ML061390018

Technical Specifications, Reduction in Reactor Coolant System Flow and Increase in Steam Generator Tube Plugging Limit

ML061390018
Person / Time
Site:	Saint Lucie
Issue date:	05/16/2006
From:	Plant Licensing Branch III-2
To:
Moroney B, NRR/DORL, 415-3974
Shared Package
ML061120005	List: ML061110372 ML061390018
References
TAC MC8757
Download: ML061390018 (5)

Text

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 CFR 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 AND 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.

Renewed License No. NPF-16 Amendment No. 145

- 3a -

B. Technical Specifications The Technical Specifications contained in Appendix A and B, as revised through Amendment No. 145 are hereby incorporated in the renewed license. FPL shall I operate the facility in accordance with the Technical Specifications.

Appendix B, the Environmental Protection Plan (Non-Radiological), contains environmental conditions of the renewed license. If significant detrimental effects or evidence of irreversible damage are detected by the monitoring programs required by Appendix B of this license, FPL will provide the Commission with an analysis of the problem and plan of action to be taken subject to Commission approval to eliminate or significantly reduce the detrimental effects or damage.

C. Updated Final Safety Analysis Report FPL's Updated Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21 (d), as revised on March 28, 2003, describes certain future activities to be completed before the period of extended operation. FPL shall complete these activities no later than April 6, 2023, and shall notify Renewed License No. NPF-16 Amendment No. 145

POWER DISTRIBUTION LIMITS DNB PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB-related parameters shall be maintained within the limits shown on Table 3.2-2:

a. Cold Leg Temperature

b. Pressurizer Pressure

c. Reactor Coolant System Total Flow Rate

d. AXIAL SHAPE INDEX APPLICABILITY: MODE 1.

ACTION:

With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to < 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.2.5.1 Each of the parameters of Table 3.2-2 shall be verified to be within their limits by instrument readout at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4.2.5.2 The Reactor Coolant System total flow rate shall be determined to be within its limit by measurement* at least once per 18 months.

• Not required to be performed until THER 1 AL POWER is > 80% of RATED THERMAL POWER.

ST. LU CIE -UNIT 2 3/4 2-14 Amendment No. 8g, 1 4 5

TABLE 3.2-2 DNB MARGIN LIMITS FOUR REACTOR COOLANT PUMPS PARAMETER OPERATING Cold Leg Temperature (Narrow Range) Within the limits specified in the COLR Table 3.2-2 Pressurizer Pressure* Within the limits specified in the COLR Table 3.2-2 Reactor Coolant Flow Rate** > 335,000 gpm and > the limit specified I in the COLR Table 3.2-2 AXIAL SHAPE INDEX COLR Figure 3.2-4

• Limit not applicable during either a THERMAL POWER ramp increase in excess of 5% of RATED THERMAL POWER or a THERMAL POWER step increase of greater than 10% of RATED THERMAL POWER.

• • Commencing with the startup for Cycle 16 and until the Combustion Engineering Model 3410 Steam Generators are replaced, Reactor Coolant Flow Rate will also be limited in accordance with Renewed Operating License Paragraph 3.A.

ST. LUCIE - UNIT 2 3/4 2-15 Amendment No.. 92.844. 438.

145

X REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

10. Tube Renair refers to sleeving with Westinghouse Leak Limiting Alloy 800 sleeves as described in WCAP-15918-P Revision 2 (with range of conditions as revised in Appendix A of WCAP-16489-NP, Revision 0),

which are used to maintain a tube in service. Leak Limiting Alloy 800 Sleeves are applicable only to the original steam generators. The pressure boundary portion of the original tube wall in the sleeve/tube assembly (i.e., the sleeve-to-tube joint) shall be inspected prior to installation of each sleeve.

b. The steam generator shall be determined OPERABLE after completing the corresponding actions (plug or repair all tubes exceeding the Plugging or Repair Limit and all tubes containing through-wall cracks) required by Table 4.4-2.

4.4.5.5 Reports

a. Within 15 days following the completion of each inservice inspection of steam generator tubes, the number of tubes plugged or repaired in each steam generator shall be reported to the Commission in a Special Report pursuant to Specification 6.9.2.

b. The complete results of the steam generator tube inservice inspection shall be submitted to the Commission in a Special Report pursuant to Specification 6.9.2 within 12 months following completion of the inspection. This Special Report shall include:

1. Number and extent of tubes and sleeves inspected.

2. Location and percent of wall-thickness penetration for each indication of an imperfection.

3. Identification of tubes plugged or repaired.

c. Following each inspection and within 120 days after the reactor coolant system reenters MODE 4, the following information concerning indications found in the tubesheet region (including the expansion transition) shall be reported to the Commission in a special report pursuant to Specification 6.9.2. This Special Report shall include:

1. Number of total indications, location of each indication, orientation of each indication, severity of each indication, and whether the indications initiated from the inside or outside diameter.

2. The cumulative number of indications detected in the tubesheet region as a function of elevation within the tubesheet.

3. Projected end-of-cycle accident inducted leakage from tubesheet indications. This leakage shall be combined with the postulated end-of-cycle accident induced leakage from all other sources. If the preliminary estimated-totalprojected end-of-cycle-accident-induced leakage-from all sources exceeds the leakage limit, the NRC staff shall be notified prior to Unit restart.

ST. LUCIE - UNIT 2 3/4 4-15 Amendment No. 43, 44Q, 444, 145