Text

Amend 10 to License NPF-2,incorporating Changes to Tech Specs Re Radial Peaking Factor Limits & Setpoint for Reactor Trip After Turbine Trip.Tech Spec Changes Encl
	ML19261C340
Person / Time
Site:	Farley
Issue date:	03/02/1979
From:	Schwencer A; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML19261C341	List: ML19261C340; ML19261C342; ML19261C344;
References
	NUDOCS 7903220294
	Download: ML19261C340 (41)
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.f hdCLIAR REGU'.ATOR'. COMMIS3 ON WASHINGTON, D. C. 20555 g

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ALABAMA POWER COMPANY DOCKET NO. 50-348 JOSEPH M. FARLEY NUCLEAR PLANT, UNIT 10. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.10 License No. NPF-2 1.

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

A.

The application for amendment by Alabama Power Co'mpany (the licensee)datedNovember 15, 1978, supplemented by letters dated December 21, 1978 and January 12, 1979, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission'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 Comission; 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 Comission's regulations; D.

The issuance of this amendment will not be inimical to the comon 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 Comission's regulations and all applicable requirements have been satisfied.

7103D}Odfli

. 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. 10, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications."

3.

This license amendment is effective as indicated in the letter of transmittal for this amendment.

FOR THE NUCLEAR REGULATORY COMMISSION

/.

pe2WOL-A. Schwencer, Chief Operating Reactors Branch #1 Division of Operating Reactors

Attachment:

Changes to the Technical Specifications Date of Issuance: March 2, 1979

ATTACHMENT TO LICENSE AMENDMENT NO.10 FACILITY OPERATING LICENSE NO. NPF-2 DOCKET NO. 50-348 Replace or revise as indicated 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.

Pages III l-3 B 2-7 3/4 1-16 3/4 1-17 3/4 1-18 3/4 1-19 3/4 1-20 3/4 1-26 (Delete) 3/4 2-7 3/4 2-11 3/4 3-7 3/4 3-8 3/4 10-1 3/4 10-2 3/4 10'3 B 3/4 2-1 8 3/4 2-2 B 3/4 2-4 8 3/4 2-5 5-4 6-3 6-5 6-11

INDEX LIMITIflG CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION Page 3/4.0 APPLICABILITY.............................................

3/4 0-1 3,4.1 REACTIVITY CONTROL SYSTEMS 7/4.1.1 B0 RATION CONTROL 00*F.........................

3/4 1-1 Shutdown Margin - T avg Shutdown Margin - T

200*F.........................

3/41-3 avg -

Baron Dilution.........................................

3/4 1-4 Moderator Temperature Coefficient......................

3/4 1-5 Minimum Temperature for Cri ti cali ty....................

3/4 1-6 3/4.1.2 B0 RATION SYSTEMS Flow Paths - Shutdown..................................

3/4 1-7 Fl ow Pa th s - Opera ti ng.................................

3/4 1 8 Charging Pump - Shutdown...............................

3/4 1-9 Charging Pumps - Operating.............................

3/4 1-10 Boric Acid Transfer Pumps -

Shutdown...................

3/4 1-11 Boric Acid Transfer Pumps - Operating..................

3/4 1-12 Borated Water Sources -

Shutdown.......................

3/4 1-13 Borated Water Sources - Operating......................

3/4 1-14 3/ 4.1. 3 MOVABLE CONTROL ASSEMBLIES Group Height...........................................

3/4 1-16 Posi tion Indicator Channel s Operating................,

3/4 1-19 Position Indicator Channels Shutdown...................

3/4 1-20 Rod Drop Time..........................................

3/4 1-21 Shutdown Rod Insertion Limit...........

3/4 1-22 Control Rod Insertion Limits...........................

3/4 1-23 FARLEY - UNIT 1 III Amendment No.10

INDEX i

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS SECTION Pace 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE...................................

3/42-1 3/4.2.2 HEAT FLUX HOT CHANNEL FACT 0R...........................

3/4 2-5 3/4.2.3 NUCLEAR ENTHALPY H0T CHANNEL FACT 0R....................

3/4 2-9 3/4.2.4 QUADRANT POWE R T I LT RAT I 0..............................

3/4 2-11 3/4.2.5 DNB PARAMETERS.........................................

3/4 2-13 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION....................

3/43-1 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM.............

3/4 3-14 INSTRUMENTATION 3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring...................................

3/4 3-35 Movable Incore Detectors...............................

3/4 3-39 Seismic Instrumentation................................

3/4 3-40 Meteorological Instrumentation.........................

3/4 3-43 Remote Shutdown Instrumenta ti on........................

3/4 3-46 Chl ori ne De tection Systems.............................

3/4 3-49 High Energy Line Break Isolation Sensors...............

3/4 3-50 Post Accident Instrumentation..........................

3/4 3-53 Fire Detection Instrumentation.........................

3/4 3-56 l

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS N o rma l 0 p e r a t i o n.......................................

3/4 4-1 3/4.4.2 SAFETY VALVES - SHUTD0WN...............................

3/44-4 3/4.4.3 SAFETY VALVES - 0PERATING..............................

3/4 4-5 3/4.4.4 PRESSURIZER............................................

3/4 4-6 3/4.4.5 STEAM GENERATORS......................................

3/4 4-7 FARLEY - UNIT 1 IV Amendment No. 4

DE:!NITIONS CHANNEL FUNCTIONAL TEST 1.11 A CHANNEL FUNCTIONAL TEST shall be the injection of a simulated signal into the channel as close to the primary sensor as practicable to verify OPERABILITY including alarm and/or trip functions.

