Text

Proposed Tech Specs,Revising Sections 3,4 & 6 as Supplemental Response to Generic Ltr 83-43 Re LER Reporting Requirements of 10CFR50.72 & 50.73
	ML20099L329
Person / Time
Site:	Farley
Issue date:	11/26/1984
From:	; ALABAMA POWER CO.
To:	;
Shared Package
ML20099L319	List: ML20099L315; ML20099L329;
References
	GL-83-43, NUDOCS 8412010088
	Download: ML20099L329 (106)
	v • d • e

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INDEX DEFINITIONS.

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SECTION- PAGE

. 1.0 DEFINITIONS I

1.1 ACTION ........................................................... 1-1 1.2 AX I AL FL UX D I FF ERE NCE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 T.3 CHANNEL' CALIBRATION .............................................. 1-1

' 1.4 ' CHANNEL CHECK .................................................... 1 1.5 CHANNEL FUNCTION TEST ............................................ 1-1 1.6 CONTA IN MENT INTE GRIT Y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-2 1.7 CONTROLLED LEAKAGE ............................................... 1-2 1.8 CORE ALTERATION ................................................... 1-2 1.9 DOSE EQUIVALENT I-131 ............................................ 1-2 1.10 E-AVERAGE DIS INTEGRATION ENERGY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 4

1.11 ENGINEERED SAFETY FEATURES RESPONSE TIME ......................... 1 -

f 1.12 FREQUENCY NOTATION ............................................... 1-3 4

1.13 CASE 0VS RADWASTE TREATMENT SYSTEM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 1.14 IDENTIFIED LEAKAGE ............................................... 1-3 1.15 LIQUID RADWASTE TREATMENT SYSTEM ................................. 1-4 1.16 MAJOR CHANGES TO RADI0 ACTIVE WASTE TREATMENT SYSTEMS. . . . . . . . . . . . . . 1-4 1.17 0FFSITE DOSE CALCUL ATION MANU AL (0DCM) . . . . . . . . . . . . . . .' . . . . . . . . . . . . 1-4 1.18.0PERABLE - OPERABILITY ........................................... 1-4 1.19 OPE RATION AL MOD E - MO DE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5 1.20 PHYSICS TESTS .................................................... 1-5 1.21 PRESSURE BOUNDARY LEAKAGE ........................................- 1-5 1.22 PROCESS CONTROL PROGRAM (PCP) .................................... 1-5 !

1.23 PURGE-PURGING .................................................... 1-5

1. 24 QUAD RANT POW ER T ILT RATI O . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5 1.25 RATED THERMAL POWER .............................................. 1-5 1.26 REACTOR . TRIP SYSTEM RESPDNSE TIME . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-6 1.27 REPORTABLE EVENT .................................................- 1-6 1.28 SHUTDOWN MARGIN .................................................. 1-6 i 1.29~ SOLIDIFICATION ................................................... 1-6 1.30 SOURCE CHECK ..................................................... 1-6 4

1.31 STAGGERED TEST BASIS ............................................. 1-6 1.32 THERMAL POWER . .................................................. 1-7 1.33 UNIDENTIFIED LEAKAGE .............................................

1-7

1. 34 VENTIL ATION EXHAUST TREATMENT SYSTEM . . . . . . . . . . . . . . . . . . . . . . . . . , . . . 1-7 1.35 VENTING .......................................................... 1-7

. TABLE 1.1 OP ERAT I ON AL M ODE S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-8 TABLE - 1. 2 FRE QU ENCY NGTAT ION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

1-9 i

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- FARLEY-UNIT 1 I AMENDMENT:NO.

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.m._ . , , _ . , _ , . _ . . . . - . . . , ,.-_..._ u _ _ _ . , _ _ , . . . .

1 INDEX ADMINISTRATIVE CONTROLS

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

. 6-10 Audits ........................................................... 6-11 Au th o r i ty . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-12 Records ......................................................... 6-12

. 6.5.3 TECHNICAL REVIEW AND CONTROL Activities ...................................................... 6-12 Records ..................................................'......, 6-13 6.6 REPORTABLE EVENT ACTION ........................................... 6-14 16.7 SAFETY LIMIT VIOLATION ............................................ 6-14 6.8 PROCEDURES AND PROGRAMS ........................................... 6-14 6.9 REPORTING REQUIREMENTS 6.9.1 ROUTINE REPORTS Startup Report ................................................... 6-15a An nu al Re po r t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 Annual Radiological Environmental Operating Report . .. . .... . .... .. 6-17 Semi annual Radi oacti ve Ef fluent Release Report . . . . . . . . . . . . . . . . . . . 6-17 Mon thly . 0pe ra ti ng Repo rt ~ . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .'. 19.

Radi al Peaki ng Factor Limi t Report . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 6-19 Annual Diesel Generator Reliabil ity Data Report . . . . . . . . . . . . . . . . . . 6-19 Annual Reactor Coolant' System Specific Activity Report ........... 6-20 Annual Seal ed Source Leakage Report . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-20 6.9.2 SPECIAL REPORTS .................................................. 6 6.10 RECORD RETENTION .................................................. ~6-20

'6.11 RADIATION PROTECTION PR0 GRAM ......................................- 6-21 6.12 HIGH RADIATION AREA ............................................... 6-22 1FARLEY-UNIT'1 XIX AMENDMENT:NO.

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- 1 INDEX ADMINISTRATIVE CONTROLS j SECTION PAGE 6.13 PROCESS CONTROL' PROGRAM .....................~..................... 6-23 -

6.14 0FFSITE-DOSE CALCULATION MANUAL .................................. 6-23 6.15 MAJOR CHANGES TO RADI0 ACTIVE WASTE TREATMENT SYSTEMS . . . . . . . . . . . . . 6-24 i_

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FARLEY-UNIT 2 XX AMENDMEMT NO.

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DEFINITIONS REACTOR TRIP SYSTEM RESPONSE TIME 1.26 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until

- loss of stationary gripper coil voltage.

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

SHUTDOWN MARGIN 1.28 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical froa its present condition assuning 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 withdrawn.

SOLIDIFICATION 1.29 SOLIDIFICATION shall be the conversion of radioactive wastes from liquid i systems to a homogeneous (uniformly distributed), monolithic, immobilized solid with definite volume and shape, bounded by a stable surface of distinct outline on all sides (free-standing).

, SOURCE CHECK 4 1.30 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.

STAGGERED TEST BASIS 1.31 A STAGGERED TEST BASIS shall consist of:

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

b. The testing of one system, subsystem, train or other designated i

component at the beginning of each subinterval.

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FARLEY-UNIT 1 1-6 AMENDMENT NO.

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREENTS 3/4.0 - APPLICABILITY LIMITING CONDITION FOR OPERATION 3.0.1 Compliance with the Limiting Conditions for Operation contained in the succeeding specifications is required during the OPERATIONAL MODES or other conditions specified therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be met.

3.0.2 A condition prohibited by the Technical Specifications shall exist when- {

the requirements of the Limiting Condition for Operation and associated ACTION requirements are not met within the specified time intervals. If the Limiting Condition for Operation is restored prior ',o the expiration of the specified time intervals, completion of the ACTION requirements is not required.

3.0.3 When a Limiting Conditi for Operation is not met, except as provided in the associated ACTION requirements, within one hour ACTION shall be initiated to place the unit in a MODE in which the specification does not apply by placing it, as applicable, in:

1. At least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,

2. At least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and

3. At least COLD SHUTOOWN within the subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Where corrective measures are completed that permit operation under the ACTION requirements, the ACTION may be taken in accordance with the specified time limits as measured from the time of failure to meet the Limiting Condition for Operation. Exceptions to these requirements are stated in the individual specifications.

3.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the conditions of the Limiting Condition for Operation are met without reliance on provisions contained in the ACTION requirements. This provision shall not prevent passage through OPERATIONAL MODES as required to comply with ACTION requirements. Exceptions to these requirements are stated in the individual specifications.

3.0.5 When a system, subsystem, train, component or device is detennined to be inoperable solely because its emergency power source is inoperable, or solely because its normal power source is inoperable, it may be considered OPERABLE for the purpose of satisfying the requirements of its applicable Limiting Condition for Operation, provided: (1) its corresponding normal or emergency power source is OPERABLE; and (2) all of its redundant system (s), subsystem (s), train (s),

component (s) and device (s) are OPERABLE, or likewise satisfy the requiremerits of this specification. Unless both conditions (1) and (2) are satisfied within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , ACTION shall be initiated to place the unit in a MODE in which the applicable Limiting Condition for Operation does not apply by placing it, as applicable, in:

FARLEY-UNIT 1 3/4 0-1 AMENDMENT N0.

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REACTIVITY CONTROL SYSTEMS MODERATOR TENPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION.

3.1.1.3 The moderator temperature coefficient (MTC) shall be:

a. Less than or equal to 0.5 x 10-4 delta k/k/*F for the all rods withdrawn, beginning of cycle life (BOL), below 70% THERMAL POWER condition. Less than or equal to 0 delta k/k/*F at or above 70%

THERMAL POWER.

b. Less negative than -3.9 x 10-4 delta k/k/*F for the all rods withdrawn, end of cycle life (E0L), RATED THERMAL POWER condition.

APPLICABILITY: Specification 3.1.1.3.a - MODES 1 and 2* only#

Specification 3.1.1.3.b - MODES 1, 2 and 3 only#

l ACTION:

a. With the MTC more positive than the limit of 3.1.1.3.a above, operation l in MODES 1 and 2 may proceed provided:

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1. Control rod withdrawal limits are established and maintained sufficient to restore the MTC to within its limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 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 . These withdrawal limits shall be in addition to the insertion limits of Specification 3.1.3.6.

2. The control rods are maintained within the withdrawal limits established above until a subsequent calculation verifies that the MTC has been restored to within its limit for the all rods withdrawn condition.

3. A Special Report is prepared and submitted to the Commission pursuant to Specification 6.9.2 within 10 days, describing the {

value of the measured MTC, the interim control rod withdrawal limits and the predicted average core burnup necessary for restoring the positive MTC to within its limit for the all rods withdrawn condition.

b. With the liTC more negative than the limit of 3.1.1.3.b above, be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

With Kef f greater than or equal to 1.0

See Special Test Exception 3.10.3 FARLEY-UNIT 1 3/4 1-4 AMENDMENT N0.

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS.(Continued)

Rn

.2. When thexy F c is less than or equal to the F xy limit for the appropriate measured core plane, additional power distribution maps shall be taken and Fxyccompared to FffPand xy F L ' at least once per-31 EFPD.

e. The Fxy limit for RATED THERMAL POWER (FffP) shall be provided for all core planes containing bank "D" control rods and all unrodded core planes in a Radial Peaking Factor Limit Report per Specification 6.9.1.11.

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f. The F xy limits of e, above, are not applicable in the following core planes regions as measured in percent of core height from the bottom of the fuel:

1. Lower core region from 0 to 15%, inclusive.

2. Upper core region from 85 to 100%, inclusive.

3. Grid plane regions at 17.8 i 2%, 32.1 i 2%, 46.412%, 60.6 i 2%

and 74.9 2%, inclusive.

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

g. L4th Fxyc exceeding F y L the effects of Fxy on Fq (Z) shall be evaluated to determine if Fq (Z) is within its limits.

4.2.2.3 When F determinations, an overall measured F qshall (Z) qbe(Z) is measured obtained for other from_a power than map distribution Fxy and increased by-3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty.

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l FARLEY-UNIT 1 3/42-6 AMENDMENT NO.

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INSTRUFENTATION FIRE DETECTION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.9 As a minimum, the fire detection instrumentation for each fire detection zone shown in Table 3.3-12 shall be OPERABLE.

APPLICABILITY: Whenever equipment protected by the fire detection instrument is required to be OPERABLE.

ACTION:

With the number of OPERABLE fire detection instrument (s) less than the minimum number OPERABLE requirement of Table 3.3-12:

a. Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months establish a fire watch patrol to inspect the zone (s) with the inoperable instrument (s) at least once per hour, unless the instrument (s) is located inside the containment, then monitor the containment air temperature at least once per hour at the locations listed in Specification 4.6.1.5.

b. Restore the inoperable instrument (s) to OPERABLE status within 14 days or prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 30 days outlining the action taken, l the cause of the inoperability and the plans and schedule for restoring the instrument (s) to OPERABLE status.

c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.9.1 Each of the above required fire detection instruments which are accessible during plant operation shall be demonstrated OPERABLE at least once per 6 months by performance of a function test which includes subjecting the detector to test aerosol. Fire detectors which are not accessible during plant operation shall be demonstrated OPERABLE by the performance of this functional test during each COLD SHUTDOWN exceeding 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months unless performed in the previous 6 months.

4.3.3.9.2 The NFPA Standard 72D supervised circuits supervision associated with the detector alarms of each of the above required fire detection instruments shall be demonstrated OPERABLE at least once per 6 months.

FARLEY-UNIT 1 3/4 3-59 AMENDMENT NO. I

1 INSTRUMEllTATION RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUfENTATION LIMITING CONDITION FOR OPERATION 3.3.3.10 The radioactive liquid effluent monitoring 'i.1strumentation channels shown in Table 3.3-13 shall be OPERACLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.11.1.1 are not exceeded. The alarm /

trip setpoints of these channels shall be determine + in accordance with the OFFSITE DOSE CALCULATION MANUAL (0DCM).

APPLICABILITY: At all times.

ACTION:

a. With a radioactive liquid effluent monitoring instrumentation channel alarm / trip setpoint less conservative than required by the above specification, immediately suspend the release of radioactive liquid effluents monitored by the affected channel or declare the channel inoperabl e.

b. With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-13.

c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.10 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3-8.

FARLEY-UNIT 1 3/4 3-61 AMENDMENT NO.

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4 INSTRUMENTATION RADI0 ACTIVE GASEGUS EFFLUENT MONITORING INSTRUMENTATION

' LIMITING CONDITION FOR OPERATION 3.3.3.11 The radioactive gaseous effluent monitoring instrumentation channels shown in Table 3.3-14 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.11.2.1 are not exceeded. The alarm /

trip setpoints of these channels shall be determined in accordance with the ODCM.

APPLICABILITY: As shown in ' Tabl e 3.3-14.

ACTION:

a. With a radioactive gaseous effluent monitoring instrumentation channel alarm /Arip setpoint less conservative than required by the above Specification, immediately suspend the release of radioactive gaseous effluents monitored by the affected channel or declare the channel inoperable.

b. With less than the minimum number of raaioactive gaseous effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Tabl e 3.3-14.

c. The provir. ions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.11 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3-9.

FARLEY-UNIT 1 3/4 3-66 AMENDMENT N0.

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-REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

I

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

b. The steam generator shall be determined OPERABLE after completing the corresponding actions (plug all tubes exceeding the plugging limit and all tubes containing through-wall cracks) required by Table 4.4-2.

4.4.6.5 , Reports

a. Following each inservice inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the Commission within 15 days.

b. The complete results of the stean generator tube inservice inspection shall be submitted to the Commission in a Special Report pursuant to Specification 6.9.2 within 12 months following the completion of the inspection. This Special Report shall include:

1. Number and extent of tubes inspected.

2. Location and percent of wall-thickness penetration for each indication of an imperfection.

3. Identification of tubes plugged.

c. Results of steam generator tube inspections which fall into Category C-3 shall be considered a REPORTABLE EVENT and shall be reported pursuant to 10CFR50.73 prior to resumption of plant operation. The written report shall provide a description of investigatior.s conducted to determine the cause of the tube degradation and corrective measures taken to prevent recurrence.

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FARLEY - UNIT 1 3/4 4-13 AMENDMENT N0.

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TABLE 4.4-2 STEAM GENERATOR TURE INSPECTION 5o G IST SAPPLE INSPECTION 2ND SAf9LE. INSPECTION 3RD SAWLE INSPECTION

?

