ML19337A561

Proposed Tech Specs 3.4.6.2,4.4.6.1 & 3.4-1,permitting Leak Testing of Valves in Mode 3 & Allowing Util 30 Days to Accumulate Data at Rated Sys Pressure to Establish Baseline Leakage Limits for RHR Sys Supply Line Valves

ML19337A561
Person / Time
Site:	Sequoyah
Issue date:	09/24/1980
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML19337A552	List: ML19337A551 ML19337A561
References
NUDOCS 8009290280
Download: ML19337A561 (5)

Text

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ENCLOSURE 1 PROPOSED CllANGES TO SEQUOYAH NUCLEAR PLANT UNIT 1 .

TECHNICAL SPECIFICATIONS

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• REACTOR COOLANT SYSTEM

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OPERATIONAL LEAKAGE LIVITING CONDITION FOR OPERATION

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3.4.6.2 Reactor Coolant System leakage shall be limited to:

a. No PRES 5URE 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 LEAKAGE from the Reactor Coolant System,

e. 40 GPM CONTROLLED LEAKAGE at a Reactor Coolant System pressure of 2235 + 20 psig.

f. 1 GPM leakage from any Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1.*

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

a. With any PRESSURE BOUNDARY 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, and leakage from Reactor Coolant System Pressure Isolation Valves, 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 than 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 /> by use of at least two closed manual or deactivated automatic 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 />.

SURVEILLANCE REQUIREMENTS 4.4.6.1 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by:

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• Specific exceptions to the 1 GPM leakage limit and the MODE 3 and 4 l applicability are listed on Table 3.4-1.

SEQUOYAH - UNIT 1 3/4 4-14 l

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TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES VALVE NUMBER FUNCTION 63-586 Baron Injection 63-587 Boron Injection 63-588 Boron Injection 63-589 Baron Injection 63-581 Boron Injection 63-560 f 63-561 Accumulator Accumulator Discharge l) (l)

Discharge 63-562 Accumulator Discharge (l)63-563 Accumulator Discharge (l)63-622 Accumulator Discharge 63-623 Accumulator Discharge 63-624 Accumulator Discharge 63-625 Accumulator Discharge 63-551 Safety Injection (Cold Leg)63-553 Safety Injection (Cold Leg)63-557 Safety Injection (Cold Leg)63-555 Safety Injection (Cold Leg) .63-632 Residual Heat Removal (Cold Leg)(1)63-633 Residual Heat Removal (Cold Leg)(1)63-634 Residual Heat Removal (Cold Leg)(1)63-635 Residual Heat Removal (Cold Leg)(1)63-641 Residual Heat Removal / Safety Injection (Hot Leg)63-644 Residual Heat Removal / Safety Injection (Hot Leg)63-558 Safety Injection (Hot Leg)63-559 Safety Injection (Hot Leg)63-543 Safety Injection (Hot Leg)63-545 Safety Injection (Hot Leg)63-547 Safety Injection (Hot Leg)

63-549 Safety Injection (Hot Leg) 63-640- Residual Heat Removal (Hot Leg)63-643 Residual Heat Removal (Hot Leg)87-558 . Upper Head Injection 87-599 Upper Head Injection 87-560 Upper Head Injection 87-561 Upper Head Injection 87-562 Upper Head Injection 87-563 Upper Head Injection FCV-74-1 Residual Heat Removal (1)(2)

FCV-74-2 Residual Heat Removal (1)(2)

(1) These valves must be tested prior to entering MODE 2.

(2) The leakage limit for these valves is 3 GPM. This value will be finalized within 30 days of issuance of this amendment.

I SEQUOYAH - UNIT 1 3/4 4-15a

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ENCLOSURE 2 REASONS AND JUSTIFICATIONS FOR PROPOSED CHANGES TO SEQUOYAH NUCLEAR PLANT UNIT 1 TECHNICAL SPECIFICATIONS c

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1. Specification 3.4.6.2.f. Page 3/4 4-14 Add a festnote to this specification to note that there are exceptions to the one GPM leakage limit and the requirement to test before enter-ing MODE 4. These exceptions are identified in Table 3.4-1.

2. Table 3.4-1, Page 3/4 4-15a The leakage test procedure for the cold leg injection line check

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valves requires that the Reactor Coolant System (RCS) pressure be greater than the cold leg accumulator pres.ure. The plant must be in MODE 3 to achieve this pressure. The specific test procedures were discussed with A. Schwencer of the NRC staff by telephone on September 23, 1980. The affected check valves are:

63-560, -561, -562, -563 Cold Leg Injection Line 63-632, -633, -634, -635 RHR Cold Leg The Residual Heat Removal (FHR) pumps or two Reactor Coolant Pumps (RCP) must be operating in MODE 5. This is a requirement of technical specification 3.4.1 3 In order to test the RHR supply line valves, the RHR system must be isolated. Two RCP's must be started. The pump heat from the RCP's will cause the RCS to heat up and exceed MODE 5.

TVA proposes to test the valves before entering MODE 2 to allow for normal plant heatup and isolation of the RHR system. The affected valves are:

FCV-74-1 and FCV-74-2 RHR Supply Line The present specification imposes a one GPM leakage limit on the RHR supply line valves FCV-74-1 and FCV-74-2. The baseline extrapolation, based on an ASME low to high pressure test correlation, for FCV-74-1 resulted in an projected leakage of 1.84 GPM at rated system pressure.

TVA believes that the test correlation is overly conservative. We request that the leakage limits for FCV-74-1 and FCV-74-2 be increased to three GPM for 30 days. During this period, TVA would make leakage measurements at low pressure and at rated system pressure.

Aftar the data has been analyzed, TVA would like to meet with the

'm r . a dis;ueo the results and establish a final baseline leakage value for these valves. ~Without the waiver, the plant would be pro-hibited from starting up and collecting data at rated system pressure.

Section 5.2.6.1 of Safety Evaluation Report (SER) Supplement No. 2 for Sequoyah states in part that "(L)eakage rates higher than one GPM may be considered acceptable if the leak rate changes are below one GPM above the previous test leak rate." TVA's proposal of a three GPM valve is consistent with the SER statement. The fact that these

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valves are gate valves in which valve position is known and that the leakage will be limited to one GPM above a preliminary baseline

, provides reasonable assurance that the design pressure of the RHR system will not be exceeded thus reducing the probability of an occur-rence of an intersystem LOCA. In addition, TVA believes valuable data will be collected which can be used to establish a more effective technical specification.

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