakage to an individual SG, the entire primary to secondary leakage should beconservatively assumed to be from one SG) during the cycle preceding theinspection which is the subject of the report,

j. For refuelin and the subsequent inspe.ti. n ycle, t The calculatedaccident induced leakage rate from the portion of the tubes below 4-31 15.21inches from the top of the tubesheet for the most limiting accident in the mostlimiting SG. In addition, if the calculated accident induced leakage rate from themost limiting accident is less than 2-.50 2.49 times the maximum operationalprimary to secondary leakage rate, the report should describe how it wasdetermined, and

k. For refrefingo-utage, 1,,The results of monitoring for tube axial displacement(slippage). If slippage is discovered, the implications of the discovery andcorrective action shall be provided.

3.0 TECHNICAL EVALUATION

3.1 Background

Seabrook Station is a four loop Westinghouse designed plant with Model F SGs having5626 tubes in each SG. A total of 173 tubes are currently plugged in all four SG. Thedesign of the SG includes Alloy 600 thermally treated tubing, full depth hydraulicallyexpanded tubesheet joints, and stainless steel tube support plates with broached holequatrefoils.

The steam generator inspection scope is governed by TS 6.7.6.k, Steam Generator (SG)Program; Nuclear Energy Institute (NEI) 97-06, Steam Generator Program Guidelines[Reference 3]; EPRI 1003138, Pressurized Water Reactor Steam Generator ExaminationGuidelines [Reference 4]; EPRI 1012987, Steam Generator Integrity AssessmentGuidelines [Reference 5]; Seabrook Station "Steam Generator Management ReferenceManual;" and the results of the degradation assessments required by the SG Program.Criterion IX, "Control of Special Processes" of 10 CFR Part 50, Appendix B, requires inpart that nondestructive testing be accomplished by qualified personnel using qualifiedprocedures in accordance with the applicable criteria. The inspection techniques andequipment are capable of reliably detecting known and potential specific degradationmechanisms applicable to Seabrook. The inspection techniques, essential variables andequipment are qualified to Appendix H, "Performance Demonstration for Eddy CurrentExamination" of the EPRI Steam Generator Examination Guidelines.

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Catawba Nuclear Station, Unit 2, (Catawba) reported indication of cracking followingnondestructive eddy current examination of the SG tubes during their fall 2004 outage.NRC Information Notice (IN) 2005-09, "Indications in Thermally Treated Alloy 600Steam Generator Tubes and Tube-to-Tubesheet Welds" [Reference 6], provided industrynotification of the Catawba issue. IN 2005-09 noted that Catawba reported crack-likeindications in the tubes approximately seven inches below the top of the hot leg tubesheetin one tube, and just above the tube-to-tubesheet welds in a region of the tube known asthe tack expansion in several other tubes. Indications were also reported in the tube-endwelds, also known as tube-to-tubesheet welds, which join the tube to the tubesheet.

NextEra policies and programs, as well as TS 6.7.6.k, require the use of applicableindustry operating experience in the operation and maintenance of Seabrook. Theexperience at Catawba, as noted in IN 2005-09, shows the importance of monitoring alltube locations (such as bulges, dents, dings, and other anomalies from the manufacture ofthe SGs) with techniques capable of finding potential forms of degradation that may beoccurring at these locations (as discussed in Generic Letter 2004-001, "Requirements forSteam Generator Tube Inspections"). Since the Seabrook Westinghouse Model F SGswere fabricated with Alloy 600 thermally treated tubes similar to the Catawba Unit 2Westinghouse Model D5 SGs, a potential exists for Seabrook to identify tube indicationssimilar to those reported at Catawba within the hot leg tubesheet region if similarinspections are performed.

Potential inspection plans for the tubes and tube welds underwent intensive industrydiscussions in March 2005. The findings in the Catawba SG tubes present three distinctissues with regard to the SG tubes at Seabrook:

1) Indications in internal bulges and over expansions within the hot leg tubesheet,

2) Indications at the elevation of the tack expansion transition, and

3) Indications in the tube-to-tubesheet welds and propagation of these indicationsinto adjacent tube material.

Prior to each SG tube inspection, a degradation assessment, which includes a review ofoperating experience, is performed to identify degradation mechanisms that have apotential to be present in the Seabrook SGs. A validation assessment is also performed toverify that the eddy current techniques utilized are capable of detecting those flaw typesthat are identified in the degradation assessment. Based on the Catawba operatingexperience, Seabrook revised the SG inspection plan for the fall 2006 refueling outage(OR1 1) to include sampling of bulges and over expansions within the tubesheet regiondown to 17 inches from the top of the tubesheet on the hot leg side. The sample wasbased on the guidance contained in EPRI 1003138, "Pressurized Water Reactor SteamGenerator Examination Guidelines," Revision 7; and TS 6.7.6.k, Steam Generator (SG)Program. According to EPRI SG examination guidelines, the inspection plan is expandedif necessary due to confirmed degradation in the region required to be examined (i.e., a

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tube crack). Primary Water Stress Corrosion Cracking (PWSCC) was not detectedduring the OR11 examination.

In OR13, fall 2009, the SG inspection plan included a sampling plan of bulges and overexpansions within the tubesheet region on the hot leg side down to the depth of 13.1inches below the top of the tubesheet. PWSCC was not detected in the tubesheet regionin OR13.

In OR14, spring 2011, the SG inspection was limited to the top to tubesheet expansiontransition based on finding a single indication of axial outside diameter stress corrosioncracking in OR13. There were no crack indications detected in the OR14 limitedinspection.

As a result of these potential issues and to prevent unnecessarily plugging tubes in theSeabrook SGs, NextEra is proposing changes to TS 6.7.6.k to limit the steam generatortube inspection and repair (plugging) to the safety significant portion of the tubes.

3.2 Licensing Basis Analysis (H*Analysis)

On May 28, 2009, Westinghouse WCAP-17071-P, Revision 0, "H*: Alternate RepairCriteria for the Tubesheet Expansion Region in Steam Generators with HydraulicallyExpanded Tubes (Model F)," [Reference 9] was submitted as Attachment 4 to NextEra'srequest to change TS 6.7.6.k, "Steam Generator (SG) Program" to supportimplementation of a permanent alternate repair criterion for SG tubes [Reference 17]. OnAugust 13, 2009, NextEra received a request for additional information (RAI) [Reference14], which contained 24 questions. On September 1, 2009, NextEra received a secondRAI letter [Reference 15], which clarified previously received RAI #4, #21, and #24, andadded RAI #25.

On September 16, 2009 [Reference 18], NextEra provided the provided responses toquestions 1 through 25 of the August 13, 2009 and September 1, 2009 letters andincluded the following documents:

" Westinghouse letter LTR-SGMP-09-100 P-Attachment, Revision 0, "Response toNRC Request for Additional Information on H*; Model F and Model D5 SteamGenerators," [Reference 10], and

* Westinghouse letter SGMP-09-109-P Attachment, Revision 0 "Response to NRCRequest for Additional Information on H*; RAI #4; Model F and Model D5Steam Generators," [Reference 11]

On September 18, 2009, NextEra submitted a request [Reference 19] to revise thepermanent alternate repair criteria amendment request to a one-time change applicable torefueling outage 13 and the subsequent inspection cycle. This request was made in

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response to a September 2, 2009 teleconference between NRC staff and industrypersonnel, in which the NRC staff indicated that their concerns with eccentricity of thetube sheet tube bore in normal and accident conditions (RAI question 4 of the September1, 2009 letter) had not been resolved.

On December 23, 2009, the NRC provided a letter [Reference 16] documenting theunresolved issues relating to tubesheet bore eccentricity. The letter contained 14questions that required resolution before the NRC could complete its review of apermanent amendment request.

The following documents have been prepared by Westinghouse to provide finalresolution of the remaining questions identified in the December 23, 2009 NRC letter insupport of the permanent H* amendment for Seabrook Station.

" WCAP-17330-P, Rev. 1, "H*: Resolution of NRC Technical Issue RegardingTubesheet Bore Eccentricity (Model F/Model D5) [Reference 8]

* LTR-SGMP- 10-78 P-Attachment, "Effects of Tubesheet Bore Eccentricity andDilation on Tube-to-Tubesheet Contact Pressure and Their Relative Importance toH*," [Reference 13]. This document, which is applicable to Seabrook Model F SGs,was transmitted to the NRC by Westinghouse letter LTR-NRC- 10-68 on November 9,2010.

