ML061310495
| ML061310495 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 05/11/2006 |
| From: | Grecheck E Dominion, Virginia Electric & Power Co (VEPCO) |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 06-142B, TAC MC7526, TAC MC7527 | |
| Download: ML061310495 (8) | |
Text
VIRGINIA ELECTRIC AND POWER COMPANY
- RICHMOND, VIRGINIA 23261 May 11, 2006 U.S. Nucllear Regulatory Commission Serial No.
06-142B Attention:: Document Control Desk NL&OS/ETS RO Washington, D.C. 20555 Docket Nos. 50-338/339 License Nos. NPF-417 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)
By letter dated July 5, 2005 (Serial No.05-419), Dominion submitted proposed license amendments for North Anna Unit Nos. 1 and 2. The proposed changes would add a reference in Technical Specification 5.6.5. b, Core Operating Limits Report (COLR), to allow the use of an alternate methodology to perform a thermal-hydraulics analysis to predict the critical heat flux and departure from nucleate boiling ratio (DNBR) for the Advanced Mark-BW fuel.
In addition, Dominion requested the Nuclear Regulatory Commission (NRC) staffs approval of the site/fuel type/code specific Statistical Design Limits obtained by the plant specific implementation of the NRC-approved methodology documented in Topical Report VEP-NE-2-A, Statistical DNBR Evaluation Methodology. In a facsimile dated April 25, 2006, the NRC staff requested further additional information to complete the review. The attachment to this letter provides the requested information.
Dominion continues to request approval of this license amendment request by September 1, 2006. This requested schedule permits in-house performance of DNB analyses with DOM-NAF-2 and the VIPRE-D/BWU code/correlation set in support of the use of AREVA AMBW fuel at North Anna Power Station Units 1 and 2 for operating cycles 20 and 19, respectively. This change will be implemented within 60 days of NRC approval.
If you have any questions or require additional information, please contact Mr. Thomias Shaub at (804) 273-2763.
Very truly yours, Eugene 5;. Grecheck Vice President - Nuclear Support Services Attachment Commitments made in this letter: None
06-1 42B Docket Nos. 50-338/339 Page 2 of 3 cc:
U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW Suite 23T85 Atlanta, Georgia 30303 Mr. J. T. Reece NFC Senior Resident Inspector Nolrth Anna Power Station Mr. Stephen R. Monarque NRC Project Manager - Surry and North Anna U. S. Nuclear Regulatory Commission Orie White Flint North 1 1555 Rockville Pike Rockville, Maryland 20852 Mr. J. E. Reasor, Jr.
Old Dominion Electric Cooperative lnrisbrook Corporate Center 42101 Dominion Blvd.
Suite 300 Glen Allen, Virginia 23060 Commissioner Bureau of Radiological Health 1500 East Main Street Suite 240 Richmond, Virginia 23218
06-1 428 Docket Nos. 50-338/339 Page 3 of 3 COMMONWEALTH OF VIRGINIA
)
)
COUNTY OF HENRICO
)
The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Eugene S. Grecheck, who is Vice President -
Nuclear Support Services, 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 2006.
My Commission Expires:
Y Notary Public (SEAL)
ATTACHMENT Serial No. 06-142B 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)
Virginia Electric and Power Company (Dominion)
North Anna Power Station Units 1 and 2
Serial No. 06-1426 Docket Nos. 50-338/339 Virqinia Electric and Power Companv (Dominion)
North Anna Power Station Unit Nos. 1 and 2 Response to Request for Additional Information on Proposed Technical Specification Changes on Addition of Analvtical Methodologv to the Core Operatinq Limits Report (Tac Nos. MC7526 And MC7527)
Background
By letter dated July 5, 2005, Virginia Electric and Power Company (Dominion) submitted proposed license amendments to add a reference in Technical Specification 5.6.5. b, Core Operating Limits Report (COLR), to permit the use of an alternate methodology to perform a thermal-hydraulic analysis to predict the critical heat flux (CHF) and departure from nucleate boiling (DNB) ratio (DNBR) for the Advanced Mark-BW (AMBW) fuel at North Anna Power Station, Unit Nos. 1 and 2 (North Anna 1 and 2).
In a letter dated February 14, 2006, the NRC staff requested additional information to complete the review of the proposed Technical Specification and statistical design limit.
This information was provided in a letter dated March 30, 2006. In a subsequent facsimile, dated April 25, 2006, the NRC requested further information to clarify information provided in the original July 5, 2005 submittal.
NRC Question 1 Table 3.21-1 lists the DNBR limits to be 1.20 (above 700 psia) for the BWU-Z correlation and 1.18 (above 594 psia) for the BWU-ZM. Appendix A to topical report DOM-NAF-2 lists the corrected standard deviations to be 0.0919 (Table A.4.1-2) and 0.0875 (7able A.4.2-2) for BWU-Z and BWU-ZZM, respectively. In combining the two correlations into B WU-ZZM, Section 3.1.3 states that %because additional experimental code/correlations uncertainty....
In this implementation, B WU-ZM code/correlation uncertainilies were used to obtain the BWU-ZZM SDL, because they are slightly more conservative.
Clarify the apparent inconsistency in stating that the B WU-ZM uncertainty is slightly larger and more conservative.
Do mi niori Response The randomized DNBR distribution is obtained from the unrandomized MDNBR results by correcting for the code/correlation uncertainty using equation 1.
where:
0 s(hf/P) is the standard deviation of the codekorrelation M/P database for the CHF correlation under study (see Table 1-1 below).
Page 1 of 4
Serial No. 06-1428 Docket Nos. 50-338/339 Average WP S( WP)
N Kis 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:
BWU-2 BWU-ZM 0.9950 1.01 38 0.0907 0.0875 528 148 K = /-
2. (n - 1)
(I/=
- 1.645)2
[equation 21 (Reference 2, Page 37, equation 2.4.5)
K t
K*S(M/P) 1.05390 1.1 0820 0.09559 0.09697 To randomize the MDNBR results obtained for the BWUZZM correlation, the BWU-ZM code/correlation uncertainties were used, as they happen to be slightly more conservative than the code/correlation uncertainties for BWU-Z. This is because the Equation 1 accounts not only for the standard deviation of the CHF experimental database, but also for the size of the database. When taking both into account BWU-ZM is slightly more conservative, i.e. the product K
- S(M/P) is larger.
Equation 1 differs slightly from Equation 2.4.1 in Reference 2 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, and it was defined in Equation 7 in the request for additional information (RAI) for Reference 2. This equation has been used in previous implemenitations of the Statistical DNBR Evaluation Methodology, such as Reference 3 (see Equation 3.2.2-1) and Reference 4.
NRC Question 2 Explain hlow the numbers in Columns 2 and 3 (randomized DNB SDNBR and Total DNB ST&
in Tables 3.1.6-3 and 3.1.6-4 are obtained.
Dominion Response The Randomized DNBR distribution is obtained for each statepoint using Equation 1 as described in detail in the response to Question #1. The Randomized DNBR distribution is then evaluated to determine the Randomized DNBR SDNBR, which is shown in Column 2 of Tables 3.1.6-3 and 3.1.6-4.
Page 2 of 4
Serial No. 06-142B Docket Nos. 50-338/339 Column 3, the Total STotal, is obtained using the Root-Sum-Square method according to equation 3:
where:
0 S D ~ ~ B R is the standard deviation for the Randomized DNBR distribution.
0 The factor { i F - l. O i i s the uncertainty in the standard deviation of the 2,000 Monte Carlo simulations, and provides a 95% upper confidence limit on the standard deviation.
is the uncertainty in the mean of the correlation. N is the number of Ak experimental datapoints in the original correlation database.
0 F, is the code uncertainty, that has been defined as 5% (20 value), i.e. 5.0%/1.645
=3.04% (1 (J value). See Section 3.1.5 in Reference 1.
0 FM is the model uncertainty, which is 0.0 in our case as we are running the Monte Cairlo simulation with the production model (see Section 3.1.4 in Reference 1).
Note that Uhis equation differs slightly from the equation listed in Reference 2. It has a new factor applied to the Randomized DNBR SDNBR, the yfi factor to correct for the uncertainty in the mean of the correlation. This factor has been used in previous imp1emeni:ations of the Statistical DNBR Evaluation Methodology, such as Reference 3 and Reference 4.
NRC Question 3 In Section 3.1.7, it is stated that the SDLs for the BWU-ZZM and BWU-N correlations are increased to 1.34 and 1.38, respectively, so that 99.9% of fuel rods in the core would not experience DNB. However, Statepoint B in Table 3.1.7-I and Statepoint A in Table 3.7.1-2 show the rods in DNB are 0.092% and 0.091%, respectively, for the chosen SDL for BWU-ZZM and BWU-N.
Clarify whiy it is not necessary to increase the SDLs further so that 0.1 % of the rods core-wide would experience DNB.
Page 3 of 4
Serial No. 06-142B Docket Nos. 50-338/339 Do mi nioin Response The Pin Peak SDL95195 values in Tabl& 3.1.6-3 anc 3.1.6-4 result in less than 0.1 Yo of the rods in DNB on a core-wide basis. According to page 52 of Reference 2, the applicaticin of a higher SDL for a fixed DNBR standard deviation will yield a lower number alf rods in DNB.
Statepoirnt B in Table 3.1.7-1 shows the rods in DNB is 0.092% with the applied SDL of 1.34. The SDL that would result in 0.1 Yo of the rods in DNB on a core-wide basis for Statepoinit B in Table 3.1.7-1 would be less than 1.34, but greater than 1.33.
Statepoinit A in Table 3.1.7-2 shows the rods in DNB is 0.091 Yo with the applied SDL of 1.38. The SDL that would result in 0.1 Yo of the rods in DNB on a core-wide basis for Statepoirnt A in Table 3.1.7-2 would be less than 1.38, but greater than 1.37.
It is conslervative to use the largest Full Core SDL99.9 from all the evaluated statepoints for application to safety analysis. Thus, the SDLs for the BWU-Z/ZM and BWU-N correlations were appropriately selected as 1.34 and 1.38, respectively.
References
- 1. Letter from E. S. Grecheck (Dominion) to US NRC Document Control Desk, 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. 05-41 9, July 5, 2005.
- 2. Topical Report, VEP-NE-2-A, Statistical DNBR Evaluation Methodology, June 1987.
- 3. Letter from W. L. Stewart (VEPCO) to US NRC Document Control Desk, Virginia Electriic and Power Company North Anna Power Station Units 1 and 2 - Proposed Technical Specification Changes, Serial No.87-231, June 17, 1987.
- 4. Letter from W. L. Stewart (VEPCO) to US NRC Document Control Desk, Virginia Electric and Power Company Surry Power Station Units 1 and 2 - Proposed Techniical Specification Changes - FAh Increase/Statistical DNBR Methodology, Serial No. 91 -374, July 8, 1991.
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