ML17289A344

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Proposed TS Amends to 3/4.4.2, Safety Relief Valves & 3/4/3/7/5, Accident Monitoring Instrumentation. Table 3.3.7.5-1, Accident Monitoring Instrumentation Included W/Amended Action Statements
ML17289A344
Person / Time
Site: Columbia Energy Northwest icon.png
Issue date: 02/21/1992
From:
WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:
Shared Package
ML17289A343 List:
References
NUDOCS 9202280011
Download: ML17289A344 (11)
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Text

REACTOR COOLANT SYSTEM 3/4.4.2 SAFETY/RELIEF VALVES LIMITING CONDITION FOR OPERATION ACTION: (Continued) valve(s) within 2 minutes or if suppression pool average water tempera-ture is llO'F or greater, place the reactor mode switch in th'e Shut-down position.

~ ~ a I I gL ( ~sorel)er vntLcl sf l<e if<~ po>A'sons snag[ s'c'gr aa C. Nit~hone or more safety/relief valves inoperable, restore to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next l2 hours and in COLD SHUTDOM within the following 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s~~=

SURVEILLANCE RE UIREMENTS 4.4.2 The OPERABLE by performance

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each safety/relief valve shall be demonstrated

a. CHANNEL CHECK at l=ast once per 31 days, and a
b. CHANNEL CALIBRATION at least once per 18 months.**
    • The provisions of Specification 4.0.4 are not applicable provided the surveillance is performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure is adequate to perform the test.

Refueling u for no later th M '1-4I~

, or until the first forced n ect repair/repl acement

'V2022800li 92022i PDR ADOCK 05000397 P PDR MASHINGTON NUCLEAR - UNIT 2 3/4 4-7a Amendment No. N,96

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TABLE 3.3.7.5-1 ACCIDENT MONITORING INSTRUMENTATION MINIMUM APPLICABLE REQUIRED NUMBER CNANNELS OPERAT IONAI INSTRUHENT OF CHANNELS OPERABLE CONDITIONS ACTION

l. Reactor Vessel Pressure 1, 2 80
2. Reactor Vessel Water Level 1, 2 BO
3. Suppression Chamber Water Level 80
4. Suppression Chamber Water Temperature 2/sector i/sector 1, 2 80
5. Suppression Chamber Air Temperature 1, 2 SO Z~
6. Drywell Pressure 1, 2 SO Q.
7. Drywell Air Temperature 1, 2 80 re U
8. Drywell Oxygen Concentration 1, 2 80
9. Drywell Hydrogen Concentration 1, 2 80
10. Safety/Relief Valve Position Indicators + @va 1 ve 1/val ve 1, 2 ll. Suppression Chamber Pressure 1, 2
12. Condensate Storage Tank Level 1, 2 80
13. Hain Steam Line Isolation Valve Leakage Control System Pressure 1, 2 80 no later than Hay c en uration to effect repair/

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CONTFIlOLLED COPY Table 3.3.7.5-1 (Continued)

ACCIDENT HOHITORIHG IHSTRUHEHTATIOH ACTIOH STATEHEHTS ACTION 80 "

a. kith the number of OPERABLE accident monitoring instrumentation channels less than the Required Humber of Channels shown in Table
3. 3.7.5-1, restore the inoperable channel(s) to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
b. arith the number of OPERABLE accident monitoring instrumentation channels less than the Minimum Channels OPERABLE requirements'of Table 3.3.7.5-1, restore the inoperable channel(s) to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at 1east HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

ACTIOH 81 kith the number of OPERABLE accident monitoring instrumentation channels less than required by the Hinimum Channels OPERABLE requirement, either restore the inoperable channel(s) to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or:

a. Initiate the preplanned alternate method of moni.oring the appropriate parameter(s), and In lieu of any other report required by Specification,6.9.1, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 14 days following the even. outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

gC. TXOrJ WASHINGTON NUCLA~R - UNIT 2 3/4 3-73

t Insert to page 3/4 3-73 Table 3.3.7.5-1 (Continued)

