ML13179A014

Proposed License Amendment Request Addition of an Analytical Methodology to Core Operating Limits Reports and an Increase to Minimum Temperature for Criticality

ML13179A014
Person / Time
Site:	Surry, North Anna
Issue date:	06/26/2013
From:	Heacock D Virginia Electric & Power Co (VEPCO)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML13179A014 (42)

Text

VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 June 26, 2013 10 CFR 2.390 10 CFR 50.90 U. S. Nuclear Regulatory Commission Serial No.13-145 ATTN: Document Control Desk NL&OS/ETS RO Washington, D. C. 20555 Docket Nos. 50-280/281 50-338/339 License Nos. DPR-32/37 NPF-4/7 VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS 1 AND 2 NORTH ANNA POWER STATION UNITS 1 AND 2 PROPOSED LICENSE AMENDMENT REQUEST (LAR)

ADDITION OF AN ANALYTICAL METHODOLOGY TO THE NORTH ANNA AND SURRY CORE OPERATING LIMITS REPORTS (COLRS) AND AN INCREASE TO THE SURRY MINIMUM TEMPERATURE FOR CRITICALITY Pursuant to 10 CFR 50.90, Virginia Electric and Power Company (Dominion) requests amendments, in the form of changes to the Technical Specifications (TS) to Facility Operating License Numbers NPF-4 and NPF-7 for North Anna Power Station Units 1 and 2, and the Technical Specifications (TS) to Facility Operating License Numbers DPR-32 and DPR-37 for Surry Power Station Units 1 and 2, respectively. The proposed LAR requests approval of the following items: 1) generic application of Appendix D, "Qualification of the ABB-NV and WLOP Critical Heat Flux (CHF) Correlations in the Dominion VIPRE-D Computer Code," to Fleet Report DOM-NAF-2-A, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code," 2) the plant specific application of Appendix D to DOM-NAF-2-A to North Anna and Surry Power Stations (in accordance with Section 2.1 of DOM-NAF-2-A), and 3) an increase in the Surry Power Station TS Minimum Temperature for Criticality. The plant specific application of Appendix D to DOM-NAF-2-A requires the inclusion of the appendix to the TS list of references for determining core operating limits (i.e., the TS list of COLR references).

A discussion of the proposed changes is provided in Attachment 1. The marked-up and typed proposed TS pages are provided in Attachments 2 and 3 for North Anna and Attachments 4 and 5 for Surry, respectively. Attachment 6 documents the qualification of the ABB-NV and WLOP CHF Correlations in the Dominion VIPRE-D computer code and the associated code/correlation deterministic design limits.

We have evaluated the proposed amendments and have determined that they do not involve a significant hazards consideration as defined in 10CFR50.92. The basis for our determination is included in Attachment 1. We have also determined that operation with Attachment 6 to this letter contains Proprietary Information which is to be withheld from public disclosure under 10 CFR 2.390.

Upon removal of the attachment, this letter is decontrolled.

Serial No.13-145 Docket Nos. 50-280/281 and 50-338/339 Page 2 of 5 the proposed changes will not result in any significant increase in the amount of effluents that may be released offsite and no significant increase in individual or cumulative occupational radiation exposure. Therefore, the proposed amendments are eligible for categorical exclusion from an environmental assessment as set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment is needed in connection with the approval of the proposed changes. The basis for our determination is also included in Attachment 1. The proposed amendments have been reviewed and approved by the Facility Safety Review Committee.

For reference, the licensing history of DOM-NAF-2-A is as follows. In letters dated September 30, 2004 and January 13, 2005 [Serial Nos.04-606 (ML042800118) and 05-020 (ML050180257), respectively] Dominion submitted DOM-NAF-2, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code," and associated Appendices A and B, as supplemented by letters dated June 30, 2005 and September 8, 2005 [Serial Nos.05-328 (ML051870070) and 05-020A (ML052510484), respectively] for NRC review and approval. NRC approval of DOM-NAF-2 including Appendixes A and B was obtained in letter dated April 4, 2006 (ML060790496) as revised by the NRC in letter dated June 23, 2006 (ML061740212). In letter dated April 4, 2008 [Serial No. 08-0174, (ML080980229)],

Dominion submitted Appendix C of DOM-NAF-2-A for NRC review and approval. NRC approval of Appendix C of DOM-NAF-2-A was obtained in letter dated April 22, 2009 (ML091030639). In letter dated August 28, 2009 and supplemented by letter dated November 20, 2009 [Serial Nos.09-528 (ML092430338) and 09-528A (ML093310330),

respectively) Dominion requested approval of the removal of a restriction on the WRB-1 CHF correlation. NRC approval was obtained in letter dated June 21, 2010 (ML101620034). Dominion provided the latest approved version of DOM-NAF-2 Rev. 0.2-P-A to the NRC in letter dated August 20, 2010 [Serial No.10-486 (ML102390421)].

Continuing with this modular approach to Fleet Report DOM-NAF-2, Dominion is now submitting Appendix D to this Fleet Report, "Qualification of the ABB-NV and WLOP CHF Correlations in the Dominion VIPRE-D Computer Code," for NRC review and approval.

Associated with our request for NRC approval of DOM-NAF-2, Appendix D, the following three documents are also provided as Attachments 6, 7 and 8, respectively:

1. DOM-NAF-2-P, Rev. 0.3 Appendix D, "Qualification of the ABB-NV and WLOP CHF Correlations in the Dominion VIPRE-D Computer Code" (Proprietary)

2. DOM-NAF-2-NP, Rev. 0.3 Appendix D, "Qualification of the ABB-NV and WLOP CHF Correlations in the Dominion VIPRE-D Computer Code" (Non-Proprietary)

3. Westinghouse Application for Withholding Proprietary Information from Public Disclosure CAW-13-3687, accompanying Affidavit, and Proprietary Information Notice.

Item 1 above contains information proprietary to Westinghouse Electric Company LLC and 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

Serial No.13-145 Docket Nos. 50-280/281 and 50-338/339 Page 3 of 5 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-13-3687 and should be addressed to James A. Gresham, Manager, Regulatory Compliance, Westinghouse Electric Company, Suite 428, 1000 Westinghouse Drive, Cranberry Township, Pennsylvania 16066.

Approval of the proposed amendments is requested by June 27, 2014. Dominion also requests a 60-day implementation period following NRC approval of the requested license amendments.

If you have any questions or require additional information, please contact Mr. Thomas Shaub at (804) 273-2763.

Sincerely, David A. Heacock President and Chief Nuclear Officer COMMONWEALTH OF VIRGINIA

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COUNTY OF HENRICO The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by D. A. Heacock, who is President and Chief Nuclear Officer, of Virginia Electric and Power Company. 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 0 U ý--ay of A/.,, 2013.

My Commission Expires: 5IL2hJ/* .

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Serial No.13-145 Docket Nos. 50-280/281 and 50-338/339 Page 4 of 5 Commitments made in this letter: None Attachments:

1. Discussion of Change

2. Proposed North Anna Technical Specifications Pages (Mark-Up)

3. Proposed North Anna Technical Specifications Pages (Typed)

4. Proposed Surry Technical Specifications and Bases Pages (Mark-Up)

5. Proposed Surry Technical Specifications and Bases Pages (Typed)

6. Qualification of the ABB-NV and WLOP CHF Correlations in the Dominion VIPRE-D Computer Code (Proprietary)

7. Qualification of the ABB-NV and WLOP CHF Correlations in the Dominion VIPRE-D Computer Code (Non-Proprietary)

8. Westinghouse Electric Company LLC, Application for Withholding Proprietary Information from Public Disclosure and the Accompanying Affidavit

Serial No.13-145 Docket Nos. 50-280/281 and 50-338/339 Page 5 of 5 cc: U.S. Nuclear Regulatory Commission Region II Marquis One Tower 245 Peachtree Center Avenue, NE Suite 1200 Atlanta, Georgia 30303-1257 NRC Senior Resident Inspector North Anna Power Station NRC Senior Resident Inspector Surry Power Station Ms. K. R. Cotton Gross NRC Project Manager U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 08 G-9A 11555 Rockville Pike Rockville, Maryland 20852-2738 Dr. V. Sreenivas NRC Project Manager U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 08 G-9A 11555 Rockville Pike Rockville, Maryland 20852-2738 Mr. J. E. Reasor, Jr.

Old Dominion Electric Cooperative Innsbrook Corporate Center, Suite 300 4201 Dominion Blvd.

