NL-12-0561, License Amendment Request to Revise Technical Specification Sections, 5.5.9, Steam Generator (SG) Program and 5.6.10, Steam Generator Tube Inspection Report.

License Amendment Request to Revise Technical Specification Sections, 5.5.9, Steam Generator (SG) Program and 5.6.10, Steam Generator Tube Inspection Report.

ML12087A307
Person / Time
Site:	Vogtle
Issue date:	03/22/2012
From:	Ajluni M Southern Nuclear Operating Co, Southern Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-12-0561, CAW-12-3404
Download: ML12087A307 (102)

Text

Mark J. Ajiluni, P.E. Southern Nuclear Nuclear Licensing Director Operating Company, Inc.

40 Inverness Center Parkway Post Office Box 1295 Birmingham, Alabama 35201 Tel 205.992.7673 Fax 205.992.7885 March 22, 2012 SOUTHERN k COMPANY Docket Nos.: 50-424 NL-12-0561 50-425 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Vogtle Electric Generating Plant - Units 1 and 2 License Amendment Request to Revise Technical Specification Sections 5.5.9, "Steam Generator (SG) Program" and 5.6.10, "Steam Generator Tube Inspection Report" Ladies and Gentlemen:

Pursuant to 10 CFR 50.90, Southern Nuclear Operating Company (SNC) hereby requests an amendment to Facility Operating License Nos. NPF-68 and NPF-81 for Vogtle Electric Generating Plant Units 1 and 2 (VEGP). This amendment request proposes to revise VEGP Technical Specification (TS) 5.5.9, "Steam Generator (SG) Program," to permanently exclude portions of the tube below the top of the SG tubesheet from periodic SG tube inspections. This change is supported by the analysis described in section 4 of Enclosure 1. 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 permanent reporting requirements in lieu of those which had been previously established on a one-cycle basis. provides the basis for the proposed change. Enclosures 2 and 3 contain the markup of the proposed Technical Specifications and the clean-typed proposed Technical Specifications, respectively. Enclosures 4 and 5 contain proprietary and non-proprietary Westinghouse Electric Company LLC information, respectively, which supports the analysis described in Enclosure 1. On page 34 of both the proprietary Enclosure 4 and non-proprietary Enclosure 5, the title of the Appendix A cover page should be LTR-SGDA-1 1-87 instead of LTR-SGMP-11-87. On page 39 of Enclosure 5, the table number should be Table 2 instead of Table 15. The table is properly numbered in the proprietary Enclosure 4.

There is no technical impact on the report; all technical data has been redacted in this table of the non-proprietary attachment. Westinghouse letter LTR-SGMMP-11-28 Errata, Rev. 1 acknowledges these typographical errors and is provided in . Enclosure 6 contains the supporting affidavit signed by Westinghouse Electric Company LLC, the owner of the information. This affidavit sets forth the basis on which the information in Enclosure 4 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. As mentioned in Enclosure 1, all 00ý

ýflj

U. S. Nuclear Regulatory Commission NL-12-0561 Page 2 other supporting Westinghouse documents have previously been sent to the NRC by SNC or by another utility. Enclosure 7 provides the SNC specific responses to questions 12 and 13 from the Duke Energy response to request for additional information and question 15 from the Dominion response to request for additional information.

On September 24, 2009, the NRC issued VEGP Amendment Numbers 157 and 138 (Units 1 and 2, respectively) for SG Interim Alternate Repair Criteria (Reference 18 of Enclosure 1). As a condition of approval, SNC made the following regulatory commitments:

" To revise the Steam Generator Program Strategic Plan procedure to include monitoring for tube slippage as part of the steam generator tube inspection program for Unit 1 and Unit 2. Slippage monitoring will occur for each inspection of the Vogtle 1 and 2 steam generators.

• For the condition monitoring (CM) assessment, the component of operational leakage from the prior cycle from below the H* distance will be multiplied by a factor of 2.48 and added to the total accident 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 accident induced leakage and the accident induced leakage from sources other than the tubesheet expansion region will be divided by 2.48 and compared to the observed operational leakage. An administrative limit will be established to not exceed the calculated value.

The program/procedure changes needed to meet these commitments were completed in accordance with the NRC approval of amendment numbers 157 and 138 (Units 1 and 2, respectively). These changes remained in place for amendment numbers 160 and 142 (Units 1 and 2, respectively) (reference 24 of enclosure 1) and will remain in place for this License Amendment Request.

Therefore, no new NRC commitments are required.

It has been determined that this amendment application does not involve a significant hazard consideration as determined per 10 CFR 50.92. Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs to be prepared in connection with the issuance of this amendment.

SNC requests approval of the proposed license amendments by September 15, 2012 to support implementation during the Unit 1 fall refueling outage. The proposed changes would be implemented within 30 days of issuance of the amendment.

This letter contains no new NRC commitments. If you have any questions, please contact Mr. N. J. Stringfellow at (205) 992-7037.

U. S. Nuclear Regulatory Commission NL-12-0561 Page 3 Mr. M. J. Ajluni states he is Nuclear Licensing Director of Southern Nuclear Operating Company, is authorized to execute this oath on behalf of Southern Nuclear Operating Company and to the best of his knowledge and belief, the facts set forth in this letter are true.

Respectfully submitted, M. J. Ajluni Nuclear Licensing Director Sworn to and subscribedbefore me this

• day of /> 9k ,a ,2012.

Notary Public My commission expires: / //80//-*

MJA/RMJ/lac

Enclosures:

1. Basis for Proposed Change

2. Markup of Proposed Technical Specifications

3. Clean Typed Pages for Technical Specifications

4. LTR-SGMMP-1 1-28, Rev. 1, "Response to USNRC RAI for Model D5 and Model F SG Permanent H* Submittals", February 2, 2012 (Proprietary)

5. LTR-SGMMP-1 1-28, Rev. 1, "Response to USNRC RAI for Model D5 and Model F SG Permanent H* Submittals", February 2, 2012 (Non-Proprietary)

6. Westinghouse Electric Company LLC CAW-12-3404 "Application for Withholding Proprietary Information from Public Disclosure," February 22, 2012

7. Response to Request for Additional Information Questions Specific to Vogtle Electric Generating Plant

8. LTR-SGMMP-11-28 Errata, Rev. 1, "LTR-SGMMP-11-28, Revision 0 and Revision 1, P- and NP-Attachment Errata", March 20, 2012 cc: Southern Nuclear Operatingq Company Mr. S. E. Kuczynski, Chairman, President & CEO Mr. D. G. Bost, Executive Vice President & Chief Nuclear Officer Mr. T. E. Tynan, Vice President - Vogtle Mr. B. L. Ivey, Vice President - Regulatory Affairs Mr. B. J. Adams, Vice President - Fleet Operations RType: CVC7000 U. S. Nuclear Recqulatory Commission Mr. V. M. McCree, Regional Administrator Mr. P. G. Boyle, NRR Senior Project Manager-Vogtle Mr. L. M. Cain, Senior Resident Inspector - Vogtle State of Georgia Mr. J. H. Turner, Environmental Director Protection Division

Vogtle Electric Generating Plant - Units 1 and 2 License Amendment Request to Revise Technical Specification Sections 5.5.9, "Steam Generator (SG) Program" and 5.6.10, "Steam Generator Tube Inspection Report" Enclosure 1 Basis for Proposed Change to NL-12-0561 Basis for ProposedChange Enclosure 1 Basis for Proposed Change Table of Contents

1. Summary Description

2. Detailed Description

3. Background

4. Summary of Licensing Basis Analysis (H*Analysis)

5. Technical Evaluation

6. Regulatory Evaluation 6.1 Applicable Regulatory Requirements / Criteria 6.2 Significant Hazards Consideration 6.3 Conclusion

7. Environmental Considerations

8. References El-i to NL-12-0561 Basis for ProposedChange

1. Summary Description Southern Nuclear Operating Company (SNC) proposes to revise the Vogtle Electric Generating Plant Units 1 and 2 (VEGP) Technical Specifications (TS) 5.5.9, "Steam Generator (SG)

Program," to permanently exclude portions of the SG tube below the top of the SG tubesheet from periodic tube inspections. 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. In addition, this amendment proposes to revise TS 5.6.10, "Steam Generator Tube Inspection Report" to provide permanent reporting requirements in lieu of those which were previously established on a one-cycle basis. 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 4 and provides the licensing basis for this change. Table 5-1 of WCAP 17330-P (Reference 16) provides the 95/95 H* value of 15.2 inches for plants with Model F SGs which includes VEGP.