CORE ALTERATION 1.12 CORE ALTERATION shall be the movement or manipulation of any com-ponent within the reactor pressure vessel with the vessel head removed and fuel in the vessel.

Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.

SHUTDOWN MARGIN 1.13 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming all full length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdiawn.

IDENTIFIED LEAKAGE 1.14 IDENTIFIED LEAKAGE shall be:

a.

Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or b.

Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE, or c.

Reactor coolant system leakage through a steam generator to the secondary system.

UNIDENTIFIED LEAKAGE 1.15 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE.

i FARLEY - UNIT 1 1-3 Amendment No.10

DEFIN!TIONS PRESSURE BOUNDARY LEAKAGE 1.16 PRESSURE B0UNDARY 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.

CONTROLLED LEAKAGE 1.17 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals.

OVADRANT POWER. TILT RATIO 1.18 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated 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.

DOSE EQUIVALENT I-131 1.19 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (uCi/ gram) shich alone would produce the same thyroid dose as the quantity and iso-topic mixture of I-131, I-132, I-133, I-134, and I-135 actually present.

The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TIO-14844, " Calculation of Distance Factors for Power and Test Reactor Sites."

STAGGERED TEST BASIS 1.20 A STAGGERED TEST BASIS shall consist of:

a.

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

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

FARLEY - UNIT 1 1-4

LIMITING SAFETY SYSTEM SETTINGS BASES reliability of the Reactor Protection System.

This trip is redundant.to the Steam Generator Water Level Low-Low trip.

The Steam /Feedwater Flow Mismatch portion of this trip is activated when the steam flow exceeds the feedwater flow by >_ l.55 x 106 lbs/ hour.

The Steam Generator Low Water level portion of the trip is activated when the water level drops below 25 percent, as indicated by the narrow range instrument. These trip values include sufficient allowance in excess of normal operating values to preclude spurious trips but will initiate a reactor trip before the steam generators are dry.

Therefore, the required capacity and starting time requirements of the auxiliary feedwater pumps are reduced and the resulting thermal transient en the Reactor Coolant System and steam generators is minimized.

Undervoltace and Underfrecuency - Reactor Coolant pumo Busses The Undervoltage and Underfrequency Reactor Coolant Pump bus trips provide reactor core protection against DNB as a result of loss of voltage or underfrequency to more than one reactor coolant pump. The specified set poin.ts assure a reactor trip signal is generated before the low flow trip set point is reached.

Time delays are incorporated in the underfrequency and undervoltage trips to prevent spurious reactor trips from momentary electrical power transients.

For under-voltage, the delay is set so that the time required for a signal to reach the reactor trip breakers following the simultaneous trip of two or more reactor coolant pump bus circuit breakers shall not exceed 0.9 seconds.

For underfrequency, the delay is set so that the time required for a signal to reach the reactor trip breakers after the underfrequeni:y trip sat point is reached shall not exceed 0.3 seconds.

Turbine Trip A Turbine Trip causes a direct reactor trip when operating above P-9.

Each of the turbine trips provide turbine protection and reduce the severity of the ensuing transient.

No credit was taken in the accident analyses for operation of these trips.

Their functional capability at the specified trip settings is required to enhance the overall reliability of the Reactor Protection Sys e-FARLEY - UNIT I B 2-7 Amendment No.10

4 LIMITING SAFETY SYSTEM SETTINGS BASES Safety Injection Input from ESF If a reactor trip has not already been generated by the reactor protective instrumentation, the ESF automatic actuation logic channels will initiate a reactor trip upon any signal which initiates a safety injection. This trip is provided to protect the core in the event of a LOCA. The ESF instrumentation channels which initiate a safety.

injection signal are shown in Table 3.3-3.

Reactor Coolant Pump Breaker Position Trip The Reactor Coolant Pump Breaker Position Trips are anticipatory trips which provide reactor core protection against DNB resulting from the opening of any one pump breaker above P-8 or the opening of two or more pump breakers below P-8.

These trips are blocked below P-7.

The open/close position trips assure a reactor trip signal is generated before the low flow trip set point is reached.

No credit was taken in the accident analyses for operation of these trips. Their functional capability at the open/close position settings is required to enhance the overall reliability of the Reactor Protection System.

FARLEY - UNIT 1 B 2-8

REACTIVITY CONTROL SYSTEMS SURVEILLANCE RE0VIREMENTS (Continued)

At leas' once per 7 days by:

t a.

1.

Verifying the boron concentration in eoch water source, 2.

Verifying the contained borated water volume of each water source, and 3.

Varifying the boric acid storage system solution temperature.

b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature when the outside ambient air temperature is < 35 F.

FARLEY - UNIT 1 3/a 1-15

[ REACTIVITY C0" TROL SYSTE"5 3/4.1.3 MOVABLE CONTROL ASSEMBLIES GROUP HEIGHT LIMITING CONDITION FOR OPERATION 3.1.3.1 All full length (shutdown and control) rods which are inserted in the core shall be OPERABLE and positioned within 112 steps (indicated position) of their group step counter demand position.

APPLICABILITY: MODES 1* and 2*

ACTION:

a.

With one or more full length rods inoperable due to being immovable as a result of excessive friction or mechanical interference or known to be untrippable, determine that the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

With more than one full length rod inoperable or misaligned from the group step counter demand position by more than 1 12 steps (indicated position), be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c.

With one full length rod inoperable or misaligned from its group step counter demand position by more than i 12 steps (indicated position), POWER OPERATION may continue provided that within one hour either:

1.