Glen Allen, Virginia 23060 State Health Commissioner Virginia Department of Health James Madison Building - 7 th Floor 109 Governor Street Room 730 Richmond, Virginia 23219

Serial No.13-145 Docket Nos. 50-280/281 and 50-338/339 ATTACHMENT 1 DISCUSSION OF CHANGE Virginia Electric and Power Company (Dominion)

Surry Power Station Units 1 and 2 North Anna Power Station Units I and 2

Serial No.13-145 Docket Nos. 50-280/281 and 50-338/339 Attachment 1

1.

SUMMARY

DESCRIPTION Pursuant to 10CFR50.90, Virginia Electric and Power Company (Dominion) is requesting approval of the following items: 1) generic application of Appendix D, "Qualification of the ABB-NV and WLOP Critical Heat Flux (CHF) Correlations in the Dominion VIPRE-D Computer Code," to Fleet Report DOM-NAF-2-A, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code," 2) the plant specific application of Appendix D to DOM-NAF-2-A to North Anna Power Station (North Anna) and Surry Power Station (Surry) in accordance with Section 2.1 of DOM-NAF-2-A, and

3) an increase in the Surry Technical Specifications (TS) Minimum Temperature for Criticality.

The first item, Appendix D, "Qualification of the ABB-NV and WLOP Critical Heat Flux (CHF) Correlations in the Dominion VIPRE-D Computer Code," to Fleet Report DOM-NAF-2-A documents the qualification of the W-3 Alternate CHF correlations using VIPRE-D and lists the deterministic design limits (DDLs) for each alternate correlation.

The W-3 Alternate CHF correlations are the ABB-NV and WLOP CHF correlations (Reference 1). Implementation of the W-3 Alternate CHF correlations provides improved predictive capabilities in determining the thermal-hydraulic performance of 17x17 fuel products within North Anna's cores and 15x15 fuel products within Surry's cores. Proprietary and non-proprietary versions of Appendix D to DOM-NAF-2-A are provided as Attachments 6 and 7.

The second item requests the inclusion of Appendix D of Dominion Fleet Report DOM-NAF-2-A to the North Anna TS 5.6.5.b and Surry TS 6.2.C list of NRC-approved methodologies used to determine core operating limits (i.e., the reference list of the Core Operating Limits Report (COLR)). The addition of DOM-NAF-2, Appendix D within the Surry TS also requires the addition of the ABB-NV CHF correlation and its corresponding VIPRE-D DDL to Surry TS. 2.1 .A.l.a.

The third proposed item increases the Surry TS 3.1.E.4 Minimum Temperature for Criticality limit from 522 0F to 538 0F. When increasing the TS Minimum Temperature for Criticality, it is also necessary to modify Surry TS 3.1..B.1, [Reactor Coolant System (RCS)] Heatup and Cooldown, to reflect the revised limit. The proposed increase in the TS Minimum Temperature for Criticality provides margin in verifying that a given reload cycle design meets the most-positive moderator temperature coefficient (MTC) limit.

Surry TS Bases changes reflecting the proposed Surry TS changes have been included for information. There are no North Anna TS Bases changes in support of the proposed change to the North Anna TS.

Page 1 of 18

Serial No.13-145 Docket Nos. 50-280/281 and 50-338/339 Attachment 1 The proposed TS changes have been reviewed, and it has been determined that no significant hazards consideration exists as defined in 10 CFR 50.92. In addition, it has been determined that the changes qualify for categorical exclusion from an environmental assessment as set forth in 10 CFR 51.22(c)(9); therefore, no environmental impact statement or environmental assessment is needed in connection with the approval of the proposed TS changes.

2. BACKGROUND 2.1 DOM-NAF-2 and Appendix D VIPRE-D is the Dominion version of the computer code VIPRE (Versatile Internals and Components Program for Reactors - EPRI), developed for EPRI (Electric Power Research Institute) by Battelle Pacific Northwest Laboratories in order to perform detailed thermal-hydraulic analyses to predict CHF and Departure from Nucleate Boiling (DNB) of reactor cores. VIPRE-01 has been approved by the NRC (References 2 and 3). VIPRE-D is based upon VIPRE-01 and was customized by Dominion to fit the specific needs of Dominion's nuclear plants and fuel products through the addition of several vendor specific CHF correlations. Dominion has obtained generic approval of Fleet Report DOM-NAF-2-A (Reference 4) from the NRC, which provides the necessary documentation to describe the intended uses of VIPRE-D for PWR licensing applications.

Appendix D to DOM-NAF-2 documents the qualification of the W-3 Alternate CHF correlations using VIPRE-D and lists the DDLs for each alternate correlation. The W-3 Alternate CHF correlations are the ABB-NV and WLOP CHF correlations (Reference 1).

The ABB-NV CHF correlation is a replacement for the W-3 CHF correlation for nominal operating conditions that fall outside of the primary CHF correlation applicable range, (i.e., below the first mixing vane grid (MVG) also referred to as the non-mixing vane grid (NMVG) span). The WLOP CHF correlation is a replacement for the W-3 CHF correlation at low pressure and low flow operating conditions.

2.2 Minimum Temperature for Criticality Surry Units 1 and 2 implemented the Minimum Temperature for Criticality TS by Amendments 105/105 dated December 31, 1985 (Reference 5). The TS Limiting Condition for Operation (LCO) was set at 522 0 F and has been maintained at that value since then (Reference 6). The TS Minimum Temperature for Criticality at Surry is lower than the other Westinghouse plants in the Dominion nuclear fleet. The TS Minimum Temperature for Criticality is 541OF for North Anna (Reference 7), 540°F for Kewaunee Page 2 of 18

Serial No.13-145 Docket Nos. 50-280/281 and 50-338/339 Attachment 1 (Reference 8), and 551 *F for Millstone 3 (Reference 9). This proposed change to 538°F will bring Surry more in line with the rest of the Dominion fleet.

2.3 Coupling of the Surry TS Changes The Surry implementation of the W-3 Alternate CHF correlations has been coupled with a request to increase the TS Minimum Temperature for Criticality. These two changes have been coupled because together they provide increased flexibility in loading pattern development as well as improved design margins. The W-3 Alternate CHF correlations provide improved predictive capabilities in determining the thermal-hydraulic performance at Surry, and the increase in the TS Minimum Temperature for Criticality provides margin in verifying that a given cycle design meets the most-positive MTC limit.

3. PROPOSED TECHNICAL SPECIFICATIONS CHANGES The following specific change to the North Anna Units 1 and 2 TS is proposed:

• North Anna TS 5.6.5.b Core Operating Limits Report (COLR)

- This section is revised to modify a reference that reflects the proposed change below. The modification describes the Dominion-specific analytical application of the ABB-NV and WLOP CHF correlations used in determining core limits at North Anna. The following bolded text is proposed to be added to COLR Reference 16:

16.DOM-NAF-2-A, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code," including Appendix A, "Qualification of the F-ANP BWU CHF Correlations in the Dominion VIPRE-D Computer Code," Appendix C, "Qualification of the Westinghouse WRB-2M CHF Correlation in the Dominion VIPRE-D Computer Code," and Appendix D, "Qualification of the ABB-NV and WLOP CHF Correlations in the Dominion VIPRE-D Computer Code."

The following specific changes to the Surry Units 1 and 2 TS are proposed:

• Surry TS 2.1 Safety Limit, Reactor Core

- Revise TS 2.1.A.1.a to include "-- 1.14 for ABB-NV". The change adds the ABB-NV correlation and the corresponding VIPRE-D Deterministic Design Limit.

Page 3 of 18

Serial No.13-145 Docket Nos. 50-280/281 and 50-338/339 Attachment 1

" Surry TS 3.1 Reactor Coolant System

- Revise TS 3.1.B.1 to reference the applicable specification for the Minimum Temperature for Criticality, TS 3.1.E, versus providing the limit in multiple specifications.

- Revise the TS 3.1.B Basis to reference the applicable specification for the Minimum Temperature for Criticality, TS 3.1.E, versus providing the limit in multiple specifications.

- Revise TS 3.1.E.4 to reflect the proposed Minimum Temperature for Criticality limit of 538 0 F.

- Revise the TS 3.1.E Basis to reflect the proposed Minimum Temperature for Criticality limit of 538 0 F.

• Surry TS 6.2.C Core Operating Limits Report

- This section is revised to modify a reference that reflects the proposed change below. The modification describes the Dominion-specific analytical application of the ABB-NV and WLOP CHF correlations used in determining core limits at Surry. The following bolded text is proposed to be added to COLR Reference 8:

8. DOM-NAF-2-A, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code," including Appendix B, "Qualification of the Westinghouse WRB-1 CHF Correlation in the Dominion VIPRE-D Computer Code," and Appendix D, "Qualification of the ABB-NV and WLOP CHF Correlations in the Dominion VIPRE-D Computer Code."

