ML110270089

Response Request for Additional Information (Rai), Proposed License Amendment Request (Lar), Addition of Analytical Methodology to COLR

ML110270089
Person / Time
Site:	North Anna
Issue date:	01/26/2011
From:	Price J Dominion, Virginia Electric & Power Co (VEPCO)
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
11-019
Download: ML110270089 (32)

Text

VIRGINIA E L E C T R I C AND POWER COMPANY RICHMOND, VIRGINIA 23261 January 26, 2011 10CFR50.90 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555 Serial No.

NL&OS/ETS Docket Nos.

License Nos.11-019 RO 50-338/339 NPF-4/7 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION)

NORTH ANNA POWER STATION UNITS 1 AND 2 RESPONSE REQUEST FOR ADDITIONAL INFORMATION (RAil PROPOSED LICENSE AMENDMENT REQUEST (LAR)

ADDITION OF ANALYTICAL METHODOLOGY TO COLR In a July 19, 2010 letter (Serial NO.1 0-404) supplemented by a September 9, 2010 letter (Serial No.10-523), Dominion requested 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, respectively.

The proposed amendments requested 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," into 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 including Appendix C of Fleet Report DOM-NAF-2-A into TS 5.6.5.b, Dominion also requested NRC review and approval for the use of the Dominion Topical Report VEP-NE-2-A, "Statistical DNBR Evaluation Methodology," with the Westinghouse RFA-2 fuel at North Anna using the VIPRE-DIWRB-2M code/correlation pair, as well as the SDL.

In a January 3, 2011, e-mail from Dr. V. Sreenivas (Accession Number ML110040003), the NRC requested additional information to complete the review of the proposed licensing actions. The response to this RAI is provided in the attachment to this letter.

A phone call was held between Dominion and the NRC on January 11,2011 to discuss the RAI questions. It was identified that four of the questions were outside of the scope of the proposed amendments as requested in the July 19, 2010 letter and supplemented by the September 9, 2010 letter.

Specifically, Questions 3, 4, 11, and 12 were questions identified as being outside of the scope. Dominion understands the NRC's interest in this information and the ultimate relevance of the information to subject matter.

Accordingly, Dominion can make the requested information available for an audit as mentioned in the NRC's acceptance for review letter dated September 17, 2010 (Serial NO.10-566).

Serial No.11-019 Docket Nos. 50-338/339 Page 2 of 3 The information provided in the attachment to this letter does not impact the conclusion of the significant hazards consideration determination as defined in 10 CFR 50.92 or the evaluation for eligibility for categorical exclusion as set forth in 10 CFR 51.22(c)(9).

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).

Therefore, Dominion continues to request approval of the proposed amendments by July 21, 2011 to complete analysis work required to support operation of the Westinghouse RFA-2 fuel. Dominion also continues to request 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.

Vi rice sident - Nuclear Engineering

Attachment:

Response to Request for Additional Information Proposed License Amendment Request (LAR) Addition of Analytical Methodology to COLR Commitments made in this letter: None COMMONWEALTH OF VIRGINIA VICKI l. HUll

~

NotaryPublic t

Commonwnlth of Virginia

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140542 l

My Commll.1on ExpIrn May 31. 2014 COUNTY OF HENRICO The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by J. Alan Price, who is Vice President - Nuclear Engineering of 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 t 12 day at ~~

, 2011.

My Commission Expires:

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Notary Public

Attachment:

Response to Request for Additional Information 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 Manage r 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 - ]'h Floor 109 Governor Street Room 730 Richmond, Virginia 23219 Serial No.11-019 Docket Nos. 50-338/339 Page 3 of 3

Serial No.11-019 Docket Nos. 50-338/339 ATTACHMENT RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION PROPOSED LICENSE AMENDMENT REQUEST (LAR)

ADDITION OF ANALYTICAL METHODOLOGY TO COLR Virginia Electric and Power Company (Dominion)

North Anna Power Station Units 1 and 2

Serial No.11-019 Docket Nos. 50-338/339 RAI Response Addition of Analytical Methodology to COLR VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION)

NORTH ANNA POWER STATION UNITS 1 AND 2 RESPONSE REQUEST FOR ADDITIONAL INFORMATION (RAI)

PROPOSED LICENSE AMENDMENT REQUEST (LAR)

ADDITION OF ANALYTICAL METHODOLOGY TO COLR

Background

By letter dated July 19, 2010, Dominion submitted a License Amendment Request (LAR) to the NRC to add Fleet Report DOM-NAF-2-A including Appendix C to the list of NRC-approved methodologies for determining core operating limits (Le., the reference list of the Core Operating Limits Report) in TS 5.6.5.b.

The LAR also requested the NRC's review and approval of the use of Dominion Topical Report VEP-NE-2-A using the VIPRE-DIWRB-2M code/correlation with the Westinghouse 17x17 RFA-2 fuel and the resulting Statistical Design Limit (SOL). During the NRC's acceptance review of the LAR, the NRC requested supplemental information by letter dated August 23,2010 and Dominion provided the supplemental information to the NRC, by letter dated September 9,2010.

This letter responds to the NRC Request for Additional Information (RAI) received in e-mail dated January 3, 2011.

Dominion is requested to respond to the following questions by February 21, 2011.

NRC Question 1 Table 4-1 of WCAP-15025PA indicates that the applicable range of pressure for WRB-2M is 1495 to 2425 psia, and Table 1 of Safety Evaluation for Appendix C to DOM-NAF-2 indicate that the applicable range of pressure for WRB-2M is 1405 to 2425 psia.

Please make sure that the correct pressure range for applicability of WRB-2M correlation is 1495 to 2425 psia.

Dominion's Response Dominion is aware of the typographical error in the Safety Evaluation for Appendix C of DOM-NAF-2-A. In the copy of DOM-NAF-2-A [Reference 1a] that was sent to the NRC by letter dated August 20, 2010, a note was added to the bottom of page 6 of the SER transmitted in letter dated April 22, 2009, referring the reader to a footnote in Appendix C on page C-22.

This footnote describes the typographical errors and confirms Dominion uses the approved applicability range of the WRB-2M CHF correlation from 1495 psia to 2425 psia for pressure.

References 1a.

Fleet Report, DOM-NAF-2, Rev. 0.2-P-A, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code," August 2010. [Non-Proprietary Version ADAMS Accession No. ML102390419]

Page 1 of 28

Serial No.11-019 Docket Nos. 50-338/339 RAI Response Addition of Analytical Methodology to COLR 1b.

Letter to USNRC from C. L. Funderburk (Dominion), "Virginia Electric and Power Company (Dominion), Dominion Nuclear Connecticut, INC. (DNC), Dominion Energy Kewaunee, INC. (DEK), North Anna and Surry Power Stations Units 1 and 2, Millstone Power Station Units 2 and 3, Kewaunee Power Station, DOM-NAF-2-P/NP, Revision 0.2-A, Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code," August 20, 2010, Dominion Serial No.10-486.

