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Rev 2 to Surveillance Test Procedure FNP-1-STP-9.0, RCS Leakage Test. Related Info Encl
	ML20080T296
Person / Time
Site:	Farley, 05000000
Issue date:	03/25/1979
From:	Poole D; ALABAMA POWER CO.
To:	;
Shared Package
ML20079E758	List: ML20080T285; ML20080T296; ML20080T311; ML20080T321; ML20080T345; ML20080T354; ML20080T363;
References
	FOIA-83-789 FNP-1-STP-9.0, NUDOCS 8403010137
	Download: ML20080T296 (22)
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VG . 7 FN?-1-STP-9.O March 25, 1979 Revision 2-

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FI C NUCLEAR PLANT SURVEILLANCE TEST ?ROCEDURE FNP-1-STP-9.0 .

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SURVEILUs%CE ~EST NO. TEC:-:NICAL SPEC!?ICATION REFERENCE W P-1-s'I'P-9.0 4.4.6.2d --

TITLE: MODE (s) REqt RI.w TEST:

RCS IZAKAGE TEST 1, 2 , 3 and 4 TEST RESULTS (To be completed by test perfor=ar)

PIRIOR':ED BY -

DATE/TDIE /

COMPONENT OR T uIN.TESTID (if applicable) .

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SURVE!IJ ANCE TEST PROCEDURE ST?-9.0 '

RCS *T'dCAGE TEST i

l 1.0 Purcose .. .

1 To deterrdne idan-J.fied and unidentified reactor coolant sy__ stem leakage __ .. . . .. _. _ .

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2.0 ' Acceotance Criteria -[

7 .I- ~Uniirent EIWd'1eak' Age-is 1 1~gpm. -- -

2.2 Identified 1 akage i 10 gym..

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NOTE: Asterisked steps (*) are those associated ,

with Acceptance Criteria. -

3.0 - Initial conditions -

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3.1 Reactor power and reictor coolant teinperature 1- - - -

- -'~ - ' should' be stablized and held constant for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ...

"' *"' - "N-prior to and during the test.

3.2 The pressurizer level and pressure control

pw -are-in--automatic. -- - - - - - - - -

- - - - -3 . 3 Se--leve?.-of the- VCT is near the top--of-the

_ normal operating band.

- _. . 3.4 The RCDT level is in the nornal operating band ~ -

and-the RCDT punps are OFF.

. - 3.5 The_CVCS-system .is- aligned per SOP-2.1A.--

__.4.O.. Precautions.and L ii m tat ons i'

_. _. 4.1 No. sampling of the RCS o$ CVCS shall be done during 6's ast.

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~ 4.2 Any of the following will render this test void:

l 4.2.1 Energency boration 4.2.2 Diversion of letdown to the recycle j-l , ,_

holdup ' d .

l 4.2.3 Make up from any source which doe:. not go through the beric acid blender.

! 4.3 Pressurizer temperature and RCS tenperature and i pressure final values must equal initial values.

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5.0 Inst uctions 5.1 Notify C&EP departnent of the performance of' the --

test to insure that no sa=pling of the'RCS or -

CVCS will be done during this tast.

5.2 Read and record initial readings on the data sheet.

NOTE: Batch Integrator Readings will be taken prior to, and at the conclusion of each ..

make up' evolution. --

5.3 After 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months verfy that RCS te=perature'and - i pressure and pressurizer tenperature are the ...

. same as recorded for initial readings.

. 5.4 If coolant temperature and pressure and pressurirer temperature are the same as recorded for initial values, record final values on data sheet and ,

proceed to_ step 5.6.,, ,. ,

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5.5 If any parameters in step 5.4 have changed, . . n. i .

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5.5.1 Adjust parameter to initial [ reading ?

value, -

5.5.2 Allow conditions to stabilize (approximately

. 15 minutes),

5.5.3 Record final values on the data sheet. . , . .

S.6 Calculate identified and unidentified leakages using the infornation recorded and the for=ula on the data sheet (Acceptance criteria is 3,1 gpm unidentified leakage and 5,10 gpm identified leakage).

5.7 Return RCDT system to normal per S0?-50.0.

6.0 References 6.1 P&ID D-175037 - RCS, sheet 2.

6.2 ?&ID D-175039 - CVCS, sheet 2.

6.3 ?&ID D-175042 - Waste Processing System, sheet 1.

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EACTORCOOLANTSYSTDi hfi PERATIONAL LEAKAGE ]. . . . . . .

RMITING CONDITION FOR OPERATION

.4.6.2~ . Reactor Coolant System leakage shall be limited to:

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a. No PRESSURE BOUNDARY LEAKAGE, ---

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b .~ l GPM UNIDENTIFIED LEAKAGE '

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c. 1 GPM total primary-to-secondary leakage through all steam generators and 500 gallons per day through any one steam '

generator, . ..,_

-d. 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, and {

e. 31 GPM CONTROLLED LEAKAGE at a Reactor Ccolant System

.~Z pressure of 2235 + 20 psig.

