Text

Amend 13 to License NPF-2,limiting Conditions for Operation & Surveillance Requirements of Containment Sys,Eccs & Reactivity Control Sys
	ML19207A872
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Site:	Farley
Issue date:	07/31/1979
From:	Schwencer A; Office of Nuclear Reactor Regulation
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9, UNITED STATES NUCLEAR REGULATORY COMMISSION y

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ALABAMA POWER COMPANY DOCKET NO. 50-348 JOSEPH M. FARLEY NUCLEAR PLANT, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.13 License No. NPF-2 1.

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

A.

The application for amendment by Alabama Power Company (the licensee) dated September 6,1978 (superceding your application of March 17,1977) supplemented by letters dated November 3, 9, 17, 1978 and January 4, March 21, and April 17, 1979, compl ies with the standards and requirements of the Atomic Energj 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 amer.dment 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.

7008220 M

. 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:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.13, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

3.

Delete license condition 2.C.(3)(b).

4.

This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION it'l/AW A. Schwencer, Chief Operating Reactors Branch #1 Division of Operating Reactors

Attachment:

Changes to the Technical Specifications Date of Issuance:

July 31,1979 781163

ATTACHMENT TO LICENSE AMENDMENT NO. 13 FACILITY OPERATING LICENSE N0. NPF-2 DOCKET NO. 50-348 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. Revised pages are identified by Amendment number and contain vertical lines indicating the area of change.

Corresponding overleaf pages are also provided to maintain document completeness.

Paaes V

3/4 1-9 3/4 4-2 3/4 4-3 3/4 4-30 (added) 3/4 4-31 (added) 3/4 4-32 (was 3/4 4-30) 3/4 4-33 (was 3/4 4-31)

B 3/4 1-3 B 3/4 1-4 (added)

B 3/4 4-1 B 3/4 4-2 B 3/4 4-3 B 3/4 4-11 781164

INDEX l

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMEN PAGE SECTION 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT 3/4 6-1 Co n t a i n me n t I n te g ri ty..................................

3/4 6-2 Co n t a i n me n t L e a k a g e....................................

3/4 6-4 Containment Air Locks..................................

3/4 G-5 Internal Pressure......................................

3/4 6-6 Air Temperature........................................

3/4 6-7 Containmen t Structural Integri ty.......................

3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS 3/4 6-11 Containmen t Spray Sys tems..............................

3/4 6-12 S p ray Ad di t i v e Sy s t em..................................

3/4 6-14 Conta i nmen t Co al i ng Sy stem.............................

3/4 6-15 3/4.6.3 CCMTAINMENT ISOLATION VALVES...........................

3/4.6.4 COMBUSTIBLE GAS CONTROL 3/4 6-20 Hy d r o g e n An a l y z e r s.....................................

3/4 6-21 El ectric Hydrogen Recombiners..........................

3/4 6-22 Reactor Cavity Hydrogen Dilution System...............

3/4 6-23 Hyd rogen Mi xi ng Sy s tem.................................

'iS11G5 FARLEY - UNIT 1 VI

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS Pne_

SECTION 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4 4-14 Leakage Detection Systems..............................

3/4 4-16 Op er a ti on al Le aka g e....................................

'/4 4-18 3/4.4.7 CHEMISTRY..............................................

3/4 4-21 3/4.4.8 SPECIFIC ACTIVITY......................................

3/4 4.9 PRESSURE / TEMPERATURE LIMITS 3/4 4-25 Rea ctor Cool a n t Sys tem.................................

3/4 4-29 Pressurizer............................................

3/4 4-30 Overpressure Protection Systems.......

3/4.4.10 STRUCTlJRAL INTEGRITY...................................

3/4 4-32 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4 5-1 3/4.5.1 ACCUMULATORS.......................................

3/4 5-3 3/4,5.2 ECCS SUBSYSTEMS - T,yg >_350 F.........................

3/4 5-6 3/4.5.3 ECCS SUBSYSTEMS - T,yg < 350 F.........................

3/4.5.4 BORON INJECTION SYSTEM 3/4 5-8 Boron Injection Tank...................................

3/4 5-9 Heat Tracing...........................................

