ML070190473

RAI, Extended Power Uprate - Round 12 (TAC Nos. MC3812, MC3743, MC3744) (TS-431 and TS-418)

ML070190473
Person / Time
Site:	Browns Ferry
Issue date:	01/26/2007
From:	Ellen Brown NRC/NRR/ADRO/DORL/LPLII-2
To:	Singer K Tennessee Valley Authority
Brown, E, NRR/DORL, 415-2315
References
TAC MC3743, TAC MC3744, TAC MC3812
Download: ML070190473 (7)

Text

January 26, 2007 Mr. Karl W. Singer 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, UNIT 1, 2, AND 3 REQUEST FOR ADDITIONAL INFORMATION FOR EXTENDED POWER UPRATE - ROUND 12 (TS-431 AND TS-418) (TAC NO. MC3812, MC3743 AND MC3744)

Dear Mr. Singer:

By letters dated June 28 and 25, 2004, the Tennessee Valley Authority (TVA, the licensee) submitted amendment requests for Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3, 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, and July 6, 21, 24, 26, and 31, and August 4, 16, and 18, and September 1 and 15, October 3, 5, and 13, and November 6 and 7, and December 11, 2006.

The proposed amendments would change the BFN operating licenses to increase the maximum authorized power level by approximately 20 percent above the current maximum authorized power level for Unit 1, and approximately 15 percent for Units 2 and 3. A response to the enclosed Request for Additional Information is needed before the Nuclear Regulatory Commission staff can complete the review.

A reply is expected within 30 days of the issuance of this letter. If you have any questions, please contact me at (301) 415-2315.

Sincerely,

/RA/

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

Enclosure:

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

January 26, 2007 Mr. Karl W. Singer 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, UNIT 1, 2, AND 3 REQUEST FOR ADDITIONAL INFORMATION FOR EXTENDED POWER UPRATE - ROUND 12 (TS-431 AND TS-418) (TAC NO. MC3812, MC3743 AND MC3744)

Dear Mr. Singer:

By letters dated June 28 and 25, 2004, the Tennessee Valley Authority (TVA, the licensee) submitted amendment requests for Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3, 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, and July 6, 21, 24, 26, and 31, and August 4, 16, and 18, and September 1 and 15, October 3, 5, and 13, and November 6 and 7, and December 11, 2006.

The proposed amendments would change the BFN operating licenses to increase the maximum authorized power level by approximately 20 percent above the current maximum authorized power level for Unit 1, and approximately 15 percent for Units 2 and 3. A response to the enclosed Request for Additional Information is needed before the Nuclear Regulatory Commission staff can complete the review.

A reply is expected within 30 days of the issuance of this letter. If you have any questions, please contact me at (301) 415-2315.

Sincerely,

/RA/

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

Enclosure:

Request for Additional Information cc w/enclosure: See next page Distribution: See next page ADAMS Accession No. ML070190473 NRR-088 OFFICE LPL2-2/PM LPL2-2/PM LPL2-2/PM LPL2-2/LA DSS/EEMB DSS/SBWB LPL2-2/BC NAME LRegner EBrown MChernoff BClayton KManoly GCranston MChernoff for LRaghavan DATE 1 / 24 /07 1 / 23 /07 1 / 25 /07 1 / 23 /07 1 / 24 /07 1 /25/07 1/26 /07

OFFICIAL RECORD COPY

SUBJECT:

BROWNS FERRY NUCLEAR PLANT, UNITS 1, 2, AND 3 REQUEST FOR ADDITIONAL INFORMATION FOR EXTENDED POWER UPRATE - ROUND 12 (TS-431 AND TS-418) (TAC NOS. MC3812, MC3743 AND MC3744)

Document Date: January 26, 2007 DISTRIBUTION:

PUBLIC LPL2-2 R/F RidsNrrDorlLpl2-2 RidsNrrLABClayton (Hard Copy)

RidsNrrLACGoldstein RidsNrrPMMChernoff RidsNrrPMEBrown RidsAcrsAcnwMailCenter RidsOgcRp RidsRgn2MailCenter RidsNrrDorl (CHaney/McGinty)

RidsNrrDorlDpr RidsNrrDssEemb (KManoly)

RidsNrrDssSbwb (GCranston)

RidsNrrDss (TMartin)

TScarbrough CWu TAlexion RidsNrrDeEema (KManoly)

GThomas THuang ZAbdullahi LWard LLois LRegner

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 SBWB (Unit 1 only)

57. State what reload analyses will be performed to support the actual Cycle 7 extended power uprate (EPU) operation plan, which could consist of initial105-percent power operation followed by subsequent operation at EPU conditions.

