ML20027C356

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Proposed Tech Spec Pages 3/4 4-17 & 3/4 4-19 Re RCS Operational Leakage & RCS Pressure Isolation Valves.Safety Evaluation Encl
ML20027C356
Person / Time
Site: Farley Southern Nuclear icon.png
Issue date: 10/11/1982
From:
ALABAMA POWER CO.
To:
Shared Package
ML20027C353 List:
References
NUDOCS 8210150344
Download: ML20027C356 (6)
v • d • e


Text

REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITIN5 CONDITION FOR OPERATION I

3.4.7.2 Reactor Coolant System leakage shall b'e limited to:

a. No PRESSURE BOUNDARY LEAKAGE, I
b. 1 GPM UNIDENTIFIED LEAKAGE,
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 Coolant System pressure of 2235 + 20 psig.
f. As specified in Table 3.4-1 at a Reactor Coolant System pressure of 2235 + 20 psig, a maximum allowable leakage of 3 gpm for any Reactor Coolant System Isolation Valve of nominal diameter of 2 inches and of 5 gpm (with certain limitations) for any Reactor Coolant System Isolation Valve of nominal diameter greater than 2 inches.

APPLICABILITY: MODES 1; 2, 3 and 4 ACTION:

a. With any PRESSURE B0UNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
b. With any Reactor Coolant System leakage greater than any one of the

. above limits, excluding PRESSURE B0UNDARY LEAKAGE, reduce the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. -

c. With any Reactor Coolant System Pressure Isolation Valve leakage h greater that the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 69

- use of at least two closed manual or deactiviated autom3 tic valves, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD "

SH0TDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

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SURVEILLANCE REQUIREMENTS .

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4.4.7.2.1 Reactot, Coolant System leakages shall be demonstrated to be within ,

each of the above limits by; e

a. Monitoring the containment atmosphere particulate radioactivity monitor at least once per 12' hours. ,

.b. ' Monitoring the containment air gooler condensate level system or containmerit atmosphere gasegus radioactivity monitor at least' once 9

per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. ^ 3

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TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ~ ISOLATION VALVES MAX. (#) MAX. (#)

ALLOWABLE ALLOWABLE LEAKAGE LEAKAGE TPNS DESCRIPTION (GPM) TPNS DESCRIPTION (GPM)

Q2E11V001A 12" Gate 5 Q2E21V062A 2" Check 3 Q2E11V001B 12" Gate 5 Q2E21V062B 2d Check 3 Q2E11V016A 12" Gate 5 Q2E21V062C 2" Check 3 Q2E11V016B 12" Gate 5 Q2E21V066A 2" Check 3 Q2E11V021A 6" Check 5 Q2E21V066B 2" Check 3 Q2E11V021B 6" Check 5 Q2E21V066C 2" Check 3 Q2E11V021C 6" Check 5 Q2E21V076A* 6" Check 5 Q 2E11V042A 10" Check 5 Q2E21V076B* 6" Check 5 Q2E11V042B 10" Check 5 Q2E21V077A* 6" Check 5 Q2E11V051A 6" Check 5 Q2E21V077B* 6" Check 5 Q2E11V0516 6" Check 5 Q2E21V077C 6' Check 5 Q2E11V051C 6" cneck 5 Q2E21V078A 2" Check 3 Q2E21V032A* 12" Check 5 Q2E21V0788 2" Check 3 Q2E21V032B* 12" Check 5 Q2E21V078C 2" Check 3 Q2E21V032C* 12" Check 5 Q2E21V079A 2" Check 3 Q2E21V037A* 12" Check 5 Q2E21V079B 2" Check 3 Q2E21V037B* 12" Check 5 Q2E21V079C 2" Check 3 Q2E21V037C* 12" Check 5 ,

(#) A. The limitations for maximum allowable leakage of Reactor Coolant System Isolat' ion Valves with a nominal ?iameter greater than two inches is as follows:

1. Leakage rates less than or equal to 1.0 gpm are considered acceptable. However, for intial tests, or test (following valve repair or replacement, leakage rates less than or< equal to 5.0 gpm are considered acceptable.
2. Leakage rates greater than 1.0 gpm but less' than or equal,to 5.0 gpm are considered acceptable if the latest measured rate has not exceeded the rate determined by the previous test by an amount *that reduces the margin between measured leakage rate and the maximum permissible rate of35.0 gpm by 50% or greater.
3. Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are considered unacceptable if the latest measured rate ~

exceeded the rate determined by the previous test by an amount that redu'ces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater.

4. Leakage rates greater than 5.0 gpm, are considered unacceptable.

B. The maximum allowable leakage of Reactor Coolant System Isolation '

Valves with a nominal diameter of two inches is.3 gpm.

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FARLEY - UNIT 2 3/4 4-19 O m 3 b

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REACTOR COOLANT SYSTEM . . _. _

l OPERATIONAL LEAKAGE . . _

LIMITING CONDITION FOR OPERATION 3.4,.7.2 Reactor Coolant System leakage shall be limited to: .