[ADAMS Accession No. ML102390421]

NRC Question 2 The licensee states that "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 Revision 1 of "Fuel Criterion Evaluation Process (FCEP) Notification of the RFA-2 Design" indicate that "Subsequent to the original RFA-2 notification, an additional evaluation of the CHF data has determined that WRB-2M is also applicable to the 17x17 RFA-2 fuel with or without IFMs for both the 12 foot and 14 foot cores with a 1.14 DNBR correlation limit. All DNBR correlation limits specified herein are based on a 95/95 criterion basis."

The NRC staff recognizes the fact that for certain accidents/transients, such as MSLB special correlations may have to be used since the pressure and temperature may fall outside the range of applicability of normal correlations. However, the staff could not find any technical justification for the licensee's method for using the W-3 correlation.

Please provide technical justification supported by a fuel vendor methodology and references, in support of the use of W-3 correlation for some transient analyses at NAPS.

Dominion's Response The W-3 correlation was used in the original licensing of North Anna.

The relevant discussion on the departure from nucleate boiling design basis is included in Section 4.4 of the North Anna UFSAR [Reference 2a].

The W-3 Critical Heat Flux (CHF) correlation has been approved by the NRC as the appropriate CHF correlation to use when reactor system conditions are outside the range of the fuel-specific CHF correlations (e.g. WRB-1, WRB-2, WRB-2M) in Reference 2b. In Reference 2b, the NRC approved the use of the W-3 CHF correlation in VIPRE for the analysis of the main steam line break (MSLB) accident analysis. This is also consistent with the NRC approval of the Beaver Valley Extended Power Uprate

[Reference 2c].

The W-3 CHF correlation has also been approved by the NRC for Dominion's use with the WRB-1 and WRB-2M CHF correlations in Reference 2d, for the Page 2 of 28

Serial No.11-019 Docket Nos. 50-338/339 RAI Response Addition of Analytical Methodology to COLR analysis of the rod withdrawal from subcritical (RWSC) and MSLB events and when the local conditions are outside the range of applicability of the primary CHF correlation.

References 2a.

North Anna Updated Final Safety Analysis Report, Revision 46.

2b.

Letter to H. Sepp (Westinghouse) from T. H. Essig (USNRC), "Acceptance for Referencing of Licensing Topical Report WCAP-14565, 'VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal/Hydraulic Safety Analysis (TAC No. M98666),'" January 19, 1999.

2c.

"Safety Evaluation Related to Extended Power Uprate at Beaver Valley Power Station, Unit Nos. 1 and 2," July 19, 2006, ADAMS Accession No. ML061720376.

2d.

Fleet Report, DOM-NAF-2, Rev. 0.2-P-A, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code," August 2010. [Non-Proprietary Version ADAMS Accession No. ML102390419]

NRC Question 3 FCEP Notification of the RFA-2 Design, Revision 1 indicates that the impacts on Non-LOCA Fuel Clad Temperature (Item I of Design Categories A) and all other safety analysis parameters were found to be unaffected by these changes.

This was confirmed by Westinghouse in an integrated 10 CFR 50.59 evaluation (

Reference:

"Transmittal of EVAL-01-066: GENERIC Implementation of Robust Fuel Assembly-2 (RFA-2) Design Changes," LTR-ESI-01-154, August 31, 2001.'j Please submit a copy of this document to the agency for staff's information.

Dominion's Response Dominion's submittal of July 19, 2010 [Reference 3a] only requests approval for the use of the Statistical DNBR Evaluation Methodology with the VIPRE-D/WRB-2M code/correlation for the Westinghouse 17x17 RFA-2 fuel and the resulting SOL. Thus, the requested information is not directly germane to this review.

However, Westinghouse can make LTR-ESI-01-154 available for NRC review during the audit, which was requested by the NRC in letter dated September 17, 2010 [Reference 3b].

References 3a.

Letter to USNRC from J. A. Price (Dominion), "Virginia Electric and Power

Company, North Anna Power Station Units 1 and 2,

Proposed License Amendment Request (LAR), Addition of Analytical Methodology to COLR," July 19, 2010, Dominion Serial NO.1 0-404.

Page 3 of 28

Serial No.11-019 Docket Nos. 50-338/339 RAI Response Addition of Analytical Methodology to COLR 3b.

Letter to D. A. Heacock (Dominion) from V. Sreenivas (USNRC), "North Anna Power Station Units 1 and 2: Acceptance Review of Proposed License Amendment Request, Addition of Analytical Methodology to COLR (TAC Number ME4262 and ME4263)," September 17, 2010.

NRC Question 4 The proposed new RFA-2 fuel will be co-resident with AREVA Advanced Mark-BW fuel that has different thermal and hydraulic characteristics from that of RFA-2 fuel. Provide detailed thermal-hydraulic analyses that shall demonstrate that the RFA-2 fuel is hydraulically compatible with the co-resident fuel and thereby maintains thermal-hydrodynamic stability, in accordance with SRP Section 4.4, Thermal and Hydraulic Design.

Dominion's Response Dominion's submittal of July 19, 2010 [Reference 4a] only requests approval for the use of the Statistical DNBR Evaluation Methodology with the VIPRE-D/WRB-2M code/correlation for the Westinghouse 17x17 RFA-2 fuel and the resulting SDL.

The hydraulic compatibility of the Westinghouse 17x17 RFA-2 fuel with the AREVA Advanced Mark-BW fuel is not directly germane to this review. However, as discussed in Dominion letter dated September 9, 2010 [Reference 4b] to support the fuel transition, Westinghouse analyses will confirm that hydraulic compatibility and thermal-hydraulic -criteria are met for limiting mixed core configurations. The analyses being performed include assessment of axial fit-up, grid growth, cross flow and lift force.

These analyses will be completed prior to fuel product implementation at North Anna Unit 1 (the lead unit) in spring 2012.

These analyses are still underway and can be made available for a regulatory audit following their completion, as requested by the NRC in letter dated September 17, 2010 [Reference 4c].

References 4a.

Letter to USNRC from J. A. Price (Dominion), "Virginia Electric and Power

Company, North Anna Power Station Units 1 and 2,

Proposed License Amendment Request (LAR), Addition of Analytical Methodology to COLR," July 19, 2010, Dominion Serial NO.1 0-404.

4b.

Letter to USNRC from J. A. Price (Dominion), "Virginia Electric and Power Company, North Anna Power Station Units 1 and 2, Supplemental Information to Support Proposed License Amendment Request (LAR) - Addition of Analytical Methodology to the Core Operating Limits Report (COLR)," September 9, 2010.

[ADAMS Accession No. ML102560291]

4c.

Letter to D. A. Heacock (Dominion) from V. Sreenivas (USNRC), "North Anna Power Station Units 1 and 2: Acceptance Review of Proposed License Amendment Request, Addition of Analytical Methodology to COLR (TAC Number ME4262 and ME4263)," September 17, 2010.

Page 4 of 28

Serial No.11-019 Docket Nos. 50-338/339 RAI Response Addition of Analytical Methodology to COLR NRC Question 5 Licensee states, "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 2a 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."

(a) Explain, with examples how the uncertainties were combined, in consistent with their relative dependence or independence to quantify the total uncertainty for each parameter.

(b) Explain how margins were included in the uncertainties to provide additional conservatism, and to allow for future changes in plant hardware or calibration procedures without invalidating the analysis. Provide representative examples.