The NRC previously issued the following TS amendments revising SG tube inspection requirements:

Amendment Numbers 138 and 117 (Units 1 and 2, respectively) (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 Unit 2 Refueling Outage 11 and the subsequent operating cycle.

• Amendments Numbers 146 and 126 (Units 1 and 2, respectively) (Reference 2) to exclude the portion of the tubes below 17 inches from the top of the hot leg tubesheet from the requirement to plug for Unit 1 Refueling Outage 13 and Unit 2 Refueling Outage 12 and the subsequent operating cycles.

" Amendment Numbers 150 and 130 (Units 1 and 2, respectively) (Reference 3) which approved an interim alternate repair criteria for Unit 1 Refueling Outage 14 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.

" Amendment Numbers 152 and 133 (Units 1 and 2, respectively) (Reference 4) applied the criteria of Amendments 150 and 130 to Unit 2 Refueling Outage 12 and the subsequent operating cycle. Amendment Numbers 150 and 152 (Unit 1) and 130 and 133 (Unit 2) are only applicable for Refueling Outages 14 (Unit 1) and 13 (Unit 2) and the subsequent operating cycles.

" Amendment Numbers 157 and 138 (Units 1 and 2, respectively) (Reference 18) revised TS 5.5.9, "Steam Generator (SG) Program," to exclude portions of the tubes within the tubesheet from periodic SG inspections (establish alternate repair criteria). In addition, these amendments revised TS 5.6.10, "Steam Generator Tube Inspection Report," to El-2 to NL-12-0561 Basis for ProposedChange remove reference to previous interim alternate repair criteria and provide reporting requirements specific to Refueling Outage 15 and the subsequent operating cycle and for Unit 2 during Refueling Outage 14 and the subsequent operating cycle.

Amendment Numbers 160 and 142 (Units 1 and 2, respectively) (Reference 24) revised TS 5.5.9, "Steam Generator (SG) Program," to exclude portions of the tube below the top of the SG tubesheet from periodic SG inspections for Unit 1 during Refueling Outage 16 and the subsequent operating cycle and for Unit 2 during Refueling Outage 15 and the subsequent operating cycle. In addition, this amendment revised TS 5.6.10, "Steam Generator Tube Inspection Report," to remove reference to previous interim alternate repair criteria and provide reporting requirements specific to Refueling Outages 16 (Unit

1) and 15 (Unit 2) and the subsequent operating cycles.

Approval of this amendment application is requested by September 15, 2012 to support Unit 1 Refueling Outage 17 (fall 2012), since the existing one-cycle amendment expires at the end of the current operating cycle.

2. Detailed Description Proposed Changes to Current TS TS 5.5.9.c. 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:

For Unit 1 during Refueling Outage 16 and the subsequent operating cycle and for Unit 2 during Refueling Outage 15 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, as noted in italic type:

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:

E1-3 to NL-12-0561 Basis for ProposedChange Gyele and for &Wit 2 "-~n Rfuechg1 Outag 1and the subs ,on

.peratiog.y..e,-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.

TS 5.5.9.d 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 Unit 1 during Refueling Outage 16 and the subsequent operating cycle and for Unit 2 during Refueling Outage 15 and the subsequent operating cycle, portions of the tube below 15.2 inches below the top of the tubesheet are 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 portions of the SG tube not excluded above, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less).

If definitive information, such as from examination of a pulled tube, diagnostic nondestructive 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.

E1-4 to NL-12-0561 Basis for ProposedChange This section would be revised as follows, as noted in italic type:

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. Fie- Unit , dur.ng R, fueR,,g Outage 16 and th subsequent e.pratingc...e and for Unit 2 d-'ring Rfueing Outage 15 and the subsequent e.p.ating /o,"', Portions of the tube below 15.2 inches below the top of the tubesheet are 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 portions of the SG tube not excluded above, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less).

If definitive information, such as from examination of a pulled tube, diagnostic nondestructive 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 5.6.10 h., 5.6.10 i., and 5.6.10 j. currently states:

h. For Unit 1 during Refueling Outage 16 and the subsequent operating cycle and for Unit 2 during Refueling Outage 15 and the subsequent E1-5 to NL-12-0561 Basis for ProposedChange 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; and For Unit 1 during Refueling Outage 16 and the subsequent operating cycle and for Unit 2 during Refueling Outage 15 and 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 2.48 times the maximum operational primary to secondary leakage rate, the report should describe how it was determined.

j. For Unit 1 during Refueling Outage 16 and the subsequent operating cycle and for Unit 2 during Refueling Outage 15 and the subsequent operating cycle, the results of monitoring for tube axial displacement (slippage). If slippage is discovered, the implications of the discovery and corrective action shall be provided.

TS 5.6.10 h., 5.6.10 i and 5.6.10 j. would be revised as follows, as noted in italic type:

h. For- L4 4 durlng R lfu;lng Outage 16 and, the ub6oeount eeratng cycic and for-Unit 2 dur"n Refualing Outage 15 and the subcoguenl epeA9&?g-Gyole- 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; and

i. Foer Unit 1 during Refueling Outage 16 and the subsequent operatn Gycle and for Unit 2 durig Refucling Outage 15 and tho 6ubsequoni epe.AR... G.y.*e-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 2.48 times the maximum operational primary to secondary leakage rate, the report should describe how it was determined.

j. For Unit 1 during Rofucling Outage 16 and the subsequent÷p..a.in" 6,w1, and for Unit 2 during R-fuling Outage 15 and the subcquen,

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

El-6 to NL-12-0561 Basis for Proposed Change

3. Background VEGP consists of two four-loop Westinghouse designed plants with Model F SGs, having 5626 tubes in each SG. A total (for all four SGs per unit) of 148 tubes are currently plugged on Unit 1 and 46 on Unit 2. The design of the SG includes Alloy 600 thermally treated tubing, full depth hydraulically expanded tubesheet joints, and stainless steel tube support plates with broached hole quatrefoil.

The SG inspection scope is governed by:

• TS 5.5.9, "Steam Generator (SG) Program,"

• Nuclear Energy Institute (NEI) 97-06, "Steam Generator Program Guidelines,"

(Reference 6),

• EPRI 1003138, "Pressurized Water Reactor Steam Generator Examination Guidelines,"

(Reference 7),

• EPRI 1012987, "Steam Generator Integrity Assessment Guidelines," (Reference 8),

" NMP-ES-004, "Steam Generator Program," 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 known and potential specific degradation mechanisms applicable to VEGP. The inspection techniques, essential variables, and equipment are qualified to Appendix H, "Performance Demonstration for Eddy Current Examination" of the EPRI SG examination guidelines.

Catawba Nuclear Station, Unit 2, reported indication of cracking following nondestructive eddy current examination of the SG 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 9), 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.

SNC policies and programs require the use of applicable industry operating experience in the operation and maintenance of VEGP. 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 SGs) 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 VEGP Westinghouse Model F SGs were fabricated with Alloy 600 thermally treated tubes similar to the Catawba Unit 2 Westinghouse Model D5 SGs, a potential exists for VEGP 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 SG tubes present three distinct issues with regard to the SG tubes at VEGP:

E1-7 to NL-12-0561 Basis for ProposedChange

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 SG 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 VEGP SGs. 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, VEGP revised the SG inspection plan to include sampling of bulges and overexpansions within the tubesheet region on the hot leg side during Refueling Outage 1 R14 and 2R13 for Units 1 and 2, respectively. The sample is based on the guidance contained in EPRI 1003138, "Pressurized Water Reactor Steam Generator Examination Guidelines," Revision 7, and TS 5.5.9, 'Steam Generator (SG) Program." The inspection plan was expanded according to EPRI SG examination guidelines 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 1 R14, 1 R15, and 1 R16 or in Refueling Outage 2R13, 2R1i4, and 2R15 for Units 1 and 2, respectively.

While flaws in bulges/overexpansions have been found at Unit 1, a separate inspection program for these indications has been implemented at both VEGP units. This inspection program is in accordance with VEGP's current technical specifications and industry guidance.