E Sample Size Resul t Action Required Resul t Action Required Resul t Action Required Z

y A minimum of C-1 None N/A N/A S Tubes per u N/A N/A S.G. C-2 Plug defective C-1 None N/A N/A tubes and inspect additional 25 C-2 Plug defective C-1 None tubes in this S.G. tubes and inspect additional 45 tubes C-2 Plug defective in this S.G. tubes C-3 Perform action for

{ C-3 result of first T sample

$ C-3 Perform action for N/A N/A C-3 result of first sample C-3 Inspect all tubes All other None N/A N/A in this S.G. , S.G.s are plug defective C-1 tubes and inspect 25 tubes in each Some S.G.s Perform action for N/A N/A other S.G. C-2 but no C-2 result of additional second sample y S.G.s are M C-3 5

g Notification to Additional Inspect all tubes N/A N/A y NRC pursuant to S.G. is C-3 in each S.G. and 2 10CFRSO.73 plug defective o tubes.

Notification to NRC pursuant to 10CFR50.73 g

S-3N % Where N is the number of steam generators in the unit, and n is the number of steam generators inspected IT during an inspection

' REACTOR COOLANT SYSTEM ACTION: (Continued)

, MODES 1, 2, 3, 4 and 5:

-a'. With the specificf activity of the primary coolant greater than 1.0 microcurie per gram DOSE EQUIVALENT I-131 or greater than 100/E microcuries' per gram, perform the sampling and analysis requirements of item 4e of Table 4.4-4 until the specific activity of the primary coolant is restored to within its limits.

L SURVEILLANCE REQUIREMENTS 4.4.9 The specific activity of the primary coolant shall be determined to be within the limits by performance of the sampling and analysis program of Table 4.4-4.

f FARLEY-UNIT 1 3/4 4-24 AMENDMENT N0.

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CONTAINENT SYSTEMS SURVEILLANCE . REQUIREMENTS - (Conti nued)

b. Removing one wire from each of-a dome, vertical and hoop tendon checked for lift off force and' determining that:

1. The corrosion level over the entire length of the tendon wires has not progressed since the original installation or the previous surveillance.

2. There are no changes in physical appearance of the sheathing filler material.

3. A minimum tensile strength of 240,000 pounds per square inch for 7.t l least three wire samples (one from each end and one at mid-length) !

cut from each removed wire. Failure of any one of the wire samples to meet the minimum tensile stre rth test is evidence of abnonnal degradation of the containment structure.

3 I

r 4.6.1.6.2 End Anchorages and Adjacent Concrete Surfaces The structural integrity of the end anchorages and adjacent concrete surfaces shall be demonstrated by determining through inspection that no adverse changes have occurred in the visual appearance of the end anchorage concrete exterior surfaces or the concrete crack patterns adjacent to the end anchorages.

l Inspections of the concrete shall be containment leakage rate tests only (performed during the first4.6.1.2) raference Specification Type Awhile the containment is at its maximum test pressure.

4.6 .1.6 .3 Liner Plate The structural integrity of the containment liner plate shall be determined during the shutdown for the first Type A containment leakage i rate test only (reference Specification 4.6.1.2) by a visual inspection of the !

plate and verifying no adverse changes in appearance or other abnormal degradation.

FARLEY-UNIT 1 3/4 6-9 AMENDMENT N0.

.h _ .

' PLANT SYSTEMS SURVEILLANCELREQUIREMENTS (Continued)

^

. 1. With a' half-life: greater than 30 -dayst(excludin'g Hydrogen '3), and 2.< In any form other than-. gas. . .

b. Stored sources not in use - Each sealed : source and fission detector shall be tested prior to use or transfer to another licensee unless tested within' the previous six monthr.. Sealed sources and fission detectors transferred without a certificate indicating the last' test date shall be tested prior to being placed -into use.

c. Startup sources and fission detectors - Each _ sealed startup source and c fission detector shall be tested within 31 days prior to being

subjected to core flux or installed in the core and following repair or 4

maintenance to'the source.

.ji i

J

\

FARLEY-UNIT 1 3/4 7-81 AMENDMENT NO.

v

, b

w'-l e

PLANTSYST(MS 3/4.7.11 FIRE' SUPPRESSION SYSTEMS-FIRE' SUPPRESSION WATER SYSTEM

' LIMITING' CONDITION FOR 00ERATION

.3.7.11.1-The fire suppression water system shall be OPERABLE with:

-a. Two high pressure pumps,: each with a capacity of 2500 gpm, with their discharge aligned to.the fire suppression header,

b. Separate water supplies, each with a minimum contained volume of.

250,000 ' gallons, and ,

c. An OPERABLE . flow path capable of taking suction from'each tank and j transferring' the water through distribution piping with OPERABLE- i sectionalizing control or isolation valves to the yard hydrant curb valves, the last valve ahead of the water flow alarm-device.on each -

sprinkler or hose standpipe, and the last valve ahead of the deluge valve onLeach deluge or spray system required to be_0PERABLE per Specifications 3.7.11.2, 3.7.11.4 and 3.7.11.5.

APPLICABILITY: ' At all times.

ACTION:

a. With one of the above requitad pumps and/or water supplies inoperable, restore the inoperable equipaint to OPERABLE status within 7 days or prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 30 days outlining the plans and procedures to be used to provide for the loss of redundancy in this system. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

b. With the fire suppression water system otherwise inoperable:

1. Establish a backup fire suppression water system within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and

2. Submit a Special Report in accordance with Specification 6.9.2:

a) By telephone within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b) Confirmed by telegraph, mailgram or fascimile transmission no later than the first working day following the event, and c) In writing within 14 days following the event, outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

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

o

PLANT SYSTEFG SPRAY AND/0R SPRINKLER SYSTEMS LIMIlING CONDITION FOR OPERATION 3.7.11.2 The spray and/or sprinkler systems listed in Table 3.7-5 shall be OPERABLE.

APPLICABILITY: Whenever equipment-in the spray / sprinkler protected areas is required to be OPERABLE.

ACTION:

a. With one or more of the above required spray and/or sprinkler systems ;

inoperable, within one hour establish a continuous fire watch with j backup fire suppression equipment for those areas in which redundant systems or components could be damaged; for other areas, establish an hourly fire watch patrol. Restore the system to OPERABLE status within 14 days or prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 30 days outlining the {

action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

l l

b. The provisions of Specification 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.11.2 Each of the. above required spray and/or sprinkler systems shall be i demonstrated OPERABLE:

a. At least once per 31 days by verifying that each vaive (manual, power operated or automatic) in the flow path is in its correct position.

b. At least once per 12 months by cycling each testable valve in the flow path through at least one complete cycle of full travel.

FARLEY-UNIT 1 3/4 7-85 AMENDMENT N0.

s

~

- PLANT' SYSTEMS ~

,' ' CO2__ SYSTEMS .

l1 . l LIMITING CONDITION FOR OPERATION 3.7.1153 The following hig'h pressire and low pressure C0 2 systems shall be

~ 0PERABLE.

1a. - Service Water Intake S*,ructure (each '4160 volt bus and each 600 volt load center) - HP.-

'b. Turbine Building 13 ton unit' and distribution system in the Auxiliary Building - L.P. -

~

c. Diesel Building'5 ton unit and distribution system.

~ APPLICABILITY: Whenever. equipment protected by the CO2 systems is required to be OPERABLE.-

JACTION:

a. With one or more of the above required'C02 systems inoperable, within . '

one hour establish a continuous fire watch with backup. fire suppression equipment for those areas in which redundant systems or components could be damaged; for other areas, establish an hourly fire watch .

patrol . Restore the system to OPERABLE status within 14 days or prepare and submit a Special Report to the Comission pursuant to Specification 6.9.2 within the next 30 days outlining the action taken, .

the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status,

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

. SURVEILLANCE REQUIREMENTS 4.7.11.3.1 Each of the above required CO2 systems shall be demonstrated OPERABLE at least once per 31 days by verifying that each manual valve in -the flow ath is in ~its correct position.

4.7.11.3.2 Each of the above required low pressure C02 systems shall be demonstrated OPERABLE:

i

a. At least once per 7 days by verifying the CO2 storage tank level to be l greater than 50% and pressure to be greater than 233 psig, and )

1

b. At_least once per 18 months by verifying: '

1.- The system valves and associated ventilation dampers and fire door release mechanisms actuate manually and automatically, upon receipt of a simulated actuation signal, and 2.. Flow from each nozzle during a " Puff Test."

FARLEY-UNIT;l 3/4 7-88 . AMENDMENT NO.

I

. d

PLANT SYSTENG FIRE HOSE STATIONS LIMITING CONDITION FOR OPERATION 3.7.11.4 The fire hose stations shown in Table 3.7-6 shall be OPERABLE.

APPLICABILITY: Whenever equipment in the areas protected by the fire hose stations is required to be OPERABLE.

ACTION:

a. With one or more of the fire hose stations shown in Table 3.7-6 inoperable, route

an additional equivalent capacity fire hose to the unprotected area (s) from an OPERABLE hose station within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months if the inoperable fire hose is the primary means of fire suppression; otherwise route the additional hose within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Restore the fire hose station to OPERABLE status within 14 days or prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within -l the next 30 days outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the station to OPERABLE status.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.11.4 Each of the fire ho:e stations shown in Table 3.7-6 shall be demonstrated OPERABLE:

a. At least once per 31 days by visual inspection of the fire hose stations accessible during plant operation to assure all required equipment is at the station.

b. At least once per 18 months by:

1. Removing the hose for inspection and re-racking, and

2. Inspecting all gaskets and replacing any degraded gaskets in the couplings.

c. At least once per 3 years by:

1. Partially opening each hose station valve to verify valve OPERABILITY and no flow blockage.

2. Conducting a hose hydrostatic test at a pressure at least 50 psig greater than the maximum pressure available at that hose station.

lt routing of the hose would require rendering a fire barrier penetration inoperable, hose will be routea up to but not through the penetration with sufficient hose length to reach the unprotected area (s).

FARLEY-UNIT 1 3/4 7-50 AMENDMENT NO.

PLANT SYSTEMS YARD FIRE IlYDRANTS AND HYDRANT HOSE HOUSES LIMITING CONDITION FOR CPERATION 3.7.11.5 The yard fire hydrants and associated hydrant hose houses shown in Table 3.7-7 shall be OPERABLE.

APPLICABILITY: Whenever equipment in the areas protected by the yard fire hydrants is required to be OPERABLE.

ACTION:

a. With one or more of the yard fire hydrant or associated hydrant hose houses shown in Table 3.7-7 inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months have sufficient additional lengths of 21/2 inch diameter hose located in an adjacent OPERABLE hydrant hose house to provide service to the unprotected area (s) if the inoperable fire hydrant or associated hydrant hose house is the primary means of fire suppression; otherwise provide the additional hose within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Restore the hydrant or hose house to OPERABLE status within 14 days or prepare and submit a Special Report ,

to the Commission pursuant to Specification 6.9.2 within the next 30 '

days outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the hydrant or hose house to OPERABLE status.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.11.5 Each of the yard fire hydrants and associated hydrant hose houses shown in Table 3.7-7 shall be demonstrated OPERABLE:

a. At least once per 31 days by visual inspection of the hydrant hose house to assure all required equipment is at the hose house.

b. At least once per 6 months (once during March, April or May and once during September, October or November) by visually inspecting each yard fire hydrant and verifying that the hydrant barrel is dry and that the hydrant is not damaged.

1

c. At least once per 12 months by:

1. Conducting a hose hydrostatic test at a pressure at least 50 psig greater than the maximum pressure available at any yard fire hydrant.

2. Inspecting all the gaskets and replacing any degraded gaskets in the couplings.

3. Performing a flow check of each hydrant to verify its OPERABILITY.

FARLEY-UNIT 1 3/4 7-92 AMENDMENT N0.

)

PLANT SYSTEMS -

3/4.7.12 FIRE L ARRIER PENETRATIONS LIMITING CONDITION FOR OPERATION 3.7.12 All fire barrier penetrations (including cable penetration barriers, firedoors and fire dampers) in fire zone boundaries protecting safety related areas shall be functional.

APPLICABILITY: At all times.

ACTION:

a. With one or more of the above required fire barrier penetrations non-functional, within one hour either, establish a continuous fire watch on at least one side of the affected penetration, or verify the OPERABILITY of fire detectors on at least one side of the non-functional fire barrier and establish an hourly fire watch patrol .

Restore the non-functional fire barrier penetration (s) to functional status within 7 days or prepare.and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 30 days outlining the action taken, the cause of the non-functional penetration and plans and schedule for restoring the fire barrier penetration (s) to functional status,

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicaole.

SURVEILLANCE REQUIREMENTS 4.7.12 Each of the above required fire barrier penetrations shall be verified to be fuinctional:

a. At least once per 18 months by a visual inspection.

b. Prior to returing a penetration fire barrier to functional status following repairs or maintenance by performance of a visual inspection of the affected penetration fire barrier (s).

FARLEY-UNIT 1 3/4 7-94 AMEN 0 MENT N0.

ELECTRICAL POWER SYSTEMS St'RVEILLANCE REQUIREMENTS (Continued) b) Coolant Temperature High (CTH) c) Co;1 ant Pressure Low (CPL) d) Crankcase Pressure High (CCPH)

11. Verifying'the capability to reject a load of greater than or equal to the largest single load associated with that diesel generator (approximately 1000 kw); while maintaining voltage between 3740 and 4580 volts and speed less than or equal to 75% of the difference between nominal speed and the overspeed trip setpoint.

d. At least once per 10 years or after any modifications which could affect diesel generator interdependence by starting the diesel ,

generators simultaneously, and verifying that the diesel generators 1 accelerate to at least 900 rpm, for the 2850 kw generator and 514 rpm for the 4075 kw generator, in less than or equal to 12 seconds.

e. At least once per 5 years, on a staggered basis, by verifying that the diesel generator can reject a load of 1200-2400 kw without tripping.

The diesel generator output breaker (s) must remain closed such that the diesel generator is connected to at least one emergency bus. Verify that all fuses and breakers on the energized. emergency bus (es) are not tripped. The generator voltage shall remain within 3330 and 4990 volts during and following the load rejection.

l l I l

l FARLEY-UNIT 1 3/4 8-6 AMENDMENT NO.

Table 4.11-1 (Centinu:d)-

' TABLE NOTATION

b. A' composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen.which is representative of the liquids released.

c. A batch release is the discharge of liquid wastes of a discrete volume. Prior to sampling for analyses, each batch shall be isolated, and then thoroughly mixed, by a method described in the ODCM, to assure representative sampling. .

d. A continuous release is the discharge of liquid wastes of a nondiscrete volme; e.g., from a volme of system that has an input flow during the effluent release.

e. The principal gamma emitters for which the MDC specification applies exclusively are the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, 2n-65, - Mo-99, Cs-134, C9137, Ce-141, and Ce-144. This Ifst does not noean that only these nuclides are to be detected ~and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and _ reported.

f. Sampling will be performed only if the effluent will' be discharged to the environment.

g. Deviation from the M)C requirements of Table 4.11-1 shall be reported per Specification 6.9.1.8.

FARLEY-UNIT 1 3/4 11-4 AMENDMENT N0.

RADI0 ACTIVE EFFLUENTS

- DOSE LIMITING CONDITION FOR OPERATION 3.11.1.2 The dose or dose commitment to an individual from radioactive materials

in liquid effluents released, from each reactor unit, from the site-(see Figure 5.1-4) shall be . limited

a. During any calendar quarter to less than or equal to 1.5 mrem to the total body and to less than or equal to 5 mrem to any organ, and

b. During any calendar year to less than or equal to 3 mrem to the total body and to less than or equal to 10 mrem to any organ.

APPLICABILITY: At all times. -

ACTION:

a. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the causc(s) for exceeding the limit (s) and defines the corrective actions to be taken to reduce the releases of radioactive materials in liquid effluents during the remainder of the current calendar quarter and during the remainder of the current calendar year, so that the cumulative dose or dose commitment to an individual from these releases is within 3 mrem to the total body and i 10 mrem to any organ,

b. The provisions of specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.1.2 Dose Calculations. Cumulative dose contributions from liquid effluents shall be determined in accordance with the ODCM at least once per 31 days.

4 FARLEY-UNIT 1 3/4 11-5 AMENDMENT NO.

RADI0 ACTIVE EFFLUENTS

. LIQUID WASTE TREATMENT LIMITING CONDITION FOR OPERATION 3.11.1.3 The LIQUID RADWASTE TREATMENT SYSTEM shall be OPERABLE. The appropriate portions of the system shall be used to reduce the radioactive materials in liquid wastes prior to their discharge when the projected doses due to the liquid effluent from the site (see Figure 5.1-4) when averaged over the calendar quarter would exceed 0.18 mrem to the total body or 0.6 mrem to any organ.*

APPLICABILITY: At all times.

