Text

Proposed Tech Specs Re Support for 24-month Fuel Cycle Surveillance Extensions for Unit 3
	ML20086M115
Person / Time
Site:	Millstone
Issue date:	07/18/1995
From:	; NORTHEAST NUCLEAR ENERGY CO.
To:	;
Shared Package
ML20086M112	List: B15293, Application for Amend to License NPF-49.Amend Would Support 24-month Fuel Cycle Surveillance Extension for Unit 3; ML20086M115;
References
	NUDOCS 9507240133
	Download: ML20086M115 (38)
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h)O Cl1ctr)g l 12/16/93 INro onQ ltEACTOR COOLANT SYSTEM 3/4.4.4 RELIEF VALVES j LIMITING CONDITION FOR OPERATION 3.4.4. Both mer-operated relief valves (PORVs) and their associated block valves shall m 0PERABLE.

APPLICABILITY: MODES 1, 2, and 3.

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a. With one or both PORV(s) inoperable and capable of being manually cycled, within I hour either restore the PORV(s) to OPERABLE status -

or close the associated block valve (s) with power maintained to the block valve (s); otherwise, be in at least HDT STANDBY within the ,

next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HDT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . ;

b. With one PORY inoperable and not capable of being manually cycled, within I hour either restore the PORV to OPERABLE status or close the associated block valve and remove power from the block valve; l restore the PORV to 0PERABLE status within the following 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT STANDBY wtihin the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HDT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . -

c. With both PORVs inoperable and not capable of being manually cycled, i within I hour either restore at least one PORV to OPERABLE status or close its associated block valve and remove power from the block valve and be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . j

d. With one or both block valve (s) inoperable, within I hour restore the block valve (s) to 0PERABLE status, or place its associated PORV(s) control switch to 'CLOSE." Restore at least one block valve !

to OPERABLE status within the next hour if both block valves are :

inoperable; restore any remaining inoperable block valve to operable :

status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; otherwise, be 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 HDT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

l e. Entry into an OPERATIONAL MODE is pemitted while subject to these ACTION requirements.

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l l k 33 95071s P $CK05000423 PDR MILLSTONE - UNIT 3 3/44-12 ' Amendment No. U.88 till

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12/16/93 REACTOR COOLANT SYSTEM RELIEF VALVES SURVEILLANCE REQUIREMENTS 4.4.'4.1 In addition to the requirements of Specif.is ion 4.0.5, each PORY shall be demonstrated OPERABLE at least once per 18 mon >by:

Performance of a CHANNEL CALIBRATION, and N( REFU EImtravnL6N GL.~

a.

b. Operating the valve through one complete cycle of full travel during MODES 3 or 4.

4.4.4.2 Each block valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel "

unless the block valve is closed with pwer removed in order to meet the requirements of ACTION b. or c. in Spectif W ton 3.4.4.

4.4.4.3 The emergency power supply for PORVs and block valves shall be demonstrated OPERABLE at least18once 6er 'W'M by operating the valves through a complete cycle of full travel.

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MILLSTONE - UNIT 3 3/44-13 Amendment No. 88

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. a REACTOR C00LAKr SYSTEM

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3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING C0fGITION FOR OPERATION 3.4.6.1 The following Reactor Coolant System Leakage Detection Systems shall be OPERABLE:

a. Either the Containment Atmosphere Gaseous or Particulate Radioactivity Monitoring System, and

b. The Containment Drain Sump Level or Pumped Capacity Monitoring System APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

a. With both the Containment Atmosphete Gaseous and Particulate Radioactivity Monitors INOPERABLE, operation may continue for up to :

30 days provided the containment Drain Sump. Level or Pumped Capacity Monitoring System is OPERABLE and gaseous grab samples of the containment atmosphere are obtained at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and :

analyzed for gross noble gas activity within the subsequent 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ; otherwise, be 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 .

b. With the Containment Drain Sump Level or Pumped Capacity Monitoring System IN0PERABLE, operation may continue for up to 30 days provided either the Containment Atmosphere Gaseous or Particulate ;

Radioactivity Monitoring System is OPERABLE; othemise, be in at l 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 followng 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . -

SURVEILLANCE REQUIREMENTS 4.4.6.1 The Leakage Detection Systems shall be demonstrated OPERABLE by:

a. Containment Atmosphere Gaseous and Particulate Radioactivity .

Monitoring Systems-performance of CHANNEL CHECK, CHANNEL CALIBRATION, and ANALOG CHANNEL OPERATIONAL TEST at the frequencies specified in Table 4.3-3, and

. b.

ContainmentDrainSumpLevelandPumpedCapacity{MonitoringSy perfomance of CHANNEL CALIBRATION at least once er'IB mont@

g L, P.E NU E L.IM 6 I b17 G 8 0 A L-NILLSTONE - UNIT 3 3/4 4-21 Amendment No. T/, 77, 100

e REACTOR COOLANT SYSTEN OPERATIONAL LEAKAGE SURVEILLANCE REQUIREMENTS t

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

a. Monitoring the containment atmosphere (gaseous or partf r.ulate) radioactivity 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 drain sump inventory and discharge at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ;

c. Measurement of the CONTROLLED LEAXAGE to the reactor coolant pump seals when the Reactor Coolant System pressure is 2250 1 20 psia at least once per 31 days with the modulating valve fully open. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4;

d. Performance of a Reactor Coolant System water inventory balance at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; and

e. Monitoring the Reactor Head Flange Leakoff System at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

I 4.4.6.2.2 Each Reactor Ccolant System Pressure Isolation Valve specified in i Table 3.4-1 shall be demonstrated OPERABLE by verifying leakage to be within l its limit:

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a. At least once@er18 monIR, !

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b. Prior to entering MODE 2 whenever the plant has been in COLD '

SHUTDOWN for 7 days or more and if leakage testing has not been performed in the previous 9 months,

c. Prior to returning the valve to service following maintenance, repair or replacement work on the valve, I

d. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation due to automatic or manual action or flow through the valve, and

e. As outlined in the ASME Code,Section XI, paragraph IWV-3427(b).

The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4.

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MILLSTME - UNIT 3 3/44-13 Amendment No.100 l

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. REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEM I LINITING Co W ITION FOR OPERATION ACTION (Continued)

e. In the event the PORVs, the RHR suction relief valves, or the RCS are used to mitigate an RCS pressure transient, a Special vent (s)

Report s hall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days. The report shall describe the circumstances initiating the transient, the effect of the PORVs, the RHR suction relief valves, or RCS vent (s) on the transient, and any corrective action necessary to prevent recurrence.

f. Entry into an OPERATIONAL MODE is pemitted while subject to these ACTION requirements.

l SURVEILLANCE REQUIREMENTS i

l 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE by:

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a. Performance of an ANALOG CHANNEL OPERATIONAL TEST on the PORY l

actuation channel, but excluding valve operation, within 31 days prior to entering a condition in which the PORV is required OPERABLE and at least once per 31 days thereafter when the PORV is required OPERABLE;

b. Performance of a CHANNEL CALIBRATION on the PORV actuation channel at least once ge M 8 m po an{ g%, l

c. Verifying the PORY isolation valve is open at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months when the PORV is being used for overpressure protection.

