Text

Amends 136 & 128 to Licenses NPF-2 & NPF-8,respectively, Revising pressure-temp Limit Heatup & Cooldown & Hydrostatic Testing Curves
	ML20216C165
Person / Time
Site:	Farley
Issue date:	04/09/1998
From:	Berkow H; NRC (Affiliation Not Assigned)
To:	;
Shared Package
ML20216C169	List: ML20216C165;
References
	NUDOCS 9804140354
	Download: ML20216C165 (48)
	v • d • e

I i

GAR Ef?

I k

o UNITED STATES s

g NUCLEAR REGULATORY COMMISSION 2

WASHINGTON, D.C. 30806 4 001

\\...../

i SOUTHERN NUCLEAR OPERATING COMPANY. INC.

ALABAMA POWER COMPANY DOCKET NO. 50-348 i

JOSEPH M. FARI EY NUCLEAR PLANT. UNIT 1

- AMENDMENT TO FACILITY OPERATING LICENSE I

Amendment No. 136 License No. NPF-2 l

1.

The Nuclear Regulatory Commission (the Commission) has found that-I A.

The application for amendment by Southern Nuclear Operating Company, Mc.

(SNC), dated July 23,1997, as supplemented September 30, October 27, and December 18,1997, and Febraury 12,1998, 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; 1

B.

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

C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the healtn 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 j

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 i

Operating License No. NPF-2 is hereby amended to read as follows:

90041400J4 980409 F

PDR ADOCK 05000348 P

PDR 7

i p.

l 4 i

(2)

Technical Soecifications l

The Technical Specifications contained in Appendices A and B, as revised through Amendment No.136, are hereby incorporated in the license. Southern l

Nuclear shall operate the facility in accordance with the Technical Specifications.

3.

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

' FOR THE NUCLEAR REGULATORY COMMISSION O

I

&j.

e rt N. Berkow, irector roject Directorate 11-2 4

l Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation

Attachment:

l Changes to the Technical Specifications l

Date of issuance: April 9, 1998 l

t l

l l

l

I l

l l

l ATTACHMENT TO LICENSE AMENDMENT NO.136 TO FACILITY OPERATING LICENSE NO. NPF-2 DOCKET NO. 50-348 l

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

1 Remove Insert I

l XIX XIX 1-5 1-5 3/44-3 3/4 4-3 3/4 4 4 a 3/4 44a 3/4 4-27 3/4 4-27 3/4 4-29 3/4 4-29 3/4 4-30 3/4 4-30 3/4 4-32 3/4 4-32 i

B 3/4 4-1 B 3/4 4-1 B 3/4 44 B 3/4 44 B 3/4 4-7 8 3/4 4-7 B 3/4 4-8 8 3/4 4-8 l

B 3/4 4-9 B 3/4 4-9.

B 3/4 4-10 B 3/4 4-10 l

B 3/4 4-10a B 3/4 4-10s B 3/4 4-11 B 3/4 4-11 B 3/4 4-12 B 3/4 4-12 B 3/4 4-13 B 3/4 4-13 B 3/4 4-14 83/44-14 6-20 6-20 l

6-20a I

INDEX DEFINITIONS SECTION PAGE i

I 1.0 DEFINITIONS l

1.1 ACTION.......................................................... 1-1 1.2 AXIAL FLUX DIFFERENCE...........................................

1-1

{

1.3 CHANNEL CALIBRATION............................................. 1-1 l

1.4 CHANNEL CHECK................................................... 1-1 1.5 CHANNEL FUNCTION TEST........................................... 1-1 1.6 CONTAINMENT INTEGRITY........................................... 1-2 j

1.7 CONTROLLED LEAKAGE..............................................

1-2 1.8 CORE ALTERATION.................................................. 1-2 l

1.Ba CORE OPERATING LIMITS REPORT.................................... 1-2 1.9 DOSE EQUIVALENT I-131........................................... 1-2 1.10 E -AVERAGE DISINTEGRATION ENERGY................................ 1-3 1.11 ENGINEERED SAFETY FEATURES RESPONSE TIME........................ 1-3 1.12 FREQUENCY NOTATION.............................................. 1-3 1.13 (Deleted)..................... 1-3 1.14 I DENTI FI ED LEAKAG E..............................................

1-3 1.15 I.I^"!D P.'2?'.'2T: TPI.'/""~ "' :YS T""

( Del e t ed )......................

1-4 1.16. J00 C".'?:CES TO P_'2!^.'.0"'!'1" "'lT" TT".'."'""""'

SVGINiiMs ( Del e t ed )...............................................

1-4 l

1.17 OFFSITE DOSE CALCULATION MANUAL (CDCM).......................... 1-4 1.18 OPERABLE - OPERABILITY.......................................... 1-4 1.19 OPERATIONAL MODE - MODE......................................... 1-5 1.20 PHYSICS TESTS................................................... 1-5 1.21 PRESSURE BOUNDARY LEAKAGE.......................................

1-5 l

1.21a PRESSURE TEMPERATURE LIMITS REPORT (PTLR).......................

1-5 l 1.22 PROCES S CONTROL PROGRAM ( PCP )................................... 1-5 i

1.23 PURGE-PURGING................................................... 1-5 1.24 QUADRANT POWER TILT RATIO....................................... 1-5 I

i 1.25 RATED THERMAL POWER.............................................

1-6 1.26 REACTOR TRIP SYSTEM RESPONSE TIME............................... 1-6 j

1.27 REPORTABLE EVENT................................................ 1-6 1.28 SHUTDOWN MARGIN................................................. 1-6 l

t 1.29 GOMGEM4AMGN (Deleted)........................................ 1-6 1.30 SOURCE CHECK.................................................... 1-6 1.31 STAGGERED TEST BASIS............................................ 1-6 l

1.32 THERMAL POWER................................................... 1-7 1.33 UNI DENTI FI ED LEAKAGE............................................ 1-7 l

1.34 VENTILATION EXHAUST TREATMENT SYSTEM............................

1-7 1.35 VENTING......................................................... 1-7 TABLE 1.1 OPERATIONAL MODES........................................ 1-8

- TABLE 1.2 FREQUENCY NOTATION....................................... 1-9 r

FARLEY-UNIT 1 I

AMENDMENT NO.W,W,W, f#,136 I

i-L

INDEX ADMINISTRATIVE CONTROLS I

I SECTION PAGE Review..............................................

6-10 Audits.............................................. 6-11

)

Authority........................................... 6-12 Records.............................................

6-12 6.5.3 TECHNICAL REVIEW AND CONTROL Activities.......................................... 6-12 Records...............J....~.........................

6-13 6.6 REPORTABLE EVENT ACTION.................................... 6-14 6.7 SAFETY LIMIT VIOLATION.....................................