58. For Cycle 7, address whether the Supplemental Reload License Report (SRLR) and Core Operating Limit Report (COLR) will be revised based on the actual operating plan.

Discuss any deviation from the standard reload analysis set, and address whether any analyses will be dispositioned based on a bounding analyses approach.

59. Since there is no specific operational plan for Cycle 7, explain how the maximum decrease in the shutdown margin (SDM) at cycle exposures other than begininng-of-cycle (e.g., R value) will be calculated before the restart of the plant. Address whether the R value will be calculated for the assumed operation at 105 percent and at 120 percent, or if the SDM calculation will be revised based on the actual operational plan.

60. Before implementation of the Unit 1 restart, provide the COLR and the SRLR for the finalized Cycle 7 operational plan. If separate reload SRLRs were performed, provide both of them.

61. In a request for additional information (RAI) dated August 10, 2006, the Nuclear Regulatory Commission (NRC) staff asked in SBWB question 37d. why a local critical SDM demonstration was not warranted for Unit 1 restart. The NRC staff recognizes that disabling the banked position withdrawal sequence, which mitigates a control rod (CR) drop accident, was not prudent in order to perform the local critical SDM demonstration.

However, considering the uniqueness of Unit 1 Cycle 7 and 8, performing the startup physics test may still be prudent and warranted. Discuss whether the one-rod-out subcriticality test will be performed before the startup-criticality CR pulls to ensure that there is sufficient subcriticality with the highest worth CR withdrawn. Address whether this will be performed for all the analytically-determined high worth CRs before restart of the criticality CR pulls.

Enclosure

62. State all planned startup physics tests.

63. a. Provide the rod density and the R value for the SDM calculation, based on the combined power level operation.

b. Discuss if the operation at the lower power level will decrease the additonal design SDM (e.g., the 1.5-percent design margin for EPU), considering the CR patterns and the power distributions expected relative to the projected EPU operation.

c. Provide a comparative assessment of the components of the design SDM to show that operation at the lower power level will not result in lower available design margin to account for potentially higher modeling uncertainties.

Specifically, provide assurances that the control of the additional reactivity through deep CRs for operation at the lower power levels will not result in a higher CR worth, such that the additional design margin is reduced.

64. In general, small break loss-of-coolant accidents tend to be relatively insensitive to stored energy; therefore, fuel type does not influence the difference in limiting breaks on Units 1, 2, and 3. Explain why the 0.5 ft2 break is limiting for Units 2 and 3 while the 0.06 ft2 discharge break is limiting for Unit 1. Provide peak centerline temperatures (PCTs) for breaks in the range of 0.04 - 0.07 ft2 for Units 2 and 3.

65. Provide the PCT for a top-peaked axial distribution for the 0.5 ft2 break. Explain why a top-peaked axial power distribution is less limiting for the 0.5 ft2 break for Units 2 and 3.

66. Provide the plot and detailed results for the 0.5 ft2 break from the June 12, 2006, submittal with the ANP-2541 enclosure. Ensure the two-phase level in the core region is included in the package.

EEMB 119. (Unit 1) In TVA response to RAI EEMB.39, the steam separators at Units 2 and 3 were evaluated for the EPU condition, since they are the most critical component next to the steam dryers affected by the increase steam flow within the reactor vessel. Confirm whether and how the identical quantitative evaluation can be applied for the Unit 1 steam separator at the EPU operation. If not, provide a summary evaluation for the Unit 1 separators.

120. (Unit 1) In reference to EEMB.42 of July 26, 2006, transmittal, provide a summary evaluation of recirculation pumps and their supports at the EPU condition. The requested information should include the calculated stresses and cumulative usage factors (CUFs) for the critical components of the recirculation pump in comparison with the allowable limits. Also, provide a discussion of the potential for vane passing frequency vibration at the EPU condition at Unit 1.

