Text

License Amendment Request, Revision to Technical Specification (TS) 5.5.9, Steam Generator (SG) Program, and TS 5.6.10, Steam Generator Tube Inspection Report, for a Permanent Alternate Repair Criteria
	ML12102A080
Person / Time
Site:	Wolf Creek
Issue date:	03/29/2012
From:	Broschak J; Wolf Creek
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	ET 12-0002
	Download: ML12102A080 (107)
	v • d • e

WLF CREEK

'NUCLEAR OPERATING CORPORATION John P. Broschak Vice President Engineering March 29, 2012 ET 12-0002 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555

Subject:

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

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

The proposed amendment constitutes a redefinition of the steam generator tube primary to secondary pressure boundary and defines the safety significant portion of the tube that must be inspected or plugged. Tube flaws detected below the safety significant portion of the tube are not required to be plugged. The exclusion of plugging flaws in the non-safety significant portion of the tube minimizes unnecessary tube plugging and maintains the safety margin of the steam generators to perform their safety function by maintaining the reactor coolant pressure boundary, 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 Page 2 of 5 WCNOC requests the approval of the proposed license amendment by December 20, 2012 to support implementation during the WCGS 2013 Refueling Outage (Refueling Outage 19),

currently scheduled to commence on February 4, 2013. Once approved, the amendment will be implemented prior to MODE 4 entry during startup from Refueling Outage 19.

WCAP-17330-P, Revision 1, "H*: Resolution of NRC Technical Issue Regarding Tubesheet Bore Eccentricity (Model F/Model D5)," provides the resolution of the remaining questions in support of the permanent application of the H* criterion. WCAP-17330-P, Revision 1, and Westinghouse 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, Duke Energy submitted a license amendment request for permanent application of the alternate repair 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, 2011 and 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 request for additional information. Duke Energy responded to the request for additional information on January 12, 2012.

Subsequent to the Duke Energy license amendment request, Virginia Electric and Power Company (Dominion) submitted a license amendment request for permanent application of the alternate repair criterion H* for Surry Power Station Units 1 and 2. On January 18, 2012, the NRC issued a request for additional information. Dominion responded to the request for additional 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 Amendment Requests for Permanent Application of the Alternate Repair Criterion, H*, to the Model D5 and Model F SGs") augments the responses to the Duke Energy request for additional information to include similar responses applicable to Model F steam generators. Additionally, Enclosure I addresses the Dominion request for additional information on question 14 for the Model F steam generators.

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

Enclosure I provides the proprietary Westinghouse Electric Company LLC, LTR-SGMMP-1 1-28 Rev.1 P-Attachment, "Response to USNRC Request for Additional Information Regarding the License Amendment Requests for Permanent Application of the Alternate Repair Criterion, H*,

to the Model D5 and Model F SGs." Enclosure II provides the non-proprietary Westinghouse Electric Company LLC "LTR-SGMMP-11-28 Rev.1 NP-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." As Enclosure I contains information proprietary to Westinghouse Electric Company LLC, it is supported by an affidavit signed by Westinghouse Electric Company LLC, the owner of the information. The affidavit sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR 2.390 of the Commission's regulations. Accordingly, it is respectfully requested that the information, which is proprietary to Westinghouse, be withheld from public disclosure in accordance with 2.390 of the Commission's regulations. This affidavit, along with Westinghouse authorization letter, CAW-12-3417, "Application for Withholding Proprietary Information from Public Disclosure," is contained in Enclosure Ill.

ET 12-0002 Page 3 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 changes will be implemented pursuant to TS 5.5.14, "Technical Specification (TS) Bases Control Program," at the time the amendment is implemented.

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

The Plant Safety Review Committee reviewed this amendment application. In accordance with 10 CFR 50.91, "Notice for public comment; State consultation," a copy of this amendment application, with attachments, is being provided to the designated Kansas State official.

Attachment V provides the regulatory commitments associated with this application. If you have 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 Page 4 of 5 Attachments: Evaluation IV III Markup of Technical Specification pages Proposed Changes to Technical Specification Bases (for information only)

IV Retyped Technical Specification pages V List of Regulatory Commitments VI Response to Request for Additional Information Questions Specific to Wolf Creek Nuclear Operating Corporation Enclosure I - Westinghouse Electric Company LLC, LTR-SGMMP-11-28 Rev.1 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) 11 - Westinghouse Electric Company LLC LTR-SGMMP-11-28 Rev.1 NP-Attachment and Errata, "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" (Non-proprietary)

III - Westinghouse Electric Company LLC CAW-12-3417, "Application for Withholding Proprietary Information from Public Disclosure" cc: E. E. Collins (NRC), w/a, w/e T. A. Conley (KDHE), w/a, w/e (Enclosure II only)

J. R. Hall (NRC), w/a, w/e N. F. O'Keefe (NRC), w/a, wie Senior Resident Inspector (NRC), w/a, w/e

ET 12-0002 Page 5 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 President Engineering of Wolf Creek Nuclear Operating Corporation; that he has read the foregoing document and knows the contents thereof; that he has executed the same for and on behalf of said Corporation with full power and authority to do so; and that the facts therein stated are true and correct to the best of his knowledge, information and belief.

By JoO1P. Broschak Vi g(President Engineering SUBSCRIBED and sworn to before me this f9 z day of MdrA ,2012.

Notary li-0 "b GAYLE SHEPHEARD1 M&Notary Public - State of Kansas

. /*

My Appt. Expires 2L '-2/ J Expiration Date_________

Attachment I to ET 12-0002 Page 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 Permanent Alternate 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

Attachment I to ET 12-0002 Page 2 of 23

1.

SUMMARY

DESCRIPTION Wolf Creek Nuclear Operating Corporation (WCNOC) proposes to revise Wolf Creek Generating 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 from periodic steam generator tube inspections. In addition, this amendment proposes to revise TS 5.6.10, "Steam Generator Tube Inspection Report" to remove reference to previous interim alternate repair criteria and provide reporting requirements specific to the permanent alternate repair criteria. Application of the supporting structural analysis and leakage evaluation results to exclude portions of the tubes from inspection and repair of tube indications is interpreted to constitute a redefinition of the primary to secondary pressure boundary. The proposed changes to the TS are based on the supporting structural analysis and leakage evaluation completed by Westinghouse Electric Company LLC. The documentation supporting the Westinghouse analysis is described in Section 3 and provides the licensing basis for this change. Table 5-1 of WCAP 17330-P (Reference 19) provides the 95/95 H* value of 15.21 inches for plants with Model F Steam Generators which includes WCGS.

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

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

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

Amendment Number 178 (Reference 3) which approved an interim alternate repair criteria 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 if any circumferential cracking observed in the region greater than 17 inches from the top of the tubesheet is less than a value sufficient to permit the remaining circumferential ligament to transmit the limiting axial loads. Three new reporting requirements were added 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 top of the tubesheet for Refueling Outage 17 and the subsequent operating cycle. Additionally TS 5.6.10 was revised to provide reporting requirements specific to Refueling Outage 17 and 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 top of the tubesheet for Refueling Outage 18 and the subsequent operating cycle. Additionally TS 5.6.10 was revised to provide reporting requirements specific to Refueling Outage 18 and the subsequent operating cycle.

Attachment I to ET 12-0002 Page 3 of 23 Approval of this amendment application is requested by December 20, 2012 to support the WCGS Refueling Outage 19 (February 2013), since the existing one-cycle amendment expires at 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 inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.

The following alternate tube repair criteria shall be applied as an alternative 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 inches below the top of the tubesheet do not require plugging. Tubes with service-induced flaws located in the portion of the tube from the top of the tubesheet to 15.2 inches below the top of the tubesheet shall be plugged upon detection.

This section would be revised as follows:

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

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

1. Tubes with service-induced flaws located greater than 15.21 inches below the top of the tubesheet do not require plugging.

Tubes with service-induced flaws located in the portion of the tube from the top of the tubesheet to 15.21 inches below the top of the tubesheet shall be plugged upon detection.

TS 5.5.9d. currently states:

d. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria. For

Attachment I to ET 12-0002 Page 4 of 23 Refueling Outage 18 and the subsequent operating cycle, the portion of the tube below 15.2 inches 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 meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. An assessment of degradation shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

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

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

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

This section would be revised as follows:

d. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria. The portion of the tube below 15.21 inches 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 meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. An assessment of degradation shall be performed to

Attachment I to ET 12-0002 Page 5 of 23 determine the type and location of flaws to which the tubes may be' susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

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

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

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

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) the primary to secondary LEAKAGE rate observed in each SG (if it is not practical to assign the LEAKAGE to an individual SG, the entire primary to secondary LEAKAGE should be conservatively assumed to be from one SG) during the cycle preceding the inspection which is the subject of the report; Following completion of an inspection performed in Refueling Outage 18 (and any inspections performed in the subsequent operating cycle) the calculated accident induced leakage rate from the portion of the tubes below 15.2 inches from the top of the tubesheet for the most limiting accident in the most limiting SG. In addition, if the calculated accident induced leakage rate from the most limiting accident is less than 2.50 times the maximum operational primary to secondary leak rate, the report should 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) the results of monitoring for the tube axial displacement (slippage). If

Attachment I to ET 12-0002 Page 6 of 23 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 not practical to assign the LEAKAGE to an individual SG, the entire primary to secondary LEAKAGE should be conservatively assumed to be from one SG) during the cycle preceding the inspection which is the subject of the report; The calculated accident induced leakage rate from the portion of the tubes below 15.21 inches from the top of the tubesheet for the most limiting accident in the most limiting SG. In addition, if the calculated accident induced leakage rate from the most limiting accident is less than 2.50 times the maximum operational primary to secondary leak rate, the report should describe how it was determined; and

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

3. TECHNICAL EVALUATION

Background

WCGS is a four loop Westinghouse designed plant with Model F steam generators having 5626 tubes in each steam generator. A total of 266 tubes are currently plugged in all four steam generators. The design of the steam generator includes Alloy 600 thermally treated tubing, full depth hydraulically expanded tubesheet joints, and stainless steel tube support plates with quatrefoil 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 Examination Guidelines," (Reference 6); EPRI 1019038, "Steam Generator Integrity Assessment Guidelines," (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 that nondestructive testing be accomplished by qualified personnel using qualified procedures in accordance with the applicable criteria. The inspection techniques and equipment are capable of reliably detecting the existing and potential specific degradation mechanisms applicable to WCGS. The inspection techniques, essential variables and equipment are qualified to Appendix H, "Performance Demonstration for Eddy Current Examination," and Appendix I, "NDE System Measurement Uncertainties for Tube Integrity Assessments," of the EPRI Steam Generator Examination Guidelines.

Attachment I to ET 12-0002 Page 7 of 23 Catawba Nuclear Station, Unit 2, (Catawba) reported indication of cracking following nondestructive eddy current examination of the steam generator tubes during their fall 2004 outage. NRC Information Notice (IN) 2005-09, "Indications in Thermally Treated Alloy 600 Steam Generator Tubes and Tube-to-Tubesheet Welds," (Reference 8), provided industry notification of the Catawba issue. IN 2005-09 noted that Catawba reported crack like indications in the tubes approximately seven inches below the top of the hot leg tubesheet in one tube, and just above the tube-to-tubesheet welds in a region of the tube known as the tack expansion in several other tubes. Indications were also reported in the tube-end welds, also known 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 industry operating 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) with techniques capable of finding potential forms of degradation that may be occurring at these locations (as discussed in Generic Letter 2004-001, "Requirements for Steam Generator Tube Inspections"). Since the WCGS Westinghouse Model F steam generators were fabricated with Alloy 600 thermally treated tubes similar to the Catawba Unit 2 Westinghouse Model D5 steam generators, a potential exists for WCGS to identify tube indications similar to those reported at Catawba within the hot leg tubesheet region.

