Text

Amends 137 & 129 to Licenses NPF-02 & NPF-08,respectively, Change Max Reactor Core Power Level for Facility Operation from 2652 Mwt to 2775 Mwt for Plants.Amends Also Approve Changes to TS to Implement Uprated Power Operation
	ML20217P047
Person / Time
Site:	Farley
Issue date:	04/29/1998
From:	Collins S; NRC (Affiliation Not Assigned)
To:	;
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ML20217P053	List: ML20217P047; ML20217P057;
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t UNITED STATES g

j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 3De85 40M o

SOUTHERN NUCLEAR OPERATING COMPANY. INC.

ALABAMA POWER COMPANY DOCKET NO. 50-348 JOSEPH M. FARLEY NUCI FAR PLANT. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.137 License No. NPF-2 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Southem Nuclear Operating Company, Inc.

(Southem Nuclear), dated February 14,1997, as supplemented by letters dated June 20, August 5, September 22, November 19, December 9, December 17, and December 31,1997, January 23, February 12, February 26, March 3, March 6, March 16, April 3, April 13, and two letters on April 17,1998, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been eatisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendment; and paragraph 2.C.(2) of Facility Operating License No. NPF-2 is hereby amended to read as follows:

i 9905060048 990429 I

PDR ADOCK 05000348 P

PDR

e s.

(2)

Technical Speedications I

The Technical Specifications contained in Appendices A and B, as revised through Amendment No.137, are hereby incorporated in the license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications.

' n addition, the license is amended to add the following paragraph to 2.C.(3)(h) to Facility i

Operating License No. NPF-2:

(h).

The Additional Conditions. contained in Appendix C, as revised through Amendment No.137, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the additional conditions.

3.

This license amendment is effective as of its date of issuance and shall be implemented prior to entering Mode 4 for Cycle 16 (fall 1998).

FOR THE NUCLEAR REGULATORY COMMISSION ue J.

irector Office of Nuclear Reactor Regulation Attachments:

1. Changes to Operating License

2. Changes to the Technical Specifications Date ofissuance:

April 29,1998 I

6 i

ATTACHMENTTO LICENSE AMENDMENT NO.137 TO FACILITY OPERATING LICENSE NO. NPF-2 DOCKET NO. 50-348 Replace the following pages of Operating License with the enclosed pages. The revised areas are indicated by marginallines.

Remove jnamt 4

4 6

6 12 12 Appendix C Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised areas are indicated by marginal lines.

Remove Insert 1-6 1-6 j

2-5 2-5 2-6 2-6 2-7 2-7 3/4 2-15 3/4 2-15 3/4 3-25 3/4 3-25 3/4 3-27 3/4 3-27 3/4 3-28 3/4 3-28 3/4 5-5 3/4 5-5 3/4 7-9 3/4 7-9 B 3/4 2-5 B 3/4 2-5 B 3/4 6-2 B 3/4 6-2 B 3/4 7-1 B 3/4 7-1 5-6 5-6 5-8 5-8 6-19a 6-19a 6-24 6-24

s (5)

Southern Nuclear, 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 fora, for sample analysis or instrument calibration or associated with radioactive apparatus or components;.and (6)

Southern Nuclear, pursuant to th-Act and 10 CFR Parts 30 and 70, to possess, bur

-+, separate, such byproduct and special nuclear mar ials as may be produced by the operation of the a:deility.

C.

This license shall'be deemed to~ cont'ain 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; 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 or incorporated below:

(1)

Marinum Power Level Southern Nuclear is authorized to operate the facility at steady state reactor core power levels not in excess of 2775 megawatts (thermal).

Prior l

to attaining the power level, Alabama Power company shall complete the preoperational tests, startup tests and other items identified in Attachment 2 to this license jn the sequence specified.

Attachment 2 is an integral part of this,licanse.

(2)

Technical Snecifications The Technical Specifications contained in Ap>endices A and B, as revised through Asendment No.137 are -here>y incorporated in the l license.

Southern Nuclear Operating Cospany. Inc., shall operate the facility in accordance with the Technical Specifications.

J Farley - Unit 1 kendment No. (6,137

e r 2.C.3.

2.

Identification of the procedures used to quantify (con't.)

parameters that are critical to control points; 3.

. identification of process sampling points; _

4.

A procedure for the recording and managemant of data; 5.

Procedures defining corrective actions for off control point chemistry conditions; and 6.

A procedure identifying the authority responsible for the interpretation of the data and the sequence and timing of administrative events required to initiate corrective action.

(h)

The Additional Conditions contained in Appendix C, as revised through Amendment No.137

, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the additional conditions.

(4)

Fire Protection Southem Nuclear shall implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report for the facility and as approved in the Fire Protection Safety Evaluation Reports dated February 12,1979, August 24,1983, December 30,1983, November 19,1985, September 10,1986, and December 29,1986. Southern Nuclear 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.

i i

Amendment No. 8, AA, AM 9tX 137

)

4

e H.

This license is effective as of the date of issuance and shall expire at midnight, June 25,2017.

