ML12110A114

Millstone, Unit 3 - License Amendment Request for Permanent Alternate Repair Criteria for Steam Generator Tube Inspection and Repair

ML12110A114
Person / Time
Site:	Millstone
Issue date:	04/12/2012
From:	Price J A Dominion Nuclear Connecticut, Dominion
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
12-203
Download: ML12110A114 (102)

Text

Dominion Nuclear Connecticut, Inc. / ., .5000 Dominion Boulevard, Glen Allen, VA 23060 Dominion Web Address: www.dom.com PROPRIETARY INFORMATION

-WITHHOLD UNDER 10 CFR 2.390 April 12, 2012 U.S. Nuclear Regulatory Commission Serial No.12-203 Attention:

Document Control Desk NSSL/MLC RO Washington, DC 20555 Docket No. 50-423 License No. NPF-49 DOMINION NUCLEAR CONNECTICUT.

INC.MILLSTONE POWER STATION UNIT 3 LICENSE AMENDMENT REQUEST FOR PERMANENT ALTERNATE REPAIR CRITERIA FOR STEAM GENERATOR TUBE INSPECTION AND REPAIR Pursuant to 10 CFR 50.90, Dominion Nuclear Connecticut, Inc. (DNC) hereby requests an amendment to Facility Operating License No. NPF-49 for Millstone Power Station Unit 3 (MPS3). This amendment request proposes to permanently revise Technical Specification (TS)6.8.4.g, "Steam Generator (SG) Program," to exclude a portion of the steam generator tubes below the top of the steam generator tubesheet from periodic inspections.

Inclusion of the permanent alternate repair criteria (PARC) in TS 6.8.4.g permits deletion of the previous temporary alternate repair criteria (TARC) for Cycle 15. In addition, this amendment request also proposes to revise the reporting criteria in TS 6.9.1.7, "Steam Generator Tube Inspection Report," to remove reference to the previous Cycle 15 TARC, and add reporting requirements specific to the PARC.The proposed amendment constitutes a redefinition of the steam generator tube primary-to-secondary pressure boundary and defines the safety significant portion of the tube that must be inspected and plugged, as necessary.

Tube flaws detected below the safety significant portion of the tube are not required to be plugged. Allowing flaws in the non-safety significant portion of the tube to remain in service minimizes unnecessary tube plugging while maintaining the safety margin of the steam generators in performance of the safety functions necessary to maintain reactor coolant pressure boundary, reactor coolant flow, and primary-to-secondary heat transfer.The proposed TS changes are based on the supporting structural analysis and leakage evaluation completed by Westinghouse Electric Company, LLC (Westinghouse).

The documentation supporting the Westinghouse analysis is described in Section 4.0, "Summary of Licensing Basis Analysis (H* Analysis)," of Enclosure 1 and provides the licensing basis for this change.Enclosure 1 provides a description and basis for the proposed changes. The marked-up TS pages for the proposed changes are provided in Enclosure

2. Enclosures 3 and 4 contain the proprietary and non-proprietary Westinghouse information, respectively, which supports the analysis provided in Enclosure

1. Enclosure 5 contains an errata letter from Westinghouse, dated March 20, 2012, to correct an administrative error in the' non-proprietary document provided in Enclosure

4. As Enclosure 3 contains information proprietary to Westinghouse; Enclosure 6 contains the supporting affidavit signed by Westinghouse, the owner of the information.

This affidavit sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses, with specificity, the considerations listed in paragraph (b)(4) of 10 CFR 2.390 of the Commission's regulations.

Accordingly, it is respectfully requested that the information, which is proprietary to Westinghouse, be withheld ATTACHMENT 3 CONTAINS PROPRIETARY INFORMATION THAT IS BEING WITHHELD FROM PUBLIC DISCLOSURE UNDER 10 CFR 2.390. UPON SEPARATION OF ATTACHMENT 3, THIS PAGE IS DECONTROLLED.

Serial No: 12-203 Docket No. 50-423 Page 2 of 3 from public disclosure in accordance with 2.390 of the Commission's regulations.

Enclosure 7 provides the MPS3-specific responses to the relevant plant-specific requests for additional information submitted to Catawba Units 1 and 2 (Questions 12 and 13) and Surry Units 1 and 2 (Question 15). Enclosure 8 addresses the previous commitments associated with steam generator alternate repair criteria.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.

DNC requests approval of the proposed license amendment by April 14, 2013 to support 3R15, which is scheduled to start on April 14, 2013. Once approved, the proposed changes will be implemented within 30 days of issuance of the amendment and prior to Mode 5 startup of MPS3.Should you have any questions in regard to this submittal, please contact Wanda Craft at (804)273-4687.Sincerely, J. an ice V VICKI L. HULL Vi e Pre ident -Nuclear Engineering

] Notary Public Commonwealth of Virginia 140542 COMMONWEALTH OF VIRGINIA My Commission Expires May 31. 2014 COUNTY OF HENRICO The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by J. Alan Price, who is Vice President

-Nuclear Engineering of Dominion Nuclear Connecticut, Inc. He has affirmed before me that he is duly authorized to execute and file the foregoing document in behalf of that company, and that the statements in the document are true to the best of his knowledge and belief.-7-1/Acknowledged before me this _"day of,&ra ,2012.My Commission Expires: L/'.qv .3l Zcui)/I I 11/1_6L /L AdL Notary Public Commitments:

None Serial No: 12-203 Docket No. 50-423 Page 3 of 3

Enclosures:

1. Basis for Proposed Change 2. Mark-up of Proposed Technical Specification Changes 3. Westinghouse Electric Company LLC, LTR-SGMMP-1 1-28 Rev. 1 P-Attachment,"Response to USNRC Request for Additional Information Regarding the License Amendment Requests for Permanent Application of the Alternate Repair Criterion, H*, to the Model D5 and Model F SGs," February 2, 2012. (Proprietary)

4. Westinghouse Electric Company LLC, LTR-SGMMP-1 1-28 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," February 2, 2012. (Non-Proprietary)

5. Westinghouse Electric Company LLC, LTR-SGMMP-1 1-28 Errata Rev. 1, "LTR-SGMMP-11-28 Revision 0 and Revision 1, P- and NP- Attachment Errata," March 20, 2012. (Non-Proprietary)

6. Westinghouse Electric Company LLC, CAW-1 2-3446, "Application for Withholding Proprietary Information from Public Disclosure," March 21, 2012. (Affidavit for LTR-SGMMP-1 1-28 Rev. 1 P-Attachment)

7. Millstone Power Station Unit 3, Plant-Specific Response to Requests for Additional Information

8. Millstone Power Station Unit 3, Commitments from Previous Steam Generator Alternate Repair Criteria cc: U.S. Nuclear Regulatory Commission Region I Regional Administrator 475 Allendale Road.King of Prussia, PA 19406-1415 J. S. Kim NRC Project Manager U.S. Nuclear Regulatory Commission, Mail Stop 08 C2A One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Millstone Power Station Director, Bureau of Air Management Monitoring and Radiation Division Department of Energy and Environmental Protection 79 Elm Street Hartford, CT 06106-5127 Serial No: 12-203 Docket No. 50-423 ENCLOSURE 1 Basis for Proposed Change DOMINION NUCLEAR CONNECTICUT, INC.MILLSTONE POWER STATION UNIT 3 Serial No: 12-203 Docket No. 50-423 Enclosure 1, Page"1 of 22 Basis for Proposed Change Table of Contents 1.0 Summary Description 2.0 Detailed Description

3.0 Background

4.0 Summary of Licensing Basis Analysis (H* Analysis)5.0 Technical Evaluation 6.0 Regulatory Evaluation 6.1 Applicable Regulatory Requirements

/ Criteria 6.2 No Significant Hazards Consideration

6.3 Precedents

6.4 Conclusion 7.0 Environmental Considerations 8.0 References Serial No: 12-203 Docket No. 50-423 Enclosure 1, Page 2 of 22 1.0 Summary Description Pursuant to 10 CFR 50.90, Dominion Nuclear Connecticut, Inc. (DNC) hereby requests an amendment to Facility Operating License No. NPF-49 for Millstone Power Station Unit 3 (MPS3). This amendment request proposes to permanently revise Technical Specification (TS) 6.8.4.g, "Steam Generator (SG) Program," to exclude a portion of the steam generator tubes below the top of the steam generator tubesheet from periodic inspections.

Inclusion of the permanent alternate repair criteria (PARC) in TS 6.8.4.g permits deletion of the previous temporary alternate repair criteria (TARC) for Cycle 15. In addition, this amendment request also proposes to revise the reporting criteria in TS 6.9.1.7, "Steam Generator Tube Inspection Report," to remove reference to the previous Cycle 15 TARC, and add reporting requirements specific to the PARC.The proposed TS changes are based on the supporting structural analysis and leakage evaluation completed by Westinghouse Electric Company, LLC (Westinghouse).

The documentation supporting the Westinghouse analysis is described in Section 4.0 of this enclosure and provides the licensing basis for this change.2.0 Detailed Description Proposed Changqes to TS 6.8.4..q.c:

Deleted text is struck through and added text is italicized and bold.c. Provisions for SG tube repair criteria:

Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.The following alternate tube repair criteria shall be applied as an alternative to the 40% depth-based criteria: 1. Refueling Outage 14 and the subsequent operating 7Tubes 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.

Proposed Changes to TS 6.8.4..q.d:

Deleted text is struck through and added text is italicized and bold.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 Serial No: 12-203 Docket No. 50-423 Enclosure 1, Page 3 of 22 tube repair criteria.

For Refueling Outage 14 and the subsequcnt epeFatig-eyeT Pportions 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 non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.Proposed Changies to TS 6.9.1.7.i, 6.9.1.7.m, and 6.9.1.7.k:

Deleted text is struck through and added text is italicized and bold.i. During Refueling Outage 14 and the subsequent operating cycle-, The primary to secondary LEAKAGE rate observed in each SG (if it is not practical to assign the LEAKAGE to an individual SG, the entire primary to secondary LEAKAGE should be conservatively assumed to be from one SG) during the cycle preceding the inspection which is the subject of the report, j. DuFrig Refueling Outage 14 and the ,

operating The calculated accident induced leakage rate from the portion of the tubes Serial No: 12-203 Docket No. 50-423 Enclosure 1, Page 4 of 22/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 2.49 times the maximum operational primary to secondary leakage rate, the report should describe how it was determined; and k. Refueling Outage 14 and the subsequent operat!ng Tihe results of monitoring for tube axial displacement (slippage).

If slippage is discovered, the implications of the discovery and corrective action shall be provided.3.0 Background MPS3 is a four loop Westinghouse designed plant with Model F steam generators having 5,626 tubes in each steam generator.

A total of 177 tubes are currently plugged throughout the four steam generators.

The design of the steam generators include Alloy 600 thermally treated tubing, full depth hydraulically expanded tubesheet joints, and stainless steel tube support plates with broached hole quatrefoils.

The steam generator inspection scope is governed by TS 6.8.4.g, "Steam Generator (SG)Program," Nuclear Energy Institute (NEI) 97-06, "Steam Generator Program Guidelines," (Reference 1); EPRI "Pressurized Water Reactor Steam Generator Examination Guidelines," (Reference 2); EPRI "Steam Generator Integrity Assessment Guidelines," (Reference 3); ER-AP-SGP-101, "Steam Generator Program," (Reference 4), 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 MPS3.The inspection techniques, essential variables and equipment are qualified to the EPRI Steam Generator Examination Guidelines.

Catawba Nuclear Station, Unit 2, (Catawba) reported indication of cracking following nondestructive eddy current examination of the steam generator tubes during their fall 2004 outage. NRC Information Notice (IN) 2005-09, "Indications in Thermally Treated Alloy 600 Steam Generator Tubes and Tube-to-Tubesheet Welds," (Reference 5), 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.