Potential inspection plans for the tubes and tube welds underwent intensive industry discussions in March 2005. The findings in the Catawba steam generator tubes present three distinct 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 into adjacent tube material.

Prior to each steam generator tube inspection, a degradation assessment, which includes a review of operating experience, is performed to identify degradation mechanisms that have a potential to be present in the WCGS steam generators. A validation assessment is also performed to verify that the eddy current techniques utilized are capable of detecting those flaw types that are identified in the degradation assessment. Based on operating experience discussed above, WCNOC revised the steam generator inspection plan to include sampling of bulges and overexpansions within the tubesheet region on the hot leg side in Refueling Outage 14 (Spring 2005) and Refueling Outage 15 (Fall 2006). The sample is based on the guidance contained in EPRI 1013706, "Pressurized Water Reactor Steam Generator Examination Guidelines," Revision 7, and TS 5.5.9, "Steam Generator (SG) Program." The inspection plan is expanded according to EPRI steam generator examination guidelines if necessary due to confirmed degradation in the region required to be examined (i.e. a tube crack). Degradation was not detected in the tubesheet region in Refueling Outage 14 and Refueling Outage 15.

Attachment I to ET 12-0002 Page 8 of 23 At WCGS, tube flaw indications within the tube sheet have only been found at the hot leg tube ends. 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 of the tube end. Of these seventy-six flaw indications, only eight met the tube repair criteria in the technical specifications.

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

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

Summary of Licensing Basis Analysis (H* Analysis)

On June 2, 2009, Westinghouse WCAP-17071-P, Revision 0, "H*: Alternate Repair Criteria for the Tubesheet Expansion Region in Steam Generators with Hydraulically Expanded Tubes (Model F)," (Reference 10) was submitted as Enclosure I of WCNOC request (Reference 11) to change 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 alternate repair 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 provided the responses to questions 1 through 25 of the August 11, 2009 letter and included the following documents:

Westinghouse letter LTR-SGMP-09-100 P-Attachment, Revision 0, "Response to NRC Request 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 NRC Request for Additional Information on H*; RAI #4; Model F and Model D5 Steam Generators," August 25, 2009 (Reference 16).

Attachment I to ET 12-0002 Page 9 of 23 On September 15, 2009, WCNOC submitted a request (Reference 17) to revise the permanent alternate repair criteria amendment request (Reference 11) to be an interim change applicable to Refueling Outage 17 and the subsequent operating cycle. This request was made in response to a September 2, 2009 teleconference between NRC Staff and industry personnel, in which the NRC Staff indicated that their concerns with eccentricity of the tube sheet tube bore in normal and accident conditions (RAI question 4 of the August 11, 2009 letter) have not been resolved. The September 15, 2009 letter also requested the NRC staff to provide the specific questions concerning the tubesheet bore eccentricity issue which must be resolved to support a permanent alternate repair criteria amendment request.

On December 9, 2009, the NRC provided a letter (Reference 18) documenting the currently identified and unresolved issues relating to tubesheet bore eccentricity. This letter contained 14 unresolved questions which required resolution before the NRC could complete its review of a permanent 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 of the remaining questions identified in the December 9, 2009 NRC letter in support of the permanent H* amendment for WCGS.

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

" LTR-SGMP-10-78 P-Attachment, "Effects of Tubesheet Bore Eccentricity and Dilation on Tube-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, was transmitted 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 Bottom of the Expansion Transition (BET) for Implementation of H*," was prepared to support plant determinations of BET measurements and their significant deviation assessment. This document, which is applicable to WCGS's Model F steam generators, was transmitted to the 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 Tubesheet Bore Eccentricity," September 13, 2010. This document, which is applicable to WCGS's Model 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 Regarding Tubesheet Bore Eccentricity (Model F/Model D5)," June 2011 (Reference 19) makes reference to Revision 2 of WCAP-17071-P and Revision 1 of LTR-SGMP-09-100 P-Attachment. As described above, WCNOC has previously submitted Revision 0 of these documents. These revisions (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 contained in WCAP-17071-P, Revision 0 and LTR-SGMP-09-100 P-Attachment, Revision 0, remains valid and provides part of the licensing basis for the requested amendment.

Attachment I to ET 12-0002 Page 10 of 23 The following table provides the list of the WCGS licensing basis documents for H*.

Document Revision Title Reference Number Number Number WCAP-1 7330-P 1 H*: Resolution of NRC Technical Issue Regarding 19 Tubesheet 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 10 Expansion Region in Steam Generators with Hydraulically Expanded Tubes (Model F)

LTR-SGMP-09-100 0 Response to NRC Request for Additional Information 15 P-Attachment on H*; Model F and Model D5 Steam Generators LTR -SGMP 0 Response to NRC Request for Additional Information 16 109 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 20 P-Attachment on Tube-to-Tubesheet Contact Pressure and Their Relative Importance to H*

LTR-SGMP-10-33 0 H* Response to NRC Questions Regarding 21 P-Attachment Tubesheet Bore Eccentricity LTR-SGMMP 1 Response to USNRC Request for Additional 35 28 P-Attachment Information Regarding the License Amendment Requests for Permanent Application of the Alternate Repair Criterion, H*, to the Model D5 and Model F SGs In addition, the following correspondence is also applicable to the WCGS permanent alternate repair criteria request.

A March 28, 2011 letter from the NRC to Southern Nuclear Operating Company (Reference

25) documented the summary of a February 16, 2011 public meeting regarding steam generator tube inspection permanent alternate repair criteria. Enclosure 3 of the NRC letter provided technical NRC Staff questions developed at the meeting. Responses to these questions 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 by MPR Associates. Related to the independent review, a May 26, 2011 letter from the NRC to Southern Nuclear Company (Reference 26) included a presubmittal review request for additional information. The response to the NRC presubmittal review request is provided in Southern Nuclear Operating Company letter NL-1 1-1178 (Reference 27).

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

Attachment I to ET 12-0002 Page 11 of 23 Regarding Tubesheet Bore Eccentricity (Model F/Model D5)." A supplement (Reference 29) to the license amendment request was submitted on July 11, 2011 and provided Westinghouse Electric Company LLC LTR-SGMP-1 1-58, "WCAP-1 7330-P, Revision 1 Erratum." On January 5, 2012, a request for additional information (Reference 30) was transmitted electronically to Duke 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 Power Company (Dominion) submitted a license amendment request (Reference 32) for permanent application of the alternate repair criterion H* for Surry Power Station Units 1 and 2. On January 18, 2012, the NRC issued a request for additional information (Reference 33).

Dominion responded to the request for additional information on February 14, 2012 (Reference 34).

Westinghouse Electric Company LLC, LTR-SGMMP-11-28 Rev.1 P-Attachment (Reference 35), "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," augments the responses to the Duke Energy request for additional information to include similar responses applicable to Model F steam generators.

Additionally, LTR-SGMMP-11-28 Rev.1 P-Attachment addresses the Dominion request for additional information question 14 for the Model F steam generators. Attachment VI provides WCNOC specific responses to questions 12 and 13 from the Duke Energy request for additional information and question 15 from the Dominion request for additional information.

Evaluation To preclude unnecessarily plugging tubes in the WCGS steam generators, an evaluation was performed to identify the safety significant portion of the tube within the tubesheet necessary to maintain structural and leakage integrity in both normal and accident conditions. Tube inspections will be limited to identifying and plugging degradation in the safety significant portion of the tubes. The technical evaluation for the inspection and repair methodology is provided in the H* Analysis as described above. This evaluation is based on the use of finite element model structural analysis and a bounding leak rate evaluation based on contact pressure between the tube and the tubesheet during normal and postulated accident conditions. The limited tubesheet inspection criteria were developed for the tubesheet region of the WCGS Model F steam generator considering the most stringent loads associated with plant operation, including transients and postulated accident conditions. The limited tubesheet inspection criteria were selected to prevent tube burst and axial separation due to axial pullout forces acting on the tube and to ensure that the accident induced leakage limits are not exceeded. The H* Analysis provides technical justification for limiting the inspection in the tubesheet 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 based upon evaluation and testing programs that quantified the tube-to-tubesheet radial contact pressure for bounding plant conditions as described in the H* Analysis. The tube-to-tubesheet radial contact pressure provides resistance to tube pullout and resistance to leakage during plant operation and transients.

Attachment I to ET 12-0002 Page 12 of 23 Primary-to-secondary leakage from tube degradation is assumed to occur in several design basis accidents: feedwater line break (FLB), steam line break (SLB), locked rotor, and control rod ejection. The radiological dose consequences associated with this assumed leakage are evaluated to ensure that they remain within regulatory limits (e.g. 10 CFR Part 100, 10 CFR 50.67, GDC 19). The accident induced leakage performance criteria are intended to ensure the primary-to-secondary leak rate during any accident does 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 the tubesheet. The criteria for tube burst described in NEI 97-06 (Reference 5) and NRC Regulatory Guide (RG) 1.121, "Bases for Plugging Degraded PWR Steam Generator Tubes,"

(Reference 9) are satisfied due to the constraint provided by the tubesheet. Through application of the limited tubesheet inspection scope as described below, the existing operating leakage limit provides assurance that excessive leakage (i.e., greater than accident analysis assumptions) 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 (0.1 gpm) through any one steam generator. Consequently, accident leakage is approximately 10 times the allowable leakage, if only one steam generator is leaking. Using a SLB/FLB overall leakage factor of 2.50, accident induced leakage is less than 0.6 gpm, if all 4 steam generators are leaking at 150 gpd at the beginning of the accident. Therefore, significant margin exists between the conservatively estimated accident induced leakage and the allowable accident leakage (1.0 gpm).

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

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

I The plant transient response following a full power double-ended main feedwater line rupture corresponding to "best estimate" initial conditions and operating characteristics as generally presented 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 characteristic instead of a heatup transient. The use of either the component design specification transient or the 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 to establish a conservative structural (fatigue) design basis for RCS components; however, H* does not involve component structural and fatigue issues. The best estimate transient is considered more appropriate for use in the H* leakage calculations.

Attachment I to ET 12-0002 Page 13 of 23

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 both station blackout (loss of main feedwater and reactor coolant pump coast down) and turbine trip.

" The assumptions on which the FLB safety analysis is based are specifically intended to establish a conservative basis for minimum auxiliary feedwater (AFW) capacity requirements and combines worst case assumptions which are exceptionally more severe when the FLB occurs inside containment. For example, environmental errors that are applied to reactor trip and engineered safety feature actuation would no longer be applicable. This would result in much earlier reactor trip and greatly increase the steam 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 the secondary separators, which could only result in an increased initial cooldown (because of retained liquid inventory available for cooling) when compared to the FLB transient. A SLB could not result in more limiting temperature conditions than a FLB.