FOR THE NUCLEAR REGULATORY COMMISSION i

ORIGINAL SIGNED BY:

R. C. DeYoung #or/

Roger S. Boyd, Director

'~

Division of Project Management Office of Nuclear Reactor Regulation Attachments:

1. Appendices A & B - Technical-Specifications

2. Preoperational Tests, Startup Tests and Other items Which Must Be Completed Prior to Proceeding to Succeeding Operational Modes

3. Appendix C - Additional Conditions l

Date of issuance: June 25,1977 Farley - Unit 1 Amendment No. S'1', Wg,137

i APPENDIX C ADDITIONAL CONDmONS OPERATING LICENSE NO. NPF-2 Southem Nuclear Operating Company, Inc. (SNC), shall comply with the following conditions on the schedules noted below:

Amendment Condition Completion Number Additional Condition Date 137 SNC shall complete classroom and simulator training Prior to Unit 2 entering for operations crews as described in SNC's letter Mode 2 from the dated September 22,1997, and evaluated in the staffs spring 1998 refueling Safety Evaluation dated April 29,1998.

outage 137 SNC shall complete final simulator modifications in Two years after restart accordance with ANSI /ANS 3.5-1985 and review from the Unit 1 fall results of the Cycle 16 startup testing to determine 1998 refueling outage any potential effects on operator training as described in SNC's letter dated September 22,1997, and evaluated in the staffs Safety Evaluation dated April 29,1998.

137 SNC shall provide a Steam Generator (SG) Tube Prior to the Unit 1 Rupture radiological consequences analysis that steam generator incorporates a flashing fraction, which is replacement outage appropriate for the Unit i design.

in spring 2000 i

Amendment No.137

e.

DEFINITIONS RATED THERMAL POWER 1.25' RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 2775 MWt.

l REACTOR TRIP SYSTEM RESPONSE TIME l.26 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until loss of stationary gripper coil voltage.

REPORTABLE EVENT l

1.27 A. REPORTABLE EVLNT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50.

SHUTDOWN MARGIN 1.28 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which j

the reactor is suberitical or would be suberitical from its present condition assuming all full length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest j

reactivity worth which is assumed to be fully withdrawn.

1 f'

SOLIDIFICATION l

1.29 This definition deleted. Refer to the Process Control Program.

SOURCF CHECK 1.30 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.

STAGGERED TEST BASIS 1.31 A STAGGEEED TEST BASIS shall consist of a.

A test schedule for n systems, subsystems, trains or other designated components obtained by dividing the specified test interval into n equal subintervals, b.

The testing of one system, subsystem, train or other designated component at the beginning of each subinterval.

FARLEY-UNIT 1 1-6 AMENDM3NT NO.137 i

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EMERGENCY CORE COOLING SYSTEMS l

SURVEILLhNCE REQUIREMENTS (Continued)

By verifying the correct position of each mechanical position stop e.

for the following ECCS throttle valves:

1.

Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following completion of each valve stroking operation or maintenance on the valve when the ECCS subsystems are required to be OPERABLE.

e 2.

At least once per 18 months.

valve Number 3

CVC-V-8991 A/B/C CVC-V-8989 A/B/C cvC-V-8996 A/B/C CVC-V-8994 A/B/C l

f.

At least once per 18 months, during shutdown, by:

)

l 1.

Verifying that each automatic valve in the flow path act.iates to its correct position on a safety injection test signal.

2.

Verifying that each of the following pumps start l

automatically upon receipt of a safety injection test l

signals a)

Centrifugal charging pump b)

Residual heat removal pump g.

By verifying that each of the following pumps develops the indicated alfferential pressure on recirculation flow when tested pursuant to Specification 4.0.5:

1.

Centrifugal charging pump 2 2323 paid l

2.

Residual heat removal pump 2 145 psid l

l l

l h.

Prior to entry into Mode 3 from Mode 4, verify that the mechanical l

stops on low head safety injection valve RHR-HV 603 A/B are intact.

s l

t FARLEY-UNIT 1 3/4 5-5 AMENDMENT NO.137 l

l

F-i l

\\

s s

PLANT SYSTEMS MAIN STEAM LINE ISOLATION VALES I

i LIMITING OONDITION FOR OPERATION l

(

3.7.1.5 Each main steam line isolation valve shall be OPERABLE.

i APPLICABILITY: MODES 1, 2 and 3.

I ACTION:

MODE 1 With one main steam line isolation valve inoperable, POWER l

OPERATION may continue provided the inoperable valve is restored to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; otherwise, reduce power to less than or equal to 5% of RATED THERMAL POWER within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

MODES 2 With one main steam line isolation valve inoperable, and 3 subsequent operation in MODES 2 or 3 may proceed provided the isolation valve is restored to OPERABLE status or closed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after entering MODE 2; otherwise, be in HOT STANDBY i

within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.1.5 Each main steam line-isolation valve shall be demonstrated OPERABLE by verifying full closure within 7 seconds when tested pursuant to l

Specification 4.0.5.

FARLEY-UNIT 1 3/4 7-9 AMENDMENT No.137

4 4

POWER DISTRIBUTION LIMITS BASES 3/4.2.4 OUADRANT POWER TILT RATIO The quadrant power tilt ratio limit assures that the radial power distribution satisfies the design values used in the power capability analysis. Radial power distribution measurements are made during startup testing and periodically during power operation.

The-limit of 1.02, at which corrective action is required, provides DNB and linear heat generation rate protection with x-y plane power tilts.

The two hour time allowance for operation with a tilt condition greater than 1.02 but less than.l.09 is provided to allow identification and correction of a dropped or misaligned control rod.

In the event such action does not correct the tilt, the margin for uncertainty on Fo is reinstated by reducing the maximum allowed power by 3 percent for each p5rcent of tilt,in excess,of 1.0.

For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the movable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO.

The incore detector monitoring is done with a full incore flux map or two sets of four symmetric thimbles.- The two sets of four symmetric thimbles is a unique set of eight detector locations. These locations are C-8, E-5, E-11, H-3, H-13, L-5, L-ll, and N-8.

3/4.2.5 DNB PARAMETERS The limits on the DNB related parameters assure that each of the parameters are maintained within the normal steady state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to meet the DNB design criterion throughout each analyzed transient. The indicated T value of 580.3*F is based on the average of two control board readings and 8n indication uncertainty of 2.9'F.

