ML083231102

Correction to Amendment No. 152 Regarding Alternative Source Term

ML083231102
Person / Time
Site:	Saint Lucie
Issue date:	11/17/2008
From:	Mozafari B Plant Licensing Branch II
To:	Stall J Duane Arnold
Mozafari B, NRR/ADRO/DORL, 415-2020
References
TAC MD6202
Download: ML083231102 (7)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 November 17, 2008 Mr. J. A. Stall Senior Vice President, Nuclear and Chief Nuclear Officer Florida Power and Light Company P.O. Box 14000 Juno Beach, Florida 33408-0420 SUB..IECT: ST. LUCIE PLANT, UNIT 2 - CORRECTION TO AMENDMENT NO.152 REGARDING ALTERNATIVE SOURCE TERM (TAC NO. MD6202)

Dear Mr. Stall:

On September 29, 2008, the Nuclear Regulatory Commission issued Amendment No.152 to Renewed Facility Operating License No. NPF-16 for the St. Lucie Plant, Unit No.2. This amendment was in response to your application dated July 16, 2007, as supplemented by letters dated February 14, March 18, April 14, June 2, July 11, and August 13, 2008.

The amendment modified the facility's operating licensing bases to adopt the alternative source term as allowed in 10 CFR 50.67 and described in Regulatory Guide 1.183. The licensee revised the plant licensing basis through reanalysis of the following radiological consequences of the Updated Final Safety Analysis Report Chapter 15 accidents: Loss-of-Coolant Accident, Fuel-Handling Accident, Main Steam Line Break, Steam Generator Tube Rupture, Reactor Coolant Pump Shaft Seizure, Control Element Assembly Ejection, Letdown Line Break, and Feedwater Line Break.

After issuance, it became apparent that some pages were incorrect. License page 3a was incomplete and should have been number 3; page 6-15e was missing number 147 (a previous amendment) listed at the bottom; and page 6-15f was missing number 147. That page contained rollover text from the previous page and is being re-numbered 6-15e(1). Therefore, the list of amended pages has also been changed to reflect the corrected page numbers.

The corrected pages are enclosed. We regret any inconvenience this may have caused. If you have any questions regarding this matter, please call me at (301) 415-2020.

Sincerely, IRA Eva Brown forI Brenda L. Mozafari, Senior Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-389

Enclosure:

Corrected Pages cc: See next page

Florida Power & Light Company ST. LUCIE PLANT cc:

Mr. J. Kammel Radiological Emergency Planning Administrator Division of Emergency Preparedness Department of Community Affairs 6000 Southeast Tower Drive Stuart, Florida 34997 Additional Distribution via ListServ

November 17, 2008 Mr. J. A. Stall Senior Vice President, Nuclear and Chief Nuclear Officer Florida Power and Light Company P.O. Box 14000 Juno Beach, Florida 33408-0420 SUB,JECT: ST. LUCIE PLANT, UI\IIT 2 - CORRECTION TO AMENDMEI\JT NO.152 REGARDING ALTERNATIVE SOURCE TERM (TAC NO. MD6202)

Dear Mr. Stall:

On September 29, 2008, the Nuclear Regulatory Commission issued Amendment No.152 to Renewed Facility Operating License No. I\J PF-16 for the St. Lucie Plant, Unit l\Jo. 2. This amendment was in response to your application dated July 16, 2007, as supplemented by letters dated February 14, March 18, April 14, June 2, July 11, and August 13, 2008.

The amendment modified the facility's operating licensing bases to adopt the alternative source term as allowed in 10 CFR 50.67 and described in Regulatory Guide 1.183. The licensee revised the plant licensing basis through reanalysis of the following radiological consequences of the Updated Final Safety Analysis Report Chapter 15 accidents: Loss-of-Coolant Accident, Fuel-Handling Accident, Main Steam Line Break, Steam Generator Tube Rupture, Reactor Coolant Pump Shaft Seizure, Control Element Assembly Ejection, Letdown Line Break, and Feedwater Line Break.

After issuance, it became apparent that some pages were incorrect. License page 3a was incomplete and should have been number 3; page 6-15e was missing number 147 (a previous amendment) listed at the bottom; and page 6-15f was missing number 147. That page contained rollover text from the previous page and is being re-numbered 6-15e(1). Therefore, the list of amended pages has also been changed to reflect the corrected page numbers.

The corrected pages are enclosed. We regret any inconvenience this may have caused. If you have any questions regarding this matter, please call me at (301) 415-2020.

Sincerely, IRA Eva Brown forI Brenda L. Mozafari, Senior Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket l\Jo. 50-389

Enclosure:

Corrected Pages cc: See next page DISTRIBUTION:

PUBLIC LPL2-2 R/F RidsNrrPMBMozafari RidsNrrDorlLpl2-2 RidsOgcRp RidsAcrsAcnwMailCenter RidsRgn2MailCenter GHill (2) RidsNrrLABClayton ADAMS Accession Number" ML083231102 NRR-106 OFFICE LPL2-2/PM LPL2-2/LA LPL2-2/BC NAME EBrown for BMozafari BClayton TBoyce DATE 11/17/08 11/14/08 11/17/08 OFFICIAL RECORD COpy

ATTACHMENT TO LICENSE AMENDMENT NO. 152 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-16 DOCKET NO. 50-389 Replace Page 3 of Renewed Operating License NPF-16 with the attached Page 3.

