ML070660489

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Technical Specifications Pages Re Five-Percent Uprate (TAC No. MD3048) (TS-418)
ML070660489
Person / Time
Site: Browns Ferry Tennessee Valley Authority icon.png
Issue date: 03/06/2007
From:
NRC/NRR/ADRO/DORL/LPLII-2
To:
Chernoff M, NRR/DORL, 415-4041
Shared Package
ML070680307 List:
References
TAC MD3048
Download: ML070660489 (14)
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(3) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use at any time any byproduct, source, and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, 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 equipment and instrument calibration or associated with radioactive apparatus or components; (5) Pursuant to the Act and 10 CFR Parts 30 and 70, to possess but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I:

Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; 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 or incorporated below:

(1) Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 3458 megawatts thermal.

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 269, are hereby incorporated in the renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

For Surveillance Requirements (SRs) that are new in Amendment 234 to Facility Operating License DPR-33, the first performance is due at the end of the first surveillance interval that begins at implementation of the Amendment 234. For SRs that existed prior to Amendment 234, including SRs with modified acceptance criteria and SRs whose frequency of performance is being extended, the first performance is due at the end of the first surveillance interval that begins on the date the surveillance was last performed prior to implementation of Amendment 234.

BFN-UNIT 1 Renewed License No. DPR-33 Amendment 269

(13) Browns Ferry Nuclear Plant shall implement and maintain in effect all provisions of the approved Fire Protection Program as described in the Final Safety Analysis Report for BFN as approved in the safety evaluations dated December 8, 1988; March 6, 1991; March 31, 1993; November 2, 1995; and Supplement dated November 3, 1989; subject to the following provision:

The licensee may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

(14) The licensee shall maintain the Augmented Quality Program for the Standby Liquid Control System to provide quality control elements to ensure component reliability for the required alternative source term function defined in the Updated Final Safety Analyses Report (UFSAR).

(15) The licensee is required to confirm that the conclusions made in TVA's letter dated September 17, 2004, for the turbine building remain acceptable using seismic demand accelerations based on dynamic seismic analysis prior to the restart of Unit 1.

D. The UFSAR supplement, as revised, submitted pursuant to 10 CFR 54.21(d), shall be included in the next scheduled update to the UFSAR required by 10 CFR 50.71(e)(4) following the issuance of this renewed operating license. Until that update is complete, TVA may make changes to the programs and activities described in the supplement without prior Commission approval, provided that TVA evaluates such changes pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.

E. The UFSAR supplement, as revised, describes certain future activities to be completed prior to the period of extended operation. TVA shall complete these activities no later than December 20, 2013, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.

F. All capsules in the reactor vessel that are removed and tested must meet the test procedures and reporting requirements of the most recent NRC-approved version of the Boiling Water Reactor Vessels and Internals Project (BWRVIP) Integrated Surveillance Program (ISP) appropriate for the configuration of the specimens in the capsule. Any changes to the BWRVIP ISP capsule withdrawal schedule, including spare capsules, must be approved by the NRC prior to implementation.

All capsules placed in storage must be maintained for future insertion. Any changes to storage requirements must be approved by the NRC, as required by 10 CFR Part 50, Appendix H.

BFN-UNIT 1 Renewed License No. DPR-33 Amendment 269

G. (1) During the power uprate power ascension test program and prior to exceeding 30 days of plant operation above a nominal 3293 megawatts thermal power level (100-percent OLTP) or within 30 days of satisfactory completion of steam dryer monitoring and testing that is necessary for achieving 105-percent OLTP (whichever is longer), with plant conditions stabilized at 105-percent OLTP, TVA shall trip a condensate booster pump, a condensate pump, and a main feedwater pump on an individual basis (i.e.,

one at a time). Following each pump trip, TVA shall confirm that plant response to the transient is as expected in accordance with previously established acceptance criteria. Evaluation of the test results for each test shall be completed and all discrepancies resolved in accordance with corrective action program requirements and the provisions of the power ascension test program.

(2) During the power uprate power ascension test program and prior to exceeding 30 days of plant operation above a nominal 3293 megawatts thermal power level (100-percent OLTP) or within 30 days of satisfactory completion of steam dryer monitoring and testing that is necessary in order to achieve 105-percent OLTP (whichever is longer), with plant conditions stabilized at 105-percent OLTP, TVA shall perform a MS isolation valve closure test and a turbine generator load reject test. Following each test, TVA shall confirm that plant response to the transient is as expected in accordance with previously established acceptance criteria. The evaluation of the test results for each test shall be completed, and all discrepancies resolved, prior to resumption of power operation.