The rod is restored to OPERABLE status within the above alignment requirements, or 2.

Tije rod is declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied.

POWER OPERATION may then continue provided that; a)

A revaluation of each accident analysis of Table 3.1-1 is performed within 5 days, this reevaluation shall confirm that the previously analyzed results of these accidents remain valie for the duration of operation under these conditiou, b)'

The SHUTDOWN MARGIN requirement of Specificatior.

3.1.1.1 is determined at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , and

See Special Test Exceptions 3.10.2 and 3.10.3.

FARLEY - UNIT 1 3/4 1-16 Amendment No.10

IREACT'VITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION (Continued) c)

A power distribution map is obtaingd from the movable incore detectors and F (Z) and FJ are verified to be within their limits wi9hin 72 houM.

d)

Either the THERMAL POWER level is reduced to < 75%

of RATED !HERMAL POWER within one hour and within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the high neutron flux trip setpoint is reduced to < 85% of RATED THERMAL F0WER, or e)

The remainder of the rods in the group with the inoperable rod are aligned to within + 12 steps of the inoperable rod within one hour while maintaining the rod sequence and insertion limits of Figures 3.1-1 and 3.1-2; the THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsecuent operation.

SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The position of each full length rod shall be determined to be within the group demand limit by verifying the individual rod positions at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during~ time intervals when the Rod Position Deviation Monitor is inoperable then verify the group positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4.1.3.1.2 Each full length rod not fully inserted in the core shall be determined to be OPERABLE by movement of at least 10 steps in any one direction at least once per 31 days.

FARLEY - UNIT 1 3/4 1-17 Amendment No.10

TABLE 3.1-1 ACCIDENT ANALYSES RE0VIRING REEVALUATION IN THE EVENT OF AN INOPERABLE FULL LENGTH ROD Rod Cluster Control Assembly Insertion Characteristics Rod Cluster Control Assembly Misalignment Loss Of Reactor Coolant From Small Ruptured Pipes Or From Cracks In large Pipes Which Actuates The Emergency Core Cooling System Single Rod Cluster Control Assembly Withdrawal At Full Power Major Reactor Coolant System Pipe P.uptures (Loss Of Coolant Accident)

Major Secondary System Pipe Rupture Rupture of a Control Rod Drive Mechanism Housing (Rod Cluster Control Assembly Ejection)

FARLEY - UNIT 1 3/4 1-18 Amenc ent No. 10

REACTIVITY CONTROL SYSTEMS POSITION INDICATOR CHANNELS OPERATING LIMITING CONDITION FOR OPERATION 3.1.3.2 - All shutdown and control rod position indicator channels and the demand position indication system shall be OPERABLE and capable of deter-mining the control rod positions within + 12 steps.

APPLICABILITY: MODES 1 and 2.

ACTION:

a.

With a maximum of one rod position indicator channel per group inoperable either:

1.

Determine the position of the non-indicating rod (s) in-directly by the movable incore detectors at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and immediately after any motion of the non-indicating rod which exceeds 24 steps in one direction since the last determination of the rod's position, or 2.

Reduce THERMAL POWER TO < 50% of RATED THERMAL POWER within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

b.

With a maximum of one demand position indicator per bank inoperable either:

1.

Verify that all rod position istdicators for the affected bank are OPERABLE and that the most withdrawn rod and the least withdrawn rod of the bank are within a maximum of 12 steps of each other at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , or 2.

Reduce THERMAL POWER to < 50% of RATED THERMAL POWER within 8 houts.

c.

The provision of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.1.3.2 Each rod position indicator channel shall be determined.to be OPERAELE by verifying the demand position indication system and the rod positi.on indicator channels agree within 12 steps at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the Rod Position Deviation Monitor is inoperable, then compare the demand position indication system and the rod position indicator channels at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

FARLEY - UNIT 1 3/4 1-19 Amendment No. 10

REACTIVITY CONTROL SYSTEMS POS:~IT; I:.:::ATOR CHAN:;T_3 - SFJTD0's."4 LIMITIN, CONDITION FOR OPERATION 3.1.3.3 At least one rod position indicator channel (excluding demand position indication} shall be OPERABLE for each shutdown or control rod not fully inserted.

APPLICABILITY: MODES 3,*# 4*# and 5*#

ACTION: With less than the above required position indicator channels OPERABLE, innediately open the reactor trip system breakers.

SURVEILLANCE REQUIREMENTS 4.1.3.3 Each of the above required rod position indicator channels shall be determined to be OPERABLE by verifying the demand position indication system and the rod position indicator channels agree within 12 steps at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

With the reactor trip system breakers in the closed position.

See Special Test Exception 3.10.5 FARLEY - UNIT 1 3/4 1--20 Amendment No. 7, 10

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POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued)

When the F f is less than or equal to the F 2.

RTP limit for x

theappropriatemeasuredcoreplane,adgitidXalpower dgributiogmapsshallbetakenandF compared to F

and F at least once per 31 EFPDf7 xy xy e.

The F limii:s for RATED THERMAL POWER within speciff.c core planelYshall be:

RTP 1.

F

< l.71 for all core planes containing bank "D" control rdds,and 2.

Fgpall unrodded core planes:

F 1 1.68 up to core elevations of 2.4 ft.

xy RTP F

1 1.75 for core elevations from 2.4 ft. to 7.2 ft.

P F

1 1.61 for core elevations above 7.2 ft.

f.

The F limits of e, above, are not applicable in the following core 6Yaneregionsasmeasuredinpercentofcoreheightfromthe bottom of the fuel:

1.

