wolf creek - license amendment request, revision to ...on june 30, 2011, duke energy submitted a...

WLF CREEK'NUCLEAR OPERATING CORPORATION

John P. BroschakVice President Engineering

March 29, 2012

ET 12-0002

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

Subject: Docket No. 50-482: Revision to Technical Specification (TS) 5.5.9, "SteamGenerator (SG) Program," and TS 5.6.10, "Steam Generator Tube InspectionReport," for a Permanent Alternate Repair Criteria

Gentlemen:

Pursuant to 10 CFR 50.90, Wolf Creek Nuclear Operating Corporation (WCNOC) herebyrequests an amendment to Renewed Facility Operating License No. NPF-42 for the Wolf CreekGenerating Station (WCGS). This amendment request proposes to revise WCGS TechnicalSpecification (TS) 5.5.9, "Steam Generator (SG) Program," to permanently exclude portions ofthe tube below the top of the steam generator tubesheet from periodic steam generator tubeinspections. In addition, this amendment proposes to revise TS 5.6.10, "Steam Generator TubeInspection Report" to provide permanent reporting requirements that have been previouslyestablished on a one-cycle basis.

The proposed amendment constitutes a redefinition of the steam generator tube primary tosecondary pressure boundary and defines the safety significant portion of the tube that must beinspected or plugged. Tube flaws detected below the safety significant portion of the tube arenot required to be plugged. The exclusion of plugging flaws in the non-safety significant portionof the tube minimizes unnecessary tube plugging and maintains the safety margin of the steamgenerators to perform their safety function by maintaining the reactor coolant pressureboundary, reactor coolant flow, and primary to secondary heat transfer.

A,60o

P.O. Box 411 / Burlington, KS 66839 / Phone: (620) 364-8831

An Equal Opportunity Employer M/F/HC/VET

ET 12-0002 of 5

WCNOC requests the approval of the proposed license amendment by December 20, 2012 tosupport implementation during the WCGS 2013 Refueling Outage (Refueling Outage 19),currently scheduled to commence on February 4, 2013. Once approved, the amendment will beimplemented prior to MODE 4 entry during startup from Refueling Outage 19.

WCAP-17330-P, Revision 1, "H*: Resolution of NRC Technical Issue Regarding TubesheetBore Eccentricity (Model F/Model D5)," provides the resolution of the remaining questions insupport of the permanent application of the H* criterion. WCAP-17330-P, Revision 1, andWestinghouse Electric Company LLC LTR-SGMP-11-58 are not included with this application,as these documents have been previously docketed by Duke Energy. On June 30, 2011, DukeEnergy submitted a license amendment request for permanent application of the alternaterepair criterion H* at Catawba Unit 2 based on the technical justification in WCAP-17330-P,Revision 1. A supplement to the license amendment request was submitted on July 11, 2011and provided Westinghouse Electric Company LLC LTR-SGMP-11-58, "WCAP-17330-P,Revision 1 Erratum." The NRC transmitted on January 5, 2012, by electronic mail, a requestfor additional information. Duke Energy responded to the request for additional information onJanuary 12, 2012.

Subsequent to the Duke Energy license amendment request, Virginia Electric and PowerCompany (Dominion) submitted a license amendment request for permanent application of thealternate repair criterion H* for Surry Power Station Units 1 and 2. On January 18, 2012, theNRC issued a request for additional information. Dominion responded to the request foradditional information on February 14, 2012.

Enclosure I (Westinghouse Electric Company LLC, LTR-SGMMP-11-28 Rev.1 P-Attachment,"Response to USNRC Request for Additional Information Regarding the License AmendmentRequests for Permanent Application of the Alternate Repair Criterion, H*, to the Model D5 andModel F SGs") augments the responses to the Duke Energy request for additional informationto include similar responses applicable to Model F steam generators. Additionally, Enclosure Iaddresses the Dominion request for additional information on question 14 for the Model Fsteam generators.

Attachment VI provides WCNOC specific responses to questions 12 and 13 from the DukeEnergy request for additional information and question 15 from the Dominion request foradditional information.

Enclosure I provides the proprietary Westinghouse Electric Company LLC, LTR-SGMMP-1 1-28Rev.1 P-Attachment, "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." Enclosure II provides the non-proprietary WestinghouseElectric Company LLC "LTR-SGMMP-11-28 Rev.1 NP-Attachment, "Response to USNRCRequest for Additional Information Regarding the License Amendment Requests for PermanentApplication of the Alternate Repair Criterion, H*, to the Model D5 and Model F SGs." AsEnclosure I contains information proprietary to Westinghouse Electric Company LLC, it issupported by an affidavit signed by Westinghouse Electric Company LLC, the owner of theinformation. The affidavit sets forth the basis on which the information may be withheld frompublic disclosure by the Commission and addresses with specificity the considerations listed inparagraph (b)(4) of 10 CFR 2.390 of the Commission's regulations. Accordingly, it isrespectfully requested that the information, which is proprietary to Westinghouse, be withheldfrom public disclosure in accordance with 2.390 of the Commission's regulations. This affidavit,along with Westinghouse authorization letter, CAW-12-3417, "Application for WithholdingProprietary Information from Public Disclosure," is contained in Enclosure Ill.

ET 12-0002 of 5

Attachments I through IV provide the Evaluation, Markup of TSs, Proposed TS Bases changes,and Retyped TS pages, respectively, in support of this amendment request. Attachment Ill,proposed changes to the TS Bases, is provided for information only. Final TS Bases changeswill be implemented pursuant to TS 5.5.14, "Technical Specification (TS) Bases ControlProgram," at the time the amendment is implemented.

It has been determined that this amendment application does not involve a significant hazardconsideration as determined per 10 CFR 50.92, "Issuance of amendment." Pursuant to 10CFR 51.22, "Criterion for categorical exclusion; identification of licensing and regulatory actionseligible or otherwise not requiring environmental review," Section (b), no environmental impactstatement or environmental assessment needs to be prepared in connection with the issuanceof this amendment.

The Plant Safety Review Committee reviewed this amendment application. In accordance with10 CFR 50.91, "Notice for public comment; State consultation," a copy of this amendmentapplication, with attachments, is being provided to the designated Kansas State official.

Attachment V provides the regulatory commitments associated with this application. If youhave any questions concerning this matter, please contact me at (620) 364-4085, or Mr.Gautam Sen at (620) 364-4175.

Sincerely,

P. Broschak

JPB/rlt

ET 12-0002 of 5

Attachments:IVIII

IVVVI

EvaluationMarkup of Technical Specification pagesProposed Changes to Technical Specification Bases (for informationonly)Retyped Technical Specification pagesList of Regulatory CommitmentsResponse to Request for Additional Information Questions Specific toWolf Creek Nuclear Operating Corporation

Enclosure I - Westinghouse Electric Company LLC, LTR-SGMMP-11-28 Rev.1 P-Attachment, "Response to USNRC Request for Additional InformationRegarding the License Amendment Requests for Permanent Applicationof the Alternate Repair Criterion, H*, to the Model D5 and Model F SGs"(Proprietary)

11 - Westinghouse Electric Company LLC LTR-SGMMP-11-28 Rev.1 NP-Attachment and Errata, "Response to USNRC Request for AdditionalInformation Regarding the License Amendment Requests for PermanentApplication of the Alternate Repair Criterion, H*, to the Model D5 andModel F SGs" (Non-proprietary)

III - Westinghouse Electric Company LLC CAW-12-3417, "Application forWithholding Proprietary Information from Public Disclosure"

cc: E. E. Collins (NRC), w/a, w/eT. A. Conley (KDHE), w/a, w/e (Enclosure II only)J. R. Hall (NRC), w/a, w/eN. F. O'Keefe (NRC), w/a, wieSenior Resident Inspector (NRC), w/a, w/e

ET 12-0002 of 5

STATE OF KANSAS )SS

COUNTY OF COFFEY )

John P. Broschak, of lawful age, being first duly sworn upon oath says that he is Vice PresidentEngineering of Wolf Creek Nuclear Operating Corporation; that he has read the foregoingdocument and knows the contents thereof; that he has executed the same for and on behalf ofsaid Corporation with full power and authority to do so; and that the facts therein stated are trueand correct to the best of his knowledge, information and belief.

ByJoO1P. BroschakVi g(President Engineering

SUBSCRIBED and sworn to before me this f9 z day of MdrA ,2012.

GAYLE SHEPHEARD1M&Notary Public - State of KansasMy Appt. Expires 2L '-2/ . • / J

Notary "b li-0

Expiration Date_________

Attachment I to ET 12-0002Page 1 of 23

EVALUATION

Subject: Revision to Technical Specification (TS) 5.5.9, "Steam Generator (SG) Program,"and TS 5.6.10, "Steam Generator Tube Inspection Report," for a PermanentAlternate Repair Criteria

1. SUMMARY DESCRIPTION

2. DETAILED DESCRIPTION

3. TECHNICAL EVALUATION

4. REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria

4.2 Significant Hazards Consideration

4.3 Conclusions

5. ENVIRONMENTAL CONSIDERATION

6. REFERENCES


1. SUMMARY DESCRIPTION

Wolf Creek Nuclear Operating Corporation (WCNOC) proposes to revise Wolf CreekGenerating Station (WCGS) Technical Specification (TS) 5.5.9, "Steam Generator (SG)Program," to exclude portions of the tube below the top of the steam generator tubesheet fromperiodic steam generator tube inspections. In addition, this amendment proposes to revise TS5.6.10, "Steam Generator Tube Inspection Report" to remove reference to previous interimalternate repair criteria and provide reporting requirements specific to the permanent alternaterepair criteria. Application of the supporting structural analysis and leakage evaluation resultsto exclude portions of the tubes from inspection and repair of tube indications is interpreted toconstitute a redefinition of the primary to secondary pressure boundary. The proposedchanges to the TS are based on the supporting structural analysis and leakage evaluationcompleted by Westinghouse Electric Company LLC. The documentation supporting theWestinghouse analysis is described in Section 3 and provides the licensing basis for thischange. Table 5-1 of WCAP 17330-P (Reference 19) provides the 95/95 H* value of 15.21inches for plants with Model F Steam Generators which includes WCGS.

The NRC previously issued the following amendments revising steam generator tube inspectionrequirements:

* Amendment Number 162 (Reference 1) to exclude degradation found in the portion ofthe tubes below 17 inches from the top of the hot leg tubesheet from the requirement toplug for Refueling Outage 14 and the subsequent operating cycle.

" Amendment Number 169 (Reference 2) to exclude degradation found in the portion ofthe tubes below 17 inches from the top of the hot leg tubesheet from the requirement toplug for Refueling Outage 15 and the subsequent operating cycle.

* Amendment Number 178 (Reference 3) which approved an interim alternate repaircriteria for Refueling Outage 16 and the subsequent operating cycle that requires full-length inspection of the tubes within the tubesheet but does not require plugging tubes ifany circumferential cracking observed in the region greater than 17 inches from the topof the tubesheet is less than a value sufficient to permit the remaining circumferentialligament to transmit the limiting axial loads. Three new reporting requirements wereadded to TS 5.6.10.

* Amendment Number 186 (Reference 4) revised TS 5.5.9, "Steam Generator (SG)Program," to eliminate inspection and repair of tubes more than 13.1 inches below the topof the tubesheet for Refueling Outage 17 and the subsequent operating cycle. AdditionallyTS 5.6.10 was revised to provide reporting requirements specific to Refueling Outage 17and the subsequent operating cycle.

* Amendment Number 195 (Reference 23) revised TS 5.5.9, "Steam Generator (SG)Program," to eliminate inspection and repair of tubes more than 15.2 inches below the topof the tubesheet for Refueling Outage 18 and the subsequent operating cycle. AdditionallyTS 5.6.10 was revised to provide reporting requirements specific to Refueling Outage 18and the subsequent operating cycle.


Approval of this amendment application is requested by December 20, 2012 to support theWCGS Refueling Outage 19 (February 2013), since the existing one-cycle amendment expiresat the end of the current operating cycle.

2. DETAILED DESCRIPTION

Proposed Changes to Current TS

TS 5.5.9c. currently states:

c. Provisions for SG tube repair criteria. Tubes found by inserviceinspection to contain flaws with a depth equal to or exceeding 40% of thenominal tube wall thickness shall be plugged.

The following alternate tube repair criteria shall be applied as analternative to the 40% depth-based criteria:

1. For Refueling Outage 18 and the subsequent operating cycle,tubes with service-induced flaws located greater than 15.2 inchesbelow the top of the tubesheet do not require plugging. Tubeswith service-induced flaws located in the portion of the tube fromthe top of the tubesheet to 15.2 inches below the top of thetubesheet shall be plugged upon detection.

This section would be revised as follows:

c. Provisions for SG tube repair criteria. Tubes found by inserviceinspection to contain flaws with a depth equal to or exceeding 40% of thenominal tube wall thickness shall be plugged.


1. Tubes with service-induced flaws located greater than 15.21inches below the top of the tubesheet do not require plugging.Tubes with service-induced flaws located in the portion of the tubefrom the top of the tubesheet to 15.21 inches below the top of thetubesheet shall be plugged upon detection.

TS 5.5.9d. currently states:

d. Provisions for SG tube inspections. Periodic SG tube inspections shallbe performed. The number and portions of the tubes inspected andmethods of inspection shall be performed with the objective of detectingflaws 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 tubeoutlet, and that may satisfy the applicable tube repair criteria. For


Refueling Outage 18 and the subsequent operating cycle, the portion ofthe tube below 15.2 inches from the top of the tubesheet is excluded fromthis requirement. The tube-to-tubesheet weld is not part of the tube. Inaddition to meeting the requirements of d.1, d.2, and d.3 below, theinspection scope, inspection methods, and inspection intervals shall besuch as to ensure that SG tube integrity is maintained until the next SGinspection. An assessment of degradation shall be performed todetermine the type and location of flaws to which the tubes may besusceptible and, based on this assessment, to determine whichinspection methods need to be employed and at what locations.

1. Inspect 100% of the tubes in each SG during the first refuelingoutage following SG replacement.

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

3. If crack indications are found in any portion of the SG tube notexcluded above, then the next inspection for each SG for thedegradation mechanism that caused the crack indication shall notexceed 24 effective full power months or one refueling outage(whichever is less). If definitive information, such as fromexamination of a pulled tube, diagnostic non-destructive testing,or engineering evaluation indicates that a crack-like indication isnot associated with a crack(s), then the indication need not betreated as a crack.

This section would be revised as follows:

d. Provisions for SG tube inspections. Periodic SG tube inspections shallbe performed. The number and portions of the tubes inspected andmethods of inspection shall be performed with the objective of detectingflaws 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 tubeoutlet, and that may satisfy the applicable tube repair criteria. The portionof the tube below 15.21 inches from the top of the tubesheet is excludedfrom 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, theinspection scope, inspection methods, and inspection intervals shall besuch as to ensure that SG tube integrity is maintained until the next SGinspection. 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 whichinspection methods need to be employed and at what locations.


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

3. If crack indications are found in any portion of the SG tube notexcluded above, then the next inspection for each SG for thedegradation mechanism that caused the crack indication shall notexceed 24 effective full power months or one refueling outage(whichever is less). If definitive information, such as fromexamination of a pulled tube, diagnostic non-destructive testing,or engineering evaluation indicates that a crack-like indication isnot associated with a crack(s), then the indication need not betreated as a crack.

TS 5.6.1Oh., 5.6.1Oi., and 5.6.1Oj. currently state:

h. Following completion of an inspection performed in Refueling Outage 18(and any inspections performed in the subsequent operating cycle) theprimary to secondary LEAKAGE rate observed in each SG (if it is notpractical to assign the LEAKAGE to an individual SG, the entire primaryto secondary LEAKAGE should be conservatively assumed to be fromone SG) during the cycle preceding the inspection which is the subject ofthe report;

Following completion of an inspection performed in Refueling Outage 18(and any inspections performed in the subsequent operating cycle) thecalculated accident induced leakage rate from the portion of the tubesbelow 15.2 inches from the top of the tubesheet for the most limitingaccident in the most limiting SG. In addition, if the calculated accidentinduced leakage rate from the most limiting accident is less than 2.50times the maximum operational primary to secondary leak rate, the reportshould describe how it was determined; and

j. Following completion of an inspection performed in Refueling Outage 18(and any inspections performed in the subsequent operating cycle) theresults of monitoring for the tube axial displacement (slippage). If


slippage is discovered, the implications of discovery and corrective action

shall be provided.

TS 5.6.1Oh., 5.6.10i., and 5.6.1Oj. would be revised as follows:

h. The primary to secondary LEAKAGE rate observed in each SG (if it is notpractical to assign the LEAKAGE to an individual SG, the entire primaryto secondary LEAKAGE should be conservatively assumed to be fromone SG) during the cycle preceding the inspection which is the subject ofthe report;

The calculated accident induced leakage rate from the portion of thetubes below 15.21 inches from the top of the tubesheet for the mostlimiting accident in the most limiting SG. In addition, if the calculatedaccident induced leakage rate from the most limiting accident is less than2.50 times the maximum operational primary to secondary leak rate, thereport should describe how it was determined; and

j. The results of monitoring for the tube axial displacement (slippage). Ifslippage is discovered, the implications of discovery and corrective actionshall be provided.