ACTION:

a. With the LIQUID RADWASTE TREATMENT SYSTEM inoperable for more than 31 days or with radioactive liquid waste being discharged without treatment and in excess of the abova limits, prepare and submit to the Commission within 30 days pursuant to Specification 6.9.2 a Special Report which includes the following information:

1. Identification of the inoperable equipment or subsystems and the reason for inoperability,

2. Action (s) taken to restore the inoperable equipment to OPERABLE status, and-

3. Summary description of action (s) taken to prevent a recurrence.

b. The provisions of specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.1.3.1 Doses due to liquid releases to unrestricted areas shall be projected, based on computed operational history, at least once per 31 days, in accordance with the ODCM.

4.11.1.3.2 The LIQUID RADWASTE IREATMENT SYSTEM shall be demonstrated OPERABLE by operating the LIQUID RADWASTE TREATMENT SYSTEM equipment for at least 15 l minutes at least once per 92 days unless the LIQUID RADWASTE TREATMENT SYSTEM equipment has been utilized to process radioactive liquid effluents during the previour, '2 days.

Per reactor unit i

FARLEY-UNIT 1 3/4 11-6 AMENDMENT NO.

i L--

e . .

y ,

p ?

-Table 4'.11-2 (Continu:d)

TABLE NOTATION

-b, Analyses shall also be perfonned following shutdown from >15% RATED THERMAL POWER, startup to >15% RATED THERMAL. POWER or a RRMAL POWER '

E change exceeding 15 percent of the RATED THERMAL POWER-within a one

. hour period.

c. Tritium grab samples.shall be taken' from the- plant vent stack at least

. once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the refueling canal is flooded. >

c - d. Samples shall be changed at least once per 7 days and analyses shall be

completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after changing (or after. removal fro;n sampler).~ Sampling shall:also be perfomed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months _;

for at least 2 days following each shutdown from >15% RATED THERMAL.

POWER, startup to >15%' RATED THERMAL POWER or THElD4AL POWER change

, exceeding 15 percelit of RATED THERMAL POWER in one hour and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months of changing. When samples collected for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months are analyzed, the corresponding MDC may be increased by.a factor of 10.

e. Tritium grab samples shall be taken at.least once per 7. days from the ventilation exhaust from the spent fuel pool area, whenever spent fuel is in the spent fuel pool.

f. The ratio of the sample flow rate to the sampled stream flow rate shall

. be known for_ the time period covered by each dose or dose rate i calculation made in accordance with Specifications 3.11.2.1, 3.11.2.2 l and 3.11.2.3.

i

- g. The principal gamma emitters for which the MC specification' applies j exclusively are the following radionuclides: Mn-54, Fe-59, Co-58, ;

Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144 for particulate i emissions. This list does not mean that only these nuclides are to be

detected and reported. Other peaks which are measureable and i identifiable, together with the above nuclides, shall also be j- identified and reported. ,

h. Deviations from HDC requirements of Table 4.11-2 shall be reported per i Specification 6.9.1.8. l

1. A composite particulate sample is one in which the quantity of air sampled is proportional to the quantity of air discharged.' Either a i specimen which is representative of the tir discharged may be i accumulated and analyzed or the individual samples may be analyzed _and

{ weighted in proportion to their respective volume discharged.

I.

j . The principal gamma emitters for which the NC specification applies exclusively are the following radionuclides: Kr-87, Kr-86, Xe-133, Xe-133m, Xe-135 and Xe-138 for gaseous emissions. This does not mean that only these nuclides dre to be detected and ' reported. Other peaks which are measurabl.e and identifiable, together with the above

nuclides, shall also be identified and reported.

- FARLEYdNIT1 3/4 11-11 AMENDMENT NO.

i T

. + . -,%,,,,,-,.y .---,,,y .w-,y ,r, ,. ..gr,-p,v.,-y. . + - .y ,- 9v.-y. ,,m.... ,- w,-cw ...#n..,_w.,,,,e ...-m.,, -mywm,3.,~. ,_,,-e-.

.e RADIOACTIVE EFFLUENTS LDOSE - NOBLE GASES

LIMITING CONDITION FOR OPERATION 3.11.2.2 The air dose due to noble gases released in gaseous effluents,.from each' reactor unit, from the site (see Figure 5.1-3) shall be limited to.the

?: .following:

a. During any calendar' quarter: ~Less'than or equal to 5 mrad for gamma radiation and less than or equal to 10 mrad for beta radiation, and

b. During any calendar year: Less than or' equal to 10 mrad for gamma radiation and less than or equal to 20 mrad for beta radiation.

APPLICABILITY: At all times.

ACTION a.- With the calculated air dose from radioactive noble gases in gaseous

~

effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, .a Special l Report which identifies the cause(s) for exceeding the limit (s) and defines the corrective actions to be taken to reduce the releases of radioactive noble gases in gaseous effluents during the remainder of the current calendar quarter and during the remaindef of the current calendar year, so that the cumulative dose is within 10 mrad for gamma radiation and 20 mrad for beta radiation.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not' applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.2 Dose Calcuiations Cumulative dose contributions for the current calendar quarter and current calendar year shall be determined in accordance with the ODCM at least once per 31 days.

FARLEY-UNIT 1 3/4 11-12 AMENDMENT NO.

I

RADI0 ACTIVE-EFFLUENTS DOSE - RADI0 IODINES,' RADI0 ACTIVE MATERIALS IN PARTICULATE FORM, AND RADIONUCLIDES OTHER THAN NOBLE GASES LIMITING' CONDITION FOR OPFRATION 3.11.2.3 The dose to an individual from radiofodines and radioactive materials in particulate form, and radionuclides (other than roble gases) widi half-lives greater than 8 days in gaseous effluents released, from each reactor unit, from the site' (see Figure 5.1-3) shall be limited to ,the following:

l a. During any calendar quarter: Less than or equal to 7.5 mrem to any organ, and

b. During any calendar year: Less than or equal to 15 mrem to any organ. ,

APPLICABILITY: At all times.

i ACTION:

i a. With the calculated dose from the release of radiciodines, radioactive L

materials in particulate form,.or radionuclides (other than noble gases) with half lives greater = than 8 days, in gaseous effluents exceeding any

+

of'the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies

! the cause(s) for exceeding the limit and_ defines the corrective actions to be taken to reduce the releases of-radioiodines and radioactive materials in particulate form, and radionuclides (other than' nobles j

gases) with half-lives greater than 8 days in gaseous effluents during the remainder of the current calendar quarter and during the remainder

[ of the current calendar year, so that the cumulative dose or dose j commitment to an individual from these releases is within 15 mrem to any organ.

b. The provisions of Specifications 3.0.3 and 3.0.4' are not applicable.

SURVEILLANCE REQUIREMENTS

} -

4.11.2.3 Dose Calculations- Cumulative dose contributions for the current 1

calendar. quarter and current calendar year shall be determined in accordance with l the ODCM at least once per 31 days.

, FARLEY-UNIT 1 3/4 11-13 AMENDMENT NO.

=+

e

~ RADI0 ACTIVE EFFLUENTS GASE0US RADWASTE TREATMENT

. LIMITING CONDITION FOR OPERATION 3.11.2.4 The GASEOUS:RADWASTE TREATMENT SYSTEM and the VENTILATION EXHAUST

TREATMENT SYSTEM shall be OPERABLE. The appropriate portions of the GASE0US

_RADWASTE TREATMENT SYSTEM shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected gaseous effluent air doses due'to gaseous effluent releases fron the site (see Figure 5.1-3), when averaged over the calendar quarter, would exceed 0.6 mrad'for gama radiation and 1.2 mrad for beta radiation.* The appropriate portions of the VENTILATION EXHAUST 1REATMENT. SYSTEM shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected doses due to gaseous effluent releases from the site (see Figure 5.1-3) when averaged _over the calendar quarter would exceed 0.9 mrem to any organ.*

APPLICABILITY: At all - times.

ACTION:

a.- With the GASE0US RADWASTE TREATMENT SYSTEM and/or the VENTILATION EXHAUST TREATMENT SYSTEM inoperable for more than 31 days or with gaseous waste being discharged without treatment and in excess of the above limits,~ prepare and submit to the Comission within 30 days, l pursuant to Specification 6.9.2, a Special Report which includes the following information:

1. Identification of the inoperable equipment or subsystems and the

' reason for inoperability,

2. Action (s) taken to restore the inoperable _ equipment to OPERABLE status, and

3. Summary description of action (s) taken to prevent _a recurrence,

b. The provisions.of Specifications 3.0.3 'and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.4.1 Doses due to gaseous releases from the site shall be projected, based on computed operations history, at least once per 31 days, in accordance with the

~"

ODCM..

4.11.2.4.2 The GASE0US RADWASTE TREATMENT SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM shall be demonstrated OPERABLE by operating the GASEOUS RADWASTE TREATMENT SYSTEM equipment and the VENTILATION EXHAUST TREATMENT SYSTEM equipment for at least 15 minutes, at least once per 92 days unless the appropriate system has been utilized to process radioactive gaseous effluents during .the previous 92 days .

-

Doses are per reactor unit.

FARLEY-UNIT 1 3/4 11-14 AMENDMENT NO.

t

RAul0 ACTIVE EFFLUENTS 3/4.11.3 RADWASTE SOLIDIFICATION LIMITING CONDITION FOR OPERATION 3.11.3 The radwaste solidification system shall be OPERABLE and used, as applicable in accordance with a PROCESS CONTROL PROGRAM, for the SOLIDIFICATION and packaging of radioactive wastes to ensure meeting the requirements of 10CFR Part 20 and of 10CFR Part 71 prior to shipment of radioactive wastes from the site.

APPLICABILITY: At all times.

ACTION:

a. With the packaging requirements of 10CFR Part 20 and/or 10CFR Part 71 not satisfied, suspend shipments of defectively packaged solid radioactive wastes from the-site.

b. With the radwaste solidification system inoperable for more than 31 days, prepare and submit to the Comission within 30 days, pursuant to Specification 6.9.2, a Special Report which includes the following information:

1. Identification of the inoperable equipment or subsystems and the reason for inoperability,

2. Action (s) taken to restore the inoperable equipment to OPERABLE status,

3. A description of the alternative used for SOLIDIFICATION and packaging of radioactive wastes, and

4. Summary description of action (s) taken to prevent a recurrence.

c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.3.1 The radwaste solidification system shall be demonstrated OPERABLE at least once per 92 days by:

a. Operating the radwaste solidification system at least once in the !

previous 92 days in accordance with the PROCESS CONTROL PROGRAM, or

b. Verification of the existence of a valid contract for SOLIDIFICATION to be performed by a contractor in accordance with a PROCESS CONTROL PROGRAM.

FARLEY-UNIT 1 3/4 11-17 AMENDMENT NO.

L _RADI0 ACTIVE EFFLUENTS L

3/4.11.4 TOTAL DOSE LIMITING CONDITION FOR OPERATION

,ee . . - ~ - - -s -- ... - -< .

=

-e w 3.11.4 The dose or dose commitment to any member of-the public, due to releases

~

of rad 10 activity and radiation, from uranium fuel cycle sources shall be limited to less than or equal to 25 mrem to the total body or any organ (except the

. thyroid, which shall be limited to less than or equal to 75 mrem) over 4 consecutive quarters.

i -APPLICABILITY: At all times.

ACTION:

a. -With the calculated doses from the release of-. radioactive materials'in liquid or gaseous effluents exceeding twice 'the Ifmits of Specification 3.11.1.2.a, 3.11.1.2.b, 3.11.2.2.a, 3.11.2.2.b, 3.11.2.3.a, or 3.11.2.3.b, prepare and submit a Special Report to the Director, Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington,

D.C. 20555, within 30 days, which' defines the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding the limits of Specification 3.11.4. This Special Report shall include an analysis which. estimates the. radiation exposure (dose) to a member of the public from uranium' fuel cycle sources (including all effluent pathways and direct radiation) .for a 4 consecutive quarter period that

~

includes the release (s)- covered by' this report. If the estimated i

dose (s) exceeds the limits of Specification 3.11.4, and if the release condition resulting in violation of 40CFR190 has not already been !

i corrected, tre Special Report shall include a request for a variance in accordance with the provisions of 40CFR190 and including the specified

. information of $ 190.11(b). Submittal of the report is considered a l

~

timely request, and a variance is granted until. staff action on the request is complete. The variance only-relates to the limits of 40CFR190, and does not apply in any way to the requirements for dose limitation of 10CFR Part 20, as addressed in other sections of this technical specification.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable, J

SURVEILLANCE REQUIREMENTS 4.11.4 Dose Calculations Cumulative dose contributions from liquid and gaseous effluents shall be determined in accordance with Specifications 4.11.1.2, 4.11.2.2, and 4.11.2.3, 'and in accordance with the ODCM.

l l

i FARLEY-UNIT 1 3/4 11-19 AMEND 4ENT NO.

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I3/4.121 RADIOLOGICAL-ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM-

, LIh1 TING-CONDITION FOR OPERATION 3.12.1 The radiological _ env;ronmental monitoring program shall be conducted as

specified in _ Table 3.12-1.

APPLICABILITY: At all times. 1 ACTION:

a. 'With the radiological environmental monitoring program not being conducted as specified in Table 3.12-1, prepare and submit to the Commission, in the Annual Radi-logical Operating Report, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence.

b. With the level of radioactivity in an environmental sampling medium exceeding the reporting levels of Table 3.12-2 when averaged over any calendar quarter, prepare and submit to the Commission within 30 days from the end of the affected calendar quarter-a Special Report pursuant

~to Specification 6.9.2. When more than one-of the radionuclides in Table 3.12-2 are detected in the sampling medium, this report shall be submitted if:

concentration (1) + concentration (2) + ...> 1.0 limit level -(l) limit level (2)

When-radionuclides other than those in Table 3.12-2 are detected and are the result of plant effluents, this Special Report shall be submitted if l the potential annual dose to an individual is equal to or greater than the calendar year limits of Specifications 3.11.1.2, 3.11.2.2 and 3.11.2.3. This Special Report is not required if the measured level of radioactivity 'was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual-Radiological Environmental Operating Report.

c. With milk ~ or fresh leafy vegetable samples unavailable from one or mn of the sample locations required by Table 3.12-1, prepare and submit to the Commission within 30 days, purscant to Specification 6.9.2, a Special Report which identifies the cause of the unavailability of samples and identifies locations for obtaining replacement samples. The locations from which samples were unavailable may then be deleted from those required by Table 3.12-1, provided the locations from which the replacement samples were obtained are added to the environmental monitoring program as replacement locations.

d. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

FARLEY-UNIT 1 3/4 12-1 AMEN 0 MENT NO.

l-L.

Table 4.12-1 (C:ntinu;d)

. TABLE NOTATION

b. ' LOC for drinking water.

c. - Other peaks which are measurable and identifiable, together with the radionuclides in Table 4.12-1, shall be identified and reported per Specification 6.9.1.8.

6 l

I I

FARLEY-UNIT 1 3/4 12-9 AMENDMENT NO.

RADIOLOGICAL ENVIR00 MENTAL MONITORING 3/4'.12.2' LAND USE CENSUS

_ LIMITING CONDITION FOR'0PERATION

[ 3;12.2'. A land use ' census shall be conducted and shall identify the location of the nearest milk animal and the nearest residence.

APPLICABILITY: At all times.

ACTION:

a. 'With a land use census identifying a location (s) which yields a:

calculated dose or dose commitment greater than the values currently being calculated in Specification 4.11.2.3, prepare'and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the new location (s). i

b. -With a land use census identifying a location (s) which yields a

(' calculated dose or dose connitment (via the same exposure pathway) 20

. percent greater than at a location from which samples are currently being obtained in accordance with Specification 3.12.1, prepare and submit to the Commission within 30 days pursuant to Specification 6.9.2, a Special Report which identifies.the new location. The new

~ location shall be added to the radiological environmental monitoring program within 30 days. The sampling location, excluding the control station location, having the lowest calculated dose or dose connitment (via the same exposure pathway) may be deleted from this monitoring program after (October 31) of the year in which this land use census was conducted.

c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.12.2 The land use census shall be conducted at least once per 12 months between the dates of (June 1 and October 1) using that information which will I provide' the best results, such as by a door-to-door survey, aerial sumey, or by l consulting local agriculture authorities. '

.FARLEY-UNIT 1 3/4 12-10 AENDENT NO.