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

Since the best estimate FLB transient temperature would not be expected to exceed the normal operating temperature, the viscosity ratio for the FLB transient is set to 1.0. Therefore, the leak rate factor would only be a function of the increase in pressure differential during the design basis SLB/FLB. However, per Reference 15, the FLB transient was evaluated as a heatup event. Since dynamic viscosity decreases with the increase in temperature during a postulated FLB event, the viscosity subfactor increases above 1.0. For WCGS, the resulting leak rate ratio for 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 rod ejection event, are conservatively modeled using the design specification transients to result in increased temperatures in the steam generator hot and cold legs for a period of time. As previously noted, dynamic viscosity decreases with increasing temperature. Therefore, leakage would be expected to increase due to decreasing viscosity and increasing differential pressure for the duration of time that there is a rise in RCS temperature. For transients other than a SLB and a FLB, the length of time that a plant with Model F steam generators will exceed the normal operating differential pressure across the tubesheet is less than 30 seconds. As the accident induced leakage performance criteria is defined in gallons per minute, the leak rate for a locked rotor ejection event can be integrated over a minute to compare to the limit. Time integration permits an increase in acceptable leakage during the time of peak pressure differential by approximately a factor of two because of the short duration (less than 30 seconds) of the elevated pressure differential. This translates into an effective reduction in leakage factor by the same factor of two for the locked rotor event. Therefore, for the locked rotor event, the

Attachment I to ET 12-0002 Page 14 of 23 leakage factor of 1.77 (Revised Table 9-7, Reference 15) for WCGS is adjusted downward to a factor of 0.89. Similarly, for the control rod ejection event, the duration of the elevated pressure differential is less than 10 seconds. Thus, the peak leakage factor may be reduced by a factor of six from 2.65 to 0.44.

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 source and compared to the allowable accident induced leakage limit. For the operational assessment (OA), the difference in the leakage between the allowable leakage and the accident induced leakage from sources other than the tubesheet expansion region will be divided by 2.50 and compared to the observed operational leakage. An administrative limit will be established to not exceed the calculated value.

Reference 11 redefines the primary pressure boundary. The tube-to-tubesheet weld no longer functions as a portion of this boundary. The hydraulically expanded portion of the tube into the tubesheet over the H* distance now functions as the primary pressure boundary in the area of the tube and tubesheet, maintaining the structural and leakage integrity over the full range of steam generator operating conditions, including the most limiting accident conditions. The evaluation in Reference 11 determined that degradation in tubing below this safety significant portion of the tube does not require inspection or repair (plugging). The inspection of the safety significant portion of the tubes provides a high level of confidence that the structural and leakage 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 to tube slippage and concluded that the tubes are fully restrained against motion under very conservative design and analysis assumptions such that tube slippage is not a credible event for any tube in the bundle. As a condition of approval of Amendment Number 186, WCNOC committed 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), the NRC staff requested that WCNOC perform a validation of the tube expansion from the top of tubesheet to the bottom of expansion transition (BET) to determine if there are any significant deviations that would invalidate assumptions in WCAP-17071-P (Reference 10). WCNOC has completed the validation of the tube expansion from the top of tubesheet to the BET. Based on data review and LTR-SGMP-09-111 P-Attachment, Rev. 1 (Reference 22), WCNOC did not identify 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 operating licenses to include Technical Specifications (TSs) as part of the operating license. The TSs ensures the operational capability of structures, systems, and components that are required to protect 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

Attachment I to ET 12-0002 Page 15 of 23 the Code of Federal Regulations (10 CFR 50.36) which requires that the TSs include items in the following specific categories: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation; (3) surveillance requirements per 10 CFR 50.36(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, define requirements for the reactor coolant pressure boundary (RCPB) with respect to structural and leakage integrity.

GDC 19 of 10 CFR 50, Appendix A, defines requirements for the control room and for the radiation protection of the operators working within it. Accidents involving the leakage or burst of 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 this appendix 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 control of special processes, inspection, testing, and corrective action.

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

Under 10 CFR 50.65, the Maintenance Rule, licensees classify steam generators as risk-significant components because they are relied upon to remain functional during and after design basis events. Steam generators are to be monitored under 10 CFR 50.65(a)(2) against industry established performance criteria. Meeting the performance criteria of Nuclear Energy Institute (NEI) 97-06, Revision 2, "Steam Generator Program Guidelines," provides reasonable assurance that the steam generator tubing remains capable of fulfilling its specific safety function 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 of normal operating conditions (including startup, operation in the power range, hot standby, cooldown and all anticipated transients included in the design specification) and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary to secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary to secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial loads.

Attachment I to ET 12-0002 Page 16 of 23

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 assumed in the accident analysis in terms of total leakage rate for all steam generators and leakage rate for an individual steam generator. Leakage is not to exceed 1 gpm per steam generator, except for specific types of degradation at specific locations when implementing alternate repair criteria as documented in the Steam Generator Program technical specifications.

The Reactor Coolant System (RCS) operational primary to secondary leakage through any 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 tubesheet from the secondary face that is required to maintain structural and leakage integrity over the full range of steam generator operating conditions, including the most limiting accident conditions.

The evaluation in this Attachment determined that degradation in tubing below the safety significant portion of the tube does not require plugging and serves as the bases for the tubesheet inspection program. As such, the WCGS inspection program provides a high level of confidence that the structural and leakage criteria are maintained during normal operating and accident conditions.

4.2 Significant Hazards Consideration This amendment application proposes to revise Technical Specification (TS) 5.5.9, "Steam Generator (SG) Program," to exclude portions of the tubes within the tubesheet from periodic steam generator inspections. In addition, this amendment proposes to revise Technical Specification (TS) 5.6.10, "Steam Generator Tube Inspection Report," to remove reference to previous interim alternate repair criteria and provide reporting requirements specific to the temporary alternate repair criteria. Application of the structural analysis and leak rate evaluation results, to exclude portions of the tubes from inspection and repair is interpreted to constitute a redefinition of the primary to secondary pressure boundary.

The proposed change defines the portion of the tube that must be inspected and repaired. A justification has been developed by Westinghouse Electric Company, LLC to identify the specific inspection depth below which any type of axial or circumferential primary water stress corrosion 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 the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," Part 50.92(c), as discussed below:

(1) Does the proposed change involve a significant increase in the probability or consequences 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 criteria does not have a detrimental impact on the integrity of any plant structure, system, or

Attachment I to ET 12-0002 Page 17 of 23 component that initiates an analyzed event. The proposed change will not alter the operation of, or otherwise increase the failure probability of any plant equipment that initiates an analyzed accident.

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

Addressing the SGTR event, the required structural integrity margins of the steam generator 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 the tubesheet and constraint provided by the tube-to-tubesheet joint. Tube burst cannot occur within the thickness of the tubesheet. The tube-to-tubesheet joint constraint results from the hydraulic expansion process, thermal expansion mismatch between the tube and tubesheet, from the differential pressure between the primary and secondary side, and tubesheet deflection. The structural margins against burst, as discussed in Regulatory Guide (RG) 1.121, "Bases for Plugging Degraded PWR Steam Generator Tubes," and TS 5.5.9 are maintained for both normal and postulated accident conditions.

The proposed change has no impact on the structural or leakage integrity of the portion of the tube outside of the tubesheet. The proposed change maintains structural and leakage integrity 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 the occurrence of a SGTR accident.

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

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

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

The leakage factor of 2.50 for WCGS, for a postulated SLB/FLB, has been calculated as shown 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

Attachment I to ET 12-0002 Page 18 of 23 will be multiplied by a factor of 2.50 and added to the total leakage from any other source and compared to the allowable accident induced leakage limit. For the operational assessment (OA), the difference in the leakage between the allowable leakage and the accident induced leakage from sources other than the tubesheet expansion region will be divided by 2.50 and compared to the observed operational leakage.

The probability of an SLB/FLB is unaffected by the potential failure of a steam generator tube as the failure of the tube is not an initiator for an SLB/FLB event. SLB/FLB leakage is limited by leakage flow restrictions resulting from the leakage path above potential cracks through the tube-to-tubesheet crevice. The leak rate during all postulated accident conditions that model primary-to-secondary leakage (including locked rotor and control rod ejection) has been shown to remain within the accident analysis assumptions for all axial and or circumferentially orientated cracks occurring 15.21 inches below the top of the tubesheet. 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 any one steam generator. Consequently, accident leakage is approximately 10 times the allowable leakage, if only one steam generator is leaking. Using the limiting SLB/FLB overall leakage factor of 2.50, accident induced leakage is less than 0.6 gpm, if all 4 steam generators are leaking at 150 gpd at the beginning of the accident. Therefore, significant margin exists between the conservatively estimated accident induced leakage and the allowable accident leakage (1.0 gpm).

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

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

Response: No The proposed change alters the steam generator inspection and reporting criteria. It does not introduce any new equipment, create new failure modes for existing equipment, or create any new limiting single failures. Plant operation will not be altered, and safety functions will continue to perform as previously assumed in accident analyses.

Therefore, the proposed change does not create the possibility of a new or different kind of 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. It maintains the required structural margins of the steam generator tubes for both normal and accident conditions. NEI 97-06 and RG 1.121, are used as the bases in the development of the limited tubesheet inspection depth methodology for determining that steam generator tube integrity considerations are maintained within acceptable limits. RG 1.121 describes a method acceptable to the NRC for meeting GDC 14, "Reactor Coolant Pressure Boundary," GDC 15, "Reactor Coolant System Design," GDC 31, "Fracture

Attachment I to ET 12-0002 Page 19 of 23 Prevention of Reactor Coolant Pressure Boundary," and GDC 32, "Inspection of Reactor Coolant Pressure Boundary," by reducing the probability and consequences of a SGTR.

RG 1.121 concludes that by determining the limiting safe conditions for tube wall degradation, the probability and consequences of a SGTR are reduced. This RG uses safety factors on loads for tube burst that are consistent with the requirements of Section III of the American Society of Mechanical Engineers (ASME) Code.

For axially-oriented cracking located within the tubesheet, tube burst is precluded due to the presence of the tubesheet. For circumferentially-oriented cracking, the H* Analysis documented in Section 3, defines a length of degradation-free expanded tubing that provides the necessary resistance to tube pullout due to the pressure induced forces, with applicable safety factors applied. Application of the limited hot and cold leg tubesheet inspection criteria will preclude unacceptable primary to secondary leakage during all plant conditions. The methodology for determining leakage provides for large margins between calculated and actual leakage values in the proposed limited tubesheet inspection depth criteria.

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 the tubesheet to the top of the tubesheet (secondary face) that is required to maintain structural and leakage integrity over the full range of steam generating operating conditions, including the most limiting accident conditions. The H* Analysis determined that degradation in tubing below the safety significant portion of the tube does not require plugging and serves as the basis for the limited tubesheet inspection criteria, which are intended to ensure the primary-to-secondary leak 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 and safety of the public will not be endangered by operation in the proposed manner, 2) such activities will be conducted in compliance with the Commission's regulations, and 3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5. ENVIRONMENTAL CONSIDERATION WCNOC has evaluated the proposed amendment for environmental considerations. The review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, and would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the 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 set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact

Attachment I to ET 12-0002 Page 20 of 23 statement or environmental assessment needs to be prepared in connection with the proposed amendment.