6-14 6.8 PROCEDURES AND PROGRAMS....................................

6-14 6.9 REPORTING REQUIREMENTS 6.9.1 ROUTINE REPORTS l

Startup Report......................................

6-15a Annual Report....................................... 6-16 Annual Radiological Environmental Operating Report.............................................. 6-17 i

Annual Radioactive Effluent Release j

Report.............................................. 6-17 Monthly Operating Report............................ 6-19 Core Operating Limits Report........................ 6-19 l

Annual Diesel Generator Reliability Data Report.............................................. 6-19a Annual Reactor Coolant System Specific Activity Report.....................................

6-20 Annual Sealed Source Leakage Report................. 6-20 Pressure Temperature Limits Report (PTLR)............

6-20a i

6.9.2 SPECIAL REPORTS.....................................

6-20a 6.10 RECORD RETENTION.......................................... 6-20a 6.11 RADIATION PROTECTION PROGRAM..............................

6-21a

'6.12 HIGH RADIATION AREA.......................................

6-22 L

f FARLEY-UNIT 1 XIX AMENDM NO

1 DEFINITIONS OPERATIONAL MODE - MODE 1.19 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level and average reactor coolant temperature specified in Table 1.1.

PHYSICS TESTS 1.20 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and

1) described in Chapter 14.0 of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission.

PRESSURE BOUNDARY LEAKAGE 1.21 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube leakage) through a non-isolable fault in a Reactor Coolant System Component body, pipe wall or vessel wall.

PRESSURE TEMPERATURE LIMITS REPORT (PTLR) 1.21a The PRESSURE TEMPERATURE LIMITS REPORT (PTLR) is the unit specific document that provides the reactor vessel pressure and temperature (P-T) limits, including heatup and cooldown rates, for the current reactor vessel fluence period. These P-T limits shall be determined for each fluence period or effective full-power years (EFPYs) in accordance with Specification 6.9.1.15.

Plant operation within these operating limits is addressed in LCO 3.4.10.1, RCS Pressure / Temperature Limits.

PROCESS CONTROL PROGRAM (PCP) 1.22 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, tests, and determinations to be made to ensure that 4

processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71; State regulations; burial ground requirements; and other requirements governing the disposal of solid radioactive waste.

PURGE - PURGING 1.23 PURGE or PURGING is the controlled process of discharging air or gas from a confinement to naintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is i

required to purify the confinement.

QUADRANT POWER TILT RATIO 1.24 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector cali-brated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater. With one excore detector inoperable, the remaining three detectors shall be used for computing the average.

FARLEY-UNIT 1 1-5 AMENDMENT NO.[M,hB/,136 i

i i

l l

REACTOR COOLANT SYSTEM HOT SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.1.3 a.

At least two of the Reactor Coolant and/or residual heat removal (RHR) loops listed below shall be OPERABLE:

I 1.

Reactor Coolant Loop A and its associated steam generator and Reactor Coolant pump,*

I 2.

Reactor Coolant Loop B and its associated steam generator and Reactor Coolant pump,*

3.

Reactor Coolant Loop C and its associated steam I

^

generator and Rea' tor' Coolant pump,*

~

' ~~

~

c 4.

Residual Heat Removal Loop A, 5.

Residual Heat Removal Loop B.

b.

At least one of the above Reactor Coolant and/or RHR loops shall be in operation.**

APPLICABILITY:

MODE 4.

ACTION:

a.

With less than the above required Reactor Coolant and/or RHR loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible.

b.

With no Reactor-Coolant or RHR loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required coolant loop to operation.'

A Reactor Coolant pump shall not be started with one or more of the Reactor Coolant System cold leg temperatures less than or equal to 325'F l unless 1) the pressurizer water volume is less than 770 cubic feet (24%

of wide range, cold, pressurizer level indication) or 2) the secondary water temperature of each steam generator is less than 50*F above each of the Reactor Coolant System cold leg temperatures.

- All Reactor Coolant pumps and residual heat removal pumps may be de-energized for'up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months provided 1) no operations are permitted that

>would cause dilution of the Reactor Coolant System boron concentration, l

and 2) core outlet temperature is maintained at least 10*F below saturation temperature.

FARLEY-UNIT 1 3/4 4-3 AMENDMENT NO.#1/,A36

REACTOR COOLANT SYSTEM COLD SHUTDOWN LIMITING CONDITION FOR OPERATION Two# residual heat removal (RHR) loops shall be OPERABLE

i 3.4.1.4 a.

and at least one RHR loop shall be in operation.**

I APPLICABILITY:

MODE 5.## ###

l l

ACTION:

a.

With less than the above required RHR/ Reactor Coolant loops OPERABLE, immediately initiate corrective action to return the required RHR/ Reactor coolant loops to OPERABLE 1

-status'as soon as~possible.

b.

With no coolant loop in operation, suspend all operations invo'lving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required coolant loop to operation.

SURVEILLANCE REQUIREMENTS 4.4.1.4 At least one RHR loop shall be determined to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

The normal or emergency power source nay be inoperable in MODE 5.

The RHR loop may be removed from operation for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months provided (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration,. and (2) core outlet temperature is maintained et least 10' F below saturation temperature.

Three filled Reactor Coolant. oops and at least two steam generators having levels greater than or ' qual to 10% of wide range indication may be substituted for one RHR loop.

1. A Reactor Coolant pump shall not be started with one or more of the Reactor Coolant System cold leg temp'ratures less than or equal to 325'F unless (1) the pressuriter wate. volume is less than 770 cubic l

feet (24% of wide range, cold, pressurater level indication) or (2) the secondary water temperature of each stes t generator is less than 50' F above each of the Reactor Coolant Systra cold leg temperatures.

N# The number of operating Reactor Coc1.nt pumps is limited to one at RCS temperatures less than 110*F with the exception that a second pump nay be started for the purpose of maintaining continuous flow while taking the operating pump out of service.

PARLEY-UNIT 1 3/4 4-4a AMENDMENT NO.26,136 i

~.

REACTOR COOLANT SYSTEM 3/4.4.10 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.10.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limits specified in the

-g PRESSURE TEMPERATURE LIMITS REPORT (PTLR) during heatup, cooldown, i

criticality, and inservice leak and hydrostatic' testing.

APPLICABILITY: At all times.

ACTION:

I With any of the above limits specified in the PTLR exceeded, restore the l

temperature and/or pressure to within the limit within 30 minutes; perform an i

engineering evaluation or inspection to determine the effects of the out-of-

l limit condition on the fracture toughness of the Reactor Pressure Vessel; j

determine that the Reactor Pressure Vessel remains acceptable for continued operation or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce the RCS T and pressure to less than 200*F and 500 psig, respectively, within av thefol$owing30 hours.