Dominion's Response Question (a)

Consistent with the Statistical DNBR Evaluation Methodology approved topical report

[Reference 5a], inlet temperature, pressurizer pressure, core thermal power, and vessel flow rate were selected as statistical treated parameters in the analysis.

The magnitudes and functional forms of the uncertainties for the statistical treated parameters were derived in a rigorous analysis of current North Anna plant hardware and measurement/calibration procedures, and were summarized in Table 3.2-1 of of Reference 5b.

Bounding values for the uncertainties were then assumed in the analyses of the Statistical DNBR Evaluation Methodology to cover potential plant changes.

In previous responses to a Request for Additional Information (RAI), Dominion provided the method used for combining all sensor, rack, and other component uncertainties.

That response is documented in References 5c and 5d. The method used to combine sensor, rack, and component uncertainties used by Dominion has not changed since that time; therefore, only a brief summary of the method and the values of the sensor, rack, and component uncertainties used in the uncertainty analyses have been provided herein. In Tables 5-1 through 5-9, footnotes and bolded text have been used to identify differences between values reported herein and those previous reported in References 5c and 5d. The footnotes are also used to provide examples of how margin has been used to offset changes in plant hardware or calibration procedures without invalidating the analysis for the response to part (b).

Methodology The methodology used to combine the error components for a channel is the appropriate statistical combination of those groups of components which are statistically Page 5 of 28

Serial No.11-019 Docket Nos. 50-338/339 RAI Response Addition of Analyt ical Methodology to COLR independent (i.e., not interactive).

Those errors that are interactive are added arithmetically into groups to form independent groups that can be statistically combined.

Systematic (bias) errors are combined arithmetically outside the radical.

This methodology for the combination of uncertainties is termed the 'Square Root of the Sum of the Squares' (SRSS). The calculation for the total statistical error allowance for a loop calibration by modules with systematic error(s) is defined as follows:

CSA = SE +/- [EA2+ PMA2+ PEA2+ (SeA + SMTE)2 + SD2+ SPE2+ STE2

+ (M1 + M1 MTE)2 + (M2 + M2MTE)2 +...+ (Mn + MnMTE)2 + RD2+ RTE2+ RRA2]'h where:

CSA

= Channel Statistical Accuracy SE

= Systematic Error; i.e., error due to environmental conditions EA

= Environmental Allowance PMA

= Process Measurement Accuracy PEA

= Primary Element Accuracy SCA

= Sensor Calibration Accuracy SMTE = Sensor Measuring and Test Equipment SO

= Sensor Drift SPE

= Sensor Pressure Effect STE

= Sensor Temperature Effect SPSE

= Sensor Power Supply Effect M1

= First Module Accuracy M1 MTE = First Module Measurement and Test Equipment Accuracy M2

= Second Module Accuracy M2MTE = Second Rack Measurement and Test Equipment Accuracy Mn

= nth Module Accuracy MnMTE = nth Rack Measurement and Test Equipment Accuracy RD

= Rack Drift RTE

= Rack Temperature Effects RRA

= Rack Readability Allowance Pressurizer Pressure Dominion has quantified the magnitude and distribution of uncertainty on the pressurizer pressure (system pressure) per the pressurizer pressure control system.

The current component accuracies for the pressurizer pressure control loop are shown in Table 5-1.

The pressurizer pressure uncertainty is presently quantified as a normal, two-sided, 95% probability distribution with a magnitude of +/- 3.67% of span or +/- 29.3 psi.

The pressurizer pressure uncertainty was conservatively applied in Reference 5b as a normal, two-sided, 95% probability distribution with a magnitude of +/- 30 psia and a standard deviation (a) of 15.31 psia. This assumed uncertainty represents the minimum required accuracy for the pressurizer pressure and bounds the quantified uncertainty.

Page 6 of 28

Serial No.11-019 Docket Nos. 50-338/339 RAI Response Addition of Analytical Methodology to COLR Average Temperature Dominion has quantified the magnitude and distribution of uncertainty on the average temperature (Tavg) per the Tavg rod control system. The current component accuracies for the Tavg rod control loop are shown in Table 5-2. The average temperature uncertainty is presently quantified as a normal, two-sided, 95% probability distribution with a magnitude of +/-3.22% of span or +/-3.22°F.

The average temperature uncertainty was conservatively applied in Reference 5b as a normal, two-sided, 95% probability distribution with a magnitude of +/-4.2°F and a standard deviation (0) of 2.143°F.

This assumed uncertainty represents the minimum required accuracy for Tavg and bounds the quantified uncertainty.

Core Power Dominion has quantified the magnitude and distribution of uncertainty on the core power as measured by a secondary side heat balance which is dependent upon the component accuracies for the feedwater temperature, steam pressure and feedwater flow differential pressure. The component accuracies for the feedwater temperature, steam pressure and feedwater flow differential pressure input to the plant computer are shown in Tables 5-3, 5-4 and 5-5, respectively. The core power uncertainty is quantified as a normal, two-sided, 95% probability distribution with a magnitude of +/-O.862% of 2893 MWt (pre-MUR).

In order to increase power to the approved MUR power level of 2940 MWt, the uncertainty on the power indication is reduced to 0.35% due to the use of ultrasonic flow meters.

Dominion requested a Measurement Uncertainty Recapture (MUR) uprate in Reference 5e and received NRC approval in Reference 5f. This submittal documented a core power uncertainty of 0.35% when using feedwater ultrasonic flow meters. The Statistical DNBR Evaluation Methodology did not take advantage of the reduced power uncertainty.

In the event that the UFMs are unavailable and core is operated at the pre-MUR power level, the Statistical DNBR Evaluation Methodology remains bounding due to the use of the larger power uncertainty.

The standard deviation used in the Statistical DNBR Evaluation Methodology was 0.771%,

which included additional conservatism to allow for future changes in plant hardware or calibration procedures without invalidating the analysis.

This standard deviation corresponds to a +/-1.511% uncertainty, treated as a normal, two-sided, 95% probability distribution. This assumed uncertainty represents the minimum required accuracy for core power and bounds the quantified uncertainty.

It is noted that the value of 0.711 % on page 8 of Attachment 4 of the LAR dated July 19, 2010 [Reference 5b] is a typographical error.

However, the value of 0.771%

reported in Table 3.2-1 of Attachment 4 of Reference 5b is correct.

Page 7 of 28

)

Serial No.11-019 Docket Nos. 50-338/339 RAI Response Addition of Analytical Methodology to COLR RCS Flow Dominion has quantified the magnitude and distribution of uncertainty on the RCS flow as measured by a precision side heat balance at the beginning of each cycle. The RCS flow is then monitored by the cold leg elbow taps during the cycle, which are normalized against the precision calorimetric.

The inputs to the precision flow calorimetric are from power, hot leg temperature, and cold leg temperature.

The power uncertainty is determined as previously discussed. The component accuracies for the hot and cold leg temperature input to the plant computer are shown in Table 5-7. The component accuracy for the elbow tap indicator is shown in Table 5-8. An additional uncertainty to the RCS flow is the effect of hot leg streaming. This is conservatively included both as random and systematic uncertainty.