6. REFERENCES

1. NRC letter from J. N. Donohew, USNRC, to R. A. Muench, WCNOC, "Wolf Creek Generating Station - Issuance of Exigent Amendment RE: Steam Generator (SG) Tube Surveillance 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 Creek Generating Station - Issuance of Amendment RE: Steam Generator Tube Inspections Within 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 Creek Generating Station - Issuance of Amendment RE: Revision to Technical Specification 5.5.9 on the Steam Generator Program (TAC NO. MD8054)," April 4, 2008. (ADAMS Accession No. ML080840004)

4. NRC letter from B. K. Singal, USNRC, to R. A. Muench, WCNOC, "Wolf Creek Generating Station - Issuance of Amendment RE: Revision to Technical Specification (TS) 5.5.9, "Steam Generator (SG) Program," and TS 5.6.10 "Steam Generator Tube Inspection 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 Steam Generator Tubes and Tube-to-Tubesheet Welds," April 7, 2005.

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

10. Westinghouse Electric Company LLC, WCAP-17071-P, Revision 0, "H*: Alternate Repair Criteria for the Tubesheet Expansion Region in Steam Generators with Hydraulically 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, "Steam Generator (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)

Attachment I to ET 12-0002 Page 21 of 23

12. NRC letter from B. K. Singal, USNRC, to R. A. Muench, WCNOC, "Wolf Creek Generating Station - Request for Additional Information Regarding the Permanent Alternate 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 to License Amendment Request for a Permanent Alternate Repair Criterion to Technical Specification 5.5.9, "Steam Generator (SG) Program"," August 25, 2009. (ADAMS Accession No. ML092450095)

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

15. LTR-SGMP-09-1 00, "LTR-SGMP-09-100 P-Attachment, "Response to NRC Request for Additional 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 Additional Information 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, "Steam Generator (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 Creek Generating Station - Transmittal of Unresolved Issues Regarding Permanent Alternate Repair 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 Regarding Tubesheet Bore Eccentricity (Model F/Model D5)," June 2011.

20. LTR-NRC-10-68, "Submittal of LTR-SGMP-10-78 P-Attachment and LTR-SGMP-10-78 NP-Attachment, "Effects of Tubesheet Bore Eccentricity and Dilation on Tube-to-Tubesheet Contact Pressure and Their Relative Importance to H*," (Proprietary/Non-Proprietary) for Review and Approval," November 9, 2010.

21. LTR-NRC-10-70, "Submittal of LTR-SGMP-10-33 P-Attachment and LTR-SGMP-10-33 NP-Attachment, LTR-SGMP-10-33 P-Attachment, "H* Response to NRC Questions Regarding Tubesheet Bore Eccentricity," (Proprietary/Non-Proprietary) for Review and Approval," 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 Bottom of the Expansion Transition (BET) for Implementation of H*," (Proprietary/Non-Proprietary) for Review and Approval," November 10, 2010.

Attachment I to ET 12-0002 Page 22 of 23

23. NRC letter from J. R. Hall, USNRC, to M. W. Sunseri, WCNOC, "Wolf Creek Generating Station - Issuance of Amendment RE: Changes to Technical Specification (TS) 5.5.9, "Steam Generator (SG) Program," and TS 5.6.10 "Steam Generator Tube Inspection Report," (TAC NO. ME5121)," April 6, 2011. (ADAMS Accession No. ML110840590)

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 on Technical Issues Regarding Steam Generator Tube Inspection Permanent Alternate Repair Criteria (TAC NOS. ME5417 and ME5418)," March 28, 2011. (ADAMS Accession No. ML110660648)

26. NRC letter to Southern Nuclear Operating Company, Inc., "Vogtle Electric Generating Plant Units 1 and 2 - Presubmittal Consideration of Steam Generator Alternative Repair Criteria Requirements Request for Additional Information (TAC NOS. ME 5417 and ME5418)," May 26, 2011. (ADAMS Accession No. ML11140A099)

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

28. Duke Energy Corporation, "Proposed Technical Specifications (TS) Amendment TS 3.4.13, "RCS Operational LEAKAGE," TS 5.5.9, "Steam Generator (SG) Program," TS 5.6.8, "Steam Generator (SG) Tube Inspection Report," License Amendment Request to Revise TS for Permanent Alternate Repair Criteria," June 30, 2011. (ADAMS Accession No. ML11188A107)

29. Duke Energy Corporation, "Proposed Technical Specifications (TS) Amendment TS 3.4.13, "RCS Operational LEAKAGE," TS 5.5.9, "Steam Generator (SG) Program," TS 5.6.8, "Steam Generator (SG) Tube Inspection Report," License Amendment Request to Revise TS for Permanent Alternate Repair Criteria," July 11, 2011. (ADAMS Accession No. 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 Generator License Amendment Request to Revise Technical Specification for Permanent Alternate Repair Criteria (TAC NO. ME6671)," January 5, 2012. (ADAMS Accession No. ML120090321)

31. Duke Energy Corporation, "Proposed Technical Specifications (TS) Amendment TS 3.4.13, "RCS Operational LEAKAGE," TS 5.5.9, "Steam Generator (SG) Program," TS 5.6.8, "Steam Generator (SG) Tube Inspection Report," License Amendment Request to Revise TS for Permanent Alternate Repair Criteria," January 12, 2012. (ADAMS Accession No. ML12019A250)

Attachment I to ET 12-0002 Page 23 of 23

32. Virginia Electric and Power Company (Dominion) letter Serial No.11-403, "License Amendment Request Permanent Alternate Repair Criteria for Steam Generator Tube Inspection and Repair," July 28, 2011. (ADAMS Accession No. ML112150144)

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 Generator License Amendment Request to Revise Technical Specifications for Permanent Alternate 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 to Request for Additional Information Related to License Amendment Request for Permanent 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 License Amendment Requests for Permanent Application of the Alternate Repair Criterion, H*, to the Model D5 and Model F SGs," February 2, 2012.

Attachment II to ET 12-0002 Page 1 of 4 ATTACHMENT II Markup of Technical Specification pages

Attachment II to ET 12-0002 Page 2 of 4 Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Proqram (continued)

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

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

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

i.Fri f~fe ag418Andhe ub e uetM'pp tircc 7,

(f'" ubes with service-induced flaws located greater than r.ilches E I below the top of the tubesheet do not require plugging. ubes with service-induced flaws located in the portion of the tube from the top 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 12-0002 Page 3 of 4 Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Program (continued)

d. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet and that Thne.. 'd thef's brseqtfe opi6ratifig yore portion of the tube below 2 inches 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 meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. An assessment of degradation shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

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

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

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

e. Provisions for monitoring operational primary to secondary LEAKAGE.

(continued)

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

Attachment II to ET 12-0002 Page 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 4 following completion of an inspection performed in accordance with the Specification 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 degradation mechanism;

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

e. Number of tubes plugged during the inspection outage for each active degradation mechanism;

f. Total number and percentage of tubes plugged to date;

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

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

.. primary o secondary LEAKAGE rate observed in each SG (if it is not practical to assign the LEAKAGE to an individual SG, the entire primary to secondary LEAKAGE should be conservatively assumed to be from one SG) during the cycle preceding the inspection which is the subject of the report;

i. Fgrowino ion ofoinsppion eior 6' in Refu*ling tag8 and y on forme in subs quent op6ratincyclehe//

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

j. Fowin mpl tion p*an in ecti pe*meoin Re/elin utn 1 ains ction'perfo(ied the sdbsedqcent ol*ratir cyct ) th6 results of monitoring for the tube axial displacement (slippage). If slippage is discovered, the implications of discovery and corrective action shall be provided.

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

Attachment III to ET 12-0002 Page 1 of 3 ATTACHMENT III Proposed Changes to Technical Specification Bases (for information only)

Attachment III to ET 12-0002 Page 2 of 3 SG Tube Integrity B 3.4.17 BASES APPLICABLE The steam generator tube rupture (SGTR) accident is the limiting design SAFETY basis event for SG tubes and avoiding an SGTR is the basis for this ANALYSES Specification. The analysis of an SGTR event assumes a bounding primary to secondary LEAKAGE rate equal to the operational LEAKAGE rate limits in LCO 3.4.13, "RCS Operational LEAKAGE," plus the leakage rate associated with a double-ended rupture of a single tube. The accident analysis for an SGTR assumes the contaminated secondary fluid is released to the atmosphere via SG atmospheric relief valves and safety valves.

The analysis for design basis accidents and transients other than an SGTR assume the SG tubes retain their structural integrity (i.e., they are assumed not to rupture.) In these analyses, the steam discharge to the atmosphere is based on the total primary to secondary LEAKAGE from all SGs of 1 gallon per minute or is assumed to increase to 1 gallon per minute as a result of accident induced conditions. For accidents that do not involve fuel damage, the primary coolant activity level of DOSE EQUIVALENT 1-131 is assumed to be equal to the LCO 3.4.16, "RCS Specific Activity," limits. For accidents that assume fuel damage, the primary coolant activity is a function of the amount of activity released from the damaged fuel. The dose consequences of these events are within the limits of GDC 19 (Ref. 2), 10 CFR 100 (Ref. 3) or the NRC approved licensing basis (e.g., a small fraction of these limits).

Steam generator tube integrity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

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

During a SG inspection, any inspected tube that satisfies the Steam Generator Program repair criteria is removed from service by plugging. If a 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 entire length of the tube, including the tube wall, between the tube-to-tubesheet weld at the tube inlet and the tube-to-tubesheet weld at the tube outlet.

.Reeli ut 187 dthe 9b ueCii:,

ýoerOatng ce< -im alternate 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 12-0002 Page 3 of 3 SG Tube Integrity B 3.4.17 BASES REFERENCES 1. NEI 97-06, "Steam Generator Program Guidelines."

2. 10 CFR 50 Appendix A, GDC 19.

3. 10 CFR 100.

4. ASME Boiler and Pressure Vessel Code,Section III, Subsection NB.

5. Draft Regulatory Guide 1.121, "Basis for Plugging Degraded Steam Generator Tubes," August 1976.

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

7. License Amendment No. (18W ctr ý13200.)

Wolf Creek - Unit 1 B 3.4.17-7 Revision 44

Attachment IV to ET 12-0002 Page 1 of 4 ATTACHMENT IV Retyped Technical Specification pages

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Program (continued)

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

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

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

1. Tubes with service-induced flaws located greater than 15.21 inches below the top of the tubesheet do not require plugging.

Tubes with service-induced flaws located in the portion of the tube from the top of the tubesheet to 15.21 inches below the top of the tubesheet shall be plugged upon detection.

(continued)

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

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Program (continued)

d. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria. The portion of the tube below 15.21 inches 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 meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. An assessment of degradation shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

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

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

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

e. Provisions for monitoring operational primary to secondary LEAKAGE.