3/4 5-10 3/4.5.5 REFUELING W ATER STORAGE TANK...........................

h

/811 4 FARLEY - UNIT 1 V

Amendment No.13

REACTIVITY CONTROL SYSTEMS CHARGING PUMP - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.3 At least one charging pump in the boron injection flow path required by Specification 3.1.2.1 shall be OPERABLE and capable of being powered from an OPERABLE emergency bus.

APPLICABILITY: M00ES 5* and 6.

ACTION:

With no charging pump OPERABLE, suspend all operetions involving CORE ALTERATIONS or positive reactivity changes until one charging pump is restored to OPERAPLE status.

SURVEILLANCE REQUIREMENTS 4.1. 2. 3.1 At least the above required charging pump shall be demonsi. rated OPERABLE by,erifying, that on recirculation flow, the pump develops a discharge pressure of > 2458 psig when tested pursuant to Specification 4.U.5.

4.1.2.3.2 All charging pumps, except the above required OPERABLE pump, shall be demonstrated inoperable at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that the motor cirucit breakers have been removed fror. their electrical power supply circuits.

A maximum of one centrif ugal charging pump shall be OPERABLE whenever the temperature of one or moro of the RCS cold legs is 1 180 F.

'(IS116 FARLEY - UNIT 1 3/4 1-9 Amendment No.13

REACTIVITY CONTROL SYSTEMS CHARGING PUMPS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.4 At least two charging pumps shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With only one charging pump OPERABLE, restore at least two chargina pumps to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 1% aK/K at 200 F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore at least two charging pumps to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.1.2.4 At least two charging pumps shall be demonstrated OPERABLE by verifying, that on recirculation flow, each oump develops a discharge pressure of > 2458 psig when tested pursuant to Specification 4.0.5.

2011b0 FARLEY - UNIT 1 3/4 1-10

3/4.4 REACTOR COOLANT SYSTEM REACTOR COOLANT LOOPS LIMITING CONDITION FOR OPERATION 3. 4.1.

All reactor coolant loops shall be in operation.

APPLICABILITY: As noted below, but excluding MODE 6.*

ACTION:

Above P-7, comply with either of the followir.g ACTIONS:

With one reactor coolant loop and associated pump not in a.

operation, STARTUP and/or continued POWER OPERATION may proceed provided THERMAL POWER is restricted to less than 36% of RATED THERMAL POWER and the following ESF instrumentation channels associated with the loop not in operation, are placed in their tripped condition within 1 hour:

1.

T

-- Low-Low channel used in the coincidence circuit wiE0SteamFlow-HighforSteamLineIsolation.

2.

Steam Line Pressure - Low for Safety Injection.

3.

Steam Flow-High Channel used for MSIV Isolation.

4.

Differential Pressure Between Steam Lines - High channel used for Safety Injection (trip all bistables which indicate low active loop steam pressure wich respect to the idle loop steam pressure).

b.

With one reactor coolant loop and associated pump not in operation, subsequent STARTUP and POWER OPERATION above 36%

of RATED THERMAL POWER may proceed provided:

1.

The following actions have been corB?ad with the reactor in at least HOT STANDBY.

a)

Reduce the overtempera' e

-ip setpoint to the value specified in Sp(:.ticuv.a 2.2.1 for 2 loop operation.

See Special Test Exception 3.10.4.

FARLEY - UNIT 1 3/4 4-1 781169

REACTOR COOLANT SYSTEM _

ACTION (Continued) b)

Place the following reactor trip system and ESF instrumentation channels, associated with the loop not in operation, in their tripped conditions:

1)

Overpower AT channel.

2)

Overtemperature AT channel.

3)

T

-- Low-Low channel used in the coinci-dSMe circuit with Steam Flow - High for Steam Line Isolation.

4)

Steam Line Pressure - Low channel used for Safety Injection.

5)

Steam Flow-High channel used for MSIV Isolation.

6)

Differential Pressure Between Steam Lines - High channel used for Safety Injection (trip all bistables which indicate low active loop steam pressure with respect to the idle loop steam pressure).

c)

Change the P-8 interlock setpoint from the value specified in Table 3.3-1 to 1 66% of RATED THERMAL POWER.