* LTR-SGMP-09-1 11 P-Attachment, Rev. 1, "Acceptable Value of the Location of the •Bottom of the Expansion Transition (BET) for Implementation of H*," [Reference20] was prepared to support plant determinations of BET measurements and theirsignificant deviation assessment. This document, which is applicable to SeabrookModel F SGs, was transmitted to the NRC by Westinghouse letter LTR-NRC- 10-69on November 10, 2010.

" LTR-SGMP-10-33 P-Attachment, "H* Response to NRC Questions RegardingTubesheet Bore Eccentricity," [Reference 12]. This document, which is applicable toSeabrook Model F SGs, was transmitted to the NRC by Westinghouse letter LTR-NRC- 10-70 on November 11, 2010.

Note that WCAP-17330-P, Rev. 1, "H*: Resolution of NRC Technical Issue RegardingTubesheet Bore Eccentricity (Model F/Model D5), June 2011, makes reference toRevision 2 of WCAP- 17071-P and Revision 1 of LTR-SGMP-09-100 P-Attachment. Asdescribed above, NextEra has previously submitted Revision 0 of these documents.Revisions I and 2 of WCAP-17071-P and Revision 1 of LTR-SGMP-09-100 P-Attachment were created to resolve editorial comments. The technical informationcontained in WCAP- 17071 -P, Revision 0 and LTR-SGMP-09-100 P-Attachment,Revision 0, remains valid and provides part of the licensing basis for the requestedamendment.

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A summary of the H* licensing bases documents is provided in Table 1 below.

Table 1

Summary of H* Licensing Bases Documents

DocumentNumber

RevisionNumber

Title ReferenceNumber

WCAP-17071-P

LTR-SGMP-09-100 P-Attachment

LTR -SGMP-09-109-P Attachment

WCAP-17330-P

LTR-SGMP- 10-78P-Attachment

LTR-SGMP- 10-33

P-Attachment

LTR-SGMP- 11-58

LTR-SGMMP-l 1-28 P-Attachment

0 H*: Alternate Repair Criteria for the Tubesheet ExpansionRegion in Steam Generators with Hydraulically ExpandedTubes (Model F)"

0 Response to NRC Request for Additional Information onH*; Model F and Model D5 Steam Generators

0 Response to NRC Request for Additional Information onH*; RAI #4; Model F and Model D5 Steam Generators

I 1H*: Resolution of NRC Technical Issue RegardingTubesheet Bore Eccentricity

0 Effects of Tubesheet Bore Eccentricity and Dilation onTube-to-Tubesheet Contact Pressure and Their RelativeImportance to H*

0 H* Response to NRC Questions Regarding Tubesheet

Bore Eccentricity

0 WCAP-17330-P, Revision 1 Erratum

1 Response to USNRC Request for Additional InformationRegarding the License Amendment Requests forPermanent Application of the Alternate Repair Criterion,H*, to the Model D5 and Model F SGs

9

10

11

8

13

12

21

32

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In addition, the following industry correspondence is applicable to NextEra's request forapplication of permanent alternate repair criteria:

* A March 28, 2011 letter from the NRC to Southern Nuclear Operating Company[Reference 22] documented the summary of a February 16, 2011 public meetingregarding SG tube inspection permanent alternate repair criteria. Enclosure 3 of theNRC letter provided technical NRC Staff questions developed at the meeting.Responses to these questions have been incorporated into WCAP-17330-P,Revision 1.

* Section 1.3 of WCAP-17330-P, Revision 1 identifies revisions to the report to addressrecommendations from the independent review of the H* analysis performed by MPRAssociates. Related to the independent review, a May 26, 2011 letter from the NRCto Southern Nuclear Company [Reference 23] included a pre-submittal review requestfor additional information. The response to the NRC request is provided in SouthernNuclear Operating Company letter NL- 11-1178 [Reference 24].

* On June 30, 2011, Duke Energy Letter "Duke Energy Carolina (Duke Energy)Catawba Nuclear Station, Units 1 and 2, Docket Numbers 50-413 and 50-414,Proposed Technical Specification (TS) Amendment, TS 3.4.13, "RCS OperationalLeakage," TS 5.5.9, "Steam Generator (SG) Program," TS 5.6.8 "Steam Generator(SG) Tube Inspection Report," License Amendment Request to Revise TS forPermanent Alternative Repair Criteria [Reference 25] was submitted to the NRC. OnJuly 11, 2011, a supplement [Reference 26] to the license amendment request wassubmitted, which provided Westinghouse Electric Company LLC LTR-SGMP-1 1-58,"WCAP-17330-P, Revision 1 Erratum" [Reference 21].

* On November 15, 2011, the NRC transmitted via email a Draft Request forAdditional Information Regarding the Steam Generator License Amendment toRevise Technical Specifications for Permanent Alternate Repair Criteria. On January5, 2012 the NRC issued the Request for Additional Information Regarding the SteamGenerator License Amendment to Revise Technical Specifications for PermanentAlternate Repair Criteria [Reference 27].

• On January 12, 2012, Duke Energy submitted responses to the November 15, 2011Request for Additional Information [Reference 28].

" Subsequent to the Duke Energy license amendment request, Virginia Electric andPower Company (Dominion) submitted a license amendment request [Reference 29]for permanent application of the alternate repair criterion, H*, for Surry PowerStation Units 1 and 2. On January 18, 2012, the NRC issued a request for additionalinformation [Reference 30]. Dominion responded to the request for additionalinformation on February 14, 2012 [Reference 31 ].

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Westinghouse Electric Company LLC, LTR-SGMMP- 11-28 Rev. 1 P-Attachment,"Response to USNRC Request for Additional Information Regarding the LicenseAmendment Requests for Permanent Application of the Alternate Repair Criterion, H*,to the Model D5 and Model F SGs," [Reference 32], augments the responses to the DukeEnergy request for additional information to include similar responses applicable toModel F steam generators. Additionally, LTR-SGMMP-I 1-28 Rev. 1 P-Attachmentaddresses the Dominion request for additional information question 14 for the Model Fsteam generators. Attachment 7 provides Seabrook Station specific responses toquestions 12 and 13 from the Duke Energy request for additional information andquestion 15 from the Dominion request for additional information.

3.3 Evaluation

To preclude unnecessarily plugging tubes in the Seabrook SGs, tube inspections will belimited to identifying and plugging degradation in the portion of the tube within thetubesheet necessary to maintain structural and leakage integrity in both normal andaccident conditions. The technical evaluation for the inspection and repair methodologyis provided in the H* analysis as described above. This evaluation is based on the use offinite element model structural analysis and a bounding leak rate evaluation based oncontact pressure between the tube and the tubesheet during normal and postulatedaccident conditions. The limited tubesheet inspection criteria were developed for thetubesheet region of the Seabrook Model F SG considering the most stringent loadsassociated with plant operation, including transients and postulated accident conditions.The limited tubesheet inspection criteria were selected to prevent tube pull out from thetubesheet due to axial end cap loads acting on the tube and to ensure that the accidentinduced leakage limits are not exceeded. The H* analysis provides technical justificationfor limiting the inspection in the tubesheet expansion region to less than the full depth ofthe tubesheet.

The basis for determining the portion of the tube that requires eddy current inspectionwithin the tubesheet is based upon evaluation and testing programs that quantified thetube-to-tubesheet radial contact pressure for bounding plant conditions as described in theH* analysis. The tube-to-tubesheet radial contact pressure provides resistance to tubepull out.

Primary-to-secondary leakage from tube degradation in the tubesheet area is assumed tooccur in several design basis accidents: feedwater line break (FLB), steam line break(SLB), locked rotor, and control rod ejection. The radiological dose consequencesassociated with this assumed leakage are evaluated to ensure that they remain withinregulatory limits (e.g. 10 CFR Part 100, 10 CFR 50.67, GDC 19). The accident inducedleakage performance criteria are intended to ensure the primary-to-secondary leak rateduring any accident does not exceed the primary-to-secondary leak rate assumed in theaccident analysis. Radiological dose consequences define the limiting accident for theH* analysis.