SURVEILLANCE REQUIREMENTS 4.4.10.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits specified in the PTLR at Isast once per l

hour during system heatup, cooldown, and inservice leak and hydrostatic testing' operations.

4.4.10.1.2 The reactor vessel material irradiation surveillance specimens

< shall be removed and examined, to determine changes in material properties, as required by 10CFR50, Appendix H.

I FARLEY-UNIT 1 3/4 4-27 AMENDMENTNO.f6//A,136

li-t I,

l l

THIS PAGE.INTENTIbNALLY LEFT BLANK l

I s.

e j

l

\\

l r

d t

i l

l l

l i

l l

1 i

.FARLEY - UNIT 1 3/4 4-29 AMENDMENT NO.W,M4136

. t

)'.

'1

f 1

.(

g THIS PAGE INTENTIONALLY LEFT BIANK i-i l

r l

<~ >

p i

t l-1 I

l i

i i

i I

i i

I l

i l

t l-i I

L l

l i

l l

i-i s.

\\

l i

i I

l-FARLEY - UNIT 1 3/4 4-30 AMENDMENTNo.58//fl136 l

l'

REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITIONS FOR OPERATION 3.4.10.3 At least one of the following overpressure protection systems shall be OPERABLE:

a. Two RHR relief valves with:

i 1.

A lift setting of less than or equal to 450 psig, and 2.

The associated RHR relief valve isolation valves open; or

b. The Reactor Coolant, System,(RCS).depressurized with an RCS vent of greater than or equal to 2.85 square inches.

APPLICABILITY: When the temperature of one or'more of the RCS cold legs is less than or equal to 325'F, except when the reactor vessel head is l

removed.

ACTION:

a. With one RHR relief valve inoperable, restore the inoperable valve to j

OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or perform the following:

l 1.

Establish the following requirements:

1.

Reduce pressurizer level to less than or equal to 30 j

percent (cold calibrated), and 11.

Assign a dedicated operator for RCS pressure monitoring and control, and 111.

Restore the inoperable valve to OPERABLE status within 7 days, or; 1

i 2.

Depressurize and vent the RCS through a greater than or equal to 2.85 square inch vent within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

b. With both RHR relief valves inoperable, within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months either:

1.

Restore at least one RHR relief valve to OPERABLE status, or 2.

Depressurize and vent the RCS through a greater than or equal to 2.85 square inch vent.

c. In the event a RHR relief valve or a RCS vent is used to mitigate a RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days.

The report shall describe the circumstances initiating the transient, the effect of the RHR relief valves or vent on the transient and any corrective action necessary to prevent recurrence.

d. The provisions of Specification 3.0.4 are not applicable.

I FARLEY-UNIT 1 3/4 4-32 AMENDMENTNO.II[Id8[

136

3/4.4 REACTOR COOL 1.NT SYSTEM BASES 3/4.4.1

' CACTOR COOLANT LOOPS AND COOLANT CIRCULATION T>s plant is designed to operate with all Reactor Coolant Loops in opere*aon, and meet the DNB design criterion during all normal operations 2nd anticipated transients.

In MODES 1 and 2 with one Reactor Coolant Loop not in operation this specification requires that the plant be in at least HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

In MODE 3, two Reactor Coolant Loops provide sufficient heat removal capability for' removing core heat even in the event of a bank withdrawal accident; however, a. single Reactor Coolant Loop provides sufficient decay heat removal capacity. if. a bank withdrawal accident _can be prevented;.i.e.,

by opening the Reactor Trip Breakers or shutting down the rod drive motor / generator sets.

In MODE 4, a single reactor coolant or RHR loop provides sufficient heat removal capability for removing decay heat, but single failure considerations require that at least two loops be OPERABLE. Thus, if the Reactor Coolant Loops are not OPERABLE, this specification requires two RHR loops to be OPERABLE.

In MODE 5, single f ailure considerations require two RHR loops to be OPERABLE.

The operation of one Reactor Coolant Pump or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control.

The restrictions on starting a Reactor Coolant Pump with one or more Reactor Coolant System cold legs less than or equal to 325'r are provided l

to prevent Reactor Coolant System pressure transients, caused by energy additions from the secondary system, which could exceed the limits of Appendix G to 10 CFR Part 50.

The Reactor Coolant System will be protected against overpressure transients and will not exceed the limits of Appendix G by either (1) restricting the water volume in the pressurizer and thereby providing a volume for the primary coolant to expand into, or (2) by restricting starting of the Reactor Coolant Pumps to when the secondary water temperature of each steam generator is less than 50*F above each of the Reactor Coolant System cold leg temperatures.

FARLEY-UNIT 1 B 3/4 4-1 AMENDMENT NO. [l/,laf/,/M/,

136

REACTOR COOLANT SYSTEM BASES Reducing T to less than 500*F prevents the release of activity should a avg steam generator tube rupture since the saturation pressure of the primary coolant is below the lift pressure of the atmospheric steam relief valves.

The surveillance requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take corrective action.

Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena. A reduction in frequency of isotopic analyses following power changes may be perndssible if justified by the data obtained.

3/4.4.10 PRESSURE / TEMPERATURE LIMITS _

The temperature and pressure changes during heatup and cooldown are limited to be consistent with the requirements given in the ASME Boiler and Pressure vessel Code,Section XI, Appendix G, asrequiredper10CFRPart50,AppendixG.l 1)

The reactor coolant temperature and pressure and system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with the PRESSURE TEMPERATURE LIMITS REPORT (PTLR).

l s) Allowable combinations of pressure and temperature for specific temperature change rates r.re below and to the right of the limit lines shown in the PTLR.

Limit lines for cooldown rates between l those presented may be obtained by interpolation.

b) The PTLR defines limits to assure prevention of nonductile l

failure only.

For normal operation, other inherent plant characteristics, e.g.,

pump heat addition and pressurizer heater capacity, may limit the heatup and cooldown rates that can be achieved over certain pressure-temperaturs ranges.

2)

These limit lines shall be calculated periodically using methods approved by the NRC.

L 3)

The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the steam generator is below 70*F.

FARLEY-UNIT 1 B 3/4 4-6 AMENDMENT NO. N/7f/136

REACTOR COOLANT SYSTEM E'.SES 4)

The pressurizer heatup and cooldown rates shall not exceed 100*F/hr and 200*F/hr respectively. The spray shall not be used if the temperature difference betwwen the pressurizer and the spray fluid is grester than 320*T.

5)

System preservice hydrotests and in-service leak and hydrotests shall bo performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code,Section XI.

Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RTndt, at the end of 21.9 effective full power years (EFPY) of service life. The 21.9 EFPY service life r riod is chosen such that the limiting RTndt at the 1/4T location in the core region is greater than the RTndt of the limiting unirradiated material. The selection of such a limiting RTndt assures that all components in the Reactor Coolant System will be operate.* conservatively in accordance with applicable Code requirements.

1 FARLEY-UNIT 1 B 3/4 4-7 AMENDMENT NO. $8M,136

p I

j

_ REACTOR COOLANT SYSTEM BASES l

l Values of ARTndt detennined in accordance with the NRC-approved nethodology l may be used until the next results from the material surveillance program, evaluated according to ASTM E185-82, are available. Capsules will be removed in accordance with the requirements of ASTM E185-82 and 10 CPR 50, App.ndix H.

The surveillance specimen withdrawal schedule in shown in the l

PTLR. The heatup and cooldown curves must be recalculated when the ART ndt determined from the surveillance capsule exceeds the calculated ARTndt f0f the equivalent capsule radiation exposure.

Allowable pressure-temperature relations ~arps for various heatup and cooldown rates are calculated using methods derived from Appendix G in Section XI of the ASME Boiler and Pressure Vessel Cede as required by l

Appendix G to 10 CFR Part 50 and these methods are discussed in detail in WCAP-14040-NP-A, Revision 2, and the NRC letters dated March 31, 1998 and April 3, 1998.

I Althcugh the pressurizer operates in temperature ranges above those for which there is reason for concern of non-ductile failure, operating lindts are provided to assure compatibility of operation with the fatigue shaly.=is performed in accordance with the ASME Code requirements.

The OPERABILITY of either RHR relief valve or an RCS vent opening of greater than or equal to 2.85 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS legs are less than or equal to 325'F.

Either RHR relief valve has adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either (1) the start of an idle RCP with the secondary water temperature of the steam generator less than or equal to 50*F above the RCS cold leg temperatures provided measures are taken to cushion the overpressure effects at RCS temperatures above 250*F, or (2) the start of all operable charging pumps and their injection into a water solid RCS.

In j

the case of the injection by the charging pumps, the analysis is based on the start of the maximum number of operable charging pumps allowed by the Technical Specifications.

i l

FARLEY-UNIT 1 B 3/4 4-8 AMENDMENT No. EW,#N,136

i i

i l

l l

l w-3 m

L'wa A

d 5-B g

s M

E.

M M

E FARLEY-UNIT 1 3 3/4 4-9 AMENDMENT No.MI/7Y,136 f

a l

[

l l

l t

l' 1

THIS PAGE INTENTIONALLY LErr BIANK l

I 4

1

~"IT 1 a 3/4 4-10 AMENDMENT NO.8//f,136

THIS PAGE INTENTIONALLY LEFT BLANK i

)

1 1

I t

)

1 1

i i

I i

i

{

FARLEY-UNIT 1~

B 3/4 4-10a AMENDMENT No. 71/136 iI'

r i

REACTOR COOLANT SYSTEM BASES THIS PAGE INTENTIONALLY LEFT BLANK l

o i

1 1

i

]

FARLEY-UNIT 1 B 3/4 4-11 AMENDMENT NO//f//136

REACTOR COOLANT SYSTEM BASES i

THIS PAGE INTENTIONALLY LEFT BIAtX

)

i J

j i

i 1

)

J l

l l

FARLEY-UNIT 1 B 3/4 4-12 AMENDMENTNO.$M136

i REACTOR COOLANT SYSTEM basts THIS PAGE INTENTIONALLY LEFT BLANK i

l l

i l

l i

i i

l TARLEY-UNIT 1 B 3/4 4-13 AMENDMENT No. $136

I REACTOR COOLANT SYSTEM I

RASES I

i l

i i

3/4.4.11 STRUCTURAL INTEGRITY i

J

-The inservice inspection and testing programs for ASME Code Class 1, 2 and 3 j

components ensure that the structural-integrity and operational readiness of these components will be maintained at an acceptable level throughout the life of the plant. These programs are in accordance with section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50.55a(g) except where specific written relief has been granted by the I

commission pursuant to 10 CFR Part 50.55a(g) (6) (i).

I J

AMENoxENTx0.t//#/gg, rARtEr-UNIT 1 a 3/4 4-14

l l

l

-ADMINISTRATIVE CONTROLS ANNUAL REACTOR COOLANT SYSTEM SPECIFIC ACTIVITY REPORT 6.9.1.13 This annual report is only required when the results of specific activity analyses of the primary coolant have exceeded the limits of Specification 3.4.9 during the year. The following information shall be

. included:

(1) Reactor power history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first l

sample in which the limit was exceeded (in graphic and tabular format);

l (2) Results of the last isotopic analysis for radiciodine performed prior to exceeding.the. limit, results of analysis while limit was exceeded and results of one analysis after the radiciodine activity was reduced to less than the limit. Each result should include date 2nd time of sampling and the radioiodine concentrations; (3) Clean-up flow history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior I

to the first sample,in which the limit,was exceeded;. 4), Graph of the I-131

(

concentration (micro ci/gm) and one other radioiodine isotop, concentration l

(micro Ci/gm) as a function of time for the duration of the specific activity above the steady-state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radioiodine limit.

ANNUAL SEALED SOURCE LEAKAGE REPORT 6.9.1.14 A report shall be prepared and submitted to the commission on an annual basis if sealed source or fission detector leakage tests reveal the 1

presence of greater than or equal to 0.005 microcuries of removable contamination.

i 1

FARLEY-UNIT 1 6-20 AMENDMENTNO.8/[8I,136

ADMINISTRATIVE CONTROLS PRESSURE TEMPERATURE LIMITS REPORT (PTLR) 6.9.1.15 The reactor coolant system pressure and temperature limits, including heatup and cooldown rates, shall be established and documented in the PTLR for LCO 3.4.10.1.

The analytical methods used to determine the RCS pressure and temperatute limits shall be those previously reviewed and approved by the NRC, specifically those described in the NRC letters dated March 31, 1998 and April 3, 1998.

The PTLR shall be provided to the NRC upon issuance for each reactor fluence period and for any revision or supplement thereto.

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the commission in accordance with the requirements of 10CFR50.4 within the time period specified for each report.

Reports should be submitted to the U. S. Nuclear Regulatory Commission, ATTN:

Document Control Desk, Washington, D.C.

20555.

6.10 RECORD RETENTION i

In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated.

6.10.1 The following records shall be retained for at least five years:

a.

Records and logs of unit operation covering time interval at each power level.

b.

Records and logs of principal maintenance activities, inspections, repair and replacement of principal itens of equipment related to nuclear safety.

c.