2.

THERt4AL POWER is restricted to 1 61% of RATED THERMAL POWER.

Below P-7#:

1.0, operation may proceed provided at least two With K.,f[o>lant loops and associated pumps are in operation.

a.

reactof o

A reactor coolant pump shall not be started with one or more of the RCS cold leg temperature 1 310 F 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 RCS cold leg temperature.

'781170 FARLEY - UNIT 1 3/44-2 Amendment No. f.13

REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION (Continued)

ACTION (Continued) b.

With Keff < l.0, operation may proceed provided at least one reactor coolant loop is in operation with an associated reactor coolant or residual heat removal pump.*

The provisions of Specifications 3.0.3 and 3.0.4 are not c.

applicable.

SURVEILLANCE REQUIREMENTS 4.4.1. With one reactor coolant loop and associated pump not in operation, at least once per 31 days determine that:

The applicable reactor trip system and/or ESF actuation system a.

instrumentation channels specified in the ACTION statements above have been placed in their tripped conditions, and b.

The P-8 interlock setpoint is within the following limits if the P-8 interlock was reset for 2 loop operation < 66% of RATED THERMAL POWER.

'/81171

All reactor coolant pumps and residual heat removal pumps may be deenergized for up to one (1) hour provided no operations are permitted which could cause dilution of reactor coolant system boron concentration.

I FARLEY - UNIT 1 3/4 4-3 Amendment No. 13

REACTOR COOLANT SYSTEM SAFETY VALVES - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.2 A minimum of one pressurizer code safety valve shall be OPERABLE with a lift setting of 2485 FSIG + 1%.*

APPLICABILIH: MODES 4 and 5.

ACTION:

With no pressurizer code safety valve OPERABLE, immediately suspend all operations involving positive reactivity changes and place an OPERABLE RHR loop into operation in the shutdown cooling mode.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIPEMENTS 4.4.2 No additional Surveillance Requirements other than those required by Specification 4.0.5.

The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

78.1..EN.

FARLEY - UNIT 1 3/4 4-4

REACTOR COOLANT SYSTEM PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.9.2 The pressurizer temperature shall be limited to:

A maximum cooldown of 200 F in any one hour period, a.

b.

A maximum heatup of 100 F in any one hour period, and A maxfmum spray water temperature differential of 320 F.

c.

APPLICABILITY: At all times.

ACTION:

With the pressurizer temperature limits in excess of any of the above limits, restore the temperature to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the fracture toughness proptcties of the pressurizer; determine that the pressurizer 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 pressurizer pressure to less than 500 psig within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.9.9.2 The pressurizer temperatures shall be deteniiined to be within the limits at least once per hour during system heatup or cooldown.

The spray water temperature differential shall be determined to be within the limit at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during auxiliary spray operation.

781173 FARLEY - UNIT 1 3/4 4-29 Amendment No. 8

REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITIONS FOR OPERATION 3.4.9.3 At least one of the following overpressure protection systems shall be OPERABLE:

a.

Two RHR relief valves with:

1.

A lift setting of 5.450 psig, and 2.

The associated RHR relief valve isolation valves open; or b.

A reactor coolant system vent of > 2.85 square inches.

APPLICABLITY:

When the temperature of one or more of the RCS cold legs is 5,310 F, except when the reactor vessel head is removed.

ACTION:

a.

With one RHR relief valve inoperable, either restore the inoperable valve to OPERABLE status within 7 days or depressurize and vent the RCS through a > 2.85 square inch vent within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ; maintain tee RCS in a vented condition until both RHR relief valves have be restored to OPERABLE status.

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 > 2.85 square inch vent and maintain the RCS in a vented condition until both RHR relief valves have been restored to OPERABLE status, 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.

FARLEY - UNIT 1 3/4 4-30 Amendment No.13 781174

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.9.3.1 Each RHR relief valve shall be demonstrated OPERABLE by:

a.

Verifying the RHR relief valve isolation valves (8701a, 870lb, 8702a and 8702b) are open at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months when the RHR relief valve is being used for overpressure protection.

b.

Testing in accordance with the inservice test requirements

~ - - -

for ASME Category C valves pursuant to Specification 4.0.5.