Based on these inspections, a limited number of flaws existed in the tube sheets of VEGP SGs.

The flaws that have been found are associated with residual stress conditions at either the tube ends or bulges/overexpansions within the tube sheet. No indications of a 360 degree sever have been detected in any SG at VEGP. Consequently, the level of degradation in the VEGP SGs is very limited compared to the assumption of "all tubes severed" that was utilized in the development of the permanent H*. Consequently, structural integrity will be assured for the operating period between inspections allowed by TS 5.5.9, "Steam Generator (SG) Program."

As a result of these potential issues and the possibility of unnecessarily plugging tubes in the VEGP SGs, SNC is proposing changes to TS 5.5.9 to limit the SG tube inspection and repair (plugging) to the safety significant portion of the tubes.

4. Summary of Licensing Basis Analysis (H* Analysis)

On May 19, 2009, Westinghouse WCAP-1 7071 -P, Revision 0, "H*: Alternate Repair Criteria for the Tubesheet Expansion Region in Steam Generators with Hydraulically Expanded Tubes (Model F)," (Reference 5) was submitted as enclosure 5 of SNC request to change 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 SG tubes (Reference 19).

On July 10, 2009, SNC received a request for additional information (RAI) letter, which contained twenty-four questions (Reference 22). As a result of a teleconference with NRC staff held on July 30, 2009, SNC received a second RAI letter on August 5, 2009 (Reference 23).

The August 5, 2009 letter contained three questions related to questions 4, 20 and 24 from RAI E1-8 to NL-12-0561 Basis for ProposedChange letter received on July 10, 2009. The August 5, 2009 letter also contained one additional question.

On August 28, 2009, SNC provided the following documents as enclosures to References 29 and 30 in response to questions 1 through 24 of the July 10, 2009 letter and questions 1 through 4 of the August 5, 2009 letter:

• 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 11), 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 13).

On August 28, 2009, SNC submitted Westinghouse letter LTR-SGMP-09-104 P-Attachment "White Paper on Probabilistic Assessment of H" dated August 13, 2009, (Reference 12) as supplemental information.

On September 11, 2009, SNC submitted a request to revise the May 19, 2009 amendment request to be an interim change applicable to Unit 1 during refueling outage 15 and the subsequent operating cycle and to Unit 2 during refueling outage 14 and the subsequent operating cycle (Reference 20). 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 July 10, 2009 letter and RAI question 1 of the August 5, 2009 letter) had not been resolved. The September 11, 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 November 23, 2009, the NRC provided a letter (Reference 21) documenting the currently identified and unresolved issues relating to tubesheet bore eccentricity. This letter contained fourteen 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 16) provides a discussion of the action plan to respond to the fourteen unresolved questions.

The following documents have been prepared by Westinghouse to provide final resolution of the remaining questions identified in the November 23, 2009 NRC letter in support of the permanent H* amendment for Vogtle:

• WCAP-17330-P, Revision 1, "H*: Resolution of NRC Technical Issue Regarding Tubesheet Bore Eccentricity (Model F/D5)," June 2011 (Reference 16),

" 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, 2010 (Reference 14),

" LTR-SGMP-10-33 P-Attachment, "H* Response to NRC Questions Regarding Tubesheet Bore Eccentricity," September 13, 2010 (Reference 15).

E1-9 to NL-12-0561 Basis for ProposedChange WCAP-1 7330-P, Revision 1, "H*: Resolution of NRC Technical Issue Regarding Tubesheet Bore Eccentricity (Model F/D5)," June 2011 (Reference 16) makes reference to Revision 2 of WCAP-1 7071-P and Revision 1 of LTR-SGMP-09-1 00 P-Attachment. As described above, SNC has previously submitted Revision 0 of these documents. These revisions (Revisions 1 and 2 of WCAP-17071-P and 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-09-1 00 P-Attachment, Revision 0, remains valid and provides part of the licensing basis for the requested amendment change.

As a condition for approving VEGP Amendments 157 and 138 (Units 1 and 2, respectively)

(Reference 18) for interim alternate repair criterion, the NRC required a commitment to measure the location of the bottom of the expansion transition (BET) relative to the top of the tubesheet (TTS) and report any significant deviations from the constant 0.3-inch value already included in the calculated value(s) of H*. LTR-SGMP-09-1 11, Revision 1, "Acceptable Value of the Location of the Bottom of the Expansion Transition (BET) for Implementation of H*," September 1, 2010, was prepared to support plant determinations of BET measurements and their significant deviation assessment and was submitted to the NRC per SNC letter NL-10-2104 (Reference 34).

On June 30, 2011, Duke Energy submitted a license amendment request (Reference 31) 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, "H*: Resolution of NRC Technical Issue Regarding Tubesheet Bore Eccentricity (Model F/Model D5)." A supplement (Reference 39) 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 RAI (Reference 32) was transmitted electronically to Duke Energy. Duke Energy responded to the RAI on January 12, 2012 (Reference 35).

Subsequent to the Duke Energy license amendment request, Virginia Electric and Power Company (Dominion) submitted a license amendment request (Reference 36) 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 RAI (Reference 37). Dominion responded to the RAI on February 14, 2012 (Reference 38).

Westinghouse Electric Company LLC, LTR-SGMMP-1 1-28 Rev.1 P-Attachment (Enclosure 4),

"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 RAI to include similar responses applicable to Model F steam generators. Additionally, this letter addresses the Dominion request for additional information question 14 for the Model F steam generators. On page 34 of both the proprietary Enclosure 4 and non-proprietary Enclosure 5, 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 non-proprietary LTR-SGMMP-1 1-28 Rev. 1 NP-Attachment (Enclosure 5), the table number should be Table 2 instead of Table 15. The table is properly numbered in the proprietary attachment.

There is no technical impact on the report; all technical data has been redacted in this table of the non-proprietary attachment. Westinghouse letter LTR-SGMMP-1 1-28 Errata, Rev. 1 (Reference 40) acknowledges these typographical errors and is provided in Enclosure 8.

Enclosure 7 provides VEGP specific responses to questions 12 and 13 from the Duke Energy request for additional information and question 15 from the Dominion request for additional EI-10

Enclosure 1 to NL-12-0561 Basis for ProposedChange information. The NRC questions to Duke Energy and Dominion are identified in italics in Enclosure 7.

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

Document Rev. Title Ref. Ref. That Number Number Number Submitted Document to NRC WCAP-17071-P 0 H*: Alternate Repair Criteria for the Tubesheet 5 19 Expansion Region in Steam Generators with Hydraulically Expanded Tubes (Model F)

LTR-SGMP 0 Response to NRC Request for Additional 11 29 100 P-Attachment Information on H*; Model F and Model D5 Steam Generators LTR -SGMP 0 Response to NRC Request for Additional 13 30 -

109 P-Attachment Information on H*; RAI #4; Model F and Model D5 Steam Generators WCAP-17330-P 1 H*: Resolution of NRC Technical Issue Regarding 16 31 Tubesheet Bore Eccentricity (Model F/D5)

LTR-SGMP-10-78 0 Effects of Tubesheet Bore Eccentricity and Dilation 14 34 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 15 34 P-Attachment Tubesheet Bore Eccentricity LTR-SGMMP 1 Response to USNRC Request for Additional 33 N/A for SNC 28 P-Attachment Information Regarding the Catawba License Amendment Request for Permanent Application of the Alternate Repair Criterion, H*

Westinghouse letter LTR-SGMP-1 1-58 (Reference 28) corrected a transposition of numbers in Table 3-30 of WCAP-1 7330-P, Revision 1 in the Model D5 section of the table. The data for the Model F SGs is unaffected.

In addition, the following correspondence is also applicable to the VEGP permanent alternate repair criteria request:

A March 28, 2011 letter from the NRC to SNC (Reference 25) documented the summary of a February 16, 2011 public meeting regarding SG 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-1 7330-P, Revision 1 (Reference 16).

El-11 to NL-12-0561 Basis for Proposed Change Section 1.3 of Reference 16 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 SNC (Reference 26) included a pre-submittal review request for additional information. The response to the NRC pre-submittal review request is provided in SNC letter NL-1 1-1178 (Reference 27).

5. Technical Evaluation To preclude unnecessarily plugging tubes in the VEGP SGs, 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 described in section 4. 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 VEGP Model F SG 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.