The indicated pressure value of 2209 psig is based on the average of two control board readings and an indication uncertainty of 24 psi. The indicated total RCS flow rate is based on the average of two elbow tap measurements from each loop as read on the plant computer and an uncertainty of 2.4% flow (0.1% flow is included for feedwater venturi fouling).

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months surveillance of T and pressurizer pressure through the control av board readings are sufficient t$ ensure that the parameters are restored within their limits following load changes and other expected transient operation.

The 18 month surveillance of the total RCS flow rate may be performed by one of two alternate methods. One method is a precision heat balance performed at the l

beginning of each fuel cycle. The other method is based on the Ap measurements from the cold leg elbow taps, which are correlated to past precision heat balance measurements. Correlation of the flow indication channels with selected precision loop flow measurements is documented in WCAP-14750 Use of the elbow l

tap Ap measurement method removes the requirement for performance of a precision RCS flow heat balance measurement for that cycle. The monthly surveillance of l

the total RCS flow rate is a reverification of the RCS flow requirement using process computer indications of loop elbow tap measurements that are correlated either to the precision RCS flow measurement or the elbow tap measurement at the

.beginning of the fuel cycle. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months RCS flow surveillance is a qualitative verification of significant flow degradation using the control board indicators fed by elbow tap measurements.

FARLEY-UNI" 1 B 3/4 2-5 AMENDMENT NO. 137

s, l

CONTAINMENT SYSTEMS BASES The maximum peak calculated containment internal pressure obtained from a.LOCA event is 43 psig and from a MSLB event is 53 psig. These containment analyses calculatior.s include an initial positive pressure of up to 3 psig.

The analyses results demonstrate that the maximum containment pressure will remain below the design limit of 54 psig.

3/4.6.1.5 AIR iEnrERATURE l

l The, limitations on containment average air. temperature ensure that the overall containment average air temperature does not exceed the initial temperature condition assumed in the accident analysis for a LOCA or steam line break accident.

3/4.6'.1.6 CONTAINMENT STRUCTURAL INTEGRITY l.

This limitation ensures that the structural integrity of the containment will.be maintained comparable to the original design standards for the life of the facility. Structural integrity is required to ensure that tt.a containment l-will withstand the maximum peak calculated internal pressure of 43 psig inLthe i

event of a LOCA, including an initial positive pressure of up to 3'psig.

In

. addition,' structural integrity is required to ensure that the containment will withstand the maximum peak calculated internal pressure of 53 psig in the event of a MSLB, including an initial positive pressure of up to 3 psig. As per Appendix J, Option B, the LOCA peak calculated containment internal pressure defines the Pa value for the Containment' Leakage Rate Testing Program required by Surveillance 4.6.1.2.

The measurement of the containment lift off force, visual examination of tendons, anchorages and exposed interior and exterior surfaces of the containment, and the containment leakage tests are j

sufficient to demonstrate this capability.

The surveillance requirements for demonstrating the containment's structural integrity are in compliance with the-recommendations of paragraph L

C.1.3 of Regulatory Guide 1.35 " Inservice Surveillance of Ungrouted Tendons in Prestressed Concrete Containment structure," January 1976.

)

l 3/4.6.1.7 CONTAINMENT VENTILATION SYSTEM 2ne'48-inch! containment purge supply and exhaust isolaticn valves are l

required to be closed in MODES above COLD SHUTDOWN since these valves have not been demonstrated capable of closing during a LOCA or steam line break t.

l accident. Maintaining these valves closed during plant operations ensures L

that excessivo quantities of radioactive materials will not be released via the containment purge system.

The use of the containment purge lines is restricted to the 8-inch vent supply and exhaust isolation valves to ensure that the site boundary dose guidelines of.10 CFR Part 100 would not be exceeded in the event of a loss-of-

-coolant accident during venting operations.

L FARLEY-UNIT 1 3 3/4 6-2 Amendment No. 137

3/4.7 PLANT SYSTEMS BASES

- 3/4.7.1 TURBINE CYMD 3/4.7.1.1~

SAFETY VALVES The OPERABILITY of the main steam line code safety valves ensures that i

the secondary system pressure will be limited to within 1104 (1194 psig) of its design pressure of 1085 psig during the most severe anticipated system operational transient.- The maximum relieving capacity is associated with a turbine trip from 100% RATED THERMAL POWER coincident with an assumed loss of

. condenser heat sink (i.e., no steam bypass to the condenser).

The specified valve lift settings and relieving capacities are in accordance with the requirements of Section III of the ASME Boiler and Pressure code, 1971 Edition. The total relieving capacity for all valves on all of the steam lines is at least 12,984,660 lbs/hr which is 105.8 percent c!

J the total secondary steam flow of 12,270,000 lbs/hr at 1004 RATED THERMAL POWER. A minimum of 2 OPERABLE safety valves per steam generator ensures that I

sufficient relieving capacity is'available for the allowable THERMAL POWER restriction in Table 3.7-2.

STARTUP and/or POWER OPERATION is. allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduction in secondary. system steam flow and THERMAL POWER required by the the reduced reactor _ trip settings of the Power Range Neutron Flux channels.

The reactor trip setpoint reductions are consistent with the assumptions used in the accident analysis.

.T FARLEY-UNIT 1 B 3/4 7-1 AMENDHENT NO. 137

DESICN FEATURES 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reacter shall contain 157 fuel assemblies. Each assembly shall consist of a matrix of zirconium alloy, zircaloy-4, or EIRLOS fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO2) as fuel material.

Limited substitutions of zirconium alloy, zircaloy-4, EIRLO*, or stainless steel' filler rods for fuel rods, in accordance with NRC-approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to

'those fuel designs that have been analyzed with applicable NRC-approved codes and methods, and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in non-limiting core regions.