4.4.9.3.2 Each RHR suction relief valve shall be demonstrated OPERABLE when the RHR suction relief valves are being used for cold overpressure protection as follows:

a. For RHR suction relief valve 3RHS*RV8708A, by verifying at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that 3RHS*MV8701A and 3RHS*MV8701C are open;

b. For RHR suction relief valve 3RHS*RV8708B, by verifying at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that 3RHS*MVS702B and 3RHS*MV8702C are open; and

c. Testing pursuant to Specification 4.0.5.

NILLSTONE - UNIT 3 3/44-39 Amendment No. U , 5 ,100

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No Chempe, REACTOR COOLANT SYSTEN f cg- Tn R) o oly ,

3/4.4.11 REACTOR COOLANT SYSTEN YENTS / {

.. LIMITING C'0W ITION FOR OPERATION l

3.4.11 At least one Reactor Coolant System vent path consisting of 'two )

parallel trains with two valves inseries powered from emergency busses shall '

be OPERABLE and the vent closed

at each of the following locations:

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a.- Reactor vessel head, and 1

b. Pressurizer steam space. 1 APPLICABILITY: NODES 1, 2, 3, and 4.

EIlQti:

a. With one train of the reactor vessel head vent path inoperable, STARTUP and/or POWER OPERATION may continue provided the inoperable- i train is maintained closed with power removed from the valve actuators of all valves in the inoperable train; restore the inoperable train to OPERABLE status within 30 days, or, be in 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 j 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With both trains of the reactor vessel head vent paths inoperable; maintain both trains closed with power removed from the valve actuators of all valves in the inoperable trains, and restore at least one of the trains to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in 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 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c. With any valve (s) of the pressurizer steam space vent path inoperable in NODES 1, 2, or 3, follow the- ACTION requirements of Specification 3.4.4.

d. With any valve (s) of the pressurizer steam space vent path -

inoperable in NODE 4 follow the ACTION requirements of ,

, Specification 3.4.9.3.

SURVEILLANCE REQUIRENENTS F 4.4.11.1 Each train of the reactor vessel head vent path isolation valve not required to be closed by ACTION a. or b., above, shall be demonstrated OPERABLE at least once per COLD SHUTDOWN, if not performed within the previous 92 days, by operating the valve through one complete cycle of full travel from the control room. '

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For an OPERABLE vent path using a power-operated relief valve (PORV) as the vent path, the PORV block valve is not required to be closed.

MILLSTONE - UNIT 3 3/44-43 Amendment No. 77, 100 em n .- . _

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l SURVEILLANCE REQUIREMENTS fContinued) p 4.4.11.2 Each train of the reactor vessel head vent path shall be demonstrated OPERABLE at least once geretiriiiiinIB)by g 7,erragg

' a. Verifying all manual isolation valves in each vent path are locked in the open position,

b. Cycling each vent valve through at least one complete cycle of full

. travel from the control room, and

c. Verifying flow through the Reactor Coolant System vent paths during vent 1ng.

4.4.11.3 Each train of the pressurizer steam s> ace vent path shall be demonstrated OPERABLE per the applica >1e requirement of Specifications 4.4.4.1 through 4.4.4.3 and 4.4.g.3.1. In addition, flow shall be verified through the pressurizer steam space vent path during venting at least onceder 18Dng h .

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MILLSTONE - UNIT 3 3/4 4-43a Amendment No.79, 88 sans

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1 Docket No. 50-423' B15293 ,

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Attachment 2 Millstone Nuclear Power Station, Unit No. 3 Proposed Revision to Technical Specifications ,

24-Month Fuel Cycle Reactor Coolant System Surveillance Extension Retyped Pages i

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July 1995 i

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REACTOR COOLANT SYSTEM RELIEF VALVES SURVEILLANCE REQUIREMENTS 4.4.4.I In addition to the requirements of Specification 4.0.5, each PORY shall be demonstrated OPERABLE at least once each REFUELING INTERVAL by: l

a. Performance of a CHANNEL CALIBRATION, and

b. Operating the valve through one complete cycle of full travel during MODES 3 or 4.

4.4.4.2 Each block valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel unless the block valve is closed with power removed in order to meet the requirements of ACTION b. or c. in Specification 3.4.4.

4.4.4.3 The emergency power supply for the PORVs and block valves shall be demonstrated OPERABLE at least once each REFUELING INTERVAL by operating the l valves through a complete cycle of full travel.

NILLSTONE - UNIT 3 3/4 4-13 Amendment No. pp, 0400

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l REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAXAGE ,

i LEAKAGE DETECTION SYSTEMS l l

LIMITING CONDITION FOR OPERATION j l

3.4.6.1 The following Reactor Coolant System Leakage Detection Systems shall l be OPERABLE:

a. Either the Containment Atmosphere Gaseous or Particulate Radioactivity Monitoring System, and l

b. The Containment Drain Sump Level or Pumped Capacity Monitoring System )

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

ACTION:

a. With both the Containment Atmosphere Gaseous and Particulate Radioactivity Monitors INOPERABLE, operation may continue for up to :

30 days provided the Containment Drain Sump Level or Pumped Capacity 1 Monitoring System is OPERABLE and gaseous grab samples of the containment atmosphere are obtained at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and 1 analyzed for gross noble gas activity within the subsequent 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ; otherwise, be 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 . I

)

b. With the Containment Drain Sump Level or Pumped Capacity Monitoring System IN0PERABLE, operation may continue for up to 30 days provided !

either the Containment Atmosphere Gaseous or Particulate l Radioactivity Monitoring System is OPERABLE; otherwise, be in at I I

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 followng 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.4.6.1 The Leakage Detection Systems shall be demonstrated OPERABLE by:

l

a. Containment Atmosphere Gaseous and Particulate Radioactivity Monitoring Systems-performance of CHANNEL CHECK, CHANNEL CALIBRATION, and ANALOG CHANNEL OPERATIONAL TEST at the frequencies specified in Table 4.3-3, and

b. Containment Drain Sump Level and Pumped Capacity Monitoring System-performance of CHANNEL CALIBRATION at least once each REFUELING INTERVAL.

MILLSTONE - UNIT 3 3/4 4-21 Amendment No. J/, 79, JP9, 0401

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REACTOR C0OLANT SYSTEM OPERATIONAL LEAKAGE SURVEILLANCE REQUIREMENTS l

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

a. Monitoring the containment atmosphere (gaseous or particulate) radioactivity 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 drain sump ir,ventory and discharge at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ;

c. Measurement of the CONTROLLED LEAKAGE to the reactor coolant pump seals when the Reactor Coolant System pressure is 2250 i 20 psia at least once per 31 days with the modulating valve fully open. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4;

d. Performance of a Reactor Coolant System water inventory balance at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; and

e. Monitoring the Reactor Head Flange Leakoff System at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4.4.6.2.2 Each Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1 shall be demonstrated OPERABLE by verifying leakage to be within its limit:

a. At least once each REFUELING INTERVAL, l

b. Prior to entering MODE 2 whenever the plant has been in COLD SHUTDOWN for 7 days or more and if leakage testing has not been performed in the previous 9 months,

c. Prior to returning the valve to service following maintenance, repair or replacement work on the valve,

d. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation due to automatic or manual action or flow through the valve, and

e. As outlined in the ASME Code,Section XI, paragraph IWV-3427(b).

The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4.

i MILLSTONE - UNIT 3 3/4 4-23 Amendment No. Jpp.