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.a. No PRESSURE BOUNDARY LEAKAGE,

b. 1 GPM UNIDENTIFIED LEAKAGE, .
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 LEAXAGE from the Reactor Coolant System, and
e. 31 GPM CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2235 2 20 psig.

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a. With any PRESSURE SOUNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />'.
b. With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE, reduce the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
c. With any Reactor Coolant System Pressure Isolation Valve leakage greater that the above limit, isolate the high pres'sure portion of

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the affected system from the icw pressure portion within,4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by

> - use of at least two closed manual or deactivated automat'ic valves, or Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD "

SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.-

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SURVEILLANCE REOUIREMENTS ._

4.4.7.2.1 Reactor Coolant System leakages shall be demonstrated to be within 1

, each of the above limits by; *

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( a. Monitoring the containment atmosphere particulate radioactivity ,

monitor at Teast once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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. h. Monitoring the containment air cooler condensate level system or- -

containment atmosphere gaseous radioactivity monitor at least once . '

per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. .

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Attachment 2 Safety Evaluation For One-Time Change of Unit 2 Technical Specification 3/4.4.7.2

. I. Back gr ound Technical Specification 3/4.4.7.2 currently specifies the performance of leak tests on all reactor coolant system pressure isolation valves. The acceptance criteria of 1 gpm contained in the Unit 2 Technical Specification has proven to be too restrictive and has resulted in unacceptable delays in the return to power.

Alabama Power Company requests a one-time change of Technical Specification 3/4.4.7.2 to modify the acceptance criteria to 3 gpm for reactor coolant isolation valves with a nominal diameter of two inches and 1 to 5 gpm (with certain limitations) for reactor coolant isolation valves with a nominal diameter greater than two inches.

II.

References:

P (1) FNP Unit 1 Technical Specification 3/4.3.7.3 (2) FNP Unit 2 Technical Specificaticn 3/4.4.7.2 III. Buses It is imperative that the acceptance criteria for reactor coolant system isolation valve leakage be established prior to shutdown in

. order that this critical path not be impacted. The initial valve testing is performed on the critical path of the unit outage in -

order to identify valves requiring subsequent repair during' the outage. Af ter valve repair and re-installation, the valves tare then retested for leakage as a part of the critical path of the return to power. If the results of leak retest do.not satisfy the stringent 1-gpm acceptance criteria of the current technial i specification, the critical path of the cutage may be further impacted due to valve rework, re-installation, retest, and the draindown of reactor coolant system required to repair several of these valves. The difference in acceptance criteria from 1 gpm to 3 or' 5 spm requires a variance in leakage configuration that is negligible; inconsequential, and beyond#that which can be 'affected by reasonable maintenance activities.

As an example, the disk of a six-inch check valve would be elevated l from its seat by approximately 0.00004 in. to produce a qleak rate of 1 gpm and by approximately 0.00020 in, to produce a Teak rate of 5 gpm. To correct small irregularities that produce leak rates

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between 1 g'pm an~d 3 or 5 gpm, several maintenance and retesting a iterati6ns may be necessary. ~ This activity subjects plant

. personnel to greater radiation exposures, potentially requires draining the . reactor coolant system to retest the reworked valve and extends the time before the unit is returned to power.

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Attachment 2 Page 2 The Unit I leak test acceptance criteria of 1 to 5 gpm, (under certain limitations) have been proven to be adequate in establishing the pressure retaining capability of the valves. As shown by testing experience at Farley Nuclear Plant-Unit 1, valves that did not satisfy either the acceptar.ce criteria of 1 gpm or 1 to 5 gpm were found to contain the same minor valve seating irregularities causing the valves not to seat completely under low test pressures, and no evidence of impending valve failere has been found using either acceptance criteria.

For valves with a nominal diameter of two inches, the acceptance criteria of 3 gpm is requested based on review of past leak rate data for two-inch valves. This data showed that such valves did not produce leak rates greater than 3 gpm. The 3-gpm acceptance criteria was selected for smaller valves since repair is easier, actual leakage rates are not significant, and the ability to meet and maintain lower leakage rates is facilitated. Additionally, the acceptance criteria of 3 gpm is below the 5 gpm criteria that has been proven to establish the pressure retaining capability of the valves.

This proposed.one-time change to Unit 2 Technical Specification 3/4.4.7.2 provides for a high assurance of reactor coolant system integrity through surveillance and testing requirements without unwarranted compromise to the health and safety of the public and needless jeopardizing of the timely return of the_ unit to power ~

operation.

IV. Concl usion This proposed change to the Unit 2 Technical Specifbcation-

' represents an improvement in testing required for the reactor coolant system pressure isolation valves and does not involve an unreviewed safety question as defined by 10 CFR 50.59.

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