DNC policies and programs require the use of applicable industry operating experience in the operation and maintenance of MPS3. The experience at Catawba, as noted in IN 2005-09, shows the importance of monitoring all tube indications (such as bulges, dents, dings, and other anomalies from the manufacture of the steam generators) with techniques capable of finding potential forms of degradation that may be occurring at these locations (as discussed in Generic Letter 2004-001, "Requirements for Steam Generator Tube Inspections" -Reference 6). Since the MPS3 Westinghouse Model F steam generators were fabricated Serial No: 12-203 Docket No. 50-423 Enclosure 1, Page 5 of 22 with Alloy 600 thermally treated tubes similar to the Catawba Unit 2 Westinghouse Model D5 steam generators, a potential exists for MPS3 to identify tube indications similar to those reported at Catawba within the hot leg tubesheet region.Potential inspection plans for the tubes and tube welds underwent intensive industry discussions in March 2005. The findings in the Catawba steam generator tubes present two distinct issues with regard to the steam generator tubes at MPS3: 1 1) Indications in internal bulges and overexpansions within the hot leg tubesheet; and 2) Indications at the elevation of the tack expansion transition.

Prior to each steam generator tube inspection, a degradation assessment, which includes a review of operating experience, is performed to identify degradation mechanisms that have a potential to be present in the MPS3 steam generators.

A validation assessment is also performed to verify that the eddy current techniques utilized are capable of detecting those flaw types that are identified in the degradation assessment.

Based on the Catawba operating experience, MPS3 revised the steam generator inspection plan for Refueling Outage 10 (fall 2005) and subsequent refueling outages to include sampling of bulges and overexpansions within the tubesheet region on the hot leg side. The sample was based on the guidance contained in the EPRI "Pressurized Water Reactor Steam Generator Examination Guidelines" and TS 6.8.4.g, "Steam Generator (SG) Program." Degradation was not detected in the tubesheet region in Refueling Outage 10 or Refueling Outage 11.For Refueling Outage 12 (fall 2008) and the subsequent operating cycle (Cycle 13), an interim alternate repair criteria (IARC) was approved as License Amendment (LA) 245 (Reference

7) which revised TS 6.8.4.g. The IARC required full-length inspection of the tubes within the tubesheet but did not require plugging tubes if any circumferential cracking observed in the region greater than 17 inches from the top of the tubesheet was less than a value sufficient to permit the remaining circumferential ligament to transmit the limiting axial loads. During these inspections, indications were identified in the hot leg tube ends of steam generators

'A' and 'C', which required tube end inspection scope expansions that included steam generators

'B' and 'D'. Indications were observed at the hot leg tube ends in all four steam generators and in one cold leg tube end in steam generator

'D'. These indications were within approximately 0.75 inches from the tube end. Indications with circumferential extent greater than 94 degrees and mixed-mode indications were plugged. All axial and circumferential oriented indications 94 degrees or less in circumferential extent, were left in service consistent with the criteria provided in the IARC. Axial indications and indications with circumferential extent of up to, and including 94 degrees, do not challenge the structural and leakage integrity requirements of NEI 97-06.For Refueling Outage 13 (spring 2010) and the subsequent operating cycle (Cycle 14), one-time alternate repair criteria (ARC) was approved as LA 249 (Reference

8) which again revised TS 6.8.4.g. The one-time ARC excluded portions of the tubes within the tubesheet (i.e., greater than 13.1 inches below the top of the tubesheet) from periodic steam generator tube inspections.

No indications of cracking were detected in any tube during Refueling Outage 13. During these inspections and at the request of the NRC, DNC performed a validation of the tube expansion from the top of tubesheet to the bottom of expansion transition (BET) to determine if there were any "significant" deviations that would invalidate assumptions in WCAP-17071-P (Reference 18). DNC completed the validation for MPS3 Serial No: 12-203 Docket No. 50-423 Enclosure 1, Page 6 of 22 and the results were provided to the NRC in DNC letter 10-276, dated April 26, 2010 (Reference 13). Based on review of BET values, a total of seven tubes were identified with BET values greater than 1.0 inch from the top of the tubesheet (value considered "significant" by Westinghouse).

As committed to by DNC (Reference 13), these seven tubes were removed from service during Refueling Outage 14.For Refueling Outage 14 (fall 2011) and the subsequent operating cycle (Cycle 15), the TARC was approved as LA 252 (Reference

29) which again revised TS 6.8.4.g. The TARC excluded portions of the tubes within the tubesheet (i.e., greater than 15.2 inches below the top of the tubesheet) from periodic steam generator tube inspections.

The degradation mechanisms found during Refueling Outage 14 included; anti-vibration bar wear, tube support plate wear, volumetric indications from fabrication and volumetric degradation from foreign object wear. No indications of cracking were detected during these inspections.

Based on these inspections, no indications of a 360 degree sever have been detected in any steam generator at MPS3. Consequently, the level of degradation in the MPS3 steam generators is very limited compared to the assumption of "all tubes severed" that was utilized in the development of the permanent H* value. Thus, structural integrity will be assured for the operating period between inspections allowed by TS 6.8.4.g, "Steam Generator (SG)Program." To prevent unnecessary plugging of additional tubes in the MPS3 steam generators, DNC is proposing a permanent change to TS 6.8.4.g to limit the steam generator tube inspection and repair (plugging) to the portion of tube from 15.2 inches below the top of the tubesheet on the hot leg side to 15.2 inches below the top of the tubesheet on the cold leg side. In addition, this amendment request proposes to revise TS 6.9.1.7, "Steam Generator Tube Inspection Report," to provide reporting requirements specific to this PARC.4.0 Summary of Licensing Basis Analysis (H* Analysis)The following is based on the application submitted by Southern Nuclear Operating Company, Inc. (SNC) for Vogtle Units 1 and 2, which (through spring 2011) functioned as the lead plant for application of the PARC. The Vogtle-specific information is indented and italicized.

On May 19, 2009, Westinghouse WCAP-17071-P, Revision 0, "H*: Alternate Repair Criteria for the Tubesheet Expansion Region in Steam Generators with Hydraulically Expanded Tubes (Model F)," was submitted as part of the Southern Nuclear Operating Company (SNC) request to change Technical Specification (TS) 5.5.9, "Steam Generator (SG) Program", and TS 5.6.10, "Steam Generator Tube Inspection Report" to support implementation of a PARC for steam generator tubes.As part of the review of SNC's license amendment request (LAR) dated May 19, 2009, the NRC issued two requests for additional information (RAIs) to SNC on July 10, 2009 and August 5, 2009. The July 10, 2009 RAI contained twenty-four (24) questions, with three of the questions being site-specific.

The August 5, 2009 RAI contained three questions related to Questions 4, 20 and 24 from the July 10, 2009 RAI, as well as one additional site-specific question (NRC Question 25). With the exception of the four site-specific questions, Westinghouse developed responses to the questions in L TR-Serial No: 12-203 Docket No. 50-423 Enclosure 1, Page 7 of 22 SGMP-09-100 P-Attachment (Reference 9), with the response to RAI #4 being provided under LTR-SGMP-09-109 P-Attachment (Reference 10). RAIs submitted to Wolf Creek, Byron/Braidwood, Comanche Peak and Seabrook, also included one additional question (NRC Question 26) that was not included on the Vogtle docket.This question was also addressed by Westinghouse in LTR-SGMP-09-100 P-Attachment.

SNC responses to the four site-specific RAI questions, as well as the Westinghouse RAI responses, were submitted to the NRC in two separate letters (Serial Nos. NL-09-1265 and NL-09-1375), both dated August 28, 2009.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, as supplemental information (Reference 26).On September 11, 2009, SNC submitted a request to revise the May 19, 2009 LAR to be an interim change for Vogtle Units 1 and 2. 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 tubesheet tube bore in normal and accident conditions (RAI Question 4 of the July 10, 2009 letter and RAI Question I of the August 5, 2009 letter) have 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 PARC amendment request.On November 23, 2009, the NRC issued a letter to SNC (Reference

11) documenting the currently identified and unresolved issues relating to tubesheet bore eccentricity.

This letter contained fourteen (14) questions which required resolution before the NRC could complete its review of a PARC amendment request.On November 23, 2009, DNC submitted a LAR to revise TS 6.8.4.g, "Steam Generator (SG)Program," and TS 6.9.1.7, "Steam Generator Tube Inspection Report," to support implementation of a one-time ARC for Refueling Outage 13 and the subsequent operating cycle (Cycle 14) (Reference 12). Westinghouse WCAP-17071-P, Revision 0, "H*: Alternate Repair Criteria for the Tubesheet Expansion Region in Steam Generators with Hydraulically Expanded Tubes (Model F)" was submitted as Enclosure

5. Also included in the November 23, 2009 submittal were DNC's responses to the four site-specific RAI questions, as well as the Westinghouse RAI responses provided in LTR-SGMP-09-100-P (Reference

9) and LTR-SGMP-09-109 P-Attachment (Reference 10).As a condition for approving the one-time ARC, the NRC staff requested that DNC perform a validation of the tube expansion from the top of tubesheet to the bottom of expansion transition (BET) to determine if there are any "significant" deviations that would invalidate assumptions in WCAP-17071-P (Reference 18). DNC completed the validation for MPS3 and the results were provided to the NRC in DNC letter 10-276, dated April 26, 2010 (Reference 13). Based on review of BET values, a total of seven tubes were identified with BET values greater than 1.0 inch from the top of the tubesheet (value considered "significant" by Westinghouse).

As a result, DNC committed to remove these tubes from service no later than the next scheduled inspection.

Note: LTR-SGMP-09-1 11 P-Attachment, Rev. 1,"Acceptable Value of the Location of the Bottom of the Expansion Transition (BET) for Implementation of H*," dated September 2010 (Reference

14) was subsequently developed by Westinghouse to support plant determinations of BET measurements and their significant Serial No: 12-203 Docket No. 50-423 Enclosure 1, Page 8 of 22 deviation assessment.

This document was submitted to the NRC under Westinghouse letter LTR-NRC-10-69 (Reference 15).On May 3, 2010, the NRC issued LA 249 (Reference

8) for Refueling Outage 13 and the subsequent operating cycle (Cycle 14). During the NRC's administrative processing of the supporting documentation for LA 249, the NRC discovered discrepancies with proprietary markings in some of the Westinghouse documents supporting MPS3's LAR dated November 23, 2009. On September 22, 2010, the NRC issued a letter to DNC (Reference 16)requesting affected documents be revised and resubmitted to meet the requirements of 10 CFR 2.390(b)(1)(i).

In letter dated October 7, 2010 (Reference 17), DNC resubmitted corrected documents which included WCAP-17071-P, Rev. 2 (Reference 18).and LTR-SGMP-09-100 P-Attachment, Rev. 1 (Reference 19).The following documents have been prepared by Westinghouse to provide final resolution of the remaining questions identified in the November 23, 2009 NRC letter to SNC (Reference

11) in support of the PARC amendment for MPS3:* WCAP-17330-P, Revision 1, "H*: Resolution of NRC Technical Issue Regarding Tubesheet Bore Eccentricity (Model F/Model D5)," June 2011 (Reference 24),* 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 20)," LTR-SGMP-10-33 P-Attachment, "H* Response to NRC Questions Regarding Tubesheet Bore Eccentricity," September 13, 2010 (Reference 21).On June 30, 2011, Duke Energy submitted a LAR (Reference

30) for permanent application of-the alternate repair criterion H* at Catawba Unit 2 based on the technical justification in WCAP-17330-P, Revision 1, "H*: Resolution of NRC Technical Issue Regarding Tubesheet Bore Eccentricity (Model F/Model D5)" (Reference 24). A supplement (Reference

31) to the LAR was submitted on July 11, 2011 and provided Westinghouse Electric Company LLC, LTR-SGMP-11-58, "WCAP-17330-P, Revision 1 Erratum" (Reference 37). On January 5, 2012, a RAI from the NRC (Reference

32) was transmitted electronically to Duke Energy.Duke Energy responded to the RAI on January 12, 2012 (Reference 33).Subsequent to the Duke Energy LAR, Virginia Electric and Power Company (Dominion) submitted a LAR (Reference

34) 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 35). Dominion responded to the RAI on February 14, 2012 (Reference 36).Westinghouse Electric Company LLC, LTR-SGMMP-1 1-28 Rev.1 P-Attachment (Enclosure 3), "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," augmented the responses to the Duke Energy RAI to include similar responses applicable to Model F steam generators.