0402

1 REACTOR C0OLANT SYSTEN OVERPRESSURE PROTECTION SYSTEN LIMITING CONDITION FOR OPERATION ACTION (Continued)

e. In the event the PORVs, the RHR suction relief valves, or the RCS vent (s) are used to mitigate an RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days. The report shall describe the circumstances initiating the transient, the effect of the PORVs, the RHR suction relief valves, or RCS vent (s) on the transient, and any corrective action necessary to prevent recurrence,

f. Entry into an OPERATIONAL MODE is permitted while subject to these ACTION requirements.

SURVEILLANCE REQUIRENENTS 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE by:

a. Performance of an ANALOG CHANNEL OPERATIONAL TEST on the PORV actuation channel, but excluding valve operation, within 31 days prior to entering a condition in which the PORV is required OPERABLE and at least once per 31 days thereafter when the PORV is required OPERABLE;

b. Performance of a CHANNEL CALIBRATION on the PORV actuation channel at least once each REFUELING INTERVAL; and l

c. Verifying the PORV isolation valve is open at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months when the PORV is being used for overpressure protection.

4.4.9.3.2 Each RHR suction relief valve shall be demonstrated OPERABLE when the RHR suction relief valves are being used for cold overpressure protection as follows:

a. For RHR suction relief valve 3RHS*RV8708A, by verifying at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that 3RHS*MV8701A and 3RHS*MV8701C are open;

b. For RHR suction relief valve 3RHS*RV8708B, by verifying at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that 3RHS*MV8702B and 3RHS*MV8702C are open; and

c. Testing pursuant to Specification 4.0.5.

N Amendment No. 77, pp, Jpp, gLSTONE-UNIT 3 3/4 4-39

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SURVEILLANCE REQUIREMENT (Continued) 4.4.11.2 Each train of the reactor vessel head vent shall be demonstrated OPERABLE at least once each REFUELING INTERVAL by: path l

a. Verifying all manual isolation valves in each vent path are locked in the open position,

b. Cycling each vent valve through at least one complete cycle of full travel from the control room, and

c. Verifying flow through the Reactor Coolant System vent paths during venting.

4.4.11.3 Each train of the pressurizer steam space vent path shall be demonstrated OPERABLE per. the applicable requirement of Specifications 4.4.4.1 through 4.4.4.3 and 4.4.9.3.1.

flow shall be verified through the pressurizer steam space vent pathInaddition,I during venting at least once each REFUELING INTERVAL.

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NILLSTONE - UNIT 3 3/4 4-43a Amendment No. 77, pp, oo

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Docket No. 50-423 )

B15293-s i

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Attachment 3 lt Millstone Nuclear Power Station, Unit No. 3 ;

Proposed Revision to Technical Specifications 24-Month Fuel' Cycle '

Surveillance Extensions Description of the Proposed Technical.

Specification-Changes Reactor Coolant System i

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-July 1995 i

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I U.S. Nuclear Regulatory Commission ,

B15293/ Attachment 3/Page 1 !

July 18, 1995 i

Millstone Nuclear Power Station, Unit No. 3 Description of the Proposed Technical specification Changes l

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Introduction On June 7,1995, Millstone Unit No. 3 began operating on a 24-month ;

fuel cycle instead of the previous 18-month cycle. To take '

advantage of this longer fuel cycle, NNECO is proposing to modify j the frequency of a number of the surveillance requirements in the ;

Millstone Unit No. 3 Technical Specifications. The proposed .

changes are described below: 1 Dgscription of the Proposed Chanaes

1. Surveillance Reauirements 4.4.4.1.a and 4.4.9.3.1.b. Channel l Calibration for the Power Operated Relief Valves (PORVs) ;

Surveillance Requirements 4.4.4.1.a and 4.4.9.3.1.b verifies the operability of the PORVs by performance of a channel calibration at least once per 18 months. NNECO proposes to extend the frequency of Surveillance Requirements 4.4.4.1.a i and 4.4.9.3.1.b from at least once per 18 months to at least once each refueling interval (i.e., nominal 24 months).

2. Sections 4.4.4.1 and 4.4.4.3, Relief Valves. Surveillance i Reauirements Surveillance Requirement 4.4.4.1 verifies the operability of each power operated relief valve (PORV) by a performance of a channel calibration and by operating the valve through one complete cycle of full travel during Modes 3 or 4 at least once per 18 months. Surveillance Requirement 4.4.4.3 verifies the operability of the emergency power supply for the PORVs and block valves by operating the valves through a complete cycle of full travel at least once per 18 months. NNECO proposes to extend the frequency of Surveillance Requirements 4.4.4.1 and 4.4.4.3 from at least once per 18 months to at least once each refueling interval (i.e., nominal 24 months).

3. Section 4.4.6.1.b. Channel Calibration of Containment Drain Sumo Level and Pumoed Capacity Monitorina Instrumentation Surveillance Requirement 4.4.6.1.b verifies the operability of the containment drain sump level and pump capacity monitoring instrumentation by performance of channel calibration at least once per 18 months. NNECO proposes to extend the frequency of Surveillance Requirement 4.4.6.1.b from at least once per 18 months to at least once each refueling interval (i.e.,

nominal 24 months).

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U.S. Nuclear Regulatory Commission-

.B15293/ Attachment 3/Page 2 July 18, 1995

4. Eggtion 4.4.6.2.2. Reactor ' Coolant System (RCS) Pressure Isolation Valves. Surveillance Reauirements Surveillance Requirement 4.4.6.2.2.a verifies the operability of each RCS pressure. isolation valve by verifying leakage to be within its limit at least once per 18 months. NNECO.

proposes to extend the frequency of Surveillance Requirement 4.4.6.2.2.a from at least once per 18 months to at least once each refueling interval (i.e., nominal 24 months).

5. Section 4.4.11.2 and 4.4.11.3. Reactor Coolant System Vents.

Surveillance Reauirements Surveillance Requirement 4.4.11.2 verifies the_ operability of each reactor vessel head vent path. at least once per 18 months. Surveillance Requirement 4.4.11.3 verifies the operability of each of the pressurizer steam space vent path at least once per 18 months. NNECO. proposes to~ extend the frequency of Surveillance Requirement 4.4.11.2 and 4.4.11.3 from at least once per 18 months to at least once each refueling interval (i.e., nominal 24 months).

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@ i Docket No. 50-423 L6- B15293 l

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Attachment 4 I

Millstone Nuclear Power' Station, Unit No. 3

, t Proposed Revision.to Technical Specifications 24-month Fuel Cycle s

safety Assessment and Significant' Hazards l Consideration for-Changes to Reactor Coolant System i l

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July 1995 1

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l' U.S. Nuclear Regulatory Commission B15293/ Attachment 4/Page 1 July 18, 1995

. Millstone Nuclear Power 8tation, Unit No. 3 Proposed Revision to Technical Specifications 24-month Fuel Cycle Safety Assessment and Significant Easards consideration for Changes to Reactor Coolant system Safety Assessment and Sionificant Hazards Consideration Backaround On June 7,1995, Millstone Unit No. 3 began operating on a 24-month fuel cycle instead of the previous 18-month cycle. To be consistent with this longer fuel cycle, NNECO is proposing to modify the frequency of a number of the surveillance requirements existing in'the Millstone Unit No. 3 Technical Specifications.

The safety assessment and significant hazards consideration for'the-proposed changes to sections 4.4.4.1, 4.'4.4.3. (Relief ' Valves,.

Surveillance Requirements), Section 4.4.6.1.b (Leakage Detection Systems), Section 4.4.6.2.2 (RCS Pressure Isolation Valves),

Section 4.4.9.3.1.b (Overpressure Protection L System), . Sections .