ALL REPORTABLE EVENTS submitted to the Commission.

d.

Records of surveillance activities, inspections and calibrations required by these Technical Specifications, Records of changes made to the procedures required by e.

Specification 6.8.1.

f.

Records of radioactive shipments.

g.

Records of sealed source and fission detector leak tests and results.

FARLEY-UNIT 1 6-20a AMENDMENT NO.136 l

fL p

,,m urg\\

UNITED STATES g

w B

NUCLEAR REGULATORY COMMISSION U

[

WASHINGTON, D.C. 30006 4001

\\,,*****/

SOUTHERN NUCI FAR OPERATING COMPANY. INC.

ALABAMA POWER COMPANY DOCKET NO. 50-364 i

JOSEPH M. FARI FY NUCI FAR PLANT. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE - -

l l

Amendment No. 128 l

License No. NPF-8 i

l 1.

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

l:

A.

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

l (SNC), dated July 23,1997, as supplemented Seotember 30, October 27 and i

l December 18,1997, and Febraury 12,1998, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the l

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 l

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.

l 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 amended to read as follows:

l r

2 (2)

Technical Soecifications l

The Technical Specifications contained in Appendices A and B, as revised through Amendment No.128, are hereby incorporated in the license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications.

3.

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

~ FOR THE NUCLEAR REGULATORY COMMISSION h

5 /7 rbert N. Berkow, Director Project Directorate ll-2 Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of issuance:

April 9, 1998

i ATTACHMENT TO LICENSE AMENDMENT NO.128 TO FACILITY OPERATING LICENSE NO. NPF-8 DOCKET NO. 50-364 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised areas are indicated by marginallines.

Remove IDaed I.

l XIX XIX t-5 1-5 3/44-3 3/44-3 3/4 44a 3/4 4 4 a 3/4 4-27 3/4 4-27 l

3/4 4-29 3/4 4-29 i

3/4'4-30 3/4 4-30 3/4 4-32 3/4 4-32 B 3/4 4-1 B 3/4 4-1 B 3/4 4-6 B 3/4 4-6 B 3/4 4-7 B 3/4 4-7 B 3/4 4-8 8 3/4 4-8 8 3/4 4-9 B 3/4 4-9 83/44-11 B 3/4 4-11 B 3/4 4-12 B 3/4 4-12 j

B 3/4 4-13 B 3/4 4-13 B 3/4 4-14 B 3/4 4-14 6-20 6-20 6-20a

r l

INDEX l

DEFINITIONS I

l SECTION PAGE 1.0 DEFINITIONS 1.1 ACTION.......................................................... 1-1 1.2 AXIAL FLUX DIFFERENCE...........................................

1-1 1.3 CHANNEL CALIBRATION.............................................

1-1 1.4 CHANN EL CHECK...................................................

1-1 1.5 CHANNEL FUNCTION TEST...........................................

1-1 l

1.6 CONTAINMENT INTEGRITY...........................................

1-2 1.7 CONTROLLED LEAKAGE..............................................

1-2 1.8 CORE ALTERATION................................................. 1-2

'1.Ba CORE OPERATING' LIMITS REPORT.................................... 1-2 ~

1.9 DOSE EQUIVALENT I-131...........................................

1-2 1.10 E -AVERAGE DISINTEGRATION ENERGY................................ 1-3 1.11 ENGINEERED SAFETY FEATURES RESPONSE TIME........................ 1-3 1.12 FREQUENCY NOTATION.............................................. 1-3 1.13 (Deleted)..................... 1-3 1.14 I DENTI FI ED LEAKAG E......................................... ~.....

1-3 1.15 I.! O"! O r.'.0W'2TS TP2.*.T."2"T SY&T-EM ( Delet ed )......................

1-4 1.16.

20."

C."_'."C E: T O "J." ^'.c TI'1: ".'2 T : T " " ' '"""2'T GVG9ENG (Deleted)............................................... 1-4 l

1.17 OFFSITE DOSE CALCULATION MANUAL (ODCM).......................... 1-4 1.18 O P ERAB LE - O P ERABI LI TY..........................................

1-4 1.19 OPERATIONAL MODE - MODE.........................................

1-5 1.20 PHYSICS TESTS................................................... 1-5 1.21 PRESSURE BOUNDARY LEAKAGE.......................................

1-5 1.21a PRESSURE TEMPERATURE LIMITS REPORT ( PTLR )........................ 1-5 l 1.22 PROCESS CONTROL PROGRAM ( PCP )................................... 1-5 1.23 PU RG E-PU RG I N G................................................... 1 - 5 1.24 QUADRANT POWER TILT RATIO....................................... 1-5 1.25 RATED THERMAL POWER.............................................

1-6 1.26 REACTOR TRI P SYSTEM RES PONSE TIME............................... 1-6 I

1.27 REPORTABLE EVENT................................................ 1-6 1.28 SHUTDOWN MARGIN................................................. 1-6 1

1.29 GGI49HEAHON ( Del e t ed )........................................

1-6 1.30 S OU RCE CH EC K....................................................

1-6 1.31 STAGGERED TEST BASIS............................................ 1-6 1.32 THERMAL' POWER................................................... 1-7 1.33 UN I DENTI FI ED LEAKAGE............................................ 1-7 1.34 VENTILATION EXHAUST TREATMENT SYSTEM............................

1-7 1.35 VENTING......................................................... 1-7 TABLE 1.1 OPERATIONAL MODES...........................................

1-8 TABLE 1.2 FREQUENCY NOTATION..........................................

1-9 1

i FARLEY-UNIT 2 I

AMENDMENTNO/$[$I/Ild 12P

)

l INDEX ADMINISTRA.IVE CONTROLS

[

SECTION PAGE l

l Review.................................................. 6-10

\\

Audits.................................................. 6-11 l

Authority...............................................

6-12 Records................................................. 6-12 6.5.3 TECHNICAL REVIEW AND CONTROL I

Activities.............................................. 6-12

..Re c o r d s.................................................. 6 - 13 6.6 REPORTABLE EVENT ACTION........................................ 6-14 6.7 SAFETY LIMIT VIOLATION......................................... 6-14 6.8 P ROCEDU RES AND P ROG RAMS........................................

6-14 69 REPORTING REQUIREMENTS 6.9.1 ROUTINE REPORTS startup Report.......................................... 6-15a Annual Report........................................... 6-16 Annual Radiological Environmental Operating Report...... 6-17 Annual Radioactive Ef fluent Release Report.............. 6-17 i

i Monthly Operating Report................................