3. TECHNICAL EVALUATION

Background

WCGS is a four loop Westinghouse designed plant with Model F steam generators having 5626tubes in each steam generator. A total of 266 tubes are currently plugged in all four steamgenerators. The design of the steam generator includes Alloy 600 thermally treated tubing, fulldepth hydraulically expanded tubesheet joints, and stainless steel tube support plates withquatrefoil broached holes.

The steam generator inspection scope is governed by TS 5.5.9, "Steam Generator (SG)Program;" Nuclear Energy Institute (NEI) 97-06, "Steam Generator Program Guidelines,"(Reference 5); EPRI 1013706, "Pressurized Water Reactor Steam Generator ExaminationGuidelines," (Reference 6); EPRI 1019038, "Steam Generator Integrity AssessmentGuidelines," (Reference 7); WCNOC procedure AP 29A-003, "Steam Generator Management;"and the results of the degradation assessments required by the Steam Generator Program.Criterion IX, "Control of Special Processes" of 10 CFR Part 50, Appendix B, requires in part thatnondestructive testing be accomplished by qualified personnel using qualified procedures inaccordance with the applicable criteria. The inspection techniques and equipment are capableof reliably detecting the existing and potential specific degradation mechanisms applicable toWCGS. The inspection techniques, essential variables and equipment are qualified toAppendix H, "Performance Demonstration for Eddy Current Examination," and Appendix I,"NDE System Measurement Uncertainties for Tube Integrity Assessments," of the EPRI SteamGenerator Examination Guidelines.


Catawba Nuclear Station, Unit 2, (Catawba) reported indication of cracking followingnondestructive eddy current examination of the steam generator tubes during their fall 2004outage. NRC Information Notice (IN) 2005-09, "Indications in Thermally Treated Alloy 600Steam Generator Tubes and Tube-to-Tubesheet Welds," (Reference 8), 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 tubesheet inone tube, and just above the tube-to-tubesheet welds in a region of the tube known as the tackexpansion in several other tubes. Indications were also reported in the tube-end welds, alsoknown as tube-to-tubesheet welds, which join the tube to the tubesheet.

WCNOC policies and programs, as well as TS 5.5.9, require the use of applicable industryoperating experience in the operation and maintenance of WCGS. The experience at Catawba,as noted in IN 2005-09, shows the importance of monitoring all tube locations (such as bulges,dents, dings, and other anomalies from the manufacture of the steam generators) withtechniques capable of finding potential forms of degradation that may be occurring at theselocations (as discussed in Generic Letter 2004-001, "Requirements for Steam Generator TubeInspections"). Since the WCGS Westinghouse Model F steam generators were fabricated withAlloy 600 thermally treated tubes similar to the Catawba Unit 2 Westinghouse Model D5 steamgenerators, a potential exists for WCGS to identify tube indications similar to those reported atCatawba within the hot leg tubesheet region.

Potential inspection plans for the tubes and tube welds underwent intensive industrydiscussions in March 2005. The findings in the Catawba steam generator tubes present threedistinct issues with regard to the steam generator tubes at WCGS:

1) Indications in internal bulges and overexpansions 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 indications intoadjacent tube material.

Prior to each steam generator tube inspection, a degradation assessment, which includes areview of operating experience, is performed to identify degradation mechanisms that have apotential to be present in the WCGS steam generators. A validation assessment is alsoperformed to verify that the eddy current techniques utilized are capable of detecting those flawtypes that are identified in the degradation assessment. Based on operating experiencediscussed above, WCNOC revised the steam generator inspection plan to include sampling ofbulges and overexpansions within the tubesheet region on the hot leg side in Refueling Outage14 (Spring 2005) and Refueling Outage 15 (Fall 2006). The sample is based on the guidancecontained in EPRI 1013706, "Pressurized Water Reactor Steam Generator ExaminationGuidelines," Revision 7, and TS 5.5.9, "Steam Generator (SG) Program." The inspection planis expanded according to EPRI steam generator examination guidelines if necessary due toconfirmed degradation in the region required to be examined (i.e. a tube crack). Degradationwas not detected in the tubesheet region in Refueling Outage 14 and Refueling Outage 15.


At WCGS, tube flaw indications within the tube sheet have only been found at the hot leg tubeends. Approximately 18,414 tube ends were inspected at WCGS during Refueling Outage 16(Spring 2008). Seventy-six flaw indications have been found in the inspections within 1 inch ofthe tube end. Of these seventy-six flaw indications, only eight met the tube repair criteria in thetechnical specifications.

Based on these inspections, a limited number of tube flaws existed in the tubesheet area of theWCGS steam generators. The flaws that have been found are associated with residual stressconditions at the tube ends. No indications of a 360 degree sever have been detected in anysteam generator at WCGS. Consequently, the level of degradation in the WCGS steamgenerators is very limited compared to the assumption of "all tubes severed" that was utilized inthe development of the permanent H* alternate repair criterion. Consequently, structuralintegrity will be assured for the operating period between inspections allowed by the proposedTS 5.5.9, "Steam Generator (SG) Program."

As a result of these potential issues and to prevent the unnecessarily plugging of additionaltubes in the WCGS steam generators, WCNOC is proposing changes to TS 5.5.9 to limit thesteam generator tube inspection and repair (plugging) to the safety significant portion of thetubes.

Summary of Licensing Basis Analysis (H* Analysis)

On June 2, 2009, Westinghouse WCAP-17071-P, Revision 0, "H*: Alternate Repair Criteria forthe Tubesheet Expansion Region in Steam Generators with Hydraulically Expanded Tubes(Model F)," (Reference 10) was submitted as Enclosure I of WCNOC request (Reference 11) tochange Technical Specification (TS) 5.5.9, "Steam Generator (SG) Program," and TS 5.6.10,"Steam Generator Tube Inspection Report," to support implementation of a permanent alternaterepair criterion for steam generator tubes.

On August 11, 2009, WCNOC received a request for additional information (RAI) letter(Reference 12), which contained twenty-five (25) questions.

On August 25, 2009 (Reference 13) and September 3, 2009 (Reference 14), WCNOC providedthe responses to questions 1 through 25 of the August 11, 2009 letter and included thefollowing documents:

* Westinghouse letter LTR-SGMP-09-100 P-Attachment, Revision 0, "Response to NRCRequest for Additional Information on H*; Model F and Model D5 Steam Generators,"August 12, 2009 (Reference 15), and

* Westinghouse letter LTR SGMP-09-109-P Attachment, Revision 0 "Response to NRCRequest for Additional Information on H*; RAI #4; Model F and Model D5 SteamGenerators," August 25, 2009 (Reference 16).


On September 15, 2009, WCNOC submitted a request (Reference 17) to revise the permanentalternate repair criteria amendment request (Reference 11) to be an interim change applicableto Refueling Outage 17 and the subsequent operating cycle. This request was made inresponse to a September 2, 2009 teleconference between NRC Staff and industry personnel, inwhich the NRC Staff indicated that their concerns with eccentricity of the tube sheet tube borein normal and accident conditions (RAI question 4 of the August 11, 2009 letter) have not beenresolved. The September 15, 2009 letter also requested the NRC staff to provide the specificquestions concerning the tubesheet bore eccentricity issue which must be resolved to support apermanent alternate repair criteria amendment request.

On December 9, 2009, the NRC provided a letter (Reference 18) documenting the currentlyidentified and unresolved issues relating to tubesheet bore eccentricity. This letter contained 14unresolved questions which required resolution before the NRC could complete its review of apermanent amendment request. Section 1.2 of WCAP-17330-P, Revision 1 (Reference 19)provides a discussion of the action plan to respond to the 14 unresolved questions.

The following documents have been prepared by Westinghouse to provide final resolution ofthe remaining questions identified in the December 9, 2009 NRC letter in support of thepermanent H* amendment for WCGS.

* WCAP-17330-P, Revision 1, "H*: Resolution of NRC Technical Issue Regarding TubesheetBore Eccentricity (Model F/Model D5)," June 2011 (Reference 19).

" LTR-SGMP-10-78 P-Attachment, "Effects of Tubesheet Bore Eccentricity and Dilation onTube-to-Tubesheet Contact Pressure and Their Relative Importance to H*," September 7,2009. This document, which is applicable to WCGS's Model F steam generators, wastransmitted to the NRC by Westinghouse letter LTR-NRC-10-68 on November 9, 2010(Reference 20).

* LTR-SGMP-09-1 11 P-Attachment, Rev. 1, "Acceptable Value of the Location of the Bottomof the Expansion Transition (BET) for Implementation of H*," was prepared to support plantdeterminations of BET measurements and their significant deviation assessment. Thisdocument, which is applicable to WCGS's Model F steam generators, was transmitted tothe NRC by Westinghouse letter LTR-NRC-10-69 on November 10, 2010 (Reference 22).

* LTR-SGMP-10-33 P-Attachment, "H* Response to NRC Questions Regarding TubesheetBore Eccentricity," September 13, 2010. This document, which is applicable to WCGS'sModel F steam generators, was transmitted to the NRC by Westinghouse letter LTR-NRC-10-70 on November 11, 2010 (Reference 21).

Note that WCAP-17330-P, Revision 1, "H*: Resolution of NRC Technical Issue RegardingTubesheet Bore Eccentricity (Model F/Model D5)," June 2011 (Reference 19) makes referenceto Revision 2 of WCAP-17071-P and Revision 1 of LTR-SGMP-09-100 P-Attachment. Asdescribed above, WCNOC has previously submitted Revision 0 of these documents. Theserevisions (Revisions 1 and 2 of WCAP-17071-P, Revision 1 of LTR-SGMP-09-100 P-Attachment) were created to resolve editorial comments. The technical information containedin WCAP-17071-P, Revision 0 and LTR-SGMP-09-100 P-Attachment, Revision 0, remains validand provides part of the licensing basis for the requested amendment.


The following table provides the list of the WCGS licensing basis documents for H*.

Document Revision Title ReferenceNumber Number Number

WCAP-1 7330-P 1 H*: Resolution of NRC Technical Issue Regarding 19Tubesheet Bore Eccentricity (Model F/Model D5)

LTR SGMP-11-58 0 WCAP-17330-P, Revision 1 Erratum 24

WCAP-17071-P 0 H*: Alternate Repair Criteria for the Tubesheet 10Expansion Region in Steam Generators withHydraulically Expanded Tubes (Model F)

LTR-SGMP-09-100 0 Response to NRC Request for Additional Information 15P-Attachment on H*; Model F and Model D5 Steam Generators

LTR -SGMP-09- 0 Response to NRC Request for Additional Information 16109 P-Attachment on H*; RAI #4; Model F and Model D5 Steam

Generators

LTR-SGMP-10-78 0 Effects of Tubesheet Bore Eccentricity and Dilation 20P-Attachment on Tube-to-Tubesheet Contact Pressure and Their

Relative Importance to H*

LTR-SGMP-10-33 0 H* Response to NRC Questions Regarding 21P-Attachment Tubesheet Bore Eccentricity

LTR-SGMMP-11- 1 Response to USNRC Request for Additional 3528 P-Attachment Information Regarding the License Amendment

Requests for Permanent Application of the AlternateRepair Criterion, H*, to the Model D5 and Model FSGs

In addition, the following correspondence is also applicable to the WCGS permanent alternaterepair criteria request.

* A March 28, 2011 letter from the NRC to Southern Nuclear Operating Company (Reference25) documented the summary of a February 16, 2011 public meeting regarding steamgenerator tube inspection permanent alternate repair criteria. Enclosure 3 of the NRC letterprovided technical NRC Staff questions developed at the meeting. Responses to thesequestions have been incorporated into WCAP-17330-P, Revision 1 (Reference 19).

* Section 1.3 of Reference 19 identifies revisions to the report (WCAP-17330-P, Revision 1)to address recommendations from the independent review of the H* analysis performed byMPR Associates. Related to the independent review, a May 26, 2011 letter from the NRCto Southern Nuclear Company (Reference 26) included a presubmittal review request foradditional information. The response to the NRC presubmittal review request is provided inSouthern Nuclear Operating Company letter NL-1 1-1178 (Reference 27).

On June 30, 2011, Duke Energy submitted a license amendment request (Reference 28) forpermanent application of the alternate repair criterion H* at Catawba Unit 2 based on thetechnical justification in WCAP-17330-P, Revision 1, "H*: Resolution of NRC Technical Issue


Regarding Tubesheet Bore Eccentricity (Model F/Model D5)." A supplement (Reference 29) tothe license amendment request was submitted on July 11, 2011 and provided WestinghouseElectric Company LLC LTR-SGMP-1 1-58, "WCAP-1 7330-P, Revision 1 Erratum." On January5, 2012, a request for additional information (Reference 30) was transmitted electronically toDuke Energy. Duke Energy responded to the request for additional information on January 12,2012 (Reference 31).

Subsequent to the Duke Energy license amendment request, Virginia Electric and PowerCompany (Dominion) submitted a license amendment request (Reference 32) for permanentapplication of the alternate repair criterion H* for Surry Power Station Units 1 and 2. OnJanuary 18, 2012, the NRC issued a request for additional information (Reference 33).Dominion responded to the request for additional information on February 14, 2012 (Reference34).

Westinghouse Electric Company LLC, LTR-SGMMP-11-28 Rev.1 P-Attachment (Reference35), "Response to USNRC Request for Additional Information Regarding the LicenseAmendment Requests for Permanent Application of the Alternate Repair Criterion, H*, to theModel D5 and Model F SGs," augments the responses to the Duke Energy request foradditional information to include similar responses applicable to Model F steam generators.Additionally, LTR-SGMMP-11-28 Rev.1 P-Attachment addresses the Dominion request foradditional information question 14 for the Model F steam generators. Attachment VI providesWCNOC specific responses to questions 12 and 13 from the Duke Energy request foradditional information and question 15 from the Dominion request for additional information.

Evaluation

To preclude unnecessarily plugging tubes in the WCGS steam generators, an evaluation wasperformed to identify the safety significant portion of the tube within the tubesheet necessary tomaintain structural and leakage integrity in both normal and accident conditions. Tubeinspections will be limited to identifying and plugging degradation in the safety significantportion of the tubes. The technical evaluation for the inspection and repair methodology isprovided in the H* Analysis as described above. This evaluation is based on the use of finiteelement model structural analysis and a bounding leak rate evaluation based on contactpressure between the tube and the tubesheet during normal and postulated accidentconditions. The limited tubesheet inspection criteria were developed for the tubesheet region ofthe WCGS Model F steam generator considering the most stringent loads associated with plantoperation, including transients and postulated accident conditions. The limited tubesheetinspection criteria were selected to prevent tube burst and axial separation due to axial pulloutforces acting on the tube and to ensure that the accident induced leakage limits are notexceeded. The H* Analysis provides technical justification for limiting the inspection in thetubesheet expansion region to less than the full depth of the tubesheet.

The basis for determining the safety significant portion of the tube within the tubesheet is basedupon evaluation and testing programs that quantified the tube-to-tubesheet radial contactpressure for bounding plant conditions as described in the H* Analysis. The tube-to-tubesheetradial contact pressure provides resistance to tube pullout and resistance to leakage duringplant operation and transients.


Primary-to-secondary leakage from tube degradation is assumed to occur in several designbasis accidents: feedwater line break (FLB), steam line break (SLB), locked rotor, and controlrod ejection. The radiological dose consequences associated with this assumed leakage areevaluated to ensure that they remain within regulatory limits (e.g. 10 CFR Part 100, 10 CFR50.67, GDC 19). The accident induced leakage performance criteria are intended to ensure theprimary-to-secondary leak rate during any accident does not exceed the primary-to-secondaryleak rate assumed in the accident analysis. Radiological dose consequences define the limitingaccident condition for the H* Analysis.

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 (Reference 5) and NRCRegulatory Guide (RG) 1.121, "Bases for Plugging Degraded PWR Steam Generator Tubes,"(Reference 9) are satisfied due to the constraint provided by the tubesheet. Throughapplication of the limited tubesheet inspection scope as described below, the existing operatingleakage limit provides assurance that excessive leakage (i.e., greater than accident analysisassumptions) will not occur. The accident induced leak rate limit for WCGS is 1.0 gpm. TheTS 3.4.13, "RCS Operational LEAKAGE," operational leak rate limit is 150 gpd (0.1 gpm)through any one steam generator. Consequently, accident leakage is approximately 10 timesthe allowable leakage, if only one steam generator is leaking. Using a SLB/FLB overall leakagefactor of 2.50, accident induced leakage is less than 0.6 gpm, if all 4 steam generators areleaking at 150 gpd at the beginning of the accident. Therefore, significant margin existsbetween the conservatively estimated accident induced leakage and the allowable accidentleakage (1.0 gpm).

Plant-specific operating conditions are used to generate the overall leakage factor ratios thatare to be used in the condition monitoring and operational assessments. The plant-specificdata provide the initial conditions for application of the transient input data. The results of theanalysis of the plant-specific inputs, to determine the bounding plant for each model of steamgenerator are contained in Section 6 of Reference 10.

As discussed in Reference 19, the leak rate ratio (accident induced leak rate to operational leakrate) is a product of the pressure differential subfactor and the viscosity subfactor using theDarcy flow equation.