H. The licensee must complete the thirteen (13) Unit 1 restart commitments that are discussed in Appendix F of the license renewal application, dated December 31, 2003, as supplemented by letters dated January 31, 2005, March 2, and April 21, 2006. Completion of these activities must be met prior to power operation of Unit 1.

This renewed license is effective as of the date of issuance and shall expire midnight on December 20, 2033.

FOR THE NUCLEAR REGULATORY COMMISSION Original Signed By J. E. Dyer J. E. Dyer, Director Office of Nuclear Reactor Regulation Attachments:

1. Unit 1 - Technical Specifications - Appendices A and B Date of Issuance: May 4, 2006 BFN-UNIT 1 Renewed License No. DPR-33 Amendment 269

Definitions 1.1 1.1 Definitions (continued)

OPERABLE - OPERABILITY A system, subsystem, division, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, division, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation. These tests are:

a. Described in Section 13.10, Refueling Test Program; of the FSAR;
b. Authorized under the provisions of 10 CFR 50.59; or
c. Otherwise approved by the Nuclear Regulatory Commission.

RATED THERMAL POWER RTP shall be a total reactor core heat transfer rate to (RTP) the reactor coolant of 3458 MWt.

(continued)

BFN-UNIT 1 1ý1-6 Amendment No. 2-4-,269

SLC System 3.1.7 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY Verify the concentration and temperature of Once within boron in solution are within the limits of 8 hours0.333 days <br />0.0476 weeks <br />0.011 months <br /> after Figure 3.1.7-1. discovery that SPB concentration is

> 9.2% by weight AND 12 hours0.5 days <br />0.0714 weeks <br />0.0164 months <br /> thereafter SR 3.1.7.5 Verify the minimum quantity of Boron-10 in the 31 days SLC solution tank and available for injection is

_ 203 pounds.

SR 3.1.7.6 Verify the SLC conditions satisfy the following 31 days equation:

AND (C )( Q )( E )

(13 wt. %)(86 gpm)(1 9.8 atom%) > 1 Once within 24 hours1 days <br />0.143 weeks <br />0.0329 months <br /> after where, water or boron is added to the solution C = sodium pentaborate solution concentration (weight percent)

Q = pump flow rate (gpm)

E = Boron-10 enrichment (atom percent Boron-10)

SR 3.1.7.7 Verify each pump develops a flow rate >_39 24 months gpm at a discharge pressure > 1325 psig.

(continued)

BFN-UNIT I 3.1-25 Amendment No. 2-34,25, ,2---, 269

RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 1 of 3)

Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED FUNCTION OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER TRIP REQUIRED REQUIREMENTS VALUE CONDITIONS SYSTEM ACTION D.1

1. Intermediate Range Monitors
a. Neutron Flux - High 2 3 G SR 3.3.1.1.1 <-120/125 SR 3.3.1.1.3 divisions of full SR 3.3.1.1.5 scale SR 3.3.1.1.6 SR 3.3.1.1.9 SR 3.3.1.1.14 5 (a) 3 H SR 3.3.1.1.1 < 120/125 SR 3.3.1.1.4 divisions of full SR 3.3.1.1.9 scale SR 3.3.1.1.14
b. Inop 2 3 G SR 3.3.1.1.3 NA SR 3.3.1.1.14 5 (a) 3 H SR 3.3.1.1.4 NA SR 3.3.1.1.14
2. Average Power Range Monitors 2 3 (b) G SR 3.3.1.1.1
a. Neutron Flux - High, SR 3.3.1.1.6 Setdown SR 3.3.1.1.7 SR 3.3.1.1.13 SR 3.3.1.1.16
b. Flow Biased Simulated 3 (b) F SR 3.3.1.1.1

SR 3.3.1.1.13 RTP(c)

SR 3.3.1.1.16

c. Neutron Flux - High 1 3 (b) F SR 3.3.1:1.1 *120% RTP SR 3.3.1.1.2 SR 3.3.1.1.7 SR 3.3.1.1.13 SR 3.3.1.1.16 (continued)

(a) With any control rod withdrawn from a core cell containing one or more fuel assemblies.

(b) Each APRM channel provides inputs to both trip systems.

(c) [0.66 W + 66% - 0.66 A W] RTP when reset for single loop operation per LCO 3.4.1, "Recirculation Loops Operating."