6-19 i

Core Operating Limits Report............................ 6-19 j

Annual Diesel Generator R**.iability Data Report......... 6-19a l

Annual Reactor Coolant System Specific Activity Report.. 6-20 Annual Sealed Source Leakage Report..................... 6-20 Pressure Temperature Limits Report ( PTLR)...............

6-2 0 a 6.9.2 S PECIAL RE PO RTS................................ -.............. 6-2 0 a l

l 6.10 RECO RD RET ENTI ON..............................................

6-2 0 a 6.11 RADIATION PROTECTION PROGRAM.................................. 6-21a f 12 HIGH RADIATION AREA........................................... 6-22 FARLEY-UNIT 2 XIX AMENDMENT NO.If3/I25, 128

DEFINITIONS OPERATIONAL MODE - MODE 1.19 An OPERATIONAL NODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level and average reactor coolant temperature specified in Table 1.1.

PHYSICS TESTS 1.20 PHYSICS TESTS shall be those taste performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and 1) described in Chapter 14.0 of the FSA2, 2) authorized under the provisions of 10 CFR 50.59,'or 3) otherwise approved by the commission.

PRESSURE BOUl(DARY LEAKAGE 1.21 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube leakage) through a non-isolable fau3t in a heactor coolant System Component body, pipe wall or vessel wall.

PRESSURE TEMPERATURE LIMITS REPORT iPTLR) 1.21a The PRESSURE TEMPERATURE LIMITS REPORT (PTLR) is the unit specific document that provides the reactor vessel pressure and temperature (P-T)_

limits, including heatup and cooldown rates, for the current reactor vessel l

fluence period. These P-T limits shall be determined for each fluence period or effective full-power years (EFPYs) in accordance with Specification 6.9.1.15.

Plant operation within these operating limits is addressed in LCO i

3.4.10.1, RCS Pressure / Temperature Limits.

l PROCESS CONTROL PROGRAM fPCP)

)

1.22 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, tests, and date:rminations to be made t,o ensure that processing and packaging of solid radioactive wastes based on demonstrated j

processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71; State regulations; burial ground requiresaents; and other requirements governing the disposal of solid radioactive waste.

l l

l PURGE - PURGING l

i 1.23 PURGE and PURGING is the controlled process of discharging air or gas l

from a confinement to maintain tem;perature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

QUADRANT POWER TILT RATIO 1.24 QUADRANT POFER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated ~ output to the average of the upper excore detector cali-brated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater. With one excore detector inoperable, she remaining three detectors shall be used for computing the average.

FARLEY-UNIT 2 1-5 AMENDMENT NO.M/,128

T' REACTOR COOLANT SYSTEM HOT SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.1.3 a.

At least two of the Reactor Coolant and/or residual heat removal (RHR) loops listed below shall be OPERABLE:

l 1.

Reactor Coolant Loop A and its associated steam generator and Reactor Coolant pump, l

2.

Reactor Coolant Loop B and its associated steam generator and Reactor Coolant pump, l

3.

Reactor Coolant Loop C and its associated steam generator and Reactor Coolant pump, 4.

Residual Heat Removal Loop A, 5.

Residual Heat Removal Loop B.

b.

At least one of the above Reactor Coolant and/or RHR loops shall be in operation.

APPLICABILITY:

MODE 4.

ACTION:

a.

With less than the above required Reactor Coolant and/or RHR loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible.

b.

With no Reactor Coolant or RHR loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required coolant loop to operation.

A Reactor Coolant pump shall not be started with one or more of the Reactor Coolant System cold leg temperatures less than or equal to 325'F l unless 1) the pressurizer water volume is less than 770 cubic feet (24%

of wide range, cold, pressurizer level indication) or 2) the secondary water temperature of each steam generator is less than 50*F above each of the Reactor Coolant System cold leg temperatures.

- All Reactor Coolant pumps and residual heat removal pumps may be de-energized for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months provided 1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and 2) core outlet temperature is maintained at least 10*F below saturation temperature.

l l

l FARLEY-UNIT 2 3/4 4-3 AMENDMENT NO.128

REACTOR COOLANT SYSTEM COLD SHUTDOWN LIMITING CONDITION FOR OPERATION Two# residual heat removal (RHR) loops shall be OPERABLE 3.4.1.4 a.

and at least one RHR loop shall be in operation.

APPLICABILITY:

MODE 5.## ###

l ACTION:

a.

With less than the above required RHR/ Reactor Coolant loops OPERABLE, immediately initiate corrective action to return the required RHR/ Reactor coolant loops to OPERABLE

~

status as soon as possible. _

b.

With no coolant loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and inmediately initiate corrective action to return the required coolant loop to operation.

SURVEILLANCE REQUIREMENTS 4.4.1.4 At least one RHR loop shall be determined to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

The normal or emergency power source may be inoperable in MODE 5.

The RHR loop may be removed from operation for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months provided (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is naintained at least 10*F below saturation temperature.

Three filled Reactor Coolant loops and at least two steam generators having levels greater than or equal to 10% of wide range indication may be substituted for one RHR loop.

1. A Reactor Coolant pump shall not be started with one or more of the Reactor Coolant System cold leg temperatures less than or equal to 325*F unless (1) the pressurizer water volume is less than 770 cubic l

feet (24% of wide range, cold,.pressuriter level indication) or (2) the secondary water temperature of each steam generator is less than 50*F l

above each of the Reactor Coolant System cold leg temperatures.

l L

1. 1. The number of operatirag Reactor Coolant pumps is lindted to one at RCS

. temperatures ler;s than 110'r with the exception that a second pump may be started for the purpose of maintaining continuous flow while takirg the operating pump out of service.

FARLEY-UNIT 2 3/4 4-4a AMENDMENT NO. AA(128 l'

REACTOK COOLANT SYSTEM 3/4.4.10 PRESStHtE/ TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING = CONDITION FOR OPERATION 3.4.10.1 The Reactor Coolant System (except the pressurizer) temperature l-and pressure shall be limited in accordance with the limits specified in the PRESSURE TPMPERATURE LIMITS REPORT (PTLR) during heatup, cooldown, 1

criticElity, and inservice leak and hydrostatic testing.

d APPLICABILITY:

At all times.

ACTION:

-With any of the above limits specified in the PTLR exceeded, a setore the l

temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation or inspection to determine the effects of the out-of-limit condition on the fracture toughness of the Reactor Pressure Vessel; determine that the Reactor Pressure Vessel remains acceptable for continued operation or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce the RCS T and pressure to'less than 200*F and 500 peig, respectively, within avg the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

i SURVEILLANCE REQUIREMENTS i

4.4.10.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits specified in the PTLR at least once per l

hour during system heatup, cooldown, and inservice leak and hydrostatic testing operations.