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The constraint that is provided by the tubesheet precludes tube burst for cracks within thetubesheet. The criteria for tube burst described in NEI 97-06 and NRC Regulatory Guide(RG) 1.121, "Bases for Plugging Degraded PWR Steam Generator Tubes," [Reference 7]are satisfied due to the constraint provided by the tubesheet. Through application of thelimited tubesheet inspection scope as described below, the existing operating leakagelimit provides assurance that excessive leakage (i.e., greater than accident analysisassumptions) will not occur. The assumed accident induced leak rate limit is 500 gallonsper day (gpd) through the faulted steam generator and 940 gpd for the remaining threeSGs for SLB. The accident induced leak rate is 1 gallon per minute (gpm) for the rodejection and locked rotor events. Thus, the limiting accident is SLB. Based upon thelimiting leak rate factor of 2.49, an operational leak rate of less than 200 gpd would berequired to prevent exceeding the assumed accident induced leak rate limit of 500 gpd.Therefore, the TS leak rate limit of 150 gpd provides significant margin against the 500gpd accident analysis leak rate assumption.

Plant-specific operating conditions are used to generate the overall leakage factor ratiosthat are used in the condition monitoring and operational assessments. The plant-specificdata provide the initial conditions for application of the transient input data. The resultsof the analysis of the plant-specific inputs to determine the bounding plant for each modelof SG and to assure that the design basis accident contact pressures are greater than thenormal operating pressure contact pressure are contained in section 6 of WCAP- 17071-P(Reference 9). As discussed in WCAP-17330-P, the leak rate ratio (accident inducedleak rate to operational leak rate) is a product of the pressure differential subfactor andthe viscosity subfactor using the Darcy flow equation.

The plant transient response following a full power double-ended main feedwater linerupture corresponding to "best estimate" initial conditions and operating characteristics asgenerally presented in the Updated Safety Analysis Report (UFSAR) Chapter 15.0 safetyanalysis, indicates that the transient for a Model F SG exhibits a cooldown characteristicinstead of a heatup transient. The use of either the component design specificationtransient or the Chapter 15.0 safety transient for leakage analysis for FLB is overlyconservative because:

* The assumptions on which the FLB design transient is based are specifically intendedto establish a conservative structural (fatigue) design basis for RCS components;however, H* does not involve component structural and fatigue issues. The bestestimate transient is considered more appropriate for use in the H* leakagecalculations.

* For the Model F SG, the FLB transient curve [Figure 9-5, Reference 10] represents adouble-ended rupture of the main feedwater line concurrent with both stationblackout (loss of main feedwater and reactor coolant pump coast down) and turbinetrip.

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The assumptions on which the FLB safety analysis is based are specifically intendedto establish a conservative basis for minimum auxiliary feedwater (AFW) capacityrequirements and combines worst case assumptions, which are exceptionally moresevere when the FLB occurs inside containment. For example, environmental errorsthat are applied to reactor trip and engineered safety feature actuation would nolonger be applicable. This would result in much earlier reactor trip and greatlyincrease the SG liquid mass available to provide cooling to the RCS.

A SLB event would have similarities to a FLB except that the break flow path wouldinclude the secondary separators, which could only result in an increased initial cooldown(because of retained liquid inventory available for cooling) when compared to the FLBtransient. A SLB could not result in more limiting temperature conditions than a FLB.

In accordance with plant operating procedures, the operator would take action followinga high energy secondary line break to stabilize the RCS conditions. The expectation for aSLB or FLB with credited operator action is to stop the system cooldown throughisolation of the faulted SG and control of temperature by the AFW System. Steampressure control would be established by either the SG safety valves or control system(atmospheric relief valves). For any of the steam pressure control operations, themaximum temperature would be approximately the no-load temperature and would bewell below normal operating temperature.

Since the best estimate FLB transient temperature would not be expected to exceed thenormal operating temperature, the viscosity ratio for the FLB transient can be set to 1.0.Therefore, the leak rate factor would only be a function of the increase in pressuredifferential during the design basis SLB/FLB. However, per Reference 10, the FLBtransient was evaluated as a heatup event. Since dynamic viscosity decreases with theincrease in temperature during a postulated FLB event, the viscosity subfactor increasesabove 1.0. For Seabrook, the resulting leak rate ratio for both the SLB and FLB events isconservatively determined to be 2.49.

The other design basis accidents, such as the postulated locked rotor event and the controlrod ejection event, are conservatively modeled using the design specification transientsthat result in increased temperatures in the SG hot and cold legs for a period of time. Aspreviously noted, dynamic viscosity decreases with increasing temperature. Therefore,leakage would be expected to increase due to decreasing viscosity and increasingdifferential pressure for the duration of time that there is a rise in RCS temperature. Fortransients other than a SLB and FLB, the length of time that a plant with model F SGswill exceed the normal operating differential pressure across the tubesheet is less than 30seconds. As the accident induced leakage performance criteria is defined in gallons perminute, the leak rate for a locked rotor event can be integrated over a minute forcomparison to the limit. Time integration permits an increase in acceptable leakageduring the time of peak pressure differential by approximately a factor of two because ofthe short duration (less than 30 seconds) of the elevated pressure differential. This

12

translates into an effective reduction in the leakage factor by the same factor of two forthe locked rotor event. Therefore, for the locked rotor event, the leakage factor of 1.77[Table 9-7, Reference 10] for Seabrook is adjusted downward to a factor of 0.89.Similarly, for the control rod ejection event, the duration of the elevated pressuredifferential is less than 10 seconds. Thus, the peak leakage factor is reduced by a factorof six, from 2.65 to 0.44. Due.to the short duration of the transients above NOPdifferential, no leakage factor is required for the locked rotor and control rod ejectionevents (i.e., the leakage factor is under 1.0 for both transients).

For the Condition Monitoring (CM) assessment, the component of leakage from the priorcycle from below the H* distance will be multiplied by a factor of 2.49 and added to thetotal leakage from any other source and compared to the allowable accident inducedleakage limit. For the Operational Assessment (OA), the difference between theallowable leakage and the accident induced leakage from sources other than the tubesheetexpansion region will be divided by 2.49 and compared to the observed operationalleakage.

Reference 17 redefined the primary pressure boundary. The tube to tubesheet weld nolonger functions as a portion of this boundary. The hydraulic expansion of the tube intothe tubesheet over the H* distance now functions as the primary pressure boundary in thearea of the tube and tubesheet, maintaining the structural and leakage integrity over thefull range of SG operating conditions, including the most limiting accident conditions.The evaluation in Reference 17 determined that degradation in tubing below this safetysignificant portion of the tube does not require inspection or repair (plugging). Theinspection of the safety significant portion of the tubes provides a high level ofconfidence that the structural and leakage performance criteria are maintained duringnormal operating and accident conditions.

WCAP-17071 -P, section 9.8, provides a review of leak rate susceptibility to tube slippageand concluded that the tubes are fully restrained against motion under very conservativedesign and analysis assumptions such that tube slippage is not a credible event for anytube in the bundle. NextEra committed to monitor for tube slippage as part of the steamgenerator tube inspection program in OR13. This commitment remains in place for thepermanent license amendment.

As a condition for approving NextEra's Amendment Number 123 [Reference 2], theNRC required a commitment to measure the location of the bottom of the expansiontransition (BET) relative to the top of the tubesheet (TTS) and report any significantdeviations from the constant 0.3 inch value already included in the calculated value(s) ofH*. LTR-SGMP-09-1 11 P-Attachment, Rev. 1, "Acceptable Value of the Location of theBottom of the Expansion Transition (BET) for Implementation of H*" [Reference 20]was prepared to support plant determinations of BET measurements and their significantdeviation.assessment. LTR-SGMP-09- 111 P-Attachment was submitted to the NRC byWestinghouse Electric Company, LLC in letter LTR-NRC-10-69 on November 10, 2010.

13

Based on data review and Reference 20, NextEra did not identify any significantdeviations from the top of the tubesheet to the BET.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria

General Design Criteria (GDC) 1, 2, 4, 14, 30, 31, and 32 of 10 CFR 50, Appendix A,define requirements for the reactor coolant pressure boundary (RCPB) with respect tostructural and leakage integrity.

GDC 19 of 10 CFR 50, Appendix A, defines requirements for the control room and forthe radiation protection of the operators working within it. Accidents involving theleakage or burst of SG tubing comprise a challenge to the habitability of the controlroom.

10 CFR 50, Appendix B, establishes quality assurance requirements for the design,construction, and operation of safety related components. The pertinent requirements ofthis appendix apply to all activities affecting the safety related functions of thesecomponents. These requirements are described in Criteria IX, XI, and XVI of AppendixB and include control of special processes, inspection, testing, and corrective action.