Lower core region from 0 to 15%, inclusive.

2.

Upper core region fro _m 85 to 100%, inclus'.ve.

3.

Grid plane regions at 17.8 1 2%, 32.1 1 2%, 46.4 1 2%,

60.6 1 2% and 74.9 1 2%, inclusive.

4.

Core plane regions within i 2% of core height (i 2.88 inches) about the bank demand position of the bank "D" control rods.

g.

With F exceeding F ' the effects of F C

evalualddtodetermiNE,ifF(Z)iswithiNYits1b(Z)shallbe on F it.

g 4.2.2.3 When F (Z) is measured pursuant to specification 4.lc.2.2, an n

overall measured F (Z) shall be obtained from a power dis:ribt.:for map 0

and increased by 3.. to account for manufacturing cle-ances ard further increased by 55 to account for measurement uncertaint.,.

FARLEY - UNIT 1 3/4 2-7 Amendment No.10

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lgli

+

,i-l t

[

..i

{ l ti...a g;, _._ g I

i il li i

illi lili ill ih o

2 4

6 8

10 12 11 CORE HEIGHT (FT)

Figure 3.2-2 K(Z) Normalized Fo(Z) as a Function of Core Height

POWER DISTRIBUTION LIMITS OVADRANT POWER TILT RATIO LIMITING C0!;DITION FOR OPERATION 3.2.4 THE QUADRANT POWER TILT RATIO shall not exceed 1.02.

APPLICABILITY: MODE 1 AB0VE 50% OF RATED THERMAL POWER

ACTION:

a.

With the QUADRANT POWER TILT RATIO determined to exceed 1.02 but 1 1.09:

1.

Within 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months s:

a)

Either reduce the QUADRANT POWER TILT RATIO to within its limit, or b)

Reduce THERMAL POWER at least 3% from emfED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1.0 and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within

' the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

2.

Verify that the QUADRANT POWER TILT RATIO is within its limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding the limit or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High Trip set-points to 1 55% of PATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

~

Ide. tify and correct the cause of the out of limit condition 3.

n prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL power may proceed provided that the QUADRANT POWER TILT RATIC is verified within its limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified acceptable at 95%

or greater RATED THERMAL POWER.

b.

With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to misalignment of either a shutdown or control rod:

1.

Reduce THERAL POWER at least 35 fror RATED THERMAL POWER for each l' of indicated OL'A RANT POWER TILT F17:0 in excess of 1.0, within 30 nicu ei.

2.

Verify that tne QUADRANT POWER T:LT RATIO is wi:nin its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after exceeding tne limit or

See Special Test Exception 3.10.2.

FARLEY - U.:T 1 3/4 2-11 Amendment No.10

POWER DISTRIBUTION LIMITING CONDITION FOR OPERATION (Continued) reduce THERMAL POWER to less than 50% of PATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High trip Setpoints to < 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

3.

Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified acceptable at 95%

or greater RATED THERMAL POWER.

c.

With.the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to causes other than the misalignment of either a shutdown, control or part length rod:

1.

Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High Trip Setpoints to < 55'; of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

2.

Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified at 95 or greater RATED THERMAL POWER.

SURVEILLANCE REQUIREMENTS 4.2.4 The QUADRANT POWER TILT RATIO shall be determined to be within the limit above 50% of RATED THERMAL POWER by:

Calculating the ratio at least once per 7 days when the alarm is a.

OPERABLE.

b.

Calculating the ratio at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during steady state operation when the alarm is inoperable.

c.

Using the movable incore detectors to confirn -hit the.ocwer distribution is consistent with the indicated :UAD4NT POWER TILT RATIO at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when one Power Dar.ge Channel is inoperable and THERMAL POWER is > 75 percent of PATED THERMAL POWER.

FAPLEY -UNIT 1 3/4 2-12

TABLE 3.3-1 (Continued)

ACTION 9 - With a channel associated with an operating loop inoperable, restore the inoperable channel to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel associated with an operating loop may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.1.1.1.

ACTI0fl 10 - With one channel inoperable, restore the inoperable channel to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to below P-8 within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

Operation below P-8 may continue pursuant to ACTION 11.

ACTION 11 - With less than the itinimum Number of Channels OPERABLE, operation may continue provided the inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ACTION 12 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and/or open the reactor trip breakers.

REACTOR TRIP SYSTEM INTERLOCKS DESIGNATION CONDITION AND SETPOINT FUNCTI0'i P-6 With 2 of 2 Intermediate Range))

P-6 defeats the manual Neutron Flux Channels < 6 x 10 block of source range amps.

reactor trip P-7 With 2 of 4 Power Range Neutron P-7 defeats the automatic Flux Channels > 11% of RATED block of reactor trip THERMAL POWER or 1 of 2 Turbine on: Low flow in more impulse chamber pressure channels than one primary coolcnt

> 55 psia.

loop, reactor coolant pump under-voltage and under-frequency, pressurizer low pressure, and press;rizer hign level.

FARLEY - UNIT 1 3/4 3-7 Arendment No.10

TABLE 3.3-1 (Continued)

DE SIG*lATION CONDITIO'1 AND SETPOINT FUtiCTIOf:

P-8 With 2 of 4 Power Range Neutron P-8 defeats the automatic Flux channels > 36: of RATED block of reactor trip THERMAL POWER.

on low coolant flow in

~

a single loop.

P-9 With 2 of 4 Power range fieutron P-9 defeats the automatic Flux channels > 51; of RATED block of Reactor Trip on THERMAL POWER.

Turbine Trip.

