ML110040003

E-mail Request for Additional Information Addition of Analytical Methodology to COLR

ML110040003
Person / Time
Site:	North Anna
Issue date:	01/03/2011
From:	V Sreenivas Plant Licensing Branch II
To:	Heacock D Virginia Electric & Power Co (VEPCO), Dominion
References
TAC ME4262, TAC ME4263
Download: ML110040003 (4)

Text

From: Sreenivas, V Sent: Monday, January 03, 2011 3:39 PM To: 'david.heacock@dom.com' Cc: 'Tom Shaub'; 'david.sommers@dom.com'; Panicker, Mathew; Kulesa, Gloria

Subject:

REQUEST FOR ADDITIONAL INFORMATION: NORTH ANNA POWER STATION UNITS 1 AND 2 -ADDITION OF ANALYTICAL METHODOLOGY TO COLR (TAC No ME4262, ME4263)

By letter dated July 19, 2010, Virginia Electric and Power Company (VEPCO) requests amendments, in the form of changes to the Technical Specifications (TS) to Facility Operating License Numbers NPF-4 and NPF-7, for North Anna Power Station Units 1 and 2 (NAPS), respectively. The proposed change would revise TS 5.6.5.b, CORE OPERATING LIMITS REPORT to include NRC-approved methodology, 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. The addition of the above approved methodology to TS 5.6.5.b would allow Dominion to use the VIPRE-D/WRB-2M and VIPRE-D/W-3 correlation pairs to perform licensing calculations with Westinghouse RFA-2 fuel in NAPS cores. The purpose of this email is to provide the results of the U.S. Nuclear Regulatory Commission (NRC) staffs detailed technical review of this request.

Please submit the following additional information in order to complete its detailed technical review.

REQUEST FOR ADDITIONAL INFORMATION (RAI)

VIRGINIA ELECTRIC AND POWER COMPANYNORTH ANNA POWER STATION UNITS 1 AND 2, DOCKET No. 50-338 AND 50-339

1. WCAP-15025PA, DOM-NAF-2 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.

2. NAPS LAR No.10-404 Attachment 1, Section 4.0 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 licensees 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 transients analyses at NAPS.

3. LTR-NRC-02-55 FCEP Notification of the RFA-2 Design, Revision 1 indicates that the impacts on Non-LOCA Fuel Clad Temperature (Item l 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.) Please submit a copy of this document to the agency for staffs information.

4. Thermal-Hydrodynamic stability 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.

5. LAR Attachment 4, 3.2 Uncertainty Analysis 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 2 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.

6. VEP-NE-2-A, LAR Attachment 4, 3.1 Methodology Review 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 0.1-A)

7. LAR Attachment 4, Section 3.5 Code Uncertainty 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.

8. Transition core DNBR penalty 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 Staffs 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.

9. LAR Attachment 4, Section 3.6 Monte Carlo Calculations 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.

10. LAR Attachment 4, Section 4, Applicability of W-3 correlation 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, 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:

(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 THINC IV thermal hydraulics code.

11. Reload safety analyses 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.

12. Fuel performance and material properties - Fuel Criteria Evaluation Process (FCEP)

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.

Please submit the response to these RAIs by February 21, 2011. If you have any questions please contact me.

V. Sreenivas, PH.D., C.P.M.,

Project Manager, Rm.#O8F6, LPL2-1 North Anna Power Reactors, Units 1 and 2 Division of Operating Reactor Licensing-NRR (301) 415-2597, v.sreenivas@nrc.gov

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