I

The plant transient response following a full power double-ended main feedwater line rupturecorresponding to "best estimate" initial conditions and operating characteristics as generallypresented in the Updated Safety Analysis Report (USAR) Chapter 15.0 safety analysis,indicates that the transient for a Model F steam generator exhibits a cooldown characteristicinstead of a heatup transient. The use of either the component design specification transient orthe Chapter 15.0 safety transient for leakage analysis for FLB is overly conservative because:

The assumptions on which the FLB design transient is based are specifically intended toestablish a conservative structural (fatigue) design basis for RCS components; however,H* does not involve component structural and fatigue issues. The best estimatetransient is considered more appropriate for use in the H* leakage calculations.


* For the Model F steam generator, the FLB transient curve (Figure 9-5, Reference 10)represents a double-ended rupture of the main feedwater line concurrent with bothstation blackout (loss of main feedwater and reactor coolant pump coast down) andturbine trip.

" The assumptions on which the FLB safety analysis is based are specifically intended toestablish 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 no longerbe applicable. This would result in much earlier reactor trip and greatly increase thesteam generator liquid mass available to provide cooling to the RCS.

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

In accordance with plant operating procedures, the operator would take action following a highenergy secondary line break to stabilize the RCS conditions. The expectation for a SLB or FLBwith credited operator action is to stop the system cooldown through isolation of the faultedsteam generator and control of temperature by the AFW System. Steam pressure controlwould be established by either the steam generator safety valves or control system(atmospheric relief valves). For any of the steam pressure control operations, the maximumtemperature would be approximately the no load temperature and would be well below normaloperating temperature.

Since the best estimate FLB transient temperature would not be expected to exceed the normaloperating temperature, the viscosity ratio for the FLB transient is set to 1.0. Therefore, the leakrate factor would only be a function of the increase in pressure differential during the designbasis SLB/FLB. However, per Reference 15, the FLB transient was evaluated as a heatupevent. Since dynamic viscosity decreases with the increase in temperature during a postulatedFLB event, the viscosity subfactor increases above 1.0. For WCGS, the resulting leak rate ratiofor both the SLB and FLB events is conservatively determined to be 2.50.

The other design basis accidents, such as the postulated locked rotor event and the control rodejection event, are conservatively modeled using the design specification transients to result inincreased temperatures in the steam generator hot and cold legs for a period of time. Aspreviously noted, dynamic viscosity decreases with increasing temperature. Therefore, leakagewould be expected to increase due to decreasing viscosity and increasing differential pressurefor the duration of time that there is a rise in RCS temperature. For transients other than a SLBand a FLB, the length of time that a plant with Model F steam generators will exceed the normaloperating differential pressure across the tubesheet is less than 30 seconds. As the accidentinduced leakage performance criteria is defined in gallons per minute, the leak rate for a lockedrotor ejection event can be integrated over a minute to compare to the limit. Time integrationpermits an increase in acceptable leakage during the time of peak pressure differential byapproximately a factor of two because of the short duration (less than 30 seconds) of theelevated pressure differential. This translates into an effective reduction in leakage factor bythe same factor of two for the locked rotor event. Therefore, for the locked rotor event, the


leakage factor of 1.77 (Revised Table 9-7, Reference 15) for WCGS is adjusted downward to afactor 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 may be reduced by a factorof six from 2.65 to 0.44.

For the condition monitoring (CM) assessment, the component of leakage from the prior cyclefrom below the H* distance will be multiplied by a factor of 2.50 and added to the total leakagefrom any other source and compared to the allowable accident induced leakage limit. For theoperational assessment (OA), the difference in the leakage between the allowable leakage andthe accident induced leakage from sources other than the tubesheet expansion region will bedivided by 2.50 and compared to the observed operational leakage. An administrative limit willbe established to not exceed the calculated value.

Reference 11 redefines the primary pressure boundary. The tube-to-tubesheet weld no longerfunctions as a portion of this boundary. The hydraulically expanded portion of the tube into thetubesheet over the H* distance now functions as the primary pressure boundary in the area ofthe tube and tubesheet, maintaining the structural and leakage integrity over the full range ofsteam generator operating conditions, including the most limiting accident conditions. Theevaluation in Reference 11 determined that degradation in tubing below this safety significantportion of the tube does not require inspection or repair (plugging). The inspection of the safetysignificant portion of the tubes provides a high level of confidence that the structural andleakage performance criteria are maintained during normal operating and accident conditions.

WCAP-17071-P (Reference 10), section 9.8, provides a review of leak rate susceptibility due totube slippage and concluded that the tubes are fully restrained against motion under veryconservative design and analysis assumptions such that tube slippage is not a credible eventfor any tube in the bundle. As a condition of approval of Amendment Number 186, WCNOCcommitted to monitor for tube slippage as part of the steam generator tube inspection program.This commitment will remain in place to support the permanent alternate repair criteria request,and the results of monitoring will be reported in accordance with TS 5.6.10.

As a condition for approving the WCGS Interim Alternate Repair Criterion (Reference 3), theNRC staff requested that WCNOC perform a validation of the tube expansion from the top oftubesheet to the bottom of expansion transition (BET) to determine if there are any significantdeviations that would invalidate assumptions in WCAP-17071-P (Reference 10). WCNOC hascompleted the validation of the tube expansion from the top of tubesheet to the BET. Based ondata review and LTR-SGMP-09-111 P-Attachment, Rev. 1 (Reference 22), WCNOC did notidentify any significant deviations from the top of tubesheet to the BET for WCGS.

4. REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria

Section 182a of the Atomic Energy Act requires applicants for nuclear power plant operatinglicenses to include Technical Specifications (TSs) as part of the operating license. The TSsensures the operational capability of structures, systems, and components that are required toprotect the health and safety of the public. The U.S. Nuclear Regulatory Commission's (NRC's)requirements related to the content of the TSs are contained in Section 50.36 of the Title 10 of


the Code of Federal Regulations (10 CFR 50.36) which requires that the TSs include items inthe following specific categories: (1) safety limits, limiting safety system settings, and limitingcontrol settings; (2) limiting conditions for operation; (3) surveillance requirements per 10 CFR50.36(c)(3); (4) design features; and (5) administrative controls.

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

GDC 19 of 10 CFR 50, Appendix A, defines requirements for the control room and for theradiation protection of the operators working within it. Accidents involving the leakage or burstof steam generator tubing comprise a challenge to the habitability of the control room.

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

10 CFR 100, Reactor Site Criteria, established reactor siting criteria, with respect to the risk ofpublic exposure to the release of radioactive fission products. Accidents involving leakage ortube burst of steam generator tubing may comprise a challenge to containment and thereforeinvolve an increased risk of radioactive release.

Under 10 CFR 50.65, the Maintenance Rule, licensees classify steam generators asrisk-significant components because they are relied upon to remain functional during and afterdesign basis events. Steam generators are to be monitored under 10 CFR 50.65(a)(2) againstindustry established performance criteria. Meeting the performance criteria of Nuclear EnergyInstitute (NEI) 97-06, Revision 2, "Steam Generator Program Guidelines," provides reasonableassurance that the steam generator tubing remains capable of fulfilling its specific safetyfunction of maintaining the reactor coolant pressure boundary. The NEI 97-06, Revision 2,steam generator performance criteria are:

* All in-service steam generator tubes shall retain structural integrity over the full range ofnormal operating conditions (including startup, operation in the power range, hotstandby, cooldown and all anticipated transients included in the design specification)and design basis accidents. This includes retaining a safety factor of 3.0 against burstunder normal steady state full power operation primary to secondary pressuredifferential 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 due topressure 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 basis accident,other than a steam generator tube rupture, shall not exceed the leakage rate assumedin the accident analysis in terms of total leakage rate for all steam generators andleakage rate for an individual steam generator. Leakage is not to exceed 1 gpm persteam generator, except for specific types of degradation at specific locations whenimplementing alternate repair criteria as documented in the Steam Generator Programtechnical specifications.

* The Reactor Coolant System (RCS) operational primary to secondary leakage throughany one steam generator shall be limited to 150 gallons per day.

The safety significant portion of the tube is the length of tube that is engaged in the tubesheetfrom the secondary face that is required to maintain structural and leakage integrity over the fullrange of steam generator operating conditions, including the most limiting accident conditions.The evaluation in this Attachment determined that degradation in tubing below the safetysignificant portion of the tube does not require plugging and serves as the bases for thetubesheet inspection program. As such, the WCGS inspection program provides a high level ofconfidence that the structural and leakage criteria are maintained during normal operating andaccident conditions.

4.2 Significant Hazards Consideration

This amendment application proposes to revise Technical Specification (TS) 5.5.9, "SteamGenerator (SG) Program," to exclude portions of the tubes within the tubesheet from periodicsteam generator inspections. In addition, this amendment proposes to revise TechnicalSpecification (TS) 5.6.10, "Steam Generator Tube Inspection Report," to remove reference toprevious interim alternate repair criteria and provide reporting requirements specific to thetemporary alternate repair criteria. Application of the structural analysis and leak rateevaluation results, to exclude portions of the tubes from inspection and repair is interpreted toconstitute a redefinition of the primary to secondary pressure boundary.

The proposed change defines the portion of the tube that must be inspected and repaired. Ajustification has been developed by Westinghouse Electric Company, LLC to identify thespecific inspection depth below which any type of axial or circumferential primary water stresscorrosion cracking can be shown to have no impact on Nuclear Energy Institute (NEI) 97-06,"Steam Generator Program," performance criteria.

WCNOC has evaluated whether or not a significant hazards consideration is involved with theproposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuanceof amendment," Part 50.92(c), as discussed below:

(1) Does the proposed change involve a significant increase in the probability orconsequences of an accident previously evaluated?

Response: No

The previously analyzed accidents are initiated by the failure of plant structures, systems,or components. The proposed change that alters the steam generator inspection criteriadoes not have a detrimental impact on the integrity of any plant structure, system, or


component that initiates an analyzed event. The proposed change will not alter theoperation of, or otherwise increase the failure probability of any plant equipment thatinitiates an analyzed accident.

Of the applicable accidents previously evaluated, the limiting transients with considerationto the proposed change to the steam generator tube inspection and repair criteria are thesteam generator tube rupture (SGTR) event, the steam line break (SLB) and the feedlinebreak (FLB) postulated accidents.

Addressing the SGTR event, the required structural integrity margins of the steamgenerator tubes and the tube-to-tubesheet joint over the H* distance will be maintained.Tube rupture in tubes with cracks within the tubesheet is precluded by the presence of thetubesheet and constraint provided by the tube-to-tubesheet joint. Tube burst cannotoccur within the thickness of the tubesheet. The tube-to-tubesheet joint constraint resultsfrom the hydraulic expansion process, thermal expansion mismatch between the tube andtubesheet, from the differential pressure between the primary and secondary side, andtubesheet deflection. The structural margins against burst, as discussed in RegulatoryGuide (RG) 1.121, "Bases for Plugging Degraded PWR Steam Generator Tubes," and TS5.5.9 are maintained for both normal and postulated accident conditions.

The proposed change has no impact on the structural or leakage integrity of the portion ofthe tube outside of the tubesheet. The proposed change maintains structural and leakageintegrity of the steam generator tubes consistent with the performance criteria in TS 5.5.9.Therefore, the proposed change results in no significant increase in the probability of theoccurrence of a SGTR accident.

At normal operating pressures, leakage from tube degradation below the proposed limitedinspection depth is limited by the tube-to-tubesheet joint. Consequently, negligible normaloperating leakage is expected from degradation below the inspected depth within thetubesheet region. The consequences of an SGTR event are not affected by the primaryto secondary leakage flow during the event as primary to secondary leakage flow througha postulated tube that has been pulled out of the tubesheet is essentially equivalent to asevered tube. Therefore, the proposed changes do not result in a significant increase inthe consequences of a SGTR.

The consequences of a SLB or FLB are also not significantly affected by the proposedchanges. The leakage analysis shows that the primary-to-secondary leakage during aSLB/FLB event would be less than or equal to that assumed in the Updated SafetyAnalysis Report.

Primary-to-secondary leakage from tube degradation in the tubesheet area during thelimiting accidents (i.e., SLB/FLB) is limited by flow restrictions. These restrictions resultfrom the crack and tube-to-tubesheet contact pressures that provide a restricted leakagepath above the indications and also limit the degree of potential crack face opening ascompared to free span indications.

The leakage factor of 2.50 for WCGS, for a postulated SLB/FLB, has been calculated asshown in References 10, 15 and 19. Specifically, for the condition monitoring (CM)assessment, the component of leakage from the prior cycle from below the H* distance


will be multiplied by a factor of 2.50 and added to the total leakage from any other sourceand compared to the allowable accident induced leakage limit. For the operationalassessment (OA), the difference in the leakage between the allowable leakage and theaccident induced leakage from sources other than the tubesheet expansion region will bedivided by 2.50 and compared to the observed operational leakage.

The probability of an SLB/FLB is unaffected by the potential failure of a steam generatortube as the failure of the tube is not an initiator for an SLB/FLB event. SLB/FLB leakageis limited by leakage flow restrictions resulting from the leakage path above potentialcracks through the tube-to-tubesheet crevice. The leak rate during all postulated accidentconditions that model primary-to-secondary leakage (including locked rotor and controlrod ejection) has been shown to remain within the accident analysis assumptions for allaxial and or circumferentially orientated cracks occurring 15.21 inches below the top of thetubesheet. The accident induced leak rate limit for WCGS is 1.0 gpm. The TS 3.4.13,"RCS Operational LEAKAGE," operational leak rate limit is 150 gpd (0.1 gpm) through anyone steam generator. Consequently, accident leakage is approximately 10 times theallowable leakage, if only one steam generator is leaking. Using the limiting SLB/FLBoverall leakage factor of 2.50, accident induced leakage is less than 0.6 gpm, if all 4steam generators are leaking at 150 gpd at the beginning of the accident. Therefore,significant margin exists between the conservatively estimated accident induced leakageand the allowable accident leakage (1.0 gpm).

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

(2) Does the change create the possibility of a new or different kind of accident from anyaccident previously evaluated?

Response: No

The proposed change alters the steam generator inspection and reporting criteria. It doesnot introduce any new equipment, create new failure modes for existing equipment, orcreate any new limiting single failures. Plant operation will not be altered, and safetyfunctions will continue to perform as previously assumed in accident analyses.

Therefore, the proposed change does not create the possibility of a new or different kindof accident from any accident previously evaluated.

(3) Does the change involve a significant reduction in a margin of safety?

Response: No

The proposed change alters the steam generator inspection and reporting criteria. Itmaintains the required structural margins of the steam generator tubes for both normaland accident conditions. NEI 97-06 and RG 1.121, are used as the bases in thedevelopment of the limited tubesheet inspection depth methodology for determining thatsteam generator tube integrity considerations are maintained within acceptable limits. RG1.121 describes a method acceptable to the NRC for meeting 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 ReactorCoolant Pressure Boundary," by reducing the probability and consequences of a SGTR.RG 1.121 concludes that by determining the limiting safe conditions for tube walldegradation, the probability and consequences of a SGTR are reduced. This RG usessafety factors on loads for tube burst that are consistent with the requirements of SectionIII of the American Society of Mechanical Engineers (ASME) Code.

For axially-oriented cracking located within the tubesheet, tube burst is precluded due tothe presence of the tubesheet. For circumferentially-oriented cracking, the H* Analysisdocumented in Section 3, defines a length of degradation-free expanded tubing thatprovides the necessary resistance to tube pullout due to the pressure induced forces, withapplicable safety factors applied. Application of the limited hot and cold leg tubesheetinspection criteria will preclude unacceptable primary to secondary leakage during all plantconditions. The methodology for determining leakage provides for large margins betweencalculated and actual leakage values in the proposed limited tubesheet inspection depthcriteria.

Therefore, the proposed change does not involve a significant reduction in any margin of

safety.

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 maintain structuraland leakage integrity over the full range of steam generating operating conditions, including themost limiting accident conditions. The H* Analysis determined that degradation in tubing belowthe safety significant portion of the tube does not require plugging and serves as the basis forthe limited tubesheet inspection criteria, which are intended to ensure the primary-to-secondaryleak rate during any accident does not exceed the leak rate assumed in the accident analysis.

Based on the considerations above, 1) there is a 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 to thehealth and safety of the public.

5. ENVIRONMENTAL CONSIDERATION

WCNOC has evaluated the proposed amendment for environmental considerations. Thereview has determined that the proposed amendment would change a requirement with respectto installation or use of a facility component located within the restricted area, as defined in 10CFR 20, and would change an inspection or surveillance requirement. However, the proposedamendment does not involve (i) a significant hazards consideration, (ii) a significant change inthe types or significant increase in the amounts of any effluent that may be released offsite, or(iii) a significant increase in individual or cumulative occupational radiation exposure.Accordingly, the proposed amendments meet the eligibility criterion for categorical exclusion setforth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact


statement or environmental assessment needs to be prepared in connection with the proposedamendment.