4.4.11.2 and 4.4.11.3 (Reactor Coolant Vents, Surveillance Requirements) are described below. In the near future, NNECO will be proposing additional changes to the Millstone Unit No. 3 Technical Specifications to prepare for the conversion to 24-month 4 fuel' cycles. Each of these submittals will contain evaluations that are independent and which stand alone.

I. Relief Valves. Surveillance Recuirements 4.4.4.1 and 4.4.4.3 and Reactor Coolant System Vents. Surveillance Reauirements 4.4.11.3 Safety Assessment Surveillance Requirement 4.4.4.1 verifies the operability of each power-operated relief valve (PORV) by a performance of a channel calibration (Surveillance Requirement 4.4.4'.1.a) and by operating ' the valve through one complete cycle of full travel during Modes 3 or 4 (Surveillance Requirement 4.4.11.b) at least once per 18 months. Surveillance Requirement 4.4.4.3 verifies the operability of the emergency power supply for the PORVs and block valves. by operating the valves through a complete. cycle of-full travel ~at least-once per.18 months.

Surveillance Requirement 4.4.11.3 requires that each train of the pressurizer steam space vent path shall be demonstrated operable per applicable requirement of Specification 4.4.4.1 through 4.4.4.-3 and 4.4.9.3.1. In addition, flow is verified through the pressurizer steam space vent path during venting at least once per 18 months. The first portion of the surveillance requirement is covered under Surveillance

U.S. Nuclear Regulatory Commission B15293/ Attachment 4/Page 2 July 18, 1995 Requirement 4.4.4.1 through 4.4.4.3. The flow path through the PORV and block valve is verified by either a temperature rise in the down stream pipe or a reduction in RCS pressure.

NNECO proposes to extend the frequency of Surveillance Requirements 4.4.4.1.b, 4.4.4.3 and 4.4.11.3 from at least once per 18 months to at least once each REFUELING INTERVAL (i.e., nominal 24 months). The safety assessment regarding Surveillance Requirements 4.4.4.1.a and 4.4.9.3.1.b will be covered under a separate evaluation.

The proposed change to Surveillance Requirements 4.4.4.1.b, 4.4.4.3 and 4.4.11.3 does not alter the intent or method by which the surveillances are conducted, does not involve any physical changes to the plant, does not alter the way any structure, system or component functions and does not modify the manner in which the plant is operated. As such, the proposed change to the frequency of Surveillance Requirements 4.4.4.1.b, 4.4.4.3 and 4.4.11.3 will not degrade the ability of each PORV and block to perform its intended function.

Equipment performance over the last four operating cycles was evaluated to determine the impact of extending the frequency of Surveillance Requirements 4.4.4.1.b, 4.4.4.3 and 4.4.11.3.

This evaluation included a review of surveillance results, preventive maintenance records and the frequency and type of corrective maintenance. The component covered under these surveillances are: ,

Train 'A' valves are 3RCS*MV8000A (block valve) and j 3RCS*PCV455A (PORV).

Train 'B' valves are 3RCS*MV8000B (block valve) and 3RCS*PCV456 (PORV).

The PORVs are electrically controlled, pressure actuated, poppet type relief valves with magnetically actuated position indication. Actuation is achieved utilizing a three way solenoid actuated switchgear valve. One case of not meeting the stroke time acceptance criteria occurred during 1986.

Investigation determined the slow stroke time was caused by the orifice in the solenoid valves that vent the PORVs to open i them. A design change was implemented to remove the orifices and the stroke times were found normal. All subsequent tests (from 1986 to date) met the acceptance criteria (opening and closing stroke times).

PORV maintenance problems have occurred with the indication circuits and have been traced to a weakening of the magnet that operates the position switch. This does not affect the valve operability sjnce a temperature rise or pressure drop provides a proof for the correct operation of the valve.

U.S. Nuclear Regulatory Commission B15293/ Attachment 4/Page 3 July 18, 1995 There have been no failures of the PORV's or the block valves to operate.

In addition, Surveillance Requirement 4.4.4.2 provides additional assurance that the block valves are operable.

On the basis of the above evaluation, there is a reasonable assurance that the frequency of Surveillance Requirements 4.4.4.1.b, 4.4.4.3 and 4.4.11.3 can be extended from at least once per 18 months to once each refueling interval (i.e.,

nominal 24 months) without significantly affecting valve reliability.

The proposed change has been evaluated from a PRA perspective, and it has been determined that the change does not represent a significant plant risk.

Sianificant Hazards Consideration NNECO has reviewed the proposed change in accordance with ,

10CFR50.92 and has concluded that the change does not involve a significant hazards consideration (SHC). The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not compromised. The proposed change does not involve an SHC because k the change would not: '

1. Involve a significant increase in the probability or consequences of an accident previously analyzed.

The proposed change to Surveillance Requirements 4.4.4.1.b, 4.4.4.3 and 4.4.11.3 extends the frequency for demonstrating operability of the PORVs and block valves. The proposal would extend the frequency from at least once per 18 months to at least once each refueling interval (i.e., nominal 24 month).

The proposed change to surveillance Requirements 4.4.4.1.b, 4.4.4.3 and 4.4.11.3 does not alter the intent or method by which the surveillances are conducted. In addition, the acceptance criterion for each surveillance is unchanged. As such, the proposed change to the frequency of Surveillance Requirements 4.4.4.1.b, 4. 4.4. 3 and 4. 4.11. 3 will not degrade the ability of the PORVs and block valves to perform their intended function.

An evaluation of past surveillances, preventive maintenance records and the frequency and the type of corrective i maintenance concluded that decreasing the surveillance j frequency will have little impact on safety. Since the proposed change only affects the surveillance frequency, the proposed change cannot affect the probability of any previously analyzed accident. While the proposed change can lengthen the intervals between surveillances, the increase in

U.S. Nuclear Regulatory Commission B15293/ Attachment 4/Page 4 July 18, 1995 intervals has been evaluated and it is concluded that there is no significant impact on the reliability or availability of the PORVs or block valves and consequently, there is no impact on the consequences of any analyzed accident.

2. Create the possibility of a new or different kind of accident from any previously analyzed.

The proposed change to Surveillance Requirements 4.4.4.1.b, 4.4.4.3 and 4.4.11.3 does not modify the design or operation of any plant system. The proposed change does not alter the intent or method by which the surveillance is conducted other than increasing the interval from 18 months to 24 months (nominal). The proposed change does not introduce a new failure mode. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously analyzed.

3. Involve a significant reduction in the margin of safety.

Changing the frequency of Surveillance Requirements 4.4.4.1.b, 4.4.4.3 and 4.4.11.3 from at least once per 18 months to at least once each refueling interval does not change the basis for frequency. The proposed change does not alter the intent or method by which the surveillances are conducted, does not involve any physical changes to the plant, does not alter the -

way any structure, system or component functions and does not modify the manner in which the plant is operated. Further, the previous history of reliability of the PORVs and block valves provides assurance that the change will not affect the reliability of these valves. Thus the proposed change has no impact on the margin of the safety.

II. Reactor Coolant System Pressure Boundary Isolation Valves.

Surveillance Reauirement 4.4.6.2.2.a Safety Assessment Surveillance Requirement 4.4.6.2.2.a verifies operability of the reactor coolant system pressure boundary isolation valves l by verifying leakage to be within the limit at least once per l 18 months. NNECO proposes to extend the frequency of l surveillance requirement 4.4.6.2.2.a from at least once per 18 l months to at least once each refueling interval (i.e. , nominal 24 months).

