ML102020165

Virginia Electric and Power Company North Anna Power Station Units 1 and 2 Proposed License Amendment Request (LAR) Addition of Analytical Methodology to COLR

ML102020165
Person / Time
Site:	North Anna
Issue date:	07/19/2010
From:	Price J Virginia Electric & Power Co (VEPCO)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML102020165 (43)

Text

VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 July 19, 2010 10CFR50.90 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555 VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNITS 1 AND 2 PROPOSED LICENSE AMENDMENT REQUEST(LAR)

ADDITION OF ANALYTICAL METHODOLOGY TO COLR Serial No.

NL&OS/ETS Docket Nos.

License Nos.10-404 RO 50-338/339 NPR-4/7 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 NPR-4 and NPR-7 for North Anna Power Station Units 1 and 2, respectively.

The proposed LAR requests the inclusion of NRC approved Appendix C of Dominion Fleet Report DOM-NAF-2-A, "Qualification of the Westinghouse WRB-2M CHF Correlation in the Dominion VIPRE-D Computer Code," to the Technical Specification 5.6.5.b, as a referenced analytical methodology.

Furthermore, plant specific application of the methodology requires approval of the Statistical Design Limit (SDL) for the relevant code/correlation pair.

Consequently, in addition to the inclusion of Fleet Report DOM-NAF-2-A, including Appendix C, in TS 5.6.5.b, Dominion also requests NRC review and approval of the implementation of the Dominion Topical Report VEP-NE-2-A, "Statistical DNBR Evaluation Methodology," for Westinghouse RFA-2 fuel at North Anna using the VIPRE-D/WRB-2M code/correlation pair, as well as the SDL obtained by this implementation. Pursuant to 10 CFR 50.59, the SDL discussed in Attachment 4 requires NRC review and approval since the SDL establishes a Design Basis Limit for Fission Product Barrier (DBLFPB).

Upon approval of this amendment request and the SDL documented in Attachment 4, Dominion will be capable of performing in-house DNB analyses for the intended uses described in DOM-NAF-2-A using VIPRE-D to support the use of Westinghouse RFA-2 fuel at North Anna Units 1 and 2.

Dominion is currently planning to use Westinghouse RFA-2 fuel in North Anna Units 1 and 2 commencing with North Anna Unit 1, Cycle 23 (Spring 2012) and North Anna Unit 2, Cycle 23 (Spring 2013).

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, respectively. Attachment 4 provides the technical basis for:

1) adding Appendix C of Dominion Fleet Report DOM-NAF-2-A to the list of USNRC approved methodologies for determining core operating limits, 2) the implementation of the Dominion Statistical DNBR Evaluation Methodology for Westinghouse RFA-2 fuel at North Anna with the VIPRE-D/WRB-2M code/correlation pair, and 3) the SDL obtained by this implementation.

Serial No.10-404 Docket Nos. 50-338/339 Page 2 of 3 We have evaluated the proposed amendment and have determined that it does 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 the proposed change 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 amendment is 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 change. The basis for our determination is also included in Attachment 1. The proposed amendment has been reviewed and approved by the Facility Safety Review Committee.

Approval of the proposed amendments is requested by July 21, 2011.

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, J.

la Price Vi e 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 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 My Commission Expires:

~ - - -

GINGER LYNN MELTON Notary Public Commonwtalth of Virginia 310847 My Commission expire. Apr 30.2013

~

Serial No.10-404 Docket Nos. 50-338/339 Page 3 of 3 Attachments:

1. Discussion of Change 2.

Proposed Technical Specifications Pages (Mark-Up) 3.

Proposed Technical Specifications Pages (Typed)

4. Technical Basis for Adding Appendix C of Fleet Report DOM-NAF-2-A to the List of USNRC Approved Methodologies for Determining Core Operating Parameters Commitments made in this letter:

None 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 Ms. K. R. Cotton 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 - i h Floor 109 Governor Street Room 730 Richmond, Virginia 23219

ATTACHMENT 1 DISCUSSION OF CHANGE Virginia Electric and Power Company (Dominion)

North Anna Power Station Units 1 and 2 Serial No.10-404 Docket Nos. 50-338/339

Serial No.1 0*404 Proposed License Amendment Addition of Analytical Method to COLR DISCUSSION OF CHANGE

1.0 INTRODUCTION

Virginia Electric and Power Company (Dominion) proposes to modify the North Anna Power Station Technical Specifications to include Fleet Report DOM NAF-2-A, Appendix C, "Qualification of the Westinghouse WRB 2M Critical Heat Flux (CHF) Correlation in the Dominion VIPRE-D Computer Code," to the Technical Specification S.6.S.b list of USNRC approved methodologies used to determine core operating limits {Le., the reference list of the North Anna Core Operating Limits Report (COLR)}.

Approval of this change will allow Dominion to use the VIPRE-DIWRB-2M and VIPRE-DIW-3 code/correlation pairs to perform licensing calculations of Westinghouse Robust Fuel Assembly - 2 (RFA-2) fuel in North Anna cores, using the deterministic design limits (DDLs) documented in Appendix C of the DOM-NAF-2-A Fleet Report and the statistical design limit (SDL) documented in Attachment 4.

The DDLs were approved as part of the review and approval of DOM-NAF-2-A Appendix C (Reference 1).

Dominion requests the NRC review and approval of the SDL documented herein, consistent with 10 CFR SO.S9(c)(2)(vii) the change establishes a Design Basis Limit for Fission Product Barrier (DBLFPB).

The proposed technical specification change has 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 change qualifies for categorical exclusion from an environmental assessment as set forth in 10 CFR S1.22(c)(9); therefore, no environmental impact statement or environmental assessment is needed in connection with the approval of the proposed technical specification change.

2.0 BACKGROUND

Dominion plans to purchase fuel assemblies from Westinghouse for use at North Anna Power Station, Units 1 and 2. These assemblies are planned to be inserted in Units 1 and 2, commencing with Cycle 23 for both units. The Westinghouse 17x17 RFA-2 fuel product is a replacement for the resident fuel product, which is the AREVA Advanced Mark-BW (AMBW).

The Westinghouse RFA-2 fuel product contains modified mid-grids and modified intermediate flow mixer grids (IFMs).

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 critical heat flux (CHF) and Departure from Nucleate Boiling Ratio (DNBR) of reactor cores. VIPRE-01 has been approved by the U. S. Nuclear Regulatory Commission (USNRC) (References 3 and 4). VIPRE-D is based upon VIPRE-01, MOD-02.1, and is customized by Dominion to fit the specific needs of Dominion's nuclear plants and fuel products.

In April 2006, Dominion obtained generic approval of Fleet Report DOM-NAF-2-A (Reference

1) from the USNRC.

DOM-NAF-2-A provides the necessary documentation to describe the Page 1 of 8

Serial No.1 0-404 Proposed License Amendment Addition of Analytical Method to COLR intended uses of VIPRE-D for PWR licensing applications.

Appendix C documents the qualification of the WRB-2M CHF correlation with the VIPRE-D code and listed the DDLs for the WRB-2M and W-3 CHF correlations, which was approved generically by the USNRC in April 2009 (Reference 1). The implementation of this methodology on a site specific basis requires USNRC approval of site, fuel type and code specific DNBR design limits (Le., SDL and/or DDLs) and a Technical Specification change to include DOM-NAF-2-A, with applicable Appendix, in the COLR list of references.

3.0 PROPOSED TECHNICAL SPECIFICATIONS CHANGE Analysis of the RFA-2 fuel design at North Anna with Dominion-specific, USNRC-approved methods will require a revision to the existing plant Technical Specifications. This change is administrative in nature involving the modification of a reference that supports the COLR. The proposed change is provided below.

TS 5.6.5.b, CORE OPERATING LIMITS REPORT (COLR)

This section is appended to modify a reference that reflects the proposed change below.

The modification describes a Dominion-specific analytical method used in determining core limits that are applicable to the Westinghouse RFA-2 fuel product.

The following bolded text is proposed to be added to COLR Item 19:

19.

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,"

and Appendix C,

"Qualification of the Westinghouse WRB-2M CHF Correlation in the Dominion VIPRE-D Computer Code. II

4.0 TECHNICAL EVALUATION

This package includes the technical basis and the required documentation to support the site specific application of the VIPRE-D thermal hydraulics code with the WRB-2M and W-3 CHF correlations to Westinghouse RFA-2 fuel in North Anna Power Station Units 1 and 2 cores.

Departure from nucleate boiling (DNB) analyses for the Westinghouse RFA-2 fuel product will use the USNRC-approved VIPRE-D code and the W-3 or WRB-2M CHF correlations described in DOM-NAF-2-A Appendix C (Reference 1) depending upon the transient.

The W-3 correlation is only used below the first mixing grid or when the local thermodynamic conditions are outside of the range of validity of the WRB-2M CHF correlation, such as the main steam-line break evaluation, where there is reduced temperature and pressure.

The W-3 CHF correlation is always used deterministically.

