ML050940197

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Attachment 2, Memo, Correction to License Authority Files
ML050940197
Person / Time
Site: Farley  Southern Nuclear icon.png
Issue date: 03/29/2005
From: Sean Peters
NRC/NRR/DLPM/LPD2
To: Harris J
NRC/OCIO
Peters S, NRR/DLPM, 415-1842
Shared Package
ML050890364 List:
References
TAC MC3231, TAC MC3232
Download: ML050940197 (15)
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Text

UNITED STATES X.. NUCLEAR REGULATORY COMMISSION VASHINGTON, D.C. 20555-0001 December 29, 1999 Mr. D. N. Morey Vice President - Farley Project Southern Nuclear Operating Company, Inc.

Post Office Box 1295 Birmingham, Alabama 35201-1295

SUBJECT:

JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2-ISSUANCE OF AMENDMENTS RE: STEAM GENERATOR REPLACEMENTS (TAC NOS. MA4393 and MA4394)

Dear Mr. Morey:

The Nuclear Regulatory Commission has Issued the enclosed Amendment No. 147 to Facility Operating License No. NPF-2 and Amendment No. 138 to Facility Operating License No. NPF-8 for the Joseph M. Farley Nuclear Plant, Units 1 and 2. The amendments change the Unit 1 and Unit 2 Improved Technical Specifications (ITS) in response to your application of December 1, 1998, as supplemented by your letters of April 21, July 19, October 18, and November 11, 1999. The amendments revise the ITS to address changes associated with replacing the current Westinghouse Model 51 steam generators with Westinghouse Model 54F steam generators. The Unit 1 ITS set applies after you replace the Unit 1 steam generators in spring 2000 until you replace the Unit 2 steam generators in spring 2001. The Unit 2 ITS set applies after you replace both the Unit 1 and the Unit 2 steam generators.

We are also enclosing a copy of our related safety evaluation. We will include a Notice of Issuance in the Commission's biweekly FederalRegister Notice.

Sincerely,

/ \VI' L. Mark Padovan, Project Manager, Section 1 Project Directorate II Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-348 and 50-364

Enclosures:

1. Amendment No. 147 to NPF-2
2. Amendment No. 138 to NPF-8
3. Safety Evaluation cc w/ ends: See next page ATTACHMENT 2 I t

ATTACHMENT TO LICENSE AMENDMENT No. 138 TO FACILITY OPERATING LICENSE NO. NPF-8 DOCKET NO. 50-364 Replace the following pages of Facility Operating License No. NPF-8 with the attached revised pages. The revised pages are Identified by amendment number and contain vertical lines Indicating area of changes. Pages noted with an "*' have changed only due to Information rolling over from one page to another.

Remove Insert Remove Insert 3.3.1 -17 3.3.1-17 B 3.6.5-3* B 3.6.5-3*

3.3.2-11 3.3.2-11 B 3.6.6-3 B 3.6.6-3 3.4.5-2 3.4.5-2 B 3.7.16-1 B 3.7.16-1 B 3.4.5-5 B 3.4.5-5 5.5-5 5.5-5 B 3.4.5-6 B 3.4.5-6 5.5-6 5.5-6 3.4.6-2 3.4.6-2 5.5-7 5.5-7 B 3.4.6-5 B 3.4.6-5 5.5-8 5.5-8 3.4.7-1 3.4.7-1 5.5-9 5.5-9 3.4.7-2 3.4.7-2 5.5-10* 5.5-10 B 3.4.7-1 B 3.4.7-1 5.5-11 5.5-11 B 3.4.7-2 B 3.4.7-2 5.5-12' 5.5-12 B 3.4.7-4 B 3.4.7-4 5.5-13* 5.5-13 B 3.4.7-5 B 3.4.7-5 5.5-1 5.5-14 3.4.13-1 3.4.13-1 5.5-15* 5.5-15 B 3.4.13-2 B 3.4.13-2 5.5-16 5.5-16 B 3.4.13-3* B 3.4.13-3* 5.5-17 5.5-17 B 3.4.13-4* B 3.4.13-4* 5.5-18* 5.5-18 3.4.16-1 3.4.16-1 5.5-19 5.5-19 3.4.16-2 3.4.16-2 5.5-20 Delete 3.4.16-4 3.4.16-4 5.5-21 Delete B 3.4.16-1 B 3.4.16-1 5.5-22 Delete B 3.4.16-2 B 3.4.16-2 5.5-23 Delete B 3.4.16-3 B 3.4.16-3 5.5-24 Delete B 3.6.1-2 B 3.6.1-2 5.5-25 Delete B 3.6.2-2 B 3.6.2-2 5.6-5 5.6-5 B 3.6.4-1 B 3.6.4-1 5.6-6 5.6-6 B 3.6.5-2 B 3.6.5-2