Reactor Coolant System (RCS) pressure boundary isolation valves are provided to prevent overpressurization and rupture of the emergency core cooling system (ECCS) low pressure piping which could result in a loss of coolant accident (LOCA). These valves provide isolation between RCS and the i following systems: low pressure safety injection system (SIL),

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U.S. Nuclear Regulatory Commission B15293/ Attachment 4/Page 5 July 18, 1995 high pressure safety injection (SIH) and residual heat removal (RHR). These are twenty-seven check valves and four motor operated valves (MOVs). These are listed in Table 1.

The . proposed change to Surveillance Requirement 4.4.6.2.2.a does not alter the intent or method by which the surveillances are conducted, does not involve any physical changes to the plant, does not alter the way any structure, system or component functions, and does not modify the manner in which the plant is operated. As such, the proposed change to the frequency of Surveillance Requirement 4.4. 6. 2. 2.a will not degrade the ability of each RCS pressure boundary isolation valve to perform its intended function.

Equipment performance over the last operating cycle was evaluated to determine the impact of extending the frequency of Surveillance Requirement 4.4.6.2.2.a. This evaluation included a review of surveillance results, preventive maintenance records and the frequency and type of corrective maintenance.

Surveillances for the valves in the scope of Surveillance Requirement 4.4.6.2.2.a are covered by Surveillance Procedure SP 3601F.4. Acceptance criterion for the valves in the scope of this surveillance is 0.5 gpm leakage per nominal inch of valve size up to maximum of 5 gpm at an RCS pressure of 2230-2270 psia. The summary of the surveillance test scope and results are provided below:

1. SI Accumulator check valves SIL*V15,17,19,21: 10 tests were performed with no failures

2. SIH/RCS cold leg loops 1, 2, 3 and 4, check valves RCS*V29, 70, 106, 145: 21 tests performed with two failures (one in 1988 and one 1992) . The failure in 1988 was attributed to accumulation of crud between the valve clapper and the valve seat thus preventing closure. The line was flushad and the subsequent test met the acceptance criterion. The failure in 1992 occurred because the valve did not seat properly at low RCS pressure (350 psia) . At normal RCS. pressure (2250 psia) ,

it passed the test.

3. RCS loop 1 and 4 suction valve RHS*MV8701C, . 8702C, 8701A, and 8702B: 27 tests were performed with no failures.

4. SIL to RCS check valves RVS*V30, 71, 107 and 146: 20 tests were performed with one failure. This failure occurred because the valve did not seat properly at low RCS pressure. When tested at higher RCS pressure (2250 psia), it met the acceptance criterion.

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U.S. Nuclear Regulatory Commission B15293/ Attachment 4/Page 6 July 18, 1995

5. RHR/SI to RCS loops, 1, 2, 3 and 4 check valve: 23 tests were performed with no failures.

6. SIH to RCS cold legs check valve SIH*V5: 12 tests were performed with no failures.

7. SIH/RCS loops 1, 2, 3 and 4 hot leg check valves SIL*V27, 29, 110 and 112: 15 tests were performed with one failure. The valve was retested and it passed.

8. RHR/SI to RCS loops 2 and 4 check valves SIL*V26, 28: 13 tests were performed with no failures.

9. RCS hot leg check valves RCS*V26, 69, 102, 142: 12 tests were performed with no failures.

It is noted that the number of tests for each group of valves are different because the technical specification requires the same test to be performed on other occasions per Surveillance Requirements 4.4.6.2.2.b,c, and d.

Based on the above surveillance test results, the reliability of the RCS isolation valves is considered high.

The majority of corrective maintenance was performed on the four MOVs due to their inherent complexity comparing to the check valves. The corrective maintenance on the valves involved packing leakage, and position indication adjustments.

i The only preventive maintenance that is scheduled on an 18-month frequency for the MOVs involved visual inspection of the overall valve condition including electrical connection, limit switches, torque switches and verification of the grease presence. These tasks are not considered critical to valve performance and there is no indication that the proposed extension could cause deterioration in valve condition or lubricants condition, or prevent the valves from performing their safety function (i.e., actuate as required).

Based on the engineering review of equipment performance, preventive and corrective maintenance history, the proposed change is considered acceptable. In addition, the RCS leakage is continuously monitored by verifying RCS inventory at least once every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Significant valve leakage would be identified and the appropriate corrective actiori would be implemented.

The proposed change has been evaluated from a PRA perspective, and it has been determined that the change does not represent a risk to public safety.

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U.S. Nuclear Regulatory Commission l B15293/ Attachment 4/Page 7 l July 18, 1995 Sianificant Hazards Consideration NNECO has reviewed the proposed change in accordance with 10CFR50.92 and has concluded that the change does not involve an SHC. The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not compromised. The proposed change does not involve a SHC because the change would not:

1. Involve a significant increase in the probability or consequences of an accident previously analyzed.

The prcposed change to Surveillance Requirement 4.4.6.2.2.a extends the frequency for demonstrating operability of the RCS pressure boundary isolation valves by verifying leakage to be withirc the limit. The proposal would extend the frequency from at least once per 18 months to at least once each refueling interval (i.e., nominal 24 months).

The proposed change to Surveillance Requirement 4.4.6.2.2.a does not alter the intent or method by which the surveillances are conducted. In addition, the acceptance criterion for each

' surveillance is unchanged. As such the proposed change to the frequency of Surveillance Requirement 4. 4. 6. 2. 2. a will not degrade the ability of the RCS pressure boundary isolation valves to perform their intended function.

An evaluation of past surveillances, preventive maintenance records and the frequency and the type of corrective maintenance concluded that decreasing the surveillance frequency will have little impact on safety. Since the proposed change only affects the surveillance frequency, the proposed change cannot significantly affect the probability of any previously analyzed accident. While the proposed change can lengthen the intervals between surveillances, the increase in intervals has been evaluated and it is concluded that there is no significant impact on the reliability or availability of the RCS pressure boundary isolation valves and consequently, there is no impact on the consequences of any analyzed accident.

2. Create the possibility of a new or different kind of accident from any previously analyzed.

The proposed change to surveillance Requirement 4.4.6.2.2.a does not modify the design or operation of any plant system.

The proposed change does not alter the intent or method by which the curveillances are conducted other than increasing the interval from 18 month to 24 months (nominal). The proposed change does not introduce a new failure mode.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously analyzed.

U.S. Nuclear Regulatory Commission B15293/ Attachment 4/Page 8 July 18, 1995

3. Involve a significant reduction in the margin or safety.

Changing the frequency of Surveillance Requirement 4.4.6.2.2.a from at least once per 18 months to at least once each refueling does not change the basis for frequency. The proposed change does not alter the intent or method by which the surveillance is conducted, does not involve any physical change to the plant, does not alter the way any structure, system or component functions, and does not modify the manner in which the plant is operated. Further, based on the past surveillance test results, the proposed change will not degrade the RCS pressure boundary isolation valve's ability to l perform their function. Review of the corrective and preventive maintenance records concluded that there is no indication that the proposed extension could cause deteriorization in valve condition or performance. Thus the proposed change has no impact on the margin of safety.