4. TECHNICAL EVALUATION 4.1 Appendix D of DOM-NAF-2 Appendix D of Fleet Report DOM-NAF-2, which is provided in Attachments 6 and 7, documents the qualification of the ABB-NV and WLOP CHF Correlations with the VIPRE-D computer code and the associated code/correlation DDLs.

4.2 Application of Appendix D of DOM-NAF-2 to North Anna This section includes the technical basis and the required documentation to support the site specific application of the VIPRE-D thermal hydraulics code with the ABB-NV and WLOP CHF correlations to 17x17 fuel products in North Anna Power Station Units 1 and 2 cores. Section 2.1 of Fleet Report DOM-NAF-2-A lists the information to be Page 4 of 18

Serial No.13-145 Docket Nos. 50-280/281 and 50-338/339 Attachment 1 provided to the NRC by Dominion for review and approval of any plant specific application of the VIPRE-D computer code and includes:

1) The addition of DOM-NAF-2-A and the applicable appendices to the plant's reference list of the COLR,

2) The identification of any changes to Statistical Design Limits (SDLs) for the relevant code and correlation(s),

3) The identification of any TS changes to OTAT, OPAT, FAH, or reactor protection function, as well as any changes to Reactor Core Safety Limits (RCSLs), and

4) The identification of any changes to the list of UFSAR transients for which the code and correlations apply.

This License Amendment Request (LAR) requests the addition of Appendix D of DOM-NAF-2 to the North Anna COLR list of references. Descriptions of the required TS changes are provided in Section 3. This satisfies item 1 above.

The W-3 Alternate CHF correlations will not be used statistically in licensing bases calculations at North Anna Power Station. Instead, the W-3 Alternate CHF correlations will be applied deterministically when used in licensing bases calculations for North Anna's cores. This is the same manner in which the W-3 CHF correlation is currently applied. The applicable DDLs for the W-3 Alternate CHF correlations are those documented in Appendix D of DOM-NAF-2. The DDLs are 1.14 for ABB-NV and 1.22 for WLOP. As there are no changes to the SDLs at North Anna, item 2 is satisfied.

A deterministic safety analysis limit (SAL) equal to 1.55 has been selected for both the ABB-NV and WLOP CHF correlations at North Anna for use in the analyses of the Westinghouse RFA-2 fuel design. The DDLs are fixed and any changes to their values would require NRC review and approval. However, the SALs for DNB analyses may be changed without prior NRC review and approval provided the changes meet the criteria established in Reference 10.

The difference between the SAL and the DDL is the available retained departure from nucleate boiling ratio (DNBR) margin:

Retained DNBR Margin [%] = k (SAL SALDDL) *100 10 The resulting available retained DNBR margin is 26.4% for ABB-NV and 21.2% for WLOP.

Page 5 of 18

Serial No.13-145 Docket Nos. 50-280/281 and 50-338/339 Attachment 1 This method of defining retained DNBR margin allows for all of the DNBR margin to be found in a single, clearly defined location. The retained DNBR margin can be used to offset generic DNBR penalties such as rod bow. The retained DNBR margin may also be used to offset potential DNBR penalties such as transition core penalties. The reload thermal-hydraulic evaluations prepared as part of the reload safety analysis process evaluates the applicable DNB penalties and confirms positive DNBR margin exists for the given cycle. Retained DNBR margin is tracked separately for each DNB limiting accident and applicable CHF correlation.

In determining the impact of the use of the W-3 Alternate CHF correlations on the reactor protection system and the list of UFSAR transients for which the code and correlations apply, it is important to understand how the W-3 Alternate CHF correlations will be applied at North Anna. As discussed in Section D.2 of Appendix D of DOM-NAF-2 and Section 2.1 above, the ABB-NV CHF correlation is to be applied to Updated Final Safety Analysis Report (UFSAR) transients where the limiting location of DNBR occurs below the first MVG, and the WLOP CHF correlation is to be applied to UFSAR transients when the conditions occur outside of the range of applicability of the primary CHF correlation, specifically low pressure or low flow conditions. This is consistent with the current application of the W-3 CHF correlation at North Anna.

Dominion verified acceptable DNB performance of the Westinghouse RFA-2 fuel at North Anna with the VIPRE-D/ABB-NV and VIPRE-D/WLOP code/correlation pairs using the DDLs qualified in Appendix D of DOM-NAF-2. The results of these calculations demonstrate that the minimum DNBR values are equal to or greater than the applicable SAL for all of the analyses. These calculations support the current RCSLs; the OTAT and OPAT trip functions (including FAI reset functions); FAH limits; as well as the evaluated Chapter 15 events. As such there are no changes to the OTAT, OPAT, FAH, reactor protection system, or the RCSLs due to the implementation of the W-3 Alternate CHF correlations at North Anna. This satisfies item 3.

The two notable UFSAR transients where the W-3 Alternate CHF correlation is limiting with regards to DNB performance in licensing bases calculations are the Rod Withdrawal from Subcritical (RWSC) and Main Steamline Break (MSLB) accidents. The RWSC event occurs within the applicable range of the primary CHF correlation with the exception of the limiting location of DNBR potentially being below the first MVG. Thus, the ABB-NV CHF correlation is applied to the RWSC transient. The limiting DNBR for the MSLB event occurs under low pressure conditions and is outside the applicable range of the primary CHF correlation. Thus, the WLOP CHF correlation is applied to the MSLB event. The other Chapter 15 UFSAR transients are limited by the primary CHF correlation and are not affected by the implementation of the W-3 Alternate CHF Page 6 of 18

Serial No.13-145 Docket Nos. 50-280/281 and 50-338/339 Attachment 1 correlations. There is no change to the list of UFSAR transients analyzed with VIPRE-D.

This satisfies item 4.

4.3 Application of Appendix D of DOM-NAF-2 to Surry This section includes the technical basis and the required documentation to support the site specific application of the VIPRE-D thermal hydraulics code with the ABB-NV and WLOP CHF correlations to 15x15 fuel products in Surry Power Station Units 1 and 2 cores. Section 2.1 of Fleet Report DOM-NAF-2-A lists the information to be provided to the NRC by Dominion for review and approval of any plant specific application of the VIPRE-D computer code and includes:

1) The addition of DOM-NAF-2-A and the applicable appendices to the plant's reference list of the COLR,

2) The identification of any changes to the statistical design limits (SDLs) for the relevant code and correlation(s),

3) The identification of any TS changes to OTAT, OPAT, FAH, or reactor protection function, as well as any changes to Reactor Core Safety Limits (RCSLs), and

4) The identification of any changes to the list of UFSAR transients for which the code and correlations apply.

This LAR requests the addition of Appendix D of DOM-NAF-2 to the Surry COLR list of references. The addition of Appendix D of DOM-NAF-2 also requires the addition of the ABB-NV DDL to TS 2.1.A.1.a. The DDL for WLOP is not added to TS 2.1.A.1.a at Surry, as the WLOP CHF correlation does not support the RCSLs. Descriptions of the required TS changes are provided in Section 3. This satisfies item 1 above.

The W-3 Alternate CHF correlations will not be used statistically in licensing bases calculations at Surry Power Station. Instead, the W-3 Alternate CHF correlations will be applied deterministically when used in licensing bases calculations for Surry's cores.

This is the same manner in which the W-3 CHF correlation is currently applied. The applicable DDLs for the W-3 Alternate CHF correlations are those documented in Appendix D of DOM-NAF-2. The DDLs are 1.14 for ABB-NV and 1.22 for WLOP. As there are no changes to the SDLs at Surry, item 2 is satisfied.

A deterministic SAL equal to 1.40 has been selected for both the ABB-NV and WLOP CHF correlations at Surry for use in the analyses of the Westinghouse 15x15 Upgrade fuel design. The DDLs are fixed and any changes to their values would require NRC review and approval. However, the SALs for DNB analyses may be changed without Page 7 of 18

Serial No.13-145 Docket Nos. 50-280/281 and 50-338/339 Attachment 1 prior NRC review and approval provided the changes meet the criteria established in Reference 10.

The difference between the SAL and the DDL is the available retained DNBR margin:

Retained DNBR Margin [ = \SALDDL*100

(]SAL- *10 The resulting available retained DNBR margin is 18.5% for ABB-NV and 12.8% for WLOP.