J

\\

CONTROL ROD ASSEMBLIES

5. 3. 2.

The reactor core shall contain 48 control rod assemblies. The control material shall be silver, indium and cadmium as approved by the NRC.

5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The reactor coolant system is designed and shall be maintained:

a.

In accordance with the code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements,

.b.

For a pressure of 2485 psig, and c.

For a temperature of 650*F, except for the pressurizer which is 680*F.

VOLUME 5.4.2 The total water and steam volume of the reactor coolant system is 9829 2 100 cubic feet at a nominal T of 567.2*F.

l avg 5.5 METEOROLOGICAL TOWER LOCATION 5.5.1.The meteorological tower shall be located as shown on Figure 5.5-1.

l i

l FARLEY-UNIT 1 5-6 AMENDMENT NO. 137

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ADMINISTRATIVE CONTROLS ll l

2.

WCAP-10216-P-A, Rev. lA, " Relaxation Of Constant Axial Offset Control / Fg Surveillance Technical Specification," February 1994 (E Proprietary).

(Methodology for Specifications 3.2.1 - Axial Flux Difference and-3.2.2 - Heat Flux Hot Channel Factor.)

34.

WCAP-12945-P-A, Volume 1, Revision 2, and volumes 2 through 5, Revision 1, " Code Qualification Document for Best Estimate LOCA Analysis," March 1998 (H Proprietary).

3b.

WCAP-12610-P-A,

VANTAGE + Fuel Assembly Reference Core Report,"

April 1995 (H Proprietary).

(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)

The core operating limits shall be determined so that all applicable limits (e.g.,

fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

ANNUAL DIESEL GENERATOR PELIABILITY DATA REPORT 6.9.1.12 The number of tests (valid or invalid) and the number of failures to start on demand for each diesel generator shall be submitted to the NRC annually. This report shall contain the information identified in Regulatory Position C.3.b of NRC Regulatory Guide 1.108, Revision 1, 1977.

FARLEY-UNIT 1 6-19a AMENDMENT No. 137

l ADMINISTRATIVE CONTROLS 1

6.15 MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS (Liquid, Gaseous, Solid)

This specification deleted. Refer to the Offsite Dose Calculation Manual and the Process Control Program.

6.16 CONTAINMENT LEAKAGE RATE TESTING PROGRAM A program shall be established to implement the. leakage rate testing of containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option l

B, as modified by approved exemptions. This program shall be in accordance I

with the guidelines contained in Regulatory guide 1.163, " Performance-Based Containment Leak-Test Program," dated September 1995.

l The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 43 psig.

l The maximum allowable containment leakage rate, La, at Pa, is 0.15% of containment air weight per day.

Leakage rate acceptance criteria are:

Containment overall leakage rate acceptance criterion is s 1.0 La-a.

During plant startup following testing in accordance with this program, the leakage rate acceptance criteria are s 0.60 La for the combined Type B and C-tests, and 5 0.75 La for Type A tests; b.

Air lock testing acceptance criteria are:

1) overall air lock leakage rate.is s 0.05 La when tested at 2 Pa*

2)

For each door, leakage rate is 5 0.01 La when pressurized to-2 10 psig.

The provisions of Specification 4.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program.

The provisions of Specification 4.0.3 are applicable to the containment Leakage Rate Testing Program.

l l

l FARLEY-UNIT 1 6-24 AMENDMENT NO.137 1

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4 UNITED STATES l

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NUCLEAR REGULATORY COMMISSION l

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a wAsHsucTow, p.c. sones. coot s...../

SOUTHERN NUCLEAR OPERATING COMPANY. INC.

ALABAMA POWER COMPANY DOCKET NO. 50-364 i

J.QSEPH M. FARLEY NUCLEAR PLANT. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.129 License No. NPF-8 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Southem Nuclear Operating Company, Inc.

(Southem Nuclear), dated February 14,1997, as supplemented by letters dated June 20, August 5, September 22, November 19, December 9, December 17, and December 31,1997, January 23, February 12, February 26, March 3, March 6, March 16, April 3, April 13, and two ktters on April 17,1998, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity w' h the application, the provisions of the d

Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendment; and paragraph 2.C.(2) of Facility Operating License No. NPF-8 is hereby emended to read as follows:

l L

l -

(2)

Technical Soecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.129, are hereby incorporated in the license. Southem Nuclear shall operate the facility in accordance with the Technical Specifications.

1 i

I In addition, the license is amended to add paragraph 2.C.(22) to Facility Operating License No. NPF4 as follows:

(22)-

Additional Conditions The Additional Conditions contained in Appendix C, as revised through Amendment No.129, are hereby incorporated in the license. The licensee shall l

operate the facility in accordance with the additional conditions.

3.

This license amendment is effective as of its date of issuance and shall be implemented prior to entering Mode 4 for Cycle 13 (spring 1998).

FOR THE NUCLEAR REGULATORY COMMISSION l

l ue or Office of Nuclear Reactor Regulation

Attachment:

1. Changes to the Operating License

2. Changes to the Techn! cal Specifications l.-

)

Date ofIssuance:

April 29, 1998

l 4

ATTACHMENT TO LICENSE AMENDMENT NO.129 TO FACILITY OPERATING LICENSE NO. NPF-8 DOCKET NO. 50-364 Replace the fohowing pages of the Operating License with the enclosed pages. The revised pages are indicated by marginal lines.

Ramove Insert 4

4 14 14 19 19 Appendix C Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised areas are indicated by marginal lines.