III. Reactor Vessel Head Vent Path Valves, Surveillance Reauirements 4.4.11.2.a. b and c Safety Assessment The reactor vessel head vent path is provided to remove non condensible gases or steam from the reactor vessel head. The system is designed to mitigate a possible condition of inadequate core cooling or impaired natural circulation resulting from the accumulation of the non condensible gases in the reactor coolant (RCS) system. The system consists of two parallel flow paths with redundant isolation valves in each flow path. The active portion of the system consists of four one-inch open/close solenoid-operated isolation valves j connected to the one-inch vent pipe, which is located near t center of the reactor vessel head. There are two valves in series in each parallel flow path to minimize the possibility of reactor coolant system pressure boundary leakage. The isolation valves in one flow path are powered by one vital power supply and the valves in the second flow path are powered by another vital power supply.

Surveillance Requirements 4.4.11.2.a, b and c verify operability of the reactor vessel head vent flow path by verifying that all manual isolation valves in the vent path are locked in the open position, by cycling each vent valve through at least one complete cycle of full travel from the control room, and by verifying flow through the RCS vent paths during venting at least once per 18 months. NNECO proposes to extend the frequency of Surveillancis Requirements 4.4.11.2.a, b and c from at least once per 16 months to at least once each refueling interval (i.e., nominal 24 months).

U.S. Nuclear Regulatory Commission l B15293/ Attachment 4/Page 9 July 18, 1995 The proposed change to Surveillance Requirements 4.4.11.2.a, b and c does not alter the intent or method by which the surveillances are conducted, does not involve any physical changes to the plant, does not alter the way any structure, system or component functions, and does not modify the manner in which the plant is operated. As such the proposed change to the frequency of the Surveillance Requirements 4.4.11.2.a, b and c will not degrade the ability of the reactor vessel head vent path isolation valves to perform their intended function.

Equipment performance over the last four operating cycles was evaluated to determine the impact of extending the frequency of Surveillance Requirement 4.4.11.2.a, b and c. This evaluation included a review of surveillance results, preventive maintenance records and the frequency and type of corrective maintenance. The valves covered under these surveillances are included in Table 1.

Surveillance Procedure SP 3601B.1 "RCS Vent Valve Lineup" and SP 3601B.2 "RCS Vent Path Operability Check" are used to ,

perform applicable surveillances for the valves listed in Attachment 1. There were eight surveillances performed in the past to verify the lock open position of valve RCS*V153. In every case, the valve was 'ound in the correct position.

During the 1993 refueling outage, additional manual valves (3RCS*V894, V895, V896, V897) were installed in the reactor vessel head vant flow path. This design modification will allow any maintenance / corrective maintenance of the reactor ,

vessel head vent piping solenoid valves at power. The solenoid valves had a history of high maintenance due to seat leakage so the ability to perform maintenance / corrective maintenance on these valves at power was highly desirable.

Surveillance Procedure SP3601B.1 has been updated to reflect this modification. Surveillance Procedures SP3601B.2 covers operability of solenoid valves (3RCS*SV8095A & B, 3RCS*SV8096A

& B). A review of the past surveillance indicates that no failures occurred when the valves were cycled. Based on the surveillance test results, the reliability of the reactor vessel head vent path valves is considered high.

The corrective maintenance performed on the valves involved position indication problems. In each case, the condition was corrected while the plant was operating with no adverse impact on plant operation. They do not indicate any generic valve operability problem.

Review of the preventive maintenance requirements for the reactor vessel head vent path valves revealed that there is no periodic maintenance scheduled on an 18 month frequency.

l U.S. Nuclear Regulatory Commission l B15293/ Attachment 4/Page 10 July 18, 1995 Based on the engineering review of equipment performance, preventive and corrective maintenance history, the proposed change is considered acceptable.

The proposed change has been evaluated from a PRA perspective, and it has been determined that the change does not represent a risk to public safety.

Sianificant Hazards Consideration NNECO has reviewed the proposed change in accordance with 10CFR50.92 and has concluded that the change does not involve an SHC. The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not compromised. The proposed change does not involve an SHC because the change would not:

1. Involve a significant increase in the probability or consequences of an accident previously analyzed.

The proposed change to Surveillance Requirements 4.4.11.2.a, b and c extends the frequency for demonstrating the operability of the reactor vessel head vent flow path by verifying that all manual isolation valves in the vent path are locked in the open position, by cycling each vent valve through at least one complete cycle of full travel from the control room and by verifying flow through the RCS vent paths during venting. The proposal would extend the frequency from l at least once per 18 months to at least once each refueling I interval (i.e., nominal 24 months). ;

l The proposed change to Surveillance Requirements 4.4.11.2.a, l b and c doe' not alter the intent or method by which the l surveillancer are conducted. In addition, the acceptance criterion for each surveillance is unchanged. As such, the l proposed change to L9e frequency of Surveillance Requirements 4.4.11.2.a, b and : will not degrade the ability of the reactor vessel head vent path isolation valves to remove non condensible gases er steam from the reactor vessel head. It is noted that these valves are not credited in any accident analysis and do not operate in response to an accident. While the proposed change can lengthen the intervals between surveillances, the increase in intervals has been evaluated and it is concluded that there is no significant impact on the reliability or availability of these valves and consequently, there is no impact on the consequences of any analyzed accident. The proposed change can not affect the probability i of any previously analyzed accident, since the proposed change only affects the surveillance frequency.

2. Create the possibility of a new or different kind of accident from any previously analyzed.

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U.S. Nuclear Regulatory Commission B15293/ Attachment 4/Page 11 July 18, 1995 i

The proposed change to Surveillance Requirements 4.4.11.2.a, b and c does not modify the design or operation of any plant system. The proposed change does not alter the intent or method by which the surveillances are conducted other than increasing the interval from 18 months to 24 months (nominal) .

The proposed change does not introduce a new failure mode.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously analyzed.

3. Involve a significant reduction in the margin of safety.

Changing tne frequency of Surveillance Requirements ,

4.4.11.2.a, b and c from at least once per 18 months to at least once each refueling interval does not change the basis for frequency. The proposed change does not alter the intent or method by which the surveillances are conducted, does not involve any physical changes to the plant, does not alter the way any structure, system or component functions and does not ,

modify the manner in which the plant is operated. Further, I the previous history of reliability of the reactor vessel head vent path isolation valves provides assurance that the change will not affect the reliability of these valves. Thus the proposed change has no impact on the margin of safety.

IV. Surveillance Recuirements 4.4.4.1.a and 4.4.9.3.1.b. Power ODerated Relief Valve (PORV) Hich Pressure Loaic and PORV Actuation Channel Calibration Safety Assessment Surveillance Requirement 4.4.4.1.a verifies the operability of the PORVs (high pressure logic) by performance of a channel calibration at least once per 18 months. Surveillance Requirement 4.4.9.3.1.b verifies the operability of the PORVs (low temperature operation) by a performance of a channel calibration at least once per 18 wanths. NNECO proposes to extend the frequency of Surveillance Requirements 4.4.4.1.a i and 4.4.9.3.1.b from at least once per 18 months to at least i once each refueling interval (i.e., nominal 24 months).

The proposed change to surveillance Requirements 4.4.4.1.a and 4.4.9.3.1.b does not alter the intent or method by which the surveillances are conducted, does not involve any physical changes to the plant, does not alter the way any structure, system or component functions and does not modify the manner in which the plant is operated. As such, the proposed change j to t?.e frequency of Surveillance Requirements 4.4.4.1.a and 4.4.9.3.1.b will not degrade the ability of each PORV to perform its function during Modes 1, 2 and 3 and low temperature operation in Modes 4, 5, and 6.