LPPLICABILITY: MODES 1, 2, 3 and 4 & 1 G _.= _ .

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LCTION: ==E.

c = .s a With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY ,

'within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 s hours. ,

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With any Reactor Coolant System leakage greater than any one

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of the above limits, excluding PRESSURE BOUNDARY LEAKAGE, reduce the leakage rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be . . - - -

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in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

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SURVEILLANCE REQUIREMENTS _ , _ .

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4.4.6.2 Reactor Coolant System leakages shall be demonstrated to be dEE cithin cach of the.above limits by; ~==-

a. Monitoring the containment atmosphere particulate radioactivity _ :sr monitor at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . .

b. Monitoring the containment air cooler condensate level system or centainment atmosphere gaseous radioactivity monitor at EE5.

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

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REACTOR COOLANT SYSTEM EE=

EE EEE SURVEILLANCE REQUIREMENTS (Continued) f_5 E'E 5EE

c. Measurement of the CONTROLLED LEAXAGE to the reactor coolant 55 pump seals when the Reactor Coolant System pressure is 2235 EEE

+ 20 psig at least once per 31 days with the modulating valve . EEE

'Tully open, EEE

d. Perfomance of a Reacto.r Coolant System water inventory balance

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at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months during steady state operation, and E5

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e. Monitoring the reactc.r head flange leakoff temperature at ss least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . ~

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OEFIN!TIONS -

PRESSURE BOUNDARY LEAKAGE 1.16 PRESSURE BOUNDARY LEAKAGE shall be leakage (except ' steam generator tube leakage) through a non-isolable . fault in a Reactor Coolant System component body, pipe all or vessel wall.

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

OUADRANT POWER. TILT RATIO 1.18 QUADRANT POWER TILT RATIO shall be the ratio of the maximum up)Cr excore detector calibrated output to the average of the upper excore detector calibrated cutputs, or the ratio of the maximum lower excore detector calibrated output ,to tha average of the lower excore detector calibrated outputs,'whichever is greater. With one excore detector .

inoperable, the remaining three detectors shall be used for computing the average.

DOSE EQUIVALENT I-131 1.19 DOSE EQUIVALENT I-131 shall be that concentration of 1-131 (uci/ gram) which. alone would produce the same thyroid dose as the quantity and iso-

^

topic mixture of I-131,- I-132, I-133,1-134, and I-135 actually present.

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

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

.a. A test schedule for n systems, subsystems, trains or other designated components obtained by dividing the specified test : interval into n equal subintervals, b., The testing of one system, subsystem, train er other designated conconent at the beginning of each subinterval.

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e FARLEY - UNIT 1 1-4

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'OEFINITIONS CHANNEL FUNCTIONAL TEST 1.11 A CHANNEL' FL'NCTIONAL TEST 'shall be the injection of a simulated signal into the channel as close to the primary sensor as practicable to verify OPERABILITY including alarm and/or trip functions.

CORE ALTERATION 1.12 CORE ALTERATION shall be the movement or manipulation of any com-

.ponent within the reactor pressure vessel with the vessel head removed and fuel-in the vessel. Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.

SHUTDOWN MARGIN 1.13 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be suberitical 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 resctivity worth which is assumed to be fully withdrawn.

IDENTIFIED LEAKAGE 1.14 IDENTIFIED LEIKAGE shall be: ,

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

b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere

with the operation of leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE, or

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

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

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RfACTOR COOLANT _ SYSTEM L 8ASES 3/4 4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection s,ystems required by this specification are provided to monitor and detect leakage from the Reactor Coolant Pressure Boundary. These detection systems are consistent with the recommendations of. Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems." May 1973.

3/4.4.6.2 OPERATIONAL LEA M Industry experience has shown that while a limited amount of leakage is e7pected from _the RCS, rhe unidentified portion of this leakage can be reduced to a threshold value of less than 1 GPM. This threshold value is sufficiently low to ensure early detection of additional leakage.

-The 10 GPM IDENTIFIED LEAKAGE limitation provides allowance for a .

limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the leakare detection systems.

.The CONTROLLED LEAKAGE limitation restricts operation when the total flow supplied to .the reactor coolant pump seals exceeds 31 GPM with the valve in-the supply line fully open at a nominal RCS pressure of 2235 psig. - This limitation ensures that in the event of a LOCA, the safety injection flow will not be less than assumed in the accident analyses.

4 The-total steam generator tube leakage limit of 1 GPM for all steam generata s ensures that the dosage contribution from the tube leakage will be limited to a small fraction of Part 100 limits in the event of either a steam generator tube rupture or steam line break. The 1 GPM p limit is' consistent with the assumptions used in the analysis of these accidents. The 500 gpd leakage limit per steam generator ensures that steam. generator tube integrity is maintained in the_ event of a main steam line ructure as under LOCA conditions.

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