10 CFR 100, Reactor Site Criteria, establishes reactor site criteria, with respect to the riskof public exposure to the release of radioactive fission products. Accidents involvingleakage or tube burst of SG tubing may comprise a challenge to containment andtherefore involve an increased risk of radioactive release.

10 CFR 50.67, Accident Source Term, establishes limits on the accident source term usedin design basis radiological consequence analyses with regard to radiation exposure tomembers of the public and to control room occupants.

Under 10 CFR 50.65, the Maintenance Rule, licensees classify SGs as risk significantcomponents because they are relied upon to remain functional during and after designbasis events. SGs are to be monitored under 10 CFR 50.65(a) (2) against industryestablished performance criteria. Meeting the performance criteria of NEI 97-06,Revision 3, provides reasonable assurance that the SG tubing remains capable offulfilling its specific safety function of maintaining the reactor coolant pressure boundary.The NEI 97-06, Revision 3, SG performance criteria are:

All in-service SG tubes shall retain structural integrity over the full range of normaloperating conditions (including startup, operation in the power range, hot standby,cool down, and all anticipated transients included in the design specification) anddesign basis accidents. This includes retaining a safety factor of 3.0 against burstunder normal steady state full power operation primary-to-secondary pressure

14

differential and a safety factor of 1.4 against burst applied to the design basis accidentprimary-to-secondary pressure differentials. Apart from the above requirements,additional loading conditions associated with the design and licensing basis shall alsobe evaluated to determine if the associated loads contribute significantly to burst orcollapse. In the assessment of tube integrity, those loads that do significantly affectburst or collapse shall be determined and assessed in combination with the loads dueto pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axialloads.

" The primary-to-secondary accident induced leakage rate for any design basisaccident, other than a SG tube rupture, shall not exceed the leakage rate assumed inthe accident analysis in terms of total leakage rate for all SGs and leakage rate for anindividual SG. Leakage is not to exceed 1 gpm per SG, except for specific types ofdegradation at specific locations when implementing alternate repair criteria asdocumented in the Steam Generator Program technical specifications.

" The RCS operational primary-to-secondary leakage through any one SG shall belimited to 150 gallons per day.

The safety significant portion of the tube is the length of tube that is engaged in thetubesheet from the secondary face that is required to maintain structural and leakageintegrity over the full range of SG operating conditions, including the most limitingaccident conditions. The evaluation in the H* analysis determined that degradation intubing below the safety significant portion of the tube does not require plugging andserves as the bases for the SG tube inspection program. As such, the Seabrook Stationinspection program provides a high level of confidence that the structural and leakagecriteria are maintained during normal operating and accident conditions.

4.2 No Significant Hazards Consideration

This amendment application proposes to revise Technical Specification (TS) 6.7.6.k,"Steam Generator (SG) Program," to exclude portions of the tubes within the tubesheetfrom periodic inspections and plugging. In addition, this amendment proposes torevise TS 6.8.1.7, "Steam Generator Tube Inspection Report" to provide reportingrequirements specific to the permanent alternate repair criteria. Application of thestructural analysis and leak rate evaluation results, to exclude portions of the tubesfrom inspection and repair, constitutes a redefinition of the primary-to-secondarypressure boundary.

The proposed change defines the safety significant portion of the tube that must beinspected and repaired. A justification has been developed by Westinghouse ElectricCompany, LLC to identify the specific inspection depth below which any type of axialor circumferential primary water stress corrosion cracking can be shown to have no

15

impact on Nuclear Energy Institute (NEI) 97-06, "Steam Generator ProgramGuidelines," performance criteria.

NextEra has evaluated whether or not a significant hazards consideration is involvedwith the proposed amendment by focusing on the three standards set forth in 10 CFR50.92, "Issuance of amendment," as discussed below:

1. The proposed changes do not involve a significant increase in the probabilityor consequences of an accident previously evaluated

Response: No

The previously analyzed accidents are initiated by the failure of plantstructures, systems, or components. The proposed change that alters the steamgenerator (SG) inspection and reporting criteria does not have a detrimentalimpact on the integrity of any plant structure, system, or component thatinitiates an analyzed event. The proposed change will not alter the operationof, or otherwise increase the failure probability of any plant equipment thatinitiates an analyzed accident.

Of the applicable accidents previously evaluated, the limiting transients withconsideration to the proposed change to the SG tube inspection and repaircriteria are the steam generator tube rupture (SGTR) event, the steam linebreak (SLB), and the feed line break (FLB) postulated accidents.

Addressing the SGTR event, the required structural integrity margins of theSG tubes and the tube-to-tubesheet joint over the H* distance will bemaintained. Tube rupture in tubes with cracks within the tubesheet isprecluded by the constraint provided by the presence of the tubesheet and thetube-to-tubesheet joint. Tube burst cannot occur within the thickness of thetubesheet. The tube-to-tubesheet joint constraint results from the hydraulicexpansion process, thermal expansion mismatch between the tube andtubesheet, and from the differential pressure between the primary andsecondary side, and tubesheet rotation. The structural margins against burst,as discussed in Regulatory Guide (RG) 1.121, "Bases for Plugging DegradedPWR Steam Generator Tubes," and Technical Specification (TS) 6.7.6.k, aremaintained for both normal and postulated accident conditions.

The proposed change has no impact on the structural or leakage integrity ofthe portion of the tube outside of the tubesheet. The proposed changemaintains structural and leakage integrity of the SG tubes consistent with theperformance criteria of TS 6.7.6.k. Therefore, the proposed change results inno significant increase in the probability of the occurrence of a SGTRaccident.

16

At normal operating pressures, leakage from tube degradation below theproposed limited inspection depth is limited by the tube-to-tubesheet crevice.Consequently, negligible normal operating leakage is expected fromdegradation below the inspected depth within the tubesheet region. Theconsequences of an SGTR event are not affected by the primary-to-secondaryleakage flow during the event as primary-to-secondary leakage flow through apostulated tube that has been pulled out of the tubesheet is essentiallyequivalent to a severed tube. Therefore, the proposed change does not resultin a significant increase in the consequences of a SGTR.

The consequences of a SLB or FLB are also not significantly affected by theproposed changes. The leakage analysis shows that the primary-to-secondaryleakage during a SLB/FLB event would be less than or equal to that assumedin the Updated Safety Analysis Report.

Primary-to-secondary leakage from tube degradation in the tubesheet areaduring the limiting accident (i.e., a SLB/FLB) is limited by flow restrictions.These restrictions result from the crack and tube-to-tubesheet contactpressures that provide a restricted leakage path above the indications and alsolimit the degree of potential crack face opening as compared to free spanindications.

The leakage factor of 2.49 for Seabrook Station, for a postulated SLB/FLB,has been calculated as shown in References 8, 9 and 10. For the ConditionMonitoring assessment, the component of leakage from the prior cycle frombelow the H* distance will be multiplied by a factor of 2.49 and added to thetotal leakage from any other source and compared to the allowable accidentinduced leakage limit. For the Operational Assessment, the difference in theleakage between the allowable leakage and the accident induced leakage fromsources other than the tubesheet expansion region will be divided by 2.49 andcompared to the observed operational leakage

The probability of a SLB/FLB is unaffected by the potential failure of a SGtube as the failure of the tube is not an initiator for a SLB/FLB event.SLB/FLB leakage is limited by flow restrictions resulting from the leakagepath above potential cracks through the tube-to-tubesheet crevice. The leakrate during all postulated accident conditions that model primary-to-secondaryleakage (including locked rotor and control rod ejection) has been shown toremain within the accident analysis assumptions for all axial and orcircumferentially orientated cracks occurring 15.21 inches below the top ofthe tubesheet. The assumed accident induced leak rate for Seabrook is 500gallons per day (gpd) during a postulated steam line break in the faulted loop.Using the limiting leak rate factor of 2.49, this corresponds to an acceptablelevel of operational leakage of 200 gpd. Therefore, the TS leak rate limit of

17

150 gpd provides significant added margin against the 500 gpd accidentanalysis leak rate assumption.

Therefore, the proposed change does not involve a significant increase in theprobability or consequences of an accident previously evaluated.