Verification of the RHR relief valve setpoint, of at least one c.

RHR relief valve, at least once per 18 months on a rotating basis.

4.4.9.3.2 The RCS vent shall be verified to be open at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

when the vent is being used for ov2rpressure protection.

Except when the vent pathway is provided with a valve which is locked, sealed, or otherwise secured in the open position, then verify these valves open at least once per 31 days.

781175 FARLEY - UNIT 1 3/4 4-31 Amendment No.13

REACTOR COOLANT SYSTEM _

3/4.4.10 STRUCTURAL INTEGRITY ASME CODE CLASS 1, 2 and 3 COMPONENTS LIMITING CONDITION FOR OPERATION 3.4.10.1 The structural integrity of ASME Code Class 1, 2 and 3 components shall be maintained in accordance with Specification 4.4.10.1.

APPLICABILITY: ALL MODES ACTION:

With the structural integrity of any ASME Code Class 1 component (s) a.

not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature more than 50 F above the minimum temperature required by NDT considerations.

b.

With the structural integrity of any ASME Code Class 2 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature above 200 F.

With the structural integrity of any ASME Code Class 3 component (s) c.

not conforming to the above requirements, restore the structural integrity of the affected components to within its limit or isolate the affected component (s) from service.

d.

The provisions of Specification 3.0.4 are net applicable.

SURVEILLANCE REQUIPEMENTS 4.4.10.1 The structural integrity of ASME Code Class 1, 2 and 3 com-ponents shall be demonstrated;

'/81176 FARLEY - UNIT 1 3/4 4-32 Amendment No.13 e

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS Per the requirements of Specification 4.0.5 and a.

b.

By the augmented program specified in Specifications 4.4.10.2 and 4.4.10.3 4.4.10.2 In addition to the requirements of Specification 4.0.5, the Reactor Coolant Pump flywheels shall be inspected per the recormienda-tions of Regulatory Position C.4.b of Regulatory Guide 1.14 Revision 1, August 1975.

4.4.10.3 In addition to the requirements of Specification 4.0.5 the three main steamlines from the rigid anchor points of the containment penetrations downstream to and including the main steam header shall be inspected. The extent of the inservice examinations completed during each inspecticn interval (IWA 2400, ASME Code,1974 Edition,Section XI) snall provide 100 percent volumetric examination of circumferential and longitudinal pipe welds to the extent practical.

The areas subject to examination are those defined in accordance with examination category C-G for Class 2 piping welds in Table IWC-2520.

VO1177 FARLEY - UNIT 1 3/4 4-33 Amendment No.13

REACTIVITY CONTROL SYSTEMS BASES 3/4.1.2 BORATION SYSTEMS (Continued)

With the RCS temperature below 200 F, one injection system is acceptable without single f ailure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS end positive reactivity change in the event the single injection sysi.em becomes inoperable.

The ' limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps except the required OPERABLE pump to be inoperable below 180 F provides assurance that a mass addition pressure transient can be relieved by the operation of a single RHR relief valve.

The boron capability required below 200 F is sufficient to provide a SHUTDOWN MARGIN of 1% Ak/k after xenon decay and cooldown from 200 F to 140 F.

This condition requires either 2000 gallons of 7000 ppm borated water from the boric acid storage tanks or 9,000 gallons of 2000 ppm borated water from the refueling water storage tank.

The contained water volume limits include allowance for water not available because of discharge line location and other physical characteristics.

The OPERABILITY of one boron injection system during REFUELING ensures that this system is available for reactivity control while in MODE 6.

3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that (1) acceptable power Jistribution limits are maintained, (2) the minimum SHUTDOWN MARGIN is naintained, and (3) limit the potential effects of rod misalignment and associated accident analysis.

OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby 2nsure compliance with the control rod alignment and insertion limits.

The ACTION statements which permit limited variations from the aasic requirements are accompanied by additional restrictions which ansure that the original design criteria are met.

Misalignment of a od requires measurement of peaking factors and a restriction in THERMAL

'0WER; either of these restrictions provide assurance of fuel rod integrity juring continued operation.