6. REFERENCES

1. NRC letter from J. N. Donohew, USNRC, to R. A. Muench, WCNOC, "Wolf CreekGenerating Station - Issuance of Exigent Amendment RE: Steam Generator (SG) TubeSurveillance Program (TAC NO. MC6757)," April 28, 2005. (ADAMS Accession No.ML051160100)

2. NRC letter from J. N. Donohew, USNRC, to R. A. Muench, WCNOC, "Wolf CreekGenerating Station - Issuance of Amendment RE: Steam Generator Tube InspectionsWithin The Tubesheet (TAC NO. MD2467)," October 10, 2006. (ADAMS Accession No.ML062580019)

3. NRC letter from J. N. Donohew, USNRC, to R. A. Muench, WCNOC, "Wolf CreekGenerating Station - Issuance of Amendment RE: Revision to Technical Specification5.5.9 on the Steam Generator Program (TAC NO. MD8054)," April 4, 2008. (ADAMSAccession No. ML080840004)

4. NRC letter from B. K. Singal, USNRC, to R. A. Muench, WCNOC, "Wolf CreekGenerating Station - Issuance of Amendment RE: Revision to Technical Specification(TS) 5.5.9, "Steam Generator (SG) Program," and TS 5.6.10 "Steam Generator TubeInspection Report," for Alternate Repair Criteria (TAC NO. ME1393)," October 19, 2009.(ADAMS Accession No. ML092750606)

5. NEI 97-06, "Steam Generator Program Guidelines."

6. EPRI 1013706, "Pressurized Water Reactor Steam Generator Examination Guidelines."

7. EPRI 1019038, "Steam Generator Integrity Assessment Guidelines."

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

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

10. Westinghouse Electric Company LLC, WCAP-17071-P, Revision 0, "H*: AlternateRepair Criteria for the Tubesheet Expansion Region in Steam Generators withHydraulically Expanded Tubes (Model F)," April 2009. (ADAMS Accession No.ML091590167 (Non-Proprietary))

11. WCNOC letter ET 09-0016, "Revision to Technical Specification 5.5.9, "SteamGenerator (SG) Program," and TS 5.6.10, "Steam Generator Tube Inspection Report,"for a Permanent Alternate Repair Criterion," June 2, 2009. (ADAMS Accession No.ML091590170)


12. NRC letter from B. K. Singal, USNRC, to R. A. Muench, WCNOC, "Wolf CreekGenerating Station - Request for Additional Information Regarding the PermanentAlternate Repair Criteria License Amendment Request (TAC NO. ME1393)," August 11,2009. (ADAMS Accession No. ML092100200)

13. WCNOC letter ET 09-0021, "Response to Request for Additional Information Related toLicense Amendment Request for a Permanent Alternate Repair Criterion to TechnicalSpecification 5.5.9, "Steam Generator (SG) Program"," August 25, 2009. (ADAMSAccession No. ML092450095)

14. WCNOC letter ET 09-0023, "Response to Request for Additional Information Related toLicense Amendment Request for a Permanent Alternate Repair Criterion to TechnicalSpecification 5.5.9, "Steam Generator (SG) Program"," September 3, 2009. (ADAMSAccession No. ML092590299)

15. LTR-SGMP-09-1 00, "LTR-SGMP-09-100 P-Attachment, "Response to NRC Request forAdditional Information on H*; Model F and Model D5 Steam Generators," August 12,2009. (ADAMS Accession No. ML092450095 (Non-Proprietary))

16. LTR-SGMP-09-109 P-Attachment, "Response to NRC Request For AdditionalInformation on H*; RAI #4; Model F and Model D5 Steam Generators," August 25, 2009.(ADAMS Accession No. ML092590299 (Non-Proprietary))

17. WCNOC letter ET 09-0025, "Revision to Technical Specification (TS) 5.5.9, "SteamGenerator (SG) Program," and TS 5.6.10, "Steam Generator Tube Inspection Report","September 15, 2009. (ADAMS Accession No. ML092730340)

18. NRC letter from B. K. Singal, USNRC, to R. A. Muench, WCNOC, "Wolf CreekGenerating Station - Transmittal of Unresolved Issues Regarding Permanent AlternateRepair Criteria for Steam Generators (TAC NO. ME1393)," December 9, 2009.(ADAMS Accession No. ML093360459)

19. WCAP-17330-P, Revision 1, "H*: Resolution of NRC Technical Issue RegardingTubesheet Bore Eccentricity (Model F/Model D5)," June 2011.

20. LTR-NRC-10-68, "Submittal of LTR-SGMP-10-78 P-Attachment and LTR-SGMP-10-78NP-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," November 9, 2010.

21. LTR-NRC-10-70, "Submittal of LTR-SGMP-10-33 P-Attachment and LTR-SGMP-10-33NP-Attachment, LTR-SGMP-10-33 P-Attachment, "H* Response to NRC QuestionsRegarding Tubesheet Bore Eccentricity," (Proprietary/Non-Proprietary) for Review andApproval," November 11, 2010.

22. LTR-NRC-10-69, "Submittal of LTR-SGMP-09-111 P-Attachment, Rev. 1 and LTR-SGMP-09-1 11 NP-Attachment, Rev. 1, "Acceptable Value of the Location of the Bottomof the Expansion Transition (BET) for Implementation of H*," (Proprietary/Non-Proprietary) for Review and Approval," November 10, 2010.


23. NRC letter from J. R. Hall, USNRC, to M. W. Sunseri, WCNOC, "Wolf Creek GeneratingStation - Issuance of Amendment RE: Changes to Technical Specification (TS) 5.5.9,"Steam Generator (SG) Program," and TS 5.6.10 "Steam Generator Tube InspectionReport," (TAC NO. ME5121)," April 6, 2011. (ADAMS Accession No. ML1 10840590)

24. Westinghouse Electric Company LLC LTR LTR-SGMP-11-58, "WCAP-17330-P,Revision 1 Erratum," July 6, 2011.

25. NRC letter to Southern Nuclear Operating Company, Inc., "Summary of February 16,2011 Meeting with Southern Nuclear Operating Company, Inc. and Westinghouse onTechnical Issues Regarding Steam Generator Tube Inspection Permanent AlternateRepair Criteria (TAC NOS. ME5417 and ME5418)," March 28, 2011. (ADAMSAccession No. ML1 10660648)

26. NRC letter to Southern Nuclear Operating Company, Inc., "Vogtle Electric GeneratingPlant Units 1 and 2 - Presubmittal Consideration of Steam Generator Alternative RepairCriteria Requirements Request for Additional Information (TAC NOS. ME 5417 andME5418)," May 26, 2011. (ADAMS Accession No. ML11140A099)

27. Southern Nuclear Operating Company, Inc. letter NL-11-1178, "Vogtle ElectricGenerating Plant - Response to Presubmittal Consideration of Steam GeneratorAlternative Repair Criteria Requirements Request for Additional Information," June 20,2011.

28. Duke Energy Corporation, "Proposed Technical Specifications (TS) Amendment TS3.4.13, "RCS Operational LEAKAGE," TS 5.5.9, "Steam Generator (SG) Program," TS5.6.8, "Steam Generator (SG) Tube Inspection Report," License Amendment Request toRevise TS for Permanent Alternate Repair Criteria," June 30, 2011. (ADAMS AccessionNo. ML11188A107)

29. Duke Energy Corporation, "Proposed Technical Specifications (TS) Amendment TS3.4.13, "RCS Operational LEAKAGE," TS 5.5.9, "Steam Generator (SG) Program," TS5.6.8, "Steam Generator (SG) Tube Inspection Report," License Amendment Request toRevise TS for Permanent Alternate Repair Criteria," July 11, 2011. (ADAMS AccessionNo. ML11195A067)

30. Electronic mail from NRC to Duke Energy Corporation, "Catawba Nuclear Station Unit 2(Catawba 2), Request for Additional Information (RAI) Regarding the Steam GeneratorLicense Amendment Request to Revise Technical Specification for Permanent AlternateRepair Criteria (TAC NO. ME6671)," January 5, 2012. (ADAMS Accession No.ML120090321)

31. Duke Energy Corporation, "Proposed Technical Specifications (TS) Amendment TS3.4.13, "RCS Operational LEAKAGE," TS 5.5.9, "Steam Generator (SG) Program," TS5.6.8, "Steam Generator (SG) Tube Inspection Report," License Amendment Request toRevise TS for Permanent Alternate Repair Criteria," January 12, 2012. (ADAMSAccession No. ML12019A250)


32. 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)

33. 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 PermanentAlternate Repair Criteria (TAC NOS. ME6803 and ME6804)," January 18, 2012.(ADAMS Accession No. ML12006A001)

34. 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.

35. 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*, tothe Model D5 and Model F SGs," February 2, 2012.

Attachment II to ET 12-0002Page 1 of 4

ATTACHMENT II

Markup of Technical Specification pages

Attachment II to ET 12-0002Page 2 of 4

5.5 Programs and Manuals

Programs and Manuals5.5

5.5.9 Steam Generator (SG) Proqram (continued)

3. The operational LEAKAGE performance criterion is specified inLCO 3.4.13, "RCS Operational LEAKAGE."

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


i.Fri f~fe ag 418Andhe ub e uetM'pp tircc(f'" 7, ubes with service-induced flaws located greater than r.ilches

E I below the top of the tubesheet do not require plugging. ubes withservice-induced flaws located in the portion of the tube from thetop of the tubesheet to inches below the top of the tubesheet

shall be plugged upon detection.

(continued)

Wolf Creek - Unit 1 5.0-12 Amendment No. 123, 153, 172, 178,1-86, 195

Attachment II to ET 12-0002Page 3 of 4 Programs and Manuals

5.5


5.5.9 Steam Generator (SG) Program (continued)

d. Provisions for SG tube inspections. Periodic SG tube inspections shall beperformed. The number and portions of the tubes inspected and methodsof inspection shall be performed with the objective of detecting flaws ofany type (e.g., volumetric flaws, axial and circumferential cracks) that maybe present along the length of the tube, from the tube-to-tubesheet weldat the tube inlet to the tube-to-tubesheet weld at the tube outlet and that

Thne.. 'd thef's brseqtfe opi6ratifig yore portion of the tube below 2inches from the top of the tubesheet is excluded from this requirement.The tube-to-tubesheet weld is not part of the tube. In addition to meetingthe requirements of d.1, d.2, and d.3 below, the inspection scope,inspection methods, and inspection intervals shall be such as to ensurethat SG tube integrity is maintained until the next SG inspection. Anassessment of degradation shall be performed to determine the type andlocation of flaws to which the tubes may be susceptible and, based on thisassessment, to determine which inspection methods need to be employedand at what locations.


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

3. If crack indications are found in any portion of the SG tube notexcluded above, then the next inspection for each SG for thedegradation mechanism that caused the crack indication shall notexceed 24 effective full power months or one refueling outage(whichever is less). If definitive information, such as fromexamination of a pulled tube, diagnostic non-destructive testing, orengineering evaluation indicates that a crack-like indication is notassociated with a crack(s), then the indication need not be treatedas a crack.

e. Provisions for monitoring operational primary to secondary LEAKAGE.

(continued)

Wolf Creek-Unit 1 5.0-13 Amendment No. 123, 153,172, 178,-1-6, 195

Attachment II to ET 12-0002Page 4 of 4 Reporting Requirements

5.6

5.6 Reporting Requirements

5.6.10 Steam Generator Tube Inspection Report

A report shall be submitted within 180 days after the initial entry into MODE 4following completion of an inspection performed in accordance with theSpecification 5.5.9, 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 andin-situ testing;

h. Fo in of an specti pedfoiMed u g Ouge Iand inspeo rons per the sutsequ e-iVt ope tinc

.. primary o secondary LEAKAGE rate observed in each SG (if it is notpractical to assign the LEAKAGE to an individual SG, the entire primary tosecondary LEAKAGE should be conservatively assumed to be from oneSG) during the cycle preceding the inspection which is the subject of thereport;

i. Fgrowino ion ofoinsppion 6' eior in Refu•ling tag8and y on forme in subs quent op6ratincyclehe//

ecalculated accident induced leakage rate rom the portion of the u esbelow I inches from the top of the tubesheet for the most limitingaccident in the most limiting SG. In addition, if the calculated accidentinduced leakage rate from the most limiting accident is less than 2.50times the maximum operational primary to secondary leak rate, the reportshould describe how it was determined; and

j. Fowin mpl tion p•an in ecti pe•meoin Re/elin utn 1ains ction'perfo(ied the sdbsedqcent ol•ratir cyct ) th6

results of monitoring for the tube axial displacement (slippage). Ifslippage is discovered, the implications of discovery and corrective actionshall be provided.

Wolf Creek-Unit 1 5.0-28 Amendment No. 123, 12, 158, 164,17-,,478,,79, , 195

Attachment III to ET 12-0002Page 1 of 3

ATTACHMENT III

Proposed Changes to Technical Specification Bases (for information only)

Attachment III to ET 12-0002Page 2 of 3 SG Tube Integrity

B 3.4.17

BASES

APPLICABLE The steam generator tube rupture (SGTR) accident is the limiting designSAFETY basis event for SG tubes and avoiding an SGTR is the basis for thisANALYSES Specification. The analysis of an SGTR event assumes a bounding

primary to secondary LEAKAGE rate equal to the operational LEAKAGErate limits in LCO 3.4.13, "RCS Operational LEAKAGE," plus the leakagerate associated with a double-ended rupture of a single tube. Theaccident analysis for an SGTR assumes the contaminated secondary fluidis released to the atmosphere via SG atmospheric relief valves and safetyvalves.

The analysis for design basis accidents and transients other than anSGTR assume the SG tubes retain their structural integrity (i.e., they areassumed not to rupture.) In these analyses, the steam discharge to theatmosphere is based on the total primary to secondary LEAKAGE from allSGs of 1 gallon per minute or is assumed to increase to 1 gallon perminute as a result of accident induced conditions. For accidents that donot involve fuel damage, the primary coolant activity level of DOSEEQUIVALENT 1-131 is assumed to be equal to the LCO 3.4.16, "RCSSpecific Activity," limits. For accidents that assume fuel damage, theprimary coolant activity is a function of the amount of activity releasedfrom the damaged fuel. The dose consequences of these events arewithin the limits of GDC 19 (Ref. 2), 10 CFR 100 (Ref. 3) or the NRCapproved licensing basis (e.g., a small fraction of these limits).

Steam generator tube integrity satisfies Criterion 2 of 10 CFR50.36(c)(2)(ii).

LCO The LCO requires that SG tube integrity be maintained. The LCO alsorequires that all SG tubes that satisfy the repair criteria be plugged inaccordance with the Steam Generator Program.

During a SG inspection, any inspected tube that satisfies the SteamGenerator Program repair criteria is removed from service by plugging. Ifa tube was determined to satisfy the repair criteria but was not plugged,the tube may still have tube integrity.

In the context of this Specification, a SG tube is defined as the entirelength of the tube, including the tube wall, between the tube-to-tubesheetweld at the tube inlet and the tube-to-tubesheet weld at the tube outlet.

.Reeli ut 187 dthe 9b ueCii:, ýoerOatng ce< -imalternate repair criterion or the portion of the tube below inches from

A the top of the tubesheet is specified in TS 5.5.9c.1. (Ref. 7 The tube-to-tubesheet weld is not considered part of the tube.

QL, 211

Wolf Creek - Unit 1 B 3.4.17-2 Revision 52

Attachment III to ET 12-0002Page 3 of 3 SG Tube Integrity

B 3.4.17

BASES

REFERENCES 1.

2.

3.

4.

5.

6.

7.

NEI 97-06, "Steam Generator Program Guidelines."

10 CFR 50 Appendix A, GDC 19.

10 CFR 100.

ASME Boiler and Pressure Vessel Code, Section III, SubsectionNB.

Draft Regulatory Guide 1.121, "Basis for Plugging Degraded SteamGenerator Tubes," August 1976.

EPRI, "Pressurized Water Reactor Steam Generator ExaminationGuidelines."

License Amendment No. (18W ctr ý13200.)

Wolf Creek - Unit 1 B 3.4.17-7 Revision 44

Attachment IV to ET 12-0002Page 1 of 4

ATTACHMENT IV

Retyped Technical Specification pages




3. The operational LEAKAGE performance criterion is specified inLCO 3.4.13, "RCS Operational LEAKAGE."

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


1. Tubes with service-induced flaws located greater than 15.21inches below the top of the tubesheet do not require plugging.Tubes with service-induced flaws located in the portion of the tubefrom the top of the tubesheet to 15.21 inches below the top of thetubesheet shall be plugged upon detection.

(continued)

Wolf Creek - Unit 1 5.0-12 Amendment No. 123, 153, 17"2, 178,486,.1-9,




d. Provisions for SG tube inspections. Periodic SG tube inspections shall beperformed. The number and portions of the tubes inspected and methodsof inspection shall be performed with the objective of detecting flaws ofany type (e.g., volumetric flaws, axial and circumferential cracks) that maybe present along the length of the tube, from the tube-to-tubesheet weldat the tube inlet to the tube-to-tubesheet weld at the tube outlet, and thatmay satisfy the applicable tube repair criteria. The portion of the tubebelow 15.21 inches from the top of the tubesheet is excluded from thisrequirement. The tube-to-tubesheet weld is not part of the tube. Inaddition to meeting the requirements of d.1, d.2, and d.3 below, theinspection scope, inspection methods, and inspection intervals shall besuch as to ensure that SG tube integrity is maintained until the next SGinspection. An assessment of degradation shall be performed todetermine the type and location of flaws to which the tubes may besusceptible and, based on this assessment, to determine which inspectionmethods need to be employed and at what locations.