Replace the following pages of the Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.

Remove Pages Insert Pages 1-3 1-3 3/4 3-24 3/4 3-24 3/4 3-25 3/4 3-25 3/4 6-27 3/4 6-27 3/4 6-28 3/4 6-28 3/4 6-29 3/4 6-29 3/4 7-20 3/4 7-20 6-15c 6-15c 6-15d 6-15d 6-15e 6-15e 6-15e(1 )

-3 neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required.

D. Pursuant to the Act and 10 CFR Parts 30, 40, and 70, FPL to receive, possess, and use in amounts as required any byproduct, source, or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and E. Pursuant to the Act and 10 CFR Parts 30, 40, and 70, FPL to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

3. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission's regulations: 10 CFR Part 20, Section 30.34 of 10 FR Part 30, Section 40.41 of 10 CFR Part 40, Section 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified below:

A. Maximum Power Level FPL is authorized to operate the facility at steady state reactor core power levels not in excess of 2700 megawatts (thermal).

Commencing with the startup for Cycle 16 and until the Combustion Engineering Model 3410 Steam Generators are replaced, the maximum reactor core power shall not exceed 89 percent of 2700 megawatts (thermal) if:

a. The Reactor Coolant System Flow Rate is less than 335,000 gpm but greater than or equal to 300,000 gpm, or

b. The Reactor Coolant System Flow Rate is greater than or equal to 300,000 gpm AI\JD the percentage of steam generator tubes plugged is greater than 30 percent (2520 tubes/SG) but less than or equal to 42 percent (3532 tubes/SG).

This restriction in maximum reactor core power is based on analyses provided by FPL in submittals dated October 21, 2005 and February 28, 2006, and approved by the NRC in Amendment No. 145, which limits the percent of steam generator tubes plugged to a maximum of 42 percent (3532 tubes) in either steam generator and limits the plugging asymmetry between steam generators to a maximum of 600 tubes.

S. Technical Specifications The Technical Specifications contained in Appendix A and S, as revised through Amendment No. 152 are hereby incorporated in the renewed license. FPL shall operate the facility in accordance with the Technical Specifications.

Renewed License No. NPF-16 Amendment No. 152

ADMINISTRATIVE CONTROLS (continued)

k. Ventilation Filter Testing Program (VFTP) (continued)

3. Demonstrate for each of the ESF systems that a laboratory test of a sample of the charcoal ad sorber, when obtained as described in Regulatory Guide 1.52, Revision 3, shows the methyl iodide penetration less than the value specified below when tested in accordance with ASTM 03803-1989 at a temperature of 30°C and the relative humidity specified below.

ESF Ventilation System Penetration RH Control Room Emergency Air Cleanup ,::: 0.175% 95%

Shield Building Ventilation System ,:::2.5% 95%

ECCS Area Ventilation System ,:::2.5% 95%

4. For the Control Room Emergency Air Cleanup System and the ECCS Area Ventilation System, demonstrate that the pressure drop across the combined HEPA filters and charcoal adsorbers is less than the value specified below when tested at the system f10wrate specified below. For the Shield Building Ventilation System, demonstrate that the pressure drop across the combined demisters, electric heaters, HEPA filters and charcoal adsorbers is less than the value specified below when tested at the system flowrate specified below.

ESF Ventilation System Delta P Flowrate Control Room Emergency Air Cleanup < 7.4" W.G. 2000 2: 200 cfm Shield Building Ventilation System < 8.5" W.G. 6000 2: 600 cfm ECCS Area Ventilation System < 4.35"W.G. 30,0002: 3000 cfm

5. At least once per 18 months, demonstrate that the heaters for each of the ESF systems dissipate the value specified below when tested in accordance with ASME N510-1989.

ESF Ventilation System Wattage Shield Building Ventilation System Main Heaters 30 + 3 kW Auxiliary Heaters 1.52: 0.25 kW The provisions of SR 4.0.2 and SR 4.0.3 are applicable to the VFTP test frequencies.

I. Steam Generator (SG) Program

1. A SG Program shall be established and implemented for the replacement SGs to ensure that SG tube integrity is maintained. In addition, the SG Program shall include the follOWing provisions:

a. Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged to confirm that the performance criteria are being met.

ST. LUCIE - UNIT 2 6-15e Amendment No. 4-d-9, --t4-T; 152

ADMINISTRATIVE CONTROLS (continued)

I. Steam Generator (SG) Program (continued)

1. (continued)

b. Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational leakage.

1. Structural integrity performance criterion: All in-service SG tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cooldown and all anticipated transients included in the design specification) and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

2. Accident induced leakage performance criterion: The primary-to-secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. Leakage is not to exceed 0.5 gallons per minute total through all SGs and 0.25 gallons per minute through anyone SG.

3. The operational leakage performance criterion is specified in LCO 3.4.6.2.c, "Reactor Coolant System Operational Leakage."

ST. LUCIE - UNIT 2 6-15e(1 ) Amendment No. +J9, -t41-, 152