Primary-to-secondary leakage from tube degradation in the tubesheet area 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 100, General Design Criteria (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* justification.

The constraint that is provided by the tubesheet precludes tube burst from 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," (Reference 10) 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 is 1.0 gpm. The TS operational leak rate limit is 150 gpd (0.1 gpm) through any one SG. Consequently, there is significant margin between accident leakage and allowable operational leakage. The SLB/FLB leak rate ratio is 2.48, resulting in significant margin between the conservatively estimated accident leakage and the allowable accident leakage (1.0 gpm).

EI-12 to NL-12-0561 Basis for Proposed Change 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 SG are contained in Section 6 of Reference 5.

The leak rate ratio (accident induced leak rate to operational leak rate) is directly proportional to the change in differential pressure and inversely proportional to the dynamic viscosity. Since dynamic viscosity decreases with an increase in temperature, an increase in temperature results in an increase in leak rate. However, for both the postulated SLB and FLB events, a plant cool down event would occur and the subsequent temperatures in the reactor coolant system (RCS) would not be expected to exceed the temperatures at plant no load conditions.

Thus, an increase in leakage would not be expected to occur as a result of the viscosity change.

The increase in leakage would only be a function of the increase in primary to secondary pressure differential. The resulting leak rate ratio for the SLB and FLB events is 2.48.

The other design basis accidents, such as the postulated locked rotor event and the control rod ejection event, are conservatively modeled using design specification transients which result in increased temperatures in the SG 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, as well as due to the increasing differential pressure, for the duration of time that there is a rise in RCS temperature. For transients other than a SLB and FLB, the length of time that a plant with Model F SGs 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 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 leakage factor of 1.74 (Revised Table 9-7, Reference 11) for VEGP is adjusted downward to a factor of 0.87.

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 2.62 to 0.44.

The plant transient response following a full power double-ended main feedwater line rupture corresponding to "best estimate" initial conditions and operating characteristics indicates that the transient for a Model F SG exhibits a cool-down characteristic instead of a heat-up transient as generally presented in SG design transients and in the UFSAR Chapter 15 safety analysis.

The use of either the component design specification transient or the Chapter 15 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 reactor coolant system 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.

E1-13 to NL-12-0561 Basis for Proposed Change

• For the Model F SG, the FLB transient curve (Figure 9-5, Reference 5) represents a double-ended rupture of the main feedwater tine concurrent with both loss of offsite power (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 be less severe. This would result in much earlier reactor trip and greatly increase the SG liquid mass available to provide cooling to the RCS.

A SLB event would have similarities to a FLB except that the break flow path 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 RCS 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 SG and control of temperature by the AFW system. Steam pressure control would be established by either the SG safety valves or control system (atmospheric relief valves). For any of the steam pressure control operations, the maximum RCS 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.

The leakage factor of 2.48 for VEGP, for a postulated SLB/FLB, has been calculated as shown in Revised Table 9-7 of Reference 11. Specifically, for the condition monitoring assessment, the component of leakage from the prior cycle from below the H* distance will be multiplied by a factor of 2.48 and added to the total leakage from any other source and compared to the allowable accident induced leakage limit. For the operational assessment, 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 2.48 and compared to the observed operational leakage.

Reference 5 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 SG operating conditions, including the most limiting accident conditions. The evaluation in Reference 5 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, 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 E1-14 to NL-12-0561 Basis for ProposedChange bundle. As a condition of approval of TS Amendment Numbers 157 and 138 (Units 1 and 2, respectively) (Reference 18), SNC committed to monitor for tube slippage as part of the SG tube inspection program. This requirement will remain in place to support this License Amendment Request.

As a condition for approving VEGP Amendments 157 and 138 (Units 1 and 2, respectively)

(Reference 18) for interim alternate repair criterion, the NRC staff requested that SNC perform a validation of the tube expansion from the top of tubesheet to the bottom of the expansion transition (BET) to determine if there are any significant deviations that would invalidate assumptions in WCAP-17071-P (Reference 5). SNC has completed the validation of the tube expansion from the top of tubesheet to the BET for both VEGP units. Based on data review and LTR-SGMP-09-111 P-Attachment, Revision 1 (Reference 17), SNC did not identify any significant deviations from the top of tubesheet to the BET on either VEGP unit.

6. Regulatory Evaluation 6.1 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 ensure 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 the Code of FederalRegulations(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, XI, and XVI of Appendix B and include control of special processes, inspection, testing, and corrective action.

10 CFR 100 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 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 El-15 to NL-12-0561 Basis for Proposed Change performance criteria. Meeting the performance criteria of NEI 97-06, Revision 3, 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 3, 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, cool down and all anticipated transients included in the design specification) and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design 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.

• 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 1.0 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 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 Enclosure 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 VEGP inspection program provides a high level of confidence that the structural and leakage criteria are maintained during normal operating and accident conditions.

6.2 Significant Hazards Consideration This amendment application proposes to revise Technical Specification (TS) 5.5.9, "Steam Generator (SG) Program," to permanently exclude portions of the tubes within the tubesheet from periodic steam generator 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 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.

E1-16 to NL-12-0561 Basis for Proposed Change The proposed change defines the safety significant portion of the tube that must be inspected and repaired. Ajustification 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 Guidelines," (Reference 6) performance criteria.

Southern Nuclear Operating Company (SNC) has evaluated whether 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," 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 and the steam generator inspection reporting criteria does not have a detrimental impact on the integrity of any plant structure, system, or component that initiates an analyzed event. The proposed change will not alter 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 and the feedline break (FLB) / steam line break (SLB) postulated accidents.

Tube rupture in tubes with cracks within the tubesheet is precluded by the constraint provided by the tube-to-tubesheet joint. This constraint results from the hydraulic expansion process, thermal expansion mismatch between the tube and tubesheet, and from the differential pressure between the primary and secondary side. Based on this design, the structural margins against burst, as discussed in Regulatory Guide (RG) 1.121, "Bases for Plugging Degraded PWR Steam Generator Tubes," (Reference 10) 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 integrity of the steam generator tubes and does not affect other systems, structures, components, or operational features. 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 primary water stress corrosion cracking below the proposed limited inspection depth is limited by both the tube-to-tubesheet crevice and the limited crack opening permitted by the tubesheet constraint.

Consequently, negligible normal operating leakage is expected from cracks within the tubesheet region. The consequences of an SGTR event are affected by the primary-to-secondary leakage flow during the event. However, primary-to-secondary leakage flow through a postulated broken tube is not affected by the proposed changes since the tubesheet enhances the tube integrity in the region of the hydraulic expansion by E1-17 to NL-12-0561 Basis for ProposedChange precluding tube deformation beyond its initial hydraulically expanded outside diameter.

Therefore, the proposed changes do not result in a significant increase in the consequences of a SGTR.

The consequences of a SLB/FLB are also not significantly affected by the proposed changes. During a SLB/FLB accident, the reduction in pressure above the tubesheet on the shell side of the steam generator creates an axially uniformly distributed load on the tubesheet due to the reactor coolant system pressure on the underside of the tubesheet.

The resulting bending action constrains the tubes in the tubesheet thereby restricting primary-to-secondary leakage below the midplane.

Primary-to-secondary leakage from tube degradation in the tubesheet area during the limiting accident (i.e., a 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 2.48 for Vogtle Electric Generating Plant Units 1 and 2 (VEGP), for a postulated FLB, has been calculated as shown in Revised Table 9-7 of Reference 11.

Specifically, for the condition monitoring assessment, the component of leakage from the prior cycle from below the H* distance will be multiplied by a factor of 2.48 and added to the total leakage from any other source and compared to the allowable accident induced leakage limit. For the operational assessment, 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 2.48 and compared to the observed operational leakage.

Feedline break 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 postulated accident conditions (including locked rotor) has been shown to remain within the accident analysis assumptions for all axial and or circumferentially orientated cracks occurring 15.2 inches below the top of the tubesheet. The accident induced leak rate limit is 1.0 gpm. The TS operational leak rate is 150 gpd (0.1 gpm) through any one steam generator. Consequently, there is significant margin between accident leakage and allowable operational leakage. The FLB leak rate ratio is 2.48 resulting in significant margin between the conservatively estimated accident 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 that alters the steam generator inspection criteria and the steam generator inspection reporting criteria 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 all safety functions will continue to perform as previously assumed in accident analyses.