L

POWER DISTRIBUTION LIMITS BASES The radial peaking factor Fxy(Z), is measured periodically to provide additional

, - assurance that the hot channel factor, Fq (Z), remains within its limit. The Fxy limit for RATED THERMAL POWER (FNP) as provided in the Radial Peaking Factor limit report per Specification 6.9.1.11 was determined from expected power l control maneuvers over the full range of burnup conditions in the core. ,

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

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

The two hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correction of a dropped or misaligned control rod. In the event such action does not correct the tilt, the margin for uncertainty on Fq i s reinstated by reducing the maximum

. allowed power by 3 percent for each percent of tilt in excess of 1.0.

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 f

transient and accident analyses. The limits are consistent with the initial FSAR assumptions 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 through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation. The 18 month

] periodic measurement of the RCS total flow rate is adequate to detect flow degradation and ensure correlation of the flow indication channels with measured flow such that the indicated percent flow will provide sufficient verification of flow rate on a 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months basis.

4 FARLEY-UNIT 1 B3/4 2-5 AMENDMENT NO.

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REACTOR COOLANT SYSTEM BASES- ,

3/4.4.6 STEAM GENERATORS iThe 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 also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.

1 The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible i i

corrosion of 'the steam generator tubes. If the secondary coolant chemistry is i

not maintained within these limits, localized corrosion may likely result in stress corrosion cracking. The extent of cracking during plant operation would

, be limited by the 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 nonnal operation and by postulated 1 accidents. Operating plants have demonstrated that primary-to-secondary leakage i

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 treat."qt of the i

secondary coolant. However, even if a defect should develop in servi e, it will ii be found during scheduled inservice steam generator. tube examinations. Piugging '

will be required for all tubes with imperfections exe.eding the plugging limit of

40% of the tube nominal wall thickness. ' Steam generator tube inspections of i

operating plants have demonstrated the capability to reliably tatect degradation I. that has penetrated 20% of the original tube wall thickr.ess.

j - Whenever the results of any steam generator tubing inservice inspecticn fall into l Category C-3, these results will be reported to the Consnission pursuant to i

4 10CFR50.73 prior to resumption of plant operation. Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eMv-current inspection, and revision of the Technical Specifications, if necessary.

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FARLEY-UNIT 1 B3/4 4-3 AMENDMENT NC.

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ADMINIS M J1VE CONTROLS

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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 experiments completed under the provision of Section 50.59,10CFR, to verify that such actions did not constitute an unreviewed safety question and all programs required by Specification 6.8 and changes thereto.

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

d. Proposed tests or experiments which may involve an ur. reviewed safety question as defined in Section 50.59, 10CFR.

e. Proposed changes to Technical Specifications or this Operating License.

f. Reports of violations of codes, regulations, orders, Technical Specifications, license requirements, or of internal procedures or instructions having nuclear safety significance or reports of abnormal degradation of systems designed to contain radioactive material.

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

h. ALL REPORTABLE EVENTS.

i. All recognized indications of an unanticipated deficiency in some aspect of design or operation of safety related structures, systems, or components.

j. The plant Security Plan.

k. The Emergency Plan.

1. Facility operations to detect potential nuclear safety hazards,

m. Investigations or analyses of special subjects as requested by the Chairman of the Nuclear Operations Review Board,

n. Every unexpected onsite release of radioactive material to the environs resulting from a 1ack of preplanning including the preparation and forwarding of reports covering evaluation, recommendations and disposition of the corrective action to prevent recurrence to the Plant Manager and to the Nuclear Operations Review Board.

o. Changes to the PROCESS CONTROL PROGRAM and the 0FFSITE DOSE CALCULATION MANUAL.

FARLEY-UNIT 1 6-8 AMENDMENT N0.

ADMINISTRATIVE CONTROLS ALTERNATES.

6.5.2.3 All alternate members shall be appointed in writing by the NORB Chairman to serve on a temporary basis; however, no more than two alternates shall participate as voting members in NORB activities at any one time.

CONSULTANTS 6.5.2.4 Consultants shall be utilized as determined by the NORB Chairman to provide expert advice to the NORB.

MEETING FRE00ENCY 6.5.2.5 The NORB shall meet at least once per calender quarter during the initial year of unit operation following fuel loading and at least once per six months thereafter.

QUORUM 6.5.2.6 A quorum shall consist of the Chairman or Vice Chainnan plus enough voting members to constitute a majority of the NORB. No more than a minority of the quorum shall have line responsibility for operation of the facility. For the purpose of a quorum those considered to have line responsibility will include the P1 ant Manager and personnel reporting to the P1 ant Manager.

REVIEW 6.5.2.7 The NORB shall review:

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

b. Proposed tests or experiments which involve an unreviewed safety question as defined in Section 50.59, 10CFR.

c. Violations of codes, regulations, orders, Technical Specifications, license requirements, or of internal procedures or instructions having nuclear safety significance or abnormal degradation of systems designed to contain radioactive material.

d. Significant operating abnonnalities or deviations from normal and expected performance of plant equipment that affect nuclear safety.

e. All REPORTABLE EVENTS.

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f. Recognized indications of an unanticipated deficiency in some aspect of design or operation of safety related structures, systems, or components.

I l

FARLEY-UNIT 1 6-10 AMENDMENT NO.

i i

ADMINISTRATIVE CONTROLS 6.6 REPORTABLE EVENT ACTION 6.6.1 The- following actions shall be taken for REPORTABLE EVENTS:

a. The Commission shall be notified and a report submitted pursuant to the requirements of Section 50.73 to 10CFR Part 50, and

b. Each REPORTABLE EVENT shall be reviewed by the PORC, and the results of this review shall be submitted to the NORB and the Senior Vice President.

6.7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:

a. The facility shall be placed in at least HOT STANDBY within one hour.

b. The NRC Operations Center shall be notified by telephone as soon as possible and all cases within one hour. The Senior Vice President shall be notified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ' .

c. A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the PORC. This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems or structures, and (3) corrective action taken to prevent recurrence.

d. The Safety Limit Violation Report shall be submitted to the Commission, and the Senior Vice President for NORB review within 14 days of the violation.

6.8 PROCEDURES AND PROGRAMS 6.8.1 Written procedures shall be established, implemented and maintained covering the activities referenced below:

a. The applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33, Revision 2,1978.

b. Refueling operations.

c. Surveillance and test activities of safety related equipment.

d. Security Plan implementation.

e. Emergency Plan implementation.

FARLEY-UNIT 1 6-14 AMENDMENT N0.

- ADMINISTRATIVE CONTROLS

. 6.9 REPORTING REQUIREMENTS ROUTINE REPORTS

-6.9.1-In addition to the applicable reporting requirement:: of Title 10, Code of

-Federal Regulations, the following reports shall be submitted to the Director of the Regional Office of Inspection and Enforcement unless otherwise noted.

STARTUP REPORT 6.9.1.1 A summary report of plant startup and power escalation testing shall be submitted following (1) receipt of an operating license, (2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured-by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant.

4 t

l FARLEY-UNIT 1 6-15a AMENDMENT NO.

ADMINISTRATIVE CONTROLS

e. Type of container (e.g., LSA, Type A, Type B, Large Quantity), and

f. Solidification agent (e.g., cement, urea formaldehyde).

The radioactive effluent release reports shall include unplanned releases from the site to unrestricted areas of radioactive materials in gaseous and liquid effluents on a quarterly basis.

The radioactive effluent release reports shall include any changes to the PROCESS l CONTROL PROGRAM (PCP) made during the reporting period.

MONTHLY OPERATING REPORT 6.9.1.10 Routine reports of operating statistics and shutdown experience, l including documentation of all challenges to the PORY's or safety valves, shall

! be submitted on a monthly basis to the Director, Office of Management and Program Analysis, U.S. Nuclear Regulatory Comission, Washington, D.C. 20555, with a copy to the Regional Office of Inspection and Enforcement, no later than the 15th of each month following the calendar month covered by the report.

' Any changes to the 0FFSITE DOSE CALCULATION MANUAL shall be submitted with the Monthly Operating Report within 90 days in which the change (s) was made ef fective. In addition, a report of any major changes to the radioactive waste l treatment systems shall be submitted with the Monthly Operating Report for the i period in which the change was implemented.

I i

l RADIAL PEAKING FACTOR LIMIT REPORT l 6.9.1.11 The F limit for Rated Thermal Power (F RTP) shall be provided to the j Director of the Regional Office of Inspection andinforcement, with a copy to the

( Director, Nuclear Reactor Regulation, Attention Chief of the Core Performance l Branch, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555 for all core planes containing bank "D" control rods and all unrodded core planes at least 60 l days prior to cycle initial criticality. In the event that the limit would be submitted at some other time during core life, it will be submitted 60 days prior to the date the limit would become effective unless otherwise exempted by the Commission.

Any infonnation needed to support Qwill be by request from the NRC and need not be included in this report.

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

i FARLEY-UNIT 1 6-19 AMENDMENT NO.

ADMlNISTRATIVE 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 sample 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 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 to the first sample in which the limit was exceeded; (4) Graph of the I-131 concentration (micro C1/gm) and one other radiciodine isotope concentration (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 presence of greater than or equal to 0.005 microcuries of removable contamination.

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Director of the Office of Inspection and Enforcement Regional Office within the time period specified for each report.

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 minimun 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 items of equipment related to nuclear safety.

c. ALL REPORTABLE EVENTS submitted to the Commission. j

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

e. Records of changes made to the procedures required by Specification 6.8.1.

f. Records of radioactive shipments.

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

FARLEY-UNIT 1 6-20 AMEN 0 MENT NO.

ADMINISTRATIVE CONTROLS

---==

h. Records of annual physical inventory of all sealed source material of record.

6.10.2 The following records shall be retained for the duration of the Unit Operating License:

a. Records and drawing changes reflecting unit design modifications made to systems and equipment described in the Final Safety Analysis Report.

b. Records of new and irradiated fuel inventory, fuel transfers and

, assembly burnup histories.

c. Records of radiation exposare for all individuals entering radiation control areas.

d. Records of gaseous and liquid radioactive material released to the environ s.

e. Records of transient or operational cycles for those unit components identified in Table 5.7-1.

f. Records of reactor tests and experiments.

g. Records of training and qualification for current members of the facility staff.

h. Records of in-service inspections performed pursuant to these Technical Specifications.

l

1. Records of Quality Assurance activities required by the QA Manual.

j. Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59.

k. Records of meetings of the PORC and the NORB.

1. Records of secondary water sampling and water quality.

m. Records of analyses required by the radiological environmental monitoring program.

n. Records of the service lives of all hydraulic and mechanical snubbers in service after 07-01-84 within the scope of 3/4.7.9 including the date at which the service life commences and associated installation and maintenance records.

6.11 RADIATION PROTECTION PROGRAM Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure.

FARLEY-UNIT 1 6-21 AMENDMENT NO. l

ADMINISTRATIVE CONTROLS 6.12HIGHRADIATIONARg 6.12.1 In lieu of the " control device" or " alarm signal" required by paragraph 20.203(c)(2) of 10 CFR 20, each high radiation area in which the intensity of radiation is greater than 100 mrem /hr but less than 1000 mrem /hr shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit.* Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

a. A radiation monitoring device which continuously indicates the radiation dose rate in the area.

b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alams when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate level in the area has been established and personnel have been made knowledgeable of them.

c. A health physics qualified individual (i.e., qualified in radiation protection procedures) with a radiation dose rate monitoring device who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the facility Health Physics Supervisor.

6.12.2 In addition to the requirements of 6.12.1, areas accessible to personnel with radiation levels such that a major portion of the body could receive in one hour a dose greater than 1000 mrem shall be provided with locked doors to prevent unauthorized entry, and the keys shall be maintained under the administrative control of the Shif t Foreman on duty and/or health physics supervision. Doors shall remain locked except during periods of access by personnel under an approved Radiation Work Pennit which shall specify the dose rate levels in the immediate work area and the maximum allowable stay time for individuals in that area. For individual areas accessible to personnel with radiation levels such that a majar portion of the body could receive in one hour a dose in excess of 1000 mrem ** that are located within large areas, such as PWR containment, where no enclosure exists for purposes of locking, and no enclosure can be reasonably constructed around the individual areas, then that area shall be roped off, conspicuously posted and a flashing light shall be activated as a warning device. In lieu of the stay time specification of the RWP, direct or remote (such as use of closed circuit TV cameras) continuous surveillance may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities within the area.

Health Physics personnel or personnel escorted by Health Physics personnel shall be exempt from the RWP issuance requirement during the perfonnance of their assigned radiation protection duties, provided they are following plant radiation protection procedures for entry into high radiation areas.

- Measurement made at 18" from source of radioactivity.

FARLEY-UNIT 1 6-22 AMENDMENT NO.

ADMINISTRATIVE CONTROLS 6.13. PROCESS CONTROL PROGRAM (PCP) 6.13.1' The PCP shall be approved by the Commission prior to implementation.

6.13.2 Licensee initiated changes to the PCP:

1. Shall be submitted to the Commission in the semi-annual Radioactive Effluent Release Report for the period in which the change (s) was made.

This submittal shall contain:

a. Sufficiently detailed infomation to totally support the rationale for the change without benefit of additional or supplemental j information; l b. A determination that the change did not reduce the overall conformance of the solidified waste program to exiting criteria for l solid wastes, and

c. Documentation of the fact that the change has been reviewed and found acceptable by the PORC.

2. Shall become effective upon review and approval in accordance with Speci fication 6.5.3.1.

6.14 0FFSITE DOSE CALCULATION MANUAL (0DCM) 6.14.1 The ODCM shall be approved by the Commission prior to implementation.

6.14.2 Licensee initiated changes to the ODCM:

l l 1. Shall be submitted to the Commission in the Monthly Operating Report i

within 90 days of the date the change (s) was made effective. This submittal shall contain:

a. Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information. Information submitted should consist of a package of those pages of the 00CM to be changed with each page numbered and provided with an approval and date box, together with appropriate analyses or evaluations justifying the change (s);

i b. A detemination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determinations; and l c. Documentation of the fact that the change has been reviewed and found acceptable by the PORC.

2. Shall become effective upon review and approval in accordance with Speci fication 6.5.3.1.

FARLEY-UNIT 1 6-23 AMENDMENT NO.

y ADMINISTRATIVE CONTROLS 6.15 MAJOR CHANGES TO RADI0 ACTIVE WASTE TREATMENT SYSTEMS (Liquid, Gaseous, and Solid) 6.15.1 Licensee initiated major changes to the radioactive waste systems (Liquid, Gaseous and Solid):

1. Shall be reported to the Commission in the Monthly .0perating Report for the period in which the change was implemented. The discussion of each change shall contain:

a. A sumary of the evaluation that led to the determination that the change could be made in accordance with 10CFR50.59;

b. Sufficient detailed information to totally support the reason for the change without benefit of additional or supplemental information;

c. A detailed description of the equipment, components and processes involved and the interfaces with other plant systems;

d. An evaluation of the change which shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste that differ from those previously predicted in the license application and amendments thereto; *

e. An evaluation of the change which shows the expected maximum exposures to individuals in the unrestricted area and to the general population that differ from those previously estimated in the license application and amendments thereto;

f. A comparison of the predicted releases of radioactive materials, in liquid and gaseous effluents and in solid waste, to the actual releases for the period prior to when the changes are to be made;

g. An estimate of the exposure to plant operating personnel as a result of the change; and

h. Documentation of the fact that the change was reviewed and found acceptable by the PORC.

, 2. Shall become effective upon review and approval in accordance with Specification 6.5.3.1.

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

)

4.0- Environmental Conditions 4.1 Unusual or Important Enviromnental Events Any occurrence of an unus!Ja1 or important event that indicates or could result in significant environmental impact causally related to plant operation shall be recorded and reported to the NRC in accordance with 10CFR50.72(b)(2)(vi) or by a written report per Subsection 5.4.2, as appropriate. The following are examples:

excessive bird impaction events, onsite plant or animal disease outbreaks, mortality or unusual occurrence of any species protected by the Endangered Species Act of 1973, fish kills, increase in nuisance organisms or conditions and unanticipated or emergency discharge of waste water or chemical substances.