Remove Inand 14 14 2-5 2-5 24-24 2-7 2-7 3/4 2-15 3/4 2-15 3/4 3-25 3/4 3-25

'3/4 3-27 3/4 3-27 3/4 3-28 3/4 3-28 3/4 4-12b 3/4 4-12b 3/4 5-5 3/4 5-5 3/47-9 3/4 7-9 B 3/4 2-5 B 3/4 2-5 B 3/4 4-3b B 3/4 4-3b B 3/4 & 2 B 3/4 & 2 B 3/4 7-1 B 3/4 7-1 54 54 5 54 6-19a 6-19a 6-24 6-24

-4 (6) southern Nuclear, rursuant to the Act and 10 CFR,

Parts 30, 40 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 license 65:all be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and h 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)

Marinum Power Ymval

~

southern Nuclear'is authorized to operate'the

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facility at reactor core power levels not in excess of 2775 megawatts thermal l

(2)

Technical sneelficatlana The Technical specifications contained in Appendices A and 3, as revised through Amendment No.129, are hereby incorporated in the license.

l Southern Nuclear shall operate the facility in accordance with the Technical specifications.

(3)

Initial Test Promram Alabama Power Company shall conduct the' initial test program (set forth in section 14 of the Final safety Analysis Report as amended any modifications to this program)unless suchwithout making modifications are in accordance with the provisions -

of 10 CFR section 50.59.

In addition Power Company shall not make any major, Alabama modifications to this program unless the modifications have been identified and have received prior NRC approval.

Major modifications are defined as:

Elimination of any test identified as essential a.

in section 14 of the Final safety Analysis Report, as amended; b.

Modification of test objectives, methods or

' acceptance criteria for.any test identified as essential in section 14 of the Final safety Analysis Report, as amended; c.

Performance of any test at a power level different from the level in the described program; and Farley - Unit 3 Amendment No. 83.129

required to provide for an automatic pump trip. This submittal is required within three months after NRC determination of acceptability of the small break LOCA model based on comparisons with LOFT l

test L3-6.

(ii)

If required based on (i) above, complete plant modifications to provide for automatic tripping of reactor coolant pumps within 11 months after NRC determination of model acceptability, provided there is an appropriate outage during that time interval to complete installation or during the first such scheduled outage occurring thereafter.

(3)

With respect to reliability of reactor coolant pump seal cooling (ll.K.3.25),

(i)

Prior to January 1,1982, submit results of analyses or experiments to determine consequences of a loss of cooling water to the reactor coolant pump seal coolers and describe any modifications found necessary.

(ii)

Prior to July 1,1982, complete any necessary modifications.

(4)

With respect to a revised small break LOCA model, (i)

Prior to January 1,1982, submit to the NRC a revised model to account for recent experimental

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data (ll.K.3.30).-

1 (ii)

Submit to the NRC the results of plant-specific calculations using the NRC-approved revised model prior to January 1,1983.

j (22)

Additional Conditions i

The Additional Conditions contained in Appendix C, as revised through Amendment No.129, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the additional conditions.-

Farley - Unit 2 Amendment No. 88,129

1

, J. This license is effective as of the date of issuance and shall expire at midnight, March 31,2021.

FOR THE NUCLEAR REGULATORY COMMISSION i

)

original signed by:

Harold R. Denton, Director Office of Nuclear Reactor Regulation -

Attachment:

1. Appendix A - Technical Specifications (NUREG-0697, as revised)

2. Apperidix B - Environmental Protection Plan

3. Appendix C - Additional Conditions l

Date oflasuance:

March 31,1981 l

l l.

l Farley Unit 2 Amendment No. 7'1, 88, 129 1

APPENDIX C ADDITIONAL CONDITIONS OPERATING LICENSE NO. NPF 8 Southem Nuclear Operating Company, Inc. (SNC), shall comply with the following conditions on the schedules noted below:

Amendment Condition Completion Number Addrhonal Condition Date 129 SNC shall complete classroom and simulator training Prior to Unit 2 entering for operations crews and temporary simulator Mode 2 from the modifications as described in SNC's letter dated spring 1998 refueling September 22,1997, and evaluated in the staff's Safety outage Evaluation dated April 29,1998.

129 SNC shall review the results of the Cycle 13 startup Prior to Unit 1 startup testing to determine any potential effects on operator from the fall 1998 training and incorporate these changes into licensed refueling outage operator training as described in SNC's letter dated September 22,1997, and evaluated in the staff's Safety Evaluation dated April 29,1998.

129 SNC shall provide a Steam Generator (SG) Tube Prior to the Unit 2 Rupture radiological consequences analysis that steam generator incorporates a flashing fraction, which is replacement outage appropriate for the Unit 2 design.

in spring 2001 Amendment No.129

e DEFINITIONS RATED THERMAL POWER 1.25 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 2775 MWt.

l REACTOR TRIP SYSTEM RESPONSE TIME 1.26 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until loss of stationary. gripper coil voltage.

REPORTABLE EVENT 1.27 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50.

SHUTDOWN. MARGIN 1.28 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which

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the reactor is suberitical or would be suberitical from its present condition assuming all full length rod cluster assemblies (shutdown and control) arw fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn.

SOLIDIFICATION I

1.29 This definition deleted. Refer to the Process Control Program, i

SOURCE CHECK 1.30 A SOURCE CHECK shall be the qualitative assessment of' channel response when the channel sensor is exposed to_a radioactive source.

STAGGERED TEST BASIS 1.31 A STAGGERED TEST BASIS shall consist oft i

a.

A test schedule for n systems, subsystems, trains or other designated components obtained by dividing the specified test interval into n equal subintervals, b.

The testing of one system, subsystem, train or other designated component at the beginning of esta subinterval.

FARLEY-UNIT 2 1-6 AMENDMENT NO.129

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a REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 10.

Preservice Inanection means an inspection of the full length of each tube in each steam generator performed by eddy current techniques prior to service to establish a baseline condition of the tubing. This inspection shall be performed after the field hydrostatic-test and prior to initial POWER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections.