EI-18 to NL-12-0561 Basis for Proposed Change 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 that alters the steam generator inspection criteria and the steam generator inspection reporting criteria maintains the required structural margins of the steam generator tubes for both normal and accident conditions. NEI 97-06, Revision 3, "Steam Generator Program Guidelines" (Reference 6) 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 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 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 4 of this enclosure, 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.

6,3 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.

EI-19 to NL-12-0561 Basis for ProposedChange

7. Environmental Considerations SNC has evaluated the proposed amendment for environmental considerations. The review has resulted in the determination 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, or 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 statement or environmental assessment need be prepared in connection with the proposed amendment.

El-20

,to NL-12-0561 Basis for ProposedChange

8. References

1. NRC letter from C. Gratton, USNRC, to D. E. Grissette, SNC, "Vogtle Electric Generating Plant Units 1 and 2, RE: Issuance of Amendments Regarding the Steam Generator Tube Surveillance Program (TAC Nos. MC8078 and MC8079)," September 21, 2005. (ADAMS Accession No. ML052630011)

2. NRC letter from C. Gratton, USNRC, to D. E. Grissette, SNC, "Vogtle Electric Generating Plant Units 1 and 2, "Issuance of Amendments Regarding the Steam Generator Tube Surveillance Program (TAC Nos. MD2642 and MD2643)," September 12, 2006. (ADAMS Accession No. ML062260266)

3. NRC letter from S. P. Lingam, USNRC, to T. E. Tynan, SNC, "Vogtle Electric Generating Plant Units 1 and 2, Issuance of Amendments Regarding Changes to Technical Specification (TS) Sections TS 5.5.9, "Steam Generator (SG) Program" and TS 5.6.10, "Steam Generator Tube Inspection Report," (TAC Nos. MD7450 and MD7451)." April 9, 2008. (ADAMS Accession No. ML080950247)

4. NRC letter from R. E. Martin, USNRC, to T. E. Tynan, SNC, "Vogtle Electric Generating Plant Units 1 and 2, Issuance of Amendments Regarding Steam Generator Tube Inspection Program (TAC Nos. MD9148 and MD9149)," September 16, 2008. (ADAMS Accession No. ML082530044)

5. Westinghouse Electric Company LLC, WCAP-17071-P, "H*: Alternate Repair Criteria for the Tubesheet Expansion Region in Steam Generators with Hydraulically Expanded Tubes (Model F)." (ADAMS Accession Nos. ML091470699 (Introduction through Chapter 5) and ML091470700 (Chapter 6 to end))

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

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

8. EPRI 1012987; "Steam Generator Integrity Assessment Guidelines."

9. NRC Information Notice (IN) 2005-09, "Indications in Thermally Treated Alloy 600 Steam Generator Tubes and Tube-to-Tubesheet Welds."

10. Regulatory Guide 1.121, "Bases for Plugging Degraded PWR Steam Generator Tubes,"

dated August 1976, (ADAMS Accession No. ML003739366).

11. LTR-SGMP-09-1 00, "LTR-SGMP-09-1 00 P-Attachment, "Response to NRC Request for Additional Information on H*; Model F and Model D5 Steam Generators," August 12, 2009. (ADAMS Accession No. ML0924501102)

12. LTR-SGMP-09-104 P-Attachment, 'White Paper on Probabilistic Assessment of H*,"

August 13, 2009. (ADAMS Accession No. ML092450030)

El-21 to NL-12-0561 Basis for Proposed Change

13. 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. ML092450333)

14. 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 7,2010.

15. LTR-SGMP-10-33 P-Attachment, "H* Response to NRC Questions Regarding Tubesheet Bore Eccentricity," September 13, 2010.

16. WCAP-1 7330-P, Rev. 1, "H*: Resolution of NRC Technical Issue Regarding Tubesheet Bore Eccentricity (Model F/D5)," June 2011.

17. LTR-SGMP-09-1 11, Rev. 1, "Acceptable Value of the Location of the Bottom of the Expansion Transition (BET) for Implementation of H*," September 1, 2010.

18. NRC letter from D. Wright, USNRC, to M. J. Ajluni, SNC, "Vogtle Electric Generating Plant, Units 1 And 2, Issuance Of Amendments Regarding Technical Specification (TS)

Section 5.5.9, "Steam Generator Program," And TS 5.6.10, "Steam Generator Tube Inspection Report," For Interim Alternate Repair Criteria (TAC Nos. ME1339 And ME1340)," September 24, 2009. (ADAMS Accession No. ML092170782)

19. SNC letter NL-09-0547, "Vogtle Electric Generating Plant License Amendment Request to Revise Technical Specification (TS) Sections 5.5.9, "Steam Generator (SG) Program" and TS 5.6.10, "Steam Generator Tube Inspection Report" for Permanent Alternate Repair Criteria," May 19, 2009. (ADAMS Accession No. ML091470701).

20. SNC letter NL-09-1411, "Vogtle Electric Generating Plant License Amendment Request to Revise Technical Specification (TS) Sections 5.5.9, "Steam Generator (SG) Program" and TS 5.6.10, "Steam Generator Tube Inspection Report" for Interim Alternate Repair Criteria," September 11, 2009. (ADAMS Accession No. ML092540511)

21. NRC Letter from D. Wright, USNRC, to M. J. Ajluni, SNC, "Vogtle Electric Generating Plant, Units 1 and 2 - Transmittal of Unresolved Issues Regarding Permanent Alternate Repair Criteria for Steam Generators (TAC Nos. ME 1339 and ME 1340)," November 23, 2009. (ADAMS Accession No. ML093030490)

22. NRC Letter from D. Wright, USNRC, to M. J. Ajluni, SNC, "Vogtle Electric Generating Plant, Units 1 And 2 -- Request for Additional Information Regarding Steam Generator Program (TAC Nos. ME 1339 and ME 1340)," July 10, 2009. (ADAMS Accession No. ML091880384)

23. NRC Letter from D. Wright, USNRC, to M. J. Ajluni, SNC, "Vogtle Electric Generating Plant, Units 1 And 2 -- Request for Additional Information Regarding Steam Generator Program (TAC Nos. ME 1339 and ME 1340)," August 5, 2009. (ADAMS Accession No. ML092150057)

E1-22 to NL-12-0561 Basis for ProposedChange

24. NRC Letter from P. G. Boyle, USNRC, to M. J. Ajluni, SNC, "Vogtle Electric Generating Plant, Units 1 And 2, Issuance Of Amendments Regarding Steam Generator Tube Inspection Alternate Inspection Criteria (TAC Nos. ME5067 and ME5068)," March 14, 2011. (ADAMS Accession No. ML110660264).

25. NRC Letter, "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 from P. G. Boyle, USNRC, to M. J. Ajluni, SNC, "Vogtle Electric Generating Plant, Units 1 And 2 - Presubmittal Consideration of Steam Generator Alternative Repair Criteria Requirements - Request for Additional Information (TAC Nos. ME5417 and ME5418)," May 26, 2011. (ADAMS Accession No. ML11140A099)

27. SNC letter NL-1 1-1178, "Vogtle Electric Generating Plant Response to Presubmittal Consideration of Steam Generator Alternative Repair Criteria Requirements Request for Additional Information," June 20, 2011, (ADAMS Accession No. ML111721903).

28. LTR-SGMP-1 1-58, 'WCAP-1 7330-P, Revision 1 Erratum," July 6, 2011.

29. SNC letter NL-09-1265, "Vogtle Electric Generating Plant Response to Request for Additional Information Related to License Amendment Request to Revise Technical Specification (TS) Sections 5.5.9, "Steam Generator (SG) Program" and TS 5.6.10, "Steam Generator Tube Inspection Report" for Permanent Alternate Repair Criteria,"

August 28, 2009. (ADAMS Accession No. ML092450101 (Non-Proprietary) and ML092450102 (Proprietary))

30. SNC letter NL-09-1375, 'Vogtle Electric Generating Plant Response to Request for Additional Information Related to License Amendment Request to Revise Technical Specification (TS) Sections 5.5.9, "Steam Generator (SG) Program" and TS 5.6.10, "Steam Generator Tube Inspection Report" for Permanent Alternate Repair Criteria,"

August 28, 2009. (ADAMS Accession No. ML092450333)

31. Duke Energy Letter to NRC, "Duke Energy Carolinas, LLC (Duke Energy) Catawba Nuclear Station, Units 1 and 2 Docket Numbers 50-413 and 50-414 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)

32. 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 E1-23 to NL-12-0561 Basis for ProposedChange

33. LTR-SGMMP-1 1-28 P-Attachment Revision 1, "Response to USNRC Request for Additional Information Regarding the Catawba License Amendment Request for Permanent Application of the Alternate Repair Criterion, H*," February 2, 2012

34. SNC letter NL-10-2104, 'Vogtle Electric Generating Plant - Units 1 and 2 License Amendment Request to Revise Technical Specification (TS) Sections 5.5.9, "Steam Generator (SG) Program" and TS 5.6.10, "Steam Generator Tube Inspection Report" for Temporary Alternate Repair Criteria," November 23, 2010. (ADAMS Accession No. ML103300241)

35. Duke Energy Letter to NRC, "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)

36. 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)

37. 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. ML12006AO01)

38. 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.