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

3. If crack indications are found in any portion of the SG tube notexcluded above, then the next inspection for each SG for thedegradation mechanism that caused the crack indication shall notexceed 24 effective full power months or one refueling outage(whichever is less). If definitive information, such as fromexamination of a pulled tube, diagnostic non-destructive testing, orengineering evaluation indicates that a crack-like indication is notassociated with a crack(s), then the indication need not be treatedas a crack.

e. Provisions for monitoring operational primary to secondary LEAKAGE.

(continued)

Wolf Creek - Unit 1 5.0-13 Amendment No. 123, 163, 172, 178,186, 19 ,

Reporting Requirements5.6

5.6 Reporting Requirements

5.6.10 Steam Generator Tube Inspection Report

A report shall be submitted within 180 days after the initial entry into MODE 4following completion of an inspection performed in accordance with theSpecification 5.5.9, 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 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 andin-situ testing;

h. The primary to secondary LEAKAGE rate observed in each SG (if it is notpractical to assign the LEAKAGE to an individual SG, the entire primary tosecondary LEAKAGE should be conservatively assumed to be from oneSG) during the cycle preceding the inspection which is the subject of thereport;

The calculated accident induced leakage rate from the portion of thetubes below 15.21 inches from the top of the tubesheet for the mostlimiting accident in the most limiting SG. In addition, if the calculatedaccident induced leakage rate from the most limiting accident is less than2.50 times the maximum operational primary to secondary leak rate, thereport should describe how it was determined; and

j. The results of monitoring for the tube axial displacement (slippage). Ifslippage is discovered, the implications of discovery and corrective actionshall be provided.

Wolf Creek - Unit 1 5.0-28 Amendment No. 123, 142, 158, 164,178, 179, 6, 8,196,

Attachment V to ET 12-0002Page 1 of 2

ATTACHMENT V

List of Regulatory Commitments

Attachment V to ET 12-0002Page 2 of 2

LIST OF REGULATORY COMMITMENTS

The following table identifies those actions committed to by Wolf Creek Nuclear Operating Corporationin this document. Any other statements in this letter are provided for information purposes and are notconsidered regulatory commitments. Please direct questions regarding these commitments to Mr.Gautam Sen, Manager Regulatory Affairs at Wolf Creek Generating Station, (620) 364-4175.

REGULATORY COMMITMENT DUE DATE

For the condition monitoring (CM) assessment, the component of Implementation ofleakage from the prior cycle from below the H* distance will be multiplied Amendmentby a factor of 2.50 and added to the total leakage from any other sourceand compared to the allowable accident induced leakage limit. For theoperational assessment (OA), the difference in the leakage between theallowable leakage and the accident induced leakage from sources otherthan the tubesheet expansion region will be divided by 2.50 andcompared to the observed operational leakage. An administrative limitwill be established to not exceed the calculated value.

WCNOC will monitor for tube slippage as part of the steam generator Implementation oftube inspection program. The results of this monitoring will be included Amendmentin the report required by TS 5.6.10j.

Attachment VI to ET 12-0002Page 1 of 4

ATTACHMENT VI

Response to Request for Additional Information Questions Specific to Wolf CreekNuclear Operating Corporation


Response to Request for Additional Information Questions Specific to Wolf CreekNuclear Operating Corporation

On June 30, 2011, Duke Energy submitted a license amendment request (Reference 28) forpermanent application of the alternate repair criterion H* at Catawba Unit 2 based on thetechnical justification in WCAP-17330-P, Revision 1, "H*: Resolution of NRC Technical IssueRegarding Tubesheet Bore Eccentricity (Model F/Model D5)." A supplement (Reference 29) tothe license amendment request was submitted on July 11, 2011 and provided WestinghouseElectric Company LLC LTR-SGMP-11-58, "WCAP-17330-P, Revision 1 Erratum." On January5, 2012, a request for additional information (Reference 30) was transmitted electronically toDuke Energy. Duke Energy responded to the request for additional information on January 12,2012 (Reference 31).

Subsequent to the Duke Energy license amendment request, Virginia Electric and PowerCompany (Dominion) submitted a license amendment request (Reference 32) for permanentapplication of the alternate repair criterion H* for Surry Power Station Units 1 and 2. OnJanuary 18, 2012, the NRC issued a request for additional information (Reference 33).Dominion responded to the request for additional information on February 14, 2012 (Reference34).

Westinghouse Electric Company LLC, LTR-SGMMP-1 1-28 Rev.1 P-Attachment (Enclosure II),"Response to USNRC Request for Additional Information Regarding the License AmendmentRequests for Permanent Application of the Alternate Repair Criterion, H*, to the Model D5 andModel F SGs," augments the responses to the Duke Energy request for additional informationto include similar responses applicable to Model F steam generators. Additionally, LTR-SGMMP-1 1-28 Rev.1 P-Attachment addresses the Dominion request for additional informationquestion 14 for the Model F steam generators. Provided below are Wolf Creek NuclearOperating Corporation (WCNOC) specific responses to questions 12 and 13 from the DukeEnergy request for additional information and question 15 from the Dominion request foradditional information. The NRC question is identified in italics.

12. BET measurements for Catawba 2, documented in Westinghouse letter LTR-SGMP-09-111 P-Attachment, Revision 1, range to a maximum of 0.65 inches and appear not to be afactor affecting the H* and leak rate ratio calculations. Apart from tubes with this reportedrange of BETs, are there any non-expanded or partially expanded tubes at Catawba 2? Ifso, provide revisions to the proposed technical specifications which exclude such tubesfrom the proposed H* provisions.

Response: Bottom expansion transition (BET) measurements for WCGS, documented inWestinghouse letter LTR-SGMP-09-1 11 P-Attachment, Revision 1, range to a maximum of 0.66inches. Apart from tubes with this reported range of BETs, there are no non-expanded orpartially expanded tubes in service at Wolf Creek Generating Station (WCGS). WCNOC letterET 06-0004 (Reference 1) identified a tube (R1 1, C121) in steam generator "B" that was notpreviously expanded into the hot leg tubesheet. During Refueling Outage 16 (Spring 2008) thetube was removed from service by rolling approximately 2 inches from the tube end andinstalling a mechanical plug. The one-time verification of the tube expansion to locate anysignificant deviations in the distance from the top of the tubesheet to the bottom of theexpansion transition did not identify any additional non-expanded or partially expanded


tubes. As such, revision to the technical specifications to exclude such tubes from theproposed H* provisions is not required.

13. Proposed TS 5.6.8.h through j - The proposed changes contain more words than seemnecessary, reducing the clarity of the proposed reporting requirements. For example, theproposed wording refers to "an inspection performed after each refueling outage" whichdoesn't seem to make sense. The NRC staff believes the proposed requirements can bestated more clearly and concisely as follows:

h. For Unit 2, fo!Wowing ,,mplctin of annpcton ,.,,,perf..m.d dur;g End of Cyclc 1-7Refucling Outage (and any knsections performed durig subsequent Cycle 18 operation)the primary to secondary LEAKAGE rate observed in each steam generator (if it is notpractical to assign the leakage to an individual SG, the entire primary to secondaryleakage should be conservatively assumed to be from one SG) during the cycle precedingthe inspection which is the subject of the report,

i. For Unit 2, fo" owing completion of an in....tion pe; ormed during the End of cyle 1,Refueling Outage (and any insectionS pcdormed during subsequent cyclc 18 opera"in~the calculated accident induced leakage rate from the portion of the tubes below 20 14.01inches from the top of the tubesheet for the most limiting accident in the most limiting SG.In addition, if the calculated accident leakage rate from the most limiting accident is lessthan 3.27 times the maximum primary to secondary LEAKAGE rate, the report shalldescribe how it was determined, and

j. For Unit 2, following completion of an kinection pe•formed durg the End of GycOe 1/Refuieling Outage (and any ins pcctins pcrfermcd durig subsequent Cycile 18 operation),the results of monitoring for tube axial displacement (slippage). If slippage is discovered,the implications of the discovery and corrective action shall be provided.

Provide revisions to the proposed reporting requirements as necessary to clarify theirintent.

Response: The WCNOC proposed changes to technical specification (TS) 5.6.10, "SteamGenerator Tube Inspection Report," in Attachment II are consistent with the NRC staff'srecommendation above.

15. Verify that regulatory commitments pertaining to monitoring for tube slippage and forprimary to secondary leakage, as described in Dominion letter dated December 16, 2010(NRC ADAMS Accession No. ML103550206), Attachment 1, page 10 of 23, remain inplace. In addition, revise the proposed amendment to include a revision to technicalspecification limit on primary to secondary leakage from 150 gallons per day (gpd) to 83gpd (150 divided by the proposed 1.8 leakage factor), or provide a regulatory basis for notmaking this change.

Response: The regulatory commitments pertaining to monitoring for tube slippage and forprimary to secondary leakage as described in WCNOC letter ET 09-0025 dated September 15,2009 (Reference 2) remain in place as specified in Attachment V. WCNOC is not proposingany changes to the primary to secondary LEAKAGE limit as specified in TS 3.4.13, "RCSOperational LEAKAGE," based on the following.


Primary-to-secondary leakage from tube degradation is assumed to occur in severaldesign basis accidents: feedwater line break (FLB), steam line break (SLB), locked rotor,and control rod ejection. The radiological dose consequences associated with thisassumed leakage are evaluated to ensure that they remain within regulatory limits (e.g. 10CFR Part 100, 10 CFR 50.67, GDC 19). The accident induced leakage performancecriteria are intended to ensure the primary-to-secondary leak rate during any accidentdoes not exceed the primary-to-secondary leak rate assumed in the accident analysis.Radiological dose consequences define the limiting accident condition for the H* Analysis.

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," are satisfieddue to the constraint provided by the tubesheet. Through application of the limitedtubesheet inspection scope as described below, the existing operating leakage limitprovides assurance that excessive leakage (i.e., greater than accident analysisassumptions) will not occur. The accident induced leak rate limit for WCGS is 1.0 gpm.The TS 3.4.13, "RCS Operational LEAKAGE," operational leak rate limit is 150 gpd (0.1gpm) through any one steam generator. Consequently, accident leakage is approximately10 times the allowable leakage, if only one steam generator is leaking. Using a SLB/FLBoverall leakage factor of 2.50, accident induced leakage is less than 0.6 gpm, if all 4steam generators are leaking at 150 gpd at the beginning of the accident. Therefore,significant margin exists between the conservatively estimated accident induced leakageand the allowable accident leakage (1.0 gpm).

References:

1. WCNOC letter ET 06-0004, "Revision to Technical Specification (TS) 5.5.9, "SteamGenerator (SG) Program"," February 21, 2006. (ADAMS Accession No. ML060600456)

2. WCNOC letter ET 09-0025, "Revision to Technical Specification (TS) 5.5.9, "SteamGenerator (SG) Program," and TS 5.6.10, "Steam Generator Tube Inspection Report","September 15, 2009. (ADAMS Accession No. ML092730340)

Enclosure II to ET 12-0002

Enclosure II

Westinghouse Electric Company LLC "LTR'SGMMP-11-28 Rev.1 NP-Attachment andErrata, "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-proprietay)

(53 pages)

Westinghouse Non-Proprietary Class 3

S Westinghouse

To: D.H. WarrenP.J. McDonoughH. MahdavyD.L. RogoskyL.E. MarkleN. BahtishiC.L. MitchellD.C. Beddingfield

M.W. RyanJ.J. RobertsG.R. StrussionC.W. NitchmanA.M. MrazikS.J. HydeJ. Stepanic

C. D. Cassino

Date: March 20, 2012

cc: B. J. Bedont

From:

Ext:

Fax:

H.O. Lagally724-722-5082724-722-5889

Your ref:

Our ref: LTR-SGMMP-11-28Errata, Rev. 1

Subject: LTR-SGMMP-11-28, Revision 0 and Revision 1, P- and NP-Attachment Errata

Reference: 1. LTR-SGMMP-1 1-28, Rev.0, "Response to USNRC RAI on Catawba Unit 2 Permanent H*Submittal," January 4, 2012.

2. LTR-SGMMP- 11-28, Rev.1, "Response to USNRC RAI for Model D5 and Model F SGPermanent H* Submittals," February 2, 2012.

This letter supersedes LTR-SGMMP-11-28, Rev. 1 NP Attachment Errata, "LTR-SGMMP-11-28,Revision 1 NP Attachment Errata," dated March 13, 2012.

LTR-SGMMP- 11-28, Revision 0 (Reference 1) provides responses to an NRC Request for Additional

Information (RAI) specific to the Model D5 steam generators (SGs). LTR-SGMMP-1 1-28, Revision I(Reference 2) was issued to augment Revision 0 of the same letter to provide information specific to theModel F SGs in the response to the NRC RAI. References I and 2 contain both a proprietary (P) attachmentand a non-proprietary (NP) attachment for the responses to the RAI.

For Revision 0 of LTR-SGMMP-1 1-28, the following corrections apply:

* On page 31 of both the P-Attachment and the NP-Attachment, the title of the Appendix A cover page

should be LTR-SGDA- 11-87 instead of LTR-SGMP- 11-87.

For Revision I of LTR-SGMMP-1 1-28, the following corrections apply:

" On page 34 of both the P-Attachment and the NP-Attachment, the title of the Appendix A cover page

should be LTR-SGDA-1 1-87 instead of LTR-SGMP-1 1-87.

* On page 39 of the NP-Attachment, the table number should be Table 2 instead of Table 15. The tableis properly numbered in the P-Attachment.

The technical content and the conclusions of the References I and 2 are unaffected.

of 2Our ref: LTR-SGMMP-11-28Errata, Rev. I

Electronically Approved*Prepared by: H. 0. Lagally

Steam Generator ManagementAnd Modification Programs

Electronically Approved*

Electronically Approved*Verified: G.W. Whiteman

Regulatory Compliance

Approved by: Damian A. Testa, Manager

Steam Generator ManagementAnd Modification Programs

*Electronically approved records are authenticated in the electronic document management system.

Westinghouse Electric Company LLC1000 Westinghouse Drive

Cranberry Township, PA 16066

© 2012 Westinghouse Electric Company LLCAll Rights Reserved

Westinghouse Non-Proprietary Class 3LTR-SGMMP- 11-28 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

LTR-SGMMP-1 1-28 Rev. 1 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-1 7345-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-1 7330-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

LTR-SGMMP- 11-28 Rev. 1 NP-Attachment

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 theModel 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- 1 7330-P, Revision 1 - 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 thatcorrespond 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 thecase, 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-17330-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-17330-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

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

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 cx, 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

cx, 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 1 ofWCAP-I 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. I NP-Attachment

Table 2-1

Reduced Response Surface; Model D5, 26.703 inches Radius

TS CTE T CTECase # H*+BET

(in) a,c,e

2

3

4

5

67

8

9

10

11

12

13

14

1516

17

18

19

20

21

22

23

24

25

26

27

28

29

30

31

32

33

34

35

36

37

3839 _____

5


4041

42

43

44

45 L

a,c,e

Question 3:

WCAP-1 7330-P, Revision 1 - 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 limiting 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- 1 7345-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.ce 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 CTE

variations. 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 0 th 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.

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

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 C2

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 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 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 9500th 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* withthe C2 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


Table 3-1Reduced Response Surface; Model D5, [ ]ac"e inches Radius

TS CTE T CTE H*+BETCase # n o n o(in)

Q[ ]""€'Radius)

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

a,c,e

9


39

40

41

42

43

44

45 L

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-17330-P, Revision 1 contains the data for the Model F SGs, centered on rank order9890.

Table 4-1 provides the requested information.

Table 4-1Variation of CTEs Over a Range of Rank Order Statistics for Model D5

Rank Tube Tubesheet Alpha(1 )

CTE CTE

9526 _

9527

9528

9529

9530

9531

9532

9533

9534

9535

9536

9537

9538

9539

9540

9541

9542

9543

9544

9545

9546

Notes:

1. Defined as SQRT((Tube CTE)A2 + (Tubesheet CTE)A2)

a,c,e

II


Question 5:

WCAP-17330-P, Revision 1, Table 3-29 - Provide C2 H* values for rank orders 9888and 9892. This will lend additional confidence to inferences drawn from this table onpage 3-58. In addition, provide a similar table applicable to the Model D5 SLB case.

Response:

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

Analysis code note: The structural code employed for the prior H* calculations was ANSYSWorkbench, Version 11. Version 12.1 of ANSYS Workbench was released following theissue of WCAP-17330-P, Revision 1. The updates to this version of ANSYS Workbenchinclude changes to the contact modelling and solver options. Westinghouse hasbenchmarked and configured this version of the ANSYS code and has verified the resultsand conclusions of the previous H* analyses obtained with Version 11. However, there areminor numerical differences in the results. The net difference of applying version 12.1 of theANSYS code compared to version 11 of the ANSYS code is a slight variation in the averagecircumferential contact pressure, typically on the order of ± 40 psi. Version 11 generallyproduces the lower contact pressures. Consequently, there may be small differences in thevalues provided for points already included in WCAP-17330-P, Revision 1.

Table 5-1 provides the requested additional probabilistic Model F NOP results at a [ ]a,c,e

inch radius for rank orders 9888 and 9892. Table 5-2 provides the requested probabilisticModel D5 SLB results at an [ ]a.ce inch radius for rank orders from 9533 through 9539.