This method of defining retained DNBR margin allows for all of the DNBR margin to be found in a single, clearly defined location. The retained DNBR margin can be used to offset generic DNBR penalties such as rod bow. The retained DNBR margin may also be used to offset potential DNBR penalties such as transition core penalties. The reload thermal-hydraulic evaluations prepared as part of the reload safety analysis process evaluates the applicable DNB penalties and confirms positive DNBR margin exists for the given cycle. Retained DNBR margin is tracked separately for each DNB limiting accident and applicable CHF correlation.

In determining the impact of the use of the W-3 Alternate CHF correlations on the reactor protection system and the list of UFSAR transients for which the code and correlations apply, it is important to understand how the W-3 Alternate CHF correlations will be applied at Surry. As discussed in Section D.2 of Appendix D of DOM-NAF-2 and Section 2.1 above, the ABB-NV CHF correlation is to be applied to UFSAR transients where the limiting location of DNBR occurs below the first MVG and the WLOP CHF correlation is to be applied to UFSAR transients when the conditions occur outside of the range of applicability of the primary CHF correlation, specifically low pressure or low flow conditions. This is consistent with the current application of the W-3 CHF correlation at Surry.

Dominion verified acceptable DNB performance of the Westinghouse 15x15 Upgrade fuel design at Surry with the VIPRE-D/ABB-NV and VIPRE-D/WLOP code/correlation pairs using the DDLs qualified in Appendix D of DOM-NAF-2. The results of these calculations demonstrate that the minimum DNBR values are equal to or greater than the applicable SAL for all of the analyses. These calculations support the current RCSLs; the OTAT and OPAT trip functions (including FAI reset functions); FAH limits; as well as the evaluated Chapter 14 events. As such there are no changes to the OTAT, OPAT, FAH, reactor protection system, or the RCSLs due to the implementation of the W-3 Alternate CHF correlations at Surry. This satisfies item 3.

Page 8 of 18

Serial No.13-145 Docket Nos. 50-280/281 and 50-338/339 Attachment 1 The two notable UFSAR transients where the W-3 Alternate CHF correlation is limiting with regards to DNB performance in licensing bases calculations are the RWSC and MSLB accidents. The RWSC event occurs within the applicable range of the primary CHF correlation with the exception of the limiting location of DNBR potentially being below the first MVG. Thus, the ABB-NV CHF correlation is applied to the RWSC transient. The limiting DNBR for the MSLB event occurs under low pressure conditions and is outside the applicable range of the primary CHF correlation. Thus, the WLOP CHF correlation is applied to the MSLB event. The other Chapter 14 UFSAR transients are limited by the primary CHF correlation and are not affected by the implementation of the W-3 Alternate CHF correlations. There is no change to the list of UFSAR transients analyzed with VIPRE-D. This satisfies item 4.

4.4 Increase in the Surry TS Minimum Temperature for Criticality To implement the increased Minimum Temperature for Criticality at Surry, Dominion proposes a revision to Surry TS 3.1.E.4 to reflect the proposed limit of 538 0F. In addition to this change, a revision to TS 3.1..B, [RCS] Heatup and Cooldown is also required, as the Minimum Temperature for Criticality limit is also included in this TS.

Descriptions of the required TS changes are provided in Section 3.

The Minimum Temperature for Criticality provides additional assurance that the assumptions made in the safety analyses remain bounding by maintaining the moderator temperature within the range of those analyses. The Minimum Temperature for Criticality is taken into consideration for the UFSAR Chapter 14 Safety Analyses in the reload verification of the reactivity parameters assumed in the safety analyses. The reload verification of reactivity parameters are performed at conservative conditions to ensure that the reactivity parameters assumed in the safety analyses bound the cycle design. The reload verification includes allowances for instrumentation uncertainty associated with temperature indication.

Following the increase in the Minimum Temperature for Criticality, the reload design analytical calculations will continue to demonstrate that the reload core design is bounded by the UFSAR Chapter 14 Safety Analyses. The increased Minimum Temperature for Criticality does not impact the NRC approved analytical methods used to determine the core operating limits, such as the moderator temperature coefficient.

Page 9 of 18

Serial No.13-145 Docket Nos. 50-280/281 and 50-338/339 Attachment 1 The increase in the TS Minimum Temperature for Criticality slightly decreases the operational margin associated with a Reactor Coolant System (RCS) cold water insertion transient that could cool the RCS temperature below the increased limit.

However, there is no increase in the probability of the initiating events, and the margin to the limit remains sufficient for anticipated occurrences. Should the RCS temperature fall below the proposed limit, the unit would be in an abnormal condition requiring operator action. Additionally, the approach to criticality at Surry is a very focused, methodical, and deliberate activity as governed by Surry startup procedures. Reactivity changes and overall plant response are closely monitored by the Operations staff.

Therefore, it is expected that Surry will have no difficulty staying above the increased TS Minimum Temperature for Criticality limit of 538 0F.

To be consistent with the increased Minimum Temperature for Criticality in TS 3.1.E.4, an update to TS 3.1.B.1 is also required. TS 3.1.B.1 restricts the RCS pressure and temperature to protect against non-ductile fracture of the reactor vessel during heatup and cooldown. The proposed change to TS 3.1 .B would reference the applicable TS for the Minimum Temperature for Criticality (TS 3.1.E.4) versus providing the 538 0F limit.

This proposed change is consistent with the equivalent Standard Technical Specification for RCS heatup and cooldown (TS 3.4.3) for Westinghouse Plants provided in NUREG-1431 (Reference 11). The Standard Basis of TS 3.4.3 of NUREG-1431 states, "Other Technical Specifications provide limits for operation that can be more restrictive than or can supplement these P/T limits." The Minimum Temperature for Criticality is one of the TS identified as potentially more limiting. The proposed revision to TS 3.1..B.1 is provided in Section 3 and reflects the specification provided in NUREG-1431.

5. REGUALTORY EVALUATION A Regulatory Evaluation has been performed for the proposed changes to the North Anna and Surry TS. The proposed TS changes are:

1) incorporation of Appendix D of Fleet Report DOM-NAF-2 to the North Anna TS,

2) incorporation of Appendix D of Fleet Report DOM-NAF-2 to the Surry TS and

3) increase to the Surry TS Minimum Temperature for Criticality.

Page 10 of 18

Serial No.13-145 Docket Nos. 50-280/281 and 50-338/339 Attachment 1 5.1 Applicable Regulatory Requirements/Criteria 5.1.1 Application of DOM-NAF-2 Appendix D to North Anna and Surry Section 2.1 of Fleet Report DOM-NAF-2-A lists the information to be provided to the NRC by Dominion for review and approval of any plant specific application of the VIPRE-D code:

1) The addition of DOM-NAF-2-A and the applicable appendices to the plant's reference list of the COLR,

2) The identification of any changes to SDLs for the relevant code and correlation(s),

3) The identification of any TS changes to OTAT, OPAT, FAH, or reactor protection function, as well as any changes to RCSLs and,

4) The identification of any changes to the list of UFSAR transients for which the code and correlations apply.

North Anna Technical Specification 2.1, "Safety Limits," states that "The departure from nucleate boiling ratio (DNBR) shall be maintained greater than or equal to the 95/95 DNBR criterion for the DNB correlations and methodologies specified in Section 5.6.5

[COLRI."

Surry Technical Specification 2.1, "Safety Limit, Reactor Core" states that "For transients analyzed using the deterministic methodology, the DNBR shall be maintained greater than or equal to the applicable DNB correlation limit."

In accordance with the requirements of DOM-NAF-2-A, Dominion is seeking NRC approval of the inclusion of DOM-NAF-2, Appendix D, to the North Anna Technical Specification 5.6.5.b and the Surry Technical Specification 6.2.C list of NRC approved methodologies used to determine core operating limits (i.e., the reference list of the North Anna and Surry COLRs). This change allows Dominion to use the VIPRE-D/ABB-NV and VIPRE-D/WLOP code/correlation pairs to perform licensing calculations for North Anna and Surry, using the DDLs qualified in Appendix D of DOM-NAF-2.

The ABB-NV and WLOP CHF correlations will not be used statistically in licensing basis calculations at North Anna and Surry. Acceptable DNB performance of the Westinghouse RFA-2 fuel at North Anna and the Westinghouse 15x15 Upgrade fuel design at Surry with the W-3 Alternate CHF correlations has been demonstrated.

