ML110280208

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 Temporary Alternate Repair Criteria (H*)

ML110280208
Person / Time
Site:	Millstone
Issue date:	01/20/2011
From:	Price J Dominion Nuclear Connecticut
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
10-740, FOIA/PA-2011-0115
Download: ML110280208 (31)

Text

Dominion Nuclear Connecticut, Inc.

A 5000 Dominion Boulevard, Glen Allen, Virginia 23060 Web Address: www.dom.com PROPRIETARY INFORMATION-WITHHOLD UNDER I OCFR2.390 January 20, 2011 U.S. Nuclear Regulatory Commission Serial No.10-740 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 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 TEMPORARY ALTERNATE REPAIR CRITERIA (H*)

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 revise Technical Specification (TS) 6.8.4.g, "Steam Generator (SG) Program," to exclude a portion of the tubes below the top of the steam generator tubesheet from periodic steam generator tube inspections.

Specifically, the change deletes information associated with the one-time alternate repair criteria (ARC) for Cycle 14 and adds information associated with the temporary alternate repair criteria (TARC) for Cycle 15. In addition, the 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 Cycle 14, one-time ARC, and add reporting requirements specific to the Cycle 15, TARC.

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 of Enclosure 1 and provides the licensing basis for this change. provides a description and basis for the proposed changes.

The marked-up TS pages are provided in Enclosure 2. The document contained in provides proprietary Westinghouse information which supports the analysis described in Enclosure 1. provides the non-proprietary version of the same Westinghouse document.

As Enclosure 3 contains information proprietary to Westinghouse, Enclosure 5 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 contains information being withheld from public disclosure under IOCFR2.390. Upon separation, this page is decontrolled.

4001

Serial No: 10-740 Docket No. 50-423 Page 2 of 4 Westinghouse, be withheld from public disclosure in accordance with 2.390 of the Commission's regulations.

On May 3, 2010, the NRC issued MPS3 License Amendment (LA) 249 for the steam generator interim (i.e., "one-time") ARC (Reference 8 of Enclosure 1). As a condition of approval, DNC made the following regulatory commitments:

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 program/procedure changes needed to meet the first and third commitments were completed in accordance with the NRC approval of LA 249 (Reference 27 and 28 of Enclosure 1). The changes will also apply to this license amendment request.

As such, these changes will remain in place.

Therefore, no new regulatory commitments are required. The second commitment to perform BET measurements was completed by DNC as docketed in a DNC letter dated April 26, 2010 (Serial No.10-276). Therefore, this commitment is closed.

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 amendments by September 9, 2011 to support 3R14, which is scheduled to start on October 9, 2011.

Once approved, the proposed changes will be implemented within 30 days of issuance of the amendment and prior to Mode 5 startup of MPS3.

Serial No: 10-740 Docket No. 50-423 Page 3 of 4 Should you have any questions in regard to this submittal, please contact Wanda D. Craft at (804) 273-4687.

Sincerely, J Ala Price ice resident - Nuclear Engineering COMMONWEALTH OF VIRGINIA 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.

Acknowledged before me this *_day ofJ~n, 2011.

My Commission Expires:

/41AV 31.; 20/1 Notary Public Commitments: None KI L.HU It Notary Public

Enclosures:

Commonwealth of Virginia 140542 My Commismsion Expires May 31. 2014

1.

Basis for Proposed Change I

Ei M 31. 2014

2.

Markup of Proposed Technical Specifications

3.

Westinghouse Electric Company LLC, WCAP-1 7330-P, Rev. 0, "H*: Resolution of NRC Technical Issue Regarding Tubesheet Bore Eccentricity (Model F/D5),

November 2010. (Proprietary)

4.

Westinghouse Electric Company LLC, WCAP-17330-NP, Rev. 0, "H*: Resolution of NRC Technical Issue Regarding Tubesheet Bore Eccentricity (Model F/D5),

November 2010. (Non-Proprietary)

5.