(continued)

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

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 4 following completion of an inspection performed in accordance with the Specification 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 degradation mechanism;

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

e. Number of tubes plugged during the inspection outage for each active degradation mechanism;

f. Total number and percentage of tubes plugged to date;

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

h. The primary to secondary LEAKAGE rate observed in each SG (if it is not practical to assign the LEAKAGE to an individual SG, the entire primary to secondary LEAKAGE should be conservatively assumed to be from one SG) during the cycle preceding the inspection which is the subject of the report; The calculated accident induced leakage rate from the portion of the tubes below 15.21 inches from the top of the tubesheet for the most limiting accident in the most limiting SG. In addition, if the calculated accident induced leakage rate from the most limiting accident is less than 2.50 times the maximum operational primary to secondary leak rate, the report should describe how it was determined; and

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

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

Attachment V to ET 12-0002 Page 1 of 2 ATTACHMENT V List of Regulatory Commitments

Attachment V to ET 12-0002 Page 2 of 2 LIST OF REGULATORY COMMITMENTS The following table identifies those actions committed to by Wolf Creek Nuclear Operating Corporation in this document. Any other statements in this letter are provided for information purposes and are not considered 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 of leakage from the prior cycle from below the H* distance will be multiplied Amendment by a factor of 2.50 and added to the total leakage from any other source and compared to the allowable accident induced leakage limit. For the operational assessment (OA), the difference in the leakage between the allowable leakage and the accident induced leakage from sources other than the tubesheet expansion region will be divided by 2.50 and compared to the observed operational leakage. An administrative limit will be established to not exceed the calculated value.

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

Attachment VI to ET 12-0002 Page 1 of 4 ATTACHMENT VI Response to Request for Additional Information Questions Specific to Wolf Creek Nuclear Operating Corporation

Attachment VI to ET 12-0002 Page 2 of 4 Response to Request for Additional Information Questions Specific to Wolf Creek Nuclear Operating Corporation On June 30, 2011, Duke Energy submitted a license amendment request (Reference 28) for permanent application of the alternate repair criterion H* at Catawba Unit 2 based on the technical 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) to the license amendment request was submitted on July 11, 2011 and provided Westinghouse Electric Company LLC LTR-SGMP-11-58, "WCAP-17330-P, Revision 1 Erratum." On January 5, 2012, a request for additional information (Reference 30) was transmitted electronically to Duke 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 Power Company (Dominion) submitted a license amendment request (Reference 32) for permanent application of the alternate repair criterion H* for Surry Power Station Units 1 and 2. On January 18, 2012, the NRC issued a request for additional information (Reference 33).

Dominion responded to the request for additional information on February 14, 2012 (Reference 34).

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 Amendment Requests for Permanent Application of the Alternate Repair Criterion, H*, to the Model D5 and Model F SGs," augments the responses to the Duke Energy request for additional information to 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 information question 14 for the Model F steam generators. Provided below are Wolf Creek Nuclear Operating Corporation (WCNOC) specific responses to questions 12 and 13 from the Duke Energy request for additional information and question 15 from the Dominion request for additional information. The NRC question is identified in italics.

12. BET measurements for Catawba 2, documented in Westinghouse letter LTR-SGMP 111 P-Attachment, Revision 1, range to a maximum of 0.65 inches and appearnot to be a factor affecting the H* and leak rate ratio calculations. Apart from tubes with this reported range of BETs, are there any non-expanded or partially expanded tubes at Catawba 2? If so, provide revisions to the proposed technical specifications which exclude such tubes from the proposed H*provisions.

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

Attachment VI to ET 12-0002 Page 3 of 4 tubes. As such, revision to the technical specifications to exclude such tubes from the proposed H* provisions is not required.

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

h. For Unit 2, fo!Wowing ,,mplctin of annpcton ,.,,,perf..m.d dur;g End of Cyclc 1-7 Refucling Outage (and any knsections performed durig subsequent Cycle 18 operation) the primary to secondary LEAKAGE rate observed in each steam generator (if it is not practical to assign the leakage to an individual SG, the entire primary to secondary leakage should be conservatively assumed to be from one SG) during the cycle preceding the inspection which is the subject of the report,

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 calculatedaccident induced leakage rate from the portion of the tubes below 20 14.01 inches 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 less than 3.27 times the maximum primary to secondary LEAKAGE rate, the report shall describe 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 their intent.

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

15. Verify that regulatory commitments pertaining to monitoring for tube slippage and for primary to secondary leakage, as described in Dominion letter dated December 16, 2010 (NRC ADAMS Accession No. ML103550206), Attachment 1, page 10 of 23, remain in place. In addition, revise the proposed amendment to include a revision to technical specification limit on primary to secondary leakage from 150 gallons per day (gpd) to 83 gpd (150 divided by the proposed 1.8 leakage factor), or provide a regulatorybasis for not making this change.

Response: The regulatory commitments pertaining to monitoring for tube slippage and for primary 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 proposing any changes to the primary to secondary LEAKAGE limit as specified in TS 3.4.13, "RCS Operational LEAKAGE," based on the following.

Attachment VI to ET 12-0002 Page 4 of 4 Primary-to-secondary leakage from tube degradation is assumed to occur in several design basis accidents: feedwater line break (FLB), steam line break (SLB), locked rotor, and control rod ejection. The radiological dose consequences associated with this assumed leakage are evaluated to ensure that they remain within regulatory limits (e.g. 10 CFR Part 100, 10 CFR 50.67, GDC 19). The accident induced leakage performance criteria are intended to ensure the primary-to-secondary leak rate during any accident does 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 the tubesheet. 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 satisfied due to the constraint provided by the tubesheet. Through application of the limited tubesheet inspection scope as described below, the existing operating leakage limit provides assurance that excessive leakage (i.e., greater than accident analysis assumptions) 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 (0.1 gpm) through any one steam generator. Consequently, accident leakage is approximately 10 times the allowable leakage, if only one steam generator is leaking. Using a SLB/FLB overall leakage factor of 2.50, accident induced leakage is less than 0.6 gpm, if all 4 steam generators are leaking at 150 gpd at the beginning of the accident. Therefore, significant margin exists between the conservatively estimated accident induced leakage and the allowable accident leakage (1.0 gpm).

References:

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

2. WCNOC letter ET 09-0025, "Revision to Technical Specification (TS) 5.5.9, "Steam Generator (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 and Errata, "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....

(Non-proprietay)

(53 pages)

Westinghouse Non-Proprietary Class 3 S Westinghouse To: D.H. Warren M.W. Ryan Date: March 20, 2012 P.J. McDonough J.J. Roberts H. Mahdavy G.R. Strussion D.L. Rogosky C.W. Nitchman L.E. Markle A.M. Mrazik N. Bahtishi S.J. Hyde C.L. Mitchell J. Stepanic D.C. Beddingfield cc: B. J. Bedont C. D. Cassino From: H.O. Lagally Your ref:

Ext: 724-722-5082 Our ref: LTR-SGMMP-11-28 Fax: 724-722-5889 Errata, 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 SG Permanent 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 the Model F SGs in the response to the NRC RAI. References I and 2 contain both a proprietary (P) attachment and 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 table is properly numbered in the P-Attachment.

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

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

Prepared by: H. 0. Lagally Verified: G.W. Whiteman Steam Generator Management Regulatory Compliance And Modification Programs ElectronicallyApproved*

Approved by: Damian A. Testa, Manager Steam Generator Management And Modification Programs

Electronicallyapproved recordsare authenticatedin the electronic document management system.

Westinghouse Electric Company LLC 1000 Westinghouse Drive Cranberry Township, PA 16066

Westinghouse Non-Proprietary Class 3 LTR-SGMMP- 11-28 Rev. I NP-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.

Westinghouse Electric Company LLC 1000 Westinghouse Drive Cranberry Township, PA 16066, USA

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," License Amendment Request to Revise TS for Permanent Alternate Repair Criteria, June 30, 2011.

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

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

4. USNRC Letter, "Surry Power Station Units 1 and 2 Request for Additional Information Regarding the Steam Generator License Amendment Request to Revise Technical Specification for Permanent Alternate Repair Criteria," (TAC Nos. ME6803 and ME 6804, 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 permanent application of the alternate repair criterion H* at Catawba Unit 2 based on the technical justification in WCAP-1 7330-P, Revision 1. WCAP-1 7330-P Revision 1 also includes the technical justification for the Model F SGs at Seabrook, Salem 1, Millstone 3, Vogtle Units 1 and 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 a LAR for permanent application of H* at Surry Units 1 and 2 (Reference 3). Whereas the Catawba technical justification is contained in WCAP-1 7330-P, Revision 1, the Surry technical justification is contained in WCAP-1 7345-P, Revision 2. Although the questions in Reference 2 and Reference 4 are quite similar, some of them required different numerical information for Surry than for Catawba. Further, some of the questions in Reference 2 were not repeated in Reference 4. sion 2. A separate response will be provided for the questions contained in Reference 4.

It is anticipated that several utilities with Model F steam generators (SGs) will submit LARs for the permanent application of H* for the Model F SGs. The Model F SG technical justification is also contained in WCAP-1 7330-P, Revision 1. This document augments the responses to the Reference 2 questions to include similar responses applicable to the Model F 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. Notations are made in the response to each question regarding the applicability of the response to the Model F SGs.

Questions 1 through 11 from Reference 2 are reproduced below, followed by the responses.

Questions 12 and 13 from Reference 2 will be addressed by the respective Model F utilities.

Question 14 from Reference 4 is assumed to apply for the Model F SGs and a response is provided. Question 15 from Reference 4 is specific to the Dominion Generation (Surry 1 and

2) 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 shows the same data as Figure 3-32 in Revision 0 of the WCAP, but without the data that correspondto negative tubesheet CTE variation. The footnote states that while only a few percent of the data shown in Figure 3-32 of Revision 0 reflect negative values of tubesheet CTE, these cases do result in upward scatter,but must be included to properly representthe top 10% of the Monte Carlo rank order results. This being the case, why does Figure 3-32 in Revision I properly represent the top 10% of the Monte Carlo rank order results? Why are the minimum H* values in Figure 3-36 of Revision I substantiallydifferent 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 from the same database. The title of Figure 3-36 in WCAP-17330-P, Revision 1 is correct; it applies 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 the figures 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 the data 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 of the sum of the squares of the component pairs of Monte Carlo selected values of coefficients of 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 the Revision 0 figures is the result of the Monte Carlo process that results in samples with negative variations of the tubesheet coefficient of thermal expansion with corresponding large 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 of tubesheet CTE and negative variations of tube CTE. In the Monte Carlo analysis, described further in the response to Question 3, approximately half of the H* values include a negative variation 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 the sign 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 TS CTE were assigned an H* value corresponding to a TS CTE variation of zero but with the Monte 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 "responsesurface" (i.e., H*

relationship to coefficients of thermal expansion (CTE) variabilityfor the tube and tubesheet) discussed for Model D5 steam line break (SLB) at the top of page 3-49.

Confirm that this response surface applies to a radiallocation of 26.703 inches. Is this a full response surface or "partial"response surface of the type discussedin Revision 1 of WCAP-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 a radial 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 tubesheet CTE 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 CTE Case # H*+BET (in) a,c,e 2

3 4

5 6

7 8

9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 34 35 36 37 38 39 _____

5

LTR-SGMMP- 11-28 Rev. I NP-Attachment a,c,e 40 41 42 43 44 45 L Question 3:

WCAP-1 7330-P, Revision 1 - Provide copy of the "reduced"response surfaces for bounding Model D5 SLB case discussed on page 3-58. Explain how the reduced response surfaces are used in the Monte Carlo analysis. If for a particularMonte Carlo iteration a negative variation of tubesheet CTE is randomly generated,what is done with this value (e.g., is tubesheet CTE assumed to have nominal value)? Why doesn't the use of a reduced response surface bias the rank orderingabove 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 the Model 51 F SGs, the modified question is considered more appropriate for the Model F SGs.