11.

F* Distance is the distance of the expanded portion of a tube which provides a sufficient length of undegradod tube expansion to resist pullout of the tube from the tubesheet.

The F* distance is equal to 1.60 inches plus allowance for l

eddy current uncertainty measurement and is measured down from the top of the tube sheet or the bottom of the roll transition, whichever is lower in elevation. The allowance for eddy current uncertainty is documented in the steam gen'erator eddy current inspection procedure.

12.

F* Tube is a tube:

a) with degradation equal to or greater than 40% below the F* distance, and b) which has no indication of imperfections greater than or equal to 20% of nominal wall thickness within the Fa distance,'and c) that remains inservice.

13.

Tube Exoansion is that portion'of a tube which has been increased in diameter by a rolling process such that no crevice exists between the outside diameter of the tube and the hole in the tubesheet. Tube expansion also refers to that portion of a. sleeve which has been increased in diameter by a rolling process such that no crevice exists between the outside diameter of the sleeve and the parent steam generator tube.

14.

Ig,hr,unnert Plate Recair Limit is used for the dir asition of an alloy 600 steam generator tube for i

continued service that is experiencing predominately axially oriented outside diameter stress corrosion cracking confined within the thickness of the t.ube support plates. At tube support plate intersections, the repair limit is based on maintaining steam generator tube serviceability as described below:

FARLEY-UNIT 2-3/4 4-12b AMENDMENT No. 129 1

l I

l EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Cont.' nued)

I By verifying the correct position of each mechanical position stop e.

for the following ECCS throttle valves:

1.

Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following completion of each valve stroking operation or maintenance on the valve when the ECCS subsystems are required to M OPERABLE.

2.

At least once per 18 months.

Valve Number i

J CVC-V-8991 A/B/C CVC-V-8989 A/B/C CVC-V-8996 A/B/C i

CVC-V-8994 A/B/C i

l f.

At least once per 18 months, daring shutdown, by:

1.

Verifying that'each automatic valve in the flow path actuates to'its correct position on a safety injection test signal.

2.

Verifying that each of the following pumps start automatically upon receipt of a safety injection test signals a)

Centrifugal charging pump b)

Residual heat removal pump g.

By verifying that each of the following pumps develops the indicated differential pressure on recirculation flow when tested pursuant to specification 4.0.5:

1.

Centrifugal charging pump 2 2323 psid l

2.

Residual heat removal pump 2 145 psid l

h.

Prior to entry into Mode 3 from Mode 4, verify that the mechanical stops on low head safety injection valve RHR-HV 603 A/B are intact.

l-I l

l k

FARLEY-UNIT 2 3/4 5-5 AMENDMENT NO. 129 i

r a

e PIANT SYSTEMS MAIN STEAM..LINE ISOLATION VALE.$

LIMITING CONDITION FOR OPERATION 3.7.1.5 Each main steam line isolation valve shall be OPERABLE.

l APPLICABILITY: MODES 1, 2 and 3.

ACTION:-

MODE 1-With'one main' steam line isolation' valve inoperable,' POWER l

OPERATION may continue provided the inoperable valve is restored to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; otherwise, reduce power to less than or equal to 5% of RATED THERMAL POWER within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

I MODES 2 With one main steam line isolation valve inoperable, and 3 subsequent operation in MODES 2 or 3 may proceed provided the isolation valve is restored to OPERABLE status or closed I

within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after entering MODE 2; otherwise, be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

The provisions of Specification 3.0.4 are not applicable.

1 L.

I SURVEILLANCE REQUIREMENTS i

l 4.7.1.5 Each main steam line isolation valve shall be demonstrated OPERABLE by verifying full closure within 7 seconds when tested pursuant to l

. Specification 4.0.5.

i.

l l

l 3

FARLEY-UNIT 2 3/4 7-9 AMENDMENT No.129

];

POWER DISTRIBUTION LIMITS RASES'.

- 3/4.2.4' CUADRANT POWER TILT RATIO The quadrant power tilt ratio limit assures that the radial power distribution j

satisfies the-design values used in'the power capability analysis. Radial

_ power distribution measurements'are made during startup testing and periodically during power operation.

. The limit of'1.02,'at'which corrective action is required, provides DNS and' linear heat generation: rate' protection with x-y plane power tilts.

The two hour time allowance for-operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correction of a dropped or misaligned control rod.

In the event such action does not correct the tilt, the margin for uncertainty on Fo is reinstated by reducing.

. the maximum allowed power by 3 percent for each p5rcent of tilt in excess of g

. 1.0.

4 1

For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the movable incore detectors are used-to confirm that the normalized symmetric power distribution is consistent with the QUADRANT-POWER 1

l TILT RATIO. The-incore detector monitoring is done with a full incore flux-

- map'or two sets of four symmetric thimbles. The two sets.of four symmetric

' thimbles is a unique set of eight detector locations. These locations are.

)

C-8, E-5, E-11, H-3, H-13, L-5, L-11, and N-8.

)

3/4.2.5 DhB PARAMETERS I-The limits on the.DN8 related parameters assure that each of the parameters'are maintained within.the normal steady state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to meet the DNE design criterion ~throughout each analysed transient. The indicated T value of 580.3*F is based on the average of two. control board readings' and En indication uncertainty of 2.9"F.

The indicated. pressure value of 2209 psigtis:

based on the. average of two control-board readings and an indication uncertainty of 24-psi.

The indicated total RCS flow rate is based.on-the'

.averagefof two elbow tap measurements from'each loop as read'on'the plant computer and an uncertainty of'2.44: flow (0.1% flow is included for feedwater un

, venturi fouling).