The RCS flow uncertainty is presently quantified as a normal, two-sided, 95% probability distribution with a magnitude of +/-2.390% of span.

The standard deviation used for the analysis of the Statistical DNBR Evaluation Methodology was 1.46%, which includes additional conservatism to allow for future changes in plant hardware or calibration procedures without invalidating the analysis. This standard deviation corresponds to a +/-2.862% uncertainty, treated as a normal, two-sided, 95% probability distribution. This assumed uncertainty represents the minimum required accuracy for RCS flow and bounds the quantified uncertainty.

Question (b)

The margin included in the uncertainty analyses refers to the difference between the quantified uncertainty and that used in the Statistical DNBR Evaluation Methodology. The uncertainty used in the Statistical DNBR Evaluation Methodology sets an upper bound for the allowed uncertainty of a component before an analysis is necessary.

The use of a larger uncertainty is conservative. The applied uncertainties are bounding and unchanged from those employed in Reference 5c and 5d.

The use of the margin to offset plant hardware and calibration procedure changes has been exemplified in Tables 5-1 through 5-9.

Since the time of the approval of the Statistical DNBR Evaluation Methodology for the AREVA Advanced Mark-BW fuel at North Anna, there have been changes to plant hardware and calibration procedures. Each time there is a change, the uncertainty analysis is reviewed to determine if the quantified uncertainty following the change will exceed the uncertainty assumed in the Statistical DNBR Evaluation Methodology.

One example is the multiple changes that have been made to the TAVG uncertainty. Despite the multiple changes to the TAVG uncertainty, the calculated uncertainty (+/-3.22% of span or +/-3.22°F) remained below that assumed in the development of the SOL for AREVA Advanced Mark-BW fuel (+/-4.2% of span or +/-4.2°F).

A second example where the overall uncertainty has increased is provided in the changes that have occurred to the feedwater flow uncertainty. The overall uncertainty increased from 2.274% dp span to 2.308% dp span. The feedwater flow uncertainty is an input to the core power uncertainty. Even though the change in the feedwater flow uncertainty did cause the core power uncertainty to change slightly, the uncertainty applied to core power in the Statistical DNBR Evaluation Methodology remained bounding and was unaffected.

Page 8 of 28

Serial No.11-019 Docket Nos. 50-338/339 RAI Response Addition of Analytical Methodology to COLR TABLE 5-1 Pressurizer Pressure Control Uncertainty Pressurizer Variable Pressure Uncertainty

% dp span SE 0.375 EA 0.000 PMA 0.000 PEA 0.000 SCA 0.500 SMTE 0.500 SO 0.750 SPE 0.000 STE1 1.663 SPSE 0.000 M1 0.100 M1MTE 0.153 M2 2.000 M2MTE 0.000 M3 1.002 M3MTE 0.087 RO 1.000 RTE 0.500 RRA 0.000 CSA 3.666%

Span 800 psi CSA 29.3 psi

• 1 The value of the STE term has increased from the analysis presented in References 5c and 5d.

The value increased from 1.425 to 1.663.

However, since the calculated uncertainty remained below that assumed in the development of the SOL for AREVA Advanced Mark-BW fuel, there was no impact to the SOL.

Page 9 of 28

Serial No.11-019 Docket Nos. 50-338/339 RAI Response Addition of Analytical Methodology to COLR TABLE 5-2 RCS TAVG Rod Control Signal Variable THOT TCOLD

% span

% span SCA.

0.417 SCA 0.417 SMTE3 0.167 SMTE3 0.167 SO 0.250 SO 0.250 SPE 0.000 SPE 0.000 STE 0.000 STE 0.000 SPSE 0.000 SPSE 0.000 M1,M2,M3 0.500 M4 0.500 M1MTE,M

2MTE, 0.230 M4MTE 0.230 M3MTE Output Output of 0.968 of 0.968 RTOAmp.

RTO Amp.

THOT Summator M5 0.500 M5MTE 0.120 Output of 0.835 Summator' TAVG Summator M6 0.500 M6MTE 0.090 PMA 1.700 Output of 1.956 Sumrnator" 2

Transfer equations are required to calculate the output uncertainty for these devices.

3 The value of SMTE has decreased from the analysis presented in References 5c and 5d. The value decreased from 0.170 to 0.167.

Page 10 of 28

Serial No.11-019 Docket Nos. 50-338/339 RAI Response Addition of Analytical Methodology to COLR TABLE 5-2 (continued)

RCS TAVG Rod Control Signal Variable TAVG TREF

% span

% span TAVG Summing Turbine Impulse Amp.

Pressure M7 O.SOO SeA O.SOO M7MTE 0.060 SMTE 0.20S M8 0.200 SOl 0.278 M8MTE 0.1S0 SPE 0.000 M9 0.200 STE8 1.079 M9MTE 0.1S0 SPSE 0.000 M10 0.200 M14 0.100 M10MTE 0.1S0 M14MTE 0.1S3 M11 0.2S0 M1S9 0.0 M11MTE 0.200 M1SMTE9 0.0 M12 O.SOO Output of 1.343 Lead/Lag M12MTE 0.100 TREF M13 O.SOO Summing Amp.

M13MTE 0.100 M16 0.2S0 Output of 2.331 5 M16MTE 0.100 Lead/Lag Output of Summing 0.6484 Amplifier M17 O.SOO M17MTE 0.100 Output of 0.8836 Lead/Lag 4

Transfer equations are required to calculate the output uncertainty for this device.

5 This is the Tavguncertainty input into the Tavg!Tref deviation device.

6 This is the Tref uncertainty input into the Tavg!Tref deviation device.

7 The value of SO for the turbine impulse pressure has decreased from O.SOO to 0.278 since the analysis presented in References Sc and Sd.

8 The value STE for the turbine impulse pressure has decreased from 1.367 to 1.079 since the analysis presented in References Sc and Sd.

9 The component represented by M1S and M1SMTE has been removed from the system since the analysis presented in References Sc and Sd.

Page 11 of 28

Serial No.11-019 Docket Nos. 50-338/339 RAI Response Addition of Analytical Methodology to COLR TABLE 5-2 (continued)

RCS TAVG Rod Control Signal Variable Tavg uncertainty 2.331 %

Tref uncertainty 0.883 %

Rack Effects RD 1.000 %

RTE 0.500 %

RRA 0.000 %

CA10 1.697 %

CSA 3.22%

Span 100°F CSA 3.22°F 10 The Tavg controller deadband (CA) is +/-1.5°F. This corresponds to an accuracy (1.960) of 1.697%, based on a uniform probability distribution.

Page 12 of 28

11 Serial No.11-019 Docket Nos. 50-338/339 RAI Response Addition of Analytical Methodology to COLR TABLE 5-3 Feedwater Temperature Uncertainty Feedwater Variable Temperature Uncertainty

(% span)

SE 0.000 EA 0.000 PMA 0.000 PEA 0.000 SGA 0.375 SMTE11 0.000 SD 0.375 SPE 0.000 STE 0.000 SPSE 0.500 M1 0.500 M1MTE 0.102 M2 0.000 M2MTE 0.000 M3 0.000 M3MTE 0.000 RD 1.000 RTE 0.500 RRA 0.000 GSA 1.464%

Span 200 of GSA 2.928 of SMTE included in manufacturer's accuracy for the resistance vs. temperature curve.