Westinghouse Electric Company LLC, CAW-10-3003, "Application for Withholding Proprietary Information from Public Disclosure," dated November 8, 2010. (affidavit for WCA P-I 7330-P)

Serial No: 10-740 Docket No. 50-423 Page 4 of 4 cc:

U.S. Nuclear Regulatory Commission Region I Regional Administrator 475 Allendale Road King of Prussia, PA 19406-1415 C. J. Sanders' NRC Project Manager U.S. Nuclear Regulatory Commission, Mail Stop 08B3 One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector (w/o enclosures)

Millstone Power Station Director, Bureau of Air Management (w/o enclosures)

Monitoring and Radiation Division Department of Environmental Protection 79 Elm Street Hartford, CT 06106-5127

Serial No: 10-740 Docket No. 50-423 ENCLOSURE 1.

Basis for Proposed Change DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3

Serial No: 10-740 Docket No. 50-423, Page 2 of 21 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: 10-740 Docket No. 50-423, Page 3 of 21 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 revise Technical Specification (TS) 6.8.4.g, "Steam Generator (SG) Program," to exclude a portion of the tubes below the top of the steam generator tubesheet from periodic steam generator tube inspections. Specifically, the change deletes information associated with the one-time alternate repair criteria (ARC) for Cycle 14 and adds information associated with the temporary alternate repair criteria (TARC) for Cycle 15. In addition, the 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 Cycle 14, one-time ARC, and add reporting requirements specific to the Cycle 15, TARC.

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 below and provides the licensing basis for this change.

2.0 Detailed Description Proposed Changes 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. For Refueling Outage 1-3 14 and the subsequent operating cycle, tubes with service-induced flaws located greater than 1-34 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 144 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

Serial No: 10-740 Docket No. 50-423,, Page 4 of 21 tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria. For Refueling Outage 1-3 14 and the subsequent operating cycle, portions of the tube below 4434 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 Changes to TS 6.9.1.7.i, 6.9.1.7.i, and 6.9.1.7.k: Deleted text is struck through and added text is italicized and bold.

During Refueling Outage 4-3 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,

Serial No: 10-740 Docket No. 50-423, Page 5 of 21

j.

During Refueling Outage 4-3 14 and the subsequent operating cycle, the calculated accident induced leakage rate from the portion of the tubes below 4.4 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.

During Refueling Outage 4-3 14 and the subsequent operating cycle, the results of monitoring for tube axial displacement (slippage). If slippage is discovered, the implications of the discovery and corrective action shall be provided.

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 166 tubes are currently plugged in all 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-1 01, "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

Serial No: 10-740 Docket No. 50-423, Page 6 of 21 2005-09, shows the importance of monitoring all tube locations (such as bulges, dents, dings, and other anomalies from the manufacture of the steam generators) with techniques capable of finding potential forms of degradation that may be occurring at these locations (as discussed in Generic Letter 2004-001, "Requirements for Steam Generator Tube Inspections" - Reference 6). Since the MPS3 Westinghouse Model F steam generators were fabricated with Alloy 600 thermally treated tubes similar to the Catawba Unit 2 Westinghouse Model D5 steam generators, a potential exists for MPS3 to identify tube indications similar to those reported at Catawba within the hot leg tubesheet region if similar inspections are performed during refueling outage 3R14 (fall 2011).

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)

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 3R10 (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 3R10 or refueling outage 3R1 1.

For refueling outage 3R12 (fall 2008) and the subsequent operating cycle (Cycle 13), an interim alternate repair criteria (IARC) was approved as License Amendment 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'. All 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.

Serial No: 10-740 Docket No. 50-423, Page 7 of 21 For refueling outage 3R13 (spring 2010) and the subsequent operating cycle (Cycle 14), a one-time ARC was approved as License Amendment 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. During these inspections, a robust tube sample plan utilizing diagnostic examination techniques was executed at the expansion transition and at areas of overexpansions within the hot leg tubesheet. No indications of cracking were detected in any tube during refueling outage 3R1 3.

As a result of these potential issues, and to prevent the unnecessary plugging of additional tubes in the MPS3 steam generators, DNC is proposing a 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 TARC.

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 functioned as the lead plant for application of the permanent ARC. The Vogtle-specific information is indented and italicized.

On May 19, 2609, 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 permanent ARC for steam generator'tubes.

As part of the review of SNC's license amendment request dated May 19, 2009, the NRC issued two requests for additional information (RA Is) 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 LTR-SGMP-09-100 P-Attachment (Reference 9), with the response to RAI #4 being provided under L TR-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 L TR-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.

Serial No: 10-740 Docket No. 50-423, Page 8 of 21 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 license amendment request 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 permanent ARC 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 permanent ARC amendment request.