Additionally, this letter addressed the Surry RAI Question 14 for the Model F steam generators.

Westinghouse letter LTR-SGMMP-1 1-28, Revision 1 NP Attachment Errata is included in Enclosure 5 to correct an administrative error in the non-proprietary version of LTR-SGMMP-1 1-28 Rev. 1 that is Serial No: 12-203 Docket No. 50-423 Enclosure 1, Page 9 of 22 contained in Enclosure

4. Enclosure 7 provides MPS3-specific responses to Questions 12 and 13 from the Catawba RAIs and Question 15 from the Surry RAIs. The NRC questions for Catawba and Surry are identified in italics in Enclosure 7.The following table provides the list of licensing basis documents for H* for MPS3.Document Revision Title Reference Reference That Number Number Number Submitted Document to N.RC WCAP-17071-P 2 H*: Alternate Repair Criteria for the 18 17 Tubesheet Expansion Region in Steam Generators with Hydraulically Expanded Tubes (Model F)LTR-SGMP 1 Response to NRC Request for 19 17 100 P-Attachment Additional Information on H*; Model F and Model D5 Steam Generators LTR-SGMP 0 Response to NRC Request for 10 12 109 P-Attachment Additional Information on H*; RAI #4;Model F and Model D5 Steam Generators WCAP-17330-P 1 H*: Resolution of NRC Technical 24 31 Issue Regarding Tubesheet Bore Eccentricity (Model F/Model D5)LTR-SGMP-10-78 0 Effects of Tubesheet Bore 20 22 P-Attachment Eccentricity and Dilation on Tube-to-Tubesheet Contact Pressure and Their Relative Importance to H*LTR-SGMP-1 0-33 0 H*: Response to NRC Questions 21 23 P-Attachment Regarding Tubesheet Bore Eccentricity LTR-SGMMP 1 Response to USNRC Request for Enclosure 3 This LAR 28 P-Attachment dditional Information Regarding the submittal License Amendment Requests for Permanent Application of the Alternate Repair Criterion, H*, to the Model D5 and Model F SGs Westinghouse letter LTR-SGMP-1 1-58 (Reference

37) corrected a transposition of numbers in Table 3-30 of WCAP-17330-P, Revision 1 in the Model D5 section of the table. The data for the Model F SGs is unaffected.

Serial No: 12-203 Docket No. 50-423 Enclosure 1, Page 10 of 22 In addition, the following correspondence is also applicable to the MPS3 permanent alternate repair criteria request:* A March 28, 2011 letter from the NRC to SNC (Reference

38) documented the summary of a February 16, 2011 public meeting regarding steam generator tube inspection permanent alternate repair criteria.

Enclosure 3 of the NRC letter provided technical NRC staff questions developed at the public meeting. Responses to these questions have been incorporated into WCAP-1 7330-P, Revision 1 (Reference 24).* Section 1.3 of Reference 24 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

39) included a pre-submittal consideration of steam generator alternate repair criteria requirements RAI. The response to the NRC RAIs is provided in SNC letter NL-1 1-1178 (Reference 40).5.0 Technical Evaluation To preclude unnecessarily plugging tubes in the MPS3 steam generators, an evaluation was performed to identify the safety significant portion of the tube within the tubesheet necessary to maintain structural and leakage integrity in both normal and accident conditions.

Tube inspections will be limited to identifying and plugging degradation in the safety significant portion of the tubes. The technical evaluation for the inspection and repair methodology is provided in the H* analysis described in Section 4.0. 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 MPS3 Model F steam generator considering the most stringent loads associated with plant operation, including transients and postulated accident conditions.

The limited tubesheet inspection criteria were selected to prevent tube burst and axial separation due to axial pullout forces acting on the tube and to ensure that the accident induced leakage limits are not exceeded.

The H* analysis provides technical justification for limiting the inspection in the tubesheet expansion region to less than the full depth of the tubesheet.

The basis for determining the safety significant portion of the tube within the tubesheet is based upon evaluation and testing programs that quantified the tube-to-tubesheet radial contact pressure for bounding plant conditions as described in the H* analysis.

The tube-to-tubesheet radial contact pressure provides resistance to tube pullout and resistance to leakage during plant operation and transients.

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 primary-to-secondary leakage are evaluated to ensure that they remain within regulatory limits (e.g., 10 CFR 50.67 and 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.

Serial No: 12-203 Docket No. 50-423 Enclosure 1, Page 11 of 22 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 25)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 gallon per minute (gpm). The TS operational leak rate limit is 150 gallons per day (gpd) (0.1 gpm) through any one steam generator.

Consequently, there is significant margin between accident leakage and allowable operational leakage. The SLB/FLB leak rate ratio is 2.49, resulting in significant margin between the conservatively estimated accident leakage and the allowable accident leakage (1.0 gpm).Plant-specific operating conditions are used to generate the overall leakage factor ratios that are to be used in the condition monitoring and operational assessments.

The plant-specific data provide the initial conditions for application of the transient input data. The results of the analysis of the plant-specific inputs, to determine the bounding plant for each model of steam generator, are contained in Section 6 of Reference 18.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 events and FLB events for specific break sizes and operating conditions, 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 2.49 (Table RA124-2 (Revised Table 9-7) of Reference 19).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 steam generator hot and cold legs for a period of time.As previously noted dynamic viscosity decreases with increasing temperature, therefore; leakage would be expected to increase due to decreasing viscosity, 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 steam generators will exceed the normal operating differential pressure across the tubesheet is less than 30 seconds. As the accident induced leakage performance criteria is defined in gallons per minute, the leak rate for a locked rotor 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 1.78 (Table RA124-2 (Revised Table 9-7) of Reference

19) for MPS3 is adjusted downward to a factor of 0.89.Similarly, for the control rod ejection event, the duration of the elevated pressure differential is less than 10 seconds. Thus, the peak leakage factor may be reduced by a factor of six Serial No: 12-203 Docket No. 50-423 Enclosure 1, Page 12 of 22 from 2.69 to 0.45.The plaht 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 steam generator exhibits a cool down characteristic instead of a heat-up transient as generally presented in steam generator design transients and in the UFSAR Chapter 15.0 safety analysis.

The use of either the component design specification transient or the Chapter 15.0 safety transient for leakage analysis for FLB is overly conservative because: " The assumptions on which the FLB design transient is based are specifically intended to establish a conservative structural (fatigue) design basis for RCS components; however, H* does not involve component structural and fatigue issues. The best estimate transient is considered more appropriate for use in the H* leakage calculations.

• For the Model F steam generator, the FLB transient curve (Figure 9-5, Reference 18)represents a double-ended rupture of the main feedwater line 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 significantly 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 steam generator liquid mass available to provide cooling to the RCS.A SLB event would have similarities to a FLB except that the break flow path would include the secondary separators, which could only result in an increased initial cooldown (because of retained liquid inventory available for cooling) when compared to the FLB transient.

A SLB could not result in more limiting 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 steam generator and control of temperature by the AFW system. Steam pressure control would be established by either the steam generator safety valves or control system (atmospheric relief valves). For any of the steam pressure control operations, the maximum 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 2.49 for MPS3, for a postulated SLB/FLB, has been calculated as shown in Table RA124-2 (Revised Table 9-7) of Reference 19 and includes consideration for a FLB heat-up event. Specifically, for the condition monitoring (CM) assessment, the Serial No: 12-203 Docket No. 50-423 Enclosure 1, Page 13 of 22 component of leakage from the prior cycle from below the H* distance will be multiplied by a factor of 2.49 and added to the total leakage from any other source and compared to the allowable accident induced leakage limit. For the operational assessment (OA), the difference in the leakage between the allowable leakage and the accident induced leakage from sources other than the tubesheet expansion region will be divided by 2.49 and compared to the observed operational leakage.WCAP-1 7071-P (Reference

18) redefines the primary pressure boundary.

The tube-to-tubesheet weld no longer functions as a portion of this boundary.

The hydraulically expanded portion of the tube into the tubesheet over the H* distance now functions as the primary pressure boundary in the area of the tube and tubesheet, maintaining the structural and leakage integrity over the full range of steam generator operating conditions, including the most limiting accident conditions.

The evaluation in WCAP-17071-P 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 bundle. As a condition of approval of LA 249 (Reference 8), DNC committed to monitor for tube slippage as part of the steam generator tube inspection program. This requirement will remain in place to support this LAR.Also, as a condition for approval of LA 249, the NRC staff requested that DNC perform a validation of the tube expansion from the top of tubesheet to the bottom of expansion transition (BET) to determine if there are any "significant" deviations that would invalidate assumptions in WCAP-17071-P (Reference 18). DNC completed the validation for MPS3 and the results were provided to the NRC in DNC letter 10-276, dated April 26, 2010 (Reference 13). Based on review of BET values, a total of seven tubes were identified with BET values greater than 1.0 inch from the top of the tubesheet (value considered "significant" by Westinghouse).

As committed to by DNC in Reference 13, these seven tubes were removed from service during Refueling Outage 14.6.0 Regulatory Evaluation 6.1 Applicable Regulatory Requirements/Criteria General Design Criteria (GDC) 1, 2, 4, 14, 30, 31, and 32 of 10 CFR 50, Appendix A, define requirements for the reactor coolant pressure boundary (RCPB) with respect 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.

Serial No: 12-203 Docket No. 50-423 Enclosure 1, Page 14 of 22 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, X, XI, and XVI of Appendix B and include control of special processes, inspection, test control, and corrective action.10 CFR 50.67, Accident Source Term, establishes limits on the accident source term used in design basis radiological consequence analyses with regard to radiation exposure to members of the public and to control room occupants.

Under 10 CFR 50.65, Maintenance Rule, licensees classify steam, generators as risk significant components because they are relied upon to remain functional during and after design basis events. Steam generators are to be monitored under 10 CFR 50.65(a)(2) against industry established performance criteria.

Meeting the performance criteria of NEI 97-06, Revision 3, provides reasonable assurance that the steam generator tubing remains capable of fulfilling its specific safety function of maintaining the RCPB. The steam generator performance criteria from NEI 97-06, Revision 3, 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.

• 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 (i.e., 15.2 inches from the top of the tubesheet) does not require plugging and serves as the bases for the steam generator inspection program. As such, the MPS3 inspection program provides a high level of confidence that the structural and leakage criteria are maintained during normal operating and accident conditions.

Serial No: 12-203 Docket No. 50-423 Enclosure 1, Page 15 of 22 6.2 No Significant Hazards Consideration This amendment application proposes to revise Millstone Power Station Unit 3 (MPS3)Technical Specification (TS) 6.8.4.g, "Steam Generator (SG) Program," to exclude portions of the tubes within the tubesheet from periodic steam generator inspections.

In addition, this amendment proposes to revise Technical Specification (TS) 6.9.1.7, "Steam Generator Tube Inspection Report" to remove reference to previous (Cycle 15) temporary alternate repair criteria (TARC) and provide reporting requirements specific to the permanent alternate repair criteria (PARC). Application of the structural analysis and leak rate evaluation results, to exclude portions of the tubes from inspection and repair is interpreted to constitute a redefinition' of the primary-to-secondary pressure boundary.The proposed change defines the safety significant portion of the tube that must be inspected and repaired.