Page 13 of 28

Serial No.11-019 Docket Nos. 50-338/339 RAI Response Addition of Analytical Methodology to COLR TABLE 5-4 Steam Pressure Uncertainty Steam Pressure Variable Uncertainty

% span SE 0.000 EA 0.000 PMA 0.000 PEA 0.000 SGA 0.500 SMTE 0.207 SO 0.429 SPE 0.000 STE 1.475 SPSE 0.000 M1 0.100 M1MTE 0.153 M2 0.340 M2MTE 0.030 M3 0.000 M3MTE 0.000 RO 1.000 RTE 0.500 RRA 0.000 GSA 2.076%

Span 1400 psi GSA 29.1 psi Page 14 of 28

12 13 Serial No.11-019 Docket Nos. 50-338/339 RAI Response Addition of Analytical Methodology to COLR TABLE 5-5 Feedwater Flow Uncertainty Variable Feedwater Flow Uncertainty SE 0.000 EA 0.000 PMA12 0.4247 PEA 0.000 SGA12 0.500 SMTE12 0.186 S012 0.727 SPE12 0.221 STE12 1.623 SPSE 0.000 M1 0.100 M1MTE 0.153 M2 0.340 M2MTE 0.050 RO 1.000 RTE 0.500 RRA 0.000 GSA 2.308 % dp span 1.359 % flow span Span 5.00E+06 Ibmlhr GSA13 6.79E+04Ibm/hr Multiple values have changed since the analysis presented in References 5c and 5d. The value of PMA has increased from 0.0000 to 0.4247. The value of SGA has decreased from 0.750 to 0.500. The value of SMTE has decreased from 0.187 to 0.186.

The value of SO has increased from 0.500 to 0.727. The value of SPE has decreased from 1.118 to 0.221. The value of STE has increased from 1.152 to 1.623.

The GSA has increased from 2.274% dp span to 2.308% dp span from the previous submittal in Reference 5c and 5d.

Page 15 of 28

Serial No.11-019 Docket Nos. 50-338/339 RAI Response Addition of Analytical Methodology to COLR TABLE 5-6 Core Power Uncertainty No. of Parameter Span Units Instrument CSA(%)

Process Uncertainty Channels Feedwater 200 F

1 1.464 2.928 F

Temperature Main Steamline 1400 psi 3

2.076 16.7814 psi Pressure Feedwater Flow Differential 10 Volts 2

2.308 4.80E+0414 Ibm/hr Pressure Moisture 0.1

%MCO 1

100 0.1

%MCO Carryover Flow Coefficient 0.515

% power Core Power" 100 0/0 0.862 14 15 16 Includes credit for averaging of instrument channels The flow coefficient has been decreased from 1.0% to 0.5% since the time of the analysis presented in References 5c and S.d. The decrease in the flow coefficient is due to the recalibration of the feedwater flow venturis as mentioned in Attachment 5 page 10 of Reference 5e.

The 100% Core Power level refers to the pre-MUR power level of 2893 MWt.

Page 16 of 28

Serial No.11-019 Docket Nos. 50-338/339 RAI Response Addition of Analytical Methodology to COLR TABLE 5-7 RCS Hot and Cold Leg Temperature Uncertainties THOT Uncertainty TCOLD Uncertainty Variable

% span

% span THOT TCOLD PMA 0.000 PMA 0.000 PEA 0.000 PEA 0.000 SCA 0.417 SCA 0.417 SMTE 19 0.167 SMTE 19 0.167 SD 0.250 SD 0.250 SPE 0.000 SPE 0.000 STE 0.000 STE 0.000 SPSE 0.000 SPSE 0.000 Error for 0.635 %

Error for 0.635 %

Sensor" 0.440 OF Sensor 0.762 OF (120 OF span)

(120 OF span)

RCA18 0.200 OF RCA18,2O 0.200 OF RMTE 0.000 OF RMTE 0.000 OF CSA 0.483 OF CSA 0.788 OF 17 18 19 20 Includes credit for averaging of three sensors.

The instrument loop calibration results are reviewed and verified to meet the specified uncertainty.

The value of SMTE has decreased from the analysis presented in References 5c and 5d. The value decreased from 0.170 to 0.167.

The value of RCA for TCOLD has increased from the analysis presented in References 5c and 5d. The value increased from 0.100 to 0.200.

Page 17 of 28

21 22 23 24 25 Serial No.11-019 Docket Nos. 50-338/339 RAI Response Addition of Analyt ical Methodology to COLR TABLE 5-8 ReS Elbow Tap Uncertainty Elbow Tap Variable Uncertainty

% dp span SE 0.000 EA 0.000 PMA 0.000 PEA 0.000 SCA 0.750 SMTE 0.240 SD 0.375 SPE 0.000 STE24 0.000 SPSE 0.000 M1 0.100 M1MTE 0.153 M2 1.500 M2MTE 0.030 RD 1.000 RTE 0.500 RRA25 0.833 Error for p 1.088 % dp span span terms" Error for flow 2.070 % flow span span terms" CSA, lnclcator" 2.181% flow span RSS((SCA+SMTE),SD,STE, (M1+M1MTE))

RSS((M2+M2MTE),RD,RTE, RRA)

RSS(0.5*DP Span Terms*(120/95), Flow Span Terms)

The value of STE has decreased from the previous analysis of Reference5c and 5d. The value decreased from 0.906 to 0.000.

This term has been set to zero because the RCS elbow taps are normalized at operating conditions to the precision flow calorimetric following each refueling outage.

The value of RRA has decreased from the previous analysis of Reference 5c and 5d. The value decreased from 1.000 to 0.833.

Page 18 of 28

Serial No.11-019 Docket Nos. 50-338/339 RAI Response Addition of Analytical Methodology to COLR TABLE 5-9 ReS Flow Uncertainty Variable RCS Flow Uncertainty Random Uncertainty Uncertainty, %

Components Power 1.632 THOT 0.937 Streaming 1.309 TCOLD 1.407 Elbow Taps 2.181 Total Random Uncertainty 3.463 on a loop-by-Ioop basis Total Random Uncertainty 1.999 on a plant-wide basis Streaming(Bias),Plant 1.309 Total Plant Uncertainty 2.390 References 5a.

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

5b.

Letter to USNRC from J. A. Price (Dominion), "Virginia Electric and Power

Company, North Anna Power Station Units 1 and 2,

Proposed License Amendment Request (LAR), Addition of Analytical Methodology to COLR," July 19, 2010, Dominion Serial NO.10-404.

5c.

Letter to USNRC from W. R. Matthews, "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)," March 30, 2006, Dominion Serial No.06-142.

[ADAMS Accession No. ML060900631]

Page 19 of 28

Serial No.11-019 Docket Nos. 50-338/339 RAI Response Addition of Analytical Methodology to COLR 5d.