The Dominion Statistical DNBR Evaluation Methodology in Topical Report VEP-NE-2-A (Reference 2) is applied to all statistically-treated events.

The analysis to support the implementation of DOM-NAF-2-A and VEP-NE-2-A for the Westinghouse RFA-2 fuel product Page 2 of 8

Serial No.1 0-404 Proposed License Amendment Addition of Analytical Method to COLR is summarized below. The resulting SDL is 1.25 for transients analyzed using the Statistical DNBR Evaluation Methodology and the WRB-2M CHF correlation.

In addition, the DDL is 1.14 for the WRB-2M CHF correlation (Reference 1).

As described in Appendix C of DOM-NAF-2-A, the DDL for the W-3 CHF correlation is 1.30 above 1000 psia and 1.45 below 1000 psia for steam line break events.

Upon USNRC review and approval of this change request, these DNBR design limits will be approved for application to the Westinghouse RFA-2 fuel product at North Anna.

Specifically, Attachment 4 provides the technical basis to support implementation of the Dominion Statistical DNBR Evaluation Methodology for Westinghouse HFA-2 fuel at North Anna with the VIPRE-D/WRB-2M code/correlation pair, as well as the SDL obtained by this implementation. also confirms that the existing Reactor Core Safety Limits and protection functions (over-temperature ilT (OTilT), over-power ilT (OPilT), axial power distribution (Fill),

etc) do not require revision as a

consequence of this implementation.

The Statistical DNBR Evaluation Methodology is applied to the list of UFSAR transients listed in Attachment 4. Finally, all applicable Chapter 15 analyses were evaluated with the VIPRE-D/WRB-2M code/correlation pair and the Statistical DNBR Evaluation Methodology, and they all were demonstrated to have acceptable results.

These evaluations support plant operation at the current licensed power level of 2940 MWt and will become the Analysis of Record (AOR) co-current with the transition to the 17x17 RFA-2 fuel product at North Anna.

5.0 SAFETY SIGNIFICANCE

SUMMARY

The VIPRE-01 code has been approved by the USNRC and is widely used throughout the nuclear industry for PWR safety analyses. VIPRE-D is the Dominion version of VIPRE-01.

Fleet Report DOM-NAF-2-A (Reference 1) documents the use of VIPRE-D for the thermal-hydraulic evaluation of nuclear reactor cores.

VIPRE-D includes the CHF correlations to be used for the evaluation of Westinghouse RFA-2 fuel (WRB-2M and W-3 CHF correlations).

These correlations are documented in Appendix C of DOM-NAF-2-A.

In summary, DOM-NAF-2-A describes a methodology that is fully applicable for reload design.

The application of DOM-NAF-2-A (Reference 1) in conjunction with VEP-NE-2-A (Reference 2) is used to calculate the SDL applicable to the VIPRE-D/WRB-2M code/correlation pair for Westinghouse RFA-2 fuel at North Anna. Setpoint safety analysis evaluations have been performed to verify that the existing Reactor Core Safety Limits and protection functions (OTilT,

OPilT, Fill, etc) continue to be applicable for the VIPRE-D/WRB-2M code/correlation pair and the newly calculated SDL.

The applicable Chapter 15 analyses were evaluated and were demonstrated to have acceptable results.

In conclusion, the statepoint analysis is the basis for demonstrating the acceptability of the change.

Page 3 of 8

Serial No.10-404 Proposed License Amendment Addition of Analytical Method to COLR

6.0 REGULATORY EVALUATION

6.1 Applicable Regulatory Requirements/Criteria Section 2.1 of Fleet Report DOM-NAF-2-A lists the information to be provided to the USNRC by Dominion for review and approval of any plant specific application of the VIPRE-D code:

1) Technical Specifications change request to add DOM-NAF-2-A and relevant Appendixes to the plant's COLR list.

2) Statistical Design Limit(s) for the relevant code/correlation(s)

3) Any technical specification changes related to OTilT, OPilT, Fill or other reactor protection function, as well as revised Reactor Core Safety Limits.

4) List of UFSAR transients for which the code/correlations will be applied.

The Safety Evaluation Report (SER) provided by the USNRC for Topical Report VEP-NE-2-A lists the following conditions for its use:

1) The selection and justification of the Nominal Statepoints used to perform the plant specific implementation must be included in the submittal.

2) Justification of the distribution, mean and standard deviation for the statistically treated parameters must be included in the submittal.

3) Justification of the value of model uncertainty must be included in the plant specific submittal.

4) For the relevant CHF correlations, justification of the 95/95 DNBR limit and the normality of the M/P distribution, its mean and standard deviation must be included in the submission, unless there is an approved Topical Report documenting these (such as DOM-NAF-2-A).

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 [COLR]." (Technical Basis for the Proposed Technical Specification Change) provides the justification for the 95/95 DNBR limits for application to the RFA-2 fuel product.

Based on the information contained in Attachment 4, Dominion concludes that the proposed change meets regulatory requirements and criteria.

Page 4 of 8

Serial No.10-404 Proposed License Amendment Addition of Analytical Method to COLR 6.2 Determination of No Significant Hazards Consideration Dominion proposes a change to the North Anna Power Station Units 1 and 2 Technical Specifications pursuant to 10CFR50.90.

The proposed change adds an additional appendix (Appendix C) to the VIPRE-D code currently listed in Technical Specification (TS) 5.6.5.b. Item 19. VIPRE-DIWRB-2M and VIPRE-DIW-3 code/correlation pairs will be used to perform licensing calculations of Westinghouse RFA-2 fuel in North Anna cores, using the deterministic design limits (DDLs) documented in Appendix C

of the DOM-NAF-2-A Fleet Report and the statistical design limit (SDL). using the VIPRE-DIWRB-2M code/correlation pair for RFA-2 fuel at North Anna.

In accordance with the criteria set forth in 10 CFR 50.92, Dominion has evaluated the proposed TS change and determined that the change does 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.

Approval of the proposed changes will allow Dominion to use the VIPRE-DIWRB-2M and VIPRE-DIW-3 code/correlation pairs to perform licensing calculations of Westinghouse RFA-2 fuel in North Anna Cores, using the DDLs documented in Appendix C of the DOM-NAF-2-A Fleet Report and the SDL documented herein.

Neither the code/correlation pair nor the Statistical Departure from Nucleate Boiling Ratio (DNBR)

Evaluation Methodology affect accident initiators and thus cannot increase the probability of any accident.

Further, since both the deterministic and statistical DNBR limits 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/correlation and Statistical DNBR Evaluation Methodology do not increase the potential consequences of any accident.

Finally, the full core DNB design limit provides increased assurance that the consequences of a postulated accident which includes radioactive release would be minimized because the overall number of rods in DNB would not exceed the 0.1 % level. The pertinent evaluations 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 operated in a different manner, and system interfaces will not change.

The use of the VIPRE-DIWRB-2M and VIPRE-DIW-3 code/correlation pairs to perform licensing calculations of Westinghouse RFA-2 fuel in North Anna cores 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.

Page 5 of 8

Serial No.1 0-404 Proposed License Amendment Addition of Analytical Method to COLR 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 does not involve a physical alteration of the plant (no new or different type of equipment will be installed).

The use of VIPRE-D/WRB-2M and the VIPRE-D/W-3 code/correlation pairs and the applicable fuel design limits for DNBR does not impact any of the applicable design criteria and the licensing basis criteria will continue to be 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 VIPRE-D/WRB-2M and VIPRE-D/W-3 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/WRB-2M and VIPRE-D/W-3 code/correlation pairs and the Statistical DNBR Evaluation Methodology does not involve any alteration to plant equipment or procedures that would introduce any new or unique operational modes or accident precursors.

Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3.

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

Response: No.

Approval of the proposed changes will allow Dominion to use the VIPRE-D/WRB-2M and VIPRE-D/W-3 code/correlation pairs to perform licensing calculations of Westinghouse RFA-2 fuel in North Anna cores, using the DDLs documented in Appendix C of the DOM-NAF-2-A Fleet Report and the SOL documented herein.

The SOL has been developed in accordance with the Statistical DNBR Evaluation Methodology.

North Anna TS 2.1, "Safety Limits,"

specifies that any DNBR limit established by any code/correlation must provide at least 95% non-DNB probability at a 95% confidence level. The DNBR limits meet the design basis of avoiding DNB with 95% probability at a 95% confidence level.

The required DNBR margin of safety for North Anna Power Station, which in this case is the margin between the 95/95 DNBR limit and clad failure, is therefore not reduced.

Therefore, the proposed TS change does not involve a significant reduction in a margin of safety.

Based on the above information, Dominion concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10CFR50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

Page 6 of 8

Serial No.1 0-404 Proposed License Amendment Addition of Analytical Method to COLR 6.3 Environmental Assessment The proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10CFR51.22(c)(1) as follows:

(i)

The proposed change involves no significant hazards consideration.