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 4 of 8)

Reactor Trip System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT

11. Reactor Coolant Pump (RCP)

Breaker Position

a. Single Loop 1 (g) 1 per RCP N SR 3.3.1.12 NA NA
b. Two Loops 1 (h) 1 per RCP M SR 3.3.1.12 NA NA 12Z Undervoltage l) 3 M SR3.3.1.6 2640V 22680V RCPs SR 3.3.1.10
13. Undertrequency 1(0) 3 M SR 3.3.1.6 2 56.9 Hz a 57 Hz RCPs SR 3.3.1.10
14. Steam 1.2 3 per SG E SR 3.3.1.1. 227.6% 2 28B Generator (SG) SR 3.3.1.7 I

Water Level- SR 3.3.1.10 Low Low SR 3.3.1.14 (1) Above the P-7 (Low Power Reactor Trips Block) Interlock.

(g) Above the P4 (Power Range Neutron Flux) interlock (h) Above the P-7 (Low Power Reactor Trips Block) Interlock and below the P-8 (Power Range Neutron Flux) interlock.

Farley Units 1 and 2 3.3.1-17 Amendment No. 147 (Unit 1)

Amendment No. 138 (Unit 2)

ESFAS Instrumentation 3.3.2

- .Table 3.3.2-1 (page 4 of 4)

Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT

5. Turbine Trip and Feedwater Isolation
a. Automatic Actuation 1.2 2 trains H SR 3.3.22 NA NA Logic and Actuation SR 3.3.2.3 Relays SR 3.3.2.8
b. SG Water Level- 1.2 3perSG I SR 3.32.1 S 82.4% - 82% I High High (P-1 4) SR 3.3.2.4 SR 3.3.2.7 SR 3.3.2.9
c. Safety Injecticn Refer to Function 1 (Safety Injection) for all initiation functions and requirements.
6. Auxiliary Feedwater
a. Automatic Actuation 1,23 2 trains G SR 3.3.2.2 NA NA Logic and Actuation SR 3.3.2.3 Relays SR 3.3.2.8
b. SG Water Level - 1,2.3 3 per SG D SR 3.3.2.1 2 27.6% a28% 1 Low Low SR 3.3.2.4 SR 3.3.2.7 SR 3.3.2.9's'
c. Safety Injection Refer to Function 1 (Safely Injection) for al Initiatlon functions and requirements.
d. Undervoltage 1.2 3 I SR 3.3.2.5 2 2640 volts 2 2680 volts Reactor Coolant SR 3.3.2.7 Pump SR 3.3.2.9
e. Trip of all Main 1 2 per pump J SR 3.3.2.10 NA NA Feedwater Pumps
7. ESFAS Interlocks
a. Automatic Actuation 1.2,3 2 trains L SR 3.3.2.2 NA NA Logic and Actuation SR 3.32.3 Relays SR 3.3.2.8
b. Reactor Trip, P-4 1.2,3 I per train, C SR 3.3.2.6 NA NA 2 trains
c. Pressurizer 1,2,3 3 K SR 3.3.2.4 s 2003 psig S2000 psig Pressure. P.11 SR 3.3.2.7
d. Tag. Law Low. P-12 1.2.3 1 per loop K SR 3.3.2.4 2 542.6-F . 543-F (Decreasing) SR 3.3.2.7 S 545.4-F s 545 F (increasing)

(g) Applicable to MDAFW pumps only.