Letter to USNRC from E. S. Grecheck, "Virginia Electric and Power Company (Dominion), North Anna Power Station Unit Nos. 1 and 2, Proposed Technical Specification Changes on Addition of Analytical Methodology to the Core Operating Limits Report, Administrative Correction," April 13, 2006, Dominion Serial No. 06-142A. [ADAMS Accession No. ML061040062]

5e.

Letter to USNRC from J. A. Price (Dominion), "Virginia Electric and Power Company (Dominion),

North Anna Power Station Units 1 and 2, License Amendment

Request, Measurement Uncertainty Recapture Power Uprate,"

March 26,

2009, Dominion Serial No.09-033.

[ADAMS Accession No.

ML090900055]

5f.

Letter to D. A. Heacock (Dominion) from V. Sreenivas (USNRC), "North Anna Power Station, Unit Nos. 1 and 2, Issuance of Amendments Re: Measurement Uncertainty Recapture Power Uprate (TAC Nos. ME0965 and ME0966), October 22, 2009. [ADAMS Accession No. ML092250616]

NRC Question 6 Provide detailed calculations that were performed to obtain the correlation uncertainty factor as described (very briefly) using the 95% upper confidence limit on the VIPRE-D/WRB-2M code correlation pair measured-to-predicted (M/P) CHF ratio and standard deviation (

Reference:

DOM-NAF-2, Rev O. 1-A)

Dominion's Response Section 2.4 of VEP-NE-2-A [Reference 6a] requires that each of the calculated DNBRs to be multiplied by a random variable to include the effect of the correlation uncertainty.

Equation 6-1 differs slightly from Equation 2.4.1 in Reference 6a because its original assumption of a normally distributed qualification DNBR database was found to be incorrect.

Typically the M/P distribution is found to be normal, but not the reciprocal DNBR distribution itself.

As a consequence, the randomizing factor was re-written to reflect the normality of the M/P distribution. The correction for the predicted DNBR from the thermal-hydraulic code to account correlation uncertainty was re-written as Equation 7 in the RAls for Reference 6a.

The correlation uncertainty factor is applied to all 2000 calculated DNBRs at each of the nominal statepoints to generate the Randomized DNBR distribution for use in Section 3.6 of Attachment 4 of Reference 6b. The Randomized DNBR distribution is obtained from the unrandomized DNBR results by correcting for the code/correlation uncertainty using Equation 6-1.

DNBR

=

DNBRUnrandomized Randomized.

[1.0+s(M/ P)*K*NormalRandomNumber]

[Equation 6-1]

Where:

s(M/P) is the standard deviation of the code/correlation M/P database for the WRB-2M CHF correlation taken from Appendix C of DOM-NAF-2-A (see Table 6-1 below).

Page 20 of 28

Serial No.11-019 Docket Nos. 50-338/339 RAI Response Addition of Analytical Methodology to COLR K is a sample correction factor that depends on the size of the experimental database used to obtain the code/correlation deterministic DNB limit, which is calculated as:

K=

2 (v'2n-3-1.64S)

[Equation 6-2] (Reference 6a, Page 37, Equation 2.4.5)

The 95% upper confidence limit on CHF correlation is represented by the product of the standard deviation of the code/correlation M/P database and the sample correction factor (i.e. the product S(M/P)

• K) in Equation 6-1.

Table 6-1 CHF Code/Correlation Data

[Reference 6cl WRS-2M Average M1P 1.0006 S(M/P) 0.0640 N

241 K

1.0824 K

• S(M/P) 0.06927 Equation 6-1 has been used in the previous implementation of the Statistical DNBR Evaluation Methodology for North Anna [Reference 6d].

References 6a.

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

6b.

Letter to USNRC from J. A. Price (Dominion), "Virginia Electric and Power

Company, North Anna Power Station Units 1 and 2,

Proposed License Amendment Request (LAR), Addition of Analytical Methodology to COLR," July 19, 2010, Dominion Serial NO.10-404.

6c.

Fleet Report, DOM-NAF-2, Rev. 0.2-P-A, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code," August 2010. [Non-Proprietary Version ADAMS Accession No. ML102390419]

6d.

Letter to USNRC from E. S. Grecheck (Dominion), "Virginia Electric and Power Company (Dominion), North Anna Units 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)," May 11, 2006, Dominion Serial No. 06-142B. [ADAMS Accession No. ML061310495]

Page 21 of 28

Serial No.11-019 Docket Nos. 50-338/339 RAI Response Addition of Analytical Methodology to COLR NRC Question 7 Code uncertainty that is quantified as 5% in Topical Report VEP-NE-2-A Section 2.5 is based on COBRA/THINC and COBRA/data comparisons. Explain why this 5% code uncertainty is conservative for VIPRE-D/VIPRE-W and VIPRE-D/CHF data comparisons.

Dominion's Response The 5% code uncertainty discussed in Section 2.5 of Topical Report VEP-NE-2-A

[Reference 7a]

is based on comparisons made between COBRA and another NRC-approved thermal-hydraulic code (THINC) in NRC letter dated May 27, 1982

[Reference 7b] and comparisons made between COBRA and the W-3 CHF correlation experimental database in NRC letter dated June 12, 1981 [Reference 7c]. The VIPRE-D thermal-hydraulic computer code has also under gone similar comparison testing in DOM-NAF-2-A [Reference 7d].

The VIPRE-D thermal-hydraulic computer code was compared against AREVA's LYNXT thermal-hydraulic computer code in Section 5.0 of DOM-NAF-2 with the maximum difference in predicted DNBRs between the two codes less than 5%. This comparison is more extensive than the comparison made between COBRA and THINC thermal-hydraulic computer codes.

In Appendices A through C of DOM-NAF-2, comparisons have been made between VIPRE-D and different CHF correlation experimental databases.

Figures AA.3-1 in Appendix A, Figure B.6-2 in Appendix B, and Figure C.5-1 in Appendix C of DOM-NAF-2-A provide similar comparisons to that provided in the letter dated June 12, 1981.

In all three figures, excellent agreement is shown between VIPRE-D and the CHF correlation experimental databases.

Based on the comparison of VIPRE-D/LYNXT and VIPRE-D/CHF experimental databases, the 5% code uncertainty discussed in Section 2.5 of VEP-NE-2-A was deemed appropriate for use with the VIPRE-D thermal-hydraulic computer code.

References 7a.

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

7b.

Letter to H. R. Denton (USNRC) from R. H. Leasburg (Dominion), Topical Report VEP-FRD-33, 'VEPCO Reactor Thermal-Hydraulics Analysis Using the COBRA IIlc/MIT Computer Code,'" May 27,1982.

7c.

Letter to H. R. Denton (USNRC) from W. N. Thomas (Dominion), "Topical Report VEP-FRD-33,

'VEPCO Reactor Core Thermal-Hydraulics Analysis Using the COBRA IIIC/MIT Computer Code,'" June 12, 1981.

7d.

Fleet Report, DOM-NAF-2, Rev. 0.2-P-A, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code," August 2010. [Non-Proprietary Version ADAMS Accession No. ML102390419]

Page 22 of 28

Serial No.11-019 Docket Nos. 50-338/339 RAI Response Addition of Analytical Methodology to COLR NRC Question 8 North Anna Unit 1 Cycles 23 and 24, and Unit 2 Cycles 23 and 24 will have "mixed core" configurations which will consist of RFA-2 fuel assemblies' co-resident with AREVA AMBW fuel assemblies.