P-10 With 3 of 4 Power range neutron P-10 prevent's the manual flux channels < 35 of RATED block of: Po..er range THER!%L POWER.

low setpoint reactor trip, intermediate range reactor trip, and inter ediate range rod stops.

Provides input to P-7.

FARLEY - UNIT 1 3/4 3-8 Ar.endmen: "o. 10

3/4.10 SPECIAL TEST EXCEPTIONS SHUT 00',i'; MAE3*.'

LIMITING CONDITION FOR OPERATION 3.10.1 The SHUTOOWN MARGIN requirement of Specification 5.1.1.1 may be suspended for measurement of control rod worth and shutdown margin provided reactivity equivalent to at least the highest estimated control rod worth is available for trip insertion from OPERABLE control rod (s),

APPLICABILITY: MODE 2.

Ar ION:

a.

With any full length control rod not fully inserted and with less than the above reactivity equivalent available for trip insertion, initiate and continue boration at 1 30 gpm of between 7000 and 7700 ppm boric acid solution or its equivalent until the SHUTDOWN MARGIN required by Speci-fication 3.1.1.1 is restored.

b.

With all full length control rods inserted and the reactor suberitical (K

l immediately inTNa<ea.0)bylessthantheabovere.ctivityequivalent30gpmofbetween7000and t

nd continue boration at 1 7700 ppm boric acid solution or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.

SURVEILLANCE REQUIREMENTS 4.10.1.1 The position of each full length rod either partially or fully withdrawn shall be detennined at least once per 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

4.10.1.2 Each full length rod not fully inserted shall be demonstrated capable of full insertions when tripped from at least the 50'4 withdrawn position within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to reducing the SHUTDOWN MARGIN to less than the limits of Specification 3.1.1.1.

FARLEY-UNIT l 3/4 10-1 Amendmen n. 10

lSPECIAL TEST EXCEDTIO*:S GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION 3.10.2 The group height, insertion and power distribution limits of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.1, and 3.2.4 may be suspended during the performance of PHYSICS TESTS provided:

a.

The THERMAL POWER is maintained < 855; of RATED THERMAL POWER, and b.

The limits of Specifications 3.2.2 and 3.2.3 are maintained and' determined at the frequencies specified in Specification 4.10.2.2 below.

APPLICABILITY: MODE 1 ACTION:

With any of the limits of Specifications 3.2.2 or 3.2.3 being exceeded while the requirements of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.1 and 3.2.4 are suspended, either:

a.

Reduce THERMAL POWER sufficient to satisfy the ACTION require-ments of Specifications 3.2.2 and 3.2.3, or b.

Be in HOT STAND 8Y within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.10.2.1 The THERMAL POWER shall be determined to be < 85!. of RATED THERMAL POWER at least once per hour during PHYSICS TE3ITS.

4.10.2.2 The Surveillance Requirements of Specifications.4.2.2 and 4.2.3 shall be performed at the following frequencies during PHYSICS TESTS:

a.

Specification 4.2.2 - At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.

Specification 4.2.3 - At least once per 12 ho/s.

FARLEY-UNIT 1 3/4 10-2 Amendment No.10

SPECIAL TEST EXCEPTIONS PHYSICS TESTS LIMITING C0tiDITION FOR OPERATION 3.10.3 The limitations of Specifications 3.1.1.4, 3.1.3.1, 3.1.3.5, and.3.1.3.6 may be suspended during the performance of PHYSICS TESTS provided:

a.

The THERMAL POWER does not exceed 5% of RATED THERMAL POWER, and b.

The reactor trip setpoints on the OPERABLE Intennediate and Power Range Nuclear Channels are set at 1 25% of RATED THERMAL POWER.

APPLICABILITY: MODE 2.

ACTION:

With the THERMAL POWER > 5% of RATED THERMAL POWER, immediately open the reactor trip breakers.

SURVEILLANCE REQUIREMENTS 4.10.3.1 The THERMAL POWER shall be determined to be < 5% of RATED THERMAL POWER at least once per hour during PHYSICS TESTS.

4.10.4.2 Each Intermediate and Power Range Channel shall be subjected -

to a CHANNEL Ft)N$TIONAL TEST within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months prior to initiating PHYSICS TEST

1 FARLEY-UNIT 1 3/4 10-3 Amendment No. 10

l SPECIAL TEST EXCECTION REACTOR COOLANT LOOPS i

LIMITING CONUITION FOR OPERATION l

3.10.4 The limitations of Specification 3.4.1 may be suspended during the performance of startup and PHYSICS TESTS provided:

l a.

The THERMAL POWER does not exceed the P-7 Interlock Setpoint, and b.

the Reactor Trip Setpoints on the OPERABLE Intermediate and Power Range Channels are set < 25% of RATED THERMAL POWER APPLICABLITY:

During operation below the P-7 Interlock Setpoint.

ACTION:

With the THERMAL POWER greater than the P-7 Interlock Setpoint, immediately open the reactor trip breakers.

SURVEILLANCE REQUIREMENTS 4.10.4 1 The THERMAL POWER shall be determined to be less than P-7 Interlock Setpoi ' at least once per hour during startup and PHYSICS TESTS.

4.10.4.

Each intermediate, Power Range Channel and P-7 Interlock shall be subjected to a CHANNEL FUNCTIONAL TEST within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months prior to initiating PHYSICS TESTS.

FARLE - UNIT 1 3/4 10-4

3/4.2 POWER DISTRIBUTION LIM!TS BASES The specifications of this section provide assurance of fuel integ-rity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by:

(a) maintaining the minimum DNBR in the core > 1.30 during nonnal operation and in short term transients, and (b) limiting the fission gas release, fuel pellet temperature & cladding mechanical properties to within assumed design criteria.