As described in Section 6.2 above, the proposed change involves 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 change does 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 change does 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 10CFR51.22(b),

no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.4 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 change, (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 CONCLUSION

S DOM-NAF-2-A, including Appendix C, has been demonstrated to be applicable to establish and support the reload design parameters for North Anna Power Station Units 1 and 2 for Westinghouse RFA-2 fuel. Upon USNRC approval, Fleet Report DOM-NAF-2-A, Appendix C will be added to the list of approved COLR references.

Page 7 of 8

Serial No.1 0-404 Proposed License Amendment Addition of Analytical Method to COLR

8.0 REFERENCES

1. Fleet Report, DOM-NAF-2, Rev. 0.1-A "ReactQr Core Thermal-Hydraulics Using the VIPRE-D Computer Code," R.

S. Brackmann, July 2009. [ADAMS Accession No. ML092190894]

2. Topical

Report, VEP-NE-2-A, "Statistical DNBR Evaluation Methodology,"

R.

C.

Anderson, June 1987.

3. Letter from C. E. Rossi (USNRC) to J. A. Blaisdell (UGRA Executive Committee),

"Acceptance for Referencing of Licensing Topical

Report, EPRINP-2511-CCM,

'VIPRE-01: A Thermal-Hydraulic Analysis Code for Reactor. Cores,' Volumes 1, 2, 3 and 4," May 1, 1986.

4.

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.

Page 8 of 8

Serial No.10-404 Docket Nos. 50-338/339 ATTACHMENT 2 PROPOSED TECHNICAL SPECIFICATIONS PAGES (MARK-UP)

Virginia Electric and Power Company (Dominion)

North Anna Power Station Units 1 and 2

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) b.

(continued)

14. BAW-10199P-A, liThe BWU Critical Heat Flux Correlations. II

15. BAW-10170P-A, "Statistical Core Design for Mixing Vane Cores."

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

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

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

Volume II only (SBLOCA models).

19. 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 Cod~

c.

The core operatin lmits shall be determined such that all applicable limits

.g., fuel thermal mechanical limits, core thermal hydraulic 1 mits, Emergency Core Cooling Systems (ECCS) limits, nuclear limlts such as SDM, transient analysis limits, and accident analysi limits) of the safety analysis are met.

d.

The COLR, including a y midcycle revisions or supplements, shall be provided upon issu nce for each reload cycle to the NRC.

5.6.6 PAM Report When a report is required

~

Condition B of LCO 3.3.3, "Post Accident Monitoring (PAM) I strumentation," a report shall be submitted within the followi g 14 days. The report shall outline the cause of the inoperability, nd the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

," and Appendix C, "Qualification of the Westinghouse RB-2M CHF Correlation in the Dominion VIPRE-D Computer Code."

North Anna Units 1 and 2 5.6-4 Amendments 248/228

Serial No.10-404 Docket Nos. 50-338/339 ATTACHMENT 3 PROPOSED TECHNICAL SPECIFICATIONS PAGES (TYPED)

Virginia Electric and Power Company (Dominion)

North Anna Power Station Units 1 and 2

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 5.6.6 CORE OPERATING LIMITS REPORT (COLR) b.

(continued) 14.' BAW-10199P-A, liThe BWU Critical Heat Flux Correlations."

15. BAW-10170P-A, "Statistical Core Design for Mixing Vane Cores. II

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

17. EMF-96-029 (P)(A), IIReactor Analysis System for PWRs."

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

Volume II only (SBLOCA models).

19. 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," and Appendix C,"Qualification of the Westinghouse WRB-2M CHF Correlation in the Dominion VIPRE-D Computer Code."

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.

PAM Report When a report is required by Condition B of LCO 3.3.3, IIpost 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.1 0-404 Docket Nos. 50-338/339 ATTACHMENT 4 TECHNICAL BASIS FOR ADDING APPENDIX C OF FLEET REPORT DOM-NAF-2-A TO THE LIST OF USNRC APPROVED METHODOLOGIES FOR DETERMINING CORE OPERATING PARAMETERS Virginia Electric and Power Company (Dominion)

North Anna Power Station Units 1 and 2

Serial No.1 0-404 Proposed License Amendment Addition of Analytical Method to COLR TABLE OF CONTENTS TABLE OF CONTENTS 1

1.

INTRODUCTION 2

2.

BACKGROUND 3

2.1 FLEET REPORT DOM-NAF-2-A 3

2.2 TOPICAL REPORT VEP-NE-2-A 4

2.3 WESTINGHOUSE'S FUEL CRITERIA EVALUATION PROCESS FOR THE ApPLICATION OF THE WRB-2M CHF CORRELATION TO THE RFA-2 FUEL PRODUCT 5

3.

IMPLEMENTATION OF THE STATISTICAL DNBR EVALUATION METHODOLOGy.... 6 3.1 METHODOLOGY REVIEW 6

3.2 UNCERTAINTY ANALySiS 7

3.3 CHF CORRELATIONS 10 3.4 MODEL UNCERTAINTY TERM 10 3.5 CODE UNCERTAINTY 10 3.6 MONTE CARLO CALCULATIONS 11 3.7 FULL CORE DNB PROBABILITY SUMMATION 13 3.8 VERIFICATION OF NOMINAL STATEPOINTS 15 3.9 SCOPE OF ApPLICABILITY 17 3.10

SUMMARY

OF ANALySiS 19 4.

APPLICATION OF VI PRE-DIWRB-2MIW-3 TO NAPS 20 4.1 VIPRE-DIWRB-2M SOL FOR NORTH ANNA 20 4.2 SAFETY ANALYSIS LIMITS (SAL) 20 4.3 RETAINED DNBR MARGIN 21 4.4 VERIFICATION OF EXISTING REACTOR CORE SAFETY LIMITS, PROTECTION SETPOINTS AND NAPS UFSAR CHAPTER 15 EVENTS 22 5.

CONCLUSiONS 24 6.

REFERENCES 25 Page 1 of 26

Serial No.1 0-404 Proposed License Amendment Addition of Analytical Method to COLR 1.

Introduction This report provides the plant specific application of the Statistical DNBR Methodology for North Anna Power Station (NAPS) cores containing Westinghouse 17x17 Robust Fuel Assembly - 2 (RFA-2) fuel product.

The Westinghouse RFA-2 fuel product contains modified mid-grids and modified intermediate flow mixer grids (IFMs).

Specifically, this report supports the application of U.S. Nuclear Regulatory Commission (USNRC) approved Dominion Topical Report VEP-NE-2-A, "Statistical DNBR Evaluation Methodology" (Reference 1) to NAPS, where DNBR stands for Departure from Nucleate Boiling Ratio. It provides the technical basis and documentation required by the USNRC to evaluate the plant specific application of the VEP-NE-2-A methodology to NAPS.

This application employs the VIPRE-D thermal-hydraulic computer code (DOM-NAF-2-A, Reference 2) with the Westinghouse WRB-2M Critical Heat Flux (CHF) correlation (VIPRE-D/WRB-2M code/correlation pair) for the thermal-hydraulic analysis of the Westinghouse 17x17 RFA-2 fuel product at NAPS. In particular, Dominion requests the NRC review and approval of the Statistical Design Limit (SOL) documented herein consistent with 10 CFR 50.59(c)(2)(vii) since the proposed change establishes a Design Basis Limit for a Fission Product Barrier (DBLFPB).

Dominion is also seeking approval for the inclusion of Fleet Report DOM-NAF-2-A, Appendix C, (Reference 2) to the Technical Specification (T.S.) 5.6.5.b list of USNRC-approved methodologies used to determine core operating limits (Le., the reference list of the North Anna Core Operating Limits Report (COLR)).

This would allow Dominion the use of the VIPRE-D/WRB-2M code/correlation pair to perform licensing calculations for the Westinghouse 17x17 RFA-2 fuel in North Anna's cores, using the deterministic design limit (DOL) qualified in Appendix C of Fleet Report DOM-NAF-2-A, and the SOL identified herein.

With these approvals, Dominion will be licensed to perform in-house Departure from Nucleate Boiling (DNB) analyses for the intended uses described in Fleet Report DOM-NAF-2-A to support North Anna Power Station, Units 1 and 2 operation with the Westinghouse 17x17 RFA-2 fuel product.

Page 2 of 26

Serial NO.1 0-404 Proposed License Amendment Addition of Analytical Method to COLR

===2.

Background===

Dominion is purchasing fuel assemblies from Westinghouse for use at North Anna Power Station, Units 1 and 2. These assemblies are planned to be inserted in Units 1 and 2, commencing with Cycle 23 for both units.

The fuel assemblies are designated as the Westinghouse 17x17 RFA-2 fuel product (Reference 3).

These assemblies are a replacement for the resident fuel product, which is the AREVA Advanced Mark-BW (AMBW) fuel product.

2.1 Fleet Report DOM-NAF-2-A The computer code VIPRE (Versatile Internals and Components Program for Reactors - EPRI) was developed for the Electric Power Research Institute (EPRI) by Battelle Pacific Northwest Laboratories ~o perform detailed thermal-hydraulic analyses to predict CHF and DNBR of reactor cores.