BFN-UNIT 1 3.3-6 Amendment No. 23,6, 262,269

RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 2 of 3)

Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED FUNCTION OTHER CHANNELS FROM SURVEILLANCE ALLOWABLE SPECIFIED PER TRIP REQUIRED REQUIREMENTS VALUE CONDITIONS SYSTEM ACTION D.1

2. Average Power Range Monitors (continued)
d. Inop 1,2 3 (b) G SR 3.3.1.1.16 NA
e. 2-Out-Of-4 Voter 1,2 2 G SR 3.3.1.1.1 NA SR 3.3.1.1.14 SR 3.3.1.1.16
f. OPRM Upscale 3 (b) I SR 3.3.1.1.1 NA SR 3.3.1.1.7 SR 3.3.1.1.13 SR 3.3.1.1.16 SR 3.3.1.1.17
3. Reactor Vessel Steam Dome 1,2 2 G SR 3.3.1.1.1 < 1090 psig Pressure - High(d) SR 3.3.1.1.8 SR 3.3.1.1.10 SR 3.3.1.1.14
4. Reactor Vessel Water Level - 1,2 2 G SR 3.3.1.1.1 _>528 inches Low, Level 3 (d) SR 3.3.1.1.8 above vessel SR 3.3.1.1.13 zero SR 3.3.1.1.14
5. Main Steam Isolation Valve - 1 8 F SR 3.3.1.1.8 5 10% closed Closure SR 3.3.1.1.13 SR 3.3.1.1.14
6. Drywell Pressure - High 1,2 2 G SR 3.3.1.1.8 :5 2.5 psig SR 3.3.1.1.13 SR 3.3.1.1.14
7. Scram Discharge Volume Water Level - High
a. Resistance Temperature 1,2 2 G SR 3.3.1.1.8 _<50 gallons Detector SR 3.3.1.1.13 SR 3.3.1.1.14 5 (a) 2 H SR 3.3.1.1.8 _*50 gallons SR 3.3.1.1.13 SR 3.3.1.1.14 (continued)

(a) With any control rod withdrawn from a core cell containing one or more fuel assemblies.

(b) Each APRM channel provides inputs to both trip systems.

(d) During instrument calibrations, ifthe As Found channel setpoint is conservative with respect to the Allowable Value but outside its acceptable As Found band as defined by its associated Surveillance Requirement procedure, then there shall be an initial determination to ensure confidence that the channel can perform as required before returning the channel to service in accordance with the Surveillance. If the As Found instrument channel setpoint is not conservative with respect to the Allowable Value, the channel shall be declared inoperable.

Prior to returning a channel to service, the instrument channel setpoint shall be calibrated to a value that is within the acceptable As Left tolerance of the setpoint; otherwise, the channel shall be declared inoperable.

The nominal Trip Setpoint shall be specified on design output documentation which is incorporated by reference in the Updated Final Safety Analysis Report. The methodology used to determine the nominal Trip Setpoint, the predefined As Found Tolerance, and the As Lefh Tolerance band, and a listing of the setpoint design output documentation shall be specified in Chapter 7 of the Updated Final Safety Analysis Report.

BFN-UNIT 1 3.3-7 Amendment No. 269 234, 262, 259, 257,258, 2656

ATWS-RPT Instrumentation 3.3.4.2 SURVEILLANCE REQUIREMENTS

NOTE- ----------------------------

When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours0.25 days <br />0.0357 weeks <br />0.00822 months <br /> provided the associated Function maintains ATWS-RPT trip capability.

SURVEILLANCE FREQUENCY SR 3.3.4.2.1 Perform CHANNEL CHECK of the Reactor 24 hours1 days <br />0.143 weeks <br />0.0329 months <br /> Vessel Water Level - Low Low, Level 2 Function.

SR 3.3.4.2.2 Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.4.2.3 Perform CHANNEL CALIBRATION. The 24 months Allowable Values shall be:

a. Reactor Vessel Water Level - Low Low, Level 2: Ž_471.52 inches above vessel zero; and
b. Reactor Steam Dome Pressure - High:
  • 1175 psig.

I SR 3.3.4.2.4 Perform LOGIC SYSTEM FUNCTIONAL 24 months TEST including breaker actuation.

BFN-UNIT 1 3.3-34 Amendment No. 234-,263,269

S/RVs 3.4.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1 Verify the safety function lift settings of the In accordance required 12 S/RVs are within +/- 3% of the with the Inservice setpoint as follows: Testing Program Number of Setpoint S/RVs {psiq) 4 1135 4 1145 5 1155 Following testing, lift settings shall be within

+/-+1%.