RFA-2 and AMBW assemblies exhibit different thermal-hydraulic characteristics in including pressure drop and different grid characteristics. As a response to Staff's request for supplemental information, the Licensee in the letter,10-523, dated September 9, 2010, indicated that the difference between the safety analysis (self-imposed) limit for DNB and the design limit for DNB is the available Retained DNBR margin (from statistical analysis). Statistical evaluation of DNBR, according to VEP-NE-2A, does not include any explicit evaluation of the effect of mixed-core configuration on DNBR calculation. Explain how the retained DNBR margin is used to specifically offset generic DNBR penalties such as the transition core penalty.

Dominion's Response Dominion evaluates retained DNBR margin on a cycle-specific bases consistent with Dominion's NRC-approved reload design methodology in VEP-FRD-42, Rev. 2.1-A

[Reference 8a].

The use of retained DNBR margin is consistent with Dominion's NRO*approved Statistical DNBR Evaluation Methodology in VEP-NE-2-A [Reference 8c] and is discussed in North Anna UFSAR Section 4.4.1.1 [Reference 8d].

The Reference 8a and 8c topical reports are identified as References 1 and 8 in North Anna Technical Specification 5.6.5.b, "Core Operating Limits Report" [Reference 8e].

As discussed in Section 4.3 of Dominion's LAR dated July 19, 2010 [Reference 8b],

available retained DNBR margin is the difference between the safety analysis (self-imposed) limit and the design limit.

DNBR penalties against retained margin are classified as: 1) generic fuel design issues (e.g., fuel rod bow, transition core),

2) cycle-specific violations of limits (e.g., unbounded power shapes or peaking factors),

or 3) plant operating conditions.

As stated in the LAR [Reference 8b], 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.

The available retained DNBR margin is reduced by the penalties against retained margin.

The balance of retained DNBR margin is demonstrated to be greater than 0% on a reload basis.

Reference 8a.

Topical

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

8b.

Letter to USNRC from J. A. Price (Dominion), "Virginia Electric and Power

Company, North Anna Power Station Units 1 and 2,

Proposed License Amendment Request (LAR), Addition of Analytical Methodology to COLR," July 19, 2010, Dominion Serial NO.10-404.

Page 23 of 28

Serial No.11-019 Docket Nos. 50-338/339 RAI Response Addition of Analytical Methodology to COLR 8c.

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

8d.

North Anna Updated Final Safety Analysis Report, Revision 46.

8e.

North Anna Units 1 and 2 Technical Specifications.

NRC Question 9 Table 3.6-1 lists nine sets of nominal statepoints that covers the full range of normal operation and anticipated transient conditions. Identify the transients that cover each of the set(s) of the statepoints. Also provide the range of each of the operating conditions that constitute the set of statepoints which were subjected to Monte Carlo statistical analyses.

Dominion's Response Section 3.6 of Attachment 4 of Dominion's letter dated July 19, 2010 [Reference 9a]

provides a discussion of the selection of the nominal statepoint conditions.

As discussed in Section 4.0 of VEP-NE-2-A [Reference 9b], most DNBR transients are terminated by the overtemperature liT trip function (OTLiT), which is based upon the DNBR limit and vessel-exit boiling considerations (Le., Reactor Core Safety Limit lines).

Therefore, two statepoints were selected at each of the four Reactor Core Safety Limit (RCSL) pressures (2400, 2250, 2000, and 1860 psia) because of their importance in protecting the core from DNB. For each of the RCSL lines, 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. Also from Section 4.0 of VEP-NE-2-A, another DNBR-limiting accident which is not protected by the OTLiT trip function is the Loss of Flow Accident.

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

Therefore, the selected nominal statepoints in Section 3.6 of Attachment 4 of Dominion's letter dated July 19, 2010, have been selected based on the methodology of topical report VEP-NE-2-A. The selected nominal statepoints are listed in Table 3.6-1 of of Reference 9b.

The flow rate for the low flow statepoint was -63%

which bounds (is lower than) the flow at which the minimum DNBR occurs for the loss of flow (LOFA) and locked-rotor (LOCROT) transients.

Section 3.8 of Attachment 4 of Dominion's LAR [Reference 9a] provides a verification of the selection of the nominal statepoint conditions. Therein, 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 (and SOL) for any set of conditions to which the methodology may be applied.

The Statistical DNBR Evaluation Methodology is 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, the Locked Rotor and the Single rod cluster control Page 24 of 28

Serial No.11-019 Docket Nos. 50-338/339 RAI Response Addition of Analytical Methodology to COLR assembly withdrawal at full power Accidents. The accidents to which the methodology is applicable are listed in Table 3.9-1 of Attachment 4 of Reference 9b.

The ranges of the operating conditions that constitute the set of statepoints which were subjected to Monte Carlo statistical analyses are listed in Table 9-1.

Table 9-1 Operating Ranges of Nominal Statepoints Nominal Pressurizer Inlet Temperature Power Range Flow Range Pressure Range Statepoint

[psia]

Range [OF]

[%]

[%]

A 2348.0 - 2448.4 597.1-611.7 114.8 - 121.1 94.7 - 105.6 B

2348.4 - 2450.2 606.7 - 621.0 109.3 -115.8 95.0 -107.0 C

2200.2 - 2302.7 590.0 - 603.6 114.7 -120.9 94.5 - 105.7 D

2201.3 - 2313.9 601.8 - 615.6 108.1-113.9 94.6 -105.9 E

1958.7 - 2059.9 578.6 - 592.8 114.4 -120.7 94.1 -104.6 F

1945.5 - 2059.2 590.8 - 604.9 107.8 -113.6 95.3 -104.9 G

1816.3 - 1909.6 573.6 - 588.2 115.2 - 120.8 95.0 - 106.1 H

1794.8 -1914.1 580.4 - 595.0 110.5 - 117.7 93.9 - 105.0 I

2191.7 - 2303.8 546.0 - 560.1 105.4-110.6 59.7 - 66.0 Reference 9a.

Letter to USNRC from J. A. Price (Dominion), "Virginia Electric and Power Company, North Anna Power Station Units 1 and 2,

Proposed License Amendment Request (LAR), Addition of Analytical Methodology to COLR," July 19,2010.

9b.

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

NRC Question 10 Table 4.3-1 and Section 3.3 of Attachment 4 indicate that W-3 correlation will be used in VIPRE-D code for MSLB CHF calculations when the RCS pressure drops below 1000 psia. VIPRE-01 Volume 1 Manual, Appendix D lists the applicable range of pressure for W-3 correlation is 1000 psia to 2300 psia.

Safety evaluation for topical report WCAP-9226-P-A, "Reactor Core Response to Excessive Secondary Steam release, JJ Section 2.6 indicates NRC staff approval, based on Westinghouse experimental data, justified the use of W-3 correlation in the lower pressure range of 500 -

1000 psia, presumably to be used in THINC IV thermal-hydraulic code. Provide responses to the following requests:

Page 25 of 28

Serial No.11-019 Docket Nos. 50-338/339 RAI Response Addition of Analytical Methodology to COLR (a) Provide to the agency, the Westinghouse report based on their experimental data justifying the extension of the pressure range to 500 -1000 psia.