(

4.4.10.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, as required by_10CFR50, Appendix H.

i t

FARLEY-UNIT.2 3/4 4-27 AMENDMENT No. /l4128

(-

4 i

l l.

THIs FAGE IPTEFFIop m y I,gry m 1

I l

1 l

l 1

i

+

e I

l

. '1 1

I i

l d

I

' l i

i i

i

-l

(

t I

i l

i

, j l

j..

l s

FARLEY-UNIT 2 3/4 4-29 AMENDMENT No.554M4128 l

l

t..

i l

r 1

THIS PAGE INTENTIONALI.Y LEFT BLANK l.

\\

i 1

FARLEY-WIT 2 3/4 4 30 AMENDMENT NO. TI/#f,128 3

I

l REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITIONS FOR OPERATIONS 3.4.10.3 At least one of the following overpressure protection systems I

shall be OPERABLE:

a. Two RHR relief valves with:

1.

A lift setting of less than or equal to 450 psig, and 2.

The associated RHR relief valve isolation valves opens or

b. The Reactor Coolant System (RCS) depressurized with an RCS vent of greater'than or~ equal to 2.85' square inches.

~

APPLICABILITY:

When the temperature of one or more of the RCS cold legs is less than or equal to 325'F, except.when the reactor vessel head is l

removed.

ACTION:

a. With one RHR relief valve inoperable, restore the inoperable valve to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or perform the following:

1.

Establish the following requirements:

1.

Reduce pressurizer level to less than or equal to 30 percent (cold calibrated), and 11.

Assign a dedicated operator for RCS pressure moni.scing and control, and 111.

Restore the inoperable valve to OPERABLE status within 7 days, or; 2.

Depressurize and vent the RCS through a greater than or equal to 2.85 square inch vent within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

b4 With both RHR relief valves inoperable, within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months either:

1.

Restore at least one RHR relief valve to OPERABLE status, or 2.

Depressurize and vent the RCS through a greater than or equal to 2.85 square inch vent,

c. In the event a RHR relief valve or a RCS vent is used to mitigate a RCS pressure transient, a Special Report shall be prepared and submitted l

l to'the Commission pursuant to Specification 6.9.2 within 30 days. The j

report shall describe the circumstances initiating the transient, the I

effect of the RHR relief valves or vent-on the transient and any corrective action necessary to prevent recurrence.

d. The provisions of Specification 3.0.4 are not applicable.

FARLEY-UNIT 2 3/4 4-32 AMENDMENT No.400/,128 t

l l

I l

l 3/4.4 REACTOR COOLANT SYSTEM l

BASES l'

3/4.4.1 REACTOR COOLANT LOOPS AND COOU.NT CIRCULATION The plant is designed to operate with all Reactor Coolant Loops in operation, and meet the DNB design criterion during all nornal operations and anticipated transients.

In MODES 1 and 2 with one Reactor Coolant Loop not in operation this specification requires that the plant be in at least HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

In MODE 3, two Reactor Coolant Loops provide sufficient heat removal capability for removing core heat even in the event of a bank withdrawal i

accident; however, a single Reactor Coolant Loop provides sufficient decay l

heat removal. capacity.if.a bank withdrawal accident-can be prevented; i.e.,

by opening the Reactor Trip Breakers or shutting down the-rod drive l

motor / generator sets.

In MODE 4, a single reactor coolant or RHR loop provides sufficient heat removal capability for removing decay heat, but single failure considerations require that at least two loops be OPERABLE. Thus, if the

,d Reactor Coolant Loops are not OPERABLE, this specification requires two RHR loops to be OPERABLE.

In MODE 5, single failure considerations require two RHR loops to be OPERABLE.

The operation of one Reactor Ceolant Pump or one RHR pump provides j

adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control.

The restrictions on starting a Reactor Coolant Pump with one or more Reactor Coolant System cold legs less than or equal to 325'F are provided l

to prevent Reactor Coolant System pressure transients, caused by energy additions from the secondary system, which could exceed the lindts of Appendix G to 10 CFR Part 50.

The Reactor Coolant System will be protected against overpressure transients and will not exceed the limits of Appendix G by either (1) restricting the water volume in the pressurizer and thereby providing a volume for the primary coolant to expand into, or (2) by restricting starting of the Reactor Coolant Pumps to when the secondary water temperature of each steam generator is less than 50*F above each of the Reactor Coolant System cold leg temperatures.

I l

FARLEY-UNIT 2 B 3/4 4-1 AMENDMENT NOM,M/,128 l

1 l

REACTOR COOLANT sysTr4 i

BASES e e +

Reducing T to less uhan 500'F prevents the release of activity should a a

steamgenerNortuberupturesincethesaturationpressureoftheprimary coolant is below the lift pressure of the atmospheric steam relief valves.

The surveillance muirements provide adequate assurance that excessive specific activity levele in the primary coolant will be detected in sufficient time to take corrective action. Information obtained on iodine spiking will be used to assess the parameters associated with spiking phsnomena. A reduction in frequency of isotopic analyses following power changes may be l

permissible if justified by the data obtained.

l l

3/4.4.10 PREESURE/ TEMPERATURE LIMITS The temperature and pressure changes during heatup and cooldown are limited to be consistent with the requirements given in the ASME Boiler and Pressure Vessel Code,Section XI, Appendix 0 as required per 10 CFR Part 50 Appendix G.

l I-1)

The reactor. coolant temperature and pressure and system heatup and cooldown rates (with the exception of the pressuriser) shall be limited in accordance with the PRESSURE TEMPERATURE LIMITS REPORT (PTLR).

l l

a)

Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown in the PTLR. Limit lines for cooldown l

rates between those presented may be obtained by interpolation.

b)

The PTLR defines limits to assure prevention of nonductile l

failure only. For normal operation, other inherent plant characteristics, e.g., pump heat addition and pressuriser heater capacity, may limit the heatup and cooldown rates i

that can be achieved over certain pressure-temperature IJ ranges.

2)

These limit lines shall be calculated periodically using methods approved by the NRC.

3)

The secondary side of the steam generator must not be pressurized above 200 peig if the temperature of the steam generator is below 70*F.

f 1

j FARLEY-UNIT 2 8 3/4 4-6 AMENDMENT NO.NT/128

l i

REACTOR COOLANT SYSTEM l

RASES l

4)

The pressurizer heatup and cooldown rates shall not exceed 100'F/hr and 200'F/hr respectively. The spray shall not be used if the temperature difference between the pressuriser and the spray fluid is greater than i

320'F.

5)

System preservice hydrotests and in-service leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code,Section XI.