WCAP- 17345-P, Revision 2, Section 3.4 - Confirm that the Monte Carlo analyses performed for the Model 51F SGs using the thick shell model are based upon sampling of the full H*/CTE response surfaces in Figure 8-5 of WCAP 17092 Rev 0. If this is incorrect, and only a "reduced"response surface is used, explain how the reduced response surfaces are used in the Monte Carlo analysis. If for a particularMonte Carlo iteration a negative variation of tubesheet CTE is randomly generated,what is done with this value (e.g., is tubesheet CTE assumed to have nominal value)? Why doesn't the use 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 the critical tubesheet radius of [ ]a,c,e inches. Note that the change in the maximum value of H* (see Case 45) at the critical radius of [ ]a.ce inches from the prior critical radius of 26.703 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 bias the rank ordering in a non-conservative direction; it simply limits the effort to develop a response surface to the region in parameter space where the limiting values of H* are most 6

LTR-SGMMP- 11-28 Rev. I NP-Attachment likely located. The interpolation method for the reduced response surface permits calculation of H* values with the thick-shell equation, which is the underlying calculation basis of the response surface. The Monte Carlo process randomly samples, including variances in the region 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 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 the probabilistic calculation of H* using the C2 model.

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

1. Pick a random normal deviate to represent the tubesheet CTE variation.

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

3. For each tube, assign an H* value corresponding to the current tubesheet CTE variation and the tube's CTE variation by interpolating an H*value on the response surface. If the tubesheet CTE variation is negative, interpolate as though 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, although the H* assigned to them is conservative by step 3.

Steps 1-5 represent one iteration of the Monte Carlo process. This process is repeated 10,000 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 artificially higher values of H* are assigned to the combination of randomly selected CTEs when the selected tubesheet CTE is negative. The true H* rank order of these cases is lower than the apparent value of H* for these cases. The effect is to displace the rank order of H*s with positive 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* calculation process with the C 2 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 event that the 9 5 0 0 th value contained a negative tubesheet CTE variation, the next higher rank order value with a positive tubesheet CTE was chosen. In practice, only one or two rank orders needed to be traversed to find an H*with a positive tubesheet variation. The parameters 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 the first 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 full H*/CTE response surfaces in Figure 8-5 of WCAP 17071-P, which is based on application of the thick-shell model.

The Monte Carlo process randomly samples from the response surface by means of an 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 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 the probabilistic calculation of H* using the C2 model.

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

1. Pick a random normal deviate to represent the tubesheet CTE variation.

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

3. For each tube, assign an H* value corresponding to the current tubesheet CTE variation and the tube's CTE variation by interpolating an H* value on the response surface. Ifthe tubesheet CTE variation is negative, interpolate as though 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 repeated 10,000 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 artificially higher values of H* are assigned to the combination of randomly selected CTEs when the selected tubesheet CTE is negative. The true H* rank order of these cases is lower than the apparent value of H*for these cases. The effect is to displace the rank order of H*s with positive 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 is chosen. In the event that the 9500th value contained a negative tubesheet CTE variation, the next higher rank order value with a positive tubesheet CTE was chosen. In practice, only one or two rank orders needed to be traversed to find an H* with a positive tubesheet variation. The parameters 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 the first higher rank order with a positive tubesheet CTE variation is conservative. The same process is utilized when determining the H* value for the higher probabilistic goals applicable to the Model F, that is, the 95/95 whole plant value of H*.

8

LTR-SGMMP- 11-28 Rev. I NP-Attachment Table 3-1 Reduced Response Surface; Model D5, [ ]ac"e inches Radius TS CTE T CTE H*+BET Case # n on o(in)

Q[ ]""'Radius) a,c,e 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 9

LTR-SGMMP- 11-28 Rev. I NP-Attachment 39 a,c,e 40 41 42 43 44 45 L 10

LTR-SGMMP-1 1-28 Rev. 1 NP-Attachment Question 4:

WCAP- I 7330-P, Revision 1, Table 3 Provide a similar table applicable to the Model D5 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 3-28 of WCAP-17330-P, Revision 1 contains the data for the Model F SGs, centered on rank order 9890.

Table 4-1 provides the requested information.

Table 4-1 Variation of CTEs Over a Range of Rank Order Statistics for Model D5 Rank Tube Tubesheet Alpha(1 )

CTE CTE a,c,e 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)

II

LTR-SGMMP- 11-28 Rev. 1 NP-Attachment Question 5:

WCAP-17330-P, Revision 1, Table 3 Provide C2 H* values for rank orders 9888 and 9892. This will lend additionalconfidence to inferences drawn from this table on page 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 ANSYS Workbench, Version 11. Version 12.1 of ANSYS Workbench was released following the issue of WCAP-17330-P, Revision 1. The updates to this version of ANSYS Workbench include changes to the contact modelling and solver options. Westinghouse has benchmarked and configured this version of the ANSYS code and has verified the results and conclusions of the previous H* analyses obtained with Version 11. However, there are minor numerical differences in the results. The net difference of applying version 12.1 of the ANSYS code compared to version 11 of the ANSYS code is a slight variation in the average circumferential contact pressure, typically on the order of +/- 40 psi. Version 11 generally produces the lower contact pressures. Consequently, there may be small differences in the values 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 probabilistic Model 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

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

[ ] ace ]a.c.e [ ]a.c e 9 53 8 9539 [ a,c,e [ ]a,c,e [ ]a,,e Notes:

(1) Refer to LTR-SGMP-11-58, "WCAP-17330-P Revision 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 of WCAP-17330-P, Rev. 1), the inferences drawn from these data on page 3-56 of WCAP-1 7330-P, Rev. 1 remain valid. It is expected that small variations will occur due to factors 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 C 2 model. This uncertainty is on the order of 2% of the final H* value, which is more than adequately covered by other conservatisms in the H* value that are discussed in the responses to the other questions.

13

LTR-SGMMP-1 1-28 Rev. 1 NP-Attachment Question 6:

WCAP- 1 7330-P, Revision 1, Figure 3 Should the data correspondingto the two open symbols be labeled as "dataused in probabilisticanalysis"(consistent with Figure 3-44) instead of "reduceddata?" Why does this figure show only two open symbols ratherthan 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. This question 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. No differentiation 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 that there are only two points. For clarity, the Table 6-1 provides the coordinates of the three points 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 a dark grey square.

Table 6-1 Coordinates of Three Open-symbol Points on Figure 3-45 of WCAP-17330-P, Revision I Rank H* Tube CTE Tubesheet Alpha CTE 9149 [ ]a,c,e [ ]a,c,e [ ]a,c,e 3.513 9500 ))ac,e [ ]a,c,e 3.750 9536 ]a.,c,e [ ]a,c,e ]a,c,e 3.733 14

LTR-SGMMP-1 1-28 Rev. 1 NP-Attachment ac,e Figure 6-1 Update of Figure 3-45 of WCAP-17330, Revision I 15

LTR-SGMMP- 11-28 Rev. 1 NP-Attachment Question 7:

WCAP-1 7330-P, Revision 1, Tables 3-35 to 3 The numerical methods used to generate the accumulatedpullout loads in these tables appearto contain two sources of non-conservatism. One, the distance below the top of the tubesheet (TTS) where the contact pressure transitionsfrom zero to a positive non-zero value is assumed to be the lowermost elevation for which a C2 calculation was performed and yielding a zero value contactpressure. The staff believes a more realisticand more conservative estimate of the contact pressure zero intercept value can be obtained by extrapolating the C2 results at lower elevations to the zero interceptlocation. Two, the method used to interpolate the H* distance between specific locations where C2 analyses were performed assumes that the distribution of contact pressure between these locations is a constant value equal to average value between these locations. For Table 3-35, the staff estimates that elimination of the non-conservatismsincreases the calculated H* by 0.34 inches. For Tables 3-46 and 3-48, H* increasesby 0. 15 inches. These are not trivial differences.

The staff estimates that the pullout loads corresponding to the H* distancesin Figures 3-35, 3-46, and 3-48 are overestimated by 17%, 6%, and 8%, respectively. Provide revisions 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 D5 contact pressure curve showed that extrapolation of data points provided in WCAP-17330-P, Revision 0 was unrealistically conservative. While a higher point density would always provide more certainty in the result, the current density of points was judged adequate by Westinghouse and (implicitly) by MPR in their independent review of H* methodology based on the minor effect on H*. In response to this question, another point was added to the contact 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 7-1 below. Figure 7-1 shows that the extrapolation proposed by the question is unrealistically conservative and that such an extrapolation is also inconsistent with the behavior of a real structure. A sharp break in the contact pressure curve would not be expected in the physical structure; rather, a smooth transition from zero to non-zero contact pressure would be expected. Figure 7-1 shows that addition of even more points would simply further define the smooth 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

LTR-SGMMP-1 1-28 Rev. 1 NP-Attachment a,c,e Figure 7-1 Model 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 F and 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 the difference between the mean H* and the probabilistic H* is entirely based on CTE differences, 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 of the tube/tubesheet CTE variance ratios from Table 7-2 were used.

17

LTR-SGMMP- 11-28 Rev. I NP-Attachment Table 7-3 summarizes the H* values contained in WCAP-1 7330, Revision 1 for the Model D5 and 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 Poisson contraction because neither of these are related to CTE. As can be seen from Table 7-4, the existing 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 the conservatism inherent in the current H* calculations are adequately conservative to account for small differences in judgment on the calculation process even without considering the major conservatisms identified previously (i.e., neglecting residual contact pressure).

Additional conservatism to further support this conclusion is identified below.

Table 7-1 CTE 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

LTR-SGMMP-I 1-28 Rev. I NP-Attachment Table 7-2 Ratio of CTE Variances (Refined/Used in Current H*)

Temperature Tube CTE SDs Ratios

(*F) Confidence 95% Confidence 300 [ ]a,c,e [ ].,ce 600 [ ]a,c,e [ ] a,c,,e Table 7-3 Summary of H* Lengths from WCAP-17330, Revision 1 Limiting f

Probabilistic H* Difference, Ratio Mean H*

(inches) (inches) Probabilistic - Mean Table 7-2 Table 7-2 F, 95/50 Whole Bundle _

F, 95/95 Whole Plant D5, 95/50 Whole Bundle D5, 95/95 Whole Bundle Table 7-4 Estimate of Conservatism of H* Length Related to CTE Variance Difference x Difference Dfeec Model/Case DimitincR New Probabilistic H* (Licensed H* - New F,_95/50_WholeLimiting Ratio Probabilistic H*)

F, 95/50 Whole Bundle F, 95/95 Whole Plant D5, 95/50 Whole Bundle D5, 95/95 Whole Bundle 19

LTR-SGMMP- 11-28 Rev. 1 NP-Attachment Question 8:

WCAP- I 7330-P, Revision 1, Figures3-48 and 3 These figures were generated with the thick shell model. Were "spotchecks" performed with the C2 model to determine whether adjustments to the curves in these figures are needed to approximate what the curves would look like if entirely generated with the C2 model? If not, why are the curves in their presentform 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 and the 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. The contact pressure distributions developed in WCAP-17330-P, Revision 1 assume that the crevice pressure is distributed over the full depth of the tubesheet. No "spot checks" were performed to test if the crevice pressure correction distribution, determined by the thick shell equations (shown in Figures 3-48 and 3-49 of WCAP-17330, Revision 1), required an adjustment when applied to the C2 model results. The adjustment to the final H*length in Tables 3-50 and 3-51 of WCAP-1 7330-P, Revision 1 was made to be consistent with the methodology described in WCAP-17072-P.