2. The proposed changes do not create the possibility of a new or different kindof accident from any previously evaluated

Response: No

The proposed change that alters the SG inspection and reporting criteria doesnot introduce any new equipment, create new failure modes for existingequipment, or create any new limiting single failures. Plant operation will notbe altered, and all safety functions will continue to perform as previouslyassumed in accident analyses.

Therefore, the proposed change does not create the possibility of a new ordifferent kind of accident from any. previously evaluated.

3. The proposed changes do not involve a significant reduction in the margin ofsafety.

Response: No

The proposed change that alters the SG inspection and reporting criteriamaintains the required structural margins of the SG tubes for both normal andaccident conditions. Nuclear Energy Institute 97-06, Rev. 3 "Steam GeneratorProgram Guidelines," and NRC Regulatory Guide (RG) .1. 121, "Bases forPlugging Degraded PWR Steam Generator Tubes," are used as the bases inthe development of the limited hot leg tubesheet inspection depthmethodology for determining that SG tube integrity considerations aremaintained within acceptable limits. RG 1.121 describes a method acceptableto the NRC for meeting General Design Criteria (GDC) 14, "Reactor CoolantPressure Boundary," GDC 15, "Reactor Coolant System Design," GDC 31,"Fracture Prevention of Reactor Coolant Pressure Boundary," and GDC 32,"Inspection of Reactor Coolant Pressure Boundary," by reducing theprobability and consequences of a SGTR. RG 1.121 concludes that bydetermining the limiting safe conditions for tube wall degradation, theprobability and consequences of a SGTR are reduced. This RG uses safetyfactors on loads for tube burst that are consistent with the requirements ofSection III of the American Society of Mechanical Engineers (ASME) Code.

18

For axially oriented cracking located within the tubesheet, tube burst isprecluded due to the presence of the tubesheet. For circumferentiallyoriented cracking, Westinghouse WCAP- 17071-P defines a length ofdegradation-free expanded tubing that provides the necessary resistance totube pullout due to the pressure induced forces, with applicable safety factorsapplied. Application of the limited hot and cold leg tubesheet inspectioncriteria will preclude unacceptable primary-to-secondary leakage during allplant conditions. The methodology for determining leakage as described inWCAP- 17071-P provides significant margin between the accident-inducedleakage assumption and the technical specification leakage limit duringnormal operating conditions when the proposed limited tubesheet inspectiondepth criteria is implemented.

Therefore, the proposed change does not involve a significant reduction in anymargin of safety.

Based on the above, NextEra concludes that the proposed amendment presents nosignificant hazards consideration under the standards set forth in 10 CFR 50.92(c)and, accordingly, a finding of "no significant hazards consideration" is justified.

4.3 Conclusion

The safety significant portion of the tube is the length of tube that is engaged within thetubesheet to the top of the tubesheet (secondary face) that is required to maintainstructural and leakage integrity over the full range of SG operating conditions, includingthe most limiting accident conditions. The H* analysis determined that degradation inthis distance from the top of the tubesheet does not require plugging and serves as thebasis for the limited tubesheet inspection criteria, which are intended to ensure theprimary to secondary leak rate during any accident does not exceed the leak rate assumedin the accident analysis.

Based on the considerations above, (1) there is reasonable assurance that the health andsafety of the public will not be endangered by operation in the proposed manner, (2) suchactivities will be conducted in compliance with the Commission's regulations, and (3) theissuance of the amendment will not be inimical to the common defense and security or tothe health and safety of the public

5.0 ENVIRONMENTAL CONSIDERATIONS

NextEra has evaluated the proposed amendment for environmental considerations. Thereview has determined that the proposed amendment would change a requirement withrespect to installation or use of a facility component located within the restricted area, asdefined in 10 CFR 20, and would change an inspection or surveillance requirement.However, the proposed amendment does not involve (i) a significant hazards

19

consideration, (ii) a significant change in the types or significant increase in the amountsof any effluent that may be released offsite, or (iii) a significant increase in individual orcumulative occupational radiation exposure. Accordingly, the proposed amendmentsmeet the eligibility criterion for categorical exclusion set for in 10 CFR 51.22(c) (9).Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement orenvironmental assessment needs to be prepared in connection with the proposedamendment.

6.0 REFERENCES

1. NRC Letter "Seabrook Station, Unit 1 - Issuance of Amendment RE: Limited Inspectionof the Steam Generator Tube Portion within the Tubesheet (TAC NO. MC85544),"September 29, 2006 (ML062630450)

2. NRC Letter "Seabrook Station, Unit 1 - Issuance of Amendment RE: Changes to TheSteam Generator Inspection Scope and Repair Requirements (TAC NO. ME1386),"October 13, 2009 (ML092460184)

3. NEI 97-06, "Steam Generator Program Guidelines" Revision 3, January 2011

4. EPRI 1003138, "Pressurized Water Reactor Steam Generator Examination Guidelines,"Revision

5. EPRI 1012987, Steam Generator Integrity Assessment Guidelines. Revision 3.

6. NRC Information Notice 2005-09, "Indications in Thermally Treated Alloy 600 SteamGenerator Tubes and Tube-to-Tubesheet Welds," April 7, 2005.

7. NRC Regulatory Guide 1.121, "Bases for Plugging Degraded PWR Steam GeneratorTubes," August 1976.

8. Westinghouse Electric Company LLC, WCAP-17330-P, "H*: Resolution of NRCTechnical Issue Regarding Tubesheet Bore Eccentricity", Rev. 1, June 2011.

9. Westinghouse Electric Company LLC, WCAP-17071-P, "H*: Alternate Repair Criteriafor the Tubesheet Expansion Region in Steam Generators with Hydraulically ExpandedTubes (Model F)."

10. Westinghouse Electric Company LLC, LTR-SGMP-09-100, "LTR-SGMP-09-100 P-Attachment, "Response to NRC Request for Additional Information on H*; Model F andModel D5 Steam Generators," August 12, 2009. (ADAMS Accession No.ML092450095(Non-Proprietary).

20

11. Westinghouse Electric Company LLC, LTR-SGMP-09-109 P-Attachment, "Response to

NRC Request for Additional Information on H*; RAI #4; Model F and Model D5 Steam

Generators," August 25, 2009. (ADAMS Accession No.ML092590299 (Non-

Proprietary)).

12. Westinghouse Electric Company LLC, LTR-SGMP-10-33 P-Attachment and LTR-SGMP- 10-33 NP-Attachment, LTR-SGMP- 10-33 P-Attachment, "H* Response to NRCQuestions Regarding Tubesheet Bore Eccentricity," (Proprietary/Non-Proprietary) forReview and Approval," September 13, 2010

13. Westinghouse Electric Company LLC, LTR-SGMP-10-78 P-Attachment and LTR-

SGMP-10-78 NP-Attachment, "Effects of Tubesheet Bore Eccentricity and Dilation on

Tube-to-Tubesheet Contact Pressure and Their Relative Importance to H*,"

(Proprietary/Non-Proprietary) for Review and Approval," September 7, 2010.

14. NRC Letter dated August 13, 2009 "Seabrook Station, Unit NO. 1 - Request for

Additional Information (RAI) Regarding Steam Generator Program (TAC NO. ME

1386)". (ADAMS Accession No. ML092100324).

15. NRC Letter dated September 1, 2009 "Seabrook Station, Unit NO. 1 - Second Request

for Additional Information (RAI) Regarding Steam Generator Program (TAC NO.ME1386)". (ADAMS Accession No. ML 092400135).

16. NRC Letter dated December 23, 2009 "Seabrook Station, Unit NO. 1, Transmittal of

Unresolved Issues Regarding Permanent Alternate Repair Criteria for Steam Generators(TAC NO. ME2628)". (ADAMS Accession No. ML 093421386).

17. 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 (H*)."

May 28, 2009. (ADAMS Accession No. ML091530539).

18. SBK-L-09168, "Response to Request for Additional Information Regarding Permanent

H* Alternate Repair Criteria for Steam Generator Inspections". September 16, 2009.

(ADAMS Accession No. ML092650369).

19. 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" September 18, 2009.

(ADAMS Accession No. ML092720883).

20. Westinghouse Electric Company LLC, LTR-SGMP-09-1 11 P-Attachment, Rev. 1 and

LTR-SGMP-09-1 11 NP-Attachment, Rev. 1, "Acceptable Value of the Location of the

21

Bottom of the Expansion Transition (BET) for Implementation of H*," (Proprietary/Non-

Proprietary) for Review and Approval," September 1, 2010.