No routine monitoring programs are required to implement this condition.

4.2 Environmental Monitoring 4.2.1 Aerial Remote Sensing Vegetation comunities of the site and vicinity within 1 kilometer of the cooling towers in all directions shall be aerially photographed to detect and assess the significance of damage, or lack thereof, as related to cooling tower drift dispersions. Photography shall be done by aerial overflight during May or June.

Monitoring shall include a program of low altitude false color aerial photography (either color infrared photography or multispectral or multiband photography).

The scale for full coverage shall be adequate to 4-1 AMENDMENT NO.

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

1.3 CHANNEL CALIBRATION .............................................. 1-1 1.4 CHANNEL CHECK .................................................... 1-1 l 1.5 CHANNEL FUNCTION TEST ............................................ 1-1

! 1.6 CONTAINMENT INTEGRITY ............................................ 1-2 1.7 CONTROLLED LEAXAGE ............................................... 1-2 1.8 CORE ALTERATION .................................................. 1-2 1.9 DOSE EQUIVALENT I-131 ............................................ 1-2 l 1.10 E- AV ERAGE DIS INTEGRAT ION ENERGY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 l'

1.11 ENGINEERED SAFETY FEATURES RESPONSE TIME ......................... 1-3 1.12 FREQUENCY NOTATION ............................................... 1-3 1.13 GASEOUS RADWASTE TREATMENT SYSTEM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 1.14 IDENTIFIED LEAXAGE ............................................... 1-3 1.15 LIQUID RADWASTE TREATMENT SYSTEM ................................. 1-4 l 1.16 MAJOR CHANGES TO RADI0 ACTIVE WASTE TREATMENT SYSTEMS.............. 1-4 1.17 0FFSITE DOSE CALCUL ATION MANU AL (0DCM) . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 1.18 OPERABLE - OPERABILITY ........................................... 1-4 1.19 OPERATIONAL MODE - MODE .......................................... 1-5 1.20 PHYSICS TESTS .................................................... 1-5

1. 21 P RES SU RE B 0UND AR Y L E AK AGE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5 1.22 PROCESS CONTROL PROGRAM ( PCP ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5 1.23 PURGE-PURGING .................................................... 1-5

1. 24 QU AD RANT POWER T ILT RAT I O . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5 1.25 RATED THERMAL POWER .............................................. 1-5 1.26 REACTOR TRIP SYSTEM RESPONSE TIME ................................ 1-6 1.27 REPORTABLE EVENT ................................................. 1-6 1.28 SHUTDOWN MARGIN .................................................. 1-6 1.29 SOLIDIFICATION ................................................... 1-6 1.30 SOURCE CHECK ..................................................... 1-6 1.31 STAGGERED TEST BASIS ............................................. 1-6 1.32 THERMAL POWER .................................................... 1-7

1. 3 3 U N I D E NT I F I ED L E AX AGE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-7 1.34 VENTILATION EXHAUST TREATMENT SYSTEM ............................. 1-7 1.35 VENTING .......................................................... 1-7 TABLE 1.1 OPERATIONAL MODES ........................................... 1-8 l TABLE 1.2 FREQUENCY NOTATION .......................................... 19 FARLEY-UNIT 2 I AMENDMENT NO.

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.. - - . = ._ -. . - - . . ..

INDEX ADMINISTRATIVE CONTROLS SECTION PAGE i

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

I Audits .......................................................... 6-11 A u th o r i ty . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-12 Records ......................................................... 6-12 6.5.3 TECHNICAL REVIEW AND CONTROL Activities ...................................................... 6-12 Records .......................................................... 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 i

Startup Report ................................................... 6-15a Annual Report .................................................... 6-16 Annual Radiological Envi ronmental Operating Report . . . . . . . . . . . . . . . 6-17 Semiannual Radioactive Ef fluent Rel ease Report . . . . . . . . . . . . . . . . . . . 6-17 Monthly Op era ti ng Rep ort . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-19

, Radi al Peak i ng Factor Limi t Report . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-19 i

l Annual Diesel Genera tor Reli abili ty . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-19 i

Annual Reactor Coolant System Specific Activity Report ........... 6-20 I

i Annual Seal ed Source Leakage Report . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-20 t

6 . 9 . 2 SP E C I AL REP O RT S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-20 i

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6.10 RECORD RETENTION .................................................. 6-20 I

6.11 RADIATION PROTECTION PROGRAM ...................................... 6-21 6.12 HIGH EADIATION AREA ............................................... 6-22 l

FARLEY-UNIT 2 XIX AMENDMENT NO.

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.- .~ .. -. - - -- - - - . ..

~INDEX

l. ADMINISTRATIVE CONTROLS W

-SECTION, PAGE 6.13 PROCESS CONTROL PROGRAM .......................................... 6-23 6.14 0FFSITE DOSE CALCULATION MANL g .................................. 6-23 I 6.15 MAJOR CHANGES TO RADI0 ACTIVE WASTE TREATMENT SYSTEMS . . . . . . . . . . . . . 6-24 i P

b

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o c

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FARLEY-UNIT 1 XX AENDMENT NO.

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l DEFINITIONS l

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

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

f SHUTDOWN MARGIN i

1.28 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 withdrawn.

SOLIDIFICATION 1.29 SOLIDIFICATION shall be the conversion of radioactive wastes from liquid systems to a homogeneous (uniformly distributed), monolithic, innobilized solid with definite volume and shape, bounded by a stable surface of distinct outline on all sides (free-standing).

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

STAGGERED TEST DASIS i 1.31 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 i equal subintervals,

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

FARLEY-UNIT 2 1-6 AMENDHENT NO. ,

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_ . _ _ _ _ _ _ _ . _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _______ m_ _ _ ___ - - - -

3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.0 APPLICABILITY LIMITING CONDITION FOR OPERATION 3.0.1 Compliance with the Limiting Conditions for Operation contained in the succeeding specifications is required during the OPERATIONAL MODES or other conditions specified therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be met.

3.0.2 A condition prohibited by the Technical Specifications shall exist when l the requirements of the Limiting Condition for Operation and associated ACTION requirements are not met within the specified time intervals. If the Limiting Condition for Operation is restored prior to the expiration of the specified time intervals, completion of the ACTION requirements is not required.

3.0.3 When a Limiting Condition for Operation is not met, except as provided in the associated ACTION requirements, within one hour ACTION shall be initiated to place the unit in a MODE in which the specification does not apply by placing it, as applicable, in:

1. At least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,

2. At least H0T SHUT 00WN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and

3. At least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Where corrective measures are completed that permit operation under the ACTION requirements, the ACTION may be taken in accordance with the specified time limits as measured from the time of failure to meet the Limiting Condition for Operation. Exceptions to these requirements are stated in the individual specifications.

3.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the conditions of the Limiting Condit 'n for Operation are met without reliance on provisions contained in the ALTION requirements. This provision shall not prevent passage through OPERATIONAL MODES as required to comply with ACTION requirements. Exceptions to these requirements are stated in the individual specifications.

3.0.5 When a system, subsystem, train, component or device is determined to be inoperable solely because its emergency power source is inoperable, or solely because its normal power source is inoperable, it may be considered OPERABLE for the purpose of satisfying the requirements of its applicable Limiting Condition for Operation, provided: (1) its corresponding normal or emergency power source is OPERABLE; and (2) all of its redundant system (s), subsystem (s), train (s),

l component (s) and device (s) are OPERABLE, or likewise satisfy the requirements l of this specification. Unless both conditions (1) and (2) are satisfied within l 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , ACTION shall be initiated to place the unit in a MODE in which the applicable Limiting Condition for Operation does not apply by placing it, as applicable, in:

FARLEY-UNIT 2 3/4 0-1 AMEN 0 MENT NO.

REACTIVITY CONTROL SYSTEMS MODERATOR TEWERATURE COEFFICIENT LIMITING CONCITION FOR OPERATION 3.1.1.3 The moderator teciperature coefficient (MTC) shall be:

a. Less than or equal to 0.5 x 10-4 delta k/k/*F for the all rods withdrawn, beginning of cycle life (BOL), below 70% THERMAL POWER condition. Less than or equal to 0 delta k/k/*F at or above 70%

THERMAL POWER.

b. Less negative than -3.9 x 10-4 delta k/k/*F for the all rods withdrawn, l end of cycle life (E0L), RATED THERMAL POWER condition.

l l APPLICABILITY: Specification 3.1.1.3.a - MODES 1 and 2* only#

Specification 3.1.1.3.b - MODES 1, 2 and 3 only#

ACTION:

l a. With the MTC more positive than the limit of 3.1.1.3.a above, operation j in MODES 1 and 2 may proce(d provided:

1. Control rod withdrawal limits are established and maintained sufficient to restore the MTC to within its limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 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 . These withdrawal l limits shall be in addition to the insertion limits of l Specification 3.1.3.6.

2. The control rods are maintained within the withdrawal limits established above until a subsequent calculation verifies that the MTC has been restored to within its limit for the all rods withdrawn condition.

3. A Special Report is prepared and submitted to the Commission pursuant to Specification 6.9.2 within 10 days, describing the value of the measured HTC, the interim control rod withdrawal limits and the predicted average core burnup necessary for restoring the positive HTC to within its limit for the all rods withdrawn condition.

b. With the MTC more negative than the limit of 3.1.1.3.b above, be in HOT SHUTOOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

"WIth Ke rr greater than or equal to 1.0

See Special Test Exception 3.10.3 FARLEY-UNIT 2 3/41-4 AMENDMENT NO.

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. i POWER DISTRIBUTION LIMITS !

SURVEILLANCE REQUIREMENTS (Continued)

E

2. When'the Fx c is less than or equal to the F RTP limit for the appropriate measuredc core plane, addigigal kwgr distribution maps shall be taken and Fxy compared to Fxy and F xy at least once per 31 EFPD.

e. The F limit for RATED THERMAL POWER (FhTP) shall be provided for all core pianes containing bank "D" contrdi rods and all unrodded core planes in a Radial Peaking Factor Limit Report per Specification 6.9.1.11.

l f. The Fxy limits of e, above, are not applicable in the following core planes regions as measured in percent of core height from the bottom of the fuel:

l 1. Lower core region from 0 to 15%, inclusive.

2. Upper core region from 85 to 100%, inclusive.

3. Grid plane regions at 17.8 t 2%, 32.1 12%, 46.4 t 2%, 60.6 i 2%

and 74.9 t 2%, inclusive.

i

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 FxyC exceeding Fxy L the effects of F xy on Fn (Z) shall be evaluattd to determine if Fq (Z) is within its limits.

4.2.2.3 When F measured Fq shall (Z)qbe (Z) is measured obtained for other from a power than Fxy dist.rfbution detenninations, map and increased by an over 3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty.

i FARLEY-UNIT 2 3/4 2-6 AMENDHENT NO.

INSTRU ENTATION FIRE DETECTION INSTRUtENTATION LIMITING CONDITION FOR OPERATION 3.3.3.9 As a minimum. the fire detection instrumentation for each fire I detection zone shown in Table 3.3-12 shall be OPERABLE.

APPLICABILITY: Whenever equipment protected by the fire detection instrument is required to be OPERABLE.

ACTION:

l With the number of OPERABLE fire detection instruent(s) less than the minimum number OPERABLE requirement of Table 3.3-12:

a. Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months establish a fire watch patrol to inspect the zone (s) with the inoperable instrment(s) at least once per hour, unless the instrument (s) is located inside the containment, then monitor the containment air temperature at least once per hour at the locations listed in Specification 4.6.1.5.

b. Restore the inoperable instrument (s) to OPERABLE status within 14 days or prepare and submit a Special Report to the Comission pursuant to Specification 6.9.2 within the next 30 days outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the instrumcat(s) to OPERABLE status.

c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

l l SURVEILLANCE REQUIREMENTS 4.3.3.9.1 Each of the above required fire detection instruments which are accessible during plant operation shall be demonstrated OPERABLE at least once per 6 months by performance of a function test which includes subjecting the detector to test aerosol. Fire detectors which are not accessible during plant operation shall be demonstrated OPERABLE by the performance of this functional test during each COLD SHUTDOWN exceeding 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months unless performed in the previous 6 months. :

4.3.3.9.2 The NFPA Standard 720 supervised circuits supervision associated with !

the detector alarms of each of the above required fire detection instruents I l shall be demonstrated OPERABLE at least once por 6 months.

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FARLEY-UNIT 2 3/4 3-59 AMENDMENT NO.

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INSTRUENTATION RADIDACTIVE LIQUID EFFLUENT HONITORING INSTRUENTATION LIMITING CONDITION FOR OPERATION 3.3.3.10 The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.3-13 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.11.1.1 are not exceeded. The alarm /

trip setpoints of these channels shall be determined in accordance with the OFFSITE DOSE CALCULATION MANUAL (ODCH). ,

APPLICABILITY: At all times.

ACTION:

a. With a radioactive liquid effluent monitoring instrumentation channel alarm / trip setpoint less conservative than required by the above specification, imediately suspend the release of radioactive liquid l effluents monitored by the affected channel or declare the channel inoperable.

b. With less than the minimum nunber of radioactive liquid effluent #

monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-13.

l

c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

l SURVEILLANCE REQUIREMENTS

_______ __=___-__ - - __ _ _ _ ______ __ _-_-_ _

4.3.3.10 Each radioactive liquid effluent monitoring instrumentation channel l shall be demonstrated OPERABLE by perionnance of the CHANNEL CHECK, SOURCE i CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the l frequencies shown in Table 4.3-8.

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l FARLEY-UNIT 2 3/4 3-61 AMENDMENT NO.

.-____-_____-__________________________________________-_-____-_-_________-_____-__O

. INSTRUMENTATION RADI0 ACTIVE LIQUID EFFLUENT F0NITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.10 The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.3-13 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.11.1.1 are not exceeded. The alarm /

trip setpoints of these channels shall be detennined in accordance with the 0FFSITE DOSE CALCULATION MANUAL (00CM).

APPLICABILITY: At all times.

ACTION:

a. With a radioactive liquid effluent monitoring instrumentation channel alarm / trip setpoint less conservative than required by the above specification, innediately suspend the release of radioactive liquid effluents monitored by the affected channel or declare the channel inoperable.

b. With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-13.

c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.10 Each radioactive liquid effluent monitoring instrumentation channel i shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE l CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3-8.

l O

FARLEY-UNIT 2 3/4 3-61 AMENDMENT NO.

INSTRUENTATION RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUENTATION LIMITING CONDITION FOR OPERATION

_~

l 3.3.3.11 The radioactive gaseous effluent mo11toring instrunentation channels shown in Table 3.3-14 shall be OPERABLE with their alarm / trip setpoints set to l ensure that the limits of Specification 3.11.2.1 are not exceeded. The alarm /

l trip setpoints of these channels shall be determined in accordance with the ODCM.

APPLICABILITY: As shown in Table 3.3-14.

i ACTION:

a. With a radioactive gaseous effluent monitoring instrumentation channel alarm / trip setpoint less conservative than required by the above Specification, immediately suspend the release of radioactive gaseous

effluents monitored by the affected channel or declare the channel inoperable,

b. With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-14.

c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.11 Each radioactive gaseous effluent monitoring instrumentation channel ,

shall be demonstrated OPERABLE by perfonnance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3-9.

FARLEY-UNIT 2 3/4 3-66 AMEN 0 MENT NO.

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y REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS-(Continued)

g. Preservice Inspection means an inspection of the full length of each tube in each steam generator performed by l eddy current techniques prior to service to establish a baseline condition of the tubing. This inspection shall be performed after the field hydrostatic test and prior to initial POWER OPERATION using the equipment and techniques

expected to be used during subsequent inservice inspections.

b. The steam generator shall be determined OPERABLE after l completing the corresponding actions (plug all tubes exceeding the plugging limit and all tubes containing through-wall cracks) required by Table 4.4-2.

4.4.6.5 Reports

a. Following each inservice inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the Commission within 15 days.

1 i

b. The complete results of the steam generator tube inservice inspection shall be submitted to the Commission in a Special Report pursuant to Specification 6.9.2 within 12 months following the completion of the inspection. This Special Report shall include:

1. Number and extent of tubes inspected.

I 2. Location and percent of wall-thickness penetration for each indication of an imperfection.