Farley Units 1 and 2 3.3.2-11 Amendment No. 147 (Unit 1)

Amendment No. 138 (Unit 2)

RCS Loops-MODE 3 3.4.5 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. One required RCS loop C.1 Restore required RCS 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> not in operation, and loop to operation.

reactor trip breakers closed and Rod Control OR System capable of rod withdrawal. C.2 De-energize all control 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> rod drive mechanisms (CRDMs).

D. Two required RCS loops D.1 De-energize all CRDMs. Immediately inoperable.

AND OR D.2 Suspend all operations Immediately No RCS loop in involving a reduction of operation. RCS boron concentration.

AND D.3 Initiate action to restore Immediately one RCS loop to OPERABLE status and operation.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.5.1 Verify required RCS loops are in operation. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SR 3.4.5.2 Verify steam generator secondary side water levels 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> are 2 30% (narrow range) for required RCS loops. I SR 3.4.5.3 Verify correct breaker alignment and indicated power 7 days are available to the required pump that is not in operation.

Farley Units 1 and 2 3.4.5-2 Amendment No.147 (Unit 1)

Amendment No.138 (Unit 2)

RCS Loops - MODE 4 3.4.6 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. One required RHR loop B.1 Be in MODE 5. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> inoperable.

AND Two required RCS loops inoperable.

C. Required RCS or RHR C.1 Suspend all operations Immediately loops inoperable. involving a reduction of RCS boron concentration.

OR AND No RCS or RHR loop in operation. C.2 Initiate action to restore Immediately one loop to OPERABLE status and operation.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.6.1 Verify one RHR or RCS loop is in operation. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SR 3.4.6.2 Verify SG secondary side water levels are 2 75% 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> I (wide range) for required RCS loops.

SR 3.4.6.3 Verify correct breaker alignment and indicated power 7 days are available to the required pump that is not in operation.

Farley Units 1 and 2 3.4.6-2 Amendment No.147 (Unit 1)

Amendment No.138 (Unit 2)

RCS Loops-MODE 5, Loops Filled 3.4.7 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.7 RCS Loops-MODE 5, Loops Filled LCO 3.4.7 One residual heat removal (RHR) loop shall be OPERABLE and in operation, and either:

a. One additional RHR loop shall be OPERABLE; or
b. The secondary side water level of at least two steam generators (SGs) shall be 2 75% (wide range).

- _----------- -------------------- NOTES---

1. The RHR pump of the loop in operation may not be in operation for s 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period provided:
a. No operations are permitted that would cause reduction of the RCS boron concentration; and
b. Core outlet temperature is maintained at least 10°F below saturation temperature.
2. One required RHR loop may be inoperable for 5 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing provided that the other RHR loop is OPERABLE and in operation.
3. No reactor coolant pump shall be started with one or more RCS cold leg temperatures s 3250F unless:
a. The secondary side water temperature of each SG is < 50'F above each-of the RCS cold leg temperatures; or
b. The pressurizer water volume is less than 770 cubic feet (24%

of wide range, cold, pressurizer level indication).

4. All RHR loops may be removed from operation during planned heatup to MODE 4 when at least one RCS loop is in operation.
5. The number of operating Reactor Coolant Pumps is limited to one at RCS temperatures < 1100 F with the exception that a second pump may be started for the purpose of maintaining continuous flow while taking the operating pump out of service.

Farley Units 1 and 2 3.4.7-1 Amendment No.147 (Unit 1)

Amendment No.138 (Unit 2)

RO-S .oop- 'MCDE 5- Us =.i: o 4- .

APPLICABILITY. MODE 5 with RCS .Cops fiile3.

ACTIONS ______.__

CONDITION REQUIRED ACTION COMPLETION TIME I!

A. One RHR loop IA.1 Initiate action to restore a Immediately inoperable. second RHR loop to OPERABLE status.

AND!

OR Required SGs secondary side water levels not A.2 Initiate action to restore Immediately within limits. . required SG secondary side water levels to within limits.

B. Required RHR loops S.1 Suspend all operations Immediately inoperable. involving a reduction of RCS boron concentration.

OR I AND No RHR loop in operation. B.2 Initiate action to restore Immediately one RHR loop to OPERABLE status and

_operatio..