39. 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)

40. LTR-SGMMP-11-28, Revision 1 NP Attachment Errata, March 13, 2012.

E1-24

Vogtle Electric Generating Plant - Units 1 and 2 License Amendment Request to Revise Technical Specification Sections 5.5.9, "Steam Generator (SG) Program" and 5.6.10, "Steam Generator Tube Inspection Report" Enclosure 2 Markup of Proposed Technical Specifications

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

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

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:

cyclo and for Unit 2 during Refueling Outago 16 and tho subsequent

.pe.atig .Gyl.e,_-Iubes 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.

(continued)

Vogtle Units 1 and 2 5.5-8 Amendment No. 460- (Unit 1)

Amendment No. 442 (Unit 2)

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Progqram (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. For Unit 1 during Rcfuoling Outago 16 and, the -c*,

ubsoqucnt ating cycle and for Unit 2 du*rng Ref ue'in*g Outag 15 aRd the 'ubsequontoperating G -,-, Portions of the tube below 15.2 inches below the top of the tubesheet are excluded from this requirement.

(continued)

Vogtle Units 1 and 2 5.5-9 Amendment No. 460 (Unit 1)

Amendment No. 4422 (Unit 2)

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.9 Deleted.

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. For Unit 1 duFrig Ref ueling Outago 16 and the subsequcnt operating cycle and for Unit 2 during Rcf*Ucing Outago 15 and th, subc*quent perating rcy'cle, Ihe 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; and

i. FoAr UitA 1 during Refueling Outage 16 and the 6ybsequent epcrating cycle and for Unit 2 dU*i*g Rrfucling Outagc 15 and tho subsequeRt operating cY..,"Ihe 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 2.48 times the maximum operational primary to secondary leakage rate, the report should describe how it was determined.

j. For Un~it 1 during Refueling Outage 16 and the subsequont operating cyclc and for Unit 2 d*uing Ref eling Outage 15 and the subsequent o.perating cycle, The results of monitoring for tube axial displacement (slippage). If slippage is discovered, the implications of the discovery and corrective action shall be provided.

Vogtle Units 1 and 2 5.6-6 Amendment No. 1-60 (Unit 1)

Amendment No. 442- (Unit 2)

Vogtle Electric Generating Plant - Units 1 and 2 License Amendment Request to Revise Technical Specification Sections 5.5.9, "Steam Generator (SG) Program" and 5.6.10, "Steam Generator Tube Inspection Report" Enclosure 3 Clean Typed Pages for Technical Specifications

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

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

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:

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.

(continued)

Vogtle Units 1 and 2 5.5-8 Amendment No. (Unit 1)

Amendment No. (Unit 2)

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. Portions of the tube below 15.2 inches below the top of the tubesheet are excluded from this requirement.

(continued)

Vogtle Units 1 and 2 5.5-9 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.9 Deleted.

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; and

i. 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 2.48 times the maximum operational primary to secondary leakage rate, the report should describe how it was determined.

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

Vogtle Units 1 and 2 5.6-6 Amendment No. (Unit 1)

Amendment No. (Unit 2)

Vogtle Electric Generating Plant - Units 1 and 2 License Amendment Request to Revise Technical Specification Sections 5.5.9, "Steam Generator (SG) Program" and 5.6.10, "Steam Generator Tube Inspection Report" Enclosure 5 LTR-SGMMP-11-28, Rev. 1, "Response to USNRC RAI for Model D5 and Model F SG Permanent H*Submittals", February 2, 2012 (Non-Proprietary)

Westinghouse Non-Proprietary Class 3 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.

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

© 2012 Westinghouse Electric Company LLC All Rights Reserved I

LTR-SGMMP- 11-28 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-17330-P, Revision 1. WCAP-17330-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-17345-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-17330-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-1 1-28 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-I 7330-P, Revision I - 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 orderresults? Why are the minimum H* values in Figure 3-36 of Revision I substantiallydifferent from those in Figure3-32 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 3

LTR-SGMMP- 11-28 Rev. 1 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 (x, the square root of the sum of the squares of the individual tube and tubesheet (TS) CTE components. The RSS combination of tube and tubesheet variables negates the sign of the negative variation of both the tube and TS CTE and artificially inflates the value of at, resulting in the upward data scatter shown on Figure 3-32 in WCAP-17330-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-1 7330-P, Revision 0 - Provide copy of the "responsesurface" (i.e., H*

relationshipto 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 discussed in Revision I of WCAP-17330-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-I 1-28 Rev. 1 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-1 1-28 Rev. 1 NP-Attachment a,c,e 40 41 42 43 44 45 _____ ______ _____

Question 3:

WCAP- I 7330-P, Revision I - 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 ordering above 90% in the non-conservative direction?

This question was modified in Reference 4 for the Model 51 F SG as noted below.

Because the limiting operating 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-I 7345-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 variationof 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,c,e 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-I 1-28 Rev. 1 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 C2 model ensures a conservative H* value. For instance, in order to obtain, the 95/50 full bundle H* value, the 9 5 0 0 th value in the H* rank order is chosen. In the 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.

7

LTR-SGMMP-I 1-28 Rev. 1 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 2 tubesheet and tube CTEs that represent a specific rank order of H* values for input to the C 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.

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 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 C 2 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-1 1-28 Rev. 1 NP-Attachment Table 3-1 Reduced Response Surface; Model D5, [ Ia,c,e inches Radius TS CTE T CTE H*+BET

____________~ I ~ 1a Radius)

]""'____I a,c,e 3

6 7

9 10 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. 1 NP-Attachment 39 a,c,e 40 41 42 43 44 45 [_____ _____ ________

10

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

WCAP-17330-P, Revision 1, Table 3 Provide a similar table applicableto 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-1 7330-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 H 1Tube Tubesheet Alpha(l)

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: L

1. Defined as SQRT((Tube CTE)A2 + (Tubesheet CTE)A2) 11

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 similartable 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-1 7330-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-1 7330-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,c,e inch radius for rank orders from 9533 through 9539.

Table 5-1: Model F NOP Results at [ ]ac~e inches Variation Input I 2

MC T CTE TS CTE C H*

1. na m(T in.