3. Identification of tubes plugged.

c. Results of steam generator tube inspections which fall into Category C-3 shall be considered a REPORTABLE EVENT and shall be reported pursuant to 10CFR50.73 prior to resumption of plant operation. The written report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.

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l FARLEY - UNIT 2 3/4 4-13 AMENOMENT NO.

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__m____._..__.___.__m.._ - _ . . _ _ _ _ _ . . . . _ _ _ . . - _ . _ _

_..___m. __ _ . . . . . _ -. ..

[

TANLE 4.4- 2 STEAM E NERATOR TURE INSPECTION 5

E IST SaWLE INSPECTION 29B SAMLE INSPECTION 3ND SAW1.E INSPECTION E Sample Size Resul t Action Required Result Action Required Result

~.

Action Required .

y A minimum of C-1 None N/A S Tubes per N/A N/A N/A >

S.G. C-2 Plug defective C-1 None N/A .N/A tubes and inspect additional 25 C-2 Plug defective C-1 None tubes in this S.G. tubes and inspect additional 45 tahes C-2 Plug defective in this S.G. tubes C-3 Perform action for C-3 result of first sample (so C-3 Perform action for N/A N/A p C-3 result of first sample w

C-3 Inspect all tubes All other Nnne N/A N/A in this S.G., S.G.s are plug defective C-1 tubes and inspect 25 tubes in each Some S.G.s Perform action for N/A N/A other S.G. C-2 but no C-2 result of additional second sample w S.G.s are M

~ C -3 s

A Notification to Additional Inspect all tubes N/A N/A 3 NRC pursuant to S.G. is C-3 in each S.G. and 10CFRSO.73 plug defective '

[5 tubes.

Notification to NRC pursuant to t 10CFR50.73-S-3N W ?. during Where an N isinspection the number of steam generators in the unit, and n is the number of steam generators inspected

. . _ . . ~_

.-_,~.-_____-

REACTOR COOLANT SYSTEM ACTION: (Continued)

MODES 1, 2, 3, 4 and 5:

l

a. With the specific activity of the primary coolant greater than 1.0 microcurie per gram DOSE EQUIVALENT I-131 or greater than 100/E microcuries per gram, perform the sampling and analysis requirements of item 4a of Table 4.4-4 until the specific activity of the primary l coolant is restored to within its limits.

l I

! SURVEILLANCE REQUIREMENTS 4.4.9 The specific activity of the primary coolant shall be determined to be within the limits by performance of the sampling and analysis program of Table 4.4-4.

l l FARLEY-UNIT 2 3/4 4-24 AMENDMENT NO.

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CONTAINNENT SYSTEPG SURVEILLANCE REQUIREMENTS (Continued) 4.6.1.6.2. End Anchorages and Adjacent Concrete Surfaces The structural integrity of the end anchorages of all tendons inspected pursuant to Specification 4.6.1.6.1 and adjacent concrete surfaces shall be demonstrated by determining through inspec*.lon that no apparent changes have occurred in the visual appearance of the e.id anchorage or the concrete crack patterns adjacent to the end anchorages. Inspections of the concrete shall be performed during the first Type A containment leakage rate tests (reference Specification 4.6.1.2) while the containment is at its maximum test pressure.

4.6.1.6.3 Liner Plate The structural integrity of the containment liner plate shall be determined during the shutdown for the first Type A containment leakage rate test (reference Specification 4.6.1.2) by a visual inspection of the liner plate verifying no apparent changes in appearance or other abnormal degradation.

FARLEY-UNIT 2 3/46-9 AMENDMENT tl0.

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PLANT SYSTEMS SURVEILLANCE REQUIREMENTS ~ (Continued)

1. With a half-life greater than 30 days (excluding Hydrogen 3), and

2. In any form other than gas.

b. Stored sources-not in use - Each sealed source and fission detector shall be tested prior to use or transfer to another licensee unless I tested within the previous six months. Sealed sources and fission l detectors transferred without a certificate indicating the last test

. date shall be tested prior to being placed into use.

l o c. Startup sources and fission detectors - Each sealed startup source and fission detector shall be tested within 31 days prior to being subjected to core flux or installed in the core and following repair or maintenance to the source.

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l -FARLEY-UNIT 2 3/4 7-51 AMENOMENT NO.

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l PLANT SYSTEMS 3/4.7.11 FIRE SUPPRESSION SYSTEMS 1 FIRE SUPPRESSION WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.11.1 The fire suppression water system shall be OPERABLE with:

a. Two high pressure pumps, each with a capacity of 2500 gpm, with their discharge aligned to the fire suppression header,

b. Separate water supplies, each with a minimum contained volume of 250,000 gallons, and

c. An OPERABLE flow path capable of taking suction from each tank and transferring the water through distribution piping with OPERABLE sectionalizing control or isolation valves to the yard hydrant curb valves, the last valve ahead of the water flow alarm device on each sprinkler or hose standpipe, and the last valve ahead of the deluge valve on each deluge or spray system required to be OPERABLE per Speci fications 3.7.11.2, 3.7.11.4 and 3.7.11.5.

APPLICABILITYi At all times.

ACTION:

a. With one of the above required pumps and/or water supplies inoperable, restore the inoperable equipment to OPERABLE status within 7 days or prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 30 days outlining the plans and procedures to be used to provide for the loss of redundancy in this sy stem. The provisions of Specifications 3.0.3 and 3.0.4 are not applicabl e.

l b. With the fire suppression water system otherwise inoperable:

i

1. Establish a backup fire suppression water system within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and

2. Submit a Special Report in accordance with Specification 6.9.2:

I a) By telephone within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b) Confirmed by telegraph, mailgram or fascimile transmission no later than the first working day following the event, and c) In writing within 14 days following the event, outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

FARLEY-UNIT 2 3/4 7-52 AMENDMENT NO.

PLANT SYSTEMS SPRAY AND/0R SPRINKLER SYSTEMS LIMITING CONDITION FOR OPERATION 3.7.11.2 The spray and/or sprinkler systems listed in Table 3.7-5 shall be OPERABLE.

APPLICABILITY: Whenev0r equipment in the spray / sprinkler protected areas is required to be OPERABLE.

ACTION:

a. With one or more of the above required spray and/or sprinkler systems inoperable, within one hour establish a continuous fire watch with backup fire suppression equipment for those areas in which redundant systems or components could be damaged; for other areas, establish an hourly fire watch patrol. Restore the system to OPERABLE- status within 14 days or prepare and submit a Special Report to the Connission pursuant to Specification 6.9.2 within the next 30 days outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.11.2 Each of the above required spray and/or sprinkler systems shall be demonstrated OPERABLE:

a. At least once per 31 days by verifying that each valve (manual, power operated or automatic) in the flow path is in its correct position.

b. At least once per 12 months by cycling each testable valve in the flow path through at least one complete cycle of full travel.

I FARLEY-UNIT 2 3/4 7-55 AMENDMENT NO.

PLANT SYSTEMS CO2__ SYSTEMS LIMITING CONDITION FOR OPERATION 3.7.11.3 The following high pressure and low pressure C02 systems shall be OPERABLE.

a .- Service Water Intake Structure (each 4160 volt bus and each 600 volt load center) - HP.

b. Turbine Building 13 ton unit and distribution system in the Auxiliary Building - L.P.

c. Diesel Building 5 ton unit and distribution system.

APPLICABILITY: Whenever equipment protected by the C02 systems is required to be OPERABLE.

ACTION:

a. With one or more of the above required C02 systems inoperable, within one hour establish a continuous fire watch with backup fire suppression equipment for those areas in which redundant systems or components could be damaged; for other areas, establish an hourly fire watch patrol. Restore the system to OPERABLE status within 14 days or prepare and submit a Special Report to the Comission pursuant to Specification 6.9.2 within the next 30 days outlining the action taken, l the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status,

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.11.3.1 Each of the above required C02 systems shall be demonstrated OPERABLE at least once per 31 days by verifying that each manual valve in the flow path is in its correct position.

4.7.11.3.2 Each of the above required low pressure C0g systems shall be demonstrated OPERABLE:

a. At least once per 7 days by verifying the CO2 storage tank level to be greater than 50% and pressure to be greater than 250 psig, and

b. At least once per 18 months by verifying:

1. The Jystem valves and associated ventilation dampers and fire door release mechanisms actuate manually and automatically, upon receipt of a simulated actuation signal, and

2. Flow from each nozzle during a " Puff Test."

FARLEY-UNIT 2 3/4 7-58 AMENDMENT NO.

PLANT SYSTEMS FIRE HOSE STATIONS LIMITING CONDITION FOR OPERATION 3.7.11.4 The fire hose stations shown in Table 3.7-6 shall be OPERABLE.

APPLICABILITY: Whenever equipment in the areas protecte( by the fire hose stations is required to be OPERABLE.

ACTION:

a. With one or more of the fire hose stations shown in Table 3.7-6 inoperable, route

an additional equivalent capacity fire hose to the unprotected area (s) from an OPERABLE hose station within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months if the inoperable fire hose is the primary means of fire suppression; otherwise route the additional hose within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Restore the fire hose station to OPERABLE status within 14 days or prepare and submit a l Special Report to the Commission pursuant to Specification 6.9.2 within the next 30 days outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the station to OPERABLE status.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.11.4 Each of the fire hose stations shown in Table 3.7-6 shall be demonstrated OPERABLE:

a. At least once per 31 days by visual inspection of the fire hose stations accessible during plant operation to assure all required equipment is at the station,

b. At least once per 18 months by:

1. Removing the hose for inspection and re-racking, and

2. Inspecting all gaskets and replacing any degraded gaskets in the coupl ings.

c. At least once per 3 years by:

1. Partially opening each hose station valve to verify valve OPERABILITY and no flow blockage.

2. l Conducting a hose hydrostatic test at a pressure at least 50 psig '

greater than the maximum pressure available at that hose station.

1f routing of the hose would require rendering a fire barrier penetration inoperable, hose will be routed up to but not through the penetration with sufficient hose length to reach the unprotected area (s).

FARLEY-UNIT 2 3/4 7-60 AMENDMENT N0.

PLANT SYSTEMS YARD FIRE HYDRANTS AND HYDRANT HOSE HOUSES LIMITING CONDITION FOR OPERATION 3.7.11.5 The yard fire hydrants and associated hydrant hose houses shown in Table 3.7-7 shall be OPERABLE.

APPLICABILITY: Whenever equipment in the areas protected by the yard fire hydrants is required to be OPERABLE.

ACTION:

a. With one or more of the yard fire hydrant or associated hydrant hose houses shown in Table 3.7-7 inoperable, within ~1 hour have sufficient additional lengths of 21/2 inch diameter hose located in an adjacent OPERABLE hydrant hose house to provide service to the unprotected area (s) if the inoperable fire hydrant or associated hydrant hose house is the primary means of fire. suppression; otherwise provide the additional hose within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Restore the hydrant or hose house to OPERABLE status within 14 days or prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 30 {

days outlining the action taken, the cause of the inoperability and the p1ans and schedule for restoring the hydrant or hose house to OPERABLE status.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.11.5 Each of the yard fire hydrants and associated hydrant hose houses shown in Table 3.7-7 shall be demonstrated OPERABLE:

a. At least once per 31 days by visual inspection of the hydrant hose house to assure all required equipment is at the hose house.

b. At least once per 6 months (once during March, April or May and once during September, October or November) by visually inspecting each yard fire hydrant and verifying that the hydrant barrel is dry and that the hydrant is not damaged.

c. At least once per 12 months by:

1. Conducting a hose hydrostatic test at a pressure at least 50 psig greater than the maximum pressure available at any yard fire hydrant.

2. Inspecting all the gaskets and replacing any degraded gaskets in the couplings.

3. Performing a flow check of each hydrant to verify its 0PERABILITY. ,

FARLEY-UNIT 2 3/4 7-62 AMENDMENT NO.

PLANT SYSTEFG 3/4.7.12 FIRE BARRIER PENETRATIONS LIMITING CONDITION FOR OPERATION 3.7.12 All fire barrier penetrations (including cable penetration barriers, firedoors and fire dampers) in fire zone boundaries protecting safety related areas shall be functional.

APPLICABILITY: At all times.

ACTION:

a. With one or more of the above required fire barrier penetrations non-functional, within one hour either, establish a continuous fire watch on at least one side of the affected penetration, or verify the OPERABILITY of fire detectors on at least one side of the non-functional fire barrier and establish an hourly fire watch patrol .

Restore the non-functional fire barrier penetration (s) to functional status within 7 days or prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 30 days outlinie c the action taken, the cause of the non-functional penetration and plans and schedule for restoring the fire barrier penetration (s) to functional status.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.12 Each of the above required fire barrier penetrations shall be verified to be functional:

a. At least once per 18 months by a visual inspection.

b. Prior to returning a penetration fire barrier to functional status following repairs or maintenance by performance of a visual inspection of the affected penetration fire barrier (s).

FARLEY-UNIT 2 3/4 7-64 AMENDMENT NO.

ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) b) Coolant Temperature High (CTH) c) Coolant Pressure Low-(CPL) d) Crankcase Pressure High (CCPH)

11. Verifying the capability to reject a load of greater than or equal to the largest single load associated with that diesel generator (approximately 1000 kw); while maintaining voltage between 3740 and 4580 volts and speed less than or equal to 75% of the difference between nominal speed and the overspeed trip setpoint.

d. At least once per 10 years or after any modifications which could affect diesel generator interdependence by starting the diesel generators simultaneously, and verifying that the diesel generators accelerate to at least 900 rpm, for the 2850 kw generator and 514 rpm for the 4075 kw generator, in less than or equal to 12 seconds,

e. At least once per 5 years, on a staggered basis, by verifying that the diesel generator can reject a load of 1200-2400 kw without tripping.

The diesel generator output breaker (s) must remain closed such that the diesel generator is connected to at least one emergency bus. Verify that all fuses and breakers on the energized emergency bus (es) are not tripped. The generator voltage shall remain within 3330 and 4990 volts during and following the load rejection.

FARLEY-UNIT 2 3/4 8-6 AMENDMENT N0.

s Table 4.11-1 -(Continued)

TABLE NOTATION

b. A composite sample is one in which the quantity of liquid sampled.is proportional to the quantity of liquid waste discharged and in which the method of- sampling employed results in a specimen which is representative of the liquids released.

c. A batch release is the discharge of liquid wastes of a discrete volume. Prior to sampling for analyses, each batch shall be isolated, and then thoroughly mixed, by a method described in the ODCM, to assure representative sampling.

d. A continuous release is the discharge of liquid wastes of a nondiscrete volume; e.g., from a volume of system that has an input flow during the effluent release.

e. The principal ganma emitters for which the MDC specification applies exclusively are the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported.

f. Sampling will be performed only if the effluent will be discharged to the environment.

g. Deviation from the MDC requirements of Table 4.11-1 shall be reported

~

per Specification 6.9.1.8.

l l

j FARLEY-UNIT 2 3/4 11-4 AMENDMENT N0.

L

RADI0 ACTIVE EFFLUENTS DOSE LIMITING CONDITION FOR OPERATION 3.11.1.2 The dose ,or dose commitment to an individual from radioactive materials in liquid effluents release'd, from each reactor unit, from the site (see Figure 5.1-4) shall be limited:

a. During' any calendar quarter to less than or equal to 1.5 mrem to the ~

total body and to less than or equal to 5 mrem to any organ, and

'b. During any calendar year to less than or equal to 3 mrem to the total body and to less than or' eq'lal to 10 mrem to any organ.

APPLICABILITY: At all times.

ACTION:

a. With the calculated dose from- the release. of radioactive materials in liquid effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the cause(s) for exceeding the limit (s) and defines the corrective actions to be taken to reduce the releases of radioactive materials in liquid effluents during the remainder of the current calendar quarter and during the remainder of the current calendar yaar, so that the cumulative dose or dose commitment to an individual from these releases is within 3 mrem to the total body and 10 mrem to any organ.

b. lne provisions of specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.1.2 Dose Calculations. Cumulative dose contributions from liquid effluents shall be determined in accordance with the ODCM at least once per 31 days.

FARLEY-UNIT 2 3/4 11-5 NiENDMENT NO.