SR 3.4.3.2 --------------------- NOTE ------------

Not required to be performed until 12 hours0.5 days <br />0.0714 weeks <br />0.0164 months <br /> after reactor steam pressure and flow are adequate to perform the test.

Verify each required S/RV opens when 24 months manually actuated.

BFN-UNIT 1 3.4-8 Amendment No. 234, 263, 269

Reactor Steam Dome Pressure 3.4.10 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.10 Reactor Steam Dome Pressure LCO 3.4.10 The reactor steam dome pressure shall be _<1050 psig.

APPLICABILITY: MODES 1 and 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Reactor steam dome A.1 Restore reactor steam 15 minutes pressure not within limit, dome pressure to within limit.

B. Required Action and B.1 Be in MODE 3. 12 hours0.5 days <br />0.0714 weeks <br />0.0164 months <br /> associated Completion Time not met.

BFN-UNIT 1 3.4-30 Amendment No. 2-34,269

Reactor Steam Dome Pressure 3.4.10 SURVEILLANCE REQUIREMENTS_

SURVEILLANCE FREQUENCY SR 3.4.10.1 Verify reactor steam dome pressure is 12 hours0.5 days <br />0.0714 weeks <br />0.0164 months <br />

< 1050 psig.

BFN-UNIT 1 3.4-31 Amendment No. 2-34, 269

ECCS - Operating 3.5.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY

+

SR 3.5.1.7 -----------------

NOTE -------------------------

Not required to be performed until 12 hours0.5 days <br />0.0714 weeks <br />0.0164 months <br /> after reactor steam pressure and flow are adequate to perform the test.

Verify, with reactor pressure < 1040 and 92 days

> 950 psig, the HPCI pump can develop a flow rate > 5000 gpm against a system head corresponding to reactor pressure.

SR 3.5.1.8 ------------------

NOTE -------------

Not required to be performed until 12 hours0.5 days <br />0.0714 weeks <br />0.0164 months <br /> after reactor steam pressure and flow are adequate to perform the test.

Verify, with reactor pressure < 165 psig, the 24 months HPCI pump can develop a flow rate > 5000 gpm against a system head corresponding to reactor pressure.

SR 3.5.1.9 -------------------- NOTE-------------

Vessel injection/spray may be excluded.

Verify each ECCS injection/spray subsystem 24 months actuates on an actual or simulated automatic initiation signal.

(continued)

BFN-UNIT 1 3.5-6 Amendment No. 234-,263, 269

RCIC System 3.5.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.3.1 Verify the RCIC System piping is filled with 31 days water from the pump discharge valve to the injection valve.

SR 3.5.3.2 Verify each RCIC System manual, power 31 days operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.

SR 3.5.3.3 -------------------- NOTE -------------

Not required to be performed until 12 hours0.5 days <br />0.0714 weeks <br />0.0164 months <br /> after reactor steam pressure and flow are adequate to perform the test.

Verify, with reactor pressure _<1040 psig and 92 days

> 950 psig, the RCIC pump can develop a flow rate > 600 gpm against a system head corresponding to reactor pressure.

SR 3.5.3.4 -------------------- NOTE -------------

Not required to be performed until 12 hours0.5 days <br />0.0714 weeks <br />0.0164 months <br /> after reactor steam pressure and flow are adequate to perform the test.

Verify, with reactor pressure _<165 psig, the 24 months RCIC pump can develop a flow rate >_600 gpm against a system head corresponding to reactor pressure.

(continued)

BFN-UNIT 1 3.5-13 Amendment No. 234, 263, 269

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 Safety Function Determination Program (SFDP) (continued)

A loss of safety function exists when, assuming no concurrent single failure, a safety function assumed in the accident analysis cannot be performed. For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and:

a. A required system redundant to system(s) supported by the inoperable support system is also inoperable; or
b. A required system redundant to system(s) in turn supported by the inoperable supported system is also inoperable; or
c. A required system redundant to support system(s) for the supported systems (a) and (b) above is also inoperable.

The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.

5.5.12 Primary Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995.

The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 48.5 psig. The maximum allowable primary containment leakage rate, La, shall be 2% of primary containment air weight per day at Pa.

(continued)

BFN-UNIT 1 5.0-20 Amendment No. 244-,269

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