(b) Justify the extension of the pressure range to 500 -

1000 psia in VIPRE code, since the original extension was meant for the THING IV thermal hydraulics code.

Dominion's Response Question (a)

Westinghouse provided the initial qualification of the W-3 CHF correlation to the extended pressure range of 500 to 1000 psia in letter dated March 25, 1986 [Reference 10a] in response to a NRC question.

Question (b)

The extension of the W-3 CHF correlation to a pressure range of 500 to 1000 psia has already been demonstrated for use with the VIPRE computer code, by Westinghouse in a [[letter::05000339/LER-1998-003-01, :on 980424,multiple Failures Occurred During Functional Testing of RCS Large Bore Snubbers.Cause Unknown. Large Bore Snubbers Replaced.With|letter dated October 5, 1998]] [Reference 10b]. The NRC subsequently approved the use of the W-3 correlation in VIPRE for Westinghouse in Reference 10c. In the [[letter::05000339/LER-1998-003-01, :on 980424,multiple Failures Occurred During Functional Testing of RCS Large Bore Snubbers.Cause Unknown. Large Bore Snubbers Replaced.With|letter dated October 5, 1998]], Westinghouse specifically stated that the W-3 correlation in the Westinghouse version of VIPRE is unchanged from that of the original VIPRE code.

Since Westinghouse's version of VIPRE uses the same W-3 correlation that is included in the original VIPRE code, the approval of the extension of the W-3 CHF correlation to the extended pressure range of 500 to 1000 psia applies to the W-3 CHF correlation in the original VIPRE coding.

Dominion also uses the W-3 CHF correlation from the original VIPRE coding.

Therefore, the approval in Reference 10c also applies to the W-3 CHF correlation used by Dominion. The NRC has approved Dominion's use of the W-3 correlation to the extended pressure range of 500 to 1000 psia with a 95/95 DNBR limit of 1.45 in References 10d and 10e.

References 10a.

Letter to USNRC from E.

P.

Rahe (Westinghouse),

NS-NRC-86-3116, "Westinghouse Response to NRC Additional Request on WCAP-9226-P/wCAP-9227-N-P, 'Reactor Core Response to Excessive Secondary Steam Releases,'

(Non-Proprietary)," March 25, 1986.

10b.

Letter to USNRC from H. A. Sepp (Westinghouse), "Response to Supplement Request for Additional Information on WCAP-14565, Rev. 0, 'VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal Hydraulic Safety Analysis,' [Proprietary]," October 5, 1998.

10c.

Letter to H. A. Sepp (Westinghouse) from T. H. Essig (USNRC), "Acceptance for Referencing of Licensing Topical Report WCAP-14565, 'VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal Hydraulic Safety Analysis,' (TAC No. M98666)," January 19, 1999.

Page 26 of 28

Serial No.11-019 Docket Nos. 50-338/339 RAI Response Addition of Analytical Methodology to COLR 10d.

Letter to D. A. Christian (Dominion) from C. I. Grimes (USNRC), "Millstone Power Station, Unit Nos. 2 and 3 (Millstone 2 and 3), North Anna Power Station, Unit Nos. 1 and 2 (North Anna 1 and 2), and Surry Power Station, Units Nos. 1 and 2 (Surry 1 and 2) - Approval of Dominion's Fleet Report DOM-NAF-2, 'Reactor Core Thermal-Hydraulics Using the VIRE-D Computer Code' (TAC Nos.

MC4571, MC4572, MC4574, MC4575, and MC4576)," April 4, 2006. [ADAMS Accession No. ML060790496]

10e.

Letter to D. A. Christian (Dominion) from S. R. Monarque (USNRC), "Millstone Power Station, Unit Nos. 2 and 3, North Anna Power Station, Unit Nos. 1 and 2, and Surry Power Station, Units Nos. 1 and 2 - Correction to Dominion's Fleet Report DOM-NAF-2,

'Reactor Core Thermal-Hydraulics Using the VIRE-D Computer Code,')" June 23, 2006. [ADAMS Accession No. ML061740212]

NRC Question 11 Provide summary of reload safety/licensing analyses that are being performed for North Anna power station in support of the fuel transition starting in Spring 2012.

This report should contain summaries of licensing analyses performed for potentially limiting events and analyses that are identified when disposition of other events.

Dominion 's Response Dominion's submittal of July 19, 2010 [Reference 11a] only requests approval for the use of the Statistical DNBR Evaluation Methodology with the VIPRE-DIWRB-2M code/correlation for the Westinghouse 17x17 RFA-2 fuel and the resulting SOL. Thus, the requested information includes the assessments of

LOCA, non-LOCA, and containment analysis and is not directly germane to this review.

Furthermore, these analyses will be completed prior to fuel product implementation at Unit 1 (the lead unit) in spring 2012. These analyses are presently underway and can be made available for a regulatory audit following their completion, as requested by the NRC in letter dated September 17, 2010 [Reference 11b].

References 11a.

Letter to USNRC from J. A. Price (Dominion), "Virginia Electric and Power

Company, North Anna Power Station Units 1 and 2,

Proposed License Amendment Request (LAR), Addition of Analytical Methodology to COLR," July 19, 2010, Dominion Serial NO.10-404.

11 b.

Letter to D. A. Heacock (Dominion) from V. Sreenivas (USNRC), "North Anna Power Station Units 1 and 2:

Acceptance Review of Proposed License Amendment Request, Addition of Analytical Methodology to COLR (TAC Number ME4262 and ME4263)," September 17, 2010.

Page 27 of 28

Serial No.11-019 Docket Nos. 50-338/339 RAI Response Addition of Analytical Methodology to COLR NRC Question 12 Provide details of how the change in fuel thermal conductivity with irradiation (burnup) during AOOs and design basis accidents is accounted for by the fuel vendor. Provide the methodology, computer code and procedures to track the variation of fuel thermal conductivity with burnup.

Dominion's Response Dominion's LAR submittal of July 19, 2010 [Reference 12a], only requests approval for the use of the Statistical DNBR Evaluation Methodology with the VIPRE-D/WRB-2M code/correlation for the Westinghouse 17x17 RFA-2 fuel and the resulting SDL.

The method for how the vendor tracks changes in the fuel thermal conductivity with irradiation is not directly germane to this review.

However, in accordance the NRC's acceptance review in letter dated September 17, 2010 [Reference 12b], this information can be made available for a regulatory audit.

References 12a.

Letter to USNRC from J. A. Price (Dominion), "Virginia Electric and Power

Company, North Anna Power Station Units 1 and 2,

Proposed License Amendment Request (LAR), Addition of Analytical Methodology to COLR," July 19, 2010, Dominion Serial NO.10-404.

12b.

Letter to D. A. Heacock (Dominion) from V. Sreenivas (USNRC), "North Anna Power Station Units 1 and 2: Acceptance Review of Proposed License Amendment Request, Addition of Analytical Methodology to COLR (TAC Number ME4262 and ME4263)," September 17, 2010.

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