Table 5-1: Model F NOP Results at [ ]a.,ce inches

Variation Input

MC T CTE TS CTE C2 H*# no" ma in.

9888 ]a,c,e [ ]a,c,e [ ]a,c,e

9892 ]a,c,e [ ] . [ ]a,c,e

12


Table 5-2: Model D5 SLB Results at [ ]'c' inches

Variation Input

MC T CTE TS CTE C2 H*

# no mo" in.9533 [ ]a,c,e [ ] a,c,e [ ]a,c,e

9534 a,c,e [ ]a,c,e [ ]a,c,e

9536 a,c,e ]a,c,e [ ]ace(1)

9 53 8 [ ] ace ]a.c.e [ ]a.c e

9539 [ a,c,e [ ]a,c,e [ ]a,,e

Notes:(1) Refer to LTR-SGMP-11-58, "WCAP-17330-PRevision 1 Erratum"

Although the uncertainty in the narrow range of rank order H* values for the Model D5(Table 5-2) is slightly larger than the uncertainty for the Model F (Table 5-1 and Table 3-29 ofWCAP-17330-P, Rev. 1), the inferences drawn from these data on page 3-56 ofWCAP-1 7330-P, Rev. 1 remain valid. It is expected that small variations will occur due tofactors such as variation in extremely small absolute values of the structural displacements(e.g., due to round-off effects) that are the inputs to the C2 model. This uncertainty is on theorder of 2% of the final H* value, which is more than adequately covered by otherconservatisms in the H* value that are discussed in the responses to the other questions.

13


Question 6:

WCAP- 1 7330-P, Revision 1, Figure 3-45 - Should the data corresponding to the twoopen symbols be labeled as "data used in probabilistic analysis" (consistent with Figure3-44) instead of "reduced data?" Why does this figure show only two open symbolsrather than three as are given in Figure 3-44?

Response

The question is specific to the Model D5 SGs and does not apply for the Model F SGs. Thisquestion was not included in Reference 4 for the Model 51 F SGs.

For clarity, the two (three) open symbols on Figure 3-45 of WCAP-1 7330-P, Revision 1,should be labelled the same as the three open symbols in Figure 3-44 of the report. Nodifferentiation of meaning was intended in the current labelling.

On Figure 3-45 of WCAP-17330-P, Revision 1, the two apparent open symbols are, in fact,three open symbols. Two of the points are closely overlaid, leading to the impression thatthere are only two points. For clarity, the Table 6-1 provides the coordinates of the threepoints on Figure 3-45 of WCAP-17330-P. Figure 6-1 is an update of Figure 3-45 of WCAP-17330, Revision 1 that shows the previously overlaid data points as an open triangle and adark grey square.

Table 6-1Coordinates of Three Open-symbol Points on

Figure 3-45 of WCAP-17330-P, Revision I

Rank H* Tube CTE Tubesheet AlphaCTE

9149 [ ]a,c,e [ ]a,c,e [ ]a,c,e 3.513

9500 ]]ac,e [ ]a,c,e 3.7509536 ]a.,c,e [ ]a,c,e ]a,c,e 3.733

14


ac,e

Figure 6-1Update of Figure 3-45 of WCAP-17330, Revision I

15


Question 7:

WCAP-1 7330-P, Revision 1, Tables 3-35 to 3-48 - The numerical methods used togenerate the accumulated pullout loads in these tables appear to contain two sources ofnon-conservatism. One, the distance below the top of the tubesheet (TTS) where thecontact pressure transitions from zero to a positive non-zero value is assumed to be thelowermost elevation for which a C2 calculation was performed and yielding a zero valuecontact pressure. The staff believes a more realistic and more conservative estimate ofthe contact pressure zero intercept value can be obtained by extrapolating the C2 resultsat lower elevations to the zero intercept location. Two, the method used to interpolatethe H* distance between specific locations where C2 analyses were performed assumesthat the distribution of contact pressure between these locations is a constant valueequal to average value between these locations. For Table 3-35, the staff estimates thatelimination of the non-conservatisms increases the calculated H* by 0.34 inches. ForTables 3-46 and 3-48, H* increases by 0. 15 inches. These are not trivial differences.The staff estimates that the pullout loads corresponding to the H* distances in Figures 3-35, 3-46, and 3-48 are overestimated by 17%, 6%, and 8%, respectively. Providerevisions to Tables 3-35 to 3-48, if and as needed, to address the staffs concern.

Response

This question and the response apply for both the Model D5 and Model F SGs.

Linear extrapolation of data points to determine a presumed zero contact pressure intercept,while conservative, is not realistic. The addition of a number of data points in the Model D5contact pressure curve showed that extrapolation of data points provided in WCAP-17330-P,Revision 0 was unrealistically conservative. While a higher point density would alwaysprovide more certainty in the result, the current density of points was judged adequate byWestinghouse and (implicitly) by MPR in their independent review of H* methodology basedon the minor effect on H*. In response to this question, another point was added to thecontact pressure curve for the Model D5 (Figure 3-20 of WCAP-1 7330-P, Revision 1)between the last zero point and the first non-zero point; the result is shown in Figure 7-1below. Figure 7-1 shows that the extrapolation proposed by the question is unrealisticallyconservative and that such an extrapolation is also inconsistent with the behavior of a realstructure. A sharp break in the contact pressure curve would not be expected in the physicalstructure; rather, a smooth transition from zero to non-zero contact pressure would beexpected. Figure 7-1 shows that addition of even more points would simply further define thesmooth transition in the curve as would be expected.

A similar result would be expected for the Model F SGs (Figure 3-26 of WCAP-1 7330-P,Revision 1).

16


a,c,e

Figure 7-1Model D5 Contact Pressure Profile with Added Point

Calculation of Conservatism in CTE Variances Used in Probabilistic Analysis

The CTE variances used in the probabilistic analysis were derived from a large set of

heterogeneous data across a broad range of temperatures. Since the issuance of the first H*

report, further analysis of CTE data at specific temperatures has been performed in LTR-

SGDA-1 1-87 in response to a question from the independent review by MPR Associates(Reference 5). (LTR-SGDA-1 1-87 is Reference 3-17 in WCAP-17330-P, Revision 1 and is

provided as Appendix A in this document.) The additional statistical analysis was performed

on the data to extract instrumentation uncertainty contributions (at high-confidence levels).Table 7-1 compares the values used in the analysis with the values from the more recent

statistical analysis. Values are listed at 300° and 6000, the values pertinent to the Model Fand D5 limiting conditions. As can be seen, the more accurately calculated values are

significantly lower than those used in the current technical justification of H*.

The effect of applying the more realistic CTE variations on H* can be estimated by

considering the ratio by which the standard deviations have been reduced. Since thedifference between the mean H* and the probabilistic H* is entirely based on CTEdifferences, a first-order approximation to the reduction in H* length that would result from

using the refined CTE variances can be obtained by multiplying the difference between the

current mean and probabilistic H*'s by the above ratio. For conservatism, the more limiting ofthe tube/tubesheet CTE variance ratios from Table 7-2 were used.

17


Table 7-3 summarizes the H* values contained in WCAP-1 7330, Revision 1 for the Model D5and Model F SGs and serves to provide the input for Table 7-4.

Table 7-4 shows the effects of applying the improved CTE variability values to the H*analysis. Note that the H* values in Table 7-4 do not include crevice pressure or Poissoncontraction because neither of these are related to CTE. As can be seen from Table 7-4, theexisting H* length for the Model F's is conservative by approximately [ ]ace inches and the H*length for the Model D5's is conservative by about [ ]a,c,e inch. This shows that theconservatism inherent in the current H* calculations are adequately conservative to accountfor small differences in judgment on the calculation process even without considering themajor conservatisms identified previously (i.e., neglecting residual contact pressure).Additional conservatism to further support this conclusion is identified below.

Table 7-1CTE Values Without Instrumentation Error

Tube CTE SDs, %

Temperature As Used in Improved 50% Improved 95%(F) WCAP- Confidence Confidence

17330,Rev. 1

300 2.33 [ ]a,c,e [ ]a,c,e

600 2.33 [ ]a,c,e [ ]a,c,e

Tubesheet CTE SDs, %Temperature As Used in Improved 50% Improved 95%

(F) WCAP- Confidence Confidence

17330,Rev. 1

300 1.62 [ ]a,c,e [ e

600 1.62 [ ]a,c,e [ ]a,c,e

18


Table 7-2Ratio of CTE Variances (Refined/Used in Current H*)

Tube CTE SDs RatiosTemperature

(*F) 95% ConfidenceConfidence

300 [ ]a,c,e [ ].,ce

600 [ ]a,c,e [ ] a,c,,e

Table 7-3Summary of H* Lengths from WCAP-17330, Revision 1

LimitingMean H* Probabilistic H* Difference, Ratio f(inches) (inches) Probabilistic - Mean Table 7-2

Table 7-2

F, 95/50 Whole

Bundle _

F, 95/95 Whole

Plant

D5, 95/50Whole Bundle

D5, 95/95

Whole Bundle

Table 7-4

Estimate of Conservatism of H* Length Related to CTE Variance

DifferenceDifference x DfeecModel/Case Dimitinc R New Probabilistic H* (Licensed H* - New

F,_95/50_WholeLimiting Ratio Probabilistic H*)F, 95/50 Whole

BundleF, 95/95 Whole

Plant

D5, 95/50 Whole

Bundle

D5, 95/95 Whole

Bundle

19


Question 8:

WCAP- I 7330-P, Revision 1, Figures 3-48 and 3-49 - These figures were generated withthe thick shell model. Were "spot checks" performed with the C2 model to determinewhether adjustments to the curves in these figures are needed to approximate what thecurves would look like if entirely generated with the C2 model? If not, why are thecurves in their present form conservative?

Response

This response was modified to include both the Model D5 and Model F SGS.

The Model D5 contact pressure results reported for the steam line break (SLB) condition andthe Model F contact pressure results for the normal operating (NOP) conditions in WCAP-17330-P, Revision 1 are conservative with respect to the crevice pressure distribution. Thecontact pressure distributions developed in WCAP-17330-P, Revision 1 assume that thecrevice pressure is distributed over the full depth of the tubesheet. No "spot checks" wereperformed to test if the crevice pressure correction distribution, determined by the thick shellequations (shown in Figures 3-48 and 3-49 of WCAP-17330, Revision 1), required anadjustment when applied to the C2 model results. The adjustment to the final H* length inTables 3-50 and 3-51 of WCAP-1 7330-P, Revision 1 was made to be consistent with themethodology described in WCAP-17072-P.

The contact pressure results based on application of the C2 model already represent apractical worst case with respect to crevice pressure, therefore, any further adjustment to theH* value using the curves shown in Figures 3-48 and 3-49 of WCAP-1 7330-P isunnecessary. The basis of this conclusion is explained below.

As discussed in WCAP-1 7072-P, the crevice pressure distribution was proportionallyadjusted through the thickness of the tubesheet to reflect the predicted H* tube lengthbecause the tube below any postulated 3600, 100% through-wall flaw, is assumed to beabsent. The crevice pressure at, and below, the flaw depth is in equilibrium with the primaryside pressure. Increasing the crevice pressure over the length of the predicted H*so that it isequal to the primary side pressure reduces the tube to tubesheet contact pressure andincreases the length of H*. Conversely, reducing the crevice pressure over the length of H*increases the tube to tubesheet contact pressure and decreases the length of H*.

The current contact pressure results for the Model D5 SGs and the Model F SGs show thatthere is zero contact pressure for a short distance below the top of the tubesheet. The H*length and the leakage factors are calculated based on only the length of positive contactpressure. Therefore, the pressure in the crevice below the top of the tubesheet to the pointof departure from zero contact pressure experiences the full primary to secondary pressuredifferential because that length of crevice is at the secondary side pressure condition. Duringa Model D5 steam line break, this pressure differential is equal to 2560 psid, acting towardsthe tubesheet. For the Model F, during normal operating conditions, the pressure differentialis 1453 psid, acting toward the tubesheet.

20


Figure 8-1(a) shows a comparison of the unmodified crevice pressure distribution used in theC2 analysis (i.e., the crevice pressure is distributed over the full depth of the tubesheet) andthe crevice pressure distribution that has been adjusted to reflect the final contact pressuredistribution reported in Table 3-48 in WCAP-1 7330-P, Revision 1 for the critical radius in theModel D5 SG. Similarly, Figure 8-1(b) shows the same comparison for the Model F SGsbased on the data in Table 3-46 in WCAP-17330-P, Revision 1. In effect, the normalizationof the crevice pressure distribution must be based on the shorter distance defined by thedistance between the point of departure from zero-contact pressure to the predicted H*length (i.e., the location of the assumed flaw).

When the normalization length of the crevice is decreased, the pressure differential acrossthe tube over the H* length increases. The increased pressure differential results in a largeincrease in the contact pressure between the tube and the tubesheet at the upper portion ofthe tube in the C2 analysis. This effect was not included in the current analysis for H*because including it required iterating the probabilistic contact pressure distribution at bothends of the tube portion within the tubesheet with positive contact pressure between the tubeand the tubesheet. The double iteration significantly increases the time required to performthe analysis and it is conservative to neglect it. Including the effect of the increased pressuredifferential reduces the final H* distance by more than 1 inch for the Model D5 SGs.

Figures 8-2 (a and b) are plots of the contact pressure between the tube and the tubesheetusing the probabilistic results from Tables 3-41 and 3-42 in WCAP-17330-P, Revision 1 andthe adjusted crevice pressure distribution shown in Figures 8-1(a and b). The increase incontact pressure due to adjusting the crevice pressure at the top of the tubesheet occursregardless of the predicted length of H* if the underlying contact pressure distributionincludes a length of zero contact pressure at the top of the tubesheet. Therefore, neglectingthe crevice pressure distribution adjustment in the zero contact pressure length for anypredicted H* length provides additional margin to the calculation of H*. The conservativeapplication of crevice pressure distribution in the current analysis results in an under-prediction of the actual tube to tubesheet contact pressure by about 20% and in anoverestimate of the H* length by more than 1 inch, before the additional crevice pressureadjustment from Figures 3-49 and 3-48 in WCAP-1 7330-P, Revision 1 are addedrespectively for the Model D5 and Model F SGs.

Figures 8-3 (a and b) show that no adjustment to the final probabilistic contact pressuredistribution for crevice pressure distribution is necessary. The probabilistic contact pressuredistribution is the contact pressure profile that is determined by the C2 model when theprobabilistic values of inputs (CTEs, displacements) are input to the C2 model. Theunadjusted (for crevice length) crevice pressure differential distribution, when applied to theprobabilistic contact pressure distribution, results in a near-worst-case result for H* becausethe contact pressure is much less sensitive to crevice pressure variations than it is tovariations of the other input parameters such as temperature and pressure.

For example, at the critical radius in the Model D5 tubesheet ([ ]a,c,e inch), if the appliedtubesheet displacements and temperatures throughout the tubesheet depth are kept thesame as shown in Tables 3-10 and 3-16, respectively for the Model D5 and Model F SGs, in

21


WCAP-17330-P, Revision 1, but the crevice pressure differential is held constant at 1 psithroughout the depth of the tubesheet (i.e., primary pressure in the full length of the crevice),the result is the "DP=1 psi" curve in Figures 8-3(a and b). Similarly, if the C 2 model inputsare kept the same, but the crevice pressure differential is held constant at 2560 psid for theModel D5 throughout the depth of the tubesheet (i.e., secondary pressure in the crevice), theresult is the "DP=2560 psi" curve in Figure 8-3 (a). Likewise, if the C2 model inputs are keptthe same, but the crevice pressure differential is held constant at 1453 psid for the Model Fthroughout the depth of the tubesheet (i.e., secondary pressure in the crevice), the result isthe "DP=1453 psi" curve in Figure 8-3 (b).These are the bounding conditions for crevicepressure. It is not possible for variation in crevice pressure differential to produce a contactpressure distribution less than, or greater than, the space bounded by these two curves. Thecurrent probabilistic contact pressure distribution, with the unmodified crevice pressuredifferential, is also shown on Figures 8-3 (a and b) for the Model D5 and the Model F SGs,respectively. The difference between the contact pressure distribution with the unmodifiedcrevice pressure distribution used in WCAP-17330-P, Rev. 1, and the contact pressuredistribution with the worst-case assumption of a 1 psi differential, is essentially negligible forthe Model D5 and small for the Model F.

When the modified crevice pressure differential distribution (i.e., based on the shorter crevicelength) is applied, the result is increased contact pressure as illustrated in Figures 8-4(a andb). Increased contact pressure results in a reduced H* value. However, for consistency withthe H* calculation process established in WCAP-17072-P and WCAP-17071-P, the H*distance is increased by 1.51 inches for crevice pressure distribution in the current analysismethodology, not decreased as it should be from the results shown in Figure 8-4. Therefore,the 1.51 inches from the current crevice pressure adjustment shown in Figure 3-49 in WCAP-17330-P, Revision 1 represents excess conservatism for the Model D5. Similarly, the0.68 inch from the current crevice pressure adjustment shown in Figure 3-48 in WCAP-17330-P, Revision 1 represents excess conservatism for the Model F. Further refinement ofthe crevice pressure adjustment curve as it is applied in the C 2 analysis methodology is notrequired.