In addition, those accidents analyses affected by a misaligned rod are re-evaluated to confirm that the results remain valid during future operation.

FARLEY - UNIT 1 B 3/41-3 Amendment No.13

'/81178

REACTIVITY CONTROL SYSTEMS BASES l

3 /4.1. 3 MOVALBE CONTROL ASSEMBLIES (Continued)

The maximum rod drop time restriction is consistent with the assumed rod drop time used in the accident analyses.

Measurement with T

> 541 F and with all reactor coolant pumps operating ensures that the me$l8 red drop times will be representative of insertion times experienced during a reactor trip at operating conditions.

i i

i I

781173 FARLEY - UNIT 1 B 3/4 1-4 Amendment No.13

3/4.4 REACTOR C00! ANT SYSTEM B_ASES 3/4.4.1 REACTOR COOLANT LOOPS The plant is designed to operate with all reactor coolant loops in operation, and maintain DNBR above 1.30 during all normal operations and anticipated transients. With one reactor coolant loop not in operation, THERMAL POWER is restricted to < 36 percent of RATED THERMAL POWER until the Overtemperature AT trip is reset.

Either action ensures that the DNBR will be maintained above 1.30.

A loss of flow in two loops will cause a reactor trip if operating above P-7 (11 percent of RATED THERMAL POWER) while a loss of flow in one loop will cause a reactor trip if operating above P-8 (36 percent of RATED THERMAL POWER).

A single reactor coolant loop provides sufficient heat removal capability for removing core decay heat while in HOT STANDBY; however, single failure considerations require placing a RHR loop into operation in the shutdown cooling moce if component repairs and/or corrective actions cannot be made within the allowable out-of-cervice time.

The restrictions on starting a Reactor Coolant Pump below P-7 with one or more RCS cold legs < 310 F are provided to prevent RCS pressure transients, caused by energy additions from the secondary system, which would exceed the limits of Appendix G to 10CFR Part 50.

The RCS will be protected against overpressurization 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 inta or (2) by restricting from starting the RCP's to when the secondary water temperature of eacn steam generator is less than 50 F above each of the RCS cold leg temperatures.

3/4.4.2 and 3/4.4.3 SAFETY VALVES The pressurizer code safety valves operate to prevent the RCS from above its Safety Limit of 2735 psig.

Each safety being pressu-u valve is designed to relieve 345,000 lbs per hour of saturated steam at the valve set point.

The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown.

In the event that no safety valves are OPERABLE, an operating RHR loop, connected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization.

During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of 2735 psig. The combined relief capacity of all of these valves is greater 781180 FARLEY - UNIT 1 B 3/4 4-1 Amendment No.13

REACTOR C00LANT SYSTEM BASES 3/4.4.2 and 3/4.4.3 SAFETY VALVES (Continued) than the maximum surge rate reuulting from a comp;ete loss of load assuming no reactor trip unti' the first Reactor Protective System trip set point is reached (i.e., no credit is taken for 6 direct reactor trip on the loss of load) and also assuming no operation of the power operated relief valves or steam dump valves.

Demonstration of the safety valves lift setting will occur only during shutdown and will be performed in accordance will the provisions of Section XI of the ASME Boiler and Pressure Vessel Code.

3/4.4.4 PRESSURIZER A steam bubble in the pressurizer ensures that the RCS is not a hydraulically solid system and is capable of acconnodating pressure surges during operation.

The steam bubble also protects the pressurizer code safety valves and power operated relief valves against water relief.

The power operated relief valves and steam bubble function to relieve RCS pressure during all design transients up to and including the design step load decrease with steam dump.

Operation of the power operated relief valves minimizes the undesirable opening of the spring-loaded pressurizer code safety valves.

3/4 4.5 STEAM GENERATORS The Surveillance Requirements for Inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained.

The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.

Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion.

Inservice inspection of steam generator tubing clso provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.

,lant is expected to be operated in a manner such that the Tb coolant will be maintained within those chemistry limits found second If 1.ne in negligible corrosion of the steam generator tubes.

to r L secondary coolant chemistry is not maintained within these litits, The localized corrosion may likely result in stress corrosion cracking.

extent of cracking during plant operation would be limited by the FARLEY - UNIT 1 B 3/4 4-2 Amendment No.13

.gy,

REACTOR COOLANT SYSTEM BASES STEAM GENERATORS (Continued) limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage =

500 gallons per day per steam generator).