Therefore, the existing Reactor Core Safety Limits, OTAT and OPAT trip functions Page 11 of 18

Serial No.13-145 Docket Nos. 50-280/281 and 50-338/339 Attachment 1 (including FAI reset functions), and FAH limits remain bounding and are not being changed as part of this package. The list of applicable UFSAR transients for which the code and correlations apply are also not being changed as part of this package. Thus, the four conditions of DOM-NAF-2-A have been met for the implementation of Appendix D at North Anna and Surry.

Based on this information, Dominion concludes the inclusion of DOM-NAF-2, Appendix D, in the North Anna and Surry COLR list of NRC approved methodologies meets the regulatory requirements and criteria.

5.1.2 Increase in the Surry TS Minimum Temperature for Criticality There are four requirements in Surry TS 3.1.E Minimum Temperature for Criticality that must be met prior to taking a unit at Surry critical. The four requirements are:

1. The maximum positive Moderator Temperature Coefficient limit is met,

2. The reactor is not made critical below the temperature corresponding to the non-ductile failure of the reactor vessel,

3. The reactor is subcritical by an amount equal to or greater than the potential reactivity insertion due to primary coolant depressurization when the RCS is below the temperature corresponding to non-ductile failure of the reactor vessel, and

4. The reactor is not taken critical below a given RCS temperature (currently 522-F).

The fourth item above is the TS limit for which Dominion is requesting an increase to 538 0 F. The impact of this TS change on the remaining three requirements is considered in the following paragraphs.

During the early part of a fuel cycle, the MTC may be slightly positive at RCS temperatures in the power operating range. The MTC is most positive in the earlier portion of the cycle when RCS boron concentration is highest. Throughout the cycle, the MTC must be less than the MTC limit specified in the COLR for the cycle at all times. This requirement remains regardless of the allowable RCS criticality temperature range. Thus, the MTC COLR limit will continue to be met at the proposed Minimum Temperature for Criticality limit of 538 0 F.

The increase in the Minimum Temperature for Criticality is conservative with regard to non-ductile failure as it increases the temperature at which the reactor is required to be critical, thus providing additional margin to non-ductile failure. The RCS heatup and Page 12 of 18

Serial No.13-145 Docket Nos. 50-280/281 and 50-338/339 Attachment 1 cooldown rates that protect against non-ductile failure are provided in Surry TS 3.1.B.

The proposed change does not affect the allowable RCS heatup and cooldown rates.

The reactor must maintain a specified Shutdown Reactivity Margin to preclude the possibility of an accidental criticality resulting from a change in RCS temperature or pressure. This requirement will remain regardless of the allowable RCS criticality temperature range. Thus, the requirement will continue to be met at the proposed Minimum Temperature for Criticality limit of 538°F.

The proposed increase to the Surry Minimum Temperature for Criticality has been evaluated to determine whether applicable regulations and requirements continue to be met. Dominion has determined only NRC approval of the affected TSs to reflect the higher Minimum Temperature for Criticality limit is required for implementation.

Pursuant to 10 CFR 50.90, Dominion is seeking NRC approval of a revision to Surry TS 3.1.E.4 to increase the Minimum Temperature for Criticality limit from 522 0 F to 538 0 F.

Dominion is also requesting a revision to TS 3.1.B.1 to reference the applicable specification for the Minimum Temperature for Criticality, TS 3.1.E, versus providing the limit within TS 3.1.B.1. Implementation of the increased Minimum Temperature for Criticality will provide additional margin in verifying that a given cycle's design meets the allowed most-positive MTC.

The increased Minimum Temperature for Criticality will continue to be verified against the assumptions in the safety analyses on a reload basis and does not impact the NRC approved analytical methods used to determine the core operating limits such as the MTC. RCS temperature TS requirements will continue to be met at the proposed Minimum Temperature for Criticality limit of 538 0 F. Additionally, the change to TS 3.1.B.1 discussed in Section 4.2 is consistent with the Standard Technical Specifications for Westinghouse Plants (Reference 11).

Based on this information, Dominion concludes the increase of the Surry TS 3.1.E.4 Minimum Temperature for Criticality meets the regulatory requirements.

5.2 Determination of No Significant Hazards Consideration Virginia Electric and Power Company (Dominion) proposes changes to the North Anna Power Station (North Anna) Units 1 and 2 and Surry Power Station (Surry) Units 1 and 2 Technical Specifications (TS) pursuant to 10 CFR 50.90.

The first proposed change adds an additional appendix (Appendix D) to Fleet Report DOM-NAF-2 currently listed in North Anna TS 5.6.5.b as Reference 16. The proposed change would allow Dominion to use the VIPRE-D/ABB-NV and VIPRE-D/WLOP Page 13 of 18

Serial No.13-145 Docket Nos. 50-280/281 and 50-338/339 Attachment 1 code/correlation pairs to perform licensing calculations of 17x17 fuel in North Anna cores, using the deterministic design limits (DDLs) documented in Appendix D of the Fleet Report DOM-NAF-2 and improves Dominion's thermal-hydraulic predictive capabilities.

The second proposed change adds an additional appendix (Appendix D) to Fleet Report DOM-NAF-2 currently listed in Surry TS 6.2.C as Reference 8. The proposed change would allow Dominion to use the VIPRE-D/ABB-NV and VIPRE-D/WLOP code/correlation pairs to perform licensing calculations of 15x15 fuel in Surry cores, using the DDLs documented in Appendix D of Fleet Report DOM-NAF-2 and improves Dominion's thermal-hydraulic predictive capabilities. Incorporation of Appendix D into Surry TS 6.2.C also requires the addition of the VIPRE-D/ABB-NV DDL to TS 2.1 .A.1 .a.

The third proposed change increases the Surry TS 3.1.E.4 Minimum Temperature for Criticality limit from 522 0 F to 538 0 F. Increasing the TS Minimum Temperature for Criticality also requires modification to TS 3.1..B.1 to reflect the revised Minimum Temperature for Criticality limit. When implemented with Appendix D of DOM-NAF-2, these two changes provide increased flexibility during loading pattern development as well as improved design margins at Surry.

In accordance with the criteria set forth in 10 CFR 50.92, Dominion has evaluated the proposed TS changes and determined that the changes do not represent a significant hazards consideration. The following is provided in support of this conclusion:

1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The first and second proposed changes would allow Dominion to use the VIPRE-D/ABB-NV and VIPRE-D/WLOP code/correlation pairs to perform licensing calculations for North Anna and Surry, using the DDLs documented in Appendix D of Fleet Report DOM-NAF-2. Neither code/correlation pair methodology makes any contribution to the potential accident initiators and thus cannot increase the probability of any accident. Further, since the DDLs for ABB-NV and WLOP meet the required design basis of avoiding departure from nucleate boiling (DNB) with 95% probability at a 95% confidence level, the use of the new code/correlations does not increase the potential consequences of any accident. The pertinent evaluations that need to be performed as part of the cycle specific reload safety analysis to confirm that the existing safety analyses remain applicable have been performed and determined to be acceptable. The use of a different code/correlation pair will not increase the probability of an accident because plant systems will not be Page 14 of 18

Serial No.13-145 Docket Nos. 50-280/281 and 50-338/339 Attachment 1 operated in a different manner, and system interfaces will not change. The use of the VIPRE-D/ABB-NV and VIPRE-D/WLOP code/correlation pairs to perform licensing calculations for North Anna and Surry will not result in a measurable impact on normal operating plant releases and will not increase the predicted radiological consequences of accidents postulated in the Updated Final Safety Analysis Report (UFSAR). Therefore, neither the probability of occurrence nor the consequences of any accident previously evaluated is significantly increased.

The third proposed change, an increase of the Surry Minimum Temperature for Criticality limit from 522 0F to 538 0 F, would provide Dominion with increased flexibility during loading pattern development as well as improved design margins when coupled with the second proposed change. The Minimum Temperature for Criticality is used within the reload verification process to ensure the assumptions made in the safety analysis remain bounding for the given cycle design. With implementation of the proposed change, the reload design and licensing requirements will remain in place and continue to be met at the increased Minimum Temperature for Criticality limit.

The increase in the Surry Minimum Temperature for Criticality limit will not increase the probability of an accident because plant systems will not be operated in a different manner, and system interfaces will not change. Should the reactor coolant system (RCS) temperature fall below the proposed limit, the unit would be in an abnormal condition requiring operator action. The operator actions are not changing as a result of the increased Minimum Temperature for Criticality limit. The increase in the Surry Minimum Temperature for Criticality will not result in a measurable impact on normal operating plant releases and will not increase the predicted radiological consequences of accidents postulated in the UFSAR. Therefore, neither the probability of occurrence nor the consequences of any accident previously evaluated is significantly increased.