9888 [ ]a,c,e [ ]a,c,e [ ]a,c,e 9892 [ ]a,c,e [ ]a,c,e ]ac,e 12

LTR-SGMMP-I 1-28 Rev. 1 NP-Attachment Table 5-2: Model D5 SLB Results at [ ]a"ce inches 1

MC 9533 9534 9536 ][ a,c,e Jac~e [ ]a,c,e(1) 9538 j[ a,c,e [ ]a,c,e [ ]a,c,e 9539 [ a,c,e [ ]a,ce [ ]a,c,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-1 7330-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 C2 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- 11-28 Rev. 1 NP-Attachment Question 6:

WCAP- I 7330-P, Revision 1, Figure 3 Should the data correspondingto the two open symbols be labeled as "data used in probabilisticanalysis"(consistentwith 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 [ ]a,c,e [ ]a,c,e ( ]ac,e 3.750 9536 [ ]axce [ ]a,c,e [ ]a,c,e 3.733 14

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

- a,c,e Figure 6-1 Update of Figure 3-45 of WCAP-17330, Revision 1 15

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

WCAP-I 7330-P, Revision 1, Tables 3-35 to 3 The numerical methods used to generate the accumulated pullout loads in these tables appearto contain two sources of non-conservatism. One, the distance below the top of the tubesheet (TTS) where the contact pressuretransitions from 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 contact pressure. The staff believes a more realisticand more conservative estimate of the contact pressurezero intercept value can be obtained by extrapolatingthe 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 distributionof contact pressure between these locationsis 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 correspondingto the H* distances in Figures 3-35, 3-46, and 3-48 are overestimatedby 17%, 6%, and 8%, respectively. Provide revisions to Tables 3-35 to 3-48, if and as needed, to addressthe staff's 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. I 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-1 7330-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 3000 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. 1 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 [ ]a,c,e inches and the H*

length for the Model D5's is conservative by about [ ]ace 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 Tubesheet CTE SDs, %

Temperature As Used in Improved 50% Improved 95%

(*F) WCAP 1 Confidence Confidence 17330,Rev. 1 300 1.62 [ ]a,c,e [ ]a,c,e 600 1.62 [ ]a,c,e ]a,c,e 18

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

Table 7-3 Summary of H* Lengths from WCAP-17330, Revision 1 Limiting Probabilistic H* Difference, Ratio fo 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 Model/Case Limiting Ratio New Probabilistic H* (Licensed H* - New Limiting__RatioProbabilistic H*)

F, 95/50 Whole Bundle F, 95/95 Whole Plant DS, 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, Figures 3-48 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 present form conservative?

Response

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

The Model D5 contact pressure results reported for the steam line break (SLB) condition 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-17330-P, Revision 1 was made to be consistent with the methodology described in WCAP-1 7072-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-17072-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-I 1-28 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-17330-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-17330-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. 1 NP-Attachment WCAP-1 7330-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, ifthe C2 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, ifthe C2 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 C2 analysis methodology is not required.

22

LTR-SGMMP- 11-28 Rev. 1 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. 1 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 D5: 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- 11-28 Rev. 1 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 uncertainty associatedwith the computed probabilisticH* values calculatedwith the C2 model as illustratedin Table 3-29. Depending on the response to question 8 above, there also may be some uncertainty associatedwith the H*

adjustments for the crevice pressuredistribution. 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 [ ]ace inches. The probabilistic H* value for the contact pressure distribution shown in the response to Question 8, Figure 8-2(a), is [ ]a,c,e 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 [ ]ac~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 [ ]ac,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. 1 NP-Attachment Table RAI 9-1 Conservatism in Current Model D5 H* Calculation WCAP-17330-P, Refined Source Rev 1 Calculations

_ in in - q P,,

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

Total Adjustments Final Probabilistic H* 14.01 [ ]a,¢ce Notes:

(1) Recalculated for [ ]a,c,e inches H* based on Figure 8-2(a).

(2) Crevice pressure margin ([ ]a"ce 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 [ ]ac~e Notes:

(1) Recalculated for [ ]ac"e inches H* based on Figure 8-2(b).

(2) Crevice pressure margin ([ ]ace 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- 11-28 Rev. 1 NP-Attachment Question 10:

Westinghouse letter LTR-SGMP-1O-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 reproducethe 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-10-95, 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-10-95, 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 = 12p 1 KI, but neglecting the constant because it divides out in the calculation of the resistance ratios.

29

LTR-SGMMP- 11-28 Rev. 1 NP-Attachment Table 10-1 Local Loss Coefficient for Power 2 (K=O.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.OOOOE+00 O.OOOOE+00 O.OOOOE+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+O0 O.O000E+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. 1 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 BTS from 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.OOE+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- 11-28 Rev. 1 NP-Attachment Question 11 Westinghouse letter LTR-SGMP-10-95 P - Attachment, Revision I - This report spells out the definition of Exp-2 and Power-2 in Table 5. Provide definitions of the other functions considered in the table.

Response

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

The following is a complete list of the functions with their definitions that were considered 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+12*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 l*exp(8E-05*Pc) Lower Bound Horizontal Exp-5 K = 1.1E+14*exp(1.8E-04*Pc) Upper Bound Horizontal Linear K = 6.5E+9*Pc Power-1 K = 1E+4*PA^3 Power-2 K = 0.15"(Pc) 4 5 Diagonal Bound Logarithmic K =1E+12*ln(Pc)+4E+08 Question 12 This question is a utility-specific question for which the respective utilitiesprovide 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- 11-28 Rev. 1 NP-Attachment Question 14 AW.AP 17345 F, Re-'ic. 2, Tab!- 3 50 2nd 3 51 W(AP-17330-P, Revision 1 Table 3 Are Is the footnotes in this table correct and complete? For Model 64F, Table 3-27 implies we have direct C2 calculations for rank orders 0025, 90673, .-d -0019186,9694 and 9890. Thus, for Table 3-"450, it seems at three of four cases are based on interpolated c t G. 2!

values. Sknr!"2. for A. 4- d-!f hF, Tab!l 3 27 .mnp!-. '.' hwe".. d*.rv o !nto, n for rrn!.

order6 0158=, 9697, and 0n760Ow. Thus, fop; Tabl 3 50 it o*c... ' the 'Whole p.t . 5/,

2."C bas.d on I!:ct . *2!-!2t:2n..- 2nd th. 2thcr Cf

-r. -ntA-rpo!atd V'!'c. If the staff's understandingis incorrect,clarify for which rank orders direct C2 calculations were performed and provide the H* calculations for these cases in a form similar to 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 02 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 02 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 similar question 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 application of a permanent H* ARC.

33

LTR-SGMMP- 11-28 Rev. 1 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: May 5, 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-1 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*

Approved by: D. Merkovsky Manager, SGDA

© 2011 Westinghouse Electric Company LLC All Rights Reserved

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

35

LTR-SGMMP- 11-28 Rev. 1 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 100°F to 700'F. The tubesheet data are in Table 1 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,

• 1 n-in- 1~)

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

n-i cgs =a U50 --

X.-1,0.95 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-1 1-28 Rev. 1 NP-Attachment standard formula below was used to calculate a high confidence standard deviation for the tube and tubesheet without instrumentation error:

• 95,Material - qyJ95total -- 0T2S,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 200'F and 650'F.

37

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

Temp (*F) Sample 1 Sample 2 Sample 3 Sample 4 Sample 5 Sample 6 Sample 7 Sample 8 Sample 9 Sample 10 100 125 150 175 200 225 250 275 300 325 350 375 400 425 450 475 500 525 550 575 600 625 650 675 ____________

700 38

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

Temp (°F) I Sample 111 Sample 12 Sample 13 1Sample 141 Sample 151 Sample 16 1 Sample 17 Sample 18 Sample 191 Sample 20 100 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 _____ ______________________ _

39

LTR-SGMMP- 11-28 Rev. 1 NP-Attachment Table 3 Tubesheet CTEs (gin / 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 a,c,e 125 150 175 225 250 275 300 325 350 375 400 425 450 475 500 525 550 575 600 625 650O _____ _____

675 _____ _____ ________________ _____ ____

700 _____ ______ _____ _____ _____

40

LTR-SGMMP- 11-28 Rev. 1 NP-Attachment Table 4 Tubesheet CTEs (Itin / 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 650O___________

675 __________________

700__________

41

LTR-SGMMP- 11-28 Rev. 1 NP-Attachment Table 5 Tube CTEs (Model F) (pin / in IF)

Temp (*F) Sample 1 Sample,2 Sample 3 Sample 4 Sample 5 Sample 6 Sample 7 Sample 8 Sample 9 Sample 10 100 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 42

LTR-SGMMP-11-28 Rev. 1 NP-Attachment Table 6 Tube CTEs (Model D5) (pin / in OF)

Temp (*F) Sample 11 Sample 12 Sample 13 Sample 14 Sample 15 Sample 16 Sample 17 Sample 18 Sample 19 Sample 20 100 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 7O00 43

LTR-SGMMP- 11-28 Rev. 1 NP-Attachment Table 7 Tube CTEs (Model 44F) (Iiin / 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 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 44

LTR-SGMMP-1 1-28 Rev. 1 NP-Attachment Table 8 Tube CTEs (Multiple runs on same specimen) (tin / in IF)

Temp (*F) Run i Run 2 Run 3 Run 4 Run 5 Run 6 Run 7 Run 8 Run 9 Run 10 100 a,c,e 125 150 175 200 225 250 275 300 325 350 375 400 425 450 475 500 525 550 575 600 625 650 675 700 45