VIPRE-01 was approved by the U.S. Nuclear Regulatory Commission (USNRC) in References 4 and 5 for referencing in licensing applications. VIPRE-D is the Dominion version of the VIPRE computer code based upon VIPRE-01, MOD-02.1. VIPRE-D was developed to fit the specific needs of Dominion's nuclear plants and fuel products by adding vendor specific CHF correlations and customizing its input and output. Dominion has not made any modifications to the NRC-approved constitutive models and algorithms contained in VIPRE-01.

Dominion's approved Fleet Report DOM-NAF-2-A (including Appendix C, which describes the verification and qualification of the WRB-2M CHF correlation)' (Reference 2) has been reviewed and approved by the USNRC in Reference 6. Fleet Report DOM-NAF-2-A provided the necessary documentation to describe Dominion's use of the VIPRE-D code, including modeling and qualification for Pressurized Water Reactors (PWR) thermal-hydraulic design and demonstrated that the VIPRE-D methodology is appropriate for PWR licensing applications. Appendix C qualified the WRB-2M CHF correlation with the VIPRE-D code and listed the deterministic code/correlation DNBR limits.

The WRB-2M CHF correlation is applicable for the DNBR evaluation of the Westinghouse 17x17 RFA-2 fuel product [refer to Section 2.3 for further discussion].

In addition, Section 2.1 of Fleet Report DOM-NAF-2-A listed the information to be provided to the USNRC by Dominion for the review and approval of any plant specific application of the VIPRE-D code:

1) Technical Specifications change request to add DOM-NAF-2-A and relevant Appendixes to the plant's COLR list.

2)

Statistical Design Limit(s) for the relevant code/correlation(s) (Section 4.1) 3)

Any technical specification changes related to thermal over-temperature L1T (OTL1T),

over-power L1T (OPL1T), axial power distribution (FL1I) or other reactor protection function, as well as revised Reactor Core Safety Limits (Section 4.5).

4)

List of UFSAR transients for which the code/correlations will be applied (Section 3.9).

This report provides the technical basis (items 1 through 4 above) to support implementation of the Dominion Statistical DNBR Evaluation Methodology for 17x17 RFA-2 fuel at NAPS with the VIPRE-Page 3 of 26

Serial No.10-404 Proposed License Amendment Addition of Analytical Method to COLR DtWRB-2M code/correlation pair, as well as the SOL obtained by this implementation (DOM-NAF-2-A Condition 2). This report also documents that the existing Reactor Core Safety Limits and protection functions (OTi1T, OPi1T, Fi1I, etc.) do not require revision as a consequence of this implementation (DOM-NAF-2-A Condition 3).

The list of UFSAR transients for which the code/correlation pair will be applied is also included herein (DOM-NAF-2-A Condition 4).

2.2 Topical Report VEP-NE-2-A In 1985, Virginia Power (Dominion) submitted Topical Report VEP-NE-2-A (Reference 1) to the USNRC describing a proposed methodology for the statistical treatment of key uncertainties in core thermal-hydraulic DNBR analysis. The methodology provided DNBR margin through the use of statistical rather than deterministic uncertainty treatment.

The methodology was reviewed and approved by the USNRC in May 1987, and the Safety Evaluation Report (SER) provided by the USNRC listed the following conditions for its use (Reference 7):

1) The selection and justification of the Nominal Statepoints used to perform the plant specific implementation must be included in the submittal (Sections 3.6 and 3.8).

2)

Justification of the distribution, mean and standard deviation for all the statistically treated parameters must be included in the submittal (Section 3.2).

3)

Justification of the value of model uncertainty must be included in the plant specific submittal (Section 3.4).

4)

For the relevant CHF correlations, justification of the 95/95 DNBR limit and the normality of the M/P distribution, its mean and standard deviation must be included in the submission, unless there is an approved Topical Report documenting these (such as Reference 2).

The VEP-NE-2-A methodology is currently implemented at North Anna consistent with our Licensed Amendment Request (LAR) submitted to the USNRC on July 5, 2005 (Reference 8), as supplemented by letters dated March 30, April 13, and May 11, 2006 (References 9, 10 and 11) and approved by the USNRC on July 21, 2006 (Reference 12). The USNRC approved Dominion to perform DNB analyses for AREVA AMBW fuel at NAPS using the VIPRE-D/BWU code/correlation pair.

In addition, DOM-NAF-2-A, including Appendix A (Reference 2) was added to the list of methodologies approved for the determination of core operating limits in TS 5.6.5.b. In order for Dominion to maintain vendor independence, Dominion intends to perform in-house DNB analyses for 17x17 RFA-2 fuel at North Anna Power Station, Units 1 and 2 by adding Appendix C of DOM-NAF-2-Ato the list of methodologies approved for determination of core operating limits.

Page 4 of 26

Serial No.1 0-404 Proposed License Amendment Addition of Analytical Method to COLR 2.3 Westinghouse's Fuel Criteria Evaluation Process for the Application of the WRS-2M CHF Correlation to the RFA-2 Fuel Product In WCAP-12488-A (Reference 13), Westinghouse described a process and criteria that it intends to apply to changes or improvements in existing fuel designs that will not require prior NRC review and approval when these criteria are satisfied.

Westinghouse also will apply these criteria to adjustments or improvements of fuel performance design evaluation models, based on new data, without NRC review and approval. The NRC staff reviewed the Westinghouse fuel design criteria evaluation process described in WCAP-12488 and found it acceptable for licensing applications.

Westinghouse submitted LTR-NRC-01-44 (Reference 14) to the NRC documenting the minor design changes between the RFA and RFA-2 fuel products via Westinghouse's NRC-approved Fuel Criteria Evaluation Process (FCEP) (Reference 13).

In LTR-NRC-01-44 (Reference 14),

Westinghouse notified the NRC that the WRB-1 and WRB-2 DNB correlations are applicable to the RFA-2 fuel product.

Westinghouse revised this document in LTR-NRC-02-55 (Reference 3) to include the applicability of the WRB-2M correlation to the RFA-2 fuel mid-grid/IFM grid design modifications (Le., the 17x17 RFA-2 fuel prOduct).

Page 5 of 26

Serial No.1 0-404 Proposed License Amendment Addition of Analytical Method to COLR 3.

Implementation of the Statistical DNBR Evaluation Methodology 3.1 Methodology Review In Appendix C to Fleet Report DOM-NAF-2-A (Reference 2), Dominion calculated a DDL for the VIPRE-DtWRB-2M code/correlation pair.

The Statistical DNBR Evaluation Methodology (Reference 1) is employed herein to develop an SDL for NAPS. This new limit combines the correlation uncertainty with the DNBR sensitivities to uncertainties in key DNBR analysis input parameters. Even though the new DNBR limit (the SDL) is larger than the deterministic code/correlation design

limit, its use is advantageous as the Statistical DNBR Evaluation Methodology permits the use of nominal values for operating initial conditions instead of requiring the application of evaluated uncertainties to the initial conditions for statepoint and transient analysis.

The SDL is developed by means of a Monte Carlo analysis. The variation of actual operating conditions about nominal statepoints due to parameter measurement and other key DNB uncertainties is modeled through the use of a random number generator. Two thousand random statepoints are generated for each nominal statepoint. The random statepoints are then supplied to the thermal-hydraulics code VIPRE-D, which calculates the minimum DNBR (MDNBR) for each statepoint. Each MDNBR is randomized by a code/correlation uncertainty factor as described in Reference 1 using the upper 95% confidence limit on the VIPRE-DtWRB-2M code/correlation pair measured-to-predicted (M/P) CHF ratio standard deviation (Reference 2). The standard deviation of the resultant randomized DNBR distribution is increased by a small sample correction factor to obtain a 95% upper confidence limit, and is then combined Root-Sum-Square with code and model uncertainties to obtain a total DNBR standard deviation (Stotal). In accordance with Reference 1, the SDL is then calculated as:

SDL =1 + 1.645

• Stotal

[Equation 3.1]

in which the 1.645 multiplier is the z-value for the one-sided 95% probability of a normal distribution. This SDL thus provides peak fuel rod DNB protection at greater than 95/95.

As an additional criterion, the SDL is tested to determine the full core DNB probability when the peak pin reaches the SDL. This process is performed by summing the DNB probability of each rod in the core, using a bounding fuel rod census curve and the DNB sensitivity to rod power. If necessary, the SDL is increased to reduce the full core DNB probability to 0.1 % or less.

Page 6 of 26

Serial No.1 0-404 Proposed License Amendment Addition of Analytical Method to COLR 3.2 Uncertainty Analysis This section is included herein to satisfy Condition 2

in the SER (Reference 7) of VEP-NE-2-A (Reference 1).

Consistent with VEP-NE-2-A, inlet temperature, pressurizer pressure, core thermal power, reactor vessel flow rate, core bypass flow, the nuclear enthalpy rise factor and the engineering enthalpy rise factor were selected as the statistically treated parameters in the implementation analysis. The magnitudes and functional forms of the uncertainties for the statistically treated parameters were derived in a rigorous analysis of plant hardware and measurement/calibration procedures, and have been summarized in Table 3.2-1.