Cracks having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents.

Operating plants have demon-strated that primary-to-secondary leakage of 500 gallons per day per steam generator can readily be detected by radiation monitors of steam generator blowdown.

Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged.

Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant.

However, even if a defect should develop in service, it will be found during scheduled inservice steam generator tube examinations.

Plugging will be required for all tubes with imper-fections exceeding the plugging limit of 40% degradation (60% tube wall thickness).

Steam generator tube. inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20% of the original tube wall thickness.

Whenever the results of any steam generator tubing inservice in-spection fall into Category C-3, these results will be promptly reported to the Commission pursuant to Specification 6.9.1 prior to resumption of plant operatico.

Such cases will be c'.ansidered by the Commission on a case-by-case basis and may result in e requirement for analysis, labora-tory examinations, tests, additional eddy-current inspection, and revision of the Technical Specification, if necessary.

781182 FARLEY - UNIT 1 B 3/4 4-3 Amendment No.13

REACTOR COOLANT SYSTEM BASES 3/4 4.6 REACTOR C0OLANT SYSTEM LEAKAGE 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the Reactor Coolant Pressure Boundary. These detection systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973.

3/4.4.6.2 OPERATIONAL LEAKAGE Industry experience has shown that while a limited amount of leakage is expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value of less than 1 GPM.

This threshold value is sufficiently low to ensure early detection of additional leakage.

The 10 GPM IDENTIFIED LEAKAGE limitation provides allowance for a limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the leakage detection systems.

The CONTROLLED LEAKAGE limitation restricts operation when the total flow supplied to the reactor coolant pump seals exceeds 31 GPM with the valve in the supply line fully open at a nominal RCS pressure of 2235 psig.

This limitation ensures that in the event of a LOCA, the safety injection flow will not be less than assumed in the accident analyses.

The total steam generator tube leckage limit of 1 GPM for all steam generators ensures that the dosage contribution from ti a tube leakage will be limited to a small fraction cf Part 100 limits in the event of either a steam generator tube rupture or steam line break.

The 1 GPM limit is consistent with the assumptions used in the analysis of these accidents. The 500 gpd leakage limit per steam generator ensures that steam generator tube integrity is maintained in the event of a main steam line rupture as under LOCA conditions.

781183 FARLEY-UNIT I B 3/4 4-4

REACTOR COOLANT SYSTEM BASES vessel inside radius are essentially identical, the measured transition shift for a sample can be applied with confidence to the adjacent section of the reactor vessel.

The heatup and cooldown curves must be recalcu-lated when the ART determined from the surveillance capsule is differentfromthe"SIlculatedART for the equivalent capsule radiation NDT exposure.

Thepressure-temperaturelimitlinesshownonFigure3.4-2fo[

reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature require-ments of Appendix G to 10 CFR 50.

The number of reactor vessel irradiation surveillance specimens an the frequencies for removing and testing these specimens are provided in Table 4 4-5 to assure compliance with the requirements of Appendix H to 10 CFR Part 50.

The limitations imposed on pressurizer heatup and cooldown and spray water temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASME Code requirements.

The OPERABILITY of two RHR relief valver or an RCS vent opening of t 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 cold legs are 1 310 F.

Either RHR relief valve has adequate relieving capability to protect the RCS from over-pressurization when the transient'is limited to either (1) the start of an idle RCF with the secondary water temperature of the steam generator 1 50 F above the RCS cold leg temperatures, or (2) the start of 3 charging pumps and their injection into a water solid RCS.

3/t.4.10 STRUCTURAL INTEGRITY The inspection programs for ASME Code Class 1, 2 and 3 components ensure that the structural integrity of these components will be maintained at an acceptable level throrghout the life of the plant.

To the exten' applicable, the inspection program for these components is in complianc0 with Section XI of the ASME Boiler and Pressure Vessel Code.

781184 FARLEY-UNIT 1 B 3/4 4-11 Amendment No.13

ML19207A872

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