RADI0 ACTIVE EFFLUENTS LIQUID WASTE TREATMENT LIMITING CONDITION FOR OPERATION 3.11.1.3 The LIQUID RADWASTE TREATMENT SYSTEM-shall be OPERABLE. The appropriate portions of the system shall be used to reduce the radioactive materials in liquid wastes prior to their discharge when the projected doses due to the liquid effluent from the site (see Figure 5.1-4) when averaged over the calendar quarter would exceed 0.18 mrem to the total body or 0.6 mrem to any organ.*

APPLICABILITY: At all times.

ACTION:

a. With the LIQUID RADWASTE TREATMENT SYSTEM inoperable for more than 31 days or with radioactive liquid waste being discharged without treatment and in excess of the above limits, prepare and submit to the l Commission within 30 days pursuant to Specification 6.9.2 a Special Report which includes the following information:

1. Identification of the inoperable equipment or subsystems and the reason for inoperability,

2. Action (s) taken to restore the inoperable equipment to OPERABLE status, and

3. Summary description of action (s) taken to prevent a recurrence.

b. The provisions of specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.1.3.1 Doses due to liquid releases to unrestricted areas shall be projected, based on computed operational history, at least once per 31 days, in accordance with the ODCM.

4.11.1.3.2 The LIQUID RADWASTE TREATMENT SYSTEM shall be demonstrated OPERABLE by operating the LIQUID RADWASTE TREATMENT SYSTEM equipment for at least 15 minutes at least once per 92 days unless the LIQUID RADWASTE TREATMENT SYSTEM equipment has been utilized to process radioactive liquid effluents during the previous 92 days.

Per reactor unit FARLEY-UNIT 2 3/4 11-6 AMENDMENT NO.

l

Table 4.11-2 (Continu:d)

TABLE NOTATION

b. Analyses shall also be performed following shutdown from >15% RATED THERMAL POWER, startup to >15% RATED THERMAL POWER or a 1EERMAL POWER change exceeding 15 percent of the RATED THERMAL POWER within a one hour. period.

c. Tritium grab samples shall be taken from the plant vent stack at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the refueling canal is flooded.

d. Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after changing (or after removal from sample r) . Sampling shall also be perfonned at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for at least 2 days following each shutdown from >15% RATED THERMAL POWER, startup to >15% RATED THERMAL POWER or THERhAL POWER change exceeding 15 percelit of RATED THERMAL POWER in one hour and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months of changing. When samples collected for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months are analyzed, the corresponding MDC may be increased by a factor of 10.

e. Tritium grab samples shall be taken at least once per 7 days from the ventilation exhaust from the spent fuel pool area, whenever spent fuel is in the spent fuel pool,

f. The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period covered by each dose or dose rate calculation made in accordance with Specifications 3.11.2.1, 3.11.2.2 and 3.11.2.3.

g. The principal gamma emitters for which the MDC specification applies exclusively are the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, In-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144 for particulate emissions. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measureable and identifiable, together with the above nuclides, shall also be identified and reported.

h. Deviations from MDC requirements of Table 4.11-2 shall be reported per Specification 6.9.1.8. {

1. A composite particulate sample is one in which the quantity of air sampled is proportional to the quantity of air discharged. Either a specimen which is representative of the air discharged may be accumulated and analyzed or the individual samples may be analyzed and weighted in proportion to their respective volume discharged.

j. The principal gamma emitters for which the MDC specification applies exclusively are the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135 and Xe-138 for gaseous emissions. This does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported.

FARLEY-UNIT 2 3/4 11-11 AMENDMENT N0. ,

1

1 RADI0 ACTIVE EFFLUENTS DOSE - N0BLE GASES LIMITING CONDITION FOR OPERATION 3.11.2.2 The air dose due to noble gasts released in gaseous ef?llents, from each reactor unit, from the site'(see Figure 5.1-3) shall be l'niced to the following:

a. During any calendar quarter: Less than or equal to 5 irrad for ganma radiation and less than or equal to 10 mrad for beta radiation, and

b. Juring any calendar year: Less than or equal to 10 mrad for gamma radiation and less than or equal to 20 mrad for beta radiation.

APPLICABILITY: At all times.

ACTION

a. With the calculated air dose from radioactive noble gases in gaseous '

effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special l Report which identifies the cause(s) for exceeding the limit (s) and defines the corrective actions to be taken to reduce the releases of radioactive noble gases in gaseous effluents during the remainder of the current calendar quarter and during the rt.mainder of the current calendar year, so that the cumulative dose is within 10 mrad for gamma radiation and 20 mrad for beta radiation.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.2 Dose Calculations Cumulative dose contributions for the current calendar quarter and current calendar year shall be determined in accordance with the ODCM at least once per 31 days.

FARLEY-UNIT 2 3/4 11-12 AMENDMENT NO.

i l

l RADI0 ACTIVE EFFLUENTS DOSE - RADIOI0 DINES, RADI0 ACTIVE MATERIALS IN PARTICULATE FORM, AND RADIONUCL1 DES OTHER THAN NOBLE GASES LIMITING CONDITION FOR OPERATION 3.11.2.3 The dose to an individual from radiciodines and radioactive materials in particulate form, and radionuclides (other then noble gases) with half-lives greater than 8 days in gasecus effluents released, from each reactor unit, from the site (see Figure 5.1-3) shall be limited to the following:

a. During any calendar quarter: Less than or equal to 7.5 mrem to any organ, and

b. During any calendar year: Less than or equal to 15 mrem to any organ.

APPLICABILITY: At all times.

ACTION:

a. With the calculated dose from the release of radiofodines, radioactive materials in particulate form, or radionuclides (other than noble gases) with half lives greater than 8 days, in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies l the cause(s) for exceeding the limit and defines the corrective actions to be taken to reduce the releases of radioiodines and radioactive materials in particulate form, and radionuclides (other than nobles gases) with half-lives greater than 8 days in gaseous effluents during the remainder of the current calendar quarter and during the remainder of the current calendar year, so that the cumulative dose or dose commitment to an individual from these releases is within 15 mrem to any organ.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS

=,

4.11.2.3 Dose Calculations Cumulative dose contributions for the current calendar quarter and current calendar year shall be determined in accordance with the ODCM at least once per 31 days.

FARLEY-UNIT 2 3/4 11-13 AMENDMENT NO.

r-l l

RADI0 ACTIVE EFFLUENTS GASE0US RADWASTE TREATMENT l

LIMITING CONDITION FOR OPERATION 3.11.2.4 The GASE0US RADWASTE TREATMENT SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM shall be OPERABLE. The appropriate portions of the GASEOUS RADWASTE TREATMENT SYSTEM shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected gaseous effluent air doses due to gaseous effluent releases from the site (see Figure 5.1-3), when averaged over the calendar quarter, would exceed 0.6 mrad for gamma radiation and 1.2 mrad for beta radiation.* The appropriate portions of the VENTILATION EXHAUST TREATMENT SYSTEM shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected doses due to gaseous effluent releases from the site (see Figure 5.1-3) when averaged over the calendar quarter would exceed 0.9 mrem to any organ.*

APPLICABILITY: At all times.

ACTION:

a. With the GASE0US RADWASTE TREATMENT SYSTEM and/or the VENTILATION EXHAUST TREATMENT SYSTEM inoperable for more than 31 days or with gaseous waste being discharged without treatment and in excess of the above ifmits, prepare and submit to the Commission within 30 days, l pursuant to Specification 6.9.2, a Special Report which includes the following information:

1. Identification of the inoperable equipment or subsystems and the reason for inoperability,

2. Action (s) taken to restore the inoperable equipment to OPERABLE status, and

3. Summary description of action (s) taken to prevent a recurrence.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.4.1 Doses due to gaseous releases from the site shall be projected, based on computed operations history, at least once per 31 days, in accordance with the ODCM.

4.11.2.4.2 The GASE0US RADWASTE TREATMENT SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM shall be demonstrated OPERABLE by operating the GASE0US RADWASTE TREATMENT SYSTEM equipment and the VENTILATION EXHAUST TREATMENT SYSTEM equipment for at least 15 minutes, at least once per 92 days unless the appropriate system has been utilized to process radioactive gaseous effluents during the previous 92 day s.

Doses are per reactor unit.

FARLEY-UNIT 2 3/4 11-14 AMENDMENT NO.

l:

r j- RADI0 ACTIVE EFFLUENTS

! 3/4.11.3 RADWASTE SOLIDIFICATION LIMITING CONDITION FOR OPERATION 4 .

l 3.11.3 The radwaste solidification system shall be OPERABLE and used, as 1

applicable in accordance with a PROCESS CONTROL PROGRAM, .for the SOLIDIFICATION

!- and packaging of radioactive wastes to ensure meeting the requirements of 10CFR Part 20 and of 10CFR Part 71 prior to shipment of radioactive wastes from the site.

APPLICABILITY: At all times.

ACTION:

i a. With the packaging requirements of 10CFR Part 20 and/or 10CFR Part 71 not satisfied, suspend shipments of defectively packaged solid radioactive wastes from the site. l

b. With the radwaste solidification system inoperable for more than 31 days, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which includes the following information: ,

1. Identification of the inoperable equipment or subsystens and the reason for inoperability,

2. Action (s) taken to restore the inoperable equipment to OPERABLE status,

3. A description cf the alternative used for SOLIDIFICATION and packaging of radioactive wastes, and

4. Summary description of action (s) taken to prevent a recurrence.

c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

I SURVEILLANCE REQUIREMENTS 4.11.3.1 The radwaste solidification system shall be demonstrated OPERABLE at least once per 92 days by:

a. Operating the radwaste solidification system at least once in the previous 92 days in accordance with the PROCESS CONTROL PROGRAM, or

b. Verification of the existence of a valid contract for SOLIDIFICATION to '

-be performed by a contractor in accordance with a PROCESS CONTROL PROGRAM.

l l

FARLEY-UNIT 2 3/4 11-17 AMENDMENT NO.

l t

b

RADI0 ACTIVE EFFLUENTS 3/4.11.4 TOTAL DOSE LIMITING CONDITION FOR OPERATION 3.11.4 The dose or dose commitment to any member of the public, due to releases of radioactivity and radiation, from uranium fuel cycle sources shall be limited to less than or equal to 25 mrem to the total body or any organ (except the thyroid, which shall be limited to less than or equal to 75 mrem) over 4 consecutive quarters.

APPLICABILITY: At all times.

ACTION:

a. With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Specification 3.11.1.2.a , 3.11.1.2.b , 3.11.2.2.a , 3.11.2.2.b , 3.11.2.3.a , o r 3.11.2.3.b, prepare and submit a Special Report to the Director, Nuclear Reactor Regulation, U.S. Nuclear Regulatory Comission, Washington, D.C. 20555, within 30 days, which defines the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding the limits of Specification 3.11.4. This Special Report shall include an analysis which estimates the radiation exposure (dose) to a member of the public from uranium fuel cycle sources (including all effluent pathways and direct radiation) for a 4 consecutive quarter period that includes the release (s) covered by this report. If the estimated dose (s) exceeds the limits of Specification 3.11.4, and if the release condition resulting in violation of 40CFR190 has not already been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40CFR190 and including the specified information of 5 190.11(b). Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete. The variance only relates to the limits of 40CFR190, and does not apply in any way to the requirements for dose limitation of 10CFR Part 20, as addressed in other sections of this technical specification.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 1

4.11.4 Dose Calculations Cumulative dose contributions from liquid and gaseous effluents shall be detennined in accordance with Specifications 4.11.1.2, 4.11.2.2, and 4.11.2.3, and in accordance with the ODCM.

I FARLEY-UNIT 2 3/4 11-19 AMENDMENT NO.

I 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM LIMITING CONDITION FOR OPERATION

-=

3.12.1 The radiological environmental monitoring program shall be conducted as specified in Table 3.12-1.

APPLICABILITY: At all times.

ACTION:

a. With the radiological environmental monitoring program not being conducted as specified in Table 3.12-1, prepare and subnit to the Commission, in the Annual Radiological Operating Report, a description of the reasons for not conducting the program as required and the plans

for preventing a recurrence.

b. With the level of radioactivity in an environmental sampling medium exceeding the reporting levels of Table 3.12-2 when averaged over any calendar quarter, prepare and submit to the Connission within 30 days

! from the end of the affected calendar quarter a Special Report pursuant to Specification 6.9.2. When more than one of the radionuclides in Table 3.12-2 are detected in the sampling medium, this report shall be submitted if:

concentration (1) + concentration (2) + .. .> 1.0 limit level (1) limit level (2)

When radionuclides other than those in Table 3.12-2 are detected and are the result of plant effluents, this Special Report shall be submitted if l the potential annual dose to an individual is equal to or greater than the calendar year limits of Specifications 3.11.1.2, 3.11.2.2 and 1

3.11.2.3. This Special Report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual i Radiological Environmental Operating Report.

c. With milk or fresh leafy vegetable samples unavailable from one or more of the sample locations required by Table 3.12-1, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the cause of the unavailability of samples and identifies locations for obtaining replacement samples. The locations from which samples were unavailable may then be deleted from those required by Table 3.12-1, provided the locations from which the replacement samples were obtained are added to the environmental monitoring program as replacement locations,

d. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

FARLEY-UNIT 2 3/4 12-1 AMENDMENT NO.

1 l

l 1

l t

c ;

l Table 4.12-1 (C ntinued)

TABLE NOTATION

b. M)C for drinking water.

c. Other peaks which are measurable and identifiable, together with the radionuclides in Table 4.12-1, shall be identified and reported per Specification 6.9.1.8.

1 I

1 l

i h

l l

I 4

)

i I

[

FARLEY-UNIT 2 3/4 12-9 AMENDMENT NO.

RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.2 LAND USE CENSUS LIMITING CONDITION FOR OPERATION 3.12.2 A land use census shall be conducted and shall identify the location of the nearest milk animal and the nearest residence.

APPLICABILITY: At all times.

ACTION:

a. With a land use census identifying a location (s) which yields a calculated dose or dose commitment greater than the values currently being calculated in Specification 4.11.2.3, prepare and submit to the Commission w; thin 30 days, pursuant to Specification 6.9.2, a Special

, Report which identifies the new location (s).

b. With a land use census identifying a location (s) which yields a calculated dose or dose commitment (via the same exposure pathway) 20 percent greater than at a location from which samples are currently being obtained in accordance with Specification 3.12.1, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the new location. The new location shall be added to the radiological environmental monitoring program within 30 days. The sampiing location, excluding the control station location, having the lowest calculated dose or dose commitment (via the same exposure pathway) may be deleted from this monitoring program after (October 31) of the year in which this land use census was conducted.

c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.12.2 The land use census shall be conducted at least once per 12 months between the dates of (June 1 and October 1) using that information which will provide the best results, such as by a door-to-door survey, aerial survey, or by consulting local agriculture authorities.

FARLEY-UNIT 2 3/4 12-10 AMENDMENT N0.

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l I

POWER DISTRIBUTION LIMITS BASES 3

The radial peaking factor Fxy(Z), is measured periodically to provide additional assurance that the hot channel factor, Fq (Z), remains within its limit. The F xy limit for RATED THERMAL POWER (F%P) as provided in the Radial Peaking Factor limit report per' Specification 6.9.1.11 was determined from expected power control maneuvers over the full range of burnup conditions in the core. l 3/4.2.4 00ADRANT POWER TILT RATIO The quadrant power tilt ratio limit assures that the radial power distribution satisfies the design values used in the power capability analysis. Radial power distribution measurements are made during startup testing and periodically during power operation.

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

The two hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correction of a dropped or misaligned control rod. In the event such action does not correct the tilt, the margin for uncertainty on Fq is reinstated by reducing the maximum allowed power by 3 percent for each percent of tilt in excess of 1.0.

3/4.2.5 DNB PARAMETERS The limits on the DNB related parameters assure that each of the parameters are maintained within the normal steady state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to 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 through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation. The 18 month periodic measurement of the RCS total flow rate is adequate to detect flow degradation and ensure correlation of the flow indication channels with measured flow such that the indicated percent flow will provide sufficient verification of flow rate on a 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months basis.

FARLEY-UNIT 2 B3/4 2-5 AMENDMENT NO.