In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200'F is not exceeded.

The definitions of certain hot channel and peaking factors as used in these specifications are as follows:

F (Z)

Heat Flux Hot Channel Fsctor, is defined as the maximum local g

heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing the man-ufacturing tolerances on fuel pellets and rods.

H p

Nuclear Enthalpy Rise Hot Channel Factor, is defined as the aH ratio of.the integral of linear power along the rod with the highest integrated power to the average rod power.

7(Z)

Radial Peaking Factor, is defined as the ratio of peak power F

density to average power density in the horizontal plane at core elevation Z.

3/4.2.1 AXIAL FLUX DIFFERENCE (AFD)

The limits on AXIAL FLUX DIFFERENCE assure that the F bound envelope of 2.32 times the normalized axial peaking 9a(Z) upper ctor is not exceeded during either normal operation or in the event of xenon redis-tribution following power changes.

Target flux difference is determined at equilibrium xenon conditions.

The full length rods may be positioned within the core in accordance with their respective insertion limits and should be inserted near their normal position for steady state operation at high power levels.

The value of the target flux difference obtained under these conditions divided by the l

fraction of RATED THERMAL POWER is the target flux difference at RATED THERMAL POWER for the associated core burnup conditions.

Tar;e: flux differences for other THERMAL POWER levels are obtained by multici;.ing the RATED THEPJ4AL POWER value by the appr.opriate fractional THERMAL F0WER level. The periodic updating of the target flux difference value is necessary to reflect core burnup considerations.

FARLEY - UNIT 1 B 3/4 2-1 Amendment No.10

1 POWER T5TMEUT:T LP'~5 BASES Although it is intended that the plant will'be operated with the AXIAL FLUX DIFFERENCE within the +(5)% target band about the target flux defference, during rapid plant THERMAL POWER reductions, control rod motion will cause the AFD to deviate outside of the target band at re-duced THERMAL POWER levels.

This deviation will not affect the xenon redistribution sufficiently to change the envelope of peaking factors which may be reached on a subsequent return to RATED THERMAL POWER (with the AFD within the target band) provided the time duration of the devi-ation is limited.

Accordingly, a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months penalty deviation limit cumulation during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is provided for operation outside of the target band but within the limits of Figure 3.2-1 while at THERMAL POWER levels between 50% and 905 of RATED THERMAL POWER.

From THERMAL POWER levels between 15% and 50", of RATED THERMAL POWER, deviations of the AFD outside of the target band are less significant.

The penalty of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months acutal time reflects this reduced significance.

Provisions for monitoring the AFD on an automatic basis are derived from the plant process combuter though the AFD Monitor Alarm.

The computer determines the one minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for at least 2 of 4 or 2 of 3 OPERABLE excore channels are outside the target band and the THERMAL POWER is greater than 90% of RATED THERMAL POWER.

During operation at THERMAL POWER levels between 50% and 905 and between 155 and 50%

RATED THERMAL POWER, the computer outputs an alarm message when the penalty deviation accumulates beyond the limits of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , respectively.

Figure B 3/4 2-1 shows a typical monthly target band.

ARLEY - UNIT 1 e 3/4 2-2 Amendment ;;o,10

Percent of Rated Thermal Power 5%

5%

100%

__r_=-

+

_ ??$$h E

=.

90%

-+b Z

~

~~

T

~-

80%

~

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cTarget Flux Difference 60 %

.t _. _

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30%

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30%

20%

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0

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+20h

+30%

INDICATED AXIAL FLUX DIFFERENCE Figure B 3/4 21 TYPICAL INDICATED AXIAL FLUX DIFFERENCE VERSUS THERMAL POWER FARLEY - UNIT 1 B 3/4 2-3

7k'E 3 E!STR!EUTION LI"! S BASES 3/4.2.2 and 3/4.2.3 HEAT FLUX AND NUCLEAR ENTHALPY HOT CHANNEL and RADIAL PEAKING FACTORS-N F (Z), F H and F N) 0 xy The limits on heat flux and nuclear enthalpy hot channel factors ensure that 1) the design limits on peak local power density and minimum DNBR are not exceeded and 2) in the event of a LOCA the peak fuel-clad temperature will not exceed the 2200*F ECCS acceptance criteria limit.

Each of these hot channel factors are measureable but wilt normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3.

This periodic surveillance is sufficient to ensure that the hot channel factor limits are maintained provided:

a.

Control rods in a bank move together with no individual rod insertion differing by more than + 12 steps from the group demand position.

b.

Contrni rod banks are sequenced with overlapping banks as descriLad in Specification 3.1.3.5.

c.

The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained.

d.

The axial. power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.

N The relaxation in F as a function of THERMAL POWER allows changes H

ig the radial power shape for all permissible rod insertion limits.

ab$ve,ll be maintained within its limits provided conditions a through d F

wi are maintained.

When an F measurement is taken, both experimental error and man-n ufacturing tolerance must be allowed for.

5% is the appropriate allowance for a full core map taken with the incore detector flux mapping system and 3'l is the appropriate allowance for manufacturing tolerance.

WhenF)H is measured, experimental error must be allowed for an: 4 is the appropriate allowance for a full corefap taken with the inco e detection system.

The specified limit for F^H also contains an 8% aHow-aqce for uncertainties which mean that normai operation will result in F', Un s :

1 1.55/1.08.