21. Westinghouse Electric Company, LLC LTR-SGMP-1 1-58, "WCAP-17330-P, Revision

1 Erratum," July 6, 2011.

22. NRC letter to Southern Nuclear Operating Company, Inc., "Summary of February 16,

2011 Meeting with Southern Nuclear Operating Company, Inc. and Westinghouse on

Technical Issues Regarding Steam Generator Tube Inspection Permanent Alternate

Repair Criteria (TAC NOS. ME5417 and ME5418)," March 28, 2011. (ADAMS

Accession No. ML 110660648).

23. NRC letter to Southern Nuclear Operating Company, Inc., "Vogtle Electric GeneratingPlant Units 1 and 2 - Presubmittal Consideration of Steam Generator Alternative Repair

Criteria Requirements Request for Additional Information (TAC NOS. ME 5417 and

ME5418)," May 26, 2011. (ADAMS Accession No. ML1 1 140A099).

24. Southern Nuclear Operating Company, Inc. letter NL- 11-1178, "Vogtle Electric

Generating Plant - Response to Presubmittal Consideration of Steam Generator

Alternative Repair Criteria Requirements Request for Additional Information," June 20,

2011.

25. Duke Energy Corporation, "Proposed Technical Specifications (TS) Amendment TS

3.4.13, "RCS Operational LEAKAGE," TS 5.5.9, "Steam Generator (SG) Program," TS

5.6.8, "Steam Generator (SG) Tube Inspection Report," License Amendment Request to

Revise TS for Permanent Alternate Repair Criteria," June 30, 2011 (ADAMS Accession

No. ML11188A107).



5.6.8, "Steam Generator (SG) Tube Inspection Report," License Amendment Request to

Revise TS for Permanent Alternate Repair Criteria," July 11, 2011 (ADAMS Accession

No. ML11195A067).

27. Electronic mail from NRC to Duke Energy Corporation, "Catawba Nuclear Station Unit

2 (Catawba 2), Request for Additional Information (RAI) Regarding the Steam Generator

License Amendment Request to Revise Technical Specification for Permanent Alternate

Repair Criteria (TAC NO. ME667 1)", January 5, 2012. (ADAMS Accession No.

ML120090321).



22

5.6.8, "Steam Generator (SG) Tube Inspection Report," License Amendment Request toRevise TS for Permanent Alternate Repair Criteria," January 12, 2012. (ADAMSAccession No. ML 12019A250).

29. Virginia Electric and Power Company (Dominion) letter Serial No. 11-403, "LicenseAmendment Request Permanent Alternate Repair Criteria for Steam Generator TubeInspection and Repair," July 28, 2011. (ADAMS Accession No. ML1 12150144).

30. NRC letter to Virginia Electric and Power Company (Dominion), "Surry Power Station,Unit Nos. 1 and 2 - Request for Additional Information Regarding the Steam GeneratorLicense Amendment Request to Revise Technical Specifications for Permanent AlternateRepair Criteria (TAC NOS. ME6803 and ME6804)," January 18, 2012. (ADAMSAccession No. ML12006AOO 1).

31. Virginia Electric and Power Company (Dominion) letter Serial No. 12-028, "Response toRequest for Additional Information Related to License Amendment Request forPermanent Alternate Repair Criteria for Steam Generator Tube Inspections and Repair,"February 14, 2012. (ADAMS Accession No. ML12048A676)

32. Westinghouse Electric Company LLC, LTR-SGMMP- 11-28 Rev. 1 P-Attachment,"Response to USNRC Request for Additional Information Regarding the LicenseAmendment Requests for Permanent Application of the Alternate Repair Criterion, H*,to the Model D5 and Model F SGs," February 2, 2012.

23

Attachment 2

Markup of the Technical Specifications

Mark-up of the Technical Specifications (TS)

The attached markups reflect the currently issued version of the TS and Facility OperatingLicense. At the time of submittal, the Facility Operating License was revised throughAmendment No. 127.

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

LAR Title NextEra Energy DateSeabrook Letter Submitted

LARIO-02 Application for Change to the Technical SBK-L-10074 05/14/2010Specifications for the ContainmentEnclosure Emergency Air CleanupSystem

LAR 11-04 Changes to the Technical SBK-L-1 1245 01/30/2012

Specifications for New and SpentFuel Storage

LAR 11-06 Application to Revise theApplicability of the Reactor CoolantSystem Pressure - TemperatureLimits and the Cold OverpressureProtection Setpoints

SBK-L- 11186 11/17/2011

The following TS pages are included in the attached markup:

Technical PageSpecification Title

TS 6.7.61k Steam Generator (SG) Program 6-13

TS 6.8.1.7 Steam Generator Tube Inspection Report 6-216-21a

ADMINISTRATIVE CONTROLS

PROCEDURES AND PROGRAMS

6.7.6 (Continued)

The following alternate tube repair criteria shall be applied as an alternative to the40% depth based criteria:

r",Eef~efu~egV ai-q,*ýa•ndti squsfe~prrnspe•,ý io--"-btes with ,service-induced flaws located greater than inches below the top of the

tubesheet do not require plugging. Tubes with s ice-induced flaws locatedin the portion of the tube from the top of the tubeshe to inches belowthe top of the tubesheet shall be plugged upon detection .3'1• •J

d. Provisions for SG tube inspections. Periodic SG tube inspections shall beperformed. The number and portions of the tubes inspected and methods ofinspection shall be performed with the objective of detecting flaws of any type(e.g., volumetric flaws, axial and circumferential cracks) that may be present alongthe 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 repaircriteria. ( FD-irefuelin aqeA-3-and t "ubs Fnt in t,eony Pie portionof the tube below inches from the top of the tubesheet is excluded from this

2 - requirement. Thr tube-to-tubesheet weld is not part of the tube. In addition tomeeting 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 SGtube integrity is maintained until the next SG inspection. An assessment ofdegradation shall be performed to determine the type and location of flaws towhich the tubes may be susceptible and, based on this assessment, to determinewhich inspection methods need to be employed and at what locations.

1. Inspect 100% of the tubes in each SG during the first refueling outagefollowing 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 beconsidered to begin after the first inservice inspection of the SGs. Inaddition, inspect 50% of the tubes by the refueling outage nearest themidpoint of the period and the remaining 50% by the refueling outagenearest the end of the period. No SG shall operate for more than 48effective full power months or two refueling outages (whichever is less)without being inspected.

SEABROOK - UNIT 1 6-13 Amendment No. 34, 101, 109, 415, A


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 accidentanalysis limits) of the safety analysis are met. The CORE OPERATING LIMITSREPORT for each reload cycle, including any mid-cycle revisions orsupplements thereto, shall be provided upon issuance, to the NRC DocumentControl 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 4following completion of an inspection performed in accordance withSpecification 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 degradationmechanism,

d. Location, orientation (if linear), and measured sizes (if available) of serviceinduced indications,

e. Number of tubes plugged during the inspection outage for each activedegradation 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.

j.

SEABROOK

rFo.-FeueJiJg out~ie 13 apd-the sj. bequ R lnsp."to , e, e primary tosecondary leakage rate observed in each SG (if it is not practical to assign theleakage to an individual SG, the entire primary to secondary leakage shouldbe conservatively assumed to be from one SG) during the cycle preceditnheinspection which is the subject of the report,

FaF-refueln nathea andh~L srAMseq ;t ioz1ectic"n ce- he alculatedaccident induced leakage rate from the portion of the tubes below •.inchesfrom the top of the tubesheet for the most limiting accident in the most limitingSG. In addition, if the calculated accident induced leakage rate from the mostlimiting accident is less than . times the maximum operational primary tosecondary leakage rate, eport should describe how it was determined,and

- UNIT I 6-21 Amendment No. 22, 66, 88, 104, 107, 115 ,4-2.K


6.8.1.7 (Continued)

k. (ý; .;efeelg oute-Ae results of monitoring for tube axial displacement(slippage). If slippage is discovered, the implications of the discovery andcorrective 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 RegionalAdministrator within the time period specified for each report.