A justification has been developed by Westinghouse Electric Company, LLC to identify the specific inspection depth below which any type of axial or circumferential primary water stress corrosion cracking can be shown to have no impact on the performance criteria of Nuclear Energy Institute (NEI) 97-06, "Steam Generator Program Guidelines," (Reference 1). The evaluation determined that degradation in tubing below 15.2 inches from the top of the tubesheet does not require plugging and serves as the bases for the steam generator inspection program.DNC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92,"Issuance of amendment," 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) postulated accidents.

During the SGTR event, the required structural integrity margins of the steam generator tubes and the tube-to-tubesheet joint over the H* distance will be maintained.

Tube rupture in tubes with cracks within the tubesheet is precluded by the 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 Serial No: 12-203 Docket No. 50-423 Enclosure 1, Page 16 of 22 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

25) 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 *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 steam line break (SLB) are also not significantly affected by the proposed changes. During a SLB 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 mid-plane.

Primary-to-secondary leakage from tube degradation in the tubesheet area during the limiting accident (i.e., a SLB) 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 2.49 for Millstone Power Station Unit 3 (MPS3), for a-postulated SLB/FLB, has been calculated as shown in Table RA124-2 (Revised Table 9-7) of Reference

19. Specifically, for the condition monitoring (CM) assessment, the component of leakage from the prior cycle from below the H* distance will be multiplied by a factor of 2.49 and added to the total leakage from any other source and compared to the allowable accident induced leakage limit. For the operational assessment (OA), the difference in the leakage between the allowable leakage and the accident induced leakage from sources other than the tubesheet expansion region will be divided by 2.49 and compared to the observed operational leakage.The probability of a SLB is unaffected by the potential failure of a steam generator tube as the failure of the tube is not an initiator for a SLB event. SLB 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 Serial No: 12-203 Docket No. 50-423 Enclosure 1, Page 17 of 22 (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 ratetis 150 gpd (0.1 gpm) through any one steam generator.

Consequently, there is significant margin between accident leakage and allowable operational 1eakage. The SLB/FLB leak rate ratio is only 2.49 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.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

1) and RG 1.121, "Bases for Plugging Degraded PWR Steam Generator Tubes" (Reference 25), 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 Nuclear Regulatory Commission 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 (ASME) Code.For axially oriented cracking located within the tubesheet, tube burst is precluded due to the presence of the tubesheet.

For circumferentially oriented cracking, the H* analysis, documented in Section 4.0 of this enclosure, defines a length of degradation free Serial No: 12-203 Docket No. 50-423 Enclosure 1, Page 18 of 22 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 Precedents The proposed changes to MPS3 TSs 6.8.4.g and 6.9.1.7 are similar to the following approved and pending LARs for permanent alternate repair criteria: " Duke Energy Letter to NRC, 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. ML1 1188A1 07) -Approved (see ADAMS Accession No. ML1 2054A692)" Virginia Electric and Power Company (Dominion)

Letter Serial No.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)

-Pending 6.4 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.

Serial No: 12-203 Docket No. 50-423 Enclosure 1, Page 19 of 22 7.0 Environmental Considerations DNC 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, and would change an inspection or surveillance requirement.

However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendments meet the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

8.0 References

1. NEI 97-06, Rev. 3, "Steam Generator Program Guidelines," dated January 2011.2. EPRI 1013706, "Steam Generator Management Program: Pressurized Water Reactor Steam Generator Examination Guidelines." 3. EPRI 1019038, "Steam Generator Management Program: Steam Generator Integrity Assessment Guidelines." 4. ER-AP-SGP-101, Revision 5, "Steam Generator Program." 5. NRC Information Notice (IN) 2005-09, "Indications in Thermally Treated Alloy 600 Steam Generator Tubes and Tube-to-Tubesheet Welds," dated April 7, 2005.6. NRC Generic Letter 2004-01, "Requirements for Steam Generator Tube Inspections," dated August 30, 2004.7. NRC letter "Millstone Power Station, Unit No. 3 -Issuance of Amendment Regarding Changes to Technical Specification (TS) Section 6.8.4.g, "Steam Generator Program" and Section 6.9.1.7, "Steam Generator Tube Inspection Report" (TAC No. MD8736)," dated September 30, 2008. (ADAMS Accession No. ML082321292)

8. NRC letter "Millstone Power Station, Unit No. 3 -Issuance of Amendment Re: Changes to the Steam Generator Inspection Scope and Repair Requirements (TAC No. ME2978)," dated May 3, 2010. (ADAMS Accession No. ML1 00770358)9. Westinghouse Electric Company LLC, LTR-SGMP-09-100 P-Attachment, Response to NRC Request for Additional Information on H*; Model F and Model D5 Steam Generators, dated August 12, 2009.10. Westinghouse Electric Company LLC, LTR-SGMP-09-109 P-Attach'ment, Response to NRC Request for Additional Information on H*; RAI #4; Model F and Model D5 Steam'Generators, dated August 25, 2009.

Serial No: 12-203 Docket No. 50-423 Enclosure 1, Page 20 of 22 11. NRC Letter, "Vogtle Electric Generating Plant, Units 1 and 2 -Transmittal of Unresolved Issues Regarding Permanent Alternate Repair Criteria for Steam Generators (TAC Nos. ME1 339 and ME1 340)," dated November 23, 2009. (ADAMS Accession No. ML093030490)

12. DNC letter Serial No.09-525, "Millstone Power Station Unit 3, License Amendment Request to Revise Technical Specification (TS) 6.8.4.g, "Steam Generator (SG)Program," and TS 6.9.1.7, "Steam Generator Tube Inspection Report" for One-Time Alternate Repair Criteria (H*)," dated November 23, 2009. (ADAMS Accession No.ML093620085)

13. DNC Letter 10-276, "Millstone Power Station Unit 3, License Amendment Request to Revise Technical Specification (TS) 6.8.4.g, "Steam Generator (SG) Program," and TS 6.9.1.7, "Steam Generator Tube Inspection Report" for One-Time Alternate Repair Criteria (H*) -Submittal of Supplemental Information," dated April 26, 2010 (ADAMS Accession No. ML101190416)

14. Westinghouse Electric Company LLC, LTR-SGMP-09-1 11 P-Attachment, Rev. 1,"Acceptable Value of the Location of the Bottom of the Expansion Transition (BET) for Implementation of H*," dated September 2010.15. Westinghouse letter LTR-NRC-1 0-69, "Submittal of LTR-SGMP-09-1 11 P-Attachment, Rev. 1 and LTR-SGMP-09-1 11 NP-Attachment, Rev. 1, "Acceptable Value of the Location of the Bottom of the Expansion Transition (BET) for Implementation of H*," (Proprietary/Non-Proprietary) for Review and Approval," dated November 10, 2010.16. NRC letter, "Request for Withholding Information from Public Disclosure for Millstone Power Station, Unit No. 3 (TAC No. ME2978)," dated September 22, 2010 (ADAMS Accession No. ML102160749).

17. DNC letter 10-579A, Millstone Power Station Unit 3, "Resubmittal of Westinghouse Documentation Supporting License Amendment 249, "Changes to the Steam Generator Inspection Scope and Repair Requirements (TAC No. ME2978),"" dated October 7, 2010 (ADAMS Accession No. ML1 02850435).

18. Westinghouse Electric Company LLC, WCAP-17071-P, Rev. 2, "H*: Alternate Repair Criteria for the Tubesheet Expansion Region in Steam Generators with Hydraulically Expanded Tubes (Model F)," dated September 2010.19. Westinghouse Electric Company LLC, LTR-SGMP-09-100 P-Attachment, Rev. 1,"Response to NRC Request for Additional Information on H*; Model F and Model D5 Steam Generators," dated September 7, 2010.20. Westinghouse Electric Company LLC, 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*," dated September 7, 2010.

Serial No: 12-203 Docket No. 50-423 Enclosure 1, Page 21 of 22 21. Westinghouse Electric Company LLC, LTR-SGMP-10-33 P-Attachment, "H*: Response to NRC Questions Regarding Tubesheet Bore Eccentricity," dated September 2010.22. Westinghouse letter LTR-NRC-10-68, "Submittal of LTR-SGMP-10-78 P-Attachment and LTR-SGMP-1 0-78 NP-Attachment, "Effects of Tubesheet Bore Eccentricity and Dilation on Tube-to-Tubesheet Contact Pressure and Their Relative Importance to H*," (Proprietary/Non-Proprietary) for Review and Approval," dated November 9, 2010.23. Westinghouse letter LTR-NRC-10-70, "Submittal of LTR-SGMP-10-33 P-Attachment and LTR-SGMP-10-33 NP Attachment, "H*: Response to NRC Questions Regarding Tubesheet Bore Eccentricity," (Proprietary/Non-Proprietary) for Review and Approval," dated November 11, 2010.24. Westinghouse Electric Company LLC, WCAP-17330-P, Rev. 1, "H*: Resolution of NRC Technical Issue Regarding Tubesheet Bore Eccentricity (Model F/Model D5), June 2011.25. Regulatory Guide 1.121, "Bases for Plugging Degraded PWR Steam Generator Tubes," dated August 1976, (ADAMS Accession No. ML003739366).

26. Westinghouse Electric Company LLC, LTR-SGMP-09-104-P Attachment, Rev. 1,"White Paper on Probabilistic Assessment of H*," dated August 13, 2009.27. ER-AP-SGP-1 02, Revision 3, "Steam Generator Degradation Assessment." 28. ER-AP-SGP-103, Revision 3, "Steam Generator Condition Monitoring and Operational Assessments." 29. NRC letter "Millstone Power Station, Unit No. 3 -Issuance of Amendment Re: Steam Generator Tube Inspection Alternate Repair Criteria (TAC No. ME5389)," dated October 7, 2011 (ADAMS Accession No. ML1 12580517)30. Duke Energy Letter to NRC, 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, dated June 30, 2011 (ADAMS Accession No. ML11188A107)

31. Duke Energy Letter to NRC, 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," dated July 11, 2011. (ADAMS Accession No. ML1 11 95A067)32. E-mail from USNRC (Andrew Johnson) to Duke Energy (Jon Thompson) transmitting NRC letter, "Catawba Nuclear Station Unit 2, Request for Additional Information Serial No: 12-203 Docket No. 50-423 Enclosure 1, Page 22 of 22 Regarding the Steam Generator License Amendment Request to Revise Technical Specification for Permanent Alternate Repair Criteria," dated November 15, 2011.33. Duke Energy Letter to NRC, 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," dated January 12, 2012. (ADAMS Accession No. ML12019A250)

34. Virginia Electric and Power Company (Dominion)

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

35. 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 Specification for Permanent Alternate Repair Criteria (TAC NOS. ME6803 and ME6804)," dated January 18, 2012. (ADAMS Accession No. ML12006AO01)

36. Virginia Electric and Power Company (Dominion)

Letter Serial No.12-028, "Surry Power Station, Unit Nos. 1 and 2 -Response to Request for Additional Information Related to License Amendment Request for Permanent Alternate Repair Criteria for Steam Generator Tube Inspections and Repair," dated February 14, 2012. (ADAMS Accession No. ML12048A676)

37. Westinghouse Electric Company LLC, LTR-SGMP-1 1-58, "WCAP-1 7330-P, Revision 1 Erratum," dated July 6, 2011.38. NRC Letter, "Vogtle Electric Generating Plant, Units 1 and 2 -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)," dated March 28, 2011.(ADAMS Accession No. ML110660648)

39. 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)," dated May 26,'2011. (ADAMS Accession No.ML11140A099)

40. 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," dated June 20, 2011, (ADAMS Accession No.ML111721903)

Serial No: 12-203 Docket No. 50-423 Enclosure 2 Mark-Up of Proposed Technical Specification Changes DOMINION NUCLEAR CONNECTICUT, INC.MILLSTONE POWER STATION UNIT 3 I For Infor May 31, 2007 ADMINISTRATIVE CONTROLS g. Steam Generator (SG) Program A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained.

In addition, the Steam Generator Program shall include the following provisions:.

a. Provisions for condition monitoring assessments:

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

SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.1. Structural integrity performance criterion:

All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down and all anticipated transients, included in the design specification) and design basis accidents.