The uncertainties for core thermal power, vessel flow rate, pressurizer pressure and core inlet temperature were quantified using all sensor, rack, and other component uncertainties.

Then, the uncertainties were combined in a

manner consistent with their relative dependence or independence to quantify the total uncertainty for each parameter.

Total uncertainties were quantified at the 20' level, corresponding to two-sided 95% probability.

Margin was included in these uncertainties to provide additional conservatism, and to allow for future changes in plant hardware or calibration procedures without invalidating the analysis.

The standard deviations (0')

were obtained by dividing the total uncertainty by 1.96, which is the z-value for the two-sided 95%

probability of a normal distribution.

Dominion has quantified the magnitude and distribution of uncertainty on the pressurizer pressure (system pressure) per the pressurizer pressure control system.

The pressurizer pressure uncertainty was quantified as normal, two-sided, 95% probability distribution with a magnitude of

+/- 3.67% of span or +/- 29.3 psia. The impact of parameter surveillance was considered. The current parameter surveillance limit for pressurizer pressure of 2205 psig was determined to be acceptable. With this parameter surveillance limit, the pressurizer pressure uncertainty was conservatively defined as a normal, two-sided, 95% probability distribution with a magnitude of

+/- 30.0 psia and a standard deviation (0') of 15.31 psia. The applied uncertainty is unchanged from that employed in Reference 8 and subsequently approved in Reference 12.

Dominion has quantified the magnitUde and distribution of uncertainty on the average temperature (TAVG) per the TAVG rod control system. The average temperature uncertainty was quantified as a normal, two-sided, 95% probability distribution with a magnitude of +/- 3.22% of span or +/- 3.22°F.

The impact of parameter surveillance was considered. The current parameter surveillance limit for average temperature of 591.0°F was determined to be acceptable.

With this parameter surveillance limit, the average temperature uncertainty was conservatively defined as a normal, two-sided, 95% probability distribution with a magnitude of +/- 4.2°F and a standard deviation (0) of 2.143°F.

The applied uncertainty is unchanged from that employed in Reference 8 and subsequently approved in Reference 12.

Dominion has quantified the uncertainty on core power as measured by the secondary side heat balance as 0.862% at the non-uprated power 2893 MWt. The core power uncertainty associated with the MUR uprated power (2940 MWt) is less than 0.4%, but will not be credited. The power parameter uncertainty was conservatively treated as a normal, two-sided, 95% probability Page 7 of 26

Serial No.1 0-404 Proposed License Amendment Addition of Analytical Method to COLR distribution with a magnitude of +/-1.511% and a standard deviation (0) of 0.711%.

The applied uncertainty is bounding and unchanged from that employed in Reference 8 and subsequently approved in Reference 12.

Dominion has quantified the uncertainty on the reactor coolant system (RCS) flow as 2.390%. This parameter uncertainty is treated as a normal, two-sided, 95% probability distribution with a magnitude of +/-2.862% and a standard deviation (0) of 1.46%. The applied uncertainty is bounding and unchanged from that employed in Reference 8 and subsequently approved in Reference 12.

The two-sided, 95/95 tolerance interval (95% probability, 95% confidence) for the measurement uncertainty of the nuclear enthalpy rise factor, FLiH N, is 3.39%. Conservatively, the measured FLiH N

uncertainty was defined as a normal distribution with a 4% tolerance interval for consistency with previous applications.

The magnitude and distribution of uncertainty on the engineering hot channel factor, FLiH E, was quantified as a normal probability distribution with a magnitude of +/- 3.0%. The Statistical DNBR Evaluation Methodology (Reference 1) treats the FLil uncertainty as a uniform probability distribution.

The total core bypass flow consists of separate flow paths through the thimble tubes, direct leakage to the outlet nozzle, baffle joint leakage flow, upper head spray flow and core-baffle gap flow. These five components were each quantified based on the current North Anna core configuration, their uncertainties conservatively modeled and the flows and uncertainties totaled.

The Monte Carlo analysis ultimately used a best estimate bypass flow of 5.5% with an uncertainty of 1.0%. The analysis assumed that the probability was uniformly distributed.

Page 8 of 26

Serial No.1 0-404 Proposed License Amendment Addition of Analytical Method to COLR Table 3.2-1: North Anna Parameter Uncertainties PARAMETER NOMINAL STANDARD UNCERTAINTY DISTRIBUTION VALUE DEVIATION Pressure 2250 15.31 psi

+/-30.0 psi at 2a Normal

[psia]

Temperature 553.7 2.143°F

+/-4.2°F at 2a Normal

[OF]

Power [MWt]

2,940 0.771%

+/-1.511 % at 2a Normal Flow [gpm]

295,000 1.46%

+/-2.862% at 2a Normal Fl':.HN 1.587 2.0%

+/-4.0% at 2a Normal Fl':.HE 1.0 N/A

+/-3.0%

Uniform Bypass [%]

5.5 N/A

+/-1.0%

Uniform Page 9 of 26

Serial No. 10*404 Proposed License Amendment Addition of Analytical Method to COLR 3.3 CHF Correlations The WRB-2MIW-3 CHF correlations are used for the calculation of DNBRs in the Westinghouse 17x17 RFA-2 fuel product.

Only the WRB-2M CHF correlation is applicable to the operating conditions for which the Statistical DNBR Evaluation Methodology applies.

Table 3.3-1 presents the Design Limit correlation data for VIPRE-DIWRB-2M code/correlation pair. The W-3 correlation is only used below the first mixing grid or when the local thermodynamic conditions are outside of the range of validity of the WRB-2M CHF correlation, such as the main steam-line break evaluation, where there are reduced temperature and pressure.

The W-3 CHF correlation is always used deterministically.

Table 3.3-1: CHF Code/Correlation Data (Reference 2)

WRB-2M Average M1P 1.0006 S{M1P) 0.0640 n

241 K*

1.0824 Kx S{M1P) 0.06927 3.4 Model Uncertainty Term This section is included herein to satisfy Condition 3 in the SER (Reference 7) of the Statistical DNBR Evaluation Methodology Topical Report (Reference 1).

The VIPRE-D 20-channel production model for North Anna with the 17x17 RFA-2 fuel product was used in the development of the VIPRE-DIWRB-2M code/correlation pair SOL for North Anna. Since this is the production model that Dominion intends to use for all North Anna evaluations, there is no additional uncertainty associated with the use of this model. In summary, it is concluded that no correction for model uncertainty is necessary, and the model uncertainty term is set to zero for the calculation of the total DNBR standard deviation.

3.5 Code Uncertainty The code uncertainty accounts for any differences between Dominion's VIPRE-D and Westinghouse's THINC and VIPRE (i.e. VIPRE-W) codes, with which the WRB-2M CHF data were correlated, and any effect due to the modeling of a full core with a correlation based upon bundle test data. These uncertainties are clearly independent of the correlation, the model, and parameter

• K is a sample size correction factor that gives a one-sided 95% upper confidence limit on the estimated standard deviation of a given population. It can be calculated as:

Z. en-i)

K=

2

(..JZn 1.645)

Page 10 of 26

Serial NO.1 0-404 Proposed License Amendment Addition of Analytical Method to COLR induced uncertainties. The code uncertainty was quantified at 5%; consistent with the factors specified for other thermal/hydraulic codes in Reference 1. The basis for this uncertainty is described in detail by USNRC staff in Reference 7. In Reference 7, the USNRC Staff refers to the 5% uncertainty as being a 20- value. The 5% code uncertainty is certainly conservative in light of the excellent VIPRE-DNIPRE-W and VIPRE-D/CHF data comparisons.

However, the 5%

uncertainty serves as a conservative factor that may be shown to be wholly or partially unnecessary at a later time.

A one-sided 95% confidence level on the code uncertainty is then 3.04% (= ( 5.0%) /1 '.645). The use of the 1.645 divisor (the one-sided 95% tolerance interval multiplier) is conservative since the USNRC Staff considers the 5% uncertainty to be a 20- value.

3.6 Monte Carlo Calculations In order to perform the Monte Carlo analysis, nine Nominal Statepoints covering the full range of normal operation and anticipated transient conditions were selected. These statepoints must span the range of conditions over which the statistical methodology will be applied.

Two statepoints were selected at each of the four Reactor Core Safety Limit (RCSL) pressures (2400, 2250, 2000, and 1860 psia). For each of the RCSLs, a high power statepoint at 118% and a statepoint near the intercept of the DNBR limit line with the vessel exit boiling line were chosen.

In order to apply the methodology to low flow events, a low flow statepoint is also included.

The selected Nominal Statepoints are listed in Table 3.6-1.