I U.S. Nuclear Regulatory Commission B15293/ Attachment 4/Page 12 July 18, 1995 i Table 2 provides a listing of components tested by Surveillance Requirements 4.4.4.1.a and 4.4.9.3.1.b.

Component performance over the last four operating cycles was I evaluated to determine the impact of extending the frequency of Surveillance Requirements 4.4.4.1.a and 4.4.4.9.3.b. This evaluation included a review of surveillance results, preventive maintenance records and the frequency and type of ;

the corrective maintenance. l A review of the past surveillance results for Surveillance ,

Requirement 4.4.4.1.a indicates that in all cases, the PORV ;

high pressure logic channels were calibrated within the l acceptance criteria, and there was no indication of linear !

time dependent drift with regard to the circuit components. l maintenance and/or There are no regularly scheduled calibration activities associated with these parameters of the PORV high pressure logic. A review of corrective maintenance activities did not identify any significant activities that were required to correct component failures. ;

A review of the past surveillance results for Surveillance l Requirement 4.4.9.3.1.b indicates that in all cases the PORV Channels (low temperature operation) were calibrated to be I within the applicable acceptance criterion and there was no l indication of linear time dependent drift with regard to the circuit components. There are no regularly scheduled i maintenance and/or calibration activities associated with i these parameters of the PORV actuation logic components. A

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review of corrective maintenance activities did not identify any significant activities that were required to correct component failures. A probabilistic risk assessment (PRA) review was performed to assess the impact of the technical ;

specification change on the core damage frequency (CDF). Use i of conservative assumptions showed that the maximum theoretical increase in CDF is limited to 2.5%. This change is considered insignificant. Therefore, PRA concluded that ,

the technical specification change is acceptable.

1 On the basis of the above evaluation, there is reasonable assurance that the frequency of Surveillance Requirements 4.4.4.1.a and 4.4.9.3.1.b can be extended from at least once per 18 months to once each refueling interval (i.e., nominal 24 months).

Sionificant Hazards Consideratd9_D NNECO has reviewed the proposed change in accordance with 10CFR50.92 and has concluded that the change does not involve a significant hazards considerations (SHC). This basis for this conclusion is that the three criteria of 10CFR50.92 (c) are not compromised. The proposed change does not involve an SHC because the change would not:

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,j U.S.-Nuclear Regulatory Commission. b B15293/ Attachment 4/Page 13" EJuly:18,c1995 1.. . Involve a significant increase' in the probability or . .

consequences of an' accident _previously-evaluated.= ;

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.The proposed changr .to ' Surveillance Requirements' .l

~4.4.4.1.a and 4.4.9.3.1.b extends the frequency "or j demonstrating the operability'of the PORV (high pressu e. 1 logic and- low . temperature operation ' logic) .by: l

. performance of a channel . calibration. The proposal would !

extend the frequency from at least~once-per 18 months to ;

at least once each refueling interval(i.e., nominal ~24-months)..

The proposed-change does not' alter the' intent or. method-

by which the surveillance is conducted. In addition, the ;

acceptance criterion for each surveillance is unchanged. :

As such, the proposed change will not degrade the ability !

of the PORV to perform its function during high pressure or low temperature operation.- :

i An evaluation of past surveillances,- preventive maintenance records and the frequency of the _ type of j corrective maintenances concluded that decreasing ' the -l surveillance frequency will have little impact on safety. ,

Since the proposed change only affects the surveillance ,

frequency, the proposed change can not affect the' probability of any previously analyzed accident. While i the proposed change can lengthen the intervals between q surveillances, the increase in intervals has been. .

evaluated and -it is concluded that there is no .!

significant impact on the reliability or availability of the PORVs and consequently, there.is no impact on the' !

consequences of any analyzed accident.

2. Create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed change to . Surveillance Requirements 4.4.4.1.a and 4.4.9.3.b.1 does not modify the design or ;

operation of any plant system. The proposed change does !

not alter the intent or method by which the surveillance. I is conducted other than increasing the interval from 18 j months to 24 months (nominal). The proposed change does not introduce a new failure mode. Therefore, the j

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proposed change does not create the possibility of a'new or different kind of accident from any previously analyzed.

3. Involve,a significant reduction-in a margin of safety.

Changing the frequency of Surveillance Requirements 4~4.4.1.a and 4. 4. 9. 3.1. b ' from at least once per 18 months to'once each refueling interval.does not change

'U.S. Nuclear Regulatory Commission B15293/ Attachment 4/Page 14 July 18, 1995 the basis for frequency. The proposed change does not alter the intent or method by which the surveillance is conducted, does not involve any physical changes to the plant, does not alter the way any structure, system or component functions and does not modify the manner in which the plant is operated. Further, the previous history of the PORVs provides assurances that the change will not affect the reliability of these valves. Thus the proposed change has no impact on the margin of the safety.

V. Surveillance Reauirements 4. 4. 6.1.b, Containment Drain Sumo Level and Pumo Capacity Monitorina System Channel Calibration Safety Assessment Surveillance Requirement 4.4.6.1.b requires that the containment drain sump level and pumped capacity monitoring system instrumentation be calibrated at least once per 18 months. NNECO proposes to extend the frequency of Surveillance Requirement 4.4.6.1.b from at least once per 18 months to at least once each refueling interval (i.e.,

nominal 24 months) . Surveillance Procedure SP 3447B02 is used j to perform the calibration. A review of the past surveillance results indicates that these instruments were calibrated within the acceptance criteria, and there was no indication of linear time dependent drift with regard to the circuit components.

A review of past preventive maintenance and corrective maintenance activities did not identify any significant activities that were required to correct component failures.

On the basis of the above evaluation, there is a reasonable assurance that the frequency of Surveillance Requirement 4.4.6.1.b can be extended from at least once per 18 months to once each refueling (i.e., nominal 24 months)..

With due consideration to the risk significance of the function and the operating history, a PRA review concluded that this technical specification change has negligible affect on risk.

Sianificant Hazards Consideration NNECO has reviewed the proposed change in accordance with 10CFR50.92 and has concluded that the change does not involve a significant hazards consideration (SHC). This basis for this conclusion is that the three criteria of 10CFR50.92(c) are not compromised. The proposed change does not involve an SHC because the change would not:

_ a

U.S. Nuclear Regulatory Commission B15293/ Attachment 4/Page 15 July 18, 1995 1.

Involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed change to Surveillance Requirement 4.4.6.1.b extends the frequency for demonstrating operability of the containment monitoring drain sump level and pumped capacity system by performance of a channel calibration.

from at least once Theper proposal 18 monthswould extend the frequency to at least once each refueling interval (i.e., nominal 24 months).

The proposed change does not alter the intent or method by which the surveillance is conducted. In addition, the acceptance criterion for the surveillance is unchanged.

As such, the proposed change will not degrade the ability of the containment drain sump level and pumped capacity monitoring system to perform its leak detection function.

An evaluation of past surveillances, preventive maintenance records, and the frequency of the type of corrective maintenances concluded that decreasing the surveillance frequency will have little impact on safety.

Since the proposed change only affects the surveillance frequency, the proposed change cannot affect the probability of any previously analyzed accident. While the proposed change can lengthen the intervals between surveillances, the increase in evaluated, and it intervals has been is concluded that there is no significant impact on the reliability or availability of the containment monitoring systemdrain sump level and pumped capacity on the consequence,s of any analyzed accident.and consequently, there is no i 2.