This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials.

Apart from the above requirements, additional loading conditions associated with the design basis accidents, or a combination of accidents in accordance 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 secondary loads.2. Accident induced leakage performance criterion:

The primary'to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SGL MILLSTONE

-UNIT 3 6-17a Amendment No. 238 g"ct÷br 7, 2011 ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)

Leakage is not to exceed 500 gpd per SG.3. The operational LEAKAGE performance criterion is specified in RCS LCO 3.4.6.2, "Operational LEAKAGE." c. Provisions for SG tube repair criteria:

Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40%of the nominal tube wall thickness shall be plugged.The following alternate tube repair criteria shall be applied as an alternative to the 40% depth-based criteria: 1. For Refuzeling Out.ag. 4, and the subsequent

.p.. .ting ; y..., tubes with service-induced flaws located greater than 15.2 n_.X 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.

MILLSTONE

-UNIT 3 6-17b Amendment No. 238,8-245,49, 252 Oeteber 7, 20 141 ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (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 rt1ikb-to-tubesheet weld at the tube outlet, and that may satisfy the*air criteria.

For Refucling Outage 14 and the b-aortions 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.

I. 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 non-destructive testing, or engineering evaluation indicates that a crack-like indication is not MILLSTONE

-UNIT 3 6-17c Amendment No. 69, 4-86, 24-2, 2348, 243,-245 249, 2-5-2 "7, 2011 ADMINISTRATIVE CONTROLS 6.9.1.6.c The core operating limits shall be determined so that all applicable limits (e.g. fuel'thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as SHUTDOWN MARGIN, and transient and accident analysis limits) of the safety analysis are met.6.9.1.6.d The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

STEAM GENERATOR TUBE INSPECTION REPORT 6.9.1.7 A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with TS 6.8.4.g, Steam Generator (SG)Program. The report shall include: a. The scope of inspections performed on each SQ b. Active degradation mechanisms found, c. Nondestructive examination techniques utilized for each degradation mechanism, d. Location, orientation (if linear), and measured sizes (if available) of service induced indications, e. Number of tubes plugged during the inspection outage for each active degradation mechanism, f. Total number and percentage of tubes plugged to date, g. The results of condition monitoring, including the results of tube pulls and in-situ testing, h. The effective plugging percentage for all plugging in each S D...uring Ref..eling Outage 14 and subsequent

.p.r.a.ing Iche primary to secondary LEAKAGE rate observed in each SG (if it is not practical to assign the LEAKAGE to an individual SQ 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, MILLSTONE

-UNIT 3 6-21 Amendment No. -24, 40, -50, 69, 4-04, 1-7-, 24, 24-5,2-29, 2-3-8, 245,-249.2-Oetetber 7, 2011 ADMINISTRATIVE CONTROLS* STEAM GENERATOR TUBE INSPECTION REPORT (Continued)

j. .r. , .ing Refuling Outage 14 and the subsequent operating

., 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 2.49 times the maximum operational primary to secondary leakage rate, the report should describe how it was determined; and k. Durfing Refueling Outage 11 and the subsut cptg I,, he results of monitoring for tube axial displacement (slippage).

If slippage is discovered, the implications of the discovery and corrective action shall be provided.SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the U.S. Nuclear Regulatory Commission, Document Control Desk, Washington, D.C. 20555, one copy to the Regional Administrator Region I, and one copy to the NRC Resident Inspector, within the time period specified for each report.6.10 Deleted.6.11 RADIATION PROTECTION PROGRAM.6.11.1 Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained, and adhered to for all operations involving personnel radiation exposure.6.12 HIGH RADIATION AREA A-As provided in paragraph 20.1601(c) of 10 CFR Part 20, the following controls shall be applied to high radiation areas in place of the controls required by paragraph 20.1601 (a) and (b) of 10 CFR Part 20: MILLSTONE

-UNIT 3 6-21 a Amendment No. 2--3, 245, 249, 2-65-2 Serial No: 12-203 Docket No. 50-423 Enclosure 4 Westinghouse Electric Company LLC LTR-SGMMP-11-28 Rev. 1 NP-Attachment, "Response to USNRC Request for Additional Information Regarding the License Amendment Requests for Permanent Application of the Alternate Repair Criterion, H*, to the Model D5 and Model F SGs" February 2, 2012 DOMINION NUCLEAR CONNECTICUT, INC.MILLSTONE POWER STATION UNIT 3 Westinghouse Non-Proprietary Class 3 LTR-SGMMP-1 1-28 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. ML1 12150144.4., 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

LTR-SGMMP-11-28 Rev. I 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-17330-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 correspond to 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 represent the top 10% of the Monte Carlo rank order results. This being the case, why does Figure 3-32 in Revision I properly represent the top 10% of the Monte Carlo rank order results? Why are the minimum H* values in Figure 3-36 of Revision I substantially different from those in Figure 3-32 of Revision 0?Response: This response apolies for both the Model D5 and the Model F SGs.The footnote on page 3-53 of WCAP-1 7330-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 (aL), 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-17330-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-1 1-28 Rev. I NP-Attachment prior work that the maximum values of H* are likely to occur at positive variations of tubesheet CTE and negative variations of tube CTE. In the Monte Carlo analysis, described further in the response to Question 3, approximately half of the H* values include a negative variation of tubesheet CTE and a corresponding large negative variation of tube CTE;however, the frequency of occurrence in the rank order range of interest is low As noted above, the probabilistic response surface is presented in terms of the combined variable cx, the square root of the sum of the squares of the individual tube and tubesheet (TS) CTE components.

The RSS combination of tube and tubesheet variables negates the sign of the negative variation of both the tube and TS CTE and artificially inflates the value ofresulting 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-17330-P, Revision 0 -Provide copy of the "response surface" (i.e., H*relationship to coefficients of thermal expansion (CTE) variability for 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 radial location 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."

LTR-SGMMP-11-28 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 _/_______________

S LTR-SGMMP-1 1-28 Rev. 1 NP-Attachment 40 41 42 43 44 45 a~c~e Question 3: WCAP-1 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 particular Monte 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-17345-P, Revision 2, Section 3.4' Confirm that the Monte Carlo analyses performed for the Model 51F SGs using the thick shell model are based upon sampling of the full H*/CTE response surfaces in Figure 8-5 of WCAP 17092 Rev 0. If this is incorrect, and only a "reduced" response surface is used, explain how the reduced response surfaces are used in the Monte Carlo analysis.

If for a particular Monte Carlo iteration a negative variation of tubesheet CTE is randomly generated, what is done with this value (e.g., is tubesheet CTE assumed to have nominal value)? Why doesn't the use of a reduced response surface bias the rank ordering above 90% in the non-conservative direction?

Response: Model D5 Table 3-1 provides the data for the requested response surface for the Model D5 SGs at the critical tubesheet radius of [ ]a,c,e inches. Note that the change in the maximum value of H* (see Case 45) at the critical radius of [ ]ace inches from the prior critical radius of 26.703 inches shown in the response to Question 2 is only 0.03 inch.The utilization of a reduced response surface as shown in Tables 2-1 and 3-1 does not bias the rank ordering in a non-conservative direction; it simply limits the effort to develop a response surface to the region in parameter space where the limiting values of H* are most 6 LTR-SGMMP-11-28 Rev. I NP-Attachment likely located. The interpolation method for the reduced response surface permits calculation of H* values with the thick-shell equation, which is the underlying calculation basis of the response surface. The Monte Carlo process randomly samples, including variances in the region excluded from the reduced response surface by means of the interpolation scheme.In approximately.

half of the cases, the sampling results have negative tubesheet CTEs.Because the ultimate objective is to define specific combinations of tubesheet and tube CTEs that represent a specific rank order of H* values for input to the C2 model, the salient question is how points with negative tubesheet CTEs are treated in the probabilistic calculation of H* using the C2 model.Each of the 10,000 simulations in the general Monte Carlo procedure uses the following process: 1. Pick a random normal deviate to represent the tubesheet CTE variation.

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

3. For each tube, assign an H* value corresponding to the current tubesheet CTE variation and the tube's CTE variation by interpolating an H* value on the response surface. If the tubesheet CTE variation is negative, interpolate as though the tubesheet CTE variation is zero (i.e., mean value).4. Apply sector ratios as discussed in LTR-SGMP-09-100 P Attachment, Rev. 1.5. Store the largest H* value along with the corresponding tube and tubesheet CTE variations.

Note that negative tubesheet CTE variations are retained, although the H* assigned to them is conservative by step 3.Steps 1-5 represent one iteration of the Monte Carlo process. This process is repeated 10,000 times, and the results sorted in ascending order by H* value.Step 3 of the process slightly distorts the rank order of the H* values because artificially higher values of H* are assigned to the combination of randomly selected CTEs when the selected tubesheet CTE is negative.

The true H* rank order of these cases is lower than the apparent value of H* for these cases. The effect is to displace the rank order of H*s with positive values of tubesheet CTE to lower positions in the H* vector.The manner in which these values are used in the subsequent step of the H* calculation process with the 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-1 1-28 Rev. I NP-Attachment Model F The Monte Carlo sampling for the Model F steam generators is based on sampling the full H*/CTE response surfaces in Figure 8-5 of WCAP 17071-P, which is based on application of the thick-shell model.The Monte Carlo process randomly samples from the response surface by means of an interpolation scheme. In approximately half of the cases, the sampling results have negative tubesheet CTEs. Because the ultimate objective is to define specific combinations of tubesheet and tube CTEs that represent a specific rank order of H* values for input to the C2 model, the salient question is how points with negative tubesheet CTEs are treated in the probabilistic calculation of H* using the C2 model.Each of the 10,000 simulations in the general Monte Carlo procedure uses the following process: 1. Pick a random normal deviate to represent the tubesheet CTE variation.

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

3. For each tube, assign an H* value corresponding to the current tubesheet CTE variation and the tube's CTE variation by interpolating an H* value on the response surface. 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 9500th 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.

The same process is utilized when determining the H* value for the higher probabilistic goals applicable to the Model F, that is, the 95/95 whole plant value of H*.8 LTR-SGMMP-11-28 Rev. I NP-Attachment Table 3-1 Reduced Response Surface; Model D5, []a~c~e inches Radius TS CTE T CTE H*+BET Case #n a n c (in)Q ]aeRadius)

_ a,c,e 2 3 4 5 6 7 8 9 10 ____11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 34 35 36 37 38 _.9 LTR-SGMMP-1 1-28 Rev. I NP-Attachment