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.7.1 Verify one RHR loop is in operation. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SR 3.4.7.2 Verify SG secondary side water level is > 75% (wide 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> l range) in required SGs.

Farley Units 1 and 2 3.4.7-2 Amendment No.147 (Unit 1)

Amendment No.138 (Unit2)

I

APPLICABILITY: MODES 1 and 2, MODE 3 with RCS average temperature (Tavg) > 500 0F.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. DOSE EQUIVALENT - ------- Note-----------_

1-131 > 0.5 pCi/gm. LCO 3.0.4 is not applicable. I A.1 Verify DOSE Once per4 hours EQUIVALENT 1-131 within the acceptable region of Figure 3.4.16-1.

AND A.2 Restore DOSE 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> EQUIVALENT 1-131 to within limit.

B. Gross specific activity of B.1 Be in MODE 3 with 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> the reactor coolant not Tavg < 500°F.

within limit.

Farley Units I and 2 3.4.16-1 Amendment No.147 (Unit 1)

Amendment No.138 (Unit 2)

RCS Specific Activity 3.4.16 ACTIONS CONDITION l REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Be in MODE 3 with 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Tavq < 500°F.

Time of Condition A not met.

OR DOSE EQUIVALENT 1-131 in the unacceptable region of Figure 3.4.16-1.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.16.1 Verify reactor coolant gross specific activity 7 days

< 100/E PCi/gm.

SR 3.4.16.2 ---------------------------- NOTE--------------------------

Only required to be performed in MODE 1.

Verify reactor coolant DOSE EQUIVALENT 1-131 14 days specific activity 5 0.5 pCi/gm. I AND Between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after a THERMAL POWER change of 2 15% RTP within a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period Farley Units 1 and 2 3.4.16-2 Amendment No.147 (Unit 1)

Amendment No.138 (Unit 2)

ROS Specific Activity 3.4.16 275 .I .... . I

. .I I 250 225 4-.

C:

cc 0

o - 200 o)E .... ........ ....... Unacceptable.. ..J 175 150 CZ Acceptable ... .t...z:............ .....

125 WI,- . ..... .....

LU 100 0 .. . . .. . . .. -. a......

0 75 50 25 0

20 30 40 50- 60 70 80 s0 100 Percent of RATED THERMAL POWER Figure 3.4.16-1 DOSE EQUIVALENT 1-1 31 Primary Coolant Specific Activity Limit Versus Percent of RATED THERMAL POWER with the Primary Coolant Specific Activity > 0.5 [LCi/gm DOSE EQUIVALENT 1-131. I Farley Units 1 and 2 3.4.16-4 Amendment No.147 (Unit 1)

Amendment No.138 (Unit 2)

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 November 30, 1999 Mr. D. N. Morey Vice President - Farley Project Southern Nuclear Operating Company, Inc.

Post Office Box 1295 Birmingham, Alabama 35201-1295

SUBJECT:

JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2-ISSUANCE OF AMENDMENTS RE: CONVERSION TO IMPROVED STANDARD TECHNICAL SPECIFICATIONS (TAC NOS. MA1364 and MA1365)

Dear Mr. Morey:

The Nuclear Regulatory Commission has issued the enclosed Amendment No. 146 to Facility Operating License No. NPF-2 and Amendment No. 137 to Facility Operating License No.

NPF-8 for the Joseph M. Farley Nuclear Plant, Units 1 and 2. The amendments change the Unit 1 and Unit 2 Technical Specifications (TS), TS Bases, and Facility Operating Licenses in response to your application of March 12, 1998, as supplemented by your following letters:

-April 24, 1998 - April 30, 1999 (two letters) -August 30, 1999

-August 20, 1998 - May 28, 1999 - September 15, 1999

- November 20,1998 -June 30, 1999 - September 23, 1999

-February 3, 1999 -July 27, 1999

- February 20, 1999 -August 19, 1999 The amendments fully convert your Current TS (CTS) to Improved TS (ITS) based on NUREG-1431, Standard Technical Specifications, Westinghouse Plants,' Revision 1, of April 1995. The amendments add two new Additional Conditions to Appendix C of the Unit I and Unit 2 Facility Operating Licenses. The first new Additional Condition authorizes you to relocate certain CTS requirements to Southern Nuclear Operating Company-controlled documents. The second new condition addresses the schedule for performing new and revised ITS surveillances.