6.9 (THIS SPECIFICATION NUMBER IS NOT USED)

SEABROOK - UNIT I 6-21 a Amendment No. 123 I

Attachment 4

LTR-SGMMP- 11-28 Rev. I NP-Attachment

"Response to USNRC Request for Additional Information Regarding the LicenseAmendment Requests for Permanent Application of the Alternate Repair Criterion, H*, to the

Model D5 and Model F SGs" (Non-Proprietary)

Westinghouse Non-Proprietaiy Class 3LTR-SGMMP-1 1-28 Rev. I NP-Attachment

Response to USNRC Request for Additional Information Regarding theLicense Amendment Requests for Permanent Application of theAlternate Repair Criterion, H*, to the Model D5 and Model F SGs.

Westinghouse Electric Company LLC1000 Westinghouse Drive

Cranberry Township, PA 16066, USA

© 2012 Westinghouse Electric Company LLCAll Rights Reserved

I

LTR-SGMMP-1 1-28 Rev. I NP-Attachment

References:

1. Duke Energy Letter, "Duke Energy Carolina (Duke Energy) Catawba Nuclear Station,Units 1 and 2 Docket Numbers 50-413 and 50-414, Proposed Technical Specification(TS) Amendment, TS 3.4.13, "RCS Operational Leakage," TS 5.5.9, "Steam Generator(SG) Program," TS 5.6.8, "Steam Generator (SG) Tube Inspection Report," LicenseAmendment Request to Revise TS for Permanent Alternate Repair Criteria, June 30,2011.

2. E-mail from USNRC (Andrew Johnson) to Duke Energy (Jon Thompson) transmittingNRC letter, "Catawba Nuclear Station, Request for Additional Information Regardingthe Steam Generator License Amendment Request to Revise Technical Specificationfor Permanent Alternate Repair Criteria," November 15, 2011.

3. Dominion Letter, 11-403, "Surry Power Station Units 1 and 2 - License AmendmentRequest - Permanent Alternate Repair Criteria for Steam Generator Tube Inspectionand Repair," July 28, 2011, ADAMS Accession No. ML112150144.

4. USNRC Letter, "Surry Power Station Units 1 and 2 Request for Additional InformationRegarding the Steam Generator License Amendment Request to Revise TechnicalSpecification for Permanent Alternate Repair Criteria," (TAC Nos. ME6803 and ME6804, January 18, 2012.

5. SG-SGMP-1 1-16, "H* Technical Basis Independent Review by MPR Associates:Technical Questions and Responses," April 2011.

Introduction

In Reference 1, Duke Energy submitted a license amendment request (LAR) for permanentapplication of the alternate repair criterion H* at Catawba Unit 2 based on the technicaljustification in WCAP-1 7330-P, Revision 1. WCAP-1 7330-P Revision 1 also includes thetechnical justification for the Model F SGs at Seabrook, Salem 1, Millstone 3, Vogtle Units 1and 2 and Wolf Creek. Reference 2 transmitted the NRC request for additional information(RAI) regarding the Duke Energy LAR for a permanent application of H* for Catawba Unit 2.

Subsequent to the Duke Energy LAR for Catawba, Dominion Generation also submitted aLAR for permanent application of H* at Surry Units 1 and 2 (Reference 3). Whereas theCatawba technical justification is contained in WCAP-1 7330-P, Revision 1, the Surrytechnical justification is contained in WCAP-17345-P, Revision 2. Although the questions inReference 2 and Reference 4 are quite similar, some of them required different numericalinformation for Surry than for Catawba. Further, some of the questions in Reference 2 werenot repeated in Reference 4. sion 2. A separate response will be provided for the questionscontained in Reference 4.

It is anticipated that several utilities with Model F steam generators (SGs) will submit LARsfor the permanent application of H* for the Model F SGs. The Model F SG technicaljustification is also contained in WCAP-17330-P, Revision 1. This document augments theresponses to the Reference 2 questions to include similar responses applicable to the ModelF SGs. The questions that were noted in Reference 4 to not apply for the Reference 3

2


submittal are assumed to also not apply for the submittals for the Model F SGs. Notationsare made in the response to each question regarding the applicability of the response to the

Model F SGs.

Questions 1 through 11 from Reference 2 are reproduced below, followed by the responses.Questions 12 and 13 from Reference 2 will be addressed by the respective Model F utilities.

Question 14 from Reference 4 is assumed to apply for the Model F SGs and a response isprovided. Question 15 from Reference 4 is specific to the Dominion Generation (Surry 1 and2) LAR and does not apply for the Model F SGs.

Question 1:

WCAP-17330-P, Revision I - The footnote on page 3-53 states that Figure 3-36 showsthe same data as Figure 3-32 in Revision 0 of the WCAP, but without the data that

correspond to negative tubesheet CTE variation. The footnote states that while only afew percent of the data shown in Figure 3-32 of Revision 0 reflect negative values oftubesheet CTE, these cases do result in upward scatter, but must be included toproperly represent the top 10% of the Monte Carlo rank order results.. This being the

case, why does Figure 3-32 in Revision I properly represent the top 10% of the MonteCarlo rank order results? Why are the minimum H* values in Figure 3-36 of Revision Isubstantially different from those in Figure 3-32 of Revision 0?

Response:

This response applies for both the Model D5 and the Model F SGs.

The footnote on page 3-53 of WCAP-17330-P, Revision 1 erroneously states that Figure 3-36 in WCAP-17330-P, Revision 1 and Figure 3-32 in WCAP-17330-P, Revision 0 are fromthe same database. The title of Figure 3-36 in WCAP-1 7330-P, Revision 1 is correct; itapplies to the Model D5 SG at normal operating conditions. Figure 3-32 in WCAP-17330-P,Revision 0 applies to the Model F SGs at normal operating (NOP) conditions. Because thefigures apply to different models of SGs, the H* values are also different.

A prior NRC staff question (Ref: February 2011 meeting with the NRC staff) challenged thedata scatter in Figure 3-32 in WCAP-1 7330-P, Revision 0 and other similar figures,

specifically in the context of the efficacy of the "break-line" concept. Figure 3-36 in WCAP-

17330-P, Revision 1 shows the value of H* against the value of alpha (a), the square root ofthe sum of the squares of the component pairs of Monte Carlo selected values of coefficientsof thermal expansion of the tubesheet and the tube.

The footnote on page 3-53 of WCAP-1 7330-P, Revision 1 correctly notes that scatter in theRevision 0 figures is the result of the Monte Carlo process that results in samples withnegative variations of the tubesheet coefficient of thermal expansion with correspondinglarge negative variations in tube coefficient of thermal expansion (CTE). It is known from the


prior work that the maximum values of H* are likely to occur at positive variations oftubesheet CTE and negative variations of tube CTE. In the Monte Carlo analysis, describedfurther in the response to Question 3, approximately half of the H* values include a negativevariation of tubesheet CTE and a corresponding large negative variation of tube CTE;however, the frequency of occurrence in the rank order range of interest is low

As noted above, the probabilistic response surface is presented in terms of the combined

variable (x, the square root of the sum of the squares of the individual tube and tubesheet(TS) CTE components. The RSS combination of tube and tubesheet variables negates thesign of the negative variation of both the tube and TS CTE and artificially inflates the value of

ax, resulting in the upward data scatter shown on Figure 3-32 in WCAP-1 7330-P, Revision 0.

To address this issue in the H* analysis, Monte Carlo picks with a negative variation in TSCTE were assigned an H* value corresponding to a TS CTE variation of zero but with theMonte Carlo selected value of tube CTE. The complete process used for these points,discussed in the response to Question 3, results in a conservative value of H*.

Question 2:

WCAP-17330-P, Revision 0 - Provide copy of the "response surface" (i.e., H*relationship to coefficients of thermal expansion (CTE) variability for the tube andtubesheet) discussed for Model D5 steam line break (SLB) at the top of page 3-49.Confirm that this response surface applies to a radial location of 26.703 inches. Is this afull response surface or "partial" response surface of the type discussed in Revision I ofWCAP-1 7330-P, page 3-58?

Response:

This question was eliminated in the Reference 4 RAI and is also not considered to apply for

the Model F SGs.

The data for the requested response surface is provided in Table 2-1, below. It applies to aradial location of 26.703 inches for the bounding Model D5 plant at steam line break (SLB)condition. Note that the response surface considers only positive variations in the tubesheetCTE and negative variations in the tube CTE over a wide range of standard deviations,based on the prior experience of which parameters lead to the extreme values of H*. Hence,the name "reduced response surface."