, 39 40 41 42 ____L43 445___ ____ _________a,c,e 10 LTR-SGMMP-1 1-28 Rev. I NP-Attachment Question 4: WCAP-I 7330-P, Revision 1, Table 3-28 -Provide a similar table applicable to the Model D5 SLB case, from the 9526 to 9546 rank orders.Response The question is Model D5-specific and does not appIV 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 Tube Tubesheet Alpha(')CTE CTE 9526 F 9527 9528 9529 9530 9531 9532 9533 9534 9535 9536 9537 9538 9539 9540 9541 9542 9543 9544 9545 9546 Notes: 1. Defined as SQRT((Tube CTE)A2 + (Tubesheet CTE)A2)a,c,e 1I LTR-SGMMP-1 1-28 Rev. I NP-Attachment Question 5: WCAP- I 7330-P, Revision 1, Table 3-29 -Provide C2 H* values for rank orders 9888 and 9892. This will lend additional confidence to inferences drawn from this table on page 3-58. In addition, provide a similar table applicable to the Model D5 SLB case.Response: This response applies for both the Model D5 and Model F SGs.Analysis code note: The structural code employed for the prior H* calculations was ANSYS Workbench, Version 11. Version 12.1 of ANSYS Workbench was released following the issue of WCAP-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 [ ]ac"e inch radius for rank orders from 9533 through 9539.Table 5-1: Model F NOP Results at [ ]a~c~e inches Variation Input I MC T CTE TS CTE C 2 H*# no mo in.9888 [ ]a,c,e [ ]a,c,e [ ]a,c,e 9892 [ ]a,c,e [ ]a,c,e [ ]a,c,e 12 LTR-SGMMP-1 1-28 Rev. 1 NP-Attachment Table 5-2: Model D5 SLB Results at []a,c,e inches Although the uncertainty in the narrow range of rank order H* values for the Model D5 (Table 5-2) is slightly larger than the uncertainty for the Model F (Table 5-1 and Table 3-29 of WCAP-17330-P, Rev. 1), the inferences drawn from these data on page 3-56 of WCAP-17330-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-I 1-28 Rev. I NP-Attachment Question 6: WCAP-I 7330-P, Revision 1, Figure 3-45 -Should the data corresponding to the two open symbols be labeled as "data used in probabilistic analysis" (consistent with Figure 3-44) instead of "reduced data?" Why does this figure show only two open symbols rather than three as are given in Figure 3-44?Response The question is specific to the Model D5 SGs and does not apply for the Model F SGs. This question was not included in Reference 4 for the Model 51F SGs.For clarity, the two (three) open symbols on Figure 3-45 of WCAP-17330-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 1 Rank H* Tube CTE Tubesheet Alpha_____ ____ _____ _____ ____ CTE _ _ _ _ _9149 [ ]a,c,e [ ]a,c,e [ ]ae 3.513 9500 [ ]a,c,e [ ]a,c,e [ ]a,c,e 3.750 9536 [ ]a,c,e [ ]a,c,e [ ]a,c,e 3.733 14 LTR-SGMMP-1 1-28 Rev. 1 NP-Attachment a,c,e Figure 6-1 Update of Figure 3-45 of WCAP-17330, Revision I 15 LTR-SGMMP-1 1-28 Rev. 1 NP-Attachment Question 7: WCAP-I 7330-P, Revision 1, Tables 3-35 to 3-48 -The numerical methods used to generate the accumulated pullout loads in these tables appear to contain two sources of non-conservatism.

One, the distance below the top of the tubesheet (TTS) where the contact pressure transitions 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 realistic and more conservative estimate of the contact pressure zero intercept value can be obtained by extrapolating the C2 results at lower elevations to the zero intercept location.

Two, the method used to interpolate the H* distance between specific locations where C2 analyses were performed assumes that the distribution of contact pressure between these locations is a constant value equal to average value between these locations.

For Table 3-35, the staff estimates that elimination of the non-conservatisms increases the calculated H* by 0.34 inches. For Tables 3-46 and 3-48, H* increases by 0. 15 inches. These are not trivial differences.

The staff estimates that the pullout loads corresponding to the H* distances in Figures 3-35, 3-46, and 3-48 are overestimated by 17%, 6%, and 8%, respectively.

Provide revisions to Tables 3-35 to 3-48, if and as needed, to address the 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-1 7330-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-17330-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-17330-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-11-87 is Reference 3-17 in WCAP-17330-P, Revision 1 and is provided as Appendix A in this document.)

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

Values are listed at 300° and 600', 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-1 1-28 Rev. 1 NP-Attachment Table 7-3 summarizes the H* values contained in WCAP-17330, Revision 1 for the Model D5 and Model F SGs and serves to provide the input for Table 7-4.Table 7-4 shows the effects of applying the improved CTE variability values to the H*analysis.

Note that the H* values in Table 7-4 do not include crevice pressure or Poisson contraction because neither of these are related to CTE. As can be seen from Table 7-4, the existing H* length for the Model F's is conservative by approximately

[ ]ace inches and the H*length for the Model D5's is conservative by about [ ]a,c,e inch. This shows that the conservatism inherent in the current H* calculations are adequately conservative to account for small differences in judgment on the calculation process even without considering the major conservatisms identified previously (i.e., neglecting residual contact pressure).

Additional conservatism to further support this conclusion is identified below.Table 7-1 CTE Values Without Instrumentation Error Tube CTE SDs, %Temperature As Used in Improved 50% Improved 95%(*F) WCAP- Imrvd0% Ipoe95 13,e 1CAP Confidence Confidence 17330,Rev.

1 300 2.33 [ ]a,c,e [ ]a,c,e 600 2.33 ]a,c,e [ ]a,c,e Tubesheet CUE SDs, %Temperature As Used in Improved 50% Improved 95%(°F) WCAP-17330,Rev.

1 Confidence Confidence 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. I NP-Attachment Table 7-2 Ratio of CTE Variances (Refined/Used in Current H*)Tube CTE SDs Ratios Temperature 50%(*F) 50e 95% Confidence

___________

Confidence

________30 0O [ ] ,c,e T a,c,e 600 x[ ],c,e a,c,e Table 7-3 Summary of H* Lengths from WCAP-17330, Revision 1 Limiting G Mean H* Probabilistic H* Difference, Ratio f Model/Case (inches) (inches) Probabilistic

-Mean Ratio from Table 7-2 F, 95/50 Whole ,ce Bundle F, 95/95 Whole Plant D5, 95/50 Whole Bundle D5, 95/95 Whole Bundle L Table 7-4 Estimate of Conservatism of H* Length Related to CTE Variance T Difference Difference x Model/Case Limiting Ratio New Probabilistic H* (Licensed H* -New L gProbabilistic H*) ,c'e F, 95/50 Whole Bundle F, 95/95 Whole Plant D5, 95/50 Whole Bundle D5, 95/95 Whole Bundle __1 19 LTR-SGMMP-1 1-28 Rev. 1 NP-Attachment Question 8: WCAP-I 7330-P, Revision 1, Figures 3-48 and 3-49 -These figures were generated with the thick shell model. Were "spot checks" 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-1 7330-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-17072-P.

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

The basis of this conclusion is explained below.As discussed in WCAP-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-11-28 Rev. 1 NP-Attachment Figure 8-1 (a) shows a comparison of the unmodified crevice pressure distribution used in the C2 analysis (i.e., the crevice pressure is distributed over the full depth of the tubesheet) and the crevice pressure distribution that has been adjusted to reflect the final contact pressure distribution reported in Table 3-48 in WCAP-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-1 1-28 Rev. 1 NP-Attachment WCAP-17330-P, Revision 1, but the crevice pressure differential is held constant at 1 psi throughout the depth of the tubesheet (i.e., primary pressure in the full length of the crevice), the result is the "DP=1 psi" curve in Figures 8-3(a and b). Similarly, if the C 2 model inputs are kept the same, but the crevice pressure differential is held constant at 2560 psid for the Model D5 throughout the depth of the tubesheet (i.e., secondary pressure in the crevice), the result is the "DP=2560 psi" curve in Figure 8-3 (a). Likewise, if the C 2 model inputs are kept the same, but the crevice pressure differential is held constant at 1453 psid for the Model F throughout the depth of the tubesheet (i.e., secondary pressure in the crevice), the result is the "DP=1453 psi" curve in Figure 8-3 (b).These are the bounding conditions for crevice pressure.

It is not possible for variation in crevice pressure differential to produce a contact pressure distribution less than, or greater than, the space bounded by these two curves. The current probabilistic contact pressure distribution, with the unmodified crevice pressure differential, is also shown on Figures 8-3 (a and b) for the Model D5 and the Model F SGs, respectively.

The difference between the contact pressure distribution with the unmodified crevice pressure distribution used in WCAP-1 7330-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 ac,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 ace 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-1 1-28 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 potential non-conservatisms in the H* estimate discussed in Question 7 above, there is uncertainty associated with the computed probabilistic H* values calculated with the C2 model as illustrated in Table 3-29. Depending on the response to question 8 above, there also may be some uncertainty associated with the H*adjustments for the crevice pressure distribution.

What change to the proposed H* value of 14.01 inches is needed to ensure that it is a conservative value?Response: The responses to RAI 7 and RAI 8 indicate that no adjustments to the Model D5 and Model F probabilistic H* estimates are necessary to account for the uncertainty associated with the C2 model results shown in Table 3-29 of WCAP-1 7330-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 adoustments, 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(b), is [ ]a,c,e inches.Table 9-1 and Table 9-2 summarize the adjustments to the probabilistic H* estimate compared to the adjustments that are demonstrated above in the current technical basis for H*. It is seen from Table 9-1 that a margin of [ ]a~c~e inches exists in the currently recommended H* length of 14.01 inches for the Model D5 SGs when the conservatism in the crevice pressure adjustment and the measurement error in the CTE data are quantified and the proper adjustments are made. Table 9-2 shows that a margin of [ ]a~c,e exists in the currently recommended H* length of 15.21 inches for the Model F when the conservatism in the crevice pressure adjustment and the measurement error in the CTE data are quantified and the proper adjustments are made. These previously un-quantified conservatisms significantly exceed the potential increase in the H* length if different judgments are made in the details of the H* calculation as suggested in Questions 7, 8 and 9. Based on this, it is concluded that no adjustments to the recommended probabilistic H* value of 14.01 inches for the Model D5 SGs and 15.21 inches for the Model F SGs are necessary and that the H*lengths recommended in WCAP-17330-P, Revision 1 are significantly conservative.

27 LTR-SGMMP-1 1-28 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-Unmodified H* Value Adjustments Poisson Correction Crevice Pressure and BET Adjustment CTE Uncertainty Adjustment (RAI 7)Total Adjustments Final Probabilistic H* 14.01 [ ]a,c,e Notes: (1) Recalculated for [ ]a,c,e inches H* based on Figure 8-2(a).(2) Crevice pressure margin ([ ]a,c,e inch) plus BET adder of 0.3 inch included in Pcrev correction (Figure 3-49 of WCAP-17330, Rev. 1)(3) See response to Question 7.Table RAI 9-2 Conservatism in Current Model F H* Calculation WCAP-17330-P, Refined Source Rev 1 Calculations in in ,c,e Unmodified H* Value Adjustments Poisson Correction Crevice Pressure and BET Adjustment CTE Uncertainty Adjustment (RAI 7)Total Adjustments Final Probabilistic H* 15.21 [ ]a,c,e Notes: (1) Recalculated for [ ]a,c,e inches H* based on Figure 8-2(b).(2) Crevice pressure margin ([ ]a,c,e inch) plus BET adder of 0.3 inch included in Pcrev correction (Figure 3-48 of WCAP-17330, Rev. 1)(3) See response to Question 7.28 LTR-SGMMP-1 1-28 Rev. 1 NP-Attachment Question 10: Westinghouse letter LTR-SGMP-10-95 P- Attachment, Revision I -The staff is able to reasonably reproduce the numbers in Table 5 for Exp-2 and. Power-2. It is the staffs understanding that Table 4 contains intermediate results leading to the results in Table 5. However, the staff cannot reproduce the numbers in Table 4 based on the information provided.