Create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed change to Surveillance Requirement 4.4.6.1.b does system.not modify the design or operation of any plant The proposed change does not alter the intent or method by which the surveillance in conducted other than increasing the interval from 18 months to 24 months (nominal).

failure mode.TheTherefore, proposed change does not introduce a new the proposed change does not create the possibility of a new or different kind of accident from any previously analyzed.

3.

Involve a significant reduction in a margin of safety.

Changing the frequency of Survei)1ance Requirement I 4.4.6.1.b from at least once per 18 ronths to once each refueling interval does not change the basis for frequency. The proposed change does ncet alter the intent

w :n - . D i :C.

'l l

U.S. Nuclear. Regulatory Commission l "B15293/ Attachment 4/Page 16 l July?l8,:-1995 -!

or method by which the surveillance is conducted,..does. -j not involve any physical changes-to the plant,'doestnot' alter the .w ay any structure,. system, ~ or component' functions, and does~not modify the. manner in which the plant is. operated. Further, the previous history of the- ;

containment drain sump . level' and pumped' capacity monitoring system provides assurances .that tho' change -

will not affect the reliabilityEof these. systems., Thus 1 the proposed change has no impact on the margin of the !

! safety. L 4

4 h

i.

i f

a

'I I

P t

r TABLE 1 I. Components Covered by Surveillance Reauirement 4.4.6.2.2.a EQUIPMENT DESCRIPTION VALVE TYPE 3SIL*V15 SI Tank 1A Discharge Isol. Valve Check Valve 3SIL*V17 SI Tank 1B Discharge Isol. Valve Check Valve 3SIL*V19 SI Tank 1C Discharge Isol. Valve Check Valve 3SIL*V21 SI Tank 1D Discharge Isol. Valve Check Valve 3SIL*V26 RHR/SI to RCS Loop 2, Hot Leg Check Valve 3SIL*V27 SIH to RCS Loop 2, Hot Leg Check Valve n 3SIL*V28 RHR/SI to RCS Loop 4, Hot Leg Check Valve 3SIL*V29 SIH to RCS Loop 4, Hot Leg Check Valve 3SIL*V984 RHR/SI to RCS Loop 4, Cold Leg Check Valve i 3SIL*V985 RHR/SI to RCS Loop 3, Cold Leg Check Valve 3SIL*V986 RHR/SI to RCS Loop 2, Cold Leg Check Valve 3SIL*V987 RHR/SI to RCS Loop 1, Cold Leg Check Valve 3SIH*V5 SIH to RCS Cold Legs Check Valve 3SIH*V110 SIH to RCS Loop 1, Hot Leg Check Valve <

3SIH*V112 SIH to RCS Loop 3, Hot Leg Check Valve 3RCS*V26 SIH to RCS Loop 1, Hot Leg Check Valve 3RCS*V29 SIH to RCS Loop 1, Cold Leg Check Valve 3RCS*V30 SIL to RCS Loop 1, Cold Leg Check Valve 3RCS*V69 RHR/SI to RCS Loop 2, Hot Leg Check Valve 3RCS*V70 SIH to RCS Loop 2, Cold Leg Check Valve 3RCS*V71 SIL to RCS Loop 2, Cold Leg Check Valve 3RCS*V102 SIH to RCS Loop 3, Hot Leg Check Valve 3PCG2V105 SIH to RCS Loop 3, Cold Leg Check Valve 3RLS*V107 SIL to RCS Loop 3, Cold Leg Check Valve 3RCSeV142 RHR/SI to RCS Loop 4, Hot Leg Check Valve

~

3RCS*V145 SIH to RCS Loop 4, Cold Leg Check Valve f

1

e ,

l i

EQUIPMENT DESCRIPTION VALVE TYPE 3RCS*146 SIL to RCS Loop 4, Cold Leg Check Valve 3RHS*MV8701C RCS Loop 1, Hot Leg to RHR Motor Operated l Valve ,

3RHS*MV8702C RCS Loop 4, Hot Leg to RHR Motor Operated Valve 3RHS*MV8701A RCS Loop 1, Hot Leg to RHR Motor Operated Valve 3RHS*MV8702B RCS Loop 4, Hot Leg to RHR Motor Operated Valve l

l l

l 1

i l

l l

l 2

I

)

i l

_ j

TABLE 1 (CONTINUED)

II. Comoonents Covered by Surveillance Recuirements 4.4.11.2.a. b and c EQUIPMENT DESCRIPTION ACTUATION 3RCS*V153 Reactor vessel head vent isolation valve Manual valve 3RCS*V894** Reactor vessel head vent common Manual valve

, isolation 3RCS*V895** Reactor vessel head vent A solenoids Manual valve isolation 3RCS*V896** Reactor vessel head vent B solenoids Manual valve l isolation 3RCS*V897** Reactor vessel head vent downstream Manual valve isolation 3RCSASV8095A Reactor vessel head vent A isolation Solenoid valve valve 3RCS*SV8096A Reactor vessel head vent A isolation Solenoid J valve valve 3RCS*SV8095B Reactor vessel head vent B isolation Solenoid I valve valve 3RCS*SV8096B Reactor vessel head vent B isolation Solenoid valve valve l

- NOTE: These valves have been installed in the reactor vessel head vent flow path in 1993 during the refueling outage 4.

Procedure SP 3601B.1 has been updated to reflect this change and these valves have been added into the scope of the surveillance. Until the refueling outage 4 however only valve V153 was in the scope of SP 3601B.1. l 1

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f l

TABLE 2 EQUIPMENT TESTED BY TECHNICAL SPECIFICATION REQUIREMENTS 4.4.4.1.a AND 4.4.9.3.1.b J l

EQUIPMENT DESCRIPTION l

3RCS*PI405, 3RCS*PI403 RCS WIDE RANGE PRESSURE INDICATORS j l

3RCS-PR403 RCS WIDE RANGE PRESSURE RECORDER ;

3RCS-PI405A, 3RCS-PI403A RCS WIDE RANGE PRESSURE INDICATORS 3RCS*PI405B, 3RCS*PI403B RCS WIDE RANGE PRESSURE INDICATORS 3RCS*TE413B, 413C, 423B, RCS WIDE RANGE TEMPERATURES 423C, 433B, 433C, 443B, 443C ,

3RCS*PT405, 3RCS*PT403 RCS WIDE RANGE PRESSURE 3RCS*PT405A, 3RCS*PT403B TRANSMITTERS l

3RCS*PT455, 3RCS*PT456 PRESSURIZER PRESSURE TRANSMITTERS l 3RCS*PT457, 3RCS*PT458 j 3RCS*PI455A, 3RCS*PI455B PRESSURIZER PRESSURE INDICATORS 3RCS*PI456A, 3RCS*PI456B 3RCS-PI457, 3RCS-PI458 PRESSURIZER PRESSURE INDICATORS 4

3RCS-PR455 PRESSURIZER PRESSURE RECORDERS J

e- ;

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i TABLE 2 EQUIPMENT TESTED BY TECHNICAL SPECIFICATION REQUIREMENT 4.4.6.1.b EQUIPMENT DESCRIPTION LT-42, LT-22, LT-39, Containment Sump Level LI-22, LI-39 Containment Sump Level Indication L 22, L 39 Containment Sump Level Computer Point P 23A, P 23B Containment Sump Pump Discharge ,

Pressure Computer Point F 56 Containment Sump Pump Flow Computer Point

.

ML20086M115

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