The 8% allowance is based on the following considera-tf FARLEY - UNIT 1 B 3/4 2-4 Amendment No.10

POWER DISTRIBUTION LIMITS BASES Abnormal perturbations in ghe radial power shape, such as from a.

rod. misalignment, effect FaH more directly than F,

g b.

Although rod movement has a direct influence upon limiting F to withfn its limit, such control is not readily available to limit FaH, and c.

Errors in prediction for control power shape detected during startup physics tests can be compensated for in F ing axial flux distributions.

Thiscompensation9erFby gestrict-is aH less readily available.

The radial peaking factor, Fxy (z), is measured periodically to provide additional assurance that the hot channel factor, Fg (z), remains v' thin its limit.

The Fxy (z) limits were determined from expected powe.r control maneuvers over the full range of burnup conditions in the core.

3/4.2.4 QUADRANT POWER TILT RATIO The quardrant power tilt ratio limit assures that the radial power distribution satisfies the design values used in the power capability analysis.

Radial power distribution measurements are made during start-up testing and periodically during power operation.

The limit of 1.02 at which corrective action is required provides DNB and linear heat generation rate protection with x-3 plane power tilts. A limiting tilt of 1.025 can be tolerated before the margin for uncertainty in F is depleted.

The limit of 1.02 was selected to provide 0

an allowance for the uncertainty associated with the indicated power tilt.

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

In the event such action does not correct the tilt, the margin for uncertainty on F is reinstated by reducing the power by 3 percent for each percent of ti19 in excess of 1.0.

FARLEY - UNIT 1 B 3/4 2-5 Amendment No.10

lPOWERDISTRIBUTIONLIMITS BASES 3/4.2.5 DNB PARAMETERS The limits on the DNB related parameters assure that each of the parameters are maintained within the normal steady state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the initial FSAR assunptions and have been analytically demonstrated adequate to maintain a minimum DNBR of 1.30 throughout each analyzed transient.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months periodic surveillance of these parameters though in-strument readout is sufficient to ensure that the parmneters are restored within their iimits following load changes and other expected transient operation. The 18 month periodic measurement of the RCS total flow rate is adequate to detect flew degradation and ensure correlation.of the flow indication channels with measured flow such that the indicated percent flow will provide sufficient verification o ~ flow rate on a 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months basis.

FARLEY - UI,.'T 1 B 3/4 2-6

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DESIGN FEAT'.RES DESIGN PRESSURE AND TEMPERATURE 5.2.2 The reactor containment building is designed and shall be main-tained for a maximum internal pressure of 54 psig and a temperature of 280*F.

5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 157 fuel assemblies with each fuel assembly containing 264 fuel rods clad with Zircaloy -4.

Each fuel rod shall have a ncminal active fuel length of 144 inches and contain a maximum total weight of 1766 grams uranium.

The initial core loading shall have a maximum enrichment of 3.2 weight percent U-235.

Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum enrichment of 3.5 weight percent U-235.

CONTROL ROD ASSEMBLIES 5.3.2 The reactor core shall contain 48 full length and no part length control rod assemblies. The full length control rod assemblies shall contain a nominal 142 inches of absorber material.

The nominal values of absorber material shall be 80 percent silver,15 percent indium and 5 percent cadmium. All control rods shall be clad with stainless steel tubing.

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

FARLEY - UNIT 1 5-4 Amendment No.10

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Figure 6.2-2 Facility Organization - Joseph M. Earley - Unit No.1

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Does not include the licensed Senior Reactor Operator or Se.r.ior Reactor Operator' Limited to Fuel Handling, supervising :0F.E ALTERATIONS.

Shift crew composition (including an individual qualified in radiation protection procedures) may be less than the minimum req.irements for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unexpected absence of on duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements of Table 6.2-1.

FARLEY - UNIT I 6-4

ADMINISTRATIVE CONTROLS 6.3 FACILITY STAFF QUALIFICATIONS 6.3.1 Each member of the facility staff shall meet or exceed the minimum qualificatic.s of ANSI N18.1-1971 for comparable positions, except for the Chemistry and Health Physics Supervisor who shall meet or exceed the qualifications' of Regulatory Guide 1.8, September 1975.*

6.4 TRAINING 6.4.1 A retraining and replacement training program for the facility staff shall be maintained under the direction of the Training Supervisor and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and Appendix "A" of 10 CFR Part 55.

6.4.2 A training program for the Fire Brigade shall be maintained under the direction of the Training Supervisor and shall meet or exceed the requirements of Section 27 of the NFPA Code-1976.

6.5 REVIEW AND AUDIT 6.5.1 PLANT OPERATIONS REVIEW COMMITTEE (PORC)

FUNCTION 6.5.1.1 The PORC shall function to advise the Plant Manager on all matters related to nuclear safety.

COMPOSITION 6.5.1.2 The PORC shall be composed of the:

Chairman:

Plant Manager Vice Chairman:

Assistant Plant Manager Member:

Technical Superintendent Member:

Operations Suerintendent Member: (Non-Voting)

Plant Quality Assurance Engineer Member:

Maintenance Superintendent ALTERNATES 6.5.1.3 All alternate members shall be appointed in writing by the PORC Chairman to serve on a temporary basis; however, no mc e tnan ene alter-nate shall participate as voting members in PORC activities at any :ne time.

The Minimum qualifications requirement for the Chemistry and Healtn Physics Supervisor shall become effective when the initial incumbent in this position is replaced.

FARLEY - UNIT 1 6-5 Amendment No. 4,10

D!"!STRLTI"E C0"TRC' 5 MEETING FREQUENCY 6.5.1.4 The PORC shall meet at least once per calendar month and as convened by the PORC Chairman or Vice Chairman.