On November 23, 2009, DNC submitted a license amendment request 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 3R13 and the subsequent operating cycle (Reference 12). Westinghouse WCAP-1 7071-P, Revision 0, "H*: Alternate Repair Criteria for the Tubesheet Expansion Region in Steam Generators with Hydraulically Expanded Tubes (Model F)" was submitted as Enclosure 5 of this request. Also included in this 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 interim (i.e., one-time) ARC for refueling outage 3R13 and the subsequent operating cycle (Cycle 14), the NRC staff requested that DNC perform a validation of the tube expansion from the top of tubesheet to the beginning 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 a DNC letter dated April 26, 2010,(Serial No.10-276) (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 deviation assessment. This document was submitted to the NRC under Westinghouse letter LTR-NRC-1 0-69 (Reference 15).

On May 3, 2010, the NRC issued License Amendment 249 for MPS3's refueling outage 3R13 and the subsequent operating cycle (Reference 8).

Serial No: 10-740 Docket No. 50-423, Page 9 of 21 During the NRC's administrative processing of the supporting documentation for License Amendment 249, the NRC discovered discrepancies with proprietary markings in some of the Westinghouse documents supporting MPS3's license amendment request 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).

WCAP-17330-P, Rev. 0, "H*: Resolution of NRC Technical Issue Regarding Tubesheet Bore Eccentricity (Model F/D5), dated November 2010, LTR-SGMP-10-78-P, "Effects of Tubesheet Bore Eccentricity and Dilation on Tube-to-Tubesheet Contact Pressure and Their Relative Importance to H*," dated September 7, 2009 (Reference 20), and LTR-SGMP-10-33-P, "H* Response to NRC Questions Regarding Tubesheet Bore Eccentricity," dated September 13, 2010 (Reference 21) were prepared by Westinghouse to provide final resolution of the remaining questions identified in the November 23, 2009 NRC letter (Reference 11) in support of the permanent H*amendment. WCAP-1 7330-P, Rev. 0 is provided in Enclosure 3 while documents LTR-SGMP-10-78-P and LTR-SGMP-10-33-P were submitted to the NRC under Westinghouse letters LTR-NRC-1 0-68 (Reference 22) and LTR-NRC-10-70 (Reference 23), respectively.

The following table provides the list of licensing basis documents for H* for MPS3.

Document Revision Title Reference Number Number Number WCAP-17071-P 2

H*: Alternate Repair Criteria for the Tubesheet 18 Expansion Region in Steam Generators with Hydraulically Expanded Tubes (Model F)"

LTR-SGMP 1

Response to NRC Request for Additional Information 19 100 P-Attachment on H*; Model F and Model D5 Steam Generators LTR-SGMP 0

Response to NRC Request for Additional Information 10 109 P-Attachment on H*; RAI #4; Model F and Model D5 Steam Generators WCAP-17330-P 0

H*: Resolution of NRC Technical Issue Regarding 24

-ubesheet Bore Eccentricity (Model F/D5)

LTR-SGMP-10-78 0

Effects of Tubesheet Bore Eccentricity and Dilation 20 P-Attachment on Tube-to-Tubesheet Contact Pressure and Their Relative Importance to H*

LTR-SGMP-10-33 0

H* Response to NRC Questions Regarding 21 P-Attachment Fubesheet Bore Eccentricity

Serial No: 10-740 Docket No. 50-423, Page 10 of 21 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. 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, GDC 19). The accident induced leakage performance criteria are intended to ensure the primary-to-secondary leak rate during any accident does not exceed the primary-to-secondary leak rate assumed in the accident analysis. Radiological dose consequences define the limiting accident condition for the H*

justification.

The constraint that is provided by the tubesheet precludes tube burst from cracks within the tubesheet. The criteria for tube burst described in NEI 97-06 and NRC Regulatory Guide (RG) 1.121, "Bases for Plugging Degraded PWR Steam Generator Tubes,"

(Reference 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 gpm. The TS operational leak rate limit is 150 gallons per day (gpd) (0.1 gallons per minute (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 only 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-

Serial No: 10-740 Docket No. 50-423, Page 11 of 21 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 from 2.69 to 0.45.