22


a,c,e

Figure 8-1(a): Model D5: Plot of Crevice Pressure Differential acting towards the tubesheet onthe inner diameter of the tube wall as a function of depth into the tubesheet. The zero (0)elevation is the top of the tubesheet.

a,c,e

Figure 8-1(b): Model F: Plot of Crevice Pressure Differential acting towards the tubesheet onthe inner diameter of the tube wall as a function of depth into the tubesheet. The zero (0)elevation is the top of the tubesheet.

23


a,c,e

Figure 8-2(a): Model D5: Plot of tube-to-tubesheet contact pressure for the modified andunmodified crevice pressure differential distributions shown in Figure A. The zero (0) elevationis the top of the tubesheet.

a,c,e

Figure 8-2(b): Model F: Plot of tube-to-tubesheet contact pressure for the modified andunmodified crevice pressure differential distributions shown in Figure A. The zero (0) elevationis the top of the tubesheet.

24


a,c,e

Figure 8-3(a): Model DS: Plot of tube-to-tubesheet contact pressure as a function of crevicepressure distribution. The zero (0) elevation is the top of the tubesheet.

a,c,e

Figure 8-3(b): Model F: Plot of tube-to-tubesheet contact pressure as a function of crevicepressure distribution. The zero (0) elevation is the top of the tubesheet.

25


a,c,e

Figure 8-4(a) Model D5: Composite plot showing the effect on contact pressure of adjustingcrevice pressure distribution to account for zero contact pressure near the top of thetubesheet.

a,c,e

Figure 8-4(b) Model F: Composite plot showing the effect on contact pressure of adjustingcrevice pressure distribution to account for zero contact pressure near the top of thetubesheet.

26


Question 9:

In addition to the potential non-conservatisms in the H* estimate discussed in Question 7above, there is uncertainty associated with the computed probabilistic H* valuescalculated with the C2 model as illustrated in Table 3-29. Depending on the response toquestion 8 above, there also may be some uncertainty associated with the H*adjustments for the crevice pressure distribution. What change to the proposed H* valueof 14.01 inches is needed to ensure that it is a conservative value?

Response:

The responses to RAI 7 and RAI 8 indicate that no adjustments to the Model D5 and Model Fprobabilistic H* estimates are necessary to account for the uncertainty associated with the C2

model results shown in Table 3-29 of WCAP-17330-P, Revision 1. The current Model D5 H*estimate of 14.01 inches is conservative by approximately 3.5 inches compared to thetechnically justifiable value. The current Model F H* estimate of 15.21 inches is conservativeby approximately 5.5 inches compared to the technically justifiable value. These marginsare in addition to the significant conservatism of neglecting residual contact pressure andother conservatism identified previously.

For the Model D5 SGs, the probabilistic H* value, before any adiustments, cited in Table 3-49in WCAP-17330-P, Rev. 1 is [ ]a"' inches. The probabilistic H* value for the contactpressure distribution shown in the response to Question 8, Figure 8-2(a), is [ ]ace inches.

For the Model F SGs, the probabilistic H* value, before any adiustments, cited in Table 3-49in WCAP-17330-P, Rev. 1 is [ ac"e inches. The probabilistic H* value for the contactpressure distribution shown in the response to Question 8, Figure 8-2(b), is [ ]a,c,e

inches.

Table 9-1 and Table 9-2 summarize the adjustments to the probabilistic H* estimatecompared to the adjustments that are demonstrated above in the current technical basis forH*. It is seen from Table 9-1 that a margin of [ ]a.c~e inches exists in the currentlyrecommended H* length of 14.01 inches for the Model D5 SGs when the conservatism in thecrevice pressure adjustment and the measurement error in the CTE data are quantified andthe proper adjustments are made. Table 9-2 shows that a margin of [ ]a,c,e exists in thecurrently recommended H* length of 15.21 inches for the Model F when the conservatism inthe crevice pressure adjustment and the measurement error in the CTE data are quantifiedand the proper adjustments are made. These previously un-quantified conservatismssignificantly exceed the potential increase in the H* length if different judgments are made inthe details of the H* calculation as suggested in Questions 7, 8 and 9. Based on this, it isconcluded that no adjustments to the recommended probabilistic H* value of 14.01 inches forthe Model D5 SGs and 15.21 inches for the Model F SGs are necessary and that the H*lengths recommended in WCAP-17330-P, Revision 1 are significantly conservative.

27


Table RAI 9-1Conservatism in Current Model D5 H* Calculation

WCAP-17330-P, RefinedSource Rev 1 Calculations

_ in in-

Unmodified H* Value

Adjustments

Poisson CorrectionCrevice Pressure and BETAdjustment

CTE Uncertainty Adjustment (RAI7)

Total Adjustments

Final Probabilistic H* 14.01 [ ]a,c,e

Notes:(1) Recalculated for [ ]a,c,e inches H* based on Figure 8-2(a).(2) Crevice pressure margin ([ ]a,c,e inch) plus BET adder of 0.3 inchincluded in Pcrev correction (Figure 3-49 of WCAP-17330, Rev. 1)(3) See response to Question 7.

Table RAI 9-2Conservatism in Current Model F H* Calculation

WCAP-17330-P, Refined

Source Rev 1 Calculations

_ in in c,e

Unmodified H* Value

Adjustments

Poisson Correction

Crevice Pressure and BET Adjustment

CTE Uncertainty Adjustment (RAI 7)

Total Adjustments

Final Probabilistic H* 15.21 [ ]a,c,e

Notes:(1) Recalculated for [ ]ace inches H* based on Figure 8-2(b).(2) Crevice pressure margin ([ ]a0,ce inch) plus BET adder of 0.3 inchincluded in Pcrev correction (Figure 3-48 of WCAP-17330, Rev. 1)(3) See response to Question 7.

28


Question 10:

Westinghouse letter LTR-SGMP-10-95 P - Attachment, Revision 1 - The staff is able toreasonably reproduce the numbers in Table 5 for Exp-2 and Power-2. It is the staff'sunderstanding that Table 4 contains intermediate results leading to the results in Table5. However, the staff cannot reproduce the numbers in Table 4 based on theinformation provided. Is Table 4 correctly titled? Provide a precise definition of theparameters that are listed in Table 4. Provide one example of how the parameter valueswere calculated, say for one segment at a tubesheet radius of 18.139 inches for SLB.

Response:

This response applies for all models of SG that are candidates for H*.

Table 4 in LTR-SGMP-10-95, Revision 1 is labelled correctly with regard to the definition ofthe loss coefficient function but it is based on the contact pressure results from the Thick-Shell model. Its inclusion in LTR-SGMP-1 0-95, Revision 1 is the result of a transcriptionerror.

Table 10-1, below, provides the local loss coefficients in units of (in-4) for the "Power-2"

function based on the contact pressure data contained in Table 3 of LTR-SGMP-1 0-95,Revision 1. The contact pressures in Table 3 of LTR-SGMP-1 0-95, Revision 1 are theaverage contact pressures over each segment length. The values on Table 10-1 are thesolution for K from the "Power-2" function.

Table 10-2, below, shows the segment resistances in units of (Ibf-sec/in 2) calculated from thelocal loss coefficients in Table 10-1, adjusted for units conversion and segment length. Thesegment lengths are shown on both Tables 10-1 and 10-2. Table 10-2 is the solution to theresistance equation, R = 121 iKI, but neglecting the constant because it divides out in thecalculation of the resistance ratios.

29


Table 10-1Local Loss Coefficient for Power 2 (K=0.15*(Pc) 4'5)

Segment Tubesheet RadiusLengths 4.437 10.431 18.139 26.703 42.974 49.825

from BTS toTTS Local K- NOP2.00 5.1313E+15 3.6865E+15 2.3659E+15 1.2689E+15 1.0700E+14 1.5672E+13

2.00 3.0747E+15 2.1831E+15 1.3670E+15 7.8175E+14 9.6690E+13 2.4449E+132.00 1.6627E+15 1.1207E+15 7.2723E+14 4.3233E+14 9.1542E+13 3.6160E+13

4.515 5.0019E+14 2.9683E+14 2.1225E+14 1.3996E+14 7.8376E+13 7.3598E+136.386 1.7653E+13 7.5284E+12 6.7741E+12 8.3479E+12 5.1448E+13 1.7803E+142.129 6.0972E+09 9.2123E+08 1.8742E+09 4.8467E+10 3.0885E+13 2.7622E+141.00 2.8981E+00 5.2512E-02 1.2442E-02 6.6444E+07 4.1304E+12 1.0078E+14

1.00 O.0000E+00 O.0000E+00 O.O000E+00 O.OOOOE+00 8.3625E+09 3.7119E+12

Local K-SLB2.00 5.5942E+16 4.9018E+16 3.4632E+16 2.0108E+16 2.2119E+15 2.3001E+14

2.00 2.5365E+16 2.2641E+16 1.6093E+16 9.3208E+15 1.2097E+15 1.8243E+14

2.00 9.6846E+15 8.8889E+15 6.3912E+15 3.7879E+15 6.2174E+14 1.4254E+14

4.515 1.0293E+15 1.0557E+15 7.8702E+14 5.3297E+14 1.7396E+14 9.0305E+13

6.386 3.1277E+12 4.0461E+12 3.2101E+12 2.8085E+12 1.5655E+13 7.4616E+13

2.129 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 1.0516E+12 9.0654E+13



30


Table 10-2

Segment Resistance Based on Viscosity in (Ibf-sec/inA2) Unitsfor Power 2 (K=0.15*(Pc) 4'5)

Segment Tubesheet RadiusLengths 4.437 10.431 18.139 26.703 42.974 49.825

from BTSto TTS Normal Operating Conditions2.00 1.19E+08 8.55E+07 5.49E+07 2.94E+07 2.48E+06 3.64E+05

2.00 7.13E+07 5.07E+07 3.17E+07 1.81E+07 2.24E+06 5.67E+05

2.00 3.86E+07 2.60E+07 1.69E+07 1.00E+07 2.12E+06 8.39E+05

4.515 2.62E+07 1.56E+07 1.11E+07 7.33E+06 4.11E+06 3.86E+06

6.386 1.31E+06 5.58E+05 5.02E+05 6.19E+05 3.81E+06 1.32E+07

2.129 1.51E+02 2.28E+01 4.63E+01 1.20E+03 7.63E+05 6.82E+06

1.00 3.36E-08 6.09E-10 1.44E-10 7.71E-01 4.79E+04 1.17E+06

1.00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 9.70E+01 4.31E+04

Steam Line Break Conditions2.00 3.06E+09 2.69E+09 1.90E+09 1.10E+09 1.21E+08 1.26E+07

2.00 1.39E+09 1.24E+09 8.82E+08 5.11E+08 6.63E+07 9.99E+06

2.00 5.31E+08 4.87E+08 3:50E+08 2.07E+08 3.41E+07 7.81E+06

4.515 1.27E+08 1.31E+08 9.73E+07 6.59E+07 2.15E+07 1.12E+07

6.386 5.47E+05 7.08E+05 5.61E+05 4.91E+05 2.74E+06 1.31E+07




31


Question 11

Westinghouse letter L TR-SGMP-10-95 P - Attachment, Revision 1 - This report spellsout the definition of Exp-2 and Power-2 in Table 5. Provide definitions of the otherfunctions considered in the table.

Response:

This response applies for all models of SG that are candidates for H*.

The following is a complete list of the functions with their definitions that were considered inLTR-SGMP-1 0-95, Revision 1. K is the loss coefficient as defined in Figure 1 of LTR-SGMP-10-95, Revision 1. As noted in LTR-SGMP-10-95, Revision 1, these functions are notmathematical fits to the data; rather, they are functions developed to represent variousinterpretations of the loss coefficient data.

Function Definition: Note

Exp-1 K = 1 E+1 2*exp(1.5E-03*Pc)

Exp-2 K = 3.5E+12*exp(5E-04*Pc)

Exp-3 K = 2E+12*exp(2E-04*Pc)

Exp-4 K = 6E+1 1*exp(8E-05*Pc) Lower Bound Horizontal

Exp-5 K = 1.1 E+14*exp(1.8E-04*Pc) Upper Bound Horizontal

Linear K = 6.5E+9*Pc

Power-I K = 1 E+4*PcA3

Power-2 K = 0.15*(Pc)45 Diagonal Bound

Logarithmic K = 1 E+12*In(Pc)+4E+08

Question 12

This question is a utility-specific question for which the respective utilities provide specificresponses.

Question 13

This question was a Catawba-2 - specific and does not apply to either the Model 51F or theModel F SGs.

32


Question 14

[LlfrAP 17•• F5 p, Re,.ien 2, Taob•6c 3 50 -2 3651 WCAP-17330-P, Revision I Table 3-50-Are Is the footnotes in • this table correct and complete? For Model 64F, Table 3-27 implies we have direct C2 calculations for rank orders 9025, •673, and 00019186, 9694and 9890. Thus, for Table 3-6450, it seems a4 three of four cases are based on interpolatedvalues. Sim..a.•!;, for -A-do! #4W F, Ta!b; 3 27 ifflp!!2 W.i'2 h•'. d-rot C 96!'-!ta .9..- for &Fn!&korders 0158, 0607, and 7 .Thus, fox T. .3 50, it not. ' ...... .....ey - is bacsed .f. dA.r-t -- - !2-- !2t..M.6 .nd th• other 2 2C 2r2 ..tcrpoiated '.'2!--i . If thestaff's understanding is incorrect, clarify for which rank orders direct C2 calculations wereperformed and provide the H* calculations for these cases in a form similar to Tables 3-45 to3-48.

Response

This question did not appear in Reference 2 for the Model D5 but did appear in Reference 4for the Model 51 F. With appropriate references in the question (see above), it can beconsidered to also apply for the Model F SGs.

The points that were directly calculated with the C2 model are shown on Figure 3-43 for theModel F SGs. The specific rank orders are identified in Table 3-30 of WCAP-1 7330-P,Revision 1. The range of rank orders defined by the three points for the Model F is 9186through 9890. Only one of the rank orders of interest, which define the key probabilistictargets in Table 3-50, is a point that was directly calculated using the C2 model (Model F,whole plant, 95/95). However, Figure 3-43 shows that the rank order in the range of interestis a straight line function. Consequently, because the points of interest lay within the rangeof calculated values, and the function is linear, it is appropriate to interpolate to determine theH* values.

Question 15

This question is specific to the Dominion LAR for H*. A similar question may apply for theModel F SGs in which case a response must be provided by the utility with Model F SGs thathas submitted an LAR for application of a permanent H* ARC.

33


Appendix ALTR-SGMP-11-87

(Reference 3-17 of WCAP-17330-P, Revision 1)

34


To: G. W. WhitemanB. J. BedontC. D. Cassino

Date: May5,2011

cc:

From: A. 0. RoslundExt: 724-722-6473

Your ref:

Ourref: LTR-SGDA-11-87

Fax: 724-722-5889

Subject: High-Confidence Variances for Tube and Tubesheet CTE for H*References:

1. WCAP- 17071 -P, Revision 2, "H*: Alternate Repair Criteria for the Tubesheet Expansion Regionin Steam Generators with Hydraulically Expanded Tubes (Model F)".

2. LTR-0026-0087-2, "Independent Technical Review of H* Steam Generator Tube Alternate RepairCriterion," MPR Associates, April 11, 2011.

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

The purpose of this letter is to document the methodology by which high confidence variances for tubeand tubesheet CTE for H* were calculated in response to questions from MPR in the independentreview of H*.

Electronically Approved*Prepared by: A. 0. Roslund

SGDA

Electronically Approved*

Electronically Approved*Verified: H. 0. Lagally

SGMP

Approved by: D. MerkovskyManager, SGDA

0 2011! Westinghouse Electric Company LLCAll Rights Reserved

*Electronically approved records are authenticated in the Electronic Document Management System.

35


Introduction

The calculation of H* at high probability and confidence in Reference 1 entails the use of standard

deviations for the coefficient of thermal expansion (CTE) for the tube and tubesheet, both of which are

modeled as normal distributions. The justification for modeling them as normal and the means and

standard deviations of the CTEs are contained in Appendix B of Reference 1. The standard deviations

used for the tube and tubesheet were 2.33% and 1.62%, respectively. These standard deviations are

essentially best estimate (50% confidence) from the data used. During the independent review of the

H* technical basis (References 2 and 3), it was requested that Westinghouse calculate high-confidence

variances of the standard deviations for the CTEs to show that the values used were conservative. The

data used in the following analysis were from tests that Westinghouse contracted ANTER to perform

as documented in Reference 1, Appendix B.

Methodology

ANTER tested 30 alloy 600 TT CTE specimens and 40 SA-508 tubesheet specimens. The results

were given as CTEs in 25°F increments from 100F to 700'F. The tubesheet data are in Table I

through Table 4. The tube data are in Table 5 through Table 7. In order to determine the

instrumentation error, one specimen each of the tube and tubesheet material was run ten times. These

results are shown in Table 8 and Table 9.