4

LTR-SGMMP-1 1-28 Rev. 1 NP-Attachment

Table 2-1Reduced Response Surface; Model D5, 26.703 inches Radius

r

TS CTE T CTE

Case # H*+BET

(in)

2

3

4

5

6

7

8

9

10 _11

12

13

14

15

16

17

18

19

20

21

22

23

24

25

26

27

28

29

30

31

32

33

34

35

36

37

38

39

a,c,e

5


40

41

42

43

44

45 L

a,c,e

Question 3:

WCAP- I 7330-P, Revision I - Provide copy of the "reduced" response surfaces forbounding Model D5 SLB case discussed on page 3-58. Explain how the reducedresponse surfaces are used in the Monte Carlo analysis. If for a particular Monte Carloiteration a negative variation of tubesheet CTE is randomly generated, what is done withthis value (e.g., is tubesheet CTE assumed to have nominal value)? Why doesn't theuse of a reduced response surface bias the rank ordering above 90% in the non-conservative direction?

This question was modified in Reference 4 for the Model 51 F SG as noted below.Because the limitinq operatinq condition for the Model F SGs is the same as that for theModel 51 F SGs, the modified question is considered more appropriate for the Model FSGs.

WCAP-17345-P, Revision 2, Section 3.4 - Confirm that the Monte Carlo analysesperformed for the Model 51F SGs using the thick shell model are based upon samplingof the full H*/CTE response surfaces in Figure 8-5 of WCAP 17092 Rev 0. If this isincorrect, and only a "reduced" response surface is used, explain how the reducedresponse surfaces are used in the Monte Carlo analysis. If for a particular Monte Carloiteration a negative variation of tubesheet CTE is randomly generated, what is done withthis value (e.g., is tubesheet CTE assumed to have nominal value)? Why doesn't theuse of a reduced response surface bias the rank ordering above 90% in the non-conservative direction?

Response:

Model D5

Table 3-1 provides the data for the requested response surface for the Model D5 SGs at thecritical tubesheet radius of [ ]a,c,e inches. Note that the change in the maximum value ofH* (see Case 45) at the critical radius of [ ]a.c'e inches from the prior critical radius of26.703 inches shown in the response to Question 2 is only 0.03 inch.

The utilization of a reduced response surface as shown in Tables 2-1 and 3-1 does not biasthe rank ordering in a non-conservative direction; it simply limits the effort to develop aresponse surface to the region in parameter space where the limiting values of H* are most

6


likely located. The interpolation method for the reduced response surface permits calculationof H* values with the thick-shell equation, which is the underlying calculation basis of theresponse surface. The Monte Carlo process randomly samples, including variances in theregion excluded from the reduced response surface by means of the interpolation scheme.In approximately half of the cases, the sampling results have negative tubesheet CTEs.Because the ultimate objective is to define specific combinations of tubesheet and tube CTEsthat represent a specific rank order of H* values for input to the C2 model, the salientquestion is how points with negative tubesheet CTEs are treated in the probabilisticcalculation of H* using the C2 model.

Each of the 10,000 simulations in the general Monte Carlo procedure uses the followingprocess:

1. Pick a random normal deviate to represent the tubesheet CTE variation.2. Pick a random normal deviate for each tube in the steam generator to represent

the tube CTE variation.3. For each tube, assign an H* value corresponding to the current tubesheet CTE

variation and the tube's CTE variation by interpolating an H* value on theresponse surface. If the tubesheet CTE variation is negative, interpolate asthough the tubesheet CTE variation is zero (i.e., mean value).

4. Apply sector ratios as discussed in LTR-SGMP-09-100 P Attachment, Rev. 1.

5. Store the largest H* value along with the corresponding tube and tubesheet CTEvariations. Note that negative tubesheet CTE variations are retained, althoughthe H* assigned to them is conservative by step 3.

Steps 1-5 represent one iteration of the Monte Carlo process. This process is repeated10,000 times, and the results sorted in ascending order by H* value.

Step 3 of the process slightly distorts the rank order of the H* values because artificiallyhigher values of H* are assigned to the combination of randomly selected CTEs when theselected tubesheet CTE is negative. The true H* rank order of these cases is lower than theapparent value of H* for these cases. The effect is to displace the rank order of H*s withpositive values of tubesheet CTE to lower positions in the H* vector.

The manner in which these values are used in the subsequent step of the H* calculationprocess with the C2 model ensures a conservative H* value. For instance, in order to obtain,the 95/50 full bundle H* value, the 9 5 0 0th value in the H* rank order is chosen. In the eventthat the 9 5 0 0 th value contained a negative tubesheet CTE variation, the next higher rankorder value with a positive tubesheet CTE was chosen. In practice, only one or two rankorders needed to be traversed to find an H* with a positive tubesheet variation. Theparameters associated with this value were used in the calculation of H* with the C2 model.Since higher rank orders are more conservative (larger H* distance), the process of using thefirst higher rank order with a positive tubesheet CTE variation is conservative.


Model F

The Monte Carlo sampling for the Model F steam generators is based on sampling the fullH*/CTE response surfaces in Figure 8-5 of WCAP 17071-P, which is based on application ofthe thick-shell model.

The Monte Carlo process randomly samples from the response surface by means of aninterpolation scheme. In approximately half of the cases, the sampling results have negativetubesheet CTEs. Because the ultimate objective is to define specific combinations oftubesheet and tube CTEs that represent a specific rank order of H* values for input to the C

2

model, the salient question is how points with negative tubesheet CTEs are treated in theprobabilistic calculation of H* using the C2 model.

Each of the 10,000 simulations in the general Monte Carlo procedure uses the followingprocess:

1. Pick a random normal deviate to represent the tubesheet CTE variation.2. Pick a random normal deviate for each tube in the steam generator to represent

the tube CTE variation.3. For each tube, assign an H* value corresponding to the current tubesheet CTE

variation and the tube's CTE variation by interpolating an H* value on the

response surface. If the tubesheet CTE variation is negative, interpolate asthough the tubesheet CTE variation is zero (i.e., mean value).

4. Apply sector ratios as discussed in LTR-SGMP-09-1 00 P Attachment, Rev. 1.5. Store the largest H* value along with the corresponding tube and tubesheet CTE

variations.

Steps 1-5 represent one iteration of the Monte Carlo process. This process is repeated10,000 times, and the results sorted in ascending order by H* value.

Step 3 of the process slightly distorts the rank order of the H* values because artificiallyhigher values of H* are assigned to the combination of randomly selected CTEs when theselected tubesheet CTE is negative. The true H* rank order of these cases is lower than theapparent value of H* for these cases. The effect is to displace the rank order of H*s withpositive values of tubesheet CTE to lower positions in the H* vector.

In order to obtain, the 95/50 full bundle H* value, the 9 5 0 0 th value in the H* rank order ischosen. In the event that the 9 5 0 0 th value contained a negative tubesheet CTE variation, thenext higher rank order value with a positive tubesheet CTE was chosen. In practice, onlyone or two rank orders needed to be traversed to find an H* with a positive tubesheetvariation. The parameters associated with this value were used in the calculation of H* with

the C 2 model. Since higher rank orders are more conservative (larger H* distance), theprocess of using the first higher rank order with a positive tubesheet CTE variation isconservative. The same process is utilized when determining the H* value for the higherprobabilistic goals applicable to the Model F, that is, the 95/95 whole plant value of H*.

8

LTR-SGMMP-I 1-28 Rev. I NP-Attachment

Table 3-1

Reduced Response Surface; Model D5, [ ]aace inches Radius

TS CTE T CTE ll*+BETCase #

nl a fn (in)

2~~ ~ _______________31___________ ___________

6

7

8

9

10

11

12

13

14

15

16

17

18

19

20

21

22 ____

23

24

25

26

27

28

29 ___

30 ____

31

32

33

34

35 ____

36 ____ _____ _________

37 ____

38

a~c,e

9


39

40

41

42

43

44

45

a,c,e

10


Question 4.

WCAP- I 7330-P, Revision 1, Table 3-28 - Provide a similar table applicable to the ModelD5 SLB case, from the 9526 to 9546 rank orders.

Response

The question is Model D5-specific and does not apply for the Model F. However, Table 3-28of WCAP-1 7330-P, Revision 1 contains the data for the Model F SGs, centered on rank order9890.

Table 4-1 provides the requested information.

Table 4-1Range of Rank Order Statistics for Model D5Variation of CTEs Over a

Rank H* Tube Tubesheet Alpha(z)CTE CTE

9526 _-

9527

9528

9529

9530

9531

9532

9533

9534

9535

9

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