Table 3.6-1: Nominal Statepoints for Westinghouse 17x17 RFA-2 Fuel at North Anna with VIPRE-D/WRB-2M STATE PRESSURIZER INLET POWER FLOW N

POINT PRESSURE TEMPERATURE

[%]

[%]

F~H MDNBR

[psia]

[oF]

A 2400.0 605.07 118 100 1.587 1.242 B

2400.0 613.30 113 100 1.587 1.241 C

2250.0 596.99 118 100 1.587 1.240 D

2250.0 608.45 111 100 1.587 1.241 E

2000.0 585.74 118 100 1.587 1.240 F

2000.0 598.32 111 100 1.587 1.241 G

1860.0 581.28 118 100 1.587 1.244 H

1860.0 588.98 114 100 1.587 1.242 I

2250.0 553.70 108 62.98 1.587 1.242 The Monte Carlo analysis itself consisted of 2000 calculations performed around each of the nine Nominal Statepoints. As described in Section 3.1, the DNBR standard deviation at each Nominal Statepoint was augmented by the code/correlation uncertainty, the small sample correction factor, and the code uncertainty to obtain a total DNBR standard deviation.

The Total STolal, is obtained using the Root-Sum-Square method according to Equation 3.2:

Page 11 of 26

Serial No.10-404 Proposed License Amendment Addition of Analytical Method to COLR STOTAL =

[Equation 3.2]

where:

SONBR is the standard deviation for the Randomized ONBR distribution.

The factor {~~:1 - 1.0} is the uncertainty in the standard deviation of the 2,000 Monte Carlo simulations, and provides a 95% upper confidence limit on the standard deviation.

• 1/N is the uncertainty in the mean of the correlation. N is the number of degrees of freedom in the original correlation database.

Fe is the code uncertainty, that has been defined as 5%

(20' value),

i.e.,

5.0%/1.645 = 3.04% (1 0' value). See Section 2.5 in Reference 1.

FM is the model uncertainty, which is 0.0 since the Monte Carlo simulation is run with the production model.

Note that this equation differs slightly from the equation listed in Reference 1. It has an additional factor applied to the Randomized ONBR SONBR, the 1/N factor to correct for the uncertainty in the mean of the correlation. This factor has been used in previous implementations of the Statistical ONBR Evaluation Methodology, such as Reference 8 as supplemented by Reference 11, which was subsequently approved in Reference 12.

The limiting peak fuel rod SOL was calculated to be 1.243 for VIPRE-OIWRB-2M code/correlation pair. The Monte Carlo'Statepoint analysis is summarized in Table 3.6-2.

Page 12 of 26

Serial NO.1 0-404 Proposed License Amendment Addition of Analytical Method to COLR Table 3.6-2: Peak Pin SOL Results for North Anna 17x17 RFA-2 Fuel with VIPRE-OJWRB-2M STATEPOINT Randomized DNB Total DNB Pin Peak SONBR STaTAL SDL95195 A

0.1281 0.1405 1.231 B

0.1223 0.1344 1.221 C

0.1302 0.1426 1.235 0

0.1268 0.1391 1.229 E

0.1287 0.1411 1.232 F

0.1228 0.1349 1.222 G

0.1245 0.1367 1.225 H

0.1238 0.1360 1.224 I

0.1348 0.1475 1.243 3.7 Full Core DNB Probability Summation After the development of the peak pin 95/95 ONBR limit, the data statistics are used to determine the number of rods expected in ONB. The ONB sensitivity to rod power is estimated as CJ(ONBR)/

CJ(1/F~h). The specific values of CJ(ONBR)/ CJ(1/F~h), denoted ~, are listed in Table 3.7-1.

To ensure that the calculations are conservative, a one-sided tolerance limit of ~ is used:

f3* = f3 - tea, v). se(f3) in which:

~* is the one-sided tolerance limit on ~

t(a,v) is the T-statistic with significance level a and v degrees of freedom. For 2,000 observations at a 0.05 level of significance t(0.05,2000) = 1.645.

se(~) is the standard error of ~.

The variable 1/F~h is the most statistically significant independent variable in the linear regression model, yielding R2 values larger than 99%. The value of the statistic parameter F of 1/F~h was the largest for all statepoints, which indicates that the variable 1/F~h accounts for the largest amount of the variation in the ONBR.

Table 3.7-1: a(DNBR)/ a(1/F~h) Estimation for WRB-2M STATEPOINT

~

se(~)

~*

R2 A

5.24294 0.00491 5.23486 99.9%

B 5.10419 0.00514 5.09574 99.9%

C 5.42769 0.00414 5.42088 99.9%

0 5.28417 0.00474 5.27637 99.9%

E 5.48612 0.00507 5.47778 99.9%

F 5.28143 0.00494 5.27330 99.9%

G 5.36615 0.00443 5.35887 99.9%

H 5.23293 0.00454 5.22546 99.9%

I 5.70647 0.00736 5.69437 99.8%

Page 13 of 26

Serial No.1 0-404 Proposed License Amendment Addition of Analytical Method to COLR A representative fuel rod census curve used for the determination of the SOL is listed in Table 3.7-2. The full core ONB probability summation will be re-evaluated on a reload basis to verify the applicability of the fuel rod census (Fi'lHN versus % of core with Fi'lHN greater than or equal to a given Fi'lH limit) used in the implementation analysis. The limiting full-core ONB probability summation resulted in an SOL of 1.247.

The ONB probability summation for VIPRE-O/WRB-2M code/correlation pair is summarized in Table 3.7-3.

Table 3.7-2: Representative Fuel Rod Census for a Maximum Peaking Factor Fi1h =1.587 MAXIMUM%OF FUEL RODS IN Fi1h LIMIT CORE WITH Fi1h ~

to:

0.0 1.5870 0.1 1.5866 0.2 1.5860 0.3 1.5855 0.4 1.5850 0.5 1.5840 0.6 1.5829 0.7 1.5815 0.8 1.5803 0.9 1.5786 1.0 1.5770 1.5 1.5682 2.0 1.5580 2.5 1.5500 3.0 1.5430 4.0 1.5330 5.0 1.5240 6.0 1.5150 7.0 1.5060 8.0 1.4970 9.0 1.4900 10.0 1.4860 20.0 1.4560 30.0 1.4100 40.0 1.3500 PEAK 1.5870 Page 14 of 26

Serial No.1 0-404 Proposed License Amendment Addition of Analytical Method to COLR Table 3.7-3: Full Core ONB Probability Summation for 17x17 RFA-2 Fuel with VIPRE-OJWRB-2M STATEPOINT STOTAL

% of Rods in Full Core DNB SDL99.9 A

0.1405 0.09861 1.238 B

0.1344 0.09887 1.226 C

0.1426 0.09923 1.240 0

0.1391 0.09849 1.234 E

0.1411 0.09876 1.236 F

0.1349 0.09880 1.225 G

0.1367 0.09853 1.228 H

0.1360 0.09873 1.228 I

0.1475 0.09906 1.247 3.8 Verification of Nominal Statepoints Condition 1 of the USNRC's SER for VEP-NE-2-A (Reference 7) requires that the Nominal Statepoints be shown to provide a bounding ONBR standard deviation for any set of conditions to which the methodology may potentially be applied.

It is therefore necessary to demonstrate that Slolal as calculated herein is maximized for any conceivable set of conditions at which the core may approach the SOL. To do so, a regression analysis is performed using the unrandomized ONBR standard deviations at each Nominal Statepoint as the dependent variable (Le., the raw MONBR results obtained from the Monte Carlo simulation). The Nominal Statepoint pressures, inlet temperatures, powers and flow rates are used as the independent variable. If no clear trend appears in the plot it can be concluded that the standard deviation has been maximized. If a clear trend is displayed, the regression function is determined.

This regression equation is evaluated to determine the values of the independent variable for which the standard deviation would be maximized, and it is verified that the Nominal Statepoints selected bound those conditions. In addition, the residuals of the regression are plotted again against all the independent variables, and it is verified that no trends are discernible.

Table 3.8-1 shows the R2 coefficients obtained for the verification of the nominal statepoints. The largest linear curve fit R2 coefficient is 53.65%, thus validating that there is no dependence.

An evaluation of all the data, linear fits, and R2 coefficients indicates that there are no discernible trends in the database. Therefore, it was concluded that STOTAL had been maximized for any conceivable set of conditions at which the core may approach the SOL and that the selected Nominal Statepoints provide a bounding standard deviation for any set of conditions to which the methodology may potentially be applied.

Figure 3.8-1 displays a sample regression plot for WRB-2M and clearly shows the trends discussed above.

Page 15 of 26

0.1400 0.1200 0.1000 c:I:

ca Z

0.0800 0

c:r:

2:

0.0600 C) iii 0.0400 0.0200 Serial NO.1 0-404 Proposed License Amendment Addition of Analytical Method to COLR Table 3.8-1: R2 Coefficients for the Verification of the Nominal Statepoints for North Anna 17x17 RFA-2 Fuel with VIPRE-D/WRB-2M R

2

- Linear Regression Pressure 9.83%

Temperature 41.02%

Flow Rate 53.65%

Power 1.34%

A II:--.

