ML17291A481

Proposed Tech Spec Relocating Safety/Relief Valve Position Indication Instrumentation Requirements

ML17291A481
Person / Time
Site:	Columbia
Issue date:	10/31/1994
From:	WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:
Shared Package
ML17291A480	List: ML17291A479 ML17291A481
References
NUDOCS 9411080158
Download: ML17291A481 (11)

Text

REQUEST FOR AME<2llDMENT TO TECH SPEC: RELOCATE SAFETY/RELIEF VALVE POSITION INDICATION INSTRUMENTATION REQUIREMENTS Attachment 3 TECHNICAL SPECIFICATION PAGES WITH PROPOSED CHANGES INDICATED 9411080158 941031 PDR ADOCK 05000397 P PDR TABLE 3.3.7.5-1 ACCIDENT MONITORING INSTRUMENTATION D.B O lXL INSTRUMENT'.

Reactor Vessel Pressure 3.4, 2.Reactor Vessel Mater Level Suppression Chamber Mater Level Suppression Chamber Water Temperature 5.Suppression Chamber Air Temperature 6.Drywell Pressure 7.Drywell Air Temperature 8.Drywell Oxygen Concentration 9.Drywell Hydrogen Concentration l0.Safely%11.Suppression Chamber Pressure 12.Condensate Storage Tank Level 13.Hain Steam Line Isolation Valve Leakage Control System Pressure>%Tl~44he~he-aaous44~o REQUIRED NUMBER OF CHANNELS 2/sector HINIHUM CHANNELS OPERABLE 1/sector APPLICABLE OPERATIONAL CONDITIONS 1,,2 1, 2 1, 2.1, 2 1, 2 1.'2 1 Q 2 1, 2 1, 2 1, 2 1, 2 1, 2 ACTION 80 80 80 80 80 80 80 80 80 80 80 80 J'I tR f~" gY I+

ACTION a~b.ACTION a~b.Table 3.3.7.5-1 (Continued)

ACCIDENT MONITORING INSTRUMENTATION CTION STATEM NTS 80-With the number of OPERABLE accident monitoring instrumentation channels less than the Required Number of Channels shown in Table 3.3.7.5-I, 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 />.With 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 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 />.81-With the number of OPERABL'E accident monitoring instrumentation channels less than required by the Minimum 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: Initiate the preplanned alternate method oF monitoring 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 Specifica-tion 6.9.2-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.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 Min-imum Channels OPERABLE requirement is satisfied.

Absent an OPERABLE Tailpipe Temperature monitor for the affected SRV restore the inoper-able Tailpipe Temperature 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 />.WASHINGTON NUCLEAR-UNIT 2 3/4 3-73 Amendment No.105 1 k!l TABL 4.3.7.5-1 CCIOENT MONITORING INSTRUMENTATION SURVEILLANCE R UIR MENTS~INS RUHEN 1.Reactor Vessel Pressure 2.Reactor Vessel Water Level CHANNEL CHECK CHANNEL CALIBRATION APPLICABLE

.OPERATIONAL I N 1,2 1,2 C n I X7 3.6.7.8.Suppression Chamber Mater Level Suppression Chamber Mater Temperature Suppression Chamber'Air Temperature Primary Containment Pressure Orywell Air Temperature Orywell Oxygen Concentration 1,2 1,2 1,2 1,2 1,2 1,2 m (C/1 m O~~I~CL~Q f N'~O m 9.Orywell Hydrogen Concentration 10.Safe4yfR n-In~;-o ll.Suppression Chamber Pressure 12.Condensate Storage Tank Level 13.Hain Steam Line Isolation Valve Leakage Control System Pressure 14.Neutron Flux: APRH IRN SRH 15.RCIC Flow 16.HPCS Flow 17.LPCS Flow 1,2 1,2 1,2 1,2 1,2 1,2 1,2 1,2 1,2 1,2 NNN W&N~M4~~NN INN~~&ter-+ea cue~he'w-p~men@-channels N-NNNNNN~N~

mek&an equate-4o-p J

R ACTOR COOLANT SYSTE 3 4.4.2 SAFETY RELIEF VALV S LIMITING CONDITION FOR OPERATION~CTION: (Continued) valve(s)within 2 minutes or if suppression pool average water temperature is 110'F or greater, place the reactor mode switch in the Shutdown position.c.With both the acoustic monitor and valve stem position indicator for one or more safety/relief valve(s)inoperable, restore either the acoustic monitor or valve stem position indicator 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 in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.SURVEILLANCE RE UIREMENTS 4.4.2 The position indicators for each safety/relief valve shall be demonstrated OPERABLE by performance of a: a.CHANNEL CHECK at least once per 3)days, and a b.CHANNEL CALIBRATION at least once er 18 months.~*~Re-p~~ns-af-SpeeA~~n~OA-are-not-ap~tAe-prov-Med-the

'$4 dw A~m~~~WASHINGTON NUCLEAR-UNIT 2, 3/4 4-7a Amendment No.80-,~~'128 4~l I C I~'

REACTOR COOLANT SYSTEM BASES 3 4.4.2 SAFETY R LIEF VALVES (Continued) the dual purpose safety/relief valves in their ASME Code'qualified mode (spring lift)of safety operation.

The overpressure 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 MSIV 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%of des.ign pressure.\Testing of safety/relief valves is normally performed at lower power with adequate steam pressure and flow.It is desirable to allow an increased number of valves to be out of service during testing.Therefore, an evaluation of the MSIV closure without direct scram was performed at 25%of RATED THERMAL POWER assuming only 4 safety/relief valves were operable.The results of this evalu-ation 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 110%of design pressure.TMI Action Plan Item II.D.3,"Direct Indication of Relief and Safety Valve Position," states that reactor coolant system relief and safety valves sha",: 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 dis-charge 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 impar.-..ooi-pliance to II.D.3 and entry into Limiting Condition for Operation action state-ment 3.4.2.c is required only for inoperability of both devices associated with a specific SRV, Demonstration 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 SYSTEM LEAKAGE 3 4.4.3.1 EAKAGE D TECTION SYST MS The RCS leakage detection systems required by this specification are pro-vided 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.The primary containment sump flow monitoring system monitors the UNIDENTIFIED LEAKAGE collected in the floor drain sump with a sensitivity such that 1 gpm change within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> can be measured.Alternatively, other methods for measuring flow to the sump which are capable of=detecting a change in UNIDENTIFIED LEAKAGE of 1 gpm within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> with an accuracy of+2%may be used, for up to 30 days, when the installed system is INOPERABLE.

WASHINGTON NUCLEAR-UNIT 2 B 3/4 4-la Amendment No.86-,+05, 111, 128 1I cyyw~'b I I 1 k 0 4 lf J ll~%, W II'