I l

Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RTndt, at the end of 33.8 effective full power years (EFPY) of service life. The 33.8 EFPY service life period is chosen such that the limiting RTndt at the 1/4T location in the core region is greater than the RTndt of the limiting unirradiated material. The selection of j

such a limiting RTndt assures that all components in the Reactor Coolant System j

will be operated conservatively in accordance with applicable code requirements.

l i

1

)

l l

l FARLEY-UNIT 2 B 3/4 4-7 AMENDMENT NO. $M,M/,128

i l

l REACTOR COOLANT SYSTEM i

i l

RASES l

Values of ARTndt determined in accordance with the NRC-approved methodology, l

may be used until the next results from the material surveillance program, evaluated according to ASTM E185-82, are available. Capsules will be removed-in accordance with the requirements of ASTM E185-82 and 10 CFR 50, Appendix H.

The surveillance specimen withdrawal schedule is shown in the

(

l PTLR. The heatup and cooldown curves must be recalculated when the ARTndt l

determined from the next surveillance capsule exceeds the calculated ARTndt

.for the equivalent capsule radiation exposure.

Allowable pressure-temperature relationships for various heatup and cooldown rates are' calculated'using methods derived from Appendix G in Section'XI'of l

the ASME Boiler and Pressure vessel Code as required by Appendix G to 10 CFR 50 and these methods are discussed in detail in WCAP-14040-NP-A,-

Revision 2, and the NRC letters dated March 31, 1998 and April 3, 1998.

l 1

Although the pressurizer operates in temperature ranges above those for

_ hich there is reason for concern of non-ductile failure, operating limits w

i are provided to assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements.

The OPERABILITY of either RHR relief valve or an RCS vent opening of greater L

than or equal to 2.85 square inches ensures that the RCS will be protected j

from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 325'F.

Either RHR relief valve has adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either (1) the start of an idle RCP with the secondary water temperature of the steam generator less than or equal to 50*F above the RCS cold leg temperatures provided-I measures are taken to cushion the overpressure effects at RCS temperatures above 250'F, or (2) the start of all operable charging pumps and their injection into a water solid RCS.

In the-case of the injection by the charging pumps,' the analysis is based on the start of the maximum number of l

operable charging pumps allowed by the Technical Specifications.

l FARLEY-UNIT 2 B 3/4 4-8 AMFNDH5TNO8848Ad128

i l

l L

i i

i l

i-A t'

t -

3 1

E t

i M

\\

a l

m l

~

l l

l FARLEY-UNIT 2~

g 3/4 4_g AMENDMENT No.M428

REACTOR COOIANT SYSTEM BASES i

l THIS PAGE INTENTIOhALLY LEFT BLhMK l

FARLEY-UNIT 2 3 3/4 4-11 AMENDMENT NO.128 i

l REACTOR COOLANT SYSTEM BASES THIS PAGE INTENTIONALLY LEFT BLANK l

I l

l i

1 I

l l

1 i

FARLEY-UNIT 2 5 3/4 4-12 AMENDMENT NO.128

(

l I

REACTOR COOIANT SYSTEM 1

l SASES l

THIS PACE INTENTIONALLY LEFT BIANK t 4 FARLEY-UNIT 2 B 3/4 4-13 AMENDMENT NO.128 l

j

I REACTOR COOLANT SYSTEM f

BASES is==

l l

l 3/4.4.11 STRUCTURAL INTEGRITY l

l The inservice inspection and testing programs for ASME Code Class 1, 2 and 3 components ensure that the structural integrity and operational readiness of these components will be maintained at an acceptable level throughout the life I

of the plant. These programs are in accordance with section XI of the ASME Boiler and Pressure vessel Code and applicable Addenda as required by 10 CFR Part 50.55a(g) except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50.55a(g)(6)(1).

l l

FARLEY-UNIT 2 B 3/4 4-14 AMENDMENTNO.f I//

L

I ADMINISTRATIVE CONTROLE anwuaL m m aron e thwT SYSTEM SPECIFIC ACTIVITY REPORT 6.9.1.13 This annual report is only required when the results of specific activity analyses of the primary coolant have exceeded the limits of specification 3.4.9 during the year. The following information shall be included:

(1) Reactor power history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample j

in which the limit was exceeded (in graphic and tabular format); (2) Results of the last-isotopic analysis for radiciodine performed prior to exceeding the limit, results of analysis while limit was exceeded and results of one analysis after the radiciodine activity was reduced to less than the limit. Each result should include date and time of sampling and the radiciodine concentrations; (3) clean-up flow history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded; (4) Graph of the I-131 concentration (micro ci/ga) and one, other radiciodine isotops concentration (micro ci/gn) as a function of time for the duration of the specific activity above the steady-state level; and (5) The l

time duration when the specific activity of the primary coolant exceeded the radiciodine limit.

l ANNUAL SEALED SOURCE LEAKAGE REPORT l '

annual basis if sealed source or fission detector leakage tests reveal the 6.9.1.14 A report shall be prepared and submitted to the Commission on an presence of greater than or equal to 0.005 microcuries of removable L

contamination.

I.

l l

FARLEY-UNIT 2 6-20 AMENDMENT NO. $ 84/,128

(

ADMINISTRATIVE CONTROLS PRESSURE TEMPERATURE LIMITS REPORT (PTLR) 6.9.1.15 The reactor coolant system pressure and temperature limits, including heatup and cooldown rates, shall be established and documented in the PTLR for LCO 3.4.10.1.

The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the NRC letters dated March 31, 1998 and April 3, 1998.

The PTLR shall be provided to the NRC upon issucnce for each reactor fluence period and for any revision or supplement thereto.

SPECIAL REPORTS 6.9.2 Ppecial reports shall be submitted to the Commission in accordance with the requirements of 10CFR50.4 within the time period specified for each report.

Reports should be submitted to the U. S. Nuclear Regulatory Commission, ATTN:

Document Control Desk, Washington, D.C. 20555.

6.10 RECORD RETENTION In addition to the applicable record retention requirements of Title 10, code of Federal Regulations, the following records shall be retained for at least the minimum period indicated.

6.10.1 The following records shall be retained for at least five j

years:

1 a.

Records and logs of unit operation covering time interval at each power level.

b.

Records and logs of principal maintenance activities, inspections, repair and replacement of principal itens of equipment related to nuclear safety.

c.

All REPORTABLE EVENTS submitted to the Commission, d.

Records of surveillance activities, inspections and calibrations required by these Technical Specifications.

j e.

Records of changes made to the procedures required by j

specification 6.8.1.

I f.

Records of radioactive shipments.

l g.

Records of sealed source and fission detector leak tests and results.

I l

EARLEY-UNIT 2 6-20a AMENDMENT NO.128 l

1

!

ML20216C165

Text

Attachment:

Attachment:

Navigation menu

Search