The plant transient response following a full power double-ended main feedwater line rupture corresponding to "best estimate" initial conditions and operating characteristics indicates that the transient for a Model F 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 reactor coolant system components; however, H* does not involve component structural

Serial No: 10-740 Docket No. 50-423, Page 12 of 21 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 exceptionally more severe when the FLB occurs inside containment. For example, environmental errors that are applied to reactor trip and engineered safety feature actuation would be less severe. This would result in a 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 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,

Serial No: 10-740 Docket No. 50-423, Page 13 of 21 including the most limiting accident conditions. The evaluation in WCAP-1 7071-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 License Amendment 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 license amendment request.

As a condition for approving the interim (i.e., one-time) ARC for refueling outage 3R13 and the subsequent operating cycle (Cycle 14), the NRC staff requested that DNC perform a validation of the tube expansion from the top of tubesheet to the beginning 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 a DNC letter dated April 26, 2010 (Serial No.10-276) (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.

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.

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

10 CFR 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.

Serial No: 10-740 Docket No. 50-423, Page 14 of 21 Under 10 CFR 50.65, the Maintenance Rule, licensees classify steam generators as risk significant components because they are relied upon to remain functional during and after design basis events. Steam generators are to be monitored under 10 CFR 50.65(a)(2) against industry established performance criteria. Meeting the performance criteria of NEI 97-06, Revision 2, 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 2, 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 applying a safety factor of 1.2 on the combined primary loads and 1.0 on axial loads.

The primary-to-secondary accident induced leakage rate for any design basis accident, other than a steam generator tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all steam generators and leakage rate for an individual steam generator. Leakage is not to exceed 1.0 gpm per steam generator, except for specific types of degradation at specific locations when implementing alternate repair criteria as documented in the Steam Generator Program technical specifications.

The RCS operational primary-to-secondary leakage through any one steam generator shall be limited to 150 gallons per day.

The safety significant portion of the tube is the length of tube that is engaged in the tubesheet from the secondary face that is required to maintain structural and leakage integrity over the full range of steam generator operating conditions, including the most limiting accident conditions. The evaluation in this enclosure determined that degradation in tubing below the safety significant portion of the tube (i.e., 15.2 inches from the top of the tubesheet) does not require plugging and serves as the bases for the tubesheet 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.

6.2 No Significant Hazards Consideration This amendment application proposes to revise 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

Serial No: 10-740 Docket No. 50-423, Page 15 of 21 remove reference to previous (Cycle 14) one-time alternate repair criteria (ARC) and provide reporting requirements specific to the Cycle 15 temporary ARC. 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 Nuclear Energy Institute (NEI) 97-06, "Steam Generator Program Guidelines,"

(Reference 1) performance criteria.

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 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

Serial No: 10-740 Docket No. 50-423, Page 16 of 21 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 (including locked rotor) has been shown to remain within the accident analysis assumptions for all axial and or circumferentially orientated cracks occurring 15.2 inches below the top of the tubesheet. The accident induced leak rate limit is 1.0 gpm. The TS operational leak rate is 150 gpd (0.1 gpm) through any one steam generator. Consequently, there is significant margin between accident leakage and allowable operational leakage. The SLB/FLB leak rate ratio is only 2.49

Serial No: 10-740 Docket No. 50-423, Page 17 of 21 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 2, "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. 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 of this license amendment request, defines the length of degradation-free, expanded tubing that provides the necessary resistance to tube pullout due to the pressure induced forces, with applicable safety factors applied. Application of the limited hot and cold leg tubesheet inspection criteria will preclude unacceptable primary-to-secondary leakage during all plant conditions.

The methodology for determining leakage provides for large margins between calculated and actual leakage values in the proposed limited tubesheet inspection depth criteria.

Serial No: 10-740 Docket No. 50-423, Page 18 of 21 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 proposed license amendment requests for temporary alternate repair criteria:

SNC Letter NL-10-2104, "Vogtle Electric Generating Plant - Units 1 and 2, License Amendment Request to Revise Technical Specification (TS) Sections 5.5.9, "Steam Generator (SG) Program" and TS 5.6.10, "Steam Generator Tube Inspection Report," for Temporary Alternate Repair Criteria (Reference 29).

Luminant Power letter CP-201001489, "Comanche Peak Nuclear Power Plant (CPNPP) - Docket Nos. 50-445 and 50-446, License Amendment Request 10-004, Model D5 Steam Generator Temporary Alternate Repair Criteria," dated December 1, 2010 (Reference 30).

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) and 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.