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

Response: No.

The proposed changes do not involve a physical alteration of the plant (no new or different type of equipment will be installed).

The use of the VIPRE-D/ABB-NV and VIPRE-D/WLOP code/correlation pairs and the applicable fuel design limits for DNB ratio (DNBR) does not impact any of the applicable design criteria and the pertinent licensing basis criteria will continue to be Page 15 of 18

Serial No.13-145 Docket Nos. 50-280/281 and 50-338/339 Attachment 1 met. Demonstrated adherence to these standards and criteria precludes new challenges to components and systems that could introduce a new type of accident.

Setpoint safety analysis evaluations have demonstrated that the use of the VIPRE-D/ABB-NV and VIPRE-D/WLOP code/correlation pairs is acceptable. Design and performance criteria will continue to be met, and no new single failure mechanisms will be created. The use of the VIPRE-D/ABB-NV and VIPRE-D/WLOP code/correlation pairs does not involve any alteration to plant equipment or procedures that would introduce any new or unique operational modes or accident precursors.

The increase in the Surry Minimum Temperature for Criticality does not result in any plant design changes. In addition, the minimum temperature at which the reactor is taken critical is not an accident initiator. The nominal average reactor coolant system temperature during an approach to criticality is several degrees higher than the limit proposed for the Minimum Temperature for Criticality.

Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously analyzed.

3. Does this change involve a significant reduction in a margin of safety?

Response: No.

The first two proposed changes would allow Dominion to use the VIPRE-D/ABB-NV and VIPRE-D/WLOP code/correlation pairs to perform licensing calculations for North Anna and Surry using the DDLs documented in Appendix D of Fleet Report DOM-NAF-2. North Anna TS 2.1, "Safety Limits," states that "The departure from nucleate boiling ratio (DNBR) shall be maintained greater than or equal to the 95/95 DNBR criterion for the DNB correlations and methodologies specified in Section 5.6.5 [COLR]." The DNBR limits meet the design basis of avoiding DNB with 95%

probability at a 95% confidence level. Surry TS 2.1, "Safety Limits, Reactor Core,"

specifies that "for transients analyzed using the deterministic methodology, the DNBR shall be maintained greater than or equal to the applicable DNB correlation limit." The required DNBR margin of safety for North Anna and Surry, which in this case is the margin between the 95/95 DNBR limit and clad failure, is therefore not reduced. Therefore, the proposed TS changes do not involve a significant reduction in a margin of safety.

The increased Minimum Temperature for Criticality in conjunction with the appropriate core designs will ensure the current TS limits for the most positive moderator temperature coefficient (MTC) will continue to be satisfied. The current analyses are bounding and remain applicable with the increased Minimum Page 16 of 18

Serial No.13-145 Docket Nos. 50-280/281 and 50-338/339 Attachment 1 Temperature for Criticality. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above information, Dominion concludes that the proposed changes present no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

5.3 Environmental Consideration The proposed amendments meet the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9) as follows:

(i) The proposed changes involve no significant hazards consideration.

As described in Section 5.2 above, the proposed changes involve no significant hazards consideration.

(ii) There are no significant changes in the types or significant increase in the amounts of any effluents that may be released offsite.

The proposed changes do not involve the installation of any new equipment or the modification of any equipment that may affect the types or amounts of effluents that may be released offsite. Therefore, there is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite.

(iii) There is no significant increase in individual or cumulative occupation radiation exposure.

The proposed changes do not involve physical plant changes or introduce any new modes of plant operation. Therefore, there is no significant increase in individual or cumulative occupational radiation exposure.

Based on the above, Dominion concludes that, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendments.

5.4 Regulatory Conclusion Based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by implementation of the proposed TS changes, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

Page 17 of 18

Serial No.13-145 Docket Nos. 50-280/281 and 50-338/339 Attachment 1

6. REFERENCES

1. Topical Report, WCAP-14565-P-A, Rev. 0, Addendum 2-P-A, "Addendum 2 to WCAP-14565-P-A Extended Application of ABB-NV Correlation and Modified ABB-NV Correlation WLOP for PWR Low Pressure Applications,"

April 2008.

2. Letter from C. E. Rossi (USNRC) to J. A. Blaisdell (UGRA Executive Committee), "Acceptance for Referencing of Licensing Topical Report, EPRI NP-2511-CCM, 'VIPRE-01: A Thermal-Hydraulic Analysis Code for Reactor Cores,' Volumes 1,2, 3 and 4," May 1, 1986.

3. Letter from A. C. Thadani (USNRC) to Y. Y. Yung (VIPRE-01 Maintenance Group), "Acceptance for Referencing of the Modified Licensing Topical Report, EPRI NP-2511-CCM, Revision 3, 'VIPRE-01: A Thermal Hydraulic Analysis Code for Reactor Cores,' (TAC No. M79498)," October 30, 1993.

4. Fleet Report, DOM-NAF-2-A, Rev. 0.2-P-A, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code," August 2010.

5. Letter from T. L. Chan (USNRC) to W. L. Stewart (VEPCO), Issuance of Amendment 105 to License No. DPR-32 and Amendment 105 to License No. DPR-37, Dominion Serial No.86-033, dated December 31, 1985.

6. Surry Technical Specifications through Amendment 278/278 (Unit 1/Unit 2).

7. North Anna Technical Specifications through Amendment 269/251 (Unit 1/Unit 2).

8. Kewaunee Technical Specifications through Amendment 210.

9. Millstone Unit 3 Technical Specifications through Amendment 257.

10. Technical Report, NEI 96-07, Revision 1, "Guidelines for 10 CFR 50.59 Implementation," Nuclear Energy Institute, November 2000.

11. NUREG-1431, Revision 4, "Standard Technical Specifications, Westinghouse Plants," Volumes 1 and 2.

Page 18 of 18

Serial No.13-145 Docket Nos. 50-338/339 ATTACHMENT 2 PROPOSED NORTH ANNA TECHNICAL SPECIFICATIONS PAGES (MARK-UP)

Virginia Electric and Power Company (Dominion)

North Anna Power Station Units I and 2

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

b. (continued)

13. EMF-2103 (P)(A), "Realistic Large Break LOCA Methodology for Pressurized Water Reactors."

14. EMF-96-029 (P)(A), "Reactor Analysis System for PWRs."

15. BAW-10168P-A, "RSG LOCA - BWNT Loss-of-Coolant Accident Evaluation Model for Recirculating Steam Generator Plants,"

Volume II only (SBLOCA models).

16. DOM-NAF-2-A, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code," including Appendix A, "Qualification of the F-ANP BWU CHF Correlations in the Dominion VIPRE-D Computer Code,"-afd-Appendix C, "Qualification of the Westinghouse WRB-2M CHF Correlation in the Dominion VIPRE-D Computer CGde."- <

17. WCAP-12610-P-A and CENPD-404-P-A, Addendum 1-A, "Optimized J<

ZIRLO" (Westinghouse Proprietary).

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 PAM Report When a report is required by Condition B of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

Code," and Appendix D, "Qualification of the ABB-NV and WLOP CHF Correlations in the Dominion VIPRE-D Computer Code."

North Anna Units 1 and 2 5.6-4 Amendments 267ý248

Serial No.13-145 Docket Nos. 50-338/339 ATTACHMENT 3 PROPOSED NORTH ANNA TECHNICAL SPECIFICATIONS PAGES (TYPED)

Virginia Electric and Power Company (Dominion)

North Anna Power Station Units I and 2

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

b. (continued)

13. EMF-2103 (P)(A), "Realistic Large Break LOCA Methodology for Pressurized Water Reactors."

14. EMF-96-029 (P)(A), "Reactor Analysis System for PWRs."

15. BAW-10168P-A, "RSG LOCA - BWNT Loss-of-Coolant Accident Evaluation Model for Recirculating Steam Generator Plants,"

Volume II only (SBLOCA models).

16. DOM-NAF-2-A, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code," including Appendix A, "Qualification of the F-ANP BWU CHF Correlations in the Dominion VIPRE-D Computer Code," Appendix C, "Qualification of the Westinghouse WRB-2M CHF Correlation in the Dominion VIPRE-D Computer Code," and Appendix D, "Qualification of the ABB-NV and WLOP CHF Correlations in the Dominion VIPRE-D Computer Code."