Is Table 4 correctly titled? Provide a precise definition of the parameters that are listed in Table 4. Provide one example of how the parameter values 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-10-95, 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) 6alculated 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 = 12[tKI, but neglecting the constant because it divides out in the calculation of the resistance ratios.29 LTR-SGMMP-1 1-28 Rev. 1 NP-Attachment Table 10-1 Local Loss Coefficient for Power 2 (K=0.15*(Pc) 4.s)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 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.0000E+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 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 1.0516E+12 9.0654E+13 1.00 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 4.0011E+11 1.2318E+14 1.00 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 6.2667E+11 2.0023E+14 30 LTR-SGMMP-1 1-28 Rev. I NP-Attachment Table 10-2 Segment Resistance Based on Viscosity in (Ibf-seclinA2)

Units for Power 2 (K=0.15*(Pc) 4"s)Segment Tubesheet Radius Lengths 4.437 10.431 18.139 26.703 42.974 49.825 from BTS to TTS Normal Operating Conditions 2.00 1.19E+08 8.55E+07 5.49E+07 2.94E+07 2.48E+06 3.64E+05 2.00 7.13E+07 5.07E+07 3.17E+07 1.81E+07 2.24E+06 5.67E+05 2.00 3.86E+07 2.60E+07 1.69E+07 1.00E+07 2.12E+06 8.39E+05 4.515 2.62E+07 1.56E+07 1.11E+07 7.33E+06 4.11E+06 3.86E+06 6.386 1.31E+06 5.58E+05 5.02E+05 6.19E+05 3.81E+06 1.32E+07 2.129 1.51E+02 2.28E+01 4.63E+01 1.20E+03 7.63E+05 6.82E+06 1.00 3.36E-08 6.09E-10 1.44E-10 7.71E-01 4.79E+04 1.17E+06 1.00 0.OOE+00 0.00E+00 0.OOE+00 0.OOE+00 9.70E+01 4.31E+04 Steam Line Break Conditions 2.00 3.06E+09 2.69E+09 1.90E+09 1.10E+09 1.21E+08 1.26E+07 2.00 1.39E+09 1.24E+09 8.82E+08 5.11E+08 6.63E+07 9.99E+06 2.00 5.31E+08 4.87E+08 3.50E+08 2.07E+08 3.41E+07 7.81E+06 4.515 1.27E+08 1.31E+08 9.73E+07 6.59E+07 2.15E+07 1.12E+07 6.386 5.47E+05 7.08E+05 5.61E+05 4.91E+05 2.74E+06 1.31E+07 2.129 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 6.13E+04 5.29E+06 1.00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 1.10E+04 3.37E+06 1.00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 1.72E+04 5.48E+06 31 LTR-SGMMP-I 1-28 Rev. I NP-Attachment Question 11 Westinghouse letter LTR-SGMP-10-95 P -Attachment, Revision 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 1*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*PcA3 Power-2 K = 0.15*(Pc)4.5 Diagonal Bound Logarithmic K =1 E+12*ln(Pc)+4E+08 Question 12 This question is a utility-specific question for which the respective utilities provide specific responses.

Question 13 This question was a Catawba-2

-specific and does not apply to either the Model 51F or the Model F SGs.32 LTR-SGMMP-1 1-28 Rev. I NP-Attachment Question 14 44P.A.P 47345 R, R:W.ia.n 2, Tob!cs 3 50 2.d 3 5r WCAP-17330-P, Revision I Table 3-50-Are@ Is the footnotes in teoe bles this table correct and complete?

For Model W4F, Table 3-27 implies we have direct C 2 calculations for rank orders 9025, 6-73-7, and m00019186, 9694 and 9890. Thus, for Table 3-6450, it seems a# three of four cases are based on interpolated values. im forea.de!

4. 4F, l 27 2;p!* .....wa I Ain .^ -d,.,aa for C-n o:r:915 0158, 969A7, n_ 0760. Thus, for Tab!a 2 50, it aems aa'.!l, thc ".ha!c p!ant, 05to 5" Oma. Mct C.. 2 ca!c-!at. .nd the ether araea am !ntc;pa.!atd

'!'-... If the staffs understanding is incorrect, clarify for which rank orders direct C 2 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 C 2 model are shown on Figure 3-43 for the Model F SGs. The specific rank orders are identified in Table 3-30 of WCAP-17330-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 C 2 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-1 1-28 Rev. 1 NP-Attachment To: G. W. Whiteman B. J. Bedont C. D. Cassino Date: May 5, 2011 cc: From: Ext: Fax: A. 0. Roslund 724-722-6473 724-722-5889 Your ref: Our ref: LTR-SGDA-11-87

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

H. 0. Lagally 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-1 1-28 Rev. 1 NP-Attachment Introduction The calculation of H* at high probability and confidence in Reference I 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 I through Table 4. The tube data are in Table 5 through Table 7. In order to determine the instrumentation error, one specimen each of the tube and tubesheet material was run ten times. These results are shown in Table 8 and Table 9.Best estimate (50% confidence) standard deviations were calculated from the standard formula, 0n-1 High confidence (95%) standard deviations are obtained by the standard Chi-Squared adjustment:

, 9 5-=c0 5 0 2n -i Xn- 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-l 1-28 Rev. I NP-Attachment standard formula below was used to calculate a high confidence standard deviation for the tube and tubesheet without instrumentation error: i 5 95,Material jf95,total 95,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. I NP-Attachment Table 1 Tubesheet CTEs (ptin / in IF)Temp (TF) 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 _____a,c,e 38 LTR-SGMMP-11-28 Rev. I NP-Attachment Table 15 Tubesheet CTEs (Itin / in IF)Temp (°F) Sample 11 Sample 12 Sample 13 Sample 14 Sample 15 Sample 16 Sample 17 Sample 18 Sample 19 Sample 20 100 F-125 150 175 200 225 250 275 300 325 350 375 400 425 450 475 500 525 550 575 600 625 650 675 700 ac,e 39 LTR-SGMMP-1 1-28 Rev. I NP-Attachment Table 3 Tubesheet CTEs'( in / in IF)Temp (*F) Sample 21 Sample 22 Sample 23 Sample 24 Sample 25 Sample 26 Sample 27 Sample 28 Sample 29 Sample 30 100 125 150 175 200 225 250 275 300 325 350 375 400 425 450 475 500 525 550 575 600 625 650 675 7O00 a,c,e 40 LTR-SGMMP-1 1-28 Rev. I NP-Attachment Table 4 Tuihe~heet (,min / in Temp (*F) Sample 31 Sample 32 Sample 33 Sample 34 Sample 35 Sample 36 Sample 37 Sample 38 Sample 39 Sample 40 100 125 150 175 200 225 250 275 300 325 350 375 400 425 450 475 500 525 550 575 600 625 650 675 700 a,c,e 41 LTR-SGMMP-1 1-28 Rev. 1 NP-Attachment Table 5 Tube CTEs (Model F) (gin / 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 _____ _____7O0 __________

_____ _____ _____ __________

_____ _____a,c,e 42 LTR-SGMMP-1 1-28 Rev. I NP-Attachment Table 6 Tube CTEs (Model D5) (pin / in IF)Temp (°F) Sample 11 Sample 12 Sample 13 Sample 14 Sample 15 Sample 16 Sample 17 Sample 18 Sample 19 Sample 20 100 125 150 175 200 225 250 275 300 325 350 375 400 425 450 475 500 525 550 575 600 625 650 675 700 a,c,e 43 LTR-SGMMP-1 1-28 Rev. 1 NP-Attachment Table 7 Tube CTEs (Model 44FR (ltin / in °FR Temp (°F) Sample 21 Sample 22 Sample 23 Sample 24 Sample 25 Sample 26 Sample 27 Sample 28 Sample 29 Sample 30 100 125 150 175 200 225 250 275 300 325 350.375 400 425 450 475 500 525 550 575 600 625 650 675 700 a,c,e 44 LTR-SGMMP-1 1-28 Rev. I NP-Attachment Table 8 I ube L ! s (MVultiple runs on same specimen) (piunin /iF)Temp (°F) Run I Run 2 Run 3 Run 4 Run 5 Run 6 Run 7 Run 8 Run 9 Run lO 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 a,c,e 45 LTR-SGMMP-1 1-28 Rev. I NP-Attachment Table 9 Temp (*F) Run 1 Run 2 Run 3 Run 4 Run 5 Run 6 Run 7 Run 8 Run 9 Run 10 100 125 150 175 200 225 250 275 300 325 350 375 400 425 450 475 500 525 550 575 600 625 650 675 700 a,c,e 46 LTR-SGMMP-11-28 Rev. 1 NP-Attachment Table 10 Mean and Standard Deviation, Tube Material Temperature Mean Best Estimate Standard 95% Confidence Standard (°F) (pin/in°F)

Deviation

(%) Deviation

(%)100 6.95 3.40 4.35 125 7.03 2.84 3.64 150 7.10 2.38 3.04 175 7.16 2.00 2.55 200 7.23 1.69 2.16 225 7.28 1.45 1.86 250 7.34 1.27 1.63 275 7.39 1.14 1.46 300 7.43 1.05 1.35 325 7.48 0.99 1.27 350 7.52 0.95 1.21 375 7.56, 0.92 1.17 400 7.59 0.89 1.14 425 7.63 0.87 1.12 450 7.66 0.86 1.10 475 7.69 0.85 1.08 500 7.72 0.84 1.07 525 7.76 0.83 1.07 550 7.79 0.83 1.06 575 7.82 0.82 1.05 600 7.85 0.81 1.03 625 7.88 0.79 1.01 650 7.91 0.77 0.98 675 7.94 0.74 0.95 700 7.97 0.72 0.92 47 LTR-SGMMP-11-28 Rev. 1 NP-Attachment Table 11 Mean and Standard Deviation, Tubesheet Material Temperature Mean Best Estimate Standard 95% Confidence Standard (°F) (liin/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-11-28 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-1 1-28 Rev. 1 NP-Attachment Table 13 Standard Deviation for Instrumentation Error, Tubesheet Material Temperature Best Estimate Standard 95% Confidence Standard (°F) Deviation

(%) Deviation

(%)100 2.08 1.52 125 1.82 1.32 150 1.59 1.16 175 1.40 1.02 200 1.25 0.91 225 1.13 0.82 250 1.03 0.75 275 0.95 0.69 300 0.89 0.65 325 0.85 0.62 350 0.82 0.60 375 0.79 0.58 400 0.78 0.57 425 0.78 0.57 450 0.77 0.56 475 0.78 0.57 500 0.79 0.57 525 0.79 0.58 550 0.79 0.58 575 0.80 0.58 600 0.80 0.59 625 0.80 0.58 650 0.79 0.57 675 0.77 0.56 700 0.74 0.54 50 LTR-SGMMP-11-28 Rev. 1 NP-Attachment Table 14 High-Confidence Tube and Tubesheet Standard Deviations with Instrumentation Error Removed Temperature Tube Tubesheet

(%)(*F)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 a,c,e 51 Serial No: 12-203 Docket No. 50-423 Enclosure 5 Westinghouse Electric Company LLC LTR-SGMMP-1 1-28 Errata Rev. 1"LTR-SGMMP-11-28 Revision 0 and Revision 1, P- and NP- Attachment Errata," March 20, 2012 DOMINION NUCLEAR CONNECTICUT, INC.MILLSTONE POWER STATION UNIT 3 Westinghouse Non-Proprietary Class 3* Westinghouse To: D.H. Warren P.J. McDonough H. Mahdavy D.L. Rogosky L.E. Markle N. Bahtishi C.L. Mitchell D.C. Beddingfield M.W. Ryan J.J. Roberts G.R. Strussion C.W. Nitchman A.M. Mrazik S.J. Hyde J. Stepanic C. D. Cassino Date: March 20, 2012 cc: B. J. Bedont From: Ext: Fax: H.O. Lagally 724-722-5082 724-722-5889 Your ref: Our ref: LTR-SGMMP-11-28 Errata, Rev. I

Subject:

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

Reference:

1. LTR-SGMMP-I 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-1 1-28, 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 I (Reference

2) was issued to augment Revision 0 of the same letter to provide information specific to the Model F SGs in the response to the NRC RAI. References I and 2 contain both a proprietary (P) attachment and a non-proprietary (NP) attachment for the responses to the RAI.For Revision 0 of LTR-SGMMP-11-28, the following corrections apply:* On page 31 of both the P-Attachment and the NP-Attachment, the title of the Appendix A cover page should be LTR-SGDA-11-87 instead of LTR-SGMP-11-87.For Revision I of LTR-SGMMP-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 Electronically Approved*Prepared by: H. 0. Lagally Steam Generator Management And Modification Programs Electronically Approved*Electronically Approved*Verified:

G.W. Whiteman Regulatory Compliance Approved by: Damian A. Testa, Manager Steam Generator Management And Modification Programs*Electronically approved records are authenticated in the electronic document management system.Westinghouse Electric Company LLC 1000 Westinghouse Drive Cranberry Township, PA 16066© 2012 Westinghouse Electric Company LLC All Rights Reserved Serial No: 12-203 Docket No. 50-423 Enclosure 6 Westinghouse Electric Company LLC, CAW-12-3446, "Application for Withholding Proprietary Information from Public Disclosure," March 21, 2012 (Affidavit for LTR-SGMMP-11-28 Rev. 1 P-Attachment)