Iy= -7E-07x+ 0.151 R2 = 0.5365 0.0000 20000 22000 24000 26000 28000 30000 32000 34000 36000 1/8th Core Flow Rate [gpm]

SIGMA DNBR Linear (SIGMA DNBR)

Figure 3.8-1: Variation of the Unrandomized Standard Deviation with Flow Rate for the WRB-2M CHF Correlation Page 16 of 26

Serial No.1 0-404 Proposed License Amendment Addition of Analytical Method to COLR 3.9 Scope of Applicability This section is included herein to satisfy Condition 4

in the SER (Reference 7) of VEP-NE-2-A (Reference 1).

The Statistical DNBR Evaluation Methodology may be applied to all Condition I and II DNB events (except Rod Withdrawal from Subcritical (RWFS) which is initiated from zero power), and to the Loss of Flow and the Locked Rotor Accidents.

The accidents to which the methodology is applicable are listed in Table 3.9-1.

This table corresponds to Table 2.1-1 in Reference 2. The range of application is consistent with previous applications of Dominion Statistical DNBR Evaluation Methodology applications for North Anna.

This methodology will not be applied to accidents that are initiated from zero power where the parameter uncertainties are higher.

The Statistical DNBR Evaluation Methodology provides analytical margin by permitting transient analyses to be initiated from nominal operating conditions, and by allowing core thermal limits to be generated without the application of the bypass flow, Ft.HN (measurement component) and hot channel uncertainties. These uncertainties are convoluted statistically into the DNBR limit.

Page 17 of 26

Serial No.1 0-404 Proposed License Amendment Addition of Analytical Method to COLR Table 3.9-1: UFSAR Transients Analyzed with VIPRE-D/WRB-2M/W-3 for North Anna ACCIDENT NAPS USAR APPLICATION SECTION Uncontrolled rod cluster control assembly bank 15.2.1 DET-DNB withdrawal from a subcritical condition Uncontrolled rod cluster control assembly bank 15.2.2 STAT-DNB withdrawal at power Rod cluster control assembly misalignment 15.2.3 STAT-DNB (System Malfunction or Operator Error)

Uncontrolled boron dilution 15.2.4 STAT-DNB Partial loss of forced reactor coolant flow 15.2.5 STAT-DNB Startup of an inactive reactor coolant loop 15.2.6 STAT-DNB Loss of external electrical load and/or turbine trip 15.2.7 STAT-DNB Loss of normal feedwater 15.2.8 STAT-DNB Loss of offsite power to the station auxiliaries 15.2.9 STAT-DNB Excessive heat removal due to feedwater system 15.2.10 STAT-DNB malfunctions Excessive load increase incident 152.11 STAT-DNB Accidental depressurization of the reactor coolant 15.2.12 STAT-DNB system Accidental depressurization of the main steam 15.2.13 DET-DNB system Spurious operation of the safety injection system 15.2.14 STAT-DNB at power Complete loss of forced reactor coolant flow 15.3.4 STAT-DNB Single rod cluster control assembly withdrawal at 15.3.7 STAT-DNB full power Rupture of a main steam pipe 15.4.2.1 DET-DNB Major rupture of a main feed water pipe 15.4.2.2 non-DNB Locked reactor coolant pump rotor 15.4.4 STAT-DNB Page 18 of 26

Serial NO.1 0-404 Proposed License Amendment Addition of Analytical Method to COLR 3.10 Summary of Analysis The steps of the SOL derivation analysis may be summarized as follows:

In accordance with the Statistical ONBR Evaluation Methodology, 2,000 random statepoints are generated about each nominal statepoint and VIPRE-O is then executed to obtain MONBRs.

The standard deviation for the distribution of 2,000 MONBRs is referred to as the unrandomized standard deviation.

At the limiting Nominal Statepoint (I), the standard deviation of the randomized ONBR distributions, which is the unrandomized corrected for CHF correlation uncertainty, was found to be 0.1348. This value was then combined Root Sum Square with code and model uncertainty standard deviations to obtain a total ONBR standard deviation of 0.1475, as listed in Table 3.6-2. The use of 0.1475 in Equation 3.1 yields a peak pin ONBR limit of 1.243 with at least 95% probability at a 95% confidence level. The total ONBR standard deviation was then used to obtain 99.9% ONB protection in the full core of 1.247, which occurs at Nominal Statepoint (B). Therefore the VIPRE-OIWRB-2M code/correlation pair SOL for North Anna 17x17 RFA-2 fuel is set to 1.25.

Page 19 of 26

Serial No.1 0-404 Proposed License Amendment Addition of Analytical Method to COLR 4.

Application of VIPRE-DIWRB-2MIW-3 to NAPS VIPRE-OIWRB-2M code/correlation pair together with the Statistical ONBR Evaluation Methodology will be applied to Condition I and II ONB events (except Rod Withdrawal from Subcritical, RWFS), and to the Complete Loss of Flow event and the Locked Rotor Accident. The Statistical ONBR Evaluation Methodology provides analytical margin by permitting transient analyses to be initiated from nominal operating conditions, and by allowing core thermal limits to be generated without the application of the bypass flow, FLiH N (measurement component) and FLiH E

uncertainties. These uncertainties are convoluted statistically into the ONBR limit.

The WRB-2M and W-3 CHF correlations are used for the calculation of ONBRs in the Westinghouse 17x17 RFA-2 fuel product. The W-3 correlation is only used below the first mixing grid or when the local thermodynamic conditions are outside of the range of validity of the WRB-2M CHF correlation, such as the main steam-line break evaluation, where there are reduced temperature and pressure. The W-3 CHF correlation is always used deterministically.

In addition, there are a few events that will be evaluated with the VIPRE-OIW-3 code/correlation pair and deterministic models because they do not meet the applicability requirements of the Statistical ONBR Evaluation Methodology (see the events in Table 3.9-1 labeled 'OET-ONB').

These events will be initiated from bounding operating conditions considering the nominal value and the appropriate uncertainty value, and require the application of the bypass flow, FLiH N

(measurement component) and FLiH E uncertainties. The events modeled deterministically are limited by the DOL stated in OOM-NAF-2-A (Reference 2).

4.1 VIPRE-OIWRB-2M SOL for North Anna The SOL for North Anna cores containing Westinghouse 17x17 RFA-2 fuel with the VIPRE-OIWRB-2M code/correlation pair was derived in Section 3 of this report. The SOL for VIPRE-OIWRB-2M code/correlation pair is determined to be 1.25. The SOL limit provides a peak fuel rod ONB protection with at least 95% probability at a 95% confidence level and a 99.9% ONB protection for the full core. This SOL is plant specific as it already includes the North Anna specific uncertainties for the key parameters accounted for in the application of the Statistical ONBR Evaluation Methodology. Therefore, this limit is applicable to the analysis of statistical ONB events of Westinghouse 17x17 RFA~2 fuel in North Anna cores with the VIPRE-OIWRB-2M code/correlation pair.

4.2 Safety Analysis Limits (SAL)

In the performance of in-house ONB thermal-hydraulic evaluations, design limits and safety analysis limits are used to define the available retained ONBR margin for each application. The difference between the safety analysis (self-imposed) limit and the design limit is the available retained ONBR margin.

For deterministic ONB

analyses, the design ONBR limit is set equal to the applicable code/correlation limit and it is termed the DOL. For statistical ONB analyses, the design ONBR limit is set equal to the applicable SOL. These design limits are two of the OBLFPB described in Page 20 of 26

Serial No.1 0-404 Proposed License Amendment Addition of Analytical Method to COLR Reference 15. The OOLs and SOLs are fixed and any changes to their value require USNRC review and approval. However, the safety analysis limits for deterministic and" statistical ONB analyses (SALoET and SALsTAT, respectively) may be changed without prior USNRC review and approval, provided the changes meet the criteria established in Reference 15.

A deterministic and statistical SAL equal to 1.55 has been selected for 17x17 RFA-2 fuel at NAPS with the VIPRE-OIWRB-2M code/correlation pair. This SAL is applicable for all deterministic analyses for a maximum peaking factor F.llH N equal to 1.65 and for all statistical analyses for a maximum peaking factor F.llH N equal to 1.587.

Table 4.2-1: ONBR Limits for WRB-2M and W-3 VIPRE-DIWRB-2M DOL 1.14 SOL 1.25 SAL 1.55 VIPRE-DIW-3 DOL (~1 000 psia) 1.30 DOL <<1000 psia) 1.45 SAL (~1 000 psia) 1.42 SAL <<1000 psia) 1.58 4.3 Retained DNBR Margin The difference between the safety analysis (self-imposed) limit and the design limit is the available retained ONBR margin:

(

SAL - DDL)

Retained DNBR Margin [%] =

SAL The resulting available retained ONBR margins are listed in Tables 4.3-1 and 4.3-2.