1 The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months surveillance of Tav and pressurizer pressure through the control boardreadingsare'sufficientt8ensurethattheparametersarerestored-

- within their limits following load changes and other expected transient operation.

The 18 month surveillance of the total RCS flow rate may be performed by one of two alternate methods.. One method is a precision heat balance performed at the l

l beginning of each fuel cycle..The other method is based on the Ap measurements g

from the cold leg elbow taps, which are correlated to past precision heat balance K

measurements. Correlation of the flow indication channels with selected L

- precision ~ loop flow measurements is documented in WCAP-14750. Use of the elbow l

tap'Ap measurement method removes the requirement for performance of a precision'

RCS flow heat *oalance~ measurement for that cycle. The monthly surveillance of l

the total RCS flow rate is a reverification of.the RCS flow requirement using.

]

-process computer indications of loop. elbow tap measurements that are correlated

'either to the precision F.CS. flow measurement or the elbow tap measurement at:the 1

beginning of the-fuel cycle.- The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months RCS flow surveillance'is a qualitative l

verification of significant flow degradation using the. control board indicators i

fed by? elbow. tap measurements.

]

FARLEk-UNIT 2 B 3/4 2-5 AMENDMENT NO. 129 1

i

ar c

REACTOR COOLANT SYSTEM RASES

- ~ _ - _ - _ _ _

2.

Indication of tube degradation of any type including a complete guillotine break in the tube between the bottom of the upper joint and the top of the lower roll expansion does not require that the tube be removed from service.

3.

The tube plugging limit continues to apply to the portion of the

. tube in the entire upper joint region and in the lower roll i

expansion. As noted above, the sleeve plugging limit applies to I

these areas also.

l 4.

The tube plugging limit continues to apply to that portion of the l

tube above the top of the upper joint.

b. Laser Walded l

l 1.

Indications of degradation in the length of the sleeve between the L

weld joints must La evaluated against the sleeve plugging limit.

L 2.

Indication of tube degradation of any type including a complete j

break in the tube between the upper weld joint and the. lower weld-joint does not require that the tube be removed from service.

f 3.

At the weld joint, degradation must be evaluated in both the l

l sleeve and tube.

j 4.

In a joint with more than one weld, the weld closest to the end of 1

the sleeve represents the joint to be inspected and the limit of the sleeve inspection.

i L

.5.

The tube p' lugging limit continues to apply to the portion of the i

~

tube abova the upper weld joint and below the lower weld joint.

l F* tubes do not haie to be plugged or repaired provided the remainder of the tube within the tubesheet that is above the F* distance is not degraded. The F* distance is equal to 1.60 inches plus allowance for eddy current l

uncertainty measurement and is measured down-from the top of the tubesheet or

'the bottom of the roll transition, whichever is lower in elevation.

Steam generator tube inspections of operating plants have demonstrated the i

capability to reliably detect wastage type de;;adation that has penetrated 20%

.of the original tube wall thickness.

1 Whenever the results of any steam generator tubing inservice inspection fall

[

into category c-3, these results'will be reported to the commission pursuant to 10 CFR 50.73 prior to resumption of plant operation. Such cases will be l

considered by'the commission on a case-by-case basis and may result in a L

requirement for analysis, laboratory examinations, tests, additional addy-current inspection, and revision to the Technical Specifications, if l

necessary.

I i

FARLEY-UNIT'2 B 3/4 4-3b AMENDMENT NO.- 129 l

\\

e-f l

t DONTAIMMENT SYSTEMS

^888 l

l r

l l

The maximum peak calculated containment internal pressure obtained from l

4 LOCA event is 43 poig and from a MSLB event is 53 psig. These containment analyses calculations include an initial positive pressure of up to 3 peig.

i

.The analyses'results demonstrate that the maximum containment pressure will remain below the C=.ign limit of 54 psig.

3/4.6.1.5~

AIR IakiERATURE The. limit 6tions on containment :

e air temperature ensure that the overall containment average gir tempa loes not exceed the initial temperature condition assumed in the s analysis for a LOCA or steam' i

line break accident.

3/4.6.1.6 CONTAINMENT STRUCTURAL INTEGRITY This 1bsitation ensures that the structural integrity of the containment i

will be maintained comparable to the original design' standards for the life of j

the facility.

Structural integrity is required to ensure that the containment i

will withstand the maximum peak calculated internal pressure of 43 peig in the event of a-LOCA, including an initial positive pressure of up to 3 psig. In addition,. structural integrity is required to ensure that the containment will withstand the maximum peak calculated internal pressure of 53 psig in the-L event of a MsLB,' including an initial positive. pressure of up.to 3 psig.: As per Appendix J, Option B, the LOCA peak calculated containment internal' L

pressure defines the Pa value for the Containment Leak., 2 ate Testing Program j

required by Surveillance 4.6.1.2.: The visual examination of tendons, L

anchorages and exposed interior and exterior surfaces of the containment, and the containment leakage tests along with the data obtained from Unit 1 tendon surveillance, are sufficient to demonstrate this capability.

~

[

The surveillance requirements for demonstrating the containment's l'

structural integrity are in compliance with the recommendations of paragraph l

C.1.3 of. Regulatory Guide 1.35 " Inservice Surveillance of Ungrouted Tendons in Prestressed Concrete Containment Structure," January 1976.

3/4.6.1.7 CONTAINMENT VENTILATION SYSTEM

-The 48-inch containment purge supply and exhaust isolation valves are

' required to be closed in MODES above COLD SHUTDOWN since these valves have not been demonstrated capable of closing during a LOCA or steam line break L

accident.- Maintaining these valves closed during plant operations ensures that. excessive quantities of radioactive materials will not be released via L

the containment purge system.