REACTOR COOLANT SYSTEM BASES 3/4.4.6 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 1.03, 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 also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.

The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking. The extent of cracking during plant operation would be limited by the 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 demonstrated 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 imperfections exceeding the plugging limit of 40% of the tube nominal 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 inspection fall into Category C-3, these results will be reported to the Commission pursuant to 10CFR50.73 prior to resumption of plant operation. Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary.

FARLEY-UNIT 2 B3/4 4-3 AMENDMENT NO.

(

ADMINISTRATIVE CONTROLS

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 experiments completed under the provision of Section 50.59,10CFR, to verify that such actions did not constitute an unreviewed safety question and all programs required by Specification 6.8 and changes thereto.

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

d. Proposed tests or experiments which may involve an unreviewed safety question as defined in Section 50. 59, 10C FR.

e. Proposed changes to Technical Specifications or this Operating License,

f. Reports of violations of codes, regulations, orders, Technical Specifications, license requirements, or of internal proceiures or instructions having nuclear safety significance or reports of abnomal degradation of systems designed to contain radioactive material.

g. Reports of significant operating abnomalities or deviations from normal and expected perfomance of plant equipment that affect nuclear safety.

h. ALL REPORTABLE EVENTS.

{

i. All recognized indications of an unanticipated deficiency in some aspect of design or operation of safety related structures, systems, or components.

J. The plant Security Plan.

k. The Emergency Plan.

1. Facility operations to detect potential nuclear safety hazards.

m. Investigations or analyses of special subjects as requested by the Chairman of the Nuclear Operations Review Board.

n. Every unexpected onsite release of radioactive material to the environs resulting from a lack of preplanning including the preparation and forwarding of reports covering evaluation, recomendations and disposition of the corrective action to prevent recurrence to the Plant Manager and to the Nuclear Operations Review Board.

o. Changes to the PROCESS CONTROL PROGRAM and the 0FFSITE DOSE CALCULATION MANUAL.

FARLEY-UNIT 2 6-8 AMENDMENT NO.

. ADMINISTRATIVE CONTROLS ALTERNATES 6.5.2.3 All alternate members shall be appointed in writing by the NORB Chairman to serve on a temporary basis; however, no more than two alternates shall participate as voting members in NORB activities at any one time.

CONSULTANTS 6.5.2.4 Consultants shall be utilized as detennined by the NORB Chainnan to provide expert advice to the NORB.

MEETING FREQUENCY 6.5.2.5 The NORB shall meet at least once per calender quarter during the initial year of unit operation following fuel loading and at least once per six months thereafter.

QUORUM 6.5.2.6 A quorum shall consist of the Chairman or Vice Chairman plus enough voting members to constitute a majority of the NORB. No more than a minority of the quorum shall have line responsibility for operation of the facility. For ,

the purpose of a quorum those considered to have line responsibility will l include the Plant Manager and personnel reporting to the Plant Manager.

REVIEW 6.5.2.7 The NORB shall review:

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

b. Proposed tests or experiments which involve an unreviewed safety question as defined in Section 50.59, 10CFR.

c. Violations of codes, regulations, orders, Technical Specifications, license requirements, or of internal procedures or instructions having nuclear safety significance or abnormal degradation of systems designed to contain radioactive material,

d. Significant operating abnormalities or deviations from normal and expected performance of plant equipment that affect nuclear safety,

e. All REPORTABLE EVENTS.

f. Recognized indications of an unanticipated deficiency in some aspect of design or operation of safety related structures, systems, or components.

FARLEY-UNIT 2 6-10 AMENDMENT NO.

i ADMINISTRATIVE CONTROLS i

6.6 REPORTABLE EVENT ACTION 6.6.1 The following actions shall be taken for REPORTABLE EVENTS:

a. The Comission shall be notified and a report submitted pursuant to the requirements of Section 50.73 to 10CFR Part 50, and

b. Each REPORTABLE EVENT shall be reviewed by the PORC, and the results of this review shall be submitted to the NORB and the Senior Vice President.

6.7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:

a. The facility shall be placed in at least HOT STANDBY within one hour.

b. The NRC Operations Center shall be notified by telephone as soon as possible and all cases within one hour. The Senior Vice President shall be notified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

c. A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the PORC. This report shall describe (1) appitcable circumstances preceding the violation, (2) effects of the violation upon facility components, systems or structures, and (3) corrective action taken to prevent recurrence.

d. The Safety Limit Violation Report shall be submitted to the Commission, and the Senior Vice President for NORB review within 14 days of the violation.

6.8 PROCEDURES AND PROGRAfts 6.8.1 Written procedures shall be established, implemented and miintained -

covering the activities referenced below:

a. The applicable procedures recorinended in Appendix "A" of Regulatory Guide 1.33, Revision 2,1978.

b. Refueling operations.

c. Surveillance and test activities of safety related equipment.

d. Security P1an impiementation.

e. Emergency Plan implementation.

FARLEY-UNIT 2 6-14 AMENDMENT NO.

f ADMINISTRATIVE' CONTROLS I-6.9 > REPORTING REQUIREMENTS ROUTINE REPORTS 6.9.1 In addition.to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Director of the Regional Office of Inspection and Enforcement unless otherwise noted.

STARTUP REPORT l

6.9.1.1 A summary report of plant startup and power escalation testing shall be submitted following (1) receipt of an operating license, (2) amendment to the j

license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel l supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant.

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FAktEY-UNIT 2 6-15a AMENDMENT NO.

i

ADMINISTRATIVE CONTROLS

e. Type of container (e.g., LSA, Type A, Type B, Large Quantity), and

f. Solidification agent (e.g., cement, urea formaldehyde).

The radioactive effluent release reports shall include unplanned releases from the site to unrestricted areas of radioactive materials in gaseous and liquid effluents on a quarterly basis, lj The radioactive effluent release reports shall include any changes to the PROCESS ;

CONTROL PROGRAM (PCP) made during the reporting period. '

MONTHLY OPERATING REPORT 6.9.1.10 Routine reports of operating statistics and shutdown experience, '

including documentation of all challen;es to the PORV's or safety valves shall be submitted on a monthly basis to the Director, Office of Managecent and Program Analysis, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, with a copy to the Regional Office of Inspection and Enforcement, no later than the 15th of each month following the calendar month covered by the report.

Any changes to the OFFSITE DOSE CALCULATION MANUAL shall be submitted with the Monthly Operating Report within 90 days in which the change (s) was made ef fecti ve. In addition, a report of any major changes to the radioactive waste treatment systems shall be submitted with the Monthly Operating Report for the period in which the change was implemented.

RADIAL PEAKING FACTOR LIMIT REPORT 6.9.1.11 The F limit for Rated Thennal Power (F RTP) shall be provided to the Director of thkRegional Office of Inspection andinforcement, with a copy to the Director, Nuclear Reactor Regulation, Attention Chief of the Core Performance Branch, U.S. Nuclear Regulatory Conunission, Washington, D.C. 20555 for all core planes containing bank "D" control rods and all unrodded core planes at least 60 days prior to cycle initial criticality. In the event that the limit would be submitted at some other time during core life, it will be submitted 60 days prior to the date the limit would become effective unless otherwise exempted by the Commission.

Any infonnation needed to support Fywill be by request from the NRC and need not be included in this report.

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

FARLEY-UNIT 2 6-19 AMENDMENT NO.

I e

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 sample in which the limit was exceeded (in graphic and tabular format); (2) Results of the last isotopic analysis for radioiodine performed prior .to exceeding the limit, results of analysis while limit was exceeded and results of one analysis after the radioiodine activity was redaced to less than the limit. Each result should include date and 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 to the first sample in which the-limit was exceeded; (4) Graph of the I-131 concentration (micro C1/gm) and one other radioiodine isotope concentration (micro Ci/ge) 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 presence of greater than or equal to 0.005 microcuries of removable contamination.

SPECIAL REPORTS' 6.9.2 Special reports shall be submitted to the Director of the Office of Inspection and Enforcement Regional Office within the time period specified for each report.

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 years:

a. Records and logs of unit operation covering time interval at each power l evel .

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

c. ALL REPORTABLE EVENTS submitted to the Comission.

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

e. Records of changes made to the procedures required by Specification 6.8.1.

f. Records of radioactive shipments.

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

FARLEY-UNIT 2 6-20 AMENDMENT NO.

F ADMINISTRATIVE. CONTROL'S .

h.: Records of annual physical inventory of all sealed source material: of record.-

6.10.2 The following -records shall 'be retained for.the duration of. the Unit Operating License:

a. . Records and drawing changes. reflecting unit design modifications' made to

. systems and equipment described in the Final Safety Analysis Report. .

-b. Records of new and irradiated fuel . inventory, fuel transfers 'and assembly burnup histories. .

c. Records of radiation exposure for all individuals entering radiation control areas.-

d. Records' of gaseous and liquid radioactive material released to the environs.

e. Records of transient or operational cycles for those unit components identified in Table 5.7-1. ,

f. Racords of reactor tests and e' xperiments.

g. Records of training and qualification for current members of the facility staff.

h. Records of in-service inspectior;s perfonned pursuant to these Technical Specifications.

i. Records of Quality Assurance activities required by the QA Manual.

j. Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59.

k. Records of meetings of the PORC and the NORB. .

1. Records of secondary water sampling and water quality.

m. Records of analyses required by the radiological environmental monitoring program.

1

n. Records of the service lives of all hydraulic and mechanical snubbers within .the scope of 3/4.7.9 including the date at which the service life commences and associated installation and maintenance records..

6.11 RADIATION PROTECTION PROGRAM Procedures for personnel radiation protection shall_ be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure.

FARLEY.-UNIT ~2 6-21 AMENDMENT NO. l

F ADMINISTRATIVE CONTROLS

-3 6.12 HIGH RADIATION AREA 6.12.1 In lieu of the " control device" or " alarm signal" required by paragraph 20.203(c)(2) of 10 CFR 20, each high radiation area in which the intensity of radiation is greater than 100 mrem /hr but less than 1000 mrem /hr shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit.* Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the followir.g:

a. A radiation monitoring device which continuously indicates the radiation dose rate in the area.

b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alams when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate level in the area has been established and personnel have been made knowledgeable of them.

c. A health physics qualified individual (i.e., qualified in radiation protection procedures) with a radiation dose rate monitoring device who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the facility Health Physics Supervisor.

6.12.2 In addition to the requirements of 6.12.1, areas accessible to personnel with radiation levels such that a major portion of the body could receive in one hour a dose greater than 1000 mrem shall be provided with locked doors to prevent unauthorized entry, and the keys shall be maintained under the administrative control of the Shift Foreman on duty and/or health physics supervision. Doors shall remain locked except during periods of access by personnel under an approved Radiation Work Permit which shall specify the dose rate levels in the immediate work area and the maximum allowable stay time for individuals in that area. For individual areas accessible to personnel with radiation levels such that a major portion of the body could receive in one hour a dose in excess of 1000 mrem ** that are located within large areas, such as PWR containment, where no enclosure exists for purposes of locking, and no enclosure can be reasonably constructed around the individual areas, then that area shall be roped off, conspicuously posted and a flashing light shall be activated as a waraing device. In lieu of the stay time specification of the RWP, direct or remote (such as use of closed circuit TV cameras) continuous surveillance may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities within the area.

Health Physics personnel or personnel escorted by Health Physics personnel shall be exempt from the RWP issuance requirement during the performance of their assigned radiation protection duties, provided they are following plant radiation protection procedures for entry into high radiation areas.

- Measurement made at 18" from source of radioactivity. I l

FARLEY-UNIT 2 6-22 AMENDMENT NO.

I

O _.

l ADMINISTRATIVE CONTROLS

'6.13. PROCESS CONTROLLPROGRAM (PCP)L o _

s a

6.13.1 The' PCP shall 'be appr'.-ved by the Commission prior. to implementation.

3 c6.13.2 ' I.icensee initiated changes to the PCP:

p, a V so 1. Shall be submitted to the' Commission in the semi-annual Radioactive-1 Effluent Release Report for the period in which the change (s) was made.

This submittal shall contain:

p+ a. Suffic'iently detailed information to totally support'the rationale. '

for'the change without benefit of additional or supplemental W, information; d og b.- A detemination that the change did not reduce the overall conformance of the solidified waste program to exiting criteria for -

,X solid wastes; and

[; c. . Documentation of the-fact $ hat the change'has been reviewed and found acceptable by the PORC.

2. -Shall become ' effective upon review and approval in accordance with

. Specification 6.5.3.1. ,

s e n.

1 6.14. OFFSITE DOSE-CALCULATION MANUAL (00CM) l 6.14.1 khe ODCM shall be SpircYed by the Commission prior to implementation.

,4 -

6.14'.2 Licensee initiated cnanges to the ODCM:

+ >\

1. Shall-be submitted!to the Commission in the Monthly Operating Report

~ within 90 days of the date the change (s) was made effective. This

,. submittal shall contain:

e

a. Sufficiently detatied infomation to. totally support the rationale i

for the change without benefit ofl additional or supplemental infomation. Information submitted should consist of a package of 2

, those pages of the ODCM to be changed with each page numbered and

>> provided with an approval and 'date box,.together with appropriate analyses ort evaluations justifyir.g ther change (s);

b. A detemination .that the change will not reduce the accuracy or

reliability of' dose -calculationslor setpoint deteminations; and

^

c. Documentation of the fact that tSe chlange has bee 4 reviewed and

,found acceptable by the PORC..

i ' 2. Shall becomd effective Specification 6.5.3.1!_ upon review and approval in accordance with D-3 f, FARLEY-UNIT 2 6-23 AMENDMENT NO.

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y jh' ji ' .. _. -

_ - _ , . - _ , . . _ _ . , . , _ ._.__ ,.._ .,_.e. _ _ _.__ _ . ._

g.

ADMINISTRATIVE CONTROLS 6.15 MAJOR CHANGES TO RADI0 ACTIVE WASTE TREATMENT SYSTEMS (Liquid, Gaseous, and Solid) 6.15.1 Licensee initiated major changes to the radioactive waste systems (Liquid, Gaseous and Solid):

1. Shall be reported to the Connission in the Monthly Operating Report for

- the period in which the change was implemented. The discussion of each change shall contain:

a. A summary of the evaluation that led to the determination that the change could be made in accordance with 10CFR50.59;

b. Sufficient detailed information to totally support the reason for the change without benefit of additional or supplemental information;

c. A detailed description of the equipment, components and processes involved and the interfaces with other plant systems;

d. An evaluation of the change which shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste that differ from those previously predicted in the license application and amendments thereto;

e. An evaluation of the change which shows the expected maximum exposures to individuals in the unrestricted area and to the general population that differ from those previously estimated in the license application and amendments thereto;

f. A comparison of the predicted releases of radioactive materials, in liquid and gaseous effluents and in solid waste, to the actual releases for the period prior to when the changes are to be made;

g. An estimate of the exposure to plant operating personnel as a result of the change; and

h. Documentation of the fact that the change was reviewed and found acceptable by the PORC.

2. Shall become effective upon review and approval in accordance with Specification 6.5.3.1.

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l l FARLEY-UNIT 2 6-24 AMENDMENT N0. l L

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p 4.0 Environmental Conditions 4.1 Unusual or Important Environmental Events

'Any occurrence of an unusual or important event that indicates or could result in significant environmental impact causally related to plant operation shall be recorded and reported to the NRC in accordance with 10CFR50.72(b)(2)(vi) or by a written report per Subsection 5.4.2, as appropriate. The following are examples:

excessive bird impaction events, onsite plant or animal disease outbreaks, mortality or unusual occurrence of any species protected by the Endangered Species Act of 1973, fish kills, increase in nuisance organisms or conditions and unanticipated or emergency discharge of waste water or chemical substances.

No routine monitoring programs are required to implement this condition.

4.2 Environmental Monitoring 4.2.1 Aerial Remote Sensing Vegetation comunities of the site and vicinity within 1 kilometer of the cooling towers in all directions shall be aerially photographed to detect and assess the significance of damage, or lack thereof, as related to cooling tower drift dispersions. Photography shall be done by aerial overflight during May or June.

Monitoring shall include a program of low altitude false color aerial photography (either color infrared photography or multispectral or multiband photography).

The scale for full coverage shall be adequate to 4-1 AMENDMENT NO.

- -. - .

ML20099L329

Text

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