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

As discussed in WCAP-1 7072-P, the crevice pressure distribution was proportionally adjusted through the thickness of the tubesheet to reflect the predicted H*tube length because the tube below any postulated 3600, 100% through-wall flaw, is assumed to be absent. The crevice pressure at, and below, the flaw depth is in equilibrium with the primary side pressure. Increasing the crevice pressure over the length of the predicted H*so that it is equal to the primary side pressure reduces the tube to tubesheet contact pressure and increases 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 that there 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 contact pressure. Therefore, the pressure in the crevice below the top of the tubesheet to the point of departure from zero contact pressure experiences the full primary to secondary pressure differential because that length of crevice is at the secondary side pressure condition. During a Model D5 steam line break, this pressure differential is equal to 2560 psid, acting towards the tubesheet. For the Model F, during normal operating conditions, the pressure differential is 1453 psid, acting toward the tubesheet.

20

LTR-SGMMP- 11-28 Rev. 1 NP-Attachment Figure 8-1(a) shows a comparison of the unmodified crevice pressure distribution used in the C2 analysis (i.e., the crevice pressure is distributed over the full depth of the tubesheet) and the crevice pressure distribution that has been adjusted to reflect the final contact pressure distribution reported in Table 3-48 in WCAP-1 7330-P, Revision 1 for the critical radius in the Model D5 SG. Similarly, Figure 8-1(b) shows the same comparison for the Model F SGs based on the data in Table 3-46 in WCAP-17330-P, Revision 1. In effect, the normalization of the crevice pressure distribution must be based on the shorter distance defined by the distance 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 across the tube over the H* length increases. The increased pressure differential results in a large increase in the contact pressure between the tube and the tubesheet at the upper portion of the 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 both ends of the tube portion within the tubesheet with positive contact pressure between the tube and the tubesheet. The double iteration significantly increases the time required to perform the analysis and it is conservative to neglect it. Including the effect of the increased pressure differential 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 tubesheet using the probabilistic results from Tables 3-41 and 3-42 in WCAP-17330-P, Revision 1 and the adjusted crevice pressure distribution shown in Figures 8-1(a and b). The increase in contact pressure due to adjusting the crevice pressure at the top of the tubesheet occurs regardless of the predicted length of H* if the underlying contact pressure distribution includes a length of zero contact pressure at the top of the tubesheet. Therefore, neglecting the crevice pressure distribution adjustment in the zero contact pressure length for any predicted H* length provides additional margin to the calculation of H*. The conservative application 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 an overestimate of the H* length by more than 1 inch, before the additional crevice pressure adjustment from Figures 3-49 and 3-48 in WCAP-1 7330-P, Revision 1 are added respectively for the Model D5 and Model F SGs.

Figures 8-3 (a and b) show that no adjustment to the final probabilistic contact pressure distribution for crevice pressure distribution is necessary. The probabilistic contact pressure distribution is the contact pressure profile that is determined by the C2 model when the probabilistic values of inputs (CTEs, displacements) are input to the C2 model. The unadjusted (for crevice length) crevice pressure differential distribution, when applied to the probabilistic contact pressure distribution, results in a near-worst-case result for H* because the contact pressure is much less sensitive to crevice pressure variations than it is to variations 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 applied tubesheet displacements and temperatures throughout the tubesheet depth are kept the same as shown in Tables 3-10 and 3-16, respectively for the Model D5 and Model F SGs, in 21

LTR-SGMMP-I 1-28 Rev. I NP-Attachment WCAP-17330-P, Revision 1, but the crevice pressure differential is held constant at 1 psi throughout 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 inputs are kept the same, but the crevice pressure differential is held constant at 2560 psid for the Model D5 throughout the depth of the tubesheet (i.e., secondary pressure in the crevice), the result is the "DP=2560 psi" curve in Figure 8-3 (a). Likewise, if the C 2 model inputs are kept the same, but the crevice pressure differential is held constant at 1453 psid for the Model F throughout the depth of the tubesheet (i.e., secondary pressure in the crevice), the result is the "DP=1453 psi" curve in Figure 8-3 (b).These are the bounding conditions for crevice pressure. It is not possible for variation in crevice pressure differential to produce a contact pressure distribution less than, or greater than, the space bounded by these two curves. The current probabilistic contact pressure distribution, with the unmodified crevice pressure differential, 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 unmodified crevice pressure distribution used in WCAP-17330-P, Rev. 1, and the contact pressure distribution with the worst-case assumption of a 1 psi differential, is essentially negligible for the Model D5 and small for the Model F.

When the modified crevice pressure differential distribution (i.e., based on the shorter crevice length) is applied, the result is increased contact pressure as illustrated in Figures 8-4(a and b). Increased contact pressure results in a reduced H* value. However, for consistency with the 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 analysis methodology, 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, the 0.68 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 of the crevice pressure adjustment curve as it is applied in the C 2 analysis methodology is not required.

22

LTR-SGMMP-1 1-28 Rev. I NP-Attachment a,c,e Figure 8-1(a): Model D5: Plot of Crevice Pressure Differential acting towards the tubesheet on the 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 on the 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

LTR-SGMMP-1 1-28 Rev. I NP-Attachment a,c,e Figure 8-2(a): Model D5: Plot of tube-to-tubesheet contact pressure for the modified and unmodified crevice pressure differential distributions shown in Figure A. The zero (0) elevation is the top of the tubesheet.

a,c,e Figure 8-2(b): Model F: Plot of tube-to-tubesheet contact pressure for the modified and unmodified crevice pressure differential distributions shown in Figure A. The zero (0) elevation is the top of the tubesheet.

24

LTR-SGMMP-1 1-28 Rev. 1 NP-Attachment a,c,e Figure 8-3(a): Model DS: Plot of tube-to-tubesheet contact pressure as a function of crevice pressure 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 crevice pressure distribution. The zero (0) elevation is the top of the tubesheet.

25

LTR-SGMMP-1 1-28 Rev. I NP-Attachment a,c,e Figure 8-4(a) Model D5: Composite plot showing the effect on contact pressure of adjusting crevice pressure distribution to account for zero contact pressure near the top of the tubesheet.

a,c,e Figure 8-4(b) Model F: Composite plot showing the effect on contact pressure of adjusting crevice pressure distribution to account for zero contact pressure near the top of the tubesheet.

26

LTR-SGMMP-1 1-28 Rev. 1 NP-Attachment Question 9:

In addition to the potentialnon-conservatisms in the H* estimate discussed in Question 7 above, there is uncertaintyassociated with the computed probabilisticH* values calculated with the C2 model as illustratedin Table 3-29. Depending on the response to question 8 above, there also may be some uncertaintyassociated with the H*

adjustments for the crevice pressure distribution. What change to the proposed H* value of 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 F2 probabilistic H* estimates are necessary to account for the uncertainty associated with the C 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 the technically justifiable value. The current Model F H* estimate of 15.21 inches is conservative by approximately 5.5 inches compared to the technically justifiable value. These margins are in addition to the significant conservatism of neglecting residual contact pressure and other conservatism identified previously.

For the Model D5 SGs, the probabilistic H* value, before any adiustments, cited in Table 3-49 in WCAP-17330-P, Rev. 1 is [ ]a"' inches. The probabilistic H*value for the contact pressure 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 3-49 in WCAP-17330-P, Rev. 1 is [ ac"e inches. The probabilistic H* value for the contact pressure 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* estimate compared to the adjustments that are demonstrated above in the current technical basis for H*. It is seen from Table 9-1 that a margin of [ ]a.c~e inches exists in the currently recommended H* length of 14.01 inches for the Model D5 SGs when the conservatism in the crevice pressure adjustment and the measurement error in the CTE data are quantified and the proper adjustments are made. Table 9-2 shows that a margin of [ ]a,c,e exists in the currently recommended H* length of 15.21 inches for the Model F when the conservatism in the crevice pressure adjustment and the measurement error in the CTE data are quantified and the proper adjustments are made. These previously un-quantified conservatisms significantly exceed the potential increase in the H* length if different judgments are made in the details of the H*calculation as suggested in Questions 7, 8 and 9. Based on this, it is concluded that no adjustments to the recommended probabilistic H* value of 14.01 inches for the 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

LTR-SGMMP- 11-28 Rev. I NP-Attachment Table RAI 9-1 Conservatism in Current Model D5 H* Calculation WCAP-17330-P, Refined Source Rev 1 Calculations

_ in in-Unmodified H* Value Adjustments Poisson Correction Crevice Pressure and BET Adjustment CTE Uncertainty Adjustment (RAI 7)

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 inch included in Pcrev correction (Figure 3-49 of WCAP-17330, Rev. 1)

(3) See response to Question 7.

Table RAI 9-2 Conservatism 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 inch included in Pcrev correction (Figure 3-48 of WCAP-17330, Rev. 1)

(3) See response to Question 7.

28

LTR-SGMMP-1 1-28 Rev. 1 NP-Attachment Question 10:

Westinghouse letter LTR-SGMP-10-95 P - Attachment, Revision 1 - The staff is able to reasonablyreproduce the numbers in Table 5 for Exp-2 and Power-2. It is the staff's understandingthat Table 4 contains intermediate results leading to the results in Table

5. However, the staff cannot reproduce the numbers in Table 4 based on the information provided. Is Table 4 correctly titled? Provide a precise definition of the parametersthat are listed in Table 4. Provide one example of how the parametervalues were 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 of the 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 transcription error.

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 the average contact pressures over each segment length. The values on Table 10-1 are the solution for K from the "Power-2" function.

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

29

LTR-SGMMP-1 1-28 Rev. I NP-Attachment Table 10-1 Local Loss Coefficient for Power 2 (K=0.15*(Pc) 4 '5)

Segment Tubesheet Radius Lengths 4.437 10.431 18.139 26.703 42.974 49.825 from BTS to TTS Local K- NOP 2.00 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+13 2.00 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+13 6.386 1.7653E+13 7.5284E+12 6.7741E+12 8.3479E+12 5.1448E+13 1.7803E+14 2.129 6.0972E+09 9.2123E+08 1.8742E+09 4.8467E+10 3.0885E+13 2.7622E+14 1.00 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-SLB 2.00 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 1.00 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 4.0011E+11 1.2318E+14 1.00 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 O.OOOOE+00 6.2667E+11 2.0023E+14 30

LTR-SGMMP- 11-28 Rev. I NP-Attachment Table 10-2 Segment Resistance Based on Viscosity in (Ibf-sec/inA2) Units for Power 2 (K=0.15*(Pc) 4'5)

Segment Tubesheet Radius Lengths 4.437 10.431 18.139 26.703 42.974 49.825 from BTS to TTS Normal Operating Conditions 2.00 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 Conditions 2.00 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 2.129 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 6.13E+04 5.29E+06 1.00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 1.10E+04 3.37E+06 1.00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 1.72E+04 5.48E+06 31

LTR-SGMMP-I 1-28 Rev. I NP-Attachment Question 11 Westinghouse letter LTR-SGMP-10-95 P - Attachment, Revision 1 - This report spells out the definition of Exp-2 and Power-2 in Table 5. Provide definitions of the other functions consideredin 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 in LTR-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 not mathematical fits to the data; rather, they are functions developed to represent various interpretations of the loss coefficient data.