89. (Unit 2) In the response to EEMB.40 of the July 26, 2006, submittal, Footnote 5 indicated that the current design basis stress shown above corresponds to the material stress allowable for this equation of 40,500 pounds per square inch

(psi), while the EPU stress corresponds to a location with a material stress allowable of 36,000 psi. Provide a summary of evaluation pertaining to the calculated stress for the EPU condition, which is almost equal to the allowable limit. Also specify the component material and the calculated stresses at these two maximum stress locations for current and EPU conditions for the Unit 2 feedwater loop A.

121/90. (Units 1, 2, and 3) In reference to Section 3.4 of NEDC-33101P, (a) confirm whether CUFs given in Table 3-4 are evaluated for 60 years, since the plants current licensing basis includes the 20 years of life extension at the EPU condition. If so, explain why the table indicates the evaluation was done for a 40-year life of EPU operation; (b) the CUF for the feedwater nozzle at the EPU condition is less than that for the current operating condition. Confirm whether the calculation has taken account for the increase of thermal transient due to the increase in flow at the nozzle for the EPU. Also, provide a summary of calculation for the CUF for Units 2 and 3 at feedwater nozzle, which is almost equal to 1.0 for the EPU operation; (c) the CUF for the support skirt is very small in comparison with that of the current CUF. Provide the summary of calculation and the technical basis to show that the peak stress and the number of cycles are greatly reduced for the EPU operation at the EPU condition for the Unit 1 support skirt. Also, confirm whether and how the calculated CUFs for the critical components such as main closure studs, support skirt and recirculation outlet nozzle remain unchanged for the EPU condition.

Mr. Karl W. Singer BROWNS FERRY NUCLEAR PLANT Tennessee Valley Authority cc:

Mr. Ashok S. Bhatnagar, Senior Vice President Mr. Robert G. Jones, General Manager Nuclear Operations Browns Ferry Site Operations Tennessee Valley Authority Browns Ferry Nuclear Plant 6A Lookout Place Tennessee Valley Authority 1101 Market Street P.O. Box 2000 Chattanooga, TN 37402-2801 Decatur, AL 35609 Mr. Larry S. Bryant, Vice President Mr. Larry S. Mellen Nuclear Engineering & Technical Services Browns Ferry Unit 1 Project Engineer Tennessee Valley Authority Division of Reactor Projects, Branch 6 6A Lookout Place U.S. Nuclear Regulatory Commission 1101 Market Street 61 Forsyth Street, SW.

Chattanooga, TN 37402-2801 Suite 23T85 Atlanta, GA 30303-8931 Brian OGrady, Site Vice President Browns Ferry Nuclear Plant Ms. Beth A. Wetzel, Manager Tennessee Valley Authority Corporate Nuclear Licensing P.O. Box 2000 and Industry Affairs Decatur, AL 35609 Tennessee Valley Authority 4X Blue Ridge Mr. Preston D. Swafford, Senior Vice President 1101 Market Street Nuclear Support Chattanooga, TN 37402-2801 Tennessee Valley Authority 6A Lookout Place Mr. William D. Crouch, Manager 1101 Market Street Licensing and Industry Affairs Chattanooga, TN 37402-2801 Browns Ferry Nuclear Plant Tennessee Valley Authority General Counsel P.O. Box 2000 Tennessee Valley Authority Decatur, AL 35609 ET 11A 400 West Summit Hill Drive Senior Resident Inspector Knoxville, TN 37902 U.S. Nuclear Regulatory Commission Browns Ferry Nuclear Plant Mr. John C. Fornicola, General Manager 10833 Shaw Road Nuclear Assurance Athens, AL 35611-6970 Tennessee Valley Authority 6A Lookout Place State Health Officer 1101 Market Street Alabama Dept. of Public Health Chattanooga, TN 37402-2801 RSA Tower - Administration Suite 1552 Mr. Bruce Aukland, Plant Manager P.O. Box 303017 Browns Ferry Nuclear Plant Montgomery, AL 36130-3017 Tennessee Valley Authority P.O. Box 2000 Chairman Decatur, AL 35609 Limestone County Commission 310 West Washington Street Mr. Masoud Bajestani, Vice President Athens, AL 35611 Browns Ferry Unit 1 Restart Browns Ferry Nuclear Plant Mr. Robert H. Bryan, Jr., General Manager Tennessee Valley Authority Licensing and Industry Affairs P.O. Box 2000 Tennessee Valley Authority Decatur, AL 35609 4X Blue Ridge 1101 Market Street Chattanooga, TN 37402-2801