Best estimate (50% confidence) standard deviations were calculated from the standard formula,

n- x1)° =

High confidence (95%) standard deviations are obtained by the standard Chi-Squared adjustment:

cg95s 50 2 --xn-1,o. 9 5

Results for the tube and tubesheet are in Table 10 and Table 11. Results for the tube and tubesheet

instrumentation error (multiple runs) are in Table 12 and Table 13. Note that a higher CTE variance is

conservative for the purposes of calculating H*, while a lower instrumentation variance is

conservative. Therefore, the above equation is used for adjusting material standard deviations, which

results in a higher standard deviation at high confidence. For instrumentation variance, the above

equation is used with a 0.05 instead of 0.95, which results in a high-confidence lower bound. The

36


standard formula below was used to calculate a high confidence standard deviation for the tube and

tubesheet without instrumentation error:

1 J2 2

95,Material = q2s5,total - ,925,instrumentation

Results are in Table 14. As can be seen, the standard deviation values used in the H* analyses (2.33%

for the tube and 1.62% for the tubesheet) are conservative compared to the true high-confidence

standard deviations at temperatures of 200'F and greater. The range of temperatures applicable to the

operating conditions of population of H* candidate plants is between 200IF and 650'F.

37


Table 1

Tubesheet CTEs (lain / in OF)

Temp (*F) Sample 1 Sample 2 Sample 3 Sample 4 Sample 5 Sample 6 Sample 7 Sample 8 Sample 9 Sample 10100

125

150

175200

225250275

300325

350375

400425

450475

500525

550575600625650

675

700

a,c,e

38


Table 15

Tubesheet CTEs (gin / in IF)


100 F-125

150

175200225250

275300

325

350375

400425

450

475500525

550575600

625650675

700

a,c,e

39


Table 3

Tubesheet CTEs (,Iin / in IF)


100 F125

150

175

200225

250

275

300325

350

375

400

425

450

475

500

525

550575

600

625650675

700 _____ __________ __________ __________ __________ ____

a,c,e

40


Table 4

Tubesheet CTEs (pin / in IF)

Temp (°F) Sample 31 Sample 32 Sample 33 Sample 34 Sample 35 Sample 36 Sample 37 Sample 38 Sample 39 Sample 40

100125

150175200

225250

275300

325

350375

400425450

475500525

550575

600625

650675

700

a,c,e

41


Table 5

Tube CTEs (Model F) (uin / in °FI


100

125

150

175

200

225

250

275

300

325

350

375

400

425

450

475

500

525

550

575

600

625650

675 _____ _____ _____ _____

7O0 _____ _____ _____ _____ _____

a,c,e

42


Table 6Tube CTEs (Model D5) (pin / in IF)


100 F125

150

175200

225250

275300325

350

375400

425450

475

500525

550575600

625650675700

a,c,e

43


Table 7

Tube CTEs (Model 44F) (ftin / in OF)


100 F125

150

175

200225250

275

300325

350375400425

450

475

500525550575

600625650

675IL 700

a,c,e

44


Table 8

I une C I Ls Mvultiple runs on same specimen) tpun J i Fr

Temp (°F) Run 1 Run 2 Run 3 Run 4 Run 5 Run 6 Run 7 Run 8 Run 9 Run 10

100

125

150

175

200225

250275

300325

350

375

400425

450475

500525550575600625650

675

700

a,c,e

45


Table 9

Tubesheet CTEs (Multinle runs on same snecimen) (ftin / in °F)

Temp (*F) Run 1 Run 2 Run 3 Run 4 Run 5 Run 6 Run 7 Run 8 Run 9 Run'10

100

125

150

175200225

250

275

300325

350

375

400425

450

475

500525

550

575

600625

650

675

700

a,c,e

46

LTR-SGMMP-11-28 Rev. 1 NP-Attachment

Table 10Mean and Standard Deviation, Tube Material

Temperature Mean Best Estimate Standard 95% Confidence Standard(°F) (pin/in°F) Deviation (%) Deviation (%)

100 6.95 3.40 4.35

125 7.03 2.84 3.64

150 7.10 2.38 3.04

175 7.16 2.00 2.55

200 7.23 1.69 2.16

225 7.28 1.45 1.86

250 7.34 1.27 1.63

275 7.39 1.14 1.46

300 7.43 1.05 1.35

325 7.48 0.99 1.27

350 7.52 0.95 1.21

375 7.56 0.92 1.17

400 7.59 0.89 1.14

425 7.63 0.87 1.12

450 7.66 0.86 1.10

475 7.69 0.85 1.08

500 7.72 0.84 1.07

525 7.76 0.83 1.07

550 7.79 0.83 1.06

575 7.82 0.82 1.05

600 7.85 0.81 1.03

625 7.88 0.79 1.01

650 7.91 0.77 0.98

675 7.94 0.74 0.95

700 7.97 0.72 0.92

47


Table 11Mean and Standard Deviation, Tubesheet Material

Temperature Mean Best Estimate Standard 95% Confidence Standard(°F) (pin/in*F) Deviation (%) Deviation (%)

100 6.11 2.71 3.34125 6.23 2.30 2.83

150 6.35 1.96 2.42175 6.45 1.69 2.08200 6.55 1.48 1.82225 6.63 1.31 1.62250 6.71 1.19 1.46275 6.79 1.09 1.35300 6.85 1.02 1.26325 6.91 0.97 1.19350 6.97 0.92 1.14

375 7.02 0.89 1.10400 7.07 0.86 1.06425 7.12 0.84 1.03450 7.16 0.82 1.01

475 7.20 0.80 0.99500 7.24 0.79 0.97525 7.28 0.77 0.95550 7.32 0.76 0.94575 7.35 0.76 0.93600 7.39 0.75 0.92625 7.43 0.74 0.92

650 7.48 0.75 0.92675 7.52 0.76 0.93700 7.57 0.78 0.96

48


Table 12

Standard Deviation for Instrumentation Error, Tube Material

Temperature Best Estimate Standard 95% Confidence Standard(°F) Deviation (%) Deviation (%)

100 2.28 1.66

125 2.01 1.46

150 1.77 1.29

175 1.57 1.14

200 1.39 1.01

225 1.24 0.91

250 1.12 0.81

275 1.01 0.74

300 0.92 0.67

325 0.85 0.62

350 0.79 0.58

375 0.75 0.55

400 0.71 0.52

425 0.69 0.50

450 0.67 0.49

475 0.66 0.48

500 0.65 0.48

525 0.65 0.47

550 0.64 0.47

575 0.63 0.46

600 0.62 0.46

625 0.61 0.44

650 0.59 0.43

675 0.56 0.41

700 0.53 0.38

49


Table 13

Standard Deviation for Instrumentation Error, Tubesheet Material

Temperature Best Estimate Standard 95% Confidence Standard(*F) Deviation (%) Deviation (%)

100 2.08 1.52

125 1.82 1.32

150 1.59 1.16

175 1.40 1.02

200 1.25 0.91

225 1.13 0.82

250 1.03 0.75

275 0.95 0.69

300 0.89 0.65

325 0.85 0.62

350 0.82 0.60

375 0.79 0.58

400 0.78 0.57

425 0.78 0.57

450 0.77 0.56

475 0.78 0.57

500 0.79 0.57

525 0.79 0.58

550 0.79 0.58

575 0.80 0.58600 0.80 0.59

625 0.80 0.58

650 0.79 0.57

675 0.77 0.56

700 0.74 0.54

50


Table 14High-Confidence Tube and Tubesheet Standard Deviations with Instrumentation Error Removed

Temperature Tube (%) Tubesheet (%)(*F) a,c,e

100125150175

200225250275300325350375400425450475500525550575600625650675

700

51

Enclosure III to ET 12-0002

Enclosure III

Westinghouse Electric Company LLC CAW-12-3417, "Application for WithholdingProprietary Information from Public Disclosure"

(7 pages)

WestinghouseWestinghouse Electric CompanyNuclear Services

1000 Westinghouse DriveCranberry Township, PA 16066USA

U.S. Nuclear Regulatory Commission Direct tel: (412) 374-4643Document Control Desk Direct fax: (724) 720-075411555 Rockville Pike e-mail: [email protected], MD 20852 Proj letter: SAP-12-31

CAW-12-3417

February 22, 2012

APPLICATION FOR WITHHOLDING PROPRIETARYINFORMATION FROM PUBLIC DISCLOSURE

Subject: LTR-SGMMP-1 1-28 Rev. 1 P-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" (Proprietary)

The proprietary information for which withholding is being requested in the above-referenced report isfurther identified in Affidavit CAW-12-3417 signed by the owner of the proprietary information,Westinghouse Electric Company LLC. The affidavit, which accompanies this letter, sets forth the basison which the information may be withheld from public disclosure by the Commission and addresses withspecificity the considerations listed in paragraph (b)(4) of 10 CFR Section 2.390 of the Commission'sregulations.

Accordingly, this letter authorizes the utilization of the accompanying affidavit by Wolf Creek NuclearOperating Corporation.

Correspondence with respect to the proprietary aspects of the application for withholding or theWestinghouse affidavit should reference this letter, CAW-12-3417, and should be addressed toJ. A. Gresham, Manager, Regulatory Compliance, Westinghouse Electric Company, Suite 428,1000 Westinghouse Drive, Cranberry Township, PA 16066.

Very truly yours,

JZA. GerRegulatory Compliance

Enclosures

CAW-12-3417

AFFIDAVIT

COMMONWEALTH OF PENNSYLVANIA:

ss

COUNTY OF BUTLER:

Before me, the undersigned authority, personally appeared J. A. Gresham, who, being by me duly

sworn according to law, deposes and says that he is authorized to execute this Affidavit on behalf of

Westinghouse Electric Company LLC (Westinghouse), and that the averments of fact set forth in this

Affidavit are true and correct to the best of his knowledge, information, and belief:

J. A. Gresham, Manager

Regulatory Compliance

Sworn to and subscribed before me

this 22nd day of February 2012

Notary Public

COMMONWEALTH OF PENNSYLVANIANotarial Seal

Cynthia Olesky, Notary PublicManor Boro, Westmoreland County

My Commission Expires July 16, 2014Membbr, Pknnsvlvanla Assoriation of Notaries

2 CAW-12-3417

(1) I am Manager, Regulatory Compliance, in Nuclear Services, Westinghouse Electric

Company LLC (Westinghouse), and as such, I have been specifically delegated the function of

reviewing the proprietary information sought to be withheld from public disclosure in connection

with nuclear power plant licensing and rule making proceedings, and am authorized to apply for

its withholding on behalf of Westinghouse.

(2) 1 am making this Affidavit in conformance with the provisions of 10 CFR Section 2.390 of the

Commission's regulations and in conjunction with the Westinghouse Application for Withholding

Proprietary Information from Public Disclosure accompanying this Affidavit.

(3) 1 have personal knowledge of the criteria and procedures utilized by Westinghouse in designating

information as a trade secret, privileged or as confidential commercial or financial information.

(4) Pursuant to the provisions of paragraph (b)(4) of Section 2.390 of the Commission's regulations,

the following is furnished for consideration by the Commission in determining whether the

information sought to be withheld from public disclosure should be withheld.

(i) The information sought to be withheld from public disclosure is owned and has been held

in confidence by Westinghouse.

(ii) The information is of a type customarily held in confidence by Westinghouse and not

customarily disclosed to the public. Westinghouse has a rational basis for determining

the types of information customarily held in confidence by it and, in that connection,

utilizes a system to determine when and whether to hold certain types of information in

confidence. The application of that system and the substance of that system constitutes

Westinghouse policy and provides the rational basis required.

Under that system, information is held in confidence if it falls in one or more of several

types, the release of which might result in the loss of an existing or potential competitive

advantage, as follows:

(a) The information reveals the distinguishing aspects of a process (or component,

structure, tool, method, etc.) where prevention of its use by any of

3 CAW-12-3417

Westinghouse's competitors without license from Westinghouse constitutes a

competitive economic advantage over other companies.

(b) It consists of supporting data, including test data, relative to a process (or

component, structure, tool, method, etc.), the application of which data secures a

competitive economic advantage, e.g., by optimization or improved

marketability.

(c) Its use by a competitor would reduce his expenditure of resources or improve his

competitive position in the design, manufacture, shipment, installation, assurance

of quality, or licensing a similar product.

(d) It reveals cost or price information, production capacities, budget levels, or

commercial strategies of Westinghouse, its customers or suppliers.

(e) It reveals aspects of past, present, or future Westinghouse or customer funded

development plans and programs of potential commercial value to Westinghouse.

(f) It contains patentable ideas, for which patent protection may be desirable.

There are sound policy reasons behind the Westinghouse system which include the

following:

(a) The use of such information by Westinghouse gives Westinghouse a competitive

advantage over its competitors. It is, therefore, withheld from disclosure to

protect the Westinghouse competitive position.

(b) It is information that is marketable in many ways. The extent to which such

information is available to competitors diminishes the Westinghouse ability to

sell products and services involving the use of the information.

(c) Use by our competitor would put Westinghouse at a competitive disadvantage by

reducing his expenditure of resources at our expense.

4 CAW-12-3417

(d) Each component of proprietary information pertinent to a particular competitive

advantage is potentially as valuable as the total competitive advantage. If

competitors acquire components of proprietary information, any one component

may be the key to the entire puzzle, thereby depriving Westinghouse of a

competitive advantage.

(e) Unrestricted disclosure would jeopardize the position of prominence of

Westinghouse in the world market, and thereby give a market advantage to the

competition of those countries.

(f) The Westinghouse capacity to invest corporate assets in research and

development depends upon the success in obtaining and maintaining a

competitive advantage.

(iii) The information is being transmitted to the Commission in confidence and, under the

provisions of 10 CFR Section 2.390; it is to be received in confidence by the

Commission.

(iv) The information sought to be protected is not available in public sources or available

information has not been previously employed in the same original manner or method to

the best of our knowledge andbelief.

(v) The proprietary information sought to be withheld in this submittal is that which is

appropriately marked in LTR-SGMMP- 11-28 Rev. I P-Attachment, "Response to

USNRC Request for Additional Information Regarding the License Amendment

Requests for Permanent Application of the Alternate Repair Criterion, H*, to the

Model D5 and Model F SGs" (Proprietary), for submittal to the Commission, being

transmitted by Wolf Creek Nuclear Operating Corporation and Application for

Withholding Proprietary Information from Public Disclosure, to the Document Control

Desk. The proprietary information as submitted by Westinghouse for Wolf Creek

Generating Station, is that associated with the technical justification of the H* Alternate

Repair Criteria for hydraulically expanded steam generator tubes and may be used only

for that purpose.

5 CAW-12-3417

This information is part of that which will enable Westinghouse to:

(a) License the H* Alternate Repair Criteria.

Further this information has substantial commercial value as follows:

(a) Westinghouse plans to sell the use of the information to its customers for the

purpose of licensing the H* Alternate Repair Criteria.

(b) Westinghouse can sell support and defense of the H* criteria.

(c) The information requested to be withheld reveals the distinguishing aspects of a

methodology which was developed by Westinghouse.

Public disclosure of this proprietary information is likely to cause substantial harm to the

competitive position of Westinghouse because it would enhance the ability of

competitors to provide similar technical justification and licensing defense services for

commercial power reactors without commensurate expenses. Also, public disclosure of

the information would enable others to use the information to meet NRC requirements for

licensing documentation without purchasing the right to use the information.

The development of the technology described in part by the information is the result of

applying the results of many years of experience in an intensive Westinghouse effort and

the expenditure of a considerable sum of money.

In order for competitors of Westinghouse to duplicate this information, similar technical

programs would have to be performed and a significant manpower effort, having the

requisite talent and experience, would have to be expended.

Further the deponent sayeth not.

PROPRIETARY INFORMATION NOTICE

Transmitted herewith are proprietary and/or non-proprietary versions of documents furnished to the NRCin connection with requests for generic and/or plant-specific review and approval.

In order to conform to the requirements of 10 CFR 2.390 of the Commission's regulations concerning theprotection of proprietary information so submitted to the NRC, the information which is proprietary in theproprietary versions is contained within brackets, and where the proprietary information has been deletedin the non-proprietary versions, only the brackets remain (the information that was contained within thebrackets in the proprietary versions having been deleted). The justification for claiming the informationso designated as proprietary is indicated in both versions by means of lower case letters (a) through (f)located as a superscript immediately following the brackets enclosing each item of information beingidentified as proprietary or in the margin opposite such information. These lower case letters refer to thetypes of information Westinghouse customarily holds in confidence identified in Sections (4)(ii)(a)through (4)(ii)(f) of the affidavit accompanying this transmittal pursuant to 10 CFR 2.390(b)(1).

COPYRIGHT NOTICE

The reports transmitted herewith each bear a Westinghouse copyright notice. The NRC is permitted tomake the number of copies of the information contained in these reports which are necessary for itsinternal use in connection with generic and plant-specific reviews and approvals as well as the issuance,denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license,permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on publicdisclosure to the extent such information has been identified as proprietary by Westinghouse, copyrightprotection notwithstanding. With respect to the non-proprietary versions of these reports, the NRC ispermitted to make the number of copies beyond those necessary for its internal use which are necessary inorder to have one copy available for public viewing in the appropriate docket files in the public documentroom in Washington, DC and in local public document rooms as may be required by NRC regulations ifthe number of copies submitted is insufficient for this purpose.. Copies made by the NRC must includethe copyright notice in all instances and the proprietary notice if the original was identified as proprietary.

wolf creek - license amendment request, revision to ...on june 30, 2011, duke energy submitted a...

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