Page 21 of 26

Serial No.10-404 Proposed License Amendment Addition of Analytical Method to COLR Table 4.3-1: DNBR Limits and Retained DNBR Margin for Deterministic DNB Applications DETERMINISTIC DNB APPLICATIONS DNB RETAINED CORRELATION DDL SALoET DNBR MARGIN [%]

WRB-2M 1.14 1.55 26.4 W-3 << 1000 psia) 1.45 1.61 9.9 W-3 (~1000 psia) 1.30 1.44 9.7 Table 4.3-2: DNBR Limits and Retained DNBR Margin for Statistical DNB Applications STATISTICAL DNB APPLICATIONS DNB RETAINED CORRELATION SDL SALsTAT DNBRMARGIN

[%]

WRB-2M 1.25 1.55 19.3 This method of defining retained DNBR margin allows 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 a transition core penalty.

The reload thermal-hydraulics evaluation prepared as part of the reload safety analysis process presents tables and descriptions of retained DNBR margin and applicable penalties. Retained DNBR margin is tracked separately for each CHF correlation and for statistical and deterministic analyses.

4.4 Verification of Existing Reactor Core Safety Limits, Protection Setpoints and NAPS UFSAR Chapter 15 Events This section is included herein to satisfy Condition 3 of the plant specific application list in Section 2.1 of DOM-NAF-2-A (Reference 2).

To demonstrate that the DNB performance of the Westinghouse 17x17 RFA-2 fuel is acceptable, Dominion performed calculations for full-core configurations of Westinghouse 17x17 RFA-2 fuel.

The calculations were performed using the VIPRE-DtWRB-2M and VIPRE-DtW-3 code/correlation pairs and selected statepoints including: the reactor core safety limits (RCSL), axial offset limits (AO), rod withdrawal from subcritical (RWFS), rod withdrawal at power (RWAP), loss of flow (LOFA), locked rotor events (LOCROT), hot zero power steam line break (MSLB), dropped rod limit line (DRLL), and static rod misalignment (SRM). These various statepoints provide sensitivity of DNB performance to the following: (a) power level (inclUding the impact of the part-power Page 22 of 26

Serial No.1 0-404 Proposed License Amendment Addition of Analytical Method to COLR multiplier on the allowable hot rod power F~h), pressure and temperature (RCSL); (b) limiting axial flux shapes at several axial offsets (AO); and (c) low flow (LOFA and LOCROT). The statepoints for the RWFS and MSLB were evaluated with deterministic DNB methods. The remaining statepoints were evaluated using statistical DNB methods. The evaluation criterion for these analyses is that the minimum DNBR must be equal to or greater than the applicable safety analysis limit (SAL) listed in Table 4.2-1.

The results of the calculations demonstrate that the minimum DNBR values are equal to or greater than the applicable safety analysis limit for the analyses that are performed to address statepoints of the Reactor Core Safety Limits, the OT~T, OP~T and F~I trip setpoints, as well as all the evaluated Chapter 15 events (including the LOFA and LOCROT) with an F~h of 1.587 (COLR limit of 1.65 divided by the measurement uncertainty of 1.04 = 1.587) at a Rated Thermal Power of 2940 MWt.

Page 23 of 26

Serial NO.1 0-404 Proposed License Amendment Addition of Analytical Method to COLR 5.

Conelusions Dominion's Statistical DNBR Evaluation Methodology has been used to derive an SOL.

This application employs the VIPRE-D code with the Westinghouse WRB-2M CHF correlation (VIPRE-DIWRB-2M code/correlation pair) for the thermal-hydraulic analysis of Westinghouse 17x17 RFA-2 fuel product at NAPS.

The existing Reactor Core Safety Limits, OTL\\.T, OPL\\.T and FL\\.I trip setpoints as well as the current analyses of applicable UFSAR Chapter 15 events were shown to be bounding, and will not be changed. In particular, Dominion seeks the review and approval of the SOL of 1.25 documented herein as per 10 CFR 50.59(c)(2)(vii) since it constitutes a DBLFPB.

Dominion is also seeking the approval for the inclusion of Fleet Report DOM-NAF-2-A, ~ppendix C, to the Technical Specification 5.6.5.b list of USNRC approved methodologies used to determine core operating limits (Le., the reference list of the North Anna COLR). This would allow Dominion the use of the VIPRE-DIWRB-2M code/correlation pair to perform licensing calculations for the Westinghouse 17x17 RFA-2 fuel in North Anna's cores, using the DOL qualified in Appendix C of Fleet Report DOM-NAF-2-A, and the SOL documented herein. In addition, DOM-NAF-2-A provides justification of the normality of the WRB-2M CHF M/P distributions, their means and standard deviations, as required by the SER to Reference 1.

Page 24 of 26

Serial No.1 0-404 Proposed License Amendment Addition of Analytical Method to COLR 6.

References 1.

Topical Report, VEP-NE-2-A, "Statistical DNBR Evaluation Methodology," R. C. Anderson, June 1987.

2.

Fleet Report, DOM-NAF-2, Rev. 0.1-A, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code,"

R.

S.

Brackmann, July 2009 (ADAMS Accession No.

ML092190894).

3.

Letter from H. A. Sepp, (Westinghouse) to J. S. Wermiel (NRC), "Fuel Criterion Evaluation Process (FCEP) Notification of the RFA-2 Design, Revision 1 (Proprietary)," LTR-NRC-02-55, November 13, 2002 (ADAMS Accession No. ML023190181).

4.

Letter from C. E. Rossi (NRC) 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.

5.

Letter from A.

C.

Thadani (NRC) 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.

6.

Letter from D. Wright (NRC) to D. A. Christian (Dominion), "Kewaunee Power Station, Millstone Power Station, Units 2 and 3, North Anna Power Station, Unit Nos. 1 and 2, and Surry Power Station, Unit Nos. 1 and 2 -

Appendix C to the Dominion Fleet Report DOM-NAF-2, "Qualification of the Westinghouse WRB-2M CHF Correlation in the Dominion VIPRE-D Computer Code" (TAC Nos. MD8703, MD8704, MD8705, MD8707, MD8708, MD8709)," Serial No.09-290, April 22, 2009.

7.

Letter from L. B. Engle (NRC) to W.

L. Stewart (Virginia Power), "Statistical DNBR Evaluation Methodology, VEP-NE-2, Surry Power Station, Units NO.1 & NO.2 (Surry-1 &2) and North Anna Power Station, Units NO.1 & NO.2 (NA-1&2)," Serial No.87-335 dated May 28,1987.

8.

Letter from E. S. Grecheck (Dominion) to USNRC, "Virginia Electric and Power Company, North Anna Power Station Units 1 and 2, Proposed Technical Specification Changes, Addition of Analytical Methodology to COLR," Serial No.05-419, July 5, 2005 (ADAMS Accession No. ML051890034).

9.

Letter from W. R. Mathews (Dominion) to USNRC, "Virginia Electric and Power Company (Dominion), North Anna Power Station Unit Nos. 1 and 2, Response to Request for Additional Information on Proposed Technical Specification Changes on Addition of Analytical Methodology to the Core Operating Limits Report (TAC Nos. MC7526 and MC7527)," Serial No.06-142, March 30,2006 (ADAMS Accession No. ML060900631).

Page 25 of 26

Serial NO.1 0-404 Proposed License Amendment Addition of Analytical Method to COLR

10. Letter from E. S. Grecheck (Dominion) to USNRC, "Virginia Electric and Power Company (Dominion),

North Anna Power Station Units Nos.

1 and 2,

Proposed Technical Specifications Changes on Addition of Analytical Methodology to the Core Operating Limits Report, Administrative Correction," Serial No. 06-142A, April 13, 2006 (ADAMS Accession No. ML061040062).

11. Letter from E. S. Grecheck (Dominion) to USNRC, "Virginia Electric and Power Company, (Dominion), North Anna Power Station Unit Nos. 1 and 2, Response to Request for Additional Information on Proposed Technical Specification Changes on Addition of Analytical Methodology to the Core Operating Limits Report (TAC Nos. MC7526 and MC7527)," Serial No. 06-1428, May 11,2006 (ADAMS Accession No. ML061310495).

12. Letter from S. Monarque (NRC) to D. A. Christian (Dominion), "North Anna Power Station, Unit Nos. 1 and 2, Issuance of Amendments on Changes to Analytical Methodology and Core Operating Limits Report (TAC Nos. MC7526 and MC7527)," Serial No.06-643, July 21, 2006 (ADAMS Accession No. ML062020005).

13. Technical

Report, WCAP-12488-P-A,

'Westinghouse Fuel Criteria Evaluation Process,"

October 1994.

14. Letter from H. A. Sepp (Westinghouse) to J. S. Wermiel (NRC), "Fuel Criterion Evaluation Process (FCEP) Notification of the RFA-2 Design (Proprietary)," LTR-NRC-01-44, December 19,2001.

15. Technical Report, NEI 96-07, Revision 1, "Guidelines for 10 CFR 50.59 Implementation,"

Nuclear Energy Institute, November 2000.

16. Topical Report, VEP-FRD-42, Rev. 2.1-A, "Reload Nuclear Design Methodology," August 2003.

Page 26 of 26