QUORUM 6.5.1.5 A quorum of the PORC shall consist rf the Chairman or Vice Chairman and two members including alternatrs.

RESPONSIBILITIES 6.5.1.6 The PQRC shall review:

a.

All administrative procedures and changes thereto,.

b.

The safety evaluations for 1) procedures, 2) changes. to procedures, equipment or systems and 3) tests or experi-ments completed under the provision of Section 50.59, 10 CFR, to verify that such actions did not constitute an unreviewed safety question.

c.

Proposed procedures and changes to procedures, equipment or systems which may involve an unreviewed safety question as defined in Section 50.59,10 CFR.

d.

Proposed tests or experiments which may involve an unreviewed

. safety question and defined in Section 50.59, 10 CFR.

e.

Proposed changes to Technical Specifications or this Opera-ting License.

f.

Reports of violations of codes, regulations, orders, Techni-cal Specifications, license requirements, or of internal procedures of instructions having nuclear safety significance or abnormal degradation of systems designed to contain radio-active material.

g.

Reports of significant operating abnormalities or deviations from normal and expected performance of plant equipment that affect nuclear safety.

h.

All written reports concerning events recuirinc 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months notif'-

cation to the Comission.

i.

All recognized indications of an unanticipated deficiency in some aspect of design or operation of safety related struct.:res, systems, or components.

FARLEY - UNIT 1 6-6

ADMINISTPATIVE CONTROLS a.

Minutes of each NORB meeting shall be prepared, approved and forwarded to the Senior Vice President within 14 days following each meeting.

b.

Reports of reviews encompassed by Section 6.5.2.7 above, shall be prepared, approved and forwarded to the Senior Vice President within 14 days following completion of the review, c.

Audit reports encompassed by Section 6.5.2.8 above, shall be forwarded to the Senior Vice President and to the management positions responsible for the areas audited within 30 days after completion of the audit.

6.5.3 TECHNICAL REVIEW AND CONTROL ACTIVITIES

6. 5. 3.1 Activities which affect nuclear safety shall be conducted as follows:

a.

Procedures required by Technical Specification 6.8 and other procedures which affect plant nuclear safety, and changes (other than editorial or typographical changes) thereto, shall be prepared, reviewed and approved.

Each such procedure or procedure change shall be reviewed by an individual / group other than the individual / group which prepared the procedure or pro-cedure change, but who may be from the same organization as the individual / group which prepared the procedure or procedure change.

Procedures other than Administrative Procedures will be approved by either the Technical Superintendent, the Opera-tions Superintendent, the Maintenance Superintendent or the '

Assistant Plant Manager as applicable. The Plant Manager will approve administrative procedures, security implementing procedures and emergency plan implementing procedures, Tempo-rary changes to procedures which clearly do not change the intent of the approved procedures will be approved by two members of the plant staff, at least one of whom holds a Senior Reactor Operator's License.

For changes to procedures which may involve a change in intent of the approved procedures, the person authorized above to approve the procedure shall ap-prove the change.

b.

Proposed changes or modifications to plant nuclear safety,-

.related structures, systems and components shall be reviewed as designated by the Plant Manager.

Ea:h such modification shall be reviewed by an individual /grou: other than the individual / group which designed the modification, but who may be from the same organization as the individual / group which designed the modifications.

Proposed r6difications to plant nuclear safety-related structures, systems and components shall be approved prior to implementation by the Plant Manager.

FARLEY - UNIT 1 6-11 Amendment No. f, 10

ADMIf:ISTRATIVE CONTROLS c.

Proposed tests and experiments which affect plant nuclear safety and are not addressed in the Final Safety Analysis Report shall be prepared, reviewed, and approved.

Each such test or experi-ment shall be reviewed by an individual / group other than the individual / group which prepared the proposed test or experiment.

Proposed test and experiments shall be approved before implementa-tion by the Plant Manager.

d.

Occurrences reportable pursuant t' the Technical Specification 6.0 and violations of Technical 'pecifications shall be investi-gated and a report prepared whicn evaluates the occurrence and which provides recommendations to prevent recurrence..Such re-ports shall be approved by the Plant Manager and forwarded to the Manager of Nuclear Generation; and to the Chainnan of the Nuclear Operations Review Board.

Individuals responsible for reviews performed in a'ccordance with e.

6.5.3.1.a, 6.5.3.1.b. 6.5.3.1.c and 6.5.3.1.d shall be members of the plant supervisory staff previously designated by the Plant Manager.

Each such review shall include a determination of whether or not additional, cross-disciplinary, review is necessary.

If deemed necessary, such review shall be performed by the review per-sonnel of the appropriate discipline.

f.

Each review will include a determination of whether or not an unreviewed safety question is involved.

Pursuant to 10 CFR 50.59 NRC approval of items involving unreviewed safety question will be obtained prior to Plant Manager approval for implementation.

RECORDS

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6.5.3.2 Records of the above activities shall be provided to the Plant Manager, PORC and/or NORB as necessary for required reviews.

6.6 REPORTABLE OCCURRENCE ACTION 6.6.1 The following actions shall be taken for REPORTABLE OCCURRENCES:

a.

The Commission shall be notified ar.d/or a report submitted pursuant to the requirements of Specification 6.9.

b.

Each REPORTABLE OCCURRENCE requiring 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months notification to the Commission shall be reviewed by the POP" and sub-mitted to the NORB and the Manager of Nuclear Gene-ator.

FARLEY - UNIT 1 6-12

ML19261C340

Text

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