!~

The use of the containment purge lines-is restricted to the 8-inch vent supply and exhaust isolation. valves to ensure that the site boundary dose guidelines of:10 CFR Part 100 would not be exceeded in the event'of a loss-of-coolant accident during venting operations.

m

-FARLEY-UNIT 2 3 3/4 6-2 Amendment No. 129 l

e 1

"3/4.7 PLANT SYSTEMS BASES 3/4.7.1' TURRIME CYCLE 3/4.7.1.1 EAFETY VALVES

- The OPERABILITY of the main steam line code safety valves ensures that the secondary system pressure will be limited to within.110% (1194 psig) of-its design pressure of 1085 psig during the most severe anticipated system operational transient. The maximum relieving capacity is associated with a turbine trip from 100% RATED THERMAL POWER coincident with an assumed loss of condenser heat sink (i.e., no steam bypass to the condenser).

The specified valve lift settings and relieving capacities are in accordance with the requirements of section III of the ASME Boiler and Pressure Code, 1971 Edition, The total relieving capacity for all valves on all of the steam lines is at least 12,984,660 lbs/hr which is 105.8 percent of the total secondary steam flow of 12,270,000 lbs/hr at 100% RATED THERMAL POWER.

'A minimum of 2 OPERABLE safety valves per steam generator ensures that sufficient relieving capacity is available for the allowable THERMAL POWER restriction in Table 3.7-2.

STARTUP and/or POWER OPERATION is allowablo with safety valves

. inoperable within the limitations of the ACTION requirements on the basis of the reduction in secondary system steam flow and THERMAL POWER required by the the reduced reactor trip settings of the Power Range Neutron Flux channels.

'The reactor trip setpoint reductions are consistent with the assumptions used in the accident analysis.

l FARLEY-UNIT 2 B 3/4 7-1 AMENDMENT NO.129

l t

DESIGN FEATURES 5.3 REACTOR OORE FUEL ASSEMBLIES 5.3.1 The reactor shall contain 157 fuel assamblies. Each assembly shall consist of a matrix of zirconium alloy, zircaloy-4, or IIRLO* fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO )

2 as fuel material.

Limited substitutions of zirconium alloy, sircaloy-4, EIRLO*, or stainless steel filler rods for fuel rods, in accordance with NRC-approved applications l

of fuel rod configurations, may be used.

Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC-approved codes and methods, and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in non-limiting core regions.

CONTROL ROD ASSEMBLIES l-5.3.2 The reactor core shall contain 48 control rod assemblies. The control material shall be silver, indium and cadmium as approved by the NRC.

L L4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE l

5.4.1 The reactor coolant system is designed and shall be maintained:

a.

In accordance with the. code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements, b.

For a pressure of 2485 psig, and c.

For a temperature of 650'F, except for the pressurizer which is 680*F.

VOLUME 5.4.2 The total water and steam volume of the reactor coolant system is 9829 2 100 cubic feet at a nominal T of 567.2*F.

l ayg l

5.5 METEOROLOGICAL TOWER LOCATION l

5.5.1 The meteorological tower shall be located as shown on Figure 5.5-1.

FARLEY-UNIT 2 5-6 AMENDMENT NO.129 1

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ADMINISTRATIVE CONTROLS 2.

WCAP-10216-P-A, Rev. lA, " Relaxation Of Constant Axial Offset i

l Control / Fg surveillance Technical specification," February 1994 (E Proprietary).

(Methodology for specifications 3.2.1 - Axial Flux Difference and I

3.2.2 - Heat Flux Hot Channel Factor.)

i 3a.

WCAP-12945-P-A, Volume 1, Revision 2, and volumes 2 through 5, Revision 1, " Code Qualification Document for Best Estimate LOCA Analysis,* March 1998 (E Proprietary.).

3b.

WCAP-12610-P-A, " VANTAGE + Fuel Assembly Reference Core Report,"

April 1995 (E Proprietary).

(Methodology for Specification 3.2.2 - Heat Flux Bot Channel Factor.)

The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

I ANNUAL DIESEL GENERATOR RELIABILITY DATA REPORT i

t 6.9.1.12 The number of tests (valid or invalid) and the number of failures to start on demand for each diesel generator shall be submitted to the NRC annually. This report shall contain the information identified in Regulatory Position C.3.b of NRC Regulatory Guide 1.108, Revision 1, j

1977.

f

(

FARLEY-UNIT 2 6-19a AMENDMENT NO. 129

I-ADMINISTRATIVE CONTROLS' 6.15 MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS (Liquid, Gaseous, Solid)

This specification deleted.. Refer to the Offsite Dose Calculation Manual and the Process Control Program.

6.16 CONTAINMENT LEAKAGE RATE TESTING PROGRAM l

A program shall be established to implement the leakage rate testing of containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option t

l B, as modified by approved exemptions. This program shall be in accordan,ce l

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 43 psig.

l The maximum allowable containment leakage rate, La, at Pa, is 0.15% of containment air weight per day.

I l

Leakage rate acceptance criteria are:

I Containnent overall leakage rate acceptance criterion is 51.0 La.

a.

During plant startup following testing in accordance with this program, the leakage rate acceptance criteria are 5 0.60 La for the combined Type B and C tests, and s 0.75'La for Type A tests; b.

Air lock testing acceptance criteria are:

l 1)

Overall air lock leakage rate is 5 0.05 La when tested at 2 Pa.

i l

2)

For each door, leakage rate is 5 0.01 La when pressurized to l

2 10 psig.

The provisions of Specification 4.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program.

l The provisions of Specification 4.0.3 are applicable to the Containment Leakage Rate Testing Program.

l 1

l l

FARLEr-UNIT 2 6-24 AMENDMENT NO. 129 1

l f

I

ML20217P047

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