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

Serial No: 10-740 Docket No. 50-423, Page 19 of 21 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. 2, "Steam Generator Program Guidelines," May 2005.

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, "Steam Generator Program," Revision 4.

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 Steam Generator Inspection Scope and Repair Requirements (TAC No.

ME2978)," dated May 3, 2010. (ADAMS Accession No. ML100770358)

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-Attachment, Response to NRC Request for Additional Information on H*;. RAI #4; Model F and Model D5 Steam Generators, dated August 25, 2009.

1.1. NRC Letter, "Vogtle Electric Generating Plant, Units 1 and 2 - Transmittal of Unresolved Issues Regarding Permanent Alternate Repair Criteria for Steam Generators (TAC Nos. ME 1339 and ME 1340)," November 23, 2009. (ADAMS Accession No. ML093030490)

12. DNC letter 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

Serial No: 10-740 Docket No. 50-423, Page 20 of 21 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-10-69, "Submittal of LTR-SGMP-09-1 11 P-Attachment, Rev. 1 and LTR-SGMP-09-111 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 Millstorie Power Station, Unit No. 3 (TAC No. ME2978)," dated September 22, 2010 (ADAMS Accession No. ML102160749).

17. DNC letter 10-579A, "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. ML102850435).

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.

21. Westinghouse Electric Company LLC, LTR-SGMP-10-33 P-Attachment, "H*:

Response to NRC Questions Regarding Tubesheet Bore Eccentricity," dated September 13, 2010.

22. Westinghouse letter LTR-NRC-10-68, "Submittal of LTR-SGMP-10-78 P-Attachment and LTR-SGMP10-78 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.

Serial No: 10-740 Docket No. 50-423, Page 21 of 21

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. 0, "H*: Resolution of NRC Technical Issue Regarding Tubesheet Bore Eccentricity (Model F/D5),

November 2010. (See Enclosure 3)

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-102, Revision 3, "Steam Generator Degradation Assessment."

28. ER-AP-SGP-1 03, Revision 3, "Steam Generator Condition Monitoring and Operational Assessments."

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

30. Luminant Power letter CP-201001489, "Comanche Peak Nuclear Power Plant (CPNPP) - Docket Nos. 50-445 and 50-446, License Amendment Request 10-004, Model D5 Steam Generator Temporary Alternate Repair Criteria," dated December 1,2010.

Serial No: 10-740 Docket No. 50-423 MARK-UP OF PROPOSED TECHNICAL SPECIFICATIONS DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3

For Information Onlyl 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 leakageperformance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG.

MILLSTONE - UNIT 3 6-17a Amendment No. 23 8

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:

15.2

1. For Refl44 and the subsequent op ti

ycle, tubes with service-induced flaws located gre r than 4-34 inches below the top of the tubesheet do n require plugging.

Tubes with service-induced flaws locat

• in the portion of the tube from the top of the tubesheet to 434 inches below the top of the tubesheet shall be plugged upon detection.

MILLSTONE - UNIT 3 6-17b Amendment No. 238, 2-,4,-, 2*49

M" 3, 2010 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 1cracks) that may be present along the length of the14 circumferentialcrcsthtmybprsnalnthleghote tube, from the tube-to-tubesheet weld at the tube inlet to the 2_

tube-to-tubesheet weld at the tube outlet, and that may s

, the*,,-1.2 applicable tube repair criteria. For Refueling Outage 4-3 and.

subsequent operating cycle, portions of the tube below L 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 MILLSTONE - UNIT 3 6-17c Amendment No. 69, 4-86, 24-2-, 3-8-,

243,245,249

May,20 10 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 pluggE percentage for all plugging in each SQ

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 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, 5.0, 69, 1-04, 4--74, 4

, 24-5,2-2-, 3-9, 4-5,-24

MAA-231a-80 ADMINISTRATIVE CONTROLS STEAM GENERATOR TUBE INSPECTION REPORT (Continued)

F1

j.

During Refueling Outage 4-3 and the subsequent operating cycle, th/alculated accident induced leakage rate from the portion of the tubes below 4-34 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.

During Refueling Outage 44 and the subsequent operating cycle, the results of monitoring for tube axial displacement (slippage). If slippage is discovered, the implications of the discovery and corrective action shall be provided.

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 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 6-21a Amendment No. '.- 8, 24-5, 2-49