DOMINION NUCLEAR CONNECTICUT, INC.MILLSTONE POWER STATION UNIT 3 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: NEU-12-13 CAW-12-3446 March 21, 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-12-3446 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 Dominion Nuclear Connecticut, Inc.Correspondence with respect to the proprietary aspects of the application for withholding or the Westinghouse affidavit should reference this letter, CAW-12-3446, 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, yJi. A. Gresham, Manager Regulatory Compliance Enclosures CAW- 12-3446 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: M Gresham, Manager Regulatory Compliance Sworn to and subscribed before me this 21st day of March 2012 No~tayPublic COMMONWEALTH OF PENNSYLVANIA Notarial Seal Cynthia Olesky, Notary Public Manor Boro, Westmoreland County My Commission Expires July 16, 2014.Member. Pennsylvania Association of Notaries 2 CAW-12-3446 (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) I 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-12-3446 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- 12-3446 (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 Dominion Nuclear Connecticut, Inc. Letter and Application for Withholding Proprietary Information from Public Disclosure, to the Document Control Desk. The proprietary information as submitted by Westinghouse for Millstone Unit 3, 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-12-3446 This information is part of that which will enable Westinghouse to: (a) License the H* Alternate Repair Criteria.Further this information has substantial commercial value as follows: (a) Westinghouse plans to sell the use of the information to its customers for the purpose of licensing the H* Alternate Repair Criteria.(b) Westinghouse can sell support and defense of the H* criteria.(c) The information requested to be withheld reveals the distinguishing aspects of a methodology which was developed by Westinghouse.

Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar technical justification and licensing defense services for commercial power reactors without commensurate expenses.

Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.

The development of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Westinghouse effort and the expenditure of a considerable sum of money.In order for competitors of Westinghouse to duplicate this information, similar technical programs would have to be performed and a significant manpower effort, having the requisite talent and experience, would have to be expended.Further the deponent sayeth not.

PROPRIETARY INFORMATION NOTICE Transmitted herewith are proprietary and/or non-proprietary versions of documents furnished to the NRC in connection with requests for generic and/or plant-specific review and approval.In order to conform to the requirements.of 10 CFR 2.390 of the Commission's regulations concerning the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary versions is contained within brackets, and where the proprietary information has been deleted in the non-proprietary versions, only the brackets remain (the information that was contained within the.brackets in the proprietary versions having been deleted).

The justification for claiming the information so designated as proprietary is indicated in both versions by means of lower case letters (a) through (f)located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information.

These lower case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (4)(ii)(a) through (4)(ii)(f) of the affidavit accompanying this transmittal pursuant to 10 CFR 2.390(b)(1).

COPYRIGHT NOTICE The reports transmitted herewith each bear a Westinghouse copyright notice. The NRC is permitted to make the number of copies of the information contained in these reports which are necessary for its internal use in connection with generic and plant-specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding.

With respect to the non-proprietary versions of these reports, the NRC is permitted to make the number of copies beyond those necessary for its internal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary.

Dominion Nuclear Connecticut, Inc.Letter for Transmittal to the NRC The following paragraphs should be included in your letter to the NRC: Enclosed is: 1. _ copies of LTR-SGMMP-11-28 Rev. I P-Attachment, "Response to USNRC Request for Additional Information Regarding the License Amendment Requests for Permanent Application of the Alternate Repair Criterion, H*, to the Model D5 and Model F SGs" (Proprietary)

2. -copies of LTR-SGMMP-l 1-28 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" (Non-Proprietary)

Also enclosed is the Westinghouse Application for Withholding Proprietary Information from Public Disclosure CAW- 12-3446, accompanying Affidavit, Proprietary Information Notice, and Copyright Notice.As Item I contains information proprietary to Westinghouse Electric Company LLC, it is supported by an affidavit signed by Westinghouse, the owner of the information.

The affidavit sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of Section 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 10 CFR Section 2.390 of the Commission's regulations.

Correspondence with respect to the copyright or proprietary aspects of the items listed above or the supporting Westinghouse affidavit should reference CAW-] 2-3446 and should be addressed to J. A. Gresham, Manager, Regulatory Compliance, Westinghouse Electric Company, Suite 428, 1000 Westinghouse Drive, Cranberry Township, PA 16066.

Serial No: 12-203 Docket No. 50-423 Enclosure 7 Millstone Power Station Unit 3 Plant-Specific Response to Requests for Additional Information DOMINION NUCLEAR CONNECTICUT, INC.MILLSTONE POWER STATION UNIT 3 Serial No: 12-203 Docket No. 50-423 Enclosure 7, Page 1 of 3 Provided below are the Millstone Power Station Unit 3 (MPS3) specific responses to the relevant plant-specific requests for additional information submitted to Catawba Units 1 and 2 (Questions 12 and 13) and Surry Units 1 and 2 (Question 15). The NRC questions are identified in italics.Question 12 BET measurements for Catawba 2, documented in Westinghouse letter L TR-SGMP 111 P-Attachment, Revision 1, range to a maximum of 0. 65 inches and appear not to be a factor affecting the H* and leak rate ratio calculations.

Apart from tubes with this reported range of BETs, are there any non-expanded or partially expanded tubes at Catawba 2? If so, provide revisions to the proposed technical specifications which exclude such tubes from the proposed H* provisions.

Response Bottom expansion transition (BET) measurements for MPS3, documented in Westinghouse letter LTR-SGMP-09-1 11 P-Attachment, Revision 1, vary to a maximum of 1.74 inches. DNC completed the validation for MPS3 and the results were provided to the NRC in DNC letter 10-276, dated April 26, 2010 (Reference 13). Based on review of BET values, a total of seven tubes were identified with BET values greater than 1.0 inch from the top of the tubesheet (value considered "significant" by Westinghouse).

As committed to by DNC in Reference 13, these seven tubes were removed from service during Refueling Outage 14. The remainder of in-service tubes have BET measurements that range to a maximum of 0.95 inches. Apart from the tubes within this range of BETs, there are no non-expanded or partially expanded tubes in service at MPS3. As such, revision to the technical specifications to exclude such tubes from the proposed H* provisions is not required.Question 13 Proposed TS 5.6.8. h through j- The proposed changes contain more words than seem necessary, reducing the clarity of the proposed reporting requirements.

For example, the proposed wording refers to "an inspection performed after each refueling outage" which doesn't seem to make sense. The NRC staff believes the proposed requirements can be stated more clearly and concisely as follows: h. For Unit 2, follo.ing .ompl.tion of an inspecti.n po..orm.d during End of c,,o 17 Rofuoling Qutage (and any inspections pcdornqed during subsequ'n q'e 49 e8peratn, the primary-to-secondary LEAKAGE rate observed in each steam generator (if it is not practical to assign the leakage to an individual SG, the entire primary-to-secondary leakage should be conservatively assumed to be from one SG) during the cycle preceding the inspection which is the subject of the report, i. For Unit 2, follo..ng c ,mpltion of an inpection p...orm.d during the End of .y" ,, I " Rofucling Outage (and any, inspoction pedormod during subsequent Qyclc 18 operatio the calculated accident induced leakage rate from the portion of the tubes below -20 14.01 inches from the top of the tubesheet for the most limiting accident in the most limiting SG.In addition, if the calculated accident leakage rate from the most limiting accident is less than 3.27 times the maximum primary-to-secondary LEAKAGE rate, the report shall describe how it was determined, and Serial No: 12-203 Docket No. 50-423-Enclosure 7, Page 2 of 3 f For Unit 2, fooing .ORn.pl.tion of an iWpoton d.ring the End of ,Gc, ,, Rfue;-ng ,utagc (and any pe.o.med during subsequent 18 .p.. .ti...., the results of monitoring for tube axial displacement (slippage).

If slippage is discovered, the implications of the discovery and corrective action shall be provided.Provide revisions to the proposed reporting requirements as necessary to clarify their intent.Response The MPS3 proposed changes to technical specification (TS) 6.9.1.7, "Steam Generator Tube Inspection Report," are consistent with the NRC staff's recommendation above for the Model F SGs.Question 15 Verify that regulatory commitments pertaining to monitoring for tube slippage and for primary -to-secondary leakage, as described in Dominion letter dated December 16, 2010 (NRC ADAMS Accession No. ML103550206), Attachment 1, page 10 of 23, remain in place. In addition, revise the proposed amendment to include a revision to technical specification limit on primary to secondary leakage from 150 gallons per day (gpd) to 83 gpd (150 divided by the proposed 1.8 leakage factor), or provide a regulatory basis for not making this change.Response The regulatory commitments pertaining to monitoring for tube slippage and for primary-to-secondary leakage as described in Dominion Nuclear Connecticut, Inc. (DNC) letter (Serial No.09-525) dated November 23, 2009 (Reference 12 of Enclosure

1) remain in place as discussed in Enclosure 8.DNC is not proposing any changes to the primary-to-secondary leakage limit as specified in TS 3.4.6.2, "Reactor Coolant System 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 primary-to-secondary leakage are evaluated to ensure that they remain within regulatory limits (e.g., 10 CFR 50.67, GDC 19). The accident induced leakage performance criteria are intended to ensure the primary-to-secondary leak rate during any accident does not exceed the primary-to-secondary leak rate assumed in the accident analysis.

Radiological dose consequences define the limiting accident condition for the H*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 (Reference 1 of Enclosure

1) and NRC Regulatory Guide (RG) 1.121, "Bases for Plugging Degraded PWR Steam Generator Tubes," (Reference 25 of Enclosure

1) are satisfied due to the constraint provided by the tubesheet.

Through application of the limited tubesheet inspection scope, 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 Serial No: 12-203 Docket No. 50-423 Enclosure 7, Page 3 of 3 limit is 1.0 gallon per minute (gpm). The TS operational leak rate limit is 150 gallons per day (gpd) (0.1 gpm) through any one steam generator.

Consequently, there is significant margin between accident leakage and allowable operational leakage. The SLB/FLB leak rate ratio is 2.49, resulting in significant margin between the conservatively estimated accident leakage and the allowable accident leakage (1.0 gpm).

Serial No: 12-203 Docket No. 50-423 Enclosure 8 Millstone Power Station Unit 3 Commitments From Previous Steam Generator Alternate Repair Criteria DOMINION NUCLEAR CONNECTICUT, INC.MILLSTONE POWER STATION UNIT 3 Serial No: 12-203 Docket No. 50-423 Enclosure 8, Page 1 of 1 Applicable commitments from the steam generator one-time ARC (Reference 8 of Enclosure 1), which was approved May 3, 2010, by the NRC in MPS3 License Amendment (LA) 249 include the following:

1) To monitor for tube slippage as part of the SG tube inspection program.2) To perform a one-time verification of the tube expansion to locate any significant deviations in the distance from the top of tubesheet to the bottom of the expansion transition (BET). If any significant deviations are found, the condition will be entered into the plant's corrective action program and dispositioned.

Additionally, DNC commits to notify the NRC of significant deviations.

3) For the condition monitoring (CM) assessment, the component of leakage from the prior cycle from below the H* distance will be multiplied by a factor of 2.49 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 2.49 and compared to the observed operational leakage.An administrative limit will be established to not exceed the calculated value.The status of these commitments are as follows: The second commitment to perform BET measurements was completed by DNC as docketed in DNC letter 10-276, dated April 26, 2010 (Reference 13 of Enclosure 1);therefore, this commitment is closed.The program/procedure changes (Reference 27 and 28 of Enclosure

1) needed to meet the first and third commitments were completed in accordance with the NRC approval of LA 249. These changes are in place and will remain in place for this license amendment request. Therefore, these commitments are completed and closed and no new regulatory commitments are required.