Function Definition: Note Exp-1 K = 1E+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= 1E+4*PcA3 Power-2 K = 0.15*(Pc) 45 Diagonal Bound Logarithmic K = 1E+12*In(Pc)+4E+08 Question 12 This question is a utility-specific question for which the respective utilities provide specific responses.

Question 13 This question was a Catawba specific and does not apply to either the Model 51F or the Model F SGs.

32

LTR-SGMMP-I 1-28 Rev. I NP-Attachment Question 14

[LlfrAP 17** p, F5 Re,.ien 2, Taob*6c 3 50 3651

-2 WCAP-17330-P, Revision I Table 3 Are Is the footnotes in

this table correct and complete? For Model 64F, Table 3-27 implies we have direct C2 calculationsfor rank orders 9025, *673, and 00019186, 9694 and 9890. Thus, for Table 3-6450, it seems a4 three of four cases are based on interpolated values. Sim..a.*!;, for -A-do!

4W F, Ta!b; 3 27 ifflp!!2 W.i'2h*'. d-rot C 96!'-!ta .9..- for&Fn!&k orders 0158, 0607, and 7 Thus,

. fox T. 3 50, it . not.' ...... .....

ey - is bacsed .f. dA.r-t-- - !2--!2t..M.6 .ndth* other2 2C 2r2 ..tcrpoiated'.'2!--i . If the 2

staff's understandingis incorrect,clarify for which rank orders direct C calculationswere performed and provide the H* calculationsfor these cases in a form similarto Tables 3-45 to 3-48.

Response

This question did not appear in Reference 2 for the Model D5 but did appear in Reference 4 for the Model 51 F. With appropriate references in the question (see above), it can be considered 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 the Model 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 9186 through 9890. Only one of the rank orders of interest, which define the key probabilistic targets 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 interest is a straight line function. Consequently, because the points of interest lay within the range of calculated values, and the function is linear, it is appropriate to interpolate to determine the H*values.

Question 15 This question is specific to the Dominion LAR for H*. A similarquestion may apply for the Model F SGs in which case a response must be provided by the utility with Model F SGs that has submitted an LAR for applicationof a permanent H* ARC.

33

LTR-SGMMP-1 1-28 Rev. I NP-Attachment Appendix A LTR-SGMP-11-87 (Reference 3-17 of WCAP-17330-P, Revision 1) 34

LTR-SGMMP- 11-28 Rev. 1 NP-Attachment To: G. W. Whiteman Date: May5,2011 B. J. Bedont C. D. Cassino cc:

From: A. 0. Roslund Your ref:

Ext: 724-722-6473 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 Region in Steam Generators with Hydraulically Expanded Tubes (Model F)".

2. LTR-0026-0087-2, "Independent Technical Review of H* Steam Generator Tube Alternate Repair Criterion," 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 tube and tubesheet CTE for H* were calculated in response to questions from MPR in the independent review of H*.

Electronically Approved* Electronically Approved*

Prepared by: A. 0. Roslund Verified: H. 0. Lagally SGDA SGMP Electronically Approved*

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

35

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

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

LTR-SGMMP-1 1-28 Rev. I NP-Attachment 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 10 a,c,e 100 125 150 175 200 225 250 275 300 325 350 375 400 425 450 475 500 525 550 575 600 625 650 675 700 38

LTR-SGMMP-1 1-28 Rev. I NP-Attachment Table 15 Tubesheet CTEs (gin / in IF)

Temp (*F) Sample 11 Sample 12 Sample 13 Sample 14 Sample 15 Sample 16 Sample 17 Sample 18 Sample 19 Sample 20 100 F- a,c,e 125 150 175 200 225 250 275 300 325 350 375 400 425 450 475 500 525 550 575 600 625 650 675 700 39

LTR-SGMMP-1 1-28 Rev. I NP-Attachment Table 3 Tubesheet CTEs (,Iin / in IF)

Temp (*F) Sample 21 Sample 22 Sample 23 Sample 24 Sample 25 Sample 26 Sample 27 Sample 28 Sample 29 Sample 30 100 F a,c,e 125 150 175 200 225 250 275 300 325 350 375 400 425 450 475 500 525 550 575 600 625 650 675 700 _____ __________ __________ __________ __________ ____

40

LTR-SGMMP-1 1-28 Rev. I NP-Attachment 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 40 100 a,c,e 125 150 175 200 225 250 275 300 325 350 375 400 425 450 475 500 525 550 575 600 625 650 675 700 41

LTR-SGMMP-1 1-28 Rev. I NP-Attachment Table 5 Tube CTEs (Model F) (uin / in °FI Temp (*F) Sample 1 Sample 2 Sample 3 Sample 4 Sample 5 Sample 6 Sample 7 Sample 8 Sample 9 Sample 10 100 a,c,e 125 150 175 200 225 250 275 300 325 350 375 400 425 450 475 500 525 550 575 600 625 650 675 _____ _____ _____ _____

7O0 _____ _____ _____ _____ _____

42

LTR-SGMMP-1 1-28 Rev. I NP-Attachment Table 6 Tube CTEs (Model D5) (pin / in IF)

Temp (*F) Sample 11 Sample 12 Sample 13 Sample 14 Sample 15 Sample 16 Sample 17 Sample 18 Sample 19 Sample 20 100 F a,c,e 125 150 175 200 225 250 275 300 325 350 375 400 425 450 475 500 525 550 575 600 625 650 675 700 43

LTR-SGMMP-1 1-28 Rev. I NP-Attachment Table 7 Tube CTEs (Model 44F) (ftin / in OF)

Temp (*F) Sample 21 Sample 22 Sample 23 Sample 24 Sample 25 Sample 26 Sample 27 Sample 28 Sample 29 Sample 30 100 F a,c,e 125 150 175 200 225 250 275 300 325 350 375 400 425 450 475 500 525 550 575 600 625 650 675 IL 700 44

LTR-SGMMP-1 1-28 Rev. I NP-Attachment 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 a,c,e 100 125 150 175 200 225 250 275 300 325 350 375 400 425 450 475 500 525 550 575 600 625 650 675 700 45

LTR-SGMMP-1 1-28 Rev. I NP-Attachment 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 a,c,e 100 125 150 175 200 225 250 275 300 325 350 375 400 425 450 475 500 525 550 575 600 625 650 675 700 46

LTR-SGMMP-11-28 Rev. 1 NP-Attachment Table 10 Mean 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

LTR-SGMMP-11-28 Rev. 1 NP-Attachment Table 11 Mean and Standard Deviation, Tubesheet Material Temperature Mean Best Estimate Standard 95% Confidence Standard

(°F) (pin/in*F) Deviation (%) Deviation (%)

100 6.11 2.71 3.34 125 6.23 2.30 2.83 150 6.35 1.96 2.42 175 6.45 1.69 2.08 200 6.55 1.48 1.82 225 6.63 1.31 1.62 250 6.71 1.19 1.46 275 6.79 1.09 1.35 300 6.85 1.02 1.26 325 6.91 0.97 1.19 350 6.97 0.92 1.14 375 7.02 0.89 1.10 400 7.07 0.86 1.06 425 7.12 0.84 1.03 450 7.16 0.82 1.01 475 7.20 0.80 0.99 500 7.24 0.79 0.97 525 7.28 0.77 0.95 550 7.32 0.76 0.94 575 7.35 0.76 0.93 600 7.39 0.75 0.92 625 7.43 0.74 0.92 650 7.48 0.75 0.92 675 7.52 0.76 0.93 700 7.57 0.78 0.96 48

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

LTR-SGMMP-11-28 Rev. 1 NP-Attachment 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 0.58 600 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

LTR-SGMMP-11-28 Rev. 1 NP-Attachment Table 14 High-Confidence Tube and Tubesheet Standard Deviations with Instrumentation Error Removed Temperature Tube (%) Tubesheet (%)

(*F) a,c,e 100 125 150 175 200 225 250 275 300 325 350 375 400 425 450 475 500 525 550 575 600 625 650 675 700 51

Enclosure III to ET 12-0002 Enclosure III Westinghouse Electric Company LLC CAW-12-3417, "Application for Withholding Proprietary Information from Public Disclosure" (7 pages)

Nuclear Services WestinghouseWestinghouse Electric Company 1000 Westinghouse Drive Cranberry Township, PA 16066 USA U.S. Nuclear Regulatory Commission Direct tel: (412) 374-4643 Document Control Desk Direct fax: (724) 720-0754 11555 Rockville Pike e-mail: greshaja@westinghouse.com Rockville, MD 20852 Proj letter: SAP-12-31 CAW-12-3417 February 22, 2012 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE

Subject:

LTR-SGMMP-1 1-28 Rev. 1 P-Attachment, "Response to USNRC Request for Additional Information Regarding the License Amendment Requests for Permanent Application of the Alternate 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 is further 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 basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR Section 2.390 of the Commission's regulations.

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

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

Very truly yours, JZA. Ger Regulatory 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 PENNSYLVANIA Notarial Seal Cynthia Olesky, Notary Public Manor Boro, Westmoreland County My Commission Expires July 16, 2014 Membbr, 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) 1am 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 NRC in 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 the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary versions is contained within brackets, and where the proprietary information has been deleted in the non-proprietary versions, only the brackets remain (the information that was contained within the brackets in the proprietary versions having been deleted). The justification for claiming the information so 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 being identified as proprietary or in the margin opposite such information. These lower case letters refer to the types 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 to make the number of copies of the information contained in these reports which are necessary for its internal 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 public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding. With respect to the non-proprietary versions of these reports, the NRC is permitted to make the number of copies beyond those necessary for its internal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose.. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary.

v t e Letter Sequence Request
TAC:ME8350, (Approved, Closed)
Initiation Request, Request Acceptance... Administration Withholding Request Acceptance Questions Answers Results Approval Other:
MONTHYEAR2012-12-11 [Table View]

ET 12-0002, License Amendment Request, Revision to Technical Specification (TS) 5.5.9, Steam Generator (SG) Program, and TS 5.6.10, Steam Generator Tube Inspection Report, for a Permanent Alternate Repair Criteria

Text

Subject:

Subject:

SUMMARY

SUMMARY

Background

References:

Subject:

Reference:

References:

Response

Response

Response

Response

Response

Response

Response

Response

Response

Response

Response

Response

Subject:

References:

Subject:

Navigation menu

ET 12-0002, License Amendment Request, Revision to Technical Specification (TS) 5.5.9, Steam Generator (SG) Program, and TS 5.6.10, Steam Generator Tube Inspection Report, for a Permanent Alternate Repair Criteria

Text

Subject:

Subject:

SUMMARY

SUMMARY

Background

References:

Subject:

Reference:

References:

Response

Response

Response

Response

Response

Response

Response

Response

Response

Response

Response

Response

Subject:

References:

Subject:

Navigation menu

Search