17. WCAP-12610-P-A and CENPD-404-P-A, Addendum 1-A, "Optimized ZIRLO" (Westinghouse Proprietary).

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 PAM Report When a report is required by Condition B of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

North Anna Units 1 and 2 5.6-4 Amendments

Serial No.13-145 Docket Nos. 50-280/281 ATTACHMENT 4 PROPOSED SURRY TECHNICAL SPECIFICATIONS AND BASES PAGES (MARK-UP)

(BASES CHANGES FOR NRC INFORMATION)

Virginia Electric and Power Company (Dominion)

Surry Power Station Units I and 2

TS 2.1-1 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMIT, REACTOR CORE Applicability Applies to the limiting combinations of THERMAL POWER, Reactor Coolant System pressure, .4" coolant temperature and coolant flow when a reactor is critical.

Objective To maintain the integrity of the fuel cladding.

Specification A. The combination of reactor THERMAL POWER level, pressurizer pressure, and Reactor Coolant System (RCS) highest loop average temperature shall not:

1. Exceed the limits specified in the CORE OPERATING LIMITS REPORT when full flow from three reactor coolant pumps exists, and the following Safety Limits shall not be exceeded:

a. The design limit for departure from nucleate boiling ratio (DNBR) shall be maintained > 1.27 for transients analyzed using the Statistical DNBR Evaluation Methodology and the WRB-1 DNB correlation. For transients analyzed using the deterministic methodology, the DNBR shall be maintained greater than or equal to the applicable DNB correlation limit (> 1.17 for WRB-1, > 1.30 for

< J W-3, > 1.14 for ABB-NV).

b. The peak fuel centerline temperature shall be maintained < 5080'F, decreasing by 58°F per 10,000 MWD/MTU of burnup.

2. The reactor THERMAL POWER level shall not exceed 118% of rated power.

Amendment Nos. 270 and 269

TS 3.1-6 2-8-~95-B. HEATUP AND COOLDOWN Specification

1. Unit 1 and Unit 2 reactor coolant temperature and pressure and the system heatup and cooldown (with the exception of the pressurizer) shall be limited in accordance with TS Figures 3.1-1 and 3.1-2.

Heatup:

Figure 3.1-1 may be used for heatup rates of up to 60 'F/hr.

Cooldown:

Allowable combinations of pressure and temperature for specific cooldown rates are below and to the right of the limit lines as shown in TS Figure 3.1-2. This rate shall not exceed 100lF/hr. Cooldown rates between those shown can be obtained by interpolation between the curves on Figure 3.1-2.

Core Operation:

During operation where the reactor core is in a critical condition (except for low level physics tests), vessel metal and fluid temperature shall be maintained above the reactor core criticality limits specified in 10 CFR 50 Appendix G. The reactor shall not be made critical when the reactor coolant temperature is below 5222-2s-- specified in T.S. 3.1.E. the Minimum Temperature for Criticality

2. The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the vessel is below 70'F.

Amendment Nos. 2W7 and- 2107

TS 3.1-8

4) The pressurizer heatup and cooldown rates shall not exceed 100F/hr. and 200°F/hr.

respectively. The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 320'F.

Although the pressurizer operates in temperature ranges above those for which there is reason for concern of non-ductile failure, operating limits are provided to assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements.

5) System preservice hydrotests and in-service leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code,Section XI according to the leak test limit line shown in Figure 3.1-1.

6) The reactor shall not be made critical when the reactor coolant temperature is below 522-F ini accUid*wII** witli T1ehiiii, SpecJifiatioun 3. 1.E. .

The fracture toughness properties of the ferritic materials in the reactor vessel are determined in accordance with the NRC Standard Review Plan, ASTM E185-82, and in accordance with additional reactor vessel requirements. These properties are then evaluated in accordance with Appendix G to Section III of the ASME Boiler and Pressure Vessel Code.

the Minimum Temperature for Criticality specified in Technical Specification 3.1 .E.

Amendment Nos. 147 4and 43

TS 3.1-18 0-24l6--

E. Minimum Temperature for Criticality Specifications

1. Except during LOW POWER PHYSICS TESTS, the reactor shall not be made critical at any Reactor Coolant System temperature above which the moderator temperature coefficient is more positive than the limit specified in the CORE OPERATING LIMITS REPORT. The maximum upper limit for the moderator temperature coefficient shall be:

a. + 6 pcm/°F at less than 50% of RATED POWER, or

b. + 6 pcmr°F at 50% of RATED POWER and linearly decreasing to 0 pcm/nF at RATED POWER.

2. In no case shall the reactor be made critical with the Reactor Coolant System temperature below the limiting value of RTNDT + 10°F, where the limiting value of RTNDT is as determined in Part B of this specification.

3. When the Reactor Coolant System temperature is below the minimum temperature as specified in E-2 above, the reactor shall be subcritical by an amount equal to or greater than the potential reactivity insertion due to primary coolant depressurization.

4. The reactor shall not be made critical when the Reactor Coolant System temperature is below 5-2S°F.

Basis t _

During the early part of a fuel cycle, the moderator temperature coefficient may be calculated to be slightly positive at coolant temperatures in the power operating range. The moderator temperature coefficient will be most positive near the beginning of cycle life, generally corresponding to when the boron concentration in the coolant is the greatest. 1 Later in the cycle, the boron concentration in the coolant will generally be lower and the moderator temperature coefficient will be less positive or will be negative in the power 4 operating range. At the beginning of cycle life, during pre-operational physics tests, measurements are made to determine that the moderator temperature coefficient is less A,'

than the limit specified in the CORE OPERATING LIMITS REPORT.

Amendment Nos. Bae-

TS 3.1-19 The requirement that the reactor is not to be made critical when the moderator coefficient is greater than the low power limit specified in the CORE OPERATING LIMITS REPORT has been imposed to prevent any unexpected power excursion during normal operations as a result of either an increase of moderator temperature or decrease of coolant pressure. This requirement is waived during LOW POWER PHYSICS TESTS to permit measurement of reactor moderator coefficient and other physics design parameters of interest. During physics tests, special operation precautions will be taken. In addition, the strong negative Doppler coefficient(2'( 3 )

and the small integrated Delta k/k would limit the magnitude of a power excursion resulting from a reduction of moderator density.

The requirement that the reactor is not to be made critical with a Reactor Coolant System temperature below the limiting value of RTNDT + 10F provides increased assurance that the proper relationship between Reactor Coolant System pressure and temperature will be maintained during system heatup and pressurization whenever the reactor vessel is in the nil ductility transition temperature range. Heatup to this temperature is accomplished by operating the reactor coolant pumps.

The requirement that the reactor is not to be made critical with a Reactor Coolant System temperature below 65--2-'F provides added assurance that the assumptions 538 made in the safety analyses remain bounding by maintaining the moderator temperature within the range of those analyses.

If a specified shutdown reactivity margin is maintained (TS Section 3.12), there is no possibility of an accidental criticality as a result of an increase of moderator temperature or a decrease of coolant pressure.

(1) UFSAR Figure 3.3-8 (2) UFSAR Table 3.3-1 (3) UFSAR Figure 3.3-9 Amendment Nos. 207 and TS 6.2-2 The analytical methods used to determine the core operating limits identified above shall be those previously reviewed and approved by the NRC, and identified below.

The CORE OPERATING LIMITS REPORT will contain the complete identification for each of the TS referenced topical reports used to prepare the CORE OPERATING LIMITS REPORT (i.e., report number, title, revision, date, and any supplements). The core operating limits shall be determined so that 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. The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided for information for each reload cycle to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector. Code," and Appendix D, "Qualification of the ABB-NV and WLOP CHF Correlations in the REFERENCES Dominion VIPRE-D Computer Code"

1. VEP-FRD-42-A, "Reload Nuclear Design Methodology"

2. WCAP-16009-P-A, "Realistic Large Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM),"

(Westinghouse Proprietary).