- D o V 4iDc Lo .o-')-

Mr. D. N. Morey November 30, 1999 We have also enclosed a copy of our related safety evaluation. We will include a Notice of Issuance in the Commission's biweekly Federal Register Notice.

Sincerely, Original signed by:

L. Mark Padovan, Project Manager, Section 1 Project Directorate 11 Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-348 and 50-364

Enclosures:

1. Amendment No.146to NPF-2
2. Amendment No.1 37to NPF-8
3. Safety Evaluation
4. Notice of Issuance cc w/encls: See next page DISTRIBUTION w/Technical Specifications:

Docket File MPadovan PUBLIC WBeckner GHill (4)

DISTRIBUTION wo/Technical Specifications PD 11-1 R/F CSchulten ACRS OGC RScholl (E-mail SE)

PSkinner, RII HBerkow DOCUMENT NAME: G:\PDii-1\Farley\Farley ITS amd rev5wpd *No majorchanges to SEs.

To receive a copy of this document, Indicate In the box: "C" = Copy with SE but no TS "E" = Copy with enclosures

'N" = No copy OFFICE PD11-1/PM PD11-11/A C lQMB/BC SPLB/SC SRXB/ EELB/SC NAME MPadova CHawes SE d e SEs d SEs datew SE d "d DATE 10/ It 199 l I99 /1 0/98 1 &7/26/99 W,519. &7/23/99 2 98/

EMCBISC/ EEIBC IOLB/SC COL1/SCRTSBIBC PDII-11SC SE di- SE dd SE dN' SE d WBeckner t) REmch

/8/99 6/16/99 628/28/6/28/99 \/99 ( 9 //3199 OFFICIAL RECORD COPY 4

AC Sources -Operating 3.8.1 SURVEiLLANCE REQUREMENTS SURVEILLANCE FREQUENCY SR 3.8.1.3 NOTES 7 --

1. DG loadings may Include gradual loading as recommended by the manufacturer.
2. Momentary transients outside the load range do not Invalidate this test.

3.. This'Surveillance shall be conducted on only one DG at a time:

4. This SR shall be preceded by and Immediately follow without shutdown a successful performance of SR 3.8.1.2 or SR 3.8.1.6.

Verify each DG Issynchronized and loaded and 31 days operates for 2 60 minutes at a load 2 2700 kW and '

e 2850 kW for the 2850 kW DG and 2 3875 kW and s 4075 kW for the 4075 kW DGs.

SR 3.8.1.4 Verify each day tank contains 2 900 gal of fuel oil for 31 days the 4075 kW DGs and 700 gal of fuel oil for the 2850 kW DG.

SR 3.8.1.5 Verify the fuel oil transfer system operates to transfer 31 days fuel oil from storage tank to the day tank.

SR 3.8.1.6 -geNO T E-E-_

All DG starts may be preceded by an engine prelube period.

Verify each DG starts from standby condition and 184 days achieves In: 12 seconds, voltage 2 3952 V and frequency 2 60 Hz.

Farley Units 1 and 2 3.8.1-7 Amendment No. 146 (Unit 1)

Amendment No. 137 (Unit 2)

AC Sources-Operating.

3.8.1 SURVEILLANCE REQUIREMENTS *_.

SURVEILLANCE FREQUENCY SR 3.8.1.7 NO-TD-This Surveillance shall not be performed In MODE I or 2.

Verify manual transfer of AC power sources from the 18 months normal offsfte circuit to the alternate required offsite drcuit.

SR 3.8.1.8 Verify each DG rejects a load greater than or equal to 18 months Its associated single largest post-accident load, and:

a. Following load rejection, the speed Is s 75% of the difference between nominal speed and the overspeed trip setpoint; and
b. Following load rejection, the voltage Is a 3740 V and s 4580 V.

Farley Units 1 and 2 3.8.1-8 Amendment No. 146 (Unit 1)

Amendment No. 137 (Unit 2)

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