ACCIDENT MONITORING INSTRUMENTATION ACTION STATEMENTS ACTION 82 - With the number of OPERABLE Safety/Relief Valve Position Indicator instrumentation channels less than the Minimum Channels OPERABLE requirement of Table 3.3.7.5-1, a ~ restore an inoperable channel to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and

b. verify operability and perform daily surveillance of the Tailpipe Temperature Monitoring instrument for the affected SRV until the Minimum Channels OPERABLE requirement is satisfied. Absent an OPERABLE Tailpipe Temperature monitor for the affected SRV restore the inoperable Tailpipe Tem-perature Monitor to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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TABLE 4. 3.7. 5-1 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE RE UIREHENTS APPLICABLE CHANNEL CHANNEL OPERATIONAL INSTRUt 1ENT CHECK CALIBRATION CONDITIONS

1. Reactor Vessel Pressure H 1, 2
2. Reactor Vessel Water Level H 1, 2
3. Suppression Chamber Water Level M 1, 2 Suppression -Chamber Water Temperature 1, H 2 A
5. Suppression Chamber Air Temperature Primary, Containment Pressure M 1 2 2

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H 1 i

7. Drywell Air Temperature H l., 2 XI
0. Drywell Oxygen Concentration H 1 2 Q f'I
9. Drywell Hydrogen Concentration H 1, 2 I
10. Safety/Relief Valve Position Indicators+ 1, 2 ro C7 ll. Suppression Chamber Pressure t1 1; 2 A
12. Condensate Storage Tank Level tl 1, 2 0
13. IIain Steam Line Isolation Valve M 1, 2 Leakage Control System Pressure
14. Neutron Flux:

APRH tl 1, 2 IRH tl 1 2 SINAI I M 1, 2

15. RCIC Flow H 1 2
16. HPCS Flow M 1, 2
17. LPCS Flow H 1 2

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~ CONTROLLED COPY REACTOR COOLANT SYSTEM BASES 3/4.4.2 SAFETY/RELIEF VALVES (Continued) the dual purpose safety/relief valves in their ASHE Code qualified mode (spring lift) of safety operation.

The overpr essure protection system must accommodate the most severe pres-surization transient. There are two major transients that represent the most severe abnormal operational transient resulting in a nuclear system pressure rise. The evaluation of these events with the final plant configuration has shown that the HSIV closure is slightly more severe when credit is taken only for indirect derived scrams; i.e., a flux scram. Utilizing this worse case transient as the design basis event, a minimum of 12 safety/relief valves are required to assure peak reactor pressure remains within the Code limit of 110'f design pressure.

Testing of safety/relief valves is normally performed at lower power. It is desirable to allow an increased number of valves to be out of service during testing. Therefore, an evaluation of the HSIV closure without direct scram was performed at 25K of RATEQ THERHAL POMER assuming only 4 safety/relief valves were operable. The results of this evaluation demonstrate that any 4 safety/relief valves have sufficient flow capacity to assure that the peak reactor pressure remains well below the code limit of llOX of design pressure.

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Qemonstration of the safety/relief valve lift settings will be performed in accordance with the provisions of Specification 4.0.5.

3/4.4.3 REACTOR COOLANT SYSTEH LEAKAGE 1 LEAKAGE QETECTION SYSTEMS

'/4.4.3.

The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the reactor coolant pressure boundary. These detection systems are consistent with the recommen'dations of Regulatory Guide 1.45, "Reactor Coolant Pressure Boundary Leakage Qetection Systems," May 1973.

MASHINGTON NUCLEAR " UNIT 2 B 3/4 4-la Amendment No. 80

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Insert to BASES 3/4.. SAFETY RELIEF VALVES THI Action Plan Item II.D.3 "Direct Indication of Relief and Safety Valve Position," states that reactor coolant system relief and safety valves shall be provided with a positive indication in the control room derived from a reliable valve-position detection device or a reliable indication of flow in the discharge pipe. Each WNP-2 SRV has both a valve stem position indication device and an acoustic monitor flow detection device which independently meet the requirements of Item II.D.3. Hence failure of one device does not impact compliance to II.D.3 and entry into Limiting Condition for Operation action statement 3.4.2.c is required only for inoperability of both devices associated with a specific SRV.

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