3. WCAP-10054-P-A, "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," (W Proprietary)

4. WCAP-10079-P-A, "NOTRUMP, A Nodal Transient Small Break and General Network Code," (W Proprietary)

5. WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Report," (Westinghouse Proprietary)

6. VEP-NE-2-A, "Statistical DNBR Evaluation Methodology"

7. VEP-NE-3-A, "Qualification of the WRB-1 CHF Correlation in the Virginia Power COBRA Code"

8. DOM-NAF-2-A, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code," including Appendix B, "Qualification of the Westinghouse WRB-1 CHF Correlation in the Dominion VIPRE-D Computer-6ole- <

9. WCAP-8745-P-A, "Design Bases for Thermal Overpower Delta-T and Thermal Overtemperature Delta-T Trip Function."

10. WCAP-12610-P-A and CENPD-404-P-A, Addendum 1-A, "Optimized ZIRLO,"

(Westinghouse Proprietary)

Amendment Nos. 2:74 and 270

Serial No.13-145 Docket Nos. 50-280/281 ATTACHMENT 5 PROPOSED SURRY TECHNICAL SPECIFICATIONS AND BASES PAGES (TYPED)

(BASES CHANGES FOR NRC INFORMATION)

Virginia Electric and Power Company (Dominion)

Surry Power Station Units 1 and 2

TS 2.1-1 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMIT, REACTOR CORE Applicability Applies to the limiting combinations of THERMAL POWER, Reactor Coolant System pressure, coolant temperature and coolant flow when a reactor is critical.

Objective To maintain the integrity of the fuel cladding.

Specification A. The combination of reactor THERMAL POWER level, pressurizer pressure, and Reactor Coolant System (RCS) highest loop average temperature shall not:

1. Exceed the limits specified in the CORE OPERATING LIMITS REPORT when full flow from three reactor coolant pumps exists, and the following Safety Limits shall not be exceeded:

a. The design limit for departure from nucleate boiling ratio (DNBR) shall be maintained > 1.27 for transients analyzed using the Statistical DNBR Evaluation Methodology and the WRB-I DNB correlation. For transients analyzed using the deterministic methodology, the DNBR shall be maintained greater than or equal to the applicable DNB correlation limit (> 1.17 for WRB-1, > 1.30 for W-3, > 1.14 for ABB-NV).

b. The peak fuel centerline temperature shall be maintained < 5080'F, decreasing by 58°F per 10,000 MWD/MTU of burnup.

2. The reactor THERMAL POWER level shall not exceed 118% of rated power.

Amendment Nos.

TS 3.1-6 B. HEATUP AND COOLDOWN Specification 1 Unit 1 and Unit 2 reactor coolant temperature and pressure and the system heatup and cooldown (with the exception of the pressurizer) shall be limited in accordance with TS Figures 3.1-1 and 3.1-2.

Heatup:

Figure 3.1-1 may be used for heatup rates of up to 60°F/hr.

Cooldown:

Allowable combinations of pressure and temperature for specific cooldown rates are below and to the right of the limit lines as shown in TS Figure 3.1-2. This rate shall not exceed 100lF/hr. Cooldown rates between those shown can be obtained by interpolation between the curves on Figure 3.1-2.

Core Operation:

During operation where the reactor core is in a critical condition (except for low level physics tests), vessel metal and fluid temperature shall be maintained above the reactor core criticality limits specified in 10 CFR 50 Appendix G. The reactor shall not be made critical when the reactor coolant temperature is below the Minimum Temperature for Criticality specified in T.S. 3.1 .E.

2. The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the vessel is below 70'F.

Amendment Nos.

TS 3.1-8

4) The pressurizer heatup and cooldown rates shall not exceed 100F/hr. and 200°F/hr.

respectively. The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 320'F.

Although the pressurizer operates in temperature ranges above those for which there is reason for concern of non-ductile failure, operating limits are provided to assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements.

5) System preservice hydrotests and in-service leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code,Section XI according to the leak test limit line shown in Figure 3.1-1.

6) The reactor shall not be made critical when the reactor coolant temperature is below the Minimum Temperature for Criticality specified in Technical Specification 3.1 .E.

The fracture toughness properties of the ferritic materials in the reactor vessel are determined in accordance with the NRC Standard Review Plan, ASTM E185-82, and in accordance with additional reactor vessel requirements. These properties are then evaluated in accordance with Appendix G to Section III of the ASME Boiler and Pressure Vessel Code.

Amendment Nos.

TS 3.1-18 E. Minimum Temperature for Criticality Specifications

1. Except during LOW POWER PHYSICS TESTS, the reactor shall not be made critical at any Reactor Coolant System temperature above which the moderator temperature coefficient is more positive than the limit specified in the CORE OPERATING LIMITS REPORT. The maximum upper limit for the moderator temperature coefficient shall be:

a. + 6 pcm/IF at less than 50% of RATED POWER, or

b. + 6 pcm/0 F at 50% of RATED POWER and linearly decreasing to 0 pcm/IF at RATED POWER.

2. In no case shall the reactor be made critical with the Reactor Coolant System temperature below the limiting value of RTNDT + 10'F, where the limiting value of RTNDT is as determined in Part B of this specification.

3. When the Reactor Coolant System temperature is below the minimum temperature as specified in E-2 above, the reactor shall be subcritical by an amount equal to or greater than the potential reactivity insertion due to primary coolant depressurization.

4. The reactor shall not be made critical when the Reactor Coolant System temperature is below 538'F.

Basis During the early part of a fuel cycle, the moderator temperature coefficient may be calculated to be slightly positive at coolant temperatures in the power operating range. The moderator temperature coefficient will be most positive near the beginning of cycle life, generally corresponding to when the boron concentration in the coolant is the greatest.

Later in the cycle, the boron concentration in the coolant will generally be lower and the moderator temperature coefficient will be less positive or will be negative in the power operating range. At the beginning of cycle life, during pre-operational physics tests, measurements are made to determine that the moderator temperature coefficient is less than the limit specified in the CORE OPERATING LIMITS REPORT.

Amendment Nos.

TS 3.1-19 The requirement that the reactor is not to be made critical when the moderator coefficient is greater than the low power limit specified in the CORE OPERATING LIMITS REPORT has been imposed to prevent any unexpected power excursion during normal operations as a result of either an increase of moderator temperature or decrease of coolant pressure. This requirement is waived during LOW POWER PHYSICS TESTS to permit measurement of reactor moderator coefficient and other physics design parameters of interest. During physics tests, special operation 3

precautions will be taken. In addition, the strong negative Doppler coefficient(2)()

and the small integrated Delta k/k would limit the magnitude of a power excursion resulting from a reduction of moderator density.

The requirement that the reactor is not to be made critical with a Reactor Coolant System temperature below the limiting value of RTNDT + 10F provides increased assurance that the proper relationship between Reactor Coolant System pressure and temperature will be maintained during system heatup and pressurization whenever the reactor vessel is in the nil ductility transition temperature range. Heatup to this temperature is accomplished by operating the reactor coolant pumps.

The requirement that the reactor is not to be made critical with a Reactor Coolant System temperature below 538°F provides added assurance that the assumptions made in the safety analyses remain bounding by maintaining the moderator temperature within the range of those analyses.

If a specified shutdown reactivity margin is maintained (TS Section 3.12), there is no possibility of an accidental criticality as a result of an increase of moderator temperature or a decrease of coolant pressure.

(1) UFSAR Figure 3.3-8 (2) UFSAR Table 3.3-1 (3) UFSAR Figure 3.3-9 Amendment Nos.

TS 6.2-2 The analytical methods used to determine the core operating limits identified above shall be those previously reviewed and approved by the NRC, and identified below.

The CORE OPERATING LIMITS REPORT will contain the complete identification for each of the TS referenced topical reports used to prepare the CORE OPERATING LIMITS REPORT (i.e., report number, title, revision, date, and any supplements). The core operating limits shall be determined so that 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. The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided for information for each reload cycle to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

REFERENCES

1. VEP-FRD-42-A, "Reload Nuclear Design Methodology"

2. WCAP-16009-P-A, "Realistic Large Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM),"

(Westinghouse Proprietary).

3. WCAP-10054-P-A, "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," (W Proprietary)

4. WCAP-10079-P-A, "NOTRUMP, A Nodal Transient Small Break and General Network Code," (W Proprietary)

5. WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Report," (Westinghouse Proprietary)

6. VEP-NE-2-A, "Statistical DNBR Evaluation Methodology"

7. VEP-NE-3-A, "Qualification of the WRB-1 CHF Correlation in the Virginia Power COBRA Code"

8. DOM-NAF-2-A, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code," including Appendix B, "Qualification of the Westinghouse WRB-1 CHF Correlation in the Dominion VIPRE-D Computer Code," and Appendix D, "Qualification of the ABB-NV and WLOP CHF Correlations in the Dominion VIPRE-D Computer Code"

9. WCAP-8745-P-A, "Design Bases for Thermal Overpower Delta-T and Thermal Overtemperature Delta-T Trip Function"

10. WCAP- 12610-P-A and CENPD-404-P-A, Addendum 1-A, "Optimized ZIRLO,"

(Westinghouse Proprietary)

Amendment Nos.