LTR-SGMMP- 11-28 Rev. 1 NP-Attachment Table 9 Tubesheet CTEs (Multiple runs on same specimen) (gin / in IF) 9 i i 1 1 Temnp (°F) Run 2 Run 3 Run 4 Run 6 Run 8 Run Run 10 Run 1 Run 5 Run 7 Run 9 a c e Temp- -(0F)- I- Run 1 4 Run 2 4 Run 3 4 Run 4 Run S 4 Run 6 4 Run 7 1 Run 8 + Run 9 + 10 100 125 150 175 200 225 ______ _____

250 275 ______

300 _____

325 ______

350 375 400 425 450 475 ____ _

500 _____ _____

525 _____

550 575 ______

600 625 650 675 700 46

LTR-SGMMP-1 1-28 Rev. 1 NP-Attachment Table 10 Mean and Standard Deviation, Tube Material Temperature Mean Best Estimate Standard 95% Confidence Standard

(°F) (iiin/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-1 1-28 Rev. 1 NP-Attachment Table 11 Mean and Standard Deviation, Tubesheet Material Temperature Mean Best Estimate Standard 95% Confidence Standard

(°F) (IJin/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-1 1-28 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 F 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

Vogtle Electric Generating Plant - Units 1 and 2 License Amendment Request to Revise Technical Specification Sections 5.5.9, "Steam Generator (SG) Program" and 5.6.10, "Steam Generator Tube Inspection Report" Enclosure 6 Westinghouse Electric Company LLC CAW-12-3404 "Application for Withholding Proprietary Information from Public Disclosure," February 22, 2012

Westinghouse Westinghouse Electric Company Nuclear Services 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: GP-18874 CAW-12-3404 February 22, 2012 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE

Subject:

LTR-SGMMP-I 1-28 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)

The proprietary information for which withholding is being requested in the above-referenced report is further identified in Affidavit CAW-12-3404 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 Southern Nuclear Operating Company.

Correspondence with respect to the proprietary aspects of the application for withholding or the Westinghouse affidavit should reference this letter, CAW-12-3404, 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, A. Gresham, Manager Regulatory Compliance Enclosures

CAW-12-3404 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:

A ham, Manager Regulatory Compliance Sworn to and subscribed before me this 22nd day of February 2012 0 Notary Public COMMONWEALTH OF PENNSYLVANIA Notarial Seal Cynthia Olesky, Notary Public Manor Boro, Westmorelanld County My Commission Expires July 16, 2014 Member. Pennsvlvania Assodatlon of Notaries

2 CAW- 12-3404 (1) 1 am Manager, Regulatory Compliance, in Nuclear Services, Westinghouse Electric Company LLC (Westinghouse), and as such, I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rule making proceedings, and am authorized to apply for its withholding on behalf of Westinghouse.

(2) 1 am making this Affidavit in conformance with the provisions of 10 CFR Section 2.390 of the Commission's regulations and in conjunction with the Westinghouse Application for Withholding Proprietary Information from Public Disclosure accompanying this Affidavit.

(3) I 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- 12-3404 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-12-3404 (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 and belief (v) The proprietary information sought to be withheld in this submittal is that which is appropriately marked in LTR-SGMMP-1 1-28 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 Southern Nuclear Operating Company and Application for Withholding Proprietary Information from Public Disclosure, to the Document Control Desk. The proprietary information as submitted by Westinghouse for Vogtle Units I and 2, 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-12-3404 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 11* 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.

Vogtle Electric Generating Plant - Units 1 and 2 License Amendment Request to Revise Technical Specification Sections 5.5.9, "Steam Generator (SG) Program" and 5.6.10, "Steam Generator Tube Inspection Report" Enclosure 7 Response to Request for Additional Information Questions Specific to Vogtle Electric Generating Plant to NL-1 2-0561 Response to Request for Additional Information Questions Specific to Vogtle Electric Generating Plant Provided below are the Vogtle Electric Generating Plant Units 1 and 2 (VEGP) specific responses to questions 12 and 13 from the Duke Energy request for additional information (Reference 32 of Enclosure 1) and question 15 from the Dominion request for additional information (Reference 37 of Enclosure 1). The NRC questions are identified in italics.

12. BET measurements for Catawba 2, documented in Westinghouse letter LTR-SGMP-09-1 11 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 reportedrange of BETs, are there any non-expanded or partially expanded tubes at Catawba2? If so, provide revisions to the proposed technicalspecifications which exclude such tubes from the proposedH* provisions.

Response: Bottom expansion transition (BET) measurements for VEGP, documented in Westinghouse letter LTR-SGMP-09-111 P-Attachment, Revision 1, range to a maximum of 0.80 inches. Apart from tubes with this reported range of BETS, there are no non-expanded or partially expanded tubes in service at VEGP. 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 throughj - The proposedchanges contain more words than seem necessary,reducing the clarity of the proposedreporting requirements. For example, the proposed wording refers to "aninspection performed after each refueling outage"which doesn't seem to make sense. The NRC staff believes the proposedrequirements can be stated more clearly and concisely as follows:

h. For Unit 2, f"ll.wing .. mplotien of an in.pection p.. eFmed duri. g End

. f Cyc,,

17 Refun*, ig Outage (and any in.pections pe.fomed during subsequent Cy..*e is ope.....:n), the primary to secondary LEAKAGE rate observed in each steam generator(if it is not practicalto assign the leakage to an individual SG, the entire primary to secondaryleakage 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, folowing cp....etion of an inspectin pe.formed during the End e.

Cycle 17 Refuefing Outage (an an insectionS perqFomed durig subsequeni Cycle 13operation, the calculatedaccidentinduced leakage rate from the portion of the tubes below 20 14.01 inches from the top of the tubesheet for the most limiting accidentin 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 reportshall describe how it was determined, and

j. For Unit 2, following completion of an inspection performed during the End e, Cycl 17 Refueling Outago (and any insections performed durig subsequeni Cy!ce 18,per, tie%, 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 reportingrequirements as necessary to clarify their intent.

E7-1 to NL-12-0561 Response to Request for Additional Information Questions Specific to Vogtle Electric Generating Plant Response: The VEGP proposed changes to technical specification (TS) 5.6.10, "Steam Generator Tube Inspection Report," are consistent with the NRC staff's recommendation above for the Model F SG.

15. Verify that regulatorycommitments pertainingto monitoring for tube slippage and for primary to secondary leakage, as describedin Dominion letter dated December 16, 2010 (NRC ADAMS Accession No. ML103550206), Attachment 1, page 10 of 23, remain in place. In addition, revise the proposedamendment to include a revision to technicalspecification 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 VEGP letter NL-09-1411 dated September 11, 2009 (Reference 20 of Enclosure 1) remain in place as specified in the cover letter. SNC 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:

Primary-to-secondary leakage from tube degradation in the tubesheet area 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 100, General Design Criteria (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* justification.

The constraint that is provided by the tubesheet precludes tube burst from 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,"

(Reference 10) 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 is 1.0 gpm. The TS operational leak rate limit is 150 gpd (0.1 gpm) through any one SG.

Consequently, there is significant margin between accident leakage and allowable operational leakage. The SLB/FLB leak rate ratio is 2.48, resulting in significant margin between the conservatively estimated accident leakage and the allowable accident leakage (1.0 gpm).

E7-2

Vogtle Electric Generating Plant - Units 1 and 2 License Amendment Request to Revise Technical Specification Sections 5.5.9, "Steam Generator (SG) Program" and 5.6.10, "Steam Generator Tube Inspection Report" Enclosure 8 LTR-SGMMP-11-28 Errata, Rev. 1, "LTR-SGMMP-11-28, Revision 0 and Revision 1, P- and NP-Attachment Errata", March 20, 2012

Westinghouse Non-Proprietary Class 3 B 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-1 1-28 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- 11-28, Revision 1 (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 1 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-1 1-87 instead of LTR-SGMP-1 1-87.

For Revision 1 of LTR-SGMMP- 11-28, 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 1 and 2 are unaffected.

Page 2 of 2 Our ref: LTR-SGMMP-11-28 Errata, Rev. 1 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 records are authenticatedin the electronic document management system.

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

© 2012 Westinghouse Electric Company LLC All Rights Reserved