ML080350698

Request for Additional Information for Extended Power Uprate - Round 16 (TS-431 and TS-418) (TACs MD5262, MD5263, MD5264)

ML080350698
Person / Time
Site:	Browns Ferry
Issue date:	02/29/2008
From:	Ellen Brown NRC/NRR/ADRO/DORL/LPLII-2
To:	Campbell W Tennessee Valley Authority
Brown E. A.
References
TAC MD5262, TAC MD5263, TAC MD5264, TS-418, TS-431
Download: ML080350698 (6)

Text

February 29, 2008 Mr. William R. Campbell, Jr.

Chief Nuclear Officer and Executive Vice President Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801

SUBJECT:

BROWNS FERRY NUCLEAR PLANT, UNITS 1, 2 AND 3 C REQUEST FOR ADDITIONAL INFORMATION FOR EXTENDED POWER UPRATE - ROUND 16 (TS-431 AND TS-418) (TAC NOS. MD5262, MD5263, AND MD5264)

Dear Mr. Campbell:

By letters dated June 28 and 24, 2004, the Tennessee Valley Authority (TVA, the licensee) submitted amendment requests for Browns Ferry Nuclear Plant (BFN), Unit 1 and Units 2 and 3, respectively, as supplemented by letters dated August 23, 2004, February 23, April 25, June 6, and December 19, 2005, February 1 and 28, March 7, 9, 23 and 31, April 13, May 5 and 11, June 12, 15, 23 and 27, July 6, 21, 24, 26, and 31, December 1, 5, 11 and 21, 2006, January 31, February 16, and 26, and April 6, 18 and 24, March 6, July 27, August 13, and 21, September 24, November 15 and 21, and December 14, 2007. The proposed amendment would change the BFN operating licenses for all three units to increase the maximum authorized power level by approximately 15 percent.

A response to the enclosed Request for Additional Information is needed before the Nuclear Regulatory Commission staff can complete the review. This request was discussed with Mr. James Emens of your staff on February 5, 2008, and it was agreed that TVA would respond by March 6, 2008. A proprietary version of these requests was provided in separate correspondence.

If you have any questions, please contact me at (301) 415-2315.

Sincerely,

/RA/

Eva A. Brown, Senior Project Manager Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-259, 260, and 50-296

Enclosure:

Request for Additional Information cc w/enclosure: See next page

ML080350698 NRR-088 OFFICE LPL2-2/PM LPL2-2/LA DSS/SRXB DSS/SNPB LPL2-2/BC NAME EBrown:sp BClayton GCranston*

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TBoyce DATE 2/14/08 2/14/08 2/01/08 1/31/08 2/29/08

Letter to William R. Campbell, Jr. from Eva A. Brown, dated February 29, 2008

SUBJECT:

BROWNS FERRY NUCLEAR PLANT, UNITS 1, 2, AND 3 REQUEST FOR ADDITIONAL INFORMATION FOR EXTENDED POWER UPRATE - ROUND 16 (TS-431 AND TS-418) (TAC NOS. MD5262, MD5263, AND MD5264)

DISTRIBUTION:

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RidsNrrLABClayton (Hard Copy)

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RidsNrrDssSbwb (GCranston)

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TAlexion MRazzaque TNakanishi PYarsky LWard

REQUEST FOR ADDITIONAL INFORMATION EXTENDED POWER UPRATE TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT, UNITS 1, 2, AND 3 DOCKET NOS.50- 259, 50-260, AND 50-296 EMEB 167./134. Tennessee Valley Authority (TVA) plans to apply a bump-up factor to the main steam line (MSL) strain gage signals under current licensed thermal power (CLTP) conditions and (( ))

Provide data showing how the bump-up factor is determined and how it is applied to the CLTP MSL strain gage signals.

SRXB (previously SBWB)

(Unit 1 Only)

73.

The Nuclear Regulatory Commission (NRC) is looking for TVA to demonstrate that the operating experience with reactors near equilibrium operation is directly applicable to the Cycle 8 Unit 1 core design. To address this concern regarding the applicability of the design basis eigenvalue, the licensee provided plant specific Cycle 7 (restart core) operating data for the initial startup and one mid-cycle cold critical measurement. Additional information is still needed to evaluate the performance of the design basis eigenvalue determination process with an absence of plant specific operational data.

To assist the evaluation of the methods uncertainties regarding the determination of the eigenvalue, provide any additional data for hot operating conditions and design targets. Specifically, for Cycle 7, provide a comparison of the plant operating critical eigenvalues to the hot critical design bases.

Also, provide the results of any low power hot statepoints with high control blade densities.

(Unit 1/ Units 2 and 3) 74./86.

Pellet clad interaction (PCI) and stress corrosion cracking (SCC) phenomena can cause clad perforation resulting in leaking fuel bundles and resultant increased reactor coolant activity. Therefore, the staff requests the licensee to provide the following additional information regarding PCI/SCC for Units 1, 2, and 3 at EPU conditions:

Enclosure

a.

Describe any differences in operating procedures associated with PCI/SSC at EPU conditions versus pre-EPU operations.

b.

From the standpoint of PCI/SCC, discuss which of the Anticipated Operational Occurrences (AOOs), if not mitigated, would most affect operational limitations associated with PCI/SSC.

c.

For the AOOs in part b), discuss the differences between the type of required operator action, if any, and the time to take mitigating actions between pre-EPU and EPU operations.

d.

If the EPU core will include fuels with non-barrier cladding that have less built-in PCI resistance, then demonstrate by plant-specific analyses that the peak clad stresses at EPU conditions will be comparable to those calculated for the current operating conditions.

e.

Describe operator training on PCI/SCC operating guidelines.

(Units 2 and 3 only)

87.

To address the adequacy of benchmark data associated with neutronic power prediction methods, the staff understands that the issue was addressed by the fuel vendor by increasing the power distribution uncertainties and propagating them into the safety limit minimum critical power ratio (SLMCPR) calculation. Provide the following additional information:

a.

Discuss the applicability of this approach to projected Units 2 and 3 operations using ATRIUM-10 fuel and AREVA methodologies.

b.

Justify the use of the local and radial power distribution uncertainties based on Quad Cities gamma scans in light of the harder neutron spectrum present in EPU cores.

88.

To address the adequacy of void-quality correlation bias and uncertainties, the staff understands that a plant specific calculation can be performed to assess the impact of the uncertainties on the operating limit minimum critical power ratio (OLMCPR). Provide the following additional information:

a.

Discuss how the void-quality correlation bias and uncertainties are addressed for the projected Units 2 and 3 operation at EPU conditions.

b.

Determine the net impact on the OLMCPR from a bias in the void-quality correlation within the uncertainty range based on full scale test data.

89.

To address the effect of bypass boiling on the stability oscillation power range monitor (OPRM) setpoints, a setpoint setdown was performed. Provide the following additional information:

a.

Discuss how the bypass boiling effect is addressed for the Units 2 and 3 OPRM setpoints.

b.

Determine a method for conservatively accounting for the effect of bypass void formation on OPRM and average power range monitor sensitivity.

90.

Provide the following information regarding the AREVA loss-of-coolant accident analyses:

the flow area above the hot bundle exit, power of